ML20054C328
| ML20054C328 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/12/1982 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 5211-82-059, 5211-82-59, NUDOCS 8204200380 | |
| Download: ML20054C328 (22) | |
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P.O. Box 480 Middletown, Pennsylvania 17057 717-944-7621 Writer's Direct Dial Number:
April 12, 1982 5211-82-059 Of fice of Inspection and Enforcement 9
Attn:
R. C. Haynes 61 \\
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- "f 'h Region I, Regional Administrator
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U. S. Nuclear Regulatory Commission Af 631 Park Avenue
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King of Prussia, PA 19406 C '.
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Dear Sir:
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Three Mile Island Nuclear Station, Unit 1 (TMI-e[
s Operating License No. DPR-50
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Docket No. 50-289 10 CFR 50.59 Report In accordance with the requirements of 10 CFR 50.59, enclosed please find two copies of changes to TMI-1 systems as described in the FSAR. There were neither any changes to procedures requiring a change to the FSAR, nor test or experiments performed not described in the FSAR.
Sincerely, 1
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.$ kill Director, TMI-l HDH:CWS:vj f Enclosures cc: Director, Office of Inspection and Enforcement (40 copies)
U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Director, Office of Management Information and Program Control U. S. Nuclear Regulatory Commission Washington, D.C.
20555 John F. Stolz, Office of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Washington, D.C.
20555 8204200300 820412 PDR ADOCK 05000289 P
PDR GPU Nuclear is a part of the General Pubhc Utilities System N
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1 Modification - Hittman Mobile Radwaste Solidification System (Pre-operation only)
Description of Change The Hittman Mobile Radwaste Solidification System uses cement to solidify all three types of waste in a pre shielded container.
The shielding is sufficient to protect operating personnel while solidifying spent resin.
A disposal liner with an internal mixer is used as the solidification container.
The ratios of cement, additives (if required) and waste that will produce a dry product are determined through test solidification in a laboratory in accordance with the PCP.
Safety Evaluation Summary The Hittman system will not adversely affect nuclear safety based on the following:
1.
Stabilization with cement will increase the level of nuclear safety during transport and disposal.
2.
A seismic spill containment is provided to collect spills up to 163 ft3 (contents of a full solidification liner.
3.
The venting of the gas space of a liner being filled does not constitute an unmonitored discharge because of the absence of a gaseous radioactive isotopes (at ambient te=perature and pressure) in any of the waste streams to be processed.
- 4 Filling and solidification controls are remote and separated by a 7-5 inch thick concrete shield thereby reducing the radiation hazard to operating personnel.
5.
Hoses used for transfer processes which are periodically tested at system operating pressure and visually inspected prior to use.
- This aspect is discussed at length in response to NRC Inspection Report 81-34.
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i Modification:
B/A 412237 - Reactor Coolant Pump Motor Surge Capacitor Removal Description of Chance:
The new design rc= oves the reactor coolant pump (RCP) motor surge capacitors from the motor circuit.
The capacitors were originally installed inside each RCP motor terminal cabinet and connected line to ground at cach phase.
This previous design exposed the capacitor dielectric to a possible high radiation level within the centainecnt building in the event of an accident.
Studies have shown that a hich radiation level oro-motes capacitor dielectric failure which electrically shorts the capacitor thus rendering the RCP motor inoperable.
Safety Evaluatien Sunnarv:
Removal of the surge capacitors from the RCP motor circuit increases the reliability of the RCP motor by eliminating radiatien effect as a cause of motor circuit failure.
i Since the operation of the RCP motor is not nuclear safety-related, removal of the surge capacitors does not impact the nucicar safety criteria, i
There are two potential transformers within the RCP motor terminal box which are used for pumo motor power monitoring.
This monitoring is a safety related feature, in the process of rc=oving the surge cacacitors, the potential transfor:cr leads were disconnected from the capacitor and re-connected to the phase stub bus.
This does not change the potential transformer circuit performance and thus this modification leaves no un-reviewed safety qucPtions.
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a Modificction:
B/A 432017 - Relocate Feed Punp on Waste Evaporator
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Descriptien of Chance:
The Reactor Coolant and Miscellaneous Waste Evcporator's feed pumps have been relocated to improve accessibility of the feed pumps.
The system repiping also added additional suction and drain valves and connections for future :ced pumps for both evaporators.
The repiped sections of the system.eere heat traced to prevent precipitation of Boron.
These modificatiens improve the maintainabilit:. and minimize the down-time of the evaporators.
In addition, isolation valves were added to the outlet lines of the distillate reservoir tank to provide positive isolation of the tank from the distillate pumps.
Safetv Evaluation Summan -
The subject modification adds piping, valves and heat tracing which are equal or better than the existing equipment.
The subiect nodi-fication does not chance system operation, design basis, margin or saintv or description in the IMI-I
'3n.
It is therefore concluded that the subject modification does not involve an unrevieved safety question.
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Change:
B/A 412072, High Pressure' Injection System Cross-Connection (RM-14)
Description of Change:
Piping modifications inside the Reactor Building have cross-connected pairs of the High Pressure Injection (HPI) System injection legs.
HPI leg A in train A is cross-connected to leg C in train B and leg B in train A is cross-connected to Icg D in train B.
In addition, a cavitating venturi has been installed in each HPI injection leg upstream of the HPI leg check valve nearest the Reactor Coolant Systen (RCS) cold leg and downstream of the cross-connection. The normal make-up line has been rerouted to inside the Reactor Building and connects to the "B" injection line dcwnstrean of the cross-connection and upstream of the HPI leg check valve nearest the RCS.
An additional check valve between the cross-connect and the make-up-line tie-in prevents make-up flow to the "D" injection no::le via the cross-connection. A high capacity make-up valve MU-V-217 has been installed in a bypass line around the normal make-up valve MU-V-17.
The make-up flow instrumentation has been modified to read the higher =ake-up flow rates.
Safety Evaluation Summarv:
The design and installation of the modifications described above meet or exceed the requirements invoked on the original-design and documented in the TMI-l FSAR Chapter 6.
The cross-connections and cavitating venturi in-prove plant safety by allowing mitigation of the effects of a small break Loss of Coolant Accident (LOCA) occuring in a HPI line at or near its connection to the reactor coolant system (RCS).or in a RCS cold Icg without operator action.
The HPI water losses from the broken leg are restricted to approximately one-fourth (1/4) of the total injection pump flow rate and -the remaining flow is injected into the RCS via the unbroken HPI legs.
The HPI flow through the unbroken legs satisfies small break Energency Core Cooling System (ECCS) criteria. The larger capacity make-up valve provides a means of quickly restoring pressuriner level following overcooling of the reactor coolant system without starting another nake-up pump or thermal shocking an RCS HPI no::le not normally used for make-up.
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B/A 412012 - MS-V22 A&B Spring Replacement ('RM-13H)
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Description of Project:
In order.to provide over pressure protection for' hhe Turbine Driven EEi l
Pu=p, i.e., do not exceed the maximum operating pressure of 250 PSIC for the turbine steam chest and mechanical governor, the following. modifications to the turbine main steam inlet line have been implemented:
1.
New springs have been installed on safet. valves MS-V22A&B whereby reducing their set pressures frc: t95 PSIC and 505 PSIG, respectively, to 200 PSIGland 220 ?S!C respectively.
1 2.
The set pressure for pressure controller PC-5 has been reduced from 200 PSIG to 175 PSIG.
J, 3.
The valve stem travel on main stea= pressure control valve, MS-V6, has been limited to 65% of the full valve stem travel
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distance.
4.
All three (3) turbine no :le valves have been locked in the full open position whereby ensuring sufficient steam flow to the Turbine Driven EW Pump.
Safetv Evaluation Summarv:
j The modifications to the turbine main steam inlet line have been reviewed 1
to ensure that they will provide the required over pressure protection for i
the Turbine Driven E N Pump.
Futhermore, these codificaticas will not create adverse conditions which will degrade the I W system performance.
The change to open all three (3) inlet steam no::le valves on the turbine will increase the nos lc flow above design flow, thus. ensuring turbine and pump performance.
The changes that have been implemented by this modification improve the EW system reliability and do not have any ad-verse effects on either the EW systee or other systens and components.
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Modification: B/A 412012 - EFW Backup Instrument Air Compressor System (RM-13)
Description of Project:
A backup source of instrument air has been previded in ceder to mitigate the event when the main instrument air supply and station service air supply systems are both inoperable.
Backup instrument air conpressors and systems have been installed in the Turbine and Intermediate Buildings.
Intermediate Building The backup instrument air supply system consists cf an SC gal reservoir, pres-suriced to1100 PSIG, which is supplied from an IE5CFF air :cepressor, IA-P-25.
When the main instrument air syste pressure drops te below 70 PSIC, the supply line from the backup air compressor is opened.
Transfer to the backup air supply is automatic and no operator actien is recuirec.
The power to the air compressor is supplied f ro= the 13 450V ES Mo:or Control Center. Ecuip=cnt serviced by this system are key valves and instrutents located in the EFh Main Feedwater and Main Steam systees.
Turbine Buildin; The backup instrument air supply system configuration for this building is similar to the system installed in the Intermediate Luilding.
The backup air compressor (IA-P-2A), electrical power supply (la ;S0V ES Motor Control Center), valves and instruments are different frer the valves and instru-cents that are in the Intermediate Building. Valves and instruments serviced by this system are located in the condensate Main Feedwater and Main Steam Systems.
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Safety Evaluation Summarv:
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With the installation of the backup air compressors and system, the ability of plant persennel to ef fect a controlled and orderly cooldown of TMI-l will be increased even when the normal instrument air supply and station service
_,ine ta:sup air supp,.y wi.,3 j
sir supply systems are rendered inoperatie, l
ensure that the =cin feedwater and emergency feedwater systems, as well thc7 main steam dump lines to both the condenser and atmosphere remain as, operational.
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d' Chance Modification:. Manual Control of PORV E/A 412062; LM-39 Descriptien of. Chance:
A key locked ccntrol switch to PORV (RC-RV2) has been added to allow the operator to manually open the valve from the centrol roca under certain emergency conditicas as called out in Plant procedures.
The switch employed by this rodification is of the sane cuality as the existing equipment.
The switch has two positions, ner=al 5 open and spring return to normal when relcased from cpen position.
The switch is keylocked in the normal position and valve is closed at that position and is under au:cmatic centrols.
The external viring added by this modification frer the relay cabine: to the centrol rect are all in ccaduit and/or existing trays.
This modification perferns the same functions in the same canner as the unnedified system.
The only result of this modificatien to the PORY (RC-RV2) contrcl syster is to allow the operator to open manually the PORV fren the control roca.
Safety Evaluation:
It is determined with regard to the FORV (RC-RV2) Manual opening frem the control room:
1)
The probability or consequences of accidens previously evalua:ed have not not been increased.
The modificatica described above is key-locked to preven:
inadvertent actuatien and the automatic functiening of the PORV is not changed.
2)
No accidents other than previously considered will be introduced since the codification will not change the cperation of the Safety Sys: cms.
3)
No safety targin has been reduced.
The FORV control syste= will continue to operate as designed.
For reascas presented above, imple=catation of the design changes associated with the manual opening of FORV (RC-RV2) fret the control roc: do not involve unreviewed safety considera:icas.
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Modification:
B/A No. B00221.- Raise stent generator level trans-mitters SPLT1, SPLT2 and SPLT4 (RMS).
Description of Change:
This change includes the raising of steam generator level transmit-ters SPLT1, SPLT2 and SPLT4, to 5'-9 3/4" above Reactor Building (FB) floor, from their existing lower level where they were vulnerable to high water level due to accidental flooding in the RB.
Safetv Evaluation Summarv:
Steam generator level is one of the key parameters used to accomplish safe shutdcen and maintenance of natural circulation.
The transmit-ters for these level instruments located at the lowest level of RB were vulnerable to accidental flooding in the RS.
Moving these transmitters to 5'-9 3/4" elevation above RE floor puts them above predicted post LOCA flood level and hence upholds the integrity of the level instruments for a safe shutdown. This modification does not result in any change to the system as describcd in the FSAP.
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Modification:
B/A 412012 - EFW Safety Grade Auto Start on Loss of Four (4) Reactor Coolant Purps or on Loss of Both Main Feedwater Pumps (RM-13E)
Description of Project:
In addition to the turbine-driven EFW pump, both of the motor-driven EFW pumps have been modified to receive an auto-start signal en loss of both MFW pumps or loss of four (4) RCS pumps.
This mcdification has been in-plemented by utilizing contacts from the RCS power monitors or by sensing the dif ferential pressure across the ITU pumps.
The ECS pump power monitors are a safety grade systen.
The MFW pump differential pressure switches have been seismically tested and are tied into the respective EFW initiating circuits through isolation relays. The DF switches and circuitry are cen-sidered safety grade to the extent possible, i.e., qualified compenents installed in the turbine building which is a non-seismic Class I structurc.
The actuation systen is arranged in two (2) trains which utilize tuo pressure switches to sense the loss of both MFW pumps and four (4) contacts from each of the redundant RCS pump power moniters te sense the loss of all four (4) RCS pumps. The power for each actuation train is supplied by the "A" station battery /diesc1 generator and the "E" statien battery / diesel generator. The "A" and "B" train sensors and cable have been located and routed to satisfy the safety grade separation criteria in Chapter S cf the TMI-l FSAR.
The motor-driven EFW pumps are powered frc: the redundant class lE 4160 volt pewcr supply and have been bioch loaded en their respective diesel generator.
Safety Evaluation Summary:
The safety grade auto-start of all EFW pumps based en the above signals and block loading of the meter-driven EFW pumps en the diesel generators will ensure adequate fics to the steam generaters.
Furtherrere, the ability of plant persennel to effect a contrc11ed and crderl> cooldown of TMI-l will be increased, as well as the reliability cf the Energency Feedwater System.
Fotential overcooling and everfill of the Sten: Cenerators have been analysed and found acceptabic (and have received "EC concurrence in SUREG 06SO p.Cl-2 and SUREG 06SO Supplement 3 p. 13)
See also RM-13C.
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Change Modification:
Rx. Bldg. Sump Level - Control Grade S/A 412047; LM-SA Description of Change:
This modification installed a level transmitter to nessure the water level above the containment floor up to 10 feet. This instrumentation supplements the instru-mentation in the su=p and extends the range to monitor water levels from the botto of the sump to 10 feet above the containment floor. The level indication for the new instrumentation is local outside the containment.
Safety Evaluation:
This modification does not change any of the safety analysis described in the FSAR, nor in any way increase the potential for any of the hypothetical accidents described therein.
This codification extends the range of the instrumentatien provided to monitor water levels in the containment due to large pipe breaks and thereby en-hances the nuclear safety in plant operation.
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Modification:
B/A No. B00215 - Raise pressurizer level transmitters RCI-LT1, RCI-LT2 and RCI-LT3 (LM-9).
Description of Chance:
This change includes the raising of pressurizer level transmitters RCI-LT1, RCI-LT2 and RCI-LT3 to 6'-1/2", 5'-11 3/4" and 5'-11 3/4" respectively abeve Reactor Buildin; (RB) floor from their existing lower level where they were vulnerable to high water level due to accidental flooding in the RB.
Safety Evaluation Summary:
Pressurizer level is one of the key parameters used to accomplish safe shutdown and r.aintenance of. natural circulation.
The trans-
=itters for these Icvel instruments located at the lowest level of RB were vulnerable to accidental flooding in the RB.
"oving these transmitters to these new elevations above RB floor puts them above predicted post LOCA flood icvel and hence uphelds the integrity of the level instruments for a safe shutdcwn. This modificatien does not result in any change to the system as described in the FSAR.
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Change Modification: Modification of Power Supply to ICS/NNI System.
B/A #412029; RM-17.
Description of Chance:
This modification installed a non-automatic transfer switch on elevation 322' of Inverter Roce IA, with manual pushbutton controls and annunciator located in the main centrol room. This manual transfer switch will pro-vide a means for the operator to remotcly cennect the ICS/NNI supply bus to an alternate, regulated feed upon loss of input voltage to the bus from the Inverter.
Safetv Evaluation:
This modification adds a remote switching capability of transferring an auxiliary supply to power the NNI/!CS system in the event the static automatic transfer switch 1A fails as a power feeder to the NNI/lCS system.
The transferred power supply does not power safety relatr.d equipment.
The components to be installed will be countcd to preclude seismically generated r.issiles.
The modification will impose an additional continuous load of 30VA on the emergency diesel generator system in the event of an off-site power failure. The added 30VA will not degrade the diesel generator system.
The implementation of the codificatirn does not involve unreviewed safety considerations with regard to the criteria 10CFE, Part 50 Section 50.59 (a) (2) or degrade any safety related equipment.
O Modification:
B/A 412012 - Manual Loader Stations for EFW Control Valves EF-V30A/B (RM-13D).
Description of Project:
The manual loader stations for EFW control valves EF-V30A&B have been installed in the control roon on the main centrol board at sections "CC" and "CL."
The =anual loader stations will provide the plant operators with the ability to control ETW flow independent of ICS, as well as upon loss of power to ICS.
Safetv Evaluatica Summarv:
This modification provides redundancy to the contrels for EFW control valves EF-V30A5B.
It will preclude thc loss of EFW flow control upon loss of power to the ICS.
Fur:hcr ore, the ability of plant operators to effect a controlled and crderly cooldown cf TMI-1 will be increased, as well as the reliability of the E=cr;;ency Feedwater System.
Modification:
B/A 412024 - Installation of Emergency Feedwater Cavitating Venturis Description of Proiect:
Cavitating venturis have been installed in the EFW system, i.e.,
in each of the EFW supply lines to the OTSG's (one (1) venturi per line) between the EFW control valves EF-V30a5E and the respective EFW containment penetration.
The venturis have been instal]cd in order to (1) limit EF'.i flow to a ruptured O!SG; (2) limit the mass and energy release from a ruptured OTSG and prevent overpressurica-tion of the Centainment Euilding; (3) ensure sufficient EFW flow to the intact OTSG; and (4) reduce the potential for overcooling of the Reactor Coolant System.
The venturis have been sized to provide at least 500 GPM EFW flow for small break LOCA or feedwater line break
' accidents.
Safetv Evaluation Summarv-This modification has been implemented in erder to limit EFW flow to a ruptured OTSO.
The cavitating venturis will improve the EFU system's ability to remove heat from the RCS in a controlled and orderly nanner.
This modification does not introduce any accidents or malfunctions not previously evaluated ner does it increase the likelihood of occurrence or consequences of any accidents.
Ne safety margins will be reduced as a result cf this modification.
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Install Alarm to Indicate Electromatic Relief Disabled Descrio: ion of Chance:
An alarm was added in the control room to alert the operator in the even; the Power Operated Relief Valve (PORV) is blocked.
The alarm occurs if the block valve is closed or if the tiDTT switch is not in the desir;ed position.
This alarm is to warn the operator if over-pressure control of the PORY is not available.
Safety Evaluation Summary:
Work was done in accordance with drhwings proviced by gal, :ne cesign requirements are compatible with the original design.
The change affects alarm circuit.s only and therefore does not affect the electromatic relief valve operation.
For the above reasons the change does not:
1.
increase the probability of an accident, 2.
create a different type accident possibility or reduce the margin of safety as defined in the technical specifications, or 3.
invcive an unreviewed safety cuestion.
Modification: B/A #412003 - Tailpipe Temperature Detector on PORV and Code Safety Valves.
Descriotion of Change:
The PORV and Code Safety Valve tailpipe ther=occuples were replaced with pairs of thermocouples forming differential detectors to monitor the discharge line temperatures relative to the local ambient tempera-ture.
The pipe thermocouple is counted on each discharge line inside of the secondary shield wall.
The ambient thermocouples are on brackets counted on the secondary shield wall adjacent to its pipe thermocouple.
The differential te=perature signals become part of the computer data base.
A pressurizer relief / code safety valve discharge line differen-tial temperature of 45 degrees Fahrenheit or more will result in the computer displaying that a pressurizer relief / code safety valve dis-charge line "high temperature" exists on the control room cathode ray tube (CRT). The CRT high temperature display clears when the differen-tial temperature drops to 40 degrees Fahrenheit or less. The differen-tial temperatures can be displayed in the control root on the CRT, printer or continuous recorder. The control root operator may confire a valve has reseated by monitoring the tailpipe temperature trend following a valve opening versus the predicted trend for a reseated valve.
Safetv Evaluation Summary:
The differential temperature indications are used by the control rect operator to detect that a relief / safety valve has opened, verify that it has reseated or detect valve seat leakage under steady state conditiens.
The differential temperature indication function as confirmation to the control grade indication provided by differential pressure transmitters connected across elbow taps downstream of the relief / code safety valves.
In addition, the relief valve will be monitored by accelerometers ccunted on the valve.
The differential temperature detectors do not affect safety related equipment and have been seismically installed to prevent tissile hazards to safety related equip =ent.
It is therefore concluded that the subject modification does not involve an unreviewed safety concern.
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Change Modification: Existing Air Operated RB-V-7 Valve Replacement by a Motor Operated Valve (PM-1, B/A 412060)
E Description of Change:
I The RB-V-7 Valve used in the nor=al reactor building cooling water system, serving to isolate the water line at the containment boundary,was replaced with a motor operated valve. The new RB-V-7 motor operated valve was powered from 480V E.S. MCC IB.
This valve was provided with status indi-cation and control facility at the control room.
Safety Evaluation Summary:
h This modification climinates the poor leakage performance associated with the original valve.
It is determined that the potentia] for the accident i
and its consequences as described in the FSAR are not increased by this
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Modification:
B/A 412012 - Main Steam Rupture Restraint to Protect EFW Pumps / Pipe (PJi-13H)
Description of Project:
The main steam rupture restraint was installed to protect the discharge pipe of the turbine-driven EFW pump (EF-P1) from a potential MSLS and pipe whip accident, as created by the 12" MS inlet line to the turbine-driven EFW pump.
The rupture restraint protects the EF-P1 pump discharge pipe between check valve EF-V13 and the common EFW system discharge manifold at valves EF-V2A&B, i.e.,
in cubicle IJ of the Intermediate Euilding on EL 295'-0".
Safetv Evaluation Summarv:
This modification protects the EFW system discharge pipe / cot =on manifold from a MSLS and pipe whip accident.
It ensures that nct single failure (MSLB) can disable the EFW system with no cperator action required. The changes that have been implemented by this modification improve the EFW systen reliability and do not have any adverse effects on either the EFW system or other systems /
components.
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i Change Modification:
Provide Auto Reset for EST Actuation System B/A 412058; LM-33 i
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1 Description of Chance:
This modification provides manual reset capabilitics in the control room for ESFAS
-bistables in HPI, LPI and reactor building isolation system channels (all safety related) for TMI-1.
The ESTAS trip bistabics were adjusted for zero deadband, thereby l
allowing reset whea the monitored variable goes above its predetermined set t
point.
The ESFAS 2 of 3 logic actuation circuits were modified with a relay contact and rc=ote reset switch located en'the control room' console which will allow the I
operator to reset the ESFAS trip channel without leaving the control room.
Safety Evaluation:
The protection functions of the ESAS system continue to function as designed and are not degraded by thesc modifications.
Remote reset of the ESTAS functions shall add to the plant control performance during emergency conditions when the operator is required to attend to many control syste=s at the main console.
Further, it is concluded that:
1)
The probability or consequences of accident, previously evaluated, have not been increased.
The modification described above will not have any effect on the operation of the ESFAS.
Accordingly, the consequences of accidents for which ESFAS is required to be operable vill not change.
i 2)
No accidents other than previously considered will be introduced.
Since a
the modification will not change the operation of ESFAS, no new accident conditions will occur as a result of the ESFAS operation.
3)
No safety margins have been reduced.
The ESFAS will continue to operate as designed to citigate the consequences of Loca's within established safety cargins.
l For the reasons presented above, impic entaticn of the design changes associated with i
the rc=ote-reset capabilitics of the bistables for all the ESFAS channels do not involve '
unreviewed safety considerations with regard to the criteria of 10 CFR, Part 50 Section 50.59 (a) (2).
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Modiffeation: B/A 412012
- E d Flew Indicatien in the Control Reen (RM-13B)
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i Description of project:
s Redundant EF'.! flow indicators have been installed on the main con-trol console in the centrol room and will provide the plant operatorn with positive EF'i flew indication f or monitori:n; OTSC level.
This safety grade modification is supplicd with Class it power and satisfies both singic failure and Seis:2ic Category I criteria.
a Two independent and redundant equipment trains provide assurance that at least one (1) flow indicator per OTSC is available to the plant operators.
Each of the En! supply lines has been provided with two (2) flow sensing devices, i.e., a pair of ultrasonic transducers, display cemputers and flow indicaters.
The i
ultrasonic t ransducere here been insted !Od betueen the En: control i
valves EF-V30.V.D and the EM? cont ain ent penetrations.
The ' Red' ultrasonic transducers FE-777 and FE-7S9 and correspondinr, tre.ns-mitters FT-77S and FT-790 provide input signals to. control censole indicators F1-779 and F1-791.
Similarly, the 'Creen' ultrasonic transducers FE-780 and FE-786 and correspending transmitters FT-7SI and FT-7S7 provide input signals to control console indicators FI-782 and F1-78S.
Safetv Evaluation Summarv:
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i This safety grade modification is designed to provide only positive En! flow indication in the.:ain control roo, and no control function.
The plant operators will use the OTSC Icvel indicatfun for control-ling En1 flow.
The changes that have been impicmented by this modification improve the En? system reliability, do not create the possibility for an accident or malfunctien, and do not have anv adverse effects en the ER1 system or other systems / components.
Furthermore, the loading of instrumentation will not degrade the operability of the diesel generators.
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a Modification:
B/A 412012 - EF-V30A&B Fail Open en Loss of Instrument Air (RM-13C)
Descriptien of Chance:
In order to provide assurance that emergency feedwater can be delivered to each steam generator when required, the failure mode of centrol valves EF-V30A&B have been modified to fail in the open position and remain in this position upon loss of air.
EF-V30A5B are air cylinder actuated centrol valves which are located in the parallel emergency feedwater lines.
Tne fail open position for EF-V30ALS has been accemplished by installing a separate Fisher trip accessary counting package which includes a 75 trip valve and 1322 cu. in, capacity air accumulator en each centrol valve.
Safety Evaluation Summary:
The purpose of this modification is to have the EF-V30 ALE control valves fail safe, i.e.,
in the open position.
This modification does not e f fect the margin of safety as defined in the basis of Tech Spec 3.4 Futhermore, the ability of plant personnel to ef fect a centro 11ed and orderly cooldown of TMI-1 will be increased, as well as, the reliability of the Emergency Feedwater System.
Ten minutes at a minimur is availabic for operator action to prevent overcooling in the extremely unlikely event of multiple failures resulting in the EFW control valves failing open.
In addition, overcooling as a result of anticipatory filling of the Stear Generators f ren 30" to 50!.
has been analyzed and found to be acceptable (fill rate is limited by procedures as well as the newly installed EF' cavitating ventures - E/A 412024).
See also RM-13E.
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