ML20041F304
| ML20041F304 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/01/1982 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20041F300 | List: |
| References | |
| NUDOCS 8203160375 | |
| Download: ML20041F304 (38) | |
Text
e 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 50.59b and Sections 6.9.1.4 and 6.9.1.5 of Appendix A to DPR-21 and DPR-65.
No common site reporting requirements occurred during 1981. 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 1981.
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MILLSTONE UNIT 1 CONTENTS CllANGES PAGES DESIGN CHANGES 1/1 - 1/13 PROCEDURE CilANGES 1/1 - 1/1 TESTS 1/1 - 1/2 CilALLENGES 1/1 - 1/1 RADIATION EXPOSURE 1/1 - 1/1 3
Page 1 of 13 PLANT DESIGN CHANGES The following list by design change number summarizes those design changes completed in 1981, relating to safety-related equipment, which could have had a potential impact on safety-related systems, could have potentially impacted the environment, or required a change to the FSAR.
None of the items constituted an unreviewed safety question.
PDCR 1-28-75 A chlorine monitor was installed at the railroad car unloading building with an annunciator in the Unit 2 control room. This change did not affect the operation of any safety-related system.
PDcR l-40-76 LPCI valves 1-LP-10A and 1-LP-10B motor starters were upgraded from NEMA size one (1) to NEMA size two (2). The motor power cables from the motor to the starter were upgraded also by replacing the existing #12 3/C with 4/C #10. This change did not adversely affect any safety system.
PDCR l-58-76 The Core Spray maintenance stop and check valve, safe ends and thermal sleeves including the piping, in loops "A" and "B" inside the Drywell were replaced. The change results in no change to the safety analysis of the Core Spray System.
PDCR 1-21-78 A complete reroute of the CRD hydraulic return line from the reactor vessel to the feedwater line was accomplished. All the old piping in the Drywell was removed and the nozzle at penetration X36 was capped. The change results in no change to the safety-related system analysis.
PDCR 1-62-78 Constant voltage tr,ansformers were installed in the power supplies to reactor level indicators 263-100A and 100B, 263-106A and 106B. This installation was performed to prevent calibration shif ts caused by variations of input power.
This change did no$ adversely af fect any safety-related equipment.
PDCR 1-88-78 An alarm window was installed to indicate the fire protection CO2 System is bypassed (when personnel are in the gas turbine cubicle) to prevent the system from being lef t accidentally out of service. This change did not af f ect the safety analysis of the system.
Page 2 of 13 PDCR l-95-78 New conduit and cable 3 have been installed for the core monitoring dunking chambers to maxe the circuitry a permanent installation. This change did not affect the operation of any safety-related system.
PDCR l-6-79 Eight vent header deflectors were installed below the torus vent header.
Vent header stresses caused by pool swell during a LOCA are much greater than assumed for the initial Millstone Unit 1 design. The Mark I Containment Long-Term Program test results indicated that these stresses are reduced to an acceptable level by use of vent header deflectors.
This change did not adversely affect safety.
PDCR l-7-79 The submergency length of the 96 torus downcomers was reduced by truncation of the downcomers. This modification results in a range of submergency from 3.0 feet to 3.33 feet with 1.0 psid between the drywell and torus air space. The decreased downcomer submergence reduces the pool swell loads that could be imposed on the torus in the initial stages of an LOCA. The downcomer tie straps were replaced. The new tie straps are designed to meet postulated induced lateral loads on the downcomers as defined in the Mark I Containment Program Load Definition Report.
This change did not adversely affect safety.
PDCR l-19-79 New cables and new conduit were installed for the dunking chambers from the 108' level fuel pool area down to the 14'6" level reactor building IRM preamplifier cabinet to institute a more permanent installation.
This change had no affect on safety.
PDCR l-39-79 A six-inch drip leg and a new steam trap assembly was installed in the sixteen-inch supply line for the Isolation Condenser. The discharge line from the steam traps was rerouted to the torus instead of the condenser via the cleanup line on the second floor of the Reactor Building in the Cleanup Heat Exchanger Room. This provided a much better drain system than had been previously installed and enabled the system boundary to be contained within the primary containment. This change resul's in no change to the safety-related system analysis.
PDCR 1-53-79 A stop valve and expaasion tank were installed on the chlorine main supply header to allow double valve isolation of the Unit I chiorinator l
without effecting the Unit 2 Chlorinator System. This change did not affest the operation of any safety-related system.
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j PDCR 1-76-79 An orifice has been installed in the off-gas cooling water line between the condenser and valve number V2-709. This orifice takes the pressure drop instead of the valve which was replaced due to severe cavitation damage. The valve can now satisfactorily throttle flow to 1000 gpm.
This change did not affect the operation of any safety-related system.
PDCR 1-101-79 The recirculation pump M.G. sets wattmeter potential transformers were reconnected from an open delta configuration to a wye configuration to obtain more accuracy in the wattmeter. This change did not affect any safety-related equipment.
PDCR 1-113-79 The cables for the APR valve acoustic monitors were installed in i
. seismically supported conduit runs. This change did not adversely affect any safety system.
I PDCR 1-8-80, PDCR 1-71-80 i
An alternate reactor protection system was installed that is redundant, diverse and independent of the existing reactor protection system and is capable of mitigating the consequences of an ATWS. This change is required per NUREG 0460 and does not affect the safety analysis.
PDCR 1-23-80 i
Fire protection system modifications to existing hose stations, new hose station installation, wet pipe sprinler system installations, deluge system expansion, sprinkler head retrofit and halon system installation were made to comply with the NRC's Branch Technical Position ASB9.5-1 as interpreted in the Fire Protection Hazards Analysis. This change improved the fire protection of safety-related equipment.
PDCR 1-27-80 i
LPCI recirculation lines were installed at torus penetrations X-210A and X-210B.
This modification provides the ability to circulate the torus water during a safety-relief valve discharge and, thus, maintain a more uniform torus water temperature for a heat sink. The three ECCS strainers in the torus were modified to withstand postulated post-LOCA submerged velocity loads. The 18-inch cleanup system relief valve discharge line in the torus was shortened by 2' - 0".
This reduces submerged velocity loads and does not affect the function of the line. The 1-inch torus vent pipe drain lines were cut and capped. The modification reduces submerged velocity loads that would be experienced during an LOCA or a i
safety / relief valve discharge through a T-quencher. The pipe supports on the internal 4-inch torus spray header were strengthened. This modifi-cation prevents movement of the spray header in an upward direction when hit by any water spray during a pool swell as a result of an LOCA. This change did not adversely affect safety.
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Page 4 of 13 PDCR l-28-80 An additional 10-inch vacuum breaker was installed on each of the six safety / relief valve discharge lines. This modification reduces torus loadings in the event of consecutive safety / relief valve actuations by maximizing the potential for inflow of air or air / steam, thus limiting the reflooding of the discharge line and causing the water level in the line to quickly return to equilibrium. This change did not adversely af fect safety.
PDCR l-29-80 Torus shell penetrations were installed to facilitate installation of the torus water temperature monitoring system and torus water level monitoring system. This change did not adversely affect safety.
PDCR l-30-80 New valve discs were installed in the ten torus-to-drywell vacuum breakers.
The improved discs were fabricated f rom a more ductile material that improves vacuum breaker operability and reliability by increasing their ability to withstand post-LOCA valve chattering, as was demonstrated in full scale testing. This change did not adversely affect safety.
PDCR l-31-80 Support saddles were installed in sixteen locations under the torus. The saddles act both to increase the torus support capacity and to incr2ase the basic frequency of the structure, thus reducing the dynamic amplifi-cation of the loads.
4" SC-28, 4" SC-29 and the 8-inch torus drain line were relocated to facilitate saddle installation. Relocation of the lines has no negative impact on the operability of the affected plant systems. This change did not adversely af fect safety.
PDCR l-33-80 A viton seat check valve was installed in the air supply piping to each of the six main steam safety / relief valves and the eight main steam isolation valves. This modification insures that the MSIV and S/RV air accumulators will maintain sufficient air pressure for the required number of valve actuations following a postulated air header break in the drywell. These check valves act as backups to the hard-seated check valves previously installed in the air supply p.iping to each of the valves. This change upgraded and improved the associated safety system analysis.
PDCR l-39-80 The secondary containment pressure transmittters have been relocated to instrument rack 2205 to allow installation of a new ATWS instrument rack.
This change did not affect the operation of any safety-related system.
Page 5 of 13 PDCR l-41-80 Additional stages of contacts were added to the control switches for valves I-IC-1, 1-IC-3 and 1-IC-4.
The contacts were wired in parallel with the contacts on the Group IV isolation reset pushbutton so that the Group IV isolation logic cannot be reset with Isolation Condenser valve switches I-IC-1, 1-IC-2, 1-IC-3 and 1-IC-4 in the open or automatic position. This was implemented to prevent the Isolation Condenser valves from going open upon resetting the isolation signal.
the system would be put back in service per procedures. This change did not adversely affect
- any safety-related equipment.
PDCR l-49-80 The torus internal catwalk and associated support system were strengthened. These modifications increase the ability of the catwalk to withstand LOCA, safety / relief valve and seismic loads and has no effect on condensation or rising torus water. The design of the extra supports considers the loads imposed upon the additional supports thus assuring they are not potential missiles. Additional stiffening members were added to the T-quncher supports in the torus. Reinforcement of the supports provides extra capability to resist loads, especially forces on the quencher tee and on the areas that result from air clearing the safety / relief valve discharge line and pushing the water out of the line upon opening of the safety / relief valve.
Circulation of the water in the torus is unchanged. The torus vent header penetration support plates for safety / relief valve discharge lines 10" MS-8d, 10" MS-8e, and 10" MS-8f were reworked. Reinforcement of the support plates increases the load-1 carrying capability of the penetration with regard to either thermal or dynamic loads. This change did not adversely affect safety.
PDCR l-53-80 The orifice located at the discharge of the FWCI condensate transfer pump was removed. A new orifice was installed in the two 12-inch lines going to the A and B condenser hotwells. These orifices are located near the i
condensers to minimize cavitation problems. A new annular type flow measurement device and Barton differential pressure gauge were installed for more accurate and reliable pump flow data. This change results in no change to the safety-related system analysis.
PDCR l-57-80 Strip heaters were installed in each 4160 volt breaker cubicle. A transformer and distribution panel were also installed to provide power for the heaters. This change was performed to prevent moisture from collecting on the circuit breaker are chutes. This change did not affect any safety-related system.
PDCR l-60-80 A new limitorque SB-3 operator and new valve yoke adapter were installed on IC-1 to stop numerous failures of the valve actuator. This change did not af fect any safety-related system.
Page 6 of 13 J
PDCR 1-61-80 The heat detector above-house heating boiler Number 3 was disconnected from the maintenance shop alarm circuit and reconnected into an existing loop circuit in the boiler room. This change did not affect any safety-related system.
PDCR 1-65-80 The Condensate Storage Tank Level Transmitter has been relocated from the moat around the tank to inside the adjacent pump house. The sensing line j
was heat traced and seismically mounted. This change protects the level j
indication from being lost when the moat becomes flooded. This change constitutes no change in a safety-related system.
l PDCR 1-72-80 An orifice plate was installed in the emergency service water cross-tie piping to the service water system to take a large pressure drop minimizing i
the af fects on the butterfly valves due to throttling. This change did i
not affect any safety-related system analysia.
t PDCR 1-73-80 Peeco flow switches for the heater drain pumps were removed and replaced i
with annular type flow measurement devices that transmit a signal to new d,
DPS switches located outside the condenser bay. This change did not affect any safety-related system.
PDCR 1-75-80 l
Replaced the four existing Crane Tilting disc feedwater check valves with Anchor / Darling swing check valves and replaced the pipe elbows in the
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drywell between 1-FW-10A & B and 11A & B.
The new check valves will provide additional assurance in meeting 10 CFR 50, Appendix "J"
test requirements. This change did not adversely affect the safety system analysis.
PDCR 1-76-80 The service water discharge piping was rerouted to the top of the discharge tunnel due to the failure of the existing 24" line. This change did not affect any safety-related system.
l PDCR 1-81-80 Hodifications were made to the condensate demineralizer regeneration j
tanks to accommodate changing the cation to anion resin ratio. This change did not af fect any safety-related system.
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PDCR l-82-80 Vent valves were installed on the service water discharge side of the Turbine Building Closed Cooling Water heat exchangers which will facilitate the heat exchangers use with adequate cooling when the service water discharge piping is out of service and the plant is shut down. The change did not affect any safety-related system.
PDCR 1-86-80 Wiring, relays and indicator lights have been installed for a fourth safety relief valve to automatically operate as part of the Automatic l
Pressure Relief (A.P.R.) System. This modification forms the basis for l
the Cycle 8 safety analysis and was completed prior to Cycle 8 start-up.
j lt was necessary to avoid power derating due to ECCS thermal limits.
This change increased the safety of the system.
I PDCR 1-94-80 l
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The power source for the ISO condenser valves, 1-IC-1, 1-IC-2, 1-IC-3 and 1-IC-4 was changed from a gas turbine generator supplied bus to a diesel j
ger.erator supplied bus to enable the ISO condenser to function as an independent ECCS backup to FWCI since the FWCI trains are supplied by a gas turbine generator supplied bus. This change lacreases the safety of the ISO Condenser System.
PDCR l-99-80 A modified torus manway cover was installed on the torus hatch. The modified cover contains electrical penetrations that are required to run instrument leads to support performance of safety / relief valve blowdown testing to be performed following the outage. This cover will be removed and the existing one will be reinstalled during a future' outage. This change did not adversely affect safety.
I PDCR 1-104-80 Isolation valves were installed in the secondary closed cooling water lines in the heating and ventilation room for the cleanup recirculation pump and reactor building sample coolers. This change had no affect on safety.
PDCR 1-107-80 4
i The Drywell head studs were machined to fit a 2-inch hex socket for case in removal and installation. This change had no affect on safety.
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Page 8 of 13 PDCR 1-109-80 Constructed various environmentally-controlled areas in the Turbine Building and Reactor Building to meet the requirements of NRC I&E Bulletin 79-01 by isolating certain electrical instrumentation and control equipment necessary to mitigate the consequences of a high energy pipe break outside the primary containment from the hostile environment resulting from the break. This change increased the safety of the affected equipments.
PDCR 1-110-80 The individual high and low-level switches from the drywell floor drain and equipment drain sumps were replaced with combination high/ low level switches. This enables easy removal of the switches when maintenance is neces sa ry.
This change did not affect any safety-related equipment.
PDCR l-113-80 Installed four channels of ultrasonic water level detectors, one detector from each channel on both north and south scram discharge headers as required by NRC in accordance with I&E Bulletin No. 80-17 Supplement I.
This change increased the safety of the scram discharge header water level system.
PDCR 1-117-80 A three-contact level switch was added to the Radwaste high conductivity sump to allow for automatic level control and provide an alarm function to minimize sump overflow. This change did not affect any safety-related equipment.
PDCR 1-118-80 Added 4 tie-ins with manual isolation valves to the Turbine Building Secondary Closed Cooling Water system supply and return lines to provide cooling water for the motor control center environmental enclosure coolers. This change did not affect safety.
PDCR 1-119-80 A fire alarm central monitoring system was installed in the Unit I control room to monitor the Refueling Outage Building. This change was part of the additions to the fire protection system.
PDCR 1-121-80 The station battery racks eere replaced as part of Electrical Equipment Seismic Upgrade to meet seismic design requirements for Class IE equipment in Nuclear Power Plants. This did not adversely affect the safety system analysis.
Page 9 of 13 PDCR 1-122-80 Relocated the core spray pumps suction pressure indicators PI-1402-40A and PI-1402-40B to avoid interference with the new environmental enclosures for instrument rack 2201A and 2201B and to allow access to read or calibrate the instruments. This change did not af fect the safety analysis of any adjacent equipment.
PDCR 1-124-80 The three jet pump beam bolt assemblies (BWR 3 type) were removed and replaced with BWR 4 type on five jet pumps because the BWR 4 type are more resistant to intergranular stress corrosion cracking. This change did not affect any safety analysis.
PDCR 1-125-80 Replaced 304SS supply line piping on both north and south end of the isolation condenser from and including the nozzle safe ends up to and including the first elbow with 316KSS piping and 304L solution annealled safe ends. This change did not adversely affect any safety-related systems.
PDCR 1-130-80 Cross-tie connections on the TBCCW supply and return lines and the RBCCW supply and return lines to provide cooling flow to the waste concentrator from the TBCCW system during the time that the RBCCW heat exchangers were out of service. This change did not affect any safety-related analysis.
PDCR 1-131-80 Modifications were made to the Heating Ventilation Air Conditioning control panel, cable separation, and steam tunnel radiation monitors as required to accommodate (E.E.Q.) Electrical Environmental Qualifications modifications. This change upgraded the Electrical Environmental Qualifications of the particular safety-related equipment and, thus, enhanced their reliability.
PDCR 1-1-81 The secondary closed cooling water lines in the steam tunnel were rerouted to permit removal and replacement of feedwater check valves.
This change did not adversely affect andy safety-related system.
PDCR 1-2-81 The fire detectors were removed from the turbine generator lif t pump area because they activated the deluge system for the seal oil pump. This was possible because the turbine generator lift pump area utilizes a different fire protection system. This change did not adversely affect the safety-related system.
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Page 10 of 13 PDCR l-3-81 As part of E.E.Q., additional smoke and heat detection circuits were added to the existing fire protection system for the reactor building, and smoke and heat detectors were installed in environmental enclosures.
This change increases the fire detection system safety analysis.
PDCR l-7-81 The 304SS piping between valve 1-CU-1 and the shutdown cooling line was replaced with 316KSS pipe because of intergranular stress corrosion cracking. This change did not adversely affect any safety-related systems.
PDCR 1-8-81 The 304SS piping between valve 1-SD-1 and the reactor recirculation piping was replaced with 316KSS pipe because of intergranular stress corrosion cracking. This change did not adversely affect any safety-related systems.
PDCR l-10-81 The power supply for diesel oil transf er pump "A" (M8-47A) was relocated from motor control center F4 (2A-4) to motor control center EF7 (22A-1).
The power supply and control cables were rerouted for the diesel in "S2" facility cable trays and conduit because motor control center EF7 is located in the diesel generator room and would be protected from the harsh environment resulting from a main steam line break. This did not affect any safety system.
_PDCR 1-25-81 An external D.C. power supply for the hotwell transmitter was installed to replace the present power supply for ease in removal and maintenance.
This change did not affect any safety-related system.
PDCR 1-26-81 Installed four pressure switches (1 out of 2 taken twice) which sense scram air header low pressure. Upon sensing low pressure, this system will initiate a reactor protection system trip, thereby causing control rod insertion. This change proved to upgrade the reactor protection system.
PDCR 1-27-81 The low-pressure coolant injection piping between valve 1-LP-11B and the 18-inch by 16-inch reducer was replaced with 316KSS pipe because of pipe cracks caused by intergranular stress corrosion cracking. This PDCR did not change any safety system analysis.
s Page 11 of 13 PDCR l-28-81 s
The auto throwover switch located in the HVAC main control panel was removed to allow the two power sources to be used for the steam tunnel radiation monitors. The elimination of the auto throwover switch deleted the potential for a single failure to fall both of the steam tunnel ventilation radiation monit ors.
This change did not affect any adjacent safety-related equipment.
PDCR l-30-81 s
The automatic isolation condenser initiation upon reactor vessel low water level has bee 3 changed from an instantaneous initiation which would automatically reset to an instantaneous initiation which requires a manual reset. This change was made to prevent excessive cycling of 1-IC-3_ valve.as a result of reactor vessel water level fluctuating at the low water level trip setpoint. This change results i_nlno change to the
- safety related system analysis. -
PDCR l_35-81 The trip point settings (drop out voltage) of the Level /2 undervoltage sensing relays have been reset to 115 volts to correspond to an RSST primary voltage of 345 kV.
This change was made to assure a minimum of 108 volts on the vital AC and instrument AC buses for all plant conditions and improve reliability. This change increases the safety of the system.
PDCR l-42-81 Wiring was added to test indicator lights for the output status lights on
- the ATWS System to verify the trip relays are in the safe condition and can therefore prevent operator error during scheduled surveillance. This change increases the safety of the ATWS System.
sPLCR l-45'-81 The Turbine Building floor an'd' equipment drain sump was cross connected to the floor drain collect tank which will eliminate leaks under the condensate demineralizer room. This change has no, affect on safety.
PDCR l-46-81
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Two 1-inch check valves were added to the Control Rod Drive Cooling Water System to prevent potential back flow through the header from the hydraulic control unit check valves. This change did not affect safety.
PDCR l-53-81 Additional hanger supports were installed on the six-inch cleanup lines on the 14 feet 6 inch elevation of the Reactor Building. This was completed to accommodate lead blankets installed on the piping as temporary shielding. This change did not affect any safety-related system.
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Page 12 of 13 PDCR 1-63-81 A lighting panel and transformer were relocated from the stack sample room south wall to the northeast corner wall. New three phase 100-amp.
distribution panels were installed on the northeast wall. These changes were made to facilitate installation of a new KAMAN stack gas radmonitor.
This change did not affect any safety-related system.
PDCR 1-65-81 The existing A.R.M. equipment which had a range of.01 to 102 mr/hr was replaced with one that has a range of.1 to 103 mr/hr because radiation rates in the area of the radwaste filter sludge pump were constantly 2
exceeding the 10 mr/hr upper limit. This change did not affect safety.
PDCR 1-77-81 Disconnected the Unit I control room office ventilation from the Unit 2 ventilation system and reconnected it to the Unit I ventilation system.
This change did not adversely affect any safety-related system.
PDCR 1-75-81 The air regulators for the feedwater valve positioners (642 A & B) were removed and replaced with PALL air filters. The filters were installed to remove fine desiccants that cause the positioners to fail.
This change will improve feedwater valve reliability.
PDCR 1-81-81 The drywell equipment drain sump level switch LA-6-2 was defective leaving no indication of drain sump level. This change provided I
annunciator / level indication by utilizing the drain sump pump auto-start level switch in lieu of level switch LA-6-2.
This change did not affect j
safety.
1 PDCR 1-83-81 This PDCR provided annunciation for the recirculation pump field breakers on control room panel 905. This change did not affect safety.
PDCR l-88-81 Temporary hoses were replaced with a hard pipeline which will connect the house heating steam system to "B" waste concentrator blow down line.
This pipeline will prevent an injury while cleaning the blow down line.
This change did not affect safety.
Page 13 of 13 PDCR 1-95-81 The high-level alarms (LS-2-7) & (LS-2-8) were removed from the " Condenser Hotwell Level High-Low" annunciator on CRP 906-14A and displayed on a 1
spare window on annunciator 950-A3. The low-level alarms (LS-2-5) &
(LS-2-6) were removed from the " Condenser Hotwell Level High-Low" annunciator on CRP 906-14A and displayed on an adjacent spare window on annunciator 905-A3. This change did not affect safety.
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Page 1 of 1 PROCEDURE CHANGES There were no procedure changes as listed in the FSAR during 1981.
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Page 1 of 2 r
4 TESTS i
There were no special tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.
The following special procedures were implemented on various safety-related systems, none of which constituted an unreviewed safety question.
SP 81-1-4 With the reactor vessel defueled and vessel head in place, this procedure provided for lowering the reactor vessel water level for. replacement of nonisolable shutdown cooling / cleanup system piping.
The intent was to 1
maintain "B" recirculation loop filled and "A" recirculation loop discharge piping filled to minimize area radiation levels.
SP 81-1-5 I
This procedure described a method to install a temporary circuit for the reactor building ventilation supply and exhaust dampers 1-HV-1 and 1-HV-4 while the original circuitry under went modifications.
SP 81-1-6 This procedure established a method to determine the adequacy of wall-mounted electrical panel supports to withstand seismic loadings. The panels were tested by applying specified loads at the top and bottom to determine i
pull-out capability of the support bolts.
1 SP 81-1-8 J
This ultrasonic test pipe weld inspection procedure was written as an alternative 10E technique to evaluate and differentiate suspect IGSCC r
1 SP 81-1-10 1
Af ter the purchase of coaxial connection kits, a special procedure was written to cover installation instructions of all Raychem Nuclear Coaxial Connection Kits.
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SP 81-1-11 i
This procedure was written to describe a method to test the length of time to bleed down the scram air header after an ATWS-initiated signal assuming failure of one solenoid valve.
SP 81-1-12 i
This procedure described a method to perform a full function test of the j
scram discharge volume header continuous monitoring system. The CMS alarm function trips at a prescribed level and verification by independent measurement indicates that CMS is monitoring accurately.
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Page 2 of 2 SP 81-1-13 This procedure provided instructions to demonstrate operability of the core spray system with the torus dewatered to meet Technical Specifications 4.5.A.
SP 81-1-15 Special procedure written to provide instructions to perform hydrostatic testing of the cleanup system valve I-CU-1.
SP 81-1-16 This special procedure delineated a method to functionally test the automatic initiation of the standby gas treatment system and the reactor building and steam tunnel ventilation isolation logic for the steam tunnel ventilation radiation monitors.
SP 81-1-18 This procedure described the method to retrieve a dropped fuel assembly in the reactor cavity and move it to the fuel pool by use of the i
frame-mounted hoist.
r SP 81-1-21 This procedure covered the functional testing of LNP logic after removal of auxiliary contact logic inputs from feeder breakers to busses 14A and 14B. The RSST supply breakers to 14A and 14B open on test initiation and the gas turbine and diesel generators sequence properly and pick up respective busses.
LNP is initiated by loss of power to busses 14C and 14D.
SP 81-1-23 Procedure to cover pneumatically testing a new run of 2" pipe from the Isolation Condenser Steam Trap to isolation valve 1-IC-58 in the cleanup pump room.
SP 81-1-25 This procedure delineated a method to local leak rate test the CRD System in its normal lineup to the vessel.
SP 81-1-29 This procedure was for testing the response time between the trip sensor actuation and scram solenoid relay deenergization of the Reactor Protection System.
SP 81-1-36 This procedure dictated instructions to cycle open and shut 1-SD-1 in an attempt to get a tighter closure of the leaking valve.
Page 1 of 1 CHALLENGES TO REACTOR COOLANT SYSTEM RELIEF VALVES In accordance with Item II-K-3-3 of the TMI action plan, the following is a listing of challenges to the Automatic Pressure Relief Valves (APR)/ Safety Valves during 1981.
In this case, all SRV's operated as designed.
Date Circumstances August 10, 1981 The plant was at 93% of rated thermal reactor power, 577 megawatts electric and increasing power to 100% of rated. An automatic scram occurred when during a high risk surveillance the first channel tripped was not reset prior to tripping the second channel. Reactor pressure increased to 1085 psig. Five of the six safety relief valves operated during the transient.
'A' (APR) was opened by the operator for approximately 3 minutes. All (SRV's) operated correctly, i
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MILLSTONE UNIT 2 CONTENTS E^82 Changes Design Changes 1-7 Procedure Changes 8
Tests 9-12 Steam Generator Tube ISI 13 Challenges to RCS Relief Valves 14 Radiation Exposure 15
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PLANT DESIGN CHANGES The following list by design change number summarizes those design changes completed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.
A summary of the safety evaluation is included for each change. None of the changes were determined to constitute an unreviewed safety question.
PDCR 2-133-79 Removed the frequency detection circuitry from all of the DC to AC inverters whose alternate source was not another inverter. Most of the time that these inverters had been out of service could be attributed to the failure of the circuitry, and unless the alternate source for an inverter was another inverter, the circuitry was not needed. The change did not affect the ability of the electrical distribution system to perform its intended function, but it did increase inverter reliability.
PDCR 2-17-80 Changed the degasifier hotwell tube bundle material as follows:
a.
Tube material changed from SA 213 TP 304 to SB 163 GR 600, b.
Tube sheet is Inconel Clad SA240 TP 304.
The material selection was recommended by the vendor and should improve equipment reliability. The design change did not affect any safety related system or systems covered in the basis of the Technical Specifi-cations.
i
[
PDCR's:
2-52-80, 2-53-80, 2-54-80, 2-55-80, 2-56-801 i
2-57-80, 2-58-80, 2-59-80, 2-60-80, 2-61-80, 2-62-80, 2-69-80, 2-73-80, 2-74-80, 2-77-80, 2-73-81, 2-118-81, 2-120-81, and 2-121-81.
Installed, modified or repaired various pipe supports as required. Each r.
PDCR number reflects the work completed within a system boundary. These changes upgraded the seismic qualification of affected pipe hangers from a safety factor of 2 to a safety factor of 4 as required by IE Bulletin 79-02.
This resulted in increasing the piping system integrity throughout the Unit. The hanger designs conformed to previously approved and l
existing standards. The changes were implemented within the requirement of the Technical Specification Limiting Conditions for Operation.
- PDCR 2-76-80 Upgraded the auxiliary feedwater flow indication to safety grade level as required by NUREG 0578.
It consisted of replacing the remote indicators and providing power for the instrument channels from the emergency buses.
The change increased the margin of ' safety and reliability by providing i,
separate and redundant auxiliary feed flow indication for each steam 1
generator, and serves to directly verify the performance of the auxiliary feed system.
PDCR 2-87-80 Installed / modified structures necessary to support the reactor vessel head area cabling. The change consisted of the following:
a.
Attachment of the head area cable support structure (HACSS) to the 38-6 elevation of containment.
b.
Support and separate the head area cables from the HACSS disconnect panels to existing cable trays.
c.
Modification of the shootout steel to facilitate the installation of the HACSS.
The structures have been analyzed and designed for all applicable loading conditions including the safe shutdown earthquake.
Neither the function nor margin to safety of any of the existing structures was altered by these changes.
PDCR 2-91-80 Replacement of the existing reactor vessel head area cabling and connectors with cable / connector arsemblies that are compatible with HACSS systems.
The change was a one for one replacement and conformed to the units original design criteria.
PDCR 2-116-80 Installed two redundant high range safety grade radiation monitors inside the containment as required by NUREG 0578. The modification consisted of seismically mounting the monitors inside containment and installing the necessary cabling / electronics to provide remote indication in the control room.
Since this change was independent of all existing equipment, any failure would not affect any other safety related equipment.
PDCR 2-134-80 Modified the circuitry controlling operation of the 4.16 and 6.9 KV switchgear compartment heaters to keep them continuously energized. The change was recommended by the vendor and serves to maintain the compartment temperature above ambient thus preventing moisture from condensation.
This PDCR would not prevent the load centers from performing their design function, but will increase their long term reliability.
PDCR 2-139-80 and 2-152-80 Revised the control circuits for various safety related equipment to prevent the equipment from changing state or position upon reset of the Engineered Safety Features signal. These modifications were in response to IE Bulletin 80-06.
The designs were reviewed with respect to single failure, physical separation, system independence, fail safe and the
ability of the equipment to accomplish its safety function, and no adverse interactions were found.
PDCR 2-142-80 and 2-142-80 Addendum 1 Installed penetration extension pieces and rubber boots on three pipes that enter the Auxiliary Building through the watertight boundary in the Refueling Water Storage Tank pipe trench. The change upgraded the boundary to meet the design requirements of Technical Specifications 5.1.3.
It did not adversely affect the safety related or other piping passing through the penetration.
PDCR 2-145-80 Floated the shield on the cable between the pulse shaper and the pulse transmitter in all four channels of the reactor coolant pump (RCP) speed sensing system. The change broke a capacitive coupling which tied the protection calculator's signal common to the equipment ground when the reactor protection system test cable was used during surveillance testing.
The common to ground tie altered the calculated setpoints, causing erroneous pre-trip and trip signals in the channel under test.
This modification was recommended by the vendor and will not prevent the RCP speed sensing system frc, processing an actual trip signal to the reactor protection system.
PDCR 2-149-80 Set the main turbine generator Underexcited Reactive Ampere Limit (URAL) at zero MVAR on the automatic voltage regulator, disabled the Swtichyard Power Monitor Relays for the 371 and 383 lines (Montville and Card Street Lines) and set the Switchyard Power Monitor Relay at 1200 MWe for the 348 line (Southington Line). The URAL setting change would automatically prevent the production of leading MVAR's.
The power monitor relay changes would automatically trip the Millstone 1 generator only if the 348 line is carrying the station output at a total load greater than 1200 MWe. The changes did not affect any safety related equipment, nor would they degrade the offsite power sources.
PDCR 2-150-80 Installed a round strainer drain fixture with a backwater (check) valve in each of the emergency diesel generator room floor drains. The change eliminated the use of floor drain plugs to present concurrent flooding of both diesel generator rooms resulting from a service water pipe rupture or fire system actuation. This was consistent with facility separation while maintaining the area drains in rooms containing safety related equipment open.
PDCR 2-158-80 Installed a roof mounted exhaust fan and missile protection shield to serve the east 480 volt switchgear room.
It was necessary to modify the ventilation in this switchgear room to maintain a negative pressure relative to the adjoining Control Room. The installation was in accordance
with the applicable codes and standards, and the design conforms to the criteria as specified in the FSAR. This PDCR only covers the mechanical installation of the fan and missle shield. The electrical portion will be addressed by a separate PDCR.
PDCR 2-161-80 Modified the reactor coolant system low temperature protection bistable to prevent actuation of a power operated relief valve during a loss of power. This removed the potential for a loss of coolant accident resulting from a single electrical failure when low temperature /over pressure protection was selected.
The change requires operator action to establish a reactor coolant system vent during a complete loss of power; however, this is acceptable and consistent with the Technical Specifications action-statement time requirements.
PDCR 2-2-81 Changed the material specifications of some of the replacement parts for the main steam safety valves. The changes were primarily specification updates as well as some material changes based on vendor experience. A comparison of the mechanical properties between the new and old materials show that they were consistent. Thus, the use of the new materials would not prevent nor impair the performance of the main steam safety valves.
J PDCR 2-3-81 Relocated the electrical penetration leads for TE 122 IIC from module F to module A within penetration ED3. This removed the only energized circuit from module F to permit insulation resistance testing of the wires in that module during plant operation. The change was implemented in accordance with previously approved procedures and retested to verify continuity and resistance to ground.
PDCR 2-7-81 l
Disconnected the electrical penetrations from the station nitrogen distribution system, and repiped as necessary to supply the penetrations from a bottled nitrogen source. This eliminated the potential for a fluid system cross contamination of the nitrogen system from entering the electrical penetrations. This change returned the penetration nitrogen system to its original design.
PDCR 2-11-81 Provided the capability of obtaining a grab sample from the waste gas decay tanks, volume control tank and waste gas surge tank without disrupting the sample flow to the oxygen analyzer. The change consisted of installing a branch connection for each sample line from the oxygen analyzer to a new hooded sample sink. The oxygen analyzer and resulting samples are not safety related nor are they used for a safety related purpose. The modification conformed to existing design standards including Seismic Class 2 over Seismic Class I criteria. The new sample sink was installed l
in and vented to the same location as the hydrogen and oxygen analyzers; thus, no new discharge paths were created.
PDCR's:
2-18-81, 2-23-81, 2-27-81, 2-29-81, 2-32-81, 2-34-81, 2-43-81, 2-60-81, 2-61-81, 2-63-81, 2-67-81, 2-89-81, 2-91-81, 2-93-81, 2-95-81, and 2-96-81.
Replaced selected mechanical snubber installed in all of the various safety related piping systems. The changes consisted of replacing inoperable International Nuclear Corp. type snubbers identified during IE Bulletin 81-01 inspections with Pacific Scientific Application type sonhbers. The replacements were equivalent or exceeded the design requirements for travel and maximum loading, and they were functionally tested for freedom of movement prior to installation.
Any required attachment modifications conformed to existing design standards.
Implementation of each PDCR was controlled by ada.inistrative procedures in accordance with the appropriate Technical Specification LCO's and Action Statements.
PDCR 2-21-81 Installed an overspeed trip test device on the Woodward governor for the steam driven auxiliary feed pump.
This provided an easier method of testing the turbine overspeed trip setpoint by utilizing the turbine governor. There was no failure mechanism in the device which could result in the inoperability of the turbine.
PDCR's:
2-30-81, 2-35-81, 2-35-81 Addendum 1, 2-36-81, and 2-50-81 Removed selected mechanical snubbers installed in the Service Water and Emergency Core Cooling Systems. The changes consisted of replacing inoperable International Nuclear Corp. type snubbers identified during IE Bulletin 81-01 inspections with rigid supports, and in the case of PDCR 2-36-81, not replacing the snubber removed. Removal of the mechanical snubbers decreased the potential for failure; hence, increased system reliability.
In all cases, system stress analysis demonstrated that the piping and support stresses remained below the allowable. Rigid restraints were designed in accordance with the appropriate mechanical standards and codes.
Implementation of each PDCR was controlled by established administrative procedures in accordance with the Technical Specification LCO's and Action Statements.
PDCR_2-45-81 Changed the annunciator logic for the reactor trip switchgear bus tie breaker (TCB-9) so that the anaunciator was alarmed when the breaker was closed vice open. TCB-9 was a normally open breaker, and the change was consistent with the philosophy of minimizing the number of alarms annunciated during steady stace operation. A spare contact in TCB-9 was used to reverse the annunciation logic. The modification would not impair or prevent TCB-9 from performing its safety related function.
6-PDCR 2-46-81 Installed a new hanger on the low pressure safety injection header, 12" - GCB-2.
A piping system review indicated that the dead weight loads resulted in some high stress conditions. The addition of the hanger relieved the dead weight stresses; thus, system integrity and reliability was improved. Stress analysis demonstrated that the system acceptability for all design conditions.
PDCR 2-51-81 Installed additional supports at two locations on the shutdown cooling system piping. The supports were necessary to facilitate routine functional testing of the snubbers at these locations without declaring that portion of piping inoperable. The supports were designed and installed in accordance with the unit design standards and applicable codes.
PDCR 2-65-81 Modified a seismic support on the reactor vessel head vent system. The hanger as originally installed was designed incorrectly, and this PDCR provided the mechanism to obtain the appropriate reviews prior to the installation of the redesigned hanger. The modified support meets all of the system design loadings, and will maintain the piping stresses below code allowables.
PDCR 2-66-81 Installed necessary cabling in containment to supply the linear range nuclear instrument channel Y detector signal to the Reactor Protection System (RPS) channel D.
Since the RPS channel D linear range detector had failed, this change restored the operability of the RPS channel until such time that the failed detector could be replaced. The core ph sics considerations with respect to the Technical Specifications LCO's i ad LSSS's were evaluated as acceptable. The change conformed to the safety channel separation requirements of the original Unit design.
Installation was in accordance with previously approved procedures. The channel Y input to the Reactor Regulating System (RRS) was deleted. Since the RRS performs no safety function and channel X can provide necessary control /indi-cation functions, this was acceptable.
PDCR 2-98-81 Opened the High Pressure Safety Injection (HPSI) System motor operated injection valves to their required throttling position. The motor i
I operators for these valves could not be relied upon to automatically position the valves; thus, the change eliminated this reliance by presetting the valves. The change enhanced HPSI system reliability. The piping system configuration (installed check valves in series, available instru-mentation, etc.) along with established procedures sufficiently addressed intrasystem LOCA considerations-Since the injection valves were maintained in their accident position and performance of other required surveillance on the HPSI system was not prevented, the safety margin in the bases of the Technical Specifications was not reduced.
PDCR 2-119-81 Blocked the Unit 2 control room ventilation supply to the Unit I control room offices. The area serviced by the Unit 2 Control Room Ventilation System was expanded, and the change was necessary in order to maintain proper system balance. The modification did not adversely affect the Control Room Ventilation System or its ability to perform any design function. The change did conform to the original design standards and did not affect any other safety related system by either function or location.
s
8-PROCEDURE ClfANGES There were no procedure changes as listed in the FSAR during 1981.
-9
!~
e TESTS s
~
.The following list by test' number summarizes those tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.
A summary of the safety evaluation is included for each test. None 6f
'the tests were evaluated as an unreviewed safety question.
i.
T79-4 Rhodium Detector Sensitivity Depletion Test - The test objective was to obtain accurate rhodium detector sensitivity factors. 'It required the installation of calibrated test detectors for comparison with inservice detectors. -This test did not impair the ability to monitor linear heat rate nor did it affect the functioning of any safety related system.
T81-1 Acoustic Valve Monitoring System'(AVMS) Operational Test - This test demonstrated that-the AVMS would respond properly when one Power Operated Relief Valve (PORV) was opened. Each PORV and it ssociated blocking-valve has been designed for full operating pressu and temperature conditions for operability. A PORV challenge wou) not result in increased accident probability considering the_ availability c: the blocking valve and single failure criteria.
1 i
T81-2 Diesel Generator A Response to Loss of 125VDC - This test was performed-by securing the 125VDC control power to the diesel and observing the response.
It did confirm that the diesel did respond properly during the j
loss of.125VDC incident on January 2, 1981. This test was performed in Mode 5 which did not require operability of this diesel. Operability was i
verified in accordance with Technical Specifications following completion of the test.
T81-3 J
Electrical Penetration Megger Checks - This test conducted insulation resistance checks or random penetration wires to determine if there were any short term effects of moisture introduction into the electrical l
penetrations during the January 7,1981 RCS/ Nitrogen cross contamination-j incident. A11'megger checks were performed on spare modules, and module j
integrity was not affected either_ physically or electrically.
T81-5 i
Nitrogen System Flush - This test controlled and documented the flushing of the carbon steel piping in the nitrogen system.
It removed any boric acid contamination that may have originated from the primary system during the cross contamination incident..The flush pressure was maintained J
within system design limits and flush water was processed through the I
normal radwaste system. The procedure did not affect any safety related i
systems.
I
- e T81-6 Electrical Penetration Relative llumidity Test - This test demonstrated that after moisture removal efforts were completed the relative humidity of all electrical penetrations were less than original design specifications.
All testing parameters were within design limits and the procedure closely paralleled Local Leak Rate Testing procedures.
T81-7 No. 1 Safety Injection Tank (SIT) Paint Adhesion Test - This test determined the adhesion properties of the paint affected by the therril transient on No. 1 SIT.
It did not affect the structural integrity of the SIT or any other safety related component. The effects on the coating of the SIT was negligible.
T81-8 SIT (1 through 4) Depressurization and Repressurization - This test was performed for data accumulation and demonstrated that residual boric acid in the nitrogen supply piping to the SIT's had little effect on the leak tightness of the SIT nitrogen valves. The conduct of this test did net deviate from any conditions of approved operating procedures. The nitrogen vented from the SIT's was processed through normal radwaste systems.
T81-9 Pressure Test of No. 1 SIT - This test demonstrated the structural integrity of No. 1 SIT when pressurized to approximately 250 psig. All test parameters were within the design limit of the SIT and other related systems.
It was performed out of mode requiring SIT operability.
T81-10 Electrical Penetration IR Testing - This test periodically meggered selected wires in various electrical penetrations to demonstrate there was no long term moisture degradation of the electrical penetrations.
The IR testing was conducted on spare penetration wires only and did not affect any required operable safety circuits.
It was in conformance to penetration design and would not result in the degradation of any penetration module.
T81-11 Auxiliary Feedwater Motor Driven Pump Test - This test demonstrated that the A Auxiliary Feed Pump could continuously operate for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> without exceeding established acceptance criteria.
It was performed in accordance with approved operating procedures. The injection of auxiliary feed during power operation did not result in any significant effects on the steam generator or feed inlet piping. The margin in Technical Specifications bases was not reduced, since the operability of the Condensate Storage and Auxiliary Feedwater Systems was not affected.
e.
T81-12 2-JH-155 Flow Test - This test verified that the A Boric Acid Pump discharge check valve could meet the Inservice Inspection (ISI) testing flow requirements.
It also provided data to allow revision of the ISI test procedure acceptance criteria. Performance did not increase the probability or create the possibility of an unanalyzed accident, e.g.,
boron dilution. Since the minimum boration flow paths were maintained, the bases of the Technical Specifications were not affected.
T81-13 RM 202, CVCS Letdown Monitor Recalibration - This test accumulated data to allow calibration of the letdown failed fuel monitor to respond to changing concentrations of Rb-88 vice Co-60.
It did not affect any safety related piping or instrument systems. The instrument affected provided no control functions nor did it have any environmental discharge path.
T81-14 Venting of the Reactor Head Vent Piping - This test provided the procedure and the vehicle for a safety evaluation to operationally depressurize the head vent piping.
It involved isolating the head vent solenoid valves from the reactor coolant system then electrically opening the solenoid valves to depressurize the piping downstream of the isolation valve. The consequences of this test were no different than those of a single failure of the piping system prior to isolation.
T81-15 Mechanical Snubber Twist and Stroke Test - This test was performed to verify the operability of safety related mechanical snubbers. After completion of the testing, it was determined that the twist test was not a valid indication of freedom of movement. Subsequent operability verification was controlled and documented by approved administrative procedures. All test conditions were within the Limiting Conditions for Operation of the Technical Specifications.
T81-16 D. C. Switchgear Room Temperature Surveillance - This test verified that the temperatures in D.C. Switchgear Room remain below established acceptance criteria following a realignment of the ventilation system.
It did not result operation of any plant systems other than by previously approved operating procedures, and was performed for data accumulation only.
T81-18 Condensate Polishing Facility (CPF) Sodium Throw Test - This test was performed to collect information on sodium concentration in the effluent of a condensate polishing demineralizer at the ammonia break point. The data was used in evaluating the operating philosophy of the CPF. The i
steam generator chemistry was maintained within the limits required by the Technical Specifications and recommended by the vendor.
T81-19, T81-23 and T81-28 Corrosion Test at Increased pli Values - This test increased the ammonia concentration (hence pil) in the condensate and feedwater systems to evaluate these effects on secondary side corrosion rates. Steam generator chemistry was maintained within the limits required by the Technical Specifications and recommended by the vendor.
T81-21 Nitrogen to Steam Generator Flush - This test controlled and documented the flushing of the contaminated piping that supplies nitrogen to the steam generators. Flush pressure was maintained within the system design limits and performance of this test did not affect any safety related system.
T81-22 LPSI Pump Test with Shutdown Cooling Established - This test accumulated base line data for use in evaluating LPSI pump performance during high flow conditions.
It was performed in accordance with established operating procedures.
T81-25 Auxiliary Building Ventilation Effectiveness - This test was performed for information only to assess the effectiveness of the ventilation system. The results were used to optimize the ventilation and establish / record a set of base line values.
T81-26 and T81-27 Makeup Isolation for Oxygen Determination - This test was performed for information only to determine the contribution of condensate storage makeup to the dissolved oxygen in the condensate system.
It was conducted such that no safety related system was affected in any way.
T81-30 Response Time Test of Containment Pressure Transmitters, PT 8238 and Pt 8239 - This test was performed to acquire data on these transmitters prior to their installation.
When in use, these transmitters provide only indication and no control function. The test was used to formalize the methods to accumulate the data and to provide a mechanism for its retention.
T81-31 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.
c
- o STEAM GENERATOR 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 Specifications 4.4.5.1.5.b.
An inservice eddy current inspection and dent assessment program utilizing multi-frequency eddy current testing commenced on December 19, 1981 and had not been completed by December 31,-1981.
The inspection was performed by Combustion Engineering Power Systems Group, System Integrity Services Personnel.
The inspection was performed in accordance with Combustion Engineering Test Procedure No. 00000-SIS-012-0 and satisfied the requirements of Nuclear Regulatory Commission Guide 1.83, Revision 1 and the Technical Specifications.
The inspection results'as of December 31, 1981 are as follows:
Steam Generator No. I 1911 tubes inspected near hot and cold leg tubesheet 950 tubes inspected near hot leg tubesheet only 228 tubes degraded greater than 20% of tube wall thickness 86 tubes degraded greater than 40% of tube wall thickness Steam Generator No. 2 1623 tubes inspected near hot and cold leg tubesheet 730 tubes inspected near hot leg tubesheet only 102 tubes degraded greater than 20% of tube wall thickness 84 tubes degraded greater than 40% of tube wall thickness The additional inspections as required by the Technical Specifications was completed in January 1982. The complete inspection results will be provided in the 1982 Annual Report.
i
- o CHALLENGES TO REACTOR COOLANT SYSTEM RELIEF VALVES In accordance with Item II.K.3.3 of the TM1 Action Plan, the following is a listing of challenges to the Power Operated Relief Valves (PORV's) during 1981. There were no challenges to the Reactor Coolant System Code Safety Valves.
In each case, all systems responded as designed.
DATE CIRCUMSTANCES January 2, 1981 During the recovery following a reactor trip, the reactor coolant system was being maintained at normal operating temperature and pressure.
The operating reactor coolant pump combination did not provide effective spray and system pressure was controlled by varying pressurizer level. Approximately two hours after the trip system pressure increased to approximately 2380 psia, and both PORV's opened for a short period of time. The open duration was not sufficient to trigger the acoustic valve monitors. The pressure was decreased and controlled manually by_the use of the auxiliary spray valve.
January 18, 1981 The plant was in hot standby at normal operating temperature and pressure.
One PORV (2-RC-402) was failed open as part of Inservice Test T81-1 to verify operability of the Acoustic Valve Monitoring System. The other PORV was deenergized closed with its associated blocking valve shut.
July 31, 1981 During normal operation at rated power, the main turbine stop valve closed. This resulted in a reactor trip due to high pressurizer pressure and the opening of both PORV's. After the trip normal pressure was automatically restored by pressurizer spray, and both PORV's shut.
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