ML19341D487
| ML19341D487 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/27/1981 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML19341D486 | List: |
| References | |
| NUDOCS 8103050676 | |
| Download: ML19341D487 (38) | |
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.j MILLSTONE UNITS 1 AND 2 ANNUAL REPORT This Annual Report has been prepared pursuant to the requirements of Title 10, Code of Federal Regulations, Section 50.59 and Sections 6.9.1.4 and 6.9.1.5 of the units' Appendix A Technical Specifications.
Common site reporting requirements are addressed in this section. Common site facility changes and tests are administered under the control of only one unit, and their evaluations are provided in the section applicable to that unit assigned the responsibility. Common site procedure changes are addressed here.
No procedure changes common to both Units' FSAR were processed during 1980.
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f4ILLSTONE UtilT 1 C0llTENTS Page CHANGES Design Changes 1/1 - 1/6 Procedure Changes 1/7 TESTS 1/8 - 1/9 RADIATION EXPOSURE 1/10
I Page 1/1 PLANT DESIGN CHANGES The following list by design change number summarizes those design changes completed in 1980, relating to safety related equipment, which could have a potential impact on safety related systems, could potentially impact the S
environment or required a change to the FSAR.
PDCR l-50 Generator Core Monitor A generator core monitor was installed on the second level of the turbine building below the generator to continuously sample and record the generator hydrogen particulate concentration and alarm when the concentration reaches a predetermined level.
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PDCR 1-50 Control Roon Annunciators Control room annunciators have been supplied with separate breakers to enable individual isolation.
j PDCR l-54 Water Quality Monitor 1
j An alarm circuitry has been added to the existing circulating water environmental annunciator to alarm on intake to discharge differential temperature.
This change i
j will help prevent exceeding the National Pollution Discharge Elimination System (HPDES).
1 PDCR l-71 Chlorine Gas Detector A chlorine gas detector has been installed in the screen house chlorine room with an alarm to warn prior to any environmental danger.
P_DCR_ l-90 Station Service Transformer Fault Switch A keylock switch has been installed in the control room used to arm the trip portion of the normal transformer fault pressure relay circuit. The installation i
of this switch eliminates the need to lift a wire in the generator lockout relay l
trip circuit during times of normal plant operation.
The fault pressure relay is used only during shutdown periods when the main and normal station service trar.sformer are backfed.
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Page 1/2 PDCR l-11 Feedwater Nozzle Thermocuple Three (3) thermocuples were installed on thu outer circumference of each of four (4) feedwater nozzles, between the reactor pressure vessel and the nozzle to provide a monitor for any~ leakage should it occur.
4 PDCR l-32 Feedwater Motor Operated Block Valve Modify the motor operators on the block valves to fully open the valves within 56 seconds. This is part of the FWCI system modifications which on FWCI system initiation will initiate an opening signal to both block valves.
PDCR l-41 Keylock Switch Addition to Reactor Manual Control System (RMCS)
A two position keylock switch was added to panel 928 to provide the control room, supervision of the refuel interlocks associated with the service platform.
PDCR l-53 Chlorination System A main header stop valve and expansion tank in the chlorine supply header was installed to provide double valve isolation of Unit 1 chlorinator without effecting the Unit 2 chlorinator system.
PDCR l-60 Hotwell to Main Steam Line Isolation Valve A valve was installed between the hotwell and main steam line " start up eductor" and a valve between the hotwell and the start up eductor steam trap to allow maintenance of the eductor nozzle and main steam isolation valve drains regardless of condenser vacuum or hotwell level.
l PDCR l-71 Accelerometers "D" Service Water Pump Two accelerometers that were previously installed on the bottcin column flange of "D" service water pump to evaluate pump performance were removed.
PDCR l-91 Motor Operated Valve Position Indicator Indication lights were added in the control cabinet of the fire pump house to provide indication of which motor operated valve, A or B, is feeding the storage tank.
PDCR l-98 Radwaste Filters Spectacle flanges were installed on the filter side of the rinse and fill valves on the floor drain filter, fuel pool ' A' filter, and fuel pool filter
'B' to prevent back leakage of water into filters and to prevent pressurizing the mters and causing leakage to the air system through check valvas.
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PDCR 1-104 Delege System Pressure Switch The delege system pressure switch has been wired into the annunciator circuit and removed from the detector circuit to improve design criterior, i
2 PDCR l-110 Cable Vault Fire Detection i
The eutectic cable fire detection system was removed from the cable vault i
because it was no longer needed due to the installed and more sensitive ionization type fire detection system.
2 PDCR l-5 Core Spray Pipe Detection The sensing lines on the core spray break detection differential pressure indicating switches were reversed and alarm set point changed to insure that a core spray pipe break will produce a change in differential pressure of sufficient magnitude to reach the alarm set point.
PDCR l-6 Isolation Condenser Supply Lines i
Additional support members were installed to the isolation condenser supply I
lines at containment penetration X-10A to increase the safety factor to i
greater than 2.
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PDCR l-7 Drywell Air Bottle Support Modification The supports on the backup drywell air supply bottles were modified to meet new seizmic specifications.
PDCR l-9 Ventilation Duct I
The ventilation duct on third floor of the Reactor Building in the area of isolation condenser penetration X-10A was rerouted to permit additional isolation condenser support members.
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PDCR 1-13 Core Spray Hanger Core spray hanger CSH-24 was modified to increase the safety factor of base plate bosting to greater than 4 as required by I&E Bulletin 79-02.
PDCR l-14 Core Spray Fill Connection r
A fill connection to core spray ' A' line between CS-4A and CS-5A was installed to quench steam and preclude future waterhammering on the system.
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PDCR l-15 Core Spray Hanger Core spray hanger CSH-18 was modified to an improved design in conjunction with system tepairs.
PDCR l-21 Domestic Water to House Heating Boiler A 2 inch gate valve with sweated ends was installed in the domestic water piping to house-heating boiler surge tank to provide a two valve isolation between house-heating boiler and domestic water supply thereby preventing the contam-ination of domestic water supply.
PDCR l-22 Isolation Concenser Initiation Circuitry The isolation condenser initiation circuity was modified to provide automatic initiation on reactor low low water level.
PDCR 1-25 Containment Atmosphere Control The drywell pressure low alarm originally installed to indicate drywell to i
torus Dp less than 0.9 lbs. was replaced with a more accurate drywell to torus Dp recorder on control room panel 903. The alarm point is set at 1 lb. which will comply wf +.h the technical specification of less than or equal to 1 lb.
PDCR 26 Primary System Oxygen Recorder A continuous oxygen monitor and recorder has been installed in the primary system continuous flow sample line. This will allow continuous monitoring of oxygen concentrations in the reactor water during all phases of plant operations.
PDCR 1-34 Load Center Transformer Space Heaters i
Space heaters have been installed in the load centers 12C,120,12E,12F transformers so that wnen the load side breaker is open the space heaters 4
will be energized and will help eliminate moisture.
PDCR l-47 Drywell Floor Drain Sump The drywell floor drain sump switch was changed to permit the operator to start the sura pump in the auto select move with shut off occurring via float control.
l PDCR 1-50 Scram Discharge Volumn Vent Reroute The north scram discharge volumn vent line was removed from the drain line and re-routed to a scupper in T.I.P. drive mechanism area.
This will insure a straight path to the atmosphere for the vent piping as already exists on the south vent system.
i Paga 1/5 PDCR l-54 Station Air Supply to Cleanup System A tell-tale drain was installed on the station air supply to clean up demin system to insure against cross-contamination of the station air.
PDCR 1-55 Reactor Manual Control System Replaced the-mechanical (RMC) Reactor Manual Control System timer with an electronic solid state timer.
PDCR l-56 Reactor Recirculation M.G. Speed Control The el 1 high limit switch in the recirc M.G. scoop tube positioners was jumps.
because with recirculation M.G. speed at the electrical limit the
_wo switch will not respond properly caused by vibration changeing set points.
PDCR l-72 Emergency Service Water Crose-Tie Piping An orifice plate was installed in the emergency service water cross-tie piping from cooling service water system to reactor building heat exchangers to reduce excessive flows and vibration at the cross-tie system.
PDCR l-82 Turbine Closed Cooling Water Vent Vents were installed at the high points of the piping on the service water outlet side of the turbine building closed cooling water heat exchangers.
l PDCR l-100 Service Water Rad Monitor A return line isolation valve was installed in the reactor building service water rad monitor piping to abilitate isolating the monitor from both ends.
PDCR l-104 Secondary Cooling Water Systems Isolation valves were installed in the supply and return lines of the secondary cooling system for isolation of the clean-up recirculation pump space cooler and reactor building sample coolers.
PDCR l-107 Drywell Head Studs The drywell head studs were machined to a hex so that a same size socket could be used to remove the lower nut, which will assist maintenance personnel in turning the stud and removing the' lower nut.
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Page 1/6 PDCR 1-116 Condensate Regeneration System Eductor An eductor for transfering resin from the shipping container to the cation resin tank was installed in the turbine building outside the regeneration room.
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l PROCEDURE CHANGES The following summarizes safety analyses for procedure changes as listed in the FSAR in accordance with the provisions of title 10, Code of Federal Regulations, Section 50.59.
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Procedure Change OP307 These two procedures were changed 412P to incorporate the design change to modify the isolation condenser initiation logic providing automatic initiation on reactor low low water
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i Page 1/8 TESTS i
The following list by test number smanarizes those tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.
None of the tests were evaluated as unreviewed safety questions.
J SP80-1-2 i
This procedure provided instructions for testing the anchor bolts on isolation condenser penetration X-10 restraint.
l SP80-1-3 i
i Drywell instrument air leakage test. - This test was performed to find the 1
amount of air leakage in the drywell instrument air header and check valve 1-1A-124. The data found was used to determine the amount of time air 1
would be available from the bottle system assuming a loss of the instrument j
air system upstream of check valve 1-IA-124, 1
i SP80-1-9 a
Hydrostatic test of Core Spray Check Valve - A hydrostatic test of the core spray keep fill check valve assemblies was performed to verify no leakage in p; ping, welds or flanges.
1 SP80-1-11 i
Generator core monitor pre-operation test procedure - This procedure j
established a method for pre-operational purging and startup of the Generator Core Monitor.
I SP80-1-12 1
This pr nedure provided instructions for testing the recently installed fire detection system.
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SP80-1-13 Isolation condenser actuation instrument functional test - This test was 3
performed to check the operability of the reactor vessel pressure switches l
263-53A, 263-53B, 263-53C, 263-53D.
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i SP80-1-16 Hydrostatic Test of Core Spray Loops A & B - This test was performed to hydrostatically test all parts of the reactor core spray loops A & B between valves 1-CS-44 and 1-CS-7A for loop A and 1-CS-78 for loop B to 1100 psi (+25-0) at ambient temperature.
SP80-1-20 Scram Discharge Volumn Vent Valve Test - This test was performed to determine that the scram discharge volumn vent valves are operable and that the vent l
system is free of any obstruction.
T80-1-01
. Performance of I&E Bulletin - This test was performed in accordance with I&E Bulletin 80-17 items 2 and 3 related to the Control Rod Drive Hydraulic System performance.
Both a manual and an automatic scram were performed per this procedure under the guidelines of I&E Bulletin 80-17.
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REGULATORY GUIDE 1.16 REPOR7 FOR 1980 CATE: 2/11/81 NORTHEAST NUCLEAR Er.ERGY CO. UNIT 1 NUtt*.,ER OF PERSONNELt >100 t'RE;11 TOTAL MAN-REM WORK & JOB FUNCTION STATION UTILITY C1HER STATION UTILITY OTHER EMPLOYEES EMPLOYEES EMPLCYEES EMPL0iEES EMPLOYEES EMPLOYEES REACTOR OPERATIONS 8 SURVEILLANCE MAINTENANCE PEPS 0tNEL 11 0
4 9.49 0.31 1.16 OPERATING PERSONNEL 45 0
0 45.18 0.C0 0.25 HEALTH PHYSICS PERSONNEL 16 4
55 15.11 1.46 21.50 SUPERVISORY PERSCtNEL 2
0 0
0.58 0.00 0.00 ENGINEERING PERSot#4EL 2
0 1
0.66 0.09 0.31 l
ROUTINE MAINTENANCE MAINTENANCE PERSCtN!L 16 0
0 4.13 0.05 0.28 OPERATING PERSONNEL 0
0 0
0.22 0.CG 0.01 i
HEALTH PHYSICS PER50t#4EL 1
0 0
0.27 0.00 0.23 SUPERVISORY PERSOtNEL 0
0 0
0.01 0.00 0.C0 ENGINEERING PERSCtNEL 0
0 2
0.0S 0.06 0.5S INSERVICE INSPECTICH itAINTENANCE PERSONNEL 3
3 162 1.18 0.75 91.60 l
OPERATING PERSONNEL 0
0 5
0.e5 0.05 1.55 i
HEALTH PHYSICS PERSOtNEL 0
1 26 0.35 0.44 7.63
$UPERVISCRY PERS0tNEL 1
0 3
0.39 0.C0 1.02 ENGIt4EERING PERSONNEL 4
5 51 2.63 2.18 3L69 SPECIAL MAINTENANCE MAINTENANCE PERS0tNEL 47 100 1544 45.19 46.53 1266.76 CPERATING PERS0ta.EL 31 1
69 9.E7 0.30 45.11 HEALTH PHYSICS PERSCtf4EL 9
4 155 4.44 2.51 93.28 SUPERVISORY PERSCNNEL 4
0 41 1.53 0.00 16.69 ENGINEERIN3 PERSONNEL 15 32 163 6.44 11.56 128.8S WASTE PP0 CESSING i
MAINTENANCE PERSotNEt.
0 1
5 0.40 0.23 2.56 OPF. RATING PER50tNEL 47 0
25 23.41 0.00 11.40 HEALTH PHYSICS PERSONNEL 9
0 4
3.6S 0.01 1.60 SUPERVISORY PERSotNEL 1
0 0
0.27 0.00 0.00 ENGINEERING PERS0tNEL 1
0 3
0.46 0.01 6.11 REFUELING MAINTENANCE PER$0tNEL 21 16 72 12.7G 4.94 32.74 OPERATING PER 01NEL 30 0
1 16.15 0.02 0.46 HEALTH PHYSICS PER;actNEL 6
1 23 1.64 0.24 E.70 SUPERVISORY PERS0tNEL 0
0 0
0.13 0.00 0.09 ENGINEERItG PER$0tNEL 6
4 31 3.02 1.80 15.87 4'
TOTAL MAINTENANCE PERS0tt'EL 98 120 1787 73.10 52.57 1395.33 y
OPERATING PERSONNEL 153 1
100 95.52 0.40 59.23 HEALTH PHYSICS PERS0tNEL 41 10 263 15.67 4.67 133.36 SUPERVISORY PERSONNEL 8
0 44 2.67 0.00 19.61 ENGINEERING PERSONNEL 28 41 251 13.50 15.71 163.44 GRAND TOTAL 328 172 2445 210.70 73.37 1791.17 TOTAL PERSONNEL EXPOSURE (SITE) 347.94 17o.72 3543.72
DECULATORY GUIDE 1.16 REPORT POR 1980 DATE: 2/11/01 NORTHEAST NUCLEAR ENERGY CO. UNIT 1 WORK 6 JOB FUNCTION TOTAL MAN-REM - ADJUSTED l
Uf1LITY OTHER EMPLOYEES EMPLOYEES j
I REACTOR OPERATIONS 6 SURVEILIANCE l
0.02 1.80 MAINTENANCE PERSONNEL 0.00 0.39 OPERATING PERSONNEL 3.03 J3.30 HEALTH PHYSICS PERSONNEL 0.00 0.00 SUPERY1SORY PERSOFNFL 0.19 0.48 ENGINEERING PERSONNEL ROUTINE MAINTENANCE 0.10 0.43 MAINTENANCE FIRSONNEL 0.00 0.02 OPERATING PERSONNEL 0.00 0.36 WEALTH PHYSICS PERSONNEL 0.00 0.00 SLTERVISORY PERSONNEL 0.12 0.90 ENGINEERING Peas 0NNEL INSERVICE INSPECT 10N MAINTENANCE PERSONNEL 1.56 142.20 0.17 3.07 OPERATING PERSONNEL 0.91 12.13 HEALTH PHYSICS PERSONNEL 0.00 1.58 SUPERY1SORY PERSONNEL 4.53 49.09 ENGINEERING PERSONNEL 5PECIAL MAINTENANCE 96.65 1962.21 MAINTENANCE PERSONNEL 0.62 69.88 OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL 5.21 144.49 0.00 28.95 SUPERVISORY PERSONNEL 24.01 199.64 ENGINEERING PERSONNEL WASTE PROCESSING C.58 4.00 MAINTDANCE PEfdONNEL 0.00 17.66 OPERATING PERSONNEL 0.02 2.79 J
HEALTH PHYSICS PERSONNEL 0.00 0.00 SLTERV150RY PERSONNEL ENGINEERING PERSONNEL 0.02 9.46 REFUEt1NG MAlh"TENANCE PERSONNEL 10.26 30.73 0.04 0.71 OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL 0.50 13.48 SUPRRV1%RY PERSONNEL 0.00 0.14 ENGINEERING PERSONNEL 3.74 24.58 TOTAL MAINTENANCE PERSONNEL 109.19 2161.37 0.83 91.75 OPERATING PERSONNEL 9.70 206.57 HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL 0.00 30.69 32.63 284.15 ENGINEERING PERSONNEL 152.40 2774.52 CRAND TOTAL TOTAL PERSONNEL EXPOSURE (SITE) 176.72 3540.72 i
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MILLSTONE UNIT 2 CONTENTS Page CHANGES DESIGN CHANGES 2/1 - 2/7 PROCEDURE CHANGES 2/8 TESTS 2/9 - 2/14 STEAM GENERATOR TUBE ISI 2/15 - 2/19 RADIATION EXPOSURE 2/20
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PLANT DESIGN CHANGE
-The following list by design change number summarizes those i
design changes completed in 1980, relating to safety related equipment, which could have a potential impact on safety related systems, could potentially impact the environment, or j
required a change to the FSAR.
PDCR 2-150-76 l
Installation of an alternate power supply for the non-vital 120 VAC panels, IAC-1 and IAC-2, to enhance plant reliability.
Since the alternate power sources are vital buses, the i
components interfacing between vital and non-vital systems were procured and installed as safety grade devices.
Thus, no malfunctions can occur which will adversely impact the correct j
functioning of the salaty system to which these components are connected.
PDCR 2 95-76 e
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Installation of equipment necessary to fill and drain the cask laydown area of the Spent Fuel Pool.
The components added were seismically supported, and the integrity of the laydown area i
wall and liner was maintained.
PDCR 2-163-78 Replaced the heat sensitive fire detectors with ionization 7
smoke sensitive type in the DC Switchgear Rooms.
This I
change required a change to the Technical Specifications and did not reduce the margin of safety as defined therein.
PDCR 2-57-79 Replaced the Service Water Pump graphite filled bearings with cutlass rubber bearings.
The change met or exceeded the design i
standards of the original bearings.
PDCR 2-125-79 and 2-22-80 j
Replaced the A and B Charging Pumps with those from a different j
manufacture
- because of failures to the old style pump cylinder i
blocks.
The change improved the integrity and operability of the charging system.
It required a data change to the FSAR.
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PDCR 2-145-79 Provided RCS temperature indication from the Reactor Protection System into the plant process computer.
The change resulted in a more accurate temperature input into the calculations for RCS flow, leakage, core heat balance, etc.
The change did not affect the safety functions of the temperature loops.
PDCR 2-171-79, 2-183-79 and 2-5-80 Installation of an acoustic valve monitoring system.
The changes provide position indication and alarm anntnciation of the Power Operated Relief and RCS Code Safety Valves as required by NUREG 0578.
The modification did not affect the operation of any system related system.
PDCR 2-174-79, 2-182-79, 2-186-79 and 2-140-80 Installation and modification of equipment necessary to automatically initiate auxiliary feedwater as required by NUREG 0578.
The changes increase the plant's operating margin of safety by making auxiliary feed available in the optimum time and do not affect the ability of any system to perform its safety related function.
PDCR 2-175-79 Modification to a.: RBCCW. system support (Hanger 305691) to increase the factor of safety as required by IE Bulletin 79-02.
The change upgraded the hanger seismically; thus, improved the system integrity.
PDCR 2-180-79 Installation of piping drains in the Containment Spray System i
which allow draining of residual water prior to ISI testing various check valves.
The change conforms to the original system design standards, and the valves added are administratively locked closed.
PDCR 2-1-80 Installation of an RCS venting system which provides a means of removing non-condensibles from the reactor vessel head and pressurizer steam space as required by NUREG 0578.
The design I
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is bounded by the existing, LOCA analysis, and is consistent with the original design standards and NRC requirements.
PDCR 2-2-80 i
Temporary installation of a spool piece in place of a service water strainer while the strainer is being refurbished.
- Thus, a service water pump is available for use if necessary; even though it is not operable with respect to the Technical Specifications.
The design and installation is such that the service water system seismic qualification will not be affected if the pump and spool piece combination must be used.
PDCR 2-6-80 Installation of space heaters in the 480V load center transformers.
This change does not adversely affect the function or operability of any vital load centers and increases reliability by reducing transformer grounds.
PDCR 2-27-80 Modification to the service water piping from the ernergency diesel generator to allow replacement of the diesel heat exchanger head.
The change does not affect service water operability or the stress analysis of the involved piping.
PDCR 2-36-80 i
Installation of a bypass switch to defeat the zero power inhibit of CMI/PDIL functions.
This change allows surveillance testing of CMI and PDIL interlocks when reactor power is 'ess than E-4%.
Any change related failure is conservative in t?1t the CMI and PDIL interlocks would be active no matter what the 4
power.
PDCR 2-40-80 Installation of vent valves on the diesel generator lube oil strainers which enhances reliability by reducing trapped air in the lube oil system following maintenance.
The change is acceptable based on the original seismic analysis as modified to include the vent assembly.
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i PDCR 2-42-80 l
Removal of the containment isolation signal to the containment purge valves.
The change removed the CIAS in the ESAS cabinets and complied with the NRC recommendation that unused circuitry be removed.
The design implemented a change to_the Technical Specifications and in no way impaired the ESAS from performing its safety related function.
PDCR 2-49-80 and 2-50 80 Installation of a bypass and strainer arrangement in the service water supply to the emergency diesel generators.
The reliability of the diesel is enhanced by reducing the probability of mussel fouling.
The design met or exceeded the standaro3 used in the original design of the service water system.
PDCR 2-63-80 Generic modification to pipe supports in the feedwater system.
This change upgraded the seismic qualification of affected pipe hangers from a safety factor of 2 to safety factor of 4 as j
required by IE Bulletin 79-02.
PDCR 2-67-80 Modification to the refuel pool drain piping to allow the use of a portable filteration unit.
This change did not affect any 1
safety related system and did reduce the total activity processed by the liquid radwaste system.
PDCR 2-70-80 Modification to prevent automatic restarting of the containment sump pumps and reopening of containment sump pump isolation 2
valves following CIAS reset.
The change results in no change to the safety analysis assumptions nor does it create the possibility of a new accident.
PDCR 2-81-80 Replaced 1;mit switches and solenoids on various valve operators outside containment.
The change upgraded the environmental qualification of the safety related valve
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operators as seguired by IE Bulletin 79-01; and thus, enchanced their reliability.
I PDCR 2-92-80
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Replaced the low voltage containment electrical penetrations with those from a different manufacturer due to degrading insulation resistance between conductors of the old style modules.
The change enhanced the safety and reliability of the plant, and did not adversely affect any safety system.
PDCR 2-93-80 and 2-94-80 Replaced the steam cenerator level transmitters with those environmentally qualified for post-LOCA conditions.
- Thus, these changes increase the reliability of the steam generator level control and indication systems.
PDCR 2-97-80 Installation of fire barriers where facility 1 and facility 2 i
safety related cables were in close proximity to each other.
i This change provides improved cable tray isolation protection and has no other effect on safety related systems.
PDCR 2-101-80 i
Installation of a reactor coolant pump oil collection system.
Since all added components are seismically supported, this change does not affect any adjacent safety related equipment.
PDCR 2-110-80 Installation of four Foxboro Spec 200 control system cabinets in the Control Room to provide centralization of certain safety related instrumentation.
This change was designed in accordance with existing criteria for safety related equipment and enhances the reliability and testability of the instrumentation contained therein.
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l PDCR 2-119-80 i
Modification to the subcooled margin monitor to delete the RCS cold leg and increase the range of the hot leg temperature inputs.
This change does not affect any safety related equipment.
PDCR 2 125-80 Modifications necessary to upgrade the components for automatic initiation of auxiliary feedwater from control grade to safety grade.
The change results in a system conforming entirely to safety Category 1E c ecifications as required by NUREG 0578.
PDCR 2-132-80 Installation of temporary demineralizers in the aerated / clean i
liquid waste system.
This change supplements the permanently installed demineralizer to minimize the cumulative release of radioactive materials in liquid effluents.
4 PDCR 2-136 Modification to the ESAS undervoltage circuitry to electrically isolate the actuation cabinets.
This change corrected a wiring error that would have caused the loss of one 4160V vital bus to i
result in a total loss of normal power (LNP).
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implementation of the change did not increase the probability of an LNP nor does it change the sequence of events once an LNP has been initiated.
I PDCR 2-138-80 Installation of a hydrostatic test plug in the top ICI flange at location L-18.
The change was necessary because the incore detector at this location was damaged and could not be replaced. This change does not degrade the RCS pressure boundry; and thus, does not constitute a safety question.
PDCR 2-148-80 Modification of the c,issile shield to allow it to seat properly on the cable tray support steel.
The change does not affect the necessary structural integrity of the missile shield.
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Page 2/7 PDCR 2-153-80 l
Temporarily removed an RCS cold leg temperature input from the RPS core protection calcu'ation.
Since the temperature inputs are actioneered in *,he RPS and four reactor coolant pumps must be in service during power operation, this change will not affect the ability of the RPS to perform its safety function.
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Page 2/8 PROCEDURE CHANGES There were no procedure changes as listed in the FSAR during 1980.
Page 2/9 TESTS The following list by test number summarizes those test performed under the provisions of Title l0, Code of Federal Regulations, Section 50.59. A su mary of the safety analysis is included for each test. None of the tests were evaluated as an unrev % ed safety question.
T79-29 and T80-7 Charging Pump P18A and Charging Pump P18B Operational Checkout -
This test provided for a controlled run in period and functional checkout of the replacement charging pumps.
Acceptance criteria demonstrated that the pump performance met the assumptions in the safety analysis.
T80-1 Feedwater Venturi Calibration (Na-?4) - This test obtained data to verify the calibration of the feedwater flow venturis.
Sodium releases were accounted for using station discharge procedures and compiled with environmental specifications.
T80-2 Verification of Terry Turbine Governor Speed Setting - This test preset the governor speed control to the high speed stop and verified that when the turbine was started in this condition it would not trip on overspeed.
T80-3 New RPS Bistable Testing - This test proved compatability of new bistable trip units with the Reactor Protection System.
Only o'e RPS channel was utilized for the testing, and existing procedures were followed to insure RPS operability including restoration of the original bistable unit at test completion.
T80-5 Pressurizer Proportional Heater Functional Test - This test determined the equilibrium pressurizer pressure when only the proportional heaters were in operation. The test was performed with the plant in Mode 3 and did not result in any abnormal system operations.
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T80 s RM 8132 Plateout Test This test determined the sampling efficiency of the Unit 2 Stack process radiation monitor.
This monitor is not safety related and was operated in accordance with approved operating procedures.
i T80-8 Turbine Valve Loop Loss Test - This test determined the loss in high pressure turbine efficiency due to throttling across the #4 control valve.
It will conform to the Technical Specifi-cation exception specifically issued for this evolution.
T80-_9_
Motor Driven Auxiliary Feed Pump Flow Test - This test verified 4
the flow capability of the pumps to insure that they net design requirements.
The pumps were operated in accordance l
with approved operating procedures.
1 T80-10 and T80-12 ESF SIAS/CIAT/EBFAS/AEAS/SRAS/MSI and Purge Vr.ive Reset Test I
for Actuation Channel 1 and Channel 2 - Thece tests determined l
the ESF component esponse to enginee ed safeguards actuation signal (ESAS) reset.
All equipment 9as operated in accordance
'(
with the applicable plant operating procedures and as permitted by the Technical Specifications.
T80-11 Auto Initiation of the Auxiliary Feedwater System Verification -
This functionally tested the equipment installed to automatically initiate auxiliary feed. The test was performed out of the modes requiring steam generator and auxiliary feed system operability.
I T80-13 Calibration of Nicolet Model 446A Spectrum Analyzer - This test verified that the analyzer was properly calibrated for use in the ISI program. No credit is taken for this instrument in the safety analysis or Technical Specifications.
e-.
~
Page 2/11 s
T80-14 Sample Volume Development-Program and Operating Procedure This test was intended to determine the relative effectiveness of the electropolishing process and an on-line ultrasonics i
technique in eliminating radioactive corrosion product deposition in liquid monitor sample volumes.
The equipment used in the performance of this test was completely independent of all plant systems.
?
I T89-15 CEA Position Indication Verification - This test determined the necessary adjustments to the Reed Switch Power Supply to insure that the Metrascope/ Backup Scanner and computer pulse counting system track with each other.
The reactor was subcritical during this test, and the CEA's were operated in accordance with the Technical Specifications.
1 T80-16 Auxiliary Feed Pump Speed Control - This test tried a procedure for increasing speed stability of the Terry Turbine during ISI testing.
It satisfied and verified Technical Specification operability requirements for the turbine driven auxiliary feed pump.
T80-17 MSI Output Relay Test - This test was performed to verify proper actuation relay performance for main steam isolation following changeout of the relays.
It was performed out of mode requiring MSI operability.
T80-18 Functional Test of the ESAS Sequencer Assembly 6N93 - This test documented the operation of the assembly with regards to the generation of blocking signals to the SIAS actuation module. This was performed on spare equipment in the I&C shop.
._ =_
Page 2/12 T80-19 and T80-32 Diesel Generator 78 and 7A Service Water Bypass System Test -
This test functionally checked the diesel service water cooling following installation of the bypass system.
It was performed on the diesel prior to declaring it operable for its safety related purpose, and did not affect the operJbility of the opposite facility diesel.
I T80-20 j
Diesel Run in After Annual Surveillance Overhaul - This test t
insured that a controlled run in/ warm up was accomplished in accordance with the vendor's representative following overhaul.
Per'ormance of this test did no: Tffect operability of the other facility diesel.
i The following tests were performed in the course of trouble shooting an apparent violation of channel separation in the
- ESAS, They consisted of simulating / initiating undervoltage conditions on the 4160V emergency buses and monitoring the effects. This testing was performed in a mode that required caly one electrical facility and did not affect the LNP circuitry or. diesel generator of the operable facility.
T80 Bus 24D LNP Test 4
T80 RSST Undervoltage Trip Circuit Test t'
T80 Bus 24D LNP Circuit Test l
T80 Bus 24D LNP Simulation Test i
j
^
T80 Bus 2iC LNP Simulation Test i
i T80 Emergency Bus 24C and 24D LNP Simulation Test 4
T80-22 Purge Valve and CIAS Actuation Module Retest - This test verified proper ope. ition of ESAS CIAS and Containment Purge
Page 2/13 l
l l
Valve circuitry following modification.
The equipment was i
operated in accordance with applicable operating procedures and the Technical Specifications, c
i T80-25 Dual Range Temperature Transmitter Testing - This functionally tested the RCS hot leg temperature transmitter, TT-lllX and TT-121X, to verify operability after installation.
No credit is taken for these temperature inputs in the safety analysis or in the basis of the Technical Specifications.
T80-28 a
}
Steam Dump Valve Testing - This functionally tested all components of the Condender Steam Dump Valve system.
The equipment is not safety related, and the test was performed j
during cold shutdown.
3 T80-29 and T80-35 RCS Head Vent Hydrostatic Test (Pressurizer Piping and Remainder of the System) - These tests verified ASME Code acceptability of the RCS head vent installation. They were performed during cold shutdown and did not affect any other safety related piping.
?
4 T80-30 MSI Reset Test - This test determined the response of components when the MSI initiation signal is reset.
Components were operated in accordance with the applicable operating procedures and the Technical Specifications.
l T80-33 4
Criticality /LPPT Cycle 4 - This test obtained data to confirm the adequacy.of the safety and transient analysis.
It was performed in accordance with the appropriate Technical Specifi-i cations including special test exceptions.
l
-,. ~ _ -., -. _ ~ _. _. -
Page 2/14 T8_0-34 Power Ascension Test Cycle 4 - This test verified core nuclear characteristics in accordance with Technical Specification requirements and confirmed related safety analysis assumpticas.
The test was performed in accordance with approved operating procedures and Technical Specification limits including special test exceptions.
Page 2/15 l
STEAM GENERA 10R TUBE INSERVICE INSPECTION RESULTS This section provides a summary of the steam generator inservice inspection results for the steam generator tubes, in accordance with Technical Specification 4.4.5.1.5.b.
In response to the specific requests for data per Technical Specification 4.4.5.1.5.b:
1.
The number and extent of tubes inspected are included in the summary.
Inspection results for tube denting inspections, in addition to the inservice inspection requirements, are also summarized.
2.
There were no tubes identified with an indication of an imperfection affecting wall-thickness.
3.
The identification of and rea on for the tubes plugged is included in the inspection results.
GENERAL DISCUSSION i
An itservice eddy current inspection and dent assessment program was performed at Millstone Point Unit 2 in Steam Generator No.1 and multi-frequency eddy current testing in Steam Generator No. 2 during the period from August 22, 1980 through September 2, 1980. The inspection was performed by Combustion Engineering Power Systems Group, System Integrity Services Personnel.
The inspection was conducted in accordance with Combustion Engineering Test Procedure No. 00000-SIS-012, 00000-NLE-082-08, 00000-ESS-073-01 and 00000-ESS-070-02 and satisfied the requirements of the Nuclear Regulatory Commission Guide 1.83 Revision 1 (July 1975) and the Plant Technical Specifications as re/ised June 14, 1979.
Steam Generator No. 2 test program was conducted by use of an MIZ-12 multi-frequency eddy current testing system.
The MIZ-12 multi-frequency eddy current tester is a highly versatile and compact eddy current inspection system.
Utilizing a " time sharing" technique, the operator may test at up to four (4) frequencies simultaneously. The MIZ-12 contains two mixing circuits, giving the operator the ability to subtract out most undesirable data (i.e., noise, support signals, etc.). The CRT display allows the operator to view the mixer output as well as the test signals. Three instrument channels were used for flaw data and one was used for dent assessment.
Approximately 875 (or 10.3%) of the tubes were tested from the hot-leg side.
Page 2/16 I
l Steam Generator No.1 eddy current test inspection consisted of 400 KHz testing for the detection of tube wall anomalias and the assessment of tube denting with both data being takn simultaneously with one pass of the eddy current probe through the tube. An electronic device called a Peak Reader (DB-1) was used.
The output of the Peak Reader was displayed on one strip chart and the normal high gain defect detection data was recorded on the other strip chart.
Approximately 1228 (or 14.6%) of the tubes were tested from the hot leg side.
Dent data analysis was accomplished by comparison to dent standard signals for dents of known dimension for which 1 (one) volt corresponds to 1 (one) mil reduction in radius for an axisymmetric dent.
The defect data was reviewed for anomalous signale Such anomalies were evaluated by phase analysis techniques for determination of the cause of the signal.
RESULTS OF THE INSPECTION Steam Generator No.1 No tube flaws were observed in either the hot or cold side in any tube inspected.
The dent assessment results are sumarized in Table 1 for steam generator No.1 hot leg side, respectively.
The overall averages of dent size data for each elevation is presented for the 1979 and 1980 inspection results.
These results will be broken down into three categories:
Tube sheet indications, eggcrate indications and support plate indication.
Analysis of the data indicated that the dent progression in steam generator No. I has shown no significant denting increase when compared to the 1979 report.
The tube sheet (Elevation 0) dent like indications show no increase in frequency when comparing the 1979 and 1980 data.
The average dent size increased approximately 0.5 mils. These dent like signals typically occur just above the start of the tube expansion at the secondary side face of the tube sheet.
The 1980 eggcrate dent signals for the nine elevations on the hot side show a small increase in frequency (45.4% vs. 36.2%)
and no increase in average size when comparing the 1980 data with 1979 results.
The tube support plates, elevations No. 10 and No. 11 show no increases in average dent size over those reported in earlier inspections.
Historical data indicates essentially all the tubes in this region are dented.
Page 2/17 l
l The sludge pile on the tube sheet on the hot side was measured.
The maximum sludge height observed was 10.5" which is the same as in the previous inspection.
In addition to the typical magnetite sludge, this inspection indicated bands of high conductivity sludge which existed during the previous inspection.
There were no +.ubes plugged in this steam generator.
Steam Generator No. 2 Data analysis indicated no tube wall flaws in any tube tested.
Table II summarizes dent assessment results for Steam Generator No. 2 hot side. These results are broken down into tube sheet indications, eggerate indications and support plate indications.
The tube sheet (Elevation 0) dent like indications show no increase in frequency when comparing the 1979 and 1980 data.
The average dent size changes approximately 1.0 mil for the hot :ide (2.17 mils vs. 3.19).
These dent like signals typically occur just above the start of the tube expansion at the secondary side face of the tube sheet.
Eggcrate dent signals for the nine elevations on the hot side show no significant increase in overall frequency and size when comparing 1979 results with 1980 data.
The tube support plates, elevations No.10 and 11 have shown a small increase in average dent size (10.07 mils in 1980 vs.
8.67 mils in 1979). As in No. 1 Steam Generator essentially all tubes in this region are dented.
Sludge measurements on the hot side indicate a maximum of 8.5" as compared to 6.2" in 1979. This inspection also indicated bands of high conductivity sludge similar to those found in Steam Generator No.1.
This condition was also present during the previous inspection.
The following tube was plugged as a result of this inspection.
Line Row Reason T
DI~
Would not pass a 0.540" diameter probe at the No. 10 support plate.
m ___
TABLE I MILLSTONE POINT UNIT 2 STEAM GENERATOR NO. 1 HOT SIDE 1979 1980 N0.
AVG.
STO.
NO.
AVG.
STD.
ELEVATION TESTED DENTED DENT _
DEV._
TESTED _
EENTED DENT DEV.
TSO 2927 40.2 2.13 1.03 1228 31.7 2.66 1.36 t
EC1 2928 45.6 1.07
.56 1227 47.3 1.10
.64 EC2 2928 74.
1.09 51 1228 87.3 1.13
.59 EC3 2928 75.8 1.24
.62.
1228 88.2 1.28
.71 EC4 2928 52.5
.98
.49 1228' 65.8
.96
.45 ECS 2919 42.9 '
.95
.46 1228 59.0
.94 49 EC6 2911 13.9
.81
.39 1228 25.6
.78
.38 EC7 2903 9.64
.79
.35 1228 15.5
.87
.50 EC8 2745 2.44
.68
.37 1112 4.4
.74
.31 EC9 2394 0.37
.60
.09 1032 0.77 1.61 2.59 42 E
TSPL10 2067 100 7.86 2.40 94T 100 8.55 2.07 j
TSPLll 729 100 6.96 2.37 356 100 7.47 2.26 ECl-9 25584 36.2 1.06
'. 5 4 10739 45.4 1.06
.60 TSPL10-ll 2796 92.7 7.64 2.42 1303 100 8.25 2.18 r
TABLE II MILLSTONE POINT UNIT 2 STEAM GENERATOR NO. 2 HOT SIDE 1980 jpg NO. ~
AVG.
STD.
NO.
AVG.
STO.
ELEVATI0fi TESTE0 DENTED DENT DEV.
TESTQ DENTED _
DENT _
DEV.
TSO 2261 30.6 2.17 1.2" 875 29.2 3.19 1.98 EC1 2261 48.7 1.28 84 2875 48.1 1.43
.93 875 67.6 1.24
.70 EC2 2261 79.8 1.19
.64 EC3 2260 72.6 1.29
.66 875 61.2 1.23
.70 EC4 2260
'J 8. 5
.92
.49 875 29.3
.92
.47 ECS 2260 15.8
.99
.55 875 23.6 1.02
.61 EC6 2260 4.7
.83
.44 875 7.2
.77
.36 EC7 2260 5.7 1.17
.83 875 8.34 1.23
.88 g
EC8 2135
. 14
.67
.29 683
.73
.74
.19 h*
EC9 2044
.04
.50 0.00 575 1.04
.78
.27 TSPL'i0 2010 1G0 8.95 2.65 500 100 10.56 2.75 TSPLll 732 100 7.80 2.23 260 100 9.13
?.77 ECl-9 20001 28.8 1.19
.68 7333 29.2 1.20
.71
- TSPL10-ll 2742 91.1 8.67 2.60 760 99.4 10.07 2.84
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RECULATORY Ct!!DE 1.16 REP 02T POR 1980 DATE6 2/12/88 HORTHEAST bOCLEAR DERGY CO. UNIT 7 i
WORK & Joe PD CTION TOTAL MAN-REX. ADJUSTED UTf.'TY OTHER D0.,JYLES EMPLOTEES REACTOR OPERAT!9N5 6 StkVE1Lt.ANCE MAINTENANCE PEPOSNEEL 0.08 3.84 OPEMTING PER50LATL 0.04 1.60 HEALTH PHY11CS PLTSONNEL S.34 49.52 SUPERV150RY PER50W'EL 0.00 2.40 i
)
ENGINEERING PERSON:sFL 4.13 6.44 ROUTINE MAINTENANCE MAINTEN*.NCE PER$0'4NEL O.00 1.02 OPEU TING PER50NhEL 0.00 0.00 HEALTH PHYSICS PER3o n EL 0.00 0,00
' SUPER 7150RY PERSONNEL 0.00 0.30 ENCINEERING PER50'3EL 0.02 0.0J INSERV1CE !XSPECT101 MAlkTENANCE PERSO:NEL 0.00 11.94 OPERAi!NG PERSonEL 0.00 0.!4 HEALTH PM51C5 PERs0'aEL 0.02 0.25
$UPERVISORY PERETEL 0.00 0.20 ENGIEEERING PERLtWNEL 0.55 9.56 SPECIAL MAINTENANCE
+
MAIKTD ANCE PERSONNEL 1.89 476.5S OPERATING PERSONNEL 0.15 12.44
[
HEALTH PHYSIG *1RsoNNEL 1.68 14.45 StPERVISORY rrbC:::EL 0.00 3.$0 ENCIEEERING PiR30NNEL 8.62 71.16 l
WA$TE PROCES$1NC MA!%TENANCE PER$0nEL 0.83 2.46 0'EUTING PER50 0TL 0.00 3.83
. HEALTH PirYSICS PER50'3EL 0.00 1.13 SUPERVISORY PER30'.7EL 0.00 0.00 ENGINEERING PERSONNEL 0.00 0.02 RETrELING MAINTENANCE PERSO::hEL 0.58 74.55 OPERATING ?!#$0NNEL 0.00 6.27 HEALTH Pinslai PERwmEl 0.00 4.I8
$UPERVISORY PER00:3EL 0.00 0.00 ENGINEERING PERSO!CEL 0.69 8.21 TOTAL MAINTENetCE PERSON':EL 3.43
$70.39 OPERATING PER$0NNEL 0.2I 24.27 HEALTH PHY$1C5 PERSONNEL 7.06 69.36 SUPERVISORY PER$0NNEL 0.00 6.10 ENGINEERING PER$0!3EL 13.63 95.40 CRA*:D TOTAL 24.32 765.75 TOTAL PERSONNEL EXPOSURE ($1TE) 176.72 3540.72 P
e
$