ML110590055
| ML110590055 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/28/2011 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TS-473 ANP-2988NP, Rev 0 | |
| Download: ML110590055 (43) | |
Text
ENCLOSURE 3 Tennessee Valley Authority Browns Ferry Nuclear Plant, Unit I Technical Specifications Change TS-473 - AREVA Fuel Transition Browns Ferry Unit I AREVA Fuel Transition Input to TVA for RAIs Non-Proprietary
ANP-2988NP Revision 0 Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs February 2011 A
AREVA AREVA NP Inc.
AREVA NP Inc.
ANP-2988NP Revision 0 February 2011 Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs
AREVA NP Inc.
ANP-2988NP Revision 0 Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs Copyright © 2011 AREVA NP Inc.
All Right Reserved AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page i Nature of Changes Item Page Description and Justification
- 1.
All New Document.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page ii Contents 1.0 2.0 3.0 In tro d u ctio n.....................................................................................................................
1 NRC Questions and AREVA Response...........................................................................
1 References....................................................................................................................
33 Tables Table 1:
[
Table 2:
[
].........................
14
]..............
. 24 Figures Figure 1:
Figure 2:
Figure 3:
Figure 4:
Figure 5:
Figure 6:
Figure 7:
Figure 8:
Figure 9:
Figure 10:
Figure 11 Illustration of Chamfered Pellet Design Change.....................................................
2 ATRIUM-10 Standard Load Chain.........................................................................
4 A TRIUM-10 Harmonized Advanced Load Chain..................................................
5 Gadolinia Fuel Rod Axial Descriptions..................................................................
7 A 10-3761B-14GV80-FAA Fuel Rod Distribution....................................................
8 A 10-3831B-12GV80-FAA Fuel Rod Distribution....................................................
9 A 10-4171B-14GV80-FAA Fuel Rod Distribution.................................................
10 Comparison of Fuel Swelling and Densification..................................................
18 Comparison of Fission Gas Release..................................................................
19 Browns Ferry Cold Critical Data for GE Cores and Transition Cores...................
28 BFN Gamma 5T. TIP Response Versus Cycle Number.......................................
32 This document contains a total of 38 pages.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 1 1.0 Introduction Tennessee Valley Authority (TVA) submitted a License Amendment Request (LAR) to change the Browns Ferry Unit 1 Technical Specifications in support of reload fuel transition to AREVA NP. In response to the LAR, the US Nuclear Regulatory Commission (NRC) has issued an initial set of questions, in the form of Request for Additional Information (RAI), Reference 1.
Based on the information provided in this report, TVA will prepare a formal response to the NRC RAIs.
2.0 NRC Questions and AREVA Response The NRC questions (i.e., RAIs) listed below are according to Reference 1:
RAI 1
ANP-2877P, Section 1.0 Explain what "chamfered pellet design" is and describe how this design reduces the occurrence of pellet chipping during manufacturing, and then reducing the pellet-clad-interaction failure due to missing pellet surface.
AREVA Response:
The chamfered pellet design includes a small chamfer on both ends of the pellet. The configuration is illustrated in Figure 1.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 2 Figure 1: Illustration of Chamfered Pellet Design Change Previously, the pellet design had theoretically sharp edges at the two pellet ends.
Hot cell investigations revealed the cause of the failures to be MPS or missing pellet surface. MPS can cause localized bending stresses in the cladding during pellet-clad interaction. The MPS is observed to extend from the pellet edge over a portion of the pellet circumferential surface similar to the chipped area illustrated in the figure above. Examinations confirmed the primary failure location in the cladding coincided with the MPS. Power changes AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 3 leading to otherwise normal cladding stress levels at the pellet-pellet interface resulted in much higher localized stresses because of the discontinuous support conditions in the immediate vicinity of the chipped area.
AREVA implemented the chamfered pellet design as one of the initiatives to improve pellet quality with the goal of eliminating MPS failures.
RAI 2
ANP-2877P, Section 2.1.4 Explain the "Harmonized Advanced Load Chain" modifications that improved the upper tie plate (UTP) connection by making it simpler and more robust.
AREVA Response:
The quarter-turn quick disconnect mechanism for latching the UTP on the ATRIUM-IC fuel assembly has evolved since the fuel design was first supplied in the early 1990s. The current configuration is designated as the Harmonized Advanced Load Chain (HALC). The HALC has been supplied for a number of years and is in use at both Browns Ferry Units 2 and 3.
The original quarter-turn quick disconnect mechanism had more parts including a small locking pin and anti-rotation spring. Although there were no failures of this design during operation, it was viewed as an overly complicated design that could be susceptible to jamming or generating loose parts. The HALC has fewer parts, is easier to latch and unlatch when required, and simpler to manufacture. Figure 2 and Figure 3 provide an exploded view of the parts in both designs.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 4 Figure 2: ATRIUM-10 Standard Load Chain AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 5 Figure 3: ATRIUM-10 Harmonized Advanced Load Chain AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 6 RAI 3: ANP-2877P, Section 2.1.5, ANP-2859(P) Appendix B (a) Provide details of the distribution of Gadolinia (U0 2+Gd 2O3) rods in the BFN Unit 1 core for the upcoming cycle, with respect to the number of Gadolinia rods and respective Gadolium (Gd) enrichment.
(b) With degraded thermal conductivity, and lower melting point of the U0 2+Gd 2O3 mixture, describe what adjustments are made in the Gadolinia rods to prevent failure of the Gadolinia rod melting. Is there any restriction on linear heat generation rate (LHGR) limit for the Gadolinia rods during normal operation and anticipated operational occurrences (AOOs)?
(c) Section 3.2.4 of ANP-2877 indicates that "for AQOs, the fuel temperatures are calculated using the same power history (as for normal operating temperatures), except that additional calculations are performed at elevated power levels as a function of exposure corresponding with the Protection Against Power Transients (PAPT) LHGR limit." Describe this process with example calculations.
(d) Describe the adjustments, and how the adjustments are applied to the fuel melt temperature, for exposure and Gadolinia content, as stated in Section 3.2.4 of ANP-2877.
Show a typical calculation.
AREVA Response:
(a) Figures 4 - 7 below depict the Gadolinia rods, their Gadolinium enrichments, the number of Gadolinia rods in each bundle design, and their locations in the bundle lattices. Figure 4 shows the rods themselves, including the respective Gadolinium enrichment, and Figures 5 - 7 show where the Gadolinia rods are loaded (highlighted with shading) in the lattices for each of the three bundle designs reported in ANP-2859(P) Appendix B. The bundle design depicted in Figure B. I of ANP-2859(P) features 14 Gadolinia rods with Gadolinium enrichments of 4.5 and 8%. The bundle design depicted in Figure B.2 of ANP-2859(P) features 12 Gadolinia rods with Gadolinium enrichments of 4.5 and 8%. The bundle design depicted in Figure B.3 of ANP-2859(P) features 14 Gadolinia rods with Gadolinium enrichments of 4.5 and 8%.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 7 Fuel Rod Type 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 7
11 13 18 19 G
G L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L L
L Pellet Type U-235 + Gd 203 /
U-234 + U-236 (wt%)
Pellet Tvme U-235 + Gd 20 3 /
.234 + U-236 (wt'3 G
4.00+4.5 Figure 4: Gadolinia Fuel Rod Axial Descriptions AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 8 I
I -------
1 2
3 4
4 4
4 3
3 1
2 5
6 7
8 6
6 9
5 3
3 6
6 10 11 12 13 7
12 6
4 7
10 9
9 9
9 6
7 10 4
8 11 9
9 7
9 4
6 12 9
Water Channel 9
8 9
4 6
13 9
10 7
9 3
9 7
6 9
9 10 6
12 6
3 5
12 7
7 8
7 12 5
3 1
3 6
10 9
9 9
6 3
2 Fuel Rod Type Quantity Fuel Rod Type Quantity 1
2 3
4 5
6 7
3 3
10 8
4 14 10 8
9 10 11 12 13 4
19 6
2 6
2 Figure 5: AIO-3761B-14GV8O-FAA Fuel Rod Distribution AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 9 J
1 2
3 4
4 4
4 3
3 1
2 14 6
7 15 6
6 16 14 3
3 6
6 10 9
12 13 7
12 6
4 7
10 9
16 9
16 10 7
10 4
15 9
16 16 7
9 4
6 12 9
Water Channel 9
15 9
4 6
13 16 9
7 9
3 16 7
10 16 9
9 9
12 6
3 14 12 7
7 15 7
12 14 3
1 3
6 10 9
9 9
6 3
2 Fuel Rod Type Quantity Fuel Rod Type Quantity 1
2 3
4 6
7 9
3 3
10 8
11 10 16 10 12 13 14 15 16 6
6 2
4 4
8 Figure 6: AIO-3831B-12GV80-FAA Fuel Rod Distribution AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 10 J
1 2
3 4
6 6
6 6
4 2
2 14 17 18 15 16 18 16 15 4
3 17 17 16 19 12 16 18 12 9
4 18 16 19 16 16 12 12 18 12 6
15 19 16 12 12 12 6
16 12 16 Water Channel 12 15 12 6
18 16 12 12 18 12 6
16 18 12 12 12 12 18 12 9
4 15 12 18 12 15 18 12 15 6
2 4
9 12 12 12 12 9
6 3
Fuel Rod Type Quantity Fuel Rod Type Quantity 1
2 3
4 6
9 12 1
4 3
6 10 4
26 14 15 16 17 18 19 1
7 12 3
11 3
Figure 7: A10-4171B-14GV80-FAA Fuel Rod Distribution AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 11 (b) Adjustments to the thermal conductivity and the fuel melt temperature for gadolinia fuel are described in Reference 2. Please note that Reference 2 extends the range of application of the thermal conductivity model to concentrations up to 8% Gd 20 3 and details on fuel melt temperature are found in Reference 3. The models are included in the approved RODEX2A code (Reference 4) used for the BFN Unit I A TRIUM-IO thermal-mechanical analyses.
AREVA uses a [
I iii. The resulting power history curve for each gadolinia concentration case is shown in Figure 3.1 of ANP-2877P.
The urania rod and the three gadolinia concentration cases along with the PLFR (part length fuel rod) case are analyzed using RODEX2A as described in ANP-2877P.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 12 I.
] for the gadolinia fuel rods also occurs in relation to the PAPT limit shown in Figure 1.1 of the ANP-2877P report.
In summary, the FDL and PAPT limits shown in Figure 1.1 in [
] limit the power levels consistent with the fuel rod thermal-mechanical analysis. In turn, the fuel rod thermal-mechanical analysis results satisfy the fuel melt criteria.
(c) A description of the basic process is as follows. The BFE1-9 calculation reported in Reference 5 is used as an example.
The RODEX2A code contains a provision to evaluate [
]
As shown in Figure 1.1 and described in ANP-2877P, the PAPT limit is f I
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 13 In the case of the urania fuel rod, the application of the factor results in raising the power level from the FDL (Fuel Design Limit or LHGR limit) up to the PAPT limit.
During the transient, f The same calculation also is performed for the gadolinia fuel rods except the steady-state power levels are those shown in Figure 3.1 for the gadolinia rods. The gadolinia power histories are increased by the [
] As described in the response to preceding RIA question, the gadolinia power levels are f I
The calculated fuel centerline temperature outputs from RODEX2A are plotted and shown in Figures 3.2, 3.3, and 3.4 of ANP-2877P for the urania, gadolinia, and PLFR (Part Length Fuel Rod), respectively. In each plot, the lower curve represents the calculated steady-state temperature versus exposure of the axial region with the lowest margin to fuel melt (i.e., at the steady-state power levels shown in Figure 3.1). The middle curve shows the calculated temperatures at the transient (PAPT) power levels for the axial region with the lowest margin to fuel melt. Although the curves are drawn as segmented lines, each symbol on the curve represents a f
]
The three figures also show the fuel melt limit for comparison (upper curves). In all cases, the calculated temperatures at the overpower levels are less than fuel melt.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 14 Although three gadolinia rod cases with varying amount of gadolinia were evaluated, only the gadolinia rod with the smallest margin to melt is shown - this is the [
] From the three different rod types that were analyzed (full length urania rod, PLFR, and gadolinia rod) the f
] gives the lowest margin to the fuel melt temperature.
(d) As described above in the response to RAI 3(b), the fuel melt model is given in References 2 and 3. The urania fuel melt temperature is [
] The melt temperature is reduced with exposure at a rate of
] The melt temperature is further reduced by f I
Using the BFE1-9 calculation as an example (Reference 5), the fuel melt curves for urania and gadolinia fuel rods are shown in Figures 3.2 and 3.3. Table 1 lists fuel melt temperature data points for the urania and gadolinia fuel rods as plotted in the two figures.
Table 1:
f
]
RAI 4: ANP-2877P, Section 3.2 Section 3.2, under bullet "Cladding Collapse," states that "The pellet/clad gap is evaluated [up to a proprietary rod exposure] to ensure the cladding does not [proprietary end state].
Section 3.2.2 "Cladding Collapse" states that gap conditions are evaluated after the first
[proprietary rod exposure stating a different end state than Section 3.2 above].
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 15 After reviewing the proprietary information contained in ANP-2877, Section 3.2, please explain why there is discrepancy between the above two statements and correct the error, if needed.
AREVA Response:
The wording in Reference 5 wasn't intended to be interpreted in a way that two conflicting exposure points are used in evaluating creep collapse. The text in the two sections is not incorrect but Section 3.2 could have been worded differently to avoid the interpretation of an error.
In the first part, Section 3.2, a brief summary is given of the design criteria. The design criteria for avoiding creep collapse restricts the [
] Once the fuel reaches an exposure [
In Section 3.2.2, the evaluation of creep collapse is summarized. The gap condition is evaluated at the first calculation step just [
] when creep collapse can be a concern.
In summary, Section 3.2 summarizes the criteria while Section 3.2.2 is describing the evaluation. The exposure point is the same in both cases and the evaluation is consistent with the criteria.
RAI 5
ANP-2877P, Section 3.2.6 Section 3.2.6 of ANP-2877 states that "the evaluation (for cladding rupture) is covered separate from this report." Identify the location of this report if it is part of the license amendment request or is contained in other docketed material. Otherwise, please provide a copy of this report.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit I AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 16 AREVA Response:
The evaluation for cladding rupture is inherently addressed as part of the LOCA analysis. The NUREG-0630 model is used and is approved in licensing topical report XN-NF-82-07(P)(A)
Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, "Exxon Nuclear Company, November 1982. In LOCA analyses, it is implemented as described in Section 2.3.8 of the approved licensing topical report EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model, "Framatome ANP, May 2001.
RAI 6
ANP-2877P, Sections 3.2.8, and 3.3.7 Discuss the impact of Gd content in Gadolinia rods (U0 2+Gd 2O3) on fuel densification, swelling and fission gas release in fuel rods.
AREVA Response:
As mentioned in Section 3.2.8 of Reference 5, AREVA does not have explicit limits for the in-reactor densification or swelling. Instead, other design criteria such as those related to cladding strain or fuel temperature protect against fuel rod failure due to densification and swelling. It is for this reason that densification and swelling are not directly compared to design limits in ANP-2877P. Also, while fuel rod internal pressure is evaluated according to the design criteria in ANP-2877P, the intermediate results on fission gas release are not provided. More discussion on these three topics is provided below as related to the AREVA analyses with the RODEX2A methodology.
The densification model in the RODEX2A code is described in Appendix K of Reference 6.
When the fuel is fabricated, it is sintered to achieve a large fraction of the theoretical density with the remainder of the volume occupied by small voids and defects. Densification is defined to occur as a result of the size change or annihilation of small voids during irradiation.
The amount of in-reactor densification is based on out-of-reactor resinter tests. These resinter tests are performed on each manufacturing lot of pellets according to guidance from NUREG 1.126. Typically, urania fuel will exhibit[
] change in density from the resinter test.
In contrast, resinter tests on the gadolinia fuel typically give f
] resinter.
The RODEX2A analyses f AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 17
[
] According to the Reg. Guide requirements for application to single or multiple pellet effects, the two resinter limits are used as input to the respective fuel rod thermal-mechanical analyses for fuel temperature and rod pressure. Although the actual resinter values are different, the analyses for urania and gadolinia fuelf
]
From the discussion above, the urania and gadolinia fuels are treated the same way with respect to densification. The [
] For the fuel rod thermal-mechanical analyses, this is a conservative treatment for gadolinia fuel.
In RODEX2A, swelling is modeled as the buildup of solid and gaseous fission products during irradiation. Some of the swelling can be [
] More discussion of the impact of gadolinia on swelling is presented below in the discussion of fission gas release.
The combined effect of swelling and densification is shown in Figure 8. The results are from the fuel rod thermal-mechanical analysis for the Browns Ferry A TRIUM-1O design. The [
case is compared to the urania fuel rod case.
AREVA NP Inc.
Browns Ferry Unit I AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 18 Figure 8: Comparison of Fuel Swelling and Densification As can be seen, the main impact of the gadolinia in the analysis is to[
] Details on the RODEX2A fission gas release model are described in References 4, 6, and 7.
Additional gadolinia rod fission gas release benchmark data were presented to the NRC in Reference 8, The RODEX2A code was found to conservatively predict the fission gas release in the expanded database of gadolinia rods [
I AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 19 I
] As part of the RODEX2A methodology, a commitment was made to the NRC to include the following additional criteria (Reference 8).
The two conditions above continue to be observed in connection with the RODEX2A methodology and were applied to the Browns Ferry A TRIUM-IO design.
The combined impact of the gadolinia design
]
on the fission gas release in the Browns Ferry A TRIUM-IO fuel rod thermal-mechanical analysis is illustrated in Figure 9. The [
] case and the urania case are plotted for comparison.
Figure 9: Comparison of Fission Gas Release The design case comparison shows [
I AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 20
RAI 7
ANP-2877P, Section 3.3.8 Section 3.3.8 states that "Mixed core conditions for liftoff are considered on a specific basis as determined by the plant and other fuel types. Analyses to date indicate a large margin to liftoff under normal operating conditions." Justify this claim by providing a summary of the analysis and calculations.
AREVA Response:
The analysis of record for Browns Ferry has a minimum [
] margin to assembly liftoff under normal operating conditions and f
] under anticipated operational occurrences (AO0). The margin to liftoff is the difference between the wet weight of the fuel assembly and hydraulic forces. The hydraulic forces are calculated from conservative thermal-hydraulic evaluations which predict a maximum core pressure drop of [
I.
Thermal-hydraulic compatibility studies have shown that the core pressure drop with a typical GNF fuel design could be a maximum of[
] the effect of mixed cores is negligible.
RAI 8
ANP-2877P, Sections 3.4.1 through 3.4.3, Table 3.3 Item 3.4.2 (a) [deleted]
(b) Item 3.4.2 of Table 3.3 indicates that violent expulsion of fuel criteria for fuel is less than 280 calories per gram (cal/g) for coolability, and is less than 170 cal/g for rod failure. Standard Review Plan (SRP, NUREG-0800) Section 4.2, Appendix B, Section C (Core coolability criteria) stipulates that for fuel rod thermal-mechanical calculations, employed to demonstrate compliance with Criteria 1 (peak radial average fuel enthalpy must remain below 230 cal/g) and Criteria 2 (peak fuel temperature must remain below incipient fuel melting conditions), must be based upon design-specific information accounting for manufacturing tolerances and modeling using the NRC-approved methods, including burnup-enhanced effects on pellet power distribution, fuel thermal conductivity, and fuel melting temperature. Provide justification for the items in 3.4.2 of Table 3.3 in support of the Standard Review Plan acceptance criteria, specifically, the difference between the SRP value for coolability (less than 230 cal/g) and the value in Table 3.3, Item 3.4.2.
AREVA Response:
(b) RESPONSE TO BE PREPARED BY TVA; AREVA INPUT NOT REQUIRED AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAls Page 21
RAI 9
ANP-2821(P) Thermal-Hydraulic (T-H) Design Report, Section 3.1 Provide a summary of detailed calculations for thermal-hydraulic characterization for the ATRIUM 10 reload fuel for BFN Unit 1. These calculations should show how the licensee obtained the loss coefficients and friction factors listed in Table 3.3 of ANP-2821; for example, upper tie plate (UTP) loss coefficient, spacer loss coefficients, lower tie plate (LTP) grid loss coefficients, orifice, and LTP loss coefficients.
AREVA Response:
The thermal-hydraulic characterization of the A TRIUM-10 reload fuel for BFN Unit I is based on the NRC approved pressure drop methodology of Reference 9. Reference 9 discusses how loss coefficients and friction factors are obtained. A summary of the detailed calculations for the A TRIUM-IC loss coefficients and friction factor listed in Table 3.3 of ANP-2821(P) is provided as follows.
t]
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 22 I
I RAI 10: ANP-2821(P), Section 3.2 Provide detailed calculations that demonstrate thermal-hydraulic compatibility of ATRIUM 10 with the co-resident GE14 fuel in BFN Unit 1. These calculations should show the following:
(a) Calculations should demonstrate that during the entire transition from a full core GE14 to a full core ATRIUM 10 fuel, there will be no major impacts on thermal-hydraulic operation of BFN Unit 1 and should demonstrate compliance over the entire licensing range of the power/flow map.
(b) Justify your selection of bottom-peaked axial power distribution as a basis for the hydraulic compatibility results, compared to the results for top-and middle-peaked axial power distributions.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 23 AREVA Response:
(a) Results of the thermal hydraulic compatibility analysis for Browns Ferry Unit 1 are presented in Tables 3.5 - 3.8 of ANP-2821(P). Analysis results are presented for three core configurations including a representative mixed core configuration for Browns Ferry Unit I (ATRIUM-10 and GE14 fuel), which is also identified as the "Transition core" in the report. These summary tables present results that show core performance as Browns Ferry Unit 1 transitions from a full core of GE14 fuel to a full core of A TRIUM-I O fuel. All fuel assemblies are explicitly modeled, both hydraulically and neutronically in the core monitoring software system (CMSS). Actual operation of Browns Ferry Units 2 and 3 transitions further demonstrate there are no hydraulic compatibility issues between GE14 and ATRIUM-10. The results demonstrate:
I AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAls ANP-2988NP Revision 0 Page 24 I
I (b) Calculations are explicitly performed for a bottom-, middle-, and top-peaked axial power distribution. The power shape dependent results for thermal hydraulic compatibility show similar trends for each of the three distributions. f I
Table 2:f RAI 11: Thermal margin performance (a) Discuss the impact of part length rods and Gadolinia (U0 2 + Gd 20 3) on the application of SPCB critical power correlation.
(b) Will any of these part length rods undergo boiling transition/dryout during normal operating conditions or during transients and accident conditions?
AREVA Response:
(a) The SPCB critical power correlation treats the part length rod no differently than a full length rod. It has an additive constant and the correlation considers if this is the limiting rod in the fuel assembly.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 25 There is no difference between the way that the CPR is calculated if the rod contains gadolinia or if the rod does not contain gadolinia. The appropriate heat generation rate is determined by the core simulator (MICROBURN-B2) for each rod used in the determination of the CPR. All fueled rods are considered in the CPR calculation.
(b) The approved methodologies and COLR tech spec limits assure that more than 99.9%
of the rods avoid boiling transition during transients. More than 99.9% of the rods also avoid boiling transition during normal operating conditions, because margin to boiling transition is greater under normal operating conditions than during transient conditions.
While the occurrence of boiling transition may not occur in some accident scenarios, the avoidance of boiling transition is not one of the acceptance criteria for accidents. Boiling transition will likely occur in some accidents. For example, during a loss of cooling accident, boiling transition is expected for most of the rods (including part length rods) in the hot assembly.
RAI 12: Mixed core and Critical Power Ratio (CPR) calculations The proposed BFN Unit 1 core with AREVA Atrium 10 and GE14 fuel designs will constitute a "mixed core."
Provide details of the impact of the mixed core on the CPR calculations, accounting for the differences in mechanical, thermal and hydraulic characteristics of the two fuel designs in the transition core at BFN Unit 1.
AREVA Response:
Tables 3.7 and 3.8 of ANP-2821(P) include CPR results for high power and average power assemblies for a mixed core configuration as Browns Ferry Unit I transitions from a full core of GE14 fuel to a full core of ATRIUM-IO fuel. The transition core configuration used in the analyses is presented in Table 3.4.
In the AREVA thermal hydraulic methodology, each fuel type is explicitly modeled. As a result, the impacts of the differences in mechanical design on geometry and loss coefficients are explicitly accounted for. The critical power performance of each fuel type is also explicitly modeled using the applicable critical power correlation for each fuel design. The SPCB critical power correlation (EMF-2209(P)(A) Revision 3 "SPCB Critical Power Correlation') is used for AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 26 ATRIUM-10 fuel and GEl4 fuel. Application of the SPCB correlation to GE14 fuel is based on the indirect process described in EMF-2245(P)(A) 'Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel."
RAI 13: ANP-2821P, Section 3.6 Stability General Design Criterion 12 of Title 10 of the Code of Federal Regulations, Part 50 Appendix A requires suppression of reactor power oscillations so that the Specified Acceptable Fuel Design Limits are not exceeded. Demonstrate, with supporting analyses and calculations, how thermal-hydraulic and neutronic stability of the mixed core will be maintained at the BFN Unit 1 throughout the upcoming and following cycles of operation.
AREVA Response:
Browns Ferry Unit 1 is a detect and suppress Option I/l plant that utilizes a PRNM-based system compliant with NEDO-32465(A). AREVA performs cycle-specific calculations to determine the required Option I/l system setpoint(s) necessary to ensure that the MCPR Safety Limit is not exceeded. This includes the calculation of the DIVOM (Delta over Initial MCPR Versus Oscillation Magnitude) as well as delta-MCPR response to a two recirculation pump trip (2RPT) event using the cycle-specific licensing basis core. The OLMCPR versus setpoint is provided for both a steady-state and the 2RPT events as described in NEDC-32465(A).
For times in which the primary OPRM system is not available, regions are defined on the power-flow map in accordance with the Backup Stability Protection (BSP) described in 0G02-0119-260 "Backup Stability Protection (BSP) for Inoperable Option /// Solution." These regions have been defined conservatively for Browns Ferry Unit 1 and are confirmed on a cycle-specific basis.
Cycle-specific stability analyses are performed based upon the actual licensing basis core design, which explicitly includes co-resident fuel designs. Both the OPRM setpoints versus OLMCPR and the backup stability regions are provided in the cycle-specific Reload Safety Analysis report. These results are based upon explicit modeling of the Browns Ferry Unit I mixed core including the A TRIUM-I and GE14 fuel types.
RAI 14: ANP-2859P Section 2.0 Provide a reference or summarize the methodology by which BFN Unit 1 is designed to achieve 71 Gigawatt-days of additional energy via final feedwater temperature reduction operation, beyond the full power capability.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAls Page 27 AREVA Response:
RESPONSE TO BE PREPARED BY TVA; AREVA INPUT NOT REQUIRED RAI 15: Shutdown Margin Describe the analysis procedure used to ensure that the shutdown margin is within the TS limit throughout the transition cycles. Specifically, address how the eigenvalue biases and uncertainties are determined and accounted for during the transition cycles.
AREVA Response:
AREVA performs a cycle specific shutdown margin calculation as part of the plant design and reload licensing analyses. The result for Browns Ferry Unit I Cycle 9 is reported in Section 3.3 of Reference 10. The shutdown margin analysis is performed using the NRC approved AREVA CASMO-4/MICROBURN-B2 neutronics analysis methodology documented in Reference 11.
For Browns Ferry Unit 1 Cycle 9, a series of shutdown margin restart calculations were performed at the desired cycle exposure points throughout the cycle using the MICROBURN-B2 control rod step out restart files. The analyses were repeated for the case of short, nominal, and long Cycle 8 EOC conditions. The calculations are performed at cold conditions with no xenon. The fuel assembly cold lattice nuclear data is generated with the CASMO-4 computer code as a function of lattice type, exposure, void history, and control. The shutdown margin is calculated with the MICROBURN-B2 reactor simulator code using nuclear data for the lattices generated by CASMO-4 calculations. A series of one-rod-out cold calculations is performed with MICROBURN-B2 in order to identify the most limiting control rod location in the core. The difference between the cold critical k-effective and the limiting one-rod-out calculated k-effective at cold conditions determines the shutdown margin.
The cold critical k-effective is determined based on data from actual startup tests for the Browns Ferry units. Figure 10 shows the data for the Browns Ferry units from benchmarking cycles with a full core of GE fuel and from the transition cycles, along with the cold critical k-effective used in the determination of shutdown margin for the Cycle 9 design. The data exhibits no dependence on unit number, and no difference between full cores of GE fuel and the transition cores. For Unit 1 Cycle 9 design and licensing purposes, the cold critical k-effective was conservatively drawn below all the data. This is more straighfforward than an approach where a AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 28 nominal average cold critical k-effective is determined and then a bias or uncertainty is applied to account for uncertainty. As Browns Ferry Unit I moves on through the transition cycles these plots will be revisited and any new data resulting from additional cold criticals at Browns Ferry will be added. The cold critical k-effective will be adjusted if new data justifies any change.
Figure 10: Browns Ferry Cold Critical Data for GE Cores and Transition Cores RAI 16: EMF-2158 Licensee has used EMF-2158 methodology to perform fuel cycle design and fuel management calculations for the Cycle 10 operation of BFN Unit 1 to generate nuclear data including cross sections, local power peaking factors, and associated uncertainties.
Section 5 of XN-NF-80-19(P)(A), Volume 1, Supplement 3, and Section 9 of EMF-2158(P)(A) together provide a very detailed description of the analyses and calculations to determine the traversing in-core probe detector (TIP) uncertainty components for boiling-water reactors.
Sections 9.4 and 9.5 provide combined uncertainties for TIP distribution calculations, TIP distribution measurement, net calculated TIP distribution and synthesized TIP distribution uncertainty. Provide details of the calculations and uncertainties listed in Chapter 9 of AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 29 EMF-2158 applicable to the D-lattice BFN Unit 1 plant. Show that the BFN Unit 1 uncertainties documented in EMF-2158 for D-lattice plants remain conservative.
AREVA Response:
The calculated and measured TIP uncertainties are the most significant components of the measured power distribution uncertainty. The uncertainty in the measured assembly power is defined in EMF-2158(P)(A) as:
SD2 + ST2
=
NIJ Where:
SB,
=
0.25 + 0.75p p 2
LPRM2 S
( S m )2 +
JR
.5Dij
( '1I 4
+
i2 8T, 5=
Relative standard deviation in the calculated TIP response distribution NIJ
=
Number of LPRM's (4) used for the measured Power determination PJ=
Correlation coefficient between neighboring radial nodes 8T*7,
=
Relative standard deviation in the radial TIP measurement SLPRM
=
Relative standard deviation in the measured LPRM response 5Su
=
TIP synthesis procedure uncertainty The value of these parameters for the Browns Ferry specific analysis are compared to the topical report values for D-lattice plants in the following table.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 30 The EMF-2158(P)(A) methodology determines the calculated TIP uncertainty (6Tu,) from separately determined uncertainty components. The two uncertainty components used to calculate 6 T, are:
the deviation between the CASMO-4/MICROBURN-B2 (C4/MB2) calculated radial TIP response and the measured radial TIP response ('5Tý.) and radial TIP measurement uncertainty (577) gT =~ V(67&)(gTr) 2 These uncertainty components are determined using traversing incore probe (TIP) measurements, which are taken at or near full power conditions for Local Power Range Monitor (LPRM) calibration.
The 5T7' is comprised of random instrument error and geometric measurement uncertainty caused by variations in the physical TIP location. A BFN specific radial TIP measurement uncertainty (8577) was calculated in accordance with the EMF-2158(P)(A) methodology using BFN gamma TIP measurements and is
]. For comparison, EMF-2158(P) (A) reports a 8T" off
] for D-Lattice plants. The BFN gamma TIP system is far less sensitive than neutron TIP systems to variations in TIP location within the comer water gap between fuel assemblies. Because 8T' is determined by comparing TIP measurements in symmetrically operated core locations, it is independent of the C4/MB2 core model and core operating conditions.
The LPRM uncertainty is based on the instrumentation and is not plant specific.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit I AREVA Fuel Revision 0 Transition Input to TVA for RAIs Page 31 The 8SS. is the uncertainty associated with update of calculated power by the core monitoring system to more closely match in-core instrumentation. A BFN specific radial synthesis uncertainty (6SSj) was calculated in accordance with the EMF-2158(P)(A) methodology using BFN gamma TIP measurements and is [
]. For comparison, EMF-2158(P)(A) reports a 8So. of[
] for D-Lattice plants. 5SSj is a function of the core monitoring system update algorithm and is independent of core operating conditions.
Utilizing the BFN specific values of 5T', 5Ttm and (S5. results in a measured assembly power distribution uncertainty of
]. BFN Safety Limit Minimum Critical Power Ratio (SLMCPR) analyses are based on the radial bundle power uncertainty value of[
] reported in the EMF-2158(P)(A) topical report rather than the BFN specific value of
]. The BFN specific value is conservative relative to the topical report value by [
] due primarily to BFN implementation of gamma TIPs for LPRM calibration.
The EMF-2158(P)(A) topical report database includes TIP measurements of cores containing many different fuel designs and identifies no correlation between C4/MB2 uncertainty and fuel design. Figure 11 demonstrates there is no significant variation in uncertainty determined from the BFN gamma TIP measurements for various mixes of fuel types. These measurements include mixed GE13 and GE14 cores operated in Unit 2 Cycle 11 and Unit 3 Cycle 13. Mixed cores of GE14 and ATRIUMTM-1O* fuel were operated in Unit 2 Cycles 14 and 15 as well as Unit 3 Cycles 12 and 13. BFN Unit 1 is an identical core design compared to BFN Units 2 and 3 (core geometry and core instrumentation) and is composed of fuel types nearly identical to those used in BFN Units 2 and 3. Data for BFN Unit I is expected to result in similar standard deviations as seen for BEFN Units 2 and 3. The uncertainties have been demonstrated to be independent of fuel type or core design.
AREVA NP Inc.
Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs ANP-2988NP Revision 0 Page 32 Figure 11 BFN Gamma 5T, TIP Response Versus Cycle Number C4/MB2 local power distributions are compared to bundle gamma scan data as reported in Tables 8.3, 8.4, and 8.5 of EMF-2158(P) (A) for IOx 10 and other orthogonal lattice designs.
These results indicate that there is no degradation in the uncertainty for IxW10 fuel relative to the other designs.
These results demonstrate that the uncertainties documented in EMF-2158 for D-lattice plants remain conservative for BFN Unit 1.
AREVA NP Inc.
ANP-2988NP Browns Ferry Unit 1 AREVA Fuel Revision 0 Transition Input to TVA for RAls Page 33 3.0 References
- 1.
Letter, C. Gratton (NRC) to R. M. Krich (TVA), "Browns Ferry Nuclear Plant, Unit 1 -
Request for Additional Information Regarding Amendment Request to Transition to AREVA Fuel (TAC No. ME3775)," USNRC, January 24, 2011. (38-9154144-000).
- 2.
XN-NF-85-92(P)(A) Revision 0, Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, Exxon Nuclear Company, Inc.,
November 1986.
- 3.
XN-NF-79-56(P)(A) Revision 1 and Supplement 1, Gadolinia Fuel Properties for LWR Fuel Safety Evaluation, Exxon Nuclear Company, November 1981.
- 4.
XN-NF-85-74(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Exxon Nuclear Company, August 1986.
- 5.
ANP-2877P Revision 0, Mechanical Design Report for Browns Ferry Unit I Reload BFEI-9 ATRIUM-10 Fuel Assemblies (105% OLTP), AREVA NP Inc., November 2009.
- 6.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 7.
EMF-85-74(P) Revision 0 Supplement 1 (P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
- 8.
Letter from R. A. Copeland (AREVA NP) to R. C. Jones (NRC), "Gadolinia Bearing Fuel Rod Design Methodology," Siemens Power Corporation, RAC:050:91, May 13, 1992.
- 9.
XN-NF-79-59(P)(A), Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies, Exxon Nuclear Company, November 1983.
- 10.
ANP-2859(P) Revision 1, Browns Ferry Unit I Cycle 9 Fuel Cycle Design (105% OLTP),
AREVA NP, October 2009.
- 11.
EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
AREVA NP Inc.
ENCLOSURE 4 Tennessee Valley Authority Browns Ferry Nuclear Plant, Unit I Technical Specifications Change TS-473 - AREVA Fuel Transition Affidavit
AFFIDAVIT STATE OF WASHINGTON
)
) ss.
COUNTY OF BENTON
)
- 1.
My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.
- 2.
I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.
- 3.
I am familiar with the AREVA NP information contained in the report ANP-2988P Revision 0, entitled, "Browns Ferry Unit 1 AREVA Fuel Transition Input to TVA for RAIs," dated February 2011 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.
- 4.
This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5.
This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is
requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."
- 6.
The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:
(a)
The information reveals details of AREVA NP's research and development plans and programs or their results.
(b)
Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c)
The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.
(d)
The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.
(e)
The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.
The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.
- 7.
In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8.
AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
w
- 9.
The foregoing statements are true and correct to the best of my knowledge, information, and belief.
A, ý SUBSCRIBED before me this day of V- *o 2011.
44A Susan K. McCoy NOTARY PUBLIC, STATE OFVAýHINGTON MY COMMISSION EXPIRES: 1/10/12