ML070440043

Request for Additional Information Regarding the National Institute of Standards and Technology Test Reactor Application for Re-Licensing

ML070440043
Person / Time
Site:	National Bureau of Standards Reactor
Issue date:	02/13/2007
From:	Mendonca M NRC/NRR/ADRA/DPR/PRTB
To:	Richards W US Dept of Commerce, National Institute of Standards & Technology (NIST)
Mendonca M, NRC/NRR/DPR/PRT, 415-1128
References
TAC MD3410
Download: ML070440043 (24)

Text

February 13, 2007Dr. Wade Richards, Manager of Operations and EngineeringNIST Center for Neutron Research National Institute of Standards and Technology U.S. Department of Commerce 100 Bureau Drive, Mail Stop 8561 Gaithersburg, MD 20899-8561

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONALINSTITUTE OF STANDARDS AND TECHNOLOGY TEST REACTOR APPLICATION FOR RE-LICENSING (TAC NO. MD3410)

Dear Dr. Richards:

We are continuing our review of your application for re-licensing of the National Institute ofStandards and Technology (NIST) test reactor dated April 9, 2004, and supplemented on October 2, 2006. After reviewing your submissions we have determined that additional information is needed. During a discussion with you on February 13, 2007, you agreed to provide a response to the enclosed Request for Additional Information no later than May 31, 2007. Your timely response is needed to support completion of the review. In accordance with 10 CFR 50.30(b), your response must be executed in a signed original under oath or affirmation.Should you have any questions regarding this review, please contact William B. Kennedy, at(301) 415-2784 or me, at (301) 415-1128.Sincerely,/RA/Marvin M. Mendonca, Senior Project ManagerResearch and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-184

Enclosure:

As stated cc w/enclosure: Please see next page National Institute of Standards and TechnologyDocket No. 50-184 cc:

Environmental Program Manager III Radiological Health Program Air & Radiation Management Adm.

Maryland Dept of the Environment 1800 Washington Blvd.

Suite 750 Baltimore, MD 21230-1724Director, Department of State Planning301 West Preston Street Baltimore, MD 21201Director, Air & Radiation Management Adm.Maryland Dept of the Environment 1800 Washington Blvd., Suite 710 Baltimore, MD 21230Director, Department of Natural ResourcesPower Plant Siting Program Energy and Coastal Zone Administration Tawes State Office Building Annapolis, MD 21401President, Montgomery County Council100 Maryland Avenue Rockville, MD 20850Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 February 13, 2007Dr. Wade Richards, Manager of Operations and Engineering NIST Center for Neutron Research National Institute of Standards and Technology U.S. Department of Commerce 100 Bureau Drive, Mail Stop 8561 Gaithersburg, MD 20899-8561

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONALINSTITUTE OF STANDARDS AND TECHNOLOGY TEST REACTOR APPLICATION FOR RE-LICENSING (TAC NO. MD3410)

Dear Dr. Richards:

We are continuing our review of your application for re-licensing of the National Institute ofStandards and Technology (NIST) test reactor dated April 9, 2004, and supplemented on October 2, 2006. After reviewing your submissions we have determined that additional information is needed. During a discussion with you on February 13, 2007, you agreed to provide a response to the enclosed Request for Additional Information no later than May 31, 2007. Your timely response is needed to support completion of the review. In accordance with 10 CFR 50.30(b), your response must be executed in a signed original under oath or affirmation.Should you have any questions regarding this review, please contact William B. Kennedy, at(301) 415-2784 or me, at (301) 415-1128.Should you have any questions regarding this review, please contact William B. Kennedy, at(301) 415-2784 or me at (301) 415-1128.Sincerely,/RA/Marvin M. Mendonca, Senior Project Manager Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-184

Enclosure:

As stated cc w/enclosure: Please see next pageDISTRIBUTION

PUBLICDPR/PRT r/fRidsNrrDprPrtaGHill (2)ACCESSION NO.: ML070440043TEMPLATE No.: NRR-106OFFICEPRTA:GEPRTA:LAPRTA:BCPRTBL:PMNAMEWKennedy:cahEHyltonDCollinsMMendonca DATE 2/13/07 02/13/07 02/13/07 02/13/07 OFFICIAL RECORD COPY EnclosureREQUEST FOR ADDITIONAL INFORMATIONREGARDING RE-LICENSING OF THENATIONAL BUREAU OF STANDARDS REACTOR DOCKET NO. 50-184Technical Questions and Comments2.1The SAR text indicates that the 100-year return wind speed of 102.5 mph is within theuncertainty limits of the 100 mph design of the Confinement Building. The 102.5 mph value is calculated based on the 90 mph 50-year return gust taken from ASCE 7-98.

However, virtually the entire country away from the coastline is rated with a 90 mph gust level. In all likelihood, a more appropriate value for the 50-year return wind speed is somewhat lower, and as a result, the 100-year return wind speed would be lower as well. In this discussion (Section 2.3.1.5, Table 2.13, and Figure 2.7) provide a more refined estimate of the 100-year return wind speed, which should be less than the design value.2.2The discussion in SAR Section 2.3.1.6 regarding snow density references a publicationof the American Meteorological Society with a range of densities of 0.07 to 0.15. Clarify the text to reflect that this range of densities is for freshly fallen snow. Verify that the correct date of the reference is 1989, and make any necessary corrections to the text and references section.

____________________________________________________________________________4.1Section 4.2.1.1, p. 4-3. In the "Fuel Composition" section, it is stated that the "fuel coreis a slug type design." Provide clarification of the term "slug type" or use more descriptive language to describe the fuel core design.4.2Section 4.2.1.2, p. 4-3. Provide sufficient overall fuel element dimensions forcomparison with the unit cell dimensions provided in this section and TS 5.3.4.3Section 4.2.1.3, p. 4-5. Provide clarification that the fabrication of NBSR fuel elementsis consistent with ANS 15.2.4.4Section 4.2.1.4, p. 4-6. The second paragraph states "Flow rates of 30 ft/sec which areover two times those seen in operation, (9.1 m/sec) were employed to measure flow conditions in each channel-" Provide clarification of whether the 9.1 m/sec is the flow rate seen in operation or the test flow rate. Also, provide discussion that justifies the use of test flow rates that are over two times the operational flow rates for both the inner and outer plenums.

4.5Section 4.2.2.1, pp. 4-9,10. The description of the "operational travel of 41º and amaximum travel of 50º" appears inconsistent with the statement "To prevent over travel during normal operation of the shim arm, installed upper and lower limit switches are set to approximately 41º and 2º, respectively." Clarify the operational shim arm travel ranges, limits, and corresponding angular positions.

24.6Section 4.2.2.2, p. 4-11. This section states that the regulating rod is "2 1/2 inches indiameter" and the last SER (NUREG-1007) says the regulating rod is "a 2.25 inch diameter solid aluminum rod." Clarify if the regulating rod design has been changed and describe any impact on the safety analysis.4.7Section 4.2.2.2, p. 4-11. This section states "The regulating rod acts as a poisondesigned with a reactivity worth approximately 0.58 ." The reactivity worth isinconsistent with the 0.58% stated elsewhere. Confirm the magnitude of this valueand clarify if the reactivity worth is derived primarily from absorption (poison) or moderator displacement.4.8Section 4.2.4, p. 4-16. This section states that the source is placed into one of theexisting experimental thimbles and does not contact the coolant. In the following section, Core Support Structure (p. 4-17), it is stated that coolant passes up through the experimental thimbles. Clarify how the source does not contact the coolant and justify why no cooling is required. Describe the source encapsulation material of construction (MOC) and the design and testing requirements.4.9Section 4.2.5, p. 4-17. This section states that the experimental thimbles are held downby poison tubes from the top plug. Describe the design of the poison tubes, including materials of construction and any age-related issues. Describe any other purpose(s) of the poison tubes.

4.10Section 4.3.1, p. 4-18. The description of the reactor vessel design discussed the useof two stainless steel O-ring gaskets at the reactor vessel flange. Describe any periodic inspection, leak testing, and replacement requirements or justify why these are not necessary.4.11Section 4.3.1, pp. 4-19 & 4-20. This section discusses "grazing tubes" as a separatevessel attachment. Relate the "grazing tubes" in the nomenclature terms used in the experimental facility descriptions in Chapter 10, e.g., radial beam tubes, through tubes, etc. Ensure nomenclature is consistent.4.12Section 4.3.1, p. 4-20. The fourth paragraph states "Since the vessel is entirely closed,there is no credible mechanism of exerting such a tensile stress, or impact, on the beam tube tips during reactor operation." Describe how all credible mechanisms for stresses resulting from pressures or impacts on the outside (non reactor side) of the beam tubes have been eliminated. Justify that the change in material properties (reduced ductility and Charpy energy) due to irradiation from past and future operations (20 years) will not reduce the design margins of safety to unacceptable levels. Describe the effect of the change in material properties (reduced ductility and Charpy energy) on the reactor vessel design rating and relief valve set pressure.

4.13Section 4.3.1, p. 4-21. The third paragraph states "The shim safety-arm drive andshock absorbing systems are mounted on the biological shield so that only the extremely small reaction between the outer faces and the balls is transmitted to the vessel." Describe what is meant by the "outer faces and balls."

34.14Section 4.3. Describe any surveillance or inspection programs for the periodicassessment of corrosion or radiation damage or why it is not needed.4.15Section 4.4.2, p. 4-23. The second paragraph states "The results yield a fast neutronflux 2.8x10

-3 n/cm 2-sec and a gamma flux of 2x10

-7 mW/cm 2 at the outside face of thebiological shield." Describe how these results were calculated, and how the subsequent 25% concrete, 75% thermal shield neutron capture gamma fractions were determined.4.16Section 4.4.3, p. 4-24. The fourth paragraph states "The radiation near the top of thecenter plug constitutes no health risk since it is in the well in the top floor that is covered with a 6-inch (15.2 cm) steel plate. This plate, an integral part of the transfer system, is always in place when fuel elements are being moved. The plate over each pick-up tool is penetrated by openings up to 6 inches (15.2 cm) in diameter that normally are plugged." It appears the dose rate of 0.5 mrem/hr stated in this section applies to an inaccessible area. Clarify what the radiation field would be in the area above the top shield plug where personnel may be located during transfer operations.4.17Section 4.5.1.2.2, p. 4-29. The fourth paragraph states "This 'loss' of material was dealtwith by adding elemental Zr and Sn, and 138Ba, to mock up those fission products."Provide the justification for this substitution.

4.18Section 4.5.1.3.1, p. 4-29. The reactivity change, , is defined and the method forcalculating presented. Elsewhere in the chapter, the values of reactivity are presented

as k/k. Provide consistent terminology or additional definitions and methodology.4.19Section 4.5.1.3.2, p. 4-30. Explain how the reactivity change of 0.34 % from "su183"to "sucold" is consistent with the reactivity temperature coefficients, e.g., the calculated moderator reactivity temperature coefficient.4.20Section 4.5.1.5.1, p. 4-34. The second paragraph states "Multiplying the differentialshim bank reactivity worth by the speed of the shim arm drives, 0.0445 °/s, one obtains the reactivity insertion rate vs. position, shown in Fig. 4.5.19." This does not appear to be what is shown in Figure 4.5.19. Clarify the statement or modify the figure to be consistent with the statement.4.21Section 4.5.1.5.3, p. 4-34. The first paragraph states "its average reactivity insertionrate is 3.8 x 10

-4 /sec." Provide the maximum differential rod worth and insertion rate, and provide a comparison with the TS 3.4 limit.4.22Section 4.5.1.6.1, p. 4-35. The second paragraph states "The fuel mass in F-5 is just138 g, so the normalized worth is 7.6 % /kg." In Figure 4.5.2A, p. 4-86, the F-5 massis given as 125g. Clarify the apparent difference.4.23Section 4.5.1.7, p. 4-36. The second paragraph states "There are only three means ofadding positive reactivity to the reactor while it is critical: (1) withdrawing the shim safety arms, (2) lowering the inlet D 2O temperature, and (3) rapidly removing experiments." Justify not including the regulating rod in this list.

44.24Section 4.5.3.2, p. 4-47. The 0.2 % limit for the pneumatic irradiation system and the1.3 % limit for movable experiments are not included in the criteria section of TS3.12. Provide justification for why these limitations are not criteria in TS 3.12 or modify the criteria accordingly.4.25Table 4.2.3, p. 4-61. Provide operating conditions and calculations for the 3.66 m/secchannel flow velocity under the "NBSR" column in the table.4.26Table 4.2.3, p. 4-61. The units for Max. Heat Flux in the first column appearinconsistent with standard heat flux units, e.g., "BTU/hr-ft 2 (W/m 2)." Also, the max. heatflux given for NBSR as 1.54 x 10 5 W/m 2 appears inconsistent with the hot spot heat fluxgiven on p. 4-54 for element H-1 and the conversion between heat flux units appears incorrect. Clarify or correct the differences, as appropriate.4.27Section 4.2.1.4, p. 4-6. This section indicates that the bypass flow was measured atsubstantially higher flow rates than the flow rates typically found during normal operation. As the dimensions of the gap for the bypass flow result from hydraulic drag, justify that the measured bypass flow rate is correct for normal operating conditions.4.28Section 4.2.5, p. 4-17. Provide clarification regarding the potential for the poison tubesto buckle due to upward coolant forces on the experimental thimbles. If buckling of the poison tubes is credible, provide analysis that shows it could not cause an accident notbounded by the maximum hypothetical accident.4.29Section 4.5.2.1.1, p. 4-37. The delayed neutron fraction is presented for steady reactorpower conditions. Describe and quantify any variation that may occur in this parameter during transient conditions.4.30Section 4.2.2.2. The regulating rod withdrawal rate has been changed since NBSR-9from 30" per minute to 120" per minute. The design of the regulating rod has also been changed. Describe how these changes affect the reactivity insertion rate of the regulating rod. Provide the evaluation that was performed to determine that the change did not impose any unreviewed safety questions.4.31Section 4.5.2.3.3. The analyses use 30 fuel elements instead of 24 fuel elementsallowed by TS 3.3 when the corner positions of the hexagonal lower grid plate are filled with plugs. Provide analyses to show that the use of 30 fuel elements in these analyses represent the limiting case. Explain how the hot channel factors account for the uncertainties in instrumentation and fuel fabrication tolerances. Describe how the uncertainties are treated (statistical vs. deterministic).4.32Section 4.6.3. Justify the assumption that the coolant within a single channel mixescompletely. Justify the assumption that the coolant mixes completely in the unfueled gap between the upper and lower core. Justify treating the uncertainties in a statistical manner. Describe the conservatism built into the correlations for DNB and OFI and quantitatively estimate the conservatism provided by these correlations for the NBSR analyses.____________________________________________________________________________

55.1Section 5.2.10, p. 5-14. In the first paragraph, the SAR states "The upper section of thethermal shield has 2-inch (5 cm) thick lead and 6-inch (15 cm) thick steel. The lead thickness was chosen to minimize the gamma ray flux at the vessel wall." As indicated in the biological shield description in Chapter 4, the lead and steel reduce the gamma ray flux in the concrete to minimize heating that may lead to cracking. Clarify the purpose of the lead shielding.5.2Section 5.2.14.1, p. 5-17. The SAR states "Maintaining the integrity of the fuel claddingrequires that it should remain below its melting temperature." The limiting criteria appears to be "blistering" temperature, as is stated in the next sentence. Provide clarification on the use of "melting temperature" vice "blistering."5.3Section 5.2.14.2, p. 5-17. This section states that if "all three parameterssimultaneously reach their safety-system settings, the burnout ratio is at least 1.3."

Provide reference to where in the SAR or elsewhere this analysis is performed or provide an analysis that demonstrates a bunrout ratio of 1.3 given those conditions.5.4Section 5.2.14.3, p. 5-18. The second paragraph states "Under this condition, the hotspot of the hottest plate remains below 160 ºF (70 ºC) (Chapter 13, Accident Analyses)."

Provide reference to where in the SAR or elsewhere the corresponding analysis and results are presented supporting this temperature and explain if this temperature is consistent with values in Table 5-5, p. 5-18 of Chp. 13.5.5Section 5.2.14.3, p. 5-18. The second paragraph states "Further, analyzing the case ofno-shutdown cooling flow (Chapter 13, Accident Analyses), the maximum temperature of the fuel plate would be less than 500 ºF (260 ºC), well below the temperature that would cause any damage." Provide reference to where in the SAR or elsewhere the corresponding analysis and results are presented supporting this temperature. Explain if this temperature is consistent with values in Table 5-10, p. 5-23 of Chp. 13, and with the temperature cited in TS 3.2 as 107 ºC (225 ºF).5.6Section 5.3.2.1.2, p. 5-21. This paragraph states "At flows of 65 gpm (250 lpm) on theprimary side-," while Section 5.4.2.3, p. 5-35, states "At flows of 35 gpm (132 lpm) on the primary side-." Both are apparently referring to the D 2O Purification HeatExchanger (HE-2). Clarify the difference between these flow rates.5.7Section 5.3.2.5, p. 5-24. The first paragraph states "The 150 psi (1 MPa) air to operatethe pneumatic control valves-." Similar wording appears in Section 5.4.2.6, p. 5-36.

Chapter 9, p. 9-12, states "The NBSR is supplied with a source of 100 psig (680 kPa) air from the main NIST compressed air facility." Clarify the difference between these air

pressures.5.8Section 5.3.8.1, p. 5-32. This paragraph states "Using this value, the limits ensure thattritium concentrations in effluents will be as low as practicable, and below concentrations allowed by 10 CFR 20.303 for liquid effluents and 10 CFR 20.106 for gaseous effluents (Chapter 11, Radiation Protection and Waste Management)." Explain the applicability of references to 10 CFR 20.303 and 10 CFR 20.106 in both the SAR and the TS, or update these references to current regulatory requirements, as applicable.

65.9Section 5.3.8.2, p. 5-33. The 2 nd and 3 rd paragraphs mention a "36 gallon/day" valueregarding primary to secondary leakage. The TS uses 40 gpd for minimum sensitivity in surveillance TS 4.5. Clarify the difference between the leakage rate sensitivity values.5.10Section 5.4.2, p. 5-34. In the 3rd paragraph, the last sentence states "Consequently,the minimum time to treat all of the primary coolant is approximately 21 1/2 hours."

Provide analysis to support the treatment time.5.11Sections 5.7.2.1 & 5.7.2.2, p. 5-42. The heat load is specified as "1.54 x 10 5 Btu/hr" andthe heat sink is specified as "60 x 10 3 Btu/hr." Explain how these two values relate toone another.5.12Section 5.2.2.6.2, p. 5-8. The temperature ranges for TR-2, TR-3, TR-4 and TR-5 haveinconsistent temperature ranges listed as the values for Fahrenheit and Celsius.

Provide clarification as to which are the correct values and the appropriate temperature range conversions.5.13Section 5.2.2.7.1, p. 5-10. Provide clarification describing methods used to preclude theintroduction of objects into the primary coolant system during maintenance associated with removal of the strainer.5.14Section 5.3.2.5, p. 5-24. Provide clarification on the response of the pneumaticallypositioned secondary valves to a loss of instrument air.

____________________________________________________________________________6.1Section 6.1.1, p. 6-1. The first paragraph states "a minimum of 28 minutes of coolantflow is always available to the core from the Inner Reserve Tank-." In Chapter 13, Appendix A, p. 5-8, the last paragraph states "For at least 20 minutes after shutdown the tank flow is more than adequate to cool the fuel elements by boiloff." Clarify these statements regarding the amount of cooling time that would be provided by the IRT.6.2Section 6.2.1.2.1, p. 6-9. The last sentence in the second paragraph states "The watermakeup capacity must be in excess of 25 gpm (95 lpm), which was calculated as adequate to prevent fuel damage." Provide an analysis and discussion of how this value was determined and compare with the flow from the D 2O Storage Tank and theEmergency Sump Pump during a loss of coolant accident.6.3Section 6.2.3, p. 6-13. Explain why the flowrates on Figure 6.4 are different from thoseon Figure 6.5 and the description on pp. 6-13 & 6-14.6.4Section 6.2.3.3.4, p. 6-18. The second paragraph states "The height of approximately100 feet (30 meters) above grade level was chosen to meet the criteria of dilution and reduced potential exposure." Describe how the stack height compares to the guidance in Regulatory Guide 1.111 and GEP stack height criteria for elevated releases. If corrections are required also apply the corrections to all affected analyses.

6.5Section 6.2.3.3.5, p. 6-19. The third and fourth paragraphs state that the EmergencyExhaust Fan motors (AC and DC) for EF-5 & EF-6 are powered from MCC DC. It 7appears from Chapter 8 that the power source for the AC motors is the A5 emergencybus. Explain and differentiate the power source and switchgear locations for these motors.____________________________________________________________________________7.1Section 7.2.1, p. 7-5. Explain why primary coolant temperature is absent from the list ofmain parameters which are monitored and provide inputs to the logic chains.7.2Section 7.2.3, p. 7-10. Provide a schematic of the control logic for confinement buildingisolation, i.e., door scram relays, fan scram relays, ventilation system alignment, etc.7.3Section 7.3.1.2, p. 7-16. Provide an explanation of the "all rods seated" contacts andthe purpose of this interlock.

____________________________________________________________________________9.1During the orientation tour, it was noted that neutron shielding for the cold neutronsource and neutron guides consists of lead shot mixed in paraffin. The quantity of shielding material was significant. The paraffin is both a large transient combustible load, but also can melt and pool resulting in more dangerous fires. The SAR does not mention the paraffin as a flammable material that is present even though it is most likely the largest single combustible source in the confinement building. Provide a description of the paraffin in the shielding blocks and the design features that prevent or mitigate its involvement in a fire.9.2Section 9.2.4.1, p. 9-7. Provide justification for the extrapolation used to determine theminimum time a fuel element must remain submerged in the primary coolant prior to transfer. Include discussion/analyses of power distribution for both the 10 MW core and the 20 MW core, decay heat for worst-case fission density and irradiation time for fuel elements in both the 10 MW core and the 20 MW core, and any assumptions made and all uncertainties (measurement, instrumentation, fabrication, etc.) in all relevant analyses. The discussion/analyses should clearly show that nowhere will the local clad temperature of a worst-case-irradiated fuel element immersed in helium reach 450 C. ____________________________________________________________________________10.1Section 10.3 of the SAR references TS 6.2(2) and 6.2(3) regarding the requirement ofthe SEC to review experimental proposals. Verify that these are the correct references and change the references if appropriate.

____________________________________________________________________________11.1Section 11.1.1.2, p. 11-3. The dose limit to members of the general public due toairborne effluents is 10 mrem/yr (10 CFR 20.1101(d)). Revise this section to reflect the appropriate dose limit.11.2Section 11.1.1.4.2, p. 11-9. Provide more detail in this section to clarify actions relatedto the disposal of the shim arms. Briefly describe the processes used to remove the shim arms from the reactor vessel (mechanical detachment and physical transfer),

including discussions of ALARA practices, and the location where the reactor shims decay for three months.

811.3Section 11.1.1.4.4, p. 11-10. Provide more detail as to the type of "materials designatedas radioactive waste" that are transferred to H wing. Describe what methods are used to control access to the H wing, or justify not requiring access control.

____________________________________________________________________________13.1Section 13.2.1, p. 13-5. Provide a discussion/analysis of potential metal-water reactionsand associated potential consequences.13.2Section 13.2.1, p. 13-6. The 1 st paragraph states "The inventory of noble gases andiodine fission products in the most heavily irradiated element is given below in Table 13.1, as determined by the computer code ORIGEN2 (Croff, 1980)." Describe or reference the assumptions on irradiation times, power levels, peaking factors, etc. to verify that this element has the maximum iodine and noble gas concentration.13.3Section 13.2.1, p. 13-6. The section states "...consideration of these effects leads to theconclusion that less than 3% of the total iodine release will be present as I 2." Providethe analyses on which this conclusion is based. Include your analyses related to the effects of temperature, pH and the presence of other fission products and chemical forms on iodine release fractions. Evaluate the effect of differences in fuel material design and configuration. Specifically, the type of fuel used at NIST (U 3 O 8) is differentthan the type of fuel for the NUREG 1465 analysis (UO 2), on which it is understood the3% is partially based. Consider reviews such as presented in "The Technology of Nuclear Reactor Safety," Volume 2, Copyright 1973 by the Massachusetts Institute of Technology, Chapter 3, "Fission Product Release" by G. W. Parker and C. J. Barton of ORNL, Section 3.3.2, "Uranium Oxide, U 3 O 8." Also, since some of isotopes haverelatively short half-lives relative to the accident duration, the daughter products may be released from solution. Describe how these parent and daughter products are accounted for in the source term and dose estimates. Provide a description of how the iodine daughter products were considered.13.4Section 13.2.2.2.2, p. 13-9 & the new calculation provided via email [Mendonca9/29/2006] following the site orientation visit. The new calculation is for a ramp insertion of 0.5% in 0.5s, whereas the previous accident scenario is for a ramp insertion of1.3% in 0.5s. SAR section 13.1.2.2.2 provides technical justification for the changein the accident scenario from the existing SAR, however this is not consistent with at least one of the bases in the TS. Specifically, the basis for TS 3.12 refers to the 1.3%

insertion transient. Correct this reference and verify that any other renewalsubmissions are consistent with the revised analyses.13.5Section 13.2.3, p. 13-10. Under the assumptions for this accident, it states "The tritiumconcentration in the primary coolant is at the maximum level permitted by the TS (5,000

µCi/ml)." The statement regarding the estimated concentration in the Basis of TS 3.6 is not a TS limit. Provide a description of how and where this limit is protected in the TS.

If there is no limitation established on this parameter in the TS, provide such.13.6Table 13.1, p. 13-16. Several of the isotopes in the fission product inventory are not inthe HOTSPOT library. Provide a description of how these were modeled in the offsite dose projections.

913.7Tables 13.3 & 13.4, p. 13-17. Provide the assumptions regarding iodine removal ratesin confinement from deposition and filtration for public and staff dose estimates. What DCFs were used for submersion, inhalation, and thyroid doses for staff doses presented in Table 13.4?13.8Table 13.2, p. 13-16. The values for removal rates from C-200 are not consistent.Determine the appropriate values and ensure that they are correct in both sets of units.13.9For each accident analysis, provide the limiting assumptions, conditions and safetysystem settings and where these limiting assumptions, conditions and safety system settings are required by Technical Specifications as required by 10CFR50.36. Compare the assumptions, conditions and safety system settings to those in ANSI 15.1 and NUREG 1537, which are applicable to test reactors.

____________________________________________________________________________Appendix A Technical Questions and Comments 13.16Section 2.2, p. 2-4. The 1st paragraph states "About 4% of the total flow in eachplenum bypasses the fuel elements and cools the in-core thimbles." Chapter 4 (SAR),

p. 4-4, states "A small amount of coolant, 4%, bypasses the external surface of the lower nozzle-preventing bulk stagnation in the moderator." In Chp. 4 (SAR), p. 4-12, the description of the regulating rod states "A fixed orifice in the nozzle of the shroud delivers a coolant water flow of 8 gpm from the outer plenum." In Chp. 4 (SAR), p. 4-50, the description of the core flow distribution states "Approximately 4% of the flow bypasses the core; this is treated conservatively in the next sections [T-H Analysis] by reducing both flows to 95% when calculating the flow through any element." In the "Core Bypass Flow" section of Appendix A, p. 4-5, the RELAP model description states "About 4% of the total primary flow bypasses the fuel elements. In RELAP5 the areas of the bypass flow junctions have been adjusted so that 4% of flow to the inner and outer plenums is bypassed." In Chp. 10, Section 10.2.6.1, p. 10-6, the description of the seven 3 1/2 in. thimbles states "The end fitting largely blocks the normal flow, but contains a small opening that allows approximately 8 gpm (0.5 liter/sec) to flow upwards through the tube to cool it, and any experiment that may be in it." In Chapter 10, Section 10.2.6.2, the description of the 2 1/2 in. thimbles states "These smaller sockets have a small hole at the bottom that allows approximately 10 gpm (0.6 liter/sec) of plenum cooling water to flow up through the experimental thimble." There appear to be some discrepancies in the above statements regarding bypass flow. Some specific considerations are:a.The 4% bypass flow is not predominately for in-core thimble cooling, since thesehave individual orifices for coolant flow.b.Chp. 4 (SAR) indicates that fuel element flow is treated as 95% full flow whileAppendix A indicates the RELAP model uses 96%.c.If six of seven 3 1/2 in. thimbles at 8 gpm and four 2 1/2 in. thimbles at 10 gpm arefed separately from the outer plenum, then this accounts for approximately 88/6400 = 1.4% of outer plenum flow not accounted for in the RELAP model.

10Provide clarification of the following comments (13.17-13.25).13.17Figure 3-5, p. 3-12. The 235U content in this figure differs from that in Figure 4.5.2A, p.4-86, in Chapter 4 of the SAR. Are these "BNL" versus "updated model" differences?13.18Figures 3-13 through 3-18, pp. 3-16 to 3-18. The orientation of the plates in thesefigures is north-south which differs from the east-west orientation in Figure 4.5.4 through 4.5.9, pp. 4-88 to 4-90. Is the orientation different in the two MCNP models? If so, provide clarification of the effect this has on the peaking factors.13.19Figure 3-28, p. 3-23. Provide analyses which demonstrate that the regulating rodmaximum reactivity differential worth and withdrawal rates will not exceed the startup accident maximum reactivity insertion rate. Alternatively, propose limits on regulating rod reactivity insertion rates to limit them to the same rate as specified for the shim rods.

Additionally, provide justification as to why the regulating rod worth should not be considered in conjunction with shim arm worth in the startup accident.13.20Figure 3-30, p. 3-24. The caption for the figure includes the description "EquilibriumCore at Startup" and the title includes the description "SU Core." In previous nomenclature, the "SU Core" is defined as the startup core prior to equilibrium fission product poison concentrations, and the "BOC Core" as the startup core with equilibrium fission product poison concentrations. For which core was this figure developed?

Provide consistent references in the renewal application documents.13.21Table 3-2, p.3-28. As previously mentioned, the description for the voided thimblesindicates 5 thimbles voided whereas p. 3-6, App. A and Chapter 4 (SAR), p. 4-39 indicates 6 thimbles voided. The values of k/k appear to be calculated as (kvoid - k base case)/ k base case instead of (kvoid - k base case)/ kvoid. What thimble volume was used for thevoid coefficients calculated for the voided thimbles case? These numbers appear to be inconsistent with those in Table 4.5.7, p. 4-67 of Chapter 4 of the SAR. What case or analysis supports the statement in Section 4.5.2.2.2, p. 4-39 of Chapter 4 of the SAR that "Finally, from the BNL analysis, if somehow only the unfueled regions between the upper and lower fuel sections were to be voided, the coefficient would be -0.025%

/l-"?13.22Tables 3-3 & 3-4, p.3-28. The values of k/k appear to be calculated as (kflooded - k base case)/ k base case instead of (kflooded - k base case)/ kflooded. Provide clarification as to which is thecorrect method for determining the values of k/k.13.23Section 4.2.2.4, p. 4-3. The fuel plate width is given as 2.3734 in. in this section, and2.436 in. on p. 4-3 of Chapter 4 of the SAR. The 2.436 in. appears consistent with the peak heat flux given in Chapter 4 on p. 4-54, element H-1.13.24Section 5.3, p. 5-3. This section states "The minimum CHFR is 1.28 and 1.18 for BOCand EOC, respectively." These values are both below the 99.9% limit values determined for CHFR on p. 4-10 of Appendix A. Provide justification to demonstrate that these provide acceptable margins.

1113.25Table 5-13, p. 5-26. This table presents CHFRs as determined by the Mirshak andCosta correlations for 500 kW operation under natural circulation. Provide justification that these correlations are applicable for natural circulation flow. Describe the flowvelocity ranges and conditions where the correlations are valid.

____________________________________________________________________________Editorial Questions and Comments 4.34Section 4.2.1.1, Fuel Composition, p. 4-3. It is stated that the aluminum powder used isATA 101 (or equivalent). Clarify the "ATA" abbreviation and add to the "Acronyms" list.4.35Section 4.2.1.2, Fuel Element Description, p. 4-4. It is stated that "the fuel plate coreframes and cladding are aluminum Alloy 6061-T0 (ASMT B209)." This is inconsistent with Table 4.2.2, which has "aluminum clad" as 6061-T6.4.36Section 4.2.1.2, Fuel Element Description, p. 4-3. It is stated that "fuel is contained infuel plates approximately 13 inches in length by 2.793 inches in width-." The width dimension is inconsistent with that in Table 4.2.3, p. 4-61 (2.415 in).4.37Section 4.2.1.2, Fuel Element Description, p. 4-4. The first line states "curvature is 5 .5inches (13.97 cm)." There is an extra space in "5 .5 inches."4.38Section 4.2.1.3, Fabrication, p. 4-5. It is stated that "Dents greater than 0.250 inch(0.06 cm) in diameter-" These dimensions are inconsistent.4.39Section 4.2.2.1, Shim Safety Arm, p. 4-9. It is stated that "Helium at just slightly aboveatmospheric pressure (15 psig) is left in the void." Is the pressure approximately twice atmospheric pressure, or 15 psia?4.40Section 4.2.2.6, Technical Specifications, p. 4-14. TS 4.3, item no. 5, states "acomparison of power range indication with flow time's delta T-." The apostrophe in "time's" appears unnecessary.4.41Section 4.2.2.6, Technical Specifications, p. 4-15. TS 4.3, the basis section states "Theshim arms shall be considered operable if they drop the top five (5º) within 220 msec."

The "top five (5º)" apparently should read "top five degrees (5º)."4.42Section 4.2.5, p. 4-16. This is a general comment about units formatting, but it occurshere because this section switches from using English units with SI units in parentheses previously, and in this section that convention is intermittently swapped. ANS-15.21-1996 states "SI units shall be used, with English units posted in parentheses, except where the regulations require a different presentation."4.43Section 4.3.1, Design, p. 4-20. The third paragraph has "2x10 23 n-cm-2-s-1." From thecontext, it appears the units should be "2x10 23 n-cm-2."4.44Section 4.4, p. 4-22. In the first paragraph of this section, the sentence "Chapter 10 ofNBSR-9 (NBS, 1966a) contains a thorough description the design considerations and 12shielding calculations for the construction of the biological shield," is apparently missingan "of" in the phrase "description of the design-."4.45Section 4.5.1.3.4, Fission Product Poisons and the Equilibrium Core, p. 4-31. Thesecond paragraph states "The reactivity difference between the SU benchmark, "su183,"

and the BOC equilibrium core, "eqlib," is keff = 0.97911, and = - 2.86 %k/k, or-$3.78." Clarify the reactivity units.4.46Section 4.5.1.5.1, The Shim Safety Arms, p. 4-33. The first paragraph states "After theinitial shim arm movement, there is a gradual withdrawal until the shim safety arms are above the core and larger withdrawal steps are needed to achieve the same negative reactivity insertion." In this context, it would appear the word "negative" should be "positive."4.47Section 4.5.1.5.1, The Shim Safety Arms, p. 4-34. The second paragraph states "Themaximum calculated rate is 4.5x10

-4 (% k/k)/s. The technical specifications limit therate to 5.0x10

-4 (% k/k)/s." The Technical specifications use the reactivity units /s. Clarify the difference between these units and those used in the TS.4.48Section 4.5.2.3.3, Hot Channels and Hot Spots from the Updated MCNP Model, p. 4-42.The last paragraph states "The rate of consumption of 235U is 1.17 times the fission rate,or 7.1 x 10 18 fis/cm 3/day." Clarify if this value and the appropriate units represent theaverage fission rate (fis/cm 3/day) or absorption rate (abs/cm 3/day).449Section 4.5.3.1.2, Moderator Dump, p. 4-46. In the first paragraph under "Basis" thephrase "with one shim arm know to be inoperable," is apparently missing an "n" in "known" as in "with one shim arm known to be inoperable."4.50Section 4.5.3.1.2, Moderator Dump, p. 4-46. In the second paragraph under "Basis" thesentence beginning "The analysis showed that the most sever accident-," is apparently missing an "e" in "severe" as in "The analysis showed that the most severe accident-."4.51Section 4.5.3.3, Safety Limits and Limiting Safety System Settings, p. 4-49. The "Basis"section for TS 2.2 uses the term "burnout ratio" whereas the term "Critical Heat Flux Ratio" is used on the previous page under section 4.5.3.3.1. When practical, use consistent terminology between the SAR and TS.4.52Section 4.6.1.2, Power Distribution in the Core, p. 4-51. In this section, the terms"horizontal strips" and "vertical strips" are used. Clarify the use of these terms as compared to the terms "slices" and "stripes" defined previously.4.53Section 4.6.2.2 & 4.6.2.3, Departure from Nucleate Boiling & Onset of Flow Instability, p.4-53. The definition of the term "T s" (both sections) is given as "saturation pressure." Itwould appear from the context this term should be "saturation temperature."4.54Section 4.6.3, Determination of Limiting Conditions, p. 4-54. The pressure at the hotspot is estimated as "3.34m D 2O, or 138.5 kPa, or 1.37 bar." The conversion from kPato bar is 1 bar = 100 kPa, so these numbers appear inconsistent.

134.55Section 4.7, References, p. 4-58. Correct the date in the reference for "NIST Center forNeutron Research (s004b)."4.56Table 4.5.5, p. 4-67. In the second column, i, the values appear to be in percentageunits, i.e. i (%).4.57Table 4.6.1, p. 4-72. Check the grammar in the statement "These are the minimumflows to assure that there be no nucleate boiling at any point in the core."

____________________________________________________________________________5.15Section 5.2.4.3, p. 5-11. In the 2nd paragraph, the phrase "and a reactor scram occurdue to-," is apparently missing an "s" at the end of "occur."5.16Section 5.2.14.3, p.5-18. The third paragraph states "Calculations show that tritiumreleases offsite are below concentrations allowed by 10 CFR 20 (Chapter 11, Radiation Protection and Waste Management)." TS 3.2 references Chapter 13 for these calculations. Clarify the difference between the locations of the supporting calculations.5.17Section 5.3.2, p. 5-20. In the 3rd paragraph, the word "Deminerizer" appears incorrect.

5.18Section 5.3.2.8, p. 5-29. In the 2nd paragraph, in the phrase "on room D100" it appearsthe word "on" should be "in."5.19Section 5.3.8.2, p. 5-32. This paragraph states "It also requires that, when the N-16monitor is inoperable, the secondary cooling water is sampled and analyzed for tritium at least monthly." The word "inoperable" should apparently be "operable" to agree with the TS.5.20Section 5.4.2.5, p. 5-36. In the 1st paragraph, should "cellulose, acetate cartridges" behyphenated as in "cellulose-acetate cartridges"?5.21Section 5.7.2.1, p. 5-42 & Section 5.7.2.6.1, p. 5-43. Two uses of nomenclature appearinconsistent with the "Cold Neutron Source" terminology used elsewhere.5.22Section 5.7.2.6.2, p. 5-43. In the 1st paragraph, the phrase "thermowell located the 11/2-inch (3.8 cm) piping" appears to be missing an "in."

____________________________________________________________________________6.7Section 6.1.1, p. 6-1. The first sentence appears to contain a typo in "Figures 6.1."

6.8Section 6.1.1, p. 6-2. The third paragraph apparently contains a typo in the phrase "thistank will start draining though the two nozzles."6.9Section 6.1.2, p. 6-2. The first paragraph apparently contains a typo in the phrase"power distribution gears."

146.10Section 6.2.3.2.2, p. 6-17. The first sentence in the second paragraph lists "filters F-26,F-27, F-59 in subsystem A." Figure 6.4 shows F-26, F-27, and F-57. Clarify the apparent mismatch.6.11Section 6.2.3.2.2, p. 6-17. The sentence "Since one of the two trains is in operationalduring an emergency...," apparently contains a typo.6.12Section 6.2.3.3.4, p. 6-18. The first paragraph states "discharge from ReactorBasement Exhaust System fan EF-27 through ACF-3." The "ACF-3" is apparently a typo for "ACV-3."6.13Section 6.2.3.4.4, p. 6-21. The last sentence uses the acronym "WSSC." Spell out theabbreviation on first use and add to the "acronyms" list.

____________________________________________________________________________7.4Section 7.2.3, p. 7-9. In the 4 th paragraph, the 1 st sentence refers to Figure 7.7 and the"relay logic ladder." It appears that this paragraph is referring to the logic diagram in Figure 7.8. If this is true, check and correct the subsequent references to Figure 7.7 in this chapter, as appropriate.7.5Section 7.3.3.1, p. 7-19, Item 6. (2) and the definition of Reactor Shutdown in the TSare not the same. Clarify the difference between the wording in the two locations.7.6Section 7.3.3.1, item 8, top of page 20. TS definition 1.3 includes an item (4)"Moderator Dump." Clarify the diference between the wording in the two locations.7.7Section 7.3.3.2, p. 7-20. In the 1 st paragraph, clarify that the 3 rd item is intended to beoperable 'in accordance with' Table 3.1 of the TS.7.8Section 7.3.3.2, p. 7-20. In the 3 rd paragraph, check and correct the wording andgrammar in the 1 st sentence "A rod withdrawal accident for the NBSR has beenanalyzed and are discussed Chapter 13 and Appendix A of this SAR-."7.9Section 7.4.1, p. 7-23. In the 2 nd paragraph, check and correct the wording andcapitalization in the sentence "A minimum of one decade of overlap is designed into the transition between the Source Range and Intermediate Range Nuclear Instrumentation and between Intermediated Range and Power range Nuclear Instrumentation." In the following sentence, check and correct the use of the word "form" in "channels form the source range."7.10Section 7.4.1, p. 7-24. In the 1 st sentence on p. 7-24, check and correct the usage of "from" and "the" in the second line, "power from directly from the the +/- 10Vdc."7.11Section 7.6.1, p. 7-26. Check and correct the word "inn" in the sentence beginning,"The instrument panels inn the control room display-."7.12Section 7.6.3, p. 7-27. In the 5 th paragraph, the last sentence appears to be missing a"the" before "reactor operator."

157.13Sections 7.7.1 through 7.7.5, pp. 7-30 & 7-31. There are multiple references toAppendix 8 (8A, 8H, 8I, 8J, 8E, 8G, 8F). Explain or correct the use of these reference numbers.7.14Section 7.7.3, p. 7-30. In the last line, check "AN47" for correctness.

7.15Section 7.8, p. 7-32. Explain the use and applicability of the ANSI/ANS 15.20 standardfor the NBSR I&C system design.7.16Section 7.8, p. 7-32. The IEEE Standard 7-4.3.2 title appears to contain an extra"Systems" after "Computers."7.17Table 7.5B, p. 7-35. Check and correct the 1 st column heading in the table.7.18Table 7.5G, p. 7-41. Check and correct the range on the D 2O IX Inlet/OutletConductivity Recorder.7.19Table 7.5G, p. 7-41. It appears there are several instances of "HE" that should be "He",i.e., to represent helium instead of heat exchanger.7.20Table 7.5I, p. 7-43. The 1 st column lists "Storage Pool IX Inlet/Outlet Conductivity" and"Thermal Shield Inlet/Outlet Conductivity."

It appears that there are only "Outlet" instruments.7.21Table 7.7B, p. 7-46. The nomenclature of the 1 st column header "Cubicle" appears to be incorrect.7.22Figures 7.4B & 7.4C, p. 7-54 & 7-55. The figure titles appear to be backwards for thesetwo figures, i.e. "Intermediate Range Channel" should go with Figure 7.4c and "Power Range Channel" with Figure 7.4b.7.23Appendix 7A, Section 5, p. 7-75. In the second paragraph, the source range channelsin the last line are referred to as "ND-1 and ND-2." These appear to be typos for "NC-1 and NC-2."7.24Appendix 7A, Section 5, p. 7-77. In the 3 rd paragraph, the 1 st sentence appears to bemissing an "of" after "rate of change."7.25Appendix 7A, Section 7, p. 7-83. In number 13, check and correct the units "pisg" in thelast sentence.7.26Appendix 7A, Section 8, p. 7-88. In number 11, the last sentence is apparently missinga "than" after "rather."

____________________________________________________________________________9.4The first sentence of Section 9.3.2 is: "The NBSR is equipped with both automatic andmanual fire detection capability." This sentence is only true when people are in the building, or more specifically in the areas where fire may occur.

16 9.5Section 9.9.4, p. 9-15. The last sentence of this section should say "Principal metalcomponents-" rather than "Principle metal components-."

____________________________________________________________________________13.10Section 13.1.4, p. 13-3. This section states "Five different scenarios for loss of primarycoolant flow have been analyzed," and in Section 13.2.4, p. 13-11, it states "Four scenarios have been given for an accident of this type [Loss of Primary Coolant Flow]-." Clarify the apparent discrepancy.13.11Section 13.2.1, p. 13-7. In the second paragraph, the phrase "for estimation of long-term (>1 day)" seems to be related to dose. Should it be "for estimation of long-term doses (>1 day)"?13.12Section 13.2.2.2.1, p. 13-8. In the first paragraph, the reactivity insertion rate, "5x10

-4k/s" appears to be inconsistent. Should the units be "5x10

-4 /s"?

13.13Section 13.2.3, p. 13-10. The 1st paragraph states "Thus, with only one operator action(which can be accomplished at any time in the first 20 minutes), the core is fully protected for several hours." In Chapter 6, p. 6-2, the time the IRT and D 2O EmergencyCooling Tank provide cooling is 2 1/2 hours. The term "several" used in the statement from section 13.2.3 appears to be subjective.

13.14Section 13.2.3, p. 13-11. The last paragraph states "For the conditions analyzed, thiswill result in a concentration approaching 1.25x10

-4 DAC." Shouldn't this be 1.25x10 4 DAC?13.15Figures 13.2, 13.3 & 13.4, p. 13-21, 13-22. Provide clarification if these are plots ofMCHFR versus time, or CFHR versus time.Appendix A 13.26Section 2.1, p.2-1. The 2nd paragraph states "The fuel elements are located on 0.177m(7 in) centers in a hexagonal array." Chapter 4, p. 4-4 indicates 0.175m and p. 4-17 indicates a 17.6 cm pitch for exp. thimbles.

13.27Section 2.1, p. 2-2. Paragraphs 7 & 8 (next to last & last) indicate reactivity worths of26%, 6 1/2%, and 0.6%. Should the units be % ?

13.28Section 2.1, p.2-3. The 2nd paragraph states "The uranium content is about 1 gm/cm 3." Data from Chapter 4, p. 4-3 indicates 1.23 gm/cm

3.

13.29Section 2.2, p. 2-3. The 1st paragraph indicates a nominal core flow of 9000 gpm.Chapter 4, p. 4-50, Table 4.1.1, p. 4-59, indicates 8700 gpm as nominal flow.

13.30Section 2.2, p. 2-4. The 1st paragraph indicates an outer plenum flow of 6700 gpm.Chapter 4, p. 4-50, Table 4.1.1, p. 4-59, indicates 6400 gpm as outer plenum flow.

1713.31Section 3.3, p. 3-4. The 1st paragraph states "Also included in this figure is the percentdecrease in the 235U content for each fuel element during a single 38-day cycle." Figure3-5, p. 3-12 shows "Decrease in 235U (grams)."

13.32Section 3.4.3, p. 3-5. The 2nd paragraph states "The D-4 element is separated from theshim arm by one row of elements-." Should the element described "D-1"?

13.33Section 3.5.2, p. 3-6. The 1st paragraph states "In the first case, the six vacantirradiation thimbles -are voided." In Table 3-2, p. 3-28, this case is described as "SU with 5 thimbles voided."

13.34Section 3.5.2, p. 3-6. The 1st paragraph states "The calculations were performed forthe SU and EOC cores for two different void cases." Table 3-2, p. 3-28 shows three

cases.

13.35Section 3.5.8, p. 3-8. The 1st paragraph states "In the present work, the maximumrelative power peaking was 1.16. In the updated model, the maximum value was 1.11."

In the SAR, Chapter 4, Figure 4.5.3, p. 4-87, the maximum peaking factor is 1.15 calculated with the updated model.

13.36Figures 3-29 & 3-32, p. 3-24 & 3-25. The y-axis labels appears to be missing the units

"(%)."13.37Figures 3-26 through 3-33, pp. 3-22 to 3-26. The y-axis labels are not discernable onprovided copy.13.38Section 4.2.3.8, p.4-6. In the 2nd paragraph, the sentence beginning states "A set ofpower factor is determined-." Should this be "A set of power factors is determined-."13.39Table 4-5, p.4-26. "Normal" primary flow in the table is given as 8800 gpm and 9000gpm in the footnote.13.40Section 5.2, p. 5-2. In the 1st paragraph, the shim arm withdrawal reactivity rate isgiven as "5 x 10

-4 k per second." Use consistent reactivity units.13.41Section 5.4, p. 5-3. In the 2nd paragraph it states "After a 0.4s delay a reactor scram isinitiated at 1.286 s." If the flow trip is initiated at 0.896 s, shouldn't the reactor scram be initiated at 1.296 s?13.42Tables 5-1 through 5-4, pp. 5-14 to 5-17. Shouldn't the column headings be CHFRinstead of MCHFR?

____________________________________________________________________________Technical Specifications Format and Content 18In Chapter 14 of the Safety Analysis Report (NBSR 14), page 14-1, the text states "TheTS have also been reformatted in accordance with the NRC-approved Standard, ANSI/ANS 15.1." The proposed Technical Specifications vary from the NUREG-1537 accepted consensus guidance of ANSI/ANS15.1. Chapter 14 and the TS should provide a one to one comparison to ANSI 15.1 and provide applicable technical specifications or justification for differences. Chapter 12 of the SAR should appropriately reflect changes to the TS. Provide justification for or modification to accommodate the following:14.1. Section 1.3, Definitions, does not contain the following definitions or the definitionsprovided differ from the guidance. Protective ActionConfinement and confinement integrity Experiments Secured experiments Moveable experiments Core excess reactivity Shutdown margin Reactivity worth Safety systems Scram time Measured value Reactor secured

Channel Channel Check Channel Test Reactor shutdown (include consideration of minimum number of control/shim rods and no work in progress)

Senior Reactor Operator Reactor Operator14.2. Each limiting condition of operation does not have a corresponding surveillancetechnical specification.14.3 Surveillance for fuel handling and storage is not provided.

14.4 Surveillance for experiments is not provided.

14.5 Provide an LCO and a surveillance specification for rod drop times.

14.6TS 6.1, Organization, is not consistent with organization chart, Figure 6.1.

14.7TS 6.1.3, Staffing, is not consistent with Section 6.1.3 of the guidance.

14.8TS 6.2, Review and Audit, is not consistent with Section 6.2 of the guidance.

14.9The TS do not have a section that corresponds to Section 6.3 of the guidance.

1914.10TS 6.3, Procedures, is not consistent with Section 6.4 of the guidance.14.11The TS do not have a section that corresponds to Section 6.5 of the guidance.

14.12TS 6.4, Required Actions, is not consistent with Section 6.6 of the guidance.

14.13TS 6.5, Reporting Requirements, is not consistent with Section 6.7 of the guidance orwith NUREG 1537.14.14.TS 6.6, Records, is not consistent with Section 6.8 of the guidance.

14.15Section 5, Design Features, has been reformatted to include Applicability, Objective,Specification, and Basis. Section 1.2.2 of the guidance states that the section should state the specification without the related information.Technical Questions and Comments 14.16The basis of TS 2.2, p. 4, states "Even in the extremely unlikely event that all threeparameters, reactor power, coolant flow, and outlet temperature simultaneously reach their Limiting Safety System Settings; the burnout ratio is at least 1.3." Provide an explanation of where these calculations and results are presented in the SAR.14.17The basis of TS 2.2, p. 4, states "Overall uncertainties in process instrumentation havebeen incorporated in the Limiting Safety System Setting." The statistical analysis in Appendix A of the SAR, Section 4.4, pp. 4-9 & 4-10, determines that to account for uncertainties in the hot channel variables, the 99.9% limit value for CHFR is 1.538.

Provide an explanation of how the statistical analysis includes the process instrumentation uncertainties mentioned above (temperature measurement is not explicitly listed in Table D-1, p. D-10 of Appendix D to NBSR Appendix A), and why the basis should not refer to an analysis in the SAR which demonstrates that at coincident LSSS values the CHFR exceeds 1.538.14.18TS 3.1, p. 8. Provide an explanation of the level of confinement integrity required duringoperations involving sawing the fuel elements and fuel transfer.14.19TS 3.1, p. 8 and TS 3.7, p. 14, contains time limits for the transfer of fuel. The time limitcalculations for fuel transfer were not found in the SAR. Provide the calculations for the fuel transfer times.14.20Provide the units for the activity level concentrations used in TS 3.11, p. 17.

14.21TS 2.2 provides for a rundown at a reactor outlet temperature of 147 ºF. The lastparameter in Table 5.3 of the SAR, p. 5-46, provides a rundown safety function at a setpoint of 130 ºF. Provide an explanation of which temperature is correct.14.22The basis of TS 3.2 cites a no shutdown cooling maximum fuel plate temperature of 107ºC (225 ºF). Section 5.2.14.3 of the SAR , p. 5-18, states "Further, analyzing the case of no-shutdown cooling flow (Chapter 13, Accident Analyses), the maximum 20temperature of the fuel plate would be less than 500 ºF (260 ºC), well below thetemperature that would cause any damage." These values appear inconsistent with Table 5-10, p. 5-23 of Chapter 13, Appendix A. Provide an explanation of which temperature is correct and where and how it was calculated.

14.23TS 2.1, Safety Limit Bases states: "Maintaining the integrity of the fuel cladding requiresthat the cladding remain below its blistering temperature 752 F (450 C )." Thetemperature quoted in F does not correlate to that for C. Provide correction. Ensurethat other conversions in the Technical Specifications are correct. Example: the basis for TS 3.1 states, "This provides a margin of safety from the lowest temperature at which blistering can occur 850 F (450 C)." Ensure the bases are consistent with the SAR.14.24TS 5.3 does not specify the aluminum alloys used for various components of the fuelelement. Provide justification for why this is acceptable.14.25NUREG 1537, Section 8.1, page 8-3, states "The technical specifications, includingtesting and surveillance provisions, ensure that the normal electrical system will be operable." Provide justification for why there is no TS for the normal electrical power system.14.26TS 4.1 requires routine verification of the confinement function of the confinementbuilding and the instrumentation that provides the confinement closure signal. Provide an explanation of why no in-service testing of the confinement building structure is specified in the SAR or TS other than the periodic confinement closure system testing and leakage testing.Editorial Questions and Comments 14.27Specification (1) of TS 2.1, p. 3, refers to Figures 2.1 and 2.2. These figures on p. 5 &6 are labeled "Table 2-1" and "Table 2-2."14.28The last sentence in the basis section of TS 2.1, p. 3, states "The analysis done in theSAR, NBSR 14, Appendix A, clearly show that the reactor can be operated at 500 kW with reduced or no flow." The tense of the sentence is not correct.14.29TS 2.1, Figures 2-1 & 2-2, pp. 5-6. When using a black and white print of the figures, itis not clear which curve represents which temperature.14.30TS 3.1, p. 8. In the 4th paragraph of the "Basis" section, it appears something ismissing in the sentence beginning "Experiments are usually-," and finishing, "in the reactor at any time."14.31TS 3.2 references Chapter 13 of the SAR for the allowable tritium releases offsite. Itappears that section 5.2.14.3 of the SAR, p. 5-18, should be the referenced.14.32TS 3.4, p. 11. In the 2nd paragraph of the "Basis" section, the word "sever" should be"severe."

2114.33TS 6.2.1 d), p. 32. In the last phrase "a change in Technical Specifications incorporatedin the facility license or an questions pursuant to 10 CFR 50.59," the "an" should be removed.14.34TS 6.3, p. 33. In the 1st paragraph, the phrase "may be approved by the Chief, NuclearEngineer, or his Deputy," should be "may be approved by the Chief Nuclear Engineer, or his Deputy."14.35TS 6.3 (1), p. 33. There should be a space between "and systems" in the 1st line.

14.36TS 6.5 (1) c), p. 34. It appears the information under "d)" belongs with "c)."