ML20245F109
| ML20245F109 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89472, NUDOCS 8908140153 | |
| Download: ML20245F109 (17) | |
Text
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Log # TXX-89472 n
C File # 10010 r
C 910.2 7UELECTRIC p,f,p$M50.34(b)
"H"" J C"hi"' dent July 31* 1989 3'-
therwwr Vice Presi U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 PREOPERATIONAL AND STARTUP TESTING AUDIT CONDUCTED JUNE 20-22, 1989 Gentlemen:
This letter provides an advance copy of changes to be included in a future FSAR amendment regarding the CPSES design response of the generator following an automatic or manual turbine trip. These changes are provided in response to inquiries by the NRC during the subject audit conducted by Messrs.
J. 7wolinski, R. Ramirez, F. Ashe, R. Gruel. and M. Fields. The results of a preliminary review indicate that these changes are acceptable.
Unless the final review identifies a concern, no further correspondcqce regarding these changes will be required.
In order to facilitate NRC staff review of these changes, the enclosure is organized as follows:
1.
Draft revised FSAR pages, with changed portions indicated by a bar in the margin, as they are to appear in a future amendment (additional pages immediately preceding and/or following the revised pages are provided if needed to understand the change).
2.
Line-by-line description / justification of each item revised.
3.
A copy of related SER/SSER sections.
4.
An index page containing the title of " bullets" which consolidate and categorize similar individual changes by subject and related SER section.
5.
The bold / overstrike version of the revised FSAR pages referented by the description / justification for each item identified above. The bold / overstrike version f acilitates review of the new text in bold 1
type font and overstriking with a slash (/) the portion of the text that is deleted.
8908140153 390731 o
PDR ADOCK 05000445 7
4 PDC 400 krth Olive Str.vt LB81 Dallas, Texas 75201
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TXX-89472 July 31,'1989 Page 2 of 2-TU Electric requests that the'NRC perform an expedited review of the above FSAR changes and inform us as to their acceptability.
Sincerely.
William J. Cahill, Jr.
By:
M-M.
- -L J@n'W. Beck..'
Vice President, Nuclear Engineering RSB/vid Enclosure <
c - Mr. R. D. Martin, Region'IV Rasident Inspectors, CPSES (3) 1 4
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f Enclosure to TXX-89472 July 31, 1989 AdJance FSAR Change Regarding the CPSES Design Response of the Generator Following an Auto / Manual Turbine Trip Item 1 Draft Revised FSAR Pages pg. 2 thru 5 Item 2 Description / Justifications for pg. 6 all FSAR Changes Item 3 Related SER/SSER Pages pg 7 thru 10 Item 4 Index Page Containing the pg. 11 Title of " Bullets" Item 5 Markup of Existing FSAR Pages pg. 12 thru 15 L
pg. 1 of 15 l
I i
1
Enclosure to TXX-89472
' July 31,1989 CPSES/FSAR
, Pg.12 ' of.15
'3..
Manual generator. trip from Control Room 4..
Electronic generator protection (EGP), including signals from primary water flows, primary water temperature, primary water tank level, and tritium control.
5.
Rotor ground protection 6.
Pilot exciter short After an automatic turbine trip, the generator breaker trip is delayed DRAFT to furnish uninterrupted power to the reactor coolant pump motors for at least 30 see without relying on the success of a bus transfer, provided the generator conditions permit this.
Likewise, following a manual turbine trip, or a turbine trip due to certain turbine faults l
or certain generator protection signals, the generator trip is delayed approximately 11.5 seconds. Generator trip (after a turbine trip) is also conditional upon detection of reverse power, except for certain generator faults, to minimize the probability and the degree of Dyerspeed after a turbine trip. The trip logic is shown functionally DRAFT on Figure 7.2-1, Sheet 16 and Figure 10.2-1.
0040.130 For a narrative and schsmatics, which describe in detail the sequence 11 of events in a turbine trip, see Reference 3, For response times the following is prcvided. These times assume 11 normal control mode of operation which is electrohydraulic control (EHC).
Approx. Time Action 11 (Milliseconds) 11 0
Load rejection 10 11 10 Load rejection is sensed by the electrical 11 control system.
20 Output signal of the valve lif t controller 11 starts to decrease.
+
10.2-25 Draft Ver*,1on
Enclosure to TXX-89472
--July 31,1989 CPSES/FSAR
' Pg. 3 of.15 I
iNormal power for the pumps is supplied'through individual buses connected to the generator. When a generator trip occurs, the buses
~
are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply coolant flow to the DRAFT core..Following any automatic turbine trip where there are no electrical or turbine faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pump buses remain connected to the network thus ensuring full flow for approximate 1y'30 seconds after the reactor trip before any transfer is made.
For a manu:1 turbine trip, or a turbine trip due to certain turbine faults or certain generator protection signals, the generator remains connected to the network for approximately 11.5 seconds.
For a complete description of the turbine / generator trip logic, see Figure 7.2-1, Sheet 16 and Figure 10.2-1.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1.
The necessary protection against a partial loss of coolant flow accident is provided by the low primary coolant flow reactor trip signal which is actuated in any reactor coolant loop by two out of three low flow signals. Above Permissive 8, low flow in any loop will actuate a reactor trip.
Between approximately 10 percent power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two loops will actuate a reactor trip. Above Permissive 7.-two or more reactor coolant pump circuit breakers opening will actuate the corresponding undervoltage relays.
This results in a reactor trip which serves as a back up to the low flow trip.
15.3.1.2 Anal _vsis of Effects and Consequences Method of Analysis 76 Draft Version 15.3-2
Enclosure to TXX-89472
= July 31,1989
- 9' CPSES/FSAR
, The ' calculated sequence of events for the two cases analyzed are shown on' Table 15.3+1.
The at
- ted reactor coolant pump will continue to coastdown, and the core ow will re.ach a new equilibrium value corresponding to the number of pumps still in operation. With the
. reactor tripped, a stable plant'conoition will eventually be attained.
Normal plant shutdown may then proceed.
l L
15.3.1.3 Conclusions The analysis shows that the DNBR will not decrease below the limiting value of 1.30 at any time during the transient. The DNBR design basis is described in Section 4.4.
All applicable ecceptance criteria are met.
15.3.2 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2.1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result fr; 4 -
simultaneous loss of electrical supplies to all reactor coolant pamps.
If the reactor is at power at the time of the accident, the immediate effect of loss of' coolant flow is a rapid increase in the coolant
' temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly.
Normal power for the reactor coolant pumps is supplied through buses from e transformer. connected to the generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply coolant flow to the core.
Following any automatic turbine trip DRAFT where there are no electrical or turbb. faults which require tripping the generator from the network, the generator remains connected to the Draft Version 15.3-5
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'i Enclosure to TXX-89472 July ' 31,- 1989 Pg. 5 of 16 CPSES/FSAR DRAFT netw'ork for approximately 30 seconds. The reactor coolant pump buses remain connected to the network, thus ensuring full flow for 30 seconds af ter the reactor trip befor.e any transfer is made.
For a manual turbine trip, or a turbine trip due to certain turbine faults or certain generatcr protection signals, the generator remains connected to the network for approximately 11.5 seconds.
For a complete description of the turbine / generator trip logic, see Figure 7.2-1, Sheet 16 and Figure 10.2-1.
s
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Draft Version 15.3-6
Epcl os'u re ' to ' TXX-89472 CPSES FSAR AMEN 0 MENT 77'
. July 31, 1989' DETAILED DESCRIPTION Pago 1 Pg. 61of 15
!FSAR-Page (as amended)
GroUD Description 10.2-25 4
Clarifies the CPSES design response of the Generator following an automatic or manual turbine trip.
Clarification:
For a manual turbine trip, or a turbine trip _due to certain turbine faults or certain generator protection signals, the generator trip'is delayed approximate ~ly 11.5 seconds compared to approximately 30 seconds for an automatic turbine' trip. The concittsions presented in the FSAR Chapter 15 Accident Analyt.s remain unchanged as a result of these clarifications.
FSAR Change Request Number: 89-510.5 Related SER Section: 15.2.2 SER/SSER impact: No 15.3-2, 5-
.4 See Page No(s):6 Clarifies the CPSES design response of the Generator following an automatic or manual turbine trip.
Clarification:
For a manual turbine trip, or a turbine trip due to certain turbine faults or certain generator protection signals, the generator' trip is delayed approximately
-11.5 seconds compared to approximately 30 seconds for an automatic turb*.ne trip. The conclusions presented in the FSAR Chapter 15 Accident Analysis remain unchanged as a result of these clarifications.
FSAR Change Request Number: 89-510.2 Related SER Section: 15.2.2 SER/SSER Impact: No l
l
Enclosure to TXX-89472.
July 31,1989 Pg. 7 o,f b (5) startup of an inactive reactor coolant pump at an incorrect tem None of these transients are limiting; the m events.
Only slight changes in primary system pressure were calculated e
the departure from nucleate boiling ratio did not fall below 1 4
, and finds these results acceptable because they do not violate the The staff limits.
15.2.2 Decreased Cooling Transients The applicant has analyzed the following events which pro system cooling:
(1) loss of external electrical load (2) turbine trip (3) inadvertent closure of main steam isolation valves (4) loss of condenser vacuum and other events resulting in turbin (5) loss of nonemergency ac power to the station auxiliaries (6) loss of normal feedwater flow (7) partial loss of forced reactor coolant flow overpressurization is the turbine triNone of these transients a pressure of approximately 2550 psia. p transient, which results in a peak RCS than 110% of the RCS design pressure, the staff finds these 15.2.3 Increased Core Reactivity Transients 15.2.3.1 Baron Dilution Events The principal means of causing an inadvertent boron dilution are t of the primary water makeup control valve and tailure of the blend either by the controller or mechanical failure.
The chemical volume and control system (CVCS) is designed to limit, even under various postu ardes, the dilution rate to values which, will allow sufficient time fo or operator response (depending on the mode of operation) to terminate tion before the shutdown margin is exhausted.
This dilution rate is indicated by instrumentation.
modes of operation.
The applicant has analyzed the baron dilution event for all,
Dilution During Refueling Uncontrolled boron dilution cannot occur during the refueling mod sources of unborated water are isolated in this mode.
15-4
Enclosure to TXX-89472 July 31, '1989 Pg. 8 of 15 ll i
10 STEAM AND POWER-CONVERSION SYSTEM 10.1 Summary Description The steam and power-conversion system is designed to remove heat energy from the primary reactor coolant loop via the steam generators and to generate electric power in the turbine generator.
After the steam passes through the low-pressure turbines, the main condensers deaerate the condensate and transfer
,, the rejected heat to the once-through circulating water system, which uses water from the Squaw Creek Reservoir. The condensate is reheated and returned as feedwater to the steam generators. The entire system is designed for the maximum-expected energy from the nuclear steam supply system (NSSS).
In normal operation of the Comanche Peak plant, the turbines and auxiliaries use all the steam from the steam generators.
A turbine bypass system is pro-vided for each turbine to discharge up to 40% of the main steam flow directly i
to the condenser during transient conditicns.
This bypass capacity, together with a 10% reactor automatic step load reduction capability, is sufficient to o
allow a 50% generator load loss without tripping the reactor or the turbine.
n i
10.2 Turbine Generator The turbine for each unit is a tandem-compound, four-flow unit consisting of one double-flow high pressure stage and two double-flow low pressure stages.
The rotational speed for each unit is 1800 rpm, and the generator nameplate j
rating is 1350 MVA with a 0.9 power factor.
Exhaust steam from the high pressure e
stage passes through one of two combination moisture-separator / reheaters con-i nected in parallel before entering the low pressure stages.
The turbine-i generator unit is manufactured by Allis-Chalmers Power Systems Inc.
10.2.1 Overspeed Protection System The turbine speed and load for normal operation are controlled by an electro-hydraulic control (EHC) system, with a backup mechanical-hydraulic control (MHC) a system, that controls the operation of hydraulic operators on the high pressure and low pressure steam control valves. Although the '.wo systems use different speed transducers, they share many components, including hydraulic fluid controls.
The EHC modulates the turbine control valves to maintain desired speed load characteristics, and it will close the intercept valves and control valves at approximately 103 to 105% of rated speed.
The MHC performs the same function as the EHC but functions only on loss of the EHC.
The MHC modulates the turbine control valves to maintain speed load characteristics, and it will clos the control valves and intercept valves at approximately 103 to 106% of raced speed.
The turbine generator control system also includes a load rejection relay (LRR) which, in conjunction with the MHC, acts as a backup to the EHC.
The LRR cir-cuitry monitors generator output, and, when it senses generator load rejection, it rapidly closes the cortrol valves for a long enough time to allow the slower
)
acting MHC to take over control of the turbine.
This system effectively limits the overspeed transient to less than 110% of ra+.ed speed.
j 10-1 i
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m Enclosure'to TXX-89472 July 31,1989 Pg. 9 of 15 In addition to the EHC and MHC systems, turbine overspeed protection is provided by two mechanical backup trip devices that cause full closure of all turbine stop and control valves and by steam extraction line nonreturn valves, at about 110% of rated speed.
to 120% of rated speed.)(The turbine disc is designed and tested for speeds up The mechanical backup system trips the turbine (1) if both the EHC and MHC fail i
to close the control valves on demand and (2) if overspeed results from load rejection and the failure of the LRR to function.
This backup system consists of two redundant mechanical devices which are activated by spring-loaded bolts on the front end of the turbine shaft.
As the speed reaches the set point at 110% of rated speed, the bolts move out under centrifugal force and trip a lever, which causes the release of the hydraulic fluid pressure on all turbine control and stop valves.
function for turbine overspeed protection.Only one of the two mechanical trip dev Although there is no backup electrical overspeed trip system as described in Section 10.2 of the SRP, the plant turbine speed centrol system (which consists of the normal speed / load centrol systems, the backup LRR, and the redundant mechanical trip devices) provides the required redundancy in turbine overspeed protection.
The staff believes that the turbine overspeed protection system design meets or exceeds the intent of Section 10.2 of the SRP for turbine overspeed protection redundancy, and is, therefore, acceptable.
There are also a number of other turbine trips associated with abnormalities:
low bearing oil pressure; low condenser vacuum; excessive vibration; abnormal thrust bearing wear; high water level in the moisture separators; manual trip from the main control room or at the turbine; loss of hydraulic control fluid pressure; steam generator high-high level or safety injection; reactor trip; and generator trip.
Turbine trips result in the closure of all the control, stop, and power-assisted extraction line check valves, and in the generation of a reactor trip signal.
The applicant has outlined an inservice inspection program for essential components of the turbine control and protection systems.
In particular, each main steam and low pressure stop-and control valve is completely dismantled i
and inspected at least once every 40 months.
The applicant discussed numerous tests that can be performed on protective l
devices and valves using the automatic turbine tester.
i frequency is biweekly.
The normal testing frequency required by Section 10.2 of the SRP.This frequency is less than the wee l
However, the automatic tu:bine tester measures valve closure times, a feature not required by the SRP, but which, the applicant contends, con result in early warning of impending valve malfunctioning, thereby allowing valves to be tested less frequently.
The l
staff agrees with the appliunt and considers biweekly testing of the above valves to be acceptable in this case.
i to less than 85% to test the valves.
The applicant will have to reduce power test procedure sper:ify remedial actions to be taken if abnormal valve clos i
times are measured.
of the turbine generator in accordance with Regulatory Guide 1,68.The i
10-2
)
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Enclosure to TXX-89472 July 31, 1989 Pg. 10,o f 15 The turbin'e-generator system meets the recommendations of Branch Technical Positions ASB.3-1,," Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," and MEB 3-1, " Postulated Break and Lcakaga Loca-tions in Fluid System Piping Outside Containment." Evaluation of protection against dynamic effects associated with the postulated pipe system failure is covered in Section 3.6 of this report.
The staff review of the turbine generator included descriptive information in Section 10.2 of the FSAR for Comanche Peak, flow charts, and diagrams.
The design criteria and bases and design of the turbine generator system meets the criteria in Section II of SRP 10.2 and industry standards.
Therefore, based on its review, the staff concludes that the turbine-generator overspeed protection system can perform its designed safety functions and is, therefore, acceptable.
10.2.2 Turbine Disc Integrity The previous review of Section 10.2.3 of the applicant's FSAR indicated that there was not-sufficient information for the staff to assess the potential for turbine disc cracking and to establish an inspection schedule.
This information has been requested.
Specifically, the staff requires information on operating temperatures and on the lubricants used during turbine assembly.
The Comanche Peak turbines were manufa'ctured by Si; mens of Germany, using a steel with above waverage yield strength for low pressure turbine discs.
This steel can contribute to the susceptibility to stress-corrosion cracking.
There is further concern because several cases of cracking in Siemens low pressure turbine discs in fossil fuel plants have been reported.
The staff has taken an interim position regarding these turbines because the applicant requires more time to gather the necessary information than the licensing period permits.
In view of the facts stated above, the staff has taken a conservative approach, assuming that the propensity for disc cracking is the same as that of Westinghouse turbines.
Therefore, the staff will require that the bares and keyways of-the low pressure turbine discs be inspected for ultrasonic indications of cracking during the first refueling outage.
The inspection schedule may be adjusted if the applicant's responses to the staff's questions are favorable.
10.3 Main Steam Supply System The function of the main steam supply system is to convey steam from the steam generators to the high pressure turbine and other auxiliary equipment for power Section 10.3.1 evaluates the safety-related portion of the main generation.
steam system (outside of containment) which includes the portion of the system between the containment up to and including the main steam isolation valves (MSIVs).
Section 10.3.2 evaluates the nonsafety-related portion of the system downstream of the MSIVs, up to and including the turbine stop valves.
10-3
EncIosuretoTXX-89472-
' July 31,1989
- Pg. 11 0f 15 -
15.2.2 Decreased Coolina Trensients SRXB 7.
The FSAR has been clarified regarding the CPSES (77) design response of the generator following an automatic or manual turbine trip.
I
Enclosure to TXX-89472 July 31g 1989 CPSES/FSAR Pg. 12,0f 15 3.
- Hanual generator trip froa Control Room i
4 Electronic generator protection (EGP), including signals from primary water flows, primary water temperature, primary water tank level, and tritium control.
5.
Rotor ground protection 6.
Pilot exciter short Af ter an automatic turbine trip, the generator breaker trip is delayed to furnish uninterrupted power to the reactor coolant pump motors for at least 33 see without relying on the success of a bus transfer, provided the fdf5f8d generator conditions permit this.
- Likewise, following a manual turbine trip, or a turbine trip due to certain turbine faults or certain generator protection signals, the generator j
trip is delayed approximately 11.5 seconds. Generator trip (after a turbine trip) is also conditienti upon detection of reverse power, except for certain generator faults, to minimite the probability and the degree of overspeed after a turbine trip. The trip logic is shown functiorrelly on Figure 7.2-1 Sheet 16 and Figure 10.2-1.
0040.130 For a narrative and schematics, which describe in detail the sequence 11 of events in a turbine trip, see Reference 3.
For response times the following is provided.
These times assume 11 normal control mode of operation which is electrohydraulic control (EHC).
Approx. Time Action 71 (Hil11 seconds) 11 0
Load rejection 10
' 11 10 Load rejection is sensed by the electrical 11 control system.
20 Output signal of the valve lift controller 11 starts to decrease.
10.2-25 Bold /0verstrike Version
Enclosure'to TXX-89472' July ~31, 1989 CPSES/FSAR
' Pg. 13, o f 15
. Normal power for the pumps is supplied through individual buses conriected to the generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply coolant flow to the core.
Following any automatic turbine trip where there are no electrical or turbine faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps buses remain connected to the network ddridtdfdf thus ensuring full flow for approximately 30 seconds after the reactor trip before any transfer is made.
For.a manual turbine trip, or a turbine trip due to certain turbine faults or certain generator protection signals, the generator remains connected to the network for approximately 11.5 seconds. For a complete description of the turbine / generator trip logic, see Figure 1.2-1, Sheet 16 and Figure 10.2-1.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1.
The necessary protection against a partial loss of coolant flow accident is provided by the low primary coolant flow reactor trip signal which is actuated in any reactor coolant loop by two out of three low flow signals. Above Permissive 8, low flow in any loop will actuate a reactor' trip.
Between approximately 10 percent power.
(Permissive 7) and the power level corresponding to Permissive 8, low flow in any two loops will actuate a reactor trip. Above Permissive
- 7. two or more reactor coolant pump circuit breakers opening will actuate the corresponding undervoltage relays. This results in a reactor trip whigh serves as a back up to the low flow trip.
15.3.1.2 Analysis gf,_gffsets and Cont 69uences
\\
Method of Analy.=11 76 j
i Bold /0verstrike 15.3-2 Version
)
Enclosure to TXX-89472 CPSES/FSAR July 31,1989 Ifkhecalculatcdsequ2ncaofeventsforthetwocasesanalyzedareshown
' 9 I on Table 15.3-1.
The affected reactor coolant pump wi11' continue to coastdown, and the core flow will reach a new equilibrium value corresponding to the number of pumps still in operation. With the reactor tripped, a stable plant condition will eventually be attained.
Normal plant shutdown may then proceed.
15.3.1.3 Conclusions The analysis shows that the DNBR will not decrease below the limiting-value of 1.30 at any time during the transient. The DNBR design basis is described in Section 4.4.
' All applicable acceptance criteria are met.
15.3.2 COMPLETE LOSS OF FORCED REACTOR COOLANT FL0i!
15.3.2.1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result frem a simultaneous loss of electrical supplies to all reactor coolant pumps.
If the reactor is at power at the i'ee of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly.
Normal power for the reactor coolant pumps is supplied through buses from a transformer connected to the generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply coolant flow to the core.
Following any automatic turbine trip where there are no electrical or turbine faults which require tripping the generator from the network, the generator remains connected to the Bold /0verstrike Version 15.3-5
Enclosure to TXX-89472 July 31,1989 -
fg. Ib of 15 CPSES/FSAR
'negyork for approximately 30 seconds. The reactor coolant pumpf buses N
- remain connected to the network dddd/dfd/, thus ensuring full flow for 30 seconds after the reactor trip before any transfer is made.
For a manual turbine trip, or a turbine trip due to certain turbine faults or certain generator protection signals, the generator remains connected to the network for approximately 11.5 seconds.
For a complete description of the turbine / generator trip logic, see Figure 7.2-1, Sheet 16 and Figure 10.2-le Bold /0verstrike 15.3-6 Version