ML20125B503
| ML20125B503 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 09/30/1991 |
| From: | Deora M, Robare D, Sozzi G DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20125B467 | List: |
| References | |
| FOIA-91-579 NUDOCS 9212100033 | |
| Download: ML20125B503 (130) | |
Text
FERMI 2 91150 i
e POWER UPRATE SAFETY- ANALYSIS Approved by:
- s_/ _.
W K. Deora Detroit Edison Co.
Approved by: / llulhMa G. L Sozzi/ t/ '
GE Nuclear Energy Approved by:
D. J. Robare GE Nuclear Energy t
Approved by: U'-d .k., -
T. B. Madden Stone & Webster Engineering Corp.
\'
September 1991 921210dO33 920622 PDR FOIA GUNTER91-579 PDR.
I'ERMI 2 91150
(
PROPRIETARY INFORMATION NOTICE Parts of this document contain proprietary information of the General Electric Company (GE) and are furnished to Detroit Edison Co. in confidence solely for the purpose or purposes stated in the transmittal letter. No other use, direct or indirect, of the document or the information it contains is authorized. The recipient shall not publish or otherwise disclose it or the information to others without the consent of GE, and shall return the document at the request of GE.
GE proprietary information is noted by "GE Proprietary Information" on those pages so designated. Non proprietary information is noted by " bars" drawn in the margin of the text of this report or is on a page not designated proprietary.
The technical bases for the information provided by GE in this report is contained in DRF No. A00-03770.
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully (
The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Detroit Edison Co. and GE, Plant Services Contract, effective December 15, 1989, as amended to the date of transmittal of this document, and nothing contained in this _ document shall be construed as charging the contract. The use of this information by anyone other Man Detroit Edison Co., or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use. GE makes no representation or warranty, express or implied, and assumes no liability as to the compte.teness, accuracy, or usefulness of the information contained in l this document, or that its use may not infringe privately owned rights.
l l
k ii September 1991
FERMI 3 91150 TABLE OF CONTLhTS Section Title Page EXE CUTIVE S UMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i 1.0 OVE RVI EW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1.1 INTRO DUCTION . . . . . . s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 PURPOSE AND APPROACH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 13 UPRATED PLANT OPERATING CONDITIONS . . . . . . . . . . . . . . . . . . . . 14 13.1 Reactor Heat Balance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 1.3.2 Reactor Performance Improvement Features . . . . . . . . . . . . . . . . . . . . . . . . 15 1.4
SUMMARY
AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.5 R E FE R EN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 2.0 REACTOR CORE AND FUEL PERFORMANCE . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 FUEL DESIGN AhT OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.1 Lead Fud Assemblies . . . . . . . . . . . . . . . . . . . . . . . . ....,........... 21 2.2 THERMAL LIMITS ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 2.2.1 Minimum Critical Power Ratio Operating Limit ..................... 2-2 2.2.2 Maximum Average Planar Linear Heat Generation Rate and Maximum Linear Heat Generation Rate Operating Limits . . . . , . . . . . . . . 2-2 2.3 REACTIVITY CHARACTERISTICS . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . 22 23.1 Power / Flow Ope rating Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.4 STAB ILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.5 REACTIVITY CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.5.1 Control Rod Drive Hydraulic System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3
- 2. 6 R E FE RE N C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 3.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS . . . . . . . . . . 3-1 3.1 NUCLEAR SYSTEM PRESSURE RELIEF . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 i September 1991
FERMI 2 91150 TABLE OF CONTENTS (Cont'd) (
Section Title Page 3.2 REACTOR OVERPRESSURE PROTECTION . . . . . . . . . . . . . . . . . . . . . . 31 33 REACTOR VESSEL AND INTERNALS , . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 33.1 Reactor Vessel Fracture Toughness . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . 32 33.2 Ret.ctor Internals and Pressure Differet tials . . . . . . . . . . . . . . . . . . . . . . . . 3-2 333 Reactor Vessel Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.4 REACTOR RECIRCULATION SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.5 REACTOR COOLANT PIPING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4 3.6 MAIN STEAMLINE FLOW RESTRICTORS . . . . . . . . . . . . . . . . . . . . . . . . 35 3.7 MAIN STEAM ISOLATION VALVES ............................ 3-5 3.8 REACTOR CORE ISOLATION COOLING . . . . . . . . . . . . . . . . . . . . . . . . . 35
['
3.9 RESIDUAL HEAT REMOVAL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.10 REACTOR WATER CLEANUP SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.11 MAIN STEAM, FEEDWATER, AND BALANCE-OF PLANT PIPING , . . 3-8
- 3.11.1. Pipe Stress Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 3.11.2 Pipe Support Evaluation . . . . . . . . . . . . . . . . ..................... 38 3.12 R E FE R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 4.0 ENGINEERED SAFETY FEATURES .............................. 4-1 l- 4.1 CONTAINMENT SYSTEM PERFORMANCE . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1.1 Containment Pressure and Temperature Respo_a .... ............ . 4-1 4.1.1.1 Long Term Suppression Pool Temperature Response . . . . . . . . . . . . . . . , 4-2 4.1.1.2 Containment Gas Temperature Response . . . . . . . . . . . . . . . . . . . . . . . . . 42
!_ 4.1.13 Short-Term Containment Pressure Response . . . . . . . . . . . . . . . . . . . . . . . 4-3
- 4.1.2 Containment Dynamic Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 1 4.1.2.1 LOCA Containment Dynamic Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.1.2.2 Safety / Relief Valve Containment Dynamic Loads . . . . . . . . . . . . . . . . . . . 4-3 4.1.23 Subcompartment Pressurization ................................ 4-4 4.13 Containme nt Isolation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 y il September 199t
FERMI 2 91150 TABLE OF CONTENTS (Cont'd)
Section Title Page 4.2 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . . . . . . . . . . . . . . . 4-4 4.2.1 High Pressure ECCS (High Pressure Coolant Injection) . . . . . . . . . . . . . . . . 45 4.2.2 RHR System (Low Pressure Coolant Injection) . . . . . . . . . . . . . . . . . . . . . . 45 4.23 1.ow Pressure Core Spray System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.2.4 Automatic Depressurization System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 43 EMERGENCY CORE COOLING SYSTEM PERFORMANCE EVALUATION . . . . . . . . . . . ................... 45 4.4 STANDBY GAS TREATMENT SYSTEM . . . . . . . . . ................ 4-6 4.5 O TH ER ES F SYSTE M S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.5.1 Main Steam Isolation Valve Leakage Control System ................. 46 4.5.2 Post.LOCA Combustible Gas Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.5.3 Emergency Cooling Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.5.4 Emergency Core Cooling Auxiliary Systems . . . . . . . . . . . . . . . . . . . . . . . . . 46 4.5.5 Main Control Room Atmosphere Control System . . . . . . . . . . . . . . . . . . . . 4 4.5.6 Standby Power System ........................................ 4-7 4.5.7 Main Steamline Flow Restrictors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . 4-7 4.5.8 C RD Velocity Limite rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 4.6 R E FE R EN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 5.0 INSTRUM ENTATION AND CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 NUCLEAR STEAM SUPPLY SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.1.1 Control Systems Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1.1.1 Neutron Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1.2 Instrument Se tpoints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 5.1.2.1 RPV High-Pressure Scram . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 5.1.2.2 High-Pressure Recirculation Pump Trip . . . . . . . . . . . . . . . . . . . . . . . . . . 52 5.1.2.3 Pressure Regulator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1.2.4 Safety / Relief Valve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1.2.5 Neutron Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1.2.6 Main Steam High Flow Isolation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 5.1.2.7 Main Steamline High Radiation Scram . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 5.1.2.8 Turbine Stop Valve Closure and Turbine Control Valve Fast Closure Scram Bypass . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 iii September 1991
_ _ _ _ _ _ _ _ _ _ I
FERMI 2 9t.tSO i
TABLE OF CONTENTS (Cont'd) l Section Title Page 5.2 BALANCE-OF PLANT - POWER CONVERSION AND AUXILI ARY SY STEM S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.2.1 Control Systems Evaluation . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.2,1,1 Pressure Control System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 5.2.1.2 T! rbine Control System . . . . . . ............................... 55 5.3 R E FE R EN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-6 6.0 ELECTRICAL POWER AND AUXILIARY SYSTEMS . . . . . . . . . . . . . . . . . . 6-1 6.1 AC POW E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6.1.1 Offsite Power Syste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1.2 Onsite Power Distribution System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 DC POWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 6.3 FUE L POO L C OO LIN G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 (
6.4 WATE R SY STE M S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.4.1 Senice Wate r Syste ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 6.4.1.1 Safe ty- R e l at e d Loa ds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 6.4.1.1,1 Emergency Equipment Senice Water System . . . . . . . . . . . . . . . . . . . . . 6-3 6.4.1.1.2 Diesel Generator Service Water System . . . . . . . . . . . . . . . . . . . . . . . . 64 6.4.1.1.3 Residual Heat Removal Service Water System . . . . . . . . . . . . . . . . . . . . 6-4 6.4.1.2 Nonsafe ty-Related Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.4.2 Main Condenser / Circulating Water / Cooling Tower Performance . . ... . . . . . 6-5 6.4.2.1 Discharge Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 6.4.3 Reactor Building Closed Cooling Water System . . . . . . . . . . . . . . . . . . . . . . 65 6.4.4 Turbine Building Closed Cooling Water System . . . . . . . . . . . . . . . . . . . . . . 65 6.4.5 Ulti m a t e H e at Sink . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 6.5 STANDBY LIQUID CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.6 POWER DEPENDENT HVAC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.7 FIR E PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.8 ' SYSTEMS NOT IMPACTED BY POWER UPRATE . . . . . . . . . . . . . . . . . . 6-9 l 6.8.1 Syste ms With No Impact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 t- t I
i 1
iv September 1991
FERAll 2 91 150 TABLE OF CONTENTS (Cont'd)
Section Title Page 6.8.2 Systems With Insignificant Impact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 6.9 R E FE REN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 7.0 MWER CONVERSION SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7-1 7.1 TURBINE /G ENERATOR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.2 CONDENSER AND STEAM JET AIR EJECTORS . . . . . . . . . . . . . . . . . . . 7-1 73 TURBINE STEAM BYPASS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.4 FEEDWATER AND CONDENSATE SYSTEMS . . . . . . . . . . . . . . . . . . . . . 72 7.4.1 Condensate De mineralizers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.4.2 S tandby Fe e dwat e r Syst e m . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . 7-3 8.0 RADWASTE SYSTEMS AND RADIATION SOURCES . . . . . . . . . . . . . . . . . 81 8.1 LIQUID WASTE MANAG EMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.2 G ASEO US WASTE MANAG E M ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.2.1 O ffgas Syste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 83 RADIATION SOURCES IN THE REACTOR CORE . . . . . . . . . . . . . . . . . 82 8 3.1 O pe ra t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 83.2 Post Ope ration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 8.4 RADIATION SOURCES IN THE COOLANT . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4.1 Coolant Activation Products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4.2 Activated Corrosion Products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 8-3 8.4 3 Fission Produ cts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... 8-3 8.5 RADIATION LEVE LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 8.5.1 Normal Ope ratio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 8.5.2 Post Ope ration /Accide nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 8.53 Offsite Doses (Normal Operation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 1
v September 1991 -
ITRMI 2 91150 TABLE OF CONTENTS (Cont'd) s Sectlon Title Page 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS . . . . . . . . . . . . . . . . . 91 9.1 REACTOR TRANSIENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2 DESIGN BASIS ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 9.2.1 Other Accidents Addressed In USAR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 93 S PECIAL EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 9.3.1 Anticipated Transients Without Scram (ATWS) . . . . . , . . . . . . . . . . . . . . . . 92 9.3.2 Station Blackout (SB0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 9.4 R E FE R EN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 10.0 ADDITIONAL ASPECTS OF POWER UPRATE . . . . . . . . . . . . . . . . . . . . . 10-1 10.1 HIGH. ENERGY LINE BREAK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 10.1.1 Temperature, Pressure, and Humidity Profiles . . . . . . . . . . . . . . . . . . . . . . 10-1 10.1.1.1 Main Steam System Line Break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 (
10.1.1.2 Fe e d wat e r Syst e m Li n e B re ak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 10.1.1.3 High Pressure Coolant Injection System Line Break . . . . . . . . . . . . . . . . 10-2 10.1.1.4 Reactor Core Isolation Cooling System Line Break . . . . . . . . . . . . . . . . . 10 2 10.1.1.5 Reactor Water Cleanup System Line Break . . . . . . . . . . . . . . . . . . . . . . . 10 2 10.1.2 Pipe Whip and Jet Impingement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.1.3 Moderate Energy Line Crack . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 l
10.2 EQUIPMENT QUALIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3 10.2.1 Equipment Qualification of Electrical Equipment ..........,........ 10-3
, 10.2.1.1 Inside Containment ........................................ 10 3 10.2.1.2 O u t sid e Co n t ainm e nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-3 l 10.2.2 EQ of Mechanical Equipment with Non-Metallic Components . . . . . . . . . . 10-4 10.2.3 Mechanical Component Design Qualification . . . . . . . . . . . . . . . . . . . . . . . 10-4 10.3 R EQ UIR E D TESTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-5 10.4 R E FE R EN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-6 i
1 vi September 199t
FERMI 2 91150 f TABLE OF CONTENTS (Cont'd)
Section Title Page 11.0 LICENSING EVALUATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 l 11.1 EVALUATION OF OTHER APPLICABLE LICENSING REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.1.1 Generic Communications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1 11.1.2 Plant Uniqu e Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1 11.1.2.1 Deviation Eve nt Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2 1 11.1.2.2 Regulatory Action Commitment Tracking . . . . . . . . . . . . . . . . . . . . . . . . 11-2 11.1.2 3 Safe ty Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2 11.1.2.4 Temporary Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-2 11.1.3 Emergency Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3 11.2 IMPACT ON TECHNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . 11-3 11.3 ENVIRONMENTAL ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-3 11.4 SIGNIFICANT HAZARDS CONSIDERATION ASSESSMENT . . . . . . . . . 11-4 1 1.4.1 In trod u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-4 11.4.2 Discussion of Issues Being Evaluated . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 11 5 ,
11.4.3 Assessment Against 10CFR50.92 Criteria . . . . . . . . . . . . . . . . . . . . . . . . . 11 13 11.5 R E FE R E N CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 18 s
vii September 199t
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FER3112 91150 LIST OF TABLES Number and Title Page Tabl e 1 - 1 G LO SS ARY O F TE R M S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Table 1-2 CURRENT AND UPRATED PLANT OPERATING CONDITIONS ......................................... 1 11 Table 31 OVERPRESSURIZATION EVENT ANALYSIS . . . . . . . . . . . . . . . . 3-11 Table 3 2 REACTOR INTERNALS AND PRESSURE DIFFERENTIAL ANALYSIS . . . . . . . . . . . . . . . . . ... 3-12 l
Table 4-1 CONTAINMENT PERFORMANCE RESULTS . . . . . . . . . . . . . . . . . 49 Table 4-2 ECCS PERFORMANCE RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 Table 5-1 ANALYTICAL LIMITS FOR SETPOINTS . . . . . . . . . , . . . . . . . . . . . 5-6 Table 6-1 FUEL POOL COOLING ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . 6-13 t Table 91 PARAMETERS USED FOR TRANSIENT ANALYSIS . . . . . . . . . . . 95 Table 9 2 TRANSIENT ANALYSIS RESULTS FOR POWER UPRATE . . . . . . 96
( Table 9-3 LOSS-OF-COOLANT ACCIDENT RADIOLOGICAL CONSEQUENCES . . . . . . . . . . . . . . . . . . . . . . . . 97 Table 9-4 MAIN STEAMLINE BREAK ACCIDENT RADIOLOGICAL CO N S EQ U EN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-8 Table 10-1 HIGH ENERGY LINE BREAK . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7 Table 11-1 TECHNICAL SPECIFICATIONS AFFECTED B Y POWER U PR ATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 21
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LIST OF FIGURES Number and Title Figure 11 UPRATED HEAT BALANCE (NOMINAL) 100% POWER /100% CORE FLOW /1N5 PSIA DOME PRESSURE Figure 12 UPRATED HEAT BALANCE (NOMINAL) 100% POWER /105% CORE FLOW (ICF)/1N5 PSIA DOME PRESSURE Figure 2-1 POWER / FLOW OPERATINO MAP FOR POWER UPRATE Figure 31 RESPONSE TO MSIV CLOSURE WITH FLUX SCRAM Figure 3 2 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE Figure 6-1 SODIUM PENTABORATE SOLUTION VOLUME CONCENTRATION REQUIREMENTS Figure 91 TURBINE / GENERATOR TRIP WITH BYPASS FAILURE - 100 % I UPRATED POWER /105% CORE FLOW Figure 9 2 TURBINE / GENERATOR TRIP WITH BYPASS AND MOISTURE SEPARATOR REHEATER FAILURE -100% UPRATED POWER /105%
CORE FLOW Figure 9-3 FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND - 100%
UPRATED POWER /105% CORE FLOW /374.5'F FEEDWATER TEMPERATURE Figure 9-4 FEEDWATER CONTROLLER FAILURE -
MAXIMUM DEMAND W/ BYPASS FAILURE /100% UPRATED POWER /105% CORE-FLOW /374.5'F FEEDWATER TEMPERATURE Figure 9-5 FEEDWATER CONTROLLER FAILURE -
MAXIMUM DEMAND W/ BYPASS AND MOISTURE SEPARATOR REHEATER FAILURE 100%
UPRATED POWER /105% CORE FLOW /374.5' F FEEDWATER TEMPERATURE Figure 111 RADIAL POWER PROFILE SCHEMATIC Figure 11-2 PCT MARGIN SCHEMATIC Figure 113 TRANSIENT MARGIN SCHEMATIC xi September 1991
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FERhti 2 91 150 EXECUTIVE
SUMMARY
Pursuant to 10CFR50.90, Detroit Edison Company hereby propous to amets ,
Operating License NPF 43 for the Fermi 2 plant by changing the plant Opertiting License !
and Technical Specifications. The proposed changes will allow Fermi 2 to increase l power level from 3293 MWt to 3430 MWt, a 4.2% increase in thermal power and a 5%
increase in steam flow. 'Dils licensing repon presents the results of the analyses and evaluations that support operation with the increase in power and steam flow. The original design power rating for Fermi 2 was 3430 Mwt and 105% steam flow; a number of transients and accidents were evaluated at that power level.
OE Nuclear Energy has published generic guidelines and evaluations for DWR power ;
plant uprates in References 1 and 2. This document follows the generic format and I addresses those issues that should be included in a Fermi 2 specific power uprate licensing l
) report. For those issues that are evaluated on a generic basis, wherever applicable, this i report refers to one of the two references, Fermi 2 specific analyses and evaluations were performed, consistent with the generic guidelines, for systems and components that might be affected to ensure their capability to support the increase in power output and steam flow. Since data is described in detail in
( the USAR, the descriptive material is referenced, as appropriate. The analyses and evaluutions resulted in determinations that the systems and components were either not affected by power uprate or had sufficient design capacity to accommodate uprate conditions.
t' In addition to the above, the effect of uprate on the emironment was assessed to verify that operation of Fermi 2 at uprated power was emironmentally acceptable with established NRC requirements and that consistency was maintained with Federal, State, and local regulations. As a result, no changes to the E 'vironmental Protection Plan or to any of the non NRC permits are required.
T
, Continuing improvements in the analytiv Mddams based on more realistic assumptions, plant performance feedback, and the latest fuel des,gns have resulted in a significant increase in the difference between calculated afety analys resul: nd licensing limits. This margin, combined with the as built systems and comp capa,..f, provides Fermi 2 with the capability to operate at uprated power and ste. flow without major modifications to the plant hardware,
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i September 1991
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FERMI 2 91150 The results of the analyses and evaluations contained herein show that operation of Fermi 2 at the proposed uprated power and steam flow can be accommodated with no significant hazards as set forth in 10CFR50.92 because the proposed changes will not:
(a) Involve a significant increase in the probability or consequences of an acent previously evaluated; or (b) Create the possibility of a new or diffeient kind of accident previously evaluated; or (c) Create a significant reduction in a margin of safety.
This proposed amendment to the Fermi 2 Operating License meets the criteria of 10CFR51.22(c)(9) for a categorical exclusion from the requirement.$ for an emironmental impact statement since it does not:
(a) Involve a significant hazards consideration, or (b) Result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (c) Result in a significant increase in individual or cunmlative occupational radiation exposure.
REFERENCES
- 1. GE Nuclear Energy," Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC 31897P 1, Class 111 (Proprietary),
June 1991.
i 2. GE Nuclear Energy, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC 31984P, Class Ill (Proprietary),
July 1991.
l l
ii Ecptember 1991
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I FERMI 2 91.l!0 l
l l
1.0 OVERVIEW l
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1.1 INTRODUCTION
I The original design power rating for Fermi 2 was 3430 MWt and 105% steam flow. l Severtl transient and accident analyses, including overpressurization, eme,gency core l ;
i cooling / loss of coolant accident, and the radiological consequences of the design basis loss- l
- of coolant and main steamline break accidents were analyzed at the original design power l
! rating. In addition, the radiological effinents resulting from normal plant operation were l evaluated at 3430 MWt. Thus for the current power uprate program, the increase is only l the 2% uncertainty factor added for consenatism.
l l
i A glossary of terms is provided in Table 11.
l t l.2 Pl'RPOSE AND APPROACil 1
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l Uprate Analysis Basis l The plant is currently licensed at 3293 MWt (100%) power. The original safety j analysis basis was that the reactor had been operating continuously at a power !cvel at least l 1.02 times the licensed power level; many of the original analyses had been performed at l 105% steam flow. The uprated power level included in .his evaluation is that power level l associated with 105Fo steam flow which is about 104.2% power. Therefore, as described l l herein, the plant has been reanalyzed at a power level of at least 1.02 times the uprated l power rating which corresponds to 1.063 times the current licensed power level.
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! ' .- - . _h 1.3 UPRATED PLANT OPERATING CONDITIONS ne following evaluations justify increasing the licensed reactor therrnal power to 3430 MWt The following descriptions provide information on the original and uprated f
plant operating conditions.
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!.5 REFERENCES 4- 1. OE Nuclear Energy," Generic Guidelines for General Electrl_c Bolling Water Reactor ,
Power Uprate," Licensing Topical Report NEDC 31897P 1, Class 111 (Proprietary), !
June 1991.
1
- 2. GE Nuclear Energy," Generic Evaluations of General Electric Bolling Water Reactor '
Power Uprate," Licensing Topical Report NEDC 31984P, Class III (Proprietary), July 1991, 1
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16 - September ' 1991__
FEIOli 3 91150
! Table 11 :
GLOSSARY OF TERMS Term Definition ABBA ASEA Brown Boveri Atom A/C Air Conditioning ADS Automatic Depressurization System '
AOO Anticipated Operating Occurrences (Moderate Frequency Transient Events)
AP Annulus Pressurization APRM Average Power Range Monitor ARTS APRM/RBM/ Technical Specifications ASME American Society of Mechanical Engineers ,
ATWS Anticipated Transients Without Scram ATWS RPT Anticipated Transients Without Scram Recirculation Pump Trip BHP Brake Horsepower BOP Balance of Plant BP Bypass i
BWR Boiling Water Reactor
( CC Collapse Criteria Code of Federal Regulations' CFR CLP Containment Long Term Program CPD Condensate Polishing Demineralizers CRD Control Rod Drive ,
CS Core Spray DBA Design Basis Accident DECO Detroit Edison Company DER Deviation Event Report DG ' Diesel Generator-DGSW Diesel Generator Service Water ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EECW . Emergt ey Equipment Cooling Water ;
EESW Emergency Equipment Service Water EFPY Effective Full Power Year EOC- End of Cycle .
EOOS Equipment Out of Senice ,
EO Equipment Qualification FPC Fuel Pool Cooling FW Feedwater FFWTR Final Feedwater Temperature Reduction ,
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I FERMI 2 91150 Table I 1 (cont'd) l GLOSSARY OF TERMS !
Definition j Term FWCF Feedwater Controller Failure FWHOS Feedwater Heater (s) Out-of Service GE GE Nuclear Energy / General Electric Company i GESTAR General Electric Standard Application for Reactor Fuel GSW General Senice Water HCU Hydraulle Control Unit HELB High Energy Line Break HPCI High Pressure Coolant Injection HVAC Heating, Ventilation, and Air Conditioning ICF Increased Core Flow IEB inspection and Enforcement Bulletin IEEE Institute of Electrical and Electronic Engineers IPE Individual Plant Examination IRM Intermediate Range Monitor JR Jet Reaction LCS Leakage Control System (
LFA Lead Fuel Assembly LHGR Unear Heat Generation Rate LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection LPRM Low Power Range Monitor LTR1 Licensing Topical Report No.1 (NEDC-31897P 1) l LTR2 Licensing Topical Report No. 2 (NEDC 31984P)
L MAPLHGR Maximum-Average Planar Linear Heat Generation Rate MCC Motor Control Center MCPR Minimum Critical Power Ratio
( MELC Moderate Energy Line Crack-
, MELLL Maximum Extended Load Line Limit l MEOD Maximum Extended Operating Domain MG Motor Generator MSIV Main Steam Isolation Valve
-MSLB Main Steamline Break MSR Moisture Separator Reheater MWD /T Megawatt Days / Tonne MWt Megawatt thermal .
- NEMA - National Electrical Manufactures Association NPSH Net Positive Suction Head 18 September .1991
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_ _ _ . . _ - _ . - . . _ _ . . _ _ . - _ _ _ _ _ _ - _ _ . _ _ . ~ . _ _ . - _ _ _ _ ..-. _
FERAll 2 91 140 i
I Table 1 1 (cont'd)
GLOSSARY OF TERMS ,
.Ittm Definition NRC Nuclear Regulatorf Commission NSSS Nuclear Steam Supply System NUREG Nuclear Regulations OLMCPR Operating Limit Minimum Critical Power Ratio PCS Pressure Control System PCT Peak Cladding Temperature PFil Partial Feedwater Heating PRA Probabilistic Risk Analysis PULD Plant Unique I.oad Definition RACTS Regulatory Action Commitment Tracking System RBCCW Reactor Building Closed Cooling Water RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RFP Reactor Feed Pump
( RF03 Refueling Outage No. 3 RiiR Residual Heat Removal RIIRSW Residual Heat Removal Service Water RIPD Reactor Internals Pressure Difference RI'T Recirculation Pump Trip RPV Reactor Pressure Valve RR&lS - Regulatory Requirements and Industry Standards RTm Reference Temperature of Nil Ductility Transition RWCU Reactor Water Cleanup RWE Rod Withdrawal Error SAR Safety Analysis Report SBFW Standby Feedwater SBO- Station Blackout -
SE Safety Evaluation-SGTS Standby Gas Treatment System SIL GE Service Information Letter SJAE Steam Jet Air Ejector SLC Standby Liquid Control SLMCPR - Safety Limit Minimum Critical Power Ratio SLO- Single loop _ Operation SPCM Suppression Pool Cooling Mode SRM- Source Range Monitor .
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19 September 1991
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i TERMI 2 91150 ;
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f Table 11 (cont'd)
GLOSSARY OF TERMS ,
Term Definition SRP Standard Review Plan :
SRV Safety / Relief Valve SRVDL Safety / Relief Valve Discharge Line :
STPM Simulated Dermal Power Monitor
- SW Service Water TB Turbine Bypass TBCCW Turbine Building Closed Cooling Water i;
TCV Turbine Control Valve TIP Traversing In Core Probe :
TM Temporary Modifications :'
TSV Turbine Stop Valve T/G Turbine / Generator T/GT Turbine / Generator Trip ,
TPM Thermal Power Monitor USAR/ .
UFSAR Updated Safety Analysis Report / Updated Final Safety Analysis Report (
UHS Ultimate Heat Sink USE Upper Shelf Energy i t
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1 i 2.0 REACTOR CORE AND FUEL, PERFORMANCE i I
l This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 4, that applies to power uprate. :
2.1 FUEL DESIGN AND OPERATION
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and thermal hydraulic characteristics. Because of this similarity, the operating limit _s of the l GE9B fuel bundle are used as a basis for establishing the LFA specific operating limits. In l addition; the LFAs are placed in core locations where they are never expected to be the l most limiting assemblies in the core on either a nodal or bundle power basis. The SVEA 96 l LFA response to power uprate conditions is very similar to the GE9B bundle and, therefore, I:
is judged to be acceptable. l l ,
21 September 1991
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- plant will be operated as described in Section 3.2 of Reference 1. t 2.3 REACrlVin' CONTROL .
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FERMI 2 91150
2.6 REFERENCES
- 1. GE Nuclear Energy,
- Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Ucensing Topleal Report NEDC 31984P, Class 111 (Proprietary),
July 1991.
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24 September 1991
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k Figure 21, POWER / FLOW OPERATING MAP FOR POWER UPRATE September 1991
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1 FER$ll 2 911.(0 i
! 3.0 REACTOR COOfANT SYSTES! AND CONNECTED SYSTE5fS This section primarily focuses on the information requested in Regulatory Guide 1.70, l Chapter 5 and to a very limited extent Chapter 3, that applies to power uprate.
l 3.1 NUCLEAR SYSTEM PRESSURE RELIEF i s
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FERMI 2 91.t50 3.11 MAIN STEAM, FEEDWATER, AND HAIANCE.0F.PIANT PIPING This sectior. addresses the adequacy of all other piping not addressed in Section 3.5 for operation at the uprated conditicm ,
l 3.11.1. Pipe Stress Evaluation ,
Operation at the uprated conditions increases pipe stresses due to slightly higher j
operating t'mperatures, pressures, and flow rates internal to the pipes. For all systems, the -
maximum stress ;evels and fatigue analysis results were reviewed based on specific increases i
in temperature, pressure, and flow rate. These piping systems have been evaluated and meet the appropriate code criteria for the uprated conditions, based on the differences '
between actual stresses and code limits in the original design. Except for a few specific i
lines, all piping has a minimum of 5% margin relative to code allowables. No new l
postulated pipt. break locations were identified.
3.11.2 Pipe Suppen Evaluation Operation at uprated conditions causes a slight increase in the pipe support loadings -
due to increases in the temperature of the affected piping systems. The original support loads due to fluid transients remain unchanged since they were evaluated for uprate flow I.
conditions. liowever, when considering the loading combination with other loads that are i
not affected by power uprate, such as scismic and deadweight, the overall combined support load increase is insignificant.
1 The supports of the systems with the most loading increases (ma.n steam and feedwater systems) were reviewed to determine if there is sufficient margin to accommodate the increased loadings. This review shows that there is adequate difference between the original design stresses and code limits of the supports to accommodate the load increase ^
within the appropriate code criteria. The existing conservatisms are introduced by the use of the lowest code allowables for various plant loading conditions; the use of. generic enveloping design loads instead of actual loads; and the conservative load application on base plates, anchor bolts, and lugs.
38 Septer'ber 1991-
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3.12 REFERENCES 3
i 1. GE Nuclear Energy," Generic Evaluations of General Electric Boiling Water t- -Reactor Power Uprate," Licensing Topical keport NEDC 31984P, Class Ill j (Proprietary), July 1991.
, 2, GE Nuclear Energy, " General Electric _ Standard Application for Reactor i
Fuel GESTAR II, United States Supplement," NEDE 24011 P A 10 US,
- March 1991.
! 3. GE Nuclear Energy, "Maximurn Extended Oper, ig Domain Analysis for
- Detroit-. Edison Company 2nrico Fermi Energy Center Unit 2,"
- NEDC 31843P, Class III, (Proprietary),-July 1990.
l l 4. - GE Nuclear Energy, " Maximum Extended I oad Line Limit and Feedwater i
Heater Out of Service Analyses for Enri %rmi Atomic Power Plant, j Unit 2," NEDC-31515, Revision 1,-(Prop." August 1989.
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FERMI 2 91150 4.0 ENGINEERED SAFETY FEATURES This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 6, that applies to power uprate.
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FERMI 2 91150 4.4 STANDBY GAS TREATMENT SYSTEM The standby gas treatment system (SGTS) is designed to minimize offsite dose rates during venting and purging of both the primary and secondary containment atmosphere under accident or abnormal conditions, wh.ile containing airborne particulates and halogens that might be present. The capacity of the SG13 was selected to provide one secondary containment air volume change per day and maintain the Reactor Building at a slight negative pressure. This capability is not impacted by the uprate.
The design of the charcoal filter beds is unaffected by uprate. The total post LOCA iodine loading increases slightly at the uprated conditions, but is well below the original-design capability of the filter.
4.5 OTHER ESF SYSTEMS Several systems / features are addressed in this section to be consistent with the NRC Safety Evaluation Report (Reference 13) and the Fermi 2 USAR.
4.5.1 Main Steam Isolation Valve Leakage Control System The containment analysis in Section 4.1 demonstrates that the peak post LOCA pressures do not increase beyond the original safety basis. Therefore, the operation of the
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MSIVLCS will not be impacted by power uprate, as noted in Sectiori 6.8.
4.5.2 Post LOCA Combustible Gas Control The hydrogen recombiner operation is required after a LOCA to maintain the containment atmosphere as a. non-combustible mixture. The combustibility of the post LOCA containment atmosphere is controlled by the concentration of oxygen. As discussed in Reference 1, a result of power uprate is that the post LOCA production of oxygen by radiolysis will increase proportionally with the power level. Sufficient capacity exists in the combustible gas control system to accommodate -the increased oxygen production. Also, recombiner operation may be initiated earlier than for the original power level. However, initiation of the recombiners is controlled procedurally based on gas concentration, and not by time.
4.5.3 Emergency Cooling Water System This system is addressed in Section 6.4.
4.5.4 Emergency Core Cooling Auxiliaq Systems These systems are addressed in Section 6.4.
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4-6 September 1991
FERMI 2 91150 4.5.5 Main Control Room Atmosphere Control System This system is not affected by power uprate, as noted in Section 6.8.-
4.5.6 Standby Power System Standby power is addressed in Sections 6.1 and 6.2.
4.5.7 Main Steamline Flow Restrictors The MSL flow restrictors are designed for power uprate, as discussed in Section 3.6.-
4.5.8 CRD Velocity Limiters The CRD velocity limiter design is not affected by power uprate.
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4 ' September 1991
- m. . _ , . - , n FER3112 91150
4.6 REFERENCES
- 1. OE Nuclear Energy," Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31897P 1, Class III (Proprietary),
June 1991,
- 2. General Electric Co.,"The GE Pressure Suppression Containment System Analytical Model, NEDO 10320, Supplement 1, May 1971,
- 3. General Electric Co., "The General Electric Mark Ill Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974,
- 4. General Electric Co., 'The General Electric Mark III Pressure Suppression Analytical Model Supplement 1," NEDO 205331, Containment System September 1975.
- 5. General Electric Co.," Maximum Discharge of Liquid-Vapor Mixtures from Vessels,"
NEDO 21052, September 1975.
- 6. Nuclear Regulatory Commission, " Standard Review Plan, Pressure - Suppression Type BWR Containments," NUREG 0800, Setion 6.2.1,1.C, Revision 6, August 1984.
- 7. U. S. Nuclear Regulatory Commission, " Mark I Containment Long Term Program Safety Evaluation Report," NUREG-0661, July 1980.
- 8. General Electric Co., " Mark I Containment Program Load Definition Report,"
NEDO 21888, Revision 0, December 1978, and Revision 2, November 1981.
- 9. GE Nuclear Energy," Elimination of Limit on Local Suppression Pool Temperature for SRV Discharge with Quenchers," NEDO-30832, Class I, December 1984.
- 10. GE Nuclear Energy," Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31984P, Class 111 (Proprietary),
July 1991.
- 11. NUTECH Engineers, Inc.,"Enrico Fermi Atomic Power Plant Unit 2, Plant Unique Analysis Report," Prepared for Detroit Edison Company, DET-04-028-1, Revision 0.
April 1982.
- 12. GE Nuclear Energy," Fermi 2 SAFER /GESTR-LOCA Loss-of Coolant Accident Analysis," NEDC 31982P, Class III (Proprietary), July 1991.
- 13. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Enrico Fermi Atomic Power Plant, Unit No. 2," NUREG 0798, x
July 1981.
4-8 September t991
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FER. 4 fl 2 91.!!0 5.0 INSTRUMENTATION AND CONTROL This section primarily focuses on the information requested in Regulatory Guide 1.70, ,
I Chapter 7, that applies to power uprate. ;..
i 5.1 NUCLEAR STEAM SUPPLY SYSTEM
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Fl:RSil 2 9t.150 5.2 BALANCE.0F. PLANT POWER CONVERSION AND AUXILIARY SYSTEMS 5.2.1 Control Systems Evaluation Operation of the plant at the uprated power level has minimal impact on the BOP system instrumentation and control devices. Based on uprated operating conditions for the power conversion and auxiliary systems, the process control valves and instrumentation have sufficient range / adjustment capability for use at the expected uprated conditions.
5.2.1.1 Pressure Control System The objective of the pressure control system (PCS) is to provide a fast and stable response to pressure and steam flow disturbances in a manner that will control the reactor pressure within its allowed high and low limits. The PCS consists of the pressure regulation system, turbine control valve system, and steam bypass valve system. . At uprated power conditions, the operating turbine inlet pressure becomes an important consideration in obtaining a desirable operating point on the turbine control valves. Adequate control valve range must be available to assure the ability of the pressure control system to respond to -
system disturbances demanding increased steam flow for the purpose of minimizing pressure excursions. Sufficient' control pressure range during system disturbances is available with power uprate. Thus, the PCS is adequate for power uprate conditions.
5.2.1.2 Turbine Control System No modification to the turbine control valves or the turbine bypass valves are expected for operation at the uprated throttle pressure conditions. No control problems-associated with the increase in steam pressure and flow are expected for power uprate operation.
i I
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5-5 September t99t
_ - _-w FER5ll 3 91150
5.3 REFERENCES
- 1. GE Nuclear Energy," Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC 31897P 1, Class III (Proprietary),
June 1991.
(
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6.0 ELECTRICAL POWER AND AUXILIARY SYSTEMS This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapters 8 and 9, that applies to power uprate.
6.1 AC POWER 6.1.1 Offsite Power System ,
Conformance to General Design Criteria 17 (10CFR50,- Appendix A), which addresses onsite and offsite electrical supply and distribution systems for safety-related components, is not affected since there are no modifications associated with the power uprate which would increase electrical loads beyond those levels previously included or revise the logic of the distribution systems, ne isolated phase bus duct is adequate for rated current output at uprated power.
The main transformers and the associated swiwayard components (rated for maximum transformer output) are adequate for the uprated transformer output.
.- 6.1.2 Onsite Power Distribution System The onsite power distribution system consists of transformers, numerous buses, and switchgear that support or are powered by the buses and switchgear.
Station loads under normal operation / distribution conditions are computed based on equipment nameplate data. Operation at the uprated level is achieved by utilizing existing equipment operating at or below the nameplate rating; therefore, under normal conditions, the electrical supply and distribution components (switchgear, MCCs, cables, etc.) are adequate. ,
Station loads under emergency operation / distribution conditions (emergency diesel generators - EDGs) are based on equipment nameplate data except for CS and RHR pumps where a conservatively high flow brake horsepower (BHP)is used. Operation at the uprated level is achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the stated pumps; therefore, under emergency conditions i
-I the electrical supply and distribution components are adequate.
The amount of power required to perform safety related functions (pump and valve loads) will not be increased with power uprate.
l 1 September 1991 j
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i L- FERA112 91150 w ,
6.2 DC POWER '
The DC loading requirements in the USAR were reviewed, and no power dependent loads were identified.
De DC power distribution system provides control and motive power for various:
, systems / components within the plant._ System loads are computed _ based on equipment nameplate data. Operation at the_ uprated level will not increase any loads beyond-nameplate rating or. revise any control logic; therefore, the DC power' distribution system is adequate.
The RCIC modification discussed in Section 3.8 will add a small DC power valve to -
the system. This additional load will be evaluated during the design process; however, -
preliminary assessment indicates it will have very little impact on the DC power distribution .
systemi 6.3 FUEL POOL COOLING e As a result of operation at the uprated power level, the spent fuel pool heat load will slightly increase.
The fuel pool cooling adequacy is determined by calculating the heat load generated by a full core discharge plus remaining spaces filled with used fuel discharged at regular intervals and calculating the bulk pool temperature (Table 6-1).
The power uprate analysis assumes an 18-month normal fuel cycle as the basis. Each reload will affect the decay heat generation ~in the. spent fuel discharged from the reactor.
The evaluation considered the' expected heat load in the spent fuel storage pool at the-uprated conditions, and confirms the capability of the fuel pool cooling system to maintain :
adequate fuel pool cooling.
This power uprate evaluation was performed on the fuel pool cooling system by reviewing the original design regt. dements and bases for the system and the current plant operating conditions as part of the power uprate program.
The maximum heat load in the fuel storage pool occurs immediately after'a full core discharge. For this case, the heat load is greater than the total design capacity of the two fuel pool heat exchangers, even for current power conditions. The RHR system, however, is more than adequate to provide the needed additional cooling capacity until the heat load
-diminishes to within the capacity of the_ fuel pool heat exchangers alone. For normalL refueling activity, the two fuel pool heat exchangers'have capacity to maintain adequate temperature in the fuel storage pool for uprate conditions, 6-2 September 19911
- _. ;- . - _ _ _ _ . . _ _ _ . . _ ~._ n. . _ -
FFRMI 3 91150 The loss of fuel pool cooling accident described in the USAR was evaluated for power uprate. The radiological consequences increase with power; however, they remain considerably smaller than 10CFR100 limits.
The normal radiation levels around the pool are expected to increase slightly primarily during fuel handling operation. This increase is acceptable and will not significantly increase the operational doses to personnel or equipment.
The fuel racks are designed for higher temperatures than are anticipated from power uprate, Therefore, the racks are acceptable for the higher local decay heat loads.
In summary, the changes are small and are within the design limits of the affected systems and components.
6A WATER SYSTEMS 6A.1 Service Water Systems 6A.1,1 Safety Related Loads
. De safety related heat loads are rejected to one of the three safety related sen> ice
( water systems. These systems are the emergency equipment senice water system (EESW),
the diesel generator service water system (DGSW) and the residual heat removal service water system (RHRSW), which are described below. All heat removed by these systems is rejected to the atmosphere via the ultimate heat sink (UHS), which includes the RHRSW cooling towers.
6A.1.1.1 Emergency Equipment Service Water System The EESW is designed to provide cooling water for the EECW during a loss of offsite power, high drywell pressure, or upon failure of the RBCCW. The EECW provides cooling to various pumps (such as the reactor recirculation, RHR, and CS pumps). The EECW heat loads also include the heat rejected by the drywell coolers and space coolers for such systems as RHR, SGTS, CS, RCIC, EECW, HPCI, and hydrogen recombiner. The EECW also removes heet from the Control Room A/C chiller condenser, the battery charger area, and the switchgear room. There is no increase in the design maximum heat load on the drywell coolers, since the original design loads were based on maximum equipment loads which were based on vessel and component temperature (575'F) greater than the anticipated operating temperatures after uprate (552*F). The design maximum heat loads of the other space coolers do not increase, due to the excess capacity in the original HVAC heat load design (Section 6.6).
(
6-3 September 1991
! FERMI 2 91150 1 6.4.1.1.2 Diesel Generator Service Water System The c'Tesel generator senice water system is designed to provide cooling water for the emergency diesel generators during testing and emergency operation. No change in heat I load is anticipated since no new or increased electrical loads are imposed on the EDGs.
6.4.1.1.3 Residual Heat Removal Service Water System l The following functions of the RHRSW system are affected to a minor degree by the l uprate:
l (a) Increased load from the RHR system when operating in the shutdown cooling mode (b) Increased load from the RHR system when operating in the fuel pool cooling -
mode (backup system)
(c)- Increased load from the RHR system wher .perating in the suppression pool cooling mode, following a postulated LOCA Heat loads from all of these functions increase in direct proportion to the percentage increase in the power level. The RHRSW takes suction from the ultimate heat sink and
( -
returns water to the ultimate heat sink via the cooling towers. .
The RHR cooling towers are designed to provide water with a temperature of 89'F (USAR Subsection 9.2.5) at design ambient conditions, with return water temperature as -
high as ll6*F. The increased heat loads after uprate. result in an RHRSW return temperature increase of less than l'F. Since the uprated maximum return temperature is calculated to be ll5'F no increase in RHRSW supply temperature resdts.
l Use of the RHRSW for flooding purposes following a LOCA is not affected by the uprate.
L 6.4.1.2 Nonsafety Related Loads The service water discharge temperature results from the heat rejected to the service I water system via the closed cooling water systems and other auxiliary heat loads. The major service water heat load in' crease from power uprate reflects an increase in main generator -
losses rejected to the stator water coolers, hydrogen coolers, and exciter coolers,in addition to increased bus cooler heat loads. The increase in service water heat loads from these sources due to uprated operation is projected to be approximately proportional to the-uprate.
d 6-4 - September 1991
- -- -._. .. ..-, . - - , - . - - - . . . . - . _ - , . . - . ~ . . - . . . . . . . . - - . . . - - . . . . _
FERAII 2 91 150 The general senice water (GSW) system is capable of supplying sufficient water to remove the additional heat loads when operating at the uprated power level. GSW is drawn directly ' rom Lake Erie; therefore, the uprate does not affect the inlet temperature to the supplied quipment. As designed, the system will experience a temperamre rise of less than l'F in the bulk outlet temperature which returns to the circulating water system and the cooling towers. This is insignificant to the design of the system.
6.4.2 Main Condenser / Circulating Water / Cooling Tower Performance The main condenser, circulating water, and cooiing tower systems are designed to remove the heat rejected to the condenser and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. Maintaining adequately low condenser pressure assures the efficient operation of the turbine / generator and minimizes wear on the turbine last stage buckets. The cooling towers also remove heat rejected to the GSW system.
De performance of the main condenser was evaluated for the 1991 retubing. This evaluation is based on a design duty over the actual yearly range of circulating water inlet temperatures and confirms that the condenser, circulating water system, and cooling towers are adequate for uprated operation.
I 6.4.2.1 Discharge Limits The state discharge limits shown in the National Pollution Discharge Elbnination System permit were compared to the current discharges and bounding analysis discharges for power uprate. De comparison demonstrates that the plant will remain within the state discharge limit during operation at uprated power and no permit change will be required.
6.4.3 Reactor Building Closed Cooling Water System The RBCCW system is designed to remove heat from the auxiliary equipment housed in the reactor building. The increase in this heat load duc to the uprate is insignificant for all loads except fuel pool cooling. Fuel pool cooling requirements increase in proportion to the power level. However, the system was conservatively designed with design heat loads which bound those anticipated for operation at the uprated power level. Therefore, there is no impact to the system design.
6.4.4 Turbine Building Closed Cooling Water System The TBCCW system provides cooling to both generator related and nongenerator related equipment in the turbine building. The generator-related equipment experiences an increase in load approximately proportional to the percentage of power uprate. However, the equipment heat loads provided by the generator manufacturer were l
6-5 September 1991
FERMI 2 91150 based on a generator rating of 1215 MWe, which is slightly higher than the uprated generator output; therefore, the design heat loads for the generator related equipment bound the actualloads at the uprated conditions. Th: nongenerator related heat loads do not significantly increase with uprate. Therefore, the TBCCW system maximum cooling capacity, as shown in USAR Subsection 9.2.7, will not be exceeded during uprated-operation.
6A.5 Ultimate Heat Sink As a result of operatbn at the uprated power level, the post LOCA ultimate heat sink (UHS) water temperature increases, primarily due to higher reactor decay heat. This results in a higher evaporation rate and, therefore, a higher minimum water inventory requirement in the RHR reservoir.
A review was performed to evaluate the increased water inventory requirement and--
temperature. It was determined that the e 'iting ultimate heat sink system will provide a sufficient quantity of water at a temperature less than 89'F (design temperature) following a design basis LOCA. The Technical Specification for' RHR reservoir water level is-adequate due to conservatism in the original water requirement calculations. Therefore, the UHS design is adequate for uprated operation.
(.
_: 1-6-6 September 1991
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FER$112 91150 6.5 STANDBY LIQUID CONTROL SYSTD1 ,
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FERMI 2 91150
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6.6 POWER DEPENDENT IIVAC Operation of the plant at the uprated power level results in a maximum increase in process fluid temperatures of approximately 6'F in the steam cycle systems, and a maximum of l'F in other auxiliary systems, with the exception of the fuel pool which increases approximately 2*F during its rnaximum loading. 'Ihis results in higher heat gains from piping, piping supports, and mechanical equipment affected by the temperature increase.
The Impact of the higher heat loads was evaluated in all affected areas: the Reactor, Auxiliary, and Turbine Buildings; drywell; areas with supplemental room coolers; areas with emergency room coolers; the Control Center; and the fuel pool area. This evaluation indicates that there is no real increase in calculated total heat gain for most areas of the plant due to high margins in calculated heat gain from pipe supports and high drywell piping temperatures used in the original design calculations. Therefore, there is no increase in maximum design temperatures for these areas.
Two veas in the Reactor and Auxiliary Buildings do have small increases in total heat gain. The RWCU heat exchanger room under normal conditions and the CRD pump room under LOCA conditions experience a total heat gain increase of 2 percent or less, which is negligible when considering inherent heat sinks and existing unit cooler margins.
Expected operating temperatures for these areas remain within the maximum design (.
temperatures of 125'F and 148'F degrees, respectively.
This evaluation concludes that the design heat loads in the original design calculations are adequate for operation at the uprated condition. Therefore, the design basis for the existing liVAC equipment and systems, the normal environmental temperatures used to establish qualified equipment life. and the area design temperature for all plant operating modes are not impacted by the uprate.
6.7 FIRE PROTECTION Operation of the plant at the uprate power level does not affect the fire suppression or detection systems. There are no physical plant configuration or combustible load changes resulting from the uprate. The safe shutdown syuems and equipment used to achieve and rnaintain cold shutdown conditions do not change, and are adequate for the uprated conditions. The operator actions required to mitigate the consequences of a fire are not affected. Therefore, the fire protection systems and analyses are not affected by plant uprate.
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6-8 September t991
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rERMI2 91180 l 1 :
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4 6.8 SYSTEMS NOT IMPACTED BY POWER UPRATE .
I
! 6.8.1 Systems With No impact i
In addition to the systems listed in Table J 1 of Reference 1 and not addressed ,
elsewhere, the following systems are not affected by operation of the plant at the uprated l
- power level
Torus Water Management System i
I Turbine Low Pressure Exhaust Spray Cooling System 4
' Mechanical Vacuum Pumps Lube Oil and Seal Water System i Station Air Systems 1 -
- Interruptible and Non interruptible Control Air Systems i
l Emergency Breathing Air System i Auxiliary Boiler System Tmbine Oil Purification Transfer and Waste System Diesel Generator Auxiliary Systems Divisions I and 11 Diesel Generator Fuel Oil Systems Divisions I and 11 Control Center Air Conditioning (Air and Water Side) Systems Primary Containment Pneumatic Supply System Nitrogen Inerting System t
Main Steam Isolation Valve I.cakage Control Syste:m
. Seismic Monitoring 1
( '
69 September 1991
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4 FERMI 3 91150 i
j 6.8.2 Systems With insignincant Irnpact i
In addition to the systems listed in Table J 2 in Reference 1, some systems are '
affected in a very minor way by operation of the plant at the uprated power level. For these systems, the effects are insignificant to the design or operation of the system and equipment: i Service Water Makeup, Decant and Overflow System RHR Complex Divisir;ns I and II RCIC Turbine Lube Oil System IIPCI Turbine Lube Oil System Heater Feed Pump Lube Oil System Reactor Feed Pump Turbine Lube Oil System Standby Feedwater Lube Oil System Turbine Gland Steam Sealing System
(
Generator Seal Oil System Turbine Lubricating Oil System hrbine Flange lleater System Generator Gas System Condensate Storage and Transfer System Condensate Makeup Demineralizer System Process Sampling Systems Equipment Drains All Doors of Auxiliary and Reactor Buildings.
Post Accident Sampling System Loose Parts Monitoring 1
4-10 September 1991
. ._ _ . - . . _- _ _ - - ~ . . - . _ . _ - . . _ . - - .-. . .- .- - - . - ..
i l'ERhfl 2 91 150
6.9 REFERENCES
- 1. GE Nuclear Energy " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC 31897P.1, Class ill (Proprietary),
June 1991.
4 I
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..~
F r.R M I 2 91 150 Table 61 .
FUEL POOL COOLING ANALYSIS
. Power Uprate Parameters Maximum Bulk Pool Temperature 127 Uprate (* F) 150 Structural Design Temperature ('F)
Bulk Pool Temper A ith %ergency
- Full Core Offloa ,, V'? 3.< nue Capacity, with Su, engmid 125 RilR Cooling ('F; k i September 1991 6 13
FEMAll 3 91 150 O
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) 7.0 POWER CONVERSION SYSTEMS i
I This section primarily focuses on the information requested in Regulatory Guide 1.70,
- Chapter 10, that applies to power uprate. _j The power conversion systems were designed to utilize the energy available from the ,
nuclear steam supply system and to accept the system and equipment flows resulting from continuous operation at 105% of rated steam flow.-
l The system operating pressures and temperatures anticipated for uprated conditions :
are shown in USAR Figure 10.12, the plant heat balance for operation at the original 105%
warranted steam flow. ,
7.1 TURHINE/ GENERATOR ,
The turbine / generator was purchased with a design capability for operating continuously at 105% of rated steam flow conditions. Both the turbine and generator were -
sized to allow operation at this higher flow with a degree of margin to allow control of important variables such as steam inlet pressure.
An engineering review of the turbine / generator was performed to provide additional t assurance that adequate margin exists in the current design to allow successful operation at the higher power level without compromising reliability or equipment life, 7.2 CONDENSER AND STEAM JET AIR EJECTORS The design of the mechanical vacuum pumps are not adversely affected by power uprate and no modification to the system is ' required. ' Die physical size of the primary condenser and evacuation time are the main factors in establishing the capabilities of the vacuum pumps. These parameters do not change. Since flow rates will not change, there will be no change to the two minute holdup time in the pump discharge line routed to the reactor building vent stack. The design capacity of the steam jet air ejectors will not be affected by the uprate since they originally were designed for operation at significantly greater than warranted flows.
7.3 TURHINE STEAM HYPASS The turbine bypass ' system was originally designed for a flow equal to 25% of the original 105% warranted steam flow. The small pressure increase due to power uprate is within the original design capability cf the turbine bypass system. Therefore this system is -
not affected by power uprate.
1 71 September - 1991
.-_ _ _. _ _ .- _ .~. . ~ ~ _ . , . . _ . , . . . _ _ . . _ . _ . _ - - . ._,._,_.a...-,2-.-
rt: n sil 2 91 150 7.4 FEEDWATER AND CONDENSATE SYSTEhlS l The feedwater and condensate systems are designed to provide a reliable supply of feedwater at the temperature, pressure, quality, and flow rate required by the reactor.
These systems are not safety related, however, their performance has a major effect on plant availability and capability to operate at the uprated condition.
For power uprate, an evaluation was performed to determine the capability of the condensate and feedwater system equiprnent to remain within the design limitations of the parameters listed below. The results showed that the design of the system equipment would not be affected by power uprate.
Pump NPSH Ability to avoid suction pressure trip Flow capacity Bearing cooling capability Rated motor horsepower Fullload motor amps Vibration Normal Operation !
The condensate and feedwater systems were originally designed for 105Cc warranted steam Dows. USAR Figures 10.11 and 10.12 show the heat balance at 100C*c and 105G warranted steam flow. Operation at the uprated power sevel does not significantly affect operating conditions of these systems. Sufficient margins exist between the calculated minimum pump suction pressures experienced and the low trip setpoint to provide adequate trip margins during steady state conditions. The existing feedwater design pressure and temperature requirement are adequate.
The feedwater heaters and associated regulating valves were originally designed for 105Cc of warranted flow conditions and, therefore, are adequate for the uprated conditions.
Transient Operation The plant response to a trip of a single reactor feedwater pump (RFP)is the same for power uprate as for current power conditions since the reactor will runback along the same inad line to the same final operating conditions. The trip of a RFP is not expected to result in a reactor scram provided the heater drain system remains in operation.
72 September 1991
i I
ri;RMIa.911$0 l
7.4.1 Condensate Demineralizers .
- The impact of power uprate on the condensate polishing demineralizers (CPD) was !
reviewed. In summan, the system is adequate for uprate operation, but will experience a l
I slight reduction in effectiveness.
\
l The reduction in effectiveness will result in slightly reduced CPD run times. l However, the reduced run times will be acceptable (refer to Section 8 for impact on i radwaste systems).
I
- 7.4.2 Standby Feedwater System The SBFW system was designed with two SDFW pumps with a combined nominal capacity of 1300 gpm and 1247 psig. This is not a safety.related system and is not required to support safe shutdown except for use as an alternate shutdown system during a fire i (NRC SRP 9.5.1). The existing system can meet all of its design objectives. No change to 4 this system is required.
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,7 3 September 1991
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i 8.0 RADWASTE SYSTE51S AND RADIATION SOURCES \
' Itis section primarily focuses on the information requested in Regulatory Guide 1.70,
- Chapters 11 and 12, that applies to power uprate, l
8.1 LIQUID WASTE hiANAGE51ENT l
The liquid radwaste system collects, monitors, processes, stores, and returns processed radioactive waste to the plant for reuse or for discharge.
The single largest source of liquid waste is from the backwash of the condensate demineralizers. With power uprate, the average time between backwash /precoat will be reduced slightly. This reduction does not affect plant safety.
The floor drain collector subsystem and the waste collector subsystem both receive periodic inputs from a variety of sources. Neither subsystem is expected to experience a significant increase in the total volume of liquid waste due to operation at the uprated condition.
The activated corrosion products in liquid wastes are expected to increase proportion +
ally to the power up, ate. Ilowever, the total volume of processed waste is not expected to
' increase appreciably since the only significant increase in processed waste is due to the more Based on a review of plant frequent backwashes of the condensate demineralizers.
operating effluent reports and the slight increase expected from power uprate, it is concluded that the requirements of 10CFR20 and 10CFR50, Appendix 1, will be met.
Therefore, power uprate will not have an adverse effect on the processing of liquid radwaste.
8.2 GASEOUS WASTE 51ANAGE.TIENT The gaseous waste systems collect, control, process, store, and dispose of gaseous radioactive waste gener ated during normal opu ation and abnormal operational occurrences.
The gaseous waste management systems include the offgas system, SGTS (Section 4.4), and various building ventilation systems. The systems are designed to meet the requirements of 10CFR20 and 10CFR50, Appendix 1.
Noncondensible radioactive gas from the main condenser systems, along with air inleakage, normally contains activation gases (principally "N, "O, and "N) and radioactive noble gas parents. The noncondensible gas is continuously removed from the main condenser by the steam jet air ejectors which discharge into the offgas system.
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8t Septernber 1991
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FERMI 2 91150 8.5 RADIATION LEVELS 8.5.1 Normal Operation The effect of power uprate on normal operation radiation levels was evaluated. The actual radiation levels in the plant are expected to increc> slightly, primarily due to increased activation products. For conservatism, many aspects of the plant were originally evaluated based on higher design radiation sources and found to be acceptable. The effect of power uprate on plant radiation levels was reevaluate'd with the conservative assumption that the design radioactive source increase is proportional to the increase in power. De slight increase in calculated radiation levels as a result of the power uprate will not effect the radiation roning or shielding in the various areas of the plant, since it is offset by conservatism in the original design, source terms used, and analytical techniques. This is demonstrated by the differences between the originally calculated and the actually experienced radiation levels.
8.5.2 Post Operation /Accideat This section evaluates the effect of the power uprate on the post operation (post shutdowr.) and post accident radiation levels. Due to the change in core inventory (Section 8.3), post operation / accident radiation levels change slightly. The slight increase g in the post operation / accident radiation levels will have no impact on the design of the plant, the Technical Support Center (TSC), or the Emergency Operations Facility (EOF).
Procedural control of the time spent in any radiation area compensates for increased radiation levels.
8.5.3 OITsite Doses (Normal Operation)
The normal offsite doses are not expected to be significantly affected by operation at the uprated power level. A review was performed to evaluate these effects based on a small general increase in releases. The calculated increase in the offsite radiological consequences, using conservative methods and assumptions, is well below any regulatory limits of 10CFR20 and 10CFR50, Appendix 1.
8-4 September 1991 j
I FElull 2 91 150 i .
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9,0 REACTOR SAFETY PERFORMANCE EVALUATIONS l I
This section primarily focuses on the information requested in Regulatory Guide 1.70, i Chapter 15, that applies to power uprate.
4 l 9.1 REACTOR TRANSIENTS i
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9.4 REFERENCES
- 1. GE Nuclear Energy,' Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC 31897P 1, Class 111 (Proprietary),
June 1991.
- 2. OE Nuclear Energy," Generic Evaluations of General Electric Bolling Water Reactor Power Uprate," Licensing Topical Report NEDC 31984P, Class 111 (Proprietary),
July 1991.
- 3. OE Nuclear Energy,
- Supplemental Reload Licensing Submittal for Reload 2 Cycle 3," 23A6522, Septernber 1990.
b, i. 9-4 Septernber 1991
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FElull 2 91 150 2 10.0 ADDITIONAL ASPECTS OF POWER UPRATE 10.1 lilGil ENERGY LINE HREAK Operation at an uprated power requires a smallincrease in the RPV dome operating pressure to supply more steam to the turbine. The slight increase in the vessel pressure and temperature results in a small increase in the mass and energy release rates following high. energy line breaks (1lELB) Evaluation of these piping systems determined that there is no change in postulated break locations. 10.1.1 Temperature, Pressure, and llumidity Profiles At the uprated power level, high energy lir.e breaks outside the primary containment cause the subcompartment pressure and temperature profiles to . increase as shown in Table 101. The relative humidity change is negligible, in most cases, the increase in the-- blowdown rate is small and the resulting profiles are bounded by the existing profiles due to the consesvatism in the original high energy line break analyses, as discussed for each break. g The high energy line break analysis evaluation was made for all systems evaluated in the USAR. The evaluation shows that the affected building and cubicles that support the safety related function are designed to withstand the resulting pressure and thermalle ',g following a high energy line break, The equipment and systems that suppo.i the safety related function are also qualified for the emironmental conditions imposed. 10.1.1.1 Main Steam System Line Break The critical parameter affecting the high energy line break analysis relative to the power uprate is the smallincrease in reactor vessel dome pressure. As a result of the power uprate, the blowdown rate increases by a small amount, and the differential pressure between subcompartments avJ peak temperature in the steam tunnel increases a very small amount. Profiles for surrounding areas are not significantly affected and are acceptable, 10.1.1.2 Feedwater System Line Break A feed. vater system line break is the critical case for flooding considerations. The flooding rate 4 dependent upon the feedwater system hardware (piping, pump, etc.),.rather than the reactor vessel conditions. Since the feedwater system hardware is not changing, the existing feedwater break flooding analysis is valid for the uprated power condition. I 10 t September 1991 _ _ _ _ _ _ _ _ _ _ _____ _ _ _ _ - - - d
~ - w FERMI 3 91 l$0' 10.1.1.3 High Pressure Coolant Injection System Line Break The steamline break in the HPCI pump / turbine-room is the limiting break for structural design and equipment qualification in the HPCI toom. The peak pressure dW 'rential between rooms increases by an amont which is structurally acceptable. The temperature and relative humidity profiles in the origh.tl design are bounding for operations
- at the uprated power level.
l 10.1.1.4 Reactor Core Isolation Cooling System Line Breal. The steamline break in tlie F.CIC pump / turbine and torus rooms are the limiting breaks fo structural design and equipment qualification in the RCIC room. The mass flow rate increases due to the increase in the reactor dome pressure relative to the original analysis, liowever, the original analysis was performed with conservative model assumptions. Inclusion of more realistic assumptions regarding pipe friction reduces the blowdown rate by more than enough to compensate for the uprate increase. Therefore the - previous HELB analysis is bounding for the uprated core condition. 10.1.1.5 Reactor Water Cleanup System Line Break The pump suction line break in RWCU pump rooms, pump holdup room, and the - discharge line break in the torus room are the limiting breaks for st.metursl design and I equipment qualification. As a result of the slightly increased vessel pressure, the bicwoown rate increases slightly. However, the original analysis v as performed with conservative model assumptions. Conservatism regarding sub. cooled flow would more than offset the e"ect of increase in the pressure thus the original HELB analysis is bounding for the uprated core condition. 10.1.2 - Pipe Whip ad * %pingement Calculations supporting the disposition of potential targets of pipe whip and jet impingement from the _ postulated HELBs have been evalu'ated and determined to be adequate for the safe shutdown effects in the uprated power conditions. Existing pipe whip restraints a*d jet impingement shields and their supporting structures are also adequate for the uprated conditions. 10.1.3 Moderate Energy Line Crack Operation at an uprated level requires a small increase in the reactor vessel pressure - during full power operation. The HPCI, RCIC, RWCU, and CRD systems have moderate I aergy piping. Of these, ths CRD system is the only one that has the limiting crack in any -
- plant area.' The moderate _ energy line crack (MELC) evaluation for the CRD piping was originally done at the system design pressure, which is unaffected by;the power uprate, Therefore the MELC evaluation for the plant is not impacted.
10-2 September .' 1991 -
FERMI 2 91150 4 $ 10.2 EQUIPMENT QUALIFICATION 10.2.1 Equipment Qualification of Electrical Equipment [ The safety related electrical equipment is evaluated in accordance with IEEE 323 to assure qualification for the normal and accident conditions expected in the area where the devices are located. 10.2.1.1 Inside Containment Qualification for safety related electrh equipment h w 'uside the containment is based on environmental conditions resulting from a man n .mline break and/or a DBA/LOCA, as well as those conditions expected during norsa operation. The current accioent and normal design conditions for. temperature, pressure, and humidity are unchanged for power uprate. The radiation levels were increased in relationship to the increase in power. Reevaluation of equipment qualification (EO) has not identified any equipment as unqualified for the uprate environmental conditions. However, the qualified life of certain equipment will be reduced based on increased emironments. The appropriate documentatio') to direct early replacement prior to exceeding its qualified life will be based ( upon aging analysis. 10.11.2 Outside Containment Temperature, pressure, and humidity environments used for qualification of equipment outside contaitanent result from a main steamline break in the pipe tunnel, DBA, or other high-energy line breaks, whichever is limiting for each plant area. The accident temperaturc, pressure, and humidity conditions resulting from a DBA/LOCA do not change with the power level, but some of the HELB profiles do_ increase by a small amount (Section 10.1). Maximum accident radiation levels used for qualification of equipment-outside containment are from a DBA/LOCA. The net result is the accident temperature, pressure, and radiation levels increase due to operation at the uprated power hvel. The accident temperatures increase less than 5'F, the pressures increase less than 1 psi. The radiation levels used in the evaluation were increased in relationship to power. The normal. temperature, pressure, and humidity conditions do not change as a result of power uprate. Reevaluation of EQ equipment has not identified any item as unqualified for.the uprate environmental conditions. However, the qualified life of certain equipment will be reduced based on increased environments. The appropriate documentation to direct early replacement prior to exceeding its qualified life will be based upon aging analysis. 10-3 September 199t _ . _ . _ . _ - ~ _ _ . , _ . , . . . - _ , . _ . . _ . _ _ _._-._ _ , _ __
7 FERatl 2 91.!!0 10.2.2 EQ of Mechanical Equipment with Non. Metallic Components Operation at the uprated power levels increases the normal process temperature up to 6*F. De accident radiation level and the normal radiation level also increase slightly due to uprate and were evaluated as discussed in Section 10.2.1.- Reevaluation of EO equipment has not identified any item as unqualified-for the uprate emironmental conditions. However, the qualified life of certain equipment will be reduced based on increased environments. The appropriate documentation to direct early replacement prior to exceeding its qualified life will be based upon aging analysis. ! 10.2.3 Mechanical Component Design Qualification - f The mechanical design of equipment / components (pumps, heat exchangers, etc.) in certain BOP system '; offected by operation at the uprated power level due to slightly increased temperatura, pressure, and some cases, flow. increases to component nozzle loads and component support loads due to the revised
- operating conditions are insignificant, and become negligible when combined with the governing seismic loads. Therefore, the components and component supports are considered to be adequately designed for the uprated conditions. (
l l l l L ' a 10-4 - September 1991 a i' !
. - ._ -. ~. - . - FERMI 3 91150 10.3 REQUIRED TESTING The testing for power uprate will be conducted based on Reference 1. Specifically the following testing will be performed as a result of power uprate at the time of implementation:
- 1. Normal design and construction acceptance testing will be performed on the RCIC one-inch steam bypass modification which is made to meet SIL 377 recommendations. Testing will be performed at power using the appropriate plant surveillance tests.
- 2. Surveillance testing will be performed on any instrumentation which requires recalibration as a result of power uprate.
- 3. Steady state data will be taken at points from 90% up to the previous rated thermal power so that aperating performance parameters can be projected for uprated power before the previous power rating is exceeded.
4 Power increases beyond the previous rating will be made along an established flow control / rod line in increments of 3% (or less). Steady state operating data will be taken and evaluated at each step.
- 5. Control system tests will be performed for the feedwater/ water level controls and the pressure controls. These operational checks will be p- te at the previous rated power condition and at each 3Fc power increment to show acceptable adjustment and operational capability. The same performance criteria shall be used as in the original power ascension tests.
As discussed in Reference 1, large transient tests will not be performed since this uprate is less than 5rc. Initial plant testing and experience during plant- operation is considered to be sufficient. (, 10-5 September 1991
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.. ~~,- , n ,,. g FERSil 2 91 150 10A REFERENCES
- 1. GE Nuclear Energy," Generic Guidelines for Genera: F.lectric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31897P 1, Class III (Proprietary),
June 1991. ( 10-6 September 1991
FERMI 2 91150 Table 101 lilGil ENERGY LINE BREAK Increase Due to Uprate Area Area Blowdown Rate a Pressure A Temperature Break Location Main steamline break in steam tunnel 1% 0.15 psi s l'F HPCI ECCS break in pump room 4% < 0.10 psi NC RCIC break in pump room or torus room NC NC NC RWCU break in pump room, holdup room, or torus room NC NC NC NC = No Change 10-7 September 1991
- - - . - - - . . - - --- .. . .. - -. -- - ~ . . - - _ . . -
FER$112 91150 11.0 LICENSING EVALUATIONS 11.1 EVALUATION OF OTHER APPLICABLE LICENSING REQUIREMENTS 11,1.1 Generic Communications The analysis, design, and implementation of power uprate was reviewed for compliance with the original licensing basis and with new regulato.ry requirements and operating experience in the nuclear industry. A generic review of the BWR power uprate program for compliance with regulatory requirements from the sources and' industry recommendations listed below is provided in Reference 1. Reference 1 also summarizes the issues resulting from the generic review that require plant - specific review. The sources listed below were reviewed and the review documented for Fermi 2 to ensure that the issues involved were incorporated, as applicable, in the power uprate evaluation. Regulatory Guides Generic Letters TMI Action Items ( Action Items (Formerly Unresolved Safety issues) New Generic Issues Information and Enforcement Circulars (IEC) Inspe'etion and Enforcement Notices (IENs) , Inspection and Enforcement Bulletins (IEBs) GE Service Information Letters (SILs) GE RAPID Information Communication Service Information Letters (RICSIL) INPO Significant Operating Experience Reports (SOER) 11.1.2 Plant Unique items Four types of items whose previous evaluation could be impacted by operation at the uprated power level were also identified. These are Deviation Event Reports (DER); Regulatory Action Commitment Tracking System (RACTS) items Safety Evaluations (SE) 11 September 1991 ___...n-..-,..-.__...__._..-,
%m , m m mm mum - ,
FERMI 2 91150
~
for work in progress and not yet integrated into the plant design b_ asis; and Temporary Modifications (TM) which are not permanent changes, but might be in effect prior to power uprate and still exist after uprate implementation. The DERs, RACTS items, and SEs have been reviewed for possible impact by power uprate and were found to be acceptable for uprate, 11.1.2.1 Deviation Event Reports Deviation Event' Reports are initiated to document plant problems that require evaluation and resolution. The subject of each open DER was reviewed to determine whether it could potentially be impacted by power uprate. Dose items _ that.could have been impacted were further evaluated by review of the actual DER. The followin'g criteria formed the basis for the conclusion that the DERs were not impacted by power uprate: (a) The subject did not impact the power generation capability of the plant. (b) The subject did not impact systems or components important to safety whose operating performance requirements would be impacted by power uprate. (c) The subject did not impact pipe stress, transient, or accident analyses affected by the power uprate. 7 i 11.1.2.2 Regulatory Action Commitment Tracking , Regulatory Action Commitment Tracking System (RACTS) commitments are those made to the NRC, usually by letter. Documents from the NRC that require evaluation and response or corrective action are assigned RACTS item numbers. All RACTS commitments were reviewed using the same criteria as that used in the DER review. 11.1.2.3 Safety Evaluations Safety Evaluations that are approved, but not included in the USAR, Revision 4' , March 1991, or in the 1990 annual Safety Evaluation Summary Report, prepared at the same time, were reviewed. The subject of each SE was evaluated to determine whether it could potentially be impacted by power uprate. The SEs were evaluated against the criteria used for DERs in'section 11.1.2.1. 11.1.2A - Temporary Modifications Temporary Modifications that are in effect at' the of time of _ power uprate-implementation will be reviewed and revised, if necessary, to include uprated conditions. L l 11 2 Septemtier 1991
, _ . _ __ _ . - ._m,.. .___,_._..a._..u_,__.._.-.,._,,.._,..__,.._
l'ERMi 3 91.t$0 i 11.1.3 Emergency Operating Procedures The plant Emergency Operating Procedures (EOP) will be reviewed for any effects of power uprate and will be updated, as necessary. This review will be based on Reference 1, Section 2.3. 11.2 IMPACT ON TECHNICAL SPECIFICATIONS Implementation of power uprate will require revision of a number of the Technical Specifications. Table 11-1 contains a list of the sections that have been identified for the effects of power uprate and a brief description of the nature of the effect. 11.3 ENVIRONMENTAL ASSESSMENT Environmental evaluations were performed to determine that power uprate would not involve an unreviewed emironmental issue and would meet the criteria established for a categorical exclusion not requiring an emironmental review. He non radiological emironmental impacts of the proposed power uprate were reviewed based on the information submitted in the Environmental Report, Operating License Stage (ER/OL); the NRC Final Emironmental Statement (FES); and the ( requirements of the Environmental Protection Plan (EPP), Section 3.0 (Appendix B to the Operating License). Based on this review, it was concluded that the proposed uprate will have insignificant impacts on the non radiological elements of concern and the plant will be operated in an emironmentally acceptable manner as established by the FES, Existing Federal, State, and local regulatory permits presently in effect will accommodate power uprate without modification. Impacts to air, water, and land resources will be essentially non-existent. The effect of power uprate on radiological effluents or offsite doses, as evaluated in the ER/OL and the FES, is insignificant since the original analyses were based on 104.2Fc l (3430 MWt) oflicensed power (3293 MWt). The analyses for power uprate were performed I at 102Fc of uprated power, resulting in a calculated increase of approximately 2Fc in effluents and doses, still well within 10CFR20 and 10CFR50, Appendix I, limits. A slight increase in occupational radiation exposures may occur due to the slight increase in radiation levels in some areas of the plant, primarily due to increased activation products. Conservative assumptions were used; the design radiation source increase is proportional to the increase in power. Even with this assumption, neither individual nor - cumulative occupational radiation exposure will be significantly increased. ( 11 3 September 1991
~ - -
Friott 2 91 150 De proposed power uprate does not require a change to the Environmental i Protection Plan or constitute an unreviewed environmental question since it does not { involve: , (a) A significant increase in any adverse environmental impact previously_ evaluated in the final environmental statement, environmental impact appraisals, or in any decisions of the Atomic Safety and Ucensing Board; or (b) A significant change in effluents or power level; or (c) A matter not previously reviewed and evaluated in the documents specified in Items (a) which may have a significant adverse erwironmental impact. The evaluations also establish that power uprate qualifies for a categorical exclusion not requiring an environmental review in accordance with 10CFR51.22(c)(9) because it does not: (a) InvcNe a significant hazards consideration, or (b) Result in a significant increase in the amounts of any effluents that may be l released offsite, or (c) Result in a significant increase in individual or cumulative occupational radiation exposure, j 11.4 SIGNIFICANT IIAZARDS CONSIDERATION ASSESS > LENT 11.4.1 Introduction l l Uprating the power level of nuclear power plants can be done safely within certain l plant specific limits up to as much as 40% power. Niany light water reactors have already been uprated world wide; many hlWe have already been added by uprate in the United l States among which are several BWRs. h1odification Summary l An increase in electrical output of a BWR is accomplished primarily by supplying higher steam flow to the turbine / generator. hiost BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, and improved fuel and core designs have resulted in a significant increase in the difference between calculated safety analysis results and the licensing limits. This available 11 4 September 199t
1 FERhli 3 91 150 l l safety analysis difference, combined with the excess as designed equipment, system, and component capability, provide GE BWRs with the potential for an increase in their thermal : power rating of between 5% and 10% without major hardware modifications and with no j significant increase in the hazards presented by the plant as approved by the NRC at the l originallicense stage, i The strategy for achieving higher power is to expand the power / flow map by l increasing core flow along existing flow control lines, liowever, there will not be an increase in the maximum recirculation flow limit over the pre uprate value. Uprated operation will also involve slightly higher reactor vessel dome pressure to provide adequate inlet pressure conditions at the turbine, accounting for the larger pressure drop through the steamlines at higher flow, and providing sufficient pressure control and turbine flow capability. Specifically, the following modifications have been identified as necessary to accommodate uprated power operation at Fermi 2: (a) The safety / relief valve setpoints are increased by 25 psig. (b) Power, pressure, and steam feedwater flow related instrument recalibrations and/or related setpoint changes are required. (c) A one inch bypass valve will be installed on the RCIC system inlet steam, as recommended by General Electric SIL 377, to reduce the tendency of the RCIC turbine to overspeed on startup due to the higher operating pressure. (d) The standby liquid control system sodium pentaborate storage tank high and low level alarms and limits are modified to ensure that an equivalent boron concentration of 900 ppm, including a 259 safety margin, can be achieved. This increases the core reload design flexibility while providing adequate shutdown capability. 11A.2 Discussion of issues Being Evaluated
Background
This assessment evaluates the responses for a power uprate license amendment. It is the intention of this safety assessment to summarize the safety significant plant reactions to licensing events and thereby conclude that no significant hazards consideration will be involved. 4 11 5 September 1991
FERMI 3 91150 l Uprate Analysis Basis l l Fermi 2 is currently licensed at 3293 MWt (100% power). The original safety l analysis was based _on the reactor operating continuously at a power level at least 1,02 times l the licensed power level. The uprated power level included in this evaluation is 3430 MWt,- l approximately 1M.2% with 105% steam flow, Several of the original analyses have already l been performed at IN,2% power, including the overpressurization analysis, ECCS/LOCA, l and DBA/LOCA. A 2% power uncertainty factor has been included in the initial conditions l used for the analyses at uprated power level. Therefore, as described herein, the plant has l been reanalyzed at a power level of at least 1.02 times the uprated power rating. l l Margin l l The above uprate analysis basis assures that the margins prescribed by the Code of l Federal Regulations are maintained by meeting the appropriate regulatory criteria. NRC l accepted computer codes and calculational techniques are used to make the calculations that l demonstrate meeting the stipulated criteria. Similarly, margin specified by application of l the American Society of Mechanical Engineers (ASME) design rules has been maintained, l as have other margin-assuring criteria used to judge the acceptability of the plant. l l Fuel Thermal Limits g I l No change is required in the basic fuel desien to achieve the uprated power levels or to meet the plant licensing limi's. l . & The fuel operating limits such as maximum _ average planar linear heat-j generation rate and operating limit minimum critical power ratio will still be met at the j uprated power level. Reload analyses will continue to meet the criteria accepted by the l NRC as specified in Reference 2, New fuel designs will meet acceptance criteria approved I by the NRC. As an example, GE fuel will meet the criteria accepted by the NRC as i specified in Reference 3. 1 l LOCA Evaluations l l The BWR design concept includes various means of supplying water to the reactor. l The basic method to deal with all Loss Of Coolant Accidents is to maintain core integrity l by providing adequate water for core cooling. Consequently, there are high and low l pressure, high and low volume, safety and nonsafety grade means of delivering water to the l vessel. These range from the large feedwater/ condensate pumps down through low pressure l emergency core cooling system pumps, high pressure coolant injection pump, reactor core l isolation cooling pump, standby liquid control pumps, and control rod drive pumps. Many l l of these water supplies are redundant in equipment and systems (e.g. there are several l l LPCI/CS pumps and completely redundant piping systems). I t-6 September 199t
ma > u.. . t . a . , FERMI 3 91150 Power uprate does not result in an increase or decrease in the available water sources nor does it change the selection of those assumed to be functional in the safety analyses. The NRC approved SAFER /GESTR - LOCA methodology was used for the uprated power analyses which is the firs't time this code has been applied at Fermi 2. Design Basis Accidents For BWR licensing evaluations, capability is demonstrated for coping with the full spectrum of hypothetical pipe break sizes including the recirculation, steam, feedwater, ECCS, and instrument lines. This break spectrum concept analytically investigates the full spectrum of large and small, high and low energy line breaks and the success of the plant systems in dealing with them while accommodating a single active equipment failure in addition to the postulated LOCA. Several of the most significant licensing assessments are made using these LOCA ground rules. These assessments are: (a) Challenges to fuel or ECCS performance analyses (Regulatory Guide 1.70, Section 63) in accordance with the rules and criteria of 10CFR50.46 and 10CFR50, Appendix K, wherein the predominant figure of merit is the fuel peak clad temperature. (b) Challenges to the containment (Regulatory Guide 1.70, Section 6.2) wherein the I primary figures of merit are the maximum containment pressure calculated during the course of the LOCA and maximum suppression pool temperature for long term cooling in accordance with 10CFP50, Appendix A, Criterion 38. (c) Calculation of the design basis accident radiological consequences (Regulatory Guide 1.70, Section 15.6.5) and comparison to the criteria of 10CFR100. (a) Challenges to Fuel Emergency core cooling systems are described in the plant USAR, Section 63. The ECCS performance evaluation was conducted through application of the 10CFR50, Appendix K, evaluation models and then showing conformance to the acceptance criteria of 10CFR50.46. As mentioned above, a complete spectrum of pipe breaks was investigated from the largest recirculation line to instrument lines. As shown schematically in Figure 11-2 and numerically in Table 4-2, the licensing safety margin is not impacted by power uprate. The increased PCT consequences for power uprate are insignificant compared to the large amount by which the results are below the regulatory criteria. l 11 7 September 1991 I
FERMI 2 91150 l (b) Challenges to the Containment l l Table 4-1 provides the worst case results of analyses of the plant containment I response to various large and small LOCAs. The impact of power uprate on the peak l values for containment pressure and temperature confirms the suitability of the plant for l operation at uprated power. Likewise, the impact of power uprate on the conditions which l affect the containment dynamic loads has been determined and the plant is jud ed satisfactory for uprated power operation. ? 7 [,j f & . ; . [, n. l (c) Design Basis Accident Radiological Consequences I l Radiological consequences of the design basis accidents that were reevaluated for l power uprate are discussed in Section 9.2 of this report. These accidents were originally l evaluated at 3430 MWt, the uprated power level. The recalculated radiological l consequences remain well within the regulatory acceptance criteria limits of 10CFR100; l therefore, the radiological safety margins established by those limits are maintained. l l Translent Analyses (' l The effects of plant transients are evaluated by investigating a number of l disturbances of process variables and malfunctions or failures of equipment according to a l scheme of postulating initiating events. These events are primarily evaluated against the ' l safety limit minimum critical power ratio etermined usinc NRC-approved methods (Figure 113 c DWpgMgggw - -~ I N2hi$ 1 l OLN1CPR is increased appropnately to' insure that the SLN1CPR is not minnge uponwhen The j the transient is initiated from the uprated power level. Licensing safety margin is not l impacted. Table 9 2 gives results for the most limiting transients for Fermi 2. I l Combined FITects l l Uprate analyses use fuel designed to present NRC approved criteria and operated l within present NRC approved limits to produce more heat in the reactor which slightly l increases reactor pressure and increases steam flow to the turbine. NRC approved design l duty cycle criteria are used to assure mechanical performance at uprate. Design basis l accidents are hypothesized to assess challenges to the fuel, containment, and offsite dose l limits. These challenges are evaluated separately in accordance with conservative regulatory l procedures such that the separate effects are more severe than any combined effects. The 11 8 September 1991
FERStl 2 91 150 offsite dose evaluation specified by Regulatory Guide 13 and Standard Review Plan 15.6.5 provides a more severe DBA radiological consequences scenario than the combined effects of the hypothetical LOCA which produces the greatest challenge to the fuel and/or containment. That is, the DBA which produces the highest PCT and/or containment pressure does not faillarge amounts of fuel and thus the source term and doses are much smaller than those calculated in the Regulatory Guide 13 evaluation. Individual Plant Examination The proposed power uprate has been assessed to have minimum impact on the Individual Plant Examination (IPE) currently underway for Fermi 2. This is not unexpected since the IPE is utilizing traditional Level 1 and Level 2 probabilistic risk assessment (PRA) techniques, which are not highly sensitive to small variations in reactor power level. Some reasons for this insensitivity are discussed below. (a) System success criteria, an integral part of the PRA,'are treated as discrete, e.g., typical success criteria would establish whether one or two pumps are needed to perform an individual function with typical margins in these determinations on the order of 20% to 50%. A 4% change in power level or corresponding decay heat level would have little effect. It is possible that the number of SRVs required to open to prevent excedence of a pressure limit I may increase by one during an overpressure transient due to the higher operating pressure, but the resulting change in potential for failure of the pressure relief function would be trivial due to the number of excess relief valves. (b) Component failure rates are also an integral component of the PRA, but these would not be expected to change since the operating environment of the components would either remain unchanged or change by very slightly. (c) Operator actions, some equipment performance, and other phenomena considered in the PRA are sensitive to timing, but the times are generally on the order of hours, whereas the effects of power uprate would be on the order of minutes. (d) Given radionuclide releases would increase fractionally in proportion to the power increase, i.e., about 4%, but this is a very small change compared to the order of magnitude radionuclide release bins used to characterize releases i in the Level 2 PRA. i For these reasons,it was determined that power uprate would have minimalimpact i on the IPE. ( 11-9 September 1991
I I'ERSti 2 91.t50 Non.LOCA Radiological Release Accidents All other radiological releases discussed in Regulatory Guide 1,70, Chapters 11 an_d 15 are either unchanged because they are not power dependent, or increase at most by the-amount of the uprate. Most of the radiological assessments presented in the US AR for the original power level were run at 3430 MWt. Thus the assessment for these events at uprate, had the calculations been done using the same methodology and assumptions,would be only a 2% increase in the calculated dose. The dose consequences for all of the non.LOCA radiological release accident events are bounded. by the design basis radiological consequences events. Equipment Qualification Plant equipment and instrumentation have been evaluated against the criteria appropriate for uprate Significant groups / types of equipment have beenjustified for uprate by generic evaluations. Some of the qualification testing / justification at the current _ power level was done at more severe conditions than the minimum required. In some cases the-qualification envelope did not change significantly due to uprate. Section 10.2 of this report discusses the plant specific EQ evaluations. Those evaluations confirm that the plant safety-related equipment will still be able to perform its intended functions. Balance-of Plant BOP systems / equipment used to perform safety related and normal _ operating functions have been reviewed for uprate in a manner comparable to that for safety related NSSS systems / equipment. This includes, but was not necessarily limited to all or portions of the main steam, feedwater, turbine, condenser, condensate, essential and non essential service water, emergency diesel generator, BOP piping, and support systems. Significant-groups / types of BOP equipment / systems are justified for uprate by generic evaluations. Plant unique evaluations justify power uprate operation for BOP systems / equipment that are not generically justified. Technical Speellication Changes
=The Technical Specifications ensure that plant and system performance parameters are maintained at the values assumed in the- safety analysis. That is, the Technical Specifications (setpoints, trip settings, etc.) are selected such that the actual equipment is maintained equal to or conservative 'with respect to the assumptions'used in the safety analyses.. The Technical Specification changes justified by the safety analyses summarized in this report are listed in Table 11-1 and summarized below. Proper account is taken of-inaccuracies introduced by instrument drift, instrument accuracy, and calibration accuracy.
For example, the reactor pressure trip is set lower in the Technical Specifications than that used in the safety analysis. This assures that the actual plant behavior will be less severe ( 11 10 September 1991
l'ERSil 2 91 150 than that represented by the safety analysis. Similarly, the Technical Specifications address equipment availability and put limits on equipment out of service such that the actual plant can be expected to have at least the complement of equipment available to deal with plant transients as that assumed in the safety analyses. Since the safety analyses show that the results are acceptable within regulatory limits, the plant can be expected to behave even more mildly to actual transients than portrayed by the safety analyses thereby assuring the public health and safety. Technical Specification setpoint changes associated with the uprated power level are made in accordance with setpoint methodology consistent with that presented in Reference 4 and continue to provide the same level of pmtection as Technical Specifications previously issued by the NRC. The Technical Specification changes for Fermi 2 uprated power operation are summarized as follows: (a) Power increase related changes including:
- 1. Redefining of rated thermal power in Technical Specification Definition 1.32 to reflect the approximately 4.2% increase to 3430 MWt.
I 2. Rescaling of Technical Specification limits for flow biased APRM rod blocks and scrams to account for the redefined rated thermal power while maintaining the same limits in terms of megawatts thermal.
- 3. Rescaling of single recirculation loop operation power limits and associated bases to maintain the same limits in terms of megawatts thermal.
- 4. Rescaling of the idle loop startup restrictions and associated bases to reflect the redefinition of rated thermal power to maintain the current rod lines and absolute power restrictions.
- 5. Rescaling of the core thermal-hydraulic stability operating restrictions and associated bases to reflect the redefinition of rated thermal power such that the same restrictions are conservatively maintained in terms of absolute power level.
(b) Pressure increase related changes including:
- 1. Increasing the dome pressure operating limit by 5 psi to 1(M5 psig.
i
- 2. Increasing the high pressure scram trip setpoints and allowable values by l 25 psi to s 1093 psig and < 1113 psig, respectively.
I k. 11-11 September 199t i
_ - ,m -- %p - FElott 2 91 150
- 3. Increasing the safety / relief valve setpoints by 25 psi such that five valves each are set at 1135,1145, and 1155 psig 1%.
- 4. Redefining the turbine first stage pressure scram bypass at 30% power to reflect the uprated 30% power pressure of 161.9 psig.
- 5. Increasing the HPCI and RCIC surveillance testing pressure by 25 psig to allow testing at the new normal operating pressure.
- 6. Increasing the standby liquid control system surveillance test pressure by 25 psig to allow the appropriate demonstration of system capability at uprated operating conditions.
- 7. Increasing the reactor coolant leakage pressure specification to 1045 psig to limit maximum allowable isolation valve leakage to 1 gpm at the uprated maximum allowable operating dome pressure.
(c) Rated steam flow increase related changes including:
- 1. The main steamline flow increase that accompanies uprated power operation results in increased steamline isolation differential pressure setpoints and allowable values at 137.9% and 139.5% rated steam flow which correspond (
to 118.9 psid and 121.9 psid, respectively.
- 2. The turbine bypass valve / system flow design capabilities in the Technical Specifications bases are changed to approximately 12%G and 25G rated steam flow, respectively.
(d) The pressure / temperature curves are modified to indicate an estimated 5'F increase in the 32 effective full power year core beltline nil ductility shift to 114'F due to increased neutron fluence. The dotted end-of life core beltline curves shift slightly, but remain non limiting. (e) The standby liquid control system high and low level alarms / limits are increased to 3040 gallons and 2560 gallons, respectively to achieve an equivalent baron 10 concentration increase from 825 to 900 parts per million (ppm). This includes a 25Fc margin over the 720 ppm calculated as necessary to shutdown the reactor. The associated bases are revised accordingly. (f) The LOCA analysis methodology identified in Technical Specification Bases 3/4.2 Reference 2,is revised to reflect the NRC-approved SAFER /GESTR methodology
- in "The GESTR LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident SAFER /GESTR Application Methodology", NEDE 237851 PA.
11-t2 September 1991
9 TERMI 3 91150 Similarly, Technical Specification 6.9.3 is revised to reference NEDE 237851 PA.' The application of the SAFER /GESTR methodology climinates the need for the single recirculation loop operation MAPLHGR multiplication factor which has been removed from the Technical Specifications and associated bases. (g) ne containment system Technical Specification bases Section 3/4.6 is revised in several places to indicate that the updated analysis results for peak containment pressure is "less than" the 56.5 psig value which was retained as the bases for these Technical Specifications. 11,4.3 Assessment Against 10CFR50.92 Criteria For this significant hazards consideration assessment, the criteria of 10CFR50.92 were applied to power uprate. The conclusions are based on the evaluations provided in this report, and are summarized as appropriate for the following safety considerations in accordance with 10CFR50.92. (1) Will the change involve a significant increase in the probability or consequences of an accident previously evaluated? The increase in power level discussed herein and associated Technical Specification i changes described above do not significantly increase the probability or consequences of an accident previously evaluated. The probability (frequency of occurrence) of design basis accidents occurring is not - affected by the increased power level because the design or regulatory criteria established for plant equipment (ASME code, IEEE standards, NEMA standards, etc.) are :till imposed.for the uprated power level. Scram setpoints are established such that there is no significant increase in scram frequency due to uprate. The changes in consequences of hypothetical accidents that occur from 1029'c of the uprated power, as opposed to those previously evaluated from 2102% of the current power, are in all cases insignificant since the accident evaluations for power uprate do not_ result in exceeding the NRC approved acceptance limits. The spectrum of hypothetical accidents and transients has been investigated and shown to meet the same regulatory criteria after uprate as before uprate in the area of core design, for example, the fuel operating limits such as maximum average planar linear heat-generation rate and minimum critical power ratio are still met at the uprated power level. Reload analyses will confirm that plant transients meet the criteria accepted by the NRC as specified in GESTAR II (Reference 2). Challenges to fuel or ECCS - performance are evaluated and shown to still meet the criteria of 10CFR50.46 and
-10CFR50, Appendix K (Section 9.2 in this report and Regulatory Guide _1.70, Section 6.3). Challenges to the containment have been evaluated and still meet
( 11 13 September ~ 1991
- _ _ - _ _ l
FERMI 2 91150 10CFR50, Appendix A, Criterion 38, long term cooling, and Criterion 50,- containment. Radiological release events have been evaluated and shown to meet the criteria of 10CFR100 (Regulatory Guide 1.70, Chapter 15). The plant was originally designed anticipating a 5% steam flow increase and many of the analyses were originally performed using 3430 MWt, including overpressurization and ECCS/LOCA. The radiological consequences of the DBA/LOCA and MSLB accidents were imiially calculated at the uprate power value. The only increase in power for the power uprate analyses is the addition of the 2% uncertainty factor, which is not a significant power increase. The increased licensed power level to 3430 MWt and the associated Technical Specification redefinition of rated thermal power have been factored into the aforementioned analyses, as applicable. The results of these analyses, as described in this repon, meet all applicable regulatory criteria and, as discussed above, do not significantly contribute any increase to the probability or consequences of any accident previously evaluated. The power related Technical Specification changes made only to reflect the redefinition of rated thermal power, while maintaining current limits in terms of megawatts thermal, result in no actual change in limits and have no effect on previously evaluated accidents. The pressure increase related Technical Specification changes described above have ( been factored into the safety analyses or have been determined to be non limiting (i.e. turbine first stage pressure - Reference 5, Appendix F.4.2.(c)). The HPCI and RCIC surveillance test pressure specification increase to allow testing at normal operating pressures fulfills the original intent of that specification. The standby liquid control system test pressure increase verifies the ability of the system to pump sodium pentaborate at uprated operating pressure conditions, and is well within the system design pressure. Increasing the leakage rate pressure is consistent with the new maximum allowable operating dome pressure and is conservative. The increase-of the high pressure scram setpoint and of the SRV setpoints by the same amount as the planned operating pressure increase maintains the same level of trip avoidance for the scram and the same simmer margin for the SRVs as originally prosided. Thus, the increase in pressure will not result in an increase in the number of unnecessary scrams or SRV actuations. Since the associated safety analyses have been reperformed and meet the applicable regulatory criteria /ASME code limits and the trip / actuation avoidance margins are maintained, the pressure-increase related Technical Specification changes do not increase the probability or consequences of a previously evaluated accident. The main steamline flow related Technical Specifications are changed to reflect the redefinition of rated main steamline flow that accompanies uprated power operation. The main steamline flow differential pressure trip setpoints are maintained at 137.9% 11 14 September 1991
4 FER3ti 3 91150 and 139.5% of uprated steam flow. These limits were retained to ensure that an adequate trip avoidance margin is maintained for the normal plant testing of MSIVs and turbine stop valves. A high assurance ofisolation protection is still provided for ! a main steamline break accident which meets the original design intent. These setpoints have no effect on the probability of the occurrence of a main steamline break. Also, since a high level of assurance of break isolation is maintained, these i setpoint changes do not significantly increase the consequences of the main steamline break accident. Similarly, the turbine bypass system capabilities are redefined in the Technical Specification bases to reCect the redefinition of rated steam flow. His bases change is informational and has no effect on the safety analyses which conservatively reDect the system capability. Thus, the bases change does not significantly affect the probability or consequences of an accident. The pressure / temperature Technical Specification curve changes reflecting the estimated effect ofinereased neutron fluence are still well within the 200* F nil ductility shift regulatory limit. Also, the upper shelf energy remains greater than 50 ft. lb. for the design life of the vessel. The assumed end-of life core beltline shift represented by the dotted curves remains non limiting. The actual effect of neutron fluence is , determined based on the results ofin vessel surveillance sample Oux wire evaluation. ! Actual fluences are expected to be less than those used in this evaluation. Since the vessel is still in compliance with regulatory requirements, the increase in fluence does ( not significantly affect reactor vessel fracture toughness. Therefore, t!"re is no significant increase in the probability or consequences of an accident previously evaluated. The standby liquid control system sodium pentaborate solution inventory increase proposed Technical Specification change provides the core reload design flexibility to maintain 18-month fuel reload cycles. Cycle length is maintained through higher reload batch sizes or higher enrichment. The proposed change does not change the permissible solution concentrations in the SLC storage tank which could impact the minimum temperature limit. The amount of solution in the tank is increased to achieve the desired 900 ppm equivalent boron-10 concentration. This increase in equivalent boron 10 concentration will provide the required design flexibility while maintaining an adequate shutdown capability. Thus, the probability or consequences of an accident previously evaluated are not significantly increased. Technical Specification changes are made to reflect the use of the NRC-approved SAFER /GESTR - LOCA methodology and to eliminate the single loop operation MAPLHGR multiplication factor. An evaluation has been performed for SLO with no reduction in MAPLHGR. The calculated PCTs are higher than for the corresponding two-loop operation condition, but remained well below the PCF limit i of 2200* F. Furthermore, the required MCPR and MAPLHGR limits at the SLO low , power / low flow operating condition ensure that the calculated SLO / PCT is below 11 15 September 1991 l
i i FERSII 2 91450 that calculated for normal two loop operation. Therefore, these Technical Specification changes will not result in a significant increase in the probability _or consequences of a_ previously evaluated accident.
?
The containment system Technical Specification bases revisions are informational only. As indicated in Table _41 of this report, peak drywell pressure was calculated - to be 49.9 psig, less than the 56.5 psig cited in the bases. The 56.5 psig is retained [
- as a conservative bounding value. Containment leakage rate testing is performed at '
the bounding 56.5 psig value which produces larger indicated leakage rates than if the lower value were used. Since these changes are merely informational and have
- no effect on plant design or operation, they will not result in a dgnificant increase
- in the probability or consequences of a previoucy evaluated accident.
, (2) Will the change create the possibility of a new or diiTerent kind of accident from any-
- accident previously evaluated?
This change will not create the possibility of a new or different kind of accident from f any accident previously evaluated. The full spectrum of accident considerations-defined in Regulatory Guide 1.70 has been reviewed and no new or different kind . of accident has been identified. Power uprate uses already developed technology and
- applies it within the capabilities of already existing plant equipment in accordance with presently existing regulatory criteria to include NRC approved codes, standards, - Y.
and methods. GE has designed BWRs of higher power levels than the uprated j power of any of the currently operating BWRs, and no new power dependent accidents have been identified. In addition, Fermi 2 was originally designed to the
- proposed uprated steam flow (105cc) and many of the accident analyses were l performed at that condition.
p ! The plant systems have been assessed and have been verified to be adequately ;
- designed and capable of performing their design intent at uprated operating .
l conditions. The only modification planned is implementation of the SIL 377 i
- recommended one inch steam inlet bypass valve modification which reduces.the. 1
! RCIC turbine tendency to overspeed before adequate governor- control valve l l hydraulic oil pressure is achieved from the turbine driven oil pump. This } modification will be designed to the same quality standards as the RCIC system and
- will be assessed by a 10CFR50.59 safety evaluation prior to implementation, since the )
. modification does not require a change to the Technical Specifications. l The changes to the Technical Specifications described above do not change the plant l configuration. The Technical Specification' operating parameters of power, pressure, flow, neutron flux, and shutdown boron concentration are increased to accommodate l power uprate, but since there is little change to plant hardware configuration, system , hardware lineups, or operating methods, no new accident scenarios have been I i 11 16 September 1991 -
_ m -# _#r FER3113 91150 identified. The change in accident evaluaticn methodo'ogy to the NRC approved SAFER /GESTR - LOCA code and resulting elimination of the MAPLHGR multiplier under single recirculation loop operation, similarly has no effect on plant configuration and does not affect operating methods in such a way as to create a new or different type of eccident from those previously evaluated. De addition of the clarification that analyzed peak pressures are less than the 56.5 psig peak accident pressure specified is informational and has no effect on the types of accidents that have been established. (3) Will the change involve a significant reduction in a margin of safety? As discussed in Section 5 of Reference 1, the safety margins prescribed by the Code of Federal Regulations have been maintained by meeting the appropriate regulatory criteria. Similarly, the margins provided by the application of the American Society of Mechanical Engineers (ASME) design acceptance criteria have been maintained where applicable, as well as otlier margin assuring acceptance criteria used to judge the acceptability of the plant. Several accident and transient analyses have been reperformed at uprated plant operating conditions consistent with the requested Technical Specification changes. De NRC approved SAFER /GESTR - LOCA methodology was used in the LOCA analysis. Additionally, Reference 1 addresses the BWR generic acceptability of analytical evaluations for the loss of feedwater transient, stability, core spray distribution, safety limit minimum critical power ratio, containment atmosphere combustibility, materials and coolant chemistry, and anticipated transients without scram (ATWS). The radiological doses have been recalculated for the events discussed in Section 9.2 of this report. As discussed in Section 5.2.3 of Reference 1, actual offsite doses for the DBA/LOCA will increase proportionately to reactor power vehen compared on a consistent basis. The doses presented in Tables 9.3 and 9.4 of this report are not comparable on a consistent basis to previous calculations since the NRC approved methodology and assumptions used have undergone extensive revision. However, as can be dete: mined from these tables, the required acceptance criteria of 10CFR100 are still met. The radiological doses resulting from the DBA/LOCA and MSLB accidents were initially analyzed at the uprated power level (3430 MWt). For the-current power uprate program, the increase is only the 2% uncertainty factor added for conservatism. ( It t7 September 1991
FERMI 2 91150 As discussed throughout this report;in each case the relevant acceptance criteria is met which preserves the margin of safety provide.d by these criteria. It is therefore concluded that the requested changes do not involve a significant reduction in any ' margin of safety. Conclusions This report describes investigations made. for a power uprate. to 3430 MWt, equivalent to 105% of original steam flow. A 5% increase in steam flow was factored into . the original design basis for Fermi 2. Several major analyses, including overpressurization, ECCS/LOCA, and the' radiological consequences of the DBA/LOCA were originally performed at 3430 MWt. The strategy for- achieving higher power is to expand the power / flow map by-increasing core flow along existing flow control lines. The predominant plant licensing c' hallenges have been reviewed to demonstrate that this uprate can be implemented without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any presently- existing regulatory limits applicable to the plant which might cause _a significant reduction in a margin of safety. Having arrived at negative declarations with respect to these considerations of 10CFR50.92, this assessment concludes that-uprate of the amount described herein does not involve a ( significant hazards consideration. l 11 18 September - 199t7
.7.
FERMI 2 91150
11.5 REFERENCES
- 1. GE Nuclear Energy, " Generic Evaluations of General Electric' Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31984P, Class III (Proprietary), July 1991.
- 2. GE Nuclear Energy, " General Electric Standard Application for Reactor Fuel -
GESTAR II, United States Supplement," NEDE 24011 P A 10-US, March 1991.
- 3. - GE Nuclear Energy, " Licensing Criteria for Fuel Designs," NEDO 31908, Amendment 22 to NEDE 24011 P A, February 1991.
- 4. General Electric Co., " Instrument Setpoint Methodology," NEDC 31336, Class III (Proprietary), October 1980.
- 5. GE Nuclear Poergy," Generic Guidelines for General Electric Boiling Water Reactor Power Uprat:,' Licensing Topical Report NEDC 31897P-1, Class III (Proprietary),'
June 1991. (
+
t 11 19 September 1991
n~n . . . _ FERMI 3 9t.150-Table 11 1 - TECHNICAL SPECIFICATIONS AFFECTED BY POWER UPRATE Location Effect Definition , 1.32 Revise value of rated thermal power definition to uprated power-level Safety Limits and Limiting Safety System Settings Table 2.2.11 Revise the APRM scram and reactor vessel steam high dome pressure of the reactor protection system setpoints Limiting Conditions for Operation and Surveillance Requirements-( . . Figure 3.1.51 Revise to implement 900 ppm minimum boron 10 concentration l i provided by SLC 1 , L 4.1.5.c Increase SLC test pressure by 25 psig to 12.5 psig-t 3.2.1 Eliminate the MAPLHGR limit for single loop operation Table 3.3,11 Revise value of turbine first stage pressure to be equivalent to that-at 30fc of rated thermal power in Notation j and Action 6 Table 3.3.2 2 Revise the main steamline high flow differential pressure setpoints Table 3.3.6 2 Revise the APRM control rod block instrument setpoints 3.4.1.1 Revise the thermal power value and eliminate the MAPLHGR limit
~
for single-loop operation 4.4.1.1.3 Revise the thermal power value for single loop operation - Figure 3.4.1.41 Revise the thermal power value for idle loop startup L l- 3.4.2.1 Revise values of SRV setpoints 3.4.3.2.d Revise leaxage rate pressure anplicability i-l. 11-21 September 1991 L
FERhli 2 91 150
~
Table 111 (continued) TECHNICAL SPECIFICATIONS AFFECTED BY POWER UPRATE Location EUect Limiting Conditions for Operation and Sunelliance Requirements '(cont'd) Figure 3.4.6.11 Revise vessel pressure / temperature cun'es 3.4.6.2/4.4.6.2 Revise the value of the reactor steam dome pressure Figure 3.4.10-1 Revise the thermal power value for core thermal hydraulic stability 4.5.1.b.3 Revise HPCI test pressure 4.7.4.b Revise RCIC test pressure BASES for Sections 3 and 4 ' i 3/4.1.5 Revise to implement 720 ppm minimum (900 ppm minimum with 25Fc margin) boron concentration for SLC 3/4.2 Revise reference-- to reflect the - NRC approved generic GE ' SAFER /GESTR LOCA analysis basis 3/4.2.1 Eliminate discussion of NIAPLHGR reduction factor for SLO: 3/4.4.1 Revise to reflect single loop operation thermal power and eliminate the NIAPLHGR reduction factor 3/4.4.10 Revise to reflect thermal power for thermal hydraulic stability. 3/4.6.1.2 & Revise to indicate that uprated analysis demonstrates maximum 3/4.6.1.6 expected pressure is less than 56.5 psig
. Revise the reactor steam dome pressure,-
3/4.6.2 Revise the containment accident pressure from "approximately" to "less than" 56.5 psig-3/4.7.9 . Revise the bypass valve capabilities to reflect the redefinition of-rated steam flow. 11 22 September 199t
.._-e_..w.,.u,--. ,e .%,.,,4+ .-,_w c,, . . + - . , .,..-,..e..-,,,4.. y a ,, ,,y,..%..w.r.~e.,, u7,..,m,y. ..~,,-c.,--, s....,,.,, e ,,-.%wm
_ . _ ., .__ _._,_. _ m__,. - FERMI 3 91150 , Table 11 1 (continued) TECilNICAL SPECIFICATIONS AFFECTED IW POWER UPRATE Location Effect AdminL .ative Controls 6.9.3 Revise to add the generic GE SAFER /GESTR - LLCA reference Note: As used in this table, the term "setpoints" include the nominal trip setpoints and allowable values (wheie applicable). ( ( 11 23 September 1991
l - o 0 . v Oi September 1991
I a 1 i O I l i l i i l i i l i i l i 1 I O i i O : September 5991
o O O September 1991
t GENERAL ELECTRIC COMPANY AFFIDAVIT I, DAVID J. ROBARE, being duly sworn, depose and state as follows:
- 1. I am Manager, Plant Licensing Services, General Electric Company, and hva been delegated the function of reviewing the information described u s'.ragraph 2 which is sought to be withheld and have been authorized h e..nply for its withholding.
- 2. the information sought to be withheld is contained in Detroit Edison Report fermi 2-91 150 " Power Uprate Safety Analysis' September 1991.
The GE Proprietary-portions of this report are identifiable by the "GE Proprietary Information" designation at the top of the page.
- 3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement of Torts, Section 757. This definition provides:
"A trade secret may consist of any formula, pattern, device or compilatiin of information which is used in one's business and which gives him an opportunity to obtain an advantage over competitors who do not know or use it...A substantial element of secrecy must exist, so that, except by the use of improper means, there would be difficulty in acquiring information...Some factors to be considered in determining whether given information is one's trade secret are (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business;'(3) the extent of measures taken by him to guard the secrecy of the informationt information to him and to his competitors; (5)(4) the value the amount of the of effort or money expanded by him developing the information; (6) the ease or difficulty with which the information (7 sid be properly acquired or duplicated by others."
- 4. Some examples of categories of-information which fit into the-definition of Proprietary Information are:
- a. Information that disclose:. a process, method or apparatus where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic
- advantage over other companies;
- b. Information consisting of supporting data and analyses, including test data, relative to a- process, method or apparatus, the application of which provide a competitive economic advantage, e.g.,
s by optimization or improved marketability; Infonatha b thh scxd wu deleted la ar. ,@c: a the Fadr. of Inknalbn Act, Ed W5 l Fo\ A. _$/-M 9
GENERAL ELECTRIC COMPANY AFFIDAVIT
- c. Information which if used by a competitor, would reduce his expenditures of resources or improve his competitive position in the design, licensingmanufacture, shipment, installation, assurance of quality or of a similar product;
- d. Information which reveals cost or price information, production capacities, budget Electric, its customers or suppliers; levels or commercial strategies of Ge
- e. Information which reveals aspects of past, present or future General Electric customer funded development plans and programs of potential commercial value to General Electric;
- f. Information which disclosh patentable subject matter for which it may be desirable to obtain patent protection;
- g. Information which General Electric must treat as proprietary according to agreements with other parties.
5. made by the Subsection Manager of the originating c who is most likely to be acquainted with the value and sensitivity of , the information in relation to industry knowledge. Access to such documents within the Company is limited on a "need to know" basis and such documents are clearly identified as proprieta y. 6. The procedure for approval of external release of such a document typically requires review by the Subsection Manager, Project Manager, Principal Scientist or other equivalent authority, by the Subsection f Manager of the cognizant Legal Operation for technical Marketing content,function (or delegate) and by the determination of the accuracy of the proprietary designation incompetitively ef accordance with the standards enumerated above. General Electric are generally limited to regulatory bodies, customersDisclosu ' and potential customers and their agents, suppliers and licensees then only with appropriate proprietary agreements. protection by applicable regulatory provisions or 7. The document mentioned in paragraph 2 above has been evaluated in accordance with the above criteria and procedures and has been found to contain information in confidence which by General is proprietary and which is customarily held Electric. 8. been held in confidence by GeneralThe information to the best of my Electric Company, no public disclosure has been made, and it is not available in public sources. (
\
A
I GENERAL ELECTRIC COMPMY AFFIDAYli
- 8. All disclosures to third parties have been made pursuant to regulatory provisions of proprietary agreements which provd de for maintenance of the information in confidence.
- 9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities. A substantial effort has been expended by General Electric to develop this information.
(
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s. GENERAL ELECTRIC COMPANY AFFIDAVIT STATEOFCALIFORNIA) ss: COUNTY OF SANTA CLARA David J. Robare, being duly sworn, deposes and says: That he has read the foregoing affidavit and the matters stated therein are truly and correct to the best of his knowledge, infor:74 tion, and belief. Executed at San Jose, California, this il day of SEntM6til 19 *)) . David J. Robare General Electric Company ( Subscribedandswornbeforemethisk day of IM/+\ut 19 4 OFFICI AL, SE A L f* PAULA F. HUS$rf .
'I notta t'svc . cwr:WA el,
. R.Qgg; y, $22 %$,$U 1m j kotary Public, State' of Capifornia N.
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ __ ._}}