ML20245F993
| ML20245F993 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 06/23/1989 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| MFN-044-89, MFN-44-89, PWM8997, NUDOCS 8906280317 | |
| Download: ML20245F993 (300) | |
Text
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- g. 405 MFN 044-89 PWM8997 June 23,1989 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L. Miller, Director Standardization and Non-Power Reactor Project Directorate
Subject:
Submittal of Proposed Technical Specifications for ABWR (SSAR Chapter 16)
Dear Mr. Miller:
Enclosed are thirty four (34) copies of the proposed Technical Specifications, including the corresponding &ses document, for the Advanced Boiling Water Reactor (ABWR). This submittal consists of the complete set of proposed Technical Specifications for ABWR with the exception of s the Instrumentation Section (Section 3.4). This is still being developed due to the differences
;v) between ABWR and past BWR designs.
These ABWR Technical Specifications are based on the BWR Owners' Group (BWROG) Improved Technical Specifications (ITS) submitted for NRC staff review on May 5,1989. This is in keeping with the NRC Staff's expressed desire that the ABWR Certification program include Technical Specifications consistent in format and content with the latest NRC direction and Industry effort. The following points are in keeping with the discussions and agreements between NRC staff, the BWROG and GE on April 5,1989. First, it is important that the ITS receive its initial review prior to the ABWR Technical Specifications, and that plan was agreed to. The ITS constitutes the ABWR submittal insofar as the ITS and ABWR specifications are common. NRC review and approval of the ITS will constitute NRC review and approval of the content of the ABWR specification which is common to both. Discussion on the common content is to be exclusively between NRC and the BWROG. NRC will also review for approval the differences unique to ABWR provided in the present submittal. Such approval, and approval of the common content resulting from the ITS review, will together constitute approval of the entire ABWR Technical Specification. To facilitate this review, the ABWR Technical Specifications have been edited as follows:
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L IOd 8906280317 890623 i q PDR ADOCK M OOO605 PDC t A _ _ _ _ _ _ _ _ _ _ _ _ i
, .b P. W. Marriott to Charles L. Miller l
Page 2 June 23,1989 (1) Table 1 cross references the corresponding sections of the ABWR Technical Specifications versus the ITS; (2) Those places where the ABWR product differs from the BWROG ITS have been denoted by underlining. This may include individual words or phrases, or entire sections. Differences in numerical values or nomenclature only, however, are typically not underlined. Section headings in the Table of Contents have also been unde.rlined to show which sections contain underlined differences. The ABWR bases document included is meant to stand alone and, therefore, does not indicate differences from the BWROG ITS Bases. (3) The content common to .ABWR and ITS is not underlined. (This is done to make the differences easier to understand and the text easier to follow, while at the same time emphasizing that the Owner's Group submittal--not this submittal- is official for ABWR as regards the common content.) It is intended that the Technical Specifications (and Bases) applicable to ABWR Instrumentation
,m (Section 3.4) will be submitted by the end of July 1989. When the review and approval of both the 1
) ITS and the ABWR - unique content are complete, GE will then amend the SSAR to incude the complete set of proposed Technical Specifications, as Chapter 16 of the submittal.
Sincerely, P. W. Marrmtt, Manager Licensing and Consulting Services cc: D. R. Wilkins (GE) F. A. Ross (DOE) J. F. Ouirk (GE) D. C. Scaletti (NRC)
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L ~ TABLE 1 - ,.
' TECH SPEC CROSS REFERENCE
- ABWR vs BWROG ITS LABWR SECTION AND TITLE BWROG ITS SECTION *
- f. 1. Through Section 3.0 s Same i L3.i . REACTIVITY CONTROL SYSTEMS a,
f1L1 . SHUTDOWN MARGIN - Same 1L2 - ~ Control Rod OPERABILITY - " 4 11 1 : Control Rod Scram Times l , ;1 .11A Control Rod Scram Accumulators - > 1.L5 3 Control Rod Drive Coupling l 1],4 Rod Pattern Control o 3.1.7 ' 3.1.7 J Standby Liquid Control System . 3.1.8 3.2 POWER DISTRIBUTION LIMITS
'3.2.1 ' VERAGE PLANAR LINEAR HEAT HEAT GENERATION RATE - Same
- 3.2.2 . . MINIMUM CRITICAL POWER RATIO 3.2.3 LINEAR HEAT GENERATION RATE 3.3 INSTRUMENTATION
, [Later]
' 3.4 REACTOR COOLANT SYSTEM 141 Recirculation Pumps Oneratine Same 141 Eaferv/ Relief Valves 3.4.4 3.4.3 Operational Leakage ' 3.4.5 3.4.4 - Specific Activity. 3.4.6
- 141 Residual Heat Removal Shutdown 3.4.7 lif' Reactor Coolant System Pressure /
Temperature Limits 3.4.8 3.4.7 Reactor Steam Dome Pressure 3.4.9 3.5 ECCS 3.. lil ECCS - Operating Same " 3.5.2 ECCS - Shutdown Same , __ _ 1
y) 3.6 CONTAINMENT SYSTEM BWROG ITS SECTION
- 3.6.1 Primary Containment Systems 3.6.1.1 Primary Containment Same (BWR/4) 3.6.1.2 Containment Air Locks " (BWR/4) 3.6.1.3 Containment Pressure " (BWR/4) 3.6.1.4 Containment Average Air Temperature " (BWR/4) 3.6.1.5 Priinary Containment [and Pressure]
Isolation Valves 3.6.1.6 (BWR/4) 3.6.1.6 Wetwell to-Drywell Vacuum Breakers 3.6.1.9 (BWR/4) 3.6.2 Suppression Pool 3.6.2.1 Suppression Pool Average
. Temperature. Same (BWR/4) 3.6.2.2 Suppression Pool Water Level 3.6.23 Residual Heat Removal Sunnression Pool Cooling 3.6.2.4 Resideal Heat Removal Wetwell Sorav 3.6.3 Hydrogen Control 3.63.1 Containment Hydrogen Recombiner Systems Same
[V3 3.6.3.2 Primary Containment Oxygen ! Concentration 3.6.3.3 (BWR/4) ! 3.6.4 Secondary Containment Systems 3.6.4.1 Secondary Containment Same 3.6.4.2 Secondary Containment Isolation Valves 3.6.4.3 Standby Gas Treatment System 3.6.3.3 (BWR/4) 3.7 PLANT SYSTEMS 3.ll Reactor Building Cooline Water / Reactor Building 3.7.1/3 * *
- Service Water System - Operatinc 322 Reactor Building Cooline Water / Reactor Building Service Water System - Shut 10wn 3.7.2/4 * "
3.73 Control Room HVAC Emergency Recirculation System 3.7.4 3.7.4 Main Condenser Offgas 3.7.5 3.8 ELECTRICAL POWER SYSTEMS 3.11 A.C. Sources - Operating Same (BWR/4) 3.8.2 A.C. Sources - Shutdown 3.8.3 Diesel Fuel Oil
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- - _ _ _ - _ - - _ _ _ w_ _ - - , - _ _ _ _ _ _ , -P .t -(
ABWR SECTION AND TITLE BWROG ITS SECTION
- 3.8.4 ; D.C. Sources - Oneratine 1M D.C. Sources - Shutdown " -
3.8.6 Battery Electrolyte 3.8.7 3.8.7 A,C Power Distribution Systems - Operating 3.8.8 1M D.C. Power Distribution Systems - Operating 3.8.9 3.8.9 A.C. and D.C. Power Distribution - Systems - Shutdown 3.8.10-3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks Same 3.9.2 Refuel Position One Rod-Out Interlock 3.9.3 - Control Rod Position 3.9.4 Control Rod Position Indication 1M Control Rod Operability- Refueling " 3.9.6 Water Level- Reactor Pressure Vessel 3.9.7 ' Water Level - Spent Fuel Storage Pools 3.9.8 Residual Heat Removal- High Water Level . 3.9.9 Residual Heat Removal- Low Water Level I 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testine Operation Same
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3.10.2 Reactor Mode Switch Interlock Testing 11M . Control Rod Withdrawal - Hot Shutdown 13.10.4 Control Rod Withdrawal - Cold Shutdown 3.10.5 . Control Rod Drive Removal- Refueling 3.10.6 Multiple Control Rod Withdrawal- Refueling 3,10,7 Control Rod Testing - Operating " AU BWROG IU Sections refer to BWR/6 version unless stated othenvise RCIC (3.5.3) is part of ECCS for ABWR and has therefore been incorporated into 3.5.1 ABWR Service Water utilizes concepts from BWR/4 and BWR/6 ITS 3.7.1 - 3.7.4 ITS 3.7.1 - 3.7.3 i O
l l' PREFACE l b(~N Technical Specifications are explicit restrictions on the operation of a commercial nuclear power plant. They.are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. -Technical specifications preserve the primary success path relied upon to detect and respond to accidents. They also complement the concept of defense in depth. Section 182a of the Atomic Energy Act of 1954, as amended (the Act), 47 U.S.C. 2011, at 2232, provides the legislative framework within which technical specifications are required. Section 182a of the Act requires in pertinent part:
"In connection with applications for licenses to operate production or utilization facilities, the applicant shall . state such technical specifications, including information on the amount, kind, and source of special nuclear material required, the place of use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will... provide adequate protection to the
' health and safety of the public. Such technical specifications shall be a part of any license issued."
() The regulatory framework implementing Section 182a of the Act is the NRC's regulation Title 10 Code of Federal Regulations Part 50, Section 50.36,
" Technical Specifications". This regulation provides in part that each operating license:
"will include technical specifications...(to) be derived from the analysis and evaluation included in the safety analysis report, and amendments thereto...and may also include such additional technical specifications as the Commission finds appropriate."
The Commission has issued an interim " Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," 52 FR 3788, February 6, 1987. This interim policy statement sets out specific criteria for the content of technical specifications. The interim policy statement specific-ally recognizes that:
"The purpose of Technical Specifications is to impose conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an j immediate threat to the public health and safety by establishing '
those conditions of operation which cannot be changed without prior Commission approval and by identifying those features which are of controlling importance to safety." This set of orooosed Technical Specifications establish these conditions and p) ( limitations for the ABWR. This set of technical specifications is intended to be used as a guide in the development of plant specific sets of technical specifications for olants whose license aoolications reference the ABWR standard plant. ABWR i 5/31/89
i fj ! b TABLE OF CONTENTS bz " PREFACE . . . . . . . . . . . . . . . . ... . . . . . . ... . . . . . . i
' TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . ii LIST OF TABLES ........................... ~ vi LIST.0F FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . vii LIST OF EFFECTIVE PAGES . . . . . . . . . . . . . . . . . . . . . . . viii
- 1. USE AND APPLICATION 1.1 DEFINITIONS .......................... 1-1 ACTIONS . . . . ... . . . . . . . . . . . . . . . . . . . . 1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ........ 1-1 CORE ALTERATION . . . . . . . . . . . . . . . . . . . . . . 1 [ CORE OPERATING LIMITS REPORT] .............. la2-DOSE EQUIVALENT I-131 . . . . . . . . . . . . . . . . . . . 1-2 LEAKAGE . . ... . . . . . . . . . . . . . . . . . . . . . . 1-3 f-LINEAR HEAT GENERATION RATE . . . . . . . . . . . . .-. . . 1-3 t MINIMUM CRITICAL POWER RATIO ............... 1-4 MODE . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 OPERABLE - OPERABILITY .................. 1-4 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . . . . . 1-6 RATED THERMAL POWER . . . . . . . . . . . . . . . . . . . . 1-6 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . 1-6~
STAGGERED TEST BASIS ................... 1-7 THERMAL POWER . . . . . . . . . . . . . . . . . . . . . . . 1-7 1.2 LOGICAL CONNECTORS . . . . . . . . . . . . . . . . . . . . . . . 1-8 1.3 COMPLETION TIMES . . . . . . . . . . . . . . . . . . . . . . . . 1-10 1.4 FREQUENCY . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-14 1.5 LEGAL CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . . 1-17 ; 1 I
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i NN ABWR 11 5/31/89 l l 1 1 _---____...---.,.____-____-_-_-____________--Q
i l l TABLE OF CONTENTS (continued) Pace
- 2. SAFETY LIMITS 2.1 Safety Limits . . . . . . . . . . . . . . . . . . . . . . 2-1 222 Safety Limit Violation . . . . . . . . . . . . . . . . . 2-1
- 3. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 3.0 APPLICABILITY Limiting Conditions for Operation . . . . . . . . . . . . . . 3-1 Surveillance Requirements . . . . . . . . . . . . . . . . . . 3-4 3.1 REACTIVITY CONTROL SYSTEMS 3.1,1 SHU1DOWN MARGIN . . . . . . . . . . . . . . . . . . 3.1-1 3.1.2 Control Rod OPERABILITY . . . . . . . . . . . . . . 3.1-S 3.1,3 Control Rod Scram Times . . . . . . . . . . . . . . 3.1-10 3.1.4 Control Rod Scram Accumulators . . . . . . . . . . . 3.1-14 3.1,5 Control Rod Drive Couclina . . . . . . . . . . . . . 3.1-16 3.1.6 Rod Pattern Control . . . . . . . . . . . . . . . . 3.1-18 -
3.1.7 Standby Liquid Control System . . . . . . . . . . . 3.1-20 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO . . . . . . . . . . . . 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE , . . . . . . . . . . . 3.2-3 3.3 INSTRUMENTATION [Later] 3.4 REACTOR COOLANT SYSTEM 3,4,1 Recirculation Pumos Operatina . . . . . . . . . . . ' 3.4-1 3.4.2 Safet v/ Relief Val ves . . . . . . . . . . . . . . . . 3.4-4 3.4.3 Operational Leakage . . . . . . . . . . . . . . . . 3.4-6 3.4.4 Specific Activity . . . . . . . . . . . . . . . . . 3.4-8 3.4 5 Residual Heat Removal - Shutdown . . . . . . . . . . 3.4-10 3.4.6 Reactor Coolant System Pressure / Temperature Limits . 3.4-12 3.4.7 Reactor Steam Dome Pressure . . . . . . . . . . . . 3.4-15 3.5 ECCS 3.5.1 ECCS - Operatina .................. 3.5-1 3.5.2 ECCS - Shutdown . . . . . . . . . . . . . . . . . . . 3.5-7 ABWR iii 5/31/89
, 1 1
j 1 1 It" 1 I 'I; TABLE OF CONTENTS 1 (continued) l P_iL91
- 3. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS (continued) 3.6 CONTAINMENT SYSTEMS-3.6.1 Primary Containment Systems 3.6.1.1 Primary Containment . . . . . . . . . . . . 3.6-1 3.6.1.2 Containment Air Locks . . . . . . . . . . . 3.6-3
'3.6.1.3 Containment Pressure ........... 3.6-7 3.6.1.4 Containment Average Air Temperature . . . . 3.6-8 3.6.1.5 Primary Containment [and Pressure]
Isolation Velves ............. 3.6-9 3.6.1.6 Wetwell-to-Drywell Vacuum Breakers .... 3.6-16 3.6.2 Suppression Pool
. 1,6.2,1 Suppression Pool Averaae Temperature. . . . 3.6-19 3.6.2.2 -Suppression Pool Water Level ....... .
3.6-22 3.6.2.3 Residual Heat Removal Suppression Pool Coolina . . . . . . . . . . . . . . . . . . 3.6-24 3.6.2.4 Residual Heat Removal Wetwell' Sorav . . . . 3.6-27 3.6.3 Hydrogen Control 3.6.3.1 Coritainment Hydrogen Recombiner Systems . . 3.6-29 3.6.3.2 Primary Containment Oxygen Concentration .............. 3.6-30 3.6.4 Secondary Containment Systems 3.6.4.1 Secondary Containment . . . . . . . . . . . 3.6-31 3.6.4.2 Secondary Containment Isolation Valves .. 3.6-34 3.6.4.3 Standby Gas Treatment System ....... 3.6-37 3.7 PLANT SYSTEMS 3 . 7. , ) Reactor Buildina Coolina Water / Reactor Buildina Service Water System - Operatina .......... 3.7-1 3.7.2 Reactor Buildina Coolina Water / Reactor Buildina Service Water System - Shutdown . .......... 3.7-3 3.7.3 Control Room HVAC Emergency Recirculation System . . 3.7-8 3.7.4 Main Condenser Offgas . . . . . . . . . . . . . . . . 3.7-1J (~'N V ABWR iv 5/31/89
TABLE OF CONTENTS l (continued) i Psoe
- 3. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS (continued) 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 A.C. Sourees - Operatina .............. 3.8-1 3.8.2 A.C. Sources - Shutdown .............. 3.8-20 3.8.3 Di e s el Fue ' 011 . . . . . . . . . . . . . . . . . . . 3.8-22 3.S.4 D.C. Sources - Operatina .............. 3.8-26 3.8.5 p.C. Sources - Shutdown .............. 3.8-31 3.8.6 Battery Electrolyte ................ 3.8-33 3.6.7 A.C. Power Distribution Systems - Ooeratina . . . . . 3.8-37 3.8.8 D.C. Power Distribution Systems - Operatina . . . . . 3.8-40 3,8,9 A.C. and D.C. Power Distribution Systems - Shutdown . 3.8-42 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks . . . . . . . . . . . 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock . . . . . . . 3.9-2 3.9.3 Control Rod Posi tion . . . . . . . . . . . . . . . . 3.9-4 3.9.4 Control Rod Position Indication . . . . . . . . . . 3.9-5 3.9.5 Control Rod OPERABILITY - Refuelina . . . . . . . . 3.9-7 3.9.6 Water Level - Reactor Pressure Vessel . . . . . . . 3.9-8 3.9.7 Water Level - Spent Fuel Storage Pools . . . . . . . 3.9-9 3.9.8 Residual Heat Removal - High Water Level . . . . . . 3.9-10 3.9.9 Residual Heat Removal - Low Water Level . . . . . . 3.9-12 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testina Operation . . 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing . . . . . . . 3.10-3 3.10.3 Control Rod Withdrawal - Hot Shutdown . . . . . . . 3.10-5 3.10.4 Control Rod Withdrawal - Cold Shutdown . . . . . . . 3.10-8 3.10.5 Control Rod Drive Removal - Refuelina . . . . . . . 3.10-11 3.10.6 Multiple Control Rod Withdrawal - Refueling . . . . 3.10-14 3.10.7 Control Rod Testina - Operatina . . . . . . . . . . 3.10-16 O
ABWR v 5/31/89
IT L-L . r* LIST OF TABLES Table No, 11111 EiLER! 1.1-1 MODES . . . . . . . . . . . . . . . . . . . . . . . 1-8' 3.1.3.1-1 Control Rod . Scram Times . . . . . . ... . . .-. . . 3.1-13 3.8.1-1 Diesel Generator Test Schedule . . . . . . . . . . . . 3.8-19 3.8.6-1 Battery Surveillance Requirements . . . . . . . . . 3.8-36 0 ABWR vi 5/31/89
LIST OF FIGURES O Fioure No. Iitig g 3.I.8-1 Sodium Pentaborate Solution Temperature Versus Concentration Requirements , 3.1-23 i O i l i l l ABWR vii 5/31/89 l l 1 t.. - _ _ _ _ _ _ - . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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1 - - 1 5. - - 1-6 - - ABWR viii 5/31/89
Definitions i 1.1 l n. ij 1.1 DEFINITIONS
................................... NOTE---------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and their Bases. g Definition ACTIONS ACTIONS shall be that part of a Technical Specification which prescribes Required Actions to be taken under designated Conditions within specified Completion Times. AVERAGE PLANAR LINEAR The AVERAGE PLANAR LINEAR HEAT HEAT GENERATION RATE GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the rum of the heat-generation rate per unit length of fuel rod (LINEAR HEAT GENERATION RATE) for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel (- bundle at that height. CORE ALTERATION A CORE' ALTERATION shall be the movement or manipulation of any fuel, sources, or reactivity control ! components within the reactor vessel with the vessel head removed and fuel in the vessel. Movement of SRNMs, LPRMs, TIPS or special movable detec-tors (including replacement) is not considered a CORE ALTERATION. Suspen-sion of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe, conservative position.
/-
C/ (continued) 1 ABWP, 11 5/31/89 -
Definitions 1.1 1.1 DEFINITIONS (continued) Term Definition [ CORE OPERATING LIMITS REPORT] The [ CORE OPERATING LIMITS REPORT (COLR)] is a reload-cycle specific document, its supplements and revi-sions, that provides core operating limits for the current operating reload cycle. These cycle specific core operating limits shall be deter-mined for each reload cycle in accord-ance with Specification [5.x]. Plant operation within these operating limits is addressed in individual specifications. DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites". i I (continued) ABWR 1-2 5/31/89 u_____--__-_----- -
4fx Definitions () - , 1.1 1.1 DEFINITIONS (continued) J.frm Definition LEAKAGE LEAKAGE shall be: A. Identified Leakage:
- 1. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or.
- 2. Leakage into the drywell atmosphere from sources that are both.specifically located and known either not to interfere with the operation of leakage
"- detection systems, or not to be Pressure Boundary g- .
Leakage. B. Pressure Boundary Leakage: Leakage through a non-isolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. C. Unidentified Leakage: All leakage which is not Identified Leakage. D. Total Leakage: Sum of the Identified and Unidentified Leakage. LINEAR HEAT GENERATION RATE The LINEAR HEAT GENERATION RATE ; (LHGR) shall be the heat generation 1 per unit length of fuel rod. It is I the integral of the heat flux over the heat transfer area associated with the unit length. L (continued) I ABWR. 1-3 5/31/89
Definitions 1.1 1.1 DEFINITIONS (continued) Term Definition MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest Critical Power Ratio (CPR) which exists in the core. The CPR shall be the ratio of that power in the assembly which is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE A MODE shall correspond to any one inclusive combination of mode switch position and average reactor coolant
' emperature specified in Table 1.1-1 eith fuel in the reactor vessel.
OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE, or have OPERABILITY, when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, compor.ent or device to perform its function (s) are also capable of performing their related support function (s). (continued) O ABWR 1-4 5/31/89
9 Definitions-1.1
,, ~\ .
t 1.1 DEFINITIONS (continued) Table 1.1-1 MODES REACTOR
' MODE SWITCH AVERAGE REACTOR MODEI ") TITLE POSITION COOLANT TEMPERATURE, Of 1 Power Operation. Run Any temperature 2 Startup Startup/ Hot Standby Any temperature 3 Hot Shutdown Shutdown '> 200 0F 0
4 . Cold Shutdown Shutdowa $ 200 F-5 Refueling Shutdown or Refuel Any temperature
/g (a) In MODES 1 through 4 fuel.is in the reactor vessel with the reactor V- vessel head closure bolts fully tensioned. In H0DE 5, fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.
(continued) ABWR l-5 5/31/89
Definitions 1.1 ) I 1.1 DEFINITIONS (continued) Term Definition ' PHYSICS TESIS PHYSICS TESTS shall be those tests performed to measure the fund ue m l l nuclear characteristics of the reactor I core and related instrumentation, and { which are: ' A. described in Chapter 14 of the i SSAR, i B. authorized under the provisions of 10 CFR 50.59, or C. otherwise approved by the Commission. RATED THERMAL POWER RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3926 Mwt. SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subtritical assuming all control rods are fully inserted except for the control rod oair of highest reactivity worth which is assumed to be fully withdrawn and the reactor is xenon free and the moderator temperature is'68'F. With a control rod not capable of being fully inserted, the reactivity worth of this control rod must also be accounted for in the determination of SHUTDOWN i MARGIN. Note: A control rod pair consists of two control rods which are connected to the same, shared Hydraulic Control Unit (HCV) scram accumulator. All control rods share an accumulator with one other rod excent for the center control rod which has its own accumulator. (continued) O ABWR l-6 5/31/89
!i ':: i j Definitions 31
../_ T -
1.1 DEFINITIONS (continued) Q. IgrLn Definition STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels,'or other designated components during the specified surveillance frequency such that all systems, subsystems, channels,.or other designated components are tested during n surveillance frequency intervals, where n-is the total number of systems, subsystems, channels or other designated components in the associated Function. THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. ABWR 1-7 5/31/89 { 4
i tcgical Connectors 1.2 1.2 LOGICAL CONNECTORS O PURPOSE Logical connectors are used in Technical Specifications to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillance and Frequencies. The only logical connectors which appear in technical specifications are AND and QB. The physical arrangement of these connectors constitute logical conventions with specific meaning. The intent of this section is to provide specific examples of logical connectors and expirin the intended meaning. EXAMPLES Example 1 demonstrates that for Condition A, both Required Actions must be completed. In this case, the logical connector AND is left justified and shows that both Required Actions A.1 and A.2 are required to be performed. Example 1. CONDITION REQUIRED ACTION A A.1
= 9 A.2 (continued)
ABWR 1-8 5/31/89 ___ -_ 1
34 kA 1 I. - Logical Connectors v 1.2 f !1.2 LOGICAL CONNECTORS fcontinued)
' EXAMPLES In' Example 2, Required Action A.1,. Required Actie A.2 and (continued) Required Action A.3:are required to be performed. For Required Action A.2, the indented position of the logical connector E indicates that A.2.1 and A.2.2 are alternate and equal choices, only one of which has to be performed. .
Example 2. CONDITION REQUIRED ACTION A. A.1 h@ A.2.1 E A.2.2 8.N_Q A.3 O ABWR 1-9 5/31/89
7 Completion Times 1.3 1.3 COMPLETION TIMES PURPOSE When a Limiting Condition for Operation is not met, the individual Technical Specifications identify equired Actions which must be completed within specified times. An understanding of the correct interpretation of these specified Completion Times is necessary for compliance with the requirements of the Technical Specifications. This section discusses the proper use and interpretation of Completion Times. COMPLETION The Completion Time is the amount of time allowed to perform a TIME Required Action and is referenced to the time it is discovered that an ACTIONS' Condition has been entered, unless otherwise specified. Required Actions must be completed within the specified Completion Times. When the ACTIONS' Condition changes, new Required Action (s) and Completion Time (s) are established from the time of discovery that the new Condition is entered. The Required Actions for the Condition which no longer exists are not required to be completed. When a Condition has associated Required Actions which result in a shutdown, the total time to reach a MODE or other specified condition is independent of what MODE the unit is in when the Condition is entered. This is acceptable because a penalty should not be incurred if a Condition is entered while the unit is in a lower MODE of operation where the consequences of a postulated event are significantly reduced. If a lower MODE of operation is reached in less time than allowed, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced. For example, if the Required Action is to be in MODE 3 in 12 hours and to be in MODE 4 in 36 hours, and the Condition is entered in MODE 3, the time allowed to reach MODE 4 is the next 36 hours because the total time to reach MODE 4 is not reduced from the allowable limit of 36 hours. Another example is if the Required Action is to be in MODE 3 in 12 hours and to be in MODE 4 in 36 hours, and MODE 3 is reached in 8 hours, the time allowed to reach MODE 4 is the next 28 hours because the total time to reach MODE 4 is not reduced from the allowable limit of 36 hours. (continued) ABWR l-10 5/31/89 j l i I J
mw - e "w. 2
" Completion Time's:
1.3.
~ 1.3 COMPLETION TIMES (continued)-
- When a Condition is entered and its Required Action.is not
' COMPLETION TIME. completed within its specified Completion. Time, a new Condition?
L is established.. This new Condition may have its own Required-
' Action (s) and Completion Time (s). . For example, if one componenttis' inoperable and is not returned to OPERABLE status
- within the Completion time, the new Condition may be:
" Required Action'and.associat5d Completion Time not met."
A new Required' Action would-then be given for this Condition,o with a new Completion Time. Some Required Actions. require completion based on an initial discovery time instead of time of entry into the Condition. Typically,- this is the same time. However, for: Example 1.3.1', in the' event the Condition of a' single inoperable-subsystem degrades into'two inoperable subsystems, Required. Action B.2 requires restoration of the first . inoperable subsystem in seven.
' days.from discovery of its inoperability. In some cases, this-may be a- shorter period of time than the 72 hours allowed in'
' Required Action B.I. Furthermore, should'one of the two subsystems'be restored to OPERABLE status and the single-
' inoperable subsystem Condition re-entered,'the Completion Time
/ for this example is based on the time of initial discovery of the remaining inoperable subsystem and not from the' time of.'
re-entry.into the Condition. Example 1.3.1 CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem A.1 Restore subsystem 7 days from n inoperable. to OPERABLE discovery of-status. inoperable subsystem B. Two B.1 Restore at least 72 hours subsystems one subsystem to inoperable. OPERABLE status. A!!D B.2 Restore first 7 days from inoperable discovery of subsystem to initial OPERABLE status. inoperable subrystem O (continued) ABWR 1-11 5/31/89
Completion Times
, 1.3 1.3 COMPLETION TIMES (continued)
COMPLETION Some Conditions allow a variable number of inoperable TIME components. These Conditions specify this allowance as "one or more" or " s x". These Conditions may be entered each time the number of inoperable components change, as long as the Condition still applies. Each entry has its own independent Required Action and Completion Time. This multiple entry of one Condition is not an exception to the requirement that only one Condition within an LCOs' ACTION may be entered at any time (LCO 3.0.2). Some Conditions allow multiple reasons for inoperability. These Conditions specify this allowance as " inoperable for one or more reasons". They may be entered each time the reasons for inoperability change, as long as the Condition still applies. Each reason for inoperability has its own independent Required Action and Completion Time. This multiple entry of one Condition is not an exception to the requirement that only one Condition within an LCOs' ACTION may be entered at any time (LC0 3.0.2). One additional exception to the Completion Time being referenced to the entry into its Condition is noted by the use of "thereafter" following the Completion Time. This requires the initiation of a periodic re-performance of the Required Action after the successful completion of its initial performance. In some cases, "immediately" and "As Soon As Practicable" are used as Completion Times. "Immediately" is generally used when a Condition cannot be permitted to continue to exist and essentially little or no time is required to complete the Required Action. Examples of when "Immediately" is used include placing the Reactor Mode Switch in the Shutdown position, and declaring a system inoperable. This is because placing a switch in a specified position requires only the time necessary to perform that action and declaring a system inoperable essentially takes no time as it is an administrative action intended to start an ACTIONS Completion Time clock. "As Soon As Practicable" is generally used when a Condition cannot be permitted to continue to exist any longer than absolutely necessary given the prevailing plant conditions and some period of time is required to complete the Required Action. This is because some judgement is required to determine the quickest and safest method to complete the Required Action (e.g., suspending operations with the potential for draining the reactor vessel) or because the Condition is such that the only ACTION that can be taken is to continue to attempt to comply with the Required Actions (e.g., restoring both RHR shutdown cooling loops to OPERABLE status). "As Soon As Practicable" is (continued) ABWR 1-12 5/31/89
- p. .
q--
.s-i t [. ,
Completion Times
. , , 1.3
, . {-
.- 1.3 COMPLETION TIMES (continued)'
- y. COMPLETION -intended to ensure that appropriate actions ~are taken'in
.. TIME .
reasonable time, but is not intended to require extraordinary lj (continued) actions be taken which do not provide any real-safety benefit. O ABWR 1-13 5/31/89
Frequency 1.4 1.4 FRE0VENCV l ! PURPOSE Each Surveillance Requirement (SR) has a specified Frequency in ; I which the Surveillance must be met to support meeting the J associated Limiting Condition for Operation (LCO). An understanding of the correct interpretation of this Frequency is necessary for compliance with the requirements of the Technical Specifications. This section discusses thE proper use and interpretation of the Surveillance Frequency. FREQUENCY In examples provided and discussed below, the Applicability of the LCO is given as MODES 1, 2, and 3. EXAMPLE Example SR 1.4.1 contains the type of SR most often encountered SR 1.4.1 throughout the Technical Specifications. It specifies an interval (31 days) during which the associated Surveillance Requirements must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Each performance must meet the requirements of SR 3.0.2. The measurement of this interval continues at all times even when the Surveillance Requirement is not required by SR 3.0.1. In cases where the interval as specified by SR 3.0.2 is exceeded while not in a MODE or other specified condition for which performance of the Surveillance Requirement is required, the Surveillance Requirement must be performed prior to entry into the MODE or other specified condition. Failure to do so would result in a "faileo" surveillance and an invalid MODE change per LCO 3.0.4. Sometimes special conditions dictate when a Surveillance is to be met. These conditions apply to the Surveillance or to the Frequency or both, and are not "other specified condition (s) in the Applicability" described in LCO 3.0.1, LCO 3.0.4, SR 3.0.1 and SR 3.0.4. They are however, "otherwise stated" conditions allowed by SR 3.0.1. Furthermore, these conditions may be stated as clarifying notes cr as part of the Surveillance Requirement itself. Examples SR 1.4.2, SR 1.4.3, and SR 1.4.4 discuss these special conditions. SR 1.4.1 Perform a (CHANNEL FUNCTIONAL 31 days TEST). (continued) ABWR l-14 5/31/89
Frequency 1.4 1.4 FRE00ENCY fcontinuhdi EXAMPLE Example SR 1.4.2 requires that teus Surveillance be performed SR 1.4.2 only at or above 25% of RTP. The phrase "Only required..." (continued) means this Surveillance may be performed in any MODE or other specified condition where unit status would allow successful completion, but is not required to be performed unless greater than or equal to 25% of RTP. The interval measurement for the Frequency of this Surveillance continues at all times as described in Example 1.4.1. However, if this Surveillance did not meet SR 3.0.2 while operation continued at less than 25% of RTP, it would not constitute a failure to meet the LCO. The Surveillance is not required below 25% of RTP, even though the LCO, per its Applicability, may be required to be met. Prior to reaching 25% of RTP, if the Surveillance were not performed within the interval as allowed by SR 3.0.2, it must still be performed prior to reaching 25% of RTP. If it is not performed prior to reaching 25% of RTP, the provisions of SR 3.0.3 would apply. SR 1.4.2 ------------NOTE---------- '- Only required with THERMAL g POWER 2 25% of RTP.
. g] ............................
Demonstrate the absolute 7 days difference between the APRM channels and the calculated power is s 2% of RTP. EXAMPLE Example SR 1.4.3 allows the option of selecting one of two SR 1.4.3 Frequencies in which to satisfactorily perform the Surveillance. The first Frequency (92 days)'is like that in Example SR 1.4.1. The second option is an example of a Frequency where the measurement of the 12 hour interval does not continue at all times. The measurement begins only after exceeding 920 psig. If reactor pressure is less than or equal to 920 psig at a time when the first Frequency cannot k met as required by SR 3.0.2, the second Frequency can be 'eleu.td and will not be considered a failure to perform a Survedin.e Requirement within the specified frequency. Upon reaching 920 psig, 12 hours (plus 25% per SR 3.0.2) would be allowed within which the Surveillance must be completed. If not performed within this interval, it would then become a failure to perform a Surveillance Requirement within the specified Frequency, and (continued) l ABWR l-15 5/31/89 I _-_ m
Frequency 1.4 1.4 FRE00ENCY (continued) EXAMPLE 1.4.3 only then would MODE changes be _ restricted per SR 3.0.4. performed, both Frequencies are met. Selection of either Once O (continued) Frequency would again be allowed, except, that as long as reactor steam dome pressure is 2 920 psig, the 92 day frequency must- continue to be met to satisfy Example SR 1.4.3. The condition of the frequency (e.g., when reactor steam dome pressure is 2 920 psig) may be expressed as a note or as prose as it is in Example SR 1.4.3. SR 1.4.3 Verify RCIC flow greater than 92 days 800 gpm with steam supplied to the turbine from 920 to 1050 .QB i psig. 12 hours when reactor steam dome pressure is 2 [ 920 ) psig EXAMPLE Example SR 1.4.4 is a Surveillance with a one time performance SR 1.4.4 Frequency followed by an Example SR 1.4.1 type Frequency. The logical connector "AND" will require that both Frequencies be met. "Once" allows a single performance to satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). "Thereafter" is used to mean future performances will be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). SR 1.4.4 Verify APLHGRS are less than Once within or equal to the required 12 hosrs limits. after 2 25% of RTP. AND ! 24 hours thereafter i O ABWR l-16 5/31/89 l
r L. o Legal Considerations L 1.5 1.5 LEGAL CONSIDERATIONS INTRODUCTION The Atomic Energy Act of 1954 requires that technical specifications be a part of operating licensen. As such, they are enforceable under federal statute as well as. Title 10 of the Code of Federal Regulations. When an applicant receives a license from the Nuclear Regulatory Commission to operate a commarcial nuclear power plant, the technical specifications are included as Appendix A to the license. Consequently, whenever a change is made to a plant's technical-specifications, an amendment to the operating. license is required. There are; however, certain' sections and additional items included with these Technical Specifications that are for , information or convenience and are not legally a part of the ! Technical Specifications or Operating License. This section I identifies the legal parts (i.e., the items that require a license amendment to change) of these Technical Specifications, and those additional parts that can be changed without requiring a license (or technical specification) amendment. LEGAL PARTS Title 10 Code of Federal Regulations, Part 50.36 delineates 7% those items which are to be included in technical (v )- specifications. These items to be included are: o Safety Limits o Limiting Safety System Settings o Limiting Conditions for Operation o Surveillance Requirements i o Design Features, and o Administrative Controls In addition, the Use and Applications Division of these Technical Specifications (comprised of Definitions, Logical Connectors, Completion Times, Frequency and Legal Considerations) is also a legal part of these Technical Specifications. , Since these Technical Specifications are issued as Appendix A to the Operating License, any change to the above defined sections constitutes a license amendment. As such, the requirements of 10 CFR 50.90, 50.91, and 50.92 apply. O (continued) ABWR l-17 5/31/89
Legal Considerations 1.5 1.5 LEGAL CONSIDERATIONS (continued) FRONT MATTER Front Matter is all the material in the front of the technical specifications used to identify and locate specific information. It includes: o Preface o Title Page o Table of Contents o List of Tables o List of Figures o List of Effective Pages None of this material is required by Title 10, Code of Federal Regulations, Part 50.36, and the Front Matter does not include any requirements on the safe operation of the plant. Therefore, the Front Matter is not a legal part of these Technical Specifications or Operating License. CROSS Cross-References are included in the body of these Technical REFERENCES Specifications to assist the user in determining other applicable requirements for a common system or component. This section is not required by Title 10, Code of Federal Regulations, Part 50.36, and is included with this set of Technical Specifications at the discretion of the Licensee. As such, they are not a legal part of these Technical Specifications or Operating License. BASES Title 10, Code of Federal Regulations, Part 50.36 includes the following statement, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall be included in the application, ! but shall not become part of the technical specifications." Therefore, the bases are not a legal part of either of these Technical Specifications or the Operating License. The bases are prepared and controlled by the Licensea, and a change / revision to the bases shall not constitute a change / amendment to the License. 4 i l i i i ABWR 1-18 5/31/89 l i
/
Safety Limitsi
~ ') 2.0
. (~/:
\~-
..2.0: SAFETY. LIMITS 2.1 SAFETY LIMITS-2.1.1 With the reactor steam dome pressure.< 785 psig or core flow
< 10% of rated core flow:
THERMAL POWER shall be 125% of RATED THERMAL' POWER. 2.1.2 'With the reactor steam dome pressure 1785 psig and core flow 1 10% of rated core flow: MINIMUM CRITICAL POWER RATIO 'shall be l' l.07 2.1.3 Reactor vessel water level shall be.> the top of active
' irradiated fuel.
2.1.4 Reactor steam dome pressure shall be 51325 psig. 2.2 SAFETY LIMIT VIOLATION With any Safety Limit not met the following actions shall be met:
./ 2 2.1 Within one hour notify the NRC Operations Center in
\, accordance with 10CFR50.72.
2.2.2 Within two hours: A .- Restore ct..apliance with all Safety Limits, and B. Insert all ' insertable control rods. 2.2,3 The Iacoropriate on-site and off-site manaaement personnel and safety review arouosi shall be notified within 24 hours. 2.2.4 A Licensee Event Report shall be prepared pursuant to 10CFR50.73. The Licensee Event Report shall be submitted to the Commission and the faooropriate on-site'and off-site manaaement personnel and safety review aroupsl within 30 days of the violation. 2.2.5 Critical operation of the unit shall not be resumed until authorized by the Commission. f lV ABWR 2-1 5/31/89
LCO Applicability'
,4 3.0 br 3.0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION (LCOs)
LCO 3.0.1 Limiting Conditions for Operation shall be met during the MODES or other specified conditions in their Applicability except as provided in LC0 3.0.2. LCO-3.0.2 Upon discovery of a failure to meet a Limiting Condition for Operation, the associated ACTIONS shall be met. Unless otherwise stated, only one Condition in each LCO's ACTIONS may be entered at any time. If the Limiting Condition for Operation is restored prior to expiration of the specified Completion Time (s), completion of the Required Act4on is not required, unless otherwise stated. LCO 3.0.3 When a Limiting Condition for Operation is not inet, and the
. associated ACTIONS are not met, or an associated ACTION is not provided, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. . Action shall be initiated within I hour to place the unit, as applicable, in:
[U~) MODE 2 within 7 hours, a.
- b. MODE 3 within 13 hours, and
- c. MODE 4 within 37 hours.
Exceptions to these requirements are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the Limiting Condition for Operation or ACTIONS, completion of the actions required by LCO 3.0.3 are not required. i LCO 3.0.3 is applicable in MODES 1, 2, and 3. t -t3 V (continued) ABWR 3-1 5/31/89
LCO Applicability 3.0 LIMITING CONDITIONS FOR OPERATION (LCOs) (continued) LC0 3.0.4 When a Limiting Condition for Operation is not met, entry into a MODE or other specified condition in the Applicability shall not be made unless the associated ACTIONS for the MODE or other specified condition to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This shall not prevent changes in MODES or other specified conditions in the Applicability which are required to comply with ACTIONS. Exceptions to LC0 3.0.4 are stated in the individual Specifications. These exceptions allow MODES or other specified conditions in the Applicability of the LC0 to be entered while meeting the associated ACTIONS which do not permit continued operation in the MODE or other specified condition for an unlimited period of time. LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control to perform required Surveillance to demonstrate its OPERABILITY or the OPERABILITY of other equipment. These Surveillance are only those which otherwise could not be performed without returning the equipment to service. LCO 3.0.6 When equipment is rendered inoperable for performance of its O LCO's Surveillance Requirements to satisfy SR 3.0.2, the Required Actions may be delayed for up to 8 hours to permit completion of the Surveillance Requirements. LCO 3.0.7 When a support system is inoperable and an LCO for that support system is specified in the Technical Specifications, the supported system is nut required to be declared inoperable solely due to support system inoperability. Only the support system LCO's ACTIONS are required to be entered. This is a clarification of the definition of OPERABILITY. When a support system is inoperable and there is not an LC0 for that support system specified in the Technical Specifications, the impact of the inoperability or degradation of the support system function on the OPERABILITY of the supported system shall be evaluated. Upon determination that the supported system is inopera'.,le, the ACTIONS of its LCO shall apply. (continued) ABWR 3-2 5/31/89 1
- _ _ = _ _ - - - - _ _ _ - - - _ _ .
LCO Applicability
. r'N O LIMITING CONDITIONS FOR OPERATION (LCOs) (continued)
LCO 3.0.8 Special Operations LCOs in Section 3.10 allow specified Technical Specification requirements to be changed to permit performance of special tests and operations. .Unless otherwise. specified, all other Technical Specification requirements remain unchanged. Compliance with Special Operations LCOs is - optional. When a Special Operation LCO is not desired to be met, entry into its MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. O O ABWR 3-3 5/31/89
SR Applicability
. 3.0 3.0 APPLICABILITY ,
SURVEILLANCE RE0VIREMENTS (SRs) SR 3.0.1 Surveillance Requirements shall- be met during the MODES or other specified conditions in the Applicability for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Satisfacto y performance of the Surveillance within the specified Frequency is required to verify the LCO is being met. Surveillance Requirements do not have to be met on inoperable equipment. SR 3.0.2 The specified Frequency for each Surveillance Requirement is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency as measured from the previous performance, or as measured from the time a specified i condition of the Frequency is met. For Frequencies specified as "once", the interval extension , above does not apply. If a Required Action's Completion Time requires periodic performance of "once per...," the Frequency extension above applies to the Completion Time intervals. Exceptions to these requirements are stated in the individual Specifications. ; l SR 3.0.3 When a Limiting Condition for Operation is not met due to failure to perform a Surveillance within the specified Frequency, the requirement to declare the equipment inoperable may be delayed for up to 24 hours from the time it is identified that the Surveillance has not been performed to permit the completion of the Surveillance. If the Surveillance is not performed within the 24 hour allowance, the Completion Time of the Required Actions begins immediately upon expiration of the 24 hour allowance. When the Surveillance is performed within the 24 hour allowance and the Surveillance Requirements are not met, the Completion Time of the Required Actions begins immediately upon the failure of the Surveillance. (continued) ABWR 3-4 5/31/89
SR Applicability 3.0 (. A.J 3.0 APPLICABILITY SURVEILLANCE REQUIREMENTS (SRs) (continued) . SR 3.0.4 Entry into a MODE or other specified condition in the Applicability shall not be made unless the Surveillance Requirements associated with a Limiting Condition for Operation have been met. This provision shall not prevent passage through, or to, MODES or other specified conditions as required to comply with the Required Actions. Exceptions to these requirements are stated in the individual Specifications.
.SR 3.0.5 Inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable.as follows:
- a. Inservice inspection of components, and inservice testing of pumps and valves, shall be in accordance with Section XI of the ASML Boiler and Pressure Vessel Code and applicable Addenda (the Code) except where relief has been requested pursuant to 10 CFR 50.55a(g)(6)(1).
- b. Test Frequencies specified in Section XI of the Code shall be applicable as follows:
Code Terminoloov Frecuency Weekly 7 days Monthly 31 days Quarterly or every 3 months 92 days Semiannually or every 6 months 184 days Every 9 months 276 days Yearly or annually 366 days Biannual or every 2 years 732 days
- c. The provisions of SR 3.0.2 are applicable to the above Code required Frequencies,
- d. Performance of Code inservice inspection and testing shall be in addition to the specified Surveillance Requirements.
1
- e. Nothing in the Code shall supersede the requirements of these Technical Specifications.
l l ABWR 3-5 5/31/89 _ _ _ _ _ - _ _________ _ L
SHUTDOWN MARGIN-9V. 3.1.1
.1 l' ,
3.1 REACTIVITY CONTROL SYSTEMS.
- 3.1.1 SHUTD0WN MARGIN LCO 3.1.1- ' SHUTDOWN MARGIN shall- be:
A. 10.38% Ak/k, with the highest worth control rod p31t analytically determined, QB B. 2 0.28% Ak/k, with the highest worth control rod pair determined by test. APPLICABILITY: MODES 1, 2, 3, 4, and S'. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-A. SHUTDOWN MARGIN less A.1 Restore SHUTDOWN MARGIN 6 hours A than specified in to required limits.
.'q) MODE 1 or 2.
B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. C. SHUTDOWN MARGIN less C.1 Fully insert all I hour than specified in insertable control rods. MODE 3. (continued) l ABWR 3.1-1 5/31/89
SHUTDOWN MARGIN 3.1.1 4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. SHUTDOWN MARGIN less D.1 Fully insert all I hour than specified in insertable control rods. MODE 4. M D.2 Ensure Secondary As soon as Containment is OPERABLE. practicable M I D.3 Ensure at least one SGTS As soon as subsystem is OPERABLE. practicable E , D.4 Ensure at least one As soon es Secondary Containment practicable Isolation Valve and-associated actuation ; instrumentation is ! OPERABLE in each associated penetration not isolated. (continued) i l l
)
l l l l O l ABWR 3.1-2 5/31/89 i
SHUTDOWN MARGIN 3.1 :.1 7 . G ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. SHUTDOWN MARGIN less E.1 Suspend CORE ALTERATIONS Immediately-than specified in except for control rod MODE 5. insertion. M E.2 Fully-insert all As soon as insertable control rods practicable in core cells containing one or more fuel assemblies. M E.3 Ensur~e Secondary As soon as Containment is OPERABLE. practicable - M E.4 Ensure at least one SGTS As soon as g subsystem is OPERABLE. practicable
'd '
g E.5 Ensure at least one As soon as Secondary Containment practicable Isolation Valve and - associated actuation instrumentation is-OPERABLE in each associated penetration not isolated.
/
1 ABWR 3.1-3 5/31/89 _ _ _ _ _ _ _ _____ __A
i SHUTDOWN MARGIN d
, 3.1.1 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY l SR 3.1.1.1 Demonstrate SHUTDOWN MARGIN. -----NOTE----- lf Only required after fuel movement or ;
control rod j replacement. j within the l RPV. j
........ ..... g
}
Once within i 4 hours after i criticality l ! i SR 3.1.1.2 Demonstrate SHUTDOWN MARGIN of each fuel Prior to movement during fuel loading sequence. fuel movement i CROSS-REFERENCES O TITLE NUMBER Reactor Protection System - Non Coincident Mode [ ] Secondary Containment Isolation Actuation Instrumentation [ ] Secondary Containment 3.6.4.1 Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 Control Rod Withdrawal - Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal - Refueling 3.10.5 O l ABWR 3.1-4 5/31/89 f __._______m___
- )
i
'~~
Control Rod OPERABILITY 3 3.1.2
, ):
3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 control Rod OPERABILITY LC0 3.1.2 All control rods shall be OPERABLE. APPLICABILITY: MODES 1 and 2.
-------------------------NOTE-------------------------
Conditions A through F may be concurrently applicable. ACTIONS CONDIT10N REQUIRED ACTION COMPLETION TIME A. One withdrawn control 611 ----------NOTE---------- rod stuck. A stuck control rod may be bypassed in RC&lS as allowed by LCO IRod Block Instrumentation, if required, to allow con-7S tinued operation. ( j m Restore stuck control I hour rod to OPERABLE status. B. Required Action and B.1 Disarm the associated I hour associated Completion control rod drive. Time of Condition A not met. AND B12 ----------NOTE---------- Not applicable when
> 10% of RTP.
Verify all inoperable 1 hour control rods not in compliance with GWSR are separated by 2 two OPERABLE control rods. A!!D
<'~x (continued)
-]
ABWR 3.1-5 5/31/89
i Control Rod OPERABILITY , 3.1.2 i ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME , l B. (continued) fL3 ----------NOTE----------
' Not applicable when s the LPSP of the RC&lS. l
........................ l l
Perform SR 3.1.2.2 24 hours j and SR 3.1.2.3 for i each withdrawn OPERABLE ! control rod. ;
-l AND j l
B.4 Demonstrate SHUTDOWN 72 hours j MARGIN is within the j limits of LC0 3.1.1. ! C. Required Actions and C.1 Be in MODE 3. 12 hours associated Completion Times of Condition B not met. I D. More than one D.1 Disarm the associated I hour ' withdrawn control rod control rod drive. stuck. AND D.2 Be in MODE 3. 12 hours (continued) 1 i O ABWR 3.1-6 5/31/89
i p i
) <
Control Rod OPERABILITY 1,q -3.1.2
' ACTIONS (continued)
-CONDITION REQUIRED' ACTION COMPLET10N TIME I i
E. Less 'than or equal' to E1 ----------HQIE----------
'eight control rods 1. Inoperable control inoperable for reasons rods may be bypassed
;other than Conditions in RC&IS as allowed A or D. by LCO [ Rod Block Instrumentation),
required, to allow insertion of inoperable control rod (s) and continued operation. 2 Inocerable control rods with failed motor drives can oniv be fully inser-ted by individual scram.
- ,-~s fully insert inoperable I hour
() control rod (s). 6HD E.2 Disarm the associated 2 hours control rod drive (s). AND g,3 .........-NOTE---------- Not applicable when -
> 10% of RTP.
Verify all inoperable 3 hours control rods not in compliance with GWSR are separated by 2 two OPERABLE control rods. (continued) n n, ABWR 3.1-7 5/31/89 L
Control Rod OPERABILITY .. 3.1.2
'ACTIO'NS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME O\ i F. Required Actions and F.1 Be in MODE 3. 12 hours associated Completion i Times of Condition E j not met. ! U 1 Greater than eight { inoperable control ] rods. j i l O O ABWR 3.1-8 5/31/89 l l
y l Control Rod OPERABILITY.; 3.1.2. ; SURVEILLANCE RE0VIREMENTS-SURVEILLANCE FREQUENCY-SR 3.1.2.1 Determine position of all control rods'. 24 hours SR '3.1.2.2 Insert each fully withdrawn control rod 7 days when at least one 1112 greater than-the LPSP-of the RC&IS SR 3.1.2.3 Insert each partially withdrbwn control 31 days when rod at least one 1t12 greater than the LPSP of the RC&lS U(M CROSS-REFERENCES TITLE NUMBER SHUTDOWN MARGIN 3.1.1 Rod Pattern Control 3.1.6 Control Rod Block Instrumentation [ ] i l l O l ABWR 3.1-9 5/31/89 . ! I
t-L Control Rod Scram Times [ 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod Scram Times LCO 3.1.3 All control rods shall have scram times less than or equal to the limits shown in Table 3.1.3-1. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ! A. One or more control A.1 Declare control rod (s) Immediately rods " slow" with scram with scram times times greater than. >[ ] seconds limits as shown in inoperable. Table 3.1.3-1. __ g A.2 Verify s 20% of the 12 hours OPERABLE control rods tested in SR 3.1.3.2 are
" slow". ;
M \ A.3 Verify s [ ] OPERABLE 12 hours control rods have scram times greater than limits. AND j i A4 Verify no more than two 12 hours OPERABLE " slow" control rods occupy adjacent locations. j B. Required Actions and B.1 Be in MODE 3. 12 hours associated Completion Times of Condition A . not met. I l
.- j O
ABWR 3.1-10 5/31/89
- l. '
s Control Rod Scram Times 3.1.3 M's_/.
. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-
............................yg35............................
During single or control rod 211r scram time tests, the CRD pumps shall be isolated from the associated scram accumulator. SR 3. I '. 3.1 Measure all control rod scram times -----NOTE-----
.with reactor steam dome pressure 2 950 Only required' psig. after fuel movement within the RPV or after l each reactor shutdown
> 120 days.
/}
(__/. Once' prior to exceeding 40% of RTP
'SR 3.1.3.2 Measure control rod scram times with 120 days of reactor steam dome pressure 2 950 psig cumulative for 210% of the control rods. operation in MODE 1 SR 3.1.3.3 Measure affected control rod scram times -----NOTE-----
at any reactor steam dome pressure. Only required . when work on ! control rod or CRD system could affect scram times. Once prior to declaring control rod (s) ' OPERABLE O (continued) ABWR 3.1-11 5/31/89
Control Rod Scram Times 3.1.3 SURVEILLANCE RE0VIREMENTS'(continued) SURVEILLANCE FREQUENCY , 1 1 SR 3.1.3.4 -----------------NOTE---------------------- l Performance of this SR satisfies SR 3.1.3.1 I and SR 3.1.3.3 for affected control rods. 1 Measure affected control rod scram times at -----NOTE----- ! reactor steam dome pressure it 950 psig. Only required I when work on l control rod or ! CRD system could affect j scram times. l Once prior to i exceeding 40% of RTP ; 1 CROSS-REFERENCES TITLE NUMBER
)
Control Rod OPERABILITY 3.1.2 i l O ABWR 3.1-12 5/31/89 l
1; l l-7 Control Rod Scram Times
'4 3.1.3-TABLE 3.1.3-1 (Page 1 of 1)
Control Rod Scram Times SCRAMTIMES")'I R0D (seconds) POSITION PERCENT REACTOR STEAM DOME PRESSURE (b) INSERTION O psig 950 psig 1050 psig 10 (c) [ ] [ 0.42 ] 40 (c) [ ] [ l'.00 ] 60 [ ] [ ] [ 1.44 ] O (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids as time zero. (b) For intermediate reactor steam dome pressures, the scram time criteria are determined by linear interpolation.
.(c) For reactor steam dome-pressure < 950 psig, only the 60% insertion scram time limit applies.
O { ABWR 3.1-13 5/31/89 l
1 Control Rod Scram Accumulators 3.1.4 3.1 REACTIVITY. CONTROL SYSTEMS
- 3. I'. 4 Control Rod Scram Accumulators LCO 3.1.4 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
......................ggg........................
Conditions A and B may be concurrently aDolicable. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME L One or more control A.1 Restore inoperable I hour scram accumulator (s) control rod scram inocerable, accumulator to OPERABLE status. 0.B A.2 Declare the associated I hour control rod (s) inoperable. E,. More than one inoo- B.1 Place the Reactor Mode 2 hours erable control rod Switch in the Shutdown scram accumulator position. associated with partially or fully withdrawn control rods. O ABWR 3.1 14 5/31/89
- r. ,
l { control Rod Scram Accumulators . 1' ou ( 'p) 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
.SR ,3.I'.4.1 Verify control rod scram accumulator 7 days pressure is t [ 1850 ) psig.
CROSS-REFERENCES TITLE NUMBER Control Rod OPERABILITY 3 .1 '. 2
, Control Rod Scram Times 3.1.3 i
O ABWR 3.1-15 5/31/89
l 1 i CRD Coupling l 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS l 3.1.5 control Rod Drive Couplina j i LCO 3.1.5 All control rods shall be coupled to their drive. APPLICABILITY: MODES 1 and 2. l 1 ACTIONS REQUIRED ACTION COMPLETION TIME CONDITION A. One or more control A.1 Recouple control rod. 2 hours rod (s) not coupled to its drive. QB A.2 Declare the control rod 2 hours inoperable. O O 3.1-16 5/31/89 ABWR
i CRD' Coupling .p ' 3.1 ~ .S .: C SURVEILLANCE REQUIREMENTS I SURVEILLANCE FREQUENCY SR 3.1.5.1- . Demonstrate each control rod does not go'to once the first 'i the overtravel position. time the ! control rod is withdrawn to " Full Out"- after fuel-movement within the RPV. 8!iQ Once prior to declaring . control rod (s) OPERABLE.when work on.
- control rod or CRD system-could affect' coupling.
CROSS-REFERENCES TITLE- NUMBER Control Rod OPERABILITY 3.1.2 i ABWR 3.1-17 5/31/89
Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS gg 3.1.6 Rod Pattern Control LC0 3.1.6 OPERABLE control rods shall comply with the requirements of the Gana Withdrawal Secuence Restrictions (GWSR). APPLICABILITY: MODES I and 2 with THERMAL POWER $ 10% of RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I A. Less than or equal to 621 ----------NOTE---------- eight OPERABLE control Affected control rods rods not in compliance may be bypassed in RC&IS with GWSR. as allowed by LCO [ Control Rod Block Instrumentation] Move affected control 8 hours rod (s) to ec* rect position. OR A.2 Declare affected control S hours { rod (s) inoperable. B2 More than eight B.1 Suspend withdrawal of Immediately ) OPERABLE control rods control rods. not in compliance with GWSR. AND i B.2 Place the Reactor Mode I hour Switch in the Shutdown position. ABWR 3.1-18 5/31/89 ,
0 l I Rod Pattern Control r;, 3.1.6-I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply. 24 hour's with QLSE. CROSS-REFERENCES
. TITLE NUMBER Control Rod OPERABILITY 3.1.2 Control Rod Block Instrumentation [ ]
Control. Rod Testing - Operating 3.10.7 v
.(
O : ABWR 3.1-19 5/31/89
l SLCS , 3.1.7 1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liouid Control System LCO 3.1.7 The Standby Liquid Control System (SLCS) shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLCS subsystem A.1 Restore inoperable 7 days from inoperable. subsystem to OPERABLE discovery of status. inoperable subsystem B. Both SLCS subsystems B.1 Restore at least one 8 hours inoperable. subsystem to OPJRABLE status, b1!D B.2 Restore the initial 7 days from inoperable subsystem discovery of to OPERABLE status. initial inoperable subsystem C. Required Actions and C.1 Be in MODE 3. 12 hours j associattd Completion j Times of Condition A 1 or B not met. ) i I i I O; I ABWR 3.1-20 5/31/89
SLCS 3.1.7
- k. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 2? hours pentaborate solution is 2 6103 gallons.
SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is within limits of Figure 3.1.7-1. SR 3.1.7.3 Verify temperature of pump suction piping 24 hours is within the limits of Figure 3.1.7-1. SR 3.1.7.4 Demonstrate by chemical analysis the 31 days concentration of boron in solution is within the limits of Figure 3.1.7-1. MLQ
~T Once within
() 24 hours after water or boron added to solution AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-1 SR 3.1.7.5 Verify each valve in flow path not 31 days locked, sealed or secured in position, is in its correct position. (continued)
/^T LJ ABWR 3.1-21 5/31/89
SLCS 3.1.7 SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.7.6- Demonstrate each pump develops a flow rate 92 days 2 50.0 gpm at a discharge pressure 2 1223 psig. SR 3.1.7.7 Demonstrate the flow path through one SLCS 18 months on subsystem from pump to reactor pressure a STAGGERED vessel is available by initiating .the TEST BASIS subsystem and pumping demineralized water to the reactor pressure vessel. SR 3.1.7.8 Demonstrate all heat-traced piping between 18 months storage tank and pump suction is unblocked. f.!!D Once within 24 hours ! after solution temperature is restored W within the limits of Figure 3.1.7-1 CROSS-
REFERENCES:
None O'; l ABWR 3.1-22 5/31/09 t . _ _ _ _ -_-
L;, , Il i. SLCS 3.1.7
'. (y' '
1.14 7, 1.12 - Go s 1.10 . SPECIFIC GRAVITY . 50
'f 30 C SATURATION o b
g y . TEMPERATURE *C g g I.0 8 - - 40 $u o
. . jm M 1.065' -
o. w r ) g
'V { l.06-
, 3g a
C l.04 - - 20 . i see -l 1.02- , ,, 13.4%
' ' ! ' '
- e i ,
.l.00 0 10 20 30 40 CONCENTRATION (t) (WEIGHT PERCENT SODIUM PENTA 80 RATE IN SOLUTION)
Figure 3.1.7-1 (Page 1 of 1)
. Sodium Pentaborate Solution Temperature Versus Concentration Requirements ABWR 3.1-23 5/31/89
(, APLHGR 3.2.1' 3.2
- G.
' POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LCO 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs).
shall be less than or equal to the limits specified in the
.[ CORE OPERATING. LIMITS REPORT).
APPLICABILITY: THERMAL POWER > 25% of RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR greater A.1 Restore APLHGR to 2 hours than-the required less than or equal to limits. the required limits. B. Required Action and B.1 Reduce THERMAL. POWER to 4 hours associated Completion < 25% of RTP.
. Time of Condition A not met.
l SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the required limits. 12 hours after 2 25% of RTP AND l 24 hours thereafter n CROSS-
REFERENCES:
None. U ABWR 3.2-1 5/31/89
h NCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO LC0 3.2.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the MCPR limit specified in the [ CORE OPERATING LIMITS REPORT). APPLICABILITY: THERMAL POWER > 25% of RTP. ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME A. MCPR less than the A.I Restore MCPR to greater 2 hours required limit. than or equal to the required limit. B. Requir d Action and B.I Reduce THERMAL POWER to 4 hours associated Completion < 25% cf RTP. Time af Condition A ' not met. ' SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify MCPR is greater than or equal Once within to the required limit. 12 hours after it 25% of RTP AN_Q 24 hours thereafter CROSS-
REFERENCES:
None O ABWR 3.2-2 5/31/89 - - _ _ _ - _ _ _ - _ _ _ . l
p LHGR 3.2.3 f] 3.2 POWER DISTRIBUTION LIMITS 3.2.3. LINEAR HEAT GENERATION RATE (Applicable to non-GE Fuel 0nlyl - LCO 3.2.3 The LINEAR HEAT GENERATION RATE (LHGR) shall be less than or equal to the limits specified in the [ CORE OPERATING LIMITS REPORT). i APPLICABILITY: THERMAL POWER > 25% of RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A .- Any LHGR greater than A.I Restore LHGR to less 2 hours the ' required limits. than or equal to the required limits. u B. Required Action and B.I Reduce THERMAL POWER to 4 hours associated Completion < 25% of RTP. fs f Time of Condition A N. not met. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to Once within the required limits. 12 hours after 2 25% of RTP AND 24 hours thereafter
.g ABWR 3.2-3 5/31/89
.++ A sl
' Recirculation Pumps Operating'-
PY 3.4.I
'34~ REACTOR COOLANT SYSTEM-3.'4.1 ' Recirculation Pumo's Doeratina.
LCO 3.4.1- At least nin'e reactor int'ernal oumos '(RIPS) shall be in operation. APPLICABI'LITY: ' MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION' COMPLETION. TIME.
......... g ....... .. _
Provisions of LCO 3.0.4
-are not aoolicable.
'A . One of the reauired. A.1 yerify reactor oower is I hour RIPS not in operation. < 95% RTP.
......... g ..........
Provisions of LCO 3.0.4 are not aoolicable. L Two of the reouired L1 Verify reactor oower is 1 hour RIPS not in operation. < 90% RTP. L Three or four of the C.1 Reduce reactor oower 4 hours i reauired RIPS not in to < 25% RTP. operation. AND C.2.1 Restore at least seven 12 hours from-RIPS to operation, initial dis-covery of less L. QB less than
- l. Ipven RIPS in ooeration.
C.2.2 Be in MODE 3. 24 hours from initial dis-covery of less (Qj than seven BJPs in oper-a+ ion. (Continued)
;,ABWR 3.4 1 5/31/89
_ _ ____ __ _ _ __-_____ _ _- D
i Recirculation Pumps Operating 3.4.1 ACTIONS (Continued) CONDITION REQUIRED ACTION COMPLETION TIME [L Less than five RIPS D.d Reduce reactor power 4 hours in operation, to < 5% RTP. SEE f i D.2.1 Restore at least seven 12 hours from RIPS to operation. initial dis-covery of less 4 QB _than seven RIPS in ooer-ation. D.2.2 Be in MODE 3. 24 hours from initial dis-covery of less than seven RIPS in operation. O O ABWR 3.4-2 5/31/89 l
l l Recirculation Pumps Operating 3.4.1 ]-
.](
SURVEILLANCE RE0'UIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.d Verify at least nine RIPS are in operation. 24 hours'
' CROSS-REFERENCES TITLE- NUMBER Reactor Coolant System Pressure / Temperature Limits 3.4.6 0
I ABWR 3.4-3 5/31/89
S/RVs 3.4.2 3.4 REACTOR COOLANT SYSTEM 3.4.2 Safety / Relief Valves LCO 3.4.2 The safety function of > 12 Safety / Relief Valves (S/RVs) shall be OPERABLE, APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 Be in MODE 3. 12 hours required S/RVs inoperable. Af!LQ , A.2 Be in MODE 4. 36 hours O O ABWR 3.4-4 5/31/89
'w S/RVs
,z j 3.4.2 SURVEILLANCE RE0VIREMENTS SURVEILLANCE -FREQUENCY SR 3.4.2.1- Demonstrate the safety function lift According to setpoints of the required S/RVs are as SR 3.0.5 follows: 28 Number of Setpoint 18 months S/RVs (osia) - 1 1150 + 11.5 2 1160 I 11.6 3 1170 I 11.7 3 1180 I 11.8 3 1190311.9 _ SR 3.4.2.2 Demonstrate each required S/RV opens when 18 months manually actuated. E
' (~l V 12 hours when reactor steam dome pressure is 2[ ] psig.
GOSS-REFERENCES , TITLE NUMBER ECCS - Operating 3.5.1 i N ABWR 3.4-5 5/31/89 i
i f.. i Operational Leakage i 3.4.3 3.4 REACTOR COOLANT SYSTEM 3.4.3 Doerational Leakace LCO 3.4.3 Reactor coolant system LEAKAGE shall be limited to: A. No Pressure Boundary Leakage, AND B. s I gpm total Unidentified Leakage averaged over any 24 hour period, a ENA C. s 25 gpm Total Leakage averaged over any 24 hour period ' APPLICABILITY: MODES 1, 2, and 3.
......................N0TE------------------------
Conditions A and B may be concurrently applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified Leakage A.1 Reduce leakage to within 4 hours
> 1 gpm. the limits.
M Total Leakage
> 25 gpm.
M ! Unidentified Leakage
> 1 gpm and Total Leakage > 25 gpm.
i (continued) e! ABWR 3.4-6 5/31/89 i
1 rc , l 4 [, j 1
, Operational Leakage 'l r~^J- 3.4.3 l ACTIONS (continuedi 1
CONDITION REQUIRED ACTION COMPLETION TIME . B. . Required Actions and B.1 Be in MODE 3. 12 hours associated Completion Times of Condition A 6!!D not met. B.2 Be in MODE.4. 36 hours DB i Any Pressure Boundary Leakage. SURVEILLANCE RE0VIREMENTS SURVEILLANCE - FREQUENCY L SR 3.4.3.1 Verify the reactor coolant system LEAKAGE 12 hours is less than or equal to the required limits. CROSS-
REFERENCES:
None i-ABWR 3.4-7 5/31/89
i Specific Activity 3.4.4 3.4 REACTOR COOLANT SYSTEM 3.4.4 Specific Activity LCO 3.4.4 The specific activity of the primary coolant shall be s 0.2 microcuries per gram DOSE EQUIVALENT I-131. APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary coolant A.1 Perform an isotopic Once per specific activity analysis for Iodine. 4 hours
> 0.2 but s 4.0 yCi per gram DOSE ASD ,
EQUIVALENT I-131. A.2 Restore specific 48 hours > activity to within limits. B. Required Actions and B.1 Perform an isotopic Once per associated Completion analysis for Iodine. 4 hours Times of Condition A not met. ANJ 08 B.2 Isolate all main steam 12 hours lines. Primary coolant specific activity
> 4.0 pCi per gram DOSE EQUIVALENT I-131. j O
ABWR 3.4-8 5/31/89 _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - )
o, 4 ,, . ,. i . I L , Specific Activity.
'3.4.4'
' SURVEILLANCE REQUIREMENTS SURVEILLAdE FREQUENCY-
. SR ' ' 3'.4.4.1 -------------NOTE-----------------
Only required in MODE 1. Demonstrate specific' activity of' primary' 31 days-1 coolant is s 0.2 pCi per gram DOSE EQUIVALENT I-131.
--CROSS-
REFERENCES:
None , l LO ' i: ABWR 3.4-9 5/31/89 i
RHR - Shutdown 3.4.5 3.4 REACTOR C0OLANT SYSTEM 3.4.5 Residual Heat Removal - Shutdown LCO 3.4.5 Two Residual Heat Removal (RHR) shutdown cooling subsystems shall be OPERABLE. APPLICABILITY: MODE 3 with reactor steam dome pressure < 135 psig, MODE 4 with heat losses to ambient not sufficient to maintain average reactor coolant temperature s 200*F. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One of the required A.1 Restore the required 2 hours from RHR shutdown cooling RHR subsystem to discovery of subsystems inoperable. OPERABLE status. inoperable subsystem B. Required Action and B.1 Provide an alternate As soon as associated Completion method capable of practicable Time of Condition A decay heat removal not met. for each required but inoperable subsystem. D.B AND l No RHR shutdown cooling subsystem jbl Restore to at least two As soon as OPERABLE. OPERABLE RHR subsystems. practicable I i l
)
l ABWR 3.4-10 5/31/89 .
o RHR - Shutdown SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify for the required RHR shutdown 31 days cooling subsystem (s) each manual, power operated, or automatic valve in the flow QB path, not locked, scaled or otherwise secured in position, is in the correct 12 hours when position or is capable of being reactor steam manually aligned in the correct position. dome pressure is < 135 psig i CROSS-REFERENCES TITLE NUMBER ECCS - Operating 3.5.1
,e m
( ) ECCS - Shutdown 3.5.2 j Residual Heat Removal Suppression Pool Cooling 3.6.2.3 l
,Q % ,Y ABWR 3.4 11 5/31/89
RCS Pressure / Temperature Limits 3.4.6 1 3.4 REACTOR COOLANT SYSTEM I 3.4.6 Reactor Coolant System Pressure / Temperature Limits ! l LC0 3.4.6 The Reactor Coolant System (RCS) temperature and reactor vessel pressure shall be maintained within the . Pressure / Temperature (PT) Limits. . APPLICABILITY: At all times. l i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i A. --------NOTE-------- A.1 Restore RCS temperature 30 minutes Required Actions A.1 and reactor vessel and A.2 must be. pressure to within the completed whenever PT Limits, i this Condition is entered. AND A.2 Determine RCS is 72 hours Operation outside the acceptable for continued PT Limits. operation. B. Required Actions and B.1 Be in MODE 3. '12 hours associated Completion Times of Condition A AND not met. B.2 Be in MODE 4. 36 hours l i i l ABWR 3.4-12 5/31/89 l I
l 1
,_ RCS Pressure / Temperature Limits / ) 3.4.6 w/
SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 -----------------NOTE----------------- Only required during system heatup, cooldown and inservice leak and hydrostatic testing. Verify RCS pressure and temperature are 30 minutes within the PT Limits. SR 3.4.6.2 Verify RCS pressure and temperature are Once within within the PT Limit Curve criticality 15 minutes limit. prior to initial control rod withdrawal for the purpose of ('~') achieving
\s_./ criticality SR 3.4.6.3 -----------------NOTE----------------- ----NOTE-----
Only required in MODES 1, 2, 3, and 4 Only reauired with reactor steam dome pressure when less than 2 25 psig. five RIPS are
...................................... in operation Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel coolant temperature is orior to each s 145'F. startuo of a reactor internal Dumo (continued) l 1
( 'r l
'\.-l ABWR 3.4-13 5/31/89
RCS Pressure / Temperature Limits 3.4.6 SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.6.4 Verify the reactor vessel flange and head 12 hours when flange temperature are it 70*F. in MODE 4 with reactor coolant system temperature s 100*F 8.NQ 30 minutes when in MODE 4 with reactor coolant system temperature s 80*F hhD 30 minutes when tensioning the reactor vessel head bolting studs CROSS-REFERENCES TITLE NUMBER Recirculation Pumps Operating 3.4.1 Inservice Leak and Hydrostatic Testing Operation 3.10.1 O ABWR 3.4-14 5/31/89
Reactor Steam Dome Pressure 3.4.7 ~ G. 3.4 REACTOR COOLANT SYSTEM 1 3.4.7 Reactor Steam Dome Pressure ] I i LC0 3.4.7 The reactor steam dome pressure shall be s 1040 psig. APPLICABILITY: H0 DES I and 2, except during anticipated transients. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Reduce reactor steam 15 minutes pressure dome pressure to
> 1040 psig. s 1040 psig.
B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify reactor steam dome pressure is 12 hours s 1040 psig. CROSS-
REFERENCES:
None O ABWR 3.4-15 5/31/89
i l ECCS - Operating 3 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS 3.5.1 ECCS - Operatina LCO 3.5.1 All ECCS injection subsystems of Divisions 1. 2_and 3 shall be OPERABLE, 8.NR 8 ADS valves shall be OPERABLE. APPLICABILITY.: MODE 1, MODES 2 and 3 except ADS and RCIC are not reauired to be OPERABLE with reactor steam dome teessure
< 50 osia for ADS and < 150 osia for RCIC.
ACTIONS i CONDITION _ REQUIRED ACTION COMPLETION TIME at One injection A.1 Restore inoperable sub- 60 days from subsystem inoDerable. system (s) to OPERABLE discovery of status. inoperable subsystem (s)
,f s s
(__/ Em Any two injection B.1 Restore at least one 14 days subsystems inoDerable. inoperable subsystem to OPERABLE status. AND B.2 Restore the initial 60 days from inoperable subsystem to discovery of OPERABLE status. initial inoperable subsystem (continued) i i 7-%
\~ / l 1
l ABWR 3.5-1 5/31/89
f ECCS - Operating 3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME L Any three iniection L1 Restore at least one 72 hours subsystems inoperable. inocerable subsystem to OPERABLE status. E L2 Restore at least two 14 days from inonerable subsystems discovery of to OPERABLE status. second inocerable E subsystem L3. Restore all inocerable 60 days from subsystems to discovery of OPERABLE status, initial inocerable subsystem D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Conditions A, E " B, or C not met. D.2 Be in MODE 4. 36 hours L One or two ADS valves L1 -------NQIE------------- inoperable. Provisions of LC0 3.0.4 are not aonlicable. Restore inoperable Prior to ADS valve (s) to startuo from OPERABLE status, any outaae of
>30 days duration AND Prior to startuo from next refuelina outaae (continued)
ABWR 3.5-2 5/31/89
h , I .: f w, ECCS - Operating'.
~ 3.5.1 ACTIONS (continued).
CONDITION REQUIRED ACTION COMPLETION TIME L L Three ADS valves L1' Restore at least 30' days inoperable. one of the inoperable
' valves to OPERABLE status.
4 L More than three G.1 Be in Mo'de 3. 12 hours
' ADS valves inoperable, b![Q DB G.2 ' Reduce reactor steam 36 hours dome pressure to Required Action and 5 50 psig.
associated Completion Time of Condition F not met. q
,D ._
O ABWR 3.53 5/31/89
( ECCS - Operating ! 3.5.1 l SURVEILLANCE Rt001REMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Demonstrate for each ECCS injection 31 days subsystem the system piping is filled j with water from the pump discharge i valve to the isolation valve. { SR 3.5.1.2 ------------------NOTE-------------------- LPFL subsystems may be considered OPERABLE during alignment to and operation in the RHR shutdown cooling mode when below 135 psig, if capable of being manually realigned and not otherwise inoperable. Verify for each ECCS injection subsystem 31 days each manual, power operated or automatic valve in the flow path not locked, sealed or otherwise secured in position is in its correct position. O SR 3.5.1.3 Verify Atmospheric Control System 31 days supply pressure to ADS valves 2 161 psig. SR 3.5.1.4 Demonstrate the following ECCS pumps develop According to the specified flow rate against a system SR 3.0.5 head corresponding to the specified reactor pressure: QR SYSTEM HEAD 92 days CORRESPONDING TO - - REACTOR SYSTEM FLOW RATE PRESSURE OF LPFL 2 4200 gpm 2 40 psig HPCF 2 800 gpm 2 1177 psig s (continued) O ABWR 3.5-4 5/31/89 j i
- _ _ - - - - _ - - - l
r! [3 N ./ ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.5 Demonstrate, with reactor pressure 92 days s 1177 psig, the RCIC pump can develop a flowrate 2 800 gpm E against a system head corresponding to a reactor pressure a 1177 psig. 12 hours when reactor steam dome pressure is 2 920 psig. SR 3.5.1.6 Demonstrate, with reactor pressure 18 months s 165 psig, the RCIC pump can develop a flow 2 800 gpm E against a system head corresponding _ to a reactor pressure h 165 psig. 12 hours when reactor
.O %/
steam dome pressure is 2 150 psig.
................N0TE-------------------
SR 3.5.1.7 Vessel injection may be excluded. Perform a system functional test 18 months for each ECCS injection subsystem including simulated automatic actuation of the system throughout its emergency operating sequence, to verify each automatic valve in the flow path actuates to its correct position.
................-N0TE--------------------
SR 3.5.1.8 Valve actuation may be excluded.
......................................... l i
Perform a system functional test for ADS, 18 months ] including simulated automatic actuation, ' C)
\s throughout its emergency operating sequence.
(continued) I l l ABWR _ 3.5-5 _ 5 _ _/31/_89__
- = - _ - h ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.9 Demonstrate each ADS valve opens when 18 months manually actuated at steam dome pressure 2 [ ] psig. QB 12 hours when reactor steam dome pressure 2[ } psig. CROSS-REFERENCES
~~
TITLE NUMBER ECCS Actuation Instrumentation [ ] Residual Heat Removal - Shutdown 3.4.5 Reactor Building Cooling Water / Reactor Building Service 3.7.1 Water - Operating Residual Heat Remeval Suppression Pool Cooling 3.6.2.3 O ABWR 3.5-6 5/31/89
i- l 1 I ECCS - Shutdown ! V(3 3.5.2 3.5- EMERGENCY CORE COOLING SYSTEMS 3.5.2 ECCS - Shutdown LCO 3.5.2 Two ECCS injection subsystems shall be OPERABLE. APPLICABILITY: MODE 4, MODE 5 except with the spent fuel pool gate removed and water level 2 23' over the top of the RPV flange. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One of the required A.1 Restore the required 4 hours from ECCS subsystems ECCS subsystems to discovery of inoperable. OPERABLE status, inoperable subsystem
) B. Required Action B.1 Suspend operations As soon as and associated Com- with a potential for practicable pletion Time of draining the reactor '
Condition A not met. vessel. C. Both of the required C.1 Suspend operations As soon as ECCS subsystems with a potential for practicable inoperable. draining the reactor vessel. AND C.2 Restore at least one 4 hours ECCS subsystem to OPERABLE status. (continued) O ABWR 3.5-7 5/31/89 ;
1 ECCS - Shutdown j
- 3. 5. 2 . j l
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l 4 D. Required Action C.2 D.I Suspend operations with As soon as . and associated a potential for drain- practicable Completion Time not ing the reactor vessel. met. M D.2 Ensure Secondary As soon as Containment is practicable OPERABLE. M D.3 Ensure at least one As soon as SGTS subsystem is practicable OPERABLE. M D.4 Ensure at least one As soon as Secondary Containment practicable Isolation Valve and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. l e ABWR 3.5-8 5/31/89
,t '
, a '
)
3 .' f :; ECCS - Shutdown 4 .
.- 3.5.2 SURVEILLANCE REQUIREMENTS
' SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify .for.each required LPFL 12 hours-subsystem the. suppression-pool water level t .is 214.63 ft.
SR 3.5.2.2 Verify'.for each required HPCF system the: A. CSP water level'2 22.8 ft. , 12 hours E B. Suppression pool water 12. hours , __ level.is 2 14.63 ft. t' SR 3.5.2.3 Demonstrate for each required ECCS -31 days-injection subsystem the system piping is filled with water from the pump discharge valve to the isolation valve. SR 3.5.2.4 ----------.--..-. NOTE--------------------- LPFL subsystems may'be considered 0PERABLE-during alignment to and operation in the RHR shutdown cooling mode if capable of being manually realigned and not otherwise inoperable. Verify for each required ECCS injection 31 days subsystem, each manual, power operated or automatic valve in the flow path not locked, sealed, or otherwise secured in position is in its correct position. l~ (continued)
.I ABWR- 3.5-9 5/31/89
t ECCS - Shutdown SURVEILLANCEREQUIREMENTS(continued) SURVEILLANCE FREQUENCY SR 3.5.2.5 Demonstrate each required ECCS pump According develops the specified flow rates against to SR 3.0.5 a system head corresponding to the specified reactor pressure: QB SYSTEM HEAD 92 days CORRESPONDING TO - - REACTOR SYSTEM FLOW RATE PRESSURE OF LPFL 2 4200 gpm 2 40 psig HPCF 2 800 gpm 2 1177 psig SR 3.5.2.6 ---------------NOTE--------------------
~~
Vessel injection may be excluded. Perform a system functional test for 18 months each required ECCS injection subsystem including simulated automatic actuation of the system throughout its emergency operating sequence, to verify each auto-matic valve in the flow path actuates to its correct position. O ABWR 3.5-10 5/31/89
k 1 p ECCS - Shutdown 3.5.2 (73j-CROSS-REFERENCES TITLE NUMBER ECCS Actuation Instrumentation [- ) ECCS - Operating 3.5.1 Residual Heat Removal - Shutdown 3.4.5 Secondary Containment 3.6.4.1
. Residual Heat Removal - High Water Level 3.9.8 Residual Heat Removal - Low Water Level 3.9.9 Reactor Building Cooling Water / Reactor Building Service 3.7.2 Water - Shutdown
-Standby Gas Treatment System 3.6.4.3 Secondary Containment Isolation Valves 3.6.4.2 l Secondary Containment Isolation Actuation Instrumentation [ ]
{ a I i l
)
l O l
.ABWR 3.5-11 5/31/89
L' . 4 l l
.i Primary Containment .j 7_ . 3.6.1.1
\) 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment LCO 3.6.1.1 The Primary Containment shall be OPERABLE.
.l APPLICABIlliv: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary Containment A.1 Restore Primary I hour inoperable. Containment to OPERABLE status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in H0DE 4. 36 hours v ABWR 3.6-1 5/31/89
Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS lh SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Demonstrate Drywell-to-Wetwell 18 months differential pressure does not decrease -- -- at a rate > [ ] inches water gauge AND per minute tested over a [ ] minute period at an initial differential pressure -----NOTE---- of [ ] psid. Only required after two l consecutive i tests fail until two consecutive tests pass. - 9 months SR 3.6.1.1.2 Perform required Type A leak rate -----NOTE----- testing in accordance with 10 CFR 50 Provisions of Appendix J and approved exemptions. SR 3.0.2 are not applicable. In accordance with 10 CFR 50 Appendix J and approved exemptions. SR 3.6.1.1.3 Perform required Type B leak rate testing -----NOTE----- except for Containment Air Locks in Provisions of accordance with 10 CFR 50 Appendix J and SR 3.0.2 are approved exemptions, not applicable. In accordance with 10 CFR 50 Appendix J and approved exemptions. ABWR 3.6-2 5/31/89 _
( i
,4 Containment Air Locks-
/ s: 3.6.1.2
< \
1 3.6 CONTAINMENT SYSTEMS 3.6.1.2 containment Air Locks LCO 3.6.1.2 Each Containment Air Lock shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.
' ACTIONS CONDITION REQUIRED ACTION. COMPLETION TIME A. One Containment Air -------------NOTE------ ------
Lock door inoperable. Provisions of LCO 3.0.4 are not applicable.
.gg ..............................
One Containment Air A.1 .----------NOTE----------
' Lock door and- Access allowed under associated interlock administrative control.
mechanism inoperable. ------------------------ [~) Close the OPERABLE door in the affected air lock. Immediately MD A.2 ----------NOTE---------- 1 Access allowed under administrative control, not to exceed one hour cumulative per year. Lock the OPERABLE air 24 hours lock door. 6ND A3 ----------NOTE---------- Access allowed under 1 administrative control, not to exceed one hour cumulative per year. Verify the OPERABLE door Once per in the affected air lock 31 days ' is locked closed. (} (continued) ABWR 3.6-3 5/31/89 L - - - - - - - - - - - - _ - - - - 1
I' Containment Air Locks 3.6.1.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME I B. Containment Air Lock -------------NOTES------------ interlock mechanism Access allowed under inoperable in une or administrative control. both Containment Air ------------------------------ Locks. B.1 Close one OPERABLE air Immediately lock door. M B.2 Lock the OPERABLE air 24 hours lock door. M B.3 Verify an OPERABLE door Once per in each affected air 31 days lock is locked clo:ed. C. One Containment Air C.1 Close one door in the Immed Dtely + Lock inoperable for affected air lock. reasons other than Condition A and B. M ; C.2 Restore inoperable air 24 hours lock to OPERABLE status. . I, (continued) O ABWR 3.6-4 5/31/89
p Cont'ainment Air Locks j'] 3.6.1.2
\ i . .
Y ~' ACTIONS fcontinued)- CONDITION REQUIRED ACTION COMPLETION TIME D. Condition A^ exists -------------NOTES------------ for both Containment 1. Provisions of LCO 3.0.4 Air Locks. are not applicable.
- 2. Access allowed under '
administrative control. l i D.1 Close one OPERABLE door Immediately l in both air locks. MD D.2 Lock one OPERABLE door . 24 hours closed in each air lock. MD D.3 - Restore one air lock 7 days n- to OPERABLE status. Y) E. One containment Air E.1 Perform Required In accordance Lock inoperable. Actions of Contidions A with for reasons of and B concurrently. Conditions A Condition A and other and B Containment Air Lock inoperable for. reasons of Condition B. F. Required Actions and F.1 Be in MODE 3. 12 hours associated Completion ,_ Times of Condition A, MQ l- 'B, C, D or L not met. F.2 Be in MODE 4. 36 hours l E Both Containment Air Locks inoperable for reasons other than Condition D or E. U.O ABWR. 3.6-5 5/31/89
7____ Containment Air Locks 3.6.1.2 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY-SR 3.6.1.2.1 Verify Containment Air Lock seal air flask 7 days pressure 2 [ ] psig. l SR 3.6.1.2.2 Demonstrate Containment Air Lock interlock -----NOTE----- mechanism is OPERABLE. Only required if not performed within previous 1 6 months Pr ar to entry into Primary Containment when Primary Containment is deinerted. SR 3.6.1.2.3 Demonstrate Containment Air Lock seal 18 months pneumatic system pressure does not decay at a rate > [ ] psig in [ ] hours from [ ] psig. SR 3.6.1.2.4 Perform required Type B containment Air -----NOTE----- Lock leak rate testing in accordance Provisions of with 10 CFR 50 Appendix J and approved SR 3.0.2 are ; exemptions. The acceptance criteria for not air lock testing are: applicable.
- a. Overall air lock leakage rate is !
s[ ] scfh when tested at Pa. In accordance with 10 CFR 50
- b. For each door, the seal leakage Appendix J and rate is s [ ] scfh when the gap approved between the door seals is pressurized exemptions to Pa.
CROSS-
REFERENCES:
None ABWR 3.6-6 5/31/89
Primary Containment Pressure 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Pressure LCO 3.6.1.3 The Primary Containment pressure shall be s 0.75 psig. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary Containment A.1 Restore pressure to I hour pressure within limits.
> 0.75 psig.
B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. O B.2 Be in MODE 4. 36 hours SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 Verify Primary Containment pressure is 12 hours s 0.75 psig. 1 CROSS-
REFERENCES:
None O ABWR 3.6-7 5/31/89
Drywell Average Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drwell Averace Air Temperature l LCO 3.6.1.4 Drywell average air temperature shall be s 135'F. APPLICABILITY: MODES 1, 2, and 3. ) j ACTIONS l l CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore average air 8 hours temperature temperature to within
> 135'F. limits. ,
B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A N A.ND not met. B.2 Be in MODE 4. 36 hours O\ < SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify Drywell average air temperature is 24 hours s 135'F. CROSS-
REFERENCES:
None O ABWR 3.6-8 5/31/89
i 1 Primary Containment and Pressure Isolation Valves ] (N 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Primary Containment and Pressure Isolation Valves LC0 3.6.1.5 The Primary Containment and Pressure Isolation Valves-shall be OPERABLE.
-APPLICABILITY: MODES 1, 2, and 3, MODES 4 and 5 when associated actuation instrumentation is required to be OPERABLE per LC0 (Primary Containment Actuation Instrumentation].
.........................N0TE-------------------------
Conditions A through D may be concurrently applicable. . ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME One or more required ----------NOTE---------- A. A.1 Primary Containment Not applicable to those 4 Isolation Valves or penetrations that have
only one isolation Pressure Isolation Valves inoperable. valve.
Verify at least one Immediately isolation valve is OPERABLE in each affected open-penetration. AND A.2.1 Restore the inoperable 8 hours valve (s) to OPERABLE status.
.QB (continued) l- ABWR 3.6-9 5/31/89
Primary Containment and Pressure Isolation Valves 3.6.1.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME i i A. (continued) A.2.2.1 ---------NOTE--------- Only applicable to Primary Containment Isolation Valves. Isolate each affected 8 hours penetration by use of at least one closed , and deactivated automatic valve, closed manual valve : or blind flange, en A.2.2.2 --------NOTE---------- )
- 1. Only applicable to Pressure Isolation Valves.
- 2. Check valves used to satisfy this Required Action must have been demonstrated to meet SR 3.6.1.5.10.
Isolate the high 8 hours pressure portion of the affected system ; from the low pressure ! portion by use of at least one closed manual or deactivated automatic or check valve. eE A.2.2.3 Verify each affected Once per penetration is 31 days isolated. (continued) O ABWR 3.6-10 5/31/89
1 l
'1.
Primary Containment and ' Pressure Isolation Valves l
' /7 3. 6.1. 5 ' -( ,) I ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME j
- B. Required Actions and B.1 Be in MODE 3. 12 hours j associated Completion Times of Condition A MD not met in MODE 1,'2, or 3. B.2 Be'in MODE 4. 36 hours C. Required Actions and C.1 Suspend CORE ALTERATIONS. Immediately associated Completion Times of Condition A not met in MODE 4 or S.
l (N v. l l k , ABWR 3.6-11 5/31/89 l L - _ _ - ---_-- _
Primary Containment and Pressure Isolation Valves 3.6.1.5 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 -----------------NOTE----------------- Valves and blind flanges in high radiation areas may be verified by use of administrative controls. Verify all manual valves and blind flanges 31 days l which are located outside the Primary j Containment and required to be closed : during accident conditions are closed. ! { , SR 3.6.1.5.2 Verify continuity of the Traversing 31 days l In-Core Probe System isolation valve j explosive charge. ;
)
l SR 3.6.1.5.3 Verify all manual valves and blind flanges -----NOTE----- which are located inside the Primary Only required Containment and required to be closed if not during accident conditions are closed. performed in the previous ; 92 days. ' Prior to entering MODE 2 or 3 , from MODE 4 l if Primary I Containment I was de-inerted l while in i MODE 4 ! 1 I (continued) {1 O l ABWR 3.6-12 5/31/89 i 1 l l
n f% Primary Containment and Pressure ' Isolation Valves
~
TV 3.6.1.5 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.5.4 -----------------NOTE----------------- MSIVs may be excluded. j
...................................... 1 Demonstrate isolation time of each According to automatic or power operated Primary SR 3.0.5 Containment Isolation Valve is within limits. QB 92 days [
SR 3.6.1.5.5 Demonstrate full closure isolation time of According to each Main Steam Line Isolation Valve is SR 3.0.5 from 3 to 5 seconds. r~3 92 days (/ - - i SR 3.6.1.5.6 Demonstrate each automatic Primary- 18 months Containment Isolation Valve actuates to its isolation position on a simulated automatic isolation signal. I SR 3.6.1.5.7 Demonstrate each reactor instrumentation 18 months line excess flow check valve actuates on-a simulated instrument line break to restrict flow to s 1 gph. SR 3.6.1.5.8 Remove and test the explosive squib from 18 months the Traversing In-Core Probe System on a isolation valves. STAGGERED TEST BASIS (continued) ! l O ! U ), ABWR 3.6-13 5/31/89 a
f Primary Containment and Pressure Isolation Values I 3.6.1.5 j SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.5.9 Demonstrate leakage rate through each 18 months Main Steam Line Isolation Valve for all four main steam lines is s 20 scf per hour when tested at [ ] psig. SR 3.6.1.5.10 -----------------NOTE----------------- Only required in MODE 1 or 2. Demonstrate the leakage rate for each 18 months Reactor Coolant System Pressure Isolation Valve is s 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm. SR 3.6.1.5.11 Demonstrate the combined leakage rate of 18 months 1 gpm times the total number of Primary Containment Isolation Valves in hydrostatically tested lines which penetrate the Primary Containment, is not exceeded when these isolation valves are tested at [ ] psig. SR 3.6.1.5.12 Perform required Type C leak rate testing -----NOTE----- in accordance with 10 CFR 50 Appendix J Provisions of and approved exemptions. SR 3.0.2 are not applicable. In accordance with 10 CFR 50 Appendix J and approved exemptions O ABWR 3.6-14 5/31/89 ______________a
.q Primary Containment; and Pressure Isolation' Valves . ;
3.6.1.5' j l CROSS-REFERENCES TITLE NUMBER'- Primary containment Isolation Actuation Instrumentation [ ]
't l
i i O AEWR 3.6-15 5/31/89 i l
Wetwell-to-Drywell VB ' 3.6.1.6 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Wetwell-to-Drywell Vacuum Breakers 9.o ! LCO 3.6.1.6 Seven Wetwell-to-Drywell Vacuum Breakers shall be j OPERABLE. ! AND All Wetwell-to-Drywell Vacuum Breakers shall be closed. I i q APPLICABILITY: MODES 1, 2, and 3. / ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Wetwell-to-Drywell A.1 Close the open vacuum 8 hours from Vacuum Breaker open. breaker, discovery of open vacuum breaker B. One required B.1 Restore the inoperable 7 days Wetwell-to-Drywell vacuum breaker to from Vacuum Breaker OPERABLE status, discovery of l inoperable for reasons inoperable other than Condition A. vacuum breaker I (continued) O ABWR 3.6-16 5/31/89
Wetwell-to-Drywell VB ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One Wetwell-to-Drywell C.1 Close the open vacuum 8 hours from Vacuum Breaker open. breaker, discovery of open vacuum breaker M MD C.2 Restore the inoperable 7 days One required vacuum breaker to from Wetwell-to-Drywell OPERABLE status. discovery of Vacuum Breaker inoperable inoperable for reasons vacuum breaker other than Condition A. D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, E B or C not met. D.2 Be in MODE 4. 36 hours O ABWR 3.C-17 5/31/89
Wetwell-to-Drywell VB 3.6.1.6 SURVEILLANCE RE0VIREMENTS SR 3.6.1.6.1 Verify all vacuum breakers are closed. 14 days
- i
.I SR 3.6.1.6.2 Perform a functional test of each required 31 days vacuum breaker.
SR 3.6.1.6.3 Demonstrate the full opening setpoint of 18 months each required vacuum breaker is 1 0.5 psid. CROSS-
REFERENCES:
None i Ol i I O; 1 ! ABWR 3.6-18 5/31/89 l i l
- g. .) Suppression Pool Average-Temperature 1 3.6.2.1
-3.6; CONTAINMENT' SYSTEMS' 3.6.2.1 Sunoression Pool Averace Temperature-LC0 3.6.2.1 Suppression Pooi average _ temperature shall be:
L s: 95'F when Ireactor oower is > 1% of RTP1 and testing which adds heat to the Suppression Pool is not_being performed. L $ 105'F when Ireactor oower is > 1% of RTP1 and testing which adds heat-to the Suppression Pool is-being performed. L s 110*F when f reactor oower is < 1% of RTPl. d APPLICABILITY: MODES 1, 2, and 3..
-ACTIONS CONDITION. REQUIRED ACTION COMPLETION TIME V -b_,. Suppression Pool. A.1 Verify average Once per hour average temperature temperature s 110*F.
> 95'F but s Il0*F. 6.ND bhD A.2 Restore average 24 hours-IReactor oower is > l*4 temperature to of RTPl. 5 95'F.
b.EQ ' Not performing testing which adds heat to the Suppression Pool. B. Required Actions and Bd Reduce power to I< l's 12 hours associated Completion of RTPl. Times of Condition A not met. l (continued) ABWR 3.6-19 5/31/89 1
i i Suppression Pool Average Temperature l 3.6.2.1
]
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME J C. Suppression Pool C.1 Suspend all testing Immediately I average temperature which adds heat to the
> 105'F. Suppression Pool.
LND Performing testing which adds heat to the Suppression Pool. D. Suppression Pool D.1 Place the Reactor Mode Immediately average temperature Switch in the Shutdown
> 110*F but position.
s 120*F. D.2 Verify average Once per temperature s 120'F. 30 minutes E. Suppression Pool E.1 Depressurize the reactor 12 hours average temperature vessel to
> 120'F. < 200 psig.
AND E.2 Be in MODE 4. 36 hours l l O ABWR 3.6-20 5/31/89 i _ _ _ .___ -
1; Suppression Pool Average Temperature l./O 3.6.2.1 lQ ~ l '- SURVEILLANCE REQUIREMENTS l' l SURVEILLANCE FREQUENCY
- SR 3.6.2.1.1 Verify the Suppression Pool average 24 hours I temperature is within the applicable limits. AND 5 minutes when performing tests which add heat to the Suppression Pool CROSS-
REFERENCES:
None O 1 l 1 O ABWR 3.6-21 5/31/89 l ___ _ -_ _ __ _ a
Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression Pool water level shall be maintained from 22.97' to 23.29'. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression Pool water A.1 Restore water level to 4 hours level < 22.97'. within limits. 0.8 suppression Pool water level > 23.29'. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 4. 36 hours O ABWR 3.6-22 5/31/B9
F
Suppression Pool Water Level A. '
3.6.2.2.-
'l N}- -
SURVEILLANCE REQUIREMENTS SURVEILLANCE. . FREQUENCY: SR 3.6.2.2.1 Verify Suppression Pool water level is 24 hours from 22.97to 23.29'.
- L _
CROSS-REFERENCES-TITLE NUMBER ECCS - Operating- ~3.5.1 O O ABWR 3.6-23 5/31/89
l RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS g ' 3.6.2.3 Residual Heat Removal SuDDreSSion Pool Coolina LCO 3.6.2.3 The Residual Heat Removal (RHR) Suppression Pool Cooling system shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. l ACTIONS j CONDITION REQUIRED tCTION COMPLETION TIME i A. One RHR Suppression A.1 ----------NOTE---------- Pool Cooling subsystem Provisions of LC0 3.0.4 inoperable. are not applicable. Restore inoperable 30 days from , subsystem to OPERABLE discovery of i status, inoperable subsystem p_. Two RHR Suppression B.1 Restore at least one 7 days from O Pool Cooling subsystems of the inoperable sub- discovery iroperable. systems to OPERABLE of second status, inoperable subsystem. A@ B.2 Restore the initial 30 days from inoperable subsystem to discovery of OPERABLE status.. initial inoperable subsystem (continued) O ABWR 3.6-24 5/31/89 i
i 7s RHR Suppression Pool Cooling g ) 3.6.2.3 L_.) ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME [ [2 All three RHR Sucores- [1 Restore at least one 8 hours j sion Pool Coolina subsystem to OPERABLE subsystems inoDerable, status. AND (12 Restore to at least two 7 days from OPERABLE subsystems. discovery of second inonerable subsystem AND [21 Restore the initial 30 days from inocerable subsystem to discovery of OPERABLE status. initial inoperable subsystem (%.j D) D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, B AND or C not met. D.2 Be in MODE 4. 36 hours
,m N
i ABWR 3.6-25 5/31/89 l
RHR Suppression Pool Cooling i SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY j SR 3.6.2.3.1 Verify each manual, automatic, or power 31 days ! operated valve in the flow path not locked, sealed or otherwise secured in position, I is in its correct position or can be I aligned to its correct position. , I SR 3.6.2.3.2 Demonstrate each RHR pump develops a flow 92 days rate 2 4200 gpm through the associated RHR heat exchanger while operating in QB the Suppression Pool Cooling mode. i According to SR 3.0.5 l CROSS-REFERENCES TITLE NUMBER Residual Heat Removal System - Shutdown 3.4.5 ECCS - Operating 3.5.1 Residual Heat Removal Wetwell Spray 3.6.2.4 4 O l ABWR 3.6-26 5/31/89
,.___ - _-_ -__ ___ ____- _ -__-___ __ __._ _ __ _ _ - = _ - - _ . . _ _ _ - _ ._ _
N<' RHR Wetwell Spray.
. .j ; 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residual' Heat Removal Wetwell Soray LCO 3.6.2.4 The Residual Heat Removal (RHR) Wetwell Spray system'shall !
be OPERABLE. l APPLICABILITY: . MODES 1, 2, and 3. v ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-a A. One RHR Wetwell Spray A.1 ----------NOTE---------- subsystem inoperable. Provisions of LCO 3.0.4 are not applicable. Restore . inoperable 7 days.from subsystem to OPERABLE discovery of. status. . inoperable subsystem O B. Both RHR Wetwell Spray B.1 Restore at least one 8 hours-
.. subsystems inoperable. subsystem to OPERABLE status.
AN_Q B.2 Restore the initial 7 days from inoperable subsystem to discovery of OPERABLE status. initial inoperable subsystem f C. Required Actions and C.1 Be in MODE 3. 12 hours associated Completion Times of Condition A AND or B not met. C.2 Be in MODE 4-. 36 hours O ABWR 3.6-27 5/31/89 I
RHR Wetwell Spray 3.6.2.4 SURVEILLANCE RE0VIREMENTS -- SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each manual, automatic, or power 31 days operated valve in the flow path is not locked, sealed or otherwise secured in position, is in its correct position or can be aligned to its correct position. SR 3.6.2.4.2 Demonstrate each RHR pump develops a flow 92 days rate 2 4200 gpm through the associated RHR heat exchanger while operating in the combined Sunoression Pool Coolino/ Wetwell Sorav Mode. CROSS-REFERENCES TITLE NUMBER O Residual Heat Removal - Shutdown 3.4.5 ECCS - Operating 3.5.1 Residual Heat Removal Suppression Pool Cooling 3.6.2.3 O ABWR 3.6-28 5/31/89
4- -g k n . Hydrogen Recombiner System 3.6.3.1 3.6' CONTAINMENT SYSTEMS < 3.6.3.1- Hydrocen Recombiner System
~
LCO 3.6.3.1 The Primary Containment' Hydrogen Recombiner System.shall be OPERABLE. APPLICABILITY: . MODES 1 and 2. ACTIONS-CONDITION . REQUIRED ACTION COMPLETION TIME fLATER). ' d i l l O ABWR 3.6-29 5/31/89
t Primary Containment Oxygen Concentration 3.6.3.2 1 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxycen Concentration LC0 3.6.3.2 The Primary Containment oxygen concentration shall be s 4% by volume. APPLICABILITY: MODE 1 with THERMAL POWER ;t 15% of RTP for > 72 hours, i MODE 1 when time. remaining with THERMAL POWER > 15% of RTP 4 prior to the next scheduled shutdown is > 72 hours. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary Containment A.1 Restore oxygen 24 hours oxygen concentration concentration to
> 4% by volume. s 4% by volume.
B. Required Action and B.1 Reduce THERMAL POWER to 8 hours associated Completion s 15% of-RTP. Time of Condition A not met. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 Verify Primary Containment oxygen 7 days i concentration is s 4% by volume. 1 CROSS-
REFERENCES:
None O ABWR 3.6-30 5/31/89 1
[ L . . Secondary Containment l r-~s 3.6.4.1
.tj 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1' The. Secondary Containment shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3, . . When handling irradiated fuel in the Secondary Containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).
.....................----N0TE-------------------------
Conditions A, B :nd C may be concurrently ' applicable. ACTIOt. CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary Containment A.1 Restore Secondary 4 hours r~v inoperable in Containment to OPERABLE (j. MODE 1, 2, or 3. status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A eg not met. B.2 Be in MODE 4. 36 hours (continued) O
)
ABWR 3.6-31 5/31/89 q
i Secondary Containment 3.6.4.1 l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary Containment C.1 ----------NOTE---------- inoperable when Provisions of LCO 3.0.3 I handling irradiated are not applicable. fuel in the Secondary ------------------------ Containment, during CORE ALTERATIONS, or Suspend handling of Immediately during OPDRVs. irradiated fuel in the Secondary Containment. E C.2 Suspend CORE ALTERATIONS. Immediately a C.3 Suspend OPDRVs. As soon as , practicable O; 1 l O ABWR 3.6-32 5/31/89
Secondary Containment 3.6.4.1
.f'VT SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify. Secondary Containment vacuum is. 24 hours 2 0.25 inches of vacuum water gauge.
SR 3.6.4.1.2 Verify all Reactor Building equipment 31 days hatches and blowort panels are closed and sealed. SR 3.6.4.1.3 Verify each' Reactor Building access door 31 days is closed, except for routine entry and exit. . SR 3.6.4.1.4 Demonstrate one Standby Gas Treatment 18 months-Subsystem will draw down the Secondary on a STAGGERED
.. Containment to 1 0.25 inches of TEST BASIS ' /N. vacuum water gauge in s [ ] seconds.
U-SR 3.6.4.1.5 Demonstrate one Standby Gas Treatment 18 months Subsystem can maintain 2 [ ] inches on a STAGGERED of vacuum water gauge-in the Secondary TEST BASIS Containment for one hour at a flow rate
< 1050 cfm.
CROSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] Secondars .ontainmvn Isolation Valves 3.6.4.2 l Standby Gas Treatment System 3.6.4.3 L 1 (d~' ABWR 3.6-33 5/31/89
i i Secondary Containment Isolation Valves
' j 3.6.4.2 3.6 CONTAINMENT SYSTEMS
]
3.6.4.2 Secondary Containment Isolation Valves LC0 3.6.4.2 The Secondary Containment Isolation Valves shall be , OPERABLE. ! APPLICABILITY: MODES I, 2, and 3, When handling irradiated fuel in the Secondary Containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ,
........................-N0TE------------------------- l Conditions A, B and C may be concurrently applicable.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 ----------NOTE---------- Sc-condary Containment Not applicable to those Isolation Valves penetrations that have- & inoperable, only one isolation W valve. ; Verify at least one Immediately isolation valve is OPERABLE in each affected open penetration. bND A.2.1 Restore the inoperable 8 hours valve (s) to OPERABLE status. ; 93 ' (contired) O ABWR 3.6-34 5/31/89
fd M , 4
,et y pg "
- _ w Secondary Containment Isolation Valves'-
0A 3.6.4.2
'i l ACTIONS (continued)
CONDITION' REQUIRED ACTION- COMPLETION TIME: o
~
A. '(continued)_ A.2.2.1 Isolate. each 'affected : 8 h'ours penetration by'use of-at least one closed. and' deactivated
-automatic valve, closed. manual valve or blind flange.
AN.D L A.2.2.2 Verify each affected '
'Once per penetration is 31 days isolated.
- B. . Required Actions and B.1 Be in' MODE 3. 12. hours?
, . associated. Completion .
Times of Condition A A.ND not met'in-MODE 1, . 2, or 3. B.2 Be in MODE 4. 36 hours. C
~
C. Required Actions and= 0.1 ----------NOTE---------- associated Completion Provisions of LCO 3.0.3 Times.of. Condition A are not ~ applicable. not met when handling ------------------------- irradiated fuel in
- the Secondary Suspend handling of Immedia+ely Containment, during irradiated fuel in the-CORE-ALTERATIONS, Secondary Containment.
or during OPDRVs. hN.D C.2 Suspend CORE ALTERATIONS. Immediately AND C.3 Suspend OPDRVs. As soon as practicable
/
,I ABWR- 3.6-35 5/31/89 '
Secondary Containment Isolation Valves SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 Verify all Reactor Building blind flanges 31 days or rupture discs which are required to be closed during accident conditions are 4 closed, i SR 3.6.4.2.2 Demonstrate the isolation time of each According to Secodary Containment Isolation Valve SR 3.0.5 is witiin limits. E i 92 days i 1 SR 3.6.4.2.3 Demonstrate each Secondary Containment 18 months l Isolation Valve actuates to its isolation ' position on a simulated automatic isolation signal. Ol
.(ROSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] l O
ABWR 3.6-36 5/31/89 i
1 l i Standby Gas Treatment System 3.6.4.3
'n) v' 3.6 CONTAINMENT SYSTEMS !
3.6.4.3 Standby Gas Treatment System LC0 3.6.4.3 The Standby Gas Treatment System (SGTS) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, When handling irradiated fuel in the Secondary Containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).
-------------------------NOTE-------------------------
Conditions A through E may be concurrently applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME L. One SGTS subsystem A.1 Restore inoperable 7 days from n inoperable for subsystem to OPERABLE discovery of
' % reasons other than status. inoperable V Condition C. subsystem L. Both SGTS subsystems B.1 Restore at least one 4 hours inoperable in MODE 1, subsystem to OPERABLE 2, or 3 for reasons status.
Other than Condition C. AN.D B.2 Restore the initial 7 days from inoperable subsystem discovery of to OPERABLE status. initial inoperable subsystem
=
L The SGTS common filter L1 Restore the common 4 hours train inocerable in filter train to M0_DE 1. 2. or 3. OPERABLE status. (continued) l U ABWR 3.6-37 5/31/89
Standby Gas Treatment System 3.6.4.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION IIME D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, B AND or C not met in MODE 1, 2, or 3. D.2 Be in MODE 4 36 hours L Required Action and E.1 ----------NOTE---------- associated Completion Provisions of LCO 3.0.3 Time of Condition A are not applicable. not met when handling ------------------------ irradiated fuel in the Secondary Containment, Suspend handling of Immediately during CORE ALTERATIONS, irradiated fuel in the or during OPDRVs. Secondary Containment.
% END Both SGTS subsystems E.2 Suspend CORE ALTERATIONS. Immediately i inoperable when handling irradiated atQ fuel in the Secondary Containment, during E.3 Suspend OPDRVs. As soon as CORE ALTERATIONS, practicable or during OPDRVs.
M The common filter train inocerable when handlina irradiated fuel in the Secondary Containment, durina CORE ALTERATIONS, or durina OPDRVs. I ) 1 e 1 ABWR 3.6-38 5/31/89 J
1 n:: I Standby Gas Treatment System .i 73 3.6.4-3 .
'- ' SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR '3.6.4.3.1 Demonstrate each SGTS subsystem operates. 31~ days stn_a with flow through the common filter train STAGGERED TEST for ;t 10 hours with heaters on. BASES.
SR 3.6.4.3.2 Demonstrate < 0.05% penetration of the 18 months SGTS HEPA filters by a D0P test at a system flow rate of 1050 to 1200 cfm. AE Once within 7 days. after painting, fire or chemical release in filter service area AND Prior to declaring subsystem OPERABLE after each complete or partial replacement of filter (continued) l O ABWR 3.6-39 5/31/89
)
L__ .-.__ _
.j
Standby Gas Treatment System 1
~
3.6.4.3 SURVEllt.ANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.4.3.3 Demonstrate < 0.05% bypass leakage 18 months through the SGTS charcoal absorber section by a halogenated hydrocarbon test at a AND system ficw rate of 1050 to 1200 cfm. Once within 7 days after painting, fire or chemical release in filter service area AND Prior to declaring subsystem OPERABLE after each complete or partial replacement of adsorber bank. l (continued) O ABWR 3.6-40 5/31/89
)
7_ l Standby Gas Treatment System
,- 3.6.4.3 SURVEILLANCE RE0VIREMENTS (continued)- j SURVEILLANCE FREQUENCY.
SR 3.6.4.3.4 -----------------NOTE----------------- Analysis to be completed within 31 days of sampling. Remove and perform a laboratory analysis ZZ.Q hours of a SGTS charcoal adsorber sample for of charcoal methyl iodide penetration. adsorber operation M 18 months M Once within 7 days after
- painting, fire or
-:[-}e.
^~ chemical release in filter service area-SR 3.6.4.3.5 Demonstrate < [ ] inches water gauge 18 months pressure drop across the combined SGTS HEPA filters and charcoal adsorber banks at a flow rate of 1050 to 1200 cfm.
SR 3.6.4.3.6 Perform a system functional test of each 18 months SGTS subsystem which demonstrates filter train startup and isolation dampers opening upon receipt of a simulated automatic initiation signal. SR 3.6.4.3.7 Demonstrate each SGTS heater dissipates 18 months from [ ] to [ ] kw.
. _ = -
ABWR 3.6 41 5/31/89
Standby Gas Treatment System 3.6.4.3 CROSS-BEFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] l 9l O ABWR 3.6-42 5/31/89 _ - _ - - _ _ _ _ _ - - - - - - - l
RCW/RSW System - Operating (~T 3.7.1 (s / -3.7 PLANT SYSTEMS 3.7.I Reactor Buildina Coolina Water (RCW)/ Reactor Buildina Service Water (RSW) System - Operatina LCO 3.7.1 The Divisions I. II. and III Reactor Buildina Coolina_ Water (RCW)/ Reactor Buildina Service Water (RSW) systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.
.................__....g91g...........................
Conditions A throuah D may be concurrent 1v aoolicable. ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME at -----HQlE---------- 611 Declare affected system Immediate1v Each of the three or comoonent inoperable. divisions may be in
.this Condition con-A() . currently.
QB , One RCW oumo and/or A2 ----------HQIl---------- one RSW oumo and/or Provisions of LCO 3.0.4 one RCW/RSW heat are not aoolicable. exchanaer inoperable ------------------------ in the same division. Restore the inoperable 60 days RCW/RSW component (s) from dis-to OPERABLE status, covery of inoperable comoonent. One RCW/RSW Division Declare affected system Immediately R. H11 inoperable for or component inoperable. reasons other than Condition A. QB B22 Restore the inocerable 7 days from RCW/RSW Division to discovery of OPERABLE statu1 inoperable . RCW/RSW l Division (Continued) ABWR 3.7 1 5/31/89
RCW/RSW System - Operating 3.7.1 ACTIONS (continued) CONDITION REQUIRED ACTION TIME L Two RCW/RSW Divisions L1 Declare affected system or Immediately inoperable for comoonent inoperable. , reasons other than Condition A. QR l C 2.1 Restore one inoperable 24 hours RCW/RSW Division to OPERABLE status. < AND C.2.2 Restore the initial 7 days from inoperable RCW/RSW discovery of Division to OPERABLE initial status. operable RCW/RSW Division. D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, MLQ B or C not met. D.2 Be in MODE 4. 36 hours i l ABWR 3.7-2 5/31/89
RCW/RSW System.-' Operating J. 3.7.1 7
\' SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify [ ultimate' heat sink) water level is 24 hours 2[ ] inches.
SR 3.7.1.2 Verify the [ ultimate heat sink) water 24 hours. temperature is s 95'F SR 3.7.1.3 Verify water' level in the RSW pump well of 14 days the intake structure is 2 [ ] feet. AND 12 hours when pump well level is s [ ] feet. V SR 3.7.1.4 Verify for each required RCW/RSW subsystem 31 days each manual, power operated or automatic valve in RCW/RSW flow paths servicing safety related systems or components not locked, sealed or otherwise secured in position is in its correct position. SR 3.7.1.5 Demonstrate each fultimate heat sink 31 days active comoonentl operates for 2 15 minutes. SR 3.7.1.6 Perform a system functional test for 18 months each required RCW/RSW subsystem including simulated automatic safety related functioning of the system. O l ABWR 3.7-3 5/31/89
RCW/RSW System - Operating 3.7 1 CROSS-REFERENCES TITLE NUMBER RHR - Shutdown 3.4.5 ECCS - Operating 3.5.1 AC Sources - Operating 3.8.1 O O ABWR 3.7-4 5/31/89 i _____ __ __ _______ _ _ -
RCW/RSW System - Shutdown l 3.7.2 i
) 3.7 PLANT SYSTEMS j 1
3.7.2 Reactor Buildina Coolina Water (RCW)/ Reactor Buildina Service Water ! (RSW) System - Shutdown I LCO 3.7.2 The Divisions I, II, and III Reactor Building Cooling Water (RCW)/ Reactor Building Service Water (RSW) systems associated with systems and components required to be OPERABLE shall be OPERABLE. 4 APPLICABILITY: MODES 4 and 5, When handling irradiated fuel in the secondary containment. ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. Required RCW/RSW A.1 Declare affected systet. Immediately Division inoperable. or component inoperable. I f~'
,,/
g) C/ ABWR 3.7-5 5/18/89
RCW/RSW System - Shutdown [ t 3.7.2
' SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the [ ultimate heat sink) water level 24 hours is 2 [ ] inches.
SR 3.7.2.2 Verify the [ ultimate heat sink) water temperature is s 95'F 24 hours SR 3.7.2.3 Verify water level in the RSW pump well of 14 days intake structure is 2 [ ] feet. AND 1 12 hours when pump well level is s [ ] feet. SR 3.7.2.4 Verify for each required RCW/RSW subsystem 31 days each manual, power operated or automatic valve in RCW/RSW flow paths servicing safety related systems or components not locked, sealed or otherwise secured in position is in its correct position. SR 3.7.2.5 Demonstrate each fultimate heat sink 31 days active comoonentl operates 2 15 minutes. ' SR 3.7.2.6 Perform a system functional test for 18 months each required RCW/RSW subsystem, including simulated automatic safety related functioning of the system. f 9 ABWR 3.7-6 5/18/89
$ i W ' 1 RCW/RSW System - Shutdown I, .
3.7.2 CROSS-REFERENCES TITLE NUMBER RHR - Shutdown 3.4.5 ECCS - Shutdown 3.5.2- , AC Sources - Shutdown .3.8.2-RHR - High Water level 3.9.8 e RHR - Low - Water level 3.9.9 I ABWR 3.7-7 5/18/89
- l
1 CR HVAC ER System
]
l 3.7.3 1 3.7 PLANT SYSTEMS 3.7.3 Control Room HVAC Emercency Recirculation System
'h '
LCO 3.7.3 The Control Room HVAC Emergency Recirculation (CRHER) system shall be OPERABLE. APPLICABILITY: MODES !, 2, 3, 4, and 5, When handling irradiated fuel in the secondary containment.
.........................N0TE-------------------------
Conditions A, B and C may be concurrently applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRHER subsystem A.1 Restore the inoperable 7 days from inoperable. subsystem to OPERABLE discovery of status, inoperable subsystem B. Required Action and B.1 Be in MODE 3. 12 hours O associated Completion Time of Condition A N 6ND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours DB Both CRHER subsystems inoperable in MODE 1, 2, or 3. (continued) O ABWR 3.7-8 5/18/89
1
'L se CR HVf.; ER System j Ds 3.7.3 j V ACTIONS (continued)-
j
)
CONDITION REQUIRED ACTION COMPLETION TIME ] l C. Required Action and C.1 Place OPERABLE subsystem Immediately associated Completion in the isolation mode of Time of Condition A operation. not met in MODE 4, 5, or when handling . QB irradiated fuel in the secondary containment. C.2.1 Suspend CORE ALTERATIONS. Immediately 08 A!iD Both CRHER subsystems C.2.2 ----------NOTE---------- inoperable in MODE 4, Provisions of LCO 3.0.3 5, or when handling are not applicable. irradiated fuel in the ------------------------ secondary containment. Suspend handling of Immediately irradiated fuel in the secondary containment. AND
+- C.2.3 Suspend operations with As soon as a potential for draining practicable the reactor vessel.
i ABWR 3.7-9 5/18/89
q CR HVAC ER System ! 3.7.3 j SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY j l i j SR 3.7.3.1 Demonstrate subsystem operates with flow 31 days I through the HEPA filters and charcoal adsorbers for 2 10 hours with heaters on. SR 3.7.3.2 Demonstrate < 0.05'/. penetration of the 18 months HEPA filters by a DOP test at a system flow rate of [ ] to [ ] cfm. AND Once within 7 days l! after I painting, fire, or chemical release in filter j service area htLQ Prior to declaring subsystem OPERABLE after each complete or partial repl acement - of a filter (continued) { l i l i l 1 i l I O ABWR 3.7-10 5/18/89 l t
CR HVAC ER System
'f . 3.7.3 SURVEILLANCE RE0VIREMENTS (continued)-
y SURVEILLANCE . FREQUENCY h SR 3.7.3.3 Demonstrate < 0.05% bypass leakage- .18 months through the adsorbet' section by a
- i. ' halogenated hydrocarbon test ~at a system M flow rate _of'[' ) to [ ] cfm.
Once within 7 days after painting, fire, or chemical release in filter service area M-Prior to declaring subsystem-OPERABLE after each O' complete or partial replacement of an adsorber bank (continued) i
=
ABWR 3.7-11 5/18/89
- - _ - - _ _ l
CR HVAC ER System. 3.7.3 SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.7.3.4 -----------------NOTE----------------- Analysis must be completed within 31 days of sampling. Remove and perform a laboratory analysis 720 hours of a charcoal adsorber sample for methyl of charcoal iodide. adsorber operation 4 E I l 18 months E __ Once within ! 7 days after painting, fire, or chemical release in filter service area SR 3.7.3.5 Demonstrate s [ ] inches water gauge 18 months pressure drop across the combined HEPA filters and charcoal adsorber banks at a system flow rate of [ ] to [ ] cfm. SR 3.7.3.6 Demonstrate heaters dissipate from 18 months [ ] to [ ] kw. (continued) l: O i ABWR 3.7-12 5/18/89 ) I
q; CR HVAC ER System. 3.7.3 Oi SURVEILLANCE REQUIREMENTS (continued)
' SURVEILLANCE FREQUENCY SR 3.7.3.7' Demonstrate each subsystem automatically 18 months switches to the isolation mode of operation on receipt.of an. actuation signal.
CROSS-
REFERENCES:
None. O O ABWR 3.7-13 5/18/89
Main Condenser Offgas 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Main Condenser Offons LCO 3.7.4 The gross gamma radioactivity rate of the noble gases measured at offgas recombiner effluent shall be s 390 ! millicuries /second, after 30 minutes decay. APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Gross gamma A.1 Restore gross gamma 72 hours radioactivity rate radioactivity rate to
> 390 mil 11 curies / within limits.
second after 30 minutes decay. B. Required Action and B.1 Isolate all main steam 12 hours associated Completion lines. Time of Condition A not met. 1 O ABWR 3.7 14 5/18/89 l J l 1
- - - - - . _--------_________J
m.. . i i~ ,
-l
'ly 1 Main Condenser Offgas'. ;
3.7.4' f~ 5
?# ? SURVEILLANCE REQUIREMENTS .-
i SURVEILLANCE FREQUENCY i 1 SR .3.7.4.1 Perform'an isotopic analysis.off a:: -----NOTE--- . 1 representative sample of gases taken'at .Only. required 'I the offgas recombiner effluent. after Steam. Jet-Air Ejectoris.in~ j operation. i 31 days AtlD Once within 4 hours' after a
;t 50%
increase in the nominal steady state fission. gas-
-'f release after s
- factoring out increases due to changes in
' THERMAL POWER level CROSS-
REFERENCES:
None I
.ABWR- 3.7-15 5/18/89 i
L1. . . . . _ . _ _ __ _ _ _ . _ _ _ _ _ . _ _ . _ _ _ . _ _ _ . _ _
h 1
.s l
1 r h A.C. Sources - Operating w- 3.8.1 A^-) - 3.8: ELECTRICAL F0WER SYSTEMS ,
'3.8.1 A.C. Sources - Operatino LCO 3.8.1 The following A.C. Electrical Power Sources shall be OPERABLE: ;
A. -Two physically independent' circuits between the off-site transmission network and the on-site Class IE distribution system, 4 8.ND L Three independent diesel generators. APPLICABILITY: MODES 1, 2,-and 3. ACTIONS CONDITION REQUIRE 3 ACTION COMPLETION TIME A, One of the required A.1 Perform SR 3.8.1.2 1 hour circuits from off-site for remaining required-inoperable. circuits from off-site. AND 1L) Once per 8 hours thereafter a!!D A.2 Restore required 14 days from inoperable circuit from discovery of off-site to OPERABLE inoperable status. circuit from off-site j B. One diesel generator B.1 Perform SR 3.8.1.2 for I hour inoperable, required circuits from off-site. AND Once per 8 hours , thereafter l l AND ; (continued)
\
i l ABWR 3.8 1 5/31/89 i
A.C. Sources - Operating i 3.8.1 l ACTIONS (continued) [ CONDITION REQUIRED ACTION COMPLETION TIME q B. (continued) B.2 Verify required redundant 2 hours systems, subsystems, trains, devices and components that depend on the OPERABLE diesel generators as a source of emergency power are OPERABLE. 1 N bND B.3 ----------NOTES----------
- 1. Only required if '
diesel generator became inoperable for any cause other than preplanned maintenance or testing.
- 2. Not required if condition which caused the diesel generator to be h
declared inoperable does not impact OPERABILITY of OPERABLE diesel generators. Perform SR 3.8.1.4 for 24 hours OPERABLE diesel generators. AND B.4 Restore inoperable 14 days from ! diesel generator to discovery of OPERABLE status, inoperable diesel generator (continued) O ABWR 3.8-2 5/31/89
p A.C.' Sources - Operating
, 3.8.1 v.
-! 1 ' ACTIONS fcontinued)
V '. CONDITION' REQUIRED ACTION COMPLETION TIME C. 'One of the required C.1 Perform SR 3.8.1.2 I hour circuits from off-site for remaining required inoperable. circuits-from off-site. E E Once per 8 hours One diesel generator thereafter inoperable. E C.2 Verify required redundsnt
- hours systems, subsystems, trains, devices and components that depend on the OPERABLE diesel generators as a source of emergency power are OPERABLE.
M
.O-C/ C.3 ----------NOTES----------
- 1. Only required if diesel generator became inoperable for any cause other than preplanned maintenance or testing.
- 2. Not required if condition which caused the diesel generator to be declared inoperable does not impact l- OPERABILITY of I
OPERABLE dies M generators. Perform SR 3.8.i.4 for 8 hours OPERABLE diesel generators. (continued) ABWR 3.8-3 5/31/89
A.C. Sources - Operating 3.8.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.4.1 Restore required 72 hours from inoperable circuit from discovery of .' off-site to OPERABLE second inoper-status, able component M C.4.2 Restore inoperable 72 hours from diesel generator to discovery of OPERABLE status, second inoper-able component M C.5 Restore initial 14 days from inoperable component' discovery of (diesel generator or. initial circuit from off-site) inoperable to OPERABLE status, component D. Two diesel generators 0.1 Perform SR 3.8.1.2 1 hour inoperable, for required circuits from off-site. AND Once oer 8 hours there-after M M Verify reauired redundant 2 hours systems. subsystems, ains. devices and components that deoend on the OPERABLE diesel aenerator as a source of emeraency oower are OPERABLE. M y .........-NOTES---------- J2 Oniv reauired if diesel cenerator(s) j' became inocerable I for any cause other ! than oreolanned maintenance or testina. i I (continued) ! i ABWR 3.8-4 5/31/89 ;
.. . _ i
D L s re p - A.C. Sources - Operating. L ! / ~N[ 3.8.1-kJ ACTIONS (continued) i, i-CONDITION REQUIRED ACTION COMPLETION. TIME
----NOTES.(cont.)--------
D. (continued) 2 . Not reauired_if-condition which. caused the diesel
.aenerator(s) to be. j declared inoperable does not imoact
? OPERABILITY of OPERABLE diesel aenerator. Perform SR 3.8.1.2 for 24 hours OPERABLE diesel-cenerator. AND Q4 Restore one of the 72 hours from reouired inoperable discovery of (('j '% diesel aenerators to- second inoper-OPERABLE statut. able diesel 6ND . generator D.5 Restore the initial 14 d'ays from inoperable diesel discovery of generator to OPERABLE initial status. inoperable diesel generator ; [2 One of the reauired Em1 Perform SR 3.8.1.2 1 hour circuits from off-site for remainino reauired inocerable, circuit from off-site. 6ND AND Once per 8 hours ) Two diesel aenerators thereafter inocerable. AND (continued) l 1 ABWR 3.8-5 5/31/89
A.C. Sources - Operating 3.8.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME L (continued) LZ Verify reauired redundant 2 hours systems, subsystems, trains, devices and components that depend on the OPERABLE diesel oenerator as a source of emeroency 00' < r are OPERABLE. AND y . . . . . - - - - - @TM - - - - - - - - - - L Only reouired if diesel aenerator(s) became inocerable for any cause other than oreolanned maintenance or testina. L Not reauired if condition which caused the diesel aenerator(s) to b_g declared inocerable does not imoact OPERABILITY of OPERABLE diesel Perform SR 3.8.1.4 for 8 hours-OPERABLE diesel aeneratol. AND E,4,1 Restore reauired 12 hours inoperable circuit from off-site to OPERABLE status. DB (certinued) O ABWR 3.8-6 5/31/89
i l A.C. Sources - Operating i
._t p-l' 3.8.1 l "s' ' /- 1~
ACTIONS (continued) q CONDITION REQUIRED ACTION COMPLETION TIME )
' L - (continued) E,4,2 Restore one of the 12 hours reauired inocerable diesel aenerators to OPERABLE status.
M LE Restore at least two of 72 hours from the reauired inoperable discovery of components (diesel cen-- second inocer-erators or circuit from able component off-site) to OPERABLE status. M LQ Restore initial 14 days from inocerable component discovery of (diesel cenerator or initial circuit from off-site) inocerabla to OPERABLE status. comoonent
%/
F. Both of the required F.1 Restore one of the 72 hours circuits from off-site required circuits from inoperable. off-site to OPERABLE status. M F.2 Restore the initial 14 days from inoperable circuit from discovery of off-site to OPERABLE initial status. inoperable circuit from off-site L Both of the reauired Ed Verify reauired redundant 2 hours circuits from off-site systems, subsystems, inoperable. trains. devices and components that depend E on the OPERABLE diesel aenerators as a () One diesel aenerator r' source of emeraency inonerable. power are OPERABLE. l M (continued) ABWR 3.8-7 5/31/89 4
f A.C. Sources - Operating h 3.8.1 ACTIONS (continued) , gg CONDITION REQUIRED ACTION COMPLETION TIME gi (continued) G22 ----------NOTES---------- 11 Only reauired if diesel oenerator-became inocerable for any cause other
. than oreolanned maintenance or testina.
- 1. Not reauired if, condition which caused the diesel aenerator to be declared inocerable does not impact OPERABILITY of OPERABLE diesel aenerators.
Perform SR 3.8.1.4 for 9 hours OPERABLE diesel aenerators. AND G,3.1 Pestore one of the re- 24 hours gyired inoperable cir-cuits from off-site to OPERABLE statuss 08 G.3.2 Restore the reauired 24 hours inocerable diesel aen-erator to OPERABLE status. ' AND l Gi Restore at least two of 72 hours from the reauired i:.ooerable discovery of components (diesel cen- second inocer-erator or circuits from able component off-site) to OPERABLE status. 1 8HQ . (continued) {gi ABWR 3.8-8 5/31/89
i' y. A.C. Sources - Operating 1/t - 3. 8 '.1
- '(f' .
ACTIONS (continued) CONDITION REQUIRED ACTION. COMPLETION TIME-G. (continued) 92 1 Restore initial 14 days from inoperable component discovery of (diesel aenerator or initial circuit from off-site) inoperable to OPERABLE status. gomponent H. Three diesel aenerators Hul Perform SR 3.8.1.2 1 hour inoperable, for reauire.d circuits from off-site. E Bt2 Restore one of the re- 2 hours autred inoperable diesel-aenerators to OPERABLE status. E () Bil ' Restore to at least two OPERABLE diesel 72 hours from discovery of aenerators. second inoper-able diesel oenerator AND H.4 Restore to three 14 days from OPERABLE diesel oener- discovery of ators. initial inoper-able diesel-aenerator c I. Required Actions and I.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, AND B, C, D, E, F, G, or H not met. I.2 Be in MODE 4. 36 hours
,,~
k. s ABWR 3.8-9 5/31/89
A.C. Sources - Operating 3.8.1 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify that the load sheddina and. 12 hours seauencina loaic (LSSL) auto-test system is operating and is not indicating a faulted condition. SR 3.8.1.2 Verify correct breaker alignment and 7 days indicated power availability for each of the required circuits from off-site. SR 3.8.1.3 Demonstrate that the response of the LSSL 31 days is within [the design criteria] by a manual insertion of each of the following test signals: A. LOCA.
- 8. Bus undervoltage.
C. Bus undervoltage followed by LOCA. D. LOCA followed by bus undervoltage. SR 3.8.1.4 ------------------NGTES--------------------
- 1. Performance of SR 3.8.1.8 satisfies this surveillance.
- 2. All diesel generator starts may be i preceded by an engine prelube period, warmup procedures and grad-ually loaded as recommended by the manufacturer.
Demonstrate diesel generator achieves the According to following voltage and frequency: Table 3.8.1-1 A. Voltage of 6210 to 7590 volts. B. Frequency of 58.8 to 61.2 Hz. (continued) ABWR 3.8-10 5/31/89
-L f- .
A.C. Sources - Operating ()~ 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY-SR 3.8.1.5 -----------------NOTES--------------------
- 1. Performance of SR 3.8.1.9' satisfies this surveillance.
- 2. All diesel generator starts may be preceded by an engine prelube period, warmup procedures and gradually loaded as recommended by the manufacturer.
- 3. Transient loads outside of this band do not invalidate this test.
Demonstrate diesel generators operate fv According to 2 60 minutes at: Table 3.8.1-1 A. Load of [ ] to [ ] kw for Diesel Generator A.
/3 (y B. Load of [ ] to [ ] kw for Diesel Generator B.
C. Load of [ ] to [ ] kw for Diesel Generator C. SR 3.8.1.6 Verify diesel generators are aligned to According to provide standby powcr to their associated Table 3.8.1-1 emergency buses. SR 3.8.1.7 Verify pressure in each required air According to start receiver is 1 [ ] psig. Table 3.8.1-1 (continued) l l A L) ABWR 3.8-11 5/31/89
A.C. Sources - Operating-SURVEILLANCE RE0VIREMENTS (continued) .___ SURVEILLANCE FREQUENCY SR 3.8.1.8 ----------------NOTE----------------------- All. diesel generator starts may be preceded by an engine prelube period. Demonstrate diesel generator starts from 184 days standby condition and achieves the following voltage and frequency in i 13 seconds: A. Voltage of 6210 to 7590 volts. B. Frequency of 58.8 to 61.2 Hz. SR 3.8.1.9 -----------------NOTES---------------------- ~~
- 1. All diesel generator starts may be preceded by an engine prelube period. ,
- 2. Transient loads outside of this band do not invalidate this test.
W Demonstrate diesel generators are loaded 184 days in s 60 se:oads after synchronization with the bus and operate for 2 60 minutes at: A. Load of [ ] to [ ] kw for i Diesel Generator A. {4 B. Loadof[ ] to ( ) kw for l Diesel Generator B. 1 C. Load of [ ] to [ ] kw for i Diesel Generator C. l I
)
SR 3.8.1.10 Demonstrate manual transfer of unit power 18 months j supply from the normal circuit to alternate ; circuit for each of the required circuits i from off-site. i (continued) O ABWR 3.8-12 5/31/89 i
l i
- A.C. Sources - Operating 3.8.1-G SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.11 Demonstrate the automatic load sequence 18 months timer is OPERABLE with the interval between each load block is [within 10% of its design interval.) SR 3.8.1.12 Demonstrate diesel generator rejects the 18 months following load while maintaining voltage of 3744 to 4576 volts and frequency of 58.8 to 61.2 Hz: A.. 2[ ] kw for Diesel Generator A.
- 8. 2[ ] kw for Diesel Generator B.
C. >[ ] kw for Diesel Generator C. x
\ SR 3.8.1.13 Demonstrate diesel generator does not 18 months trip and voltage of < ['
- ] volts is maintained during and following a load rejection of [ ] to [ ] kw.
(continued) l l-l lO l ABWR 3.8 13 5/31/89
A.C. Sources - Operating 3.8.1 i SURVEILLANCE RE0VIREMENTS fcontinued)
]
SURVEILLANCE FREQUENCY SR 3.8.1.14 -----------------NOTE---------------------- { All diesel generator starts'may be preceded 1 by an engine prelube period and warmup ) procedures recommended by the manufacturer. j I Simulate a loss of off-site power (LOSP) 18 months within 5 minutes of shutting down the diesel after the diesel has operated 2 1 hour : at 2 [ ] kw or the diesel is at normal operating temperature, and demonstrate: 1 A. Deenergization of emergency buses. B. Load shedding from emergency buses. _ C. Diesel generator auto-starts, and:
- 1. Energizes permanently connected loads in $ 13 seconds.
- 2. Energizes auto-connected shutdown loads through the load sequencing logic.
- 3. Supplies permanently and auto-connected loads for > 5~
minutes.
- 4. Achieves and maintains steady state voltage of 6210 to 7590 volts. j
- 5. Achieves and maintains steady state frequency of 58.8 to 61.2 Hz.
(continued) O AEWR 3.8-14 5/31/89
n f 's A.C. Sources ' Operating
%.) 3.8.)
SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.15 -----------------NOTE---------------------- All diesel generator starts may be preceded by an engine prelube period and warmup procedures recommended by the manufacturer. Demonstrate on an ECCS actuation test 18 months signal, without-loss of off-site power, the diesel generator auto-starts, and: A. Achieves and maintains voltage of 6210 to 7590 volts in < 13 seconds after auto-start and during
, remainder of the test.
~~
B. Achieves and maintains frequency of [ 58.8 ] to [ 61.2 ] Hz in < [ 13 ] seconds after auto-start an3 during j remainder of the test. C. Operates for > 5 minutes. (continued) l l I O l l 4
)
4 ABWR 3.8-15 5/31/89
i A.C. Sources - Operating 3.8.1 SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.16 -------------------NOTE-------------------- All diesel generator starts may be preceded by an engine prelube period and warmup procedures recommended by the manufacturer. Demonstrate on a LOSP actuation test 18 months signal in conjunction with an ECCS actuation test signal: A. Deenergization of emergency buses. B. Load shedding from emergency buses. C. Diesel generator autoestarts, and:
-- 1. Energizes permanently connected loads in 5 13 seconds.
- 2. Energizes auto-connected emergency loads through load sequencing logic.
- 3. Supplies permanently and auto-connected loads for > 5 minutes.
- 4. Achieves and maintains steady state voltage of 6210 to 7590 volts.
- 5. Achieves and maintains steady state frequency of 58.8 to 61.2 Hz.
SR 3.8.1.17 Demonstrate automatic diesel generator 18 months trips are automatically bypassed upon an ECCS actuation test signal except: A. Engine overspeed B. Generator differential current (C. Low lube oil pressure) (continued) h ABWR 3.8-16 5/31/89
.I{
. A.C. Sources.- Operating 3.8.1
! L
()- SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE- FREQUENCY SR 3.8.1.18 .---------------NOTE------------------------ Transient loads outside of this band do not invalidate this test.
. Demonstrate the diesel generators operate 18 months for > 24 hours':
A. Loaded from [ ] to [ ] kw. B. With voltage of 6210 to 7590 volts in < 13 seconds after start signal.and during remainder of the. test. C. With frequency cf 58.8 to 61.2 Hz in < 13 seconds after start signal and during remainder of the test.
/\
Verify auto-connected loads s [- 18 months Q SR 3.8.1.19 ] kw. SR 3.8.1.20 Demonstrate each diesel generator: 18 months A. Manually synchronizes with off-site power source, upon a simulated restoration of off-site power, while the diesel generator is loaded with its emergency loads. B. Transfers loads to the off-site power
-source.
C. Returns to standby status. SR 3.8.1.21 Demonstrate that with diesel generator 18 months ope ating in the test mode and connected to its bus, an ECCS actuation test signal overrides the test mode by returning the diesel generator to standby operation. (continued) t ABWR 3.8-17 5/31/89 , ______J
A.C. Sources - Operating 3.8.1 SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.22 Inspect diesel generators in accordance 18 months with procedures prepared in conjunction with manufacturer's recommendations. SR 3.8.1.23 Demonstrate when started simultaneously, 10 years all three diesel generators achieve the following voltage and frequency in 5 13 seconds: A. Voltage of 6210 to 7590 volts. B. Frequency of 58.8 to 61.2 Hz. CROSS-REFERENCES TITLE NUMBER A.C. Sources - Shutdown 3.8.2 Diesel Fuel Oil 3.8.3 O ABWR 3.8-18 5/31/89
- g. .
A.C. Sources - Operating 3.8.1 ( t G/ Table 3.8.1-1 (Page 1 of 1) Diesel Generator Test Schedule I NUMBER OF FAILURES IN NUMBER OF TESTS LAST 20 VALID FAILURES IN (a) LAST 100 VALID TESTS I*) FREQUENCY
$1 $4- 31 days 25 7 days (b) 22
.................................... NOTES-----.---------------.----.--.-......
(a) Criteria for determining number of valid failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis. Failure to satisfactorily complete SR 3.8.3.1 through SR 3.8.3.9 does not constitute a valid failure and does not require an increned frequency r3 of testing. O (b) The associated Frequency shall be maintained until 7 consecutive failure free tests have been performed and the number of valid failures in the last 20 valid tests has been reduced to one. For the purposes of determining the required Frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like new condi tion is completed. This diesel overhaul, including appropriate post-maintenance operation and testing, shall be specifically approved by the manufacturer and acceptable diesel reliability must be demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests. Ten of these tests may be slow starts (in accordance with the routine SR 3.8.1.4 and SR 3.8.1.5) and four tests shall be fast starts (in accordance with the 184 day testing requirement of SR 3.8.1. 8 and SR 3.8.1.9). If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to reset the valid failure count to zero requires NRC approval. i 1 m ABWR 3.8 19 5/31/89
l~ A.C. Sources - Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS . -l 3.8.2 A.C. Sources - Shutdown LCO 3.8.2 The following A.C. Electrical Power Sources shall be OPERABLE: A. One circuit between the off-site transmission network and the on-site Class IE distribution system, M B. One diesel generator. APPLICABILITY: MODES 4 and 5, i When handling irradiated fuel in the secondary containment. 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Less than the required A.I Suspend CORE ALTERATIONS. Immediately l A.C. power sources OPERABLE. E f A.2 ---------NOTE----------- Provisions of LCO 3.0.3 , are not applicable. Suspend handling of Immediately-irradiated fuel in the i secondary containment. ; E l A.3 Suspend operations with As soon as 1 a potential for draining practicable l the reactor vessel. J E A.4 Restore the A.C. power As soon as sources to OPERABLE practicable 3 status. . i O' ABWR 3.8-20 5/31/89
1 l A;C. Sources - Shutdown q
- 73 3.8.2. j SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR' 3.8.2.1 Perform SR 3.8.1.1 through SR 3.8.1.22 for. According to the required equipment with exception of applicable SRs SR 3.8.1.5 and SR 3.8.1.9.
CROSS-REFERENCES , TITLE NUMBER A.C. Sources - Operating 3.8.1 (_/' O ABWR 3.8-21 5/31/89
Diesel Fuel Oil 3.8.3 f l 3.8 ELECTRICAL POWER SYSTEMS l: 3.8.3 Diesel Fuel Oil LC0 > 8.3 The diesel fuel oil subsystem shall be OPERABLE for each required diesel generator. APPLICABILITY: When associated diesel generator is required to be OPERABLE.
........................N0TE--------------------------
Conditions A through F may be concurrently applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel oil level low in A.1 Restore fuel oil level I hour I one or more day tanks. in day fuel tank (s). B. Fuel oil transfer B.1 Restore fuel oil transfer 8 hours capability inoperable capability to OPERABLE for one or more diesel status, generators. O C. Fuel oil level low in C.1 Restore fuel oil level 24 hours one or more storage in storage tank (s). tanks. I l
- 0. Fuel oil properties, D.1 Restore fuel oil 72 hours except for accelerated properties to within stability testing, do limits. i not meet limits in one !
or more storage tanks. E. Fuel oil accelerated E.1 Resture fuel oil 30 days j stability testing properties to within ' properties do not meet limits. ; limits in one or more i storage tanks. , (continued) O ABWR 3.8-22 5/31/89
;m.
Diesel Fuel Oil 4 3.8.3.
. 7% .
I
- 'y ACTIONS (continued)
. CONDITION REQUIRED' ACTION COMPLETION TIME. 1 i
F. Required Actions and F.1 Declare associated Immediately1 associated Completion diesel generator (s) Times of Condition A, f aoperable. B, C,'D, or E not. met. QB 11ese1Lfuel oil-
- subsystem -inoperable for. reasons'other than ,
Condition A, B, C, D, or E. l . O ABWR 3.8-23 5/31/89
Diesel Fuel Oil 3.8.3 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel day tank contair.s According to 2[ ] gallons of fuel. Table 3.8.1-1 SR 3.8.3.2 Verify the fuel storage tanks contains 31 days 1[ ] gallons of fuel. SR 3.8.3.3 Demonstrate each required fuel transfer 31 days pump starts and transfers fuel from its storage tank to its day tank. SR 3.8.3.4 Remove accumulated water in the fuel day Once within tanks. 24 hours after 21 hour of diesel operation O SR 3.8.3.5 Demonstrate properties, other than impurity Prior to level, of new fuel oil sample are within addition of limits, new fuel oil to storage tanks SR 3.8.3.6 Demonstrate impurity level of fuel oil in Once within storage tanks is within limits. 7 days after addition of new fuel oil to storage tank SR 3.8.3.7 Demonstrate properties of fuel oil storage 92 deys tank sample are within limits. (continued) 0 ABWR 3.8-24 5/31/89 L
i I Diesel Fuel Oil (S
~~
3.8.3
'~ -
SURVEILLANCE RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.3.8 Remove acc:Jmulated water in the fuel oil 92 days storage tanks. SR 3.8.3.9 For the fuel oil system: 10 years A. Drain each fuel oil storage tank. B. Remove the sediment from the storage tank. C. Clean the storage tank. CROSS-REFERENCES C v - .). TITLE NUMBER A.C. Sources - Operating 3.8.1 ! A.C. Sources - Shutdown 3.8.2 I i a f
'N I
ABWR 3.8-25 5/31/89 j
r i D.C. Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS hl 3.8.4 D.C. Sources - Operatina LCO 3.8.4 Divisions 1, 2, 3 and 4 of the D.C. Electrical Power Sources shall be OPERABLE. ) l APPLICABILITY: MODES 1, 2, and 3. ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME L Division 4 D.C. oower A.1 Restore inoperable power 30 days source inoperable. source to OPERAdLE status. I L One of the required B.1 Restore inoperable 7 days from Division 1. 2. or 3 power source to discovery of D.C. power sources OPERABLE status. inoperable inoperable. power source L Two of the reouired L1 Restore at least one of 24 hours D.C. oower sources the inoperable oower inocerable. sources to OPERABLE status. MQ LZ Restore the initial 7 days from inocerable oower source discovery of to OPERABLE status. initial inoo-erable Divi-sion 1. 2 or 3 Dower source Q3 30 days from discovery of initial inocerable Division 4 Dower source (continued) ABWR 3.8-26 5/31/89
'~
D.C. Sources - Operating
- ' \ 3.8.4
'\ .)
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME Q Three of the reauired Qtl Restore at least one of 2 hours D.C. nower sources the inocerable power inoperable. sources to OPERABLE status. AND Q12 Restore at least two of 24 hours from the inocerable oower discovery of sources to OPERABLE second inon-itAttui. erable Dower SOUTCO. AND Q11 Restore the initial 7 days from inoperable power source discovery of to OPERABLE status, initial inon-erable Divi-r'~ x sion 1. 2 or (,,) 3 power source 08 30 days from discovery of initial inoperable Division 4 Dower source E. Required Actions and E.1 Be in MODE 3. 12 hours associated Completion Times of Condition A, AND B, C or D not met. E.2 Be in MODE 4. 36 hours
/
w l i ABWR 3.8-27 5/31/89
D.C. Sources - Operating 3.8.4 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Demonstrate the Battery Terminal Voltage 7 days
> 129 volts on float charge.
SR 3.8.4.2 Verify no visible corrosion at terminals 92 days or connectors. MQ E Once within Demonstrate the connection resistance of 31 days terminals and connectors is after a 1 150x10~6 ohms. battery discharge below 110 volts MD Once within 31 days after a battery overcharge above 150 volts SR 3.8.4.3 Verify the cells, cell plates and battery 18 months racks show no visual indication of physical damage or abnormal deterioration. SR 3.8.4.4 Verify the cell-to-cell and terminal 18 months connections are clean, tight, free of corrosion and coated with anti-corrosion material. (continued) O ABWR 3.8-28 5/31/89
)
J D.C. Sources - Operating f ') 3.8.4 V SURVEILLANCE RE0VIREMENTS (continued)
.3 SURVEILLANCE FREQUENCY ')
)
i SR 3.8.4.5. Demonstrate the resistance of each cell-and .18 months l terminal connection is 5150 x 10~6 ohms. ) 1 l SR 3.8.4.6 Demonstrate each battery charger will 18 months , supply the following amparage at a minimum ! of 125 volts for at least 10 hours: ! A. 2 500 amperes for Divisions 1, 2, and 3. B. 2 200 amperes for Division 4. I SR 3.8.4.7 ---------------NOTE----------------------- Performance of SR 3.8.4.8 satisfies this surveillance. Demonstrate battery capacity is adequate to 18 months supply and maintain in OPERABLE status.all actual or simulated emergency loads for the design duty cycle when the battery is subjected to a batte"' service test. SR 3.8.4.8 Demonstrate the battery capacity is at least 60 months (80% of the manufacturer's rating] when subjected to a performance discharge test. $_!!Q 18 months when battery shows degradation or has reached 85% of the expected life l O ABWR 3.8-29 5/31/89
1 D.C. Sources - Operating l 3.8.4 j CROSS-REFERENCES f TITLE NUMBER j f i D.C. Sources - Shutdown 3.8.5 Battery Electrolyte 3.8.6 i O l l 9 ABWR 3.8-30 5/31/89
J
~s D.C. Sources - Shutdown
( 3.8.5
) ) ,
3.8 ELECTRICAL POWER SYSTEMS 3.8.5 D. C. Sources - Shutdown LCO 3.8.5 One Division of the Division 1. 2 or 3 D.C. Electrical Power Sources shall be OPERADLE. APPLICABILITY: MODES 4 and 5, When handling irradiated fuel in the secondary containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Less than the required A.1 Suspend CORE ALTERATIONS. Immediately D.C. power sources OPERABLE. AND A.2 ----------NOTE---------- Provisions of LC0 3.0.3 are not applicable.
........................ 1 (T Suspend handling of Immediately
\- /' irradiated' fuel in the ,
secondary containment. A!!Q A.3 Suspend operations with As soon as a potential for draining practicable the reactor vessel. AND A.4 Restore at least one As soon as division (Divison 1, practicable 2 or 3) of the D.C. power system to OPERABLE status. . /~' 1 I l ABWR 3.8-31 5/31/89
D.C. Sources - Shutdown 3.8.5 i* ' SURVEILLANCE RE0VIREMENTS SR 3.8.5.1 Perform SR 3.8.4.1 through SR 3.8.4.8 According to " for the required equipment. applicable SRs l l I CROSS-REFERENCES TITLE NUMBER D.C. Sources - Operating 3.8.4 Battery Electrolyte 3.8.6 e-9 ' ABWR 3.8-32 5/31/89 .
Ba'ttery Electrolyte
/N' '
-3.8.6 Q. ,
3.8 ELECTRICAL POWER SYSTEMS-l 3.8.6 Batter _v Doctrolyte LCO 3.8.6- -Battery electrolyte.for the Divisions 1, 2, 3 and 4 batteries shall be within the limits of Table 3.8.61. 7 APPLICABILITY: When associated D.C. power sources are required to be-OPERABLE. L ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
.A . One or more cells in A.1 Demonstrate the pilot I hour one or more batteries cells' electrolyte level not within limits. and float' voltage meet '
Category C allowable values. M A. 2. Demonstrate the 24 hours parameters in Table
. !q .s 3.8.6-1 meet. Category C
'V allowable values.
E A.3 Restore the parameters 7 days to Category A and B limits of Table 3.8.6-1. B. Required Actions and B.1 Declare associated D.C. Immediately associated Completion power source inoperable. Times of Condition A not met. 03 A"erage electrolyte temperature of the pilot cells is s 60'F. O i I l ABWR 3.8-33 5/31/89 a
l r i Battery Electrolyte 3.B.6 s0RVEtttANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Demonstrate the parameters in Table 7 days 3.8.6-1 meet the Category A limits. SR 3.8.6.2 Demonstrate the parameters in Table 92 days 3.8.6-1 meet the Category B limits. MQ Once within 31 days after a battery discharge below 110 volts AND i l Once within s 31 days after a battery ; overcharge ; above 150 volts SR 3.8.6.3 Verify the average electrolyte temperature 92 days of the pilot cells > 60*F. AND Once within 31 days after a battery discharge below 110 volts AND Once within 31 days after a battery overcharge above 150 l volts ABWR 3.8-34 5/31/89
t
.c e
/ -
. ~~
Battery Electrolyte
-3.8.6-l CROSS-REFERENCES TITLE- NUMBER D.C. Sources - Operating 3.8.4 D.C. Sources - Shutdown- 3.8.5 O
o l ABWR 3.8-35 5/31/89
Bettery Electrolyte 3.8.6 Table 3.8.6-1 (Page 1 of 1) Battery Surveillance Requirements CATEGORY A CATEGORY B CATEGORY C Parameter Limits for each Limits for each Allowable designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and 5 1/4" above and 5 1/4" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage (b) > 2.13 volts > 2.13 volts > 2.07 volts Specific 2 1.195(c) > 1.190 Not mere than Gravity (,) 0.020 below the average of , all connected l cells MQ AND 1 J Average of all Average of all connected cells connected cells
> 1.200 > 1.190(c)
.....................--N0TES------------------------------
(a) Corrected for electrolyte temperature and level. (b) May be corrected for average electrolyte temperature. (c) Or battery charging current is < 2 amperes when on float charge. O 1 l ABWR 3.8-36 5/31/89
1 i A.C.. Power Dist. Systems - Operating
'(Q 3.8.7 ,
q/ 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 A.C. Power Distribution Systems - Coeratino
'LC0'3.8.7- The Division 1. 2. and 3 A.C. oower distribution systems shall be eneraized.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS-CONDITION REQUIRED ACTION. COMPLETION TIME A. One of the required A.C. A.1 Declare affected . Immediately-power distribution . equipment inoperable. Divisions deenergized. M A.2 Re-energize the . 24 hours from A.C. power distribution . discovery of Division. deenergized A.C. power distribution
~
Division ik Two of the reauired B.1 Declare affected Immediately A.C. oower distribution equipment inoperable. Divisions deeneraized. E B.2.1 Reenergize at least one 8 hours from of the deenergized A.C. discovery of power distribution second Divisions, deeneraized A.C. oower ; distribution ! division h.N.Q B,2,2 Reeneraize the initial 24 hours from deeneraized A.C. oower discovery of distribution division. initial deener-aired A.C. oower distribution division. (continued) l ABPR 3.8-3l 5/31/89 I _ _ _ _ - _ _
l A.C. Power Dist. Systems - Operating 3.8.7 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME L. All three of the L1 Daclare affect g Immediately reauired A.C. Dower eauipment inoperabi9. distribution Divisions deeneraized. 98 M Reeneraize at least one 2 hours A.C. oower distribution , Division. l D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion l Times of Condition A, BLD. l B or C not met. i 0.2 Be in MODE 4. 36 hours i l O O ABWR 3.8-38 5/31/89
l T 'T A.C. Power Dist. Systems - Operating 4 u_.) 3.8.7 l SURVEILLANCE RE0VIREMENTS i SURVEILLANCE FREQUENCY l SR 3.8.7.1 Verify correct breaker alignment and 7 days i' voltage to A.C. power distribution system.
)
CROSS-REFERENCES TITLE NUMBER A.C. and D.C. Power Distribution Systems - Shutdown 3.8.9
%>r v
i ABWR 3.8-39 5/31/89
f I D.C. Power Dist. Systems - Operacing ) 3.8.8 - 3.8 ELECTRICAL POWER SYSTEMS i 3.8.8 D.C. Power Distribution Systems - Operatina 'j LCO 3.8.8 The Division 1. 2. 3. and 4 D.C. oower distribution systems i I shall be energized. APPLICABILITY: MODES 1, 2, and 3. 1 ACTIONS ! CONDITION REQUIRED ACTION COMPLETION TIME. j A. One of the required D.C. A.1 Declare affected Immediately 3 power distribution equipment inoperable. Divisions deenergized. 03 A.2 Re-energize the 24 hours D.C. power distribution Division. 4 l I L Two of the reauired D.C. L1 Declare affected Immediately power distribution eauipment-inocerable. & Divisions deeneraized. T 0B B.2.1 Reeneraize at least one 2 hours of the deeneraized D.C. oower distribution Divisions. AND M .] Reeneraize the intial 24 hours from deeneraized 0.C. oower discovery of distribution division initial deeneraized D.C. oower distribution division. C. Required Actions and C.1 Be in MODE 3. 12 hours associated Completion Times of Condition A A@ or B not met. C.2 Be in MODE 4. 36 hours (continued) ABWR 3.8-40 5/31/89 b
D.C. Power Dist. Systems - Operating 7'^N 3.8.8 ' '~')
+
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D Three or more of the D.1 Declare affected Immediately required D.C. power equipment inoperable, distribution Divisions deenergized. QB D.2.1 Be in MODE 3. 12 hours 8BD D.2.2 Be in MODE 4. 36 hours 1 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY
,em, NJ SR 3.8.8.1 Verify correct breaker alignment and 7 days voltage to D.C. power distribution system.
CROSS-REFERENCES TITLE NUMBER A.C. and D.C. Power Distribution Systems - Shutdown 3.8.9 s l lv) ! ABWR 3.8-41 5/31/89 l
A.C. & D.C. Power Dist. Systems - Shutdown l 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 A.C. and D.C. Power Distribution Systems - Shutdown LCO 3.8.9 One comolete division of A.C. and D.C. oower distribution system shall be eneroized. APPLICABILITY: MODES 4 and 5, When handling irradiated fuel in the secondary containment. : ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 Suspend CORE ALTERATIONS. Immediately , required power l distribution buses AND deenergized. A.2 ----------NOTE---------- Provisions 9, LC0 3.0.3 l 4 are not applicable. ) Suspend handling of Immediately irradiated fuel in the secondary containment. AND A.3 Suspend operations with As soon as a potential for practicable draining the reactor vessel. AND A.4 Reenergize at least one As soon as i complete Division of practicable l A.C. and D.C. power distribution. 1 O i ABWR 3.8-42 5/31/89 l
!1 A.C. &'D.C. Power Dist. Systems Shutdown
' p)l, j,. .. 3.8.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY' SR 3.8.9.1 Perform SR 3.8.7.1 and SR 3.8.8.1 for According to the required equipment.- SR 3.8.7.1 and SR 3.8.8.1 CROSS-REFERENCES TITLE NUMBER A.C. Power Distribution Systems -' Operating 3.8.7 i
~~
D.C. Power Distribution Systems - Operating 3.8.8
. en.
f 1 ABWR 3.8-43 5/31/89
l
.r3 Refueling Equipment Interlocks 3.9.1
-Q 3.9 REFUE'ING OPERATIONS 3.9.I Refuelina Eauipment Interlocks LCO 3.9.1 The Refueling Equipment Interlocks shall be OPERABLE with.
the following inputs: l A. All-rods-in. -l B. Refuel platform position. C. Refuel platform main hoist fuel-loaded. APPLICABILITY: MODE 5 when handling fuel assemblies using equipment associated with the interlocks. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
- A. One or.more of the A.I Suspend CORE ALTERATIONS Immediately ;
required Refueling with equipment associated p, Equipment Interlocks with the inoperable
.-Q inoperable. interlock.
I SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform a [ CHANNEL FUNCTIONAL TEST). 7 days CROSS. REFERENCES TITLE NUMBER Control Rod Drive Removal - Refueling 3.10.5 k ABWR 3.g.I 5/31/89
y Refuel Position One-Rod-Out Interlock : 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock i LC0 3.9.2 The Refuel position one-rod-out interlock shall be OPERABLE. APPLICABILITY: MODE 5 with the Reactor Mode Switch in the Refuel position and any control rod withdrawn. l ACTIONS j CONDITION REQUIRED ACTION COMPLETION TIME i A. Refuel position A.1 Suspend control rod Immediately I one-rod-out interlock withdrawal . inoperable. j
~-
A.2 Fully insert all As soon as 1 insertable control rods practicable I in core cells containing one or more fuel ! assemblies. ! l I SURVEILLANCE RE0VIREMENTS 1 SURVEILLANCE FREQUENCY ! l SR 3.9.2.1 Perform a [ CHANNEL FUNCTIONAL i. 'j. 7 days QB 1 hour when l any control { rod is I withdrawn j O ABWR 3.9-2 5/31/89 j
Refuel Position One-R;d-Out Interlock O 3.9.2 CROSS-REFERENCES TITLE NUMBER Control Rod Withdrawal - Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal - Refueling 3.10.5 O O ABWR 3.9-3 5/31/89
Control Rod Position 3.9 REFUELING OPERATIONS 3.9.3 Centrol Rod Position LCO 3.9.3 All control rods shall be fully inserted. APPLICABILITY: When loading fuel assemblies into the core and not following an approved spiral reload sequence. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. All control rods not A.1 Suspend loading fuel Immediately fully inserted. assemblies into the core. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY
- SR 3.9.3.1 Verify all control rods are fully inserted. 12 hours l
i CROSS-
REFERENCES:
None O ABWR 3.9-4 5/31/89
11 L X Control Rod Position Indication 3.9.4 () 3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indication LCO 3.9.4 All control rod Full In position indicators for control rods'in core cells containing one or more' fuel assemblies shall be OPERABLE. APPLICABILITY: MODE 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One'or'more required A.I.I Declare the Refueling Immediately position indicators Equipment Interlocks inoperable. inoperable, bND A.1.2 Declare the Refuel Immediately I Position One-Rod-Out Interlock inoperable. DE A.2.I fully insert the control As soon as rod (s) associated with practicable the inoperable position indicator (s). bND A.2.2 Disarm the associated As scon as control rod drive (s). practicable O ABWR 3.9-5 5/31/89
Control Rod Position Indication 3.9.4 SURVEILLANCE RE0VIREMENTS _ SR 3.9.4.1 Verify each required control rod 7 days j Full In position indicator is OPERABLE. J CROSS-REFERENCES TITLE NUMBER Refueling Equipment Interlocks 3.9.1 Refuel Position One-Rod-Out Interlock 3.9.2 Control Rod V?thirawal - Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal - Refueling 3.10.5 O 1 O ABWR 3.9-6 5/31/89
T 3 , _f
:f :
i I l f F
' Control Rod 0PERABILITY - Refueling
-3.9.5.
y , 3.9 ' REFUELING' OPERATIONS-3.9.5 Go.rttral Rod OPERABILITY - Refuelina LC0 3.9.5 Each control rod withdrawn from 'a core cell containing one or more fuel assemblies shall.be OPERABLE. ,, APPLICABILITY: MODE 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME'- A. One or more required A.1 Fully insert inoperable As'soon as-control rods control. rods. practicable ..' inoperable. SURVEILLANCE RE0VIREMENTS SURVEILLANCE . FREQUENCY' SR 3.9.5.1. Insert each required withdrawn control rod 7 days. at least one ilta. SR 3.9.5.2 Verify each required control rod scram 7 days accumulator pressure is ;t [ 1850 ]' psig. CROSS-REFERENCES TITLE NUMBER Control Rod Withdrawal - Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal - Refueling 3.10.5 ABWR: 3.9-7 5/31/89
p Water Level - RPV 3.9.6 3.9 ' REFUELING OPERATIONS 3.9.6 Water Level - Reactor Pressure Vessel LCO 3.9.6 Water level shall be 2 23' over the top of the Reactor Pressure Vessel (RPV) flange. APPLICABILITY: When handling irradiated fuel assemblies over the RPV. When handling fuel assemblies over the RPV with irradiated fuel assemblies seated within the RPV. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-A. Water level < 23' A,1 Suspend handling fuel Immediately over the top of the assemblies over the RPV. RPV flange. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.9.6.1 Verify water level in the RPV is 24 hours 2 23' over the top of the RPV ( flange. l l i l CROSS-REFERENCES TITLE NUMBER l ECCS - Shutdown 3.5.2 Residual Heat Removal - High Water Level 3.9.8 ! ( Residual Heat Removal - Low Water Level 3.9.9 l 9j l ABWR 3.9-8 5/20/89 i
.y
, .. Water Level - Fuel Pools e 3.9.7 -
3.9 REFUELING OPERATIONS 3.9.7 Maler Level - Spent Fuel Storace Pools LCO 3.9.7 Water level shall be 123 feet over the top of irradiated fuel assemblies seated in the Spent Fuel Storage Pools. APPLICABILITY: When irradiated fuel assemblies are stored in the Spent Fuel Storage Pools. ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. Water level < 23 A.I ----------NOTE---------- Immediately feet over the top of Provisions of LCO 3.0.3 irradiated fuel- are not applicable, assemblies in the ------------------------
'~~
Spent Fuel Storage Pools. Suspend handling fuel assemblies over the O Spent Fuel Storage b Pools. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify water level is 2 23 feet over 7 days the top of irradiated fuel assemblies seated in tre Spent Fuel Storage Pools. 4 CROSS
REFERENCES:
None L) I ABWR 3.9-9 5/20/89
l RHR - High Water Level l 3.9.8 l 1 3.9 REFUELING OPERATIONS . 3.9.8 Residual Heat Removal - Hioh Water Level LCO 3.9.8 One Residual Heat Removal (RHR) shutdown cooling subsystem shall be OPERABLE. j APPLICABILITY: H0DE 5 with water level 2 [ ] feet over the top of the RPV flange and heat losses to ambient not sufficient to maintain average reactor coolant temperature s 140*F. . 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1 A. No RHR shutdown cooling A.1 Provide an alternate 8 hours subsystem OPERABLE. method capable of decay heat removal.
)
B. Required Action and B.1 Suspend operations that Immediately I associated Completion could increase reactor Sb l Time of Condition A decay heat load. not met. , AND B.2 Verify Secondary As soon as Containment is OPERABLE. practicable - A!!D B.3 Ensure at least one SGTS As soon as subsystem is OPERABLE. practicable AND B.4 Ensure at least one As soon as Secondary Containment practicable Isolation Valve and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. AND 8 (continued) O ABWR 3.9-10 5/20/09 I
J};~3 RHR - High Water Level 3**** fy) . ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.5 Provide an alternate As soon'as method capable of decay practicable heat removal. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify for the required RHR shutdown 7 days cooling subsystem each manual, power operated, or automatic valve in the flow path not locked, sealed or otherwise secured in position, is in the correct /'N position or is capable of being manually \,~_/) aligned in the correct position. ! C_EOSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] ECCS - Shutdown 3.5.2 Secondary Containment 3.6.4.1 Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 l
~~ .
A-- s ABWR 3.9 11 5/20/89
RHR - Low Water Level 1 3.9.9 l 3.9 REFUELING 'OPERATI0t:S 3.9.9 Residual' Heat Removal - Low Water Level h' LCO 3.9.9 i Two Residual Heat Removal (RHR) shutdown cooling subsystems ? shall be OPERABLE. APPLICABILITY: MODE 5 with water level < [ feet over the top of the RPV flange and heat losses] to ambient not sufficientl a maintain average reactor coolant temperature s 140*F. j ACTIONS CONDITION i REQUIRED ACTION COMPLETION TIME ! A. One of the required A.1 RHR shutdown cooling Provide an alternate 8 hours from subsystems inoperable. method capable of decay discovery of
' heat removal. inoperable subsystem E
A.2 Raise water level to " 8 hours from 2[ ] feet over the discovery of top of the RPV flange. inoperable subsystem B. Both of the required 8.1 Provide an alternate RHR shutdown cooling 2 hours subsystems inoperable. method capable of decay heat removal. (continued) I a O ABWR 3.9-12 5/20/89
Q q ? l i RHR.- Low Water Level. - 7~T . - 3,9,9 l
. \ )-
-ACTIONS (continued)
CONDITION REQUIRED ACTIOK COMPLETION TIME. . i
'C. Required Actiors and C.1 Suspend operations that.. Immediately associated Completion could increase reactor 4 Times-of Condition A decay heat load. l or B not met.
M I C.2 Ensure Secondary. . As soon as Containment is OPERABLE. practicable M C.3 Ensure at least one SGTS As soon as subsystem is OPERABLE. practicable M C.4 Ensure at least one As soon as Secondary Containment. practicable Isolatiori Valve and () (m associated actuation instrumentation is OPERABLE in each associated penetration not isolated. M C.5 Provide at least one As soon as alternate method capable practicable . of decay heat removal.
)
1 I i t i ABWR 3.9 13 5/20/89 i
RHR - Low Water Level 3.9.9 SURVEILLANCE REQUIREMENTS . SR 3.9.9.1 Verify for the required RHR shutdown 7 days cooling subsystem each manual, power operated, or automatic valve in the flow path not locked, sealed or otherwise secured'in position, is in the correct position or is capable of being manually aligned in the correct position. CROSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] ECCS - Shutdown 3.5.2 Secondary Containment 3.6.4.1 Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 O l t l 1 I O1 ABWR 3.9-14 5/20/89 , i --________-----_m
a l
,- ,s Inservice Leak and Hydro Test 3.10.1
( ;- 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testino ODePation . I LCO 3.10.1 The average reactor coolant temperature specified in Table 1.1-1 for MODE 4 operation may be changed to "Any temperature" to allow performance of an inservice leak or hydrostatic test provided the following MODE 3 LCOs are met: A LC0 ISecondary Containment Isolation Actuation Instru- l mentation. Functions for low Reactor Water Level and Hiah Refuel Floor Radiation 1. B. LC0 3.6.4.1, Secondary Containment. C. LC0 3.6.4.2, Secondary Containment Isolation Valves, t D. LC0 3.6.4.3, Standby Gas Treatment System. APPLICABILITY: MODE 4 with average reactor coolant temperature > 200*F. ACTIONS
'J CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 ----------NOTE----------
above required LCOs Required Actions to be not met. in MODE 4 include reducing average reactor coolant temperature to
$ 200*F.
Enter the applicable Immediately i Condition of the ; affected LCOs. l 0_8 A.2.1 Suspend activities that Immediately could increase the average reactor coolant temperature or pressure. AND [3 N_) (continued) ABWR 3.10-1 5/31/89
l Inservice Leak and Hydro Test t 3.10.1 ACTIONS (continued) ! CONDITION REQUIRED ACTION COMPLETION TIME l 4 A. (continued) A.2.2 Reduce average reactor 24 hours coolant temperature to s 200*F. SURVEILU NCE RE0VIREMENTS SURVEILLANCE FREQUENCY ; I SR 3.10.1.1 Perform the applicable SRs for the According to required MODE 3 LCOs. the applicable SRs CROSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] Secondary Containment 3.6.4.1 Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 O ABWR 3.10-2 5/31/89 l
Reactor Mode Switch Interlock Testing j 3.10.2 3.10 SPECIAL OPERATIf"'t 3.10.2 Reactor Mode t .ch Interlock Testina LCO 3.10.2 The Reactor Mode Switch position specified in Table 1.1-1 for MODES 3, 4 and 5 operation may be changed to irclude the Run, Startup/ Hot Standby, and Refuel position to allow : testing of instrumentation associated with the Reactor Mode Switch interlock functions provided all control rods remain fully inserted and no other CORE ALTERATIONS are in progrcss. APPLICABILITY: MODES 3 and 4 with the Reactor Mode Switch in the Run,
-Startup/ Hot Standby, or Refuel position, MODE 5 with the Reactor Mode Switch in the Run or Startup/ Hot Standby position.
ACILONS ... COND1 ? ION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Suspend CORE ALTERATIONS Immediately O LCO not met. except for control rod insertion. MD A.2 Fully insert all I hour insertable control rods. MD A.3.1 Place the Reactor Mode I hour Switch in the Shutdown position. E A.3.2 ----------NOTE---------- Not applicable in MODES 3 and 4. Place the Reactor Mode I hour Switch in the Refuel position. O ABWR 3.10-3 5/31/89
Reactor Mode Switch Interlock Testing 3.10.2 SURVEILLANCE RE0VIREMENL SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully 12 hours inserted. SR 3.10.2.2 Verify no other CORE ALTERATIONS are 24 hours in progress. i CROSS-
REFERENCES:
None i O 1 i 1
)
i i O\ ABWR 3.10-4 5/31/89 ____-_________m_ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _
1 Control Rod Withdrawal - Hot Shutdown
-- 3.10.3-pb
- Q 3.10 SPECIAL OPERATIONS: ,
1 3.10.3 control Rod Withdrawal - Hot Shutdown ) 1 LCO 3.10.3 The Reactor Mode Switch position specified in Table 1.1-1 - for MODE 3 operation may be changed to include the Refuel position to allow withdrawal of-a single control rod or control rod pair provided the following requirements ave j
' met:.
'A. LCO 3.9.2, Refuel Position One-Rod-Out- Interlock.
B. LCO 3.9.4, Control Rod Position Indication. C. All other control. rods are fully inserted. L L LCO IReactor Protection System Instrumentation, Functions for: SRNM Hiah Flux. Fast Period, and INnP: APRM Hiah Setdown Flux and INOP; React 3r :, ode Switch in Shutdown: and Manual Scram. All for Mode 5 anolicability only.1-I LC0 3.9.5, Control Rod OPERABILITY - Refueling. L All other control rods in a'five-by-five array centered on gnh control rod being withdrawn are disarmed. LCO 3.1.1, MODE 5, SHUTDOWN MARGIN, except the control rod for oairi to be withdrawn may be assumed to be the highest worth control rod pric. APPLICABILITY: MODE 3 with the Reactor Mode Switch in the Refuel position. 1 I O ABWR 3.10-5 5/31/89
1 Control Rod Withdrawal - Hot Shutdown l 3.10.3 .g i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME q i A. One or more of the A.1 --..-..---NOTE---------- I above requirements Required Actions to i not met. fully insert all insertaole control rods q include placing the Reactor Mode Switch in the Shutdown position. Enter the applicable Immediately Condition of the affected LCOs. QBR A.2.1 Fully insert all I hour insertable control rods. g l A.2.2 Place the Reactor Mode 1 hour Switch in the Shutdown position. f 9 ABWR 3.10-6 5/??/09
y . . p Control Rod Withdrawal - Hot Shutdown i' 3.10.3 n;d_/- r3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY .] 1 I SR' 3.10.3.1 Perform the applicable SRs for the According to j required LCOs. the applicable- i SRs SR 3.10.3.2 Verify all other control rods in a 24 hours five-by-five array centered on rg h control rod being withdrawn are disarmed. SR 3.10.3.3- Verify all other control rods are fully 24 hours
-l inserted.
77 CROSS-REFERENCES V. TITLE NUMBER SHUTDOWN MARGIN 3.1.1 Reactor _ Protection System Instrumentation [ ] Refuel Position One-Rod-Out Interlock 3.9.2 Control Rod Position Indication 3.9.4 Control Rod OPERABILITY - Refueling 3.9.5 l O ABWR 3.10-7 5/31/89
i. Control Rod Withdrawal - Cold Shutdown 3.10.4 3.10 SPECIAL OPERATIONS 3.10.4 Control Rod Withdrawal - Cold Shutdown LCO 3.i.0.4 The Reactor Mode Switch position specified in. Table 1.1-1 for MODE 4 operation may be changed to include the Refuel position to allow withdrawal of a single control rod or control rod pair or withdrawal and subsequent removal of the associated control rod drive (s) provided the following requirements are met: A. All other control rods are fully inserted. B. 1. LC0 3.9.2, Refuel Position One-Rod-Out Interlock. LC0 3.9.4, Control Rod Position Indication. M
- 2. A control rod withdrawal block is inserted.
L L LC0 IReactor Protection System Instrumentation, Functions for: SRNM Hiah Flux. Fast Period and INOP: APRM Hiah Setdown Flux and INOP: Reactor Mode Switch in Shutdown: and Manual Sc. am. All for Mode 5 Applicability only.1 LC0 3.9.5, Control Rod OPERABILITY - Refueling. M L All other control rods in a five-by-five array centered on gEh control rod being withdrawn are ; disarmed. LC0 3.1.1, MODE 5, SHUTDOWN MARGIN, except the control rod f or oair) to be withdrawn may be assumed to be the highest worth control rod gq10 MODE 4 with the Reactor Mode Switch in the Refuel position. APPLICABILITY: i l l I i l O \ ABWR 3.10-8 5/31/89
l s Control- Rod Withdrawal - Cold Shutdown , 3.10.4, ) []
. y'j
- 4 ACTIONS' ,
CONDITION REQUIRED ACTION COMPLETION TIME
- L One or more of the' A.1 ----------NOTE----------
.above requirement (s) Required Actions to
.not met with the fully insert all, affected control insertable: control . rods a rodM insertable. include placing the j '
Reactor Mode Switch in the Shutdown position. Enter the applicable Immediately Condition of the affected LCOs. E A.2.1 Fully insert all I hour insertable control rods.
; /S MQ L) A.2.2 Place the Reactor Mode I hour Switch in the Shutdown position.
L One or more of the IL1 Suspend- withdrawal . of Immediately
, ' above requirements not the control rodM and met with the affected removal of associated control rod M not control rod drive M . ) '
insertable. AND B.2.1 Fully insert all As soon as i' control rods. practicable M B.2.2 Satisfy the requirements As soon as of the LCO. practicable ABWR 3.10-9 5/31/89 L i- - - - _ - - - - - - - - _
3 l Control Rod Withdrawal - Cold Shutdown i 3.10.4 j SURVEILLANCE RE0VIREMENTS ) SURVEILLANCE FREQUENCY I SR 3.10.4.1 Perform the applicable SRs for the According to required LCOs. the applicable SRs l j S.P 3.10.4.2 Verify all other contrcl rods in a 24 hours . five-by-five array centered on _qhh control ' rod being withdrawn are disarmed. < l 1 SR 3.10.4.3 Verify all other control rods are fully 24 hours inserted.
~
l SR 3.10.4.4 Verify a centrol rod withdrawal block is 24 hours inserted. CROSS-REFERENCES O TITLE NUMBER SHUTDOWN MARGIN 3.1.1 Reactor Protection System Instrumentation [ ] Refuel Position One-Rod-Out Interlock 3.9.2 , Control Rod Position Indication 3.9.4 Control Rod OPERABILITY - Refueling 3.9.5 O ABWR 3.10-10 5/31/89
Control Rod Drive Removal - Refueling 3.10.5 r7X t.) . 3.10 SPECIAL OPERATIONS 3.10.5 Gpaty_pi Rod Drive Removal - Refuelina LCO 3.10.5 The requirements of LCO LRPS' Instrumentation, LCO 1BES non-coincident model, LCO 3.9.1, LCO 3.9.2, LCO 3.9.4 and LCO 3.9.5 may be suspended during MODE 5 operation to allow the removal of_ a single control rod drive or control rod drive oair associated with control rodisl withdrawn from core ce11111 containing one or more fuel assemblies provided the following requirements are inet: A. All other control rods are' fully inserted. L All other control rods in a five-by-five array centered on the control rod 111 being removed are disarmed. C. A control rod withdrawal block is inserted. L LC0 3.1.1, MODE 5, SHUTDOWN MARGIN, except the single' control rod for oair) to be withdrawn may be assumed to be the highest worth control rod p31t.
, E. No other CORE ALTERATIONS are in pro 0ress.
APPLICABILITY: MODE 5 with LC0 3.9.5 not met. i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 ' Suspend removal of the Immediately above requirements not control rod (s) and met. associated control rod drive mechanism (s). AND i A.2.1 Fully insert all As soon as control rods. practicable DLR A.2.2 Satisfy the requirements As soon as of the LCO. practicable L ABWR 3.10-11 5/31/89 i i L__.____ !
Control Rod Drive Removal - Refueling 3.10.5 SURVEILLANCE RE0VIREMENTS -- SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all other control rods are fully 24 hours inserted. SR 3.10.5.2 Verify all control rods in a five-by-five 24 hours array centered on qach control rod being removed are disarmed. SR 3.10.5.3 Verify a control rod withdrawal block is 24 hours inserted. SR 3.10.5.4 Perform the applicable SRs for LC0 3.1.J. According to the applicable SRs SR 3.10.5.5 Verify no other CORE ALTERATIONS are in progress. 24 hours h 1 O ABWR 3.10-12 5/31/89
>1 u- j 8
i I Control Rod Drive Removal -'-. Refueling ; 3.10.5 i
.,x.
- 3. j '
-CROSS-REFERENCES-TITLE NUMBER 1
SHUTDOWN MARGIN 3.1.1 ~. i Reactor Protection System-Instrumentatico ( ) Reactor Protection System Non-Coincident Mode [ ] Refueling Equipment Interlocks- 3.9.1 Refuel Position One-Rod-Out Interlock 3.9.2. Control Rod Position Indication 3.9.4 Control Rod OPERABILITY - Refueling 3.9.5 10 v l h ABWR 3.10-13 5/31/89 i
Multiple Control Rod Withdrawal - Refueling 3.10.6 3.10 SPECIAL OPERATIONS 3.10.6 Multiple Control Rod Withdrawal - Refuelina LCO 3.10.6 The " Full In" position indicators may be bypassed for any number of control rods during MODE 5 operation to allow withdrawal of those control rods, removal of associated control rod drives, or both provided the following requirements are met: A. The four fuel assemblies are removed frota the core cells associated with each caitrol rod or control rod drive to be removed. B. All other control rods in core cells containing one or more fuel assemblies are fully inserted. APPLICABILITY: MODE 5 with one or m;re " Full In" position indicators bypassed. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Suspend withdrawal of Immediately LCO not met. control rods and remr, val of associated 1 control rod drives. M 1 A.2.1 Fully insert all As soon as I control rods in core practicable i cells containing one or more fuel assemblies. - D.B A.2.2 Satisfy the requirements As soon as , of the LCO. practicable j l e! ABWR 3.10-14 5/31/89 i
I Multiple Control Rod. Withdrawal - Refueling J, 3.10.6' Q SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.6;l Verify the ~ four fuel assemblies are . . 24 hours removed from core cells associated with a control rod or control rod drive to be removed. SR 3.10.6.2 Verify all other control rods in core 24 hours cells containing one or more fuel assemblies are fully inserted. CROSS-REFERENCES TITLE NUMBER. t '
> Control Rod Position Indication 3.9.4 ABWR 3.10-15 5/31/89
__--_:___ _ _ _ _ _ _ _ __ b
E Control Rod Testing - Operating 3.10 SPECIAL OPERATIONS 3.'0.7 Control Rod Testina - Operatino LCO 3.10.7 The requirements of LCO 3.1.6 may be suspended and control rods bypassed in the Rod Control and Information System (RC&IS) as allowed by SR [ ] during MODES 1 and 2 with THERMAL POWER less than or equal to the LPSP of the , RC&IS to allaw performance of SHUTDOWN MARGIN dameratrations, control rod scram time testing, control rod I friction testino, and the Startup Test Pregram provided conformance with the control rod sequence for the specified I test is verified by a second licensed operator or other qualified member of the technical staff. APPLICABILITY: MODES I and 2 with LC0 3.1.6 not met. ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME i A. Requirements of the A.1 Suspend performance of Immediately LC0 not met. the test. SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY SR 3.10.7.1 Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the specified test. CROSS-REFERENCES TITLE NUMBER Rod Pattern Control 3.1.6 Control Rod Block Instrumentation [ ] ABWR 3.10-16 5/31/89
s TABLE OF CONTENTS l t ); , l s ; Paae B 2. SAFETY LIMITS B 2.0-1
/
B 3. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS c ,/ B 3. APPLICABILITY Limiting Conditions for Operation . . . . . . . . . . . . . . . B 3-1 Surveillance Requirements . . . . . . . . . . . . . . . . . . . B 3-7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . B 3.1-1 B 3.1.2 Control Rod OPERABILITY . . . . . . . . . . . . . . . B 3.1-7 B 3.1.3 Control Rod . Scram Times . . . . . . . . . . . . . . . B 3.1-13 B 3.1.4 Control Rod Scram Accumulators . . . . . . . . . . . . B 3.1-19 B 3.1.5 Control Rod Drive Coupling and Separation Detection. . B 3.1-23 B 3.1.6 Rod Pattern Control . . . . . . ... . . . . . . . . . B 3.1-27 B 3.1.7 Standby Liquid Control System . . . . . . . . . . . . B 3.1-31 r'3 B 3.2 POWER DISTRIBUTION LIMITS V B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . . B 3.2-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO . . . . . . . . . . . . . B 3.2-5 B 3.2.3 LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . B 3.2-9 B 3.3 INSTRUMENTATION (Later] B 3.4 REACTOR COOLANT SYSTEM B 3.4.1 Recirculation Pumps Operating . . . . . . . . . . . . B 3.4-1 B 3.4.2 Safety / Relief Valves . . . . . . . . . . . . . . . . . B 3.4 5 8 3.4.3 Operational Leakage . . . . . . . . . . . . . . . . . B 3.4-9 B 3.4.4 Specific Activity . . . . . . . . . . . . . . . . . . B 3.4-13 B 3.4.5 Residual Heat Removal - Shutdown . . . . . . . . . . . B 3.4-16 B 3.4.6 Reactor Coolant System Pressure / Temperature Limits . . B 3.4-19 B 3.4.7 Reactor Steam Dome Pressure . . . . . . . . . . . . . B 3.4-26 B 3.5 ECCS and RCIC i B 3.5.1 ECCS - Operating .................. B 3.5-1 B 3.5.2 ECCS - Shutdown . .................. B 3.5-10 f/v'N l ABWR i 5/31/89 ] l
= _ _ _
TABLE OF CONTENTS (continued) Paae B 3. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS (continued) B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Primary Containment Systems B 3.6.1.1 Primary Containment . . . . . . . . . . . . B 3.6-1 B 3.6.1.2 Containment Air Locks . . . . . . . . . . . B 3.6-5 B 3.6.1.3 Containment Pressure ........... B 3.6-11 B 3.6.1.4 Containment Average Air Temperature . . . . B 3.6-13 B 3.6.1.5 Primary Containment [and Pressure] Isolation Valves ............. B 3.6-15 B 3.6.1.6 Wetwell-to-Drywell Vacuum Breakers. . . . . B 3.6-21 B 3.6.2 Suppression Pool B 3.6.2.1 Suppression Pool Average Temperature ... B 3.6-26 B 3.6.2.2 Suppression Pool Water Level ....... B 3.6-31 B 3.6.2.3 Residual Heat Removal Suppression Pool Cooling . . . . . . . . . . . . . . . . . . B 3.6-34 8 3.6.2.4 Residual Heat Removal Wetwell Spray System .................. B 3.6-38 B 3.6.3 Hydrogen Control B 3.6.3.1 Containment Hydrogen Recombiner Systems . . B 3.6-41 B 3.6.3.2 Primary Containment Oxygen Concentration. . B 3.6-42 B 3.6.4 Secondary Containment Systems B 3.6.4.1 Secondary Containment . . . . . . . . . . . B 3.6-45 B 3.6.4.2 Secondary Containment Isolation Valves .. B 3.6-51 B 3.6.4.3 Standby Gas Treatment System ....... B 3.6-55 B 3.7 PLANT SYSTEMS B 3.7.1 Reactor Building Cooling Water / Reactor Building i Service Water System - Operating .......... B 3.7-1 B 3.7.2 Reactor Building Cooling Water / Reactor Building Service Water System - Shutdown . . . . . . ..... B 3.7-7 B 3.7.3 Control Room HVAC Emergency Recirculation System .. B 3.7-10 B 3.7.4 Main Condenser Offgas . . . . . . . . . . . . . . . . B 3.7-15 O ABWR ii 5/31/B9 i _ _ _ 3
)
1 1 WW l' TABLE OF CONTENTS
-(continued) ;
l I Pace
. B 3. LIMITING CONDITIONS FOR OPERATION-AND SURVEILLANCE RE0VIREMENTS
.(continued)
B 3.8 ELECTRICAL POWER SYSTEMS B'3.8.1 A.C. Sources - Operating . . . . . . . . . . . . . . B 3.8-1 B 3.8.2 A~.C. Sources - Shutdown . . . . .. . . . . . . .. . . B 3.8-21 B 3.8.3 Diesel Fuel Oil . . . . . . . . . . . . . . . . . . . B 3.8 B 3.8.4 D.C.-Sources - Operating . . . . . . . . . . . . . . B 3.8-28 B 3.8.5 D.C. Sources - Shutdown . . . -. . . . . . . . . . . B 3.8-34 B 3.8.6 Battery Electrolyte . . . . . . . . . . . . . . . . B 3.8-36 B 3.8.7 A.C. Power Distribution Systems - Operating . . . . B 3.8-40 B 3.8.8 D.C. Powe,* Distribution Systems - Operating . . . . B 3.8-45 B 3.8.9 A.C. and D.C. Power. Distribution Systems - Shutdown . B 3.8-49 B 3.9 REFUELING OPERATIO9S B 3.9.1 . Refueling Etpioment Interlocks . . . . . . . . . . . B 3.9-1 O B 3.9.2 Refuel Position One-Rod-Out Interlock ....... B 3.9-4
'(,)g B 3.9.3 Control Rod Position . . . . . . . . . . . . . . . . B 3.9-7 B 3.9.4 Control Rod Position Indication ....... .. B 3.9-10 B 3.9.5 Control Rod OPERABILITY - Refueling ........ B 3.9-13 B 3.9.6 Water Level - Reactor Pressure Vessel ....... B 3.9-15 B 3.9./ ' Water Level - [ Spent Fuel Storage Pools) . . . . . . B 3.9-17 B 3.9.8 Residual Heat Removal - High Water Level . . . . . . B 3.9-19 B 3.9.9 Residual Heat Removal - Low Water Level ...... B 3.9-22 B 3.10 SPECIAL OPERATIONS B 3.10.1 Inservice Leak,and Hydrostatic Testing Operation . . B 3.10-1 B 3.10.2 Reactor Mode Tsitch Interlock Testing ....... B 3.10-4 B 3.10.3 Control Rod Withdrawal - Hot Shutdown ....... B 3.10-7 B 3.10.4 Control Rod Withdrawal - Cold Shutdown . ...... B 3.10 10 B 3.10.5 Control Rod Drive Removal - Refueling ....... B 3.10-13 B 3.10.6 Multiple Control Rod Withdrawal - Refueling .... B 3.10-16 B 3.10.7 Control Rod Testing - Operating .......... B 3.10-18 l
ABWR iii 5/31/89 1 _ _ _ _ ___ _ _ _ _. _ j
I Safety Limits ; B 2.0 j 1
..n ,
Q B 2.0 SAFETY LIMITS BASES INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radio-active ~ materials to the environs. Safety Limits are estab-lished to-protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel-damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable,' a stepback j approach is used to establish a Safety Limit such that the Minimum Critical Power Ratio (MCPR) is not less than the limit specified in Specification 2.1.2. .MCPR greater than the specified limit represents a conservative margin relative i to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive material.s from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life rN of the cladding, fission product migration from this source ( is incrementally cumulative and' continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would' produce onset of transition boiling (i.e., MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling. FUEL CLADDING GE critical power correlations are applicable for all INTEGRITY critical power calculations at pressures at or above 785 psig (2.1.1) or core flows at or abovt 10% of rated flow. For operation at low pressures and low flows another basis is used as follows: (continued) ABWR B 2.0-1 5/31/89
Safety Limits-1 B 2.0 BASES (continuedl _ FijEL CLADDING Since the pressure drop in the bypass region is essentially
- INTEGRITY all elevation head, the core pressure drop at low power and (2.1.1) flows will always be greater than 4.5 psi. Analyses show (continued) that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundlo power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
MINIMUM The fuel cladding integrity Safety Limit is set such that no CRITICAL POWER fuel damage is calculated to occur if the limit is not RATIO violated. Since the parameters which result in fuel damage (2.1.2) are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not l result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel , rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical 1 power. The probability of the occurrence of boiling transition is determined using the approved General Electric i I Crith?; Power correlations. Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1. Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis. (continued) ABWR B 2.0-2 5/31/89
= - _ _ - - _ _ - _ _ - - _- _ - - . -
i Safety Limits a B 2.0 l BASES (continued) REACTOR VESSEL With fuel in the' reactor vessel during periods when the WATER LEVEL reactor is shutdown, consideration must be given to water (2.1.3) level requirements due to the effect of decay heat. If the .q water level-should drop below the top of the active i irradiated fuel during this period, the ability to' remove decay heat is reduced. 'This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action. REACTOR The Safety Limit for the reactor steam dome pressure has been STEAM DOME selected such that it is at a pressure below which it can be. PRESSURE shown that the integrity of-the system is not endangered. (2.1.4) The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code (1971 Edition, including Addenda through Winter 1972), which permits a maximum pressure transient of 110%,.1375 psig, of design pressure liSO psig. The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor cociant system is also designed to Section 111 of
/ the ASME Code and is similarly protected by the 1325 psig i Safety Limit.
REFERENCES 1. " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A, September 1988. ABWR B 2.0-3 5/31/89 ____y
$ LCO- Applicability E-B 3.0 B 3.0 APPLICABILITY - LIMITING CONDITIONS FOR OPERATION (LCOs). BASES- l l] 1 The LCO 3.0s are general requirements which apply to Section 3.0 Limiting; Conditions for 0peration and Surveillance Requirements and apply-at all times, d j unless otherwise stated. i LCO 3.0.I. 'LC0 3.0.1 establishes the Applicability statement within each
. individual specification as the requirement for when (i.e., in which MODES or other specified conditions)o conformance to the LC0 is required.
LCO 3.0.2 The Required Actions establish those remedial measures that must be taken within specified Completion Times when the 4 requirements of an LCO are not met. The Completion Times of the Required Actions are applicable from the point in time it m is discovered that the Condition exists and is entered and the " Applicability of the LC0 is met. Only one Condition in each LCO's ACTIONS may be entered at any time, unless otherwise stated (i .e. , " Conditions. . .may be concurrently applicable").. O. .If the Limiting Condition for Operation'is met prior to expir-V atio'n of the specified Completion Time (s), completion of the Required Action is not required, unless otherwise stated (i.e.,
"Reouired Actions....must be completed ~whenever this Condition ist.ttered").
The Completion Times of the Required Actions are also applicable when a system or component'is removed from service intention-ally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillance Requirements, preventive maintenance and corrective maintenance. It is not intended that this intentional entry into ACTIONS be made for operational convenience. This intentional entry shall be under appropriate administrative control. This is to limit voluntary. removal of redundant equipment from service in lieu of other alternatives that would not result in redundant' equipment being inoperable, thus limiting the time both subsys-tems/ trains of a safety function are inoperable and limiting the time other conditions exist which result in LCO 3.0.3 being entered. O (continued) ABWR B 3-1 5/31/89 l
LCO Applicability B 3.0 BASES (continued) LC0 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LC0 is not met and:
- 1. An associated Required Action and Completion Time is not met and no other Condition applies, or
- 2. The Condition is not specifically addressed by the l associated ACTIONS.
The purpose of this Specification is to delineate the time limits for placing the unit in a safe MODE or other specified , condition when operation cannot be maintained within the limits for safe operation defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience to permit (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant equipment being inoperable. One hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This time permits the 4 operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and avail-ability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the capabilities of the facility assuming only a minimum required , equipment is OPERABLE. This reduces the thermal stresses on components of the Reactor Coolant System and the potential for i a plant upset that could challenge safety systems under condi-tions for which this Specification applies. The use and i interpretation of the specified times to complete the actions - of LC0 3.0.3 shall be consistent with the discussion of ; Section 1.3, Completion Times. l A unit shutdown required per LC0 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs:
- 1) The LC0 is now met.
3
- 2) A Condition exists for which the Required Actions have now been performed.
1
- 3) ACTIONS exist which do not have expired Completion Times. These Completion Times are applicable from j the point in time there was a failure to meet that i LCO and not from the time LCO 3.0.3 is exited.
l' l j l , (continued) l I ABWR B 3-2 5/31/89
LCO Applicability-
> B 3.0 9 BASES (continued)
G' LC0 3.0.3 The requirements of LC0 3.0.3 do not apply in MODES 4 and 5 (continued) because the plant is already in the most restrictive Condition that LC0 3.0.3 would require the plant to be placed in. The requirements of LC0 3.0.3 do not apply in other specified conditions in the Applicability, unless in MODES 1, 2 or 3, because the ACTIONS of individual specifications: sufficiently define the remedial measures.to be taken. The exceptions.to LC0 3.0.3 are provided in instances where requiring a plant shutdown in accordance.with LC0.3.0.3 would-not provide appropriate remedial measures for the associated . condition of the unit. An example of this is in LC0 3.9.7, Water Level - Spent Fuel' Storage Pool. LC0 3.9.7 has an Applicability of "When irradiated fuel assemblies are' stored in the Spent Fuel Storage Pool". Therefore, this LCO can.be applicable in any or all MODES. sIf the LC0 and the Required Actions of LC0 3.9.7 are not met while in MODES 1, 2, or 3, there is no safety benefit to be gained by placing the plant in a shutdown condition. The Required Action of LC0 3.9.8 of
" Suspend further movement of fuel assemblies over the Spent Fuel Storage Pool" is the appropriate Required Action to attempt to complete in lieu of the ACTIONS of LCO 3.0.3.
LC0 3.0.4. LC0 3.0.4 establishes limitations on. MODES or other specified conditions in the Applicability changes when an LC0 is not. met. It precludes placing the unit in a-different MODE or othcr
~specified condition when the requirements of an LCO in the MODE or other specified condition to be entered are not met and continued noncompliance with these requirements would result in a shutdown to comply with the Required Actions if a change in MODES or other specified conditions was permitted. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in the MODE or'other specified condition provides an acceptable level of safety for continued operation rithout regard to the status of the plant before.or after the MODE change. Therefore, in this case, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. The provisions of this Specification l
should not, however, be interpreted as endorsing the failure to l exercise good practice in restoring systems or components to OPERABLE status before plant startup. The provisions of LC0 3.0.4 shall not prevent changes in MODES ! or other specified conditions in the Applicability which are required to comply with ACTIONS. O (continued) l ABWR B 3-3 S/31/89
I l LC0 Applicability i ' B 3.0 BASES'(continued) LC0 3.0.5 LC0 3.0.5 establishes the allowance of restoring equipment to service under administrative controls when it has been removed 1' frem service or declared inoperable to comply with ACTIONS. The purpose of this' Specification is to provide an exception to LCO 3.0.2 to allow the performance of Surveillance Requirements , to 1) demonstrate the OPERABILITY of the equipment being i returned to service, or 2) demonstrate-the OPERABILITY of other . equipment. An example of demor.strating the OPERABILITY of the equipment being returned to service is the re-opening of a containment {i isolation valve that has been closed to comply with Required Actions, and must be re-opened to perform the Surveillance Requirements. An example of demonstrating the OPERABILITY of other equipment is the taking of an inoperable channel or trip system out of the tripped condition to prevent the trip function from occur-ring during the performanc.e of a Surveillance Requirement on another channel in the otner trip system. Another similar example of demonstrating the OPERABILITY of other equipment is the taking of an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of a Surveil-lance Requirement on another channel in the same trip system. LC0 3.0.6 LC0 3.0.6 establishes the allowance of delaying Required Actions for up to 8 hours to perform Surveillance Requirements when systems are rendered inoperable for the performance of the Surveillance Requirements. The purpose of this Specification is to provide an exception to LC0 3.0.2 to allow performance of Surveillance Requirements required to demonstrate OPERABILITY. (continued) ABWR B 3-4 5/31/89
LC0 Applicability-B 3.0 [
-u BASES (continued)
LCO 3.0.7 LC0 3.0.7 establishes which ACTIONS are applicable when support systems are inoperable, depending on whether or not they have an LC0 specified in the Technical Specifications. The sup-ported system is not required to be. declared inoperable ' solely. due to support system inoperability.- Only the support system LCO's ACTIONS are required to be entered. This is _ a clarifica-tion of'the definition'of OPERABILITY but_is necessary to establish the relationship between_ the support systems and the supported system in order to to preclude cascading to _ multiple supported system ACTIONS and.to eliminate the confusion associ-ated with entering multiple LCOs' ACTIONS. Examples of support systems with LCOs specified in the Technical Specifications include cooling / service water, diesel generators and AC and DC distribution. When a support system is inoperable and there is not an LCO for that support system specified in the Technical Specifications, the licensee shall evaluate the impact of the inoperability or
' degradation of the support system function on the OPERABILITY of the supported system. This is because the inoperability or degradation of the support system function may or may not affect the OPERABILITY of the supported system, depending upon the intended function of the supported system and the level of . ,q support the support system provides. Upon determination that the supported system is inoperable, the ACTIONS of its LC0 G shall apply.
1 LC0 3.0.8 There are certain special tests and operations required to be performed at various times ec the life of the plant. These special tests and operatic.s are necessary to demonstrate select plant performance characteristics, to perform'special maintenance activities and to perform special evolutions. Special Operations LCOs in Section 3.10 allow specified , Technical Specifications requirements to be changed to permit i performance of these special tests and operations which l otherwise could not be performed if required to comply with the requirements of these Technical Specifications. Unless otherwise specified, all the other Technical Specification requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in ' effect. (continued) f ABWR B 3-5 5/31/89
LC0 Applicability B 3.0 BASES (continued) LCO 3.0.8 The Applicability of a Special Operations LC0 represents a i (continued) condition not necessarily in compliance with the normal 5 requirements of the Technical Specifications. Compliance with { Special Operations LCOs is optional. A special operation may I be performed either under the provisions of the appropriate f Special Operations LC0 or the other applicable Technical i Specification requirements. If it is desired to perform the I I special operation under the provisions of the Special Opera-tions LCO, the requirements of the Special Operations LCO shall be followed. When a Special Operations LC+ requires another LC0 to be met, only the requirements of the LC0 are required to be met regardless of that LC0's Applicability; i.e., should the requirements of the other LC0 not be met, the ACTIONS of the Special Operations LC0 apply, not the ACTIONS of the other LCO. 1 However, there are instances where the Special Operations LC0 ACTION may direct the other LCOs' ACTIONS be met. The Surveil-lances of the other LC0 are not required to be met, unless otherwise specified in the Special Operations LCO. Conditions may exist such that the Applicability of the other LC0 is met and ali that LCO's requirements are required to be met concur- , rent with the requirements of the Special Operations LCO. i O i i 1 l O ABWR B 3-6 5/31/89 l 1
SR Applicability. B 3.0 B 3.0 . APPLICABILITY. . SURVEILLANCE REQUIREMENTS (SRs).
' BASES i
The SR 3.0s are general requirements which apply to Section 3.0 Limiting
. Conditions for Operation.and Surveillance Requirements and apply'at all times, unless otherwise stated.
SR 3.0.1 SR 3.0.1 establishes the requirement that' Surveillance-Requirements must be performed during the' MODES or other specified conditions in the Applicability of the LC0 unless otherwise ' stated in an individual Surveillance Requirement. The purpose of this Specification is to ensure that Surveil-lance Requirements are performed to verify the OPERABILITY of. systems and components.and that parameters are within specified limits. Systems and components are assumed to be OPERABLE when Surveillance have been met within the specified Frequency. However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are known to be inoperable although still meeting the Surveillance Requirements. Surveillance Requirements do not have to be performed when the unit is in a MODE or other.specified condi-tion for which the requirements of the associated LC0 do not apply, unless otherwise specified. An example of this is a Surveillance Requirement whose LC0's Applicability is MODES 1, 2, and 5, and requires a CHANNEL' FUNCTIONAL TEST once during each MODE 4 of greater than 24 hours duration. Surveillance Requirements do not have to be. performed on inoperable equipment because the ACTIONS define the remedial measures'that apply. However, the Surveillance Requirements have to be met as required by SR 3.0.2 prior to restoring equipment to OPERABLE status. Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes meeting applicable Surveillance Requirements in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters having not been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily i completed to the extent possible and the equipment is rot otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other speci- ! fied condition where other necessary post maintenance tests can 1 be completed. l 0 (continued) ABWR B 3-7 5/31/89
SR Applicability B 3.0 l [ BASES (continued) l SR 3.0.1 Some examples of this process are:
-(continued) - CRD maintenance during refueling which requires scram testing at > 950 psi. . However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.3.3 is satisfied, the control rod can be considered OPERABLE. This allows startup to proceed to reach 950 psi to perform other necessary testing.
- RCIC maintenance during shutdown which requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with RCIC considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
SR 3.0 2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillance and any Required Action with a Completion Time requiring the periodic performance of the ACTION on a "once per. . ." interval . SR 3.0.2 permits an extension of the interval associated with the stated frequency to facilitate Surveillance scheduling and for consideration of plant operating conditions that may not be suitable for & W conducting the Surveillance; e.g., transient conditions or other ongoing Surveillance or maintenance activities. The 25% extension does not significantly degrade the reliability which results from performing the Surveillance at its specified frequency. This is based on engineering judgement and the recognition that the most probable result of ' any particular Surveillance Requirement being performed is the verification of conformance with the Surveillance Requirements. The exceptions to SR 3.0.2 are those Surveillance where the 1.25 times the interval specified in the Frequency does not apply. An example of where SR 3.0.2 does not apply is SR 3.6.1.1.2, whose surveillance is " Perform required Type A leak rate testing in accordance with 10 CFR 50 Appendix J and approved exemptions" and whose Frequency is "In accordance with 10 CFR 50 Appendix J and approved exemptions." The j requirements of regulations take precedence over the Technical ' Specifications. The Technical Specifications cannot in and of themselves extend a test interval specified in the regulations. Therefore, SR 3.6.1.1.2 has a note in the Frequency stating
" Provisions of SR 3.0.2 are not applicable." l 1
(continued) , ABWR B 3-8 5/31/89 m - - - - - - - _ - - - - . - - - - - - . - -
SR Applicability B 3.0 p BASES (continued) SR 3.0.3 SR 3.0.3 establishes the flexibility to defer. declaring the equipment inoperable when Surveillance have not been completed within the specified Frequency. An allowed 24 hour deferral applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2 and not at the time that the specified frequency was not met. This deferral provides an adequate time limit to complete Surveillance that have been missed. The purpose of this allowance is to permit the completion of a Surveillance before compliance with Required Actions or other remedial measures would be required that may preclude completion of a Surveil-lance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the Surveillance and the safety significance of the delay in completing the required Surveil-lance. This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of MODE changes imposed by Required Actions and for completing Surveillance Requirements that are applic-able when an exception to the requirements of SR 3.0.4 is allowed. If a Surveillance is not completed within the 24 hour allowance, the equipment is considered inoperable and the Completion Times of the Required Actions begin immediately upon expiration of the 24 hour allowance. /~] v If a Surveillance is failed within the 24 hour allowance, the equipment is considered inoperable and the Completion Times of the Required Actions begin immediately upon the failure of the Surveillance. SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable Surveillance Requirements must be met before entry into a MODE or other specified condition in the Applicability. The purpose . of this specification is to ensure that system and component j OPERABILITY requirements or parameter limits are met before { entry into a MODE or other specified condition in the l i Applicability for which these systems and components ensure safe operation of the facility. This provision applies to changes in MODES or other specified conditions in the i Applicability associated with plant shutdown as well as startup. The provisions of SR 3.0.4 shall not prevent changes in MODES l or other specified conditions in the Applicability which are j required to comply with ACTIONS. 4 ( V (continued) ABWR B 3-9 5/31/89
SR Applicability-B 3.0 BASES (continued) SR 3.0.5 SR 3.0.5 establishes the requirement that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be per-formed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. These requirements apply except when relief has been requested pursuant to 10 CFR 50.55a (g)(6)(i). This Specification includes a clarification of the Frequencies for performing the. inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to l ensure consistency in Surveillance Frequencies throughout the Technical Specifications and to remove any ambiguities relative to the Frequencies for performing the required inservice inspection and testing activities. Under the terms of this Specification, the more restrictive requirements of the Techni-cal Specifications take precedence over the ASME Boiler-and Pressure Vessel Code and applicable Addenda. The requirements of SR 3.0.4 to perform Surveillance activities before entry into a MODE or other specified condition in the Applicability takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps and valves to be tested up to one week after return to normal operation. Also, the Technical Specification definition of OPERABILITY does not allow a grace period before a component that is not capable of performing its l specified function is declared inoperable. The Technical Specifications take precedence over the ASME Boiler and Pres-sure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. i O ABWR B 3-10 5/31/89 L _ _ __ _
-i, SHUTDOWN MARGIN-
~
B 3.1.1. B 3.1 REACTIVITY CONTROL SYSTEMS l
, Wy f B 3'.l.1 ~ SHUTDOWN MARGIN' BASES l
l
. BACKGROUND SHUTDOWN MARGIN is specified to ensure: j
- a. The reactor can be made subtritical from all operating !
conditions. !
- b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits.
- c. The reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition.
General Design Criterion (GDC) 26 requires the reactivity control systems be capable of holding the core subtritical under cold conditions, APPLICABLE SHUTDOWN MARGIN is an explicit assumption in several of the f evaluations in SSAR Chapter 15, Accident Analyses. SHUTDOWN ( SAFETY ANALYSES MARGIN is assumed as an initial condition for the Control Rod Removal Error During Refueling (Ref.1) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the-REFUELING mode of operation. These interlocks prevent the withdrawal of more than one control rod, or control rod pair, from the core during refueling. (Special consideration and requirements for multiple control rod withdrawal during refuel-ing are covered in Special Operations LC0 3.10.6, Multiple Control Rod Withdrawal - Refueling.) The analysis assumes this condition is acceptable since the core will be shutdown with the highest worth control rod pair withdrawn, if adequate SHUTDOWN MARGIN has been demonstrated. Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage which could result in undue release of l radioactivity (see Bases for LCO 3.1.6). Adequate SHUTDOWN MARGIN ensures inadvertent criticalities and potential control rod withdrawals involving high worth control rods will not cause significant fuel damage. (continued) O ABWR B 3.1-1 5/31/89 l l l
SHUTDOWN MARGIN B 3.1 1 BASES fcontinued) APPLICABLE SHUTDOWN MARGIN satisfies the requirements of Selection
~ SAFETY Criterion 2 of the NRC Interim Policy Statement on Technical ANALYSES Specification Improvements as documented in Reference 4.
(continued). LCO The specified SHUTDOWN MARGIN' accounts for the uncertainty-in the demonstration of SHUTDOWN MARGIN therOby ensuring the reactor can be held subcritical during shutdown conditions and during refueling with the highest worth control rod pair withdrawn. Separate SHUTDOWN MARGIN limits are provided for demonstrations whera the highest worth control rod pair is determined analytically or by measurement. This is due to the reduced uncertainty in the SHUTDOWN MARGIN demonstration when the highest worth control rod pair is determined by measure-ment. To assure adequate SHUTDOWN MARGIN, an additional design margin is included in the design process to account for uncertainties in the design calculations (Ref. 2). APPLICABILITY During MODES I and 2, the SHUTDOWN MARGIN specification is applicable because subcriticality with the highest worth control rod pair withdrawn is assumed in the analysis. Also, the capability to reach MODE 4 conditions from any initial state is required by GDC 26. For operation in MODES 3 and 4, the SHUTDOWN MARGIN specification is required to ensure the reactor will be held subcritical with margin for a single stuck control rod. The SHUTDOWN MARGIN specification is applied to MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or of a control rod pair from loaded core cells during scram time testing. (continued) O ABWR B 3.1-2 5/31/89
y l SHUTDOWN MARGIN ' B 3.1.1' 73 BASES (continued) ( ) ACTIONS M During MODE l'or 2,' failure to meet the specified SHUTDOWN MARGIN may be caused by a control rod that cannot be inserted. Because the reactor _can still be shutdown assuming no failuresi i of additional control rods to-insert, operation is allowed to l continue.for a short time to allow restoration of SHUTDOWN i MARGIN. U l If the SHUTDOWN MARGIN cannot be restored promptly, the reactor must be in MODE 3 (i.e. insert all insertable control rods) to
- prevent the potential for further reductions in available-
. SHUTDOWN MARGIN (e.g. additional. stuck control rods).
L.1 > With SHUTDOWN MARGIN less than specified in MODE 3 the operator must insert all insertable control rods. This action results in the least reactive condition for the core. D.1. D.2. D.3. D.4 With SHUTDOWN MARGIN less than specified in MODE 4, the operator must insert all insertable control rods. This action results in the least reactive condition for the core. Actions are also taken to provide means for control of potential radioactive releases caused by an inadvertent reactivity excursion. This includes ensuring Secondary Containment is OPERABLE (LC0 3.6.4.1), at least one Standby Gas Treatment System (SGTS) subsystem is OPERABLE (LC0 3.6.4.3) and at least one Secondary Containment Isolation Valve (LC0 3.6.4.2) and associated actuatiu instrumental' m is OPERABLE in each associated penetration not isolat'L Ensuring the OPERABLE status of the components involves administrative checks, examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It does not require performing surveillance needed to demon-strate the OPERABILITY of the components. If however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, surveillance requirements may need to be performed to restore the component to OPERABLE status. (continued) I i ABWR B 3.1-3 5/31/89
SHUTDOWN MARGIN B 3.1.1 BASES (continued) ACTIONS E.1. E.2. E.3. E.4. E.5 (continued) With SHUTDOWN MARGIN less than specified in MODE 5, the operator must suspend C0'<E ALTERATIONS since these activities could reduce SHUTDOWN MARGIN. This involves the movement of fuel in the core or the withdrawal of control rods. The requirement that the activities be suspended immediately is not intended to prohibit the completion of an action that would improve or not affect SHUTDOWN MARGIN (e.g. complete the insertion of a control rod). All control rods in core cells containing one or more fuel assemblies must be insarted to place the core in the least reactive condition. Control rods in core cells with no fuel asssmblies are not required to be inserted since they have a negligible impact on core reactiv-ity. . Actions are also taken to provide means for control of potential radioactive releases caused by an inadvertent reac-tivity excursion. This includes ensuring Secondary Containment is OPERABLE (LC0 3.6.4.1), at least one Standby Gas Treatment-System (SGTS) subsystem is OPERABLE and at least one Secondary Containment Isolation Valve (LC0 3.6.4.2) and associated actuation instrumentation is OPERABLE in each associated penetration not isolated as described in the Bases for Required Actions D.2, D.3 and 0.4. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available - systems and the time required to reasonably complete the Required Actions. SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SHUTDOWN MARGIN must be demonstrated for the entire cycle length and must be performed before or during the first startup following CORE ALTERATIONS which involve changes in the core reactivity. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle demonstration must also account for changes in core reactivity during the cycle. Therefore, the specified SHUTDOWN MARGIN must be increased by a factor, R, which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning of cycle core reactivity. If the value of R is negative (that is, beginning of cycle is the most reactive point in the cycle), no correction to the beginning of cycle SHUTDOWN MARGIN is required (Ref. 3). 1 (continued) ABWR B 3.1-4 5/31/89
SHUTDOWN MARGIN P B 3.1.1 i-BASES (continued) "V SURVEILLANCE SR 3.1.1.1 (continued) REQUIREMENTS (continued) The SHUTDOWN MARGIN may be demonstrated during an in-sequer, control rod withdrawal in which the highest' worth control roa pair is analytically determined or during local criticals, where the highest worth control rod. pair is determined by testing. Local critical tests may also be performed, .but require the withdrawal of out-of-sequence (i.e., adjacent) I control rods. This testing would therefore require bypassing of the rod pattern control systems to allow the out-of-sequence-withdrawal and additional requirements must be met (see LCO' s 3'.10.7, Control Rod Testing - Operating). l Four hours after reaching criticality is provided to allow a reasonable time to perform the required calculations and have appropriate verification. SR 3.1.1.2 During MODE 5 adequate SHUTDOWN MARGIN is also required. An evaluation of each fuel movement during fuel loading shall be performed to ensure adequate SHUTDOWN MARGIN is maintained during refueling. This ensures the intermediate loading patterns are bounded by the safety analyses for the final core
,n loading pattern. For example, bounding analyses which demonstrate adequate SHUTDOWN MARGIN for the most reactive
(' configurations during the refueling may be performed to demonstrate acceptability- of the entire fuel movement sequence. Spiral offload / reload sequences inherently satisfy the surveillance requirement provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SHUTDOWN MARGIN. Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) O ABWR B 3.1-5 5/31/89
SHUTDOWN MARGIN B 3.1.1 BASES (continued) REFERENCES 1. ABWR SSAR, Section 15.4.1.1.
- 2. ABWR SSAR, Section 4.3.2.4.1.
- 3. NEDE-240ll-P-A-9, " General Electric Standard Application for Reload Fuel", September 1988. j NED0-31466, " Technical Specification Screening Criteria 4.
Application and Risk Assessment", November 1987. j i
)
O d I O ABWR B 3.1-6 5/31/89 L-_ ___ __ _ _ _ -
Control Rod OPERABILITY ' R B 3,1.2-I D- - B 3.1 REACTIVITY CONTROL SYSTEMS Q - B 3.1.2 Control Rod OPERABILITY ESES < BACKGROUND Control rods are components of the Control Rod Drive (CRD) system, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System (RPS), the CRD system provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occur-rences, specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a' malfunction in the CRD system. The CRD system is designad to satisfy the requirements of General Design Criteria 26, 27, 28 and 29.
. APPLICABLE The analytical methods and assumptions used in the evaluations S.AFETY involving control rods are presented in References 1, 2, 3 and ANALYSE 5 4. The control rods provide the primary means for rapid
[. reactivity control (reactor scram), for maintaining the reactor V subtritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD system. The capability to insert the control rods ensures the assumptions for scram reactivity in the design basis transient and accident analyses are not violated. Since the SHUTDOWN MARGIN ensures the reactor will be subtritical with the strong-est control rod pair withdrawn (assumed single failure of an HCU), the failure of an additional control rod to insert, if required, could invalidate the demonstrated SHUTDOWN MARGIN and y potentially limit the ability of the control rod drive system to hold tiie reactor subcritical. Therefore, the requirement that all control rods are OPERABLE ensures the CRD system.can perform its intended function. The control rods also protect the fuel from damage which could i result in release of radioactivity. The limits protected are l thu Safety Limit MINIMUM CRITICAL POWER RATIO (MCPR) (see Bases i for LCO 3.2.2, MCPR), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE) and the fuel damage limit (see Bases for LCO 3.1.6, Rod Pattern Control) during reactivity insertion events. (continued) l. ABWR B 3.1-7 5/31/89
-m_-__~-_a_ .-_._m
Control Rod OPERABILITY B 3.1.2 BASES (continued) APPLICABLE The negative reactivity insertion (scram) provided by the CRD SAFETY system provides the analytical basis for determination of plant ANALYSES thermal limits and provides protection against fuel damage (continued) limits during a Rod Withdrawal Error (RWE) event. Bases for LCO 3.1.3 through LCO 3.1.6 discuss in more detail how the Safety Limits are protected by the CRD system. Control Rod OPERABILITY satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6. i LCO OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the associated control rod accumulator status, the control rod coupling integrity and separation detection capability, and the ability tc determine the control rod position. Although not-all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is i required to satisfy the assumptions of the design basis transient and accident analyses. APPLICABILITY The control rod OPERABILITY requirements are applicable during MODES 1 and 2 whenever control rods may be withdrawn. In MODE 5, the OPERABILITY of withdrawn control rods is controlled by LC0 3.9.5. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LC0 3.10.3 (Control Rod Withdrawal - Hot Shutdown) and LCO 3.10.4 (Control Rod Withdrawal - Cold Shutdown) which provide adequate requirements for control rod OPERABILITY during these conditions. ACTIONS A.1 A control rod is considered stuck if it will not insert by either FMCRD drive motor torque or by scram pressure. The failure of a control rod to insert during SR 3.1.2.2 or SR 3.1.2.3 alone, however, does not necessarily mean that the i control rod is stuck, since failure of the motor drive would i also result in a failure of these tests. Verification of a stuck rod can be made by attempting to withdraw the rod. If the motor is working and the rod is actually stuck, the travel-ing nut will back down from the bottom of the drive and a rod separation alarm and rod block will result (see LCO [ Rod Block (continued) ABWR B 3.1-8 5/.11/89
i Control ' Rod ~ OPERABILITY B 3.1.2 pg Q BASES (continued)
'R ACTIONS .A_d (continued)
' (continued)
Instrumentation]). Conversely,.if the' motor' drive is known to
' be failed, the rod is not necessarily inoperable since it. is " probably still capable of . scram. However, at the next required.
performance of SR 3.1 2.2 or 3.1.2.3, there would be no way of verifying insertability, except:by' scram.- In this case, an individual scram should be attempted. If the rod scrams, the. roo is not stuck but should be considered inoperable and
. bypassed in RC&IS since it cannot be withdrawn and a separation.
situation will exist until the motor is repaired and the traveling nut is run-in to the full in position. If the rod fails to insert by individual scram, it should be considered stuck and the appropriate ACTIONS taken. The failure of a control rod pair.to insert is assumed in the design basis transient and accident analyses and therefore, with one with-drawn control rod stuck, some time is allowed to make the-control rod insertable. With a fully inserted control rod stuck, no actions ate reouired as long as the control rod remains fully inserted. As noted, a stuck control rod may be bypassed in the Rod Control and Information System (RC&IS) to allow continued operation. LC0 [ Control Rod Block Instrumen-tation] provides additional requirements when control rods are { bypassed in RC&lS to ensure compliance with the RWE analysis. B.1. B.2. B.3. B.4 With one withdrawn control rod stuck for more than the tilowed time, the control rod must be disarmed and isolated from scram pressure. The motor. drive r ay be disarmed by placing the rod in RC&IS Bypass or by manually disconnecting its power. supply. Isolating the control rod from scram prevents damage to the CRD and surrounding fuel assemblies should a scram occur. The control rod can be isolated from scram by isolating it from its associated hydraulic control unit. Two CRDs sharing an HCV can be individually isolated from scram. Below 10% of RTP, the' generic Banked Position Withdrawal-sequence (BPWS, or Ganged Withdrawal Sequence Restrictions, GWSR, for ABWR) analysis requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions including the diagonal. Out-of-sequence control rods may increase the potential reactivity worth of a control rod, or gang of control rods, during a RWE L and therefore the distribution of inoperable control rods must be controlled. O (cen14ne 8) l q ABWR B 3.1-9 5/31/89 o________. _ _ __ a
Control Rod OPERABILITY B 3.1.2 BASES (continued) ACTIONS B.1. B.2. B.3. 0.4 (continued) (continued)' With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions. Required Action B.3 of LC0 3.1.2 performs a step test on each remaining withdrawn control rod to ensure that no additional control rods are stuck. Therefore, 72 hours is allowed to perform the analysis or test in Required Action B.4. f.d If the Required Actions and associated Completion Times of Condition B cannot be met, the reactor must be in MODE 3 within 12 hours. Insertion of the remainder of the control rods eliminates the possibility of an a'Jditional failure of a control rod to insert. Prior demonstration of adequate SHUTDOWN MARGIN ensures the reactor can be held subcritical with only a single control rod withdrawn. D.1. D.2 With more than one withdrawn control rod stuck, the stuck control rods should be disarmed and isolated from scram pres-sure as described in the Bases for Required Action B.1 and the reactor must be in MODE 3 within 12 hours. The occurrence of more than one control rod without insertion capability may be an indication of a generic problem in the control rod drive system that could potentially cause additional failures of control rods to insert. Insertion of all insertable control rods eliminates the c',ssibility of an additional failure of a control rod to insect. E.1. E.2. E.3 With less than or equal to 8 control rods inoperable for any of the reasons discussed in LCO 3.1.3 through LC0 3.1.6, operation may continue provided the control rods are fully inserted and disarmed (however, they do not need to be isolated from scram). Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be disarmed by disconnecting power to the motor drive or by placing the rod in RC&IS bypass. Below 25% of RATED THERMAL POWER, the GWSR analysis requires (continued) ABWR B 3.1-10 5/31/89 _ __ _ }
1 Control Rod OPERABILITY l B 3.1.2 S_ASES (continue'd) ACTIONS . D.I. D.2 (continued)' With more than one withdrawn control ' rod stuck, the stuck: control rods should be disarmed and isolated from scram pres-sure as described in the Bases for Required Action B.1 and thel reactor must be in MODE 3 within 12. hours. The occurrence of. more than one control rod without insertion capability may be an indication of a generic problem in'the control rod drive system that could potentially-cause additional failures of control rods to insert. Insertion of all insertable control rods eliminates the possibility of an additional; failure of a control rod to insert. l E.1. E.2. E.3 With less than or equal to 8 control rods. inoperable for any of' the reasons discussed in LCO 3.3.3 through LC0 3.1.6 -operation may continue provided the control rods are fully inserted and disarmed (however, they do not need to be isolated from scram). Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequer,t operations. The control rods can be disarmed by disconnecting f power to the motor drive or by placing the rod in RC&l5 bypass. A Below 10% of RATED THERMAL POWER, the generic BPWS analysis requires inserted control rods, not in compliance with BPWS, to be separated by at least two OPERABLE control rods in all directions including the diagonal (Ref. 5). Inserted out-of-sequence control rods may increase the potential reactivity worth of other- control rods during an RWE and therefore the number and distribution of inserted inoperable control rods must be controlled. As noted, the control rods may be bypassed in the RC&lS if required to allow insertion of the inoperable control rods and continued operation. Also, as noted, control rods declared inoperable with a failed motor drive can only be inserted by scram. Control rods with failed motor drives are not inoperable for this reason alone, but must be con-sidered so upon f ailure of SR 3.1.2.2 or SR 3.1.2.3, or when not in compliance with GWSR (see LCO 3.1.6). LCO [ Control Rod Block Instrumentation] provides additional requirements when the control rods are bypassed to ensure compliance with the RWE analysis. (continued) O I ABWR B 3.1-11 5/31/89
Cont'rol Rod OPERABILITY B 3.1.2 EASES (continued) ACTIONS L1 If the Required Actions and associated Completion Times of Condition E are not met or more than 8 inoperable control rods exist, the reactor is required to be in MODE 3 within 12 hours. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i..e., scram or insertion) of the control rods. The number of control rods permitted to be inoperable when operating above 10% of RATED THERMAL POWER could be more than the value specified, but the occurrence of a large number of i inoperable control rods could be indicative of a generic problem and investigation and resolution of the potential l problem should be done. Corroletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. l SURVEILLANCE SR 3.1.2.1 l REQUIREMENTS l Determining the position of each control rod is required to l ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. SR 3.1,2.2. SR 3.1.2.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one step and observing that the control rod moves. The control rod may l then be returned to its original pos'ition. These surveillance are not required when below the actual LPSP of the RC&IS since the step insertions may not be compatible with the requirements of Rod Pattern Control (LCO 3.1.6) and the RC&IS (LCO [ Control Rod Block Instrumentation]). Partially withdrawn control rods are nnt tested weekly because of the potential power reduction I required to allow the control rod movement. (continued) O ABWR B 3.1-12 5/31/89 i L_ _ _ _ _ _ . _ . _ _
Control Rod OPERABILITY B 3.1.2 BASES (continued) SURVEILLANCE Surveillance Frequencies I REQUIREMENTS (Continued) In general, surveillance frequencies are based on industry I accepted practice and engineering judgement considering the ! unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the l system / component status. l REFERENCES 1. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel", September 1988.
- 2. ABWR SSAR, Section 4.6.1.
- 3. ABWR SSAR, Section 7.7.1.2.
- 4. ABWR SSAR, Section 15.4.1.
- 5. NED0-21231, " Banked Position Withdrawal Sequence",
January 1977, Section 7.2.
- 6. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
O
Control Rod Scram Times B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod Scram Times BASES BACKGROUND The scram function of the Control Rod Drive (CRD) system reliably controls reactivity changes during abnormal opera-tional transients to ensure specified acceptable fuel design limits are not exceeded (Ref. 1). The control rods are scrammed by positive means using hydraulic pressure exerted on the control rod drive piston. APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY control rod scram function are presented in References 2, 3, 4 ANALYSES and 5. The design basis transient and accident analyses assume all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., MINIMUM CRITICAL POWER RATIO (MCPR)). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the design basis transient and accident analyses can be met. The scram function of the CRD system prctects the Safety Limit MCPR (see Bases for LC0 3.2.2) and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1) which ensure no fuel damage if the limits are not exceeded. Above 950 psig the scram function is designed to insert negative reactivity at a rate fast enough to prevent the MCPR from becoming less than the Safety Limit MCPR during the limiting power transient analyzed in SSAR Chapter 15. Below 950 psig the scram function is assumed to function during the Rod Withdrawal Error (RWE) event (Ref. 5) and therefore also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LC0 3.1.6). For the SSAR Chapter 5 vessel overpressure protection analysis, the scram function along with the safety / relief valves ensures the peak vessel pressure is maintained within the ASME Code limits. Control Rod Scram Times satisties the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6. (continued) ABWR B 3.1 14 5/31/89
,1 j E Control Rod Scram Times y
B 3.1.3 , f BASES'(continuedi' LCO _The scram times specified in Table 3'.l.3-l' are required to~ ensure the scram reactivity assumed in the design basis trans-ient and accident analysis 'is met. : To account for single failures and " slow'. scramming control rods, the times specified in Table 3.1.3-1 are faster than those assumed in the design basis analysis. The scram times have margin' to allow.up to [ ] of the control. rods to have scram times exceeding.the
. limits (i.e., " slow" control rods) and also account for a single stuck control rod (as; allowed by;LC0 3.1.2) and an additional control rod pair (single failure criterion) failing to scram. The scram times tre specified as a function of reactor steam dome' pressure to account:for the pressure dependence of scram times.
-APPLICABILITY. The CRD scram function is' applicable during MODES' l-and 2 since a scram is only required when control rods are withdrawn, and.-
is assumed to function during transients and accidents analyzed in these conditions. These events are assumed to occur during
.startup and power operation ~. In MODE 5, the scram capability of withdrawn control rods is specified by LC0'3.9.5. In MODES 3:and 4, control. rods are only allowed.to be withdrawn-under p Special Operations LCO 3.10.3 (Control Rod Withdrawal - Hot.
Shutdown) and LC0 3.10.4- (Control Rod Withdrawal - Cold Shut-down) which provide adequate. requirements for control rod scram capability during these conditions. ACTIONS A.] for a control rod with excessive scram times (2 [ ] seconds to 60% insertion from de-energization of scram pilot valve sole-noids as time zero) the control rod must 'be declared inoperable and therefore the control rod would be fully inserted and - disarmed as required by LCO 3.1.2. Insertion of the. control rod ensures the scram reactivity is not adversely affected by the failure of the control rod to scram. An example of a control rod with excessive scram times would be a control rod with a scram solenoid pilot valve that fails to open upon receipt of a scram. signal. (continued) ABWR B 3.1-15 5/31/89 e
l
)
Control Rod Scram Times B 3.1.3 BASES (continued) ACTIONS 11 (continued) SR 3.1.3.2 is a periodic test to sample the control rod scram times during the cycle to ensure the scram times have not degraded. Because only a sample of the control rods is tested . (;t 10%). a separate limit on the number of allowed " slow" control rods in the sample (20%) is specified. This limit is chosen such that if it is exceedeo, it is an indication the ! total population of control rods, if Msted, would exceed the allowed number of " slow" control rods. Therefore, following the completion of SR 3.1.3.2 during which one or more control rods are discovered to be " slow", the number of " slow" control ' rods in the sample must be determined and verified to be less than the sample limit. More than 10% of the control rods may be tested in SR 3.1.3.2 to provide a more representative sample. If a control rod is discovered " slow" by mear.s other than SR 3.1.3.2, this Required Action is not required. A.3 A.4 4 With a control rod whose scram time exceeds the scram time limits (a " slow" control rod) but is less than [ ] seconds, continued operation is justified if the number and distribution of " slow" control rods is consistent with the assumptions of the design basis transient and accident analyses. The scram times of Table 3.1.3-1 are based on [ ] " slow" control rods, of which only two may occupy adjacent locations in any direction and therefore no degradation of the design basis scram reactivity exists if these conditions are met. If scram time data already exist for the surrounding control rods, no additional testing is required to determine if adjacent control rods are " slow". Control rods determined to be " slow" may alternatively be declared inoperable and the actions of LCO 3.1.2 followed. Inoperable control rods are not included in determining compliance with the requirements of Required Actions A.3 and A.4. B.1 Multiple adjacent " slow" control rods or an excessive number of
" slow" control rods can reduce the local scram reactivity relative to that assumed in the desigc basis transient and accident analyses. Therefore, the reactor is required to be in MODE 3 within 12 hours. Also, if an unacceptable number of control rods are determined to be " slow" during SR 3.1.3.2, suffi'ient degradation of scram reactivity may be present and the r actor must be in MODE 3 within 12 hours. Insertion of all control rods places the reactor in a condition that does not require the scram function.
(continued) ABWR B 3.1-16 5/31/89
Control Rod Scram Times. B 3.1. 3 f'Y BASES (continued) KJ ACTIONS Comoletion Tirres (continued) All Completion Times are based on industry accepted practice t and engineering judgement considering the number of available l . systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The scram reactivity used in design basis transient and acci-dent analyses is based on an assumed scram speed. Measurement of the scram times.with reactor steam dome pressure greater than 950 psig demonstrates acceptable scram times for the. transients analyzed in References 4 and 5. Scram insertion times increase with increasing reactor pressure because of the competing effects of reactor. steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure greater than 950 psig ensures the scram times will be within the specified limits at' higher pressure. Limits are specified as a function of reactor-pressure to account for the sensitivity of scram insertion
, times with-pressure and to allow a range of pressures during which scram time testing can be performed. To ensure scram time testing is performed in a reasonable time following a refueling or after a shutdown greater than 120 days, all control rods are required to be tested before exceeding 40*; of' RATED THERMAL POWER following the shutdown.
SR 3.1.3.2 Additional testing of at least a 10*/ sample of control rods is required every 120 days of cumulative operation in MODE 1. For planned testing, the control rods selected for the 105, sample should be different for each test. Data from inadvertent scrams should be used if possible to avoid. unnecessary testing at power even if the control rods with data may have been previously tested in a 10*/, sample. This frequency and number of tested control rods is based on engineering judgement considering the desire to minimize disturbances to normal plant operation, experience which shows scram times do not significantly change over an operating cycle, and the additional surveillance done on the control rod drives at more frequent intervals (LCO 3.1.2 and LC0 3.1.4). Testing of more than 10% of the control rods may be done to obtain a more representative sample. (continued) ABWR B 3.1-17 5/31/89
___-___7 Contro' Rod Scram Times B 3.1.3 BASES (cnntinued) SURVEILLANCE SR 3.1.3.2 (continued) O REQUIREMENTS (continued) When possible, scram insertion time data can be determined from full reactor core scrams. Otherwise, the scram time data is obtained during individual rod pair scrams. As noted, for testing during rod pai scrams, the test shall be performed with the charging valve closed so that the influence of the CRD pump head does not affect the control rod pair under test (during a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times). SR 3.1.3.3. SR 3.1.3.4 For FSAR Chapter 15 analyses in the startup range (primarily the RWE), low pressure ( < 950 psig) scram times are assumed to j be the same as the scram time limits at high reactor pressure f (= 950 psig). This provides assurance that the low pressure scram times will be acceptable if the more stringent high pressure scram time limits are met during surveillance testing ! at 2 950 psig. Should work on a control rod or the CRD system potentially affect the scram insertion time, testing must be I done to demonstrate adequate scram performance over the range. of applicable reactor pressures. Specific examples of work that.could affect the scram times are (but not limited to) the following: removal of any control rod drive for maintenance or modification, replacement of a control rod, and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator or isolation / check valves in the piping required for scram. For work done while the reactor is at less than 950 psig, the scram testing must be performed before declaring the control rod OPERABLE and again at reactor steam dome pressure greater than or equal to 950 psig. This testing ensures prior to withdrawing the control rod for continued operation, the control rod scram function will be demonstrated both for startup and high reactor pressure conditions. Where work has been performed at high reactor pressure, both of these require-ments will be satisfied with one test. However, for a control rod affected while shutdown, a zero pressure and high pressure test may be required. Alternatively, a test during hydrostatic pressure testing could also ratisfy both criteria. To account for the variability in scr. t imes at different reactor pres-sures, specific scram time limits are specified in Table 3.1.3-1 at high reactor pressure conditions. (continued) O ABWR B 3.1-18 5/31/89 L - - - - _ _ - __
)
Control Rod Scram Times B 3.1.3 BASES (continued) l
-(]*
Q, SURVEILLANCE Surveillance Frequencies ! I REQUIREMENTS (continued) In general, surveillance frequencies are based on industry l accepted practice and engineering judgement considering the l unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. 10 CFR Part 50, Appendix A, General Design Criterion 10.
- 2. NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel", September 1988.
- 3. ABWR SSAR, Section 4.6.1.
- 4. ABWR SSAR, Chcpter 15,
- 5. ABWR SSAR, Section 15.4.1
- 6. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
,m O
ABWR B 3.1-19 5/31/89
Control Rod Scram Accumulators B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Accumulators O BASES BACKGROUND The control rod drive (CRD) scram accumulators are part of the CRD system and are provided to ensure the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert two control rods at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.3. APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY control rod scram function are presented in References 1, 2, 3 ANALYSES and 4. The design basis transient and accident analyses assume all of the control rods scram at a specified insertion rate. OPERABILITY of each individual control rod scram accumulator ensures (along with LC0 3.1.2, LC0 3.1.3 and LC0 3.1.5) the scram reactivity assumed in the design basis transient and & accident analyses can be met. The existence of an inoperable W accumulator may invalidate prior scram time measurements for the associated control rods. The scram function of the CRD system, and therefore the OPERABILITY of the accumulators, protects the Safety Limit MINIMUM CRITICAL POWER RATIO (llCPR) (see Bases for LC0 3.2.2) and the 1% cladding plastic strain fuel design limit (see Bases for LC0 3.2.1) which ensure no fuel damage if the limit is not exceeded (see Bases for LC0 3.1.3). Also, the scram function at low reactor vessel pressure (startup conditions) provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LC0 3.1.6). Control Rod Scram Accumulators satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 5. (continued) O ,
\
ABWR B 3.1-20 5/31/89 a
l ' Control Rod Scram Accumulators B 3.1.4 (N- BASES'(continued) U . LC0 The OPERABILITY of the control rod scram accumulators is required to ensure adequate scram insertion. capability exists when needed over the entire range of reactor pressures. APPLICABILITY The control rod scram accumulators are required to be OPERABLE during MODES 1 and 2 whenever ' control rods are withdrawn and the scram function may be required. In MODE 5, the required-OPERABILITY of. accumulators associated with withdrawn control rods is specified in LC0 3.9.5. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LC0 3.10.3. (Control Rod Withdrawal - Hot Shutdown) and LC0 3.10.4 (Control Rod Withdrawal ' Cold Shutdown) which provide adequate requirements for control rod scram accumulator OPERABILITY during these conditions ACTIONS A.l. A.2 The accumulator provides the primary scram force during opera-tion at any reactor steam dome pressure. However, because of s the large number of control rods available for scram and the assumed single failure of an HCU in the safety analysis, time is allowed to restore a single failed accumulator to OPERABLE status. If the accumulator cannot be restored to OPERABLE
?.atus within the required Completion Time, the associated cwrol rods must be declared inoperable and the requirements or LCO 3.1.2 followed. This will require insertion of the control rod thereby completing the intended function of the control rod.
Ikl With two or more control rod scram accumulators inoperable, continued operation can only be justified for a short time. The importance of the accumulators in providing the scram force is such that the scram function could become severely degraded should more than one of the scram accumulators, associated with fully or partially withdrawn control rods, be inoperable at the same time. Tle most likely cause for multiple control rod scram accumu-lators to be inoperable would be if there was not a CRD pump in operation supplying adequate pressure to keep the accumulators charged. However, the RPS will initiate a scram on sensed low I 1
-f '-
(continued) ABWR B 3.1-21 5/31/89 ] l l l
l Control Rod Scram Accumulators ' B 3.1.4 i BASES (continued) ACTIONS L1 (continued) (continued) pressure in the CRD charging water header so that all rods are inserted prior to reaching a level of degradation beyond the safety analysis assumptions. Of course there may be situations where two or more accumu-lators are inoperable for other reasons. In this case, a short time is allowed for corrective action to be taken (consistent with A.1 or A.2). However, if multiple accumulators associated with withdrawn control rods exist simultaneously for longer than the required completion time, then the Reactor Mode Switch must be placed in the Shutdown position. This ensures all insertable control rods are inserted and the reactor is in a condition that does not require the active function (scram and insertion) of the control rods. Completion Times All Completion Times are baseJ on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.1.4.1 REQUIREMENTS The primary indicator of accumulator OPERABILITY is the accumu-lator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of { 1850 ) psig is well below the expected pressure of approxi-mately 2150 psig. Declaring the accumulator inopera'le o when the minimum pressure is not maintained ensures significant degradation in scram times does not occur. SR 3.1.4.1 requires the accumulator pressure be checked weekly to ensure adequate accumulator pressure exists to provide sufficient scram force. Operating experience has demonstrated that a seven day frequency for this surveillance is adequate. (continued) O ABWR B 3.1-22 5/31/89
Control Rod Scram Accumulators B 3.1.4 BASES (continued)
- 1. NEDE-24011-P-A," General Electric Standard Application b REFERENCES for Reactor Fuel", September 1988.
J
- 2. ABWR SSAR, Section 4.6.1.
- 3. ABWR SSAR, Chapter 15.
- 4. ABWR SSAR, Section 15.4.1.
1
- 5. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
i i O L O ABWR B 3.1-23 5/31/89
4 i CRD Coupling B 3.1.5 j B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control Rod Drive Counling BASES i BACKGROUND The control rod drive (CRD) and control rod are coupled through a bayonet-type coupling at the top end of the CR0 where the drive tube locks into a mating socket at the base of the l control rod. The coupling requires a 45' rotation for either engaging or disengaging. This positive means of coupling the control rod to the drive tube allows the position of the control rod to be varied and detected through the movement of the traveling nut and drive tube. CRD coupling is a requirement for control rod OPERABILITY since the lack of coupling could result in loss of control rod position information and potentially result in a large reactivity insertion should the control rod become stuck and later drop from the core. However, this could only occur in conjunction with other failures such as failure of the redundant Class lE separation detection devices (see LC0 [ Rod Block Instrumentation]). APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY CRD system are presented in References 1, 2 and 3. The CRD f ANALYSES system provides the negative scram reactivity required to ensure specified acceptable fuel design limits are ;ot exceeded during abnormal operational transients (see Bases for LC0 3.1.3). Also, if a control rod is not coupled, the position of the control rod cannot necessarily be determined. The control rod ! position (specifically the control rod pattern) is an initial condition of the Rod Withdrawal Error (RWE) event analysis. Demonstrating coupling of all control rods therefore provides protection against violating fuel damage limits during reactivity initiated accidents (see Bases for LCO 3.1.6). Additionally, the positive means of cour'*ng, along with the use of redundant separation detection devices, precludes the possibility of occurrence of a Control Rod Drop Accident. Control Rod Drive Coupling satisfies the retirements of Selectior, Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 4. (continued) ABWR B 3.1-24 5/31/89
- _ - - _ _ _ - _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ ____-_ - -_7__ - ___
CRDCoubling-B 3.1~.5 (BASES-(continued) LC0 The requirement for all control rods to-be coupled to their drive mechanisms is required to ensure OPERABILITY of the control rods such that their. intended reactivity control functions can be performed. APPLICABILITY The CRD coupling requirements are applicable during MODES 1: and 2 since under these conditions control rods or control ' rod. pair can be withdrawn and. criticality achieved. The CRDA is not applicable during MODES 3, 4 and 5.since only a single control rod or control rod pair can be withdrawn from a core cell containing fuel assemblies and SHUTDOWN MARGIN ensures -the reactor'is'subtritical under these conditions. ACTIONS .A.1. A.2 Continued operation with an uncoupled cont'rol. rod should no't be v allowed because of the increased probability of a CRDA and . therefore only a short period of time is allowed to establish and verify recoupling. Since the allowable time with :an~ . FN uncoupled. control rod is short, and control rods do not always V recouple'on the first try, multiple attempts to recouple a. control rod may be necessary. If recoupling cannot be achieved, the control rod must be - declared inoperable and the requirements of LCO 3.1.2 must be satisfied. This will ensure if the control rod cannot be recoupled in a short time, it will be inserted 'and disarmed thereby maintaining the shutdown capabilities of the. control rod. A CRDA involving this uncoupled control rod would still be precluded by the redundant separation detection devices. However, to reduce the probability that the rod may brcome stuck, it should be inserted and disarmed. Additionally, to prevent potential damage to the rod / drive coupling, after insertion, the control rod should be isolated from scram. The control. rod can be isolated from scram by isolating it from its associated hydraulic control unit and the CRD charging water header. (continued) O 'O ABWR B 3.1-25 5/31/89
CRD Coupling-B 3.1.5 l BASES (continued) ACTIONS Completion Times (continued) All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Coupling verification is performed by verifying a control rod does not go to the overtravel position when it is fully with-drawn. The overtravel pasition feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position. (Since the traveling nut will go to the overtravel position, it will be temporarily separated from the rod and drive as the control rod is backseated at the full out position. Therefore, this test also allows verifica-tion of the proper functioning of the redundant separation detection devices. See LC0 [ Control Rod Block Instrumentation].) This verification is required to be performed the first time a control rod is withdrawn to the
" Full Out" position following refueling and prior to declaring the control rod OPERABLE when work on the control rod or CRD system could affect coupling.
O REFERENCES 1. NEDE-24011-P- A-9-US, " General Electric Standard Application for Reactor Fuel", Supplement for United States, September 1988.
- 2. ABWR SSAR Section 4.6.1.
- 3. ABWR SSAR, Section 15.4.1. 1
- 4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
O ABWR B 3.1-26 5/31/89
l Rod Pattern Control D B 3.1.6
'b V
'B 3.l' REACTIVITY CONTROL SYSTEMS B 3.I' 6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup. conditions are controlled by the operator and the Rod Worth Minimizer (RWM) (LCO [ Control Rod Block Instrumentation)), such'that only specified control-rod sequences and relative positions are allowed over the operating range from all control' rods inserted to 10% of RATED THERMAL POWER. The sequences effectively limit the potential amount.and rate of reactivity increase that could occur during a control rod withdrawal, specifically the Rod Withdrawal Error (RWE) event.
APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY- RWE are summarized in References 1 and 2. RWE analyses assume ANALYSES the reactor. operator follows prescribed withdrawal sequences. These sequences define the potential initial. conditions for the RWE analysis. The RWM (LCO [ Control Rod Block Instruments-tion)) provides backup to. operator control of the withdrawal
'L sequences to ensure tbc initial conditions of the RWE analysis
-. are not violated.
Control rod patterns analyzed in References 1 and 2 follow the GWSR which is the same as the BPWS described in Reference 6. The GWSR is applicable from the condition of all control rods fully inser. d to 10% of RATED THERMAL POWER. For GWSR, the control rods are reouired to be moved in groups, with all control rods assigned to a specific group required to be within specified Lanked positions. The banked positions are defined to minimize the maximum incremental control rod worths without being overly restrictive during normal p) ant operation. Prevention or mitigation of positive react? .ty insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel. damage which could result in undue release of radioactivity (Ref. 5). Since the failure conse-quences for U07 have been shown to be insignificant below fuel energy depositions of 300 cal /gm, the fuel damage limit of 280 cal /gm provides a margin of safety to significant core damage and release of radioactivity (Ref. 3 and 4). Generic analysis of the GWSR (BPWS, Ref. 6) has demonstrated the 280 cal /gm fuel damage limit will not be violated during a postulated reactiv-ity transient while following the GWSR mode of operation. The generic analysis also evaluated the effect of fully inserted inoperable control rods not in compliance with the sequence to i - allow a limited number (8) and distribution of fully inserted inoperable control rods. (continued) ABWR B 3.1-27 5/31/89 u- - - _ - - _ -- - _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _d
)
i Rod Pattern Control i I B 3.1.6 l BASES (continued) [ APPLICABLE Rod Pattern Control satisfies the requirements of Selection SAFETY Criterion 3 of the NRC Interim Policy Statement on Technical l ! ANALYSES Specification Improvements as documented in Reference 7. (continued) i LC0 Compliance with the prescribed control rod sequences minimizes the potential consequences of a RWE by limiting the initial conditions to those consistent with GWSR. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LC0 3.1.2, consistent with the allowances for inoperable control rods in GWSR. i APPLICABILITY 1' Compliance with GWSR is required in MODES 1 and 2 when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a RWE. In MODES 3, 4 and 5, since only a total of one control rod or control rod pair can be withdrawn from core cells containing fuel assemblies, adequate SHUTDOWN MARGIN ensures the reactor will remain subtritical. (continued) O ABWR B 3.1-28 5/31/89
la 1 Rod Pattern Control B 3.1.6 G- BASES (continued)' ._
.L) ACTICNS A.1. A.2 m
With a limited ni oer of OPERABLE control rods not in compli-ance with the pic cribed control rod sequence, actions may be-taken to correct ,he control rod pattern. Noncompliance with the prescribed quence may be .the result of _ failed.synchos, drifting from a CRD purge water transient, leaking scram valves or a power reduction below 10% of RATED THERMAL POWER before establishing the correct controlfrod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequen'ce is limited to 8 to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for scram or those~ moves needed to correct the rod pattern. As noted, control rods may be bypassed in RC&lS to allow the affected control rods to be returned to their correct position. This, ensures the control rods will-be moved to the correct position. Alternatively, if the affected control rod is not moved to its correct position in the Required Completion Time, it may be declared inoperable and the requirements of LC0 3.1.2 followed.
- A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance
- with LCO 3.1.2 through LCO 3.1.5.
B.1. B.2 More than 8 out-of-sequence OPERABLE control rods indicates the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended to prevent the potential for further deviation from the prescribed sequence. Limited (1 hour) control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worths than withdrawals beyond the prescribed limits. If the control rod pattern requirements cannot be restored within I hour, the Reactor Mode Switch must be placed in the Shutdown position. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. (continued) ABWR B 3.1-29 5/31/89
Rod Pattern Control B 3.1.6 BASES (continued) p SURVEILLANCE SR 3.1.6.1 , REQUIREMENTS i The primary check on compliance with GWSR is performed by the RWM (LC0 [ Control Rod Block Instrumentation]) which provides control rod blocks to enforce the required sequence. The RWM is required to be OPERABLE when operating at less,than or equal to 10% of RATED THERMAL POWER. Should a control rod be bypassed in RC&lS, the RWM will not block movement of this control rod and therefore LCO [ Control Rod Block Instruments-tion) requires the bypassing and movement of the bypassed control rod to be verified under these conditions to ensure the bypassed control rod is returned to its correct position. Therefore, a daily check of the control rod pattern compliance with GWSR is adequate. REFERENCES 1. ABWR SSAR, Section 15.4.1.
- 2. NEDE-24011-P-A-9-US, " General Elsetric Standard Application for Reactor Fuel - Supplement for United States", Septebmer 1988.
- 3. NUREG-0800, Revision 2, Standard Review Plan, July 1981, Section 15.4.1.
- 4. Regulatory Guide 1.77.
- 5. 10 CFR Part 100.
- 6. NE00-21231, " Banked Position Withdrawal Sequence", January 1977.
- 7. NED0-31466, " Technical Specification Screening Criteria ,
Application and Risk Assessment", November 1987. ! O ABWR B 3.1-30 5/31/89
ht , ' g j ps SLCS' ]
~
B 3.1.7 [ D B 3J REACTIVITY. CONTROL SYSTEMS-b' B 3.1.7 Standby Liouid Control System i l I BASES . < BACKGROUND. The Standby Liquid' Control System.(SLCS) is designed to provide l the capability of bringing the reactor, at any time in a cycle, from full power and minimum control rod inventory (which is defined to be at the peak of the Xenon transient) to a subcritical condition with the reactor in: the' most reactive. . xenon-free state without control rod movement. The system is' '! N designed for the conditions when no control rods can be ! inserted from full power conditions. The SLCS consists of a boron solution storage tank, two 4 positive displacement pumps, two motor operated injection valves which are provided in parallel for redundancy and associated piping and valves used to transfer borated water from the storage. tank to the reactor pressure. vessel (RPV).
-The. borated solution is discharged through the 'B' High Pressure Core Flooder (HPCF) subsystem sparger.
APPLICABLE The SLCS is manually initiated from the main control room as
. SAFETY -- directed by the Emergency Operating Procedures if the operator s- ANALYSES believes the reactor cannot be shutdown or kept shutdown with the control rods. The' SLCS is used in the highly improbable event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner. The SLCS provides borated water to the reactor core to compensate for the various reactivity effects during the required condi-tions. To meet this objective, it is necessary to inject.a quantity of boron which produces.a concentration of 850 ppm of natural boron in the reactor core at 68'F. To allow for potential leakage and imperfect mixing in the reactor system, an additional 25% (220 ppm) is added to the above requirement-(Ref. 1). The volume and concentration limits in Figure 3.1.7-1 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the volume in the Residual Heat Removal shutdown cooling piping (additional 250 ppm). This quantity of solution (with a total required concentration of 1320 ppm)' is the amount which is above the pump suction shutoff level in the tank thus allowing for the portion of the tank volume which cannot be injected.
(continued) 1 ABWR B 3.1-31 5/31/89
SLCS B 3.1.7 BASES (continued) APPLICABLE The NRC Interim Policy Statement (Ref. 2) requires the SLCS be SAFETY retained in the technical specifications even though none of ANALYSES the Selection Criteria were satisfied (Ref. 3). (continued) LC0 The OPERABILITY of the SLCS is required to provide backup capability for reactivity control independent of normal reactivity control provisions from the control rods. The OPERABILITY of the SLCS is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the reactor pressure vessel, including the OPERABILITY of the pumps and valves. Two SLCS subsystems are required OPERABLE, each containing an OPERABLE pump, an explosive valve and associated piping and valves to ensure an OPERABLE flow path. APPLICABILITY The' SLCS specifications are applicable during MODES 1 and 2 since during these conditions the reactor can be critical. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LC0 3.10.3 (Control Rod Withdrawal - Hot Shutdown) and LCO 3.10.4 (Control Rod Withdrawal - Cold Shutdown) which provide adequate controls to ensure only a single control rod or control rod pair can t withdrawn. In MODE 5, only a single control rod or contro'i .. pair can be withdrawn from core cells containing fuel assemblies, and demonstration of adequate SHUTDOWH MARGIN (LC0 3.1.1) ensures the reactor will not be critical. Therefore the SLCS is not required to be OPERABLE during these conditions when only a single control rod or control rod pair can be withdrawn. ACTIONS A.1 With one SLCS subsystem inoperable, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem can result in no SLCS shutdown capability. For this reason, continued operation is permitted for a limited time only, 7 days. (continued) O ABWR B 3.1-32 5/31/89
1 $ SLCS B 3.1.7-
- BASES (continuedE ACTIONS B.1. B.2 (continued)' .
With both SLCS subsystems inoperable, no SLCS shutdown
. capability remains to handle the postulated event. Continued operation is justified because of the' low probability of an event which would require SLCS. However, continued operation of the plant is permitted for only a limited period of time, 8-hours, during which time at least one subsystem must be restored to OPERABLE status. Additionally, the initial inoperable subsystem must be restored to OPERABLE status within 7 days from. initial discovery consistent with the Completion Time of Required Action A.I.
L.1 With one or both SLCS subsystems not restored to 0PERABLE status and the associated Completion Times not met, the reactor must be in MODE 3 within 12 hours. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the O Required Action. SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3 ).7.3 are daily surveillance verifying portions of the SLCS OPERABILITY without disturbing normal plant operation. The surveillance ensure the proper solution volume and temperature (including the temperature of the pump-suction piping) are maintained. The solution temperature is important in ensuring the boron remains in solution and does not precipitate out in the storage tank or pump suction piping. Failure to meet SR 3.1.7.1, SR 3.1.7.2 or SR 3.1.7.3 will make both SLCS subsystems inoperable since the storage tank and the majority of the pump suction piping is common to both subsystems. 1 (continued) O ABWR B 3.1-33 5/31/89 i
1 SLCS ! B 3.1.7 l BASES (continued) SURVEILLANCE SR 3.1.7.4. SR 3.1.7.5 REQUIREMENTS (continued) SR 3.1.7.4 and SR 3.1.7.5 provide additional demonstration of the OPERABILITY of the SLCS. SR 3.1.7.4 provides a detailed examination of the sodium pentaborate solution by using chem-ical analysis to ensure the proper concentration of boron solution exists in the tank. SR 3.1.7.5 verifies each valve in the system (not otherwise locked, sealed or secured in posi-i tion) is in its correct position.to ensure the flow path is available to support operation of the subsystem. SR 3.1.7.4 must be performed anytime boron or water is added to the tank solution to establish acceptability of the new values for the solution concentration. Al so, SR 3.1.7.4 must be performed anytime the temperature is restored to within the limits of Figure 3.1.7-1 to ensure no significant boron precipitation occurred. SR 3.1.7.6 SR 3.1.7.6 demonstrates proper operation of the SLCS pumps and requires each pump be demonstrated to meet the minimum flow rate requirement of 50.0 gpm at a discharge pressure of 1223 psig. The minimum pump flow rate requirement ensures that when combined with the sodium pentaborate solution concentra-tion requirements, the rate of negative reactivity insertion from the SLCS will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator and xenon decay. SR 3.1.7.7. SR 3.1.7.7 SR 3.1.7.7 and SR 3.1.7.8 ensure a complete flow path from the storage tank to the reactor pressure vessel. The pump and injection valve tested should be alternated such that both complete flow paths are tested every 36 months. Surveillance Frequencies in general, surveillance frequencies a e based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) e ABWR B 3.1-34 5/31/89
I SLCS i B 3.1.7 - j BASES (continued)
*(/ )-'
REFERENCES 1. ABWR SSAR, Section 9.3.5.3.
- 2. 52FR3788, " Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,"
February 6, 1987.
- 3. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
1 0
,-~.
V ABWR B 3.1-35 5/31/89
APLHGR B 3.2.1 , n B 3.2 POWER DISTRIBUTION LIMITS
, B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE BASES BACKGROUND The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location. Limits on APLHGR are specified to assure that the fuel design limits identified in Reference I will not be exceeded during anticipated operational occurrences and that the peak cladding temperature (PCT) during the postulated design basis loss-of-coolant accident (LOCA) will not exceed the limits specified in 10 CFR 50.46.
APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY fuel design limits are presented in SSAR Chapter 4 and in ANALYSES Reference 1. The analytical methods and assumptions.used in evaluating Design Basis Accidents, anticipated operational transients and normal operation that determine the APLHGR limits are presented in the FSAR, Chapters 4, 6 and 15 and in Reference 1. r
-h '
Fuel design evaluations are performed to demonstrate that the cladding 1% plastic strain and other fuel design limits described in Reference 1 are not exceeded during anticipated operational occurrences for operation with LINEAR HEAT GENERATILJ RATES (LHGR's) up to the operating limit LHGR. APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adher-ence to fuel design limits during the limiting anticipated operational occurrences. F low dependent APLHGR limits are determined using the three dimensional BWR simulator code (Ref. 2) to analyze slow flow runout transients. The flow-dependent multiplier, MAPFACf , is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power-dependent multipliers (MAPFAC p
) are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below that where turbine stop valve
/
~f (continued)
ABWR B 3.2-1 5/31/89
APLHGR J B 3.2.1 BASES (continued) APPLICABLE closure and turbine control valve fast closure scram trips are SAFETY bypassed both high and low core flow MAPFAC p limits are ANALYSES (continued) provided for operation at power levels between 25% of RATED THERMAL POWER and the previously mentioned bypass power level. The exposure dependent APLHGR limits are reduced by MAPFACp and MAPFAC at various operating conditions to ensure that all fuel I f i design criteria are met for normal operation and anticipated j operational occurrences. A complete discussion of the analysis ; code is provided in Reference 4. ! LOCA analyses are then performed to ensure the above determined APLHGR limits are adequate to meet the PCT and maximum oxida-tion limits of 10 CFR 50.46. The analysis is performed using GE calculational models which are consistent with the require-ments of 10 CFR 50, Appendix K. A complete discussion of the analysis code used in the analysis is provided in Reference 1. The PCT following a postulated LOCA is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided l by its local peaking factor. A conservative multiplie. is i applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of APLHGR. APLHGR satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. ' LC0 The APLHGR limits specified in the [ CORE OPERATING LIMITS REPORT] are the result of the fuel design analysis and design basis accident and transient analysis. The limit is determined by multiplying the smaller of the MAPFAC and MAPFAC p factors times the exposure dependent APLHGR limi s (continued) O ABWR B 3.2-2 5/31/89
'APLHGR B 3.2.1
[. BASES'(continued)
~' APPLICABILITY. The APLHGR limits are primarilv' derived from fuel design evaluations, LOCA an6 transientianalysis that are assumed to occur from high power level conditions. Design calculations and operating experience have shown that'as power is reduced,
! margin to required APLHGR limits increases. This trend con-tinues down to the power range of 5-15% of RATED THERMAL POWER (RTP) where entry into MODE 2 occurs. When in MODE 2, the Startup Range Neutron Monitor (SRNM) scram function will' provide prompt scram initiation during any significant transi-
> ent thereby effectively removing any APLHGR' limit compliance concern in MODE 2. ,Therefore, at THERMAL POWER levels less than or equal to 25% of RTP, the reactor will be operating with substantial margin to APLHGR limits and the specification is-not required.
ACTIONS M Should any APLHGR exceed the required limits, an. initial condition of the design basis accident and transient analyses-may not be inet. Therefore, prompt action should be taken to restore the APLHGR's to within the required limits such that the plant will be operating within analyzed conditions and within design limits of th fuel rods. u If the APLHGR cannot be restored to within the required limits in two hours, it is required to reduce THERMAL POWER to < 25% of RTP. As discussed in the Bases for Applicability, operation below 25% of RTP results in sufficient margin to the required limits. Completion Times The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably complete the Required Action. (continued) O , ABWR B 3.2-3 5/31/89
APLHGR B 3.2.1 BASES (continued) SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGR's are required to be initially calculated within 12 hours after THERMAL POWER has exceeded 25% of RTP and then daily thereafter. They are compared to the specified limits to assure that the reactor is operating within the assumptions of the safety analysis. The daily surveillance requirement is based on engineering judgement considering the slow changes in power distribution. The 12 hour allowance after exceeding 25% of RTP is acceptable given the large inherent margin to operating limits at low power levels. REFERENCES 1. NED0-240ll-P-A, " General Electric Standard Application for Reactor Fuel", September 1988.
- 2. NED0-301300-A, " Steady State Nuclear Methods", May 1985.
- 3. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
- 4. NED0-24154, " Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors", October 1978.
O i i i i 9; ABWR B 3.2-4 5/31/89
MCPR" B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 KIN _lM3M CRITICAL POWER RATIO BASES BACKGROUND The MINIMUM CRITICAL POWER RATIO (MCPR) is a measure of the operating fuel assembly power relative to the fuel assembly power that would' result in the onset of boiling transition. The Safety Limit MCPR is set such that 99.9% of the fuel rods will avoid boiling transition if the limit is not violated. (refer to the Bases for LCO 2.1.2). The operating limit MCPR' is established to assure that no fuel damage results during anticipated operational occurrences. Although fuel damage would not necessarily occur if a fuel rod actually experienr.ed boiling. transition (Ref.1), the critical power at which. boiling transition is calculated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomena that is readily detected during the testing of various bundle designs. Based. on this experimental data, correlations have been developed that are used to predict critical bundle power (i.e., the bundle power level' at the onset of transition boiling) for a given set'of plant parameters (e.g., pressure, mass flux, f7 subcooling,etc.). . Since plant operating conditions and bundle U power levels are relatively easily monitored and determined, monitoring MCPk is a convenient way of ensuring .that fuel' failures due to inadequate cooling do not occur. APPLICABLE The analytical methods and assumptions used in evaluating SAFETY the anticipated operational occurrences to establish the. ANALYSES operating limit MCPR are presented in the SSAR, Chapters 4, 6 and 15, and in Reference 2. To assure that the Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to determine the largest reduction in Critical Power Ratio (CPR). The type of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion and coolant temperature decrease. The limiting transient yields the 1trgest ACPR. l When the largest ACPR is added to the Safety Limit MCPR,. the required operating limit MCPR is obtained. The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPRf and MCPR p respectively) to ensure adherence to fuel design p (continued) ABWR B 3.2-5 5/31/89 l I
-__-u_ _ __ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
c
'MCPR ' B 3.2.2 BASES (continued)
ADPLICABLE limits during the worst moderate frequency transient. Flow SAFETY dependent MCPR limits are determined by steady state thermal ANALYSES hydraulic methods with key physics response inputs benchmarked (continued) using the three dimensional BWR simulator coh (Ref. 3) to-analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System. Power dependent MCPR limits (MCPR p
) are determined mainly by the one-dimensional _ transient code (Ref. 4) for the anticipated transients that are significantly affected by power. Due to the sensitivity cf the transient response to initial core flow levels at power levels below that where the turbine stop valve closure and turbine control valve fast closure' scram trips are bypassed, a high and low flow operating limit MCPR p is provided for operating between 25% of RATED THERMAL POWER (RTP) and the previously mentioned bypass power level.
MCPR satisfies the requirements of Selection triterion 2 of the NRC Interim Policy Statement on Technical Specification ; Improvements as documented in Reference 6. LC0 The MCPR operating limits specified in the [ CORE OPERATING LIMITS REPORT] are the result of the design basis accident and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf and MCPR p limits. APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur from high power level 4 conditions. Below 25% of RTP, the reactor will be operating at minimum reactor internal pump speed and the moderator void l content will be very small. Surveillance of thermal limits below 25% of RTP is unnecessary due to the large inherent I margin that assures that the Safety Limit MCPR will not be exceeded even if a limiting transient should occur. Statis-tical analyses documented in Reference 5 indicate that the nominal value of initial MCPR expected at 25% of RTP is in excess of 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power / flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. (continued) ABWR B 3.2-6 5/31/89 __._______J
'Ii l'
MCPR B 3.2.2 ((
%/
BASES (continued) APPLICABILITY The resuP.s of these studies demonstrate that margin is (continued)- expected netween performance and MCPR requirements, and that marginn increase as power is reduced to 25% of RTP. This trend is expected to continue to the 5-15% power range where entry into MODE 2 occurs. 'l! hen in MODE 2, the Startup Range Neutron Monitor (SRNM) provides rapid scram initiation for any signif-icant power increase transient which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels less than 25% of RTP, the reactor will be operating with substantial margin to MCPR limits and the specification is not required. i ACTIONS A.1 Should any MCPR be outside the required limits, an initial condition.of the design basis transient analyses may not be met.. Therefore, prompt action should be taken to restore the MCPR's to within the required limits such that the plant will be operating within analyzed conditions. fld
-b If the MCPR cannot be restored to within the required limits in two hours, it is required to reduce. THERMAL POWER to < 25% of RTP. As discussed in the Bases for Applicability, operation below 25% of RTP results in sufficient margin to the required limits.
Completion Times The Completion Times are based on industry accepted practice j and engineering judgement considering the time to reasonably ; complete the Required Action. i SURVEILLANCE SR 3.2.2.1 REQUIREMENTS MCPR is required to be initially calculated within 12 hours after THERMAL POWER has exceeded 25% of RTP and then daily thereafter. It is compared to the specified limits to assure that the reactor is operating within the assumptions of the safety analysis. The daily surveillance requirement is based on engineering judgement considering the slow changes in power distribution. The 12 hour allowance after exceeding 25% of RTP is acceptable given the large inherent margin to operating limits at low power levels. O (continued) ABWR B 3.2-7 5/31/89
l MCPR , B 3.2 2 I BASES (continued) REFERENCES 1. NUREG-0562, " Fuel Rod Failure as a Consequence of Departure from Nucleate Boiling or Dryout", June 1979.
- 2. NE00 240ll-P-A, " General Electric Standard Application j for Reactor Fuel", September 1988.
l
- 3. NED0-30131-A, " Steady State Nuclear Methods", May 1985. f
- 4. NED0-24154, " Qualification of.the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.
- 5. "BWR/6 Generic Rod Withdrawal. Error Analysis", Appendix 158, General Electric Standard Safety Analysis Report .
(GESSAR-II).
- 6. NE00-31466, " Technical Specificati:n Screening Criteria Application and Risk Assessment", November 1987.
O l l l F e ACWR B 3.2-8 5/31/89
LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS l B 3.2.3 LINEAR HEAT GENERATION RATE BASES BACKGROUND The LINEAR HEAT GENERATION RATE (LHGR) is a measure of the heat generation rate of a fuel rod in a fuel assembly at an axial location. Limits on LHGR are specified to assure that fuel design limits will not be exceeded anywhere in the core during normal operation including anticipated operational occurrences. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to assure that fuel system damage, fuel rod failure or inability to cool the fuel will not occur during the anticipated operating conditions identified in [Ref. 1]. APPLICABLE The analytical methods and assumptions used in evaluating fuel SAFETY system design are presented in SSAR, Chapter 4 and in ANALYSES Reference 1. The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation and protection system) that fuel damage will not result in the release of radioactive O materials in excess of the guidelines of 10 CFR 20, 50 and 100. The mechanisms which could cause fuel damage during operational transients and which are considered in fuel evaluations are (1) rupture of the fuel rod cladding caused by strain from the relative expansion of the U0,, pellet and (2) severe overheating of the fuel rod cladding cauted by inadequate cooling. A value I of 1% plastic strain of the zircaloy cladding has been defined as the limit below which fuel damage caused by overtraining of the fuel cladding is not expected to occur (Ref. 2). The Safety Limit MINIMUM CRITICAL POWER RATIO (MCPR) ensures that fuel damage caused by severe overheating of the fuel rod cladding is avoided and is discussed separately in the Bases for LCO 3.2.2. Fuel design evaluations have been performed and demonstrate that the 1% plastic strain fuel design limit is not exceeded during continuous operation with LHGR's up to the operating limit specified in the [ CORE OPERATING LIMITS REPORT). The analysis also includes allowances for short term transient operation abcve the operating limit to account for anticipated operational occurrences, plus an allowance for densification power spiking. (continued) ABWR B 3.2-9 5/31/89
LHGR B 3.2.3 BASES (continued) APPLICABLE LHGR satisfies the requirements of Selection Criterion 2 of the SAFETY NRC Interim Policy Statement on Technical Specification ANALYSES Improvements'as documented in Reference 3. (continued) LCO LHGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause 1% cladding plastic strain. The operating limit to accomplish this objective is specified in the [ CORE OPERATING LIMITS REPORT]. APPLICABILITY The LHGR limit is derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels less than 25% of RATED THERMAL POWER (RTP), the reactor will be operating with substantial margin to LHGR limits and therefore, the specification is only required when operating at or above 25% of RTP. ACTIONS M Should any LHGR exceed the required limits, an initial condition of the fuel design analysis will not be met. There-fore, prompt action should be taken to restore the LHGR to within the required limits such that the plant will be operating within analyzed conditions. M If the LHGR cannot be restored to within the required limits in two hours, it is required to reduce THERMAL POWER to < 25% of RTP. Operation below 25% of RTP results in sufficient margin to the required limits. Completion Times The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably complete the Required Action. (continued) O ABWR B 3.2-10 5/31/89 _ __ _-____D
LHGR-B 3.2.3 ..r BASES-(continued) SURVEILLANCE SR 3.2.3.1 REQUIREMENTS . LHGR is required to be initially calculated within it hours after 'HERMAL' POWER has exceeded 25% of RTP and then daily. thereafter. It is compared to the specified limits to' assure that tht reactor is operating within the assumptions of the safety analysis. The daily surveillance requirement is based on engineering judgement considering the slow changes in power distribution. The 12 hour allowance after exceeding 25% of RTP is acceptable given the 'large inherent margin to operating limits at . lower power levels. REFERLNCES 1. [To be provided for non-GE fuel).
- 2. NUREG-0800, Standard Review Plan 4.2, " Fuel System
. Design",SectionII.A.2(g).
- 3. NED0-31466, " Technical Specification Screening Criteria Application and' Risk Assessment", November 1987.
O O ABWR B 3.2-11 5/31/89 =- _ _ _ _ _ - _ _ _ _ - _
Recirculation loops Operating B 3.4.1 G V B 3.4 REACTOR COOLANT SYSTEM-B 3.4.1 Recirculation Pumps ODePatinQ BASES l BACKGROUND The reactor coolant recirculation system is designed to provide a forced coolant flow.thr s.gh the core to remove heat from the-fuel. The reactor coolant recirculation system consists of ten recirculation pumps internal to the reactor vessel. The rrector internal pumps (RIPS) are driven by wet motors that protrude from the bottom of the reactor vessel. Each motor has its own external heat exchanger that is cooled by reactor building cooling water. The pump motor casings are part of the reactor coolant pressure boundary and are located in the lower drywell area. The pumps are reactor vessel internals. The recirculated coolant consists of saturated water from the steam. separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud to the inlet of the ten RIPS that are located equidistant around the botton, of the core shroud (or pump deck). The total core flow passes through the RIPS into the lower plenum, up through the orifices of the lower core plate, and then up through the core, Each pump q motor is driven by an Adjustable Speed Drive'(ASD) that is g- individually started and controlled from the main control room. The total core flow can also be controlled automatically, with all pumps in parallel, by the master flow controller. The ten RIPS are powered by four separate electrical busses. Two busses each supply three RIPS and the other two busses each supply 2 RIPS. There is also an M/G set between each of the busses supplying 3 RIPS and'the respective ASDs (2 M/G sets total). This provides for slower flow coastdown character - istics during certain transient events. APPLICABLE The operation of the reactor coolant recirculation system is an ; SAFETY initial condition assumed in the design basis Loss of Coolant i ANALYSES Accident (LOCA) (Ref. 1). During a LOCA the operating RIPS are j all assumed to trip at time zero due to a coincident loss of offsite power. The subsequent core flow coastdown will be immediate and rapid because of the relatively low inertia of the pumps. Also, for conservatism, no credit is taken for the inertia of the two M/G sets that feed six of the RIPS. (continued) O ABWR B 3.4-1 5/31/89
Recirculation Loops Operating B 3.4.1 . s BASES (continued) l APPLICABLE With at least nine of the ten RIPS in operation the LOCA l SAFETY analysis includes all potential power and flow operating points i ANALYSES from which an event might be initiated. With eight or less. J' (Continued) RIPS in operation the LOCA analysis assumptions do not include all potential operating states so that additional restrictions are necessary regarding reactor power based on the number of pumps actually operating. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2) which are analyzed in Chapter 15 of the FSAR. Recirculation Loops Operating satisfies the requirements of
. Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3.
LC0 At least nine RIPS are required to be in operation to ensure during a LOCA the assumptions of the LOCA analysis are satis-fied without restriction. With less than nine RIPS in opera-tion, all potential power and flow operating states have not been accounted for in either the LOCA or transient analysis. Therefore, certain restrictions apply depending on the number of RIPS operating.
&W APPLICABILITY Requirements for operation of the reactor coolant recirculation system are necessary during MODES 1 and 2 since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. During other conditions, the consequences of an accident are reduced and the coastdown characteristics of the reactor internal pumps are not important.
ACTIONS A.l. B.l. C.l. D.1 With less than nine RIPS in operation the reactor power level must be restricted so that the assumptions of the LOCA and transient analyses are met. With only seven or eight RIPS operating the power level is restricted to 90% and 95% of RTP, respectively. However operation may continue indefinitely. As noted, LC0 3.0.4 is not applicable for Required Actions A.1 or B.l. With less than se en pumps operating power level is restricted even further and operation may only continue for a short time. (continued) ABWR B 3.4-2 5/31/89
Recirculation loops Operating F B341 B_ASES (continued) (u) ACTIONS A.1. B.1. C.1. D.1'(continued) (continued) For the case of 5 or 6 pumps running, reactor power must be
~
reduced to less than 25% RPT because of potential stability concerns. With less than 5 pumps operating, power must be reduced to less-than 5% RPT due to the lack of detailed analysis of the actual flow distribution with less than half of the pumps in operation providing forced flow at higher po er levels. C.2.1. C.2.2. D.2.1. D.2.2 With less than seven RIPS operating the steady state power and flow characteristics of the core have not been fully analyzed. Therefore, even at reduced power levels, continued operation is allowed for~only a short time while an attempt is made to restore at least seven pumps to operating status. With less than seven pumps able to be restored to operating status within the Required Completion Time, the reactor is required to be in In this condition, the RIPS are not required to be MODE 3. operating because of the reduced severity of design basis accidents and minimal dependence on the forced flow characteristics.. A
't) ' Coroletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of availacle systems and the time required to reasonably complete the -
Required Action. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance requirement ensures that the reactor power level is within:the assumptions of the applicable analyses based on the number of pumps actually operating. Operating experience has demonstrated that a 24 hour frequency for this type of surveillance is adequate. (continued) A ABWR B 3.4-3 5/31/89
Recirculation Loops Operating B 3.4.1 i BASES (continued) - REFERENCES 1. ABWR SSAR, Section 6.3.3.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
l l 0 O O ABWR B 3.4-4 5/31/89
S/RVs B 3.4.2 B 3.4 REACTOR COOLANT SYSTEM B 3.4.2 Safety / Relief Valves BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions. As part of the nuclear pressure relief system, the size and number of safety / relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary. The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes - the ; safety mode or the relief mode. In the safety mode (or spring mode of operation), the direct action of the steam pressure in the main steam lines will act against a spring-loaded disk that will pop open when the valve inlet pressure exceeds the spring force plus the weight of the disk. In the relief mode (or power actuated mode of operation), a pneumatic piston / cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force; however, a steam inlet pressure of at least 50 psig is required to overcome the weight of the disk O which is not directly connected to the spring. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. Eight of the S/RVs that provide the relief function are part of the Automatic Depressurization System (ADS) specified in LC0 3.5.1. APPLICABLE The overpressure protection system must accommodate the most SAFETY severe pressurization transient. Evaluations have determined ANALYSES thro the most severe transient is the closure of all main steam line isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e., failure of the direct scram associ-ated with MSIV position) (Ref. 1). Twelve of the eighteen S/RVs are assumed to operate in the safety mode and no credit is taken for the relief mode of operation. The analysis results indicate the design S/RV capacity is capable of main-taining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig - 1375 psig). Refer-ence 2 discusses additional events which are expected to actuate the S/hVs. From an overpressure standpoint, these events are bounded by the MSIV closure with flux scram event described above. (continued) ABWR B 3.4-5 5/31/89
S/RVs B 3.4.2 BASES (continued) APPLICABLE S/RVs satisfy the requirements of Selection Criterion 3 of the SAFETY NRC Interim Policy Statement on Technical Specification ANALYSES Improvements as documented in Reference 3. (continued) LC0 Twelve S/RVs are required to be OPERABLE in the safety mode. The requirements of this LCO are applicable only to the capabil- ; ity of the S/RVs to mechanically open to relieve excess pres-sure. In Reference 1, an evaluation was performed to establish the parametric relationship between the peak vessel pressure and the number of OPERABLE S/RVs. The results show that with a minimum of 12 S/RVs OPERABLE in the safety mode, with setpoints in a distribution equivalent to, or conservative with respect to, the minimum requirements of SR 3.4.2.1, the ASME Code. limit of 1375 psig is not exceeded. The S/RV setpoints are established to ensure the ASME Code limit on peak reactor prersure is satisfied. The ASME code specifications require t' e lowest valve be set at or below vessel design pressure (1250 psig) and the highest safety valve be set so the total accumulated pressure does not exceed (10% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of + 1% of the nominal setpoint to account for potential setpoint drift and to provide an added degree of conservatism. Operation with less valves OPERABLE than specified, or with setpoints greater than specified, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded. APPLICABILITY The specified number of S/RVs must be OPERABLE in MODES 1, 2 and 3 since there is considerable energy in the reactor core and the limiting design basis transients are assumed to occur. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) system is capable of dissipating the heat being generated. In MODE 4, decay heat levels are low j 1 enough such that the RHR system is adequate, and reactor pressure levels are low enough such that the overpressure limit (continued) , O ABWR B 3.4-6 5/31/89 _ ._ _ __ _ a
E S/RVs B 3.4.2
~
BASES (continuedi APPLICABILITY cannot be challenged by assumed operational, transients or
.(continued) accidents. .In MODE 5, the reactor vessel head is unbolted or removed and there is no-reactor coolant pressure boundary (RCPB)'. The S/RV. function is not needed during these conditions.
' ACTIONS A.1. A.2 With any of the required S/RVs inoperable, the reactor must be in MODE 3 in 12 hours and in MODE 4 in 36 hours. With less than the minimum number of.S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. It is therefore necessary for the plant to be'in a condition where the S/RVs are 'not required. .The Completion Times allow for a controlled Shutdown of the reactor without placing undue stress on plant operators or plant systems.
SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This surveillance requirement demonstrates that the S/RVs will 5 open at the pre! es assumed in the safety analysis of Refer-
\ ence 1. The den instration of the S/RV lift settings must be performed during shutdown and in accordance with the provisions of SR 3.0.5. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
SR 3.4.2.2 A manual actuation of each S/RV is performed to verify the valve is mechanically functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valves or by change in the measured steam flow or any other method suitable to verify gross steam flow through the discharge line to the suppression pool. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the . required pres-sure is achieved to perform this test. Adequate pressure at which this test is to be performed is [ ] psig (the pressure recommended by the valve manufacturer). Plant startup is I (continued) l 5/31/89 I ABWR B 3.4 7 i
- - _ _ _ - . - ___.____m___._m__.._. ___ _ _ l
S/RVs B 3.4.2 BASES (continued) SURVEILLANCE SR 3.4.2.2 (continued) G. REQUIREMENTS (continued) allowed prior to performing this test _ because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. If the valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the S/RV is considered OPERABLE. Surveillance Frequencies in general, surveillance frequencies are based on industry 9ccepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of
,serforming the test and a likelihood of a change in the system / component status.
REFERENCES 1. ABWR SSAR, Section 5.2.2.
- 2. ABWR SSAR, Chapter 15.
- 3. NED0-31466, Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
ABWR B 3.4-8 5/31/89
-i Operational LeakageD B-3.4.3 b
V
.B 3.'4' REACTOR COOLANT SYSTEM B 3.4.3 Doerational' Leakace BASES-BACKGROUND' The reactor coolant system in' cludes systems 'and components that contain or. transport fluids to or from the reactor core. These
; systems form a major portion of the nuclear system process, barrier. The pressure contain_ing components of.the reactor coolant system,. including the portions of the system out to and:
including . isolation valves, are defined as. the Reactor Coolant - Pressure Boundary (RCPB). Limits on leakage from the RCPB are; required to ensure appropriate. action can be taken before the integrity of the nuclear system process barrier is impaired ~ ' (Ref.:1). The safety significance of leaks from the RCPB can vary widely depending on the source of the leak as well as the leakage rate: and drration. Therefore, detection of leakage in the drywell is necessary. Identified Leakage.is defined as the leakage into closed systems, such as pump seal or valve' packing. leaks that are captured, flow metered and piped to a; sump or collect-ing tank.~ Also, leakage into the containment atmosphere from. y sources that-are specifically located and known not'to inter-fere with the operation of Unidentified Leakage detection or-NJ r not to be a flaw in the RCPB are considered Identified Leakage. Unidentified Leakage is collected in the drywell floor, drain sump along with the. normal design leakage from the control rod drives, valve- flange leakage, floor drains, closed cooling water system and drywelT cooling unit drains. Methods for-separating the-Identified Leakage from the Unidentified Leakage are necessary to provide prompt and quantitative information to the operators to permit them to take corrective action. A limited amount of leakage is expected from auxiliary systems-within the drywell that cannot be made 100% leaktight. If leakage occurs from these paths, it should be detectable and isolated from the drywell atmosphere if possible, so as not to . mask' any potentially serious leak should it occur. . l APPLICABLE The ' allowable leakage rates from the reactor coolant system SAFETY have been based on the predicted and experimentally observed ! ANALYSES behavior of pipe cracks. The normally expected background leakage due to equipment design-and the detection capability of L the instrumentation for. determining system leakage were also q considered. The evidence obtained from experiments suggests for leakage somewhat greater than specified for Unidentified J (continued) ABWR B 3.4-9 5/31/89
Operational Leakage B 3.4.3 BASES (continued)
> APPLICABLE Leakage, the probability is small the imperfection or crack SAFETY associated with such leakage would grow rapidly. The ANALYSES Unidentified Leakage rate limit is established at I gpm to (continued) allow time for corrective action before the process barrier could be significantly compromises. This limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Based on crack behavior from experi-mental programs (Ref. 2 and 3) it is estimated that leak rates of hundreds of gpm will precede crack instability (Ref. 4).
There are no applicable safety analyses that assume the Total Leakage limit. The Total Leakage limit is specified based on consideration of inventory makeup capability and sump capacities. Operational Leakage satisfies the requirements of Selection Criterion 1 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 5. LC0 No Pressure Boundary Leakage is allowed since the potential exists for a break in the RCPB and a loss of substantial inventory. Leakage should not be allowed to increase signifi-cantly without a thorough examination of the source of the leak. The Total Leakage rate consists of all leakage, identified and unidentified, that . flows to the drywell floor drain and equip-ment drain sumps. The Unidentified Leakage rate limit is based on the leakage rate above background leakage rates. APPLICABILITY The potential for RCPB leakage is greatest when the reactor is pressurized. Under these conditions, high stresses are applied to the system piping resulting in the potential for i crack growth and possible failure of the RCPB. Therefore, l detection of RCPB leakage is required during MODES 1, 2 and 3. j In MODES 4 and 5, operational leakage limits are not required, i since the reactor is not pressurized and the potential for j leakage and possible pressure boundary failure is reduced. l (continued) 1 O ABWR B 3.4-10 5/31/89
.1 Operatio'nal Leakage B'3.4.3 .,
1 uYN BASES (continued) R ()'Q ACTIONSc 8.d With either the' . Total Leakage, Unidentified Leakage, or both greater than the required limits,: actions should be taken to identify the source of the leak and determine the. significance. Because the. leakage limits.are conservatively below the leakage that would constitute a critical crack size, a limited time is allowed to evaluate the situation. .If a change in Unidentified , n- Lleakage has. been adequately identified and quantified, it may ! be reclassified and considered'as Identified Leakage. However, the Total Leakage limit would remain-unchanged. B.1. B.2~. If the LEAKAGE cannot be restored to within the required limits, the reactor should be in MODE 3 and subsequently in. MODE 4. If Pressure Boundary Leakage occurs there is the potential' that the flaw in the RCPB could eventually result in a pipe' break or other LOCA. Since the area being monitored.is inaccessible, the reactor must be in' MODE'3 and subsequently..in
. MODE 4 to allow a visual inspection to determine the source.of the leak.
Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.4.3.1 REQUIREMENTS The reactor coolant system LEAKAGE is monitored by a variety of systems designed to provide alarms when leakage is indicated and to quantify the various types of leakage. ~ Leakage detec-tion is discussed in more detail in the Bases for LC0 [ Leak Detection Instrumentation]. Sump level or flow rates are typically monitored to determine actual leakage rates. How-ever, any method may be used which quantifies leakage within the guidelines of Reference 1. Operating experience has demonstrated that a 12 hour frequency for this surveillance is adequate. (continued) b ABWR 8 3.4-11 5/31/89
Operational Leakage 8 3.4.3 BASES (continued) REFERENCES 1. Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1073
- 2. GEAP-5620, " Failure Behavior in ASTM A106 Pipes Containing Axial Through-Wall Flaws," April 1968.
- 3. NUREG-76/067, " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants," October 1975.
- 4. ABWR SSAR, Section 5.2.5.5.
- 5. NEDO-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
O
?
O ABWR B 3.4-12 5/31/89
- l. 1 C - __ ___ - l
Specific Activity j B 3.4.4 1 [i_/') B 3.4 REACTOR COOLANT SYSTEM B 3.4.4 Specific Acthity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive material due to release of fission products into the coolant , and activation of crud particles in the reactor coolant. These radioactive materials in the reactor coolant could contribute to release of radioactive materials into the environment during design basis accidents. Limits on the maximum allowable level of radioactivity in the reactor coolant are established to assure, in the event of a release of any radioactive material to the environment during a i design basis accident, radiation doses are maintained within the limits of 10 CFR 100. 4 APPLICABLE Analytical methods and assumptions involving radioactive SAFETY material in the primary coolant are presented in SSAR Chapter ANALYSES 15, Accident Analyses. The specific activity in the reactor (xf coolant (source term) is an initial condition assumed for s evaluation of the consequences of an accident due to a main V steam line break (MSLB) outside the containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is postulated to be terminated by complete closure of the main steam line isolation valves (MSIVs). This release forms the basis for determining off site doses (Ref. 1). The limitations on the specific activity of the primary coolant ensure the 2 hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed 10% cf the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation. Specific Activity satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. (continued) l J h)) l ABWR B 3.4-13 5/31/89 1
Specific Activity B 3.4.4 BASES (continued) LC0 The primary coolant specific activity level nf 2 0.2 micro-curies per gram DOSE EQUIVALENT l-131 is required to ensure the source term assumed in the safety analysis of the MSLB'is not exceeded such that any release of radioactive material to the environment does not exceed 10 CFR 100 limits. APPLICABILITY Limitations on levels of primary coolant radioactivity are applicable during MODE I and MODES 2 and 3 with any main steam line not isolated since there is an escape path for release of radioactive material from the coolant to the environment in the event of a MSLB outside of the primary containment. During MODES 4 and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced. ACTIONS A.I. A.2. B.l. B.2 A primary coolant specific activity level > 0.2 pCi per gram DOSE EQUIVALENT I-131 indicates the presence of some abnormal-ity in plant operations. The range between 0.2 Ci and 4.0 yCi is acceptable for up to 48 continuous hours to account for potential iodine spiking that may occur following changes in THERMAL POWER. Increased surveillance of the reactor coolant specific activity during this period is required to closely monitor the condition and determine if additional limits are exceeded. If coolant specific activity cannot be restored s 0.2 pCi within 48 hours, or when coolant activity is > 4.0 pCi, the main steam lines are required to be isolated. This action precludes the possibility of release of radioactive material to the environment in excess of the requirements of 10 CFR 100, during a postulated MSLB accident. Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.. (continued) l O ABWR B 3.4-14 5/31/89
Specific Activity B 3.4.4
/^T A B_A.SES - ( c ont i nued ) -
SURVEILLANCE SR 3.4.4.1 REQUIREMENTS Isotopic analysis for DOSE EQUIVALENT 1-131 concentration is necessary to determine that the specified maximum primary coolant activity is s 0.2 microcuries per gram DOSE EQUIVALENT l-131 during steady state operation. Operating experience has demonstrated that a 31 day frequency for this surveillance is adequate. Surveillance Frequencies in general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. ABWR SSAR, Section 15.6.4.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987. ,
,r) 1 i
n
% l l ;
ABWR B 3.4-15 5/31/89
RHR - Shutdown B 3.4.5 l 1 i B 3.4 REACTOR COOLANT SYSTEM f B 3.4.5 Residual Heat Removal - Shutdown BASES BACKGROUND Irradiated fuel in the reactor pressure vessel (RPV) generates decay heat during normal and abnormal shutdown conditions, potentially resulting in an increase in the temperature of the l reactor coolant. This decay heat is required to be removed such that the reactor coolant temperature can be reduced to or maintained at > '00*F in preparation for performing refueling, maintenance operations or for maintaining the reactor in cold shutdown conditions. Systems capable of removing decay heat are therefore required to perform these functions. The three shutdown cooling loops of the Residual Heat Removal (RHR) system provide decay heat removal. Each loop consists of a motor driven pump, a heat exchanger, and the associated piping and valves. Each loop has its own dedicated suction from the RPV. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchanger, to the reactor via feedwater line ' A' for RHR Loop 'A' and via the RHR Low Pressure Flooder Spargers for , Loops 'B' and 'C' . The RHR heat exchangers transfer heat to the Reactor Building Cooling Water System (LCO 3.7.1 and LC0 & 3.7.2). The RHR shutdown cooling mode is a manually controlled W system. APPLICABLE Decay heat removal by operation of the shutdown cooling mode SAFETY of the RHR system is not required for mitigation of any events ANALYSES or accidents evaluated in the safety analyses. However, the NRC Interim Policy Statement (Ref. 1) requires the RHR system be retained in the Technical Specifications even though none of the selection criteria (Ref. 2) were satisfied. LCO Two shutdown cooling subsystems are required to be OPERABLE. An OPERABLE RHR shutdown cooling subsystem consists of one RHR pump, one heat exchanger, and the associated piping and valves. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODES 3 and 4, one shutdown cooling subsystem of the RHR can provide (continued) { O j 4 ABWR B 3.4-16 5/31/89 l 1 4
L RHR - Shutdown B 3.4.5
.[*y BASES (continued)
- t/
LC0 .
.the required cooling.to maintain the. desired temperature. Two (continued) subsystems are required to be OPERABLE to provide redundancy.
Operation (either continuous.or intermittent) of one subsystem can maintain and reduce the. reactor coolant temperature as required. APPLICABILITY Decay heat removal at reactor pressures above the RHR cut-in permissive pressure is typically accomplished by condensing steam from the RPV in the main condenser. When the reactor pressure is'below'the RHR cut-in permissive pressure, the RHR system may be operated in the shetdown' cooling mode. . Operation of the RHR system in the shutdown cooling mode above this pressure is not allowed because the coolant pressure.may exceed the design pressure of the shutdown cooling piping. Operation of a subsystem to remove the decay heat may be. required to either reduce or maintain coolant temperature. Continued operation of a shutdown cooling subsystem may be required to reach'and maintain reactor coolant' temperatures' s 200*f, which corresponds to MODE 4. The requirements.for decay heat removal in MODE 5 are discussed in LC0 3.9.8 and LCO 3.9.9. ACTIONS AJ. With one of the required RHR shutdown cooling subsystems inoperable for decay heat removal, the remaining OPERABLE subsystem can provide the necessary decay heat removal. However, the overall system reliability is reduced. For this reason, continued operation is allowed for a limited time only, 2 hours. B.1, B,2 If one of the inoperable subsystems cannot be restored to OPERABLE status or with all three subsystems inoperable, an alternate method of decay heat removal is required to be made j available for each required, but inoperable, subsystem to restore cooling capability. The required cooling capacity of the alternate method should be ensured ~by verifying (by calcu- . lation or demonstration) its capability to maintain or reduce { temperature. Decay heat removal by ambient losses can be co sidered as contributing to the alternate method capability. (continued) A V ABWR B 3.4-17 5/31/89 l.
RHR - Shutdoen B 3.4.5 BASES (continued) ACTIONS Comoletion Times O (Continued) All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.4.5 1 REQUIREMENTS Verification that all valves of the required RHR shutdown cooling subsystem are in the correct position ensures the proper flow path. Valves not in the correct position must be capable of manual realignment either from the control room or at the valve location. Verification of this capability is provided by actuation of the valve from the control room or the current inservice inspection reports. Operating experience has demonstrated that a 31 day frequency for this surveillance is adequate. Because some of the required valves are interlocked closed when above the RHR cut-in permissive pressure, an allowance is provided to test the valves within 12 hours after pressure has been reduced below the cut-in permissive pressure. This allows conditions to be established under which the test may be performed. REFERENCES 1. 52FR3788, Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O i ABWR B 3.4-18 5/31/89 1 E---._---___
RCS Pressure / Temperature Limits B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM B 3.4.6 Reactor Coolant System Pressure / Temperature limits BASES BACKGROUND All components in the reactor coolant system (RCS) are designed to withstand the effects of cyclic loads due to system temper-ature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with l the design assumptions and satisfy the stress limits for cyclic operation. l l The purpose of this specification is to establish operating limits that provide a wide margin to brittle failure of major piping and pressure vessel components of the Reactor Coolant Pressure Boundary (RCPB). Of the major components within the RCPB, the reactor vessel is the component most subject to brittle failure and therefore the component for which the technical specification limits is most pertinent. The basis of the pressure and temperature (PT) limits is found O in Appendix G to 10 CFR 50 (Ref. 1). Appendix G requires that limits be established, and that the limits be based on specific fracture toughness requirements for RCPB materials such that an adequate margin to brittle failure will be provided during operational occurrences. 10 CFR 50 Appendix G mandates the use of ASME Section Ill, Appendix G (Ref. 2). The concern addressed by Appendix G is that undetected flaws could exist in the RCPB components, which if subjected to unusual pressure and/or thermal stresses, could result in non-ductile (brittle) failure. Certain reactor coolant system PT combinations can cause stress concentrations at flaw loca-tions which in turn could cause flaw growth, resulting in failure before the ultimate strength of the material is attained. Flaw growth is resisted by the material toughness. Toughness is a property that varies with temperature and is lower at room temperature than at power operation. Further-more, the material toughness is affected by neutron fluence which causes the steel ductility to decrease. The effect of fluence is cumulative, and ductility steadily decreases wi*h exposure time. Only the vessel beltline region is in a hi h fluence area. Toughness is also dependent on the chemistry of the base metal and its impurities. (continued) ABWR B 3.4-19 5/31/89
RCS Pressure / Temperature Limits B 3.4.6 l' BASES (continued) BACKGROUND Linear elastic fracture mechanics (LEFM) methodology, following (continued) the guidance given by 10 CFR 50 Appendix G, ASME Section III Appendix ti, and Regulatory Guide 1.99, is used to determine the stresses and material toughness at locations within the RCPB. Although any region within the pressure boundary is subject to non-ductile failure, the regions that provide the most restric-tive limits are the vessel closure head flange, the feedwater nozzles, the control rod drive nozzles, and the vessel beltline. One indicator used to indicate the temperature effect on ductility is the nil-ductility temperature (NDT). The NDT for the steel alloy used in vessel fabrication has been established by testing. The NDT is a temperature at which it can be said that fracture failure may occur below the NDT. Ductile failure may occur above the NDT. The exact temperature value is not very precise. Consequently a reference temperature (RT ) has been established by experimental means. TheneutronemhIttle-ment effect on the material toughness is reflected by increas-ing the RT as exposure to neutron fluence increases. In effecttheNmperatureatwhichbrittlefailurecanoccur N increases. The actual shift in RT of the vessel material will be establishedperiodica1hTduring operation by removing and evaluating in accordance with ASTM E185-73 and 10 CFR 50, g Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The operating limit curves shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99 (Ref. 3 and 4). The PT curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the geometry of the reactor vessel will dictate the most restrictive limit. Across the entire pressure / temperature span of the limit curves, some locations are more restrictive, and thus the curves are composites of the most restrictive regions. The curves have been developed for heatup, in-service leak and hydrostatic testing, and cooldown in conjunction with stress analyses for a large number of operating cycles and provide a conservative margin to non-ductile failure. Although they have (continted) O ABWR B 3.4-20 5/31/89
RCS Pressure / Temperature Limits B 3.4.6 l Fg h; - BASES (continued)- [ BACKGROUND been created to provide limits for these specific normal
'(continued): ' operations, they also can be used as a basis for determining if' evaluatio.ns are necessary for abnormal transients which can l begin-from power operation. ASME Section XI Appendix E (Ref. 5) provides a recommended methodology for evaluating operating events which.cause an excursion outside the normal limits.
. APPLICABLE The. limits are not derived from design basis accident analyses' SAFETY presented in the FSAR, but are prescribed as guidance to be
' ANALYSES used during normal operation to avoid encountering pressure, temperature, and temperature-rate-of-change conditions which.
might cause undetected flaws to propagate, in turn causing non-ductile failure of the RCPB. RCS Pressure / Temperature Limits satisfies the requirements of the NRC Interim Policy Statement on Technical Specification Improvements for Nuclear. Power Reactors (Ref. 6). While nene of.the three Selection Criteria (Ref. 7) directly apply,.this specification preserves limits defining important boundaries for safe operation derived from the RCS stress analysis. Criterion 2 is the most appropriate criterion because operation outside of these boundaries is unanalyzed and may result in O RCPB failure. LCO Compliance with the following PT limits is required by this LCO:
- 1. Operation within the PT limit curves specified in the SSAR, 2 A maximum reactor coolant heatup or cooldown of 100*F in any one hour period,
- 3. A maximum temperature change of 5 10' in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves in the SSAR,
- 4. The reactor vessel flange and head flange temperature z 70*F when reactor vessel head bolting studs are under tension, (continued)
O ABWR B 3.4-21 5/31/89 j
RCS Pressure / Temperature Limits B 3.4.6 BASES (continued) LC0 5. A temperature difference between the bottom head coolant (continued) temperature and the reactor pressure vessel coolant temperature of s 145'F during reactor internal pump startup, and The above limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to non-ductile failure. APPLICABILITY The potential for violating the PT limits exists at all times when the reactor coolant system can be pressurized. The temperature rate of change limit can be potentially violated any time the reactor vessel material is a different temperature from a cooling source. ACTIONS A.1, A.2 As noted, Required Actions A.1 and A.2 must be completed whenever Condition A is entered. The purpose of the Note is to give additional emphasis to the need to restore operation to the allowable condition and to also perform an evaluation of & W the effects of any excursion outside of the allowable limits. Restoration alone is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. Restoration within the limits is appropriate because the RCPB is placed in a condition that has been verified by stress analysis. The action is in the proper direction to reduce RCPB stress. If a recirculation pump has been started and the PT limits were not met, the pump does not need to be turned off unless the PT limits are not currently being met with the pump operating. The Completion Time limit of 30 minutes is based on engineering judgement. Most violations will not be so severe that the activity cannot be accomplished in this time in a controlled manner; however, if the activity cannot be accomplished, then a controlled shutdown must be initiated per Required Actions B.1 and B.2. (continued) O ABWR B 3.4-22 5/31/89 1 1
i l RCS Pressure / Temperature Limits B 3.4.6 BASES (continued) ACTIONS A.I. A.2 (continued, l (continued) l In addition to restoration, an evaluation to determine if RCS i operation may proceed is required. The purpose of the evalu-ation is to determine if RCPB integrity is acceptable and must be accomplished before the event is remntiled. ASME Section X] Appendix E may be used to support the evalu-ation. If the evaluation cannot be accomplished in 72 Fours, or if the results of the evaluation are indeterminate or unfavorable, then the next appropriate action is to further reduce pressure and temperature as required in Condition B. l The 72 hour Completion Time is based on engineering judgement and is reasonable to accomp'lish the activities necessary. For a mild violation the evaluation should be possible within this time. More severe violations may require special, event specific stress analyses and/or inspections, which are appro-priately carried out while the RCS is in a reduced pressure and temperature condition as required by Condition B. B.l. B.2 If the Required Actions and associated Completion Times are not O met, a controlled shutdown must be initiated. This is a prudent action when the RCS remained in an unacceptable region for an extended period of incieased stress or a sufficient i severe event caused entry into an unacceptable recion. Either ' possibility indicates a need for more careful examination of the event, which is best accomplished while the RCS is in a low pressure and temperature state. With the plant at reduced pressure conditions the possibility of propagation of undetec-ted flaws is reduced. The times allowed for a controlled shutdown to MODE 4 are reasonable and avoid placing undue stress on plant operators er plant systems. SURVE]LLANCE SR 3.4.6.1 REQUIREMENTS Verification that operation is within limits is an appropriate surveillance when RCS temperature and pressure conditions are undergoing planned changes. The time period of 30 minutes is based on engineering judgement. Since temperature rate of change limits are specified in hourly increments, a hr.lf hour time period permits assessment and correction of minor deviations within a reasonable time. (continued) ABWR B 3.4-23 5/31/89
7-L RCS Pressure / Temperature Limits B 3.4.6 4 l BASES (continued) SURVEILLANCE SR 3.4.6.2 REQUIREMENTS (continued) A separate limit is used when the reactor is critical. Conse-quently, it is appropriate to verify that the RCS pressure and temperature are within the appropriate limit prior to the withdrawal of control rods that might make the reactor critical. SR 3,4.6.3 A differential temperature within the limit of this surveil-lance will ensure that thermal stresses resulting from an idle reactor internal pump startup will not exceed design allow-ances. Performing the surveillance within 15 minutes before starting the idle pump provides adequate assurance that-the limits will not be exceeded between the time of the surveil-lance and the time of the idle pump start. As noted, this surveillance is only required with less than five pumps already operating. With at least half of the pump in operation, there is already sufficient mixing that this verification is not necessary. SR 3.4.6.4 Limits on the reactor vessel flange and head flange temperature (required when the vessel head is tensioned) are generally bounded by the other PT limits during system heatup and cool-down. However, during operation in MODE 4, with RCS tempera-ture less than or equal to 100*F, surveillance of the flange temperatures is required to ensure the 70*F temperature limit is not violated. With RCS temperature less than or equal to 80*F, a more frequent check of the flange temperatures is required because of the reduced margin to the limit. The flange temperatures must also be verified to be above the limit prior to and during tensioning of the vessel head bolting studs to ensure that once the head is tensioned the limit is satisfied. Surveillance Frequencies In general, surveillance frequencies are based o industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test ant a likelihood of a change in the system / component status. , f (continued) O' ABWR B 3.4-24 5/31/89 J l
7 . . .. Lg ~
- e
.RCS Pressure / Temperature Limits.'
J B 3.4.6.-
' BASES' continued)
REFERENCES' 1. .. Code of Federal Regulations, Title 10, Part 50, Appendix G, " Fracture-Toughness Requirements."
-2. ' American Society of Mechanical Engineers (ASME), Boiler and Pressure. Vessel Code, Section -III, Appendix G,.
" Protection Against Non-Ductile Failure."
3.. USNRC Regulatory Guide 1.99, Revision 1, " Effects of x, Residual Elements on Predicated Radiation Damage to
. Reactor Vessel Materials," April 1977.
- 4. USNRC Regulatory Guide.l.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988..
- 5. American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section.XI, Appendix E,
'" Evaluation of Unanticipated Operating Events."
-6. 52FR3788, Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, USNRC, February 6, 1987- .
- 7. NED0-31466, " Technical . Specification Screening Criteria
. Application and Risk Assessment," November 1987.
O ABWR B 3.4-25 5/31/89
Reactor Steam Dome Pressure B 3.4.7 i 1 B 3.4 REACTOR COOLANT SYSTEM B 3.4.7 Reactor Steam Dome Pressure 0; BASES j BACKGROUND The reactor steam dome pressure is an assumed initial condition ! of design basis accidents and transients and is also assumed in ! the determination of compliance with reactor pressure vessel j overpressure protection criteria. APPLICABLE The reactor steam dome pressure is an initial condition of the .. f SAFETY vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system
-(primarily the safety / relief valves) during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure and therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (MINIMUM CRITICAL POWER RATIO, see Bases for LC0 3.2.2) and 1% cladding plastic strain (see Bases for LC0 3.2.1).
Reactor Steam Dome Pressure satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. LCO The specified reactor steam dome pressure limit assures the plant is operated within the assumptions of the transient analyses. Operation above the limit may result in a transient response more severe than analyzed. APPLICABILITY The reactor steam dome pressure is required to be less than or equal to the limit in MODES I and 2 where the reactor is generating significant steam and the design basis transients and accidents are bounding. The limit may be exceeded during anticipated transients since the evaluations of References 1 and 2 demonstrate that appropriate reactor and fuel limits are not exceeded. (continued) l ABWR B 3.4-26 5/31/89
, Reactor Steam Dome Pressure B 3.4.7-(N -
- BASES (continued)
APPLIC BILITY The' limit is not applicable in MODES 3,-4' and 5, because in (Continued)' these modes the reactor is shutdown. The' reactor pressure is well below the required limit and no' anticipated events will
' challenge.the overpressure limits.
- ACTIONS. A.1; B.1 I
If the reactor steam dome. pressure is greater than the: limit, prompt action should be taken to reduce the pressure to below the limit. If the operator is unable to reduce the reactor. steam dome pressure to the limit,.then the reactor is required l to-Se in MODE 3. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably' complete the Required Action. l6 SURVEILLANCE SR 3.4.7.1 V REQUIREMENTS The reactor steam dome pressure is verified to be less than or equal to the limit every 12 hours. This ensures that the initial conditions of the design basis accidents and transients are met. Operating experience has. demonstrated that the-12 hour frequency for this surveillance is adequate. REFERENCES 1. ABWR SSAR, Section 5.2.2.
- 2. ABWR SSAR, Section 15.
- 3. NE00-31466, " Technical Specification Screening Criteri' Application and Risk Assessment," November 1987.
o j 4 i ABWR B 3.4-27 5/31/89 l
ECCS Operating 1' B 3.5.1
/ 'B 3.5. EMERGENCY CORE COOLING SYSTEMS
;V ECCS - Operatina B 3.5.1 BASES BACKGROUND The.ECCS'is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to 'the environment following a loss-of-coolant
' accident (LOCA). The ECCS is supported by other systems which provide automatic ECCS initiation signals (LCO [ECCS Actuation Instrumentation]), cooling water to cool the ECCS. pumps and the rooms containing ECCS equipment (LCO 3.T.1), and electrical power (LCOs 3.8.1 and 3.8.4).- The ECCS injection network is comprised'of the High Pressure Core Flooder-(HPCF) system,.the Reactor Core Isolation Cooling (RCIC) system and the Low Pressure Flooder (LPFL) mode of the Residual Heat' Removal .(RHR)
. system. The-ECCS also consists of the Automat:c Depressuriza-tion System (ADS). The Condensate Storage Pool (CSP)' and suppression pool provide the sources of water for the ECCS although no credit is taken in the safety analyses for.the CSP.
'The ECCS injectici systems are arranged in_three separate divisions each comprised of a high pressure and ' low pressure system. ECCS Division I consists of the RCIC system, and LPFL I- subsystem "A". ECCS Division 2 consists of LPF!. subsystem "B" and HPCF subsystem "B". ECCS Division 3 consists of the HPCF subsystem "C" and the LPFL subsystem "C".
LPFL is an indeper. dent operating mode of the RHR system. In the LPFL mode, the RHR operates to provide the ECCS' function. There are three LPFL subsystems. Each LPFL subsystem (Ref. 2) consists of a motor-driven pump, piping and valves to transfer water from the suppression pool to the reactor vessel. Each LPFL subsystem has its own suction and discharge piping. The water is injected into the reactor vessel outside the core shroud, sia feedwater line B for LPFL subsystem A, and via the individual LPFL subsystem inlet nozzles for LPFL subsystems B and C. The LPFL subsystems are designed to provide core cooling at low reactor vessel pressure. Upon receipt of an initiation signal, each LPFL pump is automatically started (from normal AC power if available; otherwise, the pumps start after emergency AC power becomes available). When the reactor vessel pressure drops sufficiently, LPFL flow to the reactor vessel begins. RHR system valves in the LPFL flow path are automatically positioned to ensure the proper flow path for water from the suppression pool, through the RHR heat exchanger, to injection into the reactor vessel. RHR system valves that service other RHR functions (e.g. shutdown cooling, L fuel pool cooling) must be manually realigned when required (continued) ABWR B 3.5-1 5/31/89
l ECCS - Operating B 3.5.1 BASES (continued) O BACKGROUND following LPFL subsystem operation. A discharge test line is (continued) provided to route water from and to the suppression pool to allow full flow testing of each LPFL pump without injecting water into the reactor vessel. The HPCF system is comprised of two HPCF subsystems. A HPCF subsystem (Ref. 3) consists of a single motor. driven pump, a flooding sparger in the upper plenum of the core shroud, and piping and valves to transfer water from the suction source to the sparger. Suction piping is provided from the CSP and the suppression pool. Pump sucM on is normally aligned to the CSP source to minimize injection of suppression pool water, which is not likely to meet boiling water reactor (BWR) water quality requirements, into the reactor pressure vessel (RPV). However, if the CSP water supply is low or the suppression pool level is above a prescribed level, an. automatic transfer to the suppres-sion pool water source assures a water supply for continuous operation of the HPCF system. The HPCF system is designed to provide core cooling over a wide range of reactor vessel pressures (0 to 1177 psid, vessel to suppression pool air-space). Upon receipt of an initiation signal, the HPCF pump automatically starts (from normal AC power if available, otherwise, the pump starts after emergency AC power becomes available) and valves in the flow path begin to open. Since the HPCF system is designed to operate over the full range of expected reactor vessel pressures, HPCF flow begins as soon as the necessary valves are open. A full-flow test line is provided to route water from and to the suppression pool to allow testing of each HPCF subsystem during normal operation without injecting water in the reactor vessel. The RCIC system (Ref. 1) consists of a steam driven turbine-pump unit, piping and valves to transfer water from the suction source to the core via feedwater line A. Suction piping is provided from the CSP and the suppression pool. Pump suction is normally aligned to the CSP to minimize injection of suppression pool water into the RPV. However, if the CSP water ; supply is low or the suppression pool level is above a pre- l' scribed level, an automidic transfer to the suppression pool assures a water supply for continuous operation of the RCIC system. The steam supply to the turbine is piped from main steam line B upstream of the inboard main steam line isolation valve. The RCIC system is designed to provide core cooling over a wide range of reactor pressures. RCIC is also designed to operate following RPV isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of reactor vessel water level. (continued) ABWR B 3.5-2 5/31/89 l
- ~ - - . _ _ _ _ _ . . _ _
f I ECCS - Operating B 3.5.1 f)7 U " BASES.(continued). All ECCS pumps are provided with minimum' flow bypass lines BACKGROUND:
, (continued) which discharge.to the suppression pool. .The valves in these these lines automatically open to prevent pump damage' due to overheating when other. discharge line valves are. closed or reactor vessel pressure is greater than the LPFL pump discharge pressure following system initiation. To ensure rapid delivery of water to the reactor vessel and to minimize waterhammer effects, the ECCS discharge line keep fill systems are designed to maintain all pump discharge lines filled with water.
The ADS (Ref. 4) consists of 8 of the 18 safety / relief- valves (SRVs). It is designed to provide.depressurization of the primary system during a small break LOCA if HPCF and RCIC fail to. maintain required water level in the reactor vessel. ADS
. operation reduces the reactor vessel pressure to within the operating pressure range.of the low pressure LPFL subsystems, so'these subsystems.can provide core cooling. Each of the SRVs used for automatic depressurization is equipped with a separate dedicated pneumatic accumulator and. associated inlet check valve. The safety-related accumulators provide the pneumatic power to actuate the valves. Nitrogen is normally supplied for operation of the ADS valves from the non-safety-related Atmospheric Control System.
V _ _ . - - - APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY break sizes for a postulated LOCA. The accidents for which ANALYSES ECCS operation is required are specifically listed in SSAR Chapter 15. The required analyses and assumptions are defined in Reference 5. The results of these analyses are described in Reference 6. The ECCS subsystem design requirements ensure the criteria of Reference 7 are satisfied under all postulated LOCA conditions assuming the worst single active component failure in the ECCS. The limiting single failures are discussed in Reference 9. For any LOCA, failure of ECCS subsystems in Division 2 or 3 due to failure of its associated diesel generator is the most severe single failure. )ne ADS valve failure is analyzed as a limit-ing single failure for events requiring ADS operation. The remaining 0PERABLE ECCS subsystems' provide the capability to adequately cool the core and prevent excessive fuel damage. ECCS-Operating satisfies the requirements of~ Selection
- - Criterion 3 of the NRC Interim Policy Statement on Technical s
Specification Improvements as documented in Reference 10.
\
(continued) I ABWR B 3.5-3 5/31/89 j i
ECCS - Operating B 3.5.1 BASES (continued) O LCOs The ECCS injection subsystems are defined as the three LPFL i subsystems, the RCIC system and the two HPCF subsystems. The divisional arrangement of the ECCS subsystems and a description of what is required for the ECCS to be considered OPERABLE is provided in the Background Section. With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst single failure, the limits specified in Reference 7 could be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 7. A LPFL subsystem may be considered OPERABLE during alignment to and operation in the RHR shutdown cooling mode when below the RHR cut-in permissive pressure in MODE 3, if capable of being manually realigned to the LPFL mode and not otherwise inoper-able. At these low pressures and decay heat levels (reactor is shutdown in MODE 3) a reduced complement of ECCS subsystems can provide the required core cooling thereby allowing operation of an RHR shutdown cooling loop when necessary. APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2 and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3 the ADS function is not required when pressure is s 50 psig because the low pressure ECCS subsystems (LPFL) are capable of providing flow into the reactor vessel below this pressure. ECCS requirements for these conditions are specified in LC0 3.5.2. The RCIC system is required to be OPERABLE in MODES 1, 2 and 3 with reactor steam dome pressure 2 150 psig since RCIC is a steam driven water source which provides water for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 j with reactor steam dome pressure < 150 psig, and in MODES 4 and j 5, RCIC is not required to be OPERABLE since the other ELCS 1 3 subsystems can provide sufficient flow to the vessel. j (continued) O ABWR B 3.5-4 5/31/89 1 b__ _
ECCS - Operating B 3.5.1 7, 9ASES (continued)-
' . ACTIONS- A.l. B.1. B.2. C.1. C.2. C.3 With three or less ECCS-injection subsystems from Divisions 1, 2, and 3 inoperable, the remaining OPERABLE subsystems provide
' adequate core cooling during a LOCA. However, the overall ECCS reliability is-reduced because a single failure in one of the remaining subsystems concurrent with a LOCA...y result in the-ECCS not being able to perform its intended safety function.
'Therefore, continued operation is only allowed for a limited time.
D.1. D.2 Should the Required Actions and associated Completion Times of Conditions A, B, or C not be met, the reactor is required to be in MODE 3 and subsequently in MODE 4. Of course for the-specific case where RCIC is the only inoperable subsystem, once reactor pressure is reduced below 150 psig, the LC0 is no longer applicable and no further ACTION is required. In MODE 4, the ECCS requirements are specified in LC0 3.5.2. If untble to attain MODE 4, the reactor coolant temperature should be maintained as low as practical by use of alternate heat removal' [. methods. U)- E.1. F.1. F.2. G.I. G.2
.The LC0 requires 8 ADS valves to be OPERABLE to provide the ADS function. Analysis demonstrates that only three ADS valves are required for the ADS to successfully perform its function.
'With one or two ADS valves _ inoperable, the reliability of the ADS is relatively unaffected. Operation is allowed until the next outage of 30 days initial planned length, or until the refueling outage, whichever comes first. As noted, the provisions of LC0 3.0.4 are not applicable for this Condition to allow recovery from shutdowns of short duration prior to the next refueling outage and/or extended maintenance outage. When three ADS valves are inoperable, operation is allowed to continue for 30 days to allow time to attempt repair and/or to plan for a plant outage to affect repairs. Three ADS valves out-of-service is considered indicative of a generic problem, and continued plant operation beyond the 30 days is not allowed. If the required number of inoperable ADS valves cannot be made OPERABLE, the plant is required to be in MODE 3 and the reactor pressure reduced to s 50 psig. At these conditions the ADS function is no longer required since the reactor pressure is low enough such that the low pressure ECCS subsystems can perform their designed safety function.
n (continued) ABWR B 3.5-5 5/31/89
ECCS - Operating B 3.5.1 BASES (continued) ACTIONS These action statements assume that the valve failure is in an (continued) inaccessible area of the plant. Since the intent of this LC0 is to have all ADS valves operable at all times, all ADS valve failures which can be repaired in areas of the plant accessable during normal operation should be repaired within a reasonable period of time. Completion Times The ECCS Completion Times are based on the results of a study which evaluated the impact on- ECCS unavailability assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowable outage times (A0T). A0Ts were then chosen to provide comparable levels of ECCS availability. j i SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The pump discharge lines of the RCIC system, and the LPFL and HPCF subsystems are required to be kept full with water to minimize potential waterhammer effects when the systems are & initiated. Additionally, the lag between the receipt of the W initiation signal and the actual injectiun into the reactor vessel is minimized. One acceptable method of ensuring the lines are " full" is to vent at the high points. SR 3.5.1.2 Verification that all applicable valves are in the required position ensures proper flow paths for ECCS. However, a valve that is capable of automatic return to its ECCS position, when an ECCS initiation signal is present, can be in position for another mode of operation. This is applicable only if the valve auto-depositions and fully strokes within the time required for its ECCS function. For the RCIC system, this also includes the steam flow path for the RCIC turbine. This surveillance also includes the RCIC flow controller position. SR 3.5.1.3 The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed pneumatic supply pressure requirements for the accumulator are such that following a failure of the pneumatic supply to the accumulator, at least one valve actuation can occur with the drywell at design (continued) ABWR B 3.5-6 5/31/89
ECCS - Operating B 3.5.1 p BASES (continued) 1v) SURVEILLANCE pressure, or five valve actuations with the drywell at i REQUIREMENTS atmospheric pressure. The ECCS safety analysis assumes only (continued) one actuation to achieve the depressurization required for operation of the low pressure ECCS and therefore this require-ment provides sufficient margin to satisfy the assumptions of the safety analyses. This minimum required pressure of 151 psig is provided by the Atmospheric Control System. i SR 3.5.1.4. 3.5.1.5. 3.5.1.6 The performance requirements of the ECCS pumps are determined through application of the 10CFR50 Appendix K criteria -(Ref. 5). The pump flow rates, as determined by analysis, ensure ; that adequate core cooling is provided to satisfy the accept-ance criteria of Reference 7. Periodic surveillance is per-formed (in accordance with ASME Section XI requirements) to verify these flow rates. The pump flow rates are verified against a system head that is equivalent to the reactor vessel pressure expected during a LOCA. The total system head devel-oped is adequate. to overcome the elevation differences between the suction source and the vessel, friction losses and pressure differences present during LOCA. These values are established
^ during preoperational testing.
\~ The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC system are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating range of the system. Since the required reactor steam dome pressure must be available to perform SR 3.5.1.5 and SR 3.5.1.6, sufficient time is allowed after adequate pressure is achieved to perform these tests. SR 3.5.1.7 The ECCS subsystems are required to actuate automatically to perform their designed function. These surveillance tests demonstrate that with the required system initiation signals, the automatic initiation logic of HPCF, RCIC and LPFL will cause them to operate as designed, including actuation of all automatic valves to their required position. This test also ensures that the RCIC system will automatically restart on a reactor vessel low water level (Level 2) signal received subsequent to reactor vessel high '.sater level (Level 8) trip. This test also ensures that the HPCF subsystem and RCIC system will automatically transfer suction from the CSP to the suppression pool. Since all active components are testable and [,') v (continued) , ABWR B 3.5-7 5/31/89
)
1 ECCS - Operating l B 3.5.1 BASES (continued) O SURVEILLANCE full flow can be demonstrated by recirculation through the test REQUIREMENTS line, coolant injection into the reactor vessel is not required (continued) during the tests. SR 3.5.1.8 The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system func-tional test (logic only) is performed to demonstrate that the ADS logic operates as designed when initiated, causing proper actuation of the required components. Actual ADS valve actu-ation is excluded to prevent a reactor pressure vessel blowdown. SR 3.5.1.9 A manual actuation of ADS valves is performed to verify that the valve and solenoids are functioning properly and that no blockage exists in the SRV discharge lines. This is demon-strated by actual valve stem movement (as indicated by LVDT) or by any suitable method to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed, after the required pressure is achieved, to perform this test. Adequate pressure at which this test is to be performed is [ ] psig (the pressure recommended by the valve manufacturer). Reactor startup is allowed prior to performing this test because valve OPERABILITY and the set-points for overpressure protection are verified, per ASME requirements, prior to valve installation. Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status.. (continued) Oi ABWR B 3.5-8 5/31/89 i
1 ECCS - Operating
^
B 3.5.1 hV
~
. BASES (continued) -
REFERENCES 1. ~ABWR SSAR, Section 6.3.2.2.3.
'2. .ABWR'SSAR, Section 6.3.2.2.4'.
- 3. ABWR SSAR, Section 6.3.2.2.1.
i' .4. ABWR SSAR, Section 6.3'2.2.2..
- 5. 10CFR50, Appendix K, "ECCS Evaluation Models".
'6. _ABWR.SSAR, Section 6.3.3.
,7. 10CFR50.46, " Acceptance Criteria for; Emergency Core Cooling Systems for Light-Water-Cooled. Nuclear. Power Reactors".
8; 52FR3788,- Proposed Policy Statement on Technical Specification. Improvements for Nuclear Power Reactors,. February 6, 1987.
- 9. ABWR SSAR, Section 6.3.3.3.
.'w r 10. " Technical Specification Screening Criteria Application:
and Risk Assessment", NED0-31466, November 1987. 4
- I ABWR B 3.5-9 5/31/89
~ _ _ _ _ - _-- ____________n
ECCS - Shutdown B 3.5.2 8 3.5 EMERGENCY CORE COOLING SYSTEMS B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the High Pressure Core Flooder (HPCF) and the Low Pressure Flooder (LPFL) subsystems of the Residual Heat Removal (RHR) system are provided in the Bases for LC0 3.5.1. APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY break sizes for a postulated Loss of Coolant Accident (LOCA). ANALYSES The long-term cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one ECCS subsystem is required, post-LOCA, to maintain the peak cladding temperature below the allowable limit. To preserve the single failure criterion of Reference 2, a minimum of two ECCS subsystems are required to , be OPERABLE in MODES 4 and 5. Two OPERABLE ECCS subsystems also ensure adequate vessel inventory makeup in the event of an inadvertent vessel draindown. The consequences of an inadvert-ent draindown of the vessel, which may require ECCS subsystem operation during MODES 4 and 5, are bounded by these analyses. ECCS-Shutdown satisfies the requirements of Selection Cri-terion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. LCOs The ECCS injection subsystems are defined for the purposes of this specification as the three LPFL subsystems and the two HPCF subsystems. Each LPFL subsystem consists of one motor driven pump, piping and valves to transfer water from the i suppression pool to the reactor vessel. Each HPCF system ! consists of one motor driven pump, piping and valves to trans-fer water from the suppression pool or Condensate Storage Pool (CSP) to the reactor vessel. Any LPFL subsystem that may be aligned in the shutdown cooling mode of the RHR system in MODES 4 Jr 5, is considered OPERABLE for the ECCS function, if it can to manually realigned remotely or manually to the LPFL mode and is not otherwise inoperable. Because of low pressure and temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPFL subsystem operation to provide core cooling prior to postulated fuel uncovery. (continued) O , ABWR B 3.5-10 5/31/89
,y ^
- ECCS - SNtdown d 3.5.2 )
i J [v[ BASES (continued).
\
APPLICABILITY ECCS OPERABILITY is required:in MODES 4 and 5 to assure
. adequate coolant inventory and-sufficient heat removal' l capability for the irradiated fuel in the core in case'of an i
-inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2 and 3 are discussed in the
. Applicability section for LCO. 3.5.1. ECCS subsystems are not required to be OPERABLE during MODE 5' with the spent' fuel pool gate removed and the water level maintained greater than or. i equal to 23' above.the reactor pressure vessel flange. .This.
provides sufficient coolant inventory to allow operator ' action:
]
to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. The Automatic Depres'surization System (ADS) is not required to be OPERABLE during MODES.4 and 5. because the reactor. vessel pressure is < 50 psig and LPFL and HPCF subsystems can provide core cooling without any depressurization of the primary system. being required. Since the Reactor Core Isolation Cooling (RCIC) system . requires'- steam to operate, it is not required to be OPERABLE during MODES 4 and'5. ACT10f45 A.1. B.I With one of the two. required ECCS subsystems inoperable, the remaining OPERABLE subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, system reliability is reduced because a single failure in the remaining subsystem concurrent with a vessel draindown could result in the ECCS.not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time. With the inoperable subsystem not restored to OPERABLE status within the required Completion Time, operations that have the potential for drain-ing the reactor vessel must be suspended. This minimizes the. probability of a vessel draindown and the subsequent potential-- for ECCS actuation. (continued) i ABWR B 3.5-11 5/31/89
f f ECCS - Shutdown l B 3.5.2 BASES (continued) .l I ( ACTIONS C.l. C.2. 0.1. D.2. D.3. 9.4 With both of the required ECCS subsystems inoperable, all l coolant inventory makeup capability may be unavailable and operations that have a potential for draining the reactor vessel must be suspended. If at least one ECCS subsystem is not restored to OPERABLE status within the required Completion Times, additional actions are required to minimize any poten-tial release of radioactive materials to the environment. This includes ensuring Secondary Containment is OPERABLE, at least one Standby Gas Treatment System (SGTS) subsystem is OPERABLE and at least one Secondary Containment Isolation Valve and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. This may be performed by an administrative check, by examining logs or other inform-ation, to determine if the components are out of service for maintenance or other reasons. It does not mean to perform the surveillance needed to demonstrate OPERABILITY of the compon-ents. If however, any required component is inoperable, it must be restored to OPERABLE status. In this case, surveil-lance requirements may need to be performed to restore the component to OPERABLE status. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available. systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.5.2.1 REQUIREMENTS The minimum water level 14.63 feet required for the suppression pool is verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the RHR pumps, including consideration of a recirculation volume, vortex prevention and a safety margin for conservatism. ?ne suppression pool water level is referenced to the inside bottom of the suppression pool. With the suppression pool water level less than the required limit, all LPFL subsystems are inoperable. (continued) O ABWR B 3.5-12 5/31/89
J ECCS - Shutdown B 3.5.2 BASES (continued) SURVE1LLANCE SR 3.5.2.2 REQUIREMENTS (continued) When suppression pool level is less than 14.63 feet, HPCF is considered OPERABLE only if it can take suction from the CSP and the CSP water level is sufficient to provide the required NPSH for the HPCF pump. Therefore, a verification that either the suppression pool water level is 214.63 feet or HPCF is aligned to take suction from the CSP, and the CSP contains 285,284 gallons of water, equivalent to 22.8 feet water level, . ensures HPCF can supply makeup water to the reactor vessel. l Suppression pool water level is referenced to the inside bottom of the suppression pool. SR 3.5.2.3. 3.5.2.5. 3.5.2.6 1 The bases provided for SR 3.5.1.1, 3.5.1.4, and 3.5.1.7 are applicable. l SR 3.5.2.4 Verification that all applicable valves are in the required position ensures proper flow paths for ECCS. However, a valve that is capable of automatic return to its ECCS position, when an ECCS initiation signal is present, can be in position for 9 another mode of operation. In MODES 4 or 5, the RHR system may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, during MODES 4 or 5, RHR valves that are required for LPFL subsystem operation may be aligned for the shutdown cooling mode. The LPFL mode of the RHR however, may be considered OPERABLE for the ECCS function if all the required valves in the LPFL flow path can be manually realigned (remote or local) to allow injection into the RPV and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent vessel draindown should occur. Surveillance Frequencies In general, surveillance frequencies are based en industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) O ABWR B 3.5-13 5/31/89
ECCS - Shutdown 1 l B 3.5.2 BASES'(continued) O REFERENCES 1. ABWR SSAR, Section 6.3.
- 2. 10CFR50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors".
1
- 3. " Technical Specification Screening Criteria Application l and Risk Assessment", NE00-31466, November 1987. J i
i 1 i O
I i Primary Containment ) B 3.6.1.1 l J I
')
wJ B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES l I BACKGROUND The function of the Primary Containment is to isolate and contain fission products released from the reactor primary system in the event of a Loss of Coolant (LOCA). The Primary Containment consists of a reinforced concrete containment vessel with a steel liner which provides a leak-tight barrier surrounding the reactor primary system and includes the upper and lower dryuell and the wetwell consisting of the suppression pool and wetwell airspace. To assure Primary Containment OPERABLE, leakage test requirements have been set forth by Reference 1. These test requirements pnvide for periodic verification by tests of the leak-tight integrity of the primary containment and systems and components which penetrate the primary containment. The purpose of the leak tests are to assure that leakage through the primary containment and systems and components penetrating the primary containments shall not exceed the allowable leakage rates specified in the technical specification and used in the w safety analyses. Additionally the periodic tests performed
,) assure that proper maintenance and repairs are made during the service life of the plant.
All leakage rate requirements and surveillance requirements are ! in conformance with 10 CFR 50 Appendix J (Ref. 1) and approved exemptions. Periodic Type A tests can use the peak calculated containment pressure -f 45.0 psig (Ref. 2) or a reduced pressure of Pt, [ ] psig. The maximum allowable leakage rate for the primary containment is 0.5 percent by weight of the containment air per 24 hours at Pa, 45.0 psig or [ ] percent by weight of the containment air per 24 hours at the reduced pressure of Pt, [ ] psig (Ref. 2). The maximum allowable leakage rate is based on what is acceptable for nuclear safety considerations per 10 CFR 100. Reactor size, site location and meteorology, as well as the possible mechanisms for radioactivity generation and transport within the containment vessel are all considered in specifying the allowable leakage rate for a given containment system. Generally the lowest leakage rate which is readily obtainable and measurable is specified even though this may be lower than required by safety considerations. (continued) (] t-ABWR B 3.6-1 5/31/89
Primary Containment B 3.6.1.1 BASES (continued) BACKGROUD The maximum allowable leakage rate is used as an input to the i (continued) safety analysis of Reference ? which confirms that the specified maximum allowable leakage rate will conform to the requirements of 10 CFR 100. To maintain primary containment integrity the drywell bypass leakage must te minimized to prevent overpressurization of the wetwell during the drywell pressurization phase of a LOCA. This requires periodic testing of the drywell bypass leakage , and confirmation that the wetwell to drywell vacuum breakers- l are closed per LC0 3.6.1.6. ! This specification assures that the performance of the primary containment in the event of a LOCA meets the requirements of Reference 4 and the assumptions used in the safety analyses of References 2 and 3. i APPLICABLE Analytical methods and assumptions involving the primary SAFETY containment are presented in Reference 2 and 3. The safety ANALYSES analyses assume a non-mechanistic fission product release following a LOCA which forms the basis for determination of l offsite doses. The fission product release is in turn based on 1 an assumed leak rate from the primary containment. OPERABILITY of the Primary Containment assures that the leak rate assumed in the safety analyses is not exceeded and that the site boundary radiation dose will not exceed the limits-of 10 CFR 100 even if the non-mechanistic release were to occur. i Primary Containment satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 5. I LC0 The Primary Containment must be OPERABLE to assure that the containment conditions, passive features, and active features are consistent with those assumed in the safety analyses. In order for the Primary Containment to be considered OPERABLE, only those SRs listed in this LC0 must be met. APPLICABILITY Maintenance of Primary Containment OPERABILITY is applicable during MODES 1, 2 and 3 since it is required when sufficient energy is contained in the reactor coolant system to pressurize the containment. Primary Containment is required and is assumed in the LOCA analyses of Reference 2, where pressurization of the primary containment occurs. A LOCA in (continued) ABWR B 3.6-2 5/31/89 ________a
1 w ,
>- Primary Containment i
.B 3.6.1.1-
- m a
- n BASES (continued)-
~
Q h: *
' APPLICABILITY MODE 4 or 5;will not pressurize 'the primary containment 1and (continued)- therefore its OPERABILITY is not required. 'The analyses of:.
Reference:
3 which assumes;and requires primary containment' also assumes the' reactor _is' pressurized-(i.e., MODE 1, 2 or 3). ACTIONS' ;A.l. B.1 and B.2 q With Primary Containment inoperable'the leak rate may exceed i
.that assumed.in the design basis accident analyses. - A short )
time .is allowed to restore Primary Containment _ to_ OPERABLE status due to the low probability.of an event which would pressurize thel primary containment. However, if Primary
~
Containment cannot be. restored to OPERABLE status, the reactor I is required to be in MODE 3 and subsequently in MODE 4. t Completion Times All Completion Times are based on industry accepted practice' and engineering judgement considering the number'of available systems and the time required to reasonably ~ complete the'
~ Required Action'.
1 J t 1 _ SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS Maintaining the pressure suppression function of Primary Containment requires limiting-the leakage from.the drywell to . the wetwell. Thus, if an event were to occur which pressurized the drywell', the steam would be' directed through the downcomers into'the suppression pool. This SR measures drywell-to-wetwell differential pressure over a ten minute period-to ensure that the leakage paths which would bypass the suppression pool are within allowable limits. SR 3.6.1.1.2. SR 3.6.1.1.3 Maintaining Primary Containment OPERABLE requires _ compliance with the leak test requirements-of 10 CFR 50, Appendix J. Therefore, these two SRs reflect the leak rate testing requirements with regard to overall containment leakage (Type A and B leak tests). Other specific Appendix J requirements are addressed in the individual containment component Surveillance Requirements. Type A and 8 testing must be performed in accordance with 10 CFR 50, Appendix J or NRC approved exempt-ions to Appendix J. These periodic testing requirements verify that the containment leak rate does not exceed the leak rate-assumed in the safety analysis. The surveillance frequency is required by Appendix J, and as such, SR 3.0.2 (which' allows surveillance frequency extensions) does not apply. 3 (continued) T# ABWR B 3.6-3 5/31/89
Primary Containment B 3.6.1.1 BASES (continued) SURVEILLANCE Surveillance Frequencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. 10 CFR 50, Appendix J.
- 2. ABWR SSAR, Section 6.2.
- 3. ABWR SSAR, Section 15.6.5.
, 4. 10 CFR 50, Appendix A, GDC 16.
- 5. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
1 O i O ABWR B 3.6-4 5/31/89
Containment Air Loc.ks-B 3.6.1.2 i D. B 3.6 CONTAINMENT SYSTEMS' ( Nf B 3.6.1.2 Containment Air Locks BASES BACKGROUND' Two double door Containment Air Locks h' ave been built into the containment which provide access while' also assuring that the containment remains isolated during the process of personnel entering and exiting the primary containment. Each air lock is designed with a seal system for each door in the air lock as well as an interlock mechanism that assure one door is closed and sealed before the other door is opened. The air locks are designed to withstand the same loads, temperatures and peak
- design-internal and external pressures as the containment (Ref.
'2). As~ part:of the containment boundary, the air locks limit the release of radioactive material to the environs during normal plant operation and-through a range of incidents up to.
and including (postulated) design basis accidents. The containment, including containment air locks, is designed to maintain its functional integrity during and _ following the peak transient pressures and temperatures that would occur following any postulated loss-of-coolant accident. T' The air lock doors have inflatable seals which are maintained
\ at { -] psig by the seal air flask / pneumatic system which is-maintained at [ ] psig. Each door has two seals so that they are single failure proof in maintaining the leak tight boundary of primary containment.
APPLICABLE Primary Containment OPERABILITY, and the limiting of radio-SAFETY- active material release to the environs, is a consideration ANALYSES in the evaluation of a number of accident analyses presented in FSAR Chapters 6 and 15. The Containment Air Locks are a part of the Primary Containment boundary. Their design and maintenance.are essential to maintaining Primary Containment OPERABILITY which assures that the safety analysis assumptions are maintained. Containment Air Locks satisfy the requirements of Selection Criterion 3 of the NRC Interin Policy Statement on Technical Specification Improvements as documented in Reference 3. (continued)
-d.
ABWR B 3.6-5 5/31/89
Containment Air Locks B 3.6.1.2 BASES (continued) LC0 The Containment Air Locks are required to be OPERABLE with at l least one door closed. For the air locks to be considered OPERABLE, the door seals must not leak, the interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock tests, and both doors in each air lock must be OPERABLE. This requirement is necessary to maintain contain-ment integrity when access in and out of the containment is required. APPLICABILITY The containment air locks are required to be operable during MODES 1, 2 and 3 consistent with the requirement for when Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to pressurize containment such that a containment air lock failure would not threaten i plant safety. ! ACTIONS A.1. A.2. A.3 When one air lock door is inoperable, or one door and the interlock mechanism on the same air lock are inoperable, then the air lock will not perform its intended function of allowing personnel access while maintaining the primary containment boundary intact. It will however, maintain the boundary intact indefinitely if the remaining OPERABLE door in the air lock is closed. Therefore, for this condition, the OPERABLE .cor is closed immediately. If the components cannot be restored to OPERABLE status, the OPERABLE door must be locked and a periodic surveillance performed to confirm it is locked. The provisions of LC0 3.0.4 are not applicable and continued plant operation is justified since there is another containment air lock which is OPERABLE and because the primary containment boundary is maintained intact. (continued) O ABWR B 3.6-6 5/31/89 l
Containment Air Locks B 3.6.1.2 f'N BASES (continued) V ACTION A.1. A.2. A.3 (continued)
-(continued)' .
.If the outer door is inoperable, then it may be easily accessed to repair. H m ver, if it is-stuck in position and must be at.sessed th bogh the containment via the inner door, or if the inner door is'the one which is inoperable, then it may require a .short time (during access only) when the containment boundary is not intact. The unlimited ability to open the OPERABLE door, even if it means the primary containment boundary is temporarily not; intact, is allowed for 24 hours until the OPERABLE door is locked closed. After that, the cumulative time the OPERABLE door may be open for access under admini-strative control must not exceed one hour per year. This is acceptable due to the low probability of an event which would pressurize primary containment during the short time in which the OPERABLE door is open.
B.1. B.2. B.3 If the interlock mechanism is inoperable in one or;both air ; locks, then either OPERABLE air lock door can be closed and the' primary' containment boundary remains intact. Therefore, an OPERABLE door must be closed and if the interlock cannot be g restored to OPERABLE status, the door inust be locked closed and
;h periodically varified to be locked. However, personnel access '\~/ under administrative control. is allowed at any time during continued plant operation where the administrative control procedure will assure that when the locked door is unlocked the interlock function will be performed manually. Therefore, the integrity of the primary containment boundary will not be threatened by the failure of only the air lock interlock mechanism.
[.l. C.2 If the air lock is inoperable for reasons other than Condition A or B (e.g. a problem with the barrel or problems with both doors in an air lock) then the containment boundary may not be intact even when the doors are closed. One door in the air lock is closed to minimize the affect of any potential release to Primary Containment. Twenty-four hours is allowed to restore the air lock to OPERABLE status due to the low proba-bility of an event which could pressurize Primary Containment. a (continued) ABWR B 3.6-7 5/31/89 _ _ __o
Containment Air Locks B 3.6.1.2 BASES (continued 1 ACTIONS Ll. D.2. 0.3 STATEMENTS (continued) When Condition A (one door or one door and associated interlock inoperable) applies to both air locks, the one OPERABLE door in each air lock must be closed and locked. With a problem in both air locks, it is not possible to access primary contain-ment through an OPERABLE air lock. Furthermore, since one door is inoperable in each air lock, the primary containment boun-dary will not be intact during the short time period when the OPERABLE door is opened for access. It is not reasonable to continue plant operation with both air locks in this condition. Therefore, at least one air lock must be restored to OPERABLE status within 7 days. L.1 If Condition A applies to one air lock and Condition B applies to the other air lock, the Required Actions of both Conditions are to be followed. The actions apply to the respective air locks and provide appropriate controls for the specified conditions. F.1. F.2 If the Required Actions and associated Completion Times of Condition A, B, C, D or E are not met or both air locks are inoperable for reasons other than Condition D or E (i.e., they are more severely damaged than an inoperable door and interlock function on one air lock and an inoperable interlock function in the other air lock), then the reactor is required to be in MODE 3 and subsequently in MODE 4 where the air locks are no longer required to be OPERABLE. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. (continued) I O ABWR B 3.6-8 5/31/89
e b.,, ' Containment Air Locks B 3.6.1.2 g
/' ' BASES fcontinued) 3 l SURVEILLANCE SR 3.6.1.2.1 I' REQUIREMENTS
_(continued) The seal air flask pressure is verified to be 2 [ ] psig every 7 days to ensure the seal system remains viable. SR 3.6.1.2.2 The interlock mechanism is verified to be OPERABLE to assure that both doors cannot be inadvertently opened simultaneously. This is not verified when the containment'is inerted because of the oxygen which is released into containment when the inside air lock door is opened. It is performed prior to entry into the primary containment, unless it has been performed within the previous 6 months. If primary containment remains ir.erted for longer than 6 months. If primary containment remains inerted for longer than 6 months, then it is tested prior to subsequent entry when the containment has been de-inerted. SR 3.6.1.2.3 A seal pneumatic system test to assure that pressure does not decay > [ -] psig in a [ ] hour period is an effective leak
. rate test to verify system performance.
SR 3.6.1.2.4 Maintaining Primary Containment requires compliance with the leak test' requirements of 10 CFR 50, Appendix J (Ref. 1). Therefore, this surveillance reflects the leak rate testing requirements with regard to containment air lock ~ leakage (Type Bleaktests). Type B testing must be performed in accordance with Reference 1 or NRC approved exemptions. These periodic testing requirements verify that the containment leak rate does not exceed the leak rate assumed in the safety analysis. The SR states the overall air lock leak rate and door seal leak rate acceptance criteria as required by 10 CFR 50 Appendix J. The surveillance frequency is. required by Appendix J, and as such, SR 3.0.2 (which allows surveillance frequency extensions) does not apply. f Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) O ABWR B 3.6-9 5/31/89
i. i Containment Air Locks B 3.6.1.2 BASES (continued) REFERENCES 1. 10 CFR 50, Appendix J.
- 2. ABWR SSAR, Section 6.2.
- 3. NED0-31460, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
l 9 O ABWR B 3.6-10 5/31/89
[ Primary Containment Pres'sure l 3 3.6.'I.3~ B 3.6 CONTAINMENT. SYSTEMS' [()- B 3.6.1.3 Primary Containment Pressure SASES BACKGROUND The limit on the primary containment pressure was developed as a reasonable upper bound based on operating plant experience. A positive internal pressure assures that air does not leak in to a deinerted containment. The maximum limitations on the primary containment pressure ensure that the peak LOCA primary containment. pressure does not exceed the' maximum allowable pressure of 45 psig (Ref.1) and also prevents inadvertent scrams due to high drywell pressure. A?PLICABLE Containment performance is evaluated for the entire spectrum SAFETY of break sizes for postulated Loss-of-Coolant Accidents (LOCA). ANALYSES Among the inputs to the design basis analysis is the initial containment internal pressure (Ref.1). Analyses assume an initial primary containment pressure of 0.75 psig. This limitation assures that the peak LOCA containment internal pressure does not exceed the maximum allowable of 45 psig. j Containment Pressure satisfies the requirements of Selection
' Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2.
LC0 A maximum allowable primary containment pressure of 0.75 psig is required to assure that containment initial conditions are consistent with the safety analyses assumptions so that peak-LOCA internal containment pressures are within the design-val ue. 1 APPLICABILITY The primary containment pressure limitation is applicable during MODES 1, 2 and 3 consistent with the requirement for when the Primary Contatament must be OPERABLE. In MODE 4 and 5 there is insufficient energy in the reactor to pressurize containment beyond the calculated design basis value even if L the initial internal pressure is not within its limits. l (continued)
'f~%
J ABWR B 3.6-11 5/31/89 _ _ __ a
Primary Containment Pressure B 3.6.1.3 BASES (continued) ACTIONS A.I. B.1. B.2 With the primary containment pressure not within the limit, the conditions in the primary containment will not be consistent with those assumed in the Reference 1 safety analysis. However, continued operation is allowed for a limited time while the containment internal pressure is restored to within limits due to the low probability of an event which would pressurize the primary containment. If the required primary containment pressure is not restored to within the limit and the associated Completion Time has expired, then the reactor is required to be in MODE 3 and subsequently in MODE 4. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS The Reference 1 analysis assumes an initial primary containment pressure of 0.75 psig prior to a postulated LOCA. Surveillance of the primary containment pressure assures consistency with this assumption.
, Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status.
REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.6-12 5/31/89
Q i a Drywell Avg Air Temperature B 3.6,1.4 B 3.6 CONTAINML; - SYSTEMS [] ' A> B 3.6.1.-4 Drywell Averace Air Temperature BASES BACKGROUND The drywell;contains the reactor vessel and piping which add heat to.the airspace. Drywell coolers remove heat and maintain a suitable. environment. The average. airspace temperature effects equipment OPERABILITY and life, personnel access, and'
-the calculated response to postulated design basis events. The limitation on the drywell average air temperature was developed as. a reasonable upper bound based on. operating plant experience. The limitation on drywell temperature is used in the Reference I safety analysis. .Although used as an initial-condition for safety analyses', recent . studies have shown that the initial drywell air temperature has negligible effect;on the peak' calculated drywell pressure and temperature.
APPLICABLE Containment performance is evaluated for the entire spectrum of~ SAFETY- break sizes _for postulated Loss-of-Coolant Accidents (LOCA). ANALYSES Among the inputs to the design basis at 'vsis.is the initial Drywell average air temperature (Ref.1). Analyses assume an O initial average drywell air temperature of 135'F. Drywell Average Air Temperature satisfies the requirements of Selection Criterion 2.of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. LCO A maximum allowable drywell air-temperature of 135'T is - required to assure that the initial drywell air temperature-assumed for the safety analysis _ is not exceeded. APPLICABILITY The drywell average air temperature limitation.is applicable during MODES 1, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODE 4 and 5 there is insufficient energy in the reactor to heat up containment beyond the calculated design basis values even if the initial temperature was above its limit. (continued) I LO ABWR B 3.6-13 5/31/89 L . - _ - _ - - _ - - _ - - - - _ _
Drywell Avg Air Temperature B 3.6.1.4 BASES (continued) ACTIONS _A_,1. 8.1. B.2 With the drywell average temperature limitation not met, an initial assumption in the analysis of the containment response to a LOCA is not met. However, continued operation is allowed for a limited time while attempting to restore temperature to within limits due to the low probability of an event which would pressurize the primary containment. If the drywell average temperature is not restored within the limit and the associated Completion Time has expired, the reactor is required to be in MODE 3 and subsequently MODE 4. Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available i systems and the time required to reasonably complete the Required Action. 1 SURVEILLANCE 1R 3.6.1.4.1 REQUIREMENTS Surveillance of the drywell average air temperature assures that the drywell air temperature remains below the specified limit. Drywell air temperature is monitored in all quadrants and at various elevations. The instruments are uniformly g distributed so that an arithmetic average is an accurate representation of the actual average temperature. Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. { I l REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
I 1 I ABWR B 3.6-14 5/31/89
Primary-Containment and Pressure Isolation Valves B 3.6.1.5 l ft B 3.6 CONTAINMENT SYSTEMS
- , ("l-B 3.6.1.5 Primary Containment and Pressure Isolation Valves DASES BACKGROUND The function of the Primary Containment isolation' valves (PCIV), in combination with other accident mitigation systems, is to limit fission product release during.and following the postulated design basis accident (Ref. 4) to values'less than that which would result in offsite doses greater than 10 CFR 100 limits.. Containment isolatioa within the time limits specified for those isolation valve
- designed to close' auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
The PCIVs are required by General Design Criterion (GDC) 54, 55, 56 and 57 (Ref. 2). These GDCs set guidelines for the isolation capability of lines that penetrate primary contain-ment. The' requirements depend upon if the line connects to the containment air space, is part of the reactor coolant pressure boundary, or connects to some other component inside contain-ment. The GDC. includes requirements for design and leakage.. This LCO is intended to ensure that those GDC requirements-are
/~% satisfied during plant operation.
U The function of the Pressure Isolation Valves (PIVs) is to 1 protect the low pressure portion of systems which are connected to the primary reactor coolant system. All PIVs are designed to meet the requirements of Reference 3. APPLICABLE The analytical methods and assumptions used in evaluating the. , SAFETY radiological consequences of design basis accidents are pre-ANALYSES sented in Reference 4. The special assumptions used in , Reference 4 regarding leakage of the HSIVs is discussed in the J Bases for SR 3.6.1.5.9. The containment isolation valves are also assumed to function so that the bases for the Reference 5 containment response analysis is valid. The maximum isolation times for Primary Containment isolation valves are consistent l with those used in the Reference 4 accident analysis The Primary Containment isolation valves satisfy the require-ments of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented ! in Reference 5. (continued) ABWR B 3.6-15 5/31/89
)
Primary Containment and Pressure Isolation Valves B 3.6.1.5 BASES (continued) LC0 The OPERABILITY of the primary containment isolation valves ensure that the containment atmosphere will be isolated from the outside environment during an event with the potential for release of radioactive material to the containment atmosphere or pressurization of the containment. The OPERABILITY of the Pressure Isolation Valves ensures that low pressure portions of plant systems will be protected. By checking valve OPERABILITY. on any penetration which could contribute a large fraction of the design leakage, the total leakage is maintained at less than the design value. The isolation valves are considered OPERABLE when their isola-tion times are within limits, the valves isolate on a simulated automatic isolation signal, the valves meet leakage rate requirements, and excess flow check valves actuate within the required differential pressure range. APPLICABILITY The Primary Containment and Pressure Isolation Valve functions are required for all modes which could involve significant releases of fission products to the environment. This includes MODES 1, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to cause a significant release even if the primary containment were not isolated. However, certain valves are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LC0 [ Primary Containment Isolation Instrumentation] (this does not include the excess flow check valves which isolate the associated instrumentation). ACTIONS A.1 With one or more isolation valve inoperable, a remaining OPERABLE isolation valve in an affected penetration is adequate to perform the intended isolation function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE isolation valve could result in the loss of the isolation function for that penetration. If the affected penetration has two isolation valves, it must be verified that the other valve is OPERABLE. This action is satisfied by ' examining logs or other information, to determine if the valve is out of service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate OPERABILITY of the valve. If tht 2.'fected penetration is designed with only one isolation valve, the (continued) O ABWR B 3.6-16 5/31/89
Primary Containment and Pressure Isolation Valves-B 3.6.1.5 BASES (continued) v ACTIONS- -penetration is such that plant safety is not unduly threatened
'(Continued) by a temporary inability to isolate the penetration.
A.2.1. A.2.2.1. A.2.2.2. A.2.2.3 In addition to Required Action A.1, the inoperable valve (s) must be restored to OPERABLE status, or the line must be.. manually isolated. Acceptable isolation methods for PCIVs and PIVs are stated in.the Required-Actions. If a valve'is both a PCIV and a PIV, then the most limiting of the two Required A'ctions must be met. Only those check valves which have been demonstrated to not exceed their allowable leakage rate limit at their last surveillance may be used to meet the PIV requirement. Finally, it must be periodically verified that the affected penetration remains isolated. B.1. B.2. C.1 If the Required Actions and associated Completion Times of Condition A cannot be met in MODES 1, 2 or 3, then the reactor ~ is required to be in MODE 3 and subsequently MODE 4. If the reactor is already in MODE 4 or 5, and the valves are required to be OPERABLE, CORE ALTERATIONS must be immediately suspended. p However, this does not preclude completion of movement of a d component to a safe, conservative position. .The OPERABILITY of valves in MODES 4 and 5 is governed by the requirements for the associated actuation instrumentation. Comoletion Times All Complet' ion Times are based on industry accepted practice and engineering judgement considering the number of available-systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.1.5.1 SR 3.6.1.5.3 REQUIREMENTS The analyses in References 1 and 4 are based on a leak rate which assumes that all primary containment penetrations that are not closed by an automatic primary containment isolation valve which are required to be closed during accident condi-tions are in fact closed. The manual valves are those which do not receive an automatic isolation signal. Surveillance of i these penetrations is required to assure that this assumption is met. Those valves which are readily accessible (i.e., outside the primary containment) are surveilled every 31 days. \ (continued) ABWR B 3.6-17 5/31/89
Primary Containment and Pressure. Isolation Valves l B 3.6.1.5 BASES (continued) SURVEILLANCE Those valves which are more difficult to inspect (i.e., inside REQUIREMENTS the drywell, containment, or steam tunnel) need only be surveil (Continued) led in. MODES 4 and 5 when the primary containment is de-inerted and the' valves have not been checked in the previous 92 days. SR 3.6.1.5.2 The Transversing In-Core Probe (TIP) isolation valves are regular ball valves that are automatically actuated provided . each associated TIP is fully withdrawn (the withdrawal function also occurs automatically). As a backup, thre are also manu-ally initiated shear valves (one per line) that are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that the valves will actuate when required. SR 3.6.1.5.4. SR 3.6.1.5.5 These surveillance requirements demonstrate each power operated or automatic primary containment isolation valve is OPERABLE by verifying the isolation time to be within limits when tested pursuant to SR 3.0.5. SR 3.0.5 requires periodic inservice testing of valves in accordance with Reference 3. Section IWV-3413 of Reference 3 requires that stroke times be specified and periodically measured. The stroke time limits are speci-fied in Ref. 1. These surveillance requirements provide. assurance that the closure time of automatic isolation valves and MSIVs will not exceed that used in References 1 and 4 so that predicted consequences will not be exceeded during the postulated events. SR 3.6.1.5.6 This surveillance requirement provides assurance that the isolation valves will actuate to the proper position upon receipt of the respective containment isolation signals. This requirement complies with Reference 3 to periodically test OPERABILITY of isolation valves. (contir.ued)
)
)
J O ABWR B 3.6-18 5/31/89 l l
P'rimary' Containment and Pressure Isolation Valves l B 3.6.1.5 j
'[L/3 BASES (continued) 4 - SURVEILLANCE SR 3.6.1.5.7
- REQUIREMENTS '
(continued) This surveillance requirement demonstrates that each reactor instrumentation line excess flow check valve is OPERABLE by verifying ~ the valve reduces flow to s I gallon per hour _ on a-simulated line break. This surveillance requirement provides assurance that the instrumentation line excess flow' check valves will perform so that predicted radiological. consequences will not be exceeded during the postulated instrument line break event evaluated in Reference 5. SR 3.6.1.5.8 The TIP manual isolation shear valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The' replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch' which has been certified by having one of that batch success-fully fired. SR 3.6.1.5.9
/m) The analysis in Reference 4 is based on leakage less than the (V specified leak rate. Leakage through each MSIV should be s 20 scf/hr when tested. However, the analysis of Reference 4 assumes that only five of the eight valves meet this requirement. It is assumed that two of the other three MSIVs have leakage of 100 scf/hr and the remaining' valve 500 scf/hr.
The worst combination of leakage and failure of one MSIV to isolate would then result in a combined leakage'of 640 scf/hr. SR 3.6.1.6.10. SR 3.6.1.6.11 Leakage through each reactor coolant PIV must be < 0.5 gpm per nominal inch of valve size, though if the valve is 20 inches or larger, the maximum PIV leakage is 5 gpm. Note that the PlV surveillance _is only required in MODES I and 2; MODE 3 may be used to perform the-required demonstration. Leakage through each hydrostatically tested line which penetrates Primary containment is not to exceed I gpm when tested at [ ] psig. Surveillance of these leakage rates provides assurance that References 5 and 6 calculation assumptions are met. Note also that dual function valves must pass all applicable SRs including the Type C leak rate test (SR 3.6.1.5.12) if appropriate.
- f. (continued) l V ABWR B 3.6-19 5/31/89
Primary Containment and Pressure Isolation Valves B 3.6.1.5 BASES (continued) SURVEILLANCE SR 3.6.1.5.12 REQUIREMENTS (continued) Maintaining the isolation function requires compliance with the Type C leak rate tests of 10 CFR 50, Appendix J. These periodic testing requirements verify that the isolation valves will not leak excessively when closed. Surveillance Frequencies In general, surveillance. frequencies are based on industry accepted practica and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. 10 CFR 50, Appendix A.
- 3. ASME Code Section XI, Subsection IWV.
- 4. ABWR SSAR, Chapter 15.6.
- 5. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O\4 i ABWR B 3.6-20 5/31/89 (
)
i i
Wetwell-to-Drywell VB B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Wetwell-to-Drywell Vacuum Breakers BASES BACKGROUND The function of the Wetwell-to-Drywell Vacuum Breakers is to relieve vacuum in the drywell. There are 8 vacuum breakers located within the wetwell airspace on separate lines penetrat-ing the pedestal wall at equal distances around the pedestal. The vacuum breakers allow air and steam flow from the wetwell airspace to the lower drywell when the drywell is at a negative pressure with respect to the wetwell. Each vacuum breaker is a self-actuating valve similar to a check valve which can be remotely operated for testing purposes. Negative pressure differentials across the drywell walls are caused by rapid depressurization of the drywell. Events which cause this rapid depressurization are cooling cycles, inadvert-ent drywell spray actuation, and steam condensation from sprays or subcooled water reficod of a break following primary system ruptures. Cooling cycles result in minor pressure transients in the drywell which occur slowly and are normally controlled by heating and ventilating equipment. Spray actuation or spill of subcooled water out a break result in more significant pressure transients and becomes important in sizing the external vacuum breakers. Steam condensation following primary ruptures within the drywell results in the most severe pressure transients. Following a primary system rupture, air in the drywell is purged into the wetwell airspace, leaving the drywell full of steam. Subsequent condensation of the steam can be caused by two possible transients, ECCS flow from a line break or manual containment spray actuation following a Loss of Coolant (LOCA). These two cases determine the maximum depressurization rate of the drywell. In addition, the water leg in the ABWR vent system downtomer is controlled by the drywell to suppression thamber differential pressure. If the drywell pressure is less than the suppression chamber pressure then there will be an increase in the vent water leg. This will produce an increase in the water clearing inertia during a postulated LOCA. Consequently, this will result in an increase in the peak drywell pressure and pool swell dynamic loads during a postulated LOCA. The internal vacuum breakers limit the height of the water seg in the vent system during normal operation. (continued) ABWR B 3.6-21 5/31/89 L________-_-__-_--_-___-______ _
h Wetwell-to-Drywell VB B 3.6.1.6 BASES (continued)
- APPLICABLE Analytical methods and assumptions involving the wetwell-SAFETY to-Drywell Vacuum Breakers are presented in Reference 1 ANALYSES as part of the accident response of the containment systems.
Internal, wetwell to drywell vacuum breakers are provided as part of the primary containment to limit the negative pressure differential across the drywell and wetwell walls. The safety analyses assume the internal vacuum breakers to be closed initially and to be fully open at a differential pressure of 0.5 psid. Additionally, one of the eight internal vacuum breakers is assumed to fail in a closed position (Ref. 1). The results of the analyses show that the design pressure is not exceeded even with the worst possible accident scenario. The vacuum breaker opening differential pressure setpoint and the requirement for seven vacuum breakers to be operational are a result of the requirement on the vacuum breakers to limit the vent system water leg height. The total cross sectional area of the main vent system between the drywell and wetwell needed to fulfill this requirement has been established as a minimum of [ ] times the total break area (Ref.1). In turn, the vacuum relief capacity between the drywell and wetwell should be [ ] of the total main vent cross sectional area with the valves set to operate (begin to open) at 0.2 psid pressure differential and be fully open at 0.5 psid. Wetwell-to-Drywell Vacuum Breakers satisfy the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. LC0 Seven vacuum breakers must be OPERABLE for opening, and all eight Wetwell-to-Drywell Vacuum Breakers are required to be closed (except during testing or when the vacuum breakers are performing their intended design function). The requirement provides assurance that the drywell to suppression chamber negative pressure differential remains below the design value. The closed requirement assures that there is no excessive bypass leakage should a LOCA occur. APPLICABILITY The Suppression Chamber-to-Drywell Vacuum Breakers must be OPERABLE and closed during MODES 1, 2 and 3 consistent with the , requirement for when primary containment must be OPERABLE. In 1 MODES 4 and 5 there is insufficient energy in the reactor to heat up and pressurize the drywell to the point that a significant drywell depressurization event could occur. l l (continued) 1 ABWR B 3.6-22 5/31/89
K Wetwell-to-Drywell VB B 3.6.~1.6~ y ,
%- BASES (continued) 7):
4" -
.ACTIONSc _M An open vacuum breaker results in leakage between the drywell and suppression chamber and the potential for suppression chamber overpressure due to bypass leakage if a'LOCA were to occur. Therefore,_the open vacuum breaker must~be closed. A-short time is allowed to close the vacuum breaker due to.the low probability of an: event which would pressurize the primary
-containment. .lf vacuum breaker position indication is-not-reliable, an alternate method of verifying.the vacuum breakers are closed is to verify that a differential pressure of 0.5 psi between the torus and drywell is maintained for one hour without makeup.
M-With one of the required vacuum breakers inoperable for reasons other than Condition A (e.g., the vacuum breaker is not open, and may be stuck closed or not within its opening setpoint limit such that it would not function as designed were an event to occur that depressurized the drywell), the remaining seven OPERABLE vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced because a single failure in one of the remaining vacuum breakers could result in an excessive wetwell-to-drywell
- (G/ differential pressure during a design basis event. Therefore, with one of the 'eight required vacuum breakers inoperable, a limited time is allowed (consistent with other Primary Contain-iMnt functions) to restore the inoperable vacuum breaker to DrERABLE status so that plant conditions are consistent with those assumed for the design basis analysis.
C.1. C.? With one Suppression Chamber-to-Drywell Vacuum Breaker open and one of the required vacuum breakers otherwise inoperable, the Required Actions of Conditions A and B must be satisfied. D.1. D.2 If the Required Actions and associated Completion Times of Condition A, B or C are not met, the reactor is required to be in MODE 3 and subsequently in MODE 4. (continued) l-l'
's ABWR B 3.6-23 5/31/89
Wetwell-to-Drywell VB B 3.6.1.6 BASES (continued) ACTIONS Comoletion Times (continued) All Completion times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS Each vacuum breaker is verified to be closed (except when being tested in accordance with SR 3.6.1.6.2 or when the vacuum breakers are performing their intended design function; to ensure that this potential large bypass leakage path is not present. This can be performed by observing the valve indica-tion or, if vacuum breaker indication is not reliable, an alternate method of verifying the vacuum breakers are closed is to verify that a differential pressure o; 0.5 psi between the torus and drywell is maintained for one hour without makeup. SR 3.6.1.6.2 Each vacuum breaker must be cycled to ensure it opens ade-quately to perform its design function and returns to the fully . closed position. This assures that the safety analysis assump-tions are valid. SR 3.6.1.6.3 Verification of the vacuum breaker opening setpoint is neces-sary to assure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. Surveillance Frequencies ! In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. l (continued) 1 O ABWR B 3.6-24 5/31/89 i L- - - --. _ 1
I Wetwell-to-Drywell VB { B 3.6.1.6 i BASES (continued) o/~~'s REFERENCES 1. ABWR SSAR, Section 6.2. 4
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
I l t/ tO V ABWR B 3.6-25 5/31/89 \- - _ _ . _. _ __ __ __ __ _ - - D
Suppression Pool Avg Temp B 3.6.2.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.1 Suppression Pool Averace Temperature BASES . BACKGROUND The suppression pool is a concentric volume of water contained within the stainless steel liner of the wetwell which is located at the bottom of the containment, separated from the upper and lower drywell by the diaphragm floor and reactor pedestal, respectively. The suppression pool is designed to absorb the decay heat and sensible heat released during a reactor blowdown from safety / relief valve discharges or from a Loss-of-Coolant Accident (LOCA). The pool must also condense steam from the Reactor Core Isolation Cooling (RCIC) system turbine exhaust and provides the main emergency water supply source for the reactor vessel. The amount of energy that the pool can absorb as it condenses steam is dependent upon the-initial pool temperature. The lower the initial pool temper-ature, the more heat it can absorb without heating up exces-sively. Since it is indirectly connected with the upper and lower drywell through the connecting vents and vacuum breakers its temperature will effect both containment pressure and average air temperature. Using conservative inputs and methods, the maximum calculated containment pressure during and following a design basis accident must remain below the containment design pressure of 45 psig. In addition, the , maximum containment average air temperature must remain below 340*F. The allowable suppression pool temperatures are based on the following:
- 1. 95'F is the initial condition for the analysis determining pool performance and satisfies the post-LOCA long-term peak pool temperature of 207'F. It is also the initial condition for the long term suppression pool heatup analysis for Safety / Relief Valve (S/RV) transients.
l 2. The original basis for the reactor pressure vessel depressurization limit was the Bodega Bay Tests (Ref. 5) which were conductM with a peak pool temperature of 170*F. The fwference 3 analysis indicates that a reactor blowdown will raise the pool water temperature approxi-mately 50*F. So to stay below the 170*F limit, the operator is instructed to begin reactor depressurization when the pool temperature reaches 120*F.. (continued) O l ABWR B 3.6-26 5/21/89 l l l \ _ _____-
Suppression Pool Avg Temp B 3.6.2.1 BASES feo.ntinued) BACKGROUND 3. Il0*F and 105*F are derived from engineering judgment (continued) which considers operator reaction time and avoidance of unnecessary scrams and/or depressurizations. The limitations on suppression pool temperature are necessary to meet the requirements of Reference 2. APPLICABLE Containment performance is evaluated for the entire spectrum of SAFETY break sizes for postulated LOCAs. Inputs to the safety ANALYSES analyses. Inputs to the safety analyses include initial
- suppression pool temperature. An initial suppression pool temperature of 95'F is assumed for the analyses of References 1 and 3. Reactor shutdown at a pool temperature of 110*F and initiation of reactor vessel depressurization at a pool temper-ature of 120*F are assumed for the References 1 and 3 analyses.
The safety analyses assure that the 207'F design average pool temperature limit are not eaceeded during a postulated accident. Suppression Pool Average Temperature satisfies the requirements of Selection Criterion 2 and 3 of the NRC Interim Policy Statement on Technical Specification Improverc.ents as documented in Reference 4. LCO A limitation on the suppression pool temperature is required to assure that the containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak containment pressures and temperatures do not exceed design values during a postulated LOCA or any transient resulting in heatup of the suppression pool. The LC0 requirements are as follows: A. Temperature s 95'F when [ reactor power is > 1% of RTP] and when not testing equipment that discharges steam to the l suppression pool which ensures that licensing bases initial conditions are met. l B. Temperature $ 105*F when [ reactor power is > E of RTP] and when testing equipment that discharget steam to the suppression pool which ensures the plant has testing flexibility. The time period with temperature above 95*F is short enough to not cause a significant increase in plant risk. (continued) ABWR B 3.6-27 5/31/89
Suppression Pool Avg Temp B 3.6.2.1 4 BASES (continued) LC0 C. Temperature s 110*F when [ reactor power is s 1% of RTP] which ensures that the plant will be shutdown upon exceed-ing 110*F. The pool is designed to absorb decay heat and sensible heat with a pool cooling system operating, but could be heated beyond design limits by the steam generated if the reactor is not shutdown. Note that [1% of RATED THERMAL POWER (RTP)] has been chosen as a convenient measure of when the reactor is producing heat input approximately equal to normal system heat losses. [A parameter valve representative of 1% of RTP will be determined and substituted at a later date.) APPLICABILITY The suppression pool temperature limitations are applicable during MODES 1, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to significantly heat up the suppression pool even if the initial pool limit is not maintained. ACTIONS A,1. A.2 With the suppression pool temperature above the specified limit when not performing testing which adds heat to the suppression pool, the containment conditions exceed the conditions assumed for the Reference 1 analyses. However, containment cooling capability still exists and the containment pressure suppression function will occur at temperatures well above that assumed for safety' analyses. Therefore, continued operation is allowed for only a limited time to allow the suppression pool temperature to be restored below the limit. Additionally, when pool temperature is above 95'F, increased monitoring of sup-pression pool temperature is required to ensure it remains at or below 110'F. 161 If the Required Actions and associated Completion Times of Condition A are not met, then reactor power must be reduced [to below approximately 1% of RTP). The time specified is consis-tent with the time to be in MODE 3 for other containment actions. At this power level, heat input to the suppression pool is approximately equal to normal system heat losses. Also, the energy required to be absorbed by the suppression pool during a LOCA is greatly reduced. (continued) ABWR B 3.6-28 5/31/89
Suppression Pool Avg Temp B 3.6.2.1 BASES (continued)
?}^#-
. ACTIONS. L.l .
(continued) . _ Suppression pool temperature up to 105'F'is allowed during testing which adds heat to the-suppression pool. -If tempera-ture exceeds'105'F, the testing must be immediately suspended to preserve the pool's heat absorption capability. With testing suspended, Condition A is entered and the Required Actions and associated Completion Times are applicable. D.I. D.2 Suppression pool temperature at or above 110*F requires that the reactor be shutdown immediately. This is accomplished by placing the Reactor Mode Switch in the Shutdown position. Additionally when pool temperature is above 110*F increased monitoring of pool temperature-is required to ensure'it remains below 120'F. E.1. E.2 c Continued heat addition to the suppression pool with pool temperature above 120*F'could result in exceeding the design basis maximum allowable values for containment temperature or I pressure. Furthermore, if a blowdown were to occur when [' temperature was above 120*F the maximum allowable bulk and A . local temperatures could be exceeded very. quickly. Therefore, the reactor must be depressurized and placed in MODE 4 to limit this potential. Comoletion Times All Completion Times are based on industry accepted practice and engineering. judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE:E SR 3.6.2.1.1 REQUIREMENTS The Suppression Pool Average Temperature is regularly monitored (24 hours) to ensure that the required limits are satisfied. When heat is being added to the suppression pool by testing, it l is necessary to monitor suppression pool temperature every 5 l minutes since temperature could be increasing rapidly. This will ensure that allowable pool temperatures are not exceeded. (continued) O 1 ABWR B 3.6-29 5/31/89 i i
Suppression Pool Avg Temp B 3.6.2.1 BASES (continued) SURVEILLANCE Surveillance Frequencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the ; system / component status. I REFERENCES 1. ABWR SSAR, Section 6.2 .
- 2. 10 CFR 50, Appendix A, GDC 13.
- 3. ABWR SSAR, Chapter 15.
- 4. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
- 5. " Preliminary Hazards Summary Report, Bodega Bay Unit No.
1," Pacific Gas and Electric Co., December 28, 1962. O 4 l O ABWR B 3.6-30 5/31/89
i' Suppression Pool Water Level B 3.6.2.2 g B'3.6 [ CONTAINMENT SYSTEMS l~ -J B 3.6.2.2 Suppression Pool Water Myfl BASES BACKGROUND The suppression pool is designed to absorb the decay heat and sensible heat released during a reactor blowdown from safety / relief valve'(S/RV) discharges or from a' Loss of Coolant. Accident (LOCA). The pool must' also condense steam from the Reactor Core Isolation Cooling (RCIC) turbine exhaust and provides :the main emergency water supply source for the reactog . vessel. Thesuppressionpoolvolume-rangesbetween126,j27ft at the low water level alarm of 22.97' and [ ] ft at-the high water level alarm of 23.29'. If. the suppression pool water level is too low, it could j jeopardize steam condensation from the S/RV quenchers, main ; vents, or RCIC turbine exhaust. It could also be an inadequate emergency' makeup source. The lower volume' would also absorb less steam energy before heating up excessively. Therefore, a minimum pool water level is specified. -; If the suppression pool water level is too high, it'could result in excessive clearing loads from S/RV discharges.and excessive pool' swell loads during a LOCA.
- APPLICABLE Initial Suppression Pool Water Level effects containment pool
-SAFETY temperature response calculations, calculated drywell pressure j '
ANALYSES during vent clearing for a LOCA, calculated pool swell loads for a LOCA, and calculated loads due to S/RV discharger. i Suppression Pool Water Level satisfies the requirements of ; Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 1. LC0 A limit that Suppression Pool Water Level be from 22.97* to 23.29' is required to assure that the containment conditions i assumed for the safety analysis are met. The high or low water 1 level limits was used in the safety analysis, depending upon _ which is conservative for a particular calculation. 1 (continued) l o l ABWR B 3.6-31 5/31/89 1 1 _ ____-__--_A
f-Suppression Pool Water Level B 3.6.2.2 BASES (continued) APPLICABILITY The Suppression Pool Water Level limitations are applicable O during MODES I, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to place significant loads on the containment even if pool water level is not kept within limits. However, the suppression pool still provides the primary source of emergency makeup water during these conditions. Requirements for Suppression Pool Water L Level during MODES 4 and 5 are specified in LC0 3.5.2. ACTIONS .A_d When suppression pool water level is outside the limits, the conditions assumed for the safety analysis are not met. If water level is below the minimum level, the pressure suppres-sion function still exists as long as main vents are covered, RCIC turbine exhaust is covered, and S/RV quenchers are : covered. If water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis or as long as the drywell and containment sprays are OPERABLE. Therefore, continued operation for a limited time is allowed to restore Suppression Pool Water Level to within limits. B.1. B.2 If water level cannot be restored to within limits then the reactor is required to be in MODE 3 and subsequently in MODE 4. Requirements for suppression pool water level during operation in MODE 4 are specified in LC0 3.5.2. Completion Times ! All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS The licensing bases analysis assumes that initial Suppression Pool Water level is between the specified limits. This SR assures that the assumption is valid. (continued) ABWR B 3.6-32 5/31/89
1 Suppression Pool.' Water Level B-3 6.2.2 BASES (continuedi 73
'" Q
' SURVEILLANCE Surveillance Frequencies REQUIREMENTS (continued) In general, surveillance frequencies are based..on-industry
. accepted practice and engineering judgement considering'the unit conditions required to' perform the test, the ease of performing .the test and a likelihood of a change in the system / component status.
REFERENCES 1. NED0 31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987. V I LO 1 ABWR B 3.6-33 5/31/89
RHR Suppression Pool Cooling B 3.6.2.3 B 3.6 CONTAINMEN1
-l B 3.6.2.3 Residual Heat Removal SuDDression Pool Coolina BASES i
l BACKGROUND Following an accident, the Residual Heat Removal (RHR) system removes heat from the containment. The suppression pool is designed to absorb the sudden input of heat from the primary ' system and long term, the pool continues to absorb residual heat generated by fuel in the reactor core. Some means must be. provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by three redundant RHR Subsystems. Initially, this is accomplished by the Low Pressure Flooding mode as water is drawn from the suppression pool and pumped through the RHR heat exchangers to the vessel. Eventually this water spills out the postulated break and drains back to the suppression pool. Because this recircula- , tion path includes a pass through the RHR heat exhanger, enough containment cooling is accomplished to adequately control the short term containment response. However, long term pool temperature (and overall containment responseT is controlled by RHR in the Suppression Pool Cooling mode. Each of the three RHR subsystems contains one pump and one heat 4 exchanger which are independently controlled by plant operators. The three RHR subsystems perform the suppression pool cooling function by circulating water from the suppression pool, through the RHR heat exchangers, and returning it to the suppression pool. A suppression pool cooling subsystem is OPERABLE when the pump, heat exchanger and associated piping and valves are OPERABLE. Reactor Building Cooling Water (RCW), circulating through the shell side of the heat exchangers, exchanges heat with the suppression pool water, discharges this heat to the associated Reactor Building Service Water (RSW) loop, and ultimately to the external heat sink. Each subsystem, including the associated RCW and RSW loops, is separate and independent, including separate power supply divisions. The combined heat removal capability of two subsystems is sufficient to meet the long term post loss of Coolant Accident (LOCA) pool cooling requirement. Transient events such as turbine trip or a stuck opel safety / relief valve (S/RV) also result in discharge of steam from the reactor to the suppres- ; sion pool, increasing suppression pool water temperature. S/RV leakage increases suppression pool temperature more slowly. The suppression pool cooling mode of the RHR system is used to lower the suppression pool water bulk temperature following I such events. (continued) l ABWR B 3.6-34 5/31/89 l l l J
RHR Suppression' Pool Cooling B 3.6.2.3
.3 BASES (continued)
.l )
%- APPLICABLE.~ . Reference 1 contains 'the results of analyses used to predict-SAFETY- containment pressure and temperature following -large and small ANALYSES' break LOCAs. The intent of the analysis is to demonstrate that the heat removal capacity of the RHR _ heat exchangers is ade-quate to maintain the containment conditions,within its design limits. The time history for' suppression pool temperature is-followed to' demonstrate that the maximum suppression' pool temperature remains less than the design limit following the accident event.
- RHR Suppression Pool Cooling satisfies the requirements of Selection' Criterion 3 of the NRC Interim Policy Statement on-Technical Specification Improvements as documented in Reference 2.
LC0 Three subsystems of the RHR Suppression Pool Cooling system must be OPERABLE. Each subsystem consists of one pump,'one heat exchanger, and the necessary piping and valves to circu-late water from and back to the suppression pool to cool the suppression pool. The LC0 assures that' redundant RHR Suppres-sion Pool Cooling subsystems are available to perform the heat removal function.
.p %f APPLICABILITY .The RHR Suppression Pool Cooling function is only required to be OPERABLE in MODES 1,.2 and 3 consistent with the requirement.
for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy.in the reactor to overheat the suppression pool even if suppression pool cooling was not available. (continued) i l ABWR B 3.6-35 5/31/89
i RHR Suppression Pool Cooling B 3.6.2.3 BASES (continted) ACTIONS A.1 When one RHR Suppression Pool Cooling subsystems are inoperable, the remaining OPERABLE subsystems are adequate to perform the containment cooling function. However, the overall reliability is reduced because a single failure in the remaining subsystem could result in significantly reduced containment heat removal capability. Continued operation is permitted for a limited time due to the low probability of an unplanned event which would require suppression pool cooling be OPERABLE. As noted, the provisions of LCO 3.0.4 are not applicable. B.I. B.2 With two RHR Suppression Pool Cooling subsystems inoperable, reduced (and possibly inadequate) containment heat removal capability remains to handle either accident or transient events. Continued operation is justified because of the low probability of an event which would require suppression pool cooling and because some level of containment heat removal capability still exists. However, continued operation of the plant is permitted for only a limited period of time. The initial inoperable subsystem is also required to be restored to OPERABLE status consistent with Required Action A.I. C.1. C.2. C3 With all three RHR Suppression Pool Cooling subsystems inoperable, no containment heat removal capability remains to handle either accident or transient events. Continued operation is justified because of the low probability of an event which would require suppression pool cooling. However, continued operation of the plant is permitted for only a limited period of time. The initial and subsequent inoperable subsystems are also required to be restored to OPERABLE status consistent with Required Actions A.1, B.1 and B.2. D.1. 0.2 If the Required Actions and associated Completion Times of Condition A, B or C are not met, then the reactor is required to be in MODE 3 and subsequently MODE 4 where the RHR Suppres-sion Pool Cooling system is not required. If the reactor cannot attain MODE 4 because of inoperable RHR subsystems, the reactor coolant temperature should be maintained as low as practicable using an alternate decay heat removal method. (continued ABWR B 3.6-36 5/31/89
RHR Suppression Pool Cooling B 3.6.2.3 x BASES (continued)
;( )
5
'V ACTIONS Completion Times (continued).
All Completion Times are based on industry accepted practice and engineering judgement considering the number of available l systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Successful initiation of suppression pool cooling requires the operation of certain valves. It is verified that each manual, automatic or power operated valve not locked, sealed, or otherwise secured in position, is in its correct position or can be aligned to the position required to operate suppression pool cooling. SR 3.6.2.3.2 The components of RHR Suppression Pool Cooling subsystems A, B and C are designed to be tested during normal plant operation. Each subsystem is tested to verify that the required flow of 4200 gpm (Ref. 1) through the associated heat exchangers is p achieved. Each subsystem is tested to ensure that all will provide the requisite flow. Q Surveillance Frecuegjai In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform tne test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
ABWR B 3.6-37 5/31/89 q
\
l 1 1
) RHR Wetwell Spray j [ ! B 3.6.2.4 i I B 3.6 CONTAINMENT SYSTEMS B 3.6.2.4 Residual Heat Removal Wetwell Soray . J j ILASES , 1
)
BACKGROUND Following an accident, the Residual Hect Removal (RHR) Wetwell i spray system removes heat from the wetwell airspace. The ! suppression pool is designed to absorb the sudden input of heat I from the primary system. The heat th t results in steam in the wetwell increases primary containment pressure. Some means must be provided to remove heat from the wetwell so that the pressure and temperature inside the primary containment remain j within design limits. This function is provided by two j redunh nt RHR Wetwell Spray subsystems. l 5 Each of the RHR subsystems contains a pump and a heat exchanger which are independently controlled by plant operators. , Subsystros B and C only perform the wetwell spray function by 1 circulating water from the suppression pool, through the RHR heat exchangers, and returning it to the common wetwell spray sparger. A wetwell spray subsystem is OPERABLE when the pump, heat exchanger and associated piping and valves are OPERABLE. The wetwell spray is sufficient to condense the steam from small bypass leaks from the drywell to the wetwell airspace. Reactor Building Cooling Water (RCW), circulating through the shell side of the heat exchangers, exchanges hett with the suppression pool water and discherges this heat to the external heat sink, via the associated Reactor Building Service Water i (RSW) loop. APPLICABLE Refererce 1 contains the results of analyses used to predict SAFETY containment pressure and temperature following large and small ANALYSES break Loss of Coolant Accidents (LOCAs). The intent of the 1 analysis is to demonstrate that the heat removal. capacity of the RHR heat exchangers is adequate to maintain the containment j structure within its design limits. The time history for containment pressure is followed to demonstrate that the containment pressure remains less than the design limit following the accident event. The RHR Wetwell Spray system satisfies the requirements of Selection Criterion 3 of ti,e NRC Interim Policy Statement on Technical Specification Improvements as documented in , Reference 2. 1 (continued) O ABWR B 3.6-38 5/31/89
RHR Wetwell Spray B 3.6.2.4 q- . BASES (continued) LCO. Two subsystems of the RHR Wetwell Spray system must be. OPERABLE. Each subsystem consists of one pump,:one heat exchanger, and the necessary piping and valves to circulate water from the suppression pool to the wetwell spray sparger to reduce wetwell airspace pressure. The LC0 assures that redundant RHR Wetwell Spray subsystems are available to perform the heat removal function. APPLICABILITY The RHR Wetwell Spray function is only required to be'available in MODES 1,.2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient. energy in the reactor to overheat.the wetwell even if suppression pool cooling and wetwell spray were not available. ACTIONS A,1 When one RHR Wetwell Spray subsystem is inoperable the remaining OPERABLE subsystem is adequate to perform the wetwell spray function. However, the overall reliability is. reduced
. r' because a single failure in the remaining subsystem could $ result in no wetwell spray capabil1ty. As noted, the provisions of LC0 3.0.4 are not applicable. Continued opera-tion is permitted for a limited time due to the low probability of an event which would require wetwell sprays be OPERABLE.
B.1. B.2 - With two RHR Wetwell Spray subsystems inoperable, no wetwell spray capability remains to handle either accident or trensient events. Continued operation is justified because of the low probability of an event which would require wetwell spray. However, continued operation of the plant is permitted for only a limited period of time. The initial inoperable subsystem is also required to be restored to OPERABLE status consistent with Required Action A.1. C.1. C.2 If the Required Actions and associated Completion Times of Condition A or B are not met, the reactor is required to be in MODE 3 and subsequently MODE 4 where the RHR Wetwell Spray < system is not required. If the reactor cannot attain MODE 4 (continued) bg ABWR B 3.6-39 5/31/89 i
- . - - - - . . - - . - - . . . __..___.___._____..__._.___2 .____._2____.__m_m__m___
RHR Wetwell Spray B 3.6.2.4 BASES (continued) ACTIONS because of the inoperable RHR subsystems, the reactor coolant (continued) temperature should be maintained as low as practicable using an alternate decay heat removal method. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.2.4.1 REQUIREMENTS Successful initiation of wetwell spray requires the operation of certain valves. Each manual, automatic or power operated valve not locked, sealed, or otherwise secured in its correct position is checked to verify that it can be aligned to the position required to operate wetwell spray. SR 3.6.2.4.2 The components of RHR Suppression Pool Cooling System are designed to be tested during normal plant operation. Each subsystem is tested to verify that the required flow of 4200 gpm (Ref. 1) through the associated heat exchanger (and common spray sparger) is achieved. The B and C subsystems are tested in the combined Suppression Pool Cooling /Wetwell Spray Mode to verify proper functioning with flow to the common wetwell spray sparger. Each subsystem is tested to ensure that both will provide the requisite flow. Surveillance Frequencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. N7ED-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.6-40 5/31/89
Hydrogen Recombiner System B 3.6.3.1 ?'N B 3.6 CONTAINMENT SYSTEMS \
) Hydrocen Recombiner System B 3.6.3.1 BASES (Later)
O ! O ABWR B 3.6-41 5/31/89
i i Primary Containment Oxygen Concentration B 3.6.3.2 l B 3.6 CONTAINMENT SYSTEMS 1 B 3.6.3.2 Primary Containment OKYQen Concentration 0- j BASES . e I BACKGROUND All nuclear reactors must be designed to withstand events which generate hydrogen either due to the zirconium metal-water l 4 reaction in the core or due to radiolysis. The primary method ) to control hydrogen is to inert the-primary containment. With j the primary containment inert (oxygen concentration < 4%), a combustible mixture cannot be present in the containment for any hydrogen concentration. The capability to inert.the primary containment and maintain oxygen below 4% by volume , works together with the Hydrogen Recombiner System (1.00 i 3.6.3.1) to mitigate events which produce hydrogen. For example, an event which rapidly generates hydrogen from zir- {: conium metal-water reaction will result in an increase in the concentration of hydrogen in the containment but oxygen concen-tration will remain below 4% and no combustion can occur. Long term generation of both hydrogen and oxygen from radiolytic decomposition of water may eventually result in a combustible nixture in containment, except that the hydrogen recombiners remove hydrogen and oxygen gases faster than they can be produced from radiolysis and again no combustion can occur. O APPLICABLE The Reference 1 calculations assume that the primary contain-SAFETY ment is inerted when the design basis LOCA occurs. Thus, the ANALYSES hydrogen assumed to be released to the containment as a result of metal-water reaction in the reactor core will not produce combustible gas mixtures in the containment. Oxygen which is subsequently generated by radiolytic decomposition of water is recombined by the Hydrogen Recombiners (LC0 3.6.3.1) more rapidly than it is produced. The Oxygen Concentration that ensures the containment is inert-satisfies the requirements of Selection Criteria 2 of the NRC Interim Policy Statement on Technical Specification Improve-ments as documented in Reference 2. i LC0 The primary containment oxygen concentration is maintained below 4% by volume to ensure that an event which produces any amount of hydrogen does not result in a combustible mixture inside primary containment. (continued) O. ABWR B 3.6-42 5/31/89 i
-s - - 7
~ -
. r i,
, 3 y. a L
. Primary Containment Oxygen. Concentration-
' ~
B 3.6.3,2 c '
. BASES (continued)-
APPLICABILITY The primary containment must be inert only in MODE l since- 1 during these conditions the' highest probability for an event which could produce hydrogen exists. j Inerting primary containment is an operational problem because
~it constrains containment access. _Therefore, the primary-containment:is-inerted as late:as possible in the plant startup 1 '
and is:deinerted as. soon as possible in the plant 1 shutdown. As
'.'ong as reactor power is below 15% the potential for' an event .
which generates significant hydrogen is' low and the containment need not'be inert. Furthermore, the' probability of an event which generates hydrogen occurring within the first 72 hours cf a startup or.within theLlast 72 hours before a shutdown is low ! enough that'these " windows" with the primary containment not i inerted are also justified.
-ACTIONS. A'.1. B.1
.lf oxygen concentration exceeds 4'% by.vo'lume at any time while' a operating in MODE 1 with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must 73 be restored to below 4%. A short time (24 hours) is allowed ~
j') _ <
.with oxygen concentration above 4% due to other hydrogen mitigating systems (e.g. hydrogen recombiners) and due to the low probability and long duration of an event which would generate significant amounts of hydrogen. If it cannot be restored within-the required Completion Time, reactor power is required to be reduced to s 15% of RTP where the potential for the generation of hydrogen in the primary containment is reduced.
Completion Times All Completion Times are based on industry accepted practice - and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. .l SURVEILLANCE SR 3.6.3.2d l REQUIREMENTS-The primary containment must be determined to be inert by verifying oxygen concentration is below 4% by volume. , l
-l p
(continued) .; Q i i ABWR B 3.6-43 5/31/89
-__ __ _ _ O
1 2 Primary Containment Oxygen Concentration l B 3.6.3.2 i BASES (continued) SURVEILLANCE Surveillance Frequencies REQUIREMENTS { (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. ABWR SSAR, Section 6.2.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
I O I O. ABWR B 3.6-44 5/31/89 _a
Secondary Containment < B 3.6.4.1
;m B 3.6 CONTAINMENT SYSTEMS
-l p Secondary Containment
~
\/ - B 3.6.4.1 BASES BACKGROUND The plant- design and analyses postulate that a non-isolatable release of fission' products can occur only inside the primary containment. Since the primary containment is a high pressure leak tight barrier, the bulk of these fission products are-expected to remain entrapped within the primary containment and-should pose no apt.reciable risk to the general. population as required by Reference 1.
' As an extra margin of safety to accommodate small quantities:of fission products which are postulated to escape from primary.
containment, a secondary containment has been provided. The secondary containment acts as an additional barrier to the release of fission products through and around primary containment penetrations or.through components containing primary system fluid which may be located outside primary containment. Additionally, the secondary containment'is the required. fission product barrier for plant operations that occur outside primary containment-(e.g. handling irradiated-fuel) or occur when the primary containment need not be OPERABLE. (n) U The secondary containment is a structure wi'.h a known leakage rate which completely encloses the primary containment and those components which may be postulated to contain primary system fluid. This structure forms a control volume which serves to. dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g. due to pump / motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain secondary containment pressure less than the external pressure. Requirements for these systems are separately specified in LC0 3.6.4.2, Secondary Containment Isolation Valves, and LCO 3.6.4.3, Standby Gas Treatment System (SGTS). (continued) ABWR B 3.6 45 5/31/89
Secondary Containment I B 3.6.4.1 BASES (continued) BACKGROUND It has been established (Ref. 3) that the negative pressure (continued) differential to be uniformly maintained in the secondary containment should be no less than 0.25 inches water when compared to adjacent regions (the environment) under all wind conditions up to the wind speed at which diffusion becomes great enough to assure site boundary exposures less than those of the LOCA even if exfiltration were to occur. In this manner the in-leakage of fresh air negates the tendency of the fission products to exfiltrate through the non-leak tight secondary containment barrier. APPLICABLE There are two principal events described in Reference 2 which SAFETY take credit for Secondary Containment OPERABILITY. These are: ANALYSES o Loss of Coolant Accident Inside Containment o Fuel Handling Accident The secondary containment performs no active function in . response to either of these limiting events, however, its leak tightness function is required to limit the offsite radiation dose as required by Reference 1. Maintaining the Secondary Containment OPERABLE assures that fission products entrapped within the secondary containment structure, will be treated prior to discharge to the environment. Secondary containment satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 4. (continued) O! ABWR B 3.6-46 5/31/89 1
m. Secondary Containment B 3.6.4.1 A BASES fcontinued) LCOs- Secondary Containment provides a controlled volume into which fission products which escape primary containment or reactor coolant pressure boundary components located in secondary containment can he diluted and processed prior to release to the environment. For the Secondary Containment to be considered CPERABLE, it must have adequate leak tightness to ensure that the required vacuum can 'oe established and maintained. This OPERABILITY is defined by the Surveillance Requirements for this LCO. APPLICABILITY Secondary Containment is required for all MODES which could involve significant releases of fission products into the secondary' containment. This includes MODES 1, 2 and 3 at which time the reactor primary coolant may be above 200*F and significant releases of radioactive steam and gas from postulated pipe' ruptures or other events is possible. Fission products could escape primary containment and/or escape from those components outside primary containment which contain primary system fluid. Secondary Containment is also required for other situations under which significant radioactive' releases can be postulated such as handling of irradiated fuel in the Secondary Containment, during CORE ALTERATIONS, and during operations with a potential for draining the reactor
.( f'T) vessel.
~ ACTIONS Al The secondary containment is the final barrier to the release of radioactive products from the primary system to the environment. If the Secondary Containment is inoperable in MODE 1, 2, or 3, radioactive releases could be larger than calculated for design basis events. A short time is allowed to restore Secondary Containment to OPERABLE status due to the low probability of an event occurring in this time period that would require the Secondary Containment function.
(continued) ABWR B 3.6-47 5/31/89 ,
Secondary Containment B 3.6.4.1 BASES (continued) ACTIONS B.I. B.2 (continued) If the Required Action and associated Completion Time of Condition A are not met (in MODE 1, 2, or 3), then the plant conditions are not consistent with those assumed for licensing analysis, or plant capability is threatened if a radioactive release were to occur. Therefore, the reactor is required to be in MODE 3 and subsequently in MODE 4 to minimize the potential consequences of potential events that could occur. C.1. C.2. C.3 Handiing irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with the potential to drain t'ne reactor vessel can be postulated to cause fission product release to the secondary containment. During refueling operations the secondary containment is the only barrier to release of fission products to the environment. Handling irradiated fuel in the secondary containment and CORE ALTERA-TIONS must be immediately suspended if the secondary contain-ment is inoperable. Suspension of these activities shall not preclude the completion of movement of components to a safe, conservative position. Operations that have the potential for-draining the reactor vessel must be suspended as soon as practicable to minimize the probability of a vessel draindown and subsequent potential for fission product release. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS Assurance of of Secondary Containment OPERABILITY is provided by verifying that secondary containment vacuum is ;t 0.25 inches of vacuum water gauge. The secondary containment boundary must be intact or the negative pressure could not be maintained. (continued) O ABWR B 3.6-48 5/31/89
m= . t m- s ,
- " ?
4 } l , m +-
-) J Secondary. Containment; i
~ .B-3.6.4.1l j 3 1 L ;r: . . . .
~1 i [HASES (continued) l
@I ,e . .
i SJ SURVEILLANCE SR 3.6.4.1.2. SR 3.6.4.1.3 REQUIREMENTS 1 (continued). Opening.'of secondary containment' hatches, panels, process
~
-openings'and access-ways-.could permit'the infiltration of outside-air of.such a magnitude.as to prevent maintaining the-4 desired negati_ve~ pressure. ' Exterior environment 1' disturbances .'3 a (e.g.. wind gusts): may then permit portions 'of;the secondary-containment.to experience a positive: pressure' relative _to the.
4 environment ~.' In _this case. fission product exfiltration could
* . be postulated thus negating the secondary containment safety-function. . Permanently closingLor monitoring frequently all such openings _provides adequate assurance that exfiltration from the secondary containment will not' occur.
SR 3.6.4.1'.4. SR'3.6.4.1.5. The SGTS exhausts the secondary containment free volume through appropriate treatment equipment to the environment; To assure!
-that all untreated fission. products'are treated, these tests demonstrate that one SGTS will rapidly establish and maintain a; pressure in the secondary containment less than the' lowest postulated pressure external, to the secondary. containment
- boundary. This cannot be accomplished' rapidly, if the secondary containment boundary is not intact.. Therefore, these~' tests.are
? used to ensure secondary containment boundary integrity.. Since d Lthese.SRs are Secondary Containment tests, they need not be performed with each SGTS subsystems. However,.the SGTS subsystems are tested on a STAGGERED TEST BASIS to ensure..that-in addition to the requirements of LCO 3.6.4.3, either SGTS will perform this test.
'! Surveillance Frequencies Except as note'd, surveillance frequencies are based on industry accepted practice and engineering judgement considering the .
unit conditions required to perform the test, the ease of performing the test and likelihood of a change in the system / component status. (continued) l-O ABWR B 3.6-49 5/31/89
Secondcry Containment-B 3.6.4.1 BASES (continued) REFERENCES 1. 10 CFR 100.
- 2. ABWR SSAR, Section 6 and 15.
- 3. Standard Review Plan 6.2.3.
- 4. NEDO-31466, " Technical Specification Screening Criteria Application and Risk Assessment," Novembet' 1987.
l O O ABWR B 3.6-50 5/31/89
a N J Secondary Containment. Isolation Valves B 3.6.4.2 v A. B 3.6 CONTAINMENT. SYSTEMS ] B 3.6.4.2 Secondary Containment Isolation Valves - BASES
'l BACKGROUND. The function of the Secondary Containment Isolation Valves, O in-combination with other accident mitigation systems, is to f limit fission product release _ during and following a postulated 1 design basis accident to values less than leakage rates which l would result in offsite doses greater than those set forth in ,
Reference 1. .This includes ventilation system automatic ! isolation dampers as well as other valves designated in the Surveillance Requirements. Secondary Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of-radioactive material to the-environment will be consistent with the assumptions used in Reference 2. APPLICABLE The analytical. methods and assumptions used in evaluating the SAFETY radiological consequences of design basis accidents are ANALYSES' presented in Reference 2. The maximum isolation times for Secondary Containment Isolation Valves are the times used in-the accident analyses. The Secondary Containment Isolation valves satisfy the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 4. LC0 The Secondary Containment Isolation Valves shall be OPERABLE with isolation times less than.or equal to the times in Reference 2. The analytical valve closure times were chosen to assure that 10 CFR 100 exposure limits are not exceeded for postulated events considered in Reference 2. Closure times are selected to minimize the release of containment atmosphere to the environs, thereby mitigating the offsite radiological consequences. (continued) O ABWR B 3.6-51 5/31/89
Secondary Containment Isolation Valves B 3.6.4.2 BASES (continued) APPLICABILITY The Secondary Containment Isolation Valve function is required for all MODES and other specified conditions which could involve significant releases of fission products into the secondary containment. This includes MODES 1, 2 and 3 at which time the reactor coolant may be above 200*F and significant releases of radioactive steam and gas from postulated pipe ruptures or other events are possible. Fission gasses could ! bypass primary containment and/or escape from those components outside primary containment which contain primary system fluid. Secondary Containment Isolation Valves' are also required to be OPERABLE for other situations where radioactive gas release can be postulated including handling of irradiated fuel in the i Secondary Containment, CORE ALTERATIONS, and during operations with a potential for draining the reactor vessel. 1 ACTIONS b_d With one or more isolation valves inoperable, a remaining OPERABLE isolation valve in an affected penetration is adequate to perform the intended isolation function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE isolation valve could result in the loss of ; the isolation function for that penetration. If the affected i penetration has two isolation valves, it must be verified that < the other valve is OPERABLE. This action is satisfied by , examining logs or other information, to determine if the valve is out of service for maintenance or other reasons. It does - not mean to perform the surveillance requirements needed to l demonstrate OPERABILITY of the valve. If the affected j penetration is designed with only one isolation valve, the 4 penetration is such that plant safety is not unduly threatened by a temporary inability to isolate the penetration. l I A.2.1 A.2.2.1. A.2.2.2 j In addition to Required Action A.1, the inoperable valve (s) must be restored to OPERABLE status, or the line must be manually isolated. Acceptable isolation methods are stated in the Required Actions. Finally, the penetration must be periodically verified to be isolated. 1 (continued) l ABWR B 3.6-52 5/31/89 1 J
Secondary Containment Isolation Valves B'3.6.4.2 }"!,,1 BASES (continued) ~ (A ACTIONS' B.1. B.2
- (continued) .
If the Required Actions and associated' Completion-Times of: Condition A are not met in MODE 1, 2 or 3, the isolation valves may not be capable of performing their intended function and the reactor is required to be in MODE 3 and subsequently in MODE 4. C.1. C.2. C.3 If the Required Actions and associated Completion Times of Condition A are not met when handling irradiated. fuel in the Secondary. Containment.or during CORE ALTERATIONS these opera-tions must be immediately suspended. Suspension of these . activities shall not preclude the completion of the movement of a component to a safe, conservative position. If fuel is being handled in the secondary containment while the reactor is in MODE 1, 2, or' 3, then Conditions B and C are concurrently applicable. If the reactor'is shutdown and operations are-being performed which have a potential for draining the reactor vessel, .they must be suspended as soon as practicable to minimize the probability of a vessel draindown and subsequent potential for fission product release. r-g [ompletion Times
\ All' Completion Times are based on industry accepted practice -
and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS Not all secondary containment penetrations are designed to be closed by automatic isolation dampers / valves even though they must be closed during accident conditions. Therefore, these penetrations must be closed at all times during normal operation. This surveillance verifies that those penetrations are closed by blind flanges or rupture disks. (continued) f b) . ABWR B 3.6-53 5/31/89 ___m___ m_._ ._
~ . _ - - - - - - - - - _ - - - - _ - _ _ . - . .- Secondary Containment Isolation Valves B 3.6.4.2 i BASES (continued) SURVEILLANCE SR 3.6.4.2.2 ] REQUIREMENTS (continued) This Surveillance Requirement demonstrates each Secondary Containment Isolation Valve is OPERABLE by verifying isolation J time to be within limits when tested pursuant to SR 3.0.5. SR 3.0.5 requires periodic inservice testing of valves in accord-ante with Reference 3. Section IWV-3413.of Reference 3 requires that stroke times be specified and periodically measured. This Surveillance Requirement provides assur- * ;e that the closure time for isolation valves will not exceed that used in Chapter 15 analyses so that predicted radiological consequences will not be exceeded during the postulated events. SR 3.6.4.2.3 This Surveillance Requirement provides assurance that each Secondary Containment isolation valve is OPERABLE and will assume the proper position following containment isolation signals. Surveillance Frequencies In general, surveillance frequencies are based on industry ! accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the &i system / component status. T 1 REFERENCES 1. 10 CFR 100 1
- 2. ABWR SSAR, Section 6 and 15.
- 3. ASME Section XI, IWV-3000
- 4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.6-54 5/31/89
p q ,
.j Standby Gas: Treatment l q B 3.6.4'.3 '
[ 3: W B'3.6' CONTAINMENT-SYSTEMS. G~' :B 3.6.4.3. Standby Gas Treatment System l' 1 BASES BACKGROUND 'The function of the Standby Gas Treatment System'(SGTS) is to maintain the secondary containment at a negative pressure with 1 respect to .the environment following the design basis. accident '] and process gaseous. releases to limit the thyroid dose and the whole body dose at the plant site boundary within the . guidelines of Reference 1. The SGTS' consists of the following components: A. One, 100 percent. capacity charcoal filter. train, consist-ing of (components listed in order of air flow direction):
- 1. Prefilter
-2. High efficiency. particulate air'(HEPA) filter
- 3. Char oal adsorber-
- 4. HEPA t.lter B. ~Two fully redundant subsystems, each with its own.
ductwork, dampers, and controls, and consisting of: Demister
-.' l.
- 2. Electric heater
~
- 3. Centrifugal fan with inlet flow control vanes.
The sizing of the SGTS equipment and components is. based on the design of the reactor building -*.ructure and the assumed leakage of the secondary containment. The internal pressure of the. secondary containment is maintained at a-negative pressure of 0.25 inches water gauge (per Ref. 5) when the system is'in operation, which represents the internal pressure required to ensure zero exfiltration of air from_the building when a single nonisolated line as large as four inches or all nonseismically qualified lines two inches and smaller are open. The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the air stream to less than 70 percent (Ref. 3). The prefilter removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes elemental iodine and organic iodides, and the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal adsorber. (continued) ABWR B 3.6-55 5/31/89
< Standby Gas Treatment B 3.6.4.3 BASES (continued) BACKGROUND The SGTS automatically starts and operates in response to (continued) actuation signals indicative of conditions that could require operation of the system. Following initiation, both SGTS subsystem fans start. SGTS flows are controlled by modulating inlet vanes installed on the fans and two-position volume control dampers installed in branch ducts to individual regions of the secondary contain-ment. The system operates at full fan capacity until the SGTS boundary region is drawn down to the design negative pressure of 2 0.25 inch vacuum water gauge. At this time, motor-operated dampers automatically throttle the exhaust flow so tnat the volume of air being exhausted is equal to the volume of air infiltrated at the design negative pressure. APPLICABLE The design basis for the dGTS as presented in FSAR Chapter 15, SAFETY Accident Analysis, is to mitigate the consequences of the ANALYSIS Loss of Coolant Accident (LOCA) and the Fuel Handling Acci-dents. For all events analyzed the SGTS is shown to be auto-matically initiated to limit, via filtration and adsorption, the site boundary radioactivity dose level to well within the Reference 1 guidelines. The SGTS satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6. LC0 Both SGTS subsystems and the common, passive filter train must be OPERABLE to assure at least one subsystem is available in the event of a single active failure, to meet its safety function following a LOCA or fuel handling accident. A description of what is required for the SGTS to be considered OPERABLE is provided in the Background section. (continued) Ol , ABWR B 3.6-56 5/31/89 l ____ _ -__- ___ - -____________a
!. 3 Standby Gas Treatment l B 3'.6.4.3 L i i
.% BASES (continued)- . . APPLICABILITY Both SGTS subsystems and the filter train shall be OPERABLE during MODES 1, 2, 3 and when irradiated fuel is being handled, i during CORE' ALTERATIONS and during operations with a potential l L
for draining the reactor vessel to prevent a potential source l of radioactivity from escaping from the secondarj containment, resulting in exceeding the offsite dose limits.. This applic-ability is consistent with the requirement for when the second- , ary containment must'be OPERABLE. 1 ACTIONS Ad With one SGTS subsystem inoperable, the remaining OPERABLE subsystem is adequate to perform the required radioactivity release control function. However, the overall plant reliabil-ity is reduced because a single failure in the remaining subsystem could result in a total loss of the SGTS function. Therefore, the inoperable SGTS subsystem must be restored to OPERABLE status and continued operation with one subsystem inoperable is only allowed for a short time. B.1. B.2. C.1 A If both SGTS subsystems, or the common filter train, are inoperable in MODE 1, 2 or 3 the SGTS will not be capable of () performing its intended function. The common filter train or one inoperable subsystem (which includes the common filter train) must be restored to OPERABLE status within 4 hours. Continued operation is justified because of the low probability of an event which would require SGTS operation. The initial inoperable subsystem must also be restored within 7 days from its initial discovery consistent with Required Action A.I. D.1. D.2 If the Required Actions and Associated Completion Times of Condition A or B are not met in MODE 1, 2 or 3, the reactor is required to be in MODE 3 and subsequently in MODE 4 where the probability of requiring operation of the SGTS is reduced. (continued) d ABWR B 3.6-57 5/31/89
Standby Gas Treatment B 3.6.4.3 BASES (continued) ACTIONS E,1. E.2. E.3 (continued) If the Required Actions and associated Completion Times of Condition A or B are not met when handling irradiated fuel in the Secondary Containment, during CORE ALTERATIONS, or during operations with a potential for draining the reactor vessel, these actions are suspended to minimize the possibility of a radioactive release as. a consequence of a fuel handling acci-dent or inadvertent vessel draindown which uncovers the core. Suspension of these activities shall not preclude completion of the component to a safe, conservative position. If fuel is being handled while the reactor is in MODE 1, 2 or 3, then Conditions D and E are concurrently applicable. Comoletion Times All Completion Times are based'on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Standby systems should be checked periodically to ensure they will start and function properly. This surveillance require-ment ensures each subsystem will start on demand and continue to operate. Operation with the heaters on (automatic heater cycling to maintain temperature fulfills this requirement) for greater than or equal to 10 hours every 31 days reduces the buildup of moisture on the adsorbers and HEPA filters. (continued) O ABWR B 3.6-58 5/31/89 i
-___ A
-i Standby Gas Treatment B 3.6.4.3 i i
. BASES (continuedi ;
SURVEILLANCE SR 3.6.4.3.2. SR 3.6.4.3.3. SR 3.'6.4.3.4. SR 3.C.4.3.5 REQUIREMENTS i These surveillance requirements demonstrate that the designed-
~
.(continued). )
filtration capability of the system is maintained by verifying that the system flow, HEPA filters and charcoal adsorbers satisfy the in-place testing acceptance criteria, surveillance , intervals and procedures or Reference 3 and ANSI N510-1975. j The laboratory analysis of a representative. carbon sample (SR 3.6.4.3.4) must be performed in accordance with the testing criteria of Regulatory Position C 6.a.of
Reference:
- 3. The carbon sample to be used in this test must be obtained in -
accordance with Regulatory Position C.6.b of Reference 3. The in place acceptance criteria for SR 3.6.4.3.2 and SR 3.6.4.3.3 are defined in Regulatory Position C.5 of Reference 3. The system flow rate is verified during subsystem operation, for SR 3.6.4.3.2, SR 3.6.4.3.3 or SR 3.6.4.3.5 when tested in accordance with ANSI N510-1975 (Ref. 2). In addition to the 18 month surveillance intervals, SR 3.6.4.3.2, SR 3.6.4.3.3 and SR 3.6.4.3.4 are required whenever modifications or events which may affect the integrity of the HEPA filters or charcoal adsorbers have. occurred. For the
. purpose of this specification, " filter service crea" means the area where painting, fire or chemical release occurred and from which an operating fan is taking suction.
SR 3.6.4.3.6 The SGTS filter train and isolation damper OPERABILITY.must be demonstrated to ensure that for the LOCA and fuel handling accident events, the licensing basis assumptions remain valid. A test actuation signal is used to test each SGTS subsystem and to verify that these system components will function when required. SR 3.6.4.3.7 l: The electric heaters in each filter train are provided and sized to reduce the humidity of the treated air from 100% to-70%. The incoming 1200 cfm of air at a temperature of [ -]'F (Ref. 4) and 100% relative humidity requires [ ] kw (Ref. 5) of heat to lower its relative humidity to 70%. The test will be performed in accordance with ANSI N510-1975. (continued) 1 ABWR B 3.6-59 5/31/89
Standby Gas Treatment B 3.6.4.3 BASES (continued) SURVEILLANCE Surveillance Frequencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. 10 CFR 100, Reactor Site Criteria.
- 2. ANSI /ASME N510-1975, Testing of Nuclear Air Cleaning Systems.
- 3. USNRC Regulatory Guide 1.52, Rev. 2.
- 4. ANSI /ASME N509-1976, Nuclear Power Plant Air Cleaning Units and Components..
- 5. Standard Review Plan, 6.5.3.
- 6. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O O 4 ABWR B 3.6-60 5/31/89
l RCW/RSM System - Operating B 3.7.1 B.3.7 PLANT SYSTEMS B 3.7.1 Reactor Buildina Coolina Water / Reactor Buildina Service Water System - Operatina j BASES ,_, BACKGROUND The Reactor Building Cooling Water (RCW) System and Reactor Building Service Water (RCW) System together are designed to provide cooling water for the removal of heat from plant ; auxiliaries, such as Residual Heat Removal (RHR) system motor and pump seal coolers, heat exchangers, standby diesel generators, room coolers for Emergency Core Cooling System (ECCS) equipment etc., required for a sofe reactor shutdown following a design basis transient or accident. The RCW/RSW system also provides cooling to plant components, as required, during normal operation and reactor shutdown modes. During an accident, most equipment required for normal operation only is isolated from the RCW system; cooling is directed only to safety related equipment, the control rod drive pumps, and the instrument air compressors. The operator may, at his discretion, isolate the control rod drive pump and the instrument air compressor coolers. For the purpose of this technical specification the combined 9 RCW/RSW system consists of the [ ultimate heat sink (UHS)), three independent RCW and RSW cooling water headers (Divi-sions I, II, and III) and their associated pumps, piping, valves and instrumentation. These divisions cool equipment in Division I, II, and III respectively. These divisions are mechanically and electrically separated from each other so that failure of one division will not affect the OPERABILITY of the other divisions. The RCW and RSW systems operate together to provide the cooling water required by the RCW system loads. Operability of each RCW division is dependent upon the operability of its corresponding RSW division. Therefore, an inoperable RSW subsystem renders the associated RCW/RSW division inoperable, and ultimately all components cooled by that division must be declared inoperable. Cooling water is pumped from the [ UHS) by the three Reactor Building Service Water (RSW) Divisions to the RCW heat exchangers through the three main redundant supply headers. After removing heat from the RCW beat exchangers, the water is discharged to the [ UHS). (continued) ABWR B 3.7-1 5/31/89
RCW/RSW System - Operating B 3.7.1 BASES BACKGROUND The three RCW divisions supply cooling water to redundant (continued) equipment required for a safe reactor shutdown. The spe-cific equipment for which the RCW/RSW system supplies cooling water is listed in Reference 2. Each RCW/RSW
- division has two RCW pumps, two heat exchangers, and two RSW i pumps. One RCW pump, one RSW pump, and one heat exchanger in each division are adequate to support normal operation, and are in operation at all times. The standby pump in each division automatically starts and the standby heat exchanger is automatically valved in upon receipt of a LOCA signal.
The RCW/RSW system is designed to withstand a single active or passive failure (pump seal or valve failure) coincident with a loss of offsite power without losing the capability to supply adequate cooling water to equipment required for safe reactor shutdown. APPLICABLE Following a design basis accident or transient, the RCW/RSW { SAFETY system will realign itself and continue to operate auto- l ANALYSES matica11y and without operator action. Manual initiation of i supported systems, e.g., RHR suppression pool cooling, is however, performed for long term cooling operations. , The ability of the RCW/RSW system to support long term cooling of the reactor or containment is evaluated in SSAR Chapters 6 (Engineered Safety Features), 9 (Auxiliary Systems) and 15 (Accident Analyses). These analyses explic-itly assume the RCW/RSW will provide adequate cooling support to the equipment required for safe reactor shutdown. I These analyses include the evaluation of the long term i containment response after a design basis Loss Of Coolant Accident (LOCA). The RCW/RSW system provides cooling water ] to the RHR heat exchangers for all modes of RHR to limit the j suppression pool temperature and containment pressure following a LOCA. This ensures the containment can perform its intended function of limiting the release of radioactive materials to the environment following a LOCA. The RCW/RSW system also provides cooling to other components assumed to function during a LOCA. The safety analyses for long term cooling were performed (References 4 and 5) for a LOCA concurrent with a loss of affsite power, and minimum available diesel generator power. The worst case single failure which would affect the perfor-mance of the RCW/RSW systems is the ftilure of one of the three standby diesel generators which m ld affect one subsystem of the RCW/RSW system. (continued) ABWR B 3.7-2 5/31/89
RCW/RSW System - Operating-B 3.7.1 m bb BASES APPLICABLE RCW/RSW System - Operating satisfies the requirements of SAFETY. Selection Criterion 3 of the NRC Interim Policy Statement on. ANALYSIS Technical Specification Improvements as documented in (continued) Reference 8. LCOs The OPERABILITY of the RCW/RSW system is required to ensure the effective operation of the RHR system in removing heat from the reactor or containment and the effective operation of other safety related equipment during a design basis-accident or transient. The OPERABILITY of each independent division of the RCW/RSW system is based on 1)'having an OPERABLE [ UHS], 2) all pumps and heat exchangers in all RCW/RSW divisions OPERABLE, 3) OPERABLE RCW flow paths to all components required to be OPERABLE, 4) OPERABLE RSW flow paths capable of taking suction from the [ UHS] and transfer-ring the water to the RCW heat exchangers, as required, and
- 5) a minimum water level in the pump well of the intake structure.
Requiring all divisions OPERABLE assures adequate capability to meet cooling requirements of the equipment required for
.(O
,/ safe shutdown.
The OPERABILITY of the (VHS] is based on having greater than the minimum required water level and less than the maximum allowed water temperature. APPLICABILITY The requirements for OPERABILITY of the RCW/RSW system in MODES 1, 2, and 3 are governed by the requited OPERABILITY of the ECCS (LC0 3.5.1), the diesel generators (LC0 3.8.1) and the applicable Containment and Decay Heat removal modes of the RHR system (LCOs 3.4.5, 3.6.2.3 and 3.6.2.4). RCW/RSW system requirements for other operating modes are covered in LC0 3.7.2. ACTIONS A.I. A.2 With one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger in any division inoperable, that division is still operable to provide cooling water to all its served components. One RCW pump, one RSW pump, and one heat exchanger in each division is adequate to meet all design criteria except for long term containment cooling. As such, an extended time is allowed for a single RCW and/or RSW pump 1' (continued) ABWR B 3.7-3 5/31/89
RCW/RSW System - Operating B 3.7.1 BASES AClIONS and/or heat exchanger to be out of service, in any of the (continued) divisions, provided the RCW/RSW subsystems are otherwise OPERABLE. As noted, each Division may be in this condition concurrently. If the inoperable component (s) can not be restored to OPERABLE status within the allowed Completion Time, then Condition D applies. Alternatively, the affected system or components (i.e. those requiring all RCW/RSW pumps and heat exchangers in a division to be OPERABLE) may be declared inoperable and the appropriate actions taken. If two RCW pumps, two RSW pumps or two heat exchangers are out of service in one or more divisions, then Action B, C or D applies. Because of the level of redundancy in the RCW/RSW system design, the provisions of LC0 3.0.4 are not applicable, and Modes 1, 2, or 3 may be entered while in ' ,ndition A. B.1. B.2. C.1. C.2.1. C.2.2 The loss of both RCW pumps, both RSW pumps, or both heat exchangers in the same Division would render that division completely inoperable and the associated equipment cannot perform the intended function. With any RCW/RSW flow path inoperable (e.g. valve failure, flow blockage) the system (s) served (cooled) by that RCW/RSW division may become inoper-able; however, the RCW/RSW division may still be able to supply cooling to all other associated systems. Other types of failures, such as the failure of ar, isolation valve (which would isolate non-essential components in case of a LOCA) may reduce the capability of the division to less than its design value for the designated essential components. In these cases, the division (s) is considered inoperable and continued operation is only allowed for a shor'. period. With one division inoperable, for reasons other than Condi-tion A, there is still sufficient safety-related cooling water available to assure safe shutdown capability, even with an additional single failure. However, because overall system redundancy is reduced, continued operation is only allowed for seven days. With two divisions inoperable, the system is still capable of performing its intended function; however, system redundancy is severely degraded so that continued operation is only allowed for 24 hours at which time at least one of the inoperable Divisions must be restored to OPERABLE status. Additionally, the initial O (continued) ABWR B 3.7-4 5/31/89
RCW/RSW System - Operating ; B 3.7.1 1 7 () BASES I ACTIONS inoperable division must be restored to OPERABLE status (continued) within seven days. Alternatively, the affected system (s) or component (s) under any of these conditions, may be declared inoperable and the associated Actions taken. D.1. D.2 If the Required Actions and associated completion times of Conditions A, B, or_ C cannot be met, or with all three RCW/RSW Divisions inoperable, the reactor is required to be in MODE 3 in 12 hours and in MODE 4 in the following 24 hours. In MODE 4 the system requirements are reduced as specified in LC0 3.7.2. If MODE 4 cannot be achieved because of the inoperable RCW/RSW division (s), the reactor coolant temperature should be maintained as low as practicable using an alternate heat removal method. Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of avail-able systems and the time required to reasonably complete the Required Action. v SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This surveillance verifies the [ ultimate heat sink] has sufficient cooling water to satisfy the design basis of 30 day cooling capability with no external makeup source. With the [ ultimate heat sink) inoperable, the affected RCW subsystems must be declared inoperable. [This SR is only applicable when the UHS consists of a dedicated body of water such as a cooling tower basin or pond.] SR 3.7.1.2 Verification of the { UHS) temperature ensures that the heat removal capability of the RCW/RSW system is within the , assumptions of the design basis accident analysis. SR 3.7.1.3 This surveillance requirement verifies that the water level inside the pump wells of the intake structure is sufficient for proper operation of the RSW pumps (NPSH and pump vortex-ing are considered in determining this limit). If the water _c ,') m (continued) ABWR B 3.7-5 5/31/89
yev--- y-- ., .- I RCW/RSW System - Operating 3.7.1 BASES SURVEILLANCE level is > [ ] feet, there .s sufficient margin to the REQUIREMENTS minimum level requirement ([ ] feet) so that the surveil-(continued) lance can be performed every 14 days. However, if the level s[ ] feet, the surveillance should be performed more frequently (every 12 hours) to ensure adequate suction conditions for the RSW pumps. [This surveillance is not applicable for sea water plants.] SR 3.7.1.4 Verification of the correct alignment of all valves is essential to ensure the proper flow paths servicing safety related systems or components for the RCW/RSW subsystems. SR 3.7.1.5 This surveillance verifies the OPERABILITY of the [ ultimate heat sink active components). The 15 minute duration for operation is sufficient to monitor the steady state perform-ance of the required active components. [This SR is only i applicable for plants taking credit for UHS active components such as cooling tower fans.] SR 3.7.1.6 This surveillance verifies the automatic isolation valves of the RCW/RSW system will automatically switch to the safety or emergency position tn provide cooling water exclusively to designated eciipment during an accident event. This sur-veillance also verifies the automatic start capability of the standby RCW/RSW pumps and [ ultimate heat sink] active components and the automatic inclusion of the standby heat exchanger in the coding loop on both the RCW and RSW sides. Surveillance Frequencies In general, Surveillance Frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system /comisonent status.
~
(continued) 4 O ABWR B 3.7-6 5/31/89
gy p ; 1 RCW/RSW System - Operating hoQ 3.7.1 fy -y 7, -( . ; k/ BASES REFERENCES' 1. Regulatory Guide l~.27, "Ultim'te a Heat Sink'for Nuclear. Power Plants", Para. ;C.I. 2 '. ABWR SSAR, . Table 9.2-4.
'3.
- Regulatory Guide 1.1, " Net Positive' Suction Head for
' Emergency CJre Cooling and Containment Heat: Removal ~
System' Pumps".
- 4. ABWR SSAR, Section 6.2.1.1.3.3.1.4.
- 5. ABWR SSAR, Section 6.2.2.3.1.
- 6. NED0-31466, "Technicai Specification Screening Criteria Application and Risk Assessment", November 1987.
A : V :- ALWR B 3.7-7 5/31/89
1 1 RCW/RSW System - Shutdown B 3.7.2 B.3.7 PLANT SYSTEMS h B 3.7.2 Reactor Buildina Coolina Water (RCW)/ Reactor Buildina Service Water (RSW) System - Shutdown BASES BACKGROUND The Reactor Building Cooling Water (RCW)/ Reactor Building Service Water (RSW) System is described in the Bases for LCO 3.7.1. APPLICABLE The ability of the RCW/RSW syste;n to support long term SAFETY cooling of the reactor or containment is evaluated in SSAR ANALYSES Chapter 6 (Engineered Safety Features), 9 (Auxiliary Systems) and 15 (Accident Analyses). These analysis expli-citly assume that the SSW will provide adequate cooling support to the equipment required for safe reactor shutdown. These analyses include the evaluation of the long term containment response after a design basis accident (DBA). The consequences of any accidents or transients during shutdown conditions (MODES 4, 5 or when handling irradiated fuel in the primary or secondary containment) such as an inadvertent draindown of the reactor vessel are bounded by these analyses. In shutdown conditions, core cooling may be required if the O1 reactor vessel should inadvertently draindown. The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The consequences of an inad-vertent draindown of the vessel, which may require ECCS subsystem operation during MODES 4 cnd 5, are bounded by I these analyses. RCW/RSW System - Shutdown satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on ' Technical Specification Improvements as documented in Reference 1. LCOs The OPERABILITY of the RCW/RSW system is required to provide a coolant soerce to ensure the effective operation of the Residual Heat Removal (RHR) system in removir.g heat from the reactor or containment and the effective operation of other safety related equipment during a design basis accident or transient. In modes 4 and 5 the RCW/RSW system is required to support a reduced complement of equipment. The OPERABILITY of each independent division of the RCW/RSW system is based on having an OPERABLE [ UHS), one OPERABLE (continued) ABWR B 3.7-8 5/31/89
. _ _ _ . _ _ . _____________-_____m
pg - - RCW/RSW System'- Shutdown B 3.7.2 X/ ' BASES-LCOs .RCWland RSW pump in'the subsystem, and an OPERABLE flow path-capable of taking suction from the associated RSW [ UHS intake] and transferring the water to the appropriate RCW-heat exchanger, as required.' This reduced complement of one ) RCW pump, one RSW pump, and one heat exchanger beine required for each operable division is due to:the educed heat loads during MODES 4 and 5. Only those RCW/RSW sub-systems associated with the systems and components required-to be OPERABLE by LC0 3.4.5, LCO 3.5.2, LCO 3.8.2, LC0 3.9.8 and LC0 3.9.9 are required to be OPERABLE in MODES 4 and 5, and when handling irradiated fuel in the secondary containment. The intake structure water level requirement of SR3.7.1.6' is required to be met for the RSW to be considered operable. APPLICABILITY The requirements for OPERABILITY of the RCW/RSW system in MODES 4 and 5, and when handling irradiated fuel in the secondary containment are governed by the required OPERABILITY of the equipment serviced by the RCW/RSW system in those MODES. RCW/RSW system requirements for modes 1, 2,.
/ \.)T and 3 are specified in LCO 3.7.1.
ACT10NS A.I In Modes 4 and 5, and when handling irradiated fuel in the secondary containment, the RCW/RSW system is required to be OPERABLE to support operation of required diesel generators, ECCS subsystems and RHR shutdown cooling subsystems. If an RCW/RSW division associated with required equipment is inoperable, the capability of the supported systems to perform their intended functions cannot be assured. There-fore, the affected systems or components should be declared inoperable and the Required Actions specified in the appropriate LC0 followed. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of avail-able systems and the time required to reasonably complete the Required Actions. (continued) f N. . ABWR B 3.7-9 5/31/89
1 RCW System - Shutdown i 3.7.2 l BASES , i SURVEILLANCE SR 3.7.2.1 throuah SR 3.7.2.6 h REQUIREMENTS The Bases provided for SR 3.7.1.1 through 3.7.1.6 are i applicable. j i Surveillance Frequencies l I In general, Surveillance Frequencies are based on industry ~{' accepted practice and engineering judgement considering the unit cnnditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. REFERENCES 1. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987. O O ABWR B 3.7-10 5/31/89 _ _ _ __ _ - A
CR HVAC ER System B 3.7.3 h
.V B 3.7 PLANT SYSTEMS B 3.7.3 Control Room HVAC Emeroency Recirculation System L
BASES BACKGROUND The Control Room HVAC Emergency Recirculation (CRHER) system is designed to provide a radiologically controlled environment to ensure the control room will remain habitable for personnel during and following a design basis accident-(DBA)-(Ref. 1). To meet these requirements, this system is. designed in conjunc-tion with control room design provisions such that the radia-tion' exposure to personnel inside the control room is less than 5 rem whole body consistent with the requirements of' General Design Criterion 19 of Appendix A to 10 CFR 50. The safety related function of the CRHER system consists of two independent and redundant high efficiency air filtration subsystems. Each subsystem consists of a demister, an electric heater, prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section, a second HEPA filter and a. fan. Demisters remove entrained water droplets from the inlet stream, thereby protecting prefilters, HEPA filters, and adsorbers from water damage [ and pluggin,g. Heaters follow the demisters and are designed to heat the incoming stream to reduce the stream's relative humidity before the air stream reaches the filters and' adsorbers. The prefilters and HEPA filters remove particulate matter, which may be radioactive. The charcoal adsorbers provide a holdup period for gaseous iodine allowing time to decay. The CRHER system's safety related function is as a standby system, parts of which are also operated during normal plant operation to maintain the control room environment. Upon receipt of an actuation signal '(indicative of conditions that could result-in radiation exposure to control room personnel) the CRHER system automatically switches to the isolation mode of operation to prevent infiltration of contaminated air into the control room. A system of dampers isolates the control room and control room air flow is recirculated and processed through either of the two filter subsystems. After 10 minutes of isolation and when conditions permit, fresh air can be manually brought into the control room through the charcoal filter system. (continued) ( ABWR B 3.7-11 5/31/89
CR HVAC ER System B 3.7.3 BASES (continued) APPLICABLE The ability of the CRHER system to maintain the habitability 0F SAFETY the control room is an explicit assumption for the safety ANALYSES analyses evaluated in SSAR Chapters 6, (Engineered Safety Features) and 15, (Accident Analyses). The isolation mode of the CRHER system is assumed to operate following a Lots Of Coolant Accident (LOCA), main steam line break (MSLB) and fuel handling accident. (Ref. 2). The radiological doses to control room personnel as a result of the various design basis accidents are summarized in Reference 3. In all cases, the doses are within the limits of 10 CFR 50, Appendix A, General Design Criterion 19. Control Room HVAC Emergency Recirculation system satisfies the requirements of Selection Criterion 3 of the NRC Interim , Policy Statement on Technical Specification Improvements as ! documented in Reference 6. ! LC0 Two redundant subsystems of the CRHER system are required to enrure at least one is available assuming a single failure disables the other subsystem. The CRHER subsystem consists of two independent and redundant filtration trains. Should any component in one subsystem fail, filtration can be performed by the other subsystem. The OPERABILITY of each independent subsystem is based on having adequate system flow and OPERABLE HEPA filters, chLrcoal adsorbers and heaters. A description of what is required for CRHER to be considered OPERABLE is provided in the Background Section. APPLICABILITY The CRHER system is required to be OPERABLE in MODES 1, 2, 3, 4, 5 and when handling irradiated fuel in the primary or secondary containment to ensure the control room will be habitable for personnel during and following a design basis accident. l (continued) l O l' ABWR B 3.7-12 5/31/89 L __
1 CR HVAC ER System d B 3.7.3 f g l Cl BASES (continued) ACTIONS A_J With one CRHER subsystem inoperable, the remaining OPERABLE subsystem can maintain the habitability of the control room during the postulated design basis accidents assuming no additional failures in the OPERABLE subsystem. However, if a single active component fails concurrent with the postulated design basis accident, depending on the specific failure, the CRFA system may not be able to perform its intended safety function. Therefore, system reliability is reduced and operation is only allowed to continue for a limited time. B.1. B.2 1 In MODES 1, 2 and 3, with an .noperable subsystem not restored to OPERABLE status and the associated Completion Time , not met or both subsystems inoperable, the CRFA system may not be capable of performing its intended safety function and the reactor is required to be in MODE 3 and subsequently in MODE 4. C.1. C.2.1. C.2.2. C.2.3 m In MODES 4, 5 and when handling irradiated fuel in the ()" primary or secondary containment, if an in;. able subsystem cannot be restored to OPERABLE status and the associated Completion Time is not met, the remaining OPERABLE subsystem may be placed in the isolation mode of operation. This ensures the CRHER system is operating prior to any potential design basis accident that could require the CRHER system to actuate. This action provides a continuous check of the operation of the CRHER system. Alternatively, CORE ALTERATIONS, handling of irradiated fuel in the primary or secondary containment and operations that could drain the ! reactor vessel must be suspended. This eliminates the potential design basis accidents under these conditions that could require the CRHER system to function. Suspension of these activities shall not preclude completion of the movement of a component to a safe, conservative position. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. l- O (continued) V I ABWR B 3.7-13 5/31/89
CR HVAC ER System B 3.7.3 BASES (continued) SURVEILLANCE General i REQUIREMENTS In addition to the ANSI N510 test requirements, and norn ' ! preventive and post-maintenance testing, there are a number of specific tests which must be performed to ensure proper functioning of the CRFA system. SR 3.7.3.1 Standby systems should be checked periodically to ensure they will start and function properly. This Surveillance ! Requirement ensures each subsystem will start on demand and continue to operate. Operation with the heaters on for greater than or equal to 10 hours every 31 days reduces the buildup of moisture on the adsorbers and HEPA filters. SR 3.7.3.2. SR 3.7.3.3. SR 3.7.3.4 and SR 3.7.3.5 These Surveillance Requirements demonstrate the designed filtration capability of the system is maintained by verifying the system flow rate, HEPA filters, end charcoal adsorbers satisfy the in-place testing acceptance criteria, the surveillance intervals and procedures of Reference 4 and ANSI N510-1975. The laboratory analysis of a representative 4 carbon sample (SR 3.7.3.3) must be performed in accordance I with the testing criteria of Regulatory Position C.6.a of The carbon sample to be used in this test must I Reference 4. be obtained in accordance with Regulatory Position C.6.b of Reference 4. The in-place acceptance criteria of SR 3.7.3.2 and SR 3.7.3.3 are defined in. Regulatory Position C.5 of Reference 4. The system flow rate is verified during subsystem operation for SR 3.7.3.2, SR 3.7.3.3 or SR 3.7.3.5 when tested in accordance with ANSI N510-1975. In addition to the 18 month Frequency, SR 3.7.3.2, SR 3.7.3.3 and SR 3.7.3.4 are also required whenever major modifications or events (e.g. painting, fire or chemical release) which may affect the integrity of the HEPA filters or charcoal adsorbers have occurred. For the purpose of this specification, " filter service area" means the area where painting, fire or chemical release occurred and from j which suction is being taken by an operating fan. t-(continued) O ABWR B 3.7-14 5/31/89
hf i CR'HVAC'ER System B 3.7.3
' ry w k, ,/ BASES (continued)
SURVE1LLANCE SR 3.7.3.6 REQUIREMENTS (continced). This Surveillance Requirement verifies the duct heater performance of the CRFA system. The test'is performed in accordance with Section 14 of ANSI N510-1975 [with the exception of. the 5% current phase balance of Section 14.2.3. The offsite power system for the ABWR.may consist of a non-- transpositional grid which has an inherent unbalanced load distribution that results in unbalanced _ voltages' in the plant. Voltage unbalances exceeding the 5% criteria of ANSI N510-1975 are not atypical.] SR 3,7.3.7 This Surveillance Requirement verifies the CRHER system will automatically switch to the isolation mode of' operation to maintain con +rol room habitability on receipt of an actuation signal. Surveillance Frequencies
'In general, Surveillance Frequencies are based on industry -f-accepted practice and. engineering judgement considering the
( unit conditions required to perform the test, the ease of.
' performing the test and a likelihood of a cha'nge in the system / component status. The Surveillance Frequencies for testing of the~ HEPA filters and charcoal absorber units are consistent with the requirements of Reference 5.
REFERENCES 1. ABWR SSAR, Section 6.5.
- 2. ABWR SSAR, Section 9.4.
- 3. ABWR SSAR, Section 15.
- 4. Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Post Accident Engineered Safety-Feature -
Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants", Revision 2, March 1978.
- 5. Generic Letter 83-13, " Clarification of Surveillance ;
Requirements for HEPA Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup Systems," March 2, 1983.
- 6. NED0-31466, " Technical Specification Screening Criteria f3 - Application and Risk Assessment," November 1987.
V ABWR B 3.7-15 5/31/89 L- - - - - - - - _ - - - __ _ _ _ _ _ _ _ _ _ _ _ _________ __ _ _ _ ___ _ _ _ _
Main Condenser B 3.7.4 8 3.7 PLANT SYSTEMS B 3.7.4 Main Condenser Offaas BASES BACKGROUND During plant operation, steam from the low-pressure turbine is exhausted directly into the condenser. Air leakage and noncondensible gases are collected in the condenser, then exhausted through the steam-jet air ejectors to'the Offgas System. The offgas from the main condenser normally includes radioactive gases. The Main Condenser Offgas System has been incorporated into the plant design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radio-lytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser and the water and condensibles are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the offgas condenser. APPLICABLE The main condenser offgas gross gamma radioactivity rate is SAFETY an initial condition of the Offgas Eystem Failure Event ANALYSES (Ref. 1). The analysis assumes a gross failure in the Off-gas System that results in the rupture of the Offgas System pressure boundary. The gross gamma radioactivity rate is controlled to ensure that during the event the calculated offsite doses will be well within the limits of 10 CFR 100. Main Condenser Offgas satisfies the requirements of Selec-tion Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. LCO lo ensure compliance with the assumptions of the Offgas System Failure Event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 pCi/MWt-sec at 30 minutes decay. The LC0 is established consistent with this requirement (3926 MWt x 100 pCi/MWt-sec = 390 millicuries /second). (continued) O ABWR B 3.7-16 5/31/89
l h Main Condenser B 3.7.4 yx. h,[ BASES (continuedi APPLICABILITY The LC0 is applicable when steam is being exhausted to the main condenser. This occurs'during MODE 1 and during MODES
. 2 and 3 with any main' steam line not isolated. In MODES 4 and 5,. steam is not being exhausted to the main condenser and the requirements are not applicable.
ACTIONS. . A .1 If the offgas radioactivity rate limit is exceeded, a limited time is permitted to restore the gross gamma.radioa-ctivity rate to within the limit because of the large margin to permissible dose and exposure limits. B. l . B.2 ' If the gross gamma radioactivity rate-is not restored to within.the limits and the associated Completion Time is not met, all main steam lines must be isolated. This isolates - the condenser from the source of the radioactive steam. Comoletion Times V All Completion Times are based on industry accepted practice and engineering judgement considering the number of avail-able systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This surveillance periodically ' analyzes a sample of the offgas to ensure that the required limits are satisfied. If the measured rate of radioactivity increases significantly (by ;t 50% percent after correcting for expected increases due to changes in THERMAL POWER), the isotopic analysis is performed within 4 hours after the increase is noted to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The Surveillance Frequencies are considered adequate based on the avail-ability of instrumentation to continuously monitor the offgas. (continued) O Q ABWR B 3.7-17 5/31/89
Main ondenser B 3.7.4 s' BASES (continued) REFERENCES 1. ABWR SSAR, Section 15.7.1.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
1 O O' ABWR B 3.7-18 5/31/89
y
'A.C. Sources - Operating-B 3.8.1
();
,~~
B3.8L.ELECTRICALPOWERSYSTEMS B 3.8.1 A.C. Sources - Operatina BASES BACKGROUND. The A.C. power system' consists of the ',ff-site power sources (preferred power), [an alternate AC source (on-site combustion turbine),] and the on-site standby power-sources. As required by General Design Criteria 17 (Ref.1), the design of the A.C. power system provides-independence and redundancy to ensure an available source of power to the Engineered Safety feature (ESF) systems. 4 The on-site Class IE A.C. distribution system is divided into. three divisional, redundant 1oad groups so that proper functioning of any one group will ensure the minimum safety functions are being performed. Each divisional A.C. load group has connections to two preferred (off-site) power supplies, [to the alternate AC source,] and to a single diesel generator. Each diesel generator is exclusively connected to a single 6900-volt divisional load group. A.C. power from the two off-site sources [and the alternate AC-( source,) to the on-site electrical distribution network is (]/ supplied by physically independent circuits, each of which is capable of supplying all ESF loads through the three divisional buses. One off-site source is connected to the Startup Transformer which provides direct ties to each of the ESF buses. The second off-site source is connected to the main transformer. With the main generator off .line, the main transformer back-feeds off-site power.from the main distribution grid to the two Auxiliary Transformers and each ESF bus is tied to one of the two Auxiliary Transformers. With the main generator on line (generator output breaker closed) supplying power to the grid via the main transformer, the. Auxiliary Transformers are powered directly by the generator. [ Additionally, the on-site alternate AC source provides a direct feed, via manual connection only, to each of the ESF buses.] (continued) ABWR B 3.8-1 5/31/89
A.C. Sources - Operating B 3.8.1 BASES (continued) BACKGROUND The on-site standby A.C. power source consists of three diesel (continued) generator units, which supply standby power to the three divisional 6900-volt emergency buses. Each diesel generator starts automatically on either a LOCA signal (i.e., low reactor water level signal or high drywell pressure signal) or a loss of off-site power signal (i.e., bus undervoltage). Additionally, the diesel generators will automatically tie to their respective buses on a loss of off-site power signal. For Divisions 1, 2 and 3, the automatic diesel start and the transfer of power from normal to emergency power supplies is controlled by the Load Shedding and Sequencing Logic (LSSL). The logic actuates on Loss of Off-Site Power (LOSP) or LOCA signals. The logic starts the diesel generators and, if an undervoltage exists on a Division 1, 2 or 3. bus, it sheds all loads, except the power centers, from the affected 6.9 kV bus and then sequentially starts the vital loads. Ratings for the diesel generators satisfy the requirements of Regulatory Guide 1.9 (Ref. 2). The continuous service rating of the diesel generators is 5000 kw on an 8800 hour base with 110 percent load permissible for 2 hours in any 24 hour period. The diesel generators are rated at 6900-volts, three phase, 60 Hz, and are capable of attaining rated frequency and voltage within 13 seconds after receipt of a start signal (Ref. 3). The ESF systems which are powered from divisional power sources are listed in Reference 5. APPLICABLE The initial conditions of design basis transient and accident SAFETY analyses in the SSAR Chapters 6, Engineered Safety Features, ANALYSES and 15, Accident Analyses, assume all ESF systems are OPERABLE. The A.C. power system is designed to provide sufficient capa-city, capability, redundancy and reliability to ensure the availability of necessary power to ESF systems so that the fuel, reactor pressure vessel and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for LCO Sections 3.2 (Power Distribution Limits) and 3.6 (Containment Systems). , (continued) i e ABWR B 3.8-2 5/31/89 l 4
A.C. Sources - Operating
'B 3.8.1 gy BASES (continued)
(J l APPLICABLE- The OPERABILITY of the power sources are consistent with the SAFETY. initial assumptions of the accident analyses and are based upon ANALYSES- maintaining the minimum required on-site A.C. and D.C. power (continued) sources and associated distribution systems OPERABLE during accident conditions, with 1) an assumed loss of off-site power and 2) a single failure of one of the standby A.C. sources. A.C. Sources - Operating satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference'15. LC0 Two physically independent circuits between the off-site transmission r.etwork and the on-site Class IE distribution system, and the three independent diesel generators ensure availability of the required power to respond to accidents and transients and maintain the unit in a safe shutdown condition. The two circuits from off-site are required to be " physically independent" such that a single component fault (e.g., breaker trip) will not cause both power sources to be lost to one or more 6900-volt emergency bus (es). Thus, a physically indepen-Q t/ dent circuit consists of one incoming line to the switchyard, a circuit path (including breakers and disconnects) to an ener-gized transformer (either the startup transformer or the two auxiliary transformers), and a circuit path from the energized transformer to the 6900-volt emergency buses of Divisions 1, 2 and 3 (including the associated 6900-volt emergency bus supply breakers). Each independent diesel generator consists of the diesel generator and the associated equipment required to ensure performance of the diesel generator's safety related functions. The fuel oil system requirements are located in LCO 3.8.3. (continued) i O ABWR B 3.8-3 5/31/89 i
-I
A.C. Sources - Operating 8 3.0.1 BASES fcontinued) APPLICABILITY The A.C. power sources are required to be OPERABLE in MODES 1, 2 and 3 to ensure that:
- 1. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of antici-pated operational occurrences or abnormal transients, and,
- 2. Adeo.uate core cooling is provided, and containment integ-rity and other vital functions are maintained in the event of a postulated DBA.
A.C. power requirements for MODES 4 and 5 are covered in l LC0 3.8.2. ACTIONS The Required Actions specified for the levels of degradation of the power sources provide restrictions upon continued operation commensurate with the level of degradation. For simplicity, the Actions regarding Conditions where the available off-site A.C. sources are less than required by the LCO assume no difference between the two off-site sources. In actuality, since one source of off-site power is through the main transformer, there will be different operational impacts depending on which off-site source is lost and what the initial operational conditions are at the time it is last. If the main generator is on line w% n the main transformer or associated off-site source, is lost, then either the reactor will be tripped automatically (due to load rejection) or the reactor will be subsequently restricted in power. This will depend on whether the initial power was above or below the bypass set-point for the RPS scram on turbine stop valve closure or turbine control valve fast closure (see LC0 [RPS Actuation Instrumentation]). In any case, with the main transformer or grid unavailable, there is no incentive for extended continued operation. Therefore, the completion time associated with unavailable off-site A.C. power sources are directed primarily at the off-site source feeding the start-up transformer. Nonetheless, all associated Required Actions and Completion Times are applicable to both off-site A.C. sources. However, if the reactor is tripped upon loss of the main distribution grid, irregardless of the status of the off-site source feeding the startup transformer, then LC0 3.8.2 becomes applicable in lieu of 3.8.1. (continued) O ABWR B 3.8-4 5/31/89
l A.C. Sources - Operating B 3.8.1
) BASES (continued)
A.I. A.2 ! ACTIONS (continued) With one of the required circuits from off-site inoperable, I sufficient off-site power is available from the other reo'* ired l off-site circuit to ensure that the unit can be maintair s in a 1 safe shutdown condition following a design basis transient or accident. Even failure of the remaining required off-site circuit will not jeopardize a safe shutdown of the unit because of the redundant standby diesel generators [and alternate AC source). Operation could therefore safely continue if the availability of the remaining sources is verified. However, since the system reliability is degraded, a time limit on continued operation is imposed. To ensure a highly reliable power source, it is necessary to verify the availability of the remaining required circuit from off-site on a more frequent basis if one circuit from off-site is inoperable. The availability of the remaining required circuit from off-site must be "erified within one hour and once per 8 hours thereafter until tne inoperable circuit from off-site is restored to OPERABLE status. 4 Power operation may continue for a period that should not
<~ exceed 14 days. If the source is not restored within 14 days, a controlled shutdown must be initiated per Required Actions
( 's) I.1 and 1.2. { B.l. B.2. B.3. B.4 With one diesel generator inoperable, sufficient A.C. power sources remain available to ensure safe shutdown of the unit in the event of a transient or accident even with the addition of a single failure. Operation could therefore safely continue for some period of time if the availability of the remaining sources is verified. Completion times are consistent with those of Condition A. The specific list of equipment encompassed by Required Action B.2 is provided in [Ref. 7]. This requirement is intended to provide assurance that a loss of off-site power, during the period that one diesel generator is inoperable, will not result in a complete loss of safety function of critical systems. The term verify, as used in this context, means to administrative 1y check, by examining logs or other information, to determine if certain components are out of service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component. O O (continued) ABWR B 3.8-5 5/31/89 l l L_ _ _ _ _
A.C. Sources - Operating B 3.8.1 m BASES (continued) ACTIONS Required Action B.3 provides an allowance to avoid unnecessary (continued) testing of the OPERABLE diesel generators when a diesel generator is declared inoperable because of a component which can be tested independently to restore the inoperable diesel
, generator to OPERABLE status. Regulatory Guide 1.108 (Ref. 8) defines the diesel generator unit as consisting of the engine, generator, combustion air system, cooling water system up to the supply, fuel oil supply system, lubricating' oil system, starting energy sources, auto start controls, manual controls and the diesel generator breaker. There are many potential failures of diesel generator subsystems that could cause a diesel generator to become inoperable and therefore, require all OPERABLE diesel generators be demonstrated OPERABLE but do not require starting and loading the inoperable diesel generator to restore it to OPERABLE status. Inoperabilities of diesel generators caused by failures of equipment that are not part of the defined diesel generator unit are categorized as i invalid failures in acc< rdance with Regulatory Guide 1.108 !
since the failure would not have prevented the diesel generator from performing its intended safety function. As such, they do not impact the technical specification surveillance frequency of the diesel generator that failed. Likewise, there should be no reason to require additional testing of OPERABLE diesel generators to determine if the same invalid failure mode exists. C.I. C.2. C.3. C.4. C.5 In Condition C, a level of redundancy is lost in both the off-site power system and the on-site A.C. power system. However, adequate power system redundancy is still provided by diverse sources of power (i.e. two diesel generators and one off-site source [as well as the alternate AC source]). The reliability of the power systems in this Condition is still quite good, however, because two sources of power have been lost, continued operation is allowed for only a limited time. With the available off-site and standby A.C. power sources aach one less than the LCO, power operation may continue for 72 hours. Additionally, the initial inoperable source must be restored to OPERABLE status within 14 days of discovery of the initial inoperable source (consistent with the Completion Times of Required Actions A.2 and B.4). If either an off-site or a standby A.C. source is restored to OPERABLE status within the required time, Condition A or B is reentered (whichever is applicable) which allows operation for 14 days from the time of loss of the initial inoperable source (consistent with the loss i i ' 1 (continued) ABWR B 3.8-6 5/31/89 i 1
i. A.C. Sources - Operating
' B 3.8.1
- n. !- BASES (continued)
~V ACTIONS C.1. C.2. C.3. C.4. C.5 (continued)
(continued) of one A.C. source in Condition A or B). If neither an off-site source nor a standby source is restored within the 72 r hours or the initial inoperable source is not restored within 14 days, whichever comes first, a controlled shutdown must be initiated per: Required Actions 1.1 and 1.2. D.1. D 2. D.3 .'D.4. D.5 With two diesel generators inoperable, the remaining standby A.C. power sources.are available to feed the minimum required ESF functions. Since the off-site power system, [the alternate A.C. source,) and one diesel generator provide sufficient sources of A.C. power at this level of degradation, the risk associated with continued operation for a short time is prob-ably less than that associated with an immediate controlled shutdown (the immediate shutdown could cause grid instability which could result in total loss of off-site A.C. power). However, since any inadvertent generator trip could also result in total loss of off-site A.C. power, the time allowed for continued operation is restricted. The intent here is twofold: (a) to avoid the risk associated with an immediate controlled shutdown and (b) to subsequently minimize the risk associated (qj with this level of degradation. l With the available standby A.C. electrical supplies two less than the LCO, operation may continue for 72 nours. Addition-ally, the initial inoperable diesel generator must be restored to OPERABLE status within 14 days of discovery of the initial inoperable die el generator (consistent with the Completion Times of Required Action B.4). If all but one of the diesel' generators are restored to OPERABLE status within the required time, Condition B is reentered which allows operation for 14 days from the time of the. loss of the initial inoperable diesel generator. If no standby A.C. supply is restored within 72 hours or the initial inoperable diesel generator is not restored within 14 days, whichever comes first, a controlled shutdown must be initiated per Required Actions 1.1 and 1.2. 1 (continued)- ! I n V l ABWR B 3.8-7 5/31/89
A.C. Sources - Operating B 3.8.1 BASES (continued) ACTIONS E.1. E.2. E.3. E.4. E.5. E.6 (continued) In Condition E, individual redundancy is lost in both the off-site power system and the standby A.C. power system (i.e., design generators). However, power system redundancy is provided by diverse power sources [as well as the alternate AC source] such that, as the Condition D, it may be less risky to continue operation for a short time than to immediately initiate a shutdown. Since the level of degradation here is more severe than Condition C or D, the time is restricted further. With one of required off-site sources and two of the diesel generators inoperable, operation may continue for 12 hours of which time either the inoperable off-site source or one of the inoperable diesel generators must be restored to OPERABLE i status. A second inoperable component (diesel generator or off-site source) must be restored to OPERABLE status within 72 hours of initial discovery of the second of the remaining two inoperable components (consistent with the Completion Times of Required Actions C.4 or D.4). Au ionally, the remaining inoperable component must be restored to OPERABLE status within 14 days of discovery of the initial inoperable component (consistent with the Completion Times of Required Actions A.2 or B.4). In other words, if one of the inoperable components is restored to OPERABLE status within the required Completion Time, then Condition C or D, as appropriate is reentered which allows operation for 72 hours from the time of loss of the second inoperable component. Similarly, if a second component is restored to OPERABLE status within the required Completion i Time, then Condition A or B, as appropriate, is reentered which allows operation for 14 days from the time of loss of the initial inoperable component. If one of the three inoperable components is not restored within 12 hours, or if the second inoperable component is not restored within 72 hours, or the initial inoperable component is not restored within 14 days, whichever comes first, a con-trolled shutdown must be initiated per Required Action 1.1 and l l 1.2. (continued) i l O ABWR B 3.8-8 5/31/89
-l A.C. Sources - Operating B 3.8.1 a
-{/m)! BASES (continued) v l ACTIONS F.1. F.2 (continued)
With both of the required circuits from off-site inoperable, sufficient standby A.C. power sources [ including the Alternate AC source] are available to maintain the unit in a safe shut-down condition in the' event of a design basis transient.or accident. Additionally, since the generator will be discon-nected from the distribution grid, the reactor will have already tripped or will be operating at a low power level. If the reactor is still at power there is some benefit for con-tinued operation if the grid connection can be restored within a reasonable time. However, since the A.C. power system is degraded, a time limit on continued operation is imposed. With the available off-site A.C. power sources two less than the LCO, operation may. continue for 72 hours. Additionally, the initial inoperable circuit from off-site must be restored-to OPERABLE status within 14 days of discovery of the initial inoperable circuit from off-site (consistent with the Comple - tion Time of Required Action A.2). If one circuit from off-site is restored to OPERABLE status within the required time, Condition A is reentered which allows operation for 14 days from the time of the loss of the initial inoperable circuit e from off-site. If no circuit from off-site is; restored within i 72 hours, or the initial inoperable circuit from off-site is not restored within 14 days, whichever comes first, a controlled shutdown must be initiated per Required Actions 1.1 and 1.2. G.I. G.2. G.3. G.4. G.5 With both of the required circuits from.off-site inoperable and one of the diesel generators inoperable, sufficient standby A.C. power sources (i.e., the remaining two diesel generators [and the alternate AC source]) are still available to maintain the unit in a safe shutdown condition in the event of a design ; basis transient or accident. Additionally, as per Condition F, ! if the reactor is still operating it will be at low power ; level. There is some potential benefit to continued operation I if the grid connection can be restored within a reasonable time. However, since the A.C. power system is significantly l degraded, a time limit on continued operation is imposed.
^
i (continued) l l l D L) . ABWR B 3.8 9 5/31/89
A.C. Sources - Operating B 3.8.1 BASES (continued) ACTIONS G.I. G 2. G.3. G.4. G.5 (Continued) l (continued) ] i With both required off-site sources and one diesel generator I inoperable, operation may continue for 24 hours, at which time l either the inoperable diesel generator or one of the off-site l sources must be restored to OPERABLE status. The second inoperable component (diesel generator or off-site source) must -i be restored to OPERABLE status within 72 hours of initial dis- ) covery of the second inoperable components (consistent with the 1 Completion Times of Required Actions C.4 or F.1). . 4 Additionally, the initial inoperable component must be restored to 0FERABLE status within 14 days of discovery of the initial inoperable component (consistent with the Completion Times of Required Actions A.2 or B.4). If one_of the inoperable components is restored to OPERABLE status within the required Completion Time, then Condition C or F, as appropriate, is reentered which allows operation for 72 hours from the time of initial loss of the second inoperable component. Similarly, if a second component is restored to OPERABLE status within the required Completion Time, then Condition A or B, as appropriate is reentered which allows operation for 14 days from the time of loss of the initial inoperable component. If one or the inoperable components is not restored within 24 hours, or if the second inoperable component is not restored within 72 hours, or if the initial i inoperable component is not restored within 14 days, whichever comes first, a controlled. shutdown must be initiated per Required Action 1.1 and 1.2. H.1. H.2. H.3. H.4 With all three diesel generators inoperable, no standby A.C. sources [other than the Alternate AC source] are available to feed the minimum required ESF functions. A loss of off-site , power event under this condition would result in a station blackout situation [if not for the alternate AC source] and therefore the level of degradation is severe. However, as with Action D, there is also risk associated with a forced shutdown. Therefore, continued operation is allowed, but only for a very : short time. ! 4 (continued) Ol l' ABWR B 3.8-10 5/31/89 I l l !
A.C. Sources - Operating B 3.8.1 BASES (continued) ACTIONS H.1. H.2. H.3. H.4 (Continued) (continued) With no diesel generators available, operation may continue for two hours, at which time one diesel generator must be restored to OPERABLE status. The second diesel generator must be restored to OPERABLE status within 72 hours of discovery of the second diesel generator (consistent with the Completion Time of Required Action D.4). Additionally, the third diesel generator must be restored to OPERABLE status within 14 days of the discovery of the initial inoperable diesel generator (consistent with the Completion Time of Required Action B.4). Conditions D and B, respectively as appropriate, are reentered as diesel generators are restored to OPERABLE status within the required Completion Times. If the diesel generators are not restored within the appropriate completion times, then a controlled shutdown must be initiated per Required Actions I..I and I.2. I.1. 1.2 If the Required Actions and associated Completion Times of Condition A, B, C, D, E, F, G, or H cannot be met, the unit must be in MODE 3 in 12 hours and MODE 4 in the following 24 O hours. These times allow for a controlled shutdown of the reactor without placing undue stress on plant operators or plant systems. The allowable out-of-service times are based on the general intent of Regulatory Guide 1.93 (reference 6), recognizing the fact that this Reg. Guide addresses specifically assumed hardware configurations and plant capabilities that existed at the time it was written and revised. The ABWR design not only has e third complete division of ESF equipment, with its own standby A.C. power source, but has improved ECCS capability at the divisional level. [ Additionally, an alternate AC power source (combustion turbine) will be included on-site, that can be manually connected to each ESF bus, such that it can act as essentially a replacement off-site source.] Therefore, the out-of-service times and levels of degradation have been adjusted relative to those described in Reg. Guide 1.93. Essentially, each level of degradation for ABWR has been equated with one less level of degradation in Reg. Guide 1.93. For instance, with only one A.C. power source inoperable,14 days are allowed to restore it to OPERABLE status. With two (continued) O ABWR B 3.8-11 5/31/89
i A.C. Sources - Operating .i ' B 3.8.1 1 BASES (continued) l ACTIONS 1.1. 1.2 (Continued) j (continued) 1 A.C. power sources inoperable, 72 hours is allowed to restore 4 at least one of the two sources to OPERABLE status, and so on. , With more than two A.C. power sources inoperable, more restric- i tive Completion Time are imposed to maintain an equivalent i level of A.C. power source reliability. j 4 Overall, the strategy is consistent with the intent of Reg. Guide 1.93, with an appropriate adjustment mode for the improved design capabilities of the ABWR. Operating experience indicates that the availability of a 3 typical off-site source is higher than that of a typical j standby A.C. supply (Ref. 6). Thus, if risk is evaluated in terms of availability, the risk associated with the loss of an off-site power source (the source with the higher availability) would appear to be more severe than the risk associated with the loss of a standby A.C. supply (the source with the lower availability). However, this apparent difference in severity is offset by maintainability considerations: that is, the time required to detect and restore an unavailable off-site source is generally much less than that required to detect and restore an unavailable standby A.C. supply. Based on these considera-tions, a general distinction between operating restrictions associated with the loss of an off-site source and those restrictions associated with the loss of a standby A.C. supply is not warranted. SURVEILLANCE The A.C. power sources are designed to permit inspection and REQUIREMENTS testing of all important areas and features, especially those which have a standby function. Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions). The Surveil-lance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9 (Ref. 2), 1.108 (Ref. 8), and 1.137 (Ref.
- 9) as addressed in the FSAR.
(continued) O ABWR B 3.8-12 5/31/89 i
A.C. Sources - Operating B 3.8.1 BASES (continued) SURVEILLANCE SR 3.8.1.1 REQUIREMENTS (continued) This surveillance provides on-line indication that the auto-matic load shedding and sequencing logic (LSSL) is OPERABLE and will perform as designed upon receipt of LOCA or bus undervoltage signals. The verification of the auto-test no-fault indication helps to ensure that the appropriate loads will be shed as designed and the sequencing logic will properly sequence and control the start signals to motor feeder breakers so as to prevent an overburden of any power source by automatic load application. SR 3.8.1.2 This surveillance requirement assures proper circuit continuity for the off-site A.C. power supply to distribution network and availability of off-site A.C. power. The breaker alignment verifies that each breaker is in its correct position to ensure distribution buses and loads are connected to their preferred power source. SR 3.8.1.3 ( This surveillance demonstrates that the automatic LSSL are (]/ OPERABLE and will perform as designed upon receipt of LOCA or bus undervoltage signals. The performance of manual tests for the listed test inputs ensures that the appropriate loads will be shed as designed and the sequencing logic will properly sequence and control the start signals to motor feeder breakers so as to prevent an overburden of any power source by automatic load application.
$h 3.8.1.4. SR 3.8.1.8 These surveillance help to ensure the availability of the standby power supply to mitigate design basis transients and accidents and maintain the unit in safe shutdown conditions.
The 13 second time requirement to achieve the specified voltaae and frequency supports the assumptions in the design basis LOCA analysis (Ref. 10). (continued) (3 ABWR B 3.8-13 5/31/89
A.C. Sources - Operating B 3.S.1 BASES (continued) SURVEILLANCE SR 3.8.1.4. LR 3.8.1.8 (continued) REQUIREMENTS (continued) To minimize the wear on moving parts which do not get lubri-cated during engine idle conditions, all diesel generator starts for this surveillance may be preceded by an engine prelube period. The diesel generator start in 13 seconds from standby (cold) conditions need only be performed once per 184 days. All other engine starts for this surveillance may be preceded by warmup procedures as recommended by the manufac-turer so that the mechanical stress and wear on the diesel engine are minimized. SR 3.8.1.5. SR 3.8.1.9 These surveillance demonstrate that the standby diesel generators are capable of synchronizing and accepting greater than or equal to the maximum expected accident loads within 60 seconds. This 60 second period allows for manual loading. The 60 minute run time for the diesel generator (required by Ref. 8, para. 2.c.(2)) is to stabilize the engine teroperature. This will ensure that cooling and lubrication are adequate for extended periods of operation. The 60 second diesel generator loading requirement shall only be performed once per 184 days. All other diesel engine runs for this surveillance may include gradual loading as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine are minimized (Ref. 11). Momentary transients due to changing bus loads do not invalidate this test. The load band is provided only to avoid routine overloading of the diesel generator. SR 3.8.1.6 This surveillance verifies that breakers and switching equip-ment are properly aligned to provide standby power to the associated ESF buses. SR 3.8.1.7 This surveillance verifies that sufficient air capacity at the proper pressure for a minimum of I engine start is available without the aid of the refill air compressor. The system design requirement provides for 5 engine crank cycles (Ref. 12). However, only I start is assumed for all safety analyses. Requiring an air pressure 2 [ ] psig provides adequate margin to the diesel generator engine start lockouts at [ ] psig. (continued) I O ABWR B 3.8-14 5/31/89 l _ _ - _ _-
ps- , , ! A.C. Sources - Operating. B 3.8.1
) ,. .
). /. BASES (continued)
SURVEILLANCE SR 3.8.1.10 REQUIREMENTS (continued) Transfer of the unit power supply from the normal circuit from off-site to the alternate circuit from off-site deinonstrates the OPERABILITY of the alternate circuit distribution network to feed the shutdown loads. SR 3.8.1.11 As required by Reference 8, paragraph 2.a.(2), each diesel generator is required to demonstrate proper operation.for the
. design basis accident loading sequence to ensure that voltage and frequency are maintained within the required limits. Under accident conditions, prior to connecting the diesel generators to their appropriate bus', all loads are shed except load center feeders and those motor control centers which feed Class IE loads (referred to as permanently connected loads). Upon reaching rated voltage.and frequency, the diesel generators are then connected to their respective bus. Loads are then sequen-tially connected to the bus by the automatic sequencing logic.
The sequencing logic controls tt.e permissive and starting signals to motor breakers so as to prevent an overburdened power source by automati' load application. Reference 4 provides a summary of the automatic loading of ESF buses.
- [O3 SR 3.8.1.12 Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed which, if excessive, might result in a trip' of the engine. This surveillance demonstrates the diesel generator load response characteristics and capabil-ity to reject the largest single load without exceeding pre-determined voltage and frequency limits. For Diesel Generator A, the largest single load is 585 kw (RCW pump), and for Diesel Generators B and C, 1607 kw (HPCF B and C pumps) (Ref. 5). As required by Reference 13, paragraph 6.3.l(3), the load rejection test is acceptable if the increase in the speed of the diesel does not exceed 75% of the difference between nominal speed and the overspeed trip setpoint, or 15 percent above nominal, whichever is lower. For Diesel Generator A, B and C this represents 64.S Hz, equivalent to 75% or 15%.
(continued) ( ( ABWR B 3.8-15 5/31/89 - = _ _ _ _ . _ _ _ __
n q l. A.C. Sources - Operating i B 3.8.1 , l MSLS- (continuedi SURVEILLANCE SR 3.8.1.13 REQUIREMENTS l (continued) This surveillance demonstrates the diesel generator capability ] to reject a full load without exceeding the predetermined ! voltage limits. The generator full load rejection may occur due to a system fault or ir. advertent breaker tripping. This l , surveillance verifies proper engine-generator load response { under the simulated test conditions. This test will simulate the total connected loads that the diesel generator will experience following a full load rejection and verify that the diesel generator will not trip upon loss of the load. SR 3.8.1.14 , i As required by Reference 8, paragraph 2.a.(1), this surveil-lance demonstrates the as-designed operation of the standby power sources during loss of the the off-site power source. The diesel generators must demonstrate their functional capability at full load temperature conditions within five minutes after a one hour continuous run or after normal ! operating temperature is achieved. This demonstrates that the l diesel engine can restart from a hot condition. This test verifies all actions encountered from the loss of off-site power to shedding of the non-essential loads and energization of the emergency buses from the diesel generators. It further demonstrates the capability of the diesel generator to auto-matically achieve the required voltage and frequency within the specified time. All engine starts for this surveillance may be preceded by warmup procedures as recommended by the manufac-turer so that the mechanical stress and wear on the diesel engine are minimized. SR 3.8.1.15 This surveillance demonstrates that the diesel generator automatically starts and achieves the required voltage and frequency within the specified time from the design basis actuation signals (LOCA signals) and operates for > 5 minutes. The five minute period provides sufficient time to demonstrate OPERABILITY and achieve stability. All engine starts for this surveillance may be preceded by warmup procedures as recom-mended by the manufacturer so that the mechanical stress and wear on the diesel engine are minimized. (continued) O ABWR B 3.8-16 5/31/89 m --_
A.C. Sources - Operating B 3.8.1 BASES (continued) SURVEILLANCE SR 3.8.1.16 REQUIREMENTS (continued) This surveillance demonstrates the diesel generator operation as discussed in the Bases for SR 3.8.1.14 during loss of off-site power in conjunction with an ECCS actuation signal. FR 3.8.1.17 t This surveillance demonstrates that diesel generator non-critical protective functions (e.g. high jacket water temper-ature) are bypassed on an ECCS actuation test signal and critical protective functions (engine overspeed and generator differential current (and low engine lube oil pressure]) trip the diesel generator to avert substantial damage to the diesel generator unit. The non-critical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. This provides the operator with sufficient time to react appropriately. The diesel generator availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immed'ately detrimental to emergency operation of the diesel generator. SR 3.8.1.18 Regulatory Guide 1.108 (Ref. 8, Para. 2.a.(3)) requires demon-stration once per 18 months that the diesel generators can start and achieve the proper voltage and frequency within the specified time and also run continuously with the proper voltage and frequency. This includes demonstrating full load capability for an interval of not less than 24 hours, of which 22 hours are at a load equivalent to the continuous rating of the diesel and 2 hours at a load equivalent to the 2 hour rating of the diesel. SR 3.8.1.19 This surveillance verifies that the total auto-connected loads do not exceed the diesel generator rating. Even though the diesel generator is rated to carry overload for short duration, long term overloading may damage the diesel generator. (continued) l O ABWR B 3.8-17 5/31/89
i A.C. Sources - Operating [ B 3.8.1 BASES (continued) SURVEILLANCE SR 3.8.1.20 REQUIREMENTS (continued) As required by Reference 8, paragraph 2.a.(6), this surveil-lance assures that the manual synchronization and automatic load transfer from the diesel generator to the off-site power source can be made and the diesel generator can be returned to standby status when off-site power is restored. It also ensures that the auto-start logic is reset to allow the diesel generator to reload if a subsequent loss of off-site power occurs. The diesel generator is considered to be in standby status when the diesel generator is shutdown (diesel speed is zero and output breaker open) and it can receive an auto-start signal and operate through the loading sequence. SR 3.8.1.21 Demonstration of the test mode override ensures that the diesel , generator availability under accident conditions will not be i compromised as the result of testing. Interlocks to the LOCA and loss-of-offsite power sensing circuits cause the diesel generator to automatically reset to standby operation if either of these signals is received during operation in the test mode. Standby operation is defined as the diesel generator running at rated conditions with the diesel generator output breaker open. These provisions for automatic switchover are required by Reference 14, paragraph 6.2.6(2). SR 3.8.1.22 This surveillance verifies that the diesel engine and its auxiliary systems are in a condition that meets the manufde-turer's recommendations. The inspection at 18 month intervals ; is sufficient to detect wear and tear on engine components, and determine if replacement, maintenance or adjustment is required. SR 3.8.1.23 This surveillance demonstrates that the diesel generator starting interdependence has not been compromised fol~owing 10 years of operation. (continued) 1 I i i t Oi i I ABWR B 3.8-18 5/31/89 l
[.1 A.C. Sources - Operating B 3.8.1 7.. BASES (continued)
-Q-SURVEILLANCE Surveillance Frequencies REQUIREMENTS (continued) In general surveillance frequences are based on engineering judgement and industry accepted practice considering the unit conditions required to perform the test, the ease of performing the test and the likelihood of a charge in system or component status.
Diesel Generator Test Schedule The diesel generator test schedule (Table 3.8.1-1) implements the recommendations of Generic' Letter 84-15 (Ref. 11). The purpose of this test schedule is to provide timely test data to establish a confidence level associated with the goal to maintain diesel generator reliability above 0.95 per demand. Per Generic Letter 84-15, each diesel generator unit should be tested at least once every 31 days. Whenever a diesel gener-ator has experienced two or more failures in the last 20 demand >., the time betveen tests is reduced to seven days. Two failures in 20 demands is a failure rate of 0.1, or the thre,- hold of acceptable diesel generator performance, and hence may be an early indication of degradation of the reliability of a (7 diesel generator. However, when considered in the light of a
.V long history of tests, two failures in the last 20 demands may only be a statistically probable distribution of two random events. Increasing the test frequency will allow for a more timely accumulation of additional test data upon which to base judgement of the reliability of the diesel generator.
Reports All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Administrative Control [ 5.9.x ] within 30 days. Reports of diesel generator failures shall include the information recommended in Regulatory Posi-tion C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests of any diesel generator is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. (continued) ABWR B 3.8-19 5/31/89
A.C.-Sources - Operating B 3.8.1 BASES (continued) REFERENCES 1. 10CFR50, General Design Criteria 17, " Electric Power Systems."
- 2. Regulatory Guide 1.9, " Selection, Design, and Qualification of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," March 1971.
- 3. ABWR SSAR, Section 8.3.1.1.8.
- 4. ABWR SSAR, Table 8.3-4.
- 5. ABWR SSAR, Tables B.3-1 to 8.3-3.
- 6. Regulatory Guide 1.93, " Availability of Electric Power Sources," December 1974.
- 7. [ Specific list of equipment covered by Required Actions B.2 and C.2].
- 8. Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," August 1977.
- 9. Regulatory Guide 1.137, " Fuel Oil Systems for Standby Diesel Generators," October 1979.
- 10. ABWR SSAR, Section 6.3.3, Table 6.3.1.
- 11. Generic Letter 84-15, " Proposed Staff Actions to improve and Maintain Diesel Generator Reliability," July 2,1984.
- 12. ABWR SSAR, Section 9.5.6.
- 13. IEEE Std. 387-1977, "IEEE Standard Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations."
- 14. IEEE Std. 308-1978, "IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations."
- 15. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.8-20 5/31/89 f l - - - - - - - -
1
]
A.C. Sources Shutdown B 3.8.2 ) 1 p 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 A.C. Sources - Shutdown BASES BACKGROUND A description of off-site and on-site A.C. power sources is j provided in the Bases for LCO 3.8.1.. APPLICABLE A reduced compliment of A.C. Sources in MODES 4 and 5, and when
, ~ SAFETY- handling irradiated fuel _in the secondary containment'is ANALYSES adequate to assure power for systems required to recover from an inadvertent draindown of the vessel or a fuel bundle handling accident (Ref. 1). :
A.C. Sources - Shutdown satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. LC0 One circuit between the off-site transmission network and-the Class IE distribution system and one diesel generator ensure
. N(~l /. the availability of the required power to recover from postulated events in MODES 4 and 5 and when handling irradiated fuel (e.g., fuel handling accident, vessel draindown). A description of what is required for OPERABILITY of +he circuit from off-site and the diesel generator is provided in the Bases for LC0 3.8.1.
APPLICABILITY The OPERABILITY of A.C. Sources in MODES 4 and 5, and when moving irradiated fuel assures (1) adequate coolant inventory for the irradiated fuel in the core in case of an inadvertent draindown of the vessel, (2) mitigation of a fuel bundle
' handling accident, (3) sufficient power for required support systems (e.g., decay heat removal, refueling activities, component cooling), and (4) sufficient instrumentation and control capability for monitoring and maintaining the. unit status.
(continued) O ABWR B 3.8-21 5/31/89
A.C. Sources - Shutdown 8 3.8.2 BASES (continued) ACTIONS. A.l. A.2. A.3. A.4 With less than the required A.C. sources OPERABLE, only a maximum of one diesel generator or one off-site circuit [and the alternate AC source] is available. Consequently, a single failure could result in no available A.C. power to recover from postulated. events in MODES 4 and 5, and when handling irradiated fuel. Therefore, the requirements are to take immediate actions to suspend CORE ALTERATIONS and handling of s irradiated fuel in the secondary containment. Suspension of ] these activities shall not preclude completion of the movement l of a comperent to a safe, conservative position. Activities which could potentially result in inadvertent draindown of the vessel (e.g., undervessel control rod drive replacement) must be suspended as soon as practicable. This should preclude the occurrence of the postulated events. SURVEILLANCE The Bases provided for SR 3.8.1.1 through SR 3.8.1.22 are REQUIREMENTS applicable. REFERENCES 1. ABWR SSAR, Section 15.7.4.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.8-22 5/31/89
f L Diesel . Fuel Oil - B 3.8.3' [: y p-Q .B 3.8 ELECTRICAL POWER SYSTEMS
'B 3.8.3 Diesel Fuel Oil BASES BACKGROUND For the purpose of this LCO, the diesel fuel oil subsystem is considered to include 1) fuel oil storage, 2) fuel oil transfer capabilities, and 3) fuel oil properties.
Each diesel generator is.provided with a storage tank having a fuel capacity sufficient to operate.that diesel. for a period of 7 days while the diesel generator is supplying maximum post-LOCA load demand (Ref.1). The maximum load demand is-calculated using the sssumption that a minimum.of any two
. diesel generators is available. This on-site fuel capacity
' lasts longer than the time it would take to replenish the on-site supply from outside sources.
A[ ] gallon day tank is provided for each diesel. Each day tank is housed in'a separate room in the diesel generator i building and has a fuel capacity for approximately 8 hours of full-load operations'(Ref. 1). Fuel oil is transferred from storage tanks to the day tanks by either of two pumps per
/3 division located at the storage tanks. Redundancy of pumps and
\j piping precludes the failure of one pump or the rupture of any pipe, valve, or tank to result in the loss of more than one diesel generator. In the event that the piping between the last isolation valve and the day tank breaks, the use of one diesel can be lost. This occurs only after the 8 hour supply 7
of fuel in the day tank has been used. All of the outside tanks, pumps, and piping are underground. During operation of the diese1' generators, fuel oil pumps driven by the diesel engines transfer fuel from the day tanks to the diesel engine fuel manifolds. Level controls mounted on the day tanks automatically start and stop the storage tank transfer pumps. The level of the fuel supply in each tank .is indicated in the control room. In addition, alarms both locally and in the control room annunciate low level and high level in any day tank. In the unlikely event of a failure in one of the supply trains, the associated day tank low-level alarm annunciated when the fuel oil remaining in the tank provides one hour of full-load operation, thus allowing the operator to take corrective action to prevent the loss of the diesel. O O (continued) ABWR B 3.8-23 5/31/89
Diesel Fuel Oil B 3.8.3 ! BASES (continued) BACKGROUND For proper operation of the standby diesel generators, it is (continued) necessary to ensure the proper quality of the fuel oil. Reference 3 addresses the recommended fuel oil practices as supplemented by Reference 2, Section c.2. The fuel oil properties governed by these surveillance requirements are the water and sediment content, the kinematic viscosity, specific gravity (or API gravity), and impurity level. APPLICABLE The initial conditions of design basis transient and accident SAFETY analyses in the SSAR Chapters 6, Engineered Safety Features, ANALYSES and 15, Accident Analyses, assume all ESF systems are OPERABLE. The diesel fuel oil subsystem provides the necessary fuel supply to support operation of the diesel generators. Diesel Fuel Oil satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvement as documented in Reference 4. LC0 The diesel fuel oil subsystem is required to be OPERABLE to , ensure availability of the required A.C. power to shutdown the ! reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated design basis ' accident. This reouires that the associated day tank and storage tank for each diesel generator be OPERABLE and that the fuel oil properties be within acceptable limits. APPLICABILITY The diesel fuel oil subsystem is required to be OPERABLE when the associated diesel generator is required to be OPERABLE (Refer to LC0 3.8.1 and LC0 3.8.2). ACTIONS !L1 With the fuel oil level in the day tank low, insufficient fuel oil is available to satisfy the design basis requirement of supporting I hour of continuous operation. A limited time is provided to restore the fuel oil level via the fuel transfer pump or the diesel generator must be declared inoperable and the Required Actions of LCO 3.8.1 or LC0 3.8.2, as applicable, initiated. (continued) B 3.8-24 5/31/89 4 ABWR
4 Diesel Fuel Oil B 3.8.3 () BASES (continued) ACTIONS L1 (continued)' With the fuel oil transfer capability inoperable, sufficient fuel oil is available in the day tank to support 8 hours of continuous diesel generator operation. If the transfer capa-bility cannot be restored within this interval, the diesel generator must be declared inoperable. L1 Low level in one or more fuel storage tanks indicates that the-design basis requirement.of supporting 7 days of continuous
< operation may not be able to be satisfied. A period of 24-hours is provided to restore'the fuel oil level or the diesel generator must be declared inoperable. This 24 hour period is acceptable due to the fuel oil transfer capability between storage tanks.
D.1. E.1 With the fuel oil properties defined in the Bases for SR 3.8.3.5 and SR 3.8.3.6 not within the required limits, a period of 72 hours is allowed to restore the fuel oil proper-ties. This provides sufficient time to retest the fuel oil
.O sample to confirm that the. initial test results were valid.
Even if a diesel generator start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the diesel generator would still be capable of performing its intended function. For'the accelerated stability testing portion of the diesel fuel oil testing program, a period of 30 days is provided to restore the fuel oil properties to within limits of ANSI- N195. This test measures stability of distillate fuels under acceler-ated oxidizing conditions. It is performed in an extreme testing environment and results only in- an index of long term storage stability. If all other tested parameters are within limits of ASTM-0975, fuel oil exceeding the insolubility limits of of ANSI-N195 Appendix B will remain stable for a period ranging from months to years. Ll With the Required Actions and associated Completion Times of Conditions A, B, C, D, or E not met, the associated diesel n generator must be declared inoperable and the Actions of LC0 3.8.1 or LC0 3.8.2 followed.
' (continued)
ABWR B 3.8-25 5/31/89 l L 1 _ _ _ __o
L I 1 l Diesel Fuel Oil B 3.8.3 h BASES (continued) SURVEILLANCE SR 3.8.3.1 ; REQUIREMENTS l The verification of each diesel engine f91 day tank supply l ensures that enough fuel is on hand at the engine to sustain at l least 8 hours of full load operation for the diesel generator ' (Ref. 1). SR 3.8.3.2 i The verification of each diesel engine fuel storage tank supply demonstrates that sufficient fuel is available to sustain at least 7 days of full load operation for the diesel generator f (Ref. 1). This time period is sufficient to place the unit in { a safe shutdown condition and provides enough time to bring in ! replenishment fuel from an off-site location. l I SR 3.8.3.3 This surveillance demonstrates that each required fuel oil i transfer pump operates and transfers fuel from its associated ! storage tank to its associated day tank. This is required to q support the 7 day continuous operation of the standby powar j sources. This surveillance provides assurance that the fuel transfer pump is OPERABLE, there is no crud buildup blocking the fuel supply line and that the fuel supply line is intact. SR 3.8.3.4. SR 3.8.3.8 94 These surveillance remove, if necessary, water accumulation in the fuel oil supply due to excessive condensation or water in-leakage in the tanks and supply system. It is only required if analysis shows that there is water in the fuel oil tanks. This is necessary to ensure that the fuel oil quality is within manufacturer's recommended standards. The surveillance frequencies are established by Reference 2, paragraph 2.e. SR 3.8.3.5. SR 3.8.3.6. SR 3.8.3.7 Periodically and upon replenishment of the fuel oil, these surveillance are performed to establish that the diesel engine fuel meets the NRC guidelines on kinematic viscosity, water and sediment content, specific gravity and sediment impurity (Ref. 2). The samples must be obtained in accordance with ASTM-D270 (Ref. 2, para. c.2.c) or other applicable ASTM method. The water and sediment content and kinematic viscosity must be determined by testing in accordance with ASTM-D975 (Ref. 2, para. c.2.a) and the impurity level must be determined in accordance with ASTM-D2274 or other applicable ASTM method.. The required surveillance frequencies for sampling the fuel oil are established by Reference 2, paragraph 2.b. (continued) ABWR B 3.8-26 5/31/89
~ _. Diesel Fuel Oil B 3.8.3 ! BASES (continued) SURVEILLANCE SR 3.8.3.5. SR 3.8.3.6. SR 3.8.3.7 (continued) REQUIREMENTS 1 (continued) Typical fuel oil sample requirements are: A. Water and sediment content less than or equal to the value ; listed in ASTM-D975, Table 1. B. Kinematic viscosity specified on ASTM-D975, Table 1. C. API Gravity within 0.3*F at 60*F or a specific gravity of within 0.0016 at 60/60*F. D. Impurity level < 2 mg of insolubles per 100 ml. The other properties specified in Table 1 of ASTM-D975 and Regulatory Guide 1.137, Revision 1, October 1979, Position 2.a should be tested in accordance with ASTM-D975. Failure to meet this requirement shall not affect diesel generator OPERABILITY. SR 3.8.3.9 The draining of the fuel oil stored in the supply tanks, removal of accumulated sediment and tank cleaning is required at 10 year interval.s (Ref. 2, para. 2.f). To preclude the introduction of surfactants in the fuel system, the cleaning
.9 should be accomplished using sodium hypochlorite solutions ec their equivalent rather than soap or detergents.
REFERENCES 1. ABWR SSAR, Section 9.5.4.
- 2. Regulatory Guide 1.137, " Fuel Oil Systems for Standby Diesel Generators," October 1979.
- 3. ANSI N195-1976, " Fuel Oil Systems for Standby Diesel Generators," Appendix B.
- 4. NEDD-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.8-27 5/31/89
D.C. Sources - Operating B 3.8.4 j i 3.8 ELECTRICAL POWER SYSTEMS B 3.8.4 D.C. Sources - Operatino ) BASES l BACKGROUND The station D.C. power system provides the A.C. power system with required control power. It also provides both motive and control power to selected safety-related equipment. The D.C. j subsystems conform to the requirements of References 1 and 2. The 125-volt D.C. power systems consist of four independent i Class lE D.C. power subsystems (Divisions 1, 2, 3 and 4). Each division consists of a battery bank and its associated battery charger (s). Each of the Divisions 1, 2, and 3 D.C. power subsystems provides the control power for its associated Class IE A.C. power load group, 6.9 kV switchgear and 480-volt load centers. They also provide D.C. power for ESF valve actuation and D.C. control power for the diesel generator in their i respective divisions. Each divisional subsystem (including Division 4) also provides D.C. power for plant instrumentation, logic, alarm and indication circuits. Each 125-volt D.C. battery is separately housed in a ventilated room apart from its charger (s) and distribution center. Each subsystem is located in an area separated physically and electrically from other subsystems to ensure that a single . failure in one train does not cause a failure in the redundant train. There is no sharing between redundant Class IE trains of equipment such as batteries, battery chargers, or distribution panels. Batteries have sufficient stored energy to operate connected essential loads continuously for at least 2 hours withont recharging (Ref. 3). All batteries are sized to produce required capacity at 80 percent of nameplate rating, corresponding to warranted capacity at end-of-life-cycles and the 100 percent design demand. The voltage design limit is 1.75 volts per cell (Ref. 3) . Each battery charger has enough power-output capacity for the steady-state operation of connected loads required during normal operation while maintaining its battery in a fully charged stated. Each battery charger has sufficient capacity to restore the battery from the design minimum charge to its fully charged state in 10 hours while supplying normal steady-state loads. Two redundant battery chargers, each of which serves as a back-up to two normal divisional chargers, are provided for the four 125-volt D.C. battery banks. (continued) O ABWR B 3.8-28 5/31/89
r - i
,- i I l;
D.C. Sources - Operating _ B 3.8.4 BASES (continued) f] V ~ ~
-APPLICABLE The initial conditions of design basis transient and accident SAFETY analyses-in SSAR Chapter 6, Engineering Safety Features, and )
ANALYSES :15, . Accident Analyses, assume all ESF systems are OPERABLE.
-The D.C. power systems provide normal and emergency D.C. power for emergency auxiliaries and for control and switching during all modes of operation. The OPERABILITY of the power sources ' is consistent with the initial assumptions of the accident analyses and is based upon D.C. power sources and associated l distribution systems being OPERABLE during accident conditions with,' (1) an assumed loss of off-site power, and (2) a single-failure of.one of the standby A.C. sources.
D.C. Sources - Operating satisfies the requirements of Selec-
. tion Criterion 3 of the NRC Interim Policy Statement on Tech-nical Specification Improvements as documented in Reference 8.
- (
LCO Divisions 1, 2, 3, and 4 D.C. electrical power sources are required to be OPERABLE to ensure availability of the required-power _to shutdown the reactor and maintain it in a safe condi-tion after an anticipated operational occurrence or a postu-lated design basis accident. Loss of any one of the D.C. power Jubsystems does not prevent the minimum safety-function from fl being performed. The D.C. electrical power sources are con-O sidered OPERABLE if the 125-volt batteries and at least one battery charger per battery satisfy the applicable surveillance requirements. The battery electrolyte requirements are located
in LCO 3.8.6.
APPLICABILITY Divisions 1, 2, 3, and 4 of the D.C. power sources are required to be OPERABLE in MODES 1, 2, and 3 to ensure that:
- 1. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of antici-pated operational' occurrences or abnormal transients, and
- 2. Adequate core cooling is provided, and containment integ-rity and other vital functions are maintained in the event i of a postulated DBA.
Divisional D.C. power requirements for MODES 4 and 5 are covered in the bases for LC0 3.8.5. (continued) i
-x ABWR B 3.8-29 5/31/89 i
D.C. Sources - Operating B 3.8.4 BASES (continued) ACTIONS A.1 With the Division 4 D.C. power subsystem inoperable (e.g. inoperable battery or inoperable required battery charger), the available D.C. power sources have the capability to support a safe shutdown and to mitigate an accident condition, even with an additional single failure. However, since the overall redundancy is reduced (specifically with regards to plant instrumentation and actuation and control logic) continued power operation should not exceed 30 days. If the affected Division 4 D.C. supply is not restored within these 30 days, a controlled shutdown.must be initiated per Required Actions D.1 and 0.2. B.1 With one of the Division 1, 2 or 3 D.C. power subsystems inoperable (e.g., inoperable battery or inoperable required battery charger), the available D.C. power sources have the capacity to support a safe shutdown and to mitigate an accident condition. However. since a subsequent worst case single failure could result in only one divisional power source being OPERABLE, continued power operation should not exceed 7 days. The Division 4 D.C. power subsystem may be available; however, it does not support directly the operation of any of the'3 FSF load group divisions. Therefore, if the affected D.C. supply is not restored within these 7 days, a controlled shutdown must be initiated per Required Actions D.1 and D.2. C.l. C.2 With any two of the Division 1, 2 or 3 D.C. power subsystems inoperable, the available D.C. power source (s) has the capabil-ity to support a safe shutdown and to mitigate an accident condition. However, since the overall D.C. power system redundancy is severely reduced, continued operation should not exceed 24 hours, at which time at least one of the inoperable Division 1, 2 or 3 D.C. power subsystems should be restored to OPERABLE status. Additionally, the initial inoperable Divi-sion 1, 2 or 3 D.C. power subsystem should be restored to OPERABLE status within 7 days of the time of discovery (consis-tent with Completion Times of Required Action B.1). If either of the two inoperable Division 1, 2 or 3 D.C. source is not restored within the applicable required Completion Time, then a controlled shutdown must be initiated per Required Actions D.1 and D.2. (continued) O ABWR B 3.8-30 5/31/89
s D.C. S'ources - Operating- i B'3.8.4 I BASES (continued) u l ACTIONS __ D.1~. D.2
- (Continued) -
If at any time a Required Action and associated Completion Time-of. Condition A, B, or C is not met, then'a controlled shutdown
~
must be initiated to place the plant in a condition where a lesser degree of D.C. power. system redundancy is justified. SURVEILLANCE Thefsurveillance requirements for demonstrating the OPERABILITY- _ REQUIREMENT S' of the unit batteries are in accordance with the recommendations of References 5 and 6. j SR 3.8.4.1. SR 3.8.4.2. SR 3.8.4.3 __SR 3.8.4.4. SR 3.8.4.5-Verifying battery terminal float voltage'and performing visual inspections of the battery cells, connections or the resistance measurement of each cell and terminal connection provides an indication of physical damage or abnormal deterioration which could potentially~ degrade battery performance. SR 3.8.4.6
-q Paragraph C.I.b of Reference 7 requires that the battery V charger supply be based on the largest combined demands of the various steady-state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the' unit during which these demands occur. The minimum required amperes and duration for the battery chargers ensure that the battery. load requirements can be satisfied (Refer to SR 3.8.4.7).
SR 3.8.4.7 Paragraph C,1.c of Reference 7 requires the performance of a battery servie.e test in accordance with Reference 6 at inter-vals not to exceed 18 months. A battery service test is a special capacity test to demonstrate the capability of the battery to meet the system design requirements. (continued) , l L ABWR B 3.8-31 5/31/89
L D.C. Sources - Operating B 3.8.4 I BASES (continued) SURVEILLANCE SR 3.8.'4.8 REQUIREMENTS (continued) Paragraph '.~'J) of Reference 6 recommends a performance test for e- 1 ba'.tery at 60 month intervals. A battery performance
% is a capacity test of the battery in the "as found" condition, after being in service, to detect any change in the capacity as determined by the new battery acceptance test. In accordance with Paragraph 6 of. Reference 6, the battery should be replaced if its capacity is below 80 percent of the manufac-turer's rating. A capacity of 80 percent shows the battery rate of deterioration is increasing even if there is ample capacity to meet the load requirements. Paragraph 4.2(3) of Reference 6 recommends annual performance tests of battery capacity for any battery that shows signs of degradation or has reached 85 percent of the service life expected of the applica-tion. Degradation is indicated when the battery capacity drops more than 10 percent of rated capacity from its average on previous performance tests, or is below 90 percent of the manufacturer's rating. However, this test cannot be performed at other than shutdown conditions. Therefore, an 18 month interval is appropriate.
Surveillance Frequencies Except where noted above, surveillance frequencies are based on engineering judgement and industry accepted practice considering the unit conditions required to perform the test, the ease of performing the test and the likelihood of a change in the system component status. (continued) l l
)
i ABWR B 3.8-32 5/31/89 I
D.C. Sources - Operating B 3.8.4
.- BASES (continued) i REFERENCES 1. Regulatory Guide 1.6, " Independence Between Redundant 4 l
Standby (On-site) Power Sources and Between Their Distribution Systems," March 10, 1971.
- 2. IEEE Standard 308-1974, "IEEE- Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations."
- 3. ABWR SSAR, Section 8.3.2.
- 4. Regulatory Guide 1.93, " Availability of Electric Power Sources," December 1974.
- 5. Regulatory Guide 1.129, " Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978.
- 6. IEEE Standard 450-1980, "IEEE Recommended Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Subsystems."
- 7. Regulatory Guide 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants," February 1977.
- 8. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
~
O ABWR B 3.8-33 5/31/89
D.C. Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 D.C. Sources - Shutdown BASES BACKGROUND A description of the D.C. power sources is provided in the Bases for LCO 3.8.4. APPLICABLE A reduced compliment of D.C. Sources in MODES 4 and 5, and when SAFETY handling irradiated fuel in the secondary containment is ade-ANALYSES quate to assure sufficient instrumentation and control capa-bility is available for monitoring and maintaining the unit status, and to assure power for systems and control logic required to recover from an inadvertent draindown of the vessel or a fuel bundle handling accident (Ref. 1). D.C. Sources - Shutdown satisfies the requirements of Selection ! Criterion 3 of the NRC Interim Policy Statement on Technical ' Specification Improvements as documented in Reference 2.
~
1 LC0 At least one OPERABLE Division 1, 2 or 3 of the D.C. electrical power sources ensures the availability of the required power to recover from postulated events in MODES 4 and 5, and when handling irradiated fuel (e.g., fuel handling accident, vessel draindown). A D.C. electrical power source is considered OPERABLE if the 125-volt battery and at least one battery charger per battery are OPERABLE. The battery electrolyte requirements are located in LC0 3.8.6. APPLICABILITY The OPERABILITY of D.C. Sources in MODES 4 and 5, and when handling irradiated fuel is required to assure (1) adequate coolant inventory for the irradiated fuel in the core in case of an inadvertent draindown event, (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status, and (3) sufficient decay heat removal. (continued) l l ABWR B 3.8-34 5/31/89
'D.C. Sources - Shutdown B 3.8.5
- p. ,
BASES (continued) f3 -
*t y'-
ACTIONS A.1. A.2. A.3. A 4 With less than of the required D.C. power system'0PERABLE (e.g. the battery or required charger inoperable) adequate D.C. power is not available to support' operation of the ECCS and-decay' heat removal equipment, uninterruptable power supplies and: switchgear required to recover from a fuel handling accident or vessel draindown event. Consequently, it is required to,take immediate actions to su: pend CORE ALTERATIONS and handling of irradiated fuel,in the primary and secondary containment. Suspension of these activities shall not preclude completion of
- - the movement of a component to a safe, conservative position.
Activities which could potentially result in an inadvertent draindown of the vessel (e.g.. undervessel control rod drive replacement) must be suspended as soon as practicable. This should preclude the occurrence of the postulated events.
. SURVEILLANCE The Bases provided for SR 3.8.4.1 through SR 3.8.4.8 are REQUIREMENTS applicable.
/ REFERENCES -1. .ABWR SSAR, Section 15.7.4.
1.]) 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987. O ABWR B 3.8-35 5/31/89
Battery Electrolyte B 3.8.6 8 3.8 ELECTRICAL POWER SYSTEMS B 3.8.6 Battery Electrolyte BASES . 1 BACKGROUND Table 3.8.6-1 specifies the limits for each designated. pilot tell (Category A) and each cornected cell (Category B) for electrolyte level, float voltaga and specific gravity (sg). The limits for the designated pilot cell's float voltage and specific gravity (greater than or equal to 2.13 volts and 0.015 (sg) or a battery charger current that had stabilized at a low value) are based upon the manufacturer's recommended values and are characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity (greater than 2.13 volts and 1.190 (sg) and an average specific gravity of all the connected cells greater than or equal to 1.200 (sg) are the values recommended by the manufacturer to ensure the OPERABILITY and capability M the l battery. i The specific gravity may be corrected for electrolyte tempera- i ture and level. However, if the charging current is less .than 2 amperes and the specific gravity readings are equal to or greater than the manufacturer's recommendation, it is not necessary to correct for electrolyte level. O APPLICABLE The initial conditions of design basis transient and accident , SAFETY analyses in SSAR Chapter 6, Engineering Safety Features, and l ANALYSES 15, Accident Analyses, assume all ESF systems are OPERABLE. The D.C. power systems provide normal and emergency 0.C. power for emergency auxiliaries and for control and switching during all modes of operation. The OPERABILITY of the power sources is consistent with the initial assumptions of the accident analyses and are based upon D.C. power sources and associated distribution systems OPERABLE during accident conditions, with (1) an assumed loss of off-site power, and (2) a single failure of one of the other standby A.C. sources. Battery Electrolyte satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. I ~ (continued) O ABWR B 3.8-36 5/31/89 . 1
.J i
a
. Battery Electrolyte B 3.8.6 n '
BASES (continued) U LCOs The D.C. power system is required to be OPERABLE to ensure availability of the required power to shutdown the reactor and maintain it in a safe condition after an anticipated opera-tional occurrence or a postulated design basis accident. Battery electrolyte meeting the limits of Categories'A and B. l and the allowable values of Category C helps ensure the OPERABILITY of the D.C. sources. APPLICABILITY The battery electrolyte must be within the limits of Table 3.8.7-1 when the associated D.C. power sources are required to be OPERABLE (Refer to Applicability discussion in Bases for LC0 3.8.4 and LCO 3.8.5).
?
ACTIONS A.1. A.2. A.3 Operation with one or more cells in one or more batteries parameters' outside the normal limit (i.e., Category A limits not met or Category- B limits not met or Category A and B limits not met) but within:the allowable value (Category C) specified in Table 3.8.6-1 is permitted since sufficient capacity exists f~ to perform the intended function and maintain the margin of
- ( -
safety. The pilot cells electrolyte level and float voltage are required to be demonstrated to meet the Category C allowable values within I hour. This is because this check will provide a gross indication of the status of the remainder of the battery's cells since the pilot cells are typically the worse cells. The requirement to demonstrate the Category C allowable values are not met within 24 hours will ensure that during the time to restore the parameter to the Category A and B limits that the battery will still be capable of performing
)
its intended function. However, since the battery capacity is reduced, continued operation is only permitted for 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, is the manufacturer's recommended specific gravity, which ensures that the decrease in capacity will be less than the safety margin provided in sizing; (3) the allowable value for an individual-cell's specific gravity is the value recommended by the manufacturer to ensure overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage (greater than [ 2.07 ] volts) ensures the battery's capability to perform its design function. L 1 I d (continue ) ABWR B 3.8-37 5/31/89 L L
Battery Electrolyte B 3.8.6 BASES (continued) ACTIONS A.I. A.2. A.3 (continued) (continued) When any battery parameter is outside the Category C allowable value, sufficient capacity to supply the maximum expected load requirements is not assured and the associated D.C. power source must be declared inoperable and the applicable Actions l of LC0 3.8.4 and LCO 3.8.5 must be followed. U ! If the appropriate parameters in Table 3.8.6-1 cannot be met per the Required Actions and associated Completion Times of Conditions A.1, A.2 or A.3, the associate'd battery must'be i declare inoperable and the Required Actions of LC0 3.8.4 or LC0 3.8.5 followed as appropriate. The battery must also be declared inoperabic if the average electrolyte temperature of the pilot cells is 1 60*F. Below this temperature, the battery's capability may not be sufficient to satisfy the design basis load profile. l l SURVEILLANCE SR 3.8.6.1 REQUIREMENTS Paragraph 3.3.1 of Reference 1 requires that the battery be demonstrated to meet Category A limits on a regularly scheduled - interval. The 7 day frequency for this surveillance is consistent with the guidance of paragraph 7.4.1 of Reference 2. SR 3.8.6.2. SR 3,8.6.3 Paragraph 3.3.2 of Reference 1 requires that the battery be demonstrated to meet Category B limits and the average electrolyte temperature on a quarterly basis. In addition, within 31 days of a battery discharge below 110 volts or a battery overcharge above 150 volts, the battery must be demonstrated to meet Category B limits and the average electrolyte temperature limit. This one-time 31 day surveillance verifies that no significant degradation of the battery occurred as a consequence of the discharge or overcharge. The 31 day requirement is judged to be acceptable given the 7 day periodic check of the Category A limits. l (continued) 1
}
ABWR B 3.8-38 5/31/89
,c]
Battery Electrolyte B 3.8.6 c l : r~T BASES.(continued)' REl'ERENCES 1. IEEE Std. 450-1975, "IEEE Recommended Practice'for S-Maintenance' Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
- 2. IEEE Std. 308-1978, "!EEE Standard Criteria for. Class IE Power Systems for Nuclear Power Generating Stations."
- 3. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
~i O
ABWR B 3.8-39 5/31/89
ll A.C. Power Dist. Systems - Operating B 3.8.7 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.7. A.C. Power Distribution Systems - Ooeratina BASES BACKGROUND The A.C. on-site power system is comprised of three independent 6.9 kV ESF distribution systems with their 480-volt load centers and motor control centers (MCC),120-volt vital A.C. power system and the standby power supplies (diesel generator units). Each 6.9 kV ESF bus is normally connected to a pre- , ferred source which is one of the two Auxiliary transformers which are being supplied by the main generator if it is on-line or by the main distribution grid through the main transformer if the generator output breaker is open. Upon loss of the preferred off-site power source to the 6.9 kV ESF buses, a j transfer to the on-site emergency power sources is accomplished j by utilizing a time delayed bus undervoltage relay. Transfer to the alternate off-site power source is by manual action only. APPLICABLE The initial conditions of design basis transient and accident SAFETY analyses in FSAR Chapters 6, Engineering Safety Features, and ANALYSES 15, Accident Analyses, assume all ESF systems are OPERABLE. The electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy and relia-bility to ensure the availability of necessary power to ESF systems so that the fuel, reactor pressure vessel and contain-ment design limits are not exceeded. These limits are dis-cussed in more detail in the Bases for LC0 sections 3.2 (Power Distribution Limits) and 3.6 (Containment Systems). The energization of the A.C. power distribution buses is consistent with the initial assumptions of the accident analyses and are based upon maintaining at least one division of the on-site A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions while accounting for, (1) an assumed loss of off-site power, and (2) a single failure of one of the other standby A.C. sources. A.C. Power Distribution Systems - Operating satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 1. (continued) l O ABWR B 3.8-40 5/31/89
e , . t' A.C. Power Dist. Systems - Operating B 3.8.7 y BASES (continued)
~( "/.' LCO The A.C. power distribution system divisions listed in Table B 3.8.8-1 ensure the availability of A.C. power for the systems required to shutdown the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated design basis acc adent.
APPLICABILITY The A.C. power distribution systems are' required to be energized in MODES 1, 2, and 3 to ensure that:
- 1. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of anticipated operational occurrences or abnormal transients, and
- 2. Adequate core cooling is provided, and containment integrity and other vital functions are maintained in the event of a postulated DBA.
A.C. power distribution system requirements for MODES 4 and 5 are covered in the bases for LCO 3.8.10. ACTIONS A.1. A.2. D.1. D.2 With one division of the A.C. power distribution system not energized, the remaining two OPERABLE A.C. power distribution systems are capable of supporting the minimum safety functions necessary to shutdown the unit and maintain it in the safe shutdown condition, even with an additional single failure. Therefore, continued operation is allowed for a short time. However, since the overall redundancy is reduced, the inop-erable A.C. power distribution system buses must be reenergized within 24 hours, or a controlled shutdown must be initiated per Required Actions D.1 and D.2. An alternate approach is to immediately declare inoperable all equipment associated with the deenergized power distribution system buses and enter the ACTIONS of the corresponding LCOs. B.1. B.2. D.1. D.2 With two of three divisions of the A.C. power distribution system not energized, the remaining OPERABLE A.C. power distri-bution system is capable of supporting the minimum safety functions necessary to shutdown the unit and maintain it in the safe shutdown condition. However, an additional single failure could result in insufficient power for the minimum required safety functions. Therefore, at least one inoperable A.C. (continued) ABWR B 3.8-41 5/31/89
A.C. Power Dist. Systems - Operating B 3.8.7
' BASES (c.* inued) v
- ACTIONS 8.1. S.2.1 B.2.2. D.1. D.2 (Continued)
O ! (continued) l power distribution system buses must be reenergized within 8 ! hours, or a controlled shutdown must be initiated per Required Actions C.1 and C.2. An alternate approach is to immediately declare inoperable all equipment associated with the deener-gized power distribution system buses and enter the ACTIONS of the corresponding LCOs. C.1. C.2. D,1. D.2 With all three of the required A.C. power distribution Divisions deenergized, insufficient power for the minimum required safety functions is available to recover from a design basis accident or transient. Since the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown, con-tinued operation is allowed for two hours or a controlled shutdown must be initiated per Required Actions D.1 and D.2. An alternate approach is to immediately declare inoperable all equipment associated with the deenergized power distribution buses and enter the ACTIONS of the corresponding LCO. Completion Times The allowable out-of-service times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Actions. SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This surveillance demonstrates that the A.C. electrical power distribution system is functioning properly with all the desired circuit breakers closed and the buses energized. The verification of proper voltage availability on the divisional buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. Surveillance Frequencies In general, surveillance frequencies are based on engineering judgement and industry-accepted practice considering the unit conditions required to perform the test, the ease of performing the test and the likelihood of a change in the system component i status. ] ABWR B 3.8-42 5/31/89 l _ _ .___-___---_O
A.C. Power Dist. Systems - Operating B 3.8.7 t'~ ' ; ' BASES (continued)
\' )
~ '
REFERENEES 1. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987. s / l ABWR B 3.8-43 5/31/89
3 A.C. Power Dist. Systems -_ Operating h B 3.8.7 ) ! Table B 3.8.8-1 (Page 1 of 1) Power Distribution System Divisions VOLTAGE LEVEL DIVISION 1 DIVISION 2 DIVISION 3 6900-volt 480-volt 120-volt O O ABWR B 3.8-44 5/31/89
D.C. Power Dist. Systems - Operating B 3.8.8 B 3.8 ELECTP.1 CAL POWER SYSTEMS \ B 3.8.8 LC. Power Distribution Systems - Operatjng i BASES BACKGROUND' The 125-volt D.C. power system is comprised of four separate and independent 125-volt D.C. distribution buses. Each of these Division 1, 2, 3 and 4125-volts D.C. distribution buses is fed from its dedicated battery. Division 1 battery is rated at 4000 anpere-hours (AH), Division 2 AT 3000 AH, Division 3 at 3000 AH and Division 4 is rated at 1400 AH. The 125-volt D.C. distribution bus is fed from a normal battery charger dedicated for its associated divisional battery. A second battery charger (standby charger) will provide power to the 125-volt D.C. bus upon failure of the normal charger. Both the normal and the standby chargers are supplied by the 480-volt ESF load 3 center buses. Division 1 normal charger is fed from Division 1 480-volt ESF load center, Divisior 2 normal charger from Division 2 ESF load center, Division 3 normal charger from Division 3 ESF load center and Division 4 normal charger from Division 1 ESF load center bus. Two standby chargers are provided, one feeding either Division 1 or Division 2125-volt D.C. distribution center and the other one feeding Division 3 or Division 4125-volt D.C. distribution center. Both of these standby chargers can be fed from either of two 480-volt ESF 9 load centers. The standby charger shared by Divisions 1 and 2 D.C. subsystems can be fed from either Division 1 or Division 3 480-volt ESF load center bus. APPLICABLE The initial conditioic of design basis transient and accident SAFETY analyses in FSAR Chapters 6, Engineering Safety Features, and ANALYSES 15, Accident Analyses, assume all ESF systems are OPERABLE. The D.C. electrical power distrit.ution systems are designed to provide sufficient capacity, capi.bility, redundancy and reliability to ensure the availability of necessary power to ESF systems so that the fuel, reactor pressure vessel and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for LC0 sections 3.2 (Power Distribution Limits) and 3.6 (Containment Systems). Energization of the D C. power distribution systems is con-sistent with the initial assumptions of the accident analyses and is based upon maintaining at least one division of the on-site A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions, while accounting for (1) an assun'ed loss of off-site power, and (2) a single failure of one o' the other standby A.C. sources. O (continued) ABWR B 3.8-45 5/31/89
D.C. Power Dist. Systems - Operating B 3.8.8 BASES (continued) APPLICABLE D.C. Power Distribution Systems - Operating satisfies the SAFETY requirements of Selection Criterion 3 of the NRC Interim Policy 1 ANALYSES Statement on Technical Specification Improvements as documented (continued) in Reference 1. LC0 The D.C. power distribution system divisions listed in Table B 3.8.8-1 ensure the capability to shutdown the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated design basis accident. APPLICABILITY The D.C. power distribution systems are required to be OPERABLE in MODES 1., 2, and 3 to ensure that:
- 1. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of anticipated operational occurrences or abnormal transients, and
- 2. Adequate core cooling is provided, and containment integrity and other vital functions are maintained in the event of a postulated DBA.
D.C. Power distribution system requirements for MODES 4 and 5 are covered in the bases for LC0 3.8.9. ACTIONS A.l. B.1. B.2.1. B.2.2. C.l. C.2 With one division of the D.C. power distribution system deenergized, the remaining energized D.C. power distribution divisions are adequate to support a safe shutdown and to mitigate an accident condition. However, because the overall reliability is reduced, continued power operation should not exceed 24 hours. If the D.C. power distribution system cannot , be reenergized in 24 hours, a controlled shutdown must be initiated. With two divisions of the D.C. power distribution system deenergized, the overall system reliability is further reduced and continued operation should not exceed two hours. If at least one of the deenergized D.C. power distribution divisions cannot be reenergized within two hours, or the initially deenergized division cannot be reenergized within 24 hours of initial discovery, then a controlled shutdown must be initiated. i (continued)
)
ABWR B 3.8-46 5/31/89 i
D.C. Power Dist. Systems - Operating
- B 3.8.8 Q
/3 BASES (continued)
ACTIONS An alternate approach is to declare inoperable'all equipment associated with the deenergized power distribution system
~
(continued) immediately- and enter the ACTIONS of the corresponding LCOs.
' Q. l . D' . 2 With three or more'of the required D.C. power distribution divisions deenergized, overall reliability is further reduced and sufficient power may not be available to recover-from a design basis' accident or transient. It.is therefore'necessary-to immediately declare inoperable all equipment , associated with the deenergized power distribution system buses or initiate a controlled shutdown.
SURVEILLANCE SR 3.8.8.1-REQUIREMENTS This surveillance demonstrates that the electrical power distribution system is functioning properly with all the desired circuit breakers closed and the buses energized. The i. verification of proper voltage availability on the divisional buses ensures that the required power is readily available for motive as well as control functions for critical system loads (p) connected to these buses. Surveillance Frequencies in general, surveillance frequencies are based on engineering judgement and industry accepted practice considering the unit conditions required to perform the test, the ease of performing the test and the' likelihood of a change in the system component status. REFERENCES 1. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
- 2. Regulatory Guide 1.93, " Availability of Electric Power Sources," December 1974.
l l l 1 d
^%l
'n l
I ABWR. B 3.8-47 5/31/89 ) i 1
- -- - ___________________________J
s. D.C. Power Dist. Systems - Operating B 3.8.8 Table B 3.8.8-1 (Page 1 of 1) D.C. Power Distribution Systems [Later] O 1 i hi ABWR B 3.8-48 5/31/89 _ _ ______-___-_____a
A.C. & D.C. Power Dist. Systems - Shutdown B 3.8.9 ('N ' B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.9 A.C. and D.C. Power Distribution Systems - Shutdown BASES BACKGROUND A description of the A.C. and D.C. electrical power distribu-tion system is provided in the Bases for LCO 3.8.7 and LC0 3.8.8. APPLICABLE A reduced compliment of distribution systems in MODES 4 and 5, SAFETY and when handling irradiated fuel in the secondary containment ANALYSES is adequate to assure power for systems required to recover from an inadvertent draindown of the vessel or a fuel bundle handling accident (Ref. 1). A.C. and D.C. Power Distribution Systems - Shutdown satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2.
^ Any OPERABLE division of both A.C. and D.C. electrical power LCO V)
( distribution ensures the availability of the required power to-recover from postulated events in MODES 4 and 5, and when handling irradiated fuel (e.g., fuel handling accident, vessel draindown). APPLICABILITY The OPERABILITY of the electrical power distribution systems in MODES 4 and 5, and when moving irradiated fuel is required to assure (1) adequate coolant inventory for the irradiated fuel in the core in case of an inadvertent draindown of the vessel, (2) mitigation of a fuel bundle handling accident, (3) suffi-cient power for required support systems (e.g., decay heat removal, refueling activities, component cooling), and (4) sufficient instrumentation and control capability for monitoring and maintaining the unit status. (continued) 1' . O I' ABWR B 3.8-49 5/31/89 L - - - - - - - - - - - - _ _ _ - -
A.C. & D.C. Power Dist. Systems - Shutdown r B 3.8.9 I' BASES (continued) ACTIONS A.I. A.2. A.3. A.4 With the required A.C. or D.C. power distribution, systems not energized, sufficient power may not be available to recover from a fuel handling accident or vessel draindown event. Consequently, it is required to take immediate Actions to suspend CORE ALTERATIONS and handling of irradiated fuel in the secondary containment. Suspension of these activities shall not preclude completion of the movement of a component to a safe, conservative position. Activities which could poten-tially result in an inadvertent draindown of the vessel (e.g., undervessel control rod drive replacement) must be suspended as soon as practical. This should preclude the occurrence of the postulated events. SURVEILLANCE The Bases provided for SR 3.8.7.1 and SR 3.8.8.1 are REQUIREMENTS applicable. REFERENCES 1. ABWR SSAR, Section 15.7.4.
- 2. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.8-50 5/31/89 j 1 I 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ l
[' Refueling Equipment Interlocks B ?.R.1
,s B 3.9 REFUELING OPERATIONS ' ( )-
B 3.9.1 Refuelina Eauipment Interlocks BASES-BACKGROUND Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures to prevent the reactor from achieving criticality during refueling operations. The refueling interlock circuitry senses the condition of the refueling equipment and the control rods. Depending on the-sensed condition, interlocks are actuated to prevent the movement of the refueling equipment or withdrawal of control rods (control rod block). Two channels of instrumentation are provided to sense the refueling platform position, refueling platform main hoist condition (fuel loaded) and whether all control rods are fully inserted. With the Reactor. Mode switch in the Shutdown or Refuel position, the indicated conditions are combined in logic circuits to determine if all restrictions on refueling equipment operations and control rod movement are satisfied. Operation of refueling equipment is prevented by interrupting
/-'s the power supply to the equipment. The refueling platform is provided with two mechanical switches that open before the
('~) platform or any of its hoists are physically located-over the reactor vessel to indicate the approach of the platform toward its position over the core. The refueling platform hoists (used for fuel movement) are provided with switches that open when the hoist is fuel loaded. The switches open at a' load weight lighter than the weight of a. single fuel assembly. The refuelin'g interlocks use.the above indications to prevent operation of the refueling equipment (when fuel loaded) over
~
the core whenever any control rod is withdrawn and also prevent control rod withdrawal whenever refueling equipment (fuel loaded) is over the core. , (continued) ; i l i l l ABWR B 3.9-1 5/31/89
y. Refueling Equipment Interlocks B 3.9.1 BASES (continued) APPLICABLE The refueling interlocks are explicitly assumed in the FSAR SAFETY analysis for the Control Rod Removal Error During Refueling ANALYSES (Ref. 1) Theses analyses evaluate the consequences of control rod withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Criticality (and therefore subsequent prompt reactivity excursions) is prevented during the insertion of fuel, provided all control rods are fully inserted during the fuel insertion. The refueling. interlocks accomplish this by preventing loading of fuel into the core with any control rod withdrawn. Refueling Equipn.ent Interlocks satisfy the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. LC0 To prevent criticality during refueling, the refueling interlocks ensure that fuel assemblies are not loaded with any control rod withdrawn. ^ prevent these conditions from developing, the all-rot , refueling platform position and refueling platform main ..a st fuel loaded inputs are required OPERABLE. These inputs are combined in logic circuits which h provide equipment and control rod blocks to prevent operations that could result in criticality during refueling operations. APPLICABILITY The refueling interlocks provide protection against prompt reactivity excursions during MODE 5. The interlocks are only required to be OPERABLE when the refueling equipment is being used to handle fuel. In all other MODES, the refueling interlocks are not required to be OPERABLE. During these other conditions, the Reactor Protection System (LC0 [ ]), Control Rod Block Instrumentation (LC0 [ ]) and Control Rod OPERABILITY (LC0 3.1.2) provide mitigation of potential reactivity excursions. (continued) O ABWR B 3.9-2 5/31/89
f l Refueling Equipment Interlocks 'j B 3.9. l' X BASES (continued) 7 ACTIONS ad' With any refueling equipment interlock inoperable (does not include the one-rod-out interlock), performance of CORE ALTERATIONS with the affected refueling equipment should be immediately suspended. This action ensures that operations.are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel. into a cell with a withdrawn control rod).- Suspension.of CORE ALTERATIONS shall not preclude completion of the movement of'a component to a safe conservative position. CORE ALTERATIONS normally performed with other equipment not affected by these refueling interlocks is not prohibited. SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated refuel.ing interlock will function properly-when a simulated or actual signal indicative of. a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such th't a the entire channel is tested. y
.0perating experience has demonstrated that a seven day frequency for this surveillance is adequate.
REFERENCES 1. ABWR SSAR, Section 15.4.1.1.
- 2. NED0-31466, " Technical Specification. Screening Criteria Application and Risk Assessment," November 198L i
ABWR B 3.9-3 5/31/89
Refuel Position One-Rod-Out Interlock B 3.9.2 L B 3.9 REFUELING OPERATI9NS B 3.9.2 Refuel Position One-Rod-Out Interlock BASES BACKGROUND Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures which prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod is permitted to be withdrawn from a core cell containing one or more fuel assemblies. This is enforced by a logic circuit consisting of redundant channels that uses the all-rods-in signal (which receives input from the control rod Full In position indicators as discussed in LCO 3.9.4) and a rod selection signal to prevent the selection of a second control rod for movement when any other control rod is not fully inserted. APPLICABLE The refueling one-rod-out interlock is explicitly assumed in , SAFETY the FSAR analysis for the Control Rod Removal Error During l ANALYSES Refueling (Ref. 1). This analysis evaluates the consequences of control rod withdrawal during refueling. A prompt reactiv-ity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. The Refuel position one-rod-out interlock, together with requirements for adequate SHUTDOWN MARGIN during refueling (LCO 3.1.1) prevent criticality by preventing with-drawal of more than one control rod. With only one control rod ' withdrawn, the core will remain subcritical thereby preventing any prompt critical excursion. Refuel Position One-Rod-Out Interlock satisfies the require-ments of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 2. (continued) 1 0 ABWR B 3.9-4 5/31/89 1 i
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued) LC0 To prevent criticality during MODE 5, the Refuel position one-rod out interlock ensures no more than one control rod may be withdrawn. To prevent this condition from developing, both channels of the Refuel Position one-rod-out interlock are required to be OPERABLE. APPLICABILITY The Refuel position one-rod out interlock provides protection against prompt reactivity excursions only when the Reactor Mode Switch is in the Refuel position. With the Reactor Mode Switch in other positions, the Refuel position one-rod-out interlock is bypassed and not required to be OPERABLE. During these conditions, the Reactor Protection System (LC0 [ ]) and Control Rod OPERABILITY (LC0 3.1.2) provide mitigation of potential reactivity excursions. With the Reactor Mode Switch in the Shutdown position, a control rod block (LC0 [ Control Rod Block Instrumentation)) ensures all control reds are inserted, thereby preventing criticality during refueling er shutdown conditions. ACTIONS A.1. A.2 With the one-rod-out interlock inoperable (one or both channels), the refueling interlocks may not be capable of preventing more than one control rod from being withdrawn. These conditions may lead to criticality and therefore control rod withdrawal must be suspended and any withdrawn control rod from a core cell containing one or more fuel assemblies must be fully inscried as soon as practicable. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted. The Completion Time of "as soon as practicable" is provided to allow some judgement to determine the quickest and safest method to complete the Required Action or restore compliance with the LC0. In general, under these conditions only a single control rod will be withdrawn from a core cell containing one or more fuel assemblies and will be readily inserted in a short period of time. (continued) O
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES (continued) SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated Refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a i required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested. An allowance is provided to withdraw a single control rod for the purpose of testing the interlocks. Operating experience has demonstrated that a seven day frequency for this surveillance is adequate. REFERENCES 1. ABWR SSAR, Section 15.4.1.1.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O 1 t j O! l ABWR B 3.9-6 5/31/89
._____-___________O
L l ! Centrol Rod Position 1 B 3.9.3 l B 3.9 REFUELING 7); t
'" )
B 3.9.3 Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD system. During refueling, movement of control rods i.; limited by the refueling interlocks (LC0 3.9.1 and 3.9.2) or the control rod block function of the Reactor Mode Switch Shutdown position (LC0 [ Control Rod Block Instrumentation]). The refueling interlocks will allow a single control rod to be withdrawn at any one time unless the Full In position indica- i tion of one or more control rods has been bypassed as allowed ! by LC0 3.10.6. Under these conditions, multiple control rods may be withdrawn and it is necessary to preclude loading fuel assemblies into the core during a fuel shuffle or to follow fuel loading sequences specifically designed to minimize the probability of criticality during refueling to prevent an inadvertent criticality. Requiring all control rods to be inserted when loading fuel assemblies into the core during a fuel shuffle provides this protection. (3
; i APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY during refueling is provided by refueling interlocks (LC0 ANALYSES 3.9.1 and 3.9.2), SHUTDOWN MARGIN (LCO 3.1.1), Startup Range Neutron Monitor (SRNM) neutron flux scram (LC0 [RPS Actuation Instrumentation]), Average Power Range Monitor (APRM) nautron flux scram (LC0 [RPS Actuation Instrumentation]) and control rod block instrumentation (LC0 [ Control Rod Block Instruments-tion]). Explicit safety analyses in the SSAR (Ref.1) only assume the functioning of the refueling interlocks and adequate SHUTDOWN MARGIN. However, the control rod Full In position indicators are allowed to be bypassed in accordance with LC0 3.10.6 if all fuel assemblies are removed from the associated core cell. With more than one control rod Full In position indicator bypassed, the control rods could be withdrawn and the refueling interlocks would not prevent fuel loading. There-fore, prior to fuel reload, all control rods must be .'ully inserted to ensure an inadvertent criticality does not occur consistent with the analysis of Reference 1. An exception to this requirement is allowed if fuel is being loaded in an approved spiral reload sequence that does not use a complete set of blade guides. The approved spiral reload sequence typically involves reloading such that fuel is always being loaded on the periphery of the fueled zone. During the spiral
(~g V (continued) ABWR B 3.9-7 5/31/89
r J l Control Rod Position B 3.9.3 BASES (continued) APPLICABLE reloading, all control rods in core cells containing fuel are - { SAFETY fully inserted and the control rod in the next cell to be l ANALYSES loaded is fully inserted. This minimizes the reactivity 1 insertion of each loaded fuel assembly and greatly reduces the I probability of a reactivity excursion. j Control Rod Position satisfies- the requirements of Selection i Criterion 3 of the NRC Interim Policy Statement on Technical ] Specification Improvements as documented in Reference 2. LC0 All control rods must be fully inserted during a fuel reload not following an approved spiral reload sequence to prevent an inadvertent criticality during refueling. APPLICABILITY loading fuel into core cells with the associated control rods withdrawn may result in inadvertent criticalities during refueling. Therefore, prior to loading fuel into a core cell, the associated control rod must be inserted. To ensure that a fuel loading error does not result in loadiag fuel into a core cell with the associated control rod withdrawn, all control rods must be inserted prior to loading fuel. This requirement is not applicable when fuel is being loaded while following an approved spiral reload sequence which is specifically designed to minimize the probability of an inadvertent criticality. ACTIONS A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. All control rods would have to be fully inserted before resuming fuel loading operations. Suspension of these activities shall not preclude the completion of movement of a component to a safe, conservative position. SURVEILLANCE SR 3.9.3.1 i i REQUIREMENTS During refueling, to ensure the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensures this condition is maintained. Operating experience has demonstrated that a 12 hour frequency for this surveillance is adequate. (continued) ABWR B 3.9-8 5/31/89
Control Rod Position B 3.9.3 BASES (continued) (~') (/ ABWR SSAR, Section 15.4.1.1. REFERENCES 1.
- 2. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
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OG l ABWR B 3.9-9 5/31/J.B \
' Control Rod Position Indication i B 3.9.4 i B 3.9 REFUELING B 3.9.4 Control Rod Position Indication I
BASES BACKGROUND During refueling, movement of control rods is controlled by the refueling interlocks (LCO 3.9.1 and 3.9.2) or the control rod block function of the Reactor Mode Switch Shutdown position i (LC0 [ Control Rod Block Instrumentation]). The refueling j interlocks will allow only a single control rod to be withdrawn J at any one time. The refueling interlocks use the Full In position indicators to determine the position of all control I If the [ Full In] position signal is not present for rods. every control rod, then the all-rods-in permissive for the refueling equipment interlocks is not present and fuel loading is prevented. Also, the Refuel position one-rod-out interlock will not allow the withdrawal of a second control rod. There-fore, the Full In position indicators provide necessary infor-mation to the refueling interlocks to prevent inadvertent criticalities during refueling operations. APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY during refueling is provided by refueling interlocks (LC0 ANALYSES 3.9.1 and 3.9.2), SHUTDOWN MARGIN (LC0 3.1.1), Startup Range Neutron Monitor (SRNM) neutron flux scram (LCO [RPS Actuation Instrumentation]), Average Power Range Monitor (APRM) neutron flux scram (LCO [RPS Actuation Instrumentation]) and control rod block instrumentation (LC0 [ Control Rod Block Instruments-tion]). Explicit safety analyses in the SSAR (Ref. 1) only assume the functioning of the refueling interlocks and adequate SHUTDOWN MARGIN. The Full In position indicators are required to be OPERABLE such that the refueling interlocks can perform their intended function. The refueling interlocks ensure no more than one control rod can be withdrawn and fuel cannot be loaded with any control rod withdrawn. Only those control rods in core cells containing one or more fuel assemblies must be inserted since control rods withdrawn from control cells which do not contain fuel assemblies have a negligible impact on reactivity control and are therefore not required to be , inserted. These control rods are therefore not required to have OPERABLE full In position indication and can be bypassed in the refueling interlocks logic once the fuel has been removed from the cell as allowed by LC0 3.10.6. LCO 3.9.3 ensures these control rods are fully inserted during fuel loading (even if fuel is not scheduled to be loaded into the affected cell) when applicable. Otherwise, the control rod must be fully inserted and the position indication OPERABLE before loading fuel assemblies into the cell. (continued) ABWR B 3.9-10 5/31/89
Control Rod Position Indication B 3.9.4 BASES (continued) APPLICABLE Control Rod Position Indication satisfies the requirements of SAFETY Selection Criterion 3 of the NRC Interim Policy Statement on ANALYSES Technical Specification Improvements as documented in Reference (continued) 2. LC0 All control rod Full In position indicators for control rods in core cells containing one or more fuel assemblies must be OPERABLE to provide required input to the refueling interlocks. The indicator is considered OPERABLE if it is providing the correct position indication to the refueling interlock logic. APPLICABILITY During MODE 5, control rods from core cells containing one or more fuel assemblies must have OPERABLE Full In position indication to ensure the applicable refueling interlocks will be OPERABLE. Control rod position indication in MODES 1 and 2 is located in LC0 3.1.2, Control Rod OPERABILITY. In MODES 3 and 4, with the Reactor Mode Switch in the Refuel position, the Full In position indicators are required to be OPERABLE to ensure the Refuel position one-rod-out interlock is OPERABLE as required by LCO 3.10.3 and LC0 3.10.4. ACTIONS A.1.1. A.1.2. A.2.1. A.2.2 With the Full In position indication inoperable for a control rod in a core cell containing one or more fuel assemblies, the affected refueling interlocks must be declared inoperable and the Required Actions of the affected LCOs will provide compen-sating utions. Alternatively, the associated contrr>l rod may be fully inserted and disarmed to ensure the control rod is not l withdrawn. Under these conditions, the Full In indication may be bypassed for the affected control rod. Alternate methods should be used to ensure the control rod is fully inserted (e.g., use the 100% inserted position indication). If all fuel is removed from a core cell, the Full In position indication may be bypassed in the logic since the control rod may be withdrawn and the position indication is not required to OPERABLE (LC0 3.10.6). Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. (continued) ABWR B 3.9-11 5/31/89
Control Rod Position Indication B 3.9.4 l BASES (continued) SURVEILLANCE SR 3.9.4.1 i REQUIREMENTS The Full In position indicators are periodically verified to be j OPERABLE by verifying for all required control rods that the Full In position indication is available for all control rods I fully inserted. Operating experience has demonstrated that the seven day frequency for this surveillance is adequate. i REFERENCES 1. ABWR SSAR, Section 15.4.1.1. 1
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O i O ABWR B 3.9-12 5/31/89
Control Rod OPERABILITY - Refueling , B 3.9 5 ys B 3.9 REFUELING .
! )
M B 3.9.5 Control Rod OPERABILITY - Refuelina BASES BACKGROUND Control rods are' components of the Control Rod Drive (CRD) L system, which is the primary reactivity control system for-the reactor. In conjunction with the Reactor Protection System 1 (RPS), the CRD system provides the means for the reliable control of reactivity changes during operation. In addition, the control rods provide the capability to maintain the reactor-
.subtritical under all conditions and to limit the potential i amount and rate of reactivity increase caused by.a malfunction- )
in the CRD system. ! i
' APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY during refueling is provided by refueling interlocks (LC0 i ANALYSES 3.9.1 and 3.9.2), SHUTDOWN MARGIN (LC0 3.1.1), Startup Range Neutron Monitor (SRNM) neutron flux scram (LC0 [RPS Actuation Instrumentation]), Average Power Range Monitor (APRM) neutron flux scram (LCO [RPS Actuation Instrumentation]) and control-rod block instrumentation (LC0 [ Control Rod Block Instruments-tiun]). Explicit safety analyses in the SSAR (Ref.1) only
(.'j q assume the functioning of the' refueling interlocks and adequate SHUTDOWN MARGIN. Control rod scram provides backup protection should a prompt reactivity excursion occur. Control Rod OPERABILITY - Refueling satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy' Statement on Technical Specification Improvements as documented in Reference 2. LCO Each control rod withdrawn from a core cell containing one or more fuel assemblies must be OPERABLE. The control rod is considered OPERABLE if it is capable of being automatically inserted upon receipt of a scram signal. Inserted control rods have already completed their reactivity control function. Control rods withdrawn from control cells which do not contain fuel assemblies have a negligible impact on reactivity control and are therefcre not required to be OPERABLE. (continued) l O v i ABWR B 3.9-13 5/31/89 1
-____--________-_________O
Control Rod OPERABILITY - Refueling B 3.9.5 BASES (continued) APPLICABILITY During MODE 5, control rods withdrawn from a core cell contain-ing one or more fuel assemblies must be OPERABLE to ensure during a scram, the control rods will insert and provide the required negative reactivity to maintain the reactor subcriti-cal. Control rod requirements in MODES 1. and 2 are located in LCOs 3.1.2, 3.1.3, 3.1.4 and 3.1.5. During MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LC0 3.10.3 (Control Rod Withdrawal - Hot Shutdown) and LC0 3.10.4 (Control Rod Withdrawal - Cold Shutdown) which j provide adequate requirements for control rod OPERABILITY during these conditions. ACTIONS A.1 With a required control rod inoperable, the control rod must be fully inserted as soon as practicable. Inserting the control i rod ensures the shutdown and scram capabilities are not adversely affected since inserted control rods have completed their required reactivity control function. SURVEILLANCE SR 3.9.5.1, SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal rec;uiring a reactor shutdown occurs. Because no explicit analyses exist for automatic shutdown during refueling, the shutdown function is considered satisfied if the associated CRD scram accumulator is properly charged and the withdrawn control rod is capable of insertion. Operating experience has demonstrated that the seven day frequency for these surveillance is adequate. j l REFERENCES 1. ABWR SSAR, Section 15.4.1.1.
- 2. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.9-14 5/31/89 l l
l Water Level - RPV l' B 3.9.6 y B 3.9 REFUELING OPERATIONS
, t
B 3.9.6 Water tevel - Reactor Pressure Vessel BASES BACKGROUND Fuel movement and reactor servicing operations that consist of handling various loads such as fuel assemblies, control rods, fuel channels or other reactor equipment are performed within the Reactor Pressure Vessel (RPV), during refueling, from platforms above the RPV cavity. The water level above the RPV flange is maintained at a depth of 23'to provide adequate radiological shielding for radiation protection under normal and accident conditions, i APPLICABLE The water level above the RPV flange is an explicit assumption SAFETY of the fuel handling accident (Ref. 1). A fuel handling ANALYSES accident is evaluated to ensure the radiological consequences (calculated whole-body and thyroid doses at the exclusion area and low population zone boundaries) are s 25% of the 10 CFR Part 100 exposure guidelines (Ref. 2). A fuel handling acci-dent could release a fraction of the fission product inventory by breaching the fuel rod cladding (Ref. 3). The water level I
- (7 above the RPV flange provides for absorption of water soluble _) fission product gases and transport delays of soluble and insoluble gases which must pass through the water before being released to the containment atmosphere. This absorption and transport delay reduce the potential radioactivity of the release during a fuel handling accident. Water Level - Reactor Pressure vessel satisfies the require-ments of Selection Criterion 3 of the NRC Interim Policy l Statement on Technical Specification Improvements as documented in Reference 4. LC0 A depth of 2 23' of water over the top of the RPV flange is required to ensure adequate absorption and transport delay of fission product gases is available so offsite doses will meet the acceptance criteria of Reference 2 in the event of a fuel handling accident within the RPV. (continued) i A k Y) ABWR B 3.9-15 5/31/89
' Water tevel - RPV B 3.9.6 HAIES (continuejl ;
APPLICABILITY A fuel handling accident is postulated to occur only during MODE 5 when operations requiring handling of irradiated fuel assemblies in the RPV are performed. During other conditions, handling 'of irradiated fuel within the RPV is not possible and , therefore the fuel handling accident is not a concern. Requirements for fuel handling accidents in the Spent Fuel Storage Pool is covered by .00 3.9-.7. ACTIONS A.J With inadequate water above the RPV flange, all handling of fuel assemblies must be suspended immediately. Suspending further movement of fuel assemblies precludes the possibility of a fuel handling accident from an unanalyzed condition. j
. Suspension of fuel handling shall not preclude completion of movement to a safe, conservative position. ;
1 SURVEILLANCE SR 32 9.6,1 REQUIREMENTS The minimum required water level, 23', above the top of the RPV flange is verified periodically to ensure that the water soluble fiss' ion product gases removal assumptions of the fuel handling accident analysis are maintained so that if the fuel handling accident occurs, radiation exposures will be main-tained 5 25% of the limits specified in 10 CFR Part 100. Operating experience has demonstrated that the 24 hour frequency for this surveillance is adequate. REFERENCES 1. ABWR SSAR, Section 15.7.4.
- 2. NRC Standard Review Plan 15.7.4.
- 3. NRC Regulatory Guide 1.25.
- 4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.9-16 5/31/89 L _ ______________________a
m Water Level - Fuel Pool B 3.9.7 ] W -B 3.9 -REFUELING OPERATIONS B 3'.9.7 Water Level .Soent Fuel Storace Pool ) BASES BACKGROUND The spent fuel storage racks in the Spent Fuel Storage Pool store irradiated fuel assemblies received from the reactor pressure vessel (RPV). Refueling operations. consist of l handling various loads such as fuel assemblies, control rods, fuel channels and reactor servicing equipment over the~ fuel assemblies. stored in the Spent Fuel Storage Pool racks. The water level above the top of the stored fuel in the Spent Fuel Storage Pool is maintained at a depth of 23 feet to provide adequate radiological shielding for radiation protection under normt) and accident conditions. APPLICABLE The water level above the irradiated fuel assemblies is an w SAFETY explicit assumption of the fuel handling accident '(Ref.1). A ANALYSES fuel handling accident is evaluated to ensure the radiological consequences (calculated whole-body and thyroid doses at the exclusion area and low population zone boundaries) are s 25% of the 10 CFR 100 exposure guidelines (Ref. 2). A fuel handling accident could release a fraction of the fission product Q Q inventory by breaching the fuel rod cladding (Ref. 3). The water level in the Spent Fuel Storage Pool provides for absorp-tion of water soluble fission product gases and transport delays of soluble and insoluble gases which must pass through the water before being released to'the Secondary Containment
-atmosphere. This absorption and transport delay reduce the l;;
potential radioactivity of the release during a fuel handling accident. Water Level - Spent Fuel Storage Pool satisfies the require-ments of Selection criterion 3 of the NRC Interim Policy 1 Statement on Technical Specification Improvements as documented in Reference 4. i> l LCOs A depth of 2 23 feet of water over the top of the stored spent
- 1. fuel is required to ensure adequate absorption and transport delay of fission product gases is available so offsite doses will meet the acceptance criteria of Reference 3 in the event of a fuel handling accident in the Spent Fuel Storage Pool.
(continued) u ABWR B 3.9-17 5/31/89 j:
Water Level - Fuel Pool i B 3.9.7 1 BASES (continued) APPLICABILITY A fuel handling accident is postulated to occur during any MODE whenever irradiated fuel assemblies are being handled in the 1 Spent Fuel Storage Pool. I ACTIONS .AL1 With inadequate water above the spent fuel, all handling of fuel assemblies within the Spent fuel Storage Pool must be suspended immediately. Suspending further movement of fuel , assemblies precludes the possibility of a fuel handling accident occurring from an unanalyzed condition. Suspension of fuel handling shall not preclude completion of the movement to a safe, conservative position. SC2VEILLANCE SR 3.9.7.1 REQUIREMENTS The minimum required water level, 23 feet above the top of the stored spent fuel, is verified periodically to ensure that the water soluble fission product gases removal assumptions of the fuel handling accident analysis (Ref.1) are maintained so that if the fuel handling accident occurs, radiation exposure will be maintained s 25% of the limits specified in 10 CFR Part 100. Operating experience has demonstrated that the 24 hour frequency for this surveillance is adequate. REFERENCES 1. ABWR SSAR, Section 15.7.4.
- 2. NRC Standard Review Plan 15.7.4.
- 3. NRC Regulatory Guide 1.25.
- 4. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O ABWR B 3.9-18 5/31/89
RHR - High Water Level .j B 3.9.8
- p- B 3.9 REFUELING OPERATIONS B 3'.9.8 Residual Heat Removal - Hioh Water Level BASES i 1
BACKGROUND Refueling operations in the secondary containment are performed by personnel working above the reactor pressure vessel (RPV) cavity. The irradiated fuel in the RPV generates decay heat causing an increase of the temperature of the reactor coolant with resultant increase of the air temperature in the vicinity of the surface of the water above the reactor core. OPERABLE RHR shutdown cooling subsystems are therefore required for removing this decay heat to maintain the temperature of reactor coolant such that personnel can perform refueling operations inside the secondary containment. Each of the three shutdown cooling loops of the Residual Heat Removal (RHR) system can provide the required decay heat removal. Each loop consists of one motor driven pump, a heat exchanger, and associated piping and valves. Each loop has a dedicated suction nozzle from the reactor vessel. Each pump discharges the reactor coolant, after is has been cooled by circulation through the respective heat exchangers, to the reactor via feedwater line B for subsystem A, and via the individual RHR inlet nozzles for subsystems B and C. The RHR {V heat exchangers transfer heat to the Reactor Building Cooling Water System (LCO 3.7.2). The RHR shutdown cooling mode is a manually controlled system. In addition to the RHR shutdown cooling subsystems, the volume of water above the RPV flange provides a heat sink for decay heat removal. APPLICABLE Decay heat removal by operation of the shutdown cooling mode of - SAFETY the RHR system or any other method (e.g. volume of water above ANALYSES the RPV flange) is not required for mitigation of any events or accidents evaluated in the safety analyses. However, the NRC Interim Policy Statement (Ref. 1) requires the RHR system be retained in the Technical Specifications even though none of the selection criteria were satisfied (Ref. 2). (continued) n ABWR B 3.9-19 5/31/89
RHR - High Water Level B 3.9.8 BASES (continued) LCOs During MODE 5 with water level ;t [ ] above the RPV flange, one RHR shutdown cooling subsystem is required to be OPERABLE. Only one subsystem is required because the volume of' water above the RPV flange provides the required redundant decay heat removal capability. An OPERABLE RHR shutdown cooling subsystem consists of one RHR pump, one heat exchanger and the associated piping and valves. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. , APPLICABILITY In MODE 5, the irradiated fuel in the RPV generates decay heat that may cause an increase in the. temperature of the reactor coolant and the environment around the RPV. OPERABILITY of a decay heat removal subsystem is therefore required. When heat losses to the ambient are sufficient to remove the decay heat, no residual heat removal systems are required. Decay heat removal requirements in MODES 3 and 4 are located in LC0 3.4.5. Decay heat removal requirements in M0.r 5 when the water level is < [ ] above the RPV flange are located in LC0 3.9.9. O ACTIONS .A_d With no RHR Shutdown Cooling Subsystem OPERABLE, the volume of-water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the capa-bility to maintain the reactor coolant temperature such that personnel can perform refueling operations may be lost. Therefore, a limited time is allowed to establish an alternate method capable of decay heat removal or restore at least RHR shutdown cooling system to maintain the average reactor coolant temperature. (continued) O ABWR B 3.9-20 5/31/89
RHR - High Water Level B 3.9.8 A7 BASES (continued) ACTIONS B.1. B.2. B.3. B.4. B.5 (continued) If. an alternate method cannot be established and the associated Completion Time is not met, all operations that can potentially increase reactor decay heat generation (e.g. Toading irradiated
~
fuel bundles in the core) shall be suspended to preclude enhancing the rate of increase of. coolant temperature'. Actions are also taken to provide means for control of potential radioactive releases. This includes ensuring Secondary Con-tainment is OPERABLE (LCO 3.6.4.1), at least one Standby Gas Treatment System (SGTS) (LCO 3.6.4.3) subsystem is OPERABLE and at least one Secondary Containment Isolation Valve (LCO 3.6.4.2)'and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. Ensuring the OPERABLE status of the components involves administrative checks, such as' examining logs or other information, to deter-mine if the components are out of service for maintenance or other reasons. It does not require performing surveillance needed to demonstrate the OPERABILITY of the components. If however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, surveillance requirements may need to be performed to restore the component to OPERABLE status. Additionally, actions must continue to establish an alternate decay heat removal method as soon as practicable. {V SURVEILLANCE SR 3.9.8.1 REQUIREMENTS Verification that all valves of the required RHR shutdown cooling subsystems are in the correct position ensures the proper flow path. Valves not in the correct position must be capable of manual realignment either from'the control room or at the valve location. Verification of this capability'is provided by actuation of the valve from the control room or the current inservice inspection reports. REFERENCES 1. 52FR3788, Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.
- 2. NED0-31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
.O V
ABWR B 3.9-21 5/31/89
RHR - Low Water Level B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 Residual Heat Removal - Low Water level BASES BACKGROUND A description of the Shutdown Cooling Subsystem of the Residual Heat Removal (RHR) System is provided in the Bases for LC0 3.9.8. APPLICABLE Decay heat removal by operation of the shutdown cooling mode of SAFETY the RHR system is not required for mitigation of any events or ANALYSES accidents evaluated in the safety analyses. However, the NRC Interim Policy Statement (Ref. 1) requires the RHR system be retained in the Technical Specifications even though none of the selection criteria were satisfied (Ref. 2). LCOs During MODE 5, with water level < [ ] the RPV flange, two RHR shutdown cooling subsystems are required to be OPERABLE. The low volume of water above the flange may not provide an adequate heat sink for decay heat removal. An OPERABLE RHR shutdown cooling subsystem consists of one RHR pump, one heat exchanger and the associated piping and valves. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. Any other subsystem used for decay heat removal is considered OPERABLE when it is capable of maintaining or reducing reactor coolant temperature as required. APPLICABILITY In MODE 5 the irradiated fuel in the RPV generates decay heat I that may cause an increase in the temperature of the reactor coolant and the environment around the RPV. OPERABILITY of a decay heat removal subsystem is therefore required. When heat losses to the ambient are sufficient to remove the decay heat, I no residual heat removal systems are required. Decay heat removal requirements in MODES 3 and 4 are located in LC0 3.4.5. Lecay heat removal requirements in MODE 5 when the water level 1.s 2 [ ] above the RPV flange are located in LLO 3.9.8. (continued) O ABWR B 3.9-22 5/31/89
I RHR - Low Water Level B 3.9.9 BASES (continued) d(m ACTIONS A.1. A.2. B.1 With only one shutdown cooling subsystem OPERABLE, the remain-ing OPERABLE shutdown cooling subsystem is capable. of providing ) the required decay heat removal. However,'the overall system reliability is reduced and a limited time is allowed to restore the inoperable loop or provide an alternate method of decay heat removal. As an alternative, the RPV water level can be increased to be ;t [ ] such that LCO 3.9.9 is no longer appli-cable and the provisions of LCO 3.9.8 are satisfied. With no RHR Shutdown Cooling Subsystem OPERABLE, the capability of the RHR shutdown cooling system to provide decay heat removal -is potentially lost and an alternate method capable of decay heat removal shall be provided to maintain the average reactor coolant temperature. Because of the low decay heat levels during these' conditions a short time is provided. C.1. C.2. C.3. C.4. C.5 If an alternate m0thod cannot be established and the associated Completion Time not met, all operations that can potentially increase reactor decay heat generation (e.g. loading irradiated fuel bundles in the. core) shall be suspended to preclude enhancing the rate of increase of coolant temperature. Actions (Q/ are also taken to provide'means for control of potential radioactive' releases. ~This includes ensuring Secondary Con-tainment is OPERABLE (LC0 3.6.4.1), at least one Standby Gas Treatment' System (SGTS) (LC0 3.6.4.3) subsystem is OPERABLE and at least one Secondary Containment Isolation Valve {LC0 3.6.4.2) and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. Ensuring the OPERABLE status of the components involves administrative checks, such as examining logs or other information, to deter-mine if the components are out of service for maintenance or other reasons. It does not require performing surveillance needed to demonstrate the OPERABILITY of the components. If however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, surveillance requirements may need to be performed to restore the component to OPERABLE status. Additionally, actions must continue to establish an alternate decay heat removal method as soon as practicable. (continued) ABWR B 3.9-23 5/31/89
l RHR - Low Water Level B 3.9.9 l BASES (continued) SURVEILLANCE SR 3.9.9.1 REQUIREMENTS The bases provided for SR 3.9.8.1 are applicable. REFERENCES 1. 52FR3788, Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.
- 2. NED0 31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
O' O ' ABWR B 3.9-24 5/31/89
)
- - _ _ - _____ - -__ _ _____ _ _ _ - ___ _____ _ O
l Inservice Leak.and' Hydro Test f l' < B 3.10.1 y
.. B 3i10: 'SPECIAL OPERATIONS
':( ,
'B 3.10.1 Inservice Leak and Hydrostatic Test BASES BACKGROUND . Inservice hydrostatic testing and system leakage pressure tests required by Section XI of the ASME Boiler and Pressure ,
Vessel Code are performed prior to the_ reactor going critical after a refueling outage. Recirculation pump operation and a water-solid reactor. pressure. vessel-(except for.an air bubble fcr pressure control) are used to achieve the necessary temperatures and pressures required for these tests. The minimum temperatures (at the required pressuies) allowed for these tests are determined from the reactor pressure vessel-pressure / temperature-(PT)' limits required by.LC0 3.4.6. These limits-are. conservatively based upon the fracture toughness of' the reactor vessel, taking into account anticipated vessel-neutron fluence. With-increased reactor vessel fluence over time, the minimum allowable vessel temperature (at a given pressure) increases. Periodic updates to -the- reactor pressure vessel PT -limit curves-are performed, as necessary, based upon the results of analyses of irradiated surveillance specimens removed from the vessel. p Hydrostatic and leak testing will eventually be required with
. minimum reactor coolant temperatures above 200*F.
APPLICABLE Allowing the reactor to be considered in MODE 4 during SAFETY hydrostatic or leak testing when reactor coolant temperature is ANALYSES above 200*F, will effectively provide an exception to MODE 3 requirements. Since the hydrostatic or leak tests are performed water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be ver) low. Under these conditions, the potential. for failed fuel and subsevent increase in coolant activity above Technical Specifications limits is. mitigated. In sddition, the Secondary Containment will be OPERABLE in accordance with this Special Operation and will be capable of handling any airborne radiation or steam leaks that could occur during the performance of hydrostatic or leak testing. The consequences of a steam leak under pressure testing conditions with Secondary Containment OPERABLE will be conservatively bounded by the consequences of the main steam line break outside of Secondary Containment accident analysis described in Reference
- 1. Therefore, requiring Secondary Containment to be OPERABLE will conservatively ensure that any potential airborne radiation from steam leaks will be filtered through the Standby p%/ (continued)
ABWR. B 3.10-1 5/31/89
i Inservice Leak and Hydro Test j B 3.10.1 l 1 BASES (continued) j A:'PLICABLE Gas Treatment System (SGTS) to limit radiation releases to the SAFETY environment. ANALYSES (continued) In the event of a large primary system leak, the reactor vessel would rapidly depressurize allowing the low-pressure core cooling systems to operate. The capability of the Low Pressure Core Flooder Mode of the three RHR subsystems would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred. For the purposes of this test the protection provided by normally reouired MODE 4 applicable LCOs, in addition to the Secondary Containment requirements required to be met by this Special Operations LCO, will maintain acceptable plant operations. LC0 As described in LC0 3.0.8, compliance with this Special Operations LC0 is optional. Operation at greater than 200*F can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. However, this option may be required due to PT limits which require greater than 200*F testing while the ASME inservice test itself requires the Safety / Relief Valves (S/RVs) to be gagged preventing their OPERABILITY. If it is desired to perform this test while complying with this Special Operations LCO, then the MODE 4 applicable LCOs and specified MODE 3 LCOs must be met. This Special Operations LCO allows changing Table 1.1-1 temperature limits for MODE 4 to be "any temperature". The additional requirements for Secondary Containment LCOs to be met will provide sufficient protection for operations > 200*F for the purposes of performing an inservice leak or hydrostatic tests. This allows primary containment to be open for frequent, unobstructed access to perform inspections, and for outage activities on various systems to continue consistent with tb2 MODE 4 applicable requirements which are in effect immediately prior to and immediately after this operation. (continued) 1 O ABWR B 3.10-2 5/31/89 ! - _ _ _ ___ L
Inservice Leak and Hydro Test B 3.10.1
~
fS i BASES (continued)
' APPLICABILITY Modifying MODE 4 allowances such that thir operation can be considered as MODE 4 even though reactor. coolant temperature is greater than 200*F is adequately protected by the low decay heat level of the core and low stored energy as well as the water solid condition. The additional requirement for Secondary Containment OPERABILITY provides conservatism to the response of the facility to any event which may occur.
Operations in all other MODES are unaffected by this LCO.
' ACTIONS A.1. A.2.1. A.2.2 If an-LC0 specified in LCO 3.10.1 is not met, the ACTIONS applicable. to the stated requirements are entered and complied with. The Note to Required Action A.1 clarifies the intent of another LCO's Required Action to be in MODE 4 reducing the average reactor coolant temperature to s 200*F.
If a specified LC0 is not met and its ACTIONS are also not met, Required Action A.2.1 and A.2.2 are provided to. restore compliance with the normal MODE 4 requirements and exit this-Special Operations LC0's Applicability. Activities which'could-further increase reactor coolant temperature or pressure'are r- suspended immediately and temperature reduced to establish normal MODE 4 requirements. ( Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Actions. SURVEILLANCE- SR 3.10.1.1 REQUIREMENTS The LCOs made applicable are required to have their Surveillance met to establish the LC0 is being met. REFERENCES 1. ABWR SSAR, Section 15.6.4. (aD ABWR B 3.10-3 5/31/89
Reactor Mode Switch Interlock Testing B 3.10.2 B 3.10 SPECIAL OPERATIONS B 3.10.2 Reactor Mode Switch Interlock Testino BASES BACKGROUND The Reactor Mode Switch is a conveniently located, multi-position, keylock switch provided to select the necessary scram functions for various plant conditions (Ref. 1). The Reactor Mode Switch selects the appropriate trip relays for scram functions and provides appropriate bypasses. The mode switch positions and related scram interlock functions are summarized as follows:
- a. SHUTDOWN - Initiates a reactor scram; bypasses main steam line isolation and reactor high water level scrams.
- b. REFUEL - Selects Neutron Monitoring System (NMS) scram for low neutron flux level operation (but does not disable the Average Power Range Monitor (APRM) scram); bypasses main steam line isolation and reactor high water level scrams,
- c. STARTUP - Selects NMS scram for low neutron flux level operation (Startup Range Neutron Monitors (SRNMs);
bypasses main steam line isolation and reactor high water level scrams.
- d. RUN - Selects NMS scram for power range operation.
The Reactor Mode Switch also provides interlocks for such functions as control rod blocks, low CRD charging water header pressure trip bypass, refueling interlocks, and MSIV isolations. Operation of the Reactor Mode Switch from one position to another may be required to confirm certain aspects of these various interlocks during periodic tests and calibrations. APPLICABLE The interlock functions of the Reactor Mode Switch in MODES 3,
$AFETY 4 and 5 are provided to preclude reactivity excursions which ANALYSES could potentially result in fuel i tilure. With all control rods inserted and no CORE ALTERATIONS in progress there are no credible mechanisms for unacceptable reactivity excursions.
(continued) O ABWR B 3.10-4 5/31/89 [ _
i 1 Reactor Mode Switch Interlock Testing B 3.10.2
]
v. BASES (continued)- LC0 As described in LC0'3.0.8, compliance with.this Special-Operations LCO.is optional. MODE 3, 4 and 5 operations not , e specified in Table 1.1-1 can be performed in accordance with 1 other LCOs (i.e., LC0 3.10.3, LC0 3.10.4, LCO 3.10.7) without ! meeting this LC0 or its ACTIONS. . If testing which -involves the j Reactor Mode Switch interlocks and requires its repositioning beyond that specified in Table 1.1-1 for the current MODE of l z operation is desired, it can be performed provided all interlock functions potentially defeated are administrative 1y controlled. In MODES 3, 4 and 5 with the Reactor Mode Switch in Shutdown per Table 1.1-1, all control rods are fully inserted and a control rod block initiated. Therefore, all control rods must be verified fully inserted while in MODES 3, 4 and 5 with the Reactor Mode Switch in other than Shutdown. The-additional .LCO requirement to' preclude CORE ALTERATIONS is appropriate for MODE 5 operations, as discussed below, and is inherently met in MODES 3 and 4 by the definition of CORE ALTERATIONS which do not exist with the vessel head in place. In MODE 5, with the Reactor Mode Switch in Refuel, only one control rod (or control rod pair in special. situations) can be withdrawn under the Refuel Position one-rod-out interlock (LC0 3.9.2) and the refueling equipment interlocks (LC0 3.9.1) appropriately control other CORE ALTERATIONS. Due to the increased potential for error in controlling these multiple O(./ interlocks and the limited duration of tests involving Reactor Mode Switch position, conservative controls are required consistent with MODE 3 and 4 operations. These controls will adequately ensure the reactor. does not become critical during these tests. APPLICABILITY Periodic Reactor Mode Switch related interlock testing required while in MODES 1 and 2 is intended to be capable of being performed without the need for Special Operations exceptions. Mode switch manipulations in these MODES would likely result in plant trips. In MODES 3, 4 and 5 the interlock functions providec' by the Reactor Mode Switch in Shutdown (i.e., all control rods inserted and incapable of withdrawal), and Refueling (i.e., refueling interlocks to prevent inadvertent criticality during CORE ALTERATIONS) can be adequately administratively controlled during the performance of certain tests. (continued) ABWR B .10-5 5/31/89 t _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
Reactor Mode Switch Interlock Testing B 3.10.2 BASES (continued) ACTIONS A.1. A.2. A.3.1. A.3.2 These Required Actions are provided to restore compliance with the Technical Specifications changed by this Special Operations LCO. Restoring compliance will also result in concurrently exiting the Applicability of Special Operations LC0 3.10.2. CORE ALTERATIONS are immediately (refer to Section 1.3, Completion Times for discussion of immediately) suspended if in progress. This will preclude potential mechanisms which could lead to criticality. One hour is allowed to fully insert all insertable control rods. Placing the Reactor Mode Switch to Shutdown will ensure all inserted control rods remain inserted and result in operation in accordance with Table 1.1-1. Alternatively, if in MODE 5, the Reactor Mode Switch must be placed in the Refuel position, which will also result in operation in accordance with Table 1.1-1. Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. O SURVEILLANCE SR 3.10.2.1. SR 3.10.2.2 REQUIREMENTS Meeting the requirements of the Special Operations LCO maintains operation consistent with or conservative to operation with the Reactor Modo _ Switch in Shutdown (or Refuel for MODE 5). The function of the Reactor Mode Switch interlocks which are not in effect due to the testing in progress are adequately con.pensated for by the Special Operations LCO requirements. The administrative controls to ensure the operational requirements continue to be met, are to be periodically verified. The Surveillance performed at the 12 hour and 24 hour Frequency are intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with the Special Operations LC0 requirements. REFERENCES 1. ABWR SSAR, Section 7.2.1.1.6.2 O ABWR B 3.10-6 5/31/89 j
Control Rod Withdrawal - Hot Shutdown B 3.10.3 B 3.') SPECIAL OPERATIONS
) .
B 3.10.3 ' Control Rod ' Withdrawal - Hot Shutdowa BASES BACKGROUND In MODE 3 the Reactor Mode Switch is ir, Shutdown,. all control rods are inserted and. blocked from withdrawal. Many systems and functions are not required in these conditions due to the designed interlocks associated with the Reactor Mode Switch in Shutdown. However, circumstances arise while in MODE.3 which present the need to withdraw a single control rod or control rod pair for various tests, e.g., friction. tests,-scram timing, coupling integrity checks, etc. These single or dual control rod withdrawals can be accomplished by selecting the Refuel position for the Reactor Mode Switch. A control rod pair (those associated by a shared CRD hydraulic control unit) may be withdrawn by utilizing the Rod Test Switch, which " gangs"- the two rods together for rod position and control purposes. This Special Operations LC0 provides the appropriate additional controls to allow a single control rod, or control rod pair, withdrawal in MODE 3. With the Reactor Mode Switch in Refuel the analyses for control O
.V APPLICABLE SAFETY rod withdrawal during refueling are applicable and, provided ANALYSES the assumptions 'are met in MODE 3, these analyses will bound the consequences. Explicit safety analyses in the SSAR (Ref.
- 1) only assume the functioning of the refueling interlocks and adequate SHUTDOWN MARGIN to preclude unacceptable reactivity excursions.
Refueling interlocks restrict the movement of control rods to reinforce operational procedures which prevent the reactor from I becoming critical. These interlocks prevent the withdrawal of more than one control rod (or control-rod pair). The analyses assume this condition is acceptable since the core will be shutdown with the highest worth control rod pair withdrawn if adequate SHUTDOWN MARGIN exists. Control rod pairs have been established for each control rod drive hydraulic control unit (except 'or the center rod, which has its own accumulator). These pairs are selected and ana-lyzed so that adequate shutdown margin is maintained with any control rod pair fully withdrawn. When the rod test switch is used, the selected rod pair is subsituted for a single rod within the appropriate logic in order to satisfy the refuel mode one-rod-out interlock. The rod pair may then be withdrawn simultaneously. (continued) ABWR B 3.10-7 5/31/89 1 I
._ _. . _ _ _ _ __ _ _O
Control Rod Withdrawal - Hot Shutdown B 3.10.3 BASES (continued) APPLICABLE The control rod scram function provides backup protection e SAFETY to normal refueling procedures and the refueling interlocks ANALYSES which prevent inadvertent criticalities during refueling. Alternate backup protection can be obtained by assuring that a five-by-five array of control rods, centered on the withdrawn control rod (s), are inserted and incapable of withdrawal. LC0 As described in LCO 3.0.8, compliance with this Special Operations LC0 is optional. MODES 3 and 4 operation with the Reactor Mode Switch in Refuel can be performed in accordance with other LCOs (i.e., LC0 3.10.2, LC0 3.10.4) without meeting this Special Operations LC0 or its ACTIONS. If a single control rod or control rod pair withdrawal is desired in MODE 3, controls consistent with those required during Refueling must be implemented and this Special Operations LC0 ) applied. To back up the refueling interlocks (LCO 3.9.2 and 3.9.4), the ability to scram the withdrawn control rod (s) in the event of an inadvertent criticality is provided by this LCO's D.1 requirements. Alternately, provided a sufficient number of control rods in the vicinity of the withdrawn control rod (s) are known to be inserted and incapable of withdrawal, the possibility of criticality on withdrawal of these control rods is sufficiently precluded so as not to require scram capability of the withdrawn control rod (s). APPLICABILITY Control rod withdrawals are adequately controlled in MODES 1, 2 and 5 by existing LCOs. In MODES 3 and 4 control rod with-drawal is only allowed if performed in accordance with Special Operations LC0 3.10.3 or Special Operations LC0 3.10.4 and is limited to one control rod or control rod pair. This allowance is only provided with the Reactor Mode Switch in Refuel. 1 i During these conditions the One-Rod-Out interlock (LC0 3.9.2 and LC0 3.9.4) and scram functions (LC0 3.3.1.1 and LC0 3.9.5) or added administrative controls in D.2 of the Special Operations LCO provide mitigation of potential reactivity excursions. (continued) O ABWR 8 3.10-8 5/31/89
i Control Rod Mithdrawal - Hot Shutdown B 3.10.3 BASES (continued 1 ACTIONS A.1. A.2.1. A.2.2 l If an LCO specified in Special Operations LCO 3.10.3 is not , met, the ACTIONS applicable to the stated requirements are l entered as directed by Required Action A.I. The Note to Required Action A.1 clarifies the intent of any other LC0's Required Action to insert all control rods to'also require exiting this Special Operations Applicability by returning the Reactor Mode Switch to Shutdown. If a specified LC0 is not met and its ACTIONS as required by this Special Operations LC0's Required Action A.1 are not met, Required Action A.2.1 and A.2.2 are provided to restore compliance with the normal MODE 3 and 4 requirements and exit this Special Operations LC0's Applicability. One hour is allowed to fully insert all insertable control rods. Placing the Reactor Mode Switch in Shutdown will ensure all inserted rods remain inserted and restore operation in accordance with Table 1.1-1. Completion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the G Required Action. SURVEILLANCE SR 3.10.3.1. SR 3.10.3.2. SR 3.10.3.3 REQUIREMENTS The LCOs made applicable are required to have their Surveillance met to establish the Special Operations LC0 is being met. If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod (s) is not available, periodic verification is required to ensure the possibility of criticality remains precluded. REFERENCES 1. ABWR SSAR, Section 15.4.1. O ABWR B 3.10-9 5/31/89
Control Rod Withdrawal - Cold Shutdown B 3.10.4 8 3.10 SPECIAL OPERATIONS O B 3. d.4 Control Rod Withdrawal - Cold Shutdown BASES BACKGROUND In MODE 4, the Reactor Mode Switch is in Shutdown, all control rods are inserted and blocked from withdrawal. Many systems and functions are not required in this condition due to th, designed interlocks associated with the Reactor Mode Switch in Shutdown. However, circumstances arise while in MODE 4 which present the need to withdraw a single control rod or control rod pair for various tests, e.g., friction tests, scram timing, coupling integrity checks, etc. Situations may also require the removal of the associated control rod drive (s). These single or dual control rod withdrawals and possible subsequent removal can be accomplished by selecting the Refuel position for the Reactor Mode Switch. A control rod pair (those associ-ated by a single CRD hydraulic control unit) may be withdrawn by utilizing the Rod Test Switch, which " gangs" the two rods together for rod position and control purposes. This Special Operations LC0 provides the appropriate additional controls to allow a single or dual control rod withdrawal in MODE 4. APPLICABLE With the Reactor Mode Switch in Refuel the analyses for control SAFETY rod withdrawal during Refueling are applicable and, provided ANALYSES the assumptions are met in MODE 4, these analyses will bound the consequences. Explicit safety analyses in the SSAR (Ref.
- 1) only assume the functioning of the refueling interlocks and adequate SHUTDOWN MARGIN to preclude unacceptable reactivity excursions.
Refueling interlocks restrict the movement of control rods to reinforce operational procedures which prevent the reactor from becoming critical. These interlocks prevent the withdrawal of more than one control rod or control rod pair. The analyses assume this condition is acceptable since the core will be shutdown with the highest worth control rod pair withdrawn if adequate SHUTDOWN MARGIN exists. Control rod pairs have been established for each control rod drive hydraulic control unit (except for the center rod, which has its own accumulator). These pairs are selected and ana-lyzed so that adequate shutdown margin is maintained with any control rod pair fully withdrawn. When the rod test switch is used, the selected rod pair is substituted for a single rod within the appropriate logic in order to satisfy the refuel mode one-rod-out interlock. The rod pair may then be withrawn simultaneously. (continued) ABWR B 3.10-10 5/31/89
t Control Rod Withdrawal - Cold Shutdown L B 3.10.4 7 Y- BASES (continued) > 'O ' APPLICABLE The control rod scram function provides backup protection , SAFETY to normal refueling procedures and the refueling interlocks- l ANALYSES . which prevent inadvertent criticalities during refueling. (continued) Alternate backup protection can be obtained by assuring that a 1 five-by-five array of control rods, centered on the withdrawn control rod (s), are inserted and incapable of withdrawal. This alternate backup protection is required.when removing control-f: rod drives because the control- rod drive removal renders the withdrawn control rod (s) incapable of scram. LC0 As described in LCO 3.0.8, compliance with this Special Opera-tions LCO is optional. MODE 4 operation with the Reactor Mode Switen in Refuel can be performed-in accordance with 'other LCOs (i.e., LC0 3.10.2) without meeting this Special Operations LC0 or its ACTIONS.' If a single control rod or control rod pair withdrawal is desired-in MODE 4, controls consistent with those required during Refueling must be implemented and this Special Operations LC0 applied. The refueling interlocks of LC0 3.9.2 and LC0 3.9.4 will ensure only one control rod or control rod pair can be withdrawn. At the time CRD removal begins, the disconnection of the position indication probe (s) (PIP) will cause LCO 3.9.4 and, therefore, a,.s LC0 3.9.2 to not be met. At.this time a control rod withdrawal 4 D) block will be inserted to ensure no additional control rods can be withdrawn and compliance with the Special Operations LC0 is maintained. 1 To back up the refueling interlocks (LCO 3.9.2 and 3.9.4) or control rod withdrawal block, the ability to scram the with-drawn control rod (s) in the event of an inadvertent criticality is provided by this LCO's C.1 requirement. Alternatively, when the scram function is not OPERABLE or when a CRD is to be removed, a sufficient number of rods in the vicinity of the < withdrawn control rod are required to be inserted and incapable of withdrawal . This precludes the possibility of criticality on withdrawal of this control rod,- including its control rod 'l drive. i APPLICABILITY Control rod withdrawals are adequately controlled in MODES 1, 2 and 5 by existing LCOs. In MODES 3 and 4 control rod with-drawal is only allowed if performed in accordance with Special i Operations LC0 3.10.3 or Special Operations LC0 3.10.4 and is limited to one control rod or control rod pair. This allowance is oily provided with the reactor mode switch in Refuel. During these conditions the One-Rod-Out interlock (LC0 3.9.2 and LC0 3.9.4) and scram functions (LC0 [RPS Actuation O (continued) ABWP, B 3.10-11 5/31/89 i
Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES (continued) APPLICABILITY Instrumentation]'and LC0 3.9.5) or added administrative O (continued) controls in C.2 of the Special Operations LCO provide mitigation of potential reactivity excursions. ACTIONS A.1. A.2.1. A.2.2 If Special Operations LC0 3.10.4 requirements are not met, these Required Actions restore operation consistent with normal MODE 4 conditions (all rods inserted) or with the allowances of this Special Operations LCO. The actions are specified based on the condition of the control rod (s) being withdrawn. If a control rod is still insertable, the actions require the control rod to be inserted and the Reactor Mode Switch placed in Shutdown. If a control rod drive has been removed such that the control rod is not insertable, the actions require the most expeditious action be taken to either restore the CRD and insert its control rod, or restore compliance with the Special Operations LC0. SURVEILLANCE SR 3.10.4.1. SR 3.10.4.2. SR 3.10.4.3 SR 3.10.4.4 REQUIREMENTS The LCOs made applicable are required to have their Surveillance met to establish the Special Operations LCO is being met. If the local array of control rods is inserted and l disarmed while the scram function for a withdrawn rod is not ) available, periodic verification is required to ensure the possibility of criticality remains precluded. REFERENCES 1. ABWR SSAR, Section 15.4.1. J O ABWR B 3.10-12 5/31/89 i 1
Control Rod Drive Removal - Refueling-B 3.10.5 i (T B 3.10 SPECIAL OPERATIONS U B 3.10.5 Control Rod Drive Removal - Refuelina BASES BACKGROUND Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce opera-tional procedures which prevent the reactor from becoming critical during refueling operations. During refueling opera-tions, no more than one control rod or control rod pair is. permitted to be withdrawn from core cells containing one or more fuel assemblies. The refueling interlocks use the " Full In" position indicators to determine the position of all control rods. If the " Full In" position signal is not present for every control rod, then the all-rods-in permissive for the refueling equipment interlocks is not present and fuel loading is prevented. Also, the Refuel position one -rod-out interlock will not allow the withdrawal of a second control rod. A control rod drive pair (those associated by a shared CRD hydraulic control unit) may be removed under the control of the one-rod-out interlock by utilizing the Rod Test switch. This switch allows the CRD pair to be treated as one CRD for purposes of the one-rod-out interlock. (7 The control rod scram function provides backup protection to Ll normal refueling procedures and the refueling interlocks described above which prevent inadvertent criticalities during refueling. The requirement for this function to be OPERABLE precludes the possibility of removing the control rod drive " (CRD) once a control rod is withdrawn from a core cell contain-ing one or more fuel assemblies. This Special Operations LC0 provides controls sufficient to ensure the possibility of an inadvertent criticality is precluded while allowing a single CRD or control rod drive pair to be removed from core cells containing one or more fuel assemblies. The removal of a CRD involves disconnecting the position indication probe which causes noncompliance with LC0 3.9.4 and therefore LCOs 3.9.1 and 3.9.2. The CRD removal also requires isolation of the CRD from the CRD hydraulic system thereby causing inoperability of the control rod (LCO 3.9.5). APPLICABLE With the reactor mode switch in Refuel the analyses for control SAFETY rod withdrawal during refueling are applicable and, provided ANALYSES the assumptions are met, these analyses will bound the consequences. Explicit safety analyses in the SSAR (Ref. 1) only assume the functioning of the refueling interlocks and adequate SHUTDOWN MARGIN to preclude unacceptable reactivity
/- excursions.
(continued) ABWR B 3.10-13 5/31/89
l Control Rod Drive Removal - Refueling B 3.10.5 BASES (continued) APPLICABLE Control rod pairs have been established for each control rod SAFETY drive hydraulic control unit (except for the center rod, which ANALYSES has its own accumulator). These pairs are selected and (continued) analyzed so that adequate shutdown margin is maintained with any control rod pair fully withdrawn. When the rod test switch is used, the selected rod pair is substituted for a single rod within the appropriate logic in order to satisfy the refuel mode one-rod-out interlock. The rod pair may then be withdrawn simultaneously. Refueling interlocks restrict the movement of control rods and ! the operation of the refueling equipment to reinforce opera-tional procedures which prevent the reactor from becoming critical . These interlocks prevent the withdrawal of more than one control rod or control rod pair. The analyses assume this condition ir acceptable since the core will be shutdown with the highest worth control rod pair withdrawn if adequate SHUTDOWN MARGIN exists. By requiring all other control rods to be insertd and a control rod withdrawal block initiated, the , function of the inoperable one-rod-out interlock (LC0 3.9.2 and LC0 3.9.4) is adequately maintained. The Special Operations j LC0 requirement to suspend all CORE ALTERATIONS adequately 3 compensates for the inoperable all-rods-in permissive for the refueling equipment interlocks (LC0 3.9.1 and LC0 3.9.4). The control rod scram function provides backup protection to normal refueling procedures and the refueling interlocks which prevent inadvertent criticalities during refueling. Alternate backup protection can be obtained by assuring that a five-by-five array of control rods, centered on each withdrawn control rod, are inserted and incapable of withdrawal. LCO As described in LC0 3.0.8, compliance with this Special Operations LC0 is optional. MODE 5 operation with LC0 3.9.1, 3.9.2, 3.9.4 or 3.9.5 not met can be performed in accordance with those LCOs ACTIONS without meeting this Special Operations LC0 or its ACTIONS. If a single control rod drive or control rod drive pair removal from core cells containing one or mote fuel assemblies is desired in MODE 5, controls consistent with those required by LCOs 3.9.1, 3.9.2, 3.9.4 and 3.9.5 must be implemented. . l By requiring all other control rods to be inserted and a ; control rod withdrawal block initiated the function of the inoperable one-rod-out interlock (LC0 3.9.2 and LC0 3.9.4) is adequately maintained. The Special Operations LC0 requirement to suspend all CORE ALTERATIONS adequately compensates for the inoperable all-rods-in permissive for the refueling equipment (continued) ABWR B 3.10-14 5/31/89
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Control Rod Drive Removal - Refueling B 3.10.5
~) BASES (continued) s~-J .
LC0 interlocks (LC0 3.9.1 and LCO 3.9.4). Assuring the five-by-(continued) five array of control rods, centered on each withdrawn control rod, are inserted and incapable of withdrawal adequately satisfies the backup protection which LCOs (RPS Actuation Instrumentation] and 3.9.5 would have provided. The allowance to assume this withdrawn control rod, or control rod pair, to be the highest worth control rod pair to satisfy LCO 3.1.1 and the inability to withdraw another control rod during this operation without additional SHUTDOWN MARGIN demonstrations is conservative (i.e., the withdrawn control rod pair may not be the highest worth control rod pair). APPLICABILITY MODE 5 operations are controlled by existing LCOs. The allowance to comply with this Special Operations LC0 in lieu of the ACTIONS of LC0 3.9.1, 3.9.2, 3.9.4 or 3.9.5 is appropri-ately controlled with the additional administrative controls required by Special Operations LC0 3.10.5 which provide : mitigation of potential reactivity excursions. ACTIONS A.1. A.2.1. A.2.2 lx-,') If Special Operations LCO 3.10.5 requirements are not met, these Required Actions restore operation consistent with the normal requirements for failure to meet LCOs (RPS Instrumentation], 3.9.1, 3.9.2, 3.9.4 and 3.9.5 (all control rods inserted) or with the allowances of this Special Operations LCO. The Completion Time for Required Actions A.2.1 and A.2.2 are intended to require the most expeditious action be taken to either restore the CRD(s) and insert its control rod (s), or restore compliance with the Special Operations LCO. SURVEILLANCE SR 3.10.5.1. SR 3.10.5.2. SR 3.10.5.3. SR 3 10.5.4. SR 3.10.5.5 REQUIREMENTS Periodic verification of the administrative controls established by the LC0 is prudent to ensure the possibility of an inadvertent criticality remains precluded. REFERENCES 1. ABWR SSAR, Section 15.4.1. f"
,s ABWR B 3.10-15 5/31/89
Hultiple Control Rod Withdrawal - Refueling B 3.10.6 B 3.10 SPECIAL OPERATIONS B 3.10.6 Multiple Control Rod Withdrawal - Refuelino BASES BACKGROUND Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce opera-tional procedures which prevent the reactor from becoming critical during refueling operations. During refueling opera-tions a total of no more than one control rod or control rod pair is permitted to be withdrawn from all core cells contain-ing one or more fuel assemblies. When all four fuel assemblies are removed from a cell, the control rod may be withdrawn with no restrictions. Control rods withdrawn from cells which do not contain fuel assemblies have a negligible impact on reac-tivity control and are therefore not required to be OPERABLE (LCO 3.9.5). Any number of control rods may be withdrawn and removed from the reactor vessel if their cells contain no fuel. The refueling interlocks use the " Full In" position indicators to determine the position of all control rods. If.the " Full In" position signal is not present for every control rod, then the all-rods-in permissive for the refueling equipment inter-locks is not present and fuel loading is prevented. Also, the Refuel position one-rod-out interlock will not allow the withdrawal of additonal cor. trol rods. To allow more than one control rod (pair) to be withdrawn during refueling, these interlocks must be defeated. This Special Operations LC0 establishes the necessary controls to allow bypassing the " Full In" position indicators. APPLICABLE Explicit safety analyses in the FSAR (Ref.1) only assume the SAFETY functioning of the refueling interlocks and adequate SHUTDOWN ANALYSES MARGIN for the prevention and mitigation of prompt reactivity excursions during refueling. To allow multiple (e.g. , more than one control rod or control rod pair) control rod with-drawals, control rod removals, associated control rod drive (CRD) removal or any combination, the " Full In" position indication is allowed to be bypassed for each withdrawn control rod if all fuel has been removed from the cell. With no fuel assemblies in the core cell, the control rod has no reactivity control function and is not required to remain inserted. However, prior to reloading fuel into the cell, the control rod must be inserted to ensure an inadvertent criticality does not occur, as evaluated in the Reference 1 analysis. (continued) ABWR B 3.10-16 5/31/89 __ -_. 1
l Multiple Control Rod Withdrawal - Refueling l i-B 3.10.6 j'~k B_ASES-(continued) q LCO' As described in 'LCO 3.0.8 compliance with this Special '. Operations LC0 is optional. MODE 5 operation with " Full In"' ) u indicators not OPERABLE can be in accordance with LC0 3.9.4 l without meeting this Special Operations LC0 or its ACTIONS. If multiple control rod withdrawal or removal or CRD removal is desired, all four fuel assemblies are required to be removed j l from the associated cells. (Note - one cell with a withdrawn - control' rod, or two cells for a control rod pair, are' allowed to have fuel remaining in accordance with other. MODE 5 i applicable LCOs.) APPLICABILITY MODE 5 operations are controlled by existing LCOs. The allow-ance to comply with this Special Operations LCO in lieu of the ACTIONS of LC0 3.9.4'is appropriately controlled by requiring all fuel to be removed from cells with " Full In" indicators.not [ functioning. ACTIONS A.1. A.2.1. A.2.2
- If Special Operations LC0 3.10.6 requirements are not met, I\ these Required Actions restore operation consistent with the normal requirements for refueling (all control rods inserted or OPERABLE in core cells containing one or more fuel assemblies) or with the allowances of this Special Operations LCO. The- l Completion Time for Required Actions A.2.1 and A.2.2 are intended to require the most expeditious action be taken to-either restore the CRD(s) and insert the control rod (s), or ,
restore compliance with the Special Operations LCO. SURVEILLANCE SR 3.10.6.1. SR 3.10,02 REQUIREMENTS Periodic verification of the administrative controls estab-lished by the Special Operations LCO is prudent to ensure the i l possibility of an inadvertent criticality remains precluded. REFERENCES 1. ABWR SSAR, Section 15.4.1. l l (% l i ABWR B 3.10-17 5/31/89 _ _ _ -_ _ _ _ _ _ __O
Control Rod Testing - Operating B 3.10.7 { B 3.10 SPECIAL OPERATIONS B 3.10.7 Control Rod Testina - Operatina BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the Rod Worth Minimizer (RWM) (LC0 [ Rod Block Instrumentation]) such that only specified control rod l sequences (LC0 3.1.6) and relative positions are allowed over the operating range from all control rods inserted to the Low Power Setpoint (LPSP) of the RWM. The sequences effectively limit the potential amount and rate of reactivity increase that could occur during Rod Withdrawal Error (RWE). During these conditions,_ control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LC0 3.1.6. These tests may include SHUTDOWN MARGIN demonstrations, control rod scram time testing, control rod friction testing and testing performed during the Startup Test Program. This Special Operations LC0 provides the necessary exemption to the requirements and provides additional controls to allow the deviations from the prescribed sequences. APPLICABLE The analytical methods and assumptions used in evaluating the - SAFETY RWE are summarized in References 1 and 2. RWE-analyses assume ANALYSES the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the RWE analyses. The RWM provides backup to operator control of the withdrawal sequences to ensure the initial conditions of the RWE analyses are not violated. For special sequences developed for control rod testing, the initial control rod patterns assumed in the safety analysis of References 1 and 2 may not be preserved and therefore special RWE analyses are required to demonstrate the special sequence will not result in unacceptable consequences should a RWE occur during the test- ) ing. These analyses are dependent on the specific test being l performed. {
\
(continued) O ABWR B 3.10-18 5/31/89 j
Control Rod Testing - Operating i B 3.10.7
' "') BASES (continued)
V As described in LC0 3.0.8, compliance with this Special LC0 Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LC0 3.1.6 and during these tests no exemptions to the requirements of LC0 3.1.6 are necessary. For testing performed with a sequence not in compliance with LC0 3.1.6, the requirements of LC0 3.1.6 may be suspended provided additional controls are placed on the test to ensure the assumptions of the special safety analysis for the test sequence are satisfied. WL n deviating from the prescribed sequences of LC0 3.1.6, individual control rods must be bypassed in the Rod Control and Instrumentation System (RCIS). Assurances that the test sequence is followed can be provided by verifying conformance to the approved test sequence by a second licensed operator or other qualified member of the technical staff. These controls are consistent with those normally applied to operation in the startup range as defined in the SR [ ] when it is necessary to deviate from the prescribed sequence (e.g., an inoperable control rod that must be fully inserted). APPLICABILITY Control rod testing during MODES 1 and 2 with THERMAL POWER
.w greater than the LPSP of the RWM is adequately controlled by
; the existing LCOs on power distribution limits and control rod
'- ) block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LC0 3.2.1, LC0 3.2.2, LC0 3.2.3 and LC0 [ Rod Block Instrumentation). With THERMAL POWER less than or equal to the LPSP of the RWM, the provisions of this Special Operations LC0 are necessary to perform special tests which are not in conformance with the prescribed sequences of LC0 3.1.6. During MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LC0 3.10.3 or Special Operations LC0 3.10.4 which provide adequate controls to ensure the assumptions of the safety analysis of Reference 1 and 2 are satisfied. During these special operations and during MODE 5, the One-Rod-Out interlock (LC0 3.9.2 and LC0 3.9.4) and scram functions (LC0
[RPS Actuation Instrumentation] and LC0 3.9.5) or added administrative controls prescribed in the applicable Special Operations LCOs provide mitigation of potential reactivity excursions. (continued)
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ABWR B 3.10-19 5/31/89 i _ _ _ - - _ _ _
i Control Rod Testing - Operating ! B 3.10.7 i BASES (continued) l 1 ACTIONS Ad j 1 With the requirements of the Special Operations LC0 not met (e.g., control rod pattern is not in compliance with the , special test sequence), the testing is required to be 1 immediately suspended. Upon suspension of the special test, the provisions of LC0 3.1.6 are no longer exempted and appropriate actions are specified to restore the control rod sequence to the prescribed sequence of LC0 3.1.6, or shutdown. SURVEILLANCE SR 3.10.7.1 REQUIREMENTS Ouring performance of the special test, a second licensed operator or other qualified member of the technical staff is required to verify conformance with the approved sequence for the test. This verification must be performed during control . rod movement to prevent deviations from the specified sequence. As such, a periodic frequency is not appropriate for this surveillance. This surveillance provides adequate assurance that the specified test sequence is being followed and is also supplemented by SR [ ] which requires verification of the bypassing of control rods in RCIS and subsequent movement of these control rods. REFERENCES 1. NEDE-240ll-P-A-9-US, " General Electric Standard Application for Reactor Fuel," Supplement for United States, September 1988.
- 2. ABWR SSAR, Section 15.4.1.
O ABWR B 3.10-20 5/31/89 i
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