ML100360106

Rhode Island Atomic Energy Commission, RAI, Regarding the Rhode Island Nuclear Science Center Reactor License Renewal

ML100360106
Person / Time
Site:	Rhode Island Atomic Energy Commission
Issue date:	04/13/2010
From:	William Kennedy Research and Test Reactors Licensing Branch
To:	Tehan T State of RI, Atomic Energy Comm
Kennedy, W B, NRR/DPR, 415-2784
References
TAC ME1598
Download: ML100360106 (33)

Text

April 13, 2010 Dr. T. Tehan, Director Rhode Island Nuclear Science Center Rhode Island Atomic Energy Commission 16 Reactor Road Narragansett, RI 02882-1165

SUBJECT:

RHODE ISLAND ATOMIC ENERGY COMMISSION, REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR LICENSE RENEWAL (TAC NO. ME1598)

Dear Dr. Tehan:

The U.S. Nuclear Regulatory Commission (NRC) is continuing the review of your application for renewal of Facility Operating License No. R-95 for the Rhode Island Nuclear Science Center Reactor dated May 3, 2004, as supplemented on December 15, 2009, January 4, and January 19, 2010. During our review, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed requests for additional information no later than 60 days from 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 under oath or affirmation.

If you have any questions regarding this review, please contact me at 301-415-2784 or by electronic mail at William.Kennedy@nrc.gov. Sincerely,

/RA/ William B. Kennedy, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-193

Enclosure:

As stated Cc w/encl: See next page

Rhode Island Atomic Energy Commission Docket No. 50-193 cc: Governor Donald Carcieri State House Room 115 Providence, RI 02903 Dr. Stephen Mecca, Chairman Rhode Island Atomic Energy Commission Providence College Department of Engineering-Physics Systems River Avenue Providence, RI 02859 Dr. Harry Knickle, Chairman Nuclear and Radiation Safety Committee University of Rhode Island College of Engineering 112 Crawford Hall Kingston, RI 02881

Dr. Andrew Kadak 253 Rumstick Road Barrington, RI 02806 Dr. Bahram Nassersharif Dean of Engineering University of Rhode Island 102 Bliss Hall Kingston, RI 20881

Dr. Peter Gromet Department of Geological Sciences Brown University Providence, RI 02912 Dr. Alfred L. Allen 425 Laphan Farm Road Pascoag, RI 02859 Mr. Jack Ferruolo, Supervising Radiological Health Specialist Office of Occupational and Radiological Health Rhode Island Department of Health 3 Capitol Hill, Room 206 Providence, RI 02908-5097 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 April 13, 2010 Dr. T. Tehan, Director Rhode Island Nuclear Science Center Rhode Island Atomic Energy Commission 16 Reactor Road Narragansett, RI 02882-1165

SUBJECT:

RHODE ISLAND ATOMIC ENERGY COMMISSION, REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR LICENSE RENEWAL (TAC NO. ME1598)

Dear Dr. Tehan:

The U.S. Nuclear Regulatory Commission (NRC) is continuing the review of your application for renewal of Facility Operating License No. R-95 for the Rhode Island Nuclear Science Center Reactor dated May 3, 2004, as supplemented on December 15, 2009, January 4, and January 19, 2010. During our review, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed requests for additional information no later than 60 days from 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 under oath or affirmation.

If you have any questions regarding this review, please contact me at 301-415-2784 or by electronic mail at William.Kennedy@nrc.gov. Sincerely,

/RA/ William B. Kennedy, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-193

Enclosure:

As stated Cc w/encl: See next page DISTRIBUTION

PUBLIC DPR/PRT r/f RidsNrrDpr RidsNrrDprPrtb RidsNrrDprPrta WKennedy, NRR GLappert, NRR

ACCESSION NO.:ML100360106 NRR-088 Office PRLB:PM PRPB:LA PRLB:BC PRLB:PM Name WKennedy GLappert KBrock WKennedy Date 4/7/2010 4/12/2010 4/13/2010 4/13/2010

ENCLOSURE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL FOR THE RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR LICENSE NO. R-95 DOCKET NO. 50-193 The content of an application for research reactor license renewal is based on the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Sections 50.33, 50.34 and 50.36. Important requirements in 10 CFR 50.34, "Contents of applications; technical information," include: ...analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

...information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and shall include the following:

...description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished.

...description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.

...such items as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent.

The following requests for additional information (RAIs) are related to the Safety Analysis Report (SAR) submitted as part of the application for license renewal for the Rhode Island Nuclear Science Center (RINSC) reactor dated May 3, 2004. The number of each RAI begins with the number of the corresponding chapter of the SAR. Responses to the RAIs should be in the form of discussion and/or analysis, as appropriate, and provide sufficient information for the NRC staff to independently verify all safety-related conclusions. Responses to the RAIs may be in the form of replacement pages for the SAR. NUREG-1537, Part I, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content," dated February, 1996, contains guidance for providing sufficient information.

2.1 Chapter 2 of the SAR contains multiple section headings with no related information. Provide the omitted information. 2.2 Section 2.4.4. Provide a discussion of potential impacts of the RINSC on groundwater, or the lack thereof, including the potential for neutron activation of groundwater, leakage from the reactor pool and primary coolant system, and leakage from contaminated water systems at the facility. 4.1 Figures 4-5, 4-6, 4-7, and 4-8 were omitted from the SAR. Provide these figures.

4.2 Section 4.2.1. Provide a summary description of the fuel development and qualification program for the fuel type used at the RINSC. The description should include the fuel characteristics and parameters important to safe operation of the reactor (e.g., power density, power rate change limits, burnup, etc.). Verify that those parameters important to safety are included in or bounded by the requirements in the technical specifications (TS). 4.3 Section 4.1 states that the fuel composition is U 2 Si 2, while Section 13.2.1 states that the fuel composition is U 3 Si 2. Clarify which composition is correct for the RINSC reactor fuel. 4.4 Section 4.2.2. Provide a summary description of the program for shim safety blade and regulating blade inspection and replacement. 4.5 Section 4.2.3 states that the graphite reflectors are designed for expansion "from an integrated flux of 2X10 21 nvt (expansion based on a more than two-year, full-power operation factor)." Given that the TS do not explicitly limit the duration of full-power operation, provide a discussion of the methods used to ensure that the graphite reflectors will not be exposed to an integrated neutron flux greater than the expansion design basis (e.g., calculation of integrated flux, surveillance programs, etc.). The discussion should include consideration of current integrated flux and integrated flux during the period of the renewed license. 4.6 Section 4.2.5. Describe the design characteristics of the reactor that ensure the control blades will fully insert despite motion of the core support structure (e.g., shaking of the core due to an earthquake). The response should include tolerances between the control blades and the control blade shrouds that prevent binding of the blades within the shrouds. 4.7 Section 4.2.4. Provide a discussion of design features of the neutron startup sources that allow for reliable operation and replacement of the sources. The discussion should include calibrations, source checks, interlocks, and risk of damage to the sources.

Include a discussion of any design features and/or administrative controls that reduce the potential for damage to the sources. The discussion should also describe whether improper operation or damage to the sources could potentially lead to instrument error or mislead reactor operators. If the potential exists for damage to the neutron startup sources from operation of the reactor, propose TS requirements to ensure there will be no damage to the sources, or provide justification for not having such TS requirements. 4.8 Section 4.2.5. Provide justification for the design of the core support structure as to its ability to support the weight of the core and its ability to withstand radiation damage, mechanical stress, and chemical degradation over the period of the renewed license. 4.9 Section 4.4. Discuss the ability of the biological shield and pool liner to continue to meet their design bases during the period of the renewed license. Include considerations of radiation, chemical, and thermal degradation. Describe any surveillance programs in place to detect degradation of the biological shield and pool liner. 4.10 Section 4.5. Describe and justify the methods used to determine the reactor kinetics parameters found in Table 4-3. Provide the names of any codes used and a description of the modeling process, if applicable. 4.11 Section 4.5. Describe and justify the calculation methods for the coefficients of reactivity for temperature, void, and power. Discuss any measurements made to confirm the reactivity coefficients. Include estimates of accuracy for the coefficients. 4.12 Table 4-3 lists coolant reactivity coefficients for the coolant temperature range of 20-40 degrees Celsius (degrees C). TS 3.2.1 allows operation of the reactor with coolant temperatures up to 126 degrees Fahrenheit (degrees F) (52 degrees C).

Provide coolant reactivity coefficients over the entire coolant temperature range allowed by the TS. 4.13 Table 4-3 includes the Doppler coefficient of reactivity over the temperature range of 20-40 degrees Celsius. This temperature range appears to apply to the coolant temperature and not the fuel temperature. Provide the Doppler coefficient of reactivity over the range of fuel temperatures anticipated during all allowed modes of reactor operation and reactor transients. 4.14 Table 4.5 gives a maximum total power peaking factor of 3.06 for grid position D6. Explain how this power peaking factor accounts for localized power peaking that could be caused by a flooded experiment located in or adjacent to the core. (See RAI 14.65) 4.15 Section 4.6 references the computer code PLTEMP as the code used to determine many of the thermal-hydraulic characteristics of the reactor core. Provide a discussion of the use of this code including models of the RINSC core, applicability of the code to the thermal hydraulic conditions in the RINSC core, validation and benchmarking of the code, and code uncertainty. Provide a copy of Reference 4.6. 4.16 Section 4.6. This section makes multiple references to "Reference 4-Y." Provide the correct reference and a copy of the reference document. 4.17 Provide the values of coolant temperature and coolant height used in the analysis of Section 4.6.2. Provide justification for the use of these values. 4.18 Section 4.6.2. Confirm that the units for the values of Tsurface , T sat, and T onb found in Table 4-9 and Table 4-10 are degrees C. 4.19 Section 4.6.2. Describe the methods used to determine the values of Tsurface and T onb found in Table 4-9 and Table 4-10. Include all assumptions and correlations used in the calculations and provide justification for their use given the thermal-hydraulic characteristics of the coolant channels. 4.20 Section 4.6.2. Provide the uncertainties for the limiting safety system setting (LSSS) values for coolant height and overpower trip. Provide justification for all uncertainty values associated with the LSSS for coolant height, overpower trip, coolant temperature, and coolant flow. (See RAI 14.44) 4.21 Section 4.6.2 states that with a flow rate of 1,950 gallons per minute (gpm), the incipient boiling temperature (defined in the SAR as T onb) occurs at about 2.6 MW. From Table 4-9, it appears that the incipient temperature occurs somewhere between 1,715 gpm and 1,800 gpm. Clarify this apparent discrepancy. 4.22 Section 4.6.2 states that the reduced flow trip setting is 1,700 gpm. The requirements of TS 2.2.1 and TS 3.2.1 allow the reduced flow trip to be set at 1,600 gpm. Clarify this apparent discrepancy. 4.23 Note number 2 to Table 4-12 states that the calculations are based on a reactor inlet temperature of 42.3 degrees C. Explain the reason this temperature is used in the analysis given that it is less conservative than the coolant temperatures allowed by the proposed TS. 4.24 The rising power transient analysis of Section 4.6.4 of the SAR shows that the reactor power would reach 2.78 MW. Explain how the analysis supports the safety limit of 2.4 MW given in TS 2.1.1.1. (See RAI 14.32) 4.25 Section 4.6.4. The assumptions used in the rising power transient calculation are not consistent with the requirements of the TS. The analysis assumes a minimum reactor period of slightly greater than 7 seconds, while the TS allow a minimum reactor period of 4 seconds. Also, the surface temperature value of 122.93 degrees C appears to be based on a coolant flow rate of 1,715 gpm, which is greater than the TS requirement of 1,600 gpm. Provide a revised calculation that supports the requirements in the TS.

Include justification of all assumptions, including the assumed coolant temperature and coolant height. 4.26 Section 4.6.4 gives a surface temperature of 122.93 degrees C for normal flow at 2.6 MW. Table 4-9 indicates that this temperature corresponds to 1,715 gpm at 2.6 MW. Page 4-3 indicates that nominal flow is 1,950 gpm. Clarify this apparent discrepancy. 4.27 Section 4.6.4 states, "for a hot channel analysis, the ONB region would not present a problem for the LEU fuel." Provide justification for this conclusion. 4.28 Provide the values of coolant temperature and coolant height used in the analysis of Section 4.6.5. Provide justification for the use of these values. 4.29 Table 4-17. Explain the meaning of the negative value for Margin to Incipient Boiling at 209.1 kW. (See RAI 14.35) 4.30 Section 4.7. Clarify whether the correct reference to the figure showing the expanded core configuration is Figure 4-1 or Figure 4-2. 4.31 Section 4.8 - Clarify which figure is meant by "Figure 4" in the text. 4.32 Section 4.8. Provide a discussion of the correlation and/or calculations used to develop the Departure from Nucleate Boiling (DNB) and Departure from Nucleate Boiling Ratio (DNBR). Include all assumptions made in the analysis and justification for those assumptions. Clarify whether the term "Margin to Departure from Nucleate Boiling" in Table 4-19 is synonymous with the term "DNBR." 5.1 Section 5.5. The makeup water system operates to automatically add water to the reactor pool upon a low level indication. Excessive operation of the system could either indicate a leak in the reactor coolant, a malfunction of the pool level indicator, or a malfunction of the makeup water controls. This could result in overfilling of the pool. Describe the controls to detect abnormalities or leaks in the makeup water system. 6.1 Describe the operating parameters and design features of the confinement, including: free volume, number and type of penetrations, etc. 7.1 Provide a listing of the interlocks of the reactor protection system. 7.2 Section 7.2.5. Provide more detail regarding the interconnections of the neutron flux monitoring system, including equipment lists and performance specifications to clarify its operation. 7.3 Section 7.2.12 discusses the relay scram circuit. Provide more detail regarding the design and operation of the bridge misalignment and bridge movement safety channels required by TS 3.2.1, Table 3.1. Do these channels have set points? What is the minimum motion detectable by the bridge movement safety channel? What is the alignment tolerance associated with the bridge misalignment safety channel? Explain any interlocks that prevent reactor startup in the forced convection mode if the bridge is misaligned. (See RAI 14.71) 7.4 Section 7.2.15. Provide more detail regarding the fixed radiation monitoring instrumentation. Include a list of the positions and types of fixed monitors and indicate which have local readouts and/or alarms. 9.1 Section 9.1.2 provides no detail regarding the design specifications of the normal and emergency ventilation system other than general arrangement. TS 3.7.2 credits the ventilation system with a dilution of waste streams by a factor of 4 x 10

4. Provide sufficient details regarding both the normal and emergency ventilation system flows to confirm the appropriateness of the dilution factor. 9.2 Section 9.2 uses inconsistent units when discussing criticality protection for fuel in storage. Confirm that k eff is less than 0.8 for fuel in storage. 9.3 Figure 9-2 displays the fuel element cut-off saw. This saw is not described in the SAR. Provide a discussion of its use, when it is used, and the design features and controls in place to prevent cutting into the fissile material and control of cutting debris. 9.4 Provide a legible copy of Figure 9-7. 9.5 The references listed for this chapter lack dates and detail. Provide a more formal reference list for Chapter 9 that includes this information. 10.1 Section 10.2.1 does not describe the design features of the beam port covers or administrative controls regarding beam port use that support the assumptions made in the loss-of-coolant-accident (LOCA) presented in Section 13.2.3 of the SAR. Provide a description of the design features and administrative controls that is consistent with the assumptions made in the LOCA analysis. 10.2 Section 10.2.2 discusses administrative controls in place to limit draining of the reactor pool via the through-port. The text states that the through-port should not be opened for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following reactor shutdown. The LOCA analysis presented in Section 13.2.3 of the SAR does not analyze pool drainage through the open through-port, nor does it provide a comparison of pool drain time for a closed and open through-port. Provide justification for the statement that opening the through-port 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor shutdown is conservative. 10.3 Section 10.2.4 discusses the thermal column experiment facility. The discussion states that cooling air is required to remove heat generated in the thermal column graphite in order to prevent the graphite from overheating. However, there is no analysis of the flow rate necessary to adequately cool the graphite, and TS 3.2.1 does not provide a set point for the safety channel associated with the thermal column. Provide an analysis of air cooling of the graphite that includes the minimum flow rate necessary to cool the graphite. Explain the basis for the graphite temperature limit of 107 degrees C. (See RAI 14.74) 10.4 Section 10.2.8.2 references Section 10.8.3 which does not appear in the SAR. Provide the correct reference or provide the missing Section 10.8.3. 11.1 The bases for TS 4.1 state that "Shim safety blade inspections are the single, largest source of radiation exposure to facility personnel." However, the safety blades are not listed explicitly in Chapter 11 of the SAR as one of the facility radiation sources. Verify that this statement is accurate and describe any additional radiological controls that are used for safety blade inspections that are not included in Chapter 11. 11.2 Section 11.1.1.1. The text references calculations of airborne activity that are described in Appendix A. This document was not provided with the license renewal application. Provide a copy of the referenced Appendix A. 11.3 Section 11.1.4. of the SAR states that the Radiation Safety Office conducts routine radiation and contamination surveys. Discuss the bases of the methods and procedures used for conducting routine radiation and contamination surveys. 11.4 Section 11.1.5. Describe the provisions for the use of extremity monitoring and the conditions under which extremity monitoring is used at the RINSC. 11.5 Section 11.1.5. Describe the provisions for internal monitoring at the RINSC. Include any provisions for use of radiological respirators at the RINSC. 13.1 Section 13.1 lists nine credible accidents for research reactors based on the guidance in NUREG-1537, but only provides analyses for seven types of accidents. Provide analyses of the omitted accidents, or provide justification for not analyzing the omitted accidents. 13.2 The analysis of the maximum hypothetical accident (MHA) does not include radiation doses to personnel inside the reactor building. Provide an analysis of radiation doses to the personnel inside the reactor building. Discuss all assumptions used in the analysis, including justification for the use of the assumptions. 13.3 Table 13-3, column 2 gives the release rate of iodine isotopes from the reactor stack during the MHA. Explain how the release rate was calculated, including all assumptions regarding confinement building volume and emergency exhaust system flow rate.

Explain how the analysis is consistent with the requirements in the TS. Provide an example calculation. Explain whether the same method and assumptions used for the iodine release rate analysis was used for the whole body gamma dose analysis. If not, explain the method and assumptions used for the whole body gamma dose analysis and provide an example calculation. (See RAI 14.97) 13.4 The footnote of Table 13-5 indicates an assumed reduction of 10% of radioiodine by the reactor pool. Page 13-4 indicates a release of 1% of radioiodine from the reactor pool.

Explain this apparent inconsistency. 13.5 The footnote of Table 13-5 indicates a 50% reduction of noble gases. Explain the reason for the reduction in noble gases. 13.6 Section 13.2.2 of the SAR references Figure 13.1, but the figure does not appear in the SAR. Provide a copy of this figure. 13.7 Section 13.2.2 of the SAR states that ANL performed a PARET analysis of reactivity insertions, but there is no reference provided for the PARET analysis. Provide a copy of the referenced calculation, including initial conditions and assumptions used in the analysis. If available, provide a copy of the PARET input deck. 13.8 Section 13.2.2 states that a 200 millisecond delay was used as a conservative assumption for the time for the control blades to begin to insert following a scram. However, TS 3.2.3 specifies that the full control blade insertion time is 1 second, and does not specify a maximum control blade insertion delay time. Explain this apparent inconsistency between the SAR and the proposed TS. 13.9 Section 13.2.2 of the SAR states that during a reactivity insertion, the onset of nucleate boiling is approached, but does not occur. Provide quantitative details regarding the approach to nucleate boiling that show that the safety limits are not exceeded. (See RAI 14.62) 13.10 Section 13.2.3 presents the LOCA analysis for a break in a beam port. Provide justification that this beam port failure is the limiting initiating event for a LOCA. 13.11 The calculation of pool drain time in Section 13.2.3 makes assumptions about the design of and administrative controls for use of the beam ports and through-port. Propose TS requirements for the design and operation of the beam ports and through-port that are consistent with the assumptions made in the analysis of a LOCA, or provide justification for not including such TS requirements. 13.12 Line 32 of page 13-10 states that a coolant height of "139.4 feet (normal water level of pool)" was used as the initial coolant height in the LOCA analysis. Explain why this coolant height is consistent with the limiting safety system setting for coolant height given by TS 2.2.1 and the set point for the pool water level safety channel required by TS 3.2.1. (See RAI 14.72) 13.13 Page 13-10. Provide definitions for the terms h1, h2, and C in the equation on line 43. 13.14 Page 13-13. The boundary condition of 1,200 degrees F used in the calculation is not consistent with the cladding blistering temperature of 986 degrees F, which is the criterion for fuel damage found in the literature for U 3 Si 2 fuel. Provide an analysis using the fuel blistering temperature, or provide a discussion of why the boundary condition of 1,200 degrees F is conservative. 13.15 Page 13-13. Please provide a conclusion for the analysis ending on line 18 of this page. 13.16 Page 13-13. The analysis assumes that the decay power spatial distribution can be approximated by a sinusoidal curve. Provide justification for this assumption. 13.17 The calculation of "Heat Conduction to the Water in Core Box from the Non-Fuel Aluminum in the Element" appears incomplete. Provide the remainder of the calculation, a discussion of the results of calculation, all assumptions made in the calculation and justification for those assumptions, and any conclusions based on the calculation. 13.18 Section 13.2.4 mentions a low flow alarm and a low flow trip that are inconsistent with the requirements of the TS. Provide analyses of loss-of-coolant-flow accidents that are consistent with the requirements of the TS, or propose TS requirements that are consistent with the current analyses. 13.19 Section 13.2.4.1 states that the peak clad temperature during a loss-of-flow-accident induced by a loss of electrical power is 103 degrees C. Provide an analysis that supports this statement. Justify all assumptions made in the analysis. 13.20 Section 13.2.5 provides an analysis of a startup accident, but does not specify assumptions for coolant flow or coolant height. Explain the assumptions used in the analysis for coolant flow and coolant height. Explain how the analysis treated power peaking factors. 13.21 Figures 13-1, 13-2, 13-3, and 13-4 were not included in the license renewal application. Provide copies of these figures.

Regulation 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).

A 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.

The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34.

Regulation 10 CFR 50.36(c) states that technical specifications will include safety limits (SL), limiting safety system settings (LSSS), limiting conditions for operation (LCO), surveillance requirements, design features, and administrative controls.

In regard to SLs, 10 CFR 50.36(c)(1)(i)(A) states:

Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.

In regard to LSSS, 10 CFR 50.36(c)(1)(ii)(A) states:

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

In regard to LCOs, 10 CFR 50.36(c)(2) states:

(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of [four criteria specified by 10 CFR 50.36(c)(2)(ii)(A), (B), (C), and (D)]

In regard to surveillance requirements, 10 CFR 50.36(c)(3) states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. In regard to design features, 10 CFR 50.36(c)(4) states:

Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in 10 CFR 50.36(c)(1), (2), and (3).

In regard to administrative controls, 10 CFR 50.36(c)(5) states:

Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

American National Standards Institute/American Nuclear Society standard ANSI/ANS-15.1, 2007, "Development of Technical Specifications for Research Reactors," (ANSI/ANS-15.1) 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 following RAIs relate to the proposed TS (Chapter 14 of the SAR) submitted with the license renewal application. In responding to the following RAIs, provide a response to each individual RAI, and provide a complete set of revised proposed technical specifications that incorporate any changes made as a result of the responses to the RAIs.

General 14.1 The proposed TS contain numerous references to a version of the SAR that is different than the version of the SAR submitted with the license renewal application (e.g., TS 4.2.6 references "SAR (Part A,Section V)"). Such references are included in TS 4.1.1.b, 4.2.6, 4.2.7, 4.2.8, 5.3, 5.5, and in the bases for TS 2.1.1, 2.2.1, 3.1, 3.2, 3.9.a, 4.9.a, and 4.9.b. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.2 The "Specification" section of several proposed TS contain references to portions of the SAR. Any portion of the SAR referenced in the "Specification" section of a proposed TS will become part of the TS and license. Unless it is intended that portions of the SAR become requirements of the TS and license, revise the "Specification" sections of the proposed TS to eliminate references to the SAR. Section 1.0, "Definitions" 14.3 The proposed TS do not appear to use the term "certified operator" defined by TS 1.1. Explain the reason for including this definition, and revise the proposed TS as appropriate. 14.4 The proposed TS do not appear to use the term "class A operator" found in the definition of TS 1.1.1. Explain the reason for including this term in the definition, and revise the proposed TS as appropriate. 14.5 TS 1.1.1 does not specify that a senior reactor operator is also a reactor operator. If it is intended that a senior reactor operator can also function as a reactor operator, revise TS 1.1.1 accordingly. 14.6 The proposed TS do not appear to use the term "class B operator" found in the definition of TS 1.1.2. Explain the reason for including this term in the definition, and revise the proposed TS as appropriate. 14.7 The wording of TS 1.1.2 is non-specific in that it defines a reactor operator as, "an individual who is licensed to operate the controls of a reactor." Explain the reason for not making this definition specific to the RINSC reactor, and revise the definition as appropriate. 14.8 TS 1.4 contains two references which are more than 40 years old. Revise the definition to include valid and up-to-date references. 14.9 The proposed TS contain definitions of two distinct types of channels (i.e., instrumentation channel (TS 1.5) and measured channel (TS 1.8)). The definitions of these channel types are very similar, but include different lists of components that comprise each type of channel. Explain the physical and operational characteristics that differentiate these two channel types. Explain the reason for not consolidating the definitions of the two channel types into a single definition of "channel" that is consistent with the guidance in ANSI/ANS-15.1. 14.10 The proposed TS do not appear to use the term "instrumentation channel" defined by TS 1.5. Revise the proposed TS to use consistent terminology. 14.11 TS 1.5 contains subsections that define "channel test," "channel check," and "channel calibration." As formatted in the proposed TS, it is unclear whether these definitions apply only to instrumentation channels, or whether they apply to other types of channels defined in the proposed TS (i.e., measured channel (TS 1.8)). If it is intended that these definitions apply to all types of channels, revise the proposed TS accordingly. 14.12 The proposed TS do not appear to use the term "measured channel" defined by TS 1.8. The proposed TS use the terms "measuring channel" (TS 1.28) and "safety channel" (TS 1.28 and TS 3.2, Table 3.1). Revise the proposed TS to use consistent terminology. 14.13 TS 1.9 uses the term "measuring channel" in the definition of measured value. This term is not defined in the proposed TS (TS 1.8 defines "measured channel"). Revise the proposed TS to use consistent terminology. 14.14 TS 1.14 does not specify a reference core condition at which the excess reactivity is measured. Explain the reason for not specifying a reference core condition, and revise the proposed TS as appropriate. 14.15 TS 1.15 defines reactivity limits as "limits imposed on the reactor core excess reactivity." Contrary to this definition, TS 3.1, "Reactivity Limits," contains many limits on reactivity that are not related to the reactor core excess reactivity (e.g., TS 3.1.6). Explain this apparent inconsistency, and revise the proposed TS as appropriate. 14.16 TS 1.15 appears redundant with TS 1.14, except that TS 1.15 specifies that the excess reactivity is referenced to a reference core condition. Consider consolidating TS 1.14 and TS 1.15. 14.17 Clarify whether the word "equipment" used in TS 1.16 should be replaced with the word "experiment." If so, revise TS 1.16 as appropriate. If not, explain how alterations in equipment position or configuration could affect the reactivity worth of an experiment. 14.18 The first sentence of TS 1.17 states, "the reactor is operating whenever it is not secured or shut down." The term "reactor secured" is not defined in the TS. TS 1.19 defines the term "reactor secure." Explain this apparent inconsistency, and revise the proposed TS as appropriate. 14.19 The formatting of TS 1.19 is confusing in that it contains a subsection that defines "subcritical." The formatting implies that the reactor is secure whenever it is subcritical, which is inconsistent with the guidance in ANSI/ANS-15.1. Additionally, the definition of subcritical given by TS 1.19.1 is inconsistent with the use of the term in other definitions (e.g., TS 1.20). Revise TS 1.19 to be consistent with the guidance in ANSI/ANS-15.1, or propose separate definitions for "reactor secured" and "subcritical" that are consistent with the other proposed TS. 14.20 TS 1.24 states, "a removable experiment... can reasonably be anticipated to be moved one or more times during the life of the reactor." Clarify whether the anticipated movement of a removable experiment would be intentional movement of the experiment or could be unintentional movement of the experiment. Similar to TS 1.3.2, describe the restraining forces required for removable experiments. Explain the differences between removable experiments and secured experiments. Explain the differences between removable experiments and movable experiments. 14.21 Explain why the definition of removable experiment given in TS 1.24 is separate from the definitions of other types of experiments contained in TS 1.3, and revise the proposed TS as appropriate. 14.22 The definition of research reactor given by TS 1.26 is non-specific to the RINSC reactor. Explain the reason for including this definition, and revise the proposed TS as appropriate. 14.23 The proposed TS do not appear to use the term "rundown" defined by TS 1.27. Explain the reason for including this definition, and revise the proposed TS as appropriate. 14.24 TS 1.27 defines a "rundown" as the automatic insertion of the shim safety blades. Explain how a rundown is different from automatic insertion of the shim safety blades caused by a scram, and revise the definition accordingly. Clarify whether there are any provisions for a manually-initiated rundown. 14.25 TS 1.28 states that a safety channel is a "measuring channel," but the term "measuring channel is not defined in the proposed TS (TS 1.8 defines "measured channel"). Revise the proposed TS to use consistent terminology. 14.26 The definition of "scram time" given in TS 1.30 uses the phrase "specified control blade movement." Explain what "specified control blade movement" means, and revise the proposed TS as appropriate. 14.27 The definition of "shim safety blade" given in TS 1.31 uses the phrase "function of a safety blade." Explain the meaning of this phrase as it applies to the RINSC reactor and revise the proposed TS as appropriate. 14.28 The definition of "shutdown margin" given in TS 1.33 is inconsistent with the requirements of TS 3.1.1 in that the definition does not specify the position of the regulating blade. Also, TS 1.33 uses the phrase "most reactive position," while TS 3.1.1 uses the phrase "fully withdrawn" to describe the positions of control blades. Explain these apparent inconsistencies in the TS, and revise the proposed TS as appropriate. (See RAI 14.56) 14.29 The proposed TS do not appear to use the term "static reactivity worth" defined by TS 1.35. Explain the reason for including this definition in the proposed TS, and revise the proposed TS as appropriate. 14.30 TS 1.37 appears to be a description of allowed deferral of surveillance activities and not a definition of surveillance activities. ANSI/ANS-15.1 recommends that allowed deferral of surveillance activities be included in each TS requiring a surveillance activity. Explain why TS 1.37 is included in the definitions section of the proposed TS, and revise the proposed TS as appropriate. (See RAI 14.130) 14.31 TS 1.38 states that maximum surveillance intervals "are to provide operational flexibility and to reduce frequency." The guidance in ANSI/ANS-15.1 states that maximum surveillance intervals "are to provide operational flexibility only and are not to be used to reduce frequency." Explain this apparent inconsistency, and revise the proposed TS as appropriate. Section 2.0, "Safety Limits and Limiting Safety System Settings" 14.32 The "Applicability" section of TS 2.1.1 states that the specification applies to steady state operation. Explain the reason that the safety limits (SLs) apply only to steady-state operation. If the SLs also apply to reactor transients, revise TS 2.1.1 as appropriate. If the SLs do not apply to transients, proposed SLs in accordance with 10 CFR 50.36(c)(1)(i)(A) that apply to all reactor operations allowed by the proposed TS and all credible accidents. (See RAI 4.24) 14.33 The bases for TS 2.1.1 reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.34 Item 2 of the "Objective" section of TS 2.1.2 states, "To assure consistency with other defined safety system parameters." Explain the meaning of this statement, and revise the proposed TS as appropriate. 14.35 TS 2.1.2.1 specifies a SL of 217 kW for the true value of the reactor power during operation in the natural convection mode. The SL is based on preventing nucleate boiling in the hot channel. Table 4-17 of the SAR shows a negative margin to incipient boiling at 209.1 kW for the hot channel, which implies that incipient boiling occurs at a power level less than 209.1 kW. Explain this apparent inconsistency between the SL and Table 4-17, and revise the proposed TS as appropriate. (See RAI 4.29) 14.36 The "Applicability" section of TS 2.2.1 reads, "LEU Fuel Temperature - Forced Convection Mode." However, the "Specification" section of TS 2.2.1 gives limits for reactor thermal power, primary coolant flow through the core, height of water above the top of the core, and reactor coolant outlet temperature, and not fuel temperature. Explain this apparent inconsistency between the "Applicability" and "Specification" sections of TS 2.2.1, and revise the proposed TS as appropriate. 14.37 The "Objective" section of TS 2.2.1 appears to be both an applicability statement and an objective statement. Explain why the applicability statement is in the "Objective" section of TS 2.2.1, and revise the proposed TS as appropriate. 14.38 The "Objective" section of TS 2.2.1 contains the statement, "to assure that the maximum fuel temperature permitted is such that no damage to the fuel cladding will result in the forced convection mode." This statement appears to be inconsistent with the requirement of 10 CFR 50.36(c)(1)(ii)(A) that, "where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded." Additionally, TS 1.7 states that limiting safety system settings (LSSS) will be "chosen so that automatic protective action will correct an abnormal situation before a safety limit is exceeded," which appears to be inconsistent with the objective to limit fuel temperature. Explain these apparent inconsistencies, and revise the proposed TS as appropriate. 14.39 TS 2.2.1 gives LSSS for reactor thermal power, primary coolant flow through the core, height of water above the top of the core, and reactor coolant outlet temperature. TS 2.1.1 establishes SLs for these variables. 10 CFR 50.36(c)(1)(ii)(A) requires that, "where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded." Explain how the LSSS satisfy the requirement of 10 CFR 50.36. Include analyses, with fully justified assumptions, that show the LSSS prevent exceeding a SL for all operations allowed by the proposed TS and all credible accidents. Per 10 CFR 50.36(a)(1), these analyses shall be summarized and/or referenced in the bases for the LSSS. (See RAI 14.32) 14.40 The bases for TS 2.2.1 reference fuel temperature and fuel cladding temperature as though these parameters were the parameters for which the SLs were established. TS 2.1.1 does not establish SLs on fuel temperature or fuel cladding temperature.

TS 2.1.1 establishes SLs on reactor thermal power, reactor coolant flow through the core, reactor coolant outlet temperature, and height of water above the top of the core. Explain how the bases support each LSSS, and revise the proposed TS as appropriate. (See RAI 14.39) 14.41 The bases for TS 2.2.1 make multiple references to fuel temperature and fuel cladding temperature limits. If the intention is to have these limits be SLs for the RINSC reactor, revise the proposed TS as appropriate. 14.42 The bases for TS 2.2.1 state, "flow and temperature limits were chosen to prevent incipient boiling even if transient power rises to the 2 MW trip limit of 2.4 MW." However, the LSSS for reactor power specified by TS 2.2.1 is 2.3 MW. Explain this apparent inconsistency between the bases and the specification. 14.43 The bases for TS 2.2.1 state, "flow and temperature limits were chosen to prevent incipient boiling even if transient power rises to the 2 MW trip limit of 2.4 MW." However, Section 4.6.4 of the SAR states that during a rising power transient, the calculated fuel surface temperature would be above the onset of nucleate boiling temperature. Explain this inconsistency between the bases of TS 2.2.1 and the analysis in the SAR. 14.44 The bases for TS 2.2.1 include uncertainties associated with some of the LSSS parameters, but exclude reactor power and coolant height. Discuss the uncertainties associated with these parameters and explain how the uncertainties were incorporated into the analyses supporting the LSSS. (See RAI 4.20) 14.45 The bases for TS 2.2.1 state, "the LSSS for the pool level is set for a scram upon a 2 inch drop in water level." TS 2.2.1 specifies a LSSS of 23.7 feet, which is a true value, and not a magnitude of decrease in pool level. Explain this apparent inconsistency, and revise the proposed TS as appropriate. 14.46 The bases for TS 2.2.1 state, "the safety limit settings chosen provide acceptable safety margins to the maximum fuel cladding temperature." Explain the meaning of the phrase "safety limit settings." Provide quantitative values for the safety margins referred to as "acceptable safety margins," and explain the reasons they are considered acceptable. 14.47 The bases for TS 2.2.1 state, "the LSSS for the pool level results in a higher number since the pool level scrams upon a 2 inch drop in water level." Explain what "higher number" means in this context. 14.48 The bases for TS 2.2.1 contain the reference, "Report on the Determination of Hot Spot Factors for the RINSC Research Reactor, August 1989." Provide a copy of this reference. 14.49 The bases for TS 2.2.1 reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.50 The bases for TS 2.2.2 state, "the SAR has determined that up to 217 kW can be removed by natural convection." However, Table 4-17 of the SAR shows a negative margin to incipient boiling at 209.1 kW, which implies that incipient boiling occurs at a power level less than 209.1 kW. Explain this apparent inconsistency between the bases and Table 4-17 of the SAR. (See RAI 14.35) 14.51 The bases for TS 2.2.2 state, "with a 15% overpower trip, 115 kW will be the LSSS." This seems to be an arbitrary value with no supporting analysis or justification. Provide an analysis, with fully justified assumptions, that demonstrates the LSSS on reactor power will prevent a SL from being exceeded for all operations allowed by the proposed TS and all credible accidents. 14.52 The bases for TS 2.2.2 state, "the pool level scram (2 inch drop) is the same as the forced convection mode." TS 2.2.2 specifies a LSSS of 23.7 feet, which is a true value, and not a magnitude of decrease in pool level. Explain this apparent inconsistency, and revise the proposed TS as appropriate. 14.53 The bases for TS 2.2.2 state, "the pool temperature 130 F safety limit, having a 3% error, results in a LSSS of 126 F." Explain the basis for the 3 percent error. Provide an analysis, with fully justified assumptions, that demonstrates the LSSS on pool temperature will prevent a SL from being exceeded for all operations allowed by the proposed TS and all credible accidents. 14.54 The bases for TS 2.2.2 do not discuss uncertainties associated with reactor power and coolant height. Explain the uncertainties associated with these variables and explain how the uncertainties were treated in the analyses supporting the LSSS. Section 3.0, "Limiting Conditions for Operation" 14.55 ANSI/ANS-15.1 recommends technical specifications establish limits on fuel burnup. Explain the reason for not including such a specification, and revise the proposed TS as appropriate. 14.56 TS 3.1.1 requires the shutdown margin to be determined with the most reactive shim safety blade and the regulating blade fully withdrawn. The bases for TS 3.1.1 do not mention the position of the regulating blade. Explain this apparent inconsistency, and revise the proposed TS as appropriate. (See RAI 14.28) 14.57 The bases for TS 3.1.1 reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.58 The bases for TS 3.1.3 state that the limit on the reactivity worth of experiments prevents melting of the fuel. However, the SLs specified in TS 2.1 do not include fuel temperature. Explain how the LCO for the reactivity worth of experiments is consistent with the SLs, and revise the proposed TS as appropriate. 14.59 TS 3.1.4 does not include explicit reactivity limits for removable experiments. Explain which reactivity limit (movable or secured) applies to removable experiments or revise TS 3.1.4 to include an explicit reactivity limit for removable experiments. (See RAI 14.20) 14.60 TS 3.1.4 limits the reactivity worth of each movable experiment to 0.08 %k/k. Section 13.2.2 of the SAR appears to state that the total reactivity worth of all movable experiments is limited to 0.08 %k/k. Explain whether each movable experiment is limited to 0.08 %k/k, or whether the total reactivity worth of all movable experiments is limited to 0.08 %k/k. If the reactivity worth of each movable experiment is limited to 0.08 %k/k, explain whether multiple movable experiments could comprise the total experiment reactivity worth limit of 0.6 %k/k (e.g., ten movable experiments each with a reactivity worth of 0.06 %k/k). 14.61 The bases for TS 3.1.4 state that the individual reactivity worth of an experiment is limited to a value that will not produce a stable reactor period of less than 30 seconds. Explain whether this statement applies to all types of experiments. Provide an analysis that supports this statement, and revise the proposed TS as appropriate. 14.62 The bases for TS 3.1.4 state that the control and safety systems will protect the safety limits in the case that the reactivity associated with an experiment is inserted into the reactor. Section 13.2.2 of the SAR presents an analysis of an insertion of reactivity, but does not explicitly demonstrate that the LCO is chosen such that the LSSS will prevent the SLs from being exceeded. Provide analyses, including fully justified assumptions, that show the LCO is appropriately chosen so that the LSSS will prevent exceeding the SLs. (See RAI 13.9) 14.63 TS 3.1.5 requires the reactor to be subcritical by at least 3.0 %k/k during fuel loading changes. Explain how it is determined that the reactor is subcritical by at least 3.0 %k/k during fuel loading changes. Explain the reason for not specifying a surveillance requirement for this LCO, and revise the proposed TS as appropriate. 14.64 TS 3.1.6 limits the reactivity worth of the regulating blade. The proposed TS do not appear to specify surveillance requirements for the reactivity worth of the regulating blade. Explain the reason for not specifying a surveillance requirement for the reactivity worth of the regulating blade, and revise the proposed TS as appropriate. 14.65 TS 3.1.7 states, "Experiments which could increase reactivity by flooding, shall not remain in or adjacent to the core unless the shutdown margin required in Specification 3.1.1 would be satisfied after flooding." Explain why experiments that could reduce the shutdown margin below 1.0 %k/k by flooding would ever be allowed in or adjacent to the core, and revise the proposed TS as appropriate. (See RAI 4.14) 14.66 TS 3.1.8 states "surveillance will be conducted at initial startup and change in fuel type." Explain the reason that this surveillance requirement is included in the LCO, and revise the proposed TS as appropriate. 14.67 TS 3.1.9 specifies core configuration requirements for operation in the forced convection mode. Explain why the TS do not contain any similar core configuration requirements for operation in the natural convection mode. Explain why the proposed TS do not restrict core configurations to the three core configurations referenced in TS 4.1.b. (See RAI 14.134) 14.68 TS 3.2.1 specifies reactor safety systems and safety-related instrumentation that are required for critical reactor operation. However, the proposed TS do not contain any requirements for reactor safety systems and safety-related instrumentation that must be operable when the reactor is subcritical, but not secured. Explain why the proposed TS do not require any operable safety systems or safety-related instrumentation when the reactor is subcritical, but not secured (e.g., movement of fuel in the reactor core). Explain why the radiation monitors listed in Table 3.2 are not required during work of the types specified in TS 1.19.1.c and TS 1.19.1.d. Revise the proposed TS as appropriate. 14.69 TS 3.2.1, Table 3.1 contains a column labeled "Function" that appears to contain both the function of each safety channel and the set point. As written, it is difficult to understand if the set points are maximum or minimum set points. For example, the "Function" column states "automatic scram at 1600gpm" for the coolant flow rate safety channel. This implies the scram set point can be any value less than or equal to 1600 gpm. However, the LSSS for coolant flow rate is 1600 gpm, which means that any set point less than 1600 gpm would be inconsistent with the LSSS. Other examples are reactor power level, coolant outlet temperature, log N period, and pool temperature. Revise Table 3.1 to clearly state the maximum and minimum set points for the safety channels, and ensure the set points are consistent with the LSSS. 14.70 TS 3.2.1, Table 3.1 states that the function of the reactor power level safety channel is "automatic scram when 115% of range scale with 2.3 MW max," and this is required in both forced and natural convection operating modes. Explain how a maximum reactor power trip setting of 2.3 MW in the natural convection mode of operation is consistent with the LSSS of 115 kW specified by TS 2.2.2, and revise the proposed TS as appropriate. What is the range scale of the reactor power level safety channels? Can the scram functions be disabled by increasing the range scale? Are there scram set points at 115 kW and 2.3 MW that are independent of the channel range scale? 14.71 TS 3.2.1, Table 3.1 requires a bridge misalignment safety channel and a bridge movement safety channel. The "Function" column of Table 3.1 does not contain set points for these channels. Explain the reason that Table 3.1 does not specify set points for these channels, and revise the proposed TS as appropriate. (See RAI 7.3) 14.72 TS 3.2.1, Table 3.1 requires a pool water level safety channel with a set point at 16 inches below the suspension frame base plate elevation. TS 2.2.1 gives the LSSS for pool water level as 23.7 feet. Explain why Table 3.1 and the LSSS use different frames of reference and different units for the pool water level safety channel set point. Explain how the LCO is consistent with the LSSS, and revise the proposed TS as appropriate. 14.73 TS 3.2.1, Table 3.1 requires three detector high voltage failure safety channels. The "Function" column of Table 3.1 states, "automatic scram if Voltage decreases 50V max." Explain what "Voltage decreases 50V max" means. 14.74 TS 3.2.1, Table 3.1 requires a no flow thermal column safety channel when the reactor is operated above 100 kW in the forced convection mode. The table does not specify a set point for the safety channel and the SAR does not specify what flow rate is necessary to remove the heat generated in the graphite in the thermal column. Explain why there is no set point for the safety channel. (See RAI 10.3) 14.75 TS 3.2.1, Table 3.2, items 1 and 2 contain the acronym "FC." Define this acronym, and revise the proposed TS as appropriate. 14.76 TS 3.2.1, Table 3.2 requires a log count rate blade withdrawal interlock with a set point less than 3 counts per second. Explain why a set point less than 3 counts per second (e.g., a set point of 0 counts per second) is appropriate for this safety-related instrument, and revise the proposed TS as appropriate. 14.77 TS 3.2.1, Table 3.2 requires a servo control interlock with a set point of "30 sec (fullout)." What is the parameter to which the "30 sec" set point applies? What is the component to which the "fullout" set point applies? Revise the proposed TS as appropriate. 14.78 TS 3.2.1, Table 3.2 requires a building air gaseous exhaust (stack) monitor with a set point of "2.5 x normal particulate 2 x normal." Explain what this set point means. Clarify whether this single monitor fulfills the functions of monitoring both particulates and gaseous effluents, and revise the proposed TS as appropriate. (See RAI 14.103) 14.79 TS 3.2.1, Table 3.2, item 10 requires a radiation monitor labeled "primary demineralizer (hot DI)." Explain what "hot DI" means, and revise the proposed TS as appropriate. 14.80 TS 3.2.1, Table 3.2 contains footnote (b) which states, "The reactor shall not be continuously operated without a minimum of one radiation monitor on the experimental level of the reactor building and one monitor over the reactor pool operating and capable of warning personnel of high radiation levels." Explain what "continuously operated" means. Explain why the radiation monitors subject to footnote (b) do not need to be operating for reactor operations that are not considered "continuous." Explain how each radiation monitor located on the experimental level can individually provide adequate monitoring of the entire experimental level. 14.81 TS 3.2.2 requires all shim safety blades to be operable before the reactor is made critical. Explain why the regulating blade is not required to be operable before the reactor is made critical. 14.82 TS 3.2.2 references the surveillance requirements of TS 4.1.1 and TS 4.1.2. Explain why TS 3.2.2 references these surveillance requirements. 14.83 TS 3.2.3 references the surveillance requirements of TS 4.2.5 and TS 4.2.6 (the reference to TS 4.2.6 appears to be incorrect). Explain why TS 3.2.3 references these surveillance requirements. 14.84 TS 3.2.4 appears to be a reactivity limit. Explain the reason for not including TS 3.2.4 in TS 3.1. 14.85 TS 3.2.4 specifies a maximum reactivity insertion rate for a single control or regulating blade of 0.02 %k/k per second. The bases for the TS state that the reactivity insertion rate limit was determined in the SAR, but the SAR does not appear to contain an analysis of a ramp insertion of 0.02 %k/k per second. Section 13.2.5 provides an analysis of a startup accident, but the analyzed reactivity addition rate (0.0196 %k/k per second) appears to be less conservative than the TS limit. Explain how the SAR supports the reactivity insertion rate limit in TS 3.2.4. If the SAR does not support the TS limit, provide an analysis that supports the TS limit. Alternately, revise the proposed TS to be consistent with the analysis in the SAR. 14.86 The bases for TS 3.2.1 state, "the period scram limits the rate of rise of the reactor power to periods which are manually controllable." Table 3.1 indicates that the Log N Period trip channel set point is 4 seconds. The SAR does not appear to contain an analysis that shows how a reactor period slightly greater than 4 seconds would be manually controllable. Explain how a reactor period slightly greater than 4 seconds is manually controllable by the reactor operator. 14.87 The bases for TS 3.2 only discuss the reactor power, reactor period, and coolant flow scrams required by TS 3.2.1. Provide bases for the other safety channels and safety-related instrumentation required by TS 3.2.1, Table 3.1 and Table 3.2. 14.88 The bases for TS 3.2 do not provide bases for TS 3.2.2 and TS 3.2.3. Provide bases for these TS. 14.89 The bases for TS 3.2 reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.90 TS 3.3.a.3 appears to be a surveillance requirement and not a limiting condition for operation. Explain why TS 3.3.a does not specify a limit for primary coolant water radioactivity, and revise the proposed TS as appropriate. 14.91 The "Applicability" section of TS 3.3.b includes cycles of chloride and resistivity. TS 3.3.b does not contain any specifications related to these parameters. Explain this apparent inconsistency, and revise the proposed TS as appropriate. 14.92 TS 3.3.b.2 appears to be a surveillance requirement and not a limiting condition for operation. Explain why TS 3.3.b.2 does not specify a limit for sodium-24 in the secondary coolant, and revise the proposed TS as appropriate. 14.93 The proposed TS contain TS 3.4, 3.5, 3.6, "Confinement and Emergency Exhaust System and Emergency Power." The proposed TS is difficult to understand because it combines the requirements for three systems into one specification without clearly stating the requirements for each system. Explain the reason for combining all of these requirements into one specification, and explain the reason for the multiple numbers in the title of the TS. Revise the proposed TS to either separate the limiting conditions for operation (LCOs) for the three systems into three separate TS, or revise the proposed TS to clearly state the requirements for each of the three systems. 14.94 The "Applicability" and "Objective" sections of TS 3.4, 3.5, 3.6 mention fuel handling, handling of radioactive material, and any operation that could cause the spread of airborne radioactivity in the confinement area. The "Specification" section only contains requirements for reactor operation. Explain why the TS does not contain requirements for fuel handling, handling of radioactive material, and any operation that could cause the spread of airborne radioactivity in the confinement area. Revise the proposed TS as appropriate. 14.95 The "Specification" section of TS 3.4, 3.5, 3.6 states, "the reactor shall not be operated unless the following equipment is operable and/or conditions met." Explain the reason for using the "and/or" condition in the specification, and revise the proposed TS as appropriate. 14.96 TS 3.4, 3.5, 3.6 does not appear to contain any requirements for normal ventilation during reactor operation, fuel handling, handling of radioactive material, and any operation that could cause the spread of airborne radioactivity in the confinement area. Explain why there are no requirements for normal ventilation, and revise the proposed TS as appropriate. 14.97 TS 3.4, 3.5, 3.6 does not appear to contain any requirements for ventilation flow rates for normal ventilation or the emergency exhaust system. Explain why there are no requirements for normal ventilation or the emergency exhaust system flow rates, and revise the proposed TS as appropriate. 14.98 TS 3.4, 3.5, 3.6 requires the emergency cleanup exhaust system to be operable during reactor operation, but does not specify what constitutes operability of the system. Explain what constitutes operability of the emergency cleanup exhaust system (e.g., minimum required equipment, filtration requirements, etc.), and revise the proposed TS as appropriate. 14.99 TS 3.4, 3.5, 3.6 requires that the function of the emergency generator is "to insure power systems and other designated systems." To what "power system" does this refer? What are the "other designated systems" referenced in the function statement? Explain the reason for not specifying what equipment is required to be powered by the emergency generator. Explain why there are no LCOs regarding what constitutes operability of the emergency generator (e.g., type of generator, minimum operating time, etc.), and revise the proposed TS as appropriate. 14.100 The bases for TS 3.4, 3.5, 3.6 appear to only contain bases for operation of the emergency exhaust system. Provide bases for normal operation of the confinement and the requirements for emergency power. 14.101 ANSI/ANS-15.1 recommends technical specifications include the minimum number, type, and location of required environmental radiation monitors. Section 11.1.7 of the SAR discusses environmental monitoring at the RINSC. Explain the reason for not including such requirements, and revise the proposed TS as appropriate. 14.102 The "Applicability" section of TS 3.7.1 mentions fuel movement and handling of radioactive materials in the reactor building, but the specification only specifies requirements for reactor operation. Explain why there are no requirements for radiation monitoring systems during fuel movement and handling of radioactive materials in the reactor building, and revise the proposed TS as appropriate. 14.103 TS 3.7.1.1 states, "The particulate activity monitor and the gaseous activity monitor for the facility exhaust stack shall be operating." TS 3.2.1, Table 3.2, item 5 only requires one building air gaseous exhaust (stack) monitor, and does not require a separate particulate monitor. Explain this apparent inconsistency, and revise the proposed TS as appropriate. (See RAI 14.78) 14.104 TS 3.7.1.1 states, "The particulate activity monitor and the gaseous activity monitor for the facility exhaust stack shall be operating." This statement does not specify when the monitors are required to be operating. Explain why the TS does not specify when the monitors are required to be operating, and revise the proposed TS as appropriate. 14.105 TS 3.7.1.1 specifies that the reactor may be operated for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without either a particulate activity monitor or a gaseous activity monitor. Explain the basis for operating the reactor for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without particulate effluent activity detection capability. Explain the basis for operating the reactor for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without gaseous effluent activity detection capability. 14.106 TS 3.7.1.2 allows the reactor to be operating for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without the continuous air monitoring unit required by TS 3.2.1, Table 3.2, item 11. TS 3.2.1 states that the reactor shall not be made critical unless the unit is operating. Explain this apparent inconsistency between TS 3.7.1.2 and TS 3.2.1, and revise the proposed TS as appropriate. 14.107 The footnote to TS 3.7.1.3 states that the reactor may be operated in a steady-state power mode if an area radiation monitor or the reactor bridge radiation monitor is replaced with a portable gamma-sensitive monitor with its own alarm. How long can the reactor be operated with a portable monitor performing the function of an area radiation monitor, and why? How does the portable instrument notify the reactor operator of changing radiation conditions? Given that TS 3.2.1, Table 3.2 only requires one operating radiation monitor on the experimental level, explain how a single portable monitor provides adequate detection capability to monitor radiation conditions on the entire experimental level. 14.108 The "Bases" section of TS 3.7.1 does not provide bases for the stack effluent monitors. Provide bases for the stack effluent monitors. 14.109 The "Objective" section of TS 3.7.2.a states, "To assure containment integrity is maintained during reactor operation..." Explain what "containment integrity" means. Explain how TS 3.7.2 "assures containment integrity." 14.110 TS 3.7.2.a.1 limits the concentration of radioactive materials in the effluent released from the facility exhaust stack to 10E5 times the air effluent concentration limits in 10 CFR 20. The bases state that the limit incorporates a dilution factor of 4x10E4. Given that the release concentration limit is greater than the dilution factor, explain how the dilution factor ensures that off-site concentrations of radioactive materials will be below the air effluent concentration limits in 10 CFR 20. 14.111 Explain how TS 3.7.2.a.1 ensures that airborne effluents released from the RINSC will satisfy the ALARA dose constraint of 10 CFR 20.1101(d). 14.112 The "Bases" section of TS 3.7.2.a references a letter sent to the NRC in 1963. 10 CFR 50.36 requires that the proposed TS be derived from analysis included in the SAR. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.113 The first sentence of TS 3.7.2.b states "The liquid waste retention tank discharge shall be batch sampled and the gross activity per unit volume determined before release." This statement appears to be a surveillance requirement and redundant to the requirement specified in TS 4.7.b.2. Explain the reason for including this requirement as an LCO, and revise the proposed TS as appropriate. 14.114 TS 3.7.2.b states, "All off-site releases shall be directed into the municipal sewer system." The bases state that liquid wastes can be removed from the site by a commercial licensed organization. Explain this apparent inconsistency between the specification and the bases, and revise the proposed TS as appropriate. 14.115 TS 3.7.2.b does not contain requirements for the concentration of radioactivity in liquid wastes that can be discharged from the RINSC site. Explain why TS 3.7.2.b does not contain any such requirements, and revise the proposed TS as appropriate. 14.116 ANSI/ANS-15.1 recommends that the technical specifications specify that experiments will be designed such that they do not contribute to the failure of other experiments or reactor systems and components important to safety. Explain the reason that the proposed TS do not contain any such requirement for experiments, and revise the proposed TS as appropriate. 14.117 TS 3.8.3 states, "Fissionable materials shall have total iodine and strontium inventory less than that allowed by the facility by-product license." What facility by-product license does this specification reference? What inventory limits does that by-product license specify? Why are iodine and strontium the only elements of concern for experiments involving fissionable materials? Provide an analysis of the consequences of the failure of an experiment involving fissionable materials that shows the consequences are bounded by the analysis of the MHA presented in Chapter 13 of the SAR. Discuss all assumptions used in the analysis, including justification for the use of the assumptions. 14.118 TS 3.8.5 states, "experiments shall be designed against failure from internal and external heating at the true values associated with the LSSS for reactor power level and other process variables." ANSI/ANS-15.1 recommends that experiments also be able to withstand reactor transients. Section 4.6.4 of the SAR states that a rising power transient could result in a maximum reactor power of 2.78 MW, which is greater than the LSSS value of 2.30 MW. Explain how TS 3.8.5 ensures that experiments will be designed to withstand reactor transients, and revise the proposed TS as appropriate. 14.119 The requirements of TS 3.8.10 imply that accidents involving experiments could result in occupational and public radiation doses up to the regulatory limits. These doses would be greater than the consequences of a fuel failure accident analyzed in Section 13.2.1 of the SAR. Explain why the SAR considers the fuel failure accident to be the MHA if the failure of an experiment could have greater consequences. Provide an analysis of the occupational and public dose consequences of the worst-case failure of an experiment that is consistent with the requirements of the proposed TS. Discuss all assumptions used in the analysis, including justification for the use of the assumptions. 14.120 TS 3.8.10 contains requirements related to occupational and public radiation doses resulting from experiments. The specification states, "Experimental materials... which could off-gas... under: (1) normal operating conditions of the experiment... shall be limited in activity such that: if 100% of the gaseous activity or radioactive aerosols produced escaped to... the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the occupational limits for maximum permissible concentration." Explain the reason for allowing normal operation of experiments to result in off-site concentrations of radioactive materials up to the regulatory limits. Does this requirement pertain to the sum of all experiments or individual experiments? Explain how this requirement ensures compliance with the ALARA dose constraint of 10 CFR 20.1101(d). (See RAI 14.123) 14.121 TS 3.8.10 specifies requirements related to failure of an experiment encapsulation. Explain what specific types of encapsulation are covered by TS 3.8.10, and revise the proposed TS as appropriate. 14.122 It appears that the first sentence of the second paragraph of TS 3.8.10 explicitly excludes "fuel materials" from the requirements of TS 3.8.10. Clarify whether "fuel materials" is synonymous with "fissionable materials" as used in TS 3.8.3. If the requirements of TS 3.8.10 exclude fissionable materials, explain the reason for not including similar requirements for experiments that contain "fuel materials," and revise the proposed TS as appropriate. 14.123 The second paragraph of TS 3.8.10 states, "if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the occupational limits for maximum permissible concentration." Revise the proposed TS to use current 10 CFR Part 20 terminology (e.g., Annual Limit on Intake or Derived Air Concentration). Explain why occupational concentration limits are used as limits for the release of radioactive material to the atmosphere, and revise the proposed TS as appropriate. 14.124 The third paragraph of TS 3.8.10 contains assumptions used to calculate releases of radioactive material from experiment malfunctions. These assumptions do not appear to be derived from analyses in the SAR and the bases for TS 3.8 state that the specification is "self explanatory." Provide discussions and/or analyses that explain the assumptions required by TS 3.8.10. Revise the proposed TS as appropriate. 14.125 The third paragraph of TS 3.8.10 states, "Limits for maximum permissible concentrations are specified in the appropriate section of 10CFR20." Revise the proposed TS to use current 10 CFR Part 20 terminology and to be more specific about the section of 10 CFR Part 20 that applies to TS 3.8.10. 14.126 The bases for TS 3.8 state that several of the specifications are "self explanatory." In accordance with 10 CFR 50.36, provide bases for all of the specifications in TS 3.8. 14.127 TS 3.9.a.1 sets a limit of 1x10E22 neutrons per square centimeter on the accumulated flux for the beryllium reflectors. The SAR does not appear to contain an analysis that supports the flux limit. Provide an analysis of the flux limit for the beryllium reflectors. 14.128 The bases for TS 3.9.a reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.129 TS 3.9.b appears to be a surveillance requirement and not a LCO on the physical condition of the fuel. Explain why the TS do not specify an LCO on the physical condition of the fuel, and revise the proposed TS as appropriate. Section 4.0, "Surveillance Requirements" 14.130 TS 4.0 specifies that some surveillance requirements may be deferred during periods of reactor shutdown. As recommended in ANSI/ANS-15.1, allowed deferral of a surveillance requirement should be specified as part of the surveillance requirement.

Each surveillance requirement that may be deferred during reactor shutdown must specify whether the surveillance must be completed prior to reactor operation. Each allowed deferral must be supported by a basis statement that explains the reason deferral is warranted during reactor shutdown. Revise the proposed TS as appropriate. 14.131 TS 4.1.1 requires measurement of shim blade insertion rates. Explain the reason for not requiring measurement of shim blade withdrawal rates, and revise the proposed TS as appropriate. 14.132 TS 4.1.1 does not require surveillance of the shim safety blades following maintenance or replacement. Explain the reason for not requiring surveillance of the shim safety blades following maintenance or replacement, and revise the proposed TS as appropriate. 14.133 TS 4.1.1.b references the startup core and three other analyzed cores. Explain the reason for referencing the startup core, and revise the proposed TS as appropriate. 14.134 TS 4.1.1.b implies that there are only three allowed core configurations for the RINSC reactor. The proposed TS do not contain an LCO restricting the configuration of the RINSC core to three configurations. Explain the reason for only requiring surveillance of the shim safety blades when switching to one of the three referenced core configurations, and revise the proposed TS as appropriate. 14.135 TS 4.1.1.b references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS. 14.136 TS 4.1.2 requires inspection of the shim safety blades to detect swelling. The bases for TS 4.1.2 state that inspection will detect swelling and cracking. Explain this apparent inconsistency between the specification and the bases, and revise the proposed TS as appropriate. 14.137 TS 4.1.3 requires measurement of an experiment's reactivity worth prior to the "initial use" of the experiment. The bases for TS 4.1.3 state, "The specified surveillance relating to the reactivity worth of experiments will assure that the reactor is not operated for extended periods before determining the reactivity worth of experiments." The specification and bases imply that the reactor can be operated without determining the reactivity worth of experiments. Explain how TS 4.1.3 ensures that the experiment reactivity requirements of TS 3.1 are met, and revise the proposed TS as appropriate. 14.138 The bases for TS 4.1.3 state that the specification "provides assurance that experiment reactivity worths do not increase beyond the established limits due to core configuration changes." The specification does not appear to require any surveillance of experiment reactivity worths following core configuration changes. Explain the apparent inconsistency between the specification and the bases, and revise the proposed TS as appropriate. 14.139 ANSI/ANS-15.1 recommends annual thermal power verification. Explain the reason that the proposed TS do not contain any such requirement, and revise the proposed TS as appropriate. 14.140 ANSI/ANS-15.1 recommends annual surveillance of required interlocks. Explain the reason that the proposed TS do not contain any such requirements, and revise the proposed TS as appropriate. 14.141 TS 4.2 specifies surveillance requirements for the safety system and safety-related instrumentation required by TS 3.2.1. However, the proposed TS do not specify surveillance requirements for many of the items required by TS 3.2.1, Table 3.1 and Table 3.2. In accordance with 10 CFR 50.36(c)(3), propose surveillance requirements for the safety system and safety related instrumentation required by TS 3.2.1. 14.142 TS 4.2.1.a requires channel tests of nuclear instrumentation "prior to each reactor startup following a period when the reactor was secured." Given that the TS do not require the reactor to be secured on a periodic basis, explain the reason for not requiring periodic (e.g., quarterly) surveillance of the nuclear instrumentation, and revise the proposed TS as appropriate. 14.143 TS 4.2.2 states, "A channel calibration of the safety channels listed in Table 3.1, which can be calibrated, shall be performed annually." Revise the proposed TS to explicitly state which channels listed in Table 3.1 will be calibrated annually. 14.144 TS 4.2.3 appears to be an LCO and not a surveillance requirement. Explain the reason for including TS 4.2.3 in the surveillance requirements, and revise the proposed TS as appropriate. 14.145 TS 4.2.6 does not require surveillance of the shutdown margin following changes in control blades. Explain the reason for not requiring surveillance of the shutdown margin following control blade changes, and revise the proposed TS as appropriate. 14.146 TS 4.2.6 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS. 14.147 TS 4.2.7 does not require surveillance of the excess reactivity following changes in control blades. Explain the reason for not requiring surveillance of the excess reactivity following control blade changes, and revise the proposed TS as appropriate. 14.148 TS 4.2.7 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS. 14.149 TS 4.2.8 does not require surveillance of the reactivity insertion rate following changes in control blades. Explain the reason for not requiring surveillance of the reactivity insertion rate following control blade changes, and revise the proposed TS as appropriate. 14.150 TS 4.2.8 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS. 14.151 The "Bases" section of TS 4.2 does not contain bases for TS 4.2.6, 4.2.7, or 4.2.8. Provide bases for these specifications. 14.152 The bases for TS 4.2.3 states, "Radiation monitors are checked for proper operation in Specification 4.2.3. Calibration and setpoint verification involve..." However, TS 4.2.3 appears to be an LCO and does not specify surveillance requirements (e.g., channel tests, channel checks, or channel calibrations). Explain this apparent inconsistency between the specification and the bases for TS 4.2.3, and revise the proposed TS as appropriate. 14.153 The second paragraph of the bases for TS 4.3.a appears to be a description of the pool level detection system, not the bases for the proposed surveillance requirements. In accordance with 10 CFR 50.36, provide bases that explain the reasons for the requirements of TS 4.3.a.4, 4.3.a.5, and 4.3.a.6. 14.154 The "Bases" section of TS 4.3.b appears to be a description of how secondary coolant chemistry is controlled and how secondary coolant radioactivity is monitored, not the bases for the proposed surveillance requirements. In accordance with 10 CFR 50.36, provide bases that explain the reasons for the requirements of TS 4.3.b.1 and 4.3.b.2. 14.155 ANSI/ANS-15.1 recommends surveillance of required ventilation filters. Explain the reason that the proposed TS do not contain any such requirements, and revise the proposed TS as appropriate. 14.156 The first sentence of Specification 1 of TS 4.4, 4.5, 4.6 appears to be a surveillance requirement. The rest of Specification 1 appears to be a combination of a description of system operation and LCOs for the confinement and emergency exhaust systems (e.g., maximum emergency cleanup system flow rate, minimum differential pressure, etc.). Revise Specification 1 to include only surveillance requirements and relocate any LCOs to the appropriate sections of the proposed TS. 14.157 Specification 2.a of TS 4.4, 4.5, 4.6 requires inspection of "building ventilation blowers and dampers (including solenoid valves, pressure switches, piping, etc.)" Revise the proposed TS to explicitly state each piece of equipment that must be inspected. 14.158 Specification 2.b of TS 4.4, 4.5, 4.6 requires inspection of personnel access and reactor room overhead doors. Explain why the specification does not require inspection of the truck door, and revise the proposed TS as appropriate. 14.159 Specification 3 of TS 4.4, 4.5, 4.6 does not contain enough detail regarding the testing frequency of the emergency generator. Revise the proposed TS to include the maximum surveillance interval for testing the emergency generator. Describe the tests that comprise the emergency generator testing, and revise the proposed TS as appropriate. 14.160 ANSI/ANS-15.1 recommends technical specifications include surveillance requirements for radiation monitoring at site boundary and environmental monitoring. Section 11.1.7 of the SAR discusses environmental monitoring at the RINSC. Explain the reason for not including such surveillance requirements, and revise the proposed TS as appropriate. 14.161 TS 4.7.a.1 requires annual calibration of the particulate air monitors. The LCOs specified in TS 3.2.1, Table 3.2 do not appear to contain a requirement for particulate air monitors. Explain this apparent inconsistency, and revise the proposed TS as appropriate. (See RAI 14.103) 14.162 TS 4.7.a.3 requires a daily channel check of the "main floor monitor." The TS do not appear to contain an LCO for a "main floor monitor." Revise TS 4.7.a.3 to use terminology for radiation monitors consistent with the terminology for radiation monitors required by TS 3.2.1, Table 3.2 or TS 3.7.1, or propose an LCO for a "main floor monitor." 14.163 The bases for TS 4.8 state, "Review of the experiments... assures that the insertion of experiments will not negate the consideration implicit in the Safety Limits." Explain what "consideration implicit in the Safety Limits" means in terms of experiments. 14.164 Since the application for license renewal was submitted, TS 4.9 was amended by Amendment No. 29 to Facility Operating License No. R-95, dated December 28, 2004. Clarify whether the amended TS 4.9 that is currently in the license should replace proposed TS 4.9 contained in the application for license renewal. 14.165 The bases for TS 4.9.a reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. 14.166 The bases for TS 4.9.b state, "The fission density limit for this reactor cannot be exceeded." The proposed TS do not appear to contain a fission density limit for the fuel. Explain the reason for not including a fission density limit for the fuel. (See RAI 14.55) 14.167 The bases for TS 4.9.b state, "Burnup calculations are made quarterly (4.9.1)." To what does "(4.9.1)" refer? Explain the reason that the burnup calculations are not a required surveillance, and revise the proposed TS as appropriate. 14.168 The bases for TS 4.9.b reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended. Section 5.0, "Design Features" 14.169 In accordance with 10 CFR 50.36(a)(1), provide bases for proposed technical specifications in Section 5, "Design Features." 14.170 ANSI/ANS-15.1 recommends that the number and type of control blades be included in the technical specifications. Explain the reason that the regulating blade is not specified in TS 5.3. Explain the reason that the control blade materials are not specified in the proposed TS, and revise the proposed TS as appropriate. 14.171 Proposed TS 5.3 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced section of the SAR a requirement in the proposed TS. 14.172 TS 5.4 appears to contain LCOs for the emergency cleanup system (e.g., filter requirements). Explain the reason for including these LCOs as part of the design features of the reactor building, and revise the proposed TS as appropriate. 14.173 TS 5.4 states, "The reactor building exhaust blower operates in conjunction with additional exhaust blower(s) which provide dilution air from non-reactor building sources." Clarify whether this statement applies to normal ventilation, emergency cleanup system operation, or both. 14.174 TS 5.4 mentions dilution air from non-reactor building sources. Explain the reason for not establishing a quantitative LCO for dilution air, and revise the proposed TS as appropriate. 14.175 TS 5.4 mentions exhaust air from the reactor building. Explain the reason for not establishing a quantitative LCO for exhaust air, and revise the proposed TS as appropriate. 14.176 Proposed TS 5.5 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license.

Explain why it is necessary to make the referenced section of the SAR a requirement in the proposed TS. Section 6.0, "Administrative Controls" 14.177 Any RAIs related to administrative controls will be provided in a future letter. The following RAI relates to financial qualifications.

1. Exhibit 3, "Summary of Decommissioning Calculations," of the supplement to the application dated January 19, 2010, provided a 20-year SAFSTOR scenario for the Rhode Island Nuclear Science Center (RINSC), with three columns listed for "Assumed Decom escalation factor," with a 5 percent escalation factor for "Management Maintenance & Supervision," "Assumed Discount Rate Factor," with a 2.85 percent discount rate factor, and "Assumed Discount Rate Factor," with a 0 percent (no discounting) discount rate factor. As discussed during the phone conversation of March 3, 2010, the NRC staff requires the following supplemental information to the Rhode Island Atomic Energy Commission Request for Additional Information response dated January 19, 2010:

(a) Clarify what the costs in "Management Maintenance & Supervision

" represent and whether the annual costs associated with SAFSTOR are accounted for in the analysis. Also, identify the basis for the use of the $800,000 base cost in year 1.

(b) The RAI response states that

"[a] figure of 25% was used beginning in year 4 and continuing through year 20," however the NRC staff notes that the figure in year 4 is not 25 percent of the figure from year 2. Clarify this inconsistency.

(c) Explain how the $5,866,092 in year 1 of Column 2, Exhibit 3, was determined. Also, provide a numerical example showing how this number was determined.

(d) Explain the method used to perform discounting in Column 2, Exhibit 3. Also, provide a numerical example showing the method used to perform discounting using the 2.85 percent discount rate factor.