Text

Responses Are Provided to Questions 9, 10, 11, 12, 15.1, 23.4, 24, and 25.5; Along with a Corrected Copy of the Proposed Technical Specifications (Chapter 14) of the SAR
	ML12251A231
Person / Time
Site:	U.S. Geological Survey
Issue date:	08/30/2012
From:	Debay T; US Dept of Interior, Geological Survey (USGS)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
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S-USGS science for a changing world Department of the Interior US Geological Survey PO Box 25046 MS 974 Denver, CO 80225-0046 August 30, 2012 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Reference:

U.S. Geological Survey TRIGA Reactor (GSTR), Docket 50-274, License R-1 13, Request for Additional Information (RAI) dated September 29, 2010

Subject:

Responses are provided to Questions 9, 10, 11, 12, 15.1, 23.4, 24, and 25.5; along with a corrected copy of the proposed Technical Specifications (Chapter 14) of the SAR Mr. Wertz:

Responses to the above questions are provided in the enclosed pages. This should complete our responses to the RAI dated September 29, 2010.

Sincerely, Tim DeBey USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct.

Executed on 8130/12 Attachment Copy to:

Betty Adrian, Reactor Administrator, MS 975 USGS Reactor Operations Committe Ac'Ic

9. NUREG-1537, Part 1 Section 4.5.1, "Normal Operating Conditions," requests a description of the limiting core configuration "that would yield the highest power density using the fuel specified for the reactor." the GSTR SAR provides a description of a typical core, but not the limiting core configuration applicable to GSTR. Please provide a description of the limiting core configuration for the GSTR. If the GSTR retains the dual LSSS of 100 kilowatts (kW) and 1.1 megawatts (MW), then limiting core configurations need to be specified for each LSSS. Please also provide tabulations of control rod worths and core excess reactivities for all core configurations.

Normal Core:

The normal core consists of 4 control rods (3 FFCR), 122 fuel elements, and 1 Central Thimble Irradiation facility which is filled with water.

Limiting Core:

The limiting core configuration is going to change from the description in the SAR section 4.5.1.2. We are removing the limit of 100 kW full power if there are fewer than 100 elements in the core. The new limiting core configuration will be 110 fuel elements (not including the fuel follower control rods). It will be a mixed core of 12, 8.5, and 8 wt% fuel elements, with new 12 wt% fuel elements occupying all B ring and C ring positions except for the two control rod positions. The 8 wt% aluminum clad fuel elements are restricted to the F and G rings. All operating limits for the limiting core will be the same as the normal core configuration. A new Technical Specification restriction will be that all operational cores must have a minimum of 110 fuel elements, excluding the FFCRs.

The normal and limiting configurations will have the same LSSS at 1.1 MW.

Note: The limiting core configuration values are calculated using the MCNP model.

Parameter Worth for limiting core configuration Excess reactivity $6.92 Transient rod $2.62 Shim I rod $2.40 Shim II rod $2.29 Regulating rod $4.25 Control rod total $11.56 Parameter Worth for current core configuration Excess reactivity $4.17 Transient rod $2.20 Shim I rod $1.96 Shim II rod $1.85 Regulating rod $3.10 Control rod total $9.11

10, NUREG-1 537, Part I Section 4.5.2, "Reactor Core Physics Parameters," requests a description of the neutronic methods and-a full set of parameters that are appropriate for use in the GSTR safety analysis. The GSTR SAR does not provide sufficient information. Please provide fuel and moderator temperature, void, and power coefficients; and power distribution estimates, The GSTR neutronic safety analysis was done using MCNP v 1.6 using the Endfb VII libraries. The specific libraries used for our analysis were created using temperature-specific cross sections. The program makexfs was used to create these cross section libraries, and the temperatures used for the libraries are in Table 1.

Table 1: Fuel temperatures used to create MCNP libraries for each ring Ring Fuel Temperature (°C)

B ring 395.97 C ring 373.59 D ring 272.64 E ring 244.78 F ring 204.75 G ring 180.89 Parameters for the limiting core analysis at 1.1 MW are as follows. Table 2 is an extension to Table 4.6 in the SAR section 4.5.4, giving parameters used for the limiting core configuration. The 8.5 wt% fuel used in the limiting configuration in the D-F rings is not modeled as new 8.5 wt% fuel. A fuel-to-cladding gap thickness of 0.1 mm is used in the thermal-hydraulic analysis.

Table 2: GSTR Core Loading for Limiting Core (Extension of Table 4.6 in SAR)

Core loading info Limiting Core Date: Aug, 2012 Core Excess $6.92 Reactivity keff 1.0484 A Ring Water B Ring 6 New 12wt%

C Ring 10 New 12wt%

1 AFCR 1 I FFCR D Ring 16 STD 2 FFCR E Ring 24 STD F Ring 30 STD G Ring 24 STD 12 Water 1 AFCR - Air-Followed Control Rod

The peak fuel temperature in the B-ring is 527 *C. The peak void fraction and power coefficients are -

0.069 $/void fraction and -0.00368 $/K, respectively. The power distribution estimates for each ring are in Table 3.

Table 3: Power distribution estimates for each ring Ring Power estimate (kW)

B ring 21.68 - 22.17 C ring 9.81 - 20.73 D ring 7.14 - 12.82 E ring 9.02 - 10.36 F ring 6.14 - 7.01 G ring 4.52 - 4.85

11. NUREG-1 537, Part 1 Section 4.5.3, "Operating Limits," requests information regarding the operating limits applicable to the limiting core configuration of the GSTR. The GSTR SAR provides only an upper limit on excess reactivity and a value for the shutdown margin.

11.1 Please describe any limits on excess reactivity components, such as those due to temperature variations; poisons (e.g., xenon and samarium); and experimental worths including pulse limitations.

11.2 Please provide limits on control rod worths and describe the manner for determining a shutdown margin, including a discussion of uncertainties.

11.3 Please describe the limits on core excess reactivity for the GSTR.

11.1 A 40-hour run of the reactor at full power was performed in 1972 when the reactor core was small (similar to the limiting core) and neutron flux and power peaking were higher than recent cores. This run showed a peak, equilibrium xenon worth of $2.30, with a peak, post-shutdown xenon worth of

$3.00. These values are approximately three times the values seen during normal full-power GSTR operations of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or less. Starting the reactor up and going to full power at the peak of the post-shutdown xenon will cause a reactivity transient of <1 cent every two minutes from xenon burnup. This reactivity transient is easily accommodated by normal control rod movement. Samarium transients are very minimal (<10 cents), occur on a very slow time scale, and have no effect on reactor operations.

Experiment worths are limited by Technical Specifications to $3.00 for any single, fixed experiment.

11.2 The transient rod has a limit of $4.15 as indicated in the GSTR SAR section 4.2.2. The other control rods do not have individual limits but are restricted in aggregate by the core excess reactivitylimit and the shutdown margin limit. To calculate the shutdown margin the reactor is brought to critical at a power level below 100 watts. Using the rod worth curves, the core excess and shutdown margin are calculated. The shutdown margin is calculated as the sum of all individual control rod reactivity worths that arneeded to make the core go critical.

Error in control rod measurements are +/-0.5 sec on timing during control rod calibrations, giving a maximum error of I cent reactivity for a control rod worth $4.50.

The error in control rod calculations from the MCNP model is +/-$0.016 for control rod worths.

11.3 The limit on core excess is set at $7.00, changing from the $6.80 in the SAR. The limiting core excess (as well as the normal core) is $6.92 (from model) (Tim would like to keep the $7 excess. With the model showing $6.92 is going to be the excess for the limiting core, then we will never go above $7 dollars in the near future. This argument might be used for keeping the excess limit at $7) -Tech Specs under SAR section 14.3.1.1.2 limit the core excess to $6.80, are we switching to $7.00 even though the model shows a limit of $6.92 for our limiting configuration?

12. NUREG-1537T Part 1 Section 4.6, "Thermal-Hydraulic Design," requests a description of therrnal-hvdraulic conditions in the GSTR to demonstrate that sufficient cooling capacity exists for steady-state and pulsed operating conditions. Please describe the DNBR analysis pertaining to the GSTR addressing both the steady-state and pulsed operations.

The thermal-hydraulic analyses were done using RELAP5 mod 3 with input from the results of the MCNP model for the power production factors. The highest single element power production factor of 22.17 kW was used as the hot channel to show the worst case scenario. RELAP5 generates heat flux results which can then be compared to Bernath's critical heat flux equations. RELAP5 gives a DNBR of 6.03 while the Bernath equations give a DNBR of 1.45. It is well known the DNBR calculated by the Bernath equations is conservative when compared to the other methods that can be used to calculate the DNBR.

Using these two numbers gives a range for the DNBR for the GSTR core. The RELAP5 model gives a peak fuel temperature of 526.37 0C, a peak cladding temperature of 134.32 0C, and a max coolant temperature of 91.6 0C. All of these values are well within safe operating parameters for the GSTR.

The pulsing operation analyses were done using RELAP5 mod 3, as in the steady state model. The following table was generated using the results from RELAP5.

Pulse Reactivity ($) 1.32 2 2.50 3 Peak Power (MW) 127.33 1224.93 2696.43 4583.52 Peak Fuel Temp (0C) 438.3 637.4 761.35 879.75 Void Fraction 0 0 - .13 (1.5 sec)

Peak Clad Temp (0C) 226.85 301.85 Peak Water Temp (°C) -37 ~51 Historical data from the GSTR indicate that these calculated values are very conservative.

15.1 GSTR SAR Subsection 13.2.1.2 states that the highest power density of 22 kW was used to determine the fuel element inventory. In GSTR SAR Subsection 4.5.1.2, the GSTR core was assumed to have 100 fuel rods. In Subsection 13.2.2.2.1, the total peaking factor (PF) is 3.85. In Subsection 13.2.2.2.2, the GSTR SAR states that the GSTR routinely operates with all but two or three grid positions occupied by the fuel element. The core has 125 fuel grid positions.

Therefore, there could be as many as 122 fuel rods. Therefore, the highest density rod appears to range between 31.6 and 38.5 kW (e.g., an average rod density of 8.20 to 10 kW [1,000/100 to122 rods] multiplied by a PF of 3.85).

Please describe the method and assumptions used to determine the highest power density of 22 kW, and the fission product inventory in Table 13.1. Also, please explain why the fission product release is limited to halogens and noble gases and does not include semi-volatile fission products such as cesium.

As stated in the answer to question 9, the option of having a core with fewer than 100 elements, with an associated power limit of 100 kW, has been removed from the relicensing request. The new limiting core will have no fewer than 110 fuel elements not including the FFCRs. The new limiting core has no defined peaking factor limit, no defined limit on power production per fuel element, and no reduced full power limit. The new limiting core MCNP model gives a peak power production of 22.17 kW at a total core power of 1100 kW, which is a maximum power peaking factor of 2.51. This maximum peaking factor occurs in a 12 wt% fuel element in the B-ring, as expected.

The rupture of a fuel element would result in the release of fission products. Fuel elements are rarely, if ever, removed from the pool water, so it is assumed that the damaged fuel element is submerged in the reactor pool at the time of the accidents, and only halogens and noble gases may be released. This type of event has been analyzed by F. C. Foushee and R. H. Peters, "Summary of TRIGA Fuel Fission Product Release Experiments", Gulf Energy and Environmental Systems report A-10801, 1971. Similar conclusions are reported by S. C. Hawley and R. L. Kathren, "Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors", NUREG/CR-2387, PNL-4028 (1982).

Configuration 23.4 There does not appear to be a SR for assuring that steady state power limit of 0.1 MW is not exceeded per LCO 14.3.1.3.

Question 23.4 This surveillance requirement is no longer necessary, because we have removed the steady state power limit of 0.1 MW if fewer than 100 fuel elements in the core.

24. ANSI/ANS-15.1-2007, Section 1.2.2, "Format," recommends that the Basis "provides the background or reason for the choice of specification(s), or references a particular portion of the SAR." The Bases listed below lacked sufficient information. Please provide additional information:

24.1 In the Basis for Section 14.2.1, GSTR SAR subsection 4.5.3.1 could not be found.

24.2 In the Basis for Section 14.2.2, GSTR SAR subsections 4.5.3.1 and 4.5.3.3, could not be found. The Basis also references 1.1 MW which differs from the licensed power level of 1.0 MW used throughout the GSTR SAR.

24.3 The Basis for Section 14.3.1.1.1 provides no reference or analysis to support the selection of $.55 as the shutdown margin..

24.4 The Basis for Section 14.3.1.1.2 contains information regarding shutdown margin, nominal rod worth, total rod worth, etc., but there is no reference provided to support the values cited.

24.5 The Basis for Section 14.3.1.2 provides a reference to GSTR SAR Section 13.2.2.2.1, "Maximum Reactivity Insertion" that is supposed to provide a discussion and basis for the statements regarding temperature response to reactivity insertion. However no basis for the statement could be found in that section of the GSTR SAR.

24.6 The Basis for Section 14.3.1.2 provides a reference to GSTR SAR Section 13.2.2.2.1, "Maximum Reactivity Insertion." However, no basis is provided to substantiate the temperature limits of 325 0C in aluminum-clad fuel and 800 0C in stainless-steel clad fuel.

24.7 In the Basis for Section 14.3.3, GSTR SAR subsection 4.5.3.1.1 could not be found. In addition, values are provided for the minimum height established for water above the top of the core (Specification 1) but no such discussion existed in either GSTR SAR Sections 4.4 or 11.1.1.1. There is no basis provided for the pH range selected in Specification 4; this range is outside of the NUREG-1537 recommended range.

24.8 In the Basis for Section 14.3.5 it is stated that the worst case TEDE is well below the 10 CFR Part 20 limit for individual members of the public for 2 scenarios -

the ventilation system is operating and the system is not operating; a reference to GSTR SAR Section 13.2.1 is provided. However, in the GSTR SAR section the only scenario evaluated is for the ventilation system in operation.

24.9 The 41 Basis for Section 14.3.7.2 provides that the 4.8x10-6 uCi/mI release limit for Ar out the stack would result in an annual TEDE < 5 mrem, which is 50% of the applicable limit (10 mrem). However, the referenced GSTR SAR section 11.1 .1.1. .1 did not provide any calculations that substantiate the basis.

24.10 In the Basis for Section 14.3.8.1, GSTR SAR references 7.2.3.1 and 13.2.2 are provided as the basis for establishing the reactivity worth of any single movable experiment e.g., less than $1.00. However there is no such discussion provided in either of the GSTR SAR sections. Additionally, the acceptability for the $3.00 and $5.00 limit on a single secure experiment and the total worth of all experiments, respectfully, is not provided.

24.11 In the Bases for Sections 14.4.1, 14.4.2, 14.4.3, 14.4.5, 14.4.7, and 14.4.8, there are no references provided to support the specifications and requirements stated.

24.12 In the Basis for Section 14.5.2, GSTR SAR subsection 4.5.3.3 could not be found.

24.13 In the Basis for Section 14.5.3.2, a reference is provided in GSTR SAR Section 4.2.2 to support the following statement, "The nuclear behavior of the air- or aluminum-follower, which may be incorporated into the transient rod, is similar to a void." However, no such information was identified in GSTR SAR Section 4.2.2.

24.1 The reference of 4.5.3.1 is being changed to 4.5.4.1.

24.2 The reference of 4.5.3.1 is being changed to 4.5.4.1, and the reference of 4.5.3.3 is being changed to 4.5.4.5.

24.3 The minimum shutdown margin is being changed from $0.55 to $0.30. This value is more conservative than other values which have been approved at other similar TRIGA facilities, such as Pennsylvania State University which has a $0.25 limit and the University of Texas which has a $0.286 limit. From the MCNP model of the limiting core configuration as described in SAR 4.5.1.2, the total control rod worth is

$11.56, and the excess reactivity is $6.92. Our limit on excess reactivity from SAR 14.3.1.1.2 is $7.00. The regulating rod is the most reactive rod, with a worth of $4.25. Therefore, the limiting core configuration has a minimum shutdown margin of $0.31. We have selected to specify a minimum shutdown margin of

$0.30 based on the MCNP analysis of the limiting core configuration.

24.4 The specifications and basis for SAR 14.3.1.1.2 will be rewritten as follows:

Specifications. The maximum available excess reactivity shall not exceed $7.00.

Basis. From SAR 14.3.8.1, the absolute reactivity worth of any single securedexperiment shall be less than

$3.00. For our nominal core configuration,it takes approximately$3.00 of reactivity insertion to reach full power due to the negative temperaturefeedback. A typical xenon-135 load would be approximately

$1.00. Assuming a -$3.00 experiment is in the core, a core excess reactivity of $7.00 would be required to operate the reactorat 1000 kW with a typicalxenon-135 load. From the MCNP limiting core configuration,we have calculateda core excess reactivity of $6.92. This value is within the $7.00 limit.

24.5 The reference to SAR 13.2.2.2.1 will be deleted, and we will add a sentence at the end of the first paragraph of the Basis for SAR 14.3.1.2 that reads as follows:

A discussion of the temperatureresponse to reactivity insertioncan be found in SAR 4.2.1.10.

24.6 We will change the second paragraph of the Basis for SAR 14.3.1.2 to read as follows:

It is shown in the MCNP modelfor the limiting core configuration that reactivityinsertions up to $3.00 in operationalcores will produce pulse transientswith maximum fuel temperaturesno greaterthan 419 °C in the F and G rings (aluminum clad fuel is restricted to these positions) and no greaterthan 984 °C in the B ring (hottest pointfor stainless steel cladfuel). These peak temperatureslastfor less than two seconds.

This maintainsa safety marginfor the temperature limits of thefuel set in SAR 13.1, allowingfor uncertaintiesin measurements and/orcalculations.

24.7 The Basis for SAR 14.3.3 will be rewritten to read as follows:

Basis. The minimum height of 16 feet of water above the top of the core guaranteesthat there is sufficient water to maintain suction on the primarypump for the primarycooling system. The intake for the primarycooling pump is located 5 feet below the top of the tank liner. This location is approximately 16 feet above the top of the reactor core. The bulk water temperature limit is necessary to ensure that the ion exchange resin does not undergo severe thermal degradation.Experience at many research reactorfacilities has shown that maintainingthe conductivity within the specified limit provides acceptablecontrol of corrosion (NUREG-1537). The minimum water level of no more than 24" below the top lip of the reactortank ensuressufficient cooling water during the design reactor tank leak of 350 gpm for the aluminum clad fuel to cool to safe levels after a reactor shutdown.

The basis for pH range selected in Specification 4 being outside the NUREG-1537 recommendation range is discussed in Amendment 11 of our current license. For reference, the applicable section of the license amendment is included as Attachment 1.

24.8 Analyses in SAR 13.2.1 of the Geological Survey TRIGA Reactor (GSTR) Safety Analysis Report (SAR) give the TEDE and CEDE to members of the public and members of the reactor staff during the maximum hypothetical accident (MHA). The MHA is analyzed with the ventilation system operating, as that is the most conservative case for estimating the dose to members of the public.

When the ventilation system is operating, the reactor bay is at a lower pressure compared to the surrounding areas, and the air inside the room is exhausted through the emergency exhaust at a rate of 800 cfm. When the ventilation system is not operating, the reactor bay is not at a lower pressure and the air inside the room leaves the room at an insignificant rate. The air from the reactor room would leak into adjacent reactor facility spaces first before reaching an exhaust point, due to the design of the

HVAC system. This would greatly decrease the amount of radioactive material that would be exhausted from the facility during the MHA due to plateout of the isotopes on facility surfaces.

Thus, the analysis in SAR 13.2.1 of the GSTR SAR is the most conservative estimate for the dose to members of the public during the MHA. The statement in the GSTR SAR SAR 14.3.5, section Basis, that states "the worst case TEDE is well below the 10 CFR Part 20 limit for individual members of the public for 2 scenarios - the ventilation system is operating and the system is not operating" is correct even though SAR 13.2.1 does not directly analyze the MHA while the ventilation system is not operating.

24.9 The Basis for SAR 14.3.7.2 will be modified to read as follows:

If 41Ar is continuously dischargedat 4.8 x 10.6 pCi/ml, measurements and calculationsshow that 41Ar released to the publicly accessible areasunder the worst-case weatherconditions would result in an annual TEDE of 0.5 mrem. This is only 5% of the applicablelimit of 10 mrem. The calculation was performed with the EnvironmentalProtectionAgency's Comply code.

The report from the COMPLY code for CY2011 is included below.

COMPLY: V1.6. 8/9/2012 8:18 40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants REPORT ON COMPLIANCE WITH THE CLEAN AIR ACT LIMITS FOR RADIONUCLIDE EMISSIONS FROM THE COMPLY CODE - V1.6.

Prepared by:

USGS GSTR PO Box 25046, DFC MS-974 Alex Buehrle 303-236-4726 Prepared for:

U.S. Environmental Protection Agency Office of Radiation and Indoor Air Washington, DC 20460

COMPLY: V1.6. 8/ 9/2012 8:18 Ar-41 release 4.8e-6 uCi/mI for 1 year SCREENING LEVEL 1 DATA ENTERED:

Effluent concentration limits used.

CONCENTRATION Nuclide (curies/cu m)

AR-41 4.80E-06 NOTES:

Input parameters outside the "normal" range:

None.

RESULTS:

You are emitting 706.0 times the allowable amount given in the concentration table.

- - Failed at level 1.

COMPLY: V1.6. 8/9/2012 8:18 Ar-41 release 4.8e-6 uCi/mi for 1 year SCREENING LEVEL 2 DATA ENTERED:

Release Rate Nuclide (curies/SECOND)

AR-41 2.266E-06 Release height 6 meters.

Building height 4 meters.

The source and receptor are not on the same building.

Distance from the source to the receptor is 350 meters.

Building width 30 meters.

Default mean wind speed used (2.0 m/sec).

NOTES:

Input parameters outside the "normal" range:

None.

RESULTS:

Effective dose equivalent: 0.5 mrem/yr.

- - Comply at level 2.

This facility is in COMPLIANCE.

It may or may not be EXEMPT from reporting to the EPA.

You may contact your regional EPA office for more information.

END OF COMPLIANCE REPORT

24.10 The Basis for SAR 14.3.8.1 will be changed to read as follows:

The worst event which could possibly arise is the sudden removal of a movable experiment immediately priorto, orfollowing, a pulse transientof the maximum licensed reactivity insertion. Limiting the worth of the movable experiment to less than $1.00 will assure that the additionalincrease of transientpower and temperatureis slow enough for the high power scram to be effective and, since this transientis not a super-promptpulse, we would not violate the 1 kW Pulse Interlock which prevents pulsing above 1 kW (SAR 14.3.2.3).

The worst event that is consideredin conjunction with a single secured experiment is the sudden removal of the experiment while the reactoris operating in a critical condition at a low power level. This is equivalent to pulse-mode operation of the reactor.Hence, the reactivity limitationfor a single secured experiment at $3.00 is the same as that of a maximum allowed pulse (SAR 13.2.2.2.1 and 14.3.1.2).

After further consideration of the $5.00 limit on the total worth of all experiments, it was determined that this limit is of no value. Therefore, Specification 3 will be removed.

24.11 A sentence will added directly below 14.4 Surveillance Requirements that reads as follows:

All bases for the following surveillancerequirements can be found in the operatingprocedures within the ROM or in the R-113 reactorlicense, Amendments 1 and 11. The approved operatingprocedures are periodicallyreviewed and approved by the ROC.

24.12 The reference of 4.5.3.3 is being changed to 4.5.4.5.

24.13 The Basis for SAR 14.5.3.2 will be modified to read as follows:

Basis. The poison requirementsfor the control rods are satisfied by using neutron absorbing borated graphite, B4C powder or boron as its compounds. These materialsmust be contained in a suitable clad materialsuch as aluminum or stainless steel to ensure mechanicalstability during movement and to isolate the poison from the tank waterenvironment. Control rods (thatarefuel-followed) provide additionalreactivity to the core and increase the worth of the control rod. The use of fueled-followers has the additionaladvantage of reducingflux peaking in the water-filled regions vacated by the withdrawalof the control rods. Scram capabilitiesareprovidedfor rapidinsertion of the control rods which is the primary safety feature of the reactor. The transientcontrol rod is designedfor rapid withdrawalfrom the reactorcore which results in a reactorpulse. The nuclearbehavior of the air-or aluminum-follower, which may be incorporatedinto the transientrod, is similarto a void. A more detailed description of the control rods and their propertiescan be found in SAR 4.2.2.

25.5 ANSI/ANS-15.1-2007, Section 6.6 "Required Actions" describes requirements pertaining to actions to be taken and circumstances when they apply. GSTR AC 14.6.5.1 specifies actions pertaining to safety system setting limit (should be LSSS) violations. However, the value cited was 1.0 MW, not 1.1 MW, or 100 kW depending on the core configuration as detailed in Section 14.2.2. Please explain.

25.5 The Specifications for Section 14.2.2 will be modified to remove the 100 element limit. It will be modified to read as follows:

Specifications.The limiting safety system setting shall be a steady state thermal power of 1.1 MW, and there shall be at least 110 fuel elements in the core (not includingfuel-followed control rods).

CHAPTER 14. TECHNICAL SPECIFICATIONS Chapter 14 TECHNICAL SPECIFICATIONS Included in this document are the Technical Specifications and the "Bases" for the Technical Specifications.

These bases, which provide the technical support for the individual Technical Specifications, are included for informational purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

14.1 Definitions Audit: A quantitative examination of records, procedures or other documents.

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

Channel Calibration: A channel calibration is an adjustment of the channel such that its output cor-responds 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 include a Channel Test.

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

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

Confinement: Confinement means an enclosure on the reactor bay which controls the movement of air into it and out through a controlled path.

Control Rod: A control rod is a device fabricated from neutron absorbing material or fuel which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may Rev. 8/12 14-1 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

Types of control rods shall include:

1. Regulating Rod (Reg Rod): The regulating rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled-follower section. Its position may be varied manually or by the servo-controller.

2. Shim Rod: A shim rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled-follower section.

3. Transient Rod: The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse. It may have a voided-follower.

Core Lattice Position: The core lattice position is defined by a particular hole in the top grid plate of the core. It is specified by a letter indicating the specific ring in the grid plate and a number indicating a particular position within that ring.

Excess Reactivity: Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (kerff = 1).

Experiment: Any operation, hardware, or target (excluding devices such as detectors) which is designed to investigate non-routine reactor characteristics or which is intended for irradiation within an irradiation facility. Hardware rigidly secured to a core or shield structure so as to be a part of their design to carry out experiments is not normally considered an experiment. Specific experiments shall include:

1. Secured Experiment: A secured experiment is any experiment or component of an experiment that is held in a stationary position relative to the reactor core by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment.

2. Movable Experiment: A movable experiment is one that is not secured.

Experiment Safety Systems: Experiment safety systems are those systems, including their associated input channel(s), which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated.

Fuel Element: A fuel element is a single TRIGA fuel rod.

Instrumented Element: An instrumented element is a special fuel element in which one or more ther-mocouples have been embedded for the purpose of measuring the fuel temperatures during reactor operation.

Measured Value: The measured value is the value of a parameter as it appears on the output of a channel.

Irradiation Facilities: Irradiation facilities shall mean vertical tubes, rotating specimen rack, pneumatic transfer system, sample holding dummy fuel elements and any other in-tank irradiation facilities.

USGS Safety Analysis Report 14-2 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS Operable: A system or component shall be considered operable when it is capable of performing its intended function.

Operating: Operating means a component or system is performing its intended function.

Pulse Mode: Pulse mode shall mean any operation of the reactor with the mode selector in the pulse position.

Reactivity Worth of an Experiment: The reactivity worth of an exi)erinent is the value of the reactiv-itxy change that results from the exleriineit being inserted into or removed from its intenided position.

Reactor Facility: The physical area defined by the area that contains the Denver Federal Center Building 15, between North Center Street, 1st Street, 2nd Street, and South Center Street.

Reactor Operating: The reactor is operating whenever it is not secured or shut down. Performance of routine subcritical surveillance is not considered operating.

Reactor Operator: An individual who is licensed to manipulate the controls of a reactor.

Reactor Safety Systems: Reactor safety systems are those systems, including their associated input chan-nels, which are designed to initiate, automatically or manually, a reactor scram for the primary purpose of protecting the reactor.

Reactor Secured: The reactor is secured when

1. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material presenit in the reactor to attain criticality under optimum available conditions of moderation and reflection;

2. Or the following conditions exist:

(a) The minimum number of neut ron-absorbing control devices is fully inserted or other safety devices are in shutdown position. as required by technical specifications; (b) The console key switch is in the off position, and the key is removed from the lock; (c) No work is in progress iivolving core fuel, core structure, installed control rods. or control rod drives unless they are physically (lecoupled from tile control rods; (d) No experiments are being moved or serviced that have. on movement, a reactivity worth exceeding the maximum value allowed for a single experiment, or one dollar, whichever is smaller.

Reactor Shutdown: The reactor is shut down if it is subcritical by at least one dollar in the reference core condition with the reactivity worth of all installed experiments included.

Reference core condition: The condition of the core when it is at ambient temperature (cold, 18 'C-25

'C) and the reactivity worth of xenon is negligible (< 0.30 dollars).

Review: A qualitative examination of records, procedures or other documents.

Safety Channel: A safety channel is a measuring channel in the reactor safety system.

Rev. 8/12 14-3 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS Scram time: Scram time is the elapsed time between the initiation of a scram and the instant that the control rod reaches its fully-inserted position.

Senior Reactor Operator: Al individual who is licensed to (direct the activities of reactor operators. Such an individual is also a reactor operator.

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

Shutdown Margin: Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems and will remain subcritical without further operator action, starting from any permissible operating condition with the most reactive rod is in its most reactive position.

Shutdown Reactivity: Shutdown reactivity is the measured reactivity with all control rods inserted. The value of shutdown reactivity includes the reactivity value of all installed experiments and is determined with the reactor at ambient conditions.

Square-Wave Mode (S.-W. Mode): The square-wave mode shall mean any operation of the reactor with the mode selector in the square-wave position.

Steady-State Mode (S.-S. Mode): Steady-state mode shall mean operation of the reactor with the mode selector in the manual or auto position.

Substantive Changes: Substantive changes are changes that would provide a significant decrease in the safety of an action or event.

Surveillance Intervals: Allowable surveillance intervals shall not exceed the following:

1. Biennial - interval not to exceed 30 months

2. Annual - interval not to exceed 15 months

3. Semi-annual - interval not to exceed 7.5 months.

4. Quarterly - interval not to exceed 4 months.

5. Monthly - interval not to exceed 6 weeks.

6. Weekly - interval not to exceed 10 days.

Unscheduled Shutdown: An unsc(hed(uled shutdo)wn is deti ne(l as any unplanned shutdowii of tile reactor caused by actuation of the reactor safety systeni. operator error. equipment malfunction, or a manual shut down in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.

USGS Safety Analysis Report 14-4 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.2 Safety Limits and Limiting Safety System Setting 14.2.1 Safety Limit-Fuel Element Temperature Applicability. This specification applies to the temperature of the reactor fuel.

Objective. The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding shall result.

Specifications.

1. The temperature in an aluminum-clad TRIGA fuel element shall not exceed 500 'C under any mode of operation.

2. The temperature in a stainless-steel clad TRIGA fuel element shall not exceed 1,150 'C (if cladding temperature is at or less than 500 'C) or 950 'C (if cladding temperature greater than 500 'C) under any mode of operation.

Basis. The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification especially since it can be measured. A loss of the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy.

The safety limit for the aluminum-clad TRIGA fuel element is based on data which indicate that the zirconium hydride will undergo a phase change at 535 'C. This phase change can cause severe distortion in the fuel element and possible cladding failure. Maintaining the fuel temperature below this level will prevent this potential mechanism for cladding failure.(SAR 4.5.4)

The safety limit for the stainless-steel clad TRIGA fuel is based on data including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1,150 'C (if cladding temperature is at or less than 500 °C) or 950 'C (if cladding temperature greater than 500 °C) (SAR 4.5.4.1).

14.2.2 Limited Safety System Setting Applicability. This specification applies to the scram settings which prevent the safety limit from being reached.

Objective. The objective is to prevent the safety limits from being reached.

Specifications. The limiting safety system setting shall be a steady state thermal power of 1.1 MW, and there shall be at least 110 fuel elements in the core (not including fuel-followed control rods).

Rev. 8/12 14-5 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS Basis. The limiting safety system setting is a total core thermal power, which, if exceeded shall cause the reactor safety system to initiate a reactor scram. This setting applies to all modes of operation. In steady-state operation up to 1.1 MW, ample margins exist between this setting and the safety limits of peak fuel temperature as specified in SAR 14.2.1, as long as the aluminum-clad fuel is restricted to the F and G rings of the core assembly (SAR 4.5.4.1). Thermal and hydraulic calculations indicate that stainless-steel clad TRIGA fuel may be safely operated up to power levels of at least 1.9 MW with natural convection cooling (SAR 4.5.4.5).

High fuel temperatures are experienced during pulse transients, initiated from low power. The peak fuel temperature reached from pulse operations is controlled by limiting the energy released in the "tail" of the pulse. The pulse tail duration is limited by rod drop time settings that are tested prior to each day of pulsing operations. The peak fuel temperature estimates are obtained from calculations based on an adiabatic reactor kinetics model. Results of this model are very conservative compared to the measured values (SAR 4.5.4.2).

14.3 Limiting Conditions of Operation 14.3.1 Reactor Core Parameters 14.3.1.1 Steady-state Operation 14.3.1.1.1 Shutdown Margin Applicability. These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. They apply for all modes of operation.

Objective. The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit shall not be exceeded.

Specifications. The reactor shall not be operated unless the following conditions exist: The shutdown margin provided by control rods shall be at least $0.30 with:

1. Irradiation facilities and experiments in place and the highest-worth, non-secured experiment in its most reactive state;

2. The most reactive control rod fully-withdrawn; and

3. The reactor in the reference corecondition where there is no 135Xe poison present and the core is at ambient temperature.

Basis. The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the most reactive control rod should remain in the fully-withdrawn position.

14.3.1.1.2 Core Excess Reactivity Applicability. This specification applies to the reactivity condition of the reactor and the reactivity worths USGS Safety Analysis Report 14-6 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS of control rods and experiments. It applies for all modes of operation.

Objective. The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit shall not be exceeded.

Specifications. The maximum available excess reactivity shall not exceed $7.00.

Basis. From SAR 14.3.8.1, the absolute reactivity worth of any single secured experiment shall be less than $3.00. For our nominal core conhigurat ion, it takes approximately $3.00 of reactivity insertion to reach full power due to the negative temperature feedback. A typical ':15Xe load would be approximnately $1.00.

Assuming a -$3.00 experiment is in the core, a core excess reactivity of $7.00 would be required to operate the reactor at, 1000 kW with a typical 1: 5 Xe load. From the MCNP analysis of the limiting core configiuration.

uv have calculated a core excess reactivity of $6.92. This value is within the $7.00 limit.

14.3.1.2 Pulse Mode Operation Applicability. This specification applies to the energy generated in the reactor as a result of a pulse insertion of reactivity.

Objective. The objective is to assure that the fuel temperature safety limit shall not be exceeded.

Specifications. The reactivity to be inserted for pulse operation shall be determined and limited by a mechanical stop on the transient rod, such that the reactivity insertion shall not exceed $3.00.

Basis. The fuel temperature rise during a pulse transient has been estimated conservatively by adiabatic models. These models accurately predict pulse characteristics for operation of TRIGA cores and should be accepted with confidence, relying also on information concerning prompt neutron lifetime and prompt tem-perature coefficient of reactivity. These parameters have been established for TRIGA cores by calculations and have been confirmed in part by measurements at existing facilities. In addition, the calculations rely on flux profiles and corresponding power densities which have been calculated for a variety of TRIGA cores.

A discussion of the temperature response to reactivity insertion can be fomnd in SAR 4.2.1.10.

It is shown in the RELAP model for the limitiig core configuration that reactivity insertions tip to $3.00 in operational cores will produce pulse transients with imaLximuin fuel t(einperatures no greater than 419 'C in the F and G rings (aluminum clad fuel is restricted to these positions) and no greater than 984 'C in the B ring (hottest point for stainless steel clad fuel). These peak teniperatures last for less than two seconds. This maintains a safety margin for the temperature limits of the fuel set in SAR 13.1. allowing for uncertainties in measurements and/or calculations.

14.3.1.3 Core Configuration Limitations Applicability. This specification applies to mixed cores of aluminum-clad and stainless-steel clad types of fuel.

Objective. The objective is to assure that the fuel temperature safety limit shall not be exceeded due to power peaking effects in a mixed core.

Specifications. Aluminum-clad fuel shall only be loaded in the F and G rings of the core. If ahitinium t ead Rev. 8/12 14-7 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS fu"el is presenti int the ear-e mid there ar~e less thfin 1001 fitel eletentit; int the eer-e (ineluditig 14uel follower- eeftroel rods). theni reatetor- power haltl he littited! to a stetody 4;ate limiit of 0.1 TWA.

Basis. The limitation of power peaking effects ensures that the fuel temperature safety limit shall not be exceeded in an operational core. Keeping aluminum-clad fuel in the F and G rings limits thoses fuel temperatures to safe values for aluminum-clad fuel (SAR 4.5.1.2).

14.3.1.4 Fuel Parameters Applicability. This specification applies to all fuel elements.

Objective. The objective is to maintain integrity of the fuel element cladding.

Specifications. The reactor shall not operate with damaged fuel elements, except for the purpose of locating damaged fuel elements. A fuel element shall be considered damaged and must be removed from the core if:

1. The transverse bend exceeds 0.0625 inches over the length of the cladding;

2. Its length exceeds its original length by 0.10 inch for stainless-steel clad fuel or 0.50 inch for aluminum-clad fuel;

3. A cladding defect exists as indicated by release of fission products; or

4. Visual inspection identifies significant bulges, pitting, or corrosion.

Basis. Gross failure or obvious, significant visual deterioration of the fuel is sufficient to warrant declaration of the fuel as damaged. The elongation and bend limits are the values found acceptable to the USNRC (NUREG-1537).

14.3.2 Reactor Control And Safety System 14.3.2.1 Control Rods Applicability. This specification applies to the function of the control rods.

Objective. The objective is to determine that the control rods are operable.

Specification. The reactor shall not be operated unless the control rods are operable.

Control rods shall not be considered operable if:

1. Physical damage is apparent to the rod or rod drive assembly and it does not respond to normal control rod motion signals; or

2. The scram time exceeds 1 second for the shim and regulating rods or 2 seconds for the transient rod.

Basis. This specification assures that the reactor shall be promptly shut down when a scram signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor (SAR 13.2.2.2.1).

USGS Safety Analysis Report 14-8 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.3.2.2 Reactor Measuring Channels Applicability. This specification applies to the information which shall be available to the Reactor Operator during reactor operation.

Objective. The objective is to specify the minimum number of power measuring channels that shall be available to the operator to assure safe operation of the reactor.

Specifications. The reactor shall not be operated in the specified mode unless the minimum number of power measuring channels listed in Table 14.1 are operable.

Table 14.1: Minimum Measuring Channels Measuring Channel Effective Mode S.-S. Pulse S.-W Power Level (NP1000 and NPP1000) 2 2 Nvt-Circuit - 1 -

Power level (NM1000) 1 --

Basis. The power level monitors assure that the reactor power level is adequately monitored for steady-state, square wave and pulse modes of operation (SAR 7.2.3.1). The specifications on reactor power level indication are included in this section, since the power level is directly related to the fuel temperature.

14.3.2.3 Reactor Safety System Applicability. This specification applies to the reactor safety system channels.

Objective. The objective is to specify the minimum number of reactor safety system channels that shall be available to the operator to assure safe operation of the reactor.

Specifications. The reactor shall not be operated unless the minimum number of safety channels described in Table 14.2 and interlocks described in Table 14.3 are operable.

Rev. 8/12 14-9 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS Table 14.2: Minimum Reactor Safety Channels Safety Channel Function Effective Mode S.-S. Pulse S.-W.

Power Level SCRAM 0 1.1 MW(t) or less 2 2 1

Preset timer Console Scram Button SCRAM (<15 sec)

SCRAM 1 1 1

I High Voltage SCRAM 9 loss of nominal operating 2 1 2 voltage to required power level channels Watchdog scrams Scram upon lack of response in DAC 2 2 2 or CSC computer (one scram circuit per computer)

Table 14.3: Minimum Interlocks Interlock Function Effective Mode S.-S. Pulse S.-W.

NM1000 Power Prevents control rod withdrawal 1 Level Channel @ less than 10-7% power Transient Rod Cylinder Prevents application of air 1 unless fully inserted 1 kW Pulse Interlock Prevents pulsing above 1 kW - 1 -

Shim, and Regulating Prevents simultaneous manual 1 1 Rod Drive Circuits withdrawal of two rods Shim, and Regulating Prevents movement of any rod - 1 -

Rod Drive Circuits except transient rod Basis. The power level scrams provide protection to assure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. The high voltage scram ensures that the required power measuring channels have sufficient high voltage as required for proper functioning of their power level scrams. The interlock to prevent startup of the reactor at count rates less than 10-7% power assures that the startup is not initiated unless a reliable indication of the neutron flux level in the reactor core is available. The interlock to prevent the initiation of a pulse above 1 kW is to assure that the magnitude of the pulse will not cause the fuel element temperature safety limits to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the steady-state mode. The interlock to prevent withdrawal of the shim, safety or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive period. The interlock to prevent simultaneous withdrawal of two control rods is to limit reactivity insertion rate from the standard control rods.

USGS Safety Analysis Report 14-10 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.3.3 Reactor Primary Tank Water Applicability. This specification applies to the primary water of the reactor tank.

Objective. The objective is to assure that there is an adequate amount of water in the reactor tank for fuel cooling and shielding purposes, and that the bulk temperature of the reactor tank water remains sufficiently low to guarantee ion exchanger resin integrity.

Specifications. The reactor primary water shall exhibit the following parameters:

1. The tank water level shall be at least 16 feet above the top of the core;

2. The bulk tank water temperature shall not exceed 60 °C;

3. The conductivity of the tank water shall be less than 5 Minhos/ici when averaged over a one month period;

4. The pH of the tank water shall be in the range of 4.5 to 7.5 if there is any aluminum-clad fuel in the core.

5. When aluminum-clad fuel is in the core, the reactor shall not be operated if the tank water level is more than 24" below the top lip of the reactor tank.

6. These specifications are not required to be met if the reactor core has been defueled.

Basis. The minimuni height of 16 feet of water above the top of the core guarantees that there is sufficient water to maintain suction on the primary I)ump for the primary cooling system. The intake for the primary cooling pump is located 5 feet below the top of the tank liner. This location is approximately 16 feet above the top of the reactor core. The bulk water temperature limit is necessary to ensure that the ion exchange resin does not undergo severe thermal degradation. Experience at many research reactor facilities has shown that maintaining the conductivity within the specified limit provides acceptable control of corrosion (NUREG-1537). The minimum water level of no more than 24" below the top lip of the reactor tank ensures sufficient cooling water during the design reactor tank leak of 350 gpin for the aluminunm clad fuel to cool to safe levels after a reactor shutdown.

14.3.4 This section intentionally left blank.

14.3.5 Ventilation System Applicability, This specification applies to the operation of the facility ventilation system.

Objective. The objective is to assure that the ventilation system shall be in operation to mitigate the consequences of possible releases of radioactive materials resulting from reactor operation.

Specifications.

1. The reactor shall not be operated unless the facility ventilation system is operating and the reactor bay pressure is maintained negative with respect to surrounding areas by at least 0.1" water pressure. This 14-11 USGS Safety Analysis Report 8/12 Rev. 8/12 14-11 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS does not apply to short periods of time (not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) for system troubleshooting, maintenance and movement of personnel or equipment through open doors.

2. The reactor bay ventilation system shall operate in the emergency mode, with all exhaust air passing through a HEPA filter, whenever a high level continuous air monitor alarm is present due to airborne particulate radionuclides in the reactor bay.

Basis. The worst-case maximum total effective dose equivalent is well below the 10 CFR 20 limit for individual members of the public. This has been shown to be true for scenarios where the ventilation system continues to operate during the MHA and where the ventilation system does not operate during the MHA.

(SAR 13.2.1). Therefore, operation of the reactor for short periods while the reactor bay underpressure is not maintained because of testing or reactor bay open doors, does not compromise the control over the release of radioactive material to the unrestricted area nor should it cause occupational doses that exceed those limits given in 10 CFR 20 (SAR 11.1.1.1.5). Moreover, radiation monitors in the building, independent of the ventilation system, will give warning of high levels of radiation that might occur during operation of the reactor (SAR 11.1.6).

14.3.6 This section intentionally left blank.

14.3.7 Radiation Monitoring Systems and Effluents 14.3.7.1 Radiation Monitoring Systems Applicability. This specification applies to the radiation monitoring information which must be available to the Reactor Operator during reactor operation.

Objective. The objective is to specify the minimum radiation monitoring channels that shall be available to the operator to assure safe operation of the reactor.

Specifications. The reactor shall not be operated unless the minimum number of radiation monitoring channels listed in Table 14.4 are operating, except for time periods of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for repair and mainte-nance, provided:

1. The ventilation system is operating; and

2. The continuous air particulate radiation monitor is operating.

Each channel shall have a readout in the control room and be capable of sounding an audible alarm which can be heard in the reactor control room.

Basis. The radiation monitors provide information to operating personnel regarding routine releases of ra-dioactivity and any impending or existing danger from radiation. Their operation will provide sufficient time to evacuate the facility or take the necessary steps to prevent the spread of radioactivity to the surroundings (SAR 11.1.6).

USGS Safety Analysis Report 14-12 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS Table 14.4: Minimum Radiation Monitoring Channels Radiation Monitoring Channels Number Continuous Air Particulate Radiation Monitor 1 Area Radiation Monitor 1 14.3.7.2 Effluents 4t Applicability. This specification applies to the release rate of Ar.

41 Objective. The objective is to ensure that the concentration of the Ar in the unrestricted areas shall be below the applicable effluent concentration value in 10 CFR 20.

Specifications. The annual average concentration of alAr discharged into the unrestricted area shall not exceed 4.8 x 10-6iuCi/ml at the point of discharge.

Basis. If '"Ar is continuously discharged at 4.8 x 10-6pCi/ml, measurements and calculations show that 4"Ar released to the publicly accessible areas under the worst-case weather conditions would result in an annual TEDE of 0.5 tirem. This is only 5% of the applicable limit of 10 torem. The calculation was performed with the Environmental Protection Agencvys Comply code.

14.3.8 Limitations on Experiments 14.3.8.1 Reactivity Limits Applicability. This specification applies to experiments installed in the reactor and its irradiation facilities.

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications. The reactor shall not be operated unless the following conditions governing experiments exist:

1. Movable experiments shall have absolute reactivity worth less than $1.00;

2. The absolute reactivity worth of any single secured experiment shall be less than $3.00.

3. Total absolute e*cp.. imet.t wet-tit of all expcril'int; sa.ll be les

. th.i: $.00.

Basis. The worst event which could possibly arise is the sudden removal of a movable experiment imme-diately prior to, or following, a pulse transient of the maximum licensed reactivity insertion. Limiting the worth of the movable experiment to less tlan $1.00 will assure that the additional increase of transient power and temperature is slow enough for the high power scram to be effective and, since this transient is not a super-prompt pulse. we would riot violate the I kW Pulse Interlock which prevents pulsing above I kW (SAR 14.3.2.3).

Rev. 8/12 14-13 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS The worst event that is considered in conjunction with a single secured experiment is the sudden removal of the experiment while the reactor is operating in a critical condition at a low power level. This is equivalent to pulse-mode operation of the reactor. Hence. tlie reactivity limitatioi for a single secured experiment at

$3.00 is the same as that of a maximum allowed pulse (SA1R 13.2.2.2.1 and 14.3.1.2).

14.3.8.2 Materials Applicability. This specification applies to experiments installed in the reactor and its irradiation facilities.

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications. The reactor shall not be operated unless the following conditions governing experiments exist:

1. Explosive materials, such as gunpowder, TNT, or nitroglycerin, in quantities greater than 25 milligrams shall not be irradiated in the reactor or irradiation facilities. Explosive materials in quantities less than or equal to 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container; and

2. Each fueled experiment shall be controlled such that the total inventory of 1311-1351 in the experiment 90 is no greater than 1.5 curies and the total inventory of Sr in the experiment is no greater than 5 millicuries.

Basis. This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials (SAR 13.2.6.2). The 1.5-curie limitation on 131 1 -135I, and the 5 millicurie limit on 9OSr, assure that in the event of a failure of a fueled-experiment involving total release of the iodine, the dose in the reactor bay and in the unrestricted area will be considerably less than that allowed by 10 CFR 20 (SAR 13.2.6).

14.3.8.3 Failures and Malfunctions Applicability. This specification applies to experiments installed in the reactor and its irradiation facilities.

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications. Where the possibility exists that the failure of an experiment (except fueled experiments) under normal operating conditions of the experiment or reactor, credible accident conditions in the reactor, or possible accident conditions in the experiment could release radioactive gases or aerosols to the reactor bay or the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor bay or the unrestricted area will not result in exceeding the applicable dose limits in 10 CFR 20, assuming that:

1. 100% of the gases or aerosols escape from the experiment; USGS Safety Analysis Report 14-14 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS

2. If the effluent from an irradiation facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape;

3. If the effluent from an irradiation facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of these aerosols can escape; and

4. For materials whose boiling point is above 130 'F and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, 10% of these vapors can escape.

Basis. This specification is intended to meet the purpose of 10 CFR 20 by reducing the likelihood that released airborne radioactivity to the reactor bay or unrestricted area surrounding the GSTR will result in exceeding the total dose limits to an individual as specified in 10 CFR 20.

14.3.9 This section intentionally left blank.

14.4 Surveillance Requirements All bases for the following surveillance requirements can be found in the operating procedures within the Reactor Operations Manual or in the R-113 reactor license, Amendments 1 and 11. The approved operating l)rocedures are periodically reviewed and reapproved by the ROC.

14.4.1 Reactor Core Parameters Applicability. This specification applies to the surveillance requirements for reactor core parameters.

Objective. The objective is to verify that the reactor does not exceed the authorized limits for power, shutdown margin, core excess reactivity, specifications for fuel element condition and verification of the total reactivity worth of each control rod.

Specifications.

1. A channel calibration shall be made of the power level monitoring channels by the calorimetric method semi-annually.

2. The total reactivity worth of each control rod shall be measured following any significant change in core or control rod configuration.

3. The shutdown reactivity shall be determined prior to each day's operation, prior to each operation extending more than one day, or following any significant change in core or control rod configuration.

4. The core excess reactivity shall be determined prior to each day's operation or following any significant change in core or control rod configuration.

5. All fuel elements shall be inspected for deterioration and measured for length and bend at five year intervals or every 500 pulses.

Rev. 8/12 14-15 USCS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS

6. The core shutdown margin shall be determined at an annual frequency.

7. These checks are not required if the reactor core has been defueled.

Basis. Experience has shown that the identified frequencies will ensure performance and operability for each of these systems or components.

14.4.2 Reactor Control and Safety Systems Applicability. This specification applies to the surveillance requirements of reactor control and safety systems.

Objective. The objective is to verify performance and operability of those systems and components which are directly related to reactor safety.

Specifications.

1. The control rods shall be visually inspected for damage or deterioration biennially.

2. The scram time shall be measured semi-annually.

3. The transient rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated as necessary, semi-annually.

4. A channel check of each of the reactor safety system channels for the intended mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day.

5. A channel test of each item in Table 14.2 and 14.3 in section 14.3.2.3, other than power measuring channels, shall be performed semi-annually.

6. These checks are not required if the reactor core has been defueled.

Basis. Experience has shown that the identified frequencies will ensure performance and operability for each of these systems or components.

14.4.3 Reactor Primary Tank Water Applicability. This specification applies to the surveillance requirements for the reactor tank water.

Objective. The objective is to assure that the reactor tank water level and the bulk water temperature monitoring systems are operating, and to verify appropriate alarm settings.

Specifications.

1. A channel check of the reactor tank water level alarm shall be performed semi-annually.

2. A channel check of the reactor tank water temperature alarm shall be performed quarterly. A channel calibration of the reactor tank water temperature system shall be performed annually.

USGS Safety Analysis Report 14-16 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS

3. The reactor tank water conductivity shall be measured monthly unless the reactor tank is drained.

4. The reactor tank water pH shall be measured quarterly if Al clad fuel is in the core, and the reactor tank is filled.

Basis. Experience has shown that the frequencies of checks on systems which monitor reactor primary water can adequately keep the tank water at the proper level and maintain water quality at such a level to minimize corrosion and maintain safety.

14.4.4 This section intentionally left blank.

14.4.5 Ventilation System Applicability. This specification applies to the reactor bay confinement ventilation system.

Objective. The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the unrestricted area.

Specifications.

1. A channel check of the reactor bay ventilation system's ability to maintain a negative pressure in the reactor bay withrespect to surrounding areas shall be performed prior to each day's operation or prior to each operation extending more than one day.

2. A channel check of the reactor bay ventilation system's ability to automatically switch to the emergency mode upon actuation of the CAM high alarm shall be performed quarterly.

Basis. Experience has demonstrated that checks of the ventilation system on the prescribed frequencies are sufficient to assure proper operation of the system and its control over releases of radioactive material.

14.4.6 This section intentionally left blank.

14.4.7 Radiation Monitoring System Applicability. This specification applies to the surveillance requirements for the area radiation monitoring equipment and the air monitoring systems.

Objective. The objective is to assure that the radiation monitoring equipment is operating properly and to verify the appropriate alarm settings.

Specifications.

1. A channel check of the radiation monitoring systems in section 14.3.7.1 shall be performed prior to each day's operation or prior to each operation extending more than one day.

2. A channel test of the continuous air particulate monitor shall be performed monthly.

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CHAPTER 14. TECHNICAL SPECIFICATIONS

3. A channel calibration of the radiation monitoring systems in section 14.3.7.1 shall be performed annu-ally.

Basis. Experience has shown that an annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span.

14.4.8 Experimental Limits Applicability. This specification applies to the surveillance requirements for experiments installed in the reactor and its irradiation facilities.

Objective. The objective is to prevent the conduct of experiments which may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure.

Specifications.

1. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before routine reactor operation with said experiment.

2. An experiment shall not be installed in the reactor or its irradiation facilities unless a safety analysis has been performed and reviewed for compliance with Section 14.3.8 by the Reactor Supervisor or Reactor Operations Committee in full accord with Section 14.6.2.3 of these Technical Specifications, and the procedures which are established for this purpose.

Basis. Experience has shown that experiments which are reviewed by the staff of the GSTR and the Reactor Operations Committee can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.

14.4.9 This section intentionally left blank.

14.5 Design Features 14.5.1 Site and Facility Description Applicability. This specification applies to the U.S. Geological Survey TRIGA Reactor site location and specific facility design features.

Objective. The objective is to specify the location of specific facility design features.

Specifications.

1. The restricted area is that area inside the fence surrounding the reactor building and the reactor building itself. The unrestricted area is that area outside the reactor building and the fence surrounding the reactor building.

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CHAPTER 14. TECHNICAL SPECIFICATIONS

2. Building 15 houses the TRIGA reactor and other research laboratories of the U.S. Geological Survey.

3. The reactor facility shall be equipped with ventilation systems designed to exhaust air or other gases from the reactor bay and release them from vertical level at least 21 feet above ground level.

4. Emergency controls for the ventilation systems shall be located in the reactor control room.

Basis. The reactor building and site description are strictly defined (SAR Chapter 2). The facility is designed such that the ventilation system will normally maintain a negative pressure in the reactor bay with respect to the outside atmosphere so that there will be no uncontrolled leakage to the unrestricted environment. Controls for normal and emergency operation of the ventilation system are located in the reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure to operating personnel (SAR 9.1 and 13.2.1).

14.5.2 Reactor Coolant System Applicability. This specification applies to the tank containing the reactor and to the cooling of the core by the tank water.

Objective. The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.

Specifications.

1. The reactor core shall be cooled by natural convective water flow.

2. The tank water inlet and outlet pipes to the heat exchanger and to the demineralizer shall be equipped with siphon breaks 14 feet above the top of the core or higher.

3. A tank water level alarm shall be provided to indicate loss of coolant prior to the tank level dropping 24 inches below the top lip of the tank.

4. A bulk tank water temperature alarm shall be provided to indicate high bulk water temperature prior to the temperature exceeding 60 'C.

5. These specifications are not required to be met if the reactor core has been defueled.

Basis.

1. This specification is based on thermal and hydraulic calculations which show that the TRIGA core can operate in a safe manner at power levels up to 1.9 MW with natural convection flow of the coolant water (SAR 4.5.4.5).

2. In the event of accidental siphoning of tank water through inlet and outlet pipes of the heat exchanger or demineralizer system, the tank water level will drop to a level no less than 14 feet from the top of the core (SAR 5.2).

Rev. 8/12 14-19 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS

3. Loss-of-coolant alarm caused by a water level drop to no more than 24 inches below the top lip of the tank provides a timely warning so that corrective action can be initiated. This alarm is located in the control room (SAR 5.2).

4. The bulk water temperature alarm provides warning so that corrective action can be initiated in a timely manner to protect the quality of the ion exchange resin. The alarm is located in the control room (SAR 7.2.3.2).

14.5.3 Reactor Core and Fuel 14.5.3.1 Reactor Core Applicability. This specification applies to the configuration of fuel and in-core experiments.

Objective. The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities shall not be produced.

Specifications.

1. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator elements positioned in the reactor grid plate.

2. The TRIGA core assembly may consist of stainless-steel clad fuel elements (8.5 to 12.0 wt% uranium),

aluminum-clad fuel elements (8.0 wt% uranium), or a combination thereof. All aluminum-clad fuel elements must be located in the F or G rings.

3. The fuel shall be arranged in a close-packed configuration except for single element positions occupied by in-core experiments, irradiation facilities, graphite dummies, aluminum dummies, stainless steel dummies, control rods, and startup sources. The core may also contain two separated experiment positions in the D through E rings, each occupying a maximum of three fuel element positions.

4. Core grid positions may be empty (water filled).

5. The reflector, excluding experiments and irradiation facilities, shall be graphite. water, or a combination of graphite and water. A reflector is not required if the core has been defueled.

Basis.

1. Standard TRIGA cores have been in use for years and their characteristics are well documented.

Analytic studies performed at GSTR for a variety of mixed fuel arrangements indicate that such cores with mixed loadings would safely satisfy all operational requirements (SAR 4.2).

2. The core will be assembled in the reactor grid plate which is located in a tank of light water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of irradiation facility radiation requirements (SAR 4.2).

USGS Safety Analysis Report 14-20 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.5.3.2 Control Rods Applicability. This specification applies to the control rods used in the reactor core.

Objective. The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications.

1. The shim and regulating control rods shall have scram capability and contain borated graphite, B 4 C powder or boron, with its compounds in solid form as a poison, in aluminum or stainless steel cladding.

These rods may incorporate fueled followers.

2. The transient control rod shall have scram capability and contain borated graphite or boron, with its compounds in a solid form as a poison in an aluminum or stainless steel cladding. The transient rod drive mechanism shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an aluminum-or air-follower.

Basis. The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B 4 C powder or boron with its conmpounds in a solid form. These materials must be contained in a suitable clad nmaterial such as aluminum or stainless steel to ensure mechanical stability during movement and to isolate the poison from the tank water enviromnent. Control rods (that are fuel-followed) provide additional reactivity to the core and increase the worth of the control rod. The use of fueled-followers has the additional advantage of reducing flux peaking in the water-filled regions vacated by the withdrawal of the control rods. Scram capabilities are provided for rapid insertion of the control rods which is the primary safety feature of the reactor. The transient, control rod is designed for rapid withdrawal from the reactor core which results in a reactor pulse. The nuclear behavior of the air- or aluminum-follower, which may be incorporated into the transient rod. is similar to a void. A more detailed description of the control rods and their properties call be found in SAR. 4.2.2.

14.5.3.3 Reactor Fuel Applicability. This specification applies to the fuel elements used in the reactor core.

Objective. The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications.

1. Aluminum-clad TRIGA fuel. The individual unirradiated aluminum-clad fuel elements shall have the following characteristics:

235 (a) Uranium content: nominally 8.0 wt% enriched to a nominal 20% U; (b) Hydrogen-to-zirconium atom ratio nominally 1 to 1; and Rev. 8/12 14-21 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS (c) Cladding is aluminum of a nominal 0.030 inch thickness.

2. Stainless-steel clad TRIGA fuel. The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

235 (a) Uranium content: nominal range of of 8.5 to 12.0 wt% enriched to a nominal 20% U; (b) Hydrogen-to zirconium atom ratio nominally between 1.6 to 1 and 1.7 to 1; and (c) Cladding: 304 stainless steel, nominal 0.020 inches thick.

Basis.

1. A nominal uranium content of 8 wt% in an aluminum-clad TRIGA element is less than the traditional stainless-steel clad element design value of 8.5 wt%. Such an decrease gives a lower power density. The nominal hydrogen-to-zirconium ratio of 1.0 to 1 could result in a phase change of the ZrH if fuel temperature is allowed to exceed 535 'C. Although this would not necessarily cause a rupture of the fuel cladding, it would cause distortion and stressing of the cladding.

2. A maximum nominal uranium content of 12 wt% in a standard TRIGA element is about 50% greater than the lower-loaded nominal value of 8.5 wt%. Such an increase in loading would result in an increase in power density of less than 50%. An increase in local power density of 50% reduces the safety margin by, at most, 10%. The maximum hydrogen-to-zirconium ratio of 1.7 to 1 could result in a maximum stress under accident conditions to the fuel element cladding of about a factor of 1.5 greater than the value resulting from a hydrogen-to-zirconium ratio of 1.60. However, this increase in the cladding stress during an accident would not exceed the rupture strength of the cladding.

14.5.4 Fuel Storage Applicability. This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective. The objective is to assure that fuel which is being stored shall not become critical and shall not reach an unsafe temperature.

Specifications.

1. All fuel elements shall be stored in a geometrical array where the k-effective is less than 0.9 for all conditions of moderation.

2. Irradiated fuel elements and fuel devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the temperature of the fuel element or fueled device will not exceed design values.

Basis. The limits imposed are conservative and assure safe storage (NUREG-1537).

USGS Safety Analysis Report 14-22 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.6 Administrative Controls 14.6.1 Organization Individuals at the various management levels, in addition to being responsible for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license, technical specifications, and federal regulations. The minimum qualification for all members of the reactor operating staff shall be in accordance with ANSI/ANS 15.4, "Standard for the Selection and Training of Personnel for Research Reactors."

14.6.1.1 Structure The reactor administration shall be related to the USGS and USNRC structure as shown in Figure 14.1.

14.6.1.2 Responsibility The following specific organizational levels, and responsibilities shall exist:

1. USGS Director (Level 1): The Director is accountable for ensuring that all regulatory requirements, including implementation and enforcement, are in accordance with all requirements of the USNRC and the Code of Federal Regulations.

2. Reactor Administrator (Level 2): The Reactor Administrator is responsible to the Director and is responsible for guidance, oversight, and management support of reactor operations.

3. Reactor Supervisor (Level 3): The Reactor Supervisor reports to the Reactor Administrator and is responsible for directing the activities of the Reactor Operators and Senior Reactor Operators and for the day-to-day operation and maintenance of the reactor.

4. Reactor Operator and Senior Reactor Operator (Level 4): The Reactor Operator and Senior Reactor Operator report to the Reactor Supervisor and are primarily involved in the manipulation of reactor controls, monitoring of instrumentation, and operation and maintenance of reactor related equipment.

14.6.1.3 Staffing

1. The minimum staffing when the reactor is operating shall be:

(a) A Licensed Operator in the control room; (b) A second facility staff person present or on call; and on call means an individual who:

i. Can be reached by an available communication method within 5 minutes and ii. Is capable of getting to the reactor facility within 15 minutes under normal conditions.

Rev. 8/12 14-23 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS Line of Commnitiication ............

Line of Responsibility Figure 14.1: Administrative Structure (c) A SRO shall be reachable by any communication method and capable of getting to the reactor facility within 15 minutes under normal conditions.

2. Events requiring the direction of a Senior Reactor Operator (a) Initial approach to critical for each day's first critical operation; (b) Reactor start-up and approach to power; (c) All fuel or control-rod movements within the reactor core region; (d) Relocation of any in-core components (other than normal control rod movements) or irradiation facility with a reactivity worth greater than one dollar; (e) Recovery from unplanned or unscheduled shutdown or an unscheduled significant power reduction; and USGS Safety Analysis Report 14-24 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.6.1.4 Selection and Training of Personnel The selection, training and requalification of operations personnel shall be in accordance with ANSI/ANS 15.4, "Standard for the Selection and Training of Personnel for Research Reactors."

14.6.2 Review And Audit The ROC shall have primary responsibility for review and audit of the safety aspects of reactor facility operations.

14.6.2.1 Composition and Qualifications The Reactor Operations Committee shall be composed of at least four voting members, including the Chair-man. All members of the Committee shall be knowledgeable in subject matter related to reactor operations.

To expedite Committee business, a Committee Chairman has been appointed. The Chairman of the Reactor Operations Committee is listed by name on the Reactor Operations Committee roster.

The Committee is appointed by the Director, U.S. Geological Survey. No definite term of service is specified; but should a vacancy occur in the Committee, the Director will appoint a replacement. The remaining members of the Committee will be available to assist the Director in the selection of new members. The Reactor Supervisor and the Radiation Safety Officer are ex officio members of the Committee and the Reactor Supervisor is the only non-voting member of the Committee.

14.6.2.2 Charter and Rules The Reactor Operations Committee consists of USGS members and non-USGS members, and the Committee must meet at least semi-annually.

Criteria have been established for the conduct of the meetings and a charter for the Committee is written in the USGS Survey Manual.

A quorum for review, audit, and approval purposes shall consist of not less than one-half of the committee membership, provided that the operating staff does not constitute a majority of the committee membership.

The Chairperson or an alternate must be present at all meetings in which the official business of the committee is being conducted. Approvalst by the committee shall require an affirmative vote by a majority of the non-Survey members present and an affirmative vote by a majority of the Survey members present.

14.6.2.3 Review and Audit Function Semiannual meetings will be held to review and audit reactor operations. The following items shall be reviewed:

t "Approval" as used in this section shall mean approval for substantive issues, including Class II Experiments, 10 CFR 50.59 changes, license amendments or any other matters designated by the Reactor Administrator, the Reactor Supervisor, or the Reactor Operations Committee. Minor issues will continue to be handled by the reactor staff and the Reactor Supervisor as described elsewhere in this Reactor Operations Manual.

Rev. 8/12 14-25 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS

1. Determinations that proposed changes in equipment. systems. test, experiments, or procedures are allowed without prior authorization by the responsible authority, for example, 10 CFR 50.59 or 10 CFR 830;

2. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance;

3. All new experiments or classes of experinments that could affect reactivity or result in tile release of radioactivity;

4. Proposed changes in technical specifications, license, or charter;

5. Violations of technical specifications. license, or charter. Violations of internal procedures or instruc-tions having safety significance:

6. Operating abnormalities having safety significance;

7. Reportable occurrences listed in 14.6.6.2: and

8. Audit reports.

A written report or miiutes of the findings and re*onutieidations of the review group shall be submitted to the Reactor Administrator and the review and audit group memlbers in a timely manner after the review has been completed. Tile audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Discussions with cognizant personnel and observation of operations should be used also as appropriate. In no case shall the individual inunediately responsible for the area perform an audit in that area. The following items shall be audited:

1. Facility operations for confirmance to the technical specifications and applicable license or charter conditions: at least once per calendar year (interval between audits not to exceed 15 months);

2. The retraining and requalification program for the operating staff: at least once every other calendar year (interval between audits not to exceed 30 months);

3. The results of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety: Lt least once per calendar year (interval between audits not to exceed 15 months):

4. The reactor facility emergency plan and implementing procedures: at least once every other calendar year (interval between audits not to exceed 30 months).

Deficiencies uncovered that affect reactor safety shall inmmediately be reported to the Reactor Administrator.

A written report of the findings of lie audit shall be submitted to the Reactor Administrator and the review and audit group members within 3 months after the audit has been completed.

These meetings will also include inspections of the reactor facility and audits of reactor records by the Committee. New Class I Experiments approved by the Reactor Supervisor will be discussed. Any new Class II Experiments pending, or any business which requires Committee approval will be handled in the following manner.

USGS Safety Analysis Report 14-26 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS

1. The Reactor Supervisor will request a meeting of Survey members.

2. If the experiment is approved by them, the Reactor Supervisor will communicate the recommendations of the local group, along with the proposed experimental details to Non-Survey Members. The Reactor Supervisor shall provide adequate detail, either verbal, or in writing to the out-of-town members to allow an informed decision. The decisions, with comments and discussions, shall be documented for future review and audit purposes.

3. Approval to proceed will be given by the Reactor Operations Committee Chairman, who will sign the experiment application document to indicate Committee approval.

4. If the experiment cannot be approved in this manner, the experiment application document will be distributed to all Committee members for further discussion and will be placed on the agenda of the next full Committee meeting. Normally the non-Survey members will attend only every other semi-annual meeting. These full Committee meetings are generally scheduled each spring. Special meetings for the review of specific problems or urgent Class II Experiments will be called whenever necessary.

A Quarterly Report will be prepared by the Reactor Supervisor for the members of the Committee. This report will be in sufficient detail to allow members of the Committee to review the safety standards asso-ciated with the operation and use of the reactor facility. Each Quarterly Report will include a synopsis of all experiments approved during the period to insure that the intent and function of the Committee, as mentioned in the Technical Specifications is being maintained.

14.6.2.4 Inactive Member Status Committee members will be considered on inactive status should prolonged absences from their normal business address not allow them to participate in routine Committee business. The Committee Chairperson is responsible to insure the function of the Committee is not diluted so as to be ineffective by such actions and will recommend the appointments of alternates if the need arises.

The Reactor Supervisor will insure that any members who are unable to attend a Reactor Operations Committee meeting, or are on inactive status due to other commitments, will be kept informed of all Committee business.

14.6.3 Radiation Safety The Reactor Supervisor, in coordination with the Reactor Health Physicist, shall be responsible for imple-mentation of the radiation safety program. The requirements of the radiation safety program are established in 10 CFR 20. The program should use the guidelines of the ANSI/ANS 15.11, "Radiation Protection at Research Reactor Facilities."

Rev. 8/12 14-27 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS 14.6.4 Procedures Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgment and action should the situation require such. Procedures shall be in effect for the following items:

1. Performing experiments and maintenance;

2. Startup, operation and shutdown of the reactor;

3. Emergency situations;

4. Core changes and fuel movement;

5. Control rod removal and replacement;

6. Performing maintenance which may affect reactor safety;

7. Administrative controls;

8. Power calibration; and

9. Radiation protection.

10. Use, receipt and transfer of by-product material, if appropriate.

Substantive changes to the above procedure shall be made only with the approval of the ROC. Except for radiation protection procedures, unsubstantive changes shall be approved prior to implementation by the Reactor Supervisor and documented by the Reactor Supervisor within 90 days of implementation. Unsub-stantive changes to radiation protection procedures shall be approved and documented by the Reactor Health Physicist within 90 days of implementation.

14.6.5 Experiment Review Administrative requirements are in place at the GSTR to assure that all experiments are performed in a manner which will ensure the protection of the public. Experiment review meets the requirements of Regulaiory Guide 2.2, and Standard ANSI N401-1974 (ANS-15.6) as modified by Regulatory Guide 2.4. All experiments proposed for the reactor will be either (Class I or Class II experiments. The classification of the proposed experiments will be the responsibility of the Reactor Supervisor. Class I experiments include all experiments that have been run previously or that are minor modifications to a previous experiment. These are experiments which involve small changes in reactivity, no external shielding changes, and/or limited amounts of radioisotope production. The Reactor Supervisor has the authority to approve the following:

1. Experiments for which there exists adequate precedence for assurance of safety.

2. Experiments which represent less than that anmount of reactivity worth necessary for prompt criticality.

Rev. 8/12 14-28 USGS Safety USGS Analysis Report Safety Analysis Report 14-28 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS

3. Experiments iii which any significant reactivity worth is stable and mechanically fixed, that is, securely fastened or bolted to the reactor structure.

Class II experiments include all new experiments and major modifications of previous experiments. These experiments must be reviewed and approved by the Reavtor Operations Committee before being run. The Radiation Safety Committee may also be consulted. These experiments may involve larger changes in reactivity, external shielding changes. and/or larger amounts of radioisotope production. These include:

1. In-core experiments which involve, in an unstable form, reactivity worth greater than that necessary to produce a prompt critical condition in the reactor core.

2. Experiments involving corrosive chemicals, pressures or temperatures which, if failure should occur.

could endanger the safety of the reactor core.

3. Dynamic experiments which could introduce appreciable reactivity worth into the reactor by failure or malfunction. Included in this group are circulation systems which operate in or at, the core and by which if a failure occurred, the core could be damaged.

4. Experiments which are dynamically coupled to the reactor core and together function as a system, i.e.

to measure nuclear absorption cross sections, or study transient responses.

5. Experiments which interfere in any way with the normal function of any of the reactor safety circuits.

6. Experiments which could produce radiation levels sufficient to cause serious personnel radiation injury.

7. Experiments which by their unusual hazard could produce injury or death.

The radioisotopes produced at the GSTR may only be transferred to licensed users. Individuals associated with the U.S. Geological Survey may be approved to receive radioactive material under the authority of the USGS license by the Radiation Safety Committee. Other users must have a current Radioactive Material License. This information is verified during the approval of the experiment, prior to performance of every experiment.

14.6.6 Required Actions 14.6.6.1 Actions to Be Taken in Case of Safety System Setting Violation In the event a safety system setting limit (steady state power level of 1.1 MW) is exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the Reactor Operations Committee;

2. An immediate notification of the occurrence shall be made to the Reactor Administrator, ROC Chair-person; and Rev. 8/12 14-29 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS

3. Reports shall be made to the USNRC in accordance with Section 14.6.6.2 of these Technical Specifica-tions. The written report (required within 14 days) 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. This report shall be submitted to the ROC for review and submitted to the NRC when authorization is sought to resume operation of the reactor.

14.6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 14.6.6.2 Other than a Safety System Setting Violation For all events which are required by regulations or Technical Specifications to be reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months under Section 14.6.6.2, except a safety system setting violation, the following actions shall be taken:

1. The reactor shall be secured and the Reactor Supervisor notified;

2. Operations shall not resume unless authorized by the Reactor Supervisor;

3. The Reactor Operations Committee shall review the occurrence at their next scheduled meeting; and

4. Where appropriate, a report shall be submitted to the NRC in accordance with Section 14.6.6.2 of these Technical Specifications.

14.6.7 Reports 14.6.7.1 Annual Operating Report An annual report covering the previous calendar year shall be created and submitted by the Reactor Super-visor to the USNRC consisting of:

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 unplanned 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 of tests and experiments carried out pursuant to 10 CFR 50.59;

5. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25 percent of the concentration allowed or recommended, a statement to this effect is sufficient;

6. A summarized result of environmental surveys performed outside the facility; and USGS Safety Analysis Report 14-30 Rev. 8/12

CHAPTER 14. TECHNICAL SPECIFICATIONS

7. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25 percent of that allowed.

14.6.7.2 Special Reports In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made by the Reactor Supervisor to the NRC as follows:

1. A report within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone or fax to the NRC Operations Center followed by a written report within 14 days that describes the circumstances associated with any of the following:

(a) Any accidental release of radioactivity above applicable limits in unrestricted areas, whether or not the release resulted in property damage, personal injury, or exposure; (b) Any violation of a safety limit; (c) Operation with a safety system setting less conservative than specified in the Technical Specifica-tions.;

(d) Operation in violation of a Limiting Condition for Operation; (e) Failure of a required reactor safety system component which could render the system incapable of performing its intended safety function unless the failure is discovered during maintenance tests or periods of reactor shutdown; (f) An unanticipated or uncontrolled change in reactivity greater than $1.00; (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 a condition which could result in operation of the reactor outside the specified safety limits; or (h) A release confirmed to be fission products greater than 10CFR20 frlom a fuel element;

2. A report within 30 days in writing to the NRC, Document Control Desk, Washington, D.C. of:

(a) Permanent changes in the facility organization involving Level 1-2 personnel; (b) Significant changes in the transient or accident analyses as described in the Safety Analysis Report; 14.6.8 Records 14.6.8.1 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

1. Normal reactor operation (but not including supporting documents such as checklists, data sheets, etc., which shall be maintained for a period of at least one year);

2. Principal maintenance activities;

3. Reportable occurrences; Rev. 8/12 14-31 USGS Safety Analysis Report

CHAPTER 14. TECHNICAL SPECIFICATIONS

4. Surveillance activities required by the Technical Specifications;

5. Reactor facility radiation and contamination surveys;

6. Experiments performed with the reactor;

7. Fuel inventories, receipts, and shipments;

8. Approved changes to the operating procedures; and

9. Reactor Operations Committee meetings and audit reports.

14.6.8.2 Records to be Retained for at Least One Certification Cycle

1. Records of retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one certification cycle.

2. Records of the most recently completed certification cycle for an individual shall be maintained at least as long as that individual is employed at the facility.

14.6.8.3 Records to be Retained for the Lifetime of the Reactor Facility

1. Gaseous and liquid radioactive effluents released to the environs;

2. Offsite environmental monitoring surveys;

3. Radiation exposures for all personnel monitored; and

4. Drawings of the reactor facility.

USGS Safety Analysis Report 14-32 Rev. 8/12

ML12251A231

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