ML19292F169
| ML19292F169 | |
| Person / Time | |
|---|---|
| Site: | Vogtle, 05000426, 05000000, 05000427 |
| Issue date: | 06/06/1973 |
| From: | Harold Denton US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML19292B772 | List:
|
| References | |
| FOIA-84-624 NUDOCS 8505280166 | |
| Download: ML19292F169 (3) | |
Text
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ENCLOSURE Safety Evaluation on Amendment 10 to GESTAR II (NEDE-24011-P)
1.0 INTRODUCTION
By letter dated November 30, 1984 (Reference 1) General Electric Company submitted Amendment 10 to GESTAR II. The submittal of Amendment 10 by General Electric introduces two new fuel designs designated GE8x8E and GE8xBEB. Both designs are intended for irradiation to extended burnup levels with the latter incorporating a thin zirconiunt barrier (liner) to the inner diameter of the cladding. This submittal relies en the criteria and methods submitted in Amendment 7 (Reference 2) and the GESTR-LOCA documentation (Reference 3) to r-svide assurance that these two new designs are safe for operation in boiling water reactors. Because the review of GE's extended burnup topical report (Reference 4) is not complete many of the design bases / criteria and analysis methods reviewed in this submittal have not been approved for application to extended burnup levels. The exception to this is the GESTR-LOCA code (Reference 3) which has been approved for extended burnup.
Consequently, the approval of the GE8x8E and GE8x8EB designs for operation to extended burnup levels is contingent on NRC approval of GE's extended burnup topical reprrt (Reference 4).
It should be noted, however, that the review of Reference 4 is nearly complete and there are no outstanding issues at this time.
Our review, documented in this safety evaluation report, covers the GE8x8E and GE8xBEB fuel designs to those burnups to which BWRs with GE fuel are presently operating.
This review and safety evaluation will follow the Standard Review Plan (SRP)
(Reference 5) to ensure that all licensing requirements cf the fuel system are addressed with respect to these new fuel designs. Amendment 7 to NEDE-24011 r-A (which was approved by the staff in References 6 and 7) has updated GESTAR II to follow the format of Section 4.2 of the Standard Review Plan and was usec extensively in this review.
8505280166 841015 PDR FOIA SHOLLYB4-624 PDR
4 Those portions of the review which address Section 4.2 of the Standard Review Plan are given in this Safety Evalua u on Report in Sections 2.0 thr9 ugh 7.0.
Section 8.0 addresses the nuclear design of these new fuel bundies.
Section 9.0 addresses the thermal hydraulic design of these fuel bundles. Our conclusions are given in Section 10.0.
2.0 FUEL SYSTEM DESIGN The GE8x8E and GE8x8EB fuel designs are modifications of the GE P8x8R and BP8x8R designs. The modifications are to aspects of the fuel rod design to enable the fuel bundles to attain higher burnups and also to the design of the bundle upper tie plate. The new extended burnup design features described in Amendment 10 include changes to the peak linear heat generation rate, plenum volume, helium fill gas pressures, fuel density, fuel / cladding gap, different enrichments, additional water rods end gadolinium rod changes. Some of tnese changes are reflected in Table 2-Ic and 2-Id and Figure 2-2 of Amendment 10.
The objectives of this fuel system safety review, as described in Secticn 4.2 of the SRP, are to provide assurance that as a result of the new design features (a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (A00s), (b) fuel system damage is never so severe as to prevent control roa insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained.
"Not damaged" 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. These objectives implement General Design Criterion (GDC) 10 of 10 CFR Part 50, Appendix A (" General Design Criteria for Nuclear Power Plants") and the design limits that accamplish this are called Specified Acceptable Fuei Design Limits (SAFDLs).
" Fuel rod failure" means that the fuel rod leaks ano that the first fission product barrier (the cladding) has, therefore, been breached.
Fuel rod failure must Se accounted for in the dose analysis to show compliance with the offsite dose limits of 10 CFR Part 100 ("Peactor Site Criteria") for postulated accidents.
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"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 after an accident. The general requirements to maintain control rod insertability and core coolability appear in the General Design Criteria (e.g.,
GDC 27 and 35). Specific coolability rcquirements for the loss-of-coolant accidents are given in 10 CFR Part 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors".
In order to meet the above stated objectives and follow the format of Section 4.2, this review covers the following three main categories:
(1) Fuel System Damage Mechanisms, which are most applicable to normal oneration and anticipated operational occurrences, (2) Fuel Rod Fai.sre Mechanisms, which apply to normal operation, anticipated operational occurrences and postulated accidents, and (3) Fuel Coolability, which applies to postulated accidents.
Under it.a heading for each SAFDL, there is a Bases / Criteria section and an Evaluation section. The criteria sections address the limiting values that have been submitted by General Electric in Amendment 7 under the three major categories of failure mechanisms listed above.
It is the purpose of this review to determine if these criteria are applicable and acceptable for the GE8x8E and GE8x8EB designs submitted in Amendment 10.
These criteria along with certain definitions for fuel failure constitute the SAFDLS required by GDC 10.
The evaluation sections review the methods that General Electric used to demonstrate that the design criteria have been met for these designs. These methods may include operational experience, prototype testing ai,d analysis models.
In addition, the new features of this design are reviewed with respect to each damage mechanism to determine if they could have an adverse impact on fuel system performance.
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3.0 FUEL SYSTEM DAMAGE The design criteria in this section should not be exceeded during normal operation including anticipated operational occurrences (A00s). The evaluation portion for each damage mechanism demonstrates that the design criteria are not exceeded during normal operation including A00s.
(a) Stre:s and Strain Bases / Criteria - In keeping with the GDC 10 SAFDLs, fuel damage criteria should assure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analysis.
GE's design basis for the stress and strain of fuel assembly components is that the fuel will not fail due to stresses or strains exceeding the fuel assembly component mechanical capability.
GE employs the concept cf a design ratio that is defined at a ratio of effective stress or strain to a stress or strain limit. The stress and strain limits are conservative estimates of the ultimate tens.le stress and the corresponding strain. The effective stress or strain is calculated using von Mises' criterion. The design ratio is limited to less than or equal to unity for design purposes. This design ratio is derived from ANSI /ANS-57.5-1981 (Reference 8), which has some variations from ASME Code Secticn III which is referenced in the Standard Review Plan.
For example, while ANSI /ANS-57.5-1981 uses a full ultimate tensile stress, the ASME Code Section III (Reference 9) calls for only 70% of the same quantity.
GE has demonstrated, in response to our questions (Reference 10) during the review of Amendment 7, that a conservative approach has been developed in
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calculating the design ratios. GE performs a Monte Carlo statistical analysis which results in a design ratio distribution for either stress or strain depending on whether yielding has occurred.
In order to satisfy the design criterion, GE requires that the upper 95th percentile of the stress or strain distribution be..less than unity. The NRC recently approved this new design criterion as a result of the Amendment 7 review (Reference 6). GE has t
indicated that the same materials and material fabrication procedures are used for the GE8x8E and GE8x8EB assemblies as those used for past 8x8 designs.
Consecuently, this criterion is also found to be acceptable for the GE8x8E and GE8x8EB designs.
The effects of extended butnup operation on cladding ductility and how this may affect the above criterion will be addressed in the review of Reference 4.
Evaluation-GE has determined the design ratio distributions for the GE8x8E and GE8x8EB designs, and has stated (Reference 1) that the design criterion is met. Therefore, the new fuel designs are acceptable with respect to stress and strain limits.
(b) Strain Faticue Bases / Criteria - GE's design basis for strain fatigue is that "the fuel assembly and the fuel red cladding are evaluated to ensure that strain due to cyclic loadings will not exceed the fatigue capability." A fatigue usage limit of 1.0 is used to assure that the fatigue capability is not exceeded. The fatigue usage is defined as the ratio of the actual number of cycles at stress and the resulting strain to the allowable number of cycles at stress and strain. A fatigue curve in terms of strain amplitude and allowable number of cycles is given in Amendment 7 (Reference 2). The GESTR-MECHANICAL code (Reference 11) along with Monte Carlo error propagation of the input parameters are used to predict a distribution of fatigue usage (similar to the distributions calculated for the str'ess and strain ratios). The upper 95th percentile of the distribution of fatigue usage is required to be less than 1.0.
This methodology accounts for internal and external pressures, cladding temperatures and pellet-cladding contact. The use of this fatigue usage limit and method has been found to be conservative and acceptable by the NRC (References 6, 7, and 12). We also find the limit and method to be applicable to the GE8x8E and GEBxBEB designs.
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Evaluation -
GE has determined (Reference 1) that the 95% upper tolerance limit of the fatigue usage distributions for the GE8x8E and GE8x8EB designs meet the above criterion and thus these fuel designs are acceptable.
(c) Fretting Wear Bases / Criteria - In our Safety Evaluation Report on Amendment 7 to GESTAR II (Reference 6) which dealt with fuel criteria, we stated that instead of provid-ing a limit on fretting wear, GE considers the e#fect of fretting wear in design analysis based on testing and experience in reactor operations.
Since the SRP does not provide numerical acceptance criteria for fretting wear, and since fretting wear is addressed in the design analysis for each bundle design, the NRC staff cnncluded in Reference 6 that the intent of the SRP has been adequately met.
Evaluation - In the current GE EWR P8x8R and PP8x8R fuel design individual rods in the fuel assembly are held in position by spacers located at intervals along the length of the fuel rod, and springs are provided in each spacer cell so that the fuel rod is restrained to avoid excessive vibration. The same design is employed by GE in the GE8x8E and GE8x8EB fuel designs. Various in-pile and out-of-pile tests were described in Section 2.6.3 of NEDE-24011-P-A-6 along with the results of a continuing fuel surveillance program that had utilized nondestructive methods including eddy current measurements to locate discontinuities in the cladding and detailed visual examinations to characterize the nature of defects. As stated in NEDE-24011-P-A-6, no significant fretting wear had been observed.
GE concluded that significant fretting wear was avoided through the use of an active spring force to eliminate any clearance that would otherwise exist between the spacer structure and the fuel rods. Based on the previous good experience of BWR fuel with respect to fretting weer, and the fact that the Amendment 10 fuel designs are identical with respect to those parameters important to fretting wear (such as spring force and cladding and grid dimensions) we find the GE8x8E and GE8x8EB designs acceptable with respect to fretting wear up to current BWR burnups. An 6
evaluation of fretting wear at high burnup will be addressed in our safety evaluation on Reference 4.
In addition to the fuel rods and spacers, there is another fual system component whose functionality must be assured as an objective of the review of BWR fuel system fretting concerns, viz., the fuel assembly channel box (References 13 and 14). Since the channel box design will not change with use of these new fuel designs, our evaluation, given in Reference 15, applies to the use of current channel box de.;igns with this new fuel.
(d) External Corrosion and Crud Buildup Bases / Criteria - The GE design bases for external cladding corrosion and crud buildup are to ensure that the cladding temperature increase and cladding metal thinning due to cladding oxidation and the cladding temperature increase due to the buildup of corrosion products, do not result in fuel rod failure due to reduced cladding strength. GE does not specify a limit for external corrosion or crud thickness, however, their effects are explicitly modeled in the thermal and mechanical analyse' as a part of the GESTR-LOCA (Reference 3) and GESTR-MECHANICAL (Reference 11) codes. Because the SRP does not provide numerical limits on cladding oxidation and since GE includes cladding oxidation and crud effects in their analyses, it is concluded that GE's approach is consistent with the SRP guidelines and applicable to the GE8x8E and GE8x8EB fuel designs.
Evaluation - It has been indicated that GE explicitly models the effects of cladding corrosion and crud in the GESTR-LOCA and GESTR-MECHANICAL codes, and thus, these effects are explicitly included in their thermal and mechanical analyses. The review of this methodology has been accepted in the safety evaluation (Reference 6) of Amendment 7 and tre application of the methodology to high burnup will be addressed further in the safety evaluation of GE's extended burnup topical report (Reference 4). We find the current GE methods applicable to the new fuel desiens for the burnup range of current EWR fuel.
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(e) Roa Bow and Channel Box Deflection Bases / Criteria - Fuel rod bowing is a phenomenon that alters the pitch of adjacent fuel rods and thus affects local nuclear power peaking and heat transfer to the coolant. GE's design besis for rod bowing is that the fuel rod is evaluated to ensure that rod bowing does not ruult in fuel failure due to boiling transition.
In our SER on GESTAR 11 Amendment 7 we stated that we could not find this criterion acceptable since GE had not presented sufficient evidence that there would not be a reduction in the critical power ratio for significant amounts of bowing. However, we also found no reason to change the existing position on the effect of fuel rod bowing on critical power ratio. This position is that no reduction in critical power ratio operating limits is reouired since the amount of rod bow observed in GE BWR fuel is small, however, any rod bo.. in excess of 50% gap closure should be reported. This position is also found applicable to the GE8x8E and GE8x8EB fuel designs.
The design bues for channel box deflectior, are described in the GE topical report net.. -
M -P. We found these bases acceptable in Reference 6.
GE has not proposed any change to channel box design in Amendment 10.
Evaluation-GE has submitted a generic topical report on fuel rod bowing (Reference 16). The NRC reviewed this report and performed an independent assessment of rod bow and concluded that significant rod bow is not expected in GE BWR designs (Reference 17).
During our review of Amendment 10, we requested GE to describe the possible effects that the higher enrichments and heat rating proposed for these two new fuel designs may have on fuel rod bowing and channel box deflections. GE's response (Reference 18) has indicated that no change in rod bowing or channel box deflection characteristics is expected. This is due to the fact that both axial and transverse fast flux / fluence distributions are not expected to change in the new fuel designs.
In addition, GE has calculated that the difference in 8
maximum cladding average temperature is less than 5*C between the new and previca designs. The effect of extended burnup on rod bowing will be discussed further in the safety evaluation of Reference 4; however, it is anticipated that no further restrictions with respect to GE rod bow will be required for extended burnup operation. Consequently, the current methodology and requirements including the 50% rod bowing reporting requirement for GE rod bow stated above are found to be applicable to the GE8x8E and GE8x8EB designs. These designs are therefore acceptable with respect to fuel rod bowing.
(f) Axial Growth Bases / Criteria - The differential irradiation growth rates of fuel rods and asseubly tie rods must be considered in GE assembly designs. An axial expension space exists between the upper end plug shoulder of each fuel rod and the upper tie plate for GE assembly designs.
Failure to maintain this expansion spacing can result in fuel rod bowing and possible rod failure. An expansit
.pring is positioned over the end plug shank and rests on the bottom of the upper tie plate.
The function of the spring is to keep the rod seated in the lower tie plate and allow independent axial expansion (due to irradiation induced axial expansion of the fuel rod) of the end plug shank intc the holes of the upper tie plate.
GE has not provided a design basis or limit for this expansion spacing in either Amendment 7 or Amendment 10; however, in response to Question 1.3 of this review GE has stated (Reference 18) that the expansion spacing is sized to provide a reasonable assurance that DMtoming out of the expansion spring will not occur. This criterion is found to meet the intent of the SRP.
Evaluation - Current irradiation experience with General Electric fuel indicates that axial gap closure is not a problem (Reference 30). GE has also indicated that no metallurgical changes to the fuel or tie rod cladding material have been 9
made so that the data are applicable to the GE8X8E and GE8X8EB designs. This methodology is found to be acceptable for these new designs. The applicability of these data and methods for extended burnups will be addressed in the review of Reference 4.
(g) Fuel Rod Pressures Bases / Criteria - The SRP identifies fuel rod pressure as a potential fuel damage mechar Ism separate from the stress and strain criteria already discussed in this review. This is because the criterion for fuel rod internal pressure involves more than the cladding mechanical limits. There are a number of reasons for this distinction:
(a) outward (tensile) cladding stresses may force analytical riethods into a mode for which they were not desicned or where greater uncertainties exist, (b) the higher releases of intent'ed fission gas from the fuel associated with higher rod pressures may lead to a level of un-desirable positive thermal feedback, (c) the higher releases of fission gas associated with higher rod pressures may lead to underestimating the radiological consequences of accidents using release assumptions of Regulatory Guides 1.25 and 1.77, and (d) net positive pressures could lead to ballooning during non-LOCA boiling transition event! and such behavior is not considered in the accident analyses.
In order to simplify the analysis of fuel system damage due to excessive rod internal pressure, the SRP states that rod internal pressures should remain below the nominal reactor coolant system (RCS) pressure during normal operation unless otherwise justified. GE has elected to justify limits other than those provided in the SRP.
In the Anendment 7 submittal GE has ptoposed that the rod pressure be limited so that the instantaneous cladding creepout rate due to
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internal rod pressure greater than RCS pressure is not expected to exceed the instanteneous fuel swelling rate, i.e., the fuel-to-cladding gap does not open.
GE has shown that this new criterion is acceptable with respect to items (a) through (d) described above and this has been found acceptable in Reference 6.
This criterior. is also found acceptable for the GEEX6E and GE8XEEB designs.
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Evaluation - The Amendment 7 submittal has proposed that the GESTR-MECHANICAL coce be used to determine that the abuve criterion is met. This has been found to be acceptable in Rcference 6 and is also found tc ce acceptable for the GE8XCE and GE8XSEB designs.
l The power history used as input for the rod internal pressure calcu ations is being r.ddressed in the safety evaluation of the extended burnup topical report (Reference 4).
In response to a question from the review of Reference 4 GE has indicated that the Maximum Linear Heat Generation Rate (MLHGR) limit is used for the fuel design analysis in question. This MLHGR limit represents the bounding power possible for a pcrticular design and burnup. Although this review is not complete this power history is bounding and thus cor.servative, and is acceptable in the range of burnups for current BWR fuel. GE has presented analysis results (Reference 19) based on calculations with the GESTR-PECHANICAL code and the MLHGR power history that show the GE8X8E and GE8X8EB designs are significantly below the above criterion.
(h)
Fuel Assembly Liftoff Bases / Criteria - The SRP calls for the fuel assembly holddown capability (wet weight and spring forces) to exceed worst-case hydraulic loads for normal operation including A00s. The Amendment 7 and 10 submittals have stated that the fuel assembly is evaluated to ensure that vertical liftoff forces are not sufficient to unseat the lower tieplate from the fuel support piece to such a degtee that the resulting loss of lateral fuel bundle positioning would interfere with control blade insertion. This amendment references an approved report (Reference 90) that describes a liftoff limit to prevent control blade interference. This limit is found to be applicable to the GE8X8E and GEBX8EB fuel designs and it is therefore c'
'uded that this design criterion is accept-able.
oa evaluation (Reference 21) that GE Evaluation - Amendment 10 refers - t states is applicab'ie to the GE8X8E and GEBX8EB for BWR/2, 3 and 4 plants which concludes that the fuel assemblies can withstand worst case LOCA and seismic 11
loccings without assembly lift as well as normal operation.
For EWR/5 and 6 designs, these evaluations are plant specific. The LOCA and seismic loads are found to bound possible loads from normal operation and A00s and thus these evaluations are found to be acceptable for the GEBX8E and GE8X8EB designs.
(i) Control Material Leaching This topic concerns control blades and is outside of the scope of this review.
The issue has been satisfactorily resolved. See Reference 6 for a discussion of this topic.
4.0 FUEL PCD FAILURE In the following paragraphs, GE fuel rod failure thresholds and analyses for the failure mechanisms listed in the Standard Review Plan and Amendment 7 are reviewed with respect to the GE8X8E and GE8X8EB designs submitted in Amendment 10. When the failure thresholds are applied to normal operation including anticipated operational occurrences, they are used as limits (and hence SAFDLs) since fuel failure under those conditions should not occur according to the traditional conservative inter-pretation of General Design Criterion 10. When these thresholds are used for postulated accidents, fuel failures are permitted, but the resulting radiological doses must be within the limits required by 10 CFR 100.
(a) Hydriding Bases / Criteria - Internal hydriding as a cladding failure mechanism is precluded by centrolling the level of hydrogen impurities during fabrication. To ensure that failure will not occur due to internal cladding hydriding, GE specifier, e fabrication limit on hydrogen content, whico is less than or equal to the limit stated in SRP Section 4.2.II.A.2 (a), i.e., 2 ppm hydrogen for e U07 pellet.
Since GE uses c hydrogen limit less than or ecual to that stated in the SPP we have concluded in Reference 6 that the design limit for hydriding is acceptable and this is also found to be accepteble for the GESX8E and GE8XEEB designs.
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Evaluation - In adcition to the manufacturing controls on moisture and hydrogen impurities GE has introduced a hydrogen getter in their fuel designs as an additional assurance against internal hydriding. GE in-reactor experience has shown that internal hydriding has not been an active failure mechanism for fuel manufactured since mid-1972. Based on this information we have concluded (Reference 6) that internal hydriding is not a problem for GE fuel and thin is also found applicable to the GE6X8E and GE8X8EB designs since the same design
.and manufacturing controls will apply.
Internal hydriding with respect to extended burnup operation will be discussed further in the safety evaluation of Heference 4.
(b) Clsdding Collapse Bases / Criteria - 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, collapted cladding is assumed to be failed.
In order to define a collapse criterion to reflect the operational conditions of the reactor, GE has adopted (Reference 22) a collapse criterion that is related to an assumed pressure increase during a turbine trip without bypass; that is, if the fuel rod can sustain, without collapse, an instantaneous increase in the hot system pressure of a given magnitude, it is considered safe against collapse during norma' operation, including A00s. The maximum ovality which precedes this collapse-safe transient is defined as the design limit ovality. The report (Reference 22) that contains these limits and definitions has been reviewed and approved by NRC (Reference 22). These limits are also applicable to the GE8X8E aM GE8X8EB designs.
Evaluation - GE has indicated (Reference 1) that cladding collapse is not calculated to occur for the GE8X8E and GE8X8EB designs using the approved models for cladding collapse (Reference 22).
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(c) Overheatinc cf Cladding Bases / Criteria - As indicated in SRP Section 4.2.II.A.?, it has been traditional practice to e n ume that failures will not occur if the thermal margin criterion is satisfied. This is a conservative assumption for events that cause failures as a result of high cladding temperatures.
For BWR fuel, the thermal margin is stated in terms of the minimum value of the critical power ratio (MCPR), which corresponds to tha most limiting fuel assembly in the reactor core. As indicated in Section 5.2.; of Amendment 7, GE ensures that adequate thermal margin is maintained by selecting an MCPR based on a statistical ecalysis as follows:
" Moderate frequency transients caused by a single operator error or equipment malfunction shall be limited such that, censidering uncertainties in manufacturing and monitoring the core operating state, more than 99.9 percent of the fuel rods would be expected to avoid beiling transition."
Both the normal operation and transient thermal limits i' terms of PCPR are derived from this ar, roach, which is described fully in NEDE-10958-P-A (Reference 231 and NED0-10958-A. These design limits are consistent with the thermal margin guidelines of SRP Section 4.2.II.A.2 and thus have been found acceptable by the NRC (Reference 6). These design limits are also found to be applicable and acceptable to the GE8X8E and GE8X8EB designs.
Evaluation - See Section 9.0 for a discussion of the applicability of GE's thermal margin calculations to GE8X8E and GE8X8EB fuel.
(d) 0_verheating of Fuel Pellets Bases / Criteria - GE presented the bases and criteria for overheating of fuel pellets in Amendment 7 to GESTAR II (Reference 2). We approved these bases ard criteria in our SER on Amendment 7 (Reference 6).
GE's design basis for 14
fuel pellet everheating is that the fuel rod is evaluated to ensure that fuel rod failure due to fuel melting will not occur. To achieve the design basis, GE limits the fuel rod to (1) no fuel melting during normal steady-state operation and whole core anticipated operational occurrences, and (2) a small amount of fuel melting but not exceeding 1% cladding strain fo local anti-cipated operational occurrences such as the rod withdrawal error.
Evaluat.on - GE has stated that the above bases and criteria are met for the GE8x8E and GE8x8EB fuel. The GESTR-MECHANICAL code is used to perform the analyses. We find this acceptable.
Further discussion of overheating of pellets at extended burnup will be discussed in our review of Reference 4.
Gadolinia is mixed in with the UO in some fuel rods to act as a burnable poison 2
for reactivity control. This gadolinia lowers the melting point and thermal conductivity of the fuel.
GE currently limits the concentration of gadolinia to 61. We have agreed that this value may be increased up to 10% provided that GE carries out a successful test program to assure that the fuel rod analysis methods remain valid and that no unexpected phenomena occur for higher gadolina concentrations.
'.E has proposed and the staff has accepted (hference 24) a program to provide this verification.
(e) Excessive Fuel Enthalpy Bases / Criteria - For a severe reactivity initiated accident (RIA) in a SWR at tero or low power, fuel failure is assumed in the SRP to occur if the radially averaged fuel rod enthalpy is greater than 170 cal /g at any axial location. The 170 cal /g enthalpy criterion, developed from SPERT tests (Reference 25), is primarily intended to address cladding overheating effects, but it also indirectly addresses pellet / cladding interactions associated with severe RIAs. As indicated in Section S.2.5.1.6 of Amendment } and in Reference 26, GE uses 170 cal /g as a cladding failure threshold. This has been approved by the NRC (Ren rence 6). The applicability of this fuel enthalpy limit to 15
extended burnup fuel will be discussed in the review of Reference 4.
We find this criterion to be applicable to the GE8X8E and GE8X8EB fuel designs.
Evaluation - Bounding analyses of the Rod Drop Accident are reported in GESTAP.
II.
If these analyses are found not tc be bounc 'ng for operation during a certain cycle, a cycle-specific analysis is performed. The above criterion is applicable to the GE8x8E and GE8x8EB fuel designs.
(f) Pellet Claddina Interaction Bases / Criteria - As indicated in SRP rection 4.2.II.A.2.g, there are no specifically applicable NRC criteria for PCI failures. One design criterion used to limit failures is to restrict the cladding strain to less than 1%.
GE has stated (Peferences 1 and 2) that they employ a 1% circumferential plastic strain limit for their fuel designs during A00s and this has been found to be consistent with the SRP and thus acceptable by the NRC (Reference 6).
This strain limit is also judged to be acceptable for the GE8X8E and GE8XSEB designs.
Past operating experience has shown that the 1% cladding strain criterion is not totally effective in preventing PCI because a limit on the average strain does not prevent highly localized strains that can result in fuel failure. As a result of developmental investigations and feedback from production fuel experience, operating restrictions known as Preconditioning Interim Operating Management Recommendations (PCIOMRs) were issued by GE to the EWR operators (Reference 27). These restrictions have reduced the incidence of PCI failures and complement the 11 criterion. PCIOMRs have generally been effective in reducing PCI failures that result from operational power changes, but they are not intended to prevent PCI failures during unexpected transients and accidents.
A further discussion of PCI failures as a result of off-normal events is provided in the Amendment 7 review (Reference 6).
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Evalcation - The GESTR-MECHANICAL code is used by GE to determine that their fuel designs meet the above li cladding strain criterion. This code has been approved for this i.nalysis by the NRC (Reference 6). GE has stated (Reference
- 1) that the calculated plastic strains for the GE8X8E and GE8):SEB designs are less than the 1% strain criterion using the GESTR-MECHANICAL code at the maximum power and exposure conditions expected as a result of A00s.
(g) Claddina Ruoture Bases / Criteria - Zircaloy cladding will rupture (burst) under certain combinations of temperature, heating rate, and stress during a LOCA. There are no specific desien limits associa62d with cladding rupture other than the 10 CFR 50 Appendix K reouirement that the degree of swelling not be underestimated.
Evaluation - GE uses an empirical rupture-temperature correlation (Reference
- 28) which has been updated to include the data and models from Reference 29 as a part of the LOCA emergency core cooling system (ECCS) analysis. This correlation has been reviewed and approved in Reference 38.
The design changes made as a result of the GE8XEE and GE8X8EB designs have not changed the applicability f this rupture-temperature correlation and thus the correlation is approved for application with respect to these designs. The applicability of this correlation and other LOCA models to extended burnup fuel will be addressed in the review of Reference 4.
(h) Fuel Pod Mechanical Fracturing Bases / Criteria - The term " mechanical fracture" refers to a cladding defect that is caused by an externally applied force such as a hydraulic load or a load derived from core-plate motion. These loads are bounded by the loads of a LOCA and safe-shutdown earthquake (SSE), and the mechanical fracturing analysis 17
is usually done as a part of the seismic-and-LOCA loads analysis (see Section 5.0(d) of this SER).
The entire seismic-and-LOCA loads evaluation (including design limits) has been described by GE in a topical report that has been approved by the NRC (Reference 20).
Evaluation - The discussion of the seismic-and-LOCA loading analysis is given in Section 5.0(d) of this SER.
5.0 FUEL C00 LABILITY For accidents in which severe fuel damage might occur, core coolability must be maintained as required by several General Design Criteria (e.g., GDC 27 and 35).
In the following paragraphs, limits and methods to assure that coolability is maintained for the severe damege mechanisms listed in the Standard Review Plan are reviewed.
(a) Fragmentation of Embrittled Cladding Bases / Criteria - The most severe occurrence of cladding oxidation and possible fragnientation during an accident results from a LOCA.
In order to limit the effects of claddir.g oxidation for a LOCA GE uses (References 28 and 31) acceptance criteria of 2200'F on peak cladding temperature and 17% on maximum cladding oxidation as prescribed by 10 CFR 50.46. These criteria will be applied to the GE8X8E and GE8X8EB designs.
Evaluation - Since General Electric's methods explicitly account for design c.ianges in the fuel, and the GESX8E and GE8X8EB fuel designs are within the ranges of all the empirical relationships of the GE LOCA calculational methods, we find the GE LOCA methods acceptable for use with these new fuel designs.
Results of these analyses are reportcd on a plant-specific basis.
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(b) Violent Expulsion of Fuel Bases / Criteria - 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 dispersal of fuel 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 predt e 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. As indicated in Peferences 26 and 31, GE employs this 280 cal /g criterion as a control rod drop accident design limit and thus is consistent with the SRP. The applicability of this limit to extended burnup fuel will be addressed in the review of Reference 4.
Evaluation-Bounding analyses of the Rod Drop Accident are reported in GESTAR II.
These analyses employ methods which are applicable to the GE8x8E and GE8x8EB fuel designs.
If these analyses are not found to be bounding for operation during a certain cycle, a cycle specific analysis is performed.
The above criterion is used in all cases.
(c)
Cladding Ballooning Bases / Criteria - Zircaloy cladting 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 10 CFR 50 Appendix K requirement that the degree of swelling not be underestimated.
Evaluation - The GE cladding ballooning model for ECCS analysis is addressed in the approved report NEDE-20566-P (Reference 28). This report adopted the NUREG-0630 date base and modesing, which specifies a method acceptable to the NRC for treating cladding swelling and rupture during a LOCA. Amendments 7 and 19
10 refer to Section S.2.5.2.1.4 in NEDE-24011-P-A-6 (Reference 31) which references this report.
The design changes made as a result of the GE8x8E and GE8x8EB designs have not reduced the applicability of these me+ hods and thus are approved for application with respect to these designs. The application of the methods and models used in the LOCA analysis for extended burnup fuel will be addressed in the review of Reference 4.
(d) Fuel Assembly ~+.ructural Damage From External Forces Bases / Criteria - 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 prevent control rod insertion when required during these low probability accidents. The SPP recommends acceptance criteria to achieve these objectives.
The entire seismic-and-LOCA loads evaluation (including design limits) has been described by GE in an approved topical report (Reference 20) to which Amendments 7 and 10 make reference. These design limits are found to be applicable to the GE8x8E and GE8x8EB designs.
Evaluation - Generic analysis methods for performing combined seismic-and-LOCA v
loading analyses have been described by GE in an approved topical report (Reference 20). These analysis aethods not only include the fuel assembly structural response but also fuel rod cladding loads and assembly lift-off forces for the conbined seismic-and-LOCA events as prescribed by Appendix A to SRP 4.2.
This analysis is plant specific because it requires site specific input ground motions and thus cannot be completed in a generic tanner. Therefore, an applicant for an operating license proposing to reference the subject document (NEDE-24011-P, fcendment 10) must perform site specific analyses using Re'erence 20 analysis 20
methods in order to address the above criteria and Appendix A tc SRP Section 4.2 guidelines.
6.0 DESCRIPTION
AfiD DESIGN DRAWINGS The fuel assembly design descrip+ ton and drawings are provided in Section 2.1, Tables 2-la through d, and Figures 2-la and 2-2 of the Amendment 10 submittal.
This design description is considered to be minimal; however, through presentations made by GE (Reference 19) and responses (Reference 18) to NRC questions the Amendment 10 descriptions of the GE8x8E and GE8x8EB designs are found to be at:eptable.
7.0 TESTIf4G, INSPECTION, AND SURVEILLANCE PLANS 7.1 Testing and Inspection of New Fuel Section 2.3.1 in Amendment 7 to GESTAR II and NED0-11209-04A contain descriptions of (1) the type of quality control inspections performed during manufacturing and (2) the plan for on-site inspection and testing of new fuel a s s emb'. i cs. We concluded in Reference 6 that the new fuel testing and inspection program for GESTAR II is acceptable, and we find it to be applicable to the GE8x8E and GE8x8EB fuel designs.
7.2 On-Line Fuel System Monitorin. _
This topic is independent of the fuel design and was not addressed during this review.
7.3 Post-irradiation Surveillance In a June 27, 1984 letter (Reference 32) the staff accepted a GE proposal for post irradiation surveillance which, in part, pertained to new fuel designs.
This letter stated, in part:
"For new fuel designs, GE will conduct a general 21
visual examination of the exterior surfaces of a statistically meaningful number of fuel bundles upon discharge..... This examination will be conducted at two applications of the new design from the first year in which the design is introduced."
GE has confirmed that this surveillance program will be used for GE8x8E and GE8x8EB fuel designs.
8.0 Nuclear Design GESTAR II discusses the analytical methods used by General Electric Company to perform nuclear design and safety calculations. Since the new fuel bundles do not involve extending the ranges of the various nuclear parameters involved in core nuclear design and safety analysis, we find the use of GE's nuclear methods acceptable for the GE8x8E and GE8x8EB fuel c'esigns to the range of burnups currently used with GE fuel.
The use of nuclear analysis methods for extended burnup will be discussed in the staff SER on Reference 4.
9.0 Thermal-Hydraulic Evaluation 9.1 Applicability of the GEXL Correlation for GE 8X8E/GE 8X8EB Fuel Design In Section 5.2.1.1 of Amendment 10 (Reference 1), GE states that the GEXL correlation (Reference 23) and its uncertainty limits as previously approved by the NRC for the 8X8 (with one water rod) and 8X8R (with two water rods) fuel designs will be used to predict the critical power for the GE8X8E and GE8x8EB fuel designs. Since the GE8X8E and GE8X8EB fuel bundles will contain several features which could alter the bundle power distribution from that of the
~~
previous GE bundle designs for which GEXL was approved, the staff requested GE to provide justification for the continued use of GEXL for the new bundle designs.
In response, GE stated in Reference 18 that the major thermal-hydraulic differences be* ween the 8XER and the new GE fuel designs are (1) installation of two to six fuel rod size water rods replacing fuel rods and (2) use of an 22
improved upper tie plate design. fi also indicated that only additional water rods r:ay affect the use of the GE). correlation since the upper tie plate design is not included in the formulation of GEXL. We agree with this conclusion.
To confirm the validity of the continued use of GEXL for the GE8X8E and GE8XCEB fuei de igns, GE provided comparisons of GEXL predictions with data for 8x8 fuel bundles with four and sixteen water rods.
The staff reviewed the results of these data comparisons and agreed with GE's conclusion that use of GEXL is appropriate for the fuel bundle configura-tions for which GE presented data (four water rods) since the measured values of critical power are adequately predicted by GEXL for the data presented. However, no data comparisons for the fuel bundles with other numbers of water rods were presented. Since GE intends to apply GEXL to fuel bundles with up to eight water rods, the staff requested GE to justify that the effect of varices numbers of water rods on the critical power calculation can be adequately predicted by the GEXL correlation.
Since the effect of the number of water rods would be included in the GEXL R factor which accounts for fuel bundle local power distributions, we requested GE to submit data showing that, for different local power distributions but the same R factor, the same critical power would be obtained within an acceptable uncertainty band.
In response, GE showed in Reference 33 that (1) the experimental critical power data are within the measurement error bound for two and four water rod test bundles with approximately equal R factors, (2) GEXL compares well with an 8 x 8 fuel bundle configuration with 16 water rods, and that (3) the GEXL pre-diction is generally conservative as compered with this experimental critical power data (although conservatism in GEXL is not recuired).
In response to a staff question regarding the applicability of the uncertainties included in GETAB (Reference 18) for the GE8X8E and GE8X8EB fuel designs, GE stated that use of the uncertainties in GETAB will be directly applicable or conservative for the GE new fuel designs since (1) the methodology used to calculate the bundle pressure drop, which affects the bundle flow, remains unchanged, (2) the use of more accurate lattice physics methods and more accurate core instruments will result in smaller uncertainties for the R f actor and Traversing-Incore-Probe measurements, respectively, and (3) the higher enrichment of the GE8X8E and GE8X8EB fuel will 23
res11t in a more peaked power distribution, resulting in less fuel bundles being near limits and use of the more peaked power distribution will therefore result in a larger margin to the safety limit.
While these arguments appear plausible, GE has not presented any data or calculations to quantitatively support these positions. However, we agree with the GE conclusion that, based on the above arguments, the GE8X8E and GE8X8EB fuel bundles should not have higher uncertainties than the present GE fuel bundles. Therefore, the use of these arguments to justify that present GETAB uncertainties for the GE8X8E and GE8X8EB fuel designs is valid is acceptable but no quantitative credit may be taken for these differences.
The staff has reviewed the test data base (References 18 and 33) provided by GE to support the use of GEXL for determination of critical power for BWR cores incorporating the GE8X8E and GE8X8EB fuel designs. The staff finds that (1) the GEXL correlation adequately predicts critical power for the GE8x8E and GE8x8EB fuel designs, and (2) the use of uncertainties considered in GETAB is appropriate for the GE8X8E and GE8X8EB fuel designs. Therefore, we conclude that the GEXL correlation and uncertainties in GETAB are acceptable for the GE8x8E and GE8X8EB fuel designs.
9.2 LOCA Analysis of GE8X8E ar.d GE8X8EB Fuel Desians In Attachment I of Amendment 10 to NEDE-24011, GE states that the loss-of-coolant-accident (LOCA) analysis of GE8X8E/GE8X8EB will be performed using either the approved SAFER /GESTR code (Reference 3) or the current SAFE /REFLOOD procedure (Reference 28).
As for the use of the SAFE /REFLOOD procedure, the methodology used is generically applicable for the MAPLHGR limit determination, tet the staff was concerned that the effects of enhanced fission gas release at highe-burnup (i.e., greater than 20 MWD /kgU) were not adequately considered in the fuel performance model.
In response to this concern, GE requested (References 34 and 35) the credit for approved, but unapplied, ECCS evaluation model changes and calculated peak cledding temperature r,argin to avoid MAPLHGR penalties at higher burnups. This request 24
was approved (Reference 36) for the present GE fuel designs. The staff finds that this staff position is also applicable to the GE8X8E and GE8X8EB fuel designs since (1) a plant specific analysis for determination of MAPLHGR limits will be performed with approved methods for the specific cycle of operation and (2) GE (Reference 19) shows that the maximum fuel temperature for the GE8XSE and GE8X8EB fuel decreases with increasing exposures (for burnups greater than 30 MWD /kgU) and is less than that for the P8x8R end BP8x8R fuel designs.
9.3 Thermal Hydraulic Stability The staff has completed the generic review related to the thermal hydraulic stability for BWR cores.
In the evaluation report (Reference 37), the staff concludes that GE fuel reloads (including those with GE8X8E and GE8X8EB fuel) meet the stability criteria set forth in General Design Criteria 10 and 12 provided that the BWR has in place operating procedures and Technical Specifications which are consistent with the recommendations of CE SIL-380, to assure detection and suppression of global and local instabilities.
For the reload core without the appropriate Technical Specifications to monitor core stability, the current procedure of using the methods of NEDE 222//-P-1 to calculate a cycle specific decay ratio should be continued. The reload will be considered acceptable if the decay ratio is shown to be less than 0.80 for all possible operating conditions. BWE 2/3 type reactors using the approved GE fuel types (including GE8X8E and GE8X8EB) have been shown to ha"e adequate stability margins and, therefore, are acceptable and their r? load cycles are exempted from the current requirement to submit a cycle specific stability analysis to the NRC.
9.4 Upper Tie Plate The GE8X8E and GEBX8EB fuel designs are provided with a new upper tie plate with different flow characteristics than the upper tie plates for the P8x8R and BPBx8R designs. This will affect thermal margin and stability of the bundles, but since the difference in flew characteristics can be treated explicitly in calculations showing compliance with the pertinent SAFDL's; we find this change to be acceptable.
25
10.0 Conclusions The GE8XSE and GE8X8EB fuel system designs as described in Amendment 10 of NEDE-24011-p have been reviewed in accordance with the SRP Section 4.2.
As a result of our review, we conclude that the use of the GE8X8E and GE8X8EB fuel designs is acceptable for all BWRs.
The issues to be addressed by applicants are:
(1) BWR/5 and 6 plant-specific analysis of combined seismic-arid-LOCA loading using the GE methods (Reference
- 20) approved by the NRC or another acceptable method to demonstrate conformance to the structural acceptance guidelines described in Appendix A to SRP 4.2 (see Sections 4.0(h) and 5.0 (d) of this report) and (2) an acceptable post irradiation surveillar;e must be provided or an endorsement of the approved GE fuel surveillance program (see Section 7.0 of this report).
Vith the above provisions, we conclude that the GE8X8E and GE8X8EB fuel systems described in Amendment 10 have been designed such that (1) they Ll not be damaged as a result of normal operation and anticipated operational occurrences, (2; fuel damage during postulated accidents would not be severe enough to prevent control rod insertion when it is required, and (3) core coolability will always be maintained even after postulated accidents. This conclusion is based on two primary factors:
(1) General Electric has provided sufficient evidence that the design objectives will be met based on opera".ing experience, prototype testing, and analytical predictions.
(2) General Electric has provided for te:,tir and inspection of new fuel to ensvee 2
that it is within design tolerances at the time of cora loading. The lice" ee will perform on-line fuel failure monitoring and post-irrt.diation surveillance to detect anomalies or confirm that the fuel has performed as expected.
(3) As stated in various parts of this safety evaluation, our approval for ex-tended burnup mus2 await completion of our review of NEDE-22148-P, ' Extended Burnup Methodology" (Reference 4) which we expe't to finish shortly. We find the GE8x8E and GE8x8EB acceptable for irradiation to present GE BWR burnup love,s.
26
1.
1.0 REFERENCES
1.
Letter from J. S. Charnley (GE) to C. O. Thomas (NRC) " Submittal of Proposed Amenament 10 to GE LTR NEDE-24011-P-A-6", November 30, 1984.
2.
Letter from J. S. Charnley (GE) to F. J. Miraglia (NRC), " Proposed Revisica to GE Licensing ispical Report NEDE-24011-P-A", February 25, 1983.
3.
S. O. AFerlund, et al.
7he GESTR-LOCA and SAFER Models for The Evaluation of the Loss-of-Coola'* 5.tident, Volume I: GESTAR-LOCA A Model for the
, Prediction of Fuel Ruu.i.ermal Performance, NEDE-23785-1-P (Proprietary),
General 9lectric Company, December 1981.
4.
Extended Burnup Eva?uation Methodology, General Electric Topical Report NEDE-22148, June 1982 (Proprietary).
5.
Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-- LWR Edition, NUREG/0B00, Section 4.2, " Fuel System Design,"
Rev. 2, July 1981.
6.
Letter, C. O. Thomas (f;RC) to J. S. Charnley (GE), " Acceptance for Ref-erencing of Licensing Topical Report NEDE-24011-P Amendment 7 to Revision 6, General Electric Standard Application for Reactor Fuel," dated March 1, 1985.
7.
Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011-P Amendment 7 to Revision 6. " General lectric Standard Application for Reector fuel; SER Page Changes for Clairi-fication ', May 9,19C5.
27
8.
" Light Water Reactors Fuel Assembly Mechanical Design and Evaluation,"
AhSI/ANS-57.5-1981, published by American Nuclear Society, Approved by American Nuclear Standards Institue, May 14, 1981.
9.
" Rules for Construction of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code,Section III, 1977.
- 10. Letter from J. S. Caarnley (GE) to C. O. Thomas (NRC) April 23, 1984.
- 11. "GESTR-MECHANICAL:
Fuel Property and Model Revisions," December 1984, MFN-170-84-0, Attachment 2 in a letter from J. S. Charnely (GE) to C. O. Thomas (NRC) dated December 14, 1984.
- 12. Letter from Thomas A. Ippolito, NRC, to Richard Gridley, GE, April 16, 1979.
- 13. Safety Analysis Report for Plant Modifications to Eliminate Sianificant In-Core Vibration, EED0-21084 (Proprietary), November 1975, Genral Electric Company.
- 14. Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration, NEDE-21156, (Pro,rietary) Jenuary 1976.
15.
D. G. Eisenhut (NRC) Memorandum for Karl Goller, " Modification to Eliminate Significant In-Core Vibration," March 2, 1976.
16.
R. J. Williams, Assessment of Fuel-Rod Bowina in General Electric Soiling Water Reactors, NEDE-24284-P, August 1980.
- 17. Letter from C. O. Thimas, NRC, to J. F. Quirk, GE, May 31, 1983.
- 18. Letter from J. S. Charnley, (GE) to C. O. Thomes, (NRC), " Response to Request Nu"ber 1 for Additional Information on NEDE-24011-P-A-6, Amendment 10," March 11, 1985.
28
- 19. Letter from J. S. Charnley, GE, to R. Lobel, NRC, " Presentation on GE8x8E and GE8x8EB Fuel Designs," November 14, 1984. (Proprietary)
- 20. BWR Fuel Assembly Evaluation of Combined Safe Shutdown Earttauake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings, NEDE-21175-3-P, Amendment 3, vu.ber 1984, General Electric Company.
- 21. Letter, R. L. Gridley (GE) to D. G. Eisenhut (NRC), " Evaluation of Poten-tial Fuel Bundle lift at Operating Reactors," dated July II, 1977.
- 22. Creep Collapse Analysis of BWR Fuel Using SAFE-COLAPS Model, NEDE-20606-P-A, August 1976. General Electric Company.
23.
" General Electric BWR Thermal Analysis Basis (GETAB):
Data, Correlation and Design Application," NEDE-10958-P-A, January 1977.
- 24. Letter from J. F. Quirk, (GE), to L. S. Rubenstein (NRC), " General Electric Company Confirmatory Program in Support of Commericial Application of Gadolnia Concentration Greater than Six Weight Percent", October 18, 1983 (MFN-193-83).
25.
J. G. Grund, et al., Subassembly Test Program Outline for FY 1969 and FY 1970, IN-1313, August 1969, Idaho National Engineering Laboratory.
26.
J. Poane, et al., Rod Drop Accident Analysis for Large BWRs, NEDE-10527, March 1972; Supplement 1, July 1972; Supplement 2, January 1973.
27.
H. E. Williamson, Interim Operating Management Recommendation for Fuel Preconditioning, NEDS-10456-PC, Rev. 1, June 1975, General Electric Company.
29
.s
- 28. General Electric Company Analytical Model for Loss-of-Coolant Anaylsis in Accordance with 10 CFR 50 Appendix K, NEDE-20566-P, Volume 1, January 1976, General Electric Company.
29.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," NUREG-0630, April 1980.
- 30. Letter from J. S. Charnley, (GE), to L. S. Rubenstein, (NRC), "1983 Fuel Experience Peport", Octeber 12,1983 (Proprietary).
- 31. General Electric Standard Application For Reactor Fuel, NEDE-24011-P-A-6, April 1983.
- 32. Letter from L. S. Fubenstein (NRC) to R. L. Gridley (GE), " Acceptance of GE Proposed Fuel Surveillance Program," June 27, 1984.
33.
Letter, J. S. Charnley, GE, to C. 3. Thomas, NRC, " Supplementary Information Regarding NEDE 24011-P-A, Amendment 10," May 2, 1985.
34.
Letter from R. Engel (GE) to T. Ippolito (NRC), dated May 6, 1981.
35.
I.etter from R. Engel (GE) to T. Ippolito (NRC), dated May 28, 1981.
- 36. Memorandum from L. S. Rubenstein (NRC) to T. Novak (NRC), " Extension of General Electric Emergency Core Cooling System Performance Limits," dated June 25, 1981.
- 37. Letter from L. S. Rubenstein to D. Crutchfield, " Safety Evaluation of GE Topical Report NEDE-24011 (GESTAR) Amendment 8", dated April 17, 1985.
- 38. Letter from H. Bernard, NRC, to G. G. Sherwood, GE, " Supplementary Accepter.ce of Licensing Topical Report NEDE 20566 A(P)," May 11, 1982.
30