ML19319E592
| ML19319E592 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/31/1974 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| References | |
| NUDOCS 8004140595 | |
| Download: ML19319E592 (102) | |
Text
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SEMI-ANNUAL OPERATING REPORT
. ARKANSAS NUCLEAR ONE - UNIT I DOCKET NO. 50-313 LICENSE NO. DPR-51 3RD 6 tTH QUARTERS 1974 THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE OFFICE OF REGULATION. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD ANS MUST BE RETURNED TO THE CENTRAL RECORDS STATION 008. ANY PAGE(S)
REMOVED FOR REPRODUCTION MUST BE RETURNED TO ITS/THEIR ORIGINAL ORDER.
E O DEADLINE RETURN DATE -
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MARY JINKS, CHIEF CENTRAL RECORDS STATION O
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Those Environmental Monitoring Results as required by Technical I
Specification 6.12.2.3 (f) are not included in this report as a i result of procedural problems. These results will be provided (
- by March 21, 1975. ;
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>kvf SAFETY RELATED OfANGES IN PLANT DESIGN AND PROCEDURE DESIGN CHANGE REQUEST NO. 113 DESCRIPTION - Modification of stack radiation monitoring system by the addition of charcoal . sample filters for detemination of iodine re-leases, the addition of sample valves for portable sample collection and.the addition of flow volume recorders for stack and sample flow volume.
COMPLETION DATE - 10/31/74 SAFETY EVALUATION - This change is safety related in that it is re-quired to meet Environmental Technical Specifications. It is not considered an unreviewed safety question as it increases the effec-tiveness and efficiency of the stack monitoring system.
DESIGN CHANGE REQUEST NO. 126 DESCRIPTION - Chaage in control logic to prevent filling of waste gas decay tanks with nitrogen when draining the waste gas surge tank.
COMPLETION DATE - 8/27/74 p SAFETY EVALUATION - The Gaseous Radewaste System is safety related in f I that it is required to protect the safety of the public and of ,lant V personnel under those circumstances which would forceably produce radioactive gaseous effluents.
It does not constitute an unreviewed safety question in that the change only prevents the Waste Gas Decay Tanks from filling with nitrogen and does not affect atmospheric effluent releases.
DESIGN CHANGE REQUEST NO. 130 DESCRIPTION ~- Modification to control rod drive position indicator tube reed switches to prevent spurious power unlocks.
COMPLETION DATE - 8/28/74 SAFETY EVALUATION - This change was deemed as safety related in that it involves a safety-related system; i.e. the Control Rod Drive System.
4 No unreviewed safety question was determined from this change in that Technical Specification limits remained intact, thus no reduction in
- safety _ margins were made.
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( - DESIGN CHANGE REQUEST'NO. 142
' DESCRIPTION - Removal of parallel solenoid valve for reactor building '
purge exhaust isolation ' control valves to prevent failure of either valve from resulting in loss of instrument air.
COMPLETION DATE - 11/11/74--
SAFETY EVALUATION - This change was considered safety related in that
, the subject solenoid valves are part of the reactor building purge system and are involved with reactor building isolation.
- An unreviewed safety question was not found by the following reason:
In parallel, the solenoid valves'are unsafe due to possible result of loss of instrument air. In series, the solenoid valves are no more reliable than one would be. Deletion of the redundant solenoid valves precludes an unsafe condition and still maintains capability to cope with a single failure due to motor operated isolation valve inside the reactor building.
DESIGN CHANGE REQUEST NO.156 DESCRIPTION - Change in location of IcVel transmitter piping connec-tion to the borated water storage tank to reduce the possibility of contamination of the tank's contents.
(v) COMPLETION DATE - 11/19/74 SAFETY EVALUATION - This change is safety related in that it is related to the level indication for the BWST which provides makeup for decay h' .t. removal and provides borated water under accident conditions.
It is not considered an unreviewed safety question in that the change reduces the chance of metal pieces entering the BNST but still pro-vides a reference leg for the level transmitter which is more reliable than the existing atmospheric reference leg due to fluctuations in
, atmospheric pressure and vapor pressure within the tanks. Thus, the i
design change er.iances reliability and increases assurance of capa-bility of the dec ay heat mystem to maintain a safe condition.
DESIGN CHANGE REQUEST NO. 167 DESCRIPTION - Modification to pressurizer electromatic relief pilot
- valve _ blow off line to prevent back pressure on pilot valve when electromatic relief valve is open.
. COMPLETION DATE - 9/26/74
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(d). SAFETY EVALUATION - This change is safety related by the fact that the pilot . is needed for proper operation of the electromatic relief valve from the pressuri er. Failure of this valve would mean loss of proper operation of the electromatic relief valve.
It is not considered an unreviewed safety question in that the change increases the margin of safety by decreasing the possibility of loss of the electromatic relief valve.
DESIGN CHANGE REQUEST NO. 196 DESCRIPTION - Replacement of erroded primary makeup pump recirculation orifice with different design.
COMPLETION DATE - 11/1/74 SAFETY EVALUATION - Degradation of primary makeup pump recirculation
- line could damage primary makeup pumps and thereby degrade performance of HPI . system and other safety related functions of the primary make-up pumps.
An unreviewed safety question is not constituted as this design change ,
enhances the reliability of the makeup pumps, n
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EOUIPMENT f, FUEL PERFORMANCE Power' operation was begun with power testing on August 14, 1974. Reactor power was initially increased to 15% FP on August 16, 1974 and the generator was initially loaded on August 17, 1974. Power' testing at 15% FP continued until 8/21/74 when the Remote Shutdown Test was performed. The unit was cooled down following this test for miscellaneous secondary plant repairs and adjustments.
Power operation resumed on 9/1/74 and Integrated Control System and Turbine EH Control Adjustments were performed for several days. Feedwater control diffi-culties consumed the largest portion of the time. 25% FP Reactor power and 230 MWe gross generator output were achieved during this testing period which coded with a shutdown on 9/11/74. The unit was bsequently cooled down for repair of a failed generator exciter bearing w) i had forced the shutdown.
The Reactor was returned to power on 9/23/74. 40% FP/375 FMe~was achieved on 9/24/74 and power testing at the 40% FP plateau continued until 10/3/74 when the Reactor was tripped for test purposes. The unit was restarted on 10/4/74 and escalation toward 60% FP was begun. However, a flange steam leak at a moisture separator / reheater forced a shutdown on 10/6/74.
The unit was returned to power operation on 10/14/74 and 60% FP/535 FMe was achieved on 10/15/74. Operation was briefly interrupted due to a feedwater pump trip. Power was again returned to 60% FP when excessive RCS leakage h
U was-detected on 10/17/74. A small crack in a weld on an RCS drain line was found and the unit was cooled down for repair. The plant was returned to power.on.10/23/74 after repair of the drain line Icak and power operation at 75% FP/660 hMe was achieved. However, RCS leakage again became excessive and it was determined that the RCS drain weld had again failed. Redesign and repair of this drain line required several days.
The unit was ryturned to power operation on 11/17/74 and 75% FP/660 MWe was achieved on 11/19/74. Power testing at the 75% FP plateau continued for several days with some interruptions for secondary plant repairs and Reactor Coolant Pump power monitor relay trips. Power was increased toward 95% FP on 12/5/74 with several stops at intermediate test points. On 12/6/74, 95%
FP/800 hMe operation was achieved, but a short in the controller of a Reactor Building main chiller caused a loss of vacuum pumps and subsequent Reactor and turbine trip.
Repairs were made and the unit was returned to power on 12/7/74. 100% FP/
S75 FMe operation was reached on 12/9/74 and power testing at this plateau was carried out until the turbine trip test from 100% FP on 12/11/74.
Power operation resumed on 12/12/74 but was interrupted due to a failure of a CRD stator. The unit was cooled down for the stator repair. The plant was returned to power operation on 12/16/74 and reached 1001i FP on 12/18/74.
Final testing was performed and the unit was ' declared commercial on 12/19/74.
Pover operation continued at full power until 12/27/74 when an outage was required to repair a steam line leak. No further power operation followed p/
U in 1974.
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! The performance of the -nuclear fuel is demonstrated by the attached . figures I
..which show the relative radial. cower distribution at various ' increments of j core burnup. All measured powe'r distributions agreed very well with ex- '
l pected values indicating good fuel performance to date.
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SYl4EIRIC 1/8 CORE SPND STRIID/
CORE GRID CROSS-tter et.NCE Core Centerlines 1 2 4 10 14 21 30 37 _
H-8 H-9 F-8 H-5 N-8 H-13 B-8 H-1 3 6 5 20 29 31 45 -
G-9 F-7 E-9 K-12 C-9 B7 R-7 12 17 27 28 44 46 L-6 M-10 D-10 C-10 P-6 R-10 26 33 42 49 E-11 D-5 0-5 M-14 41 48 51 N4 0-12 D-14 52 C-13 s
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Quadrant Centerline 1 SPND SIRIn IEEER 4-8 --CCRE GRID O
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STEADY STATE RADIAL POWER DISTRIP2UTION "EI.ATIVE PEAKING FACTORS DATE: 9-26-74 TIME: 2000 Core Centerlines 1 2 4 10 14 21 30 37 1.023 1.278 1.266 1.383 1.270 1.252 1.355 3.958 -
0.98 1.29 1.27 1.43 1.12 1.28 1.35 3.96 6 b a 40 49 bl 45 1.244 1.362 1.232 1.257 1.041 1.043 3.748 1.21 1.47 1.21 1.27 1.00 0.97 1.78 12 17 27 28 44 4n 1.290 1.305 0.979 0.983 0.719 J.445 1.24 1.32 0.95 0.94 0.64 3.43 26 33 42 49 1.065 1.073 0.859 0.703 1.02 1.11 0.82 0.71
[ 41 48 51 (y 0.912 0.856 0.514 0.90 0.91 0.58 52 0.533 0.67s N, l N Qtladrant t Centerline K - SPND NINBER X.XX - MEASURED P/p-X.XXj - PREDICTED P/p Predicted Conditions: Measurement Conditions 40% FP Reactor Power 41 % FP Rod Group Pos.itions EFPD 3.97 l Gp 6 @ 75% WD BORON (PPM) 1154 i Gp 7 3 0% WD Rod Group Conditions Gp 8 0 30.8% WD Gp 5% WD 100 1
, 4.0 EFPD Gp 6% hD 74 f l (j -4.6% Imbalance Gp 7% KD Gp 8% WD 30 0 ,
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! IMBAI.ANCE -5.87%
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\ '-',h STEADY STATE RADIAL POWER DISTRIBilTION RELATIVE PEAKING FACTORS DATE: 11-21-74 TIME: 0500 i
Core Centerlines j 1 2 4 10 14 21 30. 37 1.035 1.306 1.2S2 1.405 1.134 1.225 1.311 0.90F -
1.00 1.32 1.30 1.46 1.13 1.25 1.28 0.95 3 o 5 20 29 31 45 1.260 1.372 1.249 1.266 1.041 1.006 0.737 1.25 1.51 1.24 1.27 0.99 0.92 0.74 12 li 27 28 44 4h 1.312 1.339 0.981 0.989 0.689 0.440 l 1.27 1.34 0.96 0.94 0.62 0.40 26 33 42 49 1.090 1.062 0.876 0.701 1.05 1.12 0.82 0.69 41 46 si
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s, 0.925 0.854 0.507 0.91 0.91 0.53
- 4 52 0.546 0.68 g N,
N .e Centerline X - SPND NUMBER X.XX, - MEASURED P/p X.XXl - PREDICTED P/p
. Predicted Conditions: Measurement Conditions 75% FI REACTOR POWER 75% FP Rod Grt up Positions EFPD 12.20 Cp 6 @ 75% WD BORON (PPM) 10 7 4,__,
Gp 7 @ 0% WD Rod Group Positions
,-s s Gp 8 @ 30.8% M'> Gp 5% WD 100 i 15.2 EFPD 71.1
, [( ,/ -13.0% Imbalance Gp 6% WD Gp 7% WD 0 Gp 8% WD 30.7 IMBALANCE -14.4*
O'- / STEADY STATE RADIAL POWER DISTRIBUTION i
RELATIVE PEAKING FACTORS DATE: 12-9-74 TIME: 2200 Core Centerlines
'l 2 4 10 14 21 30 37 1.094 1.292 1.238 1.393 1.192l.1.228 1.292 0.882 -
1.02 1.28 1.27 1.43 1.16 1.24 1.24 ,
0.88 3 e 5 20 29 31 45 1.222 1.327 1.227 1.268 1.048 1.013 0.734 1.22 1.45 1.23 1.27 1.00 0.92 0.74 14 li 27 28 44 46 1.275 1.337 07990 1.006 0.753 0.474 1.25 1.33 0.98 0.95 0.66 0.48 26 33 42 49 1.134 1.071 0.880 0.718 1.08 1.13 0.83 0.72 41 48 51
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0.908 0.842 0.503 0.92 0.90 0.58 ,
52 0.541 0.67 s Ns, N Quadrant Centerline A - SPND NUMBER X.XX - MEASURED P/p X.XX - PREDICTED P/p Predicted Conditions Measurement Conditions 100% FP REACTOR POWER 100% FP Rod Group Position EFPD 20.20 Gp 6 @ 87.5% ND BORON (PPM) 1079 Gp 7 @ 12.5% ND Rod Group Positions Gp 8 9 22.5% ND Gp 5% ND 100 25 EFPD Gp 6% WD 94.0 0~ -2.5% Imbalance Gp 7% WD Gp 8% ND 17.4 13.1 IMBALANCE -3.58%
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- t. i STEADY STATE RADIAL POWER DISTRIBUTION !
i RELATIVE PEAKfNG FACTORS DATE: 12-18-74 TIME: 2200 .
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Core Centerlines 1 2 4 10 14 21 30 37 I 1.067 1.289 1.256 1.413 1.193 1.210 1.274 .862 '
1.02 1.28 1.27 1.43 1.16 1.24 1.24 0.88 l
3 6 5 20 29 31 45 1.232 1.364 1.242 1.280 1.059 .955 .715 1.22 1.45 1.23 1.27 1.00 0.92 0.74 12 17 27 28 44 46 '
1.302 1.339 1.023 1.018 0.719 0.,449 l 1.25 1.33 0.98 0.95 0.66 0.48 26 33 42 49 1.150 1.119 0.883 0.701
. 1.08 1.13 0.83 0.72 41 48 51 O- 0.929 0.852 0.504 0.92 0.90 0.58 __ ;
52 0.552 0.67 s N Quadrant :
Centerline X - SPND NUMBER X.XXi - MEASURED P/p X.XXl - PREDICTED P/p Predicted Conditions: Measurement Conditions 1 100% FP REACTOR POWER 98% FP
, Rod Group Positions EFPD 24 l Gp 6 8 87.5% BORON (PPM) 1070 Gp 7 0 12.5% Rod Group i Gp 8 8 22.5% Gp 5% WD 100 25 EFPD Gp 6% WD 89 g -2.5% Imbalance Gp 7% WD 14 Gp 8% WD 12 IMBALANCE +0.16% ;
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TECllNICAL SPECIFICATION CHANGES -
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[ Specification Change Request Date Approval Date l l ETS 2.1.1 Relocate condenser 11-20-74 .11-22-74 i
- outlet temperature t
! detector.in the dis- ;
charge canal and monitor j once per shift rather ,
! than-every two hours !
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N RESULTS OF SURVEILLANCE TESTS Surveillance testing was performed per Section 4 of the . Technical Specifi-cations and any discrepancies noted have been reported in Abnormal Occurrence Reports. ;
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i l RESULTS OF. LEAK RATE TESTS PERFORMED l
! None this period.
SUM 1ARY OF CHANGES, TESTS AND EXPERIMENTS REQUIRING AUTHORIZATION FROM THE C04 FISSION PURSUANT TO 10 CFR S0.59(a) a None this period.
i CilANGES IN PLANT OPERATING ORGANIZATION INVOLVING KEY SUPERVISORY i O PERSONNEL None this period.
RESULTS OF LEAK TESTS PERFORMED ON SOURCES THAT REVEAL THE PRESSURE OF 0.005 pCi OR MORE OF REMOVABLE CONTAMINATION None this period.
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Sil:fIAllY OF POWER GENERATION AllG. SEPT. OCT. NOV. DEC. 11)TAL Gross Thermal Power 48,883 292,680 268,441 257,976 1,121,328 1,989,308
- Generated (MWil) i Gross Electrical Power 5,978 92,863 84,688 87,159 384,810 655,498 Generated (MWil)
Net Electrical Power -12,389 73,329 81,901 68,963 360,558 572,362 Generated (MWil)
Time Reactor was 371.2 468.9 203.7 209.7 544.0 1797.5 Critical (llrs . )
Time Generator was 70.1 334.6 192.0 173.5 509.4 1279.6 On Line (llrs.)
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. t i-i 1 4 SIMta'lY OF REACTOR SHUTDONNS ! { ! s i 1
- t. t This summary covers all plant shutdowns from the period of August 6 1 (initial criticality) to December 31,1974.
f The' method af shutdown I3 s red as follows: Auto - Automatic Trip of Reactor Manual - Manual . Shutdown of Reactor Manual T-ip - Manual Trip of Reactor - i 5, ~. a
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Q)) & '(/) Stt' NARY OF Rl! ACTOR SilllTDONNS Number Date Duration flethod Type Cause Correction
--------- - - - - - - - - - ------------ ------------ --------- -- - -- -----AUGU ST 1974 --------------- --------------------------
1 8 Aug. 1.4 hrs. Manual llot Power escalation testing NA 2 9 Aug. 1.1 hrs. flanual. Ilot Power escalation testing NA 3 9 Aug. 1.3 hrs. Manual llot Power escalation testing NA 4 9 Aug. 1.2 hrs. Manual- llot Power escalation testing NA 5 9 Aug. 1.1 hrs. Manual llot Power escalation testing NA 6 9 Aug. 1.4 hrs. Manual llot Power escalation testing NA 7 10 Aug. 1.9 hrs. Manual llo t Power escalation testing NA 8 10 Aug. 1.1 hrs. Manual llot Power escalation testing NA :i 9 10 Aug. 1.7 hrs. Manual Ilo t Power escalation testing NA 10 11 Aug. 2.0 hrs. Auto fligh flux while withdrawing control rod assembly 7-4 11 11 Aug. 0.7 hrs. Manual llot power escalation testing NA
' 12 11 Aug. 1.1 hrs. Manual llot Power escalation testing NA 5 13 11 Aug. 0.9 hrs. Manual llot Power escalation testing NA
, 14 11 Aug. 1.0 hrs. Manual llot Power escalation testing NA 15 12 Aug. 1.2 hrs. Manual llo t Power escalation testing NA 16 12 Aug. 1.0 hrs. Manual llot Power escalation testing NA 17 12 Aug. 1.2 hrs. Manual llot Operator training NA 18 13 Aug. 1.5 hrs. !!anual llot Operator training NA 19 13 Aug. 3.5 hrs. Manual llot Operator training NA 20 21 Aug. 241.4 hrs. flanual Trip Cold Remote Shutdown Test, maintenance 6 const ection NA _
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,3 SIMIARY OF REACTOR SliiffD0lfNS Number Date Duration Method Type Cause Correction . -
------------------------------------------ SEPTEMBER 1974 ----------- - - - - - - - - - - - - - - - - - - - - - - - - - -
20, 21 Aug. 13.2 hrs. (Cont. from Aug.1974, see previous ' entry) , 21 1 Sep. 0.8 hrs. Man'ual llot Operator training NA i 22 1 Sep. 0.4 hrs. Manual llot 3perator training NA 23 1 Sep. 0.5 hrs. Manual llot 0perator training NA 24 1 Sep. 3.3 hrs. Manual llot Operator training NA 25 5 Sep. 3.3 hrs. Planual Trip Ilot Steam generator level increased due Restarted feedwater to overspeed of feedwater pump pump
, 26 12 Sep. 1.5 hrs. Manual llot Operator training and examination NA .
-, 27 12 Sep. 0.5 hrs. Manual llot Operator training and examination NA
" 28 12 Sep. 0.5 hrs. Manual Ilot Operator training and examination NA 29 12 Sep. 0.5 hrs. Manual llot Operator training and examination NA 30 12 Sep. 0.7 hrs. Manual llot Operator training and examination NA 31 12 Sep. 0.5 hrs. Manual llot Operator training and examination NA 32 12 Sep. 0.7 hrs. Manual llot Operator training and examination NA 33 12 Sep. 0.8 hrs. Manual llot Operator training and examination NA
- 34 13 Sep. 0.7 hrs. Manual llot Operator training and examination NA 35 13 Sep. 0.6 hrs. Manual llot Operator training and examination NA 36 13 Sep. 0.6 hrs. Manual llot Operator training and examination NA i 37 13 Sep. 0.3 hrs. Manual llot Operator training and examination NA 38 13 Sep. 221.5 hrs. Manual Cold Generator exciter bearing failure
- forced premature shutdown NA i
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O N ] O d V Sil'BIARY 01 REACTOR SilllTDOWNS Number Date Duration tiethod Type Cause Correction
------------------- OC10 BER 1974 ------------ --------------------------
71anual Trip
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39 3 Oct. 13.8 hrs. flot Power escalation testing NA 40 6 Oct. 117.6 hrs. Planual Cold Reheater gasket blew out Replaced gasket 41 11 Oct. 4.1 hrs. flanual Trip flot etain feed' eater pump B tripped Opened FW turbine exhaust butterfly 42 12 Oct. 50.8 hrs. Auto Cold !bheater gasket blew out Replaced gasket 43 15 Oct. 4.0 hrs. Auto llot flain feedwater pump A tripped Vibration sensor mal-function corrected 44 17 Oct. 197.1 hrs. blanual Cold Leak in 1 " drain line below "B" Shutdown G made repairs RC pump 45 25 Oct. 152.9 hrs. flanual Cold Leak in th" drain line below "B" Shutdown G made repairs RC pump
--------- - - - - - - - - - ----------- ------------ --------- -------- NOVEFIBER 1974 ------------- --------------------------
i 45 25 Oct. 399.3 hrs. (Cont. from October 1974, see previous entry) 46 17 Nov. 1.2 hrs. Auto llot blomentary interruption of station None power during electrical storm 47 22 Nov. 23.4 hrs Auto llot liigh flux indication while in- None creasing average temperature for power escalation testing. 48 25 Nov. 81.3 hrs. Auto llot blaladjusted pump /pover relays. Removed pump / power re-lays from cabinets G bench tested G reset. 49 28 Nov. 3.1 hrs, blant.a1 llot Incorrect boric acid. sample which Took reactor suberitical, caused estimated critical position verified boron concen-to be incorrectly predicted tration in reactor coolan :. 50 29 Nov. 2.0 hrs, Auto llot Reactor coolant pump power monitor Additional testing of contact bounce when starting 3rd pump / power relays. No pump for reactor coolant pump flow corrective action. test a
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StI4tti Y OF lif4C10R SillifD0h'NS Number Date Duration flethod Type Cause Correction
--------- D ECEMBER 1974 ------------ . .--------------------------
4 51 6 Dec. 21.2 hrs Auto - llot liigh pressure and temperature of teactor 52 11 Dec. .9.5 hrs Auto llot Turbine / generator trip resulting NA from electrical fault in RB chiller ciceuit-53 12 Dec. 28.5 hrs Manual Cold Open stator winding on control Replaced stator winding rod drive mechanism for control rod assembly 6-3 54 14 Dec. 42.0 hrs Manual Cold Packing blew out on CY-1009, Replaced i Pressurizer spray block valve.
- . - 55 27 Dec. 98.8 hrs flanual Cold Cracked resistance temperature Replaced resistance detector well in main steam line temperature detector well l
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- . MAINTENANCE i I
! This section discusses all safety-related maintenance performed at Arkansas _ Nuclear One~- Unit 1.between August 6, 1974 and December 31, 1974. This . ! j includes maintenance work 'done by APSL crews and (as noted) maintenance done l j for AP6L by Bechtel Construction. r s l i t (
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ARKANSAS NilCLliAR ONE - II:;1T 1 PLAIN'IENANCE FIALFUNCTION, EFFECT PRECAllTIONS '
.IOB ON SAFE CORRECTIVE pog COA 1PONEAT CAUSE RESULT 540 CV-3823 Valve would not Potentially over- None Position switch adjusted 'None SW Discharge to close 100% fill emergency to indicate correctly timergency Pond pond from lake. when valve was closed.
549 Y-II Inverter Damaged cable in No'AC indication. None Replaced damaged cable in None SCR section of __SCR sect ion of inverter. inverter. . 557 Ilydrogen Purge Erroneous radia- liigh radiation None Repaired digital rate None G ICW Radiation tion indications. alarms in control meters. Ebni tors. room. a 559 PDT-1238 PDT-1238 out of Incorrect flu flow. None Re-calibrated PDT-1238 Observed P:r.
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and Flu tank Indicat ion - levels close-Incorrect , ly. 560 Y-25 Inverter Frequency meter No frequency indi- None - Installed new frequency Nonc not working. cation with Y-25 meter. . supplying AC load. , 563 Cas Radwaste Failure in instru- Incorrect flow None He-calibrated flow Terminated Discharge Flow mentation flow indication. - transmitter, blew out gas dis-
, string. sensing lines, checked charges, string calibration.
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. HALFUNCTIO.N, EFFECT PRECAUTIONS ~
.JO B CORRECTIVE ON SAFE FOR C0!IPONENT CAllSE RiiSilLT 564 CRD Reed Switches Sticking magnetic False assymetric None Bumped position indicator llot shutdown reed switches on rod indication in tube and sticking switch CRD absolute posi- control room. opened.
tion indicator. tube. _ 567 RCP Underpower Relays falsely lleactor trip. None \djusted sensitivity of None Relays, indicating RCP relay operation. trip. 570 Reactor Building Ilroken gear on Inner door and None Replaced gear on outer Cold shutdown Personnel llatch outer door. outer door open door.
- N Outer Door, interlock -
'[ _ defea ted. 572 CRD Reed Switches Sticking magnetic False assymetric None - Slight bump on position llot shutdown reed switch in rod indication in - indicator tube caused control rod drive control room. switch to open. absolute position indicating mecha-nism. 574 Wast Gas Com- Pressure switch Continuous waste None Re-calibrated pressure None pressor Controls on waste gas surgt gas compressor switch 4812. tank out of cali- operation. bration. w 0
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? R\LFilNCTION EFFECT Pl!!T AllTIONS ' J0P ON SAFE CORRECTIVli FOR ORDER GPERATIOy . ACTION
' ~ REACTOR SAFETY CO?tPONENT CAllSE RESOLT 577 PDT 7442 PDT 7442 out of Incorrect hydrogen None - Filled transmitter reference None Reactor Building calibration. purge flow, leg with water and adjusted Ilydrogen Purge zero.
Flow' 585 G212D Inlet to Pla tadjus tmen t Yalva would not N.mne 'djusted A salve linhag - T18n not in g-O ste Gas Decay close 100%. to perm a 100*. close, use. Tank T18D 587 Waste Gas Com- Damaged suction Compressor would None Replaced suction and dis- Ilse of pressor C-9B and discharge not compress air. charge check valves. standby check valves. _ s compressor. U$ -
. 588 SS1011A Reactor Dripping Reactor ) None Tightened packing. None Coolant Sample Coolant water on > Result Valve sample room floor. )
Eroded packing )Cause 589 CV-1215 Valve could not Unable tu switch None loasuned packing to Lno letdown cooler be operated with letdo.in collors. free valve. E29B inlet valve. hand. switch on control console. _ 590 GZ-12D Inlet to Valve stroke Valve leaking None Valve stroke adjustment. None Waste Gas Decay maladjustment. through. Tank T18D i t
N p f\ d C U ARKANSAS NUCLEAR ONE - II::IT I 41!.'TENA.NCE EFFI-CT PRECAllTIONS' JOB FIALFUNCTION CORRECTIVE ON SAFE FOR CO'IlONI.?fi' CAllSE RESllLT 591 CV-1235 Leaking valve Leaking reactor None Tightened valve bonnet. None Pressitrizer Level bonnet. ' coolant in reactor Control Valve auxiliary buildini . 592 CV-1222 Valve bonnet leal Leaking reactor None , _ Efforts to t ighten bo net N,)ne Letdown Block coolant in reactor holts did not eli.ainat e Grifice Isolation auxiliary buil1- leaLing conlition. P. e - Valve ing. pair postponed until later. 593 C98 Waste Gas llanging dischargt Compressor would None Freed discharge check Standby Compressor check valve. not compre valve. Waste Gas l Compressor in Service. u ~ 8
.596 "B" Makeup Pump Trash under valvt Relief valve None Removed relief valve and None 6 :lon Relief seat. weeping to waste. cleaned.
Valve 598 Reactor Building Leaking hydrau- Loss of hydraulic None Tightened valve pael'ing None Personnel Lock lic line fitting- oil. and tubing fittings. ilydraulic System. and valve pack-ings. . 599 C9A Wast Gas. Sticking compres- Compressor drive None Cleaned discharge.. check Standby Compressor sor discharge motor overload. valves. Compressor valve. in service. J
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C O T AREANSAS NtJCLEAR ONE - IINI'l 1 ilAINTLN\NCE EFFECT PRECAUTIONS ~ JOB FtALFilNCTION CORRECTIVE FOR ON SAFE s.'O'IPONENT CAtISE RESilLT 600 1 ,erter Y-25 Defective static Inverter would no1 None Replaced defective static None transfer and return to normal transfer and current sens-current sensing po+.er source ing board, board. - at'ter being placed on alter- _ nate source.
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ARl*ANSAS NilCLlan UNE ITNIT I llAI.'.' TEN \NCli Fl\LFilNCTION EFFEcr PRECAUTIONS' JOB ONS5hE CORRECTIVE FOR ORDER OPERATIT. ACTION REACTOR SAFETY C051PONENT CAllSE RESULT 602 CV-1207 Eroded valve Inadequate flow None Replaced plug and seat Plant shut down RC Pump Seal internals. control. ring. Injection Flow
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Control - 603 Pitted Relief ) Leaking contents None Lapp 61 valve seat. None Valve Seat )Cause waste gas decay Relief Valve on ) tank to waste gas _,. Wast Gas Decay ) Component surge tank. Tank ) i 607 Reactor Coolant liigh background Detector readi'ng False Set digital ratemeter None n Leak Detector radiation. high. indica- spectrometer to be sensi-tive to Xe-135.
'- tion of reactor coolant leakage .
609 RPS Channel "li" Incorrect RTD Channel indicating None Eliainated difference in RPS Channel Temperature external lead low temperature external RTD lead resis- "B" in manual Channel resistance com- i'n comparison to tance. bypass. pensation. other three chan- - nels. 610 CV-7402 - RH Valve could not Defective air None Defective air solenoid Nonc
- Purge Isolation solenoid valve. be opened. vs.lve replaced.
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MALFilNCTION EFFECT PRECAtITIONS'
.10B ON SAFE CORRECTIVE FOR ORDER L1PERATIO.g ACTION COSHOlENT CAllSE . REAC40R SAFETY REStILT 613 PSV-1245 Pitted valve seat. Valve leaking None Lapped seat. None Purification through.
Demineralizer ' T36B Relief Valve . .
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614 "B" Makeup Punp Leaking seal Oil leaking on None Rescaled seal plates and None Speed-Increaser plates and pipe gamip base. (l' ump irl resoldered pipe joint, Joint. l Service) tightened up loose con- l nections after removed from service for monthly - e check. x y 617 PCV-2699 - Main Valve failed to Excessive blow- -None Disassembled valve and None
. Steam Relief rescat at'd5 ired back and loss of found blowback adjusting Valve pressure fonowing
' steam pressure on ring locking pin improp-reactor and turbine main steam header. ~ erly engaged with blew-trip fiom 15% power - back adjusting ring. for rem ste shutdown test. 618 CV-2618 f CV-2668 Tight shutoff not Excessive steam None Disassembled valves, hone l None Main Steam Atmos- possible with leakage. pheric Dump Valves existing valve seat, cleaned piston rings. and cleaned up galled internals. places on pistons and cylinder walls. ge W 4
(0 e ,V T ARKANSAS NilCLEAR OM - IINIT 1 FRINTENANCE EFFECT PRECAllTIONS' JOB. FIALFUNC. TION CORRECTIVE ON SAFE FOR ORDER ~c)PERATIO ACTION ' REACTOR SAFETY CAllSE COMPONENT ._ RESilLT 620 CV-1008 Eroded bonnet Reactor coolant Magnitud e CV-1008 isolated and re- Close observa-Pressurizer Spray gasket. leak. of leak placed bonnet gasket, tion of RCS Control Valve threat- - pressure. ened ' reactor ._ shutdown . 621 CV-1213, CV-1214, Excessive packing Reactor coolant Nor.e Backseated g ives and None CV-121S, and leak. leak in reactor repacked. CV-1216. . RC building ~ Letdown Cooler E-29A, E-29B _ s S Inlet and Outlet - e Isolation Valves. - 631 C9B Waste Gas Motor cooling fan Metallic rattle- Nonc - Replaced fan. in correct None Compressor Motor loose on sha ft . ing noise when - position on shaft. motor started. 631 CV-7447 - RB Valve packing Valve torquing None Adjusted valve operator Nonc isolation Valve excessively tight. out in closing torque switch to permit Ilydrogen Purge direction. . reliable operation. System. 637 PS4812 Start /Stop Improper set point Waste Gas compres- None Re-calibrated PS4812. None Permissive Switch PS4812. sors would not for Waste Gas start. Compressors. OM Q ______--m - _ _ _ _ - _ _ - _ _ _-____m _ __ m- _ . _
[ i . } N N' TR12NSAS NIfri. EAR ONE - IIN!'l 1 ?t\lNTE'W:CE EFFECT PRECAUTIONS ~
. I 0 11 MALFUNCTION CORRECTIVE ON SAFE FOR COMPONENT CAIISE RESULT 638 Diesel Generator DC Field Ampmeter No generator field None Replaced ampmeter. Diesel Generator
#1 Field Ampmeter inoperable. current indicated. #2 in standby.
639 PSV-1249 - HU Tank Valve cap unsea. led Waste gas surge None Scaled cap None Relief Valve. tank leaking through cap. ~ -
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646 PSV-1000 Eroded seat and Valve leaking to None Replaced disc'and lapped Cold Shutdown. Pressurizer disc. reactor coolant seat. - Electromatic quench tank. Relief Valve 64 7 PSV-4812 - Waste Trash under seat. Valve leaking' .None Inspected, cleaned, and None Gas Surge Tank slightly to checked setpoint. 8 Relief Valve, station vent. 650 Reactor Building Insufficiently Excessive accu- None Removed access doors from None Coolers. sized drain lir.es mulation of water cooler housings to prevent 3 on coolers. in coolers during hausings from filling with accident condi- water. tions. 651 Control Rod Drive lindetermined. Spurious motor None~ Group S regulating supply Reactor shut System. faul t indications. programmer control system down. checked out and no faults were found. m N 4 1
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JOB CORRECTIVE ON sal'E FOR OllDER . ACTION OPEFATIO , REACTOR SAFETY e CO*1PO"ENT CAllSE RESilLT l 653 PDT1209 T. PDT1210 Loss of high Sensing lines to None Transmitters properly None llPI Flow Trans- transmitters im- pressure injection vented and checked out. mitters. properly vented. flow indication. 660 RE-8018 1.oose connection Fa i a t.re a l t ria None Tightened loote connection None Area Radiation rm radiation on <ligital on detector. Planitor in I!Il ,le t ec t o r . ratemeter. Personnel flatch. _ 661 "A" Core Flood Reference leg en liigh level indi- None Filled refercoce leg and Redundant Tank Level transmitter under cation, blew down to reference level indi-
' Transmitter filled - '
level. cation. t - o
, 665 RE-8018 Failed detector Excessive count None Replaced Gl tube in None Radiation Element tube. ratemeter fluctu- detector.
in RB Personnel ation. Access llatch. 666 RC-2236 - Process lilectronic com- Failure alarm on None Installed input comparator None Itadiation Monitor ponent fai lure control casesole, in digital r:itemeter,
, in "A" Intermedi- in digital rate-ate Cooling Watet aeter. -
Loop. 668 B-1 Core Flood Transmitter Transmitter read- None Filled reference leg and Standby trans-Tank Level Trans- reference leg ing high. blew down to reference mitter in mitter. under filt ed. level. service. E m i I t
's ARIL\NSAS .NUCll3R ONF - IINIT 1 flAINTENANCE EFFECT PRECAUTIONS' MALFUNCTION CORRECTIVE FOR JOS ON SAFE ORDER- l)PERATIO.s ACTION . REAC'IUR ' SAFETY C03fPONENT CAUSE. RESULT' 669' PDT-1247 - Makeup Leaking plug in Transmitter re- None Tightened plug in rear None .
4 Filter Differen- rear of trans- liability in of - t ransmit ter , mitter, doubt. - t ial '.Transmi tter. , 671. PDT-1209 - liigh Damaged. bellows Erroneous high None . Replaced bellows in Reactor shut Pressure Injection in transmitter. pressure injec- transmitter. - down, t. Flow Transmitter. tion flow. 676 Battery Charger Fans loose on Fan blades None Installed new fans. None DD3 and D$4 drive shaft. damaged. Coo 1ing Fan.
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u 680 MU-31B Seal Loose bonnet Slight bonnet -None Tightened bonnet bolts. None
" Return Cooler- bolts, leak.
8 Outlet Isolation ~ Valve. 681' CV-.2620 Dirty contacts on Valve torquing None Disassembled and cleaned Redundant Emergency Feed- motor operator out before reach- contact s on torque switch. emergency Water to Steam . torque switch, ing 100*. open feedwatcr Generator "B". .and/or 100% valve operable. closed position 3. - t 681 CV-2806 Service Dirty contacts Valve torquing None Disassembled and cleaned Service Water Water to Emer- on torque switch, out before reach- torque switch contacts to redundant gency Water Picp ing 100% open and assured proper oper- emergency FW r~ P7A. position. ation. pump assured. i _ __ _ - _ . _ _ _ _ _ _ _ _a -- _____.,w_2m . - . . _ _-,.m- _ _ - - -- - _. _ , _ _ . -m,
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()) ARTANSAS NtlCLEAR ONE - IlNIT 1 Fi\INTENANCE EFFECT PRECAlfTIONS ' JOB MALFUNCTION CORRECTIVE ON SAFE FOR ORDER - 0PERATIO.5 ACTION REACTOR SAFETY C0ftPONE?ff _ CAllSE RESilLT 685 CV-2216 "A" Let- Broken pins in Valve wot Id not None- Valve disassembled and None down' Cooler butterfly valve. open, replaced broken pin. Isolation Valve. 688 PSV-1000 Erod ed d '. .,c . . Valve leaking to None Disassembled valve and Reactor shut Pressurizer reactor coolant replaced disc and lapped down Electromatic quench tank. seat < _ Relief Valve. 690 CV-2668 Tight shutoff not Excessive steam None Disassembled va'Ive and Cold shut down Main Steam possible with leakage. inspected and reassembled. Atmospheric Dump existing valve Valve internals. _ s y 692 Control Rod Drive IJndetermined Spurious indica- None Replaced direction error None System tions of control cards in auxiliary and rod drive stator - regulating power supplies. faults. .
-695 Control Rod Drive Loose connection Operation of None Resoldered loose connection None Position Indica- on "on control" switch producing on switch, tion Panel test switch. po response.
Mal function. - 696 Control Rod Bent pin on elec- Improper operation None Straightened bent recep- Reactor shut Drive Group 7 tronic component of Group 7 regu- tacle pin. down. Regulating Power card. lating power Supply. supply. S
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ARl'ANSAS NUCLEAR ONE - IINIT 1 MAINTENANCE EFFECT PRECAUTIONS' JOB MALFUNCTION CORRECTIVE ON SAFE FOR COMPONENT CAtlSE RESULT 699 Area Radiation Maladjusted check Spurious alarms None Adjusted orientation of None Monitoring System. source on radia- in Area Radiation check sources in respect tion detectors. flonitoring System. to windows on detectors
. to eliminate excessive background caused by sourJe streaming.
700 ~ Reactor Coolant Dirty flow Low flow alarm. None Clean dirty flow indicator. None Leak Detector. indicator. s - 702 CV-2618 - Main Tight shutoff not Excessive steam None Disassembled, inspected, Reactor sliut Steam Atmospheric possible with leakage to installed new piston down.
. Dump Valve. existing valve atmosphere. .
rings and reassembled. c,. internals. c,. _ . 709 TE-1044 G TE-1045 Detector shorted Incorrect cold None. Replaced TE-1044, TE-1045 None Reactor Coolant out. leg temperature , (Dual Element RTD). (Redundant System Cold Leg indication. Element ser-Temperature De- viceable), tectors. 710 C9A Wast Gas Maximum pressure ' Compressor would None Adjusted pressure limit- None Compressor limiting mecha- not produce de- - ing mechanism to achieve nism out of sired discharge desired discharge pressure. adjustment. pressure. 9 b _ _ _-____ _ ._m-m__ __ ____
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J .- I ARKANSAS NUCLEAR ONE - UNIT 1 MAINTENA JCE EFFECT PRECAUTIONS' JOB - MALFUNCTION CORRECTIVE ON SAFE FOR C0!!IONENT -
. . . CAUSE RESULT 714 PSV-1001 Eroded seat and Valve leaking to None Replaced disc and recond'i- Cold shutdown. '
Pressurizer Code disc. reactor coolant tioned seat. Relief Valve. quench tank. 716~ CV-1050 - Decay Worn packing. - Dripping reactor None Backseated and repacked None
, lleat Suction Valve. coolant on floor yplve with decay heac of reactor build- sysresiiin service.
ing. . 718 CV-1207 - Reactor Eroded plug. Insufficient flow None Removed plug and recondi- Cold shutdown. Coolant Pump Seal control to reactor tioned seating surface. i Injection Flow coolant pump 8 Control Valve. seals. s
% 722 Pressurizer lleater Burned out circuit Unable to energize None Replaced burned out Reactor shut i Group 6. breaker control heater group 6. control transformer. down.
trans former. 723 Y-24 Inverter Defectiv; voltage Tripped circuit None Replaced faulty printed Reactor s' hut
- sensing and oscil- breaker. circuit boards and fuse, down.
lator printed circuit ' board s , blown-fuse. _ 726 RBV-70B Reactor Bent valve stem Leaking reactor - None Replaced valve stem and Cold shutdown. Coolant System and eroded disc. coolant into seat and repacked valve. liigh Point Vent reactor building Valve, vent header. I d g am W
rv % r) I ) (v' ! ARKANSAS NUCLEAR ONE - UNIT 1 MAINTENANCE
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EFFECT PRECAUTIONS' JOB MALFUNCTION ON SAFE CORRECTIVE FOR COMPONENT .__ CAUSE- RESULT 727 RBV-71B - Reactor Eroded seat and Reactor coolant None Reconditioned seat and Cold shistdown. Coolant Systen disc. leaking to RB disc. liigh . Point . Vent vent header. Valve. , 728 RBV.70A.- Reactor Eroded seat and Reactor coolant Vone _ Reconditioned seat and Cold shutdown. Coolant System disc. leaking to disc 7 liigh Point Vent reactor building Valve. vent header. , 729 N2 -38 and N2 -37 Eroded seat and Reactor coolant l'ossible Lapped valves and seat. Cold shutdown. Nitrogen St pply disc. leaking into aver-
, Valves to Pressu- nitrogen supply (>res sur-rizer. header and caus-- ing N ing overpressure. riitrogen , supply, header to pressuri- .
- er and both core flood tanks inside ~
i reactor - building. 730 CV-1000 Pressuri- Worn packing. Leaking reactor Mone Repacked valve. Cold shutdown. zer Electromatic coolant into Relief Isolation . reactor binilding. Valve. F 9 0
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\ v/ V ARKANSAS NUCLEAR ONE - UNIT l' MAINTENANCE MALFUNCTION EFFECT PRECAUTIONS ~
JOB CORRECTIVE ON SAFE FOR ORDER ' ACTION
- QPERATIOB REACTOR SAFETY
' COMi'ONEN T CAUSE RESULT 730 CV-1008 Worn packing. Leaking reactor None Repack valve. -Cold shutdown.
(Cont.) Pressurizer Spray coolant into Flow Control Valve . reacto'r building. 731 BW-6,B Reactor Leaking bonnet Leaking borated None Relilaced bonnet gasket. Cold shutdown. Building Spray gasket. ~ water storage Pump Suction Check tank contents - Valve. on floor near spray pump. 740 GZ-11A - Waste Ruptured dia- Gas leak into None Isolated valve Monitored gas Gas Compressor phragm. reactor auxiziary Replaced diaphragm. activity. Discharge Valve. building. t g ju 741 Power to Pump Relay contact None Adjusted contact holding
* ) Reactor shut-Relays in Reactor bounce )Cause coil to eliminate contact down.
Protection System. False sensing of ) - bounce. tripped reactor ) . coolant pumps, ) Result causing reactor ) tripa. ) 743 PSV-14 07 ."A" Valve body' loose Valve lifting at None Tightened up valve body
~ None Decay lleat Pump .and ga'g screw lower than de- -
and replaced gag screw. Discharge Relief missing. sired pressure Valve. and spraying Dil l Pump Room. 744 PSV-4812 Waste Eroded seat. Leaking gas to None' Cleaned internal parts Stack Radiation Gas Surge Tank station vent ,and lapped seats. Monitoring Relief Valve. plemP causing ' System In hi-rad alarms. Service. i
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v (v ]) v) ARKANSAS NUCLEAR ONE - UNIT 1 MAI?ffENANCE EFFECT PRECAUTIONS' JOB MALFUNCTION CORRECTI\,E ON SAFE FOR COMPONENT . CAUSE RESULT 746 Reactor Protection Maladjusted relays ) Cause None Lowered power setpoint from Reactor shut-System Pump to 801, to 50% and alecreased down. Power Relays. False sensing of ) time delay from 180 msee to tripped reactor - ) 150 mesc. coolant pumps, ) Result causing reactor ) trips. ) ~ , 750 ~ Reactor Coolant "B" phase grounded Reactor coolant None Repaired ground fault on Reactor shut-Pump Motor P32B. pump P32B tripped. "B" phase. , down. 752 CV-1000 Pressurizei Premature torque Valve failing to None Increased torque switch Reactor shut Electromatic Re- switch operation open and close setting. down. lief Isolation in open and closed completely. s Valve. direction.
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tj _ i 755 GZ-llB "B" Waste Ruptured diaphragm . Gas leaks into None - Replaced diaphragm. None Gas Compressor reactor auxiliary Discharge Valve. building. - 756 Reactor Protection False sensing of False reactor None Lowered reactor coolant Reactor shut System Pump to reactor coolant trip conditions. punp to power watt relay down. Power Relays. pump trips. trip set point from 501, to 25t. . s e
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ARKANSAS NUCLEAR ONE - UNIT 1 MAINTENANCE EFFECT PRECAUTIONS.' JOB MALFUNCTION CORRECTIVE ON SAFE FOR
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COMPONENT _ . - CAUSE RESULT l
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772 Stack Monitoring Excessively tight Pulley. shaft None Replaced sample pump. No gaseous System Sample. Pump. drive belt. bearing failure. releases while-pump out of
- service.
'779 Waste Gas System flooding Damaged compressor None Rebuilt compressor and None Compressor System. with water, diaphragas and replaced GZ-llA Diaphragm.
. suction and dis-3 charge check -- valves. Ruptured diaphragm in GZ-IIA l Waste
' Gas-Compressor _
, y ~ Discharge Isola- ,
,. tion Valve.
787 - CV-2620 Failed torque Valve torques None , Torque switch modified Cold shutdown. Emergency Feed- < witch. out in opening to allow safe operation - I water Supply to direction, until replacement parts i "B" Steam Genera- arrive. tor. 788 CV-1428 Loose hydraulic Partial loss of Non6 Tightened all fittings. None Decay lleat fittings. hydraulic fluid. Cooler Outlet i Flow Control Valve.
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V L) ARKANSAS NUCl. EAR ONE - UNIT 1 FttlNTENANCE MALFUNCTION EFFECT PRECAUTIONS' JOB ON SAFd CORRECTIVE FOR , COMPONENT . CAllSE RESill.T 791 Control Rod Drive Stator motor Assymetric rod None Replaced stator motor. Cold shutdown. Mechanism - Rod 3, burned out. condition in Group 6. control room.
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800 PSV-2699 - Main Blowback adjust- Sb psig excessive None A slight adjustment to None Steam Relief Valve ment rings slight- blowback. --uppeg and lower blowback ly out of ad- adjusting rings. Justment. . 802 P32D Reactor Transmitter body Binding movement None Relievedstrainontrans- Reactor shut Coolant Pump in strain. and failed elec- mit'ter body and replaced down.
, Bleed-Off Flow tronic components .
electronic components.
, Transmitter. -- '
- ao
.s 803 Air Flow Sensing Failed air flow Cabinet cooling None Replaced air flow sensing None
; Switch on Engi- sensing switch. fan failure . switch.
neered Safeguard alarm. System Analog Channel No. 2. 804 PSV-2699 - Main Slightly mal- Valve lifting None Increased popping pressure None Steam Relief adjusted blowback 'approximately approximately 50 psi and
- Valve. .. adjusting rings 50 psi-to low -
decreased blowback approxi-
- and popping pres- and blowing back .mately 50 psi.
sure setpoint. approximately 50 psi greater than desired before rescating. e w I Y
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ARKANSAS NUCLEAR ONE - IINIT I MAINTEMNCE MALFUNC TION EFFECT PRECAUTIONS' JOB. -CORRECTIVE ON SAFE FOR ORDER ) OPERATIOy ACTION C0!!PONENT __ CAllSE REACTUR SAFETY' RESULT 808 PSV-2699 Main Confirm adjust- Valve setpoint None Adjustments made and Steam Relief Reactor' shut. ments made on found to be 30 tested to confirm correct- down. Valve. JO #804. psig lower than setpoints, desired setpoint. 812 Nuclear Instru- Electronic com- Loss of +600 V None __ Replaced 600 V DC Power Reactor Pro-mentation Channel ponent fa i l ure. DC Power Supply Suppl?. tection System 8 Power Supply. in Channel 8. Channel D in manual bypass. 814 Engineered Safe- Channel 1 liigh Tripped ESAS None Replaced Bistable. None guard System Pressure Injectiot Analog Channel 1. Channel 1. (ESAS in Bistable malfunc-
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815 CV-1429 Valve position Valve position None Replaced linkage. None Decay lleat indicator linkage indication incor-Cooler Outlet fell off. rect. - Flow Control Valve. 816 Control R.od Drive Faulty electronic Unable to pull None Replaced faulty electronic Reactor shut Grcup 6 Regulating . component in con- Group,6 from - components, down when mal-Power Supply. trol rod drive core. - function control system. occured. m
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sm-p r~~.; ARKANSAS NtlCLEAR ONE - !! NIT 1 MAINTENANCE EFFECT PRECAUTIONS' JOB MALFUNCTION CORRECTIVE ON SAFE FOR. COBIPONENT .m CAllSE RESULT 817 Engineered Safe- Dislodged internal linable to take None Restored electrical con- Reactor shut guard System connector pin. manual control tinuity caused by dislodged down. Manual / Auto of CV-1228 after internal connector pin. Pushbutton for Engineered Safe-CV-1228. - guard System tripped. 818 - PSV-1001 Broken locking Unable to adjust None Restored locking pin on Cold shutdown. Pressurizer Code pin on top main setpoint of valve top spring washer. Relie f . Valve. spring washer. without damaging in.t ernal s . , 820 Nuclear Ins ru- liigh humidity in Nuclear instra- None Checked all cable connections None - Reactor u mentation reactor building, mentation - in reactor building. shut down. Channels 2 fr3. Channels 2 f. 3 ' malfunctioning. _ 822 Control Rod Drive Loose connections Overheating at None - Checked and tightened all Reactor s. hut-System. on DC busses. loose connections. DC busses. down. 834 Nuclear Instru- Dragging meter Disagreement in None Freed up dragging meter Nonc mentation System movement. redundant power movement. Power Range Linear - indications. - Amplifier Meter. 1 r
. 6 4
( b , b(~s . ARKANSAS NUCLEAR ONE - IINIT 1 MAINTENANCE MALFUNCTION EFFECT PRECAUTIONS' j JOB CORRECTIVE ON SAFE FOR I COMPONENT - CAllSE RESULT l 835 Control Rod Drive' Burned out stator Assymetric rod None Replaced stator motor. Reactor shut Mechanism - motor. conditions. down. Group 6, Rod 3. 837 FT-1209 4 FI-1210 Cold solder joint F16w indicator None Repaired cold solder None liigh Pressure Inj. on range potenti- failing to re- joint. Flow indicator. ometer. ' turn to zero after flow is stopped. , 842 RE-2236 Faulty electronic Erroneous radia- None Repaired electronic None Intermediate components in tion indications. components. Cooling Water digital rate- _ s N Radiation Element, meter. - 843 CV-1429 - Decay Valve position Incorrect valve None Replace valve position None lleat Cooler Out- indicator linkage position indica- indicator linkage, let Flow Control fell off. tion. - Valve. 84S RE-7442 Faulty " Test- liigh radiation None Repaired " Test-Operate- None !! Radiation, Ele- Operate-Reset" alarm. Reset" switch.
- i ment in Ilydrogen switch on digital -
0 Purge System. ratemeter. - 848 CV-1429 - Decay liydraulic pump Valve operable None Reset thermal overload and None lleat Cooler Out - tripped on over- by manual hand tested pump operation. let Flow Control load, pump only. Valve. 9
d O V ARKANSAS NUCLEAR ONE - IINIT 1 CONSTRUCTION MAINTENANCE
* (Sc 3 notel EFFECT PRECAUTIONS' MALFUNCTION CORRECTIVE ON SAFE FOR COMPONENT- CAllSE RESULT 1 Let down orifice Errosion of SliE ht increase None Replaced with better Cold shutdown.
in 'nake up 6 nrifice. in letdown flow. design orifice. purif ,;ation system. 2 RTD in RCS-loop -Moisture 4 heat Erroneous cold None. Replaced RTD Cold shutdown, leg temp indi-cation 3 RTD well in Improper initial Steam leak and None Removed and replaced with Cold shutdown.
, main steam line installation cracked RTD well shortened RTD well.
I g 4 Pressurizer dis- Inadequate Excessive None Added additional hangers Cold shutdown. charge line design movement during and hydraulic shock snubbers, sway transients suppresors, bars and modifi-cations to existing hangers.
- NOTE: This is maintenance work done for APSL by construction during report period.
I'N b,x /] . NON-SAFETY RELATED CHANGES, TESTS AND EXPERIMENTS DESIGN CHANGE REQUEST NO. 102 DESCRIPTION - Modification of Hydrogen-Oxygen sample collection system to allow proper collection of sample and purging of sample hood. COMPLETION DATE - 8/10/74 DESIGN CHANGE REQUEST NO.103 DESCRIPTION - Connection of Emergency Communication System desk set in control room to vital 120 VAC power source for better reliability in the event of an emergency. COMPLETION DATE - 8/14/74 DESIGN CHANGE REQUEST NO. 104 DESCRIPTION - Addition of domestic water supply line to the outside of the circulating water intake structure for use in washing down equip-ment. 1 COMPLETION DATE - 12/6/74 s- - DESIGN CHANGE REQUEST NO. 107 DESCRIPTION - Addition of valve and pipe tee in line to provide water or steam to main steam block valve cavities for testing. COMPLETION DATE - 8/14/74 DESIGN CHANGE REQUEST NO. 112 DESCRIPTION - Relocation of air sets in instrument air system to pre-vent bleeding down of air accumulation in a post accident loss of instrument air situation. COMPLETION DATE - 8/14/74 DESIGN CHANGE REQUEST NO. 118 DESCRIPTION - Elimination of condensate flush to the Reactor Coolant Pump standpipes and standpipe low level alarm due to re-evaluation of need for condensate flush. COMPLETION DATE - 16/9/74 O V v. m ( Y
%J DESIGN CHANGE REQUEST NO. 122 Installation of bypass around radiation detector in condenser vacuum pump exhaust line to permit quarterly calibrations of radiation de -
tector. COMPLETION DATE - 11/5/74 DESIGN CHANGE REQUEST NO. 128 Correction of incorrect wiring connections to provide proper computer monitoring of main feedwater pumps discharge conditions. COMPLETION DATE - 10/9/74 DESIGN CHANGE REQUEST NO. 133 Additional level transmitter added to resin storage tank to provide redundancy and reduce need for entry into high radiation area for maintenance. COMPLETION DATE - 10/9/74 DESIGN CHANGE REQUEST NO. 139 O)
+
k/ Addition of instrument air to drumming station equipment for opera-tion. COMPLETION DATE - 9/11/74 DESIGN CHANGE REQUEST NO. 141 Change in control logic for turbine bypass valve and turbine header pressure controls to allow only single loop control and prevent i cycling. COMPLETION DATE - 10/9/74 DESIGN CHANGE REQUEST NO. 160 Addition of heater in raw water metering valve pit to prevent freez-ing of water during cold weather. COMPLETION DATE - 10/29/74 l 1 l DESIGN CHANGE REQUEST NO. 165 Change in wiring of instrument electrical power supplies to reduce possibility of an electrical short. COMPLETION DATE - 1/19/74 l u I
('~x i
. DESIGN CHANGE REQUEST NO. 169 I Change in control logic for main feedwater turbine to prevent the bypass of low suction pressure switch when the low flow switch is bypassed.
COMPLETION DATE . 11/19/74 DESIGN CHANGE REQUEST NO. 173 Replacement of eroded orifice in reactor coolant system letdown line with different design. COMPLETION DATE - 9/26/74 DESIGN CHANGE REQUEST NO. 17S Addition of block walls in auxiliary building for improved radiation shielding. : COMPLETION DATE - 10/4/74 DESIGN CHANGE REQUEST NO. 182 ] f, Megawatt thermal conyt-eter modification to eliminate electronic noise g pickup in signal transmission lines between megawatt thermal converter
\ and integrated control system.
1 COMPLETION DATE - 11/IS/74
-DESIGN CHANGE REQUEST NO. 201 Change in control logic for main feedwater turbines to trip main turbine upon loss of both feedwater turbines.
.CMPLETION DATE - 11/8/74 DESIGN CHANGE REQUEST NO. 21S Addition of globe and check valves in chemical addition. system to assure minimum recirculation fo. boric acid pumps.
4 COMPLETION DATE - 10/31/74 tu '
- 46 ,
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ARKANSAS NUCLEAR ONE-UNIT 1 FINAL SAFETY ANALYSIS REPORT ( AMENDMENT NO. 47 1he following pages of Arkansas Power 6 Light Company's Arkansas Nuclear One-Unit 1 Final Safety Analysis Report are to be removed and, where specified, additional pages dated March 1, 1975, are to be inserted. ^ Remove Insert VOLUME I CHAPTER 1 1-1, 1-la 1-1, 1-la VOLUME II CHAPTER 7 l 7-21, 7-22 7-21, 7-22 7-23, 7-24 7-23, 7-24 Figure 7-25 Figure 7-2S CHAprER 8 8-1, 8-la 8-1, 8-la 8-6, 8-6a 8-6, 8-6a CHAPTER 9 9-7, 9-8 9-7, 9-8 9-27, 9-27a 9-27, 9-27a Figure 9-3 Figure 9-3 Figure 9-5 Figure 9-5 Figure 9-7 Figure 9-7 , Figure 9-16 Figure 9-16 CHAPTER 10
! 10-1, 10-2 10-1, 10-2 CHAPTER 11 l 11-7b , 11-8 11-7b , ll o Figure 11-1 Figure 11-1 Figure 11-3 Figure 11-3 VOLUME III_
14-11, 14-12 14-11, 14-12 O
Remove Insert VOLUME IV AEC QUESTIONS i 1.10a l 8.12 1.10a Figure 8.13-1 8.12 9.69 Figure 8.13-1 i 9.69 l 1
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[ . (v) 1 IITfRODUCTION AND SIWARY 1.1 IKTRODUCTION
'Ihis Final Safety Analysis Report is submitted in support of Arkansas Power 6 Light Company's application for a license to operate a nuclear generating station designated as Arkansas Nuclear One - Unit 1. Middle South Utilities is a holding company holding the common stock of Arkansas Power 6 Light Company, !1ississippi Power 6 Light Company, Louisiana Power 6 Light Company, Arkansas-Missouri Power Company, and New Orleans Public Service Incorporated. l 47 The Unit is located in Pope County, Arkansas, about six miles West-Northwest of Russellville, Arkansas.
Arkansas Nuclear One - Unit l'is designed to operate at core power levels up to 2568 wt which, when the 16 mt contribution frc:n the reactor coolant pumps is included, corresponds to a gross electrical output of 883 we. l47 Site parameters, principal structures, engineered safeguards, physics, and certain hypothetical accidents are all evaluated at the design power level of 2%8 wt. l The nuclear steam supply system is a pressurized ster reactor type which is similar to many other PWR's operating or under o nstruction. It uses chemical shim and control rods for reactivity contrr1 and generates steam with a small amount of superheat in once-through steam generators. The nuclear steam supply system and two fuel cores will be supplied by the Babcock & Wilcox Company. [m) The general arrangement of major equipment and structures, including the C/ Reactor, Auxiliary, and Turbine Buildings is shown on Figures 1-2 through 1-11. Arkansas Pcwer & Light Company is fully responsible for the safe operation of the station. The design, construction and aid in the testing, and startup of the unit has been and will be supplied principally by Bechtel Corporation and the Babcock & Wilcox Company. Assistance has been and will be rendered by other consultants as required. Middle So.th Services will act as a consultant to Arkansas Power & Light Campsny du-ing design, construction and operation of Unit 1 with special emphad s being placed ' n "uclear Fuel Management. A representative of 21 Middle South Services w.h be an active member of the Design Revier Board during design and construction periods. (See Appendix 1 for their technical qualif'ications . ) 1.2 DESIGN SUMIC Y, 1.2.1 SITE CHARACTERISTICS The site consists of approximately 1100 acres providing for an O.65 mile exclusion radius. The site is characterized by remoteness from population centers, freedom from flooding, sound, hard rock for structure foundations, a reliable network for emergency power and favorable conditions of hydrology, geology, seismology snd meteorology.
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L . ,! 1-1 Amendment No. 47 March 1, 197S
1.2.2 POWER LEVEL The design and license power level for the reactor core will be 2568 LWt and all physics and core thermal hydraulics information in this report is based on that power level. An additional 16 FMt is available to the cycle from the contribution of the r'eactor coolant pumps resulting in a gross electrical output of 883 FMe. 47 O 1-la Amendment No. 47
.\farch 1, 1975 l
Reactivity shutdown margin provided by the safety rods is assured by diversifi-cation _of their power buses. This feature, as shown in Figure 7-1, utilizes l } four separate buses, each having a separate trip device, to power the safety 3
~'
/ rods. 'A. failure in one bus does not reflect into the other buses , therefore, a single failure in the distribution system for the safety rods does not pre-vent a plant shutdown.
The minimum voltage required to hold a drive in a withdrawn position is h2 volt DC per coil (2 coil " hold" mode). The probability of an external DC source be-ing applied to the control rod drive mechanisms downstream from the reactor trip points such that the CRA's are held in their withdrawn positions after a trip is not considered credible for the following reasons:
- a. The secondary trip devices in the Control Lod Drive System remove all DC power from the drives.
- b. Control rod drive power cables are terminated at only three points between the control rod drive cabinets and the drive mechanisms.
Two of these terminations are made outside and inside the reactor building electrical penetrations inside junction boxes containing only control rod drive power cables. The third termination is made in bulkhead connectors (one per drive) in the area of the reactor. The only other cables terminated in this area are the control rod drive instrumentation cables. The instrumentation cables are ter-minated in bulkhead connectors of a different size and configuration, therefore mismating of connectors could not be accomplished.
- c. No cable splices are permitted between termination points described, f~3\
(,_,/ d. DC systems from the batteries are not grounded and are equipped with ground detecting circuitry. In su= mary, series redundant trip devices having adequate rating, testability and a " split-bus" arrangement insure safety of reactor trip circuits. 7.2,2.3.2 Reactivity _ Rate Limits The desired rate of change of CRA reactivity insertion and uniform reactivity distribution over the core are provided for by the control rod drive and power supply design, and the selection of rods in a group. The motor, lead screw and power supply designs are fixed to provide a uniform rate of speed of 30 in./ min. The reactivity change is then controlled by the' red group size. To insure flexibility in this area, a patch panel has been included in the power supply to enable the interchange of rod worth between rod groups. Any rod may be patched into any group with the exception of Group 8. A linear reactivity insertion rate is provided for~ by the withdraw-insert se-quence of rod groups. As described in Section 7 2.2.2.1, the sequence and sequence monitor are interlocked to provide redundancy in the sequencing of groups 5, 6, and 7 Control interlocks are in the sequencing equipment and protection against total rod withdrawal accident is provided by .he RPS as analyzed in Section ik.1.2.2. p v 7-21
Uniform and symmetrical group insertion rate is provided for by synchronous withdrawal of all rods in that group. Such synchronous withdrawal is achievad by the design of the power supply. A group pcVer supply operates synchronously by having its load (h to 12 CRA motor windings) connected in parallel on the output of the SCR's. As the programmer gates on the SCR's all rods in a group have the same motor winding energized si=ultaneously, producing synchronous motion of the entire group. Speed-limiting is accomplished through the use of 60 H synchronous programmer motors. These motovs are povered, through transformers, from the same h80 V AC source as the remainder of the CRD system. Thus, the speed of rod motien is locked to the plant's AC power frequency which, in turn, is limited by the turbine controls. The turbine speed control system closes the governor valves at 103% of rated speed. Overspeed trip is at 111% and a backup overspeed trip is set at 111 5% of rated speed. At maximum overspeed, rod speed is (1.115 x 30): 33.h in./ minute. Each control rod is provided with a rod position indication monitor (Section 7.2.2.3.4) to sense asymmetric rod patterns by comparing the individual rod position with its group average position. hhen the rod moves out-of-step from its group by a preset amount, the monitor alarms the condition to the operator, computer and 'the ICS. Depending on the power setting and the control mode, action is initiated by the ECS, if an in-limit is also pre-sent, to insert rods and reduce power. 47 T.2.2.3.3 Startup Considerations The rod drive controls receive interlock signals from the ICS and nuclear in-strumentation (NI). These inputs are used to inhibit automatic mode selection if a large error exists in the ICS reactor controls and to inhibit out motion for high startup rates, respectively. In addition to the startup considerations , dilution controls , te permit removal of reactor shutdown concentrations of boron in the reactor coolant, are pro-vided. This control bypasses the normal reactor coolant dilution controls , described in Section 7.2.2.2, providing all safety rods are withdrawn from the core and the operator initiates a continuous feed and bleed cycle. An addi-tionalinterlock on rod Group 5 inhibits the use of this circuit when rod Group 5 is more than 80% vithdrawn. 7 2.2.3.h operational Considerations The control rod assembly positioning system provides the ability to move any rod to any position required consistent with reactor safety. As noted in Sec-tion T.2.2.3.2 a uniform speed is provided by the drive system. A fixed rod position when motion is not required is obtained by the power supply ability to energize two adjacent vindings of the CRA motor stator. This static ener-gizing of the vindings maintain a latched stator and fixed rod position. Position Indication As previously described, two separate position indication signals are provided. The absolute position sensing system produces signals proportional to CRA posi-tien from the reed switch matrix located en each CRA mechanism. The relative position indication system produces a signal proporticnal to the number of CRA T-22 Amendment No. 47 O March 1, 1975
e !;/] motor power- pulses from a stepping motor and precision potentiometer for -( f. .each CRA mechanism. Position indicating readout devices mounted on the operator's console consist of_69 single rod position meters and 4 control group average position meters. Accuracy of all meters is to 17, of full scale. The operation of a selector switch permits either relative or absolute position information to be displayed on the single rod meters. The control-group-average meters display the arithmetic average of the absolute l 22 position signals of all CRA's in a group. A selector switch on the operator's console permits the group meters to display either the positions of all safety rod groups (Groups 1 - h) or the positions of all regulating rod groups (Groups 5 - 7) and the . axial power shaping rod groups (Group 8) . Indicator lights are provided on the single-rod meter panel to indicate when each rod is; (1) fully inserted, (2)- fully withdrawn, (3) under control and (4) whether a fault is present. Indicators on the operator's console show full insertion, full withdrawal, under-control and fault indication for each of the eight control rod groups. The plant computer monitors the status of each trip device in the CRD system 22 and will alarm a trip condition. , Failures which could result in unplanned control rod withdrawal are contin-uously monitored by fault detection circuits. When failures are detected, in-dicator lights and alarms alert the operator. Fault indicator lights remain
\
on until the fault condition' is cleared by the operator. A list of indicated s - faults is shown below: (1) Asymmetric rod patterns (indicator and alarm) . (2) Motor rotation faults (indicator only) . * (3) Sequence faults (indicator and alarm). (4) Rea position sensor faults. 22 (5) Safety rods not withdrawn (indicator only) . (6) Programmer lamp faults'(indicator only). Faults serious- enough to warrant immediate action produce automatic correction ! commands from the fault detection circuits. Status indicators on the opera-tor's console provide monitoring offcontrol modes. A description of each fault detector follows: (1) Asymmetric Rod Monitor
- a. Design Basis - To detect and alarm if any rod deviates from its group reference position by more than a maximum of 13 inches true position. ]:2 i b .- Circuit Operation - There are 69 asymmetric rod pattern monitors, one assigned to each control rod. These monitors continuously com-pare the individual rod absolute position signal with the absolute group reference-(average) signal. The absolute value of the dif-ference between the two signals is computed, and if this difference is less than the maximum value set by the circuit calibration, no V
T-23 Amendment No. 22 Pecember ik, 1971
output results. If, hcVever, the difference is greater than the setpoint, a relay is actuated which r.lar=s the asyc=etric ccndition. Two alarm channels are provided in each monitor which are identical except for tee setpoints. One setpoint is calibrated for a T-inch signal differential (maximum 11-inch true positien separation) and 22 initiates an alarm. The other setpoint is a 9-inch signal differ-ential (maximum 13-inch true position separation) and initiates the action described below,
- c. Corrective Action - Action taken upon detection of an asy==etric red fault depends upon the control mode and the power level in effect at the time the fault is detected. Corrective action is the same fer any asy= metric condition including " stuck-in," stuck-out," or dropped control rods.
Detection of a 7-inch signal differential is defined as an "asy= metric l 22 rods alarm." Actuation of this alarm causes the fault indicator lamp for that rod to be energized and an alarm signal to be sent to the plant computer and annunciator. If the condition is not corrected and the separation increases to a 9-inch signal difference, the following actions occur: l 22
- 1. "Asy==etric fault" lamp on the operator's console is energized.
If operation is in the manual control mode, operator acticn is required by administrative control.
- 2. If operation is in the automatic mode and an in-limit is detected ,
also, a " runback fault" signal is sent to the Integrated Control System. 'the ICS will impose a maximum reactor power level of 609 of rated power if power is initially less than 60*+. When reactor power is greater than 60". of rated power, the Con-trol Rod Drive System generates an "Out Inhibit" signal which disables the "Out" co==and circuits to all drives and the ICS initiates a runback to 60% reactor power. "Out Inhibit" alar =s are sent to the ICS, plant ann'inciator and plant cc=puter. Reactor pcwer remains limited to 60% maximum in automatic con-trol until the fault is corrected. (2) Motor Rotation Fault Detector
- a. Design Basis - To detect and prevent unwarranted "aut" motion of con-trol rods caused by a failure such that "out" motion results from an "in" co= mand.
- b. Circuit Operation - Each of the five progra=mers is equipped with a rotation fault detector. These circuits consist of a rotation direc-tion sensor and a command versus rotation comparntor. The rotatien sensor is mechanically coupled to the progra==er and produces an output signal which reflects the direction of rotation of the pro-grs==er. The ec=parator compares the rotation with the actual ecm-mand. If actual rotation is in the "out" direction but co==and is 7-2h Amendment No. 47 flarch 1, 1975
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Amendment h7 l f -1 larch 1, 1975 t , - - - - - - - - -=m I APXA'3SAS PC'ER & LIGHT W. CONTROL ROOM FICL'PE NC. I APXANSAS NUCIEAR ONE - GIT 1 LAYOITT 7 25 ) gy l
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V) 8 ELECTRICAL SYSTE!S 8.1 IESIGN BASES The electrical system for Arkansas Nuclear One - Unit 1 is designed to be electrically self-sufficient and provide adequately reliable power sources for all electrical equipment for startup, normal operation, safe shutdown 1 and handling of all emergency situations. The following criteria have been used in the system and equipment design:
- a. All components of the system are sized for operation under normal and emergency conditions.
- b. No single component failure vill prevent operation of the ,
required engineered safeguards. )
- c. Redundant sources of power and/or automatic transfer of loads are provided to ensure continuous operation of equipment as required under emergency conditions.
- d. As far as practical, the system is arranged in such a manner as to make it possible to test equipment with the plant in operation.
- e. The relevsnt ANSI, NEMA, IEEE and the National Electrical Code recommendations are used as a guide in the design. )
- f. Class I electrical equipment is seismic qualified in accordance with the IEEE " Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations" JCNPS/SC2.5 (to be designated IEEE 3hk) as published in draft form for trial use and comment on August 11, 1971. 22-0.1
- g. Electrical and physical separation of cables and equipment associated with redundant elements of the engineered safe-guards is provided. All safety related portions of the standby electrical power system conform to safety guide 6 and 9, and IEEE 308 47
- h. The electrical system of Unit 1 is independent of Unit 2, except Startup Transformer No. 2 which is common to both units.
22-8.1 To prevent the concurrent loss of all auxiliary power the various sources of power are independent of and isolated from each other. The power supply and control of equipment providing engineered safeguards is arranged to 22 minimize the possibility of a loss of their operating functions cue to p physical damage. ( x 8-1 Amendment No. 47 March 1, 1975
8.2 ELECTRICAL SYSTEM DESIGN 8.2.1 NETWORK INTERCONNECTIONS Utlit 1 generates electric power at 22 kV which is fed through an isolated phase bus to the main transfor=er bank, consisting of three single-phase transformers, where it is stepped up to 500 kV transmission voltage and delivered to the station switchyard (See Figure 8-1). The 500 kV substation design is a ring bus scheme. The 500 kV station switchyard includes one line to the Mate 1 vale 500 kV substation and one line to the Ft. Smith 500 kV substation. The 161 kV switchyard at the generating station is also a ring bus design and includes one line to the Russellville East 161 kV substation and one line to the Morrilton East 161 kV substation. A bus tie autotransformer bank consisting of three single phase autotransformers interconnects the 500 kV and 161 kV syste=s in the station switchyard. The 22 kV tertiary of the autotransformer bank supplies Startup Transformer No.1 which is identical to the Unit Auxiliary 0 l l l l 8-la Amendment No. 22 December 14, 1971 l
l il least a seven day total diesel oil inventory will be available on-site in the emergency storage tanks for diesel generator operation during. complete loss of electric power conditions. thder -imum flood conditions (no nuclear accident), the plant will be shut down and the diesel generator loaded to about 50 percent of the rating of a single unit. Then the emergency storage tank inventory will last for about seven days operai. ion even if one emergency storage tank is considered unavailable. After the ma::icum hypothetical earthquake and a simultaneous nuclear accident, three and one-half days emergency supply of diesel oil 23-8.3 vill be available even in the unlikely event one emergency storage tank has failed. Within this period, additional fuel could be delivered to the plant site by any one of three methods; truck delive:y, rail car delivery or delivery by barge from the river. In the highly unlikely event that all three of these normal supply routes are unavailable because of the earthquake, fuel could be airlifted to the plant site via helicopter. It is expected that the maximum probable flood would be above plant grade (elevation 353) for sobut two to five days. 8.2.2.h h80 Volt Auxiliary System l l The kB0 volt system contains seven load centers, each consisting of a l h160/h80 volt transformer and a k80 volt switchgear bus. Six of the load centers are paired into double ended configurations with a tie breaker between buses. Engineered Safeguard buses have two tie breakers. N Redundant loads are fed from the same load center pair, one per bus. The two main load center pairs (h buses) serve non-essential station loads. One Engineered Safeguard load center pair (2 buses) serves Engineered Safeguard or essential loads and motor control centers. The seventh, i unpaired load center serves non-essential station loads. The capacities of , load center transformers and tie breakers in paired load centers are sufficient to permit operation with one of the transformers out of service. A spare load center transformer, interchangeable with any of the seven load center transformers in service, is provided. The h80 volt system also contains 27 motor control centers,19 of which serve non-essentini station loads, and 6 of which serve Engineered & Safeguard or essential loads and 2 are designated preferred emergency motor control centers and can be manually transferred to feed from either Engineered Safeguard subsystem. These supply such i loads 'as emergency lighting, the standby battery charger and the turbine generator emergency bearing and pump. 8.2.2.5 125 Volt DC System The de system is designed to provide continuous power for control, instrumentation, reactor protection and engineered safeguard systems, safeguard actuation control systems and other loads for normal operation and orderly shutdown. d
) Two independent and physically separated 125 volt batteries and de control centers are provided for the vital instrumentation, distribution panels, i emergency lighting r.nd motors. Three battery chargers are supplied with I
86 Amendment No. 47 March 1, 1975 7
two serving as normal supplies to the de control centers with the associated battery floating on the bus. The third battery charger serves as a standby battery charger to either de control center. The nor=al supply battery chargers are supplied from separate h80 volt engineered safeguard motor control centers. The standby battery charger is supplied from the 480 volt preferred emergency motor control center. The batteries supply the load without interruption should the battery chargers fail. Two distribution panels are provided suppling the de instrumentation and control power for the unit. Each de control center is provided with a manual transfer switch to supply power to the distribution panel. The transfer switch allows the operator to select the power from either de control center to each distribution panel. Each battery is sized to carry the continuous emergency de and vital ac load for a minimum period of two (2) hours in addition to supplying power for the operation of momentary loads during the two hour period. 9 In normal operation the batteries are floates on the buses, and assu=e load without interruption on loss of a battery charger. The ungrounded O l l l Amendment No. 21 O 8 6a January 21, 1972
92 CHEMICAL ADDITION AND SAMPLING SYSTEMS ( i 9 2.1 DESIGN BASES \v/ Chemical addition and sampling operations are required to change and monitor the concentratien of various chemicals in the reactor coolant system and aux-iliary systems. The chemical addition system is designed to add boric acid to the reactor coolant system for reactivity control, lithium hydroxide for pH control, and hydrazine for oxygen control. The system also provides boric acid for other plant corponents, and is sized to be able to add sufficient boric acid to maintain the core 15 Ak/k suberitical at any time during life. The sampling system is designed to sample reactor coolant, steam generator effluent, and liquids and gases from the auxiliary systems. 9 2.2 SYSTEM DESCRIPTION AND EVALUATION 1Me chemical addition cnd sampling systems are shown schematically on Figure 9-h. The sampling system has separate sampling stations for reactor coolant and steam generator sampling. The chemical addition system permits chemical addition to the reactor coolant system and reactor auxiliary systems during normal reactor operation. The chemical addition and sampling system is required to function during 4 ,' emergency conditions as identified in the technical specifications. The steam generator feedwater quality is maintained within the limits given in Table 9-3 and the reactor coolant quality is maintained by this system with-in the limits given in Table 9-4. A brief functional description of the major system components follows. Ox Boric Acid Addition Tank
\~/'b The boric acid addition tank is provided as a source of concentrated boric acid solution. The volume of the tank provides sufficient beric acid solution to increase the reactor coolant system boron concentration to that required for cold shutdown with no Xenen. Tank trace heating and electrically heat traced transfer lines maintain the fluid te=perature above that required to assure solubility of the boric acid.
~
Bcric Acid Mix Tank The beric acid mix tank is used to mix and supply the boric acid for the boric acid addition tank. The tank is equipped with an agitator for =ixing, and electrical heaters to maintain the temperature for boric acid solubility. Boric Acid Pumes Two boric acid pu=ps are provided to transfer the concentrated boric acid solu-tion from the boric acid addition tank to the berated water storage tank, make-im +e.k, or the spent fuel storage pool (Figure 9-h). [^\ l 3 \_,,/ Amendment No. 47 9_7 March 1, 1975
Lithium Hydroxide Mix Tank Lithium hydroxide is mixed 1n this tank and added to the makeup and purifica-tion system for pH control of the reactor coolant. The tank is equipped with an agitator for mixing. Lithium Hydroxide Pumn The lithium hydroxide pump transfers lithium hydroxide from the lithium hydrox-ide mix tank to the letdown line upstream of the makeup filters. If this pump is unavailable, the hydrazine pump can be used to transfer lithium hydroxide. Hydrazine Drum A drum supplies hydrazine to the reactor coolant system; the hydrazine is used to scavenge dissolved oxygen. Hydrazine Pump The hydrazine pump transfers hydrazine from the hydrazine dru= to the letdown line upstream of the make-up filters. If this pump is unavailable, the lithium hydroxide pump can be used to transfer hydrazine. Pressurizer Sample Cooler This cooler ecols the samples taken frcs the pressurizer steam or water space. Steam Generator Samole Cooler This cooler cools the sample taken from the secondary side of either steam generator. 9 2.2.1 Mode of oneration The chemical addition system deliver, the necessary chemicals to other systems as required. Boric acid is supplied to the spent fuel pool, borated water storage tank, and makeup tank as makeup for leakage or to change the concen- 6 tration of boric acid in the associated systems. The sampling system is used to take samples to assure that desired water qualities, gas concentrations, and boric acid concentration are maintained. Sampling locations and types of sam-ples taken at each location are shown in Table 9-5 9.2.2.2 Reliability Considerations The chemical addition and sampling systems are required to function during emergency conditions as identified in the technical specifica-tions. Redundant boric acid pumps and independent boric acid addition lines are provided to guard against a single component failure render- 42 ing the system inadequate for boron addition. Boric acid is avail-able for boration from the boric acid addition tank. To prevent precipitation, heat tracing is installed on components and lines used to transfer concentrated boric acid. The pumps, tanks, coolers, and instrumentation are located in the auxiliary building and are accessible for inspection and maintenance. 9-8 Amendment No. 42 November 30, 1973
l' - All electrical gear except some limit switches are located above water (V) for greater integrity and ease of maintenance. The hydraulic system that actuates the rotating fuel basket uses storage pocl water for operation to eliminate contamination, as do the hydraulic fuel and control rod grapples on the fuel handling bridges. The fuel transfer canal and storage pool water will normally have a baron concentration of 2,270 ppm, since the borated water storage tank has a minimum boron concentration of 2270 ppm. However, the minimum boron con-
' centration required in the fuel transfer canal and storage pool water is 1800 ppm. Although this concentration is sufficient to maintain core shut- 37 down if all of the control rod assemblies were removed from the core, only a few control rods vill be removed at any one time during the fuel shuffling and replacement. Although not required for safe storage of spent fuel assen-blies, the spent fuel storage pool vater will also be borated so that the transfer canal water vill not be diluted during fuel transfer operations.
Each fuel handling bridge mast travel is designed to limit the maximum lift of a fuel assembly to a safe shielding depth. ,21 Relief valves are provided on each stud tensioner to prevent overtensioning of the studs due to excessive hydraulic pressure. Gross failures of fuel are prevented by safety margins in the design and control of the core.
'N The fuel assembly utilizes a free-standing Zircaloy fuel rod of sufficient
) length to acccmmodate the expected fission gas release from the fuel.
l Any leaking fuel assemblies will be removed from the core for verificaticn of leakage and placed in a failed fuel container. This operation is done in the fuel transfer canal. The failed fuel is then transferred in the sealed containers to the spent fuel storage pool. Offsite shipment, fol-loving a suitable decay period, vill require that fuel be transferred to a liner ccmpatible with the shipping cask design. 9 6.2.h Handling Layout of the fuel handling area in the Auxiliary Building is such that the spent fuel casks vill never be required to traverse the spent fuel pool during removal of spent fuel elements. Diverse electrical interlocks (a limit switch and a power disconnect from the main contact rails) are pro-vided on the fuel handling crane to prevent an inadvertent traverse of the pool with a cask. The electrical interlocks, crane, and crane rails vill be designed to stop
'the fuel handling crane from full speed, assuming that one electrical inter-lock fails and that no operator action is taken. An analysis was performed k0 to determine whether this would cause the spent fuel cask to sving over the spent fuel pool. It was conservatively assumed that all of the kinetic energy of the cask vas converted to potential energy in the pendulum-like swing of the cask. The resultant motion of the cask in the horizontal plane
[ s). was calculated to be 2.15 ft. Allowing for stopping distance of the crane 3x_ / 9-27 Amendment No. LO August 6, 1973
i and for margin in the relatioLship of the electrical interlocks , thn hori-zontal cask motion from the center of the cask loading area would be less than h ft. It is, therefore, considered incredible that the spent fuel cask would ever fall iato the spent fuel pool. This conclusion is based on the LO fact that it would take four independent concurrent failures to cause the cask to fall into the spent fuel pool (the four failures are as follows: 1 & 2) both, diverse, electrical interlocks must fail; 3) a mechanical device on the crane must fail, e.g. the crane hock; h) the safety slings must fail). The spent fuel cask arrives by rail and is lifted from the rail car using ,
'I the fuel handling crane until the 9" interlock stops the cask motion.
47 The cask is then moved to the cask washdevn area and cleaned prior to lowering into the cask loading pit. The cask is moved to the loading pit, the safety slings are removed, and the cask is lowered into the leading 21 pit. (Motion along the edge of the spent fuel pool is prevented by key atterlocks.) After the spent fuel has been loaded into the cask, the cask is lifted frcm the loading pit until the 9" interlock stops the cask motion the safety slings are installed. If the interlock failed to stop the cask motion (either at this point or when the cask came up from the rail car), oper-ating procedures vould require that the raising of the cask be innediately terminated and the cask lowered again until the interlock could be cali-brated or repaired. The failure of the interlock to stop cask motion at 9" or less above the relay panel room ceiling would be deternined by co= paring the cask height to a pre-determined 9" benchmark. In moving the spent fuel cask frem the cask washdown area to the railroad spur opening, the loaded fuel handling crane vill pass over the relay panel room ceiling which is designed to absorb the energy of a cask drop. This determination was cade utilizing the energy absorbing capacity concept. This method does not lend itself to precise determination of design margin, rather it allows for a determination of design adequacy. The structural strength of the relay panel room ceiling is the second level of backup protection for failure of the crane (the safety slings are the l22-5. 36 first). To protect the areas under the path of the fuel handling crane from the consequences of a dropped cask, the following provisions vill be included: 21 (a) The crane vill be provided with an interlock (independent of the stops preventing traverse of thb crane over the pool) set to limit lifting the cask to a maximum of 9 inches above floor level. To ensure that this height vill not be exceeded, especially over the relay panel room, the interlock provided on the crane vill be both inspected and 9-27a Amendment No. 47 March 1, 1975
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se Arendment No. 47 ARKANSAS PO* DER & LIGIIT FI GUFJ' N O* AIt:NiSAS HUCLEAR ONE-1EiIT 1 FIRE PIOTE.~."20N SYSTEM 9-16 e
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(g) 10 STEAM AND POWER CONVERSION SYSTEM 28-(,/ 10.3 10.1 SUbNARY DESCRIPTION The steam and power conversion system is designed to remove heat energy from the reactor coolant in two steam generators and convert it to electrical energy via the turbine generator. This system is shown schematically in Figure 10-1. The condenser transfers the heat which is unusable in the cycle to the condenser cooling water and deaerates the condensate. The closed regenerative turbine cycle heats the condensate and returns it to the steam generators. The entire system is designed for the maximum expected energy from the nuclear steam supply system. l Upon loss of full load, the system will dissipate all the energy existent or produced in the reactor coolant system through bypass valves to the 1 41 condenser, the power operated atmospheric dump valves, a.C the main steam I safety valves. The unit is designed to maintain station auxiliary load without a reactor trip upon loss of full load due to faults occuring on the system side of the S00 kV switchyard breakers. The steam bypass to the condenser and atmospheric dump valves will be utilized as necessary to ~3, 3 achieve this load reduction. 10.3 The steam and power conversion system provides steam for driving two steam generator feedwater pumps, condensate water inventory heating and building heating as required. These are completely closed systems and have no direct atmospheric interfaces. Steam used for the turbine gland sealing f-x system does not leak outward. Leakage, if any, will be air leakage ( ') inwards. Steam used for the emergency feedwater pump turbine is exhausted L./ to atmosphere. This exhausted steam could be radioactive only if both failed fuel and a steam generator tube leak exist. If, by the highly unlikely combination of events, radioactivity leakage to the atmosphere should occur the turbine driven pump would be secured and the backup electrical drivea emergency feedwater pump would be available. Under normal operating conditions, there are no radioactive contaminants present in the steam and power conversion system. It is possible for this system to become contaminated only through steam generator tube leaks, simultaneous with failed fuel in the reactor. In this event, monitoring of the steam generator shell side sample points and the condenser vacuum pumps suction air will detect any contamination. 10.1.1 TRIPS, AUTOMATIC CONTROL ACTIONS AND ALARMS Trips, automatic control actions and alarms will be initiated by deviations of system variables within the steam and power conversion system. In the case of automatic corrective action in the steam and power conversion system, appropriate corrective action will be taken to protect the reactor coolant system. The more significant malfunctions or faults which cause trips, automatic actions or alarms in the steam and power conversion system are:
- a. Turbine Trips
] \ ( ,/ 1. Generator faults. 10-1 Amendment No. 41.
- 2. Electrical faults.
- 3. Loss of condenser vacuum.
- 4. 'Ihrust bearing wear.
- 5. liigh vibration.
- 6. Exhaust hood temperature above 250 F. 28-10.3
- 7. Low bearing oil pressure.
- 8. Turbine overspeed.
- 9. Turbine protection for generator motoring.
- 10. Reactor trip.
- 11. !!anual { Emergency) Trip.
- 12. Rotor long alarm.
- 13. Rotor short alarm. 47
- 14. Trip of both main feedwater pump turbines.
- b. Automatic Control Actions
- 1. Feedwater flow lagging feedwater demand results in a reduction in power demand.
- 2. Low feedwater temperature results in a reduction in power demand.
- 3. liigh level in steam generator results in a reduction in fr.edwater flow.
l l 4. Low level in steam generator results in an increase in l ' feedwater flow. 28-10.3
- 5. Loss of one main feedwater pump results in plant runback to 55 percent full power.
l
- 6. Loss of two of the three condensate pumps results in plant runback to 55 percent full power.
- 7. Loss of two main feedwater pumps and fault in the auxiliary feedwater pump when in operation will start the turbine driven emergency feedwater pump.
- c. Principal Annunciator Alarms
- 1. Low pressure at each feedwater pump suction.
- 2. Low vacuum in condenser.
10-2 Amendment No. 47 March 1, 1975
% Piping and equipment in contact with reactor coolant vastes is constructed of corrosion-resistant materials. A dru= ming station area radiation monitor and radvaste liner cask surface radiation monitor with associated alanns are provided to detect possible leakage and ensure that radiation levels about the radvaste liner do not exceed Department of Transporation limits. The drumming station is a Seismic Class I structure. Floor drainage is piped to the dirty vaste drain tank. All vaste tanks are vented to the gas collection header and provide adequate venting for filling or draining operations. All necessary instrumentation and controls for operation of the solid 37 radvaste system are located on a local control panel in the dr'"-ing station near the process module. In addition, level indication and alarms are provided at the liquid radvaste control panel for the spent resin tank. A spent resin tank gas sample connection is also provided. In addition to the solid radvaste system drumming station, there is a lov level drumming area for low activity or potentially contaminated vastes. Soft solid vastes such as anti-contamination clothing, rags, paper, gloves, shoe coverings, and glove bags vill be compressed into 55 gallon drums by a hydraulic compgetor. Filled containers vill be monitored for surface radiation and stored in the lov level drumming area prior to shipment to an off-site disposal facility. O l47 V . l l 1 l l I 11-7; Amendment No. 47 March 1, 1975
11.1 3.4 Process Radiation Monitoring System 11.1 3.h.1 Description The process radiation monitoring system is designed to assure that ioniz.ing radiation levels are recorded, indicated and alarmed so that actfon, cither automatic or manual, can be taken to prevent radioactive reJesse from exceeding the Hmits of 10 Ci? 20. Davices are located in the various process syste=s to moniter radiaticn levels and annunciate any abnor ally high activity. Instrument ranges and sensitivities are chosen to enable monitoring within the requirements of 10 L7R 20. All electronic circuitry (except for photcmultipliers and geiger tubes) is colid state and each circuit has its own D.C. power supply and provisions for instrument operational testing while in service. All ecnitors are supplied with check sources. The check source simulates a radioactive sample and serves as a check for both the readout and dstection equipment. Radiation monitcring equipment panels are located inside the main control roon. These panels provide mounting for indicators, recorders, power cupplies and alarms for the radiation monitoring systems. The radiation monitoring panels are fed by the instrument a-c bus which, in the event of a loss of normal power, is fed by the diesel generators bus. The type of detectors used and the information displayed on these panels are listed in Table 11-7 The sensitivity and alarm conditions for each instrument are also listed. 11.1.3.L 2 System Evaluation A31 process systems which contribute to plant discharges are monitored prior to entering the various discharge systems. Each discharge system is clso monitored providing redundancy of radiation detection for plant effluents. The Reactor Building air, vaste gas, and the main condenser air discharge radiation monitoring systems are backed up by the stack gas conitoring system. The service water, radwaste liquids discharge, intermediate cooling, and decay heat radiation monitoring systems are backed up by the liquid discharge monitor system. Testing and maintenance for all systems and circuit testing of readout cquipment and power supplies can be performed from the panels located in tha control room. The circuit being tested or repaired indicates and alanus this condition in the control room. Faulty or inoperative circuits during operation are also alarmed and indicated in this manner. 11-8
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One . additional loss-of-coolant flow mechanism has been analyzed in the re-
,. actor design evaluation. This involves possible flow or leakage past the seat l ) of a reactor internals vent valve. These valves are designed to be closed dur- \m,/ ing all normal operations and during all nor=al and accident transients except for those accident transients involving reverse flow through che core. The design provides for positive closure even with no flow, and several rotational clearances are provided to ensure free motion and to prevent any tendency to stick. The valves also have a self-align =ent feature to prevent reactor cool-ant leakage. Hydrostatic and vibrational tests have been made to demonstrate that the valves vill operate as designed. However, the case of a reactor in-ternals vent valve remaining open has been analyzed. This malfunction reduces the effective core flow and results in a reduced DNBR at the steady-state de-sign overpower. The steady-state design overpower DNBR for these conditions is listed in Table 14-11. The minimum transient DNBR for these conditions is listed in Table IL-13.
14.1.2.7 Stuck-Out. Stuck-In, or Dropped Control Rod Accident 1h.1.2.7.1 Identification of Cause In the event that a control rod cannot be moved, localized pcTer peaking and shutdown margin must be considered. If a control rod is dropped into the core while operating, the resulting transient must be examined. Adequate hot shutdown margin is assured by requiring a subcriticality of 1 percent ak/k with the control rod of greatest worth fully withdrawn from the core. The nuclear analysis reported in Section 3.2.2 demonstrates that this criterion can be sati fied. This criterion has been analyzed in ter=s of the
,7-"'s minimum tripped red worth available in the loss-of-coolant-flow, rod ejection, i
V) startup, rod withdrawal, and steam-line-failure accidents. In all cases the available rod worth is sufficient to provide =argins below any damage thresh-old. If . control rod deviates from its group reference position by more than an indicated 9 inches and an in-limit is detected, the Control Rod Drive System l 47 and Integrated Control System will inhibit all rod-out motion, and the steam generator load demand is run back to 60% of rated load. The details of these actions, which occur for both a dropped or stuck rod, are described in Sections 7.2.2 and 7.2.3. 1h.1.2.7.2 Reactor Protection Criteria The criteria for reactor protection for this accident are: a. The minimum DNB ratio shall not be less than 1.3
- b. The reactor coolant system pressure shall not exceed code pressure limits.
1k.1.2.7.3 Methods.of Analysis A detailed B&W digital model has been used to analyze the transient response to a dropped control rod. This program includes fuel pin, point kinetics, pressurizer, and loop =odels, including the steam generators. (h i ; Q.) 14-11 Amendment No. 47 March 1,1975
The reactor is assumed to be operating at rated power when the control rod is dropped. To achieve the most adverse response, the most negative values of the moderator and Doppler coefficients were used along with the maximum calculated rod worth for rated power operation. In addition, no Control Rod Drive System or Integrated Control System action was assumed to occur. The parameters used in this analysis are shown in Table ik 15. Ik.l.2.7.4 Results of Analysis The results are presented in Figure lh-20. The neutron power decreases causing a rapid decrease in both the core moderator temperature and the fuel temperature. These temperature decreases overcompensate for the worth of the control rod, and the neutron power rises slightly above the initial neutron power level. The neutron power then decreases to below the initial power level and eventually levels out at the initial power level. The thermal power response is similar to the neutron power; however, the thermal power level never exceeds the initial va'.ue. Both the core moderator temperature and pressurizer pressure decr :ase during the trar.sient and level out at a value lower than the int +1a'. value. Since the thermal power never exceeds the inicial value and the pressure decreases during the transient the protection criteria are met. Cases have been run for rod drops at beginning-of-life conditions and lower rod worths. These transients are not included in this discussion because they represent less severe conditions than the end-of-life conditions and the maximum calculated rod worth. 1h.1.2.8 Loss of Electric Power lb.1.2.8.1 Identification of Caase The unit is designed to withstand the effects of loss of electric load or electric power. Emergency power systems are described in Section 8. Two types of power losses are considered:
- a. A loss of load condition caused by separation of the unit from the transmission system.
- b. A hypothetical condition resulting in a complete loss of all system and unit power except the unit batteries.
1h.1.2.8.2 Reactor Protection Criteria The criteria for reactor portection for this accident are:
- a. Fuel damage must not occur.
- b. Reactor coolant system pressure shall not exceed code pressure
! limits,
- c. (1) Resultant doses for loss ## all AC power shall not exceed l 10 CFR 100 limits.
2k-lk.h (2) Resultant doses for loss of load shall not exceed 10 CFR I 20 limits. lk-12 Amendment No. Eh February 29, 1972 l [
r w Safety Review Comittee Six (6) Members - Little Rock 27 Two (2) Members - Site (Plant Staff) Plant Safety Comittee Five (5) Members - Site (Plant Staff) ' l47 The' Chief QA Coordinator has the responsibility for devising the QA , Auditing Program for operation of ANO Unit 1. QA procedures and instructions are drafted under the supervisica of the CQAC and sub-mitted to the QAC for review. Final draf ts of QA procedures must be approved by the QAC before they are put into use. Approved procedures 4 are issued to holders of the AP&L QA Manual. A manual has been issued to each department of the Company affected by the QA Program. 4 i l I l i I 1.10a Amendment No. 47 1 ! March 1, 1975 i l I l i I l~ \
p RESPONSES TO ?phIC ITEMS OF DlTEREST ITEM 8.12' We have evaluated your responses to Requests 8.1 and 8.2. Your responses take expection to IEEE-308 for the purpose of gaining some small additional power during your system peak demand periods. We have concluded that the independence of redundant standby power supplies (diesel generator) must be maintained during all phases of plant operation; therefore, the use of both diesel generators for peaking purposes must be avoided. The use of one diesel generator for peaking could, under certain restricted conditions, provide the independence required in IEEE-308 and GDC 17 Your responses to these requests should be amended to reflect these comments.
RESPONSE
1he diesel generators will not be used for peaking. 47 O C V) 0 12 Amendment No. 47 March 1, 1975
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ARKANSAS NUCLEAR ONE-UNIT I GENERATORS FUEL OIL SYSTEM 8.13 -1 >
f'N RESPONSES TO SPECIFIC ITEMS OF INTEREST v l-k ITEM 9.69 For all tanks that contain gas under pressure (such as nitrogen. hydrogen, oxygen, air and CO2 tanks) provide the following: (1) the design and operating pressure, (2) the maximum pressure of the gas supply, (3) the location of the tank, (h) the maximum total energy if the tank should rupture, (5) the possibility of the tank or parts of the tank to act as a missile, (6) the protective measure taken to prevent a tank failure, and (7) the protective measure taken to prevent the loss of function of adjacent equipment essen-tial- for a safe shutdown condition.
RESPONSE
Table 9.69-1 lists all tanks (excluding instrument air accumulators) centain-
-ing gas under pressure and their associated maximum operating pressure, design pressure, hydrostatic test pressure, and location.
It is conceivable that parts of a tank, such as a nipple or a nozzle, could receive an accidental knock and fail. However, the probability of such an 28 cccurrence is extremely low and protection against this occurance is not provided.
/'~'N Portable gas storage bottles are hydro-tested once every five years to
(' 1 times design pressure by the supplier. Design pressure is always higher than maximum operating pressure. 47 Permanently installed tanks have relief valves set for the appropriate design pressure. Bottle storages are selected to minimize the possibility of damage to vital ; equipment due to gas generated tissiles. Except for the core flooding tanks 28 'l which are pressurized by nitrogen, there are no essential systems which depend on gas pressure for operation during an emergency. The core flooding tanks are separated from each other by missile proof valls. l l i [ T N] 9.69 Amendment No. 47
!! arch 1,1975
f m. EFFECTIVE LIST OF PAGES FOR ARKANSAS NUCLEAR ONE-UNIT I FINAL SAFETY ANALYSIS REPORT , (Through Amendment No. 46) Page Amend. Page Amend. Page Amend. No. No. No. No. No. No. VOLUME I i 21 1-15 Orig. 1-53 36 11 Orig. 1-16 Orig. 1-54 43 iii Orig. 1-17 Orig. 1-55 38 iv 21 1-18 Orig. 1-56 43 v Orig. 1-19 Orig. 1-57 43 vi Orig. 1-20 Orig. 1-58 36 i vii 22 1-21 Orig. 1-59 36 viii Orig. 1-22 Orig. 1-60 36 ix Orig. 1-23 Orig. 1-61 36 x Orig. , 1-24 Orig. 1-62 25 xi 27 1-25 Orig. 1-62a 25 4 G.i 1-1 CHAPTER I 21 1-26 1-27 Orig. Orig. 1-63 1-64 Orig. 36 1-11 Orig. 1-28 Orig. 1-65 36 4 1-111 Orig. 1-29 - Orig.
- 1-66 Orig. l 1-iv Orig. 1-30 Orig. 1-67 36 1-v Orig. 1-31 Orig. 1-68 Orig.
1-vi Orig. 1-32 Orig. 1-69 Orig. 1-vii Orig. 1-33 Orig. 1-70 25 1-viii 31 1-34 Orig. 1-70a 43 1-villa 31 1-35 37 1-70b 44 1-ix 21 1-36 37 1-70c 43 1-x 21 1-37 37 1-70d 25 1-1 21 1-38 37 1-70e 27 ! 1-la 21 1-39 Orig. 1-70f 27 l 1-2 Orig. 1-40 43 1-70g 43 l 1-3 22 1-41 43 1-70h 42 I 1-4 Orig. 1-42 42 1-701 27 1-5 Orig. 1-43 21 1-70j 43 1-6 Orig. 1-44 43 1-70k 43 1-7 Orig. 1-45 43 1-701 43 1-8 . Orig. 1-46 27 1-70m 43 1-9 Orig. 1-47 Orig. 1-70n. 43 1-10 Orig. 1-48 43 1-71 21 1-11 Orig. 1-49 Orig. 1-71a 21
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1-71e 37 2-v Orig. Fig. 1 7 Orig. 1.-72 Orig. 2-vi 21 Fig. 2-8 Orig. 1-73 Orig. 2-1 26 Fig. 2-9 Orig. 1-74 Orig. 2-2 . Orig. Fig. 2-10 Orig. 4 1-75 Orig. 2-3 Orig. Fig. 2-11 Orig. , 1-76 Orig. 2-4 Orig. Fig. 2-12 Orig. 1-77 Orig. 2-5 Orig. Fig. 2-13 Orig. 1-78 Orig. 2-6 Orig. Fig. 2-14 Orig. 1-79 30 2-7 37 Fig. 2-15 Orig. Fig. 2-16 Orig. 1-80 30 2-7a 37 1-81 30 2-8 Orig. Fig. 2-17 Orig. 1-82 30 2-9 21 Fig. 2-18 Orig.
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.9-22 Orig.' CHAPTER 10 11-21 Orig.
9-23 30 10-1 28 11-22 Orig. 9-24 Orig. 10-11 28 11-23 Orig. 9-25 Orig. 10-1 41 11-24 Orig. 9-26 Orig. 10-2 28 11-25 Orig. 9-27 40 10-3 28 11-26 Orig. 9-27a 40 10-4 28 11-27 22 i 9-27b 22 10-5 28 11-27a 22 9-28 28 10-6 28 11-28 28 9-29 28 10-7 46 11-29 37 9-30 28 , 10-8 43 11-30 36 9-30a 46 10-9 46 11-31 45 9-30b 28 10-10 46 11-31a 37 9-31 31 10-11 46 11-31b 37 9-31a 24 Fig. 10-1 38 11-31c 37 9-31b 24 Fig. 10-2 46 11-32 Orig. 9-32 Orig. Fig. 10-3 46 11-33 Orig. 9-33 22 CHAPTER 11 11-34 31 9-34 Orig. 11-1 Orig. 11-35 22 9-35 31 11-11 Orig. 11-36 22 9-36 Orig. 11-111 22 11-37 Orig. 9-37 Orig. 11-iv Orig. 11-37a 22 9-38 Orig. 11-1 Orig. 11-38 Orig. ( 9-39 Orig. 11-2 22 11-39 Orig. 9-40 46 11-3 Orig. 11-40 Orig. 9-46a 46 11-4 Orig. 11-41 45 9-41 Orig. 11-5 Orig. 11-42 45 9-42 Orig. 11-6 Orig. Fig. 11-1 44 9-43 21 11-7 37 Fig. 11-2 44 9-44 22 11-7a 37 Fig. 11-3 27 Fig. 9-1 30 11-7b 43 Fig. 11-4 Orig. Fig. 9-2 Orig. 11-8 Orig. Fig. 11-5 41
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No. No. No. No. No. No. 12-9 28 13-36 38 14-6 27 12-10 31 13-37 26 14-7 22 12-11 .31 13-38 41 14-7a 22 12-12 31 13-39 41 14-8 Orig. 12-13 44 13-40 26 14-9 Orig. Fig. 12-1 44 13-41 26 14-10 Orig. Fig. 12-2 34 13-42 26 14-11 27 CHAPTER.13 13-43 46 14-12 24 13-1 26 13-44 46 14-13 Orig. 13-11 26 13-45 46 14-14 42 13-1 37 13-46 46 14-15 Orig. 13-la 36 13-47 46 14-16 Orig. 13-1b 36 13-48 46 14-17 Orig. 13-1c 26 13-49 46 14-18 Orig. 13-1d 20 13-50 26 14 -19 27 13-2 31 13-51 26 14-20 Orig. 13-3 31 13-52 37 14-21 Orig. 13-4 Orig. 13-53 46 14-22 Orig. 13-5 41 Fig. 13-1 44 14-23 Orig. 13-6 Orig. 14-24 Orig. 13-7 38 VOLUME III 14-25 38 13-8 Orig. i 21 14-26 Orig. 13-9 26 11 Orig. 14-27 Orig. 4 [. 4 d 13-10 13-11 26 26 iii Orig. 14-28 14-29 Orig. 22 iv 21 l 13-12 26 v Orig. 14-30 22 13-13 38 vi Orig. 14-31 Orig. 13-14 31 vii 22 14-32 Orig. 13-15 31 viii Orig. 14-33 Orig. . 13-16 31 ix Crig. 14-34 Orig. l 13-17 26 x Orig. 14-34a 31 j 13-18 26 xi 27 14-35 Orig. ' 13-19 26 CHAPTER 14 14-36 Orig. 13.-20 26 14-i 31 14-37 Orig. 13-21 26 14-ii 31 14-38 Orig. 13-22 26 14-i11 Orig. 14-39 31 , 13-23 44 14-iv Orig. 14-40 Orig. l 13-24 26 14-v Orig. 14-41 Orig. l 13-25 26 14-vi Orig. 14-42 Orig, i 13-26 26 14-vii Orig. 14-43 Orig. j 13-27 26 14-viii Orig. 14-44 Orig. ;
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No. No. Page No. Amend. No. Page No. Amend. No. Fig. 14-80 37- APPENDICES Fig. A-1 36 Fig. 14-81 37 APPENDIX A Fig. A-2 36 Fig. 14 37 A-i 38 Fig. A-3 36 Fig. 14-83 37 A-ii 38 Fig. A-4 38 Fig. 14-84 ~37 A-iii 38 Fig. A-5 38 Fig. 14-85 37 A-1 Orig. Fig. A-6 36 Fig. 14-86 37 A-2 Orig. Fig. A-7 38 Fig. 14-87 37 A-3 45 Fig. A-8 38 Fig. 14-88 37 A-4 45 Fig. A-9 38 Fig. 14-89 37 A-4a 45 Fig. A-10 38 Fig. 14-90 '37 A-5 Orig. Fig. A-11 38 Fig. 14-91 37 A-6 38 Fig. A-12 38 Fig. 14-92 37 A-7 38 Fig. A-13 38 Fig. 14-93 37 A-8 38 Fig. A-14 38 Fig. 14-94 37 A-9 38 Fig. A-15 38 Fig. 14-95_ 37 A-10 38 Fig. A-16 38 Fig. 14-96 37 A-11 38 Fig. A-17 38 Fig. 14-97 37 A-12 38 APPENDIX 1 Fig. 14-98 37 A-13 38 1-i 34 Fig. 14-99 37- A-14 38 1-1 Orig. Fig. 14-100 37 A-15 38 1-2 Orig. Fig. 14-101 37 A-16 38 1-3 Orig.
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A-18 38 1-5 34 CHAPTER 15 A-19 38 1-6 34 This chapter as A-20 38 1-7 . 34 e6 stained in A-21 38 1-8 34 the FSAR is no A-22 38 1-9 34 longer valid. A-23 38 1-10 34 i It is now . A-24 38 1-11 34 superseded by A-25 38 1-12 34 the Technical A-26 38 1-13 34 L Specifications A-27 -38 1-14 34 as issued with A-28 38 1-14a 34 License No. A-29 38 1-14b 43 DPR-51. A-30 38 1-14e 34 A-31 38 1-14d 34 A-32 38 1-14e 34 A-33 38 1-14f 34 A-34 38 1-15 24 A-35 38 1-16 24 A-36_ 38 Fig. 1-1 24 A-37 38 APPENDIX 2A A-38 38 2.A-i Orig. A-39 38 2.A-ii 29 A-40 38 2.A-iii Orig. A-41 38 2. A-iv 29 I A-4 2 38 2.A-v 29 A-43 38 2.A-vi 29 EP-10 J e emw.
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' 2. A-1 Orig. 2. A-45 Orig. 3A-12a 31 2.A-2 0-ig. 2.A-46 Orig. Fig. 3A-1 36 2.A-3 cig. 2. A-47 29 Fig. 3A-2 22 2.A-4 crig. 2. A-48 29 Fig. 3A-3 22 2.A-5 Orig. 2. A-49 29 Fig. 3A-4 22 2.A-6 Orig. 2.A-50 29 Fig. 3A-5 22 2.A-7 Orig. 2.A-51 29 Fig. 3A-6 22 2.A-8 Orig. 2.A-52 29 Fig. 3A-7 22 2.A-9 Orig. 2.A-53 29 Fig. 3A-8 22 2.A-10 Orig. 2.A-54 29 Fig. 3A-9 22 2.A-11 Orig. 2.A-55 29 Fig. 3A-10 22 2.A-12 Orig. 2.A-56 29 APPENDIX 4 2.A-13 Orig. 2.A-57 31 4-1 33 2.A-14 Orig. 2.A-58 31 4-2 33 2.A-15 Orig. Fig. 2A-1 Orig. 4-3 33
- 2. A-16 29 Fig. 2A-2 Orig. APPENDIX 5 2.A-16a 29 Fig. 2A-3 Orig, i Orig.
. 2.A-16b ' 31 Fig. 2A-4 Orig. 5.A-i 28 2.A-16c 31 Fig. 2A-5 Orig. 5.A-ii Orig. 2.A-16d 29 Fig. 2A-6 Orig. 5.A-iii Orig. 2.A-16e 29 Fig. 2A-7 Orig. 5.A-iv Orig. js 2.A-17 Orig. Fig. 2A-8 Orig. 5. A-v Orig.
; ) 2.A-18 Orig. Figs 2A-9 Orig. APPENDIX 5A
(_) -2.A-19 2.A-20 Orig. Orig. Fig. 2A-10 Fig. 2A-11 Orig. Orig. 5.A-1 5 A-2 Orig. Orig. 2.A-21 Orig. Fig. 2A-12 Orig. 5.A-3 24 2.A-22 Orig. Fig. 2A-13 Orig. 5.A-4 24
- 2. A-23 Orig. Fig. 2A-14 Orig. 5.A-5 28 2.A-24 Orig. Fig. 2A-15 29 5.A-6 28 2.A-25 Orig. Fig. 2A-16 29 5.A-7 28 2.A-26 Orig. Fig. 2A-17 29 5.A-8 28 2.A-27 Orig. APPENDIX 3 5.A-9 28 2.A-28 Orig. APPENDIX 3A 5.A-10 28 2.A-29 Orig. 3A-i 22 5.A-11 28 2.A-30 Orig. 3A-ii 31 Fig. SA-1 Orig.
2.A-31 Orig. 3A-lii 22 Fig. SA-2 Orig. 2.A-32 Orig. 3A-iv 22 Fig. 5A-3 28 2.A-33 Orig. 3A-1 36 Fig. SA-4 28 2.A-34 Orig. 3A-2 22 Fig. 5A-5 28 2.A-35 Orig. 3A-3 22 APPENDIX 5B
- 2. A-36 ' Orig. 3A-4 22 5.B-1 Orig.
2.A-37 Orig. 3A-5 22 5.B-2 Orig. 2.A-38 Orig. 3A-6 22 APPENDIX SC 2.A-39 Orig. 3A-7 31 S.C-1 Orig.
- 2. A-4 0 Orig. 3A-8 22 5.C-2 Orig.
- 2. A-41 Orig. 3A-9 31 S.C-3 Orig.
-2.A-42 Orig. 3A-10 31 S . C-4 Orig.
>'~'N 2. A-4 3 Orig. 3A-11 31 f) m 2.A-44 Orig. 3A-12 31
+
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g Page Amend. Page Amend. Page Amend. No. No. No. No. No. No. APPENDIX 5D APPENDIX SL 1.3 25 5.D-1 Orig. 5.L-1 37 1.3a 25 5.D-2 Orig. -5.L-2 37 1.4 22 APPENDIX SE 5.L-3 37 1.5 25 5.E-1 Orig. 5.L-4 37 1.7 22 APPENDIX SF 5.L-5 37 1.8 25 5.F-1 Orig. 5.L-6 37 1.8a 25 5.F-2 Orig. E.L-7 37 1.10 27
-5.F-3 Orig. 5.Le8 37 1.10a 27 5.F-4 Orig. 5.L-9 37 1.11 31 5.F-5 Orig. 5,L-10 37 1.12 27 5.F-6 Orig. 5.L-11 37 1.14 27 Fig. 5F-1 Orig. 5.L-12 37 1.15 27 Fig. 5F-2 Orig. 5.L-13 37 1.15a 42 APPENDIX SG 5.L-14 37 1.18 31 5.G-1 Orig. 5.L-15 37 1.19 31 5.G-2 Orig. 5.L-16 37 1.22 27
, 5.G-3 Orig. Fig. SL-1 37 '1.22a 27 5.G-4 Orig. Fig. SL 37 1.24 27 5.G-5 Orig. Fig. SL-3 37 1.27 31 5.G-6 Orig. Fig. SL-4 37 1.28 27 5.G-7 Orig. Fig. SL-5 37 2.5 42 h APPENDIX SH 2.6 25 Q 5.H-1 5.H-2 24 34 i VOLUME IV 21 2.6a 2.6b 42 25 Fig. 5H-1 34 ii Orig. Fig. 2.6-1 25 Fig. 5H-2 34 iii Orig. Fig. 2.6-2 25 APPENDIX 5J iv 21 Fig. 2.6-3 25 5.J-1 Orig. v Orig. 2.7 28 5.J-2 23 vi Orig. 2.7a 28 5.J-2a 24 vii 22 2.7b 28 Fig. 5J-1 Orig, viii Orig. 2.7c 28 Fig. 5J-2 Orig. ix Orig. 2.7d 28 Fig. 5J-3 Orig. x Orig. 3.2 23 Fig. 5J-4 28 xi 17 3.3 28 Fig. 5J-5 Orig. AEC QUESTIONS 3.4 28 Fig. 5J-6 Orig. i 26 3.5 31 Fig. 5J-7 Orig. ii 26 3.6 28 Fig. 5J-8 Orig, iii 26 3.6a 28 Fig. SJ-9 28 iv 26 3.6b 28
~ APPENDIX 5K v 26 3.6c 28 5.K-1 ' Orig. vi 27 3.6d 28 5.K-2 Orig. vii 28 3.6e 28 5.K-3 Orig. viii 28 3.6f 28 5.K-4 Orig. ix 28 3.6g 28 5.K-5 Orig. x 28 3.6h 28 5.K-6 24 xi 28 3.61 28 5.K-7 24 xii 28 4.1 25 (O-I Fig. SK-1 Fig. SK-2 Orig.
Orig. xiii 1.1 28 26 4.la 4.lb 25 25 EP-12
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Page Amend. Page Amend. No. No. No. No. No. No. 4.2 23 Fig. 4.28-8 28 5.62 24 4.3 23 Fig. 4.28-9 28 5.63 24 4.3a 23 Fig. 4.28-10 28 5.63a 24 4.7 36 4.29 28 5.65 26 4.7a 26 4.29a 28 5.67 28 4.7b 23 4.30 28 5.69 28 4.8 25 4.30a 28 5.69a 28 4.9 23 4,30b 28 5.71 28 4.11 25 4.30c 28 5.74 28 4.12 25 4.31 28 5.75 28 4.12a 25 4.31a 32 5.77 28 4.12b 25 4.32 35 5.78 28 4.(13-14) 23 4.32A 35 5.79 33 4.15 22 4.32B 35 5.80 31 4.15a 23 4.32C 35 5.80a 28 4.16 23 4.32D 35 5.81 32 Fig. 4.16-1 23 4.32E 35 5.82 31 Fig. 4.16-2 23 4.32F 35 5.83 30 4.17 23 4.32G 35 5.84 30 , 4.18 23 4.32H 35 5.88 30 4.19 23 4.33 28 6.1 25 I, 4.20 25 4.33a 41 6.3 ~ 26 4.20a 25 4.34 31 6.4 23 k ,) 4.20b 25 4.35 28 6.4a 23 4.22 -25 4.36 28 6.5 25 4.22a 25 4.37 29 6.6 24 4.22b 25 5.4 28 6.7 23 4.22c 25 5.4a 24 6.7a 23 , 4.22d 25 5.6 28 6.8 31 1 4.22e 23 5.13 25 6.9 28 4.23 26 5.13a 25 6.12 28 4.26 25 5.13b 25 6.13 27
'4.27 26 5.13c 25 6.13a 27 l 4.27a 33 5.13d 25 6.14 28 !
4.28 28 5.13e 25 6.15 27 l 4.28a' 28 5.13f 25 6.16 38 l 4.28b 28 5.13g 25 Fig. 6.16-1 38 l 4.28c- 28 5.13h 25 6.17 28 l 4.28d 28 5.16 24 6.18 28 4.28c 28 5.17 23 Fig. 6.18-1 28 4.28f 28 5.17a 23 6.19 32 4.28g 28 5.21 23 6.19a 32 Fig. 4.28-1 28 5.21a 23 Fig. 6.19,1-1 32 Fig. 4.28-2 28 5.22 22 Fig. 6.19.)-2 32 Fig. 4.28-3 28 5.25 24 Fig. 6.19.5-1 32 Fig. 4.28-4 28 5.25a 24 Fig. 6.19.5-2 32 i Fig. 4.28-5 28 5.40 24 Fig. 6.19.5-3 32 I- 5.54 25 7.1 23
'~')j Fig. 4.28-6 28 Fig. 4.28-7 28 5.61 24 7.la 23 1
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k - Page- Amend. Page Amend. Page Amend. No. No. No. No. No. No. 7.2 23 Tab. 9-9.2.3 22 Fig. 9.25-2 44 7.3 _25 9.4 25 Fig. 9.25-3 28 7.5 25 9.4a 25 Fig. 9.25-4 44 7.5a 25 9.5 26 9.26 27 7.8 25 9.Sa 26 9.26a 27
. 7.8a 25 9.5b 26 Fig. 9.26-1 27 7.9' 42 9.Sc 26 9.27 42 7.11 25 Fig. 9.5.1-1 28 9.28 28 7.11a 25 Fib. 9.5.1-2 26 9.29 27 7.12 23 pig. 9.5.1-3 26 9.30 28 1 7.13 23 pig. 9.5.2-1 26 Fig. 9.30-1 28 7.14 23 Fig. 9.5.2-2 26 9.31 28 7.15 25 Fig. 9.5.2-3 26 9.32 28 7.16 23 pig. 9.5.2-4 28 9.33 31 7.18 23 9.6 26 9.34 28 7.19 _25 9.6a 26 9.35 28 7.20 22 9.6b 27 9.35a 38 7.21 23 9.6c 37 9.35b 28 7.21t 23 pig. 9.6.4-1 26 9.36 27 7.22 37 pig. 9.6.4-2 20 9.37 27 7.24 30 pig. 9.6.4-3 27 Fig. 9.37-1 27 7.27 28 pig. 9.6.4-4 27 Fig. 9.37-2 27 7.28 28 Fig. 9.6.4-5 27 Fig. 9.37-3 27 C_ 7.29 28 27 pig. 9.6.4-6 27 Fig. 9.37-4 27 q 7.30_ Fig. 9.'6.4-7 27 9.39 27 7.31 34 9.7 28 9.41 27 7.32 27 9.8 36 Fig. 9.41-1 27 8.6 22 9.8a 28 Fig. 9.41-2 27 8.6a 22 9.9 28 9.42 42 Fig. 8.6.1 22 9.10 28 9.43 27 8.8 28 9.11 28 9.44 28 8.10 27 9.12 42 9.44a 28 8.10a 27 -
9.13 28 9.44b 28 8.12 28 9.14 28 9.44c 28 8.13 27 9.15 28 9.44d 28 Fig. 8.13-1 44 9.16 28 9.44e 28 8.14 27 pig. 9.16-1 28 9.44f 28 8.15 27 9.17 42 Fig. 9.44-1 28 8.16 28 9.18 28 9.45 28 Fig. 8.16-1 28 9.19 42 9.45a 28 Fig.-8.16-2 28 9.20 42 9.46 28 Fig. 8.16-3. 28 9.21 27 9.47 34 7 Fig. 8.16-4 28 9.22 28 9.47a 34 Fig. 8.16-5 25 9.22a 28 9.47b 34 9.1 37 9.23 42 9.47c 34 9.la 37 9.24 28 9.47d 34 9.2 22 9.25 28 Fig. 9.47-1 28 f 9.2a 22 pig. 9.25-1 28 Fig. 9.47-2 28 EP-14 J
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g * ( Page Amend. Page Amend. Page Amend. No. No. No. No. No. No. 9.48 28 11.6j 25 12.11d 41 9.49 42 11.6k 25 12.11e 41 9.50 28 11.61 31 12.11f 41 9.50a 28 11.6m 31 12.11g 41 9.50b 28 11.6n 31 12.11h 41 9.50c 28 11.7 28 12.11i 41 9.50d- 28 11.9 42 12.11j 41 9.50e 28 11.10 28 12.11k 41 9.61 27 11.10a 37 12.111 41 9.62 27 11.11 28 12.11m 41 9.63 27 11.12 28 12.11n 41 9.64 28 11.12a 28 12.11o 41 9.65 27 11.13 28 12.11p 41 9.66 27 11.14 28 12.11.q 41 9.67- 31 11.15 28 12.11r 41 9.68 28 11.16 28 12.11s 41 9.69 28 11.17 28 12.11t 41 9.69a 28 11.17a 28 12.11u 41 9.70 28 11.19 28 iz.11v 41 9.71 33 11.19a 28 12.11w 41 9.71a 28 11.19b 28 12.11x 41 9.71b. 28 11.19c 28 12.11y 41
) 9.71c 28 11.20 28 12.112 41 ] 9.72 42 11.21 27 12.11aa 41 9.73 28 11.22 27 12.11bb 41 9.74 28, 11.22a 27 12.11cc 28 9.75 28 11.22b 27 1^.11dd 41 9.75a 28 11.23 27 14.1 31 Fig. 9.75-1 28 11.23a 27 14.la 31 Fig. 9.75-2 28 11.24 28 Fig. 14.1-1 31 Fig. 9.75-3 28 11.24a 28 Fig. 14.1-2 31 Fig. 9.75-4 28 11.25 27 14.2 24 9.76 42 11.25a 27 I4.2a I 24 9.77 28 11.26 27 14.2b 24 10.1 23 11.26a 27 14.2c 24 10.2 23 11.26b 27 14.3 42 10.4 28 12.1 26 14.3a 25 10.4a 28 12.2 25 14.4 24 11.3 43 12.4- 25 14.5 28 11.6 25 12.5 25 14.6 28 11.6a 31 12.5a 25 14.7 41 11.6b 31 12.6 36 14.7a 41 11.6c- 31 12,7 25 14.7b 41 11.6d 31 12.8 25 14.7c 41 11.6e 36 12.(9-10) 29 14.8 37 11.6f 31 12.11 41 14.8a 28 11.6g 25 12.11a 28 14.9 28 (i .
11.6h 11.61 25 25 12.11b 12.11c 28 41 14.9a Fig. 14.9-1 28 28 EP-15
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Page Amend. Page Amend. No. No. No. No. 14.10 34 26 25 14.10a 34 27 25 14.10b 34 28 25 14.10c 34 29 25 f l 14.10d 34 30 25 14.10e 34 31 25 14.10f 34 32 25 14.10g 34 33 25 l Fig. 14.10.1-1 34 34 25 l Fig. 14.10.2-1 34 35 25 l Fig. 14.10.3-1 34 36 25 14.11 46 37 25 l 14.12 33 38 25 14.12a 33 39 25 AS.4 ^28 40 25 A5.4a 28 41 25 AS.5 28 42 25 l 43 25 RESPONSES TO NEW 44 25 GENERAL DESIGN 45 25 CRITERIA 46 25 47 25 [A] 1 ' 1 2 25 25 48 49 25 25 3 25 50 25 4 25 51 25 5 25 52 26 l 6 25 53 25 I 7 25 54 25 8 25 55 25 9 25 56 25 10 25 57 25 f 11 25 58 25 12 25 59 25 13 25 60 25 14 25 15 25 ; 16 25 i 17 42 17a 42 17b 42 IS 25 19 45 t 20 25 ; 21 25 j 22 25 1 23 25 y ^~,. 24 25
) / 25 25 ty/
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