ML16200A317

Kansas State University - Response to Request for Additional Information Pertaining to Safety Analysis of 12% U-ZrH Fuel for Use in Kansas State University Triga Reactor

ML16200A317
Person / Time
Site:	Kansas State University
Issue date:	04/28/2014
From:	Geuther J Kansas State University
To:	Office of Nuclear Reactor Regulation
References
Download: ML16200A317 (180)

Text

KANSAS STATE I TRIGA Mk 11 Nuclear u N 1 v E R s 1 T y Reactor Laboratory Additional Information Pertaining to Safety Analysis of 12°/o U-ZrH Fuel for Use in Kansas State University TRI GA Reactor 28 April 2014 To Whom It May Concern, The following information is provided to address questions in the Request for Additional Information issued by the NRC in response to a license amendment request to add four 12%

weight elements to the KSU TRIGA reactor core. The RAJ was dated 3/19/2014, and was issued in response to a license amendment request dated 4/9/2012 and recorded in ADAMS under accession number ML12109A063. The questions from the RAI are repeated in italics, followed by the facility response.

1. NUREG-153 7, Part 1, Section 4.5.2, states in part that "the applicant should discuss the core physics parameters and show the methods and analyses used to determine them. " Your license amendment application provided results of the reactor core modeled using MCNP5 as a JD model by using the code output to compare the "perturbed core" with the "existing core. "

However, no information was provided in the application concerning the calculations from which the results were obtained. Provide a detailed description of the calculations performed, and the input of the MCNP5 code, including information concerning the comparison of the new core operation and performance, with that of the licensed reference core (l.imiting case).

The reactivity calculations (Table 1 in the application) pertaining to the stuck-rod shutdown margin and the maximum all-rods out reactivity were calculated as follows, where "existing" refers to the existing core and "perturbed" refers to the core model with 12% fuel.

A k perturbed - k existmg LJ.p =

k existing Shut down margin (perturbed) = Shut down margin (existing) - tip Max. excess reactivity (perturbed)= Max. excess reactivity (existing) +tip The flux peaking and fission density calculations were based on cell flux (F4) and fission density (F7) tallies. The element to average power density peaking (PP) and flux peaking (FP) were then calculated as follows:

pp= Power Densityi [MeV I g]

' Core Average Power Density [MeV I g]

2 FP = Fluxi [cm* I source particle]

' Core Average Flux [cm*2 I source particle]

Note that the calculation for power density did not account for the difference in mass density between 12% and 8.5% fuel. The 12% fuel is in fact 3.85% more dense than the average 8.5% in the model fuel due to the higher uranium composition. The difference in density can be accounted for as follows:

pp = Power Density; [Me V I g] *Fuel mass, element i [g]

' Core Average Power Density [MeV I g] *Fuel mass per element, core average[g]

However, the maximum calculated power peaking for a 12.5%-loaded fuel element in the E- or F-ring was 1.26. Accounting for differences in mass density, this value wou ld become 1.31 , sti ll far below the maximum B-ring power peaking of 1.63. The maximum fission product source inventory for a 12% element in Ring E or F, which is calculated by multiplying the power peaking times the expected increase in element lifetime relative to an 8.5%-loaded element, would increase from 1.85 to 1.92 if the difference in fuel mass density is accounted for, and therefore would still be wel l below the value of 2.0 assumed in the SAR analysis.

The fuel number densities for the 8.5% fuel were calculated based on the actual fuel depletion records at the facility, which are based on the assumptio n that each element in the core receives the same amount of burn up per kWh of operation. In other words, radial peaking effects are not taken into account to calculate fuel depletion. The mass of 235 U and total uranium mass per element used in the model were therefore based on the initial composition of the fuel when it was installed in the core minus the average burnup per element in the core for each year of operation:

Mass of uranium (g) =Initial mass (g) - Burnup (kWh) . 4.42 x 10-5 _g_

1. fuel elements kWh Mass of 235 U (g) =Initial mass (g)- Burnup (kWh) . 5.25 x io-s _g_

1. fuel elements kWh The constants in the above equations are based on the conversion of 1 gram of fuel to a total number of fissions (accounting for loss due to capture) and the number of fissions per kWh of energy. The database of depleted fuel composition is updated on an annual basis based on the reactor power history. The model did not include fission product poisons. The 12% fuel number density and mass density were based on an element with 12.5 weight percent uranium to account for manufacturing deviations from the 12% nominal value.

The MCNP input decks for the control case and one of the perturbed cases are attached (Attachment 1).

2. NUREG-153 7, Part 1, Section 4.5.2, states in part that the information should include "Technical Specifications that control important design f eatures, limiting conditions of operation, and surveillance requirements as discussed in Chapter 14 of this format and content guide. " The Bases of the attached proposed Technical Specifications (J'S) section 5.1 states that "Calculations show that 12%-loadfuel in the E- and F- rings will not exceed the temperature of 2

8%-load instrumented elements in the B-ring. " The application does not provide calculated values for the temperature and the power of the hottest rod and of the new 12% rods. Provide values of the calculated fuel temperatures and power.

The assertion that the 12%-load fuel in the E- and F-rings will not exceed the temperature of the instrumented elements in the B-ring is based on the fission density as calculated by MCNP. The fission density prov ides the heat source term (power density, q '") for the fuel. T he centerline fuel temperature is based on fuel geometry, power density, the thermal conductivity through the fuel and cladd ing, and the heat transfer coefficient to the coo lant. As ide from power density, the other terms wou ld not significantly change between 8%-loaded elements in the B-ring and 12%-loaded elements in the E- or F-rings.

General Atomics Report GA-8848 , "TRIGA Mk III Reactor Description," (1968), provides experimentally-measured maxim um fue l temperature values for a core fue led w ith 8%-loaded fuel in a circular grid plate pattern, with a B, C, D, E, and F ring, exactly like the KSU TRIGA core. Section 3.4.3 of the GA report states that "The maximum measured fuel temperature at 1.5 MW is ~460 °C wh ich is very much below the temperature limi ts for the ZrHu material." The maximum power allowed by license for the KSU reactor is 1.25 MW, less than the 1.5 MW used to make the GA measurements. Assum ing that the GA measurements were conducted at 20°C, the !:J.T between the fuel and moderator at 1.5 MW was 440°C. Therefore at the maximum moderator temperature allowed in the KSU Techn ical Specifications, 54.4°C, the maximum fuel temperature may be expected to be slightly higher, name ly 494°C.

If 494°C is taken to be the temperature of the element with the highest fission density in the core, and the temperature of the other elements is scaled by the fission density relative to the hottest element (as determ ined by MCNP), the temperature of various elements in the core at 1.5 MW is as fol lows. Note that the TS limit for steady state temperature is 750°C, mu ch hi gher than the maxim um expected fuel temperature of 494°C. Also note that this calculation contains conservatism du e to the use of the peak fue l temperature measured by GA at 1.5 MW, wh ile the KSU reactor wil l be limited to a lower power of :S 1.25 MW.

Expected Maximum Ring Temperature at 1500 kW, °C B 494 c 475 392 -----

316

--~~~--- -----------.

379 258 311

3. NUREG-153 7, Part 1, Section 4.5.1, states in part that "The applicant should give in detail the effects of changes in configuration and fuel burnup. " The application states that "the composition and density of the new core were acijusted to approximate core depletion. " Provide the calculation details and method used to support this statement.

3

Details on the methodology used to adjust the modeled fuel number densities for depletion are given in the response to bullet point 1, above.

4. NUREG-153 7, Part 1, Section 4.5.2, states in part that "The information should include ... The axial and radial distributions of neutron flux densities, justifications for the methods used, and comparison with applicable measurements." Jn figures 2 through 4 of the application, the power of one fuel element in the B-ring is higher than the other elements in the same ring. Explain the origin of this power peaking in the ring and elaborate on the effect of this peaking on the core power distribution.

The fuel composition in the model was based on the actual depleted fuel used in the reactor core; therefore there is variation in the element to average power peaking among the different elements in each ring based on the power history of each fuel element. In the B-ring, the fuel elements differ in 235 U loading by up to approximately 1.6%. The highest-load element in the B-ring had the highest calculated fission density. More importantly, this element (B-4) is directly adjacent to the pulse rod channel. The MCNP model was run with rods withdrawn, so element B-4 was next to what was essentially a water channel, which also caused an increase in the element power density.

5. NUREG- 153 7, Part 1, Section 4.5.1, states in part that "The information in the FSAR should include ... The number, types, and location of all core components on the grid plate, including fuel, control rods " ... The number of the fuel elements presented in the figures of the application are 79 (figure 1) or 78 (figures 2-4), while the number of the fuel elements in the SAR reference core is 83. Explain the apparent difference.

The core configuration considered in the license amendment included 85 fuel elements, which would occur if every lattice location contained a fuel element except for four control rod channels, the central thimble in the center lattice position, and a startup source occupying one lattice position. The figures included in the license amendment application did not display or print correctly, so that the total number of data points was slightly less than 85. Corrected version of the figures are included below.

4

B-Rin g 1.5 - ~e* **e C-Rln g 0 **

• e~e*il a 0-Ring 1.3

. : . iilg

• 1.1 s.. .*

oo

~* *.0 E-Hin g 0 .9 ~

0.7 J

• Control 0 12% Elements in D Ring 0 .5 FIGURE 1- EFFECT OF ADDING FOUR 12 % -LOADED FUEL ELEMENTS TO D-RING ON THE ELEMENT TO AVERAGE FLUX RATIO B-R in g

~ 1.6 e. C-Ri ng

• 0 0-Ring

<( 0 QI 0 u

0 1.4 0 0

.o.

0

... e *

~ 1.2 * .. e. *~ -i 0 E ~ o .*

• E-Ring QI 0 0 w

bQ 1 c

'+:i ta

~ 0.8 c

0

• Control

'iii

.:a 0.6 LI. o 12% Elements in D Ring 0.4 FIGURE 2 - EFF ECT OF ADDING FOUR 12 % -LOADED FU EL EL EMENTS TO D -RING ON THE ELEMENT TO AVERAG E HEA T RAT IO .

5

1.8 B-Rin g 1 .6 ~ C-Ring 0

1.4 - D-Ring E-Ring 1.2 0 0

0 0

1 0

F-Ri ng 0 .8

• Control 0 .6 o 12% Elemen ts in E Ring 0.4 FIGURE 3 - EFFECT OF ADD ING FOUR 12 % - LOADED FUEL ELEMENTS TO E-RING ON THE ELEMENT TO AVERAGE HEAT RATIO.

1.8 -

B-Ring C-Rin g 1.6 *0

~!>~

. 0 1.4 e e ~o.e* D-Ring e

• 0 e e 1.2 E-Ring

~

(1)

r: F-Ring c: 1 0 0

• 0

.~ 0 0 .8 0

• Control

• e e e.,,**

0.6 o 12% El em ents in F Ring 0.4 FIGURE 4- EFFECT OF ADDING FOUR 12 % -LO ADED FU EL ELEMENTS TO F-RING ON THE ELEMENT TO AVERAGE HEAT RATIO .

6

6. NUREG-1537, Part 1, Section 4.5.1, states in part that "The information in the SAR should include... The calculated core reactivity for all core configurations, including the limiting configuration that would yield the highest possible power density. " According to the application "There is some concern that local power peaking effects in 12% fuel would results in unacceptably high local fuel temperatures if the fuel is located near a control rod channel, which becomes a moderator channel upon rod withdrawal". Provide a quantitative analysis on the consequences (fuel element power and temperature) of locating the new fuel elements adjacent to control rods, including defining the criteria on the distances, in order for the above potential event or accident to be evaluated. Also, provide the administrative controls necessary to prevent an unintentional installation of the new fuel in such positions and discuss the basis for whether these controls should be included in the technical specifications.

"Adjacent" positions to control rod channels include positions in the same ring with indices one greater or one less than the control rod position index, as well as the nearest lattice position in the ring immediately interior to the control rod channel and the ring immediately exterior to the control rod channel. For example, in the diagram shown below for an obsolete core configuration, the pulse rod is in lattice position DlO. Therefore 12% fuel would be excluded from C7, D9, Dl 1, and El3. (Note that 12% fuel will only be allowed in the E- and F-rings, so that C7, D9, and Dl l would not be allowed anyway). This prohibition is written into the proposed revision to the Technical Specifications, and will be added to Procedure 10: Fuel Element Inspection and Procedure 26: Fuel Handling Procedure. In control rod positions where two lattice positions in the immediately exterior or interior rings are equidistant from the rod, a total of six positions will be excluded. For example, if a control rod was located at position E2, 12% fuel would not be allowed in F2 or F3.

7

Core 11-19 Figure 5 - ln this example core configuration, 12% elements would be excluded from positions C7, D9, Dll, and El3, which arc immediately adjacent to the pulse rod at 010.

The RAT refers to a statement that there is "some concern that local power peaking effects in 12% fuel would result in unacceptably high local fuel temperatures if the fuel is located near a control rod channel, which becomes a moderator channel upon rod withdrawal." This statement is in reference to information received from other licensees, who indicated that it was difficult to get 12% fuel 1icensed due to concern on the part of the NRC about power peaking near control rod channels. On the other hand, our MCNP calculations show that the effect of power peaking due to the additional moderator when a control rod is withdrawn from an adjacent channel is minimal. The model indicated that a 12.5% element in lattice position E-2, adjacent to a rod channel in the same ring at position E-1 , would have 66.4% of the fission heating rate (and therefore temperature) of the hottest 8.5%-loaded element in the B-ring. If a 12.5% element is loaded into position E-21, adjacent to a rod channel at position D-16, immediately interior to the fuel element, then the element at E-21 will have 69.6% of the fission heating rate of the hottest 8.5%-loaded element in the B-ring. (12.5% elements were used in the model to allow for manufacturing variation in the as-built load of a nominally 12%-loaded element). According to the MCNP model, in both cases, an 8.5% element loaded next to the rod channel would have approximately 10% less fission heating power than the 12.5% element. This calculation demonstrates that the 12%-load fuel , even if immediately adjacent to a control rod channel, would stay well below the temperature indicated by the instrumented 8.5% elements in the B-ring. The Facility elected to exclude 12% elements from these locations, however, in order to avoid a delay in the license amendment, as other licensees indicated that placing 12% fuel near water channels was a point of contention in the past.

8

7. NUREG-1537, Part 1, Section 4.5.1, states in part that "The information in the SAR should include ... Discussion of the safety considerations, including a limiting core configuration that would yield the highest power densities and fuel temperatures achievable with the planned fuel."

The application states that the only accident that can be credibly affected by increasing the fuel loading is the Maximum Hypothetical Accident (MHA). Provide a justification.

The Safety Analysis Report for the KSU TRIGA reactor contains three accidents: release of fission products from a damaged irradiated element (the MHA); the sudden loss of cooling I shielding water when operating at full power accompanied by a SCRAM; and an unexpected insertion of $3.00 ofreactivity from either 10 kW or 104 kW.

The maximum fuel temperature from an unexpected reactivity transient is based on the Fuchs-N ordheim model. The parameters of the Fuchs-Nordheim model are not affected by the use of 12% fuel in the E- or F-rings. In particular, the assumed reactivity insertion of $3.00 is less than the $4.00 excess reactivity relative to reference conditions allowed by the facility TS. Therefore it is set by the TS limit on the maximum allowable pulse reactivity, not the excess reactivity of the core. The element-to-average power peaking assumed in the pulse analysis does affect the peak fuel temperature, but the maximum element to average power peaking is not affected by the use of 12% fuel ifthe fuel is placed in the E- or F-rings, as the B-ring 8.5% elements will still be closest to the element-to-average peaking of 2.0 assumed in the SAR analysis. Therefore the use of 12% fuel does not impact the analysis of an inadvertent insertion of reactivity.

The effect of the total, sudden loss of cooling and shielding water accompanied by a SCRAM when operating at full power is analyzed by considering the maximum temperature of a bare fuel element, considering the decay of fission products as the heat source following shutdown, and the radiological consequences of the absence of shielding water around the core. As with the unanticipated insertion of reactivity, the maximum temperature following this accident would occur in the B-ring, away from the 12% elements in the outer rings, because as we have demonstrated the highest power peaking will still be in the B-ring. The fission product density calculations presented in our submittal demonstrate that the maximum power element-to-average power peaking of 1.64 in the B-ring is much less than the factor of 2.0 assumed in the SAR. The SAR analysis of the radiological consequences of the uncovered core only considers distances of

> 16.5 feet above the core (at the surface of the reactor pool) or greater; therefore the reactor core, with a diameter of 1.5 feet, is essentially a point source at relevant distances. The redistribution of fission products within the core due to moving reactivity to the E- and F-rings in the form of 12% fuel is therefore irrelevant when calculating personnel dose at a distance. The total source term is based on the reactor operating at its full license power level of 1250 kW, so the ability to increase reactivity to get closer to this power level by adding 12% fuel will not increase the actual fission product source term to a value beyond that which is assumed in the SAR analysis.

Therefore the analysis of the loss of coolant I shielding water accident is not affected by the use ofl2% fuel.

8. In the application cover letter, under the title "Request for License Amendment, " there are references to paragraphs 2.D.2 and 4.D.2 of the facility license. There are no such paragraphs in the current license. Provide references to the correct paragraphs.

9

The referenced section of the cover leader for the Request for License Amendment reads:

A detailed safety analysis (attached) demonstrates that fuel loaded to 12%

uranium by weight can be used safely in certain locations in the core (specified in the proposed amendment to the Technical Specifications). We therefore request that paragraph 2.D.2 to License R-88 be removed, and for the possession and use limit of 4.20 kg of fissile material in connection with the operation of the reactor to be increased to 4.55 kg to accommodate the fuel that will no longer be covered under 4.D.2.

This section should have read:

A detailed safety analysis (attached) demonstrates that fuel loaded to 12%

uranium by weight can be used safely in certain locations in the core (specified in the proposed amendment to the Technical Specifications). We therefore request that paragraph 2.B.4 to License R-88 be removed, and for the possession and use limit of 4.20 kg of fissile material in connection with the operation of the reactor in paragraph 2.B.2 to be increased to 4.55 kg to accommodate the fuel that will no longer be covered under paragraph 2.B.4.

9. Provide a marked up copy of the current license and Technical Specifications indicating the proposed changes.

A marked up copy of the current license, with a clean copy of the license incorporating the proposed changes, is included in Attachment 2. A marked up copy of the current Technical Specifications, along with a clean copy of the Technical Specifications incorporating the proposed changes, is included in Attachment 3.

10. Provide a clean copy of the current license and Technical Specifications incorporating the proposed changes.

A marked up copy of the current license, with a clean copy of the license incorporating the proposed changes, is included in Attachment 2. A marked up copy of the current Technical Specifications, along with a clean copy of the Technical Specifications incorporating the proposed changes, is included in Attachment 3.

I swear under penalty of perjury that the foregoing is true and correct.

Sincerely, ey A. Geuther Nuclear Reactor Facilities Manager Kansas State University 10

ATTACHMENT 1 MCNP Input - Control Case c KSU TRIGA Mark II c

c Core with no 12% elements c

c CELLS:

1 0 1 :-20 :38 $ OUTSIDE 2 6 -1 ( ( (-1 2 -38 ) : (-2 7 -38 44 ) : $ POOL ELEMENTS

(-17 18 -38 30 ) : (-18 20 -38 ) : (-7 17 -38 37 ) ) :

(-15 18 -38 43 ) ) #7 #8 3 6 34 33 -7 11 $ RSR SPACE/VOLUME 4 7 -2.699 -35 32 -9 12 #3 $ RSR WALLS 5 2 -1. 6 -36 31 -9 16 #3 #4 #7 #8 $ REFLECTOR GRAPHITE 6 7 -2.699 -37 30 -7 17 #3 #4 #5 #7 #8 $ CANNISTER/WALLS 7 7 -2.699 31 -40 90 -38 26 $ NEBP AL CASE 8 3 -0.001239 31 38 26 $ NEBP CAVITY 9 0 (-30 -7 15 ) : (-44 -2 7 ) fill=? $ CORE SPACE 10 7 -2.699 -15 18 -43 203 $ Lower grid plate 11 0 -15 18 -203 fill=4 $ CT penetration, lower grid plate c ** UNIVERSES c ******************************************************************************

c ** WATER IN CORE AREA: UNIVERSE = (8]

95 6 -1 20 U=8 $ Water above pool bottom 96 7 -2.699 -20 U=8 $ Stuff below pool bottom c ** CONTROL ROD UNIVERSES c ** NOTE: Holes in guide tubes modeled as reduced Al density c (101, 110, 120, 126) c ** PULSE ROD: UNIVERSE = (21]

101 7 -2.12 210 U=21 $ Guide Tube 102 7 -2.699 50 -68 213 -214 U=21 $ Rod cladding 103 11 -2.0 50 213 U=21 $ Pulse rod 104 9 -7.9 68 -217 218 U=21 $ SS extension 105 6 -1 -210 #102 #103 #104 U=21 $ Water c ** SAFETY ROD: UNIVERSE = (22]

110 7 -2.12 210 U=22 $ Guide Tube 111 13 -2.52 51 212 U=22 $ Safety rod 112 9 -7.9 68 -217 218 U=22 $ SS extension 113 6 -1 -210 #111 #112 U=22 $ Water c ** SHIM ROD: UNIVERSE = (23]

120 7 -2.12 210 U=23 $ Guide Tube 121 7 -2.699 52 -68 211 -212 U=23 $ Rod cladding 122 12 -2.48 52 211 U=23 $ Shim/safety rod 123 9 -7.9 68 -217 218 U=23 $ SS extension 124 6 -1 -210 #121 #122 #123 U=23 $ Water c ** REGULATING ROD: UNIVERSE = (24]

126 7 -2.12 219 U=24 $ Guide Tube 127 7 -2.699 53 -68 215 -216 U=24 $ Rod cladding 128 12 -2.48 53 215 U=24 $ Regulating rod 129 9 -7.9 68 218 -217 U=24 $ SS extension 130 6 -1 -219 #127 #128 #129 U=24 $ Water c ******************************************************************************

c ** FUEL UNIVERSES: Individual FE's modeled for burnup as of 3-31-10 c ==============================================================================

20101 6 21 201 U=201 $ Water around lower pin 20102 9 -7.9 201 U=201 $ Bottom pin 20103 9 -7.9 -14 21 U=201 $ BOTTOM SS cap 20104 2 -1. 6 14 202 U=201 $ BOTTOM Axial reflector 20105 9 -7.9 14 -4 202 U=201 $ Cladding 20106 201 -5.685 13 -10 200 -202 U=201 $ ELEMENT B-1 S/N:6315 20107 8 -6.5 13 200 U=201 $ Zirc FILLER 20108 2 -1. 6 10 202 U=201 $ TOP Axial reflector 20109 3 -0.001239 6 202 U=201 $ Air gap 20110 9 -7.9 4 -3 U=201 $ TOP SS cap 20111 9 -7.9 3 -201 U=201 $ Upper pin 20112 6 -1 3 201 U=201 $ Water around pin c ==============================================================================

20201 like 20101 but u=202 11 .

ATTACHMENT I 20202 like 20102 but u=202 20203 like 20103 but U=202 20204 like 20104 but u=202 20205 like 20105 but u=202 20206 like 20106 but mat=202 rho=-5.681 U=202 $ ELEMENT B-2 S/N:l0880 20207 like 20107 but u=202 20208 like 20108 but u=202 20209 like 20109 but u=202 20210 like 20110 but U=202 20211 like 20111 but u=202 20212 like 20112 but u=202 c ==============================================================================

20301 like 20101 but u=203 20302 like 20102 but U=203 20303 like 20103 but u=203 20304 like 20104 but u=203 20305 like 20105 but u=203 20306 like 20106 but mat=203 rho=-5.685 u=203 $ ELEMENT B-3 S/N:6577 20307 like 20107 but u=203 20308 like 20108 but U=203 20309 like 20109 but U=203 20310 like 20110 but u=203 20311 like 20111 but u=203 20312 like 20112 but u=203 c ==============================================================================

20401 like 20101 but u=204 20402 like 20102 but u=204 20403 like 20103 but U=204 20404 like 20104 but u=204 20405 like 20105 but u=204 20406 like 20106 but mat=204 rho=-5.688 u=204 $ ELEMENT B-4 S/N:2966 20407 like 20107 but u=204 20408 like 20108 but U=204 20409 like 20109 but u=204 20410 like 20110 but u=204 20411 like 20111 but u=204 20412 like 20112 but u=204 c ==============================================================================

20501 like 20101 but u=205 20502 like 20102 but u=205 20503 like 20103 but u=205 20504 like 20104 but u=205 20505 like 20105 but u=205 20506 like 20106 but mat=205 rho=-5.684 u=205 $ ELEMENT B-5 S/N:l0707 20507 like 20107 but u=205 20508 like 20108 but u=205 20509 like 20109 but u=205 20510 like 20110 but U=205 20511 like 20111 but u=205 20512 like 20112 but u=205 c ==============================================================================

20601 like 20101 but U=206 20602 like 20102 but u=206 20603 like 20103 but u=206 20604 like 20104 but u=206 20605 like 20105 but u=206 20606 like 20106 but mat=206 rho=-5.684 u=206 $ ELEMENT B-6 S/N:6525 20607 like 20107 but u=206 20608 like 20108 but u=206 20609 like 20109 but u=206 20610 like 20110 but U=206 20611 like 20111 but u=206 20612 like 20112 but u=206 c ==============================================================================

30101 like 20101 but u=301 30102 like 20102 but u=301 30103 like 20103 but u=301 30104 like 20104 but u=301 30105 like 20105 but u=301 30106 like 20106 but mat=301 rho=-5.680 u=301 $ ELEMENT C-1 S/N:ll351 30107 like 20107 but u=301 12

ATTACHMENT I 30108 like 20108 but u=301 30109 like 20109 but u=301 30110 like 20110 but u=301 30111 like 20111 but u=301 30112 like 20112 but u=301 c ==============================================================================

30201 like 20101 but u=302 30202 like 20102 but u=302 30203 like 20103 but u=302 30204 like 20104 but u=302 30205 like 20105 but u=302 30206 like 20106 but mat=302 rho=-5.684 u=302 $ ELEMENT C-2 S/N:6316 30207 like 20107 but u=302 30208 like 20108 but u=302 30209 like 20109 but u=302 30210 like 20110 but u=302 30211 like 20111 but u=302 30212 like 20112 but u=302 c ==============================================================================

30301 like 20101 but u=303 30302 like 20102 but u=303 30303 like 20103 but u=303 30304 like 20104 but u=303 30305 like 20105 but u=303 30306 like 20106 but mat=303 rho=-5.691 u=303 $ ELEMENT C-3 S/N:2963 30307 like 20107 but u=303 30308 like 20108 but u=303 30309 like 20109 but u=303 30310 like 20110 but u=303 30311 like 20111 but U=303 30312 like 20112 but u=303 c ==============================================================================

30401 like 20101 but u=304 30402 like 20102 but u=304 30403 like 20103 but u=304 30404 like 20104 but u=304 30405 like 20105 but u=304 30406 like 20106 but mat=304 rho=-5.638 u=304 $ ELEMENT C-4 S/N:3329 30407 like 20107 but u=304 30408 like 20108 but u=304 30409 like 20109 but U=304 30410 like 20110 but u=304 30411 like 20111 but u=304 30412 like 20112 but U=304 c ==============================================================================

30501 like 20101 but u=305 30502 like 20102 but u=305 30503 like 20103 but u=305 30504 like 20104 but u=305 30505 like 20105 but u=305 30506 like 20106 but mat=305 rho=-5.710 u=305 $ ELEMENT C-5 S/N:2953 30507 like 20107 but u=305 30508 like 20108 but u=305 30509 like 20109 but u=305 30510 like 20110 but u=305 30511 like 20111 but U=305 30512 like 20112 but u=305 c ==============================================================================

30601 like 20101 but u=306 30602 like 20102 but U=306 30603 like 20103 but u=306 30604 like 20104 but u=306 30605 like 20105 but u=306 30606 like 20106 but mat=306 rho=-5.649 u=306 $ ELEMENT C-6 S/N:3082 30607 like 20107 but u=306 30608 like 20108 but u=306 30609 like 20109 but u=306 30610 like 20110 but u=306 30611 like 20111 but u=306 30612 like 20112 but u=306 c ==============================================================================

13

ATTACHMENT I 30801 like 20101 but u=308 30802 like 20102 but u=308 30803 like 20103 but u=308 30804 like 20104 but u=308 30805 like 20105 but U=308 30806 like 20106 but mat=308 rho=-5.693 u=308 $ ELEMENT C-8 S/N:2933 30807 like 20107 but u=308 30808 like 20108 but u=308 30809 like 20109 but u=308 30810 like 20110 but u=308 30811 like 20111 but u=308 30812 like 20112 but u=308 c ==============================================================================

30901 like 20101 but u=309 30902 like 20102 but u=309 30903 like 20103 but u=309 30904 like 20104 but U=309 30905 like 20105 but U=309 30906 like 20106 but mat=309 rho=-5.684 u=309 $ ELEMENT C-9 S/N:6314 30907 like 20107 but u=309 30908 like 20108 but u=309 30909 like 20109 but u=309 30910 like 20110 but U=309 30911 like 20111 but U=309 30912 like 20112 but U=309 c ==============================================================================

31001 like 20101 but u=310 31002 like 20102 but u=310 31003 like 20103 but u=310 31004 like 20104 but u=310 31005 like 20105 but u=310 31006 like 20106 but mat=310 rho=-5.684 u=310 $ ELEMENT C-10 S/N:6527 31007 like 20107 but u=310 31008 like 20108 but u=310 31009 like 20109 but u=310 31010 like 20110 but u=310 31011 like 20111 but u=310 31012 like 20112 but u=310 c ==============================================================================

_31101 like 20101 but u=311 31102 like 20102 but u=311 31103 like 20103 but u=311 31104 like 20104 but u=311 31105 like 20105 but u=311 31106 like 20106 but mat=311 rho=-5.684 U=311 $ ELEMENT C-11 S/N:6317 31107 like 20107 but u=311 31108 like 20108 but U=311 31109 like 20109 but u=311 31110 like 20110 but u=311 31111 like 20111 but U=311 31112 like 20112 but u=311 c ==============================================================================

31201 like 20101 but U=312 31202 like 20102 but u=312 31203 like 20103 but u=312 31204 like 20104 but u=312 31205 like 20105 but u=312 31206 like 20106 but mat=312 rho=-5.684 u=312 $ ELEMENT C-12 S/N:6526 31207 like 20107 but u=312 31208 like 20108 but u=312 31209 like 20109 but U=312 31210 like 20110 but u=312 31211 like 20111 but U=312 31212 like 20112 but u=312 c ==============================================================================

40101 like 20101 but u=401 40102 like 20102 but u=401 40103 like 20103 but u=401 40104 like 20104 but u=401 40105 like 20105 but u=401 40106 like 20106 but mat=401 rho=-5.637 u=401 $ ELEMENT D-1 S/N:3380 14

ATTACHMENT I 40107 like 20107 but u=401 40108 like 20108 but u=401 40109 like 20109 but u=401 40110 like 20110 but u=401 40111 like 20111 but u=401 40112 like 20112 but U=401 c ==============================================================================

40201 like 20101 but u=402 40202 like 20102 but u=402 40203 like 20103 but U=402 40204 like 20104 but u=402 40205 like 20105 but u=402 40206 like 20106 but mat=402 rho=-5.655 u=402 $ ELEMENT D-2 S/N:3330 40207 like 20107 but U=402 40208 like 20108 but u=402 40209 like 20109 but u=402 40210 like 20110 but u=402 40211 like 20111 but u=402 40212 like 20112 but u=402 c ==============================================================================

40301 like 20101 but u=403 40302 like 20102 but U=403 40303 like 20103 but u=403 40304 like 20104 but u=403 40305 like 20105 but u=403 40306 like 20106 but mat=403 rho=-5.626 u=403 $ ELEMENT D-3 S/N:5001 40307 like 20107 but u=403 40308 like 20108 but U=403 40309 like 20109 but u=403 40310 like 20110 but u=403 40311 like 20111 but U=403 40312 like 20112 but u=403 c ==============================================================================

40501 like 20101 but u=405 40502 like 20102 but u=405 40503 like 20103 but u=405 40504 like 20104 but u=405 40505 like 20105 but U=405 40506 like 20106 but mat=405 rho=-5.636 u=405 $ ELEMENT D-5 S/N:3144 40507 like 20107 but U=405 40508 like 20108 but U=405 40509 like 20109 but U=405 40510 like 20110 but U=405 40511 like 20111 but U=405 40512 like 20112 but u=405 c ==============================================================================

40601 like 20101 but u=406 40602 like 20102 but u=406 40603 like 20103 but u=406 40604 like 20104 but u=406 40605 like 20105 but u=406 40606 like 20106 but mat=406 rho=-5.646 u=406 $ ELEMENT D-6 S/N 6224 40607 like 20107 but U=406 40608 like 20108 but u=406 40609 like 20109 but U=406 40610 like 20110 but U=406 40611 like 20111 but u=406 40612 like 20112 but u=406 c ==============================================================================

40701 like 20101 but u=407 40702 like 20102 but U=407 40703 like 20103 but u=407 40704 like 20104 but U=407 40705 like 20105 but u=407 40706 like 20106 but mat=407 rho=-5.646 U=407 $ ELEMENT D-7 S/N:3018 40707 like 20107 but u=407 40708 like 20108 but u=407 40709 like 20109 but u=407 40710 like 20110 but u=407 40711 like 20111 but U=407 40712 like 20112 but U=407 15

ATTACHMENT I c ==============================================================================

40801 like 20101 but u=408 40802 like 20102 but u=408 40803 like 20103 but u=408 40804 like 20104 but u=408 40805 like 20105 but u=408 40806 like 20106 but rnat=408 rho=-5.636 u=408 $ ELEMENT D-8 S/N:3105 40807 like 20107 but u=408 40808 like 20108 but u=408 40809 like 20109 but U=408 40810 like 20110 but u=408 40811 like 20111 but u=408 40812 like 20112 but U=408 c ==============================================================================

40901 like 20101 but u=409 40902 like 20102 but U=409 40903 like 20103 but u=409 40904 like 20104 but u=409 40905 like 20105 but u=409 40906 like 20106 but*rnat=409 rho=-5.642 u=409 $ ELEMENT D-9 S/N 2452 40907 like 20107 but u=409 40908 like 20108 but u=409 40909 like 20109 but u=409 40910 like 20110 but U=409 40911 like 20111 but u=409 40912 like 20112 but u=409 c ==============================================================================

41001 like 20101 but u=410 41002 like 20102 but u=410 41003 like 20103 but u=410 41004 like 20104 but u=410 41005 like 20105 but U=410 41006 like 20106 but rnat=410 rho=-5.632 u=410 $ ELEMENT D-10 S/N:2448 41007 like 20107 but u=410 41008 like 2oio0 but u=410 41009 like 20109 but u=410 41010 like 20110 but u=410 41011 like 20111 but u=410 41012 like 20112 but u=410 c ==============================================================================

41101 like 20101 but u=411 41102 like 20102 but u=411 41103 like 20103 but u=411 41104 like 20104 but u=411 41105 like 20105 but u=411 41106 like 20106 but rnat=411 rho=-5.641 u=411 $ ELEMENT D-11 S/N:2435 41107 like 20107 but U=411 41108 like 20108 but u=411 41109 like 20109 but U=411 41110 like 20110 but u=411 41111 like 20111 but u=411 41112 like 20112 but U=411 c ==============================================================================

41201 like 20101 but u=412 41202 like 20102 but U=412 41203 like 20103 but u=412 41204 like 20104 but u=412 41205 like 20105 but u=412 41206 like 20106 but rnat=412 rho=-5.655 u=412 $ ELEMENT D-12 S/N 3876 41207 like 20107 but u=412 41208 like 20108 but U=412 41209 like 20109 but u=412 41210 like 20110 but U=412 41211 like 20111 but u=412 41212 like 20112 but U=412 c ==============================================================================

41301 like 20101 but u=413 41302 like 20102 but u=413 41303 like 20103 but u=413 41304 like 20104 but U=413 41305 like 20105 but U=413 16

ATTACHMENT I 41306 like 20106 but mat=413 rho=-5.644 U=413 $ ELEMENT D-13 S/N:3696 41307 like 20107 but U=413 41308 like 20108 but u=413 41309 like 20109 but u=413 41310 like 20110 but u=413 41311 like 20111 but u=413 41312 like 20112 but u=413 c ==============================================================================

41401 like 20101 but u=414 41402 like 20102 but u=414 41403 like 20103 but u=414 41404 like 20104 but u=414 41405 like 20105 but u=414 41406 like 20106 but mat=414 rho=-5.669 u=414 $ ELEMENT D-14 S/N:3494 41407 like 20107 but u=414 41408 like 20108 but U=414 41409 like 20109 but u=414 41410 like 20110 but u=414 41411 like 20111 but u=414 41412 like 20112 but u=414 c ==============================================================================

41501 like 20101 but u=415 41502 like 20102 but u=415 41503 like 20103 but u=415 41504 like 20104 but u=415 41505 like 20105 but u=415 41506 like 20106 but mat=415 rho=-5.656 u=415 $ ELEMENT D-15 S/N:3501 41507 like 20107 but u=415 41508 like 20108 but u=415 41509 like 20109 but U=415 41510 like 20110 but u=415 41511 like 20111 but U=415 41512 like 20112 but U=415 c ==============================================================================

41701 like 20101 but U=417 41702 like 20102 but U=417 41703 like 20103 but U=417 41704 like 20104 but u=417 41705 like 20105 but U=417 41706 like 20106 but mat=417 rho=-5.656 U=417 $ ELEMENT D-17 S/N:3498 41707 like 20107 but U=417 41708 like 20108 but U=417 41709 like 20109 but U=417 41710 like 20110 but u=417 41711 like 20111 but U=417 41712 like 20112 but u=417 c ==============================================================================

41801 like 20101 but u=418 41802 like 20102 but u=418 41803 like 20103 but u=418 41804 like 20104 but u=418 41805 like 20105 but u=418 41806 like 20106 but mat=418 rho=-5.643 u=418 $ ELEMENT D-18 S/N 3336 41807 like 20107 but u=418 41808 like 20108 but u=418 41809 like 20109 but U=418 41810 like 20110 but U=418 41811 like 20111 but u=418 41812 like 20112 but u=418 c ==============================================================================

50201 like 20101 but u=502 50202 like 20102 but U=502 50203 like 20103 but u=502 50204 like 20104 but u=502 50205 like 20105 but u=502 50206 like 20106 but mat=502 rho=-5.687 U=502 $ ELEMENT E-2 S/N:2986 50207 like 20107 but u=502 50208 like 20108 but u=502 50209 like 20109 but U=502 50210 like 20110 but u=502 50211 like 20111 but u=502 17

ATTACHMENTl 50212 like 20112 but u=502 c ==============================================================================

50301 like 20101 but u=503 50302 like 20102 but U=503 50303 like 20103 but u=503 50304 like 20104 but u=503 50305 like 20105 but u=503 50306 like 20106 but mat=503 rho=-5.632 u=503 $ ELEMENT E-3 S/N:2458 50307 like 20107 but u=503 50308 like 20108 but u=503 50309 like 20109 but u=503 50310 like 20110 but u=503 50311 like 20111 but U=503 50312 like 20112 but u=503 c ==============================================================================

50401 like 20101 but u=504 50402 like 20102 but U=504 50403 like 20103 but u=504 50404 like 20104 but u=504 50405 like 20105 but u=504 50406 like 20106 but mat=504 rho=-5.642 U=504 $ ELEMENT E-4 S/N:3006 50407 like 20107 but u=504 50408 like 20108 but u=504 50409 like 20109 but U=504 50410 like 20110 but u=504 50411 like 20111 but u=504 50412 like 20112 but U=504 c ==============================================================================

50501 like 20101 but u=505 50502 like 20102 but u=505 50503 like 20103 but U=505 50504 like 20104 but u=505 50505 like 20105 but U=505 50506 like 20106 but mat=505 rho=-5.625 u=505 $ ELEMENT E-5 S/N:5014 50507 like 20107 but u=505 50508 like 20108 but u=505 50509 like 20109 but u=505 50510 like 20110 but u=505 50511 like 20111 but u=505 50512 like 20112 but u=505 c ==============================================================================

50601 like 20101 but u=506 50602 like 20102 but u=506 50603 like 20103 but u=506 50604 like 20104 but u=506 50605 like 20105 but u=506 50606 like 20106 but mat=506 rho=-5.627 u=506 $ ELEMENT E-6 S/N:4744 50607 like 20107 but u=506 50608 like 20108 but u=506 50609 like 20109 but u=506 50610 like 20110 but u=506 50611 like 20111 but u=506 50612 like 20112 but U=506 c ==============================================================================

50701 like 20101 but u=507 50702 like 20102 but U=507 50703 like 20103 but U=507 50704 like 20104 but u=507 50705 like 20105 but U=507 50706 like 20106 but mat=507 rho=-5.620 U=507 $ ELEMENT E-7 S/N:3147 50707 like 20107 but u=507 50708 like 20108 but U=507 50709 like 20109 but u=507 50710 like 20110 but u=507 50711 like 20111 but U=507 50712 like 20112 but U=507 c ==============================================================================

50801 like 20101 but u=508 50802 like 20102 but u=508 50803 like 20103 but u=508 50804 like 20104 but u=508 18

A1TACHMENT1 50805 like 20105 but U=508 50806 like 20106 but mat=508 rho=-5.624 u=508 $ ELEMENT E-8 S/N:4991 50807 like 20107 but u=508 50808 like 20108 but u=508 50809 like 20109 but u=508 50810 like 20110 but u=508 50811 like 20111 but u=508 50812 like 20112 but u=508 c ==============================================================================

50901 like 20101 but u=509 50902 like 20102 but u=509 50903 like 20103 but u=509 50904 like 20104 but u=509 50905 like 20105 but u=509 50906 like 20106 but mat=509 rho=-5.588 u=509 $ ELEMENT E-9 S/N:4742 50907 like 20107 but U=509 50908 like 20108 but u=509 50909 like 20109 but u=509 50910 like 20110 but U=509 50911 like 20111 but u=509 50912 like 20112 but U=509 c ==============================================================================

51001 like 20101 but u=510 51002 like 20102 but u=510 51003 like 20103 but u=510 51004 like 20104 but u=510 51005 like 20105 but u=510 51006 like 20106 but mat=510 rho=-5.617 u=510 $ ELEMENT E-10 S/N:4351 51007 like 20107 but u=510 51008 like 20108 but U=510 51009 like 20109 but u=510 51010 like 20110 but U=510 51011 like 20111 but U=510 51012 like 20112 but u=510 c ==============================================================================

51101 like 20101 but U=511 51102 like 20102 but U=511 51103 like 20103 but U=511 51104 like 20104 but u=511 51105 like 20105 but u=511 51106 like 20106 but mat=511 rho=-5.631 u=511 $ ELEMENT E-11 S/N:3107 51107 like 20107 but u=511 51108 like 20108 but u=511 51109 like 20109 but u=511 51110 like 20110 but u=511 51111 like 20111 but u=511 51112 like 20112 but u=511 c ==============================================================================

51201 like 20101 but U=512 51202 like 20102 but u=512 51203 like 20103 but U=512 51204 like 20104 but u=512 51205 like 20105 but U=512 51206 like 20106 but mat=512 rho=-5.656 U=512 $ ELEMENT E-12 S/N:3690 51207 like 20107 but U=512 51208 like 20108 but u=512 51209 like 20109 but u=512 51210 like 20110 but u=512 51211 like 20111 but U=512 51212 like 20112 but U=512 c ==============================================================================

51301 iike 20101 but U=513 51302 like 20102 but u=513 51303 like 20103 but u=513 51304 like 20104 but u=513 51305 like 20105 but u=513 51306 like 20106 but mat=513 rho=-5.693 U=513 $ ELEMENT E-13 S/N:2987 51307 like 20107 but u=513 51308 like 20108 but u=513 51309 like 20109 but u=513 51310 like 20110 but u=513 19

ATTACHMENT I 51311 like 20111 but u=513 51312 like 20112 but u=513 c ==============================================================================

51401 like 20101 but u=514 51402 like 20102 but u=514 51403 like 20103 but u=514 51404 like 20104 but u=514 51405 like 20105 but u=514 51406 like 20106 but mat=514 rho=-5.636 u=514 $ ELEMENT E-14 S/N:3118 51407 like 20107 but U=514 51408 like 20108 but U=514 51409 like 20109 but u=514 51410 like 20110 but u=514 51411 like 20111 but u=514 51412 like 20112 but u=514 c ==============================================================================

51501 like 20101 but u=515 51502 like 20102 but u=515 51503 like 20103 but u=515 51504 like 20104 but U=515 51505 like 20105 but u=515 51506 like 20106 but mat=515 rho=-5.692 u=515 $ ELEMENT E-15 S/N:2934 51507 like 20107 but u=515 51508 like 20108 but u=515 51509 like 20109 but u=515 51510 like 20110 but u=515 51511 like 20111 but u=515 51512 like 20112 but U=515 c ==============================================================================

51601 like 20101 but u=516 51602 like 20102 but u=516 51603 like 20103 but u=516 51604 like 20104 but u=516 51605 like 20105 but u=516 51606 like 20106 but mat=516 rho=-5.619 u=516 $ ELEMENT E-16 S/N:4343 51607 like 20107 but u=516 51608 like 20108 but u=516 51609 like 20109 but u=516 51610 like 20110 but u=516 51611 like 20111 but u=516 51612 like 20112 but u=516 c ==============================================================================

51701 like 20101 but u=517 51702 like 20102 but u=517 51703 like 20103 but u=517 51704 like 20104 but u=517 51705 like 20105 but u=517 51706 like 20106 but mat=517 rho=-5.670 u=517 $ ELEMENT E-17 S/N:3517 51707 like 20107 but U=517 51708 like 20108 but u=517 51709 like 20109 but u=517 51710 like 20110 but u=517 51711 like 20111 but u=517 51712 like 20112 but u=517 c ==============================================================================

51801 like 20101 but U=518 51802 like 20102 but u=518 51803 like 20103 but u=518 51804 like 20104 but u=518 51805 like 20105 but u=518 51806 like 20106 but mat=518 rho=-5.658 u=518 $ ELEMENT E-18 S/N:3502 51807 like 20107 but u=518 51808 like 20108 but U=518 51809 like 20109 but u=518 51810 like 20110 but u=518 51811 like 20111 but u=518 51812 like 20112 but U=518 c ==============================================================================

51901 like 20101 but U=519 51902 like 20102 but U=519 51903 like 20103 but u=519 20

ATTACHMENTl 51904 like 20104 but u=519 51905 like 20105 but u=519 51906 like 20106 but mat=519 rho=-5.657 u=519 $ ELEMENT E-19 S/N:3503 51907 like 20107 but u=519 51908 like 20108 but u=519 51909 like 20109 but u=519 51910 like 20110 but u=519 51911 like 20111 but u=519 51912 like 20112 but u=519 c ==============================================================================

52001 like 20101 but u=520 52002 like 20102 but u=520 52003 like 20103 but u=520 52004 like 20104 but u=520 52005 like 20105 but u=520 52006 like 20106 but mat=520 rho=-5.649 u=520 $ ELEMENT E-20 S/N:3009 52007 like 20107 but u=520 52008 like 20108 but u=520 52009 like 20109 but u=520 52010 like 20110 but u=520 52011 like 20111 but u=520 52012 like 20112 but U=520 c ==============================================================================

52101 like 20101 but u=521 52102 like 20102 but u=521 52103 like 20103 but U=521 52104 like 20104 but u=521 52105 like 20105 but U=521 52106 like 20106 but mat=521 rho=-5.645 u=521 $ ELEMENT E-21 S/N:3011 52107 like 20107 but u=521 52108 like 20108 but U=521 52109 like 20109 but u=521 52110 like 20110 but u=521 52111 like 20111 but u=521 52112 like 20112 but u=521 c ==============================================================================

52201 like 20101 but U=522 52202 like 20102 but U=522 52203 like 20103 but U=522 52204 like 20104 but u=522 52205 like 20105 but U=522 52206 like 20106 but mat=522 rho=-5.646 u=522 $ ELEMENT E-22 S/N:3014 52207 like 20107 but u=522 52208 like 20108 but u=522 52209 like 20109 but u=522 52210 like 20110 but U=522 52211 like 20111 but u=522 52212 like 20112 but u=522 c ==============================================================================

52301 like 20101 but u=523 52302 like 20102 but U=523 52303 like 20103 but u=523 52304 like 20104 but u=523 52305 like 20105 but U=523 52306 like 20106 but mat=523 rho=-5.641 u=523 $ ELEMENT E-23 S/N:3008 52307 like 20107 but u=523 52308 like 20108 but U=523 52309 like 20109 but u=523 52310 like 20110 but U=523 52311 like 20111 but U=523 52312 like 20112 but u=523 c ==============================================================================

52401 like 20101 but u=524 52402 like 20102 but u=524 52403 like 20103 but u=524 52404 like 20104 but u=524 52405 like 20105 but u=524 52406 like 20106 but mat=524 rho=-5.636 u=524 $ ELEMENT E-24 S/N:3111 52407 like 20107 but u=524 52408 like 20108 but U=524 52409 like 20109 but u=524 21

ATTACHMENT I 52410 like 20110 but u=524 52411 like 20111 but u=524 52412 like 20112 but u=524 c ==============================================================================

60101 like 20101 but u=601 60102 like 20102 but u=601 60103 like 20103 but u=601 60104 like 20104 but u=601 60105 like 20105 but u=601 60106 like 20106 but mat=601 rho=-5.612 u=601 $ ELEMENT F-1 S/N:5017 60107 like 20107 but u=601 60108 like 20108 but U=601 60109 like 20109 but u=601 60110 like 20110 but u=601 60111 like 20111 but U=601 60112 like 20112 but u=601 c ==============================================================================

60201 like 20101 but u=602 60202 like 20102 but u=602 60203 like 20103 but u=602 60204 like 20104 but u=602 60205 like 20105 but U=602 60206 like 20106 but mat=602 rho=-5.613 u=602 $ ELEMENT F-2 S/N:5018 60207 like 20107 but u=602 60208 like 20108 but u=602 60209 like 20109 but u=602 60210 like 20110 but u=602 60211 like 20111 but u=602 60212 like 20112 but u=602 c ==============================================================================

60301 like 20101 but u=603 60302 like 20102 but u=603 60303 like 20103 but u=603 60304 like 20104 but u=603 60305 like 20105 but u=603 60306 like 20106 but mat=603 rho=-5.626 u=603 $ ELEMENT F-3 S/N:5027 60307 like 20107 but u=603 60308 like 20108 but u=603 60309 like 20109 .but U=603 60310 like 20110 but U=603 60311 like 20111 but U=603 60312 like 20112 but U=603 c ==============================================================================

60401 like 20101 but u=604 60402 like 20102 but U=604 60403 like 20103 but U=604 60404 like 20104 but u=604 60405 like 20105 but u=604 60406 like 20106 but mat=604 rho=-5.613 u=604 $ ELEMENT F-4 S/N:5021 60407 like 20107 but u=604 60408 like 20108 but u=604 60409 like 20109 but u=604 60410 like 20110 but U=604 60411 like 20111 but u=604 60412 like 20112 but U=604 c ==============================================================================

60501 like 20101 but U=605 60502 like 20102 but U=605 60503 like 20103 but u=605 60504 like 20104 but u=605 60505 like 20105 but u=605 60506 like 20106 but mat=605 rho=-5.613 u=605 $ ELEMENT F-5 S/N:5026 60507 like 20107 but u=605 60508 like 20108 but u=605 60509 like 20109 but U=605 60510 like 20110 but u=605 60511 like 20111 but U=605 60512 like 20112 but U=605 c ==============================================================================

60601 like 20101 but U=606 60602 like 20102 but u=606 22

ATTACHMENT I 60603 like 20103 but u=606 60604 like 20104 but u=606 60605 like 20105 but u=606 60606 like 20106 but mat=606 rho=-5.697 U=606 $ ELEMENT F-6 S/N:6223 60607 like 20107 but u=606 60608 like 20108 but u=606 60609 like 20109 but u=606 60610 like 20110 but u=606 60611 like 20111 but u=606 60612 like 20112 but u=606 c ==============================================================================

60701 like 20101 but u=607 60702 like 20102 but u=607 60703 like 20103 but u=607 60704 like 20104 but u=607 60705 like 20105 but u=607 60706 like 20106 but mat=607 rho=-5.605 U=607 $ ELEMENT F-7 S/N:5256 60707 like 20107 but u=607 60708 like 20108 but U=607 60709 like 20109 but u=607 60710 like 20110 but u=607 60711 like 20111 but U=607 60712 like 20112 but u=607 c ==============================================================================

60801 like 20101 but u=608 60802 like 20102 but u=608 60803 like 20103 but U=608 60804 like 20104 but u=608 60805 like 20105 but u=608 60806 like 20106 but mat=608 rho=-5.613 u=608 $ELEMENT F-8 _S/N:5254 60807 like 20107 but U=608 60808 like 20108 but u=608 60809 like 20109 but u=608 60810 like 20110 but u=608 60811 like 20111 but u=608 60812 like 20112 but u=608 c ==============================================================================

60901 like 20101 but u=609 60902 like 20102 but u=609 60903 like 20103 but u=609 60904 like 20104 but u=609 60905 like 20105 but U=609 60906 like 20106 but mat=609 rho=-5.613 u=609 $ ELEMENT F-9 S/N:5031 60907 like 20107 but u=609 60908 like 20108 but u=609 60909 like 20109 but u=609 60910 like 20110 but u=609 60911 like 20111 but u=609 60912 like 20112 but U=609 c ==============================================================================

61101 like 20101 but u=611 61102 like 20102 but u=611 61103 like 20103 but u=611 61104 like 20104 but u=611 61105 like 20105 but U=611 61106 like 20106 but mat=611 rho=-5.599 u=611 $ ELEMENT F-11 S/N:5949 61107 like 20107 but u=611 61108 like 20108 but U=611 61109 like 20109 but U=611 61110 like 20110 but u=611 61111 like 20111 but u=611 61112 like 20112 but U=611 c ==============================================================================

61201 like 20101 but u=612 61202 like 20102 but u=612 61203 like 20103 but u=612 61204 like 20104 but u=612 61205 like 20105 but u=612 61206 like 20106 but mat=612 rho=-5.703 U=612 $ ELEMENT F-12 S/N:2900 61207 like 20107 but U=612 61208 like 20108 but u=612 23

ATTACHMENT I 61209 like 20109 but u=612 61210 like 20110 but u=612 61211 like 20111 but u=612 61212 like 20112 but u=612 c ==============================================================================

61301 like 20101 but u=613 61302 like 20102 but u=613 61303 like 20103 but u=613 61304 like 20104 but u=613 61305 like 20105 but U=613 61306 like 20106 but mat=613 rho=-5.625 U=613 $ ELEMENT F-13 S/N:4339 61307 like 20107 but u=613 61308 like 20108 but u=613 61309 like 20109 but u=613 61310 like 20110 but U=613 61311 like 20111 but u=613 61312 like 20112 but U=613 c ==============================================================================

61401 like 20101 but u=614 61402 like 20102 but u=614 61403 like 20103 but u=614 61404 like 20104 but U=614 61405 like 20105 but u=614 61406 like 20106 but mat=614 rho=-5.600 U=614 $ ELEMENT F-14 S/N:5653 61407 like 20107 but u=614 61408 like 20108 but u=614 61409 like 20109 but u=614 61410 like 20110 but u=614 61411 like 20111 but u=614 61412 like 20112 but u=614 c ==============================================================================

61501 like 20101 but U=615 61502 like 20102 but u=615 61503 like 20103 but u=615 61504 like 20104 but u=615 61505 like 20105 but U=615 61506 like 20106 but mat=615 rho=-5.606 u=615 $ ELEMENT F-15 S/N:5654 61507 like 20107 but u=615 61508 like 20108 but u=615 61509 like 20109 but u=615 61510 like 20110 but u=615 61511 like 20111 but u=615 61512 like 20112 but U=615 c ==============================================================================

61601 like 20101 but u=616 61602 like 20102 but U=616 61603 like 20103 but u=616 61604 like 20104 but u=616 61605 like 20105 but u=616 61606 like 20106 but mat=616 rho=-5.599 u=616 $ ELEMENT F-16 S/N:5655 61607 like 20107 but u=616 61608 like 20108 but u=616 61609 like 20109 but u=616 61610 like 20110 but u=616 61611 like 20111 but u=616 61612 like 20112 but u=616 c ==============================================================================

61701 like 20101 but u=617 61702 like 20102 but u=617 61703 like 20103 but u=617 61704 like 20104 but U=617 61705 like 20105 but u=617 61706 like 20106 but mat=617 rho=-5.596 u=617 $ ELEMENT F-17 S/N:5939 61707 like 20107 but u=617 61708 like 20108 but u=617 61709 like 20109 but u=617 61710 like 20110 but u=617 61711 like 20111 but u=617 61712 like 20112 but u=617 c ==============================================================================

61801 like 20101 but u=618 24

ATTACHMENT I 61802 like 20102 but u=618 61803 like 20103 but U=618 61804 like 20104 but u=618 61805 like 20105 but u=618 61806 like 20106 but mat=618 rho=-5.599 u=618 $ ELEMENT F-18 S/N:5946 61807 like 20107 but u=618 61808 like 20108 but u=618 61809 like 20109 but U=618 61810 like 20110 but U=618 61811 like 20111 but U=618 61812 like 20112 but u=618 c ==============================================================================

61901 like 20101 but u=619 61902 like 20102 but u=619 61903 like 20103 but u=619 61904 like 20104 but u=619 61905 like 20105 but u=619 61906 like 20106 but mat=619 rho=-5.634 u=619 $ ELEMENT F-19 S/N:3113 61907 like 20107 but u=619 61908 like 20108 but u=619 61909 like 20109 but U=619 61910 like 20110 but u=619 61911 like 20111 but u=619 61912 like 20112 but u=619 c ==============================================================================

62001 like 20101 but u=620 62002 like 20102 but u=620 62003 like 20103 but u=620 62004 like 20104 but u=620 62005 like 20105 but U=620 62006 like 20106 but mat=620 rho=-5.691 u=620 $ ELEMENT F-20 S/N:2949 62007 like 20107 but U=620 62008 like 20108 but u=620 62009 like 20109 but U=620 62010 like 20110 but u=620 62011 like 20111 but u=620 62012 like 20112 but u=620 c ==============================================================================

62101 like 20101 but u=621 62102 like 20102 but u=621 62103 like 20103 but u=621 62104 like 20104 but u=621 62105 like 20105 but u=621 62106 like 20106 but mat=621 rho=-5.612 u=621 $ ELEMENT F-21 S/N:5649 62107 like 20107 but u=621 62108 like 20108 but U=621 62109 like 20109 but u=621 62110 like 20110 but u=621 62111 like 20111 but U=621 62112 like 20112 but u=621 c ==============================================================================

62201 like 20101 but u=622 62202 like 20102 but u=622 62203 like 20103 but U=622 62204 like 20104 but u=622 62205 like 20105 but u=622 62206 like 20106 but mat=622 rho=-5.670 u=622 $ ELEMENT F-22 S/N:2917 62207 like 20107 but u=622 62208 like 20108 but u=622 62209 like 20109 but U=622 62210 like 20110 but u=622 62211 like 20111 but u=622 62212 like 20112 but u=622 c ==============================================================================

62301 like 20101 but u=623 62302 like 20102 but u=623 62303 like 20103 but u=623 62304 like 20104 but u=623 62305 like 20105 but u=623 62306 like 20106 but mat=623 rho=-5.599 u=623 $ ELEMENT F-23 S/N:5000 62307 like 20107 but u=623 25

ATTACHMENT I 62308 like 20108 but U=623 62309 like 20109 but u=623 62310 like 20110 but u=623 62311 like 20111 but u=623 62312 like 20112 but u=623 c ==============================================================================

62401 like 20101 but u=624 62402 like 20102 but u=624 62403 like 20103 but u=624 62404 like 20104 but u=624 62405 like 20105 but u=624 62406 like 20106 but rnat=624 rho=-5.596 U=624 $ ELEMENT F-24 S/N:5948 62407 like 20107 but u=624 62408 like 20108 but u=624 62409 like 20109 but u=624 62410 like 20110 but U=624 62411 like 20111 but u=624 62412 like 20112 but U=624 c ==============================================================================

62501 like 20101 but u=625 62502 like 20102 but u=625 62503 like 20103 but u=625 62504 like 20104 but U=625 62505 like 20105 but u=625 62506 like 20106 but rnat=625 rho=-5.603 u=625 $ ELEMENT F-25 S/N:5022 62507 like 20107 but u=625 62508 like 20108 but u=625 62509 like 20109 but u=625 62510 like 20110 but u=625 62511 like 20111 but U=625 62512 like 20112 but u=625 c ==============================================================================

62601 like 20101 but u=626 62602 like 20102 but u=626 62603 like 20103 but u=626 62604 like 20104 but u=626 62605 like 20105 but u=626 62606 like 20106 but rnat=626 rho=-5.701 u=626 $ ELEMENT F-26 S/N:2907 62607 like 20107 but u=626 62608 like 20108 but u=626 62609 like 20109 but u=626 62610 like 20110 but u=626 62611 like 20111 but u=626 62612 like 20112 but u=626 c ==============================================================================

62701 like 20101 but u=627 62702 like 20102 but u=627 62703 like 20103 but u=627 62704 like 20104 but U=627 62705 like 20105 but U=627 62706 like 20106 but rnat=627 rho=-5.609 u=627 $ ELEMENT F-27 S/N:5944 62707 like 20107 but u=627 62708 like 20108 but U=627 62709 like 20109 but U=627 62710 like 20110 but u=627 62711 like 20111 but u=627 62712 like 20112 but u=627 c ==============================================================================

62801 like 20101 but U=628 62802 like 20102 but U=628 62803 like 20103 but u=628 62804 like 20104 but u=628 62805 like 20105 but U=628 62806 like 20106 but rnat=628 rho=-5.638 u=628 $ ELEMENT F-28 S/N:3326 62807 like 20107 but u=628 62808 like 20108 but u=628 62809 like 20109 but u=628 62810 like 20110 but u=628 62811 like 20111 but u=628 62812 like 20112 but u=628 c ==============================================================================

26

ATTACHMENT 1 62901 like 20101 but u=629 62902 like 20102 but u=629 62903 like 20103 but u=629 62904 like 20104 but u=629 62905 like 20105 but U=629 62906 like 20106 but mat=629 rho=-5.682 u=629 $ ELEMENT F-29 S/N:2914 62907 like 20107 but u=629 62908 like 20108 but u=629 62909 like 20109 but u=629 62910 like 20110 but u=629 62911 like 20111 but u=629 62912 like 20112 but u=629 c ==============================================================================

63001 like 20101 but u=630 63002 like 20102 but U=630 63003 like 20103 but u=630 63004 like 20104 but u=630 63005 like 20105 but U=630 63006 like 20106 but mat=630 rho=-5.681 u=630 $ ELEMENT F-30 S/N:2909 63007 like 20107 but u=630 63008 like 20108 but u=630 63009 like 20109 but u=630 63010 like 20110 but u=630 63011 like 20111 but u=630 63012 like 20112 but u=630 c ==============================================================================

c ** RABBIT: UNIVERSE = [2]

145 7 -2. 699 (25 -12): (12 219) : (-25 -45) u=2 $ Not tube 146 3 -0.001239 12 -219 U=2 $ air 147 6 -1 #145 #146 u=2 $water c ** GRAPHITE ROD: UNIVERSE = [3]

160 6 21 201 U=3 $ Water around lower pin 161 9 -7.9 201 U=3 $ Bottom pin 162 9 -7.9 -14 21 U=3 $ BOTTOM SS cap 163 9 -7.9 14 -4 202 U=3 $ Cladding 164 2 -1.6 14 202 U=3 $ GRAPHITE 165 9 -7. 9 4 -3 U=3 $ TOP SS cap 166 9 -7.9 3 -201 U=3 $ Upper pin 167 6 -1 3 201 U=3 $ Water around pin c ** CENTRAL THIMBLE: UNIVERSE [4]

170 6 -1 12 -202 u=4 $ Water over midplane 171 7 -2.699 -12 :202 u=4 $ Below centerline c ** SOURCE: UNIVERSE = [5]

175 7 -2.699 -233 :230 :232 U=5 $ SOURCE HOLDER 176 9 -0.394473 233 -234 -231 U=5 $ SOURCE, SS SHELL 177 3 -0.001239 -232 233 -230 #176 U=5 $ SOURCE CAVITY c ** THIMBLE: UNIVERSE = [6]

180 6 21 201 U=6 $ Water around lower pin 181 7 -2.699 201 u=6 $ Bottom pin 182 7 -2.699 21 -14 U=6 $ BOTTOM cap 183 7 -2.699 14 210 U=6 $ Thimble TUBE 184 6 -1 14 -210 U=6 $ Thimble Water c ** CORE & FUEL: UNIVERSE = [7]

c ** WATER AROUND UPPER PINS & GRID PLATE ***********************************

185 6 1 5 1201 1202 1203 1204 1205 1206 1301 1302 1303 $ H20 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623 1624 1625 1626 1627 1628 1629 1630 203 U=7 c ** UPPER GRID PLATE *******************************************************

186 7 -2.699 -5 8 1201 1202 1203 1204 1205 1206 1301 $ Ul7ER GRID PLAT 1302 1303 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623 1624 1625 1626 1627 1628 1629 1630 203 U=7 27

ATTACHMENT 1 c ** CORE WATER *************************************************************

187 6 8 1201 1202 1203 1204 1205 1206 1301 1302 1303 $ CORE H20 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623'1624 1625 1626 1627 1628 1629 1630 203 U=7 c ** FUEL POSITIONS *********************************************************

c ** central thimble 1100 0 -203 fill=4 U=7 $ CT c ** B Ring 201 0 -1201 fill=201 (0 4 .05384 0 ) U=7 $ B-1 202 0 -1202 fill=202 (3.510728 2.02692 0 ) U=7 $ B-2 203 0 -1203 fill=203 (3 .510728 -2 .02692 0 ) U=7 $ B-3 204 0 -1204 fill=204 (0 -4. 05384 0 ) U=7 $ B-4 205 0 -1205 fill=205 (-3.510728 -2.02692 0 ) U=7 $ B-5 206 0 -1206 fil1=206 (-3. 510728 2. 02692' 0 ) U=7 $ B-6 c ** C Ring 301 0 -1301 fill=301 (0 7.98068 0 ) U=7 $ C-1 302 0 -1302 fill=302 (3. 99034 6. 911472 0 )* U=7 $ C-2 303 0 -1303 fill=303 (6 .911472 3 .99034 0 ) U=7 $ C-3 304 0 -1304 fil1=304 (7. 98068 0 0 ) U=7 $ C-4 305 0 -1305 fill=305 (6.911472 -3.99034 0) U=7 $ C-5 306 0 -1306 fill=306 (3. 99034 -6. 911472 0 ) U=7 $ C-6 c ******************** PULSE ROD **********************************************

307 0 -1307 fill=21 (0 -7 .98068 39.37 ) U=7 $ C-7 c *** 0.03937 cm/unit ***OUT 60% 23.622 ***OUT 100% 39.37 *** CRIT 15.622 c *****************************************************************************

c ** THIMBLE=6 WATER=8 FUEL=l 308 0 -1308 fill=308 (-3.99034 -6.911472 0 U=7 $ C-8 309 0 -1309 fill=309 (-6.911472 -3.99034 0 U=7 $ C-9 310 0 -1310 f ill=310 (-7.98068 0 0 ) U=7 $ C-10 311 0 -1311 fill=311 (-6.911472 3.99034 0 U=7 $ C-11 312 0 -1312 fill=312 (-3.99034 6.911472 0 U=7 $ C-12 c ** D Ring 401 0 -1401 fill=401 (0 11. 94562 0 ) U=7 $ D-1 402 0 -1402 fill=402 (4.085643 11.22521 0 U=7 $ D-2 403 0 -1403 fill=403 (7. 678497 9 .150876 0 U=7 $ D-3 c ****************************SAFETY ROD***************************************

404 0 -1404 fill=22 (10.34521 5.97281 33.02 ) U=7 $ D-4 C *** 0.03302 cm/unit ***OUT 60% 19.812 ***OUT 100% 33.02 *** CRIT 11.02 c *****************************************************************************

405 0 -1405 fill=405 (11.76414 2.074335 0 ) U=7 $ D-5 406 0 -1406 fill=406 (11. 76414 -2. 074335 0 ) U=7 $ D-6 407 0 -1407 fill=407 (10. 34521 -5. 97281 0 ) U=7 $ D-7 408 0 -1408 fill=408 (7 .678497 -9 .150876 0 ) U=7 $ D-8 409 0 -1409 fill=409 (4.085643 -11.22521 0) U=7 $ D-9 410 0 -1410 fill=410 (0 -11. 94562 0 ) U=7 $ D-10 411 0 -1411 fill=411 (-4.085643 -11.22521 0 ) U=7 $ D 11 412 0 -1412 fill=412 (-7 .678497 -9 .150876 0 ) U=7 $ D 12 413 0 -1413 fill=413 (-10.34521 -5.97281 0 ) U=7 $ D 13 414 0 -1414 fill=414 (-11.76414 -2.074335 0 ) U=7 $ D-14 415 0 -1415 fill=415 (-11.76414 2.074335 0 ) U=7 $ D 15 c ****************************SHIM ROD******************************************

416 0 -1416 fill=23 (-10.34521 5.97281 43.18 ) U=7 $ D-16 C *** 0.04318 cm/unit *** 60% OUT 25.908 *** 100% OUT 43.18 *** CRIT 15.908 c ******************************************************************************

417 0 -1417 fill=417 (-7.678497 9.150876 0 U=7 $ D-17 418 0 -1418 fill=418 (-4.085643 11.22511 0 ) U=7 $ D-18 c ** E Ring c ****************************REGULATING ROD************************************

501 0 -1501 fill=24 (0 15.91564 41.91 ) U=7 $ E-1 C *** 0.04191 cm/unit *** 60% OUT 25.146 *** 100% OUT 41.91 *** CRIT 0 c ******************************************************************************

502 0 -1502 fill=502 (4 .119271 15. 37333 0 ) U=7 $ E-2 503 0 -1503 fill=503 (7. 95782 13. 78335 0 ) U=7 $ E-3 504 0 -1504 fill=504 (11.25406 11.25406 0 ) U=7 $ E-4 505 0 -1505 fill=505 (13.78335 7.95782 0 ) U=7 $ E-5 506 0 -1506 fill=506 (15.37333 4.119271 0 ) U=7 $ E-6 28

ATIACHMENT 1 507 0 -1507 fill=507 (15.91564 0 0 ) U=7 $ E-7 508 0 -1508 fill=508 (15.37333 -4.119271 0 ) U=7 $ E-8 509 0 -1509 fill=509 (13. 78335 -7. 95782 0 ) U=7 $ E-9 510 0 -1510 fill=510 (11.25406 -11.25406 0 ) U=7, $ E-10 511 0 -1511 fill=5ll (7.95782 -13.78335 0 ) U=7 $ E-11 512 0 -1512 fill=512 (4 .119271 -15. 37333 0 ) U=7 $ E-12 513 0 -1513 fill=513 (0 -15.91564 0 ) U=7 $ E-13 514 0 -1514 fill=514 (-4.119271 -15.37333 0 ) U=7 $ E-14 515 0 -1515 fill=515 (-7.95782 -13.78335 0 ) U=7 $ E-15 516 0 -1516 fill=516 (-11.25406 -11.25406 0 ) U=7 $ E-16 517 0 -1517 fill=517 (-13.78335 -7.95782 0 ) U=7 $ E-17 518 0 -1518 fill=518 (-15.37333 -4.119271 0 ) U=7 $ E-18 519 0 -1519 fill=519 (-15.91564 0 0 ) U=7 $ E-19 520 0 -1520 fill=520 (-15.37333 4.119271 0 ) U=7 $ E-20 521 0 -1521 fill=521 (-13.78335 7.95782 0 ) U=7 $ E-21 522 0 -1522 fill=522 (-11.25406 11.25406 0 ) U=7 $ E-22 523 0 -1523 fill=523 (-7. 9*5792 13. 78335 ) U=7 $ E-23 524 0 -1524 fill=524 (-4.119271 15.37333 0 U=7 $ E-24 c ** F Ring 601 0 -1601 fill=601 (0 19.8882 0 ) U=7 $ F-1 602 0 -1602 fill=602 (4.134989 19.45359 0 ) U=7 $ F-2 603 0 -1603 fill=603 (8. 08926 18 .16878 0 ) U=7 $ F-3 604 0 -1604 fill=604 (11. 68999 16. 08989 0 ) U=7 $ F-4 605 0 -1605 fill=605 (14.77981 13.30878 0 ) U=7 $ F-5 606 0 -1606 fill=606 (17. 22369 9. 9441 0 ) U=7 $ F-6 607 0 -1607 fill=607 (18.9148 6.145792 0 ) U=7 $ F-7 608 0 -1608 fill=608 (19. 77925 2. 078883 0 ) U=7 $ F-8 609 0 -1609 fill=609 (19. 77925 -2. 078883 0 ) U=7 $ F-9 c ** SOURCE IS UNIVERSE 5, WATER VOID IS UNIVERSE 8 610 0 -1610 fill=5 (18.9148 -6.145792 0 ) U=7 $ F-10 SRC 5 611 0 -1611 fill=611 (17.22369 -9.9441 0 ) U=7 $ F-11 612 0 -1612 fill=612 (14. 77981 -13 .3078 0 ) U=7 $ F-12 c ** TEST FOR FUEL WORTH (fill=l) COMPARED TO WATER VOID (fill=8) 613 0 -1613 fill=613 (11. 68999 -16. 08989 0 ) U=7 $ F-13 ****

614 0 -1614 fill=614 (8.08926 -18.16878 0 ) U=7 $ F-14 615 0 -1615 fill=615 (4.134989 -19.45359 0 ) U=7 $ F-15 616 0 -1616 fill=616 (0 -19. 8882 0 ) U=7 $ F-16 617 0 -1617 fill=617 (-4.134989 -19.45359 0 ) U=7 $ F-17 618 0 -1618 fill=618 (-8.08926 -18.16878 0 ) U=7 $ F-18 c ** TEST FOR FUEL WORTH (fill=l) COMPARED TO WATER VOID (fill=8) 619 0 -1619 fill=619 (-11.68999 -16.08989 0 ) U=7 $ F-19 ****

620 0 -1620 fill=620 (-14.77981 -13.3078 0 ) U=7 $ F-20 621 0 -1621 fill=621 (-17.22369 -9.9441 0 ) U=7 $ F-21 622 0 -1622 fill=622 (-18.9148 -6.145792 0) U=7 $ F-22 623 0 -1623 fill=623 (-19.77925 -2.078883 0) U=7 $ F-23 C ** RABBIT IS UNIVERSE 2, WATER VOID IS UNIVERSE 8, RABBIT REMOVED FOR CORE III-4 624 0 -1624 fill=624 (-19. 77925 2. 078883 0 ) U=7 $ F-24 Old rabbit location 625 0 -1625 fill=625 (-18.9148 6.145792 0 ) U=7 $ F-25 626 0 -1626 fill=626 (-17.22369 9.9441 0 ) U=7 $ F-26 627 0 -1627 fill=627 (-14.77981 13.3078 0 ) U=7 $ F-27 628 0 -1628 fill=628 (-11. 68999 16. 08989 0 ) U=7 $ F-28 629 0 -1629 fill=629 (-8.08926 18.16878 0 ) U=7 $ F-29 630 0 -1630 fill=630 (-4 .134989 19 .45359 0 ) U=7 $ F-30 c SURFACES c *** ELEVATIONS **************************************************************

1 pz 50 $ POOL SURFACE 2 pz 37.2618 $TOP OF FUEL PINS 3 pz 33.782 $Top of end cap 4 pz 30.3276 $Top of gap/CLADDING 5 pz 33.79 $Top of grid plate 6 pz 27.7876 $Top of axial reflector 7 pz 27.94 $REFLECTOR CANNISTER 8 pz 31.162 $ **** Bottom top grid 9 pz 26.035 $RADIAL REFLECTOR TOP 10 pz 19.05 $ Top of fuel 11 pz 0.635 $ RSR FLOOR 12 pz 0 $ RSR SUBFLOOR -- LINE OF FUEL SYMMETRTY 13 pz -19.05 $ Bottom of fuel (38.1/2 cm) c ** BEST ESTIMATE 3.74 in= 9.4996 cm {-19.05 cm -9.4996 cm}=>

29

ATTACHMENT 1 14 pz -27.7876 $Bottom of axial reflector/CLADDING c ** APPROX CONE 3.797 cm base 3.797 cm heigth 21 pz -32 $ Bottom of plug/top of pin 15 pz -34.163 $Top of bottom grid plate 16 pz -26.035 $RADIAL REFLECTOR BOTTOM 17 pz -27.305 $ REFLECTOR CANNISTER BOTTOM 18 pz -36.07 $ Bottom of grid plate 20 pz -50 $ BOTTOM OF POOL c *** OTHER PLANES *************************************************************

25 pz -5.0 $ Bottom of thick part of rabbit 26 px 0 $ CORE 1/2 c *** z CYLINDERS **************************************************************

30 CZ 22.85 $ CORE/REFLECTOR WALL 31 CZ 23.485 $ INNER REFLECTOR 32 CZ 33.02 $ RSR INNER WALL 33 CZ 33.665 $ RSR INNER WALL 34 CZ 37.465 $ RSR OUTER VOLUME 35 CZ 38.1 $ RSR OUTER WALL 36 CZ 53.35 $ OUTER REFLECTOR 37 CZ 53.975 $ OUTER REFLECTOR WALL 38 CZ 60 $ EDGE OF POOL c *** OTHER CYLINDERS **********************************************************

40 1 c/x 0 -8.3 8.255 $ NE/SW BEAM PORT 90 1 c/x 0 -8.3 7.62 $ NE/SW BEAM PORT 43 CZ 22 $ LOWER GRID PLATE RADIUS 44 CZ 24.75 $ UPPER GRID PLATE RADIUS 45 CZ 1. 0 $ Bottom part of rabbit, above pin c *** FLUX PROBE HOLES *********************************************************

50 pz -12.4373 $ Pulse rod, span 39.37 cm 51 pz -12.6811 $Safety rod, span 33.02 cm 52 pz -26.0390 $ Shim rod, span 43.18 53 pz -21.3628 $ Regulating rod, span 41.91 68 pz 24.257 c ** UNIVERSE FUEL PIN cylinders ***********************************************

200 CZ 0.2285 $ Zirc FILLer 201 CZ 0.635 $ Pin diameter 202 CZ 1.8477 $ Fuel diameter MARK III FUEL SCHEMATIC 203 CZ 1. 985 $ CT & RABBIT c ** UNIVERSE CONTROL ROD cylinders ********************************************

210 cz 1.69165 $ Guide tube outer radius 1.89865 cm 219 cz 1.58815 $ REG guide tube surface 211 cz 1.511 $ Pulse meat 212 cz 1.587 $ Pulse rod cladding surface 213 cz 1.5113 $ Shim/safety meat 214 cz 1.5875 $ Shim/safety rod cladding surface 215 cz 1.03505 $ Rod reg meat 216 cz 1.11125 $ Reg rod cladding surface 217 cz 1.254 $ Extension tube outer radius 218 cz 0.619 $ Extesion tube inner radius c ** UNIVERSE SOURCE ***********************************************************

230 CZ 0.981 $SOURCE HOLDER CAVITY 231 cz 0.9525 $ SOURCE RADIUS c source elevations ALSO USES SOME FUEL ELEVATIONS 232 pz 3.8 $Source cavity top 233 pz -3.8 $Source cavity bottom 234 pz 2.55 $Source heigth c ******* END OF UNIVERSE SURFACE DEFINTIONS ***********************************

c ** A RING USES SURFACE 203 ***************************************************

c 1000 c/z O O 1.8985 $ A-1 = surface 203 c ** B Ring ********************************************************************

1201 c/z 0 4.05384 1. 8985$ B-1 1202 c/z 3. 510728 2.02692 1. 8985 $ B-2 1203 c/z 3. 510728 -2.02692 1.8985 $ B-3 1204 c/z 0 -4.05384 1.8985 $ B-4 1205 c/z -3. 510728 -2.02692 1.8985 $ B-5 1206 c/z -3. 510728 2.02692 1.8985 $ B-6 c ** C Ring ********************************************************************

1301 c/z o 7.98068 1.8985$ c-1 1302 c/z 3.99034 6.911472 1.8985 $ C-2 1303 c/z 6.911472 3.99034 1.8985 $ C-3 1304 c/z 7.98068 o 1.8985 $ C-4 30

ATTACHMENT 1 1305 c/z 6. 911472 -3.99034 1.8985 $ C-5 1306 c/z 3.99034 -6.911472 1. 8985 $ C-6 1307 c/z 0 -7.98068 1.8985 $ C-7 1308 c/z -3.99034 -6.911472 1.8985 $ C-8 1309 c/z -6.911472 -3.99034 1.8985 $ C-9 1310 c/z -7.98068 0 1. 8985 $ C-10 1311 c/z -6.911472 3.99034 1. 8985 $ C-11 1312 c/z -3.99034 6.911472 1. 8985 $ C-12 c ** D Ring ********************************************************************

1401 c/z 0 11.94562 1. 8985 $ D-1 1402 c/z 4.085643 11.225211 1.8985 $ D-2 1403 c/z 7.678497 9.150876 1.8985 $ D-3 1404 c/z 10.34521 5.97281 1.8985 $ D-4 1405 c/z 11. 764139 2.074335 1.8985 $ D-5 1406 c/z 11.764139 -2.074335 1.8985 $ D-6 1407 c/z 10.34521 -5.97281 1.8985 $ D-7 1408 c/z 7.678497 -9.150876 1.8985 $ D-8 1409 c/z 4 .119271 -11.225211 1.8985 $ D-9 1410 c/z 0 -11.94562 1. 8985 $ D-10 1411 c/z -4 .119271 -11. 225211 1.8985 $ D 11 1412 c/z -7.678497 -9.150876 1.8985 $ D 12 1413 c/z -10.34521 -5.97281 1.8985 $ D 13 1414 c/z -11. 764139 -2.074335 1.8985 $ D-14 1415 c/z -11.764139 2.074335 1. 8985 $ D 15 1416 c/z -10.34521 5.97281 1. 8985 $ D-16 1417 c/z -7.678497 9.150876 1.8985 $ D-17 1418 c/z -4 .119271 11.225211 1.8985 $ D-18 c ** E RING ********************************************************************

1501 c/z 0 15.91564 1. 8 98 5 $ E-1 1502 c/z 4 .119271 15.373328 1.8985 $ E-2 1503 c/z 7.95782 13.783349 1.8985 $ E-3 1504 c/z 11.254057 11. 254057 1.8985 $ E-4 1505 c/z 13.783349 7.95782 1.8985 $ E-5 1506 c/z 15.373328 4 .119271 1.8985 $ E-6 1507 c/z 15.91564 0 1.8985 $ E-7 1508 c/z 15.373328 -4.119271 1.8985 $ E-8 1509 c/z 13.783349 -7.95782 1.8985 $ E-9 1510 c/z 11.254057 -11.254057 1.8985 $ E-10 1511 c/z 7.95782 -13.783349 1.8985 $ E-11 1512 c/z 4.119271 -15.373328 1.8985 $ E-12 1513 c/z 0 -15.91564 1.8985 $ E-13 1514 c/z -4 .119271 -15.373328 1.8985 $ E-14 1515 c/z -7.95782 -13. 783349 1.8985 $ E-15 1516 c/z -11. 254057 -11. 254057 1.8985 $ E-16 1517 c/z -13.783349 -7.95782 1. 8985 $ E-17 1518 c/z -15. 373328 -4 .119271 1. 8985 $ E-18 1519 c/z -15.91564 0 1.8985 $ E-19 1520 c/z -15.373328 4 .119271 1.8985 $ E-20 1521 c/z -13.783349 7.95782 1. 8 98 5 $ E-21 1522 c/z -11. 254057 11.254057 1.8985 $ E-22 1523 c/z -7.95782 13.783349 1. 8985 $ E-23 1524 c/z -4 .119271 15.373328 1.8985 $ E-24 c ** F Ring ********************************************************************

1601 c/z 0 19.8882 1.8985 $ F-1 1602 c/z 4.134989 19.453595 1.8985 $ F-2 1603 c/z 8.08926 18.168775 1.8985 $ F-3 1604 c/z 11.689991 16.089892 1.8985 $ F-4 1605 c/z 14.779813 13.3087803 1.8985 $ F-5 1606 c/z 17.223686 9.9441 1. 8985 $ F-6 1607 c/z 18.914802 6.145792 1.8985 $ F-7 1608 c/z 19.77925 2.078883 1.8985 $ F-8 1609 c/z 19.77925 -2.078883 1.8985 $ F-9 1610 c/z 18.914802 -6.145792 1.8985 $ F-10 SOURCE 1611 c/z 17.223686 -9.9441 1.8985 $ F-11 1612 c/z 14.779813 -13.307803 1.8985 $ F-12 1613 c/z 11.689991 -16.089892 1.8985 $ F-13 1614 c/z 8.08926 -18.168775 1.8985 $ F-14 1615 c/z 4.134989 -19.453595 1.8985 $ F-15 1616 c/z 0 -19.8882 1.8985 $ F-16 1617 c/z -4.134989 -19.453595 1.8985 $ F-17 1618 c/z -8.08926 -18.168775 1.8985 $ F-18 31

ATTACHMENT 1 1619 c/z -11. 689991 -16.089892 1.8985 $ F-19 1620 c/z -14.779813 -13.307803 1.8985 $ F-20 1621 c/z -17.223686 -9.9441 1.8985 $ F-21 1622 c/z -18.914802 -6.145792 1.8985 $ F-22 1623 c/z -19.77925 -2.078883 1. 8985 $ F-23 RABBIT 1624 c/z -19.77925 2.078883 1. 8985 $ F-24 1625 c/z -18.914802 6.145792 1.8985 $ F-25 1626 c/z -17.223686 9.9441 1.8985 $ F-26 1627 c/z -14. 779813 13.307803 1.8985 $ F-27 1628 c/z -11. 689991 16.089892 1.8985 $ F-28 1629 c/z -8.08926 18.168775 1.8985 $ F-29 1630 c/z -4.134989 19.453595 1.8985 $ F-30 trl o o o o.6428 0.776 o -0.776 0.6428 o o o 1 mode n kcode 100000 1.000000 10 110 c MATERIAL CARDS c u + Zr-H rho=5.7473 g/cmA3 BOL c S(a,b)- Zr-H@ 300 K c ******************************************************************************

c This current model only accounts for the burnup of U235 c ** B-1 ** U235 left after Burnup = 37.012 g **Density= 5.685 m201 92235.66c -0.016144 92238.66c -0.066645 40000.66c -0.900788 1001. 66c -0.016422 c ** B-2 ** U235 left after Burnup 36.731 g **Density 5.681 m202 92235.66c -0.016022 92238.66c -0.066140 40000.66c -0.901405 1001.66c -0.016433 c ** B-3 ** U235 left after Burnup 37.012 g **Density 5.685 m203 92235.66c -0.016144 92238.66c -0.066645 40000.66c -0.900788 1001.66c -0.016422 c ** B-4 ** U235 left after Burnup 37.315 g **Density 5.688 m204 92235.66c -0.016276 92238.66c -0 .067191 40000.66c -0.900123 1001.66c -0.016410 c ** B-5 ** U235 left after Burnup 36.985 g ** Density 5.684 m205 92235.66c -0.016132 92238.66c -0.066598 40000.66c -0.900847 1001. 66c -0.016423 c ** B-6 ** U235 left after Burnup 37.034 g **Density 5.684 m206 92235.66c -0.016154 92238. 66c -0.066686 40000.66c -0.900739 1001. 66c -0.016421 c ** C-1 ** U235 left after Burnup 36.691 g **Density 5.680 m301 92235.66c -0.016004 92238.66c -0.066068 40000.66c -0.901493 1001.66c -0.016435 c ** C-2 ** U235 left after Burnup 37.012 g **Density 5.684 m302 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-3 ** U235 left after Burnup 37.515 g **Density 5.691 m303 92235.66c -0.016363 92238.66c -0.067552 40000.66c -0.899683 1001.66c -0.016402 c ** C-4 ** U235 left after Burnup 33.388 g **Density 5.638 m304 92235.66c -0.014563 92238.66c -0.060120 40000.66c -0.908749 1001. 66c -0.016567 32

ATIACHMENT 1 c ** C-5 ** U235 left after Burnup 39.023 g ** Density 5. 710 m305 92235.66c -0.017021 92238.66c -0.070267 40000.66c -0.896370 1001. 66c -0.016342 c ** C-6 ** U235 left after Burnup 34.287 g **Density 5.649 m306 92235.66c -0.014955 92238.66c -0.061739 40000.66c -0.906774 1001.66c -0.016531 c ** C-8 ** U235 left after Burnup 37.690 g ** Density 5.693 m308 92235.66c -0.016440 92238.66c -0.067867 40000.66c -0.899298 1001.66c -0.016395 c ** C-9 ** U235 left after Burnup 37.012 g ** Density 5.684 m309 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001. 66c -0.016422 c ** C-10 ** U235 left after Burn up 37.012 g **Density 5.684 m310 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001. 66c -0.016422 c ** C-11 ** U235 left after Burn up 37.012 g ** Density 5.684 m311 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-12 ** U235 left after Burnup 37.012 g ** Density 5.684 m312 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** D-1 ** U235 left after Burnup 33.326 g **Density 5.637 m401 92235.66c -0.014536 92238.66c -0.060009 40000.66c -0.908885 1001.66c -0.016570 c ** D-2 ** U235 left after Burnup 34.700 g ** Density 5.655 m402 92235.66c -0.015136 92238.66c -0.062483 40000.66c -0.905867 1001.66c -0.016515 c ** D-3 ** U235 left after Burnup 32.480 g **Density 5.626 m403 92235.66c -0.014167 92238.66c -0.058485 40000.66c -0.910744 1001.66c -0.016604 c ** D-5 ** U235 left after Burnup 33.266 g ** Density 5.636 m405 92235.66c -0.014510 92238.66c -0.059901 40000.66c -0.909017 1001.66c -0.016572 c ** D-6 ** U235 left after Burnup 38.012 g ** Density 5.697 m406 92235.66c -0.016580 92238.66c -0.068447 40000.66c -0.898591 1001. 66c -0.016382 c ** D-7 ** U235 left after Burnup 34.046 g ** Density 5.646 m407 92235.66c -0.014850 92238.66c -0.061305 40000.66c -0.907303 1001.66c -0.016541 c ** D-8 ** U235 left after Burnup 33.278 g ** Density 5.636 m408 92235.66c -0.014515 92238.66c -0.059922 40000.66c -0.908991 1001.66c -0.016572 c ** D-9 ** U235 left after Burnup 33.703 g ** Density 5.642 33

ATTACHMENT 1 m409 92235.66c -0.014701 92238.66c -0.060688 40000.66c -0.908057 1001.66c -0.016555 c ** D-10 ** U235 left after Burnup 32.929 g ** Density 5.632 m410 92235.66c -0.014363 92238.66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** D-11 ** U235 left after Burnup 33.658 g ** Density 5. 641 m411 92235.66c -0.014363 92238.66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** D-12 ** U235 left after Burn up 34. 711 g ** Den_sity 5.655 m412 92235.66c -0.015140 92238.66c -0.062503 40000.66c -0.905843 1001.66c -0.016514 c ** D-13 ** U235 left after Burn up 33.835 g ** Density 5.644 m413 92235.66c -0.014758 92238.66c -0.060925 40000.66c -0 .907767 1001. 66c -0.016549 c ** D-14 ** U235 left after Burnup 35.835 g ** Density 5.669 m414 92235.66c -0.015631 92238.66c -0.064527 40000.66c -0.903373 1001.66c -0.016469 c ** D-15 ** U235 left after Burnup 34.841 g ** Density 5.656 m415 92235.66c -.0. 015197 92238.66c -0.062737 40000.66c -0.905557 1001.66c -0.016509 c ** D-17 ** U235 left after Burn up 34.835 g ** Density 5.656 m417 92235.66c -0.015194 92238.66c -0.062726 40000.66c -0.905570 1001. 66c -0.016509 c ** D-18 ** U235 left after Burnup 33.822 g ** Density 5.643 m418 92235.66c -0.014753 92238.66c -0.060902 40000.66c -0.907796 1001. 66c -0.016550 c ** E-2 ** U235 left after Burnup 37.211 g **Density 5.687 m502 92235.66c -0.016231 92238.66c -0.067004 40000.66c -0.900351 1001.66c -0.016414 c ** E-3 ** U235 left after Burnup 32.929 g ** Density 5.632 m503 92235.66c -0.014363 92238.66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** E-4 ** U235 left after Burnup 33.692 g **Density 5.642 m504 92235.66c -0.014696 92238.66c -0.060668 40000.66c -0.908081 1001. 66c -0.016555 c ** E-5 ** U235 left after Burnup 32.388 g ** Density 5.625 m505 92235.66c -0.014127 92238.66c -0.058320 40000.66c -0.910946 1001.66c -0.016607 c ** E-6 ** U235 left after Burnup 32.503 g ** Density 5.627 m506 92235.66c -0.014177 92238.66c -0.058527 40000.66c -0.910693 1001. 66c -0.016603 c ** E-7 ** U235 left after Burnup 31.991 g ** Density 5.620 m507 92235.66c -0.013954 34

ATTACHMENT 1 92238.66c -0.057605 40000.66c -0 .911818 1001.66c -0.016623 c ** E-8 ** U235 left after Burnup 32.288 g **Density 5.624 m508 92235.66c -0.014084 92238.66c -0.058140 40000.66c -0 .911165 1001.66c -0.016611 c ** E-9 ** U235 left after Burnup 29.517 g **Density 5.588 m509 92235.66c -0.012875 92238.66c -0.053150 40000.66c -0. 917253 1001.66c -0.016722 c ** E-10 ** U235 left after Burn up 31. 776 g ** Density 5.617 m510 92235.66c -0.013860 92238.66c -0.057218 40000.66c -0.912290 1001.66c -0.016632 c ** E-11 ** U235 left after Burnup 32.835 g **Density 5.631 m511 92235.66c -0.014322 92238.66c -0.059125 40000.66c -0.909964 1001.66c -0.016590 c ** E-12 ** U235 left after Burnup 34.835 g ** Density 5.656 m512 92235.66c -0.015194 92238.66c -0. 062726 40000.66c -0.905570 1001.66c -0.016509 c ** E-13 ** U235 left after Burn up 37.713 g ** Density 5.693 m513 92235.66c -0.016450 92238.66c -0.067908 40000.66c -0.899248 1001.66c -0.016394 c ** E-14 ** U235 left after Burn up 33.254 g ** Density 5.636 m5.14 92235.66c -0.014505 92238.66c -0.059879 40000.66c -0.909043 1001. 66c -0.016573 c ** E-15 ** U235 left after Burn up 37.636 g **Density 5.692 m515 92235.66c -0.016416 92238.66c -0.067770 40000.66c -0.899417 1001.66c -0.016397 c ** E-16 ** U235 left after Burn up 31.948 g ** Density 5.619 m516 92235.66c -0.013935 92238.66c -0.057527 40000.66c -0.911912 1001.66c -0.016625 c ** E-17 ** U235 left after Burn up 35.937 g ** Density 5.670 m517 92235.66c -0.015675 92238.66c -0.064710 40000.66c -0.903149 1001.66c -0.016465 c ** E-18 ** U235 left after Burnup 34.960 g ** Density 5.658 m518 92235.66c -0.015249 92238.66c -0.062951 40000.66c -0.905296 1001.66c -0.016504 c ** E-19 ** U235 left after Burn up 34.855 g **Density 5.657 m519 92235.66c -0.015203 92238.66c -0.062762 40000.66c -0.905526 1001. 66c -0.016509 c ** E-20 ** U235 left after Burnup 34.234 g **Density 5.649 m520 92235.66c -0.014932 92238.66c -0.061644 40000.66c -0.906890 1001. 66c -0.016533 c ** E-21 ** U235 left after Burnup 33.977 g ** Density 5.645 m521 92235.66c -0.014820 92238.66c -0.061181 35

ATTACHMENT 1 40000.66c -0.907455 1001.66c -0.016544 c ** E-22 ** U235 left after Burn up 34.021 g **Density 5.646 m522 92235.66c -0.014839 92238.66c -0.061260 40000.66c -0.907358 1001.66c -0.016542 c ** E-23 ** U235 left after Burnup 33.599 g **Density 5.641 m523 92235.66c -0.014655 92238.66c -0.060500 40000.66c -0.908285 1001.66c -0.016559 c ** E-24 ** U235 left after Burn up 33.226 g ** Density 5.636 m524 92235.66c -0.014493 92238.66c -0.059829 40000.66c -0.909105 1001. 66c -0.016574 c ** F-1 ** U235 left after Burnup 31.327 g ** Density 5.612 m601 92235.66c -0.013664 92238.66c -0.056409 40000.66c -0.913277 1001. 66c -0.016650 c ** F-2 ** U235 left after Burnup 31.467 g ** Density 5.613 m602 92235.66c -0.013725 92238.66c -0.056661 40000.66c -0 .912969 1001.66c -0.016644 c ** F-3 ** U235 left after Burnup 32.486 g **Density 5.626 m603 92235.66c -0.014170 92238.66c -0.058496 40000.66c -0.910730 1001. 66c -0.016603 c ** F-4 ** U235 left after Burnup 31.438 g ** Density 5.613 m604 92235.66c -0. 013713 92238.66c -0.056609 40000.66c -0 .913033 1001. 66c -0.016645 c ** F-5 ** U235 left after Burnup 31.456 g ** Density 5.613 m605 92235.66c -0 .013721 92238.66c -0.056641 40000.66c -0.912993 1001.66c -0.016645 c ** F-6 ** U235 left after Burnup 38.012 g **Density 5.697 m606 92235.66c -0.016580 92238.66c -0.068447 40000.66c -0.898591 1001. 66c -0.016382 c ** F-7 ** U235 left after Burnup 30.816 g ** Density 5.605 m607 92235.66c -0.013441 92238.66c -0.055489 40000.66c -0.914399 1001. 66c -0.016670 c ** F-8 ** U235 left after Burnup 31.404 g ** Density 5.613 m608 92235.66c -0 .013698 92238.66c -0.056548 40000.66c -0.913107 1001.66c -0.016647 c ** F-9 ** U235 left after Burnup 31. 418 g ** Density 5. 613 m609 92235.66c -0.013704 92238.66c -0.056573 40000.66c -0.913077 1001.66c -0.016646 c ** F-11 ** U235 left after Burnup 30.355 g ** Density 5.599 m611 92235.66c -0.013240 92238.66c -0.054659 40000.66c -0.915412 1001.66c -0.016689 c ** F-12 ** U235 left after Burn up 38.461 g **Density 5.703 m612 92235.66c -0.016776 92238.66c -0.069255 40000.66c -0.897605 36

ATTACHMENT 1 1001.66c -0. 016364 c ** F-13 ** U235 left after Burn up 32.407 g ** Density 5.625 m613 92235.66c -0.014135 92238.66c -0.058354 40000.66c -0.910904 1001.66c -0.016607 c ** F-14 ** U235 left after Burn up 30.445 g ** Density 5.600 m614 92235.66c -0.013280 92238.66c -0.054821 40000.66c -0.915214 1001.66c -0.016685 c ** F-15 ** U235 left after Burnup 30.903 g ** Density 5.606 m615 92235.66c -0.013479 92238.66c -0.055646 40000.66c -0.914208 1001.66c -0.016667 c ** F-16 ** U235 left after Burnup 30.326 g ** Density 5.599 m616 92235.66c -0.013228 92238.66c -0.054607 40000.66c -0.915476 1001.66c -0.016690 c ** F-17 ** U235 left after Burn up 30.081 g **Density 5 .596 m617 92235.66c -0.013121 92238.66c -0.054166 40000.66c -0.916014 1001.66c -0.016700 c ** F-18 ** U235 left after Burn up 30.344 g ** Density 5.599 m618 92235.66c -0. 013236 92238.66c -0.054639 40000.66c -0.915436 1001. 66c -0.016689 c ** F-19 ** U235 left after Burnup 33.099 g **Density 5.634 m619 92235.66c -0.014437 92238.66c -0. 059600 40000.66c -0.909384 1001.66c -0.016579 c ** F-20 ** U235 left after Burn up 37.540 g ** Density 5.691 m620 92235.66c -0.016374 92238.66c -0.067597 40000.66c -0 .899628 1001.66c -0.016401 c ** F-21 ** U235 left after Burnup 31.381 g ** Density 5.612 m621 92235.66c -0.013688 92238.66c -0.056506 40000.66c -0 .913158 1001.66c -0.016648 c ** F-22 ** U235 left after Burnup 35.890 g ** Density 5.670 m622 92235.66c -0.015655 92238.66c -0.064626 40000.66c -0.903253 1001. 66c -0.016467 c ** F-23 ** U235 left after Burnup 30.368 g ** Density 5.599 m623 92235.66c -0.015655 92238.66c -0.064626 40000.66c -0.903253 1001. 66c -0.016467 c ** F-24 ** U235 left after Burnup 30.079 g ** Density 5.596 m624 92235.66c -0.013120 92238.66c -0.054162 40000.66c -0.916018 1001. 66c -0.016700 c ** F-25 ** U235 left after Burn up 30.667 g ** Density 5.603 m625 92235.66c -0.013376 92238.66c -0.055221 40000.66c -0.914726 1001.66c -0.016676 c ** F-26 ** U235 left after Burn up 38.339 g ** Density 5.701 m626 92235.66c -0.016723 92238.66c -0.069035 40000.66c -0.897873 1001.66c -0.016369 37

ATTACHMENT 1 c ** F-27 ** U235 left after Burnup 31.103 g ** Density 5.609 m627 92235.66c -0.013567 92238.66c -0.056006 40000.66c -0.913769 1001.66c -0.016659 c ** F-28 ** U235 left after Burnup 33.415 g **Density 5.638 m628 92235.66c -0.014575 92238.66c -0.060169 40000.66c -0.908690 1001.66c -0.016566 c ** F-29 ** U235 left after Burnup 36.881 g Density 5.682 m629 92235.66c -0.016087 92238.66c -0.066410 40000.66c -0.901076 1001. 66c -0.016427 c ** F-30 ** U235 left after Burn up 36.760 g **Density 5.681 m630 92235.66c -0.016034 92238.66c -0.066192 40000.66c -0.901341 1001.66c -0.016432 c ** 12 wt % New Fuel Elements ** U235 52 g ** Density 5.875 m700 92235.66c -0.022682 92238.66c -0.093634 40000.66c -0.867862 1001. 66c -0.015822 c ******************************************************************************

c graphite rho = 1.6 g/cmA3 c S(a,b)- Carbon@ 300K m2 6000.66c -1 $MAT c Air rho= 0.001239 $ RSR AIR m3 6000.66c -0.000124 7014.66c -0.755268 8016.66c -0.231781 18000 -0.012827 c Light water rho=l.00 g/cmA3 c S(a,b)- H20@ 300K m6 1001.66c 0.66667 $Water 8016.66c 0.33333 c Aluminum rho 2.699 g/cmA3 m7 13027.66c -1 $Aluminum, 2.699 g/cc c Zirconium$ Zirc Filler, 6.5 g/cc c S(a,b)- H20@ 300K ma 40000.66c -1 c Stainless Steel rho c m9 26000.55c -1.0 m9 26000.55c -0.67 $MAT 24000.50c -0.17 28000.50c -0.08849997 42000.66c -0.004452 c B4C SHIM & REG rod rho > 2.48 g/cmA3, email from GA, c NEED TO ADJUST FOR BURNUP TO DIFFERENTIATE BETWEEN NEW ROD??

c rho 1.60-1.75 g/cmA3 from B. Ma c 16% Boron from UIC TRIGA SAR roll 6000.66c -0.84 $ Pulse Rod, 1.6 g/cc, 5010.66c -0.0296 5011.66c -0.1304 c B4C rho> 2.48 g/cc, SHIM & REG, BUT THESE ARE OLD & PART BURNED ml2 6000.66c -0.22 $ Shim Rod, 2.5 g/cc 5010.66c -0.1443 5011.66c -0.6357 c B4C rho >2.48 g/cc, SAFETY ROD ml3 6000.66c -0.22 $ Shim Rod, 2.5 g/cc 5010.66c -0.1443 5011. 66c -0.6357 c ******************************************************************************

imp:n O 1 1165r $ 1, 63012 c ******************************************************************************

c SOURCE DISTRIBUTED ACROSS THE CORE VOLUME sdef ERG=Dl POS=O 0 -29 AXS=O 0 1 RAD=D2 EXT=D3 spl -3 si2 O 22.8 si3 O 45.7 c ******************************************************************************

mt201 h/zr.Olt zr/h.Olt $ uzrh-full salphabeta card 38

ATTACHMENT 1 mt202 h/zr.Olt zr/h.Olt mt203 h/zr. Olt zr/h.Olt mt204 h/zr.Olt zr/h.Olt mt205 h/zr.Olt zr/h. Olt mt206 h/zr.Olt zr/h.Olt mt301 h/zr.Olt zr/h.Olt mt302 h/zr.Olt zr/h.Olt mt303 h/zr.Olt zr/h.Olt mt304 h/zr.Olt zr/h.Olt mt305 h/zr.Olt zr/h.Olt mt306 h/zr.Olt zr/h.Olt mt308 h/zr.Olt zr/h.Olt mt309 h/zr.Olt zr/h. Olt mt310 h/zr.Olt zr/h.Olt mt311 h/zr.Olt zr/h. Olt mt312 h/zr.Olt zr/h.Olt mt401 h/zr.Olt zr/h.Olt mt402 h/zr.Olt zr/h.Olt mt403 h/zr.Olt zr/h.Olt mt405 h/zr.Olt zr/h.Olt mt406 h/zr.Olt zr/h.Olt mt407 h/zr.Olt zr/h.Olt mt408 h/zr.Olt zr/h.Olt mt409 h/zr.Olt zr/h.Olt mt410 h/zr.Olt zr/h.Olt mt411 h/zr. Olt zr/h. Olt mt412 h/zr.Olt zr/h.Olt mt413 h/zr.Olt zr/h.Olt mt414 h/zr.Olt zr/h.Olt mt415 h/zr.Olt zr/h.Olt mt417 h/zr.Olt zr/h.Olt mt418 h/zr.Olt zr/h.Olt mt502 h/zr. Olt zr/h.Olt mt503 h/zr. Olt zr/h. Olt mt504 h/zr.Olt zr/h. Olt mt505 h/zr.Oit zr/h.Olt mt506 h/zr.Olt zr/h.Olt mt507 h/zr.Olt zr/h.Olt mt508 h/zr.Olt zr/h.Olt mt509 h/zr.Olt zr/h.Olt mt510 h/zr.Olt zr/h.Olt mt511 h/zr.Olt zr/h.Olt mt512 h/zr. Olt zr/h. Olt mt513 h/zr.Olt zr/h.Olt mt514 h/zr.Olt zr/h.Olt mt515 h/zr.Olt zr/h.Olt mt516 h/zr.Olt zr/h.Olt mt517 h/zr.Olt zr/h.Olt mt518 h/zr.Olt zr/h.Olt mt519 h/zr.Olt zr/h.Olt mt520 h/zr. Olt zr/h.Olt mt521 h/zr.Olt zr/h.Olt mt522 h/zr.Olt zr/h.Olt mt523 h/zr.Olt zr/h.Olt mt524 h/zr.Olt zr/h.Olt mt601 h/zr.Olt zr/h. Olt mt602 h/zr.Olt zr/h.Olt mt603 h/zr.Olt zr/h.Olt mt604 h/zr.Olt zr/h.Olt mt605 h/zr.Olt zr/h.Olt mt606 h/zr.Olt zr/h.Olt mt607 h/zr.Olt zr/h.Olt mt608 h/zr.Olt zr/h.Olt mt609 h/zr.Olt zr/h.Olt mt611 h/zr.Olt zr/h.Olt mt612 h/zr.Olt zr/h.Olt mt613 h/zr.Olt zr/h. Olt mt614 h/zr.Olt zr/h.Olt mt615 h/zr.Olt zr/h.Olt mt616 h/zr. Olt zr/h.Olt mt617 h/zr.Olt zr/h.Olt 39

ATTACHMENT 1 mt618 h/zr.Olt zr/h.Olt mt619 h/zr.Olt zr/h.Olt mt620 h/zr.Olt zr/h.Olt mt621 h/zr.Olt zr/h.Olt mt622 h/zr.Olt zr/h.Olt mt623 h/zr.Olt zr/h.Olt mt624 h/zr. Olt zr/h.Olt mt625 h/zr.Olt zr/h.Olt mt626 h/zr.Olt zr/h.Olt mt627 h/zr.Olt zr/h.Olt mt628 h/zr.Olt zr/h.Olt mt629 h/zr. Olt zr/h. Olt mt630 h/zr. Olt zr/h.Olt mt700 h/zr.Olt zr/h.Olt mt2 grph.OlT mt6 lwtr. 01 c ************************* TALLY SPECIFICATION ********************************

c Flux average tally for active fuel region of all 85 elements c FUEL ELEMENT ORDER IMPORTANT f4:n 20106 20206 20306 20406 20506 20606 &

30106 30206 30306 30406 30506 30606 30806 30906 31006 31106 31206 &

40106 40206 40306 40506 40606 40706 40806 40906 41006 41106 41206 &

41306 41406 41506 41706 41806 &

50206 50306 50406 50506 50606 50706 50806 50906 51006 51106 51206 &

51306 51406 51506 51606 51706 51806 51906 52006 52106 52206 52306 &

52406 &

60106 60206 60306 60406 60506 60606 60706 60806 60906 61106 61206 &

61306 61406 61506 61606 61706 61806 61906 62006 62106 62206 62306 &

62406 62506 62606 62706 62806 62906 63006 f7:n 20106 20206 20306 20406 20506 20606 &

30106 30206 30306 30406 30506 30606 30806 30906 31006 31106 31206 &

40106 40206 40306 40506 40606 40706 40806 40906 41006 41106 41206 &

41306 41406 41506 41706 41806 &

50206 50306 50406 50506 50606 50706 50806 50906 51006 51106 51206 &

51306 51406 51506 51606 51706 51806 51906 52006 52106 52206 52306 &

52406 &

60106 60206 60306 60406 60506 60606 60706 60806 60906 61106 61206 &

61306 61406 61506 61606 61706 61806 61906 62006 62106 62206 62306 &

62406 62506 62606 62706 62806 62906 63006 40

ATTACHMENT 1 MCNP Input - Perturbed Case c KSU TRIGA Mark II c

c Core with 4 12% elements in the E ring c E7, Ell, El8, E23 c

c CELLS:

1 0 1 :-20 :38 $ OUTSIDE 2 6 -1 ( ( (-1 2 -38 ) : (-2 7 -38 44 ) : $ POOL ELEMENTS

(-17 18 -38 30 ) : (-18 20 -38 ) : (-7 17 -38 37 ) ) :

(-15 18 -38 43 ) ) #7 #8 3 6 34 33 -7 11 $ RSR SPACE/VOLUME 4 7 -2.699 -35 32 -9 12 #3 $ RSR WALLS 5 2 -1. 6 -36 31 -9 16 #3 #4 #7 #8 $ REFLECTOR GRAPHITE 6 7 -2.699 -37 30 -7 17 #3 #4 #5 #7 #8 $ CANNISTER/WALLS 7 7 -2.699 31 -40 90 -38 26 $ NEBP AL CASE 8 3 -0.001239 31 38 26 $ NEBP CAVITY 9 0 (-30 -7 15 ) : (-44 -2 7 ) fill=? $ CORE SPACE 10 7 -2.699 -15 18 -43 203 $ Lower grid plate 11 0 -15 18 -203 fill=4 $ CT penetration, lower grid plate c ** UNIVERSES c ******************************************************************************

c ** WATER IN CORE AREA: UNIVERSE = [8]

95 6 -1 20 U=8 $ Water above pool bottom 96 7 -2.699 -20 U=8 $ Stuff below pool bottom c ** CONTROL ROD UNIVERSES c ** NOTE: Holes in guide tubes modeled as reduced Al density c (101, 110, 120, 126) c ** PULSE ROD: UNIVERSE = [21]

101 7 -2.12 210 U=21 $ Guide Tube 102 7 -2.699 50 -68 213 -214 U=21 $ Rod cladding 103 11 -2.0 50 213 U=21 $ Pulse rod 104 9 -7.9 68 -217 218 U=21 $ SS extension 105 6 -1 -210 #102 #103 #104 U=21 $ Water c ** SAFETY ROD: UNIVERSE = [22]

110 7 -2.12 210 U=22 $ Guide Tube 111 13 -2.52 51 212 U=22 $ Safety rod 112 9 -7.9 68 -217 218 U=22 $ SS extension 113 6 -1 -210 #111 #112 U=22 $ Water c ** SHIM ROD: UNIVERSE = [23]

120 7 -2.12 210 U=23 $ Guide Tube 121 7 -2.699 52 -68 211 -212 U=23 $ Rod cladding 122 12 -2.48 52 211 U=23 $ Shim/safety rod 123 9 -7.9 68 -217 218 U=23 $ SS extension 124 6 -1 -210 #121 #122 #123 U=23 $ Water c ** REGULATING ROD: UNIVERSE = [24]

126 7 -2.12 219 U=24 $ Guide Tube 127 7 -2.699 53 -68 215 -216 U=24 $ Rod cladding 128 12 -2.48 53 215 U=24 $ Regulating rod 129 9 -7.9 68 218 -217 U=24 $ SS extension 130 6 -1 -219 #127 #128 #129 U=24 $ water c ******************************************************************************

c ** FUEL UNIVERSES: Individual FE's modeled for burnup as of 3-31-10 c ==============================================================================

20101 6 21 201 U=201 $ Water around lower pin 20102 9 -7.9 201 U=201 $ Bottom pin 20103 9 -7.9 -14 21 U=201 $ BOTTOM SS cap 20104 2 -1. 6 14 202 U=201 $ BOTTOM Axial reflector 20105 9 -7.9 14 -4 202 U=201 $ Cladding 20106 201 -5.685 13 -10 200 -202 U=201 $ ELEMENT B-1 S/N:6315 20107 8 -6.5 13 200 U=201 $ Zirc FILLER 20108 2 -1. 6 10 202 U=201 $ TOP Axial reflector 20109 3 -0.001239 6 202 U=201 $ Air gap 20110 9 -7.9 4 -3 U=201 $ TOP SS cap 20111 9 -7.9 3 -201 U=201 $ Upper pin 20112 6 -1 3 201 U=201 $ Water around pin 41

ATTACHMENT I c ==============================================================================

20201 like 20101 but u=202 20202 like 20102 but u=202 20203 like 20103 but u=202 20204 like 20104 but u=202 20205 like 20105 but u=202 20206 like 20106 but mat=202 rho=-5.681 U=202 $ELEMENT B-2 S/N:l0880 20207 like 20107 but U=202 20208 like 20108 but u=202 20209 like 20109 but u=202 20210 like 20110 but u=202 20211 like 20111 but U=202 20212 like 20112 but u=202 c ==============================================================================

20301 like 20101 but u=203 20302 like 20102 but u=203 20303 like 20103 but u=203 20304 like 20104 but u=203 20305 like 20105 but u=203 20306 like 20106 but mat=203 rho=-5.685 U=203 $ ELEMENT B-3 S/N:6577 20307 like 20107 but u=203 20308 like 20108 but u=203 20309 like 20109 but u=203 20310 like 20110 but u=203 20311 like 20111 but u=203 20312 like 20112 but u=203 c ==============================================================================

20401 like 20101 but u=204 20402 like 20102 but u=204 20403 like 20103 but u=204 20404 like 20104 but u=204 20405 like 20105 but u=204 20406 like 20106 but mat=204 rho=-5.688 U=204 $ ELEMENT B-4 S/N:2966 20407 like 20107 but u=204 20408 like 20108 but u=204 20409 like 20109 but u=204 20410 like 20110 but u=204 20411 like 20111 but u=204 20412 like 20112 but u=204 c ==============================================================================

20501 like 20101 but u=205 20502 like 20102 but u=205 20503 like 20103 but U=205 20504 like 20104 but u=205 20505 like 20105 but u=205 20506 like 20106 but mat=205 rho=-5.684 u=205 $ ELEMENT B-5 S/N:l0707 20507 like 20107 but u=205 20508 like 20108 but u=205 20509 like 20109 but u=205 20510 like 20110 but u=205 20511 like 20111 but u=205 20512 like 20112 but u=205 c ==============================================================================

20601 like 20101 but u=206 20602 like 20102 but u=206 20603 like 20103 but u=206 20604 like 20104 but U=206 20605 like 20105 but u=206 20606 like 20106 but mat=206 rho=-5.684 U=206 $ ELEMENT B-6 S/N:6525 20607 like 20107 but u=206 20608 like 20108 but u=206 20609 like 20109 but u=206 20610 like 20110 but u=206 20611 like 20111 but u=206 20612 like 20112 but u=206 c ==============================================================================

30101 like 20101 but u=301 30102 like 20102 but u=301 30103 like 20103 but u=301 30104 like 20104 but u=301 30105 like 20105 but u=301 42

ATTACHMENT!

30106 like 20106 but mat=301 rho=-5.680 u=301 $ ELEMENT C-1 S/N:11351 30107 like 20107 but u=301 30108 like 20108 but u=301 30109 like 20109 but u=301 30110 like 20110 but u=301 30111 like 20111 but u=301 30112 like 20112 but u=301 c ==============================================================================

30201 like 20101 but u=302 30202 like 20102 but u=302 30203 like 20103 but u=302 30204 like 20104 but u=302 30205 like 20105 but u=302 30206 like 20106 but mat=302 rho=-5.684 u=302 $ ELEMENT C-2 S/N:6316 30207 like 20107 but u=302 30208 like 20108 but u=302 30209 like 20109 but U=302 30210 like 20110 but U=302 30211 like 20111 but u=302 30212 like 20112 but u=302 c ==============================================================================

30301 like 20101 but U=303 30302 like 20102 but u=303 30303 like 20103 but u=303 30304 like 20104 but U=303 30305 like 20105 but u=303 30306 like 20106 but mat=303 rho=-5.691 u=303 $ ELEMENT C-3 S/N:2963 30307 like 20107 but u=303 30308 like 20108 but u=303 30309 like 20109 but u=303 30310 like 20110 but u=303 30311 like 20111 but u=303 30312 like 20112 but u=303 c ==============================================================================

30401 like 20101 but u=304 30402 like 20102 but U=304 30403 like 20103 but u=304 30404 like 20104 but u=304 30405 like 20105 but u=304 30406 like 20106 but mat=304 rho=-5.638 u=304 $ ELEMENT C-4 S/N:3329 30407 like 20107 but u=304 30408 like 20108 but u=304 30409 like 20109 but u=304 30410 like 20110 but u=304 30411 like 20111 but u=304 30412 like 20112 but u=304 c ==============================================================================

30501 like 20101 but u=305 30502 like 20102 but u=305 30503 like 20103 but u=305 30504 like 20104 but U=305 30505 like 20105 but u=305 30506 like 20106 but mat=305 rho=-5.710 U=305 $ ELEMENT C-5 S/N:2953 30507 like 20107 but u=305 30508 like 20108 but u=305 30509 like 20109 but u=305 30510 like 20110 but u=305 30511 like 20111 but U=305 30512 like 20112 but U=305 c ==============================================================================

30601 like 20101 but u=306 30602 like 20102 but u=306 30603 like 20103 but u=306 30604 like 20104 but u=306 30605 like 20105 but u=306 30606 like 20106 but mat=306 rho=-5.649 u=306 $ ELEMENT C-6 S/N:3082 30607 like 20107 but u=306 30608 like 20108 but u=306 30609 like 20109 but U=306 30610 like 20110 but u=306 30611 like 20111 but u=306 43

ATTACHMENT I 30612 like 20112 but u=306 c ==============================================================================

30801 like 20101 but u=308 30802 like 20102 but u=308 30803 like 20103 but U=308 30804 like 20104 but U=308 30805 like 20105 but u=308 30806 like 20106 but mat=308 rho=-5.693 u=308 $ ELEMENT C-8 S/N:2933 30807 like 20107 but u=308 30808 like 20108 but U=308 30809 like 20109 but u=308 30810 like 20110 but u=308 30811 like 20111 but u=308 30812 like 20112 but u=308 c ==============================================================================

30901 like 20101 but u=309 30902 like 20102 but u=309 30903 like 20103 but u=309 30904 like 20104 but u=309 30905 like 20105 but u=309 30906 like 20106 but mat=309 rho=-5.684 U=309 $ ELEMENT C-9 S/N:6314 30907 like 20107 but u=309 30908 like 20108 but u=309 30909 like 20109 but u=309 30910 like 20110 but u=309 30911 like 20111 but u=309 30912 like 20112 but U=309 c ==============================================================================

31001 like 20101 but U=310 31002 like 20102 but u=310 31003 like 20103 but u=310 31004 like 20104 but u=310 31005 like 20105 but u=310 31006 like 20106 but mat=310 rho=-5.684 u=310 $ ELEMENT C-10 S/N:6527 31007 like 20107 but u=310 31008 like 20108 but u=310 31009 like 20109 but U=310 31010 like 20110 but u=310 31011 like 20111 but u=310 31012 like 20112 but u=310 c ==============================================================================

31101 like 20101 but u=311 31102 like 20102 but u=311 31103 like 20103 but u=311 31104 like 20104 but U=311 31105 like 20105 but u=311 31106 like 20106 but mat=311 rho=-5.684 u=311 $ ELEMENT C-11 S/N:6317 31107 like 20107 but u=311 31108 like 20108 but U=311 31109 like 20109 but u=311 31110 like 20110 but u=311 31111 like 20111 but u=311 31112 like 20112 but U=311 c ==============================================================================

31201 like 20101 but u=312 31202 like 20102 but u=312 31203 like 20103 but u=312 31204 like 20104 but U=312 31205 like 20105 but u=312 31206 like 20106 but mat=312 rho=-5.684 u=312 $ ELEMENT C-12 S/N:6526 31207 like 20107 but u=312 31208 like 20108 but u=312 31209 like 20109 but U=312 31210 like 20110 but U=312 31211 like 20111 but U=312

-31212 like 20112 but u=312 c ======= ======================================================================

40101 1 ke 20101 but u=401 40102 1 ke 20102 but u=401 40103 1 ke 20103 but u=401 40104 1 ke 20104 but U=401 44

ATTACHMENT I 40105 like 20105 but U=401 40106 like 20106 but mat=401 rho=-5.637 U=401 $ ELEMENT D-1 S/N:3380 40107 like 20107 but u=401 40108 like 20108 but U=401 40109 like 20109 but u=401 40110 like 20110 but u=401 40111 like 20111 but u=401 40112 like 20112 but u=401 c ==============================================================================

40201 like 20101 but U=402 40202 like 20102 but u=402 40203 like 20103 but U=402 40204 like 20104 but u=402 40205 like 20105 but u=402 40206 like 20106 but mat=402 rho=-5.655 U=402 $ ELEMENT D-2 S/N:3330 40207 like 20107 but u=402 40208 like 20108 but u=402 40209 like 20109 but u=402 40210 like 20110 but u=402 40211 like 20111 but u=402 40212 like 20112 but u=402 c ==============================================================================

40301 like 20101 but u=403 40302 like 20102 but u=403 40303 like 20103 but u=403 40304 like 20104 but U=403 40305 like 20105 but u=403 40306 like 20106 but mat=403 rho=-5.626 U=403 $ ELEMENT D-3 S/N:5001 40307 like 20107 but u=403 40308 like 20108 but u=403 40309 like 20109 but u=403 40310 like 20110 but u=403 40311 like 20111 but U=403 40312 like 20112 but U=403 c ==============================================================================

40501 like 20101 but U=405 40502 like 20102 but U=405 40503 like 20103 but U=405 40504 like 20104 but u=405 40505 like 20105 but U=405 40506 like 20106 but mat=405 rho=-5.636 U=405 $ ELEMENT D-5 S/N:3144 40507 like 20107 but U=405 40508 like 20108 but U=405 40509 like 20109 but U=405 40510 like 20110 but u=405 40511 like 20111 but U=405 40512 like 20112 but u=405 c ==============================================================================

40601 like 20101 but u=406 40602 like 20102 but u=406 40603 like 20103 but U=406 40604 like 20104 but u=406 40605 like 20105 but u=406 40606 like 20106 but mat=406 rho=-5.646 U=406 $ ELEMENT D-6 S/N 6224 40607 like 20107 but u=406 40608 like 20108 but u=406 40609 like 20109 but u=406 40610 like 20110 but U=406 40611 like 20111 but U=406 40612 like 20112 but U=406 c ==============================================================================

40701 like 20101 but U=407 40702 like 20102 but u=407 40703 like 20103 but u=407 40704 like 20104 but u=407 40705 like 20105 but u=407 40706 like 20106 but mat=407 rho=-5.646 u=407 $ ELEMENT D-7 S/N:3018 40707 like 20107 but u=407 40708 like 20108 but u=407 40709 like 20109 but u=407 40710 like 20110 but u=407 45

ATTACHMENT I 40711 like 20111 but u=407 40712 like 20112 but U=407 c ==============================================================================

40801 like 20101 but u=408 40802 like 20102 but u=408 40803 like 20103 but U=408 40804 like 20104 but U=408 40805 like 20105 but U=408 40806 like 20106 but mat=408 rho=-5.636 U=408 $ ELEMENT D-8 S/N:3105 40807 like 20107 but u=408 40808 like 20108 but u=408 40809 like 20109 but U=408 40810 like 20110 but u=408 40811 like 20111 but u=408 40812 like 20112 but u=408 c ==============================================================================

40901 like 20101 but u=409 40902 like 20102 but U=409 40903 like 20103 but U=409 40904 like 20104 but u=409 40905 like 20105 but u=409 40906 like 20106 but mat=409 rho=-5.642 u=409 $ ELEMENT D-9 S/N 2452 40907 like 20107 but U=409 40908 like 20108 but u=409 40909 like 20109 but u=409 40910 like 20110 but u=409 40911 like 20111 but u=409 40912 like 20112 but U=409 c ==============================================================================

41001 like 20101 but U=410 41002 like 20102 but u=410 41003 like 20103 but U=410 41004 like 20104 but U=410 41005 like 20105 but u=410 41006 like 20106 but mat=410 rho=-5.632 u=410 $ ELEMENT D-10 S/N:2448 41007 like 20107 but u=410 41008 like 20108 but u=410 41009 like 20109 but u=410 41010 like 20110 but U=410 41011 like 20111 but U=410 41012 like 20112 but U=410 c ==============================================================================

41101 like 20101 but u=411 41102 like 20102 but u=411 41103 like 20103 but u=411 41104 like 20104 but U=411 41105 like 20105 but u=411 41106 like 20106 but mat=411 rho=-5.641 u=411 $ ELEMENT D-11 S/N:2435 41107 like 20107 but u=411 41108 like 20108 but U=411 41109 like 20109 but u=411 41110 like 20110 but u=411 41111 like 20111 but u=411 41112 like 20112 but u=411 c ==============================================================================

41201 like 20101 but u=412 41202 like 20102 but u=412 41203 like 20103 but u=412 41204 like 20104 but u=412 41205 like 20105 but u=412 41206 like 20106 but mat=412 rho=-5.655 U=412 $ ELEMENT D-12 S/N 3876 41207 like 20107 but u=412 41208 like 20108 but u=412 41209 like 20109 but u=412 41210 like 20110 but U=412 41211 like 20111 but U=412 41212 like 20112 but u=412 c ======= ======================================================================

41301 1 ke 20101 but U=413 41302 1 ke 20102 but u=413 41303 1 ke 20103 but u=413 46

ATTACHMENT I 41304 like 20104 but u=413 41305 like 20105 but u=413 41306 like 20106 but mat=413 rho=-5.644 u=413 $ ELEMENT D-13 S/N:3696 41307 like 20107 but u=413 41308 like 20108 but u=413 41309 like 20109 but u=413 41310 like 20110 but u=413 41311 like 20111 but u=413 41312 like 20112 but u=413 c ==============================================================================

41401 like 20101 but U=414 41402 like 20102 but u=414 41403 like 20103 but u=414 41404 like 20104 but U=414 41405 like 20105 but u=414 41406 like 20106 but mat=414 rho=-5.669 u=414 $ ELEMENT D-14 S/N:3494 41407 like 20107 but u=414 41408 like 20108 but u=414 41409 like 20109 but u=414 41410 like 20110 but U=414 41411 like 20111 but u=414 41412 like 20112 but u=414 c ==============================================================================

41501 like 20101 but U=415 41502 like 20102 but u=415 41503 like 20103 but u=415 41504 like 20104 but U=415 41505 like 20105 but u=415 41506 like 20106 but mat=415 rho=-5.656 u=415 $ ELEMENT D-15 S/N:3501 41507 like 20107 but u=415 41508 like 20108 but u=415 41509 like 20109 but U=415 41510 like 20110 but U=415 41511 like 20111 but U=415 41512 like 20112 but U=415 c ==============================================================================

41701 like 20101 but U=417 41702 like 20102 but U=417 41703 like 20103 but U=417 41704 like 20104 but U=417 41705 like 20105 but U=417 41706 like 20106 but mat=417 rho=-5.656 u=417 $ ELEMENT D-17 S/N:3498 41707 like 20107 but U=417 41708 like 20108 but u=417 41709 like 20109 but u=417 41710 like 20110 but u=417 41711 like 20111 but u=417 41712 like 20112 but u=417 c ==============================================================================

41801 like 20101 but u=418 41802 like 20102 but u=418 41803 like 20103 but u=418 41804 like 20104 but U=418 41805 like 20105 but u=418 41806 like 20106 but mat=418 rho=-5.643 U=418 $ ELEMENT D-18 S/N 3336 41807 like 20107 but U=418 41808 like 20108 but u=418 41809 like 20109 but u=418 41810 like 20110 but u=418 41811 like 20111 but u=418 41812 like 20112 but U=418 c ==============================================================================

50201 like 20101 but u=502 50202 like 20102 but u=502 50203 like 20103 but u=502 50204 like 20104 but u=502 50205 like 20105 but u=502 50206 like 20106 but mat=502 rho=-5.687 U=502 $ ELEMENT E-2 S/N:2986 50207 like 20107 but u=502 50208 like 20108 but u=502 50209 like 20109 but u=502 47

ATTACHMENT I 50210 like 20110 but u=502 50211 like 20111 but u=502 50212 like 20112 but u=502 c ==============================================================================

50301 like 20101 but u=503 50302 like 20102 but u=503 50303 like 20103 but u=503 50304 like 20104 but u=503 50305 like 20105 but u=503 50306 like 20106 but mat=503 rho=-5.632 U=503 $ ELEMENT E-3 S/N:2458 50307 like 20107 but U=503 50308 like 20108 but u=503 50309 like 20109 but u=503 50310 like 20110 but u=503 50311 like 20111 but u=503 50312 like 20112 but u=503 c ==============================================================================

50401 like 20101 but U=504 50402 like 20102 but u=504 50403 like 20103 but U=504 50404 like 20104 but U=504 50405 like 20105 but u=504 50406 like 20106 but mat=504 rho=-5.642 u=504 $ ELEMENT E-4 S/N:3006 50407 like 20107 but U=504 50408 like 20108 but u=504 50409 like 20109 but u=504 50410 like 20110 but U=504 50411 like 20111 but u=504 50412 like 20112 but U=504 c ==============================================================================

50501 like 20101 but U=505 50502 like 20102 but u=505 50503 like 20103 but u=505 50504 like 20104 but u=505 50505 like 20105 but u=505 50506 like 20106 but mat=505 rho=-5.625 u=505 $ ELEMENT E-5 S/N:5014 50507 like 20107 but u=505 50508 like 20108 but U=505 50509 like 20109 but U=505 50510 like 20110 but U=505 50511 like 20111 but U=505 50512 like 20112 but U=505 c ==============================================================================

50601 like 20101 but u=506 50602 like 20102 but u=506 50603 like 20103 but U=506 50604 like 20104 but U=506 50605 like 20105 but u=506 50606 like 20106 but mat=506 rho=-5.627 u=506 $ ELEMENT E-6 S/N:4744 50607 like 20107 but u=506 50608 like 20108 but u=506 50609 like 20109 but u=506 50610 like 20110 but U=506 50611 like 20111 but u=506 50612 like 20112 but u=506 c ==============================================================================

50701 like 20101 but u=507 50702 like 20102 but u=507 50703 like 20103 but u=507 50704 like 20104 but U=507 50705 like 20105 but u=507 50706 like 20106 but mat=700 rho=-5.892 u=507 $ ELEMENT E-7 50707 like 20107 but u=507 50708 like 20108 but U=507 50709 like 20109 but u=507 50710 like 20110 but U=507 50711 like 20111 but u=507 50712 like 20112 but u=507 c ==============================================================================

50801 like 20101 but u=508 50802 like 20102 but u=508 48

ATTACHMENT I 50803 like 20103 but U=508 50804 like 20104 but u=508 50805 like 20105 but u=508 50806 like 20106 but mat=508 rho=-5.624 u=508 $ ELEMENT E-8 S/N:4991 50807 like 20107 but u=508 50808 like 20108 but u=508 50809 like 20109 but u=508 50810 like 20110 but u=508 50811 like 20111 but u=508 50812 like 20112 but u=508 c ==============================================================================

50901 like 20101 but u=509 50902 like 20102 but u=509 50903 like 20103 but u=509 50904 like 20104 but u=509 50905 like 20105 but u=509 50906 like 20106 but mat=509 rho=-5.588 u=509 $ ELEMENT E-9 S/N:4742 50907 like 20107 but u=509 50908 like 20108 but u=509 50909 like 20109 but u=509 50910 like 20110 but u=509 50911 like 20111 but u=509 50912 like 20112 but u=509 c ==============================================================================

51001 like 20101 but u=510 51002 like 20102 but u=510 51003 like 20103 but U=510 51004 like 20104 but u=510 51005 like 20105 but u=510 51006 like 20106 but mat=510 rho=-5.617 u=510 $ ELEMENT E-10 S/N:4351 51007 like 20107 but u=510 51008 like 20108 but u=510 51009 like 20109 but u=510 51010 like 20110 but u=510 51011 like 20111 but u=510 51012 like 20112 but u=510 c ==============================================================================

51101 like 20101 but u=511 51102 like 20102 but u=511 51103 like 20103 but u=511 51104 like 20104 but u=511 51105 like 20105 but u=511 51106 like 20106 but mat=700 rho=-5.892 U=511 $ ELEMENT E-11 51107 like 20107 but u=511 51108 like 20108 but u=511 51109 like 20109 but u=511 51110 like 20110 but U=511 51111 like 20111 but u=511 51112 like 20112 but u=511 c ==============================================================================

51201 like 20101 but u=512 51202 like 20102 but u=512 51203 like 20103 but u=512 51204 like 20104 but u=512 51205 like 20105 but u=512 51206 like 20106 but mat=512 rho=-5.656 u=512 $ ELEMENT E-12 S/N:3690 51207 like 20107 but u=512 51208 like 20108 but u=512 51209 like 20109 but u=512 51210 like 20110 but u=512 51211 like 20111 but U=512 51212 like 20112 but u=512 c ==============================================================================

51301 like 20101 but u=513 51302 like 20102 but u=513 51303 like 20103 but u=513 51304 like 20104 but u=513 51305 like 20105 but u=513 51306 like 20106 but mat=513 rho=-5.693 U=513 $ ELEMENT E-13 S/N:2987 51307 like 20107 but u=513 51308 like 20108 but u=513 49

ATTACHMENT I 51309 like 20109 but u=513 51310 like 20110 but u=513 51311 like 20111 but u=513 51312 like 20112 but u=513 c ==============================================================================

51401 like 20101 but U=514 51402 like 20102 but U=514 51403 like 20103 but U=514 51404 like 20104 but U=514 51405 like 20105 but u=514 51406 like 20106 but mat=514 rho=-5.636 u=514 $ ELEMENT E-14 S/N:3118 51407 like 20107 but u=514 51408 like 20108 but U=514 51409 like 20109 but u=514 51410 like 20110 but U=514 51411 like 20111 but u=514 51412 like 20112 but u=514 c ==============================================================================

51501 like 20101 but U=515 51502 like 20102 but u=515 51503 like 20103 but u=515 51504 like 20104 but u=515 51505 like 20105 but u=515 51506 like 20106 but mat=515 rho=-5.692 U=515 $ ELEMENT E-15 S/N:2934 51507 like 20107 but u=515 51508 like 20108 but u=515 51509 like 20109 but u=515 51510 like 20110 but u=515 51511 like 20111 but U=515 51512 like 20112 but u=515 c ==============================================================================

51601 like 20101 but u=516 51602 like 20102 but u=516 51603 like 20103 but u=516 51604 like 20104 but u=516 51605 like 20105 but u=516 51606 like 20106 but mat=516 rho=-5.619 U=516 $ ELEMENT E-16 S/N:4343 51607 like 20107 but u=516 51608 like 20108 but u=516 51609 like 20109 but u=516 51610 like 20110 but u=516 51611 like 20111 but u=516 51612 like 20112 but u=516 c ==============================================================================

51701 like 20101 but u=517 51702 like 20102 but u=517 51703 like 20103 but u=517 51704 like 20104 but u=517 51705 like 20105 but u=517 51706 like 20106 but mat=517 rho=-5.670 u=517 $ ELEMENT E-17 S/N:3517 51707 like 20107 but u=517 51708 like 20108 but u=517 51709 like 20109 but U=517 51710 like 20110 but u=517 51711 like 20111 but u=517 51712 like 20112 but u=517 c ==============================================================================

51801 like 20101 but U=518 51802 like 20102 but u=518 51803 like 20103 but u=518 51804 like 20104 but u=518 51805 like 20105 but u=518 51806 like 20106 but mat=700 rho=-5.892 u=518 $ ELEMENT E-18 51807 like 20107 but U=518 51808 like 20108 but u=518 51809 like 20109 but u=518 51810 like 20110 but u=518 51811 like 20111 but U=518 51812 like 20112 but u=518 c ==============================================================================

51901 like 20101 but u=519 50

ATTACHMENTl 51902 like 20102 but u=519 51903 like 20103 but U=519 51904 like 20104 but U=519 51905 like 20105 but U=519 51906 like 20106 but mat=519 rho=-5.657 u=519 $ ELEMENT E-19 S/N:3503 51907 like 20107 but u=519 51908 like 20108 but U=519 51909 like 20109 but U=519 51910 like 20110 but u=519 51911 like 20111 but u=519 51912 like 20112 but u=519 c ==============================================================================

52001 like 20101 but u=520 52002 like 20102 but u=520 52003 like 20103 but U=520 52004 like 20104 but u=520 52005 like 20105 but u=520 52006 like 20106 but mat=520 rho=-5.649 U=520 $ ELEMENT E-20 S/N:3009 52007 like 20107 but u=520 52008 like 20108 but U=520 52009 like 20109 but U=520 52010 like 20110 but u=520 52011 like 20111 but u=520 52012 like 20112 but u=520 c ==============================================================================

52101 like 20101 but U=521 52102 like 20102 but u=521 52103 like 20103 but U=521 52104 like 20104 but u=521 52105 like 20105 but u=521 52106 like 20106 but mat=521 rho=-5.645 u=521 $ ELEMENT E-21 S/N:3011 52107 like 20107 but u=521 52108 like 20108 but u=521 52109 like 20109 but u=521 52110 like 20110 but u=521 52111 like 20111 but u=521 52112 like 20112 but u=521 c ==============================================================================

52201 like 20101 but u=522 52202 like 20102 but u=522 52203 like 20103 but u=522 52204 like 20104 but u=522 52205 like 20105 but u=522 52206 like 20106 but mat=522 rho=-5.646 U=522 $ ELEMENT E-22 S/N:3014 52207 like 20107 but u=522 52208 like 20108 but u=522 52209 like 20109 but u=522 52210 like 20110 but U=522 52211 like 20111 but u=522 52212 like 20112 but U=522 c ==============================================================================

52301 like 20101 but u=523 52302 like 20102 but u=523 52303 like 20103 but U=523 52304 like 20104 but U=523 52305 like 20105 but u=523 52306 like 20106 but mat=700 rho=-5.892 U=523 $ ELEMENT E-23 52307 like 20107 but u=523 52308 like 20108 but u=523 52309 like 20109 but u=523 52310 like 20110 but u=523 52311 like 20111 but u=523 52312 like 20112 but u=523 c ==============================================================================

52401 like 20101 but u=524 52402 like 20102 but U=524 52403 like 20103 but u=524 52404 like 20104 but u=524 52405 like 20105 but U=524 52406 like 20106 but mat=524 rho=-5.636 u=524 $ ELEMENT E-24 S/N:3111 52407 like 20107 but U=524 51

ATTACHMENT I 52408 like 20108 but u=524 52409 like 20109 but u=524 52410 like 20110 but u=524 52411 like 20111 but U=524 52412 like 20112 but u=524 c ==============================================================================

60101 like 20101 but u=601 60102 like 20102 but u=601 60103 like 20103 but U=601 60104 like 20104 but U=601 60105 like 20105 but u=601 60106 like 20106 but mat=601 rho=-5.612 u=601 $ ELEMENT F-1 S/N:5017 60107 like 20107 but u=601 60108 like 20108 but u=601 60109 like 20109 but u=601 60110 like 20110 but u=601 60111 like 20111 but u=601 60112 like 20112 but u=601 c ==============================================================================

60201 like 20101 but u=602 60202 like 20102 but u=602 60203 like 20103 but u=602 60204 like 20104 but u=602 60205 like 20105 but U=602 60206 like 20106 but mat=602 rho=-5.613 U=602 $ ELEMENT F-2 S/N:5018 60207 like 20107 but u=602 60208 like 20108 but u=602 60209 like 20109 but u=602 60210 like 20110 but u=602 60211 like 20111 but u=602 60212 like 20112 but u=602 c ==============================================================================

60301 like 20101 but u=603 60302 like 20102 but u=603 60303 like 20103 but U=603 60304 like 20104 but u=603 60305 like 20105 but u=603 60306 like 20106 but mat=603 rho=-5.626 u=603 $ ELEMENT F-3 S/N:5027 60307 like 20107 but u=603 60308 like 20108 but u=603 60309 like 20109 but u=603 60310 like 20110 but u=603 60311 like 20111 but U=603 60312 like 20112 but u=603 c ==============================================================================

60401 like 20101 but U=604 60402 like 20102 but u=604 60403 like 20103 but u=604 60404 like 20104 but u=604 60405 like 20105 but u=604 60406 like 20106 but mat=604 rho=-5.613 u=604 $ ELEMENT F-4 S/N:5021 60407 like 20107 but u=604 60408 like 20108 but u=604 60409 like 20109 but u=604 60410 like 20110 but u=604 60411 like 20111 but U=604 60412 like 20112 but u=604 c ==============================================================================

60501 like 20101 but u=605 60502 like 20102 but u=605 60503 like 20103 but u=605 60504 like 20104 but u=605 60505 like 20105 but u=605 60506 like 20106 but mat=605 rho=-5.613 u=605 $ ELEMENT F-5 S/N:5026 60507 like 20107 but u=605 60508 like 20108 but u=605 60509 like 20109 but u=605 60510 like 20110 but u=605 60511 like 20111 but u=605 60512 like 20112 but u=605 c ==============================================================================

52

ATTACHMENT I 60601 like 20101 but U=606 60602 like 20102 but u=606 60603 like 20103 but u=606 60604 like 20104 but u=606 60605 like 20105 but u=606 60606 like 20106 but mat=606 rho=-5.697 u=606 $ ELEMENT F-6 S/N:6223 60607 like 20107 but u=606 60608 like 20108 but u=606 60609 like 20109 but u=606 60610 like 20110 but u=606 60611 like 20111 but U=606 60612 like 20112 but u=606 c ==============================================================================

60701 like 20101 but U=607 60702 like 20102 but u=607 60703 like 20103 but u=607 60704 like 20104 but u=607 60705 like 20105 but U=607 60706 like 20106 but mat=607 rho=-5.605 u=607 $ ELEMENT F-7 S/N:5256 60707 like 20107 but u=607 60708 like 20108 but u=607 60709 like 20109 but U=607 60710 like 20110 but u=607 60711 like 20111 but u=607 60712 like 20112 but U=607 c ==============================================================================

60801 like 20101 but u=608 60802 like 20102 but U=608 60803 like 20103 but u=608 60804 like 20104 but u=608 60805 like 20105 but u=608 60806 like 20106 but mat=608 rho=-5.613 u=608 $ ELEMENT F-8 S/N:5254 60807 like 20107 but u=608 60808 like 20108 but u=608 60809 like 20109 but u=608 60810 like 20110 but u=608 60811 like 20111 but u=608 60812 like 20112 but u=608 c ==============================================================================

60901 like 20101 but u=609 60902 like 20102 but U=609 60903 like 20103 but u=609 60904 like 20104 but u=609 60905 like 20105 but u=609 60906 like 20106 but mat=609 rho=-5.613 u=609 $ ELEMENT F-9 S/N:5031 60907 like 20107 but U=609 60908 like 20108 but u=609 60909 like 20109 but u=609 60910 like 20110 but u=609 60911 like 20111 but u=609 60912 like 20112 but u=609 c ==============================================================================

61101 like 20101 but U=611 61102 like 20102 but u=611 61103 like 20103 but u=611 61104 like 20104 but u=611 61105 like 20105 but u=611 61106 like 20106 but mat=611 rho=-5.599 u=611 $ ELEMENT F-11 S/N:5949 61107 like 20107 but u=611 61108 like 20108 but u=611 61109 like 20109 but u=611 61110 like 20110 but u=611 61111 like 20111 but u=611 61112 like 20112 but u=611 c ==============================================================================

61201 like 20101 but U=612 61202 like 20102 but u=612 61203 like 20103 but u=612 61204 like 20104 but u=612 61205 like 20105 but u=612 61206 like 20106 but mat=612 rho=-5.703 u=612 $ ELEMENT F-12 S/N:2900 53

ATTACHMENT I 61207 like 20107 but u=612 61208 like 20108 but u=612 61209 like 20109 but u=612 61210 like 20110 but u=612 61211 like 20111 but U=612 61212 like 20112 but U=612 c ==============================================================================

61301 like 20101 but u=613 61302 like 20102 but u=613 61303 like 20103 but u=613 61304 like 20104 but u=613 61305 like 20105 but u=613 61306 like 20106 but mat=613 rho=-5.625 U=613 $ ELEMENT F-13 S/N:4339 61307 like 20107 but U=613 61308 like 20108 but u=613 61309 like 20109 but U=613 61310 like 20110 but u=613 61311 like 20111 but u=613 61312 like 20112 but u=613 c ==============================================================================

61401 like 20101 but U=614 61402 like 20102 but u=614 61403 like 20103 but u=614 61404 like 20104 but u=614 61405 like 20105 but u=614 61406 like 20106 but mat=614 rho=-5.600 u=614 $ ELEMENT F-14 S/N:5653 61407 like 20107 but u=614 61408 like 20108 but U=614 61409 like 20109 but u=614 61410 like 20110 but u=614 61411 like 20111 but u=614 61412 like 20112 but u=614 c ==============================================================================

61501 like 20101 but U=615 61502 like 20102 but U=615 61503 like 20103 but u=615 61504 like 20104 but U=615 61505 like 20105 but u=615 61506 like 20106 but mat=615 rho=-5.606 u=615 $ ELEMENT F-15 S/N:5654 61507 like 20107 but u=615 61508 like 20108 but u=615 61509 like 20109 but u=615 61510 like 20110 but u=615 61511 like 20111 but u=615 61512 like 20112 but u=615 c ==============================================================================

61601 like 20101 but u=616 61602 like 20102 but u=616 61603 like 20103 but u=616 61604 like 20104 but u=616 61605 like 20105 but u=616 61606 like 20106 but mat=616 rho=-5.599 u=616 $ ELEMENT F-16 S/N:5655 61607 like 20107 but u=616 61608 like 20108 but u=616 61609 like 20109 but U=616 61610 like 20110 but u=616 61611 like 20111 but u=616 61612 like 20112 but u=616 c ==============================================================================

61701 like 20101 but U=617 61702 like 20102 but u=617 61703 like 20103 but u=617 61704 like 20104 but u=617 61705 like 20105 but u=617 61706 like 20106 but mat=617 rho=-5.596 U=617 $ ELEMENT F-17 S/N:5939 61707 like 20107 but U=617 61708 like 20108 but U=617 61709 like 20109 but u=617 61710 like 20110 but u=617 61711 like 20111 but u=617 61712 like 20112 but U=617 54

ATTACHMENT I c ==============================================================================

61801 like 20101 but u=618 61802 like 20102 but U=618 61803 like 20103 but u=618 61804 like 20104 but u=618 61805 like 20105 but u=618 61806 like 20106 but mat=618 rho=-5.599 u=618 $ ELEMENT F-18 S/N:5946 61807 like 20107 but U=618 61808 like 20108 but u=618 61809 like 20109 but u=618 61810 like 20110 but u=618 61811 like 20111 but u=618 61812 like 20112 but U=618 c ==============================================================================

61901 like 20101 but u=619 61902 like 20102 but u=619 61903 like 20103 but u=619 61904 like 20104 but u=619 61905 like 20105 but u=619 61906 like 20106 but mat=619 rho=-5.634 u=619 $ ELEMENT F-19 S/N:3113 61907 like 20107 but u=619 61908 like 20108 but u=619 61909 like 20109 but u=619 61910 like 20110 but u=619 61911 like 20111 but u=619 61912 like 20112 but u=619 c ==============================================================================

62001 like 20101 but u=620 62002 like 20102 but U=620 62003 like 20103 but U=620 62004 like 20104 but u=620 62005 like 20105 but u=620 62006 like 20106 but mat=620 rho=-5.691 u=620 $ ELEMENT F-20 S/N:2949 62007 like 20107 but U=620 62008 like 20108 but u=620 62009 like 20109 but u=620 62010 like 20110 but u=620 62011 like 20111 but u=620 62012 like 20112 but u=620 c ==============================================================================

62101 like 20101 but u=621 62102 like 20102 but u=621 62103 like 20103 but U=621 62104 like 20104 but u=621 62105 like 20105 but U=621 62106 like 20106 but mat=621 rho=-5.612 U=621 $ ELEMENT F-21 S/N:5649 62107 like 20107 but U=621 62108 like 20108 but u=621 62109 like 20109 but u=621 62110 like 20110 but U=621 62111 like 20111 but u=621 62112 like 20112 but u=621 c ==============================================================================

62201 like 20101 but U=622 62202 like 20102 but U=622 62203 like 20103 but u=622 62204 like 20104 but U=622 62205 like 20105 but u=622 62206 like 20106 but mat=622 rho=-5.670 u=622 $ ELEMENT F-22 S/N:2917 62207 like 20107 but U=622 62208 like 20108 but u=622 62209 like 20109 but u=622 62210 like 20110 but U=622 62211 like 20111 but u=622 62212 like 20112 but u=622 c ==============================================================================

62301 like 20101 but u=623 62302 like 20102 but u=623 62303 like 20103 but u=623 62304 like 20104 but U=623 62305 like 20105 but U=623 55

ATTACHMENT I 62306 like 20106 but mat=623 rho=-5.599 u=623 $ ELEMENT F-23 S/N:5000 62307 like 20107 but u=623 62308 like 20108 but u=623 62309 like 20109 but u=623 62310 like 20110 but u=623 62311 like 20111 but u=623 62312 like 20112 but u=623 c ==============================================================================

62401 like 20101 but u=624 62402 like 20102 but u=624 62403 like 20103 but u=624 62404 like 20104 but u=624 62405 like 20105 but U=624 62406 like 20106 but mat=624 rho=-5.596 u=624 $ ELEMENT F-24 S/N:5948 62407 like 20107 but U=624 62408 like 20108 but u=624 62409 like 20109 but u=624 62410 like 20110 but u=624 62411 like 20111 but u=624 62412 like 20112 but u=624 c ==============================================================================

62501 like 20101 but U=625 62502 like 20102 but u=625 62503 like 20103 but u=625 62504 like 20104 but u=625 62505 like 20105 but u=625 62506 like 20106 but mat=625 rho=-5.603 u=625 $ ELEMENT F-25 S/N:5022 62507 like 20107 but u=625 62508 like 20108 but u=625 62509 like 20109 but U=625 62510 like 20110 but u=625 62511 like 20111 but U=625 62512 like 20112 but U=625 c ==============================================================================

62601 like 20101 but u=626 62602 like 20102 but u=626 62603 like 20103 but u=626 62604 like 20104 but U=626 62605 like 20105 but U=626 62606 like 20106 but mat=626 rho=-5.701 u=626 $ ELEMENT F-26 S/N:2907 62607 like 20107 but U=626 62608 like 20108 but u=626 62609 like 20109 but U=626 62610 like 20110 but u=626 62611 like 20111 but u=626 62612 like 20112 but u=626 c ==============================================================================

62701 like 20101 but u=627 62702 like 20102 but U=627 62703 like 20103 but u=627 62704 like 20104 but u=627 62705 like 20105 but U=627 62706 like 20106 but mat=627 rho=-5.609 u=627 $ ELEMENT F-27 S/N:5944 62707 like 20107 but u=627 62708 like 20108 but U=627 62709 like 20109 but u=627 62710 like 20110 but U=627 62711 like 20111 but u=627 62712 like 20112 but U=627 c ==============================================================================

62801 like 20101 but u=628 62802 like 20102 but u=628 62803 like 20103 but U=628 62804 like 20104 but u=628 62805 like 20105 but U=628 62806 like 20106 but mat=628 rho=-5.638 u=628 $ ELEMENT F-28 S/N:3326 62807 like 20107 but u=628 62808 like 20108 but u=628 62809 like 20109 but u=628 62810 like 20110 but u=628 62811 like 20111 but U=628 56

ATTACHMENT 1 62812 like 20112 but U=628 c ==============================================================================

62901 like 20101 but u=629 62902 like 20102 but U=629 62903 like 20103 but u=629 62904 like 20104 but u=629 62905 like 20105 but U=629 62906 like 20106 but mat=629 rho=-5.682 u=629 $ ELEMENT F-29 S/N:2914 62907 like 20107 but u=629 62908 like 20108 but u=629 62909 like 20109 but U=629 62910 like 20110 but u=629 62911 like 20111 but U=629 62912 like 20112 but u=629 c ==============================================================================

63001 like 20101 but U=630 63002 like 20102 but u=630 63003 like 20103 but u=630 63004 like 20104 but U=630 63005 like 20105 but U=630 63006 like 20106 but mat=630 rho=-5.681 u=630 $ ELEMENT F-30 S/N:2909 63007 like 20107 but u=630 63008 like 20108 but u=630 63009 like 20109 but U=630 63010 like 20110 but u=630 63011 like 20111 but u=630 63012 like 20112 but U=630 c ==============================================================================

c ** RABBIT: UNIVERSE = [2]

145 7 -2.699 (25-12):(12 219): (-25 -45) u=2 $ Not tube 146 3 -0.001239 12 -219 u=2 $ air 147 6 -1 #145 #146 u=2 $water c ** GRAPHITE ROD: UNIVERSE = [3]

160 6 21 201 u=3 $ Water around lower pin 161 9 -7.9 201 u=3 $ Bottom pin 162 9 -7.9 -14 21 U=3 $ BOTTOM SS cap 163 9 -7.9 14 -4 202 u=3 $ Cladding 164 2 -1.6 14 202 U=3 $ GRAPHITE 165 9 -7.9 4 -3 U=3 $ TOP SS cap 166 9 -7.9 3 -201 u=3 $ Upper pin 167 6 -1 3 201 u=3 $ Water around pin c ** CENTRAL THIMBLE: UNIVERSE [4]

170 6 -1 12 -202 u=4 $ Water over midplane 171 7 -2.699 -12 :202 u=4 $ Below centerline c ** SOURCE: UNIVERSE = [5]

175 7 -2.699 -233 :230 :232 U=5 $ SOURCE HOLDER 176 9 -0.394473 233 -234 -231 U=5 $ SOURCE, SS SHELL 177 3 -0.001239 -232 233 -230 #176 U=5 $ SOURCE CAVITY c ** THIMBLE: UNIVERSE = [6]

180 6 21 201 u=6 $ Water around lower pin 181 7 -2.699 201 u=6 $ Bottom pin 182 7 -2.699 21 -14 U=6 $ BOTTOM cap 183 7 -2.699 14 210 U=6 $ Thimble TUBE 184 6 -1 14 -210 u=6 $ Thimble Water c ** CORE & FUEL: UNIVERSE = [7]

c ** WATER AROUND UPPER PINS & GRID PLATE ***********************************

185 6 1 5 1201 1202 1203 1204 1205 1206 1301 1302 1303 $ H20 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623 1624 1625 1626 1627 1628 1629 1630 203 U=7 c ** UPPER GRID PLATE *******************************************************

186 7 -2.699 -5 8 1201 1202 1203 1204 1205 1206 1301 $ Ul7ER GRID PLAT 1302 1303 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 57

ATTACHMENT 1 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623 1624 1625 1626 1627 1628 1629 1630 203 U=7 c ** CORE WATER *************************************************************

187 6 8 1201 1202 1203 1204 1205 1206 1301 1302 1303 $ CORE H20 1304 1305 1306 1307 1308 1309 1310 1311 1312 1401 1402 1403 1404 1405 1406 1407 1408 1409 1410 1411 1412 1413 1414 1415 1416 1417 1418 1501 1502 1503 1504 1505 1506 1507 1508 1509 1510 1511 1512 1513 1514 1515 1516 1517 1518 1519 1520 1521 1522 1523 1524 1601 1602 1603 1604 1605 1606 1607 1608 1609 1610 1611 1612 1613 1614 1615 1616 1617 1618 1619 1620 1621 1622 1623 1624 1625 1626 1627 1628 1629 1630 203 U=7 c ** FUEL POSITIONS *********************************************************

c ** central thimble 1100 0 -203 fill=4 U=7 $ CT c ** B Ring 2,01 0 -1201 fill=201 (0 4.05384 0 ) U=7 $ B-1 202 0 -1202 fill=202 (3.510728 2.02692 0 ) U=7 $ B-2 203 0 -1203 fill=203 (3 .510728 -2 .02692 0 ) U=7 $ B-3 204 0 -1204 fill=204 (0 -4 .05384 0 ) U=7 $ B-4 205 0 -1205 fill=205 (-3. 510728 -2. 02692 0 ) U=7 $ B-5 206 0 -1206 fill=206 (-3.510728 2.02692 0 ) U=7 $ B-6 c ** C Ring 301 0 -1301 fill=301 (0 7. 98068 0 ) U=7 $ C-1 302 0 -1302 fill=302 (3.99034 6.911472 0 U=7 $ C-2 303 0 -1303 fill=303 (6.911472 3.99034 0 U=7 $ C-3 304 0 -1304 fill=304 (7. 98068 0 0 ) U=7 $ C-4 305 0 -1305 fill=305 (6.911472 -3.99034 0 ) U=7 $ C-5 306 0 -1306 fill=306 (3.99034 -6.911472 0 ) U=7 $ C-6 c ******************** PULSE ROD **********************************************

307 0 -1307 fill=21 (0 -7.98068 39.37 ) U=7 $ C-7 c *** 0.03937 cm/unit ***OUT 60% 23.622 ***OUT 100% 39.37 *** CRIT 15.622 c *****************************************************************************

c ** THIMBLE=6 WATER=8 FUEL=l 308 0 -1308 fill=308 (-3.99034 -6.911472 0 U=7 $ C-8 309 0 -1309 fill=309 (-6.911472 -3.99034 0 U=7 $ C-9 310 0 -1310 fill=310 (-7.98068 0 0 ) U=7 $ C-10 311 0 -1311 fill=311 (-6.911472 3.99034 0 U=7 $ C-11 312 0 -1312 fill=312 (-3.99034 6.911472 0 U=7 $ C-12 c ** D Ring 401 0 -1401 fill=401 (0 11.94562 0 ) U=7 $ D-1 402 0 -1402 fill=402 (4.085643 11.22521 0 ) U=7 $ D-2 403 0 -1403 fill=403 (7. 678497 9 .150876 0 ) U=7 $ D-3 c ****************************SAFETY ROD***************************************

404 0 -1404 fill=22 (10. 34521 5. 97281 33. 02 ) U=7 $ D-4 c *** 0.03302 cm/unit ***OUT 60% 19.812 ***OUT 100% 33.02 *** CRIT 11.02 c *****************************************************************************

405 0 -1405 fill=405 (11.76414 2.074335 0 ) U=7 $ D-5 406 0 -1406 fill=406 (11. 76414 -2. 074335 0 ) U=7 $ D-6 407 0 -1407 fill=407 (10.34521 -5.97281 0 ) U=7 $ D-7 408 0 -1408 fill=408 (7.678497 -9.150876 0 ) U=7 $ D-8 409 0 -1409 fill=409 (4.085643 -11.22521 0 ) U=7 $ D-9 410 0 -1410 fill=410 (0 -11.94562 0 ) U=7 $ D-10 411 0 -1411 fill=411 (-4.085643 -11.22521 0 ) U=7 $ D 11 412 0 -1412 fill=412 (-7.678497 -9.150876 0 ) U=7 $ D 12 413 0 -1413 fill=413 (-10. 34521 -5. 97281 0 ) U=7 $ D 13 414 0 -1414 fill=414 (-11.76414 -2.074335 0 ) U=7 $ D-14 415 0 -1415 fill=415 (-11.76414 2.074335 0 ) U=7 $ D 15 c ****************************SHIM ROD******************************************

416 0 -1416 fill=23 (-10.34521 5.97281 43.18 ) U=7 $ D-16 C *** 0.04318 cm/unit *** 60% OUT 25.908 *** 100% OUT 43.18 *** CRIT 15.908 c ******************************************************************************

417 0 -1417 fill=417 (-7.678497 9.150876 0 ) U=7 $ D-17 418 0 -1418 fill=418 (-4. 085643 11. 22511 0 ) U=7 $ D-18 c ** E Ring c ****************************REGULATING ROD*************************************

501 0 -1501 fill=24 (0 15.91564 41.91 ) U=7 $ E-1 c *** 0.04191 cm/unit *** 60% OUT 25.146 *** 100% OUT 41.91 *** CRIT 0 c ******************************************************************************

502 0 -1502 fill=502 (4.119271 15.37333 0 ) U=7 $ E-2 503 0 -1503 fill=503 (7. 95782 13. 78335 0 ) U=7 $ E-3 504 0 -1504 fill=504 (11.25406 11.25406 0 ) U=7 $ E-4 58

ATTACHMENT 1 505 0 -1505 fill=505 (13.78335 7.95782 0 ) U=7 $ E-5 506 0 -1506 fill=506 (15.37333 4.119271 0 ) U=7 $ E-6 507 0 -1507 fill=507 (15. 91564 0 0 ) U=7 $ E-7 508 0 -1508 fill=508 (15. 37333 -4 .119271 0 ) U=7 $ E-8 509 0 -1509 fill=509 (13. 78335 -7. 95782 0 ) U=7 $ E-9 510 0 -1510 fill=510 (11.25406 -11.25406 0 ) U=7 $ E-10 511 0 -1511 fill=511 (7.95782 -13.78335 0 ) U=7 $ E-11 512 0 -1512 fill=512 (4.119271 -15.37333 0 ) U=7 $ E-12 513 0 -1513 fill=513 (0 -15. 91564 0 ) U=7 $ E-13 514 0 -1514 fill=514 (-4,119271 -15.37333 0 ) U=7 $ E-14 515 0 -1515 fill=515 (-7.95782 -13.78335 0 ) U=7 $ E-15 516 0 -1516 fill=516 (-11.25406 -11.25406 0 ) U=7 $ E-16 517 0 -1517 fill=517 (-13.78335 -7.95782 0 ) U=7 $ E-17 518 0 -1518 fill=518 (-15.37333 -4 .119271 0 ) U=7 $ E-18 519 0 -1519 fill=519 (-15.91564 0 0 ) U=7 $ E-19 520 0 -1520 fill=520 (-15.37333 4.119271 0 ) U=7 $ E-20 521 0 -1521 fill=521 (-13.78335 7.95782 0 ) U=7 $ E-21 522 0 -1522 fill=522 (-11.25406 11.25406 0 ) U=7 $ E-22 523 0 -1523 fill=523 (-7 .95782 13. 78335 ) U=7 $ E-23 524 0 -1524 fill=524 (-4.119271 15.37333 0 U=7 $ E-24 c ** F Ring 601 0 -1601 fill=601 (0 19.8882 0 ) U=7 $ F-1 602 0 -1602 fill=602 (4.134989 19.45359 0 ) U=7 $ F-2 603 0 -1603 fill=603 (8.08926 18.16878 0 ) U=7 $ F-3 604 0 -1604 fill=604 (11. 68999 16. 08989 0 ) U=7 $ F-4 605 0 -1605 fill=605 (14.77981 13.30878 0 ) U=7 $ F-5 606 0 -1606 fill=606 (17. 22369 9. 9441 0 ) U=7 $ F-6 607 0 -1607 fill=607 (18. 9148 6 .145792 0 ) U=7 $ F-7 608 0 -1608 fill=608 (19. 77925 2 .078883 0 ) U=7 $ F-8 609 0 -1609 fill=609 (19. 77925 -2. 078883 0 ) U=7 $ F-9 c ** SOURCE IS UNIVERSE 5, WATER VOID IS UNIVERSE 8 610 0 -1610 fill=5 (18. 9148 -6 .145792 0 ) U=7 $ F-10 SRC 5 611 0 -1611 fill=611 (17.22369 -9.9441 0 ) U=7 $ F-11 612 0 -1612 fill=612 (14.77981 -13.3078 0 ) U=7 $ F-12 c ** TEST FOR FUEL WORTH (fill=l) COMPARED TO WATER VOID (fill=8) 613 0 -1613 fill=613 (11. 68999 -16. 08989 0 ) U=7 $ F-13 ****

614 0 -1614 fill=614 (8. 08926 -18 .16878 0 ) U=7 $ F-14 615 0 -1615 fill=615 (4 .134989 -19. 45359 0 ) U=7 $ F-15 616 0 -1616 fill=616 (0 -19. 8882 0 ) U=7 $ F-16 617 0 -1617 fill=617 (-4.134989 -19.45359 0 ) U=7 $ F-17 618 0 -1618 fill=618 (-8.08926 -18.16878 0 ) U=7 $ F-18 c ** TEST FOR FUEL WORTH (fill=l) COMPARED TO WATER VOID (fill=8) 619 0 -1619 fill=619 (-11.68999 -16.08989 0 ) U=7 $ F-19 ****

620 0 -1620 fill=620 (-14. 77981 -13 .3078 0 ) U=7 $ F-20 621 0 -1621 fill=621 (-17. 22369 -9. 9441 0 ) U=7 $ F-21 622 0 -1622 fill=622 (-18.9148 -6.145792 0 ) U=7 $ F-22 623 0 -1623 fill=623 (-19.77925 -2.078883 0 ) U=7 $ F-23 C ** RABBIT IS UNIVERSE 2, WATER VOID IS UNIVERSE 8, RABBIT REMOVED FOR CORE III-4 624 0 -1624 fill=624 (-19.77925 2.078883 0 ) U=7 $ F-24 Old rabbit location 625 0 -1625 fill=625 (-18.9148 6.145792 0 ) U=7 $ F-25 626 0 -1626 fill=626 0-17.22369 9.9441 0 ) U=7 $ F-26 627 0 -1627 fill=627 (-14.77981 13.3078 0 ) U=7 $ F-27 628 0 -1628 fill=628 (-11.68999 16.08989 0) U=7 $ F-28 629 0 -1629 fill=629 (-8.08926 18.16878 0 ) U=7 $ F-29 630 0 -1630 fill=630 (-4.134989 19.45359 0 ) U=7 $ F-30 c SURFACES c *** ELEVATIONS **************************************************************

1 pz 50 $ POOL SURFACE 2 pz 37.2618 $TOP OF FUEL PINS 3 pz 33.782 $Top of end cap 4 pz 30.3276 $Top of gap/CLADDING 5 pz 33.79 $Top of grid plate 6 pz 27.7876 $Top of axial reflector 7 pz 27.94 $ REFLECTOR CANNISTER 8 pz 31.162 $ **** Bottom top grid 9 pz 26.035 $RADIAL REFLECTOR TOP 10 pz 19.05 $Top of fuel 11 pz 0.635 $ RSR FLOOR 12 pz 0 $ RSR SUBFLOOR -- LINE OF FUEL SYMMETRTY 59

ATTACHMENT 1 13 pz -19.05 $ Bottom of fuel (38.1/2 cm) c ** BEST ESTIMATE 3.74 in= 9.4996 cm {-19.05 cm -9.4996 cm}=>

14 pz -27.7876 $Bottom of axial reflector/CLADDING c ** APPROX CONE 3.797 cm base 3.797 cm heigth 21 pz -32 $ Bottom of plug/top of pin 15 pz -34.163 $ Top of bottom grid plate 16 pz -26.035 $ RADIAL REFLECTOR BOTTOM 17 pz -27.305 $REFLECTOR CANNISTER BOTTOM 18 pz -36.07 $ Bottom of grid plate 20 pz -50 $ BOTTOM OF POOL c *** OTHER PLANES *************************************************************

25 pz -5.0 $ Bottom of thick part of rabbit 26 px 0 $ CORE 1/2 c *** Z CYLINDERS **************************************************************

30 CZ 22.85 $ CORE/REFLECTOR WALL 31 CZ 23.485 $ INNER REFLECTOR 32 CZ 33.02 $ RSR INNER WALL 33 CZ 33.665 $ RSR INNER WALL 34 CZ 37.465 $ RSR OUTER VOLUME 35 CZ 38.1 $ RSR OUTER WALL 36 CZ 53.35 $ OUTER REFLECTOR 37 CZ 53.975 $ OUTER REFLECTOR WALL 38 CZ 60 $ EDGE OF POOL c *** OTHER CYLINDERS **********************************************************

40 1 c/x 0 -8.3 8.255 $ NE/SW BEAM PORT 90 1 c/x 0 -8.3 7.62 $ NE/SW BEAM PORT 43 CZ 22 $ LOWER GRID PLATE RADIUS 44 CZ 24.75 $ UPPER GRID PLATE RADIUS 45 CZ 1.0 $ Bottom part of rabbit, above pin c *** FLUX PROBE HOLES *********************************************************

50 pz -12.4373 $ Pulse rod, span 39.37 cm 51 pz -12.6811 $ Safety rod, span 33.02 cm 52 pz -26.0390 $Shim rod, span 43.18 53 pz -21.3628 $ Regulating rod, span 41.91 68 pz 24.257 c ** UNIVERSE FUEL PIN cylinders ***********************************************

200 CZ 0.2285 $ Zirc FILLer 201 CZ 0.635 $ Pin diameter 202 CZ 1.8477 $ Fuel diameter MARK III FUEL SCHEMATIC 203 CZ 1.985 $ CT & RABBIT c ** UNIVERSE CONTROL ROD cylinders ********************************************

210 cz 1.69165 $ Guide tube outer radius 1.89865 cm 219 cz 1.58815 $ REG guide tube surface 211 cz 1.511 $ Pulse meat 212 cz 1.587 $ Pulse rod cladding surface 213 cz 1.5113 $ Shim/safety meat 214 cz 1.5875 $ Shim/safety rod cladding surface 215 cz 1.03505 $ Rod reg meat 216 cz 1.11125 $Reg rod cladding surface 217 cz 1.254 $ Extension tube outer radius 218 cz 0.619 $ Extesion tube inner radius c ** UNIVERSE SOURCE ***********************************************************

230 cz 0.981 $SOURCE HOLDER CAVITY 231 cz 0.9525 $ SOURCE RADIUS c source elevations ALSO USES SOME FUEL ELEVATIONS 232 pz 3.8 $ Source cavity top 233 pz -3.8 $ Source cavity bottom 234 pz 2.55 $Source heigth c ******* END OF UNIVERSE SURFACE DEFINTIONS ***********************************

c ** A RING USES SURFACE 203 ***************************************************

c 1000 c/z O O 1.8985 $ A-1 = surface 203 c ** B Ring ********************************************************************

1201 c/z 0 4.05384 1.8985$ B-1 1202 c/z 3. 510728 2.02692 1.8985 $ B-2 1203 c/z 3. 510728 -2.02692 1.8985 $ B-3 1204 c/z 0 -4.05384 1.8985 $ B-4 1205 c/z -3.510728 -2.02692 1. 8985 $ B-5 1206 c/z -3.510728 2.02692 1.8985 $ B-6 c ** c Ring ********************************************************************

1301 c/z O 7.98068 1.8985$ c-1 1302 c/z 3.99034 6.911472 1.8985 $ C-2 60

ATTACHMENT 1 1303 c/z 6.911472 3.99034 1. 8985 $ C-3 1304 c/z 7.98068 0 1.8985 $ C-4 1305 c/z 6.911472 -3.99034 1.8985 $ C-5 1306 c/z 3.99034 -6.911472 1.8985 $ C-6 1307 c/z 0 -7.98068 1.8985 $ C-7 1308 c/z -3.99034 -6 .911472 1.8985 $ C-8 1309 c/z -6.911472 -3.99034 1.8985 $ C-9 1310 c/z -7.98068 0 1.8985 $ C-10 1311 c/z -6.911472 3.99034 1.8985 $ C-11 1312 c/z -3.99034 6.911472 1.8985 $ C-12 c ** D Ring ********************************************************************

1401 c/z 0 11. 94562 1.8985 $ D-1 1402 c/z 4.085643 11.225211 1.8985 $ D-2 1403 c/z 7.678497 9.150876 1. 8985 $ D-3 1404 c/z 10.34521 5.97281 1.8985 $ D-4 1405 c/z 11.764139 2.074335 1.8985 $ D-5 1406 c/z 11. 764139 -2.074335 1.8985 $ D-6 1407 c/z 10.34521 -5.97281 1.8985 $ D-7 1408 c/z 7.678497 -9.150876 1.8985 $ D-8 1409 c/z 4 .119271 -11.225211 1.8985 $ D-9 1410 c/z 0 -11. 94562 1. 8985 $ D-10 1411 c/z -4.119271 -11.225211 1.8985 $ D 11 1412 c/z -7.678497 -9.150876 1. 8985 $ D 12 1413 c/z -10.34521 -5.97281 1.8985 $ D 13 1414 c/z -11. 764139 -2.074335 1.8985 $ D-14 1415 c/z -11. 764139 2.074335 1.8985 $ D 15 1416 c/z -10.34521 5.97281 1.8985 $ D-16 1417 c/z -7.678497 9.150876 1.8985 $ D-17 1418 c/z -4 .119271 11. 225211 1.8985 $ D-18 c ** E RING ********************************************************************

1501 c/z 0 15.91564 1.8985 $ E-1 1502 c/z 4 .119271 15.373328 1. 8985 $ E-2 1503 c/z 7.95782 13. 783349 1.8985 $ E-3 1504 c/z 11.254057 11.254057 1. 8985 $ E-4 1505 c/z 13.783349 7.95782 1.8985 $ E-5 1506 c/z 15.373328 4.119271 1.8985 $ E-6 1507 c/z 15.91564 0 1. 8985 $ E-7 1508 c/z 15.373328 -4 .119271 1.8985 $ E-8 1509 c/z 13.783349 -7.95782 1.8985 $ E-9 1510 c/z 11.254057 -11. 254057 1.8985 $ E-10 1511 c/z 7.95782 -13.783349 1.8985 $ E-11 1512 c/z 4 .119271 -15.373328 1.8985 $ E-12 1513 c/z 0 -15.91564 1.8985 $ E-13 1514 c/z -4 .119271 -15.373328 1.8985 $ E-14 1515 c/z -7.95782 -13.783349 1.8985 $ E-15 1516 c/z -11. 254057 -11. 254057 1.8985 $ E-16 1517 c/z -13. 783349 -7.95782 1.8985 $ E-17 1518 c/z -15.373328 -4 .119271 1.8985 $ E-18 1519 c/z -15.91564 0 1.8985 $ E-19 1520 c/z -15.373328 4 .119271 1.8985 $ E-20 1521 c/z -13.783349 7.95782 1.8985 $ E-21 1522 c/z -11. 254057 11.254057 1.8985 $ E-22 1523 c/z -7.95782 13.783349 1.8985 $ E-23 1524 c/z -4 .119271 15.373328 1.8985 $ E-24 c ** F Ring ********************************************************************

1601 c/z 0 19.8882 1.8985 $ F-1 1602 c/z 4.134989 19.453595 1.8985 $ F-2 1603 c/z 8.08926 18.168775 1.8985 $ F-3 1604 c/z 11.689991 16.089892 1.8985 $ F-4 1605 c/z 14.779813 13.3087803 1.8985 $ F-5 1606 c/z 17.223686 9.9441 1.8985 $ F-6 1607 c/z 18.914802 6.145792 1.8985 $ F-7 1608 c/z 19. 77925 2.078883 1.8985 $ F-8 1609 c/z 19. 77925 -2.078883 1. 8985 $ F-9 1610 c/z 18.914802 -6.145792 1. 8985 $ F-10 SOURCE 1611 c/z 17.223686 -9.9441 1.8985 $ F-11 1612 c/z 14. 779813 -13 .307803 1.8985 $ F-12 1613 c/z 11.689991 -16.089892 1.8985 $ F-13 1614 c/z 8.08926 -18.168775 1.8985 $ F-14 1615 c/z 4.134989 -19.453595 1.8985 $ F-15 1616 c/z 0 -19.8882 1. 8985 $ F-16 61

ATTACHMENT 1 1617 c/z -4.134989 -19.453595 1. 8985 $ F-17 1618 c/z -8.08926 -18.168775 1.8985 $ F-18 1619 c/z -11. 689991 -16.089892 1.8985 $ F-19 1620 c/z -14.779813 -13.307803 1.8985 $ F-20 1621 c/z -17.223686 -9.9441 1.8985 $ F-21 1622 c/z -18.914802 -6.145792 1.8985 $ F-22 1623 c/z -19.77925 -2.078883 1. 8985 $ F-23 RABBIT 1624 c/z -19.77925 2.078883 1. 8985 $ F-24 1625 c/z -18.914802 6.145792 1. 8985 $ F-25 1626 c/z -17.223686 9.9441 1.8985 $ F-26 1627 c/z -14.779813 13.307803 1.8985 $ F-27 1628 c/z -11. 689991 16.089892 1.8985 $ F-28 1629 c/z -8.08926 18.168775 1.8985 $ F-29 1630 c/z -4.134989 19.453595 1.8985 $ F-30 trl O O O 0.6428 0.776 O -0.776 0.6428 o O O 1 mode n kcode 100000 1.000000 10 110 c MATERIAL CARDS c u + Zr-H rho=5.7473 g/cmA3 BOL c S(a,b)- Zr-H@ 300 K c ******************************************************************************

c This current model only accounts for the burnup of U235 c ** B-1 ** U235 left after Burnup = 37.012 g **Density= 5.685 m201 92235.66c -0.016144 92238.66c -0.066645 40000.66c -0.900788 1001. 66c -0.016422 c ** B-2 ** U235 left after Burnup 36.731 g **Density 5.681 m202 92235.66c -0.016022 92238.66c -0.066140 40000.66c -0.901405 1001.66c -0.016433 c ** B-3 ** U235 left after Burnup 37.012 g **Density 5.685 m203 92235.66c -0.016144 92238.66c -0.066645 40000.66c -0.900788 1001.66c -0.016422 c ** B-4 ** U235 left after Burnup 37.315 g **Density 5.688 m204 92235.66c -0.016276 92238.66c -0 .067191 40000.66c -0.900123 1001.66c -0.016410 c ** B-5 ** U235 left after Burnup 36.985 g ** Density 5.684 m205 92235.66c -0.016132 92238.66c -0.066598 40000.66c -0.900847 1001. 66c -0.016423 c ** B-6 ** U235 left after Burnup 37.034 g **Density 5.684 m206 92235.66c -0.016154 92238.66c -0.066686 40000.66c -0.900739 1001.66c -0.016421 c ** C-1 ** U235 left after Burnup 36.691 g **Density 5.680 m301 92235.66c -0.016004 92238.66c -0.066068 40000.66c -0.901493 1001. 66c -0.016435 c ** C-2 ** U235 left after Burnup 37.012 g **Density 5.684 m302 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-3 ** U235 left after Burnup 37.515 g ** Density 5.691 m303 92235.66c -0.016363 92238.66c -0.067552 40000.66c -0.899683 1001.66c -0.016402 c ** C-4 ** U235 left after Burnup 33.388 g **Density 5.638 m304 92235.66c -0.014563 92238.66c -0.060120 62

ATTACHMENT 1 40000.66c -0.908749 1001.66c -0.016567 c ** C-5 ** U235 left after Burnup 39.023 g ** Density 5. 710 m305 92235.66c -0.017021 92238.66c -0.070267 40000.66c -0 .896370 1001.66c -0.016342 c ** C-6 ** U235 left after Burnup 34.287 g ** Density 5.649 m306 92235.66c -0.014955 92238.66c -0.061739 40000.66c -0.906774 1001. 66c -0.016531 c ** C-8 ** U235 left after Burnup 37.690 g ** Density 5.693 m308 92235.66c -0.016440 92238.66c -0.067867 40000.66c -0.899298 1001.66c -0.016395 c ** C-9 ** U235 left after Burnup 37.012 g ** Density 5.684 m309 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-10 ** U235 left after Burn up 37.012 g **Density 5.684 m310 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-11 ** U235 left after Burn up 37.012 g **Density 5.684 m311 92235.66c -0. 016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** C-12 ** U235 left after Burnup 37.012 g **Density 5.684 m312 92235.66c -0.016144 92238.66c -0.066646 40000.66c -0.900788 1001.66c -0.016422 c ** D-1 ** U235 left after Burnup 33.326 g **Density 5.637 m401 92235.66c -0.014536 92238.66c -0.060009 40000.66c -0.908885 1001.66c -0.016570 c ** D-2 ** U235 left after Burnup 34.700 g **Density 5.655 m402 92235.66c -0.015136 92238. 66c -0.062483 40000.66c -0.905867 1001.66c -0.016515 c ** D-3 ** U235 left after Burnup 32.480 g **Density 5.626 m403 92235.66c -0.014167 92238.66c -0.058485 40000.66c -0.910744 1001.66c -0.016604 c ** D-5 ** U235 left after Burnup 33.266 g **Density 5.636 m405 92235.66c -0.014510 92238.66c -0.059901 40000.66c -0.909017 1001.66c -0.016572 c ** D-6 ** U235 left after Burnup 38.012 g ** Density 5.697 m406 92235.66c -0.016580 92238.66c -0.068447 40000.66c -0.898591 1001.66c -0.016382 c ** D-7 ** U235 left after Burnup 34.046 g ** Density 5. 646 m407 92235.66c -0.014850 92238.66c -0.061305 40000.66c -0.907303 1001.66c -0.016541 c ** D-8 ** U235 left after Burnup 33.278 g ** Density 5.636 m408 92235.66c -0.014515 92238.66c -0.059922 40000.66c -0.908991 63

ATTACHMENT 1 1001.66c -0.016572 c ** D-9 ** U235 left after Burnup 33.703 g ** Density 5.642 m409 92235.66c -0.014701 92238. 66c -0.060688 40000.66c -0.908057 1001.66c -0.016555 c ** D-10 ** U235 left after Burn up 32.929 g **Density 5.632 m410 92235.66c -0.014363 92238.66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** D-11 ** U235 left after Burnup 33.658 g **Density 5.641 m411 92235.66c -0.014363 92238. 66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** D-12 ** U235 left after Burnup 34. 711 g ** Density 5.655 m412 92235.66c -0.015140 92238. 66c -0.062503 40000.66c -0.905843 1001.66c -0.016514 c ** D-13 ** U235 left after Burnup 33.835 g ** Density 5.644 m413 92235.66c -0.014758 92238 .66c -0.060925 40000.66c -0.907767 1001.66c -0.016549 c ** D-14 ** U235 left after Burn up 35.835 g ** Density 5.669 m414 92235.66c -0.015631 92238.66c -0.064527 40000.66c -0.903373 1001. 66c -0.016469 c ** D-15 ** U235 left after Burn up 34.841 g ** Density 5.656 m415 92235.66c -0.015197 92238.66c -0.062737 40000.66c -0.905557 1001.66c -0.016509 c ** D-17 ** U235 left after Burnup 34.835 g **Density 5.656 m417 92235.66c -0.015194 92238.66c -0 .062726 40000.66c -0.905570 1001. 66c -0.016509 c ** D-18 ** U235 left after Burn up 33.822 g **Density 5.643 m418 92235.66c -0.014753 92238.66c -0.060902 40000.66c -0.907796 1001.66c -0.016550 c ** E-2 ** U235 left after Burnup 37.211 g **Density 5.687 m502 92235.66c -0.016231 92238.66c -0.067004 40000.66c -0.900351 1001. 66c -0.016414 c ** E-3 ** U235 left after Burnup 32.929 g ** Density 5.632 m503 92235.66c -0.014363 92238.66c -0.059294 40000.66c -0.909757 1001.66c -0.016586 c ** E-4 ** U235 left after Burnup 33.692 g ** Density 5.642 m504 92235.66c -0.014696 92238.66c -0.060668 40000.66c -0.908081 1001.66c -0.016555 c ** E-5 ** U235 left after Burnup 32.388 g ** Density 5.625 m505 92235.66c -0.014127 92238.66c -0.058320 40000.66c -0.910946 1001.66c -0.016607 c ** E-6 ** U235 left after Burnup 32.503 g ** Density 5.627 m506 92235.66c -0.014177 92238.66c -0.058527 40000.66c -0.910693 1001.66c -0.016603 64

ATTACHMENT 1 c ** E-7 ** U235 left after Burnup 31.991 g ** Density 5.620 m507 92235.66c -0.013954 92238 .66c -0.057605 40000.66c -0. 911818 1001.66c -0.016623 c ** E-8 ** U235 left after Burnup 32.288 g ** Density 5.624 m508 92235.66c -0.014084 92238.66c -0.058140 40000.66c -0.911165 1001.66c -0.016611 c ** E-9 ** U235 left after Burnup 29.517 g ** Density 5.588 m509 92235.66c -0.012875 92238. 66c -0.053150 40000.66c -0.917253 1001. 66c -0.016722 c ** E-10 ** U235 left after Burnup 31.776 g ** Density 5.617 m510 92235.66c -0.013860 92238.66c ~o. 051210 40000.66c -0.912290 1001.66c -0.016632 c ** E-11 ** U235 left after Burn up 32.835 g **Density 5. 631 m511 92235.66c -0.014322 92238.66c -0.059125 40000.66c -0.909964 1001. 66c -0.016590 c ** E-12 ** U235 left after Burnup 34.835 g ** Density 5.656 m512 92235.66c -0.015194 92238.66c -0. 062726 40000.66c -0.905570 1001.66c -0.016509 c ** E-13 ** U235 left after Burnup 37. 713 g ** Density 5.693 m513 92235.66c -0. 016450 92238.66c -0.067908 40000.66c -0. 899248 1001. 66c -0.016394 c ** E-14 ** U235 left after Burnup 33.254 g **Density 5.636 m514 92235.66c -0.014505 92238.66c -0.059879 40000.66c -0.909043 1001. 66c -0.016573 c ** E-15 ** U235 left after Burn up 37.636 g ** Density 5.692 m515 92235.66c -0.016416 92238.66c -0.067770 40000.66c -0.899417 1001. 66c -0.016397 c ** E-16 ** U235 left after Burnup 31.948 g ** Density 5.619 m516 92235.66c -0.013935 92238.66c -0.057527 40000.66c -0 .911912 1001. 66c -0.016625 c ** E-17 ** U235 left after Burnup 35.937 g ** Density 5.670 m517 92235.66c -0.015675 92238.66c -0.064710 40000.66c -0.903149 1001.66c -0.016465 c ** E-18 ** U235 left after Burnup 34.960 g **Density 5.658 m518 92235.66c -0.015249 92238.66c -0.062951 40000.66c -0 .905296 1001.66c -0.016504 c ** E-19 ** U235 left after Burnup 34.855 g ** Density 5.657 m519 92235.66c -0.015203 92238.66c -0.062762 40000.66c -0.905526 1001.66c -0.016509 c ** E-20 ** U235 left after Burnup 34.234 g ** Density 5.649 m520 92235.66c -0.014932 92238.66c -0.061644 40000*. 66c -0.906890 1001. 66c -0.016533 c ** E-21 ** U235 left after Burnup 33.977 g ** Density 5.645 65

ATTACHMENT 1 m521 92235.66c -0.014820 92238.66c -0.061181 40000.66c -0.907455 1001.66c -0.016544 c ** E-22 ** U235 left after Burnup 34.021 g **Density 5.646 m522 92235.66c -0.014839 92238.66c -0.061260 40000.66c -0.907358 1001.66c -0.016542 c ** E-23 ** U235 left after Burnup 33.599 g ** Density 5.641 m523 92235.66c -0.014655 92238.66c -0.060500 40000.66c -0.908285 1001.66c -0.016559 c ** E-24 ** U235 left after Burnup 33.226 g **Density 5.636 m524 92235.66c -0.014493 92238.66c -0.059829 40000.66c -0.909105 1001.66c -0.016574 c ** F-1 ** U235 left after Burnup 31.327 g ** Density 5.612 m601 92235.66c -0.013664 92238.66c -0.056409 40000.66c -0.913277 1001. 66c -0.016650 c ** F-2 ** U235 left after Burnup 31.467 g ** Density 5.613 m602 92235.66c -0.013725 92238.66c -0.056661 40000.66c -0.912969 1001.66c -0.016644 c ** F-3 ** U235 left after Burnup 32.486 g ** Density 5.626 m603 92235.66c -0.014170 92238.66c -0.058496 40000.66c -0.910730 1001. 66c -0.016603 c ** F-4 ** U235 left after Burnup 31. 438 g ** Density 5.613 m604 92235.66c -0.013713 92238. 66c -0.056609 40000.66c -0.913033 1001.66c -0.016645 c ** F-5 ** U235 left after Burnup 31.456 g ** Density 5.613 m605 92235.66c -0.013721 92238.66c -0.056641 40000.66c -0.912993 1001.66c -0.016645 c ** F-6 ** U235 left after Burnup 38.012 g **Density 5.697 m606 92235.66c -0.016580 92238.66c -0.068447 40000.66c -0.898591 1001.66c -0.016382 c ** F-7 ** U235 left after Burnup 30.816 g ** Density 5.605 m607 92235.66c -0.013441 92238.66c -0.055489 40000.66c -0.914399 1001.66c -0.016670 c ** F-8 ** U235 left after Burnup 31. 404 g ** Density 5.613 m608 92235.66c -0.013698 92238.66c -0.056548 40000.66c -0.913107 1001. 66c -0.016647 c ** F-9 ** U235 left after Burnup 31. 418 g ** Density 5.613 m609 92235.66c -0.013704 92238.66c -0.056573 40000.66c -0.913077 1001. 66c -0.016646 c ** F-11 ** U235 left after Burnup 30.355 g **Density 5.599 m611 92235.66c -0. 013240 92238.66c -0.054659 40000.66c -0.915412 1001.66c -0.016689 c ** F-12 ** U235 left after Burn up 38.461 g **Density 5.703 m612 92235.66c -0.016776 66

ATTACHMENT 1 92238.66c -0.069255 40000.66c -0.897605 1001.66c -0.016364 c ** F-13 ** U235 left after Burnup 32.407 g **Density 5.625 m613 92235.66c -0.014135 92238.66c -0.058354 40000.66c -0.910904 1001.66c -0.016607 c ** F-14 ** U235 left after Burnup 30.445 g **Density 5.600 m614 92235.66c -0.013280 92238.66c -0.054821 40000.66c -0.915214 1001.66c -0.016685 c ** F-15 ** U235 left after Burnup 30.903 g ** Density 5.606 m615 92235.66c -0.013479 92238.66c -0.055646 40000.66c -0.914208 1001. 66c -0.016667 c ** F-16 ** U235 left after Burnup 30.326 g ** Density 5.599 m616 92235. 66c -0.013228 92238.66c -0.054607 40000.66c -0.915476 1001. 66c -0.016690 c ** F-17 ** U235 left after Burnup 30.081 g ** Density 5 .596 m617 92235. 66c -0.013121 92238.66c -0.054166 40000.66c -0.916014 1001.66c -0.016700 c ** F-18 ** U235 left after Burn up 30.344 g ** Density 5.599 m618 92235.66c -0.013236 92238.66c -0.054639 40000.66c -0.915436 1001.66c -0.016689 c ** F-19 ** U235 left after Burnup 33.099 g ** Density 5.634 m619 92235.66c -0.014437 92238.66c -0.059600 40000.66c -0.909384 1001.66c -0.016579 c ** F-20 ** U235 left after Burn up 37.540 g **Density 5.691 m620 92235.66c -0.016374 92238.66c -0.067597 40000.66c -0.899628 1001.66c -0.016401 c ** F-21 ** U235 left after Burnup 31.381 g ** Density 5.612 m621 92235.66c -0.013688 92238.66c -0.056506 40000.66c -0.913158 1001.66c -0.016648 c ** F-22 ** U235 left after Burnup 35.890 g ** Density 5.670 m622 92235.66c -0.015655 92238.66c -0.064626 40000.66c -0.903253 1001.66c -0.016467 c ** F-23 ** U235 left after Burn up 30.368 g ** Density 5.599 m623 92235.66c -0.015655 92238.66c -0.064626 40000.66c -0.903253 1001.66c -0.016467 c ** F-24 ** U235 left after Burn up 30.079 g ** Density 5.596 m624 92235.66c -0.013120 92238.66c -0.054162 40000.66c -0.916018 1001.66c -0.016700 c ** F-25 ** U235 left after Burnup 30.667 g **Density 5.603 m625 92235.66c -0.013376 92238.66c -0.055221 40000.66c -0.914726 1001. 66c -0.016676 c ** F-26 ** U235 left after Burn up 38.339 g **Density 5.701 m626 92235.66c -0.016723 92238.66c -0.069035 67

ATTACHMENT 1 40000.66c -0.897873 1001.66c -0.016369 c ** F-27 ** U235 left after Burn up 31.103 g ** Density 5.609 m627 92235.66c -0.013567 92238.66c -0.056006 40000.66c -0. 913769 1001. 66c -0.016659 c ** F-28 ** U235 left after Burnup 33.415 g **Density 5.638 m628 92235.66c -0.014575 92238.66c -0.060169 40000.66c -0.908690 1001.66c -0.016566 c ** F-29 ** U235 left after Burn up 36.881 g **Density 5.682 m629 92235.66c -0.016087 92238.66c -0.066410 40000.66c -0.901076 1001. 66c -0.016427 c ** F-30 ** U235 left after Burnup 36.760 g ** Density 5.681 m630 92235.66c -0.016034 92238.66c -0.066192 40000.66c -0.901341 1001.66c -0.016432 c ** 12.3 wt % New Fuel Elements ** U235 53.3 g **Density 5.892 m700 92235.66c -0.023249 92238.66c -0.095975 40000.66c -0.867862 l001.66c -0.015822 c ** 12 wt % New Fuel Elements ** U235 = 52 g ** Density 5.875 c m700 92235.66c -0.022682 c 92238.66c -0.093634 c 40000.66c -0.867862 c 1001.66c -0.015822 c ******************************************************************************

c graphite rho= 1.6 g/cmA3 c S(a,b)- Carbon@ 300K m2 6000.66c -1 $MAT c Air rho= 0.001239 $ RSR AIR m3 6000.66c -0.000124 7014.66c -0.755268 8016.66c -0.231781 18000 -0.012827 c Light water rho=l.00 g/cmA3 c S(a,b)- H20@ 300K m6 1001.66c 0.66667 $Water 8016.66c 0.33333 c Aluminum rho 2.699 g/cmA3 m7 13027.66c -1 $ Aluminum, 2.699 g/cc c Zirconium $ Zirc Filler, 6.5 g/cc c S(a,b)- H20@ 300K ms 40000.66c -1 c Stainless Steel rho 7.9 g/cmA3 c m9 26000.55c -1.0 m9 26000.55c -0.67 $MAT 24000.50c -0.17 28000.50c -0.08849997 42000.66c -0.004452 c B4C SHIM & REG rod rho> 2.48 g/cmA3, email from GA, c NEED TO ADJUST FOR BURNUP TO DIFFERENTIATE BETWEEN NEW ROD??

c rho 1.60-1.75 g/cmA3 from B. Ma c 16% Boron from UIC TRIGA SAR mll 6000.66c -0.84 $ Pulse Rod, 1.6 g/cc, 5010.66c -0.0296 5011.66c -0.1304 c B4C rho > 2.48 g/cc, SHIM & REG, BUT THESE ARE OLD & PART BURNED ml2 6000.66c -0.22 $ Shim Rod, 2.5 g/cc 5010.66c -0.1443 5011.66c -0.6357 c B4C rho >2.48 g/cc, SAFETY ROD ml3 6000.66c -0.22 $Shim Rod, 2.5 g/cc 5010.66c -0.1443 5011.66c -0.6357 c ******************************************************************************

imp:n 0 1 1165r $ 1, 63012 c ******************************************************************************

68

ATTACHMENT 1 c SOURCE DISTRIBUTED ACROSS THE CORE VOLUME sdef ERG=Dl POS=O 0 -29 AXS=O 0 1 RAD=D2 EXT=D3 spl -3 si2 o 22.8 si3 O 45.7 c ******************************************************************************

mt201 h/zr.Olt zr/h.Olt $ uzrh-full salphabeta card mt202 h/zr.Olt zr/h.Olt mt203 h/zr.Olt zr/h.Olt mt204 h/zr.Olt zr/h.Olt mt205 h/zr. Olt zr/h.Olt mt206 h/zr.Olt zr/h.Olt mt301 h/zr.Olt zr/h.Olt mt302 h/zr.Olt zr/h.Olt mt303 h/zr.Olt zr/h.Olt mt304 h/zr.Olt zr/h.Olt mt305 h/zr.Olt zr/h.Olt mt306 h/zr. Olt zr/h.Olt mt308 h/zr.Olt zr/h.Olt mt309 h/zr.Olt zr/h. Olt mt310 h/zr.Olt zr/h.Olt mt311 h/zr.Olt zr/h.Olt mt312 h/zr.Olt zr/h.Olt mt401 h/zr.Olt zr/h.Olt mt402 h/zr. Olt zr/h.Olt mt403 h/zr.Olt zr/h.Olt mt405 h/zr.Olt zr/h.Olt mt406 h/zr.Olt zr/h.Olt mt407 h/zr.Olt zr/h.Olt mt408 h/zr.Olt zr/h.Olt mt409 h/zr.Olt zr/h.Olt mt410 h/zr.Olt zr/h.Olt mt411 h/zr.Olt zr/h.Olt mt412 h/zr.Olt zr/h.Olt mt413 h/zr.Olt zr/h.Olt mt414 h/zr.Olt zr/h.Olt mt415 h/zr.Olt zr/h.Olt mt417 h/zr.Olt zr/h.Olt mt418 h/zr.Olt zr/h.Olt mt502 h/zr.Olt zr/h.Olt mt503 h/zr.Olt zr/h.Olt mt504 h/zr.Olt zr/h.Olt mt505 h/zr.Olt zr/h.Olt mt506 h/zr.Olt zr/h.Olt mt507 h/zr.Olt zr/h.Olt mt508 h/zr.Olt zr/h.Olt mt509 h/zr.Olt zr/h.Olt mt510 h/zr. Olt zr/h.Olt mt511 h/zr.Olt zr/h.Olt mt512 h/zr.Olt zr/h.Olt mt513 h/zr.Olt zr/h.Olt mt514 h/zr.Olt zr/h.Olt mt515 h/zr.Olt zr/h.Olt mt516 h/zr.Olt zr/h.Olt mt517 h/zr.Olt zr/h.Olt mt518 h/zr.Olt zr/h.Olt mt519 h/zr.Olt zr/h.Olt mt520 h/zr.Olt zr/h. Olt mt521 h/zr.Olt zr/h.Olt mt522 h/zr.Olt zr/h.Olt mt523 h/zr.Olt zr/h.Olt mt524 h/zr.Olt zr/h.Olt mt601 h/zr.Olt zr/h.Olt mt602 h/zr.Olt zr/h.Olt mt603 h/zr.Olt zr/h.Olt mt604 h/zr.Olt zr/h.Olt mt605 h/zr.Olt zr/h.Olt mt606 h/zr.Olt zr/h.Olt mt607 h/zr.Olt zr/h.Olt mt608 h/zr.Olt zr/h. Olt mt609 h/zr.Olt zr/h.Olt 69

ATTACHMENT 1 mt611 h/zr.Olt zr/h.Olt mt612 h/zr.Olt zr/h.Olt mt613 h/zr. Olt zr/h.Olt mt614 h/zr.Olt zr/h.Olt mt615 h/zr.Olt zr/h.Olt mt616 h/zr.Olt zr/h.Olt mt617 h/zr.Olt zr/h.Olt mt618 h/zr.Olt zr/h.Olt mt619 h/zr.Olt zr/h.Olt mt620 h/zr.Olt zr/h.Olt mt621 h/zr.Olt zr/h.Olt mt622 h/zr.Olt zr/h.Olt mt623 h/zr.Olt zr/h.Olt mt624 h/zr.Olt zr/h.Olt mt625 h/zr.Olt zr/h.Olt mt626 h/zr.Olt zr/h.Olt mt627 h/zr. Olt zr/h.Olt mt628 h/zr.Olt zr/h.Olt mt629 h/zr.Olt zr/h.Olt mt630 h/zr.Olt zr/h.Olt mt700 h/zr.Olt zr/h. Olt mt2 grph.OlT mt6 lwtr.01 c ************************* TALLY SPECIFICATION ********************************

c Flux average tally for active fuel region of all 85 elements c FUEL ELEMENT ORDER IMPORTANT f4 :n 20106 20206 20306 20406 20506 20606 &

30106 30206 30306 30406 30506 30606 30806 30906 31006 31106 31206 &

40106 40206 40306 40506 40606 40706 40806 40906 41006 41106 41206 &

41306 41406 41506 41706 41806 &

50206 50306 50406 50506 50606 50706 50806 50906 51006 51106 51206* &

51306 51406 51506 51606 51706 51806 51906 52006 52106 52206 52306 &

52406 &

60106 60206 60306 60406 60506 60606 60706 60806 60906 61106 61206 &

61306 61406 61506 61606 61706 61806 61906 62006 62106 62206 62306 &

62406 62506 62606 62706 62806 62906 63006 f7:n 20106 20206 20306 20406 20506 20606 &

30106 30206 30306 30406 30506 30606 30806 30906 31006 31106 31206 &

40106 40206 40306 40506 40606 40706 40806 40906 41006 41106 41206 &

41306 41406 41506 41706 41806 &

50206 50306 50406 50506 50606 50706 50806 50906 51006 51106 51206 &

51306 51406 51506 51606 51706 51806 51906 52006 52106 52206 52306 &

52406 &

60106 60206 60306 60406 60506 60606 60706 60806 60906 61106 61206 &

61306 61406 61506 61606 61706 61806 61906 62006 62106 62206 62306 &

62406 62506 62606 62706 62806 62906 63006 70

ATTACHMENT 1 71

ATTACHMENi-:,.a-" ~

Proposed Wording of Facility License UNITED STATES NUCLEAR REGULATORY COMMISSION RENEWAL OF FACILITY LICENSE NO. R-88 DOCKET NO. 50-188 KANSAS STATE UNIVERSITY NUCLEAR REACTOR FACILITY A. This license applies to the TRIGA research reactor (herein "the reactor"), owned by the Kansas State University and located on its campus in Manhattan, Kansas, and is described in the licensee's application for renewal dated September 12, 2002, as supplemented.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Kansas State University:

1. Pursuant Section 104c of the Act and 10 CPR Part 5 0, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the reactor as a utilization facility at the designated location in accordance with the procedures and limitation described in the application and in this license;

2. Pursuant to the Act and 10CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use up to 4.55 kilograms of contained uranium-235 at enrichments less than 20% in connection with the operation of the reactor and up to 90 grams of uranium-235 at any enrichment for fission chambers and reactor experiments.

3. Pursuant to the Act and 10 CPR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material, and Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use a 2-curie sealed americium-beryllium neutron source and a 7-curie sealed polonium-berylliumneutron source (Monsanto Item NS-1, Serial No. N-693) for reactor startup; and to possess, use, but not separate, such byproduct material as may be produced by the operation of the reactor.

C. This license shall be deemed to contain and be subject to the conditions specified in Parts 20, 30, 40, 50, 51, 55, 70, and 73of10 CPR Chapter I, to all applicable provisions of the Act, and to the rules, regulations and orders of the Commission now, or hereafter in effect, and is subject to the additional conditions specified or incorporated below:

72

ATTACHMENT)" a.,

1. Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels up to a maximum of 1,250 kW (thermal) and in pulse mode with reactivity insertions not to exceed $3 .00 with all stainless-steel clad fuel elements.

73

ATTACHMENT J' 2-Proposed Wording of Facility License (Mark-Up Copy)

UNITED STATES NUCLEARREGill..ATORY C01vf11ISSION RENE\:llAL OF FACILITY LICENSE NO. R-SS DOCKET NO. 50-188 KA..:"'\l'S.AS STATE UNIVERSITY NUCLEAR REACTOR FACILITY A. TI1is license applies to the TRIGA research reactor (herein <<the reactor'), owned by the Kansas St.1.te University and located on its campus in Manli.1tta11, Kansas, and is described in the licensee's application for renewal dated September 12, 2002, as supplemented.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Kam;as State University:

L Pursuant Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,>' to possess, use, and operate the reactor as a utilization facility at the designated location in accordance >vith the procedures and limitation described in the application and :in Uris Hcen.se;

2. PursU<.mt to the Act and lOG'FR Part 70, "Domestic Licensing of Special Nuclear lvfuterial," to receive. possess, and use up to 1.204.55 kilograms of contained uranium-235 at enrichments less than 20% U.1 connection i.vifh the operation of the reactor and up to 90 grams of uranium-235 at any enrichment for fission d1ambers and reactor experiments.

3. Pursuant to the Act and 10 CFR Part 30, ..Rules of General Applicability to D0111estic Licensing of Byproduct Materh11,. and Part 70, ""Domestic Liceni;ing of Special Nl.JClear Material," to receive, possess, and use a 2-curie sealed americium-beryllium neutron source and a 7-mrie sealed polonium-beryllium. neutron source (lvfonsanto Item NS-1, Serial No. N-693) for reactor startup~ and to possess, use, but not separate, such byproduct material as may be produced by the operation of the reactor.

C. This license shall be deemed to contain and be subject to the conditions specified in Parts 20, 30, 40, 50, 51, 55, 70, and 73 of 10 CFR Chapter I, to all applicable provisions of the Act, and to the rnles, regulations and orders of the Conunission 74

ATTACHMENT7Z.*-

now, or hereafter in effect, and is subject to the addition.'11 conditions specified or

incorporated belo\v:

1. 1\lfuximum Power Level The licensee is authorized to operate the facility at steady-state power levels up to a ma.'rirnum of 1,2:50 kW (thennal) and in pulse mode \Vith re-activity :insertions not to exceed $3. 00 \Vith all stainless-steel clad fuel elements.

75

ATTACHMENT 3 Updated Technical Specifications A clean copy of the proposed technical specifications, as well as a mark-up copy showing changes, follows this page. Note that the only page with changes is TS-38, Section 5.1, Reactor Fuel. (The numbering of some subsequent pages has also changed).

76

TECHNICAL SPECIFICATIONS Table of Contents

1. D EFINITION S ................................................................................................................. TS-1

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Elem ent Tem perature Safety Lim it ..................................................................... TS-8 2.1.1. A pplicability ...................................................................................................... TS-8 2.1.2. Objective ............................................................................................................. TS-8 2.1.3. Specification ...................................................................................................... TS-8 2.1.4. Actions ................................................................................................................ TS-8 2.1.5. Basis ................................................................................................................... TS-8 2.2 Lim iting Safety System Settings ............................................................................... TS-10 2.2.1. A pplicability ..................................................................................................... TS-10 2.2.3. Objective ........................................................................................................... TS-10 2.2.4. Specification ...................................................................................................... TS-10 2.2.5. Actions .............................................................................................................. TS-10 2.2.6. Basis ................................................................................................................. TS-10

3. LIMITING CONDITIONS FOR OPERATIONS ........................................................ TS-1 1 3.1 CO RE REA CTIV ITY ................................................................................................ TS-1 1 3.1.1. Applicability ..................................................................................................... TS-1 1 3.1.3. Objective ........................................................................................................... TS-1 1 3.1.4. Specification ...................................................................................................... TS-1 1 3.1.5. Actions .............................................................................................................. TS-12 3.1.6. Basis ................................................................................................................. TS-13 3.2 PU LSED M O D E O PERA TION S.............................................................................. TS-13 3.2.1. A pplicability ..................................................................................................... TS-13 3.2.3. Objective ........................................................................................................... TS-13 3.2.4. Specification ...................................................................................................... TS-13 3.2.5. Actions .............................................................................................................. TS-13 3.2.6. Basis ................................................................................................................. TS-13 3.3 M EA SU RIN G CH ANN ELS ..................................................................................... TS-14 3.3.1. Applicability ..................................................................................................... TS-14 3.3.3. Objective ........................................................................................................... TS-14 3.3.4. Specification ...................................................................................................... TS- 14 3.3.5. Actions .............................................................................................................. TS-14 3.3.6. Bases ................................................................................................................. TS-16 3.4. SAFETY CHANNEL AND CONTROL ROD OPERABILITY .............................. TS-18 3.4.1. A pplicability ..................................................................................................... TS-18 3.4.3. Objective ........................................................................................................... TS-18 3.4.4. Specification ...................................................................................................... TS-18 3.4.5. Actions .............................................................................................................. TS-18 3.4.6. Basis ................................................................................................................. TS-19 3.5 G A SEO US EFFLU EN T CON TRO L ........................................................................ TS-20 3.5.1. A pplicability ..................................................................................................... TS-20 3.5.3. Objective ........................................................................................................... TS-20 3.5.4. Specification ...................................................................................................... TS-20 3.5.5. Actions .............................................................................................................. TS-20 3.5.6. Basis ................................................... TS-21 3.6 LIM ITA TIO N S ON EXPERIM ENTS .......................................................................... TS-22 3.6.1. A pplicability ..................................................................................................... TS-22 3.6.3. Objective ........................................................................................................... TS-22 K-State Reactor TS-1 Original (4/14)

TECHNICAL SPECIFICATIONS 3.6.4. Specification ...................................................................................................... TS-22 3.6.5. Actions .............................................................................................................. TS-22 3.6.6. Basis ................................................................................................................. TS-23 3.7 FUEL INTEG RITY .................................................................................................. TS-24 3.7.1. Applicability ..................................................................................................... TS-24 3.7.3. Objective ........................................................................................................... TS-24 3.7.4. Specification ...................................................................................................... TS-24 3.7.5. Actions .............................................................................................................. TS-24 3.7.6. Basis ................................................................................................................. TS-24 3.8 REACTOR POOL W ATER ......................................................................................... TS-25 3.8.1. Applicability ..................................................................................................... TS-25 3.8.3. Objective ........................................................................................................... TS-25 3.8.4. Specification ..................................................................................................... TS-25 3.8.5. Actions .............................................................................................................. TS-25 3.8.6. Basis ................................................................................................................. TS-26 3.9 M aintenance Retest Requirements ................................................................................ TS-27 3.9.1. Applicability ..................................................................................................... TS-27 3.9.3. Objective ........................................................................................................... TS-27 3.9.4. Specification ...................................................................................................... TS-27 3.9.5. Actions .............................................................................................................. TS-27 3.9.6. Basis ................................................................................................................. TS-27

4. SURVIELLANCES ......................................................................................................... TS-28 4.1 CORE REACTIV ITY ................................................................................................ TS-28 4.1.1. Objective ........................................................................................................... TS-28 4.1.2. Specification ...................................................................................................... TS-28 4.1.3. Basis ................................................................................................................. TS-28 4.2 PULSE M ODE ............................................................................................................. TS-29 4.2.1. Objective ........................................................................................................... TS-29 4.2.2. Specification ...................................................................................................... TS-29 4.2.3. Basis ................................................................................................................. TS-29 4.3 M EASURING CHANN ELS ..................................................................................... TS-30 4.3.1. Objective ........................................................................................................... TS-30 4.3.2. Specification ...................................................................................................... TS-30 4.3.3. Basis ................................................................................................................. TS-30 4.4 SAFETY CHANNEL AND CONTROL ROD OPERABILITY .............................. TS-31 4.4.1. Objective ........................................................................................................... TS-31 4.4.2. Specification ...................................................................................................... TS-31 4.4.3. Basis ................................................................................................................. TS-32 4.5 GA SEOUS EFFLUENT CONTROL ........................................................................ TS-33 4.5.1. Objective ........................................................................................................... TS-33 4.5.2. Specification ...................................................................................................... TS-33 4.5.3. Basis ................................................................................................................. TS-33 4.6 LIM ITATION S ON EXPERIM ENTS ....................................................................... TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis ................................................................................................................. TS-34 4.7 FU EL INTEGRITY ................................................................................................... TS-35 4.7.1. Objective ........................................................................................................... TS-35 4.7.2. Specification ...................................................................................................... TS-35 4.7.3. Basis ................................................................................................................. TS-35 4.8 REACTOR POOL W ATER ...................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor TS-2 Original (4/14)

TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36 4.8.3. Basis ................................................................................................................. TS-36 4.9 M A INTENANCE RETEST REQUIREM ENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................................................................... TS-37 4.10.3. Basis ............................................................................................................... TS-37

5. DESIG N FEATURES ...................................................................................................... TS-38 5.1 REACTOR FUEL ...................................................................................................... TS-38 5.1.1. Applicability ..................................................................................................... TS-38 5.1.2. Objective ........................................................................................................... TS-38 5.1.3. Specification ...................................................................................................... TS-38 5.1.4. Basis ................................................................................................................. TS-38 5.2 REACTOR FUEL AND FUELED DEVICES IN STORAGE ................................. TS-38 5.2.1. Applicability ..................................................................................................... TS-38 5.2.2. Objective ........................................................................................................... TS-39 5.2.3. Specification ...................................................................................................... TS-39 5.2.4. Basis ................................................................................................................. TS-39 5.3 REACTOR BU ILD IN G ............................................................................................ TS-39 5.3.1. Applicability ..................................................................................................... TS-39 5.3.2. Objective ........................................................................................................... TS-39 5.3.3. Specification ...................................................................................................... TS-39 5.3.4. Basis ................................................................................................................. TS-40 5.4 EXPERIM ENTS ........................................................................................................ TS-40 5.4. 1. Applicability ..................................................................................................... TS-40 5.4.2. Objective ........................................................................................................... TS-40 5.4.3. Specification ...................................................................................................... TS-40 5.4.4. Basis ................................................................................................................. TS-41

6. ADM INISTR ATIV E CONTRO LS ................................................................................ TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL ......................... TS-44 6.2 REV IEW AN D AU DIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS ................................................. TS-47 6.5 EM ERGENCY PLAN AN D PROCEDURES ........................................................... TS-48 6.6 OPERATO R REQ UALIFICATION ......................................................................... TS-48 6.7 PHYSICAL SECU RITY PLAN ................................................................................ TS-48 6.8 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS VIOLATED .... TS-48 6.9 ACTION TO BE TAKEN IN THE EVENT OF A REPO RTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERA TING RECORDS ............................................................................ TS-49 6.11 REPORTIN G REQU IREM ENTS ........................................................... TS-50 K-State Reactor TS-3 Original (4/14)

TECHNICAL SPECIFICATIONS

1. DEFINITIONS The following frequently used terms are defined to aid in the uniform interpretation of these specifications. Capitalization is used in the body of the Technical Specifications to identify defined terms.

ACTION Actions are steps to be accomplished in the event a required condition identified in a "Specification" section is not met, as stated in the "Condition" column of "Actions."

In using Action Statements, the following guidance applies:

• Where multiple conditions exist in an LCO, actions are linked to the (failure to meet a "Specification") "Condition" by letters and number.

• Where multiple action steps are required to address a condition, COMPLETION TIME for each action is linked to the action by letter and number.

• AND in an Action Statement means all steps need to be performed to complete the action; OR indicates options and alternatives, only one of which needs to be performed to complete the action.

• If a "Condition" exists, the "Action" consists of completing all steps associated with the selected option (if applicable) except where the "Condition" is corrected prior to completion of the steps ANNUAL 12 months, not to exceed 15 months CHANNEL A channel calibration is an adjustment of the channel to that its output CALIBRATION responds, with acceptable range and accuracy, to known values of the parameter that the channel measures.

BIENNIAL Every two years, not to exceed a 28 month interval CHANNEL A channel check is a qualitative verification of acceptable performance by CHECK observation of channel behavior. This verification shall include comparison of the channel with expected values, other independent channels, or other methods of measuring the same variable.

CHANNEL TEST A channel test is the introduction of an input signal into a channel to verify that it is operable. A functional test of operability is a channel test.

CONTROL ROD A standard control rod is one having an electric motor drive and scram (STANDARD) capability.

CONTROL ROD A transient rod is one that is pneumatically operated and has scram (TRANSIENT) capability.

DAILY Prior to initial operation each day (when the reactor is operated), or before TS-4 Original (4/14)

K-State Reactor K-State Reactor TS-4 Original (4/14)

TECHNICAL SPECIFICATIONS an operation extending more than 1 day ENSURE Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition EXHAUST The air volume in the reactor bay atmosphere between the pool surface and PLENUM the reactor bay exhaust fan EXPERIMENT An EXPERIMENT is (1) any apparatus, device, or material placed in the reactor core region (in an EXPERIMENTAL FACILITY associated with the reactor, or in line with a beam of radiation emanating from the reactor) or (2) any in-core operation designed to measure reactor characteristics.

EXPERIMENTAL Experimental facilities are the beamports, thermal column, pneumatic FACILITY transfer system, central thimble, rotary specimen rack, and the in-core facilities (including non-contiguous single-element positions, and, in the E and F rings, as many as three contiguous fuel-element positions).

IMMEDIATE Without delay, and not exceeding one hour.

NOTE:

IMMEDIATE permits activities to restorerequiredconditionsfor up to one hour: this does not permit or implv deferring or postponingaction INDEPENDENT INDEPENDENT Experiments are those not connected by a mechanical, EXPERIMENT chemical, or electrical link to another experiment LIMITING CONDITION FOR The lowest functional capability or performance levels of equipment OPERATION required for safe operation of the facility.

(LCO)

LIMITING Settings for automatic protective devices related to those variables having SAFETY SYSTEM significant safety functions. Where a limiting safety system setting is SETTING (LSSS) specified for a variable on which a safety limit placed, the setting shall be chosen so that the automatic protective action will correct the abnormal situation before a safety limit is exceeded.

MEASURED The measured value of a parameter is the value as it appears at the output VALUE of a MEASURING CHANNEL.

MEASURING A MEASURING CHANNEL is the combination of sensor, lines, CHANNEL amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.

MOVABLE A MOVABLE EXPERIMENT is one that may be moved into, out-of or EXPERIMENT near the reactor while the reactor is OPERATING.

NONSECURED NONSECURED Experiments are these that should not move while the EXPERIMENT reactor is OPERATING, but are held in place with less restraint than a secured experiment.

TS-5 Original (4/14)

K-State Reactor K-State Reactor TS-5 Original (4/14)

TECHNICAL SPECIFICATIONS OPERABLE A system or component is OPERABLE when it is capable of performing its intended function in a normal manner OPERATING A system or component is OPERATING when it is performing its intended function in a normal manner.

PULSE MODE The reactor is in the PULSE MODE when the reactor mode selection switch is in the pulse position and the key switch is in the "on" position.

NOTE:

In the PULSE MODE, reactorpower may be increasedon a period of much less than 1 second by motion of the transientcontrol rod.

REACTOR The REACTOR SAFETY SYSTEM is that combination of MEASURING SAFETY SYSTEM C1HANNELS and associated circuitry that is designed to initiate reactor scram or that provides information that requires manual protective action to be initiated.

REACTOR The reactor is secured when the conditions of either item (1) or item (2) are SECURED MODE satisfied:

(1) There is insufficient moderator or insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderation and reflection (2) All of the following:

a. The console key is it the OFF position and the key is removed from the lock

b. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives (unless the drive is physically decoupled from the control rod)

c. No experiments are being moved or serviced that have, on movement, a reactivity worth greater than $1.00 REACTOR The reactor is shutdown if it is subcritical by at least $1.00 in the SHUTDOWN REFERENCE CORE CONDITION with the reactivity worth of all experiments included.

RING A ring is one of the five concentric bands of fuel elements surrounding the central opening (thimble) of the core. The letters B through F, with the letter B used to designate the innermost ring, REFERENCE The condition of the core when it is at ambient temperature (cold) and the CORE reactivity worth of xenon is negligible (<$0.30)

CONDITION SAFETY A safety channel is a MEASURING CHANNEL in the REACTOR CHANNEL SAFETY SYSTEM SECURED A secured EXPERIMENT is an EXPERIMENT held firmly in place by a EXPERIMENT mechanical device or by gravity providing that the weight of the EXPERIMENT is such that it cannot be moved by force of less than 60 lb.

TS-6 Original (4/14)

K-State Reactor K-State Reactor TS-6 Original (4/14)

TECHNICAL SPECIFICATIONS SECURED EXPERIMENT A secured EXPERIMENT with movable parts is one that contains parts WITH MOVABLE that are intended to be moved while the reactor is OPERATING.

PARTS SHALL Indicates specified action is required/(not to be performed)

(SHALL NOT)

SEMIANNUAL Every six months, with intervals not greater than 8 months SHUTDOWN The shutdown margin is the minimum shutdown reactivity necessary to MARGIN provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating condition, and that the reactor will remain subcritical without further operator action STANDARD THERMOCOUPLE A standard thermocouple fuel element is stainless steel clad fuel element FUEL ELEMENT containing three sheathed thermocouples embedded in the fuel element.

STEADY-STATE The reactor is in the steady-state mode when the reactor mode selector MODE switch is in either the manual or automatic position and the key switch is in the "on" position.

TECHNICAL A violation of a Safety Limit occurs when the Safety Limit value is SPECIFICATION exceeded.

VIOLATION A violation of a Limiting Safety System Setting or Limiting Condition for Operation) occurs when a "Condition" exists which does not meet a "Specification" and the corresponding "Action" has not been met within the required "Completion Time."

If the "Action" statement of an LSSS or LCO is completed or the "Specification" is restored within the prescribed "Completion Time," a violation has not occurred.

NOTE

"'Condition," "Specification," "Action, "and "Completion Time" refer to applicabletitles of sections in individual Technical Specifications K-State Reactor TS-7 Original (4/14)

TECHNICAL SPECIFICATIONS

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Fuel Element Temperature Safety Limit 2.1.1 Applicability This specification applies when the reactor in STEADY STATE MODE and the PULSE MODE.

2.1.2 Objective This SAFETY LIMIT ensures fuel element cladding integrity 2.1.3 Specification (1) Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed 1150'C.

(2) Steady state fuel temperature shall not exceed 750'C.

2.1.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A Stainless steel clad, high- A.] Establish SHUTDOWN A.] IMMEDIATE hydride fuel element condition temperature exceeds 1150 0 C.

OR AND Fuel temperature exceeds 750 0 C in steady state A.2 Report per Section 6.8 A.2 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> conditions 2.1.5 Bases Safety Analysis Report, Section 3.5.1 (Fuel System) identifies design and operating constraints for TRIGA fuel that will ensure cladding integrity is not challenged.

NUREG 1282 identifies the safety limit for the high-hydride (ZrH,. 7) fuel elements with stainless steel cladding based on the stress in the cladding (resulting from the hydrogen pressure from the dissociation of the zirconium hydride). This stress will remain below the yield strength of the stainless steel cladding with fuel temperatures below 1,1501C. A change in yield strength occurs for stainless steel cladding temperatures of 5000 C, but there is no scenario for fuel cladding to achieve 500 0 C while submerged; consequently the safety limit during reactor operations is 1,150 0 C.

TS-8 Original (4/14)

K-State Reactor K-State Reactor TS-8 Original (4/14)

TECHNICAL SPECIFICATIONS Therefore, the important process variable for a TRIGA reactor is the fuel element temperature.

This parameter is well suited as a single specification, and it is readily measured. During operation, fission product gases and dissociation of the hydrogen and zirconium builds up gas inventory in internal components and spaces of the fuel elements. Fuel temperature acting on these gases controls fuel element internal pressure. Limiting the maximum temperature prevents excessive internal pressures that could be generated by heating these gases.

Fuel growth and deformation can occur during normal operations, as described in General Atomics technical report E-1 17-833. Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients. Operating with maximum long-term, steady state fuel temperature of 750'C does not have significant time- and temperature-dependent fuel growth.

TS-9 Original (4/14)

Reactor K-State Reactor TS-9 Original (4/14)

TECHNICAL SPECIFICATIONS 2.2 Limiting Safety System Settings (LSSS) 2.2.1 Applicability This specification applies when the reactor in STEADY STATE MODE 2.2.2 Objective The objective of this specification is to ensure the safety limit is not exceeded.

2.2.3 Specifications (1) Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME than A. 1 IMMEDIATE A. I Reduce power to less 1,250 kW (th)

A. Steady state power level OR exceeds 1,250 kW (th)

A.2. Establish REACTOR SHUTDOWN condition 2.2.5 Bases Analysis in Chapter 4 demonstrates that if operating thermal (th) power is 1,250 kW, the maximum steady state fuel temperature is less than the safety limit for steady state operations by a large margin. For normal pool temperature, calculations in Chapter 4 demonstrate that the heat flux of the hottest area of the fuel rod generating the highest power level in the core during operations is less than the critical heat flux by a large margin up to the maximum permitted cooling temperatures; margin remains even at temperatures approaching bulk boiling for atmospheric conditions. Therefore, steady state operations at a maximum of 1,250 kW meet requirements for safe operation with respect to maximum fuel temperature and thermal hydraulics by a wide margin. Steady state operation of 1,250 kW was assumed in analyzing the loss of cooling and maximum hypothetical accidents. The analysis assumptions are protected by assuring that the maximum steady state operating power level is 1,250 kW.

In 1968 the reactor was licensed to operate at 250 kW with a minimum reactor safety system scram set point required by Technical Specifications at 110% of rated full power, with the scram set point set conservatively at 104%. In 1993 the original TRIGA power level channels were replaced with more reliable, solid state instrumentation. The actual safety system setting will be chosen to ensure that a scram will occur at a level that does not exceed 1,250 kW.

K-State Reactor TS-10 Original (4/14)

TECHNICAL SPECIFICATIONS

3. Limiting Conditions for Operation (LCO) 3.1 Core Reactivity 3.1.1 Applicability These specifications are required prior to entering STEADY STATE MODE or PULSING MODE in OPERATING conditions; reactivity limits on experiments are specified in Section 3.8.

3.1.2 Objective This LCO ensures the reactivity control system is OPERABLE, and that an accidental or inadvertent pulse does not result in exceeding the safety limit.

3.1.3 Specification The maximum available core reactivity (excess reactivity) with all control rods fully withdrawn is less than $4.00 when:

(1) 1. REFERENCE CORE CONDITIONS exists

2. No experiments with net negative reactivity worth are in place The reactor is capable of being made subcritical by a SHUTDOWN MARGIN more than

$0.50 under REFERENCE CORE CONDITIONS and under the following conditions:

(2) 1. The highest worth control rod is fully withdrawn

2. The highest worth NONSECURED EXPERIMENT is in its most positive reactive state, and each SECURED EXPERIMENT with movable parts is in its most reactive state.

3.1.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.] ENSURE REACTOR A.1 IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meet LCO operations TS-11 Original (4/14)

K-State Reactor K-State Reactor TS-1 1 Original (4/14)

TECHNICAL SPECIFICATIONS B. La ENSURE control rods B. I IMMEDIATE fully inserted AND B.l.b Secure electrical power to the control rod circuits B. The reactor is not subcritical by more than AND

$0.50 under specified conditions B. L.c Secure all work on in- B.2 Prior to continued core experiments or operations installed control rod drives AND B.2 Configure reactor to meet LCO 3.1.5 Bases The value for excess reactivity was used in establishing core conditions for calculations (Table 13.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis. Since the fundamental protection for the KSU reactor is the maximum power level and fuel temperature that can be achieved with the available positive core reactivity, experiments with positive reactivity are included in determining excess reactivity. Since experiments with negative reactivity will increase available reactivity if they are removed during operation, they are not credited in determining excess reactivity.

Analysis (Chapter 13) shows fuel temperature will not exceed 1,1 50'C for the stainless-steel-clad fuel in the event of inadvertent or accidental pulsing of the reactor. Section 13.2 demonstrates that a $3.00 reactivity insertion from critical, zero power conditions leads to maximum fuel temperature of 746°C, while a $1.00 reactivity insertion from a worst-case steady state operation at 107 kW leads to a maximum fuel temperature of 869°C, well below the safety limit.

The limiting SHUTDOWN MARGIN is necessary so that the reactor can be shut down from any operating condition, and will remain shut down after cool down and xenon decay, even if one control rod (including the transient control rod) should remain in the fully withdrawn position.

TS-12 Original (4/14)

Reactor K-State Reactor TS-12 Original (4/14)

TECHNICAL SPECIFICATIONS 3.2 PULSED MODE Operations 3.2.1 Applicability These specifications apply to operation of the reactor in the PULSE MODE.

3.2.2 Objective This Limiting Condition for Operation prevents fuel temperature safety limit from being exceeded during PULSE MODE operation.

3.2.3 Specification The transient rod drive is positioned for reactivity insertion (upon withdrawal) less than or (1) equal to $3.00 3.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I Position the transient rod drive A. 1 IMMEDIATE A. With all stainless steel clad for pulse rod worth less than fuel elements, the worth of or equal to $3.00 the pulse rod in the OR transient rod drive position OR is greater than $3.00 in the PULSE MODE A.2 Place reactor in STEADY A.2 IMMEDIATE STATE MODE 3.2.5 Bases The value for pulsed reactivity with all stainless steel elements in the core was used in establishing core conditions for calculations (Table 13.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis.

K-State Reactor TS-13 Original (4/14)

TECHNICAL SPECIFICATIONS 3.3 MEASURING CHANNELS 3.3.1 Applicability This specification applies to the reactor MEASURING CHANNELS during STEADY STATE MODE and PULSE MODE operations.

3.3.2 Objective The objective is to require that sufficient information is available to the operator to ensure safe operation of the reactor 3.3.3 Specifications (1) The MEASURING CHANNELS specified in TABLE I SHALL be OPERATING The neutron count rate on the startup channel is greater than the minimum detector (2) sensitivity TABLE 1: MINIMUM MEASURING CHANNEL COMPLEMENT Minimum Number Operable MEASURING CHANNEL STEADY STATE PULSE MODE MODE Reactor power leveIl~ 2 1 Primary Pool Water Temperature I I Reactor Bay Differential Pressure I I Fuel Temperature 1 1 22 foot Area radiation monitor 1 1 0 or 12 foot Area monitor I Continuous air radiation monitor 21 121 EXHAUST PLENUM radiation monitor[21 11 NOTE[l]: One "Startup Channel" required to have range that indicates <1 0 W NOTE[2]: High-level alarms audible in the control room may be used 3.3.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. 1 Reactor power channels A.I.1 Restore channel to operation A.1.1 IMMEDIATE not OPERATING (min 2 OR for STEADY STATE, I A.1.2 ENSURE reactor is A.1.2 IMMEDIATE PULSE MODE) SHUTDOWN K-State Reactor TS-14 Original (4/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less AND A.2. IMMEDIATE than 90% of required operating value A.2.2 Enter REACTOR SECURED mode B. Primary water temperature, B. 1 Restore channel to operation A. I IMMEDIATE reactor bay differential OR pressure or fuel temperature CHANNEL B.2 ENSURE reactor is A.2 IMMEDIATE not operable SHUTDOWN C. 1 Restore MEASURING C.1 IMMEDIATE CHANNEL OR C.2 ENSURE reactor is shutdown C.2 IMMEDIATE 22 foot Area radiation OR C.

monitor OPERTIN is not isC.3 ENSURE personnel are not C.3 IMMEDIATE on the 22 foot level OR C.4 ENSURE personnel on 22 C.4 IMMEDIATE foot level are using portable survey meters to monitor dose rates D.I Restore MEASURING D.1 IMMEDIATE CHANNEL OR D.2 ENSURE reactor is shutdown D.2 IMMEDIATE OR D. 0 or 12 foot Area monitor is not OPERATING D.3 ENSURE personnel are not in D.3 IMMEDIATE the reactor bay OR D.4 ENSURE personnel entering D.4 IMMEDIATE reactor bay are using portable survey meters to monitor dose rates TS-155 Original (4/14)

Reactor K-State Reactor TS-1 Original (4/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME E.1 Restore MEASURING E.1 IMMEDIATE CHANNEL OR E.2 ENSURE reactor is shutdown E.2. IMMEDIATE E. Continuous air radiation OR monitor is not OPERATING E.3.a ENSURE EXHAUST E.3.a. IMMEDIATE PLENUM radiation monitor is OPERATING AND E.3.b Restore MEASURING E.3.b Within 30 days CHANNEL F. 1 Restore MEASURING F.1 IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown F.2. IMMEDIATE F. Exhaust plenum radiation OR monitor is not OPERATING F.3.a ENSURE continuous air F.3.a. IMMEDIATE radiation monitor is OPERATING AND F.3.b Restore MEASURING F.3.b Within 30 days CHANNEL G. I Do not perform a reactor G. I IMMEDIATE G. The neutron count rate on startup the startup channel is not OR greater than the minimum G.2 Perform a neutron-source detector sensitivity check on the startup channel G.2 IMMEDIATE prior to startup 3.3.5 Bases Maximum steady state power level is 1,250 kW; neutron detectors measure reactor power level.

Chapter 4 and 13 discuss normal and accident heat removal capabilities. Chapter 7 discusses radiation detection and monitoring systems, and neutron and power level detection systems.

According to General Atomics, detector voltages less than 90% of required operating value do not provide reliable, accurate nuclear instrumentation. Therefore, if operating voltage falls below the minimum value the power level channel is inoperable.

K-State Reactor TS-16 Original (4/14)

TECHNICAL SPECIFICATIONS Primary water temperature indication is required to assure water temperature limits are met, protecting primary cleanup resin integrity. The reactor bay differential pressure indictor is required to control reactor bay atmosphere radioactive contaminants. Fuel temperature indication provides a means of observing that the safety limits are met.

The 22-foot and 0-foot area radiation monitors provide information about radiation hazards in the reactor bay. A loss of reactor pool water (Chapter 13), changes in shielding effectiveness (Chapter 11), and releases of radioactive material to the restricted area (Chapter 11) could cause changes in radiation levels within the reactor bay detectable by these monitors. Portable survey instruments will detect changes in radiation levels.

The air monitors (continuous air- and exhaust plenum radiation-monitor) provide indication of airborne contaminants in the reactor bay prior to discharge of gaseous effluent. Iodine channels provide evidence of fuel element failure. The air monitors provide similar information on independent channels; the continuous air monitor (CAM) has maximum sensitivity to iodine and particulate activity, while the air monitoring system (AMS) has individual channels for radioactive particulate, iodine, noble gas and iodine.

When filters in the air monitoring system begin to load, there are frequent, sporadic trips of the AMS alarms. Although the filters are changed on a regular basis, changing air quality makes these trips difficult to prevent. Short outages of the AMS system have resulted in unnecessary shutdowns, exercising the shutdown mechanisms unnecessarily, creating stressful situations, and preventing the ability to fully discharge the mission of the facility while the CAM also monitors conditions of airborne contamination monitored by the AMS. The AMS detector has failure modes than cannot be corrected on site; AMS failures have caused longer outages at the K-State reactor. The facility has experienced approximately two-week outages, with one week dedicated to testing and troubleshooting and (sometimes) one-week for shipment and repair at the vendor facility.

Permitting operation using a single channel of atmospheric monitoring will reduce unnecessary shutdowns while maintaining the ability to detect abnormal conditions as they develop. Relative indications ensure discharges are routine; abnormal indications trigger investigation or action to prevent the release of radioactive material to the surrounding environment. Ensuring the alternate airborne contamination monitor is functioning during outages of one system provides the contamination monitoring required for detecting abnormal conditions. Limiting the outage for a single unit to a maximum of 30 days ensures radioactive atmospheric contaminants are monitored while permitting maintenance and repair outages on the other system.

Chapter 13 discusses inventories and releases of radioactive material from fuel element failure into the reactor bay, and to the environment. Particulate and noble gas channels monitor more routine discharges. Chapter II and SAR Appendix A discuss routine discharges of radioactive gasses generated from normal operations into the reactor bay and into the environment. Chapter 3 identifies design bases for the confinement and ventilation system.

Chapter 7 discusses air-monitoring systems.

Experience has shown that subcritical multiplication with the neutron source used in the reactor does not provide enough neutron flux to correspond to an indicated power level of 10 Watts.

Therefore an indicated power of 10 Watts or more indicates operating in a potential critical condition, and at least one neutron channel is required with sensitivity at a neutron flux level corresponding to reactor power levels less than 10 Watts ("Startup Channel"). If the indicated neutron level is less than the minimum sensitivity for both the log-wide range and the multirange linear power level channels, a neutron source will be used to determine that at least one of the channels is responding to neutrons to ensure that the channel is functioning prior to startup.

TS-17 Original (4/14)

K-State Reactor K-State Reactor TS-17 Original (4/14)

TECHNICAL SPECIFICATIONS 3.4 Safety Channel and Control Rod Operability 3.4.1 Applicability This specification applies to the reactor MEASURING Channels during STEADY STATE MODE and PULSE MODE operations.

3.4.2 Objective The objectives are to require the minimum number of REACTOR SAFETY SYSTEM channels that must be OPERABLE in order to ensure that the fuel temperature safety limit is not exceeded, and to ensure prompt shutdown in the event of a scram signal.

3.4.3 Specifications (1) The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLE CONTROL RODS (STANDARD) are capable of 90% of full reactivity insertion from the (2) fully withdrawn position in less than I sec.

TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Safety System Channel Minimum NubrSEDPUE Function Required OPERATING Mode orItrokNumber STEADY PULSE Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar I Scram YES YES CONTROL ROD Prevent withdrawal of standard NA (STANDARD) position 1 rods in the PULSE MODE YES interlock Prevent inadvertent pulsing Pulse rod interlock I while in STEADY STATE YES NA MODE 3.4.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I Restore channel or interlock A]. IMMEDIATE A. Any required SAFETY to operation SYSTEM CHANNEL or interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN 3.4.5 Bases The power level scram is provided to ensure that reactor operation stays within the licensed limits of 1,250 kW, preventing abnormally high fuel temperature. The power level scram is not credited K-State Reactor TS-18 Original (4/14)

TECHNICAL SPECIFICATIONS in analysis, but provides defense in depth to assure that the reactor is not operated in conditions beyond the assumptions used in analysis (Table 13.2.1.4).

The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs.

The CONTROL ROD (STANDARD) interlock function is to prevent withdrawing control rods (other than the pulse rod) when the reactor is in the PULSE MODE. This will ensure the reactivity addition rate during a pulse is limited to the reactivity added by the pulse rod.

The pulse rod interlock function prevents air from being applied to the transient rod drive when it is withdrawn while disconnected from the control rod to prevent inadvertent pulses during STEADY STATE MODE operations. The control rod interlock prevents inadvertent pulses which would be likely to exceed the maximum range of the power level instruments configured for steady state operations.

K-State Reactor TS-19 Original (4/14)

TECHNICAL SPECIFICATIONS 3.5 Gaseous Effluent Control 3.5.1 Applicability This specification applies to gaseous effluent in STEADY STATE MODE and PULSE MODE.

3.5.2 Objective The objective is to ensure that exposures to the public resulting from gaseous effluents released during normal operations and accident conditions are within limits and ALARA.

3.5.3 Specification (1) The reactor bay ventilation exhaust system SHALL maintain in-leakage to the reactor bay Releases of Ar-41 from the reactor bay exhaust plenum to an unrestricted environment (2) SHALL NOT exceed 30 Ci per year.

3.5.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.1 ENSURE reactor is A.1 IMMEDIATE SHUTDOWN OR A.2.a Do not OPERATE in the A.2.a IMMEDIATE PULSE MODE AND A.2.b Secure EXPERIMENT A.2.b IMMEDIATE A. The reactor bay ventilation operations for exhaust system is not EXPERIMENT with failure OPERABLE modes that could result in the release of radioactive gases or aerosols.

A.2.c ENSURE no irradiated fuel A.2.b IMMEDIATE handing AND A.2.d Restore the reactor bay A.2.d Within 30 days ventilation exhaust system to OPERABLE K-State Reactor TS-20 Original (4/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME Calculated releases of Ar-41 from the reactor bay exhaust Do not operate. IMMEDIATE plenum exceed 30 Ci per year.

3.5.5 Bases The confinement and ventilation system is described in Section 3.5.4. Routine operations produce radioactive gas, principally Argon 41, in the reactor bay. If the reactor bay ventilation system is secured, SAR Chapter 11 Appendix A demonstrates reactor bay concentration of 0.746 Bq ml-'(2.0lx05 tiCi ml-'), well below the IOCFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10-3 1tCi h/mL. Therefore, the reduction in concentration of Argon 41 from operation of the confinement and ventilation system is a defense in depth measure, and not required to assure meeting personnel exposure limits. Consequently, the ventilation system can be secured without causing significant personnel hazard from normal operations. Thirty days for a confinement and ventilation system outage is selected as a reasonable interval to allow major repairs and work to be accomplished, if required. During this interval, experiment activities that might cause airborne radionuclide levels to be elevated are prohibited.

It is shown in Section 13.2.2 of the Safety Analysis Report that, if the reactor were to be operating at full steady-state power, fuel element failure would not occur even if all the reactor tank water were to be lost instantaneously.

Section 13.2.4 addresses the maximum hypothetical fission product inventory release. Using unrealistically conservative assumptions, concentrations for a few nuclides of iodine would be in excess of occupational derived air concentrations for a matter of hours or days. 90Sr activity available for release from fuel rods previously used at other facilities is estimated to be at most about 4 times the ALI. In either case (radio-iodine or -Sr), there is no credible scenario for accidental inhalation or ingestion of the undiluted nuclides that might be released from a damaged fuel element. Finally, fuel element failure during a fuel handling accident is likely to be observed and mitigated immediately.

SAR Appendix A shows the release of 30 Ci per year of Ar-41 from normal operations would result in less than 10 mrem annual exposure to any person in unrestricted areas.

K-State Reactor TS-21 Original (4/14)

TECHNICAL SPECIFICATIONS 3.6 Limitations on Experiments 3.6.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.6.2 Objectives These Limiting Conditions for Operation prevent reactivity excursions that might cause the fuel temperature to exceed the safety limit (with possible resultant damage to the reactor), and the excessive release of radioactive materials in the event of an EXPERIMENT failure 3.6.3 Specifications If all fuel elements are stainless steel clad, the reactivity worth of any individual (1) EXPERIMENT SHALL NOT exceed $2.00 If two or more experiments in the reactor are interrelated so that operation or failure of (2) one can induce reactivity-affecting change in the other(s), the sum of the absolute reactivity of such experiments SHALL NOT exceed $2.00.

Irradiation holders and vials SHALL prevent release of encapsulated material in the reactor pool and core area 3.6.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.1 ENSURE the reactor is A.] IMMEDIATE SHUTDOWN A. INDEPENDENT EXPERIMENT worth is AND greater than $2.00 A.2 Remove the experiment A.2 Prior to continued operations C.] ENSURE the reactor is C.] IMMEDIATE SHUTDOWN C. An irradiation holder or vial AND releases material capable of causing damage to the C.2 Inspect the affected area C.2 Prior to continued reactor fuel or structure into operation the pool or core area AND C.3 Obtain RSC review and C.3 Prior to continued approval operation TS-22 Original (4/14)

K-State Reactor K-State Reactor TS-22 Original (4/14)

TECHNICAL SPECIFICATIONS 3.6.5 Bases Specifications 3.7(1) through 3.7(3) are conservatively chosen based on prior operation at 250 kW to limit reactivity additions to maximum values that are less than an addition which could cause temperature to challenge the safety limit.

Experiments are approved with expectations that there is reasonable assurance the facility will not be damaged during normal or failure conditions. If an irradiation capsule which contains material with potential for challenging the fuel cladding or pool wall, the facility will be inspected to ensure that continued operation is acceptable.

TS-23 Original (4/14)

K-State Reactor K-State Reactor TS-23 Original (4/14)

TECHNICAL SPECIFICATIONS 3.7 Fuel Integrity 3.7.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.7.2 Objective The objective is to prevent the use of damaged fuel in the KSU TRIGA reactor.

3.7.3 Specifications Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over (1) manufactured length (2) Fuel elements in the reactor core SHALL NOT be laterally bent more than 1/8 in.

3.7.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. Any fuel element is elongated greater than 1/8 Do not insert the fuel element into in. over manufactured IMMEDIATE length, or bent laterally greater than 1/8 in.

3.7.5 Bases The above limits on the allowable distortion of a fuel element have been shown to correspond to strains that are considerably lower than the strain expected to cause rupture of a fuel element and have been successfully applied at TRIGA installations. Fuel cladding integrity is important since it represents the only process barrier for fission product release from the TRIGA reactor.

TS-24 Original (4/14)

K-State Reactor K-State Reactor TS-24 Original (4/14)

TECHNICAL SPECIFICATIONS 3.8 Reactor Pool Water 3.8.1 Applicability This specification applies to operations in STEADY STATE MODE, PULSE MODE, and SECURED MODE.

3.8.2 Objective The objective is to set acceptable limits on the water quality, temperature, conductivity, and level in the reactor pool.

3.8.3 Specifications (1) Water temperature at the exit of the reactor pool SHALL NOT exceed 130'F with flow through the primary cleanup loop (2) Water conductivity SHALL be less than 5 jimho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core 3.8.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I ENSURE the reactor is A.1 IMMEDIATE SHUTDOWN AND A. Water temperature at the A.2 Secure flow through the A.2 IMMEDIATE exit of the reactor pool demineralizer exceeds 130°F AND A.3 Reduce water temperature to A.3 IMMEDIATE less than 130°F B.I ENSURE the reactor is B.1 IMMEDIATE SHUTDOWN B. Water conductivity is greater than 5 Iimho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 _imho/cm TS-25 Original (4/14)

K-State Reactor K-State Reactor TS-25 Original (4/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME C.I ENSURE the reactor is C.1 IMMEDIATE C. Water level above the core SHUTDOWN SHALL be at least 13 ft from the top of the core for AND all operating conditions C.2 Restore water level C.2 IMMEDIATE 3.8.5 Bases The resin used in the mixed bed deionizer limits the water temperature of the reactor pool. Resin in use (as described in Section 5.4) maintains mechanical and chemical integrity at temperatures below 130'F (54.41C).

Maintaining low water conductivity over a prolonged period prevents possible corrosion, deionizer degradation, or slow leakage of fission products from degraded cladding. Although fuel degradation does not occur over short time intervals, long-term integrity of the fuel is important, and a 4-week interval was selected as an appropriate maximum time for high conductivity.

The top of the core is 16 feet below the top of the primary coolant tank. The lowest suction of primary cooling flow into the forced cooling loop is 3.5 feet below the top of the primary coolant tank (water level is maintained about 6 inches below the top of the tank).

The principle contributor to radiation dose rates at the pool surface is Nitrogen 16 generated in the reactor core and dispersed in the pool. Calculations in Chapter 11 show the pool surface radiation dose rates from Nitrogen 16 with 13 feet of water above the core are acceptable.

For normal pool temperature, calculations in Chapter 4 assuming 16 feet and 13 feet above the core demonstrate that the heat flux of the hottest area of the fuel rod generating the highest power level in the core during operations is less than the critical heat flux by a large margin up to the maximum permitted cooling temperatures; margin remains even at temperatures approaching bulk boiling for atmospheric conditions. Therefore, pool levels greater than 13 feet above the core meet requirements for safe operation with respect to maximum fuel temperature and thermal hydraulics by a wide margin.

Therefore, a minimum pool level of 13 feet above the core is adequate to provide shielding and support the core cooling.

TS-26 Original (4/14)

K-State Reactor K-State Reactor TS-26 Original (4/14)

TECHNICAL SPECIFICATIONS 3.9 Maintenance Retest Requirements 3.9.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.9.2 Objective The objective is to ensure Technical Specification requirements are met following maintenance that occurs within surveillance test intervals.

3.9.3 Specifications Maintenance activities SHALL NOT change, defeat or alter equipment or systems in a way that prevents the systems or equipment from being OPERABLE or otherwise prevent the systems or equipment from fulfilling the safety basis 3.9.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME Maintenance is performed that has the potential to change a Perform surveillance Prior to continued, setpoint, calibration, flow rate, normal operation in or other parameter that is OR STEADY STATE measured or verified in MODE or PULSE meeting a surveillance or Operate only to perform retest MODE operability requirement 3.9.5 Bases Operation of the K-State reactor will comply with the requirements of Technical Specifications.

This specification ensures that if maintenance might challenge a Technical Specifications requirement, the requirement is verified prior to resumption of normal operations.

TS-27 Original (4/14)

Reactor K-State Reactor TS-27 Original (4/14)

TECHNICAL SPECIFICATIONS

4. Surveillance Requirements 4.1 Core Reactivity 4.1.1 Objective This surveillance ensures that the minimum SHUTDOWN MARGIN requirements and maximum excess reactivity limits of section 3.1 are met.

4.1.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SHUTDOWN MARGIN Determination SEMIANNUAL SEMIANNUAL Excess Reactivity Determination Following Insertion of experiments with measurable positive reactivity Control Rod Reactivity Worth determination BIENNIAL 4.1.3 Basis Experience has shown verification of the minimum allowed SHUTDOWN MARGIN at the specified frequency is adequate to assure that the limiting safety system setting is met When core reactivity parameters are affected by operations or maintenance, additional activity is required to ensure changes are incorporated in reactivity evaluations.

TS-28 Original (4/14)

K-State Reactor K-State Reactor TS-28 Original (4/14)

TECHNICAL SPECIFICATIONS 4.2 PULSE MODE 4.2.1 Objectives The verification that the pulse rod position does not exceed a reactivity value corresponding to

$3.00 assures that the limiting condition for operation is met.

4.2.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ENSURE Transient Pulse Rod position corresponds to reactivity Protoplig pea on not greater than $3.00Protopligpeaon 4.2.3 Basis Verifying pulse rod position corresponds to less than $3.00 ensures that the maximum pulsed reactivity meets the limiting condition for operation.

TS-29 Original (4/14)

K-State Reactor K-State Reactor TS-29 Original (4/14)

TECHNICAL SPECIFICATIONS 4.3 MEASURING CHANNELS 4.3.1 Objectives Surveillances on MEASURING CHANNELS at specified frequencies ensure instrument problems are identified and corrected before they can affect operations.

4.3.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Reactor power level MEASURING CHANNEL CHANNEL TEST DAILY Calorimetric calibration ANNUAL CHANNEL CHECK high voltage to required power level instruments Primary pool water temperature CHANNEL CALIBRATION ANNUAL Reactor Bay differential pressure CHANNEL CALIBRATION ANNUAL Fuel temperature CHANNEL CALIBRATION ANNUAL 22 Foot Area radiation monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL Continuous Air Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL EXHAUST PLENUM Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL Startup Count Rate DAILY 4.3.3 Basis The DAILY CHANNEL CHECKS will ensure that the SAFETY SYSTEM and MEASURING CHANNELS are operable. The required periodic calibrations and verifications will permit any long-term drift of the channels to be corrected.

TS-30 Original (4/14)

K-State Reactor K-State Reactor TS-30 Original (4/14)

TECHNICAL SPECIFICATIONS 4.4 Safety Channel and Control Rod Operability 4.4.1 Objective The objectives of these surveillance requirements are to ensure the REACTOR SAFETY SYSTEM will function as required. Surveillances related to safety system MEASURING CHANNELS ensure appropriate signals are reliably transmitted to the shutdown system; the surveillances in this section ensure the control rod system is capable of providing the necessary actions to respond to these signals.

4.4.2 Specifications SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Manual scram SHALL be tested by releasing partially withdrawn DAILY CONTROL RODS (STANDARD)

CONTROL ROD (STANDARD) drop times SHALL be measured to have a drop time from the fully withdrawn position of less than ANNUAL I sec.

The control rods SHALL be visually inspected for corrosion and BIENNIAL mechanical damage at intervals CONTROL ROD (STANDARD) position interlock functional test SEMIANNUAL Pulse rod interlock functional test SEMIANNUAL On each day that PULSE MODE operation of the reactor is Prior to pulsing operations planned, a functional performance check of the CONTROL ROD each day a pulse is planned (TRANSIENT) system SHALL be performed.

The CONTROL ROD (TRANSIENT) rod drive cylinder and the associated air supply system SHALL be inspected, cleaned, and SEMIANNUAL lubricated, as necessary.

4.4.3 Basis Manual and automatic scrams are not credited in accident analysis, although the systems function to assure long-term safe shutdown conditions. The manual scram and control rod drop timing surveillances are intended to monitor for potential degradation that might interfere with the operation of the control rod systems. The verification of high voltage to the power level monitoring channels assures that the instrument channel providing an overpower trip will function on demand.

The control rod inspections (visual inspections and transient drive system inspections) are similarly intended to identify potential degradation that lead to control rod degradation or inoperability.

A test of the interlock that prevents the pulse rod from coupling to the drive in the state state mode unless the drive is fully down assures that pulses will occur only when in pulsing mode. A K-State Reactor TS-31 Original (4/14)

TECHNICAL SPECIFICATIONS test of the interlock that prevents standard control rod motion while in the pulse mode assures that the interlock will function as required.

The functional checks of the control rod drive system assure the control rod drive system operates as intended for any pulsing operations. The inspection of the pulse rod mechanism will assure degradation of the pulse rod drive will be detected prior to malfunctions.

TS-32 Original (4/14)

K-State Reactor K-State Reactor TS-32 Original (4/14)

TECHNICAL SPECIFICATIONS 4.5 Gaseous Effluent Control 4.5.1 Objectives These surveillances ensure that routine releases are normal, and (in conjunction with MEASURING CHANNEL surveillances) that instruments will alert the facility if conditions indicate abnormal releases.

4.5.2 Specification SURVIELLANCE REQUIREMENTS SU RVEILLANCE FREQUENCY Perform CHANNEL TEST of air monitor ANN-UAL Verify negative reactor bay differential pressure DAILY 4.5.3 Basis The continuous air monitor provides indication that levels of radioactive airborne contamination in the reactor bay are normal.

If the reactor bay differential pressure gage indicates a negative pressure, the reactor bay exhaust fan is controlling airflow by directing effluent out of confinement.

TS-33 Original (4/14)

K-State Reactor K-State Reactor TS-33 Original (4/14)

TECHNICAL SPECIFICATIONS 4.6 Limitations on Experiments 4.6.1 Objectives This surveillance ensures that experiments do not have significant negative impact on safety of the public, personnel or the facility.

4.6.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Prior to inserting a new Experiments SHALL be evaluated and approved prior to experiment for purposes implementation. other than determination of reactivity worth Initial insertion of a new experim nwe able Measure and record experiment worth of the EXPERIMENT of the estimated worth is greater than experiment where absolute

$0.40). the absolute value (where value of the estimated worth is greater than $0.40 4.6.3 Basis These surveillances allow determination that the limits of 3.7 are met.

Experiments with an absolute value of the estimated significant reactivity worth (greater than

$0.40) will be measured to assure that maximum experiment reactivity worths are met. If an absolute value of the estimate indicates less than $0.40 reactivity worth, even a 100% error will result in actual reactivity less than the assumptions used in analysis for inadvertent pulsing at low power operations in the Safety Analysis Report (13.2.3, Case I).

TS-34 Original (4/14)

K-State Reactor K-State Reactor TS-34 Original (4/14)

TECHNICAL SPECIFICATIONS 4.7 Fuel Integrity 4.7.1 Objective The objective is to ensure that the dimensions of the fuel elements remain within acceptable limits.

4.7.2 Applicability This specification applies to the surveillance requirements for the fuel elements in the reactor core.

4.7.3 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 500 pulses of magnitude equal to or less than a pulse insertion of 3.00$

The standard fuel elements SHALL be visually inspected for cor- AND rosion and mechanical damage, and measured for length and bend Following the exceeding of a limited safety system set point with potential for causing degradation B, C, D, E, and F RING elements comprising approximately 1/3 of the core SHALL be visually inspected annually for corrosion and ANNUAL mechanical damage such that the entire core SHALL be inspected at 3-year intervals, but not to exceed 38 months 4.7.4 Basis The most severe stresses induced in the fuel elements result from pulse operation of the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply.

Triennial visual inspection of fuel elements combined with measurements at intervals determined by pulsing as described is considered adequate to identify potential degradation of fuel prior to catastrophic fuel element failure.

TS-35 Original (4/14)

K-State Reactor K-State Reactor TS-35 Original (4/14)

TECHNICAL SPECIFICATIONS 4.8 Reactor Pool Water This specification applies to the water contained in the KSU TRIGA reactor pool.

4.8.1 Objective The objective is to provide surveillance of reactor primary coolant water quality, pool level, temperature and (in conjunction with MEASURING CHANNEL surveillances), and conductivity.

4.8.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify reactor pool water level above the inlet line vacuum breaker DAILY Verify reactor pool water temperature channel operable DAILY DAILY Measure reactor Pool water conductivity At least every 20 days 4.9.3 Bases Surveillance of the reactor pool will ensure that the water level is adequate before reactor operation. Evaporation occurs over longer periods of time, and daily checks are adequate to identify the need for water replacement.

Water temperature must be monitored to ensure that the limit of the deionizer will not be exceeded. A daily check on the instrument prior to reactor operation is adequate to ensure the instrument is operable when it will be needed.

Water conductivity must be checked to ensure that the deionizer is performing properly and to detect any increase in water impurities. A daily check is adequate to verify water quality is appropriate and also to provide data useful in trend analysis. If the reactor is not operated for long periods of time, the requirement for checks at least every 20 days will ensure water quality is maintained in a manner that does not permit fuel degradation.

TS-36 Original (4/14)

K-State Reactor K-State Reactor TS-36 Original (4/14)

TECHNICAL SPECIFICATIONS 4.9 Maintenance Retest Requirements 4.9.1 Objective The objective is to ensure that a system is OPERABLE within specified limits before being used after maintenance has been performed.

4.9.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Following maintenance of Evaluate potential for maintenance activities to affect operability systems of equipment and function of equipment required by Technical Specifications required by Technical Specifications Perform surveillance to assure affected function meets Prior to resumption of requirements normal operations 4.9.3 Bases This specification ensures that work on the system or component has been properly carried out and that the system or component has been properly reinstalled or reconnected before reliance for safety is placed on it.

TS-37 Original (4/14)

K-State Reactor K-State Reactor TS-37 Original (4/14)

TECHNICAL SPECIFICATIONS

5. Design Features 5.1 Reactor Fuel 5.1.1 Applicability This specification applies to the fuel elements used in the reactor core.

5.1.2 Objective The objective is to ensure 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 mechanical integrity.

5.1.3 Specification (I) The high-hydride fuel element shall contain uranium-zirconium hydride, clad in 0.020 in.

of 304 stainless steel. It shall contain a maximum of 12.5 weight percent uranium which has a maximum enrichment of 20%. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom.

(2) For the loading process, the elements shall be placed in a close packed array except for experimental facilities or for single positions occupied by control rods and a neutron startup source.

(3) Up to four elements with greater than 9.0 weight percent uranium may be installed in the core. These elements may only be placed in the E- and F-rings of the core lattice, and may not be adjacent to control rods or water channels.

5.1.4 Bases These types of fuel elements have a long history of successful use in TRIGA reactors.

Calculations show that 12%-load fuel in the E- and F-rings will not exceed the temperature of 8%-load instrumented elements in the B-ring. Additionally the power peaking and fission product inventory assumptions in the SAR will not be challenged by 12% fuel in the E- and F-rings. Local power and temperature peaking effects during pulsing are avoided by prohibiting placement of the 12%-load fuel near water and control rod channels.

TS-38 Original (4/14)

Reactor K-State Reactor TS-38 Original (4/14)

TECHNICAL SPECIFICATIONS 5.2 Reactor Fuel and Fueled Devices in Storage 5.2.1 Applicability This specification applies to reactor fuel elements in storage 5.2.2 Objective The objective is to ensure fuel elements or fueled devices in storage are maintained Subcritical in a safe condition.

5.2.3 Specification (I) All fuel elements or fueled devices shall be in a safe, stable geometry (2) The k~ff of all fuel elements or fueled devices in storage is less than 0.8 (3) Irradiated fuel elements or fueled devices will be stored in an array which will permit sufficient natural convection cooling by air or water such that the fuel element or fueled device will not exceed design values.

5.2.4 Bases This specification is based on American Nuclear Society standard 15.1, section 5.4.

5.3 Reactor Building 5.3.1 Applicability This specification applies to the building that houses the TRIGA reactor facility.

5.3.2 Objective The objective is to ensure that provisions are made to restrict the amount of release of radioactivity into the environment.

5.3.3 Specification (1) The reactor shall be housed in a closed room designed to restrict leakage when the reactor is in operation, when the facility is unmanned, or when spent fuel is being handled exterior to a cask.

TS-39 Original (4/14)

Reactor K-State Reactor TS-39 Original (4/14)

TECHNICAL SPECIFICATIONS (2) The minimum free volume of the reactor room shall be approximately 144,000 cubic feet.

(3) The building shall be equipped with a ventilation system capable of exhausting air or other gases from the reactor room at a minimum of 30 ft. above ground level.

5.3.4 Bases To control the escape of gaseous effluent, the reactor room contains no windows that can be opened. The room air is exhausted through an independent exhaust system, and discharged at roof level to provide dilution.

5.4 Experiments 5.4.1 Applicability This specification applies to the design of experiments.

5.4.2 Objective The objective is to ensure that experiments are designed to meet criteria.

5.4.3 Specifications (1) EXPERIMENT with a design reactivity worth greater than $1.00 SHALL be securely fastened (as defined in Section 1, Secured Experiment).

(2) Design shall ensure that failure of an EXPERIMENT SHALL NOT lead to a direct failure of a fuel element or of other experiments that could result in a measurable increase in reactivity or a measurable release of radioactivity due to the associated failure.

(3) EXPERIMENT SHALL be designed so that it does not cause bulk boiling of core water (4) EXPERIMENT design SHALL ensure no interference with control rods or shadowing of reactor control instrumentation.

(5) EXPERIMENT design shall minimize the potential for industrial hazards, such as fire or the release of hazardous and toxic materials.

(6) Each fueled experiment shall be limited such that the total inventory of iodine isotopes 131 through 135 in the experiment is not greater than 5 millicuries except as the fueled experiment is a standard TRIGA instrumented element in which instance the iodine inventory limit is removed.

(7) Where the possibility exists that the failure of an EXPERIMENT (except fueled EXPERIMENTS) could release radioactive gases or aerosols to the reactor bay or atmosphere, the quantity and type of material shall be limited such that the airborne K-State Reactor TS-40 Original (4/14)

TECHNICAL SPECIFICATIONS concentration of radioactivity averaged over a year will not exceed the limits of Table II of Appendix B of 10 CFR Part 20 assuming 100% of the gases or aerosols escape.

(8) The following assumptions shall be used in experiment design:

a. If effluents from an experimental facility exhaust through a hold-up tank which closes automatically at a high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

b. If effluents from an experimental facility exhaust through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of the aerosols produced will escape.

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

(9) Use of explosive solid or liquid material with a National Fire Protection Association Reactivity (Stability) index of 2, 3, or 4 in the reactor pool or biological shielding SHALL NOT exceed the equivalent of 25 milligrams of TNT without prior NRC approval.

5.4.4 Basis Designing the experiment to reactivity and thermal-hydraulic conditions ensure that the experiment is not capable of breaching fission product barriers or interfering with the control systems (interferences from other - than reactivity - effects with the control and safety systems are also prohibited). Design constraints on industrial hazards ensure personnel safety and continuity of operations. Design constraints limiting the release of radioactive gasses prevent unacceptable personnel exposure during off-normal experiment conditions.

TS-41 Original (4/14)

Reactor K-State Reactor TS-41 Original (4/14)

TECHNICAL SPECIFICATIONS

6. Administrative Controls 6.1 Organization and Responsibilities of Personnel a) Structure.

The reactor organization is related to the University structure as shown in SAR Figure 12.1 and Technical Specifications Figure TS. 1 below.

Kansas State University (KSU) holds the license for the KSU TRIGA Reactor, located in the KSU Nuclear Reactor Facility in Ward Hall on the campus of Kansas State University. The chief administrating officer for KSU is the President. Environment, safety and health oversight functions are administered through the Vice President for Administration and Finance, while reactor line management functions are through the Provost Chief Academic Officer.

Figure TS.1: Organization and Management Structure for the K-State Reactor Radiation protection functions are divided between the University Radiation Safety Officer (RSO) and the reactor staff and management, with management and authority for the RSO separate from line management and authority for facility operations. Day-to-day radiation protection functions implemented by facility staff and management are guided K-State Reactor TS-42 Original (4/14)

TECHNICAL SPECIFICATIONS by approved administrative controls (Reactor Radiation Protection Program or RPP, Facility Operating Manual, operating and experiment procedures); these controls are reviewed and approved by the RSO as part of the Reactor Safeguards Committee (with specific veto authority). The RSO has specific oversight functions assigned though the RPP. The RSO provides routine support for personnel monitoring, radiological analysis, and radioactive material inventory control. The RSO provides guidance on request for non-routine operations such as transportation and implementation of new experiments.

b) Responsibility.

The President of the University shall be responsible for the appointment of responsible and competent persons as members of the TRIGA Reactor Safeguards Committee upon the recommendation of the ex officio Chairperson of the Committee.

The KSU Nuclear Reactor Facility shall be under the supervision of the Nuclear Reactor Facility Manager, who shall have the overall responsibility for safe, efficient, and competent use of its facilities in conformity with all applicable laws, regulations, terms of facility licenses, and provisions of the Reactor Safeguards Committee. The Manager also has responsibility for maintenance and modification of laboratories associated with the Reactor Facility. The Manager shall have education and/or experience commensurate with the responsibilities of the position and shall report to the Head of the Department of Mechanical and Nuclear Engineering.

A Reactor Supervisor may serve as the deputy of the Nuclear Reactor Facility Manager in all matters relating to the enforcement of established rules and procedures (but not in matters such as establishment of rules, appointments, and similar administrative functions). The Supervisor should have at least two years of technical training beyond high school and shall possess a Senior Reactor Operator's license. The Supervisor shall have had reactor OPERATING experience and have a demonstrated competence in supervision. The Supervisor is appointed by the Nuclear Reactor Facility Manager and is responsible for enforcing all applicable rules, procedures, and regulations, for ensuring adequate exchange of information between OPERATING personnel when shifts change, and for reporting all malfunctions, accidents, and other potentially hazardous occurrences and situations to the Reactor Nuclear Reactor Facility Manager. The Nuclear Reactor Facility Manager may also serve as Reactor Supervisor.

The Reactor Operator shall be responsible for the safe and proper operation of the reactor, under the direction of the Reactor Supervisor. Reactor Operators shall possess an Operator's or Senior Operator's license and shall be appointed by the Nuclear Reactor Facility Manager.

The University Radiation Safety Officer (RSO), or a designated alternate, shall (in addition to other duties defined by the Director of Environmental Health and Safety, Division of Public Safety) be responsible for overseeing the safety of Reactor Facility operations from the standpoint of radiation protection. The RSO and/or designated alternate shall be appointed by the Director of Environmental Health and Safety, Division of Public Safety, with the approval of the University Radiation Safety Committee, and shall report to the Director of Environmental Health and Safety, whose organization is independent of the Reactor Facility organization, as shown on SAR Figure 12.1.

The Nuclear Reactor Facility Manager, with the approval of the Reactor Safeguards Committee, may designate an appropriately qualified member of the Facility organization as Reactor Facility Safety Officer (RFSO) with duties including those of an intra-Facility K-State Reactor TS-43 Original (4/14)

TECHNICAL SPECIFICATIONS Radiation Safety Officer. The University Radiation Safety Officer may, with the concurrence of the Nuclear Reactor Facility Manager, authorize the RFSO to perform some of the specific duties of the RSO at the Nuclear Reactor Facility.

c). Staffing.

Whenever the reactor is not secured, the reactor shall be under the direction of a (USNRC licensed) Senior Operator (designated as Reactor Supervisor). The Supervisor shall be on call, within twenty minutes travel time to the facility.

Whenever the reactor is not secured, a (USNRC licensed) Reactor Operator (or Senior Reactor Operator) who meets requirements of the Operator Requalification Program shall be at the reactor control console, and directly responsible for control manipulations.

In addition to the above requirements, during fuel movement a senior operator shall be inside the reactor bay directing fuel operations.

6.2 Review and Audit a) There will be a Reactor Safeguards Committee which shall review TRIGA reactor operations to assure that the reactor facility is operated and used in a manner within the terms of the facility license and consistent with the safety of the public and of persons within the Laboratory.

b) The responsibilities of the Committee include, but are not limited to, the following:

1. Review and approval of rules, procedures, and proposed Technical Specifications;

2. Review and approval of all proposed changes in the facility that could have a significant effect on safety and of all proposed changes in rules, procedures, and Technical Specifications, in accordance with procedures in Section 6.3;

3. Review and approval of experiments using the reactor in accordance with procedures and criteria in Section 6.4;

4. Determine whether changes in the facility as described in the safety analysis report (as updated), changes in the procedures as described in the final safety analysis report (as updated), and the conduct of tests or experiments not described in the safety analysis report (as updated) may be accomplished in accordance with 10 CFR 50.59 without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90.

5. Review of abnormal performance of plant equipment and OPERATING anomalies;

6. Review of unusual or abnonnal occurrences and incidents which are reportable under 10 CFR20 and 10 CFR50;

7. Inspection of the facility, review of safety measures, and audit of operations at a frequency not less than once a year, including operation and operations records of the facility; K-State Reactor TS-44 Original (4/14)

TECHNICAL SPECIFICATIONS

8. Requalification of the Nuclear Reactor Facility Manager and/or the Reactor Supervisor,

9. Review of container failures where released materials have the potential for damaging reactor fuel or structural components including:

a) results of physical inspection b) evaluation of consequences c) need for corrective actions c) The Committee shall be composed of:

1. one or more persons proficient in reactor and nuclear science or engineering,

2. one or more persons proficient in chemistry, geology, or chemical engineering,

3. one person proficient in biological effects of radiation,

4. the Nuclear Reactor Facility Manager, ex officio,

5. the University Radiation Safety Officer, ex officio, and,

6. The Head of the Department of Mechanical and Nuclear Engineering, ex officio, or a designated deputy, to serve as chairperson of the Committee.

The same individual may serve under more than one category above, but the minimum membership shall be seven. At least five members shall be faculty members. The Reactor Supervisor, if other than the Nuclear Reactor Facility Manager, shall attend and participate in Committee meetings, but shall not be a voting member.

d) The Committee shall have a written statement defining its authority and responsibilities, the subjects within its purview, and other such administrative provisions as are required for its effective functioning. Minutes of all meetings and records of all formal actions of the Committee shall be kept.

e) A quorum shall consist of not less than a majority of the full Committee and shall include all ex officio members.

f) Any permissive action of the Committee requires affirmative vote of the University Radiation Safety Officer as well as a majority vote of the members present.

g) The Committee shall meet a minimum of two times a year. Additional meetings may be called by any member, and the Committee may be polled in lieu of a meeting. Such a poll shall constitute Committee action subject to the same requirements as for an actual meeting.

6.3 Procedures a ) Written procedures, reviewed and approved by the Reactor Safeguards Committee, shall be followed for the activities listed below. The procedures shall be adequate to K-State Reactor TS-45 Original (4/14)

TECHNICAL SPECIFICATIONS assure the safety of the reactor, persons within the Laboratory, and the public, but should not preclude the use of independent judgment and action should the situation require it. The activities are:

1. Startup, operation, and shutdown of the reactor, including (a) startup checkout procedures to test the reactor instrumentation and safety systems, area monitors, and continuous air monitors, (b) prohibition of routine operations with failed (or leaking) fuel except to find leaking elements, and (b) shutdown procedures to assure that the reactor is secured before OPERATING personnel go off duty.

2. Installation or removal of fuel elements, control rods, and other core components that significantly affect reactivity or reactor safety.

3. Preventive or corrective maintenance activities which could have a significant effect on the safety of the reactor or personnel.

4. Periodic inspection, testing or calibration of auxiliary systems or instrumentation that relate to reactor operation.

b) Substantive changes in the above procedures shall be made only with the approval of the Reactor Safeguards Committee, and shall be issued to the OPERATING personnel in written form. The Nuclear Reactor Facility Manager may make temporary changes that do not change the original intent. The change and the reasons thereof shall be noted in the log book, and shall be subsequently reviewed by the Reactor Safeguards Committee.

c) Determination as to whether a proposed activity in categories (1), (2) and (3) in Section 6.2b above does or does not have a significant safety effect and therefore does or does not require approved written procedures shall require the concurrence of I. the Nuclear Reactor Facility Manager, and

2. at least one other member of the Reactor Safeguards Committee, to be selected for relevant expertise by the Nuclear Reactor Facility Manager. If the Manager and the Committee member disagree, or if in their judgment the case warrants it, the proposal shall be submitted to the full Committee, and

3. the University Radiation Safety Officer, or his/her deputy, who may withhold agreement until approval by the University Radiation Safety Committee is obtained.

The Rector Safeguards Committee shall subsequently review determinations that written procedures are not required. The time at which determinations are made, and the review and approval of written procedures, if required, are carried out, shall be a reasonable interval before the proposed activity is to be undertaken.

d) Determination that a proposed change in the facility does or does not have a significant safety effect and therefore does or does not require review and approval by the full Reactor Safeguards Committee shall be made in the same manner as for proposed activities under (c) above.

K-State Reactor TS-46 Original (4/14)

TECHNICAL SPECIFICATIONS 6.4 Review of Proposals for Experiments a) All proposals for new experiments involving the reactor shall be reviewed with respect to safety in accordance with the procedures in (b) below and on the basis of criteria in (c) below.

b) Procedures:

1. Proposed reactor operations by an experimenter are reviewed by the Reactor Supervisor, who may determine that the operation is described by a previously approved EXPERIMENT or procedure. If the Reactor Supervisor determines that the proposed operation has not been approved by the Reactor Safeguards Committee, the experimenter shall describe the proposed EXPERIMENT in written form in sufficient detail for consideration of safety aspects. If potentially hazardous operations are involved, proposed procedures and safety measures including protective and monitoring equipment shall be described.

2. If the experimenter is a student, approval by his/her research supervisor is required. If the experimenter is a staff or faculty member, his/her own signature is sufficient.

3. The proposal is then to be submitted to the Reactor Safeguards Committee for consideration and approval. The Committee may find that the experiment, or portions thereof, may only be performed in the presence of the University Radiation Safety Officer or Deputy thereto.

4. The scope of the EXPERIMENT and the procedures and safety measures as described in the approved proposal, Including any amendments or conditions added by those reviewing and approving it, shall be binding on the experimenter and the OPERATING personnel. Minor deviations shall be allowed only in the manner described in Section 6 above. Recorded affirmative votes on proposed new or revised experiments or procedures must indicated that the Committee determines that the proposed actions do not involve changes in the facility as designed, changes in Technical Specifications, changes that under the guidance of 10 CFR 50.59 require prior approval of the NRC, and could be taken without endangering the health and safety of workers or the public or constituting a significant hazard to the integrity of the reactor core.

5. Transmission to the Reactor Supervisor for scheduling.

c) Criteria that shall be met before approval can be granted shall include:

1. The EXPERIMENT must meet the applicable Limiting Conditions for Operation and Design Description specifications.

2. It must not involve violation of any condition of the facility license or of Federal, State, University, or Facility regulations and procedures.

3. The conduct of tests or experiments not described in the safety analysis report (as updated) must be evaluated in accordance with 10 CFR 50.59 to determine if the test K-State Reactor TS-47 Original (4/14)

TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90.

4. In the safety review the basic criterion is that there shall be no hazard to the reactor, personnel or public. The review SHALL determine that there is reasonable assurance that the experiment can be performed with no significant risk to the safety of the reactor, personnel or the public.

6.5 Emergency Plan and Procedures An emergency plan shall be established and followed in accordance with NRC regulations. The plan shall be reviewed and approved by the Reactor Safeguards Committee prior to its submission to the NRC. In addition, emergency procedures that have been reviewed and approved by the Reactor Safeguards Committee shall be established to cover all foreseeable emergency conditions potentially hazardous to persons within the Laboratory or to the public, including, but not limited to, those involving an uncontrolled reactor excursion or an uncontrolled release of radioactivity.

6.6 Operator Requalification An operator requalification program shall be established and followed in accordance with NRC regulations.

6.7 Physical Security Plan Administrative controls for protection of the reactor plant shall be established and followed in accordance with NRC regulations.

6.8 Action To Be Taken In The Event A Safety Limit Is Exceeded In the event a safety limit is exceeded:

a) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

b) An immediate report of the occurrence shall be made to the Chair of the Reactor Safeguards Committee, and reports shall be made to the NRC in accordance with Section 6.11 of these specifications.

c) A report shall be made to 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 Reactor Safeguards Committee for review, and a suitable similar report submitted to the NRC when authorization to resume operation of the reactor is sought.

6.9 Action To Be Taken In The Event Of A Reportable Occurrence a ) A reportable occurrence is any of the following conditions:

TS-48 Original (4/14)

K-State Reactor K-State Reactor TS-48 Original (4/14)

TECHNICAL SPECIFICATIONS I. any actual safety system setting less conservative than specified in Section 2.2, Limiting Safety System Settings;

2. VIOLATION OF SL, LSSS OR LCO; NOTES Violation of an LSSS or LCO occurs throughfailure to comply with an "Action" statement when "Specification" is not met;failure to comply with the "Specification" is not by itselfa violation.

Surveillance Requirements must be met for all equipment/components/conditionsto be consideredoperable.

Failureto perform a surveillancewithin the requiredtime interval or failure of a surveillance test shall result in the /component/condition being inoperable

3. incidents or conditions that prevented or could have prevented the performance of the intended safety functions of an engineered safety feature or the REACTOR SAFETY SYSTEM;

4. release of fission products from the fuel that cause airborne contamination levels in the reactor bay to exceed IOCFR20 limits for releases to unrestricted areas;

5. an uncontrolled or unanticipated change in reactivity greater than $1.00;

6. an observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy has caused the existence or development of an unsafe condition in connection with the operation of the reactor;

7. an uncontrolled or unanticipated release of radioactivity.

b) In the event of a reportable occurrence, the following actions shall be taken:

1. The reactor shall be shut down at once. The Reactor Supervisor shall be notified and corrective action taken before operations are resumed; the decision to resume shall require approval following the procedures in Section 6.3.

2. A report shall be made to include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safeguards Committee for review.

3. A report shall be submitted to the NRC in accordance with Section 6.11 of these specifications.

6.10 Plant Operating Records a ) In addition to the requirements of applicable regulations, in 10 CFR 20 and 50, records and logs shall be prepared and retained for a period of at least 5 years for the following items as a minimum.

K-State Reactor TS-49 Original (4/14)

TECHNICAL SPECIFICATIONS

1. normal plant operation, including power levels;

3. principal maintenance activities;

4. reportable occurrences;

5. equipment and component surveillance activities;

6. experiments performed with the reactor;

7. all emergency reactor scrams, including reasons for emergency shutdowns.

b) The following records shall be maintained for the life of the facility:

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

2. offsite environmental monitoring surveys;

3. fuel inventories and transfers;

4. facility radiation and contamination surveys;

5. radiation exposures for all personnel;

6. updated, corrected, and as-built drawings of the facility.

6.11 Reporting Requirements All written reports shall be sent within the prescribed interval to the United States Nuclear Regulatory Commission, Washington, D.C., 20555, Attn: Document Control Desk.

In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the US. Nuclear Regulatory Commission (NRC) as follows:

a ) A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and fax or electronic mail to the NRC Operations Center and the USNRC Region IV of;

1. any accidental release of radioactivity above permissible limits in unrestricted areas, whether or not the release resulted in property damage, personal injury, or exposure;

2. any violation of a safety limit;

3. any reportable occurrences as defined in Section 6.9 of these specifications.

b) A report within 10 days in writing of:

1. any accidental release of radioactivity above permissible limits in unrestricted areas, whether or not the release resulted in property damage, personal injury or exposure; the written report (and, to the extent possible, the preliminary telephone and K-State Reactor TS-50 Original (4/14)

TECHNICAL SPECIFICATIONS telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent recurrence of the event;

2. any violation of a safety limit;

3. any reportable occurrence as defined in Section 1.1 of these specifications.

c) A report within 30 days in writing of:

1. any significant variation of a MEASURED VALUE from a corresponding predicted or previously MEASURED VALUE of safety-connected OPERATING characteristics occurring during operation of the reactor;

2. any significant change in the transient or accident analysis as described in the Safety Analysis Report.

3. a change in personnel for the Department of Mechanical and Nuclear Engineering Chair, or a change in reactor manager d) A report within 60 days after criticality of the reactor in writing to the US Nuclear Regulatory Commission, resulting from a receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level or the installation of a new core, describing the MEASURED VALUE of the OPERATING conditions or characteristics of the reactor under the new conditions.

e) A routine report in writing to the US. Nuclear Regulatory Commission within 60 days after completion of the first calendar year of OPERATING and at intervals not to exceed 12 months, thereafter, providing the following information:

1. a brief narrative summary of OPERATING experience (including experiments performed), changes in facility design, performance characteristics, and OPERATING procedures related to reactor safety occurring during the reporting period; and results of surveillance tests and inspections;

2. a tabulation showing the energy generated by the reactor (in megawatt-hours);

3. the number of emergency shutdowns and inadvertent scrams, including the reasons thereof and corrective action, if any, taken;

4. discussion of the major maintenance operations performed during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required;

5. a summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of 10 CFR 50.59;

6. 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 before the point of such release or discharge;

7. a description of any environmental surveys performed outside the facility; TS-51 Original (4/14)

Reactor K-State Reactor TS-51 Original (4/14)

TECHNICAL SPECIFICATIONS

8. a summary of radiation exposures received by facility personnel and visitors, including the dates and time of significant exposure, and a brief summary of the results of radiation and contamination surveys performed within the facility.

TS-52 Original (4/14)

Reactor K-State Reactor TS-52 Original (4/14)

TECHNICAL SPECIFICATIONS Table of Contents

1. D EFIN ITIO N S ................................................................................................................. TS-1

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Elem ent Tem perature Safety Lim it ..................................................................... TS-8 2.1.1. Applicability ...................................................................................................... TS-8 2.1.2. Objective ............................................................................................................. TS-8 2.1.3. Specification ...................................................................................................... TS-8 2.1.4. Actions ................................................................................................................ TS-8 2.1.5. Basis ................................................................................................................... TS-8 2.2 Lim iting Safety System Settings ............................................................................... TS-10 2.2.1. A pplicability ..................................................................................................... TS-10 2.2.3. Objective ........................................................................................................... TS-10 2.2.4. Specification ...................................................................................................... TS-10 2.2.5. Actions .............................................................................................................. TS-10 2.2.6. Basis ................................................................................................................. TS-10

3. LIMITING CONDITIONS FOR OPERATIONS ........................................................ TS-1 1 3.1 CO RE REACTIV ITY ................................................................................................ TS-11 3.1.1. A pplicability ..................................................................................................... TS-1 1 3.1.3. Objective ........................................................................................................... TS-11 3.1.4. Specification ...................................................................................................... TS-I 1 3.1.5. Actions .............................................................................................................. TS-12 3.1.6. Basis ................................................................................................................. TS-13 3.2 PU LSED M O D E O PERA TIO N S .............................................................................. TS-13 3.2.1. A pplicability ..................................................................................................... TS-13 3.2.3. Objective . . . ............................................

. . . . . ...... TS-13 3.2.4. Specification ...................................................................................................... TS-13 3.2.5. Actions .............................................................................................................. TS-13 3.2.6. Basis ................................................................................................................. TS-13 3.3 M EA SU RIN G CH ANN ELS ..................................................................................... TS-14 3.3.1. A pplicability ..................................................................................................... TS-14 3.3.3. Objective ........................................................................................................... TS-14 3.3.4. Specification ...................................................................................................... TS-14 3.3.5. Actions .............................................................................................................. TS-14 3.3.6. Bases ................................................................................................................. TS-16 3.4. SAFETY CHANNEL AND CONTROL ROD OPERABILITY .............................. TS-18 3.4.1. A pplicability ..................................................................................................... TS-18 3.4.3. Objective ........................................................................................................... TS-18 3.4.4. Specification ...................................................................................................... TS- 18 3.4.5. Actions .............................................................................................................. TS-18 3.4.6. Basis ................................................................................................................. TS-19 3.5 G A SEO U S EFFLU EN T CO N TRO L ........................................................................ TS-20 3.5.1. A pplicability ..................................................................................................... TS-20 3.5.3. Objective ........................................................................................................... TS-20 3.5.4. Specification ...................................................................................................... TS-20 3.5.5. Actions .............................................................................................................. TS-20 3.5.6. Basis ................................................................................................................. TS-21 3.6 LIM ITA TIO N S ON EXPERIM ENTS .......................................................................... TS-22 3.6.1. A pplicability ..................................................................................................... TS-22 3.6.3. Objective ........................................................................................................... TS-22 I K-State Reactor TS-1 Original (9/074/14)

TECHNICAL SPECIFICATIONS 3.6.4. Specification ...................................................................................................... TS-22 3.6.5. Actions .............................................................................................................. TS-22 3.6.6. Basis ................................................................................................................. TS-23 3.7 FUEL INTEG RITY .................................................................................................. TS-24 3.7.1. Applicability ..................................................................................................... TS-24 3.7.3. Objective ........................................................................................................... TS-24 3.7.4. Specification ...................................................................................................... TS-24 3.7.5. Actions .............................................................................................................. TS-24 3.7.6. Basis ................................................................................................................. TS-24 3.8 REACTOR POOL W ATER ......................................................................................... TS-25 3.8.1. Applicability ..................................................................................................... TS-25 3.8.3. Objective ........................................................................................................... TS-25 3.8.4. Specification ...................................................................................................... TS-25 3.8.5. Actions .............................................................................................................. TS-25 3.8.6. Basis ................................................................................................................. TS-26 3.9 M aintenance Retest Requirements ................................................................................ TS-27 3.9.1. Applicability ..................................................................................................... TS-27 3.9.3. Objective ........................................................................................................... TS-27 3.9.4. Specification ...................................................................................................... TS-27 3.9.5. Actions .............................................................................................................. TS-27 3.9.6. Basis ................................................................................................................. TS-27

4. SURVIELLANCES ......................................................................................................... TS-28 4.1 CORE REACTIV ITY ................................................................................................ TS-28 4.1 1. Objective ........................................................................................................... TS-28 4.1.2. Specification ...................................................................................................... TS-28 4.1.3. Basis ................................................................................................................. TS-28 4.2 PULSE MODE ............................................................................................................. TS-29 4.2.1. Objective ........................................................................................................... TS-29 4.2.2. Specification ...................................................................................................... TS-29 4.2.3. Basis ................................................................................................................. TS-29 4.3 M EASURING CHANN ELS ..................................................................................... TS-30 4.3.1. Objective ........................................................................................................... TS-30 4.3.2. Specification ...................................................................................................... TS-30 4.3.3. Basis ................................................................................................................. TS-30 4.4 SAFETY CHANNEL AND CONTROL ROD OPERABILITY .............................. TS-31 4.4.1. Objective ........................................................................................................... TS-31 4.4.2. Specification ...................................................................................................... TS-31 4.4.3. Basis ................................................................................................................. TS-32 4.5 GASEOUS EFFLUENT CONTROL ........................................................................ TS-33 4.5.1. Objective ........................................................................................................... TS-33 4.5.2. Specification ...................................................................................................... TS-33 4.5.3. Basis ................................................................................................................. TS-33 4.6 LIM ITATION S ON EXPERIM ENTS ....................................................................... TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis ................................................................................................................. TS-34 4.7 FUEL IN TEGRITY ................................................................................................... TS-35 4.7.1. Objective ........................................................................................................... TS-35 4.7.2. Specification ...................................................................................................... TS-35 4.7.3. Basis ................................................................................................................. TS-35 4.8 REACTOR POOL W ATER ...................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor TS-2 Original (9/O74j/14)

TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36 4.8.3. Basis ................................................................................................................. TS-36 4.9 M AIN TENANCE RETEST REQ UIREM ENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................................................................... TS-37 4.10.3. Basis ............................................................................................................... TS-37

5. DESIG N FEATURES ...................................................................................................... TS-38 5.1 REACTOR FUEL ...................................................................................................... TS-38 5.1.1. Applicability ..................................................................................................... TS-38 5.1.2. Objective ........................................................................................................... TS-38 5.1.3. Specification ...................................................................................................... TS-38 5.1.4. Basis ................................................................................................................. TS-38 5.2 REACTOR FUEL AND FUELED DEVICES IN STORAGE ................................. TS-38 5.2.1. Applicability ..................................................................................................... TS-38 5.2.2. Objective ........................................................................................................... TS-39 5.2.3. Specification ...................................................................................................... TS-39 5.2.4. Basis ................................................................................................................. TS-39 5.3 REACTOR BUILD IN G ............................................................................................ TS-39 5.3.1. Applicability ..................................................................................................... TS-39 5.3.2. Objective ........................................................................................................... TS-39 5.3.3. Specification ...................................................................................................... TS-39 5.3.4. Basis ................................................................................................................. TS-40 5.4 EXPERIM ENTS ........................................................................................................ TS-40 5.4.1. Applicability ..................................................................................................... TS-40 5.4.2. Objective ........................................................................................................... TS-40 5.4.3. Specification ...................................................................................................... TS-40 5.4.4. Basis ................................................................................................................. TS-41

6. ADM INISTRATIVE CO NTRO LS ................................................................................ TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL ......................... TS-44 6.2 REV IEW AN D AUDIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS ................................................. TS-47 6.5 EM ERG ENCY PLAN AN D PROCEDURES ........................................................... TS-48 6.6 OPERATOR REQUALIFICATION ......................................................................... TS-48 6.7 PHYSICA L SECURITY PLAN ................................................................................ TS-48 6.8 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS VIOLATED .... TS-48 6.9 ACTION TO BE TAKEN IN THE EVENT OF A REPO RTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERA TING RECORDS ........................................................................... TS-49 6.11 REPORTIN G REQ UIREM ENTS ........................................................... TS-50 I K-State Reactor TS-3 Original (0-4j/14)

TECHNICAL SPECIFICATIONS

1. DEFINITIONS The following frequently used terms are defined to aid in the uniform interpretation of these specifications. Capitalization is used in the body of the Technical Specifications to identify defined terms.

ACTION Actions are steps to be accomplished in the event a required condition identified in a "Specification" section is not met, as stated in the "Condition" column of "Actions."

In using Action Statements, the following guidance applies:

• Where multiple conditions exist in an LCO, actions are linked to the (failure to meet a "Specification") "Condition" by letters and number.

• Where multiple action steps are required to address a condition, COMPLETION TIME for each action is linked to the action by letter and number.

• AND in an Action Statement means all steps need to be performed to complete the action; OR indicates options and alternatives, only one of which needs to be performed to complete the action.

• If a "Condition" exists, the "Action" consists of completing all steps associated with the selected option (if applicable) except where the "Condition" is corrected prior to completion of the steps ANN-UAL 12 months, not to exceed 15 months CHIANNEL A channel calibration is an adjustment of the channel to that its output CALIBRATION responds, with acceptable range and accuracy, to known values of the parameter that the channel measures.

BIENNIAL Every two years, not to exceed a 28 month interval CHANNEL A channel check is a qualitative verification of acceptable performance by CHECK observation of channel behavior. This verification shall include comparison of the channel with expected values, other independent channels, or other methods of measuring the same variable.

CHANNEL TEST A channel test is the introduction of an input signal into a channel to verify that it is operable. A functional test of operability is a channel test.

CONTROL ROD A standard control rod is one having an electric motor drive and scram (STANDARD) capability.

CONTROL ROD A transient rod is one that is pneumatically operated and has scram (TRANSIENT) capability.

DAILY Prior to initial operation each day (when the reactor is operated), or before TS-4 Original (QI§4114)

I I K-State Reactor TS-4 Original (G/G74/14)

TECHNICAL SPECIFICATIONS an operation extending more than I day ENSURE Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition EXHAUST The air volume in the reactor bay atmosphere between the pool surface and PLENUM the reactor bay exhaust fan EXPERIMENT An EXPERIMENT is (1) any apparatus, device, or material placed in the reactor core region (in an EXPERIMENTAL FACILITY associated with the reactor, or in line with a beam of radiation emanating from the reactor) or (2) any in-core operation designed to measure reactor characteristics.

EXPERIMENTAL Experimental facilities are the beamports, thermal column, pneumatic FACILITY transfer system, central thimble, rotary specimen rack, and the in-core facilities (including non-contiguous single-element positions, and, in the E and F rings, as many as three contiguous fuiel-element positions).

IMMEDIATE Without delay, and not exceeding one hour.

NOTE.:

IMMEDIATE permits activities to restore requiredconditionsfor up to one hour: this does not permit or imply deferring or postponingaction INDEPENDENT INDEPENDENT Experiments are those not connected by a mechanical, EXPERIMENT chemical, or electrical link to another experiment LIMITING CONDITION FOR The lowest functional capability or performance levels of equipment OPERATION required for safe operation of the facility.

(LCO)

LIMITING Settings for automatic protective devices related to those variables having SAFETY SYSTEM significant safety functions. Where a limiting safety system setting is SETTING (LSSS) specified for a variable on which a safety limit placed, the setting shall be chosen so that the automatic protective action will correct the abnormal situation before a safety limit is exceeded.

MEASURED The measured value of a parameter is the value as it appears at the output VALUE of a MEASURING CHANNEL.

MEASURING A MEASURING CHANNEL is the combination of sensor, lines, CHANNEL amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.

MOVABLE A MOVABLE EXPERIMENT is one that may be moved into, out-of or EXPERIMENT near the reactor while the reactor is OPERATING.

NONSECURED NONSECURED Experiments are these that should not move while the EXPERIMENT reactor is OPERATING, but are held in place with less restraint than a secured experiment.

TS-5 Original (9/04J1A)

I I K-State Reactor TS-5 Original (9/974/14)

TECHNICAL SPECIFICATIONS OPERABLE A system or component is OPERABLE when it is capable of performing its intended function in a normal manner OPERATING A system or component is OPERATING when it is performing its intended function in a normal manner.

PULSE MODE The reactor is in the PULSE MODE when the reactor mode selection switch is in the pulse position and the key switch is in the "on" position.

NOTE.-

In the PULSE MODE, reactorpower may be increasedon a periodof much less than I second by motion of the transientcontrolrod.

REACTOR The REACTOR SAFETY SYSTEM is that combination of MEASURING SAFETY SYSTEM CHANNELS and associated circuitry that is designed to initiate reactor scram or that provides information that requires manual protective action to be initiated.

REACTOR The reactor is secured when the conditions of either item (1) or item (2) are SECURED MODE satisfied:

(1) There is insufficient moderator or insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderation and reflection (2) All of the following:

a. The console key is it the OFF position and the key is removed from the lock

b. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives (unless the drive is physically decoupled from the control rod)

c. No experiments are being moved or serviced that have, on movement, a reactivity worth greater than $1.00 REACTOR The reactor is shutdown if it is subcritical by at least $1.00 in the SHUTDOWN REFERENCE CORE CONDITION with the reactivity worth of all experiments included.

RING A ring is one of the five concentric bands of fuel elements surrounding the central opening (thimble) of the core. The letters B through F, with the letter B used to designate the innermost ring, REFERENCE The condition of the core when it is at ambient temperature (cold) and the CORE reactivity worth of xenon is negligible (<$0.30)

CONDITION SAFETY A safety channel is a MEASURING CHANNEL in the REACTOR CHANNEL SAFETY SYSTEM SECURED A secured EXPERIMENT is an EXPERIMENT held firmly in place by a EXPERIMENT mechanical device or by gravity providing that the weight of the EXPERIMENT is such that it cannot be moved by force of less than 60 lb.

TS-6 Original (Q104j14)

I IK-State Reactor TS-6 Original (WG74/14)

TECHNICAL SPECIFICATIONS SECURED EXPERIMENT A secured EXPERIMENT with movable parts is one that contains parts WITH MOVABLE that are intended to be moved while the reactor is OPERATING.

PARTS SHALL Indicates specified action is required/(not to be performed)

(SHALL NOT)

SEMIANNUAL Every six months, with intervals not greater than 8 months SHUTDOWN The shutdown margin is the minimum shutdown reactivity necessary to MARGIN provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating condition, and that the reactor will remain subcritical without further operator action STANDARD THERMOCOUPLE A standard thermocouple fuel element is stainless steel clad fuel element FUEL ELEMENT containing three sheathed thermocouples embedded in the fuel element.

STEADY-STATE The reactor is in the steady-state mode when the reactor mode selector MODE switch is in either the manual or automatic position and the key switch is in the "on" position.

TECHNICAL A violation of a Safety Limit occurs when the Safety Limit value is SPECIFICATION exceeded.

VIOLATION A violation of a Limiting Safety System Setting or Limiting Condition for Operation) occurs when a "Condition" exists which does not meet a "Specification" and the corresponding "Action" has not been met within the required "Completion Time."

If the "Action" statement of an LSSS or LCO is completed or the "Specification" is restored within the prescribed "Completion Time," a violation has not occurred.

NOTE "Condition," "Specification," "Action, "and "Completion Time" refer to applicabletitles of sections in individual Technical Specifications TS-7 Original (9/074114)

I I K-State Reactor TS-7 Original (9/07L/14)

TECHNICAL SPECIFICATIONS

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Fuel Element Temperature Safety Limit 2.1.1 Applicability This specification applies when the reactor in STEADY STATE MODE and the PULSE MODE.

2.1.2 Objective This SAFETY LIMIT ensures fuel element cladding integrity 2.1.3 Specification (1) Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed I11501C.

(2) Steady state fuel temperature shall not exceed 750'C.

2.1.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A Stainless steel clad, high- A.1 Establish SHUTDOWN A.] IMMEDIATE hydride fuel element condition temperature exceeds 1 150°C.

OR AND Fuel temperature exceeds 750'C in steady state A.2 Report per Section 6.8 A.2 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> conditions 2.1.5 Bases Safety Analysis Report, Section 3.5.1 (Fuel System) identifies design and operating constraints for TRIGA fuel that will ensure cladding integrity is not challenged.

NUREG 1282 identifies the safety limit for the high-hydride (ZrH,. 7) fuel elements with stainless steel cladding based on the stress in the cladding (resulting from the hydrogen pressure from the dissociation of the zirconium hydride). This stress will remain below the yield strength of the stainless steel cladding with fuel temperatures below 1,1501C. A change in yield strength occurs for stainless steel cladding temperatures of 5001C, but there is no scenario for fuel cladding to achieve 500 0 C while submerged; consequently the safety limit during reactor operations is 1,150 0 C.

TS-8 Original (9/074114)

I IK-State Reactor Reactor TS-8 Original (9M74/14)

TECHNICAL SPECIFICATIONS Therefore, the important process variable for a TRIGA reactor is the fuel element temperature.

This parameter is well suited as a single specification, and it is readily measured. During operation, fission product gases and dissociation of the hydrogen and zirconium builds up gas inventory in internal components and spaces of the fuel elements. Fuel temperature acting on these gases controls fuel element internal pressure. Limiting the maximum temperature prevents excessive internal pressures that could be generated by heating these gases.

Fuel growth and deformation can occur during normal operations, as described in General Atomics technical report E-1 17-833. Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients. Operating with maximum long-term, steady state fuel temperature of 750'C does not have significant time- and temperature-dependent fuel growth.

TS-9 Original (9/04114)

I I K-State Reactor TS-9 Original (QIG74ý/14)

TECHNICAL SPECIFICATIONS 2.2 Limiting Safety System Settings (LSSS) 2.2.1 Applicability This specification applies when the reactor in STEADY STATE MODE 2.2.2 Objective The objective of this specification is to ensure the safety limit is not exceeded.

2.2.3 Specifications (1) Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. 1 Reduce power to less than A.1 IMMEDIATE 1,250 kW (th)

A. Steady state power level OR exceeds 1,250 kW (th)

A.2. Establish REACTOR SHUTDOWN condition 2.2.5 Bases Analysis in Chapter 4 demonstrates that if operating thermal (th) power is 1,250 kW, the maximum steady state fuel temperature is less than the safety limit for steady state operations by a large margin. For normal pool temperature, calculations in Chapter 4 demonstrate that the heat flux of the hottest area of the fuel rod generating the highest power level in the core during operations is less than the critical heat flux by a large margin up to the maximum permitted cooling temperatures; margin remains even at temperatures approaching bulk boiling for atmospheric conditions. Therefore, steady state operations at a maximum of 1,250 kW meet requirements for safe operation with respect to maximum fuel temperature and thermal hydraulics by a wide margin. Steady state operation of 1,250 kW was assumed in analyzing the loss of cooling and maximum hypothetical accidents. The analysis assumptions are protected by assuring that the maximum steady state operating power level is 1,250 kW.

In 1968 the reactor was licensed to operate at 250 kW with a minimum reactor safety system scram set point required by Technical Specifications at 110% of rated full power, with the scram set point set conservatively at 104%. In 1993 the original TRIGA power level channels were replaced with more reliable, solid state instrumentation. The actual safety system setting will be chosen to ensure that a scram will occur at a level that does not exceed 1,250 kW.

TS-1O Original (9104114)

I IK-State Reactor TS-10 Original (91074/14)

TECHNICAL SPECIFICATIONS

3. Limiting Conditions for Operation (LCO) 3.1 Core Reactivity 3.1.1 Applicability These specifications are required prior to entering STEADY STATE MODE or PULSING MODE in OPERATING conditions; reactivity limits on experiments are specified in Section 3.8.

3.1.2 Objective This LCO ensures the reactivity control system is OPERABLE, and that an accidental or inadvertent pulse does not result in exceeding the safety limit.

3.1.3 Specification The maximum available core reactivity (excess reactivity) with all control rods fully withdrawn is less than $4.00 when:

(1) 1. REFERENCE CORE CONDITIONS exists

2. No experiments with net negative reactivity worth are in place The reactor is capable of being made subcritical by a SHUTDOWN MARGIN more than

$0.50 under REFERENCE CORE CONDITIONS and under the following conditions:

(2) 1. The highest worth control rod is fully withdrawn

2. The highest worth NONSECURED EXPERIMENT is in its most positive reactive state, and each SECURED EXPERIMENT with movable parts is in its most reactive state.

3.1.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.1 ENSURE REACTOR A.] IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meet LCO operations TS-11 Original (9IO~4ji4)

I I K-State Reactor TS-1 1 Original (OW4,/14)

TECHNICAL SPECIFICATIONS B. l.a ENSURE control rods B.] IMMEDIATE fully inserted AND B.l.b Secure electrical power to the control rod circuits B. The reactor is not subcritical by more than AND

$0.50 under specified conditions B. .c Secure all work on in- B.2 Prior to continued core experiments or operations installed control rod drives AND B.2 Configure reactor to meet LCO 3.1.5 Bases The value for excess reactivity was used in establishing core conditions for calculations (Table 13.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis. Since the fundamental protection for the KSU reactor is the maximum power level and fuel temperature that can be achieved with the available positive core reactivity, experiments with positive reactivity are included in determining excess reactivity. Since experiments with negative reactivity will increase available reactivity if they are removed during operation, they are not credited in determining excess reactivity.

Analysis (Chapter 13) shows fuel temperature will not exceed 1,150 0 C for the stainless-steel-clad fuel in the event of inadvertent or accidental pulsing of the reactor. Section 13.2 demonstrates that a $3.00 reactivity insertion from critical, zero power conditions leads to maximum fuel temperature of 746°C, while a $1.00 reactivity insertion from a worst-case steady state operation at 107 kW leads to a maximum fuel temperature of 869°C, well below the safety limit.

The limiting SHUTDOWN MARGIN is necessary so that the reactor can be shut down from any operating condition, and will remain shut down after cool down and xenon decay, even if one control rod (including the transient control rod) should remain in the fully withdrawn position.

TS-12 Original (910-74/14)

I IK-State Reactor K-State Reactor TS-12 Original (WG ý/14)

TECHNICAL SPECIFICATIONS 3.2 PULSED MODE Operations 3.2.1 Applicability These specifications apply to operation of the reactor in the PULSE MODE.

3.2.2 Objective This Limiting Condition for Operation prevents fuel temperature safety limit from being exceeded during PULSE MODE operation.

3.2.3 Specification The transient rod drive is positioned for reactivity insertion (upon withdrawal) less than or

()equal to $3.00 3.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I Position the transient rod drive A. 1 IMMEDIATE A. With all stainless steel clad for pulse rod worth less than fuel elements, the worth of or equal to $3.00 the pulse rod in the OR transient rod drive position OR is greater than $3.00 in the PULSE MODE A.2 Place reactor in STEADY A.2 IMMEDIATE STATE MODE 3.2.5 Bases The value for pulsed reactivity with all stainless steel elements in the core was used in establishing core conditions for calculations (Table 13.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis.

TS-13 Original (9IO~4/14)

I I K-State Reactor TS-1 3 Original (9/G74/14)

TECHNICAL SPECIFICATIONS 3.3 MEASURING CHANNELS 3.3.1 Applicability This specification applies to the reactor MEASURING CHANNELS during STEADY STATE MODE and PULSE MODE operations.

3.3.2 Objective The objective is to require that sufficient information is available to the operator to ensure safe operation of the reactor 3.3.3 Specifications (1) The MEASURING CHANNELS specified in TABLE I SHALL be OPERATING (2) The neutron count rate on the startup channel is greater than the minimum detector sensitivity TABLE 1: MINIMUM MEASURING CHANNEL COMPLEMENT Minimum Number Operable MEASURING CHANNEL STEADY STATE PULSE MODE MODE Reactor power level'i] 2 1 Primary Pool Water Temperature 1 1 Reactor Bay Differential Pressure 1 1 Fuel Temperature 1 1 22 foot Area radiation monitor 1 1 0 or 12 foot Area monitor 1 1 Continuous air radiation monitor[21 1 1 EXHAUST PLENUM radiation monitor[21 1 1 NOTE[I]: One "Startup Channel" required to have range that indicates <10 W NOTE[2]: High-level alarms audible in the control room may be used 3.3.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I Reactor power channels A.1.1 Restore channel to operation A. 1.1 IMMEDIATE not OPERATING (min 2 OR for STEADY STATE, I A. 1.2 ENSURE reactor is A.1.2 IMMEDIATE PULSE MODE) SHUTDOWN TS-14 Original (Q104/14)

I I K-State Reactor TS-14 Original (9/074/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less AND A.2. IMMEDIATE than 90% of required operating value A.2.2 Enter REACTOR SECURED mode B. Primary water temperature, B. I Restore channel to operation A.1 IMMEDIATE reactor bay differential OR pressure or fuel temperature CHANNEL B.2 ENSURE reactor is A.2 IMMEDIATE not operable SHUTDOWN C.I Restore MEASURING C.I IMMEDIATE CHANNEL OR C.2 ENSURE reactor is shutdown C.2 IMMEDIATE OR C. 22 foot Area radiation monitor OPERTIN is not isC.3 ENSURE personnel are not C.3 IMMEDIATE on the 22 foot level OR C.4 ENSURE personnel on 22 C.4 IMMEDIATE foot level are using portable survey meters to monitor dose rates D.1 Restore MEASURING D.1 IMMEDIATE CHANNEL OR D.2 ENSURE reactor is shutdown D.2 IMMEDIATE OR D. 0 or 12 foot Area monitor is not OPERATING D.3 ENSURE personnel are not in D.3 IMMEDIATE the reactor bay OR D.4 ENSURE personnel entering D.4 IMMEDIATE reactor bay are using portable survey meters to monitor dose rates TS-15 Original (Q1074/14)

I I K-State Reactor TS-15 Original (GiG74/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME E. I Restore MEASURING E.I IMMEDIATE CHANNEL OR E.2 ENSURE reactor is shutdown E.2. IMMEDIATE E. Continuous air radiation OR monitor is not OPERATING E.3.a ENSURE EXHAUST E.3.a. IMMEDIATE PLENUM radiation monitor is OPERATING AND E.3.b Restore MEASURING E.3.b Within 30 days CHANNEL F.I Restore MEASURING F.1 IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown F.2. IMMEDIATE F. Exhaust plenum radiation OR monitor is not OPERATING F.3.a ENSURE continuous air F.3.a. IMMEDIATE radiation monitor is OPERATING AND F.3.b Restore MEASURING F.3.b Within 30 days CHANNEL G.1 Do not perform a reactor G.] IMMEDIATE G. The neutron count rate on startup the startup channel is not OR greater than the minimum G.2 Perform a neutron-source detector sensitivity check on the startup channel G.2 IMMEDIATE prior to startup 3.3.5 Bases Maximum steady state power level is 1,250 kW; neutron detectors measure reactor power level.

Chapter 4 and 13 discuss normal and accident heat removal capabilities. Chapter 7 discusses radiation detection and monitoring systems, and neutron and power level detection systems.

According to General Atomics, detector voltages less than 90% of required operating value do not provide reliable, accurate nuclear instrumentation. Therefore, if operating voltage falls below the minimum value the power level channel is inoperable.

I K-State Reactor TS-16 Original (9/074/14)

TECHNICAL SPECIFICATIONS Primary water temperature indication is required to assure water temperature limits are met, protecting primary cleanup resin integrity. The reactor bay differential pressure indictor is required to control reactor bay atmosphere radioactive contaminants. Fuel temperature indication provides a means of observing that the safety limits are met.

The 22-foot and 0-foot area radiation monitors provide information about radiation hazards in the reactor bay. A loss of reactor pool water (Chapter 13), changes in shielding effectiveness (Chapter H1), and releases of radioactive material to the restricted area (Chapter I I) could cause changes in radiation levels within the reactor bay detectable by these monitors. Portable survey instruments will detect changes in radiation levels.

The air monitors (continuous air- and exhaust plenum radiation-monitor) provide indication of airborne contaminants in the reactor bay prior to discharge of gaseous effluent. Iodine channels provide evidence of fuel element failure. The air monitors provide similar information on independent channels; the continuous air monitor (CAM) has maximum sensitivity to iodine and particulate activity, while the air monitoring system (AMS) has individual channels for radioactive particulate, iodine, noble gas and iodine.

When filters in the air monitoring system begin to load, there are frequent, sporadic trips of the AMS alarms. Although the filters are changed on a regular basis, changing air quality makes these trips difficult to prevent. Short outages of the AMS system have resulted in unnecessary shutdowns, exercising the shutdown mechanisms unnecessarily, creating stressful situations, and preventing the ability to fully discharge the mission of the facility while the CAM also monitors conditions of airborne contamination monitored by the AMS. The AMS detector has failure modes than cannot be corrected on site; AMS failures have caused longer outages at the K-State reactor. The facility has experienced approximately two-week outages, with one week dedicated to testing and troubleshooting and (sometimes) one-week for shipment and repair at the vendor facility.

Permitting operation using a single channel of atmospheric monitoring will reduce unnecessary shutdowns while maintaining the ability to detect abnormal conditions as they develop. Relative indications ensure discharges are routine; abnormal indications trigger investigation or action to prevent the release of radioactive material to the surrounding environment. Ensuring the alternate airborne contamination monitor is functioning during outages of one system provides the contamination monitoring required for detecting abnormal conditions. Limiting the outage for a single unit to a maximum of 30 days ensures radioactive atmospheric contaminants are monitored while permitting maintenance and repair outages on the other system.

Chapter 13 discusses inventories and releases of radioactive material from fuel element failure into the reactor bay, and to the environment. Particulate and noble gas channels monitor more routine discharges. Chapter II and SAR Appendix A discuss routine discharges of radioactive gasses generated from normal operations into the reactor bay and into the environment. Chapter 3 identifies design bases for the confinement and ventilation system.

Chapter 7 discusses air-monitoring systems.

Experience has shown that subcritical multiplication with the neutron source used in the reactor does not provide enough neutron flux to correspond to an indicated power level of 10 Watts.

Therefore an indicated power of 10 Watts or more indicates operating in a potential critical condition, and at least one neutron channel is required with sensitivity at a neutron flux level corresponding to reactor power levels less than 10 Watts ("Startup Channel"). If the indicated neutron level is less than the minimum sensitivity for both the log-wide range and the multirange linear power level channels, a neutron source will be used to determine that at least one of the channels is responding to neutrons to ensure that the channel is functioning prior to startup.

TS-17 Original (Q/Q~/~4)

I K-State Reactor K-State Reactor TS-17 Original (W974,/14)

TECHNICAL SPECIFICATIONS 3.4 Safety Channel and Control Rod Operability 3.4.1 Applicability This specification applies to the reactor MEASURING Channels during STEADY STATE MODE and PULSE MODE operations.

3.4.2 Objective The objectives are to require the minimum number of REACTOR SAFETY SYSTEM channels that must be OPERABLE in order to ensure that the fuel temperature safety limit is not exceeded, and to ensure prompt shutdown in the event of a scram signal.

3.4.3 Specifications (1) The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLE CONTROL RODS (STANDARD) are capable of 90% of full reactivity insertion from the (2) fully withdrawn position in less than 1 sec.

TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum Function Required OPERATING Mode Safety System Channel Number STEADY PULSE Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar 1 Scram YES YES CONTROL ROD Prevent withdrawal of standard NA (STANDARD) position I rods in the PULSE MODE YES interlock Prevent inadvertent pulsing Pulse rod interlock 1 while in STEADY STATE YES NA MODE 3.4.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I Restore channel or interlock AI. IMMEDIATE A. Any required SAFETY to operation SYSTEM CHANNEL or interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN 3.4.5 Bases The power level scram is provided to ensure that reactor operation stays within the licensed limits of 1,250 kW, preventing abnormally high fuel temperature. The power level scram is not credited K-State Reactor TS-18 Original (91074/14)

TECHNICAL SPECIFICATIONS in analysis, but provides defense in depth to assure that the reactor is not operated in conditions beyond the assumptions used in analysis (Table 13.2.1.4).

The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs.

The CONTROL ROD (STANDARD) interlock function is to prevent withdrawing control rods (other than the pulse rod) when the reactor is in the PULSE MODE. This will ensure the reactivity addition rate during a pulse is limited to the reactivity added by the pulse rod.

The pulse rod interlock function prevents air from being applied to the transient rod drive when it is withdrawn while disconnected from the control rod to prevent inadvertent pulses during STEADY STATE MODE operations. The control rod interlock prevents inadvertent pulses which would be likely to exceed the maximum range of the power level instruments configured for steady state operations.

TS-19 Original (91074j1.4)

I I K-State Reactor TS-19 Original (9/074/14)

TECHNICAL SPECIFICATIONS 3.5 Gaseous Effluent Control 3.5.1 Applicability This specification applies to gaseous effluent in STEADY STATE MODE and PULSE MODE.

3.5.2 Objective The objective is to ensure that exposures to the public resulting from gaseous effluents released during normal operations and accident conditions are within limits and ALARA.

3.5.3 Specification (1) The reactor bay ventilation exhaust system SHALL maintain in-leakage to the reactor bay Releases of Ar-41 from the reactor bay exhaust plenum to an unrestricted environment (2) SHALL NOT exceed 30 Ci per year.

3.5.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.] ENSURE reactor is A.1 IMMEDIATE SHUTDOWN OR A.2.a Do not OPERATE in the A.2.a IMMEDIATE PULSE MODE AND A.2.b Secure EXPERIMENT A.2.b IMMEDIATE A. The reactor bay ventilation operations for exhaust system is not EXPERIMENT with failure OPERABLE modes that could result in the release of radioactive gases or aerosols.

A.2.c ENSURE no irradiated fuel A.2.b IMMEDIATE handing AND A.2.d Restore the reactor bay A.2.d Within 30 days ventilation exhaust system to OPERABLE I K-State Reactor TS-20 Original (WG74ý/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME Calculated releases ofAr-41 from the reactor bay exhaust Do not operate. IMMEDIATE plenum exceed 30 Ci per year.

3.5.5 Bases The confinement and ventilation system is described in Section 3.5.4. Routine operations produce radioactive gas, principally Argon 41, in the reactor bay. If the reactor bay ventilation system is secured, SAR Chapter 11 Appendix A demonstrates reactor bay concentration of 0.746 Bq m[l (2.01x 10-5 gCi ml'), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10-3 1 Ci h/mL. Therefore, the reduction in concentration of Argon 41 from operation of the confinement and ventilation system is a defense in depth measure, and not required to assure meeting personnel exposure limits. Consequently, the ventilation system can be secured without causing significant personnel hazard from normal operations. Thirty days for a confinement and ventilation system outage is selected as a reasonable interval to allow major repairs and work to be accomplished, if required. During this interval, experiment activities that might cause airborne radionuclide levels to be elevated are prohibited.

It is shown in Section 13.2.2 of the Safety Analysis Report that, if the reactor were to be operating at full steady-state power, fuel element failure would not occur even if all the reactor tank water were to be lost instantaneously.

Section 13.2.4 addresses the maximum hypothetical fission product inventory release. Using unrealistically conservative assumptions, concentrations for a few nuclides of iodine would be in excess of occupational derived air concentrations for a matter of hours or days. 9°Sr activity available for release from fuel rods previously used at other facilities is estimated to be at most about 4 times the ALI. In either case (radio-iodine or -Sr), there is no credible scenario for accidental inhalation or ingestion of the undiluted nuclides that might be released from a damaged fuel element. Finally, fuel element failure during a fuel handling accident is likely to be observed and mitigated immediately.

SAR Appendix A shows the release of 30 Ci per year of Ar-41 from normal operations would result in less than 10 mrem annual exposure to any person in unrestricted areas.

TS-21 Original (Q~Q-74ii4)

Reactor I K-State Reactor TS-21 Original (W974./14)

TECHNICAL SPECIFICATIONS 3.6 Limitations on Experiments 3.6.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.6.2 Objectives These Limiting Conditions for Operation prevent reactivity excursions that might cause the fuel temperature to exceed the safety limit (with possible resultant damage to the reactor), and the excessive release of radioactive materials in the event of an EXPERIMENT failure 3.6.3 Specifications If all fuel elements are stainless steel clad, the reactivity worth of any individual (1) EXPERIMENT SHALL NOT exceed $2.00 If two or more experiments in the reactor are interrelated so that operation or failure of (2) one can induce reactivity-affecting change in the other(s), the sum of the absolute reactivity of such experiments SHALL NOT exceed $2.00.

Irradiation holders and vials SHALL prevent release of encapsulated material in the reactor pool and core area 3.6.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I ENSURE the reactor is A. 1 IMMEDIATE SHUTDOWN A. INDEPENDENT EXPERIMENT worth is AND greater than $2.00 A.2 Remove the experiment A.2 Prior to continued operations C. I ENSURE the reactor is C. I IMMEDIATE SHUTDOWN C. An irradiation holder or vial AND releases material capable of causing damage to the C.2 Inspect the affected area C.2 Prior to continued reactor fuel or structure into operation the pool or core area AND C.3 Obtain RSC review and C.3 Prior to continued approval operation TS-22 Original (Q107-4114)

I I K-State Reactor TS-22 Original (WG74/14)

TECHNICAL SPECIFICATIONS 3.6.5 Bases Specifications 3.7(1) through 3.7(3) are conservatively chosen based on prior operation at 250 kW to limit reactivity additions to maximum values that are less than an addition which could cause temperature to challenge the safety limit.

Experiments are approved with expectations that there is reasonable assurance the facility will not be damaged during normal or failure conditions. If an irradiation capsule which contains material with potential for challenging the fuel cladding or pool wall, the facility will be inspected to ensure that continued operation is acceptable.

TS-23 Original (9/O~4iiA)

I I K-State Reactor TS-23 Original (9/074/14)

TECHNICAL SPECIFICATIONS 3.7 Fuel Integrity 3.7.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.7.2 Objective The objective is to prevent the use of damaged fuel in the KSU TRIGA reactor.

3.7.3 Specifications Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over (1) manufactured length (2) Fuel elements in the reactor core SHALL NOT be laterally bent more than 1/8 in.

3.7.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. Any fuel element is elongated greater than 1/8 Do not insert the fuel element into in. over manufactured IMMEDIATE length, or bent laterally greater than 1/8 in.

3.7.5 Bases The above limits on the allowable distortion of a fuel element have been shown to correspond to strains that are considerably lower than the strain expected to cause rupture of a fuel element and have been successfidly applied at TRIGA installations. Fuel cladding integrity is important since it represents the only process barrier for fission product release from the TRIGA reactor.

TS-24 Original (~~O-~4ii4)

I I K-State Reactor TS-24 Original (GID74,/14)

TECHNICAL SPECIFICATIONS 3.8 Reactor Pool Water 3.8.1 Applicability This specification applies to operations in STEADY STATE MODE, PULSE MODE, and SECURED MODE.

3.8.2 Objective The objective is to set acceptable limits on the water quality, temperature, conductivity, and level in the reactor pool.

3.8.3 Specifications (1) Water temperature at the exit of the reactor pool SHALL NOT exceed 130'F with flow through the primary cleanup loop (2) Water conductivity SHALL be less than 5 ýtmho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core 3.8.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I ENSURE the reactor is A.1 IMMEDIATE SHUTDOWN AND A. Water temperature at the A.2 Secure flow through the A.2 IMMEDIATE exit of the reactor pool demineralizer exceeds 130'F AND A.3 Reduce water temperature to A.3 IMMEDIATE less than 130'F B.I ENSURE the reactor is B.1 IMMEDIATE SHUTDOWN B. Water conductivity is greater than 5 ýtmho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 _tmho/cm TS-25 Original (QIQ~4ji4)

I I K-State Reactor TS-25 Original (WG74/14)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME C. I ENSURE the reactor is C.1 IMMEDIATE C. Water level above the core SHUTDOWN SHALL be at least 13 ft from the top of the core for AND all operating conditions C.2 Restore water level C.2 IMMEDIATE 3.8.5 Bases The resin used in the mixed bed deionizer limits the water temperature of the reactor pool. Resin in use (as described in Section 5.4) maintains mechanical and chemical integrity at temperatures below 130'F (54.4'C).

Maintaining low water conductivity over a prolonged period prevents possible corrosion, deionizer degradation, or slow leakage of fission products from degraded cladding. Although fuel degradation does not occur over short time intervals, long-term integrity of the fuel is important, and a 4-week interval was selected as an appropriate maximum time for high conductivity.

The top of the core is 16 feet below the top of the primary coolant tank. The lowest suction of primary cooling flow into the forced cooling loop is 3.5 feet below the top of the primary coolant tank (water level is maintained about 6 inches below the top of the tank).

The principle contributor to radiation dose rates at the pool surface is Nitrogen 16 generated in the reactor core and dispersed in the pool. Calculations in Chapter 11 show the pool surface radiation dose rates from Nitrogen 16 with 13 feet of water above the core are acceptable.

For normal pool temperature, calculations in Chapter 4 assuming 16 feet and 13 feet above the core demonstrate that the heat flux of the hottest area of the fuel rod generating the highest power level in the core during operations is less than the critical heat flux by a large margin up to the maximum permitted cooling temperatures; margin remains even at temperatures approaching bulk boiling for atmospheric conditions. Therefore, pool levels greater than 13 feet above the core meet requirements for safe operation with respect to maximum fuel temperature and thermal hydraulics by a wide margin.

Therefore, a minimum pool level of 13 feet above the core is adequate to provide shielding and support the core cooling.

TS-26 Original (Q/074J1A)

I I K-State Reactor TS-26 Original (W074,/14)

TECHNICAL SPECIFICATIONS 3.9 Maintenance Retest Requirements 3.9.1 Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.

3.9.2 Objective The objective is to ensure Technical Specification requirements are met following maintenance that occurs within surveillance test intervals.

3.9.3 Specifications Maintenance activities SHALL NOT change, defeat or alter equipment or systems in a way that prevents the systems or equipment from being OPERABLE or otherwise prevent the systems or equipment from fulfilling the safety basis 3.9.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME Maintenance is performed that has the potential to change a Perform surveillance Prior to continued, setpoint, calibration, flow rate, normal operation in or other parameter that is OR STEADY STATE measured or verified in MODE or PULSE meeting a surveillance or Operate only to perform retest MODE operability requirement 3.9.5 Bases Operation of the K-State reactor will comply with the requirements of Technical Specifications.

This specification ensures that if maintenance might challenge a Technical Specifications requirement, the requirement is verified prior to resumption of normal operations.

TS-27 Original (9/07-4/14)

I IK-State Reactor TS-27 Original (W074,/14)

TECHNICAL SPECIFICATIONS

4. Surveillance Requirements 4.1 Core Reactivity 4.1.1 Objective This surveillance ensures that the minimum SHUTDOWN MARGIN requirements and maximum excess reactivity limits of section 3.1 are met.

4.1.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SHUTDOWN MARGIN Determination SEMIANNUAL SEMIANNUAL Excess Reactivity Determination Following Insertion of experiments with measurable positive reactivity Control Rod Reactivity Worth determination BIENNIAL 4.1.3 Basis Experience has shown verification of the minimum allowed SHUTDOWN MARGIN at the specified frequency is adequate to assure that the limiting safety system setting is met When core reactivity parameters are affected by operations or maintenance, additional activity is required to ensure changes are incorporated in reactivity evaluations.

TS-28 Original (9/07-4/14)

I I K-State Reactor TS-28 Original (9iG74/14)

TECHNICAL SPECIFICATIONS 4.2 PULSE MODE 4.2.1 Objectives The verification that the pulse rod position does not exceed a reactivity value corresponding to

$3.00 assures that the limiting condition for operation is met.

4.2.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ENSURE Transient Pulse Rod position corresponds to reactivity Prior to pulsing operations not greater than $3.00 4.2.3 Basis Verifying pulse rod position corresponds to less than $3.00 ensures that the maximum pulsed reactivity meets the limiting condition for operation.

TS-29 Original (9L074114)

I I K-State Reactor TS-29 Original (9/074/14)

TECHNICAL SPECIFICATIONS 4.3 MEASURING CHANNELS 4.3.1 Objectives Surveillances on MEASURING CHANNELS at specified frequencies ensure instrument problems are identified and corrected before they can affect operations.

4.3.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Reactor power level MEASURING CHANNEL CHANNEL TEST DAILY Calorimetric calibration ANNUAL CHANNEL CHECK high voltage to required power level DAILY instruments Primary pool water temperature CHANNEL CALIBRATION ANNUAL Reactor Bay differential pressure CHANNEL CALIBRATION ANNUAL Fuel temperature CHANNEL CALIBRATION ANNUAL 22 Foot Area radiation monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL Continuous Air Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL EXHAUST PLENUM Radiation Monitor CHANNEL CHECK DAILY CHANNEL CALIBRATION ANNUAL Startup Count Rate DAILY 4.3.3 Basis The DAILY CHANNEL CHECKS will ensure that the SAFETY SYSTEM and MEASURING CHANNELS are operable. The required periodic calibrations and verifications will permit any long-term drift of the channels to be corrected.

TS-30 Original (9/074j.14)

I Reactor I K-State Reactor TS-30 Original (GIG74/14)

TECHNICAL SPECIFICATIONS 4.4 Safety Channel and Control Rod Operability 4.4.1 Objective The objectives of these surveillance requirements are to ensure the REACTOR SAFETY SYSTEM will function as required. Surveillances related to safety system MEASURING CHANNELS ensure appropriate signals are reliably transmitted to the shutdown system; the surveillances in this section ensure the control rod system is capable of providing the necessary actions to respond to these signals.

4.4.2 Specifications SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Manual scram SHALL be tested by releasing partially withdrawn DAILY CONTROL RODS (STANDARD)

CONTROL ROD (STANDARD) drop times SHALL be measured to have a drop time from the fully withdrawn position of less than ANNUAL I sec.

The control rods SHALL be visually inspected for corrosion and BIENNIAL mechanical damage at intervals CONTROL ROD (STANDARD) position interlock functional test SEMIANNUAL Pulse rod interlock functional test SEMIANNUAL On each day that PULSE MODE operation of the reactor is Prior to pulsing operations planned, a functional performance check of the CONTROL ROD each day a pulse is planned (TRANSIENT) system SHALL be performed.

The CONTROL ROD (TRANSIENT) rod drive cylinder and the associated air supply system SHALL be inspected, cleaned, and SEMIANNUAL lubricated, as necessary.

4.4.3 Basis Manual and automatic scrams are not credited in accident analysis, although the systems function to assure long-term safe shutdown conditions. The manual scram and control rod drop timing surveillances are intended to monitor for potential degradation that might interfere with the operation of the control rod systems. The verification of high voltage to the power level monitoring channels assures that the instrument channel providing an overpower trip will function on demand.

The control rod inspections (visual inspections and transient drive system inspections) are similarly intended to identify potential degradation that lead to control rod degradation or inoperability.

A test of the interlock that prevents the pulse rod from coupling to the drive in the state state mode unless the drive is fully down assures that pulses will occur only when in pulsing mode. A I K-State Reactor TS-31 Original (9/074,/14)

TECHNICAL SPECIFICATIONS test of the interlock that prevents standard control rod motion while in the pulse mode assures that the interlock will function as required.

The functional checks of the control rod drive system assure the control rod drive system operates as intended for any pulsing operations. The inspection of the pulse rod mechanism will assure degradation of the pulse rod drive will be detected prior to malfunctions.

TS-32 Original (Q/04/14)

I I K-State Reactor K-State Reactor TS-32 Original (9/074/14)

TECHNICAL SPECIFICATIONS 4.5 Gaseous Effluent Control 4.5.1 Objectives These surveillances ensure that routine releases are normal, and (in conjunction with MEASURING CHANNEL surveillances) that instruments will alert the facility if conditions indicate abnormal releases.

4.5.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL TEST of air monitor ANNUAL Verify negative reactor bay differential pressure DAILY 4.5.3 Basis The continuous air monitor provides indication that levels of radioactive airborne contamination in the reactor bay are normal.

If the reactor bay differential pressure gage indicates a negative pressure, the reactor bay exhaust fan is controlling airflow by directing effluent out of confinement.

TS-33 Original (~~O-74jiA)

I I K-State Reactor TS-33 Original (WQ74/14)

TECHNICAL SPECIFICATIONS 4.6 Limitations on Experiments 4.6.1 Objectives This surveillance ensures that experiments do not have significant negative impact on safety of the public, personnel or the facility.

4.6.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Prior to inserting a new Experiments SHALL be evaluated and approved prior to experiment for purposes implementation. other than determination of reactivity worth Initial insertion of absolute a new Measure and record experiment worth of the EXPERIMENT experiment where value of the estimated worth is greater than valueo heestima te

$0.40). the absolute (where value of the estimated worth is greater than $0.40 4.6.3 Basis These surveillances allow determination that the limits of 3.7 are met.

Experiments with an absolute value of the estimated significant reactivity worth (greater than

$0.40) will be measured to assure that maximum experiment reactivity worths are met. If an absolute value of the estimate indicates less than $0.40 reactivity worth, even a 100% error will result in actual reactivity less than the assumptions used in analysis for inadvertent pulsing at low power operations in the Safety Analysis Report (13.2.3, Case I).

TS-34 Original (Q/§~/'~4)

I I K-State Reactor TS-34 Original (WG74/14)

TECHNICAL SPECIFICATIONS 4.7 Fuel Integrity 4.7.1 Objective The objective is to ensure that the dimensions of the fuel elements remain within acceptable limits.

4.7.2 Applicability This specification applies to the surveillance requirements for the fuel elements in the reactor core.

4.7.3 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 500 pulses of magnitude equal to or less than a pulse insertion of 3.00$

The standard fuel elements SHALL be visually inspected for cor- AND rosion and mechanical damage, and measured for length and bend Following the exceeding of a limited safety system set point with potential for causing degradation B, C, D, E, and F RING elements comprising approximately 1/3 of the core SHALL be visually inspected annually for corrosion and ANNUAL mechanical damage such that the entire core SHALL be inspected at 3-year intervals, but not to exceed 38 months 4.7.4 Basis The most severe stresses induced in the fuel elements result from pulse operation of the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply.

Triennial visual inspection of fuel elements combined with measurements at intervals determined by pulsing as described is considered adequate to identify potential degradation of fuel prior to catastrophic fuel element failure.

TS-35 Original (Q~Q~4/i.4)

I I K-State Reactor TS-35 Original (9WA,/14)

TECHNICAL SPECIFICATIONS 4.8 Reactor Pool Water This specification applies to the water contained in the KSU TRIGA reactor pool.

4.8.1 Objective The objective is to provide surveillance of reactor primary coolant water quality, pool level, temperature and (in conjunction with MEASURING CHANNEL surveillances), and conductivity.

4.8.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify reactor pool water level above the inlet line vacuum breaker DAILY Verify reactor pool water temperature channel operable DAILY DAILY Measure reactor Pool water conductivity At least every 20 days 4.9.3 Bases Surveillance of the reactor pool will ensure that the water level is adequate before reactor operation. Evaporation occurs over longer periods of time, and daily checks are adequate to identify the need for water replacement.

Water temperature must be monitored to ensure that the limit of the deionizer will not be exceeded. A daily check on the instrument prior to reactor operation is adequate to ensure the instrument is operable when it will be needed.

Water conductivity must be checked to ensure that the deionizer is performing properly and to detect any increase in water impurities. A daily check is adequate to verify water quality is appropriate and also to provide data useful in trend analysis. If the reactor is not operated for long periods of time, the requirement for checks at least every 20 days will ensure water quality is maintained in a manner that does not permit fuel degradation.

TS-36 Original (9/07-4/14)

Reactor I K-State Reactor TS-36 Original (91G74/14)

TECHNICAL SPECIFICATIONS 4.9 Maintenance Retest Requirements 4.9.1 Objective The objective is to ensure that a system is OPERABLE within specified limits before being used after maintenance has been performed.

4.9.2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Following maintenance of Evaluate potential for maintenance activities to affect operability systems of equipment and function of equipment required by Technical Specifications required by Technical Specifications Perform surveillance to assure affected function meets Prior to resumption of requirements normal operations 4.9.3 Bases This specification ensures that work on the system or component has been properly carried out and that the system or component has been properly reinstalled or reconnected before reliance for safety is placed on it.

TS-37 Original (9IO~4ji4)

I I K-State Reactor TS-37 Original (WG74/14)

TECHNICAL SPECIFICATIONS

5. Design Features 5.1 Reactor Fuel 5.1.1 Applicability This specification applies to the fuel elements used in the reactor core.

5.1.2 Objective The objective is to ensure 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 mechanical integrity.

5.1.3 Specification (I) The high-hydride fuel element shall contain uranium-zirconium hydride, clad in 0.020 in.

of 304 stainless steel. It shall contain a maximum of 9"012.5 weight percent uranium which has a maximum enrichment of 20%. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom.

(2) For the loading process, the elements shall be placed in a close packed array except for experimental facilities or for single positions occupied by control rods and a neutron startup source.

(3) Up to four elements with greater than 9.0 weight percent uranium may be installed in the core. These elements may only be placed in the E- and F-rings of the core lattice, and may not be adjacent to control rods or water channels.

5.1.4 Bases These types of fuel elements have a long history of successful use in TRIGA reactors.

Calculations show that 12%-load fuel in the E- and F-rings will not exceed the temperature of 8%-load instrumented elements in the B-ring. Additionally the power peaking and fission product inventory assumptions in the SAR will not be challenged by 12% fuel in the E- and F-rings. Local power and temperature peaking effects during pulsing are avoided by prohibiting placement of the 12%-load fuel near water and control rod channels.

TS-38 Original (9/04/14)

I I K-State Reactor TS-38 Original (WG74/14)

TECHNICAL SPECIFICATIONS 5.2 Reactor Fuel and Fueled Devices in Storage 5.2.1 Applicability This specification applies to reactor fuel elements in storage 5.2.2 Objective The objective is to ensure fuel elements or fueled devices in storage are maintained Subcritical in a safe condition.

5.2.3 Specification (1) All fuel elements or fueled devices shall be in a safe, stable geometry (2) The keff of all fuel elements or fueled devices in storage is less than 0.8 (3) Irradiated fuel elements or fueled devices will be stored in an array which will permit sufficient natural convection cooling by air or water such that the fuel element or fueled device will not exceed design values.

5.2.4 Bases This specification is based on American Nuclear Society standard 15. 1, section 5.4.

5.3 Reactor Building 5.3.1 Applicability This specification applies to the building that houses the TRIGA reactor facility.

5.3.2 Objective The objective is to ensure that provisions are made to restrict the amount of release of radioactivity into the environment.

5.3.3 Specification (I) The reactor shall be housed in a closed room designed to restrict leakage when the reactor is in operation, when the facility is unmanned, or when spent fuel is being handled exterior to a cask.

(2) The minimum free volume of the reactor room shall be approximately 144,000 cubic feet.

I K-State Reactor TS-39 Original (910g4/14)

TECHNICAL SPECIFICATIONS (3) The building shall be equipped with a ventilation system capable of exhausting air or other gases from the reactor room at a minimum of 30 ft. above ground level.

5.3.4 Bases To control the escape of gaseous effluent, the reactor room contains no windows that can be opened. The room air is exhausted through an independent exhaust system, and discharged at roof level to provide dilution.

5.4 Experiments 5.4.1 Applicability This specification applies to the design of experiments.

5.4.2 Objective The objective is to ensure that experiments are designed to meet criteria.

5.4.3 Specifications (1) EXPERIMENT with a design reactivity worth greater than $1.00 SHALL be securely fastened (as defined in Section 1,Secured Experiment).

(2) Design shall ensure that failure of an EXPERIMENT SHALL NOT lead to a direct failure of a fuel element or of other experiments that could result in a measurable increase in reactivity or a measurable release of radioactivity due to the associated failure.

(3) EXPERIMENT SHALL be designed so that it does not cause bulk boiling of core water (4) EXPERIMENT design SHALL ensure no interference with control rods or shadowing of reactor control instrumentation.

(5) EXPERIMENT design shall minimize the potential for industrial hazards, such as fire or the release of hazardous and toxic materials.

(6) Each fueled experiment shall be limited such that the total inventory of iodine isotopes 131 through 135 in the experiment is not greater than 5 millicuries except as the fueled experiment is a standard TRIGA instrumented element in which instance the iodine inventory limit is removed.

(7) Where the possibility exists that the failure of an EXPERIMENT (except fueled EXPERIMENTS) could release radioactive gases or aerosols to the reactor bay or atmosphere, the quantity and type of material shall be limited such that the airborne concentration of radioactivity averaged over a year will not exceed the limits of Table II of Appendix B of 10 CFR Part 20 assuming 100% of the gases or aerosols escape.

I K-State Reactor TS-40 Original (9107L/14)

TECHNICAL SPECIFICATIONS (8) The following assumptions shall be used in experiment design:

a. If effluents from an experimental facility exhaust through a hold-up tank which closes automatically at a high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

b. If effluents from an experimental facility exhaust through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of the aerosols produced will escape.

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

(9) Use of explosive solid or liquid material with a National Fire Protection Association Reactivity (Stability) index of 2, 3, or 4 in the reactor pool or biological shielding SHALL NOT exceed the equivalent of 25 milligrams of TNT without prior NRC approval.

5.4.4 Basis Designing the experiment to reactivity and thermal-hydraulic conditions ensure that the experiment is not capable of breaching fission product barriers or interfering with the control systems (interferences from other - than reactivity - effects with the control and safety systems are also prohibited). Design constraints on industrial hazards ensure personnel safety and continuity of operations. Design constraints limiting the release of radioactive gasses prevent unacceptable personnel exposure during off-normal experiment conditions.

TS-41 Original (Q/§4/1A)

I I K-State Reactor TS-41 Original (QW4/14)

TECHNICAL SPECIFICATIONS

6. Administrative Controls 6.1 Organization and Responsibilities of Personnel a) Structure.

The reactor organization is related to the University structure as shown in SAR Figure 12.1 and Technical Specifications Figure TS. I below.

Kansas State University (KSU) holds the license for the KSU TRIGA Reactor, located in the KSU Nuclear Reactor Facility in Ward Hall on the campus of Kansas State University. The chief administrating officer for KSU is the President. Environment, safety and health oversight functions are administered through the Vice President for Administration and Finance, while reactor line management functions are through the Provost Chief Academic Officer.

Figure TS.I: Organization and Management Structure for the K-State Reactor Radiation protection functions are divided between the University Radiation Safety Officer (RSO) and the reactor staff and management, with management and authority for the RSO separate from line management and authority for facility operations. Day-to-day radiation protection functions implemented by facility staff and management are guided I K-State Reactor TS-42 Original (91074ý/14)

TECHNICAL SPECIFICATIONS by approved administrative controls (Reactor Radiation Protection Program or RPP, Facility Operating Manual, operating and experiment procedures); these controls are reviewed and approved by the RSO as part of the Reactor Safeguards Committee (with specific veto authority). The RSO has specific oversight functions assigned though the RPP. The RSO provides routine support for personnel monitoring, radiological analysis, and radioactive material inventory control. The RSO provides guidance on request for non-routine operations such as transportation and implementation of new experiments.

b) Responsibility.

The President of the University shall be responsible for the appointment of responsible and competent persons as members of the TRIGA Reactor Safeguards Committee upon the recommendation of the ex officio Chairperson of the Committee.

The KSU Nuclear Reactor Facility shall be under the supervision of the Nuclear Reactor Facility Manager, who shall have the overall responsibility for safe, efficient, and competent use of its facilities in conformity with all applicable laws, regulations, terms of facility licenses, and provisions of the Reactor Safeguards Committee. The Manager also has responsibility for maintenance and modification of laboratories associated with the Reactor Facility. The Manager shall have education and/or experience commensurate with the responsibilities of the position and shall report to the Head of the Department of Mechanical and Nuclear Engineering.

A Reactor Supervisor may serve as the deputy of the Nuclear Reactor Facility Manager in all matters relating to the enforcement of established rules and procedures (but not in matters such as establishment of rules, appointments, and similar administrative functions). The Supervisor should have at least two years of technical training beyond high school and shall possess a Senior Reactor Operator's license. The Supervisor shall have had reactor OPERATING experience and have a demonstrated competence in supervision. The Supervisor is appointed by the Nuclear Reactor Facility Manager and is responsible for enforcing all applicable rules, procedures, and regulations, for ensuring adequate exchange of information between OPERATING personnel when shifts change, and for reporting all malfunctions, accidents, and other potentially hazardous occurrences and situations to the Reactor Nuclear Reactor Facility Manager. The Nuclear Reactor Facility Manager may also serve as Reactor Supervisor.

The Reactor Operator shall be responsible for the safe and proper operation of the reactor, under the direction of the Reactor Supervisor. Reactor Operators shall possess an Operator's or Senior Operator's license and shall be appointed by the Nuclear Reactor Facility Manager.

The University Radiation Safety Officer (RSO), or a designated alternate, shall (in addition to other duties defined by the Director of Environmental Health and Safety, Division of Public Safety) be responsible for overseeing the safety of Reactor Facility operations from the standpoint of radiation protection. The RSO and/or designated alternate shall be appointed by the Director of Environmental Health and Safety, Division of Public Safety, with the approval of the University Radiation Safety Committee, and shall report to the Director of Environmental Health and Safety, whose organization is independent of the Reactor Facility organization, as shown on SAR Figure 12.1.

The Nuclear Reactor Facility Manager, with the approval of the Reactor Safeguards Committee, may designate an appropriately qualified member of the Facility organization as Reactor Facility Safety Officer (RFSO) with duties including those of an intra-Facility I K-State Reactor TS-43 Original (9Q74/14)

TECHNICAL SPECIFICATIONS Radiation Safety Officer. The University Radiation Safety Officer may, with the concurrence of the Nuclear Reactor Facility Manager, authorize the RFSO to perform some of the specific duties of the RSO at the Nuclear Reactor Facility.

c). Staffing.

Whenever the reactor is not secured, the reactor shall be under the direction of a (USNRC licensed) Senior Operator (designated as Reactor Supervisor). The Supervisor shall be on call, within twenty minutes travel time to the facility.

Whenever the reactor is not secured, a (USNRC licensed) Reactor Operator (or Senior Reactor Operator) who meets requirements of the Operator Requalification Program shall be at the reactor control console, and directly responsible for control manipulations.

In addition to the above requirements, during fuel movement a senior operator shall be inside the reactor bay directing fuel operations.

6.2 Review and Audit a) There will be a Reactor Safeguards Committee which shall review TRIGA reactor operations to assure that the reactor facility is operated and used in a manner within the terms of the facility license and consistent with the safety of the public and of persons within the Laboratory.

b) The responsibilities of the Committee include, but are not limited to, the following:

1. Review and approval of rules, procedures, and proposed Technical Specifications;

2. Review and approval of all proposed changes in the facility that could have a significant effect on safety and of all proposed changes in rules, procedures, and Technical Specifications, in accordance with procedures in Section 6.3;

3. Review and approval of experiments using the reactor in accordance with procedures and criteria in Section 6.4;

4. Determine whether changes in the facility as described in the safety analysis report (as updated), changes in the procedures as described in the final safety analysis report (as updated), and the conduct of tests or experiments not described in the safety analysis report (as updated) may be accomplished in accordance with 10 CFR 50.59 without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90.

5. Review of abnormal performance of plant equipment and OPERATING anomalies;

6. Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR 20 and 10 CFR50;

7. Inspection of the facility, review of safety measures, and audit of operations at a frequency not less than once a year, including operation and operations records of the facility; I K-State Reactor TS-44 Original (91G74/14)

TECHNICAL SPECIFICATIONS

8. Requalification of the Nuclear Reactor Facility Manager and/or the Reactor Supervisor,

9. Review of container failures where released materials have the potential for damaging reactor fuel or structural components including:

a) results of physical inspection b) evaluation of consequences c) need for corrective actions c) The Committee shall be composed of:

1. one or more persons proficient in reactor and nuclear science or engineering,

2. one or more persons proficient in chemistry, geology, or chemical engineering,

3. one person proficient in biological effects of radiation,

4. the Nuclear Reactor Facility Manager, ex officio,

5. the University Radiation Safety Officer, ex officio, and,

6. The Head of the Department of Mechanical and Nuclear Engineering, ex officio, or a designated deputy, to serve as chairperson of the Committee.

The same individual may serve under more than one category above, but the minimum membership shall be seven. At least five members shall be faculty members. The Reactor Supervisor, if other than the Nuclear Reactor Facility Manager, shall attend and participate in Committee meetings, but shall not be a voting member.

d) The Committee shall have a written statement defining its authority and responsibilities, the subjects within its purview, and other such administrative provisions as are required for its effective functioning. Minutes of all meetings and records of all formal actions of the Committee shall be kept.

e) A quorum shall consist of not less than a majority of the full Committee and shall include all ex officio members.

f) Any permissive action of the Committee requires affirmative vote of the University Radiation Safety Officer as well as a majority vote of the members present.

g) The Committee shall meet a minimum of two times a year. Additional meetings may be called by any member, and the Committee may be polled in lieu of a meeting. Such a poll shall constitute Committee action subject to the same requirements as for an actual meeting.

6.3 Procedures a ) Written procedures, reviewed and approved by the Reactor Safeguards Committee, shall be followed for the activities listed below. The procedures shall be adequate to I K-State Reactor TS-45 Original (W974,/14)

TECHNICAL SPECIFICATIONS assure the safety of the reactor, persons within the Laboratory, and the public, but should not preclude the use of independent judgment and action should the situation require it. The activities are:

1. Startup, operation, and shutdown of the reactor, including (a) startup checkout procedures to test the reactor instrumentation and safety systems, area monitors, and continuous air monitors, (b) prohibition of routine operations with failed (or leaking) fuel except to find leaking elements, and (b) shutdown procedures to assure that the reactor is secured before OPERATING personnel go off duty.

2. Installation or removal of fuel elements, control rods, and other core components that significantly affect reactivity or reactor safety.

3. Preventive or corrective maintenance activities which could have a significant effect on the safety of the reactor or personnel.

4. Periodic inspection, testing or calibration of auxiliary systems or instrumentation that relate to reactor operation.

b) Substantive changes in the above procedures shall be made only with the approval of the Reactor Safeguards Committee, and shall be issued to the OPERATING personnel in written form. The Nuclear Reactor Facility Manager may make temporary changes that do not change the original intent. The change and the reasons thereof shall be noted in the log book, and shall be subsequently reviewed by the Reactor Safeguards Committee.

c) Determination as to whether a proposed activity in categories (1). (2) and (3) in Section 6.2b above does or does not have a significant safety effect and therefore does or does not require approved written procedures shall require the concurrence of I. the Nuclear Reactor Facility Manager, and

2. at least one other member of the Reactor Safeguards Committee, to be selected for relevant expertise by the Nuclear Reactor Facility Manager. If the Manager and the Committee member disagree, or if in their judgment the case warrants it, the proposal shall be submitted to the full Committee, and

3. the University Radiation Safety Officer, or his/her deputy, who may withhold agreement until approval by the University Radiation Safety Committee is obtained.

The Rector Safeguards Committee shall subsequently review determinations that written procedures are not required. The time at which determinations are made, and the review and approval of written procedures, if required, are carried out, shall be a reasonable interval before the proposed activity is to be undertaken.

d) Determination that a proposed change in the facility does or does not have a significant safety effect and therefore does or does not require review and approval by the full Reactor Safeguards Committee shall be made in the same manner as for proposed activities under (c) above.

I K-State Reactor TS-46 Original (WG74/14)

TECHNICAL SPECIFICATIONS 6.4 Review of Proposals for Experiments a) All proposals for new experiments involving the reactor shall be reviewed with respect to safety in accordance with the procedures in (b) below and on the basis of criteria in (c) below.

b) Procedures:

1. Proposed reactor operations by an experimenter are reviewed by the Reactor Supervisor, who may determine that the operation is described by a previously approved EXPERIMENT or procedure. If the Reactor Supervisor determines that the proposed operation has not been approved by the Reactor Safeguards Committee, the experimenter shall describe the proposed EXPERIMENT in written form in sufficient detail for consideration of safety aspects. If potentially hazardous operations are involved, proposed procedures and safety measures including protective and monitoring equipment shall be described.

2. If the experimenter is a student, approval by his/her research supervisor is required. If the experimenter is a staff or faculty member, his/her own signature is sufficient.

3. The proposal is then to be submitted to the Reactor Safeguards Committee for consideration and approval. The Committee may find that the experiment, or portions thereof, may only be performed in the presence of the University Radiation Safety Officer or Deputy thereto.

4. The scope of the EXPERIMENT and the procedures and safety measures as described in the approved proposal, Including any amendments or conditions added by those reviewing and approving it, shall be binding on the experimenter and the OPERATING personnel. Minor deviations shall be allowed only in the manner described in Section 6 above. Recorded affirmative votes on proposed new or revised experiments or procedures must indicated that the Committee determines that the proposed actions do not involve changes in the facility as designed, changes in Technical Specifications, changes that under the guidance of 10 CFR 50.59 require prior approval of the NRC, and could be taken without endangering the health and safety of workers or the public or constituting a significant hazard to the integrity of the reactor core.

5. Transmission to the Reactor Supervisor for scheduling.

c) Criteria that shall be met before approval can be granted shall include:

1. The EXPERIMENT must meet the applicable Limiting Conditions for Operation and Design Description specifications.

2. It must not involve violation of any condition of the facility license or of Federal, State, University, or Facility regulations and procedures.

3. The conduct of tests or experiments not described in the safety analysis report (as updated) must be evaluated in accordance with 10 CFR 50.59 to determine if the test I K-State Reactor TS-47 Original (WG74,/14)

TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90.

4. In the safety review the basic criterion is that there shall be no hazard to the reactor, personnel or public. The review SHALL determine that there is reasonable assurance that the experiment can be performed with no significant risk to the safety of the reactor, personnel or the public.

6.5 Emergency Plan and Procedures An emergency plan shall be established and followed in accordance with NRC regulations. The plan shall be reviewed and approved by the Reactor Safeguards Committee prior to its submission to the NRC. In addition, emergency procedures that have been reviewed and approved by the Reactor Safeguards Committee shall be established to cover all foreseeable emergency conditions potentially hazardous to persons within the Laboratory or to the public, including, but not limited to, those involving an uncontrolled reactor excursion or an uncontrolled release of radioactivity.

6.6 Operator Requalification An operator requalification program shall be established and followed in accordance with NRC regulations.

6.7 Physical Security Plan Administrative controls for protection of the reactor plant shall be established and followed in accordance with NRC regulations.

6.8 Action To Be Taken In The Event A Safety Limit Is Exceeded In the event a safety limit is exceeded:

a) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

b) An immediate report of the occurrence shall be made to the Chair of the Reactor Safeguards Committee, and reports shall be made to the NRC in accordance with Section 6.11 of these specifications.

c) A report shall be made to 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 Reactor Safeguards Committee for review, and a suitable similar report submitted to the NRC when authorization to resume operation of the reactor is sought.

6.9 Action To Be Taken In The Event Of A Reportable Occurrence a ) A reportable occurrence is any of the following conditions:

TS-48 Original (Q~Q-74LiA)

I Reactor I K-State Reactor TS-48 Original (W974,/14)

TECHNICAL SPECIFICATIONS

1. any actual safety system setting less conservative than specified in Section 2.2, Limiting Safety System Settings;

2. VIOLATION OF SL, LSSS OR LCO; NOTES Violation of an LSSS or LCO occurs throughfailure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itselfa violation.

Surveillance Requirements must be met for all equipment/comnponents/conditionsto be consideredoperable.

Failureto petform a surveillance within the requiredtime intervalorfailure of a surveillance test shall result in the /component/condition being inoperable

3. incidents or conditions that prevented or could have prevented the performance of the intended safety functions of an engineered safety feature or the REACTOR SAFETY SYSTEM;

4. release of fission products from the fuel that cause airborne contamination levels in the reactor bay to exceed 10CFR20 limits for releases to unrestricted areas;

5. an uncontrolled or unanticipated change in reactivity greater than $1.00;

6. an observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy has caused the existence or development of an unsafe condition in connection with the operation of the reactor;

7. an uncontrolled or unanticipated release of radioactivity.

b) In the event of a reportable occurrence, the following actions shall be taken:

I. The reactor shall be shut down at once. The Reactor Supervisor shall be notified and corrective action taken before operations are resumed; the decision to resume shall require approval following the procedures in Section 6.3.

2. A report shall be made to include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safeguards Committee for review.

3. A report shall be submitted to the NRC in accordance with Section 6.11 of these specifications.

6.10 Plant Operating Records a ) In addition to the requirements of applicable regulations, in 10 CFR 20 and 50, records and logs shall be prepared and retained for a period of at least 5 years for the following items as a minimum.

I K-State Reactor TS-49 Original (W974,/14)

TECHNICAL SPECIFICATIONS

1. normal plant operation, including power levels;

3. principal maintenance activities;

4. reportable occurrences;

5. equipment and component surveillance activities;

6. experiments performed with the reactor;

7. all emergency reactor scrams, including reasons for emergency shutdowns.

b) The following records shall be maintained for the life of the facility:

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

2. offsite environmental monitoring surveys;

3. fuel inventories and transfers;

4. facility radiation and contamination surveys;

5. radiation exposures for all personnel;

6. updated, corrected, and as-built drawings of the facility.

6.11 Reporting Requirements All written reports shall be sent within the prescribed interval to the United States Nuclear Regulatory Commission, Washington, D.C., 20555, Attn: Document Control Desk.

In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the US. Nuclear Regulatory Commission (NRC) as follows:

a ) A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and fax or electronic mail to the NRC Operations Center and the USNRC Region IV of; I. any accidental release of radioactivity above permissible limits in unrestricted areas, whether or not the release resulted in property damage, personal injury, or exposure;

2. any violation of a safety limit;

3. any reportable occurrences as defined in Section 6.9 of these specifications.

b) A report within 10 days in writing of:

1. any accidental release of radioactivity above permissible limits in unrestricted areas, whether or not the release resulted in property damage, personal injury or exposure; the written report (and, to the extent possible, the preliminary telephone and I K-State Reactor TS-50 Original (GiG74/14)

TECHNICAL SPECIFICATIONS telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent recurrence of the event;

2. any violation of a safety limit;

3. any reportable occurrence as defined in Section 1.1 of these specifications.

c) A report within 30 days in writing of:

1. any significant variation of a MEASURED VALUE from a corresponding predicted or previously MEASURED VALUE of safety-connected OPERATING characteristics occurring during operation of the reactor;

2. any significant change in the transient or accident analysis as described in the Safety Analysis Report.

3. a change in personnel for the Department of Mechanical and Nuclear Engineering Chair, or a change in reactor manager d) A report within 60 days after criticality of the reactor in writing to the US Nuclear Regulatory Commission, resulting from a receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level or the installation of a new core, describing the MEASURED VALUE of the OPERATING conditions or characteristics of the reactor under the new conditions.

e) A routine report in writing to the US. Nuclear Regulatory Commission within 60 days after completion of the first calendar year of OPERATING and at intervals not to exceed 12 months, thereafter, providing the following information:

1. a brief narrative summary of OPERATING experience (including experiments performed), changes in facility design, performance characteristics, and OPERATING procedures related to reactor safety occurring during the reporting period; and results of surveillance tests and inspections;

2. a tabulation showing the energy generated by the reactor (in megawatt-hours);

3. the number of emergency shutdowns and inadvertent scrams, including the reasons thereof and corrective action, if any, taken;

4. discussion of the major maintenance operations performed during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required;

5. a summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of 10 CFR 50.59;

6. 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 before the point of such release or discharge;

7. a description of any environmental surveys performed outside the facility; TS-51 Original (QIQ7~4I14)

I I K-State Reactor TS-51 Original (GAG74,114)

TECHNICAL SPECIFICATIONS

8. a summary of radiation exposures received by facility personnel and visitors, including the dates and time of significant exposure, and a brief summary of the results of radiation and contamination surveys performed within the facility.

TS-52 Original (9/07-4/14)

I K-State Reactor K-State Reactor TS-52 Original (GIG74/14)