ML20245A635
| ML20245A635 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/11/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duquesne Light Co, Ohio Edison Co, Pennsylvania Power Co |
| Shared Package | |
| ML20245A637 | List: |
| References | |
| DPR-66-A-139, NPF-73-A-014 NUDOCS 8904250307 | |
| Download: ML20245A635 (33) | |
Text
-
1 o esog'o,(
.i, UNITED STATES f
,'v.c[i NUCLEAR REGULATORY COMMISSION j
WASWNGTON,D. C. 20S55
)
DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWFR. STATION, UNIT NO. 1 l
l AMENDFFKT TO FACILITY OPERATING LICENSE j
Amendment No. 139 License No. DPR-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, et al.
(the licensee) dated January 5,1989 complies with the standards and j
l requirements of the Atomic Energy Act of 1954, as amended (the Act)
)
and the Comission's rules and regulations set forth in 10 CFR l
I Chapter I-B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission-1 l
l C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all apolicable requirements have been satisfied.
l l
8904250307 890411 PDR ADOCK 05000334 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. DPR-66 is hereby amended to read as follows:
~~
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.139, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective on issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 7
John F. Stolz. Director Project Directorate I-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation I
Attachment:
Changes to the Technical Specifications l
Date of Issuahce: April 11, 1989 I
ATTACHMENT TO LICENSE AMENDMENT NO. 139 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following pages of the Appendix A (Technteel Specifications) with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert VII VII' VIII VIII IX IX X
X XIII XIII l
XIV XIV l
XV XV XVI XVI XIX XIX XX XX XXI XXI l
XXII XXII XXIII XXIII XXIV XXI V '
l XXV XXV XXVI XXVI 3/4 4-8 3/4 4-8 3/4 11-22 3/4 11-22 6-13a 6-13a 6-15 6-15 6-22 6-22 B 3/4 2-5 B 3/4 2-5 l
l i
~
1 INDEX
- .!!'ITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS TECTICN PAGE 3/4.4.9.2 Pressurizer.
3/4 4-27 3/4.4.9.3 Overpressure Protection Systems.
3/4 4-27a 3/4.4.10 STRUCTURAL INTEGRITY - ASME Code Class 1,
2 and 3 Components 3/4 4-28 l
3/4.4.11 RELIEF VALVES 3/4 4-29 3/4'.4.12 Reactor Coolant System Vents 3/4 4-32 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 1 3 /4. 5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T 2 350'F 3/4 5-3 avg J/4.E.3 ECCS SUBSYSTEMS - Tavg < 350'F 3/4 5-6 3/4.5.4 BORON INJECTION SYSTEM 3/4.5.4.1.1 Boron Injection Tank 2 350*F 3/4 5-7 3/4.5.4.1.2 Boron Injection Tank < 350*F 3/4 5-7a 3/4.E CONTAINMENT SYSTEMS, 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 Containment Integrity.
3/4 6-1 3/4.6.1.2 Containment Leakage.
3/4 6-2 3/4.6.1.3 Containment Air Locks.
3/4 6-5 3/4.6.1.4 Internal Pressure.
3/4 6-G 3/4.6.1.5 Air Temperature.
3/4 6-8 3/4.6.1.6 Containment Structural Integrity 3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 Containment Quench Spray System.
3/4 6-11 BEAVER VALLEY - UNIT 1 VII Miendment No. 144, 139 i
)
J INDEX l
LIMITING CONDITIONS FOR OPERATION AND' SURVEILLANCE REQUIREMENTS SECTION PAGE
)
3/4.6.2.2 Containment Recirculation Spray System 3/4 6-13 3/4.6.2.3 Chemical Addition System
. ' ~.
~
3/4 6-15
)
3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-17 i
/ 4. 6. 4 COMBUSTIBLE GAS CONTROL I
3/4.6.4.1 Hydrogen Analyzers 3/4 6-20 3/4.6.4.2 Electric Hydrogen Recombiners.
3/4 6-21 1
3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.5.1 Steam Jet Air Ejector.
3/4 6-25 3/4.7 PLANT SYSTEMS 1
./4.7.1 TUREINE CYCLE 3/4.7.1.1 Safety Valves.
3/4 7-1 3/4.7.1.2 Auxiliary Feedwater System 3/4 7-5 3/4.7.1.3 Primary Plant Demineralized Water (PPDW) 3/4 7-7 3/4.7.1.4 Activity 3/4 7-8 3/4.7.1.5 Main Steam Line Isolation Valves 3/4 7-10 3/4.7.?
STEAM GENERATOR PRESSURE / TEMPERATURE prMITATION 3/4 7-11 3/4.7.3 COMPONENT COOLING WATER SYSTEM 3/4 7-12 3/4.7.4 REACTOR PLANT RIVER WATER SYSTEM 3/4 7-13 l
3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER.
3/4 7-14 3/4.7.6 FLOOD PROTECTION 3/4 7-15 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS 3/4 7-16 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM 3/4 7-19 BEAVER VALLEY - UNIT 1 VIII klendment No. 7S0,139
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 5
EECTION PAGE 3/4.7.9 SEALED SOURCE CONTAMINATION.
.7 3/4 7-22 3/4.7.12 SNUBBERS 3/4 7-26 3/4.7.13 AUXILIARY RIVER WATER SYSTEM 3/4 7-34
-j l
3/4.7.14 FIRE SUPPRESSION SYSTEMS 3/4.7.14.1 Fire Suppression Water System.
3/4 7-35 3/4.7.14.2 Spray and/or Sprinkler System.
3/4 7-39 l
l 3/4.7.14.3 Low Pressure CO System.
3/4 7-41 2
3/4.7.14.4 Fire Hose Stations 3/4 7-42 1
3/4.7.14.5 Halon Systems 3/4 7-43 3/4.7.15 FIRE RATED ASSEMBLIES 3/4 7-44 l
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 3/4.8.1.1 Operating.
3/4 8-1 3/4.8.1.2 Shutdown 3/4 8-5 l
l 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS 1
3/4.E.2.1 A.C. Distribution - Operating.
3/4 8-6
'3/4.8.2.2 A.C. Distribution - Shutdcwn 3/4 8-7 3/4.8.2.3 D.C. Distribution - Operating.
3/4 B-8 3/4.8.2.4 D.C. Distribution - Shutdown 3/4 8-10 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.
3/4 9-1 3 /4. 9. 2 INSTRUMENTATION 3/4 9-2 3/4.9.3 DECAY TIME 3/4 9-3 BEAVER VALLEY - UNIT 1 IX Amendment No. li4,139 l
E_______-------_---
INDEX l.
i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS CFOT70N PAGE
]
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 COMMUNICATIONS 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY,
3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING 3/4 9-7 3/4.9.8.1 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.
3/4 9-8 3/4.9.8.2
-LOW WATER LEVEL,
3/4 9-8a 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION 1
SYSTEM 7/4 9-9 3/4.9.10 WATER LEVEL-REACTOR VESSEL 3/4 9-10 I
3/4.9.11 STORAGE POOL WATER LEVEL 3/4 9-11
{
3/4.9.12 FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT 3/4 9-12 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM'- FUEL
. \\
i STORAGE 3/4 9-13 l
3/4.9.14 FUEL STORAGE - SPENT FUEL STORAGE POOL 3/4 9-14
.i 3/4.9.15 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS 3/4 9-16 3/4.10 SPECIAL TEST EXCEPTIONS l
3/4.10.1 SHUTDOWN MARGIN.
3/4 10...........
3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.
3/4 10-2 BEAVER VALLEY - UNIT 1 X
Amendment No. itC,139
_________Q
INDEX 8
BASES SFCTION PAGE 3/4.3.3.7 Chlorine Detection Systems B 3/4 3-3 3/4.3.3.8 Accident Monitoring Instrumentation ~.
B 3/4 3-3 3/4.3.3.9 Radioactive Liquid Effluent Monitoring Instrumentation.
B 3/4 3-4 3/4.3.3.10 Radioactive Gaseous Effluent Monitoring Instrumentation.
B 3/4 3-4 3 /4. 4 REACTOP COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS B 3/4 4-1 2.'4.4.2 and 3/4.4.3 SAFETY VALVES B 3/4 4-la 3/4.4.4 PRESSURIZER.
B 3/4 4-2 af 4.4.S SIEAM GENERATORS B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-3 3/4.4.6.1 Leakage Detection System B 3/4 4-3 3/4.4.6.2 Operational Leakage.
B 3/4 4-3 3/4.4.7 CHEMISTRY.
B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.
B 3/4 4-4
? /3,4
^
PRESSURE.' TEMPERATURE LIMITS B 3/4 4-E 3/4.4.10 STRUCTURAL INTEGRITY B 3/4 4-10 3/4.4.11 RELIEF VALVES B 3/4 4-10 3/4.4.12 REACTOR COOLANT SYSTEM VENTS B 3/4 4-11 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS B 3/4 5-I 3 /4. 5. 2 and 3/4.5.3 ECCS SUBSYSTEMS B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM B 3/4 5-1 I
BEAVER VALLEY - UNIT 1 XIII Amendment No. 110, 139
INDEX 3
EASES SECTION EA_GE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT
" ~ ~
3/4.6.1.1 Containment Integrity.
B 3/4 6-1 3/4.6.1.2 Containment Leakage.
B 3/4 6-1 3/4.6.1.3 Containment Airlocks B 3/4 6-1 3
3/4.6.1.4 and 3/4.6.1.5 Internal Pressure and Air Temperature.
B 3/4 6-2 3/4.6.1.6 Containment Structural Integrity B 3/4 6-2 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 Containment Quench and Recirculation Spray Systems.
B 3/4 6-2 3/4.6.2.3 Chemical Addition System B 3/4 6-3 l
l 1
3/4.6.3 CONTAINMENT ISOLATION VALVES B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL.
B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.5.1 Steam Jet Air Ejector.
B 3/4 6-3 i
3 / 4. ~7 PI W
C Y.CTD'c 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 Safety Valves.
B 3/4 7-1 3/4.7.1.2 Auxiliary Feedwater Pumps.
B 3/4 7-2 3/4.7.1.3 Primary Plant Demineralized Water.
B 3/4 7-2 3/4.7.1.4 Activity B 3/4 7-3 3/4.7.1.5 Main Steam Line Isolation Valves B 3/4 7-3 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION B 3/4 7-4 BEAVER VALLEY - UNIT 1 XI V Amendment No. 740, 139
s-q j
/
INDEX BASES 1
SECTIOU PAGE 3/4.7.3 COMPONENT COOLING WATER SYSTEM B 3/4 7-4 3/4.7.4 RIVER WATER SYSTEM B.3/4 7-4 3/4.7.5 ULTIMATE HEAT SINK B 3/4 7-4 3/4.7.6 FLOOD PROTECTION B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEM B 3/4 7-5 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION.
B 3/4 7-5 I
i 3/4.7.12 SNUBBERS B 3/4 7-6 1
3/4.7.13 AUXILIARY RIVER WATER SYSTEM B 3/4 7-7 3/4.7.14 FIRE SUPPRESSION SYSTEMS B.3/4 7-7 3/4.7.15 FIRE RATED ASSEMBLIES B 3/4 7-7 2/4.8 ELECTRICAL POWER SYSTEME 3/4.8.1 AND 3/4.8.2 A.C.
- SOURCES, D.C.
SOURCES AND i
ONSITE POWER DISTRIBUTION SYSTEMS B 3/4 8-1
? /4.9 PEEL'ELING OPEP3.TIONS 3/4.9.1 BORON CONCENTRATION.
B 3/4 9-1 3/4.9.2 INSTRUMENTATION B 3/4 9-1 3/4.9.3 DECAY TIME B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS B 3/4 9-1 3/4.9.5 COMMUNICATIONS B 3/4 9-2 3/4.9.6 MANIPULATOR CRANE OPERABILITY.
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4 9-2 BEAVER VALLEY - UNIT 1 XV Amendment No. TM,139
P
.INDEX 4
J BASES I
2 ETCTION PAGE
~
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.
B 3/4 9-2
' '4. 9. 9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM B 3/4 9-2 3/4.9.10 and,3/4.9.11 WATER LEVEL-REACTOR VESSEL AND STORAGE POOL B 3/4 9-3
)
i 3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM.
B 3/4 9-3
)
l 3/4.9.14 TUEL STORAGE - SPENT FUEL STORAGE. POOL B 3/4 9-3 j
3/4.9.15 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/~.10.1 SHUTDOWN MARGIN B 3/4 10-1 1
3/4.10.2 GROUP HEIGHT, INSERTION AND POWER i
DISTRIBUTION LIMITS ~
. B 3/4 10-1 l
3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS-REACTOR i
CRITICALITY B 3/4 10-1 l
3/4.10.4 PHYSICS TESTS B 3/4 10-1 3/4.10.5 NO FLOW TESTS B 3/4 10-1 1
i 1
BEAVER VALLEY - UNIT 1 XVI Miendment No. 140, 139 j
.J
I e
\\
INDEX ADMINISTRATIVE CONTROLS a
SECTION PAGE 6.1 RESPONSIBILITY 6-1 6.2 OPOAMT?ATION 6.2.1 Corporate 6-1 l
6.2.2 Facility Staff 6-1 I
6.3 FACILITY STAFF QUALIFICATIONS 6-5
)
6.4 TRAINING 6-5
)
6.5 REVIEW AND AUDIT 6.5.1 ONSITE SAFETY COMMITTEE (OSC) 6.5.1.1 Function 6-5 6.5.1.2 Composition.
6-5 6.5.1.3 Alternates 6-6 6.5.1.4 Meeting Frequency.
6-6 6.5.1.5 Quorum 6-6 6.5.1.6 Responsibilities 6-6 6.5.1.7 Authority.
6-7 6.5.1.8 Records.
6-7 6.5.2 OFFSITE REVIEW COMMITTEE (ORC) 6.5.2.1 Function 6-7 6.5.2 2 Composition.
6-8 6.5.2.3 Alternates 6-8 6.5.2.4 Consultants 6-8 6.5.2.5 Meeting Frequency.
6-9 BEAVER VALLEY - UNIT XIX Amenchent No. 144, 139
INDEX ADMINISTRATIVE CONTROLS SIC _T_T_Q1:
PhGE 6.5.2.6 Quorum 6-9 I
6.5.2.7 Review 6-9 6.5.2.8 Audits 6-10 l
6.5.2.9 Authority.
6-11 l
6.5.2.10 Records 6-11 l
6,6 REPORTABLE EVENT ACTION 6-11 1
6,7 S AFETY LIMIT VIOLATION 6-12 6.F PROCEDURES 6-12 l
)
6.9 REPORTING REQUIREMENTS 6-13 6.9.1 Routine Reports 6-13a 6.9.1,2,3 Startup Reports.
6-13a 6.9.1.4,5 Annual Reports 6-14 6.9.1.6 Monthly Operating Report 6-15 6.9.1.10,11 Annual Radiological Environmental Report 6-15 i
6.9.1.12,13 Semi-Annual Radioactive Effluent i
Release Report 6-21 6.9.1.14 Radial Peaking Factor Limit Report 6-22 6.9.2 SPECIAL REPORTS 6-22 1220 RECORD RETENTION 6-23 f.11 RADIATIO" PPOTECTION PROGRAM 6-25 BEAVER VALLEY - UNIT 1 XX Amendment No. 140, 139
1
)
INDEX ADMINISTRATIVE CONTROLS 6.22 HIGH RADIATION AREA 6-25 l
6.24 PROCESS CONTROL PROGRAM (PCP) 6-27 6.15 OFFSITE DOSE CALCULATION MANUAL.
6-27 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-28 l
6.17 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM'.
6-31 l
l 1
i i
BEAVER VALLEY - UNIT 1 XXI Miendment No. 140, 139
TABLE'INDEX TABLE TITLE PAGE 2.2-1 Reactor Trip System Instrumentation Trip 2-6 l
Setpointe 3.1-1 Accide.it Analyses Requiring Re-evaluation 3/4 1-19A in thn event of an Inoperable Full or Part.,
Leng',n Rod 3.2-1 DN3 Parameters 3/4 2-13 3.3-1 Reactor Trip System Instrumentation 3/4 3-2 3.3-2 Reactor Trip System Instrumentation Response 3/4 3-9 Times 4.3-1 Reactor Trip System Instrumentation 3/4 3-11 Surveillance Requirements 3.3-3 Engineered Safety Features Actuation System 3/4 3-15 Instrumentation 3.3-4 Engineered Safety Features Actuation System 3/4 3-22 Instrumentation Trip Setpoints 3.3-5 Engineered Safety Feature Response Times 3/4 3-25 4.3-2 Engineered Safety Feature Actuation System 3/4 3-29 Instrumentation Surveillance Requirements 3.3-6 Radiation Monitoring Instrumentation 3/4 3-34 4.3-3 Radiation Monitoring Instrumentation 3/4 3-36 Surveillance Requirements 3.3-7 Seismic Monitoring Instrumentation 3/4 3-39 4.3-4 Seismic Monitoring Instrumentation 3/4 3-40 Surveillance Requirements 3.3-8 Meteorological Monitorinc Instrumentation 3/4 3-42 l
4.3-5 Meteorological Monitoring Instrumentation 3/4 3-43 Surveillance Requirements 3.3-9 Remote Shutdown Panel Monitoring 3/4 3-45 Instrumentation 4.3-6 Remote Shutdown Monitoring Instrumentation 3/4 3-46 Surveillance Requirements 3.3-10 Fire Detection Instruments 3/4 3-48 BEAVER VALLEY - UNIT 1 XXII kiendment No. TSO,139
______m-___
v j
Table Index (cont.)
)
TABLE TITLE PAGE 3.3-11 Accident Monitoring Instrumentation 3/4 3-51 4.3-7 Accident Monitoring Instrumentation 3/4 3-52 Surveillance Requirements 1
3.3-12 Radioactive Liquid Effluent Monitoring "
3/4 3-54 Instrumentation 4.3-12 Radioactive Liquid Effluent Monitoring 3/4 3-57 Instrumentation Surveillance Requirements 3.3-13 Radioactive Gaseous Effluent Monitoring 3/4 3-60 Instrumentation 4.3-13 Radioactive Gaseous Effluent Monitoring 3/4 3-65 Instrumentation Surveillance Requirements 4.4-1 Miriinum Number of Steam Generators to be 3/4 4-10c Inspected During Inservice Inspection 4.4-2 Steam Generator Tube Inspection 3/4 4-10d 4.4-3 Reactor Coolant System Pressure Isolation 3/4 4-14b l
Valves 3.4-1 Reactor Coolant System Chemistry Limits 3/4 4-16 4.4-10 Reactor Coolant System Chemistry Limits 3/4 4-17 l
Surveillance Requirements 4.4-12 Primary Coolant Specific Activity Sample 3/4 4-20 and Analysis Program 4.4-5 Reactor Vessel Material Irradiation 3/4 4-26 l
Surveillance Schedule 3.6-1 Containment Penetrations 3/4 6-19a 3.7-1 Maximum Allowable Power Range Neutron Flux 3/4 7-2 High Setpoint With Inoperable Steamline Safety Valves During 3 Loop Operation 3.7-2 Maximum Allowable Power Range Neutron Flux 3/4 7-3 High Setpoint with Inoperable Steamline Safety Valves During 2 Loop Operation 3.7-3 Steamline Safety Valves Per Loop 3/4 7-4 l
4.7-2 Secondary Coolant System Specific Activity 3/4 7-9 Sample and Analysis Program l
r BEAVER VALLEY - UNIT 1 XXIII Miendment No. Ykb 139
S'able Index.(cont. )
'VLE TITLE g
3.5-1 Battery Surveillance Requirements 3/4 8-9a 3.9-1 Beaver Valley Fuel Assembly Minimum Burnup 3/4 9-15 Initial U235 Enrichment For Storage in'~
vs.
Region 2 Spent Fuel Racks 4.11-1 Radioactive Liouid Waste Sampling and Analysis 3/4 11-3 Program 4.11-2 Radioactive Gaseous Waste Sampling and Analysis 3/4 11-13 Program l 3.12-1 Radiological Environmental Monitoring Program 3/4 12-3 3.12-2 Reporting Levels for Radioactivity Concentrations 3/4 12-6 in Environmental Samples 4.12-1 Maximum Values for the Lower Limits of Detection 3/4 12-7 6 3/4.4-1 Reactor Vessel Toughness Data (unitradiated)
B 3/4 4-7 6.2-1 Minimum Shift Crew Composition 6-4 6.9-1 Environmental Radiological Monitoring Program 6-20 Summary l
BEAVER VALLEY - UNIT 1 XXIV Amendment No. W,139 l
l l
Ficure Index l
t'?"FT TITLE PAGE 2.1-1 Reactor Core Safety Limit - Three Loops in 2-2 Operation i
2.1-2 Reactor Core Safety Limit - Two Loops in 2-3 Operation (One Loop Isolated) 2.1-3 Reactor Core Safety Limit - Two Loops in 2-4 i
Operation (No Isolated Loop) 3.1-1 Rod Group Insertion Limits Versus Thermal 3/4 1-24 Power - Three Loop Operation l
3.1-2 Rod Group Insertion Limits Versus Thermal 3/4 1-25 Power - Two Loop operation i
3.2-1 Axial Flux Difference Limits as a Function 3/4 2-4 of Rated Thermal Power i
3.2-2 K(z) - Normalized F (z) as a function of 3/4 2-7 q
Core Height I
a.4-1 Dose Equivalent 1-131 Primary Coolant 3/4 4-21.
Specific Activity Limit Versus Percent of Rated Thermal Power with the Primary Coolant j
specific Activity > 1.0, ci/ gram Dose Equivalent I-131 3.4-2 Beaver Valley Unit No. 1 Reactor Coolant 3/4 4-24 1
System Heatup Limitations Applicable for the First 9.5'EFPY l
3.4-3 Beaver Valley Unit No. 1 Reactor Coolant 3/4 4-25 System Cooldown Limitations Applicable for the First 9.5 ETPY l
3.6-1 Maximum Allowable Primary Containment Air 3/4 6-7 l
Pressure Versus River Water Temperature and RWST Water Temperature l
B 3/4.2-1 Typical Indicated Axial Flux Difference B 3/4 2-3 Versus Thermal Power at BOL.
I B 3/4.4-1 Fast Neutron Fluence (E>l Mev) as a Function B 3/4 4-6a I
of Full Power Service Life B 3/4.4-2 Effect of Fluence, Copper Content, and B 3/4 4-6b Phoschorus Content on ART for Reactor Vess'lSteelsPerReg.Gukke1.99 l
e BEAVER VALLEY - UNIT i XXV Amendment No. 740,139
6 Tigyre I'ndex (cont.)
EJGURE TITLI PAGE 5.1-1 Site Boundary for Gaseous and Liquid 5-lb i
Effluents for the Beaver Valley Power j
Station I
1 5.2-3 Exclusion Area - Beaver Valley Power Stat.icn 5-id f.1-4 Low Population Zone - Beaver Valley Power 5-le
)
Station 5.1-5 Gaseous Release Points - Beaver Valley Power
.5-2 l
Station I
5.1-6 Liquid Release Points - Beaver Valley Power 5-3
)
Station i
6.2-1 Corporate Organization (Partial) 6-2 6.2-2 racility organization 6-3 j
i l
l BEAVER VALLEY - UNIT 1 XXVI
/Wendment No.'144,139 l
E
REAyTOP COOLANT SYSTEM STEAM GENERATOPS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2,
3 and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T
above 200*F.
avg SURVEILLANCE REQUIREMENTS 4.4.5.1 Steam Generator Samole Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and jnspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Stear Generator Tube Samole Gelection and Insnection - The stean generator tube minimum cample
- size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
Steam generator tubes shall be examined in accordance with Article 8 of Section V
(" Eddy Current Examination of Tubular Products") and Appendix IV to Section XI
(" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing")
of the applicable year and addenda of the ASME Boiler and Pressure Vessel Code required by 10CFR50, Section 50.55a(g).
The tubes selected for each inservicc inspc: tion shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50%
of the tubes inspected shall be from these critical areas.
b.
The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
2.
All nonplugged tubes that previously had detectable wall penetrations (>20%), and 2.
Tubes in those areas where experience has indicated potential problems.
BEAVER VALLEY - UNIT 1 3/4 4-8 Amendment No. 'R,139
V RADIOACTIVE EFFLUENTS
'3/4.11.3 SOLID RADIOACTIVE WASTE L3!:3T:N3 CCNDITION FOR OPERATION 3.11.3.1 The solid radwaste system shall be used, as applicable, to solidify and package radioactive wastes, and to -ensure meeting the requirements of 10 CFR Part 20, 10 CFR Part, 61~
and of 10 CFR Part 71.
Methods utilized to meet these requirements chall be described in facility procedures and in the Process Control Program (PCP).
APPLICABILITY At all times.
ACTION:
a.
With the applicable requirements of 10 CFR Part 20, 10 CFR Part 61 and 10 CFR Part 71 not satisfied, suspend affected shipments of solid radioactive wastes from the site.
b.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3.1.1 Prior to shipment, solidification shall be verified in accordance with Station Operating Procedures.
4.11.3.1.2 Reports.
The semi-annual Radioactive Effluent Release Report in Specification 6.9.1.12 shall include the following information for each type of solid waste shipped offsite during the report period:
a.
container volume; b.
total curie quantity (determined by measurement or estim tc);
c.
principal radionuclides (determined by measurement or estimate):
d.
type of waste (e.g.,
spent
- resin, compacted dry waste evaporator bottoms);
e.
type of container (e.g.,
- LSA, Type A,
Type B,
Large Quantity);
f.
solidification agent (e.g., cement, urea formaldehyde); and g.
classification and other requirements specified by 10 CFR Part 61.
BEAVER VALLEY - UNIT 1 3/4 11-22 Amendment No. M, M,139
ApfD,NIST'RATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the U.
S.
Nuclear Regulatory Commis,sion, Document Control Desk.
STARTUP REPORTS 6.9.1.1 A
summary report of plant startup and power escalation testing will be submitted following (1) receipt of an operating
- license, (2) amendment to the license involving a planned increase in power
- level, (3) installation of fuel that has a different design or has been manufactured by a
different fuel
- supplier, and (4) modifications that may have significantly altered the
- nuclear, thermal, or hydraulic performance of'the plant.
l l
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a
description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were regaired to obtain satisfactory operation shall also be described.
Any additional specific details requested in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days l
following completion of the startup test
- program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is
)
earliest.
If the Startup Report does not cover all three events (i.e.,
initial criticality, completion of startup test program, and resumption or commencement of commercial power operation),
supplementary reports sha13 be submitted at least every three months until all three events have been completed.
l 1
l BEAVER VALLEY - UNIT 1 6-13a
/Wendment No. 794, 139 l
ADMI.NISThtATIVE CONTROLS t__
MCM7M1Y OTERATING REPORT j
l 6.9.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a montl.ly basis no later than the f
15th of each month following the calendar month covered by the report.
6.9.1.7 DELETED by Amendment No. 84 6.9.1.8 DELETED by Amendment No. 84
)
6.9.1.9 DELETES by Amendment No. 84 ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT 3 6.9.1.10 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year j
uhall be submitted prior to May 1
of each year and will include reporting any deviations not reported under 6.9.2 with respect to the
(
Radiological Effluent Technical Specifications.
6.9.1.11 The annual radiological environmental reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report
- period, including a
comparison with preoperational studies, operational controls (as appropriate),
and previous environmental surveillance
- reports, and an assessment of the observed impacts of the plant operation on the environment.
The reports shall also include the results of the land use censuses required by Specification 3.12.2.
If harr.ful effects or evidence of irreversible l
damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.
The annual radiological environmental operating reports shall include summarized and tabu 3ated results in the format of Table 6.9-1 of all radiological environmental samples taken during the report period.
In the event that some results are not available for inclusion with the
- report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted i
os soon as possible in a supplementary report.
3 A
sinole submitta3 may be made for a multiple unit site.
The submittal should combine these sections that are common to both units.
BEAVER VALLEY - UNIT 1 6-15 Amendment No. DR,1Ms (next page is 6-20) 714, 139
s-j j
e
'ADMI.NISTRATIVE CONTROLS i
The radioactive effluent release report to be submitted 60 days after January 1
of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR
- 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable-method for calculating the-dose contribution from liquid and gaseous affluents are given in Regulatory Guide 1.109, Revision 1.
The SKYSHINE code (availabic from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.
1 The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Ragulatory Guide 1.21.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be performed in accordance with the ODCM.
The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month I
period.
l BADIAL PEAFING FACTOR LIM 17 REPORT 6.9.1.14 The F
limit for Rated Thermal Power (FRTP) shall be i
provided for all cbre planes containing bank "D" control rods and all
(
unrodded core planes at least 60 days prior to cycle initial I
criticality.
In the event that the limit would.be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.
RTP Any information needed to support F will be by request from the NRC and need not be included in this report.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, within the time period cpacified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
I BEAVER VALLEY - UNIT 1 6-22 Amendment No,'%, h% 139
4 1
POWE,P DISTRIBUTION LIMITS BASES Fuel rod bowing reduces the value of the DNB ratio.
Credit is available to offset this reduction in the generic margin.
The generic design
- margins, totaling 9.1% DNBR, und' completely offsets any rod bow penalties
(< 1.3% for the worst case-which occurs at a burnup of 24,000 MWD /MTU).
This margin includes the following:
1.
Design Limit DNBR of 1.30 vs. 1.28 2.
Grid Spacing (K ) of 0.046 vs. 0.059 s
3.
Thermal Diffusion Coefficient of 0.038 vs. 0.059
)
4.
DNBR Multiplier of 0.865 vs. 0.88 5.
Pitch reduction The radial peaking factor F
(Z) is measured periodically to provide assurance that the net channel
- factor, Fo forRatedTherm81 Power (F{y(Z), remg gs within its limit.
The F
limit
)
I as provided in the RadEa1 Peaking Factor Limit Report per l
x l
specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.
l Sg4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio, limit assures that the radial power l
distribution satisfies the design values used in the power capability l
analysis.
Radial power d' v "sution measurements are made during startup testing and periodical 2 uuring power operation.
)
2
}
The limit of 1.02 at which corrective action is required provides DNB l
and linear heat generation rate protection with x-y plane power
{
- tilts, i
The two-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the ovant such action does not correct the
- tilt, the margin for uncertainty on F
is reinstated by reducing the maximum allowed g
power by 3 percent for each percent of tilt in excess of 1.0.
BEAVER VALLEY - UNIT 1 B 3/4 2-5 Amendment No. 31,139 1
l
>* aucugA **
UNITED STATES p.v.? ( i NUCLEAR REGULATORY COMMISSION c
E WASHINGTON, D. C. 20555
'Y
+..,+
DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY
(~
THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendesnt No. 14 License No. NPF-73 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, et al.
(the licensee) dated January 5, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The isruance of this amendment is in accordance with 10 CFR Part 51 of the Conmission's regulations and all applicable requirements have been satisfied.
l 8
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications conteined in Appendix A, as revised through Amendment No. 14, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. DLCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective on issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A
John F. Stolz, Director Project Directorate I a Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 11, 1989
4 e
4 ATTACHMENT TO LICENSE AMENDMENT NO. 14 FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the following pages of the Appendix A (Technical Specifications) with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
l i
Remove Insert I
3/4 4-11 3/4 4-11 6-13 6-13 6-15 6-15
{
l 6-18 6-18 I
B 3/4 2-4 B 3/4 2-4 i
i J
l 4
4 4
REANORCOOLANTSYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200 F.
SURVEILLANCE RE0VIREMENTS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The in-service inspection of steam generator tubes shall be performed at the frequen-j cies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
Steam generator tubes shall be examined in accordance with Article 8 of Section V (" Eddy Current Examination of Tubular Products") and Appendix IV to Section XI (" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing") of the applicable year and addenda of the ASME Boiler and Pressure q
i l
Vessel Code required by 10 CFR 50, Section 50.55a(g).
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except; Whero experience in similar plants with similar water chemistry indicates a.
critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; b.
The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1)
All nonplugged tubes that previously had detectable wall l
penetrations (> than 20%),
2)
Tubes in those areas where experience has indicated potential problems.
BEAVER VALLEY - UNIT 2 3/4 4-11 Amendment No. 14
,- ~
MMINI[TRATIVECONTROLS PROCEDURE (Continued) 1 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
{
The intent of the original procedure is not altered.
a.
b.
The change is approved by two (2) members of the plant _ management staff, at least one (1) of whom holds a Senior Reactor Operator's License on the unit affected.
The change is documented, reviewed by the OSC and approved by the c.
Plant Manager, predesignated alternate or a predesignated Manager to whom the Plant Manager has assigned in writing the responsibility for review and approval of specific subjects, within 14 days of I
l implementation.
6.8,4 A Post-Accident monitoring program shall be established, implemented, and maintained, The program will provide the capability to obtain and analyze reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and containment atmosphere samples following an accident.
The pro-gram shall include the following:
l 1
(i) Training of personnel, I
(ii) Procedures for sampling and analysis, and (iii) Provisions for maintenance of sampling and analysis equipment.
6.8.5 A program for monitoring of secondary water chemistry to inhibit steam 1
generator tube degradation shall be implemented.
This program shall be f
described in the station chemistry manual and shall include:
Identification of a sampling schedule for the critical parameters and
{
a.
control points for these parameters; b.
Identification of the procedures used to measure the values of the critical parameters; Identification for process sampling points; c.
i d.
Procedures for the recording and management of data; Procedures defining corrective actions for off control point chemistry e.
j conditions; and
{
f.
A procedure identifying:
1) the authority responsible for the interpretation of the data, and 2) the sequence and timing of administrative events required to initiate corrective action.
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS l
6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Centrol Desk.
l BEAVER VALLEY - UNIT 2 6-13 Amendment No. %, 14
ADMINI TR TIVE CONTROLS ANNUAL REPORTS (Continued) b.
Documentation of all challenges to the pressurizer power operated relief valves (PORVS) or pressurizer safety valves.
c.
The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8.
The following information shall be included:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit.
Each result should include date and time of sampling and the radiciodine con-centrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131-concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
l MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience shall" be submitted on a monthly basis no later than the 15th of each month following l
i l
the calendar month covered by the report.
- 6. 9.1. 7 This item intentionally blank i
6.9.1.8 This item intentionally blank j
l
- 6. 9.1. 9 This item intentionally blank ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT 3 6.9.1.10 Routine radiological environmental operating reports covering the l
operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year and will include reporting any deviations not reported under 6.9.2 with respect to the Radiological Effluent Technical Specifications.
6.9.1.11 The annual radiological environmental reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological J
environmental surveillance activities for the report period, including a com-parison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports, and an assessment of the observed i
impacts of the plant operation on the environment.
The reports shall also include the results of the land use censuses required by Specification 3.12.2.
If harmful effects or evidence of irreversible damage are detected by the 8A single submittal may be made for a multiple unit site.
The submittal should combine those sections that are common to both units.
BEAVER VALLEY - UNIT 2 6-15 Amendment No. 14 i
4
@WINISTR TIVE CONTROLS SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses,to the likely most exposed real incividual from reactor releases for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1.
The SKYSHINE Code (available from Radiation Shielding Information Center, (ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.
The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21.
In-addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be per-l formed in accordance with ODCM.
1 The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month period.
RADIAL PEAKING FACTOR LIMIT REPORT RTP 6.9.1.14 The F limit for Rated Thermal Power (F shall be provided for all core planes iontaining bank "D" control rods a E a)ll unrodded core planes at least 60 days prior to cycle initial criticality.
In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless i
l otherwise exempted by the Commission.
4 RTP Any information needed to support F will be by request from the NRC and x
need not be included in this report.
SPECIAL REPORTS.
i 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk within the time period specified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
a.
ECCS Actuation, Specifications 3.5.2 and 3.5.3.
b.
Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
Inoperable Meteorological Monitoring Instrumentation, c.
Specification 3.3.3.4.
l l
l 1
BEAVER VALLEY - UNIT 2 6-16 Amendment No. 14
m -
u
.d gs l
POWER' DISTRIBUTION LIMITS MSES
)
3/4.2.2 and 3/4 2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS F (Z) n AND Fgg (Continued)
~
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 f
c.
are maintained.
1 d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFtRENCE is mair.tained within the limits.
j The relaxation in F as a function of THERMAL POWER allows changes in H
the radial power shape for all permissible rod insertion limits.
F will be l
AH maintained within its limits provided conditions a thru d above, are maintained.
l h
When an F measurement is taken, both experimental error and manufacturing q
tolerance must be allowed for.
5% is the appropriate experimental error allow-l ance for a full core map taken with the incore detector flux mapping system and j
3% is the appropriate allowance for manufacturing tolerance.
1 The specified limit of F contains an 8% allowance for uncertainties which H
l means that normal, full power, three loop operation will result in F H < 1.55/1.08.
)
)
Fuel rod bowing reduces the value of the DNB ratio.
Credit is available to offset this reduction in the generic margin.
The generic design margins, totaling 9.1% DNBR, and completely offsets any rod bow penalties (< 1.3% for the l worst case which occurs at a burnup of 24,000 MWD /MTU).
l
{
This margin includes the following:
1.
Design Limit DNBR of 1.30 vs. 1.28 2.
Grid Spacing (K ) f 0.046 vs. 0.059 s
3.
Thermal Diffusion Coefficient of 0.038 vs. 0.059 4.
DNBR Multiplier of 0.865 vs. 0.88 5.
Pitch reduction j
The radial peaking factor Fxy (Z) is measured periodically to provide assurance that the hot channel factor, Fq (2), remains within its lir.it. The F
limit for Rated Thermal Power (FRTP) as provided in the Radial Peaking xy x
' Factor Limit Report per Specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.
3/4.2.4 QUADRANT POWER TILT RATIO The Quadrant Power Tilt Ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
BEAVER VALLEY - UNIT 2 B 3/4 2-4 Amendment No. 14
_ _ -____ ______