ML20116L568
| ML20116L568 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 10/29/1992 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116L570 | List: |
| References | |
| NUDOCS 9211190009 | |
| Download: ML20116L568 (18) | |
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/ UNITED STATES l' NUCLEAR REGULATORY COMMISSION t n { I W AsMNoT ON, D. C. 20$!.5 %e....$ COMMONWEAliN EplS901{_(OMPAt4Y QQ(_Kl] fl0,_50 237 QRLSDDI NiltMAR POWER STAT _lRL1 UN'T_Z Mi[@ MENT TO FACILITY OPERATINQ LICENSE Amendment No. 120 License No. DPR-19 1. The fluclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Commonwealth Edison Company (the licensee) dated August 9, 1991, 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 (i) that the activities authorizaJ by this amendment can be conducted without endangering the health and safety of the public, and (ii) thiit 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 thc health and safety of the public; and E. The issuance of thh amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is arnended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. OPR-19 is hereby amended to read as follows: 9211190009 921029 PDR ADOCK 05000237 p PDh
. (2) lechnical Specificatior11 The Technical Specifications contained in Appendix A, as revised through Amendment flo. 120. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance, to be implemented within 30 days. FOR THE NUCLEAR REGULATO MMISSION / Rict r arrett, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: October 29, 1992
s ATTACHMENT TO LICENSE AMENDMENT NO. 120 TACillTV OPERATING LICENSE _!io. OPR-19 DOCKET NO 50-237 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are c identified by the captioned amendment number and contain marginal lines indicating _the area of change. RuiQVI It!1LR_1 iv iv vi vi vii vil viii viil-3/4.10-6 3/4.10-6 3/4.10-7 3/4.10-7 3/4.11-1 3/4.11 3/4.11-3 B 3/4.11-4 I {
ORESDEN II OPR-19 ~ Amendment No. 120 (Table of Contents, Cont'd.) P,agg 3.9.C Diesel Fuel 3/4.9 - 5 3.9.0 Diesel Generator Operability 3/4.9 - 5 Limiting Conditions for Operation Bases (3.9) B 3/4.9 - 7 Surveillance Requirement Bases (4.9) B 3/4.9 - 8 3.10 Refueling 3/4.10- 1 3.10.A Refueling Interlocks 3/4.10- 1 3.10.B Core Monitoring 3/4.10- 1 3.10.C Fuel Storage Pool Water Level 3/4.10- 2 3.10.0 Control Rod and Control Rod Drive Maintenance 3/4.10- 3 3.10.E Extended Core Maintenance 3/4.10- 4 3.10.F Spent Fuel Cask Handling 3/4.10- 5 Limiting Conditions for Operation Bases (3.10) B 3/4.10- 8 Surveillance Requirement Bases (4.10) B 3/4.10-11 3.11 High Energy Piping Integrity - Deleted Per Amendment 120 3.12 Fire Protection Systems - Sections 3.12.A through 3.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 106) 4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactor Protection System 3/4.1 - 1 4.2 Protective Instrumentation 3/4. 2 - 1 4.2.A Primary Containment Isolation Functions 3/4.2 - 1
- 4. 2. B Core and Containment Cooling Systems --
Initiation and Control 3/4.2 - 1 4.2.0 Control Rod Block Actuation 3/4.2 - 2 4.2.0 Refueling Floor Radiation Monitors 3/4.2 - 2 4.2.E as,t Accident Instrumentation _3/4.2 - 3 4.2.F Hadioactive Liquid Effluent Instrumentation 3/4.2 - 4 4.2.G Radioactive Gaseous Effluent Instrumentation 3/4.2 - 5 - 4. 3 Reactivity Control 3/4.3-- 1 4.3.A Reactivity Limitations 3/4.3.- 1 4.3.B Control Rods 3/4.3 - 4 4.3.C Scram Insertion Times 3/4 3 > iv-
ORESDEN II DPR-15 Amendment No.120 (Table of Con.ents, Cont'd.) Pag 4.9 Auxiliary Electrical Systems 3/4.9 - 1 4.9.A Station Batteries 3/4.9 - 1 4.9.B (N/A) 4.9.C Diesel fuel 3/4.9 - Sa 4.9.D Diesel Generator Operability 3/4.9 - Sa 4.10 Refueling 3/4.10- 1 4.10.A Refueling Interlocks 3/4.10- 1 4.10.8 Core Monitoring 3/4.10- 1 e 4.10.C Fuel Storage Pool Water Level 3/4.10- 2 4.10.0 Control Rod Drive and Control Rod Drive Maintenance 3/4.10- 3 4.10.E Extended Core Maintenance 3/4.10- 4 4.10.F Spent Fuel Cask Handling 3/4.10- 5 4.10.G Fuel Storage Reactivity Limit 3/4.10- 8 4.11 High Energy Piping Integrity - Deleted per Amendment 120 4.12 Fire Protection Systems - Sections 4.12.A through 4.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 106) 5.0 Design Features 5-1 5.1 Site 5-1 5.2 Reactor 5-1 5.3 Reactor Vessel 5-1 5.4 Containment 5- .1
- 5. 5 Fuel Storage 5-1 5.6 Seismic Design 5-2 6.0 Administrative Controls 6-1 6.1 Organization, Review, Investigation and Audit 6-1 6.2 Procedures and Programs 6-13 6.3 Action to be Taken in the Event of a REPORTABLE EVENT in Plant Operation 6-15 6.4 Action to be taken in the Event a Safety Limit is Exceeded 6-15
- 6. 5 Plant Operating Records 6-15 6.6 Reporting Requirements 6 -' 17 6.7 Environmental Qualification 6-22 6.8 Offsite Dose Calculation Manual (ODCM) 6-24-6.9 Process Control Program (PCP) 6-25 6.10 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, Solid) 6 - 25.
6.11 Radiation Protection Program 6-26 6.12 High Radiation Area-6- 26 vi
0RESDEN II DPR-19 Amendment No. 120 List of Tables fag ) Table 3.1.1 Reactor Protectit' System (Scram) 3/4.1-5 Instrumentation Requirements Table 4.1.1 Scram Instrumentation Functional Tests 3/4.1-8 Table 4.1.2 Scram Instrumentation Calibration 3/4.1-10 Table 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 3/4.2-8 Table 3.2.2 Instrumentation that Initiates or Controls the Core and Containment Cooling System 3/4.2-10 Table 3.2.3 Instrumentation that Initiates Rod Block' 3/4.2 Table 3.2.4 Radioactive Liquid Effluent Monitoring Instrumentation 3/4.2. Table 3.2.5 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4.2-15 Table 3.2.6 Post Accident Monitoring Instrumentation Requirements 3/4.2-17 Table 3.2.7 Instrumentation That Initiates Recirculation Pump Trip 3/4.2-18a Table 4.2.1 Minimum Test and. Calibration Frequency for Core and Containment. Cooling Systems- . 4 Instrumentation,-Rod Blocks, and Isolations 3/4.2-19 Table 4.2.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2-22 Table 4.2.3 Radioactive Gaseous Effluent Monitoring- ' I Instrumentation Surveillance Requirements . 3/4.2-24 Table 4.2.4 Post Accident Monitoring Instrumentation Surveillance Requirements 3/4.2-26 Table 4.2.5 Minimum Test and Calibration Frequency for the Recirculation Pump Trip 3/4.2-27a Table 4.5.1 Surveillance of the HPCI Subsystem 3/4.5-~7a Table 4.6.2 Neutron Flux and Sample Withdrawal 8'3/4.6-26: Table 3.7.1-Primary Containment. Isolation 3/4.7-31: Table ~4.8.1 Radioactive Gaseous Waste Sampling and Analysis Program 3/4.8-22 Table 4.8.2. Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released From1the Site to Unrestricted Areas in Liquid Waste-3/4.8-24 Table 4.8.3 Radioactive Ltquid Waste Sampling and Analysis Program 3/4.8-25 Table 4.8.4 Radiological-Environmen_tal Monitoring Program-3/4.8-27 Table 4.8.5 Reporting LevelsLfor Radioactivity Concentrations-in-Envirormental-Samples 1 3/4.8-28 Table:4.8.6 Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program. 3/4.8-29 vii -9 t ,y 9 -i e g-9 r= g+ y--,,---,-*-,,y yy--g, -y ygyw-c -r-wg-yspww - vi-tg -ge%vw-w*ss -+ -ty - --w = de m ew-e wr* are '-e 7 'e+"n+"v"*N=**'w-*-be "-**d
ORESDEN II OPR-19 Amendment No. 120 List of Tables (continued) } 5 Table 3.12-1 Deleted Table 3.12-2 Deleted Table 3.12-3 Deleted Table 3.12-4 Deleted Table 6.1.1 Minimum Shift Manning Chart 6-4 Table 6.6.1 Special Reports 6-23 List of Figures Figure 2.1-3 APRM Bias Scram Relationship to Normal Operating Conditions B 1/2.1-17 F1gure 4.1.1 Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.1-18 Figure 4.2.2 Test Interval vs System Unavailability B 3/4.2-38 Figure 3.4.1 Deleted 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.6.1 Minimum Reactor Vessel Metal Temperature. 3/4.6-23 Figure 3.6.2 Thermal Power vs Core Flow limits for Thermal Hydraulic Stability Surveillance in Single Loop Operating 3/4.6-24 Figure 4.6.1 Minimum Reactor Pressurization Temperature B 3/4.6-29 Figure 4.6.2 Chloride Stress Corrosion Test-Results at 500'F B 3/4.6-31 Figure 4.8.1 Owner Contrclied/ Unrestricted Area Boundary B 3/4.8-38 Figure 4.8.2 Detail of Central Complex B 3/4.8-39 Figure 6.1-1 Offsite Organization - Deleted Figure 6.1-2 Station Organization - Deleted viii
i p :,. j DRESDEN II DPR-19 Amendment No. 120' t 3.10 LIMITING CONDITIONS FOR OPERATION 4.10 SURVEILLANCE REQUIREMENTS (Cont'd.) (Cont'd.) t a. Twelve (12) randomly distributed broken wires in one lay-or four-(4) broken wires in one strand of one rope-lay. t b. Wear of one-third the e original diameter of outside individual-i wire. j c. Kinking, crushing, or any other damage resulting in dis' r-tion of the ropt d. Evidence of any. type of heat damage, e. Reductions from nominal diameter of-more than' '1/16. inch for a rope diameter from 7/8" to-1 1/4" inclusive. ~ 2. Fuel cask handling in 2. Prior to operation other than the in the RESTRICTED MODE' RESTRICTED MODE will be permitted in a.. the controlled area emergency-or equipment limit switches will failure situatOns lbe tested: only to the extent. necessary to get-the b. the "two-block" cask to the closest .lirwit-switches will acceptable stable be: tested; location.. 3/4.10-6 . m..
- x.,
^ DRESDEN II DPR Amendment No.120 3.10 LIMITING CONDITIONS FOR OPERATION 4.10 SURVEILLANCE REQUIREMENTS (Cont'd.) (Cont'd.) c. the " inching hoist" controls will be
- tested, w
3. Operation with a failed 3. The empty spent fuel controlled area limit cask will be' lifted switch is permissible free of all support by for 48 hours providing a maximum of 1 foot and an operator is on the left hanging for_5. refueling floor to assure minutes prior to any the crane is operated series of fuel cask within the restricted handling operations. zone painted on the floor. 3/4.10-7
'o UNITED STATES g l' NUCLEAR REGULATORY COMMISSION o -{ ,i WASHING TON, D. C. 20505 . \\..../ COMMONWEALTH EDISr*C QMPANY DOCKET NO. 4 DRESDEN NUCLEAR POWER SufIOLJ) NIT 3 AMENDMENT TO FACIllTY OPERATING LICENSE Amendment No. 116 License No. DPR-25 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Commonwealth Edison Company (the licensee) dated August 9, 1991, 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 (i) that the activities authorized by this amendment can be conducted without endangering the iiealth 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 issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.8. of Facility Operating License No. DPR-25 is hereby amended to read as follows:
9 B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.116, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is ef fective as of the date of its issuance, to be implemented within 30 days. FOR THE NUCLEAR REGULATORY : ISS10N u/ g Richa d J arrett, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 29, 1992
ATTACHMENT TO LJCENSE AMEN 0 MENT-NO 116 FACILITY OPERATING LICENSE NO. DPR-25 7 DOCKET NO. 50-249 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain-marginal lines indicating the area of change. RlriOVE INSERT iv iv vi vi vii-vii viii viii 3/4.10-6 3/4.10 3/4.10-7 3/4.10-7 3/4.11-1 3/4.11-2 3/4.11-3 B 3/4.11-4 wie a in..iiin d
DRESDEN III DPR-25 Amendment No.116 (Table of Contents, Cont'd.) P,agg 3.9.C Diesel Fuel 3/4.9-5 3.9.0 Diesel Generator Operability 3/4.9-5 Limiting Conditions for Operation Bases (3.9) B 3/4.9-7 Surveillance Requirement Bases (4.9) B 3/4.9-8 3.10 Refueling 3/4.10-1 3.10.A Refueling Interlocks 3/4.10-1 3.10.8 Core Monitoring 3/4.10-1 3.10.C Fuel Storage Pool Water Level 3/4.10-2 3.10.D Control Rod and Control Rod Drive Maintanance 3/4.10-3 3.10 E Extended Core Maintenance 3/4.10-4 3.10.F Cpent Fuel Cask Handling 3/4.10-5 Limiting Conditions for Opera: sn Bases (3.10) B 3/4.10-8 Surveillance Requirement Bases (4.10) B 3/4.10-11 3.11 High Energy Piping Integrity,,, Deleted Per Amendment 116 3.12 Fire Protection Systems - Sections 3.12.A through 3.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 101) 4.0 SURVEILLANCE RE0VIREMENTS 4.1 Reactor Protection System 3/4.1-1 4.2 Protective Instrumentation 3/4.2-1 4.2.A Primary Containment Isolation Functions 3/4.2-1 4.2.B Core and Containment Cooling Systems -- Initiation and Control 3/4.2-1 4.2.C Control Rod Block Actuation 3/4.2-2 4.2.0 Refueling Floor Radiation Monitors 3/4.2-2 4.2.E Post Accident Instrumentation 3/4.2-3 4.2.F Radioactive Liquid Effluent Instrumentation 3/4.2-4 4.2.G Radioactive Gaseous Effluent Instrumentation 3/4.2-5 4.3 Reactivity Control 3/4.3-1 4.3.A Reactivity Limitations 3/4.3-1 4.3.B Control Rods 3/4.3-4 4.3.C Scram Insertion Times 3/4.3-10 iv 4 l
-DRESDEN III- 'DPR-#G Amendment No. 116 (Table of Contents, Cont'd.) P,,agg 4.9 Auxiliary Electrical Systems 3/4.9-1 4.9.A Station Batteries 3/4.9 4.9.B (N/A) 4.9.C Diesel Fuel-3/4.9-5a 4.9.D Diesel Generator Operability 3/4.9-5a 4.10. Refueling 3/4.10-1 4.10.A Refueling Interlocks 3/4.10-1 4.10.B Core Monitoring _ - 3/4.-10 4.10.C Fuel Storage Pool Water Level 3/4.10-2: 4.10.0 Control Rod Drive and Control Rod Drive Maintenance-3/4.10-3 4.10.E Extended Core Maintenance 3/4.10-4-- 4.10 F Spent Fuel Cask Handling 3/4.10 4.10.G Fuel Storage Reactivity Limit -3/4.10 4.11 High Energy Piping-Integrity - Deleted per Amendment 116-4.'12 Fire Protection Systems:- Sections 4.12.A'through 4.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 101)~ 5.0 Design Features -S-1 5.1 Site 5-1
- 5. 2 Reactor 5-1.
5.3 Reactor Vessel 5-1 ~ . 5. 4 Containment 5 5.5 Fuel Storage 5-1 5.6 _ Seismic Design - 5 6.0 Administrative Controls-6-1 6.1 Organization,_ Review, Investigation and Audit 6 1-6.2 Procedures and Programs 6 13~
- 6. 3 -
Action to be Taken-in the Event of a REPORTABLE EVENT in Plant Operation . 6-15 ' 6.4 Action to be taken:in the: Event a-Safety-Limit:is Exceeded 6-15,- 6.5 Plant Operating Records 6,
- 6.6 Reporting' Requirements 6-17 6.7' Environmental-Qualification
'6-21' 6.8 Offsite Dose Calculation Manual (ODCM): 23' 6.9 Process Control Program (PCP)
- 6 6.10 Major Changes.to' Radioactive Waste Treatment
. Systems (Liquid, Gaseous, Solid). .6 6.11-Radiation Protection Program-6, 6.12-High Radiation Area 25-l- i i l vi
DRESDEN III DPR-25 List of Tables Amendment No. 116 Table 3.1.1 Reactor Protection System (Scram) 3/4.1-5 Instrumentation Requirements Table 4.1.1 Scram Instrumentation Functional Tests 3/4.1-8 Table 4.1.2 Scram Instrumentation Calibration 3/4.1-10 Table 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 3/4.2-8 Table 3.2.2 Instrumentation that Initiates or Controls the Coro and Containment Cooling System 3/4.2-10 Table 3.2.3 Instrumentation that Initiates Rod Block 3/4.2-12 Table 3.2.4 Radioactive Liquid Effluent Monitoring Instrumentation 3/4.2-14 Table 3.2.5 Radioactive Gaseous Effluent 3/4.2-15 Monitoring Instrumentation Table 3.2.6 Post Accident Monitoring Instrumentation Requirements 3/4.2-17 Table 3.2.7 Instrumentation That Initiates Recircula-tion Pump Trip 3/4.2-18a Table 4.2.1 Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations 3/4.2-19 Table 4.2.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2-22 Table 4.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4.2-24 Table 4.2.4 Post Accident Monitoring Instrumentation Surveillance Requirements 3/4.2-26 Table 4.2.5 Minimum Test and Calibration Frequency for Recirculation Pump Trip 3/4.2-27a Table 4.5.1 Surveillance of HPCI Subsystem 3/4.5-7a Table 4.6.2 Neutron Flux and Sample Withdrawal B 3/4.6-30 Table 3.7.1 Primary Containment Isolation 3/4.7-31 Table 4.8.1-Radioactive Gaseous Waste Sampling and Analysis Program 3/4.8-22 Table 4.8.2 Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released from the Site to Unrestricted Areas in Liquid Waste 3/4.8-24 Table 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 3/4.8 fable 4.8.4 Radioactive Environmental Monitoring Program 3/4.8-27 Table 4.8.5 Reporting Levels for Radioactivity Concentrations in Environmental Samples 3/4.8-28 Table 4.8.6 Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program 3/4.8-29 Table 4.11-1 Deleted. Table 3.12-1 Deleted.- vii
DRESDEN III-DPR-25 Amendment No. 116 List of Tables (continued) Page Table 3.12-2 Deleted Table 3.12-3 Deleted Table 3.12-4 Deleted Table 6.1.1 Minimum Shift Manning Chart 6-4 Table 6.6.1 Special Reports 6-22 List of Figures Figure 2.1-3 APRM Bias Scram Relationship to Normal Operating Conditions B 1/2.1-17 Figure 4.1.1 Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.1-18 Figure 4.2.2 Test Interval vs System Unavailability B 3/4.2-38 Figure 3.4.1 Deleted 3/4.4-4 Figure 3.4.2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 Figure 3.6.1 Minimum Reactor Vessel Metal Temperature 3/4.6-23 Figure 3.6.2 Thermal Power vs Core Flow Limits for Thermal Hydraulic Stability Surveillance in Single Loop Operating -3/4.6-24 Figure 4.6.1 Minimum Reactor Pressurization Temperature .B 3/4.6-29: Figure 4.6.2 Chloride Stress Corrosion Test Results at 500*F B 3/4.6-31 Figure 4.8.1 Owner Controlled / Unrestricted Area Boundary B 3/4.8-38 Figure 4.8.2 Detail of Central Complex B 3/4.8-39 Figure 6.1-1 Offsite Organization - Deleted Figure 6.1-2 Station Organization - Deleted viii
DRESDEN III DPR-25 Amendment Ho. 116 3.10 LIMITING CONDITIONS FOR OPERATION 4.10 SURVEILLANCE REQUIREMENTS { Cont'd.) (Cont'd.) ~~ a. Twelve (12) randomly distributed broken wires in one lay or four(4) broken wires in one strand of one rope lay. b. Wear of one-third. the original diameter of outside individual wire. c. Kinking, crushing, or..any other damage resulting in distortion of the rope. d. Evidence of any type of heat damage.- e. Reductions from-nominal diameter of more than 1/16 inch for a rop'e diameter from 7/8 to 1 1/4"- inclusive. 2. Fuel cask handling in 2. ' Prior to other than the-operation in the RESTRICTED MODE will RESTRIC'IED MODE be permitted in emergency or equipment a. the controlled area failure situations limit switches will-only to the extent be tested: necessary to get the cask to the closest b. the "two-block" acceptable stable limit switches will-
- location, be tested; 3/4.10-6
4 DRESDEN III DPR-25 Amendment No.116 3.10 LIMITING CONDITIONS FOR OPERATION-4.10 SURVEILLANCE REQUIREMENTS. (Cont'd.) (Cont'd.) c. the " inching hoist" controls will be tested. 3. Operation with a 3. The empty spent fuel failed controlled area cask will be lifted limit switch is free of all support by permissible for 48 a maximum of 1 foot and hourc providing an left hanging for 5 operator is on the minutes prior to any refueling floor to series of-fuel cask assure the crane is handling operations. operated within the restricted zone painted on the floor. 3/4.10-7 i}}