1CAN048906, Forwards Results of Util Reanalysis for Qualification of Equipment Inside Reactor Bldg,Including Svc Water Degradations.All Environmentally Qualified Components Capable of Performing Intended Safety Functions

Forwards Results of Util Reanalysis for Qualification of Equipment Inside Reactor Bldg,Including Svc Water Degradations.All Environmentally Qualified Components Capable of Performing Intended Safety Functions

ML20244E169
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/14/1989
From:	James D ARKANSAS POWER & LIGHT CO.
To:	Calvo J NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
1CAN048906, 1CAN48906, NUDOCS 8904240324
Download: ML20244E169 (14)

Text

-

7 ,

ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK, ARKANSAS 72203 [501) 3774000 April 14,1989 1CAN048906 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, DC 20555 ATTN: Mr. Jose A. Calvo, Director Project Director, Region IV Division of Reactor Projects III, IV, V and Special Projects

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Additional Information Associated with Service Water Degradations

Dear Mr. Calvo:

In Arkansas Power & Light Company's (AP&L) meeting with the NRC Region IV Staff on April 3, 1989, we provided details on our approach toward the resolution of current issues pertaining to the service water (SW) system at Arkansas Nuclear One - Unit 1 (ANO-1). These issues include: emergency cooling pond lineup testing; sluice gate leakage; snap rings; and, containment response /EQ. AP&L issued a Licensee Event Report on March 30, 1989 (ICAN038901) concerning the containment response /EQ issue.

The purpose of this letter is to formally provide the results of AP&L's reanalysis for the qualification of equipment inside the ANO-1 Reactor Building, considering the limiting factors that have been recently identified af fecting the Reactor Building's post-LOCA temperature profile.

This letter is also in response to the request made by Mr. C. Craig Harbuck, in his letter of November 29, 1988 (ICNA118805) to provide the supporting analysis to revise the 70 F lake temperature criteria.

I. Containment Temperature Response Curve Analysis The original pressure / temperature transient analyses for ANO-1 were performed using the computer program COPATTA as described in Section 14.2.2.5.5.2 of the ANO-1 Safety Analysis Reports (SAR). Subsequent to this design basis accident (DBA) analyses, revised DBA analyses using COPATTA have been performed as a result of various deviations identified from design conditions to demonstrate compliance with the [

design basis for ANO-1. l 8904240324 890414  : (

PDR ADDCK 05000313 P PNV MEMBEA MOOLE SOUTH UTILITIES SYSTEM l

ICAN048906 April 14, 1989

. A. Assumptions In order to provide a basis for long-term operation of the plant, considering tne identified limiting conditions, a COPATTA analysis i was performed to address SW temperatures in excess of 70 F. The 70 F case had previously been performed to justify interim operation. The results of this analysis and AP&L's commitment not to operate the plant if the SW temperature exceeded 70 F were presented in AP&L's letter of December 5, 1988 (ICAN128807).

Several assumptions to justify the operation of the plant with a maximum SW temperature of 70 F were included in this new analysis.

These assumptions are as follows:

• One train of Emergency Core Cooling System (ECCS) in operation.

• Initial containment temperature of 140 F.

• SW flow to the decay heat removal heat exchangers of 1600 gpm.

• One reactor building cooler available.

• Reduction in the heat transfer coefficient associated with reduced flow to the decay heat removal heat exchangers.

New assumptions and additional information were also utilized in this analysis. These include:

• SW is taken from the Dardanelle Reservoir at a constant temperature of 95 F (previously 85 F).

• Reactor building cooler using 95 F SW. (Performance data corresponding to SW temperature of 95 F was recently obtained from vendor.)

The basis for these assumptions are as follows:

• A COPATTA parametric analysis was performed, demonstrating that the assumption of one ECCS train is conservative. This is also consistent with limiting single failure assumption of one diesel generator.

• The initial containment temperature of 140 F reflects a more l realistic average containment temperature during normal operation.

• During the eighth refueling outage (IR8), SW flows to the decay heat removal heat exchangers were measured at greater than 1,700 gpm for either loop's operation (limiting case).

ICAN048906 April 14, 1989

• Currently, all four reactor building coolers are available; however, technical specifications allow one cooler in each train to be out of service. Assuming a single failure of one diesel generator, only one cooler is assured available.

• Based on a reduction of heat exchanger efficiency due to reduced cooling water flow, the heat transfer coefficient was reduced accordingly.

• Operating experience has shown that the maximum temperature reached in the Dardanelle Reservoir over the past ten years has not exceeded 92 F. The design basis for ANO-1 was established with the Dardanelle Reservoir providing the ultimate heat sink (UHS) for normal shutdown and emergency recovery following a design basis LOCA. Consistent with this design basis, the original DBA LOCA containment pressure / temperature analysis - and all subsequent reanalyses submitted to the NRC - have been performed with a constant SW temperature associated with use of the Dardanelle Reservoir as the UHS. Subsequent to the original design, an emergency cooling pond (ECP) was added to the site. However, from a DBA LOCA containment pressure /teroperature analysis perspective, the basis remains for the utilization of the Dardanelle Reservoir for LOCA recovery.

B. Results of Analyses The results of the analysis to justify operation with lake temperatures in exce!,s of 70 F indicates the analysis is acceptable for determining long-term environmental conditions, as well as, containment peak pressure and temperature post-DBA dose analysis, reactor building energy absorption, and ECCS analyses.

Attachment 1 provides a detailed description of the results of the analyses.

II. Methods Used In Qualifying Equipment The containment temperature and pressure profiles generated from this analysis were used in making assessments of the environmental qualifications of components inside the ANO-1 Reactor Building. Table 1 provides a list of the components affected by this analysis. IE Bulletin 79-01B Section 4.1(1) specifies the utilization of these profiles for EQ assessments.

For some of the components, the test temperature profile completely enveloped the analysis profile, demonstrating the qualification of the component. For other devices, additional test documentation was purchased to envelope the new profiles.

l The test temperature profile for the remaining components did not i

completely envelope the analysis profile due to the relatively short I

duration of tests typical of IE Bulletin 79-01B vintage equipment.

Since the degradation of the component is primarily a time dependent ,

I L - - - - - - - - - - - - _ - - - - - - - - - - - - - - _

ICAN648986 April 14, 1989 functio,1 of temperature effects (i.e., oxidation), the test profile and the analysis profile were :ompared using an equivalent thermal degradation technique based on the Arrhenuis Methodology. In using this technique, each curve was dissected into discreet increments of time at a specific temperature and the. equivalent life at a specified reference temperature of 120 F was calculated. If the snount of degradation from the test exceeds the degradation from the postulated DBA conditions and the peak temperature is enveloped by the test profile, the test is then considered to demonstrate the ability of this component to perform its safety function during the DBA. Each qualification file was individually reevaluated using the described approach and qualification was demonstrated. The EQ files have been updated as required.

III. Conclusions As documented above, detailed evaluations were performed which justified the acceptability of interim operation using revised DBA analyses. This analysis demonstrated the acceptability of plant operation subject to the lone restriction that the Dardanelle Reservoir SW temperature not exceed 70 F. Additional analysis and evaluations were performed as. described above which justify unrestricted operation; therefore, the November 1988 justification for interim operation is considered removed.

Utilizing the techniques described in Section II, it has been determined that all EQ components are capable of performing their l intended safety function as designed based upon post-LOCA reactor building temperature profiles in accordance with IE Bulletin 79-01B and are, therefore, qualified.

Very truly yo rs,

/P ale E. ames Supersi or, Licensing DEJ:RWC:de Attachments t

a

TABLE 1 COMPONENTS LOCATED INSIDE THE ANO-1 REACTOR BUILDING 4

.l

- _ _ . _ _ _ _ _ _ _ _ _ _ _ __.___________.______-_____-_--___-m__ _ _

, s .

i

\-

Page ho. 1 04/05/89 E0 COMPONENTS INSIDE CONTAINMENT EQUIMPENT _ COMPONENT SERVICE TAG NUMBER

==::::::::::::::: ====== ssnsassassassan=sas.3,ses=====..... ............................ ................

l CV 1000 ZS 1000 MOTOR VALVE OPERATOR WITH LIMIT SWITCH PZR VENT BLOCK VALVE CV 1050 ZS 1050 MOTOR VALVE OPERATOR WITH LIMli SWITCH RC LOOP A HOT LEG f0 DH REMOVAL VLV CV 1053 ZS 1053 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RC QUENCH TAWK ISOLATION VALVE CV 1054 ZS 1054 MOTOR VALVE OPERATOR WITH LIMIT SWITCH QUENCH TANK SAMPLE ISOLATION VALVE CV 1214 ZS 1214 MOTOR VALVE OPERATOR WITH LIMIT SWITCH LD COOLER E29A DISCH ISOL VALVE CV 1216 25 1216 MOTOR VALVE OPERATOR WITH LIMIT SWITCH LD COOLER E29B DISCH ISOL VALVE CV 1270 ZS 1270 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RCP 32D SEAL BLEED 0FF ISOL VALVE CV 1271 25 1271 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RCP 32C SEAL BLEED 0FF ISOL VALVE CV 1272 251272 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RCP 328 SEAL BLEED 0FF ISOL VALVE CV 1273 ZS 1273 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RCP 32A SEAL BLEEDOFF !$0L VALVE CV 1410 ZS 1410 MOTOR VALVE OPERATOR WITH LIMIT SWITCH DHR FROM RCS BLOCK VALVE CV 1414 ZS 1414 MOTOR VALVE OPERATOR WITH LIMIT SWITCH DHR LOOP A SUMP ISOLATION VALVE CV 1415 ZS 1415 MOTOR VALVE OPERATOR WITH LIMIT SWITCH DHR LOOP B SUMP ISOLATION VALVE CV 1814 25 1814 MOTOR VALVE OPERATOR WITH LIMIT SWITCH $$ PZR GAS SPACE ISCLATION VALVE CV 1816 25 1816 MOTOR VALVE OPERATOR WITH LIMIT SWITCH $$ PZR LIQUID SPACE ISOL VALVE CV 1820 ZS 1820 MOTOR VALVE OPERATOR WITH LIMIT SWITCH SS SCA ISOLATION VALVE CV 1826 ZS 1826 MOTOR VALVE OPERATOR WITH LIMIT SWITCH $$ SGB ISOLATION VALVE CV 2215 IS 2215 MOTOR VALVE OPERATOR WITH LIMIT SWITCH LETDOWN COOLERS ISOLAfl0N VALVE CV 2221 ZS 2221- MOTOR VALVE OPERATOR WITH LIMIT SWITCH CRD/RCP COOLANT ISOLAI!ON VALVE CV 2415 ZS 2415A MOTOR VALVE OPERATOR WITH LIMIT SWITCH CORE FLOOD TANKT-2A BLOCK VLV CV 2416 ZS 2416 MOTOR VALVE OPERATOR 'J1TH LIMIT SWITCH CORE FLOOD TANK T 28 CV 2418 15 2418 MOTOR VALVE OPERATOR WITH LIMIT SWITCH CORE FLOOD TANK T-2A CV 2419 ZS 2419A MOTOR VALVE OPERATOR Wl1H LIMIT SWITCH CORE FLOOD TANKT 2B BLOCK VLV CV 4446 25 4446 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB SUMP TO ISOL SUMP CV 4803 25 4803 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB VENT HEADER VALVE CV 5612 ZS 5612 MOTOR VALVE OPERATOR WITH LIMIT SWITCH ISOLATION VALVE TO RB FIRE WATER CV 6205 ZS 6205 MOTOR VALVE OPERATOR WITH LIMIT SWITCH CW TO RB ISOL VALVE CV 7444 25 7444 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB ISOLATION VALVE AIR IN CV 7446 ZS 7446 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB ISOLATION VALVE AIR IN CV 7448 IS 7448 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB ISOLAfl0N VLV M2 PURGE OUT CV 7450 ZS 7450 MOTOR VALVE OPERATOR WITH LIMIT SWITCH RB ISOLAfl0N VLV H2 PURGE OUT CV 7453 ZS 7453 MOTOR VALVE OPERATOR WITH LIMIT SWITCH A!R PARTICULATE MON SYS ISOL VLV CV 7470 ZS 7470 MOTOR VALVE OPERATOR WITH LIMIT SWITCH COOLING UNIT VSFI A BACKDRAFT DAMPER CV 7471 ZS 7471 MOTOR VALVE OPERATOR WITH LIMIT SWITCH COOLING UNIT VSFIB BACKDRAFT DAMPER CV 7472 25 7472 MOTOR VALVE OPERATOR WITH LIMIT SWITCH COOLING UNIT VSFIC BACK0 RAFT DAMPER CV 7473 ZS 7473 MOTOR VALVE OPEkATOR WITH LIMIT SWITCH COOLING UNIT VSFID BACKDRAFT DAMPER GEN 1002 ELECTRICAL CABLE VARIOUS GEN 1003 PENETRATION ASSEMBLIES ELECTRICAL SERVICE CEN 1004 PENETRATION ASSEMBLIES ELECTRICAL SERVICE GEN 1005 PENETRATION ASSEMBLIES ELECTRICAL SERVICE CEN 1006 PENETRATION ASSEMBLIES ELECTRICAL SERVICE CEN 1007 HARDLINE CCAXIAL CABLE CABLE FOR ACOUSTIC VALVE POS. IND, CEN 1003 CABLE CDUPLER CABLE CONN ACOUSTIC VLV POS IND GEN 1010 ELECTRICAL CABLE 600V POWER AND CONTROL GEN 1C11 PENETRATION ASSEMBLIES ELECTRICAL SERVICE GEN 1012 PENETRATION ASSEMBLIES ELECTRICAL SERVICE CEN 1013 PENETRATION ASSEMBLIES ELECTRICAL SERVICE CEN 1014 PENETRATION ASSEMBLIES ELECTRICAL SERVICE

Fage No. 2 04/05/S9 EQ COMPONENTS INSIDE CONTAINMENT

'EQUIMPENT COMPONENT SERVICE TAG NUMBER-

==== ssesssssssssssssssa esass===ss==s sssssss============se ==seses===ssssssss==ssessan==sssss===ssasemma GEN 1015 PENETRATION ASSEMBLIES ELECTRICAL SERVICE GEN 1018 CABLE SPLICE CONTROL & POWER (<480 V) & INSTRUMENT GEN 1020 ELECTRICAL CONNECTOR (ECSA) VARIOUS INSTRUMENTS GEN 1021 ELECTRICAL CABLE VARIOUS POWER AND CONTROL GEN 1022 COAXIAL CABLE VARIOUS INSTRUMENTATION GEN 1023 ELECTRICAL CABLE VARIOUS INSTRUMENTATION-GEN 1024 INSULATING TAPE VARIOUS CLASS 1E MOTORS GEN 1025 FEEDTHROUGH/ ADAPTER MODULE ASS'Y VARIOUS INSTRUMENTATION GEN 1026 XL'E FIREWALL !!! CABLE VARIOUS CONTROL, POWER & INSTRUMENTATION APPL GEN 1027 ELECTRIC CABLE 600 VOLT POWER AND CONTROL GEN 1028 ELECTRICAL CABLE VARIOUS

. GEN 1029 ELECTRIC CABLE TRANSMIT INSTRUMENTAL!DN S!GNALS GEN 1C30 ELECTRICAL CABLE MONITORING GEN 1031 ELECTRIC CABLE POWER AND CONTROL CEN 1032 ELECTRICAL CABLE VARIOUS GEN 1034 SIS WIRE: XLPE FIREWALL !!! VARIOUS GEM 1035 ELECTRIC CABLE ELECTRICAL DISTRIBUTION GEN 1036 ELECTRICAL CABLE VARIOUS GEN 1037 ELECTRICAL CABLE ELECTRICAL DISTRIBUTION GEN 1038 ELECTRICAL CABLE ELECTRICAL DISTRIBUTION GEN 1039 TERMINAL BLOCK VARIOUS GEN 1040 NUCLEAR SERV 1CE CONNECTOR VARIOUS GEN 1042 TERMINAL BLOCK VARIOUS GEN 1043 INSTRUMENTATION CABLE VARIOUS GEN 1044 INSULATING TAPE VARIOUS LIMITOROUE OPERATORS GEN 1045 NPKV SPLICE K!T 1E PWR, CNTRL, & INST CIRCUITS 480V & LESS GEN 1C46 CABLE SPLICES VARIOUS LIMITORQUE ACTUATORS GEN 1047 COAXIAL FEEDTHROUGH ADAPTOR MODULE VARIOUS INSTRUMENTATION LE 1405 B LEVEL SENSOR CONTAINMENT FL300 LEVEL LE 5645 A&B LEVEL SENSCR CONTAINMENT FLDOD LEVEL LE 5646 A&B LEVEL SENSOR CONTAINMENT FLD00 LEVEL LT 1001 LEVEL TRANSMITTER PRESZ LEVEL LT 1002 LEVEL TRANSMITTER PRESZ LEVEL LT 1189 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICATION LT 1190 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDI AsiON LT 1191 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICAT!DN LT 1192 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICAT!DN LT 1193 DIFFERENTI AL PRESSL'RE TRANSMITTER HOT LEG LEVEL INDICAfl0N LT 1194 DIFFERENTIAL ** ESSURE TRANSMITTER HOT LEG LEVEL INDICAfl0N LT 1195 DIFFERENTIAL .T;SURE TRANSMITTER HOT LEC LEVEL 14DICATION LT 1196 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICATION LT 1197 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICATION LT 1198 DIFFERENTIAL PRESSURE TRANSMITTER HOT LEG LEVEL INDICATION l- LT 2415 LEVEL TRANSMITTER CORE FLOOD TANK T2A LT 2416 LEVEL TRANSMITTER CORE FLDOD TANK T2A LT 2418 LEVEL TRANSMITTER CORE FLD00 TANK T2B l LT 2419 LEVEL TRANSMITTER CORE FLOOD TANK T2B LT 2617 LEVEL, TRANSMITTER FTEAM GENERATOR E248 LEVEL LO RANGE l

1

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ __J

l '

Page No. 3 04/05/89

'E3 COMPONENTS INSIDE CONTAINMENT EQUIMPENT COMPONENT SERVICE

' TAG

l. NUMBER l

l LT 2618 LEVEL TRANSMITTER STEAM GENERATOR E24A LEVEL LO RANCE LT 2619 LEVEL TRANSMITTER STEAM GENERATOR E24B LEVEL HI RANGE LT 2620 LEVEL TRANSMITTER STEAM CENERATOR E24A LEVEL UPPER RANGE LT 2621 LEVEL TRANSMITTER STEAM GENERATOR E248 LEVEL LO RANGE LT 2622 LEVEL TRANSMlifER STEAM GENERATOR LEVEL LT 2623 LEVEL TRANSMITTER STEAM GENERATOR E24B LEVEL UPPER RANGE LT 2624 LEVEL TRANSMITTER STEAM GENERATOR LEVEL LT 2667 LEVEL TRANSMITTER STEAM CENERATOR E24p LEVEL LO RANCE LT 2668 LEVEL TRANSMITTER STEAM GENERATOR E2tA LEVEL LO RANGE LT 2669 LEVEL TRANSMITTER STEAM GENERATOR E24-8 LEVEL UPPER RANGE LT 2670 LEVEL TRANSMITTER STEAM CENERATOR E24A LEVEL UPPER RANGE LT 2671 LEVEL TRANSMITTER STEAM CENERATOR E248 LEVEL LO RANCE LT 2672 LEVEL TRANSMITTER STEAM CENERATOR LEVEL LT 2673 LEVEL TRANSMITTER STEAM GENERATOR E24B LEVEL UPPER RANGE LT 2674 LEVEL TRANSMITTER STEAM GENERATOR LEVEL M SSA HTDROCEN RECnMBINER REACTOR BLDG CONTAINMENT VENTILATION M SSB HYDROGEN RECOMBINER REACTOR BLDG CONTAINMENT VENTILATION NE 501 DETECTOR EXCORE NEUTRON FLUX NE 502 DETECTOR EXCORE NEUTRON FLUX PT 1020 PRESSURE TRANSMITTER RC LOOP A HOT LEG PT 1021 PRESSURE TRANSMITTER LOOP A HOT LEG TO RPS PT 1022 PRESSURE TRANSMITTER RC LOOP A HOT LEG PT 1023 PRESSURE TRANSMITTER LOOP A HOT LEG PT 1038 PRESSURE TRANSMITTER LOOP B HOT LEG PT 1039 PRESSURE TRANSMITTER LOOP A HOT LEG PT 1040 PRESSURE TRANSMITTER RC LOOP B HOT L2G PT 1041 PRESSURE TRANSMITTER ICCMDS TSAT ANDINTLK FCA HEAT DECAY CV1410 PT 1042 PRESSURE TRANSMITTER WIDE RANGE RCS PRESS TO SPDC AND ICCMDS TSAT PT 2400 PRESSURE TRANSMITTER RB PRESSURE REACTOR TRIP PT 2401 PRESSURE TRANSMITTER RB PRESSURE REACTOR TRIP PT 2402 PRESSURE TRANSMITTER RB PRESSURE REACTOR TRIP PT 2403 PRESSURE TRANSMITTER RB PRESSUPE REACTOR TRIP PT 2405 PRESSURE TRANSMITTER SPRAY REACTOR BLD NUC ENG SFEGRD(1)

PT 2406 PRESSURE TRANSMITTER SPRAY REACTOR BLD NUC ENG SFEGRD(2)

PT 2407 PRESSURE TRANSMITTER SPRAY REACTOR BLD NUC ENG SFEGRD(3)

PT 2412 PRESSURE TRANSMITTER GENERAL CONTAINMENT AREA PT 2413 PRESSURE TRANSMITTER GENERAL CONTAINMENT AREA RE 8060 RADIATION ELEMENT CONT POST LOCA AREA RAD MONITOR RE 8061 RADIATION ELEMENT CONT POST LOCA AREA RAD MONITOR SV 1071 SOLEWotD VALVE OPERATOR WITH LIMIT SW RCS REACTOR HEAD ZS 1071 SV 1072 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS REACTOR HEAD 23 1072 SV 1073 SOLEN 0!D VALVE OPERATOR WITH LIMIT SW RCS REACTOR HEAD ZS 1073 SV 1074 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS REACTOR HEAD ZS 1074 SV 1G77 $0LEN,0!D VALVE OPERATOR WITH LIMIT SW RCS PRESSURIZER ZS 1077

i i

Page No. 4 04/05/87 EQ COMPONENTS INSIDE CONTAINMENT

.EOUIMPENT COMPONENT SERVICE TAG NUMBER assasssesasssssssssssssa sessssammass===sassamasssssssssssas ass ===sessenesssss==sssans===sassess== ssumes SV 1079 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS PRESSURIZER

.ZS 1079 SV 1081 SOLEN 0!D VALVE OPERATOR WITH LIMIT SW RCS STEAM GEN E24A 25 1081 SV 1082 $0LEN0!D VALVE OPERATOR WITH LIMIT SW RCS STEAM GEN E24A

.25.1082 SV 1083 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS STEAM GEN E24A ZS 1083 SV 1084 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS STM GEN E24A

'.ZS 1084 SV 1091 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS STM GEN E24B ZS 1091 SV 1092 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS STM GEN E248 ZS 1092 SV 1093 SOLENOID VALVE OPERATOR WITH LIMIT SW RCS STM GEN E24B ZS,1093 SV 1094 $0LEN010 VALVE OPERATOR WITH L!MIT SW RCS STM CEN E248 ZS 1094 SV 1840 SOLENOID VALVE & POS SW RCS HOT LEG SAMPLE ZS 1840 SV 7410 ELECTRIC MOTOR - VSF1A BYPASS DAMPER SV 7411 ELECTRIC MOTOR VSF1B BYPASS DAMPER SV 7412 ELECTRIC MOTOR VSFIC BYPASS DAMPER SV 7413 ELECTRIC HOTOF VSF1D BYPAS$ DAMPER TE 1001 RESISTANCE TEMPERATURE DEVICE PRZ. TEMP. COMPENSATION TE 1002A RESISTANCE TEMPERATURE DEVICE PRZ. TEMP, TE 1012 TEMPERATURE SENSOR RC LOOP A HOT LEG TE 1013 TEMPERATURE SENSOR RC LOOP A HOT LEG TE 1040 TEMPERATURE SENSOR RC LOOP B HOT LEG TE 1041 TEMPERATURE SENSOR RC LOOP B HOT LEG TE 1111 RESISTANCE TEMPERATURE DEVICE RC LOOP A HOT LEG TE 1112 RESISTANCE TEMPERATURE DEVICE RC LOOP A HOT LEG TE 1139 RESISTANCE TEMPERATURE DEVICE RC LOOP B HOT LEG TE 1140 RESISTANCE TEMPERATURE DEVICE RC LOOP B HOT LEG TE 1151 CORE EXIT THERMOCOUPLE REACTOR TE 1153 CORE EXIT THERMOCOUPLE REACTOR TE 1154 CORE EXIT THERMOCOUPLE REACTOR TE 1156 CORE EXIT THERMOCOUPLE REACTOR TE 1157 CORE EXIT THERMOCOUPLE REACTOR

'TE 1159 CORE EXIT THERMOCOUPLE REACTOR TE 1160 CORE EXIT THERMOCOUPLE REACTOR TE 1162 CORE EXIT THERMOCOUPLE REACTOR TE 1164 CORE EXIT THERMOCOUPLE REACTOR TE 1165 CORE EXIT THERMOCOUPLE REACTOR TE 1166 CORE EXIT THERMOCOUPLE REACTOR TE 1167 CORE EXIT THERMOCOUPLE REACTOR TE 1169 CORE EXIT THERMOCOUPLE REACTOR TE 1170 CORE , EXIT THERMOCOUPLE REACTOR

.Page No. 5 04/05/89 EQ COMPONENTS INSIDE CONTAINMENT EQUIMPENT COMPONENT SERVICE

-TAG NUMBER ssessssssssssssssssssssa assassasssssssssssssssssssssssssssa aussssssssssssssssssssssssssssssssomassanssas TE 1173 CORE EXIT THERMOCOUPLE REACTOR TE 1174 CORE EXIT THERMOCOUPLE REACTOR TE 1189 RTD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1190 RTD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1191 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1192 RTD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1193 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1194 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1195 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER i

TE 1196 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1197 RfD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1198 RTD REF. LEG OF HOT LEG LEVEL TRANSMITTER TE 1152 CORE EXIT THERMOCOUPLE REACTOR TE-1155 CORE EXIT THERMOCOUPLE REACTOR TE-1158 CORE EXIT THERMOCOUPLE REACTOR TE-1161 CORE EXIT THERMOCOUPLE REACTOR TE-1163 CORE EXIT THERMOCOUPLE REACTOR TE-1168 CORE EXIT THERMOCOUPLE REACTOR TE-1171 CORE EXIT THERMOCOUPLE REACTOR TE 1172 CORE EXIT. THERMOCOUPLE REACTOR UE 1187 RL1 AND ASSOC. CABLING AND CONNECTORS REACTOR VESSEL UE 1188 RLI AND ASSOC. CABLING AND CONNECTORS REACTOR VESSEL VBE 1000A ACCELEROMETER MONITORS POSIT!DN OF PZR RELIEF VLV VBE 10008 ACCELEROMETER MONITORS POSIT!DN OF PZR RELIEF VLV VBE 1001A ACCELEROMETER MONITORS POSITl0N OF PER RELIEF VLV VBE 1001B ACCELEROMETER MONITORS POSIT 10N OF PZR RELIEF VLV VBE 1002A ACCELEROMETER MONITORS POSIT!DN OF PER RELIEF VLV VBE 1002B ACCELEROMETER MONITORS POSITION OF PZR RELIEF VLV VBY 100CA PREAMPLIFIER PRESSURIZER RELIEF VLV FLOW MONITOR VBY 10008 PREAMPLIFIER PRESSURIZER RELIEF VLV FLOW MONITOR VBY 1001A PREAMPLIFIER PRESSURIZER RELIEF VLV FLOW MONITOR VBY 10018 PREAMPLIFIER PRESSURIZER RELIEF VLV FLOW MONITOR VBY 1002A PREAMPLIFIER PRESSURIZER RELIEF VLV FLOW MONITOR VBY 10028 PREAMPLIFIER PRESSUR!ZER RELIEF VLV FLOW MONITOR VSFM 1A ELECTRIC MOTOR VSF 1A BLOWER VSFM 1B ELECTRIC MOTOR VSF 18 BLOWER VSFM 1C ELECTRIC MOTOR VSF 1C BLOWER VSFM 1D ELECTRIC MOTOR VSF 1D BLOWER ZS 7406 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH ZS 7407 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH ZS 7408 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH IS 7409 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH ZS 7410 LIMIT SVITCH RB COOLING UNIT DAMPER SWITCH ZS 7411 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH 25 7412 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH ZS 7413 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH ZS 7420-2 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH ZS 7421-1 LIMIT, SWITCH RB COOLING UNIT DAMPER SWITCH

Page No. 6 04/05/89 EQ COMPONENTS INSIDE CONTAINMENT EQUIMPENT COMPONENT SERVICE TAG NUMBER l

ZS 7421 2 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH ZS 7422-1 LIMIT SWITCH RS COOLING UNIT DAMPER SWITCH ZS 7422-2 LIMIT SWITCH RS COOLING UNIT DAMPER SWITCH ZS 7423 1 LIMIT SWITCH RS COOLING UNIT DAMPER SWITCH ZS 7423 2 LIMIT SWITCH RS COOLING UNIT DAMPER SWITCH ZS 7424 LIMIT SWITCH RB COOLING UNIT DAMPER MOTOR SWITCH ZS 7425 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH ZS 7426 LIMIT SWITCH R8 COOLING UNIT DAMPER MOTOR SWITCH ZS 7427 LIMIT SWITCH R8 COOLING UNIT DAMPER SWITCH ZS 7420-1 LIMIT SWITCH RB COOLING UNIT DAMPER SWITCH t

9 ATTACHMENT 1 DETAILED DESCRIPTION OF ANALYSES

l l 'The November 1988 analysis, to justify operation with a maximum service water temperature of 70 F, used input assumptions and refined analysis methodology which addresses each of the issues discussed in the Summary of Analyses section and provides a temperature and pressure profile which has been acceptably evaluated for environmental qualification (EQ) impact. The results of this analysis were evaluated against the original SAR acceptance criteria and also factored into a reassessment of EQ component evaluations.

Lake temperatures at the time of this analysis was developed, were well below 70 F. Therefore, this analysis was considered adequate until lake temperatures exceeded this value.

To support continued plant operation with lake temperatures in excess of 70 F (up to 95 F), the model associated with the current analysis will be compared with the November 1988 analysis.

l

1. Long-term Environmental Conditions - The original SAR DBA containment temperature and pressure profiles were utilized for environmental qualification assessments as specified by IE Bulletin 79-01B, Section 4.1(1). The new profiles have been considered regarding possible impact upon qualification. In particular, the results from the November 1988 analysis and the results of the current analysis have been evaluated for all ANO-1 EQ items inside containment. The results demonstrate that all of the affected components are still capable of performing their intended safety functions as designed, and are therefore still qualified.

2. Containment Peak Pressure and Temperature - The 1987 analysis on elevated temperature included a COPATTA analysis that completely envelopes most input parameters for the November 1988 analysis.

This analysis was based on greater SW flow to the DHR cooler, lower initial containment temperature, and better RB cooler performance characteristics. As documented in the 1987 analysis (AP&L correspondence to NRC of August 28, 1987 (1CAN088707)) and accepted by the NRC (ICNA108703), the peak containment temperature and pressure conditions were not significantly affected.

Therefore, results from the November 1988 analysis were also considered acceptable since the impact upon peak pressures and temperatures will be smaller than those for the 1987 analysis.

This was expected since the only varied parameter affecting the results for the " peak" conditions would be the initial temperature which is 10 F lower in the November 1988 analysis. Since the significant input parameters of the current analysis are also bounded by the 1987 analysis, the peak conditions for this analysis are acceptable as well.

3. Post-DBA Dose Analysis (Containment Leakage) - The design bases for dose calculations is the containment leakage assumptions of SAR Table 14-49:

• 0.2%/ day the first day

• 0.1%/ day thereafter

.. ~

Since the driving mechanism for leakage is containment pressure, the leakage assumptions are considered valid if the pressure at 24-hours is less than half of the peak value and decreasing (i.e.,

following the second peak). As was demonstrated in the 1987 analysis, the pressure is well below 50% of the peak at 24-hours.

Therefore, the dose calculation basis remains valid. The November 1988 analysis was bounded by the 1987 analysis and was, therefore, also acceptable. The current analysis is also bounded by the 1987 analysis and is considered acceptable.

4. Reactor Building Energy Absorption - Previous analyses examined reactor building design margin by examining total energy released and maximum (design) energy capability of the reactor building (ANO-1 SAR, Section 14.2.2.5.5.5). As expected, the energy margin is at a minimum at the time of peak pressure and increases as the I building pressure decreases. Since it is bounded in the peak l pressure range by the 1987 analysis and is not significantly I different from the original SAR analysis, the November 1988 analysis was considered acceptable. The margin, at times after the peak, is indeed reduced from the original analysis; however, i

since the remaining margin is still significantly greater than

[ during the " peak" conditions, the differences are not significant.

The peak conditions of the current analysis are also bounded by l the 1987 analysis and the energy margin will not be affected.

i

5. ECCS Analyses - Since a portion of the DBA analysis inputs are I provided by Babcock and Wilcox (B&W) and based on separate analyses, an evaluation was made to determine whether the B&W analyses were impacted in any way. As documented by B&W (B&W correspondence to AP&L dated November 29, 1984 (APL-84-747)),

their calculations of mass and energy release (provided for initial blowdown data) are not affected by building cooling. A review of the SAR, Section 14.2.2 confirms that the ECCS analyses are limiting with respect to core conditions (i.e., peak clad temperature, DLR, etc.) at times early in the event. In particular, the critical times for ECCS systems are well in advance of sump recirculation (B 25 minutes). Therefore, neither the SW flow nor temperature assumption is important to the B&W analyses. The other concerns addressed in the Summary of Analyses section are specific to the COPATTA analyses which are used for the reactor building DBA.

6. Effects On Other Analyses - The Main Steam Line Break and Steam Generator Tube Rupture analyses do involve cooldown considerations which would be impacted by degraded flow through the DHR coolers.

However, the accident consequences under consideration are radiation releases which are not affected when cooldown is transferred from the faulted steam generator to the coolers.

Therefore, accident doses are not affected.

_ _ - _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _