ML16279A523

Rhode Island Atomic Energy Commission Consolidated Responses to Request for Additional Information Regarding License Renewal(Redacted)

ML16279A523
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Site:	Rhode Island Atomic Energy Commission
Issue date:	04/28/2014
From:	Rhode Island Nuclear Science Center
To:	Office of Nuclear Reactor Regulation
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ML14126A190	List: ML14126A192 ML14126A195 ML16279A523
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RHODE ISLAND ATOMIC ENERGY COMMISSION RESEARCH REACTOR LICENSE NO. R-95 DOCKET NO. 50-193 RESPONSES TO NRC STAFF REQUEST FOR ADDITIONAL INFORMATION FOR LICENSE RENEWAL REVIEW REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

• REDACTED TEXT AND FIGURES BLACKED OUR OR DENOTED BY BRACKETS

Overall RAI Responses 140423 21 Chapter 2 of the SAR contains multiple section headings with no related information.

Provide the omitted information.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 2.3.1 General and Local Climate The average annual temperature in Rhode Island is 507F (10°C). At Providence the temperature ranges from an average of 28°F (-2'C) in January to 731F (23°C) in July. The record high temperature, 1047F (40'C), was registered in Providence on 2 August 1975; the record low, -23°F (-3 1C), at Kingston on 11 January 1942. In Providence, the average annual precipitation (1971-2000) was 46.5 in (118 cm); snowfall averages 37 in (94 cm) a year.

2.3.2.3 Humidity Rhode Island has a humid climate, with cold winters and short summers. The humidity varies depending on wind direction and ocean temperature.

2.4.3 Sanitary Sewer System The Rhode Island Nuclear Science Center is connected to the Narragansett /

South Kingstown municipal sewer system, which has a final outflow to the ocean, 2.4.4 Ground Water Ground water and site drainage flows directly into the Narragansett Bay. The original site study report performed by General Electric cited this as being one of the reasons that the reactor was built at this site.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The major potential impact that the facility could have on groundwater arises from the fact that there is a small amount of tritium production in the reactor pool water. If the pool were to have a significant leak, this would be released to the ground or sewer system. The tritium concentration in the pool has been measured to be 3 X 10-4 pCi / cc.

10 CFR 20 Appendix B Table 3 indicates that the concentration limit for the release of tritium to the sewer system is 1 X 10-' [tCi / cc. Consequently, the concentration in the I

reactor pool is two orders of magnitude less than the release limit. This indicates that there is no significant potential facility impact on the groundwater.

4.1 Figures 4-5, 4-6, 4-7, and 4-8 were omitted from the SAR. Provide these figures.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 The figures of Chapter 4, "Reactor Description", were incorrectly referenced throughout the chapter. The following table shows each reference, which page it is found on, the figure it references and the page they can be found, the corrected reference in an updated version of the SAR, and a description of the figure.

Original Original Current Text Page Figure Page Figure Page Description 4-1 4-3 4-1 4-25 4-1 4-4 14 Element Core 4-1 4-3 - 4-2 4-5 Core Assembly Cutaway View of Flow 5-3 4-6 5-3 5-11 4-3 4-9 Channels 4-1 4-7 4-4 4-28 4-4 4-11 Start-up to Equilibrium Cores 4-6 4-7 4-4 4-28 4-4 4-11 Start-up to Equilibrium Cores 4-6 4-8 4-4 4-28 4-4 4-11 Start-up to Equilibrium Cores Hot Channel Fuel Surface 4-7 4-14 - Graph 4-8 4-16 - LEU core Flow vs DP Graph 4-1 4-21 4-2 4-26 4-5 4-26 17 Element Core 4 4-21 4-1,4-2 4-25, 4-26 4-1,4-5 4-4, 4-26 14,17 Element Cores Power Peaking Factors in 4-5 4-5 4-3 4-27 4-6 4-27 Thermal Hydraulic Calculations The "Hot Channel Fuel Surface Graph" and "LEU Core Flow vs DP Graph" figures are part of the new analysis being performed by Argonne National Labs. See the updated analysis for this data. The "Power Peaking Factors in Thermal Hydraulic Calculations" figure is also outdated.

2

14 Fuel Element Core Figure 4-1 3

Reactor Core Assembly Figure 4-2 4

Cutaway view of Flow Channels Figure 4-3 5

Startup-to-Equilibrium Core Configurations with Remaining Uranium (grams of U 235)

Figure 4-4.

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17 Fuel Element Core Design Figure 4-5 7

Power Peaking Factors in Thermal Hydraulic Calculations E 0 C E 0 3 G G GG G 8 0.807 0.30 0.807 0.23 .:-44 0.823 V.768 1.169 1 748 1.7G$ 1.151 1.766 1.536 1 536 1532 1.53$ 1.5; 1.536 2.231 1.727 7

2.192 1 57U 2.192 2.231 5.9% 6.61A 5.8% 4J% 6.6% 4.8%

1.011 1 261 1011 1.012 1317 1.052 1630 163 1630 1$.31 1.372 1.631 1,56 1.53 1.3 1.536 1.536 1.538 2531 3.0IN 2.531 2_M35 2.179 2.635 7.2% 9.0% 7.2% 8.2% 7.7% 6.2%

1151 1.151 1.239 1.233 1654 165 5 1.536 1.536 1.5a6 0 1as 2.925 2925 3.122 8.122

&2% 8.2% 7.3% 7.3%

1.009 1258 1.009 1.134 1.413 1.134 1.632 1.583 1.632 4 1.591 1,500 1.591 1.53S 1.536 1.53M 4 1.5 136 1.536 2.530 3.019 2.530 2.770 3.256 2.770 72% 90% 7.2% &7% 6.3% 6.7%

0.818 0961 0.818 0.9*3 1.180 0.993 1.7At 1312 1 749 1.654 1.410 1,654 3 3 11.36 1 *X 1.536 1.536 1.536 1M.5$

2.197 1 WS 2.197 2.538 2.M 2.331 5.8% 69% 5.8% 5A)% 6.9% 69 Reog. 2 Reg. 2 Rod Rod

0. o0.602 0.526 2.228 2.073 2.228 GG 1.536 1.536 1.536 1.800 1.917 1.800 3.1% 3-5% 3.4%

Startup Core Startup Core with 3 more Fuel Element3 Radial peawng factor Planar peafcrg factor Axdia oftkirV~ factor Total peaking farior

% of total pom~vr rinelement Figure 4-6 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, bum up, etc.). Verify that those parameters important 8

to safety are included in or bounded by the requirements in the technical specifications (TS).

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Fuel qualification for reactor operation at 2 MW was established during the conversion from HEU fuel to LEU fuel. NUREG 1313 has a total review of the fuel qualification for research reactors, provided that the core in question meets the parameters under which the tests were done. These parameters were taken into consideration when the current safety analysis was performed. This qualification is valid up to 100% fuel burn-up.

4.3 Section 4.1 states that the fuel composition is U2Si2, while Section 13.2.1 states that the fuel composition is U3Si2. Clarify which composition is correct for the RNSC reactor fuel First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 U 3 Si 2 is correct.

4.4 Section 4.2.2. Provide a summary description of the program for shim safety blade and regulating blade inspection and replacement.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 RINSC performs an annual inspection of the shim safety blades by raising each blade to it's full out position and making a visual inspection of each blade. Drive speeds and blade drop times are measured in order to verify that there is no indication that blade motion is hindered. An annual control rod calibration is performed to determine that the reactivity worth of the control blades provide sufficient shutdown margin to meet RINSC Technical Specification Limits. The current control blades have been in use since 1967. At present, there has been no indication that the control blade worth have been diminished significantly. Consequently, there is no program or plan to replace control blades over the next twenty year licensing period.

4.5 Section 4.2.3 states that the graphite reflectors are designed for expansion "from an integrated flux of 2X1021 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.

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Sixth Response to RAI Dated April 13, 2010 Submitted December 7, 2010 The graphite reflector element is a block contained in a 3-inch square aluminum can with handles to allow remote handling. It should be noted that graphite has been used for many years as a reflector and crumbling or other catastrophic failures of graphite pieces have never occurred. AGOT reactor-grade graphite is used to avoid trace boron contamination. Graphite 2 is known to undergo several changes when exposed to neutron irradiation:

• Dimensional change due to neutron induced swelling

• Elastic modulus change as measured by the impulse excitation technique

• Coefficient of thermal expansion change

• Thermal conductivity change

• Electrical resistivity change

• Irradiation-induced creep SAR Section 4.2.3 discusses the expansion of graphite due to irradiation and gas evolution. The design allows for a maximum increase in graphite dimensions of 1.1%

due to irradiation growth and gas evolution. The SAR suggests that the graphite will be fine 10212 neutrons /cm 2 . Our maximum flux at the center of theupcore to anis integrated flux estimated to be of 2 Xn!cm 1013 -s. Therefore to reach 2 X 1021 nvt:

[2 X 1021 n/cm 2] / [1013 n / cm 2-s] = 2 X 108 seconds or about 5.6 X l04 hours of operation. At 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />s/day operation, this amounts to 8,000 days of full-power operation or about 22 years of full-power operation. It should be noted that the graphite reflectors are now on the periphery rather than at the center of the core (LEU core configuration).

The neutron flux at that location is at least an order of magnitude less than that at the center of the core. Thus, a conservative estimate of the time needed to reach the integrated flux would be 80,000 days or more of full-power operation. That amounts to about 219 years or well beyond the requested period of the renewed license.

2. Marsden, B.J., Preston, S. D., Wickham, A. J., "Evaluation of graphite safety issues for the British production piles at Windscale," IAEA-TECDOC-1043, September 1997.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The core is suspended from a bridge that is mounted over the top of the reactor pool.

General Electric Drawing 198E299 shows how the suspension frame holds the control rod housings and core grid box together. The reactor pool sits on a military gun pad.

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The pool is constructed of a large mass of reinforced concrete. Consequently, in the event of an earthquake, the pool, bridge, and core are expected to move as a unit. The shim safety control blades fit inside a shroud, which is part of the core grid box. When the shim safety control blades are not fully inserted into the core, each blade is suspended by an electromagnet which holds it in its withdrawn position. When fully withdrawn, the ends of the blades remain inside the shroud, which prevents misalignment on release. A significant earthquake would likely shake the shim safety blades free from the magnets. However, the reactor is fitted with a seismic scram device, which scrams the reactor upon detection of an earth tremor. General Electric Drawing 197E647 shows that the total spacing between the control blades and the shrouds is 0.125 inches.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 There are three neutron sources that are available for use as a start-up source. The first is a pair of PuBe sources that are stored together in a common container, the second is an SbBe source, and the third consists of the Be reflectors in the core. The reactor Start-Up channel has a neutron count interlock of 3 cps, which is the minimum neutron count rate that must be present in the core in order to start the reactor tip. Any one of the available neutron sources may be used as a start-up source, however given the typical reactor operating schedule, the Be reflector elements are generally used as the neutron start-up source. Gamma decay from fission fragments interact with the Be to produce a sufficient level of photo neutrons that the external sources of neutrons are generally not needed in order to have a neutron count rate of at least 3 cps in the core. It is not anticipated that there is risk of damage to the sources. The PuBe sources are leak tested every six months. The Be reflector elements are inspected as part of the fuel element inspection program.

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.

Sixth Response to RAI Dated April 13, 2010 Submitted December 7, 2010 Relative to commercial power reactors, the RINSC reactor operates at very low power; temperature, and pressure. Consequently, damage to the core support structure due to 11

radiation exposure and thermal aging will be significantly less than the damage typical of power reactors. No reports of significant radiation damage to core components of small research reactors have been published. Since the power industry does damage studies to show that their facilities can continue to operate safely with extended lifetimes, it is reasonable to assume that research reactors can safely operate within similar lifetimes.

The reactor core support consists of a suspension framne which is bolted to a moveable bridge, operated by a hand crank, which can relocate the entire core plus core support structure to various positions in the reactor pool. The four comers of the structure are occupied by the suspension posts. These comer posts connect the grid plate to the reactor bridge. The core suspension system includes a locating plate, made of heavy steel that spans the upper end of the suspension frame to provide support and location for the control blade drive mechanisms. The control blade drive guide tubes are flanged to the bottom of this locating plate. Core elements are contained in a grid box that is enclosed on four sides to confine the flow of cooling water between elements (See Fig.

4-2 of the SAR). The grid box assembly, including the drive mechanisms, is supported by the suspension frame. The elements that make up the core sit on a 7 x 9 grid plate with the four comer positions occupied by the suspension frame comer posts.

This core support system was designed to support the weight of the core plus control and cooling elements. The design has satisfactorily supported the weight of the components for over forty years and there is no credible reason why it should not continue to function as designed. No appreciable deterioration of any components of the support structure has been seen during inspections.

The core support structural materials are predominantly made of 6061-T6 Al. In order to minimize corrosion of the aluminum, reactor pool water pH and conductivity (resistivity) levels are measured weekly to verify that the values are within the RINSC Technical Specification limits (pH between 5.5 and 7.5; resistivity greater than 500 kf2/cm). As described in Section 5.5.1 of the SAR, make-up water for the pool passes through a five micron filter, an activated charcoal filter, two mixed bed demineralizers and a one micron filter before entering the pool. The pH and conductivity are measured weekly to verify that the water in the pool is within the specification limits.

The core fuel cladding material is also 6061-T6 Al. We perform an annual fuel element inspection that would provide another indication of whether or not aluminum core materials are beginning to suffer from corrosion, radiation damage or thermal stress.

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.

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Sixth Response to. RAI Dated April 13, 2010 Submitted December 7, 2010 Relative to commercial power reactors, the RINSC reactor operates at very low power, temperature, and pressure. Consequently, damage to the biological shield and the pool liner due to radiation exposure and thermal aging will be significantly less than the damage to similar structures typical of power reactors. No reports of significant radiation damage to biological shields or pool liners of small research reactors have been published. Since the power industry does damage studies to show that their facilities can continue to operate safely with extended lifetimes, it is reasonable to assume that research reactors can safely operate within similar lifetimes.

The reactor pool is surrounded by thick reinforced concrete. The inner surface of the concrete is lined with 1/44-in. thick aluminum. The liner is the primary pool water containment vessel. For the same reasons given in response to RAI 4.8, corrosion of the aluminum liner is expected to be minimal over the extended lifetime of the reactor. The same monitoring (pH and conductivity) used for the aluminum fuel clad would also alert operators to any corrosion of the liner. Required annual inspections of the pool supplement weekly water monitoring to confirm the integrity of the pool liner.

Any significant degradation of the liner, whether from chemical or mechanical causes, that could lead to pool water leakage would be detected by routine monitoring of the makeup water system.

The combination of the water pool and the surrounding concrete provide a biological shield for facility workers that keeps the dose rate below 1 mrem/hr at all points above and outside the pool area (SAR, Section 4.4). Water level monitors and radiation monitors ensure that the water depth is sufficient (approximately 24 ft) to shield personnel near the top of the pool. During routine operations radiation surveys are performed to monitor dose rates (SAR Ch. 11).

The radiation attenuation properties of the pool water are based on the nuclear properties of the water and the attenuation level will not change over time as long as the water level is maintained. The aluminum liner is a minor contributor to the biological shielding relative to the water and concrete, but as explained above it is not expected to deteriorate over the lifetime of the facility. While concrete is susceptible to thermal and radiation damage, the low power and low temperature of the RINSC reactor will not lead to any degradation of the concrete over the lifetime of the reactor.

3 A survey of aging effects on concrete was performed at the Idaho National Laboratory .

According to this report, for conditions of radiation flux up to 2 x 1019 nvt (thermal) and temperatures to 120 'C, radiation danmage to the type of concrete used in our facility was insignificant, while other types show considerable loss of strength (specifically high alumina cement concrete). All effects on concrete due to radiation, per se, were too slight to reliably measure because of the gross effects from the increased temperature during exposure. Generally speaking, the threshold of degradation in the concrete is approximately 95 'C.

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The neutron flux at the core end of the beam ports (basically the inner surface of the concrete shielding) is approximately 1 to 4 x 1012 n/cm 2-s. Assuming 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of operation per week, ten years of full-power operation would just approach the 1019 nvt threshold for the most susceptible type of concrete (not used at the RINSC). However, as noted in the referenced report, radiation damage is basically not measureable compared to temperature effects. The safety limit for the pool water temperature is 130

'F (54.4 'C). This is substantially below the 95 'C threshold value for thermal damage to the concrete. Based on these considerations it can be concluded that the biological shield will not deteriorate over the extended lifetime of the facility.

3. Literature Review of the EJfects of Radiation and Temperature on the Aging of Concrete, INEEL/EA*-04-02319, September 2004 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'.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The reactor kinetics parameters were recomputed for the equilibrium core using standard perturbation methods with Argonne's VARI3D computer code'. VARI3D uses the DIF3D. diffusion theory code to compute real and adjoint flux solutions for use in perturbation theory.

The kinetic parameters were recalculated here because the models that were used to calculate the values in Table 4-3 in the early 1990's are no longer available. The table below shows that the recomputed values of the delayed neutron fraction and the prompt neutron generation time agree very well with those in Table 4-3.

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Equilibrium Core Kinetic Parameters 2004 2010 Calculation Report With equilibrium No Xe Xe Delayed Neutron Fraction, P3-eff 0.764 0.755 0.756

(%)

Neutron Generation Time, ýts 68.3 69.4 68.6 Delayed Parameters by Family Group 3(1) 2.6580E-04 2.6591E-04 P3(2) 1.3707E-03 1.3713E-03 13(3) 1.3188E-03 1.3194E-03 P3(4) 2.8985E-03 2.8996E-03 P3(5) 1.1990E-03 1.1995E-03 P3(6) 5.0074E-04 5.0095E-04

),(I) 1.3337E-02 1.3337E-02 X(2) 3.2712E-02 3.2712E-02 X(3) 1.2075E-0 1 1.2075E-01 X*(4) 3.0279E-01 3.0279E-01 X(5) 8.4966E-01 8.4966E-01 X(6) 2.8538E+00 2.8538E+00 IA.P. Olson to J. Matos, "Availability of the VARI3D Code on Linux", Argonne National Laboratory Internal Memorandum, February 11, 2004.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Coefficients of reactivity for temperature, void, and power were recomputed for the equilibrium core for these RAI responses because the models and results in Section 5 were originally computed in the early 1990's and are no longer available.

Neutron cross sections in seven energy groups as functions of moderator temperature, fuel temperature, and coolant void fraction were prepared using the WIM/ANL cross section generation codel. Keff values were computed using the DIF3D diffusion theory code. Coefficients of reactivity were determined from these data.

RINSC does not make measurements of these parameters. The accepted values for these coefficients are the values provided by the model.

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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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Separate reactivity coefficients were recalculated for this RAI for increasing coolant temperature only and increasing coolant density only using the methods described in RAI 4.11. The reactivity coefficients are shown below for a coolant temperature range between 20 'C and 100 'C.

Equilibrium Fission Products Change Coolant Temperature Only Change Coolant Density Only Cumulative Cumulative Reactivity Reactivity Change x Water Change x Water Water 103 Temp. Reactivity, 10 3 Temp. Density Reactivity, Rel. to 20 oC oC mg/ml 0

C dk/k Rel. to 20 'C dk/k 20 0.027602 0.00 20 0.99811 0.027249 0.00 30 0.026447 -1.1551 30 0.99564 0.026636 -0.6131 40 0.025298 -2.3044 40 0.99227 0.025881 -1.3678 50 0.024155 -3.4478 50 0.98810 0.024985 -2.2641 60 0.0230 17 -4.5855 60 0.98323 0.023947 -3.3020 70 0.021885 -5.7174 70 0.97773 0.022767 -4.4815 80 0.020759 -6.8434 80 0.97171 0.021446 -5.8026 90 0.019639 -7.9636 90 0.96525 0.019983 -7.2653 100 0.018524 -9.0781 100 0.95845 0.018379 -8.8696 16

Cumulative Reactivity Change as a Function of Coolant Temperature Only 0.0

-1.0 0

• -2.0

-3.0

-4.0

-5.0 C6.0 y=2.S03EOx-08*70- 7OE4x +2.328E+00

-7.0

-9.0

-10.0 Coolant Temperature, C Cum ulative Reactivity Change Versus Coolant Tem perature Corresponding to Different Coolant Densities 0

-3

-4

-5

-7

-8 .... ... ....

-9

-10 Coolant Tem perature, C 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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Reactivity coefficients were recalculated for this RAI for increasing the fuel temperature only using the methods described in RAI 4.11. The reactivity coefficients are shown below for a fuel temperature range between 20 'C and 600 'C.

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Equilibrium Fission Products Change Fuel Temperature Only Cumulative Reactivity Change x 103 Reactivity, Rel. to 20 Ak/k 0C 20 0.027479 0.0 30 0.027280 -0.19854 40 0.027083 -0.39578 50 0.026887 -0.59172 60 0.026693 -0.78635 70 0.026499 -0.97968 80 0.026307 -1.17170 90 0.026117 -1.36243 100 0.025927 -1.55185 150 0.025000 -2.47940 200 0.024105 -3.37437 300 0.022412 -5.06657 400 0.020851 -6.62845 500 0.019419 -8.06001 600 0.018118 -9.36125 Cumulative Reactivity Change as a Function of Fuel Temperature 2

0 0 200 300 400 8oo 600 7 10

-2 CD I-- -4

-6 01 y = 6.16E-06x"- 2.018E-02x + 4.010E-01 a: -8

-10 Fuel Temperature, C 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) 18

Tenth Response to RAI Dated April 13, 2010 Submitted July 15, 2011 All of the neutronics analysis was done with the assumption that the central flux trap was filled with water. Consequently, power peaking under this condition is already taken into account.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 In Section 4.6 the original PLTEMP code' was used to determine many of the thermal-hydraulic characteristics of the reactor core. This code has been superseded by the PLTEMP/ANL V4.0 code.2 The analysis of steady-state forced-convection operation has been redone using the newer code.

Although both the old and the newer code have models that can be used to obtain the flow distribution in a reactor core, a more direct, transparent, and tractable approach was taken in the new analysis. A hydraulics model of the RINSC core was developed based on engineering fundamentals. The equations used in the analysis are provided in the memo dated 3 September 2010 from Earl E. Feldman to James E. Matos entitled "Steady State Thermal Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor". The application of the equations are explained in detail. Key internediate results for a reactor flow rate of 1580 gpm are given in tables so that one can verify that the analysis was performed correctly. The hydraulics analysis yielded the flow rate for each fueled channel in the reactor as a function of total reactor flow rate.

In the next phase of the analysis, the individual channel flow rates obtained in the previous phase were used to perform a thermal analysis of the core. The limiting channel was identified as an internal channel in element D6 and next to the highest power fuel plate. The new PLTEMP code was used to perform thermal analysis of this limiting channel. The highest power fuel plate bounded by two channels, each with half of the flow area of single channel, was modeled. This model is simple and easy to check. For the channel flow rate corresponding to a reactor flow rate of 1580 gpm the PLTEMP results for the powers at which the onset of nucleate boiling and the onset of flow instability were predicted to occur were verified by a hand calculation. Key intermediate results used in the verification are given in tables so that one can verify that the results are correct.

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In the new analysis, key results such as coolant flow rates, bulk coolant and clad surface temperatures, the conditions that produce onset of nucleate boiling and the conditions that cause flow instability are either performed by a hand calculation or are verified by one. Thus, PLTEMP code validation, benchmarking, and code uncertainty are less relevant because results that are essential to demonstrate the safety of the RINSC core have been hand calculated or hand checked. Moreover, the PLTEMP/ANL code used in the new analysis is based on an evolutionary sequence of "PLTEMP" codes in use at ANL for the past 26 years. 3-9 Validation of PLTEMP/ANL has followed standard practice in any code development task, where comparisons are made with other codes, with measurements, and with hand calculations where possible. Many examples of validation are given in the PLTEMP/ANL V4.0 manual 2, and in References 3-9.

The users guide to the PLTEMP/ANL V4.0 code. 2 is provided.

Also provided is Reference 4-6:

K. Mishima, K. Kanda and T. Shibata, "Thermal-hydraulic Analysis for Core Conversion to. Use of Low-Enrichment-uranium Fuels in KUR," KURRI-TR-258 (1984).

References:

1. Kaichiro Mishima, Keiji Kanda, and Toshikazu Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use of Low-Enriched Uranium Fuels in the KUR," Proceedings of the 1984 InternationalMeeting on Reduced Enrichmentfor Research and Test Reactors, Argonne National Laboratory, October 15-18, 1984.

2. Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

3. K. Mishima, K. Kanda and T. Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use of Low-enrichment Uranium Fuels in the KUR," KURRI-TR-258, Research Reactor Institute, Kyoto University, Dec. 7, 1984.

4. K. Mishima, K. Kanda and T. Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use of Low-Enriched Uranium Fuels in the KUR,"

ANL/RERTR/TM-6, CONF-8410173, p. 375, 1984.

5. W. L. Woodruff and K. Mishima, "Neutronics and Thermal-Hydraulics Analysis of KUHFR," ANL/RERTR/TM-3, CONF-801144, p. 579, 1980.

6. W. L. Woodruff, "Some Neutronics and Thermal-hydraulics Codes for Reactor Analysis Using Personal Computers," Proc. Int. Mtg. on Reduced Enrichment for Research and Test Reactors, Newport, RI, Sept. 23-27, 1990, CONF-9009108 (ANL/RERTR/TM- 18), Argonne National Laboratory (1993).

7. W. L. Woodruff, J. R. Deen and C. Papastergiou., "Transient Analyses and Thermal-hydraulic Safety Margins for the Greek Research Reactor (GRR1)," Proc. Int. Mtg.

on Reduced Enrichment for Research and Test Reactors, Williamsburg, VA, Sept.

19-23, 1994. CONF-9409107 (ANL/RERTR/TM-20), Argonne National Laboratory (1997).

20

8. W. L. Woodruff, "A Kinetics and Thermal-hydraulics Capability for the Analysis of Research Reactors," Nucl. Technol., Volume 64, 196 (1983).

W. L. Woodruff and R. S. Smith, "A Users Guide for the ANL Version of the PARET Code, PARET/ANL (2001 Rev.)," ANL/RERTR/TM-16, Mar. 2001.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 In Section 4.6 the original PLTEMP code' was used to determine many of the thermal-hydraulic characteristics of the reactor core. This code has been superseded by the PLTEMP/ANL V4.0 code.2 The analysis of steady-state forced-convection operation has been redone using the newer. code and is attached as a new Section 4.6 of the.SAR, "4.6 Steady-State Thermal-Hydraulic Analysis" (Reference BB).

Although both the old and the newer code have models that can be used to obtain the flow distribution in a reactor core, a more direct, transparent, and tractable approach was taken in the new analysis. A hydraulics model of the RINSC. core was developed based on engineering fundamentals. The equations used in the analysis are provided. Their application is explained in detail. Key intermediate results for a reactor flow rate of 1580 gpm are given in tables so that one can verify that the analysis was performed correctly. The hydraulics analysis yielded the flow rate for each fueled channel in the reactor as a function of total reactor flow rate.

In the next phase of the analysis, the individual channel flow rates obtained in the previous phase were used to perform a thermal analysis of the core. The limiting channel was identified as an internal channel in element D6 and next to the highest power fuel plate. The new PLTEMP. code was used to perform thermal analysis of this limiting channel. The highest power fuel plate bounded by two channels, each with half of the flow area of single channel, was modeled. This model is simple and easy to check. For the channel flow rate corresponding to a reactor flow rate of 1580 gpm the PLTEMP results for the powers at which the onset of nucleate boiling and the onset of flow instability were predicted to occur were verified by a hand calculation. Key intermediate results used in the verification are given in tables so that one can verify that the results are correct.

In the new analysis, key results such as coolant flow rates, bulk coolant and clad surface temperatures, the conditions that produce onset of nucleate boiling and the conditions that cause flow instability are either performed by a hand calculation or are verified by one. Thus, PLTEMP code validation, benchmarking, and code uncertainty are less relevant because results that are essential to demonstrate the safety of the RINSC core have been hand calculated or hand checked. Moreover, the PLTEMP/ANL code used in the new analysis is based on an evolutionary sequence of "PLTEMP" codes in use at ANL for the past 26 years. 39 Validation of PLTEMP/ANL has followed standard practice in any code development task, where comparisons are made with other codes, with measurements, and with hand calculations where possible. Many examples of validation are given in the PLTEMP/ANL V4.0 manual 2, and in References 3-9.

21

As a copy of the original PLTEMP code,1 reference 4.6 in the 2004 SAR is provided.

Also provided is Reference 4-6:

K. Mishima, K. Kanda and T. Shibata, "Thermal-hydraulic Analysis for Core Conversion to Use of Low-Enrichment-uranium Fuels in KUR," KURRI-TR-258 (1984).

References:

9. Kaichiro Mishima, Keiji Kanda, and Toshikazu Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use oaf Low-Enriched Uranium Fuels in the KUR," Proceedings of the 1984 InternationalMeeting on Reduced Enrichmentfor Research and Test Reactors, Argonne National Laboratory, October 15-18, 1984.

10. Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

11. K. Mishima, K. Kanda and T. Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use of Low-enrichtnent Uranium Fuels in the KUR," KURRI-TR-258, Research Reactor Institute, Kyoto University, Dec. 7, 1984.

12. K. Mishima, K. Kanda and T. Shibata, "Thermal-Hydraulic Analysis for Core Conversion to the Use of Low-Enriched Uranium Fuels in the KUR,"

ANL/RERTRITM-6, CONF-8410173, p. 375, 1984.

13. W. L. Woodruff and K. Mishima, "Neutronics and Thermal-Hydraulics Analysis of KUHFR," ANL/RERTRITM-3, CONF-801144, p. 579, 1980.

14. W. L. Woodruff, "Some Neutronics and Thermal-hydraulics Codes for Reactor Analysis Using Personal Computers," Proc. Int. Mtg. on Reduced Enrichment for Research and Test Reactors, Newport, RI, Sept. 23-27, 1990, CONF-9009108 (ANL/RERTR/TM-18), Argonne National Laboratory (1993).

15. W. L. Woodruff, J. R. Deen and C. Papastergiou, "Transient Analyses and Thermal-hydraulic Safety Margins for the Greek Research Reactor (GRR1)," Proc. Int. Mtg.

on Reduced Enrichment for Research and Test Reactors, Williamsburg, VA, Sept.

19-23, 1994, CONF-9409107 (ANL/RERTR/TM-20), Argonne National Laboratory (1997).

16. W. L. Woodruff, "A Kinetics and Thermal-hydraulics Capability for the Analysis of Research Reactors," Nucl. Technol., Volume 64, 196 (1983).

9. W. L. Woodruff and R. S. Smith, "A Users Guide for the ANL Version of the PARET Code, PARET/ANL (2001 Rev.)," ANL/RERTR/TM-16, Mar. 2001.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Reference 4-Y will be re-numbered in the Final Safety Analysis Report, and will refer to "Report on the Determination of the Hot Spot Factors for the Rhode Island Nuclear 22

Science Center Research Reactor Using LEU Fuel", Eugene Spring, August 24, 1989. A copy of this report is attached.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Section 4.6 in the 2004 SAR will be superseded by a new Section 4.6. In the new analysis the values of inlet coolant temperature and coolant height used in the analysis are based on the Technical Specifications Safety Limit values for reactor power (2.4 MW), reactor flow (1580 gpm), outlet temperature (1250 F), and pool water depth (23.54 feet above the active core). These are slightly more limiting than was necessary since the original Limiting Safety System Setting values would have been sufficient.

The added pressure at the top of the active core due to the weight of the water at a depth of 23.54 feet is 989.8 kg/m3 x 9.80665 m/s2 x 23.54 ft x 0.0254 x 12 rn/ft = 69645 Pa =

0.696 bar. Narragansett, Rhode Island is at 20 feet above sea level, where atmospheric pressure is 1.013 bar. Thus, the absolute pressure at the top of the active core is 1.013 +

0.696 bar =.1.709 bar. A pressure of 1.7 bar was used in the analysis. The enthalpy at the core exit was obtained for this pressure and the 125' F via the NIST Steam Tables.

The power to flow ratio of 2.4 MW / 1580 gpm yielded the enthalpy rise from core inlet to core outlet. The outlet enthalpy minus the enthalpy rise yielded the inlet enthalpy.

The inlet enthalpy and 1.7 bar pressure yielded the inlet temperature of 114.5' F. This' was rounded up to 115'F and used as the inlet temperature for the new Section 4.6 thermal analyses. The density of water at 11 5F and 1.7 bar is 989.8 kg/m3. which is the value use above to determine the added pressure due to the depth of the water.

Thus, the direct answer to the question is 115' F and 23.54 feet of water above the active core. The justification is that these values are consistent with the Technical Specifications Safety Limit values, which bound the Limiting Safety System Setting values.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 4.6 in the 2004 SAR has been superseded by a new Section 4.6 and is attached as "4.6 Steady-State Thermal-Hydraulic Analysis" (Reference BB). In the new analysis the values of inlet coolant temperature and coolant height used in the analysis are based on the Technical Specifications Safety Limit values for reactor power (2.4 MW), reactor flow (1580 gpm), outlet temperature (125' F), and pool water depth (23.54 feet above the active core). These are slightly more limiting than was necessary since Limiting Safety System Setting values would have been sufficient.

The added pressure at the top of the active core due to the weight of the water at a depth of 23.54 feet is 989.8 kg/m3 x 9.80665 m/s2 x 23.54 ft x 0.0254 x 12 m/ft = 69645 Pa =

0.696 bar. Narragansett, Rhode Island is at 20 feet above sea level, where atmospheric pressure is 1.013 bar. Thus, the absolute pressure at the top of the active core is 1.013 +

23

0.696 bar = 1.709 bar. A pressure of 1.7 bar was used in the analysis. The enthalpy at the core exit was obtained for this pressure and the 1250 F via the NIST Steam Tables.

The power to flow ratio of 2.4 MW / 1580 gpm yielded the enthalpy rise from core inlet to core outlet. The outlet enthalpy minus the enthalpy rise yielded the inlet enthalpy.

The inlet enthalpy and 1.7 bar pressure yielded the inlet temperature of 114.50 F. This was rounded up to 115'F and used as the inlet temperature for the new Section 4.6 thermal analyses. The density of water at 1157F and 1.7 bar is 989.8 kg/im 3, which is the value use above to determine the added pressure due to the depth of the water.

Thus, the direct answer to the question is 1150 F and 23.54 feet of water above the active core. The justification is that these values are consistent with the Technical Specifications Safety Limit values, which bound the Limiting Safety System Setting values.

4.18 Section 4.6.2. Confirm that the units for the values of Tsurface, Tsat, and Tonb found in Table 4-9 and Table 4-10 are degrees C.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 It is confirmed that the values of Tsurface, Tsat, and Tonb 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 Tonb 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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Section 4.6 in the 2004 SAR will be superseded by a new Section 4.6. In the new analysis the values of the clad surface temperature (Tsrface) and the values of the surface temperature that cause onset of nucleate boiling (Tob) are calculated by the PLTEMP/ANL V4.0 code.' A copy of the manual of the code is provided. The key heat transfer correlations used in the analysis of onset of nucleate boiling are the Bergles and Rohsenow correlation, which is used to determine the amount of superheat to reach onset of nucleate boiling at the clad surface, and the Dittus-Boelter correlation, which is used to determine the Nusselt number. These two correlations and a full description of the models, assumptions, and correlations of the PLTEMP/ANL code can be found in the PLTEMP/ANL V4.0 code manual.

In the analysis for onset of nucleate boiling, the PLTEMP/ANL V4.0 code was used to analyze the limiting channel in the reactor. This analysis was performed once for each of eight reactor flow rates, spanning the flow range from 1000 to 2200 gpm. Since the model represented only the limiting channel, it was easy to hand-check the PLTEMP/ANL V4.0 results. This hand-check was performed for a reactor flow rate of 1580 gpm. At this flow rate the limiting channel has a flow rate of 0.2210 kg/s. For this 24

flow rate PLTEMP/ANL V4.0 found that the channel power is 22.80 kW, which corresponds to a reactor power of 4.72 MW. The code results also indicate the axial locations where onset of nucleate boiling was first reached. Key values of the verification are provided in the new Section 4.6 in a table, which is repeated below as Table 4.19-1. Additional specifics of the analysis can be found in the memo dated 3 September 2010 from Earl E. Feldman to James E. Matos entitled "Steady State Thermal Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor", including a listing of the PLTEMP/ANL V4.0 input used for this analysis.

25

Table 4.19 Verification of PLTEMP/ANL Onset of Nucleate Boiling Prediction at the Limiting Axial Location for a Core Flow Rate of 1580 gpm Quantity Hand PLTEMP Calculation Channel Dimensions Thickness, in 0.088 Width, in 2.62 Heated Length, in 23.25 Heat Width, in 2.395 Wetted Perimeter, m 0.1376 Hydraulic Diameter, m 0.004325 Heat Transfer Area (2 Faces), m2 0.07185 Channel Flow Rate, kg/s 0.2210 Channel Power, kW 22.80 Pressure, bar 1.7 Inlet Temperature, C 46.11 Saturation Temperature, C 115.15 Cp @ 55 C, kJ/kg-C 4.1828 At Onset of Nucleate Boiling Location Layer 13 13 Channel Power to Middle of Layer, kW 15.64 Channel Power to Exit of Layer, kW 16.56 Local Peak-to-Average Power 1.2894 Bulk Temperature at Middle ofLayer w/o Hot Chan. Fac., C 63.03 63.01 Random Hot Channel Factor on AT bulk 1.24 Bulk Temperature with Hot Channel Factor, C 67.09 67.06 Bulk Temperature at Exit of Layer w/o Hot Chan. Fac., C 68.32 Viscosity, Pa-s 4.391 OE-4 Reynolds Number 14634 Thermal Conductivity, W/m-C 0.65817 Cp, kJ/kg-C 4.1867 Prandtl Number 2.793 Nusselt Number without Hot Channel Factors 74.55 Global Film Coefficient Hot Channel Factor 1.2 Film Coefficient with Hot Channel Factor, W/m'-C 9454 Heat Flux without Hot Channel Factors, MW/mi 0.4092 0.4091 Random Hot Channel Factor on AT film 1.28 Film Temperature Rise with Hot Channel Factor, C 55.77 Clad Surface Temperature with All Hot Channel Factors, C 122.5 122.5 Random Hot Channel Factor on Heat Flux 1.23 Heat Flux with Hot Channel Factors, MW/m 2 0.4779 ATsaturation based on Bergles and Rohsenow, C 7.41 Surface Temperature For Onset of Nucleate Boiling, C 122.6 122.6 Table 4.19 Verification of PLTEMP/ANL Onset of Nucleate Boiling Prediction at the Limiting Axial Location for a Core Flow Rate of 1580 gpm 26

Reference:

Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 4.6 in the 2004 SAR has been superseded by a new Section 4.6 and is attached as "4.6 Steady-State Thermal-Hydraulic Analysis" (Reference BB). In the new analysis the values of the clad surface temperature (Tsurface) and the values of the surface temperature that cause onset of nucleate boiling (Tonb) are calculated by the PLTEMP/ANL V4.0 code.' A copy of the manual of the code is provided. The key heat transfer correlations used in the analysis of onset of nucleate boiling are the Bergles and Rohenow correlation, which is used to determine the amount of superheat to reach onset of nucleate boiling at the clad surface, and the Dittus-Boelter correlation, which is use to determine the Nusselt number. These two correlations and a full description of the models, assumptions, and correlations of the PLTEMP/ANL code can be found in Reference 1.

In the analysis for onset of nucleate boiling, the PLTEMP/ANL V4.0 code was used to analyze the limiting channel in the reactor. This analysis was performed once for each of eight reactor flow rates, spanning the flow range from 1000 to 2200 gpm. Since the model represented only the limiting channel, it was easy to hand-check the PLTEMP/ANL V4.0 results. This hand-check was performed for a reactor flow rate of 1580 gpm. At this flow rate the limiting channel has a flow rate of 0.2210 kg/s. For this flow rate PLTEMP/ANL V4.0 found that the channel power is 22.80 kW, which corresponds to a reactor power of 4.72 MW. The code results also indicate the axial locations where onset of nucleate boiling was first reached. Key values of the verification are provided in the new Section 4.6 in a table, which is repeated below as Table 4.19-1. Additional specifics of the analysis can be found in the new Section 4.6, including a listing of the PLTEMP/ANL V4.0 input used for this analysis.

Table 4.19 Verification of PLTEMP/ANL Onset of Nucleate Boiling Prediction at the Limiting Axial Location for a Core Flow Rate of 1580 gpm Hand Quantity Calculation PLTEMP Channel Dimensions Thickness, in 0.088 Width, in 2.62 Heated Length, in 23.25 Heat Width, in 2.395 Wetted Perimeter, m 0.1376 27

Quantity Hand Calculation PLTEMP Hydraulic Diameter, m 0.004325 Heat Transfer Area (2 Faces), m-' 0.07185 Channel Flow Rate, kg/s 0.2210 Channel Power, kW 22.80 Pressure, bar 1.7 Inlet Temperature, C 46.11 Saturation Temperature, C 115.15 Cp @. 55 C, kJ/kg-C 4.1828 At Onset of Nucleate Boiling Location Layer 13 13 Channel Power to Middle of Layer, kW 15.64 Channel Power to Exit of Layer, kW 16.56 Local Peak-to-Average Power 1.2894 Bulk Temperature at Middle of Layer w/o Hot Chan. 63.03 63.01 Fac., C Random Hot Channel Factor on AT bulk 1.24 Bulk Temperature with Hot Channel Factor, C 67.09 67.06 Bulk Temperature at Exit of Layer w/o Hot Chan. Fac., 68.32 C

Viscosity, Pa-s 4.3910E-4 Reynolds Number 14634 Thermal Conductivity, W/m-C 0.65817 Cp, kJ/kg-C 4.1867 Prandtl Number 2.793 Nusselt Number without Hot Channel Factors 74.55 Global Film Coefficient Hot Channel Factor 1.2 Film Coefficient with Hot Channel Factor, W/m 2 -C 9454 Heat Flux without Hot Channel Factors, MW/m2 0.4092 0.4091 Random Hot Channel Factor on AT film 1.28 Film Temperature Rise with Hot Channel Factor, C 55.77 Clad Surface Temperature with All Hot Channel 122.5 122.5 Factors, C Random Hot Channel Factor on Heat Flux 1.23)

Heat Flux with Hot Channel Factors, MW/m 2 0.4779 ATsaturation based on Bergles and Rohsenow, C 7.41 Surface Temperature For Onset of Nucleate Boiling, C 122.6 122.6 28

Reference:

Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 Pool Level-A low pool level is determined by the change in state of a float switch. The height of the switch is set so that there is no uncertainty that the switch will change state before the height of the water level above the core is less than 23.7 ft. However, in order to be conservative, an error of 0.5 inches is assumed.

Temperature Measurement:

Inlet, outlet, and bulk pool temperatures at RINSC are measured with an RTD sensor. The technical manual for the meters associated with these sensors indicates that they have an accuracy of + or - 0.5 C. See the reference entitled "Flow and Temperature Meter Specifications".

The normal operating temperature range of the primary coolant is between 90 F and 110 F. Consequently, an error of 0.5 C is:

5/9(90 F - 32) = 32 C 1.5 C/32 C=0.015= 1.5%

Flow Measurement:

Coolant flow rate is measured by looking at the pressure differential across an orifice plate. This differential is transmitted to a meter as a voltage signal. The errors associated with this measurement are:

+ or - 2% of the upper range for the orifice plate (See the reference entitled "Flow Measurement Uncertainty")

+ or - 0.25% accuracy for the transmitter (See the reference entitled "Flow Measurement Uncertainty")

29

+ or - 0.05% accuracy for the meter voltage input (See the reference entitled "Flow and Temperature Meter Specifications")

Power Level:

From the "Report on the Determination of the Hot Spot Factors for the Rhode Island Nuclear Science Center Research Reactor Using LEU Fuel", Eugene Spring, August 24, 1989 which is used for hot spot analysis, the error associated with power level is 10%.

421 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 Tonb) 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.

Fourth Response to RAL Dated April 13, 2010 Submitted September 8, 2010 Section 4.6 in the 2004 SAR will be superseded by a new Section 4.6. In the new analysis Table 4-9 is longer relevant. A more detailed explanation, if one is needed, follows.

Table 4.9 represents a double search. First a range of solutions for various core pressure drops are searched and interpolated to find the one that provides the correct total core flow rate. Then at the desired core pressure drop and flow rate, solutions for a range of assumed reactor power levels are searched and interpolated to find the one that provides the clad surface temperature corresponding to the onset of nucleate boiling. In the new analysis, on the other hand, for each specific reactor flow rate, the flow rate of each fueled channel is determined via a custom-developed hydraulics model. (This model and its governing equation are described in the memo dated 3 September 2010 from Earl E. Feldman to James E. Matos entitled "Steady State Thermal Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor" in considerable detail. Key intermediate results for a reactor flow rate of 1580 gpm are given in tables so that one can verify that the analysis was performed correctly.) The new analysis uses the PLTEMP/ANL V4.0 code 1 to find the power at which onset of nucleate boiling occurs. This code has an internal search method, which determines the power level at which onset of nucleate boiling is first achieved and provides the power level value as a code output. Thus, in the new analysis the double search is not used and Table 4-9, or another similar table, is not needed.

In the new analysis, the onset of nucleate boiling for a reactor flow rates of 1715, 1800, and 1,950 gpm are predicted to occur at a reactor powers of 5.1, 5.3, and 5.7 MW.

respectively. The increased allowed power is, in part, attributable to 1) a reduction in the hot channel factors due to more use of statistical, rather than multiplicative, methods of combining uncertainty factors and 2) improvements in the determination of the reactor power distribution.

30

Reference:

Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 4.6 in the 2004 SAR has been superseded by a new Section 4.6 and is attached as "4.6 Steady-State Thermal-Hydraulic Analysis" Reference BB. In the new analysis Table 4-9 is longer relevant. A more detailed explanation, if one is needed, follows.

Table 4.9 represents a double search. First a range of solutions for various core pressure drops are searched and interpolated to find the one that provides the correct total core flow rate. Then at the desired core pressure drop and flow rate, solutions for a range of assumed reactor power levels are searched and interpolated to find the one that provides the clad surface temperature corresponding to the onset of nucleate boiling. In the new analysis, on the other hand, for each specific reactor flow rate, the flow rate of each fueled channel is determined via a custom-developed hydraulics model. (This model and its governing equation are described in the new Section 4.6 in considerable detail.

Key intermediate results for a reactor flow rate of 1580 gpm are given in tables so that one can verify that the analysis was performed correctly.) The new Section 4.6 analysis uses .the PLTEMP/ANL V4.0 code' to find the power at which onset of nucleate boiling occurs. This code has an internal search method, which determines the power level at which onset of nucleate boiling is first achieved and provides the power level value as a code output. Thus, in the new analysis the double search is not used and Table .4-9, or another similar table, is not needed.

In the new analysis, the onset of nucleate boiling for a reactor flow rates of 1715, 1800, and 1,950 gpm are predicted to occur at a reactor powers of 5.1, 5.3, and 5.7 MW, respectively. The increased allowed power is, in part, attributable to 1) a reduction in the hot channel factors due to more use of statistical, rather than multiplicative, methods of combining uncertainty factors and 2) improvements in the determination of the reactor power distribution.

Reference:

Arne P. Olson and Kalimullah, "A Users Guide to the PLTEMP/ANL V4.0 Code,"

Global Threat Reduction Initiative (GTRI) - Conversion Program, Nuclear Engineering Division, Argonne National Laboratory, March 24, 2010.

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.

31

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Section 4.6 in the 2004 SAR will be superseded by a new Section 4.6. See the response to RAI 14.36 and 14.32.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 It is true that an inlet temperature of 42.30 C (1080 F) it is less conservative than the coolant temperatures allowed by the proposed TS. Section 4.6 in the 2004 SAR will be superseded by a new Section 4.6 developed for these RAI responses. In the new analysis described in the memo dated 3 September 2010 from Earl E. Feldman to James E. Matos entitled "Steady State Thermal Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor", the inlet temperature is assumed to be 115°F (46.110 C). As explained in the answer to Question 4.17, this is based on the Technical Specifications Safety Limit values for reactor power (2.4 MW), reactor flow (1580 gpm), outlet temperature (1250 F), and pool water depth (23.54 feet above the active core).

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 It is true that an inlet temperature of 42.30 C (1080 F) it is less conservative than the coolant temperatures allowed by the proposed TS. Section 4.6 in the 2004 SAR has been superseded by a new Section 4.6 developed for these RAI responses and is attached as 74.6 Steady-State Thermal-Hydraulic Analysis" Reference BB. In the new analysis the inlet temperature is assumed to be 115'F (46.11' C). As explained in the answer to Question 4.17, this is based on the Technical Specifications Safety Limit values for reactor power (2.4 MW), reactor flow (1580 gpm), outlet temperature (1250 F), and pool water depth (23.54 feet above the active core).

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)

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The rising power transient analysis of Section 4.6.4 of the SAR is misplaced and will be moved to the section on reactivity insertion accident analyses in Chapter 13.

The analysis in the SAR was redone using the PARET/ANL code for a slow insertion of reactivity from high power. Since the power level is 2.0 +/- 0.2 MW, including 10%

uncertainty in the power level measurement, an initial power of 1.8 MW was selected to 32

achieve the maximum rise in power after the trip on power is initiated at 2.3 MW. The description of the transient is given below.

The reactor is initially operating at 1.8 MW, 122 F coolant inlet temperature, and 1740 gpm. There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715E+5 Pa. Then a long, slow ramp reactivity insertion begins at a ramp rate of 0.02 % Ak/k / s, continuing for 100 s. Power rises slowly. The power trip at 2.3 MW is actuated at 6.774 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.313 MW at 6.874 s. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for fuel meat centerline, and clad surface are: 76.7 C and 75.9 C. The peak coolant temperature of 59.6 C is reached at 6.90 s. These peak fuel and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-1 313). The peak coolant temperature is well below the saturation temperature of 115.4 C.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the response to RAI 4.24.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 4.24. A new analysis has been completed for which this discrepancy no longer exits.

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.

33

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 4.24.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 4.6.5 of the 2004 RINSC Reactor SAR provides an analysis of the thermal behavior of the LEU core during steady-state operation in the natural-convection mode.

This analysis has been completely redone and is replaced by section 4.7 of Reference AA.

Reference AA refers to the completely redone analysis of the thermal behavior of the LEU core during steady-state operation in the forced-convection mode, which is provided in sections 4.6.1 through 4.6.12 of the attached Reference BB and is a replacement for sections 4.6.1, 4.6.2, 4.6.3, and 4.8 of the 2004 RINSC Reactor SAR.

In the Reference AA analysis of natural convection, the inlet (and pool) coolant temperature is 1300 F and the coolant height is 23 feet 6.5 inches (23.54 feet) above the active core. These are the Safety Limit values and are more restrictive than the Limiting Trip values of 1280 F and 23 feet 9.1 inches, respectively. (See the table at the end of the response to RAI 14.52.)

AA. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman and M.

Kalimullah to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Natural-Convective Flow in the Rhode Island Nuclear Science Center (RINSC)

Reactor," November 8, 2010.

BB. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Forced-Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor," September

.3,2010.

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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Table 4-17 is part of section 4.6.5 of the 2004 RINSC Reactor SAR, which provides an analysis of the thermal behavior of the LEU core during steady-state operation. in the natural-convection mode. This analysis has been completely redone and is replaced by section 4.7 of Reference AA.

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As stated in the response to RAI 4.28, Reference AA refers to the completely redone analysis of the thermal behavior of the LEU core in the forced-convection mode, which is provided in Reference BB.

The meaning of the negative value of Margin to Incipient Boiling at 209.1 kW in the 2004 RINSC SAR implies that the incipient boiling occurs before 209.1 kW is reached.

However, in the analysis of section 4.7 of Reference AA, incipient boiling, which is referred to as "onset of nucleate boiling", is predicted to occur at 369 kW with all uncertainties included. The Limiting Trip value of power is 125 kW and the Safety Limit power is 200 kW. (See the table at the end of the response to RAI 14.52.) Thus, the Reference AA analysis shows a large margin to incipient boiling.

AA. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman and M.

Kalimullah to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Natural-Convective Flow in the Rhode Island Nuclear Science Center (RINSC)

Reactor," November 8, 2010.

BB. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Forced-Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor," September 3,2010.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 Section 4.7 will be modified to say:

A modification to the standard 14 element core shown in Figure 4-1 was analyzed and approved. The objective was to increase the neutron flux in the thermal column experimental facility. Several options were considered, including shifting the entire fuel matrix toward the thermal column. However, the RERTR group at Argonne National Laboratory suggested that an increase in the thermal column neutron flux could be obtained by expanding the core into the 17 element core shown in Figure 4-

2. This core is achieved by replacing three graphite reflectors on the thermal column side of the core with fuel.

4.31 Section 4.8 - Clarify which figure is meant by "Figure 4" in the text.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The first Paragraph (lines 25-29) will be modified to say:

Two fresh core models were chosen for the thermal hydraulic calculations that would produce the highest possible fuel cladding temperatures under normal reactor 35

operation at 2 MW reactor power. Figure 4-1 show the standard 14 element core, while Figure 4-2 shows the expanded 17 element core. Both cores were analyzed.

The reference to Figure 4-5 in the third paragraph, line 29 will be changed to "Figure 4-3,.

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."

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Sections 4.6.1, 4.6.2, 4.6.3, and 4.8 of the 2004 RINSC Reactor SAR provide the analysis of the thermal behavior of the LEU core during steady-state operation in the forced-convection mode. This analysis has been completely redone in sections 4.6.1 through 4.6.12 of the attached Reference BB and is a replacement for sections 4.6.1, 4.6.2, 4.6.3, and 4.8 of the 2004 RINSC Reactor SAR All assumptions made in the analysis and justification for those assumptions are provided in detail in Reference BB, which includes the.hydraulic modeling needed to obtain the individual channel flow rates, the determination of the hot channel factors, a description of the PLTEMP/ANL V4.0 code, which was used in the analysis, and a listing of the code input used in the analysis. In Reference BB section 4.6.10, "Results of Steady-State Thermal Analysis," and section 4.6.11. "Discussion" describe the determination of the power at which departure from nucleate boiling (DNB) is predicted to occur and include a discussion of the correlations and the calculations that were used.

The new analysis does not use either the term "departure from nucleate boiling ratio" or the term "DNBR". Instead, the new analysis indicates the power in kW or MW at which departure from nucleate boiling is predicted to occur.

BB. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Forced-Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor," September 3,2010.

4.33 The responses to RAI 4.24 and 4.25 present a ramp reactivity transient analysis with an initial power level of 1.8 megawatts (MW) and an initial flow rate of 1740 gallons per minute (gpm). Explain why the initial conditions used in the analysis are the most conservative initial conditions allowed by the proposed TS. If there are more conservative initial conditions, provide an analysis and discussion of the ramp reactivity transient that shows the safety limit (SL) will not be exceeded. Include an explanation of why the assumptions used in the analysis are the most conservative assumptions for any operation allowed by the proposed TS.

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Response to December 17, 2012 RAI Dated March 15, 2013 The question has to do with the fact that the transient analysis that was performed used a coolant flow rate of 1740 gpm, and the original proposed LSSS was 1600 gpm.

Consequently, the analysis does not bound the condition in which the flow rate is at the LSSS. In response, a new analysis was run with the flow rate set at the safety limit of 1580 gpm:

Case 3A: Slow Insertion of 0.02 % Ak/k / Second Reactivity From 1.8 MW Power The reactor is initially operating at 1.8 MW, 125 'F coolant outlet temperature, and 1580 gpm (the true flow minimum). There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715 x 105 Pa. The coolant inlet temperature for which an outlet temperature of 125 'F is reached was iteratively determined to be 113.6 'F (45.925 'C). Starting from this initial condition, a long, slow ramp reactivity insertion begins at a ramp rate of 0.02 % Ak/k / s [TS 3.2.4],

continuing for 100 s. Power rises slowly. The power trip at 2.3 MW is actuated at 6.8206 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.3133 MW at 6.921 s. The reactor power drops rapidly to shutdown conditions. The reactor period remains longer than the period trip set point, so the reactor does not trip on period.

Peak temperatures for fuel meat centerline, and clad surface are: 79.7 'C and 78.9 'C.

The peak coolant temperature of 61.6 'C is reached at 6.90 s. These peak fuel and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1 537, Part I, Appendix 14.1 and NUREG-1313). The peak coolant temperature is well below the saturation temperature of 115.4 'C.

If this event were to occur at a smaller ramp rate, the consequences would be reduced.

Therefore this is the most-limiting event for these power and flow conditions.

For other initial power and coolant inlet temperatures, we need to find most-limiting conditions. Table XX gives results for six cases that all have T(outlet)=125.0 F and 1580 gpm in steady-state prior to the transient.

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Table XX. Consequences for a Slow Insertion of 0.02 % Ak/k / Second Reactivity From a Range of Initial Powers. The Coolant Steady-State Initial Conditions are: Outlet Temperature is Always 125 F, and the Flow Rate is 1580 gpm.

Initial 0.00001 0.1 1.0 1.8 2.0 2.2 Power, MW T(inlet), C 51.67 51.35 48.48 45.93 45.29 44.65 Fuel Max. 80.7 83.2 81.7 79.7 79.2 78.7 Temp,, C Clad 80.0 82.5 81.0 78.9 78.4 78.0 Surface Max.

Temp., C Coolant 65.1 66.4 64.0 61.6 61.0 60.4 Max.

Temp., C Coolant 65.2 66.4 64.0 61.6 61.0 60.4 Exit Max Temp., C IIIIII It is clear from the Table XX that the most-limiting condition regarding peak fuel and clad temperatures attained occurs near an initial power of 0.1 MW. Fuel and clad temperatures never exceed 84 C, for the entire range of powers. These peak fuel and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG- 13 13). The peak coolant temperature is well below the saturation temperature of 115.4 'C. Since the peak clad surface temperature is well below the saturation temperature, there can be no incipient boiling.

The analysis shows that there are cases in which power could peak as high as 2.495 MW, which is greater than the proposed safety limit of 2.4 MW. However, the relevant issue is not the peak power level, but instead is the peak fuel cladding temperature.

Therefore, the proposed solution is to:

1. Replace the safety limit for power level with a limit of 530 C on cladding temperature.

2. Maintain the LSSS's for T outlet, H, and Flow, but remove the SL's for each of them.

3. In addition to making these changes for the forced convection mode of operation, we should also do this for natural convection mode operation.

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As submitted on 29 September 2011, the proposed Technical Specifications have the following Safety Limits and Limiting Safety System Settings which incorporate the proposed solution to this RAI question:

2.0 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits Applicability:

This specification applies to fuel that is loaded in the core.

Objective:

The objective of this specification is to ensure that the integrity of the fuel cladding is not damaged due to overheating.

Specifications:

The true value of the reactor fuel cladding shall be less than or equal to 530 C.

Bases:

NUREG 1313 shows that the integrity of the fuel cladding will not be damaged due to overheating provided that the cladding temperature does not exceed 530 C.

2.2 Limiting Safety System Settings 2.2.1 Limiting Safety System Settings for Natural Convection Mode of Operation Applicability:

These specifications apply to the safety channels that monitor variables that directly impact fuel cladding temperature during natural convection mode operation of the reactor.

Objective:

The objective of these specifications is to ensure that the safety limit for the reactor cannot be exceeded during natural convection mode operation.

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Specifications:

2.2.1.1 The limiting safety system setting for reactor thermal power shall be 115 kW.

2.2.1.2 The limiting safety system setting for the height of coolant above the top of the fuel meat shall be 23 ft 9.6 inches.

2.2.1.3 The limiting safety system setting for the bulk pool temperature shall be 125 OF.

Bases:

This combination of specifications was set to prevent the cladding temperature from approaching the 530 oC value at which damage to the fuel cladding could occur, even under transient conditions.

The thermal-hydraulic analysis for steady state power operation under natural convection cooling conditions shows that the fuel cladding temperature will remain significantly below the threshold for cladding damage during steady state operation of the reactor if the following combination of limits are in place:

- The steady state power level is less than 200 kW,

- The coolant height above the fuel meat is at least 23 ft 6.5 in, and

- The bulk pool temperature is no greater than 130 OF.

The transient analysis for natural convection cooling was performted for the most conservativecase in which all of the safety channels are at their respective limiting trip value when the transient is terminated. The analysis shows that the peak fuel cladding temperature will be approximately. 67.5 °C during a transient in which the following combination of limits are in place:

- The initial power level is no greater than 100 kW,

- The coolant height above the fuel meat is at least 23 ft 9.1 in,

- The bulk pool temperature is no greater than 128 OF, and

- The transient is terminated by an over power trip at 125 kW.

In both, the steady state and transient analyses, the predicted peak cladding temperature is significantly below the damage threshold temperature of 530 °C. The safety margins are:

40

- Margin for the transient bounded by the limiting conditions is 530 °C - 67.5 °C = 462.5 oc.

Measurement uncertainty was based on the nominal operating values of 100 kW and 108 OF for the power and pool temperature respectively, and has been determined to be:

- Power Level +/- 10 kW

- Coolant Height 0.5 in

- Temperature 3 OF Consequently, the bases for these specifications are:

Specification 2.2.1.1 sets the limiting safety system setting for reactor thermal power to be 115 kW. The analyses show that cladding damage will not occur under any condition if initial power is no greater than 200 kW.

Taking into consideration a 10 kW measurement error, if the LSSS is 115 kW, then the Limiting Trip Value could be as high as 125 kW, which still leaves a safety margin of 75 kW.

Specification 2.2.1.2 sets the limiting safety system setting for the height of coolant above the top of the fuel meat to be 23 ft 9.6 in. The analyses show that cladding damage will not occur under any condition if the height is no less than 23 ft 6.5 in. Taking. into consideration a 0.5 in measurement error, if the LSSS is 23 ft 9.6 in, then the Limiting Trip Value could be as low as 23 ft 9.1 in, which still leaves a safety margin of 2.6 in.

Specification 2.2.1.3 sets the limiting safety system setting for the bulk pool temperature to be 125 OF. The analyses show that cladding damage will not occur under any condition if the pool temperature is no greater than 130 OF.

Taking into consideration a 3 OF in measurement error, if the LSSS is 125 OF, then the Limiting Trip Value could be as high as 128 OF, which still leaves a safety margin of 2 OF.

2.2.2 Limiting Safety System Seftings for Forced Convection Mode of Operation 41

Applicability:

These specifications apply to the safety channels that monitor variables that directly impact fuel cladding temperature during forced convection mode operation of the reactor.

Objective:

The objective of these specifications is to ensure that the safety limit for the reactor cannot be exceeded during forced convection mode operation.

Specifications:

2.2.2.1 The limiting safety system setting for reactor thermal power shall be 2.1 MW.

2.2.2.2 The limiting safety system setting for the height of coolant above the top of the fuel meat shall be 23 ft 9.6 inches.

2.2.2.3 The limiting safety system setting for the primary coolant outlet temperature shall be 120 OF.

2.2.2.4 The limiting safety system setting for the primary coolant flow rate shall be 1800 gpm.

Bases:

This combination of specifications was set to prevent the cladding temperature from approaching the 530 °C value at which damage to the fuel cladding could occur, even under transient conditions.

The thermal-hydraulic analysis for steady state power operation under forced convection cooling conditions shows that the fuel cladding temperature will remain significantly below the threshold for cladding damage during operation of the reactor if the following combination of limits are in place:

- The steady state power level is less than 2.4 MW,

- The coolant height above the fuel meat is at least 23 ft 6.5 in,

- The primary coolant outlet temperature is no greater than 125 OF, and

- The coolant flow rate through the core is at least 1580 gpm.

The transient analysis for forced convection cooling was performed for the most conservative case in which all of the safety channels are at their respective limiting trip value when the 42

transient is terminated. The analysis shows that the peak fuel cladding temperature will be no greater than 85 °C during a transient in which the following combination of limits are in place:

- The initial power level is no greater than 2.2 MW,

- The coolant height above the fuel meat is at least 23 ft 9.1in,

- The primary coolant inlet temperature is no greater than 123 OF,

- The coolant flow rate through the core is a least 1740 gpm, and

- The transient is terminated by an over power trip at 2.3 MW.

In both, the steady state and transient analyses, the predicted peak cladding temperature is significantly below the damage threshold temperature of 530 °C. The safety margins are:

- Margin for the transient bounded by the limiting conditions is 530 oC - 85 °C = 445 C.

Measurement uncertainty was based on the nominal operating values of 2 MW, 1950 gpm, and 90 OF to 115 OF for the power, flow and outlet temperature respectively, and has been determined to be:

- Power Level +/- 0.2 MW

- Coolant Height 0.5 in

- Temperature 3 OF

- Flow Rate +/- 60 gpm Consequently, the bases for these specifications are:

Specification 2.2.2.1 sets the limiting safety system setting for reactor thermal power to be 2.1 MW. The analyses show that cladding damage will not occur under any condition if initial power is no greater than 2.4 MW.

Taking into consideration a 0.2 MW measurement enror, if the LSSS is 2.1 MW, then the Limiting Trip Value could be as high as 2.3 MW, which still leaves a safety margin of 0.1 MW.

Specification 2.2.2.2 sets the limiting safety system setting for the height of coolant above the top of the fuel meat to 43

be 23 ft 9.6 inches. The analyses show that cladding damage will not occur under any condition if the height is no less than 23 ft 6.5 in. Taking into consideration a 0.5 in measurement error, if the LSSS is 23 ft 9.6 in, then the Limiting Trip Value could be as low as 23 ft 9.1 in, which still leaves a safety margin of 2.6 in.

Specification 2.2.2.3 sets the limiting safety system setting for the primary coolant outlet temperature to be 120 OF.

The analyses show that cladding damage will not occur under any condition if the primary coolant outlet temperature is no greater than 125 OF. Taking into consideration a 3 OF in measurement error, if the LSSS is 120 OF, then the Limiting Trip Value could be as high as 123 OF, which still leaves a safety margin of 2 OF.

Specification 2.2.2.4 sets the limiting safety system setting for the primary coolant flow rate to be 1800 gpm. The analyses show that cladding damage will not occur under any condition if the primary coolant flow rate is at least 1580 gpm. Taking into consideration a 60 gpm in measurement error, if the LSSS is 1800 gpm, then the Limiting Trip Value could be as lw as 1740 gpm, which still leaves a safety margin of 160 gpm.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Make-up water into the reactor primary system is metered. As part of securing the facility at the end of each weekday, the make-up water meter reading is recorded and checked to verify that abnormal amounts of water have not been added. The Make-Up water System solenoid valve opens and closes to control the fill. A float switch is used to sense pool level. When the float switch senses that the pool level has dropped one inch, the solenoid valve is opened to allow flow into the pool. If the float switch senses that the pool level has dropped two inches, a reactor scram is initiated, and an alarm is sent to a contracted alarm company that monitors the facility security, fire, and other parameter alarms 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, seven days per week. In addition to the normal float switch position that closes the make-up supply valve, a second overflow switch position exists as a secondary overflow protection.

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6.1 Describe the operating parameters and design features of the confinement, including:

free volume, number and type of penetrations, etc.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The volume of the confinement building is approximately 203,695 cubic feet. The volume of the pool structure and water is approximately 21,740 cubic feet, leaving 181,955 cubic feet of open space. The control room takes up about 3,612 cubic feet of this space, with the air changed regularly with the confinement air for temperature and humidity control.

During normal operation air enters the confinement building through the Confinement Air Intake opening. This air will disperse throughout the confinement building until it is collected at the Normal Ventilation Intake, where air is drawn in by way of the Confinement Exhaust Blower. A sample of this air is collected through a copper tube and delivered to a Continuous Air Monitor in the basement, known as the Stack Monitor. This monitor samples the air for gaseous and particulate contamination then returns the sample to the exhaust line. The air is then sent up the stack, where it is diluted with air from outside the EPZ pulled in by the Dilution Blower, then released to the atmosphere at 115 feet above ground. Air from the various irradiation and experiment facilities are collected and combined with the confinement exhaust before the Confinement Exhaust Blower.

At five locations throughout the facility there are red evacuation buttons. Three are within the confinement building: one to the right of the portal entrance, one to the right of the roll-up door, and one in the control room. In addition to an audible alarm heard throughout the building, pressing any of these buttons will cause the confinement air handling systems to go into emergency mode. When this is initiated power is cut to the Confinement Exhaust Blower, the Off Gas Blower, and the Rabbit Blower. As the flow rate drops pneumatic switches trigger the pneumatic Exhaust and Intake Dampers to close. As this occurs, the Emergency Blower is energized, pulling air through the Emergency Ventilation Intake. This air is passed through a series of filters, including a charcoal filter, removing the majority of Iodine that may be present. This air is then directed to the Emergency Ventilation Exhaust where the air is diluted with air from outside the EPZ that is pulled in by the Dilution Blower. This air then travels up the stack and is released 115 feet above ground level. Because the normal sampling location for stack exhaust activity no longer represents the air being released to the environment, sampling is instead taken from approximately half way up the stack. This is achieved by way of a motorized three-way valve located at the Stack Monitor that is activated during emergency mode.

All other major penetrations through the confinement remain closed during operation, with the exception of the Lab Wing Portal, which utilizes a secondary door to maintain confinement integrity. During operation, the confinement building maintains at least a one-half inch of negative pressure in relation to the outside environment. Any other penetrations not mentioned, such as for coolant piping or wiring, are sealed or provide 45

negligible effects on potential radioactive release or unauthorized access. A summary of penetrations are provided in the following table:

Summary of Confinement Penetrations 7.1 Provide a listing of the interlocks of the reactor protection system.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Table 3.1will be revised to be:

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Protection Cooling Channels Function Set Point Mode Required Over Power Both 2 Scram by Power Level Less than or 105% of Equal.to Licensed Power Low Pool Level Both 1 Scram by Pool Level Less than or 23 ft 9.6 Drop Equal to in Primary Coolant Inlet Forced 1 Alarm by Inlet Temp Less than or 111 F Temperature Equal to Primary Coolant Outlet Forced 1 Alarm by Outlet Temp Less than or 117 F Temperature Equal to Forced I Scram by Outlet Temp Less than or 120 F Equal to Pool Temperature Both I Scram by Pool Temp Less than or 125 F Equal to Primary Coolant Flow Primary Flow Less than or 1800 Rate Both 1 Scram by Rate Equal to gpm Rate of Change of Less than or 4 Power Both 1 Scram by Period Equal to seconds Seismic Disturbance Both 1 Scranm if Seismic Disturbance Detected Bridge Low Power Position Forced 1 Scram if Bridge Not Seated at HP End Bridge Movement Both 1 Scram if Bridge Movement Detected Coolant Gates Open Forced 1 Scram if Inlet Gate Open Forced 1 Scram if Outlet Gate Open Detector HV Less than or Detector HV Failure Both 1 Scram if Decrease Equal to 50 V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V No Flow Thermal Column Forced 1 Scram by No Flow Detected Manual Scram Both 1 Scram by Button Depressed Both 1 Scram by Button Depressed No Automatic Servo Control Interlock Both 1 Servo if Regulating Blade not Full Out No Automatic 30 Both 1 Servo if Period Less than seconds Shim Safety No SS Withdrawal Both 1 Withdrawal if Count Rate Less than 3 cps No SS Both 1 Withdrawal if Test/Select SW not Off Rod Control Loss of [Less than or 10 Communication Both 1 Scram if Communication Equal to seconds 47

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Neutron flux is monitored by three independent instruments:

Neutron Flux Monitor Wide Range Monitor #1 Wide Range Monitor #2 The Neutron Flux Monitor consists of a fission chamber located in grid position B-9.

This instrument contains the Start-Up Channel, a Wide Range Channel, and a Linear Power Channel. The Start-Up and Wide Range Channels each have a Period Channel associated with them.

Each of the Wide Range Monitor Instruments has a Linear Power Channel. Each monitor uses a compensated ion chamber to detect neutrons. The detector for WRM #1 is located in grid position A-l, and the detector for WRM #2 is located in grid position G-9.

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)

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The bridge misalignment scram consists of a position switch mounted at the high power end of the reactor pool. When the bridge is not fully seated at the high power end of the pool, the switch is in a state that initiates the "Bridge Lo Pwr Pos" scram. This scram is only active when the reactor is being operated in the forced convection cooling mode.

For natural convection cooled operation at 100 kW or less, this scram is bypassed.

The bridge movement scram consists of a limit switch mounted on the gear that turns the wheels of the bridge. The switch is in the non-scram state when the switch actuator is positioned so that it is resting on the top of a gear tooth. If the bridge moves and the switch actuator rests in a valley between two gear teeth, the switch changes state to initiate a scram. The gear teeth are spaced V., inches apart, so movement of less than 1/4 inches will cause a scram.

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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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Available Radiation Instrumentation:

Instrument Position Detector Readout Alarm Setpoint

_ Type Stack Gaseous Stack Gamma Local and Local and 2.5 X Normal Monitor Sensitive Control Room Control Room Stack Particulate Stack Gamma Local and Local and 2 X Normal Monitor Sensitive Control Room Control Room Main Floor Reactor Main Gamma Local Only Local Only 2 X Normal Particulate Monitor Floor . Sensitive Reactor Bridge Area Reactor Bridge Gamma Local and Local and 2 X Normal Monitor Area Sensitive Control Room Control Room Fuel Safe Area Fuel Safe Area Gamma Local and Local and Higher of 2 X Monitor Sensitive Control Room Control Room Normal or 5 mR-hr Thermal Column Thermal Column Neutron Local and Local and Higher of 2 X Area Monitor Area Sensitive Control Room Control Room Normal or 2 mR/hr Heat Exchanger Heat Exchanger Gamma Local and Local and 2 X Normal Area Monitor Area Sensitive Control Room Control Room Primary Clean-Up Primary Clean- Gamma Local and Local and 2 X Normal Demineralizer Area Up Demineralizer Sensitive Control Room Control Room Monitor Area Required Radiation Instrumentation:

1. Area Radiation Monitors:

A. A minimum of one radiation monitor that is capable of warning personnel of high radiation levels shall be at the experimental level. The Thermal Column Area Monitor, or equivalent may serve in this capacity.

B. A minimum of one radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool. The Reactor Bridge Area Monitor, or equivalent may serve in this capacity.

C. Alarm set points may be adjusted higher with the approval of the Director or Assistant Director.

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D. If either of these detectors fail during operation, the staff shall have one hour to either repair the detector, or find an acceptable replacement without having to shut the reactor down.

2. Air Monitors:

A. A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous effluent shall be operating. The Stack Gaseous Monitor, or equivalent may serve in this capacity.

B. A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement particulate effluent shall be operating. The Stack Particulate Monitor may serve in this capacity.

C. Alarm set points may be adjusted higher with the approval of the Director or Assistant Director.

D. If either of these detectors fail during operation, the staff shall have six hours to either repair the detector, or find an acceptable replacement without having to shut the reactor down.

7.5 The response to RAI 7.3 omitted the alignment tolerance for the bridge misalignment scram (now called the "bridge low power position" scram, as specified in proposed TS 3.2.1.3). Provide the alignment tolerance. If misalignment could allow core bypass flow, provide a discussion of the impact on the thermal hydraulic analyses for normal and accident conditions. Explain how the primary coolant system instrumentation would detect core bypass flow.

Response to December 17, 2012 RAI Dated March 15, 2013 The response to RAI 7.3 indicated that the bridge misalignment scram is tripped by a limit switch. A gear system is used to move the bridge. When the bridge is fully seated in the high power section of the pool, the switch actuator rests on top of one of the gear teeth. If the bridge moves, the switch actuator drops into a valley between gear teeth.

This action trips the scram. The gear teeth are spaced V2 inches apart, so movement of less than 1/4/inches will cause a scram. Therefore the alignment tolerance is 1/4 inches.

7.6 Based on discussions and firsthand observations during NRC staff visits to the RINSC, it is clear that there have been modifications to the instrumentation and control (I&C) systems that are not fully described in the SAR. Revise the SAR to include descriptions and analyses of the current I&C systems. The revisions should be consistent with the guidance in NUREG-1537 and not restricted to Chapter 7 of the SAR if the modifications affect descriptions or analyses presented elsewhere in the SAR.

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Response to December 17, 2012 RAI Dated March 15, 2013 Changes have been made to the control rod drive system. Also, additional display systems have been added to the instrumentation racks. Consequently, the Safety Analysis Report (SAR) descriptions of the following instrumentation and control components, as submitted in 2004, need to be updated:

7.2 Operator Controls 7.2.1 Control Console The control console serves as a central point for location of operating controls and instrumentation. The operator is provided with a vantage point from which to conveniently observe reactor performance and the pool area. The operator can adjust operations to varying requirements when needed for tests, experiments and power level requirements.

The control console consists of a desk-type cabinet 71 inches wide, 44 inches high, and 31 inches deep. Located on the right of the console control panel is the reactor control computer. This computer operates the software that is the primary means of manipulating the reactor controls.

The central portion of the console is occupied by the annunciator panel. In addition to the alarm and scram indicator lights are the annunciator acknowledge switch, the annunciator reset switch, and the scram reset switch. Mounted below .the annunciator panel is the manual scram switch.

Below the scram switch in the center portion of the control panel are the original reactor controls, including the blade select switch, control blade manual control switch, manual rundown switch, regulating rod manual control switch. Also located in this section is the auto/manual select switch. This switch designates which system selects the control blade to be manipulated.

The start-up count rate and period, logarithmic power level and period, and the linear power level indicators are located on the left side of the control panel. Below the linear indicators are mounted the linear range switches and the reactor on indicator switch.

7.2.2 Instrument Rack The instrument rack is a four bay relay-rack cabinet located to the right of the control console. Three of the bays are a single unit, with the fourth bay being adjacent but independent. The instrument rack holds the various instrumentation and computers used in reactor operation.

The first rack contains a computer display for core configuration and position, as well as other elements and racks that are stored in the pool. Beneath the display is the master 51

control switch and power level select switch. Five fuses distribute power to and from the switch to various reactor control circuits. The magnet power supplies are beneath the master switch. The control power circuit breaker is at the bottom of the rack.

The second rack is dedicated to reactor power level and contains the nuclear instrumentation and power level display computer. The reactor trip logic is located above the power level display computer at the top of the rack.

The third rack contains the area radiation monitoring base units and display.

The fourth rack contains cooling and air handling controls and displays. The controls include the primary and secondary pump controls for both systems I and 2, and all the operator controllable confinement blowers. The displays include flow rates, temperatures, conductivity, and pH. At the bottom rack is a separate control power circuit breaker for the rack and the pneumatic tube control system breaker.

7.2.3 Power Distribution System Power for operation of the reactor is supplied from the main breaker panel, located at ground level on the South confinement wall, through the control power circuit breaker on the bottom left of the instrument rack. This breaker feeds a single power strip in the main instrument rack, as well as the master switch. The power strip feeds three uninterruptable power supplies (UPS), one in each bay of the main instrument rack. The UPS power the various nuclear instrumentation, computers, radiation monitors, and other equipment located in the instrument racks. A second control power breaker is located in the fourth rack. This breaker feeds a power strip, UPS, and instrumentation in similar fashion to the control power circuit breaker in the main instrument rack.

7.2.4 Master Switch The master switch for reactor control is located in the first bay of the instrument rack and is powered directly from the control power circuit breaker.

The master switch is locked in the "Off" position to prevent reactor start-up. When unlocked and turned to the "On" position the master switch energizes the 24 Volt DC power supply, control blade drive motors and control circuits, power level interlocks, magnet power supplies, trip actuation circuits, "Reactor On" light, and the control panel annunciator. When the switch is first turned to the "On" position there is a 10 second delay interlock that prevents blade withdrawal and an alarm buzzer to notify personnel.

The "Test" position is historical setting that allowed for the control blade drive motors to be energized and manipulated without energizing the magnets. The regulating rod is the only reactor control that can be manipulated in this position.

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Since the switch is powered directly from the breaker, if facility AC power supply is lost then the switch is no longer powered. This will deenergize trip actuator amplifiers, and thus the control blade magnets, causing a reactor scram.

7.2.5 Control Blade Magnet Power Supply Each of the four reactor control blades is held out by an electromagnet when the reactor is in operation. These magnets are positioned above the reactor pool normal water level, directly above the control blade armature.

The magnet power supplies are located directly beneath the master switch and power level select switch. They supply 24 volt DC power to the magnets at less than 1 amp each. The power supplies receive a 12 volt DC signal from the nuclear instrumentation, by means of the logic element. The power supplies also receive a 120 volt AC signal from the reactor control console, through the annunciator. A loss in the DC or AC signal signifies an electronic or mechanical scram, respectively. After the initiation of a scram signal the magnet power supplies deenergize the magnets, allowing the control blades to be dropped in the core. Power can be restored to the magnets by restoring the DC and AC signals and pressing the scram reset button on the annunciator.

The control blade magnets and control blade magnet power supplies are independent from any of the operator controls or control blade drive systems. The initiation of a scram cannot be overridden or ignored by an operator. The independent system means that no other system will interfere with the initiation of a scram.

7.2.6 Control Blade Drive System The reactor controls are manipulated primarily by the reactor control computer. This computer is the interface and display for the control system, through which the operator can access all features of the reactor controls. The system operates similarly to the original reactor controls.

There is a select button for each control blade which activates a pair of relays operating in a binary system to designate which blade is manipulated. This mechanism replaces the original mechanical selector switch in preventing multiple blades being withdrawn simultaneously. In the event of a malfunction where both relays are energized, the system would allow only Blade #4 to be selected. Likewise, if both relays were deenergized, Blade #1 would be selected. The select button for the selected blade changes state to indicate that is the blade being manipulated.

The selected blade can be manipulated with manual withdraw and insert buttons, mimicking the action of the original blade movement toggle switch. Additionally, the selected blade can be manipulated using the auto blade position feature. This allows the operator to enter a specific withdrawal position between 0 and 26 inches. The start button begins moving the blade toward the specified position; the stop button stops the blade where it is. The blade will automatically stop moving once it reaches the specified 53

position. Blade movements are still subject to the various alarms, scrams, and reactivity insertion rates during the use of the auto blade position feature. All movements use separate blade up and blade down relays. Should a malfunction occur and both relays become energized the blade down function would override and cause the selected control blade to insert.

The manual rundown button functions similarly to the original manual rundown switch by inserting all four blades simultaneously. Unlike the original switch, the manual rundown button will reset itself once all blades are fully inserted.

Each control blade has a separate controller located in an electrical cabinet on the North side of the reactor bridge. Each controller powers the associated stepper motor that is connected to the original control blade drive gear through a gear reducer. A digital encoder located underneath the gear reducer measures the angular movement of the control blade drive. The angular movement of the control blade drive gear correlates to the vertical movement of the control blade and is displayed for each blade on the reactor control computer.

Each control blade assembly has a limit switch located at the top and bottom of the travel range to indicate full-in and full-out position. The activation of these switches overrides the control blade movement commands and prevents control blades from moving beyond the desired travel range. The reactor control computer displays when either one of these positions have been reached.

The original reactor controls remain as a redundant system in parallel with the reactor control computer. A four position selector switch allows the operator to choose which control blade is manipulated, allowing for only one control blade to be withdrawn at a time. The selected control blade is inserted or withdrawn using a momentary toggle switch that defaults to the off position. The manual rundown switch is a maintained toggle switch that will insert all control blades to their full-in position. A two position selector switch allows the operator to change between use of the reactor control computer or the original control switches. The position of this switch determines which system will control which control blade is currently selected for manipulation. The manipulation controls for both systems are always active, with the original switch controls taking priority over the reactor control computer. However, this switch ensures only one control blade may be manipulated at a time. The reactor control computer is the sole control blade position display.

7.2.7 Regulating Rod Drive System The regulating rod is also controlled by the reactor control computer. The controls allow for manual insertion and withdrawal, similar to the original reactor controls. The reactor control computer also allows for the regulating rod to move to the full-in or full-out position through a single command. A stop button allows the operator to stop the regulating rod at whatever position it is currently in.

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The commands for manipulation of the regulating rod are sent to a pair of relays in the instrument rack, one relay for each direction of movement. The regulating rod controller is housed on the North side of the reactor bridge with the controllers for the control blades. It receives the signal from the instrument rack for manipulation of the regulating rod. Signals for regulating rod position and position limits are sent back to instrument rack and reactor control computer.

The reactor control computer also allows for the automatic manipulation of the regulating rod. Under certain conditions the operator may place the regulating rod in automatic mode. In this operational mode the operator will input the desired reactor power level in terms of percent of full power into the reactor control computer. The reactor control computer compares the desired power level to the current power level from Wide Range

1. 1. The computer may then output an insert or withdraw signal to the regulating rod to adjust the current power level until it agrees with the desired power level.

The original regulating rod control remains as a redundant system in parallel to the reactor control computer. The regulating rod can be manipulated using a momentary toggle switch to insert or withdraw the regulating rod. The reactor control computer is the sole regulating rod display. The reactor control computer is the only means of engaging the regulating rod automatic mode or adjusting the desired power level setting.

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 104 . Provide sufficient details regarding both the normal and emergency ventilation system flows to confirm the appropriateness of the dilution factor.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Chapter 13 will be re-written based on the basis document entitled "Fuel Damage Radiological Assessment". In this analysis, no credit is taken for dilution air.

Consequently, there is no longer a need to justify a dilution factor of 4 X 104.

9.2 Section 9.2 uses inconsistent units when discussing criticality protection for fuel in storage. Confirm that Keff is less than 0.8 for fuel in storage.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 P.8-5 Line 27, Line 42, and Line 43 should not refer to Ak/k. We are saying that keff is less than 0.8, which is a pure number, not a reactivity.

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.

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Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 References to this saw will be removed from the SAR. This saw is not related to the safety margin associated with the operation of the reactor.

9.4 Provide a legible copy of Figure 9-7.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 Figure 9-7 was a schematic of the Bay Campus Water System. The original drawing is somewhat illegible. Consequently this figure will be removed from the SAR. We will reference the original drawing as needed.

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.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 Most of the references are sufficient as listed. The SAAR will not refer to specific revisions of the RINSC Emergency Plan, RINSC Operating Procedures, and RINSC Security Plan because these documents are revised and updated on a regular basis. The reference entitled "TRTR-5 Fabrication Requirements" will be removed. The reference to the BMI Cask Letter (Certificate) of Compliance will be removed because that shipping cask is no longer in service. References entitled "RINSC Quality Assurance Program", and "NRC Approval Letter" have been removed because they are not referenced in the text. The-new list of references will include:

9-1 RINSC Emergency Plan 9-2 RINSC Operating Procedures 9-3 Removed - Need to remove this reference from SAR Page 9-5 Line 38 9-4 RINSC Security Plan 9-5 Removed - Need to remove this reference from SAR Page 9-7 Line 3 9-6 Removed - Not used 9-7 Removed - Not used 9-8 RINSC Safety Manual 9-9 IAEA-TEC-643, April 1992, Appendix N-3.1 "Nuclear Criticality Assessment of LEU & HEU Fuel Storage", Argonne National Laboratory The references will be re-numbered so that they appear in a numerical order that is consistent with the order in which they are referenced in the text.

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.

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Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 13.2.3, Loss of Coolant Accident (LOCA), has been completely replaced with the analyses provided in the re-written section of RINSC SAR Chapter 13.2.3 "Loss of Coolant Accident".

If one of the beam ports is severed, all six beam ports are flooded because of the common interconnected drain lines. The revised LOCA analysis assumes that water can flow from each of the six beam ports onto the reactor floor. Administrative controls are needed to guarantee that the flow resistance in each beam port is equal to or more restrictive than a round half-inch diameter, sharp-edged orifice at the exterior (beyond the reactor shielding) end of the beam port. This could be accomplished by having a cover on the exterior of each beam port that seals the beam port, except for an optional hole with a diameter of up to one half inch.

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.

Tenth Response to RAI Dated April 13, 2010 Submitted July 15, 2011, The LOCA Analysis presented in section 13.2.3 of the SAR has been revised and is entitled "Section 13.2.3 Loss of Coolant Accident (LOCA) of the Safety Analysis Report of the Rhode Island Nuclear Science Center Reactor, Submitted May 3, 2004".

In this analysis, the LOCA model is one in which an eight inch beam port extension is sheared off, and water drains through six sharp edged round half inch diameter round holes in the pool wall. It was conservatively assumed that since the drain lines of all of the beam ports are tied together by a common drain line, it would be possible for the drain line to back up and allow the un-damaged beam ports to fill with water, in which case each beam port would act as a drain path to confinement. Administratively, the area of each beam port that is open to confinement has been limited to a one half inch diameter hole. Consequently, the drain model considered a system which has six, half inch diameter holes, which corresponds to one for each beam port. The through port was not considered.

In response to this RAI question, the drain model for this analysis has been refined to include the through port experimental facility, and to provide a more realistic and less conservative representation of the experimental facility, drain piping system. This analysis is entitled "LOCA Analysis Addendum. In this analysis, it is shown that each experimental facility drain, including the through port has a one half inch diameter orifice plate welded into the drain line. In each case, this empties into a one inch 57

diameter drain line. All of these drain lines empty into a common two inch drain line that is at an elevation below all of the experimental facilities. The common drain line empties into a five inch line, which is reduced back into a two inch line that opens to atmosphere.

Experiment Drain System Since the drain line diameter gets progressively larger, and it opens to atmosphere, it is not possible for the drain line to get backed up. This fact, coupled with the fact that the elevation of the common drain line is below the lowest elevation of any of the experimental facilities means that for the Design Basis Accident in which one beam port is sheared off. there are only two drain paths for the coolant water:

1. Damaged Port Drain

2. Area of Port Open to Confinement Shearing off a through port is not considered to be credible because there is virtually no access to the through port from the top of the pool. As shown in the photograph below, this port runs across the back wall of the pool, underneath the thermal column extension:

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Consequently, the drain model in the Addendum has been changed from a tank in which six, one half inch diameter holes are at the elevation of the bottom of an eight inch beam port, to a tank in which there are two, one half inch diameter drain holes at the centerline elevation of the common two inch drain line. Under this scenario, the amount of time that it takes for the reactor pool level to drop from the pool level scram set point, to the top of the grid box is 19.34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />.

The analysis done in "Section 13.2.3 Loss of Coolant Accident (LOCA) of the Safety Analysis Report of the Rhode Island Nuclear Science Center Reactor, Submitted May 3.

2004" shows that if the pool level does not drop below the elevation of the bottom of the eight inch beam ports before the decay power fraction is 0.827% after infinite reactor operation, then the fuel cladding will not be damaged. For the drain model in the addendum, it is impossible for the pool level to drop below this point due to shearing an experimental port because the through port is inaccessible, and the eight inch ports have the lowest drain level.

The revised analysis from 2004 includes a reference that provides data for the amount of time that it takes for decay power to reach various power fractions under various operating histories (Stillman). The analysis in the addendum indicates that the amount of time that it takes for the power fraction to reach 0.827% after infinite reactor operation is 16232 seconds (4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />).

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All of this means that it takes 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the decay power to reach the point at which the fuel cladding will not become damaged, given that the pool level is not below the elevation of the bottom of the eight inch beam ports. If a Design Basis Accident occurs, it will take 19.34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> for the pool level to drain to the top of the grid box, which is well above the elevation of the bottom of the eight inch beam ports. As a result, the power fraction will decay to a harmless level by the time that the pool level reaches the top of the core box. As part of the addendum analysis, the maximum allowable drain area between an experimental port and confinement was determined to be 1.48 in2 .

The through port is at a lower elevation than the beam ports. As stated earlier, a catastrophic failure of the through port is not considered to be credible. This is why the Maximum Credible Accident for the facility has always been the catastrophic failure of a beam port rather than the through port. Therefore, if the through port were to develop a leak, the pool drain time would be no greater than the time calculated in the LOCA Addendum analysis, provided that the area open between confinement and the through port is no greater than 1.48 in2 .

Section 2.3.3 of the Safeguards Report for the Rhode Island Open Pool Reactor (4 April 1962) indicates that the original criteria for using the through port was that both ends of the port would have gate valves that could be closed in the event of a leak. Like the beam ports, the through port also has a one inch diameter drain line that can be used to isolate the port from the experimental drain system. Consequently, provided that the gate valves are in place, it is possible to close off all of the potential pool water drain pathways associated with this experimental facility.

Since the drain time estimation in the LOCA Addendum is on the order of hours for the pool level to reach the top of the core box, there is sufficient time to perform mitigating actions. In the event that it were deemed to be worthwhile to reopen the through port after it has been isolated due to a leak, it would be possible to move the core to the opposite end of the pool, and isolate that section of the pool from the end with the through port so that the high power section of the pool with the ports could be drained independently form the low power section of the pool where the core is positioned. This action makes the twelve hour delay time prior to opening the port, irrelevant.

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) 60

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The temperature limit of 107 'C is cited in the original reactor operating manual

[Operation and Maintenance Manual, One-Megawatt Open Pool Reactor for Rhode Island Atomic Energy Commission, Providence, R.I., General Electric Document GEI-77793, October 1962] but no basis is given. Since the ignition temperature of graphite is well above this temperature (ranging from approximately 400 'C upwards, depending on the specific type and form of the graphite) it is reasonable to assume the limit was placed on the graphite temperature to preclude any unexpected releases of the stored lattice energy (i.e., Wigner energy) induced by neutron irradiation.

Numerous references address the release of stored energy in graphite including:

1. Radiation Defects in Graphite, R. H. Telling, University of Sussex, December 18, 2003;

2. Evaluation of Graphite Safety Issues for the British Production Piles at Windscale: Graphite Sampling in Preparation for the Dismantling of Pile 1 and the Further Safe Storage of Pile 2, B. J. Marsden et al., AEA Technology plc;

3. Nuclear Engineering Handbook, Harold Etherington, ed., McGraw-Hill Book Company, 1958.

As neutron-irradiated graphite is annealed (heated above irradiation temperatures) little or no stored energy is released below approximately 100-125 'C. A significant release peak occurs at approximately 200 °C, so limiting the RINSC thermal column graphite to temperatures below 107 'C provides a margin of nearly a factor of two to this energy release temperature. It is not credible that the entire thermal column contains lattice defects. The threshold energy for the lowest form of induced defect is approximately 1 eV (see Telling paper). After passing through the first several inches of graphite, the neutron flux in the thermal column is, as the name indicates, thermalized to an energy spectrum with a peak near 0.025 eV, which is below the threshold for inducing lattice defects. The figure below shows the energy release rate as a function of temperature (from Ref. 2 above).

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2.000

-' .. ISpcifi beat of I - ~ oneidiated Smphit.

1.500 . . . . . . . . N ,,  ::. , . .* :*.., ..: :,,  :.:.. .- - .

- 33315712

/5 t tw* 1.000 . . . . .. . . , *. .. ..  : . .- . , .. : :

V 0.300 0.000

-0.5000 Temperature (*C)

Figure 8. Typical curves for rate of release of stored energy for a number of blocks in Pile 2 channel '3/57TR.. [Evaluation of Graphite Safety Issues for the British Production Piles at Windscale: Graphite Sampling in Preparation for the Dismantling of Pile 1 and the Further Safe Storage of Pile 2, B. J. Marsden et al., AEA Technology plc.]

The only graphite that is exposed to neutrons of energies of 1 eV or greater is the graphite in the first several inches toward the core. This section of the graphite protrudes into the reactor pool via the thermal column extension. Consequently, the cooling for this is provided by the reactor pool rather than by airflow through the thermal column. The purpose of the airflow is to prevent Ar-4 1 from being released into the reactor room. As a result, no airflow rate specification is needed.

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.

First Response to RAI Dated April. 13, 20 10 Submitted June 10, 2010 The reference should be 10.2.7.

10.5 Based on discussions and firsthand observations during NRC staff visits to the R"NSC, it is clear that there have been modifications to the pneumatic tube irradiation system described in Section 10.2.3 of the SAR. Revise the SAR to include descriptions and analyses of the current pneumatic system. The revision should be consistent with the guidance in NUREG- 153 7 and not restricted to Chapter 10 of the SAR if the modification affects descriptions or analyses presented elsewhere in the SAR.

Response to December 17, 201 2 RAI Dated March 15, 2013 Changes have been made to the pneumatic tube irradiation (Rabbit) system. The fundamental change that has been made is that the send / receive stations for this system 62

have been moved outside confinement. Consequently, the Safety Analysis Report (SAR) descriptions of the system, as submitted in 2004, needs to be updated:

10.2.3 Pneumatic System 10.2.3.1 Description The pneumatic tube sample irradiation system provides rapid transfer of samples to place them adjacent to the core for gamma and neutron irradiations. It is commonly referred to as the "Rabbit" system. Used primarily for neutron activation, the rabbit system can expose samples to an average thermal neutron flux of approximately 2.83 X 1012 neutrons/cm'-s. There are two semi-independent parallel systems that can be operated simultaneously.

The systems use a closed loop and air pressure to move sample containers from the send/receive station to the terminus at the end of transfer pipe. The air pressure is supplied by a blower located on the landing on the way up to the pool surface. The blower draws air through the solenoid cabinet that is located next to it to create a vacuum.

Within the solenoid cabinet there are eight solenoid valves that control the flow of air.

With the opening and closing of specific valves the air flow in the transfer pipes can be directed towards or away from the core to send or retrieve samples. The loop is completed by the air pipe for each system, which runs parallel to the transfer pipe. These pipes join at the send/receive station and at the terminus adjacent to the core to close the loop. Before passing through the bioshield into the pool the transfer pipe for each system meets with the air pipe to form a single double walled pipe with concentric chambers.

The outer chamber performs the function of the air pipe to the terminus. Air drawn through the blower is exhausted to the confinement off-gas system.

An upgrade moved the system from two separate send/receive stations in the confinement building to a single station outside of confinement. The transfer and air pipes pass through the confinement wall to the send/receive station located on the outside of the southeast wall of the confinement building. The send/receive station includes the send terminal, the return box, and a sample storage container. The send terminal uses the spring loaded doors from the original system, which includes the joint for the transfer and air pipes. Users open the spring loaded doors and place the sample container vertically in the transfer pipe where air pressure draws it upwards. Returning samples fall past the doors and joint into the return box. To maintain air pressure the doors have gaskets and the transfer pipe extends directly into the return box. The entire send terminal is contained in a lockable steel housing. The return box is made of clear acrylic to allow users to visually verify that all samples have returned intact. An internal wall keeps the two systems separate. All the return box panels and openings are sealed to maintain air pressure in the systems. A front facing sliding door allows users to retrieve samples for immediate analysis. Samples requiring a decay period can be dropped through a bottom sliding door into the storage container. The storage box is a lead lined steel container.

The box is approximately 44" long, 28" wide, and 22" high to accommodate a large 63

number of samples. A lockable lead lined sliding door on the top of the box allows access to the samples.

The solenoid valves are controlled by the control panel located adjacent to the send/receive station. The panel is powered through the pneumatic tube control breaker in the control room. The panel allows users to manually send and return samples. Samples can also be irradiated for a predetermined amount of time by using a timer system which is triggered by a switch in the transfer pipe. A reset switch closes all the solenoid valves to stop air flow.

Standard sample containers are about 2 inches wide and 2 to 4 inches long and made of polyurethane or polypropylene. However, the dimensions, materials, and quantity of containers used in each experiment may vary.

10.2.3.2 Evaluation The rabbit system return box and storage container minimize personnel exposure by providing shielding and allowing users to place samples into storage without directly handling activated samples. Samples for immediate analysis can be opened in a shielded fume hood located near the send/receive station to minimize exposure. Area radiation monitors alert users if a sample is returned with a higher activity than expected. The samples and containers used in the rabbit system, as well as the procedures for use of the system, are controlled by the facility radiation safety program.

The activation of argon in the transfer and air pipes poses an immersion hazard if gases accumulate. When either rabbit system is in use there is a constant exhaust of air due to the closed loop system. When all the solenoid valves are closed there is no air flow and therefore no risk of argon-41 buildup. By sealing the spring loaded doors and return box the amount of gas from the rabbit systems allowed to exit into the area of the send/receive station is minimal. Area radiation monitors alert users if dose levels increase at the send/receive station. The rabbit blower exhausts to the confinement off-gas system, and eventually the confinement stack where it is sampled and monitored for gaseous radiation levels.

An experiment containing fissionable materials has the potential to release fission fragments if the container it is in fails. Administrative controls require all rabbit experiments that contain fissionable materials be doubly encapsulated. All rabbit experiments containing fissionable material will be opened inside the confinement room.

This will ensure that any release of fission products will be handled by the confinement's air handling system. See Proposed Technical Specifications 3.8.1.4.2 and 3.8.1.4.3.

Manual ball valves are located in the transfer and air pipes. These valves can be closed by hand in the event of a pipe rupture to prevent siphoning of the pool water.

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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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 P. 14-32 Line 43 Revise the statement that "Shim safety blade inspections are the single largest source of radiation exposure to the facility personnel" to say: "Shim safety blade inspections have the potential to be the single largest source of radiation exposure to the facility personnel".

In keeping with ALARA principles, the shim safety blades are inspected in place. When a blade is inspected, it is raised to it's full out position. A visual inspection is made to the extent possible. Annual measurements are made of the blade drive times and drop times. These measurements provide assurance that there is no significant swelling.

We do not take credit for using a camera system because we do not want to be committed to having a functioning radiation resistant camera system available.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Please see attached 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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 A survey is an evaluation of the radiation hazards associated with the presence of radioactive materials and/or radiation sources under a given set of circumstances. The Radiation Safety Office conducts routine radiation and contamination surveys described in standard procedures to evaluate basic radiological conditions at the RINSC. Surveys require the use of a calibrated survey meter with an appropriate detector as well as a wipe test for removable contamination. Wipe tests may be counted using a liquid scintillation counter, proportional counter, gamma counter or other suitable radiation instrument depending on the isotopes thought to be present. An appropriately calibrated survey meter may also be used to count wipe tests.

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Routine survey frequencies are determined by an evaluation of the radiological hazards likely to be present in the area, its frequency of routine entry or use, and ALARA considerations for the surveyor. The routine survey frequencies for all areas of the RINSC are reviewed and approved by the NRSC. Each survey consists of measurements of fixed and loose contamnination and radiation levels. In accordance with Technical Specification requirements:

" Weekly surveys are completed at intervals not exceeding ten days.

" Monthly surveys are completed at intervals not exceeding six weeks.

" Quarterly surveys are completed at intervals not exceeding four months.

" Annual surveys are completed at intervals not exceeding fifteen months.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The Radiation Safety Office issues personnel monitoring devices to individuals working in controlled areas of the facility, employs bioassay techniques and keeps records of doses received by all individuals for whom monitoring is required. Since the primary airborne contaminant is Argon-41 and it is an immersion hazard, personnel monitoring devices are sufficient to monitor exposure. Respirators are not used for routine entry into controlled areas since airborne levels of other contaminants during routine operations do not exceed 10% of derived air concentrations in 10 CFR Part 20.

The use of radiation monitoring devices for external dose is required for adults who are likely to receive an anmual dose in excess of any of the following (each evaluated separately):

• 0.5 rem (5 mSv) deep-dose equivalent

• 1.5 rems (15 mSv) eye dose equivalent

• 5 rems (50 mSv) shallow-dose equivalent to the skin

• 5 rems (50 mSv) shallow-dose equivalent to any extremity The use of radiation monitoring devices for external dose is required for minors who are likely to receive an annual dose in excess of any of the following (each evaluated separately):

• 0.05 rem (0.5 mSv) deep-dose equivalent

• 0.15 rem (1.5 mSv) eye-dose equivalent 0 0.5 rem (5 mSv) shallow-dose equivalent to the skin

• 0.5 rems (5 mSv) shallow-dose equivalent to any extremity The use of radiation monitoring devices for external dose is required for declared pregnant women who are likely to receive an annual dose from occupational exposure in 66

excess of 0.05 rem (0.5 mSv) deep-dose equivalent, although the dose limit applies to the entire gestation period.

The use of radiation monitoring devices for external dose is required for individuals entering a high or a very high radiation area.

The monitoring records include (whenever applicable):

(a) The deep dose equivalent to the whole body, lens dose equivalent, shallow dose equivalent to the skin, and shallow dose equivalent to the extremities; (b) The estimated intake of radionuclides; (c) The committed effective dose equivalent (CEDE) assigned to the intake of radionuclides and the information used to assess the CEDE; (d) The total effective dose equivalent; and (e) The total of the deep dose equivalent and the committed dose to the organ receiving the highest total dose.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Bioassay is the determination of kinds, quantities or concentrations, and, in some cases, the locations of radioactive material in the human body. Bioassays may be conducted by direct measurement (in vivo counting) or by analysis and evaluations of materials excreted or removed from the human body.

The primary methods of bioassay used are the liquid scintillation counting of urine samples for a wide variety of radioisotopes and in vivo counting for gamma-emitting radioisotopes.

The primary purpose of the bioassay is the determination of the committed dose equivalent (CDE) and the committed effective dose equivalent (CEDE). CDE is the dose equivalent to organs or tissues that will be received from an intake of radioactive material by an individual during the 50-year period following the intake. The CEDE is the dose equivalent to the whole body from the internal uptake of radioisotopes.

Bioassays also act as an independent check on the adequacy of working habits and engineered safety features. When we are determining whether a potential intake should be evaluated, we consider the following circumstances:

" The presence of unusually high levels of facial and/or nasal contamination

" Entry into airborne radioactivity areas without appropriate exposure controls

" Operational events with a reasonable likelihood that a worker was exposed to unknown quantities of airborne radioactive material 67

" Known or suspected incidents of a worker ingesting radioactive material

" Incidents that result in contamination of wounds or other skin absorption

• Evidence of damage to or failure of a respiratory protective device Bioassays may be required for anyone handling or using unsealed radioactive sources.

Bioassays are required for adults likely to receive an annual intake in excess of 10 percent of the applicable annual intake limits.1 Bioassays will also be required for minors and declared pregnant women likely to receive an annual intake in excess of I percent of the annual intake limits for adults.

The annual limit on intake (ALI) is the activity of an intake of radioactive material that if taken alone would irradiate a person, 2 to the limit set for each year of occupational exposure.

The dose equivalents are recorded annually on a clear, legible record containing all of the information required by NRC Form 5.The accuracy and precision of our dosimetry service is independently tested by the National Institute of Standards and Technology, in accordance with American National Standards Institute ANSI N13.11-2009. Our dosimetry service is fully accredited in all testing categories for Ionizing Radiation Dosimetry by NVLAP (United States Department of Commerce, NIST, National Voluntary Laboratory Accreditation Program) for satisfactory compliance with criteria established under Title 15, Part 285, Code of Federal Regulations.

Found in Rhode Island Department of Health's Rules and Regulations For The Control of Radiation, Part A, Appendix B and 10 CFR 20, Appendix B.

Represented by Reference Man 11.6 The response to RAI 11.3 didn't fully address NUREG-1537 Section 11.1.4 and 11.1.6 in regards to the bases for the methods and procedures used for conducting radiation and contamination surveys. Provide a more in-depth description of the nominal frequencies at which the facility is surveyed for these hazards. The description should include additional surveys that may be used during non-routine activities.

Response to December 17, 2012 RAI Dated March 15, 2013 As stated in our original response to RAI 11.3, we conduct routine surveys of facility areas weekly, monthly, quarterly, or annually. Survey frequency is a function of the radionuclides used, quantities authorized, experiments conducted and/or occupancy of the area in question. All areas of the facility are routinely surveyed. Routine surveys are supplemented by surveys taken by the individual users of their work areas and themselves ("frisks"). Each routine survey consists of measurements of fixed and removable contamnination and radiation levels. We survey accessible areas with gamma or neutron fields monthly and verify/update posting. It is inappropriate to specify routine 68

frequencies for individual facility areas in the FSAR since the presence and/or use of radioactive materials in an area may change. Typically areas routinely using unsealed gamma emitters, and beta and alpha emitters capable of being detected by survey meters are surveyed at least weekly, but more often when any procedure is likely to produce significant radiation and/or contamination.

Action levels for removable contamination have been adapted from NRC Regulatory Guide 8.23, Table 2. We follow the prudent ALARA practice of immediately cleaning areas where removable contamination was identified. Action levels for external radiation fields follow NRC guidance for posting radiation, high radiation and very high radiation levels. We follow prudent ALARA practice by posting areas with measurable radiation levels and isolating those with high or very high radiation levels.

Instruments used for our surveys are calibrated at least annually and immediately following any maintenance (including replacement of batteries). Portable survey instruments are supplemented by area radiation and effluent monitors described in other sections of the FSAR.

Survey frequencies are reviewed and approved, and survey completions are audited by the Nuclear and Radiation Safety Committee.

11.7 The response to RAI 11.4 didn't fully describe the provisions for the use of extremity monitoring and the conditions under which extremity monitoring is used. Provide this information.

Response to December 17, 2012 RAI Dated March 15, 2013 It is our policy to assign extremity monitoring to any individual likely to receive a measurable radiation dose to the extremities. As stated in our original response to RAI 11.4, wve assign extremity monitors to adults who could receive an annual dose equivalent to the hand in excess of 5 reins and minors who could receive an annual dose equivalent to the hand in excess of 500 mrem. In practice, we assign extremity monitors to anyone qualified as a radiation worker since he/she may handle radioactive materials and could reach an extremity dose equivalent threshold where monitoring is required.

In our radiation worker training, we recommend that extremity dosimeters be worn under any protective gloves on the hand likely to receive the greatest exposure (typically the dominant hand) with the dosimeter face (ring badge) facing the radiation source (typically toward the palm of the hand).

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.

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Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The two credible accidents from the list of nine credible accidents shown in NUREG-1537 are accidents involving mishandling or malfunction of fuel and experiment malfunction. The mishandling or malfunction of fuel is an initiating event for the Maximum Hypothetical Accident (MHA), which was analyzed. The major accident involving experiments is related to an unanticipated reactivity insertion. A new analysis for a rapid insertion of the maximum reactivity worth of all experiments of 0.6% Ak/k (TS 3.1.3) is provided in RAI 13.7.

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.

Third Response to RAI Dated April 13, 2010 Submitted August 18, 2010 The answer to this question is covered in the basis document entitled "Radiological Assessment Attachment".

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)

Third Response to RAI Dated April 13, 2010 Submitted August 18, 2010 The answer to this question is covered in the basis document entitled "Radiological Assessment Attachment".

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.

Third Response to RAI Dated April 13, 2010 Submitted August 18, 2010 The answer to this question is covered in the basis document entitled "Radiological Assessment Attachment".

13.5 The footnote of Table 13-5 indicates a 50% reduction of noble gases. Explain the reason for the reduction in noble gases.

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Third Response to RAI Dated April 13, 2010 Submitted August 18, 2010 The answer to this question is covered in the basis document entitled "Radiological Assessment Attachment".

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.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 ANL generated a new analysis for a 0.6 % dk/k reactivity insertion accident since total experiment worth is limited to this much excess reactivity. This section of the SAR will be re-written based on this analysis. See the response to RAI question 13.7.

Figure 13.1 of the SAR references a graph of the core power and peak cladding temperature with time. This figure was not re-generated for the new analysis.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 For these RAI, four reactivity insertion transients were re-analyzed for forced convection cooling mode, and one reactivity insertion was analyzed for natural convection mode using the PARET/ANL Version 7.5 code.

PARET/ANL Version 7.5 solves the point-kinetics equations for reactor power versus time, while computing thermal-hydraulic conditions in one or more fuel channels.

Feedback firom the thermal-hydraulic solution is continuously fed back into the point-kinetics equations. The reactor is modeled using two hydraulic channels: a hot channel, and an average channel. The hot channel represents conditions in the coolant channel between fuel plates that is most limiting. This channel typically has fuel plates having the highest power density adjacent to it. The average channel represents all other coolant channels adjacent to fuel plates. Reactivity feedback from fuel heatup (Doppler Effect),

water heatup, and water density change, are accounted for using feedback coefficients derived from the neutronics models. The input to the PARET model consists of several categories of information:

1. geometry of the channels (fuel meat, clad, coolant)

2. delayed neutron kinetics data

3. reactivity insertion definition

4. control system response

5. initial operating conditions of power and flow

6. solution options such as time step sizes and edit selection.

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The neutronics codes used to generate input for the PARET models were:

WIMS/ANL for multi-group neutron cross sections; REBUS-PC (which includes DIF3D as the neutronics solver) for power density information, and VARI3D (which also includes DIF3D as the neutronics solver for real and adjoint flux), to provide the reactor kinetics delayed neutron fractions, decay constants, and prompt neutron lifetime. Data on reactor power distribution is provided in Reference 2, and data on the reactor kinetics parameters and reactivity feedback coefficients is provided in RAI 4.10.

The forced convection transients are assumed to take place under the following assumptions [Technical Specifications, Revised Section 2.1.1 in RAI 14.36, "Safety Limits in the Forced Convection Mode]:

Measured Parameter Limiting Trip Value Safety Limit P 2.3 MW 2.4 MW m 1740 gpm 1580 gpm H 23'9.1" 23' 6.5" To 123 F 125 F The natural convection transients are assumed to take place under the following assumptions [Technical Specifications, Revised Section 2.1.2 in RAI 14.52, "Safety Limits in the Natural Convection Mode]":

Measured Parameter Limiting Trip Value Safety Limit P 125 kW 200 kW H 23' 9.1" 23' 6.5" To 128 F 130 F The period trip at 4 seconds is assumed to fail. The power trip is functional. The time delay for control blades to begin to move after a trip is assumed to be 100 ms. The time to full insertion is the maximum allowed of 1.0 second [TS 3.2.3].

Case 1: Rapid Insertion of 0.6% Ak/k Reactivity From Very Low Power The reactor is initially operating at 10 watts, 123 °F coolant inlet temperature, and 1740 gpm. There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715 x l0s Pa. Then 0.6% Ak/k reactivity, the total reactivity worth of all experiments [TS 3.1.3], is inserted as a very short ramp of 0.1 second duration, starting at 0.0 seconds. The reactor power rises rapidly. The power trip at 2.3 MW is actuated at 10.179 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.423 MW at 10.279 s and the control blades are inserted. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for fuel meat centerline, clad surface, and coolant are: 79.8 'C; 79.1

°C; and 63.6 'C, respectively, at 10.30 s. These peak fuel and clad surface temperatures 72

are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-1313).

The peak coolant temperature is well below the saturation temperature of 115.4 'C.

Case 2: Slow Insertion of 0.02 % Ak/k /Second Reactivity From Very Low Power The reactor is initially operating at 10 watts, 123 'F coolant inlet temperature, and 1740 gpm. There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715 x 105 Pa. Then a long, slow ramp reactivity insertion begins at a ramp rate of 0.02 % Ak/k / s (TS 3.2.4), continuing for 100 s. Power rises slowly. The power trip at 2.3 MW is actuated at 32.198 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.509 MW at 32.298 s. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for fuel meat centerline, and clad surface are: 79.1 'C; 78.9 'C. The peak coolant temperature of 62.8 'C is reached at 32.40 s. These peak fuel and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-13 13). The peak coolant temperature is well below the saturation temperature of 115.4 'C.

The safety limit on power of 2.4 MW is exceeded (the power briefly reaches 2.509 MW). However, the safety limit on power does not apply to transients. In this case, the fuel meat and cladding reach peak temperatures of about 79 C, far below the maximum allowed temperature of 530 'C Case 3: Slow Insertion of 0.02 % Ak/k / Second Reactivity From 1.8 MW Power The reactor is initially operating at 1.8 MW, 123 'F coolant inlet temperature, and 1740 gpm. There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715 x 105 Pa. The coolant inlet temperature for which an outlet temperature of 123 'F is reached was iteratively determined to be 113.6 'F (45.34 °C).

Starting from this initial condition, a long, slow ramp reactivity insertion begins at a ramp rate of 0.02 % Ak/k / s [TS 3.2.4], continuing for 100 s. Power rises slowly. The power trip at 2.3 MW is actuated at 6.774 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.313 MW at 6.874 s. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for ftiel meat centerline, and clad surface are: 76.7 'C; 75.9 'C; and 59.6 Y'. The peak coolant temperature of 59.6 'C is reached at 6.90 s. These peak fuel and *.. e .,tmperatures are far below the maximum temperature of 530 'C for L,ELI slici,2 'u.- that the NRC finds acceptable as fuel and clad temperature limits not 73

to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-1313). The peak coolant temperature is well below the saturation temperature of 115.4 'C.

Case 4: Slow Insertion of 0.02 % Ak/k /second Reactivity From 2.2 MW Power The reactor is initially operating at 2.2 MW, 123 'F coolant inlet temperature, and 1740 gpm. The coolant inlet temperature for which an outlet temperature of 123 'F is reached was iteratively determined to be 111.5 'F (44.19 °C). There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.715 x 105 Pa. Starting from this initial condition, a long slow ramp reactivity insertion begins at a ramp rate of 0.02 % Ak/k / s [TS 3.2.4], continuing for 100 s. Power rises slowly. The power trip at 2.3 MW is actuated at 2.498 s. Since no actual negative reactivity from the control system occurs for 100 ms after the trip, the reactor power continues to rise from the trip level of 2.3 MW to a maximum of 2.308 MW at 2.598 s. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for fuel meat centerline, and clad surface are: 75.9 'C; 75.1 'C; and 58.5 'C. The peak coolant temperature of 58.5 'C is reached at 2.600 s. These peak fuel and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-1313). The peak coolant temperature is well below the saturation temperature of 115.4 'C.

Case 5: Rapid Insertion of 0.6% Ak/k Reactivity From 100 kW Under Natural Convection Cooling The reactor was brought up to 100 kW under natural convection conditions with a maximum outlet temperature of 128 'F. There is a water head of 23' 9.1" above the top of the fuel meat, which provides a pressure of 1.719 x 105 Pa. Power and flow are allowed to stabilize out to 360 s, at, xvhich time the power is 100 kW. Then a very short reactivity ramp of 0.6% Ak/k is inserted over 0.1 s, starting at 360.00 s. The power trip at 125 kW is actuated at 360.036 s. The reactor power continues to rise from the trip level of 115 kW to a maximum of 404 kW at 360.140 s. The reactor power drops rapidly to shutdown conditions.

Peak temperatures for fuel meat centerline (65.7 °C) and clad surface (65.7 °C) occur at 360.18 s, whereas the peak coolant temperature (62.2 °C) occurs at 59.4 s during the rise to power. These peak fuel centerline and clad surface temperatures are far below the maximum temperature of 530 'C for LEU silicide fuel that the NRC finds acceptable as fuel and clad temperature limits not to be exceeded under any conditions of operation (See NUREG-1537, Part I, Appendix 14.1 and NUREG-1313). The peak coolant temperature is well below the saturation temperature of 115.4 'C.

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The safety limit on power of 200 kW is exceeded (the power briefly reaches 404 kW).

However, the safety limit on power does not apply to transients. In this case, the fuel centerline and clad surface reach peak temperatures of about 66 'C, far below the maximum allowed temperature of 530 'C.

Reference

1. A. P. Olson, A USERS GUIDE TO THE PARET/ANL V7.5 CODE, May 1, 2010, GTRI-Conversion Program, Nuclear Engineering Division, Argonne National Laboratory Internal Memorandum, May 1, 2010.

2. Memo dated 3 September 2010 from Earl E. Feldman to James E. Matos entitled "Steady State Thermal Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor" 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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 This question confuses the fact that the 200 msec cited is a very conservative estimate of the delay from the time that the scram signal is initiated, to the time that the blades begin to drop, while the 1 sec insertion time represents the total amount of time that it takes from the initiation of the scram signal, to full insertion of the control blade.

In an effort to be consistent, all of the new analysis will be done with the assumption that the delay time between the initiation of the scram signal, and the time that the control blades begin to drop is 100 msec. This is considered to be conservative 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)

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the analysis of reactivity insertions provided in RAI 13.7 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.

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First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The only open penetrations into the pool are the rabbit, through port, and beam port tubes. The rabbit tubes enter through the pool wall at an elevation that is close to the top of the pool. Consequently, shearing open a rabbit tube will not lead to significant draining of the pool. Dropping something into the reactor pool, and shearing the through port is not considered to be a credible accident scenario because it runs underneath the thermal column extension. As a result, the beam ports are used for the LOCA analysis, and the assumption is made that the beam port extension for one of the largest beam ports is sheared off. In all of these cases, the likelihood of dropping anything into the pool that causes this kind of damage is very low because there is a steel plate bridge over the top of the core.

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.

Tenth Response to RAI Dated April 13, 2010 Submitted July 15, 2011 The Addendum LOCA analysis shows that as long as the area between each individual experimental port and confinement is no greater than 1.48 in2 , then there is sufficient pool drain time to allow for decay power to reach the point at which the fuel cladding cannot be compromised. However, this assumes that the water level will not drop below the elevation of the bottom of the eight inch beam ports. The elevation of the bottom of the through port is below the elevation of the bottom of the eight inch beam port, and an analysis for a LOCA in which the fuel is completely un submerged has not been performed. The answer to RAI question 10.2 shows that administrative controls on the use of the through port will prevent conditions from occurring that could lead to a LOCA that has not been analyzed. Therefore, the administrative controls will be set conservatively to say that:

1. Each beam port shall have no more than an area of 1.25 in 2 open to confinement during reactor operation.

2. When the reactor is in operation, the drain valve to the through port shall be closed.

3. When the through port is in use, gate valves shall be installed on the end(s) of the port that will be used for access.

4. When the through port is not being monitored for a leak condition, the ends of the port shall be closed.

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The bases for these specifications will be that:

Specification 1:

The LOCA analysis shows that as long as the pool level does not drain through an area greater than 1.48 in2 to confinement, then there will be sufficient time for decay power to drop to a point which will not damage the fuel cladding, provided that the pool level does not drop below the elevation of the bottom of the eight inch beam ports. It also shows that if any single port has a catastrophic failure, the un-damaged ports do not become pool drain pathways. Consequently, 2

limiting the areas of each experimental port that is open to confinement to 1.25 in is conservative.

Specification 2:

The through port has three potential pool leak pathways. The first is the through port drain. By keeping this drain closed during operation, that potential leak pathway is blocked, and the potential for an unnoticed pool leak though this experimental facility is prevented..

Specification 3:

If the end(s) of the through port that will be used for access have gate valves mounted to them, then in the event of a leak, the port can be easily isolated so that the leak is stopped.

Specification 4:

The LOCA analysis has shown that the amount of time available for performing mitigating actions in the event of a pool leak is on the order of hours.

Consequently, as long as reactor personnel will become aware of a pool leak though the through port reasonably quickly, and the gate valves are in place, the consequence of the leak can be mitigated quickly by closing the valves.

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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Because the normal water level of the pool is greater than 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, it should not bc used for the initial coolant height in the LOCA analysis. Section 13.',, Loss of' Coolant Accident (LOCA), has been 77

completely replaced and the replacement is attached. In the new analysis the coolant level at which the scram occurs is 23.54 feet above the top of the active core, which in the new analysis is taken to be the top of the fuel meat. This level is the minimum pool level that is permitted by the Safety Limits while operating at any force convection power level, TS 2.1.1.

13.13 Page 13-10. Provide definitions for the terms hi, h2, and C in the equation on line 43.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 C = coefficient of discharge, which is dependent on the type of orifice through which the water is draining. We assume the orifice is a standard sharp-edged orifice, which has a discharge coefficient of 0.61. The reference for this is Mark's Mechanical Engineer's Handbook, Theodore Baumeister (editor), McGraw-Hill, New York, 1958.

P. 3-62.

hl = upper elevation of the water level h2 = drain elevation Fifth Response Submitted to RAI Dated April 13, 2010 November 26, 2010 Revised 08/26/10 hl is the initial water level. h2 is the final water level, which is located at the bottom of the failed beam port. C is the orifice coefficient for the assumed 1/2,--inch diameter hole through which water flowing through the failed beam port exits the pool.

However, section 13.2.3, Loss of Coolant Accident (LOCA), has been completely replaced and the replacement is attached. In the new analysis of the draining of the pool, section 13.2.3.2. "Drain Time," the variables are defined and the, model is derived from first principles. In the new analysis, hl, h2, and C are replaced by hl, hf, and Cd, respectively.

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 U3si2 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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Section 13.2.3, Loss of Coolant Accident (LOCA), has been completely replaced and the replacement is attached. In the new analysis in section 13.2.3.14, 986- F (530° C) is identified as the temperature limit for the fuel plates during the LOCA. In section 13.2.3.13 of the new analysis, a maximum fuel temperature of 486' C, which is less than the temperature limit, is predicted during the postulated LOCA.

78

13.15 Page 13-13. Please provide a conclusion for the analysis ending on line 18 of this page.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The conclusion for the analysis ending on line 18 of this page, as is indicated symbolically on lines 13 through 18 of page 13-13, is that 0.013 Btu/sec would be conducted vertically downward to the submerged portion of the fuel plate through the fuel meat part of the fuel plate if the maximum fuel temperature (at x = 2.0') is 1200' F and the temperature of the fuel plate at the water surface (at x = 0.7') is 212'F.

However, section 13.2.3, Loss of Coolant Accident (LOCA), has been completely replaced and the replacement is attached. The analysis ending on line 18 of Page 13-13 of the 2004 RINSC safety analysis report belongs to the fuel assembly heat transfer model of that report. A new heat transfer model is provided in the new analysis in section 13.2.3.4, "Development of Heat Transfer Model".

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The power distribution in the reactor along the length of the fuel tends to have a shape that can be approximated by a sinusoid with the peak value near the middle of the length. In such situations, often a chopped cosine is typically used. However, section 13.2.3, Loss of Coolant Accident (LOCA), has been completely replaced and the replacement is attached. The new analysis assumes that the axial distribution of the decay power spatial distribution is uniform. The new analysis explains why a uniform distribution bounds the worst case. Specifically, the new section 13.2.3.4 states:

An issue is the axial distribution of the power along the length of the fuel plate.

Typically, the distribution will tend to be in the shape of a symmetric chopped cosine shape whose peak is near the center of the 23.25-inch fuel meat length.

Since the center of the fuel meat length is 4 inches above the water level, within the exposed length of the fuel plate the power is skewed toward the bottom. The more that power is skewed toward the bottom, the lower the peak solid temperature will be. The reason for this is that heat generated lower in the exposed portion of the fuel plate length has a shorter conduction path to the surface of the water than does heat generated higher in the exposed portion.

Moreover, as is obvious, lower fuel temperatures will also be produced if a greater portion of the heat produced over the entire fuel plate is generated below the surface of the water. Figure 4.6-5 (of the currently replaced section 4.6 of the RINSC reactor safety analysis report, "4.6 Steady-State Thermal-Hydraulic Analysis" Reference BB), provides the axial power shape for the fuel meat length of the highest power plate, which is the one next to the beryllium reflector in assembly D6. In order to keep the model simple and avoid the issue of the precise axial power shape, we will use a bounding approach and assume that the 79

heat generation rate in the fuel plate is uniformly distributed over the entire length of the fuel meat. The water level is at 0.672, i.e., 15.625 inches/23.25 inches, in the Figure 4.6-5. Numerical integration of the axial power in the figure shows that 56.6 % of the power is generated in the exposed portion of the fuel meat length and the remaining 43.4% is generated in the submerged portion.

This is to be compared with the uniform distribution, in which 67.2% of the power is generated in the exposed part of the fuel meat length and the remaining 32.8% is generated in the submerged portion.

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.

Fifth Response to RAL Dated April 13, 2010 Submitted November 26, 2010 The missing text from "Heat Conduction to the Water in Core Box from the Non-Fuel Aluminum in the Element" is provided in Figure 13.17-1 in the box below. It was copied from Appendix D of the Safety Analysis Report for the Low Enriched Fuel Conversion of the Rhode Island Nuclear Science Center Research Reactor, Revision 1, December, 1992. However, Section 13.2.3, Loss of Coolant Accident (LOCA), of the 2004 RINSC safety analysis report has been completely replaced and the replacement is attached. Although the new analysis does not have a "Heat Conduction to the Water in Core Box from the Non-Fuel Aluminum in the Element" section, it is complete, the results of the calculations are discussed and all assumptions made in the calculations and justifications for those assumptions, and any conclusions based on the calculations are provided. See, for example, section 13.2.3.4, "Development of Heat Transfer Model," of the attached replacement.

80

Figure 13.17-1 Missing Text Q = 131X0O09x(12QO-212) = 89.604 Btu/hr x 1/3600 I = (2-.7) x 1.3' Q = .02489 Btu/sec total heat conducted = fuel + aluminum

= .013 + .02489 = .03789 etu/sec From the original SAR it was assumed that about 30% of the heat was used in steam format ion therefore .3 x .03789 = .011367 Btu/sec and the total heat removal= .03789 + .011367

= .049 Btu/sec Since the heat generation of.0397 Btu/sec is less than that required to reach the plate melting point (.049 Btu/sec), it is assumed that the fuel does not reach the melting point.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the response to Question 13.19.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 RELAP5 cases were run to analyze loss-of-coolant-flow accidents in the RINSC reactor.

These cases were consistent with the Limiting Safety System Settings in the Forced Convection Mode. Cases were run both with and without proper opening of the natural circulation gate valves during the transient.

The RELAP5 model included the peak fuel channel, representing the highest power stripe in the highest power plate and its coolant. This model also included the average fuel channel, representing the rest of the fuel plates and their coolant. Bypass flow around the graphite reflectors, through the beryllium reflectors, and inside the control blade shrouds was represented in the model. Also bypass flow through the gamma shield was represented. The model included the pump, the heat exchanger and the 81

associated piping. The model also included the coolant ducts between the piping and the core box, as well as the natural circulation gate valves in the duct walls. The pool was also represented.

The initial steady-state conditions for the calculations were set at the hottest conditions that might not trip a scram. These conditions were taken from the limiting trip values in the Limiting Safety System Settings in the Forced Convection Mode. These conditions are listed in the table below.

Initial Steady-State Conditions for the LOF Cases Parameter Value Reactor power 2.3 MW Total pump flow 1740 GPM Height of water above the top of the core 23 ft. 9.1 in.

Outlet temperature 123 F The transient was initiated by a loss of power to the pump. RINSC personnel have measured the pump flow after a pump trip. The measured results are shown in the figure below. Also shown is a smoothed fit to the data. The smoothed fit was used to specify the pump flow in the RELAP5 calculations.

1800 1600 Solid Line = Data X--_Eit o- -Data_--

1400 n1200 1000 0

IL cL 800 E

o- rnn 400

- -- - - - - - - - --- - - - - - - .

- - - - -- ...

,. .. .

200 0

13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Time (s)

Measured Pump Coastdown Curve 82

Key parameters of the transient calculation are given in the table below.

Key Parameters for the Loss-of-Flow Transients Parameter Value Justification Pump coast-down As shown in figure above Measured Time when natural 9 seconds after start of Measured circulation gate valves open pump coast-down Scram reactivity 1 % Ak/k Limiting Conditions for Operation Scram reactivity insertion 1 second Limiting Conditions for time I Operation For the base case, the one in which the natural circulation gate valves open properly, the timing of transient events is shown in the table below.

Loss-of-Flow Transient Timing, Gate Valves Open Properly Time, s Event 0.0 Pump trips 0.0 Low flow scram tripped immediately 0.1 Control blades start moving 1.0 Control blades fully inserted 5.3.1 Flow reversal in peak fuel channel 8.01 Flow reversal in average fuel channel 9.0 Natural circulation gate valves open 9.41 Clad surface temperature peaks at 115.62 C in middle of core, peak fuel temperature = 115.73 C, coolant saturation temperature =

115.90 C at this point.

The normalized power and pump flow for the base case are shown in the figure below.

For the first second, while the control blades are inserting, the power and flow fall at about the same rate. After that the power falls slower, and the pump flow drops to zero.

83

1.0

. . . . . . . . . . . . . . .

.9 ........ ........

0 0- .8 . . . .. ....

... .... .... .... . ... . . . . ... . .. . . . . . . . .

=.

6 .6

"* 4 '. . . '. . . . . . " " . . . " . . . . '. . . . '. . . .. . . ... . . . - -.- .-

-..

-. . .-

(V N

o .2.. . . . . . .

0.0 0 2 4 6 8 10 12.14 16 18 20 .22 .24 .26 .28 .30 Time (s)

Power and Pump Flow for the RINSC LOF Flow rates through the fuel channels, the reflectors, the control rod sheaths and the outlet duct are shown in the figure below. In this figure positive flow rates are upward, and

• negative flow rates are downward. As indicated in the table above, flow reversal occurs in the average fuel channel at about 8 seconds. During normal forced flow operation both flow around the graphite reflectors and flow through the control rod sheaths

• contribute significantly to bypass flow, but after the pump coasts down the flow through the control blade sheaths dominates the bypass flow. During forced flow the flow through the outlet duct is upward. During normal natural circulation operation the flow in the outlet duct is downward. During the loss-of-flow transient the flow in the outlet duct is upward until the gate valves open. It then takes a few seconds for the flow to reverse in the outlet duct.

84

5 . . . . .

4. . . ~~. . . ... -- -- --- - - - - - - - -- - - -- - - -

3 ~.

0.:

- verage Fuel Element nt" Hot Fuel Element.

00 .-Be Reflector LLr hit 2 -- - - - -- - - --- - - --- .. . onrl.lade

. Shea -- - . - ..-- -

0

-3 -....... ....... ,

-- -- - utlet Duct.

-5 0 2 4 6 8 10 12 14 16 18 20 22 24 26 .28 .30 Time (s)

Flow Rates for the RINSC LOF Clad surface temperatures at a number of nodes in the peak fuel channel are shown in the figure below. The highest clad temperature occurs at node. 15 which is at the middle of the core. The peak clad surface temperature almost reaches the saturation temperature, where sub-cooled boiling would start. As indicated in the transient timing table above, the peak fuel centerline temperature is a fraction of a degree hotter than the peak clad surface temperature.

85

120 Saturation - -"

...

115 .........

-Node 18 105 100 T95

• -90 .........

E

~)85 D 80

-- ":Node 15, Middle of Core 5 __7 ---

,- - -- - --- --. . . .

70 - . . No 12 55 ,Node.,.Top.of.Core 30; 50 0 2 4 6 8 10 12 14 16 18 20 .22 .24 .26 .28 .30 Time (s)

Clad Surface Temperatures in the Peak Fuel Channel The peak coolant temperatures in this transient are close to the onset of sub-cooled boiling. If the initial power level were higher and the onset of sub-cooled boiling occurred, then the coolant heat transfer coefficient would go up and would limit the rise in fuel and clad temperatures.

The failure of the natural circulation gate valves to open during a loss-of-flow transient would be an extremely unlikely event, but the loss-of-flow transient was repeated with failure of the natural circulation gate valves to open. In the case in which the gate valves do not open the peak clad surface temperature was a fraction of a degree lower (115.61 C vs 115.62 C) and occurred at the same time (9.41 seconds).

The smallness and the sign of the difference in peak clad surface temperature due to the failure of the gate valves to open can be explained by two factors. First, in both cases the peak temperature occurred only 0.41 second after the gate valves should open. The flow rate figure above indicates that the change in the outlet duct flow rate in the first 0.41 second after the gate valves open is small, so the change. in peak clad temperature should be small. Second, close examination of the flow rate figure near 9 seconds indicates that if the gate valves open the outlet duct upward flow rate immediately after the gate valves open is slightly higher than what would be obtained by extrapolating the flow rate from before the gate valves open. This leads to slightly more coolant being 86

sucked out of the outlet plenum and slightly less up-flow through the fuel channels, causing slightly higher temperatures in the fuel channels if the gate valves do open.

The question of why the opening of the gate valves initially leads to slightly more up-flow in the outlet duct requires consideration of the factors that determine the flow rate in the outlet duct. Before the gate valves open the flow rate in the outlet duct is caused by the pump sucking coolant upward through the duct at a rate determined by the pump coast-down. When the gate valves first open at 9 seconds the coolant in the outlet duct is still somewhat hotter than the pool temperature. The natural circulation head in the outlet duct leads to natural circulation flow up the duct and partly out the gate valve.

The natural circulation up-flow in the duct is partially reduced by the up-flow through the fuel channels. The fuel channel flow sucks coolant out of the outlet plenum and lowers the pressure at the bottom of the duct. The natural circulation duct flow happens to be initially somewhat higher than the pump flow at that time in the coast-down. As time progresses after the gate valves open, the coolant temperatures in the outlet duct decrease and the natural circulation up-flow through the fuel channels increases, leading to flow reversal and down-flow in the outlet duct.

In conclusion, a loss-of-coolant-flow transient in the RINSC reactor would result in peak fuel and clad temperatures that are hundreds of degrees below the temperatures at which damage to the fuel plates would occur.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The forced convection transients are assumed to take place under the following assumptions [Technical Specifications, Revised Section 2.1.1 in RAI 14.36, "Safety Limits in the Forced Convection Mode]:

Measured Parameter Limiting Trip Value Safety Limit P 2.3 MW 2.4 MW m 1740 gpm 1580 gpm H 23' 9.1" 23' 6.5" To 123 F 125 F The natural convection transients are assumed to take place under the following assumptions [Technical Specifications, Revised Section 2.1.2 in RAI 14.52, "Safety Limits in the Natural Convection Mode]":

87

Measured Parameter Limiting Trip Value Safety Limit P 125 kW 200 kW H 23' 9.1" 23' 6.5" T, 128 F 130 F In each case, the Limiting Trip Values are used in the analysis because they are the permitted operating conditions that produce the most extreme fuel and clad temperatures. The Startup Accident has been analyzed under both forced-flow and natural-convection flow conditions. See RAI 13.7, Case 1 and Case 5.

Treatment of Power Peaking Factors:

Within a giyen fuel plate, there is a variation of power density across the width of the plate and along the length of the plate in the axial direction. Each of the 308 fuel plates in the core is in a different neutron flux environment. The power density is fully 3-dimensional.

In the original neutronics analysis, the DIF3D code edited an estimate of the local peak power density with each fuel volume, based on neutron flux and current gradients. The purpose of this method was to enable the computational model to be small enough to be tractable at that time. The present analysis takes advantage of advances in computational capabilities of the latest version of DIF3D, running on much faster computers with much more memory than existed before.

The new neutronics models captured the 3D effects by a two-step process. In the first step, which is equivalent to the method used in the original analysis, a 3D core model was defined in X-Y-Z geometry, with Z being the axial dimension. There were 17 axial nodes along the length of the fuel meat. In the X-Y plane, this model represented each fuel assembly as three components: two side plates, and a homogenized region consisting of all of the fuel meat, clad, and coolant water associated with the 22 fuel plates inside the envelope of the fuel assembly. Water between the fuel assemblies was separately modeled. The WIMS/ANL code was used to obtain multi-group neutron cross sections for the various compositions in the reactor. Then the DIF3D multi-group diffusion theory neutronics code calculated the neutron flux and power distribution in the core. From this result, the power per fuel assembly was obtained. Fuel assemblies D6 and E6 were found to be the two most-limiting assemblies.

In the second step, the spatial mesh was refined in the X-Y plane through D6 and E6 in order to capture additional spatial detail regarding power distribution with the fuel assembly. The spatial mesh spanning D6 and E6 was increased in order to provide a mesh interval for every plate, and one for each of nine stripes across the width of the plate. In this way, it was possible to determine the precise location and value of the peak power density, rather than using an estimate based on neutron flux and current.

Assembly D6, plate 1 (the plate closest to core center), was found to have the largest power density.

88

The peaking factors for the location with the highest power density can be found as follows. The neutronics results yield a 3D array of power density values by mesh interval. The ratio of any value in this (X,Y,Z) array to the average core power density is the local power peaking factor. The relative power density profile along the hottest stripe of the hottest fuel plate (plate 1) is plotted in Fig. 4.6-5. Figure 4.6-6 shows the variation of relative power density across the width of the plate. Stripe 9 has the largest power density. When this local power peaking factor is divided by the axial power peaking factor from Fig. 4.6-5, one obtains the radial power peaking factor. See Reference BB.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 A revised Figure 13-1 is provided based on the new transient analyses that were done for these RAI. See the answer to RAI Question 13.19.

120 Saturation, 115

-110 105 . .. .. ... . . . . ..

... - - - - - - - - - - - -

L.,- 100 ----- ----- ------d le O O - -re- --- -

95 d) 90 285 p380 75 70 ~Node 15, Middlof Core-65 60 55 50 0 2 4 6 8 10 12 14 16 18 20 .22 .24 .26 .28 .30 Time (s)

Clad Surface Temperatures in the Peak Fuel Channel 89

Beam Port Figure 13-2 90

Beam Port Vent and Drain Connection Figure 13-3 Figure 13-4 (Schematic Diagram for the Postulated Loss of Coolant Calculations),

referenced on page 13-10, line 1, has been deemed unnecessary and out-dated and has thus been removed from the SAR. Chapter 13 will be corrected accordingly.

13.22 The responses to RAI 13.4 and RAI 13.5 reference empirical data in a report on a fuel failure at the University of Virginia. Provide a copy of the report and explain why the report is applicable to the RINSC reactor (for example, similar fuel composition, similar operating characteristics, etc.).

Response to December 17, 2012 RAI Dated March 15, 2013 The reference entitled "UVAR-18, Part I Revised Safety Analysis Report in Support of Amendment to License R-66 for Two Megawatt Operation University of Virginia Reactor", October 1970 has been included with the proposed answers to the RAI questions.

Page 83 of this report briefly discusses a fission plate failure that occurred at the University of Virginia Research Reactor in 1968. The noble gas concentration due to this failure was measured in the reactor room. The iodine isotope concentration in the reactor 91

room was too small to be measured, but was inferred to be smaller than the noble gas concentration in the reactor room by at least 10%. In both cases, the concentrations were in terms of the total fission plate inventory. Consequently, these concentrations represent the release fraction that got through the plate walls, and through the water in the reactor pool. This report has no data on the specific enrichment or fuel type of the fission plate.

However, since the data is in terms of the percentage of the total quantity available, it represents a release fraction that can be applied to any situation, as long as the source term is known, and there is a comparable water column through which the fission products must pass in order to reach confinement air. For the RINSC analysis, the source term has been calculated from first principles, and applied to the saturation activites of the fission products. The UVA data is used to estimate the fraction of the noble gasses and iodine isotopes that would be released into the reactor room. A refined analysis of this is in the document entitled "Fuel Failure Addendum".

Page 62 of the report indicates that the pool level is at least 19.75 feet above the active core, and that the maximum depth of the pool is 26 feet 4 inches. Page 21 of the report indicates that the active fuel element meat length is 24 inches. Therefore, the maximum amount of water over the fuel would be 19.75 feet + 2 feet = 21.75 feet. The fission plate would be positioned next to the core, so we can assume that there was a maximum of 21.75 feet of water through which the noble gases and iodine isotopes would pass before being released to confinement. The proposed safety limit for the RINSC reactor pool height above the fuel meat is 23 feet 6.5 inches. Therefore the minimum water column height through which gases and isotopes from a fuel failure in the RINSC reactor would pass is greater than the maximum height for the University of Virginia fission plate failure. In both cases, there is a very large volume of water in which fission products may be dissolved, so saturation is not credible.

13.23 The response to RAI 13.7 provides a reactivity transient analysis that shows the safety limit on reactor power will be exceeded. The response makes the statement, "However, the safety limit on power does not apply to transients." This statement is inconsistent with the regulations in 10 CFR 50.36, "Technical Specifications," for safety limits and limiting safety system-settings. Propose new limiting safety system settings that prevent the reactor power safety limit from being exceeded. Alternately, propose a different safety limit(s) that prevents the uncontrolled release of radioactive material from the fuel and will not be exceeded during any reactor transient. Provide discussion and analyses that support the proposed TS for all operations allowed by the TS and reactor license.

Include estimates of the safety margins provided by the SL(s) and LSSS(s). (Note: The responses to the RAIs maintain the SLs for reactor power and primary coolant flow, height, and temperature. The revised proposed TS submitted September 29, 2011, contain an SL for fuel cladding temperature only.)

Response to December 17, 2012 RAI Dated March 15, 2013 Technical Specification 2.1 has been changed so that the safety limit is based on fuel cladding temperature rather than power level. Consequently, there is no longer a safety 92

limit on power of 2.4 MW for forced convection mode cooling. The new safety limit on fuel cladding temperature is:

The true value of the reactor fuel cladding shall be less than or equal to 530 C.

Fuel cladding temperature is a function of power level integrated over time. The transient analyses showed that in all cases, the cladding temperature was less than 80 C, which is well below the proposed safety limit.

13.24 The response to RAI 13.11 describes administrative controls that slow draining of the reactor pool in the case of a beam port break. These controls form the bases for assumptions in the pool drainage analysis presented in the response to RAI 10.2.2 Proposed TS 3.9.3.1 states, "Each beam port shall have no more than an area of 1.25 in open to confinement during reactor operation." However, this administrative control does not prevent activities that could increase the cross-sectional area of drainage pathways immediately following reactor operation, and thus invalidate the assumptions used in the drainage analysis. Revise the proposed TS to include requirements that are always consistent with assumptions in the analysis or revise the analysis to be consistent with the proposed TS.

Response to December 17, 2012 RAI Dated March 15, 2013 The LOCA analysis shows that as long as part of the fuel meat remains submerged in water at a level that is no lower than the elevation of the bottom of the eight inch beam ports, and the power fraction is no greater than 0.827%, then there is sufficient cooling capacity to prevent the fuel cladding temperature from reaching the blister point. The amount of time that it would take for the power fraction to drop below the point at which blister temperature of the cladding cannot be reached with a coolant level no lower than the elevation of the bottom of the eight inch beam ports, after infinite reactor operation is 16232 seconds (4.5 hrs). It has been shown that as long as the open area between a beam port and confinement is no greater than 1.25 in2 the drain time will be at least 4.5 hrs.

Consequently there is an administrative control that limits the area open between each beam port and confinement to 1.25 in 2 (See the LOCA Analysis Addendum). The thrust of this question has to do with the fact that there is no administrative control that requires that this area limit be maintained for a period of time after shutdown. If this limit were put into place, experimenters would have a waiting period before they would have access to their experimental samples or equipment. As a result, the staff is interested in finding an alternative mechanism for administratively controlling potential pool leak pathways.

Two administrative controls are proposed:

1. If there is no need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then the 1.25 in2 area opening to confinement shall be maintained until that time period has passed.

93

2. If there is a need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then:

A. The reactor shall be moved to the low power section of the pool where it is at the opposite end of the pool from the beam port extensions.

B. The pool gate shall be positioned so that the high power section of the pool is isolated in such a way that if a beam port extension were sheared off, the pool level in the low power section would not be affected.

Technical Specification 3.93 will be re-written to say:

3.9.3 Experimental Facilities 3.9.3.1 Experimental Facility Configuration During Reactor Operation Applicability:

These specifications apply to the reactor experimental facilities during reactor operation.

Objective:

These specifications ensure that in the event of a Maximum Credible Accident, the rate at which the pool level would decrease would be low enough to make certain that the fuel cladding would not be damaged due to insufficient cooling.

Specifications:

3.9.3.1.1. Each beam port shall have no more than an area of 1.25 in2 open to confinement during reactor operation.

3.9.3.1.2. When the reactor is in operation, the drain valve to the through port shall be closed.

3.9.3.1.3. When the through port is in use, gate valves shall be installed on the end(s) of the port that will be used for access.

3.9.3.1.4. When the through port is not being monitored for a leak condition, the ends of the port shall be co-!o cd.

Bases:

Specification 3.9.3.1.1:

The LOCA analysis shows that as long as the pool level does not drain through an area greater than 1.48 in 2 to confinement, then there will be sufficient time for decay power to drop to a point which will not damage the fuel cladding, provided that the pool level does not drop below the elevation of the bottom of the eight inch beam ports. It also shows that if any single port has a catastrophic failure, the un-damaged ports do not become pool drain pathways.

Consequently, limiting the areas of each experimental port that is open to confinement to 1.25 in2 is conservative.

Specification 3.9.3.1.2:

Shearing the through port is not considered to be a credible accident. Consequently, a leak in the through port is not anticipated to be catastrophic. The through port has three potential pool leak pathways. The first is the through port drain. By keeping this drain closed during operation, that potential leak pathway is blocked, and the potential for an unnoticed pool leak though this experimental facility is prevented.

Specification 3.9.3.1.3:

If the end(s) of the through port that will be used for access have gate valves mounted to them, then in the event of a leak, the port can be easily isolated so that the leak is stopped.

Specification 3.9.3.1.4:

The LOCA analysis has shown that the amount of time available for performing mitigating actions in the event of a non-.catastrophic pool leak is on the order of hours.

Consequently, as long as reactor personnel will become aware of a pool leak though the through port reasonably quickly, and the gate valves are in place, the consequence of the leak can be mitigated quickly by closing the valves.

3.9.3.2 Experimental Facility Configuration Within 4.5 Hours After Shutdown 95

Applicability:

These specifications apply to the reactor experimental facilities for the 4.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period after reactor shutdown.

Objective:

These specifications ensure that in the event of a Maximum Credible Accident, the rate at which the pool level would decrease would be low enough to make certain that the fuel cladding would not be damaged due to insufficient cooling.

Specifications:

3.9.3.2.1. If there is no need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then the 1.25 in2 area opening to confinement shall be maintained until that time period has passed.

3.9.3.2.2. If there is a need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then:

3.9.3.2.2.1. The reactor shall be moved to the low' power section of the pool where it is at the opposite end of the pool from the beam port extensions.

3.9.3.2.2.2. The pool gate shall be positioned so that the high power section of the pool is isolated in such a way that if a beam port extension were sheared off, the pool level in the low power section would not be affected.

Bases:

Specification 3.9.3.2.1 The LOCA analysis shows that if the reactor were operated for an infinite amount of time at 2 MW, the amount of time that it would take for the power fraction to decay after shutdown to a point where the fuel cladding blister temperature could not be reached, even if the pool level were at the elevation of the bottom of the 8 inch beam ports, would be 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The analysis also shows that the maximum area of an opening between a beam port and confinement that limits this drain time to 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is 1.48 96

in2 . Consequently, maintaining the limit on the area open between confinement and the beamn ports to 1.25 in 2 for a period of 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown ensures that in the event of a catastrophic beam port failure, the drain time would provide sufficient time for power to decay to a point below which the fuel could not be damaged.

Specification 3.9.3.2.2 In the event that access to a beam port is needed within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown, a provision is made so that the core can be isolated from the beam port end of the pool. With the core in the low power end of the pool, and the pool gate in place, if a beam port extension were sheared off, and a catastrophic beam port failure were to occur, the coolant level above the core would not be affected.

13.25 RAI 14.117 requested an analysis of the consequences of a failure of an experiment that contains fissionable material. The purpose of the questions was to understand the potential radiological consequences of failure of such an experiment and to determine whether the TS requirements provide reasonable assurance that the radiological consequences would be within the regulatory limits in 10 CFR Part 20. The response to the RAI states that the quantity of fissionable materials used in experiments shall be limited by the reactivity worth of the experiment. Proposed TS 3.8.1.4 is not sufficient to ensure the radiological consequences of an experiment failure will be within the regulatory limits because the reactivity worth of a material depends on many different factors, and not just the quantity of material. Revise the proposed TS to include requirements for experiments that contain fissionable material that ensure failure of the experiment will not result in exceeding the limits in 10 CFR Part 20. The regulation 10 CFR 50.36(b), states that the proposed TS must be "derived from the analyses and evaluation included in the safety analysis report." In accordance with that requirement, provide an analysis of the failure of an experiment that contains fissionable material.

Response to December 17, 2012 RAI Dated March 15, 2013 As noted in RAI question 13.25, "the reactivity worth of a material depends on many different factors, and not just the quantity of material". Likewise, there are many different factors that determine the radiological consequence of experiment failure, which go beyond the quantity of material in the experiment. Consequently, the Proposed Technical Specifications have been written to say:

3.8.1.4. Fissionable Materials

1. The quantity of fissionable materials used in experiments shall not cause the experiment reactivity worth limits to be exceeded.

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2. Fissionable materials shall be doubly encapsulated.

3. Containers for experiments that have fissionable material shall be opened inside confinement.

Failure of experiments that contain fissionable materials have the potential to have an impact on reactor criticality, or on radioactive material release. The consequence of experiment failure on criticality is bounded by limiting the reactivity worths of experiments. The analysis for this is in SAR Chapter 13 as part of the transient analysis. The radioactive material release is bounded by the analysis in SAR Chapter 13 for the Maximum Hypothetical Accident involving a fuel element failure. Double encapsulation of fissionable materials reduces the probability of the release of radioactive material. The requirement that experiments containing fissionable materials be opened inside confinement ensures that in the event of a fission product gas release, the mitigating actions of the confinement system would be available.

3.8.2.1. Experiment design shall be reviewed to ensure that credible failure of any experiment will not result in releases or exposures in excess of limits established in 10 CFR 20.

6.2.3 Review Function The NRSC shall review the following items:

6.2.3.5 New experiments 6.5 Experiments Review and Approval 6.5.1 All new experiments shall be reviewed and approved by the NRSC prior to bringing the reactor to power with the experiment loaded.

6.5.2 Substantive changes to previously approved experiments shall be reviewed and approved by the NRSC prior to bringing the reactor to power with the experiment loaded.

6.5.3 Minor changes that do not significantly alter the experiment may be approved by a Senior Reactor Operator or upper management.

This combination of Technical Specifications ensures that all experiments that contain fissionable material will be reviewed and approved by the NRSC prior to installation, and that as part of that review, analyses will be done to make sure that experiment failure will not be able to result in a radiological release in excess of limits established in 10 CFR 20.

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Section 1.0, "Definitions" 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.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 The Rhode Island Nuclear Science Center Technical Specifications contain numerous referencing errors due to discrepancies between document versions. Below is a table of the sections in which these errors can be found, their page numbers as can. be found in Chapter 14 of the SAR, the current reference, and the corrected referenced section. The Corrected SAR Reference column indicates where the information should have been located. In some cases, these sections of the SAR have been revised. Any information that has been moved or omitted will be addressed in the future.

Corrected SAR Page Reference rrece TS Section Reference 2.1.1 14-12 Part B Sections 4.6-4.8 2.2.1 14-14 Part B Section 13.2.3 2.2.1 14-14 Part A,Section XI Section 13.2.5 Part A,Section IX and 2.2.1 14-14 Part B Section X and Section 13.2.3 Appendix D 3.1 14-17 Part A,Section V Section 4.5 3.2 14-18 Section XI Section 4.5 3.9a 14-310 Part A Section VIII Section 4.2.3 4.1. lb 14-32 Part A,Section V Section 4.5 4.2.6 14-33 Part A.Section V Section 4.5 4.2.7 14-34 Part A,Section V Section 4.5 4.2.8 14-34 Part A.Section V Section 4.5 4.9a 14-39 Part A,Section VIII Section 4.2.3 4.9b 14-40 Part A,Section VI Section 4.5 Figure 4, Revision 1, Figure 4-1, Chapter 5.3 14-41 Section V, Dec. 1992 4 5.5 14-42 Part A,Section XII Section 9.2.3 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.

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Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 This is a general comment. Specifications that contain references to portions of the SAR will be revised accordingly on a case by case basis.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The reference to "certified" operator will be removed. In its place, "Operator" will be defined as: "An individual authorized by the U. S. Nuclear Regulatory Commission to carry out the responsibilities associated with the position requiring the certification".

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.

First Response to RAI Dated April 13. 2010 Submitted June 10, 2010 The reference to "Class A Operator" will be removed because it is not used anywhere in the document.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Senior Reactor Operator" will be changed to "An individual licensed under 10 CFR Part 55 to manipulate the controls of the RINSC reactor and to direct the licensed activities of reactor operators".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The reference to "Class B Operator" will be removed because it is not used anywhere in the document.

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.

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First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of Reactor Operator will be changed to "An individual licensed under 10 CFR Part 55 to manipulate the controls of the RINSC reactor".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 Updated references will be used in the definition. The new definition will be:

"Explosive material is any solid or liquid which is categorized as a severe, dangerous, or very dangerous explosion hazard in Sax's Dangerous Properties Of Industrial Materials by Richard J. Lewis, Sr., 11" Ed. (2004), or is given an identification of Reactivity (Stability) Index of 2, 3, or 4 by the National Fire Protection Association (NFPA) in its publication NFPA 704: Standard System for the Identification of the Hazards of Materials for Emergency Response, 2007 Edition".

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.

First Response to RAI Dated April 13. 2010 Submitted June 10, 2010 The definition of "Measured Channel" will be removed because it is redundant. The definition of "Channel" will be consistent with ANSI/ANS 15. 1:

"1.5 Channel A channel is the combination of sensor, line, amplifier, and output device which are connected for the purpose of measuring the value of a parameter.

1.5.1 Channel Test Channel test is the introduction of a signal into the channel for verification that it is operable.

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1.5.2 Channel Check Channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.

1.5.3 Channel Calibration Channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test."

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Instrumentation Channel" will be removed because it is redundant.

See RAI question 14.9.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 See RAI questions 14.9 and 14.10.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 TS 1.8 "Measured Channel" definition has been removed. See RAI question 14.9.

TS 1.28 is now TS 1.24 and has been revised to say "A safety channel is a channel in the reactor safety system."

TS 3.2 Table 3.1 The term "Safety Channel" is still valid in this section of TS.

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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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The term "Measured Channel" has been, deleted. See RAI question 14.9. Reference to "Measuring Channel" has also been removed.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 TS 1.14 is a generic definition of excess reactivity, rather than a reference to a specific RINSC core value. This is due to the fact that the amount of excess reactivity within the core depends on core condition (temperature, poisons, etc...). This definition will not be changed.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 This definition has been revised to encompass all of the limits placed on reactivity:

"Reactivity limits are those limits placed on the reactivity worths of reactor configurations, components, and experiments".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition in TS 1.14 provides a generic description of how "Reactivity Excess" is defined. The definition in TS 1.15 provides a general description of the types of reactivity limits that are imposed on core configurations, core components, and experiments. The reference to a "Reference Core" has been removed. See RAI question 14.15.

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 103

in equipment position or configuration could affect the reactivity worth of an experiment.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The word "equipment" has been replaced with the word "experiment". The revised definition is:

"The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 "Reactor Secured" is defined as:

"The reactor is secured when the following conditions are met:

a. The reactor is shutdown.

b. The master switch is in the off position and the key is removed from the lock.

c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods.

d. No experiments are being moved or serviced."

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 See the revised definition of "Reactor Secured" in RAI question 14.18.

The definition for "Subcritical" has been given its own TS heading. The definition has been revised to say:

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"There is insufficient fissile material or moderator present in the reactor, control rods or adjacent experiments, to attain criticality under optimum available conditions of moderation and reflection".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 "Removable Experiment" has been deleted from the TS. All RINSC experiments are categorized as either "fixed" experiments or "moveable" experiments. It is assumed that all experiments could be removed from the core, experimental facilities, or facility.

1421 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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 See RAI question 14.20.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Research Reactor" has been deleted because it is not used.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Rundown" has been deleted because it is not used.

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.

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First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 See RAI question 14.23.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Measuring Channel" has been deleted because it is not used. See RAI question 14.9.

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Scram Time" has been modified to be:

"Scram time is the elapsed time between the initiation of a scram signal and the time when the blades are fully inserted in the core".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Shim Safety Blade" has been modified to be:

"A shim safety blade is a control blade fabricated from a neutron absorbing material which is used to compensate for fuel bum-up, temperature, and poison effects. A shim safety blade is magnetically coupled to its drive unit allowing it to fully insert into the core due to gravity when the magnet is de-energized."

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) 106

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Shutdown Margin" will be changed to:

"Shutdown Margin shall mean the minimum amount of negative reactivity inserted into the core when the most reactive control blade and the regulating rod are fully withdrawn, and the remaining control blades are fully inserted into the core".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 This definition has been deleted because it is not used.

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)

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition of "Surveillance Activities" has been changed to:

"Regularly scheduled activities that verify the integrity and operability of facility infrastructure and equipment, and that ensure the safe operation of the reactor".

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.

First Response to RAI Dated April 13, 2010 Submitted June 10, 2010 The definition has been changed to:

"Maximum intervals are to provide operational flexibility, not to reduce frequency.

Established frequencies shall be maintained over the long tenn. Allowable surveillance intervals shall not exceed the following:

1. 5 years (interval not to exceed 6 years).

2. 2 years (interval not to exceed 2 1/2 y'ears).

3. Annual (interval not to exceed 15 months).

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4. Semiannual (interval not to exceed 7 1/2 months).

5. Quarterly (interval not to exceed 4 months).

6. Monthly (interval not to exceed 6 weeks).

7. Weekly (interval not to exceed 10 days).

8. Daily (must be done during the calendar day).

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 According to 10 CFR 50.36(c),(1) Safety limits, limiting safety system settings, and limiting control settings.

(i)(A) 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. If any safety limit is exceeded, the reactor must be shut down.

The SLs of Section 2.1 and the associated LSSSs of Section 2.2 are established at conservative levels that effectively preclude any damage to the fuel cladding, the primary barrier to release of radioactivity, under normal and credible abnormal conditions. As supported by the thermal-hydraulic analysis in the SAR, these settings in conjunction with the Section 3.1 LCO on maximum excess reactivity also preclude damage to the fuel cladding under credible reactivity transients.

There is no CFR requirement to specifically address transients, but the SL must address all operations, not just steady state operation. TS 2.1.1 will be revised as follows.

2.1.1 Safety Limits in the Forced Convection Mode Applicability:

This specification applies to the interrelated variables associated with core thermal and hydraulic performance when operating in forced convection mode.

These variables are:

Reactor Thermal Power, P Reactor Coolant Flow through the Core, m Reactor Coolant Outlet Temperature, To Height of Water above the Top of the Core, H 108

Objective:

To assure that the integrity of the fuel clad is maintained.

Specifications:

1. The true value of reactor power (P) shall not exceed 2.4 MW.

2. The true value of reactor coolant flow (in) shall not be less than 1580 gpm.

3. The true value of the reactor coolant outlet temperature (TO) shall not exceed 125 OF.

4. The true value of water height above the active core (H) shall not be less than 2'3 ft 6.5 in. while the reactor is operating at any power level.

Bases:

The basis for forced convection safety limits is to ensure that the calculated maximum cladding temperature in the hot channel of the core will not be exceeded. Thermal hydraulic analyses show that if the safety limits are not exceeded the integrity of the fuel cladding will be maintained.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the revised TS 2.1.1 in RAI 14.32.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The objective of the natural convection mode safety limits is to assure that the integrity of the fuel cladding is maintained. The second item listed as an objective (P.14-13 Line 14) will be removed.

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) 109

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to 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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 All references to fuel temperature have been removed. Section 2.2.1 is revised as follows.

2.2.1 Limiting Safety System Settings in the Forced Convection Mode Applicability:

These LSSSs apply to the setpoints for the safety channels monitoring reactor power, primary coolant flow, pool level and core outlet temperature.

Objective:

To assure that the integrity of the fuel cladding is maintained in the forced convection mode.

Specifications:

The limiting safety system settings for reactor thermal power (P), primary coolant flow through the core (m), height of water above the top of the core (H),

and reactor coolant outlet temperature (TO) shall be as follows:

Measured Parameter LSSS P 2.1 MW m 1800 gpm H 23 ft 9.6 in To 120 OF Bases:

These specifications were set to prevent coolant temperatures from approaching the value at which damage to fuel cladding could occur (see NUREG-1313, "Safety Evaluation Report related to the Evaluation of Low-Enriched Uranium 110

Silicide-Alumintun Dispersion Fuel for Use in Nonpower Reactors"). Flow and temperature limits were chosen to ensure that the integrity of the cladding is maintained even under transient conditions. The uncertainty in the flow measurement is +/- 3%. The uncertainty in the temperature measurement is +/- 2%.

The uncertainty in the measured power level is +/- 10% (see RAI 4.20 response).

The uncertainty in the measurement of the pool height is estimated to be 0.5 in.

At the limits of the uncertainty bands, there are still margins of 0.1 MW, 160 gpm, 2 OF and 2.6 in. to the SL values for power, flow, temperature and pool height, respectively. The following table summarizes the bases for the LSSS settings.

Measured Measurement Limiting Safety Parameter LSSS Value Uncertainty Trip Value Safety Limit Margin P 2.1 MW +/- 10% (+/- 0.2 2.3 MW 2.4MW 0.1 MW MW*)

m 1800 gpm +/- 3% (+/- 60 gpm) 1740 gpm 1580 gpm 160 gpm H 23 ft 9.6 in. +/- 0.5 in. 23 ft 9.1 in. 23 ft 6.5 in. 2.6 in.

To 120 OF +/- 2% (3 OF*) 123 OF 125 OF 20 F

• Uncertainties in measured values (+/- 0.2 MW, +/- 60 gpm, 2 OF) are based on the nominal operating values of 2 MW, 1950 gpm, and 90 OF to 115 OF for the power, flow and outlet temperature, respectively.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36.

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.

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Second Response to RAI Dated April 13, 2010 Submitted August 6. 2010 See rewritten TS 2.2.1 in response to 14.36.

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)

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 See rewritten TS 2.2.1 in response to 14.36. A summary of the new transient analyses is included in the Bases.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS -2.21 in response to 14.36.

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.

Second Response to RAI Dated April 13, 2010 Submitted Auga-ti 6, 2010 112

See rewritten TS 2.2.1 in response to 14.36.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 See rewritten TS 2.2.1 in response to 14.36. The revised thermal hydraulic analysis for the SAR is consistent with the revised TS.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36. Supporting analyses use the limiting values provided in the table included in the Bases statement.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36.

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.

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Second Response to RAI Dated April 13. 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36. Reference is no longer used for determining the flow and temperature measurement uncertainties, and is not included.

See RAI question 4.20 for flow and temperature measurement uncertainty analysis.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 See rewritten TS 2.2.1 in response to 14.36. References to a previous SAR have been removed.

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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 As indicated in the response to RAI 14.52, the bases for TS 2.2.2 has been revised and no longer states, "the SAR has determined that up to 217 kW can be removed by natural convection." Moreover, the analysis of the thermal behavior of the LEU core during steady-state operation in the natural-convection mode, which was in section 4.6.5 of the 2004 RINSC Reactor SAR and included Table 4-17, has been completely redone and is replaced by section 4.7 of Reference AA.

As stated in the response to RAI 4.28, Reference AA refers to the completely redone analysis of the thermal behavior of the LEU core in theJbrced-convection mode, which is provided in Reference BB.

In the analysis of Reference AA, onset of nucleate boiling is predicted to occur at 369 kW with all uncertainties included. Thus, there is no inconsistency between the bases for TS 2.2.2 and the power at which onset of nucleate boiling is predicted to occur in the analysis of section 4.7 of Reference AA.

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AA. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman and M.

Kalimullah to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Natural-Convective Flow in the Rhode Island Nuclear Science Center (RINSC)

Reactor," November 8, 2010.

BB. Argonne National Laboratory intra-laboratory memo, Earl E. Feldman to James E. Matos, "Steady State Thermal-Hydraulic Analysis for Forced-Convective Flow in the Rhode Island Nuclear Science Center (RINSC) Reactor," September

3. 2010.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 See rewritten TS 2.2.2 in response to 14.52. A summary of the revised thermal/transient analyses is included in the response to this RAI.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The pool level scram will be in reference to pool level elevation. In the new safety analysis, the pool level safety limit has been set at 23 ft 6.5 inches above the top of the reactor core. The LSSS has been set at 23 ft 9.6 inches above the top of the core. The safety margin that is provided by this is shown in the bases section.

Revise Section 2.2.2 as follows.

2.2.2 Limiting Safety System Settings in the Natural Convection Mode Applicability:

These LSSSs apply to the setpoints for the safety channels monitoring reactor power, pool level and pool water temperature.

Objective:

To assure that the integrity of the fuel cladding is maintained in the natural convection mode.

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Specifications:

The limiting safety system settings for reactor thermal power (P),

height of water above the top of the core (H), and reactor pool water temperature (Tp) shall be as follows:

Measured Parameter LSSS P 115 kW H 23 ft 9.6 in.

Tp 125 OF Bases:

These specifications were set to prevent coolant temperatures from approaching the value at which damage to fuel cladding could occur (see NUREG- 1313, "Safety Evaluation Report related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Nonpower Reactors"). Power and temperature limits were chosen to ensure that the integrity of the cladding is maintained even under transient conditions. The uncertainty in the pool temperature measurements is +/- 2%. The uncertainty in the measured power level under natural convection conditions is +L10% (see RAI 4.20 response). The uncertainty in the measurement of the pool height is estimated to be 0.5 in. At the limits of the uncertainty bands, there are still margins of 75 kW. 2 OF and 2.6 in. to the SL values for power, pool temperature and pool height, respectively. The following table summarizes the bases for the LSSS settings.

Measured Measurement Limiting Safety Parameter LSSS Value Uncertainty Trip Value Safety Limit Margin P 115 kW +/- 10%-(+/- 10 kW*) 125 kW 200kW 75kW H 23 ft 9.6 in. +/- 0.5 in. 23 ft 9.1 in. 23 ft 6.5 in. 2.6 in.

Tp 125 OF +/- 2% (3 OF*) 128 OF 130 OF 2 OF

• Uncertainties in measured values (+/- 10 kW, 3 OF) are based on the nominal operating values of 100 kW and 108 OF for the power and pool temperature, respectively.

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 116

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the response to RAI Question 4.20 for the power and coolant height measurement uncertainty estimate.

See rewritten TS 2.2.2 in response to 14.52 for how they were treated in the analysis supporting the LSSS.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the response to RAI Question 4.20 for the power and coolant height measurement uncertainty estimate.

See rewritten TS 2.2.2 in response to 14.52 for how they were treated in the analysis supporting the LSSS.

14.55 ANSIANS-15.1 recommends technical specifications establish limits on fuel bumup.

Explain the reason for not including such a specification, and revise the proposed TS as appropriate.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The type of fuel used at RINSC has been qualified to 98% bum-up. Consequently, no limit on fuel burn-up is necessary. The reference for this is NUREG 1313.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The definition of "Shutdown Margin" will be changed to:

"Shutdown Margin shall mean the minimum amount of negative reactivity inserted into the core when the most reactive control blade and the regulating rod are fully withdrawn, and the remaining control blades are fully inserted into the core".

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The basis for TS 3.1.1 (P.14-17 Line 6) will be changed to:

Specification 3.1.1 assures that the reactor can be shutdown from any operating condition and will remain subcritical after cool down and xenon decay even if the blade of the highest reactivity worth and the regulating blade are in the fully withdrawn position.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 This reference has to do with predictions that were made about what the shutdown margin would be, prior to when the LEU core was configured, and the shutdown margin was measured. Since this core has been in operation for more than fifteen years, and the shutdown margin for it has been measured at least annually, this reference is no longer relevant. Consequently it will be removed.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 In the Bases section for LCO 3.1.3, rewrite the paragraph for Specification 3.1.3 as follows.

Specification 3.1.3 limits the reactivity worth of experiments to values of reactivity which, if introduced as positive step changes, would preclude violating any Safety Limit. Transient analysis demonstrates that this LCO on reactivity for experiments results in no challenge to fuel integrity under credible postulated transients.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The reference to "Removable" experiments has been deleted. See the answer to RAI question 14.20.

14.60 TS 3.1.4 limits the reactivity worth of each movable experiment to 0.08 %Ak/k. Section 13.2.2 of the SAR appears to state that the total reactivity worth of all movable experiments 118

is limited to 0.08 %zAk/k. Explain whether each movable experiment is limited to 0.08

%Ak/k, or whether the total reactivity worth of all movable experiments is limited to 0.08 %Ak/k. If the reactivity worth of each movable experiment is limited to 0.08

%Ak/k, explain whether multiple movable experiments could comprise the total experiment reactivity worth limit of 0.6 %Ak/k (e.g., ten movable experiments each with a reactivity worth of 0.06 %Ak/k).

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 TS 3,1.3 limits the total reactivity worth of all experiments in the core to 0.6% dK/K.

TS 3.1.4 limits the reactivity worth of any individual moveable experiment to be 0.08%

dK/K, and any fixed experiment to be 0.6% dK/K.

An additional limit will be added to clarify that the maximum total reactivity worth of all moveable experiments in the core is 0.08%.

Rewrite these Technical Specifications as follows:

3.1.3 The total reactivity worth of experiments shall not exceed:

Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 3.1.4 The maximum reactivity worth of any individual experiment shall not exceed:

Fixed 0.6 % dK.K Moveable 0.08 % dK/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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 It is assumed that fixed experiments will not produce a reactor period because they are fixed. The total reactivity of fixed experiments is limited to 0.6 % dK/K in order to assure that if a failure occurred in which the experiment reactivity was inserted into the reactor, there would be insufficient reactivity to produce a prompt critical condition. The prompt critical condition occurs when p = P3. For the RINSC U-235 fuelled core, 03=

0.0065. Consequently, if the reactivity insertion is p = 0.6 % dK/K,. it will be less than the 0.65 % dK/K necessary to cause prompt criticality. The reference for this may be found in Glasstone, Sanmual; Sesonske, Alexander (1980), Nuclear Reactor Engineering (3rd ed.), Van Nostrand Reinhold.

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The reactivity of moveable experiments is limited to 0.08 %dK/K in order to assure that they will not produce a stable period less than 30 seconds, and to assure that the reactivity can be compensated for by the action of the control and safety systems without exceeding any safety limits. The following diagram from Lamarsh; John R., Baratta, Anthony J.,

(2001), Introdcuction to ANuclear Reactor Engineering(3rd ed.), Prentice Hall, shows that for a U-235 fueled core, with a positive reactivity insertion of 0.08 %dK/K = 0.0008 = 8 E 1 0 -4, the period would be approximately 80 seconds, which is easily compensated for by the action of the control and safety systems.

1000 100 1 11il 10 9

0 0

10-ls c 10- 2 D " 190-

-

10_110)-4 rrrJH 3 lit 10-2 I INIlor, 10-Reactivity (absolute value) 4 . 1 , Is . 1 2 46 Figure 7.2 Reactor period as a function of positive and negative Reactivity, dollars reactivity for a 235U-fueled reactor.

This was confirmed by an analysis performed by Argonne National Laboratory. In that analysis, the initial assumptions were that:

1. Reactor Power was at 10 Watts

2. Coolant Inlet Temperature was 123 F

3. Coolant Flow Rate was 1740 gpm

4. Height of the Coolant above the Fuel Meat was 23 ft 9.1 in 120

5. Water Pressure at the Top of the Fuel Meat was 1.715 X 105 Pa A reactivity insertion of +0.08% dK/K was added over a time span of 0.1 seconds. The following sequence of events were predicted to occur:

1. At t = 0.0 seconds the reactivity insertion begins, and reactor power begins to increase.

2. At t = 0.1 seconds the reactivity insertion ends, and reactor power continues to increase.

3. At t = 30 seconds a stable period of about 75 seconds is reached.

4. At t = 1166.6 seconds power reaches 2.3 MW, and the over power trip is actuated.

At this point, feedback reactivity from Doppler, water expansion, and voids cause the period to decrease to approximately 1375 seconds, effectively making power constant at 2.3 MW.

5. The model assumed that there would be a 100 msec delay between the time that the trip was actuated, and the time that negative reactivity from the control system would begin to be inserted. Consequently, at t = 1166.7 seconds, the reactor power is still approximately 2.3 MW, but negative reactivity begins to be inserted.

6. Reactor power drops rapidly to shutdown conditions.

At t = 1167 seconds, peak temperatures are estimated to be:

1. Fuel Meat Centerline is 81.4 C

2. Clad Surface is 80.6 C

3. Coolant is 64.9 C These temperatures are well below the maximum fuel cladding temperature limit of 530 C suggested in NUREG 1313.

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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the analysis of a rapid insertion of 0.6 %Ak/k from very low power in RAI 13.7. A rapid insertion of 0.08 %Ak/k from very low power for the moveable experiments is bounded by the 0.6 % Ak/k insertion case.

14.63 TS 3.1.5 requires the reactor to be subcritical by at least 3.0 %Ak/k during fuel loading changes. Explain how it is determined that the reactor is subcritical by at least 3.0 121

%Ak/k during fuel loading changes. Explain the reason for not specifying a surveillance requirement for this LCO, and revise the proposed TS as appropriate.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 RINSC is currently operating with its equilibrium core. The minimum shutdown reactivity for this core occurs just after re-fueling operations, in which four irradiated fuel elements are replaced with four fresh fuel elements. This operation was performed in October 2008. The data for the new core configuration with the fresh fuel indicated that the shutdown reactivity was -7.07% dK/K (See the reference entitled "Core Change Summary from RINSC Core LEU #3 to LEU #4"). As operation of the reactor continues, the shutdown reactivity will become more subcritical as fuel burn-up occurs.

TS 3.1.3 limits the total worth of all experiments to 0.6% dK/K. Therefore, if re-fuelling has just occurred, and an experiment worth +0.6% dK/K has been added, the shutdown reactivity would be approximately:

-7% dK/K + 0.6% dK/K = -6.4% dK/K Consequently, it is not anticipated that the reactor will ever be subcritical by less than 3%

dK/K during fuel loading operations.

Add the following surveillance item:

4.1.1.4 Prior to fuel loading changes, core reactivity shall be verified to be shutdown by a minimum of 3 %dK/K by using existing core data, or by making new core reactivity measurements.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 TS 4.1.1 will be modified to say (P14-32 Line 26):

Shim safety blade and regulating rod reactivities and insertion rates will be measured:

a. Annually

b. Whenever the core configuration is changed to an uncharacterized core 122

The reference to a previous SAR will be removed. This reference has to do with predictions that were made about core characteristics prior to when the LEU core was configured and tested.

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 %Ak/k by flooding would ever be allowed in or adjacent to the core, and revise the proposed TS as appropriate. (See RAI 4.14)

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 Technical Specification 1.16 takes into consideration credible malfunction in the definition of the reactivity worth of experiments. See the answer to RAI question 14.17.

Technical Specification 3.1.7 makes clear that flooding is a credible malfunction.

As discussed in the answer to RAI question 14.137, in order to determine the reactivity worth of a new experiment for which there is no data based on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core. In that case, it is possible that an experiment could be found to have enough positive reactivity that if additional positive reactivity were added due to flooding, the shutdown margin would be less than 1.0 %

dK!K. In that event, Technical Specification 3.1.7 requires that the experiment be removed immediately.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 This LCO has to do with the fact that the temperature coefficient must be negative. The statement "surveillance will be conducted at initial startup and change in fuel type" was meant to indicate that the temperature coefficient would be verified to be negative at initial start-up, and if there was a change in fuel type. This was verified during the initial startup with the LEU fuel. Any change in fuel type would require a change in the license.

Consequently, this surveillance is no longer necessary. As a result, it will be removed.

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) 123

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Technical Specification 3.1.9 requires that all of the core grid positions be filled with fuel elements, experiments or experiment baskets, or reflector elements during operation in forced convection mode. Under conditions of forced coolant flow, the cooling water will obviously follow the path of least resistance. If any grid position is open, some of the cooling water that would normally be forced between the fuel plates will instead go through the open grid position. reducing expected cooling to the fuel. In the natural convection mode, there is no driving force to preferentially redirect coolant flow through the open grid position. Coolant circulation depends on the temperature differences induced by heat transfer from the fuel to the adjacent water. Since there would be no fuel in the open grid position, there would be no heat transfer directly to the water in that position. In this mode, the open grid position actually provides a larger sink for heat generated in elements adjacent to the open channel.

Technical Specification 4.1.1 .b requires that the shim safety blade reactivity worths and insertion rates are measured whenever the core is changed from the start-up core to "the three other cores as analyzed and specified in SAR Part A,Section V". In that analysis.

the RINSC LEU core was initially configured with a start-up core that was reflected with graphite next to the fuel, surrounded by beryllium reflector elements:

LEU Core #1 Startup Core Configuration 124

As bum-up occurred, the core configuration was altered to a more efficient neutron reflection configuration in which some of the beryllium elements were moved next to the fuel, and the corresponding graphite elements were to the outer edge of the element grid. This was core LEU 2:

LEU Core #2 Configuration As further burn-up occurred, the configuration was altered to the most efficient neutron reflection configuration in which all of the beryllium elements were moved to positions next to the fuel. This was core LEU 3:

125

LEU Core #3 Configuration - 2 May 03 From this point forward, bum-up is offset by perfonning a fuel element change in which four elements from the center of the core are removed, the remaining fuel is shuffled inward in such a way that only the four corners of the fuelled part of the core are vacant, and new fuel is place in the comers. This core is the equilibrium core:

126

LEU Core #4 Configuration - 24 October 2008 Thus, the "three other cores" that were analyzed other than the start-up core are LEU 2, LEU 3, and the equilibrium cores. The reactor is currently operating with the equilibrium core.

TS 4.1.1 .b will be rewritten to indicate that blade worths shall be measured when any new core is installed in the reactor.

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. .d. Revise the proposed TS as appropriate.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 1.20 defines the reactor as "shutdowf *when it is subcritical by at least the shutdowvn margin with the reactivity of all installed experiments included.

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The term "Reactor Secured" was defined as part of the answer to RAI question 14.18 to be:

"The reactor is secured when the following conditions are met:

a. The reactor is shutdown.

b. The master switch is in the off position and the key is removed from the lock.

c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods.

d. No experiments are being moved or serviced."

There is no Technical Specification requirement for safety systems and safety related instrumentation that must be operable when the reactor is subcritical but not secured because it is impossible to do a pre-start checkout to verify the operability of the safety related instrumentation without the reactor being in a non-secured state. Condition b cannot be met because the master switch cannot be in the off position with the key removed in order to perform the pre-start checkout.

Radiation monitors are required for work of the types specified in TS 1.19.1.c and TS 1.19.1 .d. Technical Specification 1.17 defines the reactor to be in operation whenever it is not secured or shutdown. The answer to RAI question 7.4 provides the list of the radiation monitoring instrumentation that is required to be in operation whenever the reactor is in operation (e.g., movement of fuel in the reactor core). It is possible to verify that these instruments are operable prior to taking the reactor into an unsecured state.

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 T 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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The table is being revised to make the set points more clear. See the revised table in RAI Question 7.1.

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 128

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?

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 This was intended to communicate that there are a minimum of two over power trips that have scram set points that trip when power is at a maximum of 115% on any range. Consequently, if the power range on one of these instruments is set at 2 W, and power goes above 2.3 W (115% of 2 W), a scram will occur. The "2.3 MW max" was intended to communicate that the maximum available range for these instruments is the 2 MW scale. For natural convection mode cooling, coolant flow, and inlet and outlet temperature alarms and scrams are bypassed. When these are bypassed, the bypass switch sets the over power scram to 115% of 100 kW. Table 3.1 has been revised to say that the over power scram in both cooling modes will trip by 115% of licensed power, which historically has been limited to 100 kW for natural convection mode cooling, and 2 MW for forced convection mode cooling. See the answer to RAI question 7.1.

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)

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The bridge misalignment and bridge movement safety channels each utilize separate limit switches. The bridge movement switch is located adjacent to one of the gears used to move the bridge, with the lever arm of the switch on top of a gear tooth when correctly positioned, depressing the switch. As the bridge moves away from the high powered section the roller-wheel of the arm falls into the valley of a gear, releasing the switch.

The bridge misalignment switch is attached to the end of the track at the high power section. When the bridge is moved away from the high power section the switch is released and the bridge misalignment scram is triggered. This channel is not necessary or functional during natural convection cooled operation.

Since these are both limit switches which are simply used as state / change state indicators, no set points have ever been established.

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 129

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The pool level safety limit and limiting safety system setting are now defined in terms of height above the top of the core. In the new safety analysis, the pool level safety limit has been set at 23 ft 6.5 inches above the top of the reactor core. The LSSS has been set at 23 ft 9.6 inches above the top of the core. Table 3.1 has been modified.

See the answer to RAI question 7.1. TS 2.2.1 was modified to remove this inconsistency as part of the answer to RAI question 14.36.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 This was intended to communicate that there are three channels that use high voltage detectors, and that if the high voltage decreased on any of these channels by more than 50V, a scram would occur. Table 3.1 has been revised to make this more clear. See RAI answer 7.1.

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)

Second Response to RAI Dated April 13, 2010 Submitted August 6. 2010 The no-flow Reactor Safety System Component/Channel entry in Table 3.1 is labeled incorrectly. The safety channel does not apply to the heat generated in the graphite, but to the heat generated in the gamma shield at the front of the thermal column. The flow refers to the gamma shield water coolant which is taken off the primary coolant circuit.

The piping and instrumentation diagram (PID) on page 21 of the 1962 Safeguards Report [B. J. Tharpe, Safeguards Report for Rhode Island Open Pool Reactor, General Electric Document APED-3872, April 4, 1962] and shows the interconnection of the gamma shield cooling to the primary coolant loop. This figure is the same as Reference Drawing 762D 192 in the reactor operating manual [Operation and Maintenance Manual, One-Megawatt Open Pool Reactor for Rhode Island Atomic Energy Commission, Providence, R.I., General Electric Document GEI-77793, October 1962]. The 1992 Safety Analysis Report [Safety Analysis Report for the Low Enriched Fuel Conversion of the Rhode Island Nuclear Science Center Research Reactor, Change 1 dated January 13, 1993] states that the "thermal shield is cooled by water which is currently forced 130

around the shield using the pressure difference between the inlet and outlet primary coolant lines."

No flow rate is specified for the gamnma shield because primary coolant flow rate is monitored. As long as the minimum primary flow rate is maintained, there is sufficient flow through the gamma shield. Additionally, there is a No Flow Thermal Column Flow Scram that serves as an auxiliary check that there is coolant flow through the gamma shield. The facility has a 43 year history of operating experience that shows that this coolant system is sufficient.

See also the response to 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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 FC stands for "Forced Convection" which represents the operating mode for which the inlet and outlet coolant temperature alarms are required. Tables 3.1 and 3.2 has been combined to indicate all of the safety channels and non-radiation monitoring safety related instrumentation that is required. The table has been revised to make the operating mode for which the channels and instruments are required, and the trip set points more clear. See the table in RAI Answer 7.1.

14.76 TS 3.2.1, Table 3.2 requires a log count rate blade withdrawal interlock with a set pointless 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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The purpose of this interlock is to ensure that this channel is functioning and detecting neutrons. Historically, a minimum count rate of 3 cps has been acceptable to indicate that this instrument is functional. This table will be updated to make this clear.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The table will be revised to make it clear that there are two servo control interlocks in place. The first interlock prevents the operator from putting the rod control system into servo control if the Log N period is less than 30 seconds. The second interlock prevents the operator from putting the system into servo control if the regulating rod is not fully withdrawn (full out).

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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)

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The building air Stack Monitor consists of two monitors housed in one unit. One monitor is a gaseous monitor, and the other is a particulate monitor. The two channels are entirely independent of each other.

All of the radiation monitors in the confinement room have set points that are in terms of "normal" radiation levels. The purpose of defining set points in terms of "normal" radiation levels is to account for the fact that the radiation levels vary in the confinement room, depending on what kinds of experiments are being performed.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The primary demineralizer is the demineralizer that is used to clean up the primary pool water, as opposed to the make-tip demineralizer. Since the reactor pool water has a small amount of Na-24 in it, some of the sodium accuumulates in the demineralizer, making it radioactively "hot". The term "hot DI" has been used to refer to this demineralizer for the last fifty years. None of the current RINSC staff has knowledge about the origin of this term, but it is surmised that this term came about because this demineralizer has a tendency to be radioactively "hot", and it is a demineralizer (DI).

No revision to the Technical Specifications is necessary.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 This question was addressed on September 22, 1995 when NRC approved Amendment Number 20 to the R-95 License. A copy of this amendment has been enclosed.

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In the NRC Safety Evaluation supporting that amendment, "Continuous" operation was defined as operation for more than one 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> shift. The justification provided for allowing operation up to one 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> shift, was that:

The purpose of the Stack Gaseous and Stack Particulate Monitors is to provide an alarm function to inform operations personnel of potential radiological releases from the stack.

There are alternative radiation monitors with alarms that would be able to indicate a potential radiological release.

As long as there is at least one monitor over the reactor pool and one monitor on the experimental level that would ensure that radiological releases would be detected and alarmed, NRC deemed that this would acceptably meet the monitoring requirements.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 This specification requires the shim safety blades to be operable in accordance with TS 4.1.1 and 4.1.2. TS 4.1.1 defines when reactivity worths and insertion rates shall be measured. As part of the answer to RAI Question 14.64, these parameters are also required to be measured for the regulating rod as well. TS 4.1.2 defines when visual inspections of the shim safety blades are required to be performed. It is not possible to do visual inspections of the regulating blade because it is housed in a shroud. Consequently, in order to include the regulating blade in this specification to .the extent possible, the following additional specification will be added:

3.2.5 The regulating rod is operable in accordance with Technical Specification 4.1.1.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 3.2 has bee re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87. The new proposed shim safety LCO specifications are covered in Technical Specifications 3.2.1.1 and 3.2.1.2. The corresponding surveillance requirements are covered in Technical Specifications 4.2.1 and 4.2.2, which were submitted as part of the answer to RAI question 14.141.

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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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 3.2 has bee re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87. The original TS 3.2.3 referred to the LCO regarding shim safety drop times. This is now covered in Specification 3.2.1.1.

The original references to TS 4.2.5 and (incorrectly) 4.2.6 had to do with the surveillance requirement for shim safety drop times. These reference have been removed, though the surveillances are included as Specifications 4.2.1.1 an 4.2.1.2 in the revised version of Technical Specification 4.2 submitted as part of the answer to RAI question 14.141.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 TS 3.2.4 will be moved to TS 3.1.11.

14.85 TS 3.2.4 specifies a maximum reactivity insertion rate for a single control or regulating blade of 0.02 %Ak/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% Ak/k per second. Section 13.2.5 provides an analysis of a startup accident, but the analyzed reactivity addition rate (0.0196% Ak/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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Analysis of a reactivity insertion of 0.02 %Ak/k per second is provided in the response to RAI 13.7.

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.

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Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 The 4 second period limit serves as an auxiliary protection to assure that the reactor fuel would not be damaged in the event that there was a power transient.

As part of the answer to RAI question 13.7. an analysis was performed for a rapid insertion of 0.6 % dK/K reactivity from very low power. Effectively in this analysis, a step insertion of 0.6 % dK/K reactivity is inserted at low power and the power increases until the true power reaches the limiting safety system setting of 2.3 MW, at which point one of the over power trips cause a scram. It is assumed that it takes 100 ms for the control blades to start dropping into the core, and that it takes 1 second for full insertion.

The analysis shows that the peak fuel temperature is well below the temperature required to damage the fuel.

An insertion of 0.6 % dK/K corresponds to a period of less than 1 second.

Consequently, the consequences of a power excursion due to a 4 second period is covered by this analysis.

See the answer to RAI question 14.87 for the new basis given for the 4 second period scram.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical specification 3.2 has been re-written to conform more closely to ANSI 15.1.

Some of the specifications that had been in section 3.2 have been moved. The following table provides a summary of how things have been changed:

Original Specification New Location Location 3.2.1 Minimum Safety Instrumentation 3.2.1.3 3.2.1.4 3.2.1.5 3.2.2 Operability of Shim Safety Blades 3.2.1.1 3.2.3 Scram Time 3.2.1.1 3.2.4 Reactivity Insertion Rate 3.2.1.2 The radiation monitoring instrumentation described in the new RINSC Technical Specification 3.2.1.3 was taken from the description given as part of the answer to RAI question 7.4. References to specific radiation monitoring instrumentation have been removed in order to allow for more flexibility in using alternative monitoring equipment. References to specific radiation alarm setpoints have been removed.

RINSC has a radiation safety program, which has safety committee oversight to ensure 135

that ALARA principles are met. Radiation levels inside the reactor room are contingent on the number, and types of experiments that are in progress. Rather than defining setpoints with the caveat that they can be adjusted higher with the approval of the approval of the facility Director or Assistant Director, setpoints will be set in a manner that ensures that the goals of the Radiation Safety Program are met. Table 3.2 will be replaced with Specifications 3.2.1.3 and 3.2.1.4.

The reactor safety and safety related instrumentation described in the new RINSC Technical Specification 3.2.1.5 was taken from the description given as part of the answer to RAI question 7.1.

The bases for 3.2.1.1 and 3.2.1.2 refer to transient analyses that were part of the answer to RAI question 13.7.

The basis for Specification 3.2.1.4 is consistent with the answer given for RAI question 14.80 regarding the justification for being able to operate for six hours without the stack gaseous or particulate monitor.

The bases for Specification 3.2.1.5 regarding the safety limits, limiting trip values, and limiting safety system settings are consistent with the answer given for RAI question 14.36, except that the cooling modes for which the pool temperature, and primary coolant flow rate channels are required have been corrected. The basis regarding the inlet temperature channel is consistent with the answer given for RAI question 4.23.

The basis regarding the outlet temperature channel is consistent with the answer given for RAI question 14.36. The basis regarding the pool temperature channel refers, to the basis for Specification 2.2.2, which was updated as part of the answer to RAI question 14.52.

The new versions of Technical Specification 3.2 is:

3.2 Reactor Safety System Applicability:

This specification applies to the reactor safety system and safety related instrumentation required for critical operation of the reactor.

Objective:

The objective of this specification is to define the minimum set of safety system and safety related channels that must be operable in order for the reactor to be made critical.

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Specification:

3.2.1 The reactor shall not be made critical unless:

3.2.1.1 All shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated.

3.2.1.2 The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02% dK/K per second.

3.2.1.3 The following area radiation monitoring instrumentation is operable:

3.2.1.3.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels shall be at the experimental level.

3.2.1.3.2 A minimum of ohe radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.

3.2.1.3.3 If either of these detectors fail during operation, the staff shall have one hour to either repair the detector, or find an acceptable replacement without having to shut the reactor down.

3.2.1.4 The following air radiation monitoring instrumentation is operable:

3.2.1.4.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous effluent shall be operating.

3.2.1.4.2 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement particulate effluent shall be operating.

3.2.1.4.3 If either of these detectors fail during operation, the staff shall have six hours to either repair the detector, or find an acceptable 137

replacement without having to shut the reactor down.

3.2.1.5 The following reactor safety and safety related instrumentation is operable and capable of performing its intended function:

Protection Cooling Channels Function Set Point Mode Required Over Power Both 2 Scram by Power Level Less than or 105% of Equal to Licensed Power Low Pool Level Both 1 Scram by Pool Level Less than or 23 ft 9.6 Drop Equal to in Primary Coolant Inlet Forced 1 Alarm by Inlet Temp Less than or 111 F Temperature Equal to Primary Coolant Outlet Forced 1 Alarm by Outlet Temp Less than or 117 F Temperature Equal to Forced 1 Scram by Outlet Temp Less than or 120 F Equal to Pool Temperature Natural 1 Scram by Pool Temp Less than or 125 F Equal to Primary Coolant Flow Primary Flow Less than or 1800 Rate Forced 1 Scram by Rate Equal to gpm Rate of Change of Less than or 4 Power Both 1 Scram by Period Equal to seconds Seismic Disturbance Both 1 Scram if Seismic Disturbance Detected Bridge Low Power Position Forced 1 Scram if Bridge Not Seated at HP End Bridge Movement Both 1 Scram if Bridge Movement Detected Coolant Gates Open Forced 1 Scram if Inlet Gate Open Forced 1 Scram if Outlet Gate Open Detector HV Less than or Detector HV Failure Both 1 Scram if Decrease Equal to 50V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V No Flow Thermal Column Forced 1 Scram by No Flow Detected Manual Scram Both 1 Scram by Button Depressed Both I Scram by Button Depressed No Automatic Servo Control Interlock Both 1 Servo if Regulating Blade not Full Out 138

Protection Cooling Channels Function Set Point Mode Required No Automatic 30 Both 1 Servo if Period Less than seconds Shim Safety No SS Withdrawal Both 1 Withdrawal if Count Rate Less than 3 cps No SS Both 1 Withdrawal if Test / Select SW not Off Rod Control Loss of Less than or 10 Communication Both 1 Scram if Communication Equal to seconds Basis:

Specification 3.2.1.1 requires that all shim safety blades be capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated. As part of the Safety Analysis, Argonne National Laboratory analyzed a variety of power transients in which it was assumed that the time between the initiation of a scram signal, and full insertion of all of the shim safety rods was one second.

The analysis showed that if the reactor is operated within the safety limits, this time delay will not cause an over power excursion to damage the fuel.

Specification 3.2.1.2 requires that the. reactivity insertion rates of individual shim safety and regulating rods do not exceed 0.02% dK/K per second. As part of the Safety Analysis, Argonne National Laboratory analyzed. ramp insertions of 0.02% dK/K reactivity from a variety of initial power levels. The reactivity insertions are stopped by the over power trip. In all cases, peak fuel and cladding temperatures due to the power overshoot are well below the temperatures required to damage the fuel or cladding. Consequently, this limit ensures that an over power condition due to a reactivity insertion from raising a control rod will not damage the fuel or cladding.

Specification 3.2.1.3 identifies the area radiation monitoring instrumentation that is required to be operable when the reactor is operated. Radiation monitors that are capable of warning personnel of high radiation levels at the experimental elevation, and over the pool serve to ensure that personnel inside the reactor room are made awaie when dose rates are higher than anticipated. Additionally, these monitor alarms provide an indication of a potential fuel failure. In the event of a failure of either of these monitors, the operations staff is afforded the opportunity to rely on alternative monitoring instrumentation without having to shut the reactor down. This configuration has been in use for the life of the facility, without any indication that it is insufficient.

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Specification 3.2.1.4 identifies the air radiation monitoring instrumentation that is required to be operable when the reactor is operated. Radiation monitors that are capable of warning personnel of high gaseous and particulate airborne radioactive material levels ensure that personnel are made aware of potential radiological releases from the stack. In the event of a failure of either of these monitors, the operations staff is afforded the opportunity to rely on alternative monitoring instrumentation without having to shut the reactor down. This configuration has been in use for the life of the facility, without any indication that it is insufficient.

Specification 3.2.1.5 identifies the safety and safety related instrumentation that is required to be operable when the reactor is operated.

Two independent power level channels are required for both forced and natural convection cooling modes of operation, each of which must be capable of scramming the reactor by 105% licensed power.

The basis section of Specification 2.2.1 shows that this ensures that the power level safety limit of 2.4 MW will not be exceeded. Having two independent power level channels ensures that at least one over power protection will be available in the event of an over power excursion.

One low pool level channel is required for both forced and natural convection cooling modes of operation. This channel ensures that the reactor will not be in operation if the pool level is below the safety limit of 23 ft 6.5 inches above the top of the core.

One primary inlet coolant temperature channel is required for forced convection cooling mode operation. This channel alerts the operator in the event that the inlet temperature reaches 111 F. The steady state thermal hydraulic analysis that was done by Argonne National Laboratory for forced convection flow predicts that the inlet temperature would be 115 F for operation at 2.4 MW, with a primary flow of 1580 gpm and an outlet temperature of 125 F.

One primary outlet temperature channel is required for forced convection cooling mode operation. This channel is capable of scramming the reactor when the temperature reaches 120 F. The basis section of Specification 2.2.1 shows that this ensures that the coolant outlet temperature safety limit of 125 F will not be exceeded.

One pool temperature channel is required for natural convection cooling mode of operation. This channel is capable of scramming the 140

reactor when the temperature reaches 125 F. The basis section of Specification 2.2.2 shows that this ensures that the pool temperature safety limit of 130 F will not be exceeded. This channel provides the over temperature protection when the reactor is operated in the natural convection cooling mode.

One primary coolant flow rate channel is required for forced convection cooling mode operation. This channel assures that the reactor will not be operated at power levels above 100 kW with a primary coolant flow rate that is less than the safety limit of 1580 gpm. The basis section of Specification 2.2.1 shows that if this channel is set to scram at a limiting safety system setting of 1800 gpm., the safety limit will not be exceeded.

One rate of change of power channel is required for both cooling modes of operation. The 4 second period limit serves as an auxiliary protection to assure that the reactor fuel would not be damaged in the event that there was a power transient. As part of the Safety Analysis. Argonne National Laboratory analyzed a power excursion involving a period of less than 1 second, which was stopped by an over power scram when the true power reached the limiting safety system setting of 2.3 MW. The analysis showed.that peak fuel temperatures stayed well below the temperature required to damage the fuel. A 4 second period limit provides an additional layer of protection against this type of transient.

One seismic disturbance scram is required for both modes of operation. In the event of a seismic disturbance. the shim safety blade magnets would be likely to drop the blades due to the vibration caused by the disturbance. However, this scram ensures that the blades will be dropped in the event of a disturbance.

One bridge low power position scram is required for forced convection cooling mode operation. In order for the forced convection cooling system to work, the reactor must be seated against the high power section pool wall. This scram ensures that the reactor is properly positioned in the pool so that the coolant ducts are properly coupled with the cooling system piping.

One bridge movement scram is required for both modes of operation.

This scranm assures that the reactor will be shut down in the event that the bridge moves during operation.

One coolant gate open scram on each coolant duct is required during forced convection cooling mode operation. These scrams ensure that 141

coolant flow through the inlet and outlet ducts are not bypassed during forced convection cooling.

One detector HV failure scram is required for each of the power channels, and the period channel. These channels rely on detectors that require high voltage in order to be operable. These scrams assure that the reactor will not be operated when one of these detectors does not have proper high voltage.

One no flow thermal column scram is required during forced convection cooling mode operation. This scram ensures that there is coolant flow through the thermal column gamma shield during operations above 100 kW.

Two manual scram buttons are required to be operational during both modes of operation. One manual scram button is located in the control room, which provides the operator with a mechanism for manually scramming the reactor. The second scram button is on the reactor bridge, which provides anyone directly over the core with a mechanism for scramming the reactor if there were a reason to do so.

One servo control interlock that prevents the regulating blade from being put into automatic servo mode unless the blade is fully withdrawn is required for both modes of operation. As a result of this interlock, when the regulating blade is transferred to automatic servo control, the blade is unable to insert additional reactivity into the core.

One servo control interlock that prevents the regulating blade from being put into automatic mode if the period is less than 30 seconds is required for both modes of operation. This interlock limits the power overshoot that occurs when the regulating blade is put into automatic mode.

One shim safety interlock that prevents shim safety withdrawal if the start up neutron count rate is less than 3 cps is required for both modes of operation. This interlock ensures that the start up channel, which is the most sensitive indication of subcritical multiplication, is operational during reactor start-ups.

One shim safety interlock that prevents shim safety withdrawal if the neutron flux monitor test / select switch is not in the off position is required for both modes of operation. This interlock prevents shim safety withdrawal when this instrument is receiving test signals rather than actual signals from the detector that is part of the neutron flux monitor channel.

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One rod control communication scram is required for both modes of operation. The control rod drive system has a communication link between the digital display in the control room, and the stepper motor controllers out at the pool top. There is a watchdog feature that verifies that this communication link is not broken. In the event that the link is broken, a scram will occur within ten seconds of the break.

All of the scram signals are sent independently of this link. The transient analysis performed by Argonne National Laboratory shows that if the control rod drive communication were lost while the reactor were on a period, the over power, and period trips would prevent the power from reaching a level that could damage the fuel cladding.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Teclmical Specification 3.2 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87 for the bases to these specifications.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Techmical Specification 3.2 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87 for the revised bases to these specifications.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The primary radioactive contaminants of the coolant water are tritium and sodium-24 during normal. operation. The presence of other radioactive materials would be an indication of either an incipient fuel leak or a problem with an experiment. Specification 3.3.a.3 should be changed to read: "Except for tritium and sodium-24, the radioactivity in the primary coolant shall be maintained at levels statistically indistinguishable from background."

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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.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 The specification refers to secondary coolant water. There is no need for a technical specification dealing with either chlorides or resistivity in the secondary coolant water.

SAR Section 5.1, "Summary Description," starting at line 7 states: "The RINSC reactor is an open pool type reactor that uses de-mineralized water for primary coolant, shielding, and reactor moderator; and city water for secondary coolant. SAR Section 5.3.2, "Secondary Coolant System Operation," states: "City water is used as secondary coolant for both loops." SAR Section 5.5.2, "Secondary Makeup Water System," states "City water supplies the makeup water to the secondary coolant system." Starting at line 32 of SAR Section 5.5.2, the description states: "Historically the blow-down interval has been set such that the pH of the secondary water has been maintained between 5.5 and 9.0, which has kept mineral buildup and corrosion to a minimum." Since city water is being used, the applicability section of the technical specification should be reworded to say: "This specification applies to limiting conditions for secondary coolant pH and radioactivity." Please remove the words, "cycles of chloride," and "resistivity" from TS 3.3.b.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 For the secondary coolant to have measurable levels of radioactivity, a primary-to-secondary leak must be present. The presence of sodium-24 in the secondary coolant would be but one indication of such a leak. Specification 3.3.b.2 should be changed to read: "The radioactivity in the secondary coolant shall be maintained at levels statistically indistinguishable from background."

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.

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Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 It is difficult to define the components that should be included as part of the confinement system versus those that should be included as part of the ventilation system because the ventilation blowers and filters are critical components of the confinement system.

However, these systems have been broken apart in an attempt to make this section of the Technical Specifications follow the fornat outlined in ANSI 15.1. These specifications will be written as follows:

3.4 Confinement 3.4.1 Operations That Require Confinement Applicability:

This specification applies to the operations for which the components of the confinement system must be operable.

Objective:

To assure that operations that have the potential to release airborne radioactive material are performed under conditions in which the release to the environment would be detected, and be limited to levels below 10 CFR 20 limits.

Specification:

1. The confinement system shall be operable whenever:

1. The reactor is operating.

2. Irradiated fuel handling is in progress.

3. Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container.

4. Any work on the core or control rods that could cause a reactivity change of more than 0.65%

dK/K is in progress.

5. Any experiment movement that could cause a reactivity change of more than 0.65% dK/K is in progress.

Bases:

The purpose of the confinement system is to mitigate the consequences of airborne radioactive material release. During operation of the reactor, the production of radioactive gasses or 145

airborne particulates is possible. Though unlikely to occur, fuel cladding failure represents the greatest possible source of airborne radioactivity. The potential causes of fuel cladding damage or failure are:

1. Damage during fuel handling operations.

2. Fuel cladding damage due to an unanticipated reactivity excursion.

Additionally, fission products could be released due to damage to a sufficiently fueled experiment that has been irradiated long enough to build up a significant fission fragment inventory. In the event that the experiment is not adequately contained, it is conceivable that it could be damaged during handling operations to the extent that there could be fission fragment release.

These specifications ensure that the confinement system will be operable during conditions for which there is any potential for fuel cladding damage or failure to occur, as well as for experiment failures in which fission products could potentially be released.

3.4.2 Components Required to Achieve Confinement 3.4.2.1 Normal Operating Mode Confinement Applicability:

This specification describes the components of the confinement system that are necessary in order for the system to perform its intended function under normal operating conditions.

Objective:

To assure that the confinement system is capable of detecting a release of airborne radioactive material.

Specification:

1. The following confinement system components shall be operable:

1. Normal Personnel Access Door

2. Roll Up Door

3. Roof Hatch 146

Bases:

The personnel access door, roll up door, and roof hatch represent the major potential air access ways through confinement. If these components are operable, the major potential air pathways are capable of being controlled to ensure that any airborne radiological release would be detected either by the confinement radiation monitoring system, or by the stack effluent monitoring system.

Under normal operating conditions, the normal operating mode ventilation system controls the general airflow from outside confinement, through confinement, and back out to the environment through the stack.

3.4.2.2 Emergency Operating Mode Confinement Applicability:

This specification describes the components of the confinement system that are necessary in order for the system to perform its intended function under emergency operating conditions.

Objective:

To assure that the confinement system is capable of mitigating the consequences of a possible release of airborne radioactive material.

Specification:

1. The following emergency confinement system components shall be operable:

1. Emergency Confinement System Buttons

2. Confinement Air Intake Damper

3. Confinement Air Exhaust Damper

4. Emergency Personnel Access Door Bases:

Under emergency conditions, operability of any of the emergency confinement system buttons allows the path of the airflow from confinement, through the ventilation svsteru to be changed so that it goes through the 147

emergency filter. Operability of the confinement air intake and exhaust dampers allows the confinement building to be isolated from the outside so that no exhaust confinement air escapes through a pathway other than the emergency pathway. Emergency mode operation of the ventilation system ensures that under emergency conditions, confinement air will be drawn through the emergency filter before being exhausted through the stack.

Operability of the filter minimizes the environmental consequence of a potential airborne radioactivity release.

Emergency mode operation of the ventilation system also ensures that dilution air will be added to the confinement air from the emergency filter. Operability of the emergency personnel access door allows the reactor operator to have a confinement egress route that does not require the individual to go through the main confinement room. When the door is shut, confinement is maintained.

3.4.3 Conditions Required to Achieve Confinement 3.4.3.1 Normal Operating Mode Confinement Applicability:

This specification describes the conditions necessary to assure that normal operating mode confinement is achieved.

Objective:

To assure that the confinement system is functioning sufficiently to prevent airflow from inside confinement to the environment through an uncontrolled pathway.

Specification:

The following conditions shall be met in order to ensure that the normal confinement is achieved:

1. The Normal Personnel Access Door is closed, except for entry and exit.

2. The Roll Up Door is closed.

3. The Roof Hatch is closed.

4. The Emergency Personnel Access Door is closed, except for entry and exit.

5. The Confinement Dampers are Open.

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6. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.

Bases:

Normal confinement is maintained by keeping all of the doors and the roof hatch closed, except for entry and exit.

A negative differential pressure of 0.5 inches of water makes certain that the confinement system is performing its intended function adequately by ensuring that confinement airflow is directed through a defined pathway that is monitored for radiological release. The differential pressure is achieved by circulating air from outside confinement, through the intake damper, and ultimately back out of confinement through the exhaust damper.

3.4.3.2 Emergency Operating Mode Confinement Applicability:

This specification describes the conditions necessary to assure that emergency operating mode confinement is achieved.

Objective:

To assure that the confinement system is functioning sufficiently to prevent airflow from inside confinement to the environment through an uncontrolled pathway, and to assure that the confinement airflow pathway to the environment goes through the emergency filter and is mixed with dilution air prior to being exhausted out of the stack.

Specification:

The following conditions shall be met in order to ensure that the emergency confinement is achieved:

1. The Normal Personnel Access Door is closed, except for entry and exit.

2. The Roll Up Door is closed.

3. The Roof Hatch is closed.

4. The Emergency Personnel Access Door is closed, except for entry and exit.

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5. The Confinement Dampers are Closed.

6. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.

Bases:

Emergency confinement is maintained by closing the confinement intake and exhaust dampers, and by keeping all of the doors and the roof hatch closed, except for entry and exit. This causes all of the make-up confinement air to be drawn in through the spaces around the confinement penetrations, and directed through the Emergency Filter before being exhausted to the stack. A negative differential pressure of 0.5 inches of water makes certain that the confinement system is performing its intended function adequately by ensuring that confinement airflow is directed through the defined pathway that includes the emergency air filter, prior to being released to the environment.

3.5 Ventilation System 3.5.1 Ventilation System Components Required for Normal Operating Mode Applicability:

This specification describes the ventilation system components that must be operating in order to assure that the normal operating mode confinement is functioning.

Objective:

To assure that the normal mode confinement system is capable of performing its intended function.

Specification:

1. The following normal mode ventilation system components shall be operating:

1. Confinement Exhaust Blower

2. Confinement Exhaust Filter System, which shall include:

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1. Roughing Filter

2. Absolute Filter

3. Confinement Exhaust Stack

2. The Confinement Exhaust Filter System Absolute Filter shall be certified by the manufacturer to have a minimum efficiency of 99.97% for removing 0.3 micron diameter particulates.

Bases:

The Confinement Exhaust Blower produces a differential pressure across confinement to ensure that all confinement air pathways are through controlled pathways. The Confinement Exhaust Filter System ensures that the majority of the radioactive particulates that would be likely to be released in the event of a fuel failure would be filtered out prior to being released to the environment, until the emergency operating mode ventilation system is activated. The Confinement Exhaust Stack ensures that the plume of confinement air that is released to the environment, is released at an elevation of 115 feet above ground level, which provides for an opportunity for the air to disperse prior to the plume reaching ground level.

3.5.2 Ventilation System Components Required for Emergency Operating Mode Applicability:

This specification describes the ventilation system components that must be operating in order to assure that the emergency operating mode confinement is functioning.

Objective:

To assure that the emergency mode confinement system is capable of performing its intended function.

Specification:

1. The following emergency mode ventilation system components shall be operating:

1. Emergency Blower 151

2. Emergency Filter System, which shall include:

1. Emergency Filter Intake System Roughing Filter

2. Emergency Filter System Intake Absolute Filter

3. Emergency Filter System Charcoal Filter

4. Emergency Filter System Exhaust Absolute Filter

3. Dilution Blower

4. Confinement Exhaust Stack The exhaust rate through the emergency filter shall be less than or equal to 1500 cfm.

3. The emergency filter is at least 99% efficient at removing iodine.

4. The Emergency Filter System Exhaust Absolute Filter shall be certified by the manufacturer to have a minimum efficiency of 99.97% for removing 0.3 micron diameter particulates.

Bases:

Under emergency conditions, the Confinement Exhaust Blower turns off, and differential pressure across confinement is maintained by the Emergency Blower. The Emergency Blower directs confinement air through the Emergency Filter to remove any radioactive iodine that would be expected to be released during a fuel failure. An airflow limit of 1500 cfm though the filter ensures that the flow rate is low enough to allow the filter to adsorb at least 99 % of the iodine that would be expected to be released in the event of a fuel cladding failure. The Emergency Filter System Exhaust Absolute Filter prevents charcoal particulates from the charcoal filter from being released to the building exhaust air stream. The Dilution Blower provides a non-contaminated source of air to mix with the confinement air, so that any airborne radioactivity that is released is diluted prior to release. The Confinement Exhaust Stack ensures that the plume of confinement air that is released to the environment, is released at an elevation of 115 feet above ground level, which provides for an opportunity for the air to disperse prior to the plume reaching ground level.

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3.6 Emergency Power 3.6.1 Required Emergency Power Sources Applicability:

This specification describes the emergency electrical power sources that are necessary in order to ensure that power is available to confinement system components that are necessary to ensure that the confinement system is able to perform its intended function in the event of an electrical power outage.

Objective:

To assure that the confinement system is able to perform its intended function even, when normal electrical power is unavailable.

Specification:

1. An emergency electrical power source shall be operable whenever the confinement system is required to be operable.

Bases:

Operability of the emergency electrical power source ensures that the blower systems that are necessary in order to maintain emergency operation mode confinement will remain operable, even in the event of a facility electrical power outage.

3.6.2 Components Required to be Supplied with Emergency Power Applicability:

This specification describes the confinement system components that are required to be connected to an emergency electrical power source.

Objective:

To assure that the confinement system is able to perform its intended function even, when normal electrical power is unavailable.

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Specification:

1. The following confinement system components shall be connected to an emergency power source:

1. Emergency Blower

2. Dilution Blower Bases:

In the event of a power outage, the reactor will scram due to the loss of magnet current to the shim safety blades. The confinement air intake and exhaust dampers are pneumatically operated and will fail closed to isolate the confinement room. The confinement exhaust blower will shut off due to loss of power. As long as the emergency and dilution blowers continue to be operable, the emergency confinement system will continue to perform its intended function. In the event of a power outage, the emergency power source will supply the emergency and dilution blowers with electricity so that they will continue to operate, and the emergency confinement system will continue to be functional.

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. Expldin 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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 14.93. These Technical Specifications have been re-written to conform to ANSI 15.1. Section 3.4.1 addresses the conditions under which the confinement system is required to be operable.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The intention of the "and/or" condition was to recognize that some of the items in the specification are equipment that must be operable, and that some of the items are conditions that must be met. The condition in the specification (P. 14-24 Lines 38-39) 154

wiii be changed to "or" since each of the items listed is either equipment that must be operable, or a condition that must be met.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 14.93. These Technical Specifications have been re-written to conform to ANSI 15.1. Section 3.5.1.1 specifies the normal operating mode ventilation components that must be operating in order to achieve normal mode confinement.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 14.93. These Technical Specifications have been re-written to conform to ANSI 15.1. The specifications that are related to ventilation flow rates are:

Specification 3.4.3.1.6:

This specification requires that the differential pressure across confinement be at least 0.5 inches of water under normal confinement / ventilation conditions. No flow rate is specified because the determination of whether or not sufficient confinement exists is based on this differential pressure.

Specification 3.4.3.2.6:

This specification requires that the differential pressure across confinement be at least 0.5 inches of water under emergency confinement / ventilation conditions. No flow rate is specified because the determination of whether or not sufficient confinement exists is based on this differential pressure.

Specification 3.5.2.2:

This specification sets a limit on the maximum emergency ventilation flow rate. The maximum flow rate is limited to 1500 cfm.

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.,

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minimum required equipment, filtration requirements, etc.), and revise the proposed TS as appropriate.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 14,93. These Technical Specifications have been re-written to conform to ANSI 15.1. The specifications that are related to the operability of the emergency cleanup exhaust system are:

Specification 3.4.2.2:

This section specifies the confinement system components that must be operable in order for the emergency cleanup exhaust system to be operable.

Specification 3.4.3.2:

This section specifies the conditions that are required in order to achieve emergency confinement.

Specification 3.5.2:

This section specifies the ventilation system components that must be operable in order for the emergency cleanup exhaust system to be operable.

Specification 3.5.2.3:

This section specifies the emergency filter efficiency that is required in order for the emergency exhaust system to be operable.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI 14.93. These Technical Specifications have been re-written to conform to ANSI 15.1. The verbiage has been changed to make it clear that the function of the emergency power system is to insure that the confinement system will be capable of performing its intended function in the event of a facility power failure.

Specification 3.6.2 describes the confinement / ventilation system components that are required to be supplied with emergency power.

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The LCO for the emergency power supply is that it is capable of supplying the emergency and dilution blowers with enough power that they are capable of operating in the event of a facility power failure. The power source is not specified in the Technical Specifications so that it will be possible to replace the generator that is currently used if need be, without modifying the RINSC Technical Specifications.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The emergency power system is unrelated to the normal operation of the confinement system. Under normal conditions, the differential pressure that controls the confinement air pathway is generated by the confinement exhaust blower. This blower is not supplied with emergency power.

In the event of a power outage, the reactor will scram due to the loss of magnet current to the shim safety blades. The confinement air intake and exhaust dampers are pneumatically operated and will fail closed to isolate the confinement room. The confinement exhaust blower will shut off due to loss of power. Confinement is only maintained if the emergency confinement system is turned on.

Loss of facility power also causes the lighting in confinement to shut off. Except for minimal emergency lighting that switches on when a power failure is detected, there is no other lighting in confinement during facility power outages. Consequently, there is no reasonable opportunity to continue any of the activities that would have the potential to cause an airborne release of radioactivity during a facility power outage. If it were absolutely necessary for operations that require confinement to continue, confinement would have to be provided via the emergency system.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 As noted in the comment, ANSI 15.1 recommends that the technical specifications include the minimum number, type and location of required environmental radiation monitors. In our view, the key phrase in the ANSI standard is its referral to "required environmental monitors" presupposing that the safety analysis has determined the need for such monitors. Our safety analysis did not establish a specific need for such monitors. In addition, it is unclear as to which technical specification is referred to by this comment. Section 11.1.7 is not a technical specification but merely describes a 157

monitoring program using integrating dosimeters that has been in existence for over twenty years. The results from that monitoring are included in our annual report to the NRC to show compliance with basic 10 CFR 20 requirements. Technical Specification 6.8.4.e. already requires "a description of any environmental surveys performed outside the facility." Since this set of comments does not include administrative controls, e.g.,

annual reporting requirements, it is unclear as to where the suggested technical specification would go. RINSC has an operating history of over forty years that suggests that the current environmental monitoring system is sufficient, is documented and shared with the NRC annually and that no additional technical specification is needed.

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.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 The facility radiation monitoring system is described in SAR Section 7.2.15 and summarized in Table 3.2 of the technical specifications. The radiation monitoring system is powered and "on" all of the time. Thus, it is unnecessary to have separate requirements for fuel movement or handling radioactive materials.

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)

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the answer to RAI question 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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 This comment appears to be taken out of context. Specification 3.7.1.1 reads: "When the reactor is operating, gaseous and particulate sampling of the stack effluent shall be monitored by a stack monitor with a readout in the control room. The particulate activity monitor and the gaseous activity monitor for the facility exhaust stack shall be operating.

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If either unit is out of service for more than one shift (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), either the reactor shall be shut down or the unit shall be replaced by one of comparable monitoring capability." It is clear that the monitor is required during normal operation of the reactor. If, for some reason, either monitor is out of service for more than six hours, we either shut the reactor down or replace the defective monitor with a comparable one.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The Basis for operating the reactor for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without a gaseous activity monitor is covered in the answer to RAI Question 14.80.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The initial idea behind this apparent inconsistency was that the reactor would not be started unless the stack gaseous and particulate air monitoring systems were both working at the time of the reactor checkout. If one of the systems failed during operation, it could be replaced with another instrument for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This allows the staff six hours to work on getting the failed system back in operation before having to suspend reactor operations.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The practice has been to limit reactor operation to one shift when a portable monitor is performing the function of an area radiation monitor on the experimental level. This allows time to repair or replace the defective radiation monitor without immediately 159

shutting down the reactor. It is apparent that the comment fails to consider the earlier discussion in Technical Specification 3.2. The note (b) in Table 3.2 states that the reactor cannot 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 areas. If one looks carefully at the asterisk, one would note that the radiation monitors are on the reactor bridge, next to the fuel safe and at the thermal column. Since the reactor bridge and fuel safe are on level 5 while the thermal column is on level 3, there is only one radiation monitor on the experimental level normally, i.e., the one by the thermal column. This configuration has existed for many years and the safety analysis did not identify a need for any additional monitoring. Based on the safety analysis in Chapter 13, the critical area radiation monitor is the one on the reactor bridge since it is the first warning of a fuel failure (MHA). This technical specification will be changed and updated when the technical issues relating to the MHA have been resolved.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The "Bases" portion of the specification should read: "A continuing evaluation of contamination levels within the reactor building will be made to prevent airborne gaseous and particulate radioactivity from reaching the derived air concentration levels in 10 CFR 20 and to assure the safety of personnel. A continuing evaluation of gaseous and particulate activity in the facility exhaust will be made to assure that airborne effluent releases remain within 10 CFR 20 limits offsite. This is accomplished by the monitoring systems described in Table 3.2."

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."

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The "Applicablity" section of TS 3.7.2.a has a typo (P. 14-27 Lines 2-3). It should be changed to:

This specification applies to the monitoring of airborne effluents from the Rhode Island Nuclear Science Center (RINSC).

The "Objective" section of TS 3.7.2.a (P. 14-27 Lines 13-16) should be changed to:

To assure that the release of airborne radioactive material from the RINSC will not cause the public to receive doses that are greater than the limits established in 10 CFR 20.

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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 4x1OE4. 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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 In our view, the pertinent regulation is 10 CFR 20.1101(d) which states, in part, "to implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon-222 and its daughters, shall be established by licensees other than those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions." Thus, are requesting4ethat the Technical Specification be changed to read:

Limiting Condition for Operation: The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will not exceed 10 mrem as calculated using a generally-accepted computer program.

Surveillance.Requirement: Airborne effluents shall be monitored by a continuous air monitor installed, calibrated and maintained in accordance with ANSI 13.1. The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will be calculated using a generally-accepted computer program.

Records: Records of calibration, annual releases and effective dose equivalent calculations shall be maintained for at least three years.

Basis: 10 CFR 20.1101(d) states, in part, "to implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon-222 and its daughters, shall be established by licensees other than those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions."

Since the Rhode Island Nuclear Science Center is located on Narragansett Bay, the wind does not blow in the same direction more than about 10% of the time as shown in the following table taken from historical wind rose data.

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Wind Blowing From Frequency  % Wind Blowing From Frequency  %

North 6.20 E-02 6.02 South 5.80 E-02 5.80 North/Northeast 5.80 E-02 5.80 South/Southwest 8.40 E-02 8.40 Northeast 4.40 E-02 4.40 Southwest 1.05 E-01 10.50 East/Northeast 1.30 E-02 1.30 West/Southwest 6.40 E-02 6.40 East 1.20 E-02 1.20 West 6.80 E-02 6.80 East/Southeast 1.30 E-02 1.30 West/Northwest 9.50 E-02 9.50 Southeast 5.80 E-02 6.80 Northwest 1.04 E-01 10.40 South/Southeast 4.90 E-02 4.90 North/Northwest 6.80 E-02 6.80 Thus, during routine operations, no individual would be in the pathway of the plume more than about 10% of the time. Calculations of annual dose equivalent due to the primary airborne effluent. Argon-41, using the COMPLY Code show less than the allowable ALARA limitation given in 10 CFR 20.1101 for the hypothetical maximum exposed individual member of the general public.

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).

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please see our response to RAI 14.110 above. Compliance with 10CFR20.1 101 dose limits for individual members of the public from gaseous effluents (10omrem/y) is currently demonstrated by calculation through the use of the COMPLY code as generally described in Section 11.1.7 of the SAR.

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.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 Response: The calculation of the accident x/Q is provided in Chapter 13, Section 13.2.1.for short-term releases. The dispersion factor given in the technical specification for nornmal operations was calculated from historic wind rose data provided in the referenced letter. That data has been updated and is summarized below:

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Wind From Frequency N 0.062 NNE 0.058 NE 0.044 ENE 0.013 E 0.012 ESE 0.013 SE 0.058 SSE 0.049 S 0.058 SSW 0.084 SW 0.105 WSW 0.064 W 0.068 WNW 0.095 NW 0.104 NNW 0.068 It should be noted that the wind pattern is heavily influenced by Narragansett Bay.

In our atmospheric dispersion model, we determined the radionuclide concentrations at ground-level receptors beneath an elevated, buoyant plume of dispersing airborne effluents using two major steps: First, we calculated the height to which the plume rises at a given downwind distance from the plume source. The calculated plume rise was then added to the height of the plume's source point to obtain the so-called "effective stack height", also known as the plume centerline height or simply the emission height.

The stack at the Rhode Island Nuclear Science Center is 35 meters high. The effective stack height is determined by the buoyancy of the airborne effluent resulting from the effluent's temperature relative to the temperature of the immediate atmosphere. The ground-level radionuclide concentration beneath the plume at a given downwind distance was then predicted using the Gaussian dispersion equation. It should be noted that our airborne effluents are lighter than the surrounding air because they are generally at a higher temperature than the ambient air into which they are discharged. The dilution factor given in the specification was based on a dispersion factor (X/Q = 10-5 sec/mi3).

However, please change the specification to read: "The annual total dose equivalent to the maximally exposed individual from radioactive materials discharged to the atmosphere shall not exceed 10 millirems using a generally accepted atmospheric dispersion model."

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.

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Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please replace the technical specification with the following:

Limiting Condition for Operation: Releases from the liquid waste retention tank shall meet requirements in 10 CFR 20.2003.

Surveillance Requirement: "The liquid waste retention tank discharge shall be batch sampled and the gross activity per unit volume determined before release."

Records: Records of releases shall be maintained for at least three years.

Basis: 10 CFR 20.2003 permitc discharges to the sanitary sewer provided that conditions in 10 CFR 20.2003 (a) are met.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please see the response to RAI 14.113.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please see the response to RAI 14.113.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification section 3.8 will be re-written so that it more closely conforms to ANSI 15.1. The reactivity limits on experiments are covered in the re-written section of Technical Specifications 3.1.3 and 3.1.4. See the answer to RAI question 14.137. See the answer to RAI 13.7 for the transient analysis associated with a step reactivity insertion of the maximum worth of an experiment.

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Technical Specification 3.8 should be revised to say:

3.8 Experiments 3.8.1 Experiment Materials Applicability:

This specification describes the limitations on the types of materials that may be irradiated or installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

Objective:

The objective of this specification is to prevent damage to the reactor, reactor pool, and reactor experimental facilities.

Specification:

1. Corrosives Materials

1. Corrosive materials shall be doubly contained in corrosion resistant containers.

2. Highly Water Reactive Materials

1. Highly water reactive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

3. Explosive Materials

1. Explosive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

4. Fissionable Materials

1. The quantity of fissionable materials used in experiments shall not cause the experiment reactivity worth limits to be exceeded.

Basis:

ANSI 15.1 recommends that the kinds of materials used in experiments be taken into consideration in order to limit the 165

possibility of damage to the reactor, reactor pool, or reactor experimental facilities. Specifically, ANSI suggests that:

Damage could arise as a result of corrosive materials reacting with core, or experimental facility materials. Specification 3.8.1.1 reduces the possibility of this by requiring that corrosive materials be doubly contained so that the likelihood of container breach is minimized.

Damage could arise as a result of highly water reactive materials reacting with the pool water. Specification 3.8.1.2 makes this scenario impossible by prohibiting the use of highly water reactive materials in experiments.

Damage could arise as a result of explosive materials reacting inside and experimental facility. Specification 3.8.1.3 makes this scenario impossible by prohibiting the use of explosive materials in experiments.

Failure of experiments that contain fissionable materials have the potential to have an impact on reactor criticality, or on radioactive material release. The consequence of experiment failure on criticality is bounded by limiting the reactivity worths of experiments. The analysis for this is in SAR Chapter 13 as part of the transient analysis. The radioactive material release is bounded by the analysis in SAR Chapter 13 for the Maximum Hypothetical Accident involving a fuel element failure.

3.8.2 Experiment Failures or Malfunctions Applicability:

This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

Objective:

The objective of this specification is to ensure that experiments cannot fail in such a way that they contribute to the failure of other experiments, core components, or principle barriers to the release of radioactive material.

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Specification:

1. Experiment design shall be reviewed to ensure that credible failure of any experiment will not result in releases or exposures in excess of limits established in 10 CFR 20.

2. Experiment design shall be reviewed to ensure that no reactor transient can cause the experiment to fail in such a way that it contributes to an accident.

3. Experiment design shall be reviewed to ensure that credible failure of any experiment will not contribute to the failure of:

1. Other Experiments

2. Core Components

3. Principle physical barriers to uncontrolled release of radioactivity

4. Experiments which could increase reactivity by flooding shall not remain in the core, or adjacent to the core unless the minimum core shutdown margin required would be satisfied with the experiment in the flooded condition.

Basis:

ANSI 15.1 recommends that experiment design be taken into consideration in order to limit the possibility that an experiment failure or malfunction could result in other failures, accidents, or significant releases of radioactive material.

Experiments are reviewed by the RINSC Nuclear and Radiation Safety Committee prior to being authorized to be installed in the reactor pool, or inside the reactor experimental facilities. These specifications ensure that experimental design is considered as part of the review, in order to minimize the possibility' of these types of problems due to experiment failure or malfunction.

In order to determine the reactivity worth of a new experiment for which there is no data based on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core. In that case. it is possible that an experiment could be found to have enough ,v,  ;.-ivitv that if additional positive reactivity were Idddc' cCueC t'o ,loding, the shutdown margin would be less 167

than 1.0 % dK/K. In that event, Technical Specification 3.8.2.4 requires that the experiment be removed immediately.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Please change proposed technical specification 3.8.3 to read: "Each experiment containing fissionable materials shall be limited to a maximum reactivity worth of 0.60

% AK/K if secured or 0.08 % AK/K if moveable. The total reactivity of all experiments shall not exceed 0.60 % AK/K." The basis for the proposed technical specification can be found in Chapter 10 of the SAR. SAR Section 10.3, "Experiment Review," states that the reactivity worth of any single independent experiment or combination of connected experiments that can be added to the core simultaneously cannot exceed 0.60% AK/K and the calculated reactivity worth of any single independent experiment not rigidly fixed in place or the combination of connected or related experiments added to the core simultaneously cannot exceed 0.08% AK/K. Positive reactivity is the result of the insertion of either fissile materials or reflector materials into the core.

To address the questions posed in the RAI, when the SAR was written, the facility had an Agreement State broad-scope radioactive materials license and that is the license to which the quoted statement refers. When the facility was governed by both licenses, the broad-scope license allowed inventory was limiting. Among the numerous radionuclides

.formed when fissile material is fissioned in an experiment, the limitations on strontium and iodine were the most restrictive of the inventory limits in the broad-scope license.

During the long delay between when the SAR was submitted and the NRC completed its review and issued this RAI, the broad-scope license was dropped. It should be noted that our current technical specifications contain the same restriction and we were asked by the NRC not to submit any requests for amendments to our license while we were awaiting review of our SAR. Thus, since the limiting broad-scope license inventory items no longer exist, the questions posed in the first portion of this RAI are essentially moot at this point.

It is our contention that the limitations on the reactivity worth of an experiment essentially assures that the consequences of failure of that experiment will remain within the dose equivalent consequences of the fuel element failure. It should be noted that the RINSC is currently licensed to increase the core fuel elements from fourteen to seventeen. Each additional fuel element provides approximately 275 grams of uranium-168

235 far exceeding the reactivity of any single experiment or combination of experiments containing fissionable material. Additionally, there are technical specification limits on core excess reactivity and core shutdown margin that must be met taking the experiment into account. The inventory of radioactivity in the core is dependent on core power level and the RINSC is limited to 2 MW. The MHA assumes the failure of a fuel element containing the fission products from far more fissionable material than any single experiment or combination of experiments. Thus, the MHA bounds the radiological consequences of the failure of an experiment containing fissionable materials.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. There is no specific discussion about internal or external heating. Refer to the answer provided for RAI question 14.116.

14.119 The requirements of TS 3.8.10 imply that accidents involving experiments could result in occupantional 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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Once more the RAI shows the inherent confusion in the guidance provided by the NRC in NUREG-1537, Part 1. The NUREG uses ANSI/ANS-15.1-1990 as its recommended (required) guide. In keeping with ANSI/ANS-15.1-1990, Section 3.8.3 (1), credible failure of any experiment shall not result in releases or exposures in excess of established limits nor in excess of annual limits established in -10 CFR 20. Proposed technical specification 6.5.9, "Operating Procedures," states, in part: "Experiment review on a case-by-case basis assuring that section 3.8.3(2) of ANSI/ANS 15.1 is satisfied." Experiments are reviewed by the Nuclear and Radiation Safety Committee prior to initiation (see proposed technical specification 6.4.2.b).

14.120 TS 3.8.10 contains requirements related to occupational and public radiation doses resulting from experiments. The specification states, "Experimental materials... which 169

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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26. 2010 Please change TS 3.8.10 to read:

"The radiation dose received by any individual member of the public shall not exceed 100mrem (lmSv) in any calendar year as a result of all experiments conducted at the facility.

The radiation dose in unrestricted areas shall not exceed 2mrem (0.02mSv) in any one hour from any single experiment or set of experiments.

Annual air emissions of radioactive materials from routine operations and all experiments conducted shall not result in doses greater than 10tmrem (0.1mSv) total effective dose equivalent (TEDE)"

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 As originally submitted, specification 3.8.1 indicates that sample materials that are going to be irradiated must be contained in containers or encapsulation materials that do not react with water, and do not induce corrosion of core and core structural materials. The type of encapsulation is not relevant, as long as it is able to contain the sample material, and as long as the material used for the containers will not react with any of the core, core structure, or coolant materials.

As originally submitted, specification 3.8.10 describes actions to be taken in the event of the failure of a sample container.

Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. There is no discussion about types of encapsulation. Refer to the answer provided for RAI question 14.116.

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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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. The radiological release of all experiments is now covered in Specification 3.8.2.1. Refer to the answer provided for RAI question 14.116.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please see the response to RAI 14.120.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Please see the response to RAI 14.120.

14.125 The third paragraph of TS 3.8.10 states, "Limits for maximum permissible concentrations are specified in the appropriate section of 10 CFR 20." 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.

Seventh Response to RAI Dated April 13, 2010 Submitted December 14, 2010 Please change the wording in TS 3.8.10 to read: "Limits for derived air concentrations for occupational exposure may be found in 10 CFR Part 20, Appendix B, Table 1, Column 3 and limits for derived air concentrations for airborne effluent releases may be found in 10 CFR Part 20, Appendix B, Table 2, Column 1."

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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.

Eighth Response to RAI Dated April 13. 2010 Submitted January 24, 2011 Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.116.

14.127 TS 3.9.a.1 sets a limit of lx10E22 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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 This limit is based on an analysis that was done by the University of Missouri Research Reactor (MURR). In their analysis, they note that the HFIR Reactor has noticed the presence of small cracks at fast fluences of 1.8 X 10ý2 nvt, and suggest that "a value of 1 X 1022 nvt (>IMeV) could be used as a conservative lower limit for determining when replacement of a beryllium reflector should be considered." The RINSC limit of 1 X 10`2 nvt is even more conservative than what this analysis considers because it is not limited to fast neutron flux. See the reference entitled "Be N Fluence".

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 TS 3.9a references "Part A Section VIII" of the SAR. This section is from a previous SAR. TS 3.9a should reference Section 4.2.3, 'Neutron Moderators and Reflectors'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 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.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 As submitted Technical Specification 3.9.b requires that the fuel elements be inspected for physical defects and reactor core box fit. This is a surveillance requirement rather than an LCO. Consequently, this specification has been moved to become Teclhical Specification 4.9.b. See the answer to RAI question 14.167.

14.130 TS 4.0 specifies that some surveillance requirements may be deferred during periods of reactor shutdown. As recommended in ANSIIANS-15.1, allowed deferral of a 172

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.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 This question is best addressed after the set of RINSC surveillance items have been determined. The rationale and basis for deferring any given surveillance item, and the determination of how long the item may be deferred depends on what the complete set of surveillance items include, and therefore what other related operability checks, calibrations, and inspections are being performed during an outage.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 TS 4.2.8 requires that reactivity insertion rates be determined annually and whenever a new core is configured. In order to make this determination, shim safety blade withdrawal rates must be determined. No requirement is specified for measuring the withdrawal rates, because it would be redundant. They are determined as part of the reactivity insertion rate measurement.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specifications 4.1 and 4.2 have been re-written in order to make them conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138 for the revised version of Technical Specification 4.1, and refer to the answer provided for 14.141 for the revised version of Technical Specification 4.2. Specification 4.1.1.2 provides the surveillance requirements for shim safety blade reactivity worths.

Specification 4.2.2 provides the surveillance requirements for shim safety blade reactivity insertion rates. In both cases, a surveillance requirement has been added to cover activities that could have an effect on these parameters.

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.

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Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 There is no reason for specific core configurations to be referenced in this section of the specification. Technical Specification 4.1 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138. The surveillance requirements have been written so that they do not refer to specific core configurations.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 4.1 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138. The surveillance requirements have been written so that they do not refer to specific core configurations.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specifications 4.1 and 4.2 have been re-written in order to make them conform. more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138 for the revised version of Technical Specification 4.1, and refer to the answer provided for 14.141 for the revised version of Technical Specification 4.2. Specification 4.1.1.2 provides the surveillance requirements for shim safety blade reactivity worths.

Specification 4.2.2 provides the surveillance requirements for shim safety blade reactivity insertion rates. References to the SAR have been removed.

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.

Ninth Response to RAI Dated April 13, 2010 Submitted February 24, 2011 The purpose of inspecting the control blades is to ensure that they do not swell. Swelling could cause the blade insertion rate to increase to an extent that scram drop times could be impacted. In an effort to make the basis consistent with the specification, the paragraph that describes the basis for this (P. 14-32 Lines 43-46) will be modified to say:

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Shim safety blade inspections have the potential to be the single largest source of radiation exposure to the facility personnel. In order to minimize personnel radiation exposure and provide an inspection frequency that will detect early evidence of swelling, an annual inspection interval was selected for Specification 4.1.2.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 In order to determine the reactivity worth of an experiment, one may either:

A. Estimated it based on previous experience due to the similarity of material, quantity of material, position in the core, etc. with other experiments for which reactivity worth has been determined.

B. Measure it by determining the critical control rod heights with, and without the experiment loaded, and calculate the reactivity difference to determine the reactivity effect of the experiment.

For experiments that are not similar to any previously performed experiments, option B is the only way to determine the reactivity effect.

The reactor is defined in TS 1.17 to be operating "'whenever it is not secured or shutdown". Consequently, it is not possible to measure the reactivity effect without the reactor being in operation. The basis for this specification acknowledges this fact.

The answer to RAI question 14.60 that was sent with RINSC's second submission of answers should be updated to say:

3.1.3 The total reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth:

Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 3.1.4 The maximum reactivity worth of any individual experiment shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth:

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Fixed 0.6 % dK.K Moveable 0.08 % dK/K 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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specifications 3.1 and 4.1 have been re-written in order to make them conform more closely to ANSI 15.1. Some of the specifications that had been in section 3.1 have been moved. The table that follows provides a summary of how things have been changed. The new Technical Specification 4.1.1.3.2 addresses the issue regarding experiment worth changes due to core configuration changes.

Original Specification New Location Location 3.1.1 Shutdown Margin Reactivity 3.1.1.1.1 3.1.2 Core Excess Reactivity 3.1.1.1.2 3.1.3 Total Experiment Reactivity Worth 3.1.1..3.1 3.1.4 Individual Experiment Reactivity Worth 3.1.1.3.2 3.1.5 Criticality During Fuel Loading 3.1.1.1.4 3.1.6 Regulating Rod Reactivity Worth 3.1.1.2.1 3.1.7 Flooded Experiment 3.8.2.4 3.1.8 Negative Temperature Coefficient 3.1.1.1.3 Temperature Coefficient Surveillance 4.1.1.1.3 3.1.9 FC Mode Operation Core Grid Filled 3.1.2.1 3.1.10 FC Mode Operation Coolant Gate Stored 3.1.2.2 The basis section of Specifications 3.1.1.3.1 and 3.1.1.3.2 refer to analyses performed for reactivity insertions. The determination of a period due to a 0.08 % dK/K reactivity insertion is part of the answer to RAI question 14.61.

The basis section of Specification 4.1.1.1.3 refers to an analysis that estimates the temperature coefficient. This analysis is part of the answer to RAI questions 4.12 and 4.13.

The new versions of Technical Specification 3.1 and 4.1 are:

3.1 Core Parameters 176

3.1.1 Reactivity Limits Applicability:

This specification applies to all core configurations, including configurations that have experiments installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

Objective:

The objective of this specification is to make certain that core reactivity parameters will not exceed the limits used in the safety analysis to ensure that a reactor transient will not result in damage to the fuel.

Specification:

3.1.1..1 Core 3.1.1.1.1 The core shutdown margin shall be at least 1.0 %

dK/K.

3.1.1.1.2 The core excess reactivity shall not exceed 4.7 %

dK/K.

3.1.1.1.3 The temperature coefficient shall be negative.

3.1.1.1.4 The reactor shall be subcritical by at least 3.0

%dK/K during fuel loading changes.

3.1.1.2 Control Rods 3.1.1.2.1 The reactivity worth of the regulating rod shall not exceed 0.6 % dK/K.

3.1.1.3 Experiments 3.1.1.3.1 The total reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for. the purpose of measuring experiment reactivity worth:

Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 177

3.1.1.3.2 The maximum reactivity worth of any individual experiment shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth:

Fixed 0.6 % dK.K Moveable 0.08 % dK/K Basis:

Specification 3.1.1.1.1 provides a limit for the minimum shutdown reactivity margin that. must be available for all core configurations. The shutdown margin is necessary to ensure that the reactor can be made subcritical from any operating condition, and to ensure that it will remain subcritical after cool down and xenon decay, even if the most reactive control rod failed in the fully withdrawn position. No credit is taken for the negative reactivity worth of the regulating rod because it would not be available as part of the negative reactivity insertion in the event of a scram.

Specification 3.1.1.1.2 provides a maximum limit for excess reactivity available for all core configurations. Excess reactivity is necessary to overcome the negative reactivity effects of coolant temperature increase, coolant void increase, fuel temperature increase, and xenon build-up that occur during sustained operations. Excess reactivity is also required to be available in order to overcome any negative reactivity effects of experiments that are installed in the core.

Specification 3.1.1.1.3 requires that the temperature coefficient be negative. This requirement ensures that a temperature rise due to a reactor transient will not cause a further increase in reactivity.

Specification 3.1.1.1.4 provides a limit for the minimum core shutdown reactivity during fuel loading changes. This limit takes advantage of the negative reactivity that can be added to the core above and beyond the shutdown margin by the insertion of the highest reactivity worth, and regulating control rods. This limit assures that the core will remain subcritical during these operations.

Specification 3.1.1.2.1 provides a limit for the reactivity worth of the regulating rod. The reactivity limit is set to a value less than 178

the delayed neutron fraction so that a failure of the automatic servo system could not result in a prompt critical condition.

Specification 3.1.1.3.1 provides total reactivity limits for all experiments installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The limit on total experiment worth is set to a value less than the delayed neutron fraction so that an experiment failure could not result in a prompt critical condition. The limit on total moveable experiment worth is set to a value that will not produce a stable period of less than 30 seconds, so that the reactivity insertion can be easily compensated for by the action of the control and safety systems. As part of the Safety Analysis, Argonne National Laboratory modeled a reactivity insertion of + 0.08 % dK/K over a 0.1 second interval, and determined that this reactivity insertion resulted in a stable period of approximately 75 seconds.

Specification 3.1.1.3.2 provides total reactivity limits for any individual experiment installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The reactivity limits for both, fixed and moveable experiments are the same as the limits for total fixed and moveable experiments. Consequently, the safety analysis done for Specification 3.1.1.3.1 applies to this specification as well.

3.1.2 Core Configuration Limits Applicability:

This specification applies to core configurations during operations above 0.1 MW when the reactor is in the forced convection cooling mode, Objective:

The objective of this specification is to ensure that there is sufficient coolant to remove heat from the fuel elements when the reactor is in operation at power levels greater than 0.1 MW.

Specification:

3.1.2.1 All core grid positions shall contain fuel elements, baskets, reflector elements, or experimental facilities during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.

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3.1.2.2 The pool gate shall be in its storage location during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.

Basis:

Specification 3.1.2.1 requires that all of the core grid spaces be filled when the reactor is operated at higher power levels that require forced convection cooling. This requirement prevents the degradation of coolant flow through the fuel channels due to flow bypassing the actively fueled region of the core through unoccupied grid plate positions.

Specification 3.1.2.2 requires that the pool gate that is used for separating the sections of the pool, be in its storage location when the reactor is in operation at higher power levels that require forced convection cooling. This requirement ensures that there will be a sufficient heat sink for high power operations, and ensures that the full volume of the pool water will be available in the event of a loss of coolant accident.

4.1 Core Parameter Surveillance 4.1.1 Reactivity Limit Surveillance Applicability:

This specification applies to the surveillance requirements for reactivity limits.

Objective:

The objective of this specification is to ensure that reactivity limits are not exceeded.

Specification:

4.1.1.1 Core Reactivity Limit Surveillance 4.1.1.1.1 The core shutdown margin shall be determined:

Annually Whenever the core reflection is changed Whenever the core fuel loading is changed 180

4.1 .1.1.2 The core excess reactivity shall be determined:

Annually Whenever the core reflection is changed Whenever the core fuel loading is changed 4.1.1.1.3 The temperature coefficient shall be shown to be negative at the initial start-up after a fuel type change.

4.1.1.2 Control Rod Reactivity Limit Surveillance 4.1.1.2.1 The reactivity worth of the regulating rod shall be determined:

Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.2.2 The reactivity worth of the shim safety rods shall be determined:

Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.3 Experiment Reactivity Limit Surveillance 4.1.1.3.1 The reactivity worth of new experiments shall be determined prior to the experiments initial use.

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4.1.1.3.2 The reactivity worth of any on going experiments shall be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously.

Basis:

Specification 4.1.1.1.1 requires that the core shutdown margin be determined annually, and whenever there is a change in core loading or core reflection. The annual measurement of the shutdown margin provides a snapshot of how the shutdown margin is increasing due to fuel bum-up. Measurements made whenever the core. loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.

Specification 4.1.1.1.2 requires that the core excess reactivity be determined annually, and whenever there is a change in core loading or core reflection. The annual measurement of the excess reactivity provides a snapshot of how it is decreasing due to fuel burn-up. Measurements made whenever the core loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.

Specification 4.1.1.1.3 requires that the temperature coefficient be shown to be negative at the initial start-up after a fuel type change. A negative temperature coefficient makes power increases self limiting by inserting a negative reactivity effect as fuel and coolant temperatures rise. As part of the Safety Analysis, Argonne National Laboratory determined that for the equilibrium core, the temperature and void coefficients are negative over a temperature range of 20 C to 100 C. The fuel temperature coefficient was determined to be negative over a temperature range of 20 C to 600 C.

Specification 4.1.1.2.1 requires that the regulating rod reactivity be determined annually, and whenever there is a change in core loading or core reflection. These determinations provide assurance that the rod worth does not exceed its reactivity limit due to fuel bum-up, changes in core configuration, or control rod degradation.

Specification 4.1.1.2.2 requires that the shim safety rod reactivities be determined annually, and whenever there is a change in core loading or core reflection. These determinations 182

provide assurance that the rod worths do not degrade due to rod changes, or changes in core configuration.

Specification 4.1.1.3.1 requires that the reactivity worth of new experiments be determined prior to initial use. This ensures that reactivity worth limits are not exceeded.

Specification 4.1.1.3.2 requires that the reactivity worth of on going experiments be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously. This provides assurance that core configuration changes do not cause experiment reactivity worth limits to be exceeded, without requiring that experiment worths be re-determined every time that a recurring core configuration change, such as equilibrium core re-fuelling, occurs.

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.

Eighth Response to RAL Dated April 13, 2010 Submitted January 24, 2011 RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. Specification 4.2.7.5 requires that the power level channels be calibrated annually. This calibration is done by thermal power verification.

14.140 ANSIIANS-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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. Interlock surveillance is covered in Specification 4.2.6.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Specification 4.2 has been revised to more closely reflect what ANSI 15.1 suggests should be covered by this specification. The following table shows how the locations of the original specifications have changed.

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Specification Original ANSI New Location Standard Location Channel Test of Neutron Flux Level Safeties 4.2.1 4.2.5 4.2.4.2 Channel Test of Period Safety 4.2.1 4.2.5 4.2.4.3 Channel Calibration of Channels in Table 3.1 4. 2.2 4.2.5 4.2.7 Radiation Monitors in Table 3.2 Operable 4.2.3 4.2.3 Rod Drop Time 4.2.4 4.2.4 4.2.1.1 Rod Drop Time 4.2.5 4.2.4 4.2.1.2 Shutdown Margin 4.2.6 4.1.2 4.1.1.1.1 Excess Reactivity 4.2.7 4.1.1 4.1.1.1.2 Reactivity Insertion Rate 4.2.8 4.2.2 4.2.2 Control Rod Reactivity Worth 4.1.1 4.2.1 4.1.1.2 Power Calibration 4.2.8 4.2.7.5 The following is the new proposed Specification:

4.2 Reactor Control and Safety System Applicability:

This specification applies to the safety and safety related instrumentation.

Objective:

The objective of this specification is to ensure that the safety and safety related instrumentation is operable, and calibrated when in use.

Specification:

4.2.1 Shim safety drop times shall be measured:

4.2. 1.1 Annually 4.2.1.2 Whenever maintenance is performed which could affect the drop time of the blade 4.2.2 All shim safety reactivity insertion rates shall be measured:

4.2.2.1 Annually 4.2.2.2 Whenever maintenance is performed which could affect the reactivity insertion rate of the blade 4.2.3 The following radiation monitors shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the 184

shutdown condition, and after the channel has been repaired or de-energized:

4.2.3.1 The experimental level area radiation monitor 4.2.3.2 The pool top area radiation monitor 4.2.3.3 The gaseous effluent air monitor 4.2.3.4 The particulate air monitor 4.2.4 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized:

4.2.4.1 Control room manual scram button 4.2.4.2 Power level channels 4.2.4.3 Period channel 4.2.4.4 Rod control communication watchdog scram 4.2.5 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor, is started up from the shutdown condition, and after the channel has been repaired or de-energized for which reactor power level will be greater than 100 kW:

4.2.5.1 All of the reactor safety and safety related instrumentation listed in 4.2.4 4.2.5.2 Primary coolant flow rate scram 4.2.6 The following reactor safety and safety related instrumentation alarms, scrams, and interlocks shall be tested annually:

4.2.6.1 The following detector HV failure scrams:

4.6.2.1.1 Power level channels 4.6.2.1.2 Period channel 4.2.6.2 The following shim safety withdrawal interlocks:

4.2.6.2.1 Start-up count rate 4.2.6.2.2 Test / Select switch position 4.2.6.3 The following servo control interlocks:

4.2.6.3.1 Regulating blade not full out 4.2.6.3.2 Period less than 30 seconds 185

4.2.6.4 The following coolant system channel temperature alarms and scrams:

4.2.6.4.1 Primary inlet temperature alarm 4.2.6.4.2 Primary outlet temperature alarm 4.2.6.4.3 Primary outlet temperature scram 4.2.6.4.4 Pool temperature alarm 4.2.6.4.5 Pool temperature scram 4.2.6.5 The following coolant system channel flow scrams:

4.2.6.5.1 Primary flow scram 4.2.6.5.2 Inlet and outlet coolant gates open scrams 4.2.6.5.3 No flow thermal column scram 4.2.6.6 Low pool level scram 4.2.6.7 The following bridge scrams:

4.2.6.7.1 Bridge manual scram 4.2.6.7.2 Bridge movement scram 4.2.6.7.3 Bridge low power position scram 4.2.6.8 Seismic scram 4.2.7 The following reactor safety and safety related instrumentation shall be calibrated annually:

4.2.7.1 The experimental level area radiation monitor 4.2.7.2 The pool top area radiation monitor 4.2.7.3 The gaseous effluent air monitor 4.2.7.4 The particulate air monitor 4.2.7.5 Power level channels 4.2.7.6 Primary flow channel 4.2.7.7 Primary inlet temperature channel 4.2.7.8 Primary outlet temperature channel 4.2.7.9 Pool temperature channel Basis:

Specification 4.2.1 defines the surveillance interval for measuring the shim safety drop times. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.

The requirement that this parameter be measured after maintenance is performed which could affect the drop time of the blade assures that the reactor will not be 186

operated with a shim safety blade that does not meet the LCO requirements due to maintenance activities.

Specification 4.2.2 requires that all shim safety reactivity insertion rates shall be measured annually. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.

Specification 4.2.3 indicates the radiation monitors that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.

Specification 4.2.4 indicates the reactor safety and safety related instrumentation that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.

Specification 4.2.5 provides for the fact that if the reactor is operated at power levels less than or equal to 100 kW, the forced cooling system is not required to be operational. However, for operations above 100 kW, this specification requires that the primary coolant flow rate scram be verified to be operable prior to the initial start-up of the reactor. This requirement is consistent with the historical facility requirements.

Specification 4.2.6 defines the surveillance interval for testing the reactor safety and safety related instrumentation alarms, scrams, and interlocks that are not tested as part of the requirements of Specifications 4.2.4 and 4.2.5. The annual requirement is consistent with the historical facility frequency.

Specification 4.2.7 defines the surveillance interval for calibrating the safety and safety related instrumentation. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 While we do not want to limit our ability to operate the reactor for extended runs over multiple days, the current typical operating schedule at RINSC is one shift per day. Our desire is to set this surveillance such that these channel checks are performed once prior to the initial start-up of the day, so that if there are multiple start-ups for the day, additional channel checks are not required.

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In the event that there were a multi-day operation, it is not considered likely that RINSC could operate for a quarter of a year without re-fuelling. RINSC reached its equilibrium core in October of 2008. Based on operating data, we expect to have to refuel after 1550 MWH of operation. Therefore, if we were start with a fresh core, and operate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week, we would reach our limit of 1550 MWH of operation in 2.1 months. Consequently, quarterly surveillance in lieu of prior to initial start-up of the day is redundant.

If the wording is changed to make sure that pre-start checkouts are performed prior to the initial start-up each day that the reactor is started up from the shutdown condition, rather than after it has been secured, these conditions can be met.

RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. The new proposed specification regarding the operability of the Neutron Flux Level Safety and Period Safety Channels is:

4.2.4 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized:

4.2.4.2 Power level channels 4.2.4.3 Period channel 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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Table 3.1 was updated as part of the answer to RAI question 7.1. RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. The new proposed specification regarding channel calibrations is:

4.2.7 The following reactor safety and safety related instrumentation shall be calibrated annually:

4.2.7.1 The experimental level area radiation monitor 4.2.7.2 The pool top area radiation monitor 4.2.7.3 The gaseous effluent air monitor 4.2.7.4 The particulate air monitor 4.2.7.5 Power level chalnels 4.2.7.6 Primary flow channel 4.2.7.7 Primary inlet temperature channel 4.2.7.8 Primary outlet temperature channel 4.2.7.9 Pool temperature channel 188

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 The LCO regarding the required radiation monitoring instrumentation is covered in the new proposed Specifications 3.2.1.3 and 3.2.1.4 which were submitted as part of the answer to RAI question 14.87. The corresponding surveillance requirements are part of the revised Specification 4.2 which is part of the answer to RAI question 14.141.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 Technical Specification 4.2.6 will be changed to say:

The shutdown margin shall be determined in accordance with operating procedures:

Annually, When a new core is configured, Following control blade changes.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 4.2 has been re-written as part of the answer to RAI 14.141.

The reference to the SAR has been removed.

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 IS as appropriate.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 Technical Specification 4.2.7 will be changed to say:

The excess reactivity shall be determined in accordance with operating procedures:

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Annually, When a new core is configured, Following control blade changes.

14.148 TS 4.2.7 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed IS will become part of the IS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 4.2 has been re-written as part, of the answer to RAI 14.14 1.

The reference to the SAR has been removed.

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 IS as appropriate.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 Technical Specification 4.2.8 will be changed to say:

The excess reactivity shall be determined in accordance with operating procedures:

Annually, When a new core is configured, Following control blade changes.

14.150 TS 4.2.8 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed IS will become part of the IS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Technical Specification 4.2 has been re-written as part of the answer to RAI 14.141.

The reference to the SAR has been removed.

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.

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Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141.

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 IS 4.2.3, and revise the proposed IS as appropriate.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141.

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.

Tenth Response to RAI Dated April 13, 2010 Submitted July 15, 2011 Technical Specifications 3.3 and 4.3 have been re-written in order to make them more consistent with ANSI 15.1. The following table provides- a summary of how the specifications have changed:

Original Specification New Location Location 3.3.a. I Primary pH Note 1 3.3.a.2 Primary Conductivity 3.3.1.1 3.3.a.3 Primary Radiological Analysis 3.3.1.2 3.3.b. 1 Secondary pH Note 2 3.31.b.2 Secondary Radiological Analysis 3.3.2 4.3.a. I Primary pH Surveillance Note 1 4.3 .a.2 Primary Conductivity Surveillance 4.3.1.1 4.3.a.3 Primary Radiological Analysis Surveillance 4.3.1.2 4.3.a.4 Pool Level Scram Test 4.2.6.6 4.3.a.5 Pool Inspection - Primary System Inspection 4.3.1.4 4.3.a.6 Pool Level Verification 4.3.1.3 4.3.b.1 Secondary pH Surveillance Note 2 4.3.b.2 Secondary Radiological Analysis Surveillance 4.3.2.1 N/A Secondary System Inspection 4.3.2 2 191

Note 1 ANSI 15.1 recommends that either pH or Conductivity be monitored. pH and conductivity are related, so monitoring them both is redundant. CO 2 dissolves into the pure pool water, which pushes the equilibrium pH down to an average of approximately 5.6. Consequently, conductivity is a better measure of water quality. RINSC proposes to use a conductivity measurement rather than pH to monitor water quality.

Note 2 The purpose of measuring pH and conductivity is to reduce activation products in the coolant, and to minimize corrosion. Activation products are not an issue on the secondary side of the cooling system because the coolant is not exposed to a neutron flux. Corrosion on the secondary side of the cooling system is no longer an issue because the aluminum piping has been replaced with PVC piping, and the cooling towers are made of non-corrosive materials as well. RINSC proposes to remove this surveillance.

3.3 Coolant Systems 3.3.1 Primary Coolant System 3.3.1.1 Primary Coolant Conductivity Applicability:

This specification applies to the primary coolant.

Objective:

The objective of this specification is to maintain the primary coolant in a condition that minimizes corrosion of the fuel cladding, core structural materials, and primary coolant system components, as well as to minimize activation products produced as a result of impurities in the coolant.

Specifications:

The primary coolant conductivity shall be < 2 ýtmho / cm when averaged over a quarter of a year.

Bases:

Specification 3.3.1.1 is based on empirical data from the facility history. Over the lifetime of the facility, primary coolant conductivity has been maintained within the limit specified, and no corrosion on the fuel cladding, core structural materials, or primary coolant system components have been noted.

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3.3.1.2 Primary Coolant Activity Applicability:

This specification applies to the primary coolant.

Objective:

The objective of this specification is to provide a mechanism for detecting a potential fuel cladding leak.

Specification:

Cs-137 and 1-131 activity in the primary coolant shall be maintained at levels that are indistinguishable from background.

Basis:

Specification 3.3.1.2 provides a mechanism for detecting a potential fuel cladding leak by requiring that periodic primary coolant analysis be performed to test for .the presence of Cs-137 or 1-131. These isotopes are prominent fission products.

Consequently, if either of these isotopes are detected in the primary coolant, it may indicate a fuel cladding leak.

3.3.2 Secondary Coolant System Applicability:

This specification applies to the secondary coolant.

Objective:

The objective of this specification is to provide a mechanism for detecting a potential primary to secondary system leak.

Specifications:

Na-24 activity in the secondary coolant shall be maintained at levels that are indistinguishable from background.

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Bases:

Specification 3.3.2.1 provides a mechanism for detecting a potential primary to secondary system leak by requiring that periodic secondary coolant analysis be performed to test for the presence of Na-24. This isotope is produced by the activation of the aluminum structural materials in the primary pool, and a small concentration of it is present in the primary coolant during, and immediately following operation of the reactor. If this isotope is found in the secondary coolant, it may indicate a primary to secondary system leak.

4.3 Coolant Systems 4.3.1 Primary Coolant System 4.3.1.1 Primary Coolant Conductivity Surveillance Applicability:

This specification applies to the surveillance of the primary coolant.

Objective:

The objective of this specification is to provide a periodic verification that the primary coolant conductivity is within prescribed limits.

Specification:

The conductivity of the primary coolant shall be tested monthly.

Basis:

Specification 4.3.1.1 requires that the conductivity of the primary coolant be tested on a monthly basis. ANSI 15.1 recommends that this be performed on a weekly to quarterly schedule. Specification 3.1.1.1 sets a limit on the average conductivity when averaged over one quarter of a year.

Consequently, a monthly measurement falls within the ANSI recommended schedule, and allows for a running average based on three data points per quarter.

194

4.3.1.2 Primary Coolant Activity Surveillance Applicability:

This specification applies to the surveillance of the primary coolant.

Objective:

The objective of this specification is to provide a periodic verification that the Cs- 137 and 1-13 1 activity in the primary coolant is not significantly above background.

Specifications:

Cs-137 and 1-131 activity in the primary coolant shall be measured annually.

Basis:

Specification 4.3.1.2 requires that the Cs- 137 and 1-131 activity in the primary coolant be tested on an annual basis. This schedule is consistent with the schedule recommended by ANSI 15.1.

4.3.1.3 Primary Coolant Level Inspection Surveillance Applicability:

This specification applies to the surveillance of the primary coolant.

Objective:

The objective of this specification is to ensure that the coolant level is at an adequate height above the core during reactor operation.

Specification:

The primary coolant level shall be verified to be greater than or equal to the Limiting Safety System Setting value prior to the initial start-up each day that the reactor is started up from the shutdown condition.

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Basis:

Specification 4.3.1.3 requires that the primary coolant level be inspected prior to the first reactor start-up of each day. A float switch system is used to monitor the pool level 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week. This system is tied into the facility alarm system, which is monitored by an offsite alarm company. In the event that the pool level reaches one inch greater than the LSSS, the automatic pool fill is started. If the pool level drops to the LSSS, then a scram occurs, the operator receives an alarm, and the alarm company receives an alarm.

A daily verification of the pool level prior to starting the reactor up provides adequate assurance that the float switch is working to maintain the pool level.

4.3.1.4 Primary Coolant System Inspection Surveillance Applicability:

This specification applies to the surveillance of the primary cooling system components.

Objective:

The objective of this specification is to provide a periodic verification that there are no obvious defects in any of the system components.

Specifications:

The components of the primary coolant system shall be inspected annually.

Basis:

Specification 4.3.1.3 requires that the primary coolant system be inspected on an annual basis. This schedule is consistent with the historical inspection schedule for the facility.

4.3.2 Secondary Coolant System 4.3.2.1 Secondary Coolant Activity Surveillance 196

Applicability:

This specification applies to the surveillance of the secondary coolant.

Objective:

The objective of this specification is to provide a periodic verification that the Na-24 activity in the primary coolant is not significantly above background.

Specification:

Na-24 activity in the secondary coolant shall be measured annually.

Basis:

Specification 4.3.2.1 requires that the Na-24 activity in the primary coolant be tested on an annual basis. This schedule is consistent with the schedule recommended by ANSI 15.1.

4.3.2.2 Secondary Coolant System Inspection Surveillance Applicability:

This specification applies to the surveillance of the secondary cooling system components.

Objective:

The objective of this specification is to provide a periodic verification that. there are no obvious defects in any of the system components.

Specification:

The components of the secondary coolant system shall be inspected annually.

Basis:

Specification 4.3.2.2 requires that the primary coolant system be inspected on an annual basis. This schedule is consistent with the historical inspection schedule for the facility.

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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.

Tenth Response to RAI Dated April 13, 2010 Submitted July 15, 2011 Technical Specifications 3.3 and 4.3 have been re-written in order to make them more consistent with ANSI 15.1. See the answer to RAI question 14.153.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 ANSI/ANS standards are recommendations, not requirements. ANSI/ANS 15.1 Section 4.5.2 recommends that filter efficiency measurements be made annually to biennially, or following major maintenance. The only required ventilation filter at RINSC is the Emergency Exhaust Filter, which is tested annually as required by RINSC TS 4.4, 4.5, 4.6 Specification 4.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 Technical Specifications 4.4, 4.5, 4.6 have been broken apart in an attempt to make this section of the Technical Specifications follow the format outlined in ANSI 15.1. These specifications will be written as follows:

4.4 Confinement System Surveillance 4.4.1 Normal Operating Mode Confinement System Applicability:

This specification describes the surveillance requirements for the normal operating mode confinement system.

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Objective:

The objective of this specification is to verify that the normal operating mode confinement system is functional prior to reactor start-up.

Specification:

1. The conditions required to achieve normal operating mode confinement that are specified in section 3.4.3.1 shall be verified to be met prior to the each day of reactor start-up.

Bases:

If the conditions specified in section 3.4.3.1 are met, then the normal operating mode confinement system is functioning. By ensuring that the normal operating mode confinement system is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would be directed to the stack, mixed with dilution air, and detected by the stack radiation monitor system.

4.4.2 Emergency Operating Mode Confinement System Applicability:

This specification describes the surveillance requirements for the emergency operating mode confinement system.

Objective:

The objective of this specification is to verify that the emergency operating mode confinement system is functional.

Specification:

1. A functional test of the emergency operating mode confinement system shall be performed:

1. Quarterly

2. After any maintenance that could affect the operability of the system

2. The functional test of the emergency operating mode confinement system shall verify that the conditions required to achieve emergency operating mode confinement are met when an 199

evacuation button is depressed. The following actions shall occur when an evacuation button is depressed:

1. The evacuation horn sounds

2. The following dampers close:

1. Confinement Air Intake Damper

2. Confinement Air Exhaust Damper

3. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.

4. The confinement room HVAC and air conditioners de-energize.

Bases:

A periodic functional test of the emergency confinement system ensures that in the event of an airborne radioactivity release, the emergency confinement system is capable of being activated. The testing periods that are specified conform to ANSI 15.1 recommendations.

4.5 Ventilation System Surveillance 4.5.1 Normal Operating Mode Ventilation System Applicability:

This specification describes the surveillance requirements for the normal operating mode ventilation system.

Objective:

The objective of this specification is to verify that the normal operating mode ventilation system is operable prior to reactor start-up.

Specification:

1. The confinement exhaust blower shall be verified to be in operation prior to each day of reactor start-up:

Bases:

By ensuring that the normal operating mode ventilation system is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would 200

be directed to the stack and be detected by the stack radiation monitor svstem.

4.5.2 Emergency Operating Mode Ventilation System Applicability:

This specification describes the surveillance requirements for the emergency operating mode ventilation system.

Objective:

The objective of this specification is to verify that the emergency operating mode ventilation system is operational and functional.

Specification:

1. A test of the operability of the emergency operating mode ventilation system shall be performed:

1. Quarterly 2.. After any maintenance that could affect the operability of the system

2. The test of the operability of the emergency operating mode ventilation system shall verify that the following actions occur when an evacuation button is depressed:

1. The following blowers are de-energized:

I. Confinement Exhaust Blower

2. Rabbit System Blower

3. Off Gas System Blower

2. The following blowers are energized:

1. Emergency Exhaust Blower

2. Dilution Blower

3. The flow rate at the exhaust of the emergency exhaust blower shall be verified to be less than or equal to 1500 cfrn:

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1. Annually

2. After any maintenance that could affect the operability of the system

4. The emergency filter efficiency shall be verified to be at least 99% efficient for removing iodine:

1. Biennially

2. After any maintenance that could affect the operability of the system Bases:

A periodic test of the operability of the emergency ventilation system ensures that in the event of an airborne radioactivity release, the emergency confinement system is capable of being activated. The verification of the emergency exhaust blower flow rate, and the emergency filter efficiency ensure that the filter will perform its intended function. The testing periods that are specified conform to ANSI 15.1 recommendations.

4.6 Emergency Power System Surveillance Applicability:

This specification describes the surveillance requirements for the emergency power system.

Objective:

The objective of this specification is to verify, that the emergency power system is operable and functional.

Specification:

1. An operability test to verify that the emergency power system starts in the event of a facility power outage shall be performed quarterly.

2. A functional test of the emergency power system under load shall be performed:

1. Biennially

2. Following emergency system load changes 202

Bases:

Periodic tests of the emergency power system ensures that in the event of a facility power outage, the emergency power system would automatically start, and be capable of handling the load required to power the emergency confinement system. The testing periods that are specified conform to ANSI 15.1 recommendations.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI question 14.156. These technical specifications have been completely re-written to conform to ANSI 15.1.

Specification 4.5.1.1 requires that the ventilation components for normal operation be verified to be operation each day of reactor operation.

Specification 4.5.2 indicates the surveillance requirements for verifying the operability and functionality of the emergency operation mode ventilation system components.

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.

Second Response to RAI Dated April 13, 2010 Submitted August 6, 2010 The truck door IS the overhead door.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 See the response to RAI question 14.156. These technical specifications have been completely re-written to conform to ANSI 15.1.

Specification 4.6.1 requires that the starting capability of the emergency power system be verified quarterly.

Specification 4.6.2 requires that the capability of the emergency power system to operate under load be tested biennially, and whenever the emergency system load changes.

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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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 ANSI!ANS-15.1 recommends but does not require a technical specification including surveillance requirements for radiation monitoring at the site boundary and environmental monitoring. The section entitled "Environmental Effects of Facility Operation" of Appendix 12.1 to NUREG 1537 Part 2, indicates that "Yearly doses to unrestricted areas will be at or below established guidelines in 10CFR Part 20." It is our contention that as long as we can demonstrate that our annual doses to individuals in unrestricted areas meet those criteria, we do not need a technical specification governing radiation monitoring at our site boundary or additional environmental monitoring. Over forty years of operating experience support that claim.

14.161 TS 4.7.a.I 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)

Fourth Response to RAI Dated April 13. 2010 Submitted September 8, 2010 See the answer to RAI question 7.4, which has been revised to require "a minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement particulate effluent".

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."

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 See the answer to RAI question 7.4, which shows the revision to RINSC Technical Specification Table 3.2. As this revision is written, the required air monitoring instrumentation includes "a minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement particulate effluent". The Stack Particulate Monitor may serve to meet this requirement. Consequently, the Main Floor Air Monitor may not be required to be operational during reactor operation.

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Revise TS 4.7.a.3 to say:

1. A channel check of the particulate air monitor shall be performed for each day of operation, or once for each operation that lasts for multiple days.

2. The particulate air monitor shall be calibrated annually.

3. A channel check of the gaseous air monitor shall be performed for each day of operation, or once for each operation that lasts for multiple days.

4. The gaseous monitor shall be calibrated annually.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 This statement was intended to mean that the safety review of experiments will ensure that the installation of experiments will not put the reactor in a condition that makes reaching a safety limit credible.

Technical Specification 4.8 will be re-written to say:

4.8 Experiments Applicability:

This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.

Objective:

The objective of this specification is to ensure that experiments have been reviewed to verify that the design is within the limitations of the RINSC Technical Specifications and 10 CFR 50.59.

Specification:

4.8.1 Experiments shall be reviewed to ensure that the design is within the limitations of the RINSC Technical Specifications and 10 CFR 50.59 prior to the experiments initial use.

Basis:

This specification ensures that all experiments will be reviewed to verify that the experiment designs are within the limitations of the RINSC Technical Specifications and 10 CFR 50.59 prior to its initial use.

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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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 The amended version of this specification should be used. See the answer to RAI question 14.167.

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.

Eighth Response to RAL Dated April 13, 2010 Submitted January 24, 2011 TS 4.9a references "Part A Section VIII" of the SAR. This section is from a previous SAR. TS 4.9a should reference Section 4.2.3, "Neutron Moderators and Reflectors'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 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)

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 NUREG-1313 (p. 7) states that LEU silicide fuel with up to 4.8 g U/cm 3 was irradiation tested in the 30 MW ORR reactor to bumups up to 98% of the contained 23'U. No indication of unusual conditions were observed on the fuel plates that were tested.

Consequently, LEU silicide fuel has no bumup limit and no fission density limit. Also see the response to RAI 4.2.

14.167 The bases for TS 4.9.b state, "Bumup calculations are made quarterly (4.9.1)." To what does "(4.9.1)" refer? Explain the reason that the bumup calculations are not a required surveillance, and revise the proposed TS as appropriate.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 Burn-up calculation data is not used to assess the physical condition of the fuel. The fuel is qualified to 98% bum-up (See the answer to RAI 14.55). Consequently, this statement should be removed.

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Change this section to say:

4.9.a Beryllium Reflectors Applicability:

This specification applies to the surveillance of the standard and plug type beryllium reflectors.

Objective:

To prevent physical damage to the beryllium reflectors in the core from accumulated neutron flux exposure.

Specification:

1. The maximum neutron fluence of any beryllium reflector shall be:

a. Less than or equal to 1 E 22 neutrons / cm2, and

b. The fluence shall be determined annually.

2. The beryllium reflectors shall be visually inspected and functionally fit into the core grid box on a rotating basis not to exceed five years such that:

a. The surveillance each year shall include at least one fifth of the beryllium reflectors,

b. If a beryllium reflector is removed from use and the time since its last surveillance exceeds five years, it shall be visually inspected and finctionally fit into the core grid box prior to being placed in use, and

c. If damage is discovered, then the surveillance shall be expanded to include all of the beryllium reflectors prior to use, and annually thereafter.

Bases:

The neutron fluence limit is based on an analysis that was done by the University of Missouri Research Reactor (MURR). In their analysis, they note that the HFIR Reactor has noticed the presence of small cracks at fast fluences of 1.8 X 1022 nvt, and they suggest that "a value of 1 X 1022 nvt

(>IMeV) could be used as a conservative lower limit for determining when replacement of a beryllium reflector should be considered." The RINSC limit of 1 X 1022 nvt is even more conservative than what this analysis considers because the RINSC limit is not limited to fast neutron flux.

207

Reflector elements are visually inspected and functionally fit into the core grid box in order to verif, that there are no observable fuel defects or swelling. The rotating inspection schedule ensures that all of the reflectors in the core will be inspected at least once every five years. Since core element handling represents one of the highest risk opportunities for mechanically damaging the fuel cladding, this schedule is deemed appropriate, given the limited amount of information that is gained from these inspections. The discovery of a damaged reflector triggers an increase in the inspection schedule to an annual period.

4.9.b LEU Fuel Elements Applicability:

This specification applies to the surveillance of the LEU fuel elements.

Objective:

To verify the physical condition of the fuel elements in order to prevent operation with danmaged fuel elements.

Specification:

The fuel elements shall be visually inspected and functionally fit into the core grid box on a rotating basis not to exceed five years such that:

1. The surveillance each year shall include at least one fifth of the fuel elements,

2. The surveillance each year shall include fuel elements that represent a cross section with respect to bum-up,

3. If a fuel element is removed from use and the time since its last surveillance exceeds five years, it shall be visually inspected and functionally fit into the core grid box prior to being placed in use, and

4. If damage is detected by Technical Specification 4.3.3 or otherwise discovered, then the surveillance shall be expanded to include all of the fuel elements prior to use, and annually thereafter.

Bases:

RINSC Technical Specification 4.3.3 requires periodic pool water analysis to test for the presence of radioactivity that could potentially indicate a fuel cladding failure. Fuel elements are visually inspected and functionally fit into the core grid box in order to verify that there are no observable fuel defects or swelling. The rotating inspection schedule ensures that all of the fuel elements in the core will be inspected at least once every five years.

Since fuel handling represents one of the highest risk opportunities for 208

mechanically damaging the fuel cladding, this schedule is deemed appropriate, given the limited amount of information that is gained from these inspections. The pool water analysis is the most sensitive mechanism for detecting fuel cladding failure. A detected fuel failure triggers an increase in the inspection schedule to an annual period. Fuel inspections include a cross section of elements with respect to burn-up history in order to ensure that each inspection includes high bum-up elements that would be most likely to start to fail over time.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 TS 4.9b references "Part A Section VI" of the SAR. This section is from a previous SAR. TS 4.9b should reference Section 4.5, 'Nuclear Design'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 14.169 In accordance with 10 CFR 50.36(a)(1), provide bases for proposed technical specifications in Section 5, "Design Features."

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. The number and type of control blades have been included in section 5.3.

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.

209

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. References to the SAR have been removed.

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.

Fifth Response Submitted to RAI Dated April 13, 2010 November 26, 2010 Chapter 14, Section 5 of the RINSC Safety Analysis Report has been re-written to conform to ANSI 15.1. The items in Section 5.4 that appeared to be LCOs have been moved to Chapter 14, Sections 3 and 4. The re-written version of Section 3 is included as part of the answer to RAI 14.93. The re-wTitten version of Section 4 is included as part of the answer to RAI 14.156. The LCOs that have been moved from Section 5, have been moved to the following sections:

1. Section 3.4.2.2.1.1 requires that the confinement and clean up systems become activated when an evacuation button is depressed.

2. Section 3.4.3.2 covers the requirement that gas leaks between confinement and the outside shall be inward.

3. Section 3.4.3.2.6 requires that a negative differential pressure be maintained between confinement and the outside.

4. Section 3.5.1.1.2 requires that the confinement room exhaust air go through a roughing filter and an absolute filter prior to being released to the environment.

5. Section 3.5.2.1.2 describes the components of the emergency filter system that are required.

6. Sections 3.5.2.1.2 and 3.5.2.1.4 require the confinement exhaust to exit through an emergency filter system and a stack.

7. Section 3.5.2.3 requires that the charcoal filter in the emergency filter

- system be capable of removing 99% of the radioiodines likely to be present in the event of a fuel failure.

8. Section 3.5.2.4 requires that the exhaust absolute filter for the emergency filter system be certified by the manufacturer to be capable of removing 0.3 micron particulates.

9. Section 3.6 requires that emergency power be available to operate the clean-up system in the event of a facility power failure.

10. Section 4.4.2.2.2 requires that the dampers close when an evacuation button is depressed.

11. Section 4.4.2.2.4 requires that the confinement room vent fans and HVAC system shut off when an evacuation button is depressed.

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Chapter 14, Section 5 has been re-written to be the following:

5.0 Design Features 5.1 Site and Facility Description The Rhode Island Nuclear Science Center (RINSC) is located on a 3 acre section of a 27 acre auxiliary campus of the University of Rhode Island.

The 27 acre site was formerly a military reservation prior to becoming the Bay Campus of the university. The parcel of land is located in the town of Narragansett, Rhode Island, on the west shore of the Narragansett Bay, approximately 22 miles south of Providence, and approximately 6 miles north of the entrance of the bay from the Atlantic Ocean.

The facility is one of a number of buildings located on the Bay Campus of the university. The RINSC facility consists of a reactor room and an office wing with one entrance between them. The reactor room acts as the confinement space. The reactor pool is constructed on top of a military gun pad.

A more detailed description of the site and the facility is located in Chapter 1.

5.2 Reactor Coolant System The RINSC reactor is located in a concrete pool that is lined with aluminum. For operations up to 100 kW, natural convection cooling is possible. For operations above 100 kW, forced convection cooling must be applied. In both cases, the coolant is light water provided by the local town water supply.

The primary section of the forced convection cooling system takes water from the pool outlet line, and directs it to a delay tank where its progress through the cooling system is held up for approximately 70 seconds in order to reduce the N- 16 concentration in the water. From the delay tank, the forced cooling system is divided into two loops. Each primary loop consists of a pump and a heat exchanger.

For each loop, the water from the delay tank goes through a primary pump, through a primary heat exchanger, and back to the pool via the pool inlet line, where the two loops recombine. The piping for the primary cooling system is aluminum. Nominal temperatures and pressures are less than 130 F and less than 100 psig respectively.

The secondary sides of the primary heat exchangers use city water to remove the heat from the primary sides. For each loop, secondary water 211

from the heat exchanger is circulated to a cooling tower, through the secondary pump, and back to the heat exchanger Both of the cooling towers use air cooling to reduce the temperature of the secondary water.

A more detailed description of the reactor coolant system is located in Chapter 5.

5.3 Reactor Core and Fuel The RINSC fuel is MTR plate type fuel that has a nominal enrichment of 19.75% U-235. The chemical composition of the fuel is U3 Si2 . Each fuel assembly consists of 22 fuel plates, bound by side plates that hold the plates evenly spaced apart. At each end of the assembly, the side plates are attached to square end boxes, that are capable of being inserted into a core grid box. The cladding, side plates, and end boxes are aluminum.

Each fresh fuel assembly is loaded with 275g U-235 nominal.

The core grid box is consists of a 5 15/16 inch thick grid plate that has a 9 X 7 array of square holes, and a box that has four walls that surround the grid plate in such a way that the plate serves as the bottom of the box with the top end open. The grid box is suspended from the top of the pool by four corner posts that occupy the corner grid spaces. The box is oriented so that the open end faces up toward the top of the pool. The reactor core is configured by inserting fuel elements end boxes into grid spaces. so that each fuel assembly is standing up inside the box.

The standard core consists of 14 assemblies in a 3 X 5 array in the center of the grid box, with the central grid space available as an experimental facility. The remaining grid spaces are either filled with graphite or beryllium reflector assemblies, or incore experimental facilities. A non-standard core configuration with 17 fuel elements is also possible. In this configuration, the standard core configuration has been modified so that the three central reflector assemblies on the thermal column edge of the core are substituted with fuel assemblies.

Both core configurations include 4 shim safety control blades, arid a regulating rod. The shim safety blades are located between the fuel and the reflector assemblies on both of the edges of the fuel array that consist of 5 assemblies. There are two blades on each side of the fuel. The blades are housed in shrouds that are part of the core grid box. The shrouds ensure that the blades have unfettered movement in and out of the core. The regulating rod is positioned one grid space out from the fuel, along the central axis of the fuel on the thermal column side of the core.

212

A more detailed description of the reactor core and fuel is located in Chapter 4, as well as a description of core parameters.

5.4 Fissionable Material Storage Irradiated fuel is stored in two types of fuel storage racks in the reactor pool:

Fixed racks that are mounted on the pool wall Moveable racks that rest on the pool floor Each fixed rack has 9 spaces for fuel storage arranged in a linear array.

Each moveable rack has 18 spaces for fuel storage arranged in a 9 X 2 array.

Non-irradiated fuel is typically stored in the RINSC fuel safe.

Non-fuel fissionable materials are either kept where they are in use, or are stored in the reactor pool or fuel safe depending on size constraints and what is most reasonable from an ALARA standpoint.

A more detailed description of the fuel storage racks is located in Chapter 9.

Applicability:

This specification applies to the fissionable material storage facilities used for storing materials while they are not in use, or in an approved shipping container.

Objective:

The objective of this specification is to ensure that it is impossible for fissionable material to achieve a critical configuration.

Specification:

1. Fissionable material that is not in use or not in an approved shipping container shall be in storage.

2. Fissionable material storage facilities shall have keff <=

0.9, for all conditions of moderation and reflection using light water.

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Bases:

These specifications conform to ANSI 15.1. They ensure that fissionable material that is not in use will remain in a configuration that cannot achieve criticality.

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.

Fourth Response to RAI Dated April 13, 2010 Submitted September 8, 2010 This statement applies to both, normal and emergency operation. Under normal operation, air exits confinement through the normal exhaust intake, goes through the exhaust blower, and enters the base of the stack. The experimental gas systems (off gas and rabbit) tie into the normal exhaust system so that air from these systems go through the exhaust blower and enter the base of the stack. This part of the system is shown in green on the following diagram:

Normal Intake Exhaust Blower Delay Tank DelayTankMotorized Moisture Separator C3-Way Valve Beam Ports, Thermal Off Gas Blower Column, etc..

Rabbit Gate Rabbit Filter Rabbit Blower Exhaust Stack Valve Box Contaminated Water Storage Tanks Gaseous/Particulate RINSC Normal and Emergency Ventilation Systems Lab 102 Hoods Dilution Absolute - Normal Ventilation Filter Dilution Blower - Emergency Ventilation SAlways On

- Major Vent Emergency Charcoal - Minor Vent Filter Emergency Blower Emergency Intake rmoFniLlveer - Air Sampling Line 214

Dilution air is provided by suction on laboratory fume hoods. This air enters the system through the fume hoods, goes through the dilution blower, and enters the base of the stack, where it mixes with confinement air. This system is always on, regardless of whether the confinement exhaust system is in normal operating mode, or emergency operating mode. This part of the system is shown in blue.

Under emergency conditions, the normal exhaust system is shut off, dilution air continues to be provided, and air exits confinement through the emergency exhaust intake. The air goes through the emergency charcoal filter, the emergency blower, and enters the base of the stack, where it mixes with dilution air. This part of the system is shown in red.

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The new analysis for the maximum hypothetical accident in Chapter 13 does not take credit for dilution air. Consequently, no LCO is necessary. See the basis document entitled "Fuel Damage Radiological Assessment".

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.

Fifth Response to RAI Dated April 13, 2010 Submitted November 26, 2010 The confinement system is dependent on a negative differential pressure of 0.5 inches of water across confinement. This differential pressure is achieved by the confinement exhaust blower. As long as the blower has a sufficient flow rate to maintain the 0.5 inch differential pressure, the blower exhaust flow rate is adequate for achieving its intended function. The blower flow rate is not measured. The underlying LCO parameter that is measured is the differential pressure. Consequently, the LCO is the differential pressure rather than the exhaust blower flow rate.

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.

Eighth Response to RAI Dated April 13, 2010 Submitted January 24, 2011 Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. References to the SAR have been removed.

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14.177 Proposed TS 1.21 defines "Reactivity Worth of an Experiment." The definition implies that the reactivity change due to flooding is included in the reactivity worth of the experiment. This seems contrary to the statement in the response to RAI 14.65 that, "an experiment could be found to have enough positive reactivity that if additional positive reactivity were added due to flooding, the shutdown margin would be less than 1.0%

delta k/k." Explain this apparent discrepancy, and revise the proposed TS as appropriate.

Response to December 17, 2012 RAI Dated March 15, 2013 The issue surrounding this question is that we don't know what the reactivity worth of an experiment is until we put the experiment in the core and determine it. We need to make it clear that putting an experiment in the core to detennine its reactivity worth is not a violation of Technical Specifications when we discover that its worth exceeds Technical Specification limits, as long as it is removed immediately.

Technical Specification 1.21. will remain the same, and will continue to define the reactivity worth of an experiment in such a way that the definition takes reactivity worth chanues due to flooding into consideration. The definition will continue to be:

1.21 Reactivity Worth of an Experiment The reactivity worth of an experiment is the maximum absolute value of the reacti'ity change that would occur as a result of:

A. Insertion or removal from the core, B. Intended or anticipated changes in position, or C. Credible malfunctions that alter experiment position or configuration.

Technical Specification 3.1.1.3 has been written in 'such a way as to take reactivity worth determination into account. The proposed Technical Specifications will be revised to explicitly state that experiments may be loaded in the core for the purpose of measuring experiment reactivity worth by doing criticality studies.

3.1.1.3 Experiments 3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies:

Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 3.1.1.3.2 The maximum reactivity worth of any individual experiment shall not exceed the following limits, except when the operation of the reactor is 216

for the purpose of measuring experiment reactivity worth by doing criticality studies:

Fixed 0.6 % dK.K Moveable 0.08 % dK/K Bases:

Specification 3.1.1.3.1 provides total reactivity limits for all experiments installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The limit on total experiment worth is set to a value less than the delayed neutron fraction so that an experiment failure could not result in a prompt critical condition. The limit on total moveable experiment worth is set to a value that will not produce a stable period of less than 30 seconds, so that the reactivity insertion can be easily compensated for by the action of the control and safety systems.

As part of the Safety Analysis, Argonne National Laboratory modeled a reactivity insertion of + 0.08 % dK/K over a 0.1 second interval, and determined that this reactivity insertion resulted in a stable period of approximately 75 seconds. An allowance is made for measuring the reactivity worth of experiments. The reactor can be made critical with experiments of unknown reactivity, so that the criticality data can be used to determine whether or not the reactivity worth of an experiment is within the limits prescribed by this specification.

Specification 3.1.1.3.2 provides total reactivity limits for any individual experiment installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The reactivity limits for both, fixed and moveable experiments are the same as the limits for total fixed and moveable experiments. Consequently, the safety analysis done for Specification 3.1.1.3.1 applies to this specification as well. An allowance is made for measuring the reactivity worth of experiments.

The reactor can be made critical with experiments of unknown reactivity, so that the criticality data can be used to determine whether or not the reactivity worth of an experiment is within the limits prescribed by this specification.

14.178 Proposed TS 1.31.7 states that abnormal and significant degradation of the fuel cladding is a reportable occurrence. Section 6.7.2(c)(v) of ANS-15.1 recommends that abnormal and significant degradation of the coolant boundary also be a reportable occurrence.

Explain why the proposed TS do not include abnormal and significant degradation of the coolant boundary as a reportable occurrence.

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.Response to December 17, 2012 RAI Dated March 15, 2013 Technical Specification 1.31 will be revised to say:

1.31 Reportable Occurrence A reportable occurrence is any of the following:

1. A violation of a safety limit,

2. An uncontrolled or unplanned release of radioactive material which results in concentrations of radioactive materials inside or outside the restricted area in excess of the limits specified in Appendix B of 10CFR20.

3. Operation with a safety system setting less conservative than the limiting setting established in the Technical Specifications,

4. Operation in violation of a limiting condition for operation established in the Technical Specifications,

5. A reactor safety system component malfunction or other component or system malfunction which could, or threaten to, render the safety system incapable of performing its intended safety functions,

6. An uncontrolled or unanticipated change in reactivity in excess of 0.75 % dK/K,

7. Abnormal and significant degradation of the fuel cladding,

8. Abnormal and significant degradation of the primary coolant boundary, or

9. An observed inadequacy in the imnplementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the facility.

14.179 Proposed TS 3.1.1.3.1 specifies limits for the total reactivity worth of experiments.

Section 3.8.1 (2) of ANS-15.1 recommends that the TS should specify the "sum of the absolute values of the reactivity worths of all experiments." Proposed TS 3.1.1.3.1 does not sum the absolute values of the reactivity of all experiments. Explain this apparent discrepancy, and revise the proposed TS as appropriate.

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Response to December 17, 2012 RAI Dated March 15, 2013 We will change specification 3.1.1.3.1 to say:

3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies:

Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 14.180 The responses to RAI 14.65 and RAI 14.137 state that, "in order to determine the reactivity worth of a new experiment for which there is no data on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core." Revise proposed TS 3.1.1.3.1 and 3.1.1.3.2 to explicitly state the allowed reactor operations when testing the reactivity worth of new experiments.

Response to December 17, 2012 RAI Dated March 15, 2013 See the response to RAI 14.177.

14.181 Proposed TS 3.7.1.1 and 3.7.1.2 specify radiation monitoring instrumentation required during reactor operation and fuel movement. Explain why the instrumentation is not required during all conditions that require confinement specified in proposed TS 3.4.1.

Revise the proposed TS as appropriate.

Response to December 17, 2012 RAI Dated March 15, 2013 Technical Specification 3.7.1.1 and 3.7.1.2 will be revised to say:

3.7.1.1. The following air radiation monitoring instrumentation shall be operable whenever:

The reactor is operating, Irradiated fuel handling is in progress, Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container, Any work on the core or control rods that could cause a reactivity change of more than 0.65% dKJK is in progress, or 219

Any experiment movement that could cause a reactivity change of more than 0.65% dK/K is in progress:

3.7.1.1.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous or particulate effluent shall be operating.

3.7.1.1.2 If this detector fails during operation, a suitable alternative gaseous or particulate air monitor may be used, or an hourly grab sample analysis may be made in lieu of having a functioning monitor.

3.7.1.2 The following fission product radiation monitoring instrumentation shall be operable whenever:

The reactor is operating, Irradiated fuel handling is in progress, Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container, Any work on the core or control rods that could cause a reactivity change of more than 0.65% dK!K is in progress, or Any experiment movement that could cause a reactivity change of more than 0.65% dK!K is in progress:

3.7.1.2.1 A minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.

3.7.1.2.2 If this detector fails, a suitable alternative meter with alarming capability may be placed at the top of the bridge in lieu of the normal detector.

14.182 Current TS 3.7.1 contains requirements for radiation monitors that include explicit requirements for using portable monitors in the event that the installed instrumentation fails. Proposed TS 3.7.1 1.2 and 3.7.1.2.2 do not contain similar requirements. Revise the proposed TS to include explicit requirements for the use of replacement monitors.

Provide analyses and evaluation in the SAR that justify the proposed TS. (Note: The response to RAI 14.80 which references an NRC safety evaluation is not an adequate basis for a TS. As required by 10 CFR 50.36(b), TS "will be derived from the analyses and evaluation included in the safety analysis report.")

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Response to December 17, 2012 RAtI Dated March 15, 2013 The current TS describing air monitoring (TS 3.7.1) says:

When the reactor is operating, gaseous and particulate sampling of the stack effluent shall be monitored by a stack monitor with a readout in the control room.

The particulate activity monitor and the gaseous activity monitor for the facility exhaust stack shall be operating. If either unit is out of service for more than one shift (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), either the reactor shall be shut down or the unit shall be replaced by one of comparable monitoring capability.

The proposed TS says:

3.7.1.1. The following air radiation monitoring instrumentation shall be operable when the reactor is in operation, and during fuel handling operations:

3.7.1.1.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous or particulate effluent shall be operating.

3.7.1.1.2 If this detector fails during operation, a suitable alternative gaseous or particulate air monitor may be used, or an hourly grab sample analysis may be made in lieu of having, a functioning monitor.

The only difference between the current and proposed TS's is the proposed TS does not allow for the running of the reactor for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with a dysfunctional detector. The requirements governing the capabilities of the gas monitor are the same.

The current TS (3.7.1) which describes the required area radiation monitor states:

3. The reactor shall not be continuously* operated without a minimum of one area radiation monitor (Table 3.2.8) on the experimental level of the reactor building and one area monitor (Table 3.2.6) over the reactor pool (reactor bridge) operating and capable of warning personnel of high radiation levels.

• In order to continue operation of the reactor, replacement of an inoperative monitor must be made within 15 minutes of recognition of failure, except that the reactor may be operated in a steady-state power mode if the monitor is replaced with portable gamma-sensitive instruments having their own alarm.

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The proposed TS states:

3.7.1.2 The following fission product radiation monitoring instrumentation shall be operable when the reactor is in operation, and during fuel handling operations:

3.7.1.2.1 A minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.

3.7.1.2.2 If this detector fails, a suitable alternative meter with alarming capability may be placed at the top of the bridge in lieu of the normal detector.

The differences in the requirements governing the replacement detectors is that the new TS does not require that the replacement detector be "gamma sensitive" only that it be "suitable". The new TS also does not give a requirement that the detector be replaced within 15 minutes of recognition of failure. The new TS will be rewritten to read:

3.7.1.2 The following fission product radiation monitoring instrumentation shall be operable whenever:

The reactor is operating, Irradiated fuel handling is in progress, Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container, Any work on the core or control rods that could cause a reactivity change of more than 0.65% dK/K is in progress, or Any experiment movement that could cause a reactivity change of more than 0.65% dK/K is in progress:

3.7.1.2.1 A minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.

3.7.1.2.2 If this detector fails, a suitable gamma sensitive alternative meter with alarming capability may be placed at the top of the bridge in lieu of the normal detector.

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3.7.1.2.3 If failure of the detector occurs during operations than the replacement detector must be put in place within 15 minutes of the recognition of failure.

14.183 The regulation 10 CFR 50.54(m)(1) requires a Senior Reactor Operator to be present at the facility during recovery from an unplanned or unscheduled significant reduction in power. Proposed TS 6.1.3.2 does not include this requirement. Revise the proposed TS to correct this discrepancy.

Response to December 17, 2012 RAI Dated March 15, 2013 6.1.3.2 A Senior Reactor Operator shall be present in the facility during any of the following operations:

6.1.3.2.1 The initial reactor start-up and approach to power for the day, 6.1.3.2.2 Fuel element, reflector element, or control rod core position changes, 6.1.3.2.3 Experiment installation or removal for experiments that have a reactivity worth greater than 0.75 %dK/K, 6.1.3.2'.4 Recovery from an unscheduled significant reduction in power, and 6.1.3.2.5 Recovery from an unscheduled shutdown.

14.184 Proposed TS 6.1.3.2.2 requires a Senior Reactor Operator to be present at the facility during fuel or control rod position changes. Explain the reason for not including reflector position changes in proposed TS 6.1.3.2.2, and revise the proposed TS as appropriate.

Response to December 17, 2012 RAI Dated March 15, 2013 This will be incorporated into Proposed Technical Specification 6.1.3.2.2. See the response to RAI 14.183.

14.185 Section 6.2.2 (4) of ANS-15.1 states that the charter for the review and audit group should include provisions for dissemination, review, and approval of meeting minutes in a timely manner. Proposed TS 6.2.2 does not include any similar provisions. Revise the proposed TS to include provisions for dissemination, review, and approval of meeting minutes in a timely manner, or justify omitting such provisions from the proposed TS.

Response to December 17, 2012 RAI Dated March 15, 2013 This will be incorporated into Proposed Technical Specification 6.2.2:

6.2.1 Nuclear and Radiation Safety Committee (NRSC) Composition and Qualifications 223

6.2.1.1 Composition The NRSC shall be comprised of a minimum of seven individuals:

6.2.1.1.1 The Director 6.2.1.1.2 The Assistant Director for Operations 6.2.1.1.3 The Assistant Director for Reactor and Radiation Safety 6.2.1.1.4 Four members that are not RIAEC conmmissioners or staff 6.2.1.2 Qualification The collective qualification of the NRSC members shall represent a broad spectrum of expertise in science and engineering.

6.2.1.3 Alternates Qualified alternates may serve in the absence of regular members.

6.2.2 Nuclear and Radiation Safety Committee Charter The NRSC shall have a written Charter that specifies:

6.2.2.1 Meeting frequency of not less than once per year, 6.2.2.2 Quorum shall consist of a minimum of seven (7) members, including the Assistant Director for Radiation and Reactor Safety, and the Director or Assistant Director for Operations.

6.2.2.3 NRSC Minutes shall be reviewed and approved at the next committee meeting.

6.2.2.4 If deficiencies that affect reactor safety are found, a written report shall be submitted to the RIAEC Commissioners within three months after the NRSC has completed its audit.

14.186 Section 6.2.2 (2) of ANS-15.1 states that a quorum should consist of at least half of the voting membership of the review and audit group. Proposed TS 6.2.2.2 does not specify a minimum number of NRSC members for a quorum. What is the minimum number of NRSC members needed for a quorum? Revise the proposed TS as appropriate.

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Response to December 17, 2012 RAI Dated March 15, 2013 The review and audit group is the Nuclear and Radiation Safety Committee (NRSC).

This committee is made of individuals from the RINSC staff, the local university community, and the nearby nuclear power industry. Unfortunately, with the varied schedules of the individuals on this committee, there have been occasions in which it has been difficult to have a quorum. Consequently, to this point in time, quorum has been defined in the current Technical Specification 6.4.4 to be:

A quorum of the NRSC shall consist of not less than four (4) members and shall include the Radiation Safety Officer or designee, the Director or the Assistant Director for Operations and the Chairman or designee.

ANSI 6.2.2 (2) suggests that in order to have a quorum, at least half of the voting membership be present, and that the RINSC staff not constitute a majority of the member present. RINSC recently expanded the membership of the NRSC in an effort to increase the likelihood that at least half of the members would be able to make a meeting. It is unclear at present whether or not having more members will make it easier or more difficult to get half of the members together at any given time. The present NRSC Charter states that:

The Director, Assistant Director for Operations, and the Radiation Safety Officer shall be ex officio members of the committee.

Consequently, the RINSC staff proposes that the minimum number of members required to have a quorum be seven members. This would guarantee that the RINSC representation on the Committee would not constitute a majority. Also, at present there are thirteen NRSC members. As a result, with the current number of members, quorum would be more than half the membership. However, if it is determinedthat quorum is difficult to obtain, additional members could be added without increasing the minimum number required for quorum, This would help alleviate the problem of not being able to get a quorum. See proposed Technical Specification 6.2.2.2 in the response to RAI 14.185.

14.187Proposed TS 6.2.2.2 implies that members of the Nuclear and Radiation Safety Committee could also be members of the Rhode Island Atomic Energy Commission.

Proposed TS 6.2.1.1 does not include provisions for members of the RIAEC to be members of the NRSC. Clarify whether members of the RIAEC can also be members of the NRSC, and revise the proposed TS as appropriate.

Response to December 17, 2012 RAI Dated March 15, 2013 See Technical Specification 6.2.1.1.4 in the response to RAI 14.185.

14.188 Section 6.2.3 (5) of ANS-15.1 states that the review and audit committee shall review violations of the license. Proposed TS 6.2.3 does not include such a requirement.

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Explain why no such provision exists in the proposed TS, or revise the proposed TS to include such a provision.

Response to December 17, 2012 RAI Dated March 15, 2013 Proposed Techmical Specification 6.2.3.1 says that the NRSC shall review proposed changes to the Technical Specifications or License, and violations of the Technical Specifications. In order to make this more clear, the wording will be changed to:

The NRSC shall review the following items:

6.2.3.1 Proposed changes to the Technical Specifications or License, and violations of the Technical Specifications or License, 14.189 Section 6.2.3 (8) of ANS-15.1 states that the review and audit group shall review audit reports. Proposed TS 6.2.3 does not include such a requirement. Explain why no such provision exists in the proposed TS, or revise the proposed TS to include such a provision.

Response to December 17, 2012 RAI Dated March 15, 2013

'ANSI 15.1 Section 6.2.3 (8) indicates that the NRSC shall review audit reports.

Historically, the NRSC has been the group that performs the audits of both, the Facility Operations Program, and .the Facility Radiation Safety Program. Since this committee performs these audits, there is no provision for them to '4review" an audit report. Any issues discovered are discussed, and recorded in the NRSC meeting minutes.

14.190 Section 6.2.3 of ANS-15.1 contains provisions for distribution of minutes of the review and audit group meetings. Proposed TS 6.2.3 does not include such provisions. Explain why no such provisions exist in the proposed TS, or revise the proposed TS to include such provisions.

Response to December 17, 2012 RAI Dated March 15, 2013 The Director and the two Assistant Directors are ex-officio members of the review and audit committee. Consequently, the management of the facility is directly involved and engaged in the Safety Committee findings, and receive the minutes that contain those findings.

14.191 Section 6.2.4 of ANS-15.1 states, "In no case shall the individual immediately responsible for the area perform an audit in that area." Proposed TS 6.2.4 doesn't contain any provisions to ensure that audits are performed by an individual or group that is independent of the area under audit. Explain the controls in place at the RINSC to ensure audits are independent, and revise the proposed TS as appropriate.

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Response to December 17, 2012 RAI Dated March 15, 2013 Technical Specification 6.2.4 will be revised to say:

6.2.4 Audit Function The non-RIAEC staff members of the NRSC shall audit the following items:

This will ensure that the audits are performed by the part of the NRSC that is completely independent of the RIAEC staff.

14.192 Section 6.2.4 of ANS-15.1 contains provisions for reporting deficiencies uncovered by audits and preparing and distributing audit reports. Proposed TS 6.2.4 doesn't contain any such provisions. Explain why no such provisions exist in the proposed TS, or revise the proposed TS to include such provisions.

Response to December 17, 2012 RAI Dated March 15, 2013 An additional Techmical Specification will be added to make it clear that deficiencies that affect reactor safety are to be reported to level I management. The proposed additional specification is:

6.2.2.4 If deficiencies that affect reactor safety are found, a written report shall be submitted to the RIAEC Commissioners within three months after the NRSC has completed its audit.

14.193 Section 6.4 of ANS-15.1 states, "procedures shall be reviewed by the review group and approved by Level 2 management or designated alternates..." Proposed TS 6.4.2 states that the NRSC reviews and approves procedures. Proposed TS 6.2.2.2 specifies that an NRSC quorum consists of a majority of non-RINSC and non-RIAEC members. These TS imply that a group with a majority of non-RINSC members gives the final approval for procedures. Explain why the approval of procedures is controlled by individuals outside the RINSC operating organization and not Level 2 management.

Response to December 17, 2012 RAI Dated March 15, 2013 Historically, as part of the review function of the NRSC, procedures have been reviewed and approved by the committee. Having it done this way prevents the potential for conflicts between what management is willing to allow in procedures, versus what the safety committee views as being safe. The facility Director and Assistant Directors are part of the committee, which provides an opportunity for administrative, operations, and health physics management to express their concerns, or rationale for their points of view regarding facility procedures. A provision is made for making minor changes without obtaining prior approval for the non-RIAEC staff committee members.

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14.194 Section 6.4 of ANS- 15.1 lists eight activities that require written procedures. The current TS 6.5 lists nine activities that require procedures and is consistent with the list in ANS-15.1. Proposed TS 6.4.2 only lists five activities that require procedures (proposed TS 6.4.2.1 through proposed TS 6.4.2.5). Provide a justification for no longer requiring procedures for the activities that are in the current TS, but not in the proposed TS (for example, surveillance checks, calibrations, and inspections required by the TS).

Response to December 17, 2012 RAI Dated March 15, 2013 This section of the Teclnical Specifications will be re-written to say:

6.4 Procedures 6.4.1 Written procedures shall be adequate to assure the safe operation of the reactor, but should not preclude the use of independent judgment and action should the situation require such.

6.4.2 The procedures for the following activities shall be reviewed and approved by the NRSC:

6.4.2.1 Startup, operation and shutdown of the reactor, 6.4.2.2 Fuel loading, utloading, and movement within the reactor, 6.4.2.3 Maintenance of major components of systems that could have an effect on reactor safety, 6.4.2.4 Surveillance checks, calibrations, and inspections that are required by the Technical Specifications, or have a significant effect on reactor safety, 6.4.2.5 Radiation control, 6.4.2.6 Administrative controls for operations, maintenance, and experiments that could affect reactor safety or core reactivity, 6.4.2.7 Implementation of the Emergency and Security Plans, and 6.4.2.8 Receipt, use, and transfer of byproduct material.

6.4.3 Substantive changes to the above procedures shall be made only with the approval of the NRSC. Temporary changes to the procedures that do not change their original intent may be made by a Senior Operator.

Temporary changes to procedures shall be documented and subsequently reviewed by the NRSC.

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14.195 Section 6.5 of ANS-15.1 states that new experiments and substantive changes to previously approved experiments should be approved in writing by Level 2. Proposed TS 6.5 specifies that the NRSC approves new experiments and substantive changes to previously approved experiments. Explain why the approval of experiments is controlled by individuals outside the RINSC operating organization and not Level 2 management.

(See RAI 14.193.)

Response to December 17, 2012 RAI Dated March 15, 2013 Historically, as part of the review function of the NRSC, experiments have been reviewed and approved by the committee. Having it done this way prevents the potential for conflicts between what management is willing to allow in procedures, versus what the safety comnmittee views as being safe. The facility Director and Assistant Directors are part of the committee, which provides an opportunity for administrative, operations, and health physics management to express their concerns, or rationale for their points of view with regard to proposed experiments. A provision is made for making minor changes without obtaining prior approval for the non-RIAEC staff committee members.

14.196 Proposed TS 6.6.1 and 6.6.2 state that the NRC will be notified in accordance with proposed TS 6.7.2 in the event of a safety limit violation or other reportable occurrence.

The regulations in 10 CFR 50.36(c)(7)(ii) and 10 CFR Part 20, Appendix D, require a licensee to make initial notification to the NRC Headquarters Operations Center. Revise the proposed TS to explicitly state that initial notification will be made to the NRC Headquarters Operations Center.

Response to December 17, 2012 RAI Dated March 15, 2013 Technical Specification 6.7.2 will be re-written to say:

6.7.2 Special Reports 6.7.2.1 Reporting Requirements for Reportable Occurrences In the event of a reportable occurrence, the following notifications shall be made:

6.7.2.1.1 Within one working day after the occurrence has been discovered, the NRC Headquarters Operation Center shall be notified by telephone at the number listed in 10 CFR 20 Appendix D, and 6.7.2.1.2 Within 14 days after the occurrence has been discovered, a written report that describes the circumstances of the event shall be sent to the NRC Document Control Desk at the address listed in 10 CFR 50.4.

6.7.2.2 Reporting Requirements for Unusual Events 229

Within 30 days following an unusual event, a written report that describes the circumstances of the event shall be sent to the NRC Document Control Desk at the address listed in 10 CFR 50.4.

14.197 Section 6.6.2(1) of ANS-15.1 states that reactor operation shall not be resumed following a reportable occurrence unless authorized by Level 2 management (RINSC Director). Proposed TS 6.6.2.2 states that the SRO can authorize restart of the reactor. Explain the reason for assigning restart authority to the SRO instead of Level 2 management.

Response to December 17, 2012 RAI Dated March 15, 2013 Section 6.6.2(l) of ANS-15.1 states that reactor operation shall not be resumed following a reportable occurrence unless authorized by Level 2 management. Section 6.1.1 of ANS-15.1 states that the idealized levels of the organization are:

Level I - Individual who is responsible for the reactor facility license (Unit or organization head)

Level 2 - Individual who is responsible for the reactor facility operation (facility Director or Administrator)

Level 3 - Individual who is responsible for the day to day operation of the reactor (Senior Reactor Operator)

Level 4 - Operating staff The job descriptions of the various titles associated with the organization chart are:

The Director is the organization head, and the administrator for the facility. The job title states that this individual directs the administrative and technical programs at the facility on a day to day basis. Consequently, this is the individual that is responsible for the overall license of the facility.

The job descriptions for the Assistant Directors indicate that they manage the operations and radiation safety programs at the facility. These are the individuals that are responsible for the overall operation of the reactor facility.

The Reactor Supervisor job description says that this individual supervises all phases of reactor operation, and is responsible for the operation, maintenance, and calibration of the equipment associated with the reactor. This is the individual that is responsible for the day to day operation of the reactor.

230

Therefore, based on the organization chart that is in the current facility Technical Specifications, the individuals that would be associated with each. of the ANSI organization levels would be:

231

As a result, level 2'management in the organization is at the Assistant Director level. Consequently, proposed TS 6.6.2.2 will be changed to say:

6.6.2 Action to be Taken in the Event of a Reportable Occurrence Other Than a Safety Limit Violation 6.6.2.1 The Senior Reactor Operator shall be notified promptly and corrective action shall be taken immediately to place the facility in a safe condition until the cause of the reportable occurrence is determined and corrected.

6.6.2.2 The occurrence shall be reported to the Director or Assistant Director.

6.6.2.3 If the reactor is shutdown, operations shall not be resumed without authorization from the Director or Assistant Director for Operations.

6.6.2.4 The occurrence, and corrective action taken shall be reviewed bv the NRSC during its next scheduled meeting.

6.6.2.5 Notification shall be made to the NRC in accordance with Paragraph 6.7.2 of these specifications.

14.198 Proposed TS 6.7.1.7 states that the annual report will include, "a summary of annual radiation exposures in excess of 500 mrem received by... visitors." This appears to be inconsistent with the requirements of 10 CFR 20.1301, "Dose limits for individual members of the public," and 10 CFR 20.2203, "Reports of exposures, radiation levels, and concentrations of radioactive material exceeding the constraints or limits." Specifically, 10 CFR 20.2203(a)(2)(iv) requires a written report within 30 days after learning of doses in excess of the limits for an individual member of the public in 10 CFR 20.1301. Explain how the proposed TS meet the regulatory requirements, or revise the proposed TS, as appropriate.

Response to December 17, 2012 RAI Dated March 15, 2013 This specification is with regard to the annual report that is sent to the NRC once per year, and gives a general overview of the facility operations for the year. It does not preclude special reports that are required by regulation, such as the reporting requirements of 10 CFR 20.

14.199 The regulations in 10 CFR 50.36 require that records of the results of each review of exceeding the safety limit, the automatic safety system not functioning as required by the limiting safety system settings, or any limiting condition for operation not being met be retained by the licensee until the NRC terminates the 233

license for the facility. Proposed TS 6.8.1.3 requires records of reportable occurrences be retained for five years. The regulations in 10 CFR 50.36 require some records categorized in the proposed TS as records of reportable occurrences to be retained for the life of the facility. Revise the proposed TS to include a requirement that records of the results of each review of exceeding the safety limit, the automatic safety system not functioning as required by the limiting safety system settings, or any limiting condition for operation not being met be retained until the NRC terminates the license for the RINSC reactor.

Response to December 17, 2012 RAI Dated March 15, 2013 ANSI 15.1 section 6.8.1 (3) suggests that reportable occurrence records should be kept for a period of five years.

The proposed Technical Specifications will be changed so that reportable occurrence records are kept for the life of the facility:

6.8.1 Records to be retained for a period of at least five years 6.8.1.1 Reactor operating records, 6.8.1.2 Principal maintenance activities, 6.8.1.3 Surveillance activities required by the Technical Specifications, 6.8.1.4 Facility radiation monitoring surveys, 6.8.1.5 Experiments performed with the reactor, 6.8.1.6 Fuel inventories and transfers, 6.8.1.7 Changes to procedures, and 6.8.1.8 NRSC meeting minutes, including audit findings.

6.8.2 Records to be retained for a period of at least one certification cycle Current Reactor Operator re-qualification records shall be maintained for each individual licensed to operate the reactor until their license is terminated.

6.8.3 Records to be retained for the life of the facility 6.8.3.1 Gaseous and liquid radioactive effluents released to the environs, 6.S.5 Off-3it- enviromnental monitoring surveys, 234

6.8.3.3 Personnel radiation exposures, 6.8.3.4 Drawings of the reactor facility, and 6.8.3.5 Reportable occurrences.

235

The following RAIs relates to financial qualifications:

November 24, 2009 RAIs

1. The U.S. Nuclear Regulatory Commission (NRC) staff will analyze the financial statements for the current year, which are required by 10 CFR 50.71(b). to determine if the applicant is financially qualified to operate the RINSC. Since RIAECs financial statements were not included with the application, please provide a copy of the latest financial statements for the NRC staffs review.

2. Pursuant to 10 CFR 50.33(f)(2), the regulation states that "[t]he applicant shall submit estimates for total annual operating costs for each of the first five years of operations of the facility." Since the information included in the application is now out of date, please provide the following additional information:

(a) Projected operating costs of the RINSC for each of the fiscal years (FY) 2011 thru FY2015 (the first five year period after the projected license renewal).

(b) Confirm that RIAEC's primary source of funding to cover the operating costs for the above FY will be funded by an annual appropriation from the state of Rhode Island's budget as described in the application.

Second Response to RAI Dated November 24, 2009 Submitted January 4, 2010 236

STAT19 OF RHODE ISLARO AND PROVIDENCC PLANTA'TONS O RHODE ISLAND ATOMIC ENERGY COMMISSION Rhode Island Nuclear Science Center 16 Reactor Road Narragansett. RI 028821165 Mr. William B. Kcnncdy, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D.C. 20555-0001 Januarv 4, 2010 Re: Letter dated November 24, 2009 Dbcket No. 50-193

Dear Mr. Kennedy:

We are responding to the deconu-missioning aspects of the subject letter in a separate commumicadion to you from the Chairman of the Rhode Island Atomic Energy Commission (RIAEC). 'Ibis letter addresses questions relating to our financial resources to continue safe operatidn of the Rhode Island Nuclear Science Center (RINSC) during the requested twenty year license renewaL period.

The Rhode island Atomic Energy Commission's primary source of funding to cover operating costs is provided by an annual appropriation from the Rhode Island Legislature.

Other sourves of revenue include federal grants and payments from the University of Rhode Island for radiation satety services. The latter is used primarily to cover personnel costs under an historical agreement with the University of Rhode island. Personnel costs account for approximately 85% of the total cxpenditure.*. The General Revenue numbers shown in the table approximate annual operating costs- All other sources of rev,.enue are sent to the Rhode Island General Fund and returned to the RINSC as needed.

For Fiscal Year (FY) 2011, the RIAEC h*s requested S984, 116 from the general revenues of the state of Rhode Island. For Fiscal Years 2012 through 2015, the RIAEC expects an annual increase of approximately 3% to that appropriation. That projection is based on historic data that generally shows that annual growth.

Aoc o 237

Ihe following table provides both historic and projected financial data.

Fiscal Year 1Total'Expendicures General Revenue Other 2007 __$!.87,486  !,$827,654 $259,832

$1,474,56t 42008 1 S834,.1.0. $640:460 2009 , $1,183.832 $786,847 $396,985 2010 1[1.140.115 $775,346 .$364,769 2011 I$1,500,685 $884,116 [$616,569 2012 S1.545.706 $910,639 $635,066 2013 i$1,592.077 $937,959 $654,118 2014 sl..39:839 $966,097 $673.742 2015 $1,689,034 $99508.0 $693,954 If you hayc additional qutstions. please cozi1act M.e.

Verv truly yours, TermT' eh,J NbDirmcor Rhode 166.d Atomic Energy Commiss~ion I certiry wider peralty of dury that the m-presenatio Male aw Var trueand cu. ct.

Ex"cuted on: ~ oO By:

Auxaemtirms: RIAISC Budgets 238

STATE OF RHODE ISLAND ANO PROVIDENGE PLANTATI"nS j5P RHODE ISLAND ATOMIC ENERGY COMMISSION Rhode Island Nuclear Science Center 16 Reactor Road Narragansetn, RI 02.82.1165 October 28, 2009 The Honorable Governor Donald L. CarcieTi Office of the Governor State House Room 143 Providence, RI 02903

Dear Governor Carcicri:

This letver transmits reised pages to the Rhode Island Atomic Energy Commission (RIAEC) FY 2010.,2011 Budget These pages reflect a full 3% raise in FY20 ! and do not reflect unpaid dmas in 2010 or 2011 and the half year delay in the pay raise.

.STicrclY.,p P

Enclosure:

FY20 11 Budget Copy to: Budget Officer Senate Fiscal Staff House Fiscal Staff 239

Rhode Island Atomic Energy Commission FY2010 FY2011 Grad FTE Cs Ex - Cost Classified Director 0150 A 1.0 151.900 1.0 156,364 Assistant Director for Operations 0139.A 1.0 81,982 *1.0. W0.350 Assistant Director for Reactor Safety 0139A 1.0 67,061 1.0 92,793 Reactor Supervisor. Nuclear Science Ctr. 0132 A 1,0 -67,592 1..0 71.787 Senior Facility EngineeT 0132 A 1.0 56,865 1.0 60715 Health Physicist 0130 A 1.0 64,783 1,0 69,833 Principal Reactor Operator 0124 A *1.0 1.0 57,123 53,470 0.6 20.621 0.6 21,577 Senior Word Processing Typist 0109K Subtotal 7.6 $584,275 7.6 $620,842 Unclassified lnformation Systems Specialist 0816A 1.0, 36,565 1,0 38,703 Turnover (68,405)

Total Salaries &S8 $56012 8.6 $592,840 Benefits Retirement 126,436 128,239 Medical 82,258 65,568 FICA. 42037 41.274 Retiree Health 32,837 39.957 Payroll Accrual 3.009 .3,242 Total Salaries and Benefits 8.6 $851,489 8.6 $891,168 Cost Per FTE Position .99,010 103,624 Temporary and Seasonal 12,000 12,000 Statewide Benefit Assessment 26,527 25,315 Payroll Costs 8.6 $590,016 816. $928,483 Purchased Services Training and Educational Services $1.2,257 $12,257 Security Services 2,691 2.691 Buildings and Gr6unds Maintenance 43 43 Information Technology 167 167 240

Rhode Island Atomic Energy Commission FY 2010 .. FYZ2011 fi~d .EIF. Cost Total .$5,158 $15,168 Total Personnel 86 $905,174 8.8 $943,641 Distribution by Source of Funds Gem~ral Revenue 7.0 709,059 7.0 738,273 Federal Funbds

• 10,000 10.001)

O~ther Funds 1.6 186.115 1.6 195.368, Total: All Funds 8.6 $905,174. 8.8 $943,641 241

4, State of RFod. Island

• RHODE ISLAND ATOMIC ENERGY 0.1.r. COMMISSION Line Sequence Sunimnaxy AoW*y 0M 00400 MAWM0ATOMIC ElKRGY 0000.3010

" 20M $2M0 F FY2010 IN77000(tFo21 Nab,0 AMU*~. A.Wak" Acbuab. E04840 Awoot. I=.

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~'romobcmeg". I0 1110.5 01072$27 "9024T.1 sim0.11 S150300 Catawy: 40 Owe"01 S.71740 oodev,0 040100 0408019Mbftnono 8, ft941 10 so Ww00 PAN~ 02.01 640200 On.'aw. 00~n 00b so so2.00 S2.5000 02.00 643200 a F-0.40. $15.140 so 014,500 03.316 $3.314 042010 0UoFw : ppop.0*Ixos.,y so so 114.20) 0P,200 00.20 544030 C.l*s U46.0 I'.O, so so 02,472 3$00345 05W345 04402 Ckv: 0oo004a $27-M0 to 30 so to 242

SI.a of RIOd IMad SC

• RHODE ISLAND ATOMIC ENERGY

. COMMISION Line Sequence Smt!ma#y

?y 20M Iy 2m Fy 20,0 yýl IY21 W.b.,W Ao....t Adi.'.Cd RAoob P~, 7 Aitom Erpqy Co.mwvoC.q4 TOW lot 4 C"W W~.632 so $*5,752 $7.91it .1 41 L" SPONSOREDRESEARCH11027 1.4 4604 37.0 243

.-. StaoIooRtxo~IoWaW SRIIODE ISLAND A TOMIC ENVERGY SCommISiONl Line Sequence Su~nmnry A64~y 02 RHOMIQIX 139ATOMIC ENERG2Y CMOS.I06IN Fy "M9 FY 2M0 02010 r 2010 " '2011 14l.A9otLA100C A10114 16dN 96*4 I'lgr.W 07 AI6*o I g7 C-ft0900 Ujfs.q0.6.66 2910101 R I ATV.94CCHU0Y62C0~AVM61550 0.b.wv0 10Is 0...14.61 6110 low ftg w o.. $21M0,0 $419,241 S404.440 14$0 47$,1722 616200 M.4i.1 V.-1~ 96.W. 8-~o s2.0 $4.300 16.0 34.004 $4,00.

619W r.aolAoWW to so 12,484 $2,441 MO.60 am20100 w-Wmm ~a RM01 .. S1107=3 0105961 190.10 $101,410 $100,260 629110 79:9 S."~ 06.wky $4,9"0 M52430 $0 00 so 921129 PLCA9.6416 so 30 Mol.l4 =0.364 034.00 62-M V90 a 14496196600 $332-,W 007.03 1406 947,10 $49,062 IMA1M 06161 o,... $4,942 04.294 0.2.097 03,46 92,402 624130 140n6.0 3790 900 11707 We0 996 620100 A-eo l 710"ft6.19 :10.696 19.022 $0,9232 ULM70 =.296 4n20200 I--.l N0667."1 62901s $27.010 1.75 129,M46 M2.102 1070200 Us1.o6L"-.0y 0009 soI so so . 90 TOWS. C* 94 43T to smm29 9679.09 3m6.100 17.01,0 $722,11 0.a9110.9. No146.11.99 032160 Tr ySU1101000 $0 00 so so 630900 T1-.4n .1N99.91906119 $1,023 sm6 so $43 602006.96199 sl 7601osw6110 Si224 $0 so $119 31.141 633 GsMx496"qW*.U 042 to t0 63MM S6.006 0.1*6 31.941 51.76 02,062 11's,, $102 02960 $116 Pro*="0 $WM10 3925 s70 to $75D 244

• RHRhODE ISLAND ATOMIC ENERGY

• COMMISSION Line Sequence S,~mrnary ArPLT 08 11300D$MSANDATOWO C04118101 EKVPPGY i FV Me8 FY 2DO 'Y 2040 Fl88IO yOl H.I.381AmW:U AdkWIS, AdmoK Emed R.*..4 RoMw*

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$1,210 SIA04 52060 64040M36*4401064* so so 041208M .. ei48. Con*4A EnUIP so 1031 10 6311 Il,382 so 21.302 0414600 M.,10140W 4O. 6846 £33 $23 so 0208 1225 "43.w S.M4yE.Pm., s1.857 to so $1,887 11,857 643110 of41400opp6i & Equip 64.745 $L976 so

$2,000 643120 Cet1o suo911*48464 & Eq.* "m3 30 SIA42 $10 Mi Ii.068 64337 J406041 4 Bop$Im 9 Equip $500 640140 K6141160mdWW*46 SuWhed.L Equip so to

$4.238 1116 to 643150 pawa% opo 4Elmb $1800 0431680t.I Supl 8,416$.. $3.055 so S11,00 "6440 68iWlWo60U.W4Supplies &E" sti.431 "'no3 6432M8 134.4411 Fbet so 1038 3554 843300 Ub.oI* 3384 SIM

$10 so 643419 Posts" 144 #040114&M SI 067 S700 6434710PlFFF" so SSW0 12000 1004 643611 Pl44tdo40him so so 060 443M2 pml*,g - Da. V.40.1 338 so $238 644V0 Lkm..I *. En -444 63m 3490 so 680 843710 St1.10,".14 3138 $0 so $M3 245

Z j"rFODEs ,LAvDATOMIC ENERGY 0:.96.: COMMISSION Line Sequence Slfmay A"Mo, 952RHODE0 ISLANDATOMICENERGYCOMMISSIONF 71 f76F 00FSI Y3l N4.4a400901 Ad.140. AMool: E04dot 01.Wd; 0041 Pmw4 13 07 ~ .o.wC,4~

843910 hn"ý.o Pro.'l1#C.Ov3 $4.490 11,2011 s4.2 $440 4.490 *0 6.43=5 30o Cýa Reim* SUM90 1,096 30 90.301 "t00

$423m "toe08 suu (".4u) 30 $0 $1.76D 11.760 11.790 84MI3 Cý"O Pbwo.o Ph.no.o.oWid s0 so $190 1190 $10190 41443M P0441No" G0I8 $113 so SD $113 3115

$44400 Pod' G.sol,828..IFLAI 1.410 3547 U3m 3 50 044&V0 CA,0341j404.wd so so VWW50 9 so 1144M5 (105 Oo..4..d S10.32 $91.712 so so so 6445M Rmlt3o0L90. Etýso, $4,240 $140 S1m.6 S3.900 53.000 04471 $.-I34'T~o4..0 27 so s0 377 027 447180 DOMtHfd 3.675 so so "IS3 "I's 640110 78081915F fCEI4TREX) so so $35.30 15,160 $1. Ia0 GAII012Td39061 O60*4 S10 0143 to *704 074$

TOW. WC*Ogqol 40 Opuo30q Sum""o uoa emqdr 1 7 $100,034 300.10 us4."? smow R IATOMICENERGYCOWS4S0ION 0.11 74.0 7.4 .3.409010 246

~\ STATE OF RHODE SA*SNDAND PROVIDENCE PLATATIONS RHODE ISLAND ATOMIC ENERGY COMMISSION Rhode island Nuctear Sciene Center 16 Reaclor Road Natraganseft, RI 02882:1165 September 5, 2008 The Honorable Govemor Donald L. Carcieri Office of the Governor State House Room 143 Providence, RI 02903

Dear Governor Carcded:

This letter tansmits the Rhode Island Atomic Energy Comiission (RLIAEC) FY 2010 Budget Request. We have ,instituted nu*merous ost saving measures to meet the budget reduction targets for 2009 and 2010 which include eliminating state funded maintenance, computer equipment and supplies, training, outside printing, out of state travel, ove.-dine and equipment.. We view these cuts as deleterious to the long-term future of the Rhode ýIsand Nuclear Science.Center. We have utilized federal funds, where legal to ofr-sct shortfalls in state funding. Unfor.unately, LhlDeparunent of Energy dropped all support for research reactors this year and it will be at least a year until funding is restored. We have offset this funding loss by instituting an overhead charge on the UnLversitv of Rhode Island sponsored research account to make up for revenue shortfalls.

Still, it will be difficult to ichieve the required personnel turnover needed to meet budget targets. Any additional cuts to. personnel or basic operating budget items will result in a situation where we may not be in compliance with state and federal regulations regarding the safe operation of the fI cipate the need for supplemental funding during the 2009 and 2010 budget years due to the loss of federal funds. A Supplemental Appropriation Request will be submitted when we have a better feel for the federal finding picture.

Sirncerely, Ten'y $Ta Ph.D, Direclt AEC

Enclosure:

FY20 10 Budget Copy to: Budget Officer Senate Fiscal Staff Houwe Fiscal Staff 247

248 TABLE OF CONTENTS NARRATIVE INFORMATION Agency Description ....................... ........ ... ................................................................................. Page 1 O rganization. Chart .................................................................................................................. ................. .. Page 7 AGENCY

SUMMARY

Program Summary ......... .......................................... ....... Page 9 REVENUE ESTIMATES AND DATA Estimated Departmental Revenues ......................................................................... ........... ........... Page 10 Receipt Account Information ............ .............................................. Page 11 PERSONNEL SUPPLEMENT DATA P ositions ...................................................................................................  :.................................................. P age 12 Total Payroll .................................................................................................................. Page 13 SOC DETAIL RI Atom ic Energy Com m ission ........................................................................................................................ Page 14 BO C Detail Backup inform ation...................................................................................................... z................ Page 20 FEDERALIRESTRICTED ACCOUNT ESTIMATES 2915101 ....................................... ............................... . . .................... ....... ...... ........................... . Pag 21 249

2915102 ....................... ............ ;,.................................................................. ................... ........................... Page 22 2915013 ........... ..................................................................................................................... Page 23 2915104 ................ ;.......... ........................................... :...... :..................... ............................... Page 24 PROGRAM PERFORMANCE MEASUREMENTS Commercial and Research Use Availability ............................................... Page 25 MinbrityiFemale as a percentage of workforce .............................................................................................. Page 26 250

BR-10 Narrative Information Revised Bt-24-08 Agency: Rhode Island Atomic Energy Commtssion (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center PROGRAM TITLE: Rhode Island Nuclear Science Center Operations PROGRAM EXPLANATION:

The Rhoda island Nuclear Science Center (RINSC) Is used. for medical, biological, environmental and materials research.

education and commercial activities. In addition, the staff runs the radiation safety program for the University of Rhode Island. The Director serves on the State Radiation Advisory Commission and has takon over responsibility for low-level radioactive waste disposal activities from the Department of Environmental Management. In this capacity, he serves as the Governor's Representative to the National Low Level Waste Forum.

The Center's state-of-the-art analytical laboratories and equipment are currently being used for several environmental monitoring projects. A new classroom and educational counting laboratory were completed this year with federal grant funds. Local colleges are utilizing the facilities for classes and student projects. The facility continues the process of improving security and upgrading reactor equipment that have been financed through Department of Energy (DOE)

Grants. A high security gate has been installed at the rear of the lab building and work is proceeding on improving security in the rear of the facility.

In 1993, the reactor was converted to a new low enriched uranium fuel system that has greatly reduced security requirements and associated costs while providing a significant Improvement in performance. The use of grant monies has resulted in the addition of upgraded mechanical and electronic equipment necessary to substantially Increase reactor capability. These improvements will permit the RINSC to compete successfully for production of medical isotopes, will provide the neutron flux to do meaningful research in neutron scattering science and will provide the necessary neutron flux to conduct Neutron Capture Therapy (NCT) that Is a promising new method of curing deep cancers.

This year, RINSC upgraded the reactor control systems using industrial automation software and hardware. The rod controls system was replaced with controllers, encoders and new stepper motors. This project has Increased the reliability of the control system and allowed removal of older controls, meters and chart recorders.

RINSC and University of Rhode Island have upgraded the large anglo neutron diffractometer with new computer controls from Brookhaven National Lab and data acquisition with monies from the Department of Energy. Innovations in Nuclear Education and Intrastruclure (INIE) Grant. RINSC was awarded an additional DOE. National Energy Research Inflilute (NERI) Grant that provided significant improvements to our neutron scattering science capability. The custom electronics and computer system built by Oak Ridge National Lab for a second diffractometer will allow RINSC to complete with the Page I 251

BR-1O Narrative Information RevIsed 8-24-08 Agency: Rhode Island Atomic Energy Commission (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center National Labs and other universities for neutron science research initiatives like the Global Nuclear Energy Partnership JGNEP).

RINSC and University of Rhode Island were awarded a grant in collaboration with MIT and other universities for GNEP activities. Part of this DOE grant will provide funds to URI in support of starting a Minor in Nuclear Engineering in the Spring 2008. As part of this effort; RINSC is converting available space into a dassroom and another place into a multiple station counting lab. Spring 2008, URI will teach a Nuclear tpitection and Measurement course at RINSC. RINSC has purcha.,,ed six singla channel analyzers for the laboratory experiments associated with the course. The monies for this equipment will come from the DOE sponsored INIE program. Providence College Physics Department will also make use of this new Laboratory Equipment for their Modem Physics Course.

RINSC has completed an educational license agreement with IndoSoft Ltd. RINSC will be able to use IndoSofl software to provide remote classroom training. RINSC and URrs work with neutron scattering science will be made available to the URI Physics Students on the main campus through this program. The program Is expected to provide a remote classroom at the URI main campus. Remote training wiu also be made available to Providence College and Brown University providing remote laborutory training for Neutron Activation Analysis.

Engineering and design and fabrication work is currently in progress for the construction of a cancer treatment facility that is being funded by Dr. Karl Oil of Purdue University and a commercial Company. Researchers at Brown University and RINSC have a collaborative effort with the Massachusetts Institute of Technology (MIT) to develop a successful treatment.,

for one of the most deadly forms of brain cancer. The results of this work has allowed Dr. Leith associated with RINSC and Brown University io apply for a larger NIH grants to support and continue this work.

Dr. Alex Pszenny is currently conducting a major study of pollution outflow, which has expanded from only the Northeast United States to the West Coast and Cape Verde off the African Coast under the multi-agency CHAOS Program. This study uses the reactor to analyze samples for Bromine and Iodine as a means of determining Ozone depietion and its effect on pollution values. The neutron activation process for trace metal analysis has been shown as a viable method to further improve the pollution model and funding has been requested to continue this type of analysis. This project is a long term effort that will extend for several years.

Or. Jennifer Perry at Pomona College..AnthropologyDepartmnent and URI, Graduate School of Oceanography (GSO) are conducting experiments with prehistoric stone samples. A graduate Student from GSO Narragansett Bay Campus is participating in this new approach to use neutron activation analysis to categorize prehistoric stone toots.

Page 2 252

BR-10 Narrative Information Revised 8-24-08:

Agency: Rhode island AtomicEnergy Commission (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center A new building for the production of sensors for weapons of mass destruction has been comnpleted and Is being utilized by SubChemn Systems to do cutting edge research in underwater sensor development. This company has several grants from economic development agencies and utilizos several URI students in the research efforts.

RI Consultants has constructed a laboratory in the facility for. proof of concept work in the area of radiopharmaceutical applications. They recently developed a new method of utilizing radioisotopes to prevent clogging of arteries after angioplasty and new targets for accelerators to produce specialized radiclsotopes. They are currently developing a new phosphorus radioisotope with a company in South Africa. RINSC has given RI consultants authorization to perform .QA work on the new phosphorus radioisotopes being developed in South Africa.

BioPAL. Incorporated Is making extensive use of the reactor to conduct analysis of medical samples for a variety of treatment and research purposes Including the area of stem cell research. They have pioneered the use of neutron activation analysis (NAA) to provide an alternative technique tor In viva cell tracking and quantifying studies. .ioPal uses non-radloactlive microspheres for measuring regional blood flow, gastrointestinal motility, particle handling by the pulmonary system, health of the liver and injected protein distribution. Nano-materials can be used to label and track stem cells. The microspheres have up to six different labels, that can be examined post experiment using NAA to identity the labels. No other technology can provide quantitative cell labeling in such a simple and non-intrusive manner.

Infascitex .tnc. is utilizing the reactor to develop a method of hardening glass for use In high performance solar cells.

Future work will Include lazer analysis with URI.

RINSC Is located at the University at Rhode Island. Bay Campus. in Narragansett.' TheCenter contains a state-of-the-art nuclear counting system, laboratories, a mass spectrometer, a class 100 'clean room and facilities for handling and storage of radioactive material. The Rhode Island Nuclear Science Center has operated on a daily basis without incident since 1062, STATUTORY HISTORY' The Rhode Island Atomic Energy Commission was established in 1958 by Title 42, Chapter 27 of Rhode Island General Laws for the purpose of advising the governor and general assembly on matters relating to atomic energy and to construct and operate a nuclear reactor in the state.

Page 3 253

BR-10 Narrative Information Revised 8-24-08 Agency: Rhode Island Atomic Energy Commission (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center PROPOSED BUDGET YEAR OPERATIQ_:

The current level of operations and maintenanceof the facility fliclude:

.1. Operation of the reactor and associated research facilities to support projects in the areas of neutron activation analysis, neutron scattering and neutron damage studies, medical isotope development, and commercial Initiatives and cancer research.

2. Efforts to meet fede ral security requirementswhich continue to become more stringent

3. Operation of the URI Radiation Safety Program.

4. Tours and briefings for educational institutions and other interested individuals and groups throughout the state.

5. Technical and administrative support for the low level radioactive waste program and coordination with other state and federal agencies in the areas of nuclear technology.

6. Classroom and laboratory training for educational Institutions utilizing the new facilities at RINSC The current initiativesof the facility include:

1. Upgrading of two beam instruments to Support educailonal and research initiatives by the URI Physics Department 2- Increasing the classroom and laboratory space and equipment. Suport of URI efforts to provide a Nuclear Engineering Minor.

3. Development of a now cancer treatment protocol using gadolinium in cooperation with Brown University and MIT.,

4. Environmental monitoring studies for various state and Federal agencies. This work will be done by Microinorganics inc. This company has two laboratories at RINSC that specialize in trace metal analysis of tow level environmental samples.

5. Completion of socurity.upgrade projects required by the Nuclear Regulatory Commission.

6. Continued participation with State and Federal agencies to identify the most cost-effective method of ensuring safe disposal of low level radioactive waste generated by hospitals and universities throughout lie state and technical input to State agencies regarding current problems with the disposition of high-level nuclear waste in adjoining states.

7. Continuation of educational programs funded from the Department of Energy Reactor Sharing Grants and submission of new grants for graduate student aid under the Nuclear Engineering Education Research Grant Program and submission of new grants to the National Cancer institute to support the NCT work in progress.

Page 4 254

BR-10 Narrative Information Revised 8-24-o8 Agency: Rhode Island Atomic Energy Commission (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center

8. Efforts to increase the scope of commercial work wili continue in the area of isotope production, radio-pharmaceutical production, and cancer research and therapy.

LONG-TERNM TREN ,DS:

The new methods for medical and biological diagnostics developed by BioPAL, incotporated have significant commercial potenlial. The response to their products has been very positive on an International scope. They have invested a significant amount of money in advancod analytical equipment that has enabled them to meet the steady growth in demand. The significant capabilities of a research reactor in the area of activation analysis has been the other reason that they have been able to grow so rapidly.

Continued research at the Massachusetts Institute of Technology and other facilities worldwide in the use of Neutron Capture Therapy to cure brain and skin cancer holds real promise. for utilization of the RINSC reactor in the field of cancer treatment. A research reactor is required to produce the necessary neutron beam and the future improvements to the RINSC beam will make it one of the few facilities that can perform this type of treatment. A fifteen-year lease has been signed with the Neutron Therapy Company to construct a stateiof-the-art treatment facility with a highly advanced neutron filter. Also, In collaboration with Brown University and the Rhode Island Cancer Council. a team was formed to develop a treatment capability in Rhode Island. Current etforts focus on the development of gadolinium as the active agent In this therapy. A multi-year grant from the Department of Energy has been awarded to RINSC and MIT for this work and grants have been submitted to the National Institute of Health for funding of this research.

Continuing efforts to upgrade equipment will ensure the safe and reliable operation of the reactor for many years. The Center for Atmospheric Chemistry Studies building (CACS) that is located adjacent to RINSC provides much additional laboratory and office space for faculty and students that utilize the reactor for research. The combination of a modem reactor facility, and the CACS building provides the support necessary for a worid-cdass environmental research program.

The most disturbing long-term trend Is the lack of staff to support new initiatives. Previous budget and staff cuts combined with retiremenls and loss of experienced operators to higher paying industry jobs has made it difficult to handle the significant Increase in research work arnd commercial activities. The staff has continued to function because of the significant experience level and technical expertise of the remaining personnel. However, the extremely small staff and significant maintenance, administrative and watch standing demands associated with daily reactor operation limit the amount of additional new work that can be undertaken. Subsequent to the events of 9/11. the staff has been required to spend much lime meeting nuw federal requirements in thearea of security from terrorist attacks.

There Is currently no funding In the budget to support startup costs associated with commercial projects. Staff cuts made in the last few years has made it impossible to run the reactor more than one shift a day. A longer operational time Is required to economically provide neutrons for potential new commercial operations. Due to continued budget cuts and Page 5 255

BR-10 Narrative Information Revised 8-24-08 Agency: Rhode IslandAtomic Energy Commission (RIAEC)

Program: Operation of the Rhode Island Nuclear Science Center loss of researchers in the area of environmental chemistry, RINSC has been forced to terminate support for other state agencies in the area of radiological emergency response. In the event of a major radiological accident, RINSC would not be able to meet the technical and legal requirements that are necessary to do this work. Budget cuts have also necessitated the termination of support for other state agencies. in the area of orphaned radioactive materi.. Such material is extremely expensive to dispose of and there is no funding for this program.

The recent loss of the Assistant Direclor of Operations to a federal research reactor which doubled his salary highlights the fact that current pay scales at RINSC are not competitive with the resurgent nuclear Industry. Several senior staff members are close to retirement and it will be very difficult to hire qualified replacements.

PERFORMANCE MEASURES:

Since the charter of the Center is to support education and research, there Is no .simple indicator .that best reflects performance. Much of the research and training is unique and not easily quantifiable. The, large amount of effort that is.

expended on the URI safety prograam Is not captured in these performance measures and the large amount of. time and effort expended in project development Is also missed, However, two statistics have been developed to capture the operational performance of the reactor.sinco this Is easy to quantify. The number of samples irradiated by the reactor provides a simple measure of. reactor usage and Is therefore provided, The current operating hours of the reactor as compared to the historical record of seven hour days/ 5 days a week provides an indicator of the effects of reducod staff.

and funding while, also, reflecting the potential of the reactor to Increase research support should Increased support-be obtained.

-. Sample-hours: This parameter Melates to samples that are irradiated in the reactor and counted on the Center's computer system after being processed In the various laboratories. Reactor operating Hours: This parameter compares current operatlons to the historical level of 7 hrslday for 5 days/wk.

2. Reactor operating Hours- This number compares current operations to the historical level of 7 hrs/day:.for.

Sdays/wk. .

Page 6 256

FIGUREOrganization RIAEC r,-I Chart TECHNICAL SPECIFICATIONS Dorket 50 .193Licento R-95 Rhode Island Atomic Energy Conmmission RHODE ISLAND

---

ATOMIC ENERGY ---------- Lin. of COMMISSION (RIAEC) Communicatlon*

• IRADATIN SFET *DIRECTOR qD CANCER COUN( CIL

---

---

1 -----

II ASST.

• DIRETIOR COR F AST DIRE.CTOR FOR C"I'.

• ~ ~ P&IATO &RECO RE ACTrOR II FCLT E AR TYPIST SSTMS "OERTIN OPERATORIN I ..... l *' E R .:* _// _ RErC O " : " FACILcL IT * "

N ILNT .. 1FAC

"

• H.,l PEiA E EC RO I S UPERVISORCA EN.S IN E O

,Epage 7

HEALTH

Prepared by tenry tehn 913,2008 Agency Summary A,*..

R{Atoini. EP.aegyCoi-aon IExplanation FYZ 2007 MYOOR kY2009 , FYZ009 FYZOID spent Spent " C£awed Revised Request Expenditures ty Program*

Atonic ueig Coanission 1,095,360 1,087.498 1,491.463 1,201,04" 1.043M921 Total Exzperditurce 1,095,360 1,087,458 1,481,463 1,201,048 1,043,921 E E!XprFmdiluIesbyCategory salazy/rag03b5.lts, 855.896 888,855 927,759 995,057 1,019,137 Contctd smic. 12,031. 17,672 22.000 ."0

.j..Ung .upiliý M mpe 211,207 161.58a 455.764 175.942 1U0.419 A.sinLamcc, Orants and [amfits 0 0 20,940 .92,937 30,000

.Sohbli OpM.rtOi -qspoarea 1,079.134 1,068,1 5 1,424,463 1,263,936 1,230006 Caita] Inprovcmwens 16.226 19,313 55.0*00 50'D00 50,000 turIover-s*%pe:ail 'y "* 0 (11 82,0 2) (236,085)

Total Expenditures 1,095,360 1..087,483 1,481,463 1,201,048 1,043,921.

PaSgi 258

Piepred bytaiy tehan9!312O0ORae Page 9 Agency SummaryAgny PI Atomic FUncgyComnuttaion lExplavation lrY2O07 . VY2008 Y21 FY2009 MO1  :ý i spent Spent E~nacted Rm~lsed Requtst Elpeodkimcs by Pawld StateGenerf.Ilcvrne 799,460 927,655 819,869 824,470 . 744,492 PedolGeams 136,215 101.942. 365.940 920.37 . -

Restricted Reccipts OtheT(List Eucit): URI I 59,f)8S 157.891 240,4554 233,641 249.429 Capital )rujccts 0 0 55,000 50.000 50.000 Total F-zpendl(twel .1,095,360 1,097,489 1,481,463 1,2011,001 1,043,921 FTE Positions 8.6 8.6 8.6 11.6 .8.6 PagnO9 259

[FORM BR-lý.fmztedO~p~rtentI Revones ___ _

RI ATOMIC ENERGY COMMISO _______ ___

MON30 2007 FY20O8 FY00 FY2009 012010

,~.i Aetual AC1..i ac~I BM . .. requoli RIC~utfl.L(~ 50da; 000

-fikwaI o5 rs3o .bdled No4Tlhroopy w-- fo Cor03rnr(labre~l S 1

$__ __ 31 $1+ 511 1 7 BPA~nf~e l) 20513 51.00 13,321 2.0,0W 122481 20.0001 3 I ý hvn vt- Inc'3 (0AI&II 13t': ,i ~' St 251101 ~~340,~020,OIDE03IOI357.00 S35,00 535.500 $50.'011 so 3 so_

~~. F" $45.10001 343.523 43.5"5 35ý-0.000 100.00 1 20033 J eo~~3.o50r $100.000 1 $5.300 105,0001 $03000n(I___ Sol 50 0q~ -iq4- 0:1 SBI2 $0 so i w~~eheed(SW1Min 02008l So S __ $0 762-$632,4751 U2.4751 I

0 ralV1~l iboraota , (558.549.2 rev~~sed. Mkro.Ioar3)o~~ks isa p)eni dam~rngedueto sp.ghinke, sysemn failura which dama erdqduceapo dteodIiellbrab-IA~rrflogenani so(lectfo manta110inin -633--f _ -- p --

PogO10 260

Form BR-9 Receipt Account Number: 29100981 (28-14-605)

Receipt Account Information Receipt Account Title: Stale Greeral Account R.I.G.L

Reference:

42-27-2 Description of Receipts: (include infonnation on renewal dates, term of license, number of licenses issued, etc.)

Fees may be charged for the use of reactor facilities. Fees are established by the Rhode Island Atomic Energy Commission (RIAEC). and are intended to offset the costs associated with the operation and maintenance of the Rhode Island Nuclear Science Center. The receipts are generated from usage of irradiation and laboratory services and the rental of laboratory spaces. These fees are deposited as general revenue.

Basis for FY 2009 Revised Estimate-(1) URI is paying an annual overhead charge on the Radiation safty budget starting in FY08 (2) HioPAL. Inc. Is projectlng a reduced irradiation schedule. BioPAL also rents laboratory space tor analysis equipment and storage of their Irradiated samples.

(3) Mlcrolnorganics, Inc. rents laboratory space for work relating to environmental analyses. Their space has been reduced due to development of a counting lab and they receive credit for maintaining the clean room Basis for FY 2010 Estimate:

(1) URIoverhead fees have increased due to the 40% rate.

(2) Reactor fee charges for support and laboratory space relating to the BIoPAL group should Increase.

Page 11 261

P26 o 2 262

F..m IIH.2. R-wd U200-1 263

r RIFANS NA RIFANS Natural Account Name B10CI 2006 Actual 1 2007 Actual I 2005 ACtI.l j 2009 re*i*eU 2010 reqtueted 022*.010000 !2910101 R I ATOMIC ENERGY 611000 R oulrwr nes 2 S44,510 145 47.226 $482,128 S510.473 529,139 611300 Classified Limited 0220 $11,147 1 $11,788 $11,788 S11.788 12000 611999 Conliact Reserve 02031 , , 0 613100 Uncla fijd e mdlPanent U2.7 -S39.961 $39,469 34,1066 $34.929 36.6" 614600 Overtime: Other fl.o. Sossona1( 0285 $ 82

$.000 $0 $0 _0 I - ý -1 - - -

0190000 Payroll and Emroree Benoets Accrual 0274 $0 82,320 $2.7t8 $2.830 _ 2. 82 612010 Cash Bonus Henilh Main Org (HMO* Part 0217 1 1 1 020100 IEr.lDos e Renvar -,Slat, Coniribaio" .0280 $71.843 $02,606 $107.234 $121,308 120.243 162,120 Jdic-dre (FICA) Hmpal In~surances Tax 0261 $34,917 536075 $35,890 1 36.317 I 37,416 12411D mplre " Cost of Empployeo Medical InsurwCO 0200 $%5,0(7 $50,785 S62,635 1 .79.49 1'. 58,636 824E20 mErpkf Dental Insu**e Cost of Employee 0297 $4,3286 $4.197 -4.691 8

$,5335 1 5.61%

I24130 .*moloy Cost of Emprse Vaslorr Insurarnce 0298 $003 8060 8 6023 1 8866 1.001 1626100 IAqesse .Pi Frine Benerlls Fund Assessrment )0263 $18.3181 $51874 8166aq19I 24 S38061 2.3021 620300 Reireel Hoahln.urr,,o 102941. $11.131 812,0? S20,1666 $22,669 $23.416 039600 Socufftyj Sarices 02651 8 Sl 2.000 $2,DO( $2,000

.O $2.000 1640100 IDUdis Mjaintenance & Re*airs 0363 $85.54 , S86752 36.0013 80.000 s6.000 640200 GroJnles MaInte nance 0301 $8551 $285,20 $2.000 82.060 2,000 541.500 JMoilntnance/Rovair5: AftL:rrcaEquipment 02620 $3.4?14 2,50

$__ ___ 2.wo0 1 2,5io 2,5000 04250 pavg Supplies/Expeosts , 10436 V0 $110 w $1,04 $1.0001 1,000 643120 ConcterSu.lisandSofwarnad EquipmonE (less,, 0442 So0 $1,0 __ $1,000 81.0 . 1.000 432130 Janiteal Services 0433 $8"5 $1,240 $1,000 51,000 1.000 643140 Kitlch-n/Household Suples and E utpm6t1 (less th.n 0433 $852 $1.000 $81.0a0 $1,000 1,000 00 643150 Prram Supl]*os and Equipment(l ess thanOl 0) 0323 $2.657 $1.8751 ,1,870 $1.657 1.,875 643160 SC0 i/f0011t831 Supplies . 0421 ' 4.441 $1.500 $1,W0 1, 500C4 ':' 1,501 643180 SirldinPlant/lachtr.rr SOuppliosand Equipment (1 0437 $7.431 S13.746 $13.741 S13.746 13,146 043202 M"0 end Fees 0324 54,776 $7,261 $19,001 , 19.00A 19.010 043300 SubS*o.ptons 0324 .. 4.9 $5,000 5,000 85.00 .. 5 .000 643401 Postal, Fmieht and Oeery Serices 0321 8301 $70! 8700 S700 700 643430 Fr 0325 S52.0 $5w6 $890-

• 5o00 S 64,3613 TIMsllo Adwarlisl 0332 sc so $0 6436Z0 P*riting - Outside Vendors 03031 S3.323!575 "I5 $703575 643710 . liatffTralning 04.41 , S50( 8300 3 8300 300 Page 14 264

RIFANS NA I RIFANS Natural Aceount NaCe 80; 2006 Actual 2007 Actual 2008 Aclual 2009 reted 20'10 ecuested1 643810 insurance: PropenytCasualty 032B S,,120 9.0771 $9,100 $,9100 12.000 143820 Insurance". Prot-slonal and Occupalional 0326 $8.120 $4,000j 14.000 64.000t 0 643920,, Medical Supplies (non-Rx) 0434 13.939 3.500l $3.5 $3,50(A "3.500 I43931 Plharmacelc*als 0439 $0 s100 $100 $100 10o 64DO4 Fuel: Geson n seilelFuel 0408 . 560 S300 $3 $3C0 4001 644520 Electricýiy- Cen-lre Ufiilres Fund (fo SF utm 0400 S44.0561 44.590 $46.8 $49.160 53,000 6A4700 W.Ier - Eupepmlures for Water Consuptonliu 0411 $1.029 $B282 $2.000 $2,.0 2.000

'4 200 Renml/Leae: Fvd1nn . . 0382 i 1 .g,50, $3.264 'S2.7601 62,460 2,760 640200 Maaf Allowance . PersonallyOwned Vol*elhCs 0341 so $1 1 St 0 64"G3120 )O-elStale Travel: Lodgng 0342 $1 ,691 S0 Sol S Sl 0 648110 Central Telephone Servces (CENTIREX 0327 $4.2301 $3.435 $3.1" $3.154 3,150 649.300 Wooer me0008 5r,teonic 03221 u- _____ $01 _____ so______ 0 S531130 Educatton Ser loos 0263 $7r 60 0$_ 0

$41211 l uldeid & Olner*S*e cte-es 058 1 $01 60 So, Sol$ 0 661605 FrUmiture and i pmante 0068 S ,$7.24 $7.247 S7.247 7.247 861671 Medxcal iu&Ll ul n*e, 0057 $I $0 so01 $0 $$

$01 ..

061701 C--,Lern F.Ie ment .SIiOto$4.809

• 0060 $59 10l so 5 0111 661821 Conmputer User l.core ($ 1.000.000 or noe " 0049 00 _ 8 . 0 TOT RI ATOMIC E1IF*d Y COMMISSION TOT $825.130 $171,67e $924.62 $1,0005,8471 1,026.07:

Page 15 265

RIFANS NA RIFANS Natnral Amount NaMe BOCI 2006 Adal 1 2007 Actual 2008 Actual 200) revised 12010 requestedI 0282050110 2915101 REACTOR SHAR,FUEL ASS AND _ ._*_

6317t0 LeciEdPOf/Art Seres... . 02633l $V $20.000 $2.0001 7.000 133310 Other . Irs,, 0363 $0 * . $( $2,00(1 S2,0006 2,000 634410 Offe xpee ,rs 0323 $ $( S2.000 $2.000 2.116 634940 ,Erclinal & Reerý.ona Spp & E , 0432 $0 S( $0 so ___ ,,

634980 Bilding & Machiner Supp & Exp 0437 $0 $21.000 ,, 0 so _ 0 639510 IF reicht Cartage and Eress 0325 S $1 S_] $0 a0 640200 IGrouorrs Maintommr-o 0361 $45,.38f $0. $01 $0 _1 "3150 . Prrogm Supplies and Equipment (less th* 5,00. 0432 $2.700 1 25.000 __..... $S "I O 043920 Madical Sulpies (non-Rx) 0434 $1,7331 $72.00 $72,0001 $3.0C( 2,000 164320 lout-of-Stat Travel: Lo00i., 0342 $5.936 $3.000 $3.00 $3.000 2.0 049120 Fees: SM'le Ardid 04591 $651 so $01 $0 _ 01 050990 Orther _0589 O$ , _ $S Sol m 0 1661421 QifoirýPriumur & EuIii 106581 $0 ____ ____ $01 W___Sol___ 01 661431 Other E0uip Additional 0650 $So 80 Sol set 0' 661501 Autonrotiv Equipment 0651 $ Sol _

$0 w 0 601*21 Ouilding & Ptrrt 0quivrerrlt 0372 $. $0 0 661621 OBudir & P1r1t0Equiprinei 02 $0 " i $$ 60 617701 ICompul Equipment 0660* $ $0 $0 IN " 0

=601405 Frumriuwe mi,'dEotripmernt 06537 00 L..iiA *j $5 _____ ____

REACTOR SHAREUE. ASS AND 1NSTRUMENTATN)N TOT $'.734. $100.000 $10001 s0.000 10.00C Pag.e 10 266

RIFANS NA RIFANS Natural Account N a, BOC1 2001 AcluoI 20O7 ACtual 2009 Actual 2009 revnoed 2010 'equesled 03282050200 2915102 REACTOR INSTRUMEN FA'rION 1_ __

63 1710 LecliE indWP fla rt Se rvice s 0263 $0 $0 .

so _ (:_

633310 Other Repairs 03lo3 I ...... Sol so 1 , 01 03 4 410 . 1 Ex, ,,e pe n s e f0 3 23 1 $0 _ _ _ $0 1 $01 0_

93 4 60 0 1 Or d i ,ll &r M ach ine r Su p & Ep xe 0 4 3 71 $ 01 $ 01 so[$ So 01 639510 Fro ht Cartage and Express 0325 $0 . $So $01 So 01 040200 Gruundo Maintun n.ce 03611 30 Sol $0 Sol 0 643150 Pr oram Supples and Equl ime nt (lese than $5000 10432 $4,7311 $67.00 $147.000 $01 _ 0 6i43920 Medical Su pmrOs(noa-Rx) 0434$ $0 $ Sol $

8646320 Out-ol-State Travel: l.odging 0342 $4 $3.00 _ $3,000 _ So _ 0 649120 Fees: Sn, li Aulit 10459 $0 $_ _ $S $0 0 661421 01010 Fwutrq & Equip, 0658 So $ r so 1 0 Sol 0 6614.31 "0 her Equip Addtornal . 065 $O $C $0 _ $_ 0 661601 A rkcro.. noaband Rel ated Equ upmont D051 $23W0 $_ $0 Sol 061605 Friture and ES fpmon..t 0651 $0 $0 ,_ so 0 o6 16 0 Furnituro a rd E q uipm ent 0 6 52 U S $ 0l so, $ _

'61621 Bulding & Plant Equlpmont 0372 $0 90 $0 0 06 17 0 1 Com pulerE q uip n.rd . . 0560 $0 50 s0 0 REACTOR INSTR *UME ,NTATION PROGRAM TOT 35.02& $70,0 0 $150 ,000S Page 17 267

RIFANS NA RIFANS4Nattal Account Name .BOCI 2006 Actual I 2007 Actual 206 Actual 2006 revised 2010 requsted 1 10282050300 12915103 RI GAOOLINIUM RESEARCH 640200 Grcndo MMaintenance 0361 0.4150 Malntoance.01 ;mlrs: Medle.l Equipment - 0363 $1.071 $3,000 $3,0001 3.00 3,000:

643150 Propram S ppO*s and Equipmmn (less than S5000) 0432 S0

$0 $. So*0 a_

043150. Pr qram Sup.,les and Equipment iless ihan S50000 0323 $45,354 $100.000 $100.000 $23,0001 3.001 643180 BialdinglPlanrlrM;admn Supplies and Equl pr.t (I 04371 $3.676 $514.500 514.50 14.5000 3.5001 643430 Freirhl 0325 S364 55I0 5l50I $5000 500 643920 Medical Suppiles (nr'o-nR) 0 0434 $ $ So S

$,_o , i 045.120 Out.oI-State Trael: Ldr, 03421 $2.4761. 37.000 S7.0 7.0001 5,00 640120 Fees: Single Audit 0459 71 -0 50 $01 01 653130 . Educaion Services 0265 $14.235 i 0.0 62000 $20.0 __] 0 009950 Ott1er 0509 Sol. $0 S $0 0 65.1231 HistMiC8l1ildilgs . "" 06S1 $0.45 .5 .$0 .0 0 661421 Ottlce Furnlture & Equ 066 SO * $ _$0_ 50 $1 _ 0 061601 Aircr, oals and Related Equipment 0651 $0 ., ,,,,O)$0 _ _ $5 0 661605 Fonniture and Equipmont 0652 , s0 ,$m "01 0 661605 Fumilure nrd Equlpmrmt . 0657 so So $0_ " 0 681W05 Fumarnro and qu)pent . 0659 * $

S - so . $01 0 W01621 E34udic2n &ptanE ni ~enf 10372 so so O 0 651701 Corn EtorE r emnt,5,00 to $4,909) 0660 53,532 5,00 55,000 .o Rl GADOLINIUM RESEARCH TOT[ $76,433 . 5150.000 $150000 $73,000 20.

Pa0 e 18 268

R3FANS:NA: ' RIFANS NaturAccount Name BOC 2,, 6 Actual 1 2007 ActualI 2008Actus 20091Mied 2010mrnuestud 0282000400 2g05101 ,RCP - RINSC REACTOR

• 60123; H*Lodc Ouklirs 0661 [ $0 " 90,000 -90.00i $50,00o 50.000 0Ž82095511 i2805101 , URI SPONSORED RESEARCH 611000 Reqular Walas 0230 8100.491 $111.7671 $116.141 $116.620 119.717 1e9 iPu r duE aoeeeneflts Accrual oren 0274 .. $ $491 $614 S547 571 620100 Empron1esý Reliremen;l- Slate Contstnlruln 0280 $1,.301 $20.573 $24,123 $11,12 13.799 621120 M adro IFICA) Hompita Insuranc Tax 0261 $8 188 88,548E $8.8858 S4.4321 4908 824110 Emrpoyer Cost of Eml.lcvce Medical insurance 0295 $17,687 _$21.42 24.294 $24,317 26324 624120 Erap. er Cost of Emp~re Detlnsutarai 0297 S 1. $$1.11 $1.36 $1.684 1786 824120 E-.pIo.,or C~nI ot Err. 8rly-n V~uiCn Irurance 0208 $221 $268 $277 $40E 329 626100g Asssed FriNgr Benefits Fund Assess.mrent 0293 $4.155 $4.247 $4.414 $4.43 4508 626300 Retinae Heozth Insuronce 0294 S2,527 $2.69. 81,518 *1592

-4540 URI SPONSORED RESEARCH TOT , $150.8 86 171, SB4, 65 $165,152 173933 ITOTAL - ALL ACOUNTS ITOT 1 $1.074*002 1 $1.362*884 $1.50D.318 S$.230.006 1.280.008 Poag9 269

Form BR-6b Revised 9.2000 Page 20 270

BIR-7 Revisdl"/t99 Federal/Restricted Account Estimates Account Nmber 2i20-50100 CFDA #* 81.114 Account Name: Reactor Sharing, Fuel, & histrumentatiorn Statutory Refelrence: DE-F007-0t1D04154 1DE-FG07-02IDt4389 Explanation of Federal Grinutor Rcstricled Receipt Account Granting Agency: DOE Reactor Sharing funds arc dcrivcd from a Department of Energy Grant to offset the cost of tmaking the RINSC reactor available for use by other college and univcrsitics that do not have a reactor.

Grant Period: FY2005 FY2006 lrY2G07 FY2008 VYZ009 FYMoto Actual actual AAttaa actual revived reouest Balance from Prior Year 92,163 78,419 106,039 137,653 87,53 11,528 Plus: New Receipts/Grant Award 19.666 36.700 35.500 50,00M 0 0 MItius: ludirect Cost Recovery Equal, 'o'otat Available 111,829 115,119 141,539 i17,653 87,653 1t.528 Minus: Expenditures 33,410 9,080 3,886 100.'W0 76,125M 11,528:

Equal:lalanre Forward (to new year) 78.419 106.039 137,653 87653 11528 0 Explanatin oflMethodolugy page 21 271

DR-7 Revised 7V92 Federal/Restricted Account Estimates Account Number 2820-50200 CFDA Ot 91.114 Account Name: Reactor Instrucmntation Statutory

Reference:

DE-PSO7-90F.R12396 DE-FG07-a21D142U2 Explanation of Federal Grant or Restricled Receipt Account Granting Agency: DOIu Fands are derived from Dcpartment of F.ncry Grants to provide for equipment far umiverity research reactors The staffcomplcted installation and testingof the ncw digital rod control systcni in June 2008 and all. funds havew been expetided. rhe cost overrun was covered by transfering fanhds frometie Gadolenium fund Grant Peri'od: FY-2005 FY2006 VY2001 I'Y2908 FY,2009 .Y2010 Actual Actual Actual Actual .Revised request Balance from Prior Year 37,190 45,246 40,216 40,216 40,216 0 Plus: New Receipts/Grant Award 0 43,529 150,000 0 0 Minus: Indirect Cost Recovery Equal: Total Available 37,i0 45.246 83,745 190,216 40,216. 0 Minus: Expenditures (.,055) 5.030 43,529 150.000. 58,707* .

Equal: Balance Forward (to ncev year) 4ý24 6 n. 40.21k 40.216 118.491) balance trasfc'rrld*.1TR

.2915103 to cover costs Explanatlion of Methodology page 22 272

DR-7 Reyised 7/98 Federal/Restricted Account Estimates Account Numbcr 2820-500300 Cl'DA #: 81.114 Account Name: GadoliniumResearch Statutory

Reference:

DE.FG07-02ID14420 Explanation of Federal Grant or Restricted Receipt Account Granting Agency: DOE RINSC is supporting cancer research ofdeep esatedtumors using cells and rats by exposing them to a neutron beam that is generated in a research reactor GruaI Period: PYZOO5 iY2006 " O'2007 FY2008 FYZ009 PY2010 913012002 to 9/29/2007 Actual Actual actual actual Revised request Balance from Mrior Year 114,206 125,624 142,740 194,274 119,274 .71,230 Plus: New ReceiptrsCrant Award 50,000 92,500 85,000 85,000 0 0 Milnus: Indirect Cost Recovery Equah Total Available 164,206 218,124 227.740 269,274 119,274 71,230 Minus: Expenditures 38,582 75,384 43,466 150,000 48,044 71.230 Vqual: Balance Forward (io new year) 125,624 14.ZOA2 !A& JLt,2L

.7_4 71230 0 Explauatlon orNMethodology pg 23 273

B -,~ cRviscd Wi98 Federal/Restricted Account Estimates Accouut Number 2820-15104 CFDA 0: 81,114 Aecount Name. Nuclear Eaoergv Research Statutory Rcereroce; DE-F(T07-021D14420 Explanation of Federal Grant or Restricted Receipt Account Granting Agency:. DOE

• This ant suppored the dveClupmoleL Of a new neutroni scattCeng instrunen.t at RJNSC in conjunction with the Physics department at URI Grant Period: FY2009 Balance fromt Prior Year 0 Plus; New Rece9i2lGrant7 Award 100.000 Minus: Indirect Cost Recovery Equal: Total Available 100.0("

Minus: Expendlture . 100000 Equal: BulauceFvrward *to hew year) .

Explanation of Methodology 24 pg 274

Persons with Disabilities as a Percentage of the Worl FY2007 FY2008 FY2009 FY2010 General Government Administratfon 2.3% 2.3% 3.0%

Revenue 2.4% 2.4% 1.0%

BLsiness Regufation Labor &Training 2-5% 2.6% 2.7%

Legislature 1.$. n s., it.,

Office of t'he Lieutenarnt Governor Secretary of State 1.% -

General Treasurer 1.2% 12% 1.2%

Eoard of E[ections Rdhode Island Etics Commission 8,3% 8.3% 8.3%

Govemcrs Office Commission for Human Rights 33.3% 33.3% 33.3%

Public UWilities Commission 2.2% 2.2% 2.2%

RP<de Island Commission on Women Human Seavices Office of Health and Human Services Children. Youth, and Families 5.6% 5.0% 5.0%

Elderly Affairs 14,5% 1.5% 10.5%

Health. 1.2% 1.3% 1,5%

Human Services 3.0% 3.0% 3.0%.

1.0% 1.0% .1.0%

Mental Health. Retardation, &Hospitals Office of the Child Advocale Cormmission on Me Deaf &Hard of Hearing 67,0%, 67.0% 67.0%

RI Deveaopmental Disabilities Council Goverr4os Commission on Disabilities 100.0% 1).00% 100.0%

Commission for Human Rights .33.3% 33.3% 33.3%

Office of the Mental Health Advocate Education Elemerntry ani Secnda*y 6.5% 4.0% 6.5%

Higher Education - Board of Govemn-rs RI State Council on the Arts I1.% t.,0% 1.0% 11.0%-

RJ Atemic Energy Commission Higher Educatlon Assistance ALJrm=iy 7,2% 7.2% 7.7%

Kisborical Preservatlon and Heritage Commission I.

240 275

Females as a Percentage of the Workforce FY2007 FY2008 -FY2009 FY20io General Government Administration 37-9% 37.9% 41.5%

Revenre 59.1%. 59.1% 49.0%

Business Regulaton 54.0% 54,6%

Labor & Training 67.0% 65,2% 58.0%

Legislature n.s$ n.s.

Office ofthe Lieutenant Govemof Q0.C% 44.0% 55.0%

Secretary of State 57.1% 59.3% 59.3%

General Treasurer 61.5% 61.5%

Board of Elections 42.9% 37,5% 37.5%

Rhode Island Ethics Commission 58.3% 58.9% 58.3%

Governors Office 49.4% 54,3% 511%

Commission for Human Rights 56.7% 68.7% 66.7%,

Public Utilities Commission 35.5% 35.5% 35.5%

Rhode Istand Commission on Women 100.0% t-O.C*. 100.0%

Human Services Office of Healm and Human Seivicos 80.0% .80.0% 100.0%

Crhldren, Youth, and Families 64.2% 65.1% *5.1%

Efderty Affairs 89.0% 89.0% a9.0%

Health 66.7% 87.3% 88.0%

HLrnan Services 76.0% 73.0% 78.0%

Mental Health, Retardation, & Hospitals 65.5% 66.1% 66.0%

Office of the Child Advocate 100.0% 100.0%. 100.0%

Commlssýon on.the Deaf & Hard of Hearng 33.0% 33.0% 33.0%

RI Developmental Disabilities Council Towo0% I Do.0% 100.0%

Gvvernc',s Comnmi-ssion on Disabilities 28.8%

28.8% 28.6%

C,ommission for Human Rights .65.7% 68.7% 69.7%

Of*te of the Mental Healt Advocate 75.0% 7M.0% 50.0%

Education Elementary ait, Secondary 74.6% 71.0% 74.6%

Higher Education - Board of Governors RI State Council on the Arts 57.1%

69.6%

70.6%

6g.8%

70.e%

69.8% I.

RI Atomic Energy Commission Higher Education Assistance Authority Historical Preservation and Heritage Commission 33.3%

73.6%

66.6%

33.3%

73.8%

66.6%

33.3a%

.79.5%

66.7%.

33.3%

'1 237

• 1~

276

Minorities as a Percentage of the Workforce Females as a Percentage of the I FYZCM FYnM FYMII FyiDIO - -

GOWN1a1wW,4,o5 A~M-ww"lc q-4..

'9.3 q435 14A%

411 11.1%

T,4%

T--

C411%

C4.41I-.

54.443.Rol11

&S.d Sk-1n1 11.7 U4.0% M1-111S&M54.S 7.1%

14 11 "7163

% RR'W3+/-W & I-W11 ,CC%

151111 443115. m4~.the4 545L.w5a.w v42%

1014.7 tic%

IM'

• 1111%

3311%

MCI$% IMPS1

.41~ f Y;tiC F 4 171C% DIS115 do -IN ELI4.*IJ 4.851 .14 C144 711.3

ýk th.s T a]

4I- Cwa11.WA-155 911%  ; Emw c

Ell %

11.4% = A1 2".44 Me1% F,.04 Tdi14 .,4..ltm Alale" 524 277

RHODE ISLAND ATOMIC ENERGY COMMISSION CAPITAL BUDGET

SUMMARY

FOR FISCAL YEAR 2011-2015

"

• TerryTJPhND., Drector " E JUNE 19.2009 278

FY2011 - FY 24015 S, Budgot CapItal lniprovematPlani Rrquest Syp:er 9fALC*M f"~d Project Summary Report Agency 052 RHODE ISL*D ATOMIC ENERGY CO HMISSION Priority Ranldng Project ID Profect Name Total Funding I AECAP Atomlic EaerU Aswftm.ection $330,000,00 Prtntd Ay, Tery Tehan qtIall

June~It, 2009 1~.

279

FY 2011 - FY 2015

• SaOffice fi~ CapitalhItprve~mentPlan Request System

~

I I~

• Budget Vm Y~w~

P'ojcl!IDvs.rder Agen.Y 052 JWODE ISL,4ND ATOM*IC ENERGY COMMJWION Ordsr Project ID Project Name 1 AECAP Atomic Energy Asset Prot4ectcfl 2 CTC Cancer Truatmemt Center 3 LPL Landscape and Parking Loot 4 RINSC7 Front Paring Lot Repazr, Resurtace & Landscape 5 RINSC5 Decommissioni ng RINSC and Reactor ty*y, AMC 1; "09 PHhw y: Tony Tchr w Pg I of)l 280

FY22O11-FY2015 Budge CapitalImprovement Plan Request System

~~ P~rojectRe-quest -Nzarrtive wd Jusuficution(Fo 2)

Agency 052 RHODEISLAM!, A TOMIC'ENERGY COMAMISSION Project ID/Name: AECAP Atomic EAergy Aseet Protection Project

Description:

Complete scecrity lighting and landscaping at rcar of building. Insatal new ceiling and floor in tbc former clean room which is being converted into a counting laboratory for the students who utilize the new classroom that was recently completed, Project Justifiration:

The lighting is a security issue and landscaping around the new parking lot is a safety issue. A new ceiling and floor are needed becaue the staff has had to remove old air handling equipment from 6c ceiling and benches from the floor which has aeft asbestos tile exposed.

Project Status, awaiting funding StratrtkCriteria S art y Coa*ces ................................. ... ... ... ...... ..................... ........................... M inor Safety Narrative Lijghng isneeded for secury of tbe fariliry. Landscaping is needed to live! off aromd the pavtd ae. Asbevos floor files need to be replaced Code Violtio s ..............

. . ............ .............. ..... .... Mince Code Narrative asbertos tiles in cwiting lab Enaergy IM Pa ............ ............... ........ _......... .. ........................ None EnergD Narrative Cu'tamr S" ...-................................................ Major Customer Narrative Need sauf =en to new classroom 3 ,00v aphie Slpfkicm ...... . .......... .. .... .. ................................... . Local

etgrapbk Narrative ge..... ............................. ...... ...... . . . . ..... M edium gene) Narrative ced safe access to classroom and use of *onting lab er Iim saclig ... ........... . . . .. . . . . . ....... ........ .... . ............................ .... 0 er Narrattive t availablestff is doing much of lab %mk dey, Padr J,%20" Pitted By

Terry Teii n;" I Wf2 281

'f4Sune fth"TeM*AE FY2011 -FY2015 Budget CapitalImprovement Plan Request System

~~ Project Requesi -Narratjiva z Justifiafion (Form 2)

Agesq 052 RHODEISL.AVD A TOMIC EMRGY COMMISSION Project [D/Name: AECAP Atomic Energy Asset Prolectiom Asset Mans t ... . . .... .. . . ........ Rehabilitation Asset Narradve The rear area will be safe and secure and ik* new Labwill repLace a 30 year old coming room Opcrtong ....... ... .... ...... None Operatling Narrative

?,,or)*

Histork Pre rvati ..... . ........... .................................. NO Ifisloric Narrative Budget Narrative Critical Lif Sfety ............................................................. No Critical Life Safety Narrati the paring lot hw trembet arouid tbe Pahemcot aid Do lghting Critical L&pJ Libility ......... . ............. yes Critical Legal Liabitlity Narrative OSHA- tri and fill Prior Binding Cunemhtb t ...... . .... .... ............... NO Prior ektdiogCormmit .t'rrctiv, 282

3- of,,* s4 FY;OlI - FY2015 Budget

' CapitalImprovement Plan Request S*Wem Sotffice ProjectRequest - NarrativeandJustification(Form 2)

• WPW , O 40ki Agency 052 RHODE ISLAND A TOMC ENERGY COA(0,LSSION Project [D'MName: CTC Cancer Triatment Center Project

Description:

Dr. Karl ott of Purdue University has developed a neutron beam for tmating brain cancers. He has completed all engineering design wrok for installing the filter in the centers nuclear reactor thermal column.

Project Justification:

Dr. Ott has a fiftern year lease to complete this filter which has dhe potential to cure very difficult brain tumors.

Project Status:

Dr. Ott has expended considerable funds and effor on this project. He is lookLig for additional outside funding and a treatmnent protocol to finsih the project.

Stratelie riteria S f...... ....... ..... ........ N one Safety Narrative Codeolv latino --- ................... . .... ... ......... ............. ....... .............. . N one Code Narrative

[is rgy Impact .. . . . .N...... ..... . . . .......... .... .... ..... ..

... .............. ole Energy Narrative c mstW- * ,,ce - ....... . ........ ..... . ......... .. Ma..........

a...............................................

jor Customer Narrative RINSC will do important medical research in reatment of rancerous atuors CWaphic Ssipiticac . ... . ...... Regional Geographic Narrativr RINSC wiD serve all remrchers i he area of bra*n and deep stated tumors

,eacy Mition ...... . .................... High kgeney Narrative UNSC will be at the ciring edge oftcsner rescamh Iser Filsni..g 0 4erNarrstfive he pmrject is funded throvh outside financing by NTC Company

WMant e ..................

t ............................................ New Construction me Narrative year lese r t ... ............ ....... e.....................

crating Narrative i d.If- A.-. IN !10140 MIA-d A.- T- TA- PWaeof 283

4Budget #00"FY201J-Aw oft"* FY 2015 CaptllImprovenmentPlani Request System

• *Project Request - Nawih~e and Justiflcadon (Farm 2) geq 05.2 MJODE ISL4ND ATOMIC EIVERGY COWMMISSO Project ID/Name: CTC Cancer Treatm cut Center Hutirie ruevatt. .No Historic Narrative 8I~dp w _e

......................... N Budget Narrative Cruitcal ULe Safety Narrative Crilical Lega Liabiiity N Critical Legal Unbililty Narrativt Prior Olvding CommirtmentS Prior Bindiag C&ui~tment Narrative 284

FY2011 - FF2015 a'S~rudget CapitalImpro ement Pln Request System ProjectRequest - Narrazfieand J'ustif ation (Form 2)

Agency 0M2 RHODE S!LA4V A TO.MC ENERGY COMMISSION Project ID/Name: LPL Landscape and Parking Lot Project

Description:

Paving of therar parking lot has been completed with asset protection fxnds and a new security gate has been installed with federal funds. Landscaping and lighting are stilt required to complete the work Project Jutification:

This project resolves parking and security issues.

Project Status:

The remaining work will be completed this year Strafe&l Criteria Safety Concerns None Sjarey NarratIve Coade Vlolatl** ..

Code Narrativie LEaersylvc -- l Everty Narrative Ciatnarsef .rvite ,ý. Minor Customeri Narrtive Provide addluonal parkinig Gbolraphic Sipukaitt Local Geographic Narrating More parking for (TJIGraduate Schtool of oceanogrphy.

Agency Mlsslwi Low Agency Narrative Nuetr parking services.

User Financin 0 User Narratl~ve

'lone New Construction jiset Narrativ~e rovide additional pariing None paitialg Narrative

,sdav. Jane IL 2009 Prinmedlv' Tern Tekan Pat I 4fz 285

Z Sam4 SAOIsla" FY201) - FYZOIS I&Budget Capita IntprovecmentPlan Request System

,w4Officti 4: **'," DV-sfAI,-- ProjectRequest -Narrative and Justification(Form 2)

Ageny 052 RHODE ISLAIND ATOMIC ENERGY COMMI0SION Project IDIName: LL Landscape and Parklbg Lot Husterir preaw Otion..... --- ...... NO Historic Narrative No Budget Narrative CrkicakllAaSakty No Critical Life Way~tNarrative Critical Legal Lla~.ity .......... .....- No Critical Lergal Liability Narrative Prior Bindling C4,mioret _ ......... No Prior Bindndvg Conwitameut Narrative 286

Ss ek#1xA**Jx1Wd FY2011 - FY 2015

, Budget Capilallmprovement MAuRtquest Syre ProjectRequest - Narratveand Jusrtification(Form 2) 4gency 052 RHODE ISL"ND ATOMIC ENERGY C0M*ISSION Project ED/Nam¢: RINSC. DtcommslilonIng RUI{SC and Reactor Project

Description:

The Nuclear Regulator Commission requires an approved Decommissioning Plan prior to shutting down the nuclear reactor. This plan would cost $500,000-800,000 to develop and approvaJ would take around two years, Until the nuclear fuel is removed, the facility must comply with a licese rqiirements.

Project Justillcation:

The state is responsible for safely dcoommissioning the reactor and all costs associated with decommissioning. Since the state has no disposal site for class B & C radioactive waste. The ability to complete decommissioning is questionable at this time.

Project Statms:

The RIAEC does not plan to dcoommission tie reactor as long as it has adequate finds to operate it safely.

Strategic Criteria Safe.t C vers ............... M............Major Safety Narrative The resoor must be deconmnssoed safely Code Vilatio ... I.............................. . ........ . ... N one Cade Narrative Erwrty bapact .............. .................... .. ... .. .. .N one Energy Narrativa Cust e r serv sl ..- . . . . . . . . ... ..' ....... . . Major Cust*nuer Narrativ No More we of thwe reacwo Geto rapWc Signlfi ne .e........... .. .......... . . . . . . . . . . ......... Regional Geonr2pak Narrative All uses would lose strvice geeMi ......... .. . ............. -igh

.......................

%genay Narrative

-be agercy would Do longer perfoim its educational and rsearcth missions.

F s g. ............. ......... ......

strNa rratillve one, Suta respcasib lity by law me ltaaaget .. .... .. Infrastructure

set Narrative will saily dscomituaion the nuclear reactor 287

- - .4, wfd4*. FY20!) -FY2015 SBodger CapitalJmprovemtent Plan Requetst Syxtem 444 Dfrimt.l fdtrvE ProjectRequest - Narrativ'e and Jw~ifieution (Form 2)

Agency 052 RH!ODE ISLANZ) ATIOMC EN'ERGYCOMMISSION Project Warnie: RJ.NSCS Decommissioning RLNSC and Reactor Opeapl - .r _..... ....--------...... Major Operating Narrative This project will be very expensive and has pofetiaW for excess cost and ovcm'lmas.

Hioad Frae rvefis ..-----....... ---I..... .......... .. NO Do*dge Narrative C ritical~ifeSate ..... _.... ----- ..... ..... --- - -------- e Critical Life Safet~yNarrat~ve The nuelear reactwr Must be decommeissionied iafely CrkkclaLegalUmbility ................ ..... ..... ,

Critical Legal Liability Narrative Tbe Nuclear mg~uatoty Comm~ission can, imposec aumwtal sanedonos aed fnoes if he reacntr is not decommisasioned satfly, rrwn Binding Caui itmmct ..... ......................... Yes prior Bidinag Cominkmiaa Narwaivec By W. tKime tsatm pavcuhk f&tdemLs~iainsg.

288

%~Budget CapitalImprovement Plun Request Systev

~ * ~ , ProkcrReipues - Nrraive andlbuification (Form 2)

Agency 0D52 *AHO ISLAND ATOMIC ENERGY COMMISSION Project MINmtme: RE OSC7 Front Parking Lot Repair, Resurfaee & Landscape Project Destriptioin The tuont parking lot has broken pavement which is a safety hazard. It noeds tD be repaved Project Justification:

This is a safety item Project Status:

Awaiting funds Strateg~de Criteria sr.y cnc ..... . . Minor Salfty Narrative Parking lot is i a dilapidated condition.

Cad* V1014000 .................... .............

............. .. . . .. . . ..... None Code Narrative tripa.d all on btokett pavement Eacre Impact ........ None gnery Narrative C ustl n er s rv ................ . .......... ... ......... ... ................................ .. M inor Customer Narrative Bete parking and safet pwking lot G *oaphi Slgmn~ * ................................ ... . . . . ... ...... ...... Local Geographc Narrative fletter parking fbr users A genry M asion ... ...... ........ .... . .. ..... .... L-.....................

L..........

Low Agency Narrative Bettar parkng L~ser Fieecif .................................................................................................

Jser N.arratlive

'One mat Nt4a gement ............ ...... Rehabilitation

.ssft Narrative astare ft parkig lct to good ccndition pa titng ...... ........................... ............... N e peratleg NarrAtive 289

-r - . .5o*/Aso~h FY20II-FY2#US

~Budget CapitalImprovement Plan ReqiuestSysrtem

~~ ~ Project Request - Narrativeand Justificailon (Form 2)

ES2 RHODE ISLJDATOMW ENZRGY COMMISSION 405 Project EIDName: MISC7 Front Parking Lot Repair, Resurface & Landsca pe

.....

..... ------------ No Higoric Narraive, Budget KoTfaIlvc Critial Life S~mety . . ... -----.. No Critical Life Safety Narrative critital Legal Liability ...

Critical Legal Liability Narrativ Priar hidingctoln ....... No Prior binding Comwitiaent Narrative Ri Bddy: Terry Tehan Page 2of2 290

FY 20I1 - FY 2015 ome.f CapitalImproveement Plan Request System Agency Request vs. Plan F~ernS~ns I FY 2010 FY1l0 F2011 172012 FY 2013 FF2014 PY W00 rong FY M.11 Tov4 ALM-ey 552RHIODE ISL4ANDA rMIc ENER 6Y CVNI Pnroet: AECAP Alornic Energy Asset Pr.20ectlon Stoiw Plan F5av 130, Sod 30 so 34 V£~

o,00 ml cap"~ PIM.Vnwd two'sm0 500..sSO Me so,SO 30,. mom so.01 so 1"o30,00 1130E 55.0m0 55005 M5.00 "Am50 $0 000o05 50 $0 533500 SMJaF,d. so sqsoosMo re cap5a PweP100 floko00 Mom00 Mow05 P0.000 Volo w0.5 0S 300 Prumcd By: Trry~ Teh fs.af 291

FY 2011 - FY 2015

• Office

• 0 pve'4.w~

. - A -- *- CapitalImprovement Plan Request System Plan - ProjectFinancing r"*FY 7010 F 260W FY2:0!! FY 2012 FY20!) FY20 14 FF2015 pastFY2015 TOM0 P.'.&ý M-.,: AYCAP Ab.kE-W ~A-4 Pm.~Om "SO k.nmw AECAP status: Ram"4 MI CptaI Pt.. Fmn OM0,01 SWIM0 $0100w $S,000 SD $50,100 00 so 1 so SO 530,010O 1130.000 SWA.00 $'t0000 10 000,010

-AV, J.- Is, M" PFrbd r: Tre", T ,

~*'3 ~c ~ ' - -~

292

• , *udget FY 2011 - FY 2015

• Office CapitalImprovement PlanRequest System

• • DqI,..n i..

Project Costs Fi tr.g P? "T 2010 FY 2010 F 2011 FY2012 Fr291 FY 2914 FY2015 Po, F Y 20 1 T.IAE 0752 WHOVE I.VL4NVDArOMIC FNU~GY COMMISSION

?rolet tNamc-. AECAP Atomic En ergy Asset Pfectm~on Pm-jcc( ldmeeir AECAP o". S100.904 1MON~ 350.m 150. -DD SSI.0 $,400 so 144 *M3000

$4440w 350.00 50w 1. w5 110.00 530.DO w0 SIM15.01 1.nd. J.- IS,2M9 Fyburd By. 2myi lhan rq~ 04 1 7-T I-293

%-0.f41kk W-4..

• Budge O4Mau@e FY 2011 - FY 2015 DP fOf wiw.* CapitalImprovement Plan Request System ProjectCosts FMo.edq Pt'rl201# F Zola #5Y2011 FY2012 FY 2011 FT 2014 Fly2015l POWf r2013 Tww*

Agree; 952 RHODE ISL4NDA TOMJC ENERGY COMMISY1iON ProjectNamrw AECAP AtomicEnoly AssetPrf&0cbmo Project Itkatithe AECAP CTU500.0 MU 150 105* M5.M0 MAN00 $50000 V0.000 so 1400.00)

SUM"o 1100.00 1500(0 MAO 150M00 DOM00 M0000 3550 0 01.0 Prqojeqi Name CTC Cancor Troolnsenf Center ak2&350.0*0 50 so so 50 so so0 I.0.

515000 50 so00. 00 so so 50it 20 515.00 Projetd Nwme: LPL Lansolcape and Par~ffg Lot Proeimt Idafifler LPL Priority R*!hitu: 3 ________ ______________ _____

01.* s1'.MW0 so $0 so so to so vi 150.05 t.AMotI S56000 30 so so 30 to so so 3500.50 Project Name: RINSC5 Decwmnwissin'og RINSC and Reador' Proeto ldeosriter RINSCS 50 5:0 00 - - .10 1 - - .50 0

.0....... 00.

J-.dsIs.0

. New0 r~mwor.~ Terny ToA" No.1.12 294

Buda0Jgsr FY 2011 - FY 2015 CapitalImprovement Plan Request System Project Costs

'12 FT 2011) 7", FY 201 21 $ P",FY 2015 2?oo ParkmnLo: Repar, PRsudaco & adsp Project Name, RINSC7 Fmnrd Project Ideulifir RINSC7 Priority Rkhqz 236,OM 10 so 20 so vxw00

...

... ...

15..... ....

...

... 0 03 0 Swwwot to $a $a Th..,,&y. i'm )J~ 209 N*.~odBy.~ To.,y r0900. Pqo 1 .f 2 295