ML17132A241: Difference between revisions

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In this letter the NRC proposed discussion topics for a February 13, 2017 public meeting. On February 13, 2017, MURR met with the NRC to further discuss the topics proposed by the NRC in the February 2, 2017 letter. In consideration of the February 13, 2017 meeting discussions, MURR provides herein its licensing approach for this amendment request [i.e., Part 1 ("in-pool")]
In this letter the NRC proposed discussion topics for a February 13, 2017 public meeting. On February 13, 2017, MURR met with the NRC to further discuss the topics proposed by the NRC in the February 2, 2017 letter. In consideration of the February 13, 2017 meeting discussions, MURR provides herein its licensing approach for this amendment request [i.e., Part 1 ("in-pool")]
and discusses the subsequent to-be submitted Part 2 (i.e., "ex-pool")
and discusses the subsequent to-be submitted Part 2 (i.e., "ex-pool")
license amendment request, which collectively address the SGE at MURR. 3. Basis for the License Amendment-Part 1: "In-Pool" Activities 3.1 Construction Permit Not Required for "In-pool" Activities Implementation of Part 1 of the SGE experiment, or the "in-pool" activities, will be performed under an amendment to Renewed Facility Operating License No. R-103. MURR concludes that a construction permit is not required because implementation of the experiment will not necessitate a "material alteration" to the MURR facility per 10 C.F.R. § 50.92(a).
license amendment request, which collectively address the SGE at MURR. 3. Basis for the License Amendment-Part 1: "In-Pool" Activities  
 
===3.1 Construction===
 
Permit Not Required for "In-pool" Activities Implementation of Part 1 of the SGE experiment, or the "in-pool" activities, will be performed under an amendment to Renewed Facility Operating License No. R-103. MURR concludes that a construction permit is not required because implementation of the experiment will not necessitate a "material alteration" to the MURR facility per 10 C.F.R. § 50.92(a).
The main components necessary for the "in-pool" activities will consist of LEU target assemblies, a separate and dedicated cooling system for heat removal (forced and natural convection), and instrumentation and control equipment to support these systems. NRC regulation 10 C.F.R. § 50.92(a), Issuance of Amendment states, in part: In determining whether an amendment to a license, construction permit, or early site permit will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses, construction permits, or early site permits to the extent applicable and appropriate.
The main components necessary for the "in-pool" activities will consist of LEU target assemblies, a separate and dedicated cooling system for heat removal (forced and natural convection), and instrumentation and control equipment to support these systems. NRC regulation 10 C.F.R. § 50.92(a), Issuance of Amendment states, in part: In determining whether an amendment to a license, construction permit, or early site permit will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses, construction permits, or early site permits to the extent applicable and appropriate.
If the application involves the material alteration of a licensed facility, a construction permit will be issued before the issuance of the amendment to the license .... 3of11 (Emphasis added.) MURR is not aware of any NRC definition of "material alteration." The NRC Staff considered defining "material alteration" in 1995 -1996, but apparently abandoned the effort because of "little regulatory need to clarify the term." 2 As the NRC's Executive Director for Operations explained in SECY-96-024, In only one instance has a construction permit been issued before an amendment of an operating license, that is, an amendment to the operating license of a research reactor at the University of Maryland.
If the application involves the material alteration of a licensed facility, a construction permit will be issued before the issuance of the amendment to the license .... 3of11 (Emphasis added.) MURR is not aware of any NRC definition of "material alteration." The NRC Staff considered defining "material alteration" in 1995 -1996, but apparently abandoned the effort because of "little regulatory need to clarify the term." 2 As the NRC's Executive Director for Operations explained in SECY-96-024, In only one instance has a construction permit been issued before an amendment of an operating license, that is, an amendment to the operating license of a research reactor at the University of Maryland.
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The nominal operation cycle will load and remove 11 target rods per week for processing. At the end of the irradiation cycle , the reactor is shut down and the TA is allowed to cool by forced flow [Sa , d , e] from the TCS . At this time the l evel of stored and decay heat in the cartridge is [Sa , d , e] sufficiently l ow that direct conduction to the reactor pool water is sufficient to maintain the target rods in a safe condition.
The nominal operation cycle will load and remove 11 target rods per week for processing. At the end of the irradiation cycle , the reactor is shut down and the TA is allowed to cool by forced flow [Sa , d , e] from the TCS . At this time the l evel of stored and decay heat in the cartridge is [Sa , d , e] sufficiently l ow that direct conduction to the reactor pool water is sufficient to maintain the target rods in a safe condition.
The cartridge(s) is t h en moved to t h e loading/unloading station and the target rods are transferred to in-poo l storage or a sh i elded transfer cask which has been moved to an underwater position near the loading/unloading station for safe transfer.
The cartridge(s) is t h en moved to t h e loading/unloading station and the target rods are transferred to in-poo l storage or a sh i elded transfer cask which has been moved to an underwater position near the loading/unloading station for safe transfer.
Meanwhile a previously prepared fresh cartridge(s) is loaded into the TA positions No. SA or/and No. SB. Page 12of190 ATTACHMENT 1 1.3 Design for Loss of Forced Cooling L oss of forced cooling (LOFC) to the TA could occur by a power failure , pump failure or coolant pipe b reak. If a LOFC event occurs , a low flow signal (8S% of normal flow) or a high flow signal (11 S% of n ormal flow) is sent from the SGE TEF to the MURR reactor safety system to scram th e reactor. Decay h eat removal from the target rods is provided by pool water flowing through automatically-actuated decay h eat removal valves into the target cartridge.
Meanwhile a previously prepared fresh cartridge(s) is loaded into the TA positions No. SA or/and No. SB. Page 12of190 ATTACHMENT 1 1.3 Design for Loss of Forced Cooling L oss of forced cooling (LOFC) to the TA could occur by a power failure , pump failure or coolant pipe b reak. If a LOFC event occurs , a low flow signal (8S% of normal flow) or a high flow signal (11 S% of n ormal flow) is sent from the SGE TEF to the MURR reactor safety system to scram th e reactor. Decay h eat removal from the target rods is provided by pool water flowing through automatically-actuated decay h eat removal valves into the target cartridge.  
1.4 Target Experimental Facility Safety Design Approach The safety objectives for the SGE TEF are (1) minimization of exposure of occupational workers to ra diation from normal system operation, (2) prevention of release of radionuclides to the MURR reactor c ontainment building and the environment from operation of the system , and (3) control of target rod c ladding temperatures to prevent cladding failure leading to a fission product release to the reactor c ontainment building and potentially , to the environment.
 
===1.4 Target===
Experimental Facility Safety Design Approach The safety objectives for the SGE TEF are (1) minimization of exposure of occupational workers to ra diation from normal system operation, (2) prevention of release of radionuclides to the MURR reactor c ontainment building and the environment from operation of the system , and (3) control of target rod c ladding temperatures to prevent cladding failure leading to a fission product release to the reactor c ontainment building and potentially , to the environment.
To meet these objectives, the SGE TEF has been desi gned to the following top-level requirements to e nsure that the safety objectives are met:
To meet these objectives, the SGE TEF has been desi gned to the following top-level requirements to e nsure that the safety objectives are met:
* The target rods and cartridges shall be capable of sustained operation with the reactor operating at a steady-state power level of 10.0 MW, and through anticipated operating transients where the reactor power reaches its scram power limit , without jeopardizing the integrity of barriers that would release radionuclides to the reactor pool water, target cooling water , or the reactor containment building.
* The target rods and cartridges shall be capable of sustained operation with the reactor operating at a steady-state power level of 10.0 MW, and through anticipated operating transients where the reactor power reaches its scram power limit , without jeopardizing the integrity of barriers that would release radionuclides to the reactor pool water, target cooling water , or the reactor containment building.
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The target rods are located and held concentric to the water channels by 11 cup features that are fabricated into the lower cartridge flange and can be seen in Figure 8. This ensures even flow velocities around the target rods through the cartridge.
The target rods are located and held concentric to the water channels by 11 cup features that are fabricated into the lower cartridge flange and can be seen in Figure 8. This ensures even flow velocities around the target rods through the cartridge.
The water cutouts on the cup features are smaller than the pointed end cap of the target rod. This eliminates the possibility of the target rod getting stuck on one of these water bypass features when inserting them into the cartridge and ensures the operator can always find the center of the cartridge and guide the target rod into its final position. Page 21 of 190 [Sa , b d , e , t]
The water cutouts on the cup features are smaller than the pointed end cap of the target rod. This eliminates the possibility of the target rod getting stuck on one of these water bypass features when inserting them into the cartridge and ensures the operator can always find the center of the cartridge and guide the target rod into its final position. Page 21 of 190 [Sa , b d , e , t]
ATTACHMENT 1 [Sa , b d , e , f] Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum The water cutouts are designed to create near-uniform coolant flow over the target rod s as soon as the water enters the cartridge.
ATTACHMENT 1 [Sa , b d , e , f] Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum The water cutouts are designed to create near-uniform coolant flow over the target rod s as soon as the water enters the cartridge.  
2.1.5 Target Rods Figure 9 shows an individual target rod assembly , which consists of an upper end cap , cladding , a spring , 100 U0 2 target pellets and a lower end cap. The U0 2 target pellets have a nominal active length of 23.6 inches (600 mm) in the cold state. A nominal radial gap of 50 µm exists between the pellet outside diameter (OD) and the cladding inside diameter (ID). This nominal gap offers the best dimensional balance between cladding strain and thermal conductance to the cladding wall. The end caps are fabricated from Zircaloy-4 bar with integrated features designed to optimize installation and extraction from the cartridge.
 
====2.1.5 Target====
Rods Figure 9 shows an individual target rod assembly , which consists of an upper end cap , cladding , a spring , 100 U0 2 target pellets and a lower end cap. The U0 2 target pellets have a nominal active length of 23.6 inches (600 mm) in the cold state. A nominal radial gap of 50 µm exists between the pellet outside diameter (OD) and the cladding inside diameter (ID). This nominal gap offers the best dimensional balance between cladding strain and thermal conductance to the cladding wall. The end caps are fabricated from Zircaloy-4 bar with integrated features designed to optimize installation and extraction from the cartridge.
Both end cap s are welded to the cladding autogenously (no weld rod) by a standard automated orbital weld head. The stainless steel spring is held captive by the upper end cap, to aid in ea s ier recovery post-irradiation.
Both end cap s are welded to the cladding autogenously (no weld rod) by a standard automated orbital weld head. The stainless steel spring is held captive by the upper end cap, to aid in ea s ier recovery post-irradiation.
The individual pellet/clad components and dimensions are listed in Table 3. The Zircaloy-4 (UNS R60804) cladding will be fabricated and inspected in accordance with seamless alloy tubes for nuclear reactor fuel cladding applications per American Society for Testing and Materials (ASTM) B811. The Zircaloy-4 (UNS R60804) end caps and stainless steel spring will be fabricated from bar material in accordance to ASTM B351. Page 22 of 190 ATTACHMENT 1 The filler rods have external dimensions that are identical to the target rods , except that they are fabricate d from solid stai nl ess steel (SS3 1 6L). The filler rods ensure that the required flow co nditi ons for safe cooling of the target rods are met. T h e SS3 1 6L acts as a neutron absorber to r educe power p eaking in the target rod next to the filler rod. 2 6.5 0 5 i n C l adding (Zircaloy-4) Spring (3 1 6L SST) U02 pellet (x 100) . I i-*-*-*-*-*-*-*-*-*-*-*-*-*-*-* I Cladding (Zircalo y-4) L o w er End Cap (Zircaloy-4)
The individual pellet/clad components and dimensions are listed in Table 3. The Zircaloy-4 (UNS R60804) cladding will be fabricated and inspected in accordance with seamless alloy tubes for nuclear reactor fuel cladding applications per American Society for Testing and Materials (ASTM) B811. The Zircaloy-4 (UNS R60804) end caps and stainless steel spring will be fabricated from bar material in accordance to ASTM B351. Page 22 of 190 ATTACHMENT 1 The filler rods have external dimensions that are identical to the target rods , except that they are fabricate d from solid stai nl ess steel (SS3 1 6L). The filler rods ensure that the required flow co nditi ons for safe cooling of the target rods are met. T h e SS3 1 6L acts as a neutron absorber to r educe power p eaking in the target rod next to the filler rod. 2 6.5 0 5 i n C l adding (Zircaloy-4) Spring (3 1 6L SST) U02 pellet (x 100) . I i-*-*-*-*-*-*-*-*-*-*-*-*-*-*-* I Cladding (Zircalo y-4) L o w er End Cap (Zircaloy-4)
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* Natural Convection:
* Natural Convection:
In the event that pumped flow to the T As is interrupted , the reactor is s hut down and the TA power reverts to decay heat. The TCS transitions to natural circulation flo w to the T As in which valves on the target cooling inlet pipes are opened to allow pool water to flow through the T As as driven by natural circulation from target deca y heat. The TCS i s divided into two subsys tems , the target flow assembly and the water cooling module. The flow interface begins at the reactor pool and ends at the TA inlet. The heat rejection interface is with the MURR secondary coolant system for forced cooling operation and with the reactor pool for natural convection operation. The TCS instrumentation also includes inputs to the MURR reactor safety syste m to scram the reactor if forced flow falls below 85%, or above 115% of normal flow , and on high coolant temperature to the TA. 3.1 Functions The primary function of the TCS is to provide cooling water to the TAs at a sufficient mass flow rate to maintain the target material within its required temperature range for safe operation.
In the event that pumped flow to the T As is interrupted , the reactor is s hut down and the TA power reverts to decay heat. The TCS transitions to natural circulation flo w to the T As in which valves on the target cooling inlet pipes are opened to allow pool water to flow through the T As as driven by natural circulation from target deca y heat. The TCS i s divided into two subsys tems , the target flow assembly and the water cooling module. The flow interface begins at the reactor pool and ends at the TA inlet. The heat rejection interface is with the MURR secondary coolant system for forced cooling operation and with the reactor pool for natural convection operation. The TCS instrumentation also includes inputs to the MURR reactor safety syste m to scram the reactor if forced flow falls below 85%, or above 115% of normal flow , and on high coolant temperature to the TA. 3.1 Functions The primary function of the TCS is to provide cooling water to the TAs at a sufficient mass flow rate to maintain the target material within its required temperature range for safe operation.
The is designed to perform the following functions:
The is designed to perform the following functions:  
3.1.1 Process Functions
 
====3.1.1 Process====
Functions
* Transfer and reject the target heat to the MURR secondary coolant system during normal operation.
* Transfer and reject the target heat to the MURR secondary coolant system during normal operation.
* Maintain sufficient cooling water flow rate and temperature conditions through the target to maintain a large CHFR margin and such that any s ubcooled nucleate boiling has a negligible impact on target rod integrity.
* Maintain sufficient cooling water flow rate and temperature conditions through the target to maintain a large CHFR margin and such that any s ubcooled nucleate boiling has a negligible impact on target rod integrity.
* Minimize net exchange of heat between the TCS and the re actor pool water during normal operation. Page 36 of 190 ATTACHMENT 1
* Minimize net exchange of heat between the TCS and the re actor pool water during normal operation. Page 36 of 190 ATTACHMENT 1
* Provide for reliable transition of target heat removal to the reactor pool via natural circulation in the event of an LOFC.
* Provide for reliable transition of target heat removal to the reactor pool via natural circulation in the event of an LOFC.
* Provide instrumentation to assure that flow , temperatures and pressures are within specified conditions for operation.
* Provide instrumentation to assure that flow , temperatures and pressures are within specified conditions for operation.  
3.1.2 Safety Functions
 
====3.1.2 Safety====
Functions
* Provide a signal to the MURR reactor safety system to enable a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation or for high heat exchanger outlet temperature.
* Provide a signal to the MURR reactor safety system to enable a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation or for high heat exchanger outlet temperature.
* Provide a corresponding signal to open the decay heat removal valves to initiate natural circulation flow to the T As that coincides with a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation. Accident analysis shows that even if no natural circulation occurs after an LOFC event , target decay heat w ill be transported directly to the pool. The targets will remain in a safe condition with no radioactivity released to the reactor pool. 3.2 Target Cooling System Principal Design Parameters The principal design condition for the TCS corresponds to both T As loaded with 11 target rods each and the reactor operating at 10 MW 1* The target power levels will vary slightly due to shifts in the driving neutron flux distribution caused by changes in burnup of reactor fuel elements , position of control rods and age of the beryllium reflector.
* Provide a corresponding signal to open the decay heat removal valves to initiate natural circulation flow to the T As that coincides with a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation. Accident analysis shows that even if no natural circulation occurs after an LOFC event , target decay heat w ill be transported directly to the pool. The targets will remain in a safe condition with no radioactivity released to the reactor pool. 3.2 Target Cooling System Principal Design Parameters The principal design condition for the TCS corresponds to both T As loaded with 11 target rods each and the reactor operating at 10 MW 1* The target power levels will vary slightly due to shifts in the driving neutron flux distribution caused by changes in burnup of reactor fuel elements , position of control rods and age of the beryllium reflector.
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The table shows that the minimum safety factor is 7.8 at the worst location.
The table shows that the minimum safety factor is 7.8 at the worst location.
Page 48 of 190 ATTACHMENT 1 Table 8 Results of Rigid Structure Seismic Analysis Location Material Stress Stress Allowable Safety Factor (ksi) (ksi) Y-pipe , before flexible piping SS316L 0.7 16.7 24 Inside pool , on expansion loop 6061-T6 1.0 16.7 17 At heat exchanger interface SS316L 3.2 25 7.8 At SS to Al pipe interface 6061-T6 0.2 25.I 126 3-inch to 2-inch Y-pipe SS316L 0.3 16.7 56 All Al piping 6061-T6 0.3 16.7 56 Y-piping support SS316L 0.9 22.2 25 Inside pool , on expansion loop 6061-T6 1.0 22.2 22 The second type assumes that the biological shield/pool and Costar Tower combined with the bridge will l aterally displace during a seismic event. To demonstrate the effectiveness of the flexible couplings , a hypothetical forced displacement of 2 inches is applied. By comparison , an initial seismic evaluation of the Costar Tower showed a less than I /8 111 inch lateral movement at the location where the bridge connects.
Page 48 of 190 ATTACHMENT 1 Table 8 Results of Rigid Structure Seismic Analysis Location Material Stress Stress Allowable Safety Factor (ksi) (ksi) Y-pipe , before flexible piping SS316L 0.7 16.7 24 Inside pool , on expansion loop 6061-T6 1.0 16.7 17 At heat exchanger interface SS316L 3.2 25 7.8 At SS to Al pipe interface 6061-T6 0.2 25.I 126 3-inch to 2-inch Y-pipe SS316L 0.3 16.7 56 All Al piping 6061-T6 0.3 16.7 56 Y-piping support SS316L 0.9 22.2 25 Inside pool , on expansion loop 6061-T6 1.0 22.2 22 The second type assumes that the biological shield/pool and Costar Tower combined with the bridge will l aterally displace during a seismic event. To demonstrate the effectiveness of the flexible couplings , a hypothetical forced displacement of 2 inches is applied. By comparison , an initial seismic evaluation of the Costar Tower showed a less than I /8 111 inch lateral movement at the location where the bridge connects.
The AutoPipe model showed that the displacement has negligible effects on the stress ratio (max ratio is 0.05) demonstrating that the flexible piping design works as intended.
The AutoPipe model showed that the displacement has negligible effects on the stress ratio (max ratio is 0.05) demonstrating that the flexible piping design works as intended.  
3.4.4 Secondary Cooling for the Target Cooling System The heat from the TCS is transferred to the MURR secondary coolant system by means of one of the TCS heat exchangers. The heat is then dissipated to the atmosphere through a cooling tower with seasonal additional cooling provided by a chill water system. The secondary coolant loop that serves the TCS consists of one (1) coolant circulation pump , two (2) heat exchangers , two (2) automatic temperature control valves , one (I) chill water heat exchanger , temperature and flow instrumentation , and associated v alves and piping. A piping and instrumentation diagram is shown in Figure 28. The secondary coolant loop that serves the TCS shares a cooling tower , radiation monitoring instrumentation, and chemical control systems with the existing 15 MW capacity MURR secondary coolant system. Only the secondary coolant loop serving the TCS will be described in this section. The secondary coolant system is designed such that a failure or malfunction of any system component will not lead to target rod damage or an uncontrolled release of radioactivity to the environment.
 
3.4.5 Circulation Pump The secondary coolant circulation pump for the TCS is a centrifugal , single-stage pump that is connected to a variable speed drive (VSD) unit through a coupling.
====3.4.4 Secondary====
 
Cooling for the Target Cooling System The heat from the TCS is transferred to the MURR secondary coolant system by means of one of the TCS heat exchangers. The heat is then dissipated to the atmosphere through a cooling tower with seasonal additional cooling provided by a chill water system. The secondary coolant loop that serves the TCS consists of one (1) coolant circulation pump , two (2) heat exchangers , two (2) automatic temperature control valves , one (I) chill water heat exchanger , temperature and flow instrumentation , and associated v alves and piping. A piping and instrumentation diagram is shown in Figure 28. The secondary coolant loop that serves the TCS shares a cooling tower , radiation monitoring instrumentation, and chemical control systems with the existing 15 MW capacity MURR secondary coolant system. Only the secondary coolant loop serving the TCS will be described in this section. The secondary coolant system is designed such that a failure or malfunction of any system component will not lead to target rod damage or an uncontrolled release of radioactivity to the environment.  
 
====3.4.5 Circulation====
 
Pump The secondary coolant circulation pump for the TCS is a centrifugal , single-stage pump that is connected to a variable speed drive (VSD) unit through a coupling.
This pump , designated SP-5 , is capable of supplying 500 gpm at a discharge pressure of 50 psig. Three (3) other secondary coolant Page 49 of 190 ATTACHMENT 1 pumps are installed in a parallel configuration with each pump capable of supplying 2 , 200 gpm to the secondary coolant system. Jn case pump SP-5 failed , TCS secondary coolant flow can be established using a cross-connect pipe to the discharge header of the other three (3) secondary coolant pumps. Additiona l detail of the SP-5 control system is provided in Instrumentation and Control section of this document. z: ; ... ..... 0 rn GO z (') ::c --4 :;: . ' N .. z rn °' Figure 28 Secondary Cooling for Target Cooling System Page 50 of 190 ATTACHMENT 1 3.4.6 Heat Exchangers The secondary coolant flow makes a single pass on one of the TCS heat exchanger s. The TCS heat exchangers are plate-type heat exchangers.
This pump , designated SP-5 , is capable of supplying 500 gpm at a discharge pressure of 50 psig. Three (3) other secondary coolant Page 49 of 190 ATTACHMENT 1 pumps are installed in a parallel configuration with each pump capable of supplying 2 , 200 gpm to the secondary coolant system. Jn case pump SP-5 failed , TCS secondary coolant flow can be established using a cross-connect pipe to the discharge header of the other three (3) secondary coolant pumps. Additiona l detail of the SP-5 control system is provided in Instrumentation and Control section of this document. z: ; ... ..... 0 rn GO z (') ::c --4 :;: . ' N .. z rn °' Figure 28 Secondary Cooling for Target Cooling System Page 50 of 190 ATTACHMENT 1 3.4.6 Heat Exchangers The secondary coolant flow makes a single pass on one of the TCS heat exchanger s. The TCS heat exchangers are plate-type heat exchangers.
On one s ide of the plates is secondary coolant , and the other s ide has TCS coolant. Only one TCS heat exchanger is required for TCS operation.
On one s ide of the plates is secondary coolant , and the other s ide has TCS coolant. Only one TCS heat exchanger is required for TCS operation.
The other TCS heat exc hanger is an installed spare for operational reliability.
The other TCS heat exc hanger is an installed spare for operational reliability.
D uring the summer months , additional cooling to the secondary coolant supply to the TCS may be needed to ensure a sufficient secondar y cooling water temperature to maintain proper TCS heat exchanger outlet te mperature.
D uring the summer months , additional cooling to the secondary coolant supply to the TCS may be needed to ensure a sufficient secondar y cooling water temperature to maintain proper TCS heat exchanger outlet te mperature.
This additional cooling to the secondary coolant system is provided by a chill water system v ia a plate-type heat exchanger.
This additional cooling to the secondary coolant system is provided by a chill water system v ia a plate-type heat exchanger.  
3.4.7 Automatic Temperature Control Valves The amount of secondary coolant passing through the TCS heat exchanger is controlled by an automatic te mperature control valve located in a bypass line of its associated heat exchanger.
 
====3.4.7 Automatic====
 
Temperature Control Valves The amount of secondary coolant passing through the TCS heat exchanger is controlled by an automatic te mperature control valve located in a bypass line of its associated heat exchanger.
The valve is an electro-mechanical butterfly valve which responds to the TCS heat exchanger outlet temperature which maintains a constant cold leg temperature to the T As. Since only one TCS heat exchanger is operated at any time , only one automatic temperature control valve is in operation at any time. The other valve is an i nstalled spare for operational reliability.
The valve is an electro-mechanical butterfly valve which responds to the TCS heat exchanger outlet temperature which maintains a constant cold leg temperature to the T As. Since only one TCS heat exchanger is operated at any time , only one automatic temperature control valve is in operation at any time. The other valve is an i nstalled spare for operational reliability.
The automatic temperature control valves for the TCS heat exc hangers are designated as S-3A and S-3B. Additional detail on the operation of the automatic temperature control valves and system is provided in Instrumentation and Control section of this document.
The automatic temperature control valves for the TCS heat exc hangers are designated as S-3A and S-3B. Additional detail on the operation of the automatic temperature control valves and system is provided in Instrumentation and Control section of this document.  
3.4.8 Instrumentation The secondary coolant inlet and outlet temperatures to both the chill water and TCS heat exchangers and the total flow in the secondary s y stem TCS loop are displayed and recorded in the reactor control room. 3.4.9 Electrical Power System Supporting the TCS Electrical power to the TCS pumps and their Instrumentation and Controls Uninterruptible Power Supply (l&C UPS) will be supplied by Motor Control Center No. 4 (MCC-4) through a new 480/277 V distribution panel. Normal Electrical Power System Substation B supplies 480-volt , 3-phase , 60-cycle electrical power to MCC-4 through a 250-amp breaker as described in Chapter 8 of the MURR Safety Analysis Report (Reference 1 I). The new 480/277 V , 200-amp panel , designated HVP-4 , will be connected to MCC-4 in the reactor containment building.
 
====3.4.8 Instrumentation====
 
The secondary coolant inlet and outlet temperatures to both the chill water and TCS heat exchangers and the total flow in the secondary s y stem TCS loop are displayed and recorded in the reactor control room. 3.4.9 Electrical Power System Supporting the TCS Electrical power to the TCS pumps and their Instrumentation and Controls Uninterruptible Power Supply (l&C UPS) will be supplied by Motor Control Center No. 4 (MCC-4) through a new 480/277 V distribution panel. Normal Electrical Power System Substation B supplies 480-volt , 3-phase , 60-cycle electrical power to MCC-4 through a 250-amp breaker as described in Chapter 8 of the MURR Safety Analysis Report (Reference 1 I). The new 480/277 V , 200-amp panel , designated HVP-4 , will be connected to MCC-4 in the reactor containment building.
Electrical power to the TCS pumps will be provided by HVP-4. HVP-4 will also s upply electrical power to a new 120/208-V Distribution Center , designated 120/208V Distribution Center 3 , through a 112.5-kV A transformer. A Lighting Panel on l 20/208V Distribution Center 3 will s upply electrical power to the TCS pumps I&C UPS as shown in Figure 29. Page 51of190 ATTACHMENT 1 Location of the new electrical wiring will ensure that no electromagnetic interference will exist between the electrical power service and any safety-related I&C circuits.
Electrical power to the TCS pumps will be provided by HVP-4. HVP-4 will also s upply electrical power to a new 120/208-V Distribution Center , designated 120/208V Distribution Center 3 , through a 112.5-kV A transformer. A Lighting Panel on l 20/208V Distribution Center 3 will s upply electrical power to the TCS pumps I&C UPS as shown in Figure 29. Page 51of190 ATTACHMENT 1 Location of the new electrical wiring will ensure that no electromagnetic interference will exist between the electrical power service and any safety-related I&C circuits.
1 I COLLING 3 PUMP 5 7 9 11 13 15 17 19 21 23 25 27 29 FroM MCC 114 Breo.ker A6 I HVP-4 480/277 200A 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 PUMP I 480-277V ) 112 KVA 3PH I I TCS !LC 3 UPS Panel 5 7 9 Figure 29 120/208V DISTRIBUTION CENTER-3 400A 2 4 6 9 10 Electric Power Supply to TCS Pumps and I&C UPS 4. Instrumentation and Control System This section describes the operating characteristics of the TEF Instrumentation and Control (l&C) Systems. These systems assure that the TEF can be safely operated , monitored , and shut down as warranted.
1 I COLLING 3 PUMP 5 7 9 11 13 15 17 19 21 23 25 27 29 FroM MCC 114 Breo.ker A6 I HVP-4 480/277 200A 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 PUMP I 480-277V ) 112 KVA 3PH I I TCS !LC 3 UPS Panel 5 7 9 Figure 29 120/208V DISTRIBUTION CENTER-3 400A 2 4 6 9 10 Electric Power Supply to TCS Pumps and I&C UPS 4. Instrumentation and Control System This section describes the operating characteristics of the TEF Instrumentation and Control (l&C) Systems. These systems assure that the TEF can be safely operated , monitored , and shut down as warranted.  
4.1 Summary Description The TEF I&C Systems are comprised of the sensors , electronic circuitry , displays , and actuating devices available to provide the information and means to safely control the TEF and avoid or mitigate potential accidents that could affect the TEF or the reactor. The TEF I&C systems include the following:
 
===4.1 Summary===
Description The TEF I&C Systems are comprised of the sensors , electronic circuitry , displays , and actuating devices available to provide the information and means to safely control the TEF and avoid or mitigate potential accidents that could affect the TEF or the reactor. The TEF I&C systems include the following:
* Target Cooling System (TCS) Control System
* Target Cooling System (TCS) Control System
* TCS Protection System Page 52 of 190 ATTACHMENT 1
* TCS Protection System Page 52 of 190 ATTACHMENT 1
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The basis for the TCS high flow scram is to ensure target fission heat generation is terminated upon a TCS pipe break between the flow measuring element and a TA which would reduce flow resistance and increase flow indication at the associated flow transmitter. This condition causes low flow to the T As. The basis for the TCS high temperature scram is to ensure target fission heat generation is terminated upon a high temperature condition that is not associated with a change in TCS flow rate. Insufficient secondary coolant flow and high secondary coolant temperature could cause such a high TCS temperature to occur. The TCS Protection System is designed with all analog components.
The basis for the TCS high flow scram is to ensure target fission heat generation is terminated upon a TCS pipe break between the flow measuring element and a TA which would reduce flow resistance and increase flow indication at the associated flow transmitter. This condition causes low flow to the T As. The basis for the TCS high temperature scram is to ensure target fission heat generation is terminated upon a high temperature condition that is not associated with a change in TCS flow rate. Insufficient secondary coolant flow and high secondary coolant temperature could cause such a high TCS temperature to occur. The TCS Protection System is designed with all analog components.
The components and system are designed to be fail-safe , meaning that any loss of power or signal will cause a reactor scram. In addition , all TCS Protection System signals are separate from TCS Control System signals. Keyed bypass switches to the TCS Protection System allows for reactor operation with the TCS not operating , one TCS branch operating , or both TCS branches operating.
The components and system are designed to be fail-safe , meaning that any loss of power or signal will cause a reactor scram. In addition , all TCS Protection System signals are separate from TCS Control System signals. Keyed bypass switches to the TCS Protection System allows for reactor operation with the TCS not operating , one TCS branch operating , or both TCS branches operating.
The keyed bypass removes the protection signal inputs for any part of the TCS that is not operating.
The keyed bypass removes the protection signal inputs for any part of the TCS that is not operating.  
4.3.1 Target Cooling System Low and High Flow Rate Protection TCS flow is measured at the following locations with the indicated TCS Flow Elements (TCFEs) and TCS Flow Transmitters (TCFTs):
 
====4.3.1 Target====
Cooling System Low and High Flow Rate Protection TCS flow is measured at the following locations with the indicated TCS Flow Elements (TCFEs) and TCS Flow Transmitters (TCFTs):
* Downstream ofTCS combined cold leg header on branch A
* Downstream ofTCS combined cold leg header on branch A
* Downstream ofTCS combined cold leg header on branch B Flow elements TCFE-lA and TCFE-lB for Target A and Target B , respectively , are located in each TCS branch. The differential pressure caused by TCS flow is measured b y flow transmitters TCFT-1 A and TCFT-2A for TCS branch A and TCFT-lB and TCFT-2B for TCS branch B. The output signal (4 to 20 mA) generated by each flow transmitter is directed to a square root converter which provides a linear output signal for the two dual alann trip units and a chart recorder in series with the signal. In addition to providing flow indication and recording , the recorder will initiate a " TCS Lo Flow" alarm when TCS branch flow to either target decreases to 90% of its normal value. If TCS branch flow decreases to 85% normal flow , then the following actions are initiated by the system: 1. Reactor scram; 2. Opening of the two (2) TDHRVs in the associated branch; and Page 56 of 190 ATTACHMENT 1 3. Actuation of the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-1 A opens two (2) contacts:  
* Downstream ofTCS combined cold leg header on branch B Flow elements TCFE-lA and TCFE-lB for Target A and Target B , respectively , are located in each TCS branch. The differential pressure caused by TCS flow is measured b y flow transmitters TCFT-1 A and TCFT-2A for TCS branch A and TCFT-lB and TCFT-2B for TCS branch B. The output signal (4 to 20 mA) generated by each flow transmitter is directed to a square root converter which provides a linear output signal for the two dual alann trip units and a chart recorder in series with the signal. In addition to providing flow indication and recording , the recorder will initiate a " TCS Lo Flow" alarm when TCS branch flow to either target decreases to 90% of its normal value. If TCS branch flow decreases to 85% normal flow , then the following actions are initiated by the system: 1. Reactor scram; 2. Opening of the two (2) TDHRVs in the associated branch; and Page 56 of 190 ATTACHMENT 1 3. Actuation of the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-1 A opens two (2) contacts:  
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4: T""" * > 0::: I a I-1S30 Target A Bypass ATTACHMENT 1 4: N I > 0::: I a I-1S31 Target A Bypass CD T""" I > 0::: I a I-1S30 Target B Bypass CD N > 0::: I a I--1S31 Target B Bypass Figure 33 TDHRVs-TCS Pump Interlock Circuit TCS Pump Control Finally , if a single switch fails or is placed in the incorrect operating position , the worst condition that could be possible would be to have two (2) T As operating with one key switch in the 'Normal' position and the other key switch in 'Both AB Bypass.' In this condition , the high and low flow scrams and the high heat exchanger outlet temperature scram would be disabled for only one ( 1) of the two (2) reductant reactor safety system legs. Any other single switch mis-position or single switch failure with any operating condition would cause less protection to be bypassed.
4: T""" * > 0::: I a I-1S30 Target A Bypass ATTACHMENT 1 4: N I > 0::: I a I-1S31 Target A Bypass CD T""" I > 0::: I a I-1S30 Target B Bypass CD N > 0::: I a I--1S31 Target B Bypass Figure 33 TDHRVs-TCS Pump Interlock Circuit TCS Pump Control Finally , if a single switch fails or is placed in the incorrect operating position , the worst condition that could be possible would be to have two (2) T As operating with one key switch in the 'Normal' position and the other key switch in 'Both AB Bypass.' In this condition , the high and low flow scrams and the high heat exchanger outlet temperature scram would be disabled for only one ( 1) of the two (2) reductant reactor safety system legs. Any other single switch mis-position or single switch failure with any operating condition would cause less protection to be bypassed.
In addition , any single switch position or single switch failure with only one (1) TA operating would cause the operating TCS pump to stop due to the open TDHRVs in the secured TA cooling branch. The stopped TCS pump would reduce target flow and cause a reactor scram. Administratively , the keys to switches 1 S30 and 1 S3 I will require Reactor Manager's permission to use. Since the MURR is not normally operated with bypass keys installed at the reactor console , the presence of keys will be a noticeable exception from normal operations.
In addition , any single switch position or single switch failure with only one (1) TA operating would cause the operating TCS pump to stop due to the open TDHRVs in the secured TA cooling branch. The stopped TCS pump would reduce target flow and cause a reactor scram. Administratively , the keys to switches 1 S30 and 1 S3 I will require Reactor Manager's permission to use. Since the MURR is not normally operated with bypass keys installed at the reactor console , the presence of keys will be a noticeable exception from normal operations.
Therefore , operator attention to correct operation of the bypass keys will be peaked when they are implemented.
Therefore , operator attention to correct operation of the bypass keys will be peaked when they are implemented.  
4.3.4 Target Cooling System Parameter Indication, Recording, and Alarm System The TCS Parameter Indication , Recording , and Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS , to record TCS parameter data for long-term retention or review , and to actuate alarms to alert the operator to abnormal parameters. No Page 62 of 190 ATTACHMENT 1 safety-related functions are associated with the TCS Parameter Indication , Recordin g, and Alarm System. Every parameter is detected by a se nsor. The se nsor s i g n al i s converted to a 4-20mA current loop signa l going to paperles s chart recorder s which also serve as param e ter indication a nd alarm units. 4.3.5 Target Cooling System Secondary Coolant Control System The TCS Secondary Coolant Control System is de s igned to control the TCS secondary coolant circulation pump (SP-5) speed and the position of the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B). From the TCS Secondary Coolant Control System , the operator can adjust SP-5 to make co urse secondary coolant flow rate changes to the WCM. However , fine secondary coolant flow rate adjustment through a TCS heat exchanger is made by either S-3A or S-3B that are located in a secondary coo lant pipe that bypasses the associated heat exchanger.
 
====4.3.4 Target====
Cooling System Parameter Indication, Recording, and Alarm System The TCS Parameter Indication , Recording , and Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS , to record TCS parameter data for long-term retention or review , and to actuate alarms to alert the operator to abnormal parameters. No Page 62 of 190 ATTACHMENT 1 safety-related functions are associated with the TCS Parameter Indication , Recordin g, and Alarm System. Every parameter is detected by a se nsor. The se nsor s i g n al i s converted to a 4-20mA current loop signa l going to paperles s chart recorder s which also serve as param e ter indication a nd alarm units. 4.3.5 Target Cooling System Secondary Coolant Control System The TCS Secondary Coolant Control System is de s igned to control the TCS secondary coolant circulation pump (SP-5) speed and the position of the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B). From the TCS Secondary Coolant Control System , the operator can adjust SP-5 to make co urse secondary coolant flow rate changes to the WCM. However , fine secondary coolant flow rate adjustment through a TCS heat exchanger is made by either S-3A or S-3B that are located in a secondary coo lant pipe that bypasses the associated heat exchanger.
For simplicity , further explanation of how the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B) operate , S-3A will be assumed operating and its operation will be the example described. S-3B's controls and operation are the same as S-3A. During steady-state operation , SP-5 provides constant secondary coolant flow to the WCM. S-3A and its associated heat exchanger are in use , and S-3A is in 'Automatic' control. S-3A is located in a secondary coolant bypass line that bypas ses the TCS heat exchanger.
For simplicity , further explanation of how the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B) operate , S-3A will be assumed operating and its operation will be the example described. S-3B's controls and operation are the same as S-3A. During steady-state operation , SP-5 provides constant secondary coolant flow to the WCM. S-3A and its associated heat exchanger are in use , and S-3A is in 'Automatic' control. S-3A is located in a secondary coolant bypass line that bypas ses the TCS heat exchanger.
Therefore , as S-3A closes , less secondary coolant bypasses the heat exchanger resulting in more cooling flow to the heat exchanger.
Therefore , as S-3A closes , less secondary coolant bypasses the heat exchanger resulting in more cooling flow to the heat exchanger.
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The physics model also provides the distribution of power densities in the U0 2 pellet s for the thermal-hydraulic design of the TA. The target physics model incorporated both the T As and the MURR reactor core and examined the full range of MURR operations to develop the target physics de sign cases. 5.1 Analytical Methods The nuclear design calculations were conducted using Monte Carlo code MCNP6 (Reference  
The physics model also provides the distribution of power densities in the U0 2 pellet s for the thermal-hydraulic design of the TA. The target physics model incorporated both the T As and the MURR reactor core and examined the full range of MURR operations to develop the target physics de sign cases. 5.1 Analytical Methods The nuclear design calculations were conducted using Monte Carlo code MCNP6 (Reference  
: 12) and ENDF/B-Vll.l (References 13 and 14) to obtain eigenvalue and neutron flux distributions.
: 12) and ENDF/B-Vll.l (References 13 and 14) to obtain eigenvalue and neutron flux distributions.
The MCNP6 calculations also produce heating rates for the reactor fuel elements and target rods , which are used for the nuclear design and analyses to obtain the TA power. For the determination of the TA uranium-23S enrichment distribution in MCNP6, 100 million source histories (10,000 x I0,000 neutrons) were used to obtain converged eigenvalue and power distributions.
The MCNP6 calculations also produce heating rates for the reactor fuel elements and target rods , which are used for the nuclear design and analyses to obtain the TA power. For the determination of the TA uranium-23S enrichment distribution in MCNP6, 100 million source histories (10,000 x I0,000 neutrons) were used to obtain converged eigenvalue and power distributions.  
5.2 Target Assembly Physics Model The T As are loaded in graphite reflector positions No. SA and No. SB. The physics design and analyses of the TAs utilized the "MURR 201S reflector model" (Reference IS), which includes the reactor core and associated structures such as detailed modeling of the control blades (CB), regulating blade , beam tubes , beryllium (Be) reflector , graphite reflector , and experimental holes. The MCNP6 model of the MURR core with two (2) T As is shown in Figure 34. The target rod physical dimensions used for the nuclear design calculations are summarized in Table 3. The physical dimensions utilized in the nuclear design are exactly the same as those of the mechanical design , which includes 11 target rods per target rod cartridge (flow channel), cartridge top and bottom fi xture, cartridge guide rail , neutron shield, channel housing , and water plenum. The cartridge locking device , water diffuser , and incoming water pipe are not included in the physics model as they are expected to have little influence on the analyses.
 
===5.2 Target===
Assembly Physics Model The T As are loaded in graphite reflector positions No. SA and No. SB. The physics design and analyses of the TAs utilized the "MURR 201S reflector model" (Reference IS), which includes the reactor core and associated structures such as detailed modeling of the control blades (CB), regulating blade , beam tubes , beryllium (Be) reflector , graphite reflector , and experimental holes. The MCNP6 model of the MURR core with two (2) T As is shown in Figure 34. The target rod physical dimensions used for the nuclear design calculations are summarized in Table 3. The physical dimensions utilized in the nuclear design are exactly the same as those of the mechanical design , which includes 11 target rods per target rod cartridge (flow channel), cartridge top and bottom fi xture, cartridge guide rail , neutron shield, channel housing , and water plenum. The cartridge locking device , water diffuser , and incoming water pipe are not included in the physics model as they are expected to have little influence on the analyses.
The phy sics model of the TA is shown in Figure 3S at the axial mid-plane , and target numbering is shown in Figure 36 for the baseline TA model. --The target rod model includes target pellets, pellet-clad gap, cladding, top plenum spring , and the top and bottom endcaps. In the physics model, four (4) pellets are defined as a single axial node (2.4 cm). The effective isotopic densities of U0 2 pellet consider 1.78% of open pore due to the pellet's dish and chamfer design. The plenum spring was homogeneously mixed with the helium filling gas. Due to the complexity of the geometry of the top and bottom fixtures , the physics model of these components was designed to preserve the outer dimension and mass of the Zircaloy-4 material.
The phy sics model of the TA is shown in Figure 3S at the axial mid-plane , and target numbering is shown in Figure 36 for the baseline TA model. --The target rod model includes target pellets, pellet-clad gap, cladding, top plenum spring , and the top and bottom endcaps. In the physics model, four (4) pellets are defined as a single axial node (2.4 cm). The effective isotopic densities of U0 2 pellet consider 1.78% of open pore due to the pellet's dish and chamfer design. The plenum spring was homogeneously mixed with the helium filling gas. Due to the complexity of the geometry of the top and bottom fixtures , the physics model of these components was designed to preserve the outer dimension and mass of the Zircaloy-4 material.
Page 6S of 190 [Sa , b d, e , fJ ATTACHMENT 1 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane  
Page 6S of 190 [Sa , b d, e , fJ ATTACHMENT 1 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane  
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* i *
* i *
* I [Sa , d , e , f1 0 5 10 1 5 20 25 A x ial no de Figure 63 Power Envelope for Partial Loading for the Extreme Burnup Core Case * * [Sa , d , e , f1 0 5 10 15 20 25 Axial no de Figure 64 Power Envelope of the Partial Loading for the Maximum Burn up Core Case Page 100of190 ATTACHMENT 1 [Sa , d , e , f] Pellet l i near power (kW/m) Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case 5.3.6 Target Assembly Material Depletion As the fissile material bums in the target rods from irradiation by the incident neutron flux from the reactor core, the TA power steadily decreases. Table 27 l ists the isotopic mass of two (2) T As for the average bumup core , which is the most likely core state. As the fissile uranium depletes , the TA power also decreases linearly as shown in Figure 66. At the end of a -irradiation , the TA power drops to-* which corresponds to a 1.7% reduction when compared to the initial power of -for the commercial loading. The target peJlet bumup is [Sa , d, e, f1 -Megawatt day per ton heavy metal (MWd/tHM).
* I [Sa , d , e , f1 0 5 10 1 5 20 25 A x ial no de Figure 63 Power Envelope for Partial Loading for the Extreme Burnup Core Case * * [Sa , d , e , f1 0 5 10 15 20 25 Axial no de Figure 64 Power Envelope of the Partial Loading for the Maximum Burn up Core Case Page 100of190 ATTACHMENT 1 [Sa , d , e , f] Pellet l i near power (kW/m) Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case 5.3.6 Target Assembly Material Depletion As the fissile material bums in the target rods from irradiation by the incident neutron flux from the reactor core, the TA power steadily decreases. Table 27 l ists the isotopic mass of two (2) T As for the average bumup core , which is the most likely core state. As the fissile uranium depletes , the TA power also decreases linearly as shown in Figure 66. At the end of a -irradiation , the TA power drops to-* which corresponds to a 1.7% reduction when compared to the initial power of -for the commercial loading. The target peJlet bumup is [Sa , d, e, f1 -Megawatt day per ton heavy metal (MWd/tHM).
Material Content and Burnup of the Base Target Loading for Average Burnup Core State Calendar Operation Burn up 23su 23su 239Pu Day (MWd/tHM) (grams) (grams) (grams) I -I 461 1,874 0 * -.. 4S7 1 , 874 0.21 [Sa, d , e , f] I -.. 4S7 1 , 874 0.21 .. -.. 4S3 1 , 874 0.42 Page 101 of 190 ATTACHMENT 1 [5a , b , e , f] Operation time (day) Figure 66 Change of2-Target Assembly Power for-Operation 5.3.7 Summary of Target Assembly Nuclear Design Evaluation The nuclear performance of the TA has been evaluated using a single target pellet enrichment of 19. 75 wt% and an 11-rod target loading per cartridge housed in the TA. The evaluation results of the baseline TA model are as follows:
Material Content and Burnup of the Base Target Loading for Average Burnup Core State Calendar Operation Burn up 23su 23su 239Pu Day (MWd/tHM) (grams) (grams) (grams) I -I 461 1,874 0 * -.. 4S7 1 , 874 0.21 [Sa, d , e , f] I -.. 4S7 1 , 874 0.21 .. -.. 4S3 1 , 874 0.42 Page 101 of 190 ATTACHMENT 1 [5a , b , e , f] Operation time (day) Figure 66 Change of2-Target Assembly Power for-Operation  
 
====5.3.7 Summary====
of Target Assembly Nuclear Design Evaluation The nuclear performance of the TA has been evaluated using a single target pellet enrichment of 19. 75 wt% and an 11-rod target loading per cartridge housed in the TA. The evaluation results of the baseline TA model are as follows:
* Two stand-alone target cartridges together are sufficiently subcritical , with a K eff below 0.659.
* Two stand-alone target cartridges together are sufficiently subcritical , with a K eff below 0.659.
* The MURR driver fuel element power peaking factor is Jess than 3.27 when two (2) fresh TAs are loaded.
* The MURR driver fuel element power peaking factor is Jess than 3.27 when two (2) fresh TAs are loaded.
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* Space is e m pty o r mo Dy. vs ed fo r co rt *d ge t h o i s brought up for servic i ng Figure 73 Representation of the Cartridge Loading/Unloading Station Page 1 25 of 1 90 [5 a , b , d ,e,t]
* Space is e m pty o r mo Dy. vs ed fo r co rt *d ge t h o i s brought up for servic i ng Figure 73 Representation of the Cartridge Loading/Unloading Station Page 1 25 of 1 90 [5 a , b , d ,e,t]
ATTACHMENT 1 During normal operations , no more than 22 target rods will be either in a cartridge or in target storage at the cartridge lo ad in g/unloadin g sta tion. Of those 22 target rods , only 11 of them will be fresh. However , the performed criticality analysis assumes 33 target storage positions and two (2) nearb y target cartridges with 11 position s eac h filled with fresh target rods for a total of 55 fresh target rod s in the station at the same time. The criticality analysis shows a maximum K e rr of 0.57 under all conditions of moderation and reflection.
ATTACHMENT 1 During normal operations , no more than 22 target rods will be either in a cartridge or in target storage at the cartridge lo ad in g/unloadin g sta tion. Of those 22 target rods , only 11 of them will be fresh. However , the performed criticality analysis assumes 33 target storage positions and two (2) nearb y target cartridges with 11 position s eac h filled with fresh target rods for a total of 55 fresh target rod s in the station at the same time. The criticality analysis shows a maximum K e rr of 0.57 under all conditions of moderation and reflection.
MURR Technical Specification 5.4.a requires that fueled devices outside the reactor core be stored such that the K e rr i s le ss than 0.9 under all conditions of moderation and r e flection. Administrative controls s imilar to the MURR ad ministrative controls for reactor fuel handling will b e implemented to ensure that the location of any tar ge t rod is known at all times and that the target rod s are placed in the correct irradiation and storage locations.
MURR Technical Specification 5.4.a requires that fueled devices outside the reactor core be stored such that the K e rr i s le ss than 0.9 under all conditions of moderation and r e flection. Administrative controls s imilar to the MURR ad ministrative controls for reactor fuel handling will b e implemented to ensure that the location of any tar ge t rod is known at all times and that the target rod s are placed in the correct irradiation and storage locations.  
7.1.2 Cartridge Loading/Unloading Station Operation The normal sequence of operation at the cartrid ge loadin g/unloading station will be: I. Lower 11 or les s fresh target rods into the sta tion storage location. 2. Insert 11 or less fresh target rods into a cartridge at the station using the remote handlin g tool shown in Figure 71 . 3. Bring an irradiated target cartridge from the target housing. 4. Remove the diffuser from the cartridge and place into storage ba sket. 5. Remove irradiated target rods from the irradiated cartridge and place them int o the station storage location. 6. Move target rods as necessary to complete the placement of the next 11 target rods to be irradiated. (Note: T he next 11 target rods to be irradiated ma y be all fresh or may be a mixture of fresh and previously irradiated target rods.) 7. Move the diffuser from the storage ba ske t and in s tall it onto the target cartridge to be irradiated.  
 
====7.1.2 Cartridge====
 
Loading/Unloading Station Operation The normal sequence of operation at the cartrid ge loadin g/unloading station will be: I. Lower 11 or les s fresh target rods into the sta tion storage location. 2. Insert 11 or less fresh target rods into a cartridge at the station using the remote handlin g tool shown in Figure 71 . 3. Bring an irradiated target cartridge from the target housing. 4. Remove the diffuser from the cartridge and place into storage ba sket. 5. Remove irradiated target rods from the irradiated cartridge and place them int o the station storage location. 6. Move target rods as necessary to complete the placement of the next 11 target rods to be irradiated. (Note: T he next 11 target rods to be irradiated ma y be all fresh or may be a mixture of fresh and previously irradiated target rods.) 7. Move the diffuser from the storage ba ske t and in s tall it onto the target cartridge to be irradiated.  
: 8. Move the loaded tar get cartridge to be irradiated into the target housing at the reactor reflector (target rod remote handling tool shown in Figure 74). Pa ge 126of190 ATTACHMENT 1 Used when removal of lndMdual target rods In shallow end of the pool Is desired Spring-loaded positive lock mechanism Procedure:  
: 8. Move the loaded tar get cartridge to be irradiated into the target housing at the reactor reflector (target rod remote handling tool shown in Figure 74). Pa ge 126of190 ATTACHMENT 1 Used when removal of lndMdual target rods In shallow end of the pool Is desired Spring-loaded positive lock mechanism Procedure:  
: l. Bring tool down to target rod a. PuU on exterior release to claw tool to engage end cop 2. Push down over tar get rod unt il tt bottoms out a. Le t go of exterior release to lock t ool In place 3. Remove t arget rod o. Pul on ex te rior release to f r ee target rod Threaded portion to attach to remo v al tube/r od Exterior release C ompression spring Sliding release mechanism female e nd cop receptacle Figure 74 Target Rod Remote Handling Tool 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing 7.2.1 Target Cartridge Description A detailed de scri ption for the target cartridge and how it fits into the target hou si n g is provided in Section 2.1.4. 7.2.2 Cartridge Installation and Removal Into/From the Target Housing Operation T he nonnal sequence of operation for transferrin g the irradiated target cartridge from the target housin g to the cartridge loading/unloading station will be: I. Shut down the reactor. 2. To allow sufficient cooling and deca y , keep the target cartrid ge in the tar ge t housing for a minimum of one (1) hour after reactor shutdown prior to unlatchin g the cartridge.  
: l. Bring tool down to target rod a. PuU on exterior release to claw tool to engage end cop 2. Push down over tar get rod unt il tt bottoms out a. Le t go of exterior release to lock t ool In place 3. Remove t arget rod o. Pul on ex te rior release to f r ee target rod Threaded portion to attach to remo v al tube/r od Exterior release C ompression spring Sliding release mechanism female e nd cop receptacle Figure 74 Target Rod Remote Handling Tool 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing 7.2.1 Target Cartridge Description A detailed de scri ption for the target cartridge and how it fits into the target hou si n g is provided in Section 2.1.4. 7.2.2 Cartridge Installation and Removal Into/From the Target Housing Operation T he nonnal sequence of operation for transferrin g the irradiated target cartridge from the target housin g to the cartridge loading/unloading station will be: I. Shut down the reactor. 2. To allow sufficient cooling and deca y , keep the target cartrid ge in the tar ge t housing for a minimum of one (1) hour after reactor shutdown prior to unlatchin g the cartridge.  
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Currently, during nonnal operation of MURR, argon-41 (Ar-41) is the principal source of airborne radioactivity  
Currently, during nonnal operation of MURR, argon-41 (Ar-41) is the principal source of airborne radioactivity  
(> 99%) released through the facility ventilation exhaust stack. However, similar to the irradiation of reactor fuel, there exists the possibility to release any one of the fission products created during the fission process from the TEF. Such release would likely be limited to noble gases (krypton and xenon) and the halogens iodine and bromine that exist in the "gap" between the LEU pellets and the target cladding.
(> 99%) released through the facility ventilation exhaust stack. However, similar to the irradiation of reactor fuel, there exists the possibility to release any one of the fission products created during the fission process from the TEF. Such release would likely be limited to noble gases (krypton and xenon) and the halogens iodine and bromine that exist in the "gap" between the LEU pellets and the target cladding.
These releases could occur during two (2) points of the irradiation and use of the targets; (1) during the actual irradiation of the target rods in the graphite reflector region (this would occur if a leak occurs in the cladding of the target rod, thereby allowing the gap fission gases to escape), and (2) during handling of the target rods in the reactor pool pre-and post-irradiation.
These releases could occur during two (2) points of the irradiation and use of the targets; (1) during the actual irradiation of the target rods in the graphite reflector region (this would occur if a leak occurs in the cladding of the target rod, thereby allowing the gap fission gases to escape), and (2) during handling of the target rods in the reactor pool pre-and post-irradiation.  
3.2 Liquid Sources Liquid -Liquid sources would include primary and pool coolant used in the reactor coolant systems and 1he TEF target cooling water, which is essentially reactor pool water that has been pumped to a separate cooling system to facilitate target heat removal. The water becomes activated and contaminated by direct neutron activation of the water molecule itself (H-3) along with any activation of impurities contained in the water due to its use as a coolant. The pool water would also contain contaminants due to its direct Page 129of190 ATTACHMENT 1 contact with the reactor and pool coolant system components.
 
===3.2 Liquid===
Sources Liquid -Liquid sources would include primary and pool coolant used in the reactor coolant systems and 1he TEF target cooling water, which is essentially reactor pool water that has been pumped to a separate cooling system to facilitate target heat removal. The water becomes activated and contaminated by direct neutron activation of the water molecule itself (H-3) along with any activation of impurities contained in the water due to its use as a coolant. The pool water would also contain contaminants due to its direct Page 129of190 ATTACHMENT 1 contact with the reactor and pool coolant system components.
Since primary and pool coolant is, by design, contained to the maximum extent possible, there are no substantial releases of these liquids directly into the environment.
Since primary and pool coolant is, by design, contained to the maximum extent possible, there are no substantial releases of these liquids directly into the environment.
However, certain reactor maintenance activities along with TEF cooling system maintenance could result in small volumes of liquid (containing mainly tritium) being directed to the existing liquid waste retention system. Limited and strictly controlled quantities of liquid radioactive waste are released to the sanitary sewer in accordance with the requirements of 10 CFR 20.2003; this effluent may contain small quantities of pool and primary water. Thus the operation of the TEF would add little to existing liquid waste effluent streams. All potentially radioactive aqueous liquid wastes are directed to a liquid waste retention and disposal system located at a level below the main grade of the MURR laboratory building.
However, certain reactor maintenance activities along with TEF cooling system maintenance could result in small volumes of liquid (containing mainly tritium) being directed to the existing liquid waste retention system. Limited and strictly controlled quantities of liquid radioactive waste are released to the sanitary sewer in accordance with the requirements of 10 CFR 20.2003; this effluent may contain small quantities of pool and primary water. Thus the operation of the TEF would add little to existing liquid waste effluent streams. All potentially radioactive aqueous liquid wastes are directed to a liquid waste retention and disposal system located at a level below the main grade of the MURR laboratory building.
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This same shielding will be used to provide radiation exposure protection from the two (2) target assemblies that will be operating in the graphite reflector region directly outside the reactor pressure vessel and beryllium reflector region. It should be noted however that one TEF assembly (11 target rods) used for the production of Mo-99 via the SGE process contains approximately 263 grams of U-235; approximately 30% of the U-235 contained in one single fresh MURR fuel element. Thus, two (2) full target assemblies with 11 target rods each would contain approximately 7.5 % of the U-235 that is contained in the reactor core with eight (8) fresh fuel elements.
This same shielding will be used to provide radiation exposure protection from the two (2) target assemblies that will be operating in the graphite reflector region directly outside the reactor pressure vessel and beryllium reflector region. It should be noted however that one TEF assembly (11 target rods) used for the production of Mo-99 via the SGE process contains approximately 263 grams of U-235; approximately 30% of the U-235 contained in one single fresh MURR fuel element. Thus, two (2) full target assemblies with 11 target rods each would contain approximately 7.5 % of the U-235 that is contained in the reactor core with eight (8) fresh fuel elements.
This does not count the fuel elements in the reactor pool that are being cooled for either reuse within the reactor or awaiting sufficient cooling and decay for shipment as spent fuel. Thus the addition of the two (2) target assemblies adds less than 7.5% to the entire source term that is contained in the pool due to the use and storage of fuel elements located there. Similar calculations were performed using the computer program MicroShield to predict the dose resulting from the movement of a target cartridge to a position underneath the pool water surface that allows handling and removal of the individual target rods for eventual placement into the transfer cask. These doses are summarized in Table 36 below and provide the worst-case estimate of exposure rates from target cartridge movement activities (e.g., one hour after the end of a three week irradiation of 22 target rods). MURR staff is not expected to be present in the fields given in Table 36 below on a routine or regular basis. Target cartridge movement activities are anticipated to happen twice a week for periods not greater than 30 minutes per occurrence.
This does not count the fuel elements in the reactor pool that are being cooled for either reuse within the reactor or awaiting sufficient cooling and decay for shipment as spent fuel. Thus the addition of the two (2) target assemblies adds less than 7.5% to the entire source term that is contained in the pool due to the use and storage of fuel elements located there. Similar calculations were performed using the computer program MicroShield to predict the dose resulting from the movement of a target cartridge to a position underneath the pool water surface that allows handling and removal of the individual target rods for eventual placement into the transfer cask. These doses are summarized in Table 36 below and provide the worst-case estimate of exposure rates from target cartridge movement activities (e.g., one hour after the end of a three week irradiation of 22 target rods). MURR staff is not expected to be present in the fields given in Table 36 below on a routine or regular basis. Target cartridge movement activities are anticipated to happen twice a week for periods not greater than 30 minutes per occurrence.
It is also expected that each handling activity will be rotated amongst several staff members throughout the year. The maximum expected dose rates are presented in Table 36. Page 133of190 ATTACHMENT 1 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI Dose Location Surface Exposure Rate 1 foot Exposure Rate 1 Exposure Rate (mR/hr) (mR/hr) (mR/hr) Reactor Pool 127 100 61 Biological Shield at 18 13 6.5 Handling Station In conclusion, it is not apparent that any additional major changes are required to the MURR RPP due to the addition of the irradiation facilities that will be used for the SGE TEF. 9. Conduct of Operations 9.1 Procedures Standard Operating Procedures (SOP) will be developed that are specific for operation and maintenance of the SGE TEF and also reactor operation with the TEF. The following topics will be incorporated into procedures for TEF systems installation, testing, operation, and maintenance.
It is also expected that each handling activity will be rotated amongst several staff members throughout the year. The maximum expected dose rates are presented in Table 36. Page 133of190 ATTACHMENT 1 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI Dose Location Surface Exposure Rate 1 foot Exposure Rate 1 Exposure Rate (mR/hr) (mR/hr) (mR/hr) Reactor Pool 127 100 61 Biological Shield at 18 13 6.5 Handling Station In conclusion, it is not apparent that any additional major changes are required to the MURR RPP due to the addition of the irradiation facilities that will be used for the SGE TEF. 9. Conduct of Operations  
 
===9.1 Procedures===
 
Standard Operating Procedures (SOP) will be developed that are specific for operation and maintenance of the SGE TEF and also reactor operation with the TEF. The following topics will be incorporated into procedures for TEF systems installation, testing, operation, and maintenance.
* TEF component/systems installation and testing
* TEF component/systems installation and testing
* TEF startup plan
* TEF startup plan
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* Planned equipment replacement 6 TEF Maintenance and Inspection
* Planned equipment replacement 6 TEF Maintenance and Inspection
* Equipment performance monitoring
* Equipment performance monitoring
* Instrumentation checkout and calibration 9.2 Target Experimental Facility Commissioning Plan A high-level commissioning plan is presented in this section. A detailed startup and commissioning plan will be prepared and approved by the Reactor Safety Procedure Review Subcommittee (a subcommittee of the Reactor Advisory Committee) prior to startup of the TEF. Appropriate hold points will be identified in the plan for review of data and for approval to proceed to the next step of the startup plan. Page 135 of 190 ATTACHMENT 1 9.2.1 Assumptions
* Instrumentation checkout and calibration  
 
===9.2 Target===
Experimental Facility Commissioning Plan A high-level commissioning plan is presented in this section. A detailed startup and commissioning plan will be prepared and approved by the Reactor Safety Procedure Review Subcommittee (a subcommittee of the Reactor Advisory Committee) prior to startup of the TEF. Appropriate hold points will be identified in the plan for review of data and for approval to proceed to the next step of the startup plan. Page 135 of 190 ATTACHMENT 1 9.2.1 Assumptions
* Secondary Coolant System modifications complete
* Secondary Coolant System modifications complete
* TCS installation complete outside of the reactor pool
* TCS installation complete outside of the reactor pool
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-The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and (4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, (4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and (5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation.
-The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and (4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, (4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and (5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation.
1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target cartridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-enriched uranium dioxide pellets nominally enriched to 19.75% in the isotope uranium-235 with a nominal active length of 600 millimeters.
1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target cartridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-enriched uranium dioxide pellets nominally enriched to 19.75% in the isotope uranium-235 with a nominal active length of 600 millimeters.
2 of3 __J ATTACHMENT 2 IV. The following new TS Sections will be added (see attached pages): 3.11 Selective Gas Extraction Target Experimental Facility (Limiting Conditions of Operations) 4.11 Selective Gas Extraction Target Experimental Facility (Surveillance Requirements) 5.7 Selective Gas Extraction Target Experimental Facility (Design Features) 3 of3 1 DEFINITIONS  
2 of3 __J ATTACHMENT 2 IV. The following new TS Sections will be added (see attached pages): 3.11 Selective Gas Extraction Target Experimental Facility (Limiting Conditions of Operations) 4.11 Selective Gas Extraction Target Experimental Facility (Surveillance Requirements)  
 
===5.7 Selective===
 
Gas Extraction Target Experimental Facility (Design Features) 3 of3 1 DEFINITIONS  
-Continued ATTACHMENT 2 IB;JIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS DocketNo.
-Continued ATTACHMENT 2 IB;JIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS DocketNo.
50-186, License No. R-103 the principal physical barriers which guard against the uncontrolled release of radioactivity.
50-186, License No. R-103 the principal physical barriers which guard against the uncontrolled release of radioactivity.
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The manual scram provides assurance that the reactor can be shut down by the operator if an automatic function fails to initiate a reactor scram or if the operator detects an impending unsafe condition prior to the initiation of an automatic scram. The center test hole scram provides assurance that the reactor cannot be operated unless the removable experiment sample canister or the strainer is inserted and latched in the center test hole. This is required anytime the reactivity worth of the center test hole removable experiment sample canister and the contained experiments or the strainer exceeds the limit of Specification 3.8.b (Ref. Section 13.2.2 of the SAR). The center test hole scram may be bypassed if the total reactivity worth of the removable experiment sample canister and the contained experiments or the strainer does not exceed the limit of Specification 3.8.b and is authorized by the Reactor Manager. The SGE target assembly coolant high and low flow scrams provides assurance that the reactor will be shut down on a loss of target coolant or forced cooling flow (Ref. Section 10 of Amendment No. X). The SGE heat exchanger high outlet water temperature scram provides assurance that the reactor will be shut down on a loss of secondary cooling flow to the SGE Target Experimental Facility.
The manual scram provides assurance that the reactor can be shut down by the operator if an automatic function fails to initiate a reactor scram or if the operator detects an impending unsafe condition prior to the initiation of an automatic scram. The center test hole scram provides assurance that the reactor cannot be operated unless the removable experiment sample canister or the strainer is inserted and latched in the center test hole. This is required anytime the reactivity worth of the center test hole removable experiment sample canister and the contained experiments or the strainer exceeds the limit of Specification 3.8.b (Ref. Section 13.2.2 of the SAR). The center test hole scram may be bypassed if the total reactivity worth of the removable experiment sample canister and the contained experiments or the strainer does not exceed the limit of Specification 3.8.b and is authorized by the Reactor Manager. The SGE target assembly coolant high and low flow scrams provides assurance that the reactor will be shut down on a loss of target coolant or forced cooling flow (Ref. Section 10 of Amendment No. X). The SGE heat exchanger high outlet water temperature scram provides assurance that the reactor will be shut down on a loss of secondary cooling flow to the SGE Target Experimental Facility.
A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued  
A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued  
: h. Specification 3 .2.h assures that certain system conditions have been met prior to conducting a reactor startup (Master Control Switch lSl in the "ON" position; No Nuclear Instrument anomaly; Shim rods bottomed and in contact with their electromagnets; Source range level indication greater than 20 cps or intermediate range level recorder indication greater than 2 x 10-5% power; and Thermal Column door closed) or placing the reactor in automatic control at power (Reactor period as indicated by Intermediate Range Channels 2 and 3 greater than 3 5 seconds; Indicated reactor power level greater than the "auto control prohibit" set point on the wide range neutron flux monitor recorder; Regulating blade position greater than 60% withdrawn; and Range Selector Switch 1 S2 in the 5-kW red scale position or above) (Ref. Sections 7.5.3.1 and 7.5.4 of the SAR). A-X 3.8 Experiments Applicability:
: h. Specification 3 .2.h assures that certain system conditions have been met prior to conducting a reactor startup (Master Control Switch lSl in the "ON" position; No Nuclear Instrument anomaly; Shim rods bottomed and in contact with their electromagnets; Source range level indication greater than 20 cps or intermediate range level recorder indication greater than 2 x 10-5% power; and Thermal Column door closed) or placing the reactor in automatic control at power (Reactor period as indicated by Intermediate Range Channels 2 and 3 greater than 3 5 seconds; Indicated reactor power level greater than the "auto control prohibit" set point on the wide range neutron flux monitor recorder; Regulating blade position greater than 60% withdrawn; and Range Selector Switch 1 S2 in the 5-kW red scale position or above) (Ref. Sections 7.5.3.1 and 7.5.4 of the SAR). A-X
 
===3.8 Experiments===
 
Applicability:
ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 This specification applies to all experiments which directly utilize neutrons or other radiation produced by the reactor. Radioactive sources shall meet the requirements for experiments.
ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 This specification applies to all experiments which directly utilize neutrons or other radiation produced by the reactor. Radioactive sources shall meet the requirements for experiments.
Objective:
Objective:
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: 3. 8 .h provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). L Specification 3.8.i is intended to reduce the likelihood of damage to reactor or pool components resulting from the detonation of explosive materials (Ref. Section 13.2.6 of the SAR). J. Specification 3.8.j provides assurance that no chemical reaction will take place to adversely affect the reactor or its components.  
: 3. 8 .h provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). L Specification 3.8.i is intended to reduce the likelihood of damage to reactor or pool components resulting from the detonation of explosive materials (Ref. Section 13.2.6 of the SAR). J. Specification 3.8.j provides assurance that no chemical reaction will take place to adversely affect the reactor or its components.  
: k. The extremely low temperatures of the cryogenic liquids present structural problems that enhance the potential of an experiment failure. Specification 3.8.k provides for the proper review of proposed experiments containing or using cryogenic materials.  
: k. The extremely low temperatures of the cryogenic liquids present structural problems that enhance the potential of an experiment failure. Specification 3.8.k provides for the proper review of proposed experiments containing or using cryogenic materials.  
: 1. Specification 3.8.1 provides assurance that the integrity of the beamports will be maintained for all loop-type experiments.  
: 1. Specification  
 
====3.8.1 provides====
assurance that the integrity of the beamports will be maintained for all loop-type experiments.  
: m. Specification 3.8.m assures that corrosive materials which are chemically incompatible with reactor components, highly flammable materials, and toxic materials are adequately controlled and that this information is disseminated to all reactor users. A-X ATTACHMENT2 UNNERSITY OF MISSOURI RESEARCH REACTOR 1 TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments  
: m. Specification 3.8.m assures that corrosive materials which are chemically incompatible with reactor components, highly flammable materials, and toxic materials are adequately controlled and that this information is disseminated to all reactor users. A-X ATTACHMENT2 UNNERSITY OF MISSOURI RESEARCH REACTOR 1 TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments  
-Continued  
-Continued  
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Math Solid Works 2016 for the detailed mechanical design and 3D visualization.
Math Solid Works 2016 for the detailed mechanical design and 3D visualization.
PIPE for structural analysis of the Target Cooling System. Auto The use of al 1 software for the development of the SGE TEF has been subjected to the rigorous software ance (QA) verification and validation procedures as required by the GA QA Program and nents (Section 3), including preparation of verification and validation reports as required by quality assur subtier docm the applicable requirements in the engineering procedures (Ref. 1). 2.2 Desig n Documentation Design bases , including all assumptions for the design bases, as well as documentation of software alculations that fonn the bases of the design and safety information presented in the license re provided in the following reports listed in Table 1. 2 This table also includes other cuments. verification c amendment a applicable do Table 1 Design and Software Validation Reports and Other Applicable Documents 2 REPORT NO. 30441ROO 017 30441ROO 019 30441ROO 021 30441ROO 022 30441ROO 030 30441ROO 031 30441ROO 032 30441ROO 033 30441ROO 035 30441ROO 038 DESIGN CALCULATION REPORT TITLE ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report Target System Cooling Calculation Report Target Assembly Thermal Analysis Source Term Analysis Design Calculation Report Mo-99 Target Cooling System Seismic Analysis Design Calculation Report Mo-99 Target Assembly Nuclear Design for Once-Through Operation RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations Cooling of the MURR Beryllium Reflector Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report All reports 1 isted in Table 1 are considered GA Proprietary Information and will be appropriately marked. 2 of5 ATTACHMENT3 Table 1 (con't) Design and Software Validation Reports and Other Applicable Documents REPORT NO. DESIGN CALCULATION REPORT TITLE 30441M00043 ANSYS Thennal Model ofRB-MSS Target Rod REPORT NO. SOFTWARE VERIFICATION TEST 30441R00002 MCNP6 (Version 1.0) Verification 30441R00003 ORIGEN2 (Version 2.2) Verification 30441R00006 MCODE (Version 2.4) Verification 30441R00020 ANSYS Workbench (Version 16.0) Verification 30441R00023 OTSTM (Version 1.0) Verification 30441R00024 RELAP (Version 5 Mod. 3.3 P03) Verification 30441R00025 FRAPCON (Version 4.0) Verification 30441R00026 FRAPTRAN (Version 2.0) Verification 30441R00028 ANSYS-FLUENT (Version 16.0) Verification 30441R00029 SINDA (Version 2.2) Verification 30441R00034 OTSTM (Version 1.0) Software Design Requirements, Description and Verification REPORT NO. OTHER APPLICABLE DOCUMENTS 30441S00001 Molybdenum-99 Supply System Requirements Document QAPD-30441-II Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System (RB-MSS) / 3. Quality Assurance GA conducted the analysis and design, and will fabricate the required SSCs under a QA program that ensures conformance, through a documented system of QA requirements, specifications and inspections, with the applicable requirements of ASME NQA-1 2008/-la-2009 Addenda that are contained in the GA Quality Assurance Manual (QAM) (Ref. 2) and the accompanying Quality Division Instructions (QDI) (Ref. 3). GA, as the prime contractor for the supply of the SGE TEP SSCs, will follow a graded approach in the implementation of its QA program commensurate with the Quality Assurance Levels (QAL) requirements on both safety and performance of the facility.
PIPE for structural analysis of the Target Cooling System. Auto The use of al 1 software for the development of the SGE TEF has been subjected to the rigorous software ance (QA) verification and validation procedures as required by the GA QA Program and nents (Section 3), including preparation of verification and validation reports as required by quality assur subtier docm the applicable requirements in the engineering procedures (Ref. 1). 2.2 Desig n Documentation Design bases , including all assumptions for the design bases, as well as documentation of software alculations that fonn the bases of the design and safety information presented in the license re provided in the following reports listed in Table 1. 2 This table also includes other cuments. verification c amendment a applicable do Table 1 Design and Software Validation Reports and Other Applicable Documents 2 REPORT NO. 30441ROO 017 30441ROO 019 30441ROO 021 30441ROO 022 30441ROO 030 30441ROO 031 30441ROO 032 30441ROO 033 30441ROO 035 30441ROO 038 DESIGN CALCULATION REPORT TITLE ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report Target System Cooling Calculation Report Target Assembly Thermal Analysis Source Term Analysis Design Calculation Report Mo-99 Target Cooling System Seismic Analysis Design Calculation Report Mo-99 Target Assembly Nuclear Design for Once-Through Operation RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations Cooling of the MURR Beryllium Reflector Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report All reports 1 isted in Table 1 are considered GA Proprietary Information and will be appropriately marked. 2 of5 ATTACHMENT3 Table 1 (con't) Design and Software Validation Reports and Other Applicable Documents REPORT NO. DESIGN CALCULATION REPORT TITLE 30441M00043 ANSYS Thennal Model ofRB-MSS Target Rod REPORT NO. SOFTWARE VERIFICATION TEST 30441R00002 MCNP6 (Version 1.0) Verification 30441R00003 ORIGEN2 (Version 2.2) Verification 30441R00006 MCODE (Version 2.4) Verification 30441R00020 ANSYS Workbench (Version 16.0) Verification 30441R00023 OTSTM (Version 1.0) Verification 30441R00024 RELAP (Version 5 Mod. 3.3 P03) Verification 30441R00025 FRAPCON (Version 4.0) Verification 30441R00026 FRAPTRAN (Version 2.0) Verification 30441R00028 ANSYS-FLUENT (Version 16.0) Verification 30441R00029 SINDA (Version 2.2) Verification 30441R00034 OTSTM (Version 1.0) Software Design Requirements, Description and Verification REPORT NO. OTHER APPLICABLE DOCUMENTS 30441S00001 Molybdenum-99 Supply System Requirements Document QAPD-30441-II Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System (RB-MSS) / 3. Quality Assurance GA conducted the analysis and design, and will fabricate the required SSCs under a QA program that ensures conformance, through a documented system of QA requirements, specifications and inspections, with the applicable requirements of ASME NQA-1 2008/-la-2009 Addenda that are contained in the GA Quality Assurance Manual (QAM) (Ref. 2) and the accompanying Quality Division Instructions (QDI) (Ref. 3). GA, as the prime contractor for the supply of the SGE TEP SSCs, will follow a graded approach in the implementation of its QA program commensurate with the Quality Assurance Levels (QAL) requirements on both safety and performance of the facility.
The QA requirements from GA's QA program, and the customer requirements, are implemented via project specific Quality Assurance Project Documents (QAPD) which establishes GA's QA requirements applicable to SGE project specific design, analysis, and fabrication related activities (Refs. 1 and 4). 3 of5 ATTACHMENT3 3.1 Quality Assurance Level of Safety-Related Components Project/Resource Procedures Manual, procedure EP-4010 (Ref. 5), discusses the determination and assignment of QA levels. Three characteristics make SSCs safety-related.
The QA requirements from GA's QA program, and the customer requirements, are implemented via project specific Quality Assurance Project Documents (QAPD) which establishes GA's QA requirements applicable to SGE project specific design, analysis, and fabrication related activities (Refs. 1 and 4). 3 of5 ATTACHMENT3  
 
===3.1 Quality===
Assurance Level of Safety-Related Components Project/Resource Procedures Manual, procedure EP-4010 (Ref. 5), discusses the determination and assignment of QA levels. Three characteristics make SSCs safety-related.
If an SSC has one or more of these characteristics, then it is considered to be safety-related.  
If an SSC has one or more of these characteristics, then it is considered to be safety-related.  
: 1. Maintains the integrity of the reactor coolant boundary (the reactor vessel and associated piping that circulates the reactor coolant).  
: 1. Maintains the integrity of the reactor coolant boundary (the reactor vessel and associated piping that circulates the reactor coolant).  
: 2. Has the capability to shut down the reactor and maintain it in a safe condition.  
: 2. Has the capability to shut down the reactor and maintain it in a safe condition.  
: 3. Has the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures.
: 3. Has the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures.
EP-4010 also states that components which contain radioactive materials are classified as QAL I as well as SSCs designated engineered safety features.
EP-4010 also states that components which contain radioactive materials are classified as QAL I as well as SSCs designated engineered safety features.  
3.1.1 Selective Gas Extraction Target Experimental Facility QAL 1 Components As the result of the engineering detennination and assessment to identify all of the equipment required for the SGE TEF, the following components are considered to meet the QAL I designation:
 
====3.1.1 Selective====
 
Gas Extraction Target Experimental Facility QAL 1 Components As the result of the engineering detennination and assessment to identify all of the equipment required for the SGE TEF, the following components are considered to meet the QAL I designation:
* The target rods' cladding, endcaps (upper and lower) and welding of the endcaps.
* The target rods' cladding, endcaps (upper and lower) and welding of the endcaps.
* The target pellets.
* The target pellets.
* The flow meters in the target cooling system.
* The flow meters in the target cooling system.
* Temperature sensor downstream of the heat exchanger in the target cooling system, including its signal conditioner.
* Temperature sensor downstream of the heat exchanger in the target cooling system, including its signal conditioner.  
3.2 Manufacture of Low-Enriched Uranium Target Rods In addition to following the graded approach for SSCs under the QAL approach mandated by GA's NQA-1 compliant QA program and implemented via the project specific QAPDs, the fabrication of enriched uranium (LEU) target rods will be subjected to a the following Quality Control procedures:
 
3.2.1 Target Pellet Fabrication
===3.2 Manufacture===
 
of Low-Enriched Uranium Target Rods In addition to following the graded approach for SSCs under the QAL approach mandated by GA's NQA-1 compliant QA program and implemented via the project specific QAPDs, the fabrication of enriched uranium (LEU) target rods will be subjected to a the following Quality Control procedures:  
 
====3.2.1 Target====
Pellet Fabrication
* Qualification of the manufacturing process.
* Qualification of the manufacturing process.
* Ensuring traceability of pellet lots through a system of fabrication travelers.
* Ensuring traceability of pellet lots through a system of fabrication travelers.

Revision as of 03:43, 9 October 2018

University of Missouri-Columbia Research Reactor - Requesting NRC Approval to Amend Technical Specifications. Redacted Attachments 1-3 Re License Amendment Request, Revised and New Technical Specifications and Codes and Standards Enclosed
ML17132A241
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Site: University of Missouri-Columbia
Issue date: 05/03/2017
From: Meffert B A
Univ of Missouri
To:
Office of Nuclear Reactor Regulation
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Text

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER May 3, 2017 U.S. Nuclear }\egulatory Commission Attention:

Document Control Desk Mail Station Pl-37 Washington, DC 20555-0001

REFERENCE:

Docket No. 50-186

SUBJECT:

University of Missouri-Columbia Research Reactor Renewed Facility Operating License No. R-103 Written communication as specified by 10 C.F.R. 50.4(b)(l) requesting U.S. Nuclear Regulatory Commission approval to amend the Technical Specifications appended to Renewed Facility Operating License No. R-103 pursuant to 10 C.F.R. § 50.59(c) and 10 C.F.R. § 50.90 1. Introduction By letter dated March 27, 2017, the University of Missouri Research Reactor (MURR) submitted a license amendment request (LAR) in order to conduct an experiment that would produce molybdenum-99 (Mo-99) in large quantities as part of its role in supplying critical medical radioisotopes to the domestic and international community.

The experiment would utilize General Atomics' (GA) Selective Gas Extraction (SGE) process, which consists of irradiating target rods containing low-enriched uranium (LEU) pellets in the reactor graphite reflector region in order to produce fission product Mo-99. The Mo-99 would then be extracted from the LEU in*hot cells using the SGE technology (which does not require chemical dissolution of the LEU targets, thus avoiding the creation of large quantities of mixed liquid radioactive waste).1 Also included with the LAR submittal was an affidavit executed by Keith E. Asmussen which requested that certain information contained in six (6) attachments (Attachment 1, 4, 5, 6, 7 and 8) to the LAR be withheld from public disclosure pursuant to the 10 C.F.R. § 2.390. By letter dated April 19, 2017, the U.S. Nuclear Regulatory Commission (NRC) staff stated that they had reviewed the LAR application and affidavit

[Agencywide Documents Access and Management System (ADAMS) Accession Nos. MLl 7089A228 and MLl 7089A229]

in accordance with the requirements of 10 C.F.R. § 2.390 and concluded that insufficient justification was provided to determine that all the Currently, much of the supply ofMo-99 is produced using highly enriched uranium (HEU) targets. Because of global nonproliferation concerns, the future availability of HEU target material is uncertain.

Research and development of suitable LEU targets for producing Mo-99 is the focus of various U.S. and foreign organizations.

1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: http://web.missouri.edu/-murrwww Fighting Cancer with Tomorrow's Technology information sought to be withheld from public disclosure was proprietary information.

As such, the NRC staff concluded that the information to be withheld from public disclosure does not meet the requirements of 10 C.F.R. § 2.390. By this letter, MURR is formally withdrawing the LAR, all of its attachments (redacted and unredacted versions);

and the affidavit that was submitted to the NRC by letter dated March 27, 2017. In its place, MURR is submitting all new attachments (both redacted and unredacted versions) and revised affidavit to resolve the deficiencies noted in the NRC letter dated April 19, 2017. The remainder of this introductory letter remains essentially unchanged from the March 27, 2017 MURR letter. As discussed previously and described in detail below, this amendment request would only apply to the "in-pool" portion of the project, or Part 1, where the LEU target rods would be irradiated in the reflector region and cooled by a dedicated cooling system. The "ex-pool" portion of the project, or Part 2 (which will be addressed in a subsequent license amendment), will involve the installation of (1) the hot cells necessary for the extraction of Mo-99 from the LEU target rods using the SGE process, and (2) the hot cells that would store the radioactive waste for decay and eventual shipment.

It is also MURR's understanding, based on conversations with the NRC, that the NRC will not approve the Part 1 amendment request for implementation until the NRC also approves the Part 2 amendment request. 2. Licensing Approach Background Representatives from MURR met with NRC staff on three separate occasions

-April 27, 2015, June 2, 2016 and February 13, 2017 -to discuss MURR's approach to licensing the SGE experimental facility and provide an overview of the technical details of the project. At the end of the April 27, 2015 public meeting, the NRC suggested that MURR submit a written document outlining how MURR proposes to license the SGE experimental facility and the bases for that approach.

By letter dated September 11, 2015, MURR provided a summary of MURR's licensing approach and the bases for requesting NRC approval pursuant to 10 C.F.R. § 50.90 by amending MURR's existing Facility Operating License. The September 11, 2015 letter outlined why MURR believed that the SGE activities at MURR would (1) be classified as an "experiment," (2) not require MURR to be categorized as a "testing facility," (3) not cause MURR to be categorized as a "production facility," (4) not require a construction permit per 10 C.F.R. § 50.90, and (5) not increase MURR's reactor core licensed power level limit of 10 MWt. During the June 2, 2016 public meeting, MURR presented a project update and restated our intended licensing approach, consistent with our September 11, 2015 letter. Also during that meeting, MURR proposed a 2-part licensing submission (In-Pool vs. Ex-Pool activities) for review efficiency such that the in-pool review could commence more promptly (while MURR completed its Part 2 information).

MURR representatives left the June 2, 2016 meeting with the understanding that (1) the NRC was amenable to a 2-part submission based on the clear separation of in-pool and ex-pool activities, (2) the NRC would consider whether a construction permit was necessary based on any "alterations" or "material alterations" proposed by MURR, and (3) that the NRC staff felt strongly at the time that the ex-pool activities would require a "production facility" license. 2of11 I ___ __J By letter dated January 17, 2017, MURR stated its position and associated bases for conclusions on the following four questions and requested formal responses from the NRC: 1. Will a Construction Permit be required?

(10 C.F.R. § 50.23) 2. Will a change be required to MURR's licensed maximum operating power limit of 10 MWt due to the heat produced from the irradiation of the LEU targets? 3. Will a Production Facility license be required?

4. Will a change from a class 104c license (10 C.F.R. 50.21) to class 103 (10 C.F.R. 50.22) be required? (non-commercial vs. commercial)

By letter dated February 2, 2017, the NRC stated that they had reviewed MURR's request, as well as the information provided by MURR and its partners, GA and Nordion (Canada), in the June 2, 2016 public meeting, and that the NRC needed additional information to better understand MURR's licensing approach.

In this letter the NRC proposed discussion topics for a February 13, 2017 public meeting. On February 13, 2017, MURR met with the NRC to further discuss the topics proposed by the NRC in the February 2, 2017 letter. In consideration of the February 13, 2017 meeting discussions, MURR provides herein its licensing approach for this amendment request [i.e., Part 1 ("in-pool")]

and discusses the subsequent to-be submitted Part 2 (i.e., "ex-pool")

license amendment request, which collectively address the SGE at MURR. 3. Basis for the License Amendment-Part 1: "In-Pool" Activities

3.1 Construction

Permit Not Required for "In-pool" Activities Implementation of Part 1 of the SGE experiment, or the "in-pool" activities, will be performed under an amendment to Renewed Facility Operating License No. R-103. MURR concludes that a construction permit is not required because implementation of the experiment will not necessitate a "material alteration" to the MURR facility per 10 C.F.R. § 50.92(a).

The main components necessary for the "in-pool" activities will consist of LEU target assemblies, a separate and dedicated cooling system for heat removal (forced and natural convection), and instrumentation and control equipment to support these systems. NRC regulation 10 C.F.R. § 50.92(a), Issuance of Amendment states, in part: In determining whether an amendment to a license, construction permit, or early site permit will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses, construction permits, or early site permits to the extent applicable and appropriate.

If the application involves the material alteration of a licensed facility, a construction permit will be issued before the issuance of the amendment to the license .... 3of11 (Emphasis added.) MURR is not aware of any NRC definition of "material alteration." The NRC Staff considered defining "material alteration" in 1995 -1996, but apparently abandoned the effort because of "little regulatory need to clarify the term." 2 As the NRC's Executive Director for Operations explained in SECY-96-024, In only one instance has a construction permit been issued before an amendment of an operating license, that is, an amendment to the operating license of a research reactor at the University of Maryland.

The material alteration was the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement with those of a different design. The change rendered major portions of the safety analysis inapplicable.

3 The Staff concluded that there was an "apparent lack of need and industry or public interest in this topic." Material alterations over four decades ago at the University of Maryland which required a construction permit (issued March 25, 1970) prior to the issuance of the amendment (March 2, 1971) involved significant modifications to the reactor itself and its support systems. These modifications included the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement of those configurations with those of a different design, and major revisions to the associated accident analyses.

Significant modifications also occurred at the Massachusetts Institute of Technology Reactor (MITR) where a construction permit was issued on April 9, 1973 prior to issuance of the license amendment (July 23, 1975). In this case the material alteration consisted of replacing the heavy-water moderated, cooled and reflected core with a light-water moderated, cooled, water reflected core, and modifications of installed systems to meet regulatory requirements.

The change also rendered major portions of the safety analysis inapplicable.

These examples involve significantly greater modifications than what will exist at MURR .. In contrast to the University of Maryland and MITR examples, the SGE experiment does not involve the complete removal of major systems, strm;tures, or components (SSCs) such as control rods or core instrumentation.

The neutron producing core region of MURR will be minimally affected by the addition of the equipment for the experiment because the target rods will be located in the graphite reflector region of the reactor and used in a manner similar to other reflector region experiments.

The installation and operation of this experiment utilizes a supporting target cooling system, and associated instruments and controls, that will better ensure that existing SSCs along with the design bases function of the SSCs as described in the MURR Safety Analysis Report (SAR) will not be adversely affected by the SGE experiment.

As such, these changes do not constitute a significant modification to the facility when comparing them to previous NRC actions. SECY-96-024, "Semiannual Status Report on Implementation of Regulatory Review Group Recommendations," at 5 (Feb. 2, 1996). Id. 4of11 3.2 The Experiment Will Not Result In a Change to MURR's Operating Power Limit of 10 MW It is MURR' s position that no change to the licensed steady-state operating power level limit of 10 MWt will be required.

MURR uses the following calorimetric procedure to determine and maintain reactor power within the licensed limit using both primary and pool coolant system flow rates and differential temperatures:

Reactor Power Level = [((total average primary coolant flow of each loop -primary demineralizer flow) x (Th -Tc)Primary)

+ ((average pool coolant flow) x (Th -Tc)Poo1)]

x (a unit conversion factor) An additional backup calorimetric procedure, using secondary coolant system flow rate and differential temperature, is also performed:

Reactor Power Level= [(secondary coolant flow) x (Th -Tc)secondary]

x (a unit conversion factor) The SGE experiment will generate approximately 562 kWt of fission power from two (2) target assemblies.

A dedicated cooling system will provide forced flow to each target assembly.

This system will draw coolant from the MURR reactor pool, pump it through dedicated heat exchangers which will lower the temperature of the coolant to the required value, then circulate the coolant through the target assemblies.

The coolant leaving the target assemblies will then reenter the reactor pool at, or slightly above, the temperature of the bulk reactor pool water. Because the system is designed such that no heat is removed from the pool coolant system, the power generated by the target assemblies will not be part of MURR's 10 MWt reactor power level limit. MURR's secondary coolant system, which cools both the primary and pool coolant systems, will also provide the cooling water to the SGE experiment heat exchangers.

The SGE experiment's cooling system is instrumented with flow, pressure and temperature monitoring equipment. Flow to each target assembly will be measured by flow elements.

Temperature will be measured at four ( 4) separate locations:

heat exchanger inlet, heat exchanger outlet, target assembly inlets and target assembly outlets. Secondary coolant to the SGE experimental facility cooling system will also be instrumented with heat exchanger inlet and outlet temperature elements and a flow element. MURR will use the same type of calorimetric procedure to determine SGE experimental facility power (individual target assembly power and total power) using coolant system flow rates and differential temperatures:

Target Assembly A Power= [(target assembly coolant flow rate) x (target assembly differential temperature

-Th -T charget Assembly A] x (a unit conversion factor) Target Assembly B Power= [(target assembly coolant flow rate) x (target assembly differential temperature -Th -Tc)TargetAssemblyB]

x (a unit conversion factor) 5of11 SGE Experimental Facility Total Power = Target Assembly A Power + Target Assembly B Power An additional backup calorimetric procedure, using secondary coolant system flow rate and differential temperature across the SGE experiment heat exchanger, will also performed:

SGE Experimental Facility Total Power = [(secondary coolant flow rate through SGE experimental heat exchanger) x (Th -Tc)secondary]

x (a unit conversion factor) Redundant N-16 power level monitoring systems will also be installed on the primary coolant system piping exiting the reactor core. The N-16 monitoring systems will be calibrated at 100% reactor power operation without the SGE experimental facility operating.

This will provide an additional method to the currently installed nuclear instrumentation to monitor reactor power. 4. Basis for the License Amendment -Part 2: "Ex-Pool" Activities This discussion is included with the Part 1 license amendment submittal to provide the NRC with a more complete picture of the total request (i.e., Part 1 and Part 2) at this time. In the Part 2 license amendment request, MURR plans to submit an application that includes a request for the Commission to approve an amendment to its existing utilization facility license to expand the authorized activities to include possession, use, and operation of the facility under a combined production and utilization facility license. Presently, Renewed Facility Operating License No. R-103 authorizes MURR to "possess, use, and operate the facility as a utilization facility." 4 The Commission has previously endorsed the use of a single, combined production and utilization facility license for medical isotope production facilities.

5 The combined facility will be comprised almost entirely of SSCs that exist at MURR already. Only a few physical changes would be necessary to conduct the "ex-pool" activities

-namely, installation of hot cells in the MURR facility.

Hot cells routinely have been installed at MURR over the years without the need for a construction permit, and usually have been implemented via the 10 C.F.R. § 50.59 process.6 As explained further below, for the SGE experiment, such minor physical changes to the existing facility should not trigger the 10 C.F.R. § 50.lO(c) requirement for a construction permit to "begin" construction of the combined facility, nor should it trigger the 10 C.F.R. § 50.92(a) requirement for a construction permit for "material alterations" to the facility.

4.ll A Separate Production Facility License is Not Required The main components of the "ex-pool" phase of the SGE experiment will consist of hot cells and ancillary equipment to support those hot cells. LEU target rods irradiated during the "in-pool" phase will be transferred to the hot cells for processing and extraction of the Mo-99. NRC regulations at 10 C.F .R. § 4 Renewed Facility Operating License No. R-103 if 2.B.(2). See SRM-SECY-09-0101, Licensing of a Babcock and Wilcox Medical Isotope Production System at page 1 (Oct. 9, 2009). See, e.g., Safety Evaluation Report, Renewal of the Facility Operating License for the University of Missouri-Columbia Research Reactor at pages 10-4 to 10-7 (Jan. 2017). 6of11 50.lO(b) require a license to "transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, or use any production or utilization facility." The NRC defines "production facility" in 10 C.F.R. § 50.2 to include "[a]ny facility designed or used for the processing of irradiated materials containing special nuclear material," subject to certain exceptions that are inapplicable to the SGE experiment.

Notwithstanding these activities being part of an "experiment" as defined by the facility Technical Specifications, because of the quantity of batch material (and associated SNM) processed as part of the experiment in the hot cells, an NRC "production facility" license appears to be the most efficient pathway to licensing this SGE experiment.

MURR is not aware of any NRC regulation or policy that would prohibit including both a production and utilization facility in one Part 50 license. In fact, as discussed below, the NRC previously considered su.ch issues related to licensing medical isotope production facilities, and recommended using a single, combined production and utilization facility license in appropriate circumstances.

7 4.1.1 The Commission Previously Has Approved the Use of a Combined License In 2007, Babcock and Wilcox (B& W) nqtified the Commission of its intent to seek a license to construct and operate a Medical Isotope Production System (MIPS). Prior to submitting its application, B& W requested the NRC' s views regarding several issues, including the possibility of licensing the MIPS under a single combined utilization and production facility license. In SECY-09-0101, the Staff provided the Commission with an assessment of licensing issues regarding the B& W request. In its evaluation of this request, NRC staff "examined whether it is legally feasible to issue one 10 C.F.R. Part 50 operating license for the entire MIPS system that incorporates both the utilization and extraction I purification portions of the facility." 8 "The staff conclude[d]

that there is no legal impediment under section 161.h of the AEA to issuing one 10 C.F .R. Part 5 0 operating license for the entire MIPS facility (i.e., numerous reactors and one or more production facilities)." 9 The NRC Staff also concluded that this licensing strategy should be considered on a case-by-case basis, and that it would need to be implemented

via Commission order." The Commission ultimately approved the Staffs approach.10 Similarly, no legal impediment exists to utilizing the current Part 50 license for utilization and production activities at MURR. 4.Jl.2 A Combined License is Appropriate for the SGE Experiment The SGE experiment is an appropriate scenario for using the single license approach.

The benefit of this approach is that it provides a practical means to regulate common systems shared among and between the production and utilization facilities.

The proposed "ex-pool" activities will share certain common systems with the existing reactor at MURR. Dividing up those systems among separate licenses would be unnecessarily burdensome with no commensurate enhancement to safety or security.

Furthermore, 7 9 IO SECY-09-0101, Licensing ofa Babcock and Wilcox Medical Isotope Production System at page 3 (July 9, 2009). Id. Id. SRM-SECY-09-0101 at I. 7of11 approval of a single, combined production and utilization facility license would streamline the requirements applicable to MURR and promote regulatory efficiency.

4.2 A Construction Permit is Not Required for "Ex-Pool" Activities NRC regulations contain two provisions related to the potential requirement to obtain a construction permit. For new production or utilization facilities, 10 C.F.R. § 50.IO(c) requires a construction permit before breaking ground on a new facility (such as the MIPS, discussed above). More specifically, 10 C.F.R. § 50.IO(c) states: No person may begin the construction of a production or utilization facility on a site on which the facility is to be operated until that person has been issued . . . . a construction permit under this part .... For existing facilities, 10 C.F.R. § 50.92(a) requires issuance of a construction permit prior to approval of a license amendment if the modification would entail a "material alteration" of the facility.

As explained below, the minimal changes proposed to MURR as part of the SGE experiment do not come remotely close to the "material alteration" threshold.

Thus, the SGE experiment should not require a construction permit under either§ 50.IO(c) or§ 50.92(a).

4.2.l 10 C.F.R. § 50.lO(c) Does Not Apply in this Instance The NRC's definition of "construction" in§ 50.lO(a) (which applies to§ 50.IO(c)), is broad and if read verbatim, could include nearly any physical alteration involving SSCs related to radiological health and safety. However, in practice, the NRC has not adopted this interpretation.

Further, it appears that the § 50.IO(c) requirement for a construction permit applies only to initial construction of a facility.

By its own terms, the regulation prohibits "begin[ning]" construction, but could be argued to be inapplicable after initial facility construction is complete.

NRC-regulated power reactor licensees frequently engage in physical facility modifications that technically would fall within the definition of "construction," but based on research performed by MURR,* traditionally have not been evaluated within the scope of § 50.IO(c) -we suggest, because the modification is not the "begin[ning]" of construction for the facility.

Rather, modifications to existing facilities are covered under 10 C.F.R. § 50.92(a), in the license amendment context. Construction of MURR "was completed in substantial conformity with the Construction Permit No. CPRR-68, issued on November 21, 1961." 11 Essentially, the only physical change to the existing facility necessary to conduct the "ex-pool" activities is installation of the hot cells. The hot cells will share ce1tain common systems with the existing reactor utilization facility, including exhaust ventilation, compressed air and electrical power. Accordingly, assuming the Commission approves an order authorizing a combined production and utilization facility license, part of the "production facility" apparatus already will exist. In other words, construction previously "began" under the 1961 Construction Permit, albeit via operation of the assumed Commission order approving the existing 11 Renewed Facility Operating License No. R-103 1.B. 8of11 apparatus for a dual purpose. Given the pre-existence of a portion of the production facility, it would be inconsistent to now state that a new construction would "begin" with the further modifications of the original structure, i.e., installation of the hot cells. Accordingly, MURR suggests that the language of § 50.lO(c) does not apply to its license modification requests for inclusion ofMo-99 activities.

The discussions herein are consistent with the NRC's statutory mandate for minimal regulation of research reactors.

The Atomic Energy Act of 1954, as amended (AEA), accords special status to research facilities such as MURR. Specifically, Congress "directed" the Commission "to impose only such minimum amount of regulation" upon such facilities as is necessary to promote the common defense and security, protect public health and safety, and permit widespread and diverse research and development.

12 Also note that the NRC has concluded in Part 52 space that its review of proposed facility alterations concurrent with the application is acceptable.

In that context, the Commission has noted that, There is no safety or regulatory benefit in req,uiring the licensee to concurrently submit an application for a new Construction Permit in addition to a license amendment, inasmuch as NRC review of the alteration is assured. 13 . Undoubtedly, NRC review of the proposed alteration for the SGE experiment "is assured" as part of the f01ihcoming "ex-pool" license amendment application.

Accordingly, MURR concludes that a broad interpretation of § 50.lO(c) -one that would require the additional, cumulative step of obtaining a construction permit -would impose additional burdens of time and expense upon MURR (i.e., discourage research and development activities) without any safety or security benefit, contrary to the statutory commandment of AEA § 104.c.

The Proposed Hot Cell Installation is Not a "Material Alteration" The act of installing hot cells is not the type of"material alteration" contemplated in 10 C.F.R. § 50.92(a) for existing facilities.

As explained in the "in-pool" discussion, above, MURR has identified only two circumstances where the NRC has required issuance of a construction permit for an amendment of an operating license for a non-power reactor facility.

In one instance, the alteration included the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement with those of a different design; and in the other, replacing the existing heavy-water moderated, cooled and reflected core with a light-water moderated, cooled, heavy-water reflected core. Both modifications rendered major portions of the safety analyses inapplicable.

Neither circumstance is remotely analogous to the mere installation of hot cells in the existing reactor building.

Hot cells are routinely installed in MURR for various experiments.

MURR has installed numerous hot cells throughout the life of the facility.

Notably, the NRC has never required a construction permit for such installations.

In fact, as part of the recent license renewal review, "the NRC staff observed the installation and use of additional hot cells and fume hoods," finding such installations were evaluated in 12 Atomic Energy Act of 1954, as amended,§ 104.c (42 U.S.C. § 2134(c)).

13 Licenses, Certifications, and Approvals for Nuclear Power Plants; Final Rule, 72 Fed. Reg. 49,352, 49,408 (Aug. 28, 2007). 9of11 accordance with 10 C.F.R. § 50.59 and adequately described in the MURR SAR.14 Even when Technical Specification changes are necessary for experiments conducted in hot cells, the NRC has not required a construction permit as a condition of the license amendment.

15 Accordingly, the SGE experiment hot cell installation will not remotely affect a "material alteration" of the facility, as contemplated in 10 C.F.R. § 50.92(a), and a construction permit will not be required.

5. Summary In summary, MURR concludes that a combined production and utilization facility license is consistent ,with NRC policy. MURR also concludes that a construction permit is not required for the SGE experiment-related "ex-pool" activities under either 10 C.F.R. § 50.lO(c) or 10 C.F.R. § 50.92(a), because MURR is an existing facility, and consistent with previous NRC interpretations of its regulations, installation of hot cells would not be a "material alteration" of MURR, respectively.

If there are any questions regarding this response, please contact me at (573) 882-5118 or MeffertB@missouri.edu.

I declare under penalty of perjury that the foregoing is true and correct. Sincerely, Reactor Manager ENDORSEMENT:

Reviewed and Approved State of M;SSQtJln

/} /#'

tb(__V('

a:d swomto before me this o\;1 A. Butler, P.E. Z:

  • 1J;p5 D1rector Notary {;if My Commiulon Expires: Mardi 26, 2019 xc: Reactor Advisory Committee JACQUEUNELMATYAS .

MyCcmmissionExplres Reactor Safety Subcommittee i :;:'

March 26, 2019 Isotope Use Subcommittee HowanlCounty Dr. Mark Mcintosh, Vice Chancellor for Research, Graduate Studies Mr. Alexander Adams, U.S. Nuclear Regulatory Commission Mr. Geoffrey Wertz, U.S. Nuclear Regulatory Commission Mr. Johnny Eads, U.S. Nuclear Regulatory Commission Attachments:

1. Attachment 1 -License Amendment Request to Implement Selective Gas Extraction (SGE) Target Experimental Facility (TEF) at the University of Missouri Research Reactor 2. Attachment 2 -Revised and New Technical Specifications
3. Attachment 3 -Codes and Standards
4. Attachment 4-GA Design Report No. 30441R00021, "Target Assembly Thermal Analysis" 14 See, e.g., Safety Evaluation Report, Renewal of the Facility Operating License for the University of Missouri-Columbia Research Reactor at 10-4 to 10-7 (Jan. 2017). 15 See, e.g., Amendment No. 37 to Amended Facility License No. R-103 (Mar. 11, 2016) (ML16032A424).

10of11

5. Attachment 5 -GA Design Report No. 30441R00022, "Source Term Analysis Design Calculation Report" 6. Attachment 6 -GA Design Report No. 30441R0003 l, "Mo-99 Target Assembly Nuclear Design for Once-Through Operation" 7. Attachment 7 -GA Design Report No. 30441R00032, "RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report" 8. Attachment 8 -GA Design Report No. 30441R00017, "ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report" 11of11 c GENERAL ATOMICS 1' 1 1 " AFFIDAVIT OF CHRISTINA A/BACK I, Christlpa,,.b,,,..Back, Vice President of Nuclear Technologies and Materials, General Atomics, do and state: (2) (3) . I have been delegated the function of reviewing information described in paragraph 3 which General Atomics requests be withheld from public disclosure or publication and I am authorized to execute this affidavit on behalf of General Atomics. The affidavit is submitted under the provisions of 1 OCFR Part 2.390 to withhold documents containing confidential commercial and proprietary information (as set forth in paragraph 3 following) of General Atomics from public disclosure or publication.

General Atomics (GA) has partnered with The University of Missouri Research Reactor (MURR) and Nordion to develop the Reactor-Based Mo-99 Supply System (RB-MSS) Project using Its Selective Gas Extraction process. The information sought to be withheld is related to the analysis, design, development and licensing of General Atomics' technology for deployment at MURR. This information is contained In the following documents submitted as supporting information for MURR's "License Amendment Request to Implement Selective Gas Extraction Target Experimental Facility at the University of Missouri Research Reactor. n License Amendment Request TITLE Attachment No. License* Amendment Request to Implement Selective Gas 1 Extraction (SGE) Target Experiment Facility (TEF) at the University of Missouri Research Reactor 4 GA Report 30441 R00021: "Target Assembly Thermal Analysis" 5 GA Report 30441R00022: "Source Term Analysis Design Calculation Report" 6 GA Report 30441 R00031: "Mo-99 Target Assembly Nuclear Design for Once-Through Operation" GA Report 30441 R00032: "RELAP Accident Analysis and 7 FRAPTRAN Target Rod Transient Analysis Design Calculation Report" ' GA Report 30441R00017: "ANSYS Target Cartridge, 8 Housing Structural Analysis Design Calculation Report" 3550 GENERAL ATOMICS COURT, SAN DIEGO, CA 92121-1122 PO BOX 85608, SAN DIEGO, CA 92186-5608 (858) 455-3000 I (4) (5) I l_ / ; ( \ Nonproprietary redacted versions of these documents are being simultlaneously provided where the proprietary information on each page has been redacted.

  • --, ' In making this application for withholding of proprietary information of it is the owner, General Atomics relies upon the exemption from disclosure set Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade 18 USC Sec.1905, and NRC regulations 10CFR Part9.17(a)(4), a2.390(a)(4), and -, 10CFR2.390(b) for "trade secrets and commercial or financial information obtained from, a person and privileged or confidential." The material for which exemption from \ disclosure is hereby sought is all "confidential commercial information," and/or also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983). Pursuant to Paragraph (4) above, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Atomics' competitors without license from General Atomics constitutes a competitive economic advantage over other companies;
b. Information which, if used by a competitor, would reduce his or her expenditure of resources or improve his or her competitive position in the design, manufacture, shipment, installation, assurance or quality, or licensing of a similar product. c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
d. The Information sought to be withheld is being submitted to the NRC in confidence.

The information is of a sort customarily held in confidence by General Atomics, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by General Atomics. No public disclosure has been made, and it is not available In public sources. e. The information classified as proprietary was developed and compiled by General Atomics at a significant cost to .General Atomics. This information is classified as proprietary because it contains detailed data and analytical results not available elsewhere.

This information would provide other parties, including competitors, with information from General Atomics technical database and the results of evaluations performed by General Atomics. Release of this information would improve a competitor's position without the competitor having to expend similar resources for the development of the database.

A significant effort has been expended by General Atomics to develop this information.

1' f. The research, development, engineering, and analytical costs associated with General Atomics' unique selective gas extraction system and process comprise a substantial investment of time and resources by General Atomics, whose precise value is difficult to quantify, but clearly is substantial arid beyond the original development cost. Public disclosure of the information sought to be withheld is likely to cause substantial harm to General Atomics' competitive position.

GENERAL ATOMICS Christina A. Back Vice President, Nuclear Technologies and Materials A notary public or other officer completing this certificate verifies only the identity of the individual who signed the document to which this certificate is attached, and not the truthfulness, accuracy, or validity of that document.

State of California County of San Diego On ma'{ I , 2-o I '=!: before me, Karen R. Schultz, Notary Public, personally appeared Christina A. Back, whcf proved to me on the basis of satisfactory evidence to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument.

I certify under PENAL TY OF PERJURY under the laws of the State of California that the foregoing paragraph is true and correct. WITNESS my hand and official seal.

ATTACHMENT 1 LICENSE AMENDMENT REQUEST TO IMPLEMENT SELECTIVE GAS EXTRACTION (SGE) TARGET EXPERIMENTAL*FACILITY (TEF) AT THE UNIVERSITY OF MISSOURI RESEARCH REACTOR l ATTACHMENT 1 TABLE OF CONTENTS TABLE OF CONTENTS ..............................................................................................................................

2 LIST OF FIGURES ......................................................................................................................................

3 LIST OF TABLES ........................................................................................................................................

7 1. Selective Gas Extraction Target Experimental Facility Overview ....................................................

10 I. I Proposed Experiment Description

..........................................................................................................

10 1.2 Normal Operation

..................................................................................................................................

12 1.3 Design for Loss ofForced Cooling ........................................................................................................

13 1.4 Target Experimental Facility Safety Design Approach .........................................................................

13 1.5 DesignBases

..........................................................................................................................................

15 1.6 Target Assembly Maximum Power .......................................................................................................

16 1. 7 Target Rod Peak Linear Power ..............................................................................................................

16 1.8 Codes and Standards for Design, Fabrication and Operations

...............................................................

17 2. Detailed Selective Gas Extraction Target Assembly Description and Structural Analysis ........ : ..... 17 2.1 Target Assembly ...........................................................

........................................................................

17 2.2 Structural Analysis of Target Assembly ................................................................................................

28 3. Target Cooling System ......................................................................................................................

36 3.1 Functions

..............................................................................................................................................

.36 3.2 Target Cooling System Principal Design Parameters

............................................................................

37 3.3 Target Cooling System Design ..............................................................................................................

38 3.4 Target Cooling System Performance

....................................................................................................

.45 4. Instrumentation and Control System .................................................................................................

52 4.1 Summary Description

............................................................................................................................

52 4.2 Target Cooling System Control System Description

.............................................................................

54 4.3 Target Cooling System Protection System Description

.........................................................................

55 5. Target Assembly Nuclear Design Analysis ......................................................................................

65 5.1 Analytical Methods ................................................................................................................................

65 5.2 Target Assembly Physics Model ...........................................................................................................

65 5.3 Target Assembly Flux and Power ..........................................................................................................

76 6. Target Assembly Thermal Hydraulic Design Analysis .............. , ....................................................

103 6.1 Thermal-Hydraulic Design Basis .........................................................................................................

104 Page 2of190 ATTACHMENT 1 7. In-Pool Target Transfer System ......................................................................................................

124 7. I Cartridge Loading/Unloading Station Design ......................................................................................

I24 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing ...............................

I27 8. Radiological Protection Evaluation for the SGE Target Experimental Facility Operations

...........

128 8.I Airborne Sources .................................................................................................................................

I29 8.2 Liquid Sources ...................................

.................................................................................................

I29 8.3 Solid Sources .......................................................................................................................................

I3 I 8.4 Radioactive Waste Management Program ...........................................................................................

I32 9. Conduct of Operations

....................................................................................................................

134 9. I Procedures

...........................................................................................................................................

I34 l 0. Target Experimental Facility Accident Analyses ............................................................................

139 I 0.1 Target Experimental Facility Maximum Hypothetical Accident..

.......................................................

139 I0.2 Insertion ofExcess Reactivity

.............................................................................................................

I55 I0.3 Control Blade Withdrawal

...................................................................................................................

160 I0.4 Loss of Target Coolant ........................................................................................................................

I63 10.5 Pipe Break Locations Out of the Reactor Pool ....................................................................................

I64 IO .6 Pipe Break Locations in the Reactor Pool ...........................................................................................

I 70 I0.7 Loss of Target Flow .............................................................................................................................

175 I 0.8 Mishandling of Target Cartridge or Target Rods .................................................................................

177 I0.9 Loss of Primary Coolant Flow .............................................................................................................

I 77 10.10 Loss of Primary Coolant ......................................................................................................................

178 IO.I I Loss of Pool Coolant ...........................................................................................................................

I78 IO.I2 Loss ofOffsite Electrical Power ..........................................................................................................

I78 11. Technical Specifications

.................................................................................................................

178 12. Proposed Confirmatory Testing ......................................................................................................

178 I2. I Summary Description of Planned Tests ...............................................................................................

I 79 13. References

.......................................................................................................................................

188 LIST OF FIGURES Figure 1 Selective Gas Extraction Process Scope, Functional Relationships and Interfaces

.......................

11 Figure 2 Layout of SGE Experimental Facility in the MURR Graphite Reflector Region and Containment.

...............................................................................................................................

11 Page 3of190 ATTACHMENT 1 Figure 3 Location of Target Assemblies in MURR Graphite Reflector Positions No. SA and No. SB ....... 12 Figure 4 Target Assembly (front and back views) .......................................................................................

18 Figure S Illustration of Target Housing Elevation and Plan Views .............................................................

19 Figure 6 Cartridge Configuration and Target System Section View ...........................................................

20 Figure 7 Cartridge Upper and Lower Sections .............................................................................................

21 Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum ........ 22 Figure 9 Target Rod Arrangement

...............................................................................................................

23 Figure 10 ..................................................................................................

24 Figure 11 Inlet and Outlet Plenums with Method of Attachment (water flow is shown in blue) ................

2S Figure 12 Upper Target Cartridge Arrangement (water flow is shown in blue) ..........................................

26 Figure 13 Target Pellet Geometry ................................................................................................................

27 Figure 14 Allowable Strength vs Temperature for Al 6061 T6 and SS316L. ...............................................

29 Figure 15 Pellet and Cladding Details ..........................................................................................................

30 Figure 16 Allowable Stress for Irradiated Zircaloy-4 Based on 2/3rd Yield [Geelhood 2008] ...................

32 Figure 17 Fatigue Chart for Zircaloy-4 at 3SO °C (un-irradiated and irradiated)

.........................................

32 Figure 18 Deflection Due to Target Rod Bowing for Worst-Case Front-to-Back Power Skew ..................

3S Figure 19 Process Flow Diagram of Target Cooling System .......................................................................

39 Figure 20 Elevation View of Target Cooling System ..................................................................................

40 Figure 21 Target Cooling System Superimposed onto MURR Reactor Pool and Biological Shield ...........

41 Figure 22 Schematic for the Water Cooling Module ...................................................................................

42 Figure 23 Target Cooling System ................................................................................................................

43 Figure 24 State Points within the Target Cooling System ...........................................................................

4S Figure 2S Target Cooling System P&ID ......................................................................................................

46 Figure 26 Location and Type of Supports on Above Pool Piping ...............................................................

47 Figure 27 Location and Type of Support on In-Pool Piping ........................................................................

48 Figure 28 Secondary Cooling for Target Cooling System ...........................................................................

SO Figure 29 Electric Power Supply to TCS Pumps and I&C UPS ..................................................................

S2 Figure 30 Target Cooling System Control System Architecture

..................................................................

S4 Figure 31 Target Cooling System Control Panel .........................................................................................

SS Figure 32 Target Cooling System Protection System Relay Inputs to the MURR Reactor Safety System ..............................

.........................................................................................................

60 Figure 33 TDHRVs -TCS Pump Interlock Circuit..

...................................................................................

62 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane

...................................................................................................................................

66 Page 4of190 ATTACHMENT 1 Figure 3S MCNP6 Model of Target Assembly at Axial Mid-Plane

............................................................

66 Figure 36 Target Rod Numbering for Two Target Assemblies (baseline model) ........................................

67 Figure 37 Model of the Target Cartridge SS316L Neutron Absorber ..........................................................

67 Figure 38 MURR Control Blade Travel between January 13, 2014 and September lS, 201S ....................

69 Figure 39 Target Rod 6 Axial Neutron Flux Distribution in Position No. SA .............................................

77 Figure 40 Target Rod 17 Axial Neutron Flux Distribution in Position No. SB ...........................................

77 Figure 41 Target Assembly Azimuthal Neutron Flux Distribution

..............................................................

78 Figure 42 Power Envelope of the Base Target Loading for the Extreme Bumup Core Case ......................

79 Figure 43 Power Envelope of the Base Target Loading for the Maximum Bumup Core Case ...................

79 Figure 44 Target Assembly Pellet Linear Power Distribution for Base Target Loading Case ....................

80 Figure 4S Variation of Target Power vs. Control Blade Position ................................................................

83 Figure 46 Variation of Peak Linear Power vs. Control Blade Position ........................................................

83 Figure 4 7 Variation of Driver Fuel Element Peaking Factor vs. Control Blade Position ............................

84 Figure 48 Target Rod 6 Axial Neutron Flux for Staggered Loading Case (fresh assembly in position No. SA) .......................................................................................................................................

89 Figure 49 Target Rod 17 Axial Neutron Flux for Staggered Loading Case (fresh assembly in position No. SA) .......................................................................................................................................

90 Figure SO Target Assembly Azimuthal Neutron Flux for Staggered Loading Case (fresh assembly in position No. SA) .........................................................................................................................

90 Figure Sl Target Rod 6 Axial Neutron Flux for Staggered Loading Case (fresh assembly in position No. SB) .......................................................................................................................................

91 Figure S2 Target Rod 17 Axial Neutron Flux for Staggered Loading Case (fresh assembly in position No. SB) .......................................................................................................................................

91 Figure S3 Target Assembly Azimuthal Neutron Flux for Staggered Loading Case (fresh assembly in position No. SB) .........................................................................................................................

92 Figure S4 Power Envelope of the Staggered Loading for the Extreme Bumup Core Case (fresh assembly in position No. SA) .....................................................................................................

94 Figure SS Power Envelope of the Staggered Loading for the Maximum Bumup Core Case (fresh assembly in position No. SA) .....................................................................................................

94 Figure S6 Target Assembly Pellet Linear Power Distribution for the Staggered Loading Case (fresh assembly in position No. SA) .....................................................................................................

9S Figure S7 Power Envelope of the Staggered Loading for the Extreme Bumup Core Case (fresh assembly in position No. SB) ......................................................................................................

96 Figure S8 Power Envelope of the Staggered Loading for the Maximum Bumup Core Case (fresh assembly in position No. SB) ......................................................................................................

96 Figure S9 Target Assembly Pellet Linear Power Distribution for the Staggered Loading Case (fresh assembly in position No. SB) ......................................................................................................

97 Figure 60 Target Rod 6 Axial Neutron Flux for Partial Loading Case ........................................................

98 Page S of 190 ATTACHMENT 1 Figure 61 Target Rod 17 Axial Neutron Flux for Partial Loading Case ......................................................

99 Figure 62 Target Assembly Azimuthal Neutron Flux for Partial Loading Case ..........................................

99 Figure 63 Power Envelope for Partial Loading for the Extreme Burnup Core Case .................................

100 Figure 64 Power Envelope of the Partial Loading for the Maximum Burn up Core Case ..........................

100 Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case ...........................

101 Figure 66 Change of 2-Target Assembly Power for 2-Week Operation

....................................................

102 Figure 67 Vapor Fraction at Cladding Wall for Worst-Case Conditions in FLUENT RPI Wall Boiling Model ...........................................................................................................................

107 Figure 68 Heat Flux and CHFR as a Function of Axial Location for Peak Power Target Rods ................

108 Figure 69 U0 2 Thermal Conductivity and Thermal Expansion Coefficient at 95% Theoretical Density ...................................................................

..................................................................

111 Figure 70 Zircaloy-4 Thermal Conductivity and Radial Thermal Expansion Coefficient..

.......................

111 Figure 71 Helium Thermal Conductivity in Small Gaps ...........................................................................

112 Figure 72 Target Rod Loading/Unloading/Storage Location .....................................................................

124 Figure 73 Representation of the Cartridge Loading/Unloading Station .....................................................

125 Figure 74 Target Rod Remote Handling Tool.. ..........................................................................................

127 Figure 75 Reactor Power, Fuel and Cladding Temperatures vs. Time for a Positive Reactivity Step Insertion of 0.006 .6.k/k .............................................................................................................

156 Figure 76 Cladding Strains at Peak Pellet Location During a Positive 0.006 .6.k/k Reactivity Insertion

... 158 Figure 77 Peak Target Pellet Temperature During a Positive 0.006 .6.k/k Reactivity Insertion

.................

158 Figure 78 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a Positive 0.006 .6.k/k Reactivity Insertion

..................................................................................

159 Figure 79 Power Transient During a 0.0003 .6.k/k per Second Reactivity Insertion

..................................

160 Figure 80 Cladding Strains at Peak Pellet Location During a 0.0003 .6.k/k per Second Reactivity Insertion

....................................................................................................................................

161 Figure 81 Peak Target Pellet Temperature During a 0.0003 .6.k/k per Second Reactivity Insertion

..........

162 Figure 82 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a 0.0003 .6.k/k per Second Reactivity Insertion

...........................................................................

163 Figure 83 Pipe Break Locations Out of the Reactor Pool ..........................................................................

164 Figure 84 Mass Flow Transient During a LOCA in Air Without Decay Heat Removal Valves Opening ....................................................................................................................................

166 Figure 85 Target Power During a LOCA in Air Without Decay Heat Removal Valves Opening .............

167 Figure 86 Coolant Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening ....................................................................................................................................

168 Figure 87 Maximum Cladding ID Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening .........................................................................................................

168 Figure 88 Mass Flow Transient During a LOCA in Air With Decay Heat Removal Valves Opening ...... 169 Page 6of190 ATTACHMENT 1 Figure 89 Maximum Cladding ID Temperatures During a LOCA in Air With Decay Heat Removal Valves Opening ........................................................................................................................

170 Figure 90 Pipe Break Location in the Reactor Pool ...................................................................................

171 Figure 91 Mass Flow Transient During a LOCA in the Reactor Pool .......................................................

172 Figure 92 Target Power During a LOCA in the Reactor Pool ...................................................................

173 Figure 93 Maximum Cladding ID and Coolant Temperatures During a LOCA in the Reactor Pool ........ 174 Figure 94 Cladding OD Temperature Profile in Target Rod 17 During a LOCA in the Reactor Pool ...... 174 Figure 95 Cladding Fractional Strains in Rod 17 during a LOCA in the Reactor Pool .............................

175 Figure 96 Mass Flow Transient During Loss of Pump Flow .....................................................................

176 Figure 97 Maximum Cladding ID and Coolant Temperatures During Loss of Pump Flow ......................

177 Figure 98 Conceptual Design of Test Capsule ...........................................................................................

180 Figure 99 Visualization of Potential Pellet Deformation over Course of Irradiation (not to scale) ...........

181 Figure 100 Conceptual Schematic of CHF Test Flow Section ...................................................................

184 Figure 101 Uninstrumented Critical Heat Flux Test Section in Low Pressure Flow Loop Rig ..................

185 Figure 102 GA Test Setup Conceptual Arrangement

.................................................................................

186 LIST OF TABLES Table 1 Temperature Limits for Target Assembly Design ...........................................................................

14 Table 2 Design Calculation Reports .............................................................................................................

15 Table 3 Target and Filler Rod Dimensions (cold) ........................................................................................

23 Table 4 FRAPCON Results Summary .........................................................................................................

31 Table 5 Maximum Irradiation Damage to Target Cartridge Materials

........................................................

33 Table 6 Target Cooling System Design Parameters for Two Targets .........................................................

38 Table 7 Major Component Specifications

....................................................................................................

44 Table 8 Results of Rigid Structure Seismic Analysis ...................................................................................

49 Table 9 Definition of MURR Driver Core Bumup States (MWD) ..............................................................

68 Table 10 Critical Control Blade Positions

....................................................................................................

70 Table 11 Expected Control Blade Travelling Range for Target-Loaded Core ............................................

70 Table 12 Keff and Reactivity Insertion Values for Single Target Assembly ................................................

71 Table 13 MURR Core Maximum Power Peaking Due to Target Assembly Loading .................................

72 Table 14 Reactivity Coefficients of the Reactor Core with Two Fresh Target Assemblies

........................

73 Table 15 Reactivity Control Device Worth with Two Fresh Target Assemblies

.........................................

74 Table 16 Kinetic Parameters of the Reactor Core with and without Target Assemblies

.............................

75 Page 7of190 ATTACHMENT 1 Table 17 Component Radiation Heating Due to Target Assembly Loading for Maximum Burnup Core ............................................................................................................................................

76 Table 18 Calculated Target Power Level and Linear Power ........................................................................

78 Table 19 Target Assembly Power and Core Peaking Factors ......................................................................

82 Table 20 Effect of Regulating Blade on Pellet Peak Linear Power and Target Assembly Power ...............

85 Table 21 Material Impurities for Uncertainty Analysis ...............................................................................

85 Table 22 Uncertainties in Core Eigenvalue (Bias) .......................................................................................

87 Table 23 Uncertainties in Peak Linear Power (Bias) ...................................................................................

88 Table 24 Uncertainties in Target Power (Bias) ............................................................................................

88 Table 25 Calculated Target Power Level and Pellet Linear Power for the Staggered Loading ...................

93 Table 26 Calculated Target Power Level and Pellet Linear Power ..............................................................

98 Table 27 Material Content and Burnup of the Base Target Loading for Average Burnup Core State ...... 101 Table 28 Variation of Coolant Inlet Temperature with Active Target Rod Count.. ...................................

110 Table 29 Predicted Thermal Performance for 11 Active Target Rods, 10 MWt Reactor Power, 100% Flow ..........................................................................................................................................

113 Table 30 Predicted Thermal Performance for 11 Active Target Rods, 11.5 MWt Reactor Power, 85% Flow ..........................................................................................................................................

115 Table 31 Predicted Thermal Performance for 3 Active Target Rods, 10 MWt Reactor Power, 100% Flow ..........................................................................................................................................

117 Table 32 Predicted Thermal Performance for 3 Active Target Rods, 11.5 MWt Reactor Power, 85% Flow ..........................................................................................................................................

119 Table 33 Predicted Thermal Performance for 3 Active Target Rods, Worst-Case Operations

..................

121 Table 34 Predominant Radionuclides in the MURR Pool Coolant and Their Measured Concentrations at 10 MW .........................................................................................................

130 Table 35 Representative Radioactive Sources at MURR ...........................................................................

132 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI ...................................................................................................................................

134 Table 37 Standard Operating Procedures

...................................................................................................

135 Table 38 Activity of Volatile Fission Products in the Gap Gas of the Hottest Target Rod .......................

141 Table 39 Iodine and Noble Gas Activities Released to MURR Reactor Pool ...........................................

142 Table 40 Iodine Concentrations in Pool Water ..........................................................................................

142 Table 41 Average Iodine Concentrations in the Containment Building Air During ..................................

143 Table 42 Average Noble Gas Concentrations in the Containment Building Air during the 5 Minute Evacuation Period .....................................................................................................................

144 Table 43 Derived Air Concentration Values and 5-Minute Exposures for Iodine .....................................

146 Table 44 Derived Air Concentration Values and 5 Minute Exposures

-Noble Gases ..............................

147 Page 8of190 ATTACHMENT 1 Table 45 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building ...................

147 Table 46 Average Containment Building Leakage Rate ............................................................................

149 Table 4 7 Average Iodine Concentrations in Air Exiting the Exhaust Stack ..............................................

150 Table 48 Noble Gas Concentrations in Air Exiting the Exhaust Stack ......................................................

151 Table 49 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Iodine ..................................................................

153 Table 50 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases ........................................................

153 Table 51 Dose from Iodines and Noble Gases in the Unrestricted Area ....................................................

154 Table 52 Radiation Shine through the Containment Building ...................................................................

154 Table 53 Capsule Dimensions and Attributes

............................................................................................

179 Table 54 Irradiation Test Schedule ............................................................................................................

181 Table 55 Critical Heat Flux Test Schedule ................................................................................................

184 Table 56 Testing Specifications

.................................................................................................................

187 Table 57 System Integration Test Schedule ...............................................................................................

187 Page 9of190 ATTACHMENT 1 1. Selective Gas Extraction Target Experimental Facility Overview 1.1 Proposed Experiment Description The Selective Gas Extraction (SGE) experiment employs a first-of-a-kind concept for radioisotope production.

It is a reactor-driven, low-enriched uranium (LEU)-based system that selectively removes specific isotopes of interest, viz., molybdenum-99 (Mo-99), in gaseous fonn, that are produced from fission during irradiation of Zircaloy-4 clad target rods containing LEU. The LEU is in the fonn of uranium dioxide (U0 2) pellets that are nominally enriched to 19.75% in the isotope uranium-235 (U-235). The target rods are contained in Al6061 cartridges that ensure unifonn cooling water flow around the target rods and will be located in the graphite reflector region of the University of Missouri Research Reactor (MURR). The SGE Target Experimental Facility (TEF) will be operated by MURR staff in concert with MURR's routine reactor operations.

During SGE experiment operation, one (1) or two (2) target cartridges, holding 11 target rods each, are placed in pennanently installed support assemblies in the reactor graphite reflector region. Fission product isotopes, including Mo-99, are generated during target irradiation.

At the end of irradiation (0.5 to 3 weeks), and following a short period of cooling (to reduce decay heat), the target rods are removed from their respective cartridge and transferred to a loading/unlo'ading and storage location within the reactor pool. Subsequent target rod transfer to a bank of hot cells inside the reactor containment building will be addressed in the Part 2 license amendment submission.

Figure 1 shows a functional block diagram of the SGE experiment overview for this Part 1 License Amendment submission as well as for the Part 2 submission.

The in-pool (IP) portion of SGE process, the TEF, is the focus of this reactor License Amendment request. A description of these IP systems and associated analyses are presented in this document as follows: 1. Target System including the Target Rods, Cartridge, and Housing (1 or 2 Units) 2. Target Cooling System (TCS) 3. Cartridge Loading/Unloading and Storage Station 4. Instrumentation and Control (I&C) Systems 5. Analyses of Potential Accidents and their Impact on the MURR Facility The remaining SGE process (ex-pool) systems and associated safety analyses will be described in the Part 2 submission.

Figure 2 and Figure 3 show the layout of the TEF irradiation systems within the MURR reactor pool and containment building.

Page 10of190 --1 ATTACHMENT 1 C.ORection System IDtv Chemi s try) Hole.* Target Cooling T arget Coolin.a Selective Gas Mo-99Trap He*ting Remowl SY<tern ,_ tn s trumentaUon

& Control I"& Transport LO*A System System ,

  • Extraction Syst em System :t.a.a IU**I (C W.) J.O< .

s,.rtuo l Tnnttu C. ....... & Containment cnc.,MllWon Iii Structure Ftt<h Target Rods -Sto.-qe$y tt ORm (U,.,IJ !Cl.Cl Tr*nsf<< cask ... lt.oad In pool) A A A A A A A A -------_, -Waste Target SY<tem Shipping*

Interim W a ste Storage & Target Assembly Container Shippln1 System Reactor Torget Target h11et Cora loadinc & .. Housing 1 " C.rtridse

  • As*ernbl les Assemblies TypeB Station Sh l pplnc Cask ts111IDll

'5zlszc '2iali llS!I l!nilllllil!!!l i Final PUriflcalion c.oo11rcwat*

Cl.A T*1et Tr a nsfer Cell/Clpouie Weldtna Fr.,.h fuel Packinc& ,..

Mo-99 CI< De-Oo ddln8f tannl ng and Coolclown ToOl*lo1M1$

..

  • Transport* SpenlUO z System Ci.dding Waste Figure 1 Selective Gas Extraction Process Scope, Functional Relationships and Interfaces Figure 2 CARTRIDGE STATIOl\" CARTRIDGE WITH TARGET RODS TARGET GRAPHITE [5 a, b d , e , f] La y out of SGE Experimental Facility in the MURR Graphite Reflector Region and Containment Pa g e 11of190

--ATTACHMENT 1 Figure 3 Location of Target Assemblies in MURR Graphite Reflector Positions No. SA and No. SB 1.2 Normal Operation During nonnal operation , one (1) or two (2) target cartridge(s) are loaded with fresh target rods in the loading station shown in Figure 2. The cartridge(s) is then inserted into the target housing(s) located in graphite reflector positions No. SA and/or No. SB (Figure 3) whi l e the reactor is shut down. When the reactor is restarted , neutrons from the reactor will fission the LEU in the target rods generating fission products , including the radioisotope Mo-99. The target assemb l ies (TAs) are cooled by forced flow from the Target Cooling System (TCS) both when the reactor is operating and to remove decay heat during routine reactor shutdowns.

The heat removed from the T As is rejected via the TCS to the existing facility secondary coolant system without significantly affecting the reactor bulk pool temperature.

The nominal operation cycle will load and remove 11 target rods per week for processing. At the end of the irradiation cycle , the reactor is shut down and the TA is allowed to cool by forced flow [Sa , d , e] from the TCS . At this time the l evel of stored and decay heat in the cartridge is [Sa , d , e] sufficiently l ow that direct conduction to the reactor pool water is sufficient to maintain the target rods in a safe condition.

The cartridge(s) is t h en moved to t h e loading/unloading station and the target rods are transferred to in-poo l storage or a sh i elded transfer cask which has been moved to an underwater position near the loading/unloading station for safe transfer.

Meanwhile a previously prepared fresh cartridge(s) is loaded into the TA positions No. SA or/and No. SB. Page 12of190 ATTACHMENT 1 1.3 Design for Loss of Forced Cooling L oss of forced cooling (LOFC) to the TA could occur by a power failure , pump failure or coolant pipe b reak. If a LOFC event occurs , a low flow signal (8S% of normal flow) or a high flow signal (11 S% of n ormal flow) is sent from the SGE TEF to the MURR reactor safety system to scram th e reactor. Decay h eat removal from the target rods is provided by pool water flowing through automatically-actuated decay h eat removal valves into the target cartridge.

1.4 Target

Experimental Facility Safety Design Approach The safety objectives for the SGE TEF are (1) minimization of exposure of occupational workers to ra diation from normal system operation, (2) prevention of release of radionuclides to the MURR reactor c ontainment building and the environment from operation of the system , and (3) control of target rod c ladding temperatures to prevent cladding failure leading to a fission product release to the reactor c ontainment building and potentially , to the environment.

To meet these objectives, the SGE TEF has been desi gned to the following top-level requirements to e nsure that the safety objectives are met:

  • The target rods and cartridges shall be capable of sustained operation with the reactor operating at a steady-state power level of 10.0 MW, and through anticipated operating transients where the reactor power reaches its scram power limit , without jeopardizing the integrity of barriers that would release radionuclides to the reactor pool water, target cooling water , or the reactor containment building.
  • The target cartridges together shall not increase the reactivity worth by more than 0.006 when in serte d into the No. SA or No. SB position s of the graphite reflection region.
  • The target rods shall be hermetically sealed in cladding that will se rve as a barrier against the release of radioisotopes to the reactor pool water , target coo l ing water , or the reactor containment bui l ding.
  • In case of a loss of forced flow , natural circulation of the pool water through the TA shall provide an adequate backup means of removing target decay heat without damage to the cladding.
  • Normal cooling shall ensure that the minimum critical heat flux ratio (MCHFR) using the Bernath correlation shall not fall below 2.0 for normal operation.
  • The SGE TEF shall interface with the MURR reactor safety system to initiate a reactor scram under any condition that threatens the barriers against the release of radionuclides to the reactor pool water , target cooling water, or the reactor containment building.
  • The SGE TEF Instrumentation and Contro l (l&C) System shall incorporate instrumentation to ensure that essential system variables are monitored and are within specified operating ranges at all times. Page 13of190 ATTACHMENT 1
  • The SGE TEf I&C System shall be equipped with a reliable and redundant system that monitors the operation of systems required for safety. Upon detection of conditions that indicate an anomalous state , failure or impendin g failure of a barrier against the release of radionuclides to the reactor pool or the containment building, the system shall cause a reactor scra m , or generate appropriate alarms to cause operator action to the anomalous condition.
  • The SGE TEf shall satisfy the facility limits for any releases of radioactivity to the enviromnent.

Target and cladding design temperature limits are shown in Table I. Table 1 Temperature Limits for Target Assembly Design Component Normal Operation Transients Target U0 2 pellet s 1 2400 oc (4352 °f) 2860 oc (5180 °f) Zircaloy-4 target rod cladding 2 360 o c (680 °f) 900 °C 3 (1652 °f) Aluminum target cartridge housing 200 oc (392 °f) 200 oc (392 °f) ' The maximum normal operating temperature limit for th e target rods is set so that eve n at their hottest conditions of a maximum power l evel and a minimum flow rate LOFC eve nt , the maximum p e ll et temperature will still b e b e low the U0 2 melting temperature of2860 °C. 2 Zirca lo y-4 is not an ASME code material.

Genera l Atomics uses the irradiated properties of Zirca lo y-4 contained in Reference I , which shows adequate s trength to 360 °C for normal operations. 3 This design limit on Zirca lo y-4 c l addi n g i s th e predict ed temperature for onset of rapid oxidatio n of900 °C (Refere nc e 2). The cladding temperature predicted by transient analysis for the most severe LOFC eve nt (0.5 sec pipe s h ear in the pool at the target assembly water inlet) is 650 °C , well below the onset of rapid oxidation. Further , the same a nal ysis s how s that during thi s time , retraction of the contained U0 2 pellet eliminates any compromise of the c laddin g mech anica l strength. The following design features m1mm1ze the likelihood of the target cladding exceeding the design temperature limits during irradiation and subsequent processing , and causing a release of radioactivity to th e environment:

  • The U-235 loading in the target rod is limited to ensure the TAs remain substantially subcritical under normal operation and all credible off-normal conditions.
  • The MURR reactor will scram on low or high TA cooling water flow or loss of heat si nk (secondary coolant flow) thereby limiting heat generation in the target.
  • The TA is designed to permit cooling b y natural convection using reactor pool water, in the event of a LOfC transient.
  • The heat flux from the target rods is limited by design to ensure the temperature of the Zircaloy-4 cladding does not exceed the design limits.
  • The target rod cladding serves as a barrier to the release ofradionuclides. A breach of this barrier will be detected by the reactor pool coolant system radioactivity monitoring system. Page 14of190 ATTACHMENT 1
  • In the unlikely event of a leak of radionuclides from the target rods , the reactor pool serves to reduce the amount of both Iodine and noble gas release due to plating out of Iodine , decaying of noble gases by s l owing down the time it takes the gas to rise to the pool surface, and gas so lubil ity. 1.5 Design Bases During the course of target operation the total thermal power and power distribution will depend upon the combination of several independent variables including burnup and location of the fuel elements , age and position of the control blades, age of the beryllium reflector , time in reactor operating cyc l e and time in target operat in g cycle. This leads to a near-continuum of core and target states. The target nuclear design analyses are performed for a selected set of conditions that typify the variation in target operating states. No one set of conditions can form a conservative basis for all the safety analyses , but rather the safety design basis was created from a composite of several states to give the conservative worst-case conditions fo r evaluating the safety performance of the TEF operations.

Design bases , including all assumptions for the design bases, as well as documentation of software v erification calculations that form the bases of the design and safety information presented in the License Amendment are provided in the reports listed in Table 2 below. Table 2 Design Calculation Reports REPORT NO. TITLE 3044IROOO17 ANSYS Target Cartridge , Housing Structural Analysis Design Calculation Report 3044 lROOOl 9 Target System Cooling Calculation Report 3 0441 R0002 l Target Assembly Thermal Analysis 30441 R00022 Source Term Analysis Design Calculation Report 30441R00030 Mo-99 Target Cooling System Seismic Analysis Design Calculation Report 30441 R0003 l Mo-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00032 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441R00033 Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations 30441 R00035 Cooling of the MURR Beryllium Reflector 30441R00038 Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report 30441M00043 ANSYS Thermal Model ofRB-MSS Target Rod Page 15of190 ATTACHMENT 1 1.6 Target Assembly Maximum Power For normal irradiation operation , the maximum target power detennines the coolant flow rate necessary to meet the required TA outlet temperature.

The maximum target power also establishes the safety design bases for the fission product source terms and for removal of the residual decay heat from the target rods in case normal flow is lost. The maximum target power therefore establishes the decay heat loads and source terms for safety evaluations.

The calculated maximum thermal power for two (2) T As in graphite reflector positions No. SA and No. SB is -without margin for uncertainties. This value corresponds to MURR core power at [Sa , d , e , fJ 10 MW 1 and a maximum core burnup case (676 MWd). This case produces a peak power in fuel element FS , which is adjacent to the target rods. This value occurs at Day 2 of irradiation and declines with target rod burnup. Nevertheless , this value is conservatively used for the afterheat level and fission product source terms assuming a target irradiation time of three (3) weeks at 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per week. The worst-case control blade tilting and age is assumed (0 years for blades 'A' and 'D' and 8 years for blades 'B' and 'C') and the beryllium reflector age is assumed to be 0 y ears. This worst-case value conservatively corresponds to a control blade withdrawal position at 17.32 inches (44 cm) despite the fact that the "critical position" is 19.79 inches (S0.27 cm). The estimated 2cr statistical uncertainty in the total power for two (2) TAs is -* The total [Sa , d , e , fl uncertainties of the key performance parameters were estimated as a product of stati s tical error and mean-square (RMS) of uncertainties due to fabrication density , enrichment , and target rod position. Uncertainty in impurities was conservatively not inc l uded because it is always negative. An additional heat load for the required cooling is the structural heating from the cartridge

-. The heat from the cartridge is 12.7 kW --for two (2) TAs , for a total of-. The maximum fission power used for the safety analyses is-for two (2) TAs. 1. 7 Target Rod Peak Linear Power For normal operation, the peak target rod linear power determines the combination of coolant flow velocity , temperature and pressure required to achieve adequate heat transfer margin to prevent film boiling and centerline pellet melting. For abnormal operation , the peak target l inear power is a critical parameter for assessing the performance of the target decay heat removal system, especially during the transition from normal coo l ing to backup cooling after a reactor trip. [Sa , d , e , f] [Sa , d , e , f] The calculated peak linear power is -without including uncertainties.

This value corresponds

[Sa , d , e , f] to a reactor core power at 10 MW 1 and an extreme core burnup case (6S6 MWd). This peak linear power occurs at Day 2 of irradiation and declines with target burnup. Nevertheless , this value is conservatively used for the entire irradiation period. A worst-case control blade age is assumed (0 and 8 years) and the beryllium reflector age is assumed to be 0 years. This worst-case value conservatively corresponds to a Page 16of190 ATTACHMENT 1 c ontrol blade withdrawal po s ition at 11.81 inches (30 cm) despite the fact that the " critical position" i s 20.S9 inches (S2.29 cm). The estimated 2cr uncertaint y m the peak linear power i s -based on the same statistical

[Sa , d , e, fJ c ombination of parameters as for the maximum target power. The peak linear power to be used for safety analy s es is-. 1.8 Codes and Standards for Design, Fabrication and Operations The codes and standards; design methods; software; design documentation

quality assurance and s oftware V & V are listed in Attachment 3 to the License Amendment.
2. Detailed Selective Gas Extraction Target Assembly Description and Structural Analysis This section describes each of the separate subsystems comprising the Target Experimental Facility (TEF): Target Assembly (TA), Target Cooling System (TCS) and target transfer system , including the ne utronic and thermal-hydraulic design for the SGE TEF. 2.1 Target Assembly 2.1.1 Functions The functions of the TA are as follows:
  • Produce Mo-99 by the fission ofU-23S in U0 2 with a nominal enrichment of 19.7S%, but not to exceed 19.9%.
  • Provide containment , support , and positioning of the U0 2 target material for irradiation and cooling.
  • Guide , direct , and effectively distribute coo lin g water from the TCS to the surfaces of the clad target rods containing U0 2*
  • Prevent fission products from inadvertently entering the reactor pool water and/or containment building.
  • Maintain the target pellet configuration such that the T As are always sub-critical , and ensure that any neutronic coupling between the T As and the MURR reactor core will continue to meet the reactor license Technical Specification (TS) requirements.
  • Enable detachment of the target cartridge from the target housing at the end of irradiation , for short-term transfer to in-pool storage. Page 1 7 of 190 [Sa, d , e , f]

ATTACHMENT 1 2.1.2 Mechanical Design The two T As are designed to be installed in the graphite reflector positions No. SA and No. SB. The TAs are mirror images of each other and differ only in the position of the inlet cooling water pipe. Each TA consists of a water inlet section , a target housing , a lower plenum , a cartridge , an outlet diffuser , and a cartridge locking mechanism.

The housing is held laterally in place by an indicator hole in the reflector support plate and vertically in place by the TCS inlet piping which includes a compressible link. The TA components for the No. SA TA are shown in Figure 4. The modeling and mechanical design of the TA components in 3D was performed using the commercially available SolidWorks 2016 software package. 2.1.3 Target Housing Cartridge locking mechanism (x2} (Al6061} Figure 4 Target Assembly (front and back views) E <D <'> [Sa, b d , e , f] The functions of the target housing are to direct the flow of the cooling water, provide structural support , and position the cartridge within the graphite reflector region. The vertical and plan cutaways of the target housing are shown in Figure S. The target housing is fabricated from welded Al6061 plates while the lower housing plenum is fabricated from SS316L. Cooling water is fed to the target housing by the TCS through a line at the top of the assembly. The target housing then directs the water flow down Page 1 8 of 190 ATTACHMENT 1 through the lower plenum , up the inside of the cartridge and finally out through the diffuser into the reactor pool. The target housing and the lower housing plenum are bolted together and water leakage is prevented by a metal c-seal that keeps this interface watertight.

Engaging lug for locklng cartridge locklng mechanism (Al6061) water water Inlet / (Al6061) target hous i ng (Al6061) cartridge guide niDa (Al6061) cartridge mounting plate (SST316L)

C-seal lower housing plenum (SST316L)

Indicator Stub (SST316L)

Figure 5 cartridge locking plna (SST316L)

Illustration of Target Housing Elevation and Plan Views The bottom of the housing has an indicator stub that locates the housing in the reflector support plate , which bears part of the weight of the TA and attached piping. The piping is secured by brackets within the reactor pool and part of the weight load from the piping is transferred to the lower housing plenum by the TA sides. 2.1.4 Target Cartridge The functions of the cartridge are to (1) position and support the target rods containing the LEU pellets, (2) provide a cooling passage for the target rods , (3) and (4) to mix and guide the water outlet flow. The cartridge consists of an Al6061 diffuser on the top, an Al6061 cartridge flow housing , [Sa , b d , e , f1 an Al6061 lower cartridge flange , a locking mechanism , and 11 Zircaloy-4 clad target rods. The target rods are vertically oriented in a single plane positioned in the target cartridge orthogonal to the direction of the neutron flux (Figure 6). During commissioning operations , it will be necessary to load fewer than 11 U0 2 filled target rods in a cartridge.

When fewer than 11 U0 2 filled target rods are loaded , the remaining positions are filled with Page 19of190 ATTACHMENT 1 stainless steel filler rods with the same dimensions to en s ure that the flow around all rods is uniform and at design conditions.

The orientation and position of the cartridge relative to the reactor maximizes the Mo-99 production while meeting temperature limits on the target rods. The target rods are held in position by the top and bottom cartridge flanges which allow for the coolant to flow around the target rods. The cartridge is secured in place by a Jocking mechanism located on top of the diffuser. The Jocking mechanism engages and disengages to the top of the target housing. C*rtrldge locking mechanlem (x2) Certridge flow houe ln g (Al8081) l ower certrldge ll*nge (Al8081) Figure 6 !+------9.5 i n -------+! Targe t housing waler channel Neutron flux Cartridge Configuration and Target System Section View co " [5a , b d , e , f] Figure 7 shows the cartridge upper and lower sections highlighting key components and features of the design. The cartridge design is a clamshell with two seam welds running the length of the cartridge. Five (5) pairs of SS316L pins (3/32 inches in diameter) are located a lon g the center axis of the clamshell assembly to prevent excessive stresses at the weld locations due to the higher internal pressure from cooling water flow. This simplifies the fabrication process as well as allowing more contro l over the tolerances for the fit between the water cooling channe l s and the target rods. The cartridge is first located to the target housing by a pair of guide rails that lead to a pair of locating pins. These high tolerance pins are part of the target housing lower plenum and receive and place the cartridge in its final [5a , b , f] position in relation to the reactor. The cartridge is then held in place by a pair of locking features in the upper TA locking mechanism. The features are actuated via a lever that locks and unlocks the cartridge.

Page 20 of 1 90 ATTACHMENT 1 [5a , b d, e , t] Figure 7 Cartridge Upper and Lower Sections The target rods are fixed on the top (upper endcap) and laterally supported at the bottom (lower endcap) target rods are expected to thermally grow by < 2 mm (< 0.079 in), leaving including mechanical tolerance stack-ups.

The target rods are located and held concentric to the water channels by 11 cup features that are fabricated into the lower cartridge flange and can be seen in Figure 8. This ensures even flow velocities around the target rods through the cartridge.

The water cutouts on the cup features are smaller than the pointed end cap of the target rod. This eliminates the possibility of the target rod getting stuck on one of these water bypass features when inserting them into the cartridge and ensures the operator can always find the center of the cartridge and guide the target rod into its final position. Page 21 of 190 [Sa , b d , e , t]

ATTACHMENT 1 [Sa , b d , e , f] Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum The water cutouts are designed to create near-uniform coolant flow over the target rod s as soon as the water enters the cartridge.

2.1.5 Target

Rods Figure 9 shows an individual target rod assembly , which consists of an upper end cap , cladding , a spring , 100 U0 2 target pellets and a lower end cap. The U0 2 target pellets have a nominal active length of 23.6 inches (600 mm) in the cold state. A nominal radial gap of 50 µm exists between the pellet outside diameter (OD) and the cladding inside diameter (ID). This nominal gap offers the best dimensional balance between cladding strain and thermal conductance to the cladding wall. The end caps are fabricated from Zircaloy-4 bar with integrated features designed to optimize installation and extraction from the cartridge.

Both end cap s are welded to the cladding autogenously (no weld rod) by a standard automated orbital weld head. The stainless steel spring is held captive by the upper end cap, to aid in ea s ier recovery post-irradiation.

The individual pellet/clad components and dimensions are listed in Table 3. The Zircaloy-4 (UNS R60804) cladding will be fabricated and inspected in accordance with seamless alloy tubes for nuclear reactor fuel cladding applications per American Society for Testing and Materials (ASTM) B811. The Zircaloy-4 (UNS R60804) end caps and stainless steel spring will be fabricated from bar material in accordance to ASTM B351. Page 22 of 190 ATTACHMENT 1 The filler rods have external dimensions that are identical to the target rods , except that they are fabricate d from solid stai nl ess steel (SS3 1 6L). The filler rods ensure that the required flow co nditi ons for safe cooling of the target rods are met. T h e SS3 1 6L acts as a neutron absorber to r educe power p eaking in the target rod next to the filler rod. 2 6.5 0 5 i n C l adding (Zircaloy-4) Spring (3 1 6L SST) U02 pellet (x 100) . I i-*-*-*-*-*-*-*-*-*-*-*-*-*-*-* I Cladding (Zircalo y-4) L o w er End Cap (Zircaloy-4)

L.----------------*-*-*-*-i I I .. , *-*-*-*-*--*-----*-*i_*-*-*-*-*-*-*-*-*-*-*-*-*-*-*j I -----*-*-I i-.*-----*-*-*-* I Figure 9 Target Rod Arrangement Table 3 Target and Filler Rod Dimensions (cold) Nominal Value in Inches (mm) Component Target Rod Filler Rod Active target rod length (co ld) 23.6 (600) 23.6 (600) Tota l target rod length 26.505 (673) 26.505 (673) Pellet height 0.236 (6) NIA Pellet outside diameter 0.197 (5) N I A Clad ID 0.201 (5.1) NIA Page 23 of 190 ATTACHMENT 1 Figure 10 2.1. 7 Diffuser The functions of the diffuser are to provide upper support and containment of the target rods , provide water mixing for a bulk outlet water temperature measurement , direct the cooling water exiting the targets away from MURR equipment , and to guide nitrogen-16 (N-16) flow away from the reactor pool surface. The water is mixed in the diffuser's flow mixing zone. The exit temperature measurement is used to determine the power of the TA. 2.1.8 Target Assembly Cooling Water Flow Path The TA cooling water flow path is shown in Figure 11. Cooling water enters the housing from the inlet pipe and flows into the open lower plenum turning into the lower cartridge flange. The lower cartridge flange has labyrinthine features to minimize water bypassing the target rods during normal operations.

The flow then travels along the target rods , into the diffuser and is ultimately rejected to the reactor pool. Page 24 of 190 [Sa , b d , e , t] [Sa , b d , e , f]

ATTACHMENT 1 Figure 11 Inlet and Outlet Plenums with Method of Attachment (water flow is shown in blue) 2.1.9 Upper Target Housing and Cartridge The upper target housing, cartridge, and lower sect ion of the diffuser are shown in Figure 12. The lower flange of the diffuser acts as the lid that hold s down the target rod s and keeps them secured in the cartri dge by capturing the tar get rod's upper end cap. This, along with the lower cartridge flange supporti ng the target rods , properly constrains the target rods through the installation , irradiation , and transfer to the temporary load ing/u nloading statio n in the reactor pool. The lower diffuser flange is welded to the neck portion of the diffuser , which collects the water exiting the cartridge and mi xes it before guiding it to the resistance temperature detector (RTD) for measurement of the outlet temperature , and ultimately discharges it to the pool. The mixing of the flow is significant for accurate measurement of the power bein g generated by the TA. The lower diffuser flange is also desi g ned s uch that the water cutouts allow the flow to move through the sections have the same cross sectional surface area as the rest of the conical sections of the diffuser all the way to the exit tee. This allows the diffuser to have a minimal pressure drop from the cartridge exit to the reactor pool. Page 25 of 190 (Sa , b d , e , f]

ATTACHMENT 1 Figure 12 Upper Target Cartridge Arrangement (water flow is shown in blue) When it is time to remove an irradiated cartridge from its target housing for transfer to the in-pool loading/unloading station , the cartridge will be maneuvered and handled remotely in the reactor pool by an operator positioned on top of the pool , using the tools designed for cartridge handling. The cartridge is moved from the TA to the in-pool unloading station , where the diffuser is unlatched from the cartridge to access the target rods for loading into the in-pool storage location.

The diffuser is attached to the cartridge by four spr in g flexures that engage the cartridge upper flange. The flexures are engaged and disengaged by depressing a spring loaded lever located at the top of the diffuser.

Tools have been developed to perform such a function within the reactor pool. 2.1.10 Target Pellets Each target rod contains I 00 high density U0 2 pellets , with each pellet containing l .20S grams of U0 2* The pellets are fabricated using a process simi lar to li ght water reactor (L WR) fuel with the pellet dimensions being the only major difference. Manufacturing of the target pellets , the loading and welding into the Page 26 of 19 0 [Sa , b d , e , f1 [Sa , b d , e , f1 ATTACHMENT 1 cladding shall be performed in accordance with the required fabrication , inspection and quality control procedures. The pellet geometry i s shown in Figure 13. Each pellet is nominally 5.0 mm (0.2 in) in diameter and 6.0 mm (0.24 in) high and ha s a density of greater than 10.42 glee (95% TD). The pellets and Zircaloy-4 c ladding have a design tolerance to maintain a gap between the pellet and cladding. The gap directl y influences the peak pellet temperature and pellet-cladding interaction.

Dishing on both ends reduces ratcheting during temperature cycling. Figure 13 Target Pellet Geometry 2.1.11 Design Basis for Target System Materials I Smm The considerations upon which target system material selections are ba sed are as follow s:

  • Prior operating experience for target cladding materials in a nuclear reactor irradiation environment is preferable.
  • The cladding materials must have good mechanical strength at both normal operating temperatures and at expected temperatures during transients.
  • Materials in the neutron flux must have low neutron absorption cross-sections so as not to impede the rate of Mo-99 production.
  • Selected alloys must have the properties to readily s upport fabrication into required shapes and must be readily weldable.

Page 27 of 190 ATTACHMENT 1

  • The material used for the target rod cladding must not under go undesirable che mical reactions with the target pellet material or fission products within the operating temperature range. 2.1.12 NRC-Approved Fuel Cladding Alloys The target rods have leak-tight cladding that has proven prior experience for this application, per the design requirements. Alloys that have been approved by the U.S. Nuclear Regulatory Commission (NRC) for fuel cladding in power and research reactor s in the United States *include Zircaloy , SS304 , SS316L, and Al6061. Zircaloy-4 was chosen due to its lower neutron absorption, which results in higher product yield, and proven performance in L WRs. The Zircaloy-4 (UNS R60804) cladding will be fabricated and inspected in accordance to sea mless alloy tubes for nuclear reactor fuel cladding applications per ASTM B8 l l. 2.2 Structural Analysis of Target Assembly 2.2.1 General The TA components are analyzed in accordance with ASME Section VIII Division I and Il. The unirradiated allowable stresses used in the analyses for SS3 l 6L and Al6061 -T6 were obtained from the ASME Section II part D , and are shown in Figure 14. The Zircaloy-4 cladding has been analyzed structurally for both normal and off-normal operating conditions. 2.2.2 Target Housing and Cartridge The structural analysis for the TA was performed using ANSYS 2016. The calculations confirmed that the target housing and cartridge components are properly sized according to the ASME B&PV Section VIII, Division 1 and Section 11-D , for normal and off-normal conditions.

Based on this analysis , the structural design life of the target housing is conservatively estimated at 10 years while the design life of the cartridge is conservatively estimated for one (1) year. The target housing and the cartridge, fabricated from Al6061-T6 , were analyzed for a maximum design pressure of 148.2 kPa (21.5 psi) and 102.04 kPa (14.8 psi) differential , respectively , to allow for design margins. The normal expected operating differential pre ssures (100% flow) will be 93.91 kPa (13.62 psi) and 66.40 kPa (9.63 psi) for the housing and cartridge , respectively.

However, the flow conditions for which a reactor scram is initiated ( 115% of normal flow) increases the pressures to 120.11 kPa (17.42 psi) for the housing and to 83.84 kPa (12.16 psi) for the cartridge, respectively.

For this 115% flow scenario , the calculated maximum membrane plus bending stresses of 27.07 MPa (3930 psi) for the housing and 49.07 MPa (7120 psi) for the cartridge , show that both the target housing and cartridge aluminum components are well within the allowable limit of 82 MPa (11 , 893 psi) at expected operating temperatures of -50 °C (122 °F) (Figure 14). The SS316L pins in the cartridge see the higher stresses but all are still well within allowables (115 MPa). For the 115% flow condition, the highest maximum principal stress that the SS Tl 36L housing components experience is 46.84 MPa (6.79 ksi) while the cartridge pins are subjected to a maximum membrane plus bending stress of 49.07 MPa (7. 12 ksi). At these stresses , this design ensures that both Page 28 of 190 ATTACHMENTl the housing and the cartridge maintain a Factor of Safet y (FOS) > 3 for the aluminum components and a FOS > 2 for the s tainless steel component s before the m a terial begins y ieldin g (241 MPa yield for aluminum a nd 172 MP a for SS316L p e r the ASME Code). This satisfies Sect i on C(l)(c)(3) of Regulatory Guide 2.2 , " Development of Technical Specification s for Ex periments in Research Reactors ," on mechanical stress effect s for materials of con s truction for reactor experiments , which s tates that materials of con s truction and fabrication and a s sembly techniques utilized in experiment s " ... s hould b e so s p ec ifi e d an d u se d that ass uranc e i s pro v id e d that no s tr e s s failure c an oc c ur at s tr esses tw i ce thos e anti c ipat e d in th e manipulation and condu ct of th e ex p e rim e nt or twi ce tho se w hi c h would o ccur a s a r e sult of unint e nd e d but c r e dibl e chan ges of, or w ithin , th e ex p e r i m e nt." Both the housing and the cartridge therefore maintain a FOS > 2 to the yielding allowables which provides plenty of margin to failure s tresses a s mentioned in Regulatory Guide 2.2. ii D.. Ill Ill tn QI ii ; 0 120 1 0 0 80 60 t 40 20 .............. . ........................ *****

  • 60 6 1 Al (T6} ............. ,.... .. .... . ...... ,............ . ................................................................................ . ..... .. '" O L_ -------------0 so 100 150 200 250 300 350 400 450 Tem pe rature (°C) Figure 14 Allowable Strength vs Temperature for Al 6061 T6 and SS316L 2.2.3 Pellet-Clad Interaction 500 Pellet-clad interaction for the target rods have been analyzed using FRAPCON 4.0 to ensure that the target rod performance does not exceed design limiting factors for target pellet melting temperatu r e s and cladding strain cycles. Each target rod consi s ts of I 00 , 95% dense U0 2 pellets that are encapsulated b y the Zircaloy-4 cladding. The pellets (Figure I 5) have been designed for optimum performance w ith manufacturability in mind. Dishing of both ends was added to address ratcheting effects in the cladding , while chamfers were added fo r ease of insertion and to eliminate stress concentrations on the cladding from pellet rotation. The pellet Page 29 of 190 ATTACHMENT 1 to clad gap was optimized at SO microns nominal to minimize stress/strain in cladding on one hand and peak centerline temperatures on the other. When adding manufacturing tolerances , the gap varies between -microns. Additionally , the target rod void space is filled with > 9S% He , at ;:::J [S a, b , fJ atmosphere that provides good heat transfer properties to minimize pellet centerline temperatures.

6mm (nom.) Figure 15 Pellet and Cladding Details The target cartridge irradiation time can vary between [Sa, b , fJ Fuel Pellet Cladding Dishing . In order to construct a worst-case operating scenario for pellet-clad interaction analysis, it was assumed that during a --irradiation period the reactor could see a maximum of I startups and shutdown cycles , I per week. The reactor nominal operating power is 10 MW" with a reactor shutdown occurring at the end of [Sa , d , e ,] the period. For analysis purposes it is assumed that at the beginning and end of the --period , the reactor would run at l 1 S% (11.S MW 1) power for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The result , an upsurge in fission gas release , internal pressure rise , and additional thermal growth and pellet relocation are experienced , though the duration and frequency are not enough to cause any design limit to be exceeded.

The results of the FRAPCON analyses for the above operating scenario are shown in Table 4. Stresses and strains as a result of pellet-clad interaction are highest for the minimal cold gap -* The [S a , b , fJ results show that the primary pressure induced stress is < S MPa (72S psi), which is well within the primary stress limit of Zircaloy-4 of 430 MPa (62,366 psi) at temperature (Figure 16). With a factor of safety (FOS) of> 90 on primary stresses, the design meets ASME B&PV Code as well as Regulatory Page 30of190 ATTACHMENT 1 Guide 2.2 , Section C(l)(c)(3) on mechanical stress effects for materials of construction for reactor experiments (FOS against failure of > 2). Primary stresses are therefore not the driver for cladding failure. Secondary stresses as a result of thermal differential expansion and re-location of the cracked pellet on the other hand are the main drivers for cladding longevity.

A yield strain for Zircaloy-4 at temperature of about 0.66% is well within the strain range of twice of yield , 1.6%. Therefore, this meets the secondary stress intensity limit. With respect to cyclic fatigue for Zircaloy-4 , the maximum number of cycles that the cladding can sustain is 20,000 as shown in Figure 17 (Reference 3), which includes irradiated specimens.

With only six (6) expected cycles , the cladding has ample design margin. Gap Size Temperature (0 C) Pellet centerline Pellet surface Cladding ID Cladding OD Cladding Average Strain(%)

Radial Axial Hoop Gas Pressure (MPa) Cladding Internal Table 4 FRAPCON Results Summary -SOµm 2350 2359 379 419 303 301 182 182 242 242 -0.495 0.139 0.339 0.190 0.659 0.386 0.61 0.70 Page 31 of 190 -[Sa , d, e , f] 2515 529 324 177 250 0.145 0.109 0.150 0.81 ii a.. 600 500 Ill 400 Ill e Ui 100 0 *.. *****************

..... . 0 1()() ATTACHMENT 1

  • Zirc-4 (Irradiated)

....... 200 300 400 Temperature

(°C} Figure 16 500 Allowable Stress for Irradiated Zircaloy-4 Based on 2/3rd Yield [Geelhood 2008] 1 % strain (ASM E; factor of 2 on stress/strain)

"-/ 0.5%s tra in ;; 4: 100.0 >

  • I !! 10.0 := 0 .* o. 1 uOO D
  • 0 0.5 Ht I "'
  • cyc l .. In £OF -1 0000 10000000 l'a t l llUt lift Ir -----400000 cycles (A SME 20 , 000 cyc l es; fact o r of 20) FIO. 7-Cyd k p/4slic11roU.

"'""""""IHrofqcltstO rwp1-ft>rdotJdi"lhlbu.,,,;J7Pdiiiiiit

--M lrmtl141'd fowr eyc/'1 ill Oii F.Dft !'WR: ... Wlfoik;i sp<<iltttlU

,,,.,J lli/HIC;_M

!fl<<d <01 a ---. 70000 cycles '" JllTU '"""' /(/> t:y</IJ ltnH Ill 0 ... > ... JllTSJ"" ** Nphln. Figure 17 Fatigue Chart for Zircaloy-4 at 350 °C (un-irradiated and irradiated) 1 1 The reported va lu es for cyc l es in Fig ur e 4-14 were adj u s ted in accorda n ce t o ASME B&PV Co d e t o provide a fac tor of 2 o n stress/stra in and 20 on cyc l es (w hi c h ever i s more conservat i ve). With a facto r of> 2 on s tress/s tr a in , ASME B&PY cove r s the requirements of Regulatory G uid e 2.2. This means that in case of 0.5% strai n , applying a factor of 2 , or 1.0% stra in , th e quantity cycles are es tim ated at 70 , 000 , or in case of0.5%, u s ing a red u ct i o n factor of20 , th e expected cycles are 20 , 000 (400 , 000 cyc l es I 20). Note th at thes e are adj u stment fac t o r s t o the experimen tal data set t o obtain estima t es of live s of co mp onents per NUREG/C R-6815. Pa ge 32of190 ATTACHMENT 1 In case of the maximum gap of. microns , the upper limit of manufacturing tolerances , the concern is the ce nterline temperature of the target pellet. The maximum target pellet centerline temperature is 2515 °C ( 4577 °F) (Table 4), which will occur at startup when no relocation ha s occurred. This value is well withi n the pellet melting temperature of2860 °C (5180 °F). The analyses show that the target rod meets the design requirements and can be safely operated under nominal reactor conditions of 10 MWi. including the two (2) 11.5 MW 1 excursions at the beginning and end of the three (3) weeks. The design incorporates a safety factor of 3 , 333 for cladding cycles and > 25 , 000 for pressure-induced primary stress cycles. Additionally , the pellet cladding exhibits a thermal margin of safety of> 140 °C (252 °F), while the contained target pellet exhibits a thermal margin of safety of> 300 °C (540 °F). 2.2.4 Effect of Neutron Irradiation on Target Assembly Structural Materials The irradiation damage to TA materials of construction Al6061 , Zircaloy-4 and SS316L was calculated for full neutron spectrum using damage cross-sections (Reference 4). The maximum damage occurs at the active target vertical mid-plane.

The driver neutron flux from the reactor core drops off rapidly as a fu nction of vertical distance from the mid-plane.

Table 5 shows the component , location and damage in terms of displacements per atom (dpa). Table 5 Maximum Irradiation Damage to Target Cartridge Materials Fast Neutron dpa for Des i gn Life (0.1 -20 MeV) Material Component Location (weeks) Fluence Design Life (n/cm 2) (Ref. 5) Al6061 Target Cartridge Front face at active 52 2.0 x 10 2 1 1.12 target mid-plane Zircaloy-4 Target Rod Cladding Target rods 3 1.8 x 10 20 0.033 SS316L Target Cartridge Front fac e at active 52 2.2 x 10 2 1 0.38 target mid-plane Al6061 Target Housing Front face at active 520 7.6 x 10 21 4.29 target mid-plane SS316L Lower Plenum Front face 520 I .4 x 10 21 0.27 The impact of these low levels of irradiation damage on the material structural properties are as follows: Al6061 -This allo y has a high degree of radiation tolerance and can sustain damage levels in excess of 100 dpa. For damage levels of 1.12 dpa at 50 °C (122 °F), the alloy will experience less than 17% vo lumetric swelling.

The yield and ultimate strengths are increased by above 0.5 dpa. The alloy Page 33of190 [5 a , b , f]

ATTACHMENT 1 retains good ductility

(> 5% failure strain) after irradiation

however , the AL606 I tar get housing strain is very low at the maximum damage location. The use of non-irradiated properties per the ASME Section 11 code is appropriate for the use of this alloy in the TA (Reference 6). Zircaloy-4

-Zircaloy-4 is used reliably as fuel cladding for L WRs with fast fluences in the range of 10 22 n/cm 2 at 300 to 400 °C (572 to 752 °f). By comparison , the Zircaloy-4 cladding for the target rods will have a maximum fast fluence of 1.8 x 10 20 n/c m 2 (Table 5) at 100 to 250 °C, assuming a three (3) week exposure. This result s in the displacements per atom (dpa) from fast neutron (E 11 = 0.1 -20 MeV) irradiation of 0.033 dpa. B y comparison , L WRs experience peak Zircaloy dpa levels up to 20. However , the effect of irradiation at lower temperature is to lower the rate of annealing.

This increases the rate of damage and the level at which thermal-mechanical properties saturate as a function of fluence (or dpa). Reference 7 shows that at L WR temperatures , the properties tend to saturate at a fast neutron fluence of 1 x 10 2 1 n/cm 2 , but at research reactor temperatures

(< 100 °C), some properties will saturate at x 10 20 n/cm 2 , which is lower than the maximum fast fluence experienced by the target. Nevertheless , the total damage to the Zircaloy-4 cladding is not significant for its intended service. The effect on thermal conductivity is negligible (Reference 8). At the lower target operating temperatures, the tensile strength will increase slightly (+2 0%) at the expense of some loss of ductility

(-30%). At 250 °C (482 °f), the irradiation effects will be negligible.

These factors are accounted for in the design and structural analysis of the cladding.

SS316L -This alloy will experience a small increase in tensile strength and a corresponding small los s of ductility for the maximum dose of 0.38 dpa at the operating temperature range. This maximum exposure is very small and occurs at the longitudinal center of the cladding.

The yield strength increases by about 25% while the failure strain decreases by about 30% (Reference 9). Because th e SS3 l 6L cladding experiences very low strain along its length , the loss of ductility is not a concern. The fracture toughness will experience a small (< 15%) decrease (Reference JO). This is also an insignificant impact on structural design due to the small negative effect of the low dpa. Therefore , the use of non-irradiated properties per the ASME Section II code is appropriate for the use of this alloy in the TA. 2.2.5 Bowing of Target Rods The LEU target rods are driven by a directional neutron flux from the reactor core; hence, they can be expected to experience a power gradient across the U0 2 pellet. This will cause a small degree of bowing of the target rods in the radial plane of the core as a result of the following two (2) effects: 1. The reactor core facing side of the rod will have a slightly higher temperature than the opposite side causing differential thermal expansion; and 2. The reactor core facing side of the rod will be exposed to a slightly higher fast neutron flux causing a small degree of differential irradiation induced strain. Conservative evaluations of both effects have been carried out and the estimated bowing is shown to have no effect on rod heat removal. Page 34 of 190 ATTACHMENT 1 The power density distribution was analyzed for the target rod where the power density of the front half of the pellets was found to be on average 6% greater than the rear half. At the worst axial location , the front half power density was found to be 16% greater than the rear half. This worst-case assumption is extremely conservative , as onl y a few pellets throughout both targets come close to this power skew , and the average skew for rods containing those points is much lower. In addition , the entire stack was conservatively assumed to have the maximum power density of -The combination of [Sa , d , e , f1 maximum tilt with maximum power density produces a front-to-back temperature difference of about 1 6.8 °C. A thermal analysis was performed for the worst power density , showing a total axial growth of the target rod cladding of approximately 0.65 mm. Output from a FRAPCON analysis showed that the induced axial growth of the target rod cladding was 0.56 mm. Therefore , it was deemed that the effects of the irradiation induced growth could be captured conservatively in a structural model by doubling the Zircaloy-4 coefficient of thermal expansion in the structural model. Figure 18 provides a visual depiction of deflection of a target rod for a conservative, worst-case power skew. The figure shows the black wireframe of the un-deformed geometry , with the deformed body colored according to the deflection.

The left portion of the figure shows a 1 : 1 scale of the deflection at its worst point , which shows the maximum deflection of0.61 mm as slightly greater than the thickness of the cladding (0.5 mm). The right portion of the figure shows the overall deflection of the target rod by magnifying the displacement by a factor of 30. Figure 18 True scale deflection 30xmacnified deflection Core direction

... Deflection Due to Target Rod Bowing for Worst-Case Front-to-Back Power Skew Given that the distance between the target rod and the cartridge wall when centered is about 4.4 mm (0.17 in), no contact would occur. Additionally, flow through the cartridge is highly turbulent and boundary layers are very thin , so this level of deflection would not affect the heat transfer to the coolant. A more realistic average power skew scenario predicts a deflection of onl y 60 microns , i.e. a factor often Page 35 of 190

--ATTACHMENT 1 les s than the conservative case. Given these results , it can be concluded that target rod bowing due to both thermal and irradiation effects will not affect the thermal-mechanic al performance of the target rods and cartridge assembly.

3. Target Cooling System The Target Cooling System (TCS) provides cooling water flow to the two (2) T As during normal operation and during both planned and unplanned shutdowns.

The TCS has two methods of cooling , corresponding to force cooling (pumped) flow durin g normal operation and natural circulation flow during post loss of forced cooling (LOFC) sce narios.

  • Forced Cooling: The reactor is operating and producing fission power in the T As. The TCS pumps water from the reactor pool , coo l s it through a heat exchanger that interfaces with the MURR secondary coolant system. The TCS then transports the cooled water to both T As. The cooling water exits the T As and flows back to the reactor pool.
  • Natural Convection:

In the event that pumped flow to the T As is interrupted , the reactor is s hut down and the TA power reverts to decay heat. The TCS transitions to natural circulation flo w to the T As in which valves on the target cooling inlet pipes are opened to allow pool water to flow through the T As as driven by natural circulation from target deca y heat. The TCS i s divided into two subsys tems , the target flow assembly and the water cooling module. The flow interface begins at the reactor pool and ends at the TA inlet. The heat rejection interface is with the MURR secondary coolant system for forced cooling operation and with the reactor pool for natural convection operation. The TCS instrumentation also includes inputs to the MURR reactor safety syste m to scram the reactor if forced flow falls below 85%, or above 115% of normal flow , and on high coolant temperature to the TA. 3.1 Functions The primary function of the TCS is to provide cooling water to the TAs at a sufficient mass flow rate to maintain the target material within its required temperature range for safe operation.

The is designed to perform the following functions:

3.1.1 Process

Functions

  • Transfer and reject the target heat to the MURR secondary coolant system during normal operation.
  • Maintain sufficient cooling water flow rate and temperature conditions through the target to maintain a large CHFR margin and such that any s ubcooled nucleate boiling has a negligible impact on target rod integrity.
  • Minimize net exchange of heat between the TCS and the re actor pool water during normal operation. Page 36 of 190 ATTACHMENT 1
  • Provide for reliable transition of target heat removal to the reactor pool via natural circulation in the event of an LOFC.
  • Provide instrumentation to assure that flow , temperatures and pressures are within specified conditions for operation.

3.1.2 Safety

Functions

  • Provide a signal to the MURR reactor safety system to enable a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation or for high heat exchanger outlet temperature.
  • Provide a corresponding signal to open the decay heat removal valves to initiate natural circulation flow to the T As that coincides with a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation. Accident analysis shows that even if no natural circulation occurs after an LOFC event , target decay heat w ill be transported directly to the pool. The targets will remain in a safe condition with no radioactivity released to the reactor pool. 3.2 Target Cooling System Principal Design Parameters The principal design condition for the TCS corresponds to both T As loaded with 11 target rods each and the reactor operating at 10 MW 1* The target power levels will vary slightly due to shifts in the driving neutron flux distribution caused by changes in burnup of reactor fuel elements , position of control rods and age of the beryllium reflector.

The TCS is designed to accommodate the maximum target thermal power and the maximum linear heat rate. The TCS is designed to maintain an approximate 10 °C (18 °F) cooling water temperature rise across the target cartridge(s) for the maximum power condition. The corresponding flow rate combined with the cartridge design assures the target rod CHFR is greater than 2.0 for the worst-case allowable operating condition.

Table 6 summarizes the principal TCS design parameters assuming the reactor is operating at 10 MW 1 and the combined target loading is 22 target rods. The system is controlled to provide --mass flow which is divided equally to two (2) TAs through manual adjustment of the throttle v alves on the supply lines to each TA. This flow rate is held constant for all operating conditions.

It includes an additional 5% flow to compensate for any leakage that might occur through the labyrinth seal between the cartridge and the target housing. The inlet temperature is adjusted so that the outlet of each TA is at or slightly above the reactor pool temperature , which is nominally 38.89 °C (102 °F). Page 37of190 [Sa , d , e , t]

ATTACHMENT 1 Table 6 Target Cooling System Design Parameters for Two Targets 2 Parameter Target rods heat generation rate , kWt Target structure heat generation rate kWt Target coolant mass flow rate including S% leakage bypass , kg/s Target System inlet temp , °C Target System outlet temp, °C (Reference at nomina l flow) 3.3 Target Cooling System Design The target cooling system (TCS) is comprised of two subassemblies

  • The Target Flow Assembly
  • The Water Cooling Module 3.3.1 Target Flow Assembly Design One Target Assemb l y .. .. 28.S (83.4 °F) 38.9 (102 °F) Two Target Assemblies

..

  • 28.S (83.4 °F) 38.9 (102 °F) A process flow diagram of the target flow assembly is shown in Figure 19. The Water Cooling Module (WCM) draws water from the reactor pool through the inlet pipe and cools it to the desired value by rejecting the heat to the MURR secondary coolant system. The cooled water exits the WCM through a single pipe that splits into supply lines to the two (2) T As. Each supply line has a throttle valve that is manually adjusted to balance the flow between the two (2) T As or to shutoff flow to the TA. After the flow is balanced , the valves are pinned to prevent inadvertent maladjustment or closure. Safety-related flow meters on each supply line provide redundant signals to the MURR reactor safety system to enable a reactor scram on high or low flow. RTDs at the TA inlets and the inlet to the WCM are used to control the cooling water inlet temperature to the T As. Each supply line has redundant natural circu l ation bypass valves , which in normal operation are pneumatically held closed , but are spring-loaded to automatically open on a low flow signal from the affected line. This allows reactor pool water to enter into the supply line to initiate natural circulation.

[Sa , d , e , f] 2 Th e Heat Exchan ge r will h a ve a cap a city of-' or with a -15% mar g in , of th e total (rods + s tructure) target heat [Sa , d , e , f] gener a tion rate. Page 38of190 ATTACHMENT 1 After flowing through the TAs , the cooling water is returned to the pool through the diffusers located on pipe extensions from the target cartridges. The purpose of the diffusers i s to slow down and redirect the flow to allow decay of any N-16 before it can circulate to the pool surface. The diffuser s also serve as the mounting point for the RTD to measure the target outlet temperature to provide a caloric measurement of the target power. Figure 19 ---

J----,,.--.+-il Ta r ee t Cooline r S ys tem M odule ---*-

Tuget Dec a y H e at R e m ovo l Vo l ve (Held Process Flow Diagram of Target Cooling System F igure 20 shows an elevation view of the TCS , while Figure 21 shows the TCS superimposed on the reactor in-pool systems and biological shield. The WCM is located on the third level of the Costar Tower in the reactor containment building. The supply line is split at the wye just before the access bridge between the Costar Tower and the reactor biological shield. The individual supply lines to each TA are run parallel to the access bridge to the reactor pool. Flexible couplings are located on each line that crosses from the Costar Tower to the biological shield to accommodate any differential displacement of the two structures. The pipe to the wye is 4-inch SS3 l 6L Schedule 40. Downstream of the wye , the piping is 3-inch SS3 l 6L Schedule 40 up to the pool surface. There the pipe transitions to Al606 l Schedule 40 via a flanged coupling.

Page 39of190 ATTACHMENT 1 Target Cooling Pumps Target Decay Heat Removal Valves Pump Discharge Pipes Target Cooling Heat Exchangers TCS Module Typical 3" ID Size Pipes --Flex Lines 1--Target Assemblies Figure 20 Elevation View of Target Cooling System Flexible ,.---couplings The downcomer supply pipes are supported from the floor structure at the top of the biological shield and the TAs are allowed to rest on the reflector support plate. A flexible coupling is located above each TA to accommodate any growth or differential motion between the TA and the pipe upper support. A single suction line draws water from the pool and is delivered the WCM. The other pump is isolated using a ball valve. The suction line is 3-inch Al606 J Schedule 40 in the reactor pool and transitions to SS3 l 6L outside of the pool. The water suction pipe inlet is fitted with a replaceable screen to prevent blockage by debris. The decay heat removal valves on each supply downcomer are positioned at a submerged location just above the refueling level. These valves are 2-inch butterfly valves with SS3 l 6L bodies. Each valve has an actuator rod that extends to a pneumatic actuator located above the pool surface. Page 40 of 190 ATTACHMENT 1 Target Decay Heat Removal Valves TCS Piping TCS Module MURR Pool MURR Costar Tower Target Assembly Figure 21 Target Cooling System Supe rimposed onto MURR Reactor Pool and Biological Shield 3.3.2 Water Cooling Module Design Figure 22 s ho ws the process flow diagram for the WCM , which incorporates two (2) redundant and connected coolant pumps and heat exchangers. If a pump or heat exchanger is taken out of service , the installed spare component can be placed in operation to pennit continued TEF operation.

The pumps are se lf-priming , centrifugal pump s with variable frequency drives (VFD), w hich are manually controlled to provide the required flow. The heat exchangers are s tainless steel , plate-type units. The heat exchan ge rs reject heat to the MURR secondary coolant system. The temperature of the heat exchanger outlet flow is controlled by a flow bypass valve on the secondary side of the heat exchanger. A target system P&JD drawing i s presented in Figure 23. Page 41 of 190 ATTACHMENT 1 Ta rg et Coolin g System Module Schematic Tereet Cool i n1 Syste m M odule 8 0\ln d*l"I ' .... ________ _ Pum p Figure 22 Ma i n P i p e o u t l e t S chematic for th e Water Cooling Module P age 42 of 1 90 T I I I T I I I I I I I 'T I l T T To M U R R S econdar v Coo la n t Systom T e mpe:r a ture Co n t ro l Valvt ATTACHMENT 1 II ..... 1 TARGET COOLING SYSTEM :r.1 p & ID FsO-RXOPS

--Fig ure 23 Target Cooling System Page 43 of 190 ! 11111**1 I I I * !

ATTACHMENT 1 The WCM is located on the third level of the Costar Tower at approximately the same elevation as the top of the bio l ogical shield. The coolant pumps , heat exchanger s , piping and contro l panel are mounted on a sing l e pa ll et and insta ll ed as a sing l e u nit. Tab l e 7 describes major specificat i ons of coo l ant pumps and heat exchangers for the TCS. Ta ble 7 Major Com p o n e nt Specifica ti o n s U nit s S I E ngli s h P ump Nominal Flow Rate(+/- 15%) --Max Head with 25% margin 262.6 kPa 38.09 ps i a Out l et Pressure w i t h 25% margin 25 1.6 kPa 36.49 ps i a Max I n l et temperat u re 48.9 °C 120 °F Material 316 SST N I A D r ive VFD NIA NPSH < 10 psia N I A Pump Type Se l f-priming NIA In l et flange pipe size 3-inch N I A Ou tl et flange pipe size 4-inch N I A Nomina l Power 1 4.9 kW 20HP Vo lt age 46014801500 VAC N I A Current Norma l Operation ( 460 V AC) 20 amps NIA Frequency 60Hz NIA P h ases 3 N I A Hea t Ex ch a n ger Heat d uty with I 5% margin -N I A Hot side flow rate (primary) 0.0142 m 3 1s 225 gpm Hot side inlet temperature nomina l (primary) 38.89 °C 102 Of Hot side outlet temperature no mi nal (primary) 28.89 °C 84 °F Hot side pressure d rop (primary)

< 68.9 kPa < 10 psid Differential approac h temperat u re (for refe r ence) 6.67 °C 1 2 °F Cold side flow rate (secondary) 0.0142 m 3 1 s 225 gpm Co ld side in l et te m perature (seco n dary) 22.22 °C 72 °F Co ld s id e outlet temperature (for reference) 32.22 °C Of Maximum allowa bl e working pressure (MA W P) 103.4 kPa 15 psi Co l d side pressure drop < 68.9 kPa <lOpsid Materia l 316 SST N I A Type Plate and frame NIA Page 44 of 190 [Sa, d , e , f]

ATTACHMENT 1 3.4 Target Cooling System Performance 3.4.1 Thermal-Fluid Performance in Normal Operation, Mode 1 The TCS must be on Mode 1 prior to and during the start of the reactor if the T As contain active target r ods. The process flow diagram with definition of state points for one (of the two) T As is shown in F igure 24. The figure also provides the flow , temperature and pressure corresponding to each state point. HF/LF Pressure*, kPa Tareet Assembly Figure 24 State Points within the Target Cooling System 3.4.2 Thermal-Fluid Performance in Natural Circulation Flow, Mode 2 To MURR Secondary Coolant System Temperature!

Control Valve From MURR Secondary Coo l ant System [Sa , d , e , f] Figure 25 shows a schematic of the TCS operating in the natural circulation mode. In the event of an LOFC, the reactor is tripped by a low flow signal (85% of nominal flow) if the event is cause by a reduction of flow (e.g. loss of offsite power) or a high flow signal (115% of nominal flow) if the event was caused by a pipe break downstream of the sensor. When the reactor is tripped , the driver neutron flux to the targets quickly decays and the target thermal power transitions from fission heat to fission product decay heat. Coincident with the reactor trip , an actuation signal automatically opens the two (2) deca y heat removal valves on the affected downcomer pipe(s) such that the downcomer is open to the Page 45 of 190 ATTACHMENT 1 reactor pool. This allows natural circu l ation of poo l water to the TA and back to the pool throu gh the diffuser.

' TARGET COOLING SYSTEM P & ID RX OPS Fig ure 25 Target Coo lin g System P&ID Page 46 of 190 II I 11**1 I I I

  • 1 ATTACHMENT 1 The analysis of the TCS performance in the natural c ir culation mode for the set of design basis events that can lead to a LOFC is provided later in this document.

In all design basis cases , the natura l circulation fl ow to the TA provides the required cooling. 3.4.3 Structural Performance During a Seismic Event A seismic analysis was performed on the piping components of the TCS based on the acceleration and spectra given in the American Society of Civil Engineering Code 7-10 and interpreted according to American Society of Mechanical Engineering B&PV B3 l .3. The specified design spectra corresponds to an acceleration of 0.113 g. The analysis was performed using AutoPipe , a specialized nonlinear finite element piping program. The WCM and TA flanged connections where considered idealized anchors. The piping supports are rail-mount vibration damping clamps and are modeled as lin e supports with no gaps and firmly anchored to the biological shield , the Costar Tower or the bridge. F igure 26 shows the location and type of pipe supports for piping located above the reactor pool. To account for independent seismic movement of the building to the pool, flexible piping is added at the bridge level. All flexib l e piping is 11 inches in length and flanged. A guide support is added with the fl ange connections on the non-flexible piping side. Stiffness values were based on General Atomics' test data. Figure 27 shows the location and type of pipe supports for in-pool piping. All other supports shown are Guide Supports Only Figure 26 exchangers.

Rigid Ancho r connections on the ends. Flexible Joints to simu l ate flexible piping. (Not shown on Solidworks model) Location and Type of Supports on Above Pool Piping Page 47of190 ATTACHMENT 1 Guide + Line Supports All other supports shown are Guide Supports Only From these flanges all piping in the negative Y direction is aluminum Flexible Joints to simulate flexible piping with anchors at target housing interface Figure 27 Location and Type of Support on In-Pool Piping Two (2) types of analyses were performed.

The first type considers the biological shield , bridge and Costar Tower as rigid structures due to the low seismic loading. The maximum deflections of structures within areas of interest are estimated to be < 1 /16th inch. Table 8 shows the key results of the rigid structure analysis.

The table shows that the minimum safety factor is 7.8 at the worst location.

Page 48 of 190 ATTACHMENT 1 Table 8 Results of Rigid Structure Seismic Analysis Location Material Stress Stress Allowable Safety Factor (ksi) (ksi) Y-pipe , before flexible piping SS316L 0.7 16.7 24 Inside pool , on expansion loop 6061-T6 1.0 16.7 17 At heat exchanger interface SS316L 3.2 25 7.8 At SS to Al pipe interface 6061-T6 0.2 25.I 126 3-inch to 2-inch Y-pipe SS316L 0.3 16.7 56 All Al piping 6061-T6 0.3 16.7 56 Y-piping support SS316L 0.9 22.2 25 Inside pool , on expansion loop 6061-T6 1.0 22.2 22 The second type assumes that the biological shield/pool and Costar Tower combined with the bridge will l aterally displace during a seismic event. To demonstrate the effectiveness of the flexible couplings , a hypothetical forced displacement of 2 inches is applied. By comparison , an initial seismic evaluation of the Costar Tower showed a less than I /8 111 inch lateral movement at the location where the bridge connects.

The AutoPipe model showed that the displacement has negligible effects on the stress ratio (max ratio is 0.05) demonstrating that the flexible piping design works as intended.

3.4.4 Secondary

Cooling for the Target Cooling System The heat from the TCS is transferred to the MURR secondary coolant system by means of one of the TCS heat exchangers. The heat is then dissipated to the atmosphere through a cooling tower with seasonal additional cooling provided by a chill water system. The secondary coolant loop that serves the TCS consists of one (1) coolant circulation pump , two (2) heat exchangers , two (2) automatic temperature control valves , one (I) chill water heat exchanger , temperature and flow instrumentation , and associated v alves and piping. A piping and instrumentation diagram is shown in Figure 28. The secondary coolant loop that serves the TCS shares a cooling tower , radiation monitoring instrumentation, and chemical control systems with the existing 15 MW capacity MURR secondary coolant system. Only the secondary coolant loop serving the TCS will be described in this section. The secondary coolant system is designed such that a failure or malfunction of any system component will not lead to target rod damage or an uncontrolled release of radioactivity to the environment.

3.4.5 Circulation

Pump The secondary coolant circulation pump for the TCS is a centrifugal , single-stage pump that is connected to a variable speed drive (VSD) unit through a coupling.

This pump , designated SP-5 , is capable of supplying 500 gpm at a discharge pressure of 50 psig. Three (3) other secondary coolant Page 49 of 190 ATTACHMENT 1 pumps are installed in a parallel configuration with each pump capable of supplying 2 , 200 gpm to the secondary coolant system. Jn case pump SP-5 failed , TCS secondary coolant flow can be established using a cross-connect pipe to the discharge header of the other three (3) secondary coolant pumps. Additiona l detail of the SP-5 control system is provided in Instrumentation and Control section of this document. z: ; ... ..... 0 rn GO z (') ::c --4 :;: . ' N .. z rn °' Figure 28 Secondary Cooling for Target Cooling System Page 50 of 190 ATTACHMENT 1 3.4.6 Heat Exchangers The secondary coolant flow makes a single pass on one of the TCS heat exchanger s. The TCS heat exchangers are plate-type heat exchangers.

On one s ide of the plates is secondary coolant , and the other s ide has TCS coolant. Only one TCS heat exchanger is required for TCS operation.

The other TCS heat exc hanger is an installed spare for operational reliability.

D uring the summer months , additional cooling to the secondary coolant supply to the TCS may be needed to ensure a sufficient secondar y cooling water temperature to maintain proper TCS heat exchanger outlet te mperature.

This additional cooling to the secondary coolant system is provided by a chill water system v ia a plate-type heat exchanger.

3.4.7 Automatic

Temperature Control Valves The amount of secondary coolant passing through the TCS heat exchanger is controlled by an automatic te mperature control valve located in a bypass line of its associated heat exchanger.

The valve is an electro-mechanical butterfly valve which responds to the TCS heat exchanger outlet temperature which maintains a constant cold leg temperature to the T As. Since only one TCS heat exchanger is operated at any time , only one automatic temperature control valve is in operation at any time. The other valve is an i nstalled spare for operational reliability.

The automatic temperature control valves for the TCS heat exc hangers are designated as S-3A and S-3B. Additional detail on the operation of the automatic temperature control valves and system is provided in Instrumentation and Control section of this document.

3.4.8 Instrumentation

The secondary coolant inlet and outlet temperatures to both the chill water and TCS heat exchangers and the total flow in the secondary s y stem TCS loop are displayed and recorded in the reactor control room. 3.4.9 Electrical Power System Supporting the TCS Electrical power to the TCS pumps and their Instrumentation and Controls Uninterruptible Power Supply (l&C UPS) will be supplied by Motor Control Center No. 4 (MCC-4) through a new 480/277 V distribution panel. Normal Electrical Power System Substation B supplies 480-volt , 3-phase , 60-cycle electrical power to MCC-4 through a 250-amp breaker as described in Chapter 8 of the MURR Safety Analysis Report (Reference 1 I). The new 480/277 V , 200-amp panel , designated HVP-4 , will be connected to MCC-4 in the reactor containment building.

Electrical power to the TCS pumps will be provided by HVP-4. HVP-4 will also s upply electrical power to a new 120/208-V Distribution Center , designated 120/208V Distribution Center 3 , through a 112.5-kV A transformer. A Lighting Panel on l 20/208V Distribution Center 3 will s upply electrical power to the TCS pumps I&C UPS as shown in Figure 29. Page 51of190 ATTACHMENT 1 Location of the new electrical wiring will ensure that no electromagnetic interference will exist between the electrical power service and any safety-related I&C circuits.

1 I COLLING 3 PUMP 5 7 9 11 13 15 17 19 21 23 25 27 29 FroM MCC 114 Breo.ker A6 I HVP-4 480/277 200A 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 PUMP I 480-277V ) 112 KVA 3PH I I TCS !LC 3 UPS Panel 5 7 9 Figure 29 120/208V DISTRIBUTION CENTER-3 400A 2 4 6 9 10 Electric Power Supply to TCS Pumps and I&C UPS 4. Instrumentation and Control System This section describes the operating characteristics of the TEF Instrumentation and Control (l&C) Systems. These systems assure that the TEF can be safely operated , monitored , and shut down as warranted.

4.1 Summary

Description The TEF I&C Systems are comprised of the sensors , electronic circuitry , displays , and actuating devices available to provide the information and means to safely control the TEF and avoid or mitigate potential accidents that could affect the TEF or the reactor. The TEF I&C systems include the following:

  • Target Cooling System (TCS) Control System
  • TCS Protection System Page 52 of 190 ATTACHMENT 1
  • TCS Parameter Indication , Recording , and Alarm System
  • TCS Secondary Coolant Control System
  • TCS Secondary Coolant Parameter Indication , Recording , and Alarm System
  • N-16 Reactor Power Monitoring System
  • Pool Coolant Monitoring System The TCS Control System is designed to control TCS pump speed and the position of the Decay Heat Remova l System Automatic Valves. It is a hardwired system that relies on relays , controllers , push b uttons , switches , and light indicators.

The TCS Control System receives inputs manually and from the TCS Protection System. The TCS Protection System is designed to prevent operation of the TEF in regions in which target rod damage could occur. This is accomplished by the TCS Protection System which can initiate a reactor s cram via the reactor safety system, which instantaneously drops the reactor control rods and takes the r eactor subcritical.

Therefore , the power of both the TEF and the reactor are quickly reduced. Inputs w hich govern the TCS Protection System output are supplied by TCS flow and temperature transmitters.

The TCS Parameter Jndication , Recording , and Alarm System is designed to provide control room o perators with indications of operating parameters for the TCS , to record TCS parameter data for t erm retention or review , and to actuate alarms to alert the operator to abnormal parameters.

This is accomplished by parameter current loop signals going to chart recorders which also serve as parameter i ndication and alarm units. Parameter inputs into the TCS Parameter Indication , Recording , and Alarm System are multiple TCS temperature, pressure , and flow signals. The TCS Secondary Coolant Control System is designed to control the TCS secondary coolant circulation p ump (SP-5) speed and the position of the operating target cooling automatic temperature control valve (S-3A or S-3B). It is a hardwired system that relies on relays , controllers , push buttons , switches , and li ght indicators.

The TCS Secondary Coolant Control System receives inputs manually and from the TCS P arameter Indication , Recording , and Alarm System. The TCS Secondary Coolant Parameter Indication , Recording, and Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS Secondary Coo lant Control System , to record TCS secondary coolant parameter data for long-term retention or review, and to actuate alarms to alert the operator to abnormal parameters.

This is accomplished by para meter current loop signals going to chart recorders which also serve as parameter indication and alarm units. Parameter inputs into the TCS Secondary Coolant Parameter Jndication , Recording , and Alarm System are multip l e TCS Secondary Coolant Control System temperature , flow , and valve position s ignals. The N-16 Power Monitoring System (N-16 PMS) is designed to provide control room operators with i ndication of reactor core power by measuring the amount of N-16 produced in the primary coolant system. Reactor fission power is directly proportional to the amount ofN-16 atoms produced by the fast Page 53 of 190 ATTACHMENT 1 neutron flux in the reactor. Therefore , reactor core power can be measured separate from the TEF power. The N-16 PMS has two (2) reductant detectors located in mechanical equipment room 114 and two (2) redundant displays located in the con trol room. The Pool Coolant Monitoring System (PCMS) is designed to provide control room operators with indication and an alarm of elevated iodine-131 (1-131) activity levels in the pool coo l ant system. Elevated 1-131 activity levels in the pool coolant system would be an early indication of a leaking target rod in t h e reactor pool. The PCMS has a detector , display , and alarms in the entryway to mechanical equipment room 114 and has an indication and a l arm in the contro l room. 4.2 Target Cooling System Control Syste m Description The TCS Control System is designed to control TCS pump operatio n and the position of the Target Decay Heat Removal Valves (TDHRVs).

From the TCS Control System , the operator can start , stop , and change speed of the TCS pumps. ln add i tion , the TCS Contro l System contains an interlock between the TCS pumps and the TDHRV position that secures the TCS pumps when any one of the TDHRVs open. The interlock is actuated from a position indication sw it ch located on each TDHR V actuator w hi ch opens a contact d e-energizing both of the TCS pump contro l circuits.

The co ntrol system arc hi tecture for the TCS is shown in Figure 30. A larger image of the TCS Contro l Panel is s ho wn in Fig ure 3 1. MURR CONTROL ROOM Control Panel with Hardware Push Buttons/ Switches and Indicators Process Instrument Independent Hard Wired Safety System Functions:

-Hardwired isolated interface with MURR SCRAM loop MURR SCRAM MURR CONTROL SYSTEM LOOP VFDs with access i ble control pads for local maintenance control COOLING WATER MODULE Figure 30 t Flow , 4*20mA Temperature , 4-20mA Critica I P r ocess Instrument Target Cooling System Control System Architecture Page 54 of 190 ATTACHMENT 1 When the TDHRVs are in 'Manual' mode , the operator can individually change the position of each TDHRV as needed for maintenance or testing. In 'Auto' mode , low or high TCS flow will automatically open the two (2) TDHRVs in the associated TCS branch of the abnormal flow condition.

When a low or high TCS flow condition occurs in a TCS branch , one or more contacts in the TCS Protection System will open and de-energize the two (2) solenoid-operated va l ves that are applying air pressure to the TDHRV actuators in that TCS branch. Air pressure will be vented off the TDHRV actuators and spring force will rapidly open the TDHRVs , when required. TAR GET C OO L I NG SYST E M C O NTR O L P ANEL I 24 V Power Loss I I UPS Power Loss I TCTl-4 A OBO 0 T CTl-1 000 1 0 00 1 * *

  • Pool T emp I TCHX Outlet Te m p i I TCP A VFD Fault I I 000.0 1 0 I 0/o Load I I 000.0 1 TCPA Ready * [i] ' 0 I TCPA I Auto/Manual TCDP-1 I 000.0 1 I 000.0 1 I TCP B VFD Fault I 0 I %L oad I TCP B Ready [i] I TCP Selector A or B I
  • I Spee<l (X10)1 *
  • Figure 31 Target Cooling System Control Panel 4.3 Target Cooling System Protection System Description The TCS Protection System initiates a reactor scram via the reactor safety system to prevent operation of the TEF in regions in which target rod damage could occur. The TCS Protection System will initiate a reactor scram based on low TCS flow rate ( < 85% normal flow), high TCS flow rate (> 115% normal Page 55 of 190 ATTACHMENT 1 flow), and high TCS heat exchanger outlet temperature

(> 105 °F). These set points are based on the limiting conditions used in the thennal-hydraulic safety analysis for the TEF. The basis for the TCS low flow scram is to ensure target fi s sion heat generation is terminated upon a low flow condition caused by a TCP failure , TCS piping break upstream of the flow transmitters or on the TCP suction , or TCS piping blockage.

The basis for the TCS high flow scram is to ensure target fission heat generation is terminated upon a TCS pipe break between the flow measuring element and a TA which would reduce flow resistance and increase flow indication at the associated flow transmitter. This condition causes low flow to the T As. The basis for the TCS high temperature scram is to ensure target fission heat generation is terminated upon a high temperature condition that is not associated with a change in TCS flow rate. Insufficient secondary coolant flow and high secondary coolant temperature could cause such a high TCS temperature to occur. The TCS Protection System is designed with all analog components.

The components and system are designed to be fail-safe , meaning that any loss of power or signal will cause a reactor scram. In addition , all TCS Protection System signals are separate from TCS Control System signals. Keyed bypass switches to the TCS Protection System allows for reactor operation with the TCS not operating , one TCS branch operating , or both TCS branches operating.

The keyed bypass removes the protection signal inputs for any part of the TCS that is not operating.

4.3.1 Target

Cooling System Low and High Flow Rate Protection TCS flow is measured at the following locations with the indicated TCS Flow Elements (TCFEs) and TCS Flow Transmitters (TCFTs):

  • Downstream ofTCS combined cold leg header on branch A
  • Downstream ofTCS combined cold leg header on branch B Flow elements TCFE-lA and TCFE-lB for Target A and Target B , respectively , are located in each TCS branch. The differential pressure caused by TCS flow is measured b y flow transmitters TCFT-1 A and TCFT-2A for TCS branch A and TCFT-lB and TCFT-2B for TCS branch B. The output signal (4 to 20 mA) generated by each flow transmitter is directed to a square root converter which provides a linear output signal for the two dual alann trip units and a chart recorder in series with the signal. In addition to providing flow indication and recording , the recorder will initiate a " TCS Lo Flow" alarm when TCS branch flow to either target decreases to 90% of its normal value. If TCS branch flow decreases to 85% normal flow , then the following actions are initiated by the system: 1. Reactor scram; 2. Opening of the two (2) TDHRVs in the associated branch; and Page 56 of 190 ATTACHMENT 1 3. Actuation of the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-1 A opens two (2) contacts:
1. One contact opens in the " Yellow Leg" of the TCS Protection System which in turn opens contact TCS-1 in the "Yell ow Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-2A opens two (2) contacts:
1. One contact opens in the "Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the " Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the " TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-lB opens two (2) contacts:
1. One contact opens in the " Yellow Leg" of the TCS Protection System which in turn opens the contact TCS-1 in the " Yellow Leg" of the reactor safety system which causes the scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-2B opens two (2) contacts:
1. One contact opens in the " Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. In addition to a low flow condition , the TCS Protection System will provide protection on TCS high flow rate. In this case , the chart recorder will initiate a "TCS Hi Flow" alarm when TCS branch flow to either target increases to 110% of its normal value. If TCS branch flow increases to 115% normal flow , then the following actions are initiated by the system: Page 57 of 190 ATTACHMENT 1 1. Reactor s cram; 2. Openin g of the two (2) TDHRVs in the as s ociated branch; and 3. Actuation of the " TCS Flow Scram" annunciator alarm. The high flow dual alarm unit for TCFT-IA opens two (2) contacts: 1. One contact opens in the " Yellow Leg" of the TCS Protection S ys tem which in tum opens contact TCS-l in the " Yell ow Leg" of the reactor safety s y stem which causes the reactor scram (See Figure 32). 2. Another contact open s in the TCS Control System that de-energizes both s olenoid-operated valves appl y ing air pressure to TDHRV-IA and TDHRV-2A actuators which allows these valves to open. The openin g of this contact also activate s the " TCS Flow Scram" annunciator alarm. The high flow dual alann unit for TCFT-2A opens (2) two contacts:
1. One contact opens in the " Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the " Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens m the TCS Control S ys tem that de-energizes both solenoid-operated valves appl y ing air pressure to TDHRV-IA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activate s the " TCS Flow Scram" annunciator alarm. The high flow dual alarm unit for TCFT-1 B opens two (2) contacts: I. One contact opens in the "Yellow Leg" of the TCS Protection System which in tum opens contact TCS-1 in the " Yell ow Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The high flow dual alarm unit for TCFT-2B opens (2) two contacts:

I. One contact opens in the " Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the " TCS Flow Scram" annunciator alarm. Pa g e 58 of 190 ATTACHMENT 1 4.3.2 Target Cooling System High Temperature Protection TCS combined cold leg header temperature is measured for use in the TCS Protection System at the following location with the indicated TCS Temperature Elements (TCTEs) and TCS Temperature Transmitters (TCTTs):

  • Downstream of the TCS Module on the TCS combined cold leg header Resistance Temperature Detectors (RTDs) TCTE-2 and TCTE-3 are located beside each other in the TCS combined cold leg header. TCS Temperature Transmitters TCTT-2 and TCTT-3 convert the associated RTD signal into 4 to 20 mA output signal providing a linear output temperature signal for a dual alarm trip unit and a chart recorder in series with the signal. In addition to providing TCS cold leg temperature indication and recording , the chart recorder will initiate a "TCS Hi Temp" alarm when TCS cold leg temperature increases to 100 °F. If TCS cold leg temperature increases to 105 °F , a reactor scram and a " TCS Temp Scram" annunciator alarm are initiated.

The high temperature dual alarm unit for TCTT-2 opens a contact in the "Yellow Leg" of the TCS Protection System which in tum opens contact TCS-1 in the " Yellow Leg" of the reactor safety system which causes the scram. The high temperature dual alarm unit for TCTT-3 opens a contact in the "Green Leg" of the TCS Protection System which in tum opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor to scram (see Figure 32). Page 59 of 190 ATTACHMENT 1 Yellow Leg Bypass 1S30 Green Leg Bypass 1 S31 Target A Yellow Target B Yellow Both AB Yellow Target A Green Target B Green Both AB Green l S30 Normal Leg Byp ss L g Bypass Leg 1531 Normal Leg Bypass Leg Bypass Leg Bypass 2 x x x x x r;: IT A I Tqot FloW H!F I } (TCFT-1A)

I Tar got 1 Torgol f loW Law (1930) J L__ ---(1S30) 2 Heol fs11819'"01'lel (TCT E-3) High TllJ1llnlllft T erf<< B .....---+

1 r .. g111 FloW Hl"1 I argot B 3 } (TC FT-1 B') t.= -::_ __ J YlllowS*lly Log 1llp 1 2 3 x x x x x

i I llglt flow High I I } I g .. glll f lOW L OW (1S31) I L__ _ __ ____J (183 1) HMlflClm1gll(lulllt 2 Hl"1 T enperftlre (TCT&2) f;;-----1 I 3 .,.rllll"I floW H W! I Targl!IB ..L Byp:oss Torglt rtowl.Ow } (TC FT-28) I I (1S31) L __ c;,_.11111y LegT!1> ___ _J _ I *---------* (T CS-1) (T CS-2) j_ ________ _ T Yellow Safety Leg Green Safety Leg Figure 32 Target Cooling System Protection System Relay Inputs to the MURR Reactor Safety System Page 60 of 190 ATTACHMENT 1 4.3.3 Target Cooling System Protection Bypass Capability When it is desired to operate the reactor without operating one or both of the T As , two (2) bypass key s wi ll be inserted and switches wiJI be placed in the b y pas s position for the a s sociated TA's (or assemblies

') protection to be bypassed. The bypass switches will be a four-position , standardized type switch designed 1 S30 and 1 S3 l. With no key in bypass switch 1 S30 or 1 S3 l, the switch is locked in the " Norma l" position and aJI TCS protection signals can affect a reactor scram. In addition , the TDHRV to TCS pump interlock is active to s hut off the TCS pump when any TDHRV open s. When a key is in bypass switch 1 S30 , the switch is able to be positioned in the 'Target A Bypass ,' 'Target B Bypass ,' or 'Both AB Bypassed' position.

With 1 S30 in either the 'Target A Bypass' or the 'Target B Bypass' position, the " Ye llo w Leg" TCS flow protection scram for the bypassed TCS branch is disabled. With 1S30 in either the 'Both AB Bypassed' position , the " Yellow Leg" TCS flow protection scram and the " Yellow Leg" TCS heat exchanger outlet high temperature scram for the bypassed TCS branch are disabled.

However , bypass switch 1 S30 only disab l es the " Yellow Leg" protection.

Bypass switch 1 S3 l has the same functions as 1 S30 , but it only disab le s the " Green Leg" protection.

Therefore , to bypass scram protection from one or more TAs and remove the associated TDHRV to TCS pump interlock , both swi tches must ha v e their keys installed and positioned correctly. Thi s design ensures that no single switc h or single human error will inadvertently remove all scram protection or the TDHRV to TCS pump i nterlock to a TA that is being irradiated.

F or example , if only Target A i s to be irradiated , the TCS would be lined up to provide cooling flow to only TCS branch A. TCS branch B wo uld have no flow and the TDHRVs in TCS branch B would be open. Therefore , two keys are inserted and switches 1 S30 and 1 S3 l would be positioned to 'Target B Bypass' so that the no flow condition in branch B wou ld not cause a scram and the open TDHRVs in branch B would not prevent running either TCS pump. When switches I S30 and I S3 J are in 'Target B Bypass,' relays are energized that close contacts that bypass the reactor safety system contacts actuated by TCFT-1 B and TCFT-2B in both legs of the TCS Protection System (See Figure 32). In addition , these energized relays close contacts that bypass the TDHRV-lB and TDHRV-2B valve contacts in the TCS pump controllers thereby allowing the TCS pump to run while the cooling branch B TDHRVs are open (See Figure 33). If neither Target A nor Target B is to be irradiated , the TCS pumps would both be secured , and all four TDHRVs would be open. The two (2) bypass keys are inserted, and switches 1S30 and 1S31 should be positioned to 'Both AB Bypass' so that the no flow condition in both branches A and Band/or a high heat exchanger outlet temperature would not cause a reactor scram. The open TDHRVs in both branches wo uld prevent running either TCS pump. When switches I S30 and 1 S3 l are in 'Both AB Bypass ,' relays are energized that close contacts that bypass the reactor safety system contacts actuated by TCFT-1 A , TCFT-2A , TCFT-lB , TCFT-2B , TCTE-2 , and TCTE-3 in both l egs of the TCS Protection System (See F igure 32). However , TDHRV-lA, TDHRV-2A , TDHRV-lB, and TDHRV-2B valve contacts in the TCS pump controllers would be open thereby preventing either TCS pump to run (See Figure 33). Page 61 of 190

4: T""" * > 0::: I a I-1S30 Target A Bypass ATTACHMENT 1 4: N I > 0::: I a I-1S31 Target A Bypass CD T""" I > 0::: I a I-1S30 Target B Bypass CD N > 0::: I a I--1S31 Target B Bypass Figure 33 TDHRVs-TCS Pump Interlock Circuit TCS Pump Control Finally , if a single switch fails or is placed in the incorrect operating position , the worst condition that could be possible would be to have two (2) T As operating with one key switch in the 'Normal' position and the other key switch in 'Both AB Bypass.' In this condition , the high and low flow scrams and the high heat exchanger outlet temperature scram would be disabled for only one ( 1) of the two (2) reductant reactor safety system legs. Any other single switch mis-position or single switch failure with any operating condition would cause less protection to be bypassed.

In addition , any single switch position or single switch failure with only one (1) TA operating would cause the operating TCS pump to stop due to the open TDHRVs in the secured TA cooling branch. The stopped TCS pump would reduce target flow and cause a reactor scram. Administratively , the keys to switches 1 S30 and 1 S3 I will require Reactor Manager's permission to use. Since the MURR is not normally operated with bypass keys installed at the reactor console , the presence of keys will be a noticeable exception from normal operations.

Therefore , operator attention to correct operation of the bypass keys will be peaked when they are implemented.

4.3.4 Target

Cooling System Parameter Indication, Recording, and Alarm System The TCS Parameter Indication , Recording , and Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS , to record TCS parameter data for long-term retention or review , and to actuate alarms to alert the operator to abnormal parameters. No Page 62 of 190 ATTACHMENT 1 safety-related functions are associated with the TCS Parameter Indication , Recordin g, and Alarm System. Every parameter is detected by a se nsor. The se nsor s i g n al i s converted to a 4-20mA current loop signa l going to paperles s chart recorder s which also serve as param e ter indication a nd alarm units. 4.3.5 Target Cooling System Secondary Coolant Control System The TCS Secondary Coolant Control System is de s igned to control the TCS secondary coolant circulation pump (SP-5) speed and the position of the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B). From the TCS Secondary Coolant Control System , the operator can adjust SP-5 to make co urse secondary coolant flow rate changes to the WCM. However , fine secondary coolant flow rate adjustment through a TCS heat exchanger is made by either S-3A or S-3B that are located in a secondary coo lant pipe that bypasses the associated heat exchanger.

For simplicity , further explanation of how the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B) operate , S-3A will be assumed operating and its operation will be the example described. S-3B's controls and operation are the same as S-3A. During steady-state operation , SP-5 provides constant secondary coolant flow to the WCM. S-3A and its associated heat exchanger are in use , and S-3A is in 'Automatic' control. S-3A is located in a secondary coolant bypass line that bypas ses the TCS heat exchanger.

Therefore , as S-3A closes , less secondary coolant bypasses the heat exchanger resulting in more cooling flow to the heat exchanger.

The opposite effect occurs as S-3A opens. The position of S-3A is controlled by a controller that is driven by the TCS heat exchanger outlet temperature transmitter TCTT-4. TCTT-4 receives its temperature measurement fro m resistance temperature detector TCTE-4. When slight changes to the TCS heat exchanger outlet temperature occur , S-3A move s to regulate the heat exchanger outlet temperature back to the desired set point. At the TCS Secondary Coolant Control System panel , the operator ha s manual switches and buttons to s tart , stop, and adjust the speed of SP-5. During normal operation , the operator will adjust SP-5 speed as necessary to ensure proper automatic operation of S-3A. S-3A has a 'Manual' control in which TCTT-4 does not affect the valve movement.

In 'Ma nual', only the c ontrol room operator has buttons to open or close the valve. Manual co ntrol can be used for rapid c hanges in S-3A during operation , maintenance , or testing. 4.3.6 Target Cooling System Secondary Coolant Parameter Indication, Recording, and Alarm System The TCS Secondary Coolant Parameter Indication , Recording , a.nd Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS Secondary Coolant Control System, to record TCS sec ondary coolant parameter data for long-term retention or review , and to actuate alarms to alert the operator to abnormal parameter s. No safety-related functions are associated with the TCS Secondary Coolant Parameter Indication , Recording, and Alarm System. Parameter current loop signals goi ng to chart recorders which also serve as parameter indication and alarm units. Parameter input s into the TCS Secondary Coolant Parameter Indication, Recording , and Page 63 of 190 ATTACHMENT 1 Alann System are multiple TCS Secondary Coolant Control System temperature , flow , a nd valve position signals. 4.3.7 N-16 Power Monitoring System The N-16 Power Monitoring System (PMS) is designed to provide control room operators with indication of reactor core power by means of measuring the amount of N-16 produced in the primary coolant system. Reactor fission power is directly proportional to the amount of N-16 atoms produced by the fast neutron flux in the reactor. Therefore , reactor core power can be measured separate from the TEF power. The N-16 PMS has two (2) redundant detectors located in mechanical equipment room 114 and two (2) redundant displays located in the control room. The detectors will be placed in a shield with a collimator facing the primary coolant outlet pipe. The detectors are sealed stainless steel ionization chambers capable of measuring the dose rates at the designated measurement point. The N-16 PMS detectors are located on the reactor primary coolant hot leg piping at a distance corresponding to the time delay for primary coolant exiting the core to reach the measuring point of 1.35 times the N-16 half-life.

This power measurement is calibrated b y performing a series of power calorimetric calculations to determine reactor core power while not operating the TEF. This method of power measurement is used by several domestic and international research reactors and has been found to be stable and reliable.

4.3.8 Pool Coolant Monitoring System The Pool Coolant Monitoring System (PCMS) provides display and alarm functions both locally at the entryway to mechanical equipment room 114 and remotely in the control room based on I-131 activity concentration in the pool coolant system. Elevated 1-131 activity levels in the pool coolant system would be an early indication of a leaking target rod in the reactor pool. The PCMS has a detector in room 114 entryway. The PCMS continuously monitors pool coolant activity anytime the pool coolant system is in operation. The monitor receives 1 gpm of pool coolant from a Yi-inch vent line connection on the pool coolant circulation pump combined discharge header downstream of the pool coolant circulation pumps. The pool coolant water flows through a filter and a cation resin bed before it is monitored by a gamma scintillation detector in a 2.7-liter volume of the system. Once the pool coolant leaves the monitored volume, the coolant returns to the pool coolant system via Yi-inch vent piping on the pool water holdup tank. Given a 1 gpm flow rate , the overall time for the PCMS to detect I-131 in the pool coolant system is approximately 13 minutes. The 13 minutes is based on the I-131 taking more than 11 minutes to get to the PCMS detector volume and almost two (2) minutes for the detector to recognize the increased 1-131 concentration. Page 64 of 190 ATTACHMENT 1 5. Target Assembly Nuclear Design Analysis The target physics model predict s the producti on of Mo-99 as a function of the dimen sio n , location and loading of the tar get rods and the neutron flux distribution in the target position.

The physics model also provides the distribution of power densities in the U0 2 pellet s for the thermal-hydraulic design of the TA. The target physics model incorporated both the T As and the MURR reactor core and examined the full range of MURR operations to develop the target physics de sign cases. 5.1 Analytical Methods The nuclear design calculations were conducted using Monte Carlo code MCNP6 (Reference

12) and ENDF/B-Vll.l (References 13 and 14) to obtain eigenvalue and neutron flux distributions.

The MCNP6 calculations also produce heating rates for the reactor fuel elements and target rods , which are used for the nuclear design and analyses to obtain the TA power. For the determination of the TA uranium-23S enrichment distribution in MCNP6, 100 million source histories (10,000 x I0,000 neutrons) were used to obtain converged eigenvalue and power distributions.

5.2 Target

Assembly Physics Model The T As are loaded in graphite reflector positions No. SA and No. SB. The physics design and analyses of the TAs utilized the "MURR 201S reflector model" (Reference IS), which includes the reactor core and associated structures such as detailed modeling of the control blades (CB), regulating blade , beam tubes , beryllium (Be) reflector , graphite reflector , and experimental holes. The MCNP6 model of the MURR core with two (2) T As is shown in Figure 34. The target rod physical dimensions used for the nuclear design calculations are summarized in Table 3. The physical dimensions utilized in the nuclear design are exactly the same as those of the mechanical design , which includes 11 target rods per target rod cartridge (flow channel), cartridge top and bottom fi xture, cartridge guide rail , neutron shield, channel housing , and water plenum. The cartridge locking device , water diffuser , and incoming water pipe are not included in the physics model as they are expected to have little influence on the analyses.

The phy sics model of the TA is shown in Figure 3S at the axial mid-plane , and target numbering is shown in Figure 36 for the baseline TA model. --The target rod model includes target pellets, pellet-clad gap, cladding, top plenum spring , and the top and bottom endcaps. In the physics model, four (4) pellets are defined as a single axial node (2.4 cm). The effective isotopic densities of U0 2 pellet consider 1.78% of open pore due to the pellet's dish and chamfer design. The plenum spring was homogeneously mixed with the helium filling gas. Due to the complexity of the geometry of the top and bottom fixtures , the physics model of these components was designed to preserve the outer dimension and mass of the Zircaloy-4 material.

Page 6S of 190 [Sa , b d, e , fJ ATTACHMENT 1 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane

[Sa , d , e , f] Figure 35 MCNP6 Model of Target Assembly at Axial Mid-Plane Page 66 of 190 ATTACHMENT 1 [Sa , b , d , e , t] Figure 36 Target Rod Numbering for Two Target Assemblies (baseline model) [Sa, b , d , e , t] Figure 37 Page 67 of 190 ATTACHMENT 1 5.2.1 MURR Driver Core States The physics design evaluates the target cartridge performance as well as the reactivity insertion to the MURR reactor core and driver fue l power changes due to TA loading and changes of the MURR core states. The TA design was conducted for the equilibrium core represented by its minimum , average and maximum burnup as summarized in Table 9. The extreme bumup core represents an anticipated driver fuel bumup distribution that can create the worst power peaking in both the driver fuel and target pellets , which has been generated from the maximum burnup core case without xenon. The critical CB position (distance from its fully inserted position) is the lowest at beginning-of-cycle (BOC) and withdraws as the driver fuel depletes and fission products accumulate in the core. For the typical 1-week operating cycle, the CB reaches its equilibrium height about 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Day 2 state) after startup and slow l y withdraws until the end-of-cycle (EOC). Table 9 Definition of MURR Driver Core Burn up States (MWD) Driver Fuel Extreme Burnup Minimum Average Burnup Maximum Element Core Burnup Core Core Burnup Core <core ext> <core min> <core_avg>

<core max> -Fl 0 0 19 3 F2 117 20 92 122 F3 67 18 60 68 F4 142 142 130 145 F5 0 0 19 3 F6 117 20 92 123 F7 67 18 60 68 F8 142 142 130 144 Core total (MWD) 652 360 600 676

  • Depending on reactor operating conditions , especially CB insertion height , the flux level in the reflector region could appreciably change. During 2014 -2015 operations, the lowest startup and highest shutdown critical position was recorded at CB heights of 32.69 cm (12.8 7 in) and 61.6 cm (24.25), respectively (Reference 16). These two CB positions are bounding values that include various reactor perturbations such as Be reflector installation, CB replacement and adjacent experiment loading/unloading.

The CB movement during a 19-month period of operation in 2014 and 2015 is shown in Figure 38. The CB average age during this period is 4.7 to 5.7 years. In order to conservatively conduct the core and target power calculations , the limiting core configuration has been constructed to use a fresh Be reflector that will result in a higher neutron flux in the target region , in addition to the variations of driver fuel burnup. Page 68 of 190 ATTACHMENT 1

  • The CB age is tilted such that A and D are fresh and B and C are eight (8) years old , which will result in a higher power peaking in the TA.
  • The CB tip position is tilted so that the tip position of CB B and C is h i gher than that of A and D by 2.54 cm (I in), which results in a higher power peaking in the TA.
  • The central flux trap is loaded with sample materials.
  • Equi l ibrium xenon is used for the maximum and average bumup driver fue l s.
  • No xenon is used for the extreme and minimum bumup driver fuels. 70 60 -E 50 u -c .2 40 8. "8 30 ... 0 ... 20 ... c 0 u 10 0 0 10 20 30 40 50 60 7 0 80 90 Wee k s Fig u re 38 MURR Co nt ro l Bl ade Trave l betwee n Ja nu ary 13, 2 0 14 and Septembe r 15, 2015 Typica ll y, the CB travels over 10 cm (3.94 in) during the first two (2) days after startup and t h en slowly continues to withdraw. The driver fuel elements also have two (2) distinct states: clean fuel and equilibrium xenon. The xenon-free core bumup i s provided in Table I 0. The core bumup of an equilibrium xenon state core is approximately 20 MWD higher t h an that of the xenon-free core. Because the pellet enrichment is fixed at nomina ll y 19.75 wt%, the target cartridge design process searc h ed for the target rod position that sat i sfies the thermal design li mits of the target rod and cartridge d ur ing norma l operating conditions (acc i dent conditions presume operation at the a ll owable des i gn limits). T h e selected target rod position is 26.37 cm (10.38 in) from the core center. Page 69 of 190 ATTACHMENT 1 The critica l CB positions and expected CB travel range of the equilibrium core are summarized in Table 10 and Table 11. When the T As ar e loaded , the CB s are inserted deeper into the core. The estimated additional CB insertion depth is 3 to 9 cm (1.2 to 3.S in) depending on the core states. When these offset value s are applied to the l owest and highest CB position during 2014 to 201S operation , the lowest and highest cr iti ca l CB posit i on w i th a two (2) TA loading i s expected to be 33.7 cm (13.27 in) and Sl.9 cm (20.43 in), respective l y (Table 11). Here , t h e CB offset values of <core_e x t> and <core_max> were applied to BOC and EOC case s, respective l y , whi l e that of <core_avg> w a s approximate l y used for Day-2 CB position. Table 10 Critical Control Blade Positions Extreme Minimum Average Maximum Burnup Core Burnup Core Burnup Core Burnup Core Xenon state no xenon no xenon equi li bri u m equilibrium Regulating blade (cm) 2S.4 2S.4 38.l 38.l Critical CB position without 41.67 3S.09 S9.87 6S.OO target assembly Critical CB position with full (2 38.7S 32.7S S2.89 SS.28 fresh) target assemblies l oaded Control blade offset (cm) 2.92 2.34 6.98 9.72 Table 11 Expected Contro l Blade Travelling Range for Target-Loaded Core Recorded Critical CB Position without Expected Critical CB Range with Target Loading Target Loading Lowest Highest Lowest Highest BOC 36.6 42.6 33.7 39.6 Day-2 49.1 S8.8 42.1 Sl.8 EOC S3.6 61.6 43.9 Sl.9 5.2.2 Target Assemb l y Criticality A conservat i ve estimate of K eff for two (2) TAs was performed u s in g the MURR core mode l by replacing the driver fuel with water and assuming the control and regulating blades are completel y withdrawn from the core. The calculated va l ues of K eff for the target cartridge l ocated in reflector position No. SA (single TA), and No. SB (s ingle TA) are 0.6S868+/-0.00014 and 0.6S870+/-0.00018 , respectivel
y. When both SA and SB are loaded , the K eff is 0.6S87S+/-O.OOO 18 , indicating that the T As will be subcritical with a large margin for uncertainty.

The uncertainties are given in 2 standard deviations (2cr) as a 9S% confidence level. Page 70 of 190 ATTACHMENT 1 S.2.3 Impact of Target Assembly Loading on MURR Reactor Core L oading two T A s in the MURR g raphite reflector region c au ses perturbation s in neutronic and thermal p erformance of th e MURR cor e a s follow s:

  • The MURR core exces s reactivity will increase slightly , which will be compensated b y the reactivity control device s.
  • The neutron flux and power of the adjacent F5 fuel element will increase even though the total core power is maintained at 10 MW.
  • The thermal power generated by the target rods will not affect the reactor core primary or pool cooling capability.

The heat generated by the targets that is discharged to the reactor pool will be adequately removed by the reactor pool coolant s y stem without impacting the pool temperature. S.2.4 Reactivity Insertion The reactivity insertion due to target cartridges being loaded in graphite r eflector po s itions No. 5A and N o. SB is summarized in Table 12. The reactivity worth was calcu l ated b y replacing the water in the TA c artridge with fre s h , clean, cold target rods. Under the hot operating condition , the reactivity insertion due to a single cartridge loading is less than 0.31 %

The maximum reactivity insertion due to a hot t arget rod is the same as that of the cold target rod for all core bumup state s. Table 12 Kerr and Reactivity Insertion Va lue s for Single Target Assembly SA/SB filled Target in SA Target in SB with structure and water Kerr

(%) Kerr (%) (Kerr) Maximum with equilibrium 0.99368 0.99658 0.29 0.99660 0.29 xenon Average with equilibrium 0.99392 0.99673 0.28 0.99677 0.29 xenon Minimum without xenon 0.99385 0.99662 0.28 0.99690 0.31 Extreme without xenon 0.99382 0.99667 0.29 0.99685 0.31 S.2.S Driver Fuel Power Peaking The target cartridge loading in the graphite reflector positions has a relatively small impact on power peaking of the MURR core driver fuel. Table 13 compares the distribution of MURR driver fuel power w ith and without TA loading and provides the location of the largest peaking factors in the reactor core. F or most of core states , the fuel element FS (closest to the T As , see Figure 34 for driver fuel identification) peaking factor is the highest because the fuel bumup is the lowest and the fuel is located close to the TA. For the FS fuel element, the peaking factor of the inner plate is alway s higher than that Page 71 of 190 ATTACHMENT 1 of the outer plate. When two (2) TAs are loaded , the increase of peaking factor is less than 4% for the inner plate , while the maximum increase of peaking factor is 11 % for the outer plate. However , the absolute value of the outer plate peaking factor is always les s than that of the inner plate peaking factor. Table 13 MURR Core Maximum Power Peaking Due to Target Assemb l y Loading Inner Fuel Plate Outer Fuel Plate Number of Target Rods Axial Node 1 Peaking Axial Node Peaking (Fuel Elemen t) Factor (Fuel Element) Factor Maximum with 0 13 (F5) 2.541 13 (Fl) 2.064 equilibrium 22 11 (F5) 2.637 11 (F5) 2.218 xenon Average with 0 11 (F5) 2.426 12 (Fl) 1.999 equilibrium 22 8 (F5) 2.51 I I I (F5) 2.155 xenon Minimum 0 10 (Fl) 2.968 9 (F6) 2.444 without xenon 22 9 (F5) 3.031 9 (F6) 2.565 Extreme without 0 10 (F5) 2.905 9 (Fl) 2.200 xenon 22 10 (F5) 2.976 8 (F5) 2.397 1 Axial nodes I and 24 correspond to the bottom and top of MURR driver fuel plat e, equally space d. 5.2.6 Reactivity Coefficients Though the core power distribution and power peaking factors are altered when two (2) target cartridges are loaded in the graphite reflector region, the core reactivity characteristics do not change significantly due to target rods being loaded in the reflector region. The reactivity coefficients of the core with two (2) fresh T As were calcu l ated for the fuel temperature coefficient , coo l ant temperature coefficient and void reactivity and are summarized in Table I 4. The driver fuel temperature coefficient was calcu l ated by increasing the fuel temperature by 906.4 °C (1663.5 °F), i.e. from 20.44 °C (68.8 °F) to 926.84 °C (I 700.3 °F). The fuel temperature coefficients are consistent with core bumup state and kept negative.

Due to the lim itations in cross section data , however , the temperature dependence of lumped fission products (all fission products except 135 1, 135 Xe, 149 P m , 1 49 Sm) was not considered.

The coolant temperature coefficient was calculated by changing the coolant temperature from 20.44 °C (68.8 °F) to 76.84 °C (170.3 °F), where the coo lant density changes from 0.99815 to 0.97378 g/cm 3* The principal cross section data used for the calc ul ations are basically the same for both the lower and higher coo lan t temperature conditions, but the S(a,p) thermal scattering was correctly treated. The coolant void reactivity of the core was calculated b y removing 99.9% of coolant from the core. Page 72 of 190 ATTACHMENT 1 The statistical uncertainty (2o") of the reactivity coefficient is --0.026% when the error propagation rule is used. The least negative values are -1.03 x 1 o-6 and -1.69 x 10-4 for the fuel and coolant temperature coefficient , respectively. Negative reactivity coefficients of coolant void and temperature l imit the peak power during a reactivity transient.

Using the lowest magnitude of these negative reactivity coefficients for a positive reactivity i nsertion accident analysis is conservative because the subsequent power transient is larger than if nominal values were used. Tab l e 14 Reactivity Coefficients of the Reactor Core with Two Fresh Target Assemblies Core B u rnup F u el Tempe r at u re Coolant Temperature Coolant Void Coefficient Coefficient Reactivity State Maximum with -1.20 x 10-6 -1.69 x 10-4 -3.40 x 10-3 equilibrium xenon Average with -1.16 x 10-6 -1.73 x 10-4 -3.40 x 10-3 equilibrium xenon Minimum without -1.27 x 10-6 -1.85 x 10-4 -3.40 x 10-3 x enon E xtreme without -1.03 x 10-6 -1.79 x 10-4 -3.40 x 10-3 x enon 5.2.7 Reactivity Device Worth The impact of TA loading on the existing reactivity control devices (control and regulating blades) were estimated for their reactivity worth and subcriticality margin as summarized in Table 15. Reactivity worth of the regulating blade was calculated from fully inserted and fully withdrawn conditions while the CB is kept at its critical position.

For the selected core states , the lowest worth of the regulating blade is 3.41 x 10-3 The CB subcriticality margin was at first calculated for the average bumup core. Among four (4) CB's (A , B, C and D), the subcriticality margin of C is the lowest even though it is almost the same as that of B. The subcriticality margin increases as the core bumup increases due to effective neutron flux changes. The lowest subcriticality margin with the most reactive CB (i.e. A) and regulating blade fully withdrawn is 0.055 for the minimum bumup core. Page 73 of 190 ATTACHMENT 1 Table 15 Reactivity Control Device Worth with Two Fresh Target Assemblies Subcriticality Margin Regulating Blade Worth of Control Blade (Ak/k) (Ak/k) Maximum with equi librium xenon 0.109 3.55 x 10-3 Average with equilibrium xenon 0.106 3.41 x I 0-3 Minimum without xenon 0.055 3.69 x 10-3 Extreme without xenon 0.076 4.02 x io-3 5.2.8 Core Excess Reactivity The excess reactivity of the core with two (2) fresh TA s was calculated for the minimum burnup core which has the largest excess reactivity. All CB's and regulating blade are fully withdrawn from the core. The excess reactivity of the cold core (all at room temperature) is 0.072 Ak/k , while that of the hot operating core is 0.067 Ak/k. 5.2.9 Kinetic Parameters The kinetic behavior of the core is dominated by the driver fuel element s, which is slightly perturbed by the presence of two (2) T As in the graphite reflector region. The effective (or adjoint weighted) neutron generation time (A e rr) and effective delayed neutron fraction (f3 e ff) were generated for different burnup states of the core with and without (Reference

17) T As as summarized in Table 16. The neutron generation time tends to increases when the two (2) TA s are loaded , but very slightly.

The effective delayed neutron fractions of the core with and without T As are very close to each other within the uncertainty range (+/-2cr). From the viewpoint of point kinetics , the TA loading won't deteriorate the slope of power increase (or inverse reactivity period) during the reactivity induced transient.

Page 74of190 ATTACHMENT 1 Table 16 Kinetic Parameters of the Reactor Core with and without Target Assemblies Target Core B u rnup A.rr STD P.ff STD Load i ng State (µsec) (la) (l a) Maximum 62.3 0.273 0.00723 0.00015 No Target Average 60.6 0.278 0.00731 0.00015 Assemblies Minimum 53.7 0.244 0.00749 0.00016 Extreme 57.0 0.248 0.00732 0.00015 Maximum 62.7 0.237 0.00730 0.00013 Two Target Average 61.5 0.235 0.00745 0.00013 Assemblies Minimum 54.7 0.215 0.00766 0.00014 Extreme 58.4 0.220 0.00769 0.00014 5.2.10 Str u ctural Component Heating The st ructural component heating was estimated by MCNP6 code using the kinetic energy deposited by the fission fragments , prompt neutrons and delayed neutrons. In addition to the neutron data used for neutron flux calculation , the heating calculation uses photon-atomic data (mcplib04) (Reference 18), photon-nuclear data ( endf7u) (Reference 19), electron data ( el03) (Reference 20), and delayed neutron/gamma data ( cinder.dat , delay _library.dat , cindergl.dat , delay _li brary_ v2.dat) (References 21 and 22). MCNP6 adopts the CINDER'90 model to calculate delayed particle emissions from all of the radionuclides in the decay chain for a fission product. The photon emission time for delayed-neutron and delayed-photon emission was adjusted to 10 5 sec (the default value is 10 1 0 sec), under which the statistical uncertainty (1 cr) of the calculated heat deposited in the Be reflector is The Be reflector heating is higher for the maximum burnup core by 3.3% when compared with the minimum bumup core. For the maximum bumup core , the Be reflector heating is 216.2 kW and 205.7 kW with and without two (2) TAs, respectively.

The heat from the cartridge and neutron absorber shield is 12.7 kW and 7.6 kW respectively, when two (2) T As are loaded (Table 17). Page 75 of 190 ATTACHMENT 1 Table 17 Component Radiation Heating Due to Target Assembly Loading for Maximum Burn up Core Target Neutron Photon Electron Heating Total Component Heating Heating Loading (kW) (kW) (kW) (kW) No Target Be reflector 41.6 81.9 82.2 20S.7 Assemblies Be reflector 42.9 86.2 87.1 216.2 Two Target Al cartridge 0.27 6.2 6.2 12.7 Assemblies Steel shield O.OS 3.8 3.8 7.6 S.3 Target Assembly Flux and Power S.3.1 Fresh Target Assemblies in Both Positions No. SA and No. SB (Base Loading Pattern) The neutron flux in a TA is dominated by incoming neutrons from the reactor core even though the TA produces neutrons from fission reactions.

Figure 39 and Figure 40 show 4-group axial neutron flux distributions of the target rods 6 and 17 , i.e. the middle of each 11 target rod bundle (see Figure 36 for target rod numbering), when two fresh target cartridges are loaded in graphite reflector positions No. SA and No. SB , respectively. The upper energy boundaries of the 4-group are 20 Me V (Group 1 ), 0.1 Me V (Group 2), S.S3 keV (Group 3), and 0.62S eV (Group 4). The solid and hollow symbols in each figure indicate the MURR core states , i.e. <core_ext> and <core_max>, respectively.

Axial nodes 1 and 2S correspond to the bottom and top pellet , respectively. As can be seen , the neutron flux drops at the bottom and top section of the target rod. The neutron flux is suppressed even more in the top section due to the CBs, which is relaxed to a certain extent when the CBs are pulled out in the maximum bumup core. The thermal neutron flux is effectively flattened in the vertical middle section , from node S to 20 , due to the presence of the SS3 l 6L absorber.

Figure 41 shows neutron flux in the azimuthal direction , from target rod 1 to 11 (position No. SA) and from rod 12 to 22 (position No. SB), at axial middle plane. The neutron population is suppressed at the assembly edge region due to elongated neutron absorber material on the cartridge edge. For the average bumup core , the peak neutron fluxes of the target pellet are Page 76of190 [Sa , d, e , fl ATTACHMENT 1 Rod number 6

  • Group 1
  • Group 2 * * *
  • I 0 D
  • Group 3 <core ext> a * *
  • a *
  • Group4 a D ....... I/)
  • C'i a D
  • a E *
  • 0 a -core max> c a ........ D )( * * ;;:J
  • a [Sa , d , e , t] c a 0
  • ti * ::I * *
  • 8 *
  • Cl> *
  • z
  • 0 t * *
  • 0 ... *
  • t
  • i t *
  • 5 10 15 20 25 Ax ial n od e Figure 39 Target Rod 6 Axial Neutron Flux Distribution in Position No. SA R od number 1 7
  • Group 1 -* Group 2 -* * *
  • I D
  • Group 3 II u a <core_ ext> 11 *
  • I t D a
  • D
  • G r oup 4 a --D * -D -u
  • D
  • D I <core max> * * --* a D I a *
  • a [Sa , d , e , fl a
  • I ..
  • t * * * * *
  • t *
  • t * *
  • 0 ... n ... . -. .. ..
  • l .... -* ¥ lS
  • a 0 *
  • 0 & .. 0 *
  • 0 *
  • 9 ,. . . . . . . ... . ... g a 0 .. .. ¥ * * * ' i
  • w II
  • t
  • 0 5 10 15 20 25 Ax ia l node Figure 40 Target Rod 17 Axial Neutron Flux Distribution in Position No. SB P age 77 of 1 9 0 ATTACHMENT 1 Axial n o de 12 <core_ ext> * * * * * *
  • I! * *
  • a " " " a Ill B " a a a * "
  • a *
  • II <core_max>

a a "

  • B
  • a D *Group 1 [Sa , d, e , f] *Group 2 *Group 3 " Group 4 *
  • e e e 8 I I I l
  • I I I I * ! t * *
  • I 8 8 * * * * * *
  • 8 * * * * * * * * * * * *
  • 2 3 4 5 6 7 8 g 10 11 12 13 14 15 16 17 18 19 20 21 22 R od n umber Figure 41 Target Assembly Azimuthal Neutron Flux Distribution Calculated target power and pellet linear power when two (2) fresh T As are being irradiated are given in Table 18. Figure 42 and Figure 43 show the pellet linear power envelope of the extreme and maximum bumup cores respectively.

The axial power profile is initially bottom-peaked and gradually changes to middle-peaked shape as the regulating and control blades are withdrawn from the core. The distribution of pellet linear power is shown in Figure 44 for the four (4) different MURR core states. For the most probable operating core condition , i.e. the average burn up core , the peak pellet linear power and total TA power are , respectively. Table 18 Calculated Target Power Level and Linear Power Extreme Minimum Average Maximum Burnup Core Burnup Core Burnup Core Burnup Core Assemb l y Power (kW) .. .. .. .. Assembly in SA .. .. .. .. Assembly in SB .. .. .. .. Peak Linear Power (kW/m) * * *

  • Target Rod Number 18 20 22 22 Axial Node* 11 10 18 18 *A x ial node s I and 25 corr e spond t o bottom and top node , equally spaced. Page 78 of 190 [Sa, d , e , f] [Sa , d , e , f]

ATTACHMENT 1 x [Sa, d , e , f] 0 5 10 15 20 25 Axial node Figure 42 Power Envelope of the Base Target Loading for the Extreme Burnup Core Case [Sa , d , e , f] 0 5 10 15 20 25 Axial node Figure 43 Power Envelope of the Base Target Loading for the Maximum Burnup Core Case Page 79 of 190

  • <core_max>
  • <core_avg>

80 * <core_min> * <core_ext>

ATTACHMENT 1 * * * ... . * . ... 8. 60

... . ....... 'O * * * * *** .... Cl)

  • 40 +-------..------r...-----.-, ket---.. ... ...--t-------j
I z 0 * * * * .. -* -----------Pellet linear power (kW/m) Figure 44 Target Assembly Pellet Linear Power Distribution for Base Target Loading Case 5.3.2 Sensitivity to Control Blade Position for Full Target Assembly Loading [Sa , d , e , f] The peak linear power of the target rod i s dominated by the CB insertion depth. The sen s itivity o f peak linear power to CB position was estimated for the maximum , average , minimum and e x treme core bumup states. The CB position repre s ents the core response to the various reactivity perturbations seen b y the reactor core , including fuel depletion and material aging. Sensitivity calculations have been conducted for a wide range of CB positions beyond the expected minimum and maximum critical position for the target-loaded core. Jn this simulation , the regulating blade was fixed at its typical position: 2S.4 cm (I 0 in) for the extreme and minimum bumup core and 38. l cm (1 S in) for the average and maximum burnup core. lt should be noted that the CB movement i s synchronized with the regulating blade in real operation so that the actual operating range of CB will be narrower than the simulated one. The results are summarized in Table 19 for the total target power , peak target linear power and driver fuel peaking factors. The total target power is relatively low for the BOC condition , which is a relatively short term period. For the equilibrium fuel conditions (average and maximum burnup cores), the target power stays below-* The peak linear power of the target rod is maintained between for [Sa , d , e , f1 the simulated equilibrium core states and CB position.

The variations in total target power and peak linear power versus CB position are shown in Figure 4S and Figure 46 , respectively. The MURR driver fuel element power peaking occurs in fuel element FS. The driver fuel element peaking is higher for the BOC state when the CB and regulating blade are deep into the core. The calculated maximum peaking factor is 3.231 for the inner plate. For the outer plate that faces the target , Page 80 of 190 ATTACHMENT 1 the maximum peaking factor is 2.636 , which is lower than that of the inner plate for all core state and CB positions.

The variations of driver fuel element peaking factor vs CB position are shown in Figure 47. The effect of CB position has also been studied for the critical core configurations by changing the CB average age from 0 to 8 years. For all the critical core cases , the calculated total target power and the maximum linear power are less than those of non-critical core cases. For the extreme bumup core , the maximum linear power is for the non-critical and critical core , respectively , when [Sa , d , e , t] the average CB age i s the same (4 years) for both cases. The extreme burnup core becomes critical at CB height of32.28 cm (12.71 in) when the age of all the CBs is 8 years. The maximum linear power of the extreme burnup core with CB at -32 cm (12.6 in) is for the non-critical and critical core , respectively.

lt should be noted that the CB age tilt is 8 years for the non-critical case while it is 0 for the critical case , which reduces the neutron population in the TA to a certain extent; and both the total target power and the maximum linear power are reduced in the critical core when compared with the non-critical core. The critical core calculations have shown that the non-critical core calculations conservatively estimate the target power and the maximum linear power. However , it should also be noted that the critical core simulation here is limited by the range of CB age and its tilt and therefore , the variation of CB position is smaller when compared with the expected CB traveling range. This mean s that variation of CB age (as well as CB age tilt) is not sufficient to realistically model all the critical core cases. It is also true that in the actual core the CB position could be even higher or lower than those used for the critical core calculations due to Be reflector aging and other reactivity perturbations. Finally , it is also reasonable to assume that the target power and its axial shape are dominated by CB position. Therefore, the limiting core configuration has been selected from the non-critical core configurations.

The maximum linear power found from the sensitivity calculation is -for the extreme burnup core with the CBs positioned at 30 cm (11.81 in) and the regulating blade positioned at 2S.4 cm (10 in), [Sa , d , e , f1 which is used for the target thermal-hydraulics and CHF analyses.

The maximum TA power is -for the maximum bum up core with CB at 44 cm (17 .32 in) and regulating blade at 38.1 cm (1 S in), which is the recommended case for the TA cooling analysis.

Page 81 of 190 ATTACHMENT 1 Table 19 Target Assembly Power and Core Peaking Factors CB Position Target Power Peak Linear Inner Plate Outer Plate Core State Power (cm) (kW) (kW/m) Peaking Peaking 62 ..

  • 2.573 2.147 60 ..
  • 2.577 2.173 58 ..
  • 2.605 2.192 Maximum 56 ..
  • 2.622 2.208 burnup with 54 ..
  • 2.662 2.239 eq uilibrium 52 ..
  • 2.674 2.273 xe non 50 ..
  • 2.709 2.302 48 ..
  • 2.755 2.348 46 ..
  • 2.792 2.385 44 ..
  • 2.861 2.441 56 ..
  • 2.488 2.109 54 ..
  • 2.505 2.131 52 ..
  • 2.529 2.163 Average 50 ..
  • 2.575 2.192 burnup with 48 ..
  • 2.609 2.240 eq uilibrium xeno n 46 ..
  • 2.648 2.263 [Sa , d , e, f] 44 ..
  • 2.684 2.310 42 ..
  • 2.757 2.374 40 ..
  • 2.781 2.393 42 ..
  • 2.790 2.386 40 ..
  • 2.857 2.419 Minimum 38 ..
  • 2.898 2.453 burn up 36 ..
  • 2.944 2.489 without xenon 34 ..
  • 3.046 2.546 32 ..
  • 3.055 2.576 30 ..
  • 3.120 2.609 42 ..
  • 2.906 2.301 40 ..
  • 2.930 2.351 Extre me 38 ..
  • 2.991 2.403 burn up 36 ..
  • 3.053 2.475 without xenon 34 ..
  • 3.130 2.539 32 ..
  • 3.182 2.594 30 ..
  • 3.231 2.636 Page 82 of 190 ATTACHMENT 1 Equilibrium xenon case a a 0 a a 0
  • a 0 0 *
  • a 0
  • a * )( )( )( [Sa , d , e , fl No xenon case 30 35 40 45 50 55 60 65 CB position (cm) Figure 45 Variation of Target Power vs. Control Blade Position [Sa, d , e , fl 35 40 45 50 55 60 65 CB position (cm) Figure 46 Variation of Peak Linear Power vs. Control Blade Position Page 83 of 190

... 0 .. u "' -C'l c :.i "' Cll a. Cl) .. IV Q. Qj ::I -... QI > *c c 4.0 3.5 3.0 2.5 !l 2.0 f-1.5 1.0 0.5 0.0 30 I Inner plate (No xenon case) >t v x x j * *

  • I I Oute r p la t e 35 40 ATTACHMENT 1 I n ner plate (Equ ilibri um xe n on case) IS 0 0 a a ** C> 0 a ., " -! 0 0 I * * -* * * * * ,. * * * * * *
  • 45 50 55 60 65 CB position (cm) F igur e 47 Va r iat ion o f Dr iver F u e l E l e m e n t P ea kin g Fa ctor vs. Cont rol Bl a d e Posit ion The effect of regulating b l ade movement was assessed for the cr i tical core. Two (2) regu l ating blade positions , i.e. l ow (25.4 cm) and h i gh (38. l cm), were considered while the CBs were at t h eir critical [5 d fJ a, , e , positio n s. The resu l ts i n Table 20 s h ow that t h e pe l let peak l inear power and total TA p ower are respective l y. The maxim u m inner and o u ter plate peaking factors of the driver fuel are 3.043 and 2.576, respective l y, for the minimum bumup core. Page 84 of 190 ATTACHMENT 1 Table 20 Effect of Regulating Blade on Pellet Peak Linear Power and Target Assembly Power Regulating Target Peak Linear Inner Plate Outer Plate Core State Blade Power Power Peaking Peaking (cm) (kW) (kW Im) Ma x imum burnup 38.1 ..
  • 2.637 2.218 with equilibrium 3 l.7S ..
  • 2.641 2.147 x enon 2S.4 ..
  • 2.638 2.046 Average burn up 38.1 ..
  • 2.S 11 2.I SS with equilibrium 31.7S ..
  • 2.S42 2.093 xenon 2S.4 ..
  • 2.S28 1.973 [Sa , b , e , t] 38.1 ..
  • 3.037 2.S76 Minimum burnup 3 l.7S ..
  • 3.043 2.S63 without xenon 2S.4 ..
  • 3.031 2.S6S 38.1 ..
  • 2.988 2.S31 Extreme burnup 31.7S ..
  • 2.999 2.507 without xenon 2S.4 ..
  • 2.976 2.397 5.3.3 Uncertainties The neutronics analysis has an inherent uncertainty due to the solution method employed in the MCNP6 code. The standard deviation (l o) of the pellet power is -for a l most all the pellet numerical nodes [Sa , b , e , f1 when 100 million particles are used for the calculation.

For the total target power , the s tandard deviation ( 1 o) is as small as -* Other uncertaintie s considered are impurities of constituent materials , p e llet density , fissile content and target rod position due to manufacturing tolerance.

Table 21 shows impu rities of of the target pellet , c laddin g , cartridge and neutron shield. For the aluminum (cartridge) and stainles s teel (neutron shield), the impurity was assumed to be 10 part per million (ppm) n atural boron. Table 21 Material Impurities for Uncertainty Analysis Uranium Zircaloy-4 Aluminum Stainless Steel (wt%) (wt%) (ppm) (ppm) 232 u 2 x 10*1 NIA N I A NIA 234 u 0.26 NIA N I A N I A 23 6u 0.46 NIA N I A NIA Boron 3 ppm o.oooos 10 IO Page 8S of 190

ATTACHMENT 1 The uncertainties of the peak linear power and total target power were e s timated for the four (4) critical cores. In order to estimate the sensitivity of the core and target performance parameter s, the uncertaintie s of design parameters were defined as fo ll ows:

  • For the statistical uncertainty of the solution method , +/-2a value is used to estimate the e i genvalue , tota l target power and peak linear power wit h a 9S% confidence level.
  • The impurity always reduces the target power and is regarded as a bias. The uncertainty is estimated as 9S% of the performance parameter change due to the maximum impurity level of the target uranium , assuming a flat distribution of impurity between its minimum and maximum values.
  • The manufacturing tolerances of the pellet density and fissile content are. and
  • wt%, [Sa , b , e , fl respetively. The target performance is relatively insensitive to the variation of these parameters. For conservatism, the tolerance value is taken as the uncertainty.

In this simulation , only a positive value is used as a bias to estimate the power increases.

  • The uncertainty of the target rod position is -, which is 2/3 of the manufacturing/

installation tolerance-' i.e. approximately a 2a value when a triangular distribution is assumed between the minimum and maximum values with the mode (average) at 0. The -[Sa , b , e , fl means that the rod is closer to the core by -* Tho u gh the target rod could be either closer to or farther from the core , -is used as a bias to estimate the power increase of the target. The estimated uncertainties (positive components) of the core eigenvalue , peak linear power and total target power due to the statistical and manufacturing uncertainties are summarized in Table 22 , Table 23, and Table 24 , respectively.

The impact of the pellet fabrication density and fissile content uncertainty on the core eigenvalue and target peak linear power is very small , which makes it difficult to obtain consistent results by direct perturbation calculations. Therefore , l arge manufacturing uncertainties

  • [Sa , b , e , fl were used for the direct calculation and the results were l inearly interpolated. The uncertainty (positive variation) of the eigenvalue due to manufacturing tolerances (pellet density and fissile content) is estimated to be less than 11 per cent mille (pcm). The impact of target rod position uncertainty mostly dominate s the tota l uncertainty for all core states. For the peak linear power , the statist i ca l uncertainty prevails over uncertainties of the pellet density and fissile content, while it is comparable to the target rod position uncertainty.

Arithmetic summation of these uncertainties , excluding the impurity effect, resu l ts in a tota l uncertainty of-for the peak [Sa , b , e , fl linear power in the extreme burnup core, which is very conservative.

Specifica ll y , for the extreme burnup core , the u ncertainty due to the most dominant contributor , i.e. rod position uncertainty , is -which is -of the peak l inear power. For the total target power , the statistical uncertainty is relatively small , whi l e the target rod position uncertainty dominates the total uncertainty. The estimated maximum increase of target power due to [Sa , b , e , fl uncertainties associated with manufacturing tolerances (pellet density and enric h ment) is -(for the Page 86 of 190 ATTACHMENT 1 maximum burn up core), which is 1.1 % of the total power. The effect of target rod position uncertainty is 1.S% of the total target power for the maximum burnup core.

  • The total uncertainties of the key perfonnance parameters were estimated as a product of statistical error and root-mean-square (RMS) of uncertainties due to fabrication density , enrichment , and target rod position.

Here , the uncertainty due to impurity is neglected for conservatism. For the limiting core configurations , the total uncertainty (excluding the impurity effect) is applied to the target design power such as: where P and Po are power with and without uncertainty, respectively.

Subscripts s , d , e , and p refer to the uncertainties due to statistical , density, enrichment , and position , respectively.

  • For the eigenvalue , the estimated uncertainty is -for the maximum burnup core.
  • For the total target power , the estimated uncertainty is -for the maximum burnup core. The estimated upper bound for the total target fission power is then -when considering

[Sa , b , e , f1 uncertainties (cf. -without uncertainty).

  • For the peak linear power , the estimated uncertainty is -for the extreme burnup core. The estimated upper bound for the peak linear power is then -when considering uncertainties (cf. -without uncertainty). Table 22 Uncertainties in Core Eigenvalue (Bias) Core Burnup Statistical Impurity Fab r ication Fissile Target Rod Uncertainty Density Content Position State (pcm) (pcm) (pcm) (pcm) (pcm) vMaximum with * *
  • I
  • equilibrium xenon Average with
  • *
  • I
  • equilibrium xenon Minimum without
  • I I I
  • I I I
  • xenon Page 87 of 190 [Sa , b , e, t]

ATTACHMENT 1 Ta bl e 23 Uncertainties in Peak L i near Power (Bias) Co r e Burnu p Sta t istica l Im pu r i ty Fa bri ca ti on F i ss ile C on te n t Ta r get Rod S t a t e Uncertai n ty (%) De n s i ty (%) Po s iti on (%) (%) (%) Maxim u m with * -* *

  • equilibrium xenon Average with * -* *
  • equi l ibri u m xenon Minim um w i thout * -* *
  • xenon Extreme without * -* *
  • xenon Ta bl e 24 U n ce r ta in ties in Target Power (Bias) C or e Burnu p S t atistica l I m p uri ty Fab r icat i o n F i ss il e Ta r get R od U n certa in ty De n sity C ont e nt P o s iti on S tat e (%) (%) (%) (%) (%) Maximum with * -* *
  • equilibri u m xenon Average with * -* *
  • equilibrium xenon Minimum without * -* *
  • xenon Extreme without * -* *
  • xenon 5.3.4 N om inal Operat i ng Cyc l e (Staggered Loading Pattern) The nomina l operating cyc l e will load and remove 1 1 target rods per week for processing. In a nominal operating cycle , one (1) fresh TA (11 rods) is loaded in one grap hi te reflector position (No. SA or No. SB) [Sa , b , e , f] [Sa , b , e , f] -[Sa,d , e,f] . However , after a operat i on , a second TA is loaded i n the other reflector position.

T h erefore , both reflector positio n s are fully occupied during nominal operation with target rods. The difference between nominal operations and operations with a ll fresh targets is the bumup states of TAs: a fresh TA is in one reflector position and I-week-burned TA is in the other reflector Page 88 of 190 ATTACHMENT 1 position (i.e. staggered loading pattern). Therefore , it i s expected that the TA power and the impact on the MURR driver fuel elements of the nominal , or stagge red , loading s are always les s severe when compared with the fresh target rod loading case. Figure 48 and Figure 49 show 4-group axial neutron flux distribution s of the target rods 6 and 17 , respectively, when a fresh TA i s loaded in reflector position No. SA of the reference and extreme burn up core. The axial flux shape of the staggered target loading is almost the same as that of the fresh target loading. Figure SO shows the azimuthal flux variation on the axial middle plane, which shows a s light decrease of neutron flux level in position No. SB. Cil r-i E .2 c -)( ::i ;::: c: e 5 Cl) z I R od number 6

  • Group 1
  • Group 2 -* 0 *
  • I a
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  • l!I i1 I a <core_ext>
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  • 11
  • D -* g --I *
  • g g II *
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  • I * * .... 6 6 ... '"' . ' * & " I * . . t -* "
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  • 0 .. 0 & * * . .... .... .... & *
  • 9 '¥' '¥' T T T * * * *
  • I 8 * *
  • i *
  • l 0 5 1 0 15 20 Ax ia l nod e Figure 48 Target Rod 6 Axial Neutron Flux for Staggered Loading Case (fresh assembly in position No. SA) 25 [Sa , d , e , f] Fo r the nominal loading with a fresh TA in position No. SB (a 1-wee k-burned assembly is in position No. SA), the 4-group axial neutron flux distributions of target rods 6 and 17 are shown in Figure S l and Figu re S2, respectively. The azimuthal fluxes are shown in Figure S3. Page 89 of 190 ATTACHMENT 1 Ro d numb e r 17 * * *
  • D *Gro up 1 * *
  • I
  • Grou p 2 g a a a a D
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  • 5 1 0 15 2 0 2 5 Axial node Fig u re 49 Target Rod 1 7 Axia l Neutro n Fl u x for Staggered Loading Case (fresh assemb l y in position No. 5A) Axia l node 1 2 * * <co r e_ext> *
  • II
  • a a a a * * * * * * *
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  • G r oup 2 [5a , d , e , f] :s
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  • z 0 I I 0 * * * * * * * * * * * * * * * * * * * *
  • 2 3 4 5 6 7 8 9 1 0 11 12 13 1 4 15 16 17 1 8 1 9 20 2 1 22 Rod n um be r F i g ur e 50 Target Assembl y Az imu t h al Ne u tro n F lu x for Staggered Loa din g Case (fres h assemb l y in position No. 5A) Pa g e 90 of 190

..... Cl! N E 0 c: -x :::J c: c: -:::J a> z Cii l'i E c: -x :::J c: c: e -:::J a> z ATTACHMENT 1 Ro d num be r 6 *

  • G r o up 1 * * * !! a D Cl *G r o up 2
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  • 5 10 1 5 2 0 Ax i al node Fig u re Sl Target Rod 6 Axial Neutron Fl u x for Staggered Loading Case (fresh assembly in position No. SB) Rod numb er 1 7
  • Gro up 1
  • Gro up 2 * * . " . . a o
  • Group 3 <co re_e xt> *
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  • i 5 1 0 1 5 2 0 Axial node Fig u re S2 Target Rod 17 Axial Ne u tron F lu x for Staggered Loading Case (fresh assembly in position No. SB) Page 9 1 of 190 Cl * .. a [5a , d , e , f] Ill 25 [5a , d , e , f] 2 5 ATTACHMENT 1 Axia l node 12 <core_ext> * * *
  • I D * *
  • D D * * * *
  • D D
  • D D D D *
  • D <core_max>* D
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  • 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 Rod numbM Figure 53 Target Assembly Azimuthal Neutron Flux for Staggered Loading Case (fresh assembly in position No. SB)
  • D [Sa , d , e , t] t . 0
  • 22 TA power and rod linear power of the staggered TA loading is summarized in Table 2S when the fresh assembly is in graphite reflector position No. SA. The pellet linear power envelopes of the staggered loading are shown in Figure S4 and Figure SS for the extreme and maximum burnup core, respectively. The distribution of pellet linear power is shown in Figure S6. The axial power profile of the staggered loading is the same as that of the fresh target loading , but the overall linear power is slightly reduced in the burned assembly.

For the most probable operating condition, i.e. average burnup core, the power of the TA in position No. SB is lower by 11 kW when compared with that of the TA in position No. SA , which is attributed to target fuel depletion. The total TA power (position s No. SA and No. SB) is slightly lower than that of the fresh TA loading by 4%. Page 92 of 190 ATTACHMENT 1 Table 2S Calculated Target Power Level and Pellet Linear Power for the Staggered Loading Core Burnup Target Power Staggered Loading Staggered Loading State (fresh target in SA) (fresh target in SB) Assembly Power (kW) Assembly in SA .. .. Maximum with Assembl in SB equi libriu m Peak Linear Power (kW/m) xenon Target Rod Number 6 22 Axial Node 16 19 Assembly Power (kW) Assemb l y in SA .. .. Average with Assemb l in SB [Sa , d , e , f1 equilibrium xenon Peak Linear Power (kW/m) Target Rod Number s 22 Axial Node 16 18 Assembly Power (kW) Assembly in SA .. .. Minimum Assemb l in SB witho ut xenon Peak Linear Power (kW/m) Target Rod Number 6 22 Axial Node 8 10 Assembly Power (kW) Assembly in SA .. .. Extreme without Assemb l in SB xe non Peak Linear Power (kW/m) Target Rod Number 7 20 Axial Node 11 12 Page 93 of 190 ATTACHMENT 1 5 10 15 20 25 Axial node Figure S4 Power Envelope of the Staggered Loading for the Extreme Burnup Core Case (fresh assembly in position No. SA) 25 Axial node Figure SS Power Envelope of the Staggered Loading for the Maximum Burn up Core Case (fresh assembl y in po s ition No. SA) P age 94 o f 1 9 0 [Sa , d , e , f] [Sa , d , e , fJ ATTACHMENT 1 100 * <core_max>

  • <core_avg>

80 * <core_mi n> ... ... II> ... <core_ext>

  • -* * . ... * * * ! 60 .... * *
  • 0 ... .8 40 E ::I z 20 0 Pe lle t li near po wer (kW/m) Figure 56 Target Assembly Pellet Linear Power Distribution for the Staggered Loading Case (fresh assembly in position No. SA) [5a , d , e , f] The pellet linear power envelope s of the staggered loading with a fresh TA in position No. 5B are shown in Figure 57 and Figure 58 for the extreme and ma x imum bumup core , respectively , and their distribution is plotted in Figure 59. The total TA pow e r of the staggered target loading with fre s h target rod s in position No. 5A and No. 5B are almost the same. The po w er from the TA in position No. 5B is higher b y 1 8.6 kW when compared with that from the TA in position No. 5A. Page 95of190 ATTACHMENT 1 Axial node Figure S7 x' x 25 Power Envelope of the Staggered Loading for the Extreme Burn up Core Case (fresh assembly in position No. SB) 0 5 10 15 20 25 Axial node Figure SS Power Envelope of the Staggered Loading for the Maximum Burn up Core Case (fresh assembly in position No. SB) Page 96 of 190 [Sa , d , e , fl [Sa , d , e , t]

ATTACHMENT 1 100 r * <core_max>

  • <core_avg>

80 * <core_min> ... <core_ext>

... ** ..... ** ... Qi 60 Q. 0 ... Cll .c 40 E ::I z 20 0 Figure 59 Target Assembly Pellet Linear Power Distribution for the Staggered Loading Case (fresh assembly in position No. SB) 5.3.5 Target Assembly Flux and Power of Partial Target Assembly Loading [Sa, d , e , t] The number of target rods loaded in a TA may be less than 11 during commissioning.

For that purpose , the neutron flux and power distribution of the TA were simulated for the case of a 3-target rod loading for each 11-rod cartridge.

Target rod positions 5 to 7 and 16 to 18 were selected to maintain the symmetry. Other rods positions are loaded with filler rods to avoid excessive neutron thermalization while maintaining the nominal flow conditions. The 4-group axial neutron flux distributions of target rods 6 and 17 are shown in Figure 60 and Figure 61 , respectively. The azimuthal fluxes are shown in Figure 62. TA power and pellet linear power are s ummarized below in Table 26. The pellet linear power envelopes of the extreme and maximum burnup core are shown in Figure 63 and Figure 64 , and their distribution is plotted in Figure 65. Page 97 of 190 ATTACHMENT 1 Table 26 Calculated Ta r get Power Level and Pellet Linear Power for a Reference 3-Target Rod Partial Loading Extreme Minimum Average Maximum B u rn up Core B u rnup Core Burn u p Core B u rnup Core Assembly Power (kW) .. Assembly in SA * * *

  • Assembly in SB * * * * [Sa , d , e , f] Peak Linear Power (kW/m) Rod Number 18 18 18 18 Axial Node lS 10 lS 17 Rod number 6
  • Group 1 *
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  • 3 8 s * * * * -----+---5 10 15 20 25 Axial node Figure 60 Target Rod 6 Axial Neutron F l ux for Partial Loading Case Page 98 of 190 ATTACHMENT 1 "T Rod n u mber 17
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  • T 5 10 15 20 25 Axial node Figure 61 Target Rod 17 Axial Neutron Flux for Partial Loading Case l Axial node 12 <core_ext> * * *
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  • T -r ,-T T r T 2 3 4 5 6 7 8 Q 10 11 12 13 1 4 15 16 17 18 HI 20 2 1 22 Rod number Figure 62 Target Assembly Azimuthal Neutron Flux for Partial Loading Case P ag e 99 of 1 90 ATTACHMENT 1 i f i
  • I + * * *
  • i *
  • I [Sa , d , e , f1 0 5 10 1 5 20 25 A x ial no de Figure 63 Power Envelope for Partial Loading for the Extreme Burnup Core Case * * [Sa , d , e , f1 0 5 10 15 20 25 Axial no de Figure 64 Power Envelope of the Partial Loading for the Maximum Burn up Core Case Page 100of190 ATTACHMENT 1 [Sa , d , e , f] Pellet l i near power (kW/m) Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case 5.3.6 Target Assembly Material Depletion As the fissile material bums in the target rods from irradiation by the incident neutron flux from the reactor core, the TA power steadily decreases. Table 27 l ists the isotopic mass of two (2) T As for the average bumup core , which is the most likely core state. As the fissile uranium depletes , the TA power also decreases linearly as shown in Figure 66. At the end of a -irradiation , the TA power drops to-* which corresponds to a 1.7% reduction when compared to the initial power of -for the commercial loading. The target peJlet bumup is [Sa , d, e, f1 -Megawatt day per ton heavy metal (MWd/tHM).

Material Content and Burnup of the Base Target Loading for Average Burnup Core State Calendar Operation Burn up 23su 23su 239Pu Day (MWd/tHM) (grams) (grams) (grams) I -I 461 1,874 0 * -.. 4S7 1 , 874 0.21 [Sa, d , e , f] I -.. 4S7 1 , 874 0.21 .. -.. 4S3 1 , 874 0.42 Page 101 of 190 ATTACHMENT 1 [5a , b , e , f] Operation time (day) Figure 66 Change of2-Target Assembly Power for-Operation

5.3.7 Summary

of Target Assembly Nuclear Design Evaluation The nuclear performance of the TA has been evaluated using a single target pellet enrichment of 19. 75 wt% and an 11-rod target loading per cartridge housed in the TA. The evaluation results of the baseline TA model are as follows:

  • Two stand-alone target cartridges together are sufficiently subcritical , with a K eff below 0.659.
  • The MURR driver fuel element power peaking factor is Jess than 3.27 when two (2) fresh TAs are loaded.
  • The variation of target pellet linear power and total TA power due to CB movement is small. For the equilibrium core conditions , the peak linear power and total TA power change from --

and from respectively , when the CBs travel [5a , d , e , fl from position 46 cm (18.11 in) to 62 cm (24.41 in).

  • The uncertainty due to pellet fabrication does not exceed 0.0001 .'lk/k due to either a-* -.. [5a , d , e , f] The uncertainty due to mechanical tolerance of the target rod loading is and -for the peak linear power of the extreme bumup core and total target power of the maximum burnup core , respectively.

Page 102of190 ATTACHMENT 1

  • The peak linear power and total target power for operations with the staggered target loading or partially loaded target cartridges ar e les s than those of the full fresh loading of target rod s in the cartridge. Nuclear compatibility of the target cartridges and assembly with the MURR core has been assessed based on applicable MURR Technical Specifications.

In general , because the thermal power from the two TAs is relatively small of the nominal MURR core thermal power) and the target rods are loaded in the g raphite reflector region outside the Be reflector , the effect of target rods on the core reactivity characteristics is small as summarized below for representative core burnup states with the base target loading:

  • The least negative reactor core temperature coefficient of reactiv i ty is -1.69 x I 0-4 which is more negative than Technical Specification 5.3.a value of -1.08 x 10-4 (-6.0 x 10-5 The primar y coolant temperature range in this calculation was from 20.44 °C (68.79 °F) tO 76.84 o c (170.31 °F).
  • The reactor core void coefficient of reactivity is -3.4 x 10-3 which is more negative than Technical Specification 5.3.b value of -2.0 x 10-3 The primary coolant density in the calculation was changed from 100% nominal to 0.1 %.
  • The lowest subcritical i ty margin of the core with the most reactive shim blade and regulating blade fully withdrawn is 0.055 which is greater than Technical Specification 3.1.e value of 0.02
  • The excess reactivity of the cold , clean minimum burnup core with cold , fresh , base TA loading is 0.072 which i s less than Technical Specification 3.1.a value of 0.098 referenced to the cold , clean critical core.
  • A single target cartridge insertion into the MURR graphite reflector regions adds 0.31 %

to the MURR core , which is lower than allowed reactivity worth of each secured removable experiment , 0.6% (Technical Specification:

3.1.a). 6. Target Assembly Thermal Hydraulic Design Analysis This section provides information that demonstrates that adequate cooling capacity is available to keep the TA in a thermally safe condition during all operational states. The TCS design is described in detail in Section 3. For 10 MW 1 reactor power , the peak linear power , including margin for uncertainties , is -' which corresponds to a local target cartridge power density of-. The target rod with this power [5 a , d , e , fl density is located in the middle of the cartridge , and has an average power density of-. The two (2) T As generate -for the conditions that yield the highest peak linear power. Page 103 of 190 ATTACHMENT 1 The highest total power target rod is located at the edge of the cartrid g e. It has a peak linear power , including margins for uncertaint y, of-, and an average power density of-, with a [5 a , d e , f] local maximum of-. The two (2) TAs generate -for the conditions that yield the maximum fission power. Both the peak power and the maximum total heat cases were investigated in these analyses to ensure safe operations.

In addition to the power generated in the target rods , are produced directly in the cartridge itself (Table 17). Thi s heat i s also absorbed by the target cartridge rod cooling [5 a , d e , f] water and removed by the TCS. The maximum local target rod powers -at peak linear power and -at maximum [5 a , d , e , f] power) are used to determine the local peak temperatures and heat fluxes; the cartridge power is used to determine the total coolant heat absorption and hence , flow rate and outlet temperatures.

The pellet with the peak linear heat rate ha s that peak approximately 30% of the rod length above the coolant entrance , at a point where 35.4% of the target rod heat is generated.

The target rod with the highest total power has its maximum power pellet approximately 46% of the rod length above the coolant entrance, at a point where 47.9% of the rod heat is generated.

Analyses were performed for both target rods at their point of maximum heat generation rate , at the local water temperature and pressure.

Margins were added to account for uncertainties in flow rate , pressure drop and temperature measurements.

All analyses were performed assuming one (1) day of burnup ) which is the bumup [5 a , d , e , f] experienced for the maximum power density pellet (greater bumup rates increase the relocation of the pellet , with a resultant reduction in the pellet/clad gap , which results in colder U0 2 temperatures).

It has no effect on the CHFR. Varying burnup rates were considered in the transient analyses performed on the structure.

Steady-state , one-dimensional , axisymmetric analyses were performed at the location of the maximum U0 2 pellet power density. The effect of reactor power (10 MW, nominal and 11.5 MW 1 maximum), coolant flow rate (100% and 85% of nominal), pellet/cladd i ng gap 50 µm nominal and [5 a,d , e , f] ), cartridge channel diameter (14.12 mm minimum , 14.45 mm nominal , and 14.82 mm maximum) and the number of active target rods per cartridge (11 maximum , 3 minimum) were investigated to verify that all system temperatures , as well as the CHFR , were in safe regimes. 6.1 Thermal-Hydraulic Design Basis 6.1.1 Normal Conditions of Operation The mechanical design , dimensions and tolerances for the target rods are given in Table 3. The nominal (cold) gap between the pellet and the cladding is 50 µm. The minimum/maximum (cold) gap is 36/64 µm , respectively. The pellet/clad gap is filled with helium at I atm pressure when cold. Page I 04 of 190 ATTACHMENT 1 The target rods are placed in a scalloped cartridge which allows " annular" flow axially along the length of the rod. The target rod spacing is 11.63 mm , center to center , and the nominal cartridge flow " full" hole size is 14.45 mm. Thus , on each side of the rod , the " annulus" defining the flow along the target rod shall be open and intersecting the adjacent scallop over an angle of 72.77° for the nominal geometry.

This gives a cartridge flow channel with a hydraulic diameter of 10.01 mm. For the minimum channel diameter of 14.12 mm , the hydraulic diameter will be 9.62 mm. For the maximum channel diameter of 14.8 mm , the hydraulic diameter will be 10.02 mm. (Channel diameter variations are the result of manufacturing processes as well as cartridge pressure during operation.)

The target cooling water will flow through the cartridge at a flow rate that will have a nominal heat rise of 10 °C (18 °F). For a cartridge with 11 target rods , a target water mass flow rate of [5 a , d , e , f] -will produce a nominal fluid velocity of 5.20 m/sec as it flows through the section of maximum heat flux. In order to minimize the thermal impact on the reactor pool , the nominal target coolant outlet temperature is set to be the same as the pool water temperature , nominally 102 °F (38.89 °C). Thus , the nominal target cooling water inlet temperature is 28.89 °C (84 °F). The secondary coolant system , which is common to the MURR primary and pool coolant systems and the SGE TCS , will ensure that even on the hottest summer days , the required target inlet temperature is achieved. 6.1.2 Subcooled Nucleate Boiling For both the peak linear power target rod and the maximum total power rod , the nominal heat flux at the surface of the Zircaloy-4 cladding will be approximately

-' which will cause subcooled

[5 a , d , e , fl nucleate boiling at the cladding wall surface. To examine the boiling effects and ensure safe target rod operation , multiple boiling convection models using the ANSYS FLUENT computational fluid dynamics (CFD) computer code were developed (Reference 23). In addition , a literature search was carried out to find relevant experiments to validate the analyses.

The closest case found in the literature is that of Del Valle and Kenning (Reference

24) who studied subcooled nucleate boiling on flat plates under forced convection.

For their experiments , they utilized water at 116.7 kPa (16.9 psia), with subcooling between 24 and 84 °C , and coolant velocities between 0.8 and 2.0 rnlsec , at heat fluxes up to 4.6 MW/m 2* The 84 °C maximum subcooling in their experiments matches the target system well (82 -88 °C), while the coolant pressure and velocity are lower. Therefore , their results should be conservative.

Del Valle and Kenning's data show heat flux rising faster than other boiling correlations , such as Lottes and Chen (Reference 25). The Jens-Lottes correlation , used in the following thermal analysis , is an empirical correlation for fully-developed subcooled nucleate boiling , whereas the Chen model is a broader correlation which combines single phase forced convection and surface boiling effects. The Chen model is used in RELAP5 for nuclear thermal-hydraulic calculations.

Del Valle and Kenning give several important observations from their photographic study of the bubble formation in their experiment.

The photographic study was done for the 84 °C subcooling , 1. 7 mis coolant velocity condition.

These are summarized below: Page 105 of 190 ATTACHMENT 1

  • At high subcooling, the flow remains in a bubbly regime up until burnout. "All runs at 84 °C subcooling remained in the bubbly flow regime up to burnout , the bubbles retainin g their identities except for occasional coalescences leading to small , short-lived vapour patches."
  • The bubbles formed are small. " The maximum diameters were normall y distributed (class interval 0.1 mm) with a mean value of 0.4 mm -independent of heat flux." [5 a , d , e , t]
  • The bubbles are short-lived.

" Even at the maximum camera speed [of I 0 , 000 frames per second], individual bubbles appeared in only 3 -5 frames." These observations indicate that while bubble nucleation is expected , the bubbles are small and collapse quickly. This agrees with an earlier publication by Unal (Reference 26), who writes , " The bubbl e s form e d at high sub c ooling do not l e av e th e h e at e d surfac e, although th ey attain th e ir maximum diam e ter. Th ey s lide alon g th e h e at e d surfa ce or c ollap se , a s ob se rv e d vi s uall y b y Gunth e r." Therefore , it is expected that while some subcooled nucleate boiling will occur in the high heat flux regions of the target section , there wi ll be minimal vapor generation , and given the expected bubble lifetime and high subcooling , there will be no vapor present in the coolant exiting the target. The correlations above do not make predictions on vapor formation.

To corroborate the work of Del Valle and Kenning , ANSYS FLUENT CFO models were used to examine the presence of vapor in the target coolant. These models used an Eulerian multiphase model to track the balance of liquid water and vapor throughout the fluid domain of the target housing. The heat transfer at the wall was determined by what is known as the " RPI model ," developed at Rens s elaer Pol y technic Institute (RPI). The RPI model partitions the heat flux into boiling and convective effects , much like the Chen correlation (Reference 27). The results from these CFO models were in line with the observations of Del Valle and Kenning , in that the presence of vapor was confined to the cladding rod surfaces and that the total amount of vapor generated was very small. The resulting vapor volume fractions at the maximum heat flu x rod in the worst case operating conditions are shown in Figure 67. The graph shows that the maximum local vapor volume faction is about 5 parts per million. This i s lower than the I 00 parts per million shown in Del Valle and Kenning's experiments , although the difference is reasonable given that the coolant pressure and velocity are higher in the target (the latter by nearly a factor of 3). Additionally , the results show that all vapor bubbles collapse outside of the high heat flux region at the rod s urface , and water exiting the target housing contains no vapor. Note that the large downward spikes around 0.36 m are due to the presence of a structural support pin , which causes local flow acceleration that depresses the vapor generation.

Page 106of190 6 *e ta S.OOE-06 4.00E-06 ol: 3.00E-06 GI E ::s 15 > 2.00E-06 5 g,, 1.00E 06 0.1 0.2 ATTACHMENT 1 0.3 0.4 0.5 0.6 Axial pos iti o n (m) Figure 67 .....-Front path --Back path -.-Side path 0.7 0.8 Vapor Fraction at Cladding Wall for Worst-Case Conditions in FLUENT RPI Wall Boiling Model Both the CFO modelling and Del Valle and Kenning's work give compelling evidence to show that vapor fo rmation in the target is minimal and confined only to the highest heat flux regions of the target rods. To confi rm this evidence , experiments are planned to replicate the exact conditions that the target rods will experience in the MURR graphite reflector region , and provide data regarding CHF and subcooled nucleate boiling behavior , which s hould confirm that two-phase flow is not a concern in the target. The TA thermal design and analysis ensures that vapor bubbles do not coalesce together forming a vapor fi lm at the cladding surface. This phenomenon occurs when the surface heat flux reaches the CHF value at the transition from nucleate boiling to partial film boiling (Reference 28). Three correlations , Bernath (Reference 29), Macbeth (Reference

30) and Groeneveld (Reference
31) have been used to assure that the ta rget operates in a safe heat transfer regime. The Bernath correlation is very conservative (in predicting low critical heat fluxes) and is an industry-wide standard for such calculations.

The Macbeth correlation i ncorporates data at the low pressures of this system , and it is one of the correlations recognized in the NRC light water reactor fuel rod analysis code FRAPTRAN. The Groeneveld correlation , a lookup table based on the most extensive set of data available , is generally considered the most accurate.

All three (3) predictions were found to be in close agreement.

The l owest CHF of the three (3) correlations (in every case , Bernath) was used to calculate the CHFR , defined as the CHF divided by the actual heat flux. Recognizing the large amount of scatter and uncertainty in CHF data , a CHFR > 2.0 is required and has been assigned for the design. With a subcooling margin greater than 80 °C , the TA is normally operating at the transition point from convection to nucleate boiling , well below the value of heat flux where transition from nucleate to partial film boiling occurs. Page 1 07 of 1 90 ATTACHMENT 1 The thermal analysis first calculates the s in g l e phase convective h eat transfer from the cla ddin g surface. If the s urfac e temperature is hi g h er than th at re quir ed to initiate subcooled nucl eate boiling per the Jens Lottes co rrelati on (Reference 25), then the wate r at the s urface i s assumed to b e boiling. Its s urface temperature i s set at t h e Jen s-Lottes s u ggested va lu e, and the CHFR i s calcu l ated. Figure 68 s ho ws t h e CHF R , h ea t flux and e nthalp y as a function of the axia l lo catio n in the target rod for both the p eak linear power (PLP) rod and the maximum heatin g (MH) rod. a: LL :I: (.) Axial Location, cm Figure 68 .e. ftl c w c ftl 0 0 (.) ... 0 >< ::::> u::: -ftl :! [5 a, d, e, f] LEGEND -*-*-*-PLP He* A MW11n't -*-*-*-MH He ot Aux. MW hn't PLP Bernath CHFR MH Bernath CHFR ---PLP Groeneve'd a-iFR ----MH Groeneveld CHFR PLP Macbet h CHFR MH Macbeth CHFR t -uses ri!Jit Y ax i s Heat Flux and CHFR as a Function of Axial Location for Peak Power Target Rods The red dotted line shows the CHFR (Bernath) for the peak linear power target rod. The blue dotted line s how s the CHFR (Bernath) for the maximum power rod. In all these analyses , the Bernath CHFR was consistently the mo s t conservative. Although maximum heat flux occurs at different points along the PLP and MH rod s, the minimum CHF R is a lmo st the sa m e for both rods. The sma ll temperature rise of the coolant y ield s an almos t constant line for the coo l a nt enthalpy as a function of rod axia l location.

Thus , the point of th e minimum CHFR coi n ci de s with the lo cation of the m aximum h eat flux in t h e rod. 6.1.3 Design Margins for Uncertainties In addi tion to a safety margin of I 0% on he at transfer coefficie nt s , seve ral a dditional margins to account for operat in g mea s urement uncertainties were inc lud ed in these anal yses: Page I 08 of 190 ATTACHMENT 1 1. The pressure drops through the entire TA that were calculated using CfD were reduc e d by 5% for the thermal analyses as this will reduce the water saturation temperature and give a lower predicted CHFR. 2. A control system uncertainty of +2 °C (3.6 °f) was applied to the cooling water inlet temperature. As with the pressure drop reduction , this reduces the predicted CHfR. 3. An uncertainty of -5% was applied to the system flow rate to account for possible errors in measurement

-this also reduces the predicted CHfR. Multip le analyses were performed to investigate the possible variations of operating parameters and geometric tolerances.

Each analysis was performed for both the nominal parameters and with the above margins for operating uncertainties.

The first set of results represents the most accurate possible e ngineering solution of the case analyzed.

The second set of results represents the worst possible stacking of all the operating uncertainties.

The thermal design is required to have a CHFR > 2.0 for the combined application of uncertainties for the worst possible operating condition.

A CHfR of 2.0 is a 100% margin by itself. 6.1.4 Target Assembly Steady-State Operations The water flow rate to the TA has a value of and is 10 °C (18 °f) below the pool temperat ure. It is conservatively assumed that 5% of the flow leaks out at the labyrinthine seal at the j unction of the removable cartridge and the target housing. This leakage will be determined by ex-reactor testing on the unit. The target flow rate is therefore . After flowing through the cartridge , it exits to the reactor pool at a pool depth of 6.85 m (22.5 ft). The absolute pressure at the diffuser exit is 1.66 atm and the pressure of the water at the mid-height of the target (location of maximum heat flux) is approximately 2.15 atm for -flow. This sets the local water saturation temperat ure at 123 °C (253.4 °f). The nominal water inlet and outlet temperatures are 28.89 °C (84 °f) and 38.89° C (102 °f), respectively.

This corresponds to a reactor pool temperature of 38.89 °C (102 °f). The 10 °C (18 °f) temperature rise corresponds to 11 U0 2 filled target rods in the cartridge , each dissipating an average of -During ramp up to full scale operation , between 1 and 11 U0 2 filled rods will be present in the cartridge , with the balance occupied by non-heat generating filler rods. In order to maintain the TA water outlet temperature equal to the pool temperature , the TA inlet temperature is automatically adjusted so that: T o ut , mixed = T P O O L = [ n x (Tinl et + 10) + (11-n) x T inl e t] I 11. Hence , Tinl e t = T o ut , mi xed -10111 x n = TPO O L -0.9091 x n. Table 28 shows the variation in coolant temperatures as the active rod count varies. Page 109 of 190 [5 a , d , e , f] [5 a , d , e , f]

ATTACHMENT 1 Table 28 Variation of Coolant Inlet Temperature with Active Target Rod Count Ac ti ve Non Heat i ng T;" T;n Tout Tout Mixed T 0 u1 Mixed T 001 Rods Rods (Q C) (o F) (Q C) (o F) (Q C) (°F) 3 t 8 36.16 97.09 46.16 11 5.09 3 8.89 102.00 4 7 35.25 95.46 45.25 113.46 38.89 102.00 5 6 34.34 93.82 44.34 111.82 3 8.89 102.00 6 5 33.44 92.18 43.44 110.18 38.89 102.00 7 4 32.53 90.55 42.53 108.55 38.89 102.00 8 3 31.62 88.91 41.6 2 106.91 38.89 102.00 9 2 30.7 1 87.27 40.71 105.27 38.89 10 2.00 10 1 29.80 85.64 39.80 10 3.64 38.89 102.00 11 0 2 8.89 84.00 38.89 10 2.00 3 8.89 102.00 t operations with l ess th a n 11 ac ti ve rods wi ll onl y b e a ll owe d durin g syste m commissioning.

At no time wi ll l ess th an 3 act i ve rods b e irr adiated to e n s ur e c ritical h ea t flu x ratios a r e a lw ays above 2 at m aximum p owe r/minimum flow con diti ons. To achieve a cartridge mixed outlet temperature eq ual to a nominal pool temperature of 38.89 °C (10 2 °F) th e range for T;nt e* w ill be 28.89 °C (84 °F) to 36.16°C (97.09 °F). For each additional degree of initial temperature ris e of the pool , the cartridge inlet temperature will need to increase b y th e same number of de grees. For the se analyses , it was assumed that thr ee (3) active tar get rod operation wo uld not occur if the pool temperatur e exceeded 42.22 °C (108 °F). This limit is dictated b y the requir eme nt to maintain CHFR's above 2.0 for all operational scenarios.

As can be see n from Ta ble 28 , both the inlet and outlet coolant temperature s rise with decreasing active tar ge t rod counts, the most severe operating conditions are those with th e l eas t number of active rod s. As the sys tem heats up , the gap between the U0 2 pell et and the Zircaloy-4 c laddin g wi ll be set b y the relative expansion of the pellet and the cladding , as well as the phy s ic a l expansion of the U0 2 bec ause of relocations in the pellet due to the stresses experienced during fuel bumup. Changes in the gap also change the volume occupied by the helium between the pellet and the cladding , which, along w ith temperature changes , increa se the pre ss ure of the ga p helium. It is noted that the thermal conductivity of helium is a function of temperature , but in very s mall gaps s uch a s these , it i s also a function of the ga p size and pressur e. Properties used for these analyses are shown in Figure 69 , Figure 70 , and F i gu re 71. P age 1 I 0 of 190 ATTACHMENT 1 20 20 1 8 18 0 16 E 1 4 -12 > ;: u ::I 10 ,, c 0 8 0 E --1 6 [ c 14 0 u; c 12 ca c. ......................... >< . .. . .. .............. .......... w 10 ca E 8 ... (I) LEGEND T herma l Q:inclichvf1y CTEt t -us e s ri ijit Y axis .c ca E 6 ... .... 6 -0 (I) .c .... 4 4 c (I) u 2 2 :; 0 (.) 500 1 000 1 500 2000 Temperature, °C Figure 69 U02 T h ermal Co n d u ctivity and Thermal Expansion Coefficient at 95% Theoretical Density 20 20 0 --E --1 8 18 E :i LEGEND 16 c 0 iii Thl'Jrmal Ccnd.Jctrv1ty 14 c ca Rad i al CTE t c. 1 2 in 0 16 E 3: 14 i-> 12 ca t -use s ri ijl t Y a x i s 10 e (I) .c 8 .... -:;:: u ::I 1 0 ,, c 0 8 (.) 0 6 -c *** '''"'''''''''''" ""n****.., , ... ,_. .. ,,,,,, , , , ,,,,,,,,,.,, , ,,, , ca E 6 .S! u 4 = ... (I) .c 4 .... (I) 0 2 2 0 .!! 200 300 400 ,, ca 00 100 Temperature, °C a: Fig u re 70 Zircaloy-4 Thermal Cond u ctivity an d Rad i al Thermal E x pansion Coefficient Pa ge 111 o f 190 u 0.40 E i-> 0.3'.l -u :I "D c 0 0.20 u ca E ... .! I-0.10 ATTACHMENT 1 .............. .......... ............. ********** .......................... 100 200 JOO 400 500 600 700 800 900 1000 Te mpe rat u re, °C Figure 71 LEGE N D Helium Thermal Conductivity in Small Gaps (2 atm pressure used as an example) 5 µmgap 10µm gap 20µm gap 4-0µm gap Target rod thermal analyses were performed for dimensional tolerance s (50 , -µm U0 2/cladding gaps and water channel nominal and maximum reactor po w er (10 [5 a , b , f] and I 1.5 MW) and coolant flows (nominal flow rate value of IOO% and the SCRAM limited flow rate value of 85%) for both I I and three (3) active target rods and for both the middle peak power rod and the end rod with the maximum linear heating. The case for three (3) rods is important because it has the highest inlet temperature thus reducing the margin to saturation. This may be seen in the slightly lower CHFR's. Of all the cases investigated , the smallest value of the CHFR was seen in the case of the peak power target rod; w ith nominal pellet/clad geometries and the largest possible water channel. In all cases investigated , the CHFR was greater than or equal to 2.0. Table 29 shows the details of steady-state operations for I I active target rods per cartridge , for both the nominal and wor s t case uncertainties , for both the peak power and maximum heating rods at 10 MW reactor power with I 00% TCS flow. Table 30 shows the details of those same cases but at l I .5 MW 1 reactor power and 85% TCS flow. This is the worst possible case (as the reactor will SCRAM at either higher power or lower flow). Table 3 I and Table 32 show the corresponding three (3) target rod cases. Table 33 show s the wor s t (lowest CHFR) cases , those for three (3) target rods , with a -gap and [5 a , b , f] added calculational margin. The CHFR for this condition is 2.07. B y compari s on , the expected minimum CHFR i s 2.92 per Table 29. Page 1 12 of I 90 ATTACHMENT 1 Ta bl e 29 Predicted Therma l Performance for 11 Active Target Rods, 10 MWt Reactor Power, 10 0% Flow Peak Power Peak Power Max L in ear Max L in ear R o d , No A dd e d Ro d, Maxi mum Heat in g Rod Hea tin g R od, No A dd e d Max imu m Margi n Ad d ed Mar gin Ma rgin A dd ed M argi n Reactor power , MW 1 10 10 10 10 Flow% 100 100 100 100 Active rods 11 11 11 11 Pool temperature , °C 38.9 40.9 38.9 40.9 Target in l et temperature , °C 28.9 30.9 28.9 30.9 Cold gap Nominal Nominal Nomi n al Nominal Rod type Peak Linear Peak Linear Max Heating Max Heating Rod location Middle Middle End End Added error margin None Ma x imum None Maximum Scallop diameter Nominal Maximum Nominal Maximum Cold gap , µm 50 50 50 50 U0 2 ra d ia l growth , µm 40.06 40.0 1 38.7 38.66 Cladding radial growth , µm 2.92 2.9 2.88 2.86 Relocation growth , µm 5.11 5.11 4.94 4.94 Hot gap , µm 7.76 7.79 9.24 9.26 He l ium gap pres s ure , atm 1.74 1.74 1.71 1.71 Plenum height , mm (Co l d) 35 35 35 35 Plenum heig h t , mm (Hot) 32.61 32.61 32.55 32.56 Axial CTE U0 2. µm/(m°C) 9.87 9.87 9.8 7 9.87 Radial CTE U0 2. µm/(m 0 C) 1 0.47 10.47 10.44 10.44 Radia l CTE Zircaloy , µm/(m°C) 6.1 7 6.16 6.16 6.1 6 U0 2 ce n terline temperature , °C 2000.1 1998.7 1942.4 1941.1 U0 2 average temperature , °C 1117 1115.8 109 7.4 1096.4 U0 2 surface temperature , °C 423.6 422.9 433.0 432.4 Helium te m perature , °C 345.2 344.2 348.1 347.2 C l adding temperature at ID , °C 263.0 261.8 258.4 25 7.2 Mean claddi n g temperat u re , °C 206.0 204.7 203.4 202.2 Cladding temperature at OD , °C 156.1 154.8 155.4 154.1 Page 1 13 of 190 ATTACHMENT 1 Table 29 Predicted Thermal Performance for 11 Active Target Rods, 10 MWt Reactor Power, 100% Flow (contin ued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Maximum Heating Rod Heating Rod, No Added Maximum Margin Added Margin Mari:i:in Added Man?in U0 2 gap conductance , W l (m 2 C) 22160.5 22092.4 19603.2 19539.5 Coolant HTC , W l (m 2 C) ----Water (scallop)

OD , mm 14.45 14.823 14.45 14.823 D 1" mm 10.243 10.746 9.14 9.555 Re (Db) ----Local coo l ant velocity , m i s * * *

  • M as s flow per rod , kg/s ----Mass flow per target , kg/s * -* -Mass flow per target , gpm ----Coolant inlet temperature , c c 28.89 30.89 28.89 30.89 Local coo l a nt temperature, c c 32.52 34.71 34.02 36.29 Coolant outlet temperature , c c 39.14 41.68 39.61 42.17 Local saturation temperature , c c 1 24.53 I 23.15 124.11 1 22.78 Peak pellet power density , W i ce ----Heat (mo d eled) per rod , W ----Heat flux at c l adding OD , W l m 2 ----Bernath CHF , W l m 2 ----Bernath CHFR 3.17 2.93 3.18 2.92 Macbeth CHF , W l m 2 ----Macbeth CHFR 3.67 3.60 4.06 4.00 Groeneveld CHF , W l m 2 ----Groeneve ld CHFR 3.40 3.13 3.50 3.24 Minimum CHFR 3.167 2.933 3.176 2.924 Pressure at CHFR , atm 2.2588 2.1645 2.2299 2.1400 Page I 14 of 190 [5 a , d , e , f]

ATTACHMENT 1 Ta bl e 30 P r edicted Therma l Performance for 11 Active Target Rods, 11.5 MWt Reactor Power, 85% F l ow Pea k Powe r Peak Power Max L in ear Max Lin ear Rod , No Adde d Rod , Maxi m um Hea tin g Rod No Hea tin g R od, Max imu m Margi n A dd ed Ma r g in Adde d Ma r g in A dd ed Ma rgi n Reactor power , MW 1 11.5 I 1.5 11.5 11.5 Flow% 85 85 85 85 Active rods 11 11 l I 1 I Pool temperat u re , °C 38.9 40.9 38.9 40.9 Target in l et temperature , °C 28.9 30.9 28.9 30.9 Cold gap Nominal Nomina l Nomina l Nominal Rod type Peak Linear Peak Linear Ma x Heating Max Heatin g Rod location Middle Middle End End Added error margin None Ma x imum None Maximum Scallop diameter Nominal Maxim u m Nominal Maximum C old gap , µm 50 50 50 50 U02 rad i a l growth , µm 45.16 45.l I 43.79 43.7 5 C ladding radial growth , µm 3.03 3.01 2.98 2.97 Relocation growth , µm 5.11 5.1 I 4.94 4.94 Hot gap , µm 2.77 2.79 4.25 4.27 Helium gap pressure , atm 1.85 1.85 1.82 1.82 Plenum h e i ght , mm (Co l d) 35 35 35 35 Plenum h eight , mm (Ho t) 32.87 32.88 32.7 9 32.79 Axial CTE U0 2 , µm/(m 0 C) 9.86 9.86 9.86 9.86 Radial CTE U0 2 , µm/(m°C) 10.57 10.57 10.55 10.55 Radial CTE Zircalo y, µm/(m 0 C) 6. 19 6.18 6.18 6.17 U 0 2 centerli n e temperat u re , °C 2204.7 2203.6 2150.8 2149.8 U 0 2 average temperature , °C 1181.0 1180.0 1165.7 I 164.9 U 0 2 surface temperat u re , °C 379.3 378.7 394.2 393.8 H elium temperature , °C 328.5 327.8 333.9 333.2 C ladding temperature at ID , °C 277.2 276.2 272.1 2 7 1.2 M ean cladding temperature , °C 212.1 211.1 209.3 208.4 C ladding temperature at OD , °C 154.9 153.9 154.2 153.3 Pa g e 115 of 190 ATTACHMENT 1 Table 30 Predicted Thermal Performance for 11 Active Target Rods , 11.5 MWt Reactor Power, 85% Flow (continued)

Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Maximum Heating Rod No Heating Rod, Maximum Margin Added Margin Added Margin Added Mare:in U0 2 g ap conductance , W l (m 2 C) 40111.0 39965.3 32241.7 32132 Coolant HTC , W l(m 2 C) -------Water (sca llop) OD , mm 14.450 14.823 14.450 14.823 o, ,, mm 10.243 10.746 9.1 40 9.555 Re (D1i) ----Local coo lan t velocity , m i s * * *

  • Ma ss flow per rod , k g/s ----Mass flow per target , kg/s ----Mass flow per target , gpm .. .. .. .. Coolant inl et temperat ur e , °C 28.89 30.89 28.89 30.8 9 Local coo lant temperat ur e , °C 33.81 36.06 35.83 38.19 Coolant outlet temperat ur e , °C 42.76 45.49 43.39 46.16 Local saturation temperature , °C 122.06 121.00 121.74 120.73 Peak pellet power density , W i ce ----Heat (mo d e l ed) per rod , W ----Heat flux at cladding OD , W l m 2 ----Bernath CHF , Wlm 2 ----Bernath CHFR 2.47 2.31 2.47 2.29 Macbeth CHF , W l m 2 ----Macbet h CHFR 3.24 3.18 3.59 3.53 Groeneveld CHF , Wlm 2 ----Groeneveld CHFR 2.69 2.51 2.7 9 2.62 Minimum CHFR 2.473 2.307 2.465 2.286 Pressure at CHFR , atm 2.0 921 2.0240 2.0712 2.0063 Page 116 of 190 [5 a , d , e , f]

ATTACHMENT 1 Table 31 Predicted Thermal Performance for 3 Active Target Rods , 10 MWt Reactor Power, 100% Flow Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod , Maximum No Added Maximum Margin Added Margin Margin Added Margin Reactor power , MW 1 10 10 1 0 10 Flo w% 100 100 100 100 Active rods 3 3 3 3 Pool temperature , °C 42.2 44.2 42.2 44.2 Target inlet temperature , °C 39.5 41.5 39.5 41.5 Cold gap Nominal Nominal Nominal Nominal Rod type Peak Linear Peak Linear Max Heatin g MaxHeatin g Rod location Middle Middle End End Added error margin None Maximum None Maximum Scallop diameter Nominal Maximum Nominal Maximum Cold gap , µm 50 50 50 50 U02 rad i a l growth , µm 40.06 40.01 38.7 38.66 Cladding radial growth , µm 2.92 2.9 2.88 2.86 Relocation growth , µm 5.11 5.11 4.94 4.94 Hot g ap , µm 7.76 7.79 9.24 9.26 Helium ga p pressure , atm 1.74 1.74 1.71 1.71 Plenum height , mm (Co ld) 3-5 35 35 35 Plenum height , mm (Hot) 32.61 32.61 32.55 32.56 Axial CTE U0 2 , µm/(m°C) 9.87 9.87 9.87 9.87 Radial CTE U0 2 , µm/(m 0 C) 10.47 10.47 10.44 10.44 Radial CTE Zircalo y, µm/(m°C) 6.17 6.16 6.16 6.16 U0 2 center lin e temperature , °C 2000.1 1998.7 1942.4 1941.l U 0 2 average temperat ur e , °C 1117 1115.8 1097.4 1096.4 U 0 2 su rface temperature , °C 423.6 422.9 433 432.4 H elium temperature , °C 345.2 344.2 348.1 347.2 C ladding temp erature at JD , °C 263 261.8 258.4 257.2 Mean clad din g temperature , °C 206 204.7 203.4 202.2 Cladding temperature at OD , °C 156.1 154.8 155.4 154.1 Pa g e 117 of 190 ATTACHMENT 1 Table 31 Predicted Thermal Performance for 3 Active Target Rods, 10 MWt Reactor Power, 100% Flow (co ntinued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin Added Man?:in Man?:in Added Man?:in U0 2 gap conductance , W l (m 2 C) 22160.5 22092.4 19603.2 19539.5 Coolant HTC , W l(m 2 C) ------Water (sca ll op) OD , mm 14.450 14.823 14.450 14.823 D h,m m 10.243 10.7 46 9.140 9.555 Re (Oh) ----Local coo lant veloc i ty, m is * * *

  • Mass flow per rod , kg/s ----Mass flow per target , kg/s ----Mass flow per target , gpm ----Coolant inlet temperature , °C 39.49 41.49 39.49 41.49 Loca l coola nt temperatu r e, °C 43.12 45.31 44.62 46.89 Coolant outlet temperature , °C 49.74 52.28 50.21 52.77 Local saturation temperature , °C 124.53 123.15 124.11 122.78 Peak pe llet power density, W ice ----Heat (modeled) per rod , W ----Heat flux at cladding OD , Wlm 2 ----Bernath CHF, W l m 2 ----Bernath CHFR 2.90 2.68 2.9 1 2.67 CHF , W l m 2 ----Macbeth CHFR 3.53 3.47 3.91 3.85 Groeneveld CHF , W l m 2 ----Groeneveld CHFR 3.16 2.9 1 3.26 3.02 Minim um CHFR 2.905 2.683 2.909 2.671 Pressure at CHFR , atm 2.2588 2.1645 2.2299 2.1 400 Page 118 of 190 [5 a , d , e , f]

ATTACHMENT 1 Ta bl e 32 Pred i cte d Therma l Pe r forma n ce for 3 Active Target Rods, 11.5 MWt Re a c t or Powe r , 85% Flow P ea k Powe r P eak P ower Max L in ear Max Lin e ar Rod , No A dd e d R o d , Heat in g Ro d Hea t i n g R od, Max imum No A dd e d M a x imum M ar g in A dd ed M a rgin Ma r g in A dd e d Ma r g in Reactor Power , MW 1 11.5 11.5 11.5 11.5 Flow% 85 85 85 85 Active rods 3 3 3 3 Pool Temperature, c c 42.2 44.2 42.2 44.2 Target in l et temperature, cc 39.5 41.5 39.5 41.5 Cold Gap Nominal Nominal Nominal Nominal Rod Type Peak Linear Max Heating Max Heating Max Heating Rod loca t ion Middle End End End Added Error Margin None Maximum None Maximum Scallop Diameter Nomina l Maximum Nominal Maximum Cold Gap, µm 50 50 50 50 U02 radia l growth , µm 45.16 43.75 43.79 43.75 Cladding radial growth , µm 3.03 2.97 2.98 2.97 Relocation Growth , µm 5.11 4.94 4.94 4.94 Hot Gap , µm 2.77 4.27 4.25 4.27 Helium Gap pressure , atm 1.85 1.82 1.82 1.82 Plenum height , mm (Cold) 35 35 35 35 Plenum h e i ght , mm (Hot) 32.87 32.79 32.79 32.79 Axial CTE U0 2. µm/(mcC) 9.86 9.86 9.86 9.86 Radial CTE U0 2 , µm/(m c C) 10.57 10.55 10.55 10.55 Radial CTE Zircaloy , µm/(mcC) 6.19 6.17 6.18 6.17 U0 2 cente rl ine temperature, cc 2204.7 2149.8 2150.8 2 1 49.8 U 0 2 average temperature , cc 1 181.0 1164.9 1165.7 1164.9 U 0 2 surface temperat u re , cc 379.3 393.8 394.2 393.8 H elium te m perature, cc 328.5 333.2 333.9 333.2 Cladding temperature at ID, cc 277.2 271.2 272.1 271.2 Mean c l adding temperature , cc 212.1 208.4 209.3 208.4 C ladding temperature at OD , cc 154.9 153.3 154.2 153.3 Page 119of190 ATTACHMENT 1 Table 32 Predicted Thermal Performance for 3 Active Target Rods, 11.5 MWt Reactor Power , 85% Flow (co ntinu ed) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod , Maximum No Added Maximum Margin Added Margin Margin A dded Margin U0 2 gap conductance , W l (m 2 C) 40111.0 32132.0 32241.7 32 1 32.0 Coolant HTC , W l(m 2 C) ----Water (sca ll op) OD , mm 14.450 14.823 14.450 14.823 Dh , mm 10.243 9.555 9.140 9.555 Re (D h) ----Local coo l ant velocity , m i s * * *

  • Mas s flow per rod , k g/s ----Mas s flow per target , kg/s ----Ma ss flow per tar g et , gpm .. .. .. .. Coolant inlet temperature , c c 39.49 41.49 39.49 41.49 Local coo l a n t temperature , c c 44.40 48.79 46.43 48.79 Coolant outlet temperature , c c 53.36 56.75 53.99 56.75 Local s aturation temperature , c c 122.06 120.73 121.74 120.73 Peak pe ll et power density , W i ce ----Heat (modeled) per rod , W ----Heat flux at cladding OD , W l m 2 ----Bernath CHF , W l m 2 ----Bernath CHFR 2.26 2.08 2.25 2.08 Macbeth CHF , W l m 2 ----Macbeth CHFR 3.13 3.41 3.46 3.41 Groeneveld CHF , W l m 2 ----Groeneveld CHFR 2.50 2.45 2.60 2.45 Minim um CHFR 2.261 2.080 2.250 2.080 Pressure at CHFR , atm 2.0921 2.0063 2.0712 2.0063 Pa g e 120 of 1 90 [5 a , d ,e, t]

ATTACHMENT 1 Table 33 Predicted Thermal Performance for 3 Active Target Rods, Worst-Case Operations Peak Po we r Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod , Maximum No Added Ma x imum Margin Added Margin Margin Added Margin Reactor power , MW, 10 10 11.5 11.5 Flow% 100 100 85 85 Active rods 3 3 3 3 Pool temperature , cc 44.2 44.2 44.2 44.2 Target inlet temperature , cc 41.5 41.5 41.5 41.5 Cold ga p Minimum Minimum Minimum Minimum Rod type Peak Linear Max Heating Peak Linear Max Heating Rod location Middle End Middle End Added error margin Maximum Maximum Maximum Maximum Scallop diameter Maximum Maximum Maximum Maximum Cold gap , µm 36.25 36.25 36.25 36.25 U02 radia l growth , µm 33.72 32.88 43.8 40.59 Cladding radial growt h , µm 2.89 2.85 3.00 2.96 Relocation Growth , µm 3.70 3.58 3.70 3.58 Hot gap , µm 1.73 2.64 0.0 0 0.00 Helium gap pres s ure , atm 1.66 1.64 1.71 1.71 Plenum height , mm (Co ld) 35 35 35 35 Plenum h eight , mm (Hot) 33.13 33.05 32.98 33.03 Axial CTE U0 2 , µm/(mcC) 9.85 9.86 9.86 9.86 Radial CTE U0 2 , µm/(m c C) 10.26 10.25 10.52 10.44 Radial CTE Zircaloy , µm/(m cC) 6.16 6.16 6.18 6.18 U0 2 cente rlin e temperature , cc 1795.7 1749.3 2168.8 2060.6 U0 2 average temperature , cc 965.0 955.6 1150.2 1093.8 U0 2 surface temperature, cc 335.8 349.5 361.3 353.0 Helium temperature , c c 299.8 304.5 319.9 313.2 C ladding temperature at ID , cc 263.8 259.2 2 7 8.5 273.4 Mean cladd in g temperature , cc 205.0 202.5 21 1.4 208.7 C ladding temper ature at OD , c c 154.8 154.2 153.9 153.3 Page 121of190

--ATTACHMENT 1 Tab l e 33 Predicted Thermal Performance for 3 Active Target Rods, Worst-Case Operatio n s (contin u ed) P ea k Po we r P e a k Powe r Max L in ear M ax L in ear Rod , No A dd e d Rod , H eating Rod H e ating R o d , M ax imum N o Added Ma x imum Ma r g in A dd e d Man?:in Mar2in A dded M a r2in U0 2 gap co n ductance , W/(m 2 C) 49640.3 38060.4 49640.3 49640.3 Coola n t HTC , W /(m 2 C) ----Water (sca ll op) OD , mm 14.823 14.823 14.823 1 4.823 Dh , mm 10.785 9.586 10.785 9.586 Re (D1i) ----Local coo l ant velocity , mis * * *

  • Mass flow per rod , kg/s ----Mass flow per target, kg/s ----Mass flow per target , gpm --.. .. Coolant in l et temperat u re , °C 41.49 41.49 41.49 41.49 Local coo l ant temperat u re , °C 45.32 46.89 46.67 48.80 Coolant o u t l et temperature , °C 52.29 52.78 56.10 56.76 Local saturation temperature , °C 123.15 122.78 12 1.00 1 20.73 Peak pe ll et p ower density , W/cc ----Heat (mo d eled) per rod , W ----Heat flux at cladding OD, W/m 2 ----Bernath CHF, W/m 2 ----Bernath CHFR 2.67 2.66 2.09 2.07 Macbeth CHF , W/m 2 ----Macbeth CHFR 3.44 3.83 3.05 3.39 Groeneve l d CHF , W /m 2 ----Groe n evel d CHFR 2.89 3.00 2.32 2.44 Minimum CHFR 2.668 2.657 2.092 2.070 Pressure at CHFR , atm 2.1645 2.1400 2.0240 2.0063 Page 122 of 190 [5 a , d , e , f]

ATTACHMENT 1 6.1.5 Flow Induced Vibrations An analysis of the displacement of target rods caused b y vibration due to the surrounding flow was performed.

Ignorin g the stiffening effects of the tube filled with the tar get pellets , the fixed-end Zircaloy-4 tubes have a natural frequency of 26 Hz and a maximum displacement of 6S µm , using the correlations of Paidoussis , which is also the ASME B&PV recommended method (Reference 32). A maximum displacement of 4 µm i s calculated using Farmer et.a l (Reference 33). In either case , it can be concluded that vibrat ion s are not an issue for the target systems. 6.1.6 Cooling of the Beryllium Reflector during Operations with Target Rods The Be reflector rests between the TA and the outer reactor pressure vessel. Nuclear , fluid flow and heat transfer analyses were perfonned to show the effect of the T As in the two (2) graphite reflector positions.

The important consideration is the increased heating of the Be reflector and the effect on cooling of the Be surfaces.

The MURR Be reflector generates up to 200 kW of heat , primarily due to neutron scattering a nd absorption of high energy photons from the reactor core. With the target rods installed , the MCNP anal ys es show that the 22 targets rods will increase the reflector power density by as much as 24% at the reac tor core elevation centerline (its hottest point). This heat is then rejected to the MURR pool coolant sy stem. It i s not possible to measure the flow velocity aro und the Be reflector.

Therefore , a model of the reflector c ooling flow was created that accounts for these distinct flow paths. 1. Through the flux trap inside the annular core pressure vessel , which has a minimal effect on the reflector coo lin g. 2. Flow that travels bet ween a O.S6-inch (14.2 mm) gap between the outer reactor pressure vessel and the inner wa ll of the Be reflector.

The reactor regulating and contro l blades also reside in this gap and were acco unted for. 3. The flow that passes through the gaps around the graphite reflector elements (and TAs) and down through holes and slots in the reflector su pport plate. A portion of this flow contacts the outer wall of the beryllium reflector , and thus is relevant to the reflector cooling. Measurements show the pressure drop across the reflector assembly is 0.78 psi (S.38 kPa), which will serve as the basis for determining the flow rate in the reflector region. The flows through the three (3) fl ow paths were iterated in the model until all flow path s had the measured pressure drop and the tota l fl ow equaled the measured flow. The flow surrounding the Be reflector , the predicted power density in the reflector , and the resulting heat transfer were examined to determine if the pool coolant syste m provides adequate coo lin g to the Be reflector in the presence of two TAs installed in graphite reflector positions No. SA and No. SB (Reference 34). Worst-case contro l/regulating blade positions, r eflector age and estimated power densities were used. The results show that the reflector maintains a 20 °C (36 °F) margin from surface temperature to coolant saturation temperature in the presence of both T As and hence there is no impact on Page 123of190 ATTACHMENT 1 cooling of the Be reflector.

Additionally , the driving pressure head was shown to have a small effect on the flow rate , indicatin g that it is unlikely the su rface temperature of the Be will increase significantly if the pressure head changed. 7. In-Pool Target Transfer System 7.1 Cartridge Loading/Unloading Station Design This section pre sents a description of the equipment and process that will take place to move target rods from/to target storage to/from the target cartridge at the cartridge loading/unloading station. 7.1.1 Cartridge Loading/Unloading Station Description The cartridge loading/unloadin g station is located in the weir area of the reactor pool. It is an aluminum structure designed to bolt to the existing pool struc ture and contains storage for 33 target rods , four (4) target cartridges , and two (2) target diffuser s. Figure 72 shows the location of the in-pool storage area within the MURR reactor pool and how it is positioned in relation to the reactor. Figure 72 CARTRIDGE LOADING/U?'<"LOADING STATION CARTRIDGE WITH TARGET RODS TARGET GRAPHITE Posmo:s Target Rod Loading/Unloading/Storage Location The cartridge loadin g/unloading stat ion ha s three functions (Figure 73): I. It provide s a location for the removal and installation of the diffu ser from the cartridge; Page 124of190 [5 a , b d , e , t]

ATTACHMENT 1 2. It holds the cartr id ge while target rods are removed or insta ll ed into the cartridge; and 3. It provid es an approved target rod storage l ocation that place s the tar get rods in a geometry to e n s ure a l a r ge margin to critica li ty. odio t ed corlridge from down below goes h ere

  • Space is e m pty o r mo Dy. vs ed fo r co rt *d ge t h o i s brought up for servic i ng Figure 73 Representation of the Cartridge Loading/Unloading Station Page 1 25 of 1 90 [5 a , b , d ,e,t]

ATTACHMENT 1 During normal operations , no more than 22 target rods will be either in a cartridge or in target storage at the cartridge lo ad in g/unloadin g sta tion. Of those 22 target rods , only 11 of them will be fresh. However , the performed criticality analysis assumes 33 target storage positions and two (2) nearb y target cartridges with 11 position s eac h filled with fresh target rods for a total of 55 fresh target rod s in the station at the same time. The criticality analysis shows a maximum K e rr of 0.57 under all conditions of moderation and reflection.

MURR Technical Specification 5.4.a requires that fueled devices outside the reactor core be stored such that the K e rr i s le ss than 0.9 under all conditions of moderation and r e flection. Administrative controls s imilar to the MURR ad ministrative controls for reactor fuel handling will b e implemented to ensure that the location of any tar ge t rod is known at all times and that the target rod s are placed in the correct irradiation and storage locations.

7.1.2 Cartridge

Loading/Unloading Station Operation The normal sequence of operation at the cartrid ge loadin g/unloading station will be: I. Lower 11 or les s fresh target rods into the sta tion storage location. 2. Insert 11 or less fresh target rods into a cartridge at the station using the remote handlin g tool shown in Figure 71 . 3. Bring an irradiated target cartridge from the target housing. 4. Remove the diffuser from the cartridge and place into storage ba sket. 5. Remove irradiated target rods from the irradiated cartridge and place them int o the station storage location. 6. Move target rods as necessary to complete the placement of the next 11 target rods to be irradiated. (Note: T he next 11 target rods to be irradiated ma y be all fresh or may be a mixture of fresh and previously irradiated target rods.) 7. Move the diffuser from the storage ba ske t and in s tall it onto the target cartridge to be irradiated.

8. Move the loaded tar get cartridge to be irradiated into the target housing at the reactor reflector (target rod remote handling tool shown in Figure 74). Pa ge 126of190 ATTACHMENT 1 Used when removal of lndMdual target rods In shallow end of the pool Is desired Spring-loaded positive lock mechanism Procedure:
l. Bring tool down to target rod a. PuU on exterior release to claw tool to engage end cop 2. Push down over tar get rod unt il tt bottoms out a. Le t go of exterior release to lock t ool In place 3. Remove t arget rod o. Pul on ex te rior release to f r ee target rod Threaded portion to attach to remo v al tube/r od Exterior release C ompression spring Sliding release mechanism female e nd cop receptacle Figure 74 Target Rod Remote Handling Tool 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing 7.2.1 Target Cartridge Description A detailed de scri ption for the target cartridge and how it fits into the target hou si n g is provided in Section 2.1.4. 7.2.2 Cartridge Installation and Removal Into/From the Target Housing Operation T he nonnal sequence of operation for transferrin g the irradiated target cartridge from the target housin g to the cartridge loading/unloading station will be: I. Shut down the reactor. 2. To allow sufficient cooling and deca y , keep the target cartrid ge in the tar ge t housing for a minimum of one (1) hour after reactor shutdown prior to unlatchin g the cartridge.
3. Remove the RTD at the top of the diffuser and secure it out of the way of cartridge movement operations.
4. Thread the cartridge removal tool into the top of the diffuser.
5. Unpin the cartridge by turning bolt-connected levers w ith a socketed , long tool. Pa ge 127of190 ATTACHMENT 1 6. Lift the cartridge vertically with the attached cartridge removal tool. 7. Move the cartridge to the cartridge loading/unloading station. 8. Once the cartridge is attached to the station , unthread the cartridge removal tool. The normal sequence of operation for transferring the target cartridge from the cartridge loading/unloading station to the target housing will be: I. Thread the cartridge removal tool into the top of the diffuser 2. Move the cartridge to the just above the target housing in which the cartridge will be irradiated.
3. Slowly lower the cartridge into the housing making sure the cartridge rail features e n gage the target housing rails as seen in Figure 5. 4. Gently push downward to ensure cartridge is resting in the target housing with proper alignment pin engagement.
5. Secure the cartridge b y turning bolt-connected levers with a socketed , long tool until no further movement is detected.
6. Pull upward and push downward on the attached cartridge removal tool to ensure the cartridge is sec ur ed and cannot move. 7. Visually observe the proper position of the crutridge locking mechanism.
8. Unthread the cartr id ge removal tool. 9. Install the RTD at the top of the diffuser.
8. Radiological Protection Evaluation for the SGE Target Experimental Facility Operations This section discusses and analyzes the expected radiological consequences related to normal operations from the production of Mo-99 using the SGE TEF. Includ ed are the principal discussions of the MURR facility program to control radiation and expected radiation exposures due to operation , maintenance, and use of the irradiation hardware associated with the TEF. This section also outlines the methods for quantitative assessment of radiation doses in the restricted and unrestricted areas; application of these methods to all applicable radiation sources related to the operation of the TEF; and the program and provisions for protecting the health and safety of a ll individuals present at MURR , the general public and the environment.

This section does not discuss the Radiation Protection Aspects of the MURR Health Physics program as related to the processing of the LEU targets for the Selective Gaseous Extraction (SGE) of Mo-99 from the TEF. That wi ll be covered under the Part 2 License Amendment submission for processing the LEU target rods and the extraction of Mo-99. The MURR Radiation Protection Program (RPP) has been establis h ed to protect the health and safety of all individuals present at MURR , the general public and the environment.

In accordance with 10 CFR 20.110 I , this program has been developed , documented , and implemented to a level commensurate with the scope and extent of licensed activities at MURR , and is sufficient to ensure compliance with the Page 128of190 ATTACHMENT 1 regulations in 10 CFR 20. A primary component of this program is the fundamental principle of maintaining individual radiation exposures and releases of radioactive effluents as low as is reasonably achievable (ALARA). Responsibility for maintaining the MURR ALARA Program extends to all individuals who are granted access to the reactor facility.

Renewed Facility Operating License No. R-103 is the primary license that covers the authority and responsibilities associated with the reactor and the majority of radioactive materials produced at MURR. It is under the auspices of this license that operation of the TEF will occur. Radioactive Material use is further supplemented by the authorization of an NRC-issued Broad Scope license (24-00513-39) and is used to support the research and development mission of the MURR for materials and activities which may not be currently produced or covered under license No. R-103. The radiation sources that are expected to be generated during irradiation of the TEF fall under the administrative control of the MURR Radiation Protection and Radioactive Waste Management Programs and can be categorized as airborne, liquid, or solid. While each of these categories is discussed individually in the following paragraphs, the major contributors to each category can be summarized as follows: 8.1 Airborne Sources Airborne -Potential airborne sources consists mainly of the fission product gasses of the bromine, iodine, krypton and xenon species. Of these, iodine-131 (I-131) has the lowest Derived Air Concentration

{DAC) and Effluent Release limits due to potential thyroid exposure while krypton-85 (Kr-85) has the longest ha:lf-life (10.76 years); however, Kr-85 is a noble gas and as such is minimally active with regards to human and non-human biological effects. All potential airborne sources from the SGE TEF are confined by the target rod cladding and are not expected to be released during routine operations.

Currently, during nonnal operation of MURR, argon-41 (Ar-41) is the principal source of airborne radioactivity

(> 99%) released through the facility ventilation exhaust stack. However, similar to the irradiation of reactor fuel, there exists the possibility to release any one of the fission products created during the fission process from the TEF. Such release would likely be limited to noble gases (krypton and xenon) and the halogens iodine and bromine that exist in the "gap" between the LEU pellets and the target cladding.

These releases could occur during two (2) points of the irradiation and use of the targets; (1) during the actual irradiation of the target rods in the graphite reflector region (this would occur if a leak occurs in the cladding of the target rod, thereby allowing the gap fission gases to escape), and (2) during handling of the target rods in the reactor pool pre-and post-irradiation.

3.2 Liquid

Sources Liquid -Liquid sources would include primary and pool coolant used in the reactor coolant systems and 1he TEF target cooling water, which is essentially reactor pool water that has been pumped to a separate cooling system to facilitate target heat removal. The water becomes activated and contaminated by direct neutron activation of the water molecule itself (H-3) along with any activation of impurities contained in the water due to its use as a coolant. The pool water would also contain contaminants due to its direct Page 129of190 ATTACHMENT 1 contact with the reactor and pool coolant system components.

Since primary and pool coolant is, by design, contained to the maximum extent possible, there are no substantial releases of these liquids directly into the environment.

However, certain reactor maintenance activities along with TEF cooling system maintenance could result in small volumes of liquid (containing mainly tritium) being directed to the existing liquid waste retention system. Limited and strictly controlled quantities of liquid radioactive waste are released to the sanitary sewer in accordance with the requirements of 10 CFR 20.2003; this effluent may contain small quantities of pool and primary water. Thus the operation of the TEF would add little to existing liquid waste effluent streams. All potentially radioactive aqueous liquid wastes are directed to a liquid waste retention and disposal system located at a level below the main grade of the MURR laboratory building.

Liquid waste is then retained or mechanically processed until an assay indicates that activity concentration levels are less than the limits specified in 10 CFR 20, Appendix B for disposal by release into sanitary sewerage.

Tritium currently accounts for about 81 % of the total activity released each year. Historically, H-3 release is < 5% of the annual release limit of 5 Ci. Pool coolant is the only radioactive liquid source expected to be impacted by the TEF. Radioactivity in this liquid occurs by the same mechanisms as described above: neutron interactions with hydrogen and oxygen in the water and neutron interactions with system and structural components, with the subsequent transfer into the pool coolant. Table 34 contains a list of the predominant radionuclides and their typical measured concentrations present in the pool coolant at 10 MW. Table 34 Predominant Radionuclides in the MURR Pool Coolant and Their Measured Concentrations at 10 MW Radionuclide Half-Life Typical Measured Concentration 1 Magnesium-27 (2 7 Mg) 9 .45 minutes 1.10 x 10-2 Ci/ml Sodium-24 (2 4 Na) 14.96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.64 x 10-3 Ci/ml Manganese-56 (56 Mn) 2.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> 2.54 x 10-3 Ci/ml Technetium-101 (101 Tc) 14.2 minutes 4.70 x 10-5 Ci/ml Technetium-99m (99 mTc) 6.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 9.73 x 10-6 Ci/ml Antimony-122 (122 Sb) 2.70 days 1.01 x 10-5 Ci/ml Xenon-135 (135 Xe) 9.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 1.22 x 10-5 Ci/ml Silver -1 lOm (110 mAg) 248.9 days 1.10 x 10-5 Ci/ml 1 Listed values are typical of the measured concentrations that exist in the pool coolant at I 0 MW, 2 to three days after reactor startup. Due to the structural components that will be used in the target rods and assemblies, it is anticipated that any additional activity added to the pool water will be similar in nature to those found in the table above Page 130 of 190 ATTACHMENT 1 with the exception of the potential addition of isotopes of zirconium, most likely Zr-95 and Zr-97 with half-lives of 64.02 days and 16.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />, respectively; as zirconium is the main constituent in the target rod cladding, as Zircaloy-4.

Overall, however, it is anticipated that any increase in activity will not be greater than 6%, as determined by ratio of the TEF thermal power to the MURR core thermal power. It should be noted that the pool water will still be diverted to a pool clean up system (as is currently in place) that has anion-cation removal capability in order to ensure that the pool water is relatively free of contamination.

This slight increase in pool water activity is not expected to warrant additional engineering or administrative controls beyond the control currently in place at the MURR. JFor the purposes of the TEF, cooling water will flow from the WCM, located outside of the reactor pool, through piping located in the pool and then enter at the top of the TA. After traveling down through the housing the water will then pass upwards through the target rod region where it removes the fission heat from the target rods. The water will then be discharged out of the TA into the pool near the top of the assembly through a diffuser back into the bulk pool water region. Here the water will intermingle with the reactor pool water and circulate through the pool hold-up tank system prior to entering the pool coolant heat exchangers as noted earlier. A portion of that pool water will be diverted off to the TEF cooling system and then cooled further before being reintroduced into the TA. By utilizing existing reactor-related cooling and decay systems for the reduction of N-16, the addition of any N-16 produced by the T As will not add any appreciable source term to the overall inventory of N-16 that must be decayed and/or shielded in order to protect reactor staff. :S.3 Solid Sources :Solid -Solid sources are a bit more diverse, but for the most part are very typical of a research reactor :facility.

Such sources include the reactor fuel in use in the core, irradiated fuel stored in the reactor pool, and new, unirradiated fuel. Additionally, fueled target rods used to produce the Mo-99 will be an additional solid radiation source. At any one time, there will be a maximum of 22 target rods each 1:;ontaining 106.2 grams of uranium nominally enriched to 19.75% in the isotope U-235 being irradiated.

After irradiation, the target rods will be transferred to a dedicated hot cell system and then processed to extract Mo-99 through the SGE process. Any sources of radioactivity germane to the SGE process would be further described in the Part 2 licensing action and not included herein. Thus, the target rods (irradiated and unirradiated) would solely add to the current uranium inventory at MURR. The solid radioactive sources associated with normal operation of MURR are summarized in Table 35. Because the actual inventory of reactor fuel, SGE target rods and other radioactive sources continuously change as part of the normal operation of the reactor and the experimental program, the information presented in Table 35 should be considered representative rather than an exact inventory.

Disposition of the irradiated LEU-target material will be discussed in the Part 2 License Amendment submission and is expected to be coordinated through a Uranium Lease Take-Back agreement between MURR and the U.S. Department of Energy (DOE). Page 131 of 190 ATTACHMENT 1 Table 35 Representative Radioactive Sources at MURR ;* AI>proximate Total ' * < *Source . Nominal Physical Wt% *. Radionudide(s)

Activity .(grams) * ', Description Characteristics

' " ' Total u *'. 0 ,,,._,, *: " " "'*"*' '. 8MURRFuel Highly Enriched NIA In 10 MW Core -93 6,200 6,656 Elements Uranium Irradiated 45 MURR Fuel Highly Enriched NIA In Pool Storage -93 34,875 37,440 Elements Uranium Irradiated 4MURRFuel Highly Enriched NIA In Storage -New 93 3,100 3,328 Elements Uranium 22 SGE Target Low-Enriched NIA In Irradiation 19.75 526 2,662 Rods Uranium (LEU) Position 8.4 Radioactive Waste Management Program MURR has a comprehensive Radioactive Waste Management Program that supports operation of the reactor, its ancillary facilities, and their utilization programs.

All radioactive waste materials released from the facility through the Ventilation and Air Treatment System, the Radioactive Liquid Waste Retention and Disposal System, and the Solid Radioactive Waste Program are identified, assessed, and released or disposed of in conformance with all applicable regulations and in a manner that protects the health and safety of the general public and the environment.

It is the policy of the reactor facility to keep the volume of waste materials being generated to the absolute minimum by the efficient use of experiment materials, by the use of proper techniques, and by any other means available.

It is not anticipated that additional radiation safety training programs will need to be developed for this portion of the License Amendment request for the irradiation of the LEU target rods. Operations personnel that will be responsible for and perform the actual handling of the individual target rods and assemblies are well versed in the underwater manipulation of materials as demonstrated by the years of successful handling of fuel elements essential to the operation of MURR over the past 50 years of operational experience.

At this time, minimal additional Health Physics (HP) procedures are anticipated to need development for the irradiation phase of the SGE target rods. Current reviewed and approved HP procedures related to the operation of the reactor are deemed to be sufficient to protect both the staff and general public during the irradiation phase of this project as it is not significantly different than the actual current operation of the reactor. Any permanently installed radiation monitoring equipment at the reactor facility specifically installed to support the SGE TEP is discussed In detail in Section 4. Any HP-related radiation monitoring equipment Page 132of190 ATTACHMENT 1 used at the MURR is much the same as that described in the MURR SAR. Little additional equipment is anticipated to be needed in order to support the addition and operation of the SGE TEF. Additionally, no changes are anticipated for the MURR HP monitoring and survey program as a result of the irradiation of the target rods nor are any significant number of additional records anticipated to be generated as a result of this activity.

Shielding is the paramount design feature used in controlling radiation exposure during operation of the reactor. Shielding has been installed to keep radiation levels in areas occupied by all personnel ALARA. All shielding thicknesses are based on an operating power level of 10 MW. Fuel storage and handling requirements are based on 40-day continuous operation at 10 MW prior to shutting down and removing :fuel. With over 50 years of operational history, the installed radiation shielding has performed more than adequately as designed and analyzed.

This same shielding will be used to provide radiation exposure protection from the two (2) target assemblies that will be operating in the graphite reflector region directly outside the reactor pressure vessel and beryllium reflector region. It should be noted however that one TEF assembly (11 target rods) used for the production of Mo-99 via the SGE process contains approximately 263 grams of U-235; approximately 30% of the U-235 contained in one single fresh MURR fuel element. Thus, two (2) full target assemblies with 11 target rods each would contain approximately 7.5 % of the U-235 that is contained in the reactor core with eight (8) fresh fuel elements.

This does not count the fuel elements in the reactor pool that are being cooled for either reuse within the reactor or awaiting sufficient cooling and decay for shipment as spent fuel. Thus the addition of the two (2) target assemblies adds less than 7.5% to the entire source term that is contained in the pool due to the use and storage of fuel elements located there. Similar calculations were performed using the computer program MicroShield to predict the dose resulting from the movement of a target cartridge to a position underneath the pool water surface that allows handling and removal of the individual target rods for eventual placement into the transfer cask. These doses are summarized in Table 36 below and provide the worst-case estimate of exposure rates from target cartridge movement activities (e.g., one hour after the end of a three week irradiation of 22 target rods). MURR staff is not expected to be present in the fields given in Table 36 below on a routine or regular basis. Target cartridge movement activities are anticipated to happen twice a week for periods not greater than 30 minutes per occurrence.

It is also expected that each handling activity will be rotated amongst several staff members throughout the year. The maximum expected dose rates are presented in Table 36. Page 133of190 ATTACHMENT 1 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI Dose Location Surface Exposure Rate 1 foot Exposure Rate 1 Exposure Rate (mR/hr) (mR/hr) (mR/hr) Reactor Pool 127 100 61 Biological Shield at 18 13 6.5 Handling Station In conclusion, it is not apparent that any additional major changes are required to the MURR RPP due to the addition of the irradiation facilities that will be used for the SGE TEF. 9. Conduct of Operations

9.1 Procedures

Standard Operating Procedures (SOP) will be developed that are specific for operation and maintenance of the SGE TEF and also reactor operation with the TEF. The following topics will be incorporated into procedures for TEF systems installation, testing, operation, and maintenance.

  • TEF component/systems installation and testing
  • TEF startup plan
  • Target cartridge loading, unloading, and in-pool transfer
  • Target rod receiving, inspection, and storage
  • TCS operation
  • TEF preventative maintenance and inspection
  • Surveillance procedures required by the Technical Specifications Table 37 lists the SOPs that are expected to be used for routine operation of the facility.

Page 134 of 190 ATTACHMENT 1 Table 37 Standard Operating Procedures

' ' " No: Title '. Cori tents ,. ,, ';;* ,, ' : >> "" ', ,*' , " ',, ' '* ',,

  • Receiving and inspection of equipment
  • Installation checkout of Water Cooling Module
  • Piping installation 1 TEF Installation and Checkout
  • Welding and inspection procedures
  • Instrumentation and control installation and checkout
  • Target housing installation
  • Cold commissioning with filler rods
  • Safety interlock checkout 2 TEF Commissioning and Startup
  • Hot commissioning
  • Hot acceptance procedure
  • Checklists
  • Transfer from storage
  • Cartridge loading 3 Target Rod Loading and
  • Target assembly loading and unloading Unloading
  • Cartridge unloading and temporary cooling
  • Checklists
  • Material control procedure Target Rod Receiving, Inspection
  • Unpacking and inspection 4 and Storage
  • Acceptance criteria
  • Storage procedure
  • Pre-operational checkout
  • Startup procedure 5 TEF Normal Operations
  • Monitors and data recording
  • Safety interlocks and alarms
  • Shutdown and standby
  • Routine maintenance and inspection
  • Planned equipment replacement 6 TEF Maintenance and Inspection
  • Equipment performance monitoring
  • Instrumentation checkout and calibration

9.2 Target

Experimental Facility Commissioning Plan A high-level commissioning plan is presented in this section. A detailed startup and commissioning plan will be prepared and approved by the Reactor Safety Procedure Review Subcommittee (a subcommittee of the Reactor Advisory Committee) prior to startup of the TEF. Appropriate hold points will be identified in the plan for review of data and for approval to proceed to the next step of the startup plan. Page 135 of 190 ATTACHMENT 1 9.2.1 Assumptions

  • Secondary Coolant System modifications complete
  • TCS installation complete outside of the reactor pool
  • Interface with the MURR reactor safety system ready to be connected
  • Experimental Facilities ready to be installed
  • Both filler and target rods on hand 9.2.2 Phase 1: Installation and Initial Testing with One TEF Target Assembly
  • Reactor shutdown
  • Reactor startup with reference core configuration to 50 kW (log ECP control blade heights)
  • Reactor shutdown
  • Remove current graphite reflector element from position No. SA
  • Install one new experimental facility target housing o Connect forced cooling loop* o Connect interface to reactor safety system o Install cartridge with 11 filler rods loaded o Complete coolant system testing including all reactor safety system Technical Specification surveillances
  • After all cooling testing is complete, remove cartridge with 11 filler rods
  • Reactor startup to 50 kW o Calculate reactivity worth with new facility experimental facility target housing
  • Reactor shutdown
  • Install cartridge with 11 filler rods
  • Make coolant system adjustments due to flow restriction changes
  • Reactor startup to 50 kW <' o Calculate reactivity worth of cartridge and 11 filler rods
  • Continue reactor startup to 10 MW with forced flow and 11 filler rods 9.2.3 Phase 2: Loading of One Cartridge with 3 Target Rods and 8 Filler Rods
  • Reactor shutdown Page 136of190 ATTACHMENT 1
  • Reactor startup with reference configuration to 50 kW with all filler target rods
  • Reactor shutdown
  • Reactor startup with 3 target rods to 10 kW with thermocouple.

Check thermocouple temperature and target thermal calculations.

Verify excess reactivity, shutdown margin, and secured experiment reactivity.

  • Increase power to 50 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 100 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 250 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 500 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 1 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 2 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 5 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 10 MW -check thermocouple temperature and target thermal calculations
  • Operate at 10 MW all week with 3 target rods and 8 filler rods
  • Visually inspect each target rod at the end of the week after removal 9.2.4 Phase 3: Loading of One Cartridge with 11 Target Rods
  • Reactor shutdown
  • Reactor startup with reference configuration to 50 kW with all filler target rods
  • Reactor shutdown
  • Reactor startup with 11 target rods to 10 kW with thermocouple.

Check thermocouple temperature and target thermal calculations.

Verify excess reactivity, shutdown margin, and secured experiment reactivity.

Vary coolant inlet temperature between max and min temperatures to verify effect on reactivity.

  • Increase power to 50 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 100 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 250 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 500 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 1 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 2 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 5 MW -check thermocouple temperature and target thermal calculations Page 137of190 ATTACHMENT 1
  • Increase power to 10 MW -check thermocouple temperature and target thermal calculations
  • Operate at 10 MW all week with 11 target rods
  • Visually inspect each target rod at the end of the week after removal 9.2.5 Phase 4: Installation of Second TEF Target Assembly and Initial Testing and Conduct Full 22-Target Rod Experiment Run
  • Reactor shutdown
  • Reactor startup with reference configuration to 50 kW
  • Reactor shutdown
  • Remove current graphite reflector element from position No. SB
  • Install second new experimental facility with filler target rods in both target cartridges o Connect forced cooling loop o Connect interface to reactor safety system o Install cartridge with 11 filler rods loaded o Complete coolant system testing including all reactor safety system Technical Specification surveillances
  • After all cooling testing is complete, remove cartridge with 11 filler rods
  • Install both cartridges with 22 target rods
  • Reactor startup to 10 kW with thermocouples.

Check thermocouples temperature and target thermal calculations.

Verify excess reactivity, shutdown margin, and secured experiment reactivity.

  • Increase power to 50 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 100 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 250 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 500 kW -check thermocouple temperature and target thermal calculations
  • Increase power to 1 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 2 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 5 MW -check thermocouple temperature and target thermal calculations
  • Increase power to 10 MW -check thermocouple temperature and target thermal calculations
  • Operate at 10 MW all week with 22 target rods
  • Visually inspect each target rod at the end of the week after removal Page 138of190 ATTACHMENT 1 9.3 Material Control & Accounting The LEU target rods will be controlled and accounted for under MURR's existing material control and accounting procedures.

The fresh target rods will be stored in the same storage area as reactor fuel. Irradiated target rod material control will be discussed in the Part 2 License Amendment application.

10. Target Experimental Facility Accident Analyses Accidents affecting the SGE TEF while it is in the reactor pool can occur either during reactor operation or after reactor shutdown when the irradiated target rods are being transferred in the reactor pool. Accidents can be initiated by a failure either in the TEF or in the MURR reactor facility.

Potential accidents initiated by the TEF include breach of target rod cladding, loss of target coolant from a pipe !break, loss of target cooling flow, or mishandling of the target. Accidents initiated by the reactor facility include insertion of excess reactivity, loss of primary coolant flow, loss of primary coolant, loss of pool coolant or loss of offsite electrical power. The breach of target rod cladding has been identified as the TEF maximum hypothetical accident (MHA). 10.1 Target Experimental Facility Maximum Hypothetical Accident An accident resulting in a significant release of fission products to the environment from operating the SGE TEF is considered highly improbable.

The redundant safety measures, strictly controlled quality and administrative procedures in the design, fabrication and operation of the facility give high confidence in the improbable nature of such a release. Nevertheless, inadvertent physical damage to the TEF from manufacturing defects or damage from handling must be considered and analyzed for licensing purposes with a defined, enveloping event that can be used for bounding the radiation hazard. For the TEF, such an enveloping event, the MHA, is a postulated breach of a target rod cladding during full power operation.

The initiating event for this is hypothesized to be caused by a defect in the cladding or a defect in either of the welds that attach the upper and lower end fittings.

It is further postulated that this accident occurs at the worst possible time at the end of an approximate three-week irradiation which maximizes the fission product inventory in the gas gap between the target material and the cladding.

L0.1.1 Description of Maximum Hypothetical Accident Scenario The scenario for the MHA is a postulated breach from a defective weld or cladding at the end of a week irradiation.

The fission gas in the void volume of the target rod is released to the target cooling water which is then discharged to the reactor pool water, then into the reactor containment building, and ultimately a release to the outside environment.

During operation, fission of the LEU U0 2 target material produces a variety of fission products including the fission gases krypton, xenon, and iodine. These fission gases can diffuse out of the U0 2 target material and into the void spaces within the target rod. As the target rod bums its U-235, there is an accumulation of fission gases in these void spaces. A breach of a target rod at the end of its three-week design life will release this accumulated fission gas inventory.

Originally, the helium pressure inside the Page 139 of 190 ATTACHMENT 1 target rods is 101 kPa (14.7 psia). Temperature, thermal expansion and fission gas release raise the pressure to 550 kPa (80 psia). Therefore, a breach of the target rod rapidly releases the fission gap gas into the target cooling water which is then discharged to the reactor pool immediately above the target rods. 10.1.2 Source Term The radionuclide inventory in the target rods at the end of irradiation was calculated by using MCODE, a code to couple MCNP6 and ORIGEN2, to predict isotopic concentration at the end of irradiation.

A Once Through Source Tenn Model (OTSTM) was developed to predict the evolution of radionuclides after irradiation (Reference 35). This program uses an analytic solution to a generalized Bateman equation to calculate the nuclide inventory as a function of time. In addition, OTSTM calculates the decay heat from the actinides and fission products according to ANSI/ANS-5.1-2014, "Decay Heat Power in Light Water Reactors" (Reference 36), and the fraction of the volatile nuclides that may be released from a target rod gap ifthe cladding is breached according to ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel" (Reference 37). The subsequent decay of 1307 nuclides as they move through the TEF was also modeled using OTSTM. In the OTSTM fission products diffuse out of the U0 2 matrix and into the pellet-cladding gap during irradiation.

This "gap gas" can then escape into the atmosphere if the cladding is breached during an accident and is the bases for the MHA source term. The source term for the gap gas is calculated using the factor of 5X conservatism on the release to birth ratios, as recommended in ANSI/ANS-5.4-2011 (Reference

37) and to be conservative, the source term for the gap gas is calculated using power density and temperature profiles in the target rod at the beginning and the bumup at the end of a three-week irradiation.

The temperature is highest at the beginning because the power density is highest in the fresh pellets and the bumup is highest at the end, giving the highest pellet surface to volume ratio. In addition, the release fraction for the hottest, highest power target rod is used for calculating the gap gas for all the rods in a set. The total activity of all isotopes of iodine, krypton and xenon in the gap gas of the hottest target rod is 396 Ci at 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (End of Irradiation).

Note that the sum of the activities of the individual nuclides in Table 38 below is slightly less than the total activity because Table 38 does not include some short lived nuclides with minor radiological consequence.

The activity in a 22 target rod set is obtained by multiplying the activity in the hottest rod by 21.45, which is the ratio of the total fission power in a 22 rod set to the fission power in the hottest rod. Note that most of the fission products are volatile during irradiation due to the very high temperatures inside the rods. However, we assume that only isotopes of iodine, krypton and xenon are volatile after irradiation or upon target rod cladding failure because the rods are then much cooler. Nonetheless, the decay and progeny of non-volatile nuclides, such as Te-132, does contribute to the activities shown in Table 38. Page 140of190 ATTACHMENT 1 More detailed information about the source tenn calculation, including the input and output files for the codes used, can be found in GA Report 30441R00022, "Source Term Analysis Design Calculation Report" (Reference 38). However, for ease of reviewability the source tenn output applicable to the MHA is repeated in Table 38. Table 38 Activity of Volatile Fission Products in the Gap Gas of the Hottest Target Rod Fission Product Half-life Activity at EOI Activity at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (Ci) (Ci) Kr-85m 4.481 h 5.9623 2.3670 Kr-85 10.72 y 0.0316 0.0316 Kr-87 1.272 h 6.2891 0.2400 Kr-88 2.839 h 12.0723 2.7903 Kr-89 3.15 m 2.2661 0.0000 Kr-90 32.32 s 0.8907 0.0000 1-131 8.041 d 26.2934 25.8531 1-132 2.3 h 144.6951 84.4133 1-133 20.8 h 45.7961 37.9656 1-134 52.6m 26.0561 0.7470 1-135 6.611 h 30.5511 16.2867 Xe-133 5.245 d 64.8080 64.0750 Xe-135 9.089 h 8.1066 13.3210 Xe-135m 15.29 m 5.5078 2.6087 Xe-137 3.818 m 3.0510 0.0000 Xe-138 14.08 m 5.9110 0.0000 Xe-139 39.68 s 1.0359 0.0000 Totals: 389.3242 250.6993 :L0.1.3 Dose Assessment The methodology applied for the dose assessment calculations presented throughout this section are based on the fueled experiment failure methodology submitted by MURR to the NRC as part of relicensing efforts. JL0.1.4 Radionuclide Concentration in Reactor Pool Water A breach of the target rod with the highest power generation at the end of a three-week irradiation cycle will release the iodine, krypton and xenon gap activities shown in Table 39 into the MURR reactor pool. Page 141 of 190 ATTACHMENT 1 Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors.

However, for the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5 minute evacuation period is used as the basis for the stay time in the dose calculations for personnel that are in containment during target failure. Table 39 Iodine and Noble Gas Activities Released to MURR Reactor Pool Activity (uCi) 1311 -2.63 x 1o+o 7 85 Kr -3 .16 x 10+o 4 133Xe -6.48 x lO+O? 1321 -1.45 x 10+o 8 85 mKr -5.96 x 10+06 135 Xe -8.11x10+06 1331 -4.58 x 10+o 7 s1Kr -6.29 x 10+o 6 135 mXe -5.51 x 1 O+o 6 134 1-2.61x10+o 7 ssKr -1.21 x 10+o 7 137 Xe -3.05 x 10+06 135 1 -3.06 x 10+o 7 89 Kr -2.27 x 1 O+o 6 13sXe _ 5.91 x 10+06 9°Kr -8.91 x 10+o 5 139 Xe -1.04 x 10+06 The iodine released into the reactor pool over a 5 minute period is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 I) of bulk pool water, which then results in the following pool water concentrations for the iodine isotopes (Table 40). The water solubility of the krypton and xenon noble gases released into the pool over this same time period is conservatively ignored. The gas bubble rise time in the reactor pool from where the TEF is situated in the graphite reflector region to the pool surface has been measured at 17 seconds, so it is assumed the earliest any of the radioisotopes from the TEF, including radioiodine nuclides, enters into the containment building air volume is after 17 seconds. It is also assumed when the radioactivity enters the containment air volume it instantaneously forms a unifonn concentration in the isolated containment structure.

Table 40 Iodine Concentrations in Pool Water Concentration 1311 -1.31 x 10+o 3 1321 -7.23 x 10+o 3 1331 -2.29 x 10+o 3 1341 -1.30 x IO+o 3 1351 -1.53 x 10+o 3 Page 142of190 ATTACHMENT 1 10.1.5 Radionuclide Concentration in Containment When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed iodine concentrations evaporates into the containment building over the 5 minute period. Containment air with a temperature of 75 °F (23.9 °C) and 100% relative humidity contains H 2 0 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of20 gallons (75.7 L) of water vapor will not cause the containment air to become supersaturated.

It is also conservatively assumed that all of the iodine activity in the 20 gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following iodine concentrations in the 225,000 ft 3 (6,371.3 m 3) containment air volume: Example calculation of the average 131 1 concentration released into the containment air during the first minute: 131 1 concentration in pool water/gal x 20 gal/5 minx 0.5 minx EXP(-0.693 x (17+30 sec) I (8.02 day x 8.64 x 10+04 sec/day))

I (225,000 ft 3 x 28,317 ml/ft 3) 1.31 x 1 O+o 3 µCi/gal x 2 gal x 0.99995 I (6.371 x 10+09 ml) 4.13 x 10-07 µCi/ml The same calculation is used for the other iodine isotopes listed below and was performed for each I-minute interval in the 5 minute period. The average radioiodine concentrations over the 5 minute evacuation period is the average of the five 1-minute intervals in the 5 minute period (Table 41). Table 41 Average Iodine Concentrations in the Containment Building Air During the 5 Minute Evacuation Period Concentration

<uCi/cc) 1311 -2.06 x 10-06 1321 -1.12 x 10-05 1331 -3.59 x 10-06 1341 -1.95 x 10-06 1351 -2.3 8 x 10-06 As noted previously, the krypton and xenon gases released into the reactor pool from the SGE TEF during the 5 minute evacuation period following a cladding failure are assumed to have no absorption in the pool Page 143of190 ATTACHMENT 1 water, rise through the pool in 17 seconds (thus slightly decaying) and enter the containment building air volume where they are assumed to instantaneously form a uniform concentration in the isolated structure.

Based on the 225,000-ft 3 volume of containment building air, and the previously listed curie quantities of these gases released into the reactor pool, the average noble gas concentrations in the containment structure over the 5 minute evacuation period would be calculated as follows: Example calculation of the average 85 Kr concentration in containment during the first minute after the gas enters the containment air: 85 Kr activity x EXP(-0.693 x (17+ 30 sec) I (10.76 yrs x 3.156 x 10-t-0 7 sec/yr)) I (225,000 ft 3 x 28,317 ml/ft 3) 3.16 x 10+04 µCi x 0.99999 I (6.371 x 10+09 ml) 4.96 x 1 o-06 µCi/ml The same calculation is used for the other krypton and xenon isotopes listed below in Table 42 and was performed for each I-minute interval in the 5 minute period. The average concentrations over the five minute evacuation period are the average of the five I-minute intervals in the 5 minute period. Table 42 Average Noble Gas Concentrations in the Containment Building Air during the 5 Minute Evacuation Period Concentration (uCi/cc) 85 Kr -4.96 x I o*06 133 Xe -1.02 x 10*02 85 mKr -9 .29 x I 0*04 135Xe _ I.27 x Io-03 87 Kr -9.63 x 10*04 J35mxe -7.64 x 10-04 88 Kr -I .87 x I 0*03 137 Xe -2.99 x Io-04 s9Kr -2.02 x I0-04 138 Xe -8. I I x I 0-04 90Kr -l.4I x 10*05 139 Xe -2.I9 x I0-05 10.1.6 Dose Assessment in Restricted Area The objective of this calculation is to present a worst-case occupational dose assessment for an individual who remains in the containment building for 5 minutes following the MHA. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air. Page I44 of I90 ATTACHMENT 1 Based on the source term data provided, it is possible to detennine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building.

Because the airborne radioiodine source is composed of five different iodine isotopes, it 'Yill be necessary to detennine the dose contribution from each individual isotope and to then sum the results. Dose multiplication factors were established using the Derived Air Concentrations (DACs) for the "listed" isotopes in Appendix B of 10 CFR 20 and calculated values for the "unlisted" submersion isotopes (Kr-89, Kr-90, Xe-137, and Xe-139). The submersion DAC values that were calculated were done in accordance with the data and methodology as supplied in Federal Guidance Report No. 12. Example calculation of thyroid dose due to 131 1: The DAC can also be defined as 50,000 mrem (thyroid target organ limit) I 2,000 hr, or 25 hr. Additionally, 5 minutes of one DAC-hr is 8.33 x 1 o-02 DAC-hr. 131 1 concentration in containment

= 2.06 x 10-06 µCi/ml 131 1 DAC (10 CFR 20) = 2.00 x 10-08 µCi/ml Dose Multiplication Factor = (131 1 concentration)

I (131 1 DAC) = (2.06 x 10-06 µCi/ml) I (2.00 x 10-08 µCi/ml) = 1.03 x 1 o+-0 2 Therefore, the 5-minute thyroid exposure from 131 1 is: = Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr = 1.03 x 10+02 x (25 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr) = 2.15 x l0+-0 2 mrem The same calculation sequence was repeated for all iodine isotopes and is summarized in Table 43. Page 145of190 ATTACHMENT 1 Table 43 Derived Air Concentration Values and 5-Minute Exposures for Iodine Radionuclide Derived Air Concentration 5-Minute Exposure (µCi/ml) (mrem:-CDE) 1311 2.00 x 10-08 2.15 x 10-t-0 2 1321 3.00 x 10-06 7.74 x 10-1-0° 1331 1.00 x 10-01 7.47 x 10-t-0 1 1341 2.00 x 10-05 2.03 X 10-0l 1351 7.00 x 10-01 7.09x lO+oo Doses from the krypton and xenon radionuclides present in the containment building are assessed in much the same manner as the iodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes in containment were based on their average concentration in the containment air and the corresponding DAC. The whole body dose due to 85 Kr is calculated as follows: The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mrem/DAC-hr.

Additionally, 5 minutes of one DAC-hr is 8.33 x 10-02 DAC-hr. 85 Kr concentration in containment 85 Kr DAC (10 CFR 20) = 4.96 x 10-06 µCi/ml 1.00 x 10-04 µCi/ml Dose Multiplication Factor (85 Kr concentration)

I (85 Kr DAC) (4.96 x 10-06 µCi/ml) I (l.00 x 10-04 µCi/ml) = 4.96 x 10-02 Therefore, a 5 minute whole body exposure from 85 Kr is: = Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr = 4.96 x 10-02 x (2.5 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr) = 1.03 x 10-02 mrem The same calculation sequence was repeated for all noble gases and is summarized in Table 44. Page 146of190 ATTACHMENT 1 Table 44 Derived Air Concentration Values and 5 Minute Exposures

-Noble Gases Radionuclide Derived Air Concentration 5-Minute Exposure (µCi/ml) (mrem-CEDE) s5Kr 1.00 x 10-04 1.03 x 10-02 85mKr 2.00 x io-05 9.68 x lO+oo s1Kr 5.00 x 10-06 4.01 x 10+o 1 ssKr 2.00 x 10-06 1.95 x 10+o 2 s9Kr 1.90 x 1 o-06 2.22 x 10+o 1 90Kr 2.80 x io-06 1.05 x 1 o+oo 133 Xe 1.00 x 10-04 2.12 x 10+o 1 135 Xe 1.00 x 10-05 2.64 x 10+o 1 I35mxe 9.00 x 10-06 1.77 x 10+o 1 137 Xe 2.00 x io-05 3.11 x lO+oo 13sXe 4.oo x io-06 4.22 x 10+o 1 139Xe 3.70 x 10-06 1.24 x lO+oo To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 5 minute exposure in the containment building during the MHA, the doses from the iodines and noble gases must be added together.

This summation results in the following TEDE values presented in Table 45. Table 45 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building Iodine -Committed Dose Equivalent (Thyroid) 304 mrem Iodine -Committed Effective Dose Equivalent (CDE x 0.03) 9mrem Noble Gas -Committed Effective Dose Equivalent 380 mrem Total Effective Dose Equivalent 389 mrem By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for those exposed during a TEP failure and the MHA to applicable NRC dose limits in 10 CPR 20, the final values are shown to be well within the published regulatory limits, in fact, less than 10% of any occupational limit. Page 147of190 ATTACHMENT 1 10.1.7 Dose Consequences to Members of the Public As noted earlier in this chapter, the containment building ventilation system will shut down and the containment building will be isolated from the surrounding areas upon actuation of the isolation system. A breach of target rod cladding will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere.

It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure change has occurred in conjunction with the target failure. A reasonable assumption would be a pressure change on the order of0.7 inches of Hg (25.4 mm of Hg at 22 °C), which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment structure.

Making the conservative assumption that the containment building will leak at the Technical Specification leakage rate limit, I 0% of the contained volume over a 24-hour period from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the air leakage from the containment structure in standard cubic feet per minute (scfm) as a function of containment pressure can be expressed by the following equation:

LR 17 .68 x (CP-14. 7)112; where: LR leakage rate from containment (scfm); and CP containment pressure (psia). The minimum Technical Specification free volume of the containment building is 225,000-ft 3 at standard temperature and pressure.

At an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the containment structure would hold approximately 255,612 standard cubic feet (set) of air. When applying the Technical Specification leakage rate equation to the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the leak rate to decrease to zero from an initial leakage rate of approximately I 0.25 scfin, which would occur at the start of the event. Determination of the average leakage rate is subdivided into multiple intervals within the total 16.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> leak duration to provide a m'ore accurate calculation of release concentrations from the facility using the Technical Specification leakage rate equation.

The average containment building leakage rate for each of the first five (5) I-hour intervals and for the following three (3) 4-hour intervals is provided below in Table 46. Page 148of190 ATTACHMENT 1 Table 46 Average Containment Building Leakage Rate Hours: 0-1 1-2 2-3 3-4 4-5 5-9 9-12 12-16.5 scf/hr: 595.6 558.7 521.8 485.0 448.1 355.8 207.9 67.9 Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following target failure. First of all, most leakage pathways from containment discharge into the reactor laboratory building, which surrounds the containment structure.

Since the laboratory building ventilation system continues to operate during target failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 cfm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in a containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75% of the iodine released will be deposited in the containment vessel. For the purposes of this analysis a conservative value from Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," was used and provides a reduction factor of 50% for plate-out and deposition of iodine within the containment building.

Thus, due to this 50% iodine deposition in the containment building, each cubic foot of air released from containment has a iodine concentration that is fifty percent of each cubic foot within the containment building air. Example calculation of average 131 1 concentration in containment during the first hour: (131 1 concentration in pool water/gal x 20 gal) x EXP[(-0.693 x 0.5 hr) I (131 1 T 112)] I (scf in containment)

(1.31 x 1 O+o 3 µCi/gal x 20 gal) x 0.99820 I (229,800 scf) 1.14 x 10*01 µCi/scf The average 131 1 concentration in containment over the first hour is then further reduced by the 50% iodine plate-out reduction factor prior to escaping from containment at the average leak rate and further diluted in 30,500 scfm oflaboratory building exhaust ventilation prior to being released from the facility.

Page 149of190 ATTACHMENT 1 Example calculation of average 131 1 concentration in air released from the facility during the first hour: (average 131 1 concentration in containment

µCi/scf x average leak rate x 0.5 reduction factor) I (30,500 scfm x 60 min/hr x 28,317 ml/scf) (1.14 x 10*01 µCi/scfx 595.6 x 0.5) I (5.18 x 10+10) 6.56 X 10-IO µCi/ml The average of first five (5) 1-hour intervals and the following three (3) 4-hour intervals is then summed and averaged over the entire 16.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> duration to calculate an average of the air concentration released from the facility.

The same calculation is used for the other radioiodines and is provided in Table 47. Table 47 Average Iodine Concentrations in Air the Exhaust Stack . Activities

. ,, , . ,, . (uCi/mly.

'<** 1311 -3.38 X 10-IO 1321 -5.89 x 10*10 1331 -5.01 x 10*10 1341 -4.60 x 10*11 1351 -2.39 x 10" 10 Calculation of nobles gases released through the exhaust stack is identical to the calculation of the radioiodines released except no credit is taken for activity absorbed by pool water or reduction via out inside of the containment building.

Example calculation of average 85 Kr concentration in containment during the first hour: 85 Kr activity x EXP[(-0.693 x 0.5 hr) I (85 Kr Tll2)] I (scf in containment) 3.16 x 10+o 4 µCi x 0.99999 I (229,800 scf) 1.38 x 10*01 µCi/scf Example calculation of average 85 Kr concentration in air released from the facility during the first hour: (average 85 Kr concentration in containment

µCi/scf x average leak rate) I (30,500 scfm x 60 min/hr x 28,317 ml/scf) (1.38 x 10-01 µCi/scf x 595.6) I (5.18 x 10+10) 1.58 x 10-09 µCi/ml Page 150of190 ATTACHMENT 1 The average of the first five (5) 1-hour intervals and the following three (3) 4-hour intervals is then summed and averaged over the entire 16.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> duration to calculate an average of the air concentration released from the facility.

The same calculation is used for the other noble gasses and is provided in Table 48. Table 48 Noble Gas Concentrations in Air Exiting the Exhaust Stack Activities

<uCi/ml) 85 Kr -8.29 x 10*10 133 Xe -1.65 x 10-06 85 mKr -7 .67 x 1 o-08 135 Xe -1.44 x 10-01 87Kr -3.13 x 10-08 135 mxe -4.57 x 10*09 88 Kr -1.16 x 10*07 137 Xe -4.00 X 10-ll 89 Kr -9.34 x 10*12 138 Xe -4.31x10" 09 9°Kr -4.56 x 10*26 139Xe _ 7.10 x 10*23 l0.1.8 Dose Assessment in Unrestricted Area Assuming that (1) the above provided average leak rates from the reactor containment building, (2) the leak continues for about 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292 (explained below) and (5) there is no decay of any iodines or noble gases, then the following concentrations of iodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are at the point of maximum concentration in the unrestricted area assuming uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following target failure. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> that the containment building was leaking. A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis.

It is assumed that all off site (public) doses occur under these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. This point conservatively assumes a Stability Class F; which normally occurs only 11.4% of the time when the wind blows from the south. Thus, this calculation is conservative.

Page 151 of 190 ATTACHMENT 1 10 CPR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes.

Effluent Concentration Limits were calculated for each of the four (4) "unlisted" noble gases (Kr-89, Kr-90, Xe-137 and Xe-138) using the data and methodology contained in Federal Guidance Report No. 12 for submersion isotopes.

The DAC value was first calculated and then a factor of 219 was applied using 10 CPR 20, Appendix B, methodology for effluent values from submersion isotopes.

Exposure at 1 DAC equates to 5000 mrem per year whereas the Effluent Concentration Limit is 50 mrem per year. Thus, there is a factor of 100 times lower allowable dose for the Effluent Concentration Limit as compared to the DAC. Exposure at the effluent concentration limit assumes a person is in that effluent concentration

  • for 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year. Thus, the time assumed for exposure to the effluent concentration limit is a factor of 4.38 longer than the 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year that defines a DAC. No credit is taken for transit time from the stack to the receptor point. In the case of Kr-89 and Xe-137 the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four ( 4) lives. Example calculation of whole body dose in the unrestricted area due to 131 1: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hours) = 5. 71 x 10*03 mrem/hr 131 1 concentration 131 1 effluent concentration limit 131 1 Conversion Factor = 3.38 x 10-lO µCi/ml = 2.00 x 10*10 µCi/ml = 5.71 x 10*03 mrem/hr Therefore, a 16.5-hour whole body exposure from 131 1 at the maximum receptor site is: = (131 1 concentration I 292 dilution factor) I (1 31 1 effluent concentration limit) x (Conversion Factor x 16.5 hrs) = (3.38 x 10-10 µCi/ml/ 292) I (2.00 x 10*10 µCi/ml) x (5.71 x 10*03 mrem/hr x 16.5 hrs) = 5.45 x 10*04 mrem The same calculation is used for the other isotopes (iodines and noble gases) and results are listed in Table 49 and Table 50. Page 152of190 ATTACHMENT 1 Table 49 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Iodine Effluent Limit Concentration 1 at Radiation Dose Radioiodine

(µCi/ml) Maximum Receptor Site (mrem) (J!Ci/ml) 1311 2.00 X 10-lO 1.16 x 10-12 5.45 x 10-04 1321 2.00 x 10-08 2.02 x 10*12 9.51 x 10-06 1331 1.00 x io-09 1.72 x 10-12 1.62 x io-04 1341 6.00 x 10-08 1.58 x 10-13 2.48 x 10*07 1351 6.oo x io-09 8.18 x 10*13 1.28 x 10-05 Total = 0.000729 Note I: Concentrations are the average radioiodine concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. Table 50 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases Effluent Limit Concentration 1 at Radiation Dose Radioisotope

(µCi/ml) Maximum Receptor Site (mrem) (lLCi/ml) 85Kr 7.00 x 10*07 2.84 x 10-12 3.82 x 10-07 85mKr 1.00 x 10-07 2.63 x 10*10 2.48 x 10*04 87Kr 2.00 x 10-08 1.07 X 10-IO 5.06 x 10-04 88Kr 9.00 x 10-09 3.97 X 10-IO 4.15 x 10-03 89Kr 2.00 x 10-08 3.20 x 10-14 3.51 x 10-07 90Kr 2.00 x 10-08 1.56 x 10-28 1.23 x 10-21 133 Xe 5.00 x 10-07 5.64 x 10*09 1.06 x 10-03 135 Xe 1.00 x io-08 4.94 x 10-10 6.64 x io-04 135mxe 4.00 x 10-08 1.57 X 10-ll 3.69 x 10*05 137 Xe 2.00 x 10-08 1.37 x 10-13 1.42 x 10-07 138 Xe 2.00 x 10-08 1.48 X 10-ll 6.96 x 10*05 139Xe 2.00 x 10-08 2.43 x 10-25 1.43 x 10-18 Total= 0.00674 Note I: Concentrations are the average noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. Page 153of190 ATTACHMENT 1 To finalize the unrestricted dose in terms of TEDE, the doses from the iodines and noble gases must be added together.

Results are provided in Table 51. Table 51 Dose from Iodines and Noble Gases in the Unrestricted Area Committed Effective Dose Equivalent (Iodine) 0.000729 mrem Committed Effective Dose Equivalent (Noble Gases) 0.00674 mrem Total Effective Dose Equivalent 0.00747 mrem Summing the doses from the noble gases and the iodines simply substantiates earlier statements regarding the low levels in the unrestricted area should a failure of the TEF occur and subsequent containment building leakage following such an event. The overall TEDE is much less than I mrem, a value far below the I 00 mrem regulatory limit within I 0 CFR 20 for the unrestricted area. 10.1.9 Radiation Shine through Containment An evaluation of radiation shine through the containment building has also been performed to ensure whole body doses are of minimal concern during an accident scenario.

The radiation shine model uses MicroShield Ver. 8.02 (Grove Software) for shielding and exposure rate calculations to evaluate radiation shine through the containment structure under accident conditions, and to determine dose consequences to the public and MURR staff. Calculations of exposure rate from the TEF failure were performed using a Rectangular Volume -External Dose Point geometry for the representation of the containment structure.

The exposure rate values provided below represents the radiation fields at 1 foot (30 cm) from a 12-inch thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ) boundary of 150 meters ( 492.1 ft) representing the anticipated exposure rates in the restricted and unrestricted area respectively.

The airborne concentrations used in calculating the exposure rate values were derived from the total activities of the gap gas nuclides provided in Table 52 divided by the containment free volume of 225,000 scf. Therefore, all radiation shine exposure rate values provided are very conservative as solubility of the iodine nuclides in the pool water, and decay of all nuclides is ignored. The source term also assumes a homogenous mixture of the nuclides within the containment free volume. Table 52 provides the exposure rates external to the containment building at 1 foot and 150 meters from the building surface. Table 52 Radiation Shine through the Containment Building Exposure Rate at I-Foot from Containment Building Wall 67 mR/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters) 0.32 mR/hr Page 154 of 190

, L_______ ----ATTACHMENT 1 10.2 Insertion of Excess Reactivity Bec a u s e of (1) the subcriticality of the target cartridges and (2) the physical impossibility of removing a t arget cartridge without scramming the reactor , a reactivity accident associated with the removal or insertion of the target cartridges is not considered credible. Removal and insertion of the target cartrid g es on ly occurs when the MURR reactor is shutdown.

Two (2) different reactivity in s ertion accidents for the MURR reactor core would have an impact on the TA. First , a step in s ertion of positive reactivity based upon the maximum step insertion that the MURR c ore can withstand with no core damage, and second , a continuous ramp insertion of positive reactivit y based on the continuous withdrawal ofMURR's four (4) shim control blades. 10.2.1 Rapid Insertion of Positive Reactivity F or the Insertion of Excess Reactivity accident analysis , the licensed maximum power level of 10 MW was originally used in the reactor SAR as the starting assumption since MURR does not , nor can it l egally , operate above this power level. NUREG-1537 , Part 2 , page 13-9 , Standard R e vi e w Plan and Acce ptanc e Crit e r i a , states for In s ertion of E xcess Reactivity accident , that " Th e a c cid e nt sce nario a ss um e s that th e r e a c tor ha s a ma x imum load of fu e l (c on s i s t e nt with t h e t ec hni c al s p ec ifi c ation s), th e r e a c tor is op e ratin g at full lic e n se d pow e r , and the c ontrol sys t e m ... " The accident was reanalyzed at a m uch more conservative starting power level (11.5 MW) than is required by NUREG-1537 and the results are provided below. 11.5 MW was chosen , instead of the Limiting Safety System Setting (LSSS) set p oint of 12.5 MW , because the rod run-in sy s tem will initiate a rod run-in at 11.5 MW (Technical S pecification 3.2.f. l) and shut down the reactor prior to reaching the LSSS SCRAM set point of 125%. The third paragraph on Page 13-17 of the SAR lists the various reactivity coefficients assumed for the I nsertion of Exces s Reactivity accident analy s is. Fo r both the reactor SAR anal y ses , as well as for the updated analysis presented here , the CB in s ertion t imes are based on the current and relicensing Technical Specification 3.2.c requirement of insertion to t he 20% withdrawn position in les s than 0.7 seconds. The insertion rate was calculated based on the shim CB s travelling from 26 inches (fully withdrawn) to 5.2 inches (20% withdrawn or 80% inserted) in 0.7 s econds. This is a conservative assumption because monthly control blade drop time verifications perfor med at MURR have alwa ys y ielded insertion times of 0.6 seconds or less. S imilar to the reactor SAR analysis , the Reactivity Transient Analysis program PARET (V7.5), maintained and distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) w as used. For the Insertion o f Excess Reactivity accident analysis , two (2) channels were modeled in PARET; a hot channel representing worst-case condition s inside the core and an average channel representing the rest of the core experiencing

" average" conditions. As indicated earlier , the transient was started from an initial power level of 11.5 MW with core coolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3 , 200 gpm and 155 °F , respectively.

Also , pressurizer pres s ure w a s at 75 psia (LSSS value). Since the Insertion of Exce ss Reactivity transient wa s analyzed from a Pa g e 155of190 ATTACHMENT 1 starting power level of 11.5 MW , the rod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only the high power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds was incorporated into the CB scram model so that the CBs would only start to insert 0.15 seconds after the power level had exceeded the s cram s et point of 12.5 MW. The results of a step reactivity insertion of 600 pcm (+0.006 are shown below in Figure 75. As expected , due to the higher starting core power level , much lower core coolant flow rate and much higher than normal core coolant inlet temperature conditions a s sumed for this updated analysis , the peak power during the transient momentarily reaches approximately 37.4 MW compared to a value of approximately 33.0 MW reported in prior SAR analyses for the same 600 pcm step reactivity insertion.

3:;o -l'OWFRMW 100 o.m o 0.00 0.50 1.00 1.50 2.00 2.50 J.00 Time (seconds)

Figure 75 Reactor Power, Fuel and Cladding Temperatures vs. Time for a Positive Reactivity Step Insertion of 0.006 Ak/k The power generation in the TAs would follow the same proportional power transient that the reactor core experiences because the T As are driven by the neutron flux generated by the reactor core. The target rod analysis for the positive reactivity step insertion examines the maximum powered target rod at the beginning and end of a three-week irradiation.

A steady-state analysis of the three-week irradiation was performed using FRAPCON to establish the initial conditions for the transient analysis. The FRAPCON analysis assumed that the first and last day of the three-week irradiation are at 115% power and the rest of the operating period is at 100% power. The FRAPCON analysis also Page 156of190 ATTACHMENT 1 a ssumes that the TA flow rate is 85% of nominal during the first and last day of operation. Mid-week and weekend shutdowns are also included in the power history of the TA. The transient analysis was p erformed using the FRAPTRAN code. FRAPCON and FRAPTRAN are two NRC-sponsored computer c odes that can model the steady-state and transient thermal-mechanical behavior of light water reactor o xide fuel. Phenomena modeled by the codes include heat transfer through fuel and cladding to coolant , c ladding elastic/plastic deformation , fuel-cladding mechanical interaction , fission gas release , rod internal p ressure , and cladding oxidation. The FRAPCON steady-state analysis defines the bumup , U0 2 and cladding deformation , and fission gas r elease that form the starting point for the FRAPTRAN analysis.

The FRAPTRAN analysis increases the p ower to 115% and decreases the flow rate to 85% within 40 seconds. Twenty-one seconds later , the 6 00 pcm (+0.006 ilk/k) step reactivity insertion is simulated. The target cooling water inlet temperature t o the TA was assumed to be at 102 °F (38.9 °C) which corresponds to an outlet temperature at the m aximum pool temperature of 120 °F (48.9 °C) during nominal conditions. The maximum cladding s train occurs with the minimum gap tolerance of . The maximum cladding strain [5 a , b , f] a lso occurs at the end of the due to the effects ofbumup on gap closure and fission g as release. The FRAPCON and FRAPTRAN analysis also use the maximum tolerance on the scallop diameter that forms the flow area of the target cartridge.

This tolerance results in the minimum coolant v elocity consistent with the minimum gap. The analysis also uses the maximum pellet outer diameter , m inimum cladding inner diameter, minimum cladding thickness , and minimum cladding outer diameter.

The strain transient that the cladding is predicted to undergo during the positive step reactivity insertion is s hown in Figure 76. The hoop strain increases from about 0.50% to just under 0.76% but remains under t he 1 % strain limit. M aximum temperatures occur with the maximum gap tolerance of rather than the [5 a , b ,, f] minimum gap tolerance. Bumup effects reduce the gap so that the maximum pellet temperature occurs at t he beginning of irradiation.

The analysis also uses the minimum pellet outer diameter , maximum c ladding inner diameter , maximum cladding thickness , and maximum cladding outer diameter in order to obtain the maximum gap tolerance or to maximize peak pellet temperature. The peak target temperature t ransient is shown in Figure 77. The peak target temperature increases from 2524 °C ( 4575 °F) to slightly l ess than 2729 °C (4944 °F) in less than 0.4 seconds. This peak U0 2 temperature is 111 °C (200 °F) below the melting temperature of 2840 °C (5144 °F). Page 157of190 0.008 0.006 0.004 c ] 0.002 ...:l '; i 0 t E -0.002 -0.004 -0.006 0 0.5 ATTACHMENT 1 u Time (seconds)

Figure 76 2 -Hoop Strain -Axial Strain -Radial Strain 2.5 Cladding Strains at Peak Pellet Location During a Positive 0.006 Ak/k Reactivity Insertion 2800 2700 2600 ,...... u %' 2'.500 .. .e i 2400 8 2300 2200 2100 0 0.5 1.5 Time (seconds)

Figure 77 2 2.5 Peak Target Pellet Temperature During a Positive 0.006 Ak/k Reactivity Insertion Page 158 of I 90 ATTACHMENT 1 The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where t he peak pellet centerline temperature occurs is shown in Figure 78. The cladding outer diameter t emperature shows a modest increase in temperature of around 20 °C (36 °F). The pellet outer diameter a nd cladding inner diameter temperatures become more closely coupled due to closure of the gap between t he pellet and cladding and an increase in interface pressure.

Cladding hoop strain for the -gap at t he end of irradiation goes through a similar transient as for the -gap but increases from 0.14% to j ust under 0 .18%. P eak heat generation within the target pellets is a factor of 3.74 greater than nominal. Heat capacity and t hermal resistances result in the surface heat flux at the peak location being only a factor of 1.34 to 1.50 g reater than nominal , depending on the initial gap size and burnup conditions.

FRAPTRAN assesses c ritical heat flux (CHF) using one of five different CHF correlations. The Macbeth CHF correlation was c hosen because of its validity at low pressures.

The Macbeth CHFR reaches a minimum value of between 2.16 and 2.24 depending on initial gap size and bumup conditions.

The Bernath CHFR reaches a m inimum of 1.67 to 1.87 depending on the case. No target rod damage or radiological release would o ccur from this accident.

700 600 500 .-. u I?_, 400 QI .. .a f! 300 QI i:i.. E 200 100 0 0 0.5 1.5 Time (seconds)

Figure 78 2 2.5 -Pellet OD -Clad ID -Clad OD Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a Positive 0.006 Ak/k Reactivity Insertion Page 159 of 190 [5 a , b , fJ ATTACHMENT 1 10.3 Control Blade Withdrawal As presented in Sectio n 13.2.2.1.2 of the MURR SAR , a positive react1v1ty ramp insertion rate of 0.0003 which i s the Technica l Specification limit on the maximum rate of reactivity insertion for all four (4) shim CBs operating si multaneou s l y, was introduced to the reactor sta rtin g at subcritical cold conditions and at an initial power level of I 0 MW. For su bcritical cold conditions , the short period reactor sc ram terminates the transient within 150 sec before the po wer has reached 64 watts. No target pellet or claddin g dama ge would occ ur at such a low power. For full power conditions , the high power reactor sc ram terminates the transient after 4.53 seconds when reactor power rises from I 0.0 MW to the 12.5 MW hi g h power sc ram set point. The thermal-mechanical per fo nnance of the target rod during this transient was analyzed usin g FRAPCON and FRAPTRAN.

The power transient that drives the heat generation in the target rods is show n in Figure 79. The figure include s I sec of steady-state operation at 100% power and the power transient after reactor SCRAM. I :i! ti c .Q ti f ... -\ 0.1 0 2 3 4 5 6 7 Time (seconds)

Figure 79 Power Transient During a 0.0003 Ak/k per Second Reactivity Insertion The target rod analysis for the CB withdrawal examines the maximum power e d target rod at the beginning and end of a three-week irradiation.

A s tead y-state analy s is of the three-week irradiation was performed at 100% power usin g FRAPCON to establish the initial conditions for the transient analysis.

The FRAPCON analysis also assumes that the TA flow rate is I 00% of nominal during the three-week irradiation.

Mid-week and weekend s hutdo wns are also included in the power history for the TA. Pa ge I 60 of 190 ATTACHMENT 1 The FRAPCON steady-state analysis defines the burnup , U0 2 and cladding defonnation, and fission gas r elease that form the starting point for the FRAPTRAN ana l ysis. The maximum cladding strain occurs with the minimum gap tolerance of The maximum cladding strain also occurs at [5 a , b , f] t he end of the three week irradiation due to the effects of burn up on gap closure and fission gas release. The stra in transient that the cladding is predicted to under go during the control blade withdrawal is shown in Figure 80. The hoop strain increases from about 0.24% to just over 0.56% but remains we ll under the 1 % strain limit. 0.007 0.006 0.005 0.004 = ] 0.003 t'-l 'ii 0.002 15 0.001 E i;... 0 -Hoop Strain -Axial Strain -Radial Strain -0.001 -0.002 -0.003 0 2 3 4 5 6 7 Time (seconds)

Figure 80 Cladding Strains at Peak Pellet Location During a 0.0003 Ak/k per Second Reactivity Insertion For the case using the -gap between the pellet and cladding, the pellet centerline temperature

[5 a , b , f] i ncreases just over 300 °C (540 °F) to 2219 °C ( 4026 °F) at the beginning of irradiation.

Bum up effects r educe the gap so that the maximum pellet temperature decreases durin g irradiation.

The temperature of t he cladding inner diameter increases to just greater than 289 °C (552 °F), and the cladding outer diameter t emperature increases by les s than 9 °C (16 °F). The FRAPTRAN analysis of the gap case predicts the maximum pellet [5 a , b , t] t emperatures due to the higher thermal resistance between the pellet and cladding.

The maximum pellet t emperature increases from 2306 °C ( 4182 °F) to 254 7 °C ( 4617 °F) at the beginnin g of irradiation as s hown in Figure 81. The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where the peak pellet centerline temperature occurs is shown in Figure 82. The pellet outer diameter and cladding inner diameter temperatures become more closely coup led due to closure of the g ap between the pellet and cladding and an increase in interface pressure. The cladding inner diameter Page 161 of 190 ATTACHMENT 1 temperature shows a modest increase in temperature of around 35 °C (63 °F) to just below 321 °C (610 °F). Cladding hoop strain goes through a similar transient as shown in Figure 80 but increases from 0.13% to just over 0.14%. 2600 2550 2500 --u e_, 2450 .a E 2400 GI =-E 2350 2300 2250 0 2 3 4 5 6 7 Time (seconds)

Figure 81 Peak Target Pellet Temperature During a 0.0003 Ak/k per Second Reactivity Insertion Page 16 2 of 190 ATTACHMENT 1 700 600 500 6 e_.. 400 "' ... .a -Pellet OD t 300 =--Clad ID E -Clad OD 200 100 0 0 2 3 4 5 6 7 Time (seconds)

Figure 82 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a 0.0003 Ak/k per Second Reactivity Insertion 10.4 Loss of Target Coolant This accident assumes the double-ended break of one of the pipes in the TCS. The design and operation of the TCS are described in Section 3. Pipe breaks can occur in a variety of locations either in or out of the reactor pool. Depending on the location of the break , the loss of coolant will cause either an increase or a decrease in flow at the flow sensors on either pipe leg supplying cooling water to the two (2) TAs. The hi/low flow set points are at 115% and 85% of nominal flow and result in a reactor scram and opening of the decay heat removal v alves (natural circulation) located just above the refueling bridge level. Pump coast-down after pump trip can mitigate the early phase of the transient unless the offset pipe break prevents pump flow from reaching the TA. Pipe breaks in the reactor pool differ from pipe breaks out of the reactor pool (in air) because the decay heat removal valves are assumed to not operate. A pipe break in the pool establishes a natural circulation loop between the TA and reactor pool regardless of whether the decay heat removal valves open. If the break is in the air , a natural circulation loop is only established if the decay heat removal valves open. Otherwise , the break prevents pool water from reaching the TA inlet and establishing a natural circulation loop. Page 163 of 190 ATTACHMENT 1 The pipe break b efore the wye affec ts both TAs but a pip e break down stream of th e wye at the joint wit h the flexible pipe h as a lo wer cold l eg volume a nd greater impact on the TA. T h e relative lo cations of th ese pipe br eaks are s ho w n in F i g ure 83. The water in the co ld le g of the s uppl y lin e is a lmo s t 18 °F (10 °C) coo ler than the pool water. The drainin g of this wa t er through the TAs miti gates the transient during th e ea rl y pha se of the lo ss of coo lant acc ident (LOCA ). Figure 83 Pipe Break Locations Out of the Reactor Pool Pip e breaks out of and in the reactor pool are covered in the acc ident analyses that follow. 10.5 Pipe Break Locations Out of the Reactor Pool The pipe break in air ju st after the flexible pip e is a more credible pip e break than other locations closer to th e TA becau se the joint is mor e vulnerable and the remainin g pipe i s of welded construction.

The upper bridge structure protects the pipes w here they bend do wnwar d into the reactor pool. The br eak is analyzed to occur in the s uppl y lin e to TA 2 because it h as the target rod wi th the hi ghest power d ensity , and the target rod w ith the hi g h est power in a ddition to bein g the TA with the mo s t po wer. The LOCA for the SGE TEF was modeled u s in g RELAP5 mod3.3 p atc h 03. RELAP5 was developed b y the NRC to analyze thermal-h ydra ulic tran s ient s in pressuri ze d water r eacto r s. It can be u se d to analyze a variety of geometries a nd was used to analyze the LOCA and Loss of Flow Accident (LOF A) in Sections 13.2.3 and 1 3.2.4 of the MURR SAR for relicensing.

T h e RELAP5 mod el for this accide nt Page 16 4 of 190 ATTACHMENT 1 i ncludes both T As , TCS pump and heat exchanger , and decay heat removal valves. The double-ended guillotine rupture is modeled with three (3) valves -two of the valves connect to the reactor pool on either side of the break , and the third valve connects the two ends of the pipe across the break. Prior to accident initiation , the va l ves to the reactor pool are c l osed and the valve across the break is open. When t he break is initiated , the valve connecting the two (2) pipe ends is closed and the two (2) va l ves from each end connected to the reactor pool are opened. A ll three (3) valves are assumed to completely change p osition in 0.5 seconds which is a reasonable assumption for a mechanically induced failure that displaces t he two (2) ends of the pipe since the piping is not at high pressure.

The flow transient caused by a LOCA resulting from a double-ended pipe break is shown in Figure 84. The transient analysis starts with the reactor and T As at 100% power and the TCS at 100% nominal flow t o accurately simulate t h e expected response of the protection system. The decay heat removal valves are assumed to not operate. The break causes an increase in the flow measurement in the pipe supplying TA 2 and a decrease in the flow rate supplying TA 1. The reactor is scrammed at 0.05 seconds when the mass flow rate to TA 2 is greater than 115%. A low fl ow signal occurs shortly thereafter at 0.08 seconds for TA 1 when its flow rate is less than 85%. The operator is assumed to immediately trip the TCS water pump on the high flow alarm. CB movement is delayed by 0.15 seconds after the reactor scram signa l is generated.

The mass flow entering TA 2 falls quickly because of the pipe break. The flow throug h TA 1 remains above -* The flow through TA 2 drops below at around 9 seconds and experiences

[Sa , d , e , f] several slow flow reversa l s between 10 and 40 seconds after which boiling and chugging within both TAs i s predicted. Page 165of190 ATTACHMENT 1 Time (sec) Figure 84 -Target 1 Flow Meter --Target 2 Flow Meter Target 1 In let Target 2 Inlet * ** * * *

  • Target 1 OUtlet * * * * **
  • Target 2 outlet Mass Flow Transient During a LOCA in Air Without Decay Heat Removal Valves Opening [5 a , d , e , fJ In TA 2 , target rod 17 has the maximum peak power density and rod 22 has the maximum rod power. In TA 1 , rod 5 has the maximum peak power density and rod 6 has the maximum rod power though both rods are lower in power than the rods in TA 2. The heat generation in the target rods is shown in Figure 85. The heat generation drops at around 0.20 seconds when the CBs start inserting. The heat removal from the target rods presented in Figure 87and Figure 89 shows that stored energy removal is significant over the first IO seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal. Page 166 of 190 ATTACHMENT 1 1 Fission Power Fission Product Power * * * ** *
  • Heat Removal Rod 6 ---*Heat Removal Rod 5 * * ** * ** Heat Remova l Rod 22 ----Heat Removal Rod 17 Time (sec) Figure 85 Target Power During a LOCA in Air Without Decay Heat Removal Valves Opening Coolan t temperatures at the inlet and outlet of the two (2) T As are presented in Figure 86. Mass flow out of the T As due to boiling are represented by outlet temperature s between 70 and 100 °C (I 60 and 212 °F). Mass flow back into the T As are represented by the lower bounds of the temperature oscillations around 5 0 °C (I 20 °F). These flow oscillations also move water and thermal energy to the TA entrance which s lowly raises th a t temperature.

P ellet centerline temperatures do not increa se during this LOCA becau se of the reactor scram signal at 0.05 seconds, the U02 heat capacity, and the water flow through the TA caused by draining of the cooling li ne. The maximum temperatures of the cladding inner diameter (ID) are shown in Figure 87. The c ladding ID temperature increa ses by 3.0 °C (5.4 °F) prior to CB insertion and afterwards, is lower than normal operating temperature.

The increase s in cladding ID temperature at around 10 and 30 seco nds co incide with the decreases in flow rate at those times. Boiling within the T As keeps the maximum cladding temperature just above saturation temperature.

Page 167of190 0 200 400 ATTACHMENT I 600 Time (sec) 800 Figure 86 1000 1200 ** * * * *

  • Target 1 In l et --Target 1 Outlet ** * * * *
  • Target 2 Inlet --Target 2 OUtlet Coolant Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening 300 250 O' l>O ! 200 i 150 E --. I.' *
  • 0.01 'I I TI i I I ,I I I :1 I jl l l I 1 1 1 11 I i 'I I j, I I ! ' ! : i. I ' l 1 111 ! I ,. ... -I . I \ 1 lllWMl ll i 1 I i i
  • 0.1 1 10 100 1000 Time (sec) Figure 87 * ** * * ** Target Rod 6 --Target Rods *******Target Rod 22 --Target Rod 17 Maximum Cladding ID Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening Page 168 of 190 ATTACHMENT 1 Because the pellet and cladding temperatures are near norm al values , all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this L OCA. l f the one of the two decay heat removal valves for each TA were to operate, the consequences of the L OCA in air are even less significant.

The flow transient due to a LOCA in air with the deca y heat r emoval valves worki n g is shown in Figure 88. The TA inlet flow is the same as the TA outlet flow and a small natural circulation flow is established after 20 seconds. TA I flow rate is hi gher around 10 seconds due to a small contribution of pump coast-down through the intact supply line to TA I. Time (sec) Figure 88 ....... Tarisel 1 Fluw Meler --Target 2 Flow Meter * * * * * *

  • Target 1 Outlet --Target 2 Outlet Mass Flow Transient During a LOCA in Air With Decay Heat Removal Valves Opening [5 a, d , e, f] P ellet centerline temperatures also do not increase during this LOCA. The maximum temperatures of the cladding ID are show n in Figure 89. The cladding ID temperature increases by 3.0 °C (5.4 °F) prior to control blade insertion and afterwards, is lo wer than normal operating temperature.

Natural circulation is sufficient to prevent chugging and cladding temperatures steadi l y decrea se after 20 seconds. Peak vapor fractio n is less than 1 %. Page 169 of 1 90 300 250 G' tMI ! 200 ! E 150 E 100 50 --... ml "' *-,,, , .... " r .... i-**** I I I :

  • I I I I I I I I I 0.01 0.1 ATTACHMENT 1 I I 111 11 . i I I Ii I Ii: \'I , I I 1 Ii I II ' I I' ' -'!, 'I I 1 I I I I , 1!1 I ,i i I' I I
  • I ! . ..... ! I ,. I I 1
  • I I I I : I I 11 1 10 100 nme fsec) Figure 89 I I I I I I I I I I 1 I I I I * * ** * *
  • Target Rod 6 --Target Rod 5 * *** *** T;irgf't Rod n --Target Rod 17 Maximum Cladding ID Temperatures During a LOCA in Air With Decay Heat Removal Valves Opening Because the pellet and cladding temperatures are near normal values, all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOCA. The decay heat removal valves can prevent boiling and chugging in the TA but are not necessary to assure cladding integrity. 10.6 Pipe Break Locations in the Reactor Pool The target cooling lines supplying water to the T As are protected by the upper bridge structure above the pool and the refueling bridge located below the normal pool water level. The cooling lines also have lateral supports and are joined to the TA by a flexible joint. It is nearly incredible to postulate a mechanistic failure of these cooling pipes which results in a complete offset rupture within the assumed rupture time of 0.5 sec. If the break occurs near the decay heat removal valves then it is essentially the same as opening the decay heat removal valves. The worst possible break location is the connection between the inlet pipe welded to the target housing and the flexible pipe as shown in Figure 90. Direct mechanical interaction in this area is very unlikely due to the congestion around this elevation.

Page 170 of190 ATTACHMENT 1 Figure 90 Pipe Break Location in the Reactor Pool The LOCA in water used the same RELAP5 model as the LOCAs in air except that the break location w as relocated as shown in Figure 90. The break time to complete offset rupture was also kept at 0.5 seconds although it is more likely that the break time would be either longer underwater or less than a complete offset rupture. The flow transient caused by a LOCA in water resulting from a double-ended pipe break is shown in Figure 91. The transient analysis starts with the reactor and TA at 100% power and the TCS at l 00% nominal flow to accurately simulate the expected response of the protection system. The decay heat removal valves are assumed to not operate though their operation would be ineffective in this accident.

The break causes an increase in the flow measurement in the pipe supplying TA 2 and a decrease in the flow rate supplying TA 1. Page 171of190 ATTACHMENT 1 0.1 1 10 Time (sec) Figure 91 ******* Tarset 1 Flow Meter *******Target 2 Flow Meter --Target 1 Inlet --Target 2 Inlet ---*Target 1 Outlet ---*Target2 Outlet Mass Flow Transient During a LOCA in the Reactor Pool [5 a , d , e , fl The reactor is scrammed at 0.09 seconds when the mass flow rate to TA 2 is greater than 115%. A low flow signal occurs shortly thereafter at 0.13 seconds for TA I when its flow rate is less than 85%. CB movement is delayed by 0.15 seconds after the reactor scram signal is generated.

The mass flow entering TA 2 falls quickly because of the pipe break. The flow through TA 2 drops to zero at around 0.6 seconds and experiences several boiling and chugging oscillation for the next few seconds. The flow through TA I remains above zero for about 3 seconds before it also demonstrates some boiling and chugging oscillations.

Natural circulation through TA 2 after 3 seconds is sufficient to prevent further boiling and chugging.

The heat generation in the target rods is shown in Figure 92. The heat generation drops at around 0.24 seconds when the CBs start inserting.

Brief periods of transition and film boiling occur at the high heat flux locations of the target rods. The heat removal from the target rods presented in Figure 92 shows that stored energy removal is significant over the first 10 seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal and occur while significant stored energy remains to be removed. Page 172 of 190 ATTACHMENT 1 0.01 0.01 0.1 1 10 100 Time (sec) Figure 92 --Fission Power --F i ssion Product Power * ****** Heat Removal Rod 17 --Heat Removal Rod 22 Target Power During a LOCA in the Reactor Pool The maximum temperatures of the cladding ID and coolant entering and exiting TA 2 are shown in Figure 93. Cladding temperatures start experiencing rapid increases at around 0.6 seconds which corresponds to t he rapid drop in flow rate. Peak cladding temperatures reach 688 °C (1270 °F) in target rod 22 and 672 °C (1242 °F) in target rod 17. Peak cladding temperatures in rods 5 and 6 steadily decline after the reactor scram except for a brief temperature excursion of around 50 °C (90 °F) between 6 and 7 seconds. A FRAPTRAN analysis of cladding integrity during this LOCA in water transient was performed using minimum and maximum pellet-clad gap tolerances of C ladding OD temperatures as a function of time and axial location were obtained from RELAP5 and used as the boundary condition for the FRAPTRAN analysis.

FRAPCON results used in the insertion of excess reactivity transient analysis were used to define the burnup and fission gas release inputs for [5 a , b , t] F RAPTRAN at BOL and EOL conditions.

The maximum hoop strain occurs at EOL conditions in target rod 17 with the maximum gap tolerance of . The cladding temperatures used in the F RAPTRAN analysis for rod 17 are presented in Figure 94. The maximum hoop strain of 0.42% at the peak strain location along with the radial and axial strains at that location are presented in Figure 95. These strains are much less than the I% strain criteria so that cladding integrity is maintained.

Page 1 73 of 190 700 I I 600 I I I I 500 i I I I "r* 200 I I 100 ....... l. .. ---1 1 0 0.0 1 ATTACHMENT 1 I 111 I I II' 11 I . , ... ri* I I I 'I I 'l I : I I . I 1 , I \ 11: 1 Ii I I I I J : 1 I I I I *I I I I :1 : 1 I I I I I *: 1 I I I I I I I I C1 I* ! , I I I

  • I I I I! ...... r*i I r__ I: I I i I t' 11L I I 11' I 11 I Ir I ., *--I I h' *r* ..... J..J ..
  • .. . *:: . .:.: *:.: I -r r-* I I 0.1 1 10 100 Time (sec) Figure 93 i I I I I I I I I I *1 *******R od 6 --f1 od5 ******* R od 22 --R od 17 '"" TA) Ou t lP.t --TA2 Inlet Maximum Cladding ID and Coolant Temperatures During a LOCA in the Reactor Pool 750 650 550 250 150 50 0 10 20 30 nme (s e c) Figure 94 40 so ---9.98s ********* 10.Ss -10.6s ......... 10.7 s --11.3s --12.os --12.ss -----13.0 s ---14.0s 60 Cladding OD Temperature Profile in Target Rod 17 During a LOCA in the Reactor Pool Page I 7 4 of 190 0.005 0.004 c s 0.003 "' ii 15 tl 0.002 .... 0.001 0 ' ' ( I ,. ' I"'*..... ! '* .,,,, 0 1 I ' ATTACHMENT 1 '"*"' ' "' I ..... ; 2 ' I \ ,.. ' ,. " , . , ' Time (seconds)

Figure 95 .... ,.... __ *---.-..... , 3 4 ---Hoop Strain --Axial Strain Radial Strain Cladding Fractional Strains in Rod 17 during a LOCA in the Reactor Pool 10. 7 Loss of Target Flow The loss of target flow accident can be initiated by inadvertent valve closures , pump failures, or loss of e lectrical power. P ipe breaks discussed in the previous section also result in a loss of target flow. Inadvertent valve closure i s prevented by securing the valves in the open position using a locking pin since there is no safety req uirement associated with the operation of the manual isolation valves and they are only for maintenance purposes.

Therefore , the most limiting initiating event is a loss of pump flow (LOPF) which can be caused by either loss of electrical power or pump failure. The LOPF event impacts both TAs and i ncludes pump coast-down and fluid momentum to ease the transition from forced flow to natural circulation flow. The TCS has redundant 100% capacity pumps. Only one of the pumps is required to operate and the other pump is an installed spare. If the operating pump fails , there is no automatic switchover to the backup pump. Redundant flow signals in the TCS will initiate protective actions including reactor scram when flow either reduces to 85% of nominal or increases to 115% of nominal. Additional actions taken due to the flow reduction are opening of the decay heat removal valves. The LOPF was modeled using RELAP5 mod3.3 patch 03. The RELAP5 model is the same model used for the LOCA analyses described in Loss of Target Cooling. Page 175of190 ATTACHMENT 1 The flow transient caused by this LOPF is shown in Figure 96. The transient analysis starts with the TCS at 100% power and 100% nominal flow to accurately simulate the expected response of the protection system. The LOPF accident is assumed to include a loss of secondary flow as one would expect during a loss of site power. The analysis assumes that the decay heat removal valves do not open since they are not an Engineered Safety Feature (ESF). The LOPF causes a decrease in the flow measurement in the pipes supplying TA 1 and TA 2. Pump coast-down takes about 10 seconds to complete.

The reactor is scrammed at 0.14 seconds when the mass flow rate drops below 85%. CB movement is delayed by 0.15 seconds after the reactor scram signal is generated.

As natural circulation progresses, the colder water in the TCS is slowly replaced with water at reactor pool temperature which eventually results in a decline in natural circulation flow at around 600 seconds. Time (sec) Figure 96 * * * * * *

  • Target 2 Flow Meter --Target 2 Outlet Mass Flow Transient During Loss of Pump Flow [5 a , d , e , t] Maximum cladding ID temperatures and coolant temperatures are shown in Figure 97. Peak cladding temperatures rise slightly by 2.0 °C (3.6 °F) but then steadily decline after the reactor scram. Coolant exit temperature from TA 2 rises slightly at around 5 seconds and again after 600 seconds. At 600 seconds the inlet temperature starts to rise as reactor pool water has mixed with the cooler target cooling water. Eventually the TA inlet temperature will reach the maximum pool temperature of 50 °C (120 °F) at which point further increase in temperatures will cease. Because the pellet and cladding temperatures are near normal values , all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOPF. The decay heat removal valves can improve the natural circulation cooling in the TA but are not necessary to assure cladding integrity.

Page 176of190

* ATTACHMENT 1 Ttme (sec) Figure 97 ******* Rod 22 Clad ID --Rod 17 Clad ID * * *** *

  • Target 2 I nlet --Target 2 OUtlet Maximum Cladding ID and Coolant Temperatures During Loss of Pump Flow 10.8 Mishandling of Target Cartridge or Target Rods This event examine s the potential to damage the target cartridge or target rods while being removed from i ts reflector position to its reactor pool transfer location. Multiple barriers separate the fission products in t he target material from potential release location s within the TA. The robust design of the target cartridge and target rods , and careful design of the toolin g for handlin g the cartri dge and rods will help prevent a handling accident from occurring while a target cartridge is being move d from its reflector position to its storage and transfer location within the reactor pool , or removing a t arget rod from the cartridge. This movement occurs no more than twice per week. The movement is carefu lly planned and does not start until after the reactor is shut down and the target rods are sufficiently coo led so the TCS can be shutdown. Based on additional decay prior to handling , the consequences of mishandling the target cartridge or t arget rods are bounded by the TEF MHA. 10.9 Loss of Primary Coolant Flow This reactor event can be caused by a variety of initiating events which result is flow stagnation and reversal in the MURR core. For all initiating events other than loss of electrical power , the target cooling water can continue to circulate and cool the TA during this event re s ulting in no consequences to it. Page 177 of 190 ATTACHMENT 1 10.10 Loss of Primary Coolant This reactor event assumes the double-ended rupture of the largest primary coolant pipe. Low pressure initiates a reactor scram along with primary coolant pump trips , isolation valve closures , and anti-siphon valves opening. Automatic actions ensure that the core remains covered and decay heat is transferred to the pool water. Target cooling water can continue to circulate and cool the TA during this event resulting in no consequences to it. 10.11 Loss of Pool Coolant The loss of pool coolant accident postulates a break which Jowers the water level in the reactor pool until the reactor scram set point is reached. Low pool level alarms would activate before reaching that level. The TCS will continue to operate and remove stored energy from the target rods after the reactor scram. The suction line for the TCS is 37.5 inches (0.95 m) below the reactor scram pool level. When the water has dropped below the suction line , target cooling will no longer be available but the cooling water within the cooling system will continue to drain through the TAs. The accident from this point behaves similar to the LOCA in air presented earlier but with much of the stored energy already removed so that no boiling or chugging would occur. Natural circulation and heat transfer through the target cartridge walls would be sufficient to maintain target rod cladding integrity and prevent any release ofradioactivity. 10.12 Loss of Offsite Electrical Power This event would cause the reactor to scram , pumps to shut down , and valves to fail to their safe shutdown positions due to their fail-safe design.

Loss of electrical power would cause the decay heat removal valves to open thereby replacing the loss of forced flow from the shutdown of the target cooling water pumps with natural circulation flow. Pump coast-down and fluid momentum would provide additional flow during the transition from pump flow to natural circulation flow. This event is the same as the loss of target flow event analyzed earlier. 11. Technical Specifications The new Technical Specifications associated with the SGE TEF are included as Attachment 2 to the License Amendment.

12. Proposed Confir.matory Testing A variety of tests are planned to validate the design and demonstrate operation of the SGE TEF before installation and operation of the experiment in the reactor graphite reflector region. These tests are both at the component and system level , and are designed to validate not only performance and safety at the system level , but to validate analytical modeling of component behavior. Page 178of190 ATTACHMENT 1 12.1 Summary Description of Planned Tests 12.1.1 Target Pellet/C l adding Behavior Irradiation Testing An irradiation test campaign will be performed to demonstrate the safety of the target rods. The chosen l ocation for the test is the National Research Universal (NRU) reactor at Canadian Nuclear Laboratories (CNL), which can provide sufficient neutron flux and similar operating conditions to the nominal target rod design. The test will utilize capsules that replicate the geometry of the target rods. Following the i rradiation test , a series of post-irradiation examinations (P IE) will be carried out to verify the p erformance of the target rod pellets and cladding.

C NL will fabricate eight (8) LEU irradiation test capsules using components fabricated by GA. Each c apsule will contain up to 20 pellets of 5 mm diameter and 6 mm heigh t. The capsule structure will be r epresentative of the full TA' s cladding design. This will include a Zircaloy-4 cladding tube and a n ominal gap of 50 microns between the cladding and pellets (cold condition).

Two ceramic spacers3 will h old the U0 2 pellets in the center of the tube to prevent the metal end caps from reaching excessive t emperatures. A low-carbon steel retaining spring is included for handling loads. The capsule will be fi lled with helium to improve the gap thermal conductivity between the cladding and pellets. The critical dimensions and attributes are provided in Table 53 and an illustration of the conceptual capsule design is s hown in Figure 98 , all representative of production target rods. Table 53 Capsule Dimensions and Attributes Cladding ID , mm 5.100 +/- 0.025 Cladding wall thickness, mm 0.50-Pellet density, %TD 93 to 97 Pellet OD, mm 5.0-Pellet height, mm 6.0-[5 a , b , fJ U0 2 mass per capsule ,* g 25.1 Pellet enrichment, t% 9.4 Helium fill pressure , atm 1 Tube material Zircaloy-4 Ceramic spacer material Aluminum Oxide *Assumes 20 pellets per capsule at 95% theoretical density. tValue determined b y CNL neutronic s analysis.

3 The peak neutron flux in the NRU reactor is essentially uniform over th e length of the capsu le. A ceramic spacer i s therefore inc lud ed in the test capsule to prevent end effect over heating of th e capsule. Page 179 of 190 ATTACHMENT 1 Figure 98 Conceptual Design of Test Capsule The capsules are designed to interface with CNL's existing equipment for handling nuclear targets used for the production of medical isotopes in the NRU reactor. The flow channel for the target capsules is formed long rod-like structure which replaces a fuel element. This structure contains four internal channels, each of which house an assembly of four (4) target elements linked end over end in a straight line , referred to as a stringer.

The ends of the test capsules incorporate features which allow them to be incorporated into the stringer assemblies.

For the irradiation test , each stringer will consist of two (2) loaded test capsules and two 92) filler capsules of solid aluminum.

By utilizing the existing stringer design , the test capsules will allow for flexible handling , as each stringer can be removed separately.

The duration of the test irradiation period in the reactor is three (3) weeks , the maximum allowable irradiation time for target rods in MURR. The coolant flow conditions in the NRU reactor are similar to those in MURR. The coolant temperature ranges from 34 to 40 °C (93.2 to 104 °F), and the flow Reynolds number between 70 , 000 and 90 , 000 (80 , 000 for the target design in MURR). The coolant pressure is between 423 and 655 kPa (61.4 to 95 psi), which gives increased margin for boiling. Since there are multiple available irradiation locations , the target position was chosen to ensure that the targets [5 a , b , d , e , f] reach the maximum design pellet power density of-. Given the highly thermalized neutron [5 a , d , e , f] spectrum of the NRU reactor , the pellet enrichment was reduced to reach the desired power density. By matching the power density , coolant temperature and flow velocity , the test ensures that the irradiation conditions are sufficiently similar to the design condition in MURR. Page I 80 of 190 ATTACHMENT 1 Since FRAPCON/FRAPTRAN analysis shows that relocation strain occurring at shutdown contributes to t he closure of the pellet-cladding gap, the test will incorporate mu l tiple shutdown/restart cycles. -These shutdown/restart cycles will be simulated by removal of the test capsules from the flux field (the targets are compatible with online handling). The test schedule is given below in Table 54. Test Four Capsules to be Subjected to Irradiation Tests for Pellet/Cladding Behavior Table 54 Irradiation Test Schedule Tasks Irradiations Post-irradiation Examinations (PIE) Test Report Preparation Start End 11/25/16 4/10/2017 6/29/2017 8/14/2017 8/14/2017 10/3/2017 F ollowing the irradiation , the capsules will be subjected to numerous PIE tests. Only four (4) of the eight (8) capsules are to be examined -the other four (4) will serve as back-up specimens. During the i rradiation , the pellets may deform due to thermal strains , including "hour-glassing" and cracking , and r elease gaseous fission products.

A visualization of potential pellet deformation during irradiation is shown in Figure 99. This deformation and gas build-up can lead to both mechanical and chemical i nteractions between the cladding and pellets, and verifying that these phenomena do not compromise the i ntegrity of the cladding is important to the safety and operation of the target rods. \*\I I *' Figure 99 Visualization of Potential Pellet Deformation over Course of Irradiation (not to scale) Page 181of190 [5 a , d , e , t]

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-*-*--ATTACHMENT 1 The following PIE tests will be performed on the capsules following the irradiation period: ---Data collected from the PIE will be provided to the NRC as soon as it becomes available. While the test is intended to verify GA's predictions of target rod behavior , the test will be considered successful as long as the test capsule remains hermetically sealed prior to destructive examination. 12.1.2 Critical Heat Flux Tests The SGE target rods operate at very high power densities to maximize isotope yield. Analysis of the heat transfer in the target systems shows that cladding wall temperatures in the highest heat flux regions of the target exceed the coolant saturation temperature , leading to some subcooled nucleate boiling. The analysis predicts that the vapor content due to this boiling is confined to the surface of the target rods and is negligible in volume , and the NRC-accepted correlations show that the CHFR for the system will continue to provide sufficient margin. While the CHF with the various correlations shows that the CHFR for the target rod cartridge provides sufficient margin , the safest approach is to verify that there are Page 182 of 190 [5 a , b , e , f]

ATTACHMENT 1 adequate margins in the CHFR via an experiment that shows significant margin exists between the s ystem's maximum operating heat flux and the CHF for the configuration. The overa ll objectives of the test are as follow s:

  • Demonstrate that for the system's maxim um design heat flux , the cooling flow remains in the subcooled nucleate boiling regime, with minimal vapor generation at the wa11.
  • Experimentally determine the CHF for the system geometry and flow conditions , showing that sufficient safety margin is maintained. This test is being conducted at the University of Wisconsin. A representative test section has been fabricated for insertion into a low pressure flow loop to perform the heat flux testing. The geometry of t he target can be approximated by an annular channel. The cladding tube is at the center of the annulus , and a Pyrex tube will form the outer portion of the flow channel. The inner diameter of the Pyrex tube wi11 be set to match the hydraulic diameter of the design target flow channel. A 3D model of the test section design is shown in Figure 100 and the conceptual section of the test rig minus the instrumentation i s shown in Figure 101. H eating is provided by direct resistive heating of an lnconel 625 tube. The power supply is rated to 100 kW and is capable of supplying a heat flux at the wa11 of 9 MW/m 2* Instrumentation to measure p ressure drop , coolant temperature , Inconel tube temperature , and power supplied are a11 incorporated into t he test section design. Coolant temperature and pressure in the test rig are adjustable to match target rod design conditions. Since the resistive heating provides a flat heat profile , the length of the heat tube was r educed by the peaking factor so that the total power was preserved while operating at the maximum heat flux. The first test objective wi11 be achieved by operating the flow loop while providing sufficient power to the h eating rod to reach the desired heat flux. Nominal and worst-case

<Ill flow , .. heat flux) [5 a , b , f] conditions will be used. Cladding surface temperature will be calculated from internal thermocouple m easurements , power supply output , and heating tube geometry. The bubble generation at heater surface wi11 be captured with high speed camera. The second test objective wi11 be achieved by slow l y ramping the power to the heater until DNB occurs. DNB is detectable as an excursion event on the heater internal thermocouple. Coolant temperature wi11 b e adjusted to ensure that local conditions at the end of the heater where DNB occurs match the design t emperature of the coolant at the maximum heat flux lo cation in the target cartridge.

The test schedule is given below in Table 55. Page 183 of 190 ATTACHMENT 1 Table 55 Critical Heat Flux Test Schedule Test Tasks Start End Complete Assembly and Instrumentation of Test Rig and 9/16/2016 12/30/2 016 Perform Initial Shakedown Testing Critical Heat Flux Perform Flow Characterization and Tests CHFR Tests 1 /2/2 017 2/15/2017 Complete Final Report 2/15/2017 2/28/2017

[5 a,b,e,f] Figure 100 Conceptual Schematic of CHF Test Flow Section Page 184 of 190 ATTACHMENT 1 Figure 101 Uninstrumented Critical Heat Flux Test Section in Low Pressure Flow Loop Rig 12.1.3 System Integration and Cooling Flow Tests The purpose of the system integration is to verify the installation and removal procedures for the target s ystem components (target housing, target cartridge , target rods, and target loading and unloading station) and TCS piping upstream of the target housing. Operation of the pick-up tooling for the target rods and c artridges will also be demonstrated.

In addition , after system integration is complete , flow and pressure drop tests will be performed to verify the functionality of the WCM and to confirm the pressure drop t hrough the TA that was obtained by analysis.

A representation of these systems will be assembled and t ested at GA facilities.

Figure 102 illustrates the test components, which include a surrogate pool , target s ystem , and TCS. Page 185 of 190 Cooling Pump Target Cooling System Piping Target System ATTACHMENT 1 Figure 102 GA Test Setup Conceptual Arrangement The s urrogate pool will be a fiberglass tank; 6 feet (1.8 m) in diameter and 2 8 feet (8.5) tall. Thi s height allows realistic underwater simu lation of the system integration procedures.

The TCS piping in the surrogate pool will be repre se ntative of the piping and pipe s upport structure that will be used at MURR. The goal is to mimic the attachment points and structural interferences as closely as po ssi ble. The full scale production system at MURR consists of two target systems, however only one target system will be tested in this configuration at GA 's test facilities.

The test specifications and a comparison of the out-of-reactor test pool and the MURR pool are shown in Table 56. Page 1 86 of 190 Parameters Pool Diameter Pool Depth Distance from Bottom of Target to Nominal Operating Water Level WCM Inlet Pipe Inner Diameter Number of Target Assemblies WCM Outlet Pipe Inner Diameter Water Chemistry The test schedule is shown in Table 57. ATTACHMENT 1 Table 56 Testing Specifications GA Test Pool Specifications 6 feet 28 feet 26 feet 4 inches 1 3 inches nominal pH Conductivity µSiem Table 57 System Integration Test Schedule Test Tasks TA Installation Test System Integration and TA Pressure Drop Test Cooling Flow Tests Target Rod & Cartridge Installation/Removal Test Test Report Preparation Page 187of190 MURR Pool Specifications 10 feet 31 feet 26 feet 4 inches 2 3inches 5.5 nominal pH Conductivity

< 2 µSiem Start End 41412017 411012017 4/20/2017 412112017 4/24/2017 412812017 51812017 612312017 ATTACHMENT 1 13. References

1. "Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes", Tenth International Symposium, ASTM STP 1245, 1994 pp. 549-558. 2. Nekhamkin, Y., Hasan, D., Elias, E. "Zirconium Ignition in an Exposed Fuel Channel," Conference on Reactor Physics and Technology II, Israel, pp. 45-48, February 2014. 3. [ASTM STP 1245] "Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes", Tenth International Symposium, ASTM STP 1245, 1994, pp 549-558. 4. 2015 "Primary Radiation Damage Cross Sections, https://www-nds.iaea.org/CRPdpa/, 2015. 5. [304441M00014/B] "Maximum Neutron 'Damage of Al6061, Zircaloy-4 and SS316L" GA Document No. 30441M00014/B.

General Atomics Proprietary Information.

6. [King, 1973] King, T. T., A. Jostons and K. Farrell, "Neutron irradiation damage in a precipitation-hardened aluminum alloy," Effects of Radiation on Substructure and Mechanical Properties of Metals and Alloys, ASTM STP 529, American Society for Testing and Materials, 1973, pp. 165-180. 7. [Jin 2015] Jin, H.J. and T.K. Kim, "Neutron Irradiation Performance of Zircaloy-4 under Research Reactor Operating Conditions," Annals of Nuclear Energy, 25 (2015) pp. 309-315. 8. [IAEA-TECDOC-1496]

Thermal Physical Properties Database of Materials for Light Water and Heavy Water Reactors, IAEA-TECDOC-1496, June 2006. 9. [Byun 1996] Byun, T. S. and N. Hashimoto, "Strain hardening and long-range inte,rnal stress in the localized deformation of irradiated polycrystalline metals," Journal of Nuclear Materials 354 (2006) 123-1304-28.

10. Pawel, J.E. et. al., Irradiation performance of stainless steels for ITER applications," Journal of Nuclear Materials 239 (1996) 126 131. 11. University of Missouri Research Reactor Safety Analysis Report, submitted to the U.S. Nuclear Regulatory Commission August 2006. 12. [LANL 2003] "MCNP -A General Monte Carlo N-Particle Transport Core, Version 5," LA-UR-03-1987, Los Alamos National Laboratory, April 2003. 13. [Conlin 2013] Conlin, J. L. et al., "Continuous Energy Neutron Cross Section Data Tables Based upon ENDF/B-VII.1," LA-UR-13-20137, Los Alamos National Laboratory, Feb. 2013. 14. [Parsons 2012] Parsons, D. K., Conlin, J. L, "Release of Continuous Representation for ACE Data," LA-UR-12-00800, Los Alamos National Laboratory, February 2012. 15. [Kutikkad 2015] Kutikkad, K., Private Communication, University of Missouri Research Reactor, July 2015. Page 188of190 ATTACHMENT 1 16. [Peters 2016] Peters N., Private Communication, University of Missouri Research Reactor, May 2016. 17. [Kiedrowski 2010] Kiedrowski, B. C. et al., "MCNP5-1.60 Feature Enhancement

& Manual Clarifications," LA-UR-10-06217, Los Alamos national Laboratory, 2010. 18. M. C. White, "Photoatomic Data Library MCPLIB04:

A New Photoatomic Library Based On Data from ENDF/B-VI Release 8," LA-UR-03-1019, Los Alamos National Laboratory, February 2003. 19. M. C. White, "Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code," LA-13744-T, Los Alamos National Laboratory, July 2000. 20. J. F. Briesmeister, Ed., "MCNPTM -A General Monte Carlo N-Particle Transport Code, LA-13709-M, Los Alamos National Laboratory, March 2000. 21. J. W. Durkee, Jr. et al., "The MCNP6 Delayed-Particle Feature," LA-UR-12-00283, Los Alamos National Laboratory, March 2012. 22. D. B. Pelowitz, Ed., "MCNP6TM User's Manual Version 1.0, LA-CP-13-00634, Los Alamos National Laboratory, May 2013. 23. [30441R00033 2016] GA Document No. 30441R00033/B. "Forced Convection Cooling of High Power Density Nuclear Target Rod with Two-Phase Considerations" (December 2016) General Atomics Proprietary Information.

24. [Del Valle 1985] V.H. Del Valle, D.B.R. Kenning, Subcooled flow boiling at high heat flux, Int. J. Heat Mass Transfer, 28 (1985), pp. 1907-1920).
25. [Jens 1951] Jens, W.H., and Lottes, P.A, "Analysis of heat transfer, burnout, pressure drop and density data for high pressure water", ANL-4627, 1951. 26. [Unal 1976] H.C. Unal, Maximum Bubble Diameter, Maximum Bubble-Growth Time and Bubble-Growth Rate During the Subcooled Nucleate Flow Boiling of Water up to 17.72 MN/m/\2, Int. J. Heat Mass Transfer, 19 (1976), pp. 643-649. 27. [Chen 1966] Chen, J. C. "Correlation for Boiling Heat Transfer to Saturated Liquids in Convective Flow," in Ind. Eng. Chem. Proc. Dev., 5, 322, 1966. 28. [Glasstone

& Sesonske, 1967] S. Glasstone and A. Sesonske, Nuclear Reactor Engineering, Chapter 6, "Heat Transfer to Boiling Liquids: Surface and Volume Boiling", Sec. 6.129-6.149, Van Nostrand Reinhold [1967]. 29. [Bernarth 1960] Bernath, Louis, "A Theory of Local-Boiling Burnout and its Application to Existing Data", Chemical Engineering Progress Symposium Series, NO. 30, Vol 56, p.95-116, 1960. 30. [FRAPTRAN 2011] FRAPTRAN 1.5: "A Computer Code for the Transient Analysis of Oxide Fuel Rods", NUREG/CR-7023, PNNL-19400, Vol. 1, Rev. 1May2014.

Page 189of190 ATTACHMENT 1 31. [Groeneveld 2007] Groeneveld, D.C., et al, The 2006 CHF Look-Up Table, Nuclear Engineering and Design 237 (2007) p. 1909-1922.

32. * [Padoussis 2004] ASME Boiler and Pressure Vessel Code,Section III, Div 1, Appendix N, N-1345.1, p. 338, 2004. 33. [Farmer 2006] Farmer, M.T., Hoffman, E.A., Pfeiffer, P.F., Therios, 1.U, Wei, T.Y.C, Generation IV Nuclear Energy System Initiative Pin Core Subassembly Design, ANL-GENIV-070, April , 2006, p. 57-59. 34. [30441R00035 2016] "Cooling of the MURR Beryllium Reflector" GA Document No. 30441R00035/A (December 2016) General Atomics Proprietary Information.
35. GA Doc. No. 30441R00022

-Moved to Applicable Documents Table 3-2. 36. "Decay Heat Power in Light Water Reactors," ANSI/ANS-5.1-2014, published by the American Nuclear Society. 37. "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel," ANSl/ANS-5.4-2011, published by the American Nuclear Society. 38. [30441R00022 2016] "Source Term Analysis" GA Document No. 30441R00022/B (December 2016) General Atomics Proprietary Information.

Page 190of190 ATTACHMENT 2 ,Implementation of the Selective Gas Extraction (SGE) Target Experimental Facility (TEF) at the University of Research Reactor (MURR) will require minor changes to a number of existing MURR Technicaf'Bpecifications (TS) as well as new TSs. . \" J[. The following current TSs will require revision:

1. TS 3.2.g -Reactor Control and Reactor Safety Systems 2. TS 3.8.f -Experiments
3. TS 3.8.n-Experiments
4. TS 3.8.r -Experiments
5. TS 3.8.t -Experiments JII. Proposed changes to these Specifications are as follows (changes are in bold text): Specifications:

3.2.g There will be five (5) new reactor safety system instrument channels (22 through 26) and four (4) new corresponding notes (6 through 9) and revised bases. See attached TS pages. 3.8.f Each fueled experiment shall be limited such that the total inventory of iodine-131 through iodine-135 in the experiment is not greater than 150 Curies and the rriaximum strontium-90 inventory is no greater than 300 millicuries.

Exception:

SGE target rods. 3.8.n Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the containment building atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity when averaged over a year will not exceed the limits of 10 CFR 20. Exception:

Fueled experiments that produce iodine-131 through iodine-135, SGE target rods that produce iodine-131 through iodine-135, and non-fueled experiments that are intended to produce iodine-131 (See Specifications 3.8.f, 3.8.h and Amendment No. X). 3.8.r Cooling shall be provided to prevent the surface temperature of a submerged irradiated experiment from exceeding the saturation temperature of the cooling medium. Exception:

SGE target rods. 3.8.t The maximum temperature of a fueled experiment shall be restricted to at least a factor of two (2) below the melting temperature of any material in the experiment.

First-of-a-kind fueled experiments shall be instrumented to measure temperature.

Exception:

SGE targcrt rods. Bases: f. Specification 3.8.f restricts the generation of hazardous materials to levels that can be handled safely and easily. With exception of SGE target rods, analysis of fueled experiments containing a greater inventory of fission products has not been completed, 1 of3

\). L< ATTACHMENT 2 and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR and Section 10.1 of Amendment No. X). 1'" \, .. r. Specification 3.8.r is intended to reduce the likelihood of reactivity transf:bnts due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation (Ref. Sections 4.5 and 13.2.6 of the SAR and Sections 5 and 6 of Amendment No. X). t. Specification 3.8.t is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure. Amendment No. X provides the safety analysis which shows why the SGE target rod can be excepted from Specification 3.8.t III. The following five (5) new proposed TS definitions will be added: 1.37 Selective Gas Extraction (SGE) Target Assembly -The Selective Gas Extraction (SGE) Target Assembly consists of (1) a water inlet section, (2) a target housing, (3) a lower plenum, (4) a target cartridge, (5) an outlet diffuser, and (6) a target cartridge locking mechanism.

1.38 Selective Gas Extraction (SGE) Target Cartridge

-The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and (4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, (4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and (5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation.

1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target cartridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-enriched uranium dioxide pellets nominally enriched to 19.75% in the isotope uranium-235 with a nominal active length of 600 millimeters.

2 of3 __J ATTACHMENT 2 IV. The following new TS Sections will be added (see attached pages): 3.11 Selective Gas Extraction Target Experimental Facility (Limiting Conditions of Operations) 4.11 Selective Gas Extraction Target Experimental Facility (Surveillance Requirements)

5.7 Selective

Gas Extraction Target Experimental Facility (Design Features) 3 of3 1 DEFINITIONS

-Continued ATTACHMENT 2 IB;JIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS DocketNo.

50-186, License No. R-103 the principal physical barriers which guard against the uncontrolled release of radioactivity.

1.35 Scram Time -Scram time is the elapsed time between the initiation of a scram signal and insertion of the shim blades to the 20% withdrawn position.

1.36 Secured Experiment

-A secured experiment is any experiment, experimental apparatus, or component of an experiment that is held in. a *stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are normal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions.

1.37 Selective Gas Extraction (SGE) Target Assembly -The Selective Gas Extraction (SGE) Target Assembly consists of (1) a water inlet section, (2) a target housing, (3) a lower plenum, (4) a target cartridge, (5) an outlet diffuser, and ( 6) a target cartridge locking mechanism.

1.38 Selective Gas Extraction (SGE) Target Cartridge

-The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and (4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, (4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and ( 5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation.

1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target cartridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-enriched uranium dioxide pellets A-X 1 DEFINITIONS

-Continued ATTACHMENT 2 UNNERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 nominally enriched to 19.75% in the isotope uranium-235 with a nominal active length of 600 millimeters.

1.36 Secured Experiment -A secured experiment is any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are normal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions.

1.37 Senior Reactor Operator -A senior reactor operator is an individual who is licensed to direct the activities of reactor operators and manipulate the controls of a reactor. 1.38 Shim Blade (Rod) -A shim blade (rod) is a high worth control blade (rod) used for coarse adjustments in the neutron density and to compensate for routine reactivity losses. The shim blade (rod) is magnetically coupled to its drive mechanism allowing it to scram when the electromagnet is de-energized.

The shim blade (rod) also provides rod run-in functions_.

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

1.40 Shutdown Margin -Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and reactor safety systems starting from any permissible operating condition and with the most reactive shim blade and the regulating blade in the fully withdrawn positions, and that the reactor will remain subcritical without further operator action. 1.41 Surveillance Intervals

-Surveillance intervals are the maximum allowable intervals established to provide operational flexibility and not reduce frequency.

Established frequencies shall be maintained over the long term. The surveillance interval is the time between a check, test or calibration, whichever is appropriate to the item being subjected to the surveillance, and is measured from the date of the last surveillance.

Allowable surveillance intervals shall not exceed the following:

a. Biennial -interval not to exceed 2.5 years. b. Annual -interval not to exceed 15 months. A-X 1 DEFINITIONS

-Continued ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 c. Semiannual

-interval not to exceed 7.5 months. d. Quarterly

-interval not to exceed 4 months. e. Monthly -interval not to exceed 6 weeks. f. Weekly -interval not to exceed 10 days. g. Daily -interval not to exceed 1 calendar day. h. Within a shift -interval not to exceed the reactor shift. 1.42 True Value -The true value is the actual value of a parameter.

1.43 Unscheduled Shutdown -An unscheduled shutdown is defined as any unplanned shutdown, that occurs after all "Blade Full-In Lights" have cleared, caused by actuation of the reactor safety system, rod run-in system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.

1.44 Unsecured Experiment

-An unsecured experiment is any experiment which is not secured as defined by Specification 1.36, or the moving parts of secured experiments when they are in motion. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued

18. 19. 20. 21. 22. 23. 24. 25. 26. (I) Reactor Safety System Number Required (N) Instrument Channel Mode I Mode II Mode III Set Point Scram as a result of Facility Evacuation 1 1 1 actuating the facility evacuation system Scram as a result of Reactor Isolation 1 1 1 actuating the reactor isolation system Manual Scram 1 1 1 Push button on Control Console Scram as a result of Center Test Hole ies) ies) ies) removing the center test hole removable experiment test tubes or strainer SGE Target Assembly 2(6) 0(8) 0(8) 95 gpm (Min) 'A' Coolant Low Flow SGE Target Assembly 2(6) 0(8) 0(8) 129 gpm (Max) 'A' Coolant High Flow SGE Target Assembly i7) 0(8) 0(8) 95 gpm (Min) 'B' Coolant Low Flow SGE Target Assembly i7) 0(8) 0(8) 129 gpm (Max) 'B' Coolant High Flow SGE Heat Exchanger Outlet Water i9) 0(8) 0(8) 105 °F (Max) Temperature These Instrument Channels are not required when in Mode III operation below 50 kW in natural convection cooling (natural convection flange and pressure vessel cover removed).

These Instrument Channels are required when in Mode III operation with forced cooling. C 2) Flow orifice (instrumentation displayed in gpm) or heat exchanger (instrumentation displayed in psi) in each operating heat exchanger leg corresponding to the flow value in the table. C 3) Core (instrumentation displayed in psi) corresponding to the core flow value in the table. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued

<4) Trip pressure is that which corresponds to the pressurizer pressure indicated in the table with normal primary coolant flow. <5) Not required if reactivity worth of the center test hole removable experiment sample canister and its contents or the strainer is less than the reactivity limit of Specification 3.8.b .. This safety function shall only be bypassed with specific authorization from the Reactor Manager. (6) Not required if SGE target assembly 'A' is not in operation.

This safety function shall only be bypassed with specific authorization from the Reactor Manager. <7) Not required if SGE target assembly 'B' is not in operation.

This safety function shall only be bypassed with specific authorization from the Reactor Manager. (S) The SGE Target Experimental Facility shall only be operated during Mode I operation; therefore, not required for Mode II and III operation.

This safety function shall only be bypassed with specific authorization from the Reactor Manager. <9) Not required if the SGE Target Experimental Facility is secured. This safety function shall only be bypassed with specific authorization from the Reactor Manager. h. The following reactor control interlocks shall be operable whenever the reactor is in operation.

Minimum Numbers Interlock Function Onerable Prevents the control rods from Rod Withdrawal being withdrawn unless the 1. Prohibit control system logic functions 1 listed in the Bases have been satisfied Prevents placing the reactor in Automatic Control automatic control unless the 2. Prohibit control system logic functions 1 listed in the Bases have been satisfied Bases: a. Specification 3.2.a ensures that the normal method of reactivity control is used during reactor operation(Ref.

Section 4.5 of the SAR). A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued

b. Specification 3.2.b provides a restriction on the maximum neutron flux tilting that can occur in the core to ensure the validity of the power peaking factors described in Section 4.5 of the SAR. c. Specification 3 .2.c assures prompt shutdown of the reactor in the event a scram signal is received as analyzed in Section 13.2.2 of the SAR. The 20% level is defined as 20% of the shim blade full travel as measured from the fully inserted position.

Below the 20% level, the fall of the shim blade is cushioned by a dashpot assembly.

Approximately 91 % of the shim blade total wo1ih is inserted at the 20% level. d. Specification 3.2.d limits the rate of reactivity addition by the regulating blade to provide for a reasonable response from operator control (Ref. Section 4.5 of the SAR). This Specification is based on a regulating blade total reactivity worth limit of 6.0 x 10-3 Lik/k (Specification 5.3.d) and a regulating blade travel speed of 40 inches per minute. e. Specification 3.2.e assures that power increases caused by control rod motion will be safely terminated by the reactor safety system. The continuous control rod withdrawal accident is analyzed in Section 13.2.2 of the SAR. Based on a total shim blade reactivity worth of 0.1838 Lik/k and a maximum shim blade travel speed of 2 inches per minute in the inward direction, the maximum insertion rate of negative reactivity would be 2.4 x 10-4 Lik/k/sec.

Based on a maximum shim blade travel speed of 1 inch per minute in the outward direction, the maximum insertion rate of positive reactivity would be 1.2 x 10-4 Lik/k/sec (or 2.1 x 10-4 Lik/k/sec at the peak worth region of the shim blade bank). Both values are less than Specification 3.2.e limit of 3.0 x 10-4 Lik/k/sec.

The continuous rod withdrawal accident analyzed in Addenda 1 and 5 to the MURR Hazards Summary Report used reactivity insertion rates of 2.78 x 10-4 Lik/k/sec and 3.0 x 10-4 Lik/k/sec, respectively.

f. The specifications on high power level and short reactor period are provided to introduce shim blade insertion on a reactor transient before the reactor safety system trip is actuated.
  • The low pool level rod run-in provides assurance that the radiation level from direct core radiation above the pool will not exceed 2.5 mR/h (Ref. Section 11.1.5.l ofthe SAR). The vent tank low level rod run-in prevents reactor operation with a vent tank level which could result in the introduction of air into the primary coolant system (Ref. Section 9.13 of the SAR). A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued The anti-siphon system high level rod run-in provides assurance that the introduction of air to the invert loop is sufficiently rapid to prevent a siphoning action following a rupture of the primary coolant piping (Ref. Section 6.3 of the SAR). The rod not-in-contact with magnet rod run-in assures the reactor cannot be operated in violation of Specification 3.2.b due to a dropped shiin blade. The specification on the truck entry door prohibits reactor operation without the door's contribution to containment integrity as required by Specification 3.4.a. The regulating blade rod run-ins ensure termination of a transient which, in automatic control, is causing a rapid insertion of the regulating blade. g. The specifications on high power level, primary coolant flow, primary coolant pressure and reactor inlet water temperature provide for the limiting safety system settings outlined in Specifications 2.2.a, 2.2.b and 2.2.c. In Mode I and Mode II operation, the core differential temperature is approximately 17 °F; therefore, the reactor outlet water temperature scram set point at 175 °F provides a backup to the high reactor inlet water temperature scram. The core differential pressure scram provides a backup to the primary coolant low flow scrams. The reactor period scram assures protection of the fuel elements from a continuous control blade withdrawal accident as analyzed in Section 13.2.2 of the SAR. With the reflector plenum natural convection valve V547 in the open position and a pool coolant flow rate at 850 gpm, the pool coolant low flow scram assures the adequate cooling of the reactor pool, reflectors, control rods, and the flux trap (Ref. Section 5.3.5 of the SAR). The reflector high and low differential pressure scram provides a backup to the low pool coolant flow scram. The pressurizer high pressure scram provides assurance that the reactor will be shut down during a high pressure transient before the relief valve set point or the pressure limit of the primary coolant system is reached as analyzed in Section 13.2.9.4 of the SAR. The pressurizer low level scram provides assurance that the reactor will be shut down on a loss of coolant accident before pressurizer level decreases sufficiently to introduce nitrogen gas into the primary coolant system. A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued The pool water low level scram assures that the radiation level above the reactor pool from direct core radiation remains below 2.5 mR/h (Ref. Section 11.1.5.1 of the SAR). , The reactor scrams caused by the primary and pool coolant isolation valves (V507 A/B and V509) leaving their full open position provide the first line of protection for a loss of flow accident (in their respective system) initiated by an inadvertent closure of the isolation valve(s).

The power level interlock (PLI) scram provides assurance that the reactor cannot be operated with a power level greater than that authorized for the mode of operation selected on the Power Level Switch. The PLI scram also provides the interlocks to assure that the reactor cannot be operated in Mode I with a primary or pool coolant low flow scram bypassed.

The facility evacuation and reactor isolation scrams provide assurance that the reactor is shut down for any condition which initiates or leads to the initiation of a facility evacuation or an isolation of the reactor containment building.

The manual scram provides assurance that the reactor can be shut down by the operator if an automatic function fails to initiate a reactor scram or if the operator detects an impending unsafe condition prior to the initiation of an automatic scram. The center test hole scram provides assurance that the reactor cannot be operated unless the removable experiment sample canister or the strainer is inserted and latched in the center test hole. This is required anytime the reactivity worth of the center test hole removable experiment sample canister and the contained experiments or the strainer exceeds the limit of Specification 3.8.b (Ref. Section 13.2.2 of the SAR). The center test hole scram may be bypassed if the total reactivity worth of the removable experiment sample canister and the contained experiments or the strainer does not exceed the limit of Specification 3.8.b and is authorized by the Reactor Manager. The SGE target assembly coolant high and low flow scrams provides assurance that the reactor will be shut down on a loss of target coolant or forced cooling flow (Ref. Section 10 of Amendment No. X). The SGE heat exchanger high outlet water temperature scram provides assurance that the reactor will be shut down on a loss of secondary cooling flow to the SGE Target Experimental Facility.

A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued

h. Specification 3 .2.h assures that certain system conditions have been met prior to conducting a reactor startup (Master Control Switch lSl in the "ON" position; No Nuclear Instrument anomaly; Shim rods bottomed and in contact with their electromagnets; Source range level indication greater than 20 cps or intermediate range level recorder indication greater than 2 x 10-5% power; and Thermal Column door closed) or placing the reactor in automatic control at power (Reactor period as indicated by Intermediate Range Channels 2 and 3 greater than 3 5 seconds; Indicated reactor power level greater than the "auto control prohibit" set point on the wide range neutron flux monitor recorder; Regulating blade position greater than 60% withdrawn; and Range Selector Switch 1 S2 in the 5-kW red scale position or above) (Ref. Sections 7.5.3.1 and 7.5.4 of the SAR). A-X

3.8 Experiments

Applicability:

ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 This specification applies to all experiments which directly utilize neutrons or other radiation produced by the reactor. Radioactive sources shall meet the requirements for experiments.

Objective:

The objective of this specification is to prevent an accident which would jeopardize the safe operation of the reactor or would constitute a hazard to the safety of the facility staff and general public. Reactivity Limits Specification:

a. The absolute value of the reactivity worth of each secured removable experiment shall be limited to 0.006
b. The absolute value of the reactivity worth of all experiments in the center test hole shall be limited to 0.006 c. Each movable experiment or the movable parts of any individual experiment shall have a maximum absolute reactivity worth of 0.001
d. The absolute value of the reactivity worth of each unsecured experiment shall be limited to 0.0025
e. The absolute value of the reactivity worth of all unsecured experiments which are in the reactor shall be limited to 0.006 Materials Specification:
f. Each fueled experiment shall be limited such that the total inventory of iodine-131 through iodine-135 in the experiment is not greater than 150 Curies and the maximum strontium-90 inventory is no greater than 300 millicuries.

Exception:

SGE target rods. g. Fueled experiments containing inventories of iodine-131 through iodine-13 5 greater than 1.5 Curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy the requirements of Specification 3.8.s or be vented to the facility ventilation exhaust stack through high efficiency particulate air (HEP A) and charcoal filters which are continuously monitored for an increase in radiation levels. h. Each non-fueled experiment that is intended to produce iodine-131 shall be limited such that the inventory of iodine-131 is not greater than 150 Curies. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments

-Continued I. Explosive materials shall not be irradiated nor shall they be allowed to generate in any experiment in quantities over 25 milligrams of TNT-equivalent explosives.

Explosive materials shall be limited to a total quantity of 100 milligrams of equivalent explosives in the reactor containment building.

j. Corrosive materials shall be doubly encapsulated in corrosion-resistant containers to prevent interaction with reactor components or pool*water.

Should a failure of the encapsulation occur that could damage . the reactor, then the potentially damaged components shall be removed and inspected.

k. Cryogenic liquids shall not be used in any experiment within the reactor pooL 1. Fluids shall only be utilized in beamport loop experiments and shall be of types which will not chemically react in the event of leakage and shall be maintained at pressure and temperature conditions such that the integrity of the beam tube will not be impaired in the event of loop rupture. m. The normal operating procedures shall include controls on the use or exclusion of corrosive, flammable, and toxic materials in experiments or in the reactor containment building.

These procedural controls shall include a current list of those materials which shall not be used and the specific controls and procedures applicable to the use of corrosive, flammable, or toxic materials which are authorized.

Failure and Malfunctions Specification:

n. Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the containment building atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity when averaged over a year will not exceed the limits of 10 CFR 20. Exception:

Fueled experiments that produce iodine-131 through iodine-135, SGE target rods that produce iodine-131 through iodine-135, and non-fueled experiments that are intended to produce iodine-131 (See Specifications 3.8.f, 3.8.h and Amendment No. X). o. Experiments shall be designed and operated so that identifiable accidents such as a loss of primary coolant flow, loss of experiment cooling, etc., will not result in a release of fission products or radioactive materials from the experiment.

p. Experiments shall be designed such that a failure of an experiment will not lead to a direct failure of another experiment, a failure of a reactor fuel element, or to interfere with the action of the reactor safety and reactor control systems or other operating components.

A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments

-Continued

q. No experiments shall be placed in the reactor pressure vessel or water annulus surrounding the center test hole other than for reactor calibration.
r. Cooling shall be provided to prevent the surface temperature of a submerged irradiated experiment from exceeding the saturation temperature of the cooling medium. Exception:

SGE target rods. s. Irradiation containers to be used in the reactor, in which a static pressure will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2). t. The maximum temperature of a fueled experiment shall be restricted to at least a factor of two (2) below the melting temperature of any material in the experiment.

First-of-a-kind fueled experiments shall be instrumented to measure temperature.

Exception:

SGE target rods. Other Specification:

u. Only movable experiments in the center test hole shall be removed or installed with the reactor operating.

All other experiments in the center test hole shall be removed or installed only with the reactor shutdown.

Secured experiments shall be rigidly held in place during reactor operation.

v. Non-fueled experiments that are intended to produce iodine-131 shall be processed in hot cells that are vented to the exhaust stack system through charcoal filters which are continuously monitored for an increase in radiation levels. Bases: a. Specification 3.8.a provides assurance that any inadvertent insertion/removal or credible malfunction of a secured removable experiment would not introduce positive reactivity whose consequences would lead to radiation exposures in excess of the 10 CFR 20 limits. The step reactivity insertion is analyzed in Section 13.2.2 of the SAR. b. The reactivity worth of experiments in the center test hole is limited by Specification 3.8.b such that the introduction of the maximum reactivity worth of all experiments would not result in damage to the fuel plates as analyzed in Section 13.2.2 of the SAR. c. Specification 3.8.c provides assurance that the movement of movable experiments or movable parts of any experiment will not introduce reactivity transients more severe than one that can be controlled without initiating a reactor safety system action as analyzed in Section 13.2.2 of the SAR. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments

-Continued

d. Specification 3.8.d prevents the installation of an unsecured experiment which could introduce, as a positive step change, sufficient reactivity to place the reactor in a transient that would cause a violation of a limit as analyzed in Section 13.2.2 of the SAR. e. Specification 3.8.e assures that the reactivity worth of all unsecured experiments shall not exceed the maximum value authorized for a single secured removable experiment.
f. Specification 3.8.f restricts the generation of hazardous materials to levels that can be handled safely and easily. With the exception of the SGE target rods, analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR and Section 10.1 of Amendment No. X). g. Specification 3.8.g restricts the generation of hazardous materials to levels that can be handled safely and easily. Analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR). h. Specification
3. 8 .h provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). L Specification 3.8.i is intended to reduce the likelihood of damage to reactor or pool components resulting from the detonation of explosive materials (Ref. Section 13.2.6 of the SAR). J. Specification 3.8.j provides assurance that no chemical reaction will take place to adversely affect the reactor or its components.
k. The extremely low temperatures of the cryogenic liquids present structural problems that enhance the potential of an experiment failure. Specification 3.8.k provides for the proper review of proposed experiments containing or using cryogenic materials.
1. Specification

3.8.1 provides

assurance that the integrity of the beamports will be maintained for all loop-type experiments.

m. Specification 3.8.m assures that corrosive materials which are chemically incompatible with reactor components, highly flammable materials, and toxic materials are adequately controlled and that this information is disseminated to all reactor users. A-X ATTACHMENT2 UNNERSITY OF MISSOURI RESEARCH REACTOR 1 TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments

-Continued

n. The limitation on experiment materials imposed by Specification 3.8.n assures that the limits of 10 CFR 20, Appendix B, are not exceeded in the event of an experiment failure. o. -p. Specifications 3.8.o and 3.8.p provide guidance for experiment safety analysis to assure that anticipated transients will not result in radioactivity release and that experiments will not jeopardize the safe operation of the reactor. q. Specification 3.8.q is intended to reduce the likelihood of accidental voiding in the reactor core or water annulus surrounding the center test hole by restricting materials which could generate or accumulate gases or vapors (Ref. Section 4.5 of the SAR). r. Specification 3.8.r is intended to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation (Ref. Sections 4.5 and 13.2.6 of the SAR and Sections 5 and 6 of Amendment No. X). s. Specification 3.8.s is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure (Ref. Section 13.2.6 of the SAR). t. Specification 3.8.t is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure. Amendment No. X . provides the safety analysis which shows why the SGE target rod can be excepted from Specification 3.8.t. u. Specification 3.8.u is intended to limit the experiments that can be moved in the center test hole while the reactor is operating to those that will not introduce reactivity transients more severe than one that can be controlled without initiating reactor safety system action (Ref. Section 13.2.2 of the SAR). v. Specification 3.8.v provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.11 Selective Gas Extraction Target Experimental Facility Applicability:

This specification applies to the Selective Gas Extraction (SOE) Target Experimental Facility (TEP). Objective:

The objective of this specification is to reasonably assure that the health and safety of the staff and public is not endangered as a result of operation of the SOE TEP. Specification:

a. Safety Limits: (1) The temperature of a SOE target rod low-enriched uranium dioxide pellet shall not exceed 5180 °F (2860 °C) for any operating condition.

(2) The temperature of a SOE target rod Zircaloy-4 cladding shall not exceed 1652 °F (900 °C) for any operating condition.

b. Limiting Safety System Settings:

Mode I Operation SOE Target Assembly Coolant Low Flow Rate 95 gpm (Minimum)(!)

SOE Target Assembly Coolant High Flow Rate 129 gpm (Maximum)(!)

SOE Heat Exchanger Outlet Water Temperature 105 °F (Maximum)

CI) Each SOE target assembly.

c. Each SOE target cartridge shall contain eleven (11) SOE target rods. d. Each SOE target rod shall not be irradiated for greater than 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> at ten megawatts.
e. The SOE TEP shall not be operated unless the following components or systems are operable:

(1) One (1) SOE TEP coolant pump and one (1) SOE TEP coolant heat exchanger; and (2) SOE TEP Decay Heat Removal System. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.11 Selective Gas Extraction Target Experimental Facility -Continued

f. The SGE TEF shall not be operated unless the following minimum number of radiation monitoring channels are operable.
1. Radiation Monitoring Channel Number Pool Coolant Radiation Monitor 1 (!) (I) Exception:

The pool coolant system shall be sampled and analyzed at least once every four (4) hours for evidence of SGE target rod failure if the pool coolant radiation monitor is inoperable.

g. The SGE TEF shall not be operated unless the following instrument channels are operable:

Instrument Channel Number 1. SGE Target Assembly Inlet Water Temperature 1 2. SGE Target Assembly Outlet Water Temperature 1 Bases: a. The maximum normal operating temperature limit for the SGE target rods is established such that even at their hottest conditions of a maximum power level and a minimum flow rate loss of forced coolant accident, the maximum pellet temperature will still be below the uranium dioxide melting temperature of 5180 °F (2860 °C). The design limit on Zircaly-4 cladding is the predicted temperature for the onset of rapid oxidation of 1652 °F (900 °C). The cladding temperature predicted by transient analysis for the most severe loss of forced coolant accident is 1202 °F (650 °C), well below the onset of rapid oxidation temperature.

Further, the same analysis shows that during this time, retraction of the contained uranium dioxide pellet eliminates any compromise of the cladding mechanical strength.

b. The limiting safety system settings (LSSS) for the SGE TEF are set points which, if exceeded, will cause the reactor safety system to initiate a reactor scram to prevent a safety limit for the SGE target rods from being exceeded.

The LSSS help ensure that coolant flow rate and temperature will remain within the operating limits for the SGE TEF under the most severe anticipated transients.

c. Thermal steady-state and transient analyses are based on the SGE target cartridge being fully loaded with eleven (11) SGE target rods (Ref. Amendment No. X). A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .11 Selective Gas Extraction Target Experimental Facility '-Continued
d. Thermal steady-state and transient analyses are based on an SGE target rod being irradiated for no greater than 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> at ten megawatts (Ref. Amendment No. X). e. Specification 3.11.e assures that the SGE TEF can operate safely during state and transient conditions.
f. Specification 3.11.f provides for the early detection of a leaking SGE target rod so that corrective actions can be taken to minimize the release of fission products.
g. Specification 3.11.g assures that sufficient instrumentation is available to safely operate the SGE TEF. A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.11 Selective Gas Extraction Target Experimental Facility Applicability:

This specification applies to the surveillance requirements of the Select Gas Extraction (SGE) Target Experimental Facility (TEF). Objective:

The objective of this specification is to reasonably assure proper operation of the SGE TEF. Specification:

a. Each SGE target cartridge shall be verified to consist of eleven (11) SGE target rods prior to inserting the SGE target cartridge into the SGE target assembly.
b. The operating history of each SGE target rod shall be verified prior to placing it into an SGE target cartridge to ensure that the SGE target rod will not be irradiated for greater than 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> at ten megawatts.
c. The following components or systems shall be tested for operability at monthly intervals except during extended shutdown periods when the components or systems shall be tested prior to SGE TEF operation:

(1) One (1) SGE TEF coolant pump and one (1) SGE TEF coolant heat exchanger; and (2) SGE TEF decay heat removal system. d. The radiation monitor as required by Specification 3 .11.f shall be channel calibrated on a semiannual basis. e. The radiation monitor as required by Specification 3.11.f shall be checked for operability with a radiation source at monthly intervals.

f. The instrumentation required to monitor the parameters required by Specification 3 .11.g shall be channel calibrated on a semiannual basis. g. A thermal power verification of SGE TEF power, using coolant flows and differential temperatures, shall be performed weekly when the reactor is operating at 10 MW. Bases: a. Specification 4.11.a assures that the SGE target cartridges have the correct number of SGE target rods as assumed in the thermal-hydraulic and transient analyses (Ref. Amendment No. X). A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.11 Selective Gas Extraction Target Experimental Facility -Continued
b. Specification 4.11.b assures that the SGE target rods are not irradiated for greater than 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> as assumed in the thermal-hydraulic and transient analyses (Ref. Amendment No. X). c. Monthly testing of the SGE TEF coolant circulation pumps, coolant heat exchangers, and decay heat removal syste).11 is adequate to provide assurance of continued operability.
d. Semiannual channel calibration of the radiation monitoring instrumentation will assure that long-term drift of the channels will be corrected.
e. Experience has shown that monthly verification of operability of the radiation monitoring instrumentation is adequate assurance of proper operation over a long time period. f. Semiannual channel calibration of the instrument channels will assure that term drift of the channels will be corrected.
g. Thermal power verification of the SGE TEF helps ensure that the reactor state thermal power limit is not being exceeded.

A-X ATTACHMENT2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 5.7 Selective Gas Extraction Target Experimental Facility Applicability:

This specification applies to the design of the Selective Gas Extraction (SGE) Target Experimental Facility (TEF) and target rods. Objective:

The objective of this specification is to assure compatibility of the SGE TEF and target rods with the reactor. Specification:

a. The SGE TEF shall consist of not less than two (2) coolant pumps, two (2) coolant heat exchangers, plus all associated piping and valves. One (1) coolant pump and one (1) coolant heat exchanger shall be designed to provide the necessary cooling for either one (1) or two (2) operating target assemblies; the other coolant pump and heat exchanger are installed spares. b. The secondary coolant system shall be capable of continuous discharge of heat generated by the SGE TEF. c. The coolant pumps and heat exchangers of the SGE TEF shall constitute two (2) parallel systems separately instrumented to permit safe operation at ten megawatts, on either system operating, for either one (1) or two (2) operating , target assemblies.
d. The SGE TEF shall have a decay heat removal system. e. All major components of the SGE TEF in contact with pool water shall be constructed principally of aluminum alloys, stainless steel or Zircaloy-4.
f. Each SGE target rod shall contain nominally 100 low-enriched uranium dioxide pellets with a nominal active length of 23.6 inches (600 mm) and a target rod nominal total length of 26.5 inches (673 mm). g. The SGE target rod low-enriched uranium dioxide pellets shall have a nominal height of 0.236 inches (6 mm) and a nominal outside diameter of 0.197 inches (5 mm). h. The SGE target rod low-enriched uranium dioxide pellets shall be nominally enriched to 19.75% in the isotope uranium-235.
1. Each SGE target rod shall have a maximum uranium-235 loading of21.6 grams. j. The SGE target rod cladding material shall be Zircaloy-4.

A-X ATTACHMENT 2 UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 5.7 Selective Gas Extraction Target Experimental Facility -Continued

k. The SGE target rod cladding shall have a nominal inside diameter of 0.201 inches (5.1 mm) and a nominal outside diameter of 0.240 inches (6.1 mm). 1. The SGE target rods shall be contained in the SGE TEF target assemblies, in-pool target rod storage locations, or fresh target rod storage locations.
m. All SGE target rods shall be stored in a geometrical array where the value of Keff is less than 0.9 under all conditions of moderation and reflection.
n. Irradiated SGE target rods shall be stored in an array which will permit sufficient natural convection cooling such that the temperature of the target rods will not exceed design values. Bases: a. -e. The SGE TEF is described and analyzed in Amendment No. X. The SGE TEF and target rods can be safely operated at a reactor power level of ten megawatts as described in the Amendment.
f. -k._ Specifications 5.7.f, 5.7.g, 5.7.h, 5.7.i, 5.7.j and 5.3.k require uranium content, materials and dimensions of the SGE target rods to be in accordance with the design and fabrication specifications (Amendment No. X). 1. -n. The limits imposed by Specifications 5.7.1, 5.7.m and 5.7.n are conservative and assure safe SGE target rod storage. A-X ATTACHMENT 3 CODES AND STANDARDS
1. Design, Fabrication and Operation General Atomics (GA) is the prime contractor for the design and supply of the Selective Gas Extraction (SGE) Target Experimental Facility (TEP) structures, systems and components (SSC). The SGE TEP was designed and fabricated in accordance with applicable codes and standards, specifically:
  • The SGE TEP SSC and conduct of operations shall comply with all applicable U.S. Nuclear Regulatory Commission (NRC) requirements, other federal regulatory requirements, as well as local and state design codes and standards requirements.
  • SGE TEP components shall be designed to weet applicable requirements in Section VIII, Division 1 and 2 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 2015 (the term applicable refers to the method of evaluating the condition of structures and limits to stress, strains and cumulative fatigue damage).
  • Welding of the in-pool target and cooling systems shall meet the applicable reqtiirements 1 in Section IX of the ASME B&PV Code, 2015. Weld nondestructive examination (NDE) shall meet the intent of the applicable requirements in Section V of the ASME B&PV Code, 2015.
  • All materials used in the SGE TEP components shall meet the intent of applicable requirements in Section II of the ASME B&PV Code, 2015. 2. Design Methods 2.1 Software The nuclear, thennal-hydraulic, and mechanical design of the SGE TEP utilized a variety of analytical and design software packages that include:
  • Monte Carlo code MCNP6 [LANL 2003] using ENDF/B-VIl.1 for the nuclear design of the target assembly.
  • RELAP5 and FRAPTRAN to confirm thermal-hydraulic parameters of the target assembly, and for analysis of SGE TEP transient conditions.
  • MCODE, an open source linkage-program for combining MCNP6 and ORIGEN2.
  • ORIGEN2 and MATLAB for radionuclide source term calculations.
  • FRAPCON for target rod performance verification including pellet-clad mechanical interaction (PCMI), pellet-clad chemical interaction (PCCI) and internal pressure buildup from *volatile radionuclide release from the pellets.
  • ANSYS 2016 for the structural performance analysis for the target assembly.

The applicable SSCs of the SGE facility will be designed to meet all the requirements of the ASME code, but will not be code certified SSCs. 1of5

  • * * *
  • ATTACHMENT 3 ANS YS-FLUENT for single and two-phase flow modeling in the target cartridge during normal ti on. opera SIND A for modeling thermal-hydraulic performance of the target cartridge.

cad for SSC sizing analysis.

Math Solid Works 2016 for the detailed mechanical design and 3D visualization.

PIPE for structural analysis of the Target Cooling System. Auto The use of al 1 software for the development of the SGE TEF has been subjected to the rigorous software ance (QA) verification and validation procedures as required by the GA QA Program and nents (Section 3), including preparation of verification and validation reports as required by quality assur subtier docm the applicable requirements in the engineering procedures (Ref. 1). 2.2 Desig n Documentation Design bases , including all assumptions for the design bases, as well as documentation of software alculations that fonn the bases of the design and safety information presented in the license re provided in the following reports listed in Table 1. 2 This table also includes other cuments. verification c amendment a applicable do Table 1 Design and Software Validation Reports and Other Applicable Documents 2 REPORT NO. 30441ROO 017 30441ROO 019 30441ROO 021 30441ROO 022 30441ROO 030 30441ROO 031 30441ROO 032 30441ROO 033 30441ROO 035 30441ROO 038 DESIGN CALCULATION REPORT TITLE ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report Target System Cooling Calculation Report Target Assembly Thermal Analysis Source Term Analysis Design Calculation Report Mo-99 Target Cooling System Seismic Analysis Design Calculation Report Mo-99 Target Assembly Nuclear Design for Once-Through Operation RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations Cooling of the MURR Beryllium Reflector Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report All reports 1 isted in Table 1 are considered GA Proprietary Information and will be appropriately marked. 2 of5 ATTACHMENT3 Table 1 (con't) Design and Software Validation Reports and Other Applicable Documents REPORT NO. DESIGN CALCULATION REPORT TITLE 30441M00043 ANSYS Thennal Model ofRB-MSS Target Rod REPORT NO. SOFTWARE VERIFICATION TEST 30441R00002 MCNP6 (Version 1.0) Verification 30441R00003 ORIGEN2 (Version 2.2) Verification 30441R00006 MCODE (Version 2.4) Verification 30441R00020 ANSYS Workbench (Version 16.0) Verification 30441R00023 OTSTM (Version 1.0) Verification 30441R00024 RELAP (Version 5 Mod. 3.3 P03) Verification 30441R00025 FRAPCON (Version 4.0) Verification 30441R00026 FRAPTRAN (Version 2.0) Verification 30441R00028 ANSYS-FLUENT (Version 16.0) Verification 30441R00029 SINDA (Version 2.2) Verification 30441R00034 OTSTM (Version 1.0) Software Design Requirements, Description and Verification REPORT NO. OTHER APPLICABLE DOCUMENTS 30441S00001 Molybdenum-99 Supply System Requirements Document QAPD-30441-II Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System (RB-MSS) / 3. Quality Assurance GA conducted the analysis and design, and will fabricate the required SSCs under a QA program that ensures conformance, through a documented system of QA requirements, specifications and inspections, with the applicable requirements of ASME NQA-1 2008/-la-2009 Addenda that are contained in the GA Quality Assurance Manual (QAM) (Ref. 2) and the accompanying Quality Division Instructions (QDI) (Ref. 3). GA, as the prime contractor for the supply of the SGE TEP SSCs, will follow a graded approach in the implementation of its QA program commensurate with the Quality Assurance Levels (QAL) requirements on both safety and performance of the facility.

The QA requirements from GA's QA program, and the customer requirements, are implemented via project specific Quality Assurance Project Documents (QAPD) which establishes GA's QA requirements applicable to SGE project specific design, analysis, and fabrication related activities (Refs. 1 and 4). 3 of5 ATTACHMENT3

3.1 Quality

Assurance Level of Safety-Related Components Project/Resource Procedures Manual, procedure EP-4010 (Ref. 5), discusses the determination and assignment of QA levels. Three characteristics make SSCs safety-related.

If an SSC has one or more of these characteristics, then it is considered to be safety-related.

1. Maintains the integrity of the reactor coolant boundary (the reactor vessel and associated piping that circulates the reactor coolant).
2. Has the capability to shut down the reactor and maintain it in a safe condition.
3. Has the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures.

EP-4010 also states that components which contain radioactive materials are classified as QAL I as well as SSCs designated engineered safety features.

3.1.1 Selective

Gas Extraction Target Experimental Facility QAL 1 Components As the result of the engineering detennination and assessment to identify all of the equipment required for the SGE TEF, the following components are considered to meet the QAL I designation:

  • The target rods' cladding, endcaps (upper and lower) and welding of the endcaps.
  • The target pellets.
  • The flow meters in the target cooling system.
  • Temperature sensor downstream of the heat exchanger in the target cooling system, including its signal conditioner.

3.2 Manufacture

of Low-Enriched Uranium Target Rods In addition to following the graded approach for SSCs under the QAL approach mandated by GA's NQA-1 compliant QA program and implemented via the project specific QAPDs, the fabrication of enriched uranium (LEU) target rods will be subjected to a the following Quality Control procedures:

3.2.1 Target

Pellet Fabrication

  • Qualification of the manufacturing process.
  • Ensuring traceability of pellet lots through a system of fabrication travelers.
  • Verification of the LEU pellet loadings in each pellet lot using statistical sampling.
  • Measurement of pellet diameter using statistical sampling.

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ATTACHMENT 3 3.2.2 Target Rod Fabrication

  • Qualification of the manufacturing process.
  • Ensuring complete traceability of LEU pellet loadings into target rods through a system of fabrication travelers, including traceability on all cladding components.
  • 100% leak testing on target rod welding.
  • Physical measurements such as target rod length. 4. References
1. Project/Resource Procedures Manual (P/RPM), General Afomics Document GA-A15466, Second Ed., February 2015, containing Engineering Procedures EP-4020, "Design Control System Description," and EP-4070, Issue C, "Control of Scientific and Engineering Computer Programs." General Atomics Proprietary Information
2. General Atomics Quality Assurance Manual, 4th Edition, 26 August 2016. 3. General Atomics Quality Division Instructions Manual, as revised, 03 June 2016. 4. Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System, General Atomics Document QAPD-30441-II (January, 2017) and as revised. General Atomics Proprietary Information
5. Project/Resource Procedures Manual (P/RPM), General Atomics Document GA A15466, Second Ed., February 2015, containing Engineering Procedure EP-4010, Issue A, "Safety and Quality Assurance Classifications." General Atomics Proprietary Information 5of5 J
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