ML19284B316
| ML19284B316 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1974 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19284B310 | List: |
| References | |
| FOIA-92-403 NUDOCS 8011100285 | |
| Download: ML19284B316 (59) | |
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,,s a-a7-35 YANKEE ATOMIC ELECTRIC COMPANY ROWE, MASSACIIUSETTS SEMIANNUAL OPERATING REPORT FOR TllE PERIOD JULY 1, 1974 TO DECEMBER 31, 1974 2331 8 oo/oo aa g
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YANKEE ROWE SEMIANNUAL REPORT 7-1-74 TO 12-31-74 1
OPERATION
SUMMARY
a b
GENERAL The Yankee Atomic Electric Company Plant at Rowe, Massachusetts completed its refueling shutdown and commenced power operation at 0641 on August 25, 1974.
The plant operated at power throughout most of the balance of the reporting period.
Brief shutdowns and load reductions required during this period are described in the Operations chronology.
A.
Changes in Facility Design 1.
The following changes requiring authorization from the Commission were made:
a.
PDCR-74-3, Improvement of the Reactor permissive Circuitry as Required by Change No. 110, was completed during the period. This change was performed to fulfill an NRC request (Change 112) that the reliability of the scram circuitry be improved.
This request was made subsequent to Change 110 which relocated the permissive relay pressure switches from the turbine first stage pressure sensing line to the No. 1 nozzle pressure sensing line. All five criteria listed by the NRC in their request were satisfied by this improve-ment.
This Change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety analysis report.
b.
EDCR-72-3, Diverse Initiation of the Safety Injection System WL Relays, as required by Proposed Change 111 was completed during the report period.
This change will provide a diverse safety injection system initiation using vapor container pressure as the diverse parameter. At a trip point of 5 psig, this signal will parallel the existing primary system pressure and pressurizer pressure signals for Safety Injection System initiation.
This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety analysis report.
c.
EDCR-7' -2, Post Accident Vapor Container Hydrogen Control System, was co..pleted during the report period.
This system was installed to implement Change 97.
The system works as follows:
Following a hypothetical loss-of-coolant accident, hydrogen gas may be generated in the vapor container due to radiolytic decomposition of water, from material corrosion and a metal-water reaction involving the core cladding.
The post-accident hydrogen control system, which consists of the atmosphere recirculation system and the hydrogen venting system, provides a means for mixing the vapor container atmosphere to
}/
prevent local pocketing of hydrogen as well as a means for sampling and venting the. vapor container to maintain the hydrogen coacentration at an acceptable level to prevent an ignition.
This change does not impose any unreviewed safety question in that e system will operate in the same manner as covered in the exist-ig safety analysis report.
d.
Change No. 115, Core XI refueling was completed during the report period.
Core XI is loaded with 36 new 4.0% enriched Zircaloy clad fuel assemblies; 36 Zircaloy clad fuel assemblies recycled from Core X, originally enriched to 4.0%, and having an average burnup of 11,300 MUD /MTU; and four (4) stainless steel clad fuel assemblies recycled from Cores IX and X, originally enriched to 4.94%, and having an average burnup of 19,600 MWD /MTU.
A complete and thorough reanalysis of several potential adverse anticipated transients and accidents was performed and reported to the NRC in Proposed Change No. 115.
These analyses show that in only the most unlikely accidents could clad damage occur and in no case would there be a release of fission products that would endanger the health and safety of the public.
e.
EDCR-74-12, Replacement of the Yankee In-Core Instrumentation System, was completed during the report period.
This change was done to fulfill the requirements of Proposed Change 114.
The new structure contains 24 thermocouples and 22 paths for the movable flux mapping system. The new flux mapping system utilizes a movable fission chamber detector mounted on a drive cable. This enables rapid mapping of flux traces and repetition of runs if desired.
This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety analysis report.
2.
The following changes not requiring Commission approval were made:
a.
PDCR 73-16, Change to Main Coolant Check Valve Internals, was completed on Main Coolant Check Valves in the MC Loops No. 1 and 2 during the report period. This change was performed to lessen wear and subsequent maintenance on the check valve internals.
Bushings made of Stellite were incorporated at several points where excessive wear has been experienced.
This change does not impose any unreviewed safety question in that the system will operate in the came manner as covered in the exist-ing safety analysis report.
a b.
PDCR 73-26, Safety Injection Annunciator Loss of Power Alarm, was completed during the report period.
This chan 3 was performed to warn the control room operator of a malfunction of the Safety Injection Panel annunciator power supply.
This change does not impose any unreviewed safety question in that the system will oper;.e in the same manner as covered in the existing safety analysis report.
c.
EDCR 73-27, In-Core Instrumentation System Spacer Tubes, was completed during the report period.
The installation of the new in-core instrumentation system with the revised flux path arrange-ment, required that spacer tubes be installed in partially spent fuel.
In order to accomplish this, it was necessary that new design spacer tubes be procured.
The spacer tubes are not safety related.
Installation of re-designed spacer tubes and/or relocation of instrumented assemblies within the core will not alter any accident analyses, increase the potential for an accident or affect any technical specification.
d.
EDCR 73-8, Yankee Rowe Pipe Whip Fix Project, was completed during the report period.
This project was undertaken as outlined by a YAEC task force in Washington, D. C.
in response to the NRC Pipe Whip Criteria of 12/18/72.
Piping restraints and shield walls were installed as necessary to protect the existing installation for safe shutdown.
This change does not impose any safety question not previously reviewed by the existing safety analysis report or the Report on Effects of a Piping Systen Break Outside Containment at Yankee Nuclear Power Station, Rowe, Massachusetts. Operation of the syetem is the same as covered in the existing safety analysis report.
e.
EDCR 73-24, Instrument Port Canopy Seals for Reactor Vessel Head Penetrations, was completed during the report period.
The installation of the new flux mapping system into the Yankee reactor during the Core X-XI refueling required that new head port adapters be installed into the reactor vessel head instrument nozzles.
Associated with the new adapters was the need for a new canopy ceal between the adapter and port.
The installation of the new design canopy seal will not alter any safety evaluations.that were conducted for the plant, alter the basis of a technical specification or present an unreviewed safety hazard.
f.
PDCR 73-17, Installation of Charging Line Restraint, was completed during the report period.
A hydraulic shock and sway suppressor was installed on the charging line in the lower pipe chase just before the Vapor Container penetration.
This was done to reduce
e a
e the amplitude of the vibration in the charging line, hence, reduce the stress at the penetration, lessening the probability of another pipe failure at the penetration.
This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the exist-ing safety analysis report.
g.
pDCR 73-24, Pressurizer Safety'and Relief Valve Discharge Piping Restraints, was completed during the report period.
Piping restraints were added to the pressurizer safety and relief valve discharge lines to upgrade the existing restraint system. The need for this change arose from a stress analysis report.
This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the exist-ing safety analysis report.
h.
PDCR 74-5, S1 Accumulator Level Switch, LS-1; 2, 3 and 4 Relocation, was completed during the reporting period.
The Safety Injection Accumulator level switches, LS-1, 2, 3 and 4 were relocated from the 200 cu. ft. actuation setpoint to a 150 cu. ft. actuation setpoint.
This change will insure that adequate water would be available for core flooding during a loss of coolant accident.
1.
PDCR 74-6, Repair of Charging Line Relief Valve SV-209, was completed during the report period. The leaking relief valve was replaced with a spare valve. When the spare valve was installed into the line, flanged connections were used on the charging line connections to facilitate case of replacement or repair, and testing.
This change does not impose any unreviewed safety question in that the system will operate in the same maaner as covered in the existing safety analysis report.
j.
PDCR 74-4, Waste Gas Surge Drum Sagf e Connection, was completed l
during the report period. This change allows a WGSD sample to be purged without releasing radioactive gas to the atmosphere and still maintain an effective method of obtaining a waste gas surge drum sample.
This change does not impose any unreviewed safety question in that the system will operate in a n.anner that does not present a significant hazard not described or implicit in the existing safety analysis report.
B.
Performance Characteristics 1.
Unusual Occurrence Report On August 18, 1974, the analyn'i of Core XI start-up-physico measure-ments of control rod worths was completed, and indicated that measured
e control rod worths were less than the values reported in Sections 103 Nuclear Design, and 409 Steam Line Rupture Accident, of the FHSR as modified by Amendment No. 9.
Measured control rod worths were obtained with the aid of the Westinghouse reactivity computer and were confirmed by analysis of end point critical boron concentrations.
A review of the accident and transient analyses in the license, as modified by Amenduent No. 9, has been conducted.
The review has shown that the measured control rod worths exceed the worths used in the analyses of all events except a Steam Line Rupture.
In addition, peak-ing factors as measured by the in-core instrumentation are below the design values used in the accident analyses.
Thus these analyses are still applicable to the core, with the exception of the Steam Line Rupture.
For the Steam Line Rupture accident, the measured rod worths are less than the values used in the analysis by no more than 0.9% Ap.
The review of this analysis indicated that even with the measured rod worths, the core will not return to critical during the course of the accident as long as the all-rods-out, full power concentration of boron is not less than 400 ppm.
(For lower concentrations, the moderator temperature coefficient of reactivity may become sufficiently negative so that the cooldown accompanying the accident introduces enough positive reactivity to re. turn the core to criticality.)
Thus plant operation can proceed at full power down to an all-rods-out boron concentration of 400 ppm without re-analysis of any accidents. This boron concentration will not be reached for at least six months.
Plant operation below 400 ppm all-rods-out critical boron will be eddiessed in a later report.
2.
Plant Information Report a.
On August 13, 1974 at 0232, the pressurizer wide range level channel experienced an intermittent full scale deflection when switched from 118 psi to 621 psi by the Control Roon Operator during the plant heatup. This transient produced a false pressurizer wide range level scrcm signal thus scranming the reactor.
The f & C Department investigated tne problem and could not duplicate the condition by exercising the switch repeatedly.
The plant was returned to normal operation without further action, and as a pt ccaution, a new switch will be installed at a future outage.
b.
At 1705 on 8/25/74 during the acceleration phase of the turbine over-speed trip test, the No. I nozzle pressure increased to 125 psig causing the reactor / turbine-generator persmissiva circuits to actuate.
When the turbine tripped on overspeed a reactor scram was also affected via the turbine trip auxiliary relay (20TVX) and the closed permissive contacts of relays K1 and K2.
e Subsequently a critical approach was initiated and criticality attained at 1804.
A recommendation was made to close the permissive pressure switch isolation valve to the No. 1 nozzle sensing line, during the acceleration phase of the test.
c.
On September 28, 1974, at 0750, during the control rod drop test, with rod groups A, C and D at 3" and Group B being withdrawn to 90" a partial insertion from 42" to 18" occurred on rods 9, 11 and 12 while a full insertion from 42" to 0" occurred on rods 5, 6, 7 and 8.
Investigation of the Group B control rod drive system contactors revealed a sluggish contact finger on the stationery gripper relay SG3.
The finger was cleaned and readjusted. Megger and voltage readings were then taken.
Group B control rods were then withdrawn to 90" and the drop test was resumed with no further problems.
d.
At 1000 on October 2, 1974, the reactor operator noted that control rod group A could not be driven "IN" or "OUT" by use of the " Rod Control Switch" S-25.
Investigation by the Maintenance Department disclosed that the electrical supply wire to the Group A can drive brake coil was broken off.
This open circuit condition kept the brake engaged during the attempted control rod motion by preventing the cam train from moving.
At 1005 on the sar.e day, the wire was reconnected and control rod Group A "IN" and "OUT" motion capability was restored.
C.
Changes in Operating Procedures 1.
The following Operating Procedures were prepared or changed as necessitated by changes in facility design I.A. above.
I.A.l.a.:
OP-6101 Nuclear Instrument & Reactor Protection System Precritical Check OP-2100 Plant Startup from Cold Shutdown OP-2103 Reactor Startup and Shutdown OP-2101 Plant Startup from Hot Shutdown OP-2104 Scheduled Plant Shutdown to Hot Standby OP-6103 Reactor Permissive Pressure Switch Calibration and Permissive Circuit Functional Test
e I.A.l.b.:
OP-4205 Safety Injection System Operation Check OP-4609 Safety Injection Channels Calibration &
Functional Check I.A.l.c:
OP-1000.12 Core XI BOL Zero Power Physics Test Procedure, Attachment A.
OP-1000.16 Core XI BOL Hot Zero Power Physics Test Procedure 2.
The Operating Instructions for the plant were changed as necessitated by performance tests and evaluations which have shown that the pumped ECCS flow rate at runout condition is less than the flow assumed in ECCS analysis.
The plant has administrative 1y limited the peak heat rate of Core XI to 11.37 KW/ft. This converts to a maximum power of 588 MWt (98%) and 75 inches on control red group A.
This problem is addressed to the NRC in Proposed Change No. 117.
3.
The following Operating Procedures were prepared or changed to improve the safety of plant operations.
OP-2160 Hydrazine Addition to M.C. System OP-5756 Inspection and Maintenance of Safety Related Motor No.
OP-6150 Incore Thermocouple Recorder Calibration and/or Repair OP-6457 Repair and Adjustment of Accumulator Pressure Regulating Valves SI-PR-58 (SI-PR-5E'-
OP-4600 Radiation Monitoring Surveillance Check OP-1000.12 Core XI BOL Zero Power Physics Test Proc 2 dure OP-4200 Main Coolant System Inservice Inspection System Leak Test and Hydrostatic Pressure Test OP-5209 Maintenance of Pressurizer Code Safety Valves OP-4503 Test of Pressurizer Code Safety valves, Valve No.
AP-0021 Operating Memos OP-8300 Receiving Radioact.'v< Material
AP-0204 Safety Classification of Systems, Components and Structures OP-4628 Main Coolant Pressure Primary Channel Calibration AP-8006 Quality Control of Health Physics Equipment and Contracted Services OP-4704 Hot Channel Calculations OP-7101 Symmetric Offset Test Program OP-4630 Main Coolant Pump and Main Coolant Stop Valve Interlock Surveillance Check OP-5755 Inspection and Maintenance of Safety Related ACB and/or Contactor No.
OP-4604 Steam Cenerator NR Level Trip System Surveillance OP-4631 Waste Gas Instrument Leak Test OP-6150 Incore Thermocouple Recorder Calibration and/or Repair OP-2253 Steam Removal from Secondary Plant to Repair High Pressure Steam Line OP-4200 Main Coolant System Inservice Inspection System Leak Test and Hydrostatic Pressure Test OP-5000.22 SI Accumulator Level Switch LS-1, 2, 3 and 4 Relocation OP-2104 Scheduled Plant Shutdown to Hot Standby OP-5255 Maintenance of Shutdown Cooling Relief Valves OP-5257 Maintenance of the Purification Pumps OP-5256 Maintenance of the Boric Acid Pump OP-4216 Verification of Post Accident Hydrogen Vent System Flow Capability OP-4621 HPSI Pump Ammeter Calibration, No.
OP-4623 V.C. Post Accident H Analyzer Calibration OP-4225 Throttle Valve Surveillance Test
OP-4224 Control Valve Exercise OP-4206 Flow Test of Two HPSI Pumps on Normal AC Power OP-4208 Flow Test of Two LPSI Pumps on Normal AC Power OP-4205 Safety Injection System Operation Check OP-7103 Control Rod Position Check OP-4234 Low Pressure Surge Tank Safety Valve Testing OP-4612 LPSI Pump Discharge Header Pressure Instrument (PT-8, PI-8) Calibration OP-6000.31 Safety Injection Accumulator Level Switch gna 4)
Replacement OP-5213 Maintenance of Main Coolant Loop and Charging Line Relief Valves OP-3012 Large Volume Chemical Spill OP-4215 Operation Test of Post Accident Containment Vent Control System Valves OP-3010 Fire or Forced Evacuation of the Control Room OP-2140 Periodic Jogging of the Main Coolant Pump When Secured for Extended Shutdown OP-5456 Maintenance of the Emergency Boiler Feed Pump and Turbine Driver OP-2176 Charging Line Removal from Service for Maintenance of Drain and Relief Valve OP-5000.23 Waste Liquid Feed to Evaporator and Overhead Condenser Vent Interlock Piping Changes AP-0213 Material Identification and Control AP-0211 Material and Service Purchase AP-0217 Document Control (Procured Material and Service)
AP-0212 Material Receipt OP-6455 Safety Injection Accumulator Level Switch Operational Check AP-8007 Calculation of Internal Dose (8-y)
OP-4616 S1 Tank Electronic Level Indication Channel Calibration OP-5259 Maintenance of the Low Pressure Surge Tank Makeup Pump OP-5258 Maintenance of the Shutdown Cooling Pump and the LPST Cooling Pump AP-0220 Surveillance Tests and Records AP-0203 In-Service Inspections OP-9239 Fluoride Analysis AP-0211 Material and Service Purchase AP-9011 Primary Chemistry Test Frequencies and Specifications OP-4900 Chemistry Surveillance Tests OP-4232 Vapor Container Inspection While Operating at Power OP-2000.17 Charging Line Removal From Service for Maintenance of Flow Detector Isolation Valves OP-6000.24 Installation and Testing of the Evaporator Feed and Overhead Condenser Vent Interlock Sov's (PDCR 73-5)
OP-6000.32 Installation and Testing of the Waste Cas Surge Drum Sample Connection (PDCR 74-4)
OP-5260 Maintenance of the Component Cooling Pump OP-2177 Return of Charging Line to Service After Completion of Maintenance OP-1000.16 Core XI BOL Hot Zero Power Physics Test OP-2100 Plant Startup from Cold Shutdown OP-4603 Accident Emergency High Radiation Monitor Calibration Check AP-2005 Operations Department Surveillance Schedule During Power Operation OP-4703 Control Rod Drop Time Measurement
a OP-5601 Diaphragm Valve Inspection and Replacement Program OP-4507 Inspection of Rod Drive Control Equipment OP-4501 Bi-Monthly Check of the Station Batteries OP-4500 Weekly Check of the Station Batteries AP-5000 Maintenance Department Surveillance Schedule OP-4504 Weekly Check of the Diesel Generator System OP-4229 Emergency Boiler Feed Pump Capability Test OP-4207 Periodic Testing of Emergency Diesel Generators OP-2650 Safety Injection Accumulator Makeup OP-2125 Filling and Venting of the Main Coolant System OP-4203 Weekly Valve Check OP-4201 Power Range Channel Calibration AP-7301 Temperature Map F p OP-4625 V.C. <ressure Indicating System Calibration OP-6264 Maintenance of the Steam Generator NR Level Trip System No. _,
OP-2475 Vapor Container Access OP-7102 Release of Air From the VC Atmosphere When Integrity is Established OP-2158 Establishment of Hydrogen Blanket Gas on Low Pressure Surge Tank AP-2006 Special Orders OP-2157 Spent Fuel Pit Cooling, Alternate Method OP-4202 Control Rod Operability Check AP-9000 Training of Chemistry and Health Physics Department Personnel OP-4610 Vapor Container Trip Valve Test and Pressure Switch Calibration
OP-4609 Safety injection Actuation Channels Calibration and Functional Check OP-6103 Reactor Permissive Pressure Switch Calibration and Permissive Circuit Functional Test OP-3313 Security Force Actions Under Emergency Conditions OP-6452 VC Recirculation Flow Instrument Calibration OP-4634 Safety Injection Actuation Pressure Switches (PS-230 & PS-239) Monthly Surveillance Check OP-6000.33 Emergency Repair of Power Range Channel 5 and 6 OP-4707 Spent Fuel Pit Inspection OP-2100 Plant Startup from Cold Shutdown OP-2101 Plant Startup from Hot Standby or Hot 3hutdown OP-2103 Reactor Startup and Shutdown OP-2104 Scheduled Plant Shutdown to Hot Standby or Hot Shutdown OP-2106 Plant Cooldown from Hot Shutdown, with the Main Condenser Out of Service OP-4201 Power Range Channel Calibration OP-6154 Maintenance of the Nuclear Instrumentation OP-3612 Component Cooling Pump Header Low Flow OP-3613 Component Cooling Pump Low Pressure Auto Start OP-3614 Component Cooling Pump Inlet Header Temperature OP-3615 Component Cooling Surge Tank Level OP-3616 Safety Injection Tank Temperature OP-3617 Safety Injection Tank Level OP-3621 Main Coolant Cold Leg Low Temperature OP-3625 Neutron Shield Tank Surge Tank Level OP-3741 Steam Cenerator NR Level Scram System Instrument OP-3742 Main Coolant Flow Scram System Loop Lov Flow
OP-3743 Main Coolant Scram System Instrument Test AP-0220 Surveillance Tasts and Records AP-0021 Operating Memos OP-5926 Maintenance of the VC Air Cooler Fans and Post Accident Hydrogen Control System Fans OP-4204 Monthly Test of Safety Injection Pumps OP-4207 Periodic Testing of Emergency Diesel Generator OP-4211 Emergency Boiler Feed Pump Surveillance Test OP-2168 Steam Generator Feed Via the Charging Pumps OP-6459 Repair of SI and VC Recirculation System Type 555 d/p Transmitter OP-9246 Radioactive Liquid Releases AP-6001 Equipment Ilistory Card System OP-4622 LPSI Pump Ammeter Calibration No.
AP-2003 Initial and Review Qualification of Auxiliary Operators OP-3600 Pump Seal Tank (TK-32) Low Level AP-7104 Core XI Operational Limits OP-2178 Diaphragm Valve Repairs OP-5350 Maintenance of the Diesel Generator System No.
OP-4626 Pressurizer Wide Range Level Calibration OP-E200 Contamination Containment Areas AP-0211 Material & Service Purchase OP-4802 Environmental Water Sampling OP-4803 Environmental Soil and Vegetation Sampling OP-4805 Environmental Benthal Sampling OP-4806 Environmental Milk Sampling
OP-4807 Er.vironmental Air Sampling OP-4808 Environmental Crop Sampling OP-4809 Environmental Camma Radiation Monitoring OP-4810 Environmental Fish & Aquatic Plant Sampling OP-4210 Fire Pump Operability Test OP-4631 Waste Gas Instrument Leak Test AP-5000 Maintenance Department Surveillance Schedule oP-4601 Nuclear Instrumentation Surveillance Check OP-3603 Power Range loss of Power or Dropped Rod Alarm OP-3604 Recorder High Flux Alarm OP-3605 High-Low Taverage Alarm OP-3611 Fail Safe Panel Alarm OP-3619 Main Coolant Pressure Alarm OP-3620 Main Coolant Hot Leg High Temperature Alarm OP-3624 Main Coolant T.oop Low Flow Alarm OP-3629 Safety Injection Panel Alarm OP-3630 2400 Volt Motor Auto Trip Alarm OP-3631 2400 Volt Motor overload Alarm OP-3633 High Neutron Flux Alarm /High Startup Rate Scram Alarm OP-3634 High Startep Rate Alarm OP-3635 Reactor Scram Alarm OP-3643 Vapor Container High Pressure Alarm OP-3644 Pressurizer Surge Line Low Temperature Alarm OP-3645 Pressuriz;r Pressure Alarm OP-3636 Dressurizer Level (Narrow Range) Alarm OP-3647 Pressurizer Level (Wide Range) Alarm
OP-3648 Charging Pump Auto Trip Alarm OP-3649 Charging Pump Auto Start Alarm OP-3745 ACB Auto Trip Alarm OP-3746 Switchgear DC Control Loss OP-3747 Battery Charger AC Failure OP-3748 Diesel Day Tank High-Low Level OP-3749 Nitrogen Supply Low Pressure OP-3750 LPSI Accumulator High-Low Pressure OP-3/51 Diesel Control Lockout OP-3752 Diesel Generator Trouble OP-3754 Battery No.
1, 2 or 3 Critical Voltage OP-3756 LPSI Accumulator Level Instrument Dry Leg Tell Tale D.
All Surveillance Tests and Inspections required by the Technical Specifica-tions were satisfactorily completed during the report period with the following exceptions:
See attached Table, I. D.
E.
Containment Leak Test 1.
A Class A test was performed on the vapor container during May 13 - 16, 1974.
The results of this test have been submitted in a separate report.
2.
The following Class "B" test was performed during report period:
On October 23, 1974 the Vapor Container Personnel Hatch was tested according to attachment "W" of the procedure OP-4702, Vapor Container Class B & C Penetration Tests.
The inner seal and outer seal were checked and then the doors were closed tight. A pressure gage, test isolation valve and a service air hose were connected to the vent connection at the top of the hatchway.
The isolated volume, approximately 240 ft3, was pressurized to 32 psig and monitor for pressure decay for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The measured leak rate was <l.5 psig/24 hr which is the criteria for this test.
3.
The following Class "C" tests were performed during the report period.
a.
On July 12, 1974, the Cavity Fill 1 solation Valve. CS-V-601, was tested according to attachment "A" of the procedure OP-4702, Vapor Container Class B & C Penetrations Tests.
Using the pressure decay method, a leak rate of 6 psig/hr was determined from a starting pressure of 35 psig. This was below the nominal leakage rate.
b.
On July 12, 1974 the Vapor Container Heating System Trip Valve, TV-409 was tested according to attachment "K" of the procedure OP-4702, Vapor Container Class B & C Penetration Tests.
The valve was pressurized to 32 psig for one hour.
The leak rate measured to be 3.8 psig/hr which was an acceptable rate.
wr c.
On July 25, 1974, the Steam Generator Blowdown Line Trip Valves, TV-401A, TV-401B, TV-401C and TV-401D were tested according to attachment "N" of OP-4702, Vapor Container Class B & C Penetration Tests.
Using the filled steam generators (static head of 76 ft.) to pressurize the valves, leakage of 8 ml after one hour of monitor-ing was observed. This was below the nominal leakage rate.
F.
Tests and Experiments 1.
The following tests or experiments requiring authorization from the Commission were performed.
None.
2.
The following tests or experiments not requiring authorization from the Commission were performed.
a.
Core XI BOL Physics Testing was completed during the report period.
The purpose of this testing was to mensure various coefficients and parameters at low power level to verify nuclear design calculations utilized for analyzing plant transients and accidents and to provide such coefficients and parameters for routine plant operation. These tests also provided verification of the assumptions of Proposed Change No. 115 to the AEC.
G.
The following plant key staff change 9 were made during the report period.
None.
5 SURVEILLANCE TESTS AND INSPECTIONS DI."CREPANCIES TABLE 1.D DATE TEST DESCRIPTION DISCREPANCY AOR NO.
RESOLUTION 11/19/74 Steam Generator NR Level No. 3 Steam Generator Low Level The Relay would be replaced at Trip System Surveillance Alarm Failed to operate the next reactor shutdown.
The Check circuitry for the operation of the Low Level Trip Signal was checked and operated satisfac-torily. Repaired Number 3 Steam Generator Low Level alarm circuit which had failed the acceptance criteria as reported in November 19, 1974. Reran the Surveillance Check Procedure OP-4604, "S.G.
NR Level Trip System Surveillance Check".
The results were satisfactory.
12/17/74 Steam Generator NR Level No. 3 Steam Generator Low Level The Relay would be replaced Trip System Alarm Failed to Operate at the next reactor shutdown.
The circuitry for the operation of t?.c low level trip signa'. was checked and operated satisfac-torily.
YANKEE ROWE SEMIANNUAL REPORT 7-1-74 TO 12-31-74 II POWER GENERATION
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YANKEE ROWE SEMIANNUAL REPORT 7-1-74 TO 17-31-74 III SIIUTDOWNS
PLANT SHUTDOWNS TABLE III PLANT METHOD OF STATUS DURAT ION SHUTTING DURING DATE (HOURS)
CAUSE OF OUTAGE DOWN REACTOR OUTAGE CORRECTIVE ACTION 8/25/74 5.65 Turbine Trip 6 Reactor Scram resulting Turbine trip on Hot Standby NA from high No. I nozzle pressure cn overspeed and M.
P.
turbine while overspeed testing.
reactor scram via turbine trip aux.
relay.
9/28/74 12.76 The Plant was shutdown for scheduled Manually Con-Hot Standhy NA control rod drop time test and NRC trolled Deliberate.
Reactor Operator License Exam.
Shutdown 11/30/74 38.30 The Plant was shutdown for repair to Manually Con-Hot Shutdown Replaced Jumper Cables and to Channel 5 & 6 Nuclear Detector.
The trolled Deliber-Condition Detector in Thimble 7 for 12/1/74 two channels had been placed out of ate Shutdow.
Channel 5 6 6 of the service because of circuit malfunc-Nuclear Instrumentation.
tions.
The No. 1 Feedwater heater, The cable pull box was which had developed several leaking drilled to prevent further tubes just prior to the scheduled water buildup in the box shutdown, was repaired.
which was the cause for the failure of Channel 5 6 6 Detector.
The Control Rod Drop Time Test was The No. 1 fecdwater heater performed on 11/30/74.
was taken out of service and four (4) leaking tubes were plugged.
i i
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I
YANKEE ROWE SDlI ANNI'AI, REPORT 7-1-74 TO 12-31-74 IV CORRECTIVE MAINTENANCE SUBD!ARY
MAINTENAlk JEPARTICIT CORRECTIVE MAINTENANCE
SUMMARY
TABLE IV MALFUNCTION i
EFFECT ON CORRECTIVE ACTION TA::E5 SYSTEM COMPONE':T CAUSE i
RES:JLT SAFE OPERATION TO PREVFNT 2FffnTITION Safety Injection #1 HPSI Pump Normal Wear Seal Leak NOIE NA I
Primary
- 2 Main Coolanti Wear Noisy Operation NOIIE Refurbished Internals. Shoul a
give longer life.(PDCR 73-16 Check Valve l
Waste Gas
- 2 Waste Gas Wear Diaphragm NONE iia Comp.
Ruptured I
i Steam Gen.
- 1 Blowdown Pinhole at veld Leak NONE NA Line i
Charging &
MOV-523 Open Torque Switch Did not open NONE NA Volume 1
Charging System CH-V-855 Defective Valve Bonnet Leak NON JA l
l l
Pressurizer PR-V-606 Defective Valve Leak N0fF 71A Waste Disposal VD-V-983 Worn Diaphragm Leak NONE NA System Main Coolant VD-Y--711 Defective Valve Leak NO:iE NA Safety Injection SI-V-18 Valve needs lapping Leak NOIE NA (CheckValve) l l
MAIICE iAI?
DEPART EiT CORRECTIVE MAINTENANCE SUM)'JsRY TABLE IV
.!ALFUNCTION EFFECT 0.:
CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION I
I eak
- iO::E IIA Charging CH-V-613 Derectiv-valve L
l 1
Safety In.jection. HPSI Piping at Overgrinding veld l Undercut
?!OIIE I;A VC Penetration Waste Disposal Waste Gas SS !Lacse Fittings
' Valve leaks I;0IIE
- IA Sample Valve !
& Fittings l
1 i
! PR-MOV-512 Valve Seat & Seal Valve leaks
- IO IE l,
!!A Pressurizer Valves Ring l
I Safety Injection CS-V-627 Valve seat Valve leaks
- iOIiE
- iA I
I 1
i l
1 l
1
+
6 t
I l
1 i
i i
i i,
CORRECTIVE MAIL mNANCE
SUMMARY
TABLE IV MAINTEIIA'iCF DEPARTENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Pzr. Sarety FR-5V-101 & 132 Rec. 11 Hydro Valves Leak None NA Valves Charging i
- 2 Charging Leakage Hone NA Volume Control Pump Safety
- 3 LPSI Pump Packing Gland Needs Leakage None IIA Injection More Packing Safety MOV-539 Operator Malfunctio:
Does not open al' TIane NA Injection the va:/ electric-ally Charging &
MOV-525 Packing gland too Does not close None NA Volune Control tight.
tight with operator y
('
)
0 9
V.C. Penetra-Personnel Hatch Faulty piping Leaks oil Iene NA tions M
6 9
Pressurizer PR-SV-182 Faulty gasket Leaks None NA RO)
High Set Relief installation Valve b cd V4 Charging &
CH-V-613 Valve stem hung up Cannot open or None NA h
Volume Control close valve y
s
CORRECTIVE MAIN 1 ANCE SUSBiARY TABLE IV MAINTENANCE DEPARTME?iT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Main Coolant
!!cn e NA Injection Charging &
- 3 Chg. Pump Poor Packing & Rams Leaks None aA Volume ~ Control Charginr 3:
- 1 Chg. Pump Poor Gasket Leakc None an Volu~.e Control Charging &
- 3 Chg. Pump Packing Lea %s None UA Volume Control Steam Generator Fecdline Isol.
Leaks by None
- A Piping valve in #2 Chg. Pump Safety CC-V-t321 Loose Bolts Cover Leaks None NA Injection
CORRECTIVE MAIN: JANCE
SUMMARY
TABLE IV IRIN1ERNCE DEPAR'ISEIT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Charging &
lb. 1 01arging Broken weld Icak
!bne
!bne Volume Control Pump Discharge Header Flange Safety Sample Valve Improper seating Icak
!bne Fone Injection SI TarA Shutdown Shutdc.m Cooling Itrn Seal Irak None None Cooling Pump Charging &
LPST Cooling Worn Seal Icak None Ibne Volume Control Pump Ourging &
Charging Pump dinhole in weld Inak Ibne Ibne Volume Control Flow Transnitter Charging and
!b.1 Qurging Improper Seating Inak Pone Ibne Volune Ccntrol Pump Palief Valve Waste Disposal Oxygen Vent 3as Bad Diaphragm Bonnet Icaks
!bne Ibne Valve Off Overhead Con-denser
CORilECTIVE MAINTENANCE
SUMMARY
TABLE IV fGINIENANCE DEPAPHMENI' MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Charging &
Ib. 3 Charging Vbrn Packing Icak Ncne Ibne Volume Control Pump Charging &
No. 1 Charging Relief Valve Seat Valve Iraks
!bne
!bne Volune (bntrol Pump Relief Bad Valve Rod Control Control Rod Broken Wire on Control Rod Rcroved Auto-Rods-In Featun' None Group "A" Brake Cable Croup "A" In-on Group A, due to Iligh Tavt operatisu Selected Group B for Auto-Rods-In Safety Injectia l Recirculating Flange Seal Bad Flange Irak Bbne
!bne Pump Suction Flange Waste Disposal Oxygen Vent Gas Diaphragm Bad Inak Ibne
!bne Valve Charging &
lb. 1 Charging Worn Packing Leak Ibne Ibne Volume Control Pump Emergency Bb. 3 Diesel Worn waterhose Slight leak tbne
!bne Power Drergency Fower bb. 2 Diesel Pbrn Waterhose Slight leak Ibne
!bne
CORRECTIVE MAIN 1u..ANCE
SUMMARY
TABLE IV IRIMERCE DEPATGETF MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Charging &
tb. 1 Charging Vbrn Packing and Icak Ibne
!bne Volume Control Pu: p bad ram Coolant No. 2 Coolant Strn bearings &
Vibration of Ibne Ibne Purification Purification seal bad.
pump & seal leaks Pump
CORRECTIVE MAIN':
ANCE
SUMMARY
TABLE IV FRImHWCE DEPARTFENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Core Instru-Shear Valve, Icose Packing Slight Irak None None rentation H-2 No. 3 Charging Check Valve Ibrn Irakege Past Seat None None Onrging &
k Volume Control yg Charging &
No. 3 Charging Bellows Broken &
Icakage None None Volume Control Pump Iblief Stuck Disc Valve Pressurizer &
'P-Handle Vent Icose packing gland Inakage None IA.;ne Pressure Control Valve Safety Injectior No.
. Icose Outboard Pack-Icakage None None Pump ing Gland Charging &
No. 3 Charging Packing Inak Inak Ibte Increascd None None
\\bltre Cbntrol Pump sa Dmrgency Power No. 3 Diesel Icose Stud on Oil Oil Filter Icak None None System Generator Filter Past Seating Surface
CORRECTIVE MAINTENANCE
SUMMARY
TABLE IV MAINTENANCE DEPARTMENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Coolant No. 2 Coolar t Bad Motor Excessive nois 2 None None Purification Purificatior Bearing and vibration Pump Chemical Boric Acid Partially clogge 1 Does not Pump None None Shutdown Transfer suction & dis-properly System Pump charge Line Charging &
No. 3 Charg-Ram Bad Leakage None None Volume ing Pump Control Vent & Drain Loop No. 1 Packing Loose Packing Leak None None System Th Drain Valve Vent & Drain Loop No. 3 Packing Loose Packing Leak None None System Th Drain Valve v6
? MENT CORRECTIVE NAINTENANCE
SUMMARY
TABLE IV MALFUNCTION EFFECT ON CORRECTIVE ACTICN TAKEN SYSTE" l COMPONENT C.W Si RE%LT SAFE OPERATION TO PRLtJENT PE JITION _
.C.
Isolation TV h01C Diaphragm Bolts Air Leak None IIA
- 3 SG Blowdown Loose i
i I
faste Disposal Temp. Indicator Dirty Gwitch &
- Erratic Operation None NA j
f Slidevire
{
I t
I I.C. Isolation SOV h31 Cable Insulation None Hone "A
Cracked i
f
~rimary L&N Recording Defective Switch
' Err.ttic Operation None
.:A Chmmeter i
I 1
I i
Iucl m Instru-Thimble #3
! Cables & Detectors' Erratic Operation
- Ione
.entation Detector Deteriorated afety Injecticti SI-V-82 Sample Plugged line No sample None
- TA Valve i
V.C. Isolation TV h01B,TV h0lf Defective Plug i Suspect Leak None
.4 A Ring f
Safety Injectior' #3 LPSI Light Broken Wire Incorrect None NA on MCB Operation Reactor Pro-
- h St. Gen. NR Defective lamps &
Alarms in all the Ihne
- iA tection Level Trip Photo Diodes l time Sigma Relay I
t
I&C DEPARTMDT CORRECTIVE MI TENANCE
SUMMARY
TABLE IV MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKE.
SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Primary
- 3 & #h MC Flow Defective Ball Vent Suspect Leak None NA Detector Valve Main Steam TV-401C Stroke out of Leaked by None NA Calibration Radiation Gamma Alarm Defective Electron Incorrect None NA Monitoring
- 5h0 Tube Operation Safety Injection'SI-PR-59 Calibration Drift, Incorrect None NA Operation Waste Disposal HLS-34-1 Broken switch Incorrect None NA Operation Radiation Jordan #1401 Defective Meter Incorrect None NA Monitoring Operation Reactor Spare St. Gen.
Out of calibration Incorrect i
None NA Protection NR Level Ind/
Rea, ding Alarm unit
CORRECTIVE MAINh.yANCE
SUMMARY
TABLE IV I & L DEPART?E IT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Service Water VC Cooling Pumi Defective Valve Leak fione ITA Disch. Press.
Gauge Primary Charging Pump Broken Drive Corci Incorrect Opera-I!cne
?IA Disch. Press.
tion Safety Accumulator Air in transmitter Incorrect re ariin,
?!ane
- in Injection LT-5 Primary Charging Flcv Defective Vent Valv.
Possible Le9.
?!one I!A Transmitter Ball Feed &
- 3 S.C. Levels Defective Amplifier Incorrect readin, I;one
?IA Condensate Safety SI Tank sample Plugged line.
Leak
!!one
!!A Injection valve Safety Accumulator a
.,o n e aA Defective meter Incorrect rend 1nc indeetion LI-5 Safety Accumulator over pressurized Incerrect opera-Iione
!iA Injection LS-L,LS-o during hydro tion of switch Priruary Pressurizer WR Dirty Switch Incorrect Ilone ITA Level indication
5 CORRECTIVE MAIN 1 lANCE
SUMMARY
TABLE IV I & C DEPAPTIEIT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Nuclear
- 1+ Channel Defective Will not calibrati IJone
?IA Instrumentation Log M.C.
Nuclear Ch.8 Power Screw loose Erratie operation Ilone
!!A Instrunentation 7ange Meter Safety SI Super.
Incorrect wiring Incorrect opera-Ilone
- IA Injection lights tion Primary Loop #3 Flow Vent Valve Leak Small leak
- !one
'!A Trans Feed &
S.G. IIR Level Defective Pens Uo Indications Zione
- !A Condensate Recorder Radiation S.G.
- 2 Rad.
Defective Electron Incorrect Ilone iA bbnitoring Monitor Tube Operation.
Nuclear auclear Low Gain Incorrect
'!o n e NA Instrumentation Recorder Indication
CORRECTIVE MAIN 7 ANCE
SUMMARY
TABLE IV I & C DEPAim GNP MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Radiation Prinnry Vent Defective Detector Erroneous Ibne tbne abnitoring Stack Tube Readings Radiation Ib. 2 Steam Defective Electron Erroneous Ibne Ibne bbnitoring Generator Leak Tube Readings Detector Radiation V.C. Air Mylar Windw on Reading Iligher Ibne Ibne Fbnitoring Puticulate Detector Contami-
'1han Nontal nated Radiation Primuy Vent Defective Irrp3 dance Incorrect bbne Ibne obnitoring Stack Matching Device Operation Secondary Plant tb. 3 Steam
' Light Source Out Iru Alarm in All Ibne tbne Instrumentation Gen. Invel
'Ih's Time Trip System prah he Ibne Auxiliary Pri-Bleed Flw Defective Electron rary Plant Indicator Tube Instrumentatior
CORRECTIVE MAIF ^ 1ANCE
SUMMARY
TABLE IV I&C DEPARIMEtTI' MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Primary Plant Bleed Flm Defective tube Incorrect op3 ration Ibne Ibne Aux. Instrumn-Indication tation Radiation Vapor Container Lines filled with Incorrect ibne Ibne bbnitoring Air Particulate water operation Detector Radiation Vapor Container Iligh Voltage & Disc Reading low
!bne
^bne bbnitoring Air Particulate Out of Mjustment Detector Radiation No. 3 Steam Gen Defective tube.
Incorrect operati on Bbne Ibne bbnitoring
' Rad. Monitor Radiation Vapor Container High Voltage Out of Itigh Source Readin J None Ibne hbnitoring Air Particulate Adjustment Detector Radiation Vapor Container Air Bubble Between Erratic Operation
?bne
!bne bbnitoring Air Particulate Crystal & Plutotube Detector
CORRECTIVE MAIN ANCE SUMMART TABLE IV I & C DEPAIUMENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Radiation Prinury Vent Defective Electron Instrunent Pcadin<
l Mone tbne htnitor Stack bbnitor Tthe Iow Incore T/C Recorder Defective Print Inoperative None None Instrumentation BLhism Auxiliary
- 4 FC Purrp Dead Battery Incorrect Opera-
!bne Ibne Prinnry Plant Bearing Tenp.
tion Instrunentation Indicator Padiation
'C Air Particu-Air cubble Between Erratic Ooeration None None htnitor late Ibtector Crystal & Detector Tube
~
j CORRECTIVE MAIll 1ANCE
SUMMARY
TABLE IV I & C DEPARTMENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETTTION Primary Plant Vari-orifice Dial out of Brroneous None None Control Valve adjustment Indication Safety Indicating Burned out Indicated pump None (Pump was veri-None Injection light HPSI not in Auto fied by check of Pump at the Standby lineup to be in Auto MCB Standby.)
Safety Loop 2 Flow Fitting Loose Slight leak None None Injection Transmitter Secondary No. 3 S.G.
NF Optical Alarm 41 arm Inopera-None None Plant In s t ru-Low Level Light Burned Out tive mentation Alarm Nuclear Ins-Ch. 2 Low Ground Low Voltage None (Plant was None trumentation Voltage Power Power Supply shutdown)
Supply Defective Nuclear Ins-Ch. S&G Power Water Buildup Grounding and None (Removed two out The cable pull box was trumentation Ranges in Pull Box Shorting of of the six channels drilled to prevent Ch. S&6 which gives the Plant further water buildup Detector.
a high flux scram.
in the box which had Both channels of caused the failure of Reactor Protection were Ch. 5 & 6 Detector, placed in single so that anyone of the four remaining channels could initiat e a reactor trip.)
S CORRECTIVE MAIN 4ENANCE
SUMMARY
TABLE IV I & C DEPARTMENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TAKEN SYSTEM COMPONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITION Safety Level Channel Dirty Booster Erratic None None Injection LIT-401 Relay Operation Instrumenta-tion Safety SI Tank Hi Broken Lead Short None None Injection In-Level Alarm strumentatiortwitch Radiation VC Air Motor Bracket Motor vibratin7 None None Monitoring Particulate Loose Radiation Accident Circuit Drift Yellow Alarm None None Monitoring Emergency Light On Gamma Alarm Unit
Y ANKEE ROWE SEMIANNUAL REPORT 7-1-74 TO 12-31-74 V
SUMMARY
OF PLANT SYSTEM CilEMISTRY AND RAD 10 CHEMISTRY r
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a YANKEE ROWE SEMIANNUAL at 3RT 7-1-74 TO 12-31-74 VI OCCUPATIONAL PERSONNEL RADIATION EXPOSURE
TABLE VI.a OCCUPATIUi!AL RADIATIU, EXPOSURE
SUMMARY
l July 1974 to 31 Dec. 1974 i
100-250-500-750-Personnel N.M.E.
<100 250 500 750 1000 1-2 2-3 3-4 4-5 5-6
>6 Category mrem mrem mrem mrem mrem Rem Rem Rem Rem Rem Rea Permanent 43 5
8 16 13 10_,_
20
_1 0
0 0
0 620 34 26 14 18 5
18 0
0 0
0 0
Temporary
- No Measurable Exoosure A tabulation is attached listing the number of personnel receiving more than 3 Rem in the calendar year and the major cause(s) thereof.
All exposures in excess of 500 mrem during the six month period are tabulated on the attached page(s).
A table of reference symbols for the major duty functions of the tabulation is as follows:
Symbol Duty Function A
Routine surveillance and inspection (normal operation)
B Routine maintenance C
Special maintenance (describe below**)
D Routine refueling operation E
Special refueling operation (describe below)
F Other job related activities (describe below**)
- Additional Description of Duty Functions C.
1.
Cutting-up of old incore instrumentation package.
2.
Preparing for disposal of waste ion exchange resin.
3.
No. 2 Main Coolant check valve repair.
4.
No. 1 Main Coolant check valve repair.
5.
Maintenance on new incore instrumentation package.
E.
1.
Installation of new incore instrumentation F.
1.
None NOTE:
All categories are expressed as a percent. of the total exposure.
EXPOS!!PliS IN EXCESS OF 500 Mrem TABLE VI.b Total mrem per duty function
- Personnel Exposure Number Job Assignment (in Rem)
_ A B
l C
D E
F 1.
Nux. Operator
.900 0
0 24 49 27 0
2.
3ontrol Rn. Operator
.590 0
24 72 4
0 0
3.
tr<. Operator 1.150 7
9 45 17 22 0
4.
htx. Operator 1.251 6
12 38 36 8
0 5.
htx. Operator
.610 0
6 41 53 0
0 6.
htx. Operator
.960 0
12 57 31 0
0 7.
bitrol Rn Operator 1.660 0
9 64 5
8 0
8.
Shift Supervisor
.620 26 39 35 0
0 0
9.
Control Rn. Operator
.910 11 22 38 9
20 0
'10.
Shift Supervisor
.550 64 12 19 5
0 0
' l.
Control Rn. Operator
.500 0
12 24 17 47 0
12.
Aux. Operator 1.980 3
14 62 9
12 0
13.
' Control Rn. Operator 1.230 11 18 41 21 9
0 14.
Shift Supervisor
.760 46 13 22 12 7
0 15.
Control Ra. Operator 1.230 0
12 55 28 5
0 16.
Control Rn. Operator
.710 0
29 61 3
7 0
17.
Control Rn. Operator 1.740 2
13 79 6
0 0
1r>.
Aux. Operator 1.410 8
11 47 31 3
0 19.
shift Supe, isor
.980 16 17 47 7
l?
0 20.
Asst.Oper Supv.
.500 6
17 55 14 8
0 21.
Aux. Operator
.890 12 11 22 48 7
0 22.
Aux. Operator 1.610 9
5 42 27 17 0
23.
Aux. Operator 1.400 22 4
14 57 3
0
.m 1
Ibint. Supervisor
.980 2
9 48 0
41 0
25.
. Plant Mechanic 1.370 0
36 62 2
0 0
l Described on previous page
e 4
EXPOSURES IN EXCESS OF 500 Mrem TABLE VI.b Total mrem per duty function
- Personnel Exposure Number Job Assignment (in Rem)
A B
C D
E F
26.
Plant Mechanic 1.590 0
16 79 5
0 0
27.
Tech. Asst.bbint. Dept.
1.820 0
6 85 0
9 0
28.
Plant Mechanic 1.360 0
38 56 5
1 0
29.
Plant bbchanic 2.280 0
12 80 8
0 0
30.
Plant bbch. (Elec)
.910 13 58 27 0
2 0
31.
Plant bbchanic
.700 2
17 54 27 0
0 32.
Ibint. Supervisor
.550 0
0 50 0
50 0
33.
Plant bbchanic 1.270 0
29 57 9
5 0
34.
Plant bbchanic 1.530 0
l?
89 0
0 0
35.
Plant ?bchanic 1.230 0
31 64 5
0 0
36.
Tech. Asst. b'aint. Dept.
1.150 0
3 72 0
25 0
37.
Tech. Asst. Ibactor
.500 0
18 82 0
0 0
Engineering 38.
Contrul Technician
.580 0
20 80 0
0 0
39.
Contml 1bchnician 1.000 9
64 27 0
0 0
40.
Control 1bchnician
.620 4
44 52 0
0 0
41.
.570 2
11 87 0
0 0
42.
Chemist
.950 29 6
11 52 2
0 43.
1bster
.600 34 33 33 0
0 0
44.
Ibster 1.090 28 10 45 0
17 0
45.
IEPSco Laborer 1.560 0
12 87 1
0 0
46.
IEPSco Iaborer 1.570 0
0 99 1
0 0
47.
IEPSCo L 'borer 1.170 0
0 75 3
22 0
48.
IEPSCo labomr 1.210 0
35 65 0
0 0
49.
NEPSCo Iaborer 1.560 0
44 53 3
0 0
50.
IEP5Co Inborer
.960 2
16 80 2
0 0
i Described on previous lage
EXPOSURiiS IN liXCESS OF 500 Mrem TAB 1,E VI.b Total mrem per duty function
- Personnel Exposure Number Job Assignment fin Rem)
A B
C D
E F
51.
NEPSCc Laborer
.570 0
6 93 1
0 0
52.
NEPSCo Laborer 1.800 4
5 90 1
0 0
53.
NEPSCO Inborer 1.460 0
24 76 0
0 0
54.
NEPS(b Iaborer
.530 0
36 59 5
0 0
55.
NEPSCO Iaborer 1.270 0
3 90 7
0 0
56.
NEPSCo laborer 1.180 0
7 92 1
0 0
57.
NEPSCo Inborer
.730 0
0 100 0
0 0
58.
NEPSCO Iaborer
.620 0
0 93 7
0 0
59.
NEPSCo Inborer
.740 0
0 100 0
0 0
60.
NEPSCo Iaborer
.590 0
3 91 6
u 0
'l.
NEPSCo laborer
.550 0
0 96 4
0 0
62.
NEPSCo laborer
.930 0
0 100 0
0 0
63.
NEPSCn laborer 1.290 0
18 80 2
0 0
64.
NEPSCo Iaborer
.690 0
7 85 8
0 0
63.
NEPSCO Iaborer 1.100 0
12 85 3
0 0
66.
Contractor
.800 0
0 100 0
0 0
67.
Contractor 1.100 0
0 100 0
0 0
68.
Ccntractor
.930 0
0 100 0
0 0
69.
Contractor
.710 0
0 100 0
0 0
70.
Contractor 1.120 0
0 100 0
0 0
71.
Contractor 1.620 0
0 100 0
0, 0
72.
Contractor
.760 0
0 100 0
0 0
73.
Contractor 1.930 0
0 100 0
0 0
1.
Contractor 1.370 0
0 100 0
0 0
75.
Contractor
.540 0
0 100 0
0 0
Described on previous page
a EXPOSURES IN ETCESS OF 500 Mrem TABLE VI.b Total mrem per duty function
- Personnel Exposure Number Job Assignment (in Rem)
A B
C D
E F
76.
Contractor
.970 0
0 100 0
0 0
77.
Contractor
.530 0
0 100 0
0 0
78.
Contractor 1.330 0
0 100 0
0 0
79.
Contractor
.870 0
0 100 0
0 0
80.
Contractor 1.140 0
0 100 0
0 0
81.
Contractor
.570 0
0 100 0
0 0
82.
Contractor
.890 0
0 100 0
0 0
83.
Contractor 1.100 0
0 100 0
0 0
84.
Contractor 1.830 0
0 100 0
0 0
85.
Contractor 1.660 0
0 100 0
0 0
Described on previous page
E e
TABLE VI.c OCCUPATi0t1AL RADIAT10:1 EXPOSURE
SUMMARY
Exposures in Excess of 3000 Mrem
' January 1, 1974 - December 31, 1974 100-250-500-750-Personnel N.M.E.
<100 250 500 750 1000 1-2 2-3 3-4 4-5 5-6
>6 Category mrem mrem mrem mrem mrem Rem Rem Rem Rem Rem Rem Permanent 2
Temporary 5
- No Measurable Exposure A tabulation is attached listing the number of personnel receiving more than 3 Rem in the calendar year and the major cause(s) thereof.
All exposures in excess of 500 mrem during tr.e six month period are tabulated on the attached page(s).
A table of reference symbols for the major duty functions of the tabulation is as follows:
Symbol Duty Function A
D.outine surveillance and inspection (normal operation)
B Routine maintenance t,
Special maintenance (describe below**)
D Routine refueling operation E
Special refueling operation (describe below)
F Other job related activities (describe below**)
Additional Description of Duty Functions C.
- 1. No. 1 Main Coolant Pump inspection.
2.
No. 1 Main Coolant Loop Check Valve repair.
3.
No. 4 Steam Generator Tube repair.
- 4. Cutting up of old incore instrumentation package.
5.
Preparing for disposal of Waste Ion Exchange Resins.
6.
No. 2 Main Coolant Loop Check Valve repair.
- 7. Maintenance on new incore instrumentation package.
E. Installation of new incore instrumentation package.
F.
- 1. Installation Post Accident Hydrogen Removal System.
2.
Inservice Inspections.
- 3. Steam Gcnerator Blowdown Valve installation.
NOTE:
All categories are expressed as a percent of the total exposure.
9
EXPOSURES IN EXCESS OF 3000 Mrem TABLii VI..d Total mrem per duty function
- Person; el Exposure Number Job Assignment (in Rem)
A B
C D
E F
1.
Tech. Asst.bhint. Dept.
3.930 4.5 3.5 66.0 16.5 4.5 5.0 2.
Plant Mechanic 3.330 1.0 25.5 49.0 17.5 0
7.0 3.
NEPSCo laborer 3.450 0
0 79.0 14.5 0
6.5 4.
NEPSCo labomr 3.470 0
22.0 61.5 16.5 0
0 5.
NEPSCo laborer 3.060 1.0 8.0 74.5 10.5 0
6.0 6.
NEPSCb Iaborer 3.490 2.5 2.5 75.0 10.0 5.0 5.0 0.5 7.
NEPSCo laborer 3.490 0
1.5 76.0 22.0 0
j.
9 9
Described on previous page
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l
YANKEE ROWE SEMIANNUAL REPORT
,r-7-1-74 TO 12-31-74 VIII AB?;0RMAL OCCUkRENCES
i b
i A.
Report Number:
50-29/74-4
Title:
Chlorides Exceeding Specifications in the Reactor Coolant System.
Occurrence Date' July 2, 1974 Description of Occurrence:
During routine chemistry analysis of the reactor coolant, a chloride ion concentration of 0.1 ppm was detected.
This is a violation of Technical Specifications, Appendix A, Section B.2, referencing Section 106, Reactor Coolant Chemistry of the Final Hazards Summary Report which lists a chloride ion concentration limit of less than 0.1 ppm.
B.
Report Number:
50-29/74-5 litle:
Incore Instrumentation Flux Thimble Leak Occurrence Date: August 25, 1974 Description of Occurrence:
During a plant startup following a scheduled turbine overspeed trip test, a leak alarm was acknowledged on the movable incore flux map' sing system.
An entry into the vapor container confirmed that a minor leak had developed in flux thimble 11-2.
C.
Report Number:
50-29/74-6
Title:
Incore Instrumentation Flux Thimble Leak Occurrence Date:
December 1, 1974 Description of Occurrence:
During a plant startup following a scheduled maintenance shutdown, a leak alarm was acknowledged on the movable incore flux mapping system.
An entry into the vapor container confirmed that a minor leak had developed in flux thimble K-6.
O