ML20134M177
| ML20134M177 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/13/1997 |
| From: | Shell R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19355E721 | List: |
| References | |
| NUDOCS 9702200129 | |
| Download: ML20134M177 (5) | |
Text
_ __.
E*
j e
Tennessee valley Authority. Post Office Box 2000. Soddy-Daisy. Tenressee 37379-2000 l
February 13,1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk j
Washington, D.C. 20555 i
SEQUOYAH NUCLEAR PLANT (SON)- TECHNICAL SPECIFICATION (TS) CHANGE 96 FRAMATOME COGEMA FUEL (FCF) MARK-BW17 FUEL CONVERSION
Reference:
TVA's letter to NRC dated April 4,1996 on the above subject The above reference submitted the required TS revisions to support use of the FCF Mark-BW17 fuel type at SON. Enclosure 5 of the change request included FCF Topical Report No. BAW-10220P. (Enclosure 6 of the letter also included Topical Report BAW-10220NP, which is the nonprietary version of the BAW-10220P report.) This report contained the technical justification for the proposed TS revisions.
Tables 7.1-1 and 7.1-2 in section 7.1.8 of Topical Report BAW-10220P indicate that the thermal-hydraulic analyses performed to support the transition to Mark-BW17 fuel were based upon an assumed 7.0 percent core bypass flow. This assumption was based upon previous bypcss flow calculations performed by FCF to suppon fuel conversion analyses for Portland General Electric (Trojan) and Duke Power Company (McGuire).
Subsequent plant-specific core bypass flow calculations performed by FCF determined that the 7.0 percent assumed bypass flow value is not conservative at SQN. As a result, the amount of assumed core bypass flow has been increased from 7.0 percent of total reactor coolant system flow to 7.5 percent. (The increased flow results from a slightly larger reactor vessel to core barrel gap in the design of the SQN reactor vessel internalc.) FCF has analyzed the impact of 7.5 percent core bypass flow on the departure from nucleate boiling ratios (DNBRs) for the core safety limits as well as the limiting loss of reactor coolant flow transient. The increased core bypass flow results in DNBR reductions of less than one percent. Those events which were analyzed deterministically (non-statistical core design methodology) have been reanalyzed using the increased bypass flow value. FCF will apply a one percent penalty against the retained thermal margin for both the Mark-BW17 and V5H fuel types to account for the revised bypass flow assumption.
l 9702200129 970213 bI PDR ADOCK 05000327 P
PDR 2ad 9000 gri
' Mp e\\w ddskMW L\\le QDL
\\
1',
s U. S. Nuclear Regulatory Commission Page,2
- February 13,1997 Enclosed are copies of Tables 7.1-1 and 7.1-2 of Topical Reports BAW-10220P and BAW-10220NP which have been revised to reflect the 7.5 percent core bypass flow assumption. The proprietary data withholding affidavit included in the April 4,1996 letter to NRC as Enclosure 4 continues to apply to the attached information.
Please contact Jim Smith at (423) 843-6672 should you have any questions.
he k.
R. H. Shell Site Licensing and Industry Affairs Manager Enclosures cc: See page 3 k
l
4 U.S. Nuclear Regulatory Commission Page 3 February 13,1997 Enclosures cc (Enclosures):
Mr. R. W. Heman, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. Michael H. Mobley, Director Division of Radiological Health Third Floor L&C Annex Nashville, Tennessee 3743-1532 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region ll 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 l
i
Non Proprietary i
Table 7.1-1
)
Thermal-Hydraulic Analysis Design Parameters Desian Parameter Value Core Conficuration:
Number of Fuel Assemblies 193 Fuel Assembly Type 17x17 Number of Fuel Rods Per Assembly 264 Number of Control Clusters 53 Number of Absorber Rods per Control Cluster 24 Reactor Coolant System:
Rated Thermal Power, MWt 3411' Heat Generated In Fuel, %
97.4 Nominal System Pressure, psia 2280 Nominal Thermal Design Flow, gpm 360,100 Flow Fraction Effective for Heat Transfer 0.925 I
(7.5% Bypass) l Minimum Thermal Design Flow, gpm 348,000
)
Mechanical Design Flow, gpm 410,000 Hot Channel Core Inlet Flow Factor 0.95 Average Vessel Coolant Temperature (nominal) at 100%RTP, F
578.2 7-6
Non Proprietary
'~
Table 7.1-2 Core Thermal-Hydraulic Conditions 100% Power 360,100 gpm 7.5% Bypass l
Heat Balance Summary Rated Thermal Power, MWt 3411 Ave Heat Flux, BTU /hr-sq-ft 189,100 1
Stack Height, in 144 Fue'l Rod Outer Diameter, in 0.374 i
i Fuel Pins per Assembly 264 Assembly Flow Area, sq-in 38.7 Nominal Thermal Design RCS Flow, gpm 360,100 Flow Fraction Effective for Heat Transfer 0.925 i
i Core Inlet Velocity, ft/sec 14.3 I
i i
Inlet Mass Flux, Mlb/hr-sq-ft 2'43 l
Vessel Mass Flow Rate, Mlb/hr 136.19 Pressure, psia.
2280
)
Vessel Ave. Temp, F
578.2 Inlet Temperature, 'F 546.2 610.2 Vessel Outlet Temp, F
Vessel Delta Temp,
- F 64.1 Core Outlet Temp, F
614.5 Core Ave Temp, F
580.4 Core T
- Vessel T e,
F 4.3 oue oo I
I 7-8 i