ML20058P965
| ML20058P965 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/20/1993 |
| From: | Bauer J COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20058P968 | List: |
| References | |
| NUDOCS 9310260024 | |
| Download: ML20058P965 (10) | |
Text
I
[l..
1400 Opus Place Ccmmonwealth Edison Z1 Downers Grove, Illinois 60515 October 20,1993 i
Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
SUBJECT:
Byron Station Unit 2 Cycle 5 Reload
'l NRC Docket No. 50-455 l
REFERENCES:
See Attachment 3 i
Dear Dr. Murley:
Byron Unit 2 has completed its fourth cycle of operation and is conducting a l
refueling outage that began September 3,1993. Byron Unit 2 Cycle 4 attained a final i
cycle burnup of approximately 18,110 MWD /MTU. Cycle 5 is expected to commence on November 2,1993. The purpose of this letter is to summarize Commonwealth Edison Company's (CECO) safety evaluation regarding the Byron Unit 2 Cycle 5 reload core, and to provide the Cycle 5 Core Operating Limits Report (Attachment 2).
, describes the core reload redesign including a summary of CECO's safety evaluation, performed in accordance with the provisions of 10CFR50.59. No unreviewed safety issues or required Technical Specification t
changes were identified. provides the Operating Limits Report for Cycle 5 pursuant to Technical Specification 6.9.1.9. CECO and our vendor (Westinghouse) applied NRC approved reload design methodologies developed by Westinghouse as described in Reference 1. Commonwealth Edison performed the neutronic portion of the reload design using the methods and codes described in References 2 & 4 as approved in References 3 & 5, respectively.
l r
I kurdabyron'sleevam wpt 16
,1 9310260024 931020 PDR ADOCK 05000455-4 P
ppg y
L
._=. -
i
's
.i Dr. T.E. Murley October 20,1993 l
In summary, the Byron Unit 2 Cycle 5 reload design, including the development
.]
of the Core Operating Limits Report pursuant to the requirements of Technical l
Specification Section 6.9.1.9, was generated and verified by Commonwealth Edison using NRC approved methodologies.
Please direct any questions regarding this notification to this office.
P 5
Very truly yours, i
k.Yw0
+
Joseph A. Bauer i
Nuclear Licensing Administrator j
I JAB /gp
't Attachments l
cc:
R. R. Assa - Byron Project Manager, NRR l'
J. B. Martin ~ - Regional Administrator, Region 111 H. Peterson - Senior Resident inspector Syron r
f i
I 3
d
~
i 1
k:/ntatyrorValeevam.wpt'17
)
l
ATTACHMENT 1 Byron Unit 2 Cycle 5 Reload Description The Byron Unit 2, Cycle 5 reload core was designed to perform under current nominal design parameters, Technical Specifications and related bases, and current Technical Specification set points such that:
1.
Core characteristics will be less limiting than those previously reviewed and accepted; or
.n 2.
For those postulated incidents analyzed and reported in the Updated Braidwood/ Byron Final Safety Analysis Report (UFSAR) which could potentially be affected by fuel reload, re-analyses or re-evaluations have demonstrated that the results of the postulat3d events are within allowable limits.
The Byron Unit 2 Cycle 5 core is a " Low Leakage" design. Previously, Commonwealth Edison has successfully developed and operated similar " Low Leakage" designs at Byron as well as our Braidwood and Zion stations.
i During the Cycle 4/5 refueling outage, eighty-eight (88) Westinghouse VANTAGE 5 fuel assemblies have been inserted into the core. The Byron Unit 2 Cycle 5 core contains a full complement of fresh and previously irradiated 17x17 T
VANTAGE 5 fuel assemblies, therefore, the transition core penalty of 50 F for Large Break LOCA ECCS analysis has been removed. The NRC approved the use of, VANTAGE 5 at Byron Unit 2107 Cycle 3 operations and thereafter, under the provisions of 10CFR50.92 (Reference 6). The Braidwood/ Byron UFSAR describes the O
I' Westinghouse VANTAGE 5 fuel assemblies in a reload cora, and verified compatib!!ity with control rods and reactor internal interfaces. A mixture of Integral Fuel Burnable Absorber (IFBA) rods and Wet Annular Burnable Absorbers (WABAs) will be used as the burnable poison. The IFBA rods contain fuel pellets with an enriched B-10 coating. Both WABAs and IFBA burnable absorbers have previously been used by Commonwealth Edison.
The mechanical design of the Region 7 reload VANTAGE 5 fuel assemblies is the same as the Region 6A and 6B VANTAGE 5 fuel assemblies except for the use _
of: (1) a modified top nozzle (using a keyless, cuspless design) to address top nozzle spring issues and (2) a modified WABA holddown spring assembly, designed for improved manufacturability. The above design changes have no safety impact on Cycle 5 operation.
kyntatyronhvam.wpt 18
.g_
f 7 t
J) u i
c,,,,
D i
,Q; if" Atthchrnent 1 Page 2 l
[
L Byron U2C5 Reload Description The Byron Unit 2 Cycle 5 core has been designed and evaluated using NRC l
' iicarssed and approved rnethods. Commonwealth Edison requested approvai to perform the neutronic portion of PWR reload design using the methods described in Reference 2 wtCch was subs 9quently approved by the NRC in Reference 3.
SpeciUcaWy, the Byron Unit 2 Cycle 5 reload design, including the development of the core operating limits, were generated and verified by Commonwealth Edison using NRC approved muthodologies.
The reload fuel's nuclear design is evaluated generically in the UFSAR. The VANTAGE S fuel uses iba same pellet and fuel rod diameters as previously approved in past cycles. Althouuh this is the first full core of VANTAGE 5 fuel, the nuclear characteristics due to ine1/ANTAGE 5 fuel are within the range normally seen from cycle to cycle due to fuel management effects, as most reactivity parameters are insensitive to fuel type. The loading pattern dependent parameters were evaluated in detailin the CECO / Westinghouse reload safety evaluation process.
During the Byron Unit 2 Cycle 2/3 refueling outage, a number of fuel assembly guide pins (s-pins) seated on the underside of the upper core plate were damaged.
i CECO analyzed the potentialincreased peaking resulting from theoretical displacemems of fuel assernblies that may result from the deletion of the upper internal fuel assembly guide pins. Specifically, using the maximum possible relative displacerrents as determined by the Westinghouse Commercial Nuclear Fuel Division, CECO pedormed unit assembly calculations using the NRC approved 2D (TORTISE) basic design code. The shift in assembly power and resulting impacts on FNDH and other Safety Param6ter interaction List (SPIL) parameters, due to the assumed incmased water gap, were calculated for each affected assembly. CECO has
' deterrnined that all neutronic reload parameters remain within the previously estatiishu SPIL limits.
The radial nuclear heat flux hot channel limits are defined in Technical SpeciMetion 3/4.2.2 " Heat Flux Hot Channel Factor" These limits, as presented in l
the Operating Limits Report, have been generated to conservatively censider the increased local peaking that is theoretically possible in assemblies at locations where fud guide pins have been removed. CECO will consider the impact of the upper internals modification on Technical Specification 3/4.2.3 (Nuclear Enthalpy Rise Hot Channel Factor) by applying additional conservatisms to the measured FNDH during ine associated surveillance.
kunta byron /sieevam.wpl19 I
. Page 3 Byron U2C5 Reload Description i
CECO's reload safety evaluation process (RSE/SPIL review) is a verification to f
ensure that the previously reviewed and approved accident analyses are not adversely impacted by the cycle specific reload core design. CECO's Byron Unit 2 Cycle 5 Reload Safety Evaluation relied on previously reviewed and accepted analyses i
reported in the UFSAR, fuel technology reports, the VANTAGE 5 Reload Transition Safety Report (RTSR), and previous reload safety evaluation reports. A detailed review of the core characteristics was performed to determine those parameters affecting the postulated accident analyses reported in the Byron UFSAR. A 100 F Peak Clad Temperature (PCT) penalty will remain in place for Cycle 5 to address the possibility that the Chopped Cosine is not the most limiting power shape for Large Break LOCA analysis. This PCT penalty will be removed upon approval of WCAP-12909-P, " Power Shape Sensitivity Methodology". CECO has determined that all neutronic reload parameters remain within the previously established and recently revised reload safety and transient SPIL limits.
{
r A potentialissue regarding the Boron Dilution Mitigation System (BDMS) was j
identified which presented the following concerns:
- 1) The Inverse Count Rate Ratio (ICRR) data received from another utility l
indicated that previous data, used in the design of the BDMS, was not l
bounding.
l
- 2) The BDMS methodology was no longer conservative with respect to the flux doubling (24) setpoint, which included no instrument uncertainties in defining an equivalent " trip setpoint" as presented in the Technical Specifications.
The Boron Dilution SPIL limits were re-analyzed for the previous Byron Unit 1
)
Cycle 6 design using an increased Shutdown Margin (SDM), from 1.0% Ak/k to i
1.3% Ak/k. Additionally, a Technical Specification amendment change (Reference 8) was required. The BDMS Technical Specification change was subsequently approved in Reference 9. The Byron Unit 2 Cycle 5 Boron Dilution SPIL limits were updated and met.
The thermal-hydraulic design for the Cycle 5 reload core has not significantly changed from that of the previously reviewed and accepted cycle design. The Byron Unit 2 Cycle 5 reactor coolant system Hot Full Power Tave was increased 1 F from j
580"F to 581 F to offset the reduction in turbine throttle valve inlet pressure due to f
steam generator tube plugging. This ultimately results in restoration of secondary side _
operating margin. The thermal and hydraulic safety analyses are conservatively based on a nominal Hot Full Power Tave of 588.4 F. The FNDH limit of less than 1.65 for VANTAGE 5 assemblies ensures that the DNB ratio of the limiting power rod during i
k/nlabyronhvam.wpt?O
5 Page 4 l
Byron U2C5 Reload Description Condition I and Condition 11 events is greater than or equal to the DNBR limit of the 1
oNBR correlation being applied.
i The Operation of Byron Unit 2 Cycle 5 has been analyzed in accordance with NRC approved methodologies and satisfies all safety analysis limits. The margin of j
safety, as defined in the bases of the Technical Specifications, is not impacted or reduced.
Finally, verification of the Byron Unit 2 Cycle 5 reload core design will be performed per the standard reload startup physics tests. These tests include, but are not limited to:
1.
A physicalinventory of the fuelin the reactor by serial number and location prior to the replacement of the reactor head, 1
2.
A verification of fuel assembly alignment; 3.
Control rod drive tests and drop times; 4.
Critical boron concentration measurements; 5.
Control rod bank worth measurements using the rod swap technique; 6.
Moderator temperature coefficient measurements; 7.
Startup power distribution measurements using the incore flux mapping system.
In summary, CECO's use of VANTAGE 5 fuel and use of advanced neutronics methods (as described in References 7 and 2, respectively) have been previously approved by the NRC (References 6 and 3 respectively). The effects of the core -
i loading pattern, a one degree increase in core inlet temperature and the missing fuel assembly guide pins from the upper core plate, have been addressed with no impact on the conclusions of the 10CFR50.59 reload safety evaluation; that is, the reload and associated changes do not involve any unreviewed safety questions or require any Technical Specification changes.
kurdabymn'sleevammpt21 i