ML20217B164

Safety Evaluation of Topical Rept BAW-10228P. Science. Rept Acceptable for Licensing Applications,Subject to Listed Conditions in Accordance with Fcf Agreement (Reference 4)

ML20217B164
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Issue date:	10/05/1999
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TAC-MA4599, NUDOCS 9910120168
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NUCLEAR REGULATORY COMMISSION i WASHINGTON D.C. 30se6 4001 ]

SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT BAW-10228P. " SCIENCE' FRAMATOME COGEMA FUELS. INC.  ;

TAC NO. MA4599 l 1 SACKGROUND BAW-10228P describes the SCIENCE code package submitted for review by l Framatome Cogema Fuels, Inc., (FCF) for use in nuclear analysis of pressurized-water j reactor (PWR) cores (Reference 1). SCIENCE is an integrated system of codes specifically designed for performing nuclear analysis of PWRs. The SCIENCE code package consists of core physics tools that are two-dimensional (2D) lattice calculations I and three-dimensional (30) core calculations and data manipulation codes. The

. SCIENCE code package consists of the codes APOLLO 2-F, SMART, and COPILOTE. )

>- The APOLLO 2-F and SMART codes contain the physical description (models) of the SCIENCE code package, while COPILOTE serves as the interface between the user and

' the two physics codes, permitting sequencing and submittal of the calculations through interactive graphicalinterface.

2 TECHNICAL EVALUATION 2.1 Model Description The underlying function of the SCIENCE code package is the linking of the major core  !

physics codes, APOLLO 2-F and SMART. APOLLO 2-F calculates the parameters that are required by the SMART code. These parameters are the cross sections and the

. discontinuity factors, as well as the pin-to-pin reconstruction parameters (Reference 2).

For each type of composition, the parameters generated by the APOLLO 2-F code are placed in data files referred to as " data libraries." The data libraries contain information regarding the dependence of these parameters on feedback system variables, such as -

burnup, xenon, soluble boron, moderator density, fuel temperature, and spectral effects, (Reference 3). These libraries are generated in three steps:

e First APOLLO 2-F performs fuel depletion calculations and stores the data as depletion files. The stored data from the fuel depletion calculations account for the heterogeneity of the assemb!y under normal (reference) conditions and perturbed conditions. The perturbed condition is signified in this case by a change in the water density, thus indicating the spectral differences between actual conditions in the core and the normal reference core depletion.

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• Second, Al'OLLO2-F takes the results of the depletion calculations and uses them to initialize the isotopic concentrations for new calculations in which one or more physical state variables of the assembly are modified with respect to their initial value(s). These calculations are typically referred to as restart calculations.

Some of the variables are xenon level, moderator density, fuel temperature, and a variable that is representative of the control rod presence. All this generated data, representing numerous conditions that the assembly may encounter during the cycle depletion in the core, are placed in " restart" files for use by the SMART code.

e Third, APOLLO 2-F creates two data files for each type of fuel. One file contains cross sections and discontinuity facters and the other file contains reconstruction

. data (pin-by-o.n power distribution and burnep). These files form the libraries that contain parameters that are a function of the state variables, such as boron and

• xenon concentrations, bumup, moderator density, spectral history, fuel temperature profile, and control rod locations.

The fuel assembly is numerically fitted and geometrically represented by dividing the fuel j assembly into cubic regions to account for all the possible variations that go into making up the assembly. These variations are represented by polynomial expansions, utilizing determined polynomial coefficients to reproduce the assembly parameters calculated by the APOLLO 2-F restart calculations at the fitting point.

2.2 Descriotion of Codes  !

APOLLO 2-F is an assembly lattice code developed by the Commissariat a l'Energie  ;

Atomique and modified by FCF for its design needs. It solvss the 99-group transport equation for an assembly geometry and fumishes the homogenized two-group cross sections for the SMART code. The transpod equation is solved using the integral-

- differential equation form that is discretized based on the collision probability method.

FCF pointed out that a coupling of the regrouped multicell calculation and the six-group

. homogeneous calculation permits a good compromise between accuracy and calculation cost.

This coupling is providd by heterogeneous / homogeneous equivalence functions contained in the code. The assembly calculations can be carried out on various geometries (one-eighth of an assembly, one-fourth of an assembly) with different boundary conditions and synimetry. A sophisticated self-shielding model is applied to the cross sections in order to couactly take resonances into account. The flux calculations can be performed with a search for critical buckling to obtain proper spectral weighting. APOLLO 2-F contains a fuel depletion module. The reflector constants (radial or axial) are generated from one-dimensional (1D) APOLLO 2-F calculations using the code's Su option.

The SMART code solves the two-energy group diffusion equation fcr the core geometry under static or kinetic conditions. It solves a neutron balance equation using the average l

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3-flux and plovides the core power distributions for assemblies and also for each pin (on a pin-by-pin basis) in every assembly. The nodal expansion method is used to solve t%

neutron balance equation. It is based on a coupling between a coarse mesh finite difference calculation and a calculation of the neutron current at each interface.

- The calculation of the nodal currents is performed by solving a 1D diffusion equation at each calculation node interface. The solution for the adjustment of the nodal currents is obtained by making assumptions of the fast and thermal neutron flux shapes on both sides of the interface and building a system of equations that can be solved directly.

This solution is referred to as the "two-node" problem. The accuracy of the process is improved by using discontinuity factors, quadratic transverse leakage, and bumup gradients within the node. The bumup and spectral effects are modeled using a microscopic fuel depletion model.

~ The main depletion chains for the heavy nuclei and the main fission products are

. explicitly treated by SMART. The two energy group microscopic cross sections required to calculate isotopic depletion are obtained from data generated ley the APOLLO 2-F code. The raicroscopic cross sections and the isotopic densities resulting from the depletica calculations provide the macroscopic cross sections for the flux solution. The microscopic cross sections are stored in " multi-parameterized" data libraries from which the core calculation interpolates, depsnding on the local node conditions. A set of seven parameters is selected for cross section dependency; bumup, boron concentration, xenon, moderator density, fuel temperature, a spectral history parameter, and a control rod presence parameter. The SMART code calculates fuel pin information for power and burnup and reaction rates in the instrument tube by means of a pin reconstruction algorithm. The SMART code also solves the time-dependent two-energy group diffusion equation for 3D core geometry.

COPILOTE is an operating environment rather than a conventional calculational computer code. . It is the graphical user interface by which the user processes input and output, controls the flow of data from one code to another, and displays the status of the

- calculations.

2.3' Measurement Comoarisons in Section 4 of the submittal, FCF provided numerous examples comparing the results of

- APOLLO 2-F and the SMART code with measurement data. The data were collected

- from six reactors (Three Mile Island Unit 1, Oconee Units 1 and 2, McGuire Unit 1, Gravelines Un't 5, and Sequoyah Unit 1)'. These cores were selected on the basis of obtaining a wide variety of conditions, such as the type of burnable poison, fuel

- enrichment, loading patterns, and control rod patterns. Reactivity predictions versus core burnup, control rod worth, reactivity coefficients, and power distributions were l provided (on a local and global basis) and compared to measured data from operating l PWRs.  !

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4 The agreement between measured data and SCIENCE prediction is generally very good.

The SCIENCE results were just as good or better than prior FCF licensed codes l predictions. The uncertainties for single bank worths and total bank worths that are supported by the data presented are 15 percent and 10 percent, respectively. Previous FCF methodology (Reference 1) contained a bias difference between the 15x15 and the 17x17 bank worth results. This bias difference was again observed in the SCIENCE methodology. The cause of the bias is due to the different sources of the measured results. The 16x15 data were obtained exclusively from B&W plants. Nearly all of the 17x17 data were obtained from other vendors using various measurement techniques, (Reference 4).

, Good agreement was achieved for ejected rod worths, critical boron concentrations, temperature coefficients, and power Doppler coefficients in addition, review of the data shows that SCIENCE accurately predicts the core total peak and radial power peak.

The nuclear reliability factors (NRFs) that SCIENCE uses to adjust the predicted local and global power distribution, were found to be less than those previously established NRFs provided in Reference 1. The staff agree with the presented results.

' 2.4 - Qualification Methods for Future Modifications to the Science Code Packaoe FCF miends'to periodically update the SCIENCE code package to incorporate the latest analytics and computation techniques. Consequently, any code development or improvement of the SCIENCE code package would necessitate benchmarking and validating the SCIENCE code package to ensure that any new feature (s) implemented will produce results in keeping with a standard set of qualification criteria as stated in the submitted Topical Report BAW-10228P.

The method to be used to qualify SCIENCE for future changes is similar to the method presented in this submittal and previous topical report submittals. This method will require that neutron code qualification be based on the ability of the modified SCIENCE code package to predict several key neutronic parameters. Some of these parameters are critical boron or k-effective at hot zero power, critical boron or critical k-effective at hot full power,' individual bank rod worths, total rod worths, ejected rod worths, isothermal temperature coefficients, power Doppler coefficients, hot pin power, and hot pellet power (see Table 5.1 of References 1 and 3). These parameters will be recalculated with the modified SCIENCE code packaDe and compared to measured results and new statistics generated along with their associated uncertainties. Subjecting the modified code package to the listed criteria will emphasis the contributions of the implemented features to the code package rather than highlight the differences between the two code packages. ' Consequently, any modifications to the SCIENCE code package that meet the listed criteria in Tables 5-1 and 5-2 of this submittal will validate the modifications made to the SCIENCE code package. if the changes to the SCIENCE code package meet the criteria, FCF will internally document the changes to the code package and the associated results without notifying the NRC since there were no changes to the uncertainties or their application. However, if changes to the uncertainties occur, FCF L

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l will' submit supporting documentation to the NRC whenever the method changes affect the uncertainties to be applied in licensing applications. I The methodology and the data provided in this submittal, Topical Report BAW-10228P, )

form the basis for the current SCIENCE code package. Future application of the current l SCIENCE code package to data not provided in this topical report (such as new fuel j designs) will require revalidation of the SCIENCE code package. j 2.5 Ranae of Aoolicability of Benchmarkino -

.I The chosen benchmarks in this topical report include the types of fuel and poison that are typically licensed. The data presented in this topical report are sufficient to qualify i the SCIENCE code package for the typical fuel types listed in this submittal. If a new fuel design is used that contains materials (poison, mixed oxide fuel) and/or hafnium contro) i rods, etc.) outside this collection of benchmarks, additional benchmarks will have to be .

established. In accordance with its agreement, (Reference 4), FCF will submit to the  ;

NRC staff a description of the new design feature, the new benchmarks, and any impact on the current uncertainty factors. i 3 CONCLUSION The staff has reviewed the analyses in Topical Report BAW-10228P, " SCIENCE," and find it acceptable for licensing applications, subject to the following conditions in accordance with FCF's agreement (Reference 4) :

1. The SCIENCE code package shall be applied in such a manner that predicted results are wRhin the ranges of the validation criteria presented in Table 5-1 and the measurement uncertainties presented in Table 5-2.

2. Fuel designs to which the SCIENCE code package will be applied shall be within the validation bases of BAW-10228P. The bases of BAW-10228P are considered valid for the following conditions:

15x15 or 17x17 UO2 fuel designs.

. U235 enrichments less than or equal to a maximum of 5.0 w/o.

. Gadolinia loadings less than or equal to 8.3 w/o (nominal 8.0 w/o).

3. The following uncertainties shall be applied to the SCIENCE code package results:

. Maximum pin peaking uncertainty of 3.8 percent.

. Maximum pellet peaking uncertainty of 4.8 percent.

. Total rod worth uncertainty of 100 percent.

. Bank rod worth uncertainty of 15 percent.

4. The SCIENCE code package shall only be used for PWR licer.;;ing analyses by FCF.unless approved by the NRC for use by other organization.

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6-4 REFERENCES i . 1. Letter from T. A. Coleman,'Vice President of Government Relations, Framatome Cogema Fuels, Inc., to the NRC, regarding the submittal of the SCIENCE code i package, dated October 12,1998.

2.

R. Sanchez, J. Mondot, Z. Stankcwski, A. Cossic, and I. Zmijarevic,

" APOLLO 2-F: A User-oriented, Portable Modular Code for Multi-group Transport Assembly Calculations." International Topical Meeting on Advances in Reactor l Physics, Mathematics and Computation, Paris, France. April 27-30,1987. l i

3. R. Sanchez and M.Vergain, "An Acceleration Procedure for the Iterative So'ution l

of the Flux Current Equations in the APOLLO 2-F Code." International Topical  ;

Meeting on the Physics of Reactors: Operation and Design Computation.

Marseilles, France. PHYSOR 90, April 23-27,1990.  ;

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4. Letter from T. A. Coleman, Vice President of Govemment Relations, Framatome l Cogema Fuels, Inc., to the NRC, regarding the submittal of Topical Report l BAW-10228P, " SCIENCE", dated September 23,1999. 1 P

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1.0 INTRODUCTION

Framatome Cogema Fuels (FCF) has submitted to the NRC a topical report, entitled

" Mark-Bil Fuel Assembly Design Topical Report" BAW-10229P (Reference 1), for review and approval. Presented in Reference 1 is the informatio'n required to support the licensing basis for the implementation of the Mark-Bil fuel assembly as reload fuel in Babcock and Wilcox 6 pressurized water reactors (PWRs). This Technical Evaluation Report (TER) will address whether this new fuel design meets the NRC approved FCF fuel design criteria (Reference 2) and that the FCF analysis methodology used for this design applies to the Mark-Bil design up to the NRC approved rod average bumup level of 62 GWd/MTU (Reference 3).

It should be explained that Framatome Cogema Fuels was previously named the B&W Fuel Company (BWFC) a part of B&W Nuclear Technologies and prior to BWFC was named Babcock & Wilcox (B&W). Some of the references in this TER refer to these different company names depending on the date the reference was generated.

Pacific Northwest National Laboratory (PNNL) has acted as a consultant to the NRC in this review. As a result of the NRC staffs and their PNNL consultant's review of the topical report, a request for additional information (RAI) was sent by the NRC to FCF (Reference 5) re-questing clarification of the design changes,' lead test assembly data, the applicability of FCF evaluation methodology, and results oflicensing analyses for the Mark-B11 design. FCF responded to those questions in Reference 6. FCF was further questioned for clarification of their responses in a January 26,1999, conference call with NRC and PNNL. This conference call clarified their responses.

This review was based on those licensing requirements identified in Section 4.2 of the Standard Review Plan (SRP) (Reference 7) and the FCF approved fuel design criteria (Reference 2). The objectives of this fuel system safety review, as described in Section 4.2 of the SRP, are to provide assurance that 1) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs),2) fuel system damage is never so severe as to prevent control rod insenion when it is required, 3) the number of fuel rod failures is not underestimated for postulated accidents, and 4) coolability is always maintained. A "not damaged" fuel system is defined as fuel rods that do not fail, fuel system dimensions that remain within operational tolerances, and functional capabilities that are not reduced below those assumed in the safety analysis. Objective 1, above, is consistent with General Design Criterion (GDC) 10 [10 Code of Federal Regulations (CFR) 50, Appendix A] (Reference 8), and the design limits thauccomplish this are called specified acceptable fuel design limits (SAFDLs).

" Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR 100 (Reference 9) for postulated accidents. "Coolable geometry,"

means in general, that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channels to permit removal of residual heat for design basis accidents. The general requirements to maintain control rod insertability and core coolability appear repeatedly 1.1

l 2.0 FUEL SYSTEM DESIGN The Mark-Bil fuel assembly consists of a 15x15 square array of fuel rods, control rod guide tubes, and a, central instrumentation tube. The control rod guide tubes, central instrumentation tube, and eight spacer grids are mechanically fastened together with the top and .

bottom nozzles that make up the structural cage for the fuel rod assemblies.' Fuel rods are I supported at intervals along their length by the spacer grids with grid springs and dimples con-tained within the spacer grids to maintain rod-to-rod spacing. The spacer grid consists of an egg-(. crate arrangement ofinterlocking straps that contain springs and dimples that hold the fuel rods

' ~ in place. The top nozzle is designed to allow for fuel assembly reconstitution, the same as for the

. Mark-BIO assembly. Attached to the top nozzle are holddown springs and spring clamps'which )

keep the fuel assembly firmly seated on the lower core plate during normal plant operation. 1 The main differences between the Mark-Bil design and the Marl -BIO design is in the

, smaller diameter fuel rods, the use of flow mixing vanes on five of the six intermediate Zircaloy  ;

grids, and an improved grid restraint system on the central instrument tube. Due to the smaller i diameter fuel rods the spacer grid cell size was reduced proportionately in the spacer grids in order to maintain the same spacer spring loads. All but the bottom intermediate spacer grids (five out of six) have the bent out vanes on the top of the grid interior strips. These vanes provide improved thermal performance by locally increasing the intensity of flow turbulence in the subchannel. Mixinrc mes are not used on the lower intermediate grid since they are not needed in this cooler ax& gion of the assembly. A similar mixing vane grid is used in the Mark-B11 design for Westinghouse plants.

Due to the mixins vacs creating greater flow resistance in the uppermost intermediate i grids there are greater loads pln:ed on the grid restraint system. As a result the grid restraint system was redesigned to 1) increase the load-carrying capacity of the restraint system, and 2) to divide the loads between those from the lowest two intermediate spacer grids and those from the four uppermost intermediate spacer grids. The latter change reduces the loads on the uppermost sleeves that carry the increased loads due to the mixing vanes.

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approved by NRC (Reference 2). This strain criterion is also acceptable for application to the '

. Mark-Bil design up to the current Mark B operating burnup limit of 62 GWd/MTU (rod-average).

He niaterial property that could have a significant impact on the cladding strain limit at extended burnup levels is cladding ductility. The strain criterion could be impacted if cladding ductility were decreased, as a result of extended bumup operation, to levels that would allow cladding failure without the 1% cladding strain criteria being exceeded under normal operation and AOOs. Recent out-of-reactor measured clastic and plastic cladding strain values from high burnup cladding from two PWR fuel vendors (References 11,12 and 13) have shown a decrease in cladding ductilities when local burnups exceed 52 GWd/MTU and with increasing hydrogen -

- (corrosion) levels. In addition, the majority of the high burnup data (tensile or burst test) shows that when hydrogen levels start to exceed 700 ppm the uniform strains begin to fall below 1%.

As a result FCF has adopted a limit on maximum cladding corrosion that is consistent with maintaining cladding hydrogen levels below 700 ppm, and that has been approved by NRC (Reference 3). This is also found to be applicable to the Mark-Bil fuel design up to the current Mark-B opera' ting burnup limit of 62 GWd/MTU (rod-average).

Evaluation - The FCF strain analysis methods for Mark-B designs have been approved for application to Mark-B designs (Reference 2 and up to rod-average burnups of 62 GWd/MTU (Reference 3). FCF has performed bounding fuel rod cladding strain analyses using these rre. hods that determined that the Mark-Bil design meets the above strain criterion within the design operating limits. PNNL concludes that FCF strain analysis methods are applicable to the Mark-B11 design and that the design is acceptable with respect to cladding strain up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

3.3 STRAIN FATIGUE Bases / Criteria - The FCF design criterion for cladding strain fatigue is that the cumulative fatigue factor be less than 0.9 when a minimum safety factor of 2 on the stress amplitude or a minimum safety factor of 20 on the number of cycles, which ever is the most conservative, is imposed as per the O'Donnell and Langer design curve (Reference 14) for fatigue usage. This criterion is consistent with that described in Section 4.2 of the SRP and has previously been

. approved (References 2 and 3). This strain fatigue criterion is also acceptable for application to the Mark-B11 design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod- l average).

Evaluation - The FCF strain fatigue analysis methods for Mark-B designs have been approved for application to rod-average burnups of 62 GWd/MTU (References 2 and 3). FCF has performed bounding fuel rod cladding strain fatigue analyses using these methods that determined that the Mark-B11 design meets the above strain fatigue criterion within the design's operating limits. PNNL concludes that FCF strain fatigue analysis methods are applicable to the 3.2

s 3.5 OXIDATION AND CRUD BUII DUP Bases / Criteria - Section 4.2 of the SRP identifies cladding oxidation and crud buildup as

_ potential fuel system damage mechanisms. The SRP does not establish specific limits on

cladding oxidation and crud buildup but does specify that their effects be accounted for in the thermal and mechanical analyses performed for the fuel. As noted in Section 3.2, the cladding ductility can be significantly decreased at higher bumup levels where oxide thickness and hydrogen levels can become relatively large because of accelerated corrosion at rod-average

, burnups above 50 to 55 GWd/MTU. As a result FCF has adopted a limit of 100 microns on l maximum cladding corrosion that is consistent with maintaining cladding hydrogen levels below j 700 ppm and has been previously approved (Reference 3). This maximum corrosion limit is l based on a localized axial position on a fuel rod. PNNL concludes that this maximum corrosion

!- limit is applicable to and acceptable for application to the Mark-B11 design up to the current l

Mark B operating burnup limit of 62 GWd/MTU (rod-nverage).

Evaluation - Section 4.2 of the SRP states that the effects of cladding crud and oxidation l needs to be addressed in safety and design analyses, such as in the thermal and mechanical l analysis. The amount of cladding oxidation is dependent on the cladding type, fuel rod powers, l water chemistry control and primary inlet coolant temperatures, but the amount of oxidation and l

crud buildup increases with burnup and cannot be eliminated. Therefore, extended burnups result in a thicker oxide layer that provides an extra thermal barrier, cladding thinning and

ductility decrease that can affect the mechanical performance. The degree of this effect is l dependent on cladding' type, reactor coolant temperatures, power history, and the level of success l of a reactors' water chemistry program. The following is an evaluation of the FCF corrosion i

model.

l FCF has adopted a new cladding corrosion model, COROSO2 (Reference 3), that is more l conservative, i.e.,' predicts more corrosion, than the original OXIDEPC model in TACO 3 and i predicts the accelerated corrosion observed in high burnup rods much better than the OXIDEPC l model. This model has been approved by NRC with the commitment by FCF to collect more l maximum corrosion thickness data in the future (Reference 3). The Mark-Bil and the similarly l designed Mark-BW LTAs will also provide corrosion data up to extended bumup levels (see L . Section 6.0 on Fuel Surveillance) to verify the applicability of the new conosion model to the l Mark-B11 design. The best estimate or slightly conservative prediction of the COROSO2 model l is considered to be acceptable because of the conservatism in the FCF maximum corrosion limit.

Based on FCFs commitment to collect corrosion data at extended bumup levels from their Mark-B and Mark-BW LTAs, PNNL concludes that the COROSO2 model is acceptable for application to the Mark-B11 design in predicting maximum corrosion levels up to the current Mark B operating bumup limit of 62 GWd/MTU (rod-average).

It is noted that FCF performs reload / cycle specific evaluations to verify that cladding corrosion is within their design limit. These cycle specific evaluations are not within the scope of this review.

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- the margin for compressing the cruciform holddown springs to solid height up to a rod-average burnup of 62 GWd/MTU.

The FCF response (Reference 6) presented one cycle data from the Mark-B11 LTAs that indicated that the Mark-B11 shoulder gap and assembly growth data were within the scatter of the earlier Mark-B and Mark-BW data. FCF also provided the margins requested showing that both the margins for shoulder gap closure and solid compression of the holddown springs were relatively small up to a rod-average burnup of 62 GWd/MTU and 64 GWd/MTU, respectively.

However, examination of the FCF analysis methods used for predicting shoulder gap clearances and assembly growth demonstrate that they are very conservative. For example, the FCF

. bounding curves used for both of these analyses are significantly greater than the 95/95 bounds of the data. Berefore, the actual margins to the design bases for axial growth are quite large. In addition, FCF intends to collect axial growth and shoulder gap clearance data from the Mark-B11 LTAs. PNNL concludes that these axial growth analysis methods are conservative. Therefore, PNNL further concludes that they are acceptable for application to the Mark-Bil design and that the design is acceptable with respect to axial growth up to the current Mark-B operating burnup

- limit of 62 GWd/MTU (rod-average).

3.8 ROD INTERNAL PRFRSURE Bases / Criteria . Rod internal pressure is a driving force for, rather than a direct mechanism of, fuel system damage that could contribute to the loss of dimensional stability and cladding integrity. Section 4.2 of the SRP presents a rod pressure limit of maintaining rod pressures below system pressure that is sufficient to preclude fuel damage. The FCF design basis for the fuel rod internal pressure is that the fuel system will not be damaged due to excessive fuel rod internal pressure and FCF has established the " Fuel Rod Pressure Criterion" (Reference 15) to provide assurance that this design basis is met. These criteria are that the internal pressure of the FCF lead fuel rod in the reactor is limited to a value below which could cause 1) the diametral gap to increase due to outward cladding creep during steady-state operation, and 2) extensive DNB propagation to occur. This FCF design basis and the associated criteria have been found acceptable by the NRC (Reference 15) up to the current Mark-B burnup limits established in l

Reference 3. PNNL concludes these are also acceptable for application to the Mark-Bil design up to the current Mark B operating burnup limit of 62 GWd/MTU (rod average). 1 Evaluation - FCF utilizes the approved TACO 3 fuel performance code (Reference 16) for predicting end-of-life (EOL) fuel rod pressures and the methodology described in Reference 15  !

to verify that they do not exceed the FCF " Fuel Rod Pressure Criterion" during normal operation and AOOs. The TACO 3 fuel performance code is generic enough to be applicable to all FCF l

, . PWR fuel designs, and therefore is acceptable for application to the Mark-Bil design up to the l current Mark B operating burnup limit of 62 GWd/MTU (rod-average). The issue of DNB propagation (Fuel Rod Pressure Criterion 2 above) will be discussed in Section 4.3. The FCF tod pressure analyses are performed on a reload / cycle specific basis, i

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l 4.0 FUEL ROD FAILURE l

l In the following paragraphs, fuel rod failure thresholds and analysis methods for the failure mechanisms listed in the SRP will be reviewed. When the failure thresholds are applied to normal operation including AOOs, they are used as limits (and hence SAFDLs) since fuel failure under those conditions should not occur according to the traditional conservative interpretation of

' GDC 10. When these thresholds are used for postulated accidents, fuel failures are permitted, but they must be accounted for in the dose assessments required by 10 CFR 100. The basis or reason for establishing these failure thresholds is thus established by GDC 10 and Part 100 and only the threshold values and the analysis methods used to assure that they are met are reviewed below.

4.1 HYDRIDING Bases / Criteria - Internal hydriding as a cladding failure mechanism is precluded by controlling the level of hydrogen impurities in the fuel during fabrication; this is generally an early-in-life failure mechanism. FCF has not discussed their criteria for internal hydriding in the subject topical report; however, a limit on hydrogen level for FCF pellets is discussed in  ;

Reference 17. The hydrogen level of FCF fuel pellets is controlled by drying the pellets in the cladding and taking a statistical sample to ensure that the hydrogen level is below a specified level. Previous FCF design reviews, e.g., Reference 17, have shown that this level is below the value recommended in the SRP. Consequently, PNNL concludes that the FCF limit on hydrogen j in their fuel pellets is acceptable for the Mark-B11 design.  !

External hydriding of the cladding due to waterside corrosion is the other source and is discussed in Section 3.5 of this TER. As noted in Section 3.5, the icvel of external hydriding is controlled by FCF by a proprietary limit on corrosion thickness. PNNL concludes that this  !

corroshn limit is acceptable for limiting the level of external hydriding in the cladding for the Mark-B11 design up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

Evaluation - Internal hydriding is controlled by FCF by taking statistical samples following l pellet fabrication prior to loading the pellets in the fuel rods and confirming that hydrogen is below a specified level. Therefore, no analyses are necessary other than to conf'um that the statistical pellet sampling is below the specified level for Mark-B11 designs.

1 Extemal hydriding is controlled by the FCF limit on corrosion thickness discussed in Section 3.5 of this TER.

PNNL concludes that FCF has addressed the issue of hydriding in Mark-B11 designs up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

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4.4 . OVERHEATING OF FUEL PET T ETS l

Bases / Criteria - As a second method of avoiding cladding failure due to overheating, FCF precludes centerline pellet melting during nonnal operation and AOOs. This design criterion is the same as that given in the SRP and has previously been approved for FCF designs up to

, current operating limits (References 2 and 3). 'Ihis criterion for fuel melting is also acceptable for application to the Mark-B11 design up to the current Mark-B operating bumup limit of 62 ,

GWd/MTU (rod-average).

Evaluation - FCF utilizes the approved TACO-3 fuel performance code to determine the maximum linear heat generation rate (LHGR) at which a given fuel design will not achieve fuel melting at a 95% probability at a 95% confidence level. This FCF analysis methodology has been found to be acceptable to Mark-B designs up (Reference 2) to a rod everage bumup of '  !

62 GWd/MTU (Reference 3). PNNL also finds them acceptable for application to the Mark-B11 )

design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

- FCF has also performed a fuel melting analysis for the Mark-Bil fuel design that ,

_ demonstrates that the Mark-Bil design is acceptable within the design's operating limits. PNNL )

concludes that the Mark-B11 design is acceptable in relation to fuel melting up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

l 4.5 PELLET /CLADDINGINTERACTION ,

Bases / Criteria - As indicated in Section 4.2 of the SRP, there are no generally applicable criteria for pellet cladding interaction (PCI) failure. However, two acceptance criteria oflimited application are presented in the SRP for PCI: 1) less than 1% transient induced cladding strain,

'and 2) no centerline fuel melting. Both of these limits have been adopted by FCF for use in evaluating their fuel design: and have been approved by the NRC. These two criteria have been satisfactorily addressed in Sections 3.2 and 4.4 of this TER and will not be discussed further in this section.

Evaluation - As noted earlier, FCF utilizes the TACO-3 (Reference 16) code to show that their fuel meets both the cladding strain and fuel melting criteria. This code is acceptable per the recommendations in Sections 3.2 and 4.4.

I 4.6 CLADDING RUPTURE Bases /Criieria - There are no specific design limits associated with cladding rupture other j than the 10 CFR 50, Appendix K (Reference 19) requirements that the incidence of rupture not be underestimated. FCF uses a rupture temperature correlation consistent with NUREG-0630 l l guidance (Reference 20). PNNL concludes that FCF has adequately addressed cladding rupture ]

for the Mark-B11 design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod- j average). 4

- l 4.3 l

l u  ;

n;-

( l.0 INTRODUCTION

(

Framatome Cogema Fuels (FCF) has submitted to the NRC a topical report, entitled

" Mark-B11 Fuel Assembly Design Topical Report" BAW-10229P (Reference 1), for review and

, approval. Presented in Reference 1 is the information required.to support the licensing basis for the implementation of the Mark-Bil fuel assembly as reload fuel in Babcock and Wilcox

! . pressurized water reactors (PWRs). This Technical Evaluation Report (TER) will address whether this new fuel design meets the NRC approved FCF fuel design criteria (Reference 2) and that the FCF analysis methodology used for this design applies to the Mark-B11 design up to the

' NRC approved rod average bumup level of 62 GWd/MTU (Reference 3).

It should be explaineNat Framatome Cogema Fuels was previously named the B&W Fuel Company (BWFC) a part of B&W Nuclear Technologies and prior to BWFC was named Babcock & Wilcox (B&W). Some of the references in this TER refer to these different company names depending on the date the reference was generated.

Pacific Northwest National Laboratory (PNNL) has acted as a consultant to the NRC in this review. As a result of the NRC staff's and their PNNL consultant's review of the topical report, a request for additional information (RAI) was sent by the NRC to FCF (Reference 5) re-questing clarification of the design changes,' lead test assembly data, the applicability of FCF evaluation methodology, and results oflicensing analyses for the Mark-Bil design. FCF '

responded to those questions in Reference 6. FCF was further questioned for clarification of their responses in a January 26,1999, conference call with NRC and PNNL. This conference call clarified their responses.

This review was based on those licensing requirements identified in Section 4.2 of the Standard Review Plan (SRP) (Reference 7) and the FCF approved fuel design criteria (Reference 2). The objectives of this fuel system safety review, as described in Section 4.2 of the SRP, are to provide assurance that 1) the fuel system is not damaged as a result of normal operation z.ad anticipated operational occurrences (AOOs),2) fuel system damage is never so severe as to prevent control rod insertion when it is required, 3) the number of fuel rod failures is not underestimated for postulated accidents, and 4) coolability is always maintained. A "not damaged" fuel system is defined as fuel rods that do not fail, fuel system dimensions that remain within operational tolerances, and functional capabilities that are not reduced below those assumed in the safety analysis. Objective 1, above, is consistent with General Design Criterion (GDC) 10 [10 Code of Federal Regulations (CFR) 50, Appendix A] (Reference 8), and the design limits that accomplish this are called specified acceptable fuel design limits (SAFDLs). i

" Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the l cladding) has, therefore, been breached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR 100 (Reference 9) for postulated accidents. "Coolable geometry,"

means in general, that the fuel assembly retains its rod-bundle geometrical configuration with  !

l adequate coolant channels to permit removal of residual heat for design basis accidents. The 1 j general requirements to maintain control rod insertability and core coolability appear repeatedly 1.1 J

. _--.-.,-~.._n,, , . , -. . .

. 1 i

in the GDC (e.g., GDC 27 and 35). Specific coolability requirements for the loss-of-coolant accident (LOCA) are given in 10 CFR 50, Section 50.46.

In ' order to assure that the above stated objectives are met and follow the format of  !

Section 4.2 of the SRP, this review covers the following three major categories: 1) Fuel System i i

Damage Mechanisms, which are most applicable to normal operation and AOOs; 2) Fuel Rod Failure Mechanisms, which apply to normal operation, AOOs, and postulated accidents; and

3) Fuel Coolability, which is applied to postulated accidents. Specific fuel damage or failure criteria are identified under each of these categories in Section 4.2 of the SRP. The FCF fuel design criteria or SAFDLs and the applicability of FCF analysis methodologies to the Mark-Bil design are discussed in this TER under each fuel damage or failure mechanism listed in the SRP.

Th: purpose. of the design bases and/or. criteria is to provide limiting values that prevent

' fuel damage'or failure with respect to each mechanism. Reviewed in this TER is the applicability of the Mark-Bil design submitted in BAW-10229P to the FCF fuel design criteria and the applicability of FCF analysis methodologies to the Mark-Bil design are discussed. The FCF design criteria, along with certain definitions for fuel failure, constitute the SAFDLs required by GDC 10. The FCF analysis methods assure that the design limits and, thus, SAFDLs are met for a particular design application.

A description of a Mark-Bil fuel assembly is briefly discussed in the following section (Section 2.0). The fuel damage and failure mechanisms are addressed in Sections 3.0 and 4.0, respectively, while fuel coolability is addressed in Section 5.0.

e 4

l 1.2

2.0 FUEL SYSTEM DESIGN l

' The' Mark-B11 fuel assembly consists of a 15x15 square array of fuel rods, control rod i guide tubes, and a central instrumentation tube. The control rod guide tubes, central  !

instrumentation tube, and eight spacer grids are mechanically fastened together with the top and )

bottorn nozzles that make up the structural cage for the fuel rod assemblies. Fuel rods are j

' supported at intervals along their length by the spacer grids with grid springs and dimples con- l

- tained within the spacer grids to maintain rod-to-rod spacing. The spacer grid consists of an egg-

. crate arrangement ofinterlocking straps that contain springs and dimples that hold the fuel rods

' in place. The top nozzle is designed to allow for fuel assembly reconstitution, the same as for the

- Mark-BIO assembly. Attached to the top nozzle are holddown springs and spring clamps which keep ti.e fuel assembly firmly seated on the lower core plate during normal plant operation.

i The main differences between the Mark-B11 design and the Mark-BIO design is m the  ;

, smaller diameter fuel rods, the use of flow mixing vanes on five of the six intermediate Zircaloy grids, and ari improved grid restraint system on the central instrument tube. Due to the smaller 1 diameter fuel rods the spacer grid cell size was reduced proportionately in the spacer grids in i order to maintain the same spacer spring loads. All but the bottom intermediate spacer grids (five out of six) have the bent out vanes on the top of the grid interior strips. These vanes

. provide improved thermal performance by locally increasing the intensity of flow turbulence in the subchannel. Mixing vanes are net used on the lower intermediate grid since they are not needed in this cooler axial region of the assembly. A similar mixing vane grid is used in the

- Mark-B11 design for Westinghouse plants.

.Due to the mixing vanes creating greater flow resistance in the uppermost intermediate grids there are greater loads placed on the grid restraint system. As a result the grid restraint system was redesigned to 1) increase the load-carrying capacity of the restraint system, and 2) to divide the loads between those from the lowest two intermediate spacer grids and those from the

four uppermost intermediate spacer grids. The latter change reduces the loads on the uppermost

' sleeves that carry the increased loads due to the mixing vanes.

2.1

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. _. - . . . _. _. ._.___.....-_.rm.m.

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n 3.0 FUEL SYSTEM DAMAGE The design criteria presented in this section should not be exceeded during normal operation including AOOs. The evaluation portion of each damage mechanism evaluates the

- analysis methods used by FCF to demonstrate that the design criteria are not exceeded during normal operation including AOOs for the reconstituted fuel assembly design.

3.1 STRESS Bases / Criteria - In keeping with the GDC 10 SAFDLs, fuel damage criteria for cladding

. stress should ensure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analysis. The FCF design. basis for fuel rod cladding stresses is that the fuel system will be functional and will not j be damaged due to excessive stresses. The FCF criteria are based on guidelines established in - i

- Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel ,

Code (Reference 10). These criteria are consistent with the acceptance criteria established in L Section 4.2 of the SRP and have been previously approved by NRC for Mark-B designs )

(Reference 2).- These stress criteria are also acceptable for application to the Mark-B11 design up )

to the current Mark B operating burnup limit of 62 GWd/MTU (rod-average).

Evaluation - The stress analyses for the Mark-Bil fuel assembly components and fuel rod cladding are based on standard engineering stress analysis methods including finite-element analysis and calculated in accordance with the ASME code, which includes both normal and shear stress effects. Pressure and temperature inputs to the stress analyses are chosen so that the operating conditions for all normal operation and AOOs are enveloped. The input cladding wall .

thicknesses.are reduced to those minimum values allowed by fabrication specifications and further reduced by a conservative amount to allow for corrosion on the cladding inside and outside surfaces. These stress analysis methods have been approved for Mark B designs

. (Reference 2). PNNL concludes that the Mark-B stress analysis methods are acceptable for

. application to the Mark-Bil design up to the current Mark B operating btmup limit of 62 GWd/MTU (rod average).

FCF has performed bounding stress analyses using these methods that determined that the Mark-B11 design components, including the fuel rods, meet the approved FCF stress criteria.

Therefore, PNNL further concludes that the Mark-Bil design is acceptable with respect to design stress analysis.

3.2 ~ STRAIN -

Bases / Criteria - The FCF design criterion for fuel rod cladding strain is that maximum uniform hoop strain (clastic plus plastic) shall not exceed 1%. This criterion is intended to  !

preclude excessive cladding deformation from normal operation and AOOs. This is the same i criterion for cladding strain that is used in Section 4.2 of the SRP and has been previously l

3.1 l

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, )

I approved by NRC (Reference 2). His strain criterion is also acceptable for application to the j

, Mark-B11 design up to the current Mark B operating bumup limit of 62 GWd/MTU (rod-

average).

. The niaterial property that could have a significant impact on the cladding strain limit at extended burnup levels is cladding ductility. The strain criterion could be impacted if cladding ductility were decreased, as a result of extended burnup operation, to levels that would allow cladding fcilure without the 1% cladding strain criteria being exceeded under normal operation and AOOs. Recent out-of-reactor measured clastic and plastic cladding strain values from high 1 bumup cladding from two PWR fuel vendors (References 11,12 and 13) have shovm a decrease l

l in cladding ductilities when local bumups exceed 52 GWd/MTU and with increasing hydrogen l (corrosion) levels. In addition, the majority of the high burnup data (tensile or burst test) shows

that when hydrogen levels start to exceed 700 ppm the uniform strains begin to fall below 1%.

l As a result FCF has adopted a limit on maximum cladding corrosion that is consistent with maintaining cladding hydrogen levels below 700 ppm, and that has been approved by NRC -

- (Reference 3). This is also found to be applicable to the Mark-B11 fuel design up to the current l Mark-B opera' ting burnup limit of 62 GWd/MTU (rod-average).

l ,

Evaluation - The FCF strain analysis methods for Mark-B designs have been approved for application to Mark-B designs (Reference 2 and up to rod-average burnups of 62 GWd/MTU ,

(Reference 3). FCF has performed bounding fuel rod cladding strain analyses using these l methods that determined that the Mark-Bil design meets the above strain criterion within the i design operating limits. PNNL concludes that FCF strain analysis methods are applicable to the Mark-Bil design'and that the design is acceptable with respect to cladding strain up to the )

current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).  ;

3.3 STRAIN FATIGUE Bases / Criteria - ne FCF design criterion for cladding strain fatigue is that the cumulative fatigue factor be less than 0.9 when a minimum safety factor of 2 on the stress amplitude or a minimum safety factor of 20 on the number of cycles, which ever is the most conservative, is imposed as per the O'Donnell and Langer design curve (Reference 14) for fatigue usage. This criterion is consistent with that described in Section 4.2 of the SRP and has previously been

. approved (References 2 and 3). This strain fatigue criterion is also acceptable for application to the Mark-Bil design up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

Evaluation - The FCF strain fatigue analysis methods for Mark-B designs have been l approved for application to rod-average burnups of 62 GWd/MTU (References 2 and 3). FCF

- has performed bounding fuel rod cladding strain fatigue analyses using these methods that determined that the Mark-B11 design meets the above strain fatigue criterion within the design's operating limits. PNNL concludes that FCF strain fatigue analysis methods are applicable to the L 3.2 p  %

l

L .

Mark-B11 design and that the design is acceptable with respect to cladding strain fatigue up to s the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

3.4 FRETTING WEAR Bases / Criteria - Fretting wear is a concern for fuel, burnable poison rods, and guide tubes.

Fretting, or. wear, may occur on the fuel and/or burnable poison cladding surfaces in contact with the spacer grids if there is a gap between the grid spacer springs and the fuel rods or due to flow induced vibratory forces. The FCF design criterion for fretting wear is that the assembly design shall provide sufficient support to limit rod vibration and fretting wear. This criterion is l consistent with Section 4.2 of the SRP and has previously been approved for Mark-B designs up l to rod-average burnups of 62 GWd/MTU (References 2 and 3). This fretting wear criterion is I also acceptable for application to the Mark-B11 design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

Evahmtion - FCF has performed extensive flow-induced vibration testing of the Mark-B11 fuel assembly to examine the vibrational response and to verify that no flow related vibrational phenomena existed that could result in fretting wear. The vibrational response of the Mark-Bil was compared to the vibrational response of the proven in-reactor performance of the Mark-B10 assembly. The comparisons were performed under a wide range of flow conditions that could be experienced in-reactor with both assembly types having comparable vibrational responses and verylow amplitudes ofvibration. .

FCF has also performed a 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> wear test of the Mark-B11 assembly at simulated full power operating conditions of temperature, pretJ. e, flow and coolant chemistry. The grid springs of the spacer grids in this assembly were relaxed to simulate end-of-life conditions between the springs and fuel rods. 'Ibe results of this test showed that the wear between the grid springs anJ fuel rods was less than those of previous Mark-B designs for the same test conditions. FCF has also pointed out that they have not seen any evidence of fretting wear in Mark-B11 lead test assemblies (LTAs) after one cycle of operation.

FCF was questioned (Reference 5) on the cross flow conditions of a mixed core with the Mark-B11 assemblies and whether these cross flows could result in sufficient forces to induce fuel rod vibration. FCF responded (Reference 6) that they had used the LYNXT model to investigate cross flow velocities in a mixed core and found that the maximum cross flow velocities were significantly less than those experienced at the core periphery for Mark-B cores with similar pressure drop characteristics. These results suggest that cross flow velocities between different Mark-B assemblies will not result in fretting wear.

Based on the above testing and analyses, PNNL concludes that the Mark-B11 design is acceptable with reispect to fretting wear up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

A 3.3 rn- -.w-.p = ., , g _ _ _ . . , - ya _- *. - g - e , +e 7.~.-..-----.-.4 -

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3.5 OXIDATION AND CRUD BUILDUP Bases / Criteria - Section 4.2 of the SRP identifies cladding oxidation and crud buildup as potential fuel system damage mechanisms. The SRP does not establish specific limits on cladding oxidation and crud buildup but does specify that their effects be accounted for in the thermal and mechanical analyses perr ormed for the fuel. As noted in Section 3.2, the cladding )

l - ductility can be significantly_ decreased at higher burnup levels where oxide thickness and ,

i hydrogen levels can become relatively large because of accelerated corrosion at rod-average  !

l ,

burnups above 50 to 55 GWd/MTU. As a result FCF has adopted a limit of 100 microns on l maximum cladding corrosion that is consistent with maintaining cladding hydrogen levels below l 700 ppm and has been previously approved (Reference 3).- This maximum corrosion limit is l - based on a localized axial position on a fuel rod. PNNL concludes that this maximum corrosion l- limit is applicable to and acceptable for application to the Mark-B11 design up to the current Mark B operating bumup limit of 62 GWd/MTU (rod-average). ,

l Evaluation - Section 4.2 of the SRP states that the effects of cladding crud and oxidation i needs to be addressed in safety and design analyses, such as in the thermal and mechanical analysis. The amount of cladding oxidation is dependent on the cladding type, fuel rod powers, I water chemistry control and primary inlet coolant temperatures, but the amount of oxidation and crud buildup increases with burnup and cannot be eliminated. Therefore, extended burnups result in a thicker oxide layer that provides an extra thermal barrier, cladding thmning and ductility decrease that can affect the mechanical performance. The degree of this effect is dependent on cladding type, reactor coolant temperatures, power history, and the level of success of a reactors' water chemistry program. The following is an evaluation of the FCF corrosion model.

' FCF has adopted a new cladding corrosion model, COROSO2 (Reference 3), that is more conservative, i.e.,' predicts more corrosion, than the original OXIDEPC model in TACO 3 and predicts the accelerated conosion observed in high bumup rods much better than the OXIDEPC

model. This model has been appmved by NRC with the commitment by FCF to collect more maximum corrosion thickness data in the future (Reference 3). The Mark-Bil and the similarly designed Mark-BW LTAs will also provide corrosion data up to extended burnup levels (see Section 6.0 on Fuel Surveillance) to verify the applicability of the new corrosion model to the

- Mark-B11 design. The best estimate or slightly conservative prediction of the COROSO2 model is considered to be acceptable because of the conservatism in the FCF maximum corrosion limit. ,

Based on FCFs commitment to collect corrosion data at extended burnup levels from their Mark-B and Mark-BW LTAs, PNNL concludes that the COROSO2 model is acceptable for

' application to the Mark-Bil design in predicting maximum corrosion levels up to the current Mark B operating bumup limit of 62 GWd/MTU (rod-average).

It is'noted that FCF performs reload / cycle specific evaluations to verify that cladding corrosion is within their design limit. These cycle specific evaluations are not within the scope of this review.

3.4 j

(

.l

l 3.6 ROD BOWING Bases / Criteria - Fuel and burnable poison rod bowing are phenomena that alter the design-pitch dimensions between adjacent rods.; Bowing affects local nuclear power peaking and the local heat transfer to the coolant. Rather than place design limits on the amount of bowing that is permitted, the effects of bowing are included in the departure from nucleate boiling ratio (DNBR) analysis by a DNBR penalty when rod bow is greater than a predetermined amount.

This approach is censistent with Section 4.2 of the SRP and has previously been approved for Mark-B designs up to a rod-average burnup of 62 GWd/MTU (References 2 and 3). This rod bowing criterion is also acceptable for application to the Mark-Bil de;ign up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

Evaluation - The FCF methodology for rod bowing analysis has been found to be very I

conservative for current Mark-B designs up to a rod-average burnup of 62 GWd/MTU (Reference 3). Rod bowing has been found to be dependent on the distance between grid spacers, the rod moment ofinertia, material characteristics of the cladding, and flux distribution. ,

The moment ofinertia has changed a small amount with the change in cladding diameter but the effect on the rod bowing for the Mark-B11 assembly should be insignificant or a slight -

improvement. In addition, FCF intends to collect rod bow data from the Mark-B11 LTAs to confirm that the current FCF methodology remains conservative. Based on FCFs commitment to collect rod bow data from their Mark-Bil LTAs, PNNL concludes that FCF rod bow analysis methods are applicable to the Mark-B11 design up to the current Mark-B bumup operating limit l of 62 GWd/MTU (rod-average). l 3.7 AXIAL GROWTH Bases / Criteria - The FCF design basis for axial growth is that adequate clearance be maintained between the fuel rod end-cap-shoulder and the top and bottom nozzles, i.e., shoulder gap clearance, to accommodate the differences in the growth of fuel rods and the growth of the fuel assembly. Similarly, for assembly growth, FCF has a design basis that axial clearance between core plates and the bottom and top assembly nozzles should allow sufficient margin for fuel assembly irradiation growth during the assembly lifetime. These bases are consistent with j Section 4.2 of the SRP and have previously been approved (References 2 and 3). These bases are also acceptable for application to the Mark-B11 design up to the current Mark B operating I

burnup limit of 62 GWd/MTU (rod-average).

Evaluation - The FCF models used to predict shoulder gap clearance and assembly growth ,

are based on gap clearance data and axial growth data from Mark-B and Mark-BW designs and l FCF claims that they are applicable to those for the Mark-B11 design. FCF was questioned ]

(Reference 5) on the applicability of this data to the Mark-B11 design and was requested to provide their one cycle shoulder gap clearance and growth data for comparison to those data from the earlier designs. They were also requested to provide the margin to shoulder gap closure and 3.5

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[ .

the margin for compressing the cruciform holddown springs to solid height up to a rod-average bumup of 62 GWd/MTU.

l The FCF response (Reference 6) presented one cycle data from the Mark-B11 LTAs that l . indicated that the Mark-Bil shoulder gap and assembly growth data were within the scatter of l the earlier Mark-B and Mark-BW data. FCF also provided the margins requested showing that i

both the margins for shoulder gap closure and solid compression of the holddown springs were relatively small up to a rod average burnup of 62 GWd/MTU and 64 GWd/MTU, respectively.

However, examination of the FCF analysis methods used for predicting shoulder gap clearances l l and assembly growth demonstrate that they are very conservative. For example, the FCF l bounding curves used for both of these analyses are significantly greater than the 95/95 bounds of l

the~ data. Therefore, the actual margins to the design bases for axial growth are quite large. In addition, FCF intends to collect axial growth and shoulder gap clearance data from the Ma k-B11 LTAs. PNNL concludes that these axial growth analysis methods are conservative. Ther .. are, l

PNNL further concludes that they are acceptable for application to the Mark-Bil design and that i the design is acceptable with respect to axial growth up to the current Mark-B operating burnup l limit of 62 GWd/MTU (rod-average).

3.8 ROD INTERNAL PRERSURE Bases / Criteria - Rod intemal pressure is a driving force for, rather than a direct mechanism of, fuel system damage that could contribute to the loss of dimensional stability and cladding integrity. Section 4.2 of the SRP presents a rod pressure limit of maintaining rod pressures below system pressum that is sufficient to preclude fuel damage. The FCF design basis for the ,

fuel rod intemal pressure is that the fuel system will not be damaged due to excessive fuel rod -i intemal pressure and FCF has established the " Fuel Rod Pressure Criterion" (Reference 15) to provide assurance that this design basis is met. These criteria are that the intemal pressure of the FCF lead fuel rod in the reactor is limited to a value below which could cause 1) the diametral gap to increase due to outward cladding creep during steady-state operation, and 2) extensive DNB propagation to occur. This FCF design basis and ,the associated criteria have been found acceptable by the NRC (Reference 15) up to the current Mark-B burnup limits established in Reference 3: PNNL concludes these are also acceptable for application to the Mark-Bil design up to the current Mark B operating bumup limit of 62 GWd/MTU (rod-average).

Evaluation - FCF utilizes the approved TACO 3 fuel performance code (Reference 16) for predicting end-ofslife (EOL) fuel rod pressures and the methodology described in Reference 15 to verify that they do not exceed the FCF " Fuel Rod Pressure Criterion" during normal operation and AOOs. The TACO 3 fuel performance code is generic enough to be applicable to all FCF

> PWR fuel designs, and therefore is acceptable for application to the Mark-B11 design up to the current Mark B operating burnup limit of 62 GWd/MTU (rod-average). The issue of DNB l propagation (Fuel Rod Pressure Criterion 2 above) will be discussed in Section 4.3. The FCF rod pressure analyses are performed on a reload / cycle specific basis.

3.6 i

7- '

i 3.9 ASSEMBLY LIFIDFF Bases / Criteria - Section 4.2 of the SRP calls for the fuel assembly holddown capability (wet weight 3nd spring forces) to exceed worst case hydraulic loads for normal operation and AOOs. The FCF design criterion for assembly liftoffis that the holddown spring system shall be capable of maintaining fuel assembly contact with the lower support plate during normal operation and AOOs. This is consistent with the SRP guidelines and has previously been approved (References 2 and 3). This criterion is also acceptable for application to the Msk-Bil design up to the current Mark B operating burnup limit of 62 GWd/MTU (rod-average).

Evaluation - The fuel assembly liftoff forces are a function of primary coolant flow, holddown spring forces, assembly dimensional changes and friction pressure drop across the length of the assembly with the spacer grids a major contributor to the pressure drops. FCF has

_ performed several hydraulic tests in a full scale flow facility to measure the pressure drop characteristics of the Mark-Bil fuel assembly which were used to calculate the form loss coefficients.

FCF has performed several analyses of hydraulic lift forces using the form loss coefficients for a Mark-B11 assembly in both a full core and mixed core environment that demonstrates that

. e . the Mark-Bil assembly has lower lift forces than a Mark-BIO assembly for both core environments. This demonstrates that the Mark-Bil lift loads are bounded by the Mark-BIO

values. PNNL concludes that FCF has performed adequate testing and analyses to verify the lift forces for the Mark-Bil design meet the FCF design criterion and, therefore, this issue has been adequately' addressed.

9 e

{

3.7 i

l

j

' '4 0 FUEL ROD FAILURE In the follo ving paragraphs, fuel rod failure thresholds and analysis methods for the failure mechanisms listed in the SRP will be reviewed. When the failure thresholds are appl:!ed to normal operation including AOOs, they are used as limits (and hence SAFDLs) since fue' uilure under those conditions should not occur according to the traditional conservative interpretation of GDC 10. When these thresholds are used for postulated accidents, fuel failures are permitted, but they must be accounted for in the dose assessments required by 10 CFR 100. The basis er reason for establishing these failure thresholds is thus established by GDC 10 and Part 100 and i only the threshold values and the analysis methods used to assure that they are met are reviewed below.

4.1 HYDRIDING g Bases / Criteria - Internal hydriding as a t. ' ' 'ing failure mNhanism is precluded by controlling the level of hydrogen impurities in L.s fuel during fa rication; this is generally an early-in-life feilure mechanism. FCF has not discussed their cri' :ria for intemal hydriding in the subject topical report; however, a limit on hydrogen level for FCF pellets is discussed in Reference 17. The hydrogen level of FCF fuel pellets is controlled by drying the pellets in the cladding and taking a statistical sample to ensure that the hydrogen level is below a specified  ;

level. Previous FCF design reviews, e.g., Reference 17, have shown that this level is below the l value recommended in the SRP. Consequently, PNNL concludes that the FCF limit on hydrogen j' in their fuel pellets is acceptable for the Mark-Bil design.

External hydriding of the cladding due to waterside corrosion is the other source and is  ;

discussed in Section 3.5 of this TER. As noted in Section 3.5, the level of external hydriding is controlled by FCF by a proprietary limit on corrosion thickness. PNNL concludes that this  !

corrosion iimit is acceptable for limiting the level of extemal hydriding in the cladding for the Mark-B11 design up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

Evaluation - Intemal hydriding is controlled by FCF by taking statistical samples following i pellet fabrication prior to loading the p:llets in the fuel rods and confirming that hydrogen is below a specified level. Therefore, no analyses are necessary other than to confirm that the statistical pellet sampling is below the speciSed level for Mark-Bil designs.

Extemal hydriding is centrolled by the FCF limit on corrosion thickness discussed in Section 3.5 of this TER.

PNNL concludes that FCF has addressed the issue of hydriding in Mark-B11 designs up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

4.1 9

4

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'4.2 f1 ADDING COLLAPSE Bases / Criteria - If axial gaps in the fuel pellet column were to occur due to fuel densification, the potential would exist for the cladding to collapse into a gap. Because of the large local strains'that would result from collapse, the cladding is then assumed to fail. The FCF

~ design criterion is that cladding collapse is precluded during the fuel rod design lifetime. This design basis is the same as that in Section 4.2 of the SRP and has previously been approved

(

Reference:

2 and 3). His criterion is also acceptable for application to the Mark-Bil design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

Evaluation - De FCF analytical models for evaluating cladding creep collapse are the CROV and TACO 3 computer codes that have been reviewed and approved by NRC (Refer.mces 1 18 and 16). FCF has provided the results of their bounding creep collapse analysis that demonstrates tha't collapse will not occur for the Mark-Bil design up to a rod-average burnup of 70 GWd/MTU using a conservatively high average power history. PNNL concludes that these i codes and methods are conservative for evaluating cladding creep collapse in FCF PWR designs l ,

and, therefore, are acceptable for application to the Mark-B11 design. Based on the FCF )

analyses, PNNL further concludes that the Mark B11 design is acceptable with respect to cladding collapse up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-I

. average). .

4.3 OVERHEATING OF CLADDING Bases / Criteria - The FCF design limit for the prevention of fuel failures due to cladding overidig is that there will be at least a 95% probability at a S T% confidence level that departure from nucleate boiling (DNB) will not occur on a fuel rod having the minimum DNBR during normal operation and AOOs. This design limit is consistent with the thermal margin criterion of Section 4.2 of the SRP.and has previously been approved for FCF designs .

(References 2 and 3). This design limit is also acceptable for application to the Mark-Bil design up to the currerd Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

Evaluation - As stated in the SRP, Section 4.2, adequate cooling is assumed to exist when the thermal margin criterion to limit DNB or boiling transition in the core is satisfied. FCF has submitted a new CHF conelation for the Mark-B11 design. FCF utilizes NRC-approved critical heat flux (CHF) conelations for evaluating thermal margins and these analyses are performed on a reload / cycle specific basis.

As noted in Section 3.8, one of the design criteria for rod presores is that the limit on rod pressures prevent extensive DNB propagation to occur. The FCF methodologi for evaluating DNB propagation is described in Reference 15 and has been approved by NRC. PNNL concludes that this FCF analysis methodology for preventing DNB propagation due to rod over-pressures is acceptable for appliention to the Mark-Bil design.

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)e 4 9 4.4 OVERHEATING OF FUEL PFII ETS -

Bases / Criteria - As a secondl method of avoiding cladding failure due to overheating, FCF precludes centerline pellet melting during normal operation and AOOs. This design criterion is the 'same as that given in the SRP and has previously been approved for FCF designs up to ,

I cunent operating limits (References 2 and 3). This criterion for fuel melting is also acceptable for application to the Mark-B11 design up to the current Mark-B operating bumup limit of 62 ,

GWd/MTU (rod-average). l Evaluation - FCF utilizes the approved TACO-3 fuel performance code to determine the maximum linear heat generation rate (LHOR) at which a given fuel design will not achieve fuel melting at a 95% probability at a 95% confidence level. This FCF analysis methodology has been

- found to be acceptable to Mark-B designs up (Reference 2) to a rod-average bumup of 62 GWd/MTU (Reference 3). PNNL also fmds them acceptable for application to the Mark-B11 design up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average).

FCF has also performed a fuel melting analysis for the Mark-Bil fuel design that demonstrates that the Mark-B11 design is acceptable within the design's operating limits. PNNL

- concludes that the Matk-B11 design is acceptable in relation to fuel melting up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

- 4J, PFII.ET/CLADDINGINTERACHQN Bases / Criteria - As indicated in Section 4.2 of the SRP, there are no generally applicable l criteria for pellet cladding interaction (PCI) failure. ' However, two acceptance criteria oflimited application are presented in the SRP for PCI: 1) less than 1% transient induced cladding strain, j

- and 2) no centerline fuel melting. Both of these limits have been adopted by FCF for use in '

evaluating their fuel designs and have been approved by the NRC. These two criteria have been satisfactorily addressed in Sections 3.2 and 4.4 of this TER and will not be discussed further in this section.

Evaluation - As noted earlier, FCF utilizes the TACO-3 (Reference 16) code to show that their fuel meets both the cladding strain and fuel melting criteria. This code is acceptable pe- the ,

recommendations in Sections 3.2 and 4.4.

4.6 CLADDING RUP'IURE Bases / Criteria - There are no specific design limits associated with cladding rupture other than t'a 10 CFR 50, Appendix K (Reference 19) requirements that the incidence of rupture not  ;

. ' be underestimated. FCF uses a rupture temperature correlation consistent with NUREG-0630 guidance (Reference 20). PNNL concludes that FCF has adequately addressed cladding rupture for the Mark-Bil design up to the cunent Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

4.3

Evaluation - FCF has adopted the cladding deformation and rupture models from NUREG-0630 guidance (Reference 20) which has been approved by the NRC for ECCS evaluation. PNNL concludes that FCF has adequately addressed the issue of cladding rupture for the Mark-B11 design up to the current Mark-B operating bumup limit of 62 GWd/MTU (rod-average). -

l 4.7 FUEL ROD MECHANICAL FRACTURING Bases / Criteria - The term " mechanical fracture" refers to a fuel rod defect that is caused by an externally applied force such as a hydraulic load or a load derived from core-plate motion.

The design limits proposed by FCF to prevent fracturing is that the stresses due to postulated accidents in combination with the normal steady-state fuel rod stresses should not exceed the stress limits established in the approved methodology (Reference 2) for Mark-B fuel assembly designs. These design limits for fuel rod mechanical fracturing are acceptable for application to the Mark-Bil fuel design up to the current Mark-B operating burnup limit of 62 GWd/MTU

. (rod-average).

Evaluation - The mechanical fracturing analysis is done as a part of the seismic-and-LOCA '

loading analysis. A discussion of the seismic-and-LOCA loading analysis is given in Section 5.4 of this TER.-

S 4.4 D

. , -e -

5.0 FUEL COOLABILIIY For postulated accidents in which severe fuel damage might occur, core coolability must be maintained as required by several GDCs (e.g., GDC 27 and 35). In the following paragraphs, limits and methods used to assure that coolability is maintained are discussed for the severe damage mechanisms listed in the SRP.

5.1 FRAGMENTATION OF EMBRITTIFn CLADDING Bases / Criteria - The most severe occurrence of cladding oxidation and possible fragmentation during a postulated accident is the result of a LOCA. FCF has not discussed cladding embrittlement as a result of a LOCA in the subject topicai report but this has been previously presented by FCF in References 2 and 3 that have been approved by NRC. In order to

. reduce the effects of cladding oxidation during LOCA, FCF uses a limiting criteria of 2200*F on peak claddirig temperature (PCT) ard a limit of 17% on maximum claciding oxidation as prescribed in 10 CFR 50.46 and consistent with the SRP criteria. PNNL concludes that these

. criteria are also applicable to the Mark-B11 design up to the current Mark-B operating burnup limit of 62 GWd/MTU.

Evaluation - FCF has evaluated the impact of the Mark-B11 design changes on LOCA utilizing approved LOCA analysis methods. This analysis concluded that the Mark B-11 design meets the requirements of 10 CFR 50.46, and FCF will conf:rm this on a plant specific basis.

5.2 VIOLENT EXPULSION OF FUEL l Bases / Criteria - In a severe reactivity insertion accident (RIA), such as a control rod ejection accident, large and rapid deposition of energy in the fuel could result in melting, l fragmentation, and dispersal of fuel. The mechanical action associated with fuel dispersal might i be sufficient to destroy the fuel cladding and rod bundle geometry and provide significant pressure pulses in the primary system. To limit the effects of an RIA event, Regulatory Guide 1.77 (Reference 21) recommends that the radially-averaged energy deposition at the hottest axial location be restricted to less than 280 cal /g and the onset of DNB is assumed to be j the failure limit. It is noted that the NRC staff are currently reviewing the 280 cal /gm limit and l

the limit for fuel failure may be decreased to a lower limit at high burnup levels. Recent RIA i teuang has indicated that fuel expulsion and fuel failure may occur before the 280 cal /gm limit and the onset of DNB, respectively (Referer ces 22 and 23). However, further testing and  ;

evaluation is 'needed to establish limits. The fuel expulsion and failure limits for an RIA may i decrease in the future but the current limits remain valid at this time.

The FCF design criterion for this event is identical to that in Regulatory Guide 1,77, such that the peak fuel enthalpy for the hottest axial fuel rod location shall not exceed 280 cal /gm. 1 Therefore, PNNL concludes that FCF design limits for fuel dispersal are acceptable for application to the Mark-B11 design up to the current Mark-B operating bumup limit of 5.1 9

9

y, . . ,

. 62 GWd/MTU.

Evaluation - FCF verifies that this acceptance criterion is met for cach fuel cycle through design and cycle spec.ific analyses and by limiting the ejected rod worth. FCF uses NRC-approved methods to perform these analyses and the methods remain valid for the Mark-B11

' design. PNNL concludes that the analysis methodology remains acceptable for application to l the Mark-Bil fuel design up to the current Mark-B operating burnup limit of 62 GWd/MTU

_ (rod average).

5.3 CLADDING BALLOONING Bases / Criteria - Fuel cladding will balloon (swell) under certain combinations of temperature, heating rate, and stress during a LOCA. There are no specific design limits associated with cladding ballooning other than the 10 CFR 50 Appendix K requirement that the degree of swelling not be underestimated.

Evaluation - The cladding ballooning model and flow blockage model are directly coupled

. to the cladding rupture temperature model for the LOCA-emerEency core cooling system (ECCS) analysis that is plant specific. FCF has adopted the cladding rupture and ballooning models from NUREG-0630 (Reference 20) as recommended by Section 4.2 of the SRP and these models have been previously approved by the NRC. Therefore, PNNL concludes that FCF has' adequately

- addressed the issue of cladding ballooning and that these models remain acceptable for application to Mark-B11 designs up to the current Mark-B operating burnup limit of

. 62 GWd/MTU (rod-average).

- 5.4 FUEL ASSEMBLY STRUCTURAL DAMAGEJ.EDME2GIENAL.EORCES Bases / Criteria - Earthquakes and postulated pipe breaks in the reactor coolant system would result in extemal forces on the fuel assembly. . Appendix A to SRP Section 4.2 states that the fuel system coolable geometry shall be maintained and damage should not be so severe as to prevent control rod insertion during seismic and LOCA events. The FCF design basis is that the fuel assembly will maintain a geometry that is capable of being cooled under the worst case design accident and that no interference between control rods and thimble tubes will occur during a safe shutdown earthquake. This is consistent with the SRP and is therefore acceptable for application to the Mark-B11 fuel design up to the current Mark-B operating limits.

Evahmtion - FCF has performed impact tests on the Mark-B11 spacer grids to characterize the plastic deformation and clastic limits of the spacer grids. These tests show that the Mark-B11 spacer grids.are slightly stronger than the previous Mark-B Zircaloy grids. FCF has also performed dynamic pluck, axial stiffness and lateral stiffness tests on the Mark-Bil assembly that determined that the natural frequency, and axial and lateral stiffness values were close to those of previous Mark-B assemblies with Zircaloy grids.

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FCF has performed a seismic-LOCA aaalysis using approved analysis methods to

' determine the Mark-B11 fuel assembly structural response to bounding seismic-LOCA loadings.

, , These analyses demonstrate that the grid spacer loadings are well within their elastic limits and,

! therefore, th'eassembly retains a coolable geometry. Consequently, PNNL concludes that FCF i has satisfactorily addressed the issue of seismic-LOCA loads for the Mark-B11 design up to the I current Mark-B operating burnup limit of 62 GWd/MTU (rod average).

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6.0 FUEL SURVEII T ANCE FCF was q estioned about what futme fuel surveillance would be performed to verify .

< satisfactory performance of the Mark-Bil. FCF responded that their lead test assembly.(LTA)

- progam consisted of four Mark-B11 fuel assemblies being irradiated in Oconee-2. Three of the

- four assemblies will be irradiated for two cycles (assembly average burnup of 25 GWd/MTU) and one assembly for three cycles (assembly average bumup of 39 GWd/MTU). The LTAs will be placed in positions in the core periphery (where previous fretting had been observed) during

- the second cycle in order to demonstrate that the new spacer grids are not susceptible to fretting

. wear. Each Mark-Bil LTA will be subjected to the following inspections; visual, fuel assembly length and bow, guide tube distortion, spacer grid width, and fuel rod shoulder gap clearances.

The oxide thickness of the fuel rods, guide tubes, and spacer grids will also be measured.

PNNL verbally questioned FCF about the lack of high burnup Mark-B11 data, i.e., above .

an assembly average burnup of 39 GWd/MTU, particularly in regards to cladding corrosion -

because this is one of the bumup limiting parameters for FCF fuel designs. FCF responded that the mixing vane grid design, in Mark-B11 is essentially the same as used in the Mark-BW designs from which they have higher burnup data and also from European fuel designs with mixing vane grids. FCF has cladding oxidation data from the. Mark-BW design up to rod-average burnups of 54 GWd/MTU that demonstrate that their COROSO2 corrosion model adequately predicts cladding corrosion, and therefore, it is expected that it will also adequately predict cladding

- corrosion for the Mark-B11 design up to the current Mark-B operating burnup limit of 62 GWd/MTU (rod-average).

PNNL concludes that FCF has adequately addressed the issue of fuel surveillance.

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PNNL has reviewed the FCF thermal-mechanical design criteria and analyses for the Mark-Bil fuel design presented in Reference 1 in accordance with Section 4.2 of the SRP.

PNNL concludes that the Mark-Bil design as described in Reference 1 is acceptable for reload licensing applications up to a rod-average burnup of 62 GWd/MTU. 1

. - As noted in Section 4.3 of this TER the critical heat flux correlation for the Mark-Bil design is still under review and needs to be approved before the design can be used in reload applications. For those licensees that apply this reload methodology, the following plant-specific i analyses or evaluations are required: 1) cladding oxidation (Section 3.5); 2) rod internal pressures l (Section 3.8); 3) overheating of cladding (Section 4.3); and 4) ECCS related analyses (Sections '

5.1,5.2, and 5.3).

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8.0 REFERENCES

1. Framatome Cogema Fuels. Septernber 1997. Mark-Bil Fuel Assembly Design Topical Report. BAW-10229P, Framatome Cogema Fuels, Lynchburg, Virginia, transmitted by letter, J. H. Taylor (FCF) to U.S. NRC Document Control Desk, " Submittal of Topical Report BAW-10229P, Mark Bil Fuel Assembly Design Report," dated September 30, 1997, JHT/97-35.

2. - B&W Fuel Company. February 1991. Safety Criteria and Methodolorv for Acceptable '

Cvele Reload Analyses. BAW-10179P, B&W Fuel Company, Lynchburg, Virginia.

3. B&W Fuel Company. November 1992. Extended Burnup Evaluation. BAW-10186P, B&W Fuel Company, Lynchburg, Vuginia.

4. Framatome Cogema Fuels. September 1996. The BWU Critical Heat Flux Correlations BAW-10199P, Addendum 1, Framatome Cogema Fuels, Lynchburg, Virginia.

5. Letter, J. L. Birmingham (NRC) to C. F. McPhatter (FCF), " Request for Additional Information for Topical Report BAW-10229P, Mark-B11 Fuel Assembly Desien R..pp21," .

dated September 8,1998.

6. Letter, T. A. Coleman (FCF) to U.S. NRC Document Control Desk, dated November 13, 1998, GR855. doc.

7. U.S. Nuclear Regulatory Commission. July 1981. "Section 4.2, Fuel System Design." In j Standard Review Plan far the Review of Safety Analysis Reports for Nuclear Power {

Plants--LWR Edition. NUREG-0800, Revision 2, U.S. Nuclear Regulatory Commission,  !

Washington, D.C. I

\

8. United States Federal Register. " Appendix A, General Design Criteria for Nuclear Power Plants " 10 Code of Federal Regulations (CFRT. Part 50. U.S. Printing Office, Washington, D.C.

9. United States Federal Register. " Reactor Site Criteria." 10 Code of Federal Reaulations (CFR). Part 100. U.S. Printing Office, Washington, D.C. {

10. American Society of Mechanical Engineers.1983 Edition. "Section III, Nuclear Power Plant Components." ASME Code. American Society of Mechanical Engineers, New York.

11. Smith, Jr. G. P., R. C. Pirek and M. Griffiths. July 1994. Hot Cell Examination of Extended Burnup Fuel from Calvert Cliffs-1. EFR1 TR-103302-V2, Final Report, Electric Power Research Institute, Palo Alto, California.

8.1 4

e 8

. 12. - Newman,.L'. W. et al.1986. The Hot Cell Examination of Oconee Fuel Rods After Five Cveles ofIrradiation. DOFJET/34212-50 (BAW 1874), Babcock & Wilcox, Lynchburg, Virginia.

13. Garde, A. M.1989. " Effects ofIrradiation and Hydriding on the Mechanical Properties of

~ Zircaloy-4 at High Fluence." 'In Zirconium in the Nuclear Industry: Elohth International Svmposium, ASTM STP 1023, pp. 548-569, Eds. L.F.P. VanSwam and C. M. Eucken.

American Society for Testing and Materials, Philadelphia, Pennsylvania.

14. O'Donnell, W. J., and B. F. Langer. 1964. " Fatigue Design Basis for Zircaloy Components." In Nuc. Sci. Eng. 20:1.

15. D. A. Wesley, D. A. Farnsworth and G.A. Meyer, July 1995. Fuel Rod Gas Pressure Criterion (FRGPC), BAW-10183P-A, B&W Fuel Company, Lynchburg, Virginia.

I

16. Wesley, D. A., and K. J. Firth. October 1989. B.CO-3 Fuel Pin Thermal Annivsis Code.

BAW-10162P-A, Babcock & Wilcox, Lynchburg,, Virginia.

17. Babcock and Wilcox. April 1986. Extended Burnup Evalnation. BAW-10153P-A, Babcock and Wilcox, Lynchburg, Virginia.

18. Miles, T., et. al., August 1995. Procram to Determine In-Reactor Performance of B&W Fuel Cladding Creen Collanse, BAW.10084P-A Rev. 3, B&W Fuel Company, Lynchburg, Virginia. -

19. United States Federal Register. " Appendix K, ECCS Evaluation Models." 10 Code of Federal Rennlations (CFR1 Part 50. U.S. Printing Office, Washington, D.C.

20. Powers, D. A., and R. O. Meyer.1980. Cladding Swelling and Rupture Models for LOCA Annivsis. NUREG-0630, U.S. Nuclear Regulatory Commission, Washington, D.C.

21. U.S. Nuclear Regulatory Commission.1974. Assumptions Used for Evaluating a Control Rod Election Accident for Pressurized Water Reactors, Regulatory Guide 1.77, U.S.

Nuclear Regulatory Commission, Washington D.C.

22. Schmitz, F., et. al., March 1996. " New Results from Pulse Test in the CABRI Reactor,"

Proceedings of the 23rd Water Reactor Safety Information Meeting October 23-25.1995.

23. Fukets, T., et. al., March 1996. "New Results from the NSRk Experiments with High Burnup Fuel," Proceedings of the 23rd Water Reactor Safety Meeting.

October 23-25;1995

~

. 8.2

._ . . . ..