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LER 88-001-00:on 880120,reactor Scram Occurred Due to Actual Low Water Level Condition.Caused by Design & Personnel Errors.Operator Disciplined & Mod Addressed Inadvertent Feedwater Control Valve lockup.W/880217 Ltr
	ML18038A389
Person / Time
Site:	Nine Mile Point
Issue date:	02/17/1988
From:	Ronaldo Jenkins, Lempges T; NIAGARA MOHAWK POWER CORP.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	CON-IIT07-747-91, CON-IIT7-747-91 LER-88-001, LER-88-1, NMP29854, NUREG-1455, NUDOCS 8802230093
	Download: ML18038A389 (400)
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~ ACILITY NAMI 111 Nine Mile Point Unit 2 0 5 0 0 0 410 5 OF TITLI 141 AIRBOAT Reactor Scram due to a Loss of Feedwater Flow Caused by Pel sonnel Error IVINT OATt 1$ 1 LIA NUMIIN III O*TI lrl OTHIA 4ACIL(rillLNVOLVIO III MONTM Oxv vtAA >tAA ~ IOVIHvroj AIVNr(ov VONTI4 OAV VIAA roCILrTV rrAMII OOCXIT vvMIIA(51 VVrri~ 1 rrVM441 N/A 0 5 0 0 01 01 20 88 88 001 00 02 l7 88 N/A 0 5 0 0 OS OotAATINO TNII 1IOOAT II SVIMITTIOOVAIVANTT0 TNt AtOVIAIMINTIO) 10 C) A i( IC4444 o~ 4 ~~ or rrr rjoo~ro( I>>

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'Level 3! water level condition. The low water level was caused a loss of feedwater 'low to the reactor. An operator while placing a markuoby (tag out),

va'veo <<n instrument air system orefilter out of service without ensuring that

<edundant pr ef i 1 ter was in ser vice. Thi s i so ated the air comoressors f rom 1

he emainder of the system. Instrument air oressure throughout the olant decayed to the point of causing the minimum flow valves for the condensate, condensate booster, and feedwater pumps to fail ooen. As a result,, reduced feed(vater flow to the reactor caused reactor water lev'el to raoidly decrease to the Level 3 scram setpoint (159.3 inches). 8oth the Hiah Pressure .Core Soray
{HPCS) and Reactor Core Isolation Cooling (RCIC} systems actuated on low level instrumentation which restored norma) water level in the vessel, The Main Turbine-Generator also tripped due to RCIC injection.
Oesign and personnel errors were encountered during the event. Corrective actions have been imolemented and/or incorporated to minimize the potential of a recurrence.
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XIA1 SIOUIATIAL U A ItVM'4 lO H 410 88 00 02 oe 08 Nine Mile Point Unit 2 o s o o o a~~O~ ~OWaer~IOAema+{m I. OESCR IPTION OF EVENT On January 20, 1988 at 0944 hours0.0109 days 0.262 hours 0.00156 weeks 3.59192e-4 months with the reactor operating at approximately 414 oower in the natural circulation mode per startup and test procedures, Nine Mile Point Unit 2 (NMP2) experienced a scram due to an actual low (Level 3) water level condition. The low water level was caused by a loss of feedwater flow to the reactor.
A Niagara Mohawk operator, while placing a markup, valved Instrument Air System (IAS) prefilter 2A out of service without ensuring that the redundant orefilt r 28 was in service. Instrument air oressure throughout the plant began to decay due to this valving error. The licensed operators first noticed the ACRO Scram Valve Pilot Air Header Pressure Low" annunciator and immediately began to investigate. As air pressure decayed, the air operated minimum flow valves for the condensate, condensate booster, and feedwater pumps failed ooen (Figure 1). As a result, feedwater flow recirculated back to the condenser subsequently reducing feedwater flow to the vessel.
Operating feedwater pumps (P18 E PlC) and condensate booster pump (P2A) tripped on low suction pressure. Condensate pumps PlA, P18 and PlC and condensate booster pump P28 remained running throughout the event. An Operator ol aced the master feedwater controller in manual in an attempt to restore water level. ~e soon realized that the feedwater valves (LV108 8 C) were fully ooen and that no flow was going to the reactor. The reactor scrammed on Low (Level 3) water level, 159.3 inches. Reactor level continued to decrease until level reached 112 inches. At which time, the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling system actuated and restored level.
Instrument ai r was returned to normal approximately one minute after the scram.
HPCS, the HPCS Oiesel Generator, and RCIC systems initiated on a low-low water level signal (Level 2), 108.8 inches. The Main Turbine-Generator tripped due to RCIC injection. Reactor water level was restored and HPCS injection was terminated, by the operators, when level reached 195 inches. RCIC at Level 8 ( 202 . 3 inches), as designed. The reactor pressure was secure.'utomatically reduced during this time, due to the injection of cool water from the HPCS and RC IC systems, and the relative ly small amount of decay heat in the new core.
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V.t. NVCLtAN ASOVLATONY CO&HI~ON LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Ati+OVtO OIJ ~ NO SISO&l04 tXtlttt:tISIIN PAClt&T NANO III OOCKtT IM&tNQI LIN NIAAStN I ~ I iAOt ISI ltOVSNtiiI, iEVNIOH U V Nine Mile Point Unit 2 o s o o o 410 88 00 03 OF H~ ~ 4 INN44C m eaMaV ~ ANN ~WIST Approximately 5 minutes after the scram, the feedwater minimum flow valves closed from a normal control circuit signal. Feedwater system pressure increased to the point of exceeding reactor pressure due to the operating condensate and condensate booster pumps. With the master feedwater controller in manual and the level control valves fully open, feedwater flow resumed to the vessel. Operators later noticed that reactor water level was starting to increase without HPCS or RCIC injecting. An operator manipulated the 4Ianual/Auto feedwater controllers to the full closed position in an attempt to prevent excessive overfilling of the reactor. The feedwater regulation valves (FWS-LV108, C) locked up at approximately 8(K open, when a loss of control" signal was sensed by the valve hydraulic actuator control card. This caused the vessel to overfill and flood the main steam lines. The operators determined that feedwater was still injecting into the vessel and closed the feedwater containment isolation valves (MOV21A 5 8). This terminated the feedwater irjection and water level peaked at 333 inches.
The main steam lines were flooded up to the turbine stoo valves. Normal steam line drains were utilized to drain the excess water fror the steam lines . A turbine bypass valve was later opened manually to control reactor pressure and to initiate a plant cooldown.
The remainder of the scram recovery was routine.
II. CAUSE OF EVENT A root cause analysis for this event has been completed per Site Supervisory Procedure S-SUP-l, "Root Cause Evaluation Program". The root cause has been determined to be. personnel error. The operator visually verified that the isolation valves for prefilter 28 were open by observing the length of the valve stems. However, he should have physically verified the valves'csition open. These valves were in fact closed. Thus, when the operator isolated orefilter 2A for maintenance, the instrument air compressors were isolated from the rest of the system.
The root cause of the feedwater regulation valves locking uo was design deficiency. An existing ground loop in the contro i circuitry biased the setooi nt upward for the loss of control signal, causing lock-up of the f:edwater regulation valves.
III. ANALYSIS OF EVENT The isolation of the air compressors posed no adverse safety consequences to olant personnel or public safety as a result of this event. This portion of the IAS is not required to affect or support the safe shutdown of the reactor or to perform any safety-related functions associated with its operatior The loss of instrument air caused the minimum flow valves for the condensate, condensate booster, and feedwater pumps to fail open allowing flow to recirculate back to the condenser as designed (FSAR, Section 15.0.5, "Loss of Instrument Air").
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Nine Mile Point Unit 2 ibm ~ W I trI 0 I a o o 410 88 001 04 oi 08 The reactor scram which occurred as a result of the Level 3 trip was a conservative action. This poses no adverse safety conseauences: any power level. This action did not in any way adversely affect any other safety system nor the operators ability to achieve safe shutdown.
tAater level never fell below 112 inches during the transient. HPCS and RCIC initiated well above (conservative direction) the 108.8 inch Technical Specification (TS) required setpoint. 8ecause water level never reached 108.8, some Level 2 system actuations and isolations did not occur. An investigation was performed to ensure the setpoints of all Level 2 instruments were correct.
This investigation found that the guide value setooint for Primary Containment Isolation Actuation Level switches had been calculated incorrect)y. Oetai1s can be found in I.ER 88-04.
The automatic actuation of RCIC and HPCS with coolant injection was a conservative response with minimal p1ant impact and no resultant, impact on public safety, The systems operated as designed and r'estored reactor water level to Level 8, 202.3 inches.
In accordance with the requirements of TS iections 3.5. 1(f) and 6,9.2 (Emergency Core Cooling System (ECCS) Injections, the following special data is orovided:
For the HPCS nozzle,
- Total accumulated initiation cycles to date = 1
- Current usage factor value remains well below 0.70 The increasing reactor water level which resulted in the flooding of the main steam lines prompted the following concerns:
Thermal transients 'across the main steam pioe and reactor main steam nozzle walls caused by assumed colder water flowing into the hot main steam line.
The weight of water in the main steam lines.
Oynamic transient 1oads caused by water flowing into the main steam 1ines.
External loads/moments imposed on the reactor main steam nozzles.
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ACR~ NAZI I l I OOCCIT IN%III Ql LI1 NIP%41 lll ~ AOI lIl YIAA IIOMINT.AL AIVNIOII U IA 1 Nine Mile Point Unit 2 o s o o o 410 88 05 ov 08 r~~oeeeec weaewmaamaw<m Evaluation of the local stress due to the thermal transient across the pipe nozzle walls conservatively used the temperature of the feedwater as the fluid temperature entering the main steam pipe. This was conservative since it did not account for thermal mixing within the Reactor Pressure Vessel (RPV) due to the natural circulation of reactor water, the effect of core decay heat, or the effect of stored thermal energy in the reactor coolant pressure boundary metal piping, structures, and the reactor core. The results of the evaluation of the thermal transient indicated that all stresses satisfied design code allowables for the piping from the vessel to the main turbine stop valves and drywel 1 oenetrations. Pipe break exclusion criteria was also satisfied for this event.
Final evaluations to establish the actual usage factor for plant life will use fluid temperatures calculated by a Contractor Engineering group for this event.
Evaluation of the weight of water in the main steam piping indicated that the design pipe stress analysis included load cases considering water in the main steam lines, therefore, no further review was required. Pipe supports were designed for maximum design loading considering either water or steam.
Snubbers and spring hangers were sized to account for the maximum movement due to the combination of plant thermal, dynamic, and deadweight (water or steam) conditions.
Review of linear potentiometer displacement data in the ftain steam piping indicated that there were no vibratory responses during the flooding event.
From this it can be concluded that hydraulic transient affects were insignificant.
The only external 'loads/moments imposed on the main steam piping, components, and reactor nozzles was the weight of water and thermal expansion of the system. Since no seismic event'ccurred during the scram and subsequent plant shutdown, the applied loads were enveloped by the design loads'herefore, no further l valuation was required.
In conclusion, we have conservatively determined that the effects of thermal transients across the pipe and reactor main steam nozzle walls, the weight of water, and hydraulic dynamic transients, are within applicable code stress allowables for the piping and penetr ations and within design pipe support loadings. Break exclusion criteria is satisfied wi.ere applicable and RPV nozzle loadings due to this event are enveloped by existing design loads.
In summary, while the main steam line flooding and the loss of instrument air was an vndesirable transient, effects of the transient were within plant design margins.
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Nine Mile Point Unit 2 0 5 0 0 0 410 88 001 00 06 OF 08 e'~~Orreal ~~eeaee~WIm IV. CORRFCTIVE ACTIONS The individuals involved with this event participated in the LER investigation and follow-up report.
2. The operator involved in the valving error has been subjected to disciplinary action for inattentiveness to duty.
3. This event has been incorporated into the Operations Oepartment Lessons Learned Program.
4, Administrative. Procedure AP-3.3.1 has been revised regarding valve position verification policy.
5. The valve position verification policy will be incorporated into the operators requalification traini'ng program via THR 02-88-18.
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6. Modification PN2Y87MX238 - This modification incorporated a design change to the control circuitry of the manual/automatic controllers of the feedwater regulation valves. This change addresses the problem of inadvertent feedwater control valve lockup.
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7. Operating Procedure No. N2-0P-19, " Instrument and Service Air System" has been revised to provide additional direction in an event of loss of instrument air.
Operating Procedure No. N2-0P-101C, "Plant Shutdown" has been . A caution statement has been added to alert the operators to continuously monitor reactor water level if the feedwater regulation valves are in manual.
9 A review of the Nuclear Steam Supply Shutoff System (NSSSS) level instrumentation revealed that four level switches were found to have improperly calculated guide values. The Instrument and Control (IEC)
Oepartment imediately recalibrated these instruments and incorporated procedure changes as necessary. Additional information is contained in LER 88-04.
10. The guide value calculations f'r all other level, pressure, and temperature instrument channels which have TS required trip setooints have been reviewed by ILC to assure accuracy. Additional information is provided in LER 88-04.
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}1IOa1OI IXIIAIl:V$1I OOCACT IKNIll Ql Lt1 NLANIC II1 ~ AOl III YIAA IIQVIAIIAL +CVCK)A U u A Nine Mi'le Point Unit 2 o 5 o o o 410 88 00 07 OF 08 H~~OIICICIC<<~~Aee~VWl V. ADDITIONAL INFORMATION Identification of Components Referred to in this LER IEEE 803 IEEE 805 Component EI IS Funct System IO Instrument Air System (IAS) N/A LD High Pressure Core Spray System (HPCS) N/A BG Reactor Core Isolation
'Cooling System (RCIC) N/A BN Main Steam System (MSS) N/A SB Reactor Pressure Vessel RPV N/A Reactor Water Cleanup System (RWCU) N/A CE Instrument Air Prefilter FLT LD Feedwater Pump P SK Condensate Booster Pump P SG Turbine Stop Valves ISV TG Turbine Bypass Valve ISV TG Valve ISV LD Annunciator ANN 19 Manual/Auto Feedwater Controller FCO JB There has been one previous similar event where reactor water entered the MSS lines, LER 86-20.
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41'4 power, Nine H11e Point Unit 2 experienced a scram due to an actual low (Level 3} water level condition. The low water level was caused by a loss of 0 ~ eedwate1 flow to the reactor. An operator while placing a markup (tag out},
valved an instrument air system prefilter out of service without ensuring that a redundant prefilter was in service. This isolated the air compressors from he remainder of the system. Instrument air oressure throughout the plant decayed to the point of causing the minimum flow valves for the condensate, condensate booster, and feedwater pumps to fail open. As a result:, reduced feedv(ater flow to the reactor caused reactor water level to rapidly decrease to NO the Level 3 scram setpoint (159.3 inches). Both the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) systems actuated on low level Inst~umentation which restored normal water level in the vessel. The Llain Tu(bine-Generator also tripped due to RCIC injection.
Design and personnel errors'were encountered during the event. Corrective actions have been implemented and/or incorporated to minimize the potential of a recurrence,
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~ ACILITY ISAISt III DOCÃtT IILANtlll1l Lth IIIA%tAItl ~ AOl ISI VSAA tlOUSIITIAL IISVOION ll A Nine Mile Point Unit 2 o s o o o 410 88 02 OF 08 e~~a~ ~~~~~wlm I. OESCRIPTION OF EVENT On January 20, 1988 at 0944 hours0.0109 days 0.262 hours 0.00156 weeks 3.59192e-4 months with the reactor operating at approximately 41%, oower in the natural circulation mode per startup and test procedures, Nine Mile Point Unit 2 (NMP2) experienced a scram due to an actual low (Level 3) water level condition. The low water level was caused by a loss of feedwater flow to the reactor.
A Niagara Mohawk operator, while placing a markup, valved Instrument Air System (IAS) prefilter 2A out of service without ensuring that the redundant prefilter 28 was in service. Instrument air pressure throughout the plant began to decay due to this valving error. The licensed operators first noticed the "CRO Scram Valve Pilot Air Header Pressure Low" annunciator and immediately began to investigate. As air pressure decayed, the air operated minimum flow valves for the condensate, condensate booster, and feedwater pumps failed open (Figure 1). As a result, feedwater flow recirculated back to the condenser subsequently reducing feedwater flow to the vessel.
Operating feedwater pumps (P18 8 PlC) and condensate booster pump (P2A) tripped on low suction pressure. Condensate pumps PlA, P18 and PlC and condensate booster pump P28 remained running throughout the event. An Operator placed the master feedwater controller in manual in an attempt to restore water level. He soon realized that the feedwater valves (LV108 8 C) were fu lly open and that no flow was going to the reactor. The reactor scraped on Low (Level 3) water level, 159.3 inches. Reactor level continued to decrease until level reached 112 inches. At which time, the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling system actuated and restored level.
Instrument air was returned to normal approximately one minute after the scram.
HPCS, the HPCS Oiesel Generator, and RCIC systems initiated on a low-low water level signal ( Level 2), 108.8 inches. The Main Turbine-Generator tripped due to RCIC injection. Reactor water level was restored and HPCS injection was terminated, by the operators, when level reached 195 inches. RCIC at Level 8 (202.3 inches), as designed. The reactor pressure was secure.'utomatically reduced during this time, due to the injection of cool water from the HPCS and RCIC systems, and the relatively small amount of decay heat in the new core.
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Aoproximately 5 minutes after the scram, the feedwater minimum flow valves closed from a normal control circuit signal. Feedwater system pressure increased to the point of exceeding reactor pressure due to the operating condensate and condensate booster pumps. With the master feedwater controller in manual and the level control valves fully open, feedwater flow resumed to the vessel. Operators later notired that reactor water level was starting to increase without MPCS or RCIC injecting. An operator manipulated the Manual/Auto feedwater controllers to the full closed position in an attempt to prevent excessive overfilling of the reactor. The feedwater regulation valves (FWS-LV108, C) locked up at approximately 80K open, when a loss of control signal was sensed by the valve hydraulic actuator control card. This caused the vessel to overfill and flood the main steam lines. The operators determined that feedwater was still injecting into the vessel and closed the feedwater containment isolation valves (MOV21A 8). This terminated the Im feedwater irjection and water level peaked at 333 inches.
The main steam lines were flooded up to the turbine stop valves. Normal steam line drains were utilized to drain the excess water fror the steam lines. A turbine bypass valve was later opened manually to control reactor pressure and to initiate a plant cooldown.
The remainder of the scram recovery was routine.
II. CAUSE OF EVENT A root cause analysis for this event has been completed per Site Supervisory Procedure S-SUP-l, "Root Cause Evaluation Program". The root cause has been determined to be. personnel error. The operator visually verified that the isolation valves for prefi lter 28 were open by observing the length of the valve stems. However, he should have physically verified the valves'osition open. These valves were in fact closed. Thus, when the operator isolated prefilter 2A for maintenance, the instrument air compressors were isolated from the rest of the system.
The root cause of the feedwater regulation valves locking uo was design deficiency. An existing ground loop in the control circuitry biased the setpoint upward for the loss of control signal, causing lock-up of the faedwater regulation valves.
III. ANALYSIS OF EVENT The isolation of the air compressors posed no adverse safety consequences to plant personnel or public safety as a result of this event. This portion of the IAS is not requi red to affect or support the safe shutdown of the reactor or to perform any safety-related functions associated with its operatior The loss of instrument air caused the minimum flow valves for the condensate, condensate booster, and feedwater pumps to fai open allowing flow to 1
recirculate back to the condenser as designed (FSAR, Section 15.0.5, "Loss of Instrument Air").
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Nine Mile Point Unit 2 o 6 o o o 410 88 001 00 04 oi P~ASOIIOIOISOOA W~~~~WITSS The reactor scram which occurred as a result of the Level 3 trip was a conservative action. This poses no adverse safety consequenc s at any power level. This action did not in any way adversely affect any other safety system nor the operators ability to achieve safe shutdown.
Water level never fell below 112 inches during the transient. HPCS and RCIC initiated well above (conservative direction) the 108.8 inch Technical Specification (TS) required setpoint. Because water level never reached 108.8, some Level 2 system actuations and isolations did not occur. An investigation was performed to ensure the setpoints of all Level 2 instruments were correct.
This investigation found that the guide value setpoint for Primary Containment Isolation Actuation Level switches had been calculated incorrectly. Oetails can be found in LER 88-04.
The automatic actuation of RCIC and HPCS with coolant injection was a conservative response with minimal plant impact and no resultant impact on public safety. The systems operated as designed and restored reactor water level to Level 8, 202.3 inches.
In accordance with the requirements of TS Sections 3.5.1(f) and 6.9.2 (Emergency Core Cooling System (ECCS) Injections, the following special data is orovided:
For the HPCS nozzle,
- Total accumulated initiation cycles to date = 1
- Currerit usage factor value remains well below 0.70 The increasing reactor water level which resulted in the flooding of the main steam lines prompted the following concerns:
Thermal transients across the main steam pipe and reactor main steam nozzle walls caused by assumed colder water flowing into the hot main steam line.
The weight of water in the main steam lines.
Oynamic transient loads caused by water flowing into the main steam lines.
External loads/moments imposed on the reactor main steam nozzles.
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U.S. IIUCI.[A1 AIOUlATOAtt'OMMISSION UCENSEE EVENT REPORT (LER) TEXT CONTINUATION AAAAOVlO OMI tIO 3ISO&I04 g(DIAT/: 1I3IIN IIAtOll~IIAMO I I I OOCKIT IIUAMI1Ql ll1 IIVMM1 I ~ I ~ AOI IS Y(AA SAOUlttt AL ll 1 Nine Mile Point Unit 2 0 5 o o o 410 88 00 05 oe 08 OT ~ sm O MOOD ~ oaghedMC Aee ~W (ill Evaluation of the local stress due to the thermal transient across the pipe nozzle walls conservatively used the temperature of the feedwater as the fluid temperature entering the main steam pipe. This was conservative since it did not account for thermal mixing within the Reactor Pressure Vessel (RPV) due to the natural circulation of reactor water, the effect of core decay heat, or the effect of stored thermal energy in the reactor coolant pressure boundary metal piping, structures, and the reactor core. The results of the evaluation of the thermal transient indicated that all stresses satisfied design code allowables for the piping from the vessel to the main turbine stop valves and drywell oenetrations. Pipe break exclusion criteria was also satisfied for this event.
Final evaluations to establish the actual usage factor for plant life will use fluid temperatures calculated by a Contractor Engineering group for this event.
Evaluation of the weight of water in the main steam piping indicated that the design pipe stress analysis included load cases considering water in the main steam lines, therefore, no further review was required. Pipe supports were designed for maximum design loading considering either water or steam, Snubbers and spr'lng hangers were sized to account for the maximum movement due to the combination of plant thermal, dynamic, and deadweight (water or steam) conditions.
Review of linear potentiometer displacement data in the main steam piping indicated that there were no vibratory responses during the flooding event.
From this it can be concluded that hydraulic transient affects were insignificant.
The only external loads/moments imposed on the main steam piping, components, and reactor nozzles was the weight of water and thermal expansion of the system. Since no seismic event occurred during the scram and subsequent plant shutdown, the applied loads were enveloped by the design loads. Therefor e, no further ~valuation was required.
In conclusion, we have conservatively determined that the effects of thermal transients across the pipe and reactor main steam nozzle walls, the weight of water, and hydraulic dynamic transients, are within applicable code stress allowables for the piping and penetrations and within design pipe support loadings. Break exclusion criteria is satisfied wi.ere applicable and RPV nozzle loadings due to this event are enveloped by existing design loads, In summary, while the main steam line flooding and the loss of instrument air was an undesirable transient, effects of the transient were within plant design margins.
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V S HVCLSAA AIOULATOWYCOttttI~OII LICENSEE EVENT REPORT (lER) TEXT CONTINUATION AttAOVSO OIIS HO SI SOMI04 KXtllISS: S/SIIS MCLnV aaea III OOCKKT NIAHSA Ql Lll Itis&IA I ~ I ~ AOI ISI ttavt~tiAL .: +tV4ION M IJ A Nine Mile Point Unit 2 o 5 o o o 410 88 001 00 06 o~ 08 H~eaeOItttOCC ~~~~~WIBl I V. CORRECT I VE ACTIONS The individuals involved with this event participated in the LER investigation and follow-up report.
2. The operator involved in the valving error has been subjected to disciplinary action for inattentiveness to duty.
3. This event has been incorporated into the Operations Department Lessons Learned Program.
a. Administrative Procedure AP-3.3.1 has been revised regarding valve position verification policy.
5. The valve position verification policy will be incorporated into the operators requalification training program via TMR 02-88-18.
6. Modification PN2Y87MX238 - This modification incorporateo a design" change to the control circuitry of the manual/automatic controllers of the feedwater regulation valves. This change addresses the problem of inadvertent feedwater control valve lockup.
revised
7. Operating Procedure No. N2-0P-19, "Instrument and Service Air System" has been revised to provide additional direction in an event of loss of instrument air.
Operating Procedure No. N2-0P-101C, "Plant Shutdown" has been . A caution statement has been added to alert the operators to continuously monitor reactor water level if the feedwater regulation valves are in manual.
A review of the Nuclear Steam Supply Shutoff System (NSSSS) level instrumentation revealed that four level switches were found to have improperly calculated guide values. The Instrument and Control (l&C)
Department immediately recalibrated these instruments and incorporated procedure changes as necessary. Additional information is contained in LER 88-04.
10. The guide value calculations for all other level, pressure, and temperature instrument channels which have TS required trip setpoints have been reviewed by I&C to assure accuracy. Additional information is provided in LER 88-04.
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EDC 28100130 PAGE,3$ OP (3Q AFFECTED VENDOR MANUALS:
I N2E2 09 0 0 PWSUPO0 1 J N2E2 09 0OZPWSUP002 P N2E3 5 600ZPWSUPO0 1 INSERT AFTER TABLE OF CONTENTS THE FOLLOWING:
REFERENCES TO nBYPASSn MAY ALSO BE REFERRED TO AS "MAINTENANCE" TO CORRESPOND WITH LABELING CHANGES IN THE FIELD PER EDC 2E10130.
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I ELECTRICAL CHECKLIST FOR POTENTIAL AFFECT ON DESIGN DOCNENTS CHECKLIST NUNBER 5 ECL+
EDC NO. 5 REV. -g~/D/JO PREPARED BY: /Date 3-tz -9 o CHECKED BY: /Date The following items will be addressed prior to issue of electrical design documents:
1. Parent document revision checked YES NO
2. Open documents against parent document checked YES NO
3. Parent document responsible discipline code checked YE NO
4. Parent document update category checked YES NO
5. FSAR text reviewed for changes ~E NO
6. FSAR tables reviewed for changes ~ES sf NO I
7.. FSAR FIGS. reviewed for changes YES NO
8. FHEA FSAR affected (go
9. Technical Specification reviewed for changes ~ES NO
10. 50.59 checklist or Safety evaluation completed ~~ES NO ll. Restricted equipment list reviewed a~i 4Pn YES ONO If any answer to questions 1 thru ll, "NO", please explain e'
12. Appendix 'Sdetermination required YES NO
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13. Vendor drawings affected YES NO
14. Vendor Instructton manual LI >CL affected ~~wf.L1ic t'+ YE . NO
15. gilt f 5 lid tg ff YES ~NO
16. Hiring drawing affected YES
17. Panel layout drawing affected YES NO
18. Panel Bill of material drawing affected YES NO
19. Electrical single line drawing affected YES CNNI
20. ESK's affected YES ~NO
21. LSK's affected YES I
22. PEID affected YES +No
23. Protective relay setting drawing affected YES +No
24. Annunciator elementary diagram affected YES o
25. Annunciator window engraving affected YES o
26. Human factors manual review/change required YES ~NO
27. System descriptions affected YES No
28. Engineering specifications affected YES (Qo
29. Installation specifications affected YES
30. Electrical penetration assembly drawing affected YES
31. Any A.C. electrical load added/changed YES Qo
32. Any electrical calc. affected YES ~NO
33. Any D.C. loads added/changed to batteries YES
34. Battery load calc. affected YES ~NO
35. Any loads added/changed 'to diesels YES (
36. Diesel loading calc. affected YES ONQ
37. Any non 1E loads supplied from class lE buses YES ONO
38. Isolattoo devices considered for non 1E loads YES
39. Any c ".'idded/changed/deleted YES (Qo
'I
40. Cable schedule affected YES (Qo
41. Any raceway added/changed/deleted YES (Qo
42. Race~ay schedule affected YES ONQ
43. Any new load added to the ESK control circuit YES Qo
44. Control transformer rating affected YES Qo
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45. Control clrcutt fuse rating affected YES @NO
46. Starter s.fze affected YES +o i1'4
47. Minimum cable size for starter affected YES ONO I
48. Motor,'over~load'eater affected YES
49. VBS 1'oad,list affected YES (Qo
50. PGCC fuse list affected YES No
51. Motor, over>load,heater list affected YES
52. Parts"<procurement required for Unit 2 simulator YES ~NO
53. Critical drawings affected YES +o
54. GE PGCC wire list affected YES (Qo
55. Remote shutdown panel affected YES ONO
56. Isolator data base affected YES ~NO 1
57. LOOP diagram affected, ~, YES +NO
58. TLD affected ~ji j
YES ~NO
)
59. Set point
'" data sheet (SPUDS) changed YES Qo ij r
60. Set point cal cul ation.-~performed/changed YES Qo
61. LOOP calibration report (LCR) affected YES ONQ
62. PMS computer I/O List (IHC-4) affected YES ~NO
63. Radwaste computer I/O-List (LHS-120) affected YES ~NO t
64. Annunciator data basi (IHS-20) affected YES
65. ERF/SPOSlEOP impacted YES (Qo
i
66. R.G."1;9T fmpacted YES ONQ J.,,
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67. Site,'eri'glneerlng notification required for red lining "of critical drawings (Non-Mod. EOC's only) YES NO If any answer to questions 25 thru 67 'Yes'. please resolve~ before issuing design documents.
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313-367 ROT.OO g "f "I L( Li . cI NINE MILE POINT I 2IE ZE'JCZ NUCLEAII STATION or JlQ C TO BE lHCOAK)RATEO DOCUMENTS T~TEMPORARY MOCIF CATION CHANGED I ~
~OT TO BE NCCRPCRATEO TO BE WCCRPCRATEO, MOTTO BE POSTEO
~AOY HCORPORATEO
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QfRRRDWeE/ZS~ZXmZOS This Revision supercedes EDC 2E100513 Q DOCDME22ZJLTZOS OSLY Scope of this revision to correct Elgar relay contacts to reflect correct numbers, and to correct Elgar relay to make normally open contact a normally closed contact to operate correctly with existing annunciation scheme. Revised pages: 1, 13, 71, 109/3 Revision B: To correct. referenced 50.59 checklist in Licensing impact block and to clarify drawing changes (pg. 2, 106, 107, 108, 109 a 110) .
Revision A: Lift leads to disable shunt trip relay. To allow operation of 2VBB-UPS3A&3B upon loss of bypass i:nput fuse.'OR
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CHANGES TO PLC 253-1-2 SECT ION UF (MANUAL NHPCSEG N2t)456 (g7. E 2$ 9+I PE~UP.
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Add to Table of Contents (pg.
vOPIFICATION PN2Y87NXO iii) as last entry of section 7 WORK INSTRUCTIONS FOR PLC ~OPIFICATION II Insert work instructions (5 pages) at end of section II pa.g~ '7-)god w ~ z~c.
Add to list of spare parts CAPACITOR . 47microf arads-400V CAPACITOR . lmicro f arads-500V (note see letter 0e.v(4 822-474-04 821-5GA-P1 from Klgar dated 12/16/88 m s EQC.
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Repleee page e-1 W> ~ fee -4 existing drawings:
Replace 7481032 'ev F 6481032 rev E 5481015 rev D 5481017 rev 8 5481016 rev F 648-100-60 r ev C 648-100-40 rev F with:
7481032 rev J 6481032 r ev F 5481015 rev I 5481017 rev F 5481016 rev H ~G 648-100-60 r ev E 648-100-40 rev G o+
Add dr awingsc 5481032 rev C 9481015 r ev E 5431263 rev C 9481055 rev A 9481024 r ev C 84Sl052 rev~C o,qp~>(, re~ C Q Qg \ 050 r'8v 4 gQ -ZQ n c <t'QPm~c'c
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SECTION VI LIST OF DRAWINGS 6-1 GENERAL This section contains drawings f'r the major assemblies used 1'or the PLC 253-1-2, K lgar Sales Order 4093-1,2.
Drawing Rev ~ Title 7481032 J Installation Drawing 6481032 F In terconnec tion Diagram 5481015 L Main Frame Assy 5481017 F Moun ting P l a te 4ssy 5481016 H Input PLate Assy 5481018 B Hea t Sink Assy 648-100-60 E Schematic Line Regulator Control 648-100-40 G PC Assy Line Regulator Control 648-101-60 B Schematic Line Regulator Drive 648-102-40 D PC Assy Line Regulator Drive 6480003 A Schematic Static Switch Drive 5480003 D PC Assy Static Switch Dr ive 5481032 C Final Assy Power Line Conditioner 9481015 K We ldmen t Main Frame 5431263 C Input Plate Assy 9481055 4 Bracket Capacitor 9481024 C Front Cover 8481032 C Wire List 6-1
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o2SO Sr~ Geest <oat San Diego. Cciifornia 02121 reieceone (b10) reiex: 211063
~MS Pcccnihcr 1G, 1908 IHFORV~ATIQ."< ONLY Niagara Mohawk Power Corp. FAX TO: (315) 343-7135 9 Milo Point - Unit 2 ALL ATTACIIMENTS VIA FEDX Lake Road Lycoming, New, York 13093 Attention: Edward Dunn, Jr.
Project Engineer
Subject:
PO 64836, Status of Open Items
Reference:
Telecon E. Dunn/.C. Schanning 12/13/88 Payment on Invoice 4040501, $ 17,050.00 Payment on invoice referenced Iias been reccivcd; eviclcntly your accounts payable department had been referring to the Q.A. hold on Item 2, but failed to mention that check had been Issued for $ 17,050.00,.
ITEM 1) ATTACHMENT 1 - Documentation Requirements Document Category (Engineering 6 Procedural) No. 5 RECOMMENDED SPARES FOR MODIFICATION KIT FOR UPS253-1-106 PROJECT MQD 0 PN2Y87MX037 Qty 1 Fuse Blown Lamp Board 5430050-01 Qty 1 Capacitor .47uf-400V 822-474-04 Qty 1 Capacitor'; 1 uf-500V 821-5 GA-P1 Qty 1 Relay 2A-48VDC 861-1Y4-25 Document Category (Report 8 Verification DOC)
No. 3: Original arid three (3) copies of revised Certificate of Conformance are attached referencing final report number.
No. 5: Enclosed arc two (2) additional bound copies of Impell Rcport No. 01-1860-1731, Rev. 0, dated October 1988:
~ I' I
Z E pa as ~~/0 I'! MPC 12/16/88 Page 2 ITEM 2) DOCUMENTATION REQUIREMENTS No. 5: RECOMMENDED SPARES FOR MODIFICATION KIT FOR PLC253-1-2 - PROJECT MOD I PN2Y87MX037 Qty I Capacitor, .47uf-400V 822-474-04 Qty' Capacitor, .1uf-500V 821-5GA-P1 STATUS OF REVISED WORK INSTRUCTIONS/DRAWINGS Two (2) copies of revised work instructions with drawings are attached; Please note that revision levels are now indicated.
Drawing numbers marked with a 're still pending because of the modification required for the input plate (ECN currently being reviewed/'approved - copy attached). A "rod-lined" drawing (DWGN 9431263) is attached for your use in modifying the plate to specifications indicated. As soon as ECN is incorporated, final submittal of work instructions. with updated drawings will be provided.
Mounting hardware will ship within the next few. days.
Please contact me at'619)458-0298 if you require further information.
Regards, Cindy Schanning Service Administrator
/cs Attachments included in FEDX package
l
~ 2 E I i sap ~ ai: //0 VIOP.K INSTRUCTIONS FOR THE PLC MODIFICATION Pl C 25"-1-2 Insure that all parts and necessary documentation are on hand. Read and understand the instructions fully prior to starting any work and insure all questions are answered.
The following is a list of all related parts and drawings:
Ref. Description Part No. ~Qt .
A6 Fuse blown lamp board 5430050-02 K1 Relay 115VAC Coil 861-420-A P.
Contacts (DC) 861-ARD-CR TB4 Terminal Block 893-604-4X C4 Capacitor . 47uf-400V 822-474-04 Fuse Condition Placard Spacer - Lamp Board 9481054-01 1 Bracket - Capacitor 9481055-01 1 Input Plate +H~H 02 1 Control Board C44 Capacitor .l.uf-500V. 821-5GA-P1 1 CR31 Zener Diode 12V 843-474-2X 1 Document Title Drawin Interconnect Diagram 6481032-01, Rev. F Final Assembly <481032-01, Rev. C Installation Drawing 7481032-01, Rev. J Main Frame Assembly 5481015-03, Weldment Main Frame 9481015-01, Rev. E input Plate 5481016-01, Rev. H Input Plate Assembly 5431263&2, ~
(NEW)
Input Plate. ~t+BB-02, ~
( NEW)
Bracket - Capacitor 9481055-01, Rev. A Spacer - Lamp Board 9481054-01, Rev. A Mounting Plate Assembiy 5481017-02, Rev. F Front Cover 9481020-01, Rev. C Control Card Modifications 648-100-40/60 ECN 6310 ~
Appropriate wire list 8481032-01, Rev. C
ECN is currently being submitted for review, approval and incorporation.
<+ <~~~it~~)qadi te~ 4-B-B~ aorreaW @~be~~ WQ45'1243
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Dated 12/16/88
lV Z E. 0 0 Asser:.bl Instructions Remove tlute front'anel of the unit after insuring that all power is completely terminated into the unit and rechecking any possible input feeds with a volt meter.
Using Figure 0-2 as a guide, drill the appropriate holes in the upright angle iron frame member and the front cover plate. Utilizing the drawing on Figure 4-3, install the fuse blown lamp board with the appropriate mounting hardware as shown. After installation of the assembly, check to insure that the LED is properly positioned to indicate through the hole in the front cover and adjust the assembly if necessary. Mount the fuse condition indicating panel on the front cover as displayed on Figure 0-3.
Mount the terminal block and the reiay to the input plate assembly as indicated by the predrilled mounting holes and Figure 0-3. Secure this assembly to the frame by means of the unistrut mounting screws positioning it between the adjoining assemblies on each side. Recheck all screws and hardware to insure they. are all secured as required for proper operation.
Wirin instructions On the mounting plate assembly drawing, you will fInd the indication of placement for the C% capacitor. Solder one end of the capacitor to the Jl connector on pin 16 or T being sure to install some insulating sleeve material over any bare lead wire. It is important to minimize excess lead length to insure minimal possible movement in case of vibration. Again insuring that any bare lead wire is covered, attach an appropriately sized ring lug connector to the other end of the capacitor and secure it to the indicated chassIs mount sere~. Install the capacitor mounting bracket as shown on the assembly drawing utilizing some form of approved insulating material to minimize movement within the clamp. Label the wiring as indicated ln the wire list.
Reiay and terminal block wiring on input plate:
Utilizing the'appropriate hardware, connect a wire from the K1-1 terminal an the relay mounting black to TBa-l, the Kl-2 terminal to TBe-2, the Klfl)erminel to Kl-3 end from Kl-e to TBeg Fram the Kl-3 coil
< terminal run a wire to the 783-1 terminal and from the K1-6 termlnaT
! connect to the A1 panel C2 capacitor, terminal 1. Use the wire numbers as listed in the wire list.
ggv'zS>Vw 4 Fuse blown lamp board wiring:
Connect the following points with the proper wire size and hardware using the wire list as the numbering qulde. (A2)Jl-22 on the control board connects to the (A6)J1-7 terminal on the fuse blown lamp board. (A6)Jl-1 connects to the C3 capacitor (-) termination on the A2 assembly and the C3 {+) is connected tv the (A6)Jl-3 termination.
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p~ gF //'0 Cantral Card MadtfiaatIan The control card in each unit must be modified to accept the new diode and capacitor added to allow proper operation. Referring to ECN 6310, place the CQ0 capacitor and the CR31 diode as shown in the literature. Be sure that any residual solder or other contaminents are removed from both sides of:he PC board prior to attempting to use it.
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3 Brawn Deer Road w Diego. California 92121 ~ 2 eieohone (b10) 49~%
lelcne 211063 June 8, 1989 IN 0 Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, NY 13212 Attention: Angie Freeland
Subject:
Drawing number 9431263-01 Enclosed is one copy of subject drawing. Please note that the correct number for the input plate is. 9431263, not 9481263 and revise your documentation to reflect this correction.
P'lease submit further questions in writing, to my attention.
Regards,
~,
Debbie Winkle Service Administrator
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IO IA 'oc tl 4 tt7lo4+t'O~"'F i" tt MATEljJPPLS ENGR. 0 T WSS NO+ tt EOUIP; SPKC. TK . WORK COXPLCTIONr;-'; NWR Q I ts I KA I AT ~.:. IqSP~RCPORT."NOISIO;;..--,, OATC STATUS I C WILI SC INCORPORATED N WILt. NOT SC INCORPORATED PR 'E 4 FINAi;,WOR%',TRACKIIIOICLQSUN K I OAT C I NO CNANOC 4 i OKSCRIP IONIOI gU~gg~ ne aa ~~ s~"A..S,-,-I,4!!Peeuarw~nmror OKSCRIPTIOlf IOCI= REMARKS;f4tI~ g~.*yj~r~','g~- '.!"c'+j 7'T'., ss t "v Il~ H [,"'~ ~ ~l (sy q a > "> g ~.,~j 1 EDGAR CORM'>OH 9250 %mrna Qeer Ruad Obgo, Cabhreta 92191 K+M.R c waG7r Tdephone; (619) 4554085 Tdaa: 111%3 ELDS UR PSZ.QF "t ~e 7, 39tS Stor1e 4, Rbster Fngineerfrtg Carpm atTart 3 Executor Caapus P.Q Box 5200 Chewy H)ll, Nm Qrsey 08034 -Attention: 8.8. %N Sub/act': P.O No HP2 f03SA llfegara lobar Poser. Ca. Nfne Nle Nuclear Station, Vn)t 2: Disable. Ground. Fault Detector" OPS 253 1-$ 05~ UPS M3 1-17{i
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N2E20900IPWSUPOO 1 F N2E20900IPWSUP002 F N2E35600IPWSUPOO 1 INSERT AFTER TABLE OF CONTENTS THE FOLLOWING:
REFERENCES TO "BYPASS" MAY ALSO BE REFERRED TO AS "MAINTENANCE" TO CORRESPOND WITH LABELING CHANGES IN THE FIELD PER EDC 2E10130.
ELECTRICAL CHEOtLIST FOR POTENTIAL AFFECT ON OESIGN DOCUNENTS CHECKLIST NUNBER 5 ECL 2 ~+ CE C9 EDC NO. III REV. ~g~yD/gQ PREPARED BY: /Date - +0 CHECKED BY: C /Date 8-iz -P~
tt The following items will be addressed prior to issue of electrical design documents:

l. Parent document revision checked YES NO

2. Open documents against parent document checked YES NO

3. Parent document responsible discipline code checked YE NO

4. Parent document update category checked YES NO

5. FSAR text reviewed for changes ~E NO Q9

6. FSAR tables reviewed for changes ~ES e NO 7.. FSAR FIGS. reviewed for changes YES NO

8. FHEA FSAR affected +NO

9. Technical Specification reviewed for changes ~ES NO

10. 50.59 checklist or Safety evaluation completed ~~ES NO

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11. Restricted equipment list reviewed YES (QO If any answer to questions 1 thru ll, "NO", please explain

12. Appendix 'B'etermination required YES (QO

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13. Vendor drawings affected ES NO LI i 0- ~ggUP

14. Vendor instruction manual affected ~~g>o E;b+ YE . NO lf.dtli f I' l d \5 ff YES ~NO

16. Hiring drawing affected YES

17. Panel layout drawing affected " YES NO

18. Panel Bill of material drawing affected YES NO

19. Electricak single line drawing affected YES QN

20. ESK's affected YES ~NO

21. LSK's affected YES

22. PAID affected YES Qo

23. Protective relay setting drawing affected YES +No

24. Annunciator elementary diagram affected YES ~NO

25. Annunciator window engraving affected YES ~NO

26. Human factors manual review/change required YES 'NO

27. System descriptions affected YES Qo

28. Engineering specifications affected YES ONQ

29. Installation specifications affected YES

30. Electrical penetration assembly drawing affected YES

31. Any A.C. electrical load added/changed YES ONQ

32. Any electrical calc. affected YES (Qo

33. Any D.C. loads added/changed to batteries YES

34. Battery load calc. affected YES ~NO

35. Any loads added/changed to diesels YES (

36. Diesel loading calc. affected YES (Qo

37. Any non lE loads supplied from class lE buses YES ONO

38. Isolattoo.devices considered for non 1E loads YES

39. Any ca@i:added/changed/deleted YES 0

~'-.-.'.

40. Cable schedule affected YES ~NO

41. Any raceway added/changed/deleted YES NO

42. Raceway schedule affected YES ONO

43. Any new load added to the ESK control circuit YES ~NO

44. Control transformer rating affected YES ONQ

l

45. Control .circuit fuse rating affected YES @NO

46. Starter s,ice affected YES +O

47. Minimum cable size for starter affected YES ~NO

48. Motor overload heater affected YES

49. VBS load list. affected YES QNO

50. PGCC fuse list affected YES NO

51. Motor overload heater list affected YES

52. Parts procurement required for Unit 2 simulator YES ~NO

53. Critical drawings affected YES +NO

54. GE PGCC wire list affected YES ~NO

55. Remote shutdown panel affected YES (QO

56. Isolator data base affected YES NO

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57. LOOP diagram affected YES +NO 5&. TLD affected YES ~NO

59. Set point data sheet (SPDS) changed YES ONQ

60. Set point calculation performed/changed YES ONO

61. LOOP calibration report (LCR) affected YES QO

62. PMS computer I/O List (IHC-4) affected YES (QO

63. Radwaste computer I/O List (LHS-120) affected YES ~NO

64. Annunciator data base (IHS-20) affected YES

65. ERF/SPOSlEQF impacted YES ~NO

66. R.G. 1-.9T fmpacted YES (QO

67. Site engineering notification required for red lining of critical drawings (Non-Mod. EDC's only) YES NO If any answer to questions 25 thru 67 'Yes', please resolve before issuing design documents.

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ML18038A389

Text

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