ML20202H567
| ML20202H567 | |
| Person / Time | |
|---|---|
| Issue date: | 02/09/1998 |
| From: | NRC (Affiliation Not Assigned) |
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| ML20202H544 | List: |
| References | |
| NUDOCS 9802200414 | |
| Download: ML20202H567 (1) | |
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ENCLOSURE SAFETY EVALUATION OF SIEMENS POWER CORPORATION TOPICAL REPORT EMF 45 74(P), REVISION 0, SUPPLEMENTS 1 AND 2,.
"RODEX2A (BWR) FUEL ROD THERMAL-MECHANICAL EVALUATION MODEL" 1 INTRODUCTION in letters dated January 18 and August 13,1997, from H. D. Curet, Siemens Power Corporation (SPC), to the U.S. Nuclear Regulatory Commission (NRC), SPC submitted a Topical Report EMF 45-74(P), Revision 0, Supplements 1 and 2, "RODEX2A (BWR) Fuel Rod,-
Thermal Mechanical Evaluation Model," for NRC review.
EMF 45 74(P), Revision 0, supplements 1 and 2, describes an improved fuel rod performance code, RODEX2A, and the associated methodology that SPC intends to apply for fuel reload licensing applications in boiling-water reactors (BWRs). The purpose of these improved code and methodology is to extend the analytical capability to a slightly higher bumup range than the previously approved bumup range. The new code and methodology improve the analytical predictions of fission gas release, fuel rod growth, fuel assembly growth, fuel channel growth, etc.
The NRC staff was supported in this review by its consultant, Pacific Northwest National Laboratory (PNNL). Our consultant's technical evaluation report (TER), which is attached, provides technical findings relative to its review.
2 EVALUATION The staff has reviewed the enclosed TER, and concludes that the TER provides an adequate technical basis to approve EMF 45 74(P), Revision 0, Supplements 1 and 2.
Therefore, the staff agress with PNNL's conclusion that the improved RODEX2A code and the associated methodology described in EMF 45-74(P), Revision 0, Supplements 1 and 2, are acceptable. Based on our review, the staff adopts the findings in the attached TER.
3 CONCLUSIONS The staff has reviewed the SPC's improved RODEX2A code and the associated
- methodology described in EMF 45-74(P), Revision 0, Supplements 1 and 2, and finds that the code and methodology are adequate. Therefore, we conclude that (1) the RODEX2A code is acceptable for licensing applications to 62,000 mwd /MTU rod-average bumup, and L (2) the fuel rod growth, fuel assembly growth, and fuel channel growth models and analytical methods are acceptable for ATRIUM 9 and 10 fuel designs up tc 54,000 mwd /MTU assembly average bumup.
9802200414 980209
t TECHNICAL EVALUATION REPORT OF THE TOPICAL REPORT EMF-85-74(P), REVISION 0, SUPPLEMENTS 1 AND 2 "RODEX2A (BWR) FUEL ROD THERMAL-MECHANICAL EVALUATION MODEL" 4
C.E. Beyer D.D. Lanning January 1998 -
Prepared for Reactor Systems Branch Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Under contract DE-ACM-76RLO 1830 NRC JCN J 2435 Pacific Northwest National Laboratory Richland, Washington 99352 J
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TABLE OF CONTENTS o
1.0 INTRODUCTION
1.1 2.0 APPLICATION OF RODEX2A TO 62 GWd/MTU......... -................. 2.1 2.1 CLADDING STRAIN......................................... - 2.1 2.2 CLADDING OXID ATION.................................... 2.1 2.3 ROD INTERNAL PRESSURE................................. 2.2 i
2.4 FUEL MELTING.............................................
- 2. 3
- 3.0 ATRIUM 9 AND 10 DESIGNS TO 54 GWd/MTU........................... 3.1 -
3.1 FUEL E OD GROWTH.....................................
3.2 3.2 AS S EMBLY GROWTH....................................... 3.2 4
3.3 FUEL CHANNEL AND WATER CHANNEL GROWTH...........,,. 3.4 4.0 CONCLUS IONS................................................ 4.1
5.0 REFERENCES
......... 5.1
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1.0 114TRODUCTION This Technical Evaluation Report (TER) was prepared by Pacific 44orthwest National Laboratory (PNNL) under U.S. Nuclear Regulatory Commission (NRC) job code number J-2435. The TER sununarizes the review of Siemens Power Corporation (SPC) submittals j
Supptement 1 (Reference 1) and Supplement 2 (Reference 2) to EMF-35-74(P), Revision 0. The two supplements are related because both are needed to extend the burnup levels of SPC's latest Boiling Water Reactor (BWR) designs. Supplement I requested a change to the burnup limit of l
the application of the RODEX2A fuel performance code for BWR fbol designs from a peak pellet -
exposure of 60 GWd/MTU to a rod-average exposure of 62 GWd/MTU. - The original NRC approval of RODEX2A (Reference 3) limited the code application to a peak pellet burnup of 60 GWd/MTU. The original approval also limited the range ofRODEX2A application for end-of-L life (EOL) rod internal pressure analyses to between 50 and 60 GWd/MTU (peak pellet).
Supplement I has provided RODEX2A comparisons to fission gas release data and example fuel melting and (EOL) rod internal pressure analyses up to rod-average burnups of 62 GWd/MTU.
In addition, Supplcment I has provided BWR cladding corrosion data to demonstrate satisfactory 1
corrosion levels for SPC BWR designs.
In conjunction with this request, Supplement 2 requests an extension in the assembly-average exposure from 48 GWd/MTU to 54 GWd/MTU for the ATRIUM 9 and ATRIUM 10 fuel design methods to coincide with the request for RODEX2A in Supplemeat 1. Supplement 2 has presented rod, assembly and channel axial growth data and analyses to demonstrate that clearances will be maintained at the burnup levels requested.
The following review is divided into two parts; 1) tle increase in bumup limit for RODEX2A application from 60 GWd/MTU (peak pellet) to 62 GWd/MTU (rod average), and
- 2) the increase in assembly-average burnup limit from 48 to 54 GWd/MTU for the models and methods specific to the ATRIUM 9 and 10 fuel designs.
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2.0 APPLICATION OF RODEX2A TO 62 GWd/MTU The RODEX2A code is to be applied to steady-state thermal mechanical BWR fuel rod design analyses and is not used for calculating initial stored energy calculations for loss of.
coolant accident (LOCA) analyses. The initial stored energy analyses for LOCA are performed with the RODEX2 code that has been approved to a rod-average burnup of 62 GWd/MTU. The models in the SPC fuel performance codes and their analysis methods are usually not design dependent and can generally be applied to any BWR design. There are only a few RODEX2A thermal-mechanical models and analyses the code is applied to that.re impacted by 'he increase in bumup requested.- The RODEX2A models that are impacted by burnup are cladding oxidation, fission gas release, and fuel thermal conductivity. The RODEX2A analyses that are impacted at higher burnups are cladding strain, cladding oxidation, EOL rod intemal pressures and fuel
- melting, a
2.1 CLADDING MTRAIN The RODEX2A cladding strain analysis accounts for the effects of the requested bumup increase. In addition, the SPC BWR criteria for strain is reduced when burnups exceed 60 GWd/MTU. This is consistent with the SPC cladding strain criteria for PWRs that is approved by the NRC, Therefore, both the RODEX2A analysis methods and the SPC BWR design criteria for cladding strain are acceptable up to the burnup level requested.
2.2 CLADDING OhIDATION There are no SPC BWR design criteria that limit cladding corrosion other than to include the effects of oxidation in cladding stress, thermal and rod internal pressure analyses. However, the NRC has recently become more concemed with cladding corrosion at high bumup because thermal effects become significant and cladding mechanical properties such as ductility are adversely impacted. NRC has recently initiated a research program to examine high bumup effects on fuel rods including cladding corrosion, ductility, and other mechanical properties.
The RODEX2A cladding oxidation analysis accounts for the thermal effects of oxidation in the fuel thermal (e.g., melting) and mechanical (e.g., EOL rod internal pressure and cladding stress) analyses. The RODEX2A code significantly underpredicts BWR cladding corrosion. This underprediction was discussed with SPC and they noted that they have a commitment (Reference
- 4) to increase the corrosion predictions in RODEX2 A' to match the maximum corrosion levels from their BWR fuel cladding data provided in Figure 4 of Supplement 1. This figure shows SPC BWR fLel cladding corrosion data up to a rod-average burnup of 58 GWd/MTU with corrosion i
levels significantly less (factor of 2) than the level of concern for cladding corrosion. Therefore, the RODEX2A underprediction of corrosion is not of significant consequence in thermal analyses at current corrosion levels but should be corrected because it will become more significant as 2.1
i bumup levels increase. The underprediction of corrosion in the EOL rod internal pressure analyses is more than compensated by the large degree of conservatism that exists in the RODEX2A fission gas release model. However, this conclusion may not be valid should SPC elect to reduce the conservatism in the fission gas release model in the future. In addition, SPC i
conservatively accounts for cladding corrosion in their EOL cladding stress calculation by reducing the cladding thickness by an amount significantly greater than the maximum SPC has I
observed for BWR cladding corrosion. PNNL concludes that the RODEX2A cladding corrosion model is acceptable based on SPC's commitment (Reference 4) to increase the corrosion prediction for thermal analyses to match the maximum corrosion level observed in their data.
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Cisdding corrosion is acceptable for SPC BWR designs up to the burnup limit requested based on the data in Figure 4 of Reference 1.
2.3 ~
ROD INTERNAL PRESSURE RODEX2A is used for the SPC EOL rod internal pressure analysis. This analysis is strongly dependent on the prediction of fission gas release which in tum is burnup dependent.
I SPC has provided RODEX2A comparisons to fission gas release data at rod-average burnups up to 62 GWd/MTU and also provided an example rod pressure analysis up to a rod average burnup of 62 GWd/MTU in Supplement 1. The code comparisons to data appear to provide a conservative prediction, of fission gas release at high burnup levels, i.e., greater than 50 GWd/MTU, but the amount ofgood quality SPC data at high burnups is limited. The code appears to provide non-conservative calculation at rod-average bumups below 30 GWd/MTU.
SPC was requested (Reference 5) to: 1) perform code predictive comparisons to newer -
fission gas release data at rod-average bumups of 20 and 40_GWd/MTU,2) provide the input used to perform their example calculations in order that the same calculations could be performed with NRC's fuel performance code, and 3) supply their predicted gas release values from their -
example rod internal pressure analysis. SPC's response (Reference 6) provided the code-data comparisons, the example calculation input, and the predicted fission gas release for their rod pressure analysis. The RODEX2A data comparisons to the new fission gas release data resulted in a significant under prediction of the datum at 20 GWd/MTU but a reasonably good prediction (within the uncertainty of the data) of the datum et 40 GWd/MTU.
An audit calculation of the SPC rod internal pressure calculation was performed using the newly developed FRAPCON-3 code and the SPC supplied input. The FRAPCON-3 code has been verified apinst a very well qualified data base that includes fission gas release data up to rod-average burnups of 72 GWd/MTU. Comparison of the FRAPCON-3 audit calculational results and the RODEX2A rod pressure calculational results showed that RODEX2A underpredicted the fission gas release and rod internal pressures predicted by FRAPCON-3 between rod average bumups of 13 to 30 GWd/MTU. Both codes predicted similar values between 32 to 42 GWd/MTU (rod average) but at higher burnups the RODEX2A code began predicting higher rod pressures with the code predictions significantly higher than FRAPCON-3 J.
2.2 v
above 50 GWd/MTU. The large (conservative) overprediction of FGR at high burnups is of greatest significance for this review because the EOL rod pressure analyses generally establish the LHGR limits for BWR designs above 50 GWd/MTU. The underpredictions at low burnups (i.e.,
less than 30 GWd/MTU, rod average) are not of significance. This is because those analyses that are performed at low burnups such as LOCA and fuel melting either do not use RODEX2A (foi LOCA) or, as demonstrated below for fuel melting, the code remains conservative for thermal analyses. PNNL concludes that the application of the RODEX2A code for EOL rod intemal pressure analyres is acceptable up to rod average bumups of 62 GWd/MTU.
2A FUEL MFI TING The RODEX2A code is used for the SEC fuel melting analysis and SPC provided an example calculation in their submittal of Supplement 1. This analysis is strongly dependent on fuel thermal conductivity that has recently been found to decrease with fuel burnup. SPC was l
questioned on what SPC does for their fuel melting analyses to account for this thermal l
conductivity degradation with bumup (Reference 5). SPC responded (Reference 6) that the code i
accounts for this through code conservatisms and the assumed formation of fuel voids with increasing burnup. This was verified by an audit calculation using the FRAPCON-3 code and the same input as used for the RODEX2A analysis. The RODEX2A results predicted higher fuel temperatures (more conservative) than FRAPCOW 3 up to a rod average burnup of 20 GWd/MTU and then predicted similar or slightly conservative results up to the bumup level requested. The thermal predictions for the rod pressure analysis were also examined and found to provide a nearly identical pattern of comparison to FRAPCON-3 versus burnup as for the fuel melting analysis. Therefore, PNNL concludes that the SPC application of RODEX2A for thermal analyses including fuel melting are conservative and acceptable for application up to the bumup level requested.
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, 3.0 ATRIUM 9 AND 10 DHIGNS TO 54 GWd/MTU -
l In addition to the BWR fuel performance code RODEX2A, there are other SPC models
'and analysis methods that are both BWR design and burnup dependent that need to be addressed.
in the Supplement 2 submittal. The models in question for the ATRIUM 9 and 10 designs are rod
-- growth, assembly growth and channel growth. The cladding oxidation model in RODEX2A could also theoretically be dependent on rod size (and therefore design dependent) but comparison of l
ATRIUM 9 and 10 oxidation data to previous SPC 8x8 fuel designs with similar Zircaloy-2 L
cladding has demonstrated that the oxidation of the newer designs is within the scatter of the earlier 8x8 designs utdizing the same Zircaloy cladding.
The SPC analyses that are impacted by design and burnup are the following; 1) the fuel rod end cap to tie plate clearance or engagement, and 2) the engagement of the lower tie plate seal spring and the fuel channel. The rod to tie plate clearance or engagement depends on rod and assembly growth. A short description of the concern with engagement / clearance is provided in the following. Both fuel designs, ATRIUM 9 and 10, have the same fuel rod / upper tie plate configuration; i.e., there are holes in the tic plate that the fuel rod end cap shank fits through for i
1 engagement so that the tie plate provides lateral support for the fuel rod at the upper end. Below
- the end cap shank of the fuel rod there is a shoulder where the end cap is welded to the cladding.
The diameter of this shoulder is larger than the diameter of the tie plate holes. For the ATRIUM 9 design there is a concern with the en' cap shank remaining engaged in the tie plate d
1 holes because the fuel assembly (fixed to the tie plate) grows faster than the fuel rods. If the fuel rods become disengaged from the tie plate this will result in less lateral support for the fuel rods at the top which could result in rod vibration and fretting wear, For the ATRIUM 10 design the concem is with clearance between the end cap shoulder and the upper tie plate because the fuel -
rods grow faster than the assembly. If the end cap shoulder hits the tie plate this will result in fuel rod bowing and possible failure. This difference in concern between fuel rod engagement for ATRIUM 9 and fuel rod clearance for ATRIUM 10 is because the assembly growth for each design is different.-
The difference in assembly growth for the two designs can be explained by the -
differences in the structural skeleton of each assembly. The ATRIUM 9 design has tic rods made ofcold worked Zircaloy mechanically connected to the upper and lower tie plates for the structural skeleton. The ATRIUM 10 design has a central water channel made of fully annealed Zircaloy that is permanently connected to the lower and upper tie plates that make up the
- assembly skeleton. The ATRIUM 9 tie rods made of cold worked Zircaloy grow faster (by greater than a factor of 2) than the ATRIUM 10 water channel made of fully annealed Zircaloy.
The ATRIUM 9 and 10 fuel designs both use cold worked Zircaloy for the fuel cladding. The tie rods used in the ATRIUM 9 desigr, as noted earlier are also made of cold worked Zircaloy, but grow faster than the fuel rods because the tie rods have tensile stresses that the fuel rods do not experience. The tensile stresses in the tie rods are the cause of their greater growth as compared to the fuel rod growth. Therefore, the ATRIUM 9 design has a concern with fuel rod to tie plate engagement because the fuel assembly grows at a greater rate than the fuel rods while the 3.1
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" ATRIUM 10 design has a concern with rod to tie plate clearance because the fuel rods grow
-- faster than the assembly.
t The rod end cap to tie plate clearance or engagement depends on rod growth and assembly growth. In contrast, the engagement of the lower tie plate seal spring to the fuel
- channel ls dependent on assembly and fuel channel growth. The SPC ATRIUM 9 and 10 ndels and methods for evaluating fuel rod, assembly and fuel channel growth are discussed in the following sections.
L 3.1 FUEL ROD GROWTH The SPC BWR rod growth model is used in the rod to tie plate eng;gement/ clearance analyses. The same growth model is used for both designs, the methodology for application is different for the two designs. The SPC analysis of rod to tie plate engagement for ATRIUM 9 designs conservatively uses the 95% lower bounding curve of the growth model while the ATRIUM 10 analysis for rod to tie plate clearance conservatively uses the 95% upper bound curve of the model. The model has a limited amount of rod growth data above an assembly-average burnup of 48 GWd/MTU. However, SPC has also presented several rod growth data from PWR rods (with cold worked Zircaloy-4 cladding) beyond the burnup level requested that shows similar growth as predicted by the BWR growth model and that lie within the uncertainty of the BWR data. This data demonstrates that the rod growth model is a valid representation of the data up to the burnup requested. Based on this infonnation, PNNL concludes that the SPC BWR rod growth model and analysis methods are acce; table for application to ATRIUM 9 and
- 10 designs up to the bumup requested.
{
3.2 ASGMBLY GROWTH The ATRIUM 9 and 10 designs each have different assembly growth models because, as
. discussed above, the structural skeleton of these two assembly designs is significantly different with the former having tie rods made of cold worked Zircaloy and the latter having a central water channel made of fully annealed Zircaloy, The assembly growth models are used in the rod to tie plate engagement / clearance analyses, the fuel channel to lower tic plate seal spring engagement analysis, and the water channel to upper tie plate engagement. As noted above in regards to the rod to tie plate analysis, the ATRIUM 9 design is concerned with maintaining fuel rod engagement with the upper tie plate while the ATRIUM 10 design is concerned with maintaining clearance between the fuel rod end cap and the upper tie plate. It is also noted that
- either the 95% upper bound or 95% lower bound assembly grr - curves are used in these analyses depending on which gives the most conservative result for the analysis.
3.2 i
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The ATRIUM 9 assembly growth (tic rods) model is based on data only to a rod average burnup of 48 GWd/MTU. This is not of particular concern for the ATRIUM 9 fuel rod to tie plate engagement analysis because the SPC analysis in Supplement 2 demonstrates there is considerable margin left for engagement (greater than 65%) at the burnup requested. The SPC ATRIUM 9 analysis for fuel channel to seal spring engagement has demonstrated that engagement would be maintained, however, there was only an 8% margin of engagement left at the bumup requested.
The purpose ofmaintaining this engagement is to prevent coolant flow from bypassing the assembly into the gap between the fuel channels.
SPC was asked how they could be assured that this engagement could be maintained up to the burnup requested if they did not have assembly growth data up to this burnup and the consequences of bypass flow at extended burnups. SPC explained that the extrapolation to 54 GWd/MTU was relatively small and the rod growth for fuel rods (same Zircaloy fabrication as i
used for the tie rods) did not show any change in growth rate with burnup and, therefore, they could assume that the tie rod growth rate also would not change between 481o 54 GWd/MTU.
SPC also noted that a conservative 95% upper bound cerve is nsed for the assembly growth and a conservative 95% lower bound curve is used for fuel channel growth along with a statistical combination of worst case fabrication tolerances. They further explained that SPC had earlier performed an analysis (Reference 7) assuming the existence ofno seal springs and found that there was no impact on minimum critical power ratio (MCPR) margin at burnups greater than 30 GWd/MTU due to the additional bypass flow. The only other issue would be if bypass flow-were to become great enough to produce flow induced vibrations that would create fretting problems in the assembly, and/or between the control blades and fuel channels. This scenario, however, is not likely with a small amount of bypass flow. The issue of seal spring to channel engagement at extended bumups has been reviewed and deteimined that only a limited amount of bypass flow would result even if the ATRIUM 9 assembly growth were 25% greater than he 95% upper _ bound growth curve currently used by SPC, Therefore, there does not appear to be a strong technical justification for preventing bypass flow at extended burnups. _ PNNL concludes that tne ATRIUM 9 assembly growth model is acceptable for application up to the burnup _
requested based on the earlier SPC analysis (Reference 7) and the fact that only a minimum of bypass flow would result even if the assembly growth rate did increase at the burnup requested.
The ATRIUM 10 assembly growth (water channel) model has used the BWR fuel channel
- growth model because both use fully annealed Zircaloy fabricated in a similar manner. SPC has numerous fuel channel growth data with assembly-average burnups from 10 to 59 GWd/MTU.
SPC has collected assembly axial growth data from ATRIUM 10 lead fuel assemblies with assembly-average burnups between 10 to 35 GWd/MTU that demonstrate that the fuel channel growth _ data are applicable to the ATRIUM 10 assembly. The SPC ATRIUM 10 analyses for rod to tie plate clearances and tie plate seal spring to fuel channel engagement shnw acceptable margins. PNNL concludes that the ATRIUM 10 assembly growth modelis acceptable for application up to the burnup requested.
3.3
l 3.3 FUEL CHANNEL AND WATER CHANNEL GROWTH The ATRIUM 9 ar.d 10 designs both utilize the same fuel channel and fuel channel growth model. The SPC fuel channel growth model and the ATRIUM 10 assembly growth model -~. the same model because both are made from fully annealed Zircaloy. The fuel channel growth model is based on numerous channel growth data that extend to assembly-average burnups of
$9 GWd/MTU presented in Supplement 2. The SPC ATRIUM 9 and 10 analyses of the lower tie plate seal spring to fuel channel engagement conservatively use the 95% lower bound curve of the fuel channal growth model. A further discussion of the tie plate seal spring to fuel channel engagement is provided above.
The ATRIUM 9 design also has a central water channel that is made of the same material as the fuel channel (i.e., fully annealed Zircaloy) and, therefore, SPC also uses the 95% lower bound of the fuel channel growth model for calculating the ATRIUM 9 water channel to upper tie plate engagement. The water channel to upper tie plate engagement for the ATRIUM 9 design shows an acceptable margin. PNNL concludes that the ATRIUM 9 and 10 fuel / water channel model and analysis methods are acceptable for application up to the burnup requested.
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10 CONCLUSIONS PNNL has completed its review of burnup extensions for RODEX2A and the ATRIUM 9 and 10 BWR fuel designs and has reached the following conclusions:
The RODEX2A code described in Reference 1 is acceptable for application to BWR design:; up to rod-average burnups of 62 GWd/MTU.
l The fuel rod, assembly, and fuel channel growth models and analysis methods for the ATRIUM 9 and 10 fuel designs described in Reference 2 are acceptable up to a wmbly-average burnups of 54 GWd/MTU.
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4.1
e 5,0 REFERENCES 1.
Letter, II. D. Curet (SPC) to NRC Document Contro1 Desk.
Subject:
" Request for Resiew of RODE'I2 A (BWR) Fuel Rod Thermal Mechanleal Evalunion Model: Validation.
EMF 85 74(P), Revision 0, Supplement 1,"IIDC:96.048, dated January 15,1997.
2.
Letter, II. D. Curet (SPC) to NRC Document Control Desk,
Subject:
" Request for Review of RODEX2A (BWR) Fuel Rod Thermal. Mechanical Evaluation Model: Fuel Anembly Frnosure Evaluation: EMF 85 74(P), Revision 0, Supplement 2,"llDC:9'i:087, dated August 13,1997.
3.
Merckx, K. R. and S.11. Shann. August 1986. RODEX2A (BWR) - Fuel Rod Thermal-Mechanical Evaluatinn Model. XN NF 85 74 (A), Advanced Nuclear Fuels Corporation (currently the Siemens Power Corporation), Richland, Washington.
4.
Letter,11. D. Curet (SPC) to NRC Document Control Desk,
Subject:
"Resiew of Topical Reports, EMF 85 74(P), Revision 0, Supplement 1, RODEX2A (BWit) Fuel Rod Thermal.
MechanicalEvaluation Model: Validation. and EMF 85 74(P), Revision 0, Supplement 2, RODEX2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model: Fuel Atumbly Exnomure Evaluation (TAC Wo. 98478)," NRC:98:004, dated January 16,1998.
l 5.
Letter, E. Y. Wang (NRC) to 11. D. Curet (SPC),
Subject:
" Request for Additional f
Information (RAI) Related to Topical Report, EMF 85 74 Supplement 1, RODEX2A (BWR) Fuel Rod Thennal Mechanical F.valuallon Model: Validation, (TAC No. 98478),"
dated November 13,1997.
6.
Letter, II. D. Curet (SPC) to NRC Document Control Desk,
Subject:
" Request for Additional Information (RAI) Related to nie Topical Report " EMF 45 74(P), Revision 0, Supplement 1, RODEX2 A (BWR) Fuel Rod Thergl.liechanical Evaluation Model:
Validation. (TAC No. 98478),"liDC:97:141, dated December 3,1997.
7.
Letter,11. D. Curet (SPC) to NRC Document Control De.k, ATTN: Mr. T. E. Collins,
Subject:
"NRC request for Safety Assessment Related to Failed Seal Springs,"
IIDC:97:108, dated October 3,1997.
5.1
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