ML20212N200
| ML20212N200 | |
| Person / Time | |
|---|---|
| Issue date: | 07/23/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20212N170 | List: |
| References | |
| NUDOCS 8608280127 | |
| Download: ML20212N200 (27) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO EXXON NUCLEAR COMPANY TOPICAL REPORT XN-NF-85-67 (P), REVISION 1
" GENERIC MECHANICAL DESIGN FOR EXXON NUCLEAR JET PUMP BWR RELOAD FUEL"
1.0 INTRODUCTION
By letter dated July 18, 1985 (Reference 1), the Exxon Nuclear Company submitted for review XN-NF-85-67 (P), Revision 0, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel." The topical report provides a design description and a suninary of the supporting analyses and test results for the Exxon Nuclear Company (ENC) Jet Pump Boiling Water Reactor (JP-BWR) 8x8 and 9x9 reload fuel designs. Design criteria, technical bases for the criteria, and a description of mechanical fuel rod failure mechanisms were presented in a separate report, XN-NF-85-39 (Reference 2).
XN-NF-85-67 is a revision to the previously approved generic mechanical design report XN-NF-81-21 (P) (A) (Reference 3) which presented results and information for the ENC JP-BWR 8x8 reload fuel design. The main changes to XN-NF-81-21 (P) (A) were: (1) the addition of results and infomation for the ENC JP-BWR 9x9 reload fuel design, (2) the use of the RODEX2A code (Reference 4), instead of R0DEX2 (Reference 5), in most of the mechanical design evaluations (the staff evaluation of RODEX2A is documented in Reference 35), (3) results of analyses to support an increase in the hirnup limit of ENC JP-BWR 8x8 reload fuel, and (4) the use of different LHGR limits for the 8x8 fuel and revised design criteria for cladding collapse, steady state strain, and internal gas pressures in the evaluations for the ENC JP-BWR 8x8 reload fuel design. In XN-NF-85-67 (P), ENC has presented results of analyses to support operation to extended burnup exposures.
e60828o127 860723 PDR TOPRP EMVEXXN
2 The review documented in this Safety Evaluation Report covers the ENC JP-BWR 8x8 and 9x9 reload fuel designs to those burnups which have been previously approved for BWR fuel (30,000 MWD /MTV batch average exposure).
ENC's extended burnup topical reports for the 8x8 fuel Reference 6) and for the 9x9 fuel (Reference 7) cover the design bases / criteria and analysis methods for application to extended burnup levels. The staff evaluation of Reference 6 is documented in Reference 36. Reference 7 is currently under review with SER issuance projected for the fourth quarter of FY 1986. Consequently, the approval of the ENC JP-BWR 8x8 and 9x9 reload fuel designs for operation to extended burnup levels is contingent on the generic approval of the method by whicn burnup is considered in the design and analytical processes as described in References 6 and 7.
i In our review of XN-NF-85-67 (P), we requested additional information (Reference 8). In response, ENC submitted Revision 1 of XN-NF-85-67 (Reference 9) which supersedes Revision 0 of XN-NF-85-67, and also provided Reference 10 responding to our questions and including errata sheets.to XN-NF-85-67, Revision 1. WehavereviewedXN-NF-85-67(P),
Revision 1 (Reference 9), XN-NF-85-39 (P) (Reference 2), and ENC's responses to the staff's reauest for additional information (Reference 10). The report, XN-NF-85-67 (P), Revision 1, presents reload fuel design information and related safety analyses of the kind found in Section 4.2 of plant FSARs. Since the report will be referenced by some plants which are required to meet all of the guidelines of the current Standard Review Plan, the NRC staff has reviewed XN-NF-85-67 (P), Revision 1, in accordance with Section 4.2 of NUREG-0800, dated July, 1981, the latest version of the SRP.
2.0 FUEL SYSTEM DESIGN OBJECTIVES The objectives of this fuel system safety review as described in Section 4.2 of the Standard Review Plan (SRP) are to provide assurance that (a) i 1 - _
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the fuel system is not damaged as a result of normal operation and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained. A "not damaged" fuel system is defined as meaning that fuel rods do not fail, that fuel system dimensions remain within operational tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis.
Objective (a) above is consistent with General Design Criterion 10 (10 CFR 50, Appendix A), and the design limits that accomplish 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) hct, therefore, been breached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR 100 for postulated accidents. "Coolability," which is sometimes termed "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 even after a severe accident. The general requirements '
to maintain control rod insertability and core coolability appear repeatedly in the General Design Criteria (e.g., GDC 27 and 35). Specific coolability requirements for the loss-of-coolant accidents are given in 10 CFR 50 Section 50.46.
To assure that the above stated objectives are met, the following areas are examined: (a) design bases, (b) descriptionanddesigndrawings,(c) design evaluation, and (d) testing, inspection, and surveillance plans.
In assessing the adequacy of the design, operating experience, prototype testing, and analytical predictions are compared with acceptance criteria for fuel system damage, fuel rod failure, and fuel coolability.
3.0 DESIGN BASES AND EVALUATION Design bases for the safety analysis address fuel system damage mechanisms and suggest limiting values for important parameters such that damage will
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4-4 be limited to acceptable levels. For convenience, we group acceptance criteria for these design limits into three categories in the Standard Review Plan (SRP): (a) fuel system damage criteria, which are most applicable to normal operation, including anticipated operational occurrences (A00s), (b) fuel rod failure criteria, which apply to normal operatico, A00s, and postulated accidents, and (c) fuel coolability criteria, which apply to postulated accidents.
ENC methods of demonstrating that the described fuel designs meet the design criteria are also reviewed. Operating experience, prototype testing, and analytical predictions are cited as evidence that the fuel designs conform to the design criteria.
The subsection designations below follow the organization of the SRP rather than XN-NF-85-67 (P), Revision 1.
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3.1 FUEL SYSTEM DAMAGE CRITERIA The following paragraphs discuss the design bases, corresponding design limits and design evaluations for the damage mechanisms listed in the SRP.
These design limits along with certain criteria that define failure (Section 3.2) constitute the Specified Acceptable Fuel Design Limits (SAFSLs) required by General Design Criterion (GDC) 10. The design limits
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in this section should not be exceeded during normal operation and A00s.
(a) Design Stress Steady state stress design limits for fuel cladding are presented in Table 4.1 of XN-NF-85-39. For primary membrane stress, the allowable stress intensity limit is the lower of 2/3 of the minimum yield strength and 1/3 of the minimum ultimate tensile strength of Zircaloy-2 in the unirradiated I
I condition. Stress allcwables for primary bending stress and secondary stress are also presented in Table 4.1. These limits are derived from the ASME Boiler and Pressure Vessel Code,Section III, Article III-2000. ,
For the fuel assembly structural components, the design basis is that the fuel assembly is capable of withstanding all nonnal axial loads from fuel ;
handling operations without permanent deformation. Compliance is demonstrated either by a stress analysis for a normal axial lifting load !
which is a design factor multiple of the static weight of the , fuel assembly showing that the shear, bending, and tensile stresses are less -
than the yield strength of each specified material, or a demonstration ,
test which shows that loads of that design magnitude result in no detectable yielding of the fuel assembly structural components. The stress categories and strength theory presented in the ASME Boiler and Pressure Vessel Code,Section III, are used as a general guide. !
ENC has used Section III of the ASME code as general guidance for i developing design stress limits. This conforms with the SRP guidelines and is therefore acceptable.
As indicated in Section 3.4.3 of XN-NF-85-67 (P), Revision 1, the primary ;
membrane stresses are calculated using the Lame equations recomended by Sharifft and Popov (Reference 15). Primary bending stresses due to ovality are calculated with Timoshenko's equation (Reference 16). The cladding thermal stress and thermal bow are calculated using standard equations described by Geodier (Reference 17) and Timoshenko and Gere (Reference 18), respectively. Other secondary stresses, such as those ,
caused by (a) mechanical bow between spacers, (b) flow induced vibration, ,
and (c) contact from spacer dimples and springs, are also considered and calculated using conventional equations and equations described in the open literature (References 19, 20, 21, and 22). Table 3.3 of XN-NF-85-67 ,
(P), Revision 1, shows that the calculational results are well below the
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6 design limits for normal operation. Table 3.4 also shows that cladding '
stresses at the fuel rod end caps, based on ANSYS code (Reference 22) calculations, are within the design limits. >: f ENC has tested the assembly strength by having the tie plates subjected to axial tensile forces in excess of the design load. The result shows no evidence of plastic deformation. Further tests were performed to determine the load at failure. The failures occur as expected at the tie rod end caps with no detrimental effects on grid spacers or upper and lower tie plates.
On the basis of testing and analyses with standard engineering methods, we
- conclude that reasonable assurance has been provided that fuel assembly components including fuel rods, spacer grids, and upper and lower tie {
plates meet the stress design criteria.
(b) Design Strain ENC uses a maximum end of life steady state cladding strain limit to prevent cladding failure due to plastic instability and localization of strain. The value of the limit is consistent with the SRP ~
guidelines and is thus acceptable. For transient conditions, which relate to pellet-to-cladding interaction (PCI) failures, the strain criterion is discussed in Section 3.2 (f) of this Safety Evaluation Report.
For cladding steady-state strain calculations, ENC uses the RODEX2A' code (Reference 4), which is an interactive calculational procedure that considers the thermal-hydraulic environment at the cladding surface, the pressure inside the cladding, and the thennal, mechanical, and compositional state of the fuel and cladding.
Calculations were performed using conservative power histories based
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7 on the steady state LHGR limit curves presented in Figures-3.1, 3.2, and 3.3 of XN-NF-85-67 (P), Revision 1. Two alternate steady state LHGR limit curves were used in the design evaluations for the ENC JP-BWR 9x9 reload fuel design. The steady state LHGR limit curves are in terms of peak pellet power versus assembly planar exposure.
Power histories similar to those used in the previously approved XN-NF-81-21 (P) (A) (Reference 3) were derived from the steady state LHGR limit curves using conservative axial peaking factors and are presented as Figures 3.5, 3.6, and 3.7 in terms of peak pellet power versus pellet exposure. The performance of the fuel rods was analyzed to extended burnup exposure levels.
The calculated end-of-life (E0L) strain for both the 8x8 and 9x9 fuel designs is well within the design criteria limit of 1.0%.
(c) Strain Fatigue 4
The Exxon design basis for strain fatigue limits the total cumulative damage factor (CDF) to a value which accounts for a corrosive environment and other fatigue mechanisms. Exxon has used a tatigue design curve from O'Donnell and Langer that includes a safety factor ;
of 2 on stress amplitudes or a safety factor of 20 on the number of cycles, whichever is more conservative. This is consistent with the SRP guidelines and is, thus, acceptable.
Calculations were performed using the duty cycles summarized in Table 3.5 of XN-NF-85-67 (P), Revision 1, which conservatively envelope the expected plant operation, and the conservative power histories described in Section 3.1 (b) of this SER. The allowable number of cycles for a particular cycle loading was determined from the fatigue design curve of O'Donnell and Langer. The calculated cumulative damage factors for the 8x8 and 9x9 fuel designs are well i below the design limit and, are therefore acceptable.
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- (d) Fretting Wear -
! i The design basis for fretting wear is that fuel rod failures due to }
fretting shall not occur. Since the SRP does not provide numerical !
! acceptance criteria for fretting wear, and since fretting wear is !
addressed in the design analysis, we conclude that this response to the I SRP guidelines is acceptable. ;
l f
In XN-NF-85-67(P), Revision 1, results of fretting wear measurements j obtained in inspections of irradiated fuel and result: from extensive ficw j tests on ENC assemblies under various spacer spring load conditions are l presented. The results showed no evidence of significant fretting or wear {
damage due to flow vibration forces at the spacer to fuel rod contact !
points. Based on the test results which showed that the residual spacer j spring holding force can be quite low without resulting in fretting damage :
to the cladding, and successful irradiation experience of ENC fuel, we f conclude that the ENC JP-BWR reload fuel design has been adequately l
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designed with respect to fretting wear. A confirmatory evaluation'should
! be performad as part of the fuel surveillance program. l l
(e) External Corrosion and Crud Buildup The ENC fuel design basis for cladding corrosion and crud buildup is !
to prevent significant degradation of cladding strength and unacceptable temperature increases due to corrosion product buildup.
j Because of the thermal resistance of corrosion and crud layers, they i
- result in elevation of temperature within the fuel as well as the cladding. ENC uses a cladding outer surface temperature limit for BWR fuel based on corrosion data and correlations (Reference 11) which indicate generally low corrosion rates below that temperature.
We agree with the bases for this limit and conclude that this limit, l
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in conjunction with the assumptions employed in calculation of oxide layer thickness and the treatment of crud deposition effects, a's discussed below, is acceptable. ENC uses a two-stage corrosion rate
- model with modified correlations adapted from MATPRO-11 (Reference ,
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- 11) in RODEX2A to calculate temperature and oxide thickness. Using L the conservative power histories described in Section 3.1 (b) of this i SER, ENC calculated maximum outer cladding temperatures below the f design limit. In addition, the calculated maximum thickness of -
the oxide layer was well within the manufacturing tolerances for the l fuel cladding. The conservative aspects of the calculation account !
for the additional adverse effects of crud deposition by defining a l crud layer thickness in the RODEX2A input. Based on the results of' l i
outer cladding temperature and oxide layer thickness calculations and the treatment of crud buildup in the analysis, we conclude that ENC '
JP-BWR reload fuel is adequately designed with regard to external corrosion and crud buildup. t (f) Rod Bowing ,
Differential expansion between the fuel rods and lateral tnermal and flux gradients can lead to lateral creep bow of the rods in the span between spacer grids. ENC's design basis for fuel rod bowing is that the lateral displacement of the fuel rods shall not be of sufficient magnitude to impact thermal margins. To accomplish this, ENC established a design limit for minimum gap spacing at end-of-life.
This limit was reviewed and accepted in the NRC review of l XN-NF-81-21(P)(A) (Reference 3) for the ENC 8x8 JP-BWR reload fuel design. In XN-NF-85-67(P), Revision 1, ENC proposes to use the same design limit for the 9x9 JP-BWR reload fuel design. The use of the same minimum gap spacing limit for the 9x9 assembly, which has a j smaller initial pitch compared to the 8x8 assembly, restricts fuel rod bowing to a greater extent than for the 8x8 assembly. If both assemblies were at the design limit, the 9x9 assembly would have
P 10 experienced less gap closure (as measured in mils and in proportion to the initial gap). We therefore, find the design limit acceptable for the 9x9 JP-BWR reload fuel design.
ENC has collected measurements on gap spacing in irradiated fuel for 7x7 and 8x8 fuel for burnups up to 30000 MWD /MTU and for 9x9 fuel after one cycle of irradiation. For the 8x8 JP-BWR reload fuel design, the NRC has reviewed the methodology and results presented in XN-NF-81-21(P)(A). We concluded that the 95/95 closure for the worst span will not exceed the ENC design gap spacing limit for burnups to 30,000 MWD /MTU(batchaverage).
For the 9x9 JP-BWR reload fuel design. ENC used rod bowing data for ;
other fuel designs as well as the 9x9 fuel data collected after one irradiation cycle. The 95/95 gap closure at an exposure of 23,000 ,
MWD /MTU (peak assembly) was estimated using the approved methodology presented in XN-NF-75-32(P)(A), Supplements 1, 2, 3, and a (Reference 38). The correlation in the data base was modified'in accordance with NRC requirements to include cold-to-ho't and batch-to-batch l variations. The resulting estimate of 95/95 gap closure satisfies' !
the ENC design limit for fuel rod bowing up to an exposure of 23,000 MWD /MTU (peak assembly). To support operation beyond the burnup limits described above, ENC has indicated that it will provide justification for a less severe design limit and/or provide additional rod bowing data from irradiated fuel to demonstrate that the ENC 8x8 and 9x9 JP-BWR reload fuel design meet acceptable design limits for rod bowing.
We conclude that the design of ENC jet pump BWR fuel is acceptable with respect to fuel rod bowing for exposures up to 30000 MWD /MTU for 8x8 fuel and 23000 MWD /MTU for 9x9 fuel. Additional justification is :
required to ascertain that the fuel can be subjected to higher i .
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l 11 exposure limits without potential adverse impact on cooling capability due to rod bow.
(g) Axial Growth l The design basis for axial irradiation growth is that the fuel rods i are to be properly retained in the fuel assembly structure and the fuel assembly must be compatible with the fuel channel and fuel assembly supports in the reactor during the design lifetime. As i discussed below, ENC has shown that the standard fuel rods and spacer capture rods have sufficient engagement with the upper and lower tie plate throughout the assembly design life. This conforms with the SRP guidelines and is thus acceptable. I To account for irradiation induced growth and creep growth, the fuel assembly is designed with sufficient clearances. Generally, a higher growth rate is experienced by the tie rods than by the fuel rods, and differential rod growth is predicted from fuel inspection data of irradiated ENC fuel. Maximum differential growth estimates were calculated based on results of poolside measurements of irradiated fuel at Big Rock Point, Oyster Creek, and Barsebeck. As shown in Table 3.1 of XN-NF-85-67 (P), Revision 1, the nominal engagement of the end caps to the upper and lower tie plates is greater than the
' maximum differential growth estimates. We conclude that there is sufficient margin for fuel rod growth.
Fuel assembly growth is a direct result of tie rod growth. ENC calculations of assembly growth based on data obtained at Big Rock Point, Oyster Creek, and Carsebeck showed that the growth is acceptably small in comparison to reactor internal clearances.
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12 (h) Fuel Rod Pressure .
Section 4.2 of the SRP identifies excessive fuel rod internal 42 pressure as a potential fuel system damage mechanism. It calls for !
rod pressures to remain below nominal system pressure during normal l operation unless otherwise justified. In XN-NF-82-06(P), Revision 1 ,
(Reference 6), XN-NF-85-39(P) (Reference 2), and XN-NF-85-67 (P), l Revision 1, ENC proposes the use of an internal rod pressure limit l which is greater than system pressure. In addition, ENC has imposed l a second limit that requires the fuel-cladding gap to remain closed j during constant and increasing rod power operation under normal !
reactor operating conditions whenever rod internal pressures exceed j the r.ominal system pressure. The ENC design limits for fuel rod ;
pressure have been reviewed and accepted in the NRC review of !
XN-NF-82-06(P), Revision 1, and Supplements 2, 4, and 5 (Reference l 1
6).
The fuel rod internal pressure is primarily a function of the initial ,
pre-pressurization, fuel swelling, and fission gas release. ENC uses :
1 i a physically based model in R00EX2A for rod internal pressure !
j calculation with the conservative power histories described in l Section3.1(b)ofthisSER. As shown in Figures 3.17, 3.18, and {
l 3.19 of XN-NF-85-67(P), Revision 1, the calculated end-of-life rod l
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i pressures for both the 8x8 and 9x9 fue1 designs are well within the l design limit. The RODEX2A results used to generate Figures 3.17, P 3.18, and 3.19 also showed that the pellet to cladding gap closed i prior to the exposure at which rod pressure exceeded system pressure !
and remained closed at end-of-life. ENC JP-BWR reload fuel therefore i i meets the second design limit of a closed fuel-cladding gap. We [
conclude that the desig'n of ENC 8x8 and ENC 9x9 jet pump BWR fuel is acceptable with respect to internal rod pressure considerations.
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. i 13 (1) Assembly Liftoff The design criteria for fuel assembly hydraulic liftoff is that the i total worst case hydraulic loads, including buoyancy effects, causing an upward force on the fuel assembly and channel will not exceed the ;
weight of the fuel assembly and channel during normal operation and i anticipated operational cccurrences. This is consistent with the SRP f 1
guidelines and is therefore acceptable. ;
i As shown in Table 3.1 of XN-NF-85-67 (P), Revision 1, the weight of the fuel assembly and channel are well in excess of the total worst-case hydraulic loads, including buoyancy, causing a lift off ,
force. Therefore, we conclude that fuel assembly liftoff of the ENC JP-BWR reload. fuel will not occur during normal operation, including anticipated operational occurrences. ,
3.2 Fuel Rod Failure Criteria ,
i The NRC staff's evaluation of fuel rod failure thresholds of the failure l mechanisms listed in the SRP follows. When these failure thresholds are f applied to normal or transients operation, they are used as limits (and l 1
hence SAFDLs), since fuel failures under those conditions should not occur (according to the traditional. conservative interpretation of GDC 10). !
When these thresholds are applied to accident analyses, the number of fuel failures must be determined for input to the radiological dose l calculations required by 10 CFR 100. The application for these failure i thresholds is thus predetermined, and only the threshold values and design !
evaluations to demonstrate conformance to these limits are reviewed below, i
(a) Internal Hydriding l I
The absorption of hydrogen can result in premature cladding failure {
due to reduced ductility and fonnation [of hydride platelets, j l
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O 14 Hydriding as a cladding failure mechanism is precluded.by controlling the level of moisture and other hydrogenous impurities during !
fabrication. The fabrication limit for total hydrogen in ENC fuel pellets is more stringent than the ASTM limit cited in the SRP and is thus acceptable. Exxon has not reported significant fuel failures i due to hydriding. We conclude that reasonable assurance has been E provided that hydriding as a fuel failure mechanism will not be significant in the ENC JP-BWR reload fuel.
(b) Cladding Collapse If axial gaps in the fuel pellet column were to occur due to [
densification, the cladding would have the potential of collapsing -
into a gap (i.e., flattening). Because of the large local strains that would result from collapse, the cladding is assumed to fail.
ENC's design limit for cladding collapse has been reviewed and accepted in the review of XN-NF-82-06, Revision 1, and Supplements 2, >
4, and 5 (Reference 6).
ENC uses the approved COLAPX code (Reference 23) and RODEX2A to predict creep collapse. The COLAPX code calculates the geometry changes and creep deformation of the cladding a: a function of time.
Cladding creepdown was calculated with the RODEX2A code. As shown in Table 3.1 of XN-NF-85-67(P), Revision 1, ENC JP-BWR reload fuel does not exceed the design limit for cladding creep collapse for the most severe collapse condition. Thus, creep collapse is not expected to occur during the fuel rod lifetime. We conclude that the design is acceptable with respect to cladding collapse considerations.
(c) Overheating of Cladding As indicated in the SRP Section 4.2.II.A.2(d), it has been traditional practice to assume that failures will occur if the -
thermal margin criterion is violated. For BWR fuel, thermal margin is stated in terms of the minimum value of the critical power ratio 1
e '
15 (CPR) for the most limiting fuel assembly in the core. The design limit for ENC BWR fuel to prevent cladding overheating is that transition, boiling shall be prevented. This satisfies the-intent of MCPR criterion in the SRP and is thus acceptable. The review of thermal-hydraulic design methods is beyond the scope of this safety evaluation. The ENC XN-3 critical power correlation is discussed in XN-NF-512(P) (Reference 24).
I (d) Overheating of Fuel Pellets As indicated in the SRP Section 4.2.II.A.2(e), it has been traditional practice to assume that failure will occur if fuel pellet centerline melting tekes place. Thus, as a second criterion to avoid cladding failure due to overheating, ENC avoids fuel pellet .
centerline melting during normal operation and A00s as a design basis. The design limit corresponding to this design basis is that the peak linear heat generation rate during normal operation and A00s does not result in fuel centerline melting, taking into account the effects of burnup and gadolinia content. We find this acceptable.
For normal operation, ENC used R0DEX2A to calculate fuel centerline temperatures with the conservative power histories shown in Figures 3.5, 3.6, and 3.7 of XN-NF-85-67(P), Revision 1. As shown in Figures 3.20, 3.21, and 3.22, the calculated centerline temperatures remained below the irradiated 002 melting point. For anticipated operational
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occurrences, ENC used RODEX2 (Reference 5) and RAMPEX (Reference 25) to evaluate fuel centerline temperatures during transients. The results of the analysis showed that fuel centerline melting would not occur at or below the LHGR curve in Figure 3.4 for both the 8x8 and 9x9 reload fuel designs. We note that the LHGR curve in Figure 3.4 has LHGR's at least 120% greater than the steady state LHGR limit curves of Figures 3.1, 3.2, and 3.3. The 120% overpower case represents a bounding condition for transients from 100% power. Slow I
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4 16 transients may reach 120% power before APRM scram prevents any further increase. Fast transients have rod temperature distributions or a heat flux less severe than the 120% steady state overpower case due to heat capacity effects.
Based on the above considerations, we conclude that there is reasonable assurance that fuel pellet centerline melting will not occur in ENC JP-BWR relcad fuel during normal operation and anticipated operational occurrences. ;
(e) Excesjive Fuel Enthalpy i
For a severe reactivity initiated accident (RIA) in a BWR at zero or i low power, fuel failure is assumed to cccur if the radially averaged f fuel rod enthalpy is greater than 170 cal /g at any axial location.
The 170 cal /g enthalpy criterion is primarily intended to address cladding overheating effects, but it also indirectly addresses !
pellet / cladding interactions of the type associated with severe RIAs. -
This is consistent with the SRP guidelines and is therefore acceptable. .
ENC performs a detailed analysis of the control red drop accident l using the methodology presented in the approved report XN-NF-80-19 l (P)(A), Volume 1 (Reference 14). The analysis of the control rod !
drop accident for zero-power core conditions is performed on a cycie specific basis to assure that the total enthalpy is below the design limit of 170 cal /g. For the full power reactivity initiated accidents in a BWR, ENC uses MCPR, as discussed in Section 3.2(c) of t this SER, to predict failures. We conclude that the failure ;
mechanism of excessive fuel enthalpy for JP-BWR fuel has been properly addressed.
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s 17 (f) Pellet / Cladding Interaction L
The design criteria for mitigating pellet / cladding interaction (PCI) fuel failures are: (1) cladding uniform strain shall not exceed 1%
during any anticipated operational occurrence and (2) the fuel centerline temperature must remain below the melting point of the fuel. This is consistent with the SRP guidelines and is therefore acceptable.
As discussed in Section 3.2 (d) of this SER, fuel centerline melting will not occur in JP-BWR reload fuel during normal operation and anticipated operational occurrences. Cladding strain was calculated witn the RAMPEX code (Reference 25) with input from RODEX2 (Reference 5). The results of the calculation showed that cladding strain did
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not exceed 1% at or below the LHGR curve in Figure 3.4 for both the 8x8 and 9x9 fuel designs. As discussed in Section 3.2 (d), the LHGR curve in Figure 3.4 has LHGR values at least 120% greater than the steady state LHGR limit curves of Figures 3.1, 3.2, and 3.3. As discussedinSection3.1ofXN-85-67(P),Rev.1,operatingexperience. l and tests of similar ENC fuel designs demonstrate acceptable PCI performance when operating within recommended ramp rates and power limits. We conclude that the ENC JP-BWR 8x8 and 9x9 fuel designs meet the design criteria related to PCI.
(g) Cladding Rupture Zircaloy cladding will burst (rupture) under certain combinations of temperature, heating rate, and differential pressure - conditions ,
that occur during a LOCA. While there are no specific design limits associated with cladding rupture, the requirements of Appendix K to 10 CFR Part 50 must be met as those requirements relate to the incidence of rupture during a LOCA; therefore, a rupture temperature correlation must be used in the LOCA ECCS analysis. Cladding rupture 4
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18 is described analytica'ly as a part of the ECCS evaluation model ;
(Reference 12). ENC uses the cladding swelling and rupture models !
described in XN-NF-82-07 (P), Revision 1 (Reference 26). The NRC has (
, reviewed XN-NF-82-07 (P), Revision 1, and concluded that the models are acceptable without condition for use in licensing LOCA analyses ,
(Reference 27). We thus conclude that the failure mechanism of ;
l cladding rupture for JP-BWR fuel has been properly addressed. [
i (h) Fuel Rod Mechanical Fracturing 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 criteria for ENC ,
JP-BWR reload fuel for mechanical fracturing are presented in XN-NF-81-51 (P) (A) (Reference 28), which describes ENC's -
LOCA-seismic structural response analysis. The design basis is that the channeled fuel assemblies must withstand the external loads due to earthquakes and postulated pipe breaks without fracturing the fuel -
rod cladding. The design limit proposed by ENC is that the stresses due to postulated accidents in combination with the normal steady-state fuel rod stresses should not exceed the stress limits given in Table 3.1 of Reference 28. The stress allowables are !
derived from the ASME Boiler and Pressure Vessel Code,Section III, ;
Appendix F for faulted conditions. This design limit for mechanical fracturing has been reviewed and accepted in the NRC review of XN-NF-81-51 (P) (A) (Reference 28).
The mechanical fracturing analysis is done as a part of the seismic-and-LOCA loading analysis, which is described in the approved report XN-NF-81-51 (P) (A) (Reference 28) for the 8x8 fuel and in XN-NF-84-97 (P) (Reference 29) for the 9x9 fuel. As shown in Table 1.1 of Reference 29, the ENC JP-BWR 8x8 and 9x9 reload fuel designs meet the design limits describe'd above. The staff evaluation of F
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19 XN-NF-84-97(P) is documented in Reference 37. In addition, staff calculations indicate that the stresses calculated in the seismic and LOCA analysis in combination with the normal steady state fuel rod stresses meet the SRP guideline limit of 90% of the irradiated yield stress at the appropriate temperature. We thus conclude that the ENC JP-BWR 8x8 and 9x9 reload fuel designs meet the design criteria related to fuel' rod mechanical fracturing.
3.3 Fuel Coolability Criteria For major accidents in which severe damage might occur, core coolability must be maintained as required by Appendix A to Part 50 (e.g., GDC 27 and 35). The following paragraphs discuss the staff's evaluation of limits that will assure that coolability is maintained for the severe damage mechanisms listed in Section 4.2 of the SRP, and discuss the staff's evaluation of the conformance of the ENC JP-BWR reload fuel to the design criteria.
(a) Fragmentation of Embrittled Cladding ,
b The ENC design basis for ECCS evaluation (Reference 13) meets the requirements of 10 CFR 50.46 as it relates to cladding embrittlement for a LOCA; i.e., the acceptance criteria of 2200 F on peak cladding temperature and 17 percent on maximum cladding oxidation must be satisfied. The overall effects of cladding embrittlement on the ENC JP-BWR reload fuel design for the loss-of-coolant accident are .
discussed in the approved report XN-CC-33 (A), Revision 1 (Reference 12), and are not reviewed further here.
(b) Violent Expulsion of Fuel In a severe reactivity initiated accident (RIA) such as a BWR control rod drop, the large and rapid deposition of energy in the fuel can result in fuel melting, fragmentation, and violent dispe.rsal of fuel
s e ,
20 droplets or fragments into the primary coolant. The mechanical action associated with such fuel dispersal can be sufficient to destroy the cladding and rod-bundle geometry of the fuel and to produce pressure pulses in the primary system. To meet the guidelines of the SRP as it relates to the prevention of widespread fragmentation and dispersal of fuel and the avoidance of pressure pulse generation within the reactor vessel, a radially averaged enthalpy limit of 280 cal /g should be observed. ENC uses this criterion in the generic topical report, Volume 1 of XN-NF-80-19 (P)
(Reference 14), that presents ENC's rod-drop accident analysis methodology. ENC calculates a maximum radially averaged fuel enthalpy for the control rod drop accident for each cycle in which ENC fuel is used to assure that this enthalpy value is well below the 280 cal /g limit. We thus conclude that the fuel coolability criterion of violent expulsion of fuel has been properly addressed.
(c) Cladding Ballooning Zircaloy cladding will balloon (swell) under certain combinations of temperature, heating rate, and stress during a LOCA. While Appendix K to 10 CFR 50 requires that the degree of swelling during a LOCA not be underestimated, there are no design limits required for cladding swelling. ENC uses the cladding swelling and rupture models '
described in XN-NF-82-07 (P), Revision 1 (Reference 26). These approved models have been incorporated into ENC's ECCS evaluation ,
model (Reference 12). We thus conclude that the issue of fuel rod ballooning has been properly addressed.
(d) Structural Deformation from External Forces Earthquakes and postulated pipe breaks in the reactor coolant system would result in external forces on the fuel assembly. SRP Section 4.2 and associated Appendix A state that fuel system coolability should be maintained and that damage should not be so severe as to ;
J i
21 prevent control rod insertion when required during these low probability accidents. The ENC fuel assembly design basis for earthquakes and postulated pipe breaks is that the fuel assembly shall maintain a coolable geometry and retain control rod insertability during the occurrence of the design basis seismic /LOCA event. This is consistent with the SRP guidelines and is therefore acceptable.
ENC has performed representative analyses following the guidelines of SRP 4.2, Appendix A, and comparing the results to the acceptance criteria of Appendix A. These analyses were performed for the Dresden 3 reactor and are documented in the approved report XN-NF-81-51 (P) (A) (Reference 28) for the 8x8 fuel and in l
XN-NF-84-97(P) (Reference 29) for the 9x9 fuel. The staff evaluation of XN-NF-84-97(P) is documented in Reference 37. Conformance to the ;
acceptance criteria of SRP 4.2, Appendix A, can be demonstrated by l referencing XN-NF-81-51 (P) (A) or XN-NF-84-97 (P), as appropriate, and submitting justification that the analyses presented in the above topical reports bounds the particular application under review.
4.0 DESCRIPTION
AND DESIGN DRAWINGS The descriptien and design drawings of major fuel assembly components, including fuel rods, water rods, tie rods, upper and lower tie plates, spacer grids, compression springs, retaining springs, locking sleeves, and adjusting nuts are provided in Section 2 and Appendix A of XN-NF-85-67, Revision 1. In addition, design specifications are provided in Table 2.1.
The material properties of cladding and fuel are provided in Sections 4.1 and 4.2, resp:ctively, of XN-NF-85-39 (P) (Reference 2). While each parameter listed in SRP Section 4.2.II.B is not provided, enough information is available in sufficient detail to furnish a reasonably accurate representation of fuel design, and this information satisfies the intent of the SRP guidelines. ,
22 5.0 . TESTING, INSPECTION, AND SURVEILLANCE PLANS .
5.1 Testing and Inspection of New Fuel As described in SRP Section 4.2, testing and inspection plans for new fuel should include verification of significant fuel design paraneters. While details of the manufacturer's testing and inspection programs should be documented in quality control repor ts, the programs for onsite inspection (
of new fuel and control assemblies after they have been delivered to the plant should also be described in the SAR.
A discussion of the Exxon quality control program for the JP-BWR fuel is provided in XN-NF-IA (Reference 30), which addresses fuel system component parts, fuel pellets, rods and assemblies, and process control. Fuel system inspections vary for the different component parts and may include dimensions, visual appearance, audits of test reports, material :
certification, and non-destructive examinations. Pellet inspection is performed for dimensional characteristics such as diameter, density.
- length, and squareness of ends. Fuel rods, water rods, upper and lower tie plates, and spacer grid inspections consist of non-destructive ,
examination techniques such as leak testing, weld inspection and '
i dimensional measurements. Process control procedures are described in l detail. In addition, for any tests and inspections performed by other vendors on behalf of Exxon, Exxon reviews the quality control procedures and inspection plans to ensure that they are equivalent to those described I in XN-NF-1A and are perfo'medr properly te meet all Exxon requirements. We conclude that the ENC program is cautist 9t with the SRP guidelines.
t 5.2 On-Line Fuel System Monitoring r
~
Routine on-line fuel rod failure monitoring is a matter that would be arranged with the licensees. It is not addressed in XN-NF-85-67 (F), t Revision 1.
23 5.3 Post-Irradiation Surveillance Routine poolside inspection of some discharged fuel assemblies is a matter i that is normally arranged with the licensees. However, special surveillance related to the introduction of ENC JP-BWR reload fuel has ;
been arranged by ENC in order to assess, monitor, and confirm fuel performance. The surveillance program of selected 8x8 and 9x9 fuel assemblies at Dresden 2, Dresden 3, and KRB IIC includes profilometry; dimensional measurements and visual examination. In addition, Exxcn has ongoing testing and surveillance programs at Big Rock Point, Oyster Creek, '
and Barsebeck-1 for ENC BWR fuel designed and fabricated to nearly the same design bases as the ENC JP-BWR reload fuel. These programs consist of extensive dimensional inspections as well'as visual examinations. The ,
results from past inspections are presented in References 6, 7, 31, 32, 33, and 34.
The proposed post-irradiation surveillance program is consistent with the SRP guidelines for introduction of a new fuel design and is, therefore, acceptable.
6.0 EVALUATION FINDINGS The ENC JP-BWR reload 8x8 and 9x9 fuel designs described in XN-NF-85-67 (P), Revision 1, have been reviewed in accordance with Section 4.2 of the i
Standard Review Plan (NUREG-0800, July, 1981). The staff concludes that this. report may be referenced as an acceptable mechanical design of the 1 ENC,8x8 and 9x9 fuel for application to jet pump BWR reload cores with the conditions which follow:
- 1. The licensee proposing to use this fuel must assure that thd local power -limits provided in tFE plant technical specifications ara ,
within the bounds of the I HGR limit curves shown in Figures 3.1, 3.2, and 3.3 of XN-NF-85-67 (P), Revision l'.
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- 2. The approval based on review of this report is limited to previously approved burnup levels for BWR fuel (30,000 MWD /MTU batch average T exposure). Staff approval for operation to extended burnup levels is contingent on the generic approval of the method by which burnup is '
considered in the design and analytical processes as described in XN-NF-82-06 (P) and Supplements 1, 2, 4, and 5. ;
- 3. ENC has oemonstrated conformance to acceptable design limits for fuel rod bowing to exposures of 30,000 MWD /MTU for the 8x8 fuel and 23,000 -
MWD /MTV for the 9x9 fuel. Additional justification with regard to fuel rod bowing must be provided for operation bevond these burnup values.
On the basis of this review, we conclude that, with the above provisions, ,
XN-NF-85-67 (P), Revision 1, is acceptable for referencing in fuel reload licensing applications which use the ENC JP-BWR 8x8 and 9x9 reload fuel designs.
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.O REFERENCES
- 1. Letter from J. C. Chandler of the Exxon Nuclear Company to C. O. Thomas of
- the NRC,
Subject:
NF-NF-85-67 (P), Revision 0, " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel," dated July 18, 1985.
- 2. Letter from G. N. Ward of the Exxon Nuclear Company to G. Suh of the NRC,
Subject:
XN-NF-85-39 (P), "Sumary of ENC Mechanical Design Criteria, j Failure Mechanisms, and Material Properties for BWR Fuel Assemblies (January,1986)," dated February 24, 1986. >
- 3. Letter from J. C. Chandler of the Exxon Nuclear Company to C. O. Thomas of the NRC, Subjecti' XN-NF-81-21 (P) (A), Revision 1, " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel (September,1982)," -
dated September 29, 1982.
- 4. Letter from G. N. Ward of the Exxon Nuclear Company to C. O. Thomas of the NRC,
Subject:
XN-NF-85-74 (P), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical ,
Evaluation Model (November, 1985)," dated November 5, 1985.
- 5. XN-NF-81-58 (P) (A), Supplements 1 and 2, Revision 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company,
~
March 9, 1984.
- 6. XN-NF-82-06 (P), Revision 1, " Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company, June, 1982, and Supplement 2, dated June 5,1984, and Supplement 4, dated November,1985, and Supplement 5, dated November, 1985.
- 7. Letter from J. C. Chandler of the Exxon Nuclear Company to C. O. Thomas of the NRC,
Subject:
XN-NF-82-06 (P), Revision 1, Supplanent 1,
" Qualification of Exxon Nuclear Fuel for Extended Burnup, Supplement 1 Extended Burnup Qualification of ENC 9x9 BWR Fuel (April, 1984)," dated April 17, 1984.
- 8. Letter'from M. W. Hodges of the NRC to G. N. Ward of the Exxon Nuclear Company, dated April 8, 1986.
- 9. Letter from G. N. Ward of the Exxon Nuclear Company to G. C. Lainas of the NRC,
Subject:
Submittal of XN-NF-86-37 (P) :(sic, should be XN-NF-85-67 (P), Revision 1, " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel (April,1986)," dated April 21, 1986.
- 10. Letter from G. N. Ward of the Exxon Nuclear' Company to M. W. Hodges of the NRC,
Subject:
" Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel, "XN-NF-85-67(P), Rev. 1", dated June 23, 1986, w/ attachment 1, Responses to NRC questions.
,1 2 .
2
- 11. G. A. Reymann and D. L. Hagrman, "MATPRO: Version 11, A Handbook of Materials Properties for use in the Analysis of Light Water Reactor Fuel '
Rod Behavior," NUREG/CR-0497, TREE-1280, February, 1979.
- 12. XN-CC-33 (A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option User's Manual," Exxon Nucleir Company, November, 1975.
- 13. XN-NF-80-19 (P) (A), Volume 2, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors, EXEM: ECCS Evaluation, Summary Description," ,
Exxon Nuclear Company, June, 1981.
- 14. XN-NF-80-19 (P) (A), Volume 1, Supplements 1 and 2, " Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and -
Analysis," Exxon' Nuclear Company, March, 1983.
- 15. P. Shariffi and E. P. Popov, " Refined Finite Element Analysis of Elastic -
Plastic Thin Shells of Revolution," December, 1969.
- 16. S. Timoshenko, " Strength of Materials, Part 2," 1956.
- 17. J. F. Goodier, " Thermal Stress," Journal of Applied Mechanics, Volume 59, l March, 1937. l
- 18. S. Timoshenko and J. M. Gere, " Theory of Elastic Stability," McGraw-Hill, 1961.
- 19. R. J. Roark, " Formulas for Stress and Strain," McGraw-Hill, 1965.
- 20. M. P. Paidoussis and F. L. Sharp, "An Experimental Study of the Vibration of Flexible Cylinders Induced by Nominally Axial Flow," Transactions of American Nuclear Society,11(C),1968.
- 21. M. P. Paidoussis, "The Amplitude of Fluid Induced vibrations of Cylinders in Axial Flow," March,1965.
- 22. ANSYS Users Guide, Swanson Analysis Systems,'Inc., Houston, PA, 1979.
- 23. JN-72-23, Revision 1, " Cladding Collapse Calculational Procedure," Exxon Nuclear Company, November, 1972.
- 24. XN-NF-512 (F), Revision 1, "The XN-3 Critical Power Correlation," Exxon Nuclear Company, March, 1981.
- 25. XN-NF-573, "RAMPEX: Pellet - Clad Interaction Evaluation Code for Power Ramps," Exxon Nuclear Company, May, 1982.
w
.P ,
i' 3
- 26. XN-NF-82-07 (P), Revision 1, " Exxon Nuclear Company ECCS Swelling and Rupture Model," Exxon, Nuclear Company, August, 1982. j
- 27. Letter from C. O. Thomas of the NRC to G. F. Owsley of the Exxon Nuclear Company,
Subject:
Acceptance for Referencing of Topical Report XN-NF-82-07 (P), Revision 1, dated October 14, 1982.
- 28. XN-NF-81-51 (P) (A), "LOCA - Seismic Structural Response of -an Exxon Nuclear Company BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, May, 1986.
- 29. XN-NF-84-97 (P), "LOCA - Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, December,1984.
- 30. XN-NF-1A, " Quality Assurance Program Topical Report for Nuclear Fuel Design and Fabrication," Revision 3, Exxon Nuclear Company, August,1980.
- 31. XN-NF-77-49, "Non-Destructive Examinations of Exxon Nuclear Fuel at the Oyster Creek Reactor, Spring,1977," Exxon Nuclear Company, November 4, 1977. .
- 32. XN-NF-81-13, " Extended Burnup Demonstration Reactor Fuel Program; Poolside Fuel Examination; Big Rock Point Extended Burnup Fuel, February,1979,"
Exxon Nuclear Company, November, 1982.
- 33. XN-NF-84-131 (P), " Extended Burnup Demonstration Reactor Fuel Program Examination of Barsebeck-1 Fuel Assemblies Prior to Extended Burnup Cycle
- July,1984," Exxon Nuclear Company, January,1985.
- 34. XN-NF-86-32(P), Revision 1, " Examination of High Burnup Demonstration Fuel at Barsebeck-1," Exxon Nuclear Company, October, 1985.
- 35. Letter from G. C. Lainas of the NRC to G. N. Ward of the Exxon Nuclear Company,
Subject:
Acceptance for Referencing of Licensing Topical Report XN-NF-85-74, "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model,"
June 24, 1986.
- 36. Letter from E. Rossi of the NRC to G. N. Ward of the Exxon Nuclear Company,
Subject:
Acceptance for Referencing of Licensing Topical Report XN-NF-82-06(P), Revision 1, and Supplements 2,'4, and 5, July 18, 1986.
- 37. Letter from G. C. Lainas of the NRC to G. N. Ward of the Exxon Nuclear Company,
Subject:
Acceptance for Referencing of Licensing Topical Report XN-NF-84-97, dated August, 1986.
- 38. XN-NF-75-32 (P)(A),. Supplements I, 2, 3, and 4, " Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, October,1983.
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