ML19329A235
| ML19329A235 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/23/1976 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Goller K Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7912300288 | |
| Download: ML19329A235 (6) | |
Text
. _ _ _ _
i APR 2 3 B76 6
K. R. Coller, Assistant Director for Operating Reactors, DOR OCONEE 1 - PROPOSED CHANCES TO TECIDTICAL SPECIFICATIONS (TAR-4168)
Plant Name:
Oconee Unit 1 DochettNumbert g 69
j Responsible Branch ORF-1 i
and Project Manager Gary Zech Technical Review Branch:
Coos Performance Branch l
Review Status:
Complete The Core Performance Branch has reviewed the request by Duke Power Company to change the Technical Specifications of Unit 1 of the Oconee Nuclear Station. The changes are required to permit operatior >f Oconee 1 in its third cycle.
We find the proposed changes to be acceptable.
Original signedlid n,.Laasa 1 M D. F. Ross, Assistant Director for Reactor Safety Divisien of Systems Safety Office of Nuclear Reactor Regulation
Enclosure:
Review S. Hanauer Distribution:
F. Schroeder pDocketRoom R. Boyd NRR Reading File R. A. Purple CPB Reading File G. Zech D. P. Ross
.P.
Check W. Brooks W. Mcdonald A
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UNITED STATES t
NUCLEAR REGULATORY COMMISSION 4
W ASHINGTON, D. C.
20555 APR 2 3 Y376 K. R. Goller, Assistant Director for Operating Reactors, DOR 4168) j OCONEE 1 - PROPOSED CHANGES TO TECHNICAL S l
Oconee Unit 1 Plant Name:
50-269 Docket Number:
ORB-1 Responsible Branch Gary Zech and Project Manager Core Performance Branch Technical Review Branch:
Complete i
Review Status:
The Core Performance Branch has reviewed the request by Duke Po f the Oconee Company to change the Technical Specifications of Unit 1 o Nuclear Station.
1 in its third cycle.
We find the proposed changes to be acceptable.
D. F.
ss, Assistant Director for Reactor Safety Division of Systems Safety Office of Nuclear Reactor Regulation
Enclosure:
Review I
S. Hanauer F. Schroeder j
R. Boyd l
R. A. Purple G. Zech li P. Check l
W. Brooks l
W. Mcdonald R. Meyer i
M. Tokar
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Fuel Mechanical Design The Oconee 1, Cycle 3 reload fuel consists of 60 Mk-B4, Batch 5 fuel assemblies. There are also 61 once-burned (batches 4A and 4B) assemblies and 56 twice-burned (batch 3) fuel assemblies in
'the core loading for cycle 3 operation,for a total of 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide.
Pertinent fuel design parameters are listed in Table 4.2.1-1.
'+
Creep collapse calculations were performed for three-cycle I
assembly power histories for Oconee I, using the approved Babcock 4
and Wilcox computer code, CROV (Refs.1, 2).
The calculations included conservative treatment of effects of fission gas (no credit taken), cladding thickness (lower tolerance limit), initial cladding ovality (upper tolerance limit) and cladding temperature (assembly outlet temperature) on collapse time.
The most limiting assembly was found to have a collapse time greater than the maximun projected cycle 3 life -of 21,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />.
With respect to -fuel rod bowing, the model shich B&W presented to us on September 8, 1975 (Ref. 3) has been reviewed and is acceptable. This model may be used to predict bow magnitude until further information becomes available. Based on Oconee 1 first cycle data, the total rod bow magnitude vill be smaller than other PWR fuel k.,
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Fuel thermal analysis calculations that account for the effects
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of fuel densification have been performed with our approved version f}
of the BiW analytical model TAFY (Ref. 4). Fuel densification results in increases in stored energy, linear thermal output, and the probability l
e
. of local power spikes from axial gaps. During Cycle 3 operation, the highest relative assembly power levels occur in batches 4 and 5 fuel.
The fuel temperature analysis for batches 1, 2, and 3 fuel is documented in the Oconee Fuel Densification Report (Ref, 5).
This analysis is also applicable to batches 4 and 5 because they have the same linear.
heat rate capabilities to centerline melt (Ref. 6).
The batch 5 fuel' assemblies are not new in concept and they do not utilire different component materials.
Therefore, on the bases of the analysis presented in the cited reports, we conclude that for Oconee 1, Reload 2:
1.
the fuel rod mechanical design provides acceptable safety margins for normal operation,and 2.
the effects of fuel densification have been acceptably accounted for in the fuel design.
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1
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1.
A.- F. J. Eckert, H. W. Wilson, and K. E. Yoon, " Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse," BAW-10084, May 1974.
2.
A Generic Review of the B&W Cladding Creep Collapse Analysis Topical Report BAW-10084, USNRC, August 9, 1974.
3.
USNRC memorandum, S. Kim to D. Ross, " Babcock and Wileax Rod
{
Bow Model, November 26, 1975.
- 4. -
C. D. Morgan and H. S. Kao, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis," BAW-10044, May 1972.
5.
"Oconee 1 Fuel Densification Report," BAW-1388, Rev. 1, July 1973.
6.
"Oconee Unit 1 Cycle 3 Reload Report," BAW-1427, December,1975.
4
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t Table 4.2.1-1 Fus1 Dscign Paramstero ij Residual New Fuel Assembly Fuel Assembly Batch 3 Batch 4 Batch 5 1.
Fuel Assembly Type Mk-B2 Mk-B3 Mi.-B4 2.
Number 56 61 60 3.
Initial Fuel Enrichment 2.15 3.20/2.60 2.75 4.
Initial Fuel Density,
% Theoretical 93.5
> 94. 5 93.5 5.
Fuel Rods
-0.D. Inches
.430
.430
.430 l
I.D. Inches
.377
.377
.377 6.
Fuel Pellet 0.D. Inches
.370
.3685 (mean)
.370 E&'
f-
4
_Technicel Specification Chenges-By letter dated December 1, 1975, Duks Powsr Comp:ny requtsted Techni-cal Specification changes to permit operation of Oconee Nuclear Power Station, Unit i during its third fuel cycle. To support this request Duke Power Company also submitted a Cycle 3 Reload Report for Oconee Unit 1, and_by letter dated a supplement to this report wh'ich treated the effects of T
February 27, 1976, fuel rod bowing on power peaking in the core.
We have reviewed these submittals. The analyses were performed by the The 10 CFR Part 50, Appendix K design methods used by Babc.ock and Wilcox (B&W).
Fuel criteria were applied, making use of an approved calculation model.
rod bowing distributions were calculated with a B&W model based on PIE I
l results from operation of Oconee Unit 1 during its first cycle.
This model has been tentatively approved while awaiting more data from operation of The peaking factor effects have been Oconee Unit 1 and other B&W reactors.
appropriate statistical model. The calculated maximum calculated with an local power increase is 2.15% The applicant has proposed to accommodate this by reducing the allowable azimuthal tilt during operation by 1.2%
(excore measurement) which reduces the peaking factor by 2.21%.
The applicant has provided values for core physics parameters for the Cycle 3 loading including reactivity coefficients, potential ejected rod worth, shutdown margin, boron worth, neutron lifetime, and delayed neutron We have reviewed these values and find them acceptable.
fraction.
The applicant has examined each FSAR accident analysis with respect to changes in Cycle 3 parameters to determine the effects of the reload and to ensure that thermal performance during hypothetical transients is not de-The analysis shows that in most cases the consequences of transients graded.
are less severe and in no case are they more severe.
On the basis of our review we conclude that the proposed Technical 1
Specification changes are acceptable.
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