ML20039C799
| ML20039C799 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/08/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8112300247 | |
| Download: ML20039C799 (10) | |
Text
,
DISTRIBURT10N DEC 8E Docket File NRC PDR NSIC TERA Docket Nos 50-317 ORB #3 Rdg and 50-318 DEisenhut OELD Mr. A. E. Lundvall, Jr.
I&E-3 Vice President-Supply ACRS -10 Daltimore Gas & Electric Company JHeltemes P. O. Box 1475 PMKreutzer-3 Baltinore, l'oryland 21203 PM-DJaffe RAClark
Dear l'r. Lundvall:
Gray File 1:e are presently reviewing the application of Lombustion Engineering's fa-provements to their fuel evaluation model (FATES 3) as applicable to Calvert Cliffs Units 1 and 2.
In the course of performing our evaluation, we have found it necessary to request additional information. We request that you provide the responses to the enclosed questions within 30 days following receipt of this letter.
The reporting and/or record keeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, Original signed by.
Robert A. Clark Robert A. Clark, Chief Operating Peactors Division of Licensing
Enclosure:
As stated cc: See next page b
D'.
,/y;p p
r
~.
t,
\\
~
0Y12%h0 D
0 dM omce >
3.g, ORBg [
0R ORB #3:DL
...$.....?.r DJafM '
RAClark sunucus)
..I
. HOI
.12/.fbl..
kN
~
oarr y anc rc'RM 318 00 84 NGCM 024a OFFICIAL RECORD COPY uscro m _ m mo
"~
Balti[nore Gas and Electric Company cc:
James A. Biddison, Jr.
Ms. Mary Harrison", President General Counsel Calvert County Board of County Commissioners Baltimore Gas and Electr.ic Company Prince Frederick, MD 20768 P. O. Box 1475 Baltimore, MD 21203 U. S. Environmental Protection Agency Region III. Office George F. Trowbridge, Esquire Attn: Regional Radiation Representative Shaw, Pittman, Potts and Trowbridge Curtis Building (Sixth Floor) 1800 M Street, N. W.
- Sixth and Walnut Streets, Washington, D. C.
20036 Philadelphia, PA 19106 Mr. R. C. L. Olson, Principal Enaineer Mr. Ralph E. Architzel Nuclear Licensing Analysis Unit Resident Reactor Inspector Baltimore Gas and Electric Company NRC Inspection and Enforcement Room 922 - G&E Building P. O. Bos 437 P. O. Box 1475 Lusby, MD 20657 Baltimore, MD 21203 Mr. Charles B. Brinkman Mr. Leon B. Russell Manager - Washington Nuclear Operations Plant Superintendent
_ Combustion Enginearing, Inc.
Calvert Cliffs Nuclear Power Plant 4853'Cordell Avenue, Suite A-1 Maryland Routes 2 & 4' Bethesda, MD 20.014 Lusby, MD 20657 Mr. J. A. Tierman, Manager
.c Bechtel Power Corporation Nuclear Ppwer Department Attn: Mr. J. C. Judd Calvert Cliffs Nuclear Power Plant Chief Nuclear Engineer Maryland Routes 2 & 4 15740 Shady Grove Road Lusby, MD 20657 Gaithersburg, MD 20760 Mr. W. J. Lippold, Supervisor-Combustian Engineering, Inc.
Nuclear Fuel Management Attn: Mr. P. W. Kruse, Manager Baltimore Gas and Electric Company Engineering Services Calvert Cliffs Nuclear Pcwer Plant P. O. Box 500 P. O. Boz 1475 Windsor, CT 06095 Baltimore, Maryland 21203 PublicDeci[mentRoom Mr. R. E. Denton, General Supervisor Calvert County Library Training & Technical Services Prince Frederick, MD 20678 Calvert Cliffs Nuclear Power Plant
- Maryland Routes 2 & 4 Director, Department of State Planning Lusby, MD '20657 301 West Preston Street Baltimore, MD 21201 Mr. R, M. Douglass, Manager Quality Assurance Department Administrator, Power Plant Sj ting Program
---Fort Smallwood Road Complex Inergy and. Coastal Zone Administration P. O. Box 1475 Department of Natural Resources
- Bal timore, MD 21203
' " Tawes State' Of.fJce'Bui1 ding Annapolis,FC 21204 Mr. T. L., Syndor, General Supervisor l
Operations Quality Assurance Calvert Cliffs Nuclear Power Plant l
Maryland Routes 2 & 4 i
Lusby, MD 20657 l
l
]_
QUEST 10ilS ON CEN-161 (FATES-3) 1.
Ap plica tion o f CEN-161 ( Re f.
- 1) in Licensing Analyses.
From the list df licensing anaysis requirements provided.below, please identify which applications will be net with FATES-3.
Where the licensing analysis requirements will not be met with FATES-3, identify the analysis methods used.
4 A.
LOCA initial conditions.
B.
Initial conditions f o r o t,he r c o d e s.
j j
1.
Core thermal-hydraulic transient codes.
2.
Point kinetics system analysis codes.
1 1
3.
Co re-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.
=
.c
.[*
2.
Scrain fatigue limits.
3.
Fretting wear limits.
l 4.
Oxidation, hydriding, and the buildup of corrosion i
products (crud) limits.
5.
Dimensional changes such as rod bowing or irradiation growth limits.
1 6.
Fuel and burnable poison rod inte rnal pressure limits.
1 7.
Wo rs t-case hydraulic load (assembly holddown) limits.
8.
Control rod reactivity limits.
D.
Fuel rod-failure or initialization of other a'nalyser, used to calculate fuel rod failure.
1.
Overheating.
a.
Departure from Nucleate Boiling (DNB).
b.
Fuel enthaply.
~
c.
Fuel centerline melting (steady-state and transient).
2.' Pollot/Clodding Incoroction (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.
j 1.
Cladding embrittlement.
2.
Violent expulsion of fuel.
3.
Generalized cladding melting.
I 4.
St ructural def o rmation.
.c 5.
Fuel rod ballooning.
1.
Differences Between FATES-3 and Previous Fuel Performance Codes 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.
3.
To help us identify changes in operating limits that sight be espected 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 applications identified in your response to Questica 1.
~65 FATES-3 Input
~~
- w.........
A.
Provide a list of input parameters f o r t he-PWTE S-3'.c o d e..
State whether each input value used is best-estinate or conservative, or otherwise results in a best-esti:2 ate or conservative calculation.
If conservative, quantify the margin of conservatism and the, basis for the margin used.
If the margin varies f rom a9 plication to applicatioa, provide this information for each different application.
5
B.* Pro'vido the input values for a typical case of each of those applicati-ons ' identified in response to Question 1 in t
sufficient detail to permit staff audit calculations.
4.
FATES-3 Output A.
Provide a list of output parameters from the FATES-3 code.
Identify the end 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.
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 to application, provide this information for each different application.
C.
Using the input values, described in your response to Question 3 above, provide the. corresponding FATES-3 output values in sufficient detail to permit a comparison with staff audit calculations.
.c 5.
Code Conservatism 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 temperatures used for LOCA initial conditions, should exhibit an' appropriate level of conservatism, based on statistical analysts of experimental-vs.-predicted fuel temperatures.
6.
Model Verification A.
Provide additional fuel temperature model verification by r.licting 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 (o 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.
Retic of fuol conterlino temper'aturo prediction cnd measurement as a function of initisi gap size (Rods 1; 2 and 3).
5.
Ratio of rod internal pressure prediction and nea*surement as a function of burnup (Rods 1, 5 and 6).
B.
Provide additional verification of the FATES-3 fission gas 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 Reference 7.
C.
Prov'ide additional veri'fication 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 foi this prediction, the predicted release value sheuld be provided as a function of assumed final resintered density (94-98" T.D.).
7.
Fission Gas Release Model A.
The significant grain size dependence of the fission gas release.model does not appear to be supported by the gas _.
release data presented.
Does Combustion Engineering hav&
additional data to support this dependence?
3.
What is the grain size o f Combus tion Enginee ring commercial fuel and how does the model compare with gas release data for grain sizes in the range of the commercial fuel?
~
3.
Fission Gas Re leas'e Data A.
The Bellamy and Rich data ('Re f. 9) have been u s.e d to verify the fission gas release model.
How were the rod linear heat ratings determined for these data as they were r.o t given in the original publication?
If,this information is contained in Reference 9-2 of CEN-161,.please provide us with a copy of this reference.
B.
Bellamy and Rich have estimated (Ref. 9) the gap resistivity of their data to be between 1.5 and 2.0 C-cm2/W.
In Tables
~~~
9-5'and 9-6 of CEN-161,*Combsstion Engineering has provided predictions.for this range of gap resistivity.- There are, "however, s everal h~igh b'ur:dfiiEFa' include'd 1a TaYle.9-5'
~
~
(2.0 C-cm2/W), but these are missing from Table 9-6 l
(1.5 C-cm2/N).
Why were these data omitted from the study?
~
a a-9.
Fuel Swelling Model A.
Is the co'tal swelling value the " unconstrained swelling rate," pad is the solid fission-product swelling value the " constrained swelling rate"?
3.
How is fuel-cladding contact defined for the transition between unconstrained and constrained fuel swelling; e.g.,
is fuel relocation included in calculating contact pressure?
C.
Once the fuel is constrained radially, are both the radial and axial swelling calcul'ated using the constrained swelling rate?
Are there data to justify using the constrained value for both the radial and axial directions?
We would expect that the radis1 direction is under more constraint than the axial direction.
This is a safety concern mainly with regard to further loss of, void volume and, hence, higher gas pressure.
D.
Fuel swelling is a temperature dependent ph'enomenon, and the value used in FATES-3 is'apparently based on data from low temperature fuel..However, peak linear heat ratings will provide fuel temperatures very near the transition between low-and high-temperature fuel swelling.
Are there data that show the applicability of this swelling c rate at peak linear heat ratings?
E'.
What effect does the new (lowe r), as opposed ~
the old to i
(highe r), swelling rate have on fuel rod internal pressure?
- 10. Annular Pellet Model A.
Does the inclusion of the annular model.chande the other fuel behavior models (e.g.,
relocation, swelling and fission gas release)?
If not, why not?
If changes were made, what were they?-
3.
The equations used to calculate the temperature distribution in an annular fuel pellet are provided in Section 5 of CEN-161.
However, the method used to deternine the flux depression constants A,
B and C is not given.
Please explain how these constants are derived.
Are the same cons tants used f or both solid and annular f uel, or are these constants specifically derived for a given
..( s o l id o r a:Ln ula r). d e s ign ?,,:,,We..v.o uld e x p e c t that. Che u's.e of constants derived for solid pellets wodid Te siightly nonconservative for annular. pellet designs.
e
- 11. Fissiod Goo Production Modal A.
The fission gas production model in FATES-3 uses a value of 30 cc/ mwd while the American Nuclear Society Working Group 5.4 (Ref. 10) uses a value of 31 cc/ mwd.
What data and references were used to develop the 30 cc/ mwd value?
- 12. Cap Conductance Model A.
The gap conductance model in FATES-3 no longer has an artificial limit on gap conductancee, but calculates' contact conductance based on fu,el-cla'dding contact pressures.
When fuel-cladding contact is made, what is the typical range of gap conductivity calculated?
B.
When fuel thermal expansion or swelling is large enough to cause fuel-cladding contact, does the relocation model allow for relaxation of previous relocation?
If not, it appears that unreasonably high interfacial pressures and gap conductance may result.
- 13. Helium Production
.c A.
Postirradiation puncture data (Ref. 11) from an experimental experimental Halden fuel rod (Rod 8 of IFA-432) irradiated to 22,000 mwd /MtU have shown more helium present (25%) than dould be expected as a result of the initial fill gas introduced during fuel rod fabrication.
Similar behavior has been reported ( Re.f s. 5-6) for the Riso rods mentioned previously.
Does the FATES-3 code take into account helius production and release?
If noe, would the inclusion of such a codel have a significant effect on calculated end-of-life rod pressures?
O
~.......
M h
A REFERENCES 1.
" Improvements to Fuel Evaluation Model," Combustion Eng'neering i
j Report CEN-161-(B)-P (Proprietary) and CEN-161-(B)
(Non-Proprietary), July 1981.
2.
" Fuel Evaluation Model," Combustion Engineering Report CENPD-139-P-A (Proprietary) and CENPD-139-A (Non-Proprietary);
July 1974 3.
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.
4.
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.
-c F.
P.
Knudsen and C.
Bagger, " Power Ramp and Fission Gas Performance of Fuel Pins M20-1B, M2-2B and T9-3B," Riso Natio'nal f,aboratory (Denmark) Report Riso-M-2151, Decenber 1978.
Transmitted as enclosure to J.'C.
Voglewede (NRC) letter to J.
C.
Ennaco (C-E) dated June 17, 1981.
6.
C.
- Bagger, H.
Carlsen and P.
Knudsen, " Details o.f Design, Irradiation and Fission Gas Release for the Danish UO2-Zr Irradiation Test 022," Riso National Laboratory (Denmark)
Report Ri s o-ti-215 2, December 1973.
Transmitted as enclosure to J.
C. Voglewede (NRC) letter to J.
C. Ennaco (C-E) dated June 17, 1981.
7.
P.
Rnudsen (Riso) letter to J.
C.
Voglewede (NRC) dated October 7, 1981.
Transmitted as enclosure to J.
C.
Voglevede (NRC) letter to C.
E.
Beyer (PNL) dated October 26, 1981 (copy to J.
Ennaco, C-E).
8.
pWR Rod RJL, NRC Fuel Performance Data Base, da ted Fe b rua ry-
~~'
1981. ' Transmitted as~cnc1"osure'to J.'C. Vo'gl~e'wede (NRC) letter to J.
C. Ennaco (C,E),Qted.,Jone 17,.198g.
9.
R.
G.
Be llamy and J.
3.
Rich, " Grain-Boundry Gas Release and Swelling in High Burn-Up Uranium Dioxide," JOURNAL OF NUCLEAR MATERIALS, Vol. 33 (1969), pp. 64-76.
10.
Proposed ANS-5.4 Scandard on " Method for Calculating the Release of Fission Products from Oxide Fuel" transmitted by letter from S. E. Turner (ANS-5.4 Chairman / Southern Science Applications, Inc.) to J.
F. Mallay (NUPPSCO/ Nucle'ar Safety Analysis Center) dated May 28, 1980.
11.
S.
K.
- Edler, Ed.,
" Reactor Safety Research. Programs quarterly.
Report," U.
S.
Nuclear Regulatory Commission Report NUREG/CR-1454 (PNL-3380-4), October-December, 1980.
9 e
e Oe 4
0 e
MUN
=.
1 I
i
__