ML20050B362

Nonproprietary Partial Response to NRC Questions on CEN-161(B)-P,Improvements to Fuel Evaluation Model.

ML20050B362
Person / Time
Site:	Calvert Cliffs
Issue date:	03/21/1982
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML19268D137	List: ML20050B355 ML20050B362
References
CEN-193(B)-NP, CEN-193(B)-NP-S02, CEN-193(B)-NP-S2, NUDOCS 8204050247
Download: ML20050B362 (75)

Text

1

,. . _ l CEN-193(B)-NP Supplement 2-NP

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PARTIAL RESPONSE TO NRC QUESTIONS ON CEN-161 (B)-P, IMPROVEMENTS TO FUEL EVALUATION MODEL March 21,1982 i

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4 8204050247 820326 PDR ADOCK 05000317 P PDR

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l This report was prepared as an account of work sponsored by Combustion Engineering, Inc. Neither Combustion Engineering nor any person. acting on its behalf: s

, A. Makes any warranty or representation, express or implied including the

. warranties of fitness for a particular purpose or merchantability, with respect to accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or -

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process .

disclosed in this report.

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Question 1. Application of CEN-161 (R'ef.1) in Licensing Analyses From the list of licensing analysis requirements provided below, please identify which applications will be met with FATES-3. Where the licensing analysis requirements will not be met with FATES-3, identify the analysis methods used.

A. LOCA initial conditions. ,

B. Initial conditions for other codes.

1. Core thermal-hydraulic transient codes. .

2. Point kinetics system analysis codes.

3. Core-wide power distribution codes.

C. Fuel system damage limits or initialization of other analyses used to calculate fuel system damage limits.

1. Stress, strain, or loading limits.

2. Strain fatigue limits.

3. Fretting wear limits.

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4. Oxidation, hydriding, and the buildup of corrosion products (crud) limits.

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5. Dimensional changes such as rod bowing or irradiation growth limits.

6. Fuel and burnable poison rod internal pressure limits. .

7. Worst-case hydraulic load (assembly holddown) limits.

8. Control rod reactivity limits.

1-1 w _ - - - _ _ - _ _ - . _ _ . _ _ _ _ _ . . - _ . - - - - - - _ - - . - . _ _ - -

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D. Fuel rod failure or initialization of other analyses used to calculate fuel rod failure.

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. 1. Overheating. -

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, a. Departure from Nucleate Boiling (DNB).

b. Fuel enthalpy.

c. Fuel centerline melting (steady-state and transient).

2. Pellet / Cladding Interaction (PCI).

3. . Hydriding limits.

4. Cladding collapse.

5. Bursting.

6. Mechanical fracturing,

i. Fretting.

E. Fuel coolability or initialization of other analyses used to

. calculate fuel coolability.

, 1. Cladding embrittlement.

2. Violent expulsion of fuel.

3. Generalized cladding melting.

4. Structural deformation.

5. Fuel rod ballooning.

1-2

Response 1.

. FATES 3 is intended to provide predictions of the steady-state response of fuel rods, and to model interaal conditions of the fuel rods within the core from insertion to discharge. With the appropriate modeling of mechanical design data, power levels, and power distributions, these internal conditions serve as input for transient analyses where initial conditions are required.

The FATES 3 results are also used to confirm conpliance with design criteria and .

are input to reactor system setpoints. Specifically, FATES 3 calculations can be used to provide:

1. Fuel rod initial conditions for LOCA analyses.

2. Fuel rod initial conditions for non-LOCA analyses for which initial conditions at the start of the transient are important (e.g., internal gas p.ressure, internal gas composition, etc.).

3. Minimu'm value of the steady-state power-to-centerline-nelt.

4. Engineering factor on linear heat generation rate.

5. Fuel column thermal' expansion and densification for computation of the densification factor included in establishing Technical Specifications LCO and LSSS limits on linear heat rates.

6. Stored energy data for containnent analysis.

7. Minimum core average gap conductance values for input to non-LOCA transient analyses.

8. Fuel rod initial conditions for internal gas pressure for clad stress and strain co}pliance calculations.

1-3

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9. Maximum end-of-life design pressure and maximum cycle by cycle operating pressure.

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10. Maximum core average gap conductance values for input to non-LOCA transient analyses.

11. Minimum pressure histories for input to clad collapse analyses.

12. Fuel tegerature - power correlation data used to establish doppler reactivity coefficient correlations.

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'I 1-4 -

1 Question 2. Differences Between FATES-3 and Previous Fuel Performance Codes Question A.

Using the list of models provided in Section II.C.3(a) of Standard Review Plan 4.2, indicate whether the appropriate FATES-3 model description can be found in CEN.161 (Ref.1) or CEN-139 (Ref. 2). If the model is used in FATES-3, but is not described in either report, please provide a description of the model.

Response A.

Table 2A-1 lists the models provided in Section II.C.3 (a) of Standard Review Plan 4.2 3states whether the models are used in FATES 3, and gives the appropriate references for the model descriptions. The only model which is used in FATES 3 but not described in either CENPD-139 or CEN-161(B) is the burnup distribution in the fuel. A description of this model is given below.

Heat capacity (I) of the fuel and clad are not incorporated into FATES 3 but are used in external calculations for core stored energy.

Burnup distribution in the fuel. -

The fuel burnup varies at each axial segment (maximum of 20) in the fuel rod as a time integrated effect of the axial power distributions through the life of the fuel rod. The radial burnup distribution within each segment is taken in FATES 3 to be uniform.

(I)CENPD-135P, "STRIKIN-II, A Cylindrical Geometry Fuel R.od Heat Transfer Program", 8/74.

2 2-1

TABLE 2A-1 FATES 3 Models and References for Models of Standard Review Plan 4.2 Model in Section II.C.3(a) of Model Used Model Described Model Described Standard Review Plan 4.2 in FATES 3 in CENPD 139 in CEN-161(B)

Radial power distribution X X I

Fuel and cladding temperature distribution X

j X X X Burnup distribution in the fuel X Thermal ccnductivity of the fuel, cladding, X X cladding crud, and oxidation layers -

Censification of ,the fuel X X Thernal expansion of the fuel and cladding X X Fission gas production and release X X Solid and gaseous fission product swelling X X ro Fuel restructuring and relocation X X X;

/o Fuel and cladding dimensional changes X X X Fuel-to-cladding heat transfer coefficient X X X Thermal conductivity of the gas mixture X X Thermal conductivity of the Knudsen domain X X Fuel-to-cladding contact pressure X X lleat capacity of the fuel and cladding Growth and creep of the cladding 2 X X X Rod internal gas pressure and composition X X j

X Sorption of helium and other fill gases X X -

Cladding oxide and crud layer thickness Cladding-to-coolant heat transfer coefficient X X 1 - contains description of annular fuel pellet models 2 - description of clad axial irradiation growth model given in detail in CENPD-198 e

Question B.

To help us identify changes in operating limits that might be expected as a result of the application of CEN-161, describe and quantify the differences between FATES-3 predictions.and those calculated by previous codes for the

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applications identified in your response to Question i.

Response B.

C-E does not expect any changes in the operating limits to result because of the application of FATES 3, although some changes in key performance parameters (i.e., fuel temperatures and internal gas pressure) will exist. Differences in results between FATES 3 (CEN-161(B)) and the earlier version of FATES (CENPD-139) for the applications defined in Response 1 of Question 1 are illustrated by the changes in these key parameters shown in the two sets of comparisons provided below. The first set is a code comparison using a simple idealized history on identical fuel rods. The second set is a comparison of results taken from the FATES 3 licensing analysis for Calvert Cliffs 1 Cycle 6 and corollary runs using the earlier version of FATES (CENPD-139). The NRC fission gas release enhancement factor (NUREG 0418) was used in both sets.

Sample Comparison of FATES 3 with Previous Code Version Comparisons of fission gas release, rod internal pressure, and fuel temperatures are shown in Figures 2B-1, 28-2, and 28-3, respectively, for FATES 3 (CEN-161(B)) and the previous code version (CENPD-139) with the NRC fission gas release enhancement factor (NUREG-0418) applied. The fuel rod characteristics assumed in these comparisons are typical of a C-E 14x14 fuel assembly design.

It should be noted that these results are from sample calculations which are intended to show only how FATE'S3 compares to the previous FATES version (using the enhancement factor) for a simple power distribution transient. This sample calculation should not be construed to represent results which would be obtained in a licensing application.

2-3 1

The calculations were perfor"med by first assuming that the rod was operated with a relatively flat steady-state axial power distribution having a constant peak linear heat rate of 10 kw/ft out to the burnup of interest. Then an instantaneous power distribution transient leading to a peak linear heat rate of 14 kw/ft was imposed. The total gas release, internal pressure resulting from the transient, and the fuel temperatures in the fuel rod at the peak of 14 kw/ft are shown. Sufficient time was assumed to have elapsed after imposing the disturbed power distribution such that both instantaneous and restructuring (grain growth) fission gas release ~ occurred. In addition, transients at earlier burnups were assumed to have no permanent effect on fuel rod conditions at the burnup of interest, i.e., no earlier transients were assumed to have occurred. It should be noted that the assumed power history scenario neglected decreases in peaking factors that are normally inherent in highly burned fuel.

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2-4

Concarison of FATES 3 and Previous Code for Calvert Cliffs 1 Cycle 6 The fuel rod geometry data and power history data used in the Calvert Cliffs 1 Cycle 6 FATES 3 analysis for input to LOCA is provided by Response 3 to Question

3. Results of the application of the previous code (using the NRC enhancement factor, NUREG-0418) to the hot rod of that analyses are co@ared to FATES 3

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in Figures 2B 4 through 28-8.

Figure 28-4 shows the maximum fuel average temperature from both codes as a function of linear heat generation rate. The maximum generally occurs at the time of maximum gap size early in life. The average temperature and maximum centerline temperature of the mid section of the fuel rod which has an initial power level of 11.95 kw/ft (coglete power history is given in Response 3) is shown as a function of burnup in Figure 2B-5. Differences are not significant. Initial densification causes temperature increases in both codes followed by temperature decreases which are primarily due to gap closure.

Temperatures decrease , _ _ _ _ _ _ _ _ _ __ _ . ._ ._

in FATES (with NUREG 0418 applied) but continues to decrease in FATES 3 until fuel-clad contact occurs. The additional step decreases in temperature follow step decreases in power level (described in the input data of Response 38).

Figure 28-6 shows the gap conductance for the same location of the hot rod as a function of burnup. [ Z

_ Z.Z T Z._T Z _'_^~Z -Z

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the contribution to temerature differences across the gap and, therefore, fuel temperatures is not si5nificant as was seen in Figure 28 4 Figures 28-7 and 2B-8 show comparisons of fission gas . release and internal gas p ressu re. The trend and interpretation of these results is the same as for the samle cogar ison given above. [^ _ _ _ __ __ _ __

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FIGURE 281 TOTAL ROD FISSION GAS RELEASE VERSUS ROD AVERAGE BURNUP k

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ROD AVERAGE BURNUP, MWD /KGU 2-7 b

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FIGURE 28-2 ROD INTERNAL PRESSURE AT 14 KW/FT VERSUS ROD AVERAGE BURNUP 9

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1 MAXIMUM FUEL AVERAGE TEMPERTURE,*F

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FIGURE 28-5 FUEL TEMPERATURES AT MIDSECTION OF HOT ROD (AT 11,95KW/FT INITIAL POWER)

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ROD AVERAGE BURNUP, GWD/MTU 2-11

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FIGURE 23 6 GAP CONDUCTANCE AT MIDSECTION OF HOT ROD (AT 11.95 KW/FT INITIAL POWER) 1 h

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ROD AVERAGti BURNUP, GWD/MTU 2-12

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. FIGURE 28-7 FISSION GAS RELEASE VERSUS ROD AVERAGE BURNUP l

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Question 3. FATES-3 Input Question A.

Provide a list of input parameters for the FATES-3 code. State whether each input value used is best-estimate or conservative, or otherwise results in a best-estimate or conservative calculation. If conservative, quantify the margin of conservatism and the basis for the margin used. If the margin varies from application to application, provide this information for each different application.

Response A.

A list of input parameters for FATES 3 is provided by Table 38-1 (fuel rod data),' Table 38-2 (core data), and Tables 38-3 and 38-4 (power history data). The FATES 3 results are conservatively biased to the end applications defined in Response 1 of Question 1. _ _

The applications of FATES 3 defined in Response 1 represent only five categories of analyses, where imposed conservatisms vary from category to category. Applications 1 to 8 are identical with respect to conservatisms. Applications 9,10,11, and 12 have individual conservatisms imposed. These categories are discussed individually below with respect to input conservatisms.

Applications 1-8 Results are conservative when predicted temperatures (or stored energy) and internal gas pressure are higher than what the actual operating fuel rod is expected to experience for conditions at the start of a transient or conservatively represents conditions during a transient. -

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1 Input parameters of Table 38-2 are generally nominal core values [

] However, the thermal hydraulic model contains conservatisms for these applications in that fuel rod to coolant heat transfer is based on a closed channel. Beneficial effects of the lower heat input from lower power adjacent fuel rods and from cross-flow are neglected.

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results in high coolant temperatures and, therefore, high clad surface temperatures.

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Maximum cycle by cycle operatihg pressure is based on the same conservar. isms i and analyses as Applications 1-8. However, the maximum end-of-life design

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pressure is calculated using additional conservative biasing [ ~~

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g Application 10 FATES 3 is applied using the input of Application 1-8 except that no fission gas

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releaseisassumedandfuelandcladareassumedtobeincontact[

] It is used to represent an upper limit on gap conductance at core average fuel conditions.

Application 11

.__pesameinputisusedasforApplication1-8exceptthat[____.___.__,___

the lead burnup rod is used. No fission gas release is

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Application 12 All input is nominal to the extent possible to obtain best-estimate output.

3-6

Question B.

Provide the input values for a typical case of each of those applications identified in response to Question 1 in sufficient detail to permit staff audit calculations. ,

Response B.

A typical set of values for FATES 3 input is provided by Tables 3B-1, -2, -3, and -4. These are the values used to produce the LOCA analysis initial conditions for Calvert Cliffs 1 Cycle 6. These are used for input to Applications 1-8. Variations for other applications are as described in Response 3A.

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TABLE 38-1 FATES 3 Input Parameters

-Fuel Rod Data Pa rameter Value pellet diameter, i n, active fuel length, in.

. dish depth, in, land diameter, in.

bottom dish diameter, in, 3

f ractional open porosity in open in' total porosity pellet length, i n.

chamfer height, in.

chamfer diameter, i n.

grain size, microns initial density, % TD 2 variance on density %TD final density, % TD pellet surface roughness, in.

fuel enrichment, wt%

3 helium sorbtion, cm /gm of fuel clad outer diameter, in.

clad inner diameter, in.

clad length, in. 3 spring volume, in end cap volume in 3 air partial pressure, psi fill gas partial pressure, psi helium in fill gas, fraction nitrogen in fill gas, fraction clad surface roughness clad material tota' )acer height, i n.

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3-8

'f Table 38-2 FATES 3 Input Parameters .

-Core Data Pa rameter Value -

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heat transfer area per foot of rod per unig flow area, ft-I convectivegap maximum heat transfer coeff.,

conductance, BtuBju /hr-ft fbr-ft - F

-F minimum gap conductance, Btu /hr-ft 2 0-F coolant pressure, psia 2

coolant mass velocity, lbm/hr-ft coolant inlet enthalpy, Btu /lbm core average LHGR, kw/ft

! core average fast flux (>1.0 MEV),/n/cm!-sec 4

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3-9

Table 38-3 FATES 3 Input Parameters

-Rod Power Distributions Axial Distributions

. Node No.*

1 (bottom)

. 2 3

4 5

6 <

7 8

9 -

10 11 12 13 14 15 16 17 18 19 20 (Top)

• Nodes are equally spaced Radial Distributions Pa rameter Value Energy deposition in fuel, %

Energy deposition in clad, %

Flux depression constants **

A B

C

• • Computed by FATES 3 by method described in CENDP-139 e

3-10

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• FATES 3 Input Parameters

-Burnup Dependent Power Level .

- FATES 3

. Cycle

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3 4

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6 7

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9 10 11 12 13 14 15

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17 18 29 29

21 22 23 24 25 26 4

27 28 29 30 31 4

32 33 34 35 .

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i 3-11

Question 4. FATES-3 Output l

l Question A.

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Provide a list of output parameters from the FATES-3 code. Identify the end

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use of each output value produced by FATES-3 (e.g., another code, design criteria, information only) for each application identified in response to Question 1.

Response A.

Output from the FATES 3 code represents either hot operating conditions or cold shutdown conditions existing within the fuel rod as a function of burnup. A list of the important parameters is provided by Table 4A-1. The list of output parameters is identical regardless of intended application.

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4-1 e

Table AA-1 FATES 3 Output Parameters and Uses ,

. Output Parameter End Use Red axial average burnup initial conditions

. Rod internal pressure (hot) initial conditions and design criteria Rod internal pressure (cold) informative and initial conditions Fission gas generated informative <

. Fission gas release informative Void volume (hot) initial conditions Void volume (cold) informative Gas conposition initial conditions Burnup distributions -

init"al conditions Local fuel average temperature initial conditions, stored energy calculation, and axial densification factor calculation Local fuel and clad tenperature profiles stored energy calcu-1ations Gap conductances initial conditions Hot fuel dimensions informative Hot clad dimensions informative Cold fuel dimensions initial conditions Cold clad dimensions initial conditions Fuel melt temperature informative Steady-state power to fuel centerline melting setpoint, design criteria Flux depression constants another code Changes in fuel and clad dimenions (relocation, swelling,etc.) informative Fuel density , initial conditions e

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. 4-2

Question B. .

State whether the value of each output parameter is best-estimate or conservati ve. If conservative, quantify the margin of conservatism and the basis for the margin used. If the margin varies from application to application, provide this Information for each different application.

Response B.

a. pL.

r- . . . -- . . - -. . _..-. - .

d. studiesb)inthisareaindicatethatthebuildupofplutonium

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'?ccent in burned fuel can significantly increase flux depression and reduce fuel temperature.

(1) " Radial Concentration and Effect on Temperature of Plutonium ,

Formed in U0 During Irradiation", H. Carlsen and D. N. Sak, 2

Nuclear Technolocy , Vol. 55, Dec.1981, pp. 587-593. -

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Question C. .

Using the input values described in your response to Question 3 atove, provide the corresponding FATES-3 output values in sufficient detail to permit a comparison with staff audit calculations.

Response C. <

FATES 3 output for the Calvert Cliffs 1 Cycle 6 analyses are given in Fi gures 28-4, -5, -6, -7, an d -8. Maxinom radial averaged fuel terperatures are given in Figure 2B-4 as a function of local linear heat generation rate.

Fuel centerline and racial averaged fuel terperatures are given in Figure 28-5 foraxialnode10[~ .__ _ ____ _ _-- [; gap conductance for node 10 is given in Figure 28-6. Fuel tenperature and gap conductance data [ -~ ~ ~ ~ ~

.$. . lT .$ . I '_ I.$. in node 3 is given in Table 4C-1 as a function of burnup. Total rod fission gas release and rod internal pressure are given in Figures 2B-7 and 28-8,respectively. Input data for these analyses are provided by Respanse 38.

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Table 4C-1 Typical FATES 3. Output _. . _ . .

b _ . __ _ _ . . _ _ _ .

Hgap at Peak Tavg TCL at Rod Power at Peak Peak Average Local Peak Peak Node Power Power FATES Burnup Burnup Powe r Power (Btg/hr Node Node Cycle (MWD /MTU) (MWD /HTU) Node (kw/ft) -ft -F (DegF) (Deg. F) r j .

2 c 3

4 5'

6 7

8

'. 9 10 11 12 13 14 15 16 17 18 i 19 20 21 22 23 24 25 26 27 28 29 30 31 l 32 l 33 34

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36 37 38 39 J

4 4-6

, Question 5. Code Conservatism ,

Question 5. ,

Explain why the magnitudes of the conservatisms identified in response to Question 4 are appropriate for each application identified in response to Question 1. It is expected that certain output parameters, particularly fuel tegeratures used for LOCA initial conditions, should exhibit an appropriate level of conservatism, based on statistical analysis of experimental-vs- -

predicted fuel te geratures.

Response 5.

Conservatisms as exhibited by experimental versus predicted fuel temperatures are a significant and valuable contribution to conservatisms in the licensing analysis. Comparisons of experimental versus FATES 3 predicted fuel temeratures are given in Figure 9-4 of CEN-161 (B). As indicated in CEN-161 (B), only temperature data from fuel rods equipped with thermocouples are considered satisfactory for such comparisons. Because of the limited amount of thermocouple data that is available, statistical analysis is not done.

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As such, they are believed to be appropriate for the applications identified in Reference 1.

5-1

Question 6. Model Verification Question A.

Provide additional fuel temperature model verification by predicting fuel centerline thermocouple and rod internal pressure response for IFA-432 (Refs.3-4). These results may be presented graphically. In such a case, the results should show:

1. Fuel centerline temperature prediction vs. measurement (6 rods).

2. Rod internal pressure prediction vs. measurement (Rods 1, 5 and 6).

3. Ratio of fuel centerline temperature prediction and measurement as a function of burnup (6 rods).

4. Ratio of fuel centerline temperature prediction and measurement as a function of initial gap size (Rods 1, 2 and 3).

5. Ratio of rod internal pressure prediction and measurement as a function of burnup (Rods 1, 5 and 6). ,

s Response A. _

A ,

Discussion of the IFA-432 test. -

IFA-432 is an instrumented test fuel assembly designed by BNW for the USNRC for the study of thermal stored energy in UO2 fuel rods (l). The assembly j

was first inserted into the Halden Boiling Water Reactor in Norway in December 1975. Halden is a 12 MW test reactor that operates at a coolant pressure of 500 psi (low relative to LWRs) and an inlet temperature of 273C (459F).

i 6-1

Data reports on the performance of the fuel ods in IFA-432 through the period December 1975 to May 1980 have been issued (2)' (3). Post-i rradiation examination data, such as measured fission gas release, are not yet available for the six original rods in thi's test.

The IFA-432 assembly was designed to accommodate six instrumented fuel rods in a circular array. The original six rods loaded into this assembly (Rods 1-6) have been modeled using FATES 3. Only Rod 4 in this group was replaced during the reported irradiation period. The replacement rods (Rod 8, and its eventual replacement, Rod 9) were not analyzed for this response. <

! All rods were equipped with thermocouples at the centerline of hollow fuel ,

pellet columns located at each end of the fuel stack and were equipped with elongation detectors. Rods 1, 5 and 6 were equipped with pressure transducers.

j .

The fill gas was helium at one atmosphere pressure in all but one rod. Rod 4 j contained xenon at one atmosphere to determine its effect on gap conductance relative to helium. It was assumed in the FATES 3 analyses that the fill gas in all rods was contaminated with 11% nitrogen by volume, which is typical of nonpressurized fuel rods, and results from the normal incomplete evacuation of -

air prior to backfilling.

Each fuel rod contained 44 or 45 pellets,12 to 19 of which were hollow

(< 3 vol% hole) for thermocouple accommodation. The pellets were nominally about one-half inch long. The nominal pelht diameter was about 0.420 inches in rods with a standard 9 mil gap (i.e., Rods 1, 4-6). Slightly larger and slightly smaller diameter pellets were used to produce one rod (Rod 3) with a smaller diametral gap of 3 mils and one rod (Rod 2) with a larger diametral gap of 15 mils, respectively.

4 6-2

Zircaloy-2 clad with identical dimensions was used for all rods. The clad inner diameter was about 0.429 inches and the wall thickness was 34 mils, minimum.

, , Three types of UO2 pellet fuel were included among Rods 1-6. The standard fuel for the test was a 95% TD fuel that was stable with respect to in-reactor densification. The in-reactor densification expected for this fuel is 0.4% TD, based on the ex-reactor resinter tests made on archive pellets. This fuel was used in Rods 1-4. The design variables of interest among the rods containing this identical fuel type, therefore, were the larger and smaller gaps of Rods 2 <

and 3, respectively, and the alternate xenon fill gas of Rod 4.

Rod 5 was identical to Rod 1 except for 92% TD stable fuel pellets. The in-reactor densification expected for this fuel type is also about 0.4% TD.

Rod 6, like Rod 5, also contained 92% TD fuel type. But this fuel is expected to densify by about 3% TD in reactor and, therefore, is considered to be an unstable fuel type.

An unusual characteristic of all three fuel types used in the IFA-432 rods is '

the large radial variations noted in the initici grain size and in the initial porosity and pore size distribution of the fuel pellets. These radial variations in pellet microstructure have been attributed to oxygen activity gradients that existed during pellet sintering. The oxygen activity gradients could have been eliminated by bubbling the hydrogen gas used as a sintering atmosphere through water before it entered the sintering furnace. The processes used by C-E and most other fuel fabricators avcid the activity gradients and, thereby, produce fuel pellets with radially and axially uniform characteristics of grain size and porosity.

6-3 1

The 95% TD stable fuel (in Rods 1-4) had a grain size of 22 um at the pellet periphery and a large grain of typically 70 um average at pellet center ,

and mid-radius locations. Unusual plate-like patterns also existed within the .

larger grains which are possibly faults in the U02 crystal structure.

The 92% TD unstable fuel (in Rods 5 and 6) exhibited a grain size range after fabrication of 2 pm at the pellet periphery to 8 pm at the pellet centerline.

Randomly distributed regions characterized by greatly reduced grain size were scattered throughout the pellet (s) examined. These regions are believed to be incompletely sintered powder granules. c Finally, the grain size of three of four fuel pellets characterized as 92% TD

~

stable fuel was about 5 pm at the pellet centerline, increasing to about 18 um at the pellet periphery. A fourth pellet, however, had a grain-size of about 50-60 pm at the pellet centerline, decreasing to about 18 um at the periphery. These observations suggest that a significant variation in grain size may have been present from pellet to pellet in this fuel type.

FATES 3 results. -

The FATES 3 code was not re-programmed to consider an initial variation in grain size within single fuel pellets in a fuel rod, although the code will predict radial changes in grain size at each axial node when fuel temperatures and '--

operating times are sufficient to grow grains. Since a single grain size is not characteristic of the fuel pellets in each rod prior to irradiation, FATES 3 results are presented for the range of grain sizes associated with es .h

~

of the three fuel types described above.

E e

6-4

Table 6A-1 provides a summary'of the major design differences among the rods previously described. Included in the table are the lifetime-peak powers achieved by each rod at the thermocouple locations early in the irradiation of IFA-432. These powers range from 13.7 to 15.2 kw/ft for the upper thermocouple location, and from 10 to 11.1 kw/f t for the lower thermocouple location. The

~

powers of all rods decreased gradually over the irradiation. The final powers at the upper thermocouple locations were in the range of 10.4 to 11.6 kw/ft for rod-averaged burnups greater than 26,000 MWD /MTU.

Pressure transducer data are available for Rods 1 and 5 for the entire test and through most of the test for Rod 6. Only four of the original 12 thermocouples, however, were still functioning at the end of the test. Table 6A-1 includes a summary of the local burnups to which transducer and thermocouple data are available for each rod.

The IFA-432 rods operated throughout the test with a significant gradient in their axial power shape. This shape has been modeled in FATES 3 using 16 axial nodes, with thermocouples located at nodes 2 (lower TC) and 14 (upper TC). At BOL (beginning-of-life) the ratio of local power to rod-averaged power for nodes 2 and 14 were 0.82 and 1.12, respectively. The axial shape flattened out ,

slightly by EOL (end-of-life). Separate axial shapes were used in FATES 3 to represent the time-averaged shapes representative of the periods December 1975 to January 1978 and January 1978 through May 1980. *<

FATES 3 oredictions of fuel centerline temperatures for the TC locattor.s of Rods 1-6 are compared with the TC measurements in Figures 6A-1 through 6A-6. These data are replotted as ratios of predicted-to-measured. temperatures in Figures 6A-7 through 6A-9.

6-5

The data presented in these figures for Rods 1-4, all of which contained the 95% TD stable fuel, include FATES 3 predictions for initia1 fuel grain sizes of 22 and 70 pm. Similarly, the data presented for Rod 5, the only rod which contained the 92% TD stable fuel, include FATES 3 predictions for initial fuel grain sizes of 5 and 18 um. For Rod 6, however, FATES 3 predictions were -

generated for an 8 pm grain size only. FATES 3 so overpredicts measurements of both temperature and pressure that a sensitivity to a lower grain size assumption for the 92% TD unstable fuel in this rod is of no practical interest.

Among Rods 1-4, the greatest overprediction of temperature is given by FATES 3 <

for Rod 2, which had the large initial gap of 15 mils (see Figures 6A-3, 6A-7 and 6A-8). FATES 3 gives considerable overpredictions of the temperature at the lower TC location of this rod (there are no upper TC data for Rod 2) through the reported local burnup of 21,400 mwd /MTU, regardless of the assumed value for initial fuel grain size. Fission gas release values prediced by FATES 3 for the end of the reported irradiation of Rod 2 are 24.3 and 8.4% for the 22 and 70 pm fuel grain size assumptions, respectively.*

FATES 3 generally overpredicts fuel temperatures at the upper TC locations of Rods 1 and 3 for the fuel grain size assumption of 22 pm (see Figures 6A-1, 6A- -

2, 6A-4, 6A-7 and 6A-8). Note that Rod I had the standard 9 mil initial gap, while Rod 2 had the relatively small 3 mil initial gap. Generally, predicted temperatures are high for the lower TC locations of these two rods through the first four years of irradiation, and then become low. FATES 3 predictions of -

fission gas release for Rods 1 and 3 are relatively low compared to the values

~

given for Rod 2 above, and this is believed to be one reason why the late in life temperatures predicted for these rods are considerably below the

• FATES 3 qas release predictions are .,5ven for information but cannot be compared since measured gas release values are not available for the original six IFA-432 rods.

6-6

temperature predictions for Rod 2. The gas release predictions for Rods 1 and 3 are 5.2% and 0.9% release in Rod 1 for assumed fuel grain sizes of 22 um and 70 um, respectively; and 1.7% and 0.4% release in Rod 3 for the same two fuel grain sizes. Because the internal pressure in Rod 1 is underpredicted for an assumed grain size of 70 um, slightly over-predicted for an assumed grain size of 22 pm, based on the available pressure transducer data for this rod (the

~

pressure data for Rods 1, 5 and 6 are discussed later), the amount of gas released in Rods 1 and 3 are inferred to be better predicted by FATES 3 for the

~

smaller grain size assumption.

A reason why FATES 3 predictions of fuel temperatures at the lower TC tend to be <

low for Rods 1 and 3 (and for Rod 5, as discussed below) toward the end of the

reported period is the use of a single axial power shape in FATES 3 to describe the irradiation period from January 1978 through May 1980. For this period, the axial shape varied considerably, with the typical power at t

1e lower TC being a fractional 0.78 of the power at the upper TC. Near the end of this period e' specially, however, the fraction was as high as 0.81 for Rod 1, 0.86 for Rod 3 and 0.88 for Rod 5. Had a more detailed axial shape been modelled in FATES 3, the predictions of temperatures at the lower TCs in these rods would have been higher by about 25-100 C, respectively, at the end of the reported period. -

The BOL fuel temperatures reported for the upper and lower TC locations of the xenon-filled Rod 4 are considerably above the temperatures reported for the other rods which also contained the 95% TD stable fuel and are discussed -

above. Unfortunately, both TCs failed in Rod 4 within the first month of operation. As shown in Figures 6A-1 and 6A-2, FATES 3 provides overpredictions

~

of the data (which is limited) for both TCs.

6-7

FATES 3 gives high predictions of the fuel temperatures of Rod 5, regardless'of fuel grain size assumption, except for some of the lower TC data near the end of the reported period (see Figures 6A-5 and 6A-9). The. fission gas release values predicted by FATES 3 for the end of the reported irradiation of Rod 5 are 15.2% and 7.1% for the 5 t.m and 18 t.m fuel grain sizes, respectively. The

. actual (unmeasured) gas release is likely to be closer to the value predicted for the grain size of 18 um since the rod internal pressures predicted by

. FATES 3 are in better agreement with the measured pressure data for this grain size assumption (see pressure discussion below).

Due to the large assumed densification expected for the fuel in Rod 6, FATES 3 gives very high predictions of fuel temperatures at both TC locations throughout the irradiation of this rod (see Figures 6A-6 and 6A-9). For a fuel grain size assumption of 8 um, FATES 3 predicted the fission gas release to be 27.3% in Rod 6 at the end of the reported period. The (unmeasured) gas release likely is overpredicted, since the rod internal pressure is overpredicted significantly for this rod (see pressure discussion below).

FATES 3 analyses of Rods 1, 5 and 6 included selected timesteps throughout the irradiation of each of these rods where essentially zero power was simulated and coolant temperatures were specified at values between 160 and 460 F, according to the reported temperatures asssociated with the available pressure transducer measurements made during non-power rcactor operation. At BOL, the rod internal pressures at 302 F were about 21, E2 and 26 psia in Rods 1, 5 and

~

6, respectively. By the end of the reported period, the measured pressures had increased to 121 and 155 psia in Rods 1 and 5, respectively, when measured at

- 443 F. The last available data for Rod 6 indicated that this rod had a pressure about 20% higher than the pressure of Rod 5 toward the end of the

. reported irradiation. All three rods showed pressure reductions of about 15%

over the first 4000/ mwd /MTU in excess of reductions attributable to densification and halium absorption, suggesting that some helium leakage occurred into thermocouple cables during this period (N .

6-8

Comparisons of measured and FATES 3-predicted rod internal pressures for hot, zero-power operation are presented in Figures 6A-10 through 6A-12 for Rods 1, 5 and 6. In addition, the ratios of predicted-to-measured pressures for hot, zero-power operation are plotted for these rods in Figure 6A-13. The zero-power pressures in Rod I are generally underpredicted by FATES 3 by a factor of 2 or more at high burnup for an assumed fuel grain size of 70 cm, but are only slightly overpredicted at high burnups for an assumed fuel grain size of 22 pm. This demonstrates the significant sensitivity of the pressure calculation in an unpressurized rod to the predicted gas release, considering the difference in the predicted release values for the two FATES 3 analyses is <

only 4.3% release. The zero-power pressures in Rod 5 are overpredicted by FATES 3 by about a factor of 2.5 at high burnup for an assumed fuel grain size of 5 un, but only slightly over-predicted at high burnup for an :ssumed fuel grain size of 18pm. The zero-power pressures in Red 6 are overpredicted by FATES 3 by a factor of approximately three at high burnup for the single analysis made using an assumed fuel grain size of 8 um.

Comparisons of measured and FATES 3 predicted rod internal pressures for high power operation are presented in Figures 6A-14 through 6A-16 for Rods 1, 5 and

6. In addition, the ratios of predicted-to-measured pressures for high power operation are plotted for these rods in Figure 6A-17. Trends noted above for hot, zero-power operation also appear for high power operation. An important observation, however, is that FATES 3 predictions are more conservative for high power operation than for hot, zero-power operation. This is a positive Indication of the conservatism present in the FATES 3 calculation of hot

. internal pressure.

W 6-9

In summary, predictions of fu'le centerline temperatures and rod internal pressures are very sensitive to fission gas release predictions for nonpressurized fuel rods (such as these Halden rods). This sensitivity is due to the relatively rapid degradation of gap conductance which is additionally attributed to the uncertainties of reported power histories and to the unusual and significantly large variation in the grain size of the fuel types fabricated for these rods. In general, however, it appears that for those

~

cases where the predicted fission gas release is more than about 5"., FATES 3 provides overpredictions of fuel centerline temperatures and rod internal pressures. The overpredictions are greatest during the first 5000 MWD /MTU, <

especially for the unstable fuel type in Rod 6. This may in part be due to the underprediction of gap closure in FATES 3. The ratios of predicted-to-measured rod internal pressures early in life as shown in Figures 6A-13 and 6A-17 are large because these are ratios of low pressures to begin with. Also, some early leakage of fill gas into thermocouple cables occurred (4) which offset the expe'cted pressure increase concommitant with fission gas release and fuel rod void volume changes with burnup. Considering these factors, however, it still can be concluded that FATES 3 provides overall reasonable predictions of the temperatures and rod internal pressures for the stable fuel types included in the IFA-432 rods over the range of burnups achieved. -

(1) C. R. Hann et al., " Test Design, Precharacterization, and Fuel Assembly Fabrication for Instrumented Fuel Assemblies IFA-431 and IFA-432,"

Battelle Pacific Northwest Laboratories Report NUREG/CR-0332 (PNL-1988), -

November 1977. Transmitted as enclosure to J. C. Voglewede (NRC letter to J. C. Ennaco (C-E) dated June 17, 1981.

(2) C. R. Hann et al., " Data Report for the NRC/PNL Halden Assembly IFA-432,"

Battelle Pacific Northwest Laboratories Report NUREG/CR-0560 (PNL-2673), _

August 1978. Transmitted as enclosure to J. C. Voglewede (NRC) letter to J. C. Ennaco (C-E) dated June 17, 1981.

(3) E. R. Bradley et al., " Data Report for the NRC/PNL Halden Assembly IFA-432; April 1978-May 1980," NUREG/CR-1950 (also PNL-3709), April 1981.

(4) E. R. Bradley et al., "An Evaluation of the In-Pile Pressure Data from Instrumented Fuel Assemblies IFA-431 and IFA-432," PNL-3206 (NUREG/CR-1139),

October 1979.

6-10

1 i-TABLE 6A-1 k Sumary of Design and Operating Variables  !

for the Original Six Rods of t l IFA-432 and the Availability of ,

j Measurement Data From Fuel Thermocouples and Rod pressure Transducers j..-('

Burnups to Which Data are Lifetime- Available from Rod Instrumentation mwd /MTU Fuel Pellet Initial Peak Pressure Fuel Fuel Average Grain Cold Powers at Upper TC Lower TC Data to Rod Density Stability to Size Range, Fill Cap. Upper & Lower Data to Local Data to Local Rod-Average No. 1 TD Densification um Gas Mils TCs, kw/ft* Burnups of: Burnups of: Burnups of:

I 95 stable 22-70 He 9 14.3, 10.4 10,100 b D

, g 22.100 26.100 2 95 stable 22-70 He 15 13.7, 10.0 no data 21,400b no transducer stable D 3 95 22-70 He 3 15.2, 11.1 21,700 21,400 .no transducer ,

4 96 st.ble 22-70 Xe 9 14.9. 10.9 1.300 900 no transducer stable C b 5 92 5-18 He 9 14.6, 10.6 2.900 22,200 26,200D m 6 92 unstable 2-8 He 9 14.3, 1.0.4 8,000 19,200 18,900

' Occurred near BOL at an assembly-a reraged burnup of about 1300 mwd /MTU. BOL ratios of loca'j powers to robaveraged powers for upper and lowerTC locations are 1.12 and 0.82, respectively.

b instrument was still functioning at the end of the reported period (May 1980).

1. Some pellets s.ay have grains up to 60 pm local average. See discussion in text of response.

? e

, d A

Fig. 6A-1 Comparison of Heasured and FATES 3 PredictJ Centerline Fuel Temperatures for the Upper' TC Locations in Rods 1 and 4 of IFA-432 (957. TD Stable, 9 mil gap)

IHMARIE, C M

Heasured Rod 1 ~ Predicted 22 pm Fuel Grain Size (He till i

~B Standard) _,

Predicted 70 pm Fuel Grain Siza

~S E -

G Heasured

~

[] Predicted, 22 pm Fuel Grain Size

m. - (Xe 1) l C _

t O Predicted. 70 pm Fuel Grain Size

_f ~  %

a

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• gg i i n .i . i. i. i. e i i i e i i .

DEC MARCH JUNE SEP DEC MARCH JUNE SEP DEC MARCH JUNE SEP DEC MARCH JUK SEP DEC MARCH JUNE

'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 TIE L

I ^ _lk - ---__ _ _-_ ______ _ _ _ . _

Fig. CA-2. Comparison of iteasured and FATES 3 Predicted Centerlire Fuel Temperatures for the Lower TC Locations in Rods I and 4 of IFA-432 (951 TD Stable 9 nil gap)

IINUlltiE. [

M Measured Rod I <

- Predicted. 22 um Fuel Grain Size (lle fill . Predicted. 70 pm Fuel Grain Size Standard) ,

3 - G 11easured O Predicted. 22 um Fuel crain Size

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F i g. CA- t. Coraparison of !!easured and FATES 3 Predicted Centerline fuel Temperatures for the Upper and Lower TC Locations in Rod C of IFA-432 (92: TD Unstable, 9 mil gap)

IUPWIN.t zu Heasured Predicted. 8 pm

- -\ Fuel Grain Size

~/

f  % _

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e 1811 e i i , i i e i e i e i i e e e i g i SEP MC MARCH JUNE SEP DEC MARCH JUNE DEC MARCH JUNE '80 '80 DEC MARCH JUNE SEP DEC MARCH JUNE SEP '78 '78 '78 '78 '79 '79 '79 '79

'76 '76 '76 '76 '77 '77 '77 '77

'75 TIE i

a,  ? .' A - - - - - - - - - - - _

Fig. 6A-7. Ratics of FATES 3 Predicted to Measured Center 1tne Fuel Temperatures for the Upper TC Locations of Rods 1 and 3 (95% TD Stable) 18FINM BIN 18 ..

(No Upper TC Data Available for Rod 2 70 pm Fuel Grain Size 3

Lt  :- with the 15 mil gap)

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. g,  ? -*

, -- --A- - ___________

Fig. 6A-8. Ratios of FATES 3 Predicted to Measured Centerline Fuel Temperatures for the Lower TC

• Locations of Rods 1-3 (951 TD Stable)

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'75 '76 '76 '76 '76 '77 '77 '77 TIME ,

I

' I ^ _ __ _ _ _ _ _ _ _ _ _ __

4 Fig. 6A-9. Ratios of FATES 3 Predicted to Measured Centerline Fuel Temperatures for the Upper and Lower TC Locations of Rods 5 (921TD Stable. 9 mil gap) and 6 (92% TD Unstable, 9 mil gap) 19FnNIE Ell 8 li - -

tg 5 Upper TC 15 3- Rod 6 (8 pm Fuel Grain Size)

L2 -

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'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 1

28 .

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Ilg. 6A-II. Corvarlson of fleasured and FATES 3 Predicted Rod Internal Pressures at flot.

Zero Power Conditions for Rod 5 of ITA-432 WIML PEIM, PSIA B -

2

• 7 /

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DEC mRCH JUNE SEP DEC mRCH JUNE SEP KC MRCH JUNE SEP KC mRCH JUNE SEP DEC mRCH JUNE

'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 TIME

' ' ', .* ^

4

/

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v Fig. 6A-12. Corparison of tirasured and FAh53 Predicted Rod Internal Pressures at Hot.

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Zero Power Conditions for Rod 6 of IFA-432

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, . EfC m RCH JUNE SEP DEC 2RCH JUNE SEP DEC MRCH JUNE SEP DEC mRCH JUNE SEP '79 '79 '80 '80 DEC MRCH JUNE '77 '78 '78 '78 '78 '79 '79

'76 '76 '76 '76 '77 '77 '77

'75 TIME a i b n

'I e A .

Fig. 6A-13. Ratios of FATES 3 Predicted to Measured Rod Internal Pressures at Hot. Zero Powes' Condi,tions for Rods 1. 5 and 6 of IFA-432 PE5SE Bill LI _

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i ., . n . . e ..s a . i .i..i .. . i..i .. .. .. . i ro DEC MRCH JUNE SEP DEC MRCH JUNE SEP DEC MRCH JUNE SEP DEC MRCH JL5tE SEP DEC M RCH JUNE

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gg =. n . i n . s.. i . i..i..i ....e i.. .. .... . i DEC MRCH JUNE SEP DEC MRCH JUNE SEP DEC M RCH JUNE SEP MC MRCH JUNE SEP DEC M RCH JUNE

'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 TIME

,)  ? .'

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Fig. 6A-14. Corparison of Measured and FATES 3 Predicted Rod Internal Pressures at High Power Conditions for Rod 1 of IFA-432 WIWL1551E. P514 m -

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MC MRCH JUNE SEP MC MRCH JUNE SEP DEC MRCH JuME DEC MRCH JUNE SEP MC M RCH JUNE SEP

'79 '79 '79 '80 '80

'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 TIME i 3 , ~ ^ _ _ _

l Fig. 6A-15.

Comparison of Heasure2 and FATES 3 Predicted Rod Internal Pressums at Hinh Power Conditions for Rod 5 of IIA-C2 3 M E MElfm.PIM M

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.n . n..e i .i .i .i ...i .... . ....... i MC mRCH JUNE SEP DEC mRCH JUNE SEP KC mRCH JUNE SEP MC ERCH JUNE SEP MC MRCH JUNE

@ '75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 EI -

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'75 '76 '76 '76 '76 '77 '77 '77 '77 '78 '78 '78 '78 '79 '79 '79 '79 '80 '80 Tilf I

1 h ^ -_______

)

Question B. .

Provide additional verification of the FATES-3 fission g'as release model by predicting fission gas release from the high burnup RISO rods (Refs. 5-6).

Densification behavior for these rods should be based on resintering data in .

. Ref. 7. .

. Response B.

Discussion of the RISO tests. -

Five RISO test rods containing U02 fuel pellets and clad with Zircaloy-2 (one rod was clad with Zircaloy-4) are included in this group. Fuel pellet diameters were all about 0.497 inches, the diametral gaps ranged between 8.3 and 9.4 mils, and the cladding thicknesses ranged between 20.9 and 23.2 mils.

All rods were initially filled with helium at atmospheric pressure. Post-irradiation sampling of the rod gases indicated that some air contamination of the initial helium fill gas was present in all rods following fabrication.*

The stack length of fuel pellets measured approximately five inches in all rods. Each stack consisted of from six to nine enriched central pellets, the exact number depending on the particular pellet length used in each rod, and

)

• FATES 3 analyses of these rods included an assumption that the initial fill gas

- contains 11% nitrogen by volume to account for incomplete evacuation of air prior to helium backfilling.

e d

O 6-29

single non-enriched pellets located at each end to minimize local power peaking. Dishes were present on the two ends of all central pellets. The end pellets were flat ended. -

In rods PA29-4 and M2-2C from Irradiation Test 022(1), the central fuel

. pellets were enriched to 2.28 wt% U-235, and had an initial density of 94.7% TD and an initial average grain size of 8pm. In rods P20-1B, M2-2B and T9-38

. from Irradiation Test 013(2) , the central fuel pellets were enriched to 1.45 wt% U-235, and had an initial density of 94.4% TD and an initial average grain size of 25 um. Ex-reactor resinter tests indicated that the densification expected in reactor for these fuels is 1.1% TD(3), <

Fuel density, densification and grain size were not fully characterized for the end pellets. In the FATES 3 analyses,' similar density and densification assumptions were used for the center and the end pellets alike. The grain size of the end pellets, however, was assumed to be 15 um, based on a RISO estimate (4).

The irradiations were conducted in DR3 at RISO. DR3 is a 10 MW heavy-water materials testing reactor. All five rods were irradiated in water cooled rigs at a coolant pressure of about 1030 psia and coolant temperature of about 550F. Normal reactor cycles were about one month long, including a 4-5 day shutdown for exchange of experiments and maintenance. Test 022 was conducted over 46 of these cycles; Test 013 was conducted over 39 of these cycles. j The burnups achieved in, Rods PA29-A and M2-2C in Test 022 exceeded 40 GWd/MTU

. on a rod-averaged basis. Both rods operated at high powers throughout their irradiation. Rod PA29-4 operated at a lifetime-average power of 490 w/cm m

C O

6-30

(14.9 kw/ft) and a lifetime-peak power of 585 w/cm (17.8 kw/ft), determined on a rod-averaged basis. Similarly, rod M2-2C operated at a lifetime-averaged power of 460 w/cm (14.0 kw/ft) and a lifetime-peak power of 552 w/cm (16.8 kw/ft), determined on a rod-averaged basis. Local peak pellet powers of about 610 w/cm (18.6 kw/ft) occurred in both rods early in Test 022.

. The three rods in Irradiation Test 013 operated at considerably lower rod-averaged power than the rods in Test 022, described previously. Lifetime-peak

- powers of 529, 472 and 490 w/cm (16.1,14.4 and 14.9 kw/ft) occurred early in Test 013 in Rods M20-1B, M2-23 and T9-3B, respectively. The powers in all three rods'were typically less than about 335 w/cm (10.2 kw/ft) on a rod-averaged basis for the last third of the test.

Following Test 013, the three rods were ramped to powers not previously experienced since early in their irradiation. Rod M20-1B was ramped to 451 w/cm (13.7 kw/ft) on a rod-averaged basis in Test 050, and held at this power level for 552 hours0.00639 days <br />0.153 hours <br />9.126984e-4 weeks <br />2.10036e-4 months <br />. The final rod-averaged burnup achieved by M20-1B was 29.9 GWd/MTU. Rods M2-2B and T9-3B were further irradiated at relatively low powers in Test 055 and subsequently ramped to rod-averaged powers of 417 and 437 w/cm (12.7 and 13.3 kw/ft), respectively, in Test 067. These ramp powers were held for 667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br />. The- final rod-averaged burnups achieved by rods M2-2B and T9-3B were 28.7 and 30.4 GWd/MTU, respectively.

s

~~-

FATES 3 results. - m

~

FATES 3 analyses of the RISO rods included detailed modeling of the power histories and of the axial power shapes provided for each of the fuel

- rods (1),(2). The detail of the axial modeling was sufficient to distinguish between differences in design and power histories of the end and

. center fuel pellets.

FATES 3 predictions of ges release for the RISO rods are compared against <

measured values in Table 68-1. ,This table also includes a summary of some major design and operating variables associated with the test rods.

6-31

C-E considers the benefit of a smoother fuel pellet in a nonpressurized rod to be a diminishing effect as burnup increases, such that at the high burnups typical of the RISO rods, the fuel in these rods would be expected to operate at temperatures typical of fuel with initially rougher inner surfaces. The Ross-Stoute determined sensitivity of gap conductance to surface roughnesses -

was restricted to U0 Because reported fuel surface roughnesses l (2)in (2) are thea as-fabricated about state.

factor of two lower than the roughnesses of fuel typical of the FATES 3 data base, including C-E commercial fuel, additional FATES 3 predictions have been provided for a higher roughness value. Since the gap conductance model in FATES 3 is based on the work of Ross- <

Stoute, the gap conductance, and hence the fuel temperatures and fission gas release, are all very sensitive to the roughness assumption in a nonpressurized fuel rod. This is apparent in the comparison of FATES 3 predictions for the two roughness assumptions for each rod.

The measured fission gas release values for rods PA29-4 and M2-2C from Test 022 are reasonably predicted by FATES 3 for a surface roughness typical of commercial fuel pellets, as indicated in Table 6B-1. The measured values for rods M20-18, M2-2B and T9-3B from Test 013, however, are considerably greater than the FATES 3 predictions for these rods, regardless of the surface roughness -

assumption. It is important to note that the fuel temperatures predicted by FATES 3 for these latter three rods are considerably below the tempe atures typical of the former two rods from Test 022, especially at higher bu"nups. '}

The FATES 3 temperature predictions are believed to be entirely consistent with -

the differences in the reported power histories among the rods, as illustrated in Table 6B-2. These differences in predicted temperatures are the major factor contributing to the differences in gas release predictions by FATES 3 _

~

among the RISO rods (i.e., the fuel initial grain size, which is the only significant design difference among the rods, is a secondary factor due to the [

development of early grain growth in all rods). ,

The results of the post-irradiation metallographic examination of the fuel pellets, however, do not indicate any significant difference in operating temperatures between rods of the different test series. The size of the central void and the radial extent of columnar grain formation are not very different among the rods.

6-32

1 Possible uncertainties in the' assignment of pawer levels during the complicated power histories, or even some unreported short-term power excursions during Test 013, may explain the higher than expected fission gas release in this test series.

The two rods from Test 022, however, provide good correspondence between the

, measured values of temperature and gas release with the predictions of FATES 3.

(1) P. Knudsen and C. Bagger, " Power Ramp and Fission Gas Performance of Fuel <

Pins M20-18, M2-2B and T9-38," Riso National Laboratory (Denmark) Report Riso-M-2151, December 1978. Transmitted as enclosure to J. C. Voglewede (NRC) letter to J. C. Ennaco (C-E) dated June 17, 1981.

(2) C. Bagger, H. Carlsen and P. Knudsen, " Details of Design, Irradiation and Fission Gas Reiease for the Danish UO2-Zr Irradiation Test 022", Riso National Laboratory (Denmark) Report Riso-M-2152, December 1978.

Transmitted as enclosure to J. C. Voglewede (NRC) letter to J. C. Ennaco (C-E), dated June 17, 1981.

(3) P. Knudsen (Riso) letter to J. C. Voglewede (NRC) dated October 7,1981.

Transmitted as enclosure to J. C. Voglewede (NRC) letter to C. E. Beyer (PNL) dated October 26, 1981 (copy to J. Ennaco, C-E).

(4) Telex, P. Knudsen (Riso) to H. R. Freeburn (C-E), dated February 3, 1982.

/

B

=

C S

6-33

TABLE 6B-1 A Comparison of FATES 3 Predictions and .

. Measured Values of Fission Gas Release for FiveNonpressurizedRISdTestRods Test Series 022 022 013, 050 013, 055, 067 013, 055, 067 Rod Identity PA29-4 M2-2C M20-1B M2-2B . . '.T9-38 Diametral Gap, mils 9.4 9.4 8.3 8.7 9.4 Wt. Fraction of Fuel in End Pellets 0.14 0.16 0.21 0.21 , 0.21 Fuel Wt.% U-235 Center Pellets 2.28 2.28 1.45 1.45 1.45 End Pellets 0.72 0.72 0.72 0.72 0.72

$fFuelGrainSize, um Center Pellets 8 8 25 25 25 End Pellets 15 15 15 15 15 Lifetime-Peak, Rod-Averaged Power, kw/ft 17.8 16.8 16.1 14.4 14.9 E0L Rod-Averaged Burnup, GWd/MTU 45.4 43.1 29.9 28.7 30.4 Measured Fission Gas Release, % 48.1 35.6 38.9 29.5 29.6 FATES 3 Predicted Release, % 35.2 30.5 9.4 5.3 7.7 FATES 3 Predicted Release for 44.9 40.2 12.3 7.2 10.2 Commercial Fyel Surface Roughness, %

• Reported roughnesses are in the range of 4-6x10-5 in (RMS) for the RISS fuel pellets.

Commercial fuel roughnesses are typically about 8x10-5 in (RMS).

e i .6i  ! '

n __ _ _ _ _ - - - - - - - - - - - - _ - - - .

0 s

• i TABLE 68-2 ,

Relative Differences in Powers and Predicted Fuel Centerline Temperatures Among the RISO Rods Early Peak Differences a late Peak Differences h (at ?-4 Mwd /kgu) (at 28-30 Mwd /kgu)

Test Rod c c Series- . Identity Local kw[tj AFk U

Local kw/ft AFk 022 PA29-4 18.7 0 (base) 17.9 0 (base) 022 M2-2C s 17.7 -260 17.0 -240 013, 050 M20-18 16.9 -360 13.7 -1070 013,055, M2-28 15.1 -670 12.7 -1410 -

067 .

013, 055, T9-3B 15.8 -460 13.4 -1200

, 067 h

a At axial center of each rod. For the 022 Test les. this was period number 130(I) . For.

the 013 Test Series, this was period number 121 .

b At axial center of each rod. Fo 2)he022TestSeries,thiswasperiod158(I) For the 050 .

Test Series, this'was period 169 . For the 067 Test Series, this was period 180.

c FATES 3 relative predictions of fuel centerline temperatures at approximately the axial center of each rod.

? -

m i 8, n

Question C.

Provide additional verification of the FATES 3 fission gas release model by predicting fission gas release from rod RJL (Ref. 8). Because the resintering data are not available for this prediction, the predicted release value should -

be provided as a function of assumed final resintered density (94-98% TD).

Response C.

Discussion of the RJL test. -

Rod RJL is a prepressurized (450 psia helium) PWR rod containing solid pellet UO 2 fuel of 6% enrichment 235 U The fuel pellet OD is 0.3669 in. and the as-fabricated diametral gap between the U02 fuel column and the Zircaloy-4 CWSR clad is 6.3 mils. The active fuel length is about 7 ft. The fuel pellets have an initial density of 94.22% TD, have an initial grain size of 9.3 um, and are dished on both ends. The rod was irradiated in pressurized coolant (1990 psia) for about 32,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at rod-averaged powers ranging from about 2 to 10 kw/ft. The end-of-life burnup calculated by FATES 3 for the RJL rod, based on the history provided(l), is 52,800 mwd /MTV. .

FATES 3 results. -

~

The power history of the RJL rod has been modeled in FATES 3 using the detailed i power history provided(l). This history includes: 47 individual time -

steps, each having separately specified values of rod-averaged power; and ten unique axial shapes, each with local relative powers given for 15 axial nodes.

Four FATES 3 predictions are tabulated below for the end-of-life fission gas release in the RJL rod for various assumptions of the terminal fuel density attributable to in-reactor densification.

c

~

Fission gas release measurements are not reported {l) and are not otherwise available to C-E; therefore, a comparison between predictions and measurements is not included.

6-36

. e . .. .

As-Fabricated U0 2 Assumed Terminal UO 2  % Fission Gas Release Density, % TD Densi,ty, % TD* -

Predicted by FATES 3 94.22 95. 2.9 94.22 96. 2.7 94.22 97. 2.5 94.22 98. 2.4 <

• at 4000 mwd /MTU local fuel burnup (1) PWR Rod RJL, NRC Fuel Performance Data Base, dated February 1981.

Transmitted as enclosure to J. C. Voglewede (NRC) letter to J. C. Ennaco (C-E) dated June 17, 1981.

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