ML20042A016
| ML20042A016 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/22/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Rich Smith VERMONT YANKEE NUCLEAR POWER CORP. |
| Shared Package | |
| ML20042A017 | List: |
| References | |
| NUDOCS 8203220559 | |
| Download: ML20042A016 (11) | |
Text
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Distribution:
Do::ket File NSIC NRC FDd ACRS--10 Local POR AF0D Docket No. 50-271 ORB #2 Reading Gray File O. Eisenhut J. Vogleweda OELD FEB 2 21982
/'7 Mr, Robert L. Smith ooney Licensing Engineer S. Norris
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@M Vemont Yankee Nuclear Power
/
Corporation
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A 1671 Worcester Rocd i
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FreminghAm, MA 01701 m
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Dear Mr. Smith
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OFFICIAL RECORD COPY uso m im - m
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Mr. Robert L. Smith cc:
Mr. W. F. Conway Public Service Board President & Chief Operating Officer State of Vermont Vermont Yankee Nuclear Power Corp.
120 State Street 411 Western Avenue Montpelier, Vermont 05602 Drawer 2 West Brattleboro, Vermont 053'01 W. P. Murphy, Plant. Superintendent Vermont Yankee Nuclear Power Corp.
Mr. Louis H, Heider, V.P.
P.O. Box 157 Vermont Yankee Nuclear Power Corpo.
Vernon, Vermont 05354 25 Research Drive Westboro, Massachusetts 01581 Vermont Yt.1kee Decommissioning
...Al.liance.
John A. Ritscher, Esquire 5 State Street Rope & Gray Box 1117 226 Frankline Street Montpelier, Vermont 05602 Boston, Massachusetts 02110 Brooks Memorial Library Honorable John J. ' Easton 224 Main Street Attorney Genera'l Brattleboro, Vermont 05301 State of Vermont 109 State Street Resident Inspector Montpelier, Vermont 05602 c/o. U.S. NRC P.O. Box 176
~
Vermont Yankee Decommissioning Vernon, Vermont 05453 Alliance 53 Frost Street Ronald C. Haynes Brattleboro, Vermont 05301 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. E. W. Jackson 631 Park Avenue Manager of Operations King of Prussia, PA 19406 Vermont Yankee Nuclear Power Corp.
411 Western Avenue Drawer 2 West Brattleboro, Vermont 05301 Raymond N. McCandless Vermont Division of Occupational
& Radiological Health Administration Building 10 BaldWin Street Montpelier, Vermont 05602 New England Coalition on Nuclear
-s Pollution Hill and Dale Farm R.D. 2, Box 223 Putney, Vermont 05346
QUESTIONS ON YAEC-1249/1265 (FROSSTEY) 1.
Application of YAEC-1249 (Ref. 1) and YAEC-1265 (Ref. 2) in Licensing Analyses.
From the list of licensing anaysis requirements provided below, please identify which applications will be met with FROSSTEY.
Where the licensing analysis requirements will not be met with FROSSTEY, 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 load 1ng limits.
2.
Strain fatique limits.
3.
Fretting wear limits.
4.
Oxidation, hydriding, and the buildup of corrosion products (crud) limits.
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.
D.
Fuel rod failure or initialization of other analyses used t o.
-s calculate fuel rod failure.
1.
Overheating.
a.
Departure from Nucleate Boiling (DNB).
b.
Fuel enthaply.
c.
Fuel centerline melting ( s t e a,d y-s t a t e and transient).
2.-
Pellet / Cladding Interaction (PCI).
3.
Hydriding limits.
4.
Cladding collapse.
5.
Bursting.
6.
Mechanical fracturing.
7.
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.
2.
FROSSTEY Submodel Documentation.
Using the list of models provided in Section II.C'.3(a) of Standard Review Plan 4.2 (Ref. 3), indicate where the appropriate FROSSTEY model description can be found in Y AEC-1249 o r Y AEC-12 6 5.
If the model is used in FROSSTEY, but is not described in either report, please provide a description of the model.
3.
Differences Between FROSSTEY and the previous YAEC Fuel Performance Code.
To help us identify changes in operating limits that might be' expected as a result of the application of Y AEC-1249 and Y AEC-1265, describe and quantify the differences between FROSSTEY predictions and those calculated by the previous YAEC code, GAPEX (Ref. 4), for the applications identified in your response to Question 1.
4 FROSSTEY Input.
A.
Using the list of input parameters in Appendix C of YAEC-1249, state whether each input value used is best-estimate or conservative, or otherwise results in a best-estimate or conservative calculation (we would expect that only a few input parameters are intentionally conservative).
If an input parameter is.known to be conservative, please quant ^1fy 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.
B.
Provide the input values for a typical case of each of those applications of FROSSTEY identified in response to Question 1 in sufficient detail to permit staff audit calculations.
e
5.
FROSSTEY Output.
A.
Provide a list of output parameters from the FROSSTEY code.
Identify the end use of each output value produced.by FROSSTEY (e.g.,
another code, design criteria, information only) for each application identified in response to Question 1.
B.
State whether the value of each output parameter is best-estimate or conservative.
If conservative, quantify the margin of conservatism and the basis for the margin used.
If the margin varies from-application ro application, provide this information for each different application.
C.
Using the input values described in your response to Question 4 above, provide the corresponding FROSSTEY output values in sufficient detail to permit a comparison with staff audit calculations.
6.
Code Conservatism.
Explain why the magnitudes of the conservatisms identified in response to Question 5 are appropriate for each application identified in response to Question 1.
It is expected that certain output parameters, particularly fuel temperatures used for LOCA initial conditions, should exhibit an appropriate level of conservatism, based on statistical analysis of experimental-vs.-predicted fuel temperatures.
7.
Model Verification.
A.
The FROSSTEY code
's derived from the GAPCON-2 code and utilizes essentially the same experimental data (Ref. 5) originally used to verify GAPCON-2.
Many of the original references from which these data were taken do not quantify the amount of fuel densification which occurred during irradiatio'n.
What densification assumptions where made for these data used in the verification and qualification of FROSSTEY?
B.
Data from only one commercial LWR fuel rod (Rod JBYO97 from Maine Yankee Core 1 Batch B) are included in the FROSSTEY verification study.
The results show a significant underprediction of both fuel temperatures and fission gas
~
release by the FROSSTEY code unless large uncertainties in rod power levels are assumed.
In light of the fact that this rod was selected as part of the EPRI/CE code evaluation program (Ref. 6), and that this rod was obtained from a reactor to be analyzed with the FROSSTEY code, please indicate why these data should not be accepted as an indication of FROSSTEY's inability to predict fuel temperatures and fission gas release in commercial fuel rods.
C.
Provide additional fuel temperature'model verification by predicting fuel centerline thermocouple and rod internal pressure response for IFA-432 (Refs. 7-8).
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).
D.
Provide additional verification of the FROSSTEY fission gas
~
release model by predicting fission gas release from the high burnup Riso rods (Refs. 9-10).
Densification behavior for these rods should be based on resintering data in Reference 11.
E.
Provide additional verific'ation of the FROSSTEY fission gas release model by predicting fission gas release from rod RJL
( Re f. 12).
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%
T.D.).
8.
Fuel Centerline Temperature Predictions.
Figures 2.6, 2.7, 2.8, 2.11 and 2.12 of YAEC-1265 show a sharp upward bend in fuel centerline temperature as a function of rod power level.
This trend is not seen in the predictions of GAPCON-2 (Refs. 5-13) or other fuel performance codes with which we are familiar, nor is such a trend apparent in fuel performance data."
What is the reason for the sharp change in the slope of the curves?
9.
Fuel Stored Energy.
A.
The GAPCON-2 code, from which FROSSTEY was derived, calculates fuel stored energy by two methods.
Is fuel stored energy calculated by FROSSTEY?
If so, what method is used in licensing applications?
4
. B.
It is not clear that the effect of cracking on fuel thermal conductivity results in a conservative estimate of stored energy.
This modeling approach may increase the centerline temperature; however, it may also result in decreased fuel surface temperatures and lower calculated values of stored energy.
Reference 47 in YAEC-1249, which is used to characterize the conservatism in using a thermal conductivity degradation model, does not relate to the conservatism of such a model in a fuel performance code used for stored energy predictions.
Rather, this unpublished reference from PNL emphasizes that the reduction in thermal conductivity associated with pellet cracking biases, an_d_inc.reases the uncertainties in estimated stored energy derived from fuel temperature measurements.
Please demonstrate, in a quantitative manner, that the cracked fuel conductivity model in FROSSTEY indeed results in a conservative prediction of fuel stored energy.
- 10. Helium Production and Release.
Postirradiation puncture data (Ref. 14) from an e,xperimental unpressurized Halden fuel rod (Rod 8 of IFA-432) irradiated to 22,000 mwd /MtU have shown more helium present ( =25% ) than could be expected as a result of the initial fill gas introduced during fuel rod fabrication.
Similar behavior has been reported 'Refs. 9-10) for the Riso rods mentioned previously.
Does the FROSSTEY code take into account helium production and release?
If not, would 'the inclusion of such a model have a significant effect on calculated end-of-life rod pressures?
- 11. Fuel Grain Growth and Restructuring.
Section 6.5 of YAEC-1249 describes grain growth and restructuring models in FROSSTEY.
From postirradiation examination data, it is known that most commercial LWR fuels operate at temperatures below that required for grain growth and restructuring.
Are these models nevertheless used in licensing analyses performed by FROSSTEY?
- 12. Cladding Corrosion.
Equation 3.8b in Section 3.4 of YAEC-1249, which describes cl a d d.i n g -s corrosion, does not appear to agree with the corresponding equation in MATPRO-11 (Ref. 15).
Are these two e g at a t i o n s equivalent?
If not not, why not, since MATPRO-11 is referred to as the source of this model.
e
- 13. Cladding Growth.
Section 4.5.2 of YAEC-1249 states that the lower bound of an irradiation cladding growth model developed by EPRI is used in FROSSTEY.
Is the EPRI model appropriate for the cladding types used in Yankee Atomic ~ plants?
In addition, the lower bound may be cons.ervative for calculating rod internal pressures, but not rod clearance with the top grid plate of the assembly.
Explain why the choice of lower bound is appropriate for all applications of FROSSTEY.
Also please provide a. figure showing cladding growth as a function of time for fluences typical of YAEC reactor operation, since we have ha_d difficul_ty i,n c_omparing the FROSSTEY model with other irradiation ~grokth' mode,ls.with which we are familiar.
- 14. Rod Internal Void Volume.
It appears that a factor of pi (3.14) is missing in the calculation of rod internal void volume (Equation 5.25 of Section 5.4 of YAEC-1249).
Please confirm the validity of this equation in the document and the FROSSTEY code.
- 15. Fission Gas Release Model.
A.
Section 6.9 of YAEC-1249 states that the Beyer-Hann fission gas release coefficients are adjusted according to Equation 6-24.
However, Table 6.1 (where. Equation 6-24 is presented) indicates application at low temperatures.
Is Equation 6-24 actually applied to the Beyer-Hann coefficients at high temperatures?
B.
Please provide a figure, similar to Figure 6.6 or 6.7 in Y A'E C-12 4 9, of fission gas release fraction versus burnup for the combined low and high temperature release models.
The information should be presented for several isothermal conditions (800, 1000, 1200, 1400 and 1600 C) and one density (95% T.D.).
C.
Section 6.9 of YAEC-1249 also states that a new computational structure has been devised to allow an unlimited number of time-steps in a power history analysis using Soulhier and Notley's, approach (Ref. 16).
From the description provided, it is not clear how this is accomplished.
Please provide, by additional description or examples, an explanation of how this is done and why the technique results in conservative predictions for high release (>10%) fractions.
- 16. Gadolinia Properties.
It appears that the FROSSTEY code will be used to calculated the behavior of urania-gadolinia fuels.
How will the material properties in FROSSTEY be modified to account for the effects of gadolinia additions to urania?
.Also, what is the source of the relations used?
- 17. Burnup Distribution.
A.
Is the radial power distribution (flux depression) in FROSSTEY modified as a function of burnup?
If so, how is this accomplished?
B.
Are radial as well as axial burnup variations included in the local calculation of fission gas release and fuel swelling for each fuel node?
That is, is burnup calculated individually for each radial node and used in calculating the release and swelling for that node?
- 18. Gap Conductance.
A.
Table 5.1 in Section 5.2 of YAEC-1249 presents three sets of constants used in the calculation of solid gap conductance and are based on old (pre-1964) data.
Have these constants been verified against more recent data, such as that (Refs. 17-18) used to verify GAPCON-2?
Which set of constants will be used in licensing applications?
B.
The Lloyd mod.?1 (Ref. 19) or temperature jump distance model has been shown to give nonconservative (low) values relative to other accepted gap conductance models and also relativa to measurements.
In GAPCON-2, the Lloyd model values are multiplied by 1.8 in both open and closed gap regimes and this modification is accepted as predicting realistic, if not conservative, values.
In the FROSSTEY code, the Lloyd value is not modified in the open gap regime, except for the very-near-contact condition.
What is the effect of not using the 1.8 value in the FROSSTEY gap conductance model for the open gap condition?
Ok 6
e
REFERENCES 1.
" Methods for the Analysis of Oxide Fuel K3d Steady-State Thermal Effects (FR0sSTEY):
Code /Model Description Manual," Yankee Atomic Electric Company Report YAEC-1249P (Proprietary), April
~
1981.
2.
" Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY):
Code Qualification and Application," Yankee Atomic Electric Company Report YAEC-1265P (Proprietary), June
'1981.
~
"T' 3.
U.S.
Nuclear Regulatory Commission Standard Review Plan, Section 4.2,
" Fuel System Design," Revision 2 (July 1981),
U.S.
Nu clear Regulatory Commission Report NUREG-0800 (formerly
.NUREG-75/087).
4.
P.
A.
Berge on, " Maine Yankee Fuel Thermal Performance Evaluation Model," Yankee Atomic Electric Company Report YAEC-1099P (Proprietary), February 1976.
5.
C.
E.
Beyer et al.,
"GAPCON-THERMAL-2:
A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratories Report BNWL-1898, November 1975.
6.
" Light Water Reactor Fuel Rod Modeling Code Evaluation,"
Electric Power Research Institute Report EPRI NP-369 prepared by Combustion Engineering, Inc., March 1977.
7.
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 V.
L.
Rooney (NRC) letter to R.
L. Smith (VYNPCo) dated August 25, 1981.
3.
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 V.
L.
Rooney (NRC) letter to R.
L.
Smith (VYNPCo) dated August 25, 1981.
_s 9.
P.
Knudsen and C.
Bagger, " Power Ramp and Fission Gas Performance of Fuel Pins M20-1B, M2-2B and T9-3B," Riso National Laboratory (Denmark) Report Riso-M-2151, December 1978.
Transmitted as enclosure to V.
L.
Rooney (NRC) letter to R.
L.
Smith (VYNPCo) dated August 25, 1981.
10.
C.
- Bagger, H.
Carlsen and P.
Knudsen, " Details of Design, Irradiation and Fission Gas Release for the' Danish UO2-Zr Irradiation Test 022," Riso National Laboratory (Denmark)
Report Riso-M-2152, December 1978.
Transmitted as enclosure to V.
L.
Rooney (NRC) letter to.R.
L. Smith (VYNPCo) dated August 25, 1981.
. 11.
P.
Knudsen (Riso) letter to J.
C.
Voglevede (NRC) dated October 7, 1981.
Transmitted as enclosure to J.
C.
Voglewede (NRC) letter to C.
E.
Beyer (PNL) dated October 2 6, 1981 (copy enclosed).
12.
PWR Rod RJL, NRC Fuel Performance Data Base, dated February 1981.
Transmitted as enclosure to V.
L.
Rooney (NRC) letter to R.
L.
Smith (VYNPCo) dated August 25, 1981.
13.
C.
E.
Beyer et al., " User's Guide for GAPCON-THERMAL-2:
A Computer Program.for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratories Report BNWL-1897, November 1975. -
14.
S. K.
- Edler, Ed.,
" Reactor Safety Research Programs Quarterly Report,"
U.
S.
Nuclear Regulatory Commiesion Report NUREG/CR-1454 (PNL-3380-4), October-December, 1980.
15.
- D.
L.
- Hagrman, G. A. 'Reymann and R.
E.
Mason, "MATPRO-Version 11 (Revision 1):
A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior," Idaho National Engineering Laboratory Report NUREG/CR-0497 (TREE-1280, Rev. 1),
Fe bruary 1980.
16.
R.
Soulhier and M. J.
F.
Notley, " Effects of Power Changes on Fission-Product Gas Release from UO2 Fuel," NUCLEAR APPLICATIONS, Vol.
5, pp. 195-204, 1970.
17.
J.
E.
Garnier and S.
Be gej, "Ex-Reactor Determination of Thermal Gap and Contact Conductance Between Uranium Dioxide: Zircaloy-4 Interfaces - Stage I:
Low Gas Pressure," Battelle Pacific Northwest Laboratories Report NUREG/CR-0330 (PNL-2696), April 1979.
18.
J.
Garnier and S.
Begej, "Ex-Reactor Determination of Thermal Gap Conductance Between Uranium Dioxide and Zircaloy Stage II:
High Gas Pressure," Battelle Pacific Northwest Laboratories Report NUREG/CR-0330 (PNL-3232) Vol.
2, July 1980.
19.
W.
R.
- Lloyd, D.
P.
Wilkins and P.
R.
Hill, " Heat Transfer in Multi-Component Monatomic Gases in the Low, Intermediate and High Pressure Regime, NUCLEAR THERMIONICS CONFERENCE PROCEEDINGS (1973).
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