ML20211Q837
| ML20211Q837 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/24/1987 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LIC-87-119, NUDOCS 8703030199 | |
| Download: ML20211Q837 (4) | |
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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102-2247 402/536 4000 February 24, 1987 LIC-87-119 Document Control Desk U.S. Nuclear Regulatory Commission i
Washington, D.C.
20555
References:
1)
Docket #50-285 2)
Letter NRC (W. A. Paulson) to OPPD (R. L. Andrews) dated February 9, 1987 3)
XN-NF ' ?-61, Fort Calhoun Design Report Extended Burnup Analysis, Exxon Nuclear Company, Inc., October 1982 Gentlemen:
SUBJECT:
Cycle 11 Reload Evaluation In response to Reference 2, the below listed information is being fur-nished to allow the staff to complete the review for the Cycle 11 Reload Application.
The information is based upon the analysis contained in Reference 3 for Exxon Fuel.
The extended burnup analysis allows a peak assembly burnup of 43,000 MWD /MTV.
The fuel design for XN-3 (batch J), XN-4 (batch K) and XN-5 (batch L) is identical.
It is anticipated that an E0C peak assembly burnup of 44,500 MWD /MTU will be achieved for the K fuel.
Advanced Nuclear Fuels Corporation (formerly Exxon Nuclear Company) has agreed to revise the mechanical design report for batches K and L fuel to envelop the future batch burnups.
This analysis will be completed and submitted for approval by the USNRC prior to exceeding the ANF mechanical design criteria.
The cladding strain and fatigue were calculated by using the RODEX2 compu-ter code.
RODEX2 was run for the rod which attains the maximum exposure while using E0C conditions for various power histories, the axial location with the largest pellet / cladding contact pressures were chosen to undergo power ramping analysis.
The maximum exposure rod was also used for fatigue analysis. At the end of each cycle, the axial location with the largest pellet / cladding contact was ramped from 0 percent power to a prorated value of 112 percent of the 15.22 kW/f t. limit. The proration factor resulted from allowing only the i
peak power rod in the core to reach the 112 percent Fg limit.
070303019'? 070 L'4 PDR ADOCK ODOOOPBS p
I'UH
,-.y = _ _
g
v Document Control Desk February 24, 1987 Page 2 The stress intensities were used to determine stress amplitudes.
The frequency of power changes was the same as the estimated duty cycles reported in XN-NF-79-69, " Fort Calhoun Reload Fuel Design Report:
Mechanical Thermal-Hydraulic and Neutronic Analysis". The allowed frequency of cycles for the calculated stress amplitudes was determined from the irradiated zircaloy fatigue curves reported by O'Donnell and
- Langer.
The sum of the ratio of the egected duty cycles divided by the allowed duty cycles is the cumulative usage factor.
For the Fort Calhoun analysis, the factor was 0.2 'which is well below the 0.67 limit.
The cladding hoop stresses at the pellet / pellet interface during the various power escalations were calculated. All of the stresses were below the specified minimum 0.2 percent yield stress for the unirradiated fi cladding (50 ksi).
Since the yield stress was not exceeded, there was no plastic cladding strain and the transient cladding strain design criteria are satisfied.
The creep strain due to the most severe ramp meets the 1.0 percent total strain limit of the USNRC standard review plan 4.2 with substantial margin.
Instantaneous collapse pressure, P is calculated from the equation developedbyTimoshenkoandGerewhe, rep is calculated as 10,192 psi.
The equation does not account for the effect of ovality on collapse pressure. The critical pressure incorporates.i.e effect of ovality on collapse pressure and has been determined to be 4561 psi which is more than twice the reactor operating pressure.
The creep collapse of the cladding was analyzed using the technique described in XN-NF-82-06(P), Revision 1, " Qualification of Exxon Nuclear Fuel for Extended Burnup". At rod average burnup of 6,000 MWD /MTU, when densification is essentially complete, the combined creep down did not exceed the initial minimum diametral fuel cladding gap. The creep down plus the ovality was 0.0010 inch at >6000 MWD /MTU which is less than the minimum initial cold gap of 0.003 inch.
This will insure that no significant pellet axial gaps will occur during the life of the fuel and, therefore, no creep collapse can occur.
A conservative combination of rod power, axial power peaking and initial pressurization was incorporated into the R0DEX2 analysis for E0L fuel rod internal gas pressure.
It was determined that the internal rod pressure remins below the reactor core pressure of 2100 psia for a peak rod burnup 48,000 MWD /MTU.
The fuel' rod growth model was based on ANF measured data for PWR rods. A 95: percent upper confidence limit for values predicted by the growth model was used. The predicted maximum fuel rod growth was 1.25 inches, with conservative assumptions, yield on E0L clearance of 0.16 inch tie plate r
Document Control Desk February 24, 1987 Page 3 clearance. The MATPRO growth model was used to calculate the fuel assembly growth using the worse case dimensions, cold conditions and the peak rod average fast fluence. Allowing for uncertainty in the growth model with no hold down loads, the E0L clearance between the upper core plate and the fuel assembly latching plate was 0.19 inch. The attached Table I lists the Fuel Summary Behavior for fuel up to a peak assembly burnup of 43,000 MWD /HTU.
If you require any additional information or have any questions, please contact us.
Sincerely,
% d. ()J (n R. L. Andrews Division Manager Nuclear Production RLA/ WOW /bjb c:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D.C.
20036 Mr. A. C. Thadani, Project Director Mr. W. A. Paulson, NRC Project Manager Mr. P. H. Harrell, NRC Senior Resident Inspector
,S.
}
TAB 1E 1 FUEL stb 9tARY BEHAVICR Calculated Behavior Idall
't.imitina Condition)
Steady State am g 1/3 ault (28 ksi) 12.6 kai Cladding Stress am + ab i 1/2 ault (42 kai) 20.8 ksi am + ab + as 3 Oult (50 kai) 14.7 kai where: Om = primary membrane ab = primary bending as = secondary Steady State
< 1.01 0.192 Transient Cladding Response:
Plastic Strain for:
4t < 10 0,j,,2 (* 12 0.0 I l
0 21 10 g f t < 10 E<1.088 0.08 20 0.0 2 10 ft>10 E<0.2%
0.0 2
-Creep Strain:
< II 0.3 I Circumferential Ridge Stress:
< 56 kai (80% of irradiated clad SCC threshold stress) 48 kai Cisnulative Fatigue Usage Facter;
< 0.67 0.2 I i
Collapse at 3 6000 t#C/MTU Ovality + Creepdown
< Initial Ministan Pellet / Cladding Gap (.003 in.)
0.0010 t
l Corrosion Wall Thinning
< 0.002 ineb 0.0008 inch Cladd!ng ifydrogen Level
< 300 pg/g cladding 87 ps/s Fuel Rod Axial Growth Maintain Clearance 0.16 inch cisnrance l
l Fuel Assembly Axial Growth Maintain Clearance 0.19 inch I
clearance
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