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| issue date = 05/02/2017
| issue date = 05/02/2017
| title = Kansas State Univ. - Facility Response to 3/28/17 Request for Additional Information Regarding License Amendment Request
| title = Kansas State Univ. - Facility Response to 3/28/17 Request for Additional Information Regarding License Amendment Request
| author name = Geuther J A
| author name = Geuther J
| author affiliation = Kansas State Univ
| author affiliation = Kansas State Univ
| addressee name =  
| addressee name =  
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:*
{{#Wiki_filter:KANSAS STATE                  TRIGA Mk II Nuclear
* US Nuclear Regulatory Commission Washington, DC 20555-0001 2 May 2017 KANSAS STATE TRIGA Mk II Nuclear U N I V E R s 1 T y Reactor Laboratory  
* US Nuclear Regulatory Commission Washington, DC 20555-0001 U N I V E R       s 1 T y     Reactor Laboratory 2 May 2017


==Subject:==
==Subject:==
Facility Response to 3/28/17 Request for Additional Information (Acc.# ML17038A272)
Facility Response to 3/28/17 Request for Additional Information (Acc.# ML17038A272)
To Whom It May Concern, On March 28th, 2017, the NRC sent a third Request for Additional Information (RAI) to the Kansas State University nuclear reactor facility (license R-88, docket 50-188) regarding a license amendment request (LAR) to add up to four 12%-loaded fuel elements to the core. The original LAR was submitted on April 9, 2012 (Acc.# ML1219A063).
To Whom It May Concern, On March 28th, 2017, the NRC sent a third Request for Additional Information (RAI) to the Kansas State University nuclear reactor facility (license R-88, docket 50-188) regarding a license amendment request (LAR) to add up to four 12%-loaded fuel elements to the core. The original LAR was submitted on April 9, 2012 (Acc.# ML1219A063).
The resolution of the question asked in the most recent RAI required several amendments to Chapter 4 of the facility Safety Analysis Report and to the facility Technical Specifications.
The resolution of the question asked in the most recent RAI required several amendments to Chapter 4 of the facility Safety Analysis Report and to the facility Technical Specifications. These changes were reviewed and approved by the Reactor Safeguards Committee on 4/28/17 by unanimous vote, pending several minor editorial changes that have been made since the vote.
These changes were reviewed and approved by the Reactor Safeguards Committee on 4/28/17 by unanimous vote, pending several minor editorial changes that have been made since the vote. The following constitutes the facility's response to the RAI, and is organized as a numbered list of abbreviated questiops in order of appearance in the RAI, followed by the facility response.
The following constitutes the facility's response to the RAI, and is organized as a numbered list of abbreviated questiops in order of appearance in the RAI, followed by the facility response.
: 1. ... Please provide information that validates the pool boiling model used in the SAR or alternatively provide a traditional hot channel analysis using the Bernath correlation  
: 1. ... Please provide information that validates the pool boiling model used in the SAR or alternatively provide a traditional hot channel analysis using the Bernath correlation ... with core inlet conditions at the TS limit, or another correlation that has been validated against acceptable data.
... with core inlet conditions at the TS limit, or another correlation that has been validated against acceptable data. A RELAP-5 single-channel model was used to update the SAR analysis for a fuel element heated to 24 kWth in water at a pressure of 1.43 kPa, corresponding to the depth of the fuel in the reactor pool. 24 kWth is the power in a fuel element with an 85-element core operating at 1.25 MW and an average power peaking value of 1.63. The bulk water temperature used in the analysis was 49°C. This value is slightly lower than the maximum of 54°C specified in the Technical Specifications.
A RELAP-5 single-channel model was used to update the SAR analysis for a fuel element heated to 24 kWth in water at a pressure of 1.43 kPa, corresponding to the depth of the fuel in the reactor pool. 24 kWth is the power in a fuel element with an 85-element core operating at 1.25 MW and an element-to-average power peaking value of 1.63. The bulk water temperature used in the analysis was 49°C. This value is slightly lower than the maximum of 54°C specified in the Technical Specifications. However, based on this analysis, the TS will be revised to permit a maximum bulk water temperature of 44°C in order to avoid bulk boiling of the coolant.
However, based on this analysis, the TS will be revised to permit a maximum bulk water temperature of 44°C in order to avoid bulk boiling of the coolant. The departure from nucleate boiling ratio (DNBR) reached a minimum value of slightly above 2.0. This is significantly lower than the DNBR calculated in the current version of the SAR. The reactor core grid plates contain an array of 8 mm interstitial holes meant to accommodate flux wires or other experiments.
The departure from nucleate boiling ratio (DNBR) reached a minimum value of slightly above 2.0. This is significantly lower than the DNBR calculated in the current version of the SAR. The reactor core grid plates contain an array of 8 mm interstitial holes meant to accommodate flux wires or other experiments. The presence of an experiment in these holes may reduce the temperature at which bulk boiling will occur; therefore an additional experimental design constraint is being proposed for the TS that forbids the insertion of tubes or flux wires in the interstitial holes at bulk water temperatures
The presence of an experiment in these holes may reduce the temperature at which bulk boiling will occur; therefore an additional experimental design constraint is being proposed for the TS that forbids the insertion of tubes or flux wires in the interstitial holes at bulk water temperatures
* greater than 37°C. A drawing of the upper grid plate showing the interstitial holes is attached for reference.
* *
Inlet 1~-                                    Cold*Leg connector Figure 1 - RELAP single channel model
* greater than 37°C. A drawing of the upper grid plate showing the interstitial holes is attached for reference. Fi g ure 1 -R ELAP si n gle channe l mod e l .. .. .. .. .. .. .. , .... 4 000 .. "'c: 3 000 :;:--_,. -lo! LL. ::c 20 00 u 1000 * * * .. * * * * .. Inlet
                                          .. .. .. .. .. .. ..                                         .... Bernath-CHF
* Cold*Leg c o n n ec to r .... .... * * .... , .. . . .. .. " .... . ... ...........  
                                                                .... .. .... ,                          .... PG -CHF 4000
* *
                                                                              ..                       *
* Bernath-CHF PG-CHF Heat flux * .i. ........ . * * * * * *
* Heat flux
* 8.Loo ___ o .... o_s ___ o .... 1-o--o-.1'-s---o ..... 2_0--o-."" 2 s ___
                            "'c:  3000
Heated Length (m) F igu r e 2 -Cr i tica l heat flu x (CHF) versus heated l ength 2 
                                                                                                                  * .i. .~. . .... . ..
* * * + + Bernath-CHF 12 + + PG-CHF 10 ............ : .... 0:: CXl ***
                              -lo!
* B ..... ******** *+
LL.
* z 0 6 : . ..........
::c 2000 u
: ** .......* . . : . :* ..... * *
1000
* 4 ... * .......... *'* **************** ** .... . . : . . : . . . **************** . 2 0.05 0.10 0.15 0.20 0.25 0.3 0 Heated Length (m) Figure 3 -Departure from nucleate bailing ratio (DNBR) versus heated length * ** ... *= * * ** +:
                                                              * ~ * * ..
* 0.35 2 .... Propose a licensed thermal power limit that is within the range of the currently installed nuclear instrumentation or describe how the nuclear instrumentation system is capable of measurement of the full range of reactor power levels anticipated as described in the safety analyses including instrument uncertainties based on the current licensed thermal power limit. The three nuclear instruments at KSU have maximum readings of: 1.0 MW (NLW-1000, log channel), 1.20 MW {NMP-1000, multi-range linear channel) and 1.10 MW {NPP-1000, percent power I pulse channel).
8.Loo___o....o_s___o...1-o--o-.1'-s- --o.....2_0--o-.""2s___o_..3,..,o--o.,...~3.,..s__.
Two of these are required by the Limiting Conditions of Operation during steady state mode, one is required during pulse mode {TS 3.3, attached).
Heated Length (m)
Therefore readings at or slightly above 1.0 MW would be readable on the console instruments, but the console instruments are incapable of reading up to the license limit of 1.25 MW. In order to address this issue, the revised Technical Specifications have been amended as follows. The Limiting Safety System Setting has been left unchanged at 1.25 MW. However, an additional Limiting Condition of Operation has been added which sets the maximum operating power to 1.00 MW, with a requirement to reduce power if it exceeds 1.05 MW. In this way the facility will still have some margin to the LSSS if, for example, it is determined that the power channels were reading low due to detector uncertainty or flux shifting following power calibration.
Figu re 2 - Critical heat flux (CHF) versus heated length
However, the maximum power will not be allowed to exceed the useful range of the control panel instrumentation.
* 2
The new LCO is found in the attached Technical Specifications, section 3.10. Note that the upgraded control console instrumentation planned for installation in January 2018 is capable of reading 1.25 MW of power . 3 
* 12
* *
                                    + +
* 3. . .. Provide an explanation resolving the significant deviation between the reactor power observed on June 27, 2016 {735 kWt} and the maximum reactor power predicted in support of this license amendment
                                    + +
{600 kWt} and provide a revised analysis for the maximum core operating power after the addition of the four 12.5 U wt% fuel elements.
Bernath-CHF PG-CHF
The reason for the deviation between the power observed on 6/27 /2016 (735 kWth) and the predicted maximum power of 600 kWth in the original license amendment proposal is primarily due to the length of time involved in the LAR process. In 2012, when the LAR was originally submitted, the most recent operation at near full power had been 3/28/12, when the reactor operated at 550 kW with the three most reactive rods fully-withdrawn, and the regulating rod withdrawn to a position corresponding to only 0.04$ remaining.
                                                                                                                      ... * =
The reactor power is limited by core reactivity, so during the annual fuel inspections in the following years, several old fuel elements were replaced with fresh elements, resulting in an increase in the available excess reactivity.
10
Based on the increase in power level versus when the original LAR was submitted, the maximum power level would be expected to be slightly greater than 1 MW (i.e., 1.035 MW). 4 .... Provide clarification on how the MCNP calculations are integrated with the facility measurements to determine the maximum excess reactivity and the minimum shutdown margin. In order to reduce the error on the estimates for maximum excess reactivity and minimum shutdown margin, MCNP was used to determine only the change in reactivity due to loading 12% fuel, not the total reactivity.
                                                  **+                                                      **
Measured values of excess reactivity were taken from the daily reactivity balance, in which the calibrated integral control rod worth curves are used to determine how much excess reactivity remains above the critical configuration at zero power with no xenon. The estimated values were therefore calculated as follows: max. reactivity (estimated)=
B .... . *** ****
max. reactivity (measured)+
* 0::
8p (MCNP) min. shutdown margin (estimated)=
CXl z                            * *~
min. shutdown margin (measured)  
0    6                                **                          :
-8p (MCNP) 4 
:                                          + :*
* *
4 2
* 5 .... If operation with one or more control rods inoperable but fully inserted is acceptable, please provide the supporting analyses and evaluations from which operational acceptability is derived. The operation of the reactor core with one or more control rods inoperable but fully inserted was judged to be safe due to the understanding that the safety function of control rods is to provide a means to control reactivity, i.e., if a control rod was inoperable but inserted, then it was fulfilling its safety function.
                                                          . !~*********! : .
The response to the previous RAI did not consider the changes in power peaking due to the insertion of a control rod. In order to address this additional consideration, a reactor model was created in MCNP6 in which the fuel elements were all 9.0 wt%-loaded elements.
0.05          0.10    0.15    0.20          0.25            0.30          0.35 Heated Length (m )
The fuel elements were all modeled as having the highest allowable density of uranium, i.e, fresh fuel at 9.0 wt% U. The maximum element-to-core-average power peaking was calculated using fission heating cell tallies, and the results were compared between cases where all rods were fully withdrawn, versus having individual rods failed but fully inserted.
Figure 3 - Departure from nucleate bailing ratio (DNBR) versus heated length
Table 1 shows a summary of the results. The power peaking in most cases is B4, which is the fuel element located between the pulse rod channel and central thimble, such that this location is heavily moderated when the pulse rod is fully withdrawn.
* 2 . ... Propose a licensed thermal power limit that is within the range of the currently installed nuclear instrumentation or describe how the nuclear instrumentation system is capable of measurement of the full range of reactor power levels anticipated as described in the safety analyses including instrument uncertainties based on the current licensed thermal power limit.
When the pulse (i.e., transient) rod is inoperable and fully inserted, the maximum power peaking would occur in fuel element Cll. In some cases the power peaking is greater than in the all-rods-out case, but no case approached the assumed 2.00 power peaking factor used for PCT analysis.
The three nuclear instruments at KSU have maximum readings of: 1.0 MW (NLW-1000, log channel),
Rod Inserted Peak Power Factor Element None 1.558 B4 Regulating Rod 1.617 B4 Shim Rod 1.630 B4 Safety Rod 1.609 B4 Pulse Rod 1.438 Cll SAR 2.00 N/A Table 1: Peak Power Factors Using recent measurements of integral control rod worth, the peak element power was calculated for an all-rods-out case at the current license power limit of 1250 kW for a fully-loaded core, and then for a case with each of the four control rods fully inserted.
1.20 MW {NMP-1000, multi-range linear channel) and 1.10 MW {NPP-1000, percent power I pulse channel). Two of these are required by the Limiting Conditions of Operation during steady state mode, one is required during pulse mode {TS 3.3, attached). Therefore readings at or slightly above 1.0 MW would be readable on the console instruments, but the console instruments are incapable of reading up to the license limit of 1.25 MW. In order to address this issue, the revised Technical Specifications have been amended as follows. The Limiting Safety System Setting has been left unchanged at 1.25 MW.
For these cases, the power of the core was reduced based on the current set of integral control rod worth curves and an assumed power coefficient of -0.0045 $I kW, which is similar to the value of-0.0038$
However, an additional Limiting Condition of Operation has been added which sets the maximum operating power to 1.00 MW, with a requirement to reduce power if it exceeds 1.05 MW. In this way the facility will still have some margin to the LSSS if, for example, it is determined that the power channels were reading low due to detector uncertainty or flux shifting following power calibration.
/kw calculated by dividing the recent peak power of 735 kW by the present-day 2.77$ of excess reactivity.
However, the maximum power will not be allowed to exceed the useful range of the control panel instrumentation. The new LCO is found in the attached Technical Specifications, section 3.10. Note that the upgraded control console instrumentation planned for installation in January 2018 is capable of reading 1.25 MW of power .
The maximum power per element for this set of models is given in Table 2. Rod Inserted Approximate Reactor Power (kW} Regulating Rod (max, IRW 882 = 0.87$) Shim Rod (max, IRW = 738 2.13$) Safety Rod (max, IRW = 758 1.68$) 5 Average Power Per Element (kW) 10.38 8.68 8.91 Peak Element Power (kW} 16.78 14.15 14.35 
* 3
* * ** Rod Inserted Pulse Rod (max, IRW = 2.77$) 1250 kW, all rods out Approximate Reactor Power (kW) 685 1250 Average Power Per Element (kW) 7.74 14.71 Peak Element Power (kW) 11.13 22.91 Table 2: Power comparison between different core configurations, no 12% fuel installed.
* 3. ... Provide an explanation resolving the significant deviation between the reactor power observed on June 27, 2016 {735 kWt} and the maximum reactor power predicted in support of this license amendment {600 kWt} and provide a revised analysis for the maximum core operating power after the addition of the four 12.5 U wt% fuel elements.
A similar study was conducted using a peak power of 1250 kWth with all rods withdrawn, but with a single 12.3%-weighted fuel element in location E-10. This study confirms that the maximum power in a single fuel element remains well below the SAR-assumed value of 2.0 element-to-average, even with a high-load fuel element on the opposite side of the core relative to the inserted control rod. The results are depicted in Figure 4 . 6 
The reason for the deviation between the power observed on 6/27 /2016 (735 kWth) and the predicted maximum power of 600 kWth in the original license amendment proposal is primarily due to the length of time involved in the LAR process. In 2012, when the LAR was originally submitted, the most recent operation at near full power had been 3/28/12, when the reactor operated at 550 kW with the three most reactive rods fully-withdrawn, and the regulating rod withdrawn to a position corresponding to only 0.04$ remaining. The reactor power is limited by core reactivity, so during the annual fuel inspections in the following years, several old fuel elements were replaced with fresh elements, resulting in an increase in the available excess reactivity. Based on the increase in power level versus when the original LAR was submitted, the maximum power level would be expected to be slightly greater than 1 MW (i.e., 1.035 MW).
* *
* 4 . ... Provide clarification on how the MCNP calculations are integrated with the facility measurements to determine the maximum excess reactivity and the minimum shutdown margin.
* L ege nd P o w r P r E l e m en t (k W) ll*U JJO') D ow n Rod: R egula t ing. R e act or Power: 8 82.0 W. P ea E l em nt Pow r; 1 6.8 7 k W Down Rod; S a fety. Rea c t or Pow e r: 7 58.0 w. P ea E l eme nt Pow e r: 14.67 W D o wn R od; N on e. R ea ct or Po wer: 1 25 0.0 w. Peak E l em ent P o w er: 2 3.0 5k W Down R o d: Pulse. R e ac t o r Po wer; 65 8.0 W , Peak E l eme nt P ow e r: l l.3 4 k W Figure 4 -Powe r peoking map with different control rods stuck in , with o 12% element in location E-10 and a maximum core power of 1.25 MW. Therefore the case of an inoperable control rod being fully inserted may slightly increase maximum power peaking but will greatly reduce the maximum power per element. The slight increase in power peaking will also be more than offset by the reduction in core reactivity due to the insertion of the rod poison , resulting in a massive decrease to the maximum fuel element power. The proposed revision to the TS includes a specification of minimum of 3 operable control rods, and states that inoperable control rods must be fully inserted. (See attached, TS 3.4) . 7 
In order to reduce the error on the estimates for maximum excess reactivity and minimum shutdown margin, MCNP was used to determine only the change in reactivity due to loading 12% fuel, not the total reactivity. Measured values of excess reactivity were taken from the daily reactivity balance, in which the calibrated integral control rod worth curves are used to determine how much excess reactivity remains above the critical configuration at zero power with no xenon. The estimated values were therefore calculated as follows:
* *
max. reactivity (estimated)= max. reactivity (measured)+ 8p (MCNP) min. shutdown margin (estimated)= min. shutdown margin (measured) - 8p (MCNP)
* 6 .... "Provide a revision to the proposed TS describing the geometric limitation and the controls that will help ensure compliance and include information on the acceptability of the specific location where the 12.5% fuel elements will be installed.
* 4
Otherwise, describe how the current or previously proposed TS adequately address the control of this geometric limitation on the location [of] 12.5 wt% fuel elements in the proposed reactor core." MCNP results transmitted in the initial LAR and in previous RAI responses indicate that 12.5 wt% (max.) fuel in the E-and F-rings will be kept below the fission heating density (i.e., power) of the 9.0 wt% (max.) elements in the B-ring, even if placed adjacent to another 12.5%-loaded element. The additional restriction to avoided placing the 12.5% elements adjacent to control rod channels was intended to avoid conflict with the NRC over local power peaking during pulsing, although results from MCNP calculations performed at KSU have not shown reason for concern. The proposed TS have been revised to include a specific list of locations in which 12.5% fuel cannot be located. (See attached).
* 5 .... If operation with one or more control rods inoperable but fully inserted is acceptable, please provide the supporting analyses and evaluations from which operational acceptability is derived.
In order to ensure that the 12% fuel is only added to locations where it permitted, the fuel handling procedures will be revised to include guidance to check the Technical Specifications list of approved locations prior to loading 12% fuel. I swear under penalty of perjury that the foregoing is true and correct. Regards, Nuclear Reactor Facilities Manager Department of Mechanical and Nuclear Engineering Kansas State University Manhattan, KS 66506 Phone: {785)532-6657
The operation of the reactor core with one or more control rods inoperable but fully inserted was judged to be safe due to the understanding that the safety function of control rods is to provide a means to control reactivity, i.e., if a control rod was inoperable but inserted, then it was fulfilling its safety function. The response to the previous RAI did not consider the changes in power peaking due to the insertion of a control rod. In order to address this additional consideration, a reactor model was created in MCNP6 in which the fuel elements were all 9.0 wt%-loaded elements. The fuel elements were all modeled as having the highest allowable density of uranium, i.e, fresh fuel at 9.0 wt% U. The maximum element-to-core-average power peaking was calculated using fission heating cell tallies, and the results were compared between cases where all rods were fully withdrawn, versus having individual rods failed but fully inserted. Table 1 shows a summary of the results. The power peaking in most cases is B4, which is the fuel element located between the pulse rod channel and central thimble, such that this location is heavily moderated when the pulse rod is fully withdrawn. When the pulse (i.e., transient) rod is inoperable and fully inserted, the maximum power peaking would occur in fuel element Cll. In some cases the power peaking is greater than in the all-rods-out case, but no case approached the SAR-assumed 2.00 power peaking factor used for PCT analysis.
{W) {785)236-0602 (C) Fax: {785)532-7057 Email: geuther@ksu.edu Attachments (5): GA Drawing TOS21E106 -Top Grid Plate SAR Ch. 4 markup copy SAR Ch. 4 clean copy TS markup copy TS clean copy 8
Rod Inserted                              Peak Power Factor                  Element None                                      1.558                              B4 Regulating Rod                            1.617                              B4 Shim Rod                                  1.630                              B4 Safety Rod                                1.609                              B4 Pulse Rod                                1.438                              Cll SAR                                      2.00                              N/A Table 1: Peak Power Factors Using recent measurements of integral control rod worth, the peak element power was calculated for an all-rods-out case at the current license power limit of 1250 kW for a fully-loaded core, and then for a case with each of the four control rods fully inserted. For these cases, the power of the core was reduced based on the current set of integral control rod worth curves and an assumed power coefficient of -0.0045 $I kW, which is similar to the value of-0.0038$ /kw calculated by dividing the recent peak power of 735 kW by the present-day 2.77$ of excess reactivity. The maximum power per element for this set of models is given in Table 2.
! .391 D!A. (2.H.QLES) i l .! -'L 
Rod Inserted                      Approximate Reactor      Average Power Per          Peak Element Power (kW}                Element (kW)               Power (kW}
* *
Regulating Rod (max, IRW          882                      10.38                      16.78
* 4. Reactor Description
    =0.87$)
Shim Rod (max, IRW      =        738                      8.68                      14.15 2.13$)
* Safety Rod (max, IRW 1.68$)
                              =      758 5
8.91                      14.35
* Rod Inserted Pulse Rod (max, IRW =
2.77$)
1250 kW, all rods out Approximate Reactor Power (kW) 685 1250 Average Power Per Element (kW) 7.74 14.71 Peak Element Power (kW) 11.13 22.91 Table 2: Power comparison between different core configurations, no 12% fuel installed.
A similar study was conducted using a peak power of 1250 kWth with all rods withdrawn, but with a single 12.3%-weighted fuel element in location E-10. This study confirms that the maximum power in a single fuel element remains well below the SAR-assumed value of 2.0 element-to-average, even with a high-load fuel element on the opposite side of the core relative to the inserted control rod. The results are depicted in Figure 4 .
**                                                    6
* Legend Down Rod; None.
Rea ctor Power: 1250.0 w.
Peak Elem ent Power: 23.05kW Pow  r P r Element (kW) ll*U JJO')
Down Rod: Regula ting.
Reactor Power: 882 .0 W.
Pea Elem nt Pow r; 16.8 7kW Down Rod ; Safety .                   Down Rod: Pulse.
Rea ctor Power: 758.0 w.               React or Po wer; 658.0 W, Pea Element Power: 14.67 W              Peak Element Pow er : l l.34kW Figure 4 -Power peoking map with different control rods stuck in, with o 12% element in location E-10 and a maximum core power of 1.25 MW.
Therefore the case of an inoperable control rod being fully inserted may slightly increase maximum power peaking but will greatly reduce the maximum power per element. The slight increase in power peaking will also be more than offset by the reduction in core reactivity due to the insertion of the rod poison, resulting in a massive decrease to the maximum fuel element power.
The proposed revision to the TS includes a specification of minimum of 3 operable control rods, and states that inoperable control rods must be fully inserted . (See attached, TS 3.4) .
* 7
* 6 .... "Provide a revision to the proposed TS describing the geometric limitation and the controls that will help ensure compliance and include information on the acceptability of the specific location where the 12.5% fuel elements will be installed. Otherwise, describe how the current or previously proposed TS adequately address the control of this geometric limitation on the location [of] 12.5 wt% fuel elements in the proposed reactor core."
MCNP results transmitted in the initial LAR and in previous RAI responses indicate that 12.5 wt% (max.)
fuel in the E- and F-rings will be kept below the fission heating density (i.e., power) of the 9.0 wt% (max.)
elements in the B-ring, even if placed adjacent to another 12.5%-loaded element. The additional restriction to avoided placing the 12.5% elements adjacent to control rod channels was intended to avoid conflict with the NRC over local power peaking during pulsing, although results from MCNP calculations performed at KSU have not shown reason for concern. The proposed TS have been revised to include a specific list of locations in which 12.5% fuel cannot be located. (See attached). In order to ensure that the 12% fuel is only added to locations where it permitted, the fuel handling procedures will be revised to include guidance to check the Technical Specifications list of approved locations prior to loading 12% fuel.
I swear under penalty of perjury that the foregoing is true and correct.
* Regards, Nuclear Reactor Facilities Manager Department of Mechanical and Nuclear Engineering Kansas State University Manhattan, KS 66506 Phone: {785)532-6657 {W)
{785)236-0602 (C)
Fax:     {785)532-7057 Email: geuther@ksu.edu Attachments (5):           GA Drawing TOS21E106 -Top Grid Plate SAR Ch. 4 markup copy SAR Ch. 4 clean copy TS markup copy TS clean copy 8


===4.1 Summary===
  ,- .391 D!A. (2.H.QLES)    i
Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.
!                            l
The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations.
                          - 'L
Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems. The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:
: 4.      Reactor Description 4.1     Summary Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.
The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations. Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems.
The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:
* Central thimble
* Central thimble
* Rotary specimen rack
* Rotary specimen rack
* Thermalizing column with bulk shielding tank
* Thermalizing column with bulk shielding tank
* Thermal column with removable door *Beam ports Radial (2) Piercing (fast neutron) (1) Tangential (thermal neutron) (1) The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion.
* Thermal column with removable door
All cooling is by natural convection.
                *Beam ports Radial (2)
The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated graphite.
Piercing (fast neutron) (1)
A fourth control rod would be used for 1,250-kW operation.
Tangential (thermal neutron) (1)
A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m 3 (144,000 ft.3) confinement building K-State Reactor Safety Analysis Report 4-1 Original (12/04)
The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.)
*
depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.)
* CHAPTER 4 made of r e inforced concrete and s tructural s t ee l , w ith co mpo s ite s h ea thin g and a luminum s idin g. Sectional views of the rea c tor are s h own in Figures 4.1 and 4.2. C ritic a lity was first achieved o n October 1 6 , 196 2 at 8:25 p.m. In 1 968 pul s in g ca p a bilit y was added and the ma x imum steady-s tate operating power was incr ease d from 100 kW to 25 0 kW. The original a luminum-clad fue l e l e m e nts we r e r e plac e d with s t a inl ess-s te e l clad e l e m e nt s in 1 973. Coo lant sys tem was r e pl ace d (and upgr a d e d in 2000), th e r eac tor operating co n so l e was replac e d , and the contro l room was e nlar ged and modernized in I 993 , w ith s upport from th e U.S. D e partm e nt of E n ergy. A ll n e utronic instrum enta tion was r e pl aced in I 994. North South Figure 4. 1, Vert ical Sect ion Through the KS U TRJGA Reactor. K-State Reactor Safety Analysis Report 4-2 Original (12/04)
diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated graphite. A fourth control rod would be used for 1,250-kW operation. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m3 (144,000 ft. 3) confinement building K-State Reactor                                 4-1                               Original (12/04)
* *
Safety Analysis Report
* REACTOR DESCRIPTION  
* CHAPTER 4 made of reinforced concrete and structural steel, with composite sheathing and aluminum siding.
 
Sectional views of the reactor are shown in Figures 4. 1 and 4.2.
===4.2 Reactor===
Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW.
Core The General Atom ic s TRIGA reactor design began in 1956. The original de s ign goa l was a comp l e t ely and inherently safe reactor. Complete safety means that a ll the availab l e excess r eac ti v it y of the reactor can be instantaneously introduced without caus in g an accident.
The original aluminum-clad fue l elements were replaced with stainless-steel clad elements in 1973 . Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in I 993, with support from the U.S.
Inh erent safety means that a n incr ease in the temperature of the fuel immediately and a ut omatica ll y results in decreased reactivity through a prompt n egative temperature coefficient.
Department of Energy. All neutronic instrumentation was replaced in I 994.
These features were accomplished by using enriched uranium fuel in a zirconi um hydride matrix . I fT 8 ............ West ....... .. ... -. : . : .: -..... .-.-:-.""* *"" ! IT IP! ..... **: ," ... . .. . .. ... , .. .. .. . . * .. * ..... : "*. *.:*. :* .. : .:\ .. * .. * .*.*.::.*:.:-* ... * " (hi t), 0 q "0v.Lf' t 'Pu.>'i East Figure 4. 2, Horizontal Section Through the KSU TRJGA Reactor. T h e basic parameter providing the TRIGA s y s tem wit h a large safety factor in steady state and transient operat ion s is a prompt negative temperature coefficient , relatively constan t with temperature
North South Figure 4. 1, Vertical Section Through the KSU TRJGA Reactor.
(-0.0 1% iik/k°C). This coefficient i s a function of the fuel composition and core geometry.
K-State Reactor                               4-2                             Original (12/04)
As power and temperature increase , matrix changes ca u se a shift in the neutron e n e r gy s pectrum in th e fuel to higher energies.
Safety Analysis Report
T h e uranium exhib it s low er fission cross sect i on s for the higher ene r gy neutrons , thus co unt ering the power increase.
* REACTOR DESCRIPTION 4.2             Reactor Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident. Inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fuel in a zirconi um hydride matrix .
Therefore , fuel and clad temperature au tomatic a ll y limit operat ion of the reactor. K-State Reactor Safety Analysis Report 4-3 Original (12/04)
                                                                              * *"" ! IT ~
* *
                                                              ..... .-.- :-.""                   IP!
* CHAPTER4 It is more convenient to set a power level limit that is based on temperature.
                                                                      **: ,"       ~
The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120°F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response.  
                                                                        .*..*... .     .___...,...-.-.,-~
 
: *:~ ~ T~ 2.!* '". :
====4.2.1 Reactor====
                                                                      *.*.:*. :* . : .:\*.
Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted.
                                                                                        ~*
I fT 8 ~"I
                                                                                                              " (hi t), 0 q "0v.Lf' t 'Pu.>'i West                                                                                                            East Figure 4. 2, Horizontal Section Through the KSU TRJGA Reactor.
The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively constant with temperature (-0.0 1% iik/k°C). This coefficient is a function of the fuel composition and core geometry. As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase. Therefore, fuel and clad temperature automatically limit operation of the reactor.
K-State Reactor                                   4-3                                                 Original (12/04)
Safety Analysis Report
* CHAPTER4 It is more convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120°F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response.
4.2.1 Reactor Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted.
Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic.
Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic.
Additional advantages were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size. TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.l (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
Additional advantages were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size.
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material.
TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.l (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
The maximum temperature limits of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zri.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material. The maximum temperature limits of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zri.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.
: a. Dimensions and Physical Properties.
: a.       Dimensions and Physical Properties.
The KSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements.
The KSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements. Three instrumented aluminum-clad Mark II elements are still available for use in the core.
Three instrumented aluminum-clad Mark II elements are still available for use in the core. General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.)
General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.) hole drilled through the center of the active fuel section.
hole drilled through the center of the active fuel section. Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (I in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank. 1 Unless otherwise indicated, fuel properties are taken from the General Atomics report of Simnad [1980] and from authorities cited by Simnad. K-State Reactor Safety Analysis Report 4-4 Original (12/04)
Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (I in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank.
* *
1 Unless otherwise indicated, fuel properties are taken from the General Atomics report of Simnad [1980]
* REACTOR DESCRIPTION Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements.
and from authorities cited by Simnad.
They are clad in aluminum and have a graphite length of 55.88 cm (22 in.). Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility.  
K-State Reactor                                     4-4                               Original (12/04)
'Property Dimensions Outside diameter, Do= 2ro Inside diameter, D;= 2r; Overall length Length of fuel zone, L Length of graphite axial reflectors End fixtures and cladding Cladding thickness Burnable poisons Uranium content Weight percent U mu enrichment percent mu content Physical properties of fuel excluding cladding H/Zr atomic ratio Thermal conductivity (W cm-1 K-1) Heat capacity [T (J cm-3 K-1) Mechanical properties of delta phase U-ZrH 0 Elastic modulus at 20&deg;C Elastic modulus at 650&deg;C Ultimate tensile strength (to 650&deg;C) Compressive strength (20&deg;C) Compressive yield (20&deg;C) *source: Texas SAR [1991]. b. Composition and Phase Properties Mark II 1.47 in. (3.7338 cm) 1.41 in (3.6322 cm) 28.4 in. (72.136 cm) 14 in. (35.56 cm) 4 in. (10.16 cm) aluminum 0.030 in. (0.0762 cm) Sm wafers 8.0 20 36 g 1.0 0.18 Mark Ill 1.47 in. (3.7338 cm) 1.43 in. (3.6322 cm) 28.4 in. (72.136 cm) 15 in. (38.10 cm) 3.44 in (8.738 cm) 3 04 stainless steel 0.020 in. (0.0508 cm) None 8.5 20 38 g 1.6 0.18 2.04 + 0.00417T 9.1 x 10 6 psi 6.0 x 10 6 psi 24,000 psi 60,000 psi 35,000 psi The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% mu, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature.
Safety Analysis Report
For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760&deg;C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen.
* REACTOR DESCRIPTION Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of 55.88 cm (22 in.).
Over 25,000 pulses have been performed with the TRIGA fuel elements at General Atomic, with fuel temperatures reaching peaks of about l 150&deg;C. K-State Reactor Safety Analysis Report 4-5 Original (12/04)
Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility.
* *
  'Property                                           Mark II                Mark Ill Dimensions Outside diameter, Do= 2ro                         1.47 in. (3.7338 cm)    1.47 in. (3.7338 cm)
* CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455&deg;C. From 455&deg;C to about 1050&deg;C, the delta phase is supported by a broadening range of H/Zr ratios. 2837" I K-State Reactor Safety Analysis Report STAINLESS STEEL . fr TOP END-FIXTURE j I I , ., .. 0,, I STAINLESS STEEL BOTTOM ENO-FIXTURE 1 20MIL STAINLESS STEEL CLAD ZIRCONIUM HYDRIDE-8*5WT% URANIUM, 20%ENR., 38 235u 1:43"DIA Figure 4.3, TRIGA Fuel Element. 4-6 l
Inside diameter, D;= 2r;                           1.41 in (3.6322 cm)    1.43 in. (3.6322 cm)
* l t 15" \' l .i ; !' . Orig in al ( 12/04)
Overall length                                   28.4 in. (72.136 cm)    28.4 in. (72.136 cm)
* *
Length of fuel zone, L                             14 in. (35.56 cm)      15 in. (38.10 cm)
Length of graphite axial reflectors               4 in. (10.16 cm)        3.44 in (8.738 cm)
End fixtures and cladding                         aluminum                3 04 stainless steel Cladding thickness                               0.030 in. (0.0762 cm) 0.020 in. (0.0508 cm)
Burnable poisons                                   Sm wafers              None Uranium content Weight percent U                                   8.0                    8.5 mu enrichment percent                             20                      20 mu content                                         36 g                    38 g Physical properties offuel excluding cladding H/Zr atomic ratio                                 1.0                    1.6 Thermal conductivity (W cm- 1 K- 1)               0.18                    0.18 Heat capacity [T ~0&deg;C] (J cm-3 K- 1)                                       2.04 + 0.00417T Mechanical properties ofdelta phase U-ZrH0 Elastic modulus at 20&deg;C                                                   9.1 x 106 psi Elastic modulus at 650&deg;C                                                   6.0 x 106 psi Ultimate tensile strength (to 650&deg;C)                                       24,000 psi Compressive strength (20&deg;C)                                               60,000 psi Compressive yield (20&deg;C)                                                 35,000 psi
  *source: Texas SAR [1991].
: b.     Composition and Phase Properties The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% mu, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760&deg;C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen. Over 25,000 pulses have been performed with the TRIGA fuel elements at General Atomic, with fuel temperatures reaching peaks of about l 150&deg;C.
K-State Reactor                               4-5                               Original (12/04)
Safety Analysis Report
* CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455&deg;C. From 455&deg;C to about 1050&deg;C, the delta phase is supported by a broadening range of H/Zr ratios.
                                              .fr      STAINLESS STEEL TOP END-FIXTURE
                                              ~--~-=~.~
20MIL STAINLESS STEEL CLAD l
ZIRCONIUM HYDRIDE-2837"                                    8*5WT%         -~
URANIUM, l
20%ENR.,
j I                                        38 235u t
                                                                      -~
15" I                                        \'
l 1:43"DIA .i; I,.,. 0,,                              .!'
I STAINLESS STEEL
                                            ~
BOTTOM ENO-FIXTURE    1 Figure 4.3, TRIGA Fuel Element.
K-State Reactor                                4-6                             Orig in al ( 12/04)
Safety Analysis Report
* REACTOR DESCRIPTION
* REACTOR DESCRIPTION
: c. Core Layout A typical layout for a KSU TRIGA Il 250-kW core (Core II-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted .. Figure 4.4, Core Layout (250 kW). The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor Safety Analysis Report 4-7 Original (12/04)
: c. Core Layout A typical layout for a KSU TRIGA Il 250-kW core (Core II-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted ..
* *
Figure 4.4, Core Layout (250 kW).
* C H APTER 4 4.2.2 Contro l Rods The pulse rod is 3.1 75 cm. ( 1.25 in.) diam ete r. Other rods are 2.225 cm (7/8 in.) diamete r. Co ntr o l rods are 50.8 cm. (20 in.) l ong boron carbide or borated grap hit e , c l ad with a 0.0762 cm. (30*m il) aluminum sheath. The control rod drives are connected to the contro l rod c lutch es through thr ee extens ion s h afts. Th e clutch and upper extension s haft for standa rd rods ex tend through an assembly designed with s lot s that provides a h ydrau lic c u hi on (or buffer) for th e rod during a scram , and a l so limit s the bottom position of th e contro l rods so th a t the y do n ot imp act the bottom of the co ntrol rod g uid e tube (in th e core). Th e buffers for two s tandard rods are s h own in th e l eft hand picture below (s l otted tubes on the right hand side) a l ong wi th the top sec tion of the pulse/transient rod extens i on. The pulse rod drive c lut c h connects to a so lid extension shaft through a pneumatic cylinder; the dimensions of the cy lind er limit s bottom tra ve l. U pp e r Pulse , S him & R eg Rods R eg Rod S him Rod Pul se Rod Fig u re 4.5, Co n tro l Rod U pper Extension Assemb li es The bottom of the pul se rod is s h ow n on th e left hand side of Fig ure 4.5. The upper ex t e n sion s h aft is a hollow tube , the midd l e extension i s so lid. The upp er ex t e n s ion shaft is connected to th e middle exte n s ion shaft with l ock wire or a pin and l ock wire fo r standard rods , with a bolt ed co ll a r for the pulse rod (the mechanica l shock during a pulse requires a more sturdy fasten e r). Sec urin g the upper contro l rod exte n s i on to th e middle ex t ension a t o n e of seve ral holes drilled in the upper part of th e middle extension (Figu re 4.6) provide s adjustment for the control rod s nece ssary to ensure the control rod full in po s ition is above th e bottom of the gu id e tube. K-State Reactor Safety Analysis Report 4-8 Original (12/04)
The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor                                   4-7                           Original (12/04)
* *
Safety Analysis Report
* REACTOR DESCRIPTION Figure 4.6, Middle Extension Rod Alignment Hole s T h e middl e so lid exte n s ion is s imilarly connected to th e l ower exte n s i o n. The l owe r ex t e n s ion is holl ow , the middle extension fits into the l ower ex t e n s i on and a h o l e drill e d in the ove rl ap sec ur es th e low e r ex t e n s ion to th e middle ex t e n s ion. T y picall y th e l owe r exte n s i on h as a tighter fit than th e upp er exte n s i on becau se the l owe r and middle ex t e n s ion are not se par a t ed fo r in spec ti o n s and b eca u se th e int e r face wi th uppe r ex t e n sion i s u se d t o se t the bottom p os ition of the contro l rod. Pi ct ur es of the low e r connecto r fo r th e pul se rod a nd o n e s tandard rod a re s h own at th e l ef t in F i g ur e 4.7 .. Figure 4.7 , Standard & Pul se Rod Lower Coupling The bott om of the low er ex t e n s ion attache s direct l y to th e con tr o l rod. Pi ctures of th e contro l rod s tak e n duri n g the 2003 co ntrol rod in spec ti on are in Figure 4.8. The rod s m ove with in co nt ro l rod g uid e tubes , s hown in Fig ur e 4.9. The g uid e t ub es h ave perforated wa ll s. The g uid e tubes h ave a s m a ll m eta l wi r e in the tip that fits int o th e l ower grid plate; a se t sc r ew inside th e bott o m of th e g uide t ub e pushe s the w ir e aga in s t th e l owe r gr id pl a t e t o sec ur e th e g uid e tube. K-State Reactor Safety Analysis Report 4-9 Original (12/04)
* CHAPTER 4 4.2.2 Control Rods The pulse rod is 3. 175 cm. ( 1.25 in.) diameter. Other rods are 2.225 cm (7/8 in.) diameter.
* *
Control rods are 50.8 cm . (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm .
* CHAPTER 4 Pulse Rod Shim Rod Re g Rod Figure 4.8, Co ntrol Rods During 2003 In s pection K-State Reactor Safety Analysis Report Fu ll Gu id e Tube Figure 4.9 , Control Rod Guide Tubes 4-10 Origina l (12/04)
(30*m il ) aluminum sheath.
* *
The control rod drives are connected to the control rod clutches through three extension shafts.
The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydrau lic cu hi on (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (s lotted tubes on the right hand side) along wi th the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder; the dimensions of the cylinder limits bottom trave l.
Upper Pulse, Shim & Reg Rods           Reg Rod         Shim Rod               Pul se Rod Figure 4.5, Control Rod Upper Extension Assembli es The bottom of the pulse rod is shown on the left hand side of Figure 4.5 . The upper extension shaft is a hollow tube, the midd le extension is solid. The upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fasten er).
Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full in position is above the bottom of the gu ide tube.
K-State Reactor                                   4-8                               Original (12/04 )
Safety Analysis Report
* REACTOR DESCRIPTION Figure 4.6, Middle Extension Rod Alignment Holes The middle solid extension is similarly connected to the lower extension. The lower extension is holl ow, the middle extension fits into the lower extension and a hole drill ed in the overl ap secures the lower extension to the middle extension. Typically th e lower extension has a tighter fit than the upper extension because the lower and middle extension are not separated for inspecti ons and because the interface wi th upper extension is used to set the bottom position of the control rod.
Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7 ..
* Figure 4.7, Standard & Pulse Rod Lower Coupling The bottom of the lower extension attaches directly to the control rod . Pictures of the control rods taken duri ng the 2003 control rod inspecti on are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wi re in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube.
K-State Reactor                                   4-9                               Original (12/04)
Safety Analysis Report
* CHAPTER 4 Pulse Rod             Shim Rod           Reg Rod Figure 4.8, Control Rods During 2003 Inspection
* Fu ll Gu ide Tube Figure 4.9, Control Rod Guide Tubes K-State Reactor                          4-10                     Origina l (12/04)
Safety Analysis Report
* REACTOR DESCRIPTION
* REACTOR DESCRIPTION
: a. Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings.
: a.     Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings. A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (I) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.
A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (I) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch. Table 4.2, Summary o re ontrol Rod Interlocks INTERLOCK SETPOINT FUNCTION/PURPOSE Inhibit standard rod motion if nuclear Source Interlock 2 cps instrument startup charmel reading is less than instrument sensitivity/ensure nuclear instrument startup charmel is operating Pulse Rod Interlock Pulse rod inserted Prevent applying power to pulse rod unless rod inserted/prevent inadvertent pulse Multiple Rod Withdrawal Withdraw signal, Prevent withdrawal of more than 1 rod/Limit more than 1 rod maximum reactivity addition rate Pulse Mode Interlock Mode switch in Hi Prevent withdrawing standard control rods in Pulse pulse mode Pulse-Power Interlock
Table 4.2, Summary ore ontrol Rod Interlocks INTERLOCK                 SETPOINT                       FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument startup charmel reading is less Source Interlock                      2 cps than instrument sensitivity/ensure nuclear instrument startup charmel is operating Prevent applying power to pulse rod unless Pulse Rod Interlock            Pulse rod inserted rod inserted/prevent inadvertent pulse Withdraw signal,       Prevent withdrawal of more than 1 rod/Limit Multiple Rod Withdrawal more than 1 rod       maximum reactivity addition rate Mode switch in Hi       Prevent withdrawing standard control rods in Pulse Mode Interlock Pulse            pulse mode Prevent pulsing if power level is greater than Pulse-Power Interlock                !OkW
!OkW Prevent pulsing if power level is greater than !OkW NOTE: (!)Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock reqmred by Technical Specifications
                                                                    !OkW NOTE: (!)Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock reqmred by Technical Specifications
: b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation.
: b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation.
K-State Reactor Safety Analysis Report 4-11 Original (12/04)
K-State Reactor                                   4-11                                       Original (12/04)
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Safety Analysis Report
* CHAPTER4 Table 4.3, Summary o Reactor f s CRAMs Measuring Limitin2 Trip Setpoint Steady Actual Setpoint Channel State Pulse Linear Channel High 110% NIA 104% Power Power Channel 110% NIA 104% High power Detector High 90% 90% 90% Voltage 600&deg;C B Ring element High Fuel 555&deg;C C Ring element 450&deg;C Temperaturef 1 l 480&deg;C D Ring element 380&deg;C E Ring element 350&deg;C Period [IJ NIA NIA 3 sec NOTE [l]: Penod tnp and temperature tnp are not reqmred by Techmcal Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin of at least $0.50 . Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications.  
* CHAPTER4 Table 4.3, Summary of Reactor CRAMss Limitin2 Trip Setpoint Measuring Steady                                   Actual Setpoint Channel                           Pulse State Linear Channel High         110%           NIA                           104%
 
Power Power Channel High power 110%           NIA                           104%
====4.2.3 Neutron====
Detector High 90%           90%                           90%
Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator.
Voltage 600&deg;C B Ring element High Fuel           555&deg;C C Ring element                       450&deg;C Temperaturef1l      480&deg;C D Ring element 380&deg;C E Ring element                       350&deg;C Period [IJ           NIA           NIA                           3 sec NOTE [l]: Penod tnp and temperature tnp are not reqmred by Techmcal Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin of at least $0.50 .
Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors.
Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications.
The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially.
4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors.
The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3.18-cm (1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector.
The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3.18-cm (1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector.
K-State Reactor Safety Analysis Report 4-12 Original (12104)
K-State Reactor                                 4-12                               Original (12104)
* *
Safety Analysis Report
* REACTOR DESCRIPTION The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank. Four lugs are provided for lifting the assembly.
* REACTOR DESCRIPTION The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank.
A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures.
Four lugs are provided for lifting the assembly. A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within the fuel cladding provide additional reflection. Inserts are placed at both ends of the fuel meat, providing top and bottom reflection.
Graphite inserts within the fuel cladding provide additional reflection.
4.2.4 Neutron Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 106 n s- 1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor.
Inserts are placed at both ends of the fuel meat, providing top and bottom reflection.  
4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates.
 
The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core.
====4.2.4 Neutron====
Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position. The reflector assembly supports both grid plates.
Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 10 6 n s-1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor. 4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates. The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures.
4.3     Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion. Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration. All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool K-State Reactor                                   4-13                               Original (12/04)
Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core. Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position.
Safety Analysis Report
The reflector assembly supports both grid plates. 4.3 Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.)
* CHAPTER4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak.
thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion.
A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations.
Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration.
Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter 10).
All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool K-State Reactor Safety Analysis Report 4-13 Original (12/04)
4.4     Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield.
* *
The shielding configuration is similar to those at other TRI GA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank .
* CHAPTER4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak. A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation.
The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena.
The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations.
4.5     Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels with 8.5% U, 20%
Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction.
enrichment, the value is nearly constant at 0.01 % L'i.k/k per &deg;C, and varies only weakly dependent on geometry and temperature.
Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector.
Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures that the fuel and clad temperature limits will not be exceeded. The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity.
The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter 10). 4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield. The shielding configuration is similar to those at other TRI GA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank . The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena.  
Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor                                   4-14                             Original (12/04)
 
Safety Analysis Report
===4.5 Nuclear===
* REACTOR DESCRIPTION limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6.
Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors.
4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures.
This coefficient is a function of the fuel composition, core geometry, and temperature.
4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates.
For fuels with 8.5% U, 20% enrichment, the value is nearly constant at 0.01 % L'i.k/k per &deg;C, and varies only weakly dependent on geometry and temperature.
For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor).
Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures that the fuel and clad temperature limits will not be exceeded.
Table 4.4, 250 kW Core Parameters.
The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity.
13 (effective delayed neutron fraction)                         0.007
Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor Safety Analysis Report 4-14 Original (12/04)
                &#xa3; (~ffective neutron lifetime)                                   43 :S
* *
                                                                              -$0.0I7 EC- 1 UTf (prompt temperature coefficient)                   (a'J 250kW -275EC av (void coefficient)                                     -0.003 I %- 1 void
* REACTOR DESCRIPTION limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6. 4.5.1 Design Criteria -Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5).
                                                                          -$0.006 kW- 1 to -
The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity.
Up (power temperature coefficient -weighted ave)             $0.0I kw- 1 Table 4.5, Com arison of Control Rod Worths.
The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures.  
KSU TRIGA Mark II (250 kW)                             Cornell University Core II-19                       Core III-I                     (500 kW Pulse              D-10             $1.96             C-4           $2.I2           D-10          $1.88 Shim                C-3             $2.88             D-4           $1.85           D-I6        $2.20 Safety              NA               $0.0             D-16         $1.82           D-4          $1.99 Regulating        D-16             $1.58             E-I           $0.79           E-I          $0.58 TOTAL              NA               $6.42             NA           $6.58           NA          $6.65 NOTE: Core III-I has an experiment positioned to control the worth of the pulse rod K-State Reactor                                       4-15                               Original (12/04)
 
Safety Analysis Report
====4.5.2 Reactor====
* CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.
Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position.
4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.
For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates. For comparison purposes, a tabulation of total rod worth for each control element from the State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor).
Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRI GA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 1000&deg;C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470&deg;C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr* 1 ft*
Pulse Shim Safety Regulating TOTAL Table 4.4, 250 kW Core Parameters.
2 0 1 P- (2550 W m*2 K" 1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.
13 (effective delayed neutron fraction) 0.007 &#xa3; (
: a.       Spatial Power Distribution The following conservative approximations are made m characterizing the spatial distribution of the power during steady-state operations.
neutron lifetime) 43 :S -$0.0I7 EC-1 UTf (prompt temperature coefficient) (a'J 250kW -275EC av (void coefficient)  
* The hottest fuel element delivers twice the power of the average.
-0.003 I %-1 void -$0.006 kW-1 to -Up (power temperature coefficient -weighted ave) $0.0I kw-1 Table 4.5, Com arison of Control Rod Worths. KSU TRIGA Mark II (250 kW) Core II-19 Core III-I D-10 $1.96 C-4 $2.I2 C-3 $2.88 D-4 $1.85 NA $0.0 D-16 $1.82 D-16 $1.58 E-I $0.79 NA $6.42 NA $6.58 Cornell University (500 kW D-10 $1.88 D-I6 $2.20 D-4 $1.99 E-I $0.58 NA $6.65 NOTE: Core III-I has an experiment positioned to control the worth of the pulse rod K-State Reactor Safety Analysis Report 4-15 Original (12/04)
Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and R: as the physical radius and the extrapolation distance) is given 2 by:
* *
F" =     1.202 * (%.)
* CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions.
                                                  .. J, [ 2.4048 *(%.)J with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,. However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element.
A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.
2 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3 K-State Reactor                                   4-16                             Original (12/04)
4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.
Safety Analysis Report
Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRI GA fuel and cladding temperatures associated with pulsing operations.
* REACTOR DESCRIPTION Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40.
The analysis addresses the case of a fuel element at an average temperature of 1000&deg;C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470&deg;C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr*1 ft* 2 0 P-1 (2550 W m*2 K" 1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected.
Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.
This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.
: a. Spatial Power Distribution The following conservative approximations are made m characterizing the spatial distribution of the power during steady-state operations.
* The hottest fuel element delivers twice the power of the average. Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and R: as the physical radius and the extrapolation distance) is given 2 by: F" = 1.202 * (%.) .. J, [ 2.4048 * (%.) J with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,.
However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element. 2 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3 K-State Reactor Safety Analysis Report 4-16 Original (12/04)
* *
* REACTOR DESCRIPTION Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40. Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.
* The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of rr/2 times the average, and heat conduction radial only.
* The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of rr/2 times the average, and heat conduction radial only.
* The axial factor for power produced within a fuel element is given by: g(z)=l.514*co(%*
The axial factor for power produced within a fuel element is given by:
2*/ C ), ' + ext (6) in which &#xa3;=LI 2 and&#xa3;,,, is the extrapolation length in graphite, namely, 0.0275 m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor. Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.
g(z)=l.514*co(%* */ C ),                                             (6) 2
The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore:
                                                    '           + ext in which &#xa3;=LI 2 and&#xa3;,,, is the extrapolation length in graphite, namely, 0.0275
: m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor.
Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.
* The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore:
(7)
(7)
* The radial and axial distribution of the power within a fuel element is given by q"'(r,z) =
* The radial and axial distribution of the power within a fuel element is given by q"'(r,z) = q;~J(r)g(z),                                    (5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core.
(5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core. 3 GA-4361, Calculated Fluxes and Cross Sections for TRIG A Reactors (8/14/J 963), G. B. West K-State Reactor Safety Analysis Report 4-17 Original (12/04)
3 GA-4361, Calculated Fluxes and Cross Sections for TRIG A Reactors (8/14/J 963), G. B. West K-State Reactor                                 4-17                                 Original (12/04)
* *
Safety Analysis Report
* CHAPTER4 a.
* CHAPTER4
* The radial factor is given by: 2 f(r)= a+cr+er , I +br+dr 2 (7) in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679.
* The radial factor is given by:
The fit is illustrated in Figure 4.11. 1.3 1.2 1.1 L 1.0 <:' 0.90 0.80 0.70 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0 r (cm) Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod. (Data Points from Monte Carlo Calculations
2 f(r)= a+cr+er ,                                       (7)
[Ahrens 1999a]) Heat Transfer Models The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the LI.T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron), cladding inner surface and fuel outer surface -gap (g), and the fuel outer surface to centerline (ncl): Eq. 1 A standard heat resistance model for this system is: K-State Reactor Safety Analysis Report 4-18 Original (12/04)
I +br+dr 2 in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679.
* *
The fit is illustrated in Figure 4.11.
* REACTOR DESCRIPTION T = T +q"l_!_+ r, In(;{) +l+lj 0' ' h k rh 2k c ' g f Eq. 2 and heat flux is calculated directly as: "=Ul1T= T--'I'i, q 1 r ln(r I r) r 0 r ' -+ 0 0 I +-+-O-(2) h kc rA 2k 1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k1 is the fuel thermal conductivity.
1.3 1.2 1.1 L   1.0 0.90 0.80 0.70 0.0 0.20 0.40 0.60 0.80   1.0   1.2 1.4 1.6 1.8   2.0 r (cm)
The gap conductivity of 2840 W m-2 K" 1 (500 Btu h-1 ft -2 &deg;F-1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context. Parameters are cross-referenced to source in Table 4.6. Table 4.6: Thermodynamic Values Parameter Symbol Value Units Reference Fuel conductivity kr 18 Wm-1 K-1 Table 13.3 14.9 W m*1 K-1 (300 K) Table 13.3 Clad conductivity kg 16.6 W m-1 K-1 (400 K) Table 13.3 19.8 W m*1 K" 1 (600 K) Table 13.3 Gao resistance
Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod.
: h. 2840 wm-2 K-1 Appendix A Clad outer radius ro 0.018161 M Table 13.1 Fuel outer radius fj 0.018669 M Table 13.1 Active fuel length Lr 0.381 M Table 13.1 No. fuel elements N 83 NIA Chap 13 Axial peaking factor APF nl2 NIA Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that: l = 5.1858E-04 m'K 2kf vv Gap resistance has been experimentally determined as indicated, so that: K-State Reactor Safety Analysis Report l=3.6196E-04 m'K r,h, w 4-19 Original (12/04)
(Data Points from Monte Carlo Calculations [Ahrens 1999a])
* *
: a. Heat Transfer Models The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the LI.T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron),
* CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127&deg;C applies so that: r In':!._ 0 r m'K --' =3.103e k, w Table 4.7, Cladding Heat Transfer Coefficient Temp (&deg;K) Temp (0 C) m 2 Kw*1 300 27 3.457e-5 400 127 3.103e-5 600 327 2.60le-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small. The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.
cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ncl):
85000 f 75000 E 65000 " *;:; 55000 " 45000 " /:!. m 35000 J: 25000 500 700 Convection Heat Transfer Coefficient 900 1100 1300 1500 Power Level (KW) TRENCLINE:
Eq. 1 A standard heat resistance model for this system is:
y = 0.0326x + 16985 R 2=09976 1700 19'.lO Figure 4.10, Convection Hear Transfer Coefficient versus Power Level K-State Reactor Safety Analysis Report 1 h 0.0326P(watts)  
K-State Reactor                                   4-18                             Original (12/04)
+ 16985 4-20 Original (12/04)
Safety Analysis Report
* *
* REACTOR DESCRIPTION T 0
* REACTOR DESCRIPTION Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as: T =T +0.423P[ 1 +3.103e-5+3.620e-4+5.186e-4]
                                  =T
(10) <, ' 0.0326P + 16985 c. Steady-State Mode of Operation Centerline t+emperature calculations were performed on a "reference core" using model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation.
                                        +q"l_!_+ In(;{) +l+lj h
A core with more than 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 2occ and I oocc was assumed with the following results: Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel Fuel/Gap Gap/Clad Clad/Water Bulk Water cc' Centerline cc Interface cc Interface cc Interface cc 503.2 229.0 37.7 21.2 20.0 582.0 307.8 116.4 100.0 100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 cc at power levels up to 1900 kW with pool temperature at 27 cc (300 K), and 1700 kW with pool temperatures at 100 cc. K-State Reactor Safety Analysis Report 4-21 Original (12/04)
r, k c rh
*
                                                                            ' g 2kf Eq. 2 and heat flux is calculated directly as:
* CHAPTER 4 Hot Fuel-Rod Centerline Temperature at Power (Temperature Bevation over Pool Water Temperature) l----*300K  
                                "=Ul1T=                 T--'I'i,                                   (2) q             1 r ln(r I r)
--SOOK I 100 300 500 700 900 1100 1300 1500 1700 1900 Reactor Power (kW) Figure 4.11, Hot Fuel-Rod Centerline Temperature The margiH te eritisal heat fhm fer the refereHee eere v*as determiHed.
                                            -+     0     0 I +-+-O-r0 r '
Critieal heat flHx fer sat\lfated peel beiliHg is giveH by (Heat TFansfer, A. BejaH, 1993, Jeh.H Wiley & (3) wherefio is the density ef the fluid, pg-is the density of the vapor, e: is the surfaee tension of the liquid phase in sontaet *with vapor, is the eHthalpy of the sat\lfated fluid, and hg;sa< is the enthalpy ef the vaper phase with all values at sat\lfatieH eeHditiens ef temperatme and pressure.
h           kc         rA     2k1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k1 is the fuel thermal conductivity. The gap conductivity of 2840 W m-2 K" 1 (500 Btu h- 1 ft -2 &deg;F- 1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context.
Smfaee teHsioH data provided by Bejan was fit to a polyHomiaJ (usiHg temperatHre iH &deg;C) to geHerate data fer the temperature raHge of iHterest, Veith an Ri value of 0.999998:
Parameters are cross-referenced to source in Table 4.6.
o 1.000E-11
Table 4.6: Thermodynamic Values Parameter           Symbol       Value                 Units         Reference Fuel conductivity             kr           18             Wm- 1 K- 1       Table 13.3 14.9       W m* 1 K- 1 (300 K)   Table 13.3 16.6       W m- 1 K- 1 (400 K)   Table 13.3 Clad conductivity            kg 19.8       W m* 1 K" 1 (600 K)     Table 13.3 Gao resistance               h.         2840             wm-2 K- 1       Appendix A Clad outer radius             ro     0.018161                 M           Table 13.1 Fuel outer radius             fj     0.018669                 M           Table 13.1 Active fuel length           Lr       0.381                 M           Table 13.1 No. fuel elements             N           83                 NIA             Chap 13 Axial peaking factor       APF           nl2                 NIA           Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that:
* T 4 + 7.370E-09
l     = 5.1858E-04 m'K 2kf                           vv Gap resistance has been experimentally determined as indicated, so that:
* T 1 -1.969E-06
l=3.6196E-04 m'K r,h,                       w K-State Reactor                                  4-19                               Original (12/04)
* T' + 4.709E -06
Safety Analysis Report
* T + 7.1833E -02 Pressure at the eere is determiHed by baremetrie pressllfe at the faeility elevatieH, vaeuum maiHtaiHed iH the reaster bay aHd the 'Neight sf the water S\'er the sere. Baremetris pressllfe asseeiated with the Manhattan, Kansas airpert is 29.92 in. Mg. The reaeter bay is maiHtaiHed at a slight vaeUHm, with the maximum gage pressure (a iH, ef water) eerrespeHdiHg te 0.4 4 iH Mg; HemiHal baremetrie pressure eerreeted fer mai<imHm reaster bay vaeuum (a ehange ef apprmdmately 1.5%) eerrespeHds te 99.83 kPa. VariatieHs in leeal baremetrie pressure are en the erder ef the serrestieH fer reaeter bay K-State Reactor Safety Analysis Report 4-22 Original (12/04)
* CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127&deg;C applies so that:
* *
r0  In':!._
* REACTOR DESCRIPTION For Stieeooled eoiliag, the eritieal heat fhm is ealetilated ey (Ivey and Morris 1978): x ( ) P1 cp,f. TSAT -Tsub sub = SAT hg,sat -hf.sat +ehle 4.9, Critieel Heat Fill* Ratios (GHF versus Maidmum Heat FIU*) fer 13 & Hi Feet efWeteF OveF the Cere 4-a :w w 4G 4a w aa eg ea +G +a w ge w K-State Reactor Safety Analysis Report ft) &.4S &.-=1-S a,w 4.-W 4.-99 44-0 2-,94-(rn ft) &,g4. 9,.74. &.-M &.9e &.@ &..97 ,4.&% &.-00 &.-W 4-+& 4,-74 444 44&sect; 4-23 Original (12/04)
r                   m'K
*
                                                - - ' =3.103e                                                     k,                         w Table 4.7, Cladding Heat Transfer Coefficient Temp (&deg;K)           Temp (0 C)               m2 Kw* 1 300                   27                 3.457e-5 400                   127                 3.103e-5 600                   327                 2.60le-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small.
* CHAPTER 4 Table 4.9, Critieal Heat flux Raties (CHf versus Maximum Heat flux) fer B & Hi Feet efWater Over the Cere CFIFR (19 ft) As iHElieateEI iH Table 4 .9, the aernal heat flim is less thaH the eritieal heat flHJ< fer operatiHg temperatHres 1:1p to 55 &deg;C by more than a faetor of 4 eoHsiEleriHg both 13 feet anEI I e feet of water above the sore. The CHFR is greater than 2 for pool temperat1:1res ei<eeeEliHg the maidffitlm operatiHg val1:1e tlfl to 95 &deg;C, aHEI remaiHs Hear 2 at val1:1es tlfl to 99&deg;C. The Eliffenmce iH the eritical heat flt1>< ratio for 13 anEI I e feet of water is relatiwly small, '<Vith a miHim1:1m EliffereHee comp area to the mean of the tvro val1:1es of I.&% anEI a maxim1:1m of3.&e% belmv eO &deg;C, e.4% aeross all pool temperattlfes eoHsiElereEI.
The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.
It is elear from the table that there is a very *.viEle margiH betweeH the eperatiHg heat flm< anEI the eritieal heat flim eveH te t1Hrealistieally high pool *.vater temperatHre, se that film boiliHg aHEI eJ!Sessive elaEIEliHg temperatHre is Hot a eoHsiEleratieH iH steaEly state /( Formatted
Convection Heat Transfer Coefficient TRENCLINE:
[  
y = 0.0326x + 16985 R2 =09976 85000 f       75000 E
//( Formatted
              ~
[ // /,( Formatted
              ~ 65000
[
              "~
f //( Formatted
              ~    55000
[ /*,/>=====================
            "~    45000
operatioH.
              /:!."
1 if!( F tt d [ &#xa3;or the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD 3.3* //i/'>=o=r=m=a=e==================
m J:
oatcli-04 (Feldman  
35000 25000 500       700       900           1100         1300       1500 1700              19'.lO Power Level (KW)
'i*ii//l/ft[:=F=o=rm=att=ed=================[
Figure 4.10, Convection Hear Transfer Coefficient versus Power Level 1
dependent volumes, enforcing  
h     0.0326P(watts) + 16985 K-State Reactor                                          4-20                                   Original (12/04)
?oun?ary conditions, and tyvo pi12es. ?imulating the !!//if!//!!,[
Safety Analysis Report
Formatted
* REACTOR DESCRIPTION Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as:
[ col.d a_ n_ -d-*hot_ ch_ ann_ connec_ -t-ed_ via a __ s_ mg-le rnn __ ct10n .c_
1 T =T +0.423P[                           +3.103e-5+3.620e-4+5.186e-4]                 (10)
o_f ___
          <,   '             0.0326P + 16985
__ , __ H_e_a __ t 1 __ s_ a_d ___ de_ cl to_ t?dj 1 /J i/////(>=F=o=r=m=a=tt=e=d==============='=[
: c. Steady-State Mode of Operation Centerline t+emperature calculations were performed on a "reference core" using t~
flrud by the heat,.structure com12onent (su:iulatmg .a fuel element) of RELAP '":1th]//  
model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation. A core with more than 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 2occ and I oocc was assumed with the following results:
/fi!i/f>=====================-
Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel         Fuel/Gap     Gap/Clad Clad/Water Bulk Water cc' Centerline cc Interface cc Interface cc Interface cc 503.2         229.0         37.7         21.2           20.0 582.0         307.8         116.4         100.0         100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 cc at power levels up to 1900 kW with pool temperature at 27 cc (300 K), and 1700 kW with pool temperatures at 100 cc.
an appropnate ,a)(Ial_
K-State Reactor                               4-21                               Original (12/04)
profile and power le\lel. In this an __ alys1s, the 120_ we __ r_ level_ for_ the B n __ ngj//
Safety Analysis Report
is at 24 kW (corresponding to an 85-element core with <ving-to-averagepeaking factor o_f 1_._6JJJ/  
* CHAPTER 4 Hot Fuel-Rod Centerline Temperature at Power (Temperature Bevation over Pool Water Temperature) l----*300K --SOOK                       I o+tt1::t::tt:l:tt1:ttt1::t::tt:l:tl:1::t::tt:t:ttl:t&#xb5;1::t::ttjjjjt:t:tttllitttj::tttjjjj::ttj:tlli:tttt::tj::t::t:ttl:tttt:j:tt:t:ttl:tttj::ttt~
//ff!/;/,[
100          300            500            700          900            1100          1300          1500          1700          1900 Reactor Power (kW)
Formatted
Figure 4.11, Hot Fuel-Rod Centerline Temperature The margiH te eritisal heat fhm fer the refereHee eere v*as determiHed. Critieal heat flHx fer sat\lfated peel beiliHg is giveH by (Heat TFansfer, A. BejaH, 1993, Jeh.H Wiley &
[ '[hisp?\Ver level is a.1mlied to the heat struct,ure  
        ~
\Vi thin the single ch:mnel. Themodel assmpes _ _an i operatmgpressure of 143 kPa. and an operatmgtemperature of322.15 K (49.15&deg;C).
(3) wherefio is the density ef the fluid, pg- is the density of the vapor, e: is the surfaee tension of the liquid phase in sontaet *with vapor, ~ is the                                       eHthalpy of the sat\lfated fluid, and hg;sa< is the enthalpy ef the vaper phase with all                                         values at sat\lfatieH eeHditiens ef temperatme and pressure. Smfaee teHsioH data                                               provided by Bejan was fit to a polyHomiaJ (usiHg temperatHre iH &deg;C) to geHerate                                           data fer the temperature raHge of iHterest, Veith an Ri value of 0.999998:
1!//f/iii/>F=o=r=m=a=tt=e=d==============="=[
o     1.000E-11
i//!Ji//( f'i/1 1/ji 1>F=o=r=m=a=tt=e=d==============="=[
* T 4 + 7.370E- 09
Jhe version of_ the _Rl'::C:,1\P code licensed to KSU uses PG-CHF correlation which is a state of the 1!!/!/;jl i ( Formatted
* T 1 -1.969E- 06
[ art best estimate CHF correlation developed }Juclear Rese_arch institute of Rez in the Czech_J/:/f;///>====================='=
* T' + 4.709E - 06
Republic.
* T + 7.1833E -02 Pressure at the eere is determiHed by baremetrie pressllfe at the faeility elevatieH, vaeuum maiHtaiHed iH the reaster bay aHd the 'Neight sf the water S\'er the sere. Baremetris pressllfe asseeiated with the Manhattan, Kansas airpert is 29.92 in. Mg. The reaeter bay is maiHtaiHed at a slight vaeUHm, with the maximum gage pressure (a iH, ef water) eerrespeHdiHg te 0.4 4 iH Mg; HemiHal baremetrie pressure eerreeted fer mai<imHm reaster bay vaeuum (a ehange ef apprmdmately 1.5%) eerrespeHds te 99.83 kPa.
It is based on datajn the Czech Republic data bank froml 73 diffe1ent sets o[tube 23 sets .of d.ata, an4_153 sets of rod; datli. T_liert: are four __
VariatieHs in leeal baremetrie pressure are en the erder ef the serrestieH fer reaeter bay K-State Reactor                                                       4-22                                                       Original (12/04)
_cif PG.-C::H_F'  
Safety Analysis Report
!if///:{ Formatted
* REACTOR DESCRIPTION For Stieeooled eoiliag, the eritieal heat fhm is ealetilated ey (Ivey and Morris 1978):
[ correlat10n Basic', Flux'. Geometry'.
P1    x         (
and Power'. For the rod bundle 1t ,)S ariphcable 1n pressure range of 0.28 MPa to 18.73 MPa. for a mass fluxof.34.1 to,7478 f()r0.4-}.0_!Il  
cp,f. TSAT -Tsub
/f / //
                                                                                )
length and for a diameter of 0.00241 to .0.07813 m. TRIGA has an operating pressure of 0.143 J/ (! /[ Formatted
sub =   SAT hg,sat - hf.sat
[ MPa and fuel rod length of 0.381 m, thus the 012erating conditions_
*                +ehle 4.9, Critieel Heat Fill* Ratios (GHF versus Maidmum
fall outside the range of the /,i/[>F=o=r=m=a=t=te=d================[
                  ~
applicability of Jhe PG-CHF correlation.
Heat FIU*) fer 13 & Hi Feet efWeteF OveF the Cere G~i;:g \~d  ft)  G~i;:g  (rn ft)      ~          ~
and .a diffe1ent correlation is reguired to asses,.s the JI-/ .departure from nucleate boiling ratio. (DNBRratio).
4-a           ~                &,g4.            9,.74.    ~
One such correlation.which is applicable for //(>F=o=r=m=a=t=te=d================[
:w             &.4S              ~                &.-M      ~
the low pressure range observed in TRIGA reactor -facility is the. Bernath correlation.
                    ~            &.-=1-S           ~                ~          ~
the Formatted
w            ~                &.9e            &.@        ~
[ functional form of the Bernath correlation.can be presented in the following eguati(_lns.
                    ~            a,w               &..97            ~          ,4.&%
_
4G            ~                ~                ~          ~
i>F=o=r=m=a=tt=e=d================[
4a            4.-W             &.-00            &.-W      ~
K-State Reactor Safety Analysis Report 4-24 Original (12/04) \\ ""':===================='=  
w            4.-99            4-+&             4,-74     ~
\ \. [ Formatted
aa            ~                ~                444       ~
[ \  
eg            44-0              ~                44&sect;       ~
\ *[ Formatted
ea            ~                ~                ~          ~
[ '[ Formatted
                    +G            ~                ~                ~          ~
[
                  +a            ~                ~                ~          ~
* *
w            2-,94-            ~                ~          ~
* REACTOR DESCRIPTION s Inlet I Hot-leg Cold-Leg connector Figure 4.12 -RELAP single channel model used in CHF analysis K-State Reactor Safety Analysis Report 4-25 Original (12/04) J Formatted:
ge            ~                ~                ~          ~
Centered, Keep with next I I ! I ! I I I *---( Formatted:
w            ~                ~                ~          ~
Caption, Centered *---Formatted:
K-State Reactor                                4-23                                   Original (12/04)
Centered, Don't adjust space between Latin and Asian text, Don't adjust space between Asiar text and numbers, Tab stops: 3", Centered + 5.5'', Rig t
Safety Analysis Report
*
* CHAPTER 4 Table 4.9, Critieal Heat flux Raties (CHf versus Maximum Heat flux) fer B & Hi Feet efWater Over the Cere CFIFR (19 ft)
* CHAPTER4 = , if Dh :::: 0. lft D" Dh Dh hso =film coefficient at CHF Dh =hydraulic diameter (ft) v =coolant velocity (ft Is) Twso =wall temperature at burnout (&deg; C) DH =heated diameter (ft) The RELAP simulations were performed for the hot channel. i.e., a channel with a radial factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain the pressure, temperature.
As iHElieateEI iH Table 4.9, the aernal heat flim is less thaH the eritieal heat flHJ< fer operatiHg temperatHres 1:1p to 55 &deg;C by more than a faetor of 4 eoHsiEleriHg both 13 feet anEI I e feet of water above the sore. The CHFR is greater than 2 for pool temperat1:1res ei<eeeEliHg the maidffitlm operatiHg val1:1e tlfl to 95 &deg;C, aHEI remaiHs Hear 2 at val1:1es tlfl to 99&deg;C. The Eliffenmce iH the eritical heat flt1>< ratio for 13 anEI I e feet of water is relatiwly
and velocity distribution at different axial locations.
                                                                                                                                                                /( Formatted                        [
With these calculations and the functional form of the Bernath correlation.
small, '<Vith a miHim1:1m EliffereHee comp area to the mean of the tvro val1:1es of I.&% anEI a                                                   //~=================='=
the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA. the A-ring is occupied by the central thimble. not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard. The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel. K-State Reactor Safety Analysis Report 4-26 Original (12/04) Formatted:
maxim1:1m of3.&e% belmv eO &deg;C, e.4% aeross all pool temperattlfes eoHsiElereEI.                                                                 //( Formatted                        [
Justified, Don't adjust space between Latir and Asian text, Don't adjust space between Asian text and numbers 
                                                                                                                                                              ///,( Formatted                      [
* *
It is elear from the table that there is a very *.viEle margiH betweeH the eperatiHg heat flm<                                                 /j/,[>F=o=r=m=a=tt=e=d=============~[
* REACTOR DESCRIPTION  
anEI the eritieal heat flim eveH te t1Hrealistieally high pool *.vater temperatHre, se that film boiliHg aHEI eJ!Sessive elaEIEliHg temperatHre is Hot a eoHsiEleratieH iH steaEly state                                                         f//( Formatted                       [
,, .. Bernatf1-CHF A
                                                                                                                                                            /*,/>=====================
* PG-CHF ,, " Heat flux l.L :c 2000 u 1000 8.oo *
operatioH.                                                                                                                                     1if!( F     tt d                       [
* Cl.OS * * * ... . . IUO 0.15 0.20 0.25 Cl.30 Heated Length (m) Figure 4.13 -CHF versus heated length 0::. o::l z + + Bernath-CHF  
  &#xa3;or the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD 3.3* //i/'>=o=r=m=a=e==================
<> <> PG-CHF 10 .........  
oatcli-04 (Feldman 2oil8i:.A~~illih2i:~[ilie_:-iPJl4t;JJ5Rl&deg;t!s_e!lie~II1Hi~eIJ1.!i)_~~Iw-9_:-ii~i= 'i*ii//l/ft[:=F=o=rm=att=ed=================[
* .. &#xa2;> <). &#xa2;> ' ;, B ..... . 0 6 4. 2
dependent volumes, enforcing th~ pre~sure ?oun?ary conditions, and tyvo pi12es. ?imulating the !!//if!//!!,[ Formatted                                                                           [
* 8.oo 0.0S 0.10 0.15 0.20 0.25 Heated Length (m) 0.30 Figure 4.14 -DNBR versus heated length K-State Reactor Safety Analysis Report 4-27 0.35 0.35 Original (12/04) l Formatted:
col.d a_n_-d-*hot_ ch_ ann_ ~I- connec_-t-ed_ via a__ s_ mg-le rnn__ ct10n .c_ ~m-12o_n._ent o_f_ _~L-~P- _,__H_e_a__ t 1_s_ a_d___ de_ cl to_ t?dj /J i/////(>=F=o=r=m=a=tt=e=d==============='=[
Centered, Keep with next I I I I I I I *--( Formatted:
1 flrud by IJ1~0EJJOJ&deg;~tmg the heat,.structure com12onent (su:iulatmg .a fuel element) of RELAP '":1th]// /fi!i/f>=====================-
Caption, Centered i I I / I ti Formatted:
an appropnate ,a)(Ial_ po~er profile and power le\lel. In this an_ alys1s, the 120_ we__r_ level_ for_ the B n                               __ngj// /i///i/,[>F=o=r=m=a=tt=e=d==============~[
Centered, Keep with next I I Formatted:
is at 24 kW (corresponding to an 85-element core with <ving-to-averagepeaking factor o_f 1_._6JJJ/ //ff!/;/,[ Formatted                                                                          [
Caption, Centered 
  '[hisp?\Ver level is a.1mlied to the heat struct,ure \Vi thin the single ch:mnel. Themodel assmpes __an i //fJ////(~=================='=
*
operatmgpressure of 143 kPa. and an operatmgtemperature of322.15 K (49.15&deg;C).                                                                       1!//f/iii/>F=o=r=m=a=tt=e=d==============="=[
* CHAPTER4 *------------------------------------------
i//!Ji//(
_ ___ _______ _____ __________
f'i/ 11/ji 1>F=o=r=m=a=tt=e=d==============="=[
__
Jhe version of_ the _Rl'::C:,1\P code licensed to KSU uses PG-CHF correlation which is a state of the 1!!/!/;jl i ( Formatted                                                                     [
Formatted:
art best estimate CHF correlation developed .~Y }Juclear Rese_arch institute of Rez in the Czech_J/:/f;///>====================='=
Font: 11 pt, Not Bold d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code TASCOT for transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. ((;.2) and (.'.710), with the peak fuel temperature set to 746 &deg;C, i.e., a temperature rise of 719 &deg;C above 27 &deg;C ambient temperature.
Republic. It is based on datajn the Czech Republic data bank froml 73 diffe1ent sets o[tube dataJ,jf;///;{>F=o=rm~att~ed~~~~~~~~~~~~=====[
The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1% ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation.
23 sets .of *a:uml~r d.ata, an4_153 sets of rod; bun~le_ datli. T_liert: are four__~Or_lJ1S _cif ~h(: PG.-C::H_F' !if///:{ Formatted                                                             [
In the TASCOT calculations, thermal conductivity was set to 0.18 W cm-1 K-1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm-1 K-1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994) The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955) f pf ( )0.99 [ k0.79
correlat10n Basic', Flux'. Geometry'. and Power'. For the rod bundle 1t ,)S ariphcable 1n the~ltj;; /[>====~==~=~=======~
* C0.45 *A 0.51 -hb = 0.00122
pressure range of 0.28 MPa to 18.73 MPa. for a mass fluxof.34.1 to,7478 kg/s~IlJ2. f()r0.4-}.0_!Il                                               /f / //       >F=o=r=m=a=tt=e=d=============~[
* 0.75
length and for a diameter of 0.00241 to .0.07813 m. TRIGA has an operating pressure of 0.143 J/ (! /[ Formatted                                                                                 [
* Tw -T,at , a 05 * &#xb5; 0.29
MPa and fuel rod length of 0.381 m, thus the 012erating conditions_ fall outside the range of the /,i/[>F=o=r=m=a=t=te=d================[
* p 0.24 * (v _ v )
applicability of Jhe PG-CHF correlation. and .a diffe1ent correlation is reguired to asses,.s theJI-/
* To.15 (9-10) f g g v sat in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 &deg;C), and n the coolant ambient temperature (27&deg;C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases. The overall heat transfer coefficient U varies negligibly for ambient temperatures from 20 to 60 &deg;C, and has the value 0.21 W cm-1 K-1 at Tb = 27 &deg;C. Figure 4.14,2 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.108 lists temperatures and heat fluxes as function of time after a 2.1 % ($3.00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m-1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500&deg;C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse. K-State Reactor Safety Analysis Report 4-28 Original (12/04) Formatted:
  .departure from nucleate boiling ratio. (DNBRratio). One such correlation.which is applicable for //(>F=o=r=m=a=t=te=d================[
Justified, Don't adjust space between Latir and Asian text, Don't adjust space between Asian text and numbers 
the low pressure range observed in TRIGA reactor -facility is the. Bernath correlation. the ':::~i Formatted                                                                                     [
* * * ,..., u 0 Ill L.. :J .&#xb5; l1l L.. Ill 0. E Ill I-1000 800 600 400 200 REACTOR DESCRIPTION Os 2 4 8 16 32 64s 0 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0 2.2 Radius (cm) Figure 4.
functional form of the Bernath correlation.can be presented in the following eguati(_lns. _ ---------:~ i>F=o=r=m=a=tt=e=d================[
Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse. K-State Reactor Safety Analysis Report 4-29 Original (12/04)
                                                                                                                                                      \\ ""':===================='=
*
                                                                                                                                                        \ \. [ Formatted                           [
* l CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{&deg;C) Coolant Ambient Temperature. Time (s) Q" CHFR Fuel outside Clad surface (Wm-2) Temp. (0 C) Temp. (0 C) 0 953 1 3.57 xl0 5 4.2 781 224 2 7.34 xl0 5 2.0 683 432 4 8-52 xl0 5 1.7 574 498 8 7.54 xl0 5 2.0 461 443 16 5.71 xl0 5 2.6 344 342 32 3.46 xl0 5 4.3 224 218 64 1.04 xl0 5 14.4 100 84 4.6 Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by (.Wll) in which Po and Po are the density and volumetric expansion coefficient at core inlet conditions (27&deg;C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm 2 s*1 , fiT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by (-l+.12.)
                                                                                                                                                            \ '.~=================~
in which mis the coolant mass flow rate (kg s*1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m 2 , and Dh is the hydraulic diameter, namely, 0.02127 m. The friction factor /for laminar flow through the annular area is given by 100 Re-1 (Shah & London 1978), in which the Reynolds*number is given by Dhm I A&#xb5;0 in which &#xb5;,, is the dynamic viscosity at core inlet conditions.
                                                                                                                                                            \ *[ Formatted                         [
Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5), and the entrance pressure loss would be expected to be negligible.
                                                                                                                                                                '[ Formatted                       [
The exit contraction loss is given by K-State Reactor Safety Analysis Report 4-30 (-l+/-.Ll.)
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* *
* REACTOR DESCRIPTION s     I Inlet                                     JFormatted: Centered, Keep with next I
* REACTOR DESCRIPTION The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which [A, ]2 [ 3
I I
* R 2 sin60&deg; cos60&deg;] K =. -= = 0171 Ac 7f
I I
* Rz . , (H14) where R is the radius of the opening in the upper grid plate. Equations (H.12.) through (Hl4), solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4.911. Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRI GA Reactor During Steady-State Operations.
Hot-leg Cold-Leg                             I
P (kWt) m (kg s*1) AT(0 C) 50 0.047 3.1 100 0.061 4.7 200 0.077 7.5 300 0.090 9.6 400 0.100 11.5 500 0.108 13.3 750 0.125 17.2 1000 0.139 20.6 1250 0.150 23.8 4.7 Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
* connector Figure 4.12 - RELAP single channel model used in CHF analysis             * - - - ( Formatted: Caption, Centered
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500&deg;C; consequently, limits on fuel temperature change for cladding temperatures greater than 500&deg;C. A maximum temperature of 1150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zri.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.
                                                                                        * - - - Formatted: Centered, Don't adjust space between Latin and Asian text, Don't adjust space between Asiar text and numbers, Tab stops: 3", Centered + 5.5'',
 
Rig t K-State Reactor                          4-25                          Original (12/04)
===4.8 Operating===
Safety Analysis Report
 
* CHAPTER4 4
Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature.
                              ~=       ~ 6 , if Dh :::: 0. lft D"
The operating parameters established for the KSU TRlGA reactor are: K-State Reactor Safety Analysis Report 4-31 Original (12/04) 
                              ~=_!2_+90,if Dh ~O.lft Dh hso =film coefficient at CHF Dh =hydraulic diameter (ft) v =coolant velocity (ft Is)
*
Twso =wall temperature at burnout (&deg; C)
DH =heated diameter (ft)
The RELAP simulations were performed for the hot channel. i.e., a channel with a radial peaking--- Formatted: Justified, Don't adjust space between Latir factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain         and Asian text, Don't adjust space between Asian text the pressure, temperature. and velocity distribution at different axial locations. With these         and numbers calculations and the functional form of the Bernath correlation. the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA. the A-ring is occupied by the central thimble. not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard.
The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel.
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* REACTOR DESCRIPTION
                                                                        ,, .. Bernatf1-CHF l  Formatted: Centered, Keep with next I
II A
* PG-CHF
                                                                        ,, " Heat flux I
I I
I l.L
:c     2000 u
1000 8.oo       Cl.OS     IUO     0.15     0.20 0.25     Cl.30   0.35 Heated Length (m)
Figure 4.13 - CHF versus heated length                             * - - ( Formatted: Caption, Centered ti Formatted: Centered, Keep with next
                        + + Bernath-CHF
                        <> <> PG-CHF                                                                               I I
I 10 .........   *..
I
                                                                                                                /
                        &#xa2;> <). &#xa2;> '
                                  ;,                                                                         i B ..... .                                                                               I 0::.
o::l z
0     6
* 4.
2 8.oo       0.0S       0.10     0.15     0.20 0.25       0.30    0.35 Heated Length (m)
Figure 4.14 - DNBR versus heated length                               *--~( Formatted: Caption, Centered K-State Reactor                                       4-27                           Original (12/04)
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* CHAPTER4
  *------------------------------------------ _ ___ _______ _____ __________ __ :'.~x-;:~*-( Formatted: Font: 11 pt, Not Bold
: d. Pulsed Mode of Operation Formatted: Justified, Don't adjust space between Latir and Asian text, Don't adjust space between Asian text Transient calculations have been performed using a custom computer code TASCOT for                     and numbers transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. ((;.2) and (.'.710), with the peak fuel temperature set to 746 &deg;C, i.e., a temperature rise of 719 &deg;C above 27 &deg;C ambient temperature. The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1% ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation. In the TASCOT calculations, thermal conductivity was set to 0.18 W cm- 1 K- 1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm- 1 K- 1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994)
The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955) k0.79
* C0.45 *A 0.51               -
hb = 0.00122
* f        pf
                                                                                    *( Tw - T,at )0.99 , (9-10)
                                                        * (v g _v v )
0.75
[ a 05 * &#xb5; f0.29
* p g0.24
* To.15 sat in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 &deg;C),
and n the coolant ambient temperature (27&deg;C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases.
The overall heat transfer coefficient U varies negligibly for ambient temperatures from 20 to 60 &deg;C, and has the value 0.21 W cm- 1 K- 1 at Tb = 27 &deg;C.
Figure 4.14,2 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.108 lists temperatures and heat fluxes as function of time after a 2.1 % ($3.00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m- 1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500&deg;C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse.
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* REACTOR DESCRIPTION 1000 Os 800 u
0 2
600 Ill L..
:J 4
        .&#xb5; l1l L..                                                               8 Ill
: 0. 400 E
Ill                                                               16 I-200                                                          32 64s 0 ~~~~~~~~~~~~~~~~~~~~~~~~
0.0 0.20 0.40 0.60 0.80     1.0   1.2   1.4   1.6 1.8   2.0   2.2 Radius (cm)
* Figure 4. 14~, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse.
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* CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{&deg;C) Coolant Ambient Temperature.
Time (s)               Q"             CHFR Fuel outside       Clad surface (Wm-2 )                               Temp. (0 C)       Temp. (0 C) 0                                                         953 1               3.57 xl0 5           4.2                 781               224 2               7.34 xl05            2.0                 683               432 4               8-52 xl05            1.7                 574               498 8               7.54 xl0 5           2.0                 461               443 16               5.71 xl05            2.6                 344               342 32               3.46 xl05            4.3                 224               218 64               1.04 xl05            14.4                 100                 84 4.6     Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by
(.Wll) in which Po and Po are the density and volumetric expansion coefficient at core inlet conditions (27&deg;C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm2 s* 1, fiT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by
(-l+.12.)
in which mis the coolant mass flow rate (kg s* 1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m2, and Dh is the hydraulic diameter, namely, 0.02127 m. The friction factor /for laminar flow through the annular area is given by 100 Re- 1 (Shah & London 1978), in which the Reynolds*number is given by Dhm I A&#xb5; 0 in which &#xb5;,, is the dynamic viscosity at core inlet conditions.
Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5),
and the entrance pressure loss would be expected to be negligible. The exit contraction loss is given by
(-l+/-.Ll.)
K-State Reactor                                4-30                              Original (12/04)
Safety Analysis Report l
* REACTOR DESCRIPTION The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which 2        2 K =. [ -
A, ]
Ac    =[ 3*R   sin60&deg; cos60&deg;]
7f
* Rz           = 0171 (H14) where R is the radius of the opening in the upper grid plate. Equations (H.12.) through (Hl4),
solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4.911.
Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRI GA Reactor During Steady-State Operations.
P (kWt)                   m (kg s* 1)                     AT( 0 C) 50                     0.047                             3.1 100                     0.061                             4.7 200                       0.077                             7.5 300                       0.090                             9.6 400                       0.100                             11.5 500                       0.108                             13.3 750                       0.125                             17.2 1000                     0.139                             20.6 1250                     0.150                             23.8 4.7 Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500&deg;C; consequently, limits on fuel temperature change for cladding temperatures greater than 500&deg;C. A maximum temperature of 1150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zri.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.
4.8      Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature. The operating parameters established for the KSU TRlGA reactor are:
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* CHAPTER4
* CHAPTER4
* Steady-state power level
* Steady-state power level
* Fuel temperature measured by thermocouple during pulsing operations
* Fuel temperature measured by thermocouple during pulsing operations
* Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations.
* Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Taele 4.9Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100&deg;C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.
As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Taele 4.9Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100&deg;C bulk pool water temperature for 1.25 MW operation.
Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.
Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin. Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.  
4.8.3 Safety Margins FoF l,25Q kW steady state operations, the eritieal heat fhm Fatio indieated in Taele 4.9 ranges from 5.8 for pool water at room temperatl!Fe (27&deg;C) to 4.1 at iQ &deg;C (pool tenlperatHres aFe sontrolled to less than 48&deg;C foF operational eonee~s). Even at pool water temperatHFes approaehing boiling, the margin remains aeo\'e 2. TheFefore, margins to sonditions that eoHld eaHse e1ceessive temperatl!res dHFing steady state opeFations while eladding temperatHFes is below 5QQ&deg;C are e1ttremely large.
For 1.250 kWth steady-state operations. the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49&deg;C. The proposed Technical Specifications limit of 44&deg;C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation. Limiting pool inlet water temperature to no greater than 44&deg;C (or 37&deg;C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling.
Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746&deg;C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869&deg;C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15%
margin to the fuel temperature safety limit for cladding temperatures below 500&deg;C.
Analysis shows that cladding temperatures will remain below 500&deg;C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve K-State Reactor                                  4-32                              Original (12/04)
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* REACTOR DESCRIPTION 500&deg;C (invoking a 950&deg;C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.
Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500&deg;C, but the heat transfer simultaneously cools the fuel to much less than 950&deg;C.
If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 500&deg;C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500&deg;C.
4.9      Bibliography "TASCOT- A 2-D, Transient and Steady State Conduction Code for Analyhsis of a TRJGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C.,
  "Investigation of the Radial Variation of the Fission-Heat Source in a TRJGA Mark Ill Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C.,
* "A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,
Kansas State University TRJGA Mkll Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J .G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K.,
and N. Zuber, Theory and Design of Modern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F.,
  "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al.,
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* CHAPTER4 MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998.
NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRJGA Reactors," U.S. Nuclear Regulatory Commission, 1987.
  ''Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K.,
and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRJGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T.
  "Safety Analysis Report, TRJGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991.
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====4.8.3 Safety====
Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).
Margins FoF l,25Q kW steady state operations, the eritieal heat fhm Fatio indieated in Taele 4.9 ranges from 5.8 for pool water at room temperatl!Fe (27&deg;C) to 4.1 at *iQ &deg;C (pool tenlperatHres aFe sontrolled to less than 4 8&deg;C foF operational Even at pool water temperatHFes approaehing boiling, the margin remains aeo\'e 2. TheFefore, margins to sonditions that eoHld eaHse e1ceessive temperatl!res dHFing steady state opeFations while eladding temperatHFes is below 5QQ&deg;C are e1ttremely large. For 1.250 kWth steady-state operations.
The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A.1. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.
the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49&deg;C. The proposed Technical Specifications limit of 44&deg;C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation.
Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100&deg;F water, showed burnout fluxes of 0.9 to 2.0 Mbtu ft* 2 hr- 1 for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft* 2 hr- 1. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.
Limiting pool inlet water temperature to no greater than 44&deg;C (or 37&deg;C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling. Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746&deg;C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869&deg;C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15% margin to the fuel temperature safety limit for cladding temperatures below 500&deg;C. Analysis shows that cladding temperatures will remain below 500&deg;C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve K-State Reactor Safety Analysis Report 4-32 Original (12/04) 
A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A. land 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise.
* *
The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760&deg;C clad temperature for a peak initial fuel temperature of 1000&deg;C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure 4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor                                4.A-1                                 Original (9/02)
* REACTOR DESCRIPTION 500&deg;C (invoking a 950&deg;C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.
Safety Analysis Report
Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500&deg;C, but the heat transfer simultaneously cools the fuel to much less than 950&deg;C. If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature.
* CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470&deg;C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400&deg;C to 500&deg;C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965].
In worst-case air-cooling scenarios, cladding temperature can exceed 500&deg;C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500&deg;C. 4.9 Bibliography "TASCOT-A 2-D, Transient and Steady State Conduction Code for Analyhsis of a TRJGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C., "Investigation of the Radial Variation of the Fission-Heat Source in a TRJGA Mark Ill Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C., "A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C., Kansas State University TRJGA Mkll Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J .G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K., and N. Zuber, Theory and Design of Modern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F., "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al., K-State Reactor Safety Analysis Report 4-33 Original (12/04)
* RADIUS (IN.)
* *
Figure 4A.l. Representative Radial Variation of Power Within the TRIGA Fuel Rod I. l I .0 0.9 N
* CHAPTER4 MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Zirconium Hydride Fuels for TRJGA Reactors," U.S. Nuclear Regulatory Commission, 1987. ''Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K., and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRJGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T. "Safety Analysis Report, TRJGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991. K-State Reactor Safety Analysis Report 4-34 Original (12/04)
0.8
* *
: 0. 7
* Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998). The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A.1. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature.
: 0. 6
A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature.
: o. 5 0
Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies
AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.)
[Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface. Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties.
Figure 4A.2, Representative Axial Variation of Power Within the TRIG A Fuel Rod.
Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100&deg;F water, showed burnout fluxes of 0.9 to 2.0 Mbtu ft*2 hr-1 for e-folding periods from 5 to 90 milliseconds.
K-State Reactor                                 4.A-2                          Original (9/02)
On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft*2 hr-1. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.
Safety Analysis Report
A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A. land 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent.
* REACTOR DESCRIPTION
A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise. The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760&deg;C clad temperature for a peak initial fuel temperature of 1000&deg;C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure 4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor Safety Analysis Report 4.A-1 Original (9/02)
                                                      \            CURVE BASED ON
* *
                                                        \ ' , , DATA OF ELLI/.
* CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation.
x 5:z:      lOlt TW-TSAT ('FJ Figure 4A.3, Subcooled Boiling Heat Transfer for Water.
Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 4 70&deg;C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400&deg;C to 500&deg;C for TRIGA Mark F fuel elements which have been operated under film boiling conditions
* 1800 1700 1600 1500 llfOD IJOO 100 SEC 1200 0.1  0.2  0.3      o.*        0.5    0.6 0.7 o.e RADIUS (IN.)
[Coffer et al. 1965]. " .. RADIUS (IN.) Figure 4A.l. Representative Radial Variation of Power Within the TRIGA Fuel Rod I. l I .0 0.9 N 0.8 ;;: 0. 7 0. 6 o. 5 0 AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.) Figure 4A.2, Representative Axial Variation of Power Within the TRIG A Fuel Rod. K-State Reactor Safety Analysis Report 4.A-2 Original (9/02) 
Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse.
* *
K-State Reactor                                        4.A-3                                Original (9/02)
* x "' ..... ... 5 lOlt :z: REACTOR DESCRIPTION
Safety Analysis Report
\ CURVE BASED ON \',, DATA OF ELLI/. ' " ' .... ____ ,,,,, TW-TSAT ('FJ Figure 4A.3, Subcooled Boiling Heat Transfer for Water . 1800 1700 1600 1500 llfOD IJOO 1200 K-State Reactor Safety Analysis Report 100 SEC 0.1 0.2 0.3 o.* 0.5 0.6 0.7 o.e RADIUS (IN.) Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse. 4.A-3 Original (9/02) 
* CHAPTER 4 APPENDIX A 106 ONSET OF_!_!          PEAK HEAT FLUX NUCLEATE BOILING N
* *
                        ~
* L CHAPTER 4 APPENDIX A 106 ONSET OF_!_! PEAK HEAT FLUX NUCLEATE BOILING N ... "' 105 <:: .. "' w 10 4 u i 1ol 0.001 0.01 0.1 1.0 10 100 ELAPSED Tito![ FRCIH ENO OF PULSE (SEC) Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse . 1000 CLAO OUTER SURFACE TEHP / 10
                      <:: 105
......
                        ~
0.001 O.DI 0. 1 1.0 10 ELAPSEO TIME FROM ENO OF PULSE (SEC) Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element. K-State Reactor Safety Analysis Report 4.A-4 Original (9/02)
                        ~
* *
w u  10 4 i
* REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface.
1ol 0.001                0.01                    0.1                    1.0                    10              100 ELAPSED Tito![ FRCIH ENO OF PULSE (SEC)
Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.
Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse.
This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected.
1000 CLAO OUTER SURFACE TEHP
Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962] which provide insight into the mechanisms involved.
                                  /
They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient.
10 L---1~.....L.-1.....J...IL-....1.~.....1.......1....LJL..-1~.....L......1...J..1~-'-~-'--'-......""'.""-'-~~~~10*0 0.001                O.DI                      0. 1                  1.0                      10 ELAPSEO TIME FROM ENO OF PULSE (SEC)
These results were then compared with the incipient film boiling conditions of the 1000&deg;C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation.
Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element.
The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors.
K-State Reactor                                                  4.A-4                                                    Original (9/02)
The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960] which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate. The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft*2 hr 1 &deg;F-1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft*2 hr*' &deg;F*1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel. The surface heat flux at the midplane of the element is shown in Figure 4A.10 with gap conductance as a parameter.
Safety Analysis Report L
It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft*2 hr-1 &deg;F-1 according to the more conservative Bernath correlation at a local water temperature approaching saturation.
* REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.
K-State Reactor Safety Analysis Report 4.A-5 Original (9/02) 
This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected. Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962]
* *
which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000&deg;C peak fuel temperature case.
* CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000&deg;C is conservatively estimated to be 4 70&deg;C. As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470&deg;C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000&deg;C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is l l 50&deg;C would not produce a stress in the clad in excess of its ultimate strength.
For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960]
TRI GA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about l 150&deg;C without damage to the clad [Dee et al. 1966]. "' I 0 >< 7 ELAPSED TIME FROM 6 *ENO OF PULSE
which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate.
* 0.247 SEC 5 3 7 8 9 10 11 12 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A.7, Surface Heat Flux Distribution for Standard Non-Gapped (hgap= 500 Btu/h ft 2 &deg;F) Fuel Element After a Pulse. 13 K-State Reactor Safety Analysis Report 4.A-6 Original (9/02) 
The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft* 2 hr 1 &deg;F- 1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft* 2 hr*' &deg;F* 1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel.
* *
The surface heat flux at the midplane of the element is shown in Figure 4A.10 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft* 2 hr- 1 &deg;F- 1 according to the more conservative Bernath correlation at a local water temperature approaching saturation.
* ELAPSED TIME FROM END OF PULSE IS 0.314 SEC 10 11 REACTOR DESCRIPTION 13 15 0 I STANCE FROM BOTTOM OF FUEL (IN. l Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap= 375 Btu/h ft 2 &deg;F) After a Pulse. 8 7 N I-6 "-a: :> I-"' s 'o >< 4 >< :> _, "-I-3 ... "' :i:: 2 l 7 8 ELAPSED TIME FROM END OF PULSE IS 0.440 SEC 9 10 ll 12 13 DISTANCE FROM BOTTOM OF FUEL (IN.) 14 15 Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 250 Btu/h ft2 &deg;F ) After a Pulse. K-State Reactor Safety Analysis Report 4.A-7 Original (9/02)
K-State Reactor                                    4.A-5                                 Original (9/02)
* *
Safety Analysis Report
* CHAPTER 4 APPENDIX A 105 ... I a:: :r ..... :::> .... m )C. :::> _, ... .... < .... :c .... u < ... IOI+ EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR*FT2.
* CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000&deg;C is conservatively estimated to be 4 70&deg;C.
-&deg;F 500 FLOW VELOCITY a 1 FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME a ZERO) 0.01 0.1 1.0 ELAPSED TIME FROM END OF PULSE (SEC) Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse. K-State Reactor Safety Analysis Report 4.A-8 Original (9/02)
As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470&deg;C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000&deg;C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is l l 50&deg;C would not produce a stress in the clad in excess of its ultimate strength. TRI GA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about l 150&deg;C without damage to the clad [Dee et al. 1966].
* *
7 ELAPSED TIME FROM
* REACTOR DESCRIPTION Bibliography "A Theory of Local Boiling Burnout and Its Application to Existing Data, " Heat Transfer -Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No. 20.Bemath, L., Research in Improved TRIGA Reactor Performance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al., Characteristics of Large Reactivity Insertions in a High Performance TRIG A U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al. Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al., A Study of the Mechanism of Boiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E., Thermal Conductance of Metallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis of a Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J. Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172. Graff, W.J. Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H .. MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H. Choi, "Quarterly Technical Report SP ERT Project, April, May, June, 1964, "ISO 17030. Spano, A. H., "The Effect of Subcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hvdraulic CHF analysis." Technical report, Argonne National Laboratory.
                                    *ENO OF PULSE
2008, E.E. Feldman. K-State Reactor Safety Analysis Report 4.A-9 Original (9/02) 
* 0.247 SEC 6
* *
*      "'0 I
* CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure.
5 3
system models, and solution methods. *-----Formatted:
7            8              9            10            11            12          13 DISTANCE FROM BOTTOM OF FUEL (IN.)
Normal, Don't adjust space between Latin and Asian text, Don't adjust space between Asian text "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution." and numbers Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong.
Figure 4A.7, Surface Heat Flux Distribution for Standard Non-Gapped (hgap=
K-State Reactor Safety Analysis Report 4.A-10 Original (9/02) \\ '( Formatted:
500 Btu/h ft2 &deg;F) Fuel Element After a Pulse.
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K-State Reactor                                  4.A-6                                Original (9/02)
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Safety Analysis Report
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* REACTOR DESCRIPTION ELAPSED TIME FROM END OF PULSE IS 0.314 SEC 10      11                  13      15 0 I STANCE FROM BOTTOM OF FUEL (IN. l Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap= 375 Btu/h ft 2 &deg;F) After a Pulse.
* REACTOR DESCRIPTION "Onset of Stable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al., UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991. "Hydrodynamic Aspects of Boiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959. Zuber, W . K-State Reactor Safety Analysis Report 4.A-11 Original (9/02) 
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* AppendixB Water Properties at Nominal Operating Conditions Tpool oc 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 97 99 Tpool oc 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 97 99 999.21 998.32 997.16 995.75 994.12 992.29 990.27 988.07 985.70 983.18 980.50 977.69 974.74 971.66 968.45 965.12 961.68 960.27 958.84 PTl1J kg m-3 999.21 998.32 997.16 995.75 994.12 992.29 990.27 988.07 985.70 983.18 980.50 977.69 974.74 971.66 968.45 965.12 961.68 960.27 958.84 Pi,161 1 1 kg m-3 950.00 950.01 950.02 950.03 950.04 950.06 950.07 950.09 950.11 950.12 950.14 950.17 950.19 950.21 950.23 950.26 950.29 950.30 950.31 Pr,131 1 1 kg m-3 951.43 951.43 951.44 951.45 951.46 951.47 951.49 951.50 951.51 951.53 951.55 951.57 951.58 951.60 961.62 951.64 951.67 951.68 951.68 K-State Reactor Safety Analysis Report Data for 16 Feet of Water over the Core 47.79 47.74 47.69 47.62 47.54 47.46 47.36 47.25 47.14 47.02 46.89 46.76 46.62 46.47 46.32 46.16 45.99 45.92 ht,161 1 1 kJ kg-1 465.10 465.05 465.01 464.95 464.89 464.81 464.73 464.64 464.54 464.44 464.33 464.21 464.09 463.96 463.83 463.69 463.55 463.49 45.86 463.43 hg,161 1 1 kJ kg-1 2692.64 2692.63 2692.59 2692.59 2692.57 2692.54 2692.51 2692.48 2692.45 2692.41 2692.37 2692.33 2692.29 2692.24 2692.19 2692.15 2692.09 2692.07 2692.05 P 9 ,151 1 1 kg m-3 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.84 0.84 0.84 0.84 0.84 0.84 0.84 0.84 Data for 13 Feet of Water over the Core 38.83 38.79 38.75 38.69 38.63 38.56 38.48 38.39 38.30 38.20 38.10 37.99 37.88 37.76 37.63 37.50 37.37 37.31 37.26 hr,13l 1 l kJ kg-1 457.21 457.18 457.13 457.09 457.03 456.96 456.89 456.82 456.73 456.64 456.55 456.45 456.35 456.24 456.12 456.01 455.89 455.84 455.78 hg,1}11 kJ kg-1 2689.85 2689.84 2689.82 2689.80 2689.78 2689.76 2689.74 2689.71 2689.68 2689.65 2689.61 2689.58 2689.54 2689.50 2689.46 2689.42 2689.38 2689.36 2689.34 4.B-1 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.79 0.79 0.79 0.79 Tsat,161 1 1 oc 110.89 110.88 110.87 110.86 110.84 110.83 110.81 110.78 110.76 110.74 110.71 110.68 110.65 110.62 110.59 110.56 110.53 110.51 110.50 Tsat,131 1 1 oc 109.03 109.02 109.01 109.00 108.99 108.97 108.96 108.94 108.92 108.90 108.87 108.85 108.83 108.80 108.77 108.75 108.72 108.70 108.69 q "sat, 151 3 1 wm-2 1553.842 1552.078 1549.496 1547.118 1543.981 1540.446 1536.512 1532.205 1527.561 1522.575 1517.255 1511.666 1505.778 1499.613 1493.199 1486.527 1479.626 1472.944 1466.058 q"sa1,13l 3 l wm-2 1513.00 1511.32 1509.15 1505.85 1503.13 1500.16 1496.38 1492.25 1487.78 1482.99 1477.90 1472.52 1466.86 1460.95 1458.59 1448.37 1441.74 1435.27 1428.60 q"sub1 4 l wm-2 7239.19 6931.74 6622.60 6311.82 5999.91 5688.24 5376.29 5064.36 4753.90 4444.85 4136.85 3830.73 3526.89 3225.47 2926.81 2631.05 2338.47 2216.11 2095.18 q"subl 4 1 wm-2 6964.74 6857.12 6543.62 6229.30 5913.58 5597.21 5281.90 4966.66 4652.39 4339.60 4027.94 3718.83 3412.07 3107.29 2812.63 2506.90 2211.19 2087.92 1966.16 Revised 05/01 /17 
              "-      6 a:
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              ~
* REACTOR DESCRIPTION Common Data T Patm Cp(1] ()" "C kPa kJ kg-1 "1 Nm*1 15 99.83 4.23080 0.07149 20 99.83 4.23080 0.07120 25 99.83 4.23080 0.07083 30 99.83 4.23080 0.07039 35 99.83 4.23070 0.06989 40 99.83 4.23070 0.06932 45 99.83 4.23070 0.06869 50 99.83 4.23070 0.06800 55 99.83 4.23060 0.06727 60 99.83 4.23060 0.06649 65 99.83 4.23060 0.06566 70 99.83 4.23050 0.06480 75 99.83 4.23050 0.06390 80 99.83 4.23050 0.06297 85 99.83 4.23040 0.06201 90 99.83 4.23040 0.06102 95 99.83 4.23040 0.06001 97 99.83 4.23030 0.05898 99 99.83 4.23030 0.05793 NOTE[1]: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa
I-
=Heigth(ji)
              "' s
* 12(in/ft)
              'o                                ELAPSED TIME FROM END OF PULSE IS 0.440 SEC
*O. 0254(meters/in)
                ><    4 I-3
*Density(kg/m
                "':i::
: 3) *9.8066511000 NOTE[4]: q;ub =
2 l
1+0.l*(PJJ . cp.f -Tsub) [ x l Pg hg.sat h f,sat NOTE:[5}:
7  8        9          10        ll      12          13 14  15 DISTANCE FROM BOTTOM OF FUEL (IN.)
Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap =
250 Btu/h ft2 &deg;F ) After a Pulse.
K-State Reactor                                    4.A-7                            Original (9/02)
Safety Analysis Report
* CHAPTER 4 APPENDIX A EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR*FT2. -&deg;F 500
          ~
          ...I 105 a::
:r m
          )C.
:c u
            ~        IOI+
FLOW VELOCITY a 1 FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME a ZERO) 10 3 '--~~-'-~~~'--~..___.__._~~~~~~-'-~-'---'--'
0.01                        0.1                      1.0 ELAPSED TIME FROM END OF PULSE (SEC)
Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse.
K-State Reactor                                    4.A-8                        Original (9/02)
Safety Analysis Report
* REACTOR DESCRIPTION Bibliography "A Theory ofLocal Boiling Burnout and Its Application to Existing Data, " Heat Transfer -
Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No.
20.Bemath, L.,
Research in Improved TRIGA Reactor Performance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al.,
Characteristics ofLarge Reactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al.
Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al.,
A Study of the Mechanism ofBoiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E.,
Thermal Conductance ofMetallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis of a Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J.
Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172.
Graff, W.J.
Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H ..
MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998.
Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H.
Choi, "Quarterly Technical Report SPERT Project, April, May, June, 1964, "ISO 17030. Spano, A.
H.,
  "The Effect of Subcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hvdraulic CHF analysis." Technical report, Argonne National Laboratory. 2008, E.E. Feldman.
K-State Reactor                               4.A-9                              Original (9/02)
Safety Analysis Report
* CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure. system models, and solution methods. *----- Formatted: Normal, Don't adjust space between Latin and Asian text, Don't adjust space between Asian text "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution."          and numbers Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong.             ----------------------~~--(>=~Fo~r~m~a~tt~e~d~:~Fo~n~t:~lt~a~lic----------
                                                                                                      \ \ '( Formatted: Font: Italic
                                                                                                        \l Formatted: Font: Bold K-State Reactor                               4.A-10                                Original (9/02)
Safety Analysis Report
* REACTOR DESCRIPTION "Onset of Stable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al.,
UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991.
  "Hydrodynamic Aspects ofBoiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959.
Zuber, W .
K-State Reactor                             4.A-11                                Original (9/02)
Safety Analysis Report
* AppendixB Water Properties at Nominal Operating Conditions Data for 16 Feet of Water over the Core Tpool            Pi,161 11            ht,161 11    hg,161 11 P9 ,15111 Tsat,161 11  q"sat, 15131  q"sub14l oc              kg m- 3              kJ kg- 1      kJ kg- 1  kg m- 3      oc          wm-2          wm-2 15    999.21  950.00      47.79    465.10        2692.64      0.85    110.89      1553.842      7239.19 20    998.32  950.01      47.74    465.05        2692.63      0.85    110.88      1552.078      6931.74 25    997.16  950.02      47.69    465.01        2692.59      0.85    110.87      1549.496      6622.60 30    995.75  950.03      47.62    464.95        2692.59      0.85    110.86      1547.118      6311.82 35    994.12  950.04      47.54    464.89        2692.57      0.85    110.84      1543.981      5999.91 40    992.29    950.06      47.46    464.81        2692.54      0.85    110.83      1540.446      5688.24 45    990.27  950.07      47.36    464.73        2692.51      0.85    110.81      1536.512      5376.29 50    988.07  950.09      47.25    464.64        2692.48      0.85    110.78      1532.205      5064.36 55    985.70  950.11      47.14    464.54        2692.45      0.85    110.76      1527.561      4753.90 60    983.18  950.12      47.02   464.44        2692.41      0.85    110.74      1522.575      4444.85 65     980.50  950.14      46.89    464.33        2692.37      0.85    110.71      1517.255      4136.85 70     977.69   950.17      46.76    464.21        2692.33      0.84    110.68       1511.666      3830.73 75    974.74  950.19      46.62    464.09        2692.29     0.84    110.65      1505.778      3526.89 80    971.66   950.21      46.47    463.96        2692.24      0.84   110.62      1499.613      3225.47 85    968.45  950.23      46.32    463.83        2692.19      0.84    110.59      1493.199      2926.81 90    965.12   950.26      46.16    463.69        2692.15    0.84    110.56      1486.527      2631.05 95    961.68  950.29       45.99    463.55        2692.09      0.84    110.53      1479.626      2338.47 97    960.27    950.30      45.92    463.49        2692.07    0.84    110.51       1472.944      2216.11 99    958.84    950.31      45.86    463.43        2692.05    0.84    110.50      1466.058      2095.18 Data for 13 Feet of Water over the Core Tpool    PTl1J  Pr,131 11            hr,13l1l      hg,1}11            Tsat,131 11  q"sa1,13l3l  q"subl41 oc    kg  m- 3 kg  m-3              kJ kg- 1      kJ kg- 1                oc        wm-2          wm-2 15    999.21    951.43      38.83    457.21        2689.85    0.80    109.03      1513.00      6964.74 20    998.32    951.43      38.79    457.18        2689.84    0.80    109.02      1511.32      6857.12 25    997.16    951.44       38.75    457.13        2689.82    0.80    109.01      1509.15      6543.62 30    995.75    951.45       38.69    457.09        2689.80    0.80    109.00      1505.85      6229.30 35    994.12    951.46      38.63    457.03        2689.78    0.80    108.99      1503.13      5913.58 40    992.29   951.47      38.56    456.96        2689.76    0.80    108.97      1500.16      5597.21 45    990.27    951.49      38.48    456.89        2689.74    0.80    108.96      1496.38      5281.90 50    988.07    951.50      38.39    456.82        2689.71    0.80    108.94      1492.25      4966.66 55    985.70    951.51      38.30    456.73        2689.68    0.80    108.92      1487.78      4652.39 60    983.18    951.53      38.20   456.64        2689.65    0.80    108.90      1482.99      4339.60 65    980.50   951.55      38.10    456.55        2689.61    0.80    108.87      1477.90      4027.94 70    977.69    951.57      37.99    456.45        2689.58    0.80    108.85      1472.52      3718.83 75    974.74    951.58      37.88    456.35        2689.54    0.80    108.83      1466.86      3412.07 80    971.66    951.60      37.76   456.24        2689.50    0.80    108.80      1460.95      3107.29 85    968.45    961.62      37.63    456.12        2689.46     0.80    108.77      1458.59      2812.63 90    965.12    951.64      37.50    456.01        2689.42    0.79    108.75      1448.37      2506.90 95    961.68    951.67      37.37    455.89        2689.38    0.79    108.72      1441.74      2211.19 97    960.27    951.68      37.31    455.84        2689.36    0.79    108.70      1435.27      2087.92 99    958.84   951.68      37.26    455.78       2689.34    0.79    108.69      1428.60      1966.16 K-State Reactor                                  4.B-1                                 Revised 05/01 /17 Safety Analysis Report
* Common Data REACTOR DESCRIPTION T          Patm          Cp(1]        ()"
                              "C          kPa      kJ kg-1 "1      Nm*1 15        99.83      4.23080      0.07149 20          99.83      4.23080      0.07120 25          99.83      4.23080      0.07083 30          99.83      4.23080      0.07039 35          99.83       4.23070      0.06989 40          99.83      4.23070      0.06932 45          99.83      4.23070      0.06869 50          99.83      4.23070      0.06800 55          99.83      4.23060      0.06727 60          99.83      4.23060      0.06649 65          99.83      4.23060      0.06566 70          99.83      4.23050      0.06480 75          99.83      4.23050      0.06390 80          99.83      4.23050      0.06297 85         99.83      4.23040      0.06201 90          99.83      4.23040      0.06102 95          99.83      4.23040      0.06001 97          99.83      4.23030      0.05898 99          99.83      4.23030      0.05793 NOTE[1]: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa =Heigth(ji) *12(in/ft) *O. 0254(meters/in) *Density(kg/m 3) *9.8066511000 NOTE[4]: q;ub =  q~AT. [1+0.l*(PJJ x . cp.f *(T~r -Tsub)l Pg           hg.sat   h f,sat NOTE:[5}:
u = 1.000E -11
u = 1.000E -11
* T 4 + 7.370E -09
* T 4 + 7.370E - 09
* T 3 -1.969E -06
* T 3 -1.969E - 06
* T 2 + 4.709E -06
* T 2 + 4.709E - 06
* T + 7.1833E -02 K-State Reactor Safety Analysis Report 4.A-13 Original (9/02)
* T + 7.1833E - 02 K-State Reactor                               4.A-13                               Original (9/02)
: 4. Reactor Description  
Safety Analysis Report
 
: 4.       Reactor Description 4.1     Summary Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.
===4.1 Summary===
The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations. Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems.
Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.
The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:
The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations.
Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems. The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:
* Central thimble
* Central thimble
* Rotary specimen rack
* Rotary specimen rack
* Thermalizing column with bulk shielding tank
* Thermalizing column with bulk shielding tank
* Thermal column with removable door *Beam ports
* Thermal column with removable door
              *Beam ports
* Radial (2)
* Radial (2)
* Piercing (fast neutron) (1)
* Piercing (fast neutron) (1)
* Tangential (thermal neutron) (1) The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion.
* Tangential (thermal neutron) (1)
All cooling is by natural convection.
The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.)
The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated. graphite.
depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.)
A fourth control rod would be used for 1,250-kW operation.
diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated.
A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m 3 (144,000 ft.3) confinement building K-State Reactor Safety Analysis Report 4-1 Original (12/04)
graphite. A fourth control rod would be used for 1,250-kW operation. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m3 (144,000 ft. 3) confinement building K-State Reactor                                 4-1                               Original (12/04)
* *
Safety Analysis Report
* CHAPTER 4 made of reinforced concrete and structural steel, with composite sheathing and aluminum siding. Sectional views of the reactor are shown in Figures 4.1 and 4.2. Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW. The original aluminum-clad fuel elements were replaced with stainless-steel clad elements in 1973. Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in 1993, with support from the U.S. Department of Energy. All neutronic instrumentation was replaced in 1994. North 22 FT 1Q IN. I I I TJ.liG(tlfl.>J..j
!If.AM PCf?TI I .---------28 Fr .; tN.-----------------j Figure 4. 1, Vertical Section Through the KSU TRIGA Reactor . K-State Reactor Safety Analysis Report 4-2 Original (12/04) 
* *
* REACTOR DESCRIPTION


===4.2 Reactor===
CHAPTER 4 made of reinforced concrete and structural steel, with composite sheathing and aluminum siding.
Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident.
Sectional views of the reactor are shown in Figures 4.1 and 4.2.
Inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fuel in a zirconium hydride matrix. rn ei'I, ... .. -West .--11*r---_, ; .. , t2 rT r-I au .. l(-!iH1i::1..c,,wc
Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW.
' . ,. ,-. ''
The original aluminum-clad fuel elements were replaced with stainless-steel clad elements in 1973. Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in 1993, with support from the U.S.
I I 11
Department of Energy. All neutronic instrumentation was replaced in 1994.
---28 F'f 4 i'i. ---------i East Figure 4. 2, Horizontal Section Through the KSU TRIGA Reactor. The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively constant with temperature
North
(-0.01 % This coefficient is a function of the fuel composition and core geometry.
* 22 FT 1Q IN.
As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies.
I I
The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase.
I TJ.liG(tlfl.>J..j
Therefore, fuel and clad temperature automatically limit operation of the reactor . K-State Reactor Safety Analysis Report 4-3 Original (12/04)
                                                                                            !If.AM PCf?TI
* *
                                  . - - - - - - - - - 2 8 Fr .; tN.-----------------j I
* CHAPTER4 It is more convenient to set a power level limit that is based on temperature.
Figure 4. 1, Vertical Section Through the KSU TRIGA Reactor.
The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120&deg;F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response.
* K-State Reactor Safety Analysis Report 4-2                            Original (12/04)


====4.2.1 Reactor====
REACTOR DESCRIPTION
Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted.
* 4.2           Reactor Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident. Inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fuel in a zirconium hydride matrix.
I              11
                            .--11*r---_,
                    ; .., t2 rT r- Iau.. l(-!iH1i::1..c,,wc
(~"~'lllilfr;:AL ~~!<>!
                    ',. . I rn ei'I,      ''
                -,~
      ... ~' . -
West            ~-------                      --- 28 F'f 4 i'i.  ---------i          East Figure 4. 2, Horizontal Section Through the KSU TRIGA Reactor.
The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively constant with temperature (-0.01 % ~k/k C). This coefficient is a function of the fuel composition and core 0
geometry. As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase. Therefore, fuel and clad temperature automatically limit operation of the reactor .
* K-State Reactor Safety Analysis Report 4-3                    Original (12/04)
 
CHAPTER4 It is more convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120&deg;F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response.
4.2.1 Reactor Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted.
Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic.
Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic.
Additional advantages  
Additional advantages *were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size.
*were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size. TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material.
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material. The maximum temperature limits of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zr1.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.
The maximum temperature limits of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zr1.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.
: a.       Dimensions and Physical Properties.
: a. Dimensions and Physical Properties.
The KSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements. Three instrumented aluminum-clad Mark II elements are still available for use in the core.
The KSU TRI GA reactor is fueled by stainless steel clad Mark III fuel-elements.
General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.) hole drilled through the center of the active fuel section.
Three instrumented aluminum-clad Mark II elements are still available for use in the core. General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.)
Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (1 in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank.
hole drilled through the center of the active fuel section. Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (1 in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank. 1 Unless otherwise indicated, fuel properties are taken from the General Atomics report ofSimnad [1980] and from authorities cited by Simnad . K-State Reactor Safety Analysis Report 4-4 Original (12/04) 
1 Unless otherwise indicated, fuel properties are taken from the General Atomics report ofSimnad [1980]
* *
and from authorities cited by Simnad .
* REACTOR DESCRIPTION Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements.
* K-State Reactor Safety Analysis Report 4-4                                 Original (12/04)
They are clad in aluminum and have a graphite length of 55.88 cm (22 in.). Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility.
: Property Mark II Mark III Dimensions Outside diameter, Do= 2r 0 Inside diameter, D;= 2r; Overall length Length of fuel zone, L Length of graphite axial reflectors End fixtures and cladding Cladding thickness Burnable poisons Uranium content Weight percent U 235 U enrichment percent 235 U content Physical properties of fuel excluding cladding H/Zr atomic ratio Thermal conductivity (W cm-1 K-1) Heat capacity [T (J cm-3 K-1) Mechanical properties of delta phase U-ZrH" Elastic modulus at 20&deg;C Elastic modulus at 650&deg;C Ultimate tensile strength (to 650&deg;C) Compressive strength (20&deg;C) Compressive yield (20&deg;C) asource: Texas SAR [1991]. b. Composition and Phase Properties 1.47 in. (3.7338 cm) 1.41 in (3.6322 cm) 28.4 in. (72.136 cm) 14 in. (35.56 cm) 4 in. (10.16 cm) aluminum 0.030 in. (0.0762 cm) Sm wafers 8.0 20 36 g 1.0 0.18 1.47 in. (3.7338 cm) 1.43 in. (3.6322 cm) 28.4 in. (72.136 cm) 15 in. (38.10 cm) 3.44 in (8.738 cm) 304 stainless steel 0.020 in. (0.0508 cm) None 8.5 20 38 g 1.6 0.18 2.04 + 0.00417T 9.1 x 10 6 psi 6.0 x 10 6 psi 24,000 psi 60,000 psi 35,000 psi The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% 235 U, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature.
For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760&deg;C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen.
Over 25,000 pulses have been performed with the TRI GA fuel elements at General Atomic, with fuel temperatures reaching peaks of about 1150&deg;C . K-State Reactor Safety Analysis Report 4-5 Original (12/04) 
* *
* CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455&deg;C. From 455&deg;C to about 1050&deg;C, the delta phase is supported by a broadening range ofH/Zr ratios. I 28*37" K-State Reactor Safety Analysis Report 1*47"DIA STAINLESS STEEL BOTTOM END-FIXTURE 20MIL STAINLESS STEEL CLAD ZIRCONIUM HYDRIDE-8*5WT% URANIUM, 20%ENR., 38 235LJ l:.43 11 DIA Figure 4.3, TRIGA Fuel Element . 4-6 T 3*44 11 J_ Original (12/04) 
* *
* REACTOR DESCRIPTION
: c. Core Layout A typical layout for a KSU TRIGA II 250-kW core (Core 11-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted .. Figure 4.4, Core Layout (250 kW). The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor Safety Analysis Report 4-7 Original (12/04)
* * * ----*----------------------
--------------------------
-CHAPTER 4 4.2.2 Control Rods The pulse rod is 3.175 cm. (1.25 in.) diameter.
Other rods are 2.225 cm (7/8 in.) diameter.
Control rods are 50.8 cm. (20 in.) long boron carbide or borated graphite , clad with a 0.0762 cm. (30-mil) aluminum sheath. The control rod drives are connected to the control rod clutches through three extension shafts. The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydraulic cushion (or buffer) for the rod during a scram , and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (slotted tubes on the right hand side) along with the top section of the pulse/transient rod extension.
The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder; the dimensions of the cylinder limits bottom travel. Upper Pulse , Shim & Reg Rods Reg Rod Shim Rod Pulse Rod Figure 4.5, Control Rod Upper Extension Assemblies The bottom of the pulse rod is shown on the left hand side of Figure 4.5. The upper extension shaft is a hollow tube, the middle extension is solid. The upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fastener).
Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full in position is above the bottom of the guide tube . K-State Reactor Safety Analysis Report 4-8 Original (12/0 4) 
* *
* REACTOR DESCRIPTION i.4 ..
... {
* 4' It *,1cr .. . ' ----**--*---
*---<<* ..
"'
Figure 4.6, Middle Extension Rod Alignment Holes The middle solid extension is similarly connected to the lower extension.
The lower extension is hollow , the middle extension fits into the lower extension and a hole drilled in the overlap secures the lower extension to the middle extension.
Typically the lower extension has a tighter fit than the upper extension because the lower and middle extension are not separated for inspections and because the interface with upper extension is used to set the bottom position of the control rod. Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7 .. Figure 4.7, Standard & Pulse Rod Lower Coupling The bottom of the lower extension attaches directly to the control rod. Pictures of the control rods taken during the 2003 control rod inspection are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wire in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube . K-State Reactor Safety Analysis Report 4-9 Original (12/04) 
*
* CHAPTER 4 Pulse Rod Shim Rod Reg Rod Figure 4.8, Control Rods During 2003 Inspection Full Guide Tube Upper Lower Lower Detail Position in Upper Grid Plate K-State Reactor Safety Analysis Report Figure 4.9, Control Rod Guide Tubes 4-10 Original (12/04) 
* *
* REACTOR DESCRIPTION
: a. Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings.
A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (1) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch. T bl 4 2 S a e . ' ummaryo fC t IR di t I ks on ro 0 n er oc INTERLOCK SETPOINT FUNCTION/PURPOSE Inhibit standard rod motion if nuclear Source Interlock 2 cps instrument startup channel reading is less than instrument sensitivity/ensure nuclear instrument startup channel is operating Pulse Rod Interlock Pulse rod inserted Prevent applying power to pulse rod unless rod inserted/prevent inadvertent pulse Multiple Rod Withdrawal Withdraw signal, Prevent withdrawal of more than I rod/Limit more than I rod maximum reactivity addition rate Pulse Mode Interlock Mode switch in Hi Prevent withdrawing standard control rods in Pulse pulse mode Pulse-Power Interlock IO kW Prevent pulsing if power level is greater than IO kW NOTE: (1) Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock required by Technical Specifications
: b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times. for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIG A prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation . K-State Reactor Safety Analysis Report 4-11 Original (12/04) 
* *
* CHAPTER 4 T bl 4 3 S a e . ' ummarvo fR t SCRAM eac or s Measuring Trip Setpoint Steady Actual Setpoint Channel State Pulse Linear Channel High 110% NIA 104% Power Power Channel 110% NIA 104% High power Detector High 90% 90% 90% Voltage 600&deg;C B Ring element High Fuel 555&deg;C C Ring element 450&deg;C Temperature[ll 480&deg;C D Ring element 3 80&deg;C E Ring element 350&deg;C Period [IJ NIA NIA 3 sec NOTE [1]: Period trip and temperature trip are not required by Technical Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin of at least $0.50 . Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications.


====4.2.3 Neutron====
REACTOR DESCRIPTION
Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator.
* Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of 55.88 cm (22 in.).
Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors.
Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility.
The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially.
: Property                                          Mark II                Mark III Dimensions Outside diameter, Do= 2r0                          1.47 in. (3.7338 cm)  1.47 in. (3.7338 cm)
The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3 .18-cm ( 1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector.
Inside diameter, D;= 2r;                            1.41 in (3.6322 cm)    1.43 in. (3.6322 cm)
* K-State Reactor Safety Analysis Report 4-12 Original (12/04)
Overall length                                    28.4 in. (72.136 cm)   28.4 in. (72.136 cm)
* *
Length of fuel zone, L                              14 in. (35.56 cm)      15 in. (38.10 cm)
* REACTOR DESCRIPTION The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank. Four lugs are provided for lifting the assembly.
Length of graphite axial reflectors                4 in. (10.16 cm)        3.44 in (8.738 cm)
A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures.
End fixtures and cladding                          aluminum                304 stainless steel Cladding thickness                                  0.030 in. (0.0762 cm) 0.020 in. (0.0508 cm)
Graphite inserts within the fuel cladding provide additional reflection.
Burnable poisons                                    Sm wafers              None Uranium content Weight percent U                                    8.0                    8.5 235 U enrichment percent                          20                      20 235 U content                                      36 g                    38 g Physical properties offuel excluding cladding H/Zr atomic ratio                                  1.0                    1.6 Thermal conductivity (W cm- 1 K- 1)                0.18                  0.18 Heat capacity [T ~0&deg;C] (J cm-3 K- 1)                                      2.04 + 0.00417T Mechanical properties ofdelta phase U-ZrH" Elastic modulus at 20&deg;C                                                    9.1 x 106 psi Elastic modulus at 650&deg;C                                                  6.0 x 106 psi Ultimate tensile strength (to 650&deg;C)                                      24,000 psi Compressive strength (20&deg;C)                                                60,000 psi Compressive yield (20&deg;C)                                                   35,000 psi asource: Texas SAR [1991].
Inserts are placed at both ends of the fuel meat, providing top and bottom reflection.  
: b.      Composition and Phase Properties The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% 235 U, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760&deg;C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen. Over 25,000 pulses have been performed with the TRI GA fuel elements at General Atomic, with fuel temperatures reaching peaks of about 1150&deg;C .
* K-State Reactor Safety Analysis Report 4-5                              Original (12/04)


====4.2.4 Neutron====
CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455&deg;C. From 455&deg;C to about 1050&deg;C, the delta phase is supported by a broadening range ofH/Zr ratios.
Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 10 6 n s-1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor. 4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates. The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures.
I 20MIL STAINLESS STEEL CLAD ZIRCONIUM HYDRIDE-28*37"                                  8*5WT%
Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core. Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position.
URANIUM, 20%ENR.,
The reflector assembly supports both grid plates. 4.3 Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.)
38  235LJ 11 l:.43 DIA 1*47"DIA STAINLESS T
thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion.
3*44 11 STEEL BOTTOM END-FIXTURE J_
Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration.
Figure 4.3, TRIGA Fuel Element.
All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool K-State Reactor Safety Anaiysis Report 4-13 Original (12/04)
K-State Reactor                               4-6                              Original (12/04)
* *
Safety Analysis Report
* CHAPTER 4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak. A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation.
The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations.
Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction.
Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector.
The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter I 0). 4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield. The shielding configuration is similar to those at other TRIGA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank. The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena.


===4.5 Nuclear===
REACTOR DESCRIPTION
Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors.
* c. Core Layout A typical layout for a KSU TRIGA II 250-kW core (Core 11-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted ..
This coefficient is a function of the fuel composition, core geometry, and temperature.
Figure 4.4, Core Layout (250 kW).
For fuels with 8.5% U, 20% enrichment, the value is nearly constant at 0.01 %
The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge
per &deg;C, and varies only weakly dependent on geometry and temperature.
* K-State Reactor Safety Analysis Report 4-7                          Original (12/04)
Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures* that the fuel and clad temperature limits will not be exceeded.
The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity.
Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor Safety Analysis Report 4-14 Original (12/04)
* *
* REACTOR DESCRIPTION limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6. 4.5.1 Design Criteria -Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity.
The bottom grid plate, which suppmis the weight of the fuel elements, has holes for receiving the lower end fixtures.


====4.5.2 Reactor====
CHAPTER 4 4.2.2 Control Rods The pulse rod is 3.175 cm. (1.25 in.) diameter. Other rods are 2.225 cm (7/8 in.) diameter.
Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position.
Control rods are 50.8 cm. (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm.
For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates. For comparison purposes, a tabulation of total rod worth for each control element from the State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor).
(30-mil) aluminum sheath.
Pulse Shim Safety Regulating TOTAL Table 4.4; 250 kW Core Parameters.
The control rod drives are connected to the control rod clutches through three extension shafts.
(3 * (effective delayed neutron fraction) 0.007 R. (effective neutron lifetime) 43 :S -$0.017 EC-1 CXTf (prompt temperature coefficient)
The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydraulic cushion (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (slotted tubes on the right hand side) along with the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder; the dimensions of the cylinder limits bottom travel.
(@ 250kW -275EC av (void coefficient)
Upper Pulse, Shim & Reg Rods          Reg Rod          Shim Rod              Pulse Rod Figure 4.5, Control Rod Upper Extension Assemblies The bottom of the pulse rod is shown on the left hand side of Figure 4.5. The upper extension shaft is a hollow tube, the middle extension is solid. The upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fastener).
-0.003 1 %-1 void -$0.006 kw-1 to -CXp (power temperature coefficient-weighted ave) $0.01 kw-1 Table 4.5, Com arison of Control Rod Worths. KSU TRI GA Mark II (250 kW) Core II-19 Core III-1 D-10 $1.96 C-4 $2.12 C-3 $2.88 D-4 $1.85 NA $0.0 D-16 $1.82 D-16 $1.58 E-1 $0.79 NA $6.42 NA $6.58 Cornell University 500k D-10 $1.88 D-16 $2.20 D-4 $1.99 E-1 $0.58 NA $6.65 NOTE: Core III-1 has an experiment positioned to control the worth of the pulse rod K-State Reactor Safety Analysis Report 4-15 Original (12/04) 
Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full in position is above the bottom of the guide tube .
* *
* K-State Reactor Safety Analysis Report 4-8                              Original (12/04)
* CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions.
A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.
4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.
Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRI GA fuel and cladding temperatures associated with pulsing operations.
The analysis addresses the case of a fuel element at an average temperature of 1000&deg;C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470&deg;C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr-1 fr 2 0 p-1 (2550 W m-2 K-1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected.
This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.
: a. Spatial Power Distribution The following conservative approximations are made in characterizing the spatial distribution of the power during steady-state operations.
* The hottest fuel element delivers twice the power of the average. Classically, the radial hot-channel factor for a cylindrical reactor (using Ras the physical radius and Re as the physical radius and the extrapolation distance) is given 2 by: with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,.
However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element. 2 Elements ofNuclear Reactor Design, znct Edition (1983), J. Weisman, Section 6.3 K-State Reactor Safety Analysis Report 4-16 Original (12/04) 
* *
* REACTOR DESCRIPTION Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40. Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.
* The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of n/2 times the average, and heat conduction radial only.
* The axial factor for power produced within a fuel element is given by: g(z) = l.514*co( !!....*
* z ) , '2 2 f + f ext (6) in which e =LI 2 and eex1 is the extrapolation length in graphite, namely, 0.0275 m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor . Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.
The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore: " -p . !!_ . 2 -p p 0 8469 q ma' -83*7l'*D 0 *L 2 -83*Do *L = .. (7)
* The radial and axial distribution of the power within a fuel element is given by q"'(r,z) =
(5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core. 3 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G. B. West K-State Reactor Safety Analysis Report 4-17 Original (12/04) 
* *
* CHAPTER4 a.
* The radial factor is given by: 2 f (r) = a+ er + er , 1 +br+dr 2 (7) in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679.
The fit is illustrated in Figure 4.11.
* 1.3 1.2 1.1 " L 1.0 v ... 0.90 r (cm) Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod. (Data Points from Monte Carlo Calculations
[Ahrens 1999a]) Heat Transfer Models The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the t:i. T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron), cladding inner surface and fuel outer surface -gap (g), and the fuel outer surface to centerline (ricl): Eq. 1 A standard heat resistance model for this system is: K-State Reactor Safety Analysis Report 4-18 Original (12/04) 
* *
* REACTOR DESCRIPTION T =T +q"[_!_+ ro ln(X) c1
* h k rh 2k c I g f Eq. 2 and heat flux is calculated directly as: q"= Ul1T = Tmax -1 r 0 ln(r 0 Ir;) r 0 r 0 * -+ +-+-(2) h . kc ljhg 2k 1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k.r is the fuel thermal conductivity.
The gap conductivity of 2840 W m*2 K-1 (500 Btu h" 1 ft -2 &deg;F" 1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context. Parameters are cross-referenced to source in Table 4.6 . a e T bl 4 6 Th ermo 1ynam1c a ues d . V 1 Parameter Symbol Value Units Reference Fuel conductivity kr 18 Wm-lK"l Table 13.3 14.9 W m*1 K-1 (300 K) Table 13.3 Clad conductivity kg 16.6 W m*1 K-1 ( 400 K) Table 13.3 19.8 W m*1 K-1 (600 K) Table 13.3 Gap resistance ha 2840 wm-2 K-1 AooendixA Clad outer radius ro 0.018161 M Table 13.1 Fuel outer radius fj 0.018669 M Table 13.1 Active fuel length Lr 0.381 M Table 13.1 No. fuel elements N 83 NIA Chap 13 Axial peaking factor APF nl2 NIA Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that: = 5.1858E-04 m 2 K 2kf w Gap resistance has been experimentally determined as indicated, so that: K-State Reactor Safety Analysis Report m 2 K rh W ' g 4-19 Original (12104) 
* *
* CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127&deg;C applies so that: r r ln..CC. 0 r m'K --' =3.103e-5--
k, w T bl 4 7 Cl dd" H T a e . ' a mg eat fl c ffi rans er oe tc1ent Temp (&deg;K) Temp (&deg;C) m 2 K w-1 300 27 3.457e-5 400 127 3.103e-5 600 327. 2.601e-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small. The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.
85000 f 75000 E <' i!!. 65000 1: "' *;:; lE 55000 "' 0 0 ... 45000 c: .. I-'i 35000 :i:: 25000 500 700 Convection Heat Transfer Coefficient 900 1100 1300 1500 Power Level (KW) TRENDLINE:
y = 0.0326x + 16985 R 2 = 0.9976 1700 1900 Figure 4.10, Convection Hear Transfer Coefficient versus Power Level K-State Reactor Safety Analysis Report 1 h 0.0326P(watts)  
+ 16985 4-20 Original (12/04) 
* *
* REACTOR DESCRIPTION Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as: T, = T, + 0.423P[ 1 + 3.103e-5 + 3.620e-4 + 5.186e-4]
(10) < 0.0326P + 16985 c. Steady-State Mode of Operation Centerline temperature calculations were performed on a "reference core" using the model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation.
A core with more than. 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 20&deg;C and 100&deg;C was assumed with the following results: Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel Centerline
&deg;C 503.2 582.0 Fuel/Gap Interface
&deg;C 229.0 307.8 Gap/Clad Interface
&deg;C 37.7 116.4 Clad/Water Bulk Water &deg;C Interface
&deg;C 21.2 100.0 20.0 100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 &deg;C at power levels up to 1900 kW with pool temperature at 27 &deg;C (300 K), and 1700 kW with pool temperatures at I 00 &deg;C . K-State Reactor Safety Analysis Report 4-21 Original (12/04)
* *
* CHAPTER4 Hot Fuel-Rod Centerline Temperature at Power (Temperature 8evation over Pool Water Tern perature)
I-u -* 300 K --600 K I e-600 ::I 500 Cll Cl.
I-300 100 0 100 300 500 700 900 1100 1300 1500 1700 1900 Reactor Power (kW) Figure 4.11, Hot Fuel-Rod Centerline Temperature For the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD


===3.3 patch===
REACTOR DESCRIPTION
04 (Feldman 2008). A snapshot of the model is presented in Figure 4.12. It has two dependent volumes, enforcing the pressure boundary conditions, and two pipes, simulating the cold and hot channel connected via a single junction component of RELAP. Heat is added to the fluid by incorporating the heat structure component (simulating a fuel element) of RELAP with an appropriate axial power profile and power level. In this analysis, the power level for the B ring is at 24 kW (corresponding to an 85-element core with a ring-to-average peaking factor of 1.63). This power level is applied to the heat structure within the single channel. The model assumes an operating pressure of 143 kPa, and an operating temperature of 322.15 K ( 49. l 5&deg;C). The version of the RELAP code licensed to KSU uses PG-CHF correlation which is a state of the art best estimate CHF correlation developed by Nuclear Research institute of Rez in the Czech Republic.
* i.4 .. t~*n.o':t!-'
It is based on data in the Czech Republic data bank from 173 different sets of tube data, 23 sets of annular data, and 153 sets of rod bundle data. There are four forms of the PG-CHF correlation
{
'Basic', 'Flux', 'Geometry', and 'Power'. For the rod bundle it is applicable in the pressure range of0.28 MPato 18.73 MPa, for a mass flux of34.l to 7478 kg/s-m2, for 0.4-7.0 m length and for a diameter of 0.00241 to 0.07813 m. TRIGA has an operating pressure of 0.143 MPa and fuel rod length of 0.381 m, thus the operating conditions fall outside the range of the applicability of the PG-CHF correlation, and a different correlation is required to assess the . departure from nucleate boiling ratio (DNBR ratio). One such correlation which is applicable for the low pressure range observed in TRIGA reactor facility is the Bernath correlation.
* 4' I t *,1cr
The functional form of the Bernath correlation can be presented in the following equations . K-State Reactor Safety Analysis Report 4-22 Original (12/04) 
            .. <1-J.**t~t'l':t~*
* *
                                                  -*-*,~~.wy-...-.,.
* REACTOR DESCRIPTION I Outlet i .. Hot-leg Cold-Leg + I connector Figure 4.12 -RELAP single channel model used in CHF analysis K-State Reactor Safety Analysis Report 4-23 (8) Original (12/04) 
                                                                                    .-~~:~~~~: ~"
* *
                "' **-~~~
* CHAPTER 4 = , if Dh :<S; O.lft h = _!Q_ + 90, if Dh ;::: 0.1.ft D,, hBo =film coefficient at CHF D,, =hydraulic dia.meter (ft) v =coolant velocity (ft Is) TwBo =wall temperature at burnout (&deg; C) DH =heated diameter (ft) The RELAP simulations were performed for the hot channel, i.e., a channel with a radial peaking factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain the pressure, temperature, and velocity distribution at different axial locations.
Figure 4.6, Middle Extension Rod Alignment Holes The middle solid extension is similarly connected to the lower extension. The lower extension is hollow, the middle extension fits into the lower extension and a hole drilled in the overlap secures the lower extension to the middle extension. Typically the lower extension has a tighter fit than the upper extension because the lower and middle extension are not separated for inspections and because the interface with upper extension is used to set the bottom position of the control rod.
With these calculations and the functional form of the Bernath correlation, the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA, the A-ring is occupied by the central thimble, not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard. The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and Figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel. K-State Reactor Safety Analysis Report 4-24 Original (12/04) 
Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7 ..
* * * ---.,., '* :;;:: ...:.o:: u... :c u REACTOR DESCRIPTION 5000-------------------------
Figure 4.7, Standard & Pulse Rod Lower Coupling The bottom of the lower extension attaches directly to the control rod. Pictures of the control rods taken during the 2003 control rod inspection are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wire in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube .
A .. Bernatl1-CHF
* K-State Reactor Safety Analysis Report 4-9                              Original (12/04)
.. A PG-CHF 4000 *
* Heat flux .. _., A A .. Jo. .. 3000 A .I. A 0. A A A A A :A A
* 2000 1000 :r .;. * . .,, ..
__ *35-.... Heated Length {m) Figure 4.13 -CHF versus heated length + + Bernath-CHF
: u. * + PG-CHF 10 -** .. ... :
* B* * * * * . . .. ** * ...... * ..
* o! -.** * * * * * ** ** * *** * * . ....... ........ .... .. 2 8.oo 0.05 O.lG O.lS 0.20 0_25 0.30 Heated length {n1*) Figure 4.14 -DNBR versus heated length * * * . '*.
* 0.35 K-State Reactor Safety Analysis Report 4-25 Original (12/04)  
* *
* CHAPTER4 d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code T ASCOT for transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. (9) and (10), with the peak fuel temperature set to 746 &deg;C, i.e., a temperature rise of 719 &deg;C above 27 &deg;C ambient temperature.
The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1 % ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation.
In the TASCOT calculations, thermal conductivity was set to 0.18 W cm*1 K-1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm*1 K-1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994) (9) The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955) [ k 0.79
* 0.45
* 0.51 l . -* f c P.f fl, * ( )0.99 hh -0.00122 0.75 T,.. -T.at , cr o.s * &#xb5; 0.29
* p 0.24 * (v _ v )
* To.1s f g g v sat (10) in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 &deg;C), and Tb the coolant ambient temperature (27&deg;C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases. The overall heat transfer coefficient U varies. negligibly for ambient temperatures from 20 to 60 &deg;C, and has the value 0.21 W cm*1 K-1 at Tb= 27 &deg;C. Figure 4.15 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.10 lists temperatures and heat fluxes as function of time after a 2.1 % ($3 .00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m*1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500&deg;C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse . K-State Reactor Safety Analysis Report 4-26 Original (12/04) 
* * * ....... () 0 ...__, <!) L :J +' nl L <!) Q_ E <!) I-REACTOR DESCRIPTION 1000 Os 800 1 2 600 4 8 400 16 200 32 64s 0 '---'---'---'--_._----'-----'----'---'----'___J'--.__-'---'----'--'-----'----'----'---'----'___JL-....1 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0 2.2 Radius (cm) Figure 4. 15, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse . K-State Reactor Safety Analysis Report 4-27 Original (12/04) 
* *
* CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{0 C) Coolant Ambient Temperature.
Time (s) 0 1 2 4 8 16 32 64 Q" Fuel outside (W m-2) CHFR Temp. (oC) 3.57 x10 5 7.34 xl0 5 8.52 x10 5 7.54 xl0 5 5.71 xl0 5 3.46 x10 5 1.04 xl0 5 4.2 2.0 1.7 2.0 2.6 4.3 14.4 953 781 683 574 461 344 224 100 4.6 Thermal Hydraulic Design and Analysis Clad surface Temp. (&deg;C) 224 432 498 443 342 218 84 A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by (11) in which Po and 130 are the density and volumetric expansion coefficient at core inlet conditions (27&deg;C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm 2 s-1 , l':iT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by (12) in which mis the coolant mass flow rate (kg s-1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m 2 , and D1i is the hydraulic diameter, namely, 0.02127 m. The friction factor/for laminar flow through the annular area is given by 100 Re-1 (Shah & London 1978), in which the Reynolds number is given by D,,rh I A&#xb5;0 in which &#xb5;o is the dynamic viscosity at core inlet conditions.
Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5), and the entrance pressure loss would be expected to be negligible.
The exit contraction loss is given by K-State Reactor Safety Analysis Report 4-28 (13) Original (12/04)
* *
* REACTOR DESCRIPTION The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which _ _ [ 3
* R 2 sin60&deg; cos60&deg; ]-K = A -* 2 -0.171, c 11 R (14) where R is the radius of the opening in the upper grid plate. Equations (12) through (14), solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4. I 1. Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRIGA Reactor During Steady-State Operations.
P (kWt) m (kg s-1) 50 0.047 100 0.061 200 0.077 300 0.090 400 0.100 500 0.108 750 0.125 1000 0.139 1250 0.150 4. 7 Safety Limit !).T (&deg;C) 3.1 4.7 7.5 9.6 11.5 13.3 17.2 20.6 23.8 As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.
The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500&deg;C; consequently, limits on fuel temperature change for cladding temperatures greater than 500&deg;C. A maximum temperature of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zr1.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.


===4.8 Operating===
CHAPTER 4 Pulse Rod            Shim Rod                Reg Rod Figure 4.8, Control Rods During 2003 Inspection
* Full Guide Tube Upper        Lower      Lower Detail        Position in Upper Grid Plate Figure 4.9, Control Rod Guide Tubes
* K-State Reactor Safety Analysis Report 4-10                            Original (12/04)


Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature.
REACTOR DESCRIPTION
The operating parameters established for the KSU TRI GA reactor are: K-State Reactor Safety Analysis Report 4-29 Original (12/04) 
* a.      Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings. A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (1) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.
* *
Ta ble 4.2 SummaryofC on t ro IR0 dint er Iocks INTERLOCK                  'SETPOINT                      FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument startup channel reading is less Source Interlock                        2 cps than instrument sensitivity/ensure nuclear instrument startup channel is operating Prevent applying power to pulse rod unless Pulse Rod Interlock              Pulse rod inserted rod inserted/prevent inadvertent pulse Withdraw signal,      Prevent withdrawal of more than I rod/Limit Multiple Rod Withdrawal more than I rod        maximum reactivity addition rate Mode switch in Hi      Prevent withdrawing standard control rods in Pulse Mode Interlock Pulse            pulse mode Prevent pulsing if power level is greater than Pulse-Power Interlock                  IO kW IO kW NOTE: (1) Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock required by Technical Specifications
* CHAPTER 4
: b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times. for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIG A prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation .
* Steady-state power level
* K-State Reactor Safety Analysis Report 4-11                                      Original (12/04)
* Fuel temperature measured by thermocouple during pulsing operations
* Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations.
As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100&deg;C bulk pool water temperature for 1.25 MW operation.
Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin. Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.  


====4.8.3 Safety====
CHAPTER 4 T a ble 4.3 ' S ummarvofReac t or SCRAMs Limitin~  Trip Setpoint Measuring Steady                                  Actual Setpoint Channel                                Pulse State Linear Channel High        110%              NIA                      104%
Margins For 1,250 kWth steady-state operations, the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49&deg;C. The proposed Technical Specifications limit of 44&deg;C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation.
Power Power Channel High power 110%              NIA                      104%
Limiting pool inlet water temperature to no greater than 44&deg;C (or 37&deg;C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling. Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746&deg;C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869&deg;C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15% margin to the fuel temperature safety limit for cladding temperatures below 500&deg;C. Analysis shows that cladding temperatures will remain below 500&deg;C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve 500&deg;C (invoking a 950&deg;C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.
Detector High 90%              90%                      90%
Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500&deg;C, but the heat transfer simultaneously cools the fuel to much less than 950&deg;C . K-State Reactor Safety Analysis Report 4-30 Original (12/04)
Voltage 600&deg;C B Ring element High Fuel            555&deg;C C Ring element                        450&deg;C Temperature[ll      480&deg;C D Ring element 3 80&deg;C E Ring element                      350&deg;C Period [IJ            NIA              NIA                      3 sec NOTE [1]: Period trip and temperature trip are not required by Technical Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a
* *
* shutdown margin of at least $0.50 .
* REACTOR DESCRIPTION If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature.
Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications.
In worst-case air-cooling scenarios, cladding temperature can exceed 500&deg;C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500&deg;C. 4.9 Bibliography "TASCOT: A 2-D, Transient and Steady State Conduction Code for Analyhsis of a TRIGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C., "Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C., ''A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C., Kansas State University TRIGA MkII Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J.G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K., and N. Zuber, Theory and Design of Modern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F., "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al., MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Zirconium Hydride Fuels for TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987 . K-State Reactor Safety Analysis Report 4-31 Original (12/04) 
4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors.
* *
The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3 .18-cm ( 1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector. *
* CHAPTER 4 "Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K., and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRIGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T. "Safety Analysis Report, TRIGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991. K-State Reactor Safety Analysis Report 4-32 Original (12/04) 
* K-State Reactor Safety Analysis Report 4-12                          Original (12/04)
* *
* Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).
* The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A. l. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature.
A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature.
Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies
[Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface. Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties.
Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100&deg;F water, showed burnout fluxes of 0.9 to 2.0 Mbtu fr 2 hr-1 for e-folding periods from 5 to 90 milliseconds.
On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft-2 hr-1* It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.
A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A.1 and 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent.
A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise. The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760&deg;C clad temperature for a peak initial fuel temperature of 1000&deg;C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure'4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor Safety Analysis Report 4.A-1 Original (9/02) 
* *
* CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation.
Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470&deg;C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400&deg;C to 500&deg;C for TRIGA Mark F fuel elements which have been operated under film boiling conditions
[Coffer et al. 1965]. I. 2 I. 1 1.0 0.9 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 RADIUS (IN.) Figure 4A.1. Representative Radial Variation of Power Within the TRIGA Fuel Rod I.I 1.0 0.9 ;::; 0.8 L;:" 0.7 0.6 0.5 0 2 4 6 7 8 AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.) Figure 4A.2, Representative Axial Variation of Power Within the TRIGA Fuel Rod . K-State Reactor Safety Analysis Report 4.A-2 Original (9/02) 
* * * \ \ \ \ ' ' ' REACTOR DESCRIPTION CURVE BASED ON DATA OF ELLIO/. ' "' ' ...... ______ .,,, TW-TSAT (*f) Figure 4A.3, Subcooled Boiling Heat Transfer for Water . 1800 1700 1600 ... ... i= 1500 1400 1300 1200 K-State Reactor Safety Analysis Report 0 [LAPSED TIHE FROH END OF PULSE 0.1 0.2 0.3 0.4 o.s o.6 0. 7 0.8 RADIUS (IN.) Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse . 4.A-3 Original (9/02) 
* *
* CHAPTER 4 APPENDIX A ONSET OF _!_l PEAK HEAT FLUX N ... .... .;, 105 "' ... .. "' :::> ... .... ... c .... x .... 10 4 u c ... "' ;;: 103 0.001 NUCLEATE 801 LI NG 0.01 I ONSET or STABLE Fii.Joi llOlllNG 0.1 1.0 ELAPSED TIME FROM ENO OF PllLSE (SEC) 10 100 Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse . '0. 000 ,..........,.-..,......,........,.--,--..-.,......,.....,....-..--.;...,r--i,..-,-..-......,--r-r"T"T--.--..-.,.-,-, \000 100 -----CLAO OUTER SURFACE TEMP j io
.........
,0--:-._.__
..............
0.001 0.01 0. I 1.0 ELAPSED TIME FROM END OF PULSE (SEC) Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element . K-State Reactor Safety Analysis Report 4.A-4 Original (9/02) 
*
* REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface.
Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.
This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected.
Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962] which provide insight into the mechanisms involved.
They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient.
These results were then compared with the incipient film boiling conditions of the 1000&deg;C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation.
The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors.
The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960] which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate. The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft-2 hr-1 &deg;F-1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft-2 hr-1 &deg;F-1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel. The surface heat flux at the mid plane of the element is shown in Figure 4A. l 0 with gap conductance as a parameter.
It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft-2 hr-1 &deg;F-1 according to the more conservative Bernath correlation at a local water temperature approaching saturation.
K-State Reactor Safety Analysis Report 4.A-5 Original (9/02) 
* *
* CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000&deg;C is conservatively estimated to be 470&deg;C. As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470&deg;C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment.
Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000&deg;C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is 1150&deg;C would not produce a stress in the clad in excess of its ultimate strength.
TRI GA fuel with a hydrogen to zfrconium ratio of at least 1.65 has been pulsed to temperatures of about l l 50&deg;C without damage to the clad [Dee et al. 1966]. 7 ...... ELAPSED TIME FROM IN I-*END OF PULSE
* 0.2.47 SEC u.. 6 'I a: :::c: ...... ::i I-al IJ\ 5 I 0 ><: x :::> ...J u.. 4 t-c:t ..... :c 3 7 8 9 10 11 12. 13 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A. 7, Surface Heat Flux Distribution for Standard Non-Gapped
{hgap= K-State Reactor Safety Analysis Report 500 Btu/h ft 2 &deg;F) Fuel Element After a Pulse . 4.A-6 Original (9/02) 
* *
* REACTOR DESCRIPTION 8 7 ... ...... .... 6 .;, :i: ..... era "' 5' 0 ""
HEAT FLUX >< . :::> It .... .... .... .. ... :i: 3 2 7 8 ELAPSED TIHE FROH ENO OF PULSE IS 0.311t SEC 9 10 11 12 DISTANCE FROH BOTTOM OF FUEL (IN.) 13 15 Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 375 Btu/h ft2 &deg;F) After a Pulse. N I-"-. a: ::r:: -=> I-m U\ . 0 )C x => "-I-er: ..... :c B 7 6 5 4 3 1 1 7 8 ELAPSED TIME FROM ENO OF PULSE IS 0.440'SEC 9 10 11 12 13 DISTANCE FROM BOTTOM OF FUEL (IN.) 14 15 Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap= 250 Btu/h ft 2 &deg;F ) After a Pulse . K-State Reactor Safety Analysis Report 4.A-7 Original (9/02) j 
* *
* CHAPTER 4 APPENDIX A >< ::> ...J i.. ..... "' w :c w u *< i.. ! 101+ .,, EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR-FT 2 -&deg;F 500 FLOW VELOCITY a I FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME= ZERO) 0.01 0.1 1.0 ELAPSED TIHE FROM END OF PULSE (SEC} Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse . K-State Reactor Safety Analysis Report 4.A-8 Original (9/02) 
* *
* I REACTOR DESCRIPTION Bibliography "A Theory of Local Boiling Burnout and Its Application to Existing Data, " Heat Transfer -Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No. 20.Bernath, L., Research in Improved TRIGA Reactor Pe1formance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al., Characteristics of Large Reactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al. Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al., A Study of the Mechanism of Boiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E., Thermal Conductance of Metallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis of a Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J. Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172. Graff, W.J. Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H .. MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H. Choi, "Quarterly Technical Report SPERT Project, April, May, June, 1964, "ISO 17030. Spano, A. H., "The Effect ofSubcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis, " Technical report, Argonne National Laboratory, 2008, E.E. Feldman . K-State Reactor Safety Analysis Report 4.A-9 Original (9/02) 
* *
* CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure, system models, and solution methods. "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution." Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong . K-State Reactor Safety Analysis Report 4.A-10 Original (9/02) 
* *
* REACTOR DESCRIPTION "Onset of Stable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al., UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991. "Hydrodynamic Aspects of Boiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959. Zuber, W . K-State Reactor Safety Analysis Report 4.A-11 Original (9/02) 
* *
* Tpool oc 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 97 99 Tpool oc 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 97 99 Appendix B Water Properties at Nominal Operating Conditions 999.21 998.32 997.16 995.75 994.12 992.29 990.27 988.07 985.70 983.18 980.50 977.69 974.74 971.66 968.45 965.12 961.68 960.27 958.84 999.21 998.32 997.16 995.75 994.12 992.29 990.27 988.07 985.70 983.18 980.50 977.69 974.74 971.66 968.45 965.12 961.68 960.27 958.84 Data for 16 Feet of Water over tile Core Pt,16[1 1 kg m-3 950.00 950.01 950.02 950.03 950.04 950.06 950.07 950.09 950.11 950.12 950.14 950.17 950.19 950.21 950.23 950.26 950.29 950.30 950.31 951.43 951.43 951.44 951.45 951.46 951.47 951.49 951.50 951.51 951.53 951.55 951.57 951.58 951.60 961.62 951.64 951.67 951.68 951.68 47.79 47.74 47.69 47.62 47.54 47.46 47.36 47.25 47.14 47.02 46.89 46.76 ht,16[1 1 kJ kg-1 465.10 465.05 465.01 464.95 464.89 464.81 464.73 464.64 464.54 464.44 464.33 464.21 46.62 464.09 46.47 463.96 46.32 463.83 46.16 463.69 45.99 463.55 45.92 463.49 45.86 463.43 hg,161 1 1 kJ kg-1 2692.64 2692.63 2692.59 2692.59 2692.57 2692.54 2692.51 2692.48 2692.45 2692.41 2692.37 2692.33 2692.29 2692.24 2692.19 2692.15 2692.09 2692.07 2692.05 P 9 ,151 1 1 kg m-3 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.84 Tsat,161 1 1 oc 110.89 110.88 110.87 110.86 110.84 110.83 110.81 110.78 110.76 110.74 110.71 110.68 q "sat, 15!3! wm-2 1553.842 1552.078 1549.496 1547.118 1543.981 1540.446 1536.512 1532.205 1527.561 1522.575 1517.255 1511.666 0.84 110.65 1505.778 0.84 110.62 1499.613 0.84 110.59 1493.199 0.84 110.56 1486.527 0.84 110.53 1479.626 0.84 110.51 1472.944 0.84 110.50 1466.058 Data for 13 Feet of Water over tile Core 38.83 38.79 38.75 38.69 38.63 38.56 38.48 38.39 38.30 38.20 38.10 37.99 37.88 37.76 37.63 37.50 37.37 37.31 37.26 ht,13[1 1 kJ kg-1 457.21 457.18 457.13 457.09 457.03 456.96 456.89 456.82 456.73 456.64 456.55 456.45 456.35 456.24 456.12 456.01 455.89 455.84 455.78 hg,131 1 1 kJ kg-1 2689.85 2689.84 2689.82 2689.80 2689.78 2689.76 2689.74 2689.71 2689.68 2689.65 2689.61 2689.58 2689.54 2689.50 2689.46 2689.42 2689.38 2689.36 2689.34 Pg,13111 kg m-3 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.80 0.79 0.79 0.79 0.79 Tsat,131 1 1 oc 109.03 109.02 109.01 109.00 108.99 108.97 108.96 108.94 108.92 108.90 108.87 108.85 108.83 108.80 108.77 108.75 108.72 108.70 108.69 q"sa1,13!3 l Wm-2 1513.00 1511.32 1509.15 1505.85 1503.13 1500.16 1496.38 1492.25 1487.78 1482.99 1477.90 1472.52 1466.86 1460.95 1458.59 1448.37 1441.74 1435.27 1428.60 q"sub[4 1 Wm-2 7239.19 6931.74 6622.60 6311.82 5999.91 5688.24 5376.29 5064.36 4753.90 4444.85 4136.85 3830.73 3526.89 3225.47 2926.81 2631.05 2338.47 2216.11 2095.18 q"subl41 wm-2 6964.74 6857.12 6543.62 6229.30 5913.58 5597.21 5281.90 4966.66 4652.39 4339.60 4027.94 3718.83 3412.07 3107.29 2812.63 2506.90 2211.19 2087.92 1966.16 K-State Reactor Safety Analysis Report 4.B-1 Revised 05/01/17 
*
* REACTOR DESCRIPTION Common Data T Patm Cp[1] CJ oc kPa kJ kg-1 k-1 Nm-1 15 99.83 4.23080 0.07149 20 99.83 4.23080 0.07120 25 99.83 4.23080 0.07083 30 99.83 4.23080 0.07039 35 99.83 4.23070 0.06989 40 99.83 4.23070 0.06932 45 99.83 4.23070 0.06869 50 99.83 4.23070 0.06800 55 99.83 4.23060 0.06727 60 99.83 4.23060 0.06649 65 99.83 4.23060 0.06566 70 99.83 4.23050 0.06480 75 99.83 4.23050 0.06390 80 99.83 4.23050 0.06297 85 99.83 4.23040 0.06201 90 99.83 4.23040 0.06102 95 99.83 4.23040 0.06001 97 99.83 4.23030 0.05898 99 99.83 4.23030 0.05793 NOTE[1}: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa
=Heigth(ft)
* 12(inlft)
*0.0254(meterslin)
*Density(kglm
: 3) *9.8066511000 " 0.5 ( ) ( { })Y. NOTE[3}: qSAT =0.I49*pg . hg,sat-hf,sat.
*g*<J* Pt-Pg ( Y. J NOTE[4']* " -" . 1 0 l*(P1J . cp.f *(TsAT -Tsub)
* qsub -qSAT +
* Pg hg,sat -h /,sat NOTE:[5}:
<J = 1.000E-11*T 4+7.370E-09
* T 3 -1.969E-06
* T 2 + 4.709E-06
* T + 7.1833E-02 K-State Reactor Safety Analysis Report 4.A-13 Original (9/02)
* TECHNICAL SPECIFICATIONS Table of Contents I. DEFINITIONS
.................................................................................................................
TS-I 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .......................
TS-8 2.1 Fuel Element Temperature Safety Limit ......................................................................
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 Limiting Safety System Settings ................................................................................
TS-I 0 2.2.1. Applicability
......................................................................................................
TS-I 0 2.2.3. Objective
............................................................................................................
TS-10 2.2.4. Specification
.......................................................................................................
TS-I 0 2.2.5. Actions ...............................................................................................................
TS-I 0 2.2.6. Basis ..................................................................................................................
TS-10 3. LIMITING CONDITIONS FOR OPERATIONS
.........................................................
TS-11 3.1 CORE REACTIVITY
................................................................................................
TS-II 3.1.1. Applicability
......................................................................................................
TS-11 3.1.3. Objective
............................................................................................................
TS-II 3.1.4. Specification
.......................................................................................................
TS-11 3.1.5. Actions ...............................................................................................................
TS-12 3.1.6. Basis ..................................................................................................................
TS-13
* 3.2 PULSED MODE OPERATIONS
..............................................................................
TS-13 3.2.1. Applicability
......................................................................................................
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 MEASURING CHANNELS ......................................................................................
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. Applicability
......................................................................................................
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 GASEOUS EFFLUENT CONTROL .........................................................................
TS-20 3.5.1. Applicability
......................................................................................................
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 LIMITATIONS ON EXPERIMENTS
..........................................................................
TS-22 3.6.1. Applicability
......................................................................................................
TS-22 3.6.3. Objective
............................................................................................................
TS-22 K-State Reactor TS-1 Original (9tG+-4/1744)
* L_ ____ _
* TECHNICAL SPECIFICATIONS


====3.6.4. Specification====
REACTOR DESCRIPTION
* The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank.
Four lugs are provided for lifting the assembly. A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within the fuel cladding provide additional reflection. Inserts are placed at both ends of the fuel meat, providing top and bottom reflection.
4.2.4 Neutron Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 106 n s- 1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor.
4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates.
The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core.
Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position. The reflector assembly supports both grid plates.
4.3      Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion. Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration. All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool
* K-State Reactor Safety Anaiysis Report 4-13                              Original (12/04)


.......................................................................................................
CHAPTER 4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak.
TS-22 3.6.5. Actions ...............................................................................................................
A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations.
TS-22 3.6.6. Basis ..................................................................................................................
Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter I 0).
TS-23 3.7 FUEL INTEGRITY
4.4     Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield.
...................................................................................................
The shielding configuration is similar to those at other TRIGA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank.
TS-24 3.7.1. Applicability
The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena.
......................................................................................................
4.5     Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels with 8.5% U, 20%
TS-24 3.7.3. Objective
enrichment, the value is nearly constant at 0.01 % ~k/k per &deg;C, and varies only weakly dependent on geometry and temperature.
............................................................................................................
Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures*
TS-24 3. 7.4. Specification
that the fuel and clad temperature limits will not be exceeded. The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity.
.......................................................................................................
Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety
TS-24 3.7.5. Actions ...............................................................................................................
* K-State Reactor Safety Analysis Report 4-14                              Original (12/04)
TS-24 3. 7 .6. Basis ..................................................................................................................
TS-24 3.8 REACTOR POOL WATER .........................................................................................
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 Maintenance 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 REACTIVITY
................................................................................................
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 MEASURING CHANNELS ......................................................................................
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 LIMITATIONS ON EXPERIMENTS
........................................................................
TS-34 4.6.1. Objective
...........................................................................................................
TS-34 4.6.2. Specification
......................................................................................................
TS-34 4.6.3. Basis ..................................................................................................................
TS-34 4. 7 FUEL 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 WATER .......................................................................................
TS-36 4.8.1. Objective
...........................................................................................................
TS-36 K-State Reactor TS-2 Original (91G+4/1744)  
*
* TECHNICAL SPECIFICATIONS


====4.8.2. Specification====
REACTOR DESCRIPTION
* limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6.
4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity. The bottom grid plate, which suppmis the weight of the fuel elements, has holes for receiving the lower end fixtures.
4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates.
For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor).
Table 4.4; 250 kW Core Parameters.
(3 * (effective delayed neutron fraction)                        0.007 R. (effective neutron lifetime)                                  43 :S
                                                                              -$0.017 EC- 1 CXTf (prompt temperature coefficient)                    (@ 250kW -275EC av (void coefficient)                                      -0.003 1%- 1 void
                                                                            -$0.006 kw- 1 to -
CXp (power temperature coefficient- weighted ave)              $0.01 kw- 1 Table 4.5, Com arison of Control Rod Worths.
KSU TRI GA Mark II (250 kW)                            Cornell University Core II-19                        Core III-1                      500k Pulse                D-10              $1.96            C-4          $2.12          D-10          $1.88 Shim                  C-3              $2.88            D-4          $1.85          D-16          $2.20 Safety                NA                $0.0            D-16          $1.82          D-4          $1.99 Regulating          D-16              $1.58            E-1          $0.79          E-1          $0.58 TOTAL                NA              $6.42            NA            $6.58          NA            $6.65 NOTE: Core III-1 has an experiment positioned to control the worth of the pulse rod
* K-State Reactor Safety Analysis Report 4-15                                Original (12/04)


......................................................................................................
CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.
TS-36 4.8.3. Basis ..................................................................................................................
4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.
TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS
Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 1000&deg;C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470&deg;C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr- 1 fr 2 0 p- 1 (2550 W m-2 K- 1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.
.......................................................
: a.        Spatial Power Distribution The following conservative approximations are made in characterizing the spatial distribution of the power during steady-state operations.
TS-37 4.9.1. Objective
* The hottest fuel element delivers twice the power of the average.
...........................................................................................................
Classically, the radial hot-channel factor for a cylindrical reactor (using Ras the physical radius and Re as the physical radius and the extrapolation distance) is given 2 by:
TS-37 4.9.2. Specification
with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,. However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element.
......................................................................................................
2 Elements ofNuclear Reactor Design,    znct Edition (1983), J. Weisman, Section 6.3
TS-37 4.10.3. Basis ................................................................................................................
* K-State Reactor Safety Analysis Report 4-16                              Original (12/04)
TS-37 5. DESIGN 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 BUILDING .............................................................................................
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 EXPERIMENTS
.........................................................................................................
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. ADMINISTRATIVE CONTROLS .................................................................................
TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL..
........................
TS-44 6.2 REVIEW AND AUDIT .............................................................................................
TS-45 6.3 PROCEDURES
............................................................................................................
TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS
..................................................
TS-47 6.5 EMERGENCY PLAN AND PROCEDURES
...........................................................
TS-48 6.6 OPERATOR REQUALIFICATION
..........................................................................
TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE
....................................................................
TS-48 6.10 PLANT OPERATING RECORDS ............................................................................
TS-49 6.11 REPORTING REQUIREMENTS
...........................................................
TS-50 K-State Reactor TS-3 Original (9,LG+.4/1744)
* 
* *
* 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 ANNUAL CHANNEL CALIBRATION BIENNIAL CHANNEL CHECK CHANNEL TEST CONTROL ROD (STANDARD)
CONTROL ROD (TRANSIENT)
DAILY K-State Reactor 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 12 months, not to exceed 15 months A channel calibration is an adjustment of the channel to that its output responds, with acceptable range and accuracy, to known values of the parameter that the channel measures.
Every two years, not to exceed a 28 month interval A channel check is a qualitative verification of acceptable performance by 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. 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. A standard control rod is one having an electric motor drive and scram capability.
A transient rod is one that is pneumatically operated and has scram capability.
Prior to initial operation each day (when the reactor is operated), or before TS-4 Original (9fG+.4/1744) 
* *
* ENSURE EXHAUST PLENUM EXPERlMENT EXPERlMENT AL FACILITY IMMEDIATE INDEPENDENT EXPERlMENT LIMITING CONDITION FOR OPERATION (LCO) LIMITING SAFETY SYSTEM SETTING (LSSS) MEASURED VALUE MEASURING CHANNEL MOVABLE EXPERlMENT NONSECURED EXPERlMENT K-State Reactor TECHNICAL SPECIFICATIONS an operation extending more than I day Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition The air volume in the reactor bay atmosphere between the pool surface and the reactor bay exhaust fan An EXPERlMENT is (I) any apparatus, device, or material placed in the reactor core region (in an EXPERlMENT AL FACILITY associated with the reactor, or in line with a beam ofradiation emanating from the reactor) or (2) any in-core operation designed to measure reactor characteristics.
Experimental facilities are the beamports, thermal column, pneumatic transfer system, central thimble, rotary specimen rack, and the in-core facilities (including non-contiguous single-element positions, and, in the E and Frings, as many as three contiguous fuel-element positions).
Without delay, and not exceeding one hour. NOTE: IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action INDEPENDENT Experiments are those not connected by a mechanical, chemical, or electrical link to another experiment The lowest functional capability or performance levels of equipment required for safe operation of the facility.
Settings for automatic protective devices related to those variables having significant safety functions.
Where a limiting safety system setting is 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.
The measured value of a parameter is the value as it appears at the output of a MEASURING CHANNEL. A MEASURING CHANNEL is the combination of sensor, lines, amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.
A MOVABLE EXPERlMENT is one that may be moved into, out-of or near the reactor while the reactor is OPERA TING. NONSECURED Experiments are these that should not move while the reactor is OPERATING, but are held in place with less restraint than a secured experiment.
TS-5 Original {9fG.74/1744) 
* *
* OPERABLE OPERATING PULSE MODE REACTOR SAFETY SYSTEM REACTOR SECURED MODE REACTOR SHUTDOWN RING REFERENCE CORE CONDITION SAFETY CHANNEL SECURED EXPERIMENT K-State Reactor TECHNICAL SPECIFICATIONS A system or component is OPERABLE when it is capable of performing its intended function in a normal manner A system or component is OPERA TING when it is performing its intended function in a normal manner. 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, reactor power may be increased on a period of much less than l second by motion of the transient control rod. The REACTOR SAFETY SYSTEM is that combination of MEASURING CHANNELS and associated circuitry that is designed to initiate reactor scram or that provides information that requires manual protective action to be initiated.
The reactor is secured when the conditions of either item (1) or item (2) are 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 The reactor is shutdown if it is subcritical by at least $1.00 in the REFERENCE CORE CONDITION with the reactivity worth of all experiments included.
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, The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible
(<$0.30) A safety channel is a MEASURING CHANNEL m the REACTOR SAFETY SYSTEM A secured EXPERIMENT is an EXPERIMENT held firmly in place by a mechanical device or by gravity providing that the weight of the EXPERIMENT is such that it cannot be moved by force ofless than 60 lb. TS-6 
* *
* SECURED EXPERIMENT WITH MOVABLE PARTS SHALL (SHALL NOT) SEMIANNUAL SHUTDOWN MARGIN STANDARD TECHNICAL SPECIFICATIONS A secured EXPERIMENT with movable parts is one that contains parts that are intended to be moved while the reactor is OPERATING.
Indicates specified action is required/(not to be performed)
Every six months, with intervals not greater than 8 months The shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating condition, and that the reactor will remain subcritical without further operator action 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 MODE TECHNICAL SPECIFICATION VIOLATION K-State Reactor The reactor is in the steady-state mode when the reactor mode selector switch is in either the manual or automatic position and the key switch is in the "on" position.
A violation of a Safety Limit occurs when the Safety Limit value is exceeded.
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 applicable titles of sections in individual Technical Specifications TS-7 Original (fhlG+.4/1744)
* *
* 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====
REACTOR DESCRIPTION
* Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40.
Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.
* The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of n/2 times the average, and heat conduction radial only.
The axial factor for power produced within a fuel element is given by:
g(z) = l.514*co( !!....*
* z        ) ,                  (6)
                                                          '2      2 f +f  ext in which e =LI 2 and eex1 is the extrapolation length in graphite, namely, 0.0275
: m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor .
Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.
* The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore:
                            "    -      p      . !!_ . 2 -    p      p 0 8469                  (7) q  ma' -  83*7l'*D0 *L 2 - 83*Do *L = . .
* The radial and axial distribution of the power within a fuel element is given by q"'(r,z) =     q;:~f(r)g(z),                            (5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core.
3 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G. B. West
* K-State Reactor Safety Analysis Report 4-17                          Original (12/04)


(1) Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed l 150&deg;C. (2) Steady state fuel temperature shall not exceed 750&deg;C. 2.1.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A Stainless steel clad, high-A.I Establish SHUTDOWN A.I IMMEDIATE hydride fuel element condition temperature exceeds l 150&deg;C. OR AND Fuel temperature exceeds 7 5 0&deg;C in steady state A.2 Report per Section 6.8 A.2 Within 24 hours conditions
CHAPTER4
* The radial factor is given by:
2 f (r) = a+ er + er    ,                                 (7) 2 1+br+dr in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679.
The fit is illustrated in Figure 4.11.
* 1.3 1.2 1.1 "L  1.0 v
0.90
* a.
r  (cm)
Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod.
Heat Transfer Models (Data Points from Monte Carlo Calculations [Ahrens 1999a])
The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the t:i. T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron),
cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ricl):
Eq. 1 A standard heat resistance model for this system is:
* K-State Reactor Safety Analysis Report 4-18                              Original (12/04)


====2.1.5 Bases====
REACTOR DESCRIPTION
Safety Analysis Report, Section 3.5.1 (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged.
* T =T c1      *
NUREG 1282 identifies the safety limit for the high-hydride (ZrH11) 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).
                                        +q"[_!_+ ln(X) +~+~1 h
This stress will remain below the yield strength of the stainless steel cladding with fuel temperatures below 1, l 50&deg;C. A change in yield strength occurs for stainless steel cladding temperatures of 500&deg;C, but there is no scenario for fuel cladding to achieve 500&deg;C while submerged; consequently the safety limit during reactor operations is l,150&deg;C. K-State Reactor TS-8 Original (9fG+4/1744)
ro k              rh      2kf Eq. 2 c              I g and heat flux is calculated directly as:
* *
q"= Ul1T =                   Tmax - ~                                      (2) 1 r0 ln(r0 Ir;)           r      r *
* TECHNICAL SPECIFICATIONS Therefore, the important process variable for a TRIGA reactor is the fuel element temperature.
                                            -+                    + 0- + 0-h        . kc          ljhg 2k1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k.r is the fuel thermal conductivity. The gap conductivity of 2840 W m* 2 K- 1 (500 Btu h" 1 ft -2 &deg;F" 1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context.
This parameter is well suited as a single specification, and it is readily measured.
Parameters are cross-referenced to source in Table 4.6 .
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.
T abl e 4 6 Th ermo d1ynam1c    . Va1ues Parameter          Symbol          Value                Units            Reference Fuel conductivity            kr            18              Wm-lK"l           Table 13.3 14.9        W m* 1 K- 1 (300 K)    Table 13.3 16.6        W m* 1 K- 1 ( 400 K)    Table 13.3 Clad conductivity            kg 19.8       W m* 1 K- 1 (600 K)       Table 13.3 Gap resistance                ha          2840              wm-2 K- 1        AooendixA Clad outer radius            ro      0.018161                    M            Table 13.1 Fuel outer radius            fj      0.018669                    M            Table 13.1 Active fuel length            Lr          0.381                    M            Table 13.1 No. fuel elements             N            83                  NIA              Chap 13 Axial peaking factor        APF            nl2                  NIA            Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that:
Fuel temperature acting on these gases controls fuel element internal pressure.
                                          ~ = 5.1858E-04 m K 2
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-117-833.
2kf                            w Gap resistance has been experimentally determined as indicated, so that:
Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients.
                                            ~=3.6196E-04 m K 2
Operating with maximum long-term, steady state fuel temperature of750&deg;C does not have significant time-and temperature-dependent fuel growth . K-State Reactor TS-9 Original (flfG-74/1744)
rh
* *
                                              '  g W
* TECHNICAL SPECIFICATIONS
* K-State Reactor Safety Analysis Report 4-19                                  Original (12104)


===2.2 Limiting===
CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127&deg;C applies so that:
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.  
r r 0 ln..CC.
r                  m'K
                                                - - ' =3.103e-5--
k,                        w Tabl e 4.7' Cl add"mg H eat Trans fler c oe ffitc1ent Temp (&deg;K)              Temp (&deg;C)                 m2 K w- 1 300                      27                  3.457e-5 400                    127                  3.103e-5 600                    327.                 2.601e-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small.
The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.
Convection Heat Transfer Coefficient TRENDLINE:
y = 0.0326x + 16985 85000                                                                            R2 = 0.9976 f    75000 E
            ~
i!!. 65000 1:
lE    55000 0"'
0
            ~
I-c:
45000
            'i
:i::
35000 25000 500      700        900            1100        1300        1500  1700              1900 Power Level (KW)
Figure 4.10, Convection Hear Transfer Coefficient versus Power Level 1
h      0.0326P(watts) + 16985
* K-State Reactor Safety Analysis Report 4-20                                    Original (12/04)


====2.2.3 Specifications====
REACTOR DESCRIPTION
* Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as:
1 T,  = T, + 0.423P[                      + 3.103e-5 + 3.620e-4 + 5.186e-4]            (10)
          <                  0.0326P + 16985
: c. Steady-State Mode of Operation Centerline temperature calculations were performed on a "reference core" using the model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation. A core with more than. 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 20&deg;C and 100&deg;C was assumed with the following results:
Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel        Fuel/Gap      Gap/Clad Clad/Water Bulk Water &deg;C Centerline &deg;C Interface &deg;C Interface &deg;C Interface &deg;C 503.2         229.0          37.7          21.2           20.0 582.0          307.8        116.4          100.0        100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 &deg;C at power levels up to 1900 kW with pool temperature at 27 &deg;C (300 K), and 1700 kW with pool temperatures at I 00 &deg;C .
* K-State Reactor Safety Analysis Report 4-21                              Original (12/04)


I (I) I Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation I 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I Reduce power to less than A.I IMMEDIATE 1,250 kW (th) A. Steady state power level OR exceeds 1,250 kW (th) A.2. Establish REACTOR A.2. IMMEDIATE SHUTDOWN condition
CHAPTER4 Hot Fuel-Rod Centerline Temperature at Power (Temperature 8evation over Pool Water Tern perature)
I- u  -
* 300 K - - 6 0 0 K I 100-EE!lRREEl33EIREEl33IEIREEl33IEIREEl33IEI!EfftEEl33i333:REE!33~IRRE~IE!RE~~Im
              ~
e-600
::I
              ~ 500 Cll Cl.
              ~400 I-
              ~ 300
              ~ j~~~~~~~~~~~~~~~~~~~~~~~~~~~~i
              ~ 100 0
100    300    500    700      900        1100    1300  1500    1700    1900 Reactor Power (kW)
Figure 4.11, Hot Fuel-Rod Centerline Temperature
* For the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD 3.3 patch 04 (Feldman 2008). A snapshot of the model is presented in Figure 4.12. It has two time-dependent volumes, enforcing the pressure boundary conditions, and two pipes, simulating the cold and hot channel connected via a single junction component of RELAP. Heat is added to the fluid by incorporating the heat structure component (simulating a fuel element) of RELAP with an appropriate axial power profile and power level. In this analysis, the power level for the B ring is at 24 kW (corresponding to an 85-element core with a ring-to-average peaking factor of 1.63).
This power level is applied to the heat structure within the single channel. The model assumes an operating pressure of 143 kPa, and an operating temperature of 322.15 K (49. l 5&deg;C).
The version of the RELAP code licensed to KSU uses PG-CHF correlation which is a state of the art best estimate CHF correlation developed by Nuclear Research institute of Rez in the Czech Republic. It is based on data in the Czech Republic data bank from 173 different sets of tube data, 23 sets of annular data, and 153 sets of rod bundle data. There are four forms of the PG-CHF correlation 'Basic', 'Flux', 'Geometry', and 'Power'. For the rod bundle it is applicable in the pressure range of0.28 MPato 18.73 MPa, for a mass flux of34.l to 7478 kg/s-m2, for 0.4-7.0 m length and for a diameter of 0.00241 to 0.07813 m. TRIGA has an operating pressure of 0.143 MPa and fuel rod length of 0.381 m, thus the operating conditions fall outside the range of the applicability of the PG-CHF correlation, and a different correlation is required to assess the
  . departure from nucleate boiling ratio (DNBR ratio). One such correlation which is applicable for the low pressure range observed in TRIGA reactor facility is the Bernath correlation. The functional form of the Bernath correlation can be presented in the following equations .
* K-State Reactor Safety Analysis Report 4-22                                Original (12/04)


====2.2.5 Bases====
REACTOR DESCRIPTION
Analysis in Chapter 4 demonstrates.
* I  Outlet i                      ~    .
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.
Hot-leg Cold-Leg
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.
*                            +                          I connector Figure 4.12 - RELAP single channel model used in CHF analysis (8)
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.
* K-State Reactor Safety Analysis Report 4-23                          Original (12/04)
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 (9/G+4/1744)
* *
* TECHNICAL SPECIFICATIONS
: 3. Limiting Conditions for Operation (LCO) 3.1 Core Reactivity


====3.1.1 Applicability====
CHAPTER 4
                              ~ = D4~.6 , if Dh :<S; O.lft h
                              ~ = _!Q_ + 90, if Dh ;::: 0.1.ft D,,
hBo =film coefficient at CHF D,, =hydraulic dia.meter (ft) v =coolant velocity (ft Is)
TwBo =wall temperature at burnout (&deg; C)
DH =heated diameter (ft)
The RELAP simulations were performed for the hot channel, i.e., a channel with a radial peaking factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain the pressure, temperature, and velocity distribution at different axial locations. With these calculations and the functional form of the Bernath correlation, the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA, the A-ring is occupied by the central thimble, not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard.
The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and Figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel.
* K-State Reactor Safety Analysis Report 4-24                              Original (12/04)


These specifications are required prior to entering STEADY STATE MODE or PULSING MODE in OPERA TING 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) I. 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:
REACTOR DESCRIPTION
(2) 1. The highest worth control rod is fully withdrawn
* 5000-------------------------                                A A
: 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. I ENSURE REACTOR A. I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meetLCO operations K-State Reactor TS-11 Original (9fG-74/1744)  
Bernatl1-CHF PG-CHF 4000                          .. _., A
* *
                                                                                  *
* TECHNICAL SPECIFICATIONS B.1.a ENSURE control rods B.1 IMMEDIATE fully inserted AND B.1.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.1.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 meetLCO 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.
* Heat flux A
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.
                                                                  .. Jo. ..
Since experiments with negative reactivity will increase available reactivity if they are removed during operation, they are not credited in determining excess reactivity.
A .I. A 0.
Analysis (Chapter 13) shows fuel temperature will not exceed l,150&deg;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&deg;C, while a $1.00 reactivity insertion from a worst-case steady state operation at I 07 kW leads to a maximum fuel temperature of 869&deg;C, well below the safety limit. The limiting SlillTDOWN 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.
          .,.,---                                                                        A A     A A A :A  A
K-State Reactor TS-12 Original (MH4/1744)
* 3000
* *
                ~
* TECHNICAL SPECIFICATIONS
            -~-
            ...:.o::
u...
:c 2000 u
1000
:r .;. *
                        &~00---0~.o-s---o-_1-o~~-o.*1-s~~o-.2-o~~o-_-2-s~~o~.3-o~--o-__*35-....
Heated Length {m)
Figure 4.13 - CHF versus heated length
                            + + Bernath-CHF
                              +
: u.
* PG-CHF 10 -**
B*
                                                        ~
o! -                                                               .**
2                .                                  *  ..
8.oo      0.05      O.lG    O.lS          0.20        0_25        0.30        0.35 Heated length {n1*)
Figure 4.14 - DNBR versus heated length
* K-State Reactor Safety Analysis Report 4-25                                            Original (12/04)


===3.2 PULSED===
CHAPTER4
MODE Operations
: d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code TASCOT for transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. (9) and (10), with the peak fuel temperature set to 746 &deg;C, i.e., a temperature rise of 719 &deg;C above 27 &deg;C ambient temperature. The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1 % ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation. In the TASCOT calculations, thermal conductivity was set to 0.18 W cm* 1 K- 1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm* 1 K- 1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994)
(9) l The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955)
              .-
* k f0.79
* cP.f0.45 * ~ 0.51 fl,              * (T,.. -      )0.99 hh - 0.00122                                                  0.75                T.at      ,  (10)
[ cr o.s * &#xb5; 0.29
* p 0.24 * (v _ v )
* To.1s f          g            g      v      sat in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 &deg;C),
and Tb the coolant ambient temperature (27&deg;C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases.
The overall heat transfer coefficient U varies. negligibly for ambient temperatures from 20 to 60 &deg;C, and has the value 0.21 W cm* 1 K- 1 at Tb= 27 &deg;C.
Figure 4.15 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.10 lists temperatures and heat fluxes as function of time after a 2.1 % ($3 .00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m* 1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500&deg;C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse .
* K-State Reactor Safety Analysis Report 4-26                                    Original (12/04)


====3.2.1 Applicability====
REACTOR DESCRIPTION
* 1000 Os 800 1
()
0...__,
2
          <!)
600 L
:J 4
        +'
nl L                                                                                                    8
          <!)
Q_    400 E
          <!)                                                                                                  16 I-200                                                                                          32 64s 0 '---'---'---'--_._----'-----'----'---'----'___J'--.__-'---'----'--'-----'----'----'---'----'___JL-....1 0.0    0.20 0.40 0.60 0.80                    1.0    1.2       1.4      1.6      1.8      2.0    2.2 Radius (cm)
* Figure 4. 15, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse.
* K-State Reactor Safety Analysis Report 4-27                                            Original (12/04)


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.  
CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{0 C) Coolant Ambient Temperature.
Q"                                Fuel outside      Clad surface Time (s)              (W m-2)            CHFR              Temp. (oC)        Temp. (&deg;C) 0                                                          953 1                3.57 x10 5            4.2                781                224 2                7.34 xl0 5            2.0                683                432 4                        5            1.7                574                498 8.52 x10 8                7.54 xl0 5            2.0                461                443 16                5.71 xl0 5            2.6                344                342 32                3.46 x10 5            4.3                224                218 64                        5            14.4                100                84 1.04 xl0 4.6      Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by
*                                                                                                (11) in which Po and 130 are the density and volumetric expansion coefficient at core inlet conditions (27&deg;C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm2 s- 1, l':iT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by (12) in which mis the coolant mass flow rate (kg s- 1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m2, and D1i is the hydraulic diameter, namely, 0.02127 m. The friction factor/for laminar flow through the annular area is given by 100 Re- 1 (Shah & London 1978), in which the Reynolds number is given by D,,rh I A&#xb5; 0 in which &#xb5;o is the dynamic viscosity at core inlet conditions.
Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5),
and the entrance pressure loss would be expected to be negligible. The exit contraction loss is given by (13)
* K-State Reactor Safety Analysis Report 4-28                              Original (12/04)


====3.2.3 Specification====
REACTOR DESCRIPTION
* The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which
[~]
2 2
_        _ [ 3
* R sin60&deg; cos60&deg; ]-
K= A c
11
* R2        -  0.171,                    (14) where R is the radius of the opening in the upper grid plate. Equations (12) through (14), solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4. I 1.
Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRIGA Reactor During Steady-State Operations.
P (kWt)                    m (kg s- 1)                    !).T (&deg;C) 50                        0.047                            3.1 100                        0.061                            4.7 200                        0.077                            7.5 300                        0.090                            9.6 400                        0.100                            11.5 500                        0.108                            13.3 750                        0.125                            17.2 1000                      0.139                            20.6 1250                      0.150                            23.8
: 4. 7    Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500&deg;C; consequently, limits on fuel temperature change for cladding temperatures greater than 500&deg;C. A maximum temperature of l 150&deg;C (with clad< 500&deg;C) and 950&deg;C (with clad> 500&deg;C) for U-ZrH (H/Zr1.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.
4.8      Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature. The operating parameters established for the KSU TRI GA reactor are:
* K-State Reactor Safety Analysis Report 4-29                                Original (12/04)


(1) 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.I 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 IMMEDlA TE 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 I3.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis.
CHAPTER 4
K-State Reactor TS-13 Original (9fm4/1744)
* Steady-state power level
* *
* Fuel temperature measured by thermocouple during pulsing operations
* TECHNICAL SPECIFICATIONS
* Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100&deg;C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.
Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.
4.8.3 Safety Margins For 1,250 kWth steady-state operations, the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49&deg;C. The proposed Technical Specifications limit of 44&deg;C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation. Limiting pool inlet water temperature to no greater than 44&deg;C (or 37&deg;C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling.
Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746&deg;C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869&deg;C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15%
margin to the fuel temperature safety limit for cladding temperatures below 500&deg;C.
Analysis shows that cladding temperatures will remain below 500&deg;C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve 500&deg;C (invoking a 950&deg;C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.
Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500&deg;C, but the heat transfer simultaneously cools the fuel to much less than 950&deg;C .
* K-State Reactor Safety Analysis Report 4-30                                Original (12/04)


===3.3 MEASURING===
REACTOR DESCRIPTION
* If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 500&deg;C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500&deg;C.
4.9      Bibliography "TASCOT: A 2-D, Transient and Steady State Conduction Code for Analyhsis ofa TRIGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C.,
  "Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C.,
  ''A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,
Kansas State University TRIGA MkII Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J.G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K.,
and N. Zuber, Theory and Design ofModern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F.,
  "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al.,
MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998.
NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987 .
* K-State Reactor Safety Analysis Report 4-31                              Original (12/04)


CHANNELS 3.3.1 Applicability This specification applies to the reactor MEASURING CHANNELS during STEADY STATE MODE and PULSE MODE operations.  
CHAPTER 4 "Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K.,
and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRIGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T.
  "Safety Analysis Report, TRIGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991.
* K-State Reactor Safety Analysis Report 4-32                          Original (12/04)
* Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the Univer~ity of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).
* The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A. l. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.
Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100&deg;F water, showed burnout fluxes of 0.9 to 2.0 Mbtu fr 2 hr- 1 for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft- 2 hr- 1* It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.
A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A.1 and 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise.
The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760&deg;C clad temperature for a peak initial fuel temperature of 1000&deg;C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure'4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor                                4.A-1                                Original (9/02)
Safety Analysis Report


====3.3.2 Objective====
CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470&deg;C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400&deg;C to 500&deg;C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965].
I. 2 I. 1 1.0 0.9
* a.a~~~~~~~~~~~~~~~~~~~~~~~
0      0.1    0.2      0.3     0.4 RADIUS (IN.)
0.5      0.6    0.7 Figure 4A.1. Representative Radial Variation of Power Within the TRIGA Fuel Rod I.I 0.8 1.0 0.9
                  ;::;  0.8 L;:"
0.7 0.6 0.5 0              2              4              6      7    8 AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.)
Figure 4A.2, Representative Axial Variation of Power Within the TRIGA Fuel Rod .
* K-State Reactor Safety Analysis Report 4.A-2                               Original (9/02)


The objective is to require that sufficient information is available to the operator to ensure safe operation of the reactor 3.3.3 Specifications (I) 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 I: MINIMUM MEASURING CHANNEL COMPLEMENT MEASURING CHANNEL Minimum Number Operable STEADY STATE PULSE MODE MODE Reactor power leveJl 1 l 2 Primary Pool Water Temperature I Reactor Bay Differential Pressure I Fuel Temperature I 22 foot Area radiation monitor I 0 or 12 foot Area monitor I Continuous air radiation monitor1 2 l I EXHAUST PLENUM radiation monitor1 2 l I NOTE[!]: One "Startup Channel" required to have range that indicates
REACTOR DESCRIPTION
<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.I.I Restore channel to operation A. I.I IMMEDIATE not OPERATING (min 2 OR for STEADY STATE, 1 A.1.2 ENSURE reactor is A.1.2 IMMEDIATE PULSE MODE) SHUTDOWN K-State Reactor TS-14 Original (MH4/1744)
                                                \
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Establish REACTOR A.2 High voltage to reactor SHUTDOWN condition 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.l Restore channel to operation A. l IMMEDIATE reactor bay differential OR pressure or fuel temperature CHANNEL B.2 ENSURE reactor is A.2 IMMEDIATE not operable SHUTDOWN C.l Restore MEASURING C.l IMMEDIATE CHANNEL OR C.2 ENSURE reactor is shutdown C.2 IMMEDIATE C. 22 foot Area radiation OR monitor is not C.3 ENSURE personnel are not C.3 IMMEDIATE OPERATING on the 22 foot level
                                                  \              CURVE BASED ON
* OR C.4 ENSURE personnel on 22 C.4 IMMEDIATE foot level are using portable survey meters to monitor dose rates D.l Restore MEASURING D.l 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 K-State Reactor TS-15 Original (9JG+.4/1744)
                                                    \
* 
                                                      \          DATA OF ELLIO/.
* *
TW-TSAT (*f)
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME E.l Restore MEASURING E.l 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.l Restore MEASURING F.l 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.l Do not perform a reactor G.l IMMEDIATE G. The neutron count rate on startup the startup channel is not OR greater than the minimum G.2 Perform a neutron-source G.2 IMMEDIATE detector sensitivity check on the startup channel 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.
Figure 4A.3, Subcooled Boiling Heat Transfer for Water.
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.
1800 1700
Therefore, if operating voltage falls below the minimum value the power level channel is inoperable.
[LAPSED TIHE FROH END OF PULSE 1600
K-State Reactor TS-16 Original (8fG+.4/1744) 
                  ~
* *
i= 1500
* 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.
1400 1300 1200 0.1    0.2   0.3           0.4          o.s      o.6 0. 7 0.8 0
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.
RADIUS (IN.)
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.
Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse.
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.
* K-State Reactor Safety Analysis Report 4.A-3                                        Original (9/02)
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 11 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 IO 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. K-State Reactor TS-17 Original (9fG-7.4/17+4)
*
* TECHNICAL SPECIFICATIONS


===3.4 Safety===
CHAPTER 4 APPENDIX A 106.---,..--.-....-.r-r--..----.--.-.-..--r-....--r--.-.--.-----.----.-~--.--.,...-..-.-,
Channel and Control Rod Operability
ONSET OF _!_l          PEAK HEAT FLUX NUCLEATE 801 LI NG N
                        ~
                        ...."'    105 ONSET or STABLE
                        ...c I Fii.Joi llOlllNG
                        ....x
                        ....u    10 4
                        ...c 103 0.001              0.01                    0.1                  1.0                  10                      100 ELAPSED TIME FROM ENO OF PllLSE (SEC)
Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse.
*            '0. 000    ,..........,.-..,......,........,.--,--..-.,......,.....,....-..--.;...,r--i,..-,-..-......,--r-r"T"T--.--..-.,.-,-,
                \000 100 - - - - -
CLAO OUTER SURFACE TEMP j
io          l.___L._..L..L.1...l.-..L-L-.LJ~-L-__J--l-'-L---l--'--'-.........                          ,0--:-._.__..............~,oo 0.001                        0.01                      0. I                  1.0 ELAPSED TIME        FROM    END  OF  PULSE (SEC)
Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element.
* K-State Reactor Safety Analysis Report 4.A-4                                                      Original (9/02)


====3.4.1 Applicability====
REACTOR DESCRIPTION
* The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.
This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected. Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962]
which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000&deg;C peak fuel temperature case.
For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural
* convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960]
which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate.
The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft- 2 hr- 1 &deg;F- 1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft- 2 hr- 1 &deg;F- 1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel.
The surface heat flux at the mid plane of the element is shown in Figure 4A. l 0 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft- 2 hr- 1 &deg;F- 1 according to the more conservative Bernath correlation at a local water temperature approaching saturation.
K-State Reactor                                    4.A-5                                  Original (9/02)
Safety Analysis Report


This specification applies to the reactor MEASURING Channels during STEADY STATE MODE and PULSE MODE operations.  
CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000&deg;C is conservatively estimated to be 470&deg;C.
As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470&deg;C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000&deg;C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is 1150&deg;C would not produce a stress in the clad in excess of its ultimate strength. TRI GA fuel with a hydrogen to zfrconium ratio of at least 1.65 has been pulsed to temperatures of about l l 50&deg;C without damage to the clad [Dee et al. 1966].
7
        ......                          ELAPSED TIME FROM IN I-u..                            *END OF PULSE
* 0.2.47 SEC
        'I        6 a:
:::c:
::i I-al IJ\
I          5 0
x
          ...J u.. 4 t-c:t
:c 3
7            8              9            10            11            12.          13 DISTANCE FROM BOTTOM OF FUEL (IN.)
Figure 4A. 7, Surface Heat Flux Distribution for Standard Non-Gapped {hgap=
500 Btu/h ft 2 &deg;F) Fuel Element After a Pulse .
* K-State Reactor Safety Analysis Report 4.A-6                                  Original (9/02)


====3.4.2 Objective====
REACTOR DESCRIPTION 8
7
              .;,        6
:i:
                ~
era 5'
            "'0
                                  -----.'A~CCTUAL        HEAT FLUX
                ....      It
:i:
3            ELAPSED TIHE FROH ENO OF PULSE IS 0.311t SEC 2
7    8          9          10      11      12        13      15 DISTANCE FROH BOTTOM OF FUEL (IN.)
Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 375 Btu/h ft2 &deg;F) After a Pulse.
B 7
N I-
                "-    . 6
                -a:
::r::
                  =>
I-m        5 U\
                . 0                                      ELAPSED TIME FROM ENO OF PULSE IS 0.440'SEC
                    )C      4 x
                *~
                    =>
I-er:    3
:c 1
1 7      8          9        10        11      12        13 14    15 DISTANCE FROM BOTTOM OF FUEL (IN.)
Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap=
250 Btu/h ft 2 &deg;F ) After a Pulse.
* K-State Reactor Safety Analysis Report 4.A-7                          Original (9/02) j


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 (I) The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLE (2) CONTROL RODS (STANDARD) are capable of 90% of full reactivity insertion from the. fully withdrawn position in less than 1 sec. ill A minimum of three CONTROL RODS must be OPERABLE.
CHAPTER 4 APPENDIX A EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR-FT 2 -&deg;F 500
Ino12erable CONTROL RODS must be fully inserted.
          ...J i..
TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Safety System Channel Minimum Function Reauired OPERATING Mode Number STEADY PULSE or Interlock Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar I Scram YES YES CONTROL ROD Prevent withdrawal of standard (STANDARD) position 1 rods in the PULSE MODE NA 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 Al. IMMEDIATE A. Any required SAFETY to operation SYSTEM CHANNEL or OR interlock function is not OPERABLE A2. IMMEDIATE A.2 ENSURE reactor is SHUTDOWN K-State Reactor TS-18 Original (WG+4/1744)
w
-*--{ Formatted Table 
:c w
* *
u i..
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME B.l ENSURE inoRerable control Bl. IMMEDIATE rod is fully inserted B. A control rod is not OR OPERABLE.
          !.,, 101+
B2. IMMEDIATE B.J2 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.
FLOW VELOCITY a I FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME= ZERO) 1031.-~~.....J...~~~l--~l..--'-....J-~----"'--~~...J-~..;.i....--'-...J 0.01                            0.1                              1.0 ELAPSED TIHE FROM END OF PULSE (SEC}
The power level scram is not credited 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).
Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse.
The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. The CONTROL ROD (ST AND ARD) 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.
* K-State Reactor Safety Analysis Report 4.A-8                                Original (9/02)
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.
InoRerable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. ORerating with a fully-inserted control rod may cause Rower Reaking to shift. however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR ChaRter 13 is still bounding, and the reduction in maximum core Rower by having an inoRerable control rod fully inserted means that the highest temRerature in any fuel element with a inserted inoperable control rod will be lower than the highest temRerature in the B-ring with all rods withdrawn.
Therefore the reactor can be safely ORerated with an inoperable control rod Rrovided that the rod is fully inserted into the core. K-State Reactor TS-19 Original (91-G+4/1744)
* *
* TECHNICAL SPECIFICATIONS


===3.5 Gaseous===
REACTOR DESCRIPTION
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 (2) Releases of Ar-41 from the reactor bay exhaust plenum to an unrestricted environment SHALL NOT exceed 30 Ci per year. 3.5.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I ENSURE reactor is A. I IMMEDIATE SHUTDOWN OR A.2.a Do not OPERA TE 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.
* Bibliography "A Theory ofLocal Boiling Burnout and Its Application to Existing Data, " Heat Transfer -
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 (WG-74/1744) 
Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No.
* *
20.Bernath, L.,
* 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-1 (2.0lxl0-5 &#xb5;Ci mJ-1), well below the 1 OCFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10*3 &#xb5;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.
Research in Improved TRIGA Reactor Pe1formance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al.,
Thirty days for a confinement and ventilation system outage is selected as a reasonable interval to allow major repairs and work to be accomplished, ifrequired.
Characteristics of Large Reactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al.
During this interval, experiment activities that might cause airborne radionuclide levels to be elevated are prohibited.
Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al.,
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.
A Study ofthe Mechanism of Boiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E.,
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. 90 Sr activity available for release from fuel rods previously used at other facilities is estimated to be at most about 4 times the ALL 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.
Thermal Conductance ofMetallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis ofa Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J.
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 (91G+.4/1744)
Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172.
* *
Graff, W.J.
* TECHNICAL SPECIFICATIONS
Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H ..
MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998.
Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H.
Choi, "Quarterly Technical Report SPERT Project, April, May, June, 1964, "ISO 17030. Spano, A.
H.,
    "The Effect ofSubcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis, " Technical report, Argonne National Laboratory, 2008, E.E. Feldman .
K-State Reactor                               4.A-9                              Original (9/02)
Safety Analysis Report I


===3.6 Limitations===
CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure, system models, and solution methods.
  "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution."
Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong .
* K-State Reactor Safety Analysis Report 4.A-10                                Original (9/02)


on Experiments
REACTOR DESCRIPTION
*  "Onset ofStable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al.,
UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991.
  "Hydrodynamic Aspects ofBoiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959.
Zuber, W .
* K-State Reactor Safety Analysis Report 4.A-11                                Original (9/02)
* Tpool            Pt,16[11 Appendix B Water Properties at Nominal Operating Conditions Data for 16 Feet of Water over tile Core ht,16[ 11    hg,16111  P 9 ,15111 Tsat,16111  q"sat, 15!3!  q"sub[41 oc              kg m-  3            kJ    kg- 1    kJ  kg- 1  kg m-    3    oc        wm-    2    Wm-2 15    999.21    950.00      47.79    465.10          2692.64    0.85    110.89      1553.842      7239.19 20    998.32    950.01      47.74    465.05          2692.63      0.85    110.88      1552.078      6931.74 25    997.16    950.02      47.69    465.01          2692.59      0.85    110.87      1549.496      6622.60 30    995.75    950.03      47.62    464.95          2692.59      0.85    110.86      1547.118      6311.82 35    994.12    950.04      47.54    464.89          2692.57      0.85    110.84      1543.981      5999.91 40    992.29    950.06      47.46    464.81          2692.54      0.85    110.83      1540.446      5688.24 45    990.27    950.07      47.36    464.73          2692.51      0.85    110.81      1536.512      5376.29 50    988.07    950.09      47.25    464.64          2692.48      0.85    110.78      1532.205      5064.36 55    985.70    950.11      47.14    464.54          2692.45      0.85    110.76      1527.561      4753.90 60    983.18    950.12      47.02    464.44          2692.41      0.85    110.74      1522.575      4444.85 65    980.50    950.14      46.89    464.33          2692.37      0.85    110.71      1517.255      4136.85 70    977.69    950.17      46.76    464.21          2692.33      0.84    110.68      1511.666      3830.73 75    974.74    950.19      46.62    464.09          2692.29      0.84    110.65      1505.778      3526.89 80    971.66    950.21      46.47    463.96          2692.24      0.84    110.62      1499.613      3225.47 85    968.45    950.23      46.32    463.83          2692.19      0.84    110.59      1493.199      2926.81 90    965.12    950.26      46.16    463.69          2692.15      0.84    110.56      1486.527      2631.05 95    961.68    950.29      45.99    463.55          2692.09      0.84    110.53      1479.626      2338.47 97    960.27    950.30      45.92    463.49          2692.07      0.84    110.51      1472.944      2216.11 99    958.84    950.31      45.86    463.43          2692.05      0.84    110.50      1466.058      2095.18 Data for 13 Feet of Water over tile Core Tpool                                  ht,13[ 11    hg,131 11 Pg,13111  Tsat,13111  q"sa1,13!3l  q"subl41 oc                                    kJ kg-    1    kJ  kg-1  kg m-    3    oc        Wm-2          wm- 2 15    999.21    951.43      38.83  457.21          2689.85      0.80    109.03      1513.00      6964.74 20    998.32    951.43      38.79  457.18          2689.84      0.80    109.02    1511.32      6857.12 25    997.16    951.44      38.75  457.13          2689.82      0.80    109.01    1509.15      6543.62 30    995.75    951.45      38.69  457.09          2689.80      0.80    109.00    1505.85      6229.30 35    994.12    951.46      38.63    457.03        2689.78      0.80    108.99    1503.13      5913.58 40    992.29    951.47      38.56    456.96        2689.76      0.80    108.97    1500.16      5597.21 45    990.27    951.49      38.48    456.89        2689.74      0.80    108.96    1496.38      5281.90 50    988.07    951.50      38.39    456.82        2689.71      0.80    108.94    1492.25      4966.66 55    985.70    951.51      38.30    456.73        2689.68      0.80    108.92    1487.78      4652.39 60    983.18    951.53      38.20    456.64        2689.65      0.80    108.90    1482.99      4339.60 65    980.50    951.55      38.10    456.55        2689.61      0.80    108.87    1477.90      4027.94 70    977.69    951.57      37.99    456.45        2689.58      0.80    108.85    1472.52      3718.83 75    974.74    951.58      37.88    456.35        2689.54      0.80    108.83    1466.86      3412.07 80    971.66    951.60      37.76    456.24        2689.50      0.80    108.80    1460.95      3107.29 85    968.45    961.62      37.63    456.12        2689.46      0.80    108.77    1458.59      2812.63 90    965.12    951.64      37.50    456.01        2689.42      0.79    108.75    1448.37      2506.90 95    961.68    951.67      37.37    455.89        2689.38      0.79    108.72    1441.74      2211.19 97    960.27    951.68      37.31    455.84        2689.36      0.79    108.70    1435.27      2087.92 108.69    1428.60      1966.16 99    958.84    951.68      37.26    455.78        2689.34      0.79 K-State Reactor                                  4.B-1                                Revised 05/01/17 Safety Analysis Report


====3.6.1 Applicability====
REACTOR DESCRIPTION
 
* T oc 15 20 Common Data Patm kPa 99.83 99.83 Cp[1]
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 (1) If all fuel elements are stainless steel clad, the reactivity worth of any individual 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. (3) 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.l 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 K-State Reactor TS-22 Original (9m+-4/1744)
kJ kg-1 k- 1 4.23080 4.23080 CJ Nm-1 0.07149 0.07120 25        99.83      4.23080      0.07083 30        99.83      4.23080      0.07039 35        99.83      4.23070      0.06989 40          99.83      4.23070      0.06932 45          99.83      4.23070      0.06869 50        99.83      4.23070      0.06800 55        99.83      4.23060      0.06727 60          99.83      4.23060      0.06649 65          99.83        4.23060      0.06566 70        99.83        4.23050      0.06480 75        99.83        4.23050      0.06390 80        99.83        4.23050      0.06297 85        99.83        4.23040      0.06201 90        99.83        4.23040      0.06102 95        99.83        4.23040      0.06001 97        99.83        4.23030      0.05898 99        99.83        4.23030      0.05793 NOTE[1}: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa =Heigth(ft) *12(inlft) *0.0254(meterslin) *Density(kglm 3) *9.8066511000 NOTE[3}: qSAT"
* *
                  =0.I49*pg 0.5 . ( hg,sat-hf,sat.
* TECHNICAL SPECIFICATIONS
                                                ) (
 
                                                    *g*<J* {Pt-Pg })Y.
====3.6.5 Bases====
NOTE[4']* " - " . 1 0
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.
* qsub -qSAT
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 . K-State Reactor TS-23 Original (9!G+4/1744)
(
* *
                          +
* l*(P1J .
Pg Y. cp.f *(TsAT -Tsub) hg,sat - h /,sat J
NOTE:[5}:
  <J = 1.000E-11*T 4 +7.370E-09
* T 3 -1.969E-06
* T 2 + 4.709E-06
* T + 7.1833E- 02 K-State Reactor                              4.A-13                              Original (9/02)
Safety Analysis Report
* Table of Contents TECHNICAL SPECIFICATIONS I. DEFINITIONS ................................................................................................................. TS-I
: 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Element Temperature Safety Limit ...................................................................... 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 Limiting Safety System Settings ................................................................................ TS-I 0 2.2.1. Applicability ...................................................................................................... TS- I 0 2.2.3. Objective ............................................................................................................ TS-10 2.2.4. Specification ....................................................................................................... TS- I 0 2.2.5. Actions ............................................................................................................... TS- I 0 2.2.6. Basis .................................................................................................................. TS-10
: 3. LIMITING CONDITIONS FOR OPERATIONS ......................................................... TS-11 3.1 CORE REACTIVITY ................................................................................................ TS-II 3.1.1. Applicability ...................................................................................................... TS-11 3.1.3. Objective ............................................................................................................ TS-II 3.1.4. Specification ....................................................................................................... TS-11 3.1.5. Actions ............................................................................................................... TS-12 3.1.6. Basis .................................................................................................................. TS-13 3.2 PULSED MODE OPERATIONS .............................................................................. TS-13 3.2.1. Applicability ...................................................................................................... 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 MEASURING CHANNELS ...................................................................................... 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. Applicability ...................................................................................................... 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 GASEOUS EFFLUENT CONTROL ......................................................................... TS-20 3.5.1. Applicability ...................................................................................................... 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 LIMITATIONS ON EXPERIMENTS .......................................................................... TS-22 3.6.1. Applicability ...................................................................................................... TS-22 3.6.3. Objective ............................................................................................................ TS-22 K-State Reactor                                                TS-1                                        Original (9tG+-4/1744)
L_ _____
* TECHNICAL SPECIFICATIONS 3.6.4. Specification ....................................................................................................... TS-22 3.6.5. Actions ............................................................................................................... TS-22 3.6.6. Basis .................................................................................................................. TS-23 3.7 FUEL INTEGRITY ................................................................................................... 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 WATER ......................................................................................... 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 Maintenance 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 REACTIVITY ................................................................................................ 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 MEASURING CHANNELS ...................................................................................... 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 LIMITATIONS ON EXPERIMENTS ........................................................................ TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis .................................................................................................................. TS-34
: 4. 7 FUEL 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 WATER ....................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor                                                    TS-2                                        Original (91G+4/1744)
* TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36 4.8.3. Basis .................................................................................................................. TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................................................................... TS-37 4.10.3. Basis ................................................................................................................ TS-37
: 5. DESIGN 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 BUILDING ............................................................................................. 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 EXPERIMENTS ......................................................................................................... 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. ADMINISTRATIVE CONTROLS ................................................................................. TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL.. ........................ TS-44 6.2 REVIEW AND AUDIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS .................................................. TS-47 6.5 EMERGENCY PLAN AND PROCEDURES ........................................................... TS-48 6.6 OPERATOR REQUALIFICATION .......................................................................... TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERATING RECORDS ............................................................................ TS-49 6.11 REPORTING REQUIREMENTS ........................................................... TS-50 K-State Reactor                                                TS-3                                        Original (9,LG+.4/1744)
* 1. DEFINITIONS TECHNICAL SPECIFICATIONS 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 .
* ANNUAL CHANNEL
* 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 12 months, not to exceed 15 months 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 K-State Reactor                                TS-4                        Original (9fG+.4/1744)
* 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 EXPERlMENT      An EXPERlMENT is (I) any apparatus, device, or material placed in the reactor core region (in an EXPERlMENTAL FACILITY associated with the reactor, or in line with a beam ofradiation emanating from the reactor) or (2) any in-core operation designed to measure reactor characteristics.
EXPERlMENTAL 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 Frings, as many as three contiguous fuel-element positions).
IMMEDIATE Without delay, and not exceeding one hour.
NOTE:
IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action INDEPENDENT INDEPENDENT Experiments are those not connected by a mechanical, EXPERlMENT 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 EXPERlMENT is one that may be moved into, out-of or EXPERlMENT near the reactor while the reactor is OPERA TING.
NONSECURED NONSECURED Experiments are these that should not move while the EXPERlMENT reactor is OPERATING, but are held in place with less restraint than a secured experiment.
K-State Reactor                          TS-5                        Original {9fG.74/1744)
* OPERABLE TECHNICAL SPECIFICATIONS 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 OPERA TING 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, reactor power may be increased on a period of much less than l second by motion of the transient control rod.
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 m 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 ofless than 60 lb.
K-State Reactor                          TS-6
* SECURED EXPERIMENT TECHNICAL SPECIFICATIONS 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 applicable titles of sections in individual Technical Specifications K-State Reactor                          TS-7                        Original (fhlG+.4/1744)
* 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 l 150&deg;C.
(2)    Steady state fuel temperature shall not exceed 750&deg;C.
2.1.4    Actions
* CONDITION A    Stainless steel clad, high-hydride fuel element temperature exceeds l 150&deg;C.
REQUIRED ACTION A.I Establish SHUTDOWN condition AND COMPLETION TIME A.I IMMEDIATE OR Fuel temperature exceeds 75 0&deg;C in steady state        A.2 Report per Section 6.8 conditions                                                    A.2 Within 24 hours 2.1.5    Bases Safety Analysis Report, Section 3.5.1 (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged.
NUREG 1282 identifies the safety limit for the high-hydride (ZrH11) 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, l 50&deg;C. A change in yield strength occurs for stainless steel cladding temperatures of 500&deg;C, but there is no scenario for fuel cladding to achieve 500&deg;C while submerged; consequently the safety limit during reactor operations is l,150&deg;C.
K-State Reactor                                  TS-8                    Original (9fG+4/1744)
* 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-117-833. Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients. Operating with maximum long-term, steady state fuel temperature of750&deg;C does not have significant time- and temperature-dependent fuel growth .
K-State Reactor                                TS-9                        Original (flfG-74/1744)
* 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 I  (I)  I Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation I 2.2.4    Actions CONDITION                        REQUIRED ACTION                  COMPLETION TIME A.I IMMEDIATE A. I Reduce power to less than 1,250 kW (th)
A. Steady state power level OR exceeds 1,250 kW (th)
A.2. Establish REACTOR A.2. IMMEDIATE 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 (9/G+4/1744)
* 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 OPERA TING 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)
I. 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:
: 1. The highest worth control rod is fully withdrawn (2)
: 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. I ENSURE REACTOR            A. I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn                        AND exceeds $4.00 A.2 Configure reactor to      A.2 Prior to continued meetLCO                        operations K-State Reactor                              TS-11                        Original (9fG-74/1744)
* TECHNICAL SPECIFICATIONS B.1.a ENSURE control rods        B.1 IMMEDIATE fully inserted AND B.1.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.1.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 meetLCO 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 l,150&deg;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&deg;C, while a $1.00 reactivity insertion from a worst-case steady state operation at I 07 kW leads to a maximum fuel temperature of 869&deg;C, well below the safety limit.
The limiting SlillTDOWN 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.
K-State Reactor                                TS-12                        Original (MH4/1744)
* 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.I 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 IMMEDlA TE 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 I3.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis.
K-State Reactor                                  TS-13                        Original (9fm4/1744)
* 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 (I)    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 I: MINIMUM MEASURING CHANNEL COMPLEMENT Minimum Number Operable MEASURING CHANNEL STEADY STATE              PULSE MODE MODE Reactor power leveJl 1l                                  2 Primary Pool Water Temperature                            I Reactor Bay Differential Pressure                        I Fuel Temperature                                          I 22 foot Area radiation monitor                            I 0 or 12 foot Area monitor                                I Continuous air radiation monitor12l                      I EXHAUST PLENUM radiation monitor12l                      I NOTE[!]: 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.I Restore channel to operation      A. I.I IMMEDIATE A. I Reactor power channels not OPERATING (min 2                            OR for STEADY STATE, 1 A.1.2 ENSURE reactor is                A.1.2 IMMEDIATE PULSE MODE)
SHUTDOWN K-State Reactor                                TS-14                        Original (MH4/1744)
* CONDITION                  REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less                                      A.2. IMMEDIATE AND than 90% of required operating value A.2.2 Enter REACTOR SECURED mode B. Primary water temperature,  B.l Restore channel to operation    A. l IMMEDIATE reactor bay differential OR pressure or fuel temperature CHANNEL        B.2 ENSURE reactor is                A.2 IMMEDIATE not operable                    SHUTDOWN C.l Restore MEASURING                C.l IMMEDIATE CHANNEL OR C.2 ENSURE reactor is shutdown        C.2 IMMEDIATE OR C. 22 foot Area radiation monitor is not C.3 ENSURE personnel are not          C.3 IMMEDIATE OPERATING on the 22 foot level
* OR C.4 ENSURE personnel on 22 foot level are using portable survey meters to monitor dose rates D.l Restore MEASURING C.4 IMMEDIATE D.l 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 K-State Reactor                            TS-15                      Original (9JG+.4/1744)
* CONDITION                    REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME E.l Restore MEASURING                    E.l 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.l Restore MEASURING                    F.l IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown          F.2. IMMEDIATE
* F. Exhaust plenum radiation monitor is not OPERATING OR F.3.a ENSURE continuous air radiation monitor is OPERATING AND F.3.a. IMMEDIATE F.3.b Restore MEASURING                  F.3.b Within 30 days CHANNEL G.l Do not perform a reactor            G.l 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                                                G.2 IMMEDIATE check on the startup channel 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 (8fG+.4/1744)
* 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 11 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 IO 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.
K-State Reactor                                TS-17                        Original (9fG-7.4/17+4)
* 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 (I)    The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLE CONTROL RODS (STANDARD) are capable of 90% of full reactivity insertion from the.
(2)                                                                                                -*--{ Formatted Table fully withdrawn position in less than 1 sec.
A minimum of three CONTROL RODS must be OPERABLE. Ino12erable CONTROL ill    RODS must be fully inserted.
TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum                    Function            Reauired OPERATING Mode Safety System Channel Number                                        STEADY          PULSE or Interlock Operable                                    STATE MODE        MODE Reactor power level          2            Scram                              YES            NA Manual scram bar              I            Scram                              YES            YES CONTROL ROD Prevent withdrawal of standard (STANDARD) position          1            rods in the PULSE MODE NA            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        Al. IMMEDIATE to operation A. Any required SAFETY SYSTEM CHANNEL or OR interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN K-State Reactor                                  TS-18                        Original (WG+4/1744)
* CONDITION                        REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME B.l ENSURE inoRerable control          Bl. IMMEDIATE rod is fully inserted B. A control rod is not OR OPERABLE.
B2. IMMEDIATE B.J2 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 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 (ST ANDARD) 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.
InoRerable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. ORerating with a fully-inserted control rod may cause Rower Reaking to shift. however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR ChaRter 13 is still bounding, and the reduction in maximum core Rower by having an inoRerable control rod fully inserted means that the highest temRerature in any fuel element with a fully-inserted inoperable control rod will be lower than the highest temRerature in the B-ring with all rods withdrawn. Therefore the reactor can be safely ORerated with an inoperable control rod Rrovided that the rod is fully inserted into the core.
K-State Reactor                                  TS-19                        Original (91-G+4/1744)
* 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 A.I ENSURE reactor is SHUTDOWN OR A.2.a Do not OPERA TE in the COMPLETION TIME A. I IMMEDIATE 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 (WG-74/1744)
* CONDITION                      REQUIRED ACTION TECHNICAL SPECIFICATIONS 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- 1 (2.0lxl0-5 &#xb5;Ci mJ-1), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10*3 &#xb5;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, ifrequired. 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. 90 Sr activity available for release from fuel rods previously used at other facilities is estimated to be at most about 4 times the ALL 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 (91G+.4/1744)
* 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 (3) reactor pool and core area
* 3.6.4    Actions CONDITION A. INDEPENDENT EXPERIMENT worth is REQUIRED ACTION A.I ENSURE the reactor is SHUTDOWN AND COMPLETION TIME A.1 IMMEDIATE greater than $2.00 A.2 Remove the experiment              A.2 Prior to continued operations C.I    ENSURE the reactor is            C.l 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 K-State Reactor                                  TS-22                        Original (9m+-4/1744)
* 3.6.5  Bases TECHNICAL SPECIFICATIONS 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 .
K-State Reactor                               TS-23                         Original (9!G+4/1744)
* TECHNICAL SPECIFICATIONS
* TECHNICAL SPECIFICATIONS
: 3. 7 Fuel Integrity 3.7.l 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 TRI GA reactor. 3.7.3 Specifications (I) Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over 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 the upper core grid plate. 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.
: 3. 7     Fuel Integrity 3.7.l     Applicability This specification applies to operations in STEADY STATE MODE and PULSE MODE.
Fuel cladding integrity is important since it represents the only process barrier for fission product release from the TRI GA reactor. K-State Reactor TS-24 Original (g,LG;Z.4/1744)
3.7.2     Objective The objective is to prevent the use of damaged fuel in the KSU TRI GA reactor.
* *
3.7.3     Specifications Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over (I) manufactured length (2)     Fuel elements in the reactor core SHALL NOT be laterally bent more than 1/8 in.
* TECHNICAL SPECIFICATIONS  
3.7.4     Actions CONDITION                       REQUIRED ACTION                 COMPLETION TIME
 
* A. Any fuel element is elongated greater than 1/8 in. over manufactured length, or bent laterally greater than 1/8 in.
===3.8 Reactor===
Do not insert the fuel element into the upper core grid plate.
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 Watef tem13efaffife at the ei<it ehhe feaetef 13ee!Peel iHletBulk water tem1:1erature SHALL (1) NOT exceed 44$-&deg;C (l 11J.&deg;F)l;JQ 0 I' with flew thfe11gh the J3FimaFy eleaH1113 lee13 (2) Water conductivity SHALL be less than 5 &#xb5;mho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core ill Peel iHletBulk water tem1:1erature SHALL NOT exceed 4-037&deg;C (99-l-04&deg;F) with an exgeriment installed in an interstitial flux wire QOrt. 3.8.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I ENSURE the reactor is A.I IMMEDIATE SHUTDOWN AND A. Water tempeFatHfe at the A.2 See11re flev: thre11gh the A.2 IMMEDIATE eiHtffiletB ulk water tem1:1erature ef the reaeter demiHerali20er j380l-_exceeds +.W4M &deg;F.G_ MID A.J.l Reduce bulk water A.3 IMMEDIATE temperature to less than -l-J.()4Mof'.C K-State Reactor TS-25 Original (9,LG+4/1744)
IMMEDIATE 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 TRI GA reactor.
* *
K-State Reactor                                 TS-24                       Original (g,LG;Z.4/1744)
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME B.l ENSURE the reactor is B.l IMMEDIATE SHUTDOWN B. l;1,1at:eF temf eFaffife at: the iHlet sf the FeaetsF AND ooel-Bulk water B.2 Reduce bulk water tem12erature B.2 IMMEDIATE tem12erature exceeds to less than-4-0 37&deg;C. 4-037&deg;C with an ex12eriment installed in an interstitial OR flux wire port. B.3 Remove ex12eriment from flux B.3 IMMEDIATE wire QOrt B.l ENSURE the reactor is B.l IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 &#xb5;mho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 &#xb5;mho/cm C. l ENSURE the reactor is C.l IMMEDIATE
* 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.
: 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  
3.8.2     Objective The objective is to set acceptable limits on the water quality, temperature, conductivity, and level in the reactor pool.
 
3.8.3     Specifications Watef tem13efaffife at the ei<it ehhe feaetef 13ee!Peel iHletBulk water tem1:1erature SHALL (1)
====3.8.5 Bases====
NOT exceed 44$-&deg;C (l 11J.&deg;F)l;JQ0 I' with flew thfe11gh the J3FimaFy eleaH1113 lee13 (2)   Water conductivity SHALL be less than 5 &#xb5;mho/cm (3)     Water level above the core SHALL be at least 13 ft from the top of the core
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&deg;F (54.4&deg;C).
* ill 3.8.4 Peel iHletBulk water tem1:1erature SHALL NOT exceed 4-037&deg;C (99-l-04&deg;F) with an exgeriment installed in an interstitial flux wire QOrt.
While the integrity of the ion exchange resin reguires water tem12erature to remain below 54.4&deg;C, it is necessary to maintain water tem12erature below 4M&deg;C to ensure that the de12arture from nucleate boiling ratio CDNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an ex12eriment into an interstitial flux wire QOrt between fuel elements necessitates a further reduction in water tem12erature to a maximum of 4-037&deg;C in order to 12reclude excessive nucleate boiling of the water eHsuFe a DNBR sf at least ;&.-0. Maintaining low water conductivity over a prolonged period prevents possible corrosion, deionizer degradation, or slow leakage of fission products from degraded cladding.
Actions CONDITION                         REQUIRED ACTION A.I ENSURE the reactor is COMPLETION TIME A.I IMMEDIATE SHUTDOWN AND A. Water tempeFatHfe at the A.2 See11re flev: thre11gh the       A.2 IMMEDIATE eiHtffiletB ulk water demiHerali20er tem1:1erature ef the reaeter j380l-_exceeds +.W4M&deg;F.G_
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.
MID A.J.l Reduce bulk water               A.3 IMMEDIATE temperature to less than
K-State Reactor TS-26 Original (91G+4/1744)
                                            -l-J.()4Mof'.C K-State Reactor                                   TS-25                     Original (9,LG+4/1744)
Formatted:
* CONDITION                      REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME B.l ENSURE the reactor is               B.l IMMEDIATE SHUTDOWN B. l;1,1at:eF temf eFaffife at: the iHlet sf the FeaetsF AND                                           . ~--( Formatted:  Centered ooel-Bulk water tem12erature exceeds B.2 Reduce bulk water tem12erature   B.2 IMMEDIATE           . -~--( Formatted: Indent: Left: O" to less than-4-0 37&deg;C.
Centered Formatted:
4-037&deg;C with an ex12eriment installed in an interstitial flux wire port.
Indent: Left: O" --(Formatted:
OR                                          .--(Formatted: Centered B.3 Remove ex12eriment from flux     B.3 IMMEDIATE wire QOrt B.l ENSURE the reactor is               B.l IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 &#xb5;mho/cm B.2 Restore conductivity to less       B.2 Within 4 weeks than 5 &#xb5;mho/cm C. l ENSURE the reactor is             C.l IMMEDIATE
Centered 
: 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&deg;F (54.4&deg;C). While the integrity of the ion exchange resin reguires water tem12erature to remain below 54.4&deg;C, it is necessary to maintain water tem12erature below 4M&deg;C to ensure that the de12arture from nucleate boiling ratio CDNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an ex12eriment into an interstitial flux wire QOrt between fuel elements necessitates a further reduction in water tem12erature to a maximum of 4-037&deg;C in order to 12reclude excessive nucleate boiling of the water eHsuFe a DNBR sf at least
* *
  ;&.-0.
* l__ TECHNICAL SPECIFICATIONS 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.
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.
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 . K-State Reactor TS-27 Original (9tQ.7.4/1744)
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.
* TECHNICAL SPECIFICATIONS  
K-State Reactor                                   TS-26                       Original (91G+4/1744)
 
* TECHNICAL SPECIFICATIONS 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.
===3.9 Maintenance===
Therefore, a minimum pool level of 13 feet above the core is adequate to provide shielding and support the core cooling .
 
K-State Reactor                             TS-27                       Original (9tQ.7.4/1744) l__
Retest Requirements  
* 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.
====3.9.1 Applicability====
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 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.
This specification ensures that if maintenance might challenge a Technical Specifications requirement, the requirement is verified prior to resumption of normal operations.
K-State Reactor TS-28 Original (94l-+4/1744)
K-State Reactor                                 TS-28                     Original (94l-+4/1744)
*
* TECHNICAL SPECIFICATIONS 3.10 Maximum Steady State Power 3.9.1   Applicability This specification applies to operations in STEADY STA TE MODE.
* TECHNICAL SPECIFICATIONS 3.10 Maximum Steady State Power 3.9.1 Applicability This specification applies to operations in STEADY ST A TE MODE. 3.9.2 Objective The objective is to ensure that the reactor has adeguate margin to critical heat flux CCHF) and operates below the Limiting Safety System Setting of 1.250 kWth. 3.9.3 Specifications ill Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY STATE MODE. ill A reguired reactor power level scram is set to a value no greater than 1.250 kWth. 3.9.4 Actions CONDITION REOUIRED ACTION COMPLETION TIME A. Thermal power exceeds 1.050 kWth in Reduce power to a level no greater IMMEDIATE STEADY ST A TE than 1.050 kWth. MODE B. A reguired reactor B.I SHUT DOWN the reactor. B.I. IMMEDIATE . -. ''* -f Formatted:
3.9.2   Objective The objective is to ensure that the reactor has adeguate margin to critical heat flux CCHF) and operates below the Limiting Safety System Setting of 1.250 kWth.
Centered ' Formatted Table . . ---Formatted:
3.9.3   Specifications ill Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY STATE MODE.
Numbered + Level: 1 + Numbering Style: A, B, C, ... + Start at: 1 +Alignment:
                                                                                                      . ''* -f Formatted: Centered
Left +Aligned at: 0.25" + "-, Indent at: 0.5'' ',,i Formatted:
                                                                                                                ' Formatted Table ill   A reguired reactor power level scram is set to a value no greater than 1.250 kWth.
Indent: Left: 0.03" power level scram is . ---( Formatted:
3.9.4   Actions CONDITION                       REOUIRED ACTION                   COMPLETION TIME A. Thermal power                                                                                 - - - Formatted: Numbered + Level: 1 + Numbering Style: A, B, exceeds 1.050 kWth in      Reduce power to a level no greater                                        C, ... + Start at: 1 +Alignment: Left +Aligned at: 0.25" +
Centered set to a value above AND AND 1.250 kWth or above ---[ Formatted:
IMMEDIATE                  "-,     Indent at: 0.5''
Centered the maximum readable B.2 Adjust reactor 12ower level B.2. Prior to resuming value on a reguired scram setpoint to a readable value operations.
STEADY STA TE              than 1.050 kWth.
less than or egual to 1.250 kWth. 3.9.5 Bases The reactor control panel instrumentation is designed to measure up to 1.000 kWth of thermal *---( Formatted:
MODE                                                                                              ',,i Formatted: Indent: Left:   0.03" B. A reguired reactor power level scram is set to a value above B.I SHUT DOWN the reactor.
Left 12ower. The Limiting Safety System Setting ensures that automatic protective functions.
AND B.I. IMMEDIATE          . - - - ( Formatted: Centered AND         *~ ---[ Formatted: Centered 1.250 kWth or above the maximum readable       B.2 Adjust reactor 12ower level B.2. Prior to resuming value on a reguired         scram setpoint to a readable value operations.
i.e .. high '-------------------
channel.~                  less than or egual to 1.250 kWth.
power scrams. are set to no greater than 1.250 kWth. However, by specifying the maximum OPERA TING 12ower level as 1.000 kWth in STEADY ST ATE MODE. the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum 12ower readable on the reactor console instruments.
3.9.5   Bases The reactor control panel instrumentation is designed to measure up to 1.000 kWth of thermal         *---( Formatted: Left 12ower. The Limiting Safety System Setting ensures that automatic protective functions. i.e .. high           '-------------------
Action to reduce power is not reguired until power exceeds I 050kWth in STEADY ST A TE MODE to allow for slight variation in power level that is typical during normal K-State Reactor TS-29 Original (MH-4/1744)
power scrams. are set to no greater than 1.250 kWth. However, by specifying the maximum OPERA TING 12ower level as 1.000 kWth in STEADY STATE MODE. the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum 12ower readable on the reactor console instruments. Action to reduce power is not reguired until power exceeds I 050kWth in STEADY ST A TE MODE to allow for slight variation in power level that is typical during normal operation.~
* *
K-State Reactor                                 TS-29                         Original (MH-4/1744)
* TECHNICAL SPECIFICATIONS
* TECHNICAL SPECIFICATIONS
: 4. Surveillance Requirements 4.1 Core Reactivity  
: 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
====4.1.1 Objective====
* Control Rod Reactivity Worth determination 4.1.3   Basis BIENNIAL 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.
 
K-State Reactor                                 TS-30                       Original (9/07 4/1744)
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 SHUTDOWN MARGIN Determination Excess Reactivity Determination Control Rod Reactivity Worth determination  
* 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.
====4.1.3 Basis====
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 .
FREQUENCY SEMIANNUAL SEMIANNUAL Following Insertion of experiments with measurable positive reactivity BIENNIAL 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.
K-State Reactor                               TS-31                       Original (9m+4/1744)
K-State Reactor TS-30 Original (9/07 4/1744)
* TECHNICAL SPECIFICATIONS 4.3     MEASURING CHANNELS 4.3.l   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                                               I 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.
* TECHNICAL SPECIFICATIONS  
K-State Reactor                               TS-32                     Original (f),lQ.;Z.4/1744)
 
* 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.
===4.2 PULSE===
4.4.2   Specifications SURVIELLANCE REQUIREMENTS SURVEILLANCE                                     FREQUENCY Manual scram SHALL be tested by releasing partially withdrawn DAILY CONTROL RODS (STANDARD)
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 ENSURE Transient Pulse Rod position corresponds to reactivity not greater than $3.00 4.2.3 Basis FREQUENCY Prior to pulsing operations Verifying pulse rod position corresponds to less than $3.00 ensures that the maximum pulsed reactivity meets the limiting condition for operation . K-State Reactor TS-31 Original (9m+4/1744)
CONTROL ROD (STANDARD) drop times SHALL be measured to have a drop time from the fully withdrawn position ofless than   ANNUAL 1 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.
* TECHNICAL SPECIFICATIONS  
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.
===4.3 MEASURING===
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-33                       Original (flfG+4/1744)
CHANNELS 4.3.l 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 Reactor power level MEASURING CHANNEL CHANNEL TEST Calorimetric calibration CHANNEL CHECK high voltage to required power level instruments Primary pool water temperature CHANNEL CALIBRATION Reactor Bay differential pressure CHANNEL CALIBRATION Fuel temperature CHANNEL CALIBRATION 22 Foot Area radiation monitor CHANNEL CHECK CHANNEL CALIBRATION 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION Continuous Air Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION EXHAUST PLENUM Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION Startup Count Rate 4.3.3 Basis FREQUENCY DAILY ANNUAL DAILY ANNUAL ANNUAL ANNUAL DAILY ANNUAL DAILY ANNUAL I DAILY ANNUAL DAILY ANNUAL DAILY 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.
K-State Reactor TS-32 Original (f),lQ.;Z.4/1744)
* *
* 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 ofless than ANNUAL 1 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 (TRANSIENT) system SHALL be performed.
each day a pulse is planned 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-33 Original (flfG+4/1744)
* *
* 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.
* 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 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 .
The inspection of the pulse rod mechanism will assure degradation of the pulse rod drive will be detected prior to malfunctions . K-State Reactor TS-34 Original (9tG+4/1744)
K-State Reactor                               TS-34                         Original (9tG+4/1744)
* *
* 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.
* TECHNICAL SPECIFICATIONS  
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.
===4.5 Gaseous===
K-State Reactor                                 TS-35                         Original (fltG+4/1744)
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.  
* 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 Measure and record experiment worth of the EXPERIMENT experiment where absolute (where the absolute value of the estimated worth is greater than value of the estimated
====4.5.2 Specification====
  $0.40).
 
worth is greater than $0.40 4.6.3   Basis
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.
* These surveillances allow determination that the limits of3.7 are met.
K-State Reactor TS-35 Original (fltG+4/1744)
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).
* TECHNICAL SPECIFICATIONS  
K-State Reactor                                 TS-36                         Original (WG-7-4/1744)
 
* TECHNICAL SPECIFICATIONS 4.7     Fuellntegrity 4.7.1   Objective The objective is to ensure that the dimensions of the fuel elements remain within acceptable limits.
===4.6 Limitations===
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 of3.00$
on Experiments  
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.
====4.6.1 Objectives====
K-State Reactor                                 TS-37                       Original (MH-411744)
 
* TECHNICAL SPECIFICATIONS 4.8     Reactor Pool Water This specification applies to the water contained in the KSU TRI GA reactor pool.
This surveillance ensures that experiments do not have significant negative impact on safety of the public, personnel or the facility.  
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
====4.6.2 Specification====
* 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.
SURVIELLANCE REQUIREMENTS SURVEILLANCE Experiments SHALL be evaluated and approved prior to implementation.
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.
Measure and record experiment worth of the EXPERIMENT (where the absolute value of the estimated worth is greater than $0.40). 4.6.3 Basis These surveillances allow determination that the limits of3.7 are met. FREQUENCY Prior to inserting a new experiment for purposes other than determination of reactivity worth Initial insertion of a new experiment where absolute value of the estimated worth is greater than $0.40 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). K-State Reactor TS-36 Original (WG-7-4/1744)
K-State Reactor                               TS-38                       Original (91G+4/1744)
* *
* 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.
* TECHNICAL SPECIFICATIONS  
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.
 
K-State Reactor                               TS-39                     Original (WG+4/17+4)
===4.7 Fuellntegrity===
 
====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 of3.00$ The standard fuel elements SHALL be visually inspected for cor-AND rosion and mechanical damage, and measured for length and bend B, C, D, E, and F RING elements comprising approximately 1/3 of the core SHALL be visually inspected annually for corrosion and mechanical damage such that the entire core SHALL be inspected at 3-year intervals, but not to exceed 38 months 4.7.4 Basis Following the exceeding of a limited safety system set point with potential for causing degradation ANNUAL 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. K-State Reactor TS-37 Original (MH-411744)
* *
* TECHNICAL SPECIFICATIONS  
 
===4.8 Reactor===
Pool Water This specification applies to the water contained in the KSU TRI GA 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.
K-State Reactor TS-38 Original (91G+4/1744)
* *
* 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 Evaluate potential for maintenance activities to affect operability and function of equipment required by Technical Specifications Perform surveillance to assure affected function meets requirements  
 
====4.9.3 Bases====
FREQUENCY Following maintenance of systems of equipment required by Technical Specifications Prior to resumption of normal operations 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. K-State Reactor TS-39 Original (WG+4/17+4)
* *
* TECHNICAL SPECIFICATIONS
* 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. 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====
5.1.3   Specification (I) The high-hydride fuel element shall contain uranium-zirconium hydride, clad in 0.020 in.
(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 J).-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 lo four elements with !lreater than 9.0 weight percent uranium rnav be inslulleJ in the core. These elements 1113v onlY be nluced in the E-and F-rings of the core lattice. and rnav not be acijaeeffi te eeHtrel rncis ef watef ehar.nels.located in the following positions:
of 304 stainless steel. It shall contain a maximum of J).-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 .
E2,E4.E5.E6.E20.E21.E22,E24,Fl,F2.F30.  
(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 lo four elements with !lreater than 9.0 weight percent uranium rnav be inslulleJ in the core. These elements 1113v onlY be nluced in the E- and F-rings of the core lattice. and rnav not be acijaeeffi te eeHtrel rncis ef watef ehar.nels.located in the following positions:
====5.1.4 Bases====
E2,E4.E5.E6.E20.E21.E22,E24,Fl,F2.F30.
These types of fuel elements have a long history of successful use in TRIGA reactors.
5.1.4     Bases These types of fuel elements have a long history of successful use in TRIGA reactors.
Calculations show that I 2'Y.,-loacl fuel in the E-ancl F-rings will not cxccccl the temperature of' 8'Yo-loacl instrumented  
Calculations show that I2'Y.,-loacl fuel in the E- ancl F-rings will not cxccccl the temperature of' 8'Yo-loacl instrumented .:lernents in the 8-ring. Aclditionullv the po\,cr peaking ancl fission product im*cntorY assumptions in the SAR will not b-o challe1rncd lw 12% fuel in the F- ancl F-rings. Locul power and ternpcrnture peaking effects during pulsing are avoided bv prnhibiting placement of the 12'',*0-loaQ_ll1el_!lCl\L_l.\'ater <irn:l c_Qntrnl rod channels.
.:lernents in the 8-ring. Aclditionullv the po\,cr peaking ancl fission product im*cntorY assumptions in the SAR will not b-o challe1rncd lw 12% fuel in the F-ancl rings. Locul power and ternpcrnture peaking effects during pulsing are avoided bv prnhibiting placement of the 12'',*0-loaQ_ll1el_!lCl\L_l.\'ater  
K-State Reactor                                       TS-40                     Original (!;lm+-4/1744)
<irn:l c_Qntrnl rod channels.
* 5.2     Reactor Fuel and Fueled Devices in Storage TECHNICAL SPECIFICATIONS 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.
K-State Reactor TS-40 Original (!;lm+-4/1744)
5.2.3   Specification (I)     All fuel elements or fueled devices shall be in a safe, stable geometry (2)     The kcrr 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.
* TECHNICAL SPECIFICATIONS
K-State Reactor                                 TS-41                         Original (0074/1744)
 
* 5.3       Reactor Building TECHNICAL SPECIFICATIONS 5.3.l     Applicability This specification applies to the building that houses the TRIGA reactor facility.
===5.2 Reactor===
5.3.2     Objective The objective is to ensure that provisions are made to restrict the amount ofrelease of radioactivity into the environment.
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 .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.
====5.2.3 Specification====
(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.
(I) All fuel elements or fueled devices shall be in a safe, stable geometry (2) The kcrr 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. K-State Reactor TS-41 Original (0074/1744)
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.
* TECHNICAL SPECIFICATIONS
K-State Reactor                                 TS-42                         Original (WG+4/1744)
 
* 5.4     Experiments TECHNICAL SPECIFICATIONS 5.4.1   Applicability This specification applies to the design of experiments.
===5.3 Reactor===
5.4.2   Objective The objective is to ensure that experiments are designed to meet criteria.
Building 5.3.l Applicability This specification applies to the building that houses the TRIGA reactor facility.  
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.
====5.3.2 Objective====
(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 ofreactor control instrumentation.
 
(5)     EXPERIMENT design shall minimize the potential for industrial hazards, such as fire or the release of hazardous and toxic materials.
The objective is to ensure that provisions are made to restrict the amount ofrelease of radioactivity into the environment.
(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.
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. (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.
(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 ofradioactivity 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.
K-State Reactor TS-42 Original (WG+4/1744)
(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.
* TECHNICAL SPECIFICATIONS
: 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.
 
K-State Reactor                                 TS-43                         Original (WG+4/1744)
===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 ofreactor 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 ofradioactivity 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. K-State Reactor TS-43 Original (WG+4/1744)
* *
* TECHNICAL SPECIFICATIONS
* TECHNICAL SPECIFICATIONS
: c. For materials whose boiling point is above 130&deg;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.  
: c. For materials whose boiling point is above 130&deg;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====
5.4.4   Basis Designing the experiment to react!Vlty 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 .
Designing the experiment to react!Vlty 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  
K-State Reactor                               TS-44                       Original (91m4/1744)
-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 . K-State Reactor TS-44 Original (91m4/1744)
* *
* TECHNICAL SPECIFICATIONS
* TECHNICAL SPECIFICATIONS
: 6. Administrative Controls 6.1 Organization and Responsibilities of Personnel a) Structure.
: 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 reactor organization is related to the University structure as shown in SAR Figure
The chief administrating officer for KSU is the President.
: 12. 1 and Technical Specifications Figure TS. 1 below.
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. President Kansas State University Vice President for Administration and Finance Director Division of Public Safety University Police Department Department of Environmental Health and Safety University Radiation Safety Officer Reactor Safeguards Committee Provost Chief Academic Officer Dean College of Engineering Head, Department of Mechanical  
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.
& Nuclear Engineering Manager, KSU Nuclear Reactor Facility Reactor Supervisor Reactor Operators 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.
President Kansas State University Vice President for                           Provost Administration                         Chief Academic and Finance                               Officer Director                                   Dean Division of Public Safety                 College of Engineering University       Department of                 Head, Department of Police          Environmental                Mechanical & Nuclear Department      Health and Safety                  Engineering University                      Manager, KSU Radiation Safety                  Nuclear Reactor Officer                          Facility Reactor Supervisor Reactor Safeguards Committee Reactor Operators 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-45                       Original (9!W4/1744)
Day-to-day radiation protection functions implemented by facility staff and management are guided K-State Reactor TS-45 Original (9!W4/1744)
* 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.
* *
* 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.
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 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 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.
The Manager also has responsibility for maintenance and modification of laboratories associated with the Reactor Facility.
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 OPERA TING 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 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.
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.
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 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 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 OPERA TING experience and have a demonstrated competence in supervision.
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-46                       Original  (9/G+4/17~)
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-46 Original   
* *
* 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.
* 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.
c). Staffing.
Whenever the reactor is not secured, the reactor shall be under the direction of a (USNRC licensed)
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 can, within twenty minutes travel time to the facility.
Senior Operator (designated as Reactor Supervisor).
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.
The Supervisor shall be on can, within twenty minutes travel time to the facility.
In addition to the above requirements, during fuel movement a senior operator shall be inside the reactor bay directing fuel operations.
Whenever the reactor is not secured, a (USNRC licensed)
6.2   Review and Audit a) There wilJ be a Reactor Safeguards Committee which shall review TRJGA 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.
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.
* b) The responsibilities of the Committee include, but are not limited to, the following:
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 wilJ be a Reactor Safeguards Committee which shall review TRJGA 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;
: 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 JO 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;
: 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;
: 6. Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR 20 and JO 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-47 Original (9/G.f.4/1744)
: 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 JO 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 JO 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-47                         Original (9/G.f.4/1744)
* TECHNICAL SPECIFICATIONS
* 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:
: 8. Requalification of the Nuclear Reactor Facility Manager and/or the Reactor Supervisor,
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.
: 9. Review of container failures where released materials have the potential for damaging reactor fuel or structural components including:
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-48 Original (9iG-74/1744)
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,
* 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: I. 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 OPERA TING 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.
: 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-48                       Original (9iG-74/1744)
* 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:
I. 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 OPERA TING 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.
: 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.
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.
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
The Rector Safeguards Committee shall subsequently review determinations that written procedures are not required.
: 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
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.
: 3. the University Radiation Safety Officer, or his/her deputy, who may withhold agreement until approval by the University Radiation Safety Committee is obtained.
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-49 Original (MH'.4/1744)
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.
* TECHNICAL SPECIFICATIONS  
K-State Reactor                                 TS-49                       Original (MH'.4/1744)
 
* 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.
===6.4 Review===
b) Procedures:
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:
I . 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.
I . 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.
: 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.
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.
: 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.
: 2. If the experimenter is a student, approval by his/her research supervisor is required.
: 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.
If the experimenter is a staff or faculty member, his/her own signature is sufficient.
: 5. Transmission to the Reactor Supervisor for scheduling.
: 3. The proposal is then to be submitted to the Reactor Safeguards Committee for consideration and approval.
c) Criteria that shall be met before approval can be granted shall include:
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.
I. The EXPERIMENT must meet the applicable Limiting Conditions for Operation and Design Description specifications.
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: I. 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.
: 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 ifthe test K-State Reactor TS-50 Original (91-G+-4/1744)
: 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 ifthe test K-State Reactor                               TS-50                       Original (91-G+-4/1744)
* *
* TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license amendment pursuant to IO CFR Sec. 50.90.
* TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license amendment pursuant to IO 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.
: 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.
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.  
6.5       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.
===6.6 Operator===
: 6. 7     Physical Security Plan
Requalification An operator requalification program shall be established and followed in accordance with NRC regulations.
* Administrative controls for protection of the reactor plant 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.
===6.8 Action===
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.
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 ofrecurrence. 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.
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 ofrecurrence.
6.9     Action To Be Taken In The Event Of A Reportable Occurrence a) A reportable occurrence is any of the following conditions:
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:
K-State Reactor                               TS-51                         Original (Mf-74/1744)
K-State Reactor TS-51 Original (Mf-74/1744)
* *
* TECHNICAL SPECIFICATIONS I. any actual safety system setting less conservative than specified in Section 2.2, Limiting Safety System Settings;
* 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 through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.
: 2. VIOLATION OF SL, LSSS OR LCO; NOTES Violation of an LSSS or LCO occurs through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.
Surveillance Requirements must be met for all equipment/components/conditions.
Surveillance Requirements must be met for all equipment/components/conditions. to be considered operable.
to be considered operable.
Failure to perform a surveillance within the required time interval or failure of a surveillance test shall result in the /component/condition being inoperable
Failure to perform a surveillance within the required time 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 1 OCFR20 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.
: 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;
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.
: 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;
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.
: 5. an uncontrolled or unanticipated change in reactivity greater than $1.00;
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-52 Original (9,lG.74/1744)
: 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-52                     Original (9,lG.74/1744)
* TECHNICAL SPECIFICATIONS
* TECHNICAL SPECIFICATIONS
: 1. normal plant operation, including power levels; 3. principal maintenance activities;
: 1. normal plant operation, including power levels;
: 3. principal maintenance activities;
: 4. reportable occurrences;
: 4. reportable occurrences;
: 5. equipment and component surveillance activities;
: 5. equipment and component surveillance activities;
: 6. experiments performed with the reactor; 7. all emergency reactor scrams, including reasons for emergency shutdowns.
: 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:
b) The following records shall be maintained for the life of the facility:
: 1. gaseous and liquid radioactive effluents released to the environs;
: 1. gaseous and liquid radioactive effluents released to the environs;
: 2. offsite environmental monitoring surveys; 3. fuel inventories and transfers;
: 2. offsite environmental monitoring surveys;
: 4. facility radiation and contamination surveys; 5. radiation exposures for all personnel;
: 3. fuel inventories and transfers;
: 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 hours 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;
: 4. facility radiation and contamination surveys;
: 2. any violation of a safety limit; 3. any reportable occurrences as defined in Section 6.9 of these specifications.
: 5. radiation exposures for all personnel;
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-53 Original (9/.G+4/1744)
* 6.11
* *
: 6. updated, corrected, and as-built drawings of the facility .
* 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.
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.
c) A report within 30 days in writing of: I. 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 OPERA TING conditions or characteristics of the reactor under the new conditions.
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:
e) A routine report in writing to the US. Nuclear Regulatory Commission within 60 days after completion of the first calendar year of OPERA TING and at intervals not to exceed 12 months, thereafter, providing the following information:
a) A report within 24 hours 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-53                       Original (9/.G+4/1744)
* 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:
I. 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 OPERA TING 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 OPERA TING and at intervals not to exceed 12 months, thereafter, providing the following information:
I. 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;
I. 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);
: 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;
: 3. the number of emergency shutdowns and inadvertent scrams, including the reasons thereof and corrective action, if any, taken;
: 5. a summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of IO 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;
: 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;
: 7. a description of any environmental surveys performed outside the facility; K-State Reactor TS-54 Original (9fG+4/1744)
: 5. a summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of IO 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; K-State Reactor                               TS-54                       Original (9fG+4/1744)
* TECHNICAL SPECIFICATIONS
* 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 ofradiation and contamination surveys performed within the facility . K-State Reactor TS-55 Original (9fG.7-4/1744)
: 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 ofradiation and contamination surveys performed within the facility .
TECHNICAL SPECIFICATIONS
K-State Reactor                           TS-55                       Original (9fG.7-4/1744)
* Table of Contents 1. DEFINITIONS
.................................................................................................................
TS-1 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .......................
TS-8 2.1 Fuel Element Temperature Safety Limit.. ....................................................................
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 Limiting Safety System Settings ................................................................................
TS-10 2.2.1. Applicability
......................................................................................................
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-11 3 .1 CORE REACTIVITY
................................................................................................
TS-11 3 .1.1. Applicability
......................................................................................................
TS-11 3 .1.3. Objective
............................................................................................................
TS-11 3 .1. 4. Specification
.......................................................................................................
TS-11 3.1.5. Actions ...............................................................................................................
TS-12 3.1.6. Basis ..................................................................................................................
TS-13
* 3.2 PULSED MODE OPERATIONS
..............................................................................
TS-13 3 .2.1. Applicability
......................................................................................................
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 MEASURING CHANNELS ......................................................................................
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. Applicability
......................................................................................................
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 GASEOUS EFFLUENT CONTROL .......................................................
: .................
TS-20 3. 5 .1. Applicability
......................................................................................................
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 LIMITATIONS ON EXPERIMENTS
..........................................................................
TS-22 3 .6.1. Applicability
......................................................................................................
TS-22 3.6.3. Objective
............................................................................................................
TS-22 K-State Reactor TS-1 Original (4/17)
TECHNICAL SPECIFICATIONS


====3.6.4. Specification====
TECHNICAL SPECIFICATIONS Table of Contents
: 1. DEFINITIONS ................................................................................................................. TS-1
: 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Element Temperature Safety Limit.. .................................................................... 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 Limiting Safety System Settings ................................................................................ TS-10 2.2.1. Applicability ...................................................................................................... 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-11 3 .1 CORE REACTIVITY ................................................................................................ TS-11 3 .1.1. Applicability ...................................................................................................... TS-11 3 .1.3. Objective ............................................................................................................ TS-11 3 .1. 4. Specification ....................................................................................................... TS-11 3.1.5. Actions ............................................................................................................... TS-12 3.1.6. Basis .................................................................................................................. TS-13
* 3.2 PULSED MODE OPERATIONS .............................................................................. TS-13 3 .2.1. Applicability ...................................................................................................... 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 MEASURING CHANNELS ...................................................................................... 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. Applicability ...................................................................................................... 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 GASEOUS EFFLUENT CONTROL ....................................................... :................. TS-20
: 3. 5 .1. Applicability ...................................................................................................... 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 LIMITATIONS ON EXPERIMENTS .......................................................................... TS-22 3 .6.1. Applicability ...................................................................................................... TS-22 3.6.3. Objective ............................................................................................................ TS-22 K-State Reactor                                                TS-1                                                      Original (4/17)


.......................................................................................................
TECHNICAL SPECIFICATIONS 3.6.4. Specification ....................................................................................................... TS-22
TS-22
* 3.6.5. Actions ............................................................................................................... TS-22 3.6.6. Basis .................................................................................................................. TS-23 3.7 FUEL INTEGRITY ................................................................................................... TS-24
* 3.6.5. Actions ...............................................................................................................
: 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 WATER ......................................................................................... 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 Maintenance Retest Requirements ................................................................................ TS-27 3 .9 .1. Applicability ...................................................................................................... TS-2 7 3.9.3. Objective ............................................................................................................ TS-27 3.9.4. Specification ....................................................................................................... TS-27 3.9.5. Actions ............................................................................................................... TS-27 3.9.6. Basis .................................................................................................................. TS-27
TS-22 3.6.6. Basis ..................................................................................................................
: 4. SURVIELLANCES .......................................................................................................... TS-28 4.1 CORE REACTIVITY ................................................................................................ 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 MEASURING CHANNELS ...................................................................................... 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 LIMITATIONS ON EXPERIMENTS ........................................................................ TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis .................................................................................................................. TS-34 4.7 FUEL 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 WATER ....................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor                                                 TS-2                                                     Original (4/17)
TS-23 3.7 FUEL INTEGRITY  
...................................................................................................
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 WATER .........................................................................................
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 Maintenance Retest Requirements  
................................................................................
TS-27 3 .9 .1. Applicability  
......................................................................................................
TS-2 7 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 REACTIVITY  
................................................................................................
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 MEASURING CHANNELS ......................................................................................
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 LIMITATIONS ON EXPERIMENTS  
........................................................................
TS-34 4.6.1. Objective  
...........................................................................................................
TS-34 4.6.2. Specification  
......................................................................................................
TS-34 4.6.3. Basis ..................................................................................................................
TS-34 4.7 FUEL 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 WATER .......................................................................................
TS-36 4.8.1. Objective  
...........................................................................................................
TS-36 K-State Reactor TS-2 Original (4/17)
TECHNICAL SPECIFICATIONS


====4.8.2. Specification====
TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36
* 4.8.3. Basis .................................................................................................................. TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................'................................................... TS-37 4.10.3. Basis ................................................................................................................ TS-37
: 5. DESIGN FEATURES ...................................................................................................... TS-38 5.1 REACTOR FUEL ...................................................................................................... TS-38 5 .1.1. Applicability ...................................................................................................... TS-3 8 5.1.2. Objective ............................................................................................................ TS-38 5 .1.3. Specification ....................................................................................................... TS-3 8 5.1.4. Basis .................................................................................................................. TS-38 5 .2 REACTOR FUEL AND FUELED DEVICES IN STORAGE .................................. TS-3 8 5 .2.1. Applicability ...................................................................................................... TS-3 8 5.2.2. Objective ............................................................................................................ TS-39 5.2.3. Specification ....................................................................................................... TS-39 5.2.4. Basis .................................................................................................................. TS-39 5.3 REACTOR BUILDING ............................................................................................. TS-39 5.3.1. Applicability ...................................................................................................... TS-39 5.3.2. Objective ............................................................................................................ TS-39 5 .3 .3. Specification ....................................................................................................... TS-3 9 5 .3 .4. Basis .................................................................................................................. TS-40 5.4 EXPERIMENTS ......................................................................................................... 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. ADMINISTRATIVE CONTROLS ................................................................................. TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL. ......................... TS-44 6.2 REVIEW AND AUDIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS .................................................. TS-47 6.5 EMERGENCY PLAN AND PROCEDURES ........................................................... TS-48 6.6 OPERATOR REQUALIFICATION .......................................................................... TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERA TING RECORDS ............................................................................ TS-49 6.11 REPORTING REQUIREMENTS ........................................................... TS-50
* K-State Reactor                                                TS-3                                                      Original (4/17)
L_


......................................................................................................
TECHNICAL SPECIFICATIONS
TS-36
: 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.
* 4.8.3. Basis ..................................................................................................................
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."
TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS
In using Action Statements, the following guidance applies:
.......................................................
TS-37 4.9.1. Objective
...........................................................................................................
TS-37 4.9.2. Specification
...................................................
' ...................................................
TS-37 4.10.3. Basis ................................................................................................................
TS-37 5. DESIGN FEATURES ......................................................................................................
TS-38 5.1 REACTOR FUEL ......................................................................................................
TS-38 5 .1.1. Applicability
......................................................................................................
TS-3 8 5.1.2. Objective
............................................................................................................
TS-38 5 .1.3. Specification
.......................................................................................................
TS-3 8 5.1.4. Basis ..................................................................................................................
TS-38 5 .2 REACTOR FUEL AND FUELED DEVICES IN STORAGE ..................................
TS-3 8 5 .2.1. Applicability
......................................................................................................
TS-3 8 5.2.2. Objective
............................................................................................................
TS-39 5.2.3. Specification
.......................................................................................................
TS-39 5.2.4. Basis ..................................................................................................................
TS-39 5.3 REACTOR BUILDING .............................................................................................
TS-39 5.3.1. Applicability
......................................................................................................
TS-39 5.3.2. Objective
............................................................................................................
TS-39 5 .3 .3. Specification
.......................................................................................................
TS-3 9 5 .3 .4. Basis ..................................................................................................................
TS-40 5.4 EXPERIMENTS
.........................................................................................................
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. ADMINISTRATIVE CONTROLS .................................................................................
TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL.
.........................
TS-44 6.2 REVIEW AND AUDIT .............................................................................................
TS-45 6.3 PROCEDURES
............................................................................................................
TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS
..................................................
TS-47 6.5 EMERGENCY PLAN AND PROCEDURES
...........................................................
TS-48 6.6 OPERA TOR REQUALIFICATION
..........................................................................
TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE
....................................................................
TS-48 6.10 PLANT OPERA TING RECORDS ............................................................................
TS-49 6.11 REPORTING REQUIREMENTS
...........................................................
TS-50
* K-State Reactor TS-3 Original (4/17) L_ 
* *
* 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 ANNUAL CHANNEL CALIBRATION BIENNIAL CHANNEL CHECK CHANNEL TEST CONTROL ROD (STANDARD)
CONTROL ROD (TRANSIENT)
DAILY K-State Reactor 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 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.
* 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 .
* 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 12 months, not to exceed 15 months A channel calibration is an adjustment of the channel to that its output responds, with acceptable range and accuracy, to known values of the parameter that the channel measures.
* ANNUAL
Every two years, not to exceed a 28 month interval A channel check is a qualitative verification of acceptable performance by observation of channel behavior.
* 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 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.
This verification shall include comparison of the channel with expected values, other independent channels, or other methods of measuring the same variable.
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.
A channel test is the introduction of an input signal into a channel to verify that it is operable.
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.
A functional test of operability is a channel test. A standard control rod is one having an electric motor drive and scram capability.
CONTROL ROD A standard control rod is one having an electric motor drive and scram (STANDARD) capability.
A transient rod is one that is pneumatically operated and has scram capability.
CONTROL ROD A transient rod is one that is pneumatically operated and has scram (TRANSIENT) capability.
Prior to initial operation each day (when the reactor is operated), or before TS-4 Original (4/17) 
DAILY                Prior to initial operation each day (when the reactor is operated), or before K-State Reactor                                 TS-4                                 Original (4/17)
* *
* ENSURE EXHAUST PLENUM EXPERIMENT EXPERIMENTAL FACILITY IMMEDIATE INDEPENDENT EXPERIMENT LIMITING CONDITION FOR OPERATION (LCO) LIMITING SAFETY SYSTEM SETTING (LSSS) MEASURED VALUE MEASURING CHANNEL MOVABLE EXPERIMENT NON SECURED EXPERIMENT K-State Reactor TECHNICAL SPECIFICATIONS an operation extending more than 1 day Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition The air volume in the reactor bay atmosphere between the pool surface and the reactor bay exhaust fan 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 facilities are the beamports, thermal column, pneumatic transfer system, central thimble, rotary specimen rack, and the in-core facilities (including non-contiguous single-element positions, and, in the E and Frings, as many as three contiguous fuel-element positions).
Without delay, and not exceeding one hour. NOTE: IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action INDEPENDENT Experiments are those not connected by a mechanical, chemical, or electrical link to another experiment The lowest functional capability or performance levels of equipment required for safe operation of the facility.
Settings for automatic protective devices related to those variables having significant safety functions.
Where a limiting safety system setting is 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.
The measured value of a parameter is the value as it appears at the output of a MEASURING CHANNEL. A MEASURING CHANNEL is the combination of sensor, lines, amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.
A MOVABLE EXPERIMENT is one that may be moved into, out-of or near the reactor while the reactor is OPERATING.
NONSECURED Experiments are these that should not move while the reactor is OPERATING, but are held in place with less restraint than a secured experiment.
TS-5 Original (4/17) 
* *
* OPERABLE OPERATING PULSE MODE REACTOR SAFETY SYSTEM REACTOR SECURED MODE REACTOR SHUTDOWN RING REFERENCE CORE CONDITION SAFETY CHANNEL SECURED EXPERIMENT K-State Reactor TECHNICAL SPECIFICATIONS A system or component is OPERABLE when it is capable of performing its intended function in a normal manner A system or component is OPERATING when it is performing its intended function in a normal manner. 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, reactor power may be increased on a period of much less than l second by motion of the transient control rod. The REACTOR SAFETY SYSTEM is that combination of MEASURING CHANNELS and associated circuitry that is designed to initiate reactor scram or that provides information that requires manual protective action to be initiated.
The reactor is secured when the conditions of either item (1) or item (2) are 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 The reactor is shutdown if it is subcritical by at least $1.00 in the REFERENCE CORE CONDITION with the reactivity worth of all experiments included.
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, The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible
(<$0.30) A safety channel is a MEASURING CHANNEL 111 the REACTOR SAFETY SYSTEM A secured EXPERIMENT is an EXPERIMENT held firmly in place by a 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 (4/17) 
* *
* SECURED EXPERIMENT WITH MOVABLE PARTS SHALL (SHALL NOT) SEMIANNUAL SHUTDOWN MARGIN STANDARD TECHNICAL SPECIFICATIONS A secured EXPERIMENT with movable parts is one that contains parts that are intended to be moved while the reactor is OPERATING.
Indicates specified action is required/( not to be performed)
Every six months, with intervals not greater than 8 months The shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating condition, and that the reactor will remain subcritical without further operator action 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 MODE TECHNICAL SPECIFICATION VIOLATION K-State Reactor The reactor is in the steady-state mode when the reactor mode selector switch is in either the manual or automatic position and the key switch is in the "on" position.
A violation of a Safety Limit occurs when the Safety Limit value is exceeded.
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 applicable titles of sections in individual Technical Specifications TS-7 Original (4/17)
* *
* TECHNICAL SPECIFICATIONS 2 . SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Fuel Element Temperature Safety Limit 2. I. I Applicability This specification applies when the reactor in STEADY STATE MODE and the PULSE MODE. 2. I .2 Objective This SAFETY LIMIT ensures fuel element cladding integrity
: 2. I .3 Specification (I) Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed I I50&deg;C. (2) Steady state fuel temperature shall not exceed 750&deg;C. 2. I .4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A Stainless steel clad, high-A. I Establish SHUTDOWN A.I IMMEDIATE hydride fuel element / condition temperature exceeds II50&deg;C. OR AND Fuel temperature exceeds 750&deg;C in steady state A.2 Report per Section 6.8 A.2 Within 24 hours conditions


====2.1.5 Bases====
TECHNICAL SPECIFICATIONS an operation extending more than 1 day
Safety Analysis Report, Section 3.5.I (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged.
* ENSURE EXHAUST PLENUM Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition The air volume in the reactor bay atmosphere between the pool surface and 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.
NUREG 1282 identifies the safety limit for the high-hydride (ZrHu) 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).
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 Frings, as many as three contiguous fuel-element positions).
This stress will remain below the yield strength of the stainless steel cladding with fuel temperatures below l,I50&deg;C. A change in yield strength occurs for stainless steel cladding temperatures of 500&deg;C, but there is no scenario for fuel cladding to achieve 500&deg;C while submerged; consequently the safety limit during reactor operations is I,150&deg;C. K-State Reactor TS-8 Original (4/17)
IMMEDIATE Without delay, and not exceeding one hour.
* *
NOTE:
* TECHNICAL SPECIFICATIONS Therefore, the important process variable for a TRIGA reactor is the fuel element temperature.
IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action 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.
This parameter is well suited as a single specification, and it is readily measured.
(LCO)
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.
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.
Fuel temperature acting on these gases controls fuel element internal pressure.
MEASURED The measured value of a parameter is the value as it appears at the output VALUE of a MEASURING CHANNEL.
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-117-833.
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.
Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients.
MOVABLE A MOVABLE EXPERIMENT is one that may be moved into, out-of or EXPERIMENT near the reactor while the reactor is OPERATING.
Operating with maximum long-term, steady state fuel temperature of 750&deg;C does not have significant time-and temperature-dependent fuel growth . K-State Reactor TS-9 Original (4/17)
NON SECURED 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.
* *
K-State Reactor                         TS-5                                Original (4/17)
* TECHNICAL SPECIFICATIONS


===2.2 Limiting===
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.
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.  
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, reactor power may be increased on a period of much less than l second by motion of the transient control rod.
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                    111 the REACTOR CHANNEL SAFETY SYSTEM
* SECURED EXPERIMENT K-State Reactor A secured EXPERIMENT is an EXPERIMENT held firmly in place by a 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 (4/17)


====2.2.3 Specifications====
TECHNICAL SPECIFICATIONS SECURED
* EXPERIMENT WITH MOVABLE PARTS SHALL (SHALL NOT)
A secured EXPERIMENT with movable parts is one that contains parts that are intended to be moved while the reactor is OPERATING.
Indicates specified action is required/( not to be performed)
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 VIOLATION exceeded.
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 applicable titles of sections in individual Technical Specifications
* K-State Reactor                          TS-7                                Original (4/17)


j (1) j Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation j 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. l Reduce power to less than A. I IMMEDIATE 1,250 kW (th) A. Steady state power level OR exceeds 1,250 kW (th) A.2. Establish REACTOR A.2. IMMEDIATE SHUTDOWN condition
TECHNICAL SPECIFICATIONS
* 2.
2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS Fuel Element Temperature Safety Limit
: 2. I. I    Applicability This specification applies when the reactor in STEADY STATE MODE and the PULSE MODE.
: 2. I .2    Objective This SAFETY LIMIT ensures fuel element cladding integrity
: 2. I .3    Specification (I)     Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed I I50&deg;C.
(2)     Steady state fuel temperature shall not exceed 750&deg;C.
: 2. I .4   Actions
* CONDITION A      Stainless steel clad, high-hydride fuel element temperature exceeds II50&deg;C.
                                      /
REQUIRED ACTION A. I Establish SHUTDOWN condition COMPLETION TIME A.I IMMEDIATE OR                              AND Fuel temperature exceeds 750&deg;C in steady state          A.2 Report per Section 6.8 conditions                                                    A.2 Within 24 hours 2.1.5      Bases Safety Analysis Report, Section 3.5.I (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged.
NUREG 1282 identifies the safety limit for the high-hydride (ZrHu) 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 l,I50&deg;C. A change in yield strength occurs for stainless steel cladding temperatures of 500&deg;C, but there is no scenario for fuel cladding to achieve 500&deg;C while submerged; consequently the safety limit during reactor operations is I,150&deg;C.
K-State Reactor                                    TS-8                            Original (4/17)


====2.2.5 Bases====
TECHNICAL SPECIFICATIONS Therefore, the important process variable for a TRIGA reactor is the fuel element temperature.
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.
* 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.
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.
Fuel growth and deformation can occur during normal operations, as described in General Atomics technical report E-117-833. Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients. Operating with maximum long-term, steady state fuel temperature of 750&deg;C does not have significant time- and temperature-dependent fuel growth .
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.
* K-State Reactor                               TS-9                            Original (4/17)
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 (4/17) 
* *
* 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.I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meetLCO operations K-State Reactor TS-11 Original (4/17) 
* *
* TECHNICAL SPECIFICATIONS B.l.a ENSURE control rods B.1 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.1.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 meetLCO 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 l,150&deg;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&deg;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&deg;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 . K-State Reactor TS-12 Original (4/17)
* *
* TECHNICAL SPECIFICATIONS


===3.2 PULSED===
TECHNICAL SPECIFICATIONS
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.  
* 2.2 2.2.1 Limiting Safety System Settings (LSSS)
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 j  (1)  j Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation                j 2.2.4    Actions CONDITION                        REQUIRED ACTION                  COMPLETION TIME A. I IMMEDIATE A. l Reduce power to less than 1,250 kW (th)
A. Steady state power level OR exceeds 1,250 kW (th)
A.2. Establish REACTOR A.2. IMMEDIATE 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 (4/17)


====3.2.3 Specification====
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:
: 1. The highest worth control rod is fully withdrawn (2)
: 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.I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn                        AND exceeds $4.00 A.2 Configure reactor to      A.2 Prior to continued meetLCO                        operations
* K-State Reactor                              TS-11                                Original (4/17)


(1) 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.1 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 (4/17)
TECHNICAL SPECIFICATIONS B.l.a ENSURE control rods        B.1 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
* TECHNICAL SPECIFICATIONS
      $0.50 under specified conditions                    B.1.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 meetLCO 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 l,150&deg;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&deg;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&deg;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 .
* K-State Reactor                               TS-12                                Original (4/17)


===3.3 MEASURING===
TECHNICAL SPECIFICATIONS
* 3.2 3 .2.1 PULSED MODE Operations 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.1 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 the pulse rod in the transient rod drive position is greater than $3.00 in the PULSE MODE or equal to $3.00 OR A.2 Place reactor in STEADY STATE MODE OR A.2 IMMEDIATE 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 (4/17)


CHANNELS 3 .3 .1 Applicability This specification applies to the reactor MEASURING CHANNELS during STEADY STATE MODE and PULSE MODE operations.  
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 1 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 leveJPl                                    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[2l                      1                        1 EXHAUST PLENUM radiation monitor[21                      1                        1 NOTE[l]: 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.1.1 Restore channel to operation      A.1.1 IMMEDIATE A. l Reactor power channels not OPERATING (min 2                          OR for STEADY STATE, 1 A.1.2 ENSURE reactor is                A.1.2 IMMEDIATE PULSE MODE)
SHUTDOWN
* K-State Reactor                                  TS-14                                Original (4/17)


====3.3.2 Objective====
TECHNICAL SPECIFICATIONS CONDITION                  REQUIRED ACTION              COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less                                      A.2. IMMEDIATE AND 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 OR C. 22 foot Area radiation monitor is not C.3 ENSURE personnel are not        C.3 IMMEDIATE OPERATING on the 22 foot level
* OR C.4 ENSURE personnel on 22 foot level are using portable survey meters to monitor dose C.4 IMMEDIATE rates D.l 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 K-State Reactor                          TS-15                            Original (4/17)


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 1 SHALL be OPERA TING (2) The neutron count rate on the startup channel is greater than the minimum detector sensitivity TABLE 1: MINIMUM MEASURING CHANNEL COMPLEMENT MEASURING CHANNEL Reactor power leveJPl Primary Pool Water Temperature Reactor Bay Differential Pressure Fuel Temperature 22 foot Area radiation monitor 0 or 12 foot Area monitor Minimum Number Operable STEADY STATE PULSE MODE MODE 2 1 1 1 1 1 1 1 Continuous air radiation monitor[2l EXHAUST PLENUM radiation monitor[21 1 1 1 1 1 1 1 1 NOTE[l]: One "Startup Channel" required to have range that indicates
TECHNICAL SPECIFICATIONS CONDITION                     REQUIRED ACTION                 COMPLETION TIME
<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. l Reactor power channels A.1.1 Restore channel to operation A.1.1 IMMEDIATE not OPERATING (min 2 OR for STEADY STATE, 1 A.1.2 ENSURE reactor is PULSE MODE) A.1.2 IMMEDIATE SHUTDOWN K-State Reactor TS-14 Original (4/17)
* E.l Restore MEASURING CHANNEL OR E.1 IMMEDIATE 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.l IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown         F.2. IMMEDIATE
TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME
* F. Exhaust plenum radiation monitor is not OPERATING OR F.3.a ENSURE continuous air radiation monitor is OPERATING F.3.a. IMMEDIATE AND F.3.b Restore MEASURING                 F.3.b Within 30 days CHANNEL G.l   Do not perform a reactor         G.l 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                                                G.2 IMMEDIATE check on the startup channel prior to startup 3.3.5   Bases Maximum steady state power level is 1,250 kW; neutron detectors measure reactor power level.
* A.2.1 Establish REACTOR A.2 High voltage to reactor SHUTDOWN condition 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 C. 22 foot Area radiation OR monitor is not C.3 ENSURE personnel are not C.3 IMMEDIATE OPERATING on the 22 foot 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.
* OR C.4 ENSURE personnel on 22 C.4 IMMEDIATE foot level are using portable survey meters to monitor dose rates D.l 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 OPERA TING 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
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-15 Original (4/17) 
K-State Reactor                               TS-16                               Original (4/17)
* *
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME E.l 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 OPERA TING AND E.3.b Restore MEASURING E.3.b Within 30 days CHANNEL F.1 Restore MEASURING F.l 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.l Do not perform a reactor G.l IMMEDIATE G. The neutron count rate on startup the startup channel is not OR greater than the minimum G.2 Perform a neutron-source G.2 IMMEDIATE detector sensitivity check on the startup channel 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 (4/17)
* *
* 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 11 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. K-State Reactor TS-17 Original (4/17) 
* *
* TECHNICAL SPECIFICATIONS


===3.4 Safety===
TECHNICAL SPECIFICATIONS
Channel and Control Rod Operability
* 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 11 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.
K-State Reactor                                TS-17                                Original (4/17)


====3.4.1 Applicability====
TECHNICAL SPECIFICATIONS
* 3.4 3.4.1 Safety Channel and Control Rod Operability 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 of90% of full reactivity insertion from the (2) fully withdrawn position in less than 1 sec.
A minimum of three CONTROL RODS must be OPERABLE. Inoperable CONTROL (3)
RODS must be fully inserted .
TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum                    Function            Required OPERATING Mode Safety System Channel Number                                        STEADY          PULSE or Interlock Operable                                    STATE MODE        MODE Reactor power level            2            Scram                            YES            NA Manual scram bar              1            Scram                            YES            YES CONTROL ROD Prevent withdrawal of standard (STAND ARD) position          1            rods in the PULSE MODE NA            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.1 Restore channel or interlock        Al. IMMEDIATE to operation A. Any required SAFETY SYSTEM CHANNEL or OR interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN K-State Reactor                                    TS-18                                Original (4/17)


This specification applies to the reactor MEASURING Channels during STEADY STATE MODE and PULSE MODE operations.  
TECHNICAL SPECIFICATIONS CONDITION                        REQUIRED ACTION                  COMPLETION TIME
* B. A control rod is not OPERABLE.
B.1 ENSURE inoperable control rod is fully inserted OR Bl. IMMEDIATE B2. IMMEDIATE B.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 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.
Inoperable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. Operating with a fully-inserted control rod may cause power peaking to shift, however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR Chapter 13 is still bounding, and the reduction in maximum core power by having an inoperable control rod fully inserted means that the highest temperature in any fuel element with a fully-inserted inoperable control rod will be lower than the highest temperature in the B-ring with all rods withdrawn. Therefore the reactor can be safely operated with an inoperable control rod provided that the rod is fully inserted into the core .
* K-State Reactor                                  TS-19                                Original (4/17)


====3.4.2 Objective====
TECHNICAL SPECIFICATIONS
 
* 3.5 3 .5 .1 Gaseous Effluent Control Applicability This specification applies to gaseous effluent in STEADY STATE MODE and PULSE MODE.
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 (2) CONTROL RODS (STANDARD) are capable of90% of full reactivity insertion from the fully withdrawn position in less than 1 sec. (3) A minimum of three CONTROL RODS must be OPERABLE.
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.
Inoperable CONTROL RODS must be fully inserted . TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Safety System Channel Minimum Function Required OPERATING Mode Number STEADY PULSE or Interlock Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar 1 Scram YES YES CONTROL ROD Prevent withdrawal of standard (ST AND ARD) position 1 rods in the PULSE MODE NA YES interlock Prevent inadvertent pulsing Pulse rod interlock 1 while in STEADY ST ATE YES NA MODE 3.4.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.1 Restore channel or interlock Al. IMMEDIATE A. Any required SAFETY to operation SYSTEM CHANNEL or OR interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN K-State Reactor TS-18 Original (4/17) 
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.
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME B.1 ENSURE inoperable control Bl. IMMEDIATE rod is fully inserted B. A control rod is not OR OPERABLE.
3.5.4     Actions
B2. IMMEDIATE B.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.
* CONDITION                       REQUIRED ACTION A.1 ENSURE reactor is SHUTDOWN OR COMPLETION TIME A.1 IMMEDIATE 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.
The power level scram is not credited 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.
A.2.c ENSURE no irradiated fuel       A.2.b IMMEDIATE handing AND A.2.d Within 30 days A.2.d Restore the reactor bay ventilation exhaust system to OPERABLE K-State Reactor                                 TS-20                              Original (4/17)
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.
Inoperable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. Operating with a fully-inserted control rod may cause power peaking to shift, however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR Chapter 13 is still bounding, and the reduction in maximum core power by having an inoperable control rod fully inserted means that the highest temperature in any fuel element with a inserted inoperable control rod will be lower than the highest temperature in the B-ring with all rods withdrawn.
Therefore the reactor can be safely operated with an inoperable control rod provided that the rod is fully inserted into the core . K-State Reactor TS-19 Original (4/17)
* *
* TECHNICAL SPECIFICATIONS


===3.5 Gaseous===
TECHNICAL SPECIFICATIONS CONDITION                       REQUIRED ACTION                 COMPLETION TIME
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 (2) Releases of Ar-41 from the reactor bay exhaust plenum to an unrestricted environment 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 OPERA TE 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.
* Calculated releases of Ar-41 from the reactor bay exhaust plenum exceed 30 Ci per year.
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 (4/17) 
3.5.5   Bases Do not operate.                        IMMEDIATE 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- 1 (2.0lxl0-5 &#xb5;Ci ml- 1), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10-3 &#xb5;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.
* *
* 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-1 (2.0lxl0-5 &#xb5;Ci ml-1), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10-3 &#xb5;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.
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. 90 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.
Section 13.2.4 addresses the maximum hypothetical fission product inventory release. Using
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 (4/17)
* 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. 90 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 .
* TECHNICAL SPECIFICATIONS
* K-State Reactor                                 TS-21                               Original (4/17)


===3.6 Limitations===
TECHNICAL SPECIFICATIONS
* 3.6 3 .6.1 Limitations on Experiments 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 (3) reactor pool and core area
* 3.6.4    Actions CONDITION                        REQUIRED ACTION A.1 ENSURE the reactor is SHUTDOWN COMPLETION TIME A.1 IMMEDIATE A. INDEPENDENT EXPERIMENT worth is                              AND greater than $2.00 A.2 Remove the experiment              A.2 Prior to continued operations C.1    ENSURE the reactor is            C.1 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
* K-State Reactor                                    TS-22                                Original (4/17)


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 (1) If all fuel elements are stainless steel clad, the reactivity worth of any individual 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. (3) 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.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.1 ENSURE the reactor is C.1 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 K-State Reactor TS-22 Original (4/17)
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.
* TECHNICAL SPECIFICATIONS
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 .
* K-State Reactor                                 TS-23                                  Original (4/17)


====3.6.5 Bases====
TECHNICAL SPECIFICATIONS
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.
* 3. 7 Fuel Integrity
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 . K-State Reactor TS-23 Original (4/17) 
: 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 TRI GA reactor.
* TECHNICAL SPECIFICATIONS
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 Fuel Integrity
3.7.4     Actions CONDITION                       REQUIRED ACTION                 COMPLETION TIME
: 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 TRI GA reactor. 3.7.3 Specifications (1) Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over 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 the upper core grid plate. 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.
* A. Any fuel element is elongated greater than 1/8 in. over manufactured length, or bent laterally greater than 1/8 in.
Fuel cladding integrity is important since it represents the only process barrier for fission product release from the TRI GA reactor . K-State Reactor TS-24 Original (4/17)
Do not insert the fuel element into the upper core grid plate.
* *
IMMEDIATE 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 TRI GA reactor .
* TECHNICAL SPECIFICATIONS
* K-State Reactor                                 TS-24                               Original (4/17)


===3.8 Reactor===
TECHNICAL SPECIFICATIONS
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) Bulk water temperature SHALL NOT exceed 44 &deg;C (111&deg;F) (2) Water conductivity SHALL be less than 5 &#xb5;mho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core (4) Bulk water temperature SHALL NOT exceed 37&deg;C (99&deg;F) with an experiment installed in an interstitial flux wire port. 3.8.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.l ENSURE the reactor is A.1 IMMEDIATE SHUTDOWN AND A. Bulk water temperature exceeds 44 &deg;C A.2 Reduce bulk water temperature to less than 44 &deg;C A.3 IMMEDIATE K-State Reactor TS-25 Original (4/17)
* 3.8 3 .8.1 Reactor Pool Water 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.
* TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME B.1 ENSURE the reactor is B.l IMMEDIATE SHUTDOWN AND B. Bulk water temperature exceeds 3 7&deg;C with an B .2 Reduce bulk water temperature B.2 IMMEDIATE experiment installed in an to less than 3 7&deg;C. interstitial flux wire port. OR B.3 Remove experiment from flux B.3 IMMEDIATE wire port B.1 ENSURE the reactor is B.1 IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 &#xb5;mho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 &#xb5;mho/cm C. l 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
3.8.3     Specifications (1)     Bulk water temperature SHALL NOT exceed 44 &deg;C (111&deg;F)
(2)     Water conductivity SHALL be less than 5 &#xb5;mho/cm (3)     Water level above the core SHALL be at least 13 ft from the top of the core
*  (4) 3.8.4 Bulk water temperature SHALL NOT exceed 37&deg;C (99&deg;F) with an experiment installed in an interstitial flux wire port.
Actions CONDITION                         REQUIRED ACTION               COMPLETION TIME A.l ENSURE the reactor is           A.1 IMMEDIATE SHUTDOWN AND A. Bulk water temperature exceeds 44 &deg;C A.2 Reduce bulk water temperature to less than 44 &deg;C A.3 IMMEDIATE
* K-State Reactor                                 TS-25                             Original (4/17)


====3.8.5 Bases====
TECHNICAL SPECIFICATIONS CONDITION                        REQUIRED ACTION                COMPLETION TIME
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&deg;F (54.4&deg;C).
* B. Bulk water temperature exceeds 3 7&deg;C with an B.1 ENSURE the reactor is SHUTDOWN AND B .2 Reduce bulk water temperature B.l IMMEDIATE B.2 IMMEDIATE experiment installed in an          to less than 3 7&deg;C.
While the integrity of the ion exchange resin requires water temperature to remain below 54.4 &deg;C, it is necessary to maintain water temperature below 44 &deg;C to ensure that the departure from nucleate boiling ratio (DNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an experiment into an interstitial flux wire port between fuel elements necessitates a further reduction in water temperature to a maximum of 37&deg;C in order to preclude excessive nucleate boiling of the water. Maintaining low water conductivity over a prolonged period prevents possible corros10n, deionizer degradation, or slow leakage of fission products from degraded cladding.
interstitial flux wire port.
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.
OR B.3 Remove experiment from flux      B.3 IMMEDIATE wire port B.1 ENSURE the reactor is              B.1 IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 &#xb5;mho/cm B.2 Restore conductivity to less        B.2 Within 4 weeks than 5 &#xb5;mho/cm C. l 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
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.
* 3.8.5     Bases C.2 Restore water level                C.2 IMMEDIATE 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&deg;F (54.4&deg;C). While the integrity of the ion exchange resin requires water temperature to remain below 54.4 &deg;C, it is necessary to maintain water temperature below 44 &deg;C to ensure that the departure from nucleate boiling ratio (DNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an experiment into an interstitial flux wire port between fuel elements necessitates a further reduction in water temperature to a maximum of 37&deg;C in order to preclude excessive nucleate boiling of the water.
K-State Reactor TS-26 Original (4/17)
Maintaining low water conductivity over a prolonged period prevents possible corros10n, 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).
* TECHNICAL SPECIFICATIONS 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.
The principle contributor to radiation dose rates at the pool surface is Nitrogen 16
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 . K-State Reactor TS-27 Original (4/17) 
* 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.
* *
K-State Reactor                                 TS-26                               Original (4/17)
* TECHNICAL SPECIFICATIONS


===3.9 Maintenance===
TECHNICAL SPECIFICATIONS 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 .
* K-State Reactor                            TS-27                                Original (4/17)


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.  
TECHNICAL SPECIFICATIONS
* 3.9 3 .9 .1 Maintenance Retest Requirements 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 .
* K-State Reactor                                TS-28                              Original (4/17)


====3.9.3 Specifications====
TECHNICAL SPECIFICATIONS
* 3.10 Maximum Steady State Power 3 .9 .1  Applicability This specification applies to operations in STEADY STATE MODE.
3.9.2    Objective The objective is to ensure that the reactor has adequate margin to critical heat flux (CHF) and operates below the Limiting Safety System Setting of 1,250 kWth.
3.9.3    Specifications Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY (1)
STATE MODE.
(2)    A required reactor power level scram is set to a value no greater than 1,250 kWth.
3.9.4    Actions CONDITION                        REQUIRED ACTION                  COMPLETION TIME A. Thermal power exceeds 1,050 kWth in      Reduce power to a level no greater IMMEDIATE STEADY STATE                than 1,050 kWth.
MODE B. A required reactor          B.1 SHUT DOWN the reactor.
B.1. IMMEDIATE power level scram is set to a value above                        AND AND 1,250 kWth or above the maximum readable        B.2 Adjust reactor power level B.2. Prior to resuming value on a required        scram setpoint to a readable value operations.
channel.                    less than or equal to 1,250 kWth.
3.9.5    Bases The reactor control panel instrumentation is designed to measure up to 1,000 kWth of thermal power. The Limiting Safety System Setting ensures that automatic protective functions, i.e., high power scrams, are set to no greater than 1,250 kWth. However, by specifying the maximum OPERATING power level as 1,000 kWth in STEADY STATE MODE, the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum power readable on the reactor console instruments. Action to reduce power is not required until power exceeds 1050kWth in STEADY STATE MODE to allow for slight variation in power level that is typical during normal operation .
* K-State Reactor                                TS-29                                  Original (4/17)


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
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 4.1.3  Basis BIENNIAL 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 .
* K-State Reactor                                  TS-30                              Original (4/17)


====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 . K-State Reactor TS-28 Original (4/17)
TECHNICAL SPECIFICATIONS
TECHNICAL SPECIFICATIONS
* 3.10 Maximum Steady State Power *
* 4.2 4.2.1 PULSE MODE Objectives The verification that the pulse rod position does not exceed a reactivity value corresponding to
* 3 .9 .1 Applicability This specification applies to operations in STEADY STATE MODE. 3.9.2 Objective The objective is to ensure that the reactor has adequate margin to critical heat flux (CHF) and operates below the Limiting Safety System Setting of 1,250 kWth. 3.9.3 (1) (2) 3.9.4 A. B. Specifications Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY STATE MODE. A required reactor power level scram is set to a value no greater than 1,250 kWth. Actions CONDITION REQUIRED ACTION COMPLETION TIME Thermal power exceeds 1,050 kWth in Reduce power to a level no greater IMMEDIATE STEADY STATE than 1,050 kWth. MODE A required reactor B.1 SHUT DOWN the reactor. B.1. IMMEDIATE power level scram is set to a value above AND AND 1,250 kWth or above the maximum readable B.2 Adjust reactor power level B.2. Prior to resuming value on a required scram setpoint to a readable value operations.
  $3 .00 assures that the limiting condition for operation is met.
channel. less than or equal to 1,250 kWth. 3.9.5 Bases The reactor control panel instrumentation is designed to measure up to 1,000 kWth of thermal power. The Limiting Safety System Setting ensures that automatic protective functions, i.e., high power scrams, are set to no greater than 1,250 kWth. However, by specifying the maximum OPERATING power level as 1,000 kWth in STEADY STATE MODE, the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum power readable on the reactor console instruments.
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 .
Action to reduce power is not required until power exceeds 1050kWth in STEADY STATE MODE to allow for slight variation in power level that is typical during normal operation . K-State Reactor TS-29 Original (4/17)
* K-State Reactor                                 TS-31                              Original (4/17)
* *
* 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 SHUTDOWN MARGIN Determination Excess Reactivity Determination Control Rod Reactivity Worth determination


====4.1.3 Basis====
TECHNICAL SPECIFICATIONS
FREQUENCY SEMIANNUAL SEMIANNUAL Following Insertion of experiments with measurable positive reactivity BIENNIAL 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 . K-State Reactor TS-30 Original (4/17)
* 4.3 4.3. I MEASURING CHANNELS 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                                                I DAILY CHANNEL CALIBRATION                                            ANNUAL 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK                                                I 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 Jong-term drift of the channels to be corrected.
* TECHNICAL SPECIFICATIONS
K-State Reactor                               TS-32                          Original (4/17)


===4.2 PULSE===
TECHNICAL SPECIFICATIONS
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 ENSURE Transient Pulse Rod position corresponds to reactivity not greater than $3.00 4.2.3 Basis FREQUENCY Prior to pulsing operations Verifying pulse rod position corresponds to less than $3 .00 ensures that the maximum pulsed reactivity meets the limiting condition for operation . K-State Reactor TS-31 Original (4/17)
* 4.4 4.4.1 Safety Channel and Control Rod Operability 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)
* TECHNICAL SPECIFICATIONS
CONTROL ROD (STANDARD) drop times SHALL be measured to have a drop time from the fully withdrawn position of less than    ANNUAL 1 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-33                                Original (4/17)


===4.3 MEASURING===
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 .
* K-State Reactor                                TS-34                                Original (4/17)


CHANNELS 4.3. I Objectives Surveillances on MEASURING CHANNELS at specified frequencies ensure instrument problems are identified and corrected before they can affect operations.
TECHNICAL SPECIFICATIONS
4.3 .2 Specification SURVIELLANCE REQUIREMENTS SURVEILLANCE Reactor power level MEASURING CHANNEL CHANNEL TEST Calorimetric calibration CHANNEL CHECK high voltage to required power level instruments Primary pool water temperature CHANNEL CALIBRATION Reactor Bay differential pressure CHANNEL CALIBRATION Fuel temperature CHANNEL CALIBRATION 22 Foot Area radiation monitor CHANNEL CHECK CHANNEL CALIBRATION 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION Continuous Air Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION EXHAUST PLENUM Radiation Monitor CHANNEL CHECK CHANNEL CALIBRATION Startup Count Rate 4.3.3 Basis FREQUENCY DAILY ANNUAL DAILY ANNUAL ANNUAL ANNUAL I DAILY ANNUAL I DAILY ANNUAL DAILY ANNUAL DAILY ANNUAL DAILY The DAILY CHANNEL CHECKS will ensure that the SAFETY SYSTEM and MEASURING CHANNELS are operable. The required periodic calibrations and verifications will permit any Jong-term drift of the channels to be corrected.
* 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.
K-State Reactor TS-32 Original (4/17)
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.
* TECHNICAL SPECIFICATIONS
* K-State Reactor                                 TS-35                                Original (4/17)


===4.4 Safety===
TECHNICAL SPECIFICATIONS
Channel and Control Rod Operability
* 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 Measure and record experiment worth of the EXPERIMENT experiment where absolute (where the absolute value of the estimated worth is greater than value of the estimated
  $0.40).
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) .
* K-State Reactor                                TS-36                                  Original (4/17)


====4.4.1 Objective====
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 inse1tion of3.00$
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 .
* K-State Reactor                                TS-37                                Original (4/17)


The objectives of these surveillance requirements are to ensure the REACTOR SAFETY SYSTEM will function as required.
TECHNICAL SPECIFICATIONS
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)
* 4.8      Reactor Pool Water This specification applies to the water contained in the KSU TRIGA reactor pool.
CONTROL ROD (STANDARD) drop times SHALL be measured to have a drop time from the fully withdrawn position of less than ANNUAL 1 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.
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.
The CONTROL ROD (TRANSIENT) rod drive cylinder and the associated air supply system SHALL be inspected, cleaned, and SEMIANNUAL lubricated, as necessary.  
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 .
* K-State Reactor                                TS-38                              Original (4/17)


====4.4.3 Basis====
TECHNICAL SPECIFICATIONS
Manual and automatic scrams are not credited in accident analysis, although the systems function to assure long-term safe shutdown conditions.
* 4.9 4.9.1 Maintenance Retest Requirements Objective The objective is to ensure that a system is OPERABLE within specified limits before being used after maintenance has been performed.
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.
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.
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-33 Original (4/17)  
* K-State Reactor                               TS-39                              Original (4/17)
* *
* 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 . K-State Reactor TS-34 Original (4/17)
--------------------------------------------------
* *
* 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.
K-State Reactor TS-35 Original (4/17)
TECHNICAL SPECIFICATIONS
TECHNICAL SPECIFICATIONS
* 4.6 Limitations on Experiments
* 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.
* 4.6.1 Objectives This surveillance ensures that experiments do not have significant negative impact on safety of the public, personnel or the facility.
5.1.3   Specification (1) 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 .
====4.6.2 Specification====
(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 located in the following positions: E2, E4, ES, E6, E20, E21, E22, E24, Fl, F2, F30.
SURVIELLANCE REQUIREMENTS SURVEILLANCE Experiments SHALL be evaluated and approved prior to implementation.
5.1.4   Bases These types of fuel elements have a long history of successful use in TRIGA reactors.
Measure and record experiment worth of the EXPERIMENT (where the absolute value of the estimated worth is greater than $0.40). 4.6.3 Basis FREQUENCY Prior to inserting a new experiment for purposes other than determination of reactivity worth Initial insertion of a new experiment where absolute value of the estimated worth is greater than $0.40 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) . K-State Reactor TS-36 Original (4/17) 
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.
* *
5.2      Reactor Fuel and Fueled Devices in Storage
* TECHNICAL SPECIFICATIONS
* 5.2.l    Applicability K-State Reactor                                  TS-40                            Original (4/17)
: 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 inse1tion of3.00$ The standard fuel elements SHALL be visually inspected for cor-AND rosion and mechanical damage, and measured for length and bend B, C, D, E, and F RING elements comprising approximately 1/3 of the core SHALL be visually inspected annually for corrosion and mechanical damage such that the entire core SHALL be inspected at 3-year intervals, but not to exceed 38 months 4.7.4 Basis Following the exceeding of a limited safety system set point with potential for causing degradation ANNUAL 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 . K-State Reactor TS-37 Original (4/17) 
* *
* 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 . K-State Reactor TS-38 Original (4/17) 
* *
* 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 Evaluate potential for maintenance activities to affect operability and function of equipment required by Technical Specifications Perform surveillance to assure affected function meets requirements
 
====4.9.3 Bases====
FREQUENCY Following maintenance of systems of equipment required by Technical Specifications Prior to resumption of normal operations 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. K-State Reactor TS-39 Original (4/17) 
* *
* 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 (1) 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 located in the following positions: E2, E4, ES, E6, E20, E21, E22, E24, Fl, F2, F30. 5.1.4 Bases These types of fuel elements have a long history of successful use in TRI GA 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 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.  


===5.2 Reactor===
Fuel and Fueled Devices in Storage 5.2.l Applicability K-State Reactor TS-40 Original (4/17)
TECHNICAL SPECIFICATIONS This specification applies to reactor fuel elements in storage
TECHNICAL SPECIFICATIONS This specification applies to reactor fuel elements in storage
* 5.2.2 Objective  
* 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.
* The objective is to ensure fuel elements or fueled devices in storage are maintained Subcritical in a safe condition.  
5.2.4  Bases This specification is based on American Nuclear Society standard 15.1, section 5.4 .
* K-State Reactor                                  TS-41                                  Original (4/17)


====5.2.3 Specification====
TECHNICAL SPECIFICATIONS 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.
(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 3 0 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 .
* K-State Reactor                                TS-42                              Original (4/17)


(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 . K-State Reactor TS-41 Original (4/17)
TECHNICAL SPECIFICATIONS 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.
* TECHNICAL SPECIFICATIONS
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 13 1 through 13 5 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.
(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.
K-State Reactor                                 TS-43                                Original (4/17)


===5.3 Reactor===
TECHNICAL SPECIFICATIONS
Building 5 .3 .1 Applicability This specification applies to the building that houses the TRIGA reactor facility.  
: c. For materials whose boiling point is above 130&deg;F and where vapors formed by
*    (9) boiling this material could escape only through an undisturbed column of water above the core, at least 10% of these vapors will escape.
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 .
* K-State Reactor                              TS-44                              Original (4/17)


====5.3.2 Objective====
TECHNICAL SPECIFICATIONS
 
* 6. Administrative Controls 6.1    Organization and Responsibilities of Personnel a) Structure.
The objective is to ensure that provisions are made to restrict the amount of release of radioactivity into the environment.
The reactor organization is related to the University structure as shown in SAR Figure 12.1 and Technical Specifications Figure TS.1 below.
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. (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 3 0 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 . K-State Reactor TS-42 Original (4/17)
Kansas State University (KSU) holds the license for the KSU TRI GA 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.
* *
President Kansas State University Vice President for                              Provost Administration                          Chief Academic and Finance                                  Officer Director                                      Dean Division of Public Safety                    College of Engineering I                                                  I University        Department of                  Head, Department of Police          Environmental                  Mechanical & Nuclear Department      Health and Safety                    Engineering I
* TECHNICAL SPECIFICATIONS
University                        Manager, KSU Radiation Safety                    Nuclear Reactor Officer                            Facility I
 
Reactor I        Supervisor Reactor Safeguards Committee I
===5.4 Experiments===
Reactor Operators 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-45                                  Original (4/17)


5 .4.1 Applicability This specification applies to the design of experiments.
TECHNICAL SPECIFICATIONS by approved administrative controls (Reactor Radiation Protection Program or RPP,
5 .4.2 Objective The objective is to ensure that experiments are designed to meet criteria.
* 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.
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 13 1 through 13 5 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. (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. K-State Reactor TS-43 Original (4/17) 
* * * (9) 5.4.4 TECHNICAL SPECIFICATIONS
: c. For materials whose boiling point is above 130&deg;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. 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.
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 . K-State Reactor TS-44 Original (4/17) 
* *
* 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 TRI GA 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. President Kansas State University Vice President for Provost Administration Chief Academic and Finance Officer Director Dean Division of Public Safety College of Engineering I I University Department of Head, Department of Police Environmental Mechanical
& Nuclear Department Health and Safety Engineering I University Manager, KSU Radiation Safety Nuclear Reactor Officer Facility I Reactor I Supervisor Reactor Safeguards I Committee Reactor Operators 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-45 Original (4/17) 
* *
* 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.
b) Responsibility.
The President of the University shall be responsible for the appointment of responsible and competent persons as members of the TRlGA Reactor Safeguards Committee upon the recommendation of the ex officio Chairperson of the Committee.
The President of the University shall be responsible for the appointment of responsible and competent persons as members of the TRlGA 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 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.
The Manager also has responsibility for maintenance and modification of laboratories associated with the Reactor Facility.
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
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.
* 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 OPERA TING 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.
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 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 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 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 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 OPERA TING 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 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-46                               Original (4/17)
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-46 Original (4/17)
* *
* 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===
TECHNICAL SPECIFICATIONS Radiation Safety Officer. The University Radiation Safety Officer may, with the
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:
* 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;
: 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;
: 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;
: 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; K-State Reactor TS-47 Original (4/17)
: 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.
* TECHNICAL SPECIFICATIONS
: 5. Review of abnormal performance of plant equipment and OPERATING anomalies;
: 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:
: 6. Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR 20 and 10 CFR50;
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.
: 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-47                                 Original (4/17)
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-48 Original (4/17)
 
* *
TECHNICAL SPECIFICATIONS
* 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 1s 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.
: 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-48                               Original (4/17)
 
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 1s 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.
: 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 OPERA TING 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.
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 OPERA TING 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 1. 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.
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
The Rector Safeguards Committee shall subsequently review determinations that written procedures are not required.
: 1. the Nuclear Reactor Facility Manager, and
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.
: 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
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-49 Original (4/17)
: 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-49                             Original (4/17)
 
TECHNICAL SPECIFICATIONS
TECHNICAL SPECIFICATIONS
* 6.4 Review of Proposals for Experiments  
* 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:
* 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.               '
: 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.
: 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.
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.  
: 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.
' 2. If the experimenter is a student, approval by his/her research supervisor is required.
: 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 OPERA TING 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 CPR 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.
If the experimenter is a staff or faculty member, his/her own signature is sufficient.
: 5. Transmission to the Reactor Supervisor for scheduling.
: 3. The proposal is then to be submitted to the Reactor Safeguards Committee for consideration and approval.
c) Criteria that shall be met before approval can be granted shall include:
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 OPERA TING personnel.
: 1. The EXPERIMENT must meet the applicable Limiting Conditions for Operation and Design Description specifications.
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 CPR 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.
: 2. It must not involve violation of any condition of the facility license or of Federal, State, University, or Facility regulations and procedures .
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.
* 3. The conduct of tests or experiments not described in the safety analysis report (as updated) must be evaluated in accordance with 10 CPR 50.59 to determine if the test K-State Reactor                               TS-50                               Original (4/17)
: 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 CPR 50.59 to determine if the test K-State Reactor TS-50 Original (4/17)
* *
* 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===
TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license
Requalification An operator requalification program shall be established and followed in accordance with NRC regulations.
* amendment pursuant to 10 CFR Sec. 50.90 .
: 6. 7 Physical Security Plan Administrative controls for protection of the reactor plant shall be established and followed in accordance with NRC regulations.  
: 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:
K-State Reactor                                TS-51                                Original (4/17)


===6.8 Action===
TECHNICAL SPECIFICATIONS
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.
: 1. any actual safety system setting less conservative than specified in Section 2.2,
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.
* Limiting Safety System Settings;
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:
K-State Reactor TS-51 Original (4/17) 
* *
* 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 through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.
: 2. VIOLATION OF SL, LSSS OR LCO; NOTES Violation of an LSSS or LCO occurs through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.
Surveillance Requirements must be met for all equipment/components/conditions to be considered operable.
Surveillance Requirements must be met for all equipment/components/conditions to be considered operable.
Failure to perform a surveillance within the required time interval or failure of a surveillance test shall result in the /component/condition being inoperable
Failure to perform a surveillance within the required time 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 m the reactor bay to exceed 1 OCFR20 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.
: 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;
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.
: 4. release of fission products from the fuel that cause airborne contamination levels m the reactor bay to exceed 10CFR20 limits for releases to unrestricted areas;
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.
: 5. an uncontrolled or unanticipated change in reactivity greater than $1.00;
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-52 Original (4/17)
* 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.
* TECHNICAL SPECIFICATIONS
b) In the event of a reportable occurrence, the following actions shall be taken:
: 1. normal plant operation, including power levels; 3. principal maintenance activities;
: 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-52                             Original (4/17)
 
TECHNICAL SPECIFICATIONS
* 1. normal plant operation, including power levels;
: 3. principal maintenance activities;
: 4. reportable occurrences;
: 4. reportable occurrences;
: 5. equipment and component surveillance activities;
: 5. equipment and component surveillance activities;
: 6. experiments performed with the reactor; 7. all emergency reactor scrams, including reasons for emergency shutdowns.
: 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:
b) The following records shall be maintained for the life of the facility:
: 1. gaseous and liquid radioactive effluents released to the environs;
: 1. gaseous and liquid radioactive effluents released to the environs;
: 2. offsite environmental monitoring surveys; 3. fuel inventories and transfers;
: 2. offsite environmental monitoring surveys;
: 4. facility radiation and contamination surveys; 5. radiation exposures for all personnel;
: 3. fuel inventories and transfers;
: 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 hours 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;
: 4. facility radiation and contamination surveys;
: 2. any violation of a safety limit; 3. any reportable occurrences as defined in Section 6.9 of these specifications.
: 5. radiation exposures for all personnel;
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-53 Original (4/17)
* 6.11
* *
: 6. updated, corrected, and as-built drawings of the facility .
* 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.
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.
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 OPERA TING 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 OPERA TING conditions or characteristics of the reactor under the new conditions.
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 hours 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-53                             Original (4/17)
 
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 OPERA TING 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:
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 OPERA TING experience (including experiments performed), changes in facility design, performance characteristics, and OPERA TING procedures related to reactor safety occurring during the reporting period; and results of surveillance tests and inspections;
: 1. a brief narrative summary of OPERATING experience (including experiments performed), changes in facility design, performance characteristics, and OPERA TING 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);
: 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;
: 3. the number of emergency shutdowns and inadvertent scrams, including the reasons thereof and corrective action, if any, taken;
: 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;
: 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;
: 7. a description of any environmental surveys performed outside the facility; K-State Reactor TS-54 Original (4/17)
: 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;
* TECHNICAL SPECIFICATIONS
* 7. a description of any environmental surveys performed outside the facility; K-State Reactor                               TS-54                               Original (4/17)
: 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 ofradiation and contamination surveys performed within the facility . K-State Reactor TS-55 Original (4/17)}}
 
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 ofradiation and contamination surveys performed within the facility .
* K-State Reactor                           TS-55                               Original (4/17)}}

Latest revision as of 13:27, 24 February 2020

Kansas State Univ. - Facility Response to 3/28/17 Request for Additional Information Regarding License Amendment Request
ML17139C979
Person / Time
Site: Kansas State University
Issue date: 05/02/2017
From: Geuther J
Kansas State University
To:
Office of Nuclear Reactor Regulation
References
Download: ML17139C979 (211)
v • d • e


Text

{{#Wiki_filter:KANSAS STATE TRIGA Mk II Nuclear

  • US Nuclear Regulatory Commission Washington, DC 20555-0001 U N I V E R s 1 T y Reactor Laboratory 2 May 2017

Subject:

Facility Response to 3/28/17 Request for Additional Information (Acc.# ML17038A272) To Whom It May Concern, On March 28th, 2017, the NRC sent a third Request for Additional Information (RAI) to the Kansas State University nuclear reactor facility (license R-88, docket 50-188) regarding a license amendment request (LAR) to add up to four 12%-loaded fuel elements to the core. The original LAR was submitted on April 9, 2012 (Acc.# ML1219A063). The resolution of the question asked in the most recent RAI required several amendments to Chapter 4 of the facility Safety Analysis Report and to the facility Technical Specifications. These changes were reviewed and approved by the Reactor Safeguards Committee on 4/28/17 by unanimous vote, pending several minor editorial changes that have been made since the vote. The following constitutes the facility's response to the RAI, and is organized as a numbered list of abbreviated questiops in order of appearance in the RAI, followed by the facility response.

1. ... Please provide information that validates the pool boiling model used in the SAR or alternatively provide a traditional hot channel analysis using the Bernath correlation ... with core inlet conditions at the TS limit, or another correlation that has been validated against acceptable data.

A RELAP-5 single-channel model was used to update the SAR analysis for a fuel element heated to 24 kWth in water at a pressure of 1.43 kPa, corresponding to the depth of the fuel in the reactor pool. 24 kWth is the power in a fuel element with an 85-element core operating at 1.25 MW and an element-to-average power peaking value of 1.63. The bulk water temperature used in the analysis was 49°C. This value is slightly lower than the maximum of 54°C specified in the Technical Specifications. However, based on this analysis, the TS will be revised to permit a maximum bulk water temperature of 44°C in order to avoid bulk boiling of the coolant. The departure from nucleate boiling ratio (DNBR) reached a minimum value of slightly above 2.0. This is significantly lower than the DNBR calculated in the current version of the SAR. The reactor core grid plates contain an array of 8 mm interstitial holes meant to accommodate flux wires or other experiments. The presence of an experiment in these holes may reduce the temperature at which bulk boiling will occur; therefore an additional experimental design constraint is being proposed for the TS that forbids the insertion of tubes or flux wires in the interstitial holes at bulk water temperatures

  • greater than 37°C. A drawing of the upper grid plate showing the interstitial holes is attached for reference.

Inlet 1~- Cold*Leg connector Figure 1 - RELAP single channel model 

                                          .. .. .. .. .. .. ..                                          .... Bernath-CHF
                                                                .... .. .... ,                          .... PG -CHF 4000
                                                                              ..                        *
  • Heat flux
                            "'c:  3000
                                                                                                                  * .i. .~. . .... . ..
                             -lo!

LL.

c 2000 u

1000 

                                                             * ~ * * ..

8.Loo___o....o_s___o...1-o--o-.1'-s- --o.....2_0--o-.""2s___o_..3,..,o--o.,...~3.,..s__. Heated Length (m) Figu re 2 - Critical heat flux (CHF) versus heated length

  • 2
  • 12
                                   + +
                                   + +

Bernath-CHF PG-CHF 

                                                                                                                      ... * =

10 

                                                 **+                                                       **

B .... . *** ****

  • 0::

CXl z * *~ 0 6 **  :

+ :*

4 2 

                                                          . !~*********! : .

0.05 0.10 0.15 0.20 0.25 0.30 0.35 Heated Length (m ) Figure 3 - Departure from nucleate bailing ratio (DNBR) versus heated length

  • 2 . ... Propose a licensed thermal power limit that is within the range of the currently installed nuclear instrumentation or describe how the nuclear instrumentation system is capable of measurement of the full range of reactor power levels anticipated as described in the safety analyses including instrument uncertainties based on the current licensed thermal power limit.

The three nuclear instruments at KSU have maximum readings of: 1.0 MW (NLW-1000, log channel), 1.20 MW {NMP-1000, multi-range linear channel) and 1.10 MW {NPP-1000, percent power I pulse channel). Two of these are required by the Limiting Conditions of Operation during steady state mode, one is required during pulse mode {TS 3.3, attached). Therefore readings at or slightly above 1.0 MW would be readable on the console instruments, but the console instruments are incapable of reading up to the license limit of 1.25 MW. In order to address this issue, the revised Technical Specifications have been amended as follows. The Limiting Safety System Setting has been left unchanged at 1.25 MW. However, an additional Limiting Condition of Operation has been added which sets the maximum operating power to 1.00 MW, with a requirement to reduce power if it exceeds 1.05 MW. In this way the facility will still have some margin to the LSSS if, for example, it is determined that the power channels were reading low due to detector uncertainty or flux shifting following power calibration. However, the maximum power will not be allowed to exceed the useful range of the control panel instrumentation. The new LCO is found in the attached Technical Specifications, section 3.10. Note that the upgraded control console instrumentation planned for installation in January 2018 is capable of reading 1.25 MW of power .

  • 3
  • 3. ... Provide an explanation resolving the significant deviation between the reactor power observed on June 27, 2016 {735 kWt} and the maximum reactor power predicted in support of this license amendment {600 kWt} and provide a revised analysis for the maximum core operating power after the addition of the four 12.5 U wt% fuel elements.

The reason for the deviation between the power observed on 6/27 /2016 (735 kWth) and the predicted maximum power of 600 kWth in the original license amendment proposal is primarily due to the length of time involved in the LAR process. In 2012, when the LAR was originally submitted, the most recent operation at near full power had been 3/28/12, when the reactor operated at 550 kW with the three most reactive rods fully-withdrawn, and the regulating rod withdrawn to a position corresponding to only 0.04$ remaining. The reactor power is limited by core reactivity, so during the annual fuel inspections in the following years, several old fuel elements were replaced with fresh elements, resulting in an increase in the available excess reactivity. Based on the increase in power level versus when the original LAR was submitted, the maximum power level would be expected to be slightly greater than 1 MW (i.e., 1.035 MW).

  • 4 . ... Provide clarification on how the MCNP calculations are integrated with the facility measurements to determine the maximum excess reactivity and the minimum shutdown margin.

In order to reduce the error on the estimates for maximum excess reactivity and minimum shutdown margin, MCNP was used to determine only the change in reactivity due to loading 12% fuel, not the total reactivity. Measured values of excess reactivity were taken from the daily reactivity balance, in which the calibrated integral control rod worth curves are used to determine how much excess reactivity remains above the critical configuration at zero power with no xenon. The estimated values were therefore calculated as follows: max. reactivity (estimated)= max. reactivity (measured)+ 8p (MCNP) min. shutdown margin (estimated)= min. shutdown margin (measured) - 8p (MCNP)

  • 4
  • 5 .... If operation with one or more control rods inoperable but fully inserted is acceptable, please provide the supporting analyses and evaluations from which operational acceptability is derived.

The operation of the reactor core with one or more control rods inoperable but fully inserted was judged to be safe due to the understanding that the safety function of control rods is to provide a means to control reactivity, i.e., if a control rod was inoperable but inserted, then it was fulfilling its safety function. The response to the previous RAI did not consider the changes in power peaking due to the insertion of a control rod. In order to address this additional consideration, a reactor model was created in MCNP6 in which the fuel elements were all 9.0 wt%-loaded elements. The fuel elements were all modeled as having the highest allowable density of uranium, i.e, fresh fuel at 9.0 wt% U. The maximum element-to-core-average power peaking was calculated using fission heating cell tallies, and the results were compared between cases where all rods were fully withdrawn, versus having individual rods failed but fully inserted. Table 1 shows a summary of the results. The power peaking in most cases is B4, which is the fuel element located between the pulse rod channel and central thimble, such that this location is heavily moderated when the pulse rod is fully withdrawn. When the pulse (i.e., transient) rod is inoperable and fully inserted, the maximum power peaking would occur in fuel element Cll. In some cases the power peaking is greater than in the all-rods-out case, but no case approached the SAR-assumed 2.00 power peaking factor used for PCT analysis. Rod Inserted Peak Power Factor Element None 1.558 B4 Regulating Rod 1.617 B4 Shim Rod 1.630 B4 Safety Rod 1.609 B4 Pulse Rod 1.438 Cll SAR 2.00 N/A Table 1: Peak Power Factors Using recent measurements of integral control rod worth, the peak element power was calculated for an all-rods-out case at the current license power limit of 1250 kW for a fully-loaded core, and then for a case with each of the four control rods fully inserted. For these cases, the power of the core was reduced based on the current set of integral control rod worth curves and an assumed power coefficient of -0.0045 $I kW, which is similar to the value of-0.0038$ /kw calculated by dividing the recent peak power of 735 kW by the present-day 2.77$ of excess reactivity. The maximum power per element for this set of models is given in Table 2. Rod Inserted Approximate Reactor Average Power Per Peak Element Power (kW} Element (kW) Power (kW} Regulating Rod (max, IRW 882 10.38 16.78 

   =0.87$)

Shim Rod (max, IRW = 738 8.68 14.15 2.13$)

  • Safety Rod (max, IRW 1.68$)
                              =       758 5

8.91 14.35

  • Rod Inserted Pulse Rod (max, IRW =

2.77$) 1250 kW, all rods out Approximate Reactor Power (kW) 685 1250 Average Power Per Element (kW) 7.74 14.71 Peak Element Power (kW) 11.13 22.91 Table 2: Power comparison between different core configurations, no 12% fuel installed. A similar study was conducted using a peak power of 1250 kWth with all rods withdrawn, but with a single 12.3%-weighted fuel element in location E-10. This study confirms that the maximum power in a single fuel element remains well below the SAR-assumed value of 2.0 element-to-average, even with a high-load fuel element on the opposite side of the core relative to the inserted control rod. The results are depicted in Figure 4 .

    • 6
  • Legend Down Rod; None.

Rea ctor Power: 1250.0 w. Peak Elem ent Power: 23.05kW Pow r P r Element (kW) ll*U JJO') Down Rod: Regula ting. Reactor Power: 882 .0 W. Pea Elem nt Pow r; 16.8 7kW Down Rod ; Safety . Down Rod: Pulse. Rea ctor Power: 758.0 w. React or Po wer; 658.0 W, Pea Element Power: 14.67 W Peak Element Pow er : l l.34kW Figure 4 -Power peoking map with different control rods stuck in, with o 12% element in location E-10 and a maximum core power of 1.25 MW. Therefore the case of an inoperable control rod being fully inserted may slightly increase maximum power peaking but will greatly reduce the maximum power per element. The slight increase in power peaking will also be more than offset by the reduction in core reactivity due to the insertion of the rod poison, resulting in a massive decrease to the maximum fuel element power. The proposed revision to the TS includes a specification of minimum of 3 operable control rods, and states that inoperable control rods must be fully inserted . (See attached, TS 3.4) .

  • 7
  • 6 .... "Provide a revision to the proposed TS describing the geometric limitation and the controls that will help ensure compliance and include information on the acceptability of the specific location where the 12.5% fuel elements will be installed. Otherwise, describe how the current or previously proposed TS adequately address the control of this geometric limitation on the location [of] 12.5 wt% fuel elements in the proposed reactor core."

MCNP results transmitted in the initial LAR and in previous RAI responses indicate that 12.5 wt% (max.) fuel in the E- and F-rings will be kept below the fission heating density (i.e., power) of the 9.0 wt% (max.) elements in the B-ring, even if placed adjacent to another 12.5%-loaded element. The additional restriction to avoided placing the 12.5% elements adjacent to control rod channels was intended to avoid conflict with the NRC over local power peaking during pulsing, although results from MCNP calculations performed at KSU have not shown reason for concern. The proposed TS have been revised to include a specific list of locations in which 12.5% fuel cannot be located. (See attached). In order to ensure that the 12% fuel is only added to locations where it permitted, the fuel handling procedures will be revised to include guidance to check the Technical Specifications list of approved locations prior to loading 12% fuel. I swear under penalty of perjury that the foregoing is true and correct.

  • Regards, Nuclear Reactor Facilities Manager Department of Mechanical and Nuclear Engineering Kansas State University Manhattan, KS 66506 Phone: {785)532-6657 {W)

{785)236-0602 (C) Fax: {785)532-7057 Email: geuther@ksu.edu Attachments (5): GA Drawing TOS21E106 -Top Grid Plate SAR Ch. 4 markup copy SAR Ch. 4 clean copy TS markup copy TS clean copy 8 

 ,- .391 D!A. (2.H.QLES)     i

! l 

                         - 'L
4. Reactor Description 4.1 Summary Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.

The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations. Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems. The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:

  • Central thimble
  • Rotary specimen rack
  • Thermalizing column with bulk shielding tank
  • Thermal column with removable door
                *Beam ports Radial (2)

Piercing (fast neutron) (1) Tangential (thermal neutron) (1) The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated graphite. A fourth control rod would be used for 1,250-kW operation. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m3 (144,000 ft. 3) confinement building K-State Reactor 4-1 Original (12/04) Safety Analysis Report

  • CHAPTER 4 made of reinforced concrete and structural steel, with composite sheathing and aluminum siding.

Sectional views of the reactor are shown in Figures 4. 1 and 4.2. Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW. The original aluminum-clad fue l elements were replaced with stainless-steel clad elements in 1973 . Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in I 993, with support from the U.S. Department of Energy. All neutronic instrumentation was replaced in I 994. North South Figure 4. 1, Vertical Section Through the KSU TRJGA Reactor. K-State Reactor 4-2 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION 4.2 Reactor Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident. Inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fuel in a zirconi um hydride matrix .
                                                                             * *"" ! IT ~
                                                              ..... .-.- :-.""                    IP!
                                                                      **: ,"       ~
                                                                        .*..*... .     .___...,...-.-.,-~
*:~ ~ T~ 2.!* '". :
                                                                      *.*.:*. :* . : .:\*.
                                                                                       ~*

I fT 8 ~"I 

                                                                                                              " (hi t), 0 q "0v.Lf' t 'Pu.>'i West                                                                                                            East Figure 4. 2, Horizontal Section Through the KSU TRJGA Reactor.

The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively constant with temperature (-0.0 1% iik/k°C). This coefficient is a function of the fuel composition and core geometry. As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase. Therefore, fuel and clad temperature automatically limit operation of the reactor. K-State Reactor 4-3 Original (12/04) Safety Analysis Report

  • CHAPTER4 It is more convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120°F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response.

4.2.1 Reactor Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic. Additional advantages were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size. TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.l (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material. The maximum temperature limits of l 150°C (with clad< 500°C) and 950°C (with clad> 500°C) for U-ZrH (H/Zri.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.

a. Dimensions and Physical Properties.

The KSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements. Three instrumented aluminum-clad Mark II elements are still available for use in the core. General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.) hole drilled through the center of the active fuel section. Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (I in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank. 1 Unless otherwise indicated, fuel properties are taken from the General Atomics report of Simnad [1980] and from authorities cited by Simnad. K-State Reactor 4-4 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of 55.88 cm (22 in.).

Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility. 

 'Property                                           Mark II                 Mark Ill Dimensions Outside diameter, Do= 2ro                         1.47 in. (3.7338 cm)    1.47 in. (3.7338 cm)

Inside diameter, D;= 2r; 1.41 in (3.6322 cm) 1.43 in. (3.6322 cm) Overall length 28.4 in. (72.136 cm) 28.4 in. (72.136 cm) Length of fuel zone, L 14 in. (35.56 cm) 15 in. (38.10 cm) Length of graphite axial reflectors 4 in. (10.16 cm) 3.44 in (8.738 cm) End fixtures and cladding aluminum 3 04 stainless steel Cladding thickness 0.030 in. (0.0762 cm) 0.020 in. (0.0508 cm) Burnable poisons Sm wafers None Uranium content Weight percent U 8.0 8.5 mu enrichment percent 20 20 mu content 36 g 38 g Physical properties offuel excluding cladding H/Zr atomic ratio 1.0 1.6 Thermal conductivity (W cm- 1 K- 1) 0.18 0.18 Heat capacity [T ~0°C] (J cm-3 K- 1) 2.04 + 0.00417T Mechanical properties ofdelta phase U-ZrH0 Elastic modulus at 20°C 9.1 x 106 psi Elastic modulus at 650°C 6.0 x 106 psi Ultimate tensile strength (to 650°C) 24,000 psi Compressive strength (20°C) 60,000 psi Compressive yield (20°C) 35,000 psi 

  *source: Texas SAR [1991].
b. Composition and Phase Properties The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% mu, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760°C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen. Over 25,000 pulses have been performed with the TRIGA fuel elements at General Atomic, with fuel temperatures reaching peaks of about l 150°C.

K-State Reactor 4-5 Original (12/04) Safety Analysis Report

  • CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455°C. From 455°C to about 1050°C, the delta phase is supported by a broadening range of H/Zr ratios.
                                             .fr       STAINLESS STEEL TOP END-FIXTURE
                                             ~--~-=~.~

20MIL STAINLESS STEEL CLAD l ZIRCONIUM HYDRIDE-2837" 8*5WT% -~ URANIUM, l 20%ENR., j I 38 235u t 

                                                                      -~

15" I \' l 1:43"DIA .i; I,.,. 0,, .!' I STAINLESS STEEL 

                                            ~

BOTTOM ENO-FIXTURE 1 Figure 4.3, TRIGA Fuel Element. K-State Reactor 4-6 Orig in al ( 12/04) Safety Analysis Report

  • REACTOR DESCRIPTION
c. Core Layout A typical layout for a KSU TRIGA Il 250-kW core (Core II-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted ..

Figure 4.4, Core Layout (250 kW). The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor 4-7 Original (12/04) Safety Analysis Report

  • CHAPTER 4 4.2.2 Control Rods The pulse rod is 3. 175 cm. ( 1.25 in.) diameter. Other rods are 2.225 cm (7/8 in.) diameter.

Control rods are 50.8 cm . (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm . (30*m il ) aluminum sheath. The control rod drives are connected to the control rod clutches through three extension shafts. The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydrau lic cu hi on (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (s lotted tubes on the right hand side) along wi th the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder; the dimensions of the cylinder limits bottom trave l. Upper Pulse, Shim & Reg Rods Reg Rod Shim Rod Pul se Rod Figure 4.5, Control Rod Upper Extension Assembli es The bottom of the pulse rod is shown on the left hand side of Figure 4.5 . The upper extension shaft is a hollow tube, the midd le extension is solid. The upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fasten er). Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full in position is above the bottom of the gu ide tube. K-State Reactor 4-8 Original (12/04 ) Safety Analysis Report

  • REACTOR DESCRIPTION Figure 4.6, Middle Extension Rod Alignment Holes The middle solid extension is similarly connected to the lower extension. The lower extension is holl ow, the middle extension fits into the lower extension and a hole drill ed in the overl ap secures the lower extension to the middle extension. Typically th e lower extension has a tighter fit than the upper extension because the lower and middle extension are not separated for inspecti ons and because the interface wi th upper extension is used to set the bottom position of the control rod.

Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7 ..

  • Figure 4.7, Standard & Pulse Rod Lower Coupling The bottom of the lower extension attaches directly to the control rod . Pictures of the control rods taken duri ng the 2003 control rod inspecti on are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wi re in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube.

K-State Reactor 4-9 Original (12/04) Safety Analysis Report

  • CHAPTER 4 Pulse Rod Shim Rod Reg Rod Figure 4.8, Control Rods During 2003 Inspection
  • Fu ll Gu ide Tube Figure 4.9, Control Rod Guide Tubes K-State Reactor 4-10 Origina l (12/04)

Safety Analysis Report

  • REACTOR DESCRIPTION
a. Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings. A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (I) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.

Table 4.2, Summary ore ontrol Rod Interlocks INTERLOCK SETPOINT FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument startup charmel reading is less Source Interlock 2 cps than instrument sensitivity/ensure nuclear instrument startup charmel is operating Prevent applying power to pulse rod unless Pulse Rod Interlock Pulse rod inserted rod inserted/prevent inadvertent pulse Withdraw signal, Prevent withdrawal of more than 1 rod/Limit Multiple Rod Withdrawal more than 1 rod maximum reactivity addition rate Mode switch in Hi Prevent withdrawing standard control rods in Pulse Mode Interlock Pulse pulse mode Prevent pulsing if power level is greater than Pulse-Power Interlock !OkW 

                                                                   !OkW NOTE: (!)Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock reqmred by Technical Specifications
b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation.

K-State Reactor 4-11 Original (12/04) Safety Analysis Report

  • CHAPTER4 Table 4.3, Summary of Reactor CRAMss Limitin2 Trip Setpoint Measuring Steady Actual Setpoint Channel Pulse State Linear Channel High 110% NIA 104%

Power Power Channel High power 110% NIA 104% Detector High 90% 90% 90% Voltage 600°C B Ring element High Fuel 555°C C Ring element 450°C Temperaturef1l 480°C D Ring element 380°C E Ring element 350°C Period [IJ NIA NIA 3 sec NOTE [l]: Penod tnp and temperature tnp are not reqmred by Techmcal Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin of at least $0.50 . Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications. 4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors. The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3.18-cm (1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector. K-State Reactor 4-12 Original (12104) Safety Analysis Report

  • REACTOR DESCRIPTION The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank.

Four lugs are provided for lifting the assembly. A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within the fuel cladding provide additional reflection. Inserts are placed at both ends of the fuel meat, providing top and bottom reflection. 4.2.4 Neutron Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 106 n s- 1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor. 4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates. The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core. Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position. The reflector assembly supports both grid plates. 4.3 Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion. Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration. All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool K-State Reactor 4-13 Original (12/04) Safety Analysis Report

  • CHAPTER4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak.

A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations. Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter 10). 4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield. The shielding configuration is similar to those at other TRI GA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank . The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena. 4.5 Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels with 8.5% U, 20% enrichment, the value is nearly constant at 0.01 % L'i.k/k per °C, and varies only weakly dependent on geometry and temperature. Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures that the fuel and clad temperature limits will not be exceeded. The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity. Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor 4-14 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6.

4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures. 4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates. For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor). Table 4.4, 250 kW Core Parameters. 13 (effective delayed neutron fraction) 0.007 

                £ (~ffective neutron lifetime)                                   43 :S
                                                                             -$0.0I7 EC- 1 UTf (prompt temperature coefficient)                    (a'J 250kW -275EC av (void coefficient)                                     -0.003 I %- 1 void
                                                                         -$0.006 kW- 1 to -

Up (power temperature coefficient -weighted ave) $0.0I kw- 1 Table 4.5, Com arison of Control Rod Worths. KSU TRIGA Mark II (250 kW) Cornell University Core II-19 Core III-I (500 kW Pulse D-10 $1.96 C-4 $2.I2 D-10 $1.88 Shim C-3 $2.88 D-4 $1.85 D-I6 $2.20 Safety NA $0.0 D-16 $1.82 D-4 $1.99 Regulating D-16 $1.58 E-I $0.79 E-I $0.58 TOTAL NA $6.42 NA $6.58 NA $6.65 NOTE: Core III-I has an experiment positioned to control the worth of the pulse rod K-State Reactor 4-15 Original (12/04) Safety Analysis Report

  • CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.

4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures. Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRI GA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 1000°C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470°C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr* 1 ft* 2 0 1 P- (2550 W m*2 K" 1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.

a. Spatial Power Distribution The following conservative approximations are made m characterizing the spatial distribution of the power during steady-state operations.
  • The hottest fuel element delivers twice the power of the average.

Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and R: as the physical radius and the extrapolation distance) is given 2 by: F" = 1.202 * (%.) 

                                                  .. J, [ 2.4048 *(%.)J with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,. However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element.

2 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3 K-State Reactor 4-16 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40.

Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.

  • The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of rr/2 times the average, and heat conduction radial only.

The axial factor for power produced within a fuel element is given by: g(z)=l.514*co(%* */ C ), (6) 2 

                                                    '           +  ext in which £=LI 2 and£,,, is the extrapolation length in graphite, namely, 0.0275
m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor.

Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.

  • The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore:

(7)

  • The radial and axial distribution of the power within a fuel element is given by q"'(r,z) = q;~J(r)g(z), (5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core.

3 GA-4361, Calculated Fluxes and Cross Sections for TRIG A Reactors (8/14/J 963), G. B. West K-State Reactor 4-17 Original (12/04) Safety Analysis Report

  • CHAPTER4
  • The radial factor is given by:

2 f(r)= a+cr+er , (7) I +br+dr 2 in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is illustrated in Figure 4.11. 1.3 1.2 1.1 L 1.0 0.90 0.80 0.70 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0 r (cm) Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod. (Data Points from Monte Carlo Calculations [Ahrens 1999a])

a. Heat Transfer Models The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the LI.T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron),

cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ncl): Eq. 1 A standard heat resistance model for this system is: K-State Reactor 4-18 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION T 0
                                 =T
                                       +q"l_!_+ In(;{) +l+lj h

r, k c rh 

                                                                            ' g 2kf Eq. 2 and heat flux is calculated directly as:
                                "=Ul1T=                  T--'I'i,                                    (2) q              1 r ln(r I r)
                                           -+     0      0 I +-+-O-r0 r '

h kc rA 2k1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k1 is the fuel thermal conductivity. The gap conductivity of 2840 W m-2 K" 1 (500 Btu h- 1 ft -2 °F- 1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context. Parameters are cross-referenced to source in Table 4.6. Table 4.6: Thermodynamic Values Parameter Symbol Value Units Reference Fuel conductivity kr 18 Wm- 1 K- 1 Table 13.3 14.9 W m* 1 K- 1 (300 K) Table 13.3 16.6 W m- 1 K- 1 (400 K) Table 13.3 Clad conductivity kg 19.8 W m* 1 K" 1 (600 K) Table 13.3 Gao resistance h. 2840 wm-2 K- 1 Appendix A Clad outer radius ro 0.018161 M Table 13.1 Fuel outer radius fj 0.018669 M Table 13.1 Active fuel length Lr 0.381 M Table 13.1 No. fuel elements N 83 NIA Chap 13 Axial peaking factor APF nl2 NIA Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that: l = 5.1858E-04 m'K 2kf vv Gap resistance has been experimentally determined as indicated, so that: l=3.6196E-04 m'K r,h, w K-State Reactor 4-19 Original (12/04) Safety Analysis Report

  • CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127°C applies so that:

r0 In':!._ r m'K 

                                                - - ' =3.103e                                                      k,                         w Table 4.7, Cladding Heat Transfer Coefficient Temp (°K)            Temp (0 C)                m2 Kw* 1 300                    27                  3.457e-5 400                   127                  3.103e-5 600                   327                 2.60le-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small.

The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function. Convection Heat Transfer Coefficient TRENCLINE: y = 0.0326x + 16985 R2 =09976 85000 f 75000 E 

             ~
             ~ 65000
              "~
             ~     55000
            "~     45000
             /:!."

m J: 35000 25000 500 700 900 1100 1300 1500 1700 19'.lO Power Level (KW) Figure 4.10, Convection Hear Transfer Coefficient versus Power Level 1 h 0.0326P(watts) + 16985 K-State Reactor 4-20 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as:

1 T =T +0.423P[ +3.103e-5+3.620e-4+5.186e-4] (10) 

          <,   '             0.0326P + 16985
c. Steady-State Mode of Operation Centerline t+emperature calculations were performed on a "reference core" using t~

model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation. A core with more than 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 2occ and I oocc was assumed with the following results: Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel Fuel/Gap Gap/Clad Clad/Water Bulk Water cc' Centerline cc Interface cc Interface cc Interface cc 503.2 229.0 37.7 21.2 20.0 582.0 307.8 116.4 100.0 100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 cc at power levels up to 1900 kW with pool temperature at 27 cc (300 K), and 1700 kW with pool temperatures at 100 cc. K-State Reactor 4-21 Original (12/04) Safety Analysis Report

  • CHAPTER 4 Hot Fuel-Rod Centerline Temperature at Power (Temperature Bevation over Pool Water Temperature) l----*300K --SOOK I o+tt1::t::tt:l:tt1:ttt1::t::tt:l:tl:1::t::tt:t:ttl:tµ1::t::ttjjjjt:t:tttllitttj::tttjjjj::ttj:tlli:tttt::tj::t::t:ttl:tttt:j:tt:t:ttl:tttj::ttt~

100 300 500 700 900 1100 1300 1500 1700 1900 Reactor Power (kW) Figure 4.11, Hot Fuel-Rod Centerline Temperature The margiH te eritisal heat fhm fer the refereHee eere v*as determiHed. Critieal heat flHx fer sat\lfated peel beiliHg is giveH by (Heat TFansfer, A. BejaH, 1993, Jeh.H Wiley & 

        ~

(3) wherefio is the density ef the fluid, pg- is the density of the vapor, e: is the surfaee tension of the liquid phase in sontaet *with vapor, ~ is the eHthalpy of the sat\lfated fluid, and hg;sa< is the enthalpy ef the vaper phase with all values at sat\lfatieH eeHditiens ef temperatme and pressure. Smfaee teHsioH data provided by Bejan was fit to a polyHomiaJ (usiHg temperatHre iH °C) to geHerate data fer the temperature raHge of iHterest, Veith an Ri value of 0.999998: o 1.000E-11

  • T 4 + 7.370E- 09
  • T 1 -1.969E- 06
  • T' + 4.709E - 06
  • T + 7.1833E -02 Pressure at the eere is determiHed by baremetrie pressllfe at the faeility elevatieH, vaeuum maiHtaiHed iH the reaster bay aHd the 'Neight sf the water S\'er the sere. Baremetris pressllfe asseeiated with the Manhattan, Kansas airpert is 29.92 in. Mg. The reaeter bay is maiHtaiHed at a slight vaeUHm, with the maximum gage pressure (a iH, ef water) eerrespeHdiHg te 0.4 4 iH Mg; HemiHal baremetrie pressure eerreeted fer mai<imHm reaster bay vaeuum (a ehange ef apprmdmately 1.5%) eerrespeHds te 99.83 kPa.

VariatieHs in leeal baremetrie pressure are en the erder ef the serrestieH fer reaeter bay K-State Reactor 4-22 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION For Stieeooled eoiliag, the eritieal heat fhm is ealetilated ey (Ivey and Morris 1978):

P1 x ( cp,f. TSAT -Tsub 

                                                                               )

sub = SAT hg,sat - hf.sat

  • +ehle 4.9, Critieel Heat Fill* Ratios (GHF versus Maidmum
                 ~

Heat FIU*) fer 13 & Hi Feet efWeteF OveF the Cere G~i;:g \~d ft) G~i;:g (rn ft) ~ ~ 4-a ~ &,g4. 9,.74. ~

w &.4S ~ &.-M ~
                   ~             &.-=1-S           ~                ~          ~

w ~ &.9e &.@ ~ 

                   ~             a,w               &..97            ~          ,4.&%

4G ~ ~ ~ ~ 4a 4.-W &.-00 &.-W ~ w 4.-99 4-+& 4,-74 ~ aa ~ ~ 444 ~ eg 44-0 ~ 44§ ~ ea ~ ~ ~ ~ 

                   +G            ~                 ~                ~          ~
                  +a             ~                 ~                ~          ~

w 2-,94- ~ ~ ~ ge ~ ~ ~ ~ w ~ ~ ~ ~ K-State Reactor 4-23 Original (12/04) Safety Analysis Report

  • CHAPTER 4 Table 4.9, Critieal Heat flux Raties (CHf versus Maximum Heat flux) fer B & Hi Feet efWater Over the Cere CFIFR (19 ft)

As iHElieateEI iH Table 4.9, the aernal heat flim is less thaH the eritieal heat flHJ< fer operatiHg temperatHres 1:1p to 55 °C by more than a faetor of 4 eoHsiEleriHg both 13 feet anEI I e feet of water above the sore. The CHFR is greater than 2 for pool temperat1:1res ei<eeeEliHg the maidffitlm operatiHg val1:1e tlfl to 95 °C, aHEI remaiHs Hear 2 at val1:1es tlfl to 99°C. The Eliffenmce iH the eritical heat flt1>< ratio for 13 anEI I e feet of water is relatiwly 

                                                                                                                                                               /( Formatted                        [

small, '<Vith a miHim1:1m EliffereHee comp area to the mean of the tvro val1:1es of I.&% anEI a //~=================='= maxim1:1m of3.&e% belmv eO °C, e.4% aeross all pool temperattlfes eoHsiElereEI. //( Formatted [ 

                                                                                                                                                             ///,( Formatted                       [

It is elear from the table that there is a very *.viEle margiH betweeH the eperatiHg heat flm< /j/,[>F=o=r=m=a=tt=e=d=============~[ anEI the eritieal heat flim eveH te t1Hrealistieally high pool *.vater temperatHre, se that film boiliHg aHEI eJ!Sessive elaEIEliHg temperatHre is Hot a eoHsiEleratieH iH steaEly state f//( Formatted [ 

                                                                                                                                                            /*,/>=====================

operatioH. 1if!( F tt d [ 

 £or the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD 3.3* //i/'>=o=r=m=a=e==================

oatcli-04 (Feldman 2oil8i:.A~~illih2i:~[ilie_:-iPJl4t;JJ5Rl°t!s_e!lie~II1Hi~eIJ1.!i)_~~Iw-9_:-ii~i= 'i*ii//l/ft[:=F=o=rm=att=ed=================[ dependent volumes, enforcing th~ pre~sure ?oun?ary conditions, and tyvo pi12es. ?imulating the !!//if!//!!,[ Formatted [ col.d a_n_-d-*hot_ ch_ ann_ ~I- connec_-t-ed_ via a__ s_ mg-le rnn__ ct10n .c_ ~m-12o_n._ent o_f_ _~L-~P- _,__H_e_a__ t 1_s_ a_d___ de_ cl to_ t?dj /J i/////(>=F=o=r=m=a=tt=e=d==============='=[ 1 flrud by IJ1~0EJJOJ°~tmg the heat,.structure com12onent (su:iulatmg .a fuel element) of RELAP '":1th]// /fi!i/f>=====================- an appropnate ,a)(Ial_ po~er profile and power le\lel. In this an_ alys1s, the 120_ we__r_ level_ for_ the B n __ngj// /i///i/,[>F=o=r=m=a=tt=e=d==============~[ is at 24 kW (corresponding to an 85-element core with <ving-to-averagepeaking factor o_f 1_._6JJJ/ //ff!/;/,[ Formatted [ 

  '[hisp?\Ver level is a.1mlied to the heat struct,ure \Vi thin the single ch:mnel. Themodel assmpes __an i //fJ////(~=================='=

operatmgpressure of 143 kPa. and an operatmgtemperature of322.15 K (49.15°C). 1!//f/iii/>F=o=r=m=a=tt=e=d==============="=[ i//!Ji//( f'i/ 11/ji 1>F=o=r=m=a=tt=e=d==============="=[ Jhe version of_ the _Rl'::C:,1\P code licensed to KSU uses PG-CHF correlation which is a state of the 1!!/!/;jl i ( Formatted [ art best estimate CHF correlation developed .~Y }Juclear Rese_arch institute of Rez in the Czech_J/:/f;///>====================='= Republic. It is based on datajn the Czech Republic data bank froml 73 diffe1ent sets o[tube dataJ,jf;///;{>F=o=rm~att~ed~~~~~~~~~~~~=====[ 23 sets .of *a:uml~r d.ata, an4_153 sets of rod; bun~le_ datli. T_liert: are four__~Or_lJ1S _cif ~h(: PG.-C::H_F' !if///:{ Formatted [ correlat10n Basic', Flux'. Geometry'. and Power'. For the rod bundle 1t ,)S ariphcable 1n the~ltj;; /[>====~==~=~=======~ pressure range of 0.28 MPa to 18.73 MPa. for a mass fluxof.34.1 to,7478 kg/s~IlJ2. f()r0.4-}.0_!Il /f / // >F=o=r=m=a=tt=e=d=============~[ length and for a diameter of 0.00241 to .0.07813 m. TRIGA has an operating pressure of 0.143 J/ (! /[ Formatted [ MPa and fuel rod length of 0.381 m, thus the 012erating conditions_ fall outside the range of the /,i/[>F=o=r=m=a=t=te=d================[ applicability of Jhe PG-CHF correlation. and .a diffe1ent correlation is reguired to asses,.s theJI-/ 

 .departure from nucleate boiling ratio. (DNBRratio). One such correlation.which is applicable for //(>F=o=r=m=a=t=te=d================[

the low pressure range observed in TRIGA reactor -facility is the. Bernath correlation. the ':::~i Formatted [ functional form of the Bernath correlation.can be presented in the following eguati(_lns. _ ---------:~ i>F=o=r=m=a=tt=e=d================[ 

                                                                                                                                                      \\ ""':===================='=
                                                                                                                                                        \ \. [ Formatted                           [
                                                                                                                                                           \ '.~=================~
                                                                                                                                                            \ *[ Formatted                         [
                                                                                                                                                                '[ Formatted                       [

K-State Reactor 4-24 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION s I Inlet JFormatted: Centered, Keep with next I

I I I I Hot-leg Cold-Leg I

  • connector Figure 4.12 - RELAP single channel model used in CHF analysis * - - - ( Formatted: Caption, Centered
                                                                                        * - - - Formatted: Centered, Don't adjust space between Latin and Asian text, Don't adjust space between Asiar text and numbers, Tab stops: 3", Centered + 5.5,

Rig t K-State Reactor 4-25 Original (12/04) Safety Analysis Report

  • CHAPTER4 4
                              ~=       ~ 6 , if Dh :::: 0. lft D"
                              ~=_!2_+90,if Dh ~O.lft Dh hso =film coefficient at CHF Dh =hydraulic diameter (ft) v =coolant velocity (ft Is)

Twso =wall temperature at burnout (° C) DH =heated diameter (ft) The RELAP simulations were performed for the hot channel. i.e., a channel with a radial peaking--- Formatted: Justified, Don't adjust space between Latir factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain and Asian text, Don't adjust space between Asian text the pressure, temperature. and velocity distribution at different axial locations. With these and numbers calculations and the functional form of the Bernath correlation. the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA. the A-ring is occupied by the central thimble. not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard. The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel. K-State Reactor 4-26 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION
                                                                       ,, .. Bernatf1-CHF l  Formatted: Centered, Keep with next I

II A

  • PG-CHF
                                                                       ,, "  Heat flux I

I I I l.L

c 2000 u

1000 8.oo Cl.OS IUO 0.15 0.20 0.25 Cl.30 0.35 Heated Length (m) Figure 4.13 - CHF versus heated length * - - ( Formatted: Caption, Centered ti Formatted: Centered, Keep with next 

                       + + Bernath-CHF
                        <> <> PG-CHF                                                                                I I

I 10 ......... *.. I 

                                                                                                                /
                       ¢> <). ¢> '
                                  ;,                                                                         i B ..... .                                                                                I 0::.

o::l z 0 6

  • 4.

2 8.oo 0.0S 0.10 0.15 0.20 0.25 0.30 0.35 Heated Length (m) Figure 4.14 - DNBR versus heated length *--~( Formatted: Caption, Centered K-State Reactor 4-27 Original (12/04) Safety Analysis Report

  • CHAPTER4
 *------------------------------------------ _ ___ _______ _____ __________ __ :'.~x-;:~*-( Formatted: Font: 11 pt, Not Bold
d. Pulsed Mode of Operation Formatted: Justified, Don't adjust space between Latir and Asian text, Don't adjust space between Asian text Transient calculations have been performed using a custom computer code TASCOT for and numbers transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. ((;.2) and (.'.710), with the peak fuel temperature set to 746 °C, i.e., a temperature rise of 719 °C above 27 °C ambient temperature. The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1% ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation. In the TASCOT calculations, thermal conductivity was set to 0.18 W cm- 1 K- 1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm- 1 K- 1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994)

The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955) k0.79

  • C0.45 *A 0.51 -

hb = 0.00122

  • f pf
                                                                                    *( Tw - T,at )0.99 , (9-10)
                                                       * (v g _v v )

0.75 [ a 05 * µ f0.29

  • p g0.24
  • To.15 sat in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 °C),

and n the coolant ambient temperature (27°C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases. The overall heat transfer coefficient U varies negligibly for ambient temperatures from 20 to 60 °C, and has the value 0.21 W cm- 1 K- 1 at Tb = 27 °C. Figure 4.14,2 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.108 lists temperatures and heat fluxes as function of time after a 2.1 % ($3.00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m- 1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500°C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse. K-State Reactor 4-28 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION 1000 Os 800 u

0 2 600 Ill L..

J 4
       .µ l1l L..                                                               8 Ill
0. 400 E

Ill 16 I-200 32 64s 0 ~~~~~~~~~~~~~~~~~~~~~~~~ 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0 2.2 Radius (cm)

  • Figure 4. 14~, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse.

K-State Reactor 4-29 Original (12/04) Safety Analysis Report

  • CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{°C) Coolant Ambient Temperature.

Time (s) Q" CHFR Fuel outside Clad surface (Wm-2 ) Temp. (0 C) Temp. (0 C) 0 953 1 3.57 xl0 5 4.2 781 224 2 7.34 xl05 2.0 683 432 4 8-52 xl05 1.7 574 498 8 7.54 xl0 5 2.0 461 443 16 5.71 xl05 2.6 344 342 32 3.46 xl05 4.3 224 218 64 1.04 xl05 14.4 100 84 4.6 Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by (.Wll) in which Po and Po are the density and volumetric expansion coefficient at core inlet conditions (27°C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm2 s* 1, fiT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by (-l+.12.) in which mis the coolant mass flow rate (kg s* 1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m2, and Dh is the hydraulic diameter, namely, 0.02127 m. The friction factor /for laminar flow through the annular area is given by 100 Re- 1 (Shah & London 1978), in which the Reynolds*number is given by Dhm I Aµ 0 in which µ,, is the dynamic viscosity at core inlet conditions. Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5), and the entrance pressure loss would be expected to be negligible. The exit contraction loss is given by (-l+/-.Ll.) K-State Reactor 4-30 Original (12/04) Safety Analysis Report l

  • REACTOR DESCRIPTION The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which 2 2 K =. [ -

A, ] Ac =[ 3*R sin60° cos60°] 7f

  • Rz = 0171 (H14) where R is the radius of the opening in the upper grid plate. Equations (H.12.) through (Hl4),

solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4.911. Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRI GA Reactor During Steady-State Operations. P (kWt) m (kg s* 1) AT( 0 C) 50 0.047 3.1 100 0.061 4.7 200 0.077 7.5 300 0.090 9.6 400 0.100 11.5 500 0.108 13.3 750 0.125 17.2 1000 0.139 20.6 1250 0.150 23.8 4.7 Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500°C; consequently, limits on fuel temperature change for cladding temperatures greater than 500°C. A maximum temperature of 1150°C (with clad< 500°C) and 950°C (with clad> 500°C) for U-ZrH (H/Zri.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges. 4.8 Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature. The operating parameters established for the KSU TRlGA reactor are: K-State Reactor 4-31 Original (12/04) Safety Analysis Report

  • CHAPTER4
  • Steady-state power level
  • Fuel temperature measured by thermocouple during pulsing operations
  • Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Taele 4.9Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100°C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.

Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation. 4.8.3 Safety Margins FoF l,25Q kW steady state operations, the eritieal heat fhm Fatio indieated in Taele 4.9 ranges from 5.8 for pool water at room temperatl!Fe (27°C) to 4.1 at iQ °C (pool tenlperatHres aFe sontrolled to less than 48°C foF operational eonee~s). Even at pool water temperatHFes approaehing boiling, the margin remains aeo\'e 2. TheFefore, margins to sonditions that eoHld eaHse e1ceessive temperatl!res dHFing steady state opeFations while eladding temperatHFes is below 5QQ°C are e1ttremely large. For 1.250 kWth steady-state operations. the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49°C. The proposed Technical Specifications limit of 44°C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation. Limiting pool inlet water temperature to no greater than 44°C (or 37°C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling. Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746°C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869°C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15% margin to the fuel temperature safety limit for cladding temperatures below 500°C. Analysis shows that cladding temperatures will remain below 500°C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve K-State Reactor 4-32 Original (12/04) Safety Analysis Report

  • REACTOR DESCRIPTION 500°C (invoking a 950°C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.

Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500°C, but the heat transfer simultaneously cools the fuel to much less than 950°C. If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 500°C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500°C. 4.9 Bibliography "TASCOT- A 2-D, Transient and Steady State Conduction Code for Analyhsis of a TRJGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C., 

  "Investigation of the Radial Variation of the Fission-Heat Source in a TRJGA Mark Ill Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C.,
  • "A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,

Kansas State University TRJGA Mkll Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J .G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K., and N. Zuber, Theory and Design of Modern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F., 

 "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al.,

K-State Reactor 4-33 Original (12/04) Safety Analysis Report

  • CHAPTER4 MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998.

NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRJGA Reactors," U.S. Nuclear Regulatory Commission, 1987. 

 Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K.,

and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRJGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T. 

 "Safety Analysis Report, TRJGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991.

K-State Reactor 4-34 Original (12/04) Safety Analysis Report

Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998). The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A.1. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface. Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100°F water, showed burnout fluxes of 0.9 to 2.0 Mbtu ft* 2 hr- 1 for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft* 2 hr- 1. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation. A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A. land 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise. The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760°C clad temperature for a peak initial fuel temperature of 1000°C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure 4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor 4.A-1 Original (9/02) Safety Analysis Report

  • CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470°C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400°C to 500°C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965].
  • RADIUS (IN.)

Figure 4A.l. Representative Radial Variation of Power Within the TRIGA Fuel Rod I. l I .0 0.9 N 0.8

0. 7
0. 6
o. 5 0

AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.) Figure 4A.2, Representative Axial Variation of Power Within the TRIG A Fuel Rod. K-State Reactor 4.A-2 Original (9/02) Safety Analysis Report

  • REACTOR DESCRIPTION
                                                     \            CURVE BASED ON
                                                        \ ' , , DATA OF ELLI/.

x 5:z: lOlt TW-TSAT ('FJ Figure 4A.3, Subcooled Boiling Heat Transfer for Water.

  • 1800 1700 1600 1500 llfOD IJOO 100 SEC 1200 0.1 0.2 0.3 o.* 0.5 0.6 0.7 o.e RADIUS (IN.)

Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse. K-State Reactor 4.A-3 Original (9/02) Safety Analysis Report

  • CHAPTER 4 APPENDIX A 106 ONSET OF_!_! PEAK HEAT FLUX NUCLEATE BOILING N
                       ~
                      <:: 105
                       ~
                       ~

w u 10 4 i 1ol 0.001 0.01 0.1 1.0 10 100 ELAPSED Tito![ FRCIH ENO OF PULSE (SEC) Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse. 1000 CLAO OUTER SURFACE TEHP 

                                  /

10 L---1~.....L.-1.....J...IL-....1.~.....1.......1....LJL..-1~.....L......1...J..1~-'-~-'--'-......""'.""-'-~~~~10*0 0.001 O.DI 0. 1 1.0 10 ELAPSEO TIME FROM ENO OF PULSE (SEC) Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element. K-State Reactor 4.A-4 Original (9/02) Safety Analysis Report L

  • REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.

This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected. Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962] which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000°C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960] which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate. The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft* 2 hr 1 °F- 1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft* 2 hr*' °F* 1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel. The surface heat flux at the midplane of the element is shown in Figure 4A.10 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft* 2 hr- 1 °F- 1 according to the more conservative Bernath correlation at a local water temperature approaching saturation. K-State Reactor 4.A-5 Original (9/02) Safety Analysis Report

  • CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000°C is conservatively estimated to be 4 70°C.

As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470°C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000°C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is l l 50°C would not produce a stress in the clad in excess of its ultimate strength. TRI GA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about l 150°C without damage to the clad [Dee et al. 1966]. 7 ELAPSED TIME FROM 

                                   *ENO OF PULSE
  • 0.247 SEC 6
  • "'0 I

5 3 7 8 9 10 11 12 13 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A.7, Surface Heat Flux Distribution for Standard Non-Gapped (hgap= 500 Btu/h ft2 °F) Fuel Element After a Pulse. K-State Reactor 4.A-6 Original (9/02) Safety Analysis Report

  • REACTOR DESCRIPTION ELAPSED TIME FROM END OF PULSE IS 0.314 SEC 10 11 13 15 0 I STANCE FROM BOTTOM OF FUEL (IN. l Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap= 375 Btu/h ft 2 °F) After a Pulse.

8 7 N I- 

              "-      6 a:
              ~

I- 

              "' s
              'o                                ELAPSED TIME FROM END OF PULSE IS 0.440 SEC
               ><     4 I-3
               "':i::

2 l 7 8 9 10 ll 12 13 14 15 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 250 Btu/h ft2 °F ) After a Pulse. K-State Reactor 4.A-7 Original (9/02) Safety Analysis Report

  • CHAPTER 4 APPENDIX A EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR*FT2. -°F 500
         ~
         ...I 105 a::
r m
         )C.
c u
           ~         IOI+

FLOW VELOCITY a 1 FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME a ZERO) 10 3 '--~~-'-~~~'--~..___.__._~~~~~~-'-~-'---'--' 0.01 0.1 1.0 ELAPSED TIME FROM END OF PULSE (SEC) Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse. K-State Reactor 4.A-8 Original (9/02) Safety Analysis Report

  • REACTOR DESCRIPTION Bibliography "A Theory ofLocal Boiling Burnout and Its Application to Existing Data, " Heat Transfer -

Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No. 20.Bemath, L., Research in Improved TRIGA Reactor Performance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al., Characteristics ofLarge Reactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al. Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al., A Study of the Mechanism ofBoiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E., Thermal Conductance ofMetallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis of a Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J. Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172. Graff, W.J. Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H .. MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H. Choi, "Quarterly Technical Report SPERT Project, April, May, June, 1964, "ISO 17030. Spano, A. H., 

  "The Effect of Subcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hvdraulic CHF analysis." Technical report, Argonne National Laboratory. 2008, E.E. Feldman.

K-State Reactor 4.A-9 Original (9/02) Safety Analysis Report

  • CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure. system models, and solution methods. *----- Formatted: Normal, Don't adjust space between Latin and Asian text, Don't adjust space between Asian text "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution." and numbers Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong. ----------------------~~--(>=~Fo~r~m~a~tt~e~d~:~Fo~n~t:~lt~a~lic----------
                                                                                                     \ \ '( Formatted: Font: Italic
                                                                                                        \l Formatted: Font: Bold K-State Reactor                               4.A-10                                Original (9/02)

Safety Analysis Report

  • REACTOR DESCRIPTION "Onset of Stable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al.,

UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991. 

 "Hydrodynamic Aspects ofBoiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959.

Zuber, W . K-State Reactor 4.A-11 Original (9/02) Safety Analysis Report

  • AppendixB Water Properties at Nominal Operating Conditions Data for 16 Feet of Water over the Core Tpool Pi,161 11 ht,161 11 hg,161 11 P9 ,15111 Tsat,161 11 q"sat, 15131 q"sub14l oc kg m- 3 kJ kg- 1 kJ kg- 1 kg m- 3 oc wm-2 wm-2 15 999.21 950.00 47.79 465.10 2692.64 0.85 110.89 1553.842 7239.19 20 998.32 950.01 47.74 465.05 2692.63 0.85 110.88 1552.078 6931.74 25 997.16 950.02 47.69 465.01 2692.59 0.85 110.87 1549.496 6622.60 30 995.75 950.03 47.62 464.95 2692.59 0.85 110.86 1547.118 6311.82 35 994.12 950.04 47.54 464.89 2692.57 0.85 110.84 1543.981 5999.91 40 992.29 950.06 47.46 464.81 2692.54 0.85 110.83 1540.446 5688.24 45 990.27 950.07 47.36 464.73 2692.51 0.85 110.81 1536.512 5376.29 50 988.07 950.09 47.25 464.64 2692.48 0.85 110.78 1532.205 5064.36 55 985.70 950.11 47.14 464.54 2692.45 0.85 110.76 1527.561 4753.90 60 983.18 950.12 47.02 464.44 2692.41 0.85 110.74 1522.575 4444.85 65 980.50 950.14 46.89 464.33 2692.37 0.85 110.71 1517.255 4136.85 70 977.69 950.17 46.76 464.21 2692.33 0.84 110.68 1511.666 3830.73 75 974.74 950.19 46.62 464.09 2692.29 0.84 110.65 1505.778 3526.89 80 971.66 950.21 46.47 463.96 2692.24 0.84 110.62 1499.613 3225.47 85 968.45 950.23 46.32 463.83 2692.19 0.84 110.59 1493.199 2926.81 90 965.12 950.26 46.16 463.69 2692.15 0.84 110.56 1486.527 2631.05 95 961.68 950.29 45.99 463.55 2692.09 0.84 110.53 1479.626 2338.47 97 960.27 950.30 45.92 463.49 2692.07 0.84 110.51 1472.944 2216.11 99 958.84 950.31 45.86 463.43 2692.05 0.84 110.50 1466.058 2095.18 Data for 13 Feet of Water over the Core Tpool PTl1J Pr,131 11 hr,13l1l hg,1}11 Tsat,131 11 q"sa1,13l3l q"subl41 oc kg m- 3 kg m-3 kJ kg- 1 kJ kg- 1 oc wm-2 wm-2 15 999.21 951.43 38.83 457.21 2689.85 0.80 109.03 1513.00 6964.74 20 998.32 951.43 38.79 457.18 2689.84 0.80 109.02 1511.32 6857.12 25 997.16 951.44 38.75 457.13 2689.82 0.80 109.01 1509.15 6543.62 30 995.75 951.45 38.69 457.09 2689.80 0.80 109.00 1505.85 6229.30 35 994.12 951.46 38.63 457.03 2689.78 0.80 108.99 1503.13 5913.58 40 992.29 951.47 38.56 456.96 2689.76 0.80 108.97 1500.16 5597.21 45 990.27 951.49 38.48 456.89 2689.74 0.80 108.96 1496.38 5281.90 50 988.07 951.50 38.39 456.82 2689.71 0.80 108.94 1492.25 4966.66 55 985.70 951.51 38.30 456.73 2689.68 0.80 108.92 1487.78 4652.39 60 983.18 951.53 38.20 456.64 2689.65 0.80 108.90 1482.99 4339.60 65 980.50 951.55 38.10 456.55 2689.61 0.80 108.87 1477.90 4027.94 70 977.69 951.57 37.99 456.45 2689.58 0.80 108.85 1472.52 3718.83 75 974.74 951.58 37.88 456.35 2689.54 0.80 108.83 1466.86 3412.07 80 971.66 951.60 37.76 456.24 2689.50 0.80 108.80 1460.95 3107.29 85 968.45 961.62 37.63 456.12 2689.46 0.80 108.77 1458.59 2812.63 90 965.12 951.64 37.50 456.01 2689.42 0.79 108.75 1448.37 2506.90 95 961.68 951.67 37.37 455.89 2689.38 0.79 108.72 1441.74 2211.19 97 960.27 951.68 37.31 455.84 2689.36 0.79 108.70 1435.27 2087.92 99 958.84 951.68 37.26 455.78 2689.34 0.79 108.69 1428.60 1966.16 K-State Reactor 4.B-1 Revised 05/01 /17 Safety Analysis Report
  • Common Data REACTOR DESCRIPTION T Patm Cp(1] ()"
                              "C           kPa       kJ kg-1 "1      Nm*1 15         99.83       4.23080       0.07149 20          99.83       4.23080       0.07120 25          99.83       4.23080       0.07083 30          99.83       4.23080       0.07039 35          99.83       4.23070       0.06989 40          99.83       4.23070       0.06932 45          99.83       4.23070       0.06869 50          99.83       4.23070       0.06800 55          99.83       4.23060       0.06727 60          99.83       4.23060       0.06649 65          99.83       4.23060       0.06566 70          99.83       4.23050       0.06480 75          99.83       4.23050       0.06390 80          99.83       4.23050       0.06297 85          99.83       4.23040       0.06201 90          99.83       4.23040       0.06102 95          99.83       4.23040       0.06001 97          99.83       4.23030       0.05898 99          99.83       4.23030       0.05793 NOTE[1]: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa =Heigth(ji) *12(in/ft) *O. 0254(meters/in) *Density(kg/m 3) *9.8066511000 NOTE[4]: q;ub =   q~AT. [1+0.l*(PJJ x . cp.f *(T~r -Tsub)l Pg           hg.sat    h f,sat NOTE:[5}:

u = 1.000E -11

  • T 4 + 7.370E - 09
  • T 3 -1.969E - 06
  • T 2 + 4.709E - 06
  • T + 7.1833E - 02 K-State Reactor 4.A-13 Original (9/02)

Safety Analysis Report

4. Reactor Description 4.1 Summary Description The Kansas State University (KSU) Nuclear Reactor Facility, operated by the Department of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Department is also the home of the Tate Neutron Activation Analysis Laboratory.

The TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Regulatory Commission License R-88 and the regulations of Chapter 1, Title 10, Code of Federal Regulations. Chartered functions of the Nuclear Reactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems. The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool and fueled with heterogeneous elements consisting of nominally 20 percent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:

  • Central thimble
  • Rotary specimen rack
  • Thermalizing column with bulk shielding tank
  • Thermal column with removable door
              *Beam ports
  • Radial (2)
  • Piercing (fast neutron) (1)
  • Tangential (thermal neutron) (1)

The reactor was licensed in 1962 to operate at a steady-state thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kW, with fuel loading to support 500 kW steady state thermal power and with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at least 83 planned for the 1,250-kW core), each containing as much as 39 grams of 235 U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 ft.) depth. Criticality is controlled and shutdown margin assured by three control rods in the form of aluminum or stainless-steel clad boron carbide or borated. graphite. A fourth control rod would be used for 1,250-kW operation. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-m3 (144,000 ft. 3) confinement building K-State Reactor 4-1 Original (12/04) Safety Analysis Report

CHAPTER 4 made of reinforced concrete and structural steel, with composite sheathing and aluminum siding. Sectional views of the reactor are shown in Figures 4.1 and 4.2. Criticality was first achieved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW. The original aluminum-clad fuel elements were replaced with stainless-steel clad elements in 1973. Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in 1993, with support from the U.S. Department of Energy. All neutronic instrumentation was replaced in 1994. North

  • 22 FT 1Q IN.

I I I TJ.liG(tlfl.>J..j 

                                                                                            !If.AM PCf?TI
                                 . - - - - - - - - - 2 8 Fr .; tN.-----------------j I

Figure 4. 1, Vertical Section Through the KSU TRIGA Reactor.

  • K-State Reactor Safety Analysis Report 4-2 Original (12/04)

REACTOR DESCRIPTION

  • 4.2 Reactor Core The General Atomics TRIGA reactor design began in 1956. The original design goal was a completely and inherently safe reactor. Complete safety means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accident. Inherent safety means that an increase in the temperature of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fuel in a zirconium hydride matrix.

I 11 

                           .--11*r---_,
                   ; .., t2 rT r- Iau.. l(-!iH1i::1..c,,wc

(~"~'lllilfr;:AL ~~!<>! 

                   ',. . I rn ei'I,       
                -,~
      ... ~' . -

West ~------- --- 28 F'f 4 i'i. ---------i East Figure 4. 2, Horizontal Section Through the KSU TRIGA Reactor. The basic parameter providing the TRIGA system with a large safety factor in steady state and transient operations is a prompt negative temperature coefficient, relatively constant with temperature (-0.01 % ~k/k C). This coefficient is a function of the fuel composition and core 0 geometry. As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase. Therefore, fuel and clad temperature automatically limit operation of the reactor .

  • K-State Reactor Safety Analysis Report 4-3 Original (12/04)

CHAPTER4 It is more convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an 83 element core and 120°F inlet water temperature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatures which could cause loss of clad integrity could not occur. An 85-element core distributes the power over a larger volume of heat generating elements, and therefore results in a more favorable, more conservative, thermal hydraulic response. 4.2.1 Reactor Fuel 1 TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic. Additional advantages *were that ZrH has a high heat capacity, results in relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size. TRIGA fuel is designed to assure that fuel and cladding can withstand all credible environmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material. The maximum temperature limits of l 150°C (with clad< 500°C) and 950°C (with clad> 500°C) for U-ZrH (H/Zr1.6s) have been set to limit internal fuel cladding stresses that might challenge clad integrity (NUREG 1282). These limits are the principal design bases for the safety analysis.

a. Dimensions and Physical Properties.

The KSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements. Three instrumented aluminum-clad Mark II elements are still available for use in the core. General properties of TRIGA fuel are listed in Table 4.1. The Mark III elements are illustrated in Figure 4.3. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.) hole drilled through the center of the active fuel section. Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (0.3 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (1 in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor tank. 1 Unless otherwise indicated, fuel properties are taken from the General Atomics report ofSimnad [1980] and from authorities cited by Simnad .

  • K-State Reactor Safety Analysis Report 4-4 Original (12/04)

REACTOR DESCRIPTION

  • Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of 55.88 cm (22 in.).

Table 4.1, Nominal Properties of Mark II and Mark III TRIGA Fuel Elements in use at the KSU Nuclear Reactor Facility.

Property Mark II Mark III Dimensions Outside diameter, Do= 2r0 1.47 in. (3.7338 cm) 1.47 in. (3.7338 cm)

Inside diameter, D;= 2r; 1.41 in (3.6322 cm) 1.43 in. (3.6322 cm) Overall length 28.4 in. (72.136 cm) 28.4 in. (72.136 cm) Length of fuel zone, L 14 in. (35.56 cm) 15 in. (38.10 cm) Length of graphite axial reflectors 4 in. (10.16 cm) 3.44 in (8.738 cm) End fixtures and cladding aluminum 304 stainless steel Cladding thickness 0.030 in. (0.0762 cm) 0.020 in. (0.0508 cm) Burnable poisons Sm wafers None Uranium content Weight percent U 8.0 8.5 235 U enrichment percent 20 20 235 U content 36 g 38 g Physical properties offuel excluding cladding H/Zr atomic ratio 1.0 1.6 Thermal conductivity (W cm- 1 K- 1) 0.18 0.18 Heat capacity [T ~0°C] (J cm-3 K- 1) 2.04 + 0.00417T Mechanical properties ofdelta phase U-ZrH" Elastic modulus at 20°C 9.1 x 106 psi Elastic modulus at 650°C 6.0 x 106 psi Ultimate tensile strength (to 650°C) 24,000 psi Compressive strength (20°C) 60,000 psi Compressive yield (20°C) 35,000 psi asource: Texas SAR [1991].

b. Composition and Phase Properties The Mark III TRIGA fuel element in use at Kansas State University contains nominally 8.5% by weight of uranium, enriched to 20% 235 U, as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH1.6, the equilibrium hydrogen pressure is one atmosphere at about 760°C. The single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen. Over 25,000 pulses have been performed with the TRI GA fuel elements at General Atomic, with fuel temperatures reaching peaks of about 1150°C .
  • K-State Reactor Safety Analysis Report 4-5 Original (12/04)

CHAPTER4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrH1.64 and ZrH1.14 at room temperature, and closes at ZrH1.1 at 455°C. From 455°C to about 1050°C, the delta phase is supported by a broadening range ofH/Zr ratios. I 20MIL STAINLESS STEEL CLAD ZIRCONIUM HYDRIDE-28*37" 8*5WT% URANIUM, 20%ENR., 38 235LJ 11 l:.43 DIA 1*47"DIA STAINLESS T 3*44 11 STEEL BOTTOM END-FIXTURE J_ Figure 4.3, TRIGA Fuel Element. K-State Reactor 4-6 Original (12/04) Safety Analysis Report

REACTOR DESCRIPTION

  • c. Core Layout A typical layout for a KSU TRIGA II 250-kW core (Core 11-18) is illustrated in Figure 4.4. The layout for the 1,250-kW core is expected to be similar, except that the graphite elements will be replaced by fuel elements, one additional control rod will be added, and control rod positions will be adjusted ..

Figure 4.4, Core Layout (250 kW). The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is $2.00, the 1,250 kW core pulse rod reactivity worth is $3.00), balancing the remaining control rod's worth to meet minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge

  • K-State Reactor Safety Analysis Report 4-7 Original (12/04)

CHAPTER 4 4.2.2 Control Rods The pulse rod is 3.175 cm. (1.25 in.) diameter. Other rods are 2.225 cm (7/8 in.) diameter. Control rods are 50.8 cm. (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm. (30-mil) aluminum sheath. The control rod drives are connected to the control rod clutches through three extension shafts. The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydraulic cushion (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (slotted tubes on the right hand side) along with the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder; the dimensions of the cylinder limits bottom travel. Upper Pulse, Shim & Reg Rods Reg Rod Shim Rod Pulse Rod Figure 4.5, Control Rod Upper Extension Assemblies The bottom of the pulse rod is shown on the left hand side of Figure 4.5. The upper extension shaft is a hollow tube, the middle extension is solid. The upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the mechanical shock during a pulse requires a more sturdy fastener). Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full in position is above the bottom of the guide tube .

  • K-State Reactor Safety Analysis Report 4-8 Original (12/04)

REACTOR DESCRIPTION

  • i.4 .. t~*n.o':t!-'

{

  • 4' I t *,1cr
            .. <1-J.**t~t'l':t~*
                                                  -*-*,~~.wy-...-.,.
                                                                                   .-~~:~~~~: ~"
               "'  **-~~~

Figure 4.6, Middle Extension Rod Alignment Holes The middle solid extension is similarly connected to the lower extension. The lower extension is hollow, the middle extension fits into the lower extension and a hole drilled in the overlap secures the lower extension to the middle extension. Typically the lower extension has a tighter fit than the upper extension because the lower and middle extension are not separated for inspections and because the interface with upper extension is used to set the bottom position of the control rod. Pictures of the lower connector for the pulse rod and one standard rod are shown at the left in Figure 4.7 .. Figure 4.7, Standard & Pulse Rod Lower Coupling The bottom of the lower extension attaches directly to the control rod. Pictures of the control rods taken during the 2003 control rod inspection are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. The guide tubes have a small metal wire in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube .

  • K-State Reactor Safety Analysis Report 4-9 Original (12/04)

CHAPTER 4 Pulse Rod Shim Rod Reg Rod Figure 4.8, Control Rods During 2003 Inspection

  • Full Guide Tube Upper Lower Lower Detail Position in Upper Grid Plate Figure 4.9, Control Rod Guide Tubes
  • K-State Reactor Safety Analysis Report 4-10 Original (12/04)

REACTOR DESCRIPTION

  • a. Control Function While three control rods were adequate to meet Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one transient/pulsing control rod). The control-rod drives are mounted on a bridge at the top of the reactor tank. The control rod drives are coupled to the control rod through a connecting rod assembly that includes a clutch. The standard rod clutch is an electromagnet; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-energizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control, while scrams operation at limiting safety system settings. A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 4.3. Note that (1) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting setpoint depends on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.

Ta ble 4.2 SummaryofC on t ro IR0 dint er Iocks INTERLOCK 'SETPOINT FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument startup channel reading is less Source Interlock 2 cps than instrument sensitivity/ensure nuclear instrument startup channel is operating Prevent applying power to pulse rod unless Pulse Rod Interlock Pulse rod inserted rod inserted/prevent inadvertent pulse Withdraw signal, Prevent withdrawal of more than I rod/Limit Multiple Rod Withdrawal more than I rod maximum reactivity addition rate Mode switch in Hi Prevent withdrawing standard control rods in Pulse Mode Interlock Pulse pulse mode Prevent pulsing if power level is greater than Pulse-Power Interlock IO kW IO kW NOTE: (1) Pulse-Power Interlock normally set at 1 kW, (2) only Pulse Mode Interlock required by Technical Specifications

b. Evaluation of Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor; scram times. for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIG A prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation .
  • K-State Reactor Safety Analysis Report 4-11 Original (12/04)

CHAPTER 4 T a ble 4.3 ' S ummarvofReac t or SCRAMs Limitin~ Trip Setpoint Measuring Steady Actual Setpoint Channel Pulse State Linear Channel High 110% NIA 104% Power Power Channel High power 110% NIA 104% Detector High 90% 90% 90% Voltage 600°C B Ring element High Fuel 555°C C Ring element 450°C Temperature[ll 480°C D Ring element 3 80°C E Ring element 350°C Period [IJ NIA NIA 3 sec NOTE [1]: Period trip and temperature trip are not required by Technical Specifications The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for 1,250 kW operations for eight hours per day (including xenon override) and will assure a

  • shutdown margin of at least $0.50 .

Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per minute, with a total travel about 15 in. (38.1 cm). Maximum rate ofreactivity change for standard control rods is specified in Technical Specifications. 4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Demineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specimen rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3.4 in. in length, acting as top and bottom reflectors. The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is 0.457-m (18.7 in.) inside diameter, 1.066-m (42 in.) outside diameter, and 0.559-m (20 in.) height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3 .18-cm ( 1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector. *

  • K-State Reactor Safety Analysis Report 4-12 Original (12/04)

REACTOR DESCRIPTION

  • The radial reflector assembly rests on an aluminum platform at the bottom of the reactor tank.

Four lugs are provided for lifting the assembly. A radial void about 6 inches (15.24 cm) in diameter is located in the reflector such that it aligns with the radial piercing beam port (NE beam port). The reflector supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within the fuel cladding provide additional reflection. Inserts are placed at both ends of the fuel meat, providing top and bottom reflection. 4.2.4 Neutron Startup Source A 2-curie americium-beryllium startup source (approximately 2 x 106 n s- 1) is used for reactor startup. The source material is encapsulated in stainless steel and is housed in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for use in relocating the source manually from the 22-ft level (bridge level) of the reactor. 4.2.5 Core Support Structure The fuel elements are spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates. The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fuel elements, has holes for receiving the lower end fixtures. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 special holes in it. The 1.5-in. (3.8 cm) diameter holes in the upper grid plate serve to space the fuel elements and to allow withdrawal of the elements from the core. Triangular-shaped spacers on the upper end fixtures allow the cooling water to pass through the upper grid plate when the fuel elements are in position. The reflector assembly supports both grid plates. 4.3 Reactor Tank The KSU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (1/4-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-ft) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion. Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration. All collection plenums are connected to a leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual leak-off volume (and pressure gauge) is installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool

  • K-State Reactor Safety Anaiysis Report 4-13 Original (12/04)

CHAPTER 4 through an experiment facility, pressure in the volume will increase; isolating individual lines allows identification of the specific facility with the leak. A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. The bridge is mounted directly over the core area, and spans the tank. Aluminum grating with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating can be lowered when activities could cause objects or material to fall into the reactor pool. The grating normally remains up to reduce humidity at electro-mechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations. Four beam tubes run from the reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall run toward the reactor core. Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter I 0). 4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield. The shielding configuration is similar to those at other TRIGA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cm. (3 ft.) at the top of the tank. The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena. 4.5 Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fuel temperature) for TRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels with 8.5% U, 20% enrichment, the value is nearly constant at 0.01 % ~k/k per °C, and varies only weakly dependent on geometry and temperature. Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures* that the fuel and clad temperature limits will not be exceeded. The design bases analysis indicates that operation at 1,250 kW thermal power with an 83-element across a broad range of core and coolant inlet temperatures with natural convective flow will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss of clad integrity. Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety

  • K-State Reactor Safety Analysis Report 4-14 Original (12/04)

REACTOR DESCRIPTION

  • limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6.

4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk II grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E or F (outermost two rings) as required to support experiment operations or limit excess reactivity. The bottom grid plate, which suppmis the weight of the fuel elements, has holes for receiving the lower end fixtures. 4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, taking the place of a graphite dummy element or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates. For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Cornell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor). Table 4.4; 250 kW Core Parameters. (3 * (effective delayed neutron fraction) 0.007 R. (effective neutron lifetime) 43 :S 

                                                                              -$0.017 EC- 1 CXTf (prompt temperature coefficient)                     (@ 250kW -275EC av (void coefficient)                                       -0.003 1%- 1 void
                                                                           -$0.006 kw- 1 to -

CXp (power temperature coefficient- weighted ave) $0.01 kw- 1 Table 4.5, Com arison of Control Rod Worths. KSU TRI GA Mark II (250 kW) Cornell University Core II-19 Core III-1 500k Pulse D-10 $1.96 C-4 $2.12 D-10 $1.88 Shim C-3 $2.88 D-4 $1.85 D-16 $2.20 Safety NA $0.0 D-16 $1.82 D-4 $1.99 Regulating D-16 $1.58 E-1 $0.79 E-1 $0.58 TOTAL NA $6.42 NA $6.58 NA $6.65 NOTE: Core III-1 has an experiment positioned to control the worth of the pulse rod

  • K-State Reactor Safety Analysis Report 4-15 Original (12/04)

CHAPTER4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature. 4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures. Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 1000°C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 470°C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr- 1 fr 2 0 p- 1 (2550 W m-2 K- 1) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions associated with both steady-state and pulse-mode operations.

a. Spatial Power Distribution The following conservative approximations are made in characterizing the spatial distribution of the power during steady-state operations.
  • The hottest fuel element delivers twice the power of the average.

Classically, the radial hot-channel factor for a cylindrical reactor (using Ras the physical radius and Re as the physical radius and the extrapolation distance) is given 2 by: with a radial peaking factor of 1.93 for the KSU TRIGA II geometry,. However, TRI GA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element. 2 Elements ofNuclear Reactor Design, znct Edition (1983), J. Weisman, Section 6.3

  • K-State Reactor Safety Analysis Report 4-16 Original (12/04)

REACTOR DESCRIPTION

  • Calculated thermal neutron flux data 3 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection) has a small range of 1.36 to 1.40.

Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.

  • The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of n/2 times the average, and heat conduction radial only.

The axial factor for power produced within a fuel element is given by: g(z) = l.514*co( !!....*

  • z ) , (6)
                                                          '2      2 f +f  ext in which e =LI 2 and eex1 is the extrapolation length in graphite, namely, 0.0275
m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor .

Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.

  • The location on the fuel rod producing the most thermal power with thermal power distributed over 83 fuel rods is therefore:
                            "    -       p      . !!_ . 2 -     p       p 0 8469                   (7) q  ma' -  83*7l'*D0 *L 2 - 83*Do *L = . .
  • The radial and axial distribution of the power within a fuel element is given by q"'(r,z) = q;:~f(r)g(z), (5) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core.

3 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G. B. West

  • K-State Reactor Safety Analysis Report 4-17 Original (12/04)

CHAPTER4

  • The radial factor is given by:

2 f (r) = a+ er + er , (7) 2 1+br+dr in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel (Ahrens 1999a). Values are: a= 0.82446, b = -0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is illustrated in Figure 4.11.

  • 1.3 1.2 1.1 "L 1.0 v

0.90

  • a.

r (cm) Figure 4.12, Radial Variation of Power Within a TRGIA Fuel Rod. Heat Transfer Models (Data Points from Monte Carlo Calculations [Ahrens 1999a]) The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the sum of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the t:i. T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (ron), cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ricl): Eq. 1 A standard heat resistance model for this system is:

  • K-State Reactor Safety Analysis Report 4-18 Original (12/04)

REACTOR DESCRIPTION

  • T =T c1 *
                                       +q"[_!_+ ln(X) +~+~1 h

ro k rh 2kf Eq. 2 c I g and heat flux is calculated directly as: q"= Ul1T = Tmax - ~ (2) 1 r0 ln(r0 Ir;) r r * 

                                           -+                     + 0- + 0-h         . kc          ljhg 2k1 in which ro and r; are cladding inner and outer radii, hg is the gap conductivity, h is the convective heat transfer coefficient, and k.r is the fuel thermal conductivity. The gap conductivity of 2840 W m* 2 K- 1 (500 Btu h" 1 ft -2 °F" 1) is taken from Appendix A. The convective heat transfer coefficient is mode dependent and is determined in context.

Parameters are cross-referenced to source in Table 4.6 . T abl e 4 6 Th ermo d1ynam1c . Va1ues Parameter Symbol Value Units Reference Fuel conductivity kr 18 Wm-lK"l Table 13.3 14.9 W m* 1 K- 1 (300 K) Table 13.3 16.6 W m* 1 K- 1 ( 400 K) Table 13.3 Clad conductivity kg 19.8 W m* 1 K- 1 (600 K) Table 13.3 Gap resistance ha 2840 wm-2 K- 1 AooendixA Clad outer radius ro 0.018161 M Table 13.1 Fuel outer radius fj 0.018669 M Table 13.1 Active fuel length Lr 0.381 M Table 13.1 No. fuel elements N 83 NIA Chap 13 Axial peaking factor APF nl2 NIA Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that: 

                                          ~ = 5.1858E-04 m K 2

2kf w Gap resistance has been experimentally determined as indicated, so that: 

                                            ~=3.6196E-04 m K 2

rh 

                                             '  g W
  • K-State Reactor Safety Analysis Report 4-19 Original (12104)

CHAPTER4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 Kand 600 K. Under expected conditions, the value for 127°C applies so that: r r 0 ln..CC. r m'K 

                                               - - ' =3.103e-5--

k, w Tabl e 4.7' Cl add"mg H eat Trans fler c oe ffitc1ent Temp (°K) Temp (°C) m2 K w- 1 300 27 3.457e-5 400 127 3.103e-5 600 327. 2.601e-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small. The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function. Convection Heat Transfer Coefficient TRENDLINE: y = 0.0326x + 16985 85000 R2 = 0.9976 f 75000 E 

            ~

i!!. 65000 1: lE 55000 0"' 0 

            ~

I-c: 45000 

            'i
i::

35000 25000 500 700 900 1100 1300 1500 1700 1900 Power Level (KW) Figure 4.10, Convection Hear Transfer Coefficient versus Power Level 1 h 0.0326P(watts) + 16985

  • K-State Reactor Safety Analysis Report 4-20 Original (12/04)

REACTOR DESCRIPTION

  • Core centerline temperature for the fuel rod producing the maximum heat as a function of power can be calculated as:

1 T, = T, + 0.423P[ + 3.103e-5 + 3.620e-4 + 5.186e-4] (10) 

          <                  0.0326P + 16985
c. Steady-State Mode of Operation Centerline temperature calculations were performed on a "reference core" using the model as described above for the hottest location in the core were made. The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state 500 kW operations, while analysis assumes 1.25 MW operation. A core with more than. 83 elements will distribute heat production across a larger number of fuel elements, resulting in a lower heat flux per fuel rod than calculations based on the reference core. Since actual heat production will be less than heat calculated in analysis, actual temperatures will be lower. A power level of 1.25 MW steady state power at 20°C and 100°C was assumed with the following results:

Table 4.8, Calculated Temperature Data for 1,250 kW Operation Fuel Fuel/Gap Gap/Clad Clad/Water Bulk Water °C Centerline °C Interface °C Interface °C Interface °C 503.2 229.0 37.7 21.2 20.0 582.0 307.8 116.4 100.0 100.0 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations show the effects of thermal conductivity changes are minimal. The graph also shows that fuel temperature remains below about 750 °C at power levels up to 1900 kW with pool temperature at 27 °C (300 K), and 1700 kW with pool temperatures at I 00 °C .

  • K-State Reactor Safety Analysis Report 4-21 Original (12/04)

CHAPTER4 Hot Fuel-Rod Centerline Temperature at Power (Temperature 8evation over Pool Water Tern perature) I- u -

  • 300 K - - 6 0 0 K I 100-EE!lRREEl33EIREEl33IEIREEl33IEIREEl33IEI!EfftEEl33i333:REE!33~IRRE~IE!RE~~Im
             ~

e-600

I
             ~ 500 Cll Cl.
              ~400 I-
             ~ 300
             ~ j~~~~~~~~~~~~~~~~~~~~~~~~~~~~i
             ~ 100 0

100 300 500 700 900 1100 1300 1500 1700 1900 Reactor Power (kW) Figure 4.11, Hot Fuel-Rod Centerline Temperature

  • For the analysis of critical heat flux, a single channel model was built in RELAP-5/MOD 3.3 patch 04 (Feldman 2008). A snapshot of the model is presented in Figure 4.12. It has two time-dependent volumes, enforcing the pressure boundary conditions, and two pipes, simulating the cold and hot channel connected via a single junction component of RELAP. Heat is added to the fluid by incorporating the heat structure component (simulating a fuel element) of RELAP with an appropriate axial power profile and power level. In this analysis, the power level for the B ring is at 24 kW (corresponding to an 85-element core with a ring-to-average peaking factor of 1.63).

This power level is applied to the heat structure within the single channel. The model assumes an operating pressure of 143 kPa, and an operating temperature of 322.15 K (49. l 5°C). The version of the RELAP code licensed to KSU uses PG-CHF correlation which is a state of the art best estimate CHF correlation developed by Nuclear Research institute of Rez in the Czech Republic. It is based on data in the Czech Republic data bank from 173 different sets of tube data, 23 sets of annular data, and 153 sets of rod bundle data. There are four forms of the PG-CHF correlation 'Basic', 'Flux', 'Geometry', and 'Power'. For the rod bundle it is applicable in the pressure range of0.28 MPato 18.73 MPa, for a mass flux of34.l to 7478 kg/s-m2, for 0.4-7.0 m length and for a diameter of 0.00241 to 0.07813 m. TRIGA has an operating pressure of 0.143 MPa and fuel rod length of 0.381 m, thus the operating conditions fall outside the range of the applicability of the PG-CHF correlation, and a different correlation is required to assess the 

 . departure from nucleate boiling ratio (DNBR ratio). One such correlation which is applicable for the low pressure range observed in TRIGA reactor facility is the Bernath correlation. The functional form of the Bernath correlation can be presented in the following equations .
  • K-State Reactor Safety Analysis Report 4-22 Original (12/04)

REACTOR DESCRIPTION

  • I Outlet i ~ .

Hot-leg Cold-Leg

  • + I connector Figure 4.12 - RELAP single channel model used in CHF analysis (8)
  • K-State Reactor Safety Analysis Report 4-23 Original (12/04)

CHAPTER 4 

                              ~ = D4~.6 , if Dh :<S; O.lft h
                              ~ = _!Q_ + 90, if Dh ;::: 0.1.ft D,,

hBo =film coefficient at CHF D,, =hydraulic dia.meter (ft) v =coolant velocity (ft Is) TwBo =wall temperature at burnout (° C) DH =heated diameter (ft) The RELAP simulations were performed for the hot channel, i.e., a channel with a radial peaking factor of 1.63, assuming an 85-element core load and a power of 1.25 MWth, in order to obtain the pressure, temperature, and velocity distribution at different axial locations. With these calculations and the functional form of the Bernath correlation, the axial distribution of CHF was estimated in the hot channel. The methodology adopted for this analysis is described in literature (Feldman 2008). The hot channel model was based on the smallest hydraulic diameter in the core (between the A-ring and two B-ring elements) and the highest radial peaking factor. In the KSU TRIGA, the A-ring is occupied by the central thimble, not a fuel element. Since the actual hot channel would be between two B-ring elements and a C-ring element, the real hydraulic diameter will be slightly larger and the real heat flux into the channel will be slightly lower than the values assumed in the model. Therefore, this model is conservative in this regard. The axial CHF results from the PG and Bernath heat flux models are shown in Figure 4.13 and Figure 4.14. The DNBR ratio exceeds 2.0 for all locations along the heated length of the hot channel.

  • K-State Reactor Safety Analysis Report 4-24 Original (12/04)

REACTOR DESCRIPTION

  • 5000------------------------- A A

Bernatl1-CHF PG-CHF 4000 .. _., A 

                                                                                  *
  • Heat flux A
                                                                 .. Jo. ..

A .I. A 0. 

          .,.,---                                                                        A A     A A A :A  A
  • 3000
               ~
            -~-
           ...:.o::

u...

c 2000 u

1000

r .;. *
                       &~00---0~.o-s---o-_1-o~~-o.*1-s~~o-.2-o~~o-_-2-s~~o~.3-o~--o-__*35-....

Heated Length {m) Figure 4.13 - CHF versus heated length 

                           + + Bernath-CHF
                              +
u.
  • PG-CHF 10 -**

B* 

                                                        ~

o! - .** 2 . * .. 8.oo 0.05 O.lG O.lS 0.20 0_25 0.30 0.35 Heated length {n1*) Figure 4.14 - DNBR versus heated length

  • K-State Reactor Safety Analysis Report 4-25 Original (12/04)

CHAPTER4

d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code TASCOT for transient and steady state two-dimensional conduction calculations (Ahrens 1999). For these calculations, the initial axial and radial temperature distribution of fuel temperature was based on Eqs. (9) and (10), with the peak fuel temperature set to 746 °C, i.e., a temperature rise of 719 °C above 27 °C ambient temperature. The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1 % ($3.00) pulse from zero power and a 0.7% ($1.00) pulse from power operation. In the TASCOT calculations, thermal conductivity was set to 0.18 W cm* 1 K- 1 (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cm* 1 K- 1* The convective heat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chen 1963, Collier and Thome 1994)

(9) l The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955) 

             .-
  • k f0.79
  • cP.f0.45 * ~ 0.51 fl, * (T,.. - )0.99 hh - 0.00122 0.75 T.at , (10)

[ cr o.s * µ 0.29

  • p 0.24 * (v _ v )
  • To.1s f g g v sat in which Tw is the cladding outside temperature, Tsai the saturation temperature (111.9 °C),

and Tb the coolant ambient temperature (27°C). Fluid-property symbols and values are given in Appendix B. Subscripts f and g refer respectively to liquid and vapor phases. The overall heat transfer coefficient U varies. negligibly for ambient temperatures from 20 to 60 °C, and has the value 0.21 W cm* 1 K- 1 at Tb= 27 °C. Figure 4.15 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.10 lists temperatures and heat fluxes as function of time after a 2.1 % ($3 .00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m* 1 from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 of Appendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500°C. Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse .

  • K-State Reactor Safety Analysis Report 4-26 Original (12/04)

REACTOR DESCRIPTION

  • 1000 Os 800 1

() 0...__, 2 

         <!)

600 L

J 4
        +'

nl L 8 

         <!)

Q_ 400 E 

         <!)                                                                                                  16 I-200                                                                                           32 64s 0 '---'---'---'--_._----'-----'----'---'----'___J'--.__-'---'----'--'-----'----'----'---'----'___JL-....1 0.0    0.20 0.40 0.60 0.80                    1.0     1.2       1.4       1.6       1.8       2.0    2.2 Radius (cm)
  • Figure 4. 15, Midplane Radial Variation of Temperature Within the Fuel Subsequent to a $3.00 Pulse.
  • K-State Reactor Safety Analysis Report 4-27 Original (12/04)

CHAPTER 4 Table 4.10, Heat Flux and Fuel Temperatures Following a $3.00 Pulse from Zero Power, with 27{0 C) Coolant Ambient Temperature. Q" Fuel outside Clad surface Time (s) (W m-2) CHFR Temp. (oC) Temp. (°C) 0 953 1 3.57 x10 5 4.2 781 224 2 7.34 xl0 5 2.0 683 432 4 5 1.7 574 498 8.52 x10 8 7.54 xl0 5 2.0 461 443 16 5.71 xl0 5 2.6 344 342 32 3.46 x10 5 4.3 224 218 64 5 14.4 100 84 1.04 xl0 4.6 Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by

  • (11) in which Po and 130 are the density and volumetric expansion coefficient at core inlet conditions (27°C, 0.15285 Mpa), g is the acceleration of gravity, 9.8 cm2 s- 1, l':iT is the temperature rise through the core, and L is the height of the core (between gridplates), namely, 0.556 m. The frictional pressure loss is given by (12) in which mis the coolant mass flow rate (kg s- 1) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 m2, and D1i is the hydraulic diameter, namely, 0.02127 m. The friction factor/for laminar flow through the annular area is given by 100 Re- 1 (Shah & London 1978), in which the Reynolds number is given by D,,rh I Aµ 0 in which µo is the dynamic viscosity at core inlet conditions.

Entrance of coolant into the core is from the side, above the lower grid plate (see Section 4.2.5), and the entrance pressure loss would be expected to be negligible. The exit contraction loss is given by (13)

  • K-State Reactor Safety Analysis Report 4-28 Original (12/04)

REACTOR DESCRIPTION

  • The coefficient K is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which

[~] 2 2 _ _ [ 3

  • R sin60° cos60° ]-

K= A c 11

  • R2 - 0.171, (14) where R is the radius of the opening in the upper grid plate. Equations (12) through (14), solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4. I 1.

Table 4.11, Coolant Flow Rate and Temperature Rise for Natural-Convection Cooling the TRIGA Reactor During Steady-State Operations. P (kWt) m (kg s- 1)  !).T (°C) 50 0.047 3.1 100 0.061 4.7 200 0.077 7.5 300 0.090 9.6 400 0.100 11.5 500 0.108 13.3 750 0.125 17.2 1000 0.139 20.6 1250 0.150 23.8

4. 7 Safety Limit As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 500°C; consequently, limits on fuel temperature change for cladding temperatures greater than 500°C. A maximum temperature of l 150°C (with clad< 500°C) and 950°C (with clad> 500°C) for U-ZrH (H/Zr1.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.

4.8 Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits on other operating parameters, that is, limiting safety system settings, controls the fuel temperature. The operating parameters established for the KSU TRI GA reactor are:

  • K-State Reactor Safety Analysis Report 4-29 Original (12/04)

CHAPTER 4

  • Steady-state power level
  • Fuel temperature measured by thermocouple during pulsing operations
  • Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels. As indicated in Figure 4.14, the ratio of actual to critical heat flux is at least 2.0 for temperatures less than 100°C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.

Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation. 4.8.3 Safety Margins For 1,250 kWth steady-state operations, the critical heat flux ratio remains above 2.0 for a core with 85 fuel elements and a maximum radial power peaking factor of 1.63 assuming a coolant inlet temperature of 49°C. The proposed Technical Specifications limit of 44°C on pool inlet temperature ensures that the DNBR will be at least 2.0 during steady-state operation. Limiting pool inlet water temperature to no greater than 44°C (or 37°C with an experiment installed in an interstitial flux-wire port) will ensure that the pool water does not reach temperatures associated with excessive amounts of nucleate boiling. Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746°C, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of power operation can result in a maximum hot spot temperature of 869°C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15% margin to the fuel temperature safety limit for cladding temperatures below 500°C. Analysis shows that cladding temperatures will remain below 500°C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve 500°C (invoking a 950°C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding. Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause cladding temperature to rise above 500°C, but the heat transfer simultaneously cools the fuel to much less than 950°C .

  • K-State Reactor Safety Analysis Report 4-30 Original (12/04)

REACTOR DESCRIPTION

  • If fuel rods are placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 500°C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 500°C.

4.9 Bibliography "TASCOT: A 2-D, Transient and Steady State Conduction Code for Analyhsis ofa TRIGA Fuel Element," Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C., 

 "Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III Fuel Element Using MCNP," Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrens, C.,
 A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,

Kansas State University TRIGA MkII Reactor Hazards Summary Report," License R-88, Docket 50-188, 1961. Clack, R.W., J.R. Fagan, W.R. Kimel, and S.Z. Mikhail Convective Boiling and Condensation, 3rd ed., Oxford Press, New York, 1994.Collier, J.G., and J.R. Thome, "Bubble Dynamics and Boiling Heat Transfer," AIChE Journal 1, 532 (1955). Forster, H.K., and N. Zuber, Theory and Design ofModern Pressure Vessels, 2d. ed., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F., 

 "On the Relevance of the Vapour-Liquid Exchange Mechanism for Sub-Cooled Boiling Heat Transfer at High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winfrith, 1978. Ivey, H.J. and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, J. H. and V. K. Dhir, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, General Dynamics, General Atomic Division Report GA-3618, November 1962. Merten, U., et al.,

MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987 .

  • K-State Reactor Safety Analysis Report 4-31 Original (12/04)

CHAPTER 4 "Laminar Forced Convection in Ducts," p. 357, Academic Press, New York, 1978. Shah, R.K., and A.L. London, "The U-Zr-Hx Alloy: Its Properties and Use in TRIGA Fuel," Report E-117-833, General Atomics Corp., 1980. Simnad, M.T. 

 "Safety Analysis Report, TRIGA Reactor Facility, Nuclear Engineering Teaching Laboratory, University of Texas at Austin, Revision 1.01, Docket 50-602, May, 1991.
  • K-State Reactor Safety Analysis Report 4-32 Original (12/04)
  • Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the Univer~ity of Texas Reactor Facility (UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).
  • The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse is very similar to the power distribution shown in Figure 4A. l. This initial steep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.

Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [1961] for transient heating of ribbons in 100°F water, showed burnout fluxes of 0.9 to 2.0 Mbtu fr 2 hr- 1 for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft- 2 hr- 1* It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation. A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A.1 and 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, was used in order to obtain an upper limit for the clad temperature rise. The model used the data of McAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760°C clad temperature for a peak initial fuel temperature of 1000°C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure'4A.5 and the heat flux into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor 4.A-1 Original (9/02) Safety Analysis Report

CHAPTER 4 APPENDIX A assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion [9] for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470°C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400°C to 500°C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965]. I. 2 I. 1 1.0 0.9

  • a.a~~~~~~~~~~~~~~~~~~~~~~~

0 0.1 0.2 0.3 0.4 RADIUS (IN.) 0.5 0.6 0.7 Figure 4A.1. Representative Radial Variation of Power Within the TRIGA Fuel Rod I.I 0.8 1.0 0.9 

                 ;::;   0.8 L;:"

0.7 0.6 0.5 0 2 4 6 7 8 AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.) Figure 4A.2, Representative Axial Variation of Power Within the TRIGA Fuel Rod .

  • K-State Reactor Safety Analysis Report 4.A-2 Original (9/02)

REACTOR DESCRIPTION 

                                                \
                                                  \              CURVE BASED ON
                                                    \
                                                      \          DATA OF ELLIO/.

TW-TSAT (*f) Figure 4A.3, Subcooled Boiling Heat Transfer for Water. 1800 1700 [LAPSED TIHE FROH END OF PULSE 1600 

                  ~

i= 1500 

                   ~
                   ~

1400 1300 1200 0.1 0.2 0.3 0.4 o.s o.6 0. 7 0.8 0 RADIUS (IN.) Figure 4A.4, Fuel Body Temperature at the Midplane of a Well-Bonded Fuel Element After Pulse.

  • K-State Reactor Safety Analysis Report 4.A-3 Original (9/02)

CHAPTER 4 APPENDIX A 106.---,..--.-....-.r-r--..----.--.-.-..--r-....--r--.-.--.-----.----.-~--.--.,...-..-.-, ONSET OF _!_l PEAK HEAT FLUX NUCLEATE 801 LI NG N 

                       ~
                       ...."'     105 ONSET or STABLE
                        ...c I Fii.Joi llOlllNG
                        ....x
                        ....u     10 4
                        ...c 103 0.001               0.01                     0.1                   1.0                   10                       100 ELAPSED TIME FROM ENO OF PllLSE (SEC)

Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse.

  • '0. 000 ,..........,.-..,......,........,.--,--..-.,......,.....,....-..--.;...,r--i,..-,-..-......,--r-r"T"T--.--..-.,.-,-,
                \000 100 - - - - -

CLAO OUTER SURFACE TEMP j io l.___L._..L..L.1...l.-..L-L-.LJ~-L-__J--l-'-L---l--'--'-......... ,0--:-._.__..............~,oo 0.001 0.01 0. I 1.0 ELAPSED TIME FROM END OF PULSE (SEC) Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element.

  • K-State Reactor Safety Analysis Report 4.A-4 Original (9/02)

REACTOR DESCRIPTION

  • The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.

This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected. Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 1962] which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000°C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural

  • convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [1960]

which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate. The transient thermal calculations were performed using effective gap conductances of 500, 375, and 250 Btu ft- 2 hr- 1 °F- 1. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu ft- 2 hr- 1 °F- 1 is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRI GA fuel. The surface heat flux at the mid plane of the element is shown in Figure 4A. l 0 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu ft- 2 hr- 1 °F- 1 according to the more conservative Bernath correlation at a local water temperature approaching saturation. K-State Reactor 4.A-5 Original (9/02) Safety Analysis Report

CHAPTER 4 APPENDIX A This analysis indicates that after operation of the reactor at steady-state power levels of 1 MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000°C is conservatively estimated to be 470°C. As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470°C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 1000°C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is 1150°C would not produce a stress in the clad in excess of its ultimate strength. TRI GA fuel with a hydrogen to zfrconium ratio of at least 1.65 has been pulsed to temperatures of about l l 50°C without damage to the clad [Dee et al. 1966]. 7 

       ......                           ELAPSED TIME FROM IN I-u..                            *END OF PULSE
  • 0.2.47 SEC
       'I         6 a:
c:
i I-al IJ\

I 5 0 x 

          ...J u.. 4 t-c:t
c 3

7 8 9 10 11 12. 13 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A. 7, Surface Heat Flux Distribution for Standard Non-Gapped {hgap= 500 Btu/h ft 2 °F) Fuel Element After a Pulse .

  • K-State Reactor Safety Analysis Report 4.A-6 Original (9/02)

REACTOR DESCRIPTION 8 7 

              .;,         6
i:
               ~

era 5' 

            "'0
                                 -----.'A~CCTUAL         HEAT FLUX
               ....       It
i:

3 ELAPSED TIHE FROH ENO OF PULSE IS 0.311t SEC 2 7 8 9 10 11 12 13 15 DISTANCE FROH BOTTOM OF FUEL (IN.) Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 375 Btu/h ft2 °F) After a Pulse. B 7 N I- 

                "-    . 6
                -a:
r::
                 =>

I-m 5 U\ 

               . 0                                      ELAPSED TIME FROM ENO OF PULSE IS 0.440'SEC
                   )C      4 x
                *~
                   =>

I-er: 3

c 1

1 7 8 9 10 11 12 13 14 15 DISTANCE FROM BOTTOM OF FUEL (IN.) Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap= 250 Btu/h ft 2 °F ) After a Pulse.

  • K-State Reactor Safety Analysis Report 4.A-7 Original (9/02) j

CHAPTER 4 APPENDIX A EFFECTIVE HEAT TRANSFER COEFFICIENT IN GAP, BTU/HR-FT 2 -°F 500 

         ...J i..

w

c w

u i.. 

          !.,, 101+

FLOW VELOCITY a I FT/SEC GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT END OF PULSE (I.E. TIME= ZERO) 1031.-~~.....J...~~~l--~l..--'-....J-~----"'--~~...J-~..;.i....--'-...J 0.01 0.1 1.0 ELAPSED TIHE FROM END OF PULSE (SEC} Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse.

  • K-State Reactor Safety Analysis Report 4.A-8 Original (9/02)

REACTOR DESCRIPTION

  • Bibliography "A Theory ofLocal Boiling Burnout and Its Application to Existing Data, " Heat Transfer -

Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56, No. 20.Bernath, L., Research in Improved TRIGA Reactor Pe1formance, Final Report, General Dynamics, General Atomic Division Report GA-5786, October 20, 1964. Coffer, C.O., et al., Characteristics of Large Reactivity Insertions in a High Performance TRIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al. Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dee, J.B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, et al., A Study ofthe Mechanism of Boiling Heat Transfer, JPL Memorandum No. 20-88, March 1, 1954.Ellion, M.E., Thermal Conductance ofMetallic Surfaces in Contact, USAEC NY0-2130, May, 1959.Fenech, H., and W. Rohsenow, An Analysis ofa Thermal Contact Resistance, Trans. ANS 5, p. 476, 1962.Fenech, H., and J.J. Henry, "Thermal Conductance Across Metal Joints, "Machine Design, Sept. 15, 1960, pp 166-172. Graff, W.J. Heat Transmission, 3rd Ed., McGraw-Hill, 1954McAdams, -W.H .. MNRC, McClellan Nuclear Radiation Center Facility Safety Analysis Report, Rev. 2, April 1998. Heat, Mass and Momentum Transfer, Prentice-Hall, 1961, pp 231-232.Rohsenow, W., and H. Choi, "Quarterly Technical Report SPERT Project, April, May, June, 1964, "ISO 17030. Spano, A. H., 

   "The Effect ofSubcooled Liquid on Film Boiling," Heat Transfer 84, 149-156, (1962).Sparrow, E.M. and R.D. Cess, "Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis, " Technical report, Argonne National Laboratory, 2008, E.E. Feldman .

K-State Reactor 4.A-9 Original (9/02) Safety Analysis Report I

CHAPTER 4 APPENDIX A RELAP5/mod3.3 Code Manual Volume 1: Code structure, system models, and solution methods. 

 "Prediction of departure from nucleate boiling for an axially non-uniform heat ux distribution."

Journal of Nuclear Energy 21 (3): 241-248, 1967, L.S. Tong .

  • K-State Reactor Safety Analysis Report 4.A-10 Original (9/02)

REACTOR DESCRIPTION

  • "Onset ofStable Film Boiling and the Foam Limit," Int. J. Heat and Mass Transfer 6, 987-989, (1963). Speigler, P., et al.,

UTA, University of Texas at Austin TRIGA Reactor Facility Safety Analysis Report, Docket 50-602, Rev. 1.01, May 1991. 

 "Hydrodynamic Aspects ofBoiling Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959.

Zuber, W .

  • K-State Reactor Safety Analysis Report 4.A-11 Original (9/02)
  • Tpool Pt,16[11 Appendix B Water Properties at Nominal Operating Conditions Data for 16 Feet of Water over tile Core ht,16[ 11 hg,16111 P 9 ,15111 Tsat,16111 q"sat, 15!3! q"sub[41 oc kg m- 3 kJ kg- 1 kJ kg- 1 kg m- 3 oc wm- 2 Wm-2 15 999.21 950.00 47.79 465.10 2692.64 0.85 110.89 1553.842 7239.19 20 998.32 950.01 47.74 465.05 2692.63 0.85 110.88 1552.078 6931.74 25 997.16 950.02 47.69 465.01 2692.59 0.85 110.87 1549.496 6622.60 30 995.75 950.03 47.62 464.95 2692.59 0.85 110.86 1547.118 6311.82 35 994.12 950.04 47.54 464.89 2692.57 0.85 110.84 1543.981 5999.91 40 992.29 950.06 47.46 464.81 2692.54 0.85 110.83 1540.446 5688.24 45 990.27 950.07 47.36 464.73 2692.51 0.85 110.81 1536.512 5376.29 50 988.07 950.09 47.25 464.64 2692.48 0.85 110.78 1532.205 5064.36 55 985.70 950.11 47.14 464.54 2692.45 0.85 110.76 1527.561 4753.90 60 983.18 950.12 47.02 464.44 2692.41 0.85 110.74 1522.575 4444.85 65 980.50 950.14 46.89 464.33 2692.37 0.85 110.71 1517.255 4136.85 70 977.69 950.17 46.76 464.21 2692.33 0.84 110.68 1511.666 3830.73 75 974.74 950.19 46.62 464.09 2692.29 0.84 110.65 1505.778 3526.89 80 971.66 950.21 46.47 463.96 2692.24 0.84 110.62 1499.613 3225.47 85 968.45 950.23 46.32 463.83 2692.19 0.84 110.59 1493.199 2926.81 90 965.12 950.26 46.16 463.69 2692.15 0.84 110.56 1486.527 2631.05 95 961.68 950.29 45.99 463.55 2692.09 0.84 110.53 1479.626 2338.47 97 960.27 950.30 45.92 463.49 2692.07 0.84 110.51 1472.944 2216.11 99 958.84 950.31 45.86 463.43 2692.05 0.84 110.50 1466.058 2095.18 Data for 13 Feet of Water over tile Core Tpool ht,13[ 11 hg,131 11 Pg,13111 Tsat,13111 q"sa1,13!3l q"subl41 oc kJ kg- 1 kJ kg-1 kg m- 3 oc Wm-2 wm- 2 15 999.21 951.43 38.83 457.21 2689.85 0.80 109.03 1513.00 6964.74 20 998.32 951.43 38.79 457.18 2689.84 0.80 109.02 1511.32 6857.12 25 997.16 951.44 38.75 457.13 2689.82 0.80 109.01 1509.15 6543.62 30 995.75 951.45 38.69 457.09 2689.80 0.80 109.00 1505.85 6229.30 35 994.12 951.46 38.63 457.03 2689.78 0.80 108.99 1503.13 5913.58 40 992.29 951.47 38.56 456.96 2689.76 0.80 108.97 1500.16 5597.21 45 990.27 951.49 38.48 456.89 2689.74 0.80 108.96 1496.38 5281.90 50 988.07 951.50 38.39 456.82 2689.71 0.80 108.94 1492.25 4966.66 55 985.70 951.51 38.30 456.73 2689.68 0.80 108.92 1487.78 4652.39 60 983.18 951.53 38.20 456.64 2689.65 0.80 108.90 1482.99 4339.60 65 980.50 951.55 38.10 456.55 2689.61 0.80 108.87 1477.90 4027.94 70 977.69 951.57 37.99 456.45 2689.58 0.80 108.85 1472.52 3718.83 75 974.74 951.58 37.88 456.35 2689.54 0.80 108.83 1466.86 3412.07 80 971.66 951.60 37.76 456.24 2689.50 0.80 108.80 1460.95 3107.29 85 968.45 961.62 37.63 456.12 2689.46 0.80 108.77 1458.59 2812.63 90 965.12 951.64 37.50 456.01 2689.42 0.79 108.75 1448.37 2506.90 95 961.68 951.67 37.37 455.89 2689.38 0.79 108.72 1441.74 2211.19 97 960.27 951.68 37.31 455.84 2689.36 0.79 108.70 1435.27 2087.92 108.69 1428.60 1966.16 99 958.84 951.68 37.26 455.78 2689.34 0.79 K-State Reactor 4.B-1 Revised 05/01/17 Safety Analysis Report

REACTOR DESCRIPTION

  • T oc 15 20 Common Data Patm kPa 99.83 99.83 Cp[1]

kJ kg-1 k- 1 4.23080 4.23080 CJ Nm-1 0.07149 0.07120 25 99.83 4.23080 0.07083 30 99.83 4.23080 0.07039 35 99.83 4.23070 0.06989 40 99.83 4.23070 0.06932 45 99.83 4.23070 0.06869 50 99.83 4.23070 0.06800 55 99.83 4.23060 0.06727 60 99.83 4.23060 0.06649 65 99.83 4.23060 0.06566 70 99.83 4.23050 0.06480 75 99.83 4.23050 0.06390 80 99.83 4.23050 0.06297 85 99.83 4.23040 0.06201 90 99.83 4.23040 0.06102 95 99.83 4.23040 0.06001 97 99.83 4.23030 0.05898 99 99.83 4.23030 0.05793 NOTE[1}: 1967 ASME (IFC) Steam Tables & IAPWS-IF97 NOTE[2}:kPa =Heigth(ft) *12(inlft) *0.0254(meterslin) *Density(kglm 3) *9.8066511000 NOTE[3}: qSAT" 

                 =0.I49*pg 0.5 . ( hg,sat-hf,sat.
                                                ) (
                                                    *g*<J* {Pt-Pg })Y.

NOTE[4']* " - " . 1 0

  • qsub -qSAT

( 

                          +
  • l*(P1J .

Pg Y. cp.f *(TsAT -Tsub) hg,sat - h /,sat J NOTE:[5}: 

 <J = 1.000E-11*T 4 +7.370E-09
  • T 3 -1.969E-06
  • T 2 + 4.709E-06
  • T + 7.1833E- 02 K-State Reactor 4.A-13 Original (9/02)

Safety Analysis Report

  • Table of Contents TECHNICAL SPECIFICATIONS I. DEFINITIONS ................................................................................................................. TS-I
2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Element Temperature Safety Limit ...................................................................... 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 Limiting Safety System Settings ................................................................................ TS-I 0 2.2.1. Applicability ...................................................................................................... TS- I 0 2.2.3. Objective ............................................................................................................ TS-10 2.2.4. Specification ....................................................................................................... TS- I 0 2.2.5. Actions ............................................................................................................... TS- I 0 2.2.6. Basis .................................................................................................................. TS-10
3. LIMITING CONDITIONS FOR OPERATIONS ......................................................... TS-11 3.1 CORE REACTIVITY ................................................................................................ TS-II 3.1.1. Applicability ...................................................................................................... TS-11 3.1.3. Objective ............................................................................................................ TS-II 3.1.4. Specification ....................................................................................................... TS-11 3.1.5. Actions ............................................................................................................... TS-12 3.1.6. Basis .................................................................................................................. TS-13 3.2 PULSED MODE OPERATIONS .............................................................................. TS-13 3.2.1. Applicability ...................................................................................................... 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 MEASURING CHANNELS ...................................................................................... 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. Applicability ...................................................................................................... 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 GASEOUS EFFLUENT CONTROL ......................................................................... TS-20 3.5.1. Applicability ...................................................................................................... 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 LIMITATIONS ON EXPERIMENTS .......................................................................... TS-22 3.6.1. Applicability ...................................................................................................... TS-22 3.6.3. Objective ............................................................................................................ TS-22 K-State Reactor TS-1 Original (9tG+-4/1744)

L_ _____

  • TECHNICAL SPECIFICATIONS 3.6.4. Specification ....................................................................................................... TS-22 3.6.5. Actions ............................................................................................................... TS-22 3.6.6. Basis .................................................................................................................. TS-23 3.7 FUEL INTEGRITY ................................................................................................... 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 WATER ......................................................................................... 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 Maintenance 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 REACTIVITY ................................................................................................ 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 MEASURING CHANNELS ...................................................................................... 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 LIMITATIONS ON EXPERIMENTS ........................................................................ TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis .................................................................................................................. TS-34
4. 7 FUEL 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 WATER ....................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor TS-2 Original (91G+4/1744)
  • TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36 4.8.3. Basis .................................................................................................................. TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................................................................... TS-37 4.10.3. Basis ................................................................................................................ TS-37
5. DESIGN 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 BUILDING ............................................................................................. 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 EXPERIMENTS ......................................................................................................... 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. ADMINISTRATIVE CONTROLS ................................................................................. TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL.. ........................ TS-44 6.2 REVIEW AND AUDIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS .................................................. TS-47 6.5 EMERGENCY PLAN AND PROCEDURES ........................................................... TS-48 6.6 OPERATOR REQUALIFICATION .......................................................................... TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERATING RECORDS ............................................................................ TS-49 6.11 REPORTING REQUIREMENTS ........................................................... TS-50 K-State Reactor TS-3 Original (9,LG+.4/1744)
  • 1. DEFINITIONS TECHNICAL SPECIFICATIONS 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 .
  • ANNUAL CHANNEL
  • 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 12 months, not to exceed 15 months 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 K-State Reactor TS-4 Original (9fG+.4/1744)

  • 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 EXPERlMENT An EXPERlMENT is (I) any apparatus, device, or material placed in the reactor core region (in an EXPERlMENTAL FACILITY associated with the reactor, or in line with a beam ofradiation emanating from the reactor) or (2) any in-core operation designed to measure reactor characteristics.

EXPERlMENTAL 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 Frings, as many as three contiguous fuel-element positions). IMMEDIATE Without delay, and not exceeding one hour. NOTE: IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action INDEPENDENT INDEPENDENT Experiments are those not connected by a mechanical, EXPERlMENT 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 EXPERlMENT is one that may be moved into, out-of or EXPERlMENT near the reactor while the reactor is OPERA TING. NONSECURED NONSECURED Experiments are these that should not move while the EXPERlMENT reactor is OPERATING, but are held in place with less restraint than a secured experiment. K-State Reactor TS-5 Original {9fG.74/1744)

  • OPERABLE TECHNICAL SPECIFICATIONS 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 OPERA TING 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, reactor power may be increased on a period of much less than l second by motion of the transient control rod. 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 m 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 ofless than 60 lb. K-State Reactor TS-6

  • SECURED EXPERIMENT TECHNICAL SPECIFICATIONS 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 applicable titles of sections in individual Technical Specifications K-State Reactor TS-7 Original (fhlG+.4/1744)

  • 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 l 150°C. (2) Steady state fuel temperature shall not exceed 750°C. 2.1.4 Actions

  • CONDITION A Stainless steel clad, high-hydride fuel element temperature exceeds l 150°C.

REQUIRED ACTION A.I Establish SHUTDOWN condition AND COMPLETION TIME A.I IMMEDIATE OR Fuel temperature exceeds 75 0°C in steady state A.2 Report per Section 6.8 conditions A.2 Within 24 hours 2.1.5 Bases Safety Analysis Report, Section 3.5.1 (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged. NUREG 1282 identifies the safety limit for the high-hydride (ZrH11) 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, l 50°C. A change in yield strength occurs for stainless steel cladding temperatures of 500°C, but there is no scenario for fuel cladding to achieve 500°C while submerged; consequently the safety limit during reactor operations is l,150°C. K-State Reactor TS-8 Original (9fG+4/1744)

  • 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-117-833. Damage mechanisms include fission recoils and fission gases, strongly influenced by thermal gradients. Operating with maximum long-term, steady state fuel temperature of750°C does not have significant time- and temperature-dependent fuel growth . K-State Reactor TS-9 Original (flfG-74/1744)

  • 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 I (I) I Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation I 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A.I IMMEDIATE A. I Reduce power to less than 1,250 kW (th) A. Steady state power level OR exceeds 1,250 kW (th) A.2. Establish REACTOR A.2. IMMEDIATE 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 (9/G+4/1744)

  • 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 OPERA TING 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) I. 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:
1. The highest worth control rod is fully withdrawn (2)
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. I ENSURE REACTOR A. I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meetLCO operations K-State Reactor TS-11 Original (9fG-74/1744)

  • TECHNICAL SPECIFICATIONS B.1.a ENSURE control rods B.1 IMMEDIATE fully inserted AND B.1.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.1.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 meetLCO 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 l,150°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 I 07 kW leads to a maximum fuel temperature of 869°C, well below the safety limit. The limiting SlillTDOWN 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. K-State Reactor TS-12 Original (MH4/1744)

  • 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.I 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 IMMEDlA TE 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 I3.4) that demonstrate fuel temperature limits are met during potential accident scenarios under extremely conservative conditions of analysis. K-State Reactor TS-13 Original (9fm4/1744)

  • 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 (I) 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 I: MINIMUM MEASURING CHANNEL COMPLEMENT Minimum Number Operable MEASURING CHANNEL STEADY STATE PULSE MODE MODE Reactor power leveJl 1l 2 Primary Pool Water Temperature I Reactor Bay Differential Pressure I Fuel Temperature I 22 foot Area radiation monitor I 0 or 12 foot Area monitor I Continuous air radiation monitor12l I EXHAUST PLENUM radiation monitor12l I NOTE[!]: 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.I Restore channel to operation A. I.I IMMEDIATE A. I Reactor power channels not OPERATING (min 2 OR for STEADY STATE, 1 A.1.2 ENSURE reactor is A.1.2 IMMEDIATE PULSE MODE) SHUTDOWN K-State Reactor TS-14 Original (MH4/1744)

  • CONDITION REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less A.2. IMMEDIATE AND than 90% of required operating value A.2.2 Enter REACTOR SECURED mode B. Primary water temperature, B.l Restore channel to operation A. l IMMEDIATE reactor bay differential OR pressure or fuel temperature CHANNEL B.2 ENSURE reactor is A.2 IMMEDIATE not operable SHUTDOWN C.l Restore MEASURING C.l IMMEDIATE CHANNEL OR C.2 ENSURE reactor is shutdown C.2 IMMEDIATE OR C. 22 foot Area radiation monitor is not C.3 ENSURE personnel are not C.3 IMMEDIATE OPERATING on the 22 foot level
  • OR C.4 ENSURE personnel on 22 foot level are using portable survey meters to monitor dose rates D.l Restore MEASURING C.4 IMMEDIATE D.l 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 K-State Reactor TS-15 Original (9JG+.4/1744)
  • CONDITION REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME E.l Restore MEASURING E.l 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.l Restore MEASURING F.l IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown F.2. IMMEDIATE
  • F. Exhaust plenum radiation monitor is not OPERATING OR F.3.a ENSURE continuous air radiation monitor is OPERATING AND F.3.a. IMMEDIATE F.3.b Restore MEASURING F.3.b Within 30 days CHANNEL G.l Do not perform a reactor G.l 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 G.2 IMMEDIATE check on the startup channel 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 (8fG+.4/1744)

  • 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 11 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 IO 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. K-State Reactor TS-17 Original (9fG-7.4/17+4)

  • 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 (I) The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLE CONTROL RODS (STANDARD) are capable of 90% of full reactivity insertion from the. (2) -*--{ Formatted Table fully withdrawn position in less than 1 sec. A minimum of three CONTROL RODS must be OPERABLE. Ino12erable CONTROL ill RODS must be fully inserted. TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum Function Reauired OPERATING Mode Safety System Channel Number STEADY PULSE or Interlock Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar I Scram YES YES CONTROL ROD Prevent withdrawal of standard (STANDARD) position 1 rods in the PULSE MODE NA 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 Al. IMMEDIATE to operation A. Any required SAFETY SYSTEM CHANNEL or OR interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN K-State Reactor TS-18 Original (WG+4/1744)

  • CONDITION REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME B.l ENSURE inoRerable control Bl. IMMEDIATE rod is fully inserted B. A control rod is not OR OPERABLE.

B2. IMMEDIATE B.J2 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 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 (ST ANDARD) 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. InoRerable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. ORerating with a fully-inserted control rod may cause Rower Reaking to shift. however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR ChaRter 13 is still bounding, and the reduction in maximum core Rower by having an inoRerable control rod fully inserted means that the highest temRerature in any fuel element with a fully-inserted inoperable control rod will be lower than the highest temRerature in the B-ring with all rods withdrawn. Therefore the reactor can be safely ORerated with an inoperable control rod Rrovided that the rod is fully inserted into the core. K-State Reactor TS-19 Original (91-G+4/1744)

  • 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 A.I ENSURE reactor is SHUTDOWN OR A.2.a Do not OPERA TE in the COMPLETION TIME A. I IMMEDIATE 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 (WG-74/1744)

  • CONDITION REQUIRED ACTION TECHNICAL SPECIFICATIONS 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- 1 (2.0lxl0-5 µCi mJ-1), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10*3 µCi 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, ifrequired. 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. 90 Sr activity available for release from fuel rods previously used at other facilities is estimated to be at most about 4 times the ALL 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 (91G+.4/1744)

  • 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 (3) reactor pool and core area

  • 3.6.4 Actions CONDITION A. INDEPENDENT EXPERIMENT worth is REQUIRED ACTION A.I ENSURE the reactor is SHUTDOWN AND COMPLETION TIME A.1 IMMEDIATE greater than $2.00 A.2 Remove the experiment A.2 Prior to continued operations C.I ENSURE the reactor is C.l 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 K-State Reactor TS-22 Original (9m+-4/1744)
  • 3.6.5 Bases TECHNICAL SPECIFICATIONS 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 . K-State Reactor TS-23 Original (9!G+4/1744)

  • TECHNICAL SPECIFICATIONS
3. 7 Fuel Integrity 3.7.l 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 TRI GA reactor. 3.7.3 Specifications Fuel elements in the reactor core SHALL NOT be elongated more than 1/8 in. over (I) 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 in. over manufactured length, or bent laterally greater than 1/8 in.

Do not insert the fuel element into the upper core grid plate. IMMEDIATE 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 TRI GA reactor. K-State Reactor TS-24 Original (g,LG;Z.4/1744)

  • 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 Watef tem13efaffife at the ei<it ehhe feaetef 13ee!Peel iHletBulk water tem1:1erature SHALL (1) NOT exceed 44$-°C (l 11J.°F)l;JQ0 I' with flew thfe11gh the J3FimaFy eleaH1113 lee13 (2) Water conductivity SHALL be less than 5 µmho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core

  • ill 3.8.4 Peel iHletBulk water tem1:1erature SHALL NOT exceed 4-037°C (99-l-04°F) with an exgeriment installed in an interstitial flux wire QOrt.

Actions CONDITION REQUIRED ACTION A.I ENSURE the reactor is COMPLETION TIME A.I IMMEDIATE SHUTDOWN AND A. Water tempeFatHfe at the A.2 See11re flev: thre11gh the A.2 IMMEDIATE eiHtffiletB ulk water demiHerali20er tem1:1erature ef the reaeter j380l-_exceeds +.W4M°F.G_ MID A.J.l Reduce bulk water A.3 IMMEDIATE temperature to less than 

                                            -l-J.()4Mof'.C K-State Reactor                                    TS-25                     Original (9,LG+4/1744)
  • CONDITION REQUIRED ACTION TECHNICAL SPECIFICATIONS COMPLETION TIME B.l ENSURE the reactor is B.l IMMEDIATE SHUTDOWN B. l;1,1at:eF temf eFaffife at: the iHlet sf the FeaetsF AND . ~--( Formatted: Centered ooel-Bulk water tem12erature exceeds B.2 Reduce bulk water tem12erature B.2 IMMEDIATE . -~--( Formatted: Indent: Left: O" to less than-4-0 37°C.

4-037°C with an ex12eriment installed in an interstitial flux wire port. OR .--(Formatted: Centered B.3 Remove ex12eriment from flux B.3 IMMEDIATE wire QOrt B.l ENSURE the reactor is B.l IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 µmho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 µmho/cm C. l ENSURE the reactor is C.l 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). While the integrity of the ion exchange resin reguires water tem12erature to remain below 54.4°C, it is necessary to maintain water tem12erature below 4M°C to ensure that the de12arture from nucleate boiling ratio CDNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an ex12eriment into an interstitial flux wire QOrt between fuel elements necessitates a further reduction in water tem12erature to a maximum of 4-037°C in order to 12reclude excessive nucleate boiling of the water eHsuFe a DNBR sf at least
 ;&.-0.

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. K-State Reactor TS-26 Original (91G+4/1744)

  • TECHNICAL SPECIFICATIONS 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 . K-State Reactor TS-27 Original (9tQ.7.4/1744) l__

  • 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. K-State Reactor TS-28 Original (94l-+4/1744)

  • TECHNICAL SPECIFICATIONS 3.10 Maximum Steady State Power 3.9.1 Applicability This specification applies to operations in STEADY STA TE MODE.

3.9.2 Objective The objective is to ensure that the reactor has adeguate margin to critical heat flux CCHF) and operates below the Limiting Safety System Setting of 1.250 kWth. 3.9.3 Specifications ill Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY STATE MODE. 

                                                                                                      . * -f Formatted: Centered
                                                                                                               ' Formatted Table ill    A reguired reactor power level scram is set to a value no greater than 1.250 kWth.

3.9.4 Actions CONDITION REOUIRED ACTION COMPLETION TIME A. Thermal power - - - Formatted: Numbered + Level: 1 + Numbering Style: A, B, exceeds 1.050 kWth in Reduce power to a level no greater C, ... + Start at: 1 +Alignment: Left +Aligned at: 0.25" + IMMEDIATE "-, Indent at: 0.5 STEADY STA TE than 1.050 kWth. MODE ',,i Formatted: Indent: Left: 0.03" B. A reguired reactor power level scram is set to a value above B.I SHUT DOWN the reactor. AND B.I. IMMEDIATE . - - - ( Formatted: Centered AND *~ ---[ Formatted: Centered 1.250 kWth or above the maximum readable B.2 Adjust reactor 12ower level B.2. Prior to resuming value on a reguired scram setpoint to a readable value operations. channel.~ less than or egual to 1.250 kWth. 3.9.5 Bases The reactor control panel instrumentation is designed to measure up to 1.000 kWth of thermal *---( Formatted: Left 12ower. The Limiting Safety System Setting ensures that automatic protective functions. i.e .. high '------------------- power scrams. are set to no greater than 1.250 kWth. However, by specifying the maximum OPERA TING 12ower level as 1.000 kWth in STEADY STATE MODE. the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum 12ower readable on the reactor console instruments. Action to reduce power is not reguired until power exceeds I 050kWth in STEADY ST A TE MODE to allow for slight variation in power level that is typical during normal operation.~ K-State Reactor TS-29 Original (MH-4/1744)

  • 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 4.1.3 Basis BIENNIAL 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.

K-State Reactor TS-30 Original (9/07 4/1744)

  • 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 . K-State Reactor TS-31 Original (9m+4/1744)

  • TECHNICAL SPECIFICATIONS 4.3 MEASURING CHANNELS 4.3.l 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 I 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. K-State Reactor TS-32 Original (f),lQ.;Z.4/1744)

  • 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 ofless than ANNUAL 1 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-33 Original (flfG+4/1744)

  • 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 . K-State Reactor TS-34 Original (9tG+4/1744)

  • 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.

K-State Reactor TS-35 Original (fltG+4/1744)

  • 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 Measure and record experiment worth of the EXPERIMENT experiment where absolute (where the absolute value of the estimated worth is greater than value of the estimated 

 $0.40).

worth is greater than $0.40 4.6.3 Basis

  • These surveillances allow determination that the limits of3.7 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).

K-State Reactor TS-36 Original (WG-7-4/1744)

  • TECHNICAL SPECIFICATIONS 4.7 Fuellntegrity 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 of3.00$

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. K-State Reactor TS-37 Original (MH-411744)

  • TECHNICAL SPECIFICATIONS 4.8 Reactor Pool Water This specification applies to the water contained in the KSU TRI GA 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. K-State Reactor TS-38 Original (91G+4/1744)

  • 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. K-State Reactor TS-39 Original (WG+4/17+4)

  • 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 J).-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 lo four elements with !lreater than 9.0 weight percent uranium rnav be inslulleJ in the core. These elements 1113v onlY be nluced in the E- and F-rings of the core lattice. and rnav not be acijaeeffi te eeHtrel rncis ef watef ehar.nels.located in the following positions: E2,E4.E5.E6.E20.E21.E22,E24,Fl,F2.F30. 5.1.4 Bases These types of fuel elements have a long history of successful use in TRIGA reactors. Calculations show that I2'Y.,-loacl fuel in the E- ancl F-rings will not cxccccl the temperature of' 8'Yo-loacl instrumented .:lernents in the 8-ring. Aclditionullv the po\,cr peaking ancl fission product im*cntorY assumptions in the SAR will not b-o challe1rncd lw 12% fuel in the F- ancl F-rings. Locul power and ternpcrnture peaking effects during pulsing are avoided bv prnhibiting placement of the 12,*0-loaQ_ll1el_!lCl\L_l.\'ater <irn:l c_Qntrnl rod channels. K-State Reactor TS-40 Original (!;lm+-4/1744)

  • 5.2 Reactor Fuel and Fueled Devices in Storage TECHNICAL SPECIFICATIONS 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 kcrr 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.

K-State Reactor TS-41 Original (0074/1744)

  • 5.3 Reactor Building TECHNICAL SPECIFICATIONS 5.3.l 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 ofrelease 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. (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.

K-State Reactor TS-42 Original (WG+4/1744)

  • 5.4 Experiments TECHNICAL SPECIFICATIONS 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 ofreactor 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 ofradioactivity 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.

K-State Reactor TS-43 Original (WG+4/1744)

  • TECHNICAL SPECIFICATIONS
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 react!Vlty 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 . K-State Reactor TS-44 Original (91m4/1744)

  • 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. President Kansas State University Vice President for Provost Administration Chief Academic and Finance Officer Director Dean Division of Public Safety College of Engineering University Department of Head, Department of Police Environmental Mechanical & Nuclear Department Health and Safety Engineering University Manager, KSU Radiation Safety Nuclear Reactor Officer Facility Reactor Supervisor Reactor Safeguards Committee Reactor Operators 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-45 Original (9!W4/1744)

  • 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 OPERA TING 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-46 Original (9/G+4/17~)

  • 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 can, 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 wilJ be a Reactor Safeguards Committee which shall review TRJGA 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 JO 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 JO 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-47 Original (9/G.f.4/1744)
  • 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-48 Original (9iG-74/1744)

  • 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:

I. 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 OPERA TING 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-49 Original (MH'.4/1744)

  • 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: I . 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: I. 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 ifthe test K-State Reactor TS-50 Original (91-G+-4/1744)
  • TECHNICAL SPECIFICATIONS or experiment can be accomplished without obtaining prior NRC approval via license amendment pursuant to IO 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 ofrecurrence. 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: K-State Reactor TS-51 Original (Mf-74/1744)

  • 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 through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.

Surveillance Requirements must be met for all equipment/components/conditions. to be considered operable. Failure to perform a surveillance within the required time 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 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:

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-52 Original (9,lG.74/1744)

  • 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;
  • 6.11
6. updated, corrected, and as-built drawings of the facility .

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 hours 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-53 Original (9/.G+4/1744)
  • 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: I. 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 OPERA TING 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 OPERA TING and at intervals not to exceed 12 months, thereafter, providing the following information:

I. 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 IO 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; K-State Reactor TS-54 Original (9fG+4/1744)
  • 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 ofradiation and contamination surveys performed within the facility .

K-State Reactor TS-55 Original (9fG.7-4/1744)

TECHNICAL SPECIFICATIONS Table of Contents

1. DEFINITIONS ................................................................................................................. TS-1
2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ....................... TS-8 2.1 Fuel Element Temperature Safety Limit.. .................................................................... 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 Limiting Safety System Settings ................................................................................ TS-10 2.2.1. Applicability ...................................................................................................... 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-11 3 .1 CORE REACTIVITY ................................................................................................ TS-11 3 .1.1. Applicability ...................................................................................................... TS-11 3 .1.3. Objective ............................................................................................................ TS-11 3 .1. 4. Specification ....................................................................................................... TS-11 3.1.5. Actions ............................................................................................................... TS-12 3.1.6. Basis .................................................................................................................. TS-13
  • 3.2 PULSED MODE OPERATIONS .............................................................................. TS-13 3 .2.1. Applicability ...................................................................................................... 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 MEASURING CHANNELS ...................................................................................... 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. Applicability ...................................................................................................... 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 GASEOUS EFFLUENT CONTROL ....................................................... :................. TS-20
3. 5 .1. Applicability ...................................................................................................... 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 LIMITATIONS ON EXPERIMENTS .......................................................................... TS-22 3 .6.1. Applicability ...................................................................................................... TS-22 3.6.3. Objective ............................................................................................................ TS-22 K-State Reactor TS-1 Original (4/17)

TECHNICAL SPECIFICATIONS 3.6.4. Specification ....................................................................................................... TS-22

  • 3.6.5. Actions ............................................................................................................... TS-22 3.6.6. Basis .................................................................................................................. TS-23 3.7 FUEL INTEGRITY ................................................................................................... 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 WATER ......................................................................................... 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 Maintenance Retest Requirements ................................................................................ TS-27 3 .9 .1. Applicability ...................................................................................................... TS-2 7 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 REACTIVITY ................................................................................................ 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 MEASURING CHANNELS ...................................................................................... 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 LIMITATIONS ON EXPERIMENTS ........................................................................ TS-34 4.6.1. Objective ........................................................................................................... TS-34 4.6.2. Specification ...................................................................................................... TS-34 4.6.3. Basis .................................................................................................................. TS-34 4.7 FUEL 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 WATER ....................................................................................... TS-36 4.8.1. Objective ........................................................................................................... TS-36 K-State Reactor TS-2 Original (4/17)

TECHNICAL SPECIFICATIONS 4.8.2. Specification ...................................................................................................... TS-36

  • 4.8.3. Basis .................................................................................................................. TS-36 4.9 MAINTENANCE RETEST REQUIREMENTS ....................................................... TS-37 4.9.1. Objective ........................................................................................................... TS-37 4.9.2. Specification ...................................................'................................................... TS-37 4.10.3. Basis ................................................................................................................ TS-37
5. DESIGN FEATURES ...................................................................................................... TS-38 5.1 REACTOR FUEL ...................................................................................................... TS-38 5 .1.1. Applicability ...................................................................................................... TS-3 8 5.1.2. Objective ............................................................................................................ TS-38 5 .1.3. Specification ....................................................................................................... TS-3 8 5.1.4. Basis .................................................................................................................. TS-38 5 .2 REACTOR FUEL AND FUELED DEVICES IN STORAGE .................................. TS-3 8 5 .2.1. Applicability ...................................................................................................... TS-3 8 5.2.2. Objective ............................................................................................................ TS-39 5.2.3. Specification ....................................................................................................... TS-39 5.2.4. Basis .................................................................................................................. TS-39 5.3 REACTOR BUILDING ............................................................................................. TS-39 5.3.1. Applicability ...................................................................................................... TS-39 5.3.2. Objective ............................................................................................................ TS-39 5 .3 .3. Specification ....................................................................................................... TS-3 9 5 .3 .4. Basis .................................................................................................................. TS-40 5.4 EXPERIMENTS ......................................................................................................... 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. ADMINISTRATIVE CONTROLS ................................................................................. TS-42 6.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL. ......................... TS-44 6.2 REVIEW AND AUDIT ............................................................................................. TS-45 6.3 PROCEDURES ............................................................................................................ TS-45 6.4 REVIEW OF PROPOSALS FOR EXPERIMENTS .................................................. TS-47 6.5 EMERGENCY PLAN AND PROCEDURES ........................................................... TS-48 6.6 OPERATOR REQUALIFICATION .......................................................................... TS-48 6.7 PHYSICAL 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 REPORTABLE OCCURRENCE .................................................................... TS-48 6.10 PLANT OPERA TING RECORDS ............................................................................ TS-49 6.11 REPORTING REQUIREMENTS ........................................................... TS-50
  • K-State Reactor TS-3 Original (4/17)

L_

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 .
  • ANNUAL
  • 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 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 K-State Reactor TS-4 Original (4/17)

TECHNICAL SPECIFICATIONS an operation extending more than 1 day

  • ENSURE EXHAUST PLENUM Verify existence of specified condition or (if condition does not meet criteria) take action necessary to meet condition The air volume in the reactor bay atmosphere between the pool surface and 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 Frings, as many as three contiguous fuel-element positions). IMMEDIATE Without delay, and not exceeding one hour. NOTE: IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply deferring or postponing action 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. NON SECURED 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. K-State Reactor TS-5 Original (4/17)

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, reactor power may be increased on a period of much less than l second by motion of the transient control rod. 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 111 the REACTOR CHANNEL SAFETY SYSTEM

  • SECURED EXPERIMENT K-State Reactor A secured EXPERIMENT is an EXPERIMENT held firmly in place by a 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 (4/17)

TECHNICAL SPECIFICATIONS SECURED

  • EXPERIMENT WITH MOVABLE PARTS SHALL (SHALL NOT)

A secured EXPERIMENT with movable parts is one that contains parts that are intended to be moved while the reactor is OPERATING. Indicates specified action is required/( not to be performed) 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 VIOLATION exceeded.

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 applicable titles of sections in individual Technical Specifications

  • K-State Reactor TS-7 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 2.

2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS Fuel Element Temperature Safety Limit

2. I. I Applicability This specification applies when the reactor in STEADY STATE MODE and the PULSE MODE.
2. I .2 Objective This SAFETY LIMIT ensures fuel element cladding integrity
2. I .3 Specification (I) Stainless steel clad, high-hydride fuel element temperature SHALL NOT exceed I I50°C.

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

2. I .4 Actions
  • CONDITION A Stainless steel clad, high-hydride fuel element temperature exceeds II50°C.
                                      /

REQUIRED ACTION A. I Establish SHUTDOWN condition COMPLETION TIME A.I IMMEDIATE OR AND Fuel temperature exceeds 750°C in steady state A.2 Report per Section 6.8 conditions A.2 Within 24 hours 2.1.5 Bases Safety Analysis Report, Section 3.5.I (Fuel System) identifies design and operating constraints for TRI GA fuel that will ensure cladding integrity is not challenged. NUREG 1282 identifies the safety limit for the high-hydride (ZrHu) 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 l,I50°C. A change in yield strength occurs for stainless steel cladding temperatures of 500°C, but there is no scenario for fuel cladding to achieve 500°C while submerged; consequently the safety limit during reactor operations is I,150°C. K-State Reactor TS-8 Original (4/17)

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-117-833. 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 .

  • K-State Reactor TS-9 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 2.2 2.2.1 Limiting Safety System Settings (LSSS)

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 j (1) j Power level SHALL NOT exceed 1,250 kW (th) in STEADY STATE MODE of operation j 2.2.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. I IMMEDIATE A. l Reduce power to less than 1,250 kW (th) A. Steady state power level OR exceeds 1,250 kW (th) A.2. Establish REACTOR A.2. IMMEDIATE 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 (4/17)

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:
1. The highest worth control rod is fully withdrawn (2)
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.I IMMEDIATE SHUTDOWN A. Reactivity with all control rods fully withdrawn AND exceeds $4.00 A.2 Configure reactor to A.2 Prior to continued meetLCO operations

  • K-State Reactor TS-11 Original (4/17)

TECHNICAL SPECIFICATIONS B.l.a ENSURE control rods B.1 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.1.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 meetLCO 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 l,150°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 .

  • K-State Reactor TS-12 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.2 3 .2.1 PULSED MODE Operations 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.1 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 the pulse rod in the transient rod drive position is greater than $3.00 in the PULSE MODE or equal to $3.00 OR A.2 Place reactor in STEADY STATE MODE OR A.2 IMMEDIATE 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 (4/17)

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 1 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 leveJPl 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[2l 1 1 EXHAUST PLENUM radiation monitor[21 1 1 NOTE[l]: 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.1.1 Restore channel to operation A.1.1 IMMEDIATE A. l Reactor power channels not OPERATING (min 2 OR for STEADY STATE, 1 A.1.2 ENSURE reactor is A.1.2 IMMEDIATE PULSE MODE) SHUTDOWN

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

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Establish REACTOR SHUTDOWN condition A.2 High voltage to reactor power level detector less A.2. IMMEDIATE AND 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 OR C. 22 foot Area radiation monitor is not C.3 ENSURE personnel are not C.3 IMMEDIATE OPERATING on the 22 foot level

  • OR C.4 ENSURE personnel on 22 foot level are using portable survey meters to monitor dose C.4 IMMEDIATE rates D.l 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 K-State Reactor TS-15 Original (4/17)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME

  • E.l Restore MEASURING CHANNEL OR E.1 IMMEDIATE 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.l IMMEDIATE CHANNEL OR F.2 ENSURE reactor is shutdown F.2. IMMEDIATE
  • F. Exhaust plenum radiation monitor is not OPERATING OR F.3.a ENSURE continuous air radiation monitor is OPERATING F.3.a. IMMEDIATE AND F.3.b Restore MEASURING F.3.b Within 30 days CHANNEL G.l Do not perform a reactor G.l 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 G.2 IMMEDIATE check on the startup channel 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 (4/17)

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 11 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. K-State Reactor TS-17 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.4 3.4.1 Safety Channel and Control Rod Operability 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 of90% of full reactivity insertion from the (2) fully withdrawn position in less than 1 sec. A minimum of three CONTROL RODS must be OPERABLE. Inoperable CONTROL (3) RODS must be fully inserted . TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum Function Required OPERATING Mode Safety System Channel Number STEADY PULSE or Interlock Operable STATE MODE MODE Reactor power level 2 Scram YES NA Manual scram bar 1 Scram YES YES CONTROL ROD Prevent withdrawal of standard (STAND ARD) position 1 rods in the PULSE MODE NA 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.1 Restore channel or interlock Al. IMMEDIATE to operation A. Any required SAFETY SYSTEM CHANNEL or OR interlock function is not A2. IMMEDIATE OPERABLE A.2 ENSURE reactor is SHUTDOWN K-State Reactor TS-18 Original (4/17)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME

  • B. A control rod is not OPERABLE.

B.1 ENSURE inoperable control rod is fully inserted OR Bl. IMMEDIATE B2. IMMEDIATE B.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 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. Inoperable control rods that are fully inserted in the reactor will not negatively affect the minimum safety shutdown margin or maximum excess reactivity of the core. Operating with a fully-inserted control rod may cause power peaking to shift, however, in this case calculations have demonstrated that the maximum element-to-average power peaking of 2.0 assumed in SAR Chapter 13 is still bounding, and the reduction in maximum core power by having an inoperable control rod fully inserted means that the highest temperature in any fuel element with a fully-inserted inoperable control rod will be lower than the highest temperature in the B-ring with all rods withdrawn. Therefore the reactor can be safely operated with an inoperable control rod provided that the rod is fully inserted into the core .

  • K-State Reactor TS-19 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.5 3 .5 .1 Gaseous Effluent Control 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 A.1 ENSURE reactor is SHUTDOWN OR COMPLETION TIME A.1 IMMEDIATE 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 Within 30 days A.2.d Restore the reactor bay ventilation exhaust system to OPERABLE K-State Reactor TS-20 Original (4/17)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME

  • Calculated releases of Ar-41 from the reactor bay exhaust plenum exceed 30 Ci per year.

3.5.5 Bases Do not operate. IMMEDIATE 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- 1 (2.0lxl0-5 µCi ml- 1), well below the 10CFR20 annual limit of 2000 DAC hours of Argon 41 at 6 x 10-3 µCi 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. 90 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 .

  • K-State Reactor TS-21 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.6 3 .6.1 Limitations on Experiments 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 (3) reactor pool and core area

  • 3.6.4 Actions CONDITION REQUIRED ACTION A.1 ENSURE the reactor is SHUTDOWN COMPLETION TIME A.1 IMMEDIATE A. INDEPENDENT EXPERIMENT worth is AND greater than $2.00 A.2 Remove the experiment A.2 Prior to continued operations C.1 ENSURE the reactor is C.1 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
  • K-State Reactor TS-22 Original (4/17)

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 .

  • K-State Reactor TS-23 Original (4/17)

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 TRI GA 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 in. over manufactured length, or bent laterally greater than 1/8 in.

Do not insert the fuel element into the upper core grid plate. IMMEDIATE 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 TRI GA reactor .

  • K-State Reactor TS-24 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.8 3 .8.1 Reactor Pool Water 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) Bulk water temperature SHALL NOT exceed 44 °C (111°F) (2) Water conductivity SHALL be less than 5 µmho/cm (3) Water level above the core SHALL be at least 13 ft from the top of the core

  • (4) 3.8.4 Bulk water temperature SHALL NOT exceed 37°C (99°F) with an experiment installed in an interstitial flux wire port.

Actions CONDITION REQUIRED ACTION COMPLETION TIME A.l ENSURE the reactor is A.1 IMMEDIATE SHUTDOWN AND A. Bulk water temperature exceeds 44 °C A.2 Reduce bulk water temperature to less than 44 °C A.3 IMMEDIATE

  • K-State Reactor TS-25 Original (4/17)

TECHNICAL SPECIFICATIONS CONDITION REQUIRED ACTION COMPLETION TIME

  • B. Bulk water temperature exceeds 3 7°C with an B.1 ENSURE the reactor is SHUTDOWN AND B .2 Reduce bulk water temperature B.l IMMEDIATE B.2 IMMEDIATE experiment installed in an to less than 3 7°C.

interstitial flux wire port. OR B.3 Remove experiment from flux B.3 IMMEDIATE wire port B.1 ENSURE the reactor is B.1 IMMEDIATE SHUTDOWN B. Water conductivity is AND greater than 5 µmho/cm B.2 Restore conductivity to less B.2 Within 4 weeks than 5 µmho/cm C. l 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

  • 3.8.5 Bases C.2 Restore water level C.2 IMMEDIATE 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). While the integrity of the ion exchange resin requires water temperature to remain below 54.4 °C, it is necessary to maintain water temperature below 44 °C to ensure that the departure from nucleate boiling ratio (DNBR) will remain at least 2.0 for the hot channel while operating at 1250 kWth in STEADY STATE MODE and that excessive amounts of nucleate boiling will not occur. Insertion of an experiment into an interstitial flux wire port between fuel elements necessitates a further reduction in water temperature to a maximum of 37°C in order to preclude excessive nucleate boiling of the water.

Maintaining low water conductivity over a prolonged period prevents possible corros10n, 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.

K-State Reactor TS-26 Original (4/17)

TECHNICAL SPECIFICATIONS 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 .

  • K-State Reactor TS-27 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.9 3 .9 .1 Maintenance Retest Requirements 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 .

  • K-State Reactor TS-28 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 3.10 Maximum Steady State Power 3 .9 .1 Applicability This specification applies to operations in STEADY STATE MODE.

3.9.2 Objective The objective is to ensure that the reactor has adequate margin to critical heat flux (CHF) and operates below the Limiting Safety System Setting of 1,250 kWth. 3.9.3 Specifications Maximum OPERATING thermal power SHALL NOT exceed 1,000 kWth in STEADY (1) STATE MODE. (2) A required reactor power level scram is set to a value no greater than 1,250 kWth. 3.9.4 Actions CONDITION REQUIRED ACTION COMPLETION TIME A. Thermal power exceeds 1,050 kWth in Reduce power to a level no greater IMMEDIATE STEADY STATE than 1,050 kWth. MODE B. A required reactor B.1 SHUT DOWN the reactor. B.1. IMMEDIATE power level scram is set to a value above AND AND 1,250 kWth or above the maximum readable B.2 Adjust reactor power level B.2. Prior to resuming value on a required scram setpoint to a readable value operations. channel. less than or equal to 1,250 kWth. 3.9.5 Bases The reactor control panel instrumentation is designed to measure up to 1,000 kWth of thermal power. The Limiting Safety System Setting ensures that automatic protective functions, i.e., high power scrams, are set to no greater than 1,250 kWth. However, by specifying the maximum OPERATING power level as 1,000 kWth in STEADY STATE MODE, the reactor will have additional margin to critical heat flux and will still be allowed to operate at up to the maximum power readable on the reactor console instruments. Action to reduce power is not required until power exceeds 1050kWth in STEADY STATE MODE to allow for slight variation in power level that is typical during normal operation .

  • K-State Reactor TS-29 Original (4/17)

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 4.1.3 Basis BIENNIAL 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 .
  • K-State Reactor TS-30 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 4.2 4.2.1 PULSE MODE 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 .

  • K-State Reactor TS-31 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 4.3 4.3. I MEASURING CHANNELS 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 I DAILY CHANNEL CALIBRATION ANNUAL 0 or 12 Foot Area Radiation Monitor CHANNEL CHECK I 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 Jong-term drift of the channels to be corrected. K-State Reactor TS-32 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 4.4 4.4.1 Safety Channel and Control Rod Operability 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 1 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-33 Original (4/17)

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 .

  • K-State Reactor TS-34 Original (4/17)

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.

  • K-State Reactor TS-35 Original (4/17)

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 Measure and record experiment worth of the EXPERIMENT experiment where absolute (where the absolute value of the estimated worth is greater than value of the estimated 

 $0.40).

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) .
  • K-State Reactor TS-36 Original (4/17)

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 inse1tion of3.00$

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 .

  • K-State Reactor TS-37 Original (4/17)

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 .

  • K-State Reactor TS-38 Original (4/17)

TECHNICAL SPECIFICATIONS

  • 4.9 4.9.1 Maintenance Retest Requirements 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.

  • K-State Reactor TS-39 Original (4/17)

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 (1) 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 located in the following positions: E2, E4, ES, E6, E20, E21, E22, E24, Fl, F2, F30. 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. 5.2 Reactor Fuel and Fueled Devices in Storage

  • 5.2.l Applicability K-State Reactor TS-40 Original (4/17)

TECHNICAL SPECIFICATIONS 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 .

  • K-State Reactor TS-41 Original (4/17)

TECHNICAL SPECIFICATIONS 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. (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 3 0 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 .
  • K-State Reactor TS-42 Original (4/17)

TECHNICAL SPECIFICATIONS 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 13 1 through 13 5 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. (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.

K-State Reactor TS-43 Original (4/17)

TECHNICAL SPECIFICATIONS

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

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 .

  • K-State Reactor TS-44 Original (4/17)

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 TRI GA 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. President Kansas State University Vice President for Provost Administration Chief Academic and Finance Officer Director Dean Division of Public Safety College of Engineering I I University Department of Head, Department of Police Environmental Mechanical & Nuclear Department Health and Safety Engineering I University Manager, KSU Radiation Safety Nuclear Reactor Officer Facility I Reactor I Supervisor Reactor Safeguards Committee I Reactor Operators 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-45 Original (4/17)

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 TRlGA 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 OPERA TING 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-46 Original (4/17)

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; K-State Reactor TS-47 Original (4/17)

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-48 Original (4/17)

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 1s 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 OPERA TING 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

1. 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-49 Original (4/17)

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 OPERA TING 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 CPR 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 CPR 50.59 to determine if the test K-State Reactor TS-50 Original (4/17)

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:

K-State Reactor TS-51 Original (4/17)

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 through failure to comply with an "Action" statement when "Specification" is not met; failure to comply with the "Specification" is not by itself a violation.

Surveillance Requirements must be met for all equipment/components/conditions to be considered operable. Failure to perform a surveillance within the required time 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 m 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:

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-52 Original (4/17)

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;
  • 6.11
6. updated, corrected, and as-built drawings of the facility .

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 hours 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-53 Original (4/17)

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 OPERA TING 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 OPERA TING 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; K-State Reactor TS-54 Original (4/17)

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 ofradiation and contamination surveys performed within the facility .
  • K-State Reactor TS-55 Original (4/17)}}
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