ML20065J579
| ML20065J579 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/11/1994 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19304B983 | List: |
| References | |
| NUDOCS 9404180300 | |
| Download: ML20065J579 (75) | |
Text
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SSAR MODIFICATION PAGES O
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9404100300 940411 ,
23A6100 Rev. 2 ABWR StandardSafety Analysis Report O 'ie1 e< eie <e <ceeti eo)
Figure 3418 Test Sensors Unique to FS* and SS Tests. . . . . . . . . . 3B-54 Figure 3B-19 Envelope PSD at 019P for SST-1,2,3,9,11 and 12. . . ... . . . . . 3B-55 Figure 3B-20 Six-Test and Key-Segment Envelope PSDs at 019P . .. . . . . . 3 & 56 Figure 3421 ABWR CO Source Load Methodology.. . . .. .. . . 3B-57 Figure 3422 ABWR Typical Pressure Fluctuation Due to CO . . . . . . 3B-58 Figure 3423 Mark IH Typical Pressure Fluctuation Due to CO . ... . . 3B-59 Figure 3B-24 Typical Large Chug (025P) . .. . . . . .. .. . . . .3460 Figure 3B-25 PSD of Typical Large Chug (025P) . . . .. . 3B-61 Figure 3B-26 ABWR Chug Source Load Methodology . . . . 3B-62 Figure 3B-27 Comparison of PSDs Between Analysis and Test.. . . . . . . . . 3B-63 Figure 3B-28 Spatial Load Distribution for CH . .. .. . . . . . . . . . . .. . 3B-64 Figure 3B-29 . ... . 3B-65
{ Typical Wall Pressure Time History Due to CH.. .. . . . . . . . . . . .
Figure 3B-30 Circumferential Pressure Distribution on Access Tunnel Due to CH. . . .3466 Figure 3B-31 Typical Test Result of Upward Load on Horizontal Vent Due to CH .. . . 3B-67 Figure 3B-32 Typical Test Result of Moment Due on Horizontal Vent to CH.. . ...3468 Figure 3B-33 Horizontal Vent Upward Loading for Vent Pipe and Pedestal. . .3469 Figure 3B-34 Horizontal Vent Upward Loading for Structure Response Analysis.... . 3B-70 Figure 3E 1 Schematic Representation of MaterialJ-Integral R andJ-T Curves . ... . .. .. . 3E-3 8 Figure SE-2 Carbon Steel Test Specimen Orientation Code. .. . . . . .. . . . . 3E-39 l Figure 3E-3 Toughness Anisotropy of ASTM 106 Pipe (152 mm Sch. 80) . .. 3E-40 Figure 3E-4 Charpy Energies for Pipe Test Material as a Function of Orientation and Temperature.. .. .. . . . . . . . . . . . . . .. . . . . . . . . . 3 E-41 Figure SE-5 Charpy Energies for Plate Test Material as a Function of Orientation and Temperature. . . . . . . . . . ... . . . . . . . . . . . 3E-42 Figure 3E-6 Comparison of Base Metal. Weld and HAZ Charpy Energies for SA 333 Grade 6. . . . .. .. . . . . 3E-43 Figure 3E-7 Plot of 288 C True Stress-True Strain Curves for SA 333 Grade 6 Carbon l Steel . . . . . . . . . . . . 3E-44 List of Figwes - Amendment 32 3.0-xxix
23A6100 Rev. 2 ABWR St:nd:rd Safety Anzlysis Report List of Figures (Continued)
Figure SE-8 Plot of 288 C True Stress-True Strain Curves for SA 516 Grade 70 l Carbon Steel . . . . . .. . . . 3E-45 Figure SE-9 Plot of 117 C True Stress-True Strain Curves for SA 333 Grade 6 l Carbon Steel . . .. ... . . .. .. . 3E-46 Figure 3E-10 Plot of 177'C True Stress-True Strain Curves for SA 516 Grade 70 l Carbon Steel . . ... . . .... . . 3E-47 Figure 3E-11 Plot of 288 C TestJ-R Curve for Pipe Weld. . . 3E-48 Figure 3E-12 Plot of 288 CJmod, Tmod Data from TestJ-R Curve.. . . 3E-49 Figure 3E-13 Carbon SteelJ-T Ctuve for 216 C . .. . . . . . . . . . . . . . 3E-50 Figure 3E-14 Schematic Illustration of Tearing Stability Evaluation.. . . . . . . . 3E-51 Figure 3E-15 A Schematic Representation ofInstability Tension and Bending Stresses as a Function of Flaw Strength . . . . . . . . . . . . . . . . . . . . . . .. . 3E-52 Figure SE-16 SA 333 Grade 6 Stress-Strain Data at 288'C in the Ramberg-Osgood Format 3E-53 Figure SE-17 Carbon Steel Stress-Strain Data at 177'C in the Ramberg-Osgo ad Format . 3E-54 Figure 3E-18 Comparison of PICEP Predictions with Measured Leak Rate. .. . .3E-55 Figure 3E-19 Pipe Flow Model . . . .. . . . . . . .. .. . . .. . . 3E-56 Figure SE-20 Mass Flow Rates for Steam / Water Mixtures. . . . . . . . . . . . 3E-57 Figure SE-21 Friction Factors for Pipes. . . . . . . . . . . . . 3E-58
! Figure 3E-22 Leak Rate as a Function of Crack Length in Main Steam Pipe (Example). 3E-59 Figure SG-1 Horizontal Beam Model for AP Load . .. . . . . . . . . . . . 3G-8 l
Figure SG-2 Nodal Point (R/B Horizontal / Vertical Shell Model).. . . . ... . . . SG-9 l
Figure SG-3 Nodal Point No. (RPV/ Internal Vertical Shell Model) . . . . 3G-10 Figure SG-4 Floor Response Spectrum-Case: APA1, Node: 33, Horizontal.. ... . . . . SG-11 Figure SG-5 Floor Response Spectrum-Case: APA1, Node: 81, Horizontal., . . SG-12 l
Figure SG-6 Floor Response Spectrum-Case: APA1, Node: 85, Horizontal.. . 3G-13 l
Figure SG-7 Floor Response Spectrum-Case: APA2, Node: 33, Horizontal.. . . . . 3G-14 Figure SG-8 Floor Response Spectrum-Case: APA2, Node: 81, Horizontal.. . SG.15 Figure SG-9 Floor Response Spectrum-Case: APA2, Node: 85, Horizontal.. . . . SG-16 l
l 3.0 xxx List of Figures - Amendment 32
^ 'x Rcv 1 ABWR Standard Safety Analpsis Repcit
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O J J INSTABILITY POINT l INSTABILITY l STRESS OR LOAD i
l (JAPP, IAPP)
STRESS OR LOAD T b) c Figure 3E-14 Schematic illustration of Tearing Stability Evaluation Guidelines for LBB Application - Amendment 31 3E-51
23A6100 Rev.1 ABWR Standard Safety Analysis Report 9
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Ot M b p' _ _ _ ____
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4 a c,t a c,b FLAW LENGTH i
Figure 3E-15 A Schematic Representation of instability Tension and Bending Stresses as a Function of Flaw Strength 9
3E-52 Guidelines for LBB Application - Amendment 31
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23A6100 Rev. 2 ABWR s nd:rdsar:tyArctysisa:prit O
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0 0.1 0.2 0.3 0.4 0.50.6 0.7 0.8 0.9 1.0 1.1
/ i LOG E glT l l Figure 3E-16 SA 333 Grade 6 Stress Strain Data at 288 C in the Ramberg Osgood Format i l
Guide:ines for LBB Application --Amendment 32 32-53 .l
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- g- r-- , _ _ _ __ + m
23A6100Rsv 2 ABWR standardsafety Analysis Report O
STRESS STRAIN DATA AT 177*C l PlPE SA 333 GR6 PLATE SA 516 GR 70 PIPE SA 106 GR 8 (CE DATA) 100 -
a = 4.5, n = 4.7 a = 5.0 n = 4.0 12 10 9
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6 a = 2.2, n = 3.5 5 4
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1 0 0.1 0.2 0.30.4 0.5 0.60.7 0.8 0.9 1.0 1.1
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Carbon Steel Stress-Strain Data at 177 C in the Ramberg-Osgood Format 3&S4 Guidelines for LBB Application -Amendment 32
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23A6100 Rev. 4
' ABWR '
standardSafetyAnalysis Report 1
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-I Figure 601 Injection Flow HPCF Analysss Outlines - Amendment 34 GD-3/4 1
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23A6100 Rev. 4 ABWR StandardSaktyAnalysis Report
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Certain localized events are evaluated at other than the above-mentioned conditions.
These conditions are discussed pertinent to the appropriate event.
The power / flow operating map for a plant may differ from that used in the system response analysis given in this chapter. Differences in the map will not change the designation oflimiting events.The operadng map used at a plant will be provided by l the COL applicant to the USNRC for information (Subsection 4.4.2.1).
15.0.4.5 Evaluation of Results The results of the system response analyses are presented in Table 15.0-2. Based on.
these results, the limidng events have been idendfied. Reasons why the other events are not limidng are given in the event documentadon. The limidng events which establish CPR operating limit include:
(1) Limiting Pressurization Events: Inadvertent closure of one turbine control valve and generator load rejection with all bypass valve failure.
(2) Limiting Decrease in Core Coolant Temperature Events: Feedwater Controller Failure--Maximum Demand )
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\ For the core loading in Figure 4.S1, the resulting initial core MCPR operadng limit is j
1.17. The operating limit based on the plant loading pattern will be provided by the COL applicant to the USNRC for information (Subsecdon 15.0.5.2 for COL license !
information requirement).
Results of the transient analyses for individual plant reference core loading patterns will differ from the results shown in this chapter. However, the relative results between core associated events do not change. Therefore, only the results of the identified limiting l events given in Table 15.0-4 will be provided by the COL applicant to the USNRC for information (Subsection 15.0.5.1).
15.0.4.5.1 Effect of Single Failures and Operator Errors The effect of a single equipment failure or malfunction or operator error is provided in -
Appendix 15A.
15.0.4.5.2 Analysis Uncertainties The analysis uncertainties meet the criteria in Appendix 4B.
A summary of applicable accidents is provided in Table 15.0-5, which compares GE calculated amount of failed fuel to that used in worst-case radiological calculations for I the core shown in Figure 4.SI. Radiological calculations for a plant initial core will be
$ provided by the utility to the USNRC for information (see Subsection 15.0.5 for COL license information requirements).
Accident and Analysis- Amendment 34 15.0-5
23A6100 Rev. 4 ABWR Standard SafetyAnalysis Report O
15.0.4.5.3 Barrier Performance The significant areas ofinterest forinternal pressure damage are the high-pressure portions of the reactor coolant pressure boundary (i.e., the reactor vessel and the high pressure pipelines attached to the reactor vessel). The plant shall meet the criteria in Appendix 4B.
15.0.4.5.4 Radiological Consequences This chapter describes the consequences of radioactivity release for the core loading in Figure 4.3-1 during three types of events: (1) incidents of moderate frequency (anticipated operational occurrences); (2) infrequent incidents (abnormal operational occurrences); and (3) limiting faults (design basis accidents). For all events whose consequences are limiting, a detailed quantitative evaluation is presented. For nonlimiting events, a qualitative evaluation is presented or results are referenced from
- a more limiting or enveloping case or event.
15.0.5 COL License Inforrnation 15.0.5.1 Anticipated Operational Occurrences (AOO) l The results of the events identified in Subsection 15.0.4.5 for plant core loading will be provided by the COL applicant referencing the ABWR design to the USNRC for information.
15.0.5.2 Operating Limits The operating limit resulting from the analyses normally provided in this subsection will be provided by the COL applicant referencing the ABWR design to the USNRC for information.
15.0.5.3 Design Basis Accidents Results of the design basis accidents, including radiological consequences, will be prmided by the COL applicant referencing the ABWR design to the USNRC for information.
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e 1S.0 6 Accident and Anatysis - Amendment 34 l
23A6100 Rev. 4 ABWR standard safety Analysis Report m
reactor. The low level (L3) scram uip function meets the single-failure criterion. Four of the RIPS are tripped at I.evel 3.
15.2.7.3 Core and System Performance The results of this transient simulation are presented in Figure 15.2-12. Feedwater flow terminates at approximately 5 seconds. Subcooling decreases, causing a reducdon in core power level and pressure. As power levelis lowered, the turbine steam flow starts to drop otT because the pressure regulator is attempting to maintain pressure for the first 10 seconds. Water level continues to drop until, first, the recirculation flow is runback at Level 4 (L4) and then the vessel level (L3) scram trip setpoint is reached, whereupon the reactor is shut down and the four RIPS are tripped. Vessel water level continues to drop to the L2 trip. At this time, the remaining six RIPS are tripped and the RCIC operation is initiated. MCPR remains considerably above the safety limit, because increases in heat flux are not experienced. Therefore, this event does not have to be reanalyzed for specific core configurations.
15.2.7.4 Barrier Performance The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel or containment are designed;
\ therefore, these barriers maintain their integrity and ftmetion as designed.
15.2.7.5 Radiological Consequences The consequences of this event do not result in any fuel failure. Therefore, no analysis of the radiological consequences is required.
15.2.8 Feedwater Line Break Refer to Subsection 15.6.6.
15.2.9 Failure of RHR Shutdown Cooling The RIIR System performs low pressure core cooling, containment heat removal, containment spray and shutdown cooling functions. The RHR System has three independent divisions, each of which contains the necessary piping, pumps, valves, heat exchangers, instrumentation and electrical power for operation. Each division also has its own cooling water supply, diesel generator and room cooling system. For the shutdown cooling function, each division has its own suction line from and return line to the RPV. Thus, each of the three RHR divisions is completelyindependent of the other divisions in its shutdown cooling function. The RHR System reduces the primary system temperature to 51.7 C within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of plant shutdown.
O Q Normally,in evaluating component failure considerations associated with R11R System shutdown cooling mode operadon, active pumps, valves or instrumentation would be Increase in Reactor Pressure - Amendment 34 15.2 25
23A6100 Rsv. 3 ABWR St:nd:rd Stty An:Iysis R:p:rt assumed to fail. If the single active failure criterion is applied to the RHR System, one 9
of the three RHR divisions would be inoperable. However, the two operable RHR divisions could achieve cold shutdown to 100 C within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after reactor shutdown.
Failure of ofTsite power is another case which could affect the shutdown cooling l function. The plant will have two independent offsite power supplies. If either or both offsite power supplies are lost, each RHR division has its own diesel generator which will pennit operating that division at its rated capacity. Application of the single active failt te criterion would stillleave two RHR divisions operational.
The RHR System description and perfonnance evaluation in Subsection 5.4.7 describes the models, assumptions and results for shutdown cooling with two RHR dhisions operational.
15.2.10 COL License information 15.2.10.1 Radiological Effects of MSIV Closures COL applicants will evaluate the radiological effect of the inadvertent closure of MSIVs for the final plant design and the site parameters (Subsection 15.2.4.5.3) .
15.2.11 References 15.2-1 F. G. Brutshscy, et al., Behavior ofIodine in Reactor Water Dun'ng Plant Shutdown and Startup, August 1972 (NEDO-10585).
15.2-2 H. Careway, V. Nguyen, and P. Stancavage, Radiological Accident-The CONAC03 CODE, December 1981 (NEDO-21143-1).
O 15.2-26 increase in Reactor Pressure - Amendment 33
23A6100 Rev. 3 ABWR StandantSa!ctyAnalysis Report O
15.3 Decrease in Reactor Coolant System Flow Rate 15.3.1 Reactor Internal Pump Trip 15.3.1.1 Identification of Causes and Frequency Classification 15.3.1.1.1 Identification of Causes Reactorinternal pump (RIP) motor operation can be tripped offby design for intended i reduction of other transient core and RCPB effects, as well as randomly by unpredictable operational failures. Intentional tripping will occur in response to:
(1) Reactor vessel water level L3 setpoint trip (4 RIPS) i (2) Reactor vessel water level L2 setpoint trip (the other 6 RIPS) l (3) TCV fast closure or stop valve closure (the same 4 RIPS as L3 trip) i (4) High pressure setpoint trip (the same 4 RIPS as L3 trip) l (5) hiotor overcurrent protection (single pump)
D g (6) hiotor overload and short circuit protection (single pump)
Random tripping will occur in response to:
(1) Operator error.
(2) Loss of electrical power source to the pumps.
(3) Equipment or sensor failures and malfunctions which initiate the above intended trip response. However, all trip logics use redundant digital designs.
Single failures in the UAT or MPT and/or their protection circuits can result in loss of preferred power source to the plant.
Thus, the worst single-failure event is a loss of electrical power bus, which supplies power to RIPS. Since four buses are used to supply power to the RIPS, the worst single failure l can only cause three RIPS to trip.
A loss of AC power to station auxiliaries may cause RIPS to trip. However, not all RIPS would be tripped at the same time due to the M-G sets. Transients caused by a loss of AC power are discussed in Subsection 15.2.6.
The effect of an additional single failure on this event (i.e., trip of three RIPS) is the tripping of additional RIPS. For example,if an additional power bus fails at the same j time, the number of RIPS tripped are five or six, instead of three. However, the Decrease in Reactor Coolant System flow Rate - Amenament 33 15.3-1
23A6100 Rev. 3 ABWR Stand:rdStty Ar:Iysis R:p:rt probability of this occurring is low . This event should be classified as a limiting fault. In 9
this analysis, the trip of all RIPS is provided to bound the events oflow probability.
I When a mpid core flow reduction caused by a trip of all RIPS is sensed, a reactor scram !
is initiated to terminate the power generation. The core flow reduces rapidly due to the j relatively small inertia of the RIPS. However, natural circulation is still available to keep the reactor core covered and cooled.
15.3.1.1.2 Frequency Classification 15.3.1.1.2.1 Trip of Three Reactor Internal Pumps This transient event is categodzed as one of moderate frequency.
15.3.1.1.2.2 Trip of All Reactor Internal Pumps This event is categorized as an infrequent low probability event with special acceptance for fuel failure (see Subsection 15.3.1.5.2) .
15.3.1.2 Sequence of Events and Systems Operation 15.3.1.2.1 Sequence of Events 15.3.1.2.1.1 Trip of Three Reactor internal Pumps Table 15.Sl lists the sequence of events for Figure 15.SI.
15.3.1.2.1.2 Trip of All Reactor Internal Pumps Table 15.32 lists the sequence of events for Figure 15.S2.
15.3.1.2.1.3 Identification of Operator Actions 15.3.1.2.1.3.1 Trip of Three Reactor Internal Pumps Because no scram occurs for trip of three RIPS, no immediate operator action is required. As soon as possible, the operator should verify that no operating limits are being exceeded. The operator should also determine the cause of failure prior to returning the system to normal operation.
15.3.1.2.1.3.2 Trip of All Reactor Internal Pumps The operator should ascertain that the reactor scram is initiated. If the main turbine and feedwater pumps are tripped resulting from reactor water level swell, the operator should regain control of reactor water level through RCIC operation, monitoring reactor water level and pressure after shutdown. When both reactor pressure and level are under control, the operator should secure RCIC as necessary. The operator should also determine the cause of the trip prior to retuming the system to normal operation.
15.3-2 Decrease in Reactor Coolant System Flow Rate - Amendment 33
23A6100 Rtv.1 Stand:rd S:lety Analysis Reprt ABWR G transients. Maintenance planned for performance during refueling outages must be conducted in such a way that it will have litde or no impact on plant safety, on outage length or on other maintenance work.
The COL applicant will provide a complete O-RAP to be reviewed by the NRC. See Subsection 17.3.13.3 for COL license information 17.3.10 Owner / Operator's Reliability Assurance Program The O-RAP that will be prepared and implemented by the ABWR owner / operator will make use of the information provided by GE-NE. This information will help owner / operator determine activities that should be included in the O-RAP. Examples of elements that might be included in an O-RAP are:
(1) Reliability Performance Monitoring: Measurement of the performance of equipment to determine that it is accomplishing its goals and/or that it will continue to operate with low probability of failure.
(S) Reliability Methodologn Methods by which the plant owner / operator can compare plant data to the SSC data in the PRA.
(3) Problem Prioritization: Identification, for each of the risk- significant SSCs, of the importance of that item as a contributor to its system unavailability and assignment of priorities to problems that are detected with such equipment.
(4) Root Cause Analysis: Determination, for problems that occur regarding reliability of risk-significant SSCs, of the root causes, those causes which, after correction, will not recur to again degrade the reliability of equipment.
(5) Corrective Action Determination: Identification of corrective actions needed to restore equipment to its required functional capability and reliability, based on the results of problem identification and root cause analysis.
(6) Corrective Action Implementation: Carrying out identified corrective action on risk-significant equipment to restore equipment to its intended function in such a way that plant safety is not compromised during work.
(7) Corrective Action Verification: Post-corrective action tasks to be followed after maintenance on risk significant equipment to assure that such equipmentwill perform its safety functions.
(8) Plant Aging: Some of the risk-significant equipment is expected to undergo age related degradation that will require equipment replacement or refurbishment.
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17.3 5 Reliability Assurance Program During Design Phase - Amendment 31
23A6100 Rev. 4 ABWR StandardSafetyAnalysis Report O
(9) Feedback to Designer: The plant owner / operator will periodically compare performance of risk-significant equipment to that specified in the PRA and D-RAP, as mentioned in item 1, above, and, at its discretion, may feedback SSC performance data to plant or equipment designers in those cases that consistently show perfonnance below that specified.
(10) Programmatic Interfaces: Reliability assurance interfaces related to the work
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i of the several organizations and personnel groups working on risk-significant SSCs.
l The plant owner / operator's O-RAP will address the interfaces with constmction, startup testing, operations, maintenance, engineering, safety, licensing, quality assurance and procurement of replacement equipment.
17.3.11 D-RAP Implementation An example ofimplementation of the D-RAP is given for the Standby Liquid Control l
l System (SLCS). The purpose of the SLCS is to inject neutron absorbing poison into the reactor, upon demand, providing a backup rector shutdown capability independent of the control rods. The system is capable of operating over a wide range of reactor l pressure conditions. The SLCS may or may not be identified by the final PRA as a significant contributor to CDF or to offsite risk. For the purpose of this example, it is assumed that the SLCS is identified as a significant conuibutor to CDF or to offsite risk.
17.3.11.1 SLCS Description During normal operation, the SLCS is on standby, only to function in event the l operators are unable to control reactivity with the normal control rods. The SLCS consists of a boron solution storage tank, two positive displacement pumps, two motor operated injection valves (provided in parallel for redundancy), and associated piping ]
and valves used to transfer borated water from the storage tank to the reactor pressure I
vessel (RPV). The borated solution is discharged through the "If high pressure core flooder (IIPCF) subsystem sparger. A schematic diagram of the SLCS, showing major system components,is presented in Figure 17.3-7. Some locked open maintenance valves and some checknives are not shown. Key equipment performance requirements are:
'l l (1) Pump flow per pump 11.35 m3 /h per pump j l (2) Maximum reactor pressure (for 8.6 MPa injecdon) {
(3) Pumpable volume in storage tank 23,090.9 L j (minimum) )
i 17.3-6 Reliability Assurance Program During Design Phase - Amendment 34
23A6100 Rsv.1 sezrsirs sat:ty Analysis Report ABWR t i V
Design provisions to permit system testing include a test tank and associated piping and valves. The tank can be supplied with demineralized water which can be pumped in a closed loop through either pump or injected into the reactor.
The SLCS uses a dissolved solution of sodium pentaborate as the neutron-absorbing poison. This solution is held in a heated storage tank to maintain the solution above its saturation temperature. The SLCS solution tank, a test water tank, the two positive displacement pumps, and associated valving are located in the secondary containment on the floor elevation below the operating floor. This is a Seismic Category I structure, and the SLCS equipment is protected from phenomena such as earthquakes, tornados, hurricanes and floods as well as from internal postulated accident phenomena. In this area, the SLCS is not subject to conditions such as missiles, pipe whip, and discharging fluids, i
The pumps are capable of producing discharge pressure to inject the solution into the reactor when the reactor is at high pressure conditions corresponding to the system relief valve actuation. Signals indicating storage tank liquid level, tank outlet valve position, pump discharge pressure and injection valve position are available in the control room.
The pumps, heater, valves and controls are powered from the standby power supply or normal offsite power. The pumps and valves are powered and controlled from separate buses and circuits so that single active failures will not prevent system operation. The power supplied to one motor operated injection valve, storage tank discharge valve, and injection pump is from Division 1,480 VAC. The power supply to the other motor-operated injection valve, storage tank outlet valve, and injection pump is from Division II,480 VAC. The power supply to the tank heaters and heater controls is connectable to a standby power source. The standby power source is Class IE from an onsite source and is independent of the offsite power.
All components of the system which are required for injection of the neutron absorber into the reactor are classified Seismic Category I. All major mechanical components are designed to meet ASME Code requirements as shown below.
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\J 17.3'7 Reliability Assurance Program Ouring Design Phase - Amendment 31
23A6100 Rev. 4 ABWR StandardSafety Analysis Report O
Design Conditions Component Class Pressure Temperature Storage Tank 2 Static Head 66 C Pump 2 10.8 MPaG 66'C l
h !jection Valves 1 10.8 MPaG 66 C l
Piping Inboard ofInjection Valves 1 8.6 MPaG 302 C l
17.3.11.2 SLCS Operation The SLCS is initiated by one of three means: (1) manually initiated from the main control room; (2) automatically initiated if conditions of high reactor pressure and power level not below the Anticipated Transient Without Scram (ABVS) permissive power level exist for 3 minutes; or (3) automatically initiated if conditions of RPV water level below the Level 2 setpoint and power level not below the ATWS permissive power level exist for 3 minutes. The SLCS provides borated water to the reactor core to introduce negative reactivity effects during the required conditions.
To meet its negative reacthity objective, it is necessary for the SLCS to inject a quantity of boron which produces a minimum concentration of 850 ppm of natumi boron in the reactor core at 20 C. To allow for potentialleakage and imperfect mixing in the reactor system, an additional 25% (220 ppm) margin is added to the above requirement. 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. This quantity of boron solution is the amount which is above the pump suction shutofflevelin the storage tank, thus allowing for the portion of the tank volume which cannot be injected.
17.3.11.3 Major Differences from Operating BWRs The SLCS design is very similar to that of operating BWRS. Automatic actuation of the ABWR SLCS is similar to that incorporated in some operating BWRS. Because of the larger ABWR RPV volume, the pumping capacity has been increased from 9.8 to l 11.4 m3 /h per pump. Injection of SLCS solution through the HPCF sparger has been shown by boron mixing tests to give better mixing than the operating plant injection through a standpipe.
hijection valves of operating plants are leak-proof explosive valves to keep baron out of the reactor during SLCS testing. In the ABWR the injection valves are motor operated 17.3 4 Reliability Assurance Program During Design Phase- Amendment 34
23A6100 Rev. 4 ABWR StandardSafetyAnalysis Report D
and a suction pipe fHI system keeps the lines filled with distilled water at slightly higher pressure than that of the baron storage tank to preclude enuy of boron into the reactor.
The motor-operated injection valves provide the following advantages over explosive valves:
(1) Radiation exposure to personnel is potentially reduced during testing and maintenance because less work will be required at the valves.
(2) Post-injection containment isolation capability is enhanced because the motor operated valves can be closed following boron injection. Explosive valves cannot be reclosed to provide containment isolation.
17.3.11.4 SLCS Fault Tree The top level fault tree for the SLCS is shown in Figure 17.18,with the top gate defined l as failure to deliver 11.4 m3 /h of borated water from the storage tank to the RPV. Details providing input to most of the events in Figure 17.S8 are contained in the several additional branches to the fault tree.
It is assumed that the SLCS has been identified by the PRA as a system making I significant contribution to CDF. A listing of the SLCS components or events by Fussell-Vesely Importance was made, and those SSCs with greatest importance are given in Table 17.SI. No SSCs appear to be risk-significant because of aging or common cause considerations. The seven most significant components are listed in Table 17.S2, so these SSCs should be considered as risk-significant candidates for O-RAP activities 17.3.11.5 System Design Response The seven SLCS risk-significant components identified in Table 17.S2 as having high importance in the SLCS fault tree are now considered for redesign or for O-RAP J acdvities, as noted above. The flow chart of Figure 17.S1 guides the designer.
Two of the events in Table 17.12 result from flow of SLCS fluid being diverted through relief valves back to pump suction rather than into the RPV. Since gate and check ulve failures (which could result in relief valve operation) are accounted for by separate events, the relief valve failures of concern can be considered to be valve body failures or inadvertent opening of the reliefvalves. Plugging of the suction lines from the storage tank could result from some contamination of the tank fluid or collection of foreign matter in the tank.The pump failures to start upon demand could result from electrical or mechanical problems at the pumps or their control circuits.
Two AC electrical system failures that contribute to SLCS failure are identified in Table 17.12. No further details of electrical system failures or maintenance are Reliability Assurance Program During Design f - Amendment 34 17.3 9
23A6100 fitv.1 ACWR standardSainty Analysis Report included here. That leaves the five components noted above for special attention with 9
regard to reducing the risk of system failure.
(1) Redesign The design evaluation of Figure 17.Sl is used by the designer. The design assessment shows that the component failure rates are the same as those used in the PRA, so there is no need to recalculate the PRA. Also, no one SSC has a major impact on SLCS unavailability, so redesign or reselection of components is not required and the seven components are identified for consideration by the O-RAP.
Redesign consideradons, if they had been required, would have included trying to identify more reliable relief valves and pumps and suction lines less likely to plug. The latter might be achieved by using larger diameter pipes or multiple suction lines. Pump and valve reliability might be enhanced by specific design changes or by selection of a different component. Any such redesign would have to be evaluated by balancing the increase in reliability against the added complication to plant equipment and layout.
(2) Failure Mode Identification If redesign is not necessag, or after redesign has been completed, the appropriate O RAP activities would be identified for the three SLCS component types identified by the fault tree and discussed above. This begins with determining the likely failure modes that will lead to loss of function, i
following the steps in Figure 17.S2. The components of SLCS have adequate failure history to identify critical failure modes, so Assessment Paths A and C )
(Figures 17.33 and 17.S5, respectively) would be followed to define the failure modes for consideration.
For the SLCS relief valves, past experience with similar valves shows that the I
major failure modes are fluid leakage from the valve body and a spurious opening as result of failure of the spring, the spring fastener, the valve stem or f the disk. Past pump failures fall into two general categories, electrical i problems resulting in failure to start on demand and mechanical problems that cause a running pump to stop or fail to provide rated flow. The plugging of fluid lines generally results from presence of sediment or precipitation of compounds from saturated fluid.
Following the flow chart of Figure 17.13, the designer would determine more details about each failure mode, including pieceparts most likely to fail and the frequency of each failure mode category or piecepart failure. This would result in a list of the dominant failure modes to be considered for the O-RAP.
17.3-10 Reliability Assurance Program Ouring Design Phase - Amendment 31
23A6100 Rev. 3 ABWR StandardSafetyAnalysis Report 18.5 Remote Shutdown System The Remote Shutdown System (RSS) provides a means to safely shut down the plant from outside the main control room. It provides control of the plant systems needed to bring the plant to hot shutdown, with the subsequent capability to attain cold shutdown, in the event that the control room becomes uninhabitable.
The RSS design is described in Subsections 7.4.1.4 and 7.4.2.4. All of the controls and instrumentation required for RSS operation are identified in Subsection 7.4.1.4.4 and in Figure 7.4-2.
The RSS uses conventional, hardwired controls and indicators to maintain diversity from the main control room.These dedicated devices are arranged in a mimic of the interfacing systems process loops.
Evaluation of alternate design approaches for rehability and confirmation of the l adequacy of the RSS design is COL license information (Subsection 18.8.6).
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23A6100 Rev. 4 l ABWR standardsafety Analysis Report
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23A6100 Rev.1 ABWR Standard S:,fetyAndysis R; port 9 19L-5 A. Villerneur, et al. (Electricite de France), Living Probabilistic Safety Assessment of a French 1300 ADVe P%R Nuclear Power Plant Unit: Methodology, Results and Teachings, Published at TUV-Workshop on Living PSA Application, Harnburg, FRG, May 7-8,1990.
19L-6 Advanced Light Water Reactor Requirements Document, Appendix A: PRA Key Assumptions and Ground Rules, Draft, EPRI, August 1988.
19L-7 Recommended TechnicalSpecifcationsforFuelLoading, service information letter No. 372, General Electric, June 1982.
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ABWR She*down Risk Evaluation - Amendment 31 19L-27
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23A6100 Rev. 3 ABWR Stand:rdSttyAn:lysisRep rt O
Table 19L-1 ABWR Modes of Operation Average Reactor Reactor Mode Switch Coolant Temperature, Mode
- Title Position K ('C) 1 Power Operation Run Any temperature 2 Startup Startup/ Hot Standby Any temperature l 3 Hot Shutdown Shutdown >366.45 K (> 93.3'C) l 4 Cold Shutdown Shutdown s366.45 K (s 93.3 C) l 5 Refueling Shutdown or Refuel s366.45 K ( s93.3'C)'
In Modes 1 through 4, fuel is in the reactor vessel with the reactor vessel head closure bolts fully tensioned. in Mode 5, fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.
t Technical specification states "any temperature *, but in this mode the temperature will be below boiling point.
O O
19L-28 ABWR Shutdown Risk Evaluation - Amendment 33
23AG100 Rev. 3 standard safety Analysis aeport ABWR v
19Q.12.4 Reliability Goals (input to RAP)
The following assumed system unavailabilities were determined to be important in minimizing shutdown risk and are included in the ABWR Reliability Assurance Program:
Unavailability System (Per Demand)
RHR (SDC) 7E-2 per division RHR (LPFL) 4E-2 per division HPCF 5E-2 CRD 5E-4 CTG 5E-2 EDG 4E-3
/
Offsite Power 1E-3 ADS 6E-6 DC Power 7E-6 19Q.12.5 Conclusions 1 The ABWR has been evaluated for risks associated with shutdown conditions and for all postulated events, the risk has been determined to be low Multiple means of remosing (
decay heat and supplying inventory makeup have been identified that along with l appropriate Technical Specifications and outage procedures result in acceptably low shutdown risk levels for the AB\m.
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19045 ABnR Shutdown Risk Assessment- Amendment 33
23A3100 Rev. 2 ABWR StandardSafety Analysis Report O
Table 19Q-1 ABWR Features That Minimize Shutdown Risk Category Feature Shutdown Risk Capability Decay Heat Residual Heat Three independent (100% capacity) divisions of RHR and Removal Removal (RHR) support systems for normal DHR. Each RHR division has (DHR) System several DHR modes (e.g., SDC, SPC).
Reactor Coolant During shutdown, reactor coolant temperature is Temperature determined by measuring Reactor Water Cleanup (CUW)
Measurement inlet water temperature.
Shutdown Cooling The shutdown cooling mode of RHR uses suction piping Nozzle that connects directly to a nozzle on the RPV instead of to an external piping system. This reduces the probability of losing RHR pump suction due to air entrapment or cavitation.
Safety Relief Valves Can be used as alternate means of decay heat removal by venting steam to the suppression pool. They are also actuated to depressurize the RPV to allow use of low pressure RHR or other low pressure systems.
Suppression Pool A potential heat sink and makeup source for decay heat removal. Pool temperature is monitored in the control room to indicate trends in pool temperature. This large heat sink allows sufficient time for appropriate operator actions.
Reactor Water Can be used under certain conditions to remove decay heat.
Cleanup System See Subsection 19Q.7 and Attachment 190B for more (CUW) details on this feature.
RPV Boiling When the RPV head is removed, boiling is an effective (although not preferred) heat transfer method as long as RPV water level can be maintained by available makeup sources.
Condenser The main condenser (if available) can be used for DHR.
Remote Shutdown Cold Shutdown can be achieved and maintained from Panel (Two Divisions) outside the control room if the control room is uninhabitable due to fire, toxic gas, or other reasons. The remote shutdown panel is powered by Class 1E power to ensure availability following a Loss Of Preferred Power (LOPP). Controls are hard wired and thus not dependent on multiplexing systems. A minimum set of monitored parameters and controls are included to ensure the ability to achieve and maintain cold shutdown.
O 19046 ABWR Shutdown Risk Assessment- Amendment 32
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ABWR SSAR t. ( Amendment 34. Page Change Instruction (continued) 7he following pages Itave been changed, please make the specified changes in your SSAR. Pages arx listed as page pairs (front & back), exception chapter 16. Bold page numbers represent a page that has been changed by Amendment 31. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CilAFITR 2 CilAFITR 3 (Cont'd) TAB 10 TAB 33 2.0-1 thru 3/4 2.0-1 thru 3/4 2.0-3/4 2.0-3/4 33-1 thru 4 33-1 thru 4 TAB 2.2 TAB 3,4 2.2-3/4 2.2-3/4 3.4-1 thru 16 3.4-1 thru 18 TAD 23 IA.Bl 23-1,2 23-1,2 3.5-5 thru 12 3.5-5 thru 12 23-5,6 23-5,6 3.5 17/18 3.5 thru 17/18
\ TAB 3,6 CIIAPTER 3 3.6-1, 2 3.6-1, 2 TAB 3.0 3.6-7 thru 12 3.6-7 thru 12 3.6 31,32 3.6-31,32 3.0-i thru vi 3.0-1 thru vi 3.6 37,38 3.6-37,38 3.0-ix thru xil 3.0-is thru xil 3.6-41,42 3.6-41,42 3.0-xv thru xviii 3.0-xy thru xvill 3.0-xxi thru xxiv 3.0-xxl thru xxiv TAB 3,7 3.0-xxvii, xxviii 3.0 xxvii,xxvill 4 3.o.xxi e, x ar,4 1.o-vxi<,xxx 3.7-3, 4 3.7-3, 4 TAB 3,1 3.7-15 thru 26 3.715 thru 26 3.7-45 thru 48 3.7-45 thru 48 ;
3.1-1, 2 3.11,2 3.7-55 thru 78 3.7-55 thru 78 3.1-9, 10 3.1-9,10 ' 3.1-13, 14 3.1-13, 14 IAll.18 3.1-19, 20 3.1-19, 20 3.1-35 thru 62 3.135 thru 62 3.8-1 thru 26 3.81 thru 26 3 & 29,30 3 &29,30 TAB 32 3&37,38 3.8-37,38 3.8-41 thru 48 3.8-41 thru 48 3.2-5 thru 10 3.2-5 thru 10 3.8-51,52 3&51,52 3.2-17 thru 62 3.2-17 thru 62 3.8-59,60 3&59,60
% P ac3e "5. o- X g i. y shouId bc A % d%4d 32 ^ o + '3 1 aI p v ovicled M kbE m o d MCd\ow f A cby.
1 c AlnVR SSAR Amendment 34 Page Change Instruction (continued) l
'Ihe following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed by Amendment 34.
REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. GI Al'TER 3 fC.921'd) TAB Ano. 3B l TAJ}l9 3B-5 thru 70 3B-5 thru 73/74 -l l 3.9-1 thru 156 3.9-1 thru 155/156 TAB Ano. 3E TAB 3.10 3E 7,8 3E-7, 8 l 3E-9 thru 12 3E-9 thru 12 3.10-1, 2 3.10-1, 2 3E-15 thru 26 3E-15 thru 26 3E-33 thruf9/605o 3E-33 thru 59/60 50
- TAB 3.11 SE. 55 4bW 59/Go 3 E -55 +be a 3 9 /6 o TAB Ano.3G 3.11-3, 4 3.11-3,4 3G 5,6 3G-5,6
.p TAB 3.12 3G-9 thru 115/116 3G-9 thru 115/116 3.12-1 thru 3/4 3.121 thru 4 TAB Ano. 3H , TAB 313 3H.0-lii thru vili 311.0-111 thru vill 3.13 1, 2 3.13 1, 2 311.1-3 thru 10 311.1-3 thru 10 3.13-7 thru 10 3.13-7 thru 10 311.1 13 thru 86 3H.1-13 thru 86 3.13-13 thru 16 3.13-13 thru 21 3H.1-89 thru 92 311.1-89 thru 92 3H.197 thru 102 3H.1-97 thru 102 TAB Aon. 3A 3H.1105 thru 113/114 3H.1-105 thru 113/114 3A-3 thru 8 3A-3 thru 8 311.2-1 thru 10 3H.2-1 thru 10 3A-11,12 3A-11,12 3H.2-13 thru 30 3H.2-13 thru 30 3A 27,28 3A-27,28 3A-31 thru 50 3A-31 thru 50 3H3-1 thru 8 3H3-1 thru 8 3A-59 thru 62 3A 59 thru 62 3H3-11 thru 38 3113-11 thru 39 3A-71,72 3A-71,72 3A-75 thru 94 3A-75 thru 94 3H 4-1 thru 6 3H.4-1 thru 6 3A-97 thru 296 3A-97 thru 2% 3H.5-1,2 3H.5-1, 2 TAB Aco.31 31-1 thru 18 31-1 thru 18
- Pages 3G-51,szcmd 33 34 g y_ ,vi d e ci in @ s s vw o clM ecA o n p a c k q 3 c m 4-b e ev em,t 4kg we4 ch s c a d e d card s e v .
m . _ . . _ . . _ ,_.-. _ _ ___ _ _ _ . _ _ _ . _ . _ _ . . . . __ __ - _ _ __ _ e ABWR SSAR Amendment 34 - Page Change Instruction (continued) The following pages have been changed, please make the specified changes la your SSAR. Pages are listed as page pairs (front & back),creeption Chapter 16. Bold page numbers represent a page that has been changed by Amendment 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CHAFTER 5 (Cont'd) Cil AFITR 6 (Cont'd) TA.D 5.2 TAB 6.1 5.2-1 thru 20 5.21 thru 20 6.11 thru 4 6.1-1 thru 4 5.2-25 thru 36 5.2-25 thru 36 6.17,8 6.1-7, 8 5.2 39,40 5.2-39,40 5.2-45 thru 50 5.2-45 thru 50 TAB 6.2 5.2-53 thru 56 5.2-53 thru 56 5.2-71, 72 5.2-71,72 6.2-1 thru 228 6.21 thru 241/242 5.2-75,76 5.2-75,76 5.2-81, 82 5.2-81, 82 TAB 6.3 131}.13 631 thru 8 63-1 thru 8 63-11 thru 14 63-11 thru 14 53-5 thru 8 53-5 thru 8 63-17 thru 86 63-17 thru 87/88 53-11 thru 18 53-11 thru 18 5 3-21, 22 5 3-21,22 TAB 6.4 5 3-25/26 5 3-25/26 6.4-1 thru 10 6.4-1 thru 10 , TAB 5.4 TAB 6,5 5.4-3, 4 5.4-3, 4 5.4-7 thru 73 5.4-7 thru 75 6.5-1 thru 17/18 6 5-1 thru 18 TAB Anp. 5A TAB 6.6
- c. G - 5,6 6.g. L 6 SA-1, 2 5A-1, 2 6.6-1, 2 6.61,2 SA-7, 8 5A-7,8 6.6-11, 12 6.6-11, 12 6.6-29/30 6.6 29/30 TAB Aco. SB 6.6-39 thru 42 - 6.6-39 thru 42 5B-1 thru 5 5B-1 thru 5 TAB 6.7 6.7-1 thru 6 631 thru 6
,QlAFTER 6 TAB 60 6.0-i thru xii 6.0-1 thru xil i
l ABWR SSAR Amendment 34 Page Change Instruction (continued) , G, 1 1
'Ihe following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed J by Amendment 34.
l REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CII APER 6 (Cont'd) Cil Al'rER 7 (Cont'd) TAB Aon. 6A TAB 7,2 6A-3, 4 6A-3,4 7.2-5 thru 24 _ _ 7.2-5 thru 24 6A-9 thru 12 6A-9 thru 12 7.2-47,48(7.2-47, 48j --> TAB 7.2 (cont'd) TAB Ann. 6B 7.2-53,54 7.2-53, 54 6B-1, 2 6B-1, 2 7.2-57,58 7.2-57, 58 6B-5, 6 6B-5, 6 IADl3 TAB Ann.6C 73-1 thru 4 73-1 thru 4 m 6C-1 thru 3/4 6C-1 thru 3/4 73-7 thru 16 73-7 thru 16
) 7319 thru 22 73-19 thru 22
,V TAB API.JLD '* 7 3-33,34 73-33,34 3 ,' 7 3-49,50 73-49,50 S/4 6D-1,6 6D-1 [ / 73-53 thru 68 73-53 thru 68 73-71,72 73-71,72 TAB Ann.6E 73-81,82 73-81,82 73-87 thru % 73-87 thru 97/98 Add 6E-1 thru 3 TAB 7.4 CIIAl"TER 7 7.4-1 thru 4 7.41 thru 4 7.4-7, 8 7.47,8 TM 'LQ 7.4-17, 18 7.4 17,18 7.4-27 thru 30 7.4-27 thru 30 7.0-i thru vil/ sui 7.0-1 thm vil/ vill TAB 7.5 TAB 7,1 7.5-1, 2 7.5-1, 2 7.1-1 thru 4 7.1-1 thru 4 7.5-7 thru 14 7.5-7 thru 14 7.1-15 thru 24 7.1-15 thm 24 7.5-21 thru 24 7.5-21 thru 24 7.1-27,28 7.1-27,28 7.5-27 thru 29 7.5-27 thru 29 7.1-33, 34 7.1 33,34 7.1-37, 38 7.1-37, 38 A A -p q [) S/4 f rovi A t el m -4 L s S 6w o cI ds c_ ed\ o ^ P*C [
ABWR SSAR Amendment 34. Page Change Instruction (continued) d The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed by Amendment 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CIIAl'IER 9 (Cont'd) CIIAFITR 9 (Cont'd) TAB 9.1 TAB Aco. 9A (cont'd) 9.1-3, 4 9.1-3, 4 9A.4-1 thru 507 9A.4-1 thru 502 9.1-9 thru 18 9.1-9 thru 18 9.123 thru 32 9.1-23 thru 32 9A59 thru 14 9A.5-9 thru 14 9.1-35 thru 50 9.1-35 thrts 50 9A525,26 9A.5-25,26 9.155 thru 60 9.1-55 thru 60 9.1-65, 66 9.1-65, 66 9A.6-99,100 9A.6-99,100 9.1-69,70 9.1-69,70 TAB Aco. 9B TAB 9.2 9B-3 thru 14 9B-3 thru 14 9.21 thru 4 9.2-1 thru 4 9.2-11 thru 70 9.2-11 thru 70 TAB Ann.9C 9.2-73 thru 76 9.2-73 thru 76 9C-1 thru 12 9C-1 thru 12' TAB 93 TAB Ann. 9D 93-3 thru 6 93-3 thru 6 93-11 thru 14 93-11 thru 14 9D-1,2 9D-1, 2 93-17 thru 36 93-17 thru 36 9D-5,6 9D-5, 6 93-39 thru 42 93-39 thru 42 93-45 thru 48 93-45 thru 48 CHAlrIT,R 10 TAB 9.4 TAB 10.0 9.4-1 thru 57 9.4-1 thru 57 10.0-i/ii thru v/vi 10.0-l/11/ thru v/vl TAB 9.5 TAB 10.1 951 thru 90 9.5-1 thru 90 8 j 10.1 1 thru 6 10.1-1 thruQj TAB Aon. 9A ! TAB 10.2 l 9A.2-3 thru 9/10 9A.2-3 thru 9/10 10.2-1 thru 21/22 10.2-1 thru 21/22 ; 9A3-3 9A3-3 b v l
ABWR SSAR Amendment 34 - Page change instruction (continued) The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed by. Amendment 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CilAPTER 14 TAB 14.2 (Continued) TAB 14.0 14.2 61.62 14.2-61,62 14.0-i thru v/vi 14.0-1thru v/vi 14.2 71,72 14.2-71,72 14.2-75 thru 170 14.2-75 thru 170 TAB 14.2 14.2179 thru 182 14.2 179 thru 182 14.2-11 thru 18 14.2-11 thru 18 14.2-185,186 14.2-185,186 14.2-21,22 14.2-21,22 14.2193 thru 200 14.2193 thru 14.2-25 thru 30 14,2 25 thru3 0 200 14.2-33,4hsu 34 14. 2-33,64mt 34 14.2-39 thru 42 14.2 39 thru 42 TAB 14.3 I 14.2-47 thru 50 14.2 47 thru 50 \ 14.2-53,54 14.2-53,54 14.3-5 thru 59 14.4-5 thru 6 0 l l
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Amendment 34 - Page change instruction (continued) The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed by A me nd tn e n t 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CilAPTER 15 TAB 15.6 (continurd1 TAB.15.0 15.6-45 thru 47/48 15.6 45 thru-47/48 l 15.0-1 thru xviii 15.0-1thru xyl 15.0 5 thru 19 15.0-5 thru 19 4 TAB 15.7 TAB 151 15.7 1,2 15.7-1,2-10 to 15.7 5 thru 20 15.7 5 thru 2 0 15.1-3 thru d 15.1 3 thru 15.1-15 thru 18 15.1-15 thru 18 BIL153 15.1 21 thru 24 15.1-21 thru 2 4 15.8 1,2 15.8-1,2
. TAB 15.2 MB 15A
- 15.2-1 thru 12 15.2 1 thru 12 15.215 thru 24 15.215 thru 24 15 A-i/il I 5 A 1/il ,
15.2-27 thru 55/56 15.2-27 g 15 A v thru viil 15 A-v thru vill thru 55/56 15A 7 thru 10 15 A-7 thru 10 15 A 23/24 15 A 23/2 4 TAB 15.3 15 A 35/36 15 A 35/3 6 i is.3- t 4619/1o is.3. 4km t 8 15 A-41 thru 48 15 A-41 thru 4 8 l' 1 3 :hru 10 15.4 -3 L 10- 15A 57 thru 88 15A-57 thru 88 15 A 91 thru 110 15 A 91 thru 110 . TAB 15.4 15A-ll3 thru 118 15 A-113 thru i l'18 15.4 3 thru 10 15.4-3 thru 10 15 A 121 thru 15A-121 thru 15,419 thru 23/24 15.4-19 thru 127/128 127/128-
'23/24 ,
TAB 13 TAB 15.5 I 15B 11 thru 16 15 B-11 thru 16' 15.5 1 thru 4 15.5 1 thru 4 15B-35/36 15 B -35/3 6 ] l TAB 15.6 TAB 15D 15.6-3 thru 8 15.6 3 thru 8 15 D 1,2 15 D - 1,2 15.6 11,12 15.6 11,12 15.6-17 thru 42 15.6 17 thru 4 2 - 4 Pcge s is.o 5, C f rov' SkCcl 'w 4h3 'w
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i u 1 L ABWR SSAR i [- ' Amendment 34 - Page Change Instruction (continued) l l The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), exception Chapter 16. Hold page numbers represent a page that has been changed by Amendment 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CII APER 16 Pages printed on one side. CHAPER 16 (Cont'd) TAB 16.0 TAB 163.4 iii lii 3.4-3 3.4-3 vil vil 3.44 3.4-4 3.4-6 3.4-6 16.F2 16.0-2 3.4-8 3.4-8 3.4-12 thru 17 3.4-12 thru 17 TAB 16.2 3.4-23 3.4-23 2A1 2.0-1 TAB 163.5 TAB 163 3.5-1 thru 6 3.5-1 thru 6 3 510 3510 3.0-2 3.0-2 TAB 163.6 TAB 163.1 3.6-2 3.6-2 3.1-6 3.1-6 3.6-7 3.6-7 3.1 12 thru 14 3.112 thru 14 3.6-12 3.6-12 3.1-16 3.1-16 3.6-15 thru 18 3.6-15 thru 18 3.1-21 3.1-21 3.6-21 thru 23 3.6-21 thru 23 3628 3.6 28 TAB 1633 3.6-29 3.6-29 3 3-G 3,3 . g 3.6-34 3.6-34 331 thru 3 33-1 thru 3 33.g 33-5 7 TAB 163.7 33-8 33-8 33-10 thru 16 33-10 thru 16 3.7-16 3.7 16 3 3-21 3 3-21 3.7 17 3.7-17 3 3-24 3 3-24 33-28 thru 32 33-28 thru 32 TAT,163.8 33-34 thru 38 33-34 thru 38
.3-42 3 3-42 3&1 38-1
,3 52 3 3-52 3&3 3&3 3}68 thru 71 33-68 thru 71 3.8-5 thru 8 3.8-5 thru 8 3 3-74 3 3-74 3&l0 thru 13 3.8-10 thru 13
ABWR SSAR O V Amendment 34 Page Change Instruction (continued) ne following pages have been changed, please make the specified changes in your SSAR. Pages an listed as page pairs (front & back), exception Chapter 16. Bold page numbers represent a page that has been changed by Amendment 34. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CII AITER 16 (Cont'd) CIIAFTER 16 (Cont'd) TAB 16B3 TAB 163.8 (cont'd) B 3.0-5 B 3.0-5 3.8-15 3.8-15 B 3.0-11 B 3.0-11 3.8-17 3.8-17 3.8-18 3.8-18 TAB 16B33 3.8-25 thru 27 3.8-25 thru 27 N,'f.8
@B 33-10{9j',} _ 3 B 33-10 3.8-42 thru 44 3.8-42 thru 44 3.8-50 3.8-50 B 33-12 B 33-12 B 33-13 B 33-13 TAB 163.9 B 33-17 B 33-17 B 33-18 B 33-18 3.9-2 3.9-2 B 33-34 B 33-34 3.9-3 3.9-3 gB 33-35 B 33-35 3.97 3.9-7 B 33-54 B 33-54 O
g 3.9 9 3.9-11 3.9-9 3.9-11 B 33-57 B 33-60 thru 84 B 33-100 B 33-57 B 33-60 thru 84 B 33-100 TAB 16.5 B 33-102 thru 104 B 33-102 thru 1(M B 33-107 B 33-107 5.0-14 5.0-14 B 33-110 B 33-110 5.0-18 5.0-18 B 33-120 thru 124 B 33-120 thru 124 B 33-126 B 33-126 TAB 16B3.1 B 33-129 B 33-129 B 33-130 B 33-130 B 3.1-11 B 3.1-11 B 33-134 B 33-134 B 3.1-14 B 3.1-14 B 33-136 B 33-136 B 3.1-23 thru 25 B 3.1-23 thru 25 B 33-139 B 33-139 B 3.127 B 3.1-27 B 33-161 thru 163 B 33-161 thru 163 B 3.1-31 thru 33 B 3.1-31 thn 33 B 33-165 thru 167 B 33-165 thru 167 B 3.1-41 B 3.1-41 B 33-182 B 33-182 B 33-183 B 33-183 TAB 16B.2 B 33-211 B 33-211 B 33-215 B 33-215 B 2.0-2 B 2.0-2 B 33-217 B 33-217 B 2.0-3 B 2.0-3 B 33-238 B 33-238 B 2.0-8 B 2.0-8 B 33-240 B 33-240 C W A P"TER l7 T A S l'T.3 3;( 17.3-s b u io rr. 3- 6 +b r u Io w Altpagar p4 move % el d o nd $ 6% 5 7 1!' - G - Fgud
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