ML20039E882
| ML20039E882 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/30/1981 |
| From: | Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8201110580 | |
| Download: ML20039E882 (20) | |
Text
- _ _ _.
POWER AUTHORITY OF THE STATE CF NEW YORK
~
INDIAN POINT NO. 3 NUCLEAR POWER PLANT d
P. O. BOX 215 BUCHANAN.N.Y.10511
[g TELEPHONES 914 739 8200 8
n.
. n-.c; ;..
y
~mj r
!C JAN 81992Hf M,,,
/*
y,.,..,,
..a
~
a e
_ :a p
December 30, 1981 k
,< )
Docket No. 50-286
\\\\ (-
'\\
License No. DPR-64 Ronald C. Haynes, Regional Administrator Office of Inspection and Enforcement Region 1 U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Code of Federal Regulations 10 CFR 50.59 Changes, Tests and Experiments
Dear Mr. Haynes:
The following constitutes the annual report on changes, tests and experiments for Indian Point 3 Nuclear Power Plant as required by 10 CFR 50.59.
The Code of Regulations 10 CFR 50.59 a. specifies that changes to the facility as described in the safety analysis report, changes in the procedures as described in the safety analysis report and conduct of tests or experiments not described in the safety analysis report may be made without prior Commission approval provided the proposed change, test or experiment does not involve a change in the technical specifications incorporated in the license or constitute an unreviewed safety question.
A description of such changes, procedures and tests performed at Indian Point 3 ar.d a summary t,f the safety evaluations of each for the period of March 7,1980 to December 31, 1980 are contained in Atcach-ment 1. Each has been reviewed to ensure that the probability of occurrence or the consequences of an accident or malfunction of equip-ment important to safety previously evaluated ir the safety analysis report has not been increased, the possibility for an accident or malfunction of a different type than any evaluated previously in the 8201110580 811230 PDRADOCKOSOOOg R
$6 )e q
r-
. safety analysis report has not been created, nor the margin of safety as defined in the basis for any technical specification has not been reduced.
It was concluded that the changes, tests and experiments did not constitute an unreviewed safety question.
Very truly yours, QGt2-John C. Brons Resident Manager
+
RAS /gg Attachment cc:
Director, Office of Inspection and Enforcement (39 Copies) c/o Distributien Services Branch, DDC, ADM U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Robert Young, Acting Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Resident Inspector T. Kenny
f ATTACIIMENT I l
l l
l l
l l
Power Authority of the State of New York Indian Point 3 Nuclear Power Plant Docket No. 50-286 i
c
m.
(-
E A; Modification'and Evaluations
, 03-01 FP - Fire Protection - CO2 System This modification involved the'installationiof a low pressure'CO2 system which
' provides fire protection for the ' turbine generator bearings.
L This modification improves the fire protection capability of the plant and the American Nuclear Insurers (ANI) has concurred in this change.
~
This system is tied into the Plant Fire' Detection and Alarm System which annunciates
. locally and in the Control Room. Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes.
This
. modification was designed considering original separation criteria thus maintaining l
the. integrity of electrical separation. This modification was fabricated using standards equal to or better than those used during original installation. This' l
modification has therefore been deemed to not involve an unreviewed safety j.
question.-
'79-3-016 FP - Turbine Building Fire -Protection Modification This modification involved the installation'of new Fire Protection and Detection l
Systems in the Turbine Building.,The areas over which they provide ' fire protection are provided by two' types of systems, ie.
Wet-Pipe Sprinkler Systems and i
Deluge Water Spray Systems.
l l
The Wet-Pipe Sprinkler Systems provide fire protection to.the following areas:
1.
Turbine Generator Building elevation 15' south half.
2.
Turbine Generator Building elevation 15' north half.
3.
Turbine Generator Building elevation 35'-9".
These systems are tied into the Plant Fire Detection and Alarm System which annunicates locally and in the Control Room.
l The Deluge Water Spray Systems provide Fire Protection to the following areas:
1.
Turbine Generator Building - Turbine Governor Housing and bearings as l
back-up sytem to the CO2 system.
l 2.
Turbine Generator Building - Boiler Feed Pumps as a back-up system to the CO2 system.
l 3.
Turbine Generator Building - Outside wall of Main and Auxiliary Transformars forming a water curtain between the transformers and the Turbine l
Building.
4'.
Turbine Generator Building - Pipe Bridge to Auxiliary Feed Pump Building forming a water curtain between the Pipe Bridge and the Turbine Building.
This modification improves the fire protection capability of the plant and the j
American Nuclear Insurers (ANI) has concurred in this change.
i I
r r
l 1 of_17
p.
3 Thase tyrt:ms mest ths raquir:ments of ths National Fire Prot ction Association (NFPA) Standards No. I's, 15, and 70.
Fabrication of piping meets the requirements of the standard code.for Pressure Piping, ANSI B31.1.
The hangers and supports in safety-related buildings are designed to Seismic Category I criteria as specified in the FSAR.
This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation.
This modification was installed using standards equal to or better than those used during original construction.
This modification has therefore been deemed to not involve an unreviewed safety question.
79-03-021 CVCS - Charging Pump Suction and Diccharge Pul sation Dampaners
- This modification involves the install & tion of pulsation stabilizer separators in the 4" suction lines and the installation of pulsation dampeners in the 3" discharge lines of the positive displacement boric acid charging pumps No. 31, 32, and 33 to reduce hydraulic pulsations during operation. Presently the stabilizers / separators portion of the modification is installed in all three pumps.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering both seismic and thermal growth factors. This modification was fabricated using standards equal to or better than those used during original installation. The modification has been therefore deemed to not involve an unreviewed safety question.
79-03-027 EL - Provide Battery and Inverter to Power TL,trument Bus Fed Bv Outside Power The purpose of this modification was to install battery, inverter, DC distribution panel, auxiliary and associated equipment in order to have Instrument Bus 34 be supplied by a battery backed inverter system.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation.
This modification was fabricated using standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety q;estion.
79-03-029 PW - Install Domineralized Water Header Inside Containment The installation of the domineralized water header provides a system for both hot (decontamination) water, cold (Hydrostatic Testing and Flushing) water, and
~
standby fire protection capabilities.
This modification enhances the fire protection system and does not degrade the security plan or system integrity. Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering both seismic and thermal growth factors.
This modification was designed considering original separation criteria thus maintainine the integrity of electrical scparation. This modification was.
fabricated using standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
?
2 of 17
F
~
I t
t 79-3-031-CB - Equipment Pr-~~ge Syst~m through Containment Airlock at Eltv? tion 95'-0" This modification consisted of the installation of a flat car / dolly rolling on two (2) rails to allow easier passage of heavy equipment into containment during outages. The rails will be carried by structural frames temporarily bolted to the existing structure. The structural assemblies provide a single level, 7
continuous support for the dolly rails enabling the dolly to be ~ rolled manually from the exter. lor roadway to the interior of the Containment where the Polar Crane.can be used to pick up the equipment on the dolly. This system is rated to permit the movement of equipnent up to five (5) tons through the airlock.
The rails are spliced at three (3) locations.
Two (2) of these are at the interior and exterior airlock doors where they can be quickly removed as required to permit the closing of the doors. Portable stairs are installed inside and outside the Containment Building for access to the walkway through the airlock.
This modification was facricated using standards equal to or better than the original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
79-03-057-CP Inject Boric Acid into Steam Generators This modification involved the installation of piping, valves, etc., from the existing chemical feed system to the main feedwater headers. This will allow addition of boric acid into the secondary side of the steam generators via the main, as well as, auxiliary feedwater headers.
Westinghouse Electric Corporation recommended the addition of boric acid to the steam generators to reduce the rate and extent of steam generator tube denting.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering both seismic and thermal growth factors. This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation.
This modification was fabricated using standards equal.to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
79-3-075 FP-Fire Protection - Cable Tunnel This modification involved the extension of the existing dry pipe pre-action system serving elevations 43'-0" and 33'-0" of the electrical cable tunnel.
It covers the upper levels of cable trays which were not covered by the original system, the cable trays in the penetration area, and extend further into the cable spreading room entry area. New distrDbution headers parallel to the existing ones were installed and the original headers remain unchanged.
This modification complies with NRC Branch Technical Position 9.5.1 and the American Nuclear Insurers (ANI) have concurred in this change.
This system is tied into the Plant Fire Detection and Alarm System which annunciates locally and in the Control Room.
This modification meets the requirements of the National Fire Protection Association TFPA) Standard No. 13 and Branch 3 of 17
Technical-Position EMT 9.5.1.
.New portions of,the system were designed to.be compatible with the existing portion of the system. All hangers and supports in
^
safety related buildings are designed to seismic Category I criteria..This modification was installed using standards equal to.better than those used
- during original construction. This modification has been therefore deemed to not involve an unreviewed safety question.
FP-Fire Al' arm and Detection System Inside Containment 79-3-086
'This modification. involved the installation of conduit, cable and seven (7) photo-electric snoke detectors inside Containnent.
These smoke detectors were l.
installed in.the following locations:
l 1.
One at each reactor coolant pump.
2.
One in the electrical penetration area 3.
Two over the cable tray on.EL 54'6".
l This modification complies with NRC Branch Technical Position 9.5.1 and the :
American Nuclear Insurers (ANI) ~have concurred in this change.
j This system is tied into the Plant Fire Detection and Alarm System which'annunciates' locally and in the Control Room.
This modification was designed considering l
original separation criteria thus maintaining the integrity of electrical separation, i
This modification was installed using standards equal to er better than those
'used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
79-3-087 FP-Fire Protection System - Fire Pump House Electrical Ir.stallation and Fire Pump Controls This modification involved the installation of the following major electrical equipment in the Fire Pump House:
1.
One (1) 250 horse power motor driven fire pump.
l 2.
Two (2) 7.5 horse power motor driven jockey pumps.
3.
Four (4) 200 Kilo-Watt tank heaters 4.
One (1) Controller for the motor driven fire pumps.
5.
.One ' (1) Controller for the diesel driven fire pump.
l 6.
Two (2) Motor control centers designated as "MCC G" and MCC H".
7.
Control Panel.
This modification helps to satisfy the three major concerns regarding the overall fire protection scope:
1.
To have a completely independent fire protection system for the IP-3 Site.
j l
2.
To comply with NRC Branch Technical Position 9.5.1.
i I
l,.
l l
N 4 of 17 -
4
' e 3
3.'
.To comply with ' requests of American Nuclear Insurers,(ANI)
This system is tied into'the Plant Fire Detection-and Alarm System which Annunciates locally and in the Control ~ Room. This modification was designed considering
- original separation criteria thus maintaining the integrity:of electrical separation.
This. modification was installed-using standards equal to or better than those used during' original installation. This modification has been therefore deemed -
- tc not involve on unreviewed safety question.
79-3-105 'EL-New ~480 V Feeders Associated with the Extension ^ of Electrical Facilities' This modification involved the' installation of the 480 Volt: Feeders that are associated with the extension of the existing electrical distribution system and the installation of a-480 volt manual transfer switch for the Fire Pump Motor Supply Circuit.
The 480 Volt Feeders for the Fire Protection System are from the 480 volt buses 312 and 313 to the 480 volt Motor Control Centers G and H respectively. Buses 312 and 313 are located in the Turbine Hall land Motor Control Centers G and H are located in the Fire Pump House. The motor driven fire pump normal feed is bus 312 and the emergency. feed is 480 volt bus SA..These feeders run through the new manual transfer switch which is used to manually transfer,the feeders -to the motor driven fire pump from the normal feed to the emergency feed and from the emergency feed to normal feed. The electrical feeds to the remaining equipment installed as part of the additional facilities program are supplied through individual breakers.
- This modification was designed and installed-using standards equal to or better than those used during original design and installation. This modification meets the requirements of NRC Branch Technical Position 9.5.'1 for the power supplies to the motor driven fire pumps. Thia modification has been deemed to not involve an unreviewed safety question.
79-03-123 EL-Isolation Modification of Safety Related Components for Fire Protection This modification was made in response to NRC Fire Protection positions. These positions, basically, required modifications which'will assure a safe shutdown capability in case of fire dama'ge to cables in the Switchgear Room or the Cable
. Spreading Room in the Control Building.
The modifications consisted of fire retardant barriers installed between cable trays, isolating switches to isolate the control circuits of Diesel Generator No. 31 and the control circuits for feeder breakers to 480 volt buses 2A and 3A and the tie breaker between the two busses.
The conduit and cable' installations observed separation criteria thus maintaining electrical-integrity. The installation of the Fire Retardant Barriels takes into consideration the seismic criteria of the cable trays. The modification was designed and installed using standards equal to or better than those used during original installation. The modification has been therefore deemed -to not involve an unreviewed safety question.
t I
1
-t 5-ofl17 I
~ - -
t 79-3-132'SGBD-Steam Generator ~ Blowdown Sampling System Modification This modification' involved the installation of a recirculation path to route steam generator blowdown sample flow to the Steam Generator Blowdown-Tank.
~
- Efforts.to minimize waste effluent determined'that a. substantial reduction in
, liquid effluent would -be ' achieved by routing sample flow sto the Steam generator
- Blowdown Tank in111eu'of the sample sink drain. This recirculation path includes four. additional sample heat exchangers which were installed in series-with the existing sample heat exchangers, the piping, and valves necessary to properly
- route.the effluent to;the Steam Generator Blowdown Wnk.
This modification was designed ~ and installed using standards equal to or better than those used during. original construction. This modification has therefore been deemed to not involve an unreviewed safety question.
80-3-)18 FP-Fire Alarm and Detection System The purpose of this modification was to do the following:
1.
To disconnect the diesel generator building fire protection water flow alarms and the fire valve tamper alarms from the existing control and alarm circuits and permanently rewire and reconnect them to the appropriate alarm zone points on the Fire Display Control Panel (FDCP) in the control roora.
2.
To disconnect the three diesel generator building thermostat alarms from the existing alarm circuit and permanently rewire and reconnect them to 3-alarm zone points in the Fire Display Control Panel "FDCP" in the control room.
This modification consisted of (1) disconnection of 2-valve tamper switches, 1-water flow switches, and 7 thermostats for each of the three diesels; (2) Reconnection of the wires of the 6-tamper switches and 3-water flow switches to the appropriate zone alarms of the Fire Display Control Panel (FDCP) in the control room; and (3) reconnection of the wires of the 21-thermostats to zone alarms #21, #25 and #29 in the Fire Display Control Panel ("FDCP") in the control room.
This modification complies with NRC Branch Technical Position 9.5.1 and the American Nuclear Insurers (ANT) have concurred with this change.
This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation. This modification was installed in accordance with standards equal to or better than those used during original installation ~. This modification has been therefore deemed to not involve an unreviewed safety question.
a i
6 of 17
m_
F 80-3-019 FP-Fire Alarm and Detection ' System The purpose of this modification was to reconnect the electrical connections for Deluge Control Circuits from their existing Deluge Valve Control' Panels (DVCP) to the appropriate zones on the Fire Display Control Panel (FDCP) in the Control Room.
The above changes are required to consolidate all fire protection controls and monitoring on the new FDCP.
This modification consisted of reconnecting the following equipment:
(1) Existing Pressure Alarm Switches, Deluge Valve Solenoid Valves, tamper switches on gate valves and auxiliary relays for the Deluge Valve Control Panels for the main-and auxiliary transformers.
(2)
Existing Pressure Alarm Switches, Deluge Valve Solenoid Valves, tamper switches on gate valves and a new auxiliary relay for each of the Deluge Valve Control Panels for Boiler Feed Pump Oil Console, Hydrogen Seal Oil Unit, and Lube Oil Reservoir and Storage Tank.
(3)
Existing tamper switches on gate valves, existing auxiliary relay box for the DVCP for the Deluge Valve Contro) Valves in the upper and lower electrical tunnels.
80-3-020 FP Pire Control and Alarm System CO2 System Electrical and Instrumentation Installation The purpose of this modification was to complete the physical and mechanical installation of all electrical and instrumentation items and complete external wiring for equipment for the CO2 System and fire dampers associated with the CO2 system.
This modification consisted of mechanical and physical installation of the electrical equipment for the CO2 system and fire dampers in the following areas:
1.
3 Diesel Generator Rooms 2.
Control Building (El.15'0")
3.
Cable Spreading Room in Control Building (El. 33'0")
4.
Battery Rooms 7 of 17
4 The portions of the CO2 system that have not been completed are for the following -
equipment:
1.
Turbine Generator 2.
Generator Purge System 3.
Main Boiler Feed Pump #31 4.
Main Boiler Feed Pump #32 This modification complies with NRC Branch Technical Position 9.5.1 and the American Nuclear Insurers (ANI) have concurred with this change.
This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation. This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unraviewed safety question.
80-3-032 WDS-G-Vent Header Loop Seal Drain The purpose of this modification was to remove the liquid waste buildup from the gaseous waste system vent header.
This modification involved the installation of two (2) drain valves into the low points of the unintentional loop seals, and draining to individual drain tanks.
These valves will be left in the open position, and any moisture will drain by gravity into the drain tanks. The drain tank aill have a level gauge to determine tank level. Periodically, an operator will open the drain valves on the tanks and drain to the floor drain, which leads to the waste disposal system.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was installed in accordance with standards equal to or better than those used during original installation.
This modification has aeen therefore deemed to not involve an unreviewed safety question.
80-3-034 AFW-D.C. Powered Lighting for No. 32 Auxiliary Boiler Feed Pump The purpose of this modification was to provide 3.C. powered lighting at the steam driven No. 32 auxiliary boiler feed pump as directed by the Nuclear Regulatory Commission Confirmatory Order dated February 11, 1980.
There are three D.C. powered lighting fixtures in the Auxiiiary Boiler Feed Pump area to provide lighting to enable manual operation of No. 32 Auxiliary Boiler Feed Pump during loss of power incident. Two fixtures are in the pump area and one outside the north door. Each lighting fixture is an emergency battery operated lighting unit. Each unit consists of a battery charger, two 25W sealed beam heads and Exide B-2-ll5 12 volt battery. The battery charger operates from a 120 volt AC source. The beams can be aimed manually in any direction.
The units are located approximately 8 ft. above the finished (18'-6" El.) floor.
This modification was designed considering seismic factors. This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not -involve an unreviewed safety question.
8 of 17 u
I' 80-3-037 FP-UpgrJding of rxirting Electrical Pene.tration Firn Stops The purpose of this modification was to install new fire seals which pass ASTM-E-119 3-hour burn tests in the existing electrical conduit penetrations.
Penetrations that were up-graded under this modification were limited to those in the walls separating the Control Building from the Turbine Building and the Diesel Generator Building and in the walls between the Diesel Generator Rooms.
Seals in the existing penetrations consisted of several different materials and designs. The work performed involved complete removal of all existing penetration material from the seals and installation of 3-hour rated fire seal system. The general configurations of the 3-hour fire stops consists of ten (10) inches of silicone foam plus high purity alumina damming material. A one (1) inch thick dam was used on each side cf the wall penetration seals.
The annular space between the cable and the inner surface of a conduit was filled with silicone foam up to 9-1/2" deep, with 1" damming on either side, there are 2 such fire stops per each conduit run passing through a wall penetration.
These fire stops will prevent propogation of cable fire from one side of the wall to the other side of the wall.
This modification helps to satisfy the two major concerns regarding the overall fire protection scope.
I 1.
To comply with NRC Branch Technical Position 9.5.1.
2.
To comply with requests of American Nuclear Insurers (ANI)
This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-041 FP Diesel Generator and Control Building Door Alarms This modification consisted of adding 12 magnetic type door switches to various doors in the Diesel Generator Building and Control Building.
The opening of a door will annunciate in the control room on the Fire Display Control Panel (FCCP) to alert the operator of a breach of a fire door.
This modification helps to satisfy the two major concerns regarding the overall fire protection scope, 1.
To comply with NRC Branch Technical Position 9.5.1.
j 2.
To comply with requests of American Nuclear Insurers (ANI) l This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation. This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-049-ESS Automate Containment Isolation Valves and Install Phase A Signal to Containment Isolation Valves In accordance with the automatic containment isolation concerns of NUREG-0578 this modification automated two existing manual containment isolation valves and added a phase "A" isolation signal to these as well as two other autortatic containment isolation valves.
Two existing manual valves, 550 and 1610, were replaced with air operated valves and provided with a phase "A" isolation signal. Two existing air operated valves, 863 and 959, were provided with a phase "A"
isolation signal.
9 of 17
Thia modification was d::ign:d con 2idering original srpnration crit ria thus maintaining the integrity of electrical separation. This modification was installed in accordance with standards equal to or better than those used during original installation.
This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-058-ESS Modify Power Supply to Valvea 548 and 1788 The purpose of this modification was to revise the power supplies to Valves 548 and 1788 to meet existing plant design criteria.
This modification consisted of the installation of relays, cabinets, and wiring in order that Valves 548 and 1788 will be electrically powered from different power supplies other than those for valves 549 and 1789.
The power supply for Valves 548 and 1788, ill be 125 V.D.C. from distribution panel 32, circuit No.
- 20. Connections and terminations of specific wiring were done inside the Gas Analyzer Cabinet, "SN" Panel and "SO" panel. A relay cabinet was mounted along side the Gas Analyzer Cabinet whe o the relays for 548 and 1788 were located.
The relays will be energized or d' energized through a terminal block also mounted in the relay cabinet.
This terminal block is located in the Gas Analyzer Cabinet.
This modification was designed considering original separation and single failure criteria thus maintaining the integrity of electrical separation.
This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-072 FP-Fire Wall Partition Separating the Cable Spreading Room from the Electrical Tunnels This modification involved the installation of a three (3) hour fire rated seimnic Class I partitioning wall installed on the 33 ft. elevation of the Control Building. This wall sepa: stes two (2) distinct fire zones, the Cable Spreading Room and the Cable Tunnels.
The installation of this partitioning wall satisfies two (2) requirements:
1.
Provide 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated barrier between two (2) distinct fire zones (Cable Spreading Room from Cable Tunnels).
2.
In the event of activation of the CO2 System in the Cable Spreading Room, the wall will keep the CO2 contained. Unlike the Cable Spreading Room which has a CO2 Fire i totection System, the Cable Tunncis are protected via a Dry Pipe Pre-Action Sprinkler System.
Without this partition, the CO2 System might be ineffective in the cable spreading room.
The wall consists of 3/16" Steel Plate bolted on both sides of a steel channel ficme with a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated door providing access to/from the cable tunnels.
The wall is designed to Seismic I (DBE) structural requirements. Addition of the wall will not impair or affect the Control Building in any way from a structural standpoint. The steel is coated externally on both sides with a minimum of 2 inches of UL approved sprayed fire proofing material in accordance with 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> column rated design.
10 of 17
In addition, tha cable pen;trction through tha wall is fillrd with ct lecst six (6) inches of silicone elastomer. To insure that the door is in the closed position, it shall be electrically monitered from the Fire Display and Control Panel (FDCP) in the Control Ewom.
This modification helps to satisfy the two major concerns regarding the overall fire protection scope.
1.
To comply with NRC Branch Technical Position 9.5 t.
2.
To comply with requests of American Nuclear Insurers (ANI)
This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-081 WDS-L-Liquid Radwaste-Interconnections for New Facility Thin modification involved a core bore in reinforced concrete and the installation of piping, tee's and isolation valves in the existing liquid rcdwaste holdup inf'suent lines et the point of interconnection with the piping to the proposed new holdup facility.
The Waste Holdup Tank No. 31 is fed from three main sources of liquid radwaste.
Of these three, two will be interconnected with the proposed new Waste Holdup Tanks to be added at a later date.
In order to minimize impact on construction of the new facility and on plant operation, the interconnections were made as a separate modification. The inter-connections have included isolation valves and pipe caps to maintain the pressure boundary of the existing system.
The isolation valve installed in each branch connection will be maintained closed and the pipe cap will remain welded in place until completion of the new holdup facility, when these branch lines will be connected to the new Waste Holdup Tanks.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering seismic factors. This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety guestion.
80-3-086FP-Upgrade Control Room Door to Three Hour Fire Rating This modification involved the revisions which have been made to the Control Room door to upgrade it to a three hour equivalent fire rating. The present door is composed of bullet resistant steel and has a bullet resistant glass window.
A plate is located above the window and connected to a hinge by means of a fusible link.
In event of a fire, the temperature outside the Control Room door 0
would melt the fusible link at 160 F and the plate would fall into a channel covering the glass. This would provide protection from fire in the turbine building. This plate will give the door an equivalent fire rating of three hours a..d thus satisfy the fire protection requirements for the Control Room door.
11 of 17 b
q
~
C' y[
~
- This modification; complies witih NRC Branch Tebhnical Position 9.5'lJand the.
. American Nuclears Insurers (ANI) 'have concurred with this change.
This modificatior was installed in accordance with standards equal ~to or better-than those used during original installation.
This modification has been therefore deemed to not -involve an unreviewed safety question.
'80-3-087H2-Hydrogen Tube Storage Relocation Interconnection This modification involved the installation of piping, fittings and valves for an interconnection in the hydrogen gas system 'to facilitate relocation of the
~
hydrogen tube storage facility.
. Three valves were installed in the hydrogen supply line from the existing hydrogen.
I storage facility. This ' configuration provides a branch connection to allow
' future relocation of the hydrogen storage facility without. requiring a plant outage for the piping tie-in.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was installed in accordance with standards equal to or better than those used during original installation.
This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-091 WDS-L-Reactor Pit Water Level Alarm This modification consisted of the installation of a level switch at the Reactor Pit to provide two water level alarms at the Control Room.
The level switch was installed on the west side of the reactor pit. The mounting brackets, and supports associated with the switch were seismically designed.
Two ala.us are provided at the Central Control Room (CCR). The first alarm
~
point is activated when the water level reaches 2.5" and the second at 6".
Welding on this modificat.on was accomplished using appropriate plant specific procedures 'sased on applicable codes. This modification was designeo to seismic criteria. This modification was donigned considering original separation criteria thus maintaining the integrity of electrical separation.
This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-093 WDS-L-Containment Sump Overflow Indication This modification consisted of the installation of a solid state, capacitance sensing point-type (on/off) level sensor (Princo Inicruments, Inc. Model L-1540 Remote Point Level Control) at the Containment Sump to provide a water overflow alarm at the Central Control Room (CCR).
.12 ofl17
The overflow indication system consists of the following components:
1.
A sensing head and associated probe to be mounted at the edge of the Containment Sump.
2.
A remote control unit contained in a weatherproof enclosure.
3.
A 115 Volt AC Westinghouse isolation relay, type BFF31F.
4.
A 24 Volt AC Struthers-Dunn interposing, relay (DUNCO BlXBX11GlO) which was located on panel SBR' in the CCR.
The sensing head and the remote control unit are connected by a multiconductor cable. The remote control unit and the isolation relay were mounted inside a heavy duty NEMA 4 enclosure located just outside the crane wall.
Welding on this modification was accomplished using appropriate plant specific procedures based on applicable codes. This modification was designed considering seismic factors. This modification was designed considering original separation criteria thus maintaining the integrity of electrical separation.
This modification was installed in accordance with standards equal to or better than those used during original installation. This modification has been therefore deemed to not involve an unreviewed safety question.
80-3-101 WDS-L-Replacement of Reactor Cavity Sump Pumps The purpose of this modification was to replace the existing two Reactor Cavity Sump Pumps with updated model pumps.
The existing power and control cable from MCC #38 compartments SA and SB to exist.:ng local junction boxes were retained to serve the new pump motors. New cabic with the new pumps was installed from the motors to existing junction box.
The existing control stations were retained. New red and green status indicating lights were installed on the door of each sump pump MCC compartment.
A red indicating light " Pump #1 on" was installed on panel SAF at the control room.
Welding on this modification was accomplished using appropriate plant specific procedures based on appli;able codes. The modification was designed considering seismic factors. This r.odification was installed in accordance with standards equal to or better than those used during original installation. This modification.
has been therefore deemed to not involve an unreviewed safety question.
13 of 17
80-3-106 CB-Prevention of Damage Due to Water Leakage Inside Containment The_ purpose iME this evaluation is to summarize the results of an investigation of equipment, controls, and operating procedures and completed modifications, and provide basis for continued safe operation of IP3NPP in response to IE
'Information Notice No. 80-37 " Containment Cooler Leaks and Reactor Flooding at Indian Point Unit 2".
The containment sump under normal operation collects water from various drains within containment. This water is then pumped to the waste holdup tank when the sump level reaches the actuation level on the sump pump float switches. The sump pump flow meter measures instantaneous flow (an indication of proper pump performance) while the totalizer measures cumulative flow, which is used to indicate changes in sump accumulations.
When the level in the containment sump increases to the float switch actuation level, one sump pump will start.
If the level continues to increase, the second pump starts. There is constant containment sump level indication provided in the control Room via two independent level indicators.
Also, through use of the containment sump flow meter and totalizer, which is logged every shift, any increase in sump accumulation because of a leak would be detected.
If the containment sump-level should approach an overflow condition, either because the two pumps cannot keep up with the leak or due to failure of both pumps, an alarm will annunciate in the Control Room via an additional level indicator installed in the containment sump.
At this point water will overflow into the normally empty recirculation sump which will be indicated by the level sensors in the control room.
The containment fan cooler condensate drains from the cooling coils, are collected and flow into a vertical standpipe slotted Weir System. The flow rate from the fan units are measured, based on the water depth flowing over the Weir.
If the drainage rates for all five units are nearly the same, it can be concluded that their water is condensate from the containment atmosphere. A particular unit with a high drainage rate, with respect to the other units, is indicative of a possible leak in one of the cooling coils.
In series with each transmitter signal is an alarm for "C.B.
Fan Cooler Cond. High Level".
The affected unit can be determined by individually monitoring the drainage flow from each unit using a rotary selector switch.
Water collecting in the reactor vessel cump is pumped out via two sump pumps, each pump discharging into an individual check valve that joins a common header and discharges into the containment sump. The reactor vessel sump is normally kept dry.
Newly installed level alarms in the reactor vessel sump will annunciate in the Control Room if water should accumulate in this sump.
7.lso, an indicating light located in the control room will indicate when reactor vessel sump pump number 31 is running.
.14 of 17
-q p
(-; +. -) ;;
^; -
'y
-t
.-e a
g g
s
~
-s' r;g _
f l
r 1
\\
lThe containment sump and reactor $ vessel. sump both-have two pumps, therefore a
'~ double. failure is required to lose pumping capability..The containment sump-
~
pumps.are. actuated via twols'eparate independent-float switches.--Thus:the loss-
- of one float : switch will not'stop pump capability. : There are two independent
- level indicators plus a separate overflow level ~ indicator installed in the containment sump. The reactor vessel sump has-a' dual alarm level indicator.with i; '
an indicating light to inform the operator when.'No. 31 pump is running. With the > installation of: the multiple level indicator in conjunction.with. operator-training, equipment inspection, operation review, and modifications,-it'is
+
-deemed that reasonable and~ attentive: action has been taken to preclude ~an-event
like the one described'in IE Notice 80-37.from occurring.
~
It has been therefore deemed that the subject of this evaluation dces.not-involve:
4 -
an unreviawed safety question.
F i
f I
i r
i
?
s Y
4 t-
/
.s:
a
,151of 17
.;2 n__j
7; II.
Procedures There were no special procedures required to La reported during this period.
III. Test And Experiments There were no special tests or experiments required to be reported during this period.
i 16 of 17
f t
'BI Summary of Srfaty Ecludions concerning Organizational Changes
. The original ' headquarters structure for. management and technical support for Indian Point 3 Nuclear Power' Plant are contained in Section 12.1.1 of the Final Safety-Analysis Report (FSAR) and the Facility Operating License (FOL)
'Section 6.2 of Appendix A (Technical Specifications) and ' Section 5.6~ of Appendix B (Environmental Technical Specificatio';s). The FOL was amended as applicable
- to reflect' the current management structure changes as they occured.
t The following Safety Evaluations were performed to provide the basis for determining whether or not the changes involved an unreviewed safety ' question pursuant to 10 CFR 50.59:
1.
NSE-NYO-022, " Safety' Evaluation-Authority Management Title Change",
dated 3/26/80, relates to the promotion of Mr. J. D. Leonard, Resident Manager of.the J. A. FitzPatJick Nuclear Power Plant, to the position of Assistant Chief Engineer-Design and Analysis -in the New York Office.
2.
NSE-NYO-003, " Safety Evaluation - Authority Management Title Changes",
dated'4/4/80, relates to the'following title changes:
Paul J. Early from Assistant Chief Engineer-Projects to Vice-President and Assistant Chief Engineer-Projects Paul Koshiashian from Assistant Chief Engineer-Construction Supervisor to Vice-President and Assistant Chief Engineer-Project Services John D. Leonard from Assistant Chief Engineer-Design and Analysis to Vice-President and Assistant Chief Engineer-Design and Analysis.
3.
USE-NYO-004, " Safety Evaluation - Authority Management Title Change",
dated 4/25/80, relates to the change in title for Mr. Peter W. Lyon from' Manager-Nuclear Operations to Senior Vice President-Nuclear Generation.
4.
NSE-NYO-005, " Safety Evaluation - Authority Management Change",
dated 6/4/80, relates to.the promotion of Mr. J. P. Bayne, Resident Manager of Indian Point 3 to the position of Senior Vice President-Nuclear Generation.
5.
NSE-NYO-006, " Safety Evaluation -Authority Organization and Management Title Changes", dated 6/11/91, relates to changes in the organizational structure for management and technical support of the nuclear facilities.
The Licensing and Project Engineering Groups are shifted from the Engineering Department to the Nuclear Generation Division of the PcJer Operations Depat". ment.
The Director-Quality Assurance and the Safety Review Committee Chairman,will report to the Executive Vice President and Direct r,of Power Operations.
6.
NSE-NYO-007, " Safety-Evaluhtlon I uthority Organization and a,ted '9725/80, related to the creation of Management Title Changes", p' Procedures and Performance" a new department, entitled
, under an Executive Vice President reporting to the President and Chief Operating Officer.
7.
NSE-NYO-008, '" Safety Evsluation -~ Authority Organization and Management Title Changes",. dated 12/18/80, relates'to the creation of a new position, " Manager - Radiological. Health and Safety", reporting directly i.
, to the Senior Vice President-Nuclear Generation.
17 of 17 i
- -,