ML20095J743
| ML20095J743 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/27/1992 |
| From: | Chrzanowski D COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20095J749 | List: |
| References | |
| GL-88-16, NUDOCS 9205040126 | |
| Download: ML20095J743 (4) | |
Text
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Ccmmonwsalth Edison
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2 1400 Opus Place x '" 7 Downers Grove, Ulinois 60515 Apri1 27, 1992 Bd Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Co.nmission Nashington, D.C.
20555 Attn: USNRC Document Control Desk
SUBJECT:
Byron Station Unit 2 Cycle 4 Reload HRC_Dochuhm50-455 REFERENCLS:
See Attachment 3
Dear Dr. Murley:
Byron Unit 2 is completing a refueling outage that began February 28, 1992 following its third cycle of operation.
Byron Unit 2 Cycle 3 attained a final cycle burnup of approximately 17,166 MHD/MTU.
Cycle 4 is expected to commence operation in late April, 199?,
This letter is to summarize Commonwealth Edison Company's (CECO) plans and evaluations regarding the Byron Unit 2 Cycle 4 (BY2C4) reload core, and to provide the Cycle 4 Core Operating Limits Report (Attachment 2) per Generic letter 88-16. describes the Byron 2 Cycle 4 reload and CECO's reload safety evaluation review, which is being performed in accordance with the provisions of 10CFR50.59 as there are no unreviewed safety issues or necessary lechnical Specification changes.
The reload design has also evalua.'ed the impact on safety analyses resulting from the deletkn of 6 upper core plate feel assembly guide pins at some core locations during the Byron Unit 2 End-of-Cycle 2 refueling outage.
These removed fuel assembly guide pins did not pr^sent any unreviewed safety questions.
The Byron Unit 2 Cycle 4 core has been designed and evaluated using NRC approved methodologies. Commonwealth Edison performed the neutronic portion of the BY2C4 reload design utilizing NRC approved codes and methods as described in Reference 2.
The remain' der of the reload safety evaluation was performed by Hestinghouse in accordance with the methodology described in Reference 1.
In summary, the Byron Unit 2 Cycle 4 reload design, including the development of the Core Operating Limits Report (COLR) pursuant to the requirements of Technical Specification Section 6.9.1.9, was generated and verified by Commonwealth Edison using NRC approved methodology.
n Please direct any questions regarding this subject to this office.
Very truly yours, 00
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M David J. Chrzanowski g\\
gg Nuclear Licensing Administrator g
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'M cc:
A. H. Hsia - Project Manager, NRR M
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A. B. Davis - Regional Administrator, RIII 0y\\
$$o H. J. Kropp - NRC Resident Inspector, Byron ZNI.D/1764/1
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ATTACHMENT 1 Byron _2_Cy cle_LReloitd_Deitriation The Byron Unit 2 Cycle 4 core is a_ standard " Low Leakage" design and-is similar to the Cycle 3 core loading pattern.
During the End-of-Cycle 3 refuelti,9 outage, a total of 89 fuel assemblies, which are composed of 84 VANTAGE 5 fuel assemblies and 5 reinsert 0FA assemblies, will be loaded into the core.
The BY2C4 core will be composed of 168 Westinghouse 17x17 VANTAGE 5 assemblies (84 new and 84 once-burned) and 25 17x17 0FAs (5 once-burned and 20 twice-burned).
The NRC has approved the first use of VANTAGE 5 for Byron Unit 1 Cycle 4 and Byron Unit 2 Cycle 3 and thereafter, under the provisions of 10CFR50.90 in Reference _7.
The Byron UFSAR reesently reflects ihe transition to VANTAGE 5 fuel.
The BY2C4 reload core was designed to perform under cu, rent nominal design parameters, Technical Specifications and related bases, and current Technical Specificat'on setpoints such that:
1.
Core operating characteristics will be equivalent or less limiting than those previously reviewed and accepted; or 2.
For those postulated incidents analyzed and reported in the Byron Updated Final Safety Analysis Report (UFSAR) which could potentially be affected by fuel reload, reanalyses or reevaluations have been performed to demonstrate that the results of the postulated events.are within allowable limits.
The mechanical design of the new Cycle 4 fuel regions 6A and 6B VANTAGE 5 fuel assemblies is tho same as the Cycle 3 fuel regions SA end 5B VANTAGE 5 fuel assembliet except for:
snag-resistant Intermediate Flow Mixing (IFM) grids, and a.
b.
an_ increased radius on the fuel rod bottom end. plug.
The snag-resistant IFM grids prevent assembly hangup due to grid strap interference during an assembly removal.
This was accomplished by changing the grid corner geometry and adding guide tabs-on the outer grid strap. The bottom end plug has an increased radius in the. transition between the chamfer and the end of the plug There are no changes in the critical diniensions of the bottom end plug oi to the pressure drop from' the previous region. Therefore, these mechanical changes will have no effects on any of the fuel design parameters.
Fifty-three new Westinghouse Enhanced Performance RCCAs (EP-RCCAs) with Ag-In-Cd absorber materials will be installed in BY2C4 replacing the Hafnium RCCAs.
The EP-RCCAs have a thin chrome electroplate applied to-a specified Pength of the rodlet cladding surface to provide enhanced cladding wear resistance. The absorber diameter of the EP-RCCA is also reduced slightly at the lower extremity of the rodlets to better accommodate absorber-swelling upon irradiation and minimize interactions with the cladding.
ZNLD/1764/2
l AIIACMIENT.1 (cont'd) the UFSAR and in Reference 6.The reload fuel's nuclear design has bee As OFA and VANTAGE 5 f uel he. the same pellet and fuel rod diameters, most reactivity parameters are insensitive to f type.
Changes in nuclear characteristics due to the transition from 0FA t uel VANTAGE 5 fuel are within the range normally seen from cycle to fuel management effects and have been previously evaluated in Cy l o
cycle due to Unit l's first tran;ition cycle to VAN 17GE 5 fuel.
c e 4 in Byron dependent parameters have been evaluated ir. detail in the CECO /WThe loadin j
reload safety evaluation process.
estinghouse Commonwealth Edison (CECO) parameters remain within the Safety Parameter Interaction List (SPIL)ha These include, but are not limited to, SPIL items for non LOCA and LOCA limits.
considerations.
local peaking due to postulated larger inter-assembly gaps at locations as a result of remo';ed guide pins.
ore for Cycle 3, TORTISE, an NRC approved code used in CECO's priorIn the Cyc methodology, was used to calculate a " fuel rod-by-fuel rod" penalty t i
nuclear design overlayed on ANC (Reference 3) fuel rod powers to account for the potential o be water gap increase.
This penalty is applied to FN9H of the corner pins for assemblies adjacent to the increased water gap.
was assumed to exist over the entire length of the fuel e water gap significantly changed from that of the previouslThe thermal-hydraul Cycle 3 design.
o Tests and analysis have confirmed that the VANTAGE 5! reviewed and assemblies are hydraulically compatible with c'he OFA assemblies reload Region 4.
The VANTAGE 5 core limits are bounded by the OFA core limits Cycle 4 has a majority (168 e as
't of 193) of VANTAGE 5 fuel assemblies.
present Technical Specification FNDH limits of less than 1 55 for OFA Tne assemblies and 1.65 fc,r VANTAGE-5 assemblies ensure that the limiti Condition II events) ratio during Normal Operation and Operational Transient correlation being applied.is gruter than or equal to the DNPR lirait for the DNBR on I and were accounted for and found to have no effect on the six thermal analysis.
y raulic CECO's reload safety evaluation process (RSE/SPIL review) verification to ensure that the previously reviewed and approved UFSAR is a transient analyses are not adversely impacted by the cycle core design.
relied on previously reviewed and accepted analyses reporte specific reload fuel technology reports, the VANTAGE 5 Reload Transition S f t e UFSAR, previous RSE reports.
performed to determine those parameters affecting the postulated analyses reported in the Byron UFSAR.
ccident analyses presented in the UFSAR, the conclusions were not affected reload core characteristics.
e ZNLD/1764/3
AUACMENL1 (cont'd)
Furthermore, operation of the BY2C4 with six fuel assembly guide pins removed was evaluated for the effects on the UFSAR LOCA analyses.
The hydraulic effects of both larger and smaller gaps in the region below the removed guide pins during a hypothetical LOCA event was considered.
The impact of the altered fuel assembly configurations on the final Peak Clad Temperature (PCT) was negligible and all of the 10CFR50.46 criteria continue to be satisfied.
Thare are no re'oad-related changes to the current Technical Specificat cis required to ensure safe operation during Cycle 4.
Therefore, Westinghoi-and CECO have concluded in the BY2C4 reload safety evaluation that the t ee design parameters and assumptions renain appropriate and the conclusions in UFSAR remain applicabla.
Finally, the Byron Unit 2 Cycle 4 reload core design will be verified i
per the standaid reload startup physics tests.
These tests include, but are not limited to, the following:
1.
A core verification by physical inventory of the fuel in the reactor by serial number and location prior to the replacement of the reactor head; 2.
Control rod drive tests and drop times; 3.
Critical boron concentration measurements; 4.
Control / Shutdown bank worth measurements using the rod swap technique; 5.
Moderator Temperature Coefficient (MTC) measurements; 6.
Startup power distribution measurements using the incore flux mapping system; and 7.
Reactor coolant system flow measurement.
In addition, per the requirement of.the NRC SER-(Reference 3), the BY2C4 startup physics test results will be submitted to-the NRC since BY2C4 is the first Cyron Unit 2 reload fully analyzed by CECO with the advanced neutronics design methods (Reference 2).
In conclusion, CECO's use of VANTAGE 5 fuel (Reference 6) and use of advanced neutronics methods (Reference 2) have been previously approved by the NRC (References 7 and 3).
Therefore, pending completion of-the On-Site and Off-Site Reviews, no additional prior NRC review and approval of the reload core analyses or application for amendment to the Byron Unit 2 operating lu.ense, is required as a result of the cycle specific reload design for Cycle 4.
ZNLD/1764/4
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