ML24346A167
Text
Response to SDAA Audit Question Question Number: A-4.2-9 Receipt Date: 05/13/2024 Question:
Revise the CRA design evaluation in FSAR Chapter 4.2 and TR-117605-P to include the expected CRA 10B depletion over the 20 EFPY CRA design life. Currently, the technical report provides the CRA 10B depletion limit but does not state whether the limit will be met.
Please provide proposed markup to the FSAR.
Response
TR-117605-P, Revision 0, NuFuel-HTP2' Fuel and Control Rod Assembly Designs Section 6.2.6 is revised to clarify that the boron depletion provided is the calculated boron depletion that results in a predicted maximum rod internal pressure, and is not a boron depletion limit.
Updated Response Final Safety Analysis Report Section 4.3.2.5 is revised to state that a conservative calculation is performed to verify control rod depletion is below the boron depletion that produces the maximum rod internal pressure in TR-117605-P. FSAR Section 4.3.2.5 provides the conservatively calculated 1.43 percent control rod depletion for a 20 effective full power year lifetime.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 77
minimum cross-sectional area of [
] inch2 remaining after wear (uniform circumferential wear)
minimum cross-sectional area of [
] inch2 remaining after wear (azimuthally localized wear, considering [
])
Wear limits are used in conjunction with wear rates specific to the NPM design to determine an allowable wear-based design life. After initial irradiation and operation of the CRA design, inspections are performed so that actual rod wear rates can be compared with the predetermined wear limits to demonstrate acceptable performance.
The operating environment for the CRAs is expected to be less severe with respect to rod wear than the environment typical of operating PWRs. Axial flow rates in the reactor core and in the guide tubes are significantly lower and cross flows at and above the fuel assembly top nozzles are very low (approximately ((2(a),(c),ECI feet/second). The absence of outlet flow nozzles in the upper internals reduces the cross flows compared to a typical PWR. These flow conditions create a more benign flow environment, reducing mechanical interactions with the guide cards and fuel assemblies. Based on this assessment, the CRA design lifetime is not expected to be limited by control rod wear. 6.2.6 Control Rod Internal Pressure The control rod internal pressure analysis predicts the maximum internal rod pressure using a conservative model that calculates the depletion in the B4C pellets and release of helium to the rod plenum volume. The calculation includes helium backfill, residual, and sorbed gases in the determination of the final maximum internal pressure. The AIC material is not a source of gases. During normal operation, the CRAs are positioned such that the B4C pellets are located above the active fuel. Over the lifetime of the CRA, there is very low depletion, which results in insignificant helium production. In addition, the large porosity of the B4C absorber material provides sufficient volume to accommodate any helium produced. However, the analysis conservatively assumes [ ] percent release and retention of the helium due to depletion of the B4C pellets. Audit Question A-4.2-9 For a 20 EFPY control rod assembly design life, a conservatively calculated 10B depletion of [ ] percent creates a predicted maximum rod internal pressure of [ ] psia, which meets the criterion of being less than RCS pressure (2000 psia). 6.2.7 Component Melt Analysis The control rod is analyzed to ensure that each component remains below the melt temperature. The analysis uses conservative values for heating rates and gap conductance. The worst case calculated temperatures for all rod components are well below the material melt limits.
NuScale Final Safety Analysis Report Nuclear Design NuScale US460 SDAA 4.3-17 Draft Revision 2 versus CRA position is calculated by a series of steady-state calculations at various CRA positions, assuming the CRAs are at the PDIL as the initial position in order to minimize the initial reactivity insertion rate. The CRA with the highest worth is assumed stuck out of the core, and the flux distribution is assumed to be skewed to the bottom of the core. The reactivity worth versus CRA position is provided in Figure 4.3-21. Table 4.3-4 and Table 4.3-5 provide the CRA worth for the individual CRA that is stuck out of the core at HZP with the remaining CRAs starting from the all rods out or PDIL positions. Figure 4.3-22 through Figure 4.3-25 provide the differential and integral rod worth for the regulating bank at BOC, MOC, and EOC for various power levels. Audit Question A-4.2-9 The loss of CRA worth due to depletion of absorber material is negligible. A conservative calculation over a 20 effective full power year CRA lifetime demonstrates that less than 2.5 percent ofThe expected boron depletion is conservatively calculated over a 20 effective full power year CRA lifetime. The calculated expected boron depletion of 1.43 percent verifies the boron in the upper portion of the CRA is lost due to depletion is less than the depletion required to produce the maximum control rod internal pressure in Reference 4.3-7. The silver-indium-cadmium in the lower portion of the CRA is also evaluated for a loss of worth because of depletion and shown to have an insignificant impact on the available worth of the CRAs over their lifetime. Typical reactor operation with the rods withdrawn from the core while at full power limits the potential for CRA absorber depletion. Rod worth is confirmed at the beginning of each cycle during start-up physics testing. Pressurized water reactor (PWR) operating experience has identified a phenomenon associated with potential boron build-up on the fuel rods that could affect SDM. Build-up of boron in crud at the top of the core can cause the reactivity at the bottom of the core to increase. Such a redistribution of power adversely affects the worth of the CRAs. The uncertainty analysis of the CRA worth includes comparisons to operating data from existing PWRs. Also, constant monitoring of core AO and comparison of that offset to predicted values identifies build-up of boron on the cladding surface during operation. Further, post-irradiation examinations as described in Section 4.2 measure oxide build-up and crud deposition on the fuel rods to ensure boron deposits on the cladding do not adversely affect the rod worth. 4.3.2.6 Criticality of the Reactor During Refueling Criticality during a refueling is prevented by maintaining an effective neutron multiplication factor (keff) of 0.95 or less at all times. Refueling is performed with CRAs inserted in the fuel assemblies. Calculation of the required boron concentration for refueling assumes the two highest worth CRAs are not inserted. Criticality of fuel assemblies outside the reactor is precluded by adequate design of fuel transfer and storage facilities and by administrative control procedures.
NuScale Final Safety Analysis Report Nuclear Design NuScale US460 SDAA 4.3-21 Draft Revision 2 for safety analysis. Also, SIMULATE5 calculation results are used as input into the plant monitoring system. SIMULATE-3K SIMULATE-3K is used for transient reactor core analysis where a quasi-steady-state assumption is not valid. SIMULATE-3K has the same computational foundation as SIMULATE5 and requires a model that is initialized in SIMULATE5 as input, but has been extended for transient applications. MCNP6 MCNP6 is a general-purpose code that can be used for neutron, photon, electron, or coupled neutron photon electron transport. The code treats an arbitrary three dimensional configuration of materials in geometric cells bounded by first and second degree surfaces and some special fourth-degree surfaces. Point-wise (continuous energy) cross section data are available with MCNP6. The MCNP6 code is a higher fidelity code than CASMO5/SIMULATE5 and is used for code-to-code comparisons of the CASMO5, CMSLINK, SIMULATE5 suite of codes in Reference 4.3-2. 4.3.4 References 4.3-1 NuScale Power, LLC, Framatome Fuel and Structural Response Methodologies Applicability to NuScale, TR-108553-P-A, Revision 0. 4.3-2 NuScale Power, LLC, Nuclear Analysis Codes and Methods Qualification," TR-0616-48793-P-A, Revision 1. 4.3-3 Krimer, M., G. Grandi, and M. Carlsson, PWR Transient XENON Modeling and Analysis Using Studsvik CMS, Proceedings of 2010 LWR Fuel Performance/Top Fuel/WRFPM, Orlando, Florida, September 26-29, 2010. 4.3-4 Los Alamos National Laboratory, Initial MCNP6 Release Overview - MCNP6 Version 1.0, LA-UR-13-22924. 4.3-5 NuScale Power, LLC, Fluence Calculation Methodology and Results, TR-118976-P, Revision 0. 4.3-6 NuScale Power, LLC, Rod Ejection Accident Methodology, TR-0716-50350-P, Revision 32. Audit Question A-4.2-9 4.3-7 NuScale Power, LLC, NuFuel-HTP2TM Fuel and Control Rod Assembly Designs, TR-117605-P, Revision 0.}}