ML20235L272

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Control Sys Common Failure Analyses Rept Covering Limerick Generating Station,Unit 2
ML20235L272
Person / Time
Site: Limerick 
Issue date: 01/31/1989
From: Bennet D, Koslow A
GENERAL ELECTRIC CO.
To:
Shared Package
ML20235L267 List:
References
EDE-04-0189, EDE-4-189, NUDOCS 8902270449
Download: ML20235L272 (52)
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{{#Wiki_filter:-_ CONTROL SYSTEM COMMON FAILURE ANALYSES REPORT COVERING THE LIMERICK GENERATING STATION, UNIT 2 JAilUARY 1989 l PREPARED FOR THE PHILADELPHIA ELECTRIC COMPANY BY D. E. BENNETT . GENERAL ELECTRIC COMPANY, NUCLEAR ENERGY SAN JOSE, CALIFORNIA 95125 ~ Approved: h$ / /29/H A. Koslow, Manager Regulatory and Design Compliance Application Engineering Electrical Design Engineering i h. 6. W W u /ffr/r9 M. E. Urata, Principal Lidensing Engineer Licensing and Consulting Services ) (([ Obbk b>3 A EDE-04-0189 )

i 1.0 PURPOSE This report responds to the Nuclear Regulatory Commission's (NRC) Limerick Generating Station, Final Safety Analysis Report (FSAR) questions 421.10, 421.11 and 421.7 (IE Notice 79-22) for unit 2. These questions and concerns ' are defined as follows: During full reactor power operations, could a failure in a power source, sensor or sensor lina or a high energy line break zone common to two or more non-safety-related control system (NSCS) components adversely affect the primary reactor parameters, i.e., vessel water

level, pressure or reactivity?

If so, is the consequence, if any, of such a failure bounded by the current FSAR chapter 15 analyses which address this failure and the mitigation, if any, of this failure?

2.0 CONCLUSION

Using the comparative analysis methodology approach, as described in the following sections of this report, an engineering review team did not identify any substantial differences between the Reference 1, 2 and 3 failure analyses report data and their findings, as it applies to the currently documented unit 2 reactor design. As such, except as noted, this analyses report supports the original analysis conclusions that the identified common non-safety-related control system failure scenario consequences are bounded by the current FSAR Chapter 15 analyses. 3.0 METHODOLOGY 3.1 Description i These unit 2 plant analyses and report supplement the original Reference 1, 2, and 3 analyses and use a comparative methodology (unit 2, 1988 versus unit 1, 1983 plant designs). The validity of this methodology is logical. This is confirmed by the unit 2 plant's identity with the unit 1, 1983 plant, and its successful comparison to the original, accepted analyses data (see References 4 and 5). ) 3.2 Scope To maintain continuity and further support the validity of the comparative methodology, wherever possible, the same original analyses organizations, General Electric Company, Nuclear Energy (GE), the Bechtel Power Corporation ) (BPC) and participants performed the unit 2 analyses (see Table 1). 3.3 Definitions While these analyses methodology are comparative, unless otherwise indicated, the unit 1 analyses Reference 1, 2 and 3 report definitions are applicable for this report. .-( )

i TABLE 1 f ANALYSES SCOPE MAJOR ACTIVITY RESPONSIBILITY COMMON POWER COMPARISONS NSSS BOP Bus Structure GE BPC Critical Bus Loads GE BPC Combined Failure Effects GE BPC COMMON SENSOR COMPARISONS Sensor Line Structure GE BPC Critical Line Loads GE BPC Combined Failure Effects GE BPC HELB ZONE AND COMPONENT CONTENT COMPARIS0NS Zone Layout and Isolation GE BP. Critical Zone and Component Content GE BPC Plant Walkdown Confirmation GE BPC ADMINISTRATION AND DOCUMENTATION Compile Summary Report GE GE Maintain Design Review File GE GE > i

l l 4.0 COMMON POWER SOURCE FAILURE ANALYSIS 4.1 Description The applicability and use of the Reference 1 report findings for Unit 2 was I reviewed. Except for the use of unit 2 system and component numbering' schemes and current unit 2 documentation references, no significant difference between the Reference 1 report's non-safety-related control systems (NSCS) power distribution scheme and the unit 2 NSCS power i distribution scheme were identified. 4.1.1 Bus Structure The bus structure, as indicated on the current unit 2 design documents (mainly one-line diagrams), compared favorably with the bus tree and structure reviewed in Reference 1. A few differences were identified which, in turn, affect minor changes in subsequent bus loadings and combined failure effect scenarios (see Appendix A). However, these differences and changes do not significantly deviate from the Reference 1 findings and are bounded by similar FSAR Chapter 15 analyses. 4.1.2 Critical Bus Loads t For these analyses, critical busses are considered to be non-safety-related busses (power sources) that commonly supply power to two or more of the major reactor NSCS (see Table 2). Because power disruption to strategic components that comprise each major NSCS could significantly affect the system operation and reactor primary parameters (similar to bus structura changes), changes in this distribution might cause changes in the resulting failure scenarios. The analyses showed that these changes, when compared to the unit I analyses results, have insignificant effect and are bounded by FSAR Chapter 15 analyses (see Appendix A). TABLE 2 MAJOR REACTOR NON-SAFETY-RELATED CONTROL SYSTEMS (NSCS) 1. Nuclear Boiler / Steam Leak Detection Systems (B21) 2. Reactor Recirculation Flow Control System (B32) 3. Control Rod Drive / Reactor Manual Control Systems (Cll) 4. Reactor Feedwater Flow and Level Control System (C32) 5. Neutron Monitoring System (non-safety-related portion) (CSI) 6. Reactor Water Cleanup System (G31) 7. Main Turbine Steam and Extraction Steam Systems (M01,02) 8. Feedwater and Circulating Water Systems (M06,09) 4

5.0 COMMON SENSOR LINE FAILURE ANALYSIS 5.1 Description Similar to the common power failure analysis, the applicability and use of the Reference 2 report findings for the unit 2 common sensor failure-analysis was reviewed.

Again, except for the use of unit 2 system and component numbering schemes and documentation references, no significant l

differences between the Reference 2 report's non-safety-related (NSCS) sensor line distribution scheme and the scheme used for unit 2 NSCS sensor l lines were identified. 5.1.1 Sensor Line Structure Unlike the common power bus structure described in Section 4 and Reference 1, the sensor line structure services some ufety-related system sensors as well as the NSCS sensors. Thus, a cormon NSCf. sensor line failure could cause changes in both safety and non-safety related system actions. In most cases, either the safety system redundancies or accident mitigating actions tend to lessen or terminate any adverse NSCS action which might otherwise occur (see Appendix B). Similar to Reference 2, the unit 2 analysis findings support this effect and no significant differences were identified. 5.1.2 Critical Sensor Line Loads As was the case in the Reference 2 analyses, the identified common sensor lines and loads were mainly associated directly with nuclear boiler applications. Thus, the unit 2 nuclear boiler P& ids, as well as the various NSCS elementary and schematic diagram configurations were reviewed and compared with the Reference 2 common sensor line load sheets.

While, similar to the unit 2 common power bus structure, differences in line and sensor distribution configuration were identified and analyzed.
Again, these differences were found to be non-significant and are bounded by

) current Limerick FSAR, Chapter 15 analyses. 6.0 HELB ZONE AND ZONE NSCS COMP 0NENT CONTENT ANALYSIS 6.1 Description ) Because, unlike the common power and sensor analyses, the Reference 3 HELB zone and NSCS component content and the unit 2 HELB zone and NSCS component content are more subject to plant construction and layout difference, i.e., the physical zone barrier and component placement deviation freedoms allowed for non-safety-related equipment, unit 2 plant documentation alone was ) unable to confirm Reference 3 analysis findings completely. Thus, a visual plant walkdown of the critical zones and component locations, similar to that performed for the Reference 3 analysis, was required. While the HELB analysis plant walkdown did observe physical differences in some zone layouts and NSCS component content, when compared with the ) Reference 3 data, these differences and the combined failures effects were not found to be significant and again are bounded by FSAR Chapter 15 analyses (see Appendix C). 5

REFERENCES:

1. Control System Failures Evaluation Report For Limerick Generation Station, November 1983 (Unit 1) 2. Common Sensor Failure Evaluation Report For Limerick Generation ' Station, November 1983 (Unit 1) 3. High Energy Line Break Control System Failure Analysis for Limerick Station Unit 1, April 1984 4. Letter to NRC's A.Schwencer from PEC0's J.S.Kemper, December 14, 1983: " Limerick Generating Station, Unit I and Unit 2 SER issues from Instrumentation and Control System Branch" 5. Limerick Generating Station Safety Evaluation Report (SSER2), Section 7.7, Control Systems ) ) i

a 0 <s APPENDIX A l cnm,n Power Failure Analysis Q2nparisons, Unit 2 Vs Unit 1 (1983) DC Battery B_":e (See Fig *1re 1) 'Ihe effect of losirg these DC battery busses is the same as for unit 1 but with the followirg differences: Bus A, MCC 20D201, was charged to be the prefernd source for panel 2AY160 by way of UPS static inverter 2AD160. Loss of power to panel 2AY160 will result in closure of inboard reactor water clean up system valve G31-F001, 1 one out of two (1/2) trip of MS1V channels A and C and 1/2 scram of RPS channels A and C. Bus B, MCC 20D203, was changed to be the preferred source for panel 2BY160 by way of UPS static inverter 2BD160. Ioss of power to panel 2BY160 will result in 1/2 trip of MSIV channels B arxi D aM 1/2 scram of RPS channels B aM D). The loss of d:fwell cooling due to closure of drywell coclirs water supply valves will not m9m when bus 2BY160 is lost. The isolation logic for the drywell cooliry water valves has been changed from what it was when the unit 1 analysis was made. In addition, panels 2AY160 and 2BY160 are not expected to loso power if their preferred DC power sources are lost, because, the static inverters would normally automatically transfer to the alternate source of power. AC Power n - (See Figure 2) Bus 20A101 Panel 1AY160 for unit 1 had its preferred feed from Bus 114A-G-F, MCC 10B111, via pane.1 1AC246 with DC bus MCC 10D201 as the alternate feed. In unit 2, the preferred feed for 2AY160 is DC bus MCC 20D201 (see Figure 1) with MCC 20B111 as alternate source 2 and MCC 00B401 as alternate source 1. 'Ihis unit 2 charge has eliminated the 1/2 scran and 1/2 trip of MSIVs when bus 20A101 is lost. 'Ihe feedvater control loop B was on panel 10Y109 for the unit 1 analysis. ) For unit 2, feedwater wukul loop A is on panel 20Y109. Loss of 20Y109 to the feedwater wubul loop A causes reactor feedwater pump turbine (RFPT) A to lockup at its last ccrnmarded speed and RFFT B arxi C control to switch to manual but maintain last commanded speed. Loss of certain components will signal recirculation pump A to run back to limit #1; however, the loss of panel 20Y109 power for the recirculation contrcl system will also cause a ) scoop tube lockup of pump A, resulting in the pump remainire at its last commanded speed. A-1 h

) ? l l 1 I ) i ne combined effect of a bus 20A101 loss would be the loss of reacter i recirculation pump A, the loss of 2 out of 3 condensate pumps, and the loss of 2 out of 4 circulating water pumps. 'Ihe loss of the condensate pumps will cause all three feedwater pumps to trip on low suction pressure. 'Ihe A trip of the feedwater pumps will cause recirculation pump B to run back to limit #1. 'Ihis run back will not prevent an eventual low reactor water level, RPS 3 scram. 'Ihe loss of 2 out 4 circulating water pumps will also result in lower cordenser vacuum. h e loss of associated lower level power buses will result in the following ) events: Lockup of recirculation pump A and feed #ater turbine A; also, feedwater turbines B&C will go to manual control and maintain speed, while, recirculation pump B will remain at speed under master recirculation control. W ese events will not have any severe effects on the primary reactor parameters. Based on the unit 2 feedwater cudal system elementary diagram 791E408TR, Prz. 5, the ccebined effect of the above power lossec is the initiation cf a loss of feedwater flow transient event. 'Ihis is the same event as the loss of bus 10A101 for unit 1 and is addressed in the FSAR, Chapter 15 analyses. ) Bas 20A102 For the unit 1 analysis, panel 1BY160 had its preferred feed from 12 G-G-F, MCC10B112, via panel 1BC246 with DC bus MCC10D203 as the alternate feed. For unit 2, the preferred feed is from DC bus MCC 20D203 with MCC 20B112 as alternate source 2 and MCC 00B401 as alternate source 1. 'Ihe unit 2 configuration eliminates the 1/2 scram, 1/2 trip of MSIVs, loss of drywell cooling and false IDCA signal that occurred when panel 10A102 is lost on unit 1. 'Ihe feedwater control loop A was on panel 10Y110 for the unit 1 analysis. For the unit 2 analysis, feedwater control loop B was on panel 20Y110. 'Ihe loss of panel 20Y110 causes recirculation punp B to loc}aIp at its last speed, recirrallation pump A to run back to minimum speed, RFPT B to lockup at its last speed and RFFT A and C to run back to their minimum speed. 'Ihe combined effect of bus 20A102 loss would be, the loss of recirculation pump B, the loss of 1 out of 3 condensate pumps, the loss of 2 out of 4 circulating water pumps ard recirculation pump A will run back to its minimum speed. Also, RFPI B will lockup at its last speed and RFFT A and C will decrease speed, because of the mismatch in steam flow to feedaater flow caused by wagonent loss. 'Ihe mismatch of 25% steam flw vs 33% feedvater flow will decrease feedwater flow until the feedwater controller ) reactor level point matches the mismatch error. When this level is reached, the level will stabilize at lower than normal level but at a point above the low water level 3 RPS scram trip. A-2 5 ]

) Bus 20B201 'Ihe unit 1 MCC 10B219 was rob duplicated for unit 2. 'Ihe few loads that would have been fed frun this source for unit 2 have been fed fram MCC 20B213. 'Ihis raising of the bus feed one tier does not adversely effect or ) charge the analysis results. Bus 20B202 'Ihe unit 1 MCC 10B220, like the MCC above, was not duplicated for unit 2. 'Ihe loads for unit 2 are fed from the next higher tier MCC 20B214. Again, no adverse failure affect or analysis result charges were identified. 5 Bus 20B203 'Ihe unit 1 panel 10Y163, power came from MCC ODB131. 'Ihe unit 2 panel 20Y163 power canes fram MCC 20B217. No adverse failure affect or analysis result changes were identified. Bus 20B204 'Ihe Unit 1 panel 10Y164 power came from MCC OCB132. 'Ihe unit 2 panel 20Y164 power canes from MCC 20B218. Jgain, no adverse failure affect or analysis result changes were identdfied. ) ) ? ) h A-3

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APPENDIX B Common Sensor Failure' Analysis Comparisons. Unit 2 Vs Unit 1 (1983) Some of the sensors shown on unit 2, Nuclear Boiler P&ID, M-42, see Figure 8.1, reactor vessel tap N11B for line 6 and tap N11A for line 7, were re-evaluated when compared with the unit 1, 1983 Reference 2 analysis, see Table B-1 and B-2. As a result of this re-evaluation, a break in either instrument line 6 or line 7, will cause a reactor low water level (LWL) 3 RPS channels C and D, or A and B trip, which will cause a reactor low water level scram. This event is analyzed in FSAR, Chapter 15 and changes the Reference 2, Section 4.0 portion of the unit I report as follows: Instrument Line 6 (See Table B-1) A break in this line will cause a reactor low water level (LWL), RPS 3, channels C & D, scram. A plug in this line will inhibit a reactor LWL, RPS 3, channels C & D, scram but the redundant RPS 3, channels A & P. scram is available. Instrument Line 7 (See Table B-2) A break in this line wil'1 cause a reactor LWL RPS 3, channels A & B, scram. A plug in this line will inhibit a reactor LWL, RPS 3, channels A & B, scram but the redundant RPS 3, channels C & D, scram is available. s i B-1

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) APPENDIX C Hiah Enerav Line Break Failure Analysis Comoarison. Unit 2 vs Unit 1 (1983) ) 1. Zone Elimination Criteria The zone identification for the high energy line break (HELB) failure analysis for unit 2 was done by taking the unit 1 (Reference 3) study ) zone markings and transferring them to the unit 2 areas. For example, j area (2/189,200/Q15) becomes area (4/189,200/Q15) for unit 2. This allowed the unit 2 study zones to be compared directly with the original unit I study. The Figures 1 through 13 attached to this report show the zones l identified for.the unit 2 study reference, including those that were eliminated using the following criteria. (1) No high energy lines are routed through this zone. (2) All control system components which reside within this zone have. been eliminated per the criteria detailed in the original report, Reference 3. Therefore, no applicable control system components reside within this zone. (3) Zone boundaries, high energy lines, and control system components are the same as those found for the unit 1 study, therefore, no difference analysis is required. Because of the comparative nature of this study, the third elimination criteria was added to the two criteria used for the unit 1 analysis. The third criteria was determined to apply for those zones by the follnwing actions:

1) A review of the zone boundaries verified that no new areas needed to be included.
2) A review of the high energy lines showed that the same 1.ines were in the area and that the lines are in the same general location within each zone.

And 3) the same control system components are located in the same general area within the zone analyzed in unit 1. If these conditions were met, it was determined that the same pipe failure would affect the same group of control components. When this was found to be the case, then the zone was eliminated. If there was any major new grouping of the control system components, then the zone was retained for further analysis. 2. High Energy Line Break Failure Analysis Results Of the 11 zones that were included in the plant walk down (those for which no elimination criteria could be applied) no new combined failure effects events were found in any zone. The differences in the grouping of the components in unit 2 for the zones that required analysis were mitigated by the " sacrificial" approach used in postulating the failure of the control systems. This failure mode is the same as that used in the original study, see Reference 3, Section 2.4. l C-1

I 3. Zones 3.1 Turbine Enclosure The turbine enclosure was divided into the zones listed below. These zones are identified in Figures 2 through 7 attached. Also, noted below are the zones eliminated from the comparative analysis for at least one of the three elimination criteria listed in Section 1. l Zgng Eliminated (?) Reason Area 4/ Elevations 189,200/ Grid Q15 Yes 3 Areas 3,4/ Elevations 189,200,217/ Grid R20 No Areas 10,9,4/ Elevation 200/ Grid M12 No Area 8/ Elevation 200 Grid M21 Yes 2 Area 5/ Elevation 217/ Grid Q6 Yes 3 Areas 4,5/ Elevation 217/ Grid Q15 Yes 3 Areas 3,4/ Elevation 217/ Grid Q20 Yes 2 Areas 3,4/ Elevation 217/ Grid P20 Yes 2 Areas 3,4/ Elevation 217/ Grid P22 Yes 1 Areas 9/10/ Elevation 217/ Grid MIS No Area 8/ Elevation 217/ Grid M21 Yes 1 Area 5/ Elevation 239/ Grid Q7 Yes 1 Areas 4,5/ Elevation 23'/ Grid P14 Yes 3 9 Areas 3,4/ Elevation 239/ Grid P20 Yes 1 Area 10/ Elevation 239/ Grid M7 No Areas 9/10/ Elevation 239/ Grid M12 Yes 2 Area 9/ Elevation 239/ Grid M16 Yes 2 Areas 9/10/ Elevation 239/ Grid J14 Yes 1 Area 8/ Elevation 239/ Grid K22 Yes 1 Area 5/ Elevation 269/ Grid Q(n)10 Yes 2 C-2

) Zone Eliminated (?) Reason Areas 3.4.5/ Elevation 269/ Grid Q15 No Area 10/ Elevation 269/ Grid N9 No Area 9/ Elevation 269/ Grid N14 Yes 2 Area 9/ Elevation 269/ Grid N16 No Areas 9/10/ Elevation 269/ Grid M12 No Area 8/ Elevation 269/ Grid M22 Yes 1 Areas 9/10/ Elevation 302/ Grid M12 Yes 1 k Area 9/ Elevation 302/ Grid K14 Yes I Area 8/ Elevation 304/ Grid M22 Yes 1 Area 8/ Elevation 321/ Grid K(f)23 Yes 1 b Area 8/ Elevation 332/ Grid M23 Yes 1 Area 8/ Elevation 350/ Grid K(f)23 Yes 1 3.2 Reactor Enclosure ,t The reactor enclosure was divided into the zones listed below. These zones are identified in Figures 8 through 13 attached. The zones defined are consistent with the zones of influence defined in Chapter 3.6 of the FSAR, which analyzes high energy pipe breaks and resultant effects on safety related equipment. Also, noted below are the zones eliminated from the detailed analysis for at least one of the three zone elimination criteria listed in Section 1. Zo_ne Eliminated (?) Reason n Al-Area 13/ Elevations 177,201,217 Yes 2 Area 18/ Elevation 207/ Grid A1 l A2-Area 13/ Elevations 177,201,217 Yes 2 Area 18/ Elevation 207/ Grid A2 A3-Area 14/ Elevation 177/ Grid A3 Yes 1 A4-Areas 13,14,17,18/ Elevation 177 Yes 1 / Grid A4 B1-Areas 13,14,17,18/ Elevation 201 Yes 1 / Grid B1 C-3

r. Inng Eliminated (?)- Reason Cl-Areas 13,14,17,18/ Elevation 217 Yes 1 / Grid C1 D1-Areas 14,18/ Elevation 253/ Grid D1 No D2-Areas 13,17/ Elevation 253/ Grid D2' No D3-Areas 17,18/ Elevation 253/ Grid D3 No i D4-Areas 13,14/ Elevation 253/ Grid D4 Yes 2 El-Areas 14,18/ Elevation. 283/ Grid El Yes 2 E2-Areas 13,14,17,18/ Elevation 283 Yes 1 / Grid E2 F1-Areas 13,14,17,18/ Elevations 313 Yes 1 / Grid F1 and above 4. Zone Analyis Results-(See Figures 2 through 13) An HELB failure analysis was performed covering each of the eleven un-eliminated zones identified above. The results of the each analysis are as follows: Zone (3,4/189,200,217/R20) The regrouping-of components in this zone did not change the analysis;. thus, the results were identical to the unit 1 study results. Zone (4,9,10/200/M12) High Energy Line: 2" DBD-208 Function: Provide that portion of the CRD cooling water supply line from the pumps to the point that it penetrates the reactor enclosure wall. Effect of break: Loss,of the cooling water supply to the CRD system and the recirculation pump seal water system. The control systems affected were the same as those affected for.the unit I study. Combined Failure Effects: The CRD piping failure could cause the condensate recirculation valve to open, causing a possible reduction in feedwater flow. Also, the failure could cause the tripping of reactor feedwater pump turbine (RFPT) A, due to the failure of the feedwater pressure, switch, PSL-06-202A. The other results of the study were the same as the results for unit 1. Zone (9,10/217/M15) The RFPTs are package units and, as such, have the same configuration as the RFPTs on unit 1. The control system components are the condensate recirculation control, which are the same as unit 1. The results of the study were the same as the results for unit 1. Zone (10/239/M7) The regrouping of components in this zone did not change the analysis; thus, the results were identical to the unit I study results. C-4

Zone (3,4,5/269/Q15) The results were identical to the unit I study results. The regrouping of the unit 2 instrumentation in this zone did not affect the analysis resul ts. Zone (10,269/N9) l The regrouping of components in this zone did not change the analysis; thus, the results were identical to the unit I study results. Zone (9/269/N16) The regrouping of components in this zone did not change the analysis; thus, the results were identical to the unit I study results. Zone (9,10/269/M12) The regrouping of components in this zone did not change the analysis; thus, the results were identical to the unit I study results. Zone (14,18/253/D1) The HELB of the CRD, HCU and piping could cause the reactor water level transmitter LT-42-2N004B to fail. This failure will cause the feedwater control system to see a false reactor low water level. If this level. detector is selected to be the control element, then a level error will be generated which will cause the feedwater system to increase the feed water flow rate. This increase in feedwater flow will cause a reactor high level scram. This event is bounded by FSAR Chapter 15, loss of feedwater flow control analysis. Other results of this unit 2 zone study were the same as the results for unit 1. Zone (13,17/253/02) The HELB of the CRD, HCU and piping can cause the reactor water level transmitter LT-422N004A to fail. This failure will cause the feedwater control system to see a false reactor low water level. If this level detector is selected to be the control element, then a level error will be generated which will cause the feedwater system to increase the feedwater flow rate. This increase in feedwater flow rate will cause a rector high level scram. This event is bounded by FSAR, Chapter 15, loss of feedwater flow control analyses. Other results of this unit 2 zone study were the same as those for unit 1. Zone (17,18/253/D3) The regrouping of components in this zone did not change the analysis; thus, the results were identical to the unit I study results. C-5 l l J

HIGH ENERGY LINE BREAK Reference Drawinas Used for Unit 2 Study Zgne Instrument location. Fioure No. Drawina and Revision (3,4/189,200,217/R20) M0685 R6, M'-686 R12, 3,4 M-1500 R9, M-1501 R7 (4,9,10/200/M12) M-1520 R11, M-1511 RIO 2 (9,10/217/M15) M-1512 RIO, M-1521 R6 3 (10/239/M7) M-1522 R6 4 (3,4,5/269/Q15) M-1504 RIO, M-1510 R7 5 (10/269/N9) M-1524 R5 5 (9/269/N16) M-1515 R6 5 (9,10/269/M12) M-1515 R6, M-1524 R5 5 (14,18/253/D1) M-1539 R11, M-1553 R14 11 (13,17/253/D2) M-1533 R11, M-1545 R15 11 (17,18/253/03) M-1545 R15, M-1553 R14 11 C-6

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