ML13080A362

Response to Nrc'S Request for Additional Information Regarding Rhode Island Nuclear Science Center Reactor License Renewal

ML13080A362
Person / Time
Site:	Rhode Island Atomic Energy Commission
Issue date:	03/15/2013
From:	State of RI, Atomic Energy Comm
To:	Office of Nuclear Reactor Regulation
References
Download: ML13080A362 (42)

Text

OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL FOR THE RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR LICENSE NO. R-95 DOCKET NO. 50-193 The following requests for additional information (RAIs) are related to the Safety Analysis Report (SAR) submitted as part of the application for license renewal for the Rhode Island Nuclear Science Center (RINSC) reactor dated May 3, 2004, and the responses to RAIs submitted by letters dated December 15, 2009; January 4, January 19, December 7, and December 14, 2010; and January 24, February 24, July 15, September 29, and October 6, 2011.

Unless otherwise noted, RAIs that refer to the proposed technical specifications (TS) refer to the version of the TS submitted by letter dated September 29, 2011.

Responses to the RAIs should be in the form of discussion or analysis or both. The responses must provide sufficient information for the NRC staff to independently verify all safety-related conclusions. Responses to the RAIs may be in the form of replacement pages for the SAR.

NUREG-1537, Part 1, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content," dated February, 1996, contains guidance for providing sufficient information to satisfy the related regulatory requirements in Title 10 of the Code of FederalRegulations (10 CFR).

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.

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:

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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-1537, 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.

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 2

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 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-1313). 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.

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.

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,S 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.

Specifications:

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

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.

2.2.1.3 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 4

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 performed for the most conservative case 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 oC 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:

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

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.

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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 Settings for Forced Convection Mode of Operation 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.

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d A 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 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 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 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 9.1 in,

- 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:

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

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 error, 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 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.

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I 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 low as 1740 gpm, which still leaves a safety margin of 160 gpm.

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.

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 1/2 inches apart, so movement of less than /4inches 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.

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 submi ted in 2004, ncc, t0 bCupld*tcd.

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.

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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 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.

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a 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.

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 11

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 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.

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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.

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 andp limits are sent back'o 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 13

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.

10.5 Based on discussions and firsthand observations during NRC staff visits to the RINSC, 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-1537 and not restricted to Chapter 10 of the SAR if the modification affects descriptions or analyses presented elsewhere in the SAR.

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 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 14

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 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 -,*dth 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.

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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.

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.

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 contamination and radiation levels. We survey accessible areas with gamma or neutron fields monthly and verify/update posting. It is inappropriate to specify routine 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 w.. ...contamninatio.n as..,idntri.*" Action leves L-,.- -external radlation 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.

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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.

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, we assign extremity monitors to adults who could receive an annual dose equivalent to the hand in excess of 5 rems 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).

13.22 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.).

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 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 17

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.)

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 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. Q 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.

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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 in2 (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.

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.

I ,,luhcal Sp.cification, 3.91-3 will bc rc-;;rittcn to say:

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

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

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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.

3.93 1.4, When the through port is not being monitored for a leak condition, the ends of the port shall be closed.

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 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, 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 20

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 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 4.52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> after reactor shutdown, then the 1.25 in 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:

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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 in 2. Consequently, maintaining the limit on the area open between confinement and the beam ports to 1.25 in2 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.

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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.

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

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 L,... , L, hav, an impact on reactorria.,  :, orLoL

-Adioa" x 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.

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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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The following RAIs relate to the responses to RAIs provided by letters dated December 15, 2009; January 4, January 19, December 7, and December 14, 2010; and January 24, February 24, July 15, September 29, and October 6, 2011; and the revised proposed TS submitted by letter dated September 29, 2011. Unless otherwise noted, RAIs that refer to the proposed TS refer to the version of the TS submitted by letter dated September 29, 2011. In responding to the following RAIs, provide a response to each individual RAI, and provide a complete set of revised proposed technical specifications that incorporate any changes made as a result of the responses to the RAIs. NUREG-1537 and American National Standards Institute/American Nuclear Society (ANSI/ANS) standard titled, "The Development of Technical Specifications for Research Reactors," (ANS-15.1) provide guidance for developing TS that meet the requirements of 10 CFR 50.36. Please note that 10 CFR 50.36 requires that "technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34."

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.

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 determine 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 changes due to flooding into consideration. The definition will continue to be:

1.21 Reactivity Worth of an Experiment The reactiity worth of an experiment is the maximum absolute value of the reactivity 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.

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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 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 v so tne cclity data c. b. 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 26

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.

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, uncontrolled or unanti.-aeA ,h.,,n -,n . 1 A VA.7 In exce.s 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 implementation 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.

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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.

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.

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.

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% dK/K is in progress, or 28

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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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 a-nd 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.

The proposed TS states:

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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.

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.

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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.

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.

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.

This will be incorporated into Proposed Technical Specification 6.2.2:

6.2.1 Nuclear and Radiation Safety Committee (NRSC) Composition and Qualifications 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 32

6.2.1.1.4 Four members that are not RIAEC commissioners 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.

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.

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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 determined that 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.

14.187 Proposed 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.

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.

Explain why no such provision exists in the proposed TS, or revise the proposed TS to include such a provision.

Proposed Technical 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, 34

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.

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 "review" 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.

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.

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 w.il ensue 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.

An additional Technical 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:

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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.

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.

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).

This section of the Technical 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 reactoi, but should not precludc thc usC 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, unloading, and movement within the reactor, 6.4.2.3 Maintenance of major components of systems that could have an effect on reactor safety, 36

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.

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.)

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 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 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.

Technical Specification 6.7.2 will be re-written to say:

6.7.2 Special Reports 37

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 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.

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. Section 6.1.1 of ANS-15.1 states that the idealized levels of the organization are:

Level 1 - 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 38

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.

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:

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40 M

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 by 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.

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 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 41 ENCLOSURE

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.

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, 6.8.3.2 Off-site environmental monitoring surveys, 6.8.3.3 Personnel radiation exposures, 6.8.3.4 Drawings of the reactor facility, and 6.8.3.5 Reportable occurrences.

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