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Rhode Island Nuclear Science Center Response to NRC Request for Additional Information Regarding the Renewal
	ML16280A422
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Site:	Rhode Island Atomic Energy Commission
Issue date:	10/06/2016
From:	Goodwin C; State of RI and Providence Plantations, Governor
To:	; Office of Nuclear Reactor Regulation
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RHODE ISLAND NUCLEAR SCIENCE CENTER RESPONSE TO OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RENEWAL OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER RESEARCH REACTOR LICENSE NO. R-95; DOCKET NO. 50-193 Enclosure 1

1. The proposed TSs do not include an introduction section.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Technical Specifications, states, in part, that The applicant should be able to state conclusively that the technical specifications were prepared following an accepted format, that normal operation of the reactor within the limits of the technical specifications will not result in offsite radiation exposure in excess of 10 CFR Part 20 guidelines, and that the technical specifications limit the likelihood and consequences of malfunctions. Licensees should also be able to state that The technical specifications are neither derived nor justified in this chapter [14] of the SAR. They are determined by the analyses that appear in the other chapters of the SAR.

These understandings do not appear to be stated in the proposed RINSC TSs. Provide an introduction to the TSs that addresses these issues, or state your justification for a suitable alternative.

Response: Although it is not clear to us after reviewing NUREG 1537 appendix 14.1 and ANSI/ANS 15.1 that these references to a required introduction apply to the technical specification document, the following page will be added behind the cover page to address this issue:

Introduction The Rhode Island Nuclear Science Center (RINSC) Technical Specifications (TS) have been derived from the analyses and evaluation included in the RINSC Safety Analysis Report (SAR). They were prepared following the format as outlined in ANSI/ANS-15.1 and NUREG 1537 Part 1, appendix 14.1. Normal operation of the facility within the limits of these TS will ensure that offsite radiation exposure will remain below the limits outlined in 10 CFR 20 guidelines and will limit the likelihood and consequences of any malfunctions.

2. Proposed TS 1.0, Definitions, includes the following items identified by the NRC staff:

a. Proposed TS 1.6, Control Rod, states, A control rod is a device fabricated from neutron absorbing material, that is used to establish neutron flux changes and to compensate for routine reactivity losses. Proposed TS 1.32, Regulating Rod or Regulating Blade, states, The regulating rod is a control rod of low reactivity worth used primarily to maintain an intended power level. It is not required to have a scram capability, and may be controlled manually or by servo controller. Proposed TS 1.41, Shim Safety Rod or Shim Safety Blade, states, A shim safety rod is a control rod of high reactivity worth used primarily to make course adjustments to power level, and to provide a means for very fast reactor shutdown by having a scram capability. The NRC staff finds these definitions lack specific information on the neutron absorbing and cladding materials for the regulating and shim control rods.

Enclosure 1

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 1.2.2, Format, states that any information used to support the TSs should be explicitly referenced.

Provide a revised TSs 1.6, 1.32, and 1.41, as applicable, to include the neutron absorbing and cladding materials, or justify why no change is needed.

Response: We believe that the statement in section 1.2.2, Format, any information used to support the TSs should be explicitly referenced, is referring to individual specifications as addressed in paragraph 2 of section 1.2.2. In addition the requirement makes reference to sources used to support the TS which does not apply to the definitions. As stated in section 1.3 of appendix 14.1, specific definitions may be added to clarify the meaning of terms.

Section 1.3 also states that those definitions applicable to a particular facility should be included verbatim.

Having said that, TSs 1.6, 1.32 (now 1.30) and 1.41 (now 1.37) now read as follows, TS 1.6 and 1.32 are now verbatim as shown in ANSI/ANS 15.1 minus the inapplicable portions and 1.41 is facility specific but uses the same verbiage and level of detail as given in 1.6 and 1.30:

1.6 Control Rod A control rod is a device fabricated from neutron absorbing material that is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod can be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

1.30 Regulating Rod The regulating rod is a low worth control rod used primarily to maintain an intended power level that need not have scram capability. Its position may be varied manually or by servo-controller.

1.37 Shim Safety Blade A shim safety blade is a control rod of high reactivity worth used primarily to make course adjustments to power level, and to provide a means for very fast reactor shutdown by having scram capability.

b. Proposed TS 1.13 (now 1.11), Explosive Material, states, in part, that Explosive material is any solid or liquid which is either: [...]. The NRC staff finds that the proposed definition appears to be limited to material in a solid or liquid physical form, and notes that other types of materials, such as gases and metastable materials, can also have explosive potential, and seeks additional clarification in order to avoid potential confusion.

Enclosure 1

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 1.2.2, Format, states that any information used to support the TSs should be explicitly referenced.

Provide a revised TS 1.13 to indicate that the explosive materials used at the RINSC are limited to solid and liquid forms, propose a revised TS 1.13 definition that includes other forms used at RINSC, or justify why no change is needed.

Response: A revised TS definition 1.13 follows:

1.11 Explosive Material Explosive material is any material, in any form that is either chemically or otherwise energetically unstable or has the potential to produce a sudden expansion of the material usually accompanied by the production of heat and changes in pressure.

Specific references were also removed because there are many and they are constantly changing. This definition is generic enough to include all types of unstable material.

c. Proposed TS 1.28(now 1.26), Reactor Secured, defines conditions under which the RINSC reactor shall be considered secured, and includes proposed TS 1.28.3.1, which states that The reactor is shutdown. TS 1.29, Reactor Shutdown, provides a reactivity condition (0.75%k/k) for shutdown. However, the NRC staff finds that the criteria for the physical location of the control rods (e.g. fully inserted) appears to be missing.

ANSI/ANS-15.1-2007, Section 1.3, for reactor secured, item (2) (a), provides guidance that the minimum number of neutron absorbing control devices is fully inserted or other safety devices are in shutdown position, as required by technical specifications.

Provide a revised TS 1.28.3.1 to describe the physical location (e.g., all control rods fully inserted) for the neutron absorbing control devices, or justify why no change is needed.

Response: A revised TS definition 1.26 follows, it includes the requested change to 1.26.3.1:

1.26 Reactor Secured The reactor is secured when under optimal conditions of moderation and reflection either:

1.26.1 There is insufficient moderator available in the reactor to attain criticality, or 1.26.2 There is insufficient fissile material present in the reactor to attain criticality, or 1.26.3 The following conditions exist:

Enclosure 1

1.26.3.1 The minimum number of neutron absorbing control devices are fully inserted or other safety devices are in the shutdown position, as required by technical specifications, and; 1.26.3.2 The master switch is in the off position and the key is removed from the lock, and; 1.26.3.3 No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and; 1.26.3.4 No experiments are being moved or serviced that have a reactivity worth of greater than 0.6%k/k when moved.

d. The NRC staff finds that proposed TS 1.43, Shutdown Margin, may not be consistent with the guidance provided in ANSI/ANS-15.1-2007, as it is lacking descriptions regarding any permissible operating condition, and will remain subcritical without further operator action.

ANSI/ANS-15.1-2007, Section 1.3, shutdown margin includes reference to any permissible operating condition, and will remain subcritical without further operator action.

Provide a revised TS 1.43 to include the definition provided in ANSI/ANS-15.1-2007, or justify why no change is needed.

Response: A revised TS definition 1.43 (now 1.39) follows:

1.39 Shutdown Margin Shutdown Margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition and with the most reactive Shim Safety Blade and the Regulating Rod in their most reactive positions and that the reactor will remain subcritical without further operator action.

3. Proposed TSs in Section 4.0, Surveillance Requirements, have the following items identified by the NRC staff:

a. The NRC staff finds that proposed TS 4.2.3 and TS 4.2.4, which use the term verified to be operable [], is not defined in the TSs and does not clearly indicate the method used Enclosure 1

to determine the operability of the channels (i.e., channel check, channel test, and/or channel calibration).

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 4, Surveillance Requirements, states the surveillance-related specifications should clearly identifythe [surveillance] method.

Provide a revised TS 4.2.3 and TS 4.2.4 to describe the surveillance method, or justify why no change is needed.

Response: Added by performing a channel test to both TS 4.2.3 and 4.2.4, note that the word rate was also removed from 4.2.4.2. A no flow scram test is conducted prior to each start-up. The flow rate scram is tested annually and has been added to 4.2.5.5.1.

b. The NRC staff finds that proposed TS 4.2.5 is not clear as it does not state the surveillance test method used (i.e., a channel check, channel test, and/or channel calibration).

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 4, Surveillance Requirements, states the surveillance-related specifications should clearly identifythe [surveillance] method.

Provide a revised TS 4.2.5 to describe the surveillance method, or justify why no change is needed.

Response: Revised 4.2.5 to read shall be channel tested annually

c. The NRC staff finds that proposed TS 4.2.6 does not clearly state that a channel calibration is required, as defined in the proposed TSs, and this could cause potential confusion.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 4, Surveillance Requirements, states the surveillance-related specifications should clearly identifythe [surveillance] method.

Provide a revised TS 4.2.6 to clearly describe the surveillance method, or justify why no change is needed.

Response: Revised 4.2.6 to read The following reactor safety and safety related instrumentation shall have a channel calibration performed annually

d. The NRC staff finds that proposed TS 4.4.2.1 and TS 4.4.2.2 use the term functional test, which is not defined in the TSs, and use of the defined term operable could avoid potential confusion.

Enclosure 1

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 4, Surveillance Requirements, states the surveillance-related specifications should clearly identifythe [surveillance] method.

Provide a definition for functional in the proposed TSs. Provide a revised TSs 4.4.2.1 and 4.4.2.2 to use a term are already defined in the proposed TSs (e.g., operable), or justify why no change is needed.

Response: Section 4.4 has been rewritten and addresses this RAI. New section 4.4 follows:

4.4 Confinement System Applicability:

This specification describes the surveillance requirements for the Confinement System and components.

Objective:

The objective of this specification is to verify that the Confinement System is capable of performing its intended function prior to being utilized to support operations.

Specification:

4.4.1 Verify Confinement System is operable at least daily prior to any of the following conditions:

4.4.1.1 Reactor operations.

4.4.1.2 Handling of irradiated fuel.

4.4.1.3 Experiment handling for an experiment that has a significant fission product, or gaseous effluent activation product inventory, and for which the experiment is not inside a container.

4.4.1.4 Performing any work on the core or control rods that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.4.1.5 Performing any experiment movement that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.4.2 Verify the Confinement System remains operable during an initiation of a facility evacuation alarm.

4.4.2.1 Monthly 4.4.2.2 Following any maintenance that could affect the operability of the system 4.4.3 Verify the Confinement System remains operable during an initiation of a facility evacuation alarm concurrent with a loss of normal AC power to the facility.

4.4.3.1 Quarterly 4.4.3.2 Following any maintenance that could affect the operability of the system Enclosure 1

Basis:

By ensuring that the confinement system is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would be directed to the stack, mixed with dilution air, and detected by the stack radiation monitor system.

A periodic functional test of the confinement system under emergency conditions ensures that in the event of an airborne radioactivity release, the confinement system is capable of being activated. The testing periods that are specified conform to ANSI 15.1 recommendations.

e. The NRC staff finds that the specification in proposed TS 4.5.1, states in operation, which is not clearly defined in TS (e.g., operable, and operating, are defined terms in theTSs), and could cause confusion. In addition, proposed TS 4.5.1 appears to contain a typographical error as to whether the colon (:) in the proposed TS should be a period

(.).

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 1.2.2, Format, states that any information used to support the TSs should be explicitly referenced.

Provide a definition for in operation in the proposed TSs, revise TS 4.5.1 to use terms defined in the proposed TSs (e.g., operable or operating), or justify why no change is needed.

Response: Section 4.5 has been rewritten and addresses this RAI. New section 4.4 follows:

Enclosure 1

4.5 Confinement Ventilation System Applicability:

This specification describes the surveillance requirements for the Confinement Ventilation System.

Objective:

The objective of this specification is to verify that the Confinement Ventilation System is operable prior to performing activities that have any potential for a release of radioactivity and remains operable upon initiation of the Emergency Mode of operation.

Specification:

4.5.1 Verify the Confinement Ventilation System is operable at least daily prior to any of the following conditions:

4.5.1.1 Reactor operations.

4.5.1.2 Handling of irradiated fuel.

4.5.1.3 Experiment handling for an experiment that has a significant fission product, or gaseous effluent activation product inventory, and for which the experiment is not inside a container.

4.5.1.4 Performing any work on the core or control rods that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.5.1.5 Performing any experiment movement that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.5.2 Verify the Confinement Ventilation System Emergency Mode activates and maintains greater than a differential pressure of -0.5 WC during an initiation of a facility evacuation alarm.

4.5.2.1 Monthly 4.5.2.2 Following any maintenance that could affect the operability of the system 4.5.3 Verify the Confinement Ventilation System Emergency Mode activates and maintains greater than a differential pressure of -0.5 WC during an initiation of a facility evacuation alarm concurrent with a loss of normal AC power to the facility.

4.5.3.1 Quarterly 4.5.3.2 Following any maintenance that could affect the operability of the system 4.5.4 The Emergency Filter Bank shall be verified to be at least 99% efficient for removing iodine:

4.5.4.1 Biennially 4.5.4.2 Following any maintenance that could affect the operability of the system Enclosure 1

4.5.5 The ventilation flow through the Emergency Filter Bank shall be verified to be less than or equal to 1500 SCFM:

4.5.5.1 Biennially 4.5.5.2 Following any maintenance that could affect the operability of the system Bases:

By ensuring that the Confinement Ventilation System is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would be directed to the stack and be detected by the stack radiation monitor system and that the system is capable of supporting the Emergency Mode of operation.

A periodic test of the operability of the Confinement Ventilation System in the Emergency Mode of operation ensures that in the event of an airborne radioactivity release, the Confinement Ventilation System Emergency Mode will: 1) activate and realign as required,

2) maintain a flow rate through the filter bank less than or equal to 1500 SCFM and 3) remove at least 99% of the iodine from the exhaust air. The testing periods that are specified conform to ANSI 15.1 recommendations.

f. The NRC staff finds that proposed TS 4.6.2, states functional test, which is not clearly defined in the TSs (e.g., operable is defined) and could cause confusion.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 1.2.2, Format, states that any information used to support the TSs should be explicitly referenced.

Provide a definition for functional in the proposed TSs, revise TS 4.6.2 to use a defined term (e.g., operable), or justify why no changes are needed.

Response: Section 4.5 has been rewritten and addresses this RAI. New section 4.4 follows:

Enclosure 1

4.6 Emergency Power System Applicability:

These specifications describe the surveillance requirements for the Emergency Power System.

Objective:

The objective of these specifications is to verify that the emergency power system is operable and will perform its intended function.

Specifications:

4.6.1 Verify the Emergency Power System is operable at least daily prior to any of the following conditions:

4.6.1.1 The reactor is operating.

4.6.1.2 Irradiated fuel handling is in progress.

4.6.1.3 Experiment handling is in progress for an experiment that has a significant fission product, or gaseous effluent activation product inventory, and for which the experiment is not inside a container.

4.6.1.4 Any work on the core or control rods that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.6.1.5 Any experiment movement that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.6.2. Perform an operability test to verify that the Emergency Power System starts and loads in the event of a facility power outage.

4.6.2.1. Quarterly 4.6.2.2. Following emergency system load changes 4.6.3. Verify the fuel tank levels for the emergency generator are at least 50% full.

4.6.3.1 Monthly Bases:

Specification 4.6.1: By ensuring that the Emergency Power System is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that in the event of a loss of facility AC power while the Emergency Mode of operation is required to mitigate a potential release, emergency power would be available for the components of the Confinement and Confinement Ventilation Systems to perform their intended function.

Enclosure 1

Specification 4.6.2 periodically tests the emergency power system to ensure that in the event of a facility power outage, the emergency power system would automatically start, and be capable of handling the load required to power the emergency confinement systems.

Initiation of the emergency mode of operation connects the emergency blower and the dilution blower to the emergency power supply. The testing periods that are specified conform to ANSI 15.1 recommendations.

Specification 4.6.3 ensures that there is sufficient fuel to power the emergency generator under full load for approximately 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

g. The NRC staff finds that proposed TS 4.7.1.2 and TS 4.7.2.1, state calibrated annually, and the TSs would be more consistent with the guidance provided in ANSI/ANS-15.1-2007, to use the term channel calibration.

ANSI/ANS-15.1-2007, Section 1.3, provides guidance for use of the definition of channel calibration.

Revise TS 4.7.1.2 and TS 4.7.2.1 to use channel calibration, or justify why no change is needed.

Response: Added channel to 4.7.1.2. 4.7.2.1 changed to 4.7.2.2 to accommodate new 4.7.2.1 and revised as follows: (there is no installed Effluent Radiation Monitoring system) 4.7.2.1 Airborne Effluents Applicability:

This specification applies to the monitoring of airborne effluents from the Rhode Island Nuclear Science Center (RINSC).

Objective:

The objective of this specification is to assure that the release of airborne radioactive material from the RINSC will not cause the public to receive doses that are greater than the limits established in 10 CFR 20.

Specification:

The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will be calculated annually.

Basis:

10 CFR 20.1101(d) states, in part, to implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon - 222 and its daughters, shall be established by licensees other than Enclosure 1

those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions.

Since the Rhode Island Nuclear Science Center is located on Narragansett Bay, the wind does not blow in the same direction more than about 10% of the time as shown in the following table taken from historical wind rose data.

Wind Blowing From Frequency % Wind Blowing From Frequency %

North 6.20 E-02 6.02 South 5.80 E-02 5.80 North/Northeast 5.80 E-02 5.80 South/Southwest 8.40 E-02 8.40 Northeast 4.40 E-02 4.40 Southwest 1.05 E-01 10.50 East/Northeast 1.30 E-02 1.30 West/Southwest 6.40 E-02 6.40 East 1.20 E-02 1.20 West 6.80 E-02 6.80 East/Southeast 1.30 E-02 1.30 West/Northwest 9.50 E-02 9.50 Southeast 5.80 E-02 6.80 Northwest 1.04 E-01 10.40 South/Southeast 4.90 E-02 4.90 North/Northwest 6.80 E-02 6.80 Table 3.1: Historical Wind Rose Data Enclosure 1

Thus, during routine operations, no individual would be in the pathway of the plume more than about 10% of the time. Calculations of annual dose equivalent due to the primary airborne effluent, Argon - 41, using the COMPLY Code show less than the allowable ALARA limitation given in 10 CFR 20.1101 for the hypothetical maximum exposed individual member of the general public.

4.7.2.2 Liquid Effluent Sampling Applicability:

This specification applies to the monitoring of radioactive liquid effluents from the Rhode Island Nuclear Science Center.

Objective:

The objective of this specification is to assure that exposure to the public resulting from the release of liquid effluents will be within the regulatory limits and consistent with as low as reasonably achievable requirements.

Specification:

The liquid waste retention tank discharge shall be batch sampled and the gross activity per unit volume determined to be less than the limits set in 10 CFR 20 before release.

Basis:

10 CFR 20.2003 permits discharges to the sanitary sewer provided that conditions in 10 CFR 20.2003 (a) are met.

4. The requirements for LCO and SR TSs are established in 10 CFR 50.36(c)(2), Limiting conditions for operation, and 10 CFR 50.36(c)(3), Surveillance requirements, respectively.

The cited regulations state that LCOs are required to have a companion SR to provide assurance that the LCO will be met. Additionally, NUREG-1537, Part 1, Chapter 14, Appendix 14.1, and ANSI/ANS-15.1-2007 provide guidance regarding TSs, including LCOs and SRs.

a. Some proposed RINSC LCOs do not appear to have a corresponding SR, as noted below.

i. Proposed TS 3.1.1.1.4, states: The reactor shall be subcritical by at least 3.0 %k/k during fuel loading changes.

Enclosure 1

Provide an appropriate corresponding SR, or justify why no such SR should be required.

Response: LCO 3.1.1.1.3 was moved to section 5, therefore 3.1.1.1.4 and 4.1.1.1.4 are now 3.1.1.1.3 and 4.1.1.1.3.

Added:

4.1.1.1.3 The core shutdown reactivity shall be determined to remain greater than 3 %K/K prior to and during fuel loading changes.

Bases:

Specification 4.1.1.1.3 requires that the core shutdown reactivity shall be determined to remain greater than 3 %K/K prior to and during fuel loading changes. This limit on shutdown reactivity while moving fuel ensures a margin of safety while changing fuel configuration with no additional negative reactivity insertion available via control and safety systems and accounts for latest fuel burnup levels and new core loading.

ii. Proposed TS 3.1.2.1, states All core grid positions shall contain fuel elements, baskets, reflector elements, or experimental facilities during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.

Provide an appropriate corresponding SR, or justify why no such SR should be required.

Response: Added:

4.1.2.1 Prior to the first reactor start-up of the day with expected power operation greater than .1 MW, inspect the core to confirm that all grid positions contain fuel elements, baskets, reflector elements, or experimental facilities.

Bases:

Specification 4.1.2.1 requires that all of the core grid spaces be filled when the reactor is operated at higher power levels that require forced convection cooling. This inspection prior to each start-up ensures that the core configuration has not been changed and that all forced coolant flow will be through the core components as designed and not bypassed through an unoccupied grid location.

Enclosure 1

iii. Proposed TS 3.1.2.2, states The pool gate shall be in its storage location during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.

Provide an appropriate corresponding SR, or justify why no such SR should be required.

Response: Added:

4.1.2.2 Prior to the first reactor start-up of the day with expected power operation greater than .1 MW, inspect to ensure that the pool gate is in its storage location.

Bases:

Specification 4.1.2.2 requires that the pool gate that is used for separating the sections of the pool, be in its storage location when the reactor is in operation at higher power levels that require forced convection cooling. This inspection ensures that the full volume of the pool water is available to support reactor operation.

iv. Proposed TS 3.2.1.2, states The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02%k/k per second. Proposed TS 4.2.2, states All shim safety reactivity insertion rates shall be measured []. Proposed TS 4.2.2 would impose a SR for reactivity insertion rates for individual shim safety rods, but would not impose a SR for reactivity insertion rates for regulating rods.

Provide a revised TS 4.2.2 such that it would impose a SR for reactivity insertion rate for regulating rods, or justify why no such SR should be required.

Response: Changed as follows: (changed the numbering so that the LCO and the SR numbers match for ease of use.)

Enclosure 1

Specifications:

The reactor shall not be operated unless:

3.2.1 All shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated.

3.2.2 Only one shim safety blade can be raised at a time and the total reactivity insertion rates of any individual shim safety and the regulating rod does not exceed 0.02%k/k per second.

3.2.3 The instrumentation shown in Table 3.1 is operable and capable of performing its intended function:

4.2.1 Shim safety drop times shall be measured:

4.2.1.1 Annually 4.2.1.2 Whenever maintenance is performed which could affect the drop time of the blade 4.2.1.3 When a new core is configured 4.2.1.4 Following control blade changes 4.2.2 Verify that only one shim safety blade can be withdrawn at a time and measure each shim safety blade and regulating rod reactivity insertion rates:

4.2.2.1 Annually 4.2.2.2 Whenever maintenance is performed which could affect the reactivity insertion rate of the blade 4.2.2.3 When a new core is configured 4.2.2.4 Following control blade changes

v. Proposed TS 3.5.2.4, states The Emergency Filter System Exhaust Absolute Filter shall be certified by the manufacturer to have a minimum efficiency of 99.97% for removing 0.3 micron diameter particulates.

Provide an appropriate corresponding SR, which clarifies the interval at which this certification is performed, and if it is only performed once prior to initial filter installation, also clarifies the interval at which the filter is replaced; or, justify why no such SR should be required.

Provide a SR that ensures that the manufacturers certified state minimum efficiency remains effective throughout the surveillance interval or, justify why no change is needed Response: This is a design feature which was for original installation and or replacement of the emergency filter train. It is not a surveillable feature.

Enclosure 1

3.5.2.4 is being deleted and this requirement moved to section 5 of the TS. Additionally, section 3.5 has been rewritten as follows:

Enclosure 1

3.5 Confinement Ventilation System Applicability:

These specifications apply to the Confinement Ventilation System including all components listed below and any interconnecting duct work that allows the system to perform its intended function:

1. Confinement Exhaust Blower

2. Off-gas Blower

3. Rabbit Blower

4. Dilution Blower

5. Emergency Exhaust Blower

6. Confinement Ventilation Intake Damper

7. Confinement Ventilation Exhaust Damper

8. Emergency Exhaust Air Filter Bank

9. Confinement Exhaust Stack

10. Facility Evacuation System Objective:

The objective of this specification is to assure that the Confinement Ventilation System is capable of performing its intended function.

Specification:

3.5.1 The Confinement Ventilation System shall be operable whenever:

3.5.1.1 The reactor is operating.

3.5.1.2 Irradiated fuel handling is in progress.

3.5.1.3 Experiment handling is in progress for an experiment that has a significant fission product, or gaseous effluent activation product inventory, and for which the experiment is not inside a container.

3.5.1.4 Any work on the core or control rods that could cause a reactivity change of more than 0.60 %k/k is in progress.

3.5.1.5 Any experiment movement that could cause a reactivity change of more than 0.60 %k/k is in progress.

Basis:

The Confinement Ventilation System maintains a minimum differential pressure of -0.5 WC across the Confinement System discussed in Section 3.4 to ensure that all confinement air pathways are through a controlled pathway that is monitored for radiological release.

Enclosure 1

Under emergency conditions, when a facility evacuation is initiated, the Confinement Ventilation System realigns to isolate the Confinement Building and while maintaining the differential pressure of -0.5 WC directs all confinement air through an Emergency Filter Bank containing charcoal filters designed to remove any radioactive iodine that would be expected to be released during a fuel failure. An airflow limit of 1500 cfm though the filter ensures that the flow rate is low enough to allow the charcoal filter to adsorb at least 99% of the iodine that would be expected to be released in the event of a fuel cladding failure. The Emergency Filter Bank also contains absolute filters which prevent charcoal particulates from the charcoal filter from being released to the building exhaust air stream. The Dilution Blower remains running and provides a non-contaminated source of air to mix with the confinement air, so that any airborne radioactivity that is released is diluted prior to release.

Though the safety analysis does not take credit for the Confinement Exhaust Stack, the stack ensures that the plume of confinement air that is released to the environment, is released at an elevation of 115 feet above ground level, which provides for an opportunity for the air to disperse prior to the plume reaching ground level.

vi. Proposed TS 3.6.2.1 provides requirements that the Emergency Blower (TS 3.6.2.1.1) and Dilution Blower (TS 3.6.2.1.2) shall be connected to an emergency power source. Proposed TS 4.6.1, states A quarterly operability test shall be performed in order to verify that the emergency power system starts in the event of a facility power outage. Proposed TS 4.6.2, states A functional test of the emergency power system under load shall be performed: []. The NRC staff is not clear how proposed TSs 4.6.1 and 4.6.2 verify that the emergency blower and dilution blower are connected to the emergency power source.

Propose a revised TSs 4.6.1 and 4.6.2 such that they indicate that operation of the Emergency Blower and Dilution Blower demonstrate that emergency power is connected, or justify why no change is needed.

Response: Section 4.6 has been rewritten as follows:

Enclosure 1

4.6 Emergency Power System Applicability:

These specifications describe the surveillance requirements for the Emergency Power System.

Objective:

The objective of these specifications is to verify that the emergency power system is operable and will perform its intended function.

Specifications:

4.6.1 Verify the Emergency Power System is operable at least daily prior to any of the following conditions:

4.6.1.1 The reactor is operating.

4.6.1.2 Irradiated fuel handling is in progress.

4.6.1.3 Experiment handling is in progress for an experiment that has a significant fission product, or gaseous effluent activation product inventory, and for which the experiment is not inside a container.

4.6.1.4 Any work on the core or control rods that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.6.1.5 Any experiment movement that could cause a reactivity change of more than 0.60 %k/k is in progress.

4.6.2. Perform an operability test to verify that the Emergency Power System starts and loads (see TS 4.5.3) in the event of a facility power outage.

4.6.2.1. Quarterly 4.6.2.2. Following emergency system load changes 4.6.3. Verify the fuel tank levels for the emergency generator are at least 50% full.

4.6.3.1 Monthly Bases:

Specification 4.6.1: By ensuring that the Emergency Power System is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that in the event of a loss of facility AC power while the Emergency Mode of operation is required to mitigate a potential release, emergency power would be available for the components of the Confinement and Confinement Ventilation Systems to perform their intended function.

Enclosure 1

Specification 4.6.2 periodically tests the emergency power system to ensure that in the event of a facility power outage, the emergency power system would automatically start, and be capable of handling the load required to power the emergency confinement systems.

Initiation of the emergency mode of operation connects the emergency blower and the dilution blower to the emergency power supply. The testing periods that are specified conform to ANSI 15.1 recommendations.

Specification 4.6.3 ensures that there is sufficient fuel to power the emergency generator under full load for approximately 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

vii. Proposed TS 3.7.1.1.2, states If the detector described in specification 3.7.1.1.1 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. Proposed TSs 4.7.1.1.3, 4.7.1.1.4, 4.7.1.2.3, and 4.7.1.2.4 impose SRs on the gaseous effluent air monitor and the particulate air monitor. The NRC staff is not clear if the SRs required by TSs 4.7.1.1.3, 4.7.1.1.4, 4.7.1.2.3, and 4.7.1.2.4 are intended to extend to the alternative monitor that would be allowed by proposed TS 3.7.1.1.2.

Provide a revised TSs 4.7.1.1.3, 4.7.1.1.4, 4.7.1.2.3, and 4.7.1.2.4 to clearly delineate if they extend to the alternate monitor, propose a SR for TS 3.7.1.1.2, or justify why no changes are needed.

Response: 3.7.1.1.2 says if the detector fails during operation, The SRs are prior to start-up or after repair, so no, the SRs do not extend to the alternative required to continue operations after you are already started up.

viii. Proposed TS 3.7.1.1.4, states If the detector described in specification 3.7.1.1.3 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. Proposed TS 4.7.1.1.2 and TS 4.7.1.2.2 impose SRs on the pool top area radiation monitor. However, the NRC staff is not clear if the SRs imposed TSs 4.7.1.1.2 and 4.7.1.2.2 are intended to extend to the alternative meter.

Provide a revised TS 4.7.1.1.2 and TS 4.7.1.2.2 to include the alternative meter, propose another SR for the alternative meter in TS 3.7.1.1.4, or justify why no changes are needed.

Response: 3.7.1.1.4 says if the detector described in 3.7.1.1.3 fails, 3.7.1.1.3 is the permanently installed detector normally used to fulfill the requirement of having at least one monitor over the pool whenever the conditions of 3.7.1.1.3 are met. 3.7.1.1.4 is a temporary alternative in the event that the permanent detector fails while operating. The SRs are for permanent equipment, not temporary replacements. The SRs do not extend directly to Enclosure 1

the suitable alternative, however, 3.7.1.1.4 will be changed to read as follows:

3.7.1.1.4 If the detector described in specification 3.7.1.1.3 fails, within one hour, place a suitable gamma sensitive alternative meter with alarming capability meeting all of the requirements as the detector originally used to satisfy 3.7.1.1.3 in service over the pool in lieu of the normal detector.

ix. Proposed TS 3.7.1.1.5, states Passive radiation monitors provided by a certified vendor shall be used to provide area radiation monitoring inside confinement when the reactor is in operation, and during fuel handling operations.

Proposed TSs 4.7.1.1.1 and 4.7.1.2.1 would impose SRs on the experimental level area radiation monitor.

Clarify whether proposed TSs 4.7.1.1.1 and 4.7.1.2.1 are intended to provide SRs for the radiation monitor required by proposed TS 3.7.1.1.5. If so, revise TS 3.7.1.1.5 and/or TSs 4.7.1.1.1 and 4.7.1.2.1 to clarify that this is the case; if not, provide an appropriate corresponding SR for TS 3.7.1.1.5; or, justify why no such SR should be required.

Response: 3.7.1.1.5 has been deleted. These monitors are part of the RINSC radiation protection and management program and are not required to be addressed in TS.

x. Proposed TS 3.7.1.1.6, states, Passive radiation monitors provided by a certified vendor shall be used to provide environmental radiation monitoring when the reactor is in operation, and during fuel handling operations.

Provide an appropriate corresponding SR, or justify why no such SR should be required.

Response: 3.7.1.1.6 has been deleted. These monitors are part of the RINSC radiation protection and management program and are not required to be addressed in TS.

xi. Proposed TS 3.7.1.2.4, states Alarm set points may be adjusted higher with the approval of the Director or one of the Assistant Directors. This TS does not have a corresponding SR.

Provide an appropriate corresponding SR, or justify why no such SR should be required.

Response: Deleted 3.7.1.2.4. Although actual set points may change, this would be based on changing normal levels based on facility conditions. The 2.5 and 2.0 times normal levels addressed by 3.7.1.2.1 and 2 will not be changed.

Enclosure 1

xii. The specification in proposed TS 3.7.2.1, states The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will be calculated using a generally-accepted computer program.

Provide an appropriate corresponding SR (i.e., requiring that this calculation be performed, and compliance with 10 CFR Part 20, Standards for Protection Against Radiation, limits determined, at a specific interval), or justify why no such SR should be required.

Response: Added 4.7.2.1 as follows:

4.7.2.1 Airborne Effluents Applicability:

This specification applies to the monitoring of airborne effluents from the Rhode Island Nuclear Science Center (RINSC).

Objective:

The objective of this specification is to assure that the release of airborne radioactive material from the RINSC will not cause the public to receive doses that are greater than the limits established in 10 CFR 20.

Specification:

The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will be calculated annually.

Basis:

10 CFR 20.1101(d) states, in part, to implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon - 222 and its daughters, shall be established by licensees other than those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions.

Since the Rhode Island Nuclear Science Center is located on Narragansett Bay, the wind does not blow in the same direction more than about 10% of the time as shown in the following table taken from historical wind rose data.

Table 3.3: Historical Wind Rose Data Enclosure 1

Wind Blowing From Frequency % Wind Blowing From Frequency %

North 6.20 E-02 6.02 South 5.80 E-02 5.80 North/Northeast 5.80 E-02 5.80 South/Southwest 8.40 E-02 8.40 Northeast 4.40 E-02 4.40 Southwest 1.05 E-01 10.50 East/Northeast 1.30 E-02 1.30 West/Southwest 6.40 E-02 6.40 East 1.20 E-02 1.20 West 6.80 E-02 6.80 East/Southeast 1.30 E-02 1.30 West/Northwest 9.50 E-02 9.50 Southeast 5.80 E-02 6.80 Northwest 1.04 E-01 10.40 South/Southeast 4.90 E-02 4.90 North/Northwest 6.80 E-02 6.80 Thus, during routine operations, no individual would be in the pathway of the plume more than about 10% of the time. Calculations of annual dose equivalent due to the primary airborne effluent, Argon - 41, using the COMPLY Code show less than the allowable ALARA limitation given in 10 CFR 20.1101 for the hypothetical maximum exposed individual member of the general public.

xiii. Proposed TS 3.9.3, Experimental Facilities, would impose requirements on experimental facility configuration during reactor operation and within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following reactor shutdown. Although ANSI/ANS-15.1-2007 states that specific surveillance activities for experiments are typically not part of the TSs, the LCOs in proposed TS 3.9.3 relate to permanently installed experimental facilities (beam ports) at the RINSC, rather than specific experiments. Therefore, the LCOs in proposed TS 3.9.3 should have corresponding SRs.

Provide appropriate corresponding SRs for TS 3.9.3 (including TSs 3.9.3.1.1, 3.9.3.1.2, 3.9.3.1.3, 3.9.3.1.4, 3.9.3.2.1, 3.9.3.2.2.1, and 3.9.3.2.2.2), or justify why no such SRs should be required.

Response: Sections 3.9.3 and 4.9.3 have been changed as follows:

3.9.3 changed as follows:

3.9.3 Experimental Facilities 3.9.3.1 Experimental Facility Configuration during Reactor Operation, Including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Applicability:

These specifications apply to the surveillance of reactor experimental facilities during reactor operation, including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Enclosure 1

Objective:

The objective of these specifications is to 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 a 1.25 inch diameter opening to confinement during reactor operation.

3.9.3.1.2 The drain valve from the through port shall be closed when the though port is in use.

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

3.9.3.1.4 When the through port is not physically manned and monitored, the ends of the through 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 square inches (equivalent to a 1.37 inch diameter opening) 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 opening on each experimental port that is open to confinement to 1.25 inch diameter 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 drain/vent lines which join together and have a 1/2 orifice restriction and both ends, if open. By keeping the drain valve closed during through port use, this potential leak pathway is blocked, and all any leakage would have to come out one of the ends. The potential for an unnoticed pool leak though this experimental facility is minimized.

Specification 3.9.3.1.3: 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/experimental 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.

Enclosure 1

Specification 3.9.3.1.4: This specification ensures that if the through port is not being monitored for the event of a pool leak, the ends are sealed so that the through port is not a LOCA pathway.

Enclosure 1

3.9.3.2 Experimental Facility Configuration Within The 4.5 Hour Period After Shutdown Applicability:

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

Objective:

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

Specifications:

3.9.3.2.1 If there is no need to open a beam port within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after reactor shutdown, then the 1.25 inch diameter opening limit addressed in 3.9.3.1.1 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 the 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after reactor shutdown, then prior to exceeding the 1.25 inch diameter opening limit addressed in 3.9.3.1.1, the following actions will be taken:

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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is 1.48 square inches (equivalent to a 1.37 inch diameter opening). Consequently, maintaining the limit on the size of the opening between confinement and the beam ports to 1.25 inches in diameter for a period of 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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.

Enclosure 1

Specification 3.9.3.2.2: In the event that access to a beam port is needed within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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.

Added 4.9.3 as follows:

4.9.3 Experimental Facilities 4.9.3.1 Experimental Facility Configuration During Reactor Operation, Including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Applicability:

These specifications apply to the surveillance of reactor experimental facilities during reactor operation, including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Objective:

The objective of these surveillances is to 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:

Prior to operating the reactor, the following conditions will be met:

4.9.3.1.1 Each beam port shall have no more than a 1.25 inch diameter opening to confinement during reactor operation.

4.9.3.1.2 The drain valve from the through port shall be closed when the though port is in use.

4.9.3.1.3 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.

4.9.3.1.4 When the through port is not physically manned and monitored, the ends of the through port shall be closed.

Bases:

Specification 4.9.3.1.1: The LOCA analysis shows that as long as the pool level does not drain through an area greater than 1.48 square inches 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 Enclosure 1

drain pathways. Consequently, limiting the areas of each experimental port that is open to confinement to 1.25 inch diameter is conservative.

Specification 4.9.3.1.2: 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. By keeping the drain valve closed during through port use, that potential leak pathway is blocked and the through ports ends will either be closed or continuously manned IAW 4.9.3.1.3 and 4.9.3.1.4.

Specification 4.9.3.1.3: 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/experimental personnel will become aware of a pool leak though the through port reasonably quickly, with the gate valves in place, the consequence of the leak can be mitigated quickly by closing the valves.

Specification 4.9.3.1.4: This specification ensures that if the through port is not being monitored for the event of a pool leak, the ends are sealed so that the through port is not a LOCA pathway.

4.9.3.2 Experimental Facility Configuration Within The 4.5 Hour Period After Shutdown Applicability:

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

Objective:

The objective of these specifications is to 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:

4.9.3.2.1 Prior to opening a beam port verify that the reactor has not operated in the previous 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

4.9.3.2.2 If opening a beam port within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after reactor shutdown is absolutely required, then:

4.9.3.2.2.1. 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.

Enclosure 1

4.9.3.2.2.2. 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 4.9.3.2.1 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is 1.48 square inches (equivalent to a 1.37 inch diameter opening). Consequently, maintaining the limit on the size of the opening between confinement and the beam ports to 1.25 inches in diameter for a period of 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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 4.9.3.2.2 In the event that access to a beam port is needed within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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.

b. The NRC staff finds the following proposed TS SRs do not appear to have a corresponding LCO:

i. Proposed TS 4.1.1.2.2, states The reactivity worth of the shim safety rods shall be determined: [].

Identify the intended corresponding existing LCO, provide an appropriate LCO, or justify why no such LCO should be required.

Response: Deleted 4.1.1.2.2, there is no LCO for the shim safety blades, although the shim safety blade worth is necessary to perform 3.1.1.1.1 and 3.1.1.1.2, there is no limiting condition and the value is directly dependent on core design and fuel loading.

ii. Proposed TS 4.2.5.7.1 would impose a SR for the bridge manual scram.

Identify the intended corresponding existing LCO, provide an appropriate LCO, or justify why no such LCO should be required.

Enclosure 1

Response: Deleted 4.2.5.7.1, this is not a required channel listed in table 3.1 therefor no SR is required.

iii. The specification in proposed TS 4.3.1.3, states The primary coolant level shall be verified to be greater than or equal to the Limiting Safety System Setting value prior to the initial start-up each day that the reactor is started up from the shutdown condition. Proposed TS 2.2.1.2 would impose a LSSS for the coolant level, but the proposed RINSC TSs do not appear to contain an LCO for primary coolant level (additionally, the guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.3, Subsection (3), states that pool water level should be an LCO).

Provide an appropriate corresponding LCO, or justify why no such LCO should be required.

Response: Table 3.1 addresses the pool level scram, SR 4.2.5.6 performs an annual channel test to ensure the channel is operable. This is a prudent, prior to start-up check that is performed as part of the operators pool level walk around. If pool level was less than the LSSS value, it would be noticeable during the walk down and if not noted by the walk around it would not be possible to reset scram relays or the low pool level annunciator. No LCO is required.

iv. Proposed TS 4.3.1.4, states The components of the primary coolant system shall be inspected annually.

Provide an appropriate corresponding LCO, or justify why no such LCO should be required.

Response: This a maintenance SR that insures the integrity and reliability of the primary cooling systems is maintained and fully supports the required flows and temperature requirements already addressed by LCOs. No LCO is required, if no cooling system is operable, the reactor will not be operated at power levels greater than .1 MW as restricted by LCOs already in place.

v. Proposed TS 4.3.2.2, states The components of the secondary coolant system shall be inspected annually.

Provide an appropriate corresponding LCO, or justify why no such LCO should be required.

Response: See previous response for primary system.

vi. Proposed TS 4.6.3, states The fuel tank levels for the emergency generator shall be verified to be at least 50% full on a monthly basis. Provide an appropriate corresponding LCO, or justify why no such LCO should be required.

Enclosure 1

Response: LCO 3.6.1 addresses emergency power source operability whenever the confinement system is required. Both 4.6.2 and 4.6.3 support that LCO.

If all SRs are not satisfied, the Emergency Power System is not operable.

No additional LCO required.

vii. The specification in proposed TS 4.7.2.1, states The monitoring equipment used to measure the radioactive concentrations in the waste retention tanks shall be calibrated annually.

Provide an appropriate corresponding LCO, or justify why no such LCO should be required.

Response: Deleted this SR, This could be any of multiple pieces of monitoring equipment. 10 CFR 20 requires monitoring instrumentation to be calibrated annually and is standard practice. A SR is not required to ensure that this is done.

viii. Proposed TS 4.9.1.2. would impose SRs (TSs 4.9.1.2.1., 4.9.1.2.2., and 4.9.1.2.3.)

requiring that the beryllium reflectors be periodically inspected and functionally fit into the core grid box.

Provide an appropriate corresponding LCOs for TSs 4.9.1.2.1., 4.9.1.2.2., and 4.9.1.2.3., or justify why no such LCOs should be required.

Response: These SRs are in support of LCO 3.9.1. The basis for this LCO is that it is conservatively below a value at which it was determined that reflector damage could occur. These additional SRs for visual inspection are to verify that reflectors are not deteriorating in any way prior to the fluence limit being reached. No additional LCO is required.

5. Proposed TS 3.1.1.3.1 and TS 3.1.1.3.2 provide an exception for the reactivity worth of an experiment that states except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies. The guidance provided in ANSI/ANS-15.1-2007, Section 1.2.1, Purpose, states, in part, that Specific limitations and equipment requirements for safe reactor operation and for dealing with abnormal situation, typically derived from the Safety Analysis Report [...].

a. The NRC staff finds that TS 3.1.1.3.1 and TS 3.1.1.3.2 would allow experiments an unlimited reactivity, and this condition not supported by the SAR.

Provide a revised TSs 3.1.1.3.1 and 3.1.1.3.2 to remove the exception when the reactor is operated for the purpose of measuring experiment reactivity, or justify why no change is needed.

Response: Modified 3.1.1.3.1 and 3.1.1.3.2 as follows:

Enclosure 1

3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits when taking the reactor critical:

3.1.1.3.1.1 Total Moveable and Fixed 0.6 %k/k 3.1.1.3.1.2 Total Moveable 0.08 %k/k 3.1.1.3.2 The maximum reactivity worth of any individual experiment shall not exceed the following limits when taking the reactor critical:

3.1.1.3.2.1 Fixed 0.6 %k/k 3.1.1.3.2.2 Moveable 0.08 %k/k This will allow us to perform controlled criticality studies for unknown value experiments with the reactor in a subcritical state. This is a necessity because the reactivity value of all experiments cannot always be determined solely by calculation. The bases remains the same.

b. The NRC staff thinks that proposed TS 3.1.1.3.1.1 appears to contain a typographical error where 0.6%k.k should be 0.6%k/k.

Propose a revised TS 3.1.1.3.1.1 to change 0.6%k.k to 0.6%k/k, or justify why no change is needed.

Fixed.

6. The NRC staff finds some of the LCOs appear to be action statements that provide allowable deviations from other LCOs, but a completion time for the action statements is not clear.

ANSI/ANS-15.1-2007, states that deviations from LCOs may be allowed under specified conditions, such as action statements.

a. Proposed TS 3.7.1.1.2, states If the detector described in specification 3.7.1.1.1 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.

However, no completion time is provided for use of the alternative method.

Provide a revised TS 3.7.1.1.2 to include a completion time or, justify why no changes are needed.

Response: Revised as follows:

3.7.1.1.2 If the detector described in specification 3.7.1.1.1 fails during operation, within one hour, place in service a suitable alternative air monitor or begin an hourly grab sample analysis (grab sample analysis applies to particulate only) in lieu of having a functioning monitor.

b. Proposed TS 3.7.1.1.4, states If the detector described in specification 3.7.1.1.3 fails, a suitable gamma sensitive alternative meter with alarming capability may be placed at the Enclosure 1

top of the bridge in lieu of the normal detector. However, no completion time is provided for use of the alternative meter.

Provide a revised TS 3.7.1.1.4 to include a completion time, or justify no change is needed.

Response: Revised as follows:

3.7.1.1.4 If the detector described in specification 3.7.1.1.3 fails, within one hour, place a suitable gamma sensitive alternative meter with alarming capability meeting all of the requirements as the detector originally used to satisfy 3.7.1.1.3 in service over the pool in lieu of the normal detector.

c. Proposed TS 3.9.3.2.1, states If there is no need to open a beam port within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after reactor shutdown, then the 1.25 square inch area opening to confinement shall be maintained until that time period has passed. The NRC staff is not clear whether the opening must be reduced to a 1.25 square inch area before there is no need to open a beam port, or within a certain timeframe after there is no need to open a beam port.

Provide a revised TS 3.9.3.2.1 to include a completion time for reducing the opening size or eliminate the conditional logic, or justify why no changes are needed.

Response: Sections 3.9.3 and 4.9.3 have been revised as follows:

3.9.3 Experimental Facilities 3.9.3.1 Experimental Facility Configuration during Reactor Operation, Including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Applicability: These specifications apply to the surveillance of reactor experimental facilities during reactor operation, including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Objective: The objective of these specifications is to 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.

Enclosure 1

Specifications:

3.9.3.1.1 Each beam port shall have no more than a 1.25 inch diameter opening to confinement during reactor operation.

3.9.3.1.2 The drain valve from the through port shall be closed when the though port is in use.

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

3.9.3.1.4 When the through port is not physically manned and monitored, the ends of the through 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 square inches (equivalent to a 1.37 inch diameter opening) 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 opening on each experimental port that is open to confinement to 1.25 inch diameter 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 drain/vent lines which join together and have a 1/2 orifice restriction and both ends, if open. By keeping the drain valve closed during through port use, this potential leak pathway is blocked, and all any leakage would have to come out one of the ends.

The potential for an unnoticed pool leak though this experimental facility is minimized.

Specification 3.9.3.1.3:

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/experimental 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.

Specification 3.9.3.1.4:

Enclosure 1

This specification ensures that if the through port is not being monitored for the event of a pool leak, the ends are sealed so that the through port is not a LOCA pathway.

3.9.3.2 Experimental Facility Configuration Within The 4.5 Hour Period After Shutdown Applicability:

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

Objective:

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

Specifications:

3.9.3.2.1 If there is no need to open a beam port within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after reactor shutdown, then the 1.25 inch diameter opening limit addressed in 3.9.3.1.1 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 the 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after reactor shutdown, then prior to exceeding the 1.25 inch diameter opening limit addressed in 3.9.3.1.1, the following actions will be taken:

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:

Enclosure 1

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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is 1.48 square inches (equivalent to a 1.37 inch diameter opening). Consequently, maintaining the limit on the size of the opening between confinement and the beam ports to 1.25 inches in diameter for a period of 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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.

4.9.3 Experimental Facilities 4.9.3.1 Experimental Facility Configuration During Reactor Operation, Including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Applicability:

These specifications apply to the surveillance of reactor experimental facilities during reactor operation, including a 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months period after shutdown.

Objective:

The objective of these surveillances is to 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:

Prior to operating the reactor, the following conditions will be met:

4.9.3.1.1 Each beam port shall have no more than a 1.25 inch diameter opening to confinement during reactor operation.

4.9.3.1.2 The drain valve from the through port shall be closed when the though port is in use.

4.9.3.1.3 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.

Enclosure 1

4.9.3.1.4 When the through port is not physically manned and monitored, the ends of the through port shall be closed.

Enclosure 1

Bases:

Specification 4.9.3.1.1: The LOCA analysis shows that as long as the pool level does not drain through an area greater than 1.48 square inches 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 inch diameter is conservative.

Specification 4.9.3.1.2: 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. By keeping the drain valve closed during through port use, that potential leak pathway is blocked and the through ports ends will either be closed or continuously manned IAW 4.9.3.1.3 and 4.9.3.1.4.

Specification 4.9.3.1.3: 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/experimental personnel will become aware of a pool leak though the through port reasonably quickly, with the gate valves in place, the consequence of the leak can be mitigated quickly by closing the valves.

Specification 4.9.3.1.4: This specification ensures that if the through port is not being monitored for the event of a pool leak, the ends are sealed so that the through port is not a LOCA pathway.

4.9.3.2 Experimental Facility Configuration Within The 4.5 Hour Period After Shutdown Applicability:

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

Objective:

The objective of these specifications is to 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:

Enclosure 1

4.9.3.2.1 Prior to opening a beam port verify that the reactor has not operated in the previous 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

4.9.3.2.2 If opening a beam port within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after reactor shutdown is absolutely required, then:

4.9.3.2.2.1. 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.

4.9.3.2.2.2. 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 4.9.3.2.1: 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is 1.48 square inches (equivalent to a 1.37 inch diameter opening). Consequently, maintaining the limit on the size of the opening between confinement and the beam ports to 1.25 inches in diameter for a period of 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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 4.9.3.2.2: In the event that access to a beam port is needed within 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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.

d. Proposed TS 3.9.3.2.2 provides requirements on to open a beam port. However, the NRC staff is not clear whether the actions specified in proposed TS 3.9.3.2.2.1 and TS 3.9.3.2.2.2 must be performed before there is a need to open a beam port, or within a certain timeframe after there is a need to open a beam port.

Enclosure 1

Provide a revised TS 3.9.3.2.2 to provide a completion time for performing the actions specified in proposed TS 3.9.3.2.2.1 and TS 3.9.3.2.2.2, or justify why no change is needed.

Response: See response to previous RAI above.

7. The proposed TSs do not appear to include any limitations related to the fuel burnup.

ANSI/ANS-15.1-2007, Section 5.3, Reactor core and fuel, states [...] limitations on fuel burnup shall be included as appropriate [...]. NUREG-1313, Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-Power Reactors, Section 3.3, Irradiation Behavior, states that the fuel has been tested to [...] burnups of up to 98 percent of the uranium-235 [...].

Provide a proposed TSs limiting fuel burnup, or justify why these TSs are not required.

Response: Previously addressed in response to RAI 4.2, there is no burn-up limit to address.

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

8. Proposed TS 3.2.1.1, states All shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated. However, the number of the control rods is not specified.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, Subsection (1), states that licensees LCOs should specify the number and type of control rods that must be operable.

Provide a revised TSs that specifies the minimum number of shim safety rods that shall be operable, or justify why this is not required.

Response: Revised as follows:

3.2.1 All four shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated.

Enclosure 1

9. Proposed TS 3.2.1.2, states The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02%k/k per second. The accident analyses provided in the RINSC SAR, as supplemented, are based on a maximum ramp reactivity insertion rate of 0.02%k/k per second. Additionally, the RINSC SAR states that only one safety blade can be moved out of the core at a time. However, the NRC staff finds that while proposed TS 3.2.1.2 would limit the reactivity addition rate from withdrawal of an individual shim safety or regulating rod to the rate that in analyzed in the SAR, as supplemented, it would not limit the total reactivity addition rate to 0.02%k/k per second because it does not specify that only one safety blade or regulating rod can be withdrawn from the core at a time.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, Subsection (2), states The acceptable rates [of reactivity insertion] should be based on the SAR, including inadvertent addition of ramp reactivity at the maximum rate for the most conservative power, rod position, and reactor conditions.

Provide a revised TS 3.2.1.2 such that it limits the total reactivity addition rate from withdrawal of control rods (not the reactivity addition rate from withdrawal of an individual safety blade or regulating rod) to the rate analyzed in the SAR, or justify why no change is needed.

Response: Revised as follows (numbering format changed):

3.2.2 The total reactivity insertion rate of any one shim safety blade and the regulating rod simultaneously does not exceed 0.02%k/k per second.

4.2.2 Verify that only one shim safety blade can be withdrawn at a time and measure each shim safety blade and regulating rod reactivity insertion rates:

4.2.2.1 Annually 4.2.2.2 Whenever maintenance is performed which could affect the reactivity insertion rate of the blade 4.2.2.3 When a new core is configured 4.2.2.4 Following control blade changes

10. Proposed TS 3.2.1.3, states The instrumentation shown in Table 3.1 is operable and capable of providing its intended function. Table 3.1, Instrumentation Required for Reactor Operation, provides a list of protection channels and describes the function as a scram or an interlock preventing rod motion. The information provided in the proposed TS 3.2.1.3, Table 3.1, includes a column for functional description of the scrams and interlocks that are listed. However, the NRC staff finds that while the functional descriptions describe whether each item is a scram channel or interlock, the table does not appear to provide information regarding other functions that the scram channels may have, such as displays and alarms.

In addition, TS 3.2.1.3, Table 3.1, does not appear to list other channels that do not have a scram function, but which are required for operation.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, Subsection (4), states that licensees LCOs should specify all required scram channels and Enclosure 1

setpoints, the minimum number of channels, other functions performed by the channel [...].

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, Subsection (8), states Technical specifications for non-power reactors should have redundant and accurate power level monitors that cover the range from subcritical source multiplication to above full power level. Not all monitors are required to include scram capability []. The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, Subsection (8), also states Specifications in this section should cover the entire channel, including readout meters and recorders and the protective functions they perform, such as to prevent an LSSS from being exceeded.

Provide a revised TS 3.2.1.3, Table 3.1, to list the all of the instrumentation channels (with or without scram capability) required for reactor operations, including whether the channels have readout indications, trending capability, and alarms; provide other proposed TSs that cover this information; or, justify why no changes are needed.

Response: Replace table with following:

Table 3.1 Required Safety Channels Table 3.1.1 Required Safety Channel Scrams Line Protection Power Channels Function Set Point

Level Required

1. Over Power All 2 Scram before power 115% of Licensed is greater than Power

2. Rate of Change of All 1 Scram before period 4 seconds Power is less than

3. Detector HV Failure All 1 per Scram on a loss of 50 V below for Lines 1 & 2 operable HV power suggested channel operating voltage

4. Low Pool Level All 1 Scram before pool 23 feet 7 inches level is less than above the top of the fuel meat

5. Manual Scram All 1 Scram when Control Room Scram Button Depressed

6. Control Rod Drive All 1 Scram if loss of 10 seconds Communication communication for greater than

7. Seismic Disturbance All 1 Scram when Seismic Disturbance Detected

8. Bridge Movement All 1 Scram when Bridge Movement Detected

9. Pool Temperature </=100 1 Scram before 127 F kW temperature is greater than

10. Primary Coolant Inlet >100 1 Scram before 122 F Temperature kW temperature is Enclosure 1

greater than

11. Primary Coolant Flow >100 1 Scram before flow 1560 gpm Rate kW rate is less than

12. Coolant Gates Open >100 1 Scram when Inlet or outlet gate kW open

13. No Flow Thermal >100 1 Scram when No Flow Detected Column kW

14. Bridge Low Power >100 1 Scram when Bridge Not Seated Position kW at HP End Table 3.1.2 Required Safety Channel Interlocks 1.1 Servo Control All 1 Regulating rod can Regulating rod not Interlock not be placed in full out automatic servo mode if 1.2 Servo Control All 1 Regulating rod can 30 seconds Interlock not be placed in automatic servo mode if reactor period is less than

2. Shim Safety Blade All 1 No shim safety 3 counts per Withdrawal Interlock blade withdrawal if second start up channel count rate less than 2.2 Shim Safety Blade All 1 No shim safety Not in the Off Withdrawal Interlock blade withdrawal if position Neutron Flux Monitor Test / Select switch is Table 3.1.3 Required Safety Channel Indications

1. Wide Range Linear All 1 Provide indication of Power reactor power 2.1 Log Power All 1 Provide indication of reactor power 2.2 Log Power Start-up All 1 Provide indication of Counts start-up channel counts 2.3 Log Period All 1 Provide indication of rate of change in reactor power

3. Pool Temperature </=100 1 Provide indication of kW bulk pool temperature

4. Primary Coolant Inlet >100 1 Provide indication of Temperature kW primary coolant inlet temperature

5. Primary Coolant >100 1 Provide indication of Outlet Temperature kW primary coolant Enclosure 1

outlet temperature

6. Primary Coolant Flow >100 1 Provide indication of Rate kW primary coolant flow

7. Confinement Building All 1 Provide indication of Pressure Confinement Building Pressure

11. Proposed TS 3.2.1.3, Table 3.1, lists two servo control interlocks: No regulating rod automatic servo if Regulating rod not full out, and No regulating rod automatic servo if period is less than 30 seconds. The SAR states that interlocks prevent initiation of the regulating rod automatic servo if the regulating rod is not full out or the reactor period is less than 30 seconds. However, the NRC staff finds that while the two servo control interlocks in are listed and described in proposed TS 3.2.1.3, Table 3.1, it is not clear from the proposed TS that the interlocks only prevent initiation of the regulating rod automatic servo, and do not prevent continued operation of the regulating rod automatic servo once it has already been initiated The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2 Subsection (5), states Required interlocks that inhibit or prevent control rod withdrawal or reactor startup should be specified by a table []. Interlocks should be specific to the facility and should be based on the SAR.

Provide a revised TS 3.2.1.3, Table 3.1, to clarify the interlock, or justify why no change is needed.

Response: See revised table above, The reg rod interlocks only prevent us from putting the auto servo in service. Once in auto and the reg rod comes off the full out limit, there is no interlock, nor do we require one.

12. The NRC staff finds that the proposed TSs do not appear to provide any LCO for equipment needed to respond to a loss of coolant accident or for detection of the loss of pool water or heat exchanger integrity.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.3, Subsection (4), states that licensees should establish LCOs for systems that would lead to the detection of leakage or loss of coolant.

Provide a proposed LCOs for detection of leakage or loss of coolant, or justify why no changes are needed.

Response: See 3.1.4 in revised table above for low pool level scram. IAW analysis already submitted, this is the only safety related piece of equipment that must operate in the event of a LOCA to meet the conditions required. No operator action and no equipment operation is required. The 4.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after shutdown restriction to open a beam port, addressed in the TSs, also establishes the requirements to Enclosure 1

ensure we stay within the analyzed conditions. No additional LCOs are required to address loss of coolant.

13. Proposed TS 3.7.1.2.4, states, Alarm set points may be adjusted higher with the approval of the Director or one of the Assistant Directors. The NRC staff finds that TS 3.7.1.2.4 appears to allow a change to a radiation system monitoring alarm without a formal review process. Furthermore, the TS would also allow the Director or one of the Assistant Directors to change a value specified in an LCO without NRC review, as provided in 10 CFR 50.59, Changes, tests, and experiments, which require, in part, that changes to a licensee facility or facility procedures may only be made without an NRC-approved license amendment if they do not involve changes to TSs.

Provide a revised TS 3.7.1.2.4 such that it would not allow other TSs to be changed without NRC approval, or justify why no other changes are needed.

Response: See previous answer, 3.7.1.2.4 has been deleted.

14. Proposed TS 4.4.2.2 describes the functional test of the emergency operating mode confinement system that would be required by proposed TS 4.4.2.1, and proposed TS 4.5.2.2 describes the operability test of the emergency operating mode ventilation system that would be required by proposed TS 4.5.2.1.

NUREG-1537, Part 1, Chapter 14, Appendix 14.1, and ANSI/ANS-15.1-2007 provide guidance for SRs. The proposed SR TSs 4.4.2.2 and 4.5.2.2 appear to be inconsistent with the construct for SRs, since they appear to detail the procedure that would be used to meet another SR, rather than specifying the SR and an associated interval.

Provide a revised TSs 4.4.2.2 and 4.5.2.2 (and the related subsection TS) to make them consistent with the construct for SRs, or justify why the proposed TSs should remain as written.

Response: Deleted 4.4.2.2 and 4.5.2.2, also sections 4.4 and 4.5 rewritten.

15. The proposed TSs 5.1, Site and Facility Description, 5.2, Reactor Coolant System Description, and 5.3, Reactor Fuel and Core Description, each include a description section discussing DFs.

The requirements for TSs for DFs are established in 10 CFR 50.36(c)(4), Design features.

Additionally, NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 5, and ANSI/ANS-15.1-2007 provide guidance regarding DF TSs. Specifically, ANSI/ANS 15.1 2007 states that the DFs TSs should include a general description of the site and facility, information about the design of the reactor coolant system, a description of the reactor core and fuel and the licensed area.

a. The NRC staff finds that TSs 5.1, 5.2, and 5.3 provide information in the Description section, which is not consistent with the TS format as provided in the guidance in ANSI/ANS-15.1-2007 Enclosure 1

Provide a revised TSs 5.1, 5.2, and 5.3 to follow the format of ANSI/ANS-15.-2007, or justify why no change is needed.

b. The NRC staff finds that TS 5.1 does not appear to include a description of the area that is considered to be under the reactor facility license (Part 50).

c. Provide a revised TS 5.1 to include a description of the licensed area (in the Specifications section), or justify why no change is needed.

Response: New Section 5 as follows:

5.0 Design Features 5.1 Site and Facility Specifications The Rhode Island Nuclear Science Center (RINSC) open pool reactor is owned and operated by the State of Rhode Islands Atomic Energy Commission (RIAEC). The Rhode Island Legislature created the RIAEC under the General Laws of Rhode Island, which states (in part), to contract for, construct and operate a nuclear reactor within the state for the purpose of research, experimentation, to co-operate with and make available, under proper safeguards, the use of said reactor by the colleges, universities and industries of this state. The reactor facility is located in Narragansett, Rhode Island on the University of Rhode Island (URI) Narragansett Bay Campus (NBC).

As originally installed, the reactor and support systems built by General Electric Company were adequate for operation at one (1) MW thermal (t) under license number R-95 issued on July 21, 1964. At present, an amendment issued on September 10, 1968 permits operation up to a maximum of two (2) MW(t). The U.S. Nuclear Regulatory Commission (NRC) conversion order to switch from high to low enriched uranium fuel was issued on March 17, 1993 following approval of the revised Safety Analysis Report (SAR).

The Rhode Island Nuclear Science Center (RINSC) site is located on a 3 acre section of a 27 acre auxiliary campus of the University of Rhode Island. The 27 acre site was formerly a military reservation prior to becoming the Bay Campus of the university. The parcel of land is located in the town of Narragansett, Rhode Island, on the west shore of the Narragansett Bay, approximately 22 miles south of Providence, and approximately 6 miles north of the entrance of the bay from the Atlantic Ocean.

The facility is one of a number of buildings located on the NBC, and consists of a reactor room and an office wing with one entrance between them used for normal access and egress. The facility serves as the restricted area in which personnel access is controlled.

The reactor facility is composed of five basic systems: the pool and biological shielding; the reactor core, core suspension, control rod drives and drive shafts; the controls and instrumentation systems; the experiment facilities; and the process and cooling systems.

The Confinement Building acts as the confinement space. A differential pressure of -0.5 Enclosure 1

WC is maintained in the confinement during reactor operation and during emergency ventilation mode of operation (See sections 3.5 and 4.5). The confinement building air combines with dilution air taken from the environment outside of the confinement building and exhausts to a stack which discharges above the building. In the event of an accident which could involve the release of radioactive material, the confinement building air shall be exhausted through a filtration train prior to combining with the dilution air. The filter train will consist of a roughing filter, a charcoal filter for removing radioiodine and an absolute filter capable of removing charcoal dust which may be contaminated with radioiodine. Each absolute filter cartridge shall be individually tested and certified by the manufacturer to have an efficiency of not less than 99.97% when tested with 0.3 micron diameter dioctylphthalate smoke. The minimum removal efficiency of the charcoal filters shall be 99% (see section 3.5 and4.5), based on ORNL data and measurements performed locally.

5.2 Reactor Coolant System Specifications The reactor pool is made of concrete and has an aluminum liner. The primary coolant is light water that is provided by the local town water supply. One end of the pool is designated as the high power end of the pool because the primary inlet and outlet pipes extend into the pool at that end, allowing for forced convection cooling. The thickness of the pool wall is greater at the high power end than at the low power end of the pool. The central section of the pool is separated from the high power section by two pool wall extensions that protrude approximately two feet into the pool, opposite each other. This allows a pool dam to be put into place so that the high power section can be isolated from the rest of the pool, and drained without draining the rest of the pool. Likewise, there is a pair of pool wall extensions that separate the center section of the pool and the low power end of the pool, which allows the low power end to be drained without draining the rest of the pool.

The core is suspended in the pool from a moveable bridge, that allows the core to be positioned anywhere along the length of the pool, while being centered along the width.

The core may be operated up to 100 kW at any position in the pool, however, for operations above 100 kW, the core must be fully seated at the high power end so that the forced convection pipes and ducts are coupled. For 2 MW reactor operation, only one of the cooling loops is required to provide sufficient cooling.

The primary inlet and outlet pipes extend into the pool at the high power end, approximately twelve feet below the pool surface, and couple to the inlet and outlet ducts, which are attached to the core suspension frame. Forced convection cooling is achieved by bringing cooled water from the primary inlet pipe down the inlet duct which opens over the top of the core. Suction causes the water to go through the core into a plenum beneath the core, up the outlet duct, and into the primary outlet pipe.

The forced convection cooling system outlet pipe goes from the reactor pool to the delay tank, where cooling waters progress through the cooling system is delayed in order to reduce the Nitrogen - 16 concentration in the water prior to entering the heat exchanger room. From the delay tank, the forced cooling system is divided into two loops.

Enclosure 1

Each cooling loop consists of a primary and secondary system. Each primary system takes the heated water from the delay tank, through a primary pump, through the primary side of a heat exchanger, and back to the forced convection cooling system inlet piping, where the two systems merge before returning to the pool. The piping for the primary cooling system is aluminum. Nominal temperatures and pressures are less than 130 0 F and less than 100 psig respectively.

The secondary sides of the primary heat exchangers use city water to remove the heat from the primary sides. For each loop, secondary water from the heat exchanger is circulated to a cooling tower, through the secondary pump, and back to the heat exchanger. The piping for the secondary cooling system is polyvinyl chloride. Both of the cooling towers use air cooling to reduce the temperature of the secondary water.

5.3 Reactor Fuel and Core Specifications The RINSC fuel is MTR plate type fuel that has a nominal enrichment of 19.75% Uranium

- 235. The chemical composition of the fuel is U 3Si2 with an aluminum cladding. Each fuel assembly consists of 22 fuel plates, bound by side plates that hold the fuel plates evenly spaced apart. At each end of the assembly, the side plates are attached to square end boxes that are capable of being inserted into a core grid box. The cladding, side plates, and end boxes are aluminum. Each fresh fuel assembly is loaded with 275g Uranium - 235 nominal.

All core designs shall insure that the temperature coefficient is negative. This requirement ensures that a temperature rise due to a reactor transient will not cause a further increase in reactivity. A negative temperature coefficient makes power increases self-limiting by inserting negative reactivity as fuel and coolant temperatures rise. Neutron cross sections in seven energy groups as functions of moderator temperature, fuel temperature, and coolant void fraction were prepared using the WIM/ANL cross section generation code1. Keff values were computed using the DIF3D diffusion theory code.

Coefficients of reactivity were determined from these data. The coolant temperature coefficient was determined to be negative for temperatures between 20 0C (680 F) to 1000 C (2120 F). The fuel temperature coefficient was determined to be negative relative to 200 C for temperatures between 200 C and 6000 C (11120 F).

The core grid box is consists of a 5 15/16 inch thick grid plate that has a 9 X 7 array of square holes, and a box that has four walls that surround the grid plate in such a way that the plate serves as the bottom of the box with the top end open. The grid box is suspended from the top of the pool by four corner posts that occupy the corner grid spaces. The box is oriented so that the open end faces up toward the top of the pool.

The reactor core is configured by inserting fuel element end boxes into grid spaces, so that each fuel assembly is standing up inside the box.

The standard core consists of 14 assemblies in a 3 X 5 array in the center of the grid box, with the central grid space available as an experimental facility. The remaining grid spaces are either filled with graphite or beryllium reflector assemblies, or incore experimental facilities. A non-standard core configuration with 17 fuel elements is also possible. In this configuration, the standard core configuration has been modified so that Enclosure 1

the three central reflector assemblies on the thermal column edge of the core are substituted with fuel assemblies. This core configuration is more conservative than the 14 element core because the core power is spread over three additional assemblies.

Both core configurations include 4 shim safety blades, and a regulating rod. The shim safety blades are located between the fuel and the reflector assemblies on both of the edges of the fuel array that consist of 5 assemblies. There are two blades on each side of the fuel. The blade material is a boron carbide and aluminum sandwiched between two aluminum plates. The poison section is approximately 40.5 inches long, 25 inches of which provides active control of the core. The blades are housed in shrouds that are part of the core grid box. The shrouds ensure that the blades have unfettered movement in and out of the core. The shim safety blades are coupled to the drives using an electromagnet which is capable of releasing the blade upon initiation of a scram signal.

The regulating rod is positioned one grid space out from the fuel, along the central axis of the fuel on the thermal column side of the core. It is made of stainless steel and is approximately 25 inches long by 2.5 inches square. A servo-controlled drive regulates the position of the rod to control power and compensate for small changes in reactivity. It is hard coupled to the drive shaft and does not have scram capability.

5.4 Fissionable Material Storage Specifications Irradiated fuel is stored in two types of fuel storage racks in the reactor pool:

- Fixed racks that are mounted on the pool wall

- Moveable racks that rest on the pool floor Each fixed rack has 9 spaces for fuel storage arranged in a linear array. Each moveable rack has 18 spaces for fuel storage arranged in a 9 X 2 array. Both racks are made of aluminum with stainless steel hardware and are designed with a cadmium/aluminum sandwich and adequate spacing to ensure that when fully loaded, the k eff for the array remains < 0.9. This irradiated fuel is kept cool by natural convection only.

Non-irradiated fuel is typically stored in the RINSC fuel safe fuel safe as described in the RINSC Security Plan. This fuel is also stored in a cadmium lined storage rack.

Non-fuel fissionable materials are either kept where they are in use, or are stored in the reactor pool or fuel safe depending on size constraints and what is most reasonable from an ALARA standpoint.

16. The proposed TS 5.4, Fissionable Material Storage Description, provides requirements for fissionable material storage at the RINSC facility. The Description portion of proposed TS 5.4 discusses the storage of irradiated fuel. However, the NRC staff finds that while the Description portion of proposed TS 5.4 discusses irradiated fuel storage, including the arrays in which such fuel is stored, this information does not appear to be included in the Specifications portion of proposed TS 5.4. In addition, cooling of irradiated fuel that is in Enclosure 1

storage (i.e., how the stored irradiated fuel is cooled) does not appear to be addressed in either the Description or Specifications portions of proposed TS 5.4 The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 5, states that TSs for DFs should include descriptions of fuel storage facilities.

Provide a revised TS 5.4 to include the requirements for the storage and cooling of irradiated fuel (in the Specifications section), or justify why no change is needed.

Response: See previous RAI response

17. The specification in proposed TS 4.3.1.3, states The primary coolant level shall be verified to be greater than or equal to the Limiting Safety System Setting value prior to the initial start-up each day that the reactor is started up from the shutdown condition. The basis for TS 4.3.1.3 includes the statement that a float switch monitors the pool level 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day and 7 days per week and provides a monitored alarm. The basis for TS 4.3.1.3 also discusses the pool autofill feature and the associated equipment. In addition, your previous response to RAI-5.1 (see ADAMS Accession No. ML16202A008) stated that this equipment is used to maintain pool water minimum levels during operation or while in reactor shutdown condition.

The regulations in 10 CFR 50.36(c)(2) require that LCOs be established for the lowest functional capability or performance levels of equipment required for safe operation of the facility. As such, the NRC staff finds that maintaining the water level provides shielding that would otherwise potentially expose occupational workers or the public. As such, the NRC staff considers this equipment should be controlled in the TS.

Provide a revised TS 4.3.1.3, or justify why no change is needed.

Response: This equipment is not required for the safe operation of the facility. As previously addressed, with a postulated beam port shear and no operator action or emergency equipment activation, there is no damage to the core, the increased radiation levels would be skyward having no impact on the public.

18. The proposed TS 2.2.2, Limiting Safety System Settings for Forced Convection Mode of Operation, includes the specifications for thermal power, coolant height, and bulk pool temperature.

The requirements for LSSS TSs are established in 10 CFR 50.36(c)(1)(ii). Additionally, NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 2.2, and ANSI/ANS-15.1-2007 provide guidance regarding LSSS TSs. Specifically, NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 2.2, states that LSSS TSs should be set to ensure that SLs will not be exceeded, and the LSSS TSs should be based on analysis in the SAR. The 2010 steady state thermal-hydraulic analysis for forced-convective flow in the RINSC reactor (performed by Argonne National Laboratory, ADAMS Accession No. ML16062A376) demonstrates acceptable thermal-hydraulic results given the following assumed steady-state conditions: 125°Fahrenheit primary coolant inlet temperature; 2.4 megawatts thermal (MWt) power; 1580 gallons per minute (gpm) coolant flow rate; and, 23 ft 6.5 in of water above the Enclosure 1

core. The analysis also states that thermal power level has a measurement uncertainty of

+/- 0.2 MWt, and flow rate has a measurement uncertainty of +/- 60 gpm. Proposed TS 2.2.2.1, states The limiting safety system setting for reactor thermal power shall be 2.3 MW. Proposed TS 2.2.2.4, states The limiting safety system setting for the primary coolant flow rate shall be 1560 gpm. Given the assumed conditions (thermal power and flow rate) for the steady-state thermal hydraulic analysis, and given the measurement uncertainty in those parameters, the LSSSs in proposed TSs 2.2.2.1 and 2.2.2.4 do not appear to be sufficiently conservative.

Provide a revised TSs 2.2.2.1 and 2.2.2.4 such that the LSSSs are sufficiently conservative, or justify why no such changes are not required.

Response: The bases has been modified to reflect that the sole SL is to keep cladding temperatures below 530 degrees C and the values used in the analysis have been changed. This data is taken from analysis dated 2/26/16 and has already been provided. The bases for 2.2.2 now reads as follows:

This combination of specifications was set to prevent the cladding temperature from approaching the 530 C value at which damage to the fuel cladding could occur, under both steady state, and 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 2.5 MW,

- The coolant height above the fuel meat is at least 23 feet 6.5 inches,

- The primary coolant inlet temperature is no greater than 125 F, and

- The coolant flow rate through the core is at least 1500 gpm.

19. The proposed TS 3.7.1.1, Required Radiation Monitoring Systems, describes the radiation monitors that must be operable during reactor operation, during fuel or experiment handling, or when work is being performed on the core. Proposed TS 3.7.1.2, Radiation Monitoring System Alarm Set Points, provides the alarm set points for the required radiation monitors.

a. Proposed TS 3.7.1.1.1, states 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 whenever: []. Proposed TS 3.7.1.1.2, states If the detector described in specification 3.7.1.1.1 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 NRC staff is not clear if the one radiation monitor and the alternative [] monitor described in proposed TSs 3.7.1.1.1 and 3.7.1.1.2, respectively, would detect both gaseous and particulate radioactive effluents.

The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.7.1, Subsection (1), states that radiation monitors should be specified for both Enclosure 1

radioactive gas and those radioactive particulates that might be airborne in the reactor room.

Provide a revised TSs 3.7.1.1.1 and 3.7.1.1.2 to clarify that the monitors described in the proposed TSs would detect both gaseous and particulate radioactive effluents, or justify why no change is required.

b. The guidance provided in NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.7.1, Subsection (1), states that TSs should list the required radiation monitors, the function each performs [], the approximate location of each, the type of radiation detected, and the alarm or automatic action setting, as analyzed in the SAR. Proposed TS 3.7.1.1.1, states 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 whenever: []. Proposed TS 3.7.1.1.3, states A minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool whenever: [].

i. Proposed TS 3.7.1.1.1 does not appear to indicate all functions performed by the effluent monitor (i.e., whether it alarms locally, in the control room, or both, and whether it gives a readout locally and/or in the control room), the approximate location of the effluent monitor, or the type(s) of radiation detected by the effluent monitor. Your previous response to RAI-7.4 (ADAMS Accession No. ML16202A008) provided some of this information, but proposed TS 3.7.1.1.1 does not include the information.

Provide a revised TS 3.7.1.1.1 to include this information, or justify why no such revisions are required.

ii. Proposed TS 3.7.1.1.3 does not appear to indicate all functions performed by the pool top monitor (i.e., whether it alarms locally, in the control room, or both, and whether it gives a readout locally and/or in the control room). Your previous response to RAI-7.4 (see ADAMS Accession No. ML16202A008) provided this information, but proposed TS 3.7.1.1.3 does not include the information.

Provide a revised TS 3.7.1.1.3 to include this information, or justify why no such revisions are required.

iii. Proposed TS 3.7.1.2.3, states The area radiation monitors shall alarm when radiation levels are 2 times normal levels, or greater. It is not clear which of the monitors required by proposed TS 3.7.1.1 that proposed TS 3.7.1.2.3 is referring to.

Clarify which of the monitor(s) in proposed TS 3.7.1.1 the proposed TS 3.7.1.2.3 is referring to, and either provide a revised TS 3.7.1.2.3 to include this clarification, or justify why the proposed TS 3.7.1.2.3 should remain as written.

Response: Added table 3.2 which addresses these concerns.

Enclosure 1

Table 3.2 Required Radiation Monitors 3.2.1 Required Radiation Monitors Line Description Set Minimum Function Operating Mode

point Required 1.1 Confinement Building 2.5 1 Indication and alarm As per 3.7.1.1.1 Exhaust Stack Gaseous times both locally and in normal control room 1.2 Confinement Building 2 times 1 Indication and alarm As per 3.7.1.1.1 Exhaust Stack normal both locally and in Particulate control room

2. Reactor Bridge Area 2 times 1 Indication and alarm As per 3.7.1.1.3 Monitor normal both locally and in control room

3. Main Floor of 2 times 1 Indication and alarm As per 3.7.1.1.3 Confinement Building normal both locally and in (At least one of 3.2.2, control room lines 3, 6 or 7) 3.2.2 Other Available Radiation Monitors (NO MINIMUM REQUIRED)

Line Description Set Detector Function Operating Mode

point Type

1. Main Floor Particulate 2 times Alpha Indication and alarm N/A, Can be used Monitor normal Beta both locally and in as temporary Gamma control room alternate for stack particulate monitor

2. Fuel Safe Area Monitor 2 times Gamma Indication and alarm N/A normal Neutron both locally and in control room

3. Thermal Column Area 2 times Gamma Indication and alarm N/A Monitor normal Neutron both locally and in control room

4. Heat Exchanger Area 2 times Gamma Indication and alarm N/A Monitor normal Neutron both locally and in control room

5. Primary Clean-Up 2 times Gamma Indication and alarm N/A Demineralizer Area normal Neutron both locally and in Monitor control room

6. Beam Port Area 2 times Gamma Indication and alarm N/A Monitors (4 total) normal Neutron both locally and in control room

7. Dry Irradiation Facility 2 times Gamma Indication and alarm N/A Area Monitor normal Neutron both locally and in control room

8. Rabbit room Area 2 times Gamma Indication and alarm N/A Monitor normal Neutron both locally and in control room

9. Rabbit Room Noble 2 times Noble Gas Indication and alarm N/A Gas Monitor normal both locally and in control room

10. Pool Level Noble Gas 2 times Noble Gas Indication and alarm N/A Monitor normal both locally and in Enclosure 1

control room

20. The specification in proposed TS 4.4.1, states The conditions required to achieve normal operating mode confinement that are specified in section 3.4.3.1 shall be verified to be met prior to the each day of reactor start-up.

The NRC staff finds that TS 4.4.1 appears to contain an extraneous the.

Provide a revised TS 4.4.1 to remove the extraneous the, or justify why no change is needed.

Fixed.

Enclosure 1

ML16280A422

Text

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