ML16062A374

150903 RAI Responses 160301

ML16062A374
Person / Time
Site:	Rhode Island Atomic Energy Commission
Issue date:	03/01/2016
From:	Davis M J State of RI, Atomic Energy Comm
To:	Patrick Boyle Office of Nuclear Reactor Regulation
Shared Package
ML16062A372	List: ML16062A373 ML16062A374 ML16062A375 ML16062A376 ML16062A377 ML16062A378 ML16062A379 ML16062A380 ML16062A381 ML16062A382 ML16062A383
References
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Download: ML16062A374 (10)

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150903 RAI Responses 160301 In 2010, the analyses and material that was in the SAR that was submitted in 2004 was largely abandoned. The reviewers should not be looking at this document for information. None of the people that were involved in preparing that information are still associated with the RINSC facility or Argonne National Laboratory. Consequently, new analyses were done starting in 2010. RAI 4.34 a Response Our core configuration is now in the equilibrium configuration with the most efficient neutron reflection. When a refueling operation occurs, four of the most highly burned fuel elements are removed from the center of the core, the remaining elements are shuffled toward the center of the core leaving the four corner grid spaces open, and four fresh elements are positioned in the corner spaces. for the beginning of core data for the current core loading. RAI 4.34 b Response the beginning of core data for the current core loading.

RAI 4.34 c Response See RAI 4.12 Response for coolant temperature reactivity coefficient and coolant density reactivity coefficient. See RAI 4.13 Response for temperature and void reactivity coefficient. RAI 4.34 d Response At this point, RINSC has two core configurations, both of which have the reflector elements positioned in the most efficient reflector configuration possible. The standard core is a 14 element core. However a 17 element core was also approved by the NRC. The most limiting core is the 14 element core because at full power, 2MW is distributed across only 14 elements, rather than 17 elements. RAI 4.1 Response Figure 4-6 shows the power peaking factors for both cores. RAI 14.67 Response shows the first four configurations of the LEU core which transitioned from the least reflective configuration, to the most reflective configuration. See RAI 4.16 Response. RAI 4.36 a Response This question has to do with the fact that the control rod worth increases as the core becomes more compact, and / or as the core reflection is made more efficient. NRC has approved a 17 element core, as well as a 14 element core. The 14 element core is the loading that is typically in use, and is the core loading that was used for all of the thermal-hydraulic and transient analyses. In October of 2008, core reflection was changed to be the most efficient neutron reflection possible. As required by the facility Technical Specifications, control rod calibrations and reactivity insertion rates are determined annually. The results of those determinations have never shown the reactivity insertion rate to be greater than or equal to 0.02 %k/k per second. RAI 4.36 b Response Proposed Technical Specifications 160226 includes the following proposed specifications for limiting reactivity insertion rates: 3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies: Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 3.2.1.2 The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02% dK/K per second. Historically, there has never been a linear withdrawal rate. Control rod reactivity varies from year to year, so a linear withdrawal rate would lead to a variable reactivity insertion rate limit. Since it is the reactivity insertion rate that is the important parameter, we intend to keep the specification related to reactivity insertion rate, rather than withdrawal rate. NUREG 1537 is guidance only. There is no requirement to show compliance with the cited DNBR guidance.

-Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center Table 4.6-9 shows that onset of nucleate boiling is predicted to occur when the fuel cladding temperature reached 122.6 C, under the condition that the flow rate is 1580 gpm. This is well below the safety limit of 530 C. When the onset of nucleate boiling is reached, the heat transfer coefficient increases due to energy expended in heat of vaporization. Consequently, showing that the operating conditions would never allow onset of nucleate boiling to be reached clearly shows that the safety limit would never be reached, and that there is a significant margin of safety. Figure 4.6-8 shows this nicely. The computer codes that were used for this analysis were submitted, reviewed, and accepted years ago. Re-analyzing them is beyond the scope of this project at this point. Please see the enclosed letter dated March 15, 2013 indicating that these code manuals and inputs were sent to NRC on that date. Revised Jan. 20, 2016 by Arne possible to set the over power trip to be 115% for both natural convection and forced convection cooling modes of operation. These analyses used the original computer codes that were previously used, with updated inputs to build the desired safety envelope.

The cooling system was upgraded 23 years ago as part of relicensing the facility for the fuel conversion from HEU fuel to LEU fuel. The safety analysis for this change was reviewed and approved as part of that license change. The LOCA analysis was done for the two loop cooling system that is currently in place. RAI 5.2 a Response This cooling system was approved by NRC in 1992 as part of the HEU to LEU fuel conversion license change. RAI 5.2 b Response Proposed Technical Specification Section 5.2 provides a description of the cooling system. RAI 5.2 c Response This system was constructed and put into service long before any of the current employees were associated with the facility. It is unknown whether or not any functional tests were conducted at that time. The system has been in service for 23 years. RAI 5.2 d Response The system has been in service for 23 years. Again, in 2010, the analyses and material that was in the SAR that was submitted in 2004 was largely abandoned. The reviewers should not be looking at this document for information. The entire document will be re-written after the technical details have been defined.

RAI 5.2 e Response None of the analyses depend on both cooling loops to be operating. In 1992 as part of the fuel conversion from HEU to LEU fuel, the facility upgraded its cooling system. As part of that upgrade, a water line that runs from the facility fire sprinkler system to the top of the high power section of the reactor pool was added. It has two manual ball valves that are locked out to prevent them from being opened inadvertently. It could be used as a last resort for adding town water to the pool. However, none of the safety analyses take credit for the existence of this system. Consequently, when a new SAR is written, this system will not be described.

The SAR will be re-written once the technical details have been worked out. It will reference the safety limit that is based on fuel temperature rather than power level.

RAI 13.28 a Response Failure Add version of this that addresses some of the questions associated with this RAI, is attached. In that document, the error associated with the number of fissions per Watt-second was corrected on page 1. RAI 13.28 b Response The breathing volume rate is corrected to be 2 X 104 cm3 per minute in the definition of the DAC on page 16 of the new analysis. This is the breathing rate that was used for this analysis. RAI 13.28 c Response The new analysis takes the I-131 DAC from 10 CFR 20 Appendix B Table 1 Column 3, which lists it as 2E-8 . This is shown in the first table on page 23 in the new analysis.

RAI 13.28 d Response The analysis makes the incredible assumption that the fission fragment inventory of an entire fuel plate is released to the pool water. We think that this is a very conservative bounding condition. RAI 13.28 e Response The new analysis takes credit for the confinement stack, which makes building wake and horizontal meander effects irrelevant. RAI 13.28 f Response The new analysis makes the following assumptions based on NRC Regulatory Guide 1.183: A. Noble gases are unaffected by the pool water. B. The pool water retains 99.5% of the radioiodines that are released. C. The radioiodines are composed of 45% elemental, and 55% organic species. D. Activity released from the pool to confinement air occurs over a two hour period. E. All other fission products are retained either in the fuel, or in the pool water. RAI 13.28 g Response Fission product yield data was taken from Table 11.4 in: Lamarsh, John R., Introduction to Nuclear Engineering. Massachusetts: Addison-Wesley Publishing Company. 1977. The double dagger footnote associated with the table indicates that the yields given are equal to the yield of the nuclide plus the cumulative yield of the precursor. RAI 13.28 h Response indicate that the free volume of the confinement room was determined to be 5.15 X 109 cc. This is consistent with Proposed Technical Specification 5.1.1. RAI 13.28 i Response See RINSC Procedure MP 02 Emergency Air Filter Efficiency Test Rev. 3 which describes how the iodine efficiency of the charcoal filter is tested. Proposed Technical Specification 4.5.2.4 provides the surveillance interval for this test.

RAI 13.28 j Response The individual that wrote the Radiological Assessment Attachment retired a number of years ago. Please review the Fuel Failure Addendum.

150903 RAI Responses 160301 In 2010, the analyses and material that was in the SAR that was submitted in 2004 was largely abandoned. The reviewers should not be looking at this document for information. None of the people that were involved in preparing that information are still associated with the RINSC facility or Argonne National Laboratory. Consequently, new analyses were done starting in 2010. RAI 4.34 a Response Our core configuration is now in the equilibrium configuration with the most efficient neutron reflection. When a refueling operation occurs, four of the most highly burned fuel elements are removed from the center of the core, the remaining elements are shuffled toward the center of the core leaving the four corner grid spaces open, and four fresh elements are positioned in the corner spaces. for the beginning of core data for the current core loading. RAI 4.34 b Response the beginning of core data for the current core loading.

RAI 4.34 c Response See RAI 4.12 Response for coolant temperature reactivity coefficient and coolant density reactivity coefficient. See RAI 4.13 Response for temperature and void reactivity coefficient. RAI 4.34 d Response At this point, RINSC has two core configurations, both of which have the reflector elements positioned in the most efficient reflector configuration possible. The standard core is a 14 element core. However a 17 element core was also approved by the NRC. The most limiting core is the 14 element core because at full power, 2MW is distributed across only 14 elements, rather than 17 elements. RAI 4.1 Response Figure 4-6 shows the power peaking factors for both cores. RAI 14.67 Response shows the first four configurations of the LEU core which transitioned from the least reflective configuration, to the most reflective configuration. See RAI 4.16 Response. RAI 4.36 a Response This question has to do with the fact that the control rod worth increases as the core becomes more compact, and / or as the core reflection is made more efficient. NRC has approved a 17 element core, as well as a 14 element core. The 14 element core is the loading that is typically in use, and is the core loading that was used for all of the thermal-hydraulic and transient analyses. In October of 2008, core reflection was changed to be the most efficient neutron reflection possible. As required by the facility Technical Specifications, control rod calibrations and reactivity insertion rates are determined annually. The results of those determinations have never shown the reactivity insertion rate to be greater than or equal to 0.02 %k/k per second. RAI 4.36 b Response Proposed Technical Specifications 160226 includes the following proposed specifications for limiting reactivity insertion rates: 3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies: Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dK/K 3.2.1.2 The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02% dK/K per second. Historically, there has never been a linear withdrawal rate. Control rod reactivity varies from year to year, so a linear withdrawal rate would lead to a variable reactivity insertion rate limit. Since it is the reactivity insertion rate that is the important parameter, we intend to keep the specification related to reactivity insertion rate, rather than withdrawal rate. NUREG 1537 is guidance only. There is no requirement to show compliance with the cited DNBR guidance.

-Hydraulic Analysis for Forced Convective Flow in the Rhode Island Nuclear Science Center Table 4.6-9 shows that onset of nucleate boiling is predicted to occur when the fuel cladding temperature reached 122.6 C, under the condition that the flow rate is 1580 gpm. This is well below the safety limit of 530 C. When the onset of nucleate boiling is reached, the heat transfer coefficient increases due to energy expended in heat of vaporization. Consequently, showing that the operating conditions would never allow onset of nucleate boiling to be reached clearly shows that the safety limit would never be reached, and that there is a significant margin of safety. Figure 4.6-8 shows this nicely. The computer codes that were used for this analysis were submitted, reviewed, and accepted years ago. Re-analyzing them is beyond the scope of this project at this point. Please see the enclosed letter dated March 15, 2013 indicating that these code manuals and inputs were sent to NRC on that date. Revised Jan. 20, 2016 by Arne possible to set the over power trip to be 115% for both natural convection and forced convection cooling modes of operation. These analyses used the original computer codes that were previously used, with updated inputs to build the desired safety envelope.

The cooling system was upgraded 23 years ago as part of relicensing the facility for the fuel conversion from HEU fuel to LEU fuel. The safety analysis for this change was reviewed and approved as part of that license change. The LOCA analysis was done for the two loop cooling system that is currently in place. RAI 5.2 a Response This cooling system was approved by NRC in 1992 as part of the HEU to LEU fuel conversion license change. RAI 5.2 b Response Proposed Technical Specification Section 5.2 provides a description of the cooling system. RAI 5.2 c Response This system was constructed and put into service long before any of the current employees were associated with the facility. It is unknown whether or not any functional tests were conducted at that time. The system has been in service for 23 years. RAI 5.2 d Response The system has been in service for 23 years. Again, in 2010, the analyses and material that was in the SAR that was submitted in 2004 was largely abandoned. The reviewers should not be looking at this document for information. The entire document will be re-written after the technical details have been defined.

RAI 5.2 e Response None of the analyses depend on both cooling loops to be operating. In 1992 as part of the fuel conversion from HEU to LEU fuel, the facility upgraded its cooling system. As part of that upgrade, a water line that runs from the facility fire sprinkler system to the top of the high power section of the reactor pool was added. It has two manual ball valves that are locked out to prevent them from being opened inadvertently. It could be used as a last resort for adding town water to the pool. However, none of the safety analyses take credit for the existence of this system. Consequently, when a new SAR is written, this system will not be described.

The SAR will be re-written once the technical details have been worked out. It will reference the safety limit that is based on fuel temperature rather than power level.

RAI 13.28 a Response Failure Add version of this that addresses some of the questions associated with this RAI, is attached. In that document, the error associated with the number of fissions per Watt-second was corrected on page 1. RAI 13.28 b Response The breathing volume rate is corrected to be 2 X 104 cm3 per minute in the definition of the DAC on page 16 of the new analysis. This is the breathing rate that was used for this analysis. RAI 13.28 c Response The new analysis takes the I-131 DAC from 10 CFR 20 Appendix B Table 1 Column 3, which lists it as 2E-8 . This is shown in the first table on page 23 in the new analysis.

RAI 13.28 d Response The analysis makes the incredible assumption that the fission fragment inventory of an entire fuel plate is released to the pool water. We think that this is a very conservative bounding condition. RAI 13.28 e Response The new analysis takes credit for the confinement stack, which makes building wake and horizontal meander effects irrelevant. RAI 13.28 f Response The new analysis makes the following assumptions based on NRC Regulatory Guide 1.183: A. Noble gases are unaffected by the pool water. B. The pool water retains 99.5% of the radioiodines that are released. C. The radioiodines are composed of 45% elemental, and 55% organic species. D. Activity released from the pool to confinement air occurs over a two hour period. E. All other fission products are retained either in the fuel, or in the pool water. RAI 13.28 g Response Fission product yield data was taken from Table 11.4 in: Lamarsh, John R., Introduction to Nuclear Engineering. Massachusetts: Addison-Wesley Publishing Company. 1977. The double dagger footnote associated with the table indicates that the yields given are equal to the yield of the nuclide plus the cumulative yield of the precursor. RAI 13.28 h Response indicate that the free volume of the confinement room was determined to be 5.15 X 109 cc. This is consistent with Proposed Technical Specification 5.1.1. RAI 13.28 i Response See RINSC Procedure MP 02 Emergency Air Filter Efficiency Test Rev. 3 which describes how the iodine efficiency of the charcoal filter is tested. Proposed Technical Specification 4.5.2.4 provides the surveillance interval for this test.

RAI 13.28 j Response The individual that wrote the Radiological Assessment Attachment retired a number of years ago. Please review the Fuel Failure Addendum.