Text

NIST Center for Neutron Research - Statement of No Significant Hazard Resulting from License Amendment Request Dated 03/02/2017
	ML17068A164
Person / Time
Site:	National Bureau of Standards Reactor
Issue date:	03/02/2017
From:	Newton T; US Dept of Commerce, National Institute of Standards & Technology (NIST)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML17068A164 (82)
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UNITED STATES DEPARTMENT OF COMMERCE National Institute of Standards and Technology Gaithersburg, Maryland 20888-March 2, 2017 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Statement of no significant hazards resulting from license amendment request dated March 2, 2017

Reference:

NBSR Facility License TR-5, Docket 50-184, NRC letter dated March 2, 2017 Sirs/Madams:

In a license amendment request dated March 2, 2017, the NIST Center for Neutron Research (NCNR) requested an amendment to the facility license Technical Specifications (TS). As required by 10 CFR 50.91(a), the following analysis is presented to show the proposed amendment does not create a significant hazard using the criteria of 10 CFR 50.92(c).

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No, the proposed amendment would not increase the probability or consequences of an accident previously evaluated. The proposed amendment removes conformance conflicts within the Technical Specifications that would occur when operating the reactor as permitted under TS 2.2(4). The conflicts are removed from the TS by adding exception statements. When the reactor is operated under the NRC approved conditions in TS 2.2(4), steady state thermal hydraulic analysis shows that operation at less than 500 kW with natural circulation results in a CHFR and OFI ratio greater than 2. Transient analysis of reactivity insertion accidents shows that the fuel cladding temperature remains far below the safety limit. The limit of 10 kw was chosen since that was deemed adequate for any operational situation requiring natural circulation operation, such as testing of an unknown core loading.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No, the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment removes conformance conflicts within the Technical Specifications that would occur when operating the reactor as permitted under TS 2.2(4). The conflicts are removed from the TS by adding exception statements. The accident analysis was discussed in the document, NIST Response to NRC Request for Information (TAC No. MD3410), August 19, 2008, ADAMS Accession Number ML082890338. The request from the NRC was: " ... Provide justification for 500 kW power operations under natural convection flow by demonstrating that no credible accidents would f1 olO

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result in exceeding the safety limit ... ," the following was the response by NIST. "This analysis shows that there is ample margin between the maximum clad temperature in any credible accident and the safety limit of 450 °C." The details of the analysis are presented in the above reference.

The intent with this amendment is to allow, without apparent TS nonconformance, operation analyzed and evaluated by the NRC. This will allow the use of testing similar to the that which was performed in the commissioning of NBSR.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

No, the proposed amendment would not involve a significant reduction in a margin of safety.

This amendment will allow testing when commissioning a core configuration that is unknown in the most conservative manner appropriate. It removes apparent TS conflicts that would force the licensee into situations that would be less conservative and with less margin of safety.

In addition to the above analysis, included as attachments are the following documents:

1. All pages from the existing (Amend. 10) TR-5 Technical Specifications;

2. Proposed pages 12, 14, 16-19, 24, 25, 30, 39, and 55 with changes highlighted by vertical bars in the right hand margin.

3. Proposed license amendment 11 for possession and use of calibration sources with changes highlighted by vertical bars in the right hand margin.

The NCNR appreciates the time required to process these administrative license changes and looks forward to an approved license amendment in the near future. Please contact me directly at 301-975-6210 or by email at thomas.newton@nist.gov if you have any questions.

Thomas H. Newton, Deputy Director NIST Center for Neutron Research 100 Bureau Drive, MS 6100 Bldg. 235, Room K107 Gaithersburg, MD 20899 2

cc: U.S. Nuclear Regulatory Commission ATTN: Xiaosong Yin One White Flint North 11555 Rockville Pike, MS 012D20 Rockville, MD 20852 Susan T. Gray, Program Manager Power Plant Assessment Division Maryland Department of Natural Resources 580 Taylor Avenue (B wing, 3rd floor)

Tawes State Office Building, B3 Annapolis, MD 21401 3

, .. .I At the values established above, the Limiting Safety System Settings provide a significant margin from the Safety Limit. Even in the extremely unlikely event that reactor power, coolant flow, and outlet temperature simultaneously reach their Limiting Safety System Settings, the critical heat flux ratio (CHFR) is at least 2. For all other conditions the CHFR is considerably higher. This will ensure that any reactor transient caused by equipment malfunction or operator error will be terminated well before the safety limit is reached. Overall uncertainties in process instrumentation have been incorporated in the Limiting Safety System Settings.

Steady state thermal hydraulic analysis shows that operation at less than 500 kW with natural circulation results in a CHFR and OFI ratio greater than 2. Transient analysis of reactivity insertion accidents shows that the fuel cladding temperature remains far below the safety limit. The limit of 10 kw was chosen since that was deemed adequate for any operational situation requiring natural circulation operation, such as testing of an unknown core loading.

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(1) An excess reactivity limit provides adequate excess reactivity to override the xenon buildup and to overcome the temperature change in going from zero power to 20 MW, without affecting the required shutdown margin, In addition, the maximum reactivity insertion accident at startup, which assumes the insertion of 0.5% ~p into a critical core, is not affected by the total core excess reactivity.

(2) These specifications ensure that the reactor can be put into a shutdown condition from any operating condition and remain shutdown even if the maximum worth shim arm should stick in the fully withdrawn position with the regulating rod also fully withdrawn.

3 .1.3 Core Configuration Applicability: Core grid positions Objective: To ensure that effective fuel cooling is maintained during forced flow reactor operation.

Specification The reactor shall not be operated with forced coolant flow unless all grid positions are filled with full length fuel elements or thimbles.

Core grid positions shall be filled to prevent coolant flow from bypassing the fuel elements for operation of the reactor with forced coolant flow.

3.1.4 Fuel Bumup Applicability: Fuel Objective: To remain within allowable limits of bumup 14

beginning-of-life shim arm worths with the shim arms operating at the design speed of their constant speed mechanisms. The analysis shows that the most severe accident, a startup from source level, will not result in core damage.

3.2.2 Reactor Safety System Channels Applicability: Required instrument channels Objective: To provide protective action for nuclear and process variables to ensure the LSSS values are not exceeded.

Specifications The reactor shall not be operated unless the channels described in Table 3.2.2

. are operable and the information is displayed in the reactor Control Room.

Table 3.2.2 Reactor Safety System Channels Minimum Nuclear and Process Channels Required Channel Scram Major Scram Rundown (l)High Flux level 2 (2) Short period below 5% rated power 2 (3) Low reactor vessel D2 0 level I, 3 2 (4) Low flow reactor outlet 2 ' 3 1 2 3 (5) Low flow reactor inner or outer plenum ' 1 (6) Manual (outside of the Control Room) 1 (7) Manual 1 (8) Reactor Outlet Temperature 3 4

(9) Gaseous Effluent Monitors 2 1

One (1) of two (2) channels may be bypassed for tests or during the time maintenance involving the replacement of components and modules or calibrations and repairs are actually being performed.

f 2

One (1) of these two (2) flow channels may be bypassed during tests, or during the time maintenance involving the replacement of components and modules or calibrations and

minor repairs are actually being performed. However, outlet low flow may not be bypassed unless both inner and outer low-flow reactor inlet safety systems are operating.

3 May be bypassed during periods of reactor operation when a reduction in Limiting Safety System Settings are permitted by the specifications of Sections 2.2(4) and 3.3.1(1).

4 See specifications of Section 3.7.1 16

The nuclear and process channels of Table 3.2.2 initiate protective action to ensure that the safety limit is not exceeded. With these channels operable, the safety system has redundancy.

The exceptions are required to perform surveillance, maintenance or operation permitted by the specification of Sections 2.2(4), 3.3.1(1), and 3.7.1.

3.3 Coolant System 3.3.l Primary and Secondary Applicability: Primary fluid systems Objective: To prevent degradation of primary systems' materials.

Specifications The reactor shall not be operated unless:

(1) The reactor vessel coolant level is no more than 25 inches below the overflow standpipe.

Exception: To perform periodic surveillance of the effectiveness of the moderator dump or approach to critical testing for an unknown core loading, it is necessary to operate the reactor as permitted in the specifications of Section 2.2(4) and without restriction on reactor vessel level above the dump tube.

(2) The D 2 concentration in the Helium Sweep System shall not exceed 4% by volume.

(3) All materials, including those of the reactor vessel, in contact with the primary coolant shall be compatible with the 020 environment.

(1) The limiting value for reactor vessel coolant level is somewhat arbitrary because the core is in no dang.er so long as it is covered with water. However, a drop of vessel level indicates a malfunction of the reactor cooling system and possible approach to uncovering the core. Thus, a measurable value well above the minimum level is chosen in order to provide a generous margin of approximately 7 feet (2.13 m) above the fuel elements. To permit periodic testing, such as surveillance of the effectiveness of the moderator dump or approach to critical for an unknown core loading, it is necessary to operate the reactor without restriction on reactor vessel level. This is permissible under conditions when forced reactor cooling flow is not required, such as is permitted in the specifications of Section 2.2(4).

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(2) Deuterium gas will collect in the helium cover gas system because of radiolytic disassociation of D20. Damage to the primary system could occur if this gas were to reach an explosive concentration (about 7.8% by volume at 77°F (25°C) in helium if mixed with air). To ensure a substantial margin below the lowest potentially explosive value, a 4% limit is imposed.

(3) Materials of construction, being primarily low activation alloys and stainless steel, are chemically compatible with the primary coolant. The stainless steel pumps are heavy walled members and are in areas of low stress, so they should not be susceptible to chemical attack or stress corrosion fai_lures. A failure of the gaskets or valve bellows would not result in catastrophic failure of the primary system. Other materials should be compatible so as not to cause a loss of material and system integrity.

3.3.2 Emergency Core Cooling Applicability: Emergency Core Cooling System Objective:

To ensure an emergency supply of coolant.

Specifications The reactor shall not be operated, except under Section 2.2(4), unless:

(1) The D20 emergency core cooling system is operable.

(2) A source of makeup water to the D20 emergency cooling tank is available.

Basis

( 1) In the event of a loss of core coolant, the emergency core cooling system provides adequate protection against melting of the reactor core and associated release of fission products.

(2) The emergency core cooling system employs one sump pump to return spilled coolant to the overhead storage tank. Because only one sump pi.imp is used, it must be operational whenever the reactor is operational. There is sufficient D20 available to provide approximately 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of cooling on a once-through basis. In the event that the sump pump fails and the D20 supply in the overhead storage tank is exhausted, domestic water or a suitable alternative would be used to furnish water for once-through cooling. The water makeup capacity must be in excess of 25 gpm, which was found adequate in cooling calculations to prevent fuel damage.

Full operability is not available, nor is it needed, when operating as permitted by the specifications of Sections 2.2(4) and 3.3.1(1). However;the 3000 gallon D20 emergency cooling tank and a source of makeup water would be available.

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3.3.3 Moderator Dump System Applicability: Moderator dump Objective: To provide a backup shutdown mechanism.

Specification The reactor shall not be operated except under Section 2.2(4) unless the reactor moderator dump system is operable.

In the unlikely event that the shim arms cannot be inserted, an alternate means of shutting down the reactor is provided by the moderator dump. The moderator dump provides a shutdown capability for any core configuration. Hence, it is considered necessary for safe operation. It has been shown that the moderator dump provides sufficient negative reactivity to make the normal startup (SU) core subcritical even with all four shim arms fully withdrawn.

3.4 Confinement System 3.4.1 Operations that Require Confinement Applicability: Reactivity changes within the vessel and fuel movements outside of the vessel '

Objective: To provide an additional barrier to fission product releases.

Specifications Confinement shall be maintained when:

(1) The reactor is operating.

(2) Changes of components or equipment within the confines of the thermal shield, other than rod drop tests or movement of experiments, are being made which could cause a significant change in reactivity.

(3) There is movement of irradiated fuel outside a sealed container or system.

( 4) The reactor has been shl;ltdown for shorter than the time specified in the specification of Section 3.9.2.2.

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(2) One fission product monitor is operable or sample analysis for fission product activity is conducted daily. 1 (3) One secondary coolant activity monitor is operable or a D 2 0 storage tank level monitor is operable. 1 (4) Two area radiation monitors are operable on floors C-100.and C-200.

(5) The primary tritium concentration is less than or equal to 5 Ci/I.

(6) Removed to 3.7.2.

When required monitors are inoperable, then portable instruments, survey or analysis may be substituted for any of the normally installed monitors in specifications (1)- (4) for periods of one (1) week or for the duration of a reactor run.

1'. Operability of the monitors specification (2) and (3) above are not required for operation permitted by the specifications in Section 2.2(4) since these systems are neither operable nor needed without forced primary and secondary coolant flow.

Basis (1) The requirements of 10 CFR 20.1502(b) (2007) are met by regular monitoring for airborne radionuclides and bioassay of exposed personnel.

The two.primary airborne radionuclides present at the NBSR are 41 Ar and 3

H. The normal air exhaust system draws air from areas supplied by conditioned air, such as the first and second floors of the confinement building. The irradiated air exhaust system draws air from areas most likely to have contaminated air, such as waste sumps and penetrations in the biological shield. Normal and irradiated air are monitored continuously with G-M detectors sensitive to~ and y emissions and the combined air is exhausted through the stack. The stack release is monitored with a G-M detector.

(2) A fission products monitor located in the helium sweep gas will give an indication of a "pin-hole" breach in the cladding so that early preventive measures can be taken, When this monitor is not functional, daily testing will ensure that the fuel cladding is intact. These two measures ensure that there are no undetected releases of fission products to the primary coolant.

Specification (1) alone is adequate to assure detection of abnormal effluent radioactivity during operation as permitted by Section 2.2(4).

(3) Monitoring for primary water leakage into the secondary coolant is done by a secondary water monitor that is sensitive to radionuclides in the primary water. Leakage of primary to secondary would also be detected by a change in the D2 0 storage tank level.

(4) Fixed gamma area radiation monitors are positioned at selected locations in the confinement building. Typical alarm setting are less than 5 rnrem/hr 24

and adjusted as needed for non-routine activities, generally with the objective of identifying unusual changes in radiation conditions.

(5) At the end of the term of the NBSR license the maximum tritium concentration in the primary coolant is estimated to be 5 Ci/I. This value and reliable leak detection ensures that tritium concentrations in effluents shall be as low as is practicable.

(6) Removed to 3.7.2.

3.7.2 Effluents Applicability: Annual releases Objective: To minimize exposures to the public.

Specification The reactor shall not be operated unless:

( 1) The total exposure from effluents from the reactor facility to a person at

the site boundary shall not exceed 100 mrem per calendar year, less any external dose from the facility. The limit shall be established at the point of release or measurement using accepted diffusion factors to the boundary. For halogens and particulates with half-lives longer than 8 days, a reconcentration factor shall be included where appropriate.

(2) An environmental monitoring program shall be carried out and shall include as a minimum the analysis of samples from surface waters from the surrounding areas, vegetation or soil and air sampling.

Basis The criteria for determination of concentration limits specified above ensure that 10 CFR 20 (2007) limits are not exceeded at the site boundary. The allowance for dilution from the reactor building stack to the nearest site boundary is 1,000. This value of 1,000 from the diffusion view point is the minimum expected at the nearest site boundary under the least favorable meteorological conditions. This number could be increased by one or two orders of magnitude if normal variations in wind speed and direction were considered. Because these variations are not considered, a one or two order of magnitude margin is inherent in this limit.

In specifying the limits on particulates and long lived (longer than 8 days) halogens, consideration was given to the possibility of biological reconcentration in food crops or dairy products. Using available information (Soldat, J .D., Health Physics 9, p. 1170, 1963 ), a conservative (both the COMPLY and CAP88 codes indicate that 700 is at least an order of 25

3.9.2 Fuel Handling 3.9.2.l Within the Reactor Vessel Applicability: Fuel element latching Objective: To ensure that all fuel elements are latched between the reactor grid plates.

Specifications Following handling of fuel within the reactor vessel, the reactor shall not be operated, except under Section 2.2(4), until all fuel elements that have been handled are inspected to determine that they are locked in their proper positions in the core grid structure. This shall be accomplished by one of the following methods:

(1) Elevation check of the fuel element with main pump flow.

(2) Rotational check of the element head in the latching direction only.

(3) Visual inspection of the fuel element head or latching bar.

Basis Each NBSR fuel element employs a latching bar, which shall be rotated to lock the fuel element in the upper grid plate. Following fuel handling, it is necessary to ensure that this bar is properly positioned so that an element cannot be lifted out of the lower grid plate, which could lead to a reduction in flow to the element after pump flow is initiated. Any of the three methods above may be used to verify bar position. Tests have shown that flow from a primary pump will raise an unlatched element above its normal position and thus will be detected by the pickup tool under flow conditions. The efficacy of rotational checks has been confirmed by visual inspections.

Operation of the reactor in natural circulation at ~10 kW is permitted prior to verifying that the elements are latched. Until main pump flow is used the flow forces are not present to cause the lifting of the elements if they are not latched.

3.9.2.2 All Other Conditions Applicability: Refueling system Objective: To ensure the integrity of the fuel element cladding.

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(4) The voltage and specific gravity of each cell of the Vented Lead Acid (VLA) battery shall be tested annually. A discharge test of the entire battery shall be performed once every 5 years.

(5) A discharge test of the Valve-Regulated Lead Acid (VRLA) batteries shall be performed once every two years.

(1) The NBSR is equipped with two diesel power generators, each capable of supplying full emergency load; therefore, only one of the generators shall be required. The diesel generators have proven to be very reliable over decades of service. The quarterly test frequencies are consistent with industry practice and are considered adequate to ensure continued reliable emergency power for emergency equipment.

(2) This testing frequency of the operable generator will ensure that at least one of the required emergency generators will be operable.

(3) An annual test of the emergency power equipment under a simulated complete loss of outside power will ensure_ tpe source will be available when needed.

(4) and (5), Specific gravity and voltage checks of individual cells are the accepted method of ensuring that all cells of a VLA battery are in satisfactory condition. The annual frequency for these detailed checks is considered adequate to detect any significant changes in the ability of the battery to retain

'its charge. During initial installation, the station batteries were discharge tested to measure their capacity. Experience has shown that repeating these tests at the specified intervals is adequate to detect deterioration of the cells and loss of battery capacity.

4.7 Radiation Monitoring System and Effluents 4.7.1 Monitoring System Applicability: Radiation monitoring equipment Objective: To ensure operability ofradiation monitors.

Specifications (1) The gaseous effluent monitors for normal air, irradiated air and stack air shall I

be channel tested before startup, after a shutdown of longer than twenty-four (24) hours, or quarterly. Each of the above air monitors shall be channel calibrated annually.

(2) The fission products monitor shall be channel tested monthly and channel calibrated annually.

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Director, r---- - -- --------------- ~------------------

NIST Center for Neutron Research I

I Chief, Reactor Operations and Engineering I

I I Chief, Chief, Reactor Operations Reactor Engineering I

Reactor Supervisor NBSR Health Physics I

Senior Reactor Operator Reactor Operator Auxiliary Operator r--- --------- -----'

I I

I Safety Assessment Committee Safety Evaluation CommitteeSEC)

(SAC)

Recommendations and Technical Advice Administrative Reporting Channels Figure 6.1 55

Purpose for the Amendment This amendment is required to make changes to the NBSR Technical Specifications in order to formalize the exceptions needed to perform an Approach to Critical with an Unknown Core Loading Configuration. There are some exceptions that are required anytime the reactor is I

operated in natural circulation mode using Limiting Safety System Setting (LSSS), Technical Specification (TS), 2.2(4), Reactor Power, with natural circulation cooling flow, shall not exceed 10 kW. Also, there are additional exceptions that are needed to perform a safe initial start-up and rise-to-power operational phase of the reactor from an unknown core loading configuration using an approach to critical test procedure. Unfortunately, not all of the exceptions needed are explicitly stated in the TS. The amendment formalizes all of the exceptions in the TS. All exceptions shall apply and may be used anytime reactor operation with natural circulation flow under TS 2.2(4) is implemented.

Background Analyses for Operation with Natural Circulation Cooling Flow The reactor safety case for operation with natural circulation cooling flow is in the Safety Analysis Report(SAR), NBSR 14, Section 4.6.5. The following was the stated conclusion:

"The RE LAP code has been used to ~nalyze operation at 500 kW with natural convection allowed, and shown to be !completely safe. The analysis shows that the core coolant flow is stable and subcooled. The peak fuel centerline temperature is about

25 °C below the saturation temperature. The peak heat flux is at least a factor of five below the calculated CHF and the wa.11 heat flux corresponding to the OFI condition, as shown in Table 4.6.2. In addition, reactivity insertion transients (both startup accident and 0.5% reactivity insertion in 0.5 s) :Were analyzed, and shown to result in no fuel damage.

Notwithstanding the above analysis, the maximum reactor power permitted without forced flow is 10 kW, providing a very'.substantial margin to CHF."

The analysis was also discussed in the document, NIST Response to NRC Request/or Information (TAC No. MD3410}, August 19, 2008, ADAMS Accession Number ML082890338.

To the request: " ... Provide justification for 500 kW power operations under natural convection flow by demonstrating that no credible accidents would result in exceeding the safety limit. .. ," the following was the response by NIST. "This analysis shows that there is ample margin between the maximum clad temperature in any credible accident and the safety limit of 450 °C." The details of the analysis are presented in the above reference.

The LSSS was implemented as TS, 2.2(4) and was approved in the NRC NBSR Safety Evaluation Report(SER) (June 2009), page 4-12 with the following statement:

"For natural convection, the licensee used the Sudo-Kaminaga and Oh/Chapman correlations to check for DNB and OFI, respectively. These calculations were performed for both the inner and outer plenums, at both the hot spot and the exit of the hottest fuel

pg. 1 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Ffnal).docx

channel of the upper fuel section. Critical heat flux and OFI ratios calculated by the licensee show ample thermal safety margins (greater than a factor of 2) for steady-state operating conditions. The licensee used the RELAP5 code to analyze abnormal transients during 500-kW power operation with natural convection. These analyses show peak clad temperatures lower than the blistering temperature, thereby demonstrating the acceptability of the limiting safety system settings (LSSSs) required by TS 2.2, "Limiting Safety System Settings," for operation with natural convection. Although the licensee performed calculations for operation at 500 kW, TS 2.2 requires that the reactor power be limited to 10 kW during operation with natural circulation. This requirement provides a large safety margin for operation with natural circulation."

The LSSS, TS 2.2 (4) makes it clear that reactor operation at 10 kW is allowed without primary flow. The limiting safety system setting is ~ 10 kW with the premise that the reactor may be and is operating with natural circulation cooling flow. It should also be clear that Limiting Conditions of Operation (LCOs) associated with systems that are inoperable without forced circulation flow should have exceptions explicitly stated in tHe TS. All of the analyses in support of the required exceptions are in or referenced in the SAR and NRC SER otherwise TS 2.2(4) would not have been approved.

Technical Specifications Exceptions.

An unknown core loading could exist for several reasons.

The known core loading configuration prior to a typical start up is one for which there is enough information so that the estimated critical position (ECP) of the shims may be predicted with reasonable precision with all elements in place (30 elements). This is usually the case for beginning of cycle (BOC) startups with no more than 5 replacement (fresh) fuel elements and known shims. If that is the case, then this section is not required to perform a safe reactor startup. Under some circumstances, however, the information needed to determine the ECP may be missing or unknown, or not known with reasonable certainty. An unknown core loading configuration may exist after one or some combination of these changes have occurred since the last known core loading configuration:

the fuel element reactivity has been modified by fission product poisons and fuel burn up,

the fuel element distribution by position and loading may have been changed,

the shim reactivity worth has been modified by cadmium burn up,

shims have been replaced with new and their worth have not been measured by Technical Specification (TS) 4.1.2 Reactivity Limitations, which affects the:

o core excess, o maximum reactivity addition rate of the shims, o minimum shutdown margin (MSDM) of the reactor, or

new experiment or a new experimental facility has been added which has an undetermined reactivity worth.

pg.2 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final).docx

If in the determination of the Chief of Reactor Operations, the reactor core loading configuration is unknown, this section may be used to allow Technical Specification exceptions similar to those used on the initial startup testing performed on the NBSR in 1967. This will be an approach to critical using a procedure written, reviewed, and approved in accordance with TS 6.4 Procedures. The procedure shall include the actions to remove the exceptions before allowing forced cooling flow and full power (normal) reactor operation.

Reactor Powers 10 kW with Natural Circulation Cooling Flow The following TS are unknown, need to be excepted, or are not applicable when operating with natural circulation cooling flow at a thermal reactor power~ 10 kW Natural Circulation Mode under TS 2.2 (4):

a. Limiting Safety System Settings (LSSS) for full power, forced coolant flow, and reactor outlet temperature, TS 2.2 (1), (2), and (3).

Basis: When operating in Natural Circulation Mode (under LSSS TS 2.2 (4)), these Limiting Safety Systems Settings (LSSS) are not applicable since operation at a reactor thermal power of s 10 kW is sufficient to prevent reaching the safety limit. This was verified in the in the NRC Safety Evaluation Report (SER} (June 2009), Page 4-12 and in the Safety Analysis Report (SAR}, NBSR 14, Section 4.6.5

b.

Limiting Conditions for Operations (LCO) on excess reactivity and minimum shutdown margin, TS 3.1.2.

Basis: If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform an approach to critical for an unknown core loading configuration then the core reactivity surveillance measurements may not have been done for that loading. The approach to critical test is the most conservative process to reach a known core loading. When the core is in the configuration desired for normal operation (full flow and a critical position CP in the desired range) the LCOs are verified by performing the implementing procedures for the surveillance TS 4.1.2 before further operation.

c. LCO on core configuration, TS 3.1.3 Basis: This Limiting Condition of Operation (LCO) is not applicable when operating in the Natural Circulation Mode (under LSSS TS 2.2 (4)). Without flow the issue of coolant flow bypassing the fuel elements does not exist.

The fact is that the driving force of the natural circulation is the heat pg.3 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final).docx

produced by the fuel elements causing most of the flow to go through the elements as desired and little through the open grid positions.

d. LCO on shim arm reactivity worth, TS 3.2.1 (3).

Basis: If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform an approach to critical for an unknown core loading configuration then the core reactivity surveillance measurements have not been done. The approach to critical test is the conservative process to reach a known core loading. When the core is in the configuration desired for normal operation (full flow and a Shim Critical Position (CP) in the desired range) the LCOs are verified by performing the implementing procedures for the surveillance TS 4.2 before further operation.

e. LCO on required Safety System Channel for low reactor vessel D20 level, TS 3.2.2-(3).

Basis: If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform an approach to critical for an unknown core loading configuration involving fuel movement an exception is required for reactor vessel D20 level LCO. The reactor vessel D20 level is lowered to the refueling level which is at the level _of the 6" overflow pipe. The typical level of the reactor vessel for "normal" forced circulation cooling flow is _at the level of the 3" overflow pipe with the vessel level instruments (LRCl and LIA 40) indicating-154". The refueling level is 93 1/4" below the normal operating level (3" overflow pipe) at power which indicates -70" on the vessel level instruments (LRC 1 or LIA 40).

Irradiated elements are not moved within the vessel unless the water level is lowered to the refueling level. It is not consistent with safe operation or ALARA to cycle between refueling level for fuel movement and normal vessel level for operational testing in the natural circulation mode more than necessary.

There is no reactor safety issue with operation in the natural circulation mode with the level at 70". At that level there remains approximately 23" of reflector left to dump if needed to add additional negative reactivity. In addition, the operable shims, with the already low reflector level assures the availability of negative reactivity that is more than capable ofshutting down the reactor and maintaining it in a shutdown condition If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform surveillance testing of the upper reflector (moderator) pg.4 Technical Specification Exceptions TS Change version 3.2 2-15-2017{Final).docx

dump the level would be lowered another 23" to the top of the dump tube, which is 1" above the fueled section of the elements. If there is a need to perform surveillance testing of the upper reflector dump a procedure would be developed and approved using the normal process as per TS 6.4, Procedures. Part of the basis for such a procedure would be the start-up test procedure# 206 used during the commissioning of the NBSR (see also Section 7.2 i below}.

f. LCOs on required Safety System Channels for low flow reactor outlet, or inner or outer plenum, and reactor outlet temperature, TS 3.2.2 (4), (5), and (8).

Basis: If operating in Natural Circulation Mode (under LSSS TS 2.2 (4)), with a~ 10 kW reactor thermal power limitation these Reactor Safety System Channels require an exception is required for these LCOs. The instruments are not operable without primary flow and with the limitation on reactor power at~ 10 kW they are not necessary.

g. LCO on minimum reactor vessel coolant level, TS 3.3.1(1}:

Basis: If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform an approach to critical for an unknown core loading configuration involving fuel movement the minimum reactor vessel coolant level will require and exception. The reactor vessel level may be as low as 70" *(93 1/4" below the 3" overflow pipe}, which is the refueling level. bu ring the periodic testing with the ~ 10 kW reactor thermal power limitation in the natural circulation mode, degradation of the primary system's materials is not a significant issue. The critical concern of degradation is the fueled sections of the elements. This is not an issue since thefueled sections remain immersed in D20 except when the element is being moved to a new position within the core or, into or out of the vessel and only.with compliance of with TS 3.9.2.2. The movement of the elements happens each time the reactor is refueled and no significant degradation has been found when the elements have been placed in the spent fuel pool for storage before shipment.

h. LCO on the operability of emergency core cooling TS 3.3.2(1}.

Basis: If the vessel level is at the fuel transfer level to expose the drop-out chute

(-70"} as required for fuel movement, an exception is required for the emergency core cooling LCO. At the fuel transfer or refueling level there remains at least 24" of D20 over the fueled sections of the elements.

However, the upper (inner} reserve tank is empty thereby preventing pg.S Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final}.docx

immediate flow of the emergency core cooling from that source. Without the upper reserve tank being able to immediately add cooling the emergency core cooling system would be defined as not completely operable. All other components of the emergency cooling systems are operable and fully capable of cooling the core in this configuration. By TS 3.9.2.2, fuel shall not be removed from water unless the reactor has been shut down for a period equal to or longer than one hour for each megawatt of operating power level. After flowing there is no immediate need for emergency core cooling even if the reactor is operated in the natural circulation mode at a reactor thermal power :s: 10 kW. However, if desired, the emergency system may be initiated in a short time by using the appropriate procedure. The appropriate valves are opened to flow D20 to the vessel by gravity from the 3000 gallon D20 emergency tank directly into the upper (inner) reserve tank in the vessel. It has a distribution system to direct the flow of water into the top of each element even it is initially empty. That flow rate is adequate to cool the elements even ifthe reactor had been recently sh!Jtdown from a long run at 20 MW thermal.

i. LCO on operability of the moderator dump system, TS 3.3.3.

Basis: If the vessel level is at the fuel transfer level, to expose the drop-out chute

("'70") as required for fuel movement an exception is required for the moderator dump LCO. This allows fuel movement within the vessel and into and out of the transfer chute in dry conditions. The dry conditions are needed to reduce the drag on the fuel elements that the D20 would create and allow transfer to and from the fuel pool while minimizing the transfer of light water to the vessel or tritium to the pool. At the fuel transfer level, the fuel remains covered by at least 24" of D20. Also at that level there remains approximately 23" of moderator (reflector) left to dump if needed. If the moderator dump is initiated, even with the shim arms full out, the NBSR Safety Analysis Report states that: "the moderator dump will make the reactor subcritical even in its most reactive state."

(Safety Analysis Report, NBSR 14, Section 4.5.1.3.3). No matter from what level the moderator dump is initiated its full effectiveness is realized as soon as the moderator level is at the level of the dump tube. The basis will be changed to reflect state that the moderator dump system is fully operable as long as the dump valve is operable no matter the vessel level from which it is initiated.

pg.6 Technical Specification Exceptions TS Change version 3.2 2-15-2017{Final).docx

j. LCOs on the operability of the fission product and the secondary coolant activity monitors, TS 3.7.1 (2), and (3).

Basis: If operating in Natural Circulation Mode (under LSSS TS 2.2 (4)). In the case of TS 3.7.1 (2), then the fission product monitor channel operability requires an exception. The primary flow is necessary to operate the helium sweep gas system that is required to flow the gas, that covers the D20 in the vessel, though the monitor so that it can detect fission product gases in the sweep gas.

In the case of TS 3.7.1 {3) another exception is required. The purpose of the secondary coolant activity monitor is to detect leakage from the primary to the secondary coolant, when both are flowing, by detecting N-16 gammas in the secondary coolant flow. Without leakage from the primary to the secondary coolant there will be no N-16 in the secondary coolant. The monitor is not operable without flowing the secondary coolant system, which is not operated without operating the primary coolant system. While operating in the natural circulation mode there is no primary flow and little N-16 production occurs at a reactor thermal power of s 10 kW. Thereby, the secondary coolant activity monitor is

. inoperable as well as not needed during natural circulation mode operation.

k. LCO on assurance of fuel element latching prior to full flow and power operation, TS 3.9.2.1.

Basis: If the purpose of operation in Natural Circulation Mode (under LSSS TS 2.2 (4)) is to perform an approach to critical for an unknown core loading configuration involving less than 30 elements in the grid an exception on the fuel element latching LCO is required. Without primary pump flow the forces are not present to lift an unlatched element. As such there is no safety case for prohibiting operation of the reactor as permitted under Section 2.2(4) prior to performing the latch checks for each element that has been handled, i.e. moved during refueling, since the last latch check.

As long as the reactor is not operated with force cooling before all fuel elements that have been handled are inspected, using TS 3.9.2.1, to determine that they are locked (latched) in their proper positions in the core grid structure there is no safety issue.

pg. 7 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final).docx

General Basis: Just as some technical specifications do not apply if the reactor is shutdown, some technical specifications do not apply in natural circulation mode with the reactor operating at a thermal power of s 10 kW. In other words, the initial assumptions in the safety analysis are essential for determining the need for each technical specification. The safety analyses with the initial assumption of "operation at full flow and power" as compared to the initial assumption of "operation in natural circulation mode with a reactor thermal power at ~ 10 kW" is very different. The results of the analyses are that some technical specifications required for the operation under the first assumption are not needed with the latter.

The changes requested for the Technical Specifications are clarifications, by exceptions and annotations to reduce the confusion that may be present without those changes. It has not been necessary to modify the NBSR Safety Analysis Report to justify the changes. The specific safety analysis for natural circulation mode, which is the bases for TS 2.2(4), is discussed in the Safety Analysis Report, NBSR 14, Section 4.6.5 and evaluated in the NRC's NBSR Safety Evaluation Report (June 2009), page 4-12.

Change of the basis for TS 3.1.3 To reflect need for the LCO based on the SAR more accurately. The basis for this LCO is being changed to more accurately summarize the safety case in the SAR. The open grid positions shall be filled to prevent bypassing the coolant flow around the fuel elements during reactor operation with force flow. Containment of broken pieces of shims should not be discussed in the basis of this LCO. A catastrophic failure of the shims is not a credible event, considering:

The quality assurance during the manufacture of the shims,

The fact that the maximum shim replacement interval is -4 years, (456,000 MWHs (950 equivalent full power days)),

The construction materials are aluminum that is irradiated significantly below any embrittlement limit, and

45 years of operational experience with no significant physical integrity issues, i.e. no limiting corrosion, with the shims (Safety Analysis Report, NBSR 14, Section 4.2.2.1).

Also, due to the typically low excess reactivity and the available upper reflector dump there is adequate reactivity to shut down the reactor in the unlikely event that one shim would leave the core.

Change of the basis for TS 3.7.1(1) pg.8 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final).docx

The basis was updated to remove the specific reference to the use of "G-M" radiation detectors to a statement that is more generic. That is, basis was changed to reference use of detectors "as appropriate for the instrumentation design and radiation to be monitored."

pg.9 Technical Specification Exceptions TS Change version 3.2 2-15-2017(Final).docx

Appendix A License No. TR-5 Technical Specifications for the NIST Test Reactor (NBSR)

Table of Contents 1.0 Introduction ......................................................................................................................... 5 1.1 Scope ............................................................................................................................... 5 1.2 Application ...................................................................................................................... 5 1.2.1 Purpose ....................... ~ ............................................................................................ 5 1.2.2 Format .............................................................. ,....................................................... 5 1.3 Definitions ....................................................................................................................... 5 1.3.1 ALARA ...................... :............................................................................................ 5,

,1.3.2 Channel .................................................................................................................... 5 1.3 .2.1 Channel Calibration ............................................................................................. 6 1.3.2.2 Channel Check ..................................................................................................... 6 1.3.2.3 Channel Test ........................................................................................................ 6, 1.3.3 Confinement ............................................................................................................ 6 1.3.4 Core Configuration .................................................................................................. 6 1.3.5 Excess Reactivity ..................................................................................................... 6 1.3.6 Emergency Director ............................................... ,................................................. 6 1.3.7 Experiment. ......................................................... :......... :.......................................... 6 1.3.7.1

In-Reactor Vessel ............................................................................................ :... 6 1.3.7.2 Beam Tubes ......................................................................................................... 6 1.3. 7 .3 Movable Experiment ........................................................................................... 7 1.3. 7 .4

Secured Experiment. ............................................................................................ 7 1.3.8 License ........................................... :........... ,............................................................. 7 1.3.9 Measured Value ....................................................................................................... 7

1.3.10 Moderator Dump ..................................................................................................... 7 1.3.11 Natural Convection Cooling ..... :.............................................................................. 7 1.3.12 Operable ................................................................................................................... 7 1.3.13

Operating .................................................................................................... : ............ 7 1.3 .14 Protective Action ..................................................................................................... 7 1.3 .15* Reactor Operating .................................................................................................... 8 1.3 .16 Reactor Operator ...................................................................................................... 8 1.3.17 Reactor Safety System .......... :.... ~ .................................................. :: ......................... 8 1.3 .18 Reactor Secured .................................. *..................................................................... 8 1.3.19 Reactor Shutdown .................................................................................................... 8 1.3.20 Reactor Shutdown Mechanisms .............................................................................. 8 1.3.21 Reference Core Condition ....................................................................................... 9 1.3.22 Reactor Rundown .................................................................................................... 9 1.3.23 Rod, Control ............................................................................................................ 9 1.3.24 Rod Drop Mode ....................................................................................................... 9 1.3.25 Rod, Regulating ....................................................................................................... 9 1.3.26 Scram ....................................................................................................................... 9 1.3.26.1 Major Scram ......................................................... :.................................... ,..... 9 1.3.27

Scram Time ................................................................................ .-............................. 9 1.3.28 . Senior Reactor Operator .......................................................................................... 9 1.3.29 Shall, Should and May ............................................................................................. 9 1.3 .30 Shutdown Margin .................................................................................................. 10

1.3.31 Surveillance Activities ........................................................................................... I 0 1.3.32 Surveillance Intervals ............................................................................................ 10 1.3.33 Unscheduled Shutdown ......................................................................................... 10 2.0 Safety Limit and Limiting Safety System Settings ........................................................... 11 2.1 Safety Limit ................................................................................................................... 11 2.2 Limiting Safety System Settings ................................................................................... 11 3.0 Limiting Conditions for Operations .................................................................................. 13 3 .1 Reactor Core Parameters ............................................................................................... 13 3.1.1 ReactorPower ........................................................................................................ 13 3 .1.2 Reactivity Limitations ..................................................................... *...................... 13 3 .1.3 Core Configuration ................................................................................................ 14 3.1.4 Fuel Bumup ........................................................................................................... 14 3 .2 Reactor Control and Safety Systems ............................................................................. 15 3.2.1 ShimArrns ............................................................................................................. 15 3 .2.2 Reactor Safety System Channels ........................................................................... 16 3.3 Coolant System ................................................................ ,............................................. 17 3 .3 .1 Primary and Secondary .......................................................................................... 17 3.3.2 Emergency Core Cooling ....................................................................................... 18 3.3.3 Moderator Dump System.: ..................................................................................... 19 3.4 Confinement System ...................................................................................................... 19 3.4.1 Operations that Require Confinement ................................................................... 19 3.4.2 Equipment to Achieve Confinement ..................................................................... 20 3.5 Ventilation System ........................................................................................................ 22 3.6 Emergency Power System ............................................................................................. 23 3.7 Radiation Monitoring Systems and Effluents ................................................................ 23 3.7.1 Monitoring Systems and Effluent Limits .............................................................. 23 3.7.2 Effluents ................................................................................................................. 25

3.8 Experiments .................................................................................................................... 26 3.8.1 Reactivity Limits ...................................................................... ~ ............................ 26 3.8.2 Materials ...................................................................... ;........................................... 27 3.9 Facility Specific ...... :...................................................................................................... 29 3.9.1 Fuel Storage ........................................................................................................... 29 3.9.2 Fuel Handling **************:*****************************************************************************************30 3.9.2.l Within the Reactor Vessel ................................................................................. 30 3.9.2.2 All Other Conditions ......................................................................................... 30 4.0 Surveillance Requirements ................................................................................................ 32 4 .1 Reactor Core Parameters .......................................................................*........................ 32 4.1.1 Reactor Power ........................................................................................................ 32 4.1.2 Reactivity Limitations ........................................................................................... 33 4.2 Reactor Control and Safety Systems ............................................................................ .33 4.2.1 Shim Arms ............................................................................................................. 33 4.2.2 Reactor Safety System Channels ........................................................................... 34 4.3 Coolant Systems ............................................................................................................ 35 .

4.3.1 Primary and Secondary .......................................................................................... 35 4.3.2 Emergency Core Cooling System .......................................................................... 35 4.3.3 Moderator Dump System ....................................................................................... 36 3

4.4 Confinement System ...................................................................................................... 36 4.5 Ventilation System ........................................................................................................ 37 4.6 Emergency Power System ............................................................................................. 38 4.7 Radiation Monitoring System and Effluents ................................................................. 39 4.7.1 Monitoring System ................................................................................................ 39 4.7.2 Effiuents ................................................................................................................. 41

4.8 Experiments .................................................................................................................... 41 5.0 Design Features ................................................................................................................. 43 5.1 Site Description ............................................................................................................. 43 5.2 Reactor Coolant System ................................................................................................ 43 5.3 Reactor Core and Fuel ................................................................................................... 44 6.0 Administrative Controls ..................................................................................................... 45 6.1 Organization ...........................................................................................*...................... 45 6.1.1 Structure .............................................................*......................... :......................... 45 6.1.2 Responsibility *******************************************'**********'*:............................................... 45 6.1.3 Staffing .............................................................................................................. ,.... 45 6.1.4 Selection and Training of Personnel... .................................................................. .46 6.1.4.1 Selection of Personnel ........................................................ :.............................. 46 6.1.4.2 Training of Personnel ........................................................................................ 47 6.2 Review and Audit .......................................................................................................... 47 6.2.1 Composition and Qualifications ........................................................................... .48 6.2.2 Safety Evaluation Committee Charter and Rules ................................................. .48 6.2.3 SEC Review Function ........................................................................................... 48 6.2.4 SEC Audit Function ............................................................................................... 49 6.2.5 Safety Assessment Committee (SAC) .......... ;........................................................ .49 6.3 Radiation Safety ................................. :............................................................................ 49 6.4 Procedures ........................... :* ......................................................................................... 50 6.5 Experiment Review and Approval ................................................................................ 50

6.6 Required Actions ********************************************************'**************************************************51 6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 other than a Safety Limit Violation ................................................................. 51 6.7 Reports ........................................................................................................................... 52

6. 7.1 Annual Operating Report ............................. .- ......................................................... 52 6.7.2 Special Reports ...................................................................................................... 52 6.8 Records .......................................................................................................................... 53 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less than Five Years ....................................................................... 53 6.8.2 Records to be Retained for at Least One Operator Licensing Cycle ..................... 54 6.8.3 Records to be Retained for the Life of the Reactor Facility .................................. 54 4

1.0 Introduction These technical specifications apply to the National Institute of Standards and Technology (NIST) Test Reactor (NBSR) license TR-5.

1.1 Scope The following areas are addressed: Definitions, Safety Limits (SL) and Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO),

Surveillance Requirements, Design Features, and Administrative Controls.

1.2 Application The dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values as a result of the normal construction and manufacturing tolerances, or normal accuracy of instrumentation.

1.2.1 Purpose These specifications are derived from NISTIR 7102 (NBSR 14 Safety Analysis Report). They consist of specific limitations and equipment requirements for the safe operation of the reactor and for dealing with abnormal situations. These specifications represent a comprehensive envelope of safe operation. Only those operational parameters and equipment requirements directly related to verifying and preserving this safety envelope are listed.

1.2.2 Format The format of these specifications is as described in ANSI/ANS 15.1-2007.

1.3 Definitions The following terms are sufficiently important to be separately defined:

1.3.1 ALARA As 1ow As is Reasonably Achievable. The practice of making every reasonable effort to maintain exposures to radiation as far below dose limits as is practicable, consistent with the purpose and benefits of licensed activities and the mission of the NBSR.

1.3.2 Channel The combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.

5

1.3 .2 .1 Channel Calibration The adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.

1.3.2.2 Channel Check A qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same variable.

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

1.3.3 Confinement An enclosure of the C wing of the NCNR that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.

1.3 .4 Core Configuration The number, type, or arrangement of fuel elements, reflector elements and regulating or control rods occupying the core grid.

1.3.5 Excess Reactivity That amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is critical.

1.3 .6 Emergency Director The functions of the Emergency Director are defined in the NBSR Emergency Plan.

1.3.7 Experiment 1.3.7.l In-Reactor Vessel Any operation, hardware, or target (excluding devices such as detectors and foils), that is designed to investigate non-routine reactor characteristics or that is intended for irradiation within the reactor vessel.

1.3.7.2 Beam Tubes Any sample or hardware placed in a beam tube that has an unobstructed view of the reactor vessel or any materials placed in a 6

beam tube, such as filters and shields for which accident mitigation credit is taken.

1.3. 7.3 Movable Experiment Any experiment in which all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating 1.3. 7.4 Secured Experiment Any experiment, experimental apparatus, or component of an experiment that is held in a stationary p9sition relative to the reactor by mechanical means. The restraining force must be substantially greater than those to which the experiment might be subjected by hydraulic, pneuma.tic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.

1.3.8 License

The written authorization, by the Nuclear Regulatory Commission, for an individual or organization to carry out the. duties and responsibilities associated with the NBSR.

1.3.9 Measured Value The value of a parameter as it appears on the output Of a channel.

1.3 .10 Moderator Dump An action which drops the water level. to approximately one inch (2.5 cm) above the reactor core, thereby ensuring a subcritical state for an emergency shutdown under all reactor operating conditions.

1.3.11 Natural Convection Cooling The flow of primary water between the reactor core and a heat exchanger with no pumps operating..

1.3.12 Operable The condition of a system or component when it is capable of performing its intended function, as determined by testing or:indication.

1.3.13 Operating The condition of a component or system when it is performing its intended function.

1.3 .14 Protective Action The initiation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor facility having reached a specified limit.

7

1.3.15 Reador Operating

The condition of the reactor when it is not secured or shutdown.

1. 3 .16 Reactor Operator An individual licensed by the U.S. Nuclear Regulatory Commission to manipulate the controls of the NBSR.

1.3.17 Reactor Safety System Those systems designated in these technical specifications, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

1.3 .18 Reactor Secured The condition of the reactor when (a), (b), or (c) is true.

(a) (1) The Control Power key switch or the Rod Drive Power key switch is in the off position with the key removed and under the control of a licensed operator; and

.. (2) The condit"ion of the shim arms is per the specification of Section 3 .1.2(2); and (3) No work is in progress. involving core fuel, core structure, installed shim arms, or shim arm drives, unless the shim arm drive shafts are mechanically fixed; and (4) No experiments in any reactor experiment facility, or in any other way near the reactor, are beirig moved or serviced if the experiments have, on movement, reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller.

I (b) There is insufficient fissile material in the reactor core or adjacent experiments to attain criticality under optimum available conditions of

moderation and reflection. * *

(c) The reactor is in the rod drop test mode, and a senior reactor operator is in direct charge of the operation.

1.3 .19 Reactor Shutdown When the reactor-is subcritical by at least one dollar ($1.00) in the Reference Core Condition with all installed experiments in their most reactive condition .

. 1:3.20 Reactor Shutdown Mechanisms Mechanisms that can place the reactor in a shutdown condition, and include:

(a) Rundown (b) Scram .

(c) Major Scram (d) Moderator Dump 8

1.3 .21 Reference Core Condition The condition of the core when it is at ambient temperature and the reactivity worth of xenon is negligible.

1.3.22 Reactor Rundown The electrically driven insertion of all shim arms and the regulating rod at their normal operating speed.

1.3.23 Rod, Control A device, also known as a shim arm, fabricated from neutron absorbing material that is used to establish neutron flux changes and to compensate for routine reactivity losses. The shim arms, when coupled to their drives, provide reactivity control and therefore flux control. When the shim arm becomes decoupled from*its drive mechanism it provides a safety function by rapidly introducing negative reactivity into the reactor core.*

1.3 .24 Rod Drop Mode Any combination of control systems and mechanical systems that allows for the movement of only a single shim arm and ensures the reactor remains shutdown, when sufficient fissile material for criticality is present.

1.3.25 Rod, Regulating 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 automatically.

1.3.26 Scram The spring assisted gravity insertion of all shim arms.

1.3.26.1 Major Scram A scram accompanied by the immediate activation of the confinement isolation system.

1.3.27 Scram Time The elapsed time between the initiation of a scram signal and a specified movement of a control or safety device.

1.3.28 Senior Reactor Operator An individual licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.

1.3.29 Shall, Should and May The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.

9

1.3.30 Shutdown Margin The minimum shutdown reactivity necessary to provide confidence that the reactor can be shutdown by means of the control and safely systems starting from any permissible operating condition, with the most reactive shim arm in the most reactive position and the regulating rod fully withdrawn, and that the reactor will remain shutdown without further operator action.

1.3.31 Surveillance Activities Those tests, checks and calibrations done to predict the operability of the equipment described in Section 4.0.

1.3.32 Surveillance Intervals Maximum intervals are established to provide operational flexibility and not to reduce frequency. Established frequencies shall be maintained over the long term. The surveillance interval is the time between a check, test or calibration, whichever is appropriate to the item being subjected to the surveillance, and is measured from the date of the last surveillance. Surveillance intervals are:

(a) Five Year Interval not to exceed six years.

(b) Biennial Interval not to exceed two. and one-half years.

(c) Annual Interval not to exceed 15 months.

(d) Semi-annual .

Interval not to exceed seven and a half months.

(e) Quarterly Interval not to exceed four months.

(f) Monthly Interval not to exceed six weeks.

(g) Weekly Interval not to exceed ten days.

1.3.33 Unscheduled Shutdown Any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or equipment operability checks.

10

2.0 Safety Limit and Limiting Safety System Settings 2.1 Safety Limit Applicability: Fuel temperature Objective: To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products.

Specification

(

The reactor fuel cladding temperature shall not exceed 842°F (450°C) for any operating conditions of power and flow.

Maintaining the integrity of the fuel cladding requires that the cladding remain below its blistering temperature of842°F (450°C). For all reactor operating conditions that avoid either a Qeparture from nucleate .Qoiling (DNB), or exceeding the Critical Heat Flux (CHF)), or the Qnset of flow instability (OFI), cladding temperatures remain substantially below the fuel blistering temperature. Conservative calculations have shown that limiting combinations of reactor power and reactor coolant systein flow and temperature will prevent DNB and thus fuel blistering.

2.2 Limiting Safety System Settings Applicability: Power, flow, and temperature parameters Objective: To ensure protective action if any combination of the principal process variables should approach the safety limit.

Specifications (1) Reactor power shall not exceed 130% of full power.

(2) Reactor outlet temperature shall not exceed 14 7°F.

(3) Forced coolant flow shall not be less than 60 gprn/MW for the inner plenum and not less than 235 gprn/MW for the outer plenum.

(4) Reactor power, with natural circulation cooling flow, shall not exceed 10 kW.

11

At the values established above, the Limiting Safety System Settings provide a significant margin from the Safety Limit. Even in the extremely unlikely event that reactor power, coolant flow, and outlet temperature simultaneously reach their Limiting Safety System Settings, the critical heat flux ratio (CHFR) is at least 2. For all other conditions the CHFR is considerably higher. This will ensure that any reactor transient caused by equipment malfunction or operator error will be terminated well before the safety limit is reached. Overall uncertainties in process instrumentation have been incorporated in the Limiting Safety System Settings.

Steady state thermal hydraulic analysis shows that operation at 10 kW with natural circulation results in a CHFR and OFI ratio greater than 2. Transient analysis of reactivity insertion accidents shows that the fuel cladding temperature remains far below the safety limit.

12

3.0 Limiting Conditions for Operations 3.1 Reactor Core Parameters 3 .1.1 Reactor Power Applicability: Reactor power Objective: To ensure that licensed power is not exceeded and the safety limit is not exceeded through initiation of protective action at a specified power.

Specification The nominal reactor power shall not exceed 20 MW thermal. The reactor scram set point for a reactor power level safety channel shall not exceed 125% of full power.

Operational experience and thermal-hydraulic calculations demonstrate that the fuel elements may be safely operated at these power levels. The operating limits developed here are based upon well tested correlations, are conservative, and provide ample margin to ensure that there will be no damage to fuel during normal operation. In addition, the operating conditions provide ample margin for all credible accident scenarios to ensure that there will be no fuel damage.

3.1.2 Reactivity Limitations Applicability: Core reactivity and shim arm worth Objective: To ensure that the reactor can be placed in a shutdown condition at all times and that the safety limit shall not be exceeded.

Specifications (1) The maximum available excess reactivity for reference core conditions shall not exceed 15% Llp (approximately $20).

(2) The reactor shall not be operated unless the shutdown margin provided by the shim arms is greater than 0.757% Llp ($1.00) with:

(a) The reactor in any core condition, and (b) All movable experiments in their most reactive condition.

13

(1) An exce~s reactivity limit provides adequate excess reactivity to override the xenon buildup and to overcome the temperature change in going from zero power to 20 MW, without affecting the required shutdown margin. In addition, the maximum reactivity insertion accident at startup, which assumes the insertion of 0.5% ~p into a critical core, is not affected by the total core excess reactivity.

(2) These specifications ensure that the reactor can be put into a shutdown condition from any operating condition and remain shutdown even if the maximum worth shim arm should stick in the fully withdrawn position with the regulating rod also fully withdrawn.

3.1.3 Core Configuration Applicability: Core grid positions Obj~ctive: To ensure that a failed shim arm does not adversely affect core reactivity and cooling flow is maintained.

Specification The reactor shall not operate unless all grid positions are filled with full length fuel elements or thimbles.

The NBSR employs shim arm stops to prevent a broken shim arm from dropping from the reactor core. The proper operation of these stops depends bn adjacent fuel elements or experimental thimbles being in place to prevent the broken shim arm from falling from the core lattice. Furthermore, core grid positions shall be filled to prevent coolant flow from bypassing the fuel elements.

3.1.4 Fuel Burnup Applicability: Fuel Objective: To remain within allowable limits of burnup 14

Specification 27 The average fission density shall not exceed 2 x 10 fissions/m 3

Fuel elements in the NBSR are burned for seven (7) or eight (8) cycles. An eight (8) cycle fuel element.has an average fission density of approximately 1.9 x 1027 fissions/m 3

The U 30 8 - Al dispersion MTR fuels have been in widespread use for over 40 years. Extensive testing of fuel plates has been performed to determine the limits on fission density as a function of fuel loading. Several measurements of swelling in fuel plates show that NBSR fuel, which is moderately loaded at 18% is well below the curve that represents the allowable limit of burnup.

3.2 Reactor Control and Safety Systems 3.2.1 Shim Arms Applicability: Shim arms and shim arm worth Objective: To ensure proper shim arm reactivity insertion.

Specifications The reactor shall not be operated unless:

(1) All four shim arms are operable.

(2) The scram time shall not exceed 240 msec for a shim arm insertion of 5 degrees.

(3) The reactivity insertion rate for the four shim arms shall not exceed 5 x 10-4 ~p/sec.

(1) Although the NBSR could operate and maintain a substantial shutdown margin with less than the four installed shim arms, flux and shim arm worth distortions could occur by operating in this manner. Furthermore, operation of the reactor with one shim arm known to be inoperable would further reduce the shutdown margin that would be available if one of the remaining three shim arms were to suffer a mechanical failure that prevented its insertion.

(2) and (3) A shim arm withdrawal accident for the NBSR was analyzed using the maximum reactivity insertion rate, corresponding to the maximum 15

beginning-of-life shim arm worths with the shim arms operating at the design speed of their constant speed mechanisms. The analysis shows that the most severe accident, a startup from source level, will not result in:

core damage.

3.2.2 Reactor Safety System Channels Applicability: Required instrument channels Objective: To provide protective action for nuclear and process variables to ensure the LSSS values are not exceeded.

Specifications The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed in the reactor Control Room.

Table 3.2.2 Reactor Safety System Channels Minimum Nuclear and Process Channels Required Channel Scram . Major Scram Rundown

(1) High Flux level 2 (2) Short period below 5% rated power 2 (3) Low reactor vessel D 2 0 level J, 3 2 (4) Low flow reactor outlet 2' 3 1 2 3 (5) Low flow reactor inner or outer plenum ' 1 (6) Manual (outside of the Control Room) , 1 (7) Manual 1 1 (8) Reactor Outlet Temperature . 4 1

(9) Gaseous Effluent Monitors 2 1

One (1) of two (2)channels may be bypassed for tests or during the time maintenance involving the replacement of components and modules or calibrations and repairs are actually being performed.

2 One (1) of these two (2) flow channels may be bypassed during tests; or during the time maintenance involving the replacement of components and modules or calibrations and minor repairs are actually being performed. However; outlet low flow may not be bypassed unless both inner and outer low-flow reactor inlet safety systems are operating.

3 May be bypassed during periods of reactor operation (up to 10 kW) when a reduction in Limiting Safety System Setting values is permitted per the specifications of Sections 2.2 and 3.3.1.

4

. See specifications of Section 3. 7 .1 16

The nuclear and process channels of Table 3.2.2 initiate protective action to ensure that the safety limit is not exceeded. With these channels operable, the safety system has redundancy.

3.3 Coolant System 3 .3 .1 Primary and Secondary

. Applicability: Primary fluid systems Objective: To prevent degradation of primary systems' materials.

Specifications The reactor shall not be operated unless:

(1) The .reactor vessel coolant level is no more than 25 inches below the overflow standpipe.

Exception:To permit periodic surveillance of the effectiveness of the moderator dump, it is necessary to operate the reactor without restriction on reactor vessel level.

(2) The 0 2 concentration in the Helium Sweep System shall not exceed 4% by volume.

(3) All materials, including those of the reactor vessel, in contact with the primary coolant shall be compatible with the 0 2 0 environment.

(1) The limiting value for reactor vessel coolant level is somewhat arbitrary because the core is in no danger so long as it is covered with water.

However, a drop of vessel level indicates a malfunction of the reactor cooling system and possible approach to uncovering the core. Thus, a measurable value well above the minimum level is chosen in order to provide a generous margin of approximately7 feet (2.13 m) above the fuel elements. To permit periodic surveillance of the effectiveness of the moderator dump, it .is necessary to operate the reactor without restriction on reactor vessel level. This is permissible under conditions when forced reactor cooling flow is not required, such as is permitted in the specifications of Section 2.0.

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(2) Deuterium gas will collect in the helium cover gas system because of radiolytic disassociation of D20. Damage to the primary system could occur if this gas were to reach an explosive concentration (about 7.8% by volume at 77°F (25°C) in helium if mixed with air). To ensure a substantial margin below the lowest potentially explosive value, a 4% limit is imposed.

(3) Materials of construction, being primarily low activation alloys and stainless steel, are chemically compatible with the primary coolant. The stainless steel pumps are heavy walled members and are in areas of low stress, so they should not be susceptible to chemical attack or stress corrosion failures. A failure of the gaskets or valve bellows would not result in catastrophic failure of the prim~ry system. Other materials should be compatible so as not to cause a loss of material and system integrity.

3.3.2 Emergency Core Cooling Applicability: Emergency Core Cooling System Objective: To ensure an emergency.supply ofcoolaht.

Specifications The re.actor shall not be operated unless:

(1) The D20 emergency core cooling system is operable.

(2) A source of makeup water to the D20 emergency cooling tank is available.

(1) In the event of a loss of core coolant, the emergency core cooling system provides adequate protection against melting of the reactor core and associated release of fission products.

(2) The emergency core cooling system employs one sump pump to return spilled coolant to the overhead storage tank. Because only one sump pump is used, it must be operational whenever the reactor is operational. There is

sufficient D2 0 available to provide approximately 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of cooling on a once-through basis. In the event that the sump pump fails and the D 20 supply in the overhead storage tank is exhausted, domestic water or a suitable alternative would be useci to furnish water for once-through cooling. The water makeup capacity must be in excess of 25 gpm, which was found adequate in cooling calculations to prevent fuel damage.

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3.3.3 Moderator Dump System Applicability: Moderator dump

_Objective: To provide a backup shutdown mechanism.

Specification The reactor shall not be operated unless the reactor moderator dump system is operable.

In the unlikely event that the shim arms cannot be inserted, an alternate means of shutting down the reactor is provided by the moderator dump. The moderator dump provides a shutdown capability for any core configuration. Hence, it is considered necessary for safe operation. It has been shown that the moderator dump provides sufficient negative reactivity to .make the normal startup (SU) core subcritical .even with all four shim arms fully withdrawn.

3.4 Confinement System 3.4.l Operations that Require Confinement .

Applicability: Reactivity changes within the vessel and fuel movements outside of the vessel

.Objective: To provide an additional barrier to fission product releases.

Specifications Confinement shall be maintained when:

( 1) The reactor is operating...

(2) Changes of components or .equipment within the confines of the thermal shield, other than rod drop-tests or movement of experiments, are being made which could cause a significant change in reactivity.

(3) There is movement of irradiated fuel outside a sealed container or system.

(4) The reactor has been shutdown for shorter than the time specified in the specification of Section 3.9.2.2.

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(1) The confinement system is a major engineered safety feature. It is the final physical barrier to mitigate the release of radioactive particles and gasses to the environment following accidents. Confinement is stringently defined to ensure that the confinement building shall perform in accordance with its design basis. Confinement is not required when the reactor is shutdown and experiments are to be inserted or removed.

(2) Changes in the core involving such operations as irradiated fuel handling or shim arm repairs affect the reactivity of the core and could reduce the shutdown margin of the reactor. Confinement shall be required when these changes are made because they affect the status of the core.

The reactor is normally shutdown by a substantial reactivity margin.

Experiments are usually inserted and removed one at a ti~e; hence, the total reactivity change in any single operation shall be limited to the specified maximum worth of 0.5% £1p for any single ex:periment (including "fixed" experiments). Under this circumstance, the shutdown

margin would be substantial.

(3) Even when the reactor is shutdown, irradiated fuel contains fission product inventories sufficient to allow the specification of Section 3. 7 .2 to be exceeded shouid the eiement fail. This fuel poses a potential hazard in that its cladding could be damaged when it is not contained in a closed system, such as during transit or during sawing of aluminum end pieces.

Confinement integrity is not required when irradiated fuel is contained within a closed system, such as the reactor vessel; the transfer lock of the refueling system, or a sealed shipping cask, that serves as a secondary barrier of fission product release. *

(4) The specification of Seetion 3.9.2.2 restricts fuel movement for a specified period. Maintenance that would disable the confinement is prohibited during that period. Building doors could be opened, however, provided that confinement can .be rapidly re-established. Confinement integrity is no longer required after the waiting period, because a loss of all water to fuel in a sealed container or system will not cause fuel damage.

3.4.2 Equipment to Achieve Confinement Applicability: Confinement system Objective: To ensure that TS 3.4.1 can be met.

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Specifications Confinement shall mean that:

(1) All penetrations of the confinement building are either sealed or capable of being isolated. All piping penetrations within the reactor building are capable of withstanding the confinement test pressure.

(2) All automatic isolation valves in the. v~ntilation, process piping and guide tubes are either operable or can be closed.

(3) All automatic personnel a,ccess doors can be Closed and sealed.

(4) Except during passage, at least one set of the reactor building vestibule doors for each automatic personnel door is closed or attended, or the automatic door is closed and sealed.

(5) The reactor building truck door is closed and sealed.

Exception to (1) - (5): In order to provide for prompt remedial action, reactor confinement effectiveness may be _reduced for a period of no longer than 15 minutes when specifications (1) - ( 5) are not met or do not exist.

(1) and. (2) The confinement building is designed to be automatically sealed upon indication of high activity. To a~empt to operate the reactor with any of these conditions unmet is a violation of the confinement design basis.

Although tests have.shown that the confinement building can continue to operate with one or more of these closures failed, its margin of effectiveness is reduced. If.a closure device is placed in its closed or

sealed condition, th<;:n operability of the automatic closure device is not required.

(3) and (4) Tests performed on the confinement building have shown that even if one of the automatically closing personnel doors fails to operate properly, confinement design capability can be met if one set of building vestibule doors per vestibule are closed. By specifying that these doors remain closed e~cept when they are being used or attended, a backup to the normal confinement clos.ure is provided.

(5) The reactor building truck door is not provided with automatic closure devices. Tests have shown that the confinement building can continue to operate properly, although at reduced efficiency, if the truck door seal were to fail. Confinement cannot be established if the truck door is open.

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3.5 Ventilation System Applicability: Emergency and normal ventilation Objective: T~ minimize exposures outside of the confinement building.

Specifications

.The reactor shall not be operated unless:

(1) The building emergency recirculation system and emergency exhaust systems,

. including both fans, ate operable, and both the absolute and charcoal filter

efficiencies are at 99% or greater.

(2) The reactor-building ventilation system can filter exhaust air and discharge it*

above the confinement building rooflevel.

Exception to ( 1) and (2): In order to provide time for prompt remedial action, reactor ventilation may be inoperable for a period of no longer than 15 minutes when the specifications are not met or do not exist. Minor maintenance which disables a single fan and can be suspended without affecting the operability of the system may be performed during reactor operation:

the potential radiation.exposure to staff personnel and persons at the site boundary and beyond has been cakulated follo\¥ing an accidental release of fission product activity. These calculations are*based on the proper operation of the building recirculation system and the emergency exhaust system to maintain the confinement building at a negative pressure and to direct all effluents through filters and up through the reactor building stack. The emergency exhaust system is a redundant

system to ensure its operation. Because of its.importance, this redundancy should be available at all times so that any single failure would not preclude system operation when required.

The emergency exhaust system is designed to pass reactor building effluents through high-efficiency particulate filters capable of removing particles.of 0 .3. µm or greater with an effidency of at least 99% and the charcoal filters are capable of removing greater than 99% of the Iodine from the air. All discharge of the effluents is above the reactor building roof level. This system ensures filtering and dilution of gaseous effluents before these effluents reach personnel either onsite or offsite. The system can properly perform this function using various combinations of its installed fans and the building stack.

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3.6 Emergency Power System Applicability: Emergency electrical power supplies Objective: To ensure emergency power for vital equipment.

Specification The reactor shall not be operated unless at least one (1) of the diesel-powered generators and the station batteries (consisting of at least one (1) battery supplying a critical power UPS and one (1) battery supplying the 125 VDC buses) are operable, including associated distribution equipment, and the nuclear instrumentation and emergency exhaust fans can be supplied with electrical power from the diesel generator or the batteries.

Exception: In order to provide time for prompt remedial action, the Emergency Power System may be inoperable for a period of no longer than 15 minutes when the specification is not met or does not exist.

One diesel-powered generator is capable of supplying emergency power to all necessary emergency equipment. The second diesel-powered generator is provided to permit outages for maintenance and repairs.

The station batteries provide an additional source of emergency power for the nuclear instruments and the emergency exhaust fans. These fans may be powered from AC or DC power supplies. The batteries are capable of supplying this emergency load for a minimum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . By allowing this amount of time and by requiring operability of at least one diesel and the station batteries, adequate emergency power sources shall always be available.

3.7 Radiation Monitoring Systems and Effluents 3.7.1 Monitoring Systems and Effluent Limits Applicability: Radiation monitoring systems

- Objective: To detect abnormal levels or locations ofradioactivity.

Specifications The reactor shall not be operated unless:

(1) Two of three gaseous effluent monitors are operable for normal air, irradiated air, and stack air.

Amendment No.: 10 September 10, 2015 23

(2) One fission product monitor is operable or sample analysis for fission product activity is conducted daily.

(3) One secondary coolant activity monitor is operable or a D 2 0 storage tank level monitor is operable.

(4) Two area radiation monitors are operable on floors C-100 and C-200.

(5) The primary tritium concentration is less than or equal to 5 Ci/I.

(6) Removed to 3.7.2.

When required monitors are inoperable, then portable instruments, survey or analysis may be substituted for any of the normally installed monitors in specifications (1)- (4) for periods of one (1) week or for the duration of a reactor run.

(1) The requirements of 10 CFR 20.1502(b) (2007) are met by regular monitoring for airborne radionuclides and bioassay of exposed personnel.

41 The two primary airborne radionuclides present at the NBSR are Ar and 3

H. The normal air exhaust system draws air from areas supplied by conditioned air, such as the first and second floors of the coqfinement building. The irradiated air exhaust system draws air from areas most likely to have contaminated air, such as waste sumps and penetrations in the*biological shield. Normal and irradiated air are monitored continuously with G-M detectors sensitive to p and y emissions and the combined air is exhausted through the stack. The stack release is monitored with a G-M detector.

(2) A fission products monitor located in the helium sweep gas will give an indication of a "pin-hole" breach in the cladding so that early preventive measures can be taken. When this monitor is not functional, daily testing will ensure that the fuel cladding is intact. These two measures ensure that there are no undetected releases of fission products to the primary coolant.

(3) Monitoring for primary water leakage into the secondary coolant is done by a secondary water monitor that is sensitive to radionuclides in the primary water. Leakage of primary to secondary would also be detected by a change in the D 2 0 storage tank level (4) Fixed gamma area radiation monitors are positioned at selected locations in the confinement building. Typical alarm setting are less than 5 mrem/hr and adjusted as needed for non-routine activities, generally with the objective of identifying unusual changes in radiation conditions.

(5) At the end of the term of the NBSR license the maximum tritium Amendment No. 9 September 24, 2013

concentration in the primary coolant is estimated to be 5 Ci/I. This value and reliable leak detection ensures that tritium concentrations in effluents shall be as low as is practicable.

(6) Removed to 3.7.2.

3.7.2 Effluents Applicability: Annual releases Objective: To minimize exposures to the public.

Specification The reactor shall not be operated unless:

(1) The total exposure from effluents from the reactor facility to a person at the site boundary shall not exceed 100 mrem per calendar year, less any external.dose from the facility. The limit shall be es~blished at the point of release or measurement using accepted diffusion factors to the boundary. For halogens and particulates with half-lives longer than 8 days, a reconcentration factor shall be included where appropriate.

(2) An environmental monitoring program shall be carried out and shall include as a minimum the analysis of samples from surface waters from the surrounding areas, vegetation or soil and air sampling.

The criteria for determination of concentration limits specified above ensure that 10 CFR 20 (2007) limits are not exceeded at the site boundary. The allowance for dilution from the reactor building stack to the nearest site boundary is 1,000. This value of 1,000 from the diffusion view point is the minimum expected at the nearest site boundary under the least favorable meteorological conditions. This number could be increased by one or two orders of magnitude if normal variations in wind speed and direction were considered. Because these variations are not considered, a one or two order of magnitude margin is inherent in this limit.

In specifying the limits on particulates and long lived (longer than 8 days) halogens, consideration was given to the possibility of biological reconcentration in food crops or dairy products. Using available information (Soldat, J.D., Health Physics 9, p. 1170, 1963), a conservative (both the COMPLY and CAP88 codes indicate that 700 is at least an order of Amendment No. 9 September 24, 2013

magnitude higher than needed) reconcentration factor of 700 is applied. Thus, the limits for those isotopes are the Effluent Concentration Limits as speCified in Appendix B, Table II of 10 CFR 20 (2007) multiplied by the 1,000 dilution factor divided by the 700 reconcentration factor; that is, 1.4 times the Effluent Concentration Limit.

For the purpose of converting concentrations to dose, the values of 10 CFR 20, Appendix B, Table 2 (2007), represent an annual dose of 50 mrem, except for submersion gases where they represent an annual dose of 100 mrem. It should be taken into consideration that the values for submersion gases are based on an infinite hemisphere geometry which is rarely achievable and therefore tends to overestimate the dose.

Area vegetation and soil samples are collected for analysis. Grass samples are collected during the growing season, April through September, and soil samples during the non-growing season, October through March.

Thermo luminescent dosimeters or other devices also are placed around the

perimeter of the NBSR site to monitor direct radiation. The continuation of this environmental monitoring program will verify that the operation of the NBSR presents no significant risk to the public health and safety. Since 1969, when the NBSR began routine power operation, the environmental monitoring program has revealed nothing of significance, thereby confirming that operation of the NBSR has had little or no effect on the environment.

A report published in March 2003 supports the findings of previous studies conducted on the hydrology and geology of the NIST site and vicinity. No significant changes in the hydro-geologic systems or ground water use were identified. This report further verifies the assumptions and techniques developed in 1964.

3.8 Experiments 3 .8.1 Reactivity Limits Applicability: Reactivity of experiments Objective: To limit reactivity excursions.

Specifications The reactor shall not be operated unless:

(1) The absolute reactivity of any experiment shall not exceed 0.5% ~p.

(2) The sum of the absolute values of reactivity of all experiments in the reactor and experimental facilities shall not exceed 2.6% ~p.

Amendment No. 9 September 24, 2013

(3) No experiment malfunction shall affect any other experiment so as to cause its failure: Similarly, no reactor transient shall cause an experiment to fail in such a way as to contribute to an accident.

(1) The individual experiment reactivity limit is chosen so that the failure of an experimental installation or component shall not cause a reactivity increase greater than can be controlled by the regulating rod. Because the failure of individual experiments cannot be discounted during the operating life of the NBSR, failure should be within the control capability of the reactor. This limit does not include such semi-permanent structural materials as brackets, supports, and tubes that are occasionally removed or modified, but which are positively attached to reactor structures. When these components are installed, they are considered str1:1ctural members rather than part of an experiment.

(2) The combined reactivity allowance for experiments was chosen to.allow sufficient reactivity for contemplated experiments while limiting neutron flux depressions to less than 10%. Included within the specified 2.6% 8p is a 0.2% 8p allowance for the pneumatic irradiation system, 1.3% 8p for experiments that can be removed during reactor operation, and the remainder for semi-permanent experiments that can only be removed during reactor shutdown. Even if it were ass~med that one experiment with the maximum allowabJe reactivity of 0.5% 8p for movable experiments was removed* in 0.5 seconds, analysis shows that this ramp insertion into the NBSR operating at 20 MW would not result in any fuel failure leading to the release of fission products. The 0.2% 8p for the combined pneumatic irradiation systems has been shown to be bounded by the ramp in~ertion of 0.5% 8p and is well below this referenced accident as well as being within the 8p capability of the regulating rod.

(3) In addition to all reactor experiments being designed not. to fail from internal gas buildup or overheating, they shall be designed so that their failure does not affect either the reactor or other experiments. They shall also be designed to withstand, without failure, the same transients that the reactor itself can withstand without failure.

3.8.2 Materials Applicability: All materials used in experiments Objective: To prevent damage to the reactor or a significant release of radioactivity.

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Specifications (1) Explosive or metastable mate"rials capable of significant energy releases shall be irradiated in double walled containers that have been satisfactorily tested.

(2) Each experiment containing materials corrosive to reactor components or highly reactive with the.reactor or experimental coolants shall be doubly contained.

(3) All experiments performed at the NBSR shall be reviewed and authorized in accordance with the specifications of Section 6.5.

(1) In addition to all reactor-experiments* being designed not to fail from internal overheating or gas buildup, they shall also be designed to be compatible with their environment in the reactor. Specifically, their failures shall not lead to failures of the core structure or reactor fuel, or to

the failure of other experiments. Also, reactor experiments shall be able to withstand the same transients that the reactor itself can withstand, such as loss of reactor cooling flows and startup accident.

The detonation Of explosive or metastable materials within the reactor is not an intended* pan of the experimental procedure for the NBSR, but the possibility of a rapid energy release shall be considered when these materials are present. Fun testing of the contai.ner design shall be done.

{2) Experiments containing materials corrosive to reactor components or highly reactive with reactor or experimental coolants shall have an added

margin of safety to prevent the release of these materials to the reactor coolant system. This margin of safety is provided by the double encapsulation, each container being capable of containing the materials to be irradiated.

(3) An independent technical review of experiments ensures the experiment will not reduce the reactor safety margin.

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3.9 Facility Specific 3.9.l Fuel Storage Applicability: Fuel element storage Objective: To prevent inadvertent criticality and maintain fuel element cladding integrity.

Specifications (1) All fuel elements or fueled experiments shall be stored and handled in geometry such that the calculated keff shall not exceed 0.90 under optimum

conditions of water moderation and reflection.

(2) The water chemistry, level, and temperature in the spent fuel storage pool shall be maintained so as to ensure the integrity of the fuel elements.

(1) To ensure that no inadvertent criticality of stored fuel elements or fueled experiments occurs, they shall be maintained in a geometry that ensures an adequate margin below criticality exists. This margin is established as a keff of no greater than 0.90 for the storage and handling of fuel or fueled experiments.

(2) The cooling of spent fuel elements in storage at the NBSR depends upon the decay heat of the elements, the volume of water in a storage pool, and any additional cooling, such as the use of pumps and heat exchangers. A

storage pool is a stable environment, where water chemistry, temperature and level are easily monitored and the fuel is adequately shielded.

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3.9.2 Fuel Handling 3.9.2.1 Within the Reactor Vessel Applicability: Fuel element latching Objective: To ensure that all fuel elements are latched between the reactor grid plates.

Specifications Following handling of fuel within the reactor vessel, the reactor shall not be operated until all fuel elements that have been handled are inspected to determine that they are locked in their proper positions in the core grid structure. This shall be accomplished by one of the following methods:

(1) Elevation check of the fuel element with main pump flow.

(2) Rotational check of the element head in.the latching direction only.

(3) Visual inspection of the fuel element head or latching bar.

Each NBSR fuel element employs a latching bar, which shall be rotated to lock the fuel element in the upper grid plate. Following fuel handling, it is necessary to ensure that this bar is properly

positioned so that an element cannot be lifted out of the lower grid plate, which could lead to a reduction in flow to the element after pump flow is initiated. Any of the three methods above may be used to verify bar position. Tests have shown that flow from a primary pump will raise an unlatched element above its normal position and thus will be detected by the pickup tool under flow conditions. The efficacy of rotational checks has been confirmed by visual inspections.

3.9.2.2 All Other Conditions Applicability: Refueling system Objective: To ensure the integrity of the fuel element cladding.

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Specification A fuel element shall not be removed from water in the reactor vessel unless the reactor has been shutdown for a period equal to or longer than one hour for each megawatt of operating power level.

To ensure that a fuel element does not melt and release radioactive material, a time limit is specified before a fuel element may be removed from the vessel following reactor shutdown. Measurements carried out during reactor startup showed that for the hottest element placed dry in the transfer chute, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after shutdown from*

10 MW, the maximum temperature was only 550°F without

auxiliary cooling. Extrapolation of these measurements shows that 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after shutdown from 20 MW, the maximum temperature for the hottest element would be less than 800°F without auxiliary coolant. For all other power levels below 20 MW the specified waiting time would result in even lowe.r temperatures. This provides a substantial margin of safety from the safety limi.t.

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4.0 Surveillance Requirements Introduction The Surveillance frequencies denoted*herein are based on continuing operation of the reactor.

Surveiflance activities scheduled to occur during an operating cycle which can* not be performed with the reactor operating may be deferred to the end of that current reactor operating cycle. If the reactor is not operated for a reasonable time, a reactor system or measuring channel surveillance requirement may be waived during the associated time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested. Surveillance intervals shall not exceed those defined in these Technical Specifications. Discovery of noncompliance with any of the surveillance sp~cifications below shall limit reactor operations to that required to perform the surveillance.

4.1 Reactor Core Parameters 4.1.1 Reactor Power Applicability: Reactor Safety System channels Objective: To ensure operability*of the safety system channels.

Specifications (1) The reactor safety system channels shall be channel tested before each reactor startup, following a reactor shutdown that exceeds 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or quarterly.

(2) The reactor safety system channels shall be channel calibrated annually.

(3) A channel check of power range indication, with flow multiplied by~ T, shall be performed weekly when the reactor is operating above 5 MW.

(4) Following maintenance on any portion of the reactor control or reactor safety systems, the affected portion of the system shall be tested before the system is considered operable.

The channel tests, calibrations and flow ~T comparison will ensure that the indicated reactor power level is correct. The power level channel calibration is performed by comparison of nuclear channels with the thermal power measurement channel (flow times ~T). Because of the small ~T (about 15°F at 20 MW), these calibrations will not be performed below 5 MW.

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4.1.2 Reactivity Limitations Applicability: Core reactivity and shim arm worth.

Objective: To ensure that the reactor can be placed in a shutdown condition at all times and that the safety limit shall not be exceeded.

Specifications (1) The excess reactivity (reference core conditions) shall be verified annually or following any significant changes in the core or shim arm configuration.

(2) The total reactivity worth of each shim arm and the regulating rod, and the shutdown margin shall be verified annually as described in these Technical Specifications, or following any significant change in the core or shim arm configuration.

(1) Determining the core excess reactivity annually will ensure that the critical shim arm positions do not change unexpectedly.

(2) Measurements of reactivity worth of the shim arms and regulating rod over many years of operation have shown rod worths vary slowly as a result of absorber bumup, and only slightly with respect to operational core loading and experimental changes. An annual check shall ensure that adequate reactivity margins are maintained.

4.2 Reactor Control and Safety Systems 4.2.1 Shim Arms Applicability: Shim arm motion Objective: To ensure proper shim arm reactivity insertion.

Specifications (1) The withdrawal and insertion speeds of each shim arm shall be verified semi-annually.

(2) Scram times of each shim arm shall be measured semi-annually.

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The shim arm drives are constant speed mechanical devices. A reactor scram is aided by a spring that opposes drive motion during shim arm withdrawal.

Withdrawal an.d insertion speeds or scram time should not vary except as a result of mechanical wear. The surveillance frequency is chosen to provide a significant margin over the expected failure or wear rates of these devices.

4.2.2 Reactor Safety System Channels Applicability: Required instrument channels Objective: To ensure reliability of protective action for nuclear and process variables.

Specifications The Scram and Confinement Channels shall have the surveillance requirements shown in Table 4.2.2.

Table4.2.2 Surveillance Requirements for the Scram and Confinement Channels Channel

Action Required Surveillance Required (1) High Flux level Scram X, A (2) Short period below 5% rated power Scram X, A (3) Low reactor vessel' D20 level Scram X,A (4) Low flow reactor outlet Scram X;A (5) Low flow reactor inner or outer plenum Scram X,A (6) Manual (outside of the Control Room) Scram X,A (7) Manual Scram X,A (8) Normal Air Exhaust Activity High Major Scram X,A (9) Irradiated Air Activity High Major Scram XA

( 10) Stack Air Activity High Major Scram X,A (11) Reactor Coolant Outlet Temperature Rundown X,A X - Channel test before startup after a shutdown of longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or quarterly.

A - Annual Channel Calibration.

To ensure that instrument failures do not go undetected, frequent surveillance of the listed channels is required and operating experience has shown these frequencies to be adequate to ensure channel operability.

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4.3 Coolant Systems 4.3.1 Primary and Secondary Applicability: Primary fluid systems Objective: To prevent degradation of primary system materials.

Specifications .

(1) The primary cooling system relief valve shall be tested annually.

(2) Major additions, modifications, or repairs of the primary cooling system or its connected auxiliaries shall be tested before the affected portion of the system is placed into service.

(3) The 0 2 concentration in the helium sweep gas shall be verified every five (5) years.

(1) The frequency for testing the pressure at which the relief valve opens is consistent with industry practices on this type of valve for clean water service conditions.

(2) Major additions, modifications, or repairs of the primary system shall be either pressure tested or checked by X-ray, ultrasonic, gas leak test, dye penetrants or other methods.

(3) Recombination of deuterium and oxygen is accomplished primarily by the reactor. Operational experience and data suggests that the specified frequency is appropriate for verifying 0 2 levels.

4.3.2 Emergency Core Cooling System Applicability: Emergency Core Cooling System Objective: To ensure an emergency supply of coolant.

Specifications (I) Control valves in the emergency core cooling system shall be exercised quarterly.

(2) The operability of the emergency sump pump, using either heavy or light water, shall be tested annually.

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(3) The light water injection valves shall be exercised semi-annually.

The equipment in this system is not used in the course of normal operation, so its operability shall be verified periodically. The frequencies are chosen so that deterioration or wear would not be expected to be an important consideration. Moreover, the frequency should be sufficient to ensure that the pumps and valves will not fail because of corrosion buildup or other slow acting effects during extended periods of standby operation. Control and injection valves specified are those leading to or from the D2 0 emergency cooling tank.

4.3.3 Moderator Dump System Applicability: Moderator dump valve Objective: To provide a backup shutdown mechanism.

Specification The Moderator Dump valve shall be cycled annually.

The moderator dump valve is of proven dependable design. Operating the dump valve annually is and has been a reliable predictor of performance.

4.4 Confinement System Applicability: Confinement building and components Objective: To ensure the continued integrity and reliability of the confinement building.

Specifications (1) A test of the operability of the confinement closure system shall be performed quarterly. The trip feature shall be initiated by each of the radiation monitors that provides a signal for confinement closure, as well as by the manual major scram switch. A radiation source shall be used to test the trip feature of each of the radiation monitors annually.

(2) An integrated leakage test of the confinement building shall be performed annually at a gauge pressure of at least 6.0 inches of water and a vacuum of at 36

least 2.0 inches of water, with a maximum allowable leak rate of 24 cfm/inch of water.

(3) Any additions, modifications, or maintenance to the confinement building or its penetrations shall be tested to verify that the building can maintain its required leak tightness.

(1) The confinement closure system is initiated either by a signal from the

. confinement building gaseous effluent radiation detectors or manually by the major scram switch and each of these signal sources is used to initiate the test. In addition, each radiation detector is tested for proper response to ionizing radiation.

(2) A preoperational test program was conducted to measure the representative leakage characteristics at values of a gauge pressure of +7.5 inches of water and

-2.5 inches of water. The specified test pressures and vacuums are acceptable because past tests have shown leakage rates to be linear with applied pressures and vacuums. *

(3) Changes in the building or its penetrations shall be verified to withstand specified test pressures; therefore, tests shall be performed before the building Confinement System can be considered to be operable.

4.5 Ventilation System Applicability: Normal and Emergency ventilation system Objective: To ensure the operability of the ventilation system.

Specifications (1) An operability test of the emergency exhaust system, including the building static pressure controller and the vacuum relief valve, shall be performed quarterly.

(2) An operability test of the controls in the Emergency Control Station and an

inspection to determine that all instruments in the Emergency Control Station are indicating normally shall be made monthly.

(3) The efficiency of the absolute filters in both normal and emergency exhaust systems shall be verified biennially. It shall be verified that the absolute filters remove 99% of particles with diameters of 0.3 µm and greater.

(4) It shall be verified biennially that the charcoal filter banks in the emergency exhaust and recirculation systems have a removal efficiency of 99% for Iodine.

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(I) The emergency ventilation system depends on the proper operation of the emergency exhaust system fans, valves, and filters, which are not routinely in service. Because they are not continuously used, their failure rate as a result of wear should be low. Since they are not being used continuously, their condition in standby shall be checked sufficiently often to ensure that they shall function properly when needed. An operability test of the active components of the emergency exhaust system quarterly will ensure that each component will be operable if an emergency condition should arise. The quarterly frequency is considered adequate since this system receives very little wear and since the automatic controls are backed up by manual controls.

(2) The Emergency Control Station instrumentation must be operable to monitor the reactor's condition in the event the Control Room becomes uninhabitable.

Therefore, monthly checks of the instrumentation have been shown to be adequate

, to ensure operability.

(3) The biennial verification of the absolute filter efficiency has been sho~n to be appropriate for filters subject to *continuous air flow. Because the absolute filters in the emergency exhaust system will be idle except during brief periods of fan operation; deterioration should be much less than for filters subjected to c*ontinuous air flow where dust overloading and air breakthrough are possible after long periods of use. Therefo.re, a biennial frequency should be adequate to detecting filter deterioration.

(4) Biennial verification of filter banks, which are subjected to flow only during brief periods of fan operation ensures that the filters will perform as analyzed.

4.6 . Emergency Power System Applicability: Emergency electrical power supply equipment Objective: To ensure emergency power for vital equipment after the reactor is shutdown.

Specifications (I) Each diesel generator shall be tested for automatic starting and operation quarterly.

(2) Should one of the diesel generators become inoperative, the operable generator shall be started monthly.

(3) All emergency power equipment shall be tested under a simulated complete loss of outside power annually.

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(4) The voltage and specific gravity of each cell of the Vented Lead Acid (VLA) battery shall be tested annually. A discharge test of the entire battery shall be performed once every 5 years.

(5) A discharge test of the Valve-Regulated Lead Acid (VRLA) batteries shall be performed once every two years.

(1) The NBSR is equipped with two diesel power generators, each capable of supplying full emergency load; therefore, only one of the generators shall be required. The diesel generators have proven to be very reliable over decades of service. The quarterly test frequencies are consistent with industry practice and are considered adequate to ensure continued reliable emergency power for emergency equipment.

(2) This testing frequency of the operable generator will ensure that at least one of the required emergency generators will be operable.

(3) An annual test of the emergency power equipment under a simulated complete loss of outside power will ensure the source will be available when needed.

(4) and (5), Specific gravity and voltage checks of individual cells are the accepted method of ensuring t.hat all cells of a VLA battery are in satisfactory condition. The annual frequency for these detailed checks is considered adequate to detect any significant changes in the ability of the battery to retain its ~barge. During initial installation, the station batteries were discharge tested to measure their capacity.

Experience has shown that repeating these tests at the specified intervals is adequate to detect deterioration of the cells and loss of battery capacity.

4.7 Radiation Monitoring System and Effluents

4. 7 .1 Monitoring System Applicability: Radiation monitoring equipment Objective: To operability of radiation monitors. Specifications (1) The gaseous effluent monitors for normal air, irradiated air and stack air shall be channel tested before startup, after a shutdown of longer than twenty-four (24) hours, or quarterly. Each of the above air monitors shall be channel calibrated annually.

(2) The fission products monitor shall be channel tested monthly and channel calibrated annually.

Amendment No.: 10 September 10, 2015 39

(3) The secondary coolant activity monitor shall be channel tested monthly and channel calibrated annually. Analysis of the secondary water for tritium shall be conducted monthly. Should the secondary cooling water activity monitor be inoperable, analysis for tritium shall be performed daily.

(4) The Area Radiation Monitors shall be channel tested monthly and channel calibrated annually.

(5) For primary tritium concentrations of less than or equal to 4 Ci/I, the primary water shall be sampled annually. For tritium concentrations of greater than 4 Ci/I, the primary water shall be sampled quarterly.

(1) A channel test ensures the monitoring systems. will respond correctly to an input signal. An annual channel calibration ensures the detection and response capability of the channels.

(2) A channel test monthly is consici'ered reasonable for a device of this type .

.A channel calibration annually is considered adequate to ensure that a significant deterioration in accuracy from its normal setting does not occur.

(3) The secondary cooling water activity monitor usually give~ the first

indication of a primary-to-secondary leak. This monitor employs a simple radiation detector, the operability of which has been shown to be very good. Therefore;*a monthly channel test is considered reasonable. An annual channel calibration frequency is considered adequate to ensure that a significant deterioration in accuracy from its normal settings does not occur. Assuming operation of the secondary cooling water activity monitor and no detectable loss of primary coolant, a monthly sampling for tritium should .be adequate to detect small tritium leaks. If the secondary cooling water activity monitor is out of service, then sampling is the primary nieans of leak detection and more frequent sampling is required.

A daily frequency is Judged adequate since large leaks would still be

.detected by a decreasing level in the D 20 storage tank.

(4) The area radiation monitors (ARM) may give the first indication of a radioactive release resulting from an experiment or reactor malfunction. A monitor employs a simple radiation detector, the operability of which has been shown to be very good over many years. Therefore, a monthly channel test is considered reasonable. These monitors* are primarily used to detect an increase in activity over that which has previously existed, so they are normally set at some reasonable value above background and their absolute accuracy is not critical. Hence, the annual calibration 40

frequency is considered adequate to ensure that a significant deterioration in accuracy does not occur.

(5) The primary tritium concentration can be carefully monitored by annual analysis of the primary water. All new water is tested prior to addition to the system. Operational experience and well established neutron activation principles provide a good basis for predicting tritium buildup in the primary. Increasing the sampling frequency after concentrations exceed 4 Ci/I will ensure that the tritium concentration limit is not exceeded.

4.7.2 Effluents Applicability: Environmental monitoring sampling program Objective: To minimize radiation exposures outside of the confinement building.

Specifications (1) Water, and soil or vegetation samples shall be collected quarterly with the exception that grass samples are collected during the growing season, April through September, and soil samples during the non-growing season, October through March.

(2) Thermo luminescent dosimeters shall be colle~ted quarterly.

(3) Air sampling shall be done quarterly.

(1) Collecting and analyzing the water, and soil or vegetation samples on a quarterly basis will provide information that environmental limits are not being exceeded.

(2) Collecting and analyzing the thermoluminescent dosimeters on a quarterly basis will provide information that radiation limits are not being exceeded.

(3) Sampling the air on a quarterly basis will provide information that release limits are not being exceeded.

4.8 Experiments Applicability: Irradiation Experiments

Objective: To ensure that experiments conform to the limits of the specifications of Section 3.8.

Amendment No. 9 September 24, 2013

Specification The reactivity worth of any experiment installed in a pneumatic transfer tube, or in any other NBSR irradiation facility inside the thermal shield shall be estimated before reactor operation with said experiment.

Estimation of the reactivity worth based either on calculation or on previous or similar measurements ensures that the experiment is within authorized reactivity limits.

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5.0 Design Features 5 .1 Site Description Specifications (1) The NBSR complex is located within the National Institute for Standards and Technology grounds and access to the reactor shall be controlled.

(2) The reactor shall have a minimum exclusion radius of 400 meters, as measured from the reactor stack.

The location and government ownership of the NBSR site ensures auxiliary services including fire and security are available. The exclusion radius of 400 meters is the distance on which all unrestricted doses are calculated. Should this value decrease for any reason, a recalculation of the unrestricted doses would be necessary. Access to the reactor complex is controlled either by the facility staff or by NIST Police.

5.2 Reactor Coolant System Specifications (1) The reactor coolant system shall consist of a reactor vessel and a single cooling loop containing heat exchangers, pumps, and valves.

(2) The reactor vessel shall be designed in accordance with Section VIII of the American Society of Mechanical Engineers (ASME) Code for Unfired Pressure Vessels. The vessel shall be designed for 50 psig and 250°F. The heat exchangers shall be designed for 100 psig and a temperature of l 50°F. The connecting piping shall be designed for 125 psig and a temperature of 150°F.

(1) The reactor coolant system has been described and analyzed as a single cooling loop system containing heat exchangers, pumps and valves.

(2) The design temperature and pressure of the reactor vessel and other primary system components provide adequate margins over operating temperatures and pressures. The reactor vessel was designed to Section VIII, 1959 Edition of the ASME Code for Unfired Pressure Vessels. Any subsequent changes to the vessel should be made in accordance with the most recent edition of this Code.

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5.3 Reactor Core and Fuel Specifications (1) The 20 MW reactor core consists of 30, 3.0 x 3.3 inch (7.6 x 8.4 cm) MTR curved plate-type fuel elements. The NBSR MTR-type fuel element shall be such that the central 7 inches of the fuel element contains no fuel. The middle 6 inches of the aluminum in the unfueled region of each plate shall have been removed.

(2) The side plates, unfueled outer plates, and end adaptor castings of the fuel element shall be aluminum alloy.

(3) The fuel plates shall be U 30 8 dispersed in a matrix of aluminum, clad in aluminum alloy.

(1) The neutronic and thermal hydraulic analysis was based on the use of 30 NBSR MTR-type thirty-four (34) plate* fuel elements. The NBSR fuel element has a 7 inch centrally located unfueled area, in the open lattice array. The middle 6 inches of aluminum in the unfueled region has been removed. The analysis requires that the fuel be loaded in a specific pattern. Significant changes in core loading patterns would require a recalculation of the power distribution to ensure that the CHFR would be within acceptable limits.

(2) and (3) The aluminum clad dispersion fuels used in the MTR fuel elements have a 50 year record of reliability at many research reactors.

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6.0 Administrative Controls 6.1 Organization The Director, NIST Center for. Neutron Research shall be the licensee for the NBSR.

The NBSR shall be under the direct control of the Chief, Reactor Operations and Engineering. The Chief, Reactor Operations and Engineering shall be accountable to the Director, NCNR for the safe operation and maintenance of the NBSR.

6.1.1 Structure The management for operation of the NBSR shall consist of the organizational structure as shown in Figure 6.1.

6.1.2 Responsibility Responsibility for the safe operation of the NBSR shall be with the chain of command established in Figure 6.1. Individuals at the various management levels shall be responsible for the policies and operation of the NBSR, for safeguarding the public and facility personnel from undue radiation exposures, and for adhering to all requirements of the operating license and technical specifications.

6.1.3 Staffing (1) The minimum staffing when the reactor is not secured shall be:

(a) A Reactor Operator in the Control Room.

(b) A Reactor Supervisor present within the reactor exclusion area.

( c) An SRO present in the facility whenever a reactor startup is performed, fuel is bein'g moved within the reactor vessel, experiments are being placed in the reactor vessel or a recovery from an unplanned or unscheduled shutdown or a significant power reduction.

(2) A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the Control Room. This list shall be updated annually. The list shall include:

_(a) Management personnel.

(b) Health Physics personnel.

(c) Reactor Operations personnel.

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6.1.4 Selection and Training of Personnel The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection and Training of Personnel for Research Reactors (ANSI/ANS 15.4-2007).

Qualification and requalification of licensed reactor operators shall be performed in accordance with a Nuclear Regulatory Commission (NRC) approved program.

6.1.4.1 Selection of Personnel Minimum educational and experience requirements for those individuals who have line responsibility and/or authority for the safe operation of the facility are as follows:

(1) Chief, Reactor Operations and Engineering The Chief, Reactor Operations arid Engineering shall have an advanced college degree in engineering or *a science related field, or equivalent experience and training. Equivalent experience for this position requires five years experience in a responsible position in reactor operations or reactor engineering, including one year experience in senior reactor facility management or superv1smn.

(2) Chief, Reactor Operations The Chief, Reactor Operations shall have a college degree in engineering or a science related fields or a combined seven years of college level education and nuclear reactor experience. Three years of reactor operations experience is required. The individual shall demonstrate the capability to be an SRO at the NBSR.

(3) Reactor Supervisor (a) Four years experience in reactor operations, including experience in the operation and maintenance of equipment and in the supervision of technicians and/or senior'reactor operators.

(b) A high school diploma or equivalent and formal training in reactor technology and reactor operations. An additional two years of experience may be substituted for education and formal training.

46

(c) Shall have been licensed as a Senior Reactor Operator at the NBSR.

(4) Senior Reactor Operator A Senior Reactor Operator shall have a high school diploma or equivalent and one year experience in reactor operations. The individual shall be licensed as a Senior Reactor Operator.

(5) Reactor Operator A Reactor Operator shall have a high school diploma or equivalent and six months of technical training. The individual shall be licensed as a Reactor Operato~.

(6) Auxiliary Operator An Auxiliary Operator shall have a high school diploma or equivalent.

6.1.4.2 Training of Personnel (1) A training program shall be established to maintain the overall proficiency of the Reactor Operations organization. This program shall include components for both initial licensing and requalification, consistent with ANSI/ANS 15.4-2007.

(2) The training program shall be under the direction of the Chief, Reactor Operations and/or the Chief, Reactor Operations and Engineering.

(3) Records of individual reactor operations staff members' qualifications, experience, training, and requalification shall be maintained as described in the specification of Section 6.8.2.

6.2 Review and Audit The NCNR Safety Evaluation Committee (SEC) is established to provide an independent review ofNCNR reactor operations to ensure the facility is operated and maintained in such a manner that the general public, facility personnel and property shall not be exposed to undue risk.

The NCNR Safety Assessment Committee (SAC) is established to provide an independent rev.iew or audit of NCNR reactor operations. This audit is to ensure that safety reviews and reactor operations are being performed in accordance with regulatory requirements and public safety is being maintained.

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6.2.1 Composition and Qualifications The Director, NCNR, upon recommendation of the Chief, Reactor Operations and Engineering, shall appoint all members and alternates to the SEC. The

. SEC shall be composed o.f no less than_four members and membership terms are indefinite and at the discretion of the Director. Members and alternates shall be selected on their ability to provide independent judgment and to collectively provide a broad spectrum of expertise in reactor technology and operation. At least two members shall be from the NCNR and one from Health Physics. Unless otherwise designated by the Director, the SEC shall include the following ex officio members: the Chief, Reactor Operations; Chief, Reactor Engineering; and the Senior Supervisory Health Physicist.

6.2.2 Safety Evaluation Committee Charter and Rules The SEC shall conduct its review functions in accordance with a written charter and the charter shall be consistent with ANSI/ANS 15 .1-2007. This charter shall include provisions for: .

(1) Meeting frequency.

(2) Voting rules.

(3) Quorums.

(4) Method of submission and content of presentation to. the committee.

(5) Use* of subcommittees ..

(6) Review, approval and dissemination of minutes.

6.2.3 SEC Review Function The responsibilities of the SEC, or a designated subcommittee thereof, shall include but are not limited to the following:

(1) Review proposed tests or experiments significantly different from any

previously reviewed or which involve any questions pursuant to 10 CFR 50.59 and determine whether proposed changes or reactor tests or experiments have been adequately evaluated, documented, approved and recommendations sent to the NCNR director for action.

(2) Review the circumstances of all events described in Section 6.7.2 and the measures taken to preclude a recurrence and provide recommendations to the NCNR director for action.

(3) Review proposed changes to the NBSR facility equipment or procedures when such changes have safety significance, or involve an amendment to the facility license, a change in the Technical Specifications incorporated in the facility license, or questions pursuant to 10 CFR 50.59 and provide recommendations to the NCNR director for action. Review SAC reports.

48

(4) The SEC shall on a biennial basis review its charter and recommend to the NCNR director any changes necessary to ensure the continued effectiveness of the charter.

6.2.4 SEC Audit Function The responsibility of the SEC, or a designated subcommittee thereof, shall include but not be limited to the following audits:

(1) Facility operations at a frequency of once per calendar year, not to exceed fifteen (15) months.

(2) Results of actions taken to correct deficiencies that affect reactor safety at a frequency of once per calendar year, not to exceed fifteen (15) months.

(3) Requi;tlification program at a frequency of once every other calendar year, not to exceed thirty (30) months.

(4) NBSR Emergency Plan at a frequency of once every other calendar year, not to exceed thirty (30) months.

6.2.5 Safety Assessment Committee (SAC)

The Safety Assessment Committee (SAC) shall be composed of at least three senior technical personnel who collectively provide a broad spectrum of expertise in reactor technology. The Committee members shall be appointed by the Director, NIST Center for Neutron Research. Members of the SAC shall not be regular employees of NIST. At least two members shall pass on any report or recommendation of the Committee. The SAC shall meet

annually and as required. The Committee shall review or audit the NCNR rei;tctor operations and the performance of the SEC. The SAC shall report in writing to the Director, NIST Center for Neutron Research.

6.3 Radiation Safety The NIST Reactor Health Physics Group shall be responsible to support the licensee in the implementation of the radiation protection and ALARA program at the reactor using the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities, ANSI/ANS 15.11-2004. The NIST Reactor Health Physics Group leader shall report to the Director, NIST Center for Neutron Research for radiological matters concerning the NBSR.

49

6.4 Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The safety significant changes (determined by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations) to operating procedures shall be reviewed by the SEC and approved by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations. Such reviews and approvals shall be documented in a timely manner. Activities requiring written procedures are:

(1) Startup, operation, and shutdown of the reactor.

(2) Fuel loading, unloading, and fuel movement within the reactor vessel.

(3) Surveillance checks, calibrations, inspections and maintenance of equipment required by the technical specifications that may have an effect on reactor safety.

(4) Personnel radiation protection, consistent with applicable regulations or guidelines. The procedures shall include management commitment and programs to maintain exposures and releases as low as is reasonably achievable in accordance with the guidelines of ANSl/ANS.15.11-2004.

(5) Conduct of irradiations and experiments that could affect reactor safety .or core reactivity .. *

(6) Implementation of required plans such as emergency or security plans.

(7) Use receipt, and transfer of byproduct material, if appropriate.

Substantive changes to the procedures listed above shall be made effective only after documented review by the SEC and approval by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations. Minor modifications or temporary deviations to the original procedures which do not effect reactor safety or change their original intent may be made by the Reactor Supervisor in order to deal with special or unusual circumstances or conditions. Such changes shall be .documented and reported ~ithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the next working day to the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations.

6.5 Experiment Review and Approval Experiments shall be carried out in accordance with established and approved procedures. The following provisions shall be implemented:

(1) All new experiments or class of experiments shall be reviewed by the SEC and approved in writing by the Director, NCNR.

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(2) Substantive changes to previously approved experiments shall be made only after review by the SEC and approved in writing by the Director, NCNR. Minor changes that do not significantly alter the experiment safety envelope may be made in accordance with the SEC charter.

6.6 Required Actions 6.6.1 Actions to Be Taken in the Event the Safety Limit is Exceeded (1) The reactor shall be shutdown and reactor operations shall not be resumed until authorized by the NRC.

(2) An immediate notification of the occurrence shall be made to the Chief, Reactor Operations and Engineering and the Chief, Reactor Operations.

The Chief, Reactor Operations and Engineering shall inform the NCNR director.

(3) Reports shall be made to the NRC in accordance with the specifications of Section 6.7.2. A written report shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. The report shall be prepared by the Chief, Reactor Operations and Engineering and submitted to the SEC for review. The SEC shall review the report and submit it to the Director, NIST Center for Neutron Research director for approval. The Director shall then submit the report to the NRC.

6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6. 7 .2 other than a Safety Limit Violation (1) The reactor shall be secured and the Chief, Reactor Operations and Engineering and the Chief, Reactor Operations notified.

(2) Operations shall not resume unless authorized by the Chief, Reactor Operations and Engineering.

(3)*The SEC shall review the occurrence at their next scheduled meeting.

(4) Where appropriate and in addition to the initial notification, a report shall be submitted to the NRC in accordance with the specifications of Section 6.7.2.

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6.7 Reports

6. 7 .1 Annual Operating Report A report shall be submitted annually to the NRC and include:

(1) A brief summary of operating experience including the energy produced by the reactor and the hours the reactor was critical.

(2) The number of unscheduled shutdowns, including reasons therefore.

(3) A tabulation of major preventative and corrective maintenance operations having safety significance.

(4) A brief description, including a summary of the safety evaluations, of changes in the facility or in procedures and oftest and experiments carried out pursuant to 10 CFR 50.59 (2007).

(5) A summary of the nature and amount of radioactive effluents released or discharged to the environs and the sewer beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.

(6) A summary of environmental surveys performed outside the facility.

(7) A summary of significant exposures received by facility personnel and visitors.

6.7.2 Special Reports In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made by the Director, NCNR or the Chief, Reactor Operations and Engineering, to the NRC as follows:

( 1) There shall be a report within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, facsimile, or other NRC approved method, to the NRC Operations Center and confirmed in writing by facsimile or similar conveyance, to be followed by a written report within 14 days that describes the circumstances associated with any of the following:

(a) Accidental release of radioactivity above applicable limits in unrestricted areas, whether or not the release resulted in property damage, personal injury, or exposure.

(b) Violation of the safety limit.

52

(c) Operation with a safety system setting for required systems less conservative than the Limiting Safety System Setting values.

(d) Operation in violation of a Limiting Condition for Operation (LCO) established in the technical specifications unless prompt remedial action is taken as permitted by exception statements.

(e) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required.

Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable.

(f) Any change in reactivity greater than one dollar ($1.00) that could adversely affect reactor safety.

(g) An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of conditions which could result in operation of the reactor outside the safety limit.

(h) Abnormal and significant degr~dation in reactor fuel, cladding, coolant boundary, or confinement boundar)r.(excluding minor leaks) where applicable.

(2) There shall be a report submitted in writing within 30 days to the NRC, Document Control Desk, Washington D.C. 20555, of:

(a) Permanent changes in the facility organization involving the Director, NCNR, or the Chief, Reactor Operations and Engineering.

(b) Significant changes in the accident analyses as described in the Safety Analysis Report.

6.8 Records Records of this section may be in the form of logs, data sheets, or other retrievable forms. The required information may be contained in single or multiple records, or a combination thereof.

6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less than Five Years Amendment No. 9 September 24, 2013

(1) Normal reactor operation logs, not including supporting documents such as checklists and log sheets. (Supporting documents shall be retained for a period of at least one year.)

(2) Principal maintenance activities.

(3) Special Reports.

(4) Surveillance activities required by these Technical Specifications.

(5) Solid radioactive wa.ste shipped off-site.

(6) Fuel inventories and transfers.

(7) Reactor facility radiation and contamination surveys where required by applicable regulations.

6.8.2

Records to be Retained for at Least One Operator Licensing Cycle Records of retraining and requalification of licensed operations personnel shall be maintained for the period the individual is employed or until the license is renewed.

6.8.3 Records to be Retained for the Life of the Reactor Facility Annual reports as described in the specifications of Section 6. 7 .1, to the extent the reports contain all of the required information, may be used as the record of the following:

(1) Gaseous and liquid radioactive effluents released to the environs.

(2) Off-site environmental monitoring surveys required by these Technical Specifications.

(3) Radiation exposure for all personnel monitored.

(4) Drawings of the reactor facility.

Amendment No. 9 September 24, 2013

Director, NIST Center for Neutron Research I

I Chief, Reactor Operations and Engineering Chief, Chief, Reactor Operations Reactor Engineering Reactor Supervisor NBSR Health Physics I f Senior Reactor Operator Reactor Operator Auxiliary Operator Safety Assessment Committee Safety Evaluation Committee (SAC) (SEC)

Recommendations and Technical Advice Administrative Reporting Channels Figure 6.1 55

. *~- ~

UNITED STATES NUCLEAR REGULATORY COMMISSION THE NATIONAL INSTITUTE OF STANDARDS AND TECHNOLOGY DOCKET NO. 50-184 NATIONAL BUREAU OF STANDARDS TEST REACTOR RENEWED FACILITY OPERATING LICENSE License No. TR-5

1. The U,S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for renewal of Facility Operating License No. TR-5 filed by the National Institute of Standards and Technology (the licensee) dated April 9,

2004, as supplemented on October 2, 2006, May 30 and August 14, 2007, September 16, October 21, and December 8, 2008, and March 3, March 19, and April 22, 2009 (the application}, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10, Chapter 1, of the Code of Federal Regulations (10 CFR);

B. Construction of the National Bureau of Standards test reactor (the facility) was completed in substantial conformity with Construction Permit No. CPTR-5 dated April 22, 1963, the provisions of the Act, and the rules and regulations of the Commission; C. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

D. There is reasonable assurance that (i) the activities authorized by this renewed license can be conducted at the designated location without endangering the health and safety of the public, and (ii) such activities will be conducted in compliance with the rules and regulations of the Commission; E. The licensee is technically and financially qualified to engage in the activities

. authorized by this license in accordance with the rules and regulations of the Commission; F. The licensee is a Federal agency which, in accordance with 10 CFR Part 140, is not required to furnish proof of financial protection. The licensee has executed an indemnity agreement which satisfies the requirements of 10 CFR Part 140; G. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; H. The issuance of this license is in accordance with 10 CFR Part 51,.

"Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions," of the Commission's regulations and all applicable requirements; and I. The receipt, possession and use of byproduct and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30 and 10 CFR Part 70.

2. Facility Operating License No. TR-5 is hereby renewed in its entirety to read as follows:

A This license applies to the National Bureau of Standards test reactor (the reactor) that is owned by the National Institute of Standards and Technology (NIST or the licensee), located on NIST's campus one mile southwest of Gaithersburg, Maryland, and de~cribed in the licensee's application, as supplemented.

8. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the National Institute of Standards and Technology:

1. Pursuant to subsection 104c of the Act, and Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," of the Code of Federal Regulations (10 CFR Part 50), to possess, use, and operate the reactor as a utilization facility at the designated location in accordance with the procedures and limitations described in the application and this license.

2. Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use in connection with operation of the reactor:

a. up to 45.0 kilograms of contained uranium-235 of any enrichment, provided that less than 5.0 kilograms of this amount be unirradiated;

b. to possess and use, but not to separate such special nuclear material as may be produced by operation of the reactor.

3. Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use in connection with the operation of the reactor: (1) a two-curie.

americium-beryllium neutron source which may be used for reactor start-up, and (2) up to a total of 8 curies of byproduct material (Atomic number 1 through 83) and up to 100 micro curies of americium-241, in the form of instrument calibration sources.

4. Pursuant to the Act and 10 CFR Part 30 to possess, use, and transfer but not to separate, except for byproduct material produced in non-fueled experiments, such byproduct.material as may be produced by operation of the reactor.

C. This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70, 73, and 100 of the Commission's regulations; is subject to all applicable provisions of the Act and rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

1. The licensee is authorized to operate the reactor at steady-state power levels up to a maximum of 20 megawatts (thermal).

2. The technical specifications contained in Appendix A, as revised by Amendment No. 9-11, are h~reby incorporated in the license. The licensee shall operate the reactor in accordance with the technical specifications.

3. The licensee shall maintain and fully implement all of the provisions of the Commission-approved physical security plan, including changes made pursuant to the authority of 10 CFR 50.54(p). The approved physical security plan consists of a National Institute of Standards and Technology document, withheld from public disclosure pursuant to 10 CFR 73.21, entitled, "NBSR Safeguards Plan," dated May 1983, transmitted by letter dated May 5, 1983.

Amendment No. 11 September??, 2017

D. This license is effective as of the date of issuance and shall expire at midnight twenty years from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RAJ Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A, Technical Specifications Date of Issuance: July 2, 2009

v t e Letter Sequence Request
TAC:MF9371, (Open)
Initiation Request, Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML17173A884
MONTHYEAR2017-03-29 [Table View]

ML17068A164

Text

Subject:

Reference:

Attachment:

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