ML20135G151

Topical Rept Evaluation of WCAP-10125, Extended Burnup Evaluation of Westinghouse Fuel. Rept Acceptable for Licensing Zircaloy-clad Fuel Designs Up to Requested Extended Burnup Level

ML20135G151
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Issue date:	05/31/1985
From:	Office of Nuclear Reactor Regulation
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EtlCLOSURE 4

SAFETY EVALUATION REPORT ON WESTINGHOUSE ELECTRIC CORPORATION EXTENDED BURNUP TOPICAL REPORT - WCAP-10125 (Proprietary) i May 1985 i

i Prepared by Core Performance Branch and Accident Evaluation Branch Division of Systems Integration Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission i

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85091802lO Y

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I Table of Contents

1.0 INTRODUCTION

2.0 FUEL SYSTEM DAMAGE 3.0 FUEL R00 FAILURE 4.0 FUEL C00 LABILITY 5.0 NUCLEAR DESIGN 6.0 RADIOLOGICAL CONSIDERATIONS OF POSTULATED ACCIDENTS 4

WITH EXTENDED BURNUP OPERATION 7.0 REGULATORY POSITION i

8.0 REFERENCES

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1.0 INTRODUCTION

i Economics and prudent utilization of resources has led utilities to seek more efficient use of current generation light water reactors (LWRs).

Improved fuel utilization is one of the avenues being pursued for greater efficiency.

One of the greater improvements in fuel utilization is to increase the fuel discharge exposure which is currently at batch average burnups of approximately 28 mwd /kgM for BWRs and approximately 33 mwd /kgM for PWRs to batch average burnups of approximately 40 mwd /kgM and 50 mwd /kgM or above, respectively.

In response to this trend for extended burnup fuel operation, the Nuclear Regulatory Connission (NRC) has requested each fuel vendor to prepare and submit a topical report for review and approval that covers extended burnup experience, methods and test data to provide a generic basis for operation at extended burnups (Reference 1).

Westinghouse Electric Corporation Westinghouse has submitted such a report (Reference 2) requesting generic licensing approval of their criteria and methods used for licensing their fuel designs, for application at extended burnups.

In addition, Westinghouse has also provided responses (References 3, 4, 5, 6) to NRC questions concerning this submittal.

This review r:onsidered only Zircalory-clad Westinghouse fuel and did not con-sider the effects of extended burnup on Westinghouse stainless steel clad fuel.

This technical review and evaluation has been performed by Pacific Northwest Laboratory (PNL) under contract (FIN B2533) with the United States NRC. The review has been based on References 2 through 6 and Section 4.2 of the Standard Review Plan (SRP) (Reference 7) and covers the fuel assembly, fuel rods, and burnable poison rods but does not include the rod cluster control assemblies I

for extended burnup operation.

I This report follows the intent of Section 4.2 of the SRP, where appropriate for a generic review, to insure that all licensing requirements of the fuel

system are reviewed with respect to extended burnup operation. The objective of Section 4.2 and this review is to provide assurance that, as a result of extended burnup operation, (a) the 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.

"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. This objective implements General Design Criterion (GDC) 10 of 10 CFR Part 50, Appendix A (" General Design Criteria for Nuclear Power Plants") 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) has, therefore, been breach:

Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (" Reactor Site Criteria") for postulated accidents.

"Coolability," which is sometimes tenned "coolable geometry," means, in general, that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channels to pennit removal of residual heat after a severe accident. The general requirements to maintain control rod insertability and core coolability appear repeatedly in the General Design Criteria ('DC e.g.,

G 27and35). Specific coolability requirements 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 categories:

(1) Fuel System Damage Mechanisms, which are most applicable to normal operation and anticipated i

operational occurrences, (2) Fuel Rod Failure Mechanisms, which apply to normal operation, anticipated operational occurrences and postulated accidents and; (3) Fuel Coolability, which apply to postulated accidents.

l Because the purposes of each vendor's high burnup topical report are slightly different, it is useful to quote Westinghouse's goal in preparing this report, as stated in Reference 2.

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"The purposes of this topical is to justify operation of West.f_nghouse designed fuel to [the target] lead fuel rod, average burnup..."

In addition, Westinghouse stated that "The information supplied in this report supports the conicusion that Westinghouse design methods and safety analyses are valid for operation to "the [ proprietary] lead rod average burnup target.

No performance limitations have been identified which would preclude the design of Westinghouse fuel to this target burnup, and it has been shown that current design and safety evaluation criteria can be applied with no modification to these criteria".

The criteria sections in this review address limiting values for fuel damage that are acceptable under the three major categories of failure mechanisms listed above and in the SRP. The purpose of this review is to determine if

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the Westinghouse criteria are applicable to extended burnup operation of their fuel. These criteria along with certain definitions for fuel failure constitute the SAFDLs required by GDC 10.

The evaluation sections review the methods that Westinghouse uses to demonstrate that the design criteria have been met for extended burnup operation and thus are reviewed with respect to their applicability to the proposed range of extended burnup operation. These methods and data may include operating experience, prototype testing and analytical techniques. The determination that specific Westinghouse designs meet the stated criteria is not addressed in this review but will be addressed in specific licensing applications.

Westinghouse uses the ANS classification of plant conditions which divides plant conditions into four categories in accordance with anticipated frequency l

of occurrence and potential radiological consequences to the public. The four categories are as follows:

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Condition I:

Normal Operation and Operational Transients 2.

Condition II:

Faults of Moderate Frequency 3.

Condition III:

Infrequent Faults 4.

Condition IV:

Limiting Faults l

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O This approach is used consistently in Westinghouse analyses and.<fil.be used in this safety evaluation report.

2.0 FUEL SYSTEM DAMAGE The design criteria in this section should not be exceeded during normal operation including anticipated operational occurrences (A00s). The evaluation portion of each damage mechanism demonstrates that the design criteria are not exceeded during normal operation and A00s.

(a)DesignStress Bases / Criteria - The Westinghouse design basis for fuel assembly, fuel rod, burnable poison rod, and upper end fitting spring stresses is that the fuel system will be functional and will not be damaged due to excessive stresses.

I The design limit for fuel rod cladding stress under Condition I and II modes of operation is that the volume averaged effective stress calculated with the von Mises equation, considering interference due to uniform cylindrical pellet-to-cladding contact (caused by pellet thermal expansion and swelling, uniform cladding creep, and fuel rod / coolant system pressure differences), is less than the Zircaloy 0.2 percent offset yield stress with consideration of temperature and irradiation effects as described in Reference 8.

This report has been approved by the NRC (Reference 9). This criterion is applicable to extended burnup operation.

Evaluation - The Performance-Analysis-and-Design (PAD) code Version 3.3 (Reference 10) is used by Westinghouse to assure that the above criterion is met. This code has been verified against fuel rod ' data with rod average burnups l

up to approximately 57 mwd /kgM. This code takes into account those parameters l

important for determining cladding stresses at extended burnups, such as pellet thermal expansion and swelling, cladding creep and fuel rod / coolant system pressure differences. The NRC has approved (Reference 11) the use of this code for licensing applications without burnup restrictions. Consequently, this code is found acceptable for detennining cladding stress on fuel rods with extended burnups up to those requested by Westinghouse in Reference 2.

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It is noted that cladding stresses and strains due to transients (Condition II events) at extended burnups are not expected to be limiting because the power capability is reduced due to fissile material burnout, limiting the power excursion and thus stresses experienced by these rods.

(b) Cladding Design Strain Bases / Criteria - The Westinghouse design basis for fuel rod cladding strain is that the fuel system will not be damaged due to excessive cladding strain.

In order to meet this design basis the Westinghouse design limit for cladding strain during steady-state operation is that the total plastic tensile creep and u'niform cylindrical fuel pellet expansion due to fuel swelling and thermal expansion is less than 1 percent from the unirradiated condition.

For Condition II transients, the design limit for cladding strain is that the total tensile strain due to uniform cylindrical pellet thermal expansion during the transient is less than 1 percent of the pretransient value. These design strain bases and limits have been presented previously by Westinghouse (Reference 12) approved by the NRC (Reference 13) for application to current burnup fuel.

The material property that could have a significant impact on the cladding strain limit at extended burnup levels is cladding ductility. The strain criterion could be impacted if cladding ductility were decreased, as a result of extended burnup operation, to a level that would allow cladding failure without the Condition I and II cladding strain criterion being exceeded in the Westinghouse analyses.

From examination of irradiated Zircaloy cladding ductility data (References 14, 15), it has been concluded that ductility decreases with increasing fluence at low burnup levels, i.e., less than 12 mwd /kgM, but asymptotically approaches either a constant value or a small fluence dependence beyond these low burnups. Consequently, cladding ductility has either little or no change for the increased burnup levels projected for Westinghouse extended burnup operation.

In addition, Westinghouse has irradiated experimental and lead test rods with average burnups up to approximately 67 mwd /kgM with no adverse effects in cladding ductility.

From the above, we can conclude that the strain limit proposed by Westinghouse is applicable for' extended burnup application.

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Evaluation - The NRC-approved Westinghouse fuel perfomance code, PAD 3.3 (Reference 10), is used to assure that Westinghouse fuel meets the above

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criterion. As noted in the Design Stress section, this code has been verified against fuel rod data with rod average burnups up to approximately 57 mwd /kgM and takes into account those parameters important for determining cladding stresses and strains at extended burnups.

Consequently, this code is found acceptable for detennining cladding strains on fuel rods with extended burnys up to those requested by Westinghouse (Reference 2).

(c) Strain Fatigue

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Bases / Criteria - The Westinghouse design basis for fuel rod cladding fatigue is that the fuel system will not be damaged due to cladding strain fatigue.

In order to assure that this design basis is met, Westinghouse imposes a design limit for strain fatigue such that the fatigue life usage factor is less than 1.0.

That is, for a given strain range, the number of strain fatigue cycles are less than those required for failure when a minimum safety factor of 2 on the stress amplitude or a minimum safety factor of 20 on the number of cycles, whichever is the more conservative, is imposed. This criterion is the same as that given in Section 4.2 of the Standard Review Plan.

As noted.in the Cladding Design Strain section, the material property that could have a significant effect on cladding strain and thus strain fatigue at extended burnups is cladding ductility. However, as discussed above, extended burnup operation has shown little or no observable effects on cladding ductility and performance.

From this,'it is concluded that extended burnup operation does not reduce the applicability of the fatigue limits and thus the Westinghouse criterion is found acceptable for use in extended burnup applications.

Evaluation - The NRC-approved Westinghouse fuel performance code, PAD 3.3, is used to determine the strain range for the fatigue usage analysis. The Langer O'Donnell fatigue model (Reference 16) with the empirical factors in this model modified in order to conservatively bound the Westinghouse test data, is used with the strains from PAD 3.3 to assure that the above criterion is met. A description of this methodology and the Westinghouse data base is presented in WCAP-9500 (Reference 12) which has been approved by the NRC (Reference 13).

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D This methodology takes into account daily load follow operation and the additional fatigue load cycles that may result from extended burnup operation'.- In addition, the PAD 3.3 code accounts for those parameters i' portant for detenning cladding m

strains at extended burnups, see Sections 2.0(a) and 2.0(b). Therefore, the above methodology is found to model operational and material behavior parameters important for determining strain fatigue at extended burnups and thus is accept-able for extended burnup application.

(d)

Fretting Wear Bases / Criteria - Fretting wear is a concern for fuel and burnable poison rods, and the Zircaloy guide tubes. Fretting, or wear, may occur on the fuel and/or burnable rod cladding surfaces in contact with the spacer grids if there is a reduction in grid spacing loads in combination with small amplitude, flow-induced, vibratory forces. Guide tube wear may result when there is flow induced moticn between the control rod ends and the inner wall of the guide tube.

While the Standard Review Plan (SRP), Section 4.2, (Reference 7) does not provide numerical bounding-value acceptance criteria for fretting wear, it does stipulate that the allowable fretting wear should be stated in the safety analysis report and that the stress / strain and fatigue limits should presume the existence of this wear.

The Westinghouse design basis for fuel rod fretting wear is that fuel rods shall be designed'not to fail due to fretting wear during Condition I and II events.

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In order to meet this basis, Westinghouse uses a general guide for wall thickness

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reduction which is a percent of the original wall thickness (the specific value l

is proprietary) for evaluating cladding imperfections, including wear marks.

t Westinghouse indicates that the cladding stress and fatigue limits, discussed here in Sections 2.0(a) and 2.0(c), apply to fretting wear. Westinghouse also indicates (Reference 2) that fretting wear will not have a significant effect on cladding stresses and thus need not be considered in stress related analyses.

We have confirmed that fretting wear effects on the stress analysis are in-significant as long as the fretting wear in Westinghouse designed rods remains below the general guide for cladding imperfections stated in terms of percent wall thickness in the extended burnup topical report (Reference 2). These design 7

bases and criteria are found to be acceptable for extended burnup application.

The Westinghouse design bases and criteria for guide thimble tube wear is that no perforation of the tube wall should occur and'that the integrity of the guide thimble tube be maintained throughout the nomal life of a fuel assembly. As an additional design limit on guide thimble tubes,' Westinghouse has determined (Reference 12) from stress analyses that the limiting load on the fuel assembly structure is that which might occur during a fuel handling accident. A design criterion of 6 g is used for the analysis of this accident and this has previously been approved by the NRC (Reference 13). These design bases and criteria are also found to be acceptable for extended burnup application.

Evalu~ation - Westinghouse utilizes empirical data taken from operating reactors and out-of-reactor wear tests to provide assurance that the above criteria are met for both Zircaloy and Inconel grid designs.

Fuel rod fretting is affected-by the increased fluence and in-reactor residence time associated with extended burnup. The increasad fluence results in a slight decrease in grid spring forces and the increased residence time may result in a small increase in wear volume.

Fretting type failures have been observed at the bottom (Inconel) grid location of several rods in one of the Westinghouse.14x14 0FA assemblies. Westinghouse has stated (Reference 3) that the cause of these failures was traced to non-standard installation of the rods in the assembly during fabrication, rather than to a generic problem in rod or grid design. To support this, Westinghouse has shown (Reference 3) that the remaining 14x14 and 17x17 0FA assemblies with assembly average burnups up to 39 mwd /kgM have shown no indication of fretting wear indicating that Zircaloy grid spring forces continue to preclude fretting.

l Out-of-reactor tests on 0FA assemblies have indicated (Reference 2) that fuel rod fretting wear will not be a limiting concern up to the extended burnup level requested. From this it is concluded that fuel rod fretting is.not expected to be a problem for 0FA fuel designs with Zircaloy grids; however, in order to confirm this conclusion, it is recomended that additional fuel rod fretting data be obtained on Zircaloy grid assemblies up to the extended burnup level requested by Westinghouse.

Westinghouse 15x15 and 17x17 assemblies using the Inconel grid design have been l

irradiated for five cycles of operation (assembly average and lead rod average 1

burnups of approximately 55 mwd /kgM and 60 mwd /kgM, respectively) and for four 8

i cycles of operation, respectively. ' Detailed. visual examinations of these assemblies have indicated no evidence of cladding fretting. Consekuently, Westinghouse fuel designs with Incenel grids are"found to be acceptable for extended burnup operation.

(e) Oxidation and Crud Buildup l

Bases / Criteria - The Westinghouse design basis for cladding oxidation is that the fuel system will not be damaged due to excessive cladding oxidation.

In order to preclude a condition of accelerated oxidation, Westinghouse imposes

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specific temperature limits on the cladding. The temperature limits applied to cladding oxidation are that calculated cladding temperatures (at the oxide-to-metal interface) shall be less than a specific (proprietary) value during steady-state operation, and for Condition II transients the metal-to-oxide interface shall not exceed a higher (proprietary) value. These criteria have been approved by NRC (Reference 13) for current burnup levels and are also found to be applicable to extended burnup operation.

Evaluation - The SRP states that the effects of cladding crud and oxidation need to be addressed in safety analyses, such as thermal and mechanical analyses.

The major means of controlling cladding crud and oxidation is through primary coolant chemistry controls; however, this does not eliminate the ' pad to include their effects in safety analyses at extended burnups.

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Westinghouse has presented (Reference 3) two sets of oxide thickness data: 1) those induced by thick crud deposits, along with a bounding curve for the data, and 2) nominal oxide and crud thickness data along with a bounding curve (labeled for this discussion as best estimate) for these data. Westinghouse has indicated that they have primary water chemistry controls that limit the amount of crud deposits'and only those plants that have operated outside of these chemistry l

controls have been observed to have the thick crud deposits and abnomally high l

oxide thicknesses. This is consistent with past industry experience.

Westinghouse has indicated that a best estimate (proprietary) value of crud is input to PAD 3.3 and PAD 3.3, Addenda 2 and the best estimate bounding curve (from Figure 1 of Reference 3) for cladding oxide thickness is modeled in the PAD 3.3 and PAD 3.3 Addenda 2 codes. This is found to be acceptable for thermal l

l evaluations of extended burnup fuel because Westinghouse imposes water chemistry controls on their plants to maintain crud and oxide thicknesses to nominal values up to the extended burnup range requested.

For mechanical analyses Westinghouse has indicated that they reduce their cladding wall thickness by a specified (proprietary) amount to account for cladding defects and cladding oxidation. This amount is found to more than bound the cladding thickness reduction due to cladding oxidation at the extended burnups requested. Therefore, this methodology is acceptable for extended burnup j

application.

I (f) Rod Bowing Bases / Criteria - Fuel and burnable poison rod bowing are phenomena that alter the design-pitch dimensions between adjacent. rods. Bowing affects local nuclear 4

power peaking and the local heat ransfer to the coolant. Rather than piacing design limits on the amount of bowing that is permiteed, the effects of bowing are included in the safety analysis. This is consistent with the Standard Review Plan and the NRC has approved (Reference 17) this for current burnups.

It remains acceptable for extended burnups. The methods used for predicting the degree of rod bowing at extended burnups are evaluated below.

Evaluation - The Westinghouse methods for evaluating fuel and burnable poison rod bowing in 14x14, 15x15 and 17x17 assembly designs has been addressed in Reference 18 which has been approved by the NRC (Reference 17) for current burnup levels.

In response to NRC questions in this review Westinghouse has shown (Reference 3) that their conservative upper 95th percentile worst span l

closure curve from Reference.18 conservatively bound's their rod bow d'ta with a

t regional average burnups up to 48 mwd /kgM, i.e., peak rod average burnups approximately 53 mwd /kgM.

Inaddition,Westinghousehasindicated(Reference 2) that Westinghouse fuel assemblies will not be capable of achieving limiting power peaking factors et extended burnups due to the reduced power capabilities j

of these assemblies.

For example, the extended burnup assemblies are not limited by rod bow imposed penalties above assembly average burnups of approximately 33 mwd /kgM because the self-imposed decrease in power capabilities is greater I

than the penalty. Therefore, the operation of Westinghouse fuel assemblies l

1s found acceptable for this burnup range with respect to rod bowing.

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(g) Axial Growth Bases / Criteria - The core components requiring axial dimensional analyses are the neutron soruce rods, burnable poison rods, fuel rods, and fuel assemblies (thimble plugging rods are omitted because they are short and not axial growth limited). The axial growth of the first two of these components is primarily dependent upon the behavior of poison, source, or spacer pellets and their Type 304 stainless-steel cladding. The growth of the last two is mainly governed by fuel-pellet contact, and creep and irradiation growth of the 7.ircaloy-4 cladding, and Zircaloy-4 guide thimble tubes.

Failure to adequately design for axial growth of these components can lead to fuel rod-to-nozzle gap closure, rod bowing and perhaps fuel rod failure.

In addition, growth of the guide thimble tubes can result in collapse of the assembly holddown springs.

The Westinghouse design bases for core component rods are that (a) dimensional stability and cladding integrity are maintained during Condition I and II events and (b) these components do not interfere with shutdown during Condition III and IV events.

Westinghouse does not, per se, have design limits on the axial growth of their control, source, and burnable poison rods.

However, allowances are made to accomodate (a) pellet swelling due to gas production and (b) relative ther-mal expansion between the stainless-steel cladding and the encapsulated materi-al. Westinghouse does not account for irradiation growth of the stainless-steel cladding and has cited experiments (Reference 19) as justification for the insignificance of irradiation growth of stainless-steel at PWR operating conditions. This is also found to be true for extended burnup operation with the Zircaloy clad fuel rods providing the limiting conditions for irradiation growth.

For the Zircaloy cladding and fuel assembly components, the axial-dimensional tolerances that require controlling are (a) the spacing between the top and bottom of the fuel rods and the top and bottom fuel assembly nozzles, respec-tively, and (b) the spacing between the fuel assemblies and the upper and lower core plates. As noted earlier, failure to adequately design for the for-mer may result in fuel' rod bewing, and for the latter may result in collapse 11

a of the assembly holddown springs. With regard to a design basis for both rod-to-nozzle gap spacings and fuel assembly to core spacings, Westinghouse withdrew the proposed design limit in their extended burnup report (Reference 2) and indicated that they will continue to use the design limit approved in WCAP-9500 (Reference

12) which states that no axial interference shall take place due to closure of either the rod-to-nozzle gap spacing or the 'uel assembly to core spacing.

Evaluation - From the Westinghouse topical report on extended burnup (Reference

2) and responses (Reference 3) to NRC questions, Westinghouse has shown that they have both rod and assembly growth data near the burnups and fluences requested for extended burnup operation. These data indicate that the rod-to-nozzle gap spacings on Westinghouse fuel designs are approaching the above criteria at extended burriu,ns and thus should be monitored in their fuel surveillance prograrr. The models used by Westinghouse to predict rod and assembly growth appear to bound the extended burnup data and thus are found to be accpetable for extended burnup applications.

(h) Rod Internal Pressure Design Bases / Criteria - The Westinghouse design basis for fuel %d internal pressure is that the fuel system will not be damaged due to excessive fuel rod internal pressure. The Westinghouse design limits used to meet this design basis are that the internal pressure of the lead rod in the reactor will be limited to a value below that which could result in-(1) the diametral gap increasing due to outward cladding creep during steady-state operation and (2) extensive DNB propa-gation (References 2 and 20) This design basis and the associated limits have been found acceptable by the NRC (Reference 21) for current burnup levels, and they are not found to be affected by extended burnup operation. Therefore, they are also found to be acceptable for extended burnup application.

Evaluation - The models and methads used by Westinghouse to evaluate whether their designs meet the above basis and limits are examined in this section. The models used by Westinghouse are contained in the PAD 3.3 code (Reference 10) which has been approved by the NRC (Refernce 11) without restrictions of its use to high burnup fuel..As noted in Section 2.0(a) this code has been verified against fuel rod data with rod average burnups up to approximately 57 mwd /kgM. The NRC review of this code paid particular attention to those parameters important to internal rod pressure predictions, i.e., the thermal and fission gas release models.

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Therefore, the PAD 3.3 code is approved for use in the evaluation of rod internal pressures of extended burnup fuel to the target Westinghou'se burnup.

An important parameter of the methodology used in the internal rod pressure evaluations is the power history used as input to the PAD code.

Power history is very important because fission gas release and thus internal rod pressures are strongly dependent on the fuel thermal history.

In response to an NRC question on the power histories used, Westinghouse has indicated (References 5 and 6) that the power histories input to the PAD 3.3 code for calculating internal rod pressures are based on best estimate peak power histories from their fuel management calculations. The peak rod power histories are chosen by Westinghouse from their fuel management calculations based on those rods which have experienced the highest rod powers during each reactor cycle of operation, e.g., cycle 1, 2, 3, etc., along with the power history of the peak burnup rod of the fuel batch for a total of (number of cycles +1) power histories.

For example, a batch of fuel that will experience four cycles of operation will have at the maximum five peak power histories (sometimes a rod with a peak power during a specific reactor cycle corresponds to the peak burnup rod which would give four peak power histories for this case).

Each of these five peak power histories are then input separately into PAD 3.3 to calculate five different end-of-life internal rod pressures with the highest pressure being subject to the above criteria.

l In response to a subsequent NRC question concerning the conservatism in the rod power history methodology used by Westinghouse, Westinghouse presented analytical calculations (Reference 6) to demonstrate that the models used by Westinghouse are conservatively biased to bound any power uncertainties and Condition II power excursions that their fuel may experience as a result of extended burnup operation.

As a check on the conservatism in the Westinghouse methodology, we have perfomed f

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audit calculations using the GT2R2 code (Reference 22) along with bounding power l

histories of Westinghouse fuel to show that the Westinghouse methodology predicts bounding end-of-life rod pressures up to the extended burnups requested by Westing-I house (Reference 2).

From the above evaluation, it is concluded that Westinghouse models and methodology for detemining end-of-life rod pressures are adequately conservative and thus acceptable up to the extended burnups requested by Westing-house (Reference 2).

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s (i) Assembly Liftoff i

Design / Bases - The SRP calls for the fuel assembly holddown capability (wet weight and spring forces) to exceed worst-case hydraulic loads for normal operation, which includes anticipated operational occurrences. Westinghouse has stated (References 2 and 12) that they meet this criterion for all Condition I and II events with the exception of the turbine overspeed transient associated with a loss of external load. The NRC has accepted (Reference 13) this condition in the past for current burnup levels as long as the affected fuel assemblies can be shown to reseat properly in the core plate without damage or other adverse 1

effects during the event. This also remains acceptable for extended burnup assemblies as long as the came criteria are met as for current burnup fuel.

Evaluation - The fuel assembly liftoff forces are a function of primary coolant flow, spring forces and assembly dimensional changes. Westinghouse has indicated (Reference 2) that extended burnups will result in 1) additional irradiation relaxation of the holddown springs and 2) assembly length increases. These two phenomena have opposing effects on assembly holddown forces; however, Westinghouse predicts that there is a net increase in force with increased irradiation because fuel assembly growth is the dominant effect which more than compensates for the decrease in spring force. This is consistent with industry experience and thus assembly liftoff is not judged to be a problem at extended burnups.

(j) Control Material Leaching Control rods are not within the scope of this r'eview since they are treated separately and may be removed or installed in a core independent of fuel assembly burnup.

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3.0 FUEL R0D FAILURE In the following paragraphs, fuel rod failure thresholds and analysis methods for the failure mechanisms listed in the Standard Review Plan are reviewed.

When the failure thresholds are applied to normal operation including antici-pated operational occurrences, they are used as limits (and hence SAFDLs) since fuel failure under those conditions should not' occur according to the traditional conservative interpretation of General Design Criterion 10.

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

(a) Hydriding Bases / Criteria - Internal hydriding as a cladding failure mechanism is precluded by controlling the level of hydrogen impurities in the fuel during fabrication.

The moisture level in the uranium dioxide fuel is limited by Westinghouse (Ref rence 12) to less than or equal to 20 ppm, and this specification is compatible with the ASTM specification (Reference 23) which allows two micrograms of hydrogen per gram of uranium (i.e., 2 ppm). This is the same as the limit described in the Standard Review Plan and has been found acceptable by NRC (Reference 12) and continues to be acceptable for extended burnup application.

In addition, for extended burnup fuel, Westinghouse has introduced (Reference 2) a design limit on the hydrogen pickup level the value of which is proprietary.

Westinghouse has indi,cated that their test results show that the mechanical properties of Zircaloy-4 are not adversely affected at this level of hydrogen.

We agree with this assessment as long as hydride platelet orientation remains in the circumferential direction. Westinghouse has also stated that process controls and texture acceptance tests assure that Westinghouse cladding maintains the proper hydride platelet orientation. This design limit on hydrogen pickup level is found acceptable for extended burnup applications.

Evaluation - The hydrogen uptake of Zircaloy-4 during normal reactor operation to the extended burnup levels requested by Westinghouse is typically much lower than the Westinghouse criterion. The exception to this is when an abnormal amount of cladding oxidation is encountered that results in cladding failure.

Cladding oxidationisaddressedinSection2.0(e).

In this review Westinghouse has pro-vided data (References 2 and 3) on hydrogen uptake from commercial reactor operation to burnups that bound the extended burnup level requested by Westing-house. These data have shown that extended burnup operation up to the level i

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s requested by Westinghouse remains significantly below their criterion for hydrogen uptake. From this it is concluded'that hydriding is not a likely failure mechanism for Westinghouse fuel at extended burnups.

(b) Cladding 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 re.sult from collapse, the cladding is assumed to fail.

It is a Westinghouse design basis that fuel rod failures due to flattening will not occur.

In order to meet this design basis Westinghouse imposes a design limit for fuel rod clad flattening such that "the core residence time will not exceed the calculated core residence time which corresponds to a flattened rod frequency of 1.0".

This design basis and its associated criterion are essentially the same as those specified in the SRP and are found acceptable for extended burnup application.

Evaluation - The longer in-reactor residence times associated with extended-burnup fuel will increase the amount of creep of an unsupported fuel cladding.

Extensive postirradiation examinations of both test and consnercial fuel designs of current vintage by Westinghouse have not shown any evidence of cladding collapse or large local ovalities at rod atarage burnups up to approximately 62 mwd /kgM. This is primarily the result of the use of prepressurized rods and stable fuel in current generation designs.

Westinghouse utilizes a cladding collapse model (Reference 24) to show that the longer inreactor residence time associated with extended burnup fuel will not result in the collapse of an unsupported cladding with their fuel design. This method is very conservative in relation to the stable fuel employed in current designs, because it assumes a gap has fonned in the fuel column and the tube is unsupported. This method has been approved by the NRC (Reference 25) and because it explicitly accounts for the longer in-reactor residence times of ex-tended burnup fuel, it is also found acceptable for extended burnup applications.

(c) Overheating of Cladding Bases / Criteria - The Westinghouse design lim'it for the prevention of fuel failures due to overheating is that there will be at least 95% probability at

'16

s.

J a 95% confidence level that departure from nucleate boiling (DNB) will not

~

occur on a fuel rod having the minimum DNBR during normal operation and anticipated operational occurrences (Condition I'and II events). This design limit is consistent with the thermal margin criterion of SRP Section 4.2 and thus has been found acceptable by the NRC (Reference 13) for use at current burnup levels.

It is also judged to remain acceptable for extended burnup applications.

Evaluation - As stated in SRP Section 4.2, adequate cooling is assumed to exist when the thermal margin criterion to limit the departure from nucleate boiling (DNB) in the core is satisfied.

(d) Overheating of Fuel Pellets Bases / Criteria - As a second method of avoiding cladding failure due to over-heating, Westinghouse has as a design basis that the fuel rod will not fail due to fuel centerline melting during Condition I and II operation.

In order to assure that this basis is met, Westinghouse imposes a design limit on fuel temperatures such that there is a 95% probability that the peak linear heating rate (kW/ft) fuel rod will not exceed the UO2 melting temperature (Raference 2).

The melting temperature of the U02 is assumed to be 5080*F unirradiated and is decreased by 58 F per 10 mwd /kgM of exposure. A calculated centerline temper-ature of 4700 F has been selected by Westinghouse as the overpower limit. The fuel melting temperature dependence with fuel burnup is identical to that

~

proposed by Christiansen (Reference 26). Christiansen presented two sets of fuel melting data versus fuel burnup with the above relationship presented by Westinghouse being the largest decrease with burnup and believed to be the better of the two.

From this it is concluded that the Westinghouse criterion l

for fuel melting adequately acounts for the effects of extended burnup on fuel l

melting and thus is acceptable for extended burnup applications.

(e) Pellet / Cladding Interaction l

1 Bases / Criteria - As indicated in SRP Section 4.2, there are no generally 1

f applicable criteria for pellet / cladding interaction (PCI) failure. However, two acceptance criteria of limited application are presented in the SRP for PCI:

(1) less than 1% transient-induced cladding strain and (2) no centerline 17

t i

fuel melting. Both of these limits have been. adopted by Westinghouse for use in evaluating their fuel designs (References 2 and 12) and have been approved by the NRC (Reference 13) for current burnup applications. These are also found acceptable for extended burnup application.

Evaluation - Westinghouse uses the PAD 3.3 code (Reference 10) to show their fuel meets both the cladding strain and fuel melt criteria as discussed in Sections 2.0(b) and 3.0(d), respectively. As noted earlier, this code has been found acceptable for application to extended burnup fuel.

In addition, Westinghouse has presented (Reference 3) various power ramp data with rod average burnups up to approximately 46 mwd /kgM to show that PCI susceptibility does not increase at extended burnups.

In fact, the trend of these data suggests that PCI susceptibility may decrease at extended burnups.

From the above, it is concluded that Westinghouse methods adequately address the effects of PCI at extended burnups.

(f) Cladding Rupture Bases / Criteria - There are no specific design limits associated with cladding rupture other than the 10 CFR50 Appendix K requirement that the incident of rupture not be underestimated. The rupture model is an integral portion of the approved Westinghouse ECCS evaluation model (Reference 27). This is found acceptable for extend 2d burnups.

Evaluation - The cladding deformation and rupture models used by Westinghouse in their LOCA-ECCS analysis are directly coupled to their models for cladding ballooning and flow blockage. A more detailed discussion of these models and their relation to extended burnup operation is provided in the section that addresses cladding ballooning and flow blockage, see Section 4.0(c). These models have been approved by the NRC (Reference 27) for current burnup levels and for the reasons stated in Section 4.0(c) are also found acceptable for Gxtended burnup application.

Other parameters that are important to the L0CA analysis are those input to this i

analysis from the steady-state fuel performance code, PAD. There are two versions 18

4 of PAD, 3.3 (Reference 10) and PAD 3.3 Addenda 2 (Reference 28), used by Westinghouse to evaluate steady-state fuel' performance of their designs. The PAD 3.3 version h'as been verified against fuel performance data with rod average burnups up to 57 mwd /kgM and, as noted earlier, has been approved for extended burnupanalysisapplications(includingLOCA). The PAD 3.3 Addenda 2 version is a modification of PAD 3.3, in which conservatisms in the thermal model have been reduced. The PAD 3.3 Addenda 2 code has been verified against only thermal performance data at low burnups. Consequently, this code has been used for predicting early-in-life fuel thermal performance such as input for the LOCA analys.is. Westinghouse has justified (Reference 28) the use of the (low burnup)

PAD 3.3 Addenda 2 code for initializing steady-state thermal conditions input to their LOCA analysis of fuel designs at current burnup levels by showing that:

"the maximum peak clad temperature during a LOCA occurred using fuel parameters and initial conditions consistent with the time in life which exhibits the highest pellet average temperatures, near the beginning of life" (Reference 28).

Therefore, the use of the PAD 3.3 Addenda 2 code for initializing LOCA input has been approvsd by the NRC (Reference 29) for current burnup levels.

In response to an NRC question during this review, Westinghouse has responded (Reference 4) that their original statement (given above) from the PAD 3.3 Addenda 2 review remains valid for all Westinghouse plant configurations and approved ECCS Evaluation Models for extended burnups up to those requested in this review. Westinghouse has stated that this has been verified by performing a series of calculations with PAD 3.3 on those parameters sensitive to extended burnup. These calculations have taken into account the reduced fuel rod powers at extended burnups due to fissile material burnout. This effect is a real phenomenon at extended burnups, because if extended burnup fuel rods were driven to the rod powers allowed by the Technical Specifications, other lower burnup fuel in the core would exceed the Technical Specifications on power distribution peaking factors. The PAD 3.3 code continues to be used for fuel design cal-culations that are burnup dependent.

l 19

(h) Fuel Rod Mechanical Fracturing Bases / Criteria - The tenn " mechanical fracture" refers to a cladding defect that is caused by an externally applied force such as a load derived from core-plate motion or a hydraulic load. These loads are bounded by the loads of a safe-shutdown earthquake (SSE) and LOCA, and the mechanical fracturing analysis is usually done as a part of the SSE-LOCA loads analysis (see Section 5.0(d) of this SER).

Evaluation - The discussion of the SSE-LOCA loading analysis is given in Section 5.0(d) of this SER.

4.0 FUEL COOLABILITY l

For accidents in which severe fuel damage might occur, core coolability must be maintained as required by sev'eral General Design Criteria (e.g., GDC 27 and 35).

In the following paragraphs, limits and methods to essure that coolability is maintained are reviewed for the severe damage mechanisms listed in the Standard Review Plan.

i (a) Fragmentation of Embrittled Cladding Bases / Criteria - The most severe occurrence of claddir.g oxidation and possible f

fragmentation during a Condition III and IV accident results from a LOCA.

I i

Westinghouse uses the acceptance criteria of 2200*F on peak cladding temperature l

and 17% on maximum cladding oxidation as prescribed by 10CFR 50.46.

f t

f For events other than the LOCA, there are no separately establisned temperature

[

j or oxidation criteria. However, it is clear that for short-tenn events such as a locked rotor accident, the 2200*F peak cladding temperature and 17 percent

[

oxidation LOCA criteria are not really_ meaningful, because the temperature history

[

f for such an event is much shorter than that of a LOCA.

For events such as a locked rotor accident, Westinghouse uses (Reference 12) a peak cladding tem-i perature (PCT) criterion of 2700*F.

l I

I t

20 l

t

The Westinghouse 2700 F PCT limit was selected taking into consideration the short time (a few seconds) that the fuel is calculated to be in DNB for a locked-rotor type event and the fact that the peak cladding temperature and total metal-water reaction at the fuel hot spot is not expected to impact fuel coolable geometry. The NRC has concluded (Reference 13) that the 2700 F peak cladding temperature limit for short-term undercooling events such as the locked rotor is an acceptable coolability limit for Westinghouse fuel designs at current burnup levels. This coolability limit is not judged to be changed by extended burnup operation and thus is found to be acceptable for extended burnup application.

Evaluation - The cladding oxidation models used to determine the amount of cladding fragmentation and embrittlement during the LOCA are not affected by extended burnup operation; however, the steady-state fuel perfonnance PAD codes, used to provide input to the LOCA analysis, are burnup dependent. As noted earlier, Westinghouse has stated that early-in-life steady-state conditions are the most conservative for all their fuel designs. Consequently, Westinghouse has demonstrated tt.3t their LOCA analyses are insensitive to extended burnup operation because early-in-life conditions are most limiting. Therefore, the use of PAD 3.4 is found to be acceptable for LOCA analyses of Westinghouse extended burnup fuel.

(b) Violent Expulsion of Fuel Bases / Criteria - In a severe reactivity initiated accident (RIA) such as a control rod ejection accident, the large and rapid deposition of energy in the fuel could result in melting, fragmentation, and dispersal of fuel. The mechanical action associated with fuel dispersal might be sufficient to destroy fuel cladding and the rod-bundle geometry and to provide significant pressure pulses in the primary system. To limit the effects of an RIA event, Regulatory Guide 1.77 recommends that the radially-averaged energy deposition at the hottest axial location be restricted to less than 280 cal /g.

The Westinghouse design limits for this event are:

(a) Average fuel pellet enthalpy at the hot spot will be below 225 cal /g for unirradiated fuel and 200 cal /gm for irradiated fuel.

l 21

~

\\

)

(b) Average cladding temperature at the hot. spot will be below the temperature at which cladding embrittlement may be expected (2700*F).

(c) Peak reactor coolant pressure will be less than that which could cause pressures to exceed the faulted condition stress limits.

(d) Fuel melting will be limited to less than 10 percent of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits in (a),above.

These. limits are more conservative than the single 280 cal /g limit given in Regulatory Guide 1.77. They have been previously approved in the review of WCAP-7588 (Reference 31) and based on the above evaluation are found to be conservative and thus acceptable for extended burnup applications.

Evaluation - As discussed in Section 5 of this safety evaluation report, the methods used to calculate energy deposition as a result of reactivity insertion accidents (reactor physics codes) are also applicable to extended burnups. The extended burnup fuel is not expected to approach the 280 cal /gm criterion because fissile material burnout at extended burnups lowers the maximum possible fuel enthalpies when compared to maximum fuel enthalpies at lower burnups.

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

Evaluation - The Westinghouse cladding ballooning model is directly coupled to the cladding rupture temperature model for the LOCA-ECCS analysis and these are l

addressed in Revision 1 of WCAP-9220-P-A and WCAP-9221-A (Refer.ence 27). These l

models have been approved by the NRC (Reference 27) for current burnup levels.

The Westinghouse extended burnup topical report (Reference 2) references this approved report as being aoplicable to the LOCA analysis of extended burnup fuel.

42

e There is evidence that cladding oxidation at extended burnup levels and LOCA temperatures may result in reduced cladding strains (Reference 32)~from those traditionally predicted for LOCA. These data are not conclusive, however, becatse these tests were not performed with an oxidizing atmosphere nor under irradiation conditions.

Irrespective of whether these data are applicable to a LOCA, reduced cladding strains would result in less flow blockage and thus the current analysis methods would be more conservative with respect to this criterion.

In addition, the high cladding temperatures associated with the LOCA analysis will anneal any irradiation damage effects on cladding properties.

The steady-state operational input that is provided to the LOCA analysis from the FAD fuel performance codes is most limiting at early-in-life and thus are insensitive to extended burnup operation.

Consequently, the ECCS models approved for application to current burnup fuel are also found to be acceptable for application to fuel at extended burnup levels.

From this evaluation, it is concluded that the Westinghouse methodology for calculating cladding ballooning during a LOCA is acceptable for extended burnup applications.

(d)

Fuel Assembly Structural 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 j

required during these low probability accidents.

l The Westinghouse design basis is that the fuel assembly will maintain a geometry that is capable of being cooled under the worst case accident Condition IV event i

and that no interference between control rods and thimble tubes will occur during a safe shutdown earthquake (Reference 12). This is nearly identical to the design basis presented in the SRP and is therefore acceptable for extended burnup operation.

L 23

Evaluation - Generic analysis methods for performing combined SSE-LOCA loading analyses have been described by Westinghous'e in WCAP-9401-P-A (and-WCAP-9402-A) (Reference 33). These analysis methods not only include the fuel assembly I

structural response but also fuel rod cladding loads. These methods have been approved by the NRC (Reference 33) for current burnup levels.

The material property that could have an impact on these analyses at extended j

burnup levels is material ductility. These analyses could be impacted if cladding or assembly ductility were decreased, as a result of extended burnup 4

operation, to a level that would allow cladding or assembly failure not accounted for in the analysis. As noted in Section 2.0(a), the decrease in material ductility is expected to be negligible for the increased exposure and burnup levels requested and no adverse effects have been observed for rod average burnups up to 62 mwd /kgM.

j From the above evaluation, it is concluded that the above analysis methods i

are acceptable for extended applications.

It should be noted that this analysis is plant specific, because it requires site specific input ground motions and thus cannot be completed in a generic manner. Therefore, an applicant for an operating license proposing to reference the extended burnup report (Reference 2) and the SSE-LOCA analysis methods must perform site specific analyses using References 33 and 34 analysis l

methods in order to address the above criteria and Appendix A to SRP Section 4.2 guidelines.

l 5.0 NUCLEAR DESIGN Typical extended fuel burnup and increased fuel cycle length core designs utilize higher fuel enrichments, low leakage patterns and/or axial blankets.

Higher fuel enrichment is required to reduce the number of feed assemblies and offset the reactivity loss resulting from the higher fission product inventory.

The core neutron econor; is improved by reducing the radial leakage using low leakage loading patterns in which the high burnup fuel is located on the core periphery. Axial blankets are used to flattern the axial burnup distribution and improve fuel utilization. The increased power peaking resulting from the 1

l 24 w.-...-.

large reactivity differences between the fresh and high burnup fuel and the use of low leakage loading patterns is generally controlled using burnable poison rods.

These R atures affect the physics characteristics of high burnup core designs.

The increased fuel depletion in high burnup cores results in an increase in the plutonium fission fraction and the fission product inventory, the higher plutonium fission fraction in turn hardens the neutron spectrum and increases the neutron production per unit energy. The increase fission product inventory and use of burnable absorbers tends to increase absorption and also harden the neutron spectrum.

While the increased fuel burnup does affect the core physics characteristics, the changes are relatively small and the physics parameters are determined using standard calculational methods and procedures.

The high burnup neutronic effects enter through the microscopic cross sections and fuel assembly lattice group constants. The present calculations of these parameters account for substantial levels of plutonium, fission products and burnable absorbers, and these methods are expected to adequately treat the neutronics change associated with extended burnup. The depletion methods used to track the plutonium and fission product isotopics and various normalization procedures are also expected to be equally valid for high burnup fuel configurations.

The high burnup fuel physics characteristics and core configuration affect the core nuclear safety parameters. The major effect is to increase the power in the low burnup and/or centrally located fuel assemblies and to decrease the-power in the high burnup and/or peripherally located fuel assemblies. The resulting increase in the number and power of the peak powered rods is typically controlled by use of burnable poison rods.

The increased fission product inventory and use of burnable absorbers increases thermal and epithermal absorption and hardens the core neutron spectrum. These factors combine to reduce the boron and control rod worth, prompt neutron life-time and Doppler coefficient. The moderator temperature coefficient may increase or decrease depending on the particular high burnup design, and is also controlled l

using burnable absorbers as in present core designs. The delayed neutron fraction is also reduced as a result of-the increased plutonium fission fraction.

l y

A In addition to improving the neutron economy, the low leakage patterns reduce the pressure vessel damage fluence by shifting the power toward the center of the core and away from the vessel. This fluence reduction is partially offset, however, by the harder neutron spectrum and increased neutron production (per MeV) of the high burnup fuel.

The calculation of the high burnup core safety parameters is carried out using the same core and lattice methods and procedures used for present core designs.

The changes in the core safety parameters resulting from the higher fuel burnup designs tend to be relatively small as a result of the low relative importance of the high burnup fuel and the tendency for the increase in plutonium fission rates and fission product inventory to saturate. These calculated safety parameters provide the core neutronics input to the required plant transient and-accident analysis.

As the above discussion indicates, the effect of high burnup on the physics design is expected to result in relatively small changes in the predicted characteristics of the core, and also relatively small extensions in range of the methods used to calculate the characteristics. Because high burnup fuel is not subject to limiting duty and because of its low relative importance in detennining the core characteristics, we conclude that present methods are adequate for high burnup designs. To provide added assurance that these methods are adequate, we recomend that Westinghouse pay special attention to cct.Darisons of predicted and measured physics parameters (particularly power

d. tributions) which are monitored during the reactor cycle. A systematic pattern of deviation between predictions and measurements would provide an indication of potential problems. We intend to take an active role in following these comparisons.

6.0 RADIOLOGICAL CONSIDERATIONS OF POSTULATED ACCIDENTS WITH EXTENDED BURNUP OPERATION To ensure that accidents involving the movement of fuel do not constitute an offsite health and safety issue, design events are assessed. Analyses of fuel handling accidents assume release of the entire volatile radionuclide fuel assembly gap and plenum inventory under nominally 23 feet of water after the 1

26

e assembly has coole'i substantially (usually at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for BWR assemblies, 72 or 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for PWR assemblies).

For assemblies with burnup up to 38,000

~

mwd /t batch average at dis:harge, Regulatory Guide 1.25 assumptions are used.

These stipulate an inventory of ten percent of the total fuel assembly iodines 85 and r,sie gases (with the exception of 30 percent for Kr) in the gap and plenum volumes released upon clad perforation. An fodine decontamination factor (DF) of 100 (" Evaluation of Fission Product Release and Transport for a Fuel Handling Accident," G. Burley, USAEC, Revised October 5,1971) is assumed for 23 feet of water cover, and appropriate airborne radionuclide filtration / mixing, if any, is applied in the analysis before release to the atmosphere. The decontamination factor is based, in part, on an analysis of work presented in WCAP'-7518-L, " Radiological Consequences of a Fuel Handling Accident," M. J. Bell, et al, June 1970, NES Proprietary Class 2.

For fuel handling accident offsite radiological consequence evaluations involving fuel assemblies with burnup > 38,000 mwd /t batch average at discharge (extended burnup assemblies), the analysis is presently performed using Regulatory Guide 1.25 assumptions, but with modified gap and plenum fractional volatile radionuclide inventories. The fractional inventories range from a few percent (less than the R. G. 1.25 ten percent recommendation) to as much as 40-50 percent for certain high burnups/radionuclide combinations. The gap and plenum fractional inventories for the highest-power assembly are computed as a function of at least burnup, and at most time, temperature, and burnup using the GAPCON-THERMAL-2 computer code in conjunction with the ANS 5.4 fission gas release standard (model) proposed by the American Nuclear Society in " Radioactive Gas Release from LWR Fuel," C. E. Scyer, draft NUREG CR-2715, April 1982.

In generating these estimated fractional inventories, the conservative assumption of fuel assembly operation at a constant maximum-allowed peak linear heat generation rate (LHGR) for PWR's or MAPLHGR for BWR's is made. This assumption appears to be conservative within a factor of 2-3 for gap and plenum volatile inventories.

In addition to the conservative assumption regarding fuel assembly power operation noted above, there are two other significant sources of conservatism in the staff's analysis. The iodine decontamination factor (DF) assigned to the pool is taken to be a factor of 100.

It can be inferred from the report upon which this factor is based (WCAP-7518-L) that this value is probably l

27

conservative by about a' factor of three.

Finally, plateout of volatile iodine released from the fuel into the gap and fuel rod plenum has been entirely neglected. Although not well quantified, a tentative estimate suggests that about 10 percent or less of the iodine released into the gap will remain volatile at the fairly low temperatures after the fuel has been allowed to cool for about a day or more.

Because of the significance of these conservatisms, the staff intends to study and quantify them in more detail and to use the results of such evaluations to appropriately revise the staff's Standard Review Plan (SRP),

NUREG-0800.

In the interim, the staff concludes that consideration of all

~

three factors together noted above may permit a significant reduction of estimated thyroid doses compared to existing analyses. Adequate justification by licensees on a case-by-case basis, or by vendors on a generic basis, are likely to provid.e sufficient bases for departing from SRP criteria until such time as detailed changes can be made. A reduction by a factor of two is likely to be appropriate and conservative. Consequently, with regard to evaluation of thyroid doses for fuel-handling accidents involving extended-burnup fuel

(>38,000 mwd / tonne), and pending SRP revision, it is likely that justification can be provided for lower estimates of thyroid doses from fuel handling accidents by a factor of two in departures from SRP review criteria.

7.0 REGULATORY POSITION The review of Westinghouse Electric Company's submittal, as described in l '

WCAP-10125 (Proprietary) and responses to NRC questions in Reference 3 through 6, for application of their design criteria and analysis methods to extended burnups has been completed. As a result of our review, we conclude that these criteria and analysis methods are applicable to licensing of Westinghouse Zircaloy-clad fuel design:, up to the extended burnup level requested in WCAP-10125.

l From this evaluation, we have concluded that Westinghouse criteria and analysis methods, as described in the extended burnup topical report and response to questions, Reference 2 through 6. for extended burnup application are adequate such that 1) fuel damage is nct expected to occur as a result of normal oper-tion and anticipated operational occurrences (Condition I and II events), 2) 28 4

fuel damage during postulated accidents (Condition III and IV events) would not besevereenoughtopreventcontrolrodinsertionwhenitisrequiied,and3) core coolability will always be maintained even after postulated accidents (Condition III and IV events).

This conclusion is based on two primary factors:

1 1)

Westinghouse provided sufficient evidence that the design citeria will allow for safe operation of Westinghouse design fuel at the proposed extended burnup level; and 2)

, e Westinghouse analysis methods used to assure that these criteria Th are met have been based on adequate extended burnup operating experience and prototype testing.

I P

,5 I

L i

29 f

l i

8.0 REFERENCES

1. Letter,L.S.Rubenstein(NRC) tot.M.AndersoniW),datedJune2,1981.

2. P. J. Kersting (Editor), Extended Burnup Evaluation of Westinghouse Fuel, WCAP-10125 (Proprietary), Westinghouse Electric Corp., July 1982.

3. Letter,E.P. Rahe (W) toc.O. Thomas (NRC),"ResponsetoRequestNumber2 for Additional Information on WCAP-10125, (Extended Burnup Evaluation of Westinghouse Fuel)," (Proprietary), dated June 11, 1984.

4. Letter,E.P.RaheIW) toc.O. Thomas (NRC),"AdditionalInformationon WCAP-10125 (Extended Burnup Evaluation of Westinghouse Fuel)," (Proprietary),

dated October 31, 1984.

5. Letter,E.P.RaheIW) toc.O. Thomas (NRC),"ResponsetoInformalNRC Question on Power Histories and Fission Gas Release Uncertainties Used in E0L Rod Internal Pressure Calculations," (Proprietary), dated January 7, 1985.

6. Letter,E.P.RaheIW) toc.O. Thomas (NRC),"SlidesforTransientFission Gas Release at Extended Burnup Presentation to NRC on January 17, 1985,"

(Proprietary), dated February 6, 1985.

7. Standard Review Plan for the Review of Safety Analysis Peports for Nuclear Power Plants--LWR Edition, NUREG-0800 Section 4.2, " Fuel System Design," Rev.

2. July 1981.

8. M. D. Beaumont, et al., Properties of Fuel and Core Component Materials, WCAP-9179, Revision 1 (Proprietary) and WCAP-9224 (non-Proprietary),

i Westinghouse Electric Corp., 1978.

9. Letter to E. P. Rahe, Washington, from C. O. Thomas, NRC, Septebmer 29, 1982.

10. W. J. Leech, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations, WCAP-8720, Addendum 1, September 1979, (Proprietary), and WCAP-8964-A (non-Proprietary), Westinghouse Electric Co,p., August 1978.

j t

30

11. Letter, H. Bernard (NRC) to E. P. Rahe (W), " Acceptance for Referencing of Licensing Topical Report WCAP-8720, Addendum 1," dated July 20, 1982.

12. S. L. Davidson and J. A. Iorii, Reference Core Report 17x17 Optimized Fuel Assembly, WCAP-9500-A, Westinghouse Electric Corp., May 1982.

13. Letter, R. L. Tedesco (NRC) to T. M. Anderson (W), " Acceptance for Referencing of Licensing Topical Report WCAP-9500," dated May 22, 1981.

14. C. J. Baroch, "Effect of Irradiation at 130, 650, and 775 F on Tensile Properties of Zircaloy-4 at 70, 650, and 775*F," Properties of Reactor Struc'tural Alloys After Neutron or Particle Irradiation, ASTM STP 570, p.129, 1974.

15. M. Shimada, et al., " Ductility Loss of Ion-Irradiated Zircaloy-2 in Iodine Environment," Effects of Radiation on Materials:

Tenth Conference, ASTM STP 725, p. 233, 1981.

16. W. J. O'Donnell and B. F. Langer, " Fatigue Design Basis for Zircaloy Components," Nuc. Sci. Eng., Vol. 20, p. 1 (1964).

17.. Letter, C. O. Thomas (NRC) to T. M. Anderson (W), dated May 31, 1983.

18. J. Skaritka (editor), Fuel Rod Bow Evaluation, WCAP-8691, Revision 1, l

(Proprietary)andWCAP-8692, Revision 1(non-Proprietary), Westinghouse Electric Corp., July 1979.

l

19. J. P. Foster and R. V. Strain, " Empirical Swelling Equations for Solution-Annealed Type 304 Stainless Steel," Nuclear Technology, Vol. 24, p. 93, October 1974.

20. D. H. Risher, et al., Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis, WCAP-8963 (Proprietary) and WCAP-8964 (non-Proprietary),

1977.

31 e

1 4

i

21. Letter, J. F. Stolz (NRC) to T. M. Anderson (W), " Safety Evaluation of WCAP-8963," dated May 19, 1978.

~

22. M. E. Cunningham and C. E. Beyer, GT2R2:

An Updated Version of GAPCON THERMAL-2, NUREG/CR-3907 (PNL-5178), Pacific Northwest Laboratory, Richland, l

Washington, September 1984.

l

23. Standard Specifications for Sintered Uranium Dioxide Pellets, ASTM Standard C776-76, Part 45 (1977).

j

24. R. A. George. Y. C. Lee, and G. H. Eng, Revised Clad Flattening Model, WCAP-8377 (Proprietary) WCAP-8381 (non-Proprietary), Westinghouse Electric Corp., July 1974.

25. Memorandum, V. Stello (NRC) to R. De Young (NRC), " Evaluation of Westinghouse Report WCAP-8377, Revised Clad Flattening Model," dated January 14, 1975.

26. J. A. Christiansen, et al., " Melting Point of Irradiated Uranium Dioxide,"

Trans. Am. Nucl. Soc. 7 (2), p. 390, 1964.

27. Nuclear Safety, NTD, Topical Report, Westinghouse ECCS Evaluation Model, 1981 Version, Westinghouse Report WCAP-92'20-P-A (Proprietary) and WCAP-9221-A I

(non-Proprietary), Westinghouse Electric Corp., February 1982.

28. W. J. Leech, D. D. Davis and M. S. Bengvi, Revised PAD Code Thermal Safety Model, WCAP-8720, Addenda 2. Westinghouse Electric Corp., 1982.

i

29. Letter to E. P. Rahe, Washington, from C. O. Thomas, NRC, December 9, 1983.

30. P. E. MacDonald, et al., " Assessment of Light Water Reactor Fuel Damage i

During a Reactivity Initiated Accident," Nuclear Safety, Vol. 21 No. 5, p. 582, i

September 1980.

f r

i 32 i

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31. An Evaluation of the Rod Ejection Accident in Westinhougse Pressurized i

Water Reactor Using Special Kinetics Methods, Rev.1, WCAP-7588, Westinghouse Electric Corp.

32. P. Hoffman, " Influence of Iodine on the 5 train and Rupture Behavior of Zircaloy-4 Cladding Tubes at High Temperatures," Zirconium in the Nuclear Industry, ASTM STP 681, p. 409, American Society for Testing and Materials, 1979.

33.

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