ML13162A662

Redacted USGS RAI Clarification Information Needed to Support the USGS License Renewal SAR

ML13162A662
Person / Time
Site:	U.S. Geological Survey
Issue date:	05/17/2013
From:	Timothy Debey US Dept of Interior, Geological Survey (USGS)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Wertz, G
References
TAC ME1593
Download: ML13162A662 (33)

Text

U. S. GEOLOGICAL SURVEY RESEARCH REACTOR LICENSE NO. R-113 DOCKET NO. 50-274 U. S. GEOLOGICAL SURVEY RAI RESPONSES MAY 17, 2013 REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

• REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS

_USGS science for a changing world Department of the Interior US Geological Survey PO Box 25046 MS 974 Denver, CO 80225-0046 May 17, 2013 u.s. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Reference:

U.S. Geological Survey TRIGA Reactor (GSTR), Docket 50-274, License R-113, Request for Additional Information (RAI) dated October 2,2012

Subject:

Follow-up Safety Analysis responses from questions of phone conference on 3/21/13 Mr. Wertz:

Responses to questions from the reference phone conference are provided in the enclosed pages. Please contact me if further details, or corrections, are needed.

Sincerely,

~/

Tim De8ey USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct.

Executed on 5/17/13 Attachment Copy to:

Betty Adrian, Reactor Administrator, MS 975 USGS Reactor Operations Committee

9. Describe the limiting core configuration.

Please provide the results ofthe U.S. Geologic Survey TRIGA Reactor (GSTR) neutronic analyses that document the Limiting Core Configuration (LCC) and operating core.

• Core maps showing the contents of the core lattice positions for the limiting core configuration (LCC) and the operating core configuration (OCC):

0 00 0 0O0 000 0 0 ~

Control Rods 00 ~

~OO

. Fresh 12 wt% fuel CO old 8.5 wt% fuel 00 o ,.,.

.. 000

• Water filled positions 00 . 00 00 e 00 o 0 '

00 * , 00 000 000 0

0 000 0 0 .. -

o 00- _~ . '

. ~--~-

Figure 1. Limiting core configuration.

Figure 2. Operating core configuration.

1

• The enrichment and cladding type for fuel elements used at the GSTR o Aluminum & stainless steel clad <20% enriched (average 19.75%)

• As-modeled diagrams and dimensions for fuel and control elements in the GSTR:

a) Stainless steel clad fuel rod b) Aluminum clad fuel rod Figure 3. As modeled GSTR fuel elements.

2

a) fuel followed b) void followed c) fuel rod control rod control rod Figure 4. As-modeled GSTR SS-c1ad fuel- and void-followed control rods.

• The effective delayed neutron fraction (~eff) for the analysis is 0.00728, calculated using the adjoint weighted point kinetics estimators available in MCNP5 version 1.60. The difference between the calculated values of the LCC and OC are within the uncertainty bounds calculated by MCNP; therefore, the same value is used in all cases. The prompt neutron generation time is 4.28xlO- 5 seconds, calculated using the MCNP5 version 1.60 adjoint weighted point kinetics estimators.

• All of the neutronics results, except the core power distributions, are based on an operating power of 5 W. All reactivity numbers are calculated and reported in units of (~klk)/~eff.

• The calculated and measured all-control-rods-out k-effective (keff) and the excess reactivity (Pexcess) for the LCC and the OCC are shown below. The OCC calculated and measured values are comparable within uncertainty limits.

3

Calculated LCC: keff: 1.04714, Pexcess: $6.18 Calculated OCC: keff: 1.03650, Pexcess: $4.84 Measured acc: kerr: 1.03676, Pexcess: $4.87

• The Control Rod worths for each of the 4 control rods including the keffvalues determined for the LCC and the OCC are given below. The calculations for the OCC were taken with the other control rods at a critical position of 18.669 cm up from the fully inserted position. For the LCC the critical rod position was 17.3355 cm up from the fully inserted position.

o LCC:

Shim 1 Shim 2 Regulating Transient Rod Worth ($) 2.42 2.31 4.32 2.65 kerr (fully inserted) 0.98644 0.98769 0.97735 0.98561 kecc(fully withdrawn) 1.00388 1.00440 1.00832 1.00470 o OCC:

Shim 1 Shim 2 Regulating Transient Rod Worth ($) 2.16 2.25 3.36 2.06 kerc(fully inserted) 0.99944 1.00051 0.99691 1.00056 kerc(fully withdrawn) 1.01542 1.01718 1.02182 1.01578

• The comparison of the control rod worths calculated and measured from the acc. These values are acceptable within uncertainty limits:

Shim 1 Shim 2 Regulating Transient Calculated Worth ($) 2.16 2.25 3.36 2.06 Experimental Worth ($) 2.22 2.34 3.49 2.06

• The shutdown reactivity of the operating core with the highest-worth control rod (regulating rod) withdrawn is determined based on an MCNP run with the the regulating rod at 38.1 cm (fully withdrawn) and with all other rods fully inserted. This scenario yields a keff value of 0.99048, corresponding to a shutdown reactivity of -$1.32, which is in rough agreement with the value of -$1.75 calculated by subtracting the measured worth of the three remaining rods (shim 1, shim 2, and transient) from the measured all-rods-out reactivity.

• Power Distribution Graphics for the LCC and acc showing power in k W per fuel element:

4

./

.",- ....., 4-5kW 5-10kW

_ IO-15kW

_ 15-20kW

_ 20+kW

-

r*......--.. . . .

I \

1\ Water I

"'- - /

(~.~)

Fuel Followed Control Rods A - Regulating Rod B - Shim I Rod C - Shim 2 Rod Figure 5. Power profile for the limiting core with the control rods in the critical position.

5

Reactor Power: 915 kW _ 4-6

_ 6-8 8-10 10-12 Fuel Followed Control Rods A - Regulating Rod B Shim 1 Rod C - Shim 2 Rod Figure 6. Power profile for the operating core with the control rods in the critical position. (915 kW was used because this is a typical operating power level) 6

• Fuel temperature coefficient for the LCC and the OCC as a function of fuel temperature over the temperature range experienced during operation:

346.80 500.00 700 .00 1000.00 0.000

-0.002

-0.004

"""'

~ -0.006 A

'--'

......

.-!

c~ -0.008

( .)

0

~ -0.010 U

~ -0.012

'>

'.J:j

(.)

ro

~ -0.014 c:::

Q)

~

L.I....

-0 .016

-0.018 .OCC o LCC

-0 .020 Fuel Temperature (K)

Figure 7. Fuel reactivity coefficients for the operating and limiting core configurations as a function of temperature.

12 Describe the departure from nucleate boiling ratio (DNBR) analysis.

Thermal-hydraulic data for the LCC:

• Tables 1 and 2, below, detail the calculations used to determine the unit cells for the steady-state and transient thermal-hydraulic analyses.

7

Table 1. Unit cell calculations used for the steady state themlal-hydraulic analyses.

Parameter Value B-Ring Distance from Center (cm) 4.053 C-Ring Distance from Center (cm) 7.981 Inner Radius of Flow Channel (cm) 2.026 Outer Radius of Flow Ch31mel (cm) 6.017 Fuel Element Cross Section (cm2 ) 10.949 Wetted Perimeter of a Single Fuel Element (cm) 11.730 B-Ring Total Flow Area (cm2 ) 35.132 Flow Area/Rod (cm2 ) 5.855 Effective Hydraulic Diameter (cm) 1.997 Table 2. Unit cell calculations used for the transient thermal-hydraulic analyses.

Parameter Value

1. Elements 110

1. Control Rods 4 Element Cross Section (cm2 ) 10.949 l 2 Control Rod Cross Section (cm ) 9.580 Transient Rod Cross Section (cm2 ) 7.917 Central Thimble Cross Section (cm 2 ) 11.401 Core Cross Section (cm2 ) 2208.099 Total Flow Area (cm2 ) 824.209 2 2 Area/Rod (cm ) 8.382 Hydraulic Diameter (cm) 2.858 1 Shim 1, Shim 2 and Regulating Rods 2 Includes the four control rods

• The entry/exit pressure loss coefficients employed in the RELAP model are taken from the Oregon State University model:

• Inlet: 2.26 Exit: 0.63 8

• Diagrams of the steady state and transient RELAP models:

coolant sink coolant sink hot rod heat hot rod coolant average rod heat hot rod heat structure!

structure channel structure/coolant channel coolant channel 167188 Pa

....~I--I 60 °C coolant source coolant source a) steady state model b) transient model FiQUre 5. Steadv state and transient RELAP models of the GSTR

• Input assumptions used in the LCC DNBR analysis:

Hot rod element power: 22.2 k W Peaking Factor: 2.28 Inlet temperature: 333 .15 K Inlet Velocity: natural convection, computed by RELAP

• The RELAP steady-state model calculated the core flow rate, peak fuel and cladding temperatures, the location of the minimum DNBR (MDNBR), and the value of the MDNBR using the Bernath correlation:

Hot Channel Mass Flow Rate: ~0.129 kg/s Peak Fuel Temperature: 829.32 K (556.17 C)

Peak Cladding Temperature: 410.04 K (136.89 C)

MDNBR: 2.16 @ 23.812 em from the bottom of the fuel meat 9

• Reactivity feedback values used in the LCC transient analyses:

Table 4. Void reactivity feedback values for the limiting core.

Void Total Void Fraction Reactivity

(% Void) ($)

5 -0.15 Table 3. Fuel temperature reactivity feedback 10 -0.44 values for the limiting and operating core 15 -0.82 20 -1.18 configurations.

25 -1.65 DCC Total LCC Total 30 -2.16 Fuel Fuel 35 -2.75 Reactivity Reactivity 40 -3.45 Temperature (K) ($) ($) 45 -4.23 293.6 0 0 50 -5.04 400 -1.32 -1.32 60 -7.12 600 -4.98 -4.72 70 -9.77 800 -8.59 -8.06 80 -13.26 1200 -14.56 -13.71 90 -17.83 100 -23.70 The core water temperature feedback coefficient is a constant +0.008 $/K.

• Characterization of the response of GSTR to a reactivity pulse and an uncontrolled rod withdrawal transient event.

The GSTR pulse model sequence begins with the reactor at the Technical Specification limiting conditions of 60°C water temperature and a steady state power of 1 kW. Since the GSTR uses natural convection for cooling, no initial flow is assumed. These conditions are held for 1 second, after which a reactivity insertion is made that is equal to the requested pulse height ($3.00, $2.75, $2.50, or $2.00) over a 0.2 s period. This reactivity insertion is held for 1.5 seconds from the time of the beginning of the insertion, and then the pulse rod is inserted over a conservative 2 seconds. 15 seconds following the initiation of the pulse, the reactor scrams, adding -$5.00 of negative reactivity into the core over one second. The reactivity chart used in the RELAP simulation is shown in Table 5 with X replacing the pulse amount:

10

Table 5. Pulse reactivity insertion sequence.

Time (s) Inserted Reactivity ($)

o 0 1 o(pulse initiated) 1.2 X (pulse rod fully up) 2.5 X (pulse rod scram signal) 4.5 o (pulse rod fully down) 16 o (all rod scram signal) 17 -5 (all rods fully down)

These values are consistent with the Technical Specifications of the GSTR, along with information provided from the GSTR staff for values not covered in the Technical Specifications. In all of the following graphs, the limit line marks an upper bound of fuel temperature at 830 DC as recommended in report TRD 070.01006.05. It is noted that the TRD report discusses concerns with uranium loading of20 wt% and higher, although all of the GSTR fuel is less than 20 wt%. The $3.00 pulse analysis shows a peak fuel temperature of 1020 K (727 DC) which is 103 degrees below the 830 DC recommended level. The 830°C level has a 44 DC safety factor built into it so that the 727 DC temperature has no safety significance. The smaller pulse analyses show peak fuel temperatures well below the 830 DC recommended level and even farther below the 1150 DC safety limit.

Table 6 summarizes the safety critical information gathered from the pulse analyses, including the peak pulse power, the pulse full-width half-maximum (FWHM), and the peak fuel temperature.

Table 6. Summary of GSTR pulse simulation results Parameter $3.00 $2.75 $2.50 $2.00 Peak Power (MW) 1971 1793 1671 938 Peak Fuel Surface Temperature (K) 959 935 910 779 Peak Thermocouple Temperature (K) 1018 988 956 852 Peak Innermost Fuel Temperature (K) 1020 990 958 854 Peak Fuel Temperature (K) 1020 990 958 854 Pulse Full-Width Half-Maximum (s) 0.013 0.013 0.014 0.018 11

1200 I.E+ I0 1100 -i'---+~- ~~-......, ~ - . ~ - ~~ ~ ** -- .~ ~ - ~ . _ .~ _ _ ---t 1.E+09 1000 ............ ...... . .. ~ 1.E+08

~ ~ 1.E+07

~ 900 -j

'-"

Q)

.... +-........~..-...,.-_ _ ..._ - - - - - _ _- . _____ iiiio - _ ____ __ _ 1.E+06;;;-

, __ , G

.a

~

...

800

........

~

.... l.E+05 ~

~ l.E+04 ~

Q) c.. 700 .. "............. - - - -

E ....... -.....---...,..

~ 600 - ........ ..... _---- 1.E+03

.....*. innemlost fuel temperature (K)

"--._-- - .... - ...-- ... ---- 500 1.E+02

- thermocouple location temperature (K) I

- - - - fuel surface tempemp erature (K) 400 - . temperature limit (K) I.E+Ol po..,ver (W) 300 I .E+OO 0 5 10 15 20 Time (s)

Figure 9. LCC hot rod results for a $3 pulse.

14.2 Describe how the limited safety setting system (LSSS) and SCRAM setpoints protect the safety limit.

• Analysis ofthe uncontrolled rod withdrawal for the LCC:

Table 7 provides reactivity vs. time table used in the RELAP simulation of the continuous rod withdrawal accident. This simulation starts with the same limiting initial conditions as the pulse simulation (reactor critical at 1 kW, pool water at 60 C). The simulation utilizes a maximum, constant rod withdrawal speed of 0.9535 cm/s, which leads to a reactivity vs. time table as shown below, based on the detailed integral control worth curves calculated from the GSTR using the MCNP model.

For this simulation the important safety information is shown in Table 8. A SCRAM setpoint of 1.1 MW provides complete safety for the reactor, with an insignificant amount of energy produced and released, leading to a negligible temperature increase. This accident clearly does not challenge the integrity of the fuel, nor does it represent a threat to personnel or the environment. The control rod withdrawal rate and the control rod scram time are acceptable without need for change.

12

Table 7. Reactivity as a Table 8. Peak fuel temperature and power function of time for the results from the uncontrolled rod uncontrolled withdrawal withdrawal simulation.

of the regulation rod. Peak Reactivity Hot Rod Parameter Values Time (s) ($) Fuel CL Temp (K) 350 0.0 0.0 Fuel TIC Location Temp (K) 350 Fuel Surface Temp (K) 346 1.0 0.0 Peak Power (MW) 2.27 4.9 0.07 Peak Time (s) 13.5503 8.79 0.34 12.69 0.79 13.87 1.01 15.87 -5.00

• The 12 wt% fuel contains ~50% more uranium by mass than the 8.5% fuel and provides the limiting results for fuel used in the GSTR. Additionally, bumup further reduces the amount of fissile uranium in the 8.5% fuel within the GSTR limiting core configuration. The higher uranium density causes the 12 wt% fuel to produce a higher power density than the 8.5% fuel.

15.3 Methods used to determine the maximum hypothetical accident (MHA) doses

• The updated GSTR SAR Section '13.2.1.1 references a mdioactive release in water, when NUREG-1537 states that the "failure of one fuel element in air is the MHA for 8 TRIGA reactor". Please revise tile MHA analysis and all references to a release in water, or explain. [pg 201 The MHA analyzed was based on the statement in NUREG 1537, Part 2, page 13-5, that a "TRIGA fuel element would lose cladding integrity while suspended in air (or in the reactor pool) .... " For the GSTR, a MHA of a fuel element instantaneously losing cladding integrity in air is not credible.

Nevertheless, the GSTR has redone the MHA analysis to be in air. The new analysis is below.

• Please provide a reference for the source term provided in Table '13.1. [pg. 22]

A reference section has been added to the response and the correct reference is in the text.

• Although not so stated, the NRC staff interpreted the tempemture used in equation '13.1 as being the centerline temperature from the DNBR analysis After a revised fuel temperature is obtained following correction to tile DNBR model identified above, please provide a revised analysis ensurinig tllat the temperature used is in this analysis is a fuel only temperature, not tile centerline. The NRC staff notes tllat the use of a volume 8veraged fuel temperature in the limiting pin is sufficiently conservative for equation 13.1. [pg. 21]

13

The Colorado School of M~nes has perfonned the DNBR analysis for steady state operation with the limiting core configuration (LCC) and provided the GSTR with a volume averaged fuel temperature in the limiting pin of 423.22°C and a peak power of 22.18 kW +/- 0.03 kW. Note that in the prior pages ofthis response, the power is rounded to 22.2 kW. This new temperature and power are used in the analysis below.

• Table 13.5 thyroid doses are not needed. Please provide the CEDE and TEDE. [pg 26].

The CEDE and TEDE were provided in Table 13.6. In the analysis below, Table 13.5 will be removed and Table 13.6 will change to Table 13.5. We will continue to display the CEDE and TEDE values for the MHA.

• Table 13.6 indicates that the MHA occupational dose exceeds the limit of Title 10 of the Code of Federal Regu/ations Part 20 prior to 5 minutes. Please provide an analysis and conclusions that demonstrate compliance with the requirements for occupational dose (5000 millirem for the stay time authorized in the emergency plan). rpg.26-271 Table 13.6 has changed to Table 13.5 and the new analysis attached below reflects compliance with the requirements for occupational dose in 10 CFR 20.

The following sections will replace the sections given in the original SAR for the USGS.

13.2 Accident Initiating Events and Scenarios. Accident Analysis. and Determination of Consequences 13.2.1 Maximum Hypothetical Accident (MHA) 13.2.1.1 Accident Initiating Events and Scenarios A single fuel element could fail at any time during normal reactor operation or while the reactor is shutdown due to a manufacturing defect, corrosion, or handling damage. This type of accident is very infrequent, based on many years of operating experience with TRIGA fuel, and such a failure would not normally incorporate all of the necessary operating assumptions required to obtain a worst-case fuel-failure scenario. Historically, TRIGA fuel failures have shown very small fission product releases.

For the GSTR, the MHA has been required to be the cladding rupture of one highly irradiated fuel element followed by the instantaneous release of the noble gas and halogen fission products into the air. For the GSTR, with three different possible fuel types, a 12 wt% fuel element was chosen as the irradiated element since it contains the most 23S U and the highest inventory of fission products. The failed fuel element was assumed to have been operated at the highest core power density in the limiting core configuration (Leq for the extremely conservative continuous period of one year at 1 MW, which is realistically not possible. This results in all ofthe halogens and noble gases (except Kr-85) reaching their saturated activities.

This is the most severe accident and is analyzed to determine the limiting or bounding potential radiation doses to the reactor staff and to the general public in the unrestricted area. A less severe, but more credible accident, involving this same single element having a cladding failure in water will also be analyzed. This latter accident more correctly falls into the mishandling or malfunction of fuel accident category and will be addressed there.

During the lifetime of the GSTR, fuel within the core may be moved to new positions or removed. Fuel elements are moved only during periods when the reactor is shutdown. Also, the GSTR is very rarely 14

operated continuously at 1 MW for a period longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and never for a period of one year.

Nevertheless, this extremely conservative MHA has been analyzed for the GSTR.

The following scenario has been chosen for analysis:

• A 12 wt% fuel element was chosen as the irradiated element since it contains the most 235 U and the highest inventory of fission products. The failed fuel element was assumed to have been operated at the highest core power density for a continuous period of one year at 1 MW in the limiting core configuration resulting in 22.21 kW (22.18 kW + 0.03 kW uncertainty) in the element. This results in all of the halogens and noble gases (except Kr-85) reaching their saturated activities. Operating the reactor at 1 MW produces the fission product poison Xe-135, which depresses the neutron flux and power production in the "hot rod." The power production assumed in the "hot rod" is therefore very conservative. The element is removed from the core five minutes after the irradiation ends and brought into the reactor bay air. The five minute decay time comes from a conservative estimate of two operators w'orking expeditiously and without delay to remove security features that prevent fuel removal. In reality this time would be significantly longer than five minutes. This scenario assumes that the released noble gas and halogen fission products instantly and uniformly mix with the reactor room air. The fission products that have been released to the reactor room air are then exhausted at 1

the stack ventilation rate of 800 cfm (3.78 x 10 5 cm 3sec- ), through the emergency exhaust stack with no filtration taken into account. The air is assumed to be discharged at 6 meters (19.69 feet) above 8

ground, at the exit of the exhaust stack. The reactor room free volume is assumed to be 3.1 x 10 cubic centimeters. The exhaust system takes 15.6 minutes to expel one reactor room volume of air (3.84 room changes per hour). The time to discharge 95% of the fission product gases from the reactor room is 47 minutes, but this analysis conservatively assumes that all fission product gases are released instantaneously in a single pulse discharge. Similarly, it is conservatively assumed that the gas concentration in the reactor bay undergoes no dilution during the maximum assumed stay time of 5 minutes. These multiplied conservatisms make this scenario highly unrealistic.

13.2.1.2 Accident Analysis and Determination of Consequences It is assumed that the GSTR is fueled in the limiting core configuration shown in Figure 13.1, and that the reactor has operated continuously at 1 MW for a period of one year. Thus, all halogens and all noble gases (except Kr-85) are at their saturation activity. The highest-power density fuel element fails and releases the noble gases and halogens to the gap between the cladding and the fuel. This highest-power-density element has a conservative power density of 22.18 kW +/- 0.03 kW. For this analysis a worst case scenario will be assumed and the element has a conservative power density of 22.21 kW. The fission product inventory of halogen and noble gases are given in Table 13.1 for this element. The inventory assumes a saturated activity is present and is based upon the fission yield for each isotope. The saturated activities are then decay corrected for a 5 minute delay. The saturated activities in Table 13.1 were calculated from Oregon State University's (OSU) fission product inventory for the same MHA scenario by multiplying OSU's number by the ratio of the highest power density element at the GSTR (22.21 kW) to the highest power density element at the OSU reactor (15.9 kW) [11.

Considerable effort has been expended to measure and define the fission product release fractions for TRIGA fuels. Data on this aspect of fuel performance are reported. Using these data, GA developed a conservative correlation for fission product release to be 15

_ -5 3 ( -1.34 x 10 4 '

e = L) x 10 + 3.6 )( 10 exp (T '7.... .

, +- j.l ,

(13.1)

At an average fuel temperature of 423.22 DC, this release fraction is 3.08 x 10.5

• This assumed fuel temperature (423 .22 DC) is the expected volume averaged fuel temperature in the limiting pin for the LCC and will produce a conservative estimate for the fission product release .

Control Rods

. Fresh 12 wt% fuel 00 old !I.S wt% fuel

• Water tilled positions Figure 13.1 Limiting core configuration layout Once the fission products are released to the cladding gap, this activity is released when the cladding catastrophically fails. If the release is in air (MHA), then this activity is released directly into the reactor bay air. If the release occurs in the pool water, then the fission products must migrate through the water before being released to the reactor bay air. Once released into the reactor bay air, a further reduction of the halogen activity is expected to occur due t o plateout on the surfaces of the bay.

The fraction (w) of the fission product inventory released from a single fue l element that reaches the reactor room air and, subsequently, the atmosphere in the unrestricted environment is:

w =e f g h, (13.2) where :

e = the fraction released from the fuel to the fuel-cladding gap (3 .08xl0' 5 ) ;

f = the fraction released from the fuel-cladding gap to the reactor bay air (if no water is present), or to the pool water (if water is present);

g = the fraction released from the pool water to the reactor bay air (g=1.0 when no water is in the pool); and h = t he fraction released from the reactor room air to the outside unrestricted environment, due to plateout in the reactor bay.

16

Table 13.1 Saturated and Corrected Activities for H Power Den 12 wt% Fuel Element rr============r==========~===========r======~

Activity Saturated after 5 Isotope Half life Activity (Ci) minute decay(Ci}

Br-82 35.3 h Br-83 2.4 h Br-84m 6.0 min Br-84 31.8 min Br-85 2.87 min Br-86 55.5 sec Br-87 55.9 sec Total Bromine 1-131 8.02 d 1-132 2.28h 1-133 20.8 h 1-134 52.6 min 1-135 6.57 h 1-136 83.4 sec Total Iodine Kr-83m 1.86 h Kr-85m 4.48h Kr-85 10.76 yr Kr-87 76.2 min Kr-88 2.84 h Kr-89 3.15 min Total Krypton Xe-131m 11.9 d Xe-133m 2.19 d Xe-133 5.24 d Xe-135m 15.3 min Xe-135 9.1 h Xe-137 3.82 min Xe-138 14.1 min Total Xenon Total Ha Total Noble Gases 17

For the accident where the cladding failure occurs in air, it is very conservatively assumed that 25% of the halogens released to the cladding gap are eventually available for release from the reactor bay to the outside environment. This value is based on historical usage and recommendations. It uses a 50%

release of the halogens from the gap to the air with a natural reduction factor of 50% due to plateout in the reactor bay. Combining the 50% release from the gap with the 50% plateout results in the 25% total release. However, this value appears to be quite conservative, as some references quote a 1.7% release from the gap rather than 50%. In the reactor bay it is conservatively assumed that 50% of the halogens released to the cladding gap are released into the reactor bay.

For the accident in air, 100% of the noble gases are assumed to be available for release to the reactor bay and later the unrestricted environment.

For the accident in water, it is assumed that 95% of the halogens released from the cladding gap remain in the water and are removed by the demineralizer. A small fraction, 5%, of the halogens is assumed to escape from the water to the reactor room air. Combining this with the 50% release from the gap to the water, the result is that 2.5% of the halogens in the gap are released to the reactor room. Again, 50% of these plateout in the reactor bay before release to the outside environment. Thus a total of 1.25% of the halogens is available for release to the outside environment. For the noble gases released under water, 100% are assumed to be available for release to the unrestricted environment.

The experiences at Three Mile Island, along with recent experiments, indicate that the 50% halogen release fraction is much too large. Possibly as little as 0.06% of the iodine reaching the cladding gap may be released into the reactor bay due in part to a large amount of the elemental iodine reacting with cesium to form Csl, a compound much less volatile and more water soluble than elemental iodine.

The very conservative values for these various release fractions (see Equation 13.2) are given in Tables 13.2 and 13.3.

Table 13.2 Release Fraction Components f f g g Fission product No pool With pool No pool With pool h water Water water water Noble gas 1.0 1.0 1.0 1.0 1.0 Halogens 0.5 0.5 1.0 0.05 0.5 Table 13.3 Total Release Fraction w to the reactor w to the reactor w to the w to the Fission bay bay environment environment product No pool water With pool water No pool water With pool water Noble gas 3.08 E-5 3.08 E-5 3.08 E-5 3.08 E-5 Halogens 1.54 E-5 7.70 E-7 7.70 E-6 3.85 E-7 18

For the GSTR, the prevailing wind is from the west, blowing to the east. The minimum distance to the unrestricted environment (475 m) is to the north, the minimum distance to the nearest public residence (640 m) is to the north, and a public school is about 720 m to the east. To be conservative it was assumed that the wind is blowing from west to east and all recipients are east.

The DOE HOTSPOT computer code version 2.07.2 was used for areas outside of the reactor bay, assuming uniform dispersion with ICRP 30 dose conversion factors. The HotSpot Health Physics Code was created for use for safety-analysis of DOE facilities handling nuclear material. Additionally, HotSpot provides emergency response personnel and emergency planners with a fast, field-portable set of software tools for evaluating incidents involving atmospheric releases of mixed isotopes of radioactive material. HotSpot incorporates Federal Guidance Reports 11, 12, and 13 (FGR-11, FGR-12, FGR-13) Dose Conversion Factors (DCFs) for inhalation, submersion, and ground shine. The results of the Hotspot analyses are provided in Table 13.6.

Furthermore, for calculations beyond the reactor bay, it was conservatively assumed that all of the fission products were released to the unrestricted area by a discharge pulse, which would maximize the dose rate to persons exposed to the plume during the accident. Calculations inside the reactor bay assumed uniform distribution of the released fission products within the ~3.1 x 108 cc volume of the bay.

It was also assumed that the receptor breathing rate was 3.33 E-4 m 3sec- 1 (NRC "light work" rate) and that the longest isotope retention category was applicable.

Calculations for personnel inside the reactor bay conservatively assumed that all of the fission product gases were released instantly and uniformly distributed within the reactor bay. The exposures for personnel in the reactor room for short stay-times were calculated by conservatively assuming that the fission product concentration was constant for that time period. The short stay time is conservatively estimated to be no longer than 5 minutes because the GSTR has a small facility and a worker could briskly walk from one end of the reactor bay to the other within 30 seconds; leaving at least 4 minutes to finish any emergency necessary tasks. A more realistic non-conservative stay time for a worker in the reactor bay would be 2 minutes. The isotope concentrations in terms of DAC values and DAC-Hr exposures during a 2-minute stay time are given in Table 13.4 below. Values for 5 minute stay times are 2.5 times higher than the 2 minute stay time values since the fission product gas concentration is assumed to be constant during this exposure period.

Since a stochastic exposure of 2000 DAC-Hr results in a TEDE of 5000 mrem, the TEDE in mrem can be calculated by TEDE =(DAC-Hr)*5000/2000. (13.3)

Since a non-stochastic exposure of 1 annual limit on intake (ALI) gives a CDE of 50,000 mrem for the target organ (thyroid for radioiodine) the dose received to the thyroid of a person standing in the reactor room can be calculated by CDE =3.33E-4*t*C/ALI*50000, (13.4) where:

3.33E-4 = the NRC "light work" breathing rate with units of m 3 sec- 1; 19

t = the time exposed to the radionuclide; ALI =the occupational inhalation limit for the specified isotope from 10 CFR 20 Appendix B; and C =the concentration of the radionuclide in IlCi/m3.

Table 13.4 Concentrations and Exposures from Gaseous Fission Product Releases DAC*Hr Released Released DAC value of DAC*Hr Released Released DAC value of DACfrom 10 exposure Activity after Activity to Activity to diluted activity in exposure for Activity to Activity to diluted activity in CFR 20 Isotope 5 minute reactor bay reactor bay no minute Reactor Room envi ron ment reactor bay (with dec.y(Ci) Air NO POOl pool water (# time. no pool AIr WITH WITH POOl pool water) (#

WATER (mCi) DACs) water WATER (mCI) WATER (mCI) DACs) minute stay time 103.18 32.45 A summary of the the CDEThyroid and TEDE for 2- minute and 5-minute stay times in the reactor bay are shown in Table 13.5. As seen in Table 13.5 the TEDE and CDE for a 5 minute stay time in the conservative scenario where the fission products are released directly into the reactor bay air do not exceed the 10 CFR 20 occupational dose limits. The reactor staff finds this analysis to be acceptable to show that the GSTR is operating within 10 CFR 20 limits as the MHA is a highly conservative and a postulated scenario that has no actual credibility of occurring.

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Table 13.5 Occupational CDEThyroid and TEDE in the Reactor Room Following a Single Element Failure in Air and Water Reactor Room Occupancy CDEThyroid (no water) TEDE (no water) CDEThyroid (water) TEDE (water)

(minutes) (mrem) (mrem) (mrem) (mrem) 2 1657 258 83 81 5 4143 645 207 203 The results of the HOTSPOT code vers ion 2.07.2 calculations for the two scenarios (no water vs water in reactor tank) are shown in Table 13.6. As seen from the table, no water in the reactor gives the highest doses to the general public at any distance, as is expected since there is no capture of fission products by the water. The scenario with wate r in the reactor tank gives the lowest doses at any given distance since the capture and retention of fission products in the water is Significant . In all cases, doses for the general publ ic and occupational workers are we ll below the annual dose lim its specified by 10 CFR 20.

The reactor staff finds this analysis to be acceptable to show that the GSTR is operating within 10 CFR 20 limits as the MHA is a highly conservative and a postulated scenario that has no actual credibi lity of occurring. For our model we used the following inputs:

• Atmospheric Dispersion Models: General plume model,

• Mixture of isotopes from Table 13.4, when requested the D categorization for the Br isotopes was used. Br-86, Br-87, and 1-136 were not used in the calcu lation. It was assumed that those isotopes would not cause a sign ificant dose as the ir half lives are too short <<84 sec) compared to the relative time it would take to trave l out of the reactor bay and into the environment.

• Release height of 0 m for a ground release,

• A 10-meter wind speed of 3.84 m/s (average f rom Chapter 2 of the Safety Analysis Report),

• Wind is blowing from the west to the east,

• The ambient environment is moderately stable (F),

• Terra in is standard,

• Wind reference height is 10 m,

• Sample time is 10 m;n,

• Source geometry is Simple,

• Include ground shine,

• The non-respirable deposition ve locity is 8 cm/sec,

• The holdup time is 0 min,

• DCF library used was the FGR-ll corresponding to ICRP 30 series, 3

• The breathing rate is 3.33e-4 m /s,

• And all distances are on the plum center line for a conservative dose estimate at each location.

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Table 13.6 Radiation Doses to Members of the General Public Following a Single Element Failure CDEThyroid (no TEDE (no TEDE Distance CDEThyroid Locatton water) water) (water)

(m) (water) (mrem)

(mrem) (mrem) (mrem)

Building 15 11 52 53 0.0 0.0 south door Emergency 32 14 6.1 3.0e-5 4.3e-5 assembly area Building 21 east entrance (West 49 42 4.3 1.5e-4 2.2e-4 of Building 15)

Average of eastern 100 38 2.4 1.5e-4 2.3e-4 intersections Building 16 175 16 0.96 7.0e-5 1.0e-4 west entrance

- 200 13 0.76 5.6e-5 8.2e-5

- 250 8.6 0.5 3.8e-5 5.6e-5 Nearest Unrestricted 475 2.5 0.14 1.2e-5 1.8e-5 Location Residence 640 1.4 0.081 7.0e-6 1.0e-5 School 720 1.1 0.065 5.7e-6 8.3e-6 References

[1] Oregon State University TRIGA Reactor, "Safety Analysis Report," Oregon State University, Oregon, Rep. ML0714304582, 2004.

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23.1 Criteria for significant change in core configuration.

The significance in a change of core configuration is determined by the potential for the change causing a violation of the facility license or technical specifications. The shutdown margin (SDM) of the core will increase (an increase in safety margin) through negative reactivity changes such as loading samples with negative reactivity worth, buildup of stable fission product poisoning, or depletion of the uranium fuel. The SDM of the core will decrease (a decrease in safety margin) through positive reactivity changes such as loading samples with positive reactivity worth or adding new fuel elements to the core. Defining a fixed core reactivity change as having significance for the potential to cause a violation of the facility license or technical specifications is onerous and hard to justify. If a core configuration has a SDM of$1.50 (with a technical specification SDM limit of $0.30), then a positive increase in core reactivity of $0.50 has no safety significance and control rod calibrations would not be justified. Conversely, if a core configuration has a SDM of $0.35 (with the same SDM limit of $0.30), then a positive increase in core reactivity of $0.50 might result in a violation and control rod calibrations would definitely be justified. Therefore our definition of a significant change in core configuration is a positive reactivity change in core composition or arrangement that increases reactivity by a value greater than or equal to the shutdown margin. This definition is the same as the March 2011 approved Technical Specification 4.2.1 of the University of Wisconsin TRIGA reactor.

24.3 Basis for the SDM Value Control rod calibrations, which result in integral control rod worth curves, are performed by starting the subject control rod at the bottom of its motion and then perfonning a series of step insertions of approximately $0.25 until the control rod is at the top of its range. Each step insertion is preceded by having the reactor exactly critical. This process can take several hours to perform for a single control rod. The control rod worth curve is then created by starting at the reference zero point which is at the control rod's bottom stop. There is no error in this zero point.

Each step insertion then has an error of2.2% to 2.5% of the reactivity for that insertion. Since the control rods are only partially raised when doing the shutdown margin (SDM) measurement, the error in the control rod worth measurement is only the error associated with the movement of the control rod from the bottom stop (zero reference point) to that critical rod position. The error associated with the movement of the control rod from that critical rod position to the top of its range is not a factor in the SDM determination. It is erroneous to apply the error for the full travel of the control rod to the low power critical rod position. The proposed SDM and maximum excess reactivities proposeu for the GSTR license renewal are consistent with those approved for other 1 MW TRIGA reactors by the U.S. Nuclear Regulatory Commission in recent licensing activities.

The LCC configuration is not intended to produce a core with the maximum excess reactivity. It is intended to produce a core with maximum power peaking; a worst case condition. Adding more fuel elements to the LeC core will increase the excess reactivity, but that will also reduce the maximum power peaking in the core. The GSTR has historical operational experience 23

(31711969 log entry) with an excess reactivity near $7.00 ($6.96). A SDM of$2.12 was present in that core configuration. The following discussion is an analysis of this high reactivity core for its margin of safety regarding the SDM. The full travel control rod worths for that core were $2.98,

$2.98, $3.15, and $4.30. The critical rod positions for this high worth core were approximately 425 units (out of 1000 units full travel). These rod positions, for all rods except the regulating rod, represent a total inserted reactivity of -$4.25 from their fully down positions, and the associated reactivity error of 2.5% would represent a combined error of $0.1 06. This reactivity

($4.25) is the negative reactivity insertion potential for those three control rods and this is the negative reactivity available for ensure a sufficient SDM. The SDM calculation equals this combined (3-rod) reactivity value ($4.25) subtracted from the positive reactivity remaining in the regulating rod at its critical position ($2.13), since the regulating rod must be assumed to be fully withdrawn. The error in that regulating rod positive reactivity potential is 2.5% of $2.13, or

$0.053. The resulting SDM is $4.25 minus $2.13, equaling $2.12. The maximum error in the SDM calculation, assuming all rod calibration errors are in the most non-conservative direction, is $0.106 plus $0.053, equaling $0.159. This margin of safety has been accepted at other 1+ MW TRIGA facilities and is acceptable at the GSTR.

24.9 Calculational Reference for Values Given for Ar-41 Release Evaluation The distance of295 meters given as the nearest location a member of the public could stay with continuous uninterrupted occupancy is an error. After further review, the closest location a member of the public could stay to the GSTR with continuous uninterrupted occupancy is 475 meters. This location is north of the GSTR next to the 6th Avenue Frontage Road. The analysis for our Technical Specification limit for releasing Ar-41 to the environment is re-perfomled below using 475 meters as the receptor location.

• The text preceding Table 1 appears to indicate that the distance used in the analysis was 295 meters to the receptor (Denver Federal Center (DFC) fence line); however, a corresponding dose value is not provided in Table 'I for that distance. Also, doses to members of the public should not be reduced by occupancy factors. Statements regarding lOitering are not applicable to receptor locations for the public. Please explain or revise. [pg. 30]

Table 1 will be adjusted to include the annual dose calculated at 475 meters. The distance of 475 meters is the nearest distance to the facility that a member of the public could stay at indefinitely; therefore the dose calculated at 475 meters will not be reduced by the occupancy factor. All distances within 475 meters of the facility are located on the Denver Federal Center or in an area where no member of the public would be able to stay indefinitely. The estimated occupancy factor for a member of the public on the Denver Federal Center is 10%, but in this analysis we are being conservative and using an occupancy factor of22.8%.

• The calculational results in Table 2 indicate 7.75 curies (Ci) per year of Argone 4'1 (Ar-41) released which corresponds to -950 hours of full power operation. The most recent USGS Annual Report, dated January 24, 20'12, indicated that 12.607 Ci of Ar-41 was released for '119'1 megawatt-hour of operation. Please explain how Table 2 provides a best estimate of routine operation given the significantly larger release in 20'12. or revise. [pg. 30]

24

The value of7.75 Ci of Ar-41 released in 2011 was obtained from the 2011 USGS Annual Report. The most recent value (from the 2012 USGS Annual Report) of 12.607 Ci of Ar-41 released was not available when the initial RAI response was written. The Technical Specification limit for the GSTR is 4.8 x 10-6 uCi/ml. Using the parameters provided in the Basis section under Section 14.3.7.2,4.8 x 10-6 uCi/ml corresponds to a total annual release of 71.44 Ci of Ar-41. The GSTR annual releases of Ar-41 have been substantially less than the limit and can fluctuate by several Curies a year, depending on the operational schedule of the GSTR. The analysis in the Basis section under Section 14.3.7.2 shows that releasing the Technical Specification limit will not result in an over exposure to a member of the public. An analysis showing our 2012 release of 12.607 Ci of Ar-41 will be added into the Basis section under Section 14.3.7.2. In addition, an analysis showing our historical average release of5.7 Ci of Ar-41 will be added into the Basis section under Section 14.3.7.2.

• The calculational result in Table 2 for the 295 meter distance (DFC fence) appears to have been reduced by the occupancy factor of 22.8 percent. Doses to members of the public should not be affected by occupancy foctors. Pleose exploin or revise. [pg. 30)

Table 2 provides the summary of the analysis for the 2011 annual release of Ar-41, and the column showing the dose after the occupancy factor is applied should not have been extended to the 295 m distance. The distance of 295 meters will be changed to 475 meters which is the nearest distance to the facility that a member of the public could continuously occupy and the dose calculated at 475 meters will not be reduced by the occupancy factor.

• The COMPLY output file provided appears to be identical to the one provided in the previous submittal (dated November 16, 20'12) and does not reflect analysis as is now being presented. Please explain or revise. [pg 30]

The COMPLY output file previously provided showed the analysis for releasing the Technical Specification limit of 71.44 Ci (4.8 x 10-6 uCi/ml) in one year and calculated the applicable dose at 295 meters. The GSTR staff has reviewed the file submitted and determined that it did reflect the analysis being presented in the Basis section under Section 14.3.7.2. However, the analysis is being changed to show a release of the 71.44 Ci Technical Specification limit (4.8 x 10-6 uCi/ml) in one year and to calculate the applicable dose at 475 meters. The attached COMPLY file has been updated to reflect this revised analysis.

The original submitted SAR Basis section under Section 14.3.7.2 was replaced with the response to RAI #24.9. The submission to RAI 24.9 will be now modified as shown below to respond to the third RAI #24.9 dated 3-7-2013. The response below will take the place of the Basis section under Section 14.3.7.2 in the SAR.

Basis. If Ar-41 is continuously discharged at 4.8 x 10-6 )..lCi/ml, measurements and calculations show that Ar-41 released to publicly accessible areas under the worst-case weather conditions would result in an annual TEDE of 0.3 mrem. This is only 3% of the applicable limit of 10 mrem. The calculation was performed with the Environmental Protection Agency's COMPLY code. The following input paranleters were used:

25

Nominal exhaust flow: 1000 cfm, Ar-41 release in Ci/s: <<4.8 x 10-6 /lCi/ml)(1000 cfm)(1/60 min/sec)(1 1(1 x 106)Cil /lCi)(28316.85 mllft3))= 2.266e-6 Ci/s, Release height: 6 meters, Building height: 4 meters, Distance from source to the receptor: 475 meters, Building width: 30 meters, Default mean wind speed: 2.0 m/sec.

Using the above input parameters the USGS passes the EPA's Comply code at level 2. This is shown in the COMPLY code report included below. Using level 2, of EPA's COMPLY code and the above input parameters, the dose from the Ar-41 exhaust was also calculated at various distances from the exhaust stack. The calculated doses are shown in Table 1 and in the far right colunm an occupancy factor of 22.8% has been applied to the doses at all locations on the Denver Federal Center (DFC). The occupancy factor comes from the fact that the DFC is not occupied all week long and it is constantly monitored by the Federal Protective Service. Anyone loitering in an area would be questioned and asked to leave. No member of the public would be able to spend more than 22.8% of the year in anyone spot on the DFC. The occupancy factor value of22.8% is a conservative number calculated from 2000 working hours in one year (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />). The occupancy factor is not applied to the dose at the northern frontage road (475 meters) because that is the nearest distance to the GSTR that a member of the public could stay continuously.

Using the operation history of the GSTR, conservative dose estimates for personnel are listed in Table 2. The input parameters for this analysis are shown below:

Ar-41 release in Ci/year: 12.6, Ar-41 release in Ci/sec: 3.995e-7, Release height: 6 meters, Building height: 4 meters, Building width: 30 meters, Default mean wind speed: 2.0 m/sec.

The release of 12.6 Ci/year comes from the GSTR's 2012 arumal report. During 2012, the GSTR operated more than nornml, giving a maximum annual release of 12.6 Ci of Ar-41. A more typical dose estimate would use the same parameters as listed above but with a release of 5.7 Ci/year. This release value is the average release of Ar-41 at the GSTR facility from 1969 through 2012. The doses to personnel from a release of5.7 Ci of Ar-41 in one year are listed in Table 3.

In all instances, the doses to a member of the public are below the 10 mrem limit except for the dose at the Building 15 south door when the Technical Specification limit of 4.8e-6 /lCi/ml is released in one year. As shown in Table 1 the dose at this location with the occupancy factor is applied is 30.78 mrem/yr. The GSTR feels that an occupancy factor of 22.8% for the south door of Building 15 is too high because of monitoring, security protocols, and general use of that location. A more accurate and still conservative occupancy factor for the south door of Building 26

15 would be 5%, which is 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> at the door for every day of the year that public access is allowed on the Federal Center. This reduces the dose to a member of the public to 6.75 mrem/year when releasing the Technical Specification limit of 4.8e-6 ~lCi/ml of Ar-4l.

Table 1: Yearly doses due to Ar-41, release limit at several distances with occupancy factor applied. All yearly doses calculated with EPA's COMPLY code.

Distance Dose Dose with 22.8% Occupancy Location (m) (mrem/yr) Factor (mremfyr) 30.78 (6.75 for 5%

11 135 Building 15 south door occupancy) 32 16.7 Emergency assembly area 3.81 49 10.4 Building 21 east entrance 2.37 Average of eastern 100 4.1 intersections 0.93 175 1.8 Building 16 west entrance 0.41 475 0.3 Northern frontage road Not Applicable Table 2: Yearly doses due to release of 12.6 Ci of Ar-41, at several distances with occupancy factor applied. All yearly doses calculated with EPA's COMPLY code.

Dose with 22.8%

Distance Dose Location Occupancy Factor (m) (mrem/yr) (mrem/yr) 11 23.9 Building 15 south door 5.45 32 3.0 Emergency assembly area 0.68 49 1.8 Building 21 east entrance 0.41 Average of eastern 100 0.7 intersections 0.16 175 0.3 Building 16 west entrance 0.07 475 0.057 Northern frontage road Not Applicable Table 3: Yearly doses due to release of 5.7 Ci of Ar-41, at several distances with occupancy factor applied. All yearly doses calculated with EPA's COMPLY code.

Dose with 22.8%

Distance Dose Location Occupancy Factor (m) (mrem/yr) (mremfyr) 11 10.8 Building 15 south door 2.46 32 1.3 Emergency assembly area 0.30 49 0.8 Building 21 east entrance 0.18 Average of eastern 100 0.3 intersections 0.07 175 0.1 Building 16 west entrance 0.02 475 0.026 Northern frontage road Not Applicable 27

COMPLY: V1.6. 4115/2013 10:31 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 - Vl.6.

Prepared by:

USGS GSTR PO Box 25046, DFC MS-974, Denver, CO 80225 Alex Buehrle 303-236-4726 Prepared for:

U.S. Enviromnental Protection Agency Office of Radiation and Indoor Air Washington, DC 20460 COMPLY: V1.6. 4115/2013 10:31 Ar-41 release 4.8e-6 uCi/ml for 1 year SCREENING LEVEL 1 DATA ENTERED:

Effluent concentration limits used.

CONCENTRA nON Nuclide (curies/cu m)

AR-41 4.80E-06 NOTES:

Input parameters outside the "nomlal" range:

None.

RESULTS:

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

• • • Failed at level 1.

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COMPLY: V1.6. 4115/2013 10:31 Ar-41 release 4.8e-6 uCi/ml 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 475 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: 0.3 mrem/yr.

• • • Comply at level 2.

This facility is in COMPLIANCE.

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

You may contact your regional EPA office for more information.

• • • • • • • • • • END OF COMPLIANCE REPORT **********

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In addition to the above responses, we also are changing the following items in Chapter 14, Technical Specifications, of our original submission.

SAR Section :

14 . 3 . 2 . 3 Tabl e 2 - ~nimum Reactor Safety Channels Effective Mode Safety Channel Function S .- S . Pulse S .-W.

Power Level SCRAM @ 1.1 MW(t) 2 - 2 or less Preset timer SCRAM (.:::15 sec) - 1 -

Console Scram Button SCRAM 1 1 1 High Voltage SCRAM @ loss of nominal operating voltage to required power 2 1 2 level channels Watchdog scrams Scram upon loss of refresh signal 2 2 2 in DAC or CSC computer (one scram circuit per computer)

(Changed preset timer specification to be equal to or less than 15 seconds instead of just less than 15 seconds . )

14.3.1.1.2 Core Excess Reactivity Applicability. This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. It applies for all modes of operation .

Objectives . The objectives that must be simultaneously met are to assure that the reactor has sufficient reactivity to meet its mission requirements , be able to be shut down at any time , and not exceed its fuel temperature safety limit .

Specifications. The maximum available excess reactivity shall not exceed

$7 . 00 .

Basis. This amount of excess reactivity will provide the capability to operate the reactor at full power with experiments in place . Historical operation of the GSTR with an excess reactivity of >$6 . 90 has shown that sufficient 30

shutdown margin exists to provide for safe operation and meet the requirements of the other technical specif~cations .

14 . 3 . 7 . 1 Radiation Monitoring Systems Table 4 - Minimum Radi ati on Moni tori ng Channels Radiation Monit o ring Channe ls Number

• Continuous Air Particulate Radiation Monitor 1 Area Radiation Monitor (fixed or portable) 1

• monitors may be out-of-service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for calibration ,

maintenance , or repair . During this out-of-service time , no experiments or maintenance activities shal: be conducted which could directly result in alarm conditions (e . g . , airborne releases or high radiation levels) 14 . 5 . 1 Site and Faci l i ty Descr iption Applicabi Lity. 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 .

a . The restricted area is that area inside the fence surrounding the reactor portion of the building and the reactor operations area within the building itself . The unrestricted area is that area outside the reactor operations area and the fence surrounding the reactor portion of the b~ilding .

(Note - this change is being made because there is not a fence surrounding all of the reacto r building . The existing fence only surrounds a portion of the building where the reactor is located . There is a fence surrounding the Denver Federal Center , but this fence does not delineate the reactor ' s restricted area.)

14 . 6 . 1 . 3 Staffing

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

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

i . Can be reached by an available communication method within 5 minutes and 31

l l . is capable of getting to the reactor facility within 30 minutes under normal conditions , or iii . is within a 15 mile radius of the reactor facility .

(d) A SRO shall be reachable by any communication method and capable of getting to the reactor facility within 30 minutes under normal conditions or is within a 15 mile radius of the reactor facility .

(e) It is not necessary to have a SRO on call if the Reactor Operator in the control room is a SRO . If the Reactor Operator in the control room is a SRO , a second person shall be available at the facility or on call .

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 relocations 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 ; and (e) recovery from unplanned or unscheduled shutdown or an unscheduled significant power reduction .

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 :

3 . All new experiments or classes of experiments that could cause a reactivity change near a Techn~cal Specification limit or result in the release of radioactivity .

32