ML20091R764

Annual Rept of Facility Changes,Tests & Experiments Conducted W/O Prior Approval & Challanges to Primary & Secondary Sys PORVs & Safety Valves for 1983

ML20091R764
Person / Time
Site:	Ginna
Issue date:	12/31/1983
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 8406150233
Download: ML20091R764 (57)

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1983 ANNUAL REPORT

'OF FACILITY CHANGES, TESTS AND EXPERIMENTS CONDUCTED WITHOUT PRIOR APPROVAL AND CHALLENGES TO THE PRIMARY AND SECONDARY SYSTEM PORV'S AND SAFETY VALVES c

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O 1983 Annual Report of Facility Changes, Tests, and Experknents Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244-Page 1 of 55 PLANT MODIFICATIONS COMPLETED IN 1983 EWR-1601B TURBINE AUXILIARY FEEDWATER PUMP VALVES LUBE OIL COOLER MODIFICATION In response to a Nuclear Regulatory Commission (NRC) requirement that the Turbine Driven Auxiliary Feedwater Pump (TDAFP) be independent of AC power, the lube oil cooler for the TDAFP requires a cooling water source that is independent of AC power. As such, this modification shall include a piping change that shall supply auxiliary feedwater to the TDAFP lube oil cooler and the outer pump shaft bearing in place of an equivalent amount of service water.

The modifications to the piping system will not involve any changes in the operation of the existing auxiliary feedwater system.

The consequences of the loss of normal feedwater flow event

! are unchanged. Because the input to the turbine remains l unchanged, the ability.of the pump to perform its safety function remains unchanged.

In the event of a loss of AC power to the station auxiliaries, the operability of the auxiliary feedwater pump. turbine is assured by providing a cooling water source for the turbine lube oil system that does not requires AC power.

This modification will assure that the TDAFP can function properly even with the loss of the onsite emergency AC diesel generators. Therefore, the margins of ~ safety are, in. fact, increased for this event.

The consequences of a steam or feedwater system pipe break outside containment remain unchanged by this modification.

The Standby Auxiliary Feedwater. System is designed to mitigate -

the consequences of a high or moderate energy pipe break in the Intermediate Building, therefore, this system does not have to be designed to withstand pipe whip 'or ' jet impinge-ment. However, .the new lube-oil system components and' piping shall be designed to maintain their' capacity in' a post-accident environment.

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. I 1983 Annual Report of Facility Changes, l Tests,'and Experiments Conducted E Without Prior Approval i R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244

-Page 2 of-55 The consequences of a lube oil. system pipe break remain unchanged by-this modification and are negligible. The -

lube oil system is not required to meet pipe break or whip criteria because the Turbine Driven Auxiliary Feedwater 3

Pump System is not pressurized during normal . plant operation.

A portion of the new cooling water piping', where required l for assurance of pressure retaining integrity of the auxiliary feedwater system, shall be designed _to Class 3' requirements of ASME III, which include requirements for seismic- loadings.

Thus , -the margins of safety;provided for seismic-events are_not reduced by the modifications.

Heat removal greater - than heat generation due to fe,edwater-system malfunctions is a-means of increasing' core _ power above full power. This event"can be caused by -either the accidental opening of the condensate bypass valve or the accidental opening of the feedwater control'. valves.. The

results of both events would be the tripping of-the reactor =

l which would result in the activation of the auxiliary. feedwater -

system. The consequences of such a transient. are not changed -

by this modification.

i This modification will not alter the pump's capacity-to

-deliver 400 gpm of auxiliary feedwater ' to the' steam generators.

l Therefore, the' margins of safety during normal operation i

and transient' conditions anticipated ~during the ..lif e' 'o f

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the plant have not been reduced. It has~also been determined that.the adequacy of structures, systems and components

provided_ for the prevention of accidents and1the 1 mitigation-

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, .of.the consequences of accidents have not been affected.

l EWR-1869 AUXILIARY FEED PUMP ~ INSTRUMENTATION UPGRADE' i-- .

l The purpose..of:this modification-~is to upgrade 1the' flow- ~

and pr_ essure instrumentation associated with the Motor-

[ Driven and Turbine' Driven' Auxiliary Feedwater: pumps _'at

Ginna Station. f This modification ' involves ' the replacement'

. of_ the

following primary instrumentation: : PT.-2029,' FT-2001,_

FT-200 9,J PT-2 019,c PT-2 0 3 0,- FT-2 00 2,/ ' FT-2 0 0 6, FT-2007.

The instrumentation presently;used,doesInotlhave'the? desired _

' accuracy and repeatability.

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o 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted I Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 3 of 55 In addition, the existing flow transmitters are utilized to operate valves 4007, 4291 and 4008. Each of these flow transmitters have a built in switch which is actuated via a mechanical linkage. This mechanical linkage has enough inertia such that accurate and repeatable determination of switch actuation point is not possible. As part of this modification, these switches will be replaced with r electronic bis. ables, which electronically compare flow transmitter output with setpoint and change state when the setpoint is reached.

Additional channels of flow instrumentation will be added to each auxiliary feedwater pump. This additional channel will be of the opposite channel designation from that of the primary channel. The primary channel for each feedwater pump will control that particular pump's discharge ' valve, whereas the secondary channel merely indicates flow.

The class IE portion of this modification shall be d'esigned to be operational: 1) during all modes of normal plant operation, 2).after a safe shutdown carthquake, and 3) after a steam /feedwater line crack break event in the Inter-mediate Building.

The non Class IE portion of this modification shall be designed for operations during startup, hot shutdown, and power operations. -

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Qualified flame retardant cable insulation.shall.be used.

( A fire in the auxiliary feedwater pump area could have

( the potential of causing failure of the circuits. routed l to the field instrumentation, and as a result of cross-coupling .

, ( through, fire induced short. circuits) of instrumentation I

trains. Isolation between these circuits and'the instrument

buses is assured'by protective circuitry in the analog

, process instrumentation. Therefore,'there'is.no increase I in the probability of a fire and reanalysis of a -fire and its postulated effects ~ as.a result of this modification

is

not' required.

l The safety related portion of the modification shall be designed'to withstand the effects of the safe shutdown earthquake. The safety related analog ' computation instrtmen- -

tation is to be mounted in seismically qualified racks.

p 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted .I Without Prior Approval  !

R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 4 of 55 Field instrumentation will be seismically supported and also qualified. Therefore, the margins of safety during a seismic event have not been reduced.

During a high energy line break (HELB) in the Intermediate Building, the environmental conditions will rapidly approach saturation conditions (215 degrees F @ l psig). For the operator to have adequate knowledge as to the operability-of the auxiliary feedwater pumps, the pump's primary instru-

. mentaiton (i.e. pressure and flow) must be able to withstand and function during the HELB event. The field. instruments shall be environmentally qualified. Therefore, the margins of safety or the effects of a HELB in the Intermediate Building have not been reduced or changed.

The modification is designed primarily to improve indication of auxiliary feedwater flow to the operator. As part of this modification the existing bistable device that interfaces with the control system ( for each auxiliary feedwater pump) is being functionally replaced. This replacement will in no way affect the operation of the auxiliary feedwater pumps either in manual or automatic operation. Therefore the consequences and safety margins resulting from those events which require auxiliary feedwater for their mitigation-have not been changed.

Therefore, the margins of safety during normal operations and transient-conditions anticipated during the life of the plant have not ben reduced. The adequacy of structures, systems, and components provided for'the prevention of accidents and'for the mitigation of the consequences of accidents have not been affect 9d.

EWR-2606C POST ACCIDENT SAMPLING SYSTEM IMP _LEMENTATION This design consists of providing structural supports to post accident sampling system (PASS). The. support will

.be designed to insure its structural integrity in thef event of. a seismic occurrence. ,

The new PASS implementation is non-seismic;..however, the PASS support is ' designed to insure its structural integrity and the integrity of existing Seismic Category I structures in the event'of a seismic occurrence.

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted ,

Without Prior Approval '

R. E. Ginna Nuclear Power Plant e Unit No. 1 I

Docket No. 50-244 Page 5 of 55 Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the plant have not ben reduced. The. adequacy of structures, systems, and components ~provided for the prevention of accidents and for the mitigation of the consequences of accidents have not been affected.

EWR-2607A CONTAINMENT HIGH PRESSURE INDICATION The purpose of this modification is to provide a wider range indication of containment high pressure. Specifically-this modification involves the upgrading of existing containment high pressure (0-90 psig). transmitters and main control board pressure indicators.

i Transmitters PT-946, 948, and 950 measure containment pressure 4

(via tubing penetrations) and convert that measurement'

to a 10 - 50 m DC signal.

The pressure signal generated by_the pressure transmitter and the power supply ((PQ) is fed to both a duplex' alarm and.an isolation amplifier. The duplex alarm (PC), (which contains two separate, adjuste.ble alarm points) change state when the input exceeds either alarm setpoint. These state changes are fed to the logic for streamline isolation and containment spray. The isolation ' amplifier (PM) acts to isolate the transmitter signal from the signal going

. to the indicator and recorder mounted in the control room.

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-1983 Annual Report of Facility Changes, .i Tests, and Experiments Conducted

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Without Prior Approval R. E. Ginna Nuclear Power Plant 4

Unit No. 1 1 Docket.No.'50-244 i

Page 6 of 55 l This modification shall be designed to indicate containment

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pressure over the range of 10-200 psia.

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This modification shall be designed to be operable during i the following modes' of operation s normal power, s h u t d o w n ,-- l l

startup and accident.

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j The new transmitters shall be seismically. qualified with-4 an analysis to show the installation of these transmitters

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does not impose any'new significant loading on the existing

transmitter supports._ Therefore,.the margins of safety during a' seismic event have not been reduced.

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I Each of the transmitters ~ replaced by-this~ modification i feed a dual bistable device. This bistable has two outputs 1 which independently change state when the input exceeds ~

{ each output's setpoint. One output feeds to the.steamline l: isolation logic, while the other to the containment' spray

{ logic. Any event, (such as LOCA) which 'results in containment l pressurization will potentially result -in actuation' of .

O these_bistables. Specific analysis hasl demonstrated that i these new setpoints and'any.new instrument errors will=

result in actuation of this? system within the tolerances j'

established by'the original" accident analysis. In addition, testing to verify proper = actuation ^ of' this portion of the

} plant' protection-system shall be performed. Therefore, i- the' margin of safety for those events which' result 'in contain- -

i l ment pressurization and rely'on main' steam'line.and/or.

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a initiation of containment spray hasinot been reduced by this modification.

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! Therefore,- the margins- of saf_ety during normal operations-

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1 and transient conditions' anticipated -during :.the lif e L o f

the plant have:not been reduced. The adequacy' of structures, ~

t systems, and components provided--for.the prevention of' accidentsLand.for the, mitigation of the_ consequences of '

accidents'have'not been affected.

} EWR-26_08A'HIGH RANGE EFFLU_ENT MONITOR

NUREG 07 37 required ' that ' noble gas ef fluent monitors with-1 .

an upper range capacity-of'10 5 A ci/cc ~ be installed' on ' .

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-the' plant. vent 'and containment vent exhaust stacks at' Ginna '

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval ,

R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 7 of 55

a. Noble gas effluent monitors with an upper range capacity l of 10 5h-Ci/ cc (Xe-133) are considered to be practical- i and should be installed in all operating plants.-

b. Noble gas effluent monitoring shall be provided-for ,

the total range of concentration extending from normal.  :

condition (ALARA) concentrations to a maximum of lg5 A.Ci/cc (XE-133). Multiple monitors are considered to be necessary to cover the ranges of interest.

Each effluent monitor will continuously. sample the. air in-the vent and analyze it for particulate,-iodine, and- i noble gas concentrations. Control- terminals in the Control Room and Technical Support Center will provide the' automatic logging function and the operator-to-system. interface..

The modification has been reviewed to ensure.that failure of any electrical cable installed as a part of this modification  ;

will not result in the disabling of. vital equipment needed ,

to safely shut down the plant during postulated fires.' .i Cables for..this modification' will be installed per IEEE-384 and isolated at the power' source with appropriate' isolation =

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devices. No vithi . equipment cabi'es will be -used in this modification which have not been reviewed under a fire protection safe shutdown analysis.

Seismic qualification of the monitors is'not. required because none of the events for which the monitors are required to. operate are postulated to be caused by a seismic'evpnt.

The primary. purpose for -installation of - thef high range effluent monitors' is to quantify potential releases following

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a LOCA with. extensive fuel damage or a fuel handling accident.

For these types of accidents .the monitors will' not be' exposed to a harsh environment and therefore"will remain. operational., '

l A .LOCA with tuel damage' or a - fuel headling accident .'could, j cause the containment vessel to become a-radiation source to-the effluent monitors. .over the 40 year life of-the l Plant, the total- dose accumulated by the monitors, -including i the dose from a LOCA, is calculated to be"118 rad.- The dose is negligible'with respect to thresholds' for . material ~ ,

i degradation in this system. Th e r e fo re', this modification l will'.not be adversely'affected by ailoss'of. coolant accident. ,

or fuel

handling' accident. ,

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval-R. E. Ginna Nuclear Power. Plant Unit No. 1 Docket No. 50-244 Page 8 of 55 A high energy line break in the intermediate building could subject the effluent monitors' to an environment of ~ high' temperature (215 degrees F) and humidity. High energy line breaks are limited.to relatively small_ break sizes in the intermediate building by the Augmented Inservice Inspection Program. High range radiation releases are not anticipated during this situation. Alternate techniques for radiological release assessment are available under existing plant programs for routine and emergency monitoring.

These would include deployment off portable instrumentation for sample collection and radiation measurement to determine in-plant and offsite radiological levels. . Releases from small high energy line breaks outside containment may 'be-dominated by the break flow and not the releases through

• the plant to containment vents.

It has, therefore, been determined that the margins-of safety during normal operations and transient conditions anticipated during the life of the station have not been affected. It has also'been' determined that the adequacy '

of structures,, systems, and components.-provided-for the-prevention of accidents and the mitigation of the consequences of accidents have not.been'affected.

EWR-26_0_8_B_ HIGH RANGE _C_ONTAINM_ENT M_O_ NIT _0_R_

In-containment radiation level monitors with:a maximum range of 107 R/hr (Photon only)'shall be installed. A minimum of two such monitors that are physically iseparated shall be provided. Monitors shall be designed and qualified to-function in an accident ~ environment.

The modification does not increase the possibility or impact of a. fire because additional wiring and' cable.that-will

'be added ' in this modification, - which could add to the fire loading of the plant shall meet the IEEE 3 8 3 -19 7.4 . f l am e - ,

J test requirements. Because of. this there_ will- be no increase of fire loading' caused by this modification.

The. modification . does :not, increaseJ the impact of a seismic event because tho' modification shall be. seismic.

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r . 1983' Annual Report of Facility Changes, l' Tests, and Experiments Conducted F. Without Prior Approval. .

! R. E. Ginna Nuclear Power Plant

} Unit No. 1  ;

3 Docket No..'50-244 i

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o Page - 9 of . 55"  !

- t This modification does not. increase the possibility or

impact of an accident inside containment because two redundant i' systems shall be qualified to -withstand the full spectrum i of accidents inside of containment and still perform their designed . function.

i I It has, therefore, been determined that the margins of safety during normal operations and transient conditions j:

j_ anticipated during'the life of the station have not been '

j affected. It has also been determined that the adequacy -

! of structures,. systems,'and components-provided for the ,

j prevention of accidents and the mitigation of, the consequences  ;

j of accidents have not been affected.

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i EWR-2709 ZIRCONIUM GUIDE TUBE INTERLOCK MODIFICATION INSTALLATION _ l l With the installation of Zirconium Guide Tubes in the fuel

! assemblies for the 1981 refueling, a condition exists which j might lead to damage of the -rod drive mechanisms if a cooldown r were to occur with the control rod-drivesLlatched. 'This j problem arises from-the differentLthermal expansion rates j of zirconium.versus stainless steel and results in an inter-i farence-problem which would place. unnecessary stresses j .. on the rod control cluster assemblies.

i Current ' administrative procedures exist which require 'the

operator to actuate the manual reactor trip push-button. ,

1 prior to cooldown. This EWR is for the ' design and installation - ,

j of an autanatic interlock to ensure the reactor trip breakers are open prior to cooling down.

The reactor trip and annunciation -are actuated by two inde-

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pendent trains each containing a' T hot and a P7 permissive input. T hot channels 409A and 410A:will be set to provide-a trip signal whenever the hot leg' temperature:is less ,

than' 568 degrees F. In' addition, this ' signal willi be blocked by permissive circuit P7 when 2/4 power range '~ channels are greater than 8.8t' or 2/2 turbine 'first stage. pressure ~ i channels are greater than 8.88. The main function'of this-modification is to- automatically' open one ' or both of .the ' -

reactor trip breakers when T hot falls below-588; degrees.

F 'if the operator has not already - done.so, h'owever, this

' design will'also' prohibit..:the operator from closing.the f

r 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 10 of 55 breakers during startup until T hot is above 500 degrees F. This reactor trip will be annunciated on the "first out" annunciator panel for as long as the signal persists to pro' ride the operator with a visual reminder that the reactor trip breakers cannot be closed.

The trip logic in each train will be actuated by contacts from electromechanical relays wired into the trip circuit if each reactor trip breaker. For increased reliability, redundant T hot and P7 relays are used.

This modification has been reviewed to ensure that failure of any electrical cable installed as a part of this modification will not result in the disabling of vital equipment needed to safely shut down the plant during poetulated fires.

Cables for this modification will be installed per IEEE-384 and isolated at the power source with appropriate isolation devices.

New instrumentation installed for this modification shall be qualified to IEEE 323-1974 and IEEE 344-1975 and therefore will not increase the impact of a seismic event.

This modification is designed so as not to block any reactor trip signals at any time. None of the existing reactor trip channels has been modified or utilized in any way.

The only possible failure mode for this modification is to cause a trip, therefore, existing reactor trips are not afft,ted.

It has, therefore, been determined that the margins of safety during normal operations and transient conditions anticipated during the life of the station have not been affected. It has also been determined that the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents have not been affected.

EW R-2 7_7_0_ CONTAINMENT __ PERSONNEL _ HATCH CO_NTROLLED_ ACCESS The modification will provide Security personnel monitoring controlled access through the Containment personnel hatch greater convenience. The design will interface with the existing Ginna Station security system, and will provide e

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'l 1983 Annual Report of Facility Changes, l Tests, and Experiments Conducted 3-Without Prior Approval R. E. Ginna Nuclear Power Plant

Unit No. 1.

Docket No. 50-244

! Page 11 of 55 control and identification of individuals entering and 8.

leaving Containment via the personnel hatch. .

i-Modes of operation shall be controlled-from the existing i security system.

i Additional wiring and cable will be added -in ~ this addition,

which could add to the fire loading of the plant. However, i all such cable shall meet the IEEE 383-1974 . flame test j' requirements. Because of this~there will be no increase' 1 of fire loading caused by this addition.

The addition does not increase the impact if a. seismic event. None of the equipment associated with this addition i is required to be functional during a seismic event and J

failure of this equipment during a seismic event will not degrade existing Seismic Category I structures or' equipment.

1 It has, therefore, been determined that_the margins of j safety during normal operations and transient, conditions anticipated during the life of the station have not been j affected. It has also been determined that tho' adequacy.

! of structures, systems, and components provided-for-the prevention:of accidents and the ~ mitigation of the . consequences - r of accidents have not been affected.

I j EWR-3021 DIESEL GENERATOR COOLING _

i This modification provides an alternate source of cooling-water for the diesels- that ' is independent'of the service

< water. system. It consists of adding a tee, isolation' valve

-and a fire hose fitting between the diesel generator. heat . .

exchanger and the. service water isolation valve for each j generator.  :

j This modification' will provide a continuous source ofiwater-to the punps that is : independent .of ' the service water system.

Th e modification consists ' of... adding a' fire. hose z fitting

. ' and any necessary valves or fittings to the six inch condensate-i line f rom the ~ condensate supply ' tank to the' suction ?of '

the pumps.-

3 The imodification shall be designed to be used in -the ' event- '

of loss of. station service water.- ,

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 12 of 55 The proposed modifications neither penetrate any existing fire barriers nor affect any existing in plant fire suppression systems. The modifications do not increase any previously determined fire loadings.

The modifications are not affected by-a tornado or any other storm or flood, and held ~ mitigate the consequences of these events.

The consequences of this event are not increased by the diesel modification because the modification will be designed to Seismic Category I standards.

The consequences of this event'are not increased by the feedwater modificatiot: because the modification shall be attached to a non seismic portion of piping and shall be located so as not to affect adjacent seismic equipment.-

The proposed modifications do not adversely effect the diesel generators availability to : provide emergency on-site power in .the event of -loss of AC power. The proposed modifi-cations are not adversely affected by the loss of AC power.

The proposed modifications will not, be af fected by any high or moderate energy line breaks. This' modification will insure the availability of cooling water should postulated MELB'.s or HELB's cut off the- supply of, service water to the diesel' generator cooling system, as well as to the steam generators via the Standby Auxiliary Feedwater Pumps.

i-The' consequences of this event are not increased by these 4 modifications. The-feedwater modification'shall provide-j a source of feedwater 'should service water not be available 1 for this purpose.

Therefore, the margins -of safety during . normal operations and transient conditions anticipated during the life of the' plant have not been reduced. The adequacy of structures, .

, . systems, and components provided for'the prevention of.

i  : accidents _and for the~ mitigation of the consequences of accidents have not been-affected.

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without prior Aoproval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 13 of 55 EWR-3 0 3 0_ REACT _OR VESSEL HEAD _ _S!!IELD_

The purpose of this modification is to install a lead-wool blanket radiation shield around the reactor vessel head

( RVII) af ter the thermal insulation is removed for shutdown.

This would be lifted with the RVH when it is moved to the storage pedestal. The shield is composed of nine chain mounted blanket modules to surround the head for full 360 degrees, with the blankets overlapping. The chains are attached to support ring brackets, which are clanped onto the RVII lifting members. The modules would be stored in steel chests (three per chest) with casters for ease of movement insido containment. The shield would reduce the man-rem exposure by a factor of 3:1, or more if a control Rod Drive (CRD) Mechanism shield is installed. This would be placed af ter the RVII is on the storage pedestal, and consist of the same type of blanket modules. The CRD shield would be suspended from the detensioner hoists, which would not be in use at that time.

This modification shall be designed to retain structural integrity during cold shutdown and fire.

The conta.inment floor underneath the Reactor llead storage pedestal will be analyzed and determined able to safely carry the additional imposed load of the RVil Shield. The Radioactive Materials Storage Building will house the shield storage chests during normal operations. A review of the movment of the chests to and from containment will determine that the contamination limits are within those prescribed by the Radiation Protection Manual, and as per the Ginna Station Administrative Procedure A-1, and other applicable Ginna Station llealth Physics procedures and practices.

The containment crane has been analyzed to determine that the additional load of the shleid is within allowable capacity.

The current licensing issue " Control of Ileavy Loads" may require additional changes to the containment crano due to these modifications ( e.g . , in the area of load testing) or that administrative controls be imposed on the installation of the shields so that head lif t does not occur with the shields in place. The modification is acceptable at this time although additional changes may be required in the future to implement the requirements of Control of Ileavy Loads.

. l 1983 Annual Report of Facility Changes, Tests, and Experimento Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 14 of 55 The margin of safety for maintenance, normal operations

.and transient conditions will be increased as a result of this modification. The adequacy of the structures, systems and ccxnponents provided for the prevention of accidents and the mitigation of the consequences of accidents will not be affected.

EW R-3 0 57_ RADWASTE STORAGE _ FACILITY The purpose of this modification is to install a pre-engineered metal building over the existing fenced in area currently used for low level radwaste storage. The building will contain a bridge crane with a 3 ton capacity. This modification is required to allow plant personnel versatility in moving radwaste in and out of the existing bunker and to allow loading in a dry area.

Presently, all operations involving moving of any material must be accomplished by the acquisition of a crane limiting the loading operations pending its availability. In addition stringent restrictions imposed by the NRC pursuant to non-corrosive free liquids, which have been placed on the waste material packaged for transportation to burial sites eliminates loading the material during inclement weather. The problem of coordinating the acquisition of a crane with good wenther

, will be eliminated. This modification will allow handling l of radwaste when needed.

l The proposed modification will cover the existing fenced-in-area i presently used. No substantial new storage area vill be provided.

This modification shall be designed to be functionable during the following: normal operation, shutdown, startup and in case of fire but not during or after a seismic event.

i The radwaste storage building will be located over the existing storage facility approximately 300 feet northeast l of the plant superstructure.

This modification will be a separate unit. No connections of any utilitics will be made to any Seismic Category I structures. The failure of this structure will in no way affect the structural integrity of the plant.

1983 Annual Report of. Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1

. Docket No. 50-244

--Page.15 of-55 All fire sprinkler systems and extinguishers shall' be installed in accordance with NYS Building Codes and Station Operating Procedures. B-labeled doors will be used at two access points to the building in the event of a fire. l At no time will a potential- failure of the Radwaste Storage Building influence the operations of the plant. - ' Packaged '

low-level radioactive wastes stored in the Radwaste . Storage Building will consist of solidified waste evaporator bottoms, solidified liquid waste filters, and dry. waste. Spent resins will be stored in the facility only after solidification and packaging, or af ter dewatering and placanent in a certified high integrity container and shipping cask.

In the event of a fire, the. remote location of the Radwaste Storage Building in relation to -the main plant superstructure eliminates any adverse effects which may be caused by such an occurrence. The proposed building structure will contain-an adequate fire protection system and components provided for the mitigation of fire emergencies. External ^ contamination levels on the outside of stored radioactive waste packages -

are administ'ratively controlled by plant procedures.and practices and will not pose undue radiological . hazard if sprayed with fire protection system water.

During a postulated seismic event, the failure of the Radwaste Storage Building would not release significantly more radio .

active-contamination to the environment than that which would be released'from packaged waste material housed inside the plant Auxiliary Building.-

A postulated rupture or' puncture -of a low-level radioactive waste container during handling would entail: n_o significant release of airborne radioactivity from the'aadwaste storage Building. Waste evaporator bottoms, liquid' waste ~ filters and resins not otherwise stored ~in certified high integrity.-

containers and shipping casks would be. bound in - a solid. matrix ~

-with no free-standing water. This would prevent ~ the off-site release- of Lsignificant inhalable quantities Mof ' radioactive ;

material in the event-the waste container'is ruptured or

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puncturer- during handling. - OtherL material' stored -_in' the Radwaste - Building would consist of packaged dry waste ,(e.g. ,

~

tools, rags, clothing, discarded ~ equipment) or' wrapped; and/or packaged plant ancillary ' equipment. LSimilarly,

__ - = -

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 16 of 55 the nature of-the dry waste materials would preclude the off-site release significant inhalable airborne radioactivity.

The margin of safety for maintenance, normal operations and transient conditions will be increased as a result of this modification. The adequacy of the structures, systems and canponents provided for the prevention of accidents and the mitigation of the consequences of accidents will not be affected.

EWR-3163 SOUTH PLANT DOOR #25 MODIFICATION The purpose of this modification is to nullify: the problems

• associated with door closure on door #25 during high wind conditions by installation of an electrically opened sliding door. The door and frame shall be of aluminum and glass construction. equal to that existing in the area. This modification will provide ample access to the plant, while maintaining the high degree of security required for entrances.

The number of open door alarms should decrease drastically.

In order to install the sliding door, theLexisting doors and frame will be removed including the-transom. The new door will be installed, sliding horizontally upon actuation by electronic sensor. The existing access control and alarm equipment will be removed from the exterior door and installed'on the' interior door. The interior' door will be security door S25.

The addition does not increase the possibility or impact of a fire.

Additional wiring and cable will be added in the addition, which could add to the fire loading of the plant. Therefore, the design criteria requires that all . such cable meet' the IEEE 383-1974 flame test requirements.- Because of this there will be no increase of fire loading caused by this addition.

The addition does not increase the impact of a seismic

~

event. None of the equipment associated with this additions required.to be. functional during a seismic event and failure-of this equipment during a seismic event will not degrade existing seismic category I structures or equipment. ,

2

b,-

4 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted

. Without Prior Approval R. E. Ginna Nuclear. Power Plant

l. Unit No. 1 Docket No. 50-244 Page 17 of 55 It has, therefore, been determined that the margins of safety.during normal operations and transient conditions i .. anticipated-during the life of the station have not been -

affected.

EWR-3 418 REFUELING WATER STORAGE TAN _K LEVEL INDICATION _

This modification will require a second refueling water j

storage tank level indication system and a second sett of.

alarms be installed at Ginna Station.

The existing refueling water storage _ tank (RWST)' level indication system consists.of a level transmitter-(LT-920) connected to the RWST which signals. a . percent-scale level indicator on the main control board and actuates two annun-ciator alarm windows ("RWST Lo Lo Level" and "RWST Hi/L'o-3- Level") on the main control board by way of two bistables,

! one corresponding to each alarm window. There is also a differential pressure unit linked to the RWST on al separate line which actuates one of the same annunciator alarm windows j ("RWST Hi/Lo Level"_) at LT-920.

i The.second RWST level indication' system will replace the

differential pressure unit presently linked to the RWST.

! This modification shall designed with the_same-actuation

levels and actuate the same'two annunciator alarm-windows-

as the first RWST level indication system.

! ~

This modification has been reviewed to ensure that' failure i of any electrical cable installed as a part of this modification will.not' result in the disabling ofivital equipment needed to-safely shutdown.the plant during postulated fires.

Cables for- this modification will be installed per IEEE-384-1977 and. isolated at' the power source with appropriate i solation

- devices.

1

! Each of the annunciator window's is provided with two _ light l bulbs although each annum:iator? is dependent upon a single -

circuit for proper operatlon'of the " horn silence", "acknow a ~

=

ledge" and " reset" buttores.' The redundant RWST-level inputs to the alarm, the redundant'anpunciator power supplies,;

and the trouble free annunciator _ operation overMai ten year w

period provide reasonable - assurance that the ' alarms' will.

alert the operators. In addition; to thenalarms, Lredundant' e- y -sew + p, pge, 9 n., -w-wey y- e.i 3  %- ww- -gr-+ + - - g'y - w yv79* v r ~--

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4 j' 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted

Without Prior Approval R. E. Ginna Nuclear Power Plant E Unit-No. 1 Docket No. 50-244 Page 18 of~55 level indication will provide the-information necessary_

for the operator to complete the required tasks.

The new level transmitter, the signal processing modules, i and the-level indicator installed for this modification f

shall be scismically qualified. _In addition, the new-level transmitter, conduit, and ' instrument tubing shall be seismically supported. Therefore, this modification does not increase j the impact of a seismic event.

I This modification has been reviewed to' ensure that failure-I of the level transmitter, the signal processing modules,.

che level indicator, or the electrical cable installed

i. as a part of this modification will not result in disabling '

of-vital equipment needed.to safely shut down the plant during postulated fires or a seismic event._ ,

! The design requires that equipment installed' for this modi-

! fication be located away from areas--subject to HELB, therefore this modification will not be affected by. HELB.

l It has, therefore, been determined-that the margins of

! safety during normal operations and transient conditions anticipated during the life of the station have not .been' l affected. It has-also been determined that the: adequacy >

l of structures, systems, and components provided for the '

prevention of accidents and the mitigation of the consequences of accidents have not been affected.

A j - EWR-3 716_ 3A LOW PRE _SS_URE FEEDWATER HEATER

{ The purpose ' of this modification Lis to maximize T the effect-iveness' of low pressure feedwater heater 3A and .to maintain performance by increasing the reliability of the tubing.

As . o f July '15,

~

1982, -4.6% of. the 3A heater's ~ tubes.were 2

plugged. This results in decreased heat transfer.~ area and' increased feedwater-system pressure losses.

The restored-3B low pressure feedwater heater ' in storage i at Ginna Station shall be rebuiltL and used to replace the 3A heater. The existing.shell and channel head of the 3B' heater shal1~be reused. A- new tube bundle 'shall - be fabricated along with'a new tube sheet.

t

. . .- - - - . - . ~ - . - -

l f: i .* .

1 l 4

3 1983 Annual Report of Facility Changes,

Tests, and Experiments Conducted Without Prior Approval.

R. E. Ginna Nuclear Power Plant Unit No.,1-l Docket No. 50-244 Page 19 of 55-i A primary consideration in the rebuilt heater shall be-

the tube material. Current design specifies admiralty. *

(a copper zinc alloy). Copper has been determined to be i

a contributing factor in the corrosion process which causes steam genetator tube wastage.. -Stainless steel'shall be

specified as the new tube material due'to its corrosion

! resistant charact' eristics.

l The length of the heater shell and tube bundle shall be i extended approximately 44 inches to compensate for.the j change in thermal performance of. stainless steel versus admiralty tubing. The entire.-44 inch extension'can be '

i accomodated entirely within the condenser neck, the present location of heater 3A.

i Fe edwater temperature decrease, is analyzed in the FSAR'-

j' as the accidental _ opening of the' condensate bypass valve j 3959, a fail open,, air operated, diaphragm actuated control j valve; or the accidental movement' of the feedwater-control-

! valves to the full open positio_n. .. The effect of either i event would be to deliver feedwater to the-steam' generators l at-a reduced temperature. An unsafe' condition would! result-j due to excess heat. removal from the primary. system.

i i Installing a new 3A heater has no effect on.the actions

[ of valves mentioned above.

i

, Loss of normal feedwater flow, is analyzed in~.the FSAR l as that accident (pipe break, pump failure,,valveLmalfunction,

, or loss of outside ,ac power) which results in?a reduction .

in capabilityLof the secondary system to remove heat generated ?

in the core.

i e -

Neither the consequences nor the margins ofisafety have. "

<g been changes for this event since this_ modification..does i not introduce additional piping or equipment to;the plant.,

An: accident identified with the 3A' heater fis . that' of a multiple, double ended ' tube rupture.: This'accidentLwould require - the ' removal of the heater from operational-' service.

! Removal of,the 3A heater from service requires:the use 1 of' manual isolation. valves' which would also isolate the -

! 1A, 2A and 4A.' heaters. However, since turbine operating.

constraints - require load reduction -for . removal from service

a

. . .- . - - ,. . . _ , - . - . - _ - . ,- s-- - - - - . . . - , . ~ . ,_ - --._---w, . - .,,

. l 1

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 20 of 55 of two or more adjacent heater, an unsafe condition would not be created. The modification does not increase the consequences of the event, does not reduce the margins of safety provided for the event, and does not present an unreviewed safety question.

Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the plant have not been reduced.. The adequacy of structures, systems, and components provided for the prevention of accidents and for the mitigation of the consequences of accidents have not been affected.

.1983 Annual Report of. Facility Changes, i Tests, and Experiments Conducted '

Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 21 of 55 TSR-80-26 INSTALLATION OF NITROGEN S_UPPLY TO__ ACID AND CAUSTIC TANK LEVEL BUBB_L_ERS_

This modification involves the installation of a nitrogen bottle supply for the bubbler level indicating systems of the acid and caustic tanks in the water treatment room.

The present air supply to the bubbler level indicating systems is causing corrosion.

The nitrogen bottles are used to supply pressure to the bubbler level indicating systems for the acid and caustic storage tanks, The pressure regulator on the bottles reduces the pressure supplied to the level indicating system to the pressure of instrument air (100 psig).

The bubbler level indicating system will be supplied with a constant pressure of 100 psig by the nitrogen-bottle regulator.

Control of the level indicating system will be unchanged by changing the bubbler to a nitrogen bottle supply.

The modes of operation of the level indicating system are unchanged for various plant conditions.- The system is permanently valved in and indicates level in the caustic and acid' tanks independent of plant conditions.

This modification involves replacing the instrument air supply ' to the level bubbler system for the acid and caustic tanks with a bottled' nitrogen' supply. The water' treatment system is not required for safe shutdown of'the plant to.

for long.-term cooling o f. the - plant a f ter an accident.

The water treatment system is not required to be operable post-seismic event. Therefore, these-level bubbler systems are not safety related.

Therefore, a) the margins of safety during normal operations and transient conditions ~ anticipated _during~the life of the station will' be unchanged by the installation ^of this modification, b)'the-adequacy of structures, systems and components provided for the prevention'of~ accidents and

~

for the mitigation of the consequences . of accidents will be unchanged by the installation oflthis modification'.

i 1._ -

O.

O' 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 22 of 55 c

TSR-8_2-07 STATION _ EMERGENCY D/G AIR INLET FILTER _

ASSEMBL_Y_

This modification involves the replacement of the air inlet-filters for the station emergency diesel generator units with an upgraded model. The present generator manufacturer has specified a replacement filter for this application.

The replacement filter has different dimensions than the original filters, so a replacement bolt-on housing for the filter is also being installed. The housing is a bolt on flange assembly and the filters are a washable type element. The filter and housing are a vendor supplied replacement for the present assembly.

This modification will not change 1) the assumptions in any safety analysis in the FSAR and its supplements 2) the probability of occurrence of an accident, and 3) the consequences of an accident. The filter replacement is a vendor supplied like for like (like for better) replacement

and will have no adverse af fect on the operability of the diesel generator unit.

Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the station will be unchanged by the installation of this

modification. The adequacy of structures, systems and components provided for the prevention of accidents and for the mitigation of the consequences of accidents will be unchanged by the installation of this, modification.

TSR-83-01, LOCKED HIGH_ RADIATION _ AREA EXI_T LATCH This modification involves the installation of 15 new " Logan mortise" locks on the barrier doors (gates) to all locked high radiation areas, to replace the existing padlock on gates located in the auxiliary and intermediate buildings.

The new lock will have a recessed knob on-the inside of the gate so that the gate can be opened from_the inside without a key. This will prevent individuals from becoming locked inside a high radiation area. Locked high radiation areas will still remain under the control of the shift supervisor via "R" keys.

, - . - . +

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 23 of 55 This modification will not change 1) the assumptions.ir.

any safety analysis in the FSAR and its- supplements , 2 )

the-probability of occurrence of an accident, and 3) the consequences of an accident.

Therefore, a) the margins of safety during normal operations and transient conditions anticipated during the life of the station will be unchanged by the installation of this -

modification, and b) the adequacy of structures, systems and components provided for the prevention'of accidents and for the mitigation of the consequences of accidents will be unchanged by the installation of this modification..

TSR _83-03 INSTALLATION OF CONTAINMENT EQUIPMENT HATCH AREA 480 VOLT _ POWER S_UPPLY This modification involves installation of a 480 volt,.

~

3 phase supply on the exterior north wall of the auxiliary building near the containment equipment hatch. A permanent supply is needed for the we3. ding and for other 480 volt loads-in this area. Temporary cables are routed to this-area along the ground at present from the' auxiliary building MCC.

MCC lE will provide a source of power and the MCC breaker .

will provide a means for disconnecting power to the disconnect switch. The disconnect switch itself is ~ suitable for outdoor service and is the connection' point fori 480 -volt 'lo' ads in the area. The disconnect switch is equipped.with replaceable

' fusible elements.which provide interruption capability' in the event of'an overload.

The disconnect switch is' energized by closing the breaker at position 6DD of MCClE. The switch-is also equipped with a _ disconnect handle which must'be.closedLto supplyf 480 volt power to the loads.

The disconnect switch may2be energized.oride-energized-at the MCC depending 'upon the. need . for 1480. volt power .in - ,

the vicinity of the equipment hatch.-  !

This modification involves installation of .a 480 volt 3 phase disconnect switch to be located on-the exteriorinorth wall of the auxiliary building near the containment equipment ' i

' hatch. The switch will be. connected-to aJnon-vital.MCC. I l

l i

L I

2 .-

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted.

Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244  :

Page 24.of 55.

powered-from a non-vital. bus (bus 15)' so it will:have no ,

safety implications for any. failure mode. Bus 15 is not required to be operable per the requirements of Ginna Station ' l Technical Specifications.

Therefore; a) the margins of safety during normal operations and transient conditions anticipated during the lif e ' o f.

the station will be unchanged by ;the installation of this modification, b) the adequacy of' structures, systems and t components provided for the prevention of accidents and for the mitigation of consequences of accidents will be unchanged by the installation of this modification.

PLANT MODIFICATIONS PARTIALLY COMPLETED IN 1_983_ 1 SM-1444.2 .

UNDERV_OLTAGE RELAY _M_ODI_FICATIO_N_ _ INSTALLATION '

OF CABLE _ TRAY, C O N D_U_I T , _ SUPPORTS,_ WIRE

. PULLING AND TERMINATIONS The purpose of this procedure is for installation of cable trays, conduit, supports, wire pulling, and termination '

for the undervoltage relay modification.

This completed modification procedure was reviewed by the PORC committee and no unreviewed' safety questions, technical specification changes or violations.were involved'.with this change to the facility.

SM-1660._1_2_ INSTALLATION OF BOLTED SUPPORT _PG-52_7_

VALVE. #846 The purpose of this procedure is for the installation of

• removable support PG-527 on valve 846 'of the; ,RCS overpres-sure protection nitrogen system on the Intermediate floor ,

of the Auxiliary Building near Penetration 129.

-This completed modification procedure was reviewed by the PORC committee and no unreviewed safety" questions, technical specification changes or violations were. involved'with- ,

this change _tolthe. facility.

4 t'

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l

T 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 25 of 55 SM-183_2_B_._012 ACCEPTANCE TES_T _P_ROCED_URE FOR _F_U,N_CTI_ONAL TESTING THE VALVE _ _ TAMPER SWI_TCHES

- PHASE _I_

The purpose of this procedure is for acceptance testing phase I valve tamper zones of the fire signaling system, to verify electrical supervision of valves to verify valve tamper indication in the control room to verify Class B electrical supervision of valve tamper switches. To test the fire signaling system and in accordance with the man-ufacturer's operating and maintenance manuals.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-_183_2_B_.061_ CONVER_S_I_0_N PRO _CEDURE _FOR FIR _E PU_MP AU_TO STARTf RELAYS FPRl_, _ "_'/R2, _ AN_D_ FPAS_

The purpose of this procedure is for conversion of the existing fire pump control to the new fire pump auto-start relays.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes'or violations were involved with this change to the facility.

SM-183_28._78 CONVERSION _OF EXISTING _ SYST_EM_S15 FROM__

AUTOMATIC DELUGE _TO_ _P_R_E-ACTION The purpose of-this procedure is to convert system 515

• from automatic deluge to pre-action providing supervision of the new air pressure switch S15P.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

._u

i 3

e .

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 26 of 55 SM-18328.79 CONVERSION OF EXISTING SYSTEM SO3 FROM AUTOMAT _IC DE_L_UGE TO PRE-ACTidiT ~ ~~~~

The purpoco of this modificati;n procedure is to convert system S03 from automatic deluge to pra-action providing supervision of the now air pressure switch S03P.

This completed modification procedure was rovinwed by the PORC Committoo and no unreviewed safety questions, technical specification changes or vio'.ations wore involved with this change to the facility.

SM-1832B.80 CONVERSION OF EXISTING SYSTEMS SOL AND SO4 FROMAUTOMATIC__DELUGET_0_ PRE-ACTjoj The purpose of this procedure in to convert systems 801 and SO4 from automatic deluge to pro-action providing super-vision of the now air prosauro switches S01P and SO4P.

This completed modification procedure was reviewed by the PORC Committoo and no unreviewed safety questions, technical specification changes or violations were involved with '

this change to the facility.

SM-_1832B_.86 MODIFYI,NO AND,TJSlI_NO VAINE _ TAMPER SWIT_Clj,Ff FIRE SI_G_NALIN_O _ SYSTEM The purpose of this proceduro in for modifying and testing the following valve tamper switches:

V9213 (809) V5200 V5100 V9125 (S10) V5209 (S27)

V9127 (Sil) V5229 (S12)

V5173 V5230 (S13)

V5187 V5232 (S25)

To verify electrical supervision of valvos to verify valve tamper indication in the control room to verify Class D electrical supervision of valve tamper switches to test the fire signaling system in accordance with the manufacturer's operating and maintenance manuals.

9' i-1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior _ Approval R. E. Ginna Nuclear Power Plant-Unit No. 1 Docket No. 50-244 Page 27 of 55 This completed modification-procedure was reviewed by the PORC committee and,no unreviewed safety questions, technical specification changes.or' violations were involved with this change to the facility.

I SM-1833.21- INSTALLATION OF ELECTRICAL EQUIPMENT SPRAY S HIEL_D_S_

l 'The purpose 1of this procedure is for the installation of electrical. equipment spray . shields for the Fire Suppression System.

! This' completed modification procedure was reviewed by the j PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

i SM-1833.23 REPLACEMENT OF VALVES V5187 AND V5188 The purpose of this procedure is for~the repl'acement and testing of . valves V5187 and V5188.

{

This completed modification procedure was_ reviewed'by the PORC committee and-no unreviewed safety questions,. technical ~

specification changes or. violations.were involved.with 4

this. change to the facility.

SM-2421.3 HYDROGEN SYSTEM PIPING MODIFICATION'

.~ T h e purpose of-.this. procedure'is to perform the partial installation of piping . for the . hydrogen supply to L the VCT~

and hydrogen recombiner system from the new hydrogen bottle ,

storage bc31 ding.-

This completed modification _ procedure was reviewed by the -

PORC committee and no unreviewed safety questions, technical specification' changes or violations wereninvolved with this change to the facility.

-r'sav ss , y w ,r-e , ,y g w - 2, -g -

4 - - - - , . - - - + -

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 28 of 55 SM-2436.1 CAGE AND SAFETY BARRIER TO EXISTING LADDER "A" STEAM GENERATOR The purpose of this procedure is to install a protective cage and safety barrier to an existing ladder at the "A" steam generator.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2507.10 CONTAINMENT PENETRATION P-206, PRESSURIZER LIQUID SAMPLE MODIFICATION FOR_ VENT / DRAIN INSTALLATION OUTSIDE CONTAINMENT The purpose of this procedure is to install new supports .

and relocation of existing valves.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions,. technical specification changes or violations were involved with this change to the facility.

SM-2507.ll CONTAINMENT PENETRATION P-207, PRESSURIZER STEAM SAMPLE MODIFICATION FOR VENT / DRAIN INSTALLATION.

OUTSIDE CONTAINMENT-

'The purpose of this procedure'is tofinstall new supports and relocation of existing valves.

This compl'eted modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the' facility.

l

, SM-2507.14 INSTALLATION OF AOV-1599 l

The purpose of.this procedure is to provide the instructions l

to remove check valve 1599 ' and install AOV-1599

including-associated parts.- '

er - m )-r -vv V

! 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1  !

Docket No. 50-244 Page 29'of 55 This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-250_7_._15A REPLACEMENT OF_ TUBING ON PRESSURIZER STEAM SPACE SAMPLING _LINE OUTSIDE CONTAINMENT The purpose of this procedure is for the performance of the work associated with the removal and replacement and hydrostatic testing of portions of tubing on the pressurizer steam space sampling line outside of the containment building.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2507.16_ REPLACEMENT OF TUBING ON PRESSURIZER STEAM GPACE SAMPLING INSIDE CONTAINMENT (P-207)

The purpose of this procedure is for the. performance of work associated with the removal, replacement ~and hydrostatic testing of portions of tubing ~on the pressurizer steam space sampling line inside containment.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical' specification changes or violations were involved with this change to the facility.

SM-2507.17 REPLACEMENT OF TUBING ON PRESSURIZER LIQUID SPACE SAMPLING LINE INSIDE CONTAINMENT (P-206)

The purpose of this procedure is for the performance of work associated with the removal, replacement'and hydrostatic.

testing of portions of tubing on the pressurizer liquid space sampling line inside containment.

This completed modification. procedure-was, reviewed by the l .PORC committee and-no unreviewed safety questions,4. technical

specification changes or. violations were involved with this change to.the facility.

1983 Annual Report of Facility Changes, 1 Tests, and Experiments Conducted l Without Prior Approval l R. E. Ginna Nuclear Power Plant i Unit No. 1 Docket No. 50-244 Page 30 of 55 S M-2 5 0 7_ ._1_8_ CONTAINMENT PENETRAT_ ION P-100 CHARGING LINE TO HOT LEG, LOOP B The purpose of this procedure is to install a new check valve (9315) and a new drain valve (9318) in the charging line to hot leg, loop B piping.

This completed modificati.on procedure was reviewed by the PORC comm3ttee and no unreviewed safety questions, technical specification changes or violations were invol'ved with this change to the facility.

SM-2507.9 CONTAINMENT PENETRATION P-100 AUXILIARY SPRAY TO PRESSURIZER CHARGING LINE TO COLD LEG, LOOP B, CHARGING LINE TO HOT LEG, LOOP B The purpose of this procedure is to install a new valve (9313) and a new drain valve (9316) in the au'xiliary. spray to pressurize piping and a new check valve (9314).and a new drain valve (9317) in the charging line to cold leg, loop B piping and a new check valve (9315) and a new drain valve (9318) in the. charging line to hot leg, loop B piping.

This completed = modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were_ involved with this change to the facility.

SM-2512.75 SEISMIC UPGRADE OF THREE NEW PIPE SUPPORTS ON ANALYSIS LINE_FW-300; MAIN FW FROM FW REG.

VALVES TO B STEAM GENERATOR The purpose of this procedure is . for installation of three pipe supports on analysis line FW-300.

This ccmpleted modification procedure was reviewed by the ~

PORC committee and no unreviewed safety questions, technical'- -

specification. changes or violations were involved with this change to the facility.

m ,

-I i

. . , , l 1983' Annual Report of Facility Changes, i Tests, and Experiments Conducted )

Without Prior Approval  !

R. E. Ginna~ Nuclear Power Plant-Unit No. 1 i Docket No. 50-244 j Page 31 of 55 SM-2602.1 PIPE SUPPORTS MODIFICATIONS TO THE PRESSURIZER RELIEF LINE The purpose of this procedure is for the. work necessary to modify existing pipe supports.N-628, N-629, and N-630.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with-this change to the facility.

SM-2606.4 FLUSH AND HYDROSTATIC / PNEUMATIC TEST _

OF POST ACCIDENT SAMPLING SYSTEM TIE-INS The purpose of this procedure.is for the. performance of the post accident sampling system tie-ins . hydrostatic and pneumatic tests.

This completed. modification procedure was reviewed by-the-PORC committee and no unreviewed safety questions,-technical specific'ation chang'es or violations were involved with  ;

this change to the facility.

SM-2606.4A HYDROSTATIC TES_T OF POST ACCIDENT SAMPLING SYSTEM INSIDE THE CONTAINMENT BUILDING-The purpose of this test is to hydro-test'that portion of the sump "A" sample system located.inside the_ containment building.

~

~

Tnis completed modification ~ procedure was reviewed by,the '

PORC committee and no'unreviewed' safety questions,-technical.

specification' changes'or violations'were: involved'with this change to the facility.

i- SM-2606.4C FLUSH AND HYDROSTATIC / PNEUMATIC TEST OF' POST _

ACCIDENT SAMPLING SYSTEM BALANCE OF PIPING L The purpose -of this procedure s is for the. performance of l . hydrostatic / pneumatic tests of the post-accident sampling-

~

i balance of piping., _

+

g

). .

1983 Annual Report of _ Facility Changes, Tests, and Experiments Conducted.

Without Prior Approval R. E. Ginna Nuclear Power Plant-Unit No. 1 Docket No. 50-244  :

Page 32 of 55 1

This completed modification procedure was reviewed by the

[- PORC committee-and no unreviewed safety questions, technical specification changes or violations-were involved with

+

-this chang'e to the' facility.

SM-2606.5E = PASS WASTE TRANSFER PUMP AND P_AS_S WASTE _T_ANK' LEVEL SWITCHES __PREOPERATIONAL TEST.

The purpose of this_preoperational test is' to ensure the pass waste transfer pump, associated-piping and pass waste

tank control switches' will function in the manner.-intended

by design.

i.

} This completed modification procedure was reviewed by the

! PORC committee and no unreviewed safety questions, technical

i. specification changes-or violations were involved with this change to the facility.-

SM-2606.'5F POST ACCIDENT SAMPLING SYSTEM HEAT. TRACE-SYSTEM j . TEST-The purpose of: this procedure is to energize - the pass f. eat .

I trace' system and_ ensure that the ' system can- automatically -

. function to keep the temperature. of- the heat trace within l- specified tolerance levels.-

4 . .

rThis completed modification procedure'was reviewed by.the "

l PORC committee and-no unreviewed: safety questions, technical ~

! specification .' changes L or . violations were involved with .

this change to the facility.-

( SM-260_6.5H POST: ACCIDENT SAMPLING SYS'I EM INTEGRATED'(BOP):

[ START UP TEST.

i . / - .

L LThe purpose of~this[ procedure-_is;to verify the integrated, .

, performance'of.the~ sampling system?by obtaining reactorj

. coolant,. containment sump,i and' containment'airi_ samples -j

.under . actual _' operating . temperature "and ' pr.esstko i cond itions . :1 Specific _ objectives of;the.testLare:

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1 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1

' Docket No. 50-244 Page 33 of 55 Demonstrate the ability to withdraw a sample from each of the following sources:

o PRESSURIZER STEAM SPACE o PRESSURIZER LIQUID SPACE o REACTOR COOLANT HOT LEG LOOP "A" o CONTAINMENT SUMP "A" c CONTAINMENT AIR SPACE The following subsystems and/or equipment will be placed into service to verify their satisfactory performance while supporting the above sample operations:

o SAMPLE COOLER RACK, PAS-G-216 o LIQUID AND GAS SAMPLING PANEL (LSPG), PAS-G-214

. o INSTRUMENT AND PROCESS SUPPORT PANEL (IPSP), PAS-G-215 o ELECTRICAL CONTROL PANEL (ECP), PAS-G-217 o CONDENSATE WATER FLUSH' SYSTEM o WASTE HANDLING ~ SYSTEM, CONSISTING OF THE FOLLOWING EQUIPMENT:

BELLOWS EVACUATING COMPRESSOR'(PAS-C-200)

WASTE TALK (PAS-T-202)

WASTE TRANSFER PUMP (PAS-P-203)

Verification that all. instrumentation-and controls, including all electrical interlocks, function properly..

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2688.1 FIR 1 PUMP AIR RELEASE VALVE MO_D_IFICATION

- The purpose of this procedure is to per' form modification

~

- of the sir release valves at the motor' driven and diesel driven fire pumps.

This. completed modification' procedure was reviewed by'the PORC committee and no unreviewed safety questions, technical; specificati'on changes or violations were~. involved.with this change to'the facility.

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 34 of 55 SM-2 82 8_. 2_ VENT HEADER PIPING _ MODIFICATIO_N The purpose of this procedure is for the work associated with the shielding and support modification of the vent header piping.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2929.11 INSTALLATION AND TESTING OF THE 1B CONTROL ROD SHROUD FAN AUX. BREAKER The purpose of this procedure is for the installation and testing of the 1A control rod shroud fan auxiliary breaker.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2929.12 TERMINATION AND TESTING OF AUX. BREAKER FOR LIGHTING TRANSF_ORMER_lD The purpose of this procedure is for the installation of an auxiliary breaker for the ID lighting transformer.

This completed modification procedure was reviewed by the i

PORC committee and no unreviewed safety questions, technical specification-' changes or violations were involved with this change to the-facility.

SM-2929.5 PRE-INSTALLATION TESTING OF THE HFB BREAKERS

-IN THE AUX. BREAKER CABINETS The purpose of this procedure is_to test for overcurrent tripping on each phase of the HFB breakers. An insulation resistance test shall be done on the breakers and auxiliary switch wiring.

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1983 Annual Repor!. of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 35 of 55 This completed modification pr 'cedure was reviewed by the PORC committee and no unreviewed .;afety questions, technical specification changes or violations were involved with this change to the facility.

SM-2929.6 PRE-INSTALLATION TESTING OF DB-25 AIR CIRCU_I_T_

BREAKER The_ purpose of this procedure is to perform overcurrent trip tests on the DB-25 breakers prior to installation and to determine if breakers operate properly, i

This completed modification procedure was reviewed-by the PORC committee and no unreviewed safety questions, technical specification changes or- violations -were involved with this change to the facility.

SM-2929.7 INSTALLATION OF AUX. BREAKER CABINETS, CONDUITS, UNISTRUT_S_TRU_CTURE & CABLES The purpose of this procedure is to install auxiliary breaker cabinetsi conduits, and unistrut structure.- Cables will' be pulled f rom . the auxiliary breaker cabinets to the motor -

control' centers. The cabinets will then be spliced in at the motor control. centers.

This completed modification' procedure was reviewed by'the PORC committee and no.unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-2929.8 TERMINATIONS AND SPLICING OF. CABLES FOR - AUXILIARY BREAKER CABINET MCC 1D.

The purpose-of this procedure is for theLinstallation of.

. auxiliary breakers on selected' circuits-in MCC 1C. 1 This . completed modification procedure _ was reviewed by the .

PORC committee-and no unreviewed safety questions, technical specification changes'or--violations were involved with.

this. change.to the_ facility.

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 36 of 55 SM-2929.9_ TERMINATIONS AND SPLICING OF CAB _LES FOR AUXILIARY BREAKER CABINET MCC 1C The purpose of this procedure is for the installation of auxiliary breakers on selected circuits in MCC 1D.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3037.10 LOCAL TESTING OF THE BORIC ACID EVAPORATOR AND GAS STRIPPER _ PANELS _

The-purpose of this procedure is to provide the for testing the boric acid evaporator and gas stripper panels in the

" local" control. This procedure will also provide the necessary guidelines for testing the five solenoid valves associated with the gas stripper.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were -involved with this change to the facility.

i SM-3037.13 LOCAL TESTING OF THE' WASTE EVA_PORATOR PANEL The purpose of this procedure is.for testing the waste evaporator panel for operability in " local" control after terminations have been made to the radwaste. process computer.

This completed modification procedure was.. reviewed by the porc committee _and no unreviewed safety questions, technical-specification-changes or' violations were involved with this change to the facility.

SM-3037.16, TESTING OF THE REMOTE INDICATION PROVIDED BY THE RADWASTE P_ROCESS COMPUTER The purpose of this procedure is .to ensure that the remote indication provided by the radwaste process computer is within-acceptable tolerance withithe local indication of the waste disposal system panel.

~ . . _ _ . _ - . - _

4 i

l 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval

> R. E. Ginna Nuclear Power Plant Unit ~No. 1 Docket No. 50-244 Page 37 of 55

~ This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3037.2_ CABLE TERMINATION AND ENERGIZING OF THE FOX _3_

COMPUTER _

The purpose of this procedure is for the termination o'f

cables from the operating terminal in the auxiliary building and the terminals in the TSC to the Fox 3 computer.

This completed modification procedure was' reviewed by the

! PORC committee and no unreviewed safety questions, technical specification changes or violations were' involved with this change to the facility.

SM-3037.4 REMOVAL OF_ TC 205D AND ONE HALF INCH CONDENSATE i LINE TO DESUPERHEATER FOR INSTALLATION CONDUIT The purpose of this procedure is for removal of TC-205D and the on-half inch condensate return line to the desuper-heater.

This completed modification procedure was reviewed by the-PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3 037. 5_A_ INSTALLATION OF PNEUMATIC TUBING TO THE

SOLENOID VALVES OF THE RADWASTE PRO _ CESS COMPUTER

< The purpose of this procedure' is to' install pneumatic tubing -

j; to the solenoid -valves for the radwaste processjcomputer.- ,

(' - This completed modification procedure was reviewed by the '

PORC committee _and no'unreviewed safety questions, technical _- '

specification changes or~ violations were. involved:with this change to the facility.

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t 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted

Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244-Page 38 of 55

SM-3037.6 FINAL TERMINATIONS _FROM THE ULTRAFILTRATION UNIT TO THE RADWASTE_P_ROCESS C_O_MPUTER_

The purpose of this procedure is for cable terminations 4

frm the ultrafiltration unit to the radwaste process computer.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

5 SM-3037.7 LOCAL TESTING _OF THE ULTRAFILTRATION UNIT The purpose of this procedure is for testing the ultrafiltration unit for operability in " local" control after terminations have been made to the radwaste process computer.

This completed modification procedure was reviewed by the i PORC committee and no unreviewed safety questions,' technical specification changes or violations were involved with I

this change to the facility.

SM-3037.8 CABLE TERMINATIONS FOR THE PNEUMATIC INTERF_ ACE' RACK OF THE RADWASTE PROCESS COMPUTER-

) '

The purpose of this procedure'is for terminating cables from the pnetanatic interface rack to the universal input / output device of the radwaste process computer.

4 This completed modification procedure was reviewed.by the PORC committee and no unreviewed safety questions,_ technical specification changes or violations were involved with:

this change to the facility.

SM-3037.9 MODIFICATIONS AND FINAL TERMINATIONS TO THE GAS STRIPPER AND BORIC ACID EVAL' ORATOR PANELS FOR THE RADWASTE PROCESS COMP _ UTER t

The purpose of this procedure is for modifying and performing final terminations to the gas stripper and boric' acid evaporator panels for the radwaste process computer. Final. terminations to the five. solenoid valves associated.with the gas stripper and boric acid evaporator will be performed under .this

, procedure.

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted

.Without. Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 39 of 55 This complet'ed modification procedure was reviewed by the PORC committee and_no unreviewed safety questions, technical specification changes or violations were involved with~

this change to the facility.

SM-3097.1 SERVICE BUILDING ROOM INAC MODIF_I_ CATION 4

The purpose of this procedure is for the installation of a complete exhaust system and additional cooling for_the.

service building laundry room.

l,s This completed modification procedure was reviewed by the

PORC committee and no'unreviewed safety questions,. technical-

_ specification. changes or violations were involved with

this change to the facility.

SM-3100.'5 INSTALLATION OF_T_HE ELECTRIC _AL WIRING AND

' CONDUIT NECESSARY FOR THE MSR P_RESEPARATO_R MODI- ,

FICATION i

! The purpose-of this procedure is for the installation of-the wiring and conduit required for the-MSR preseparatcr modification.

4  ;

i This completed modification ~ pro'cedure was reviewed by the

{ PORC committee and no unreviewed-safety questions,_ technical specification changes or violations were involved with.

this change to the. facility.

I i

SM-3100.7 MOISTURE PRESEPARATOR MODIFICATION TEST PROCEDURE

'The purpose of.this procedure is to-verify'that the new' high pressure' turbine exhaust preseparator drainfsystem >

design incorporated -through this modification functions correctly, in accordance with original-design requirements.

. This . completed mo'dification procedure was reviewed by -the

PORC committee and'no unreviewed safety questions,~ technical

specification changes or: violations were-involved;with ,

this change to the facility.

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted i Without-Prior Approval '

R. E. Ginna Nuclear Power Plant Unit No.-1 Docket No. 50-244 l Page 40 of 55 SM-3107A INSTALLATION OF NOZZLE DAM INSERTS IN STEAM i GENERATOR CHANNEL HEAD NOZZLES

. The purpose of this procedure is to provide information relative to the' installation of nozzle dam inserts in steam generator channel head nozzles.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification _ changes or violations were involved with

this change to the facility.

j SM-3259.2 ELECTRICAL INSTALLATION OF THE CV-TRANSMITTER

! RELOCATION MODIFICATION i The purpose of this procedure-is for the performance of-

[ the electrical work associated with,the CV transmitter relocation modification. The committee reviewed this new procedure and recommended. approval.

This completed modification procedure was reviewed by the PORC committee and no unreviewed_ safety questions, technical specification changes or violations -were -involved with this change to the facility.

i l

SM-3259.4 REMOVAL OF PRESSURIZER TRANSMITTER CABINETS _-

'AND RELOCATION OF DP TRANSMITTER 432A Th'e purpose of this procedure is to relocate differentia 1' pressure transmitter 432A, to,a new11ocation and the removal'

of-the pressurizer; transmitter cabinets'.

-This completed modification- procedure wasi reviewed by the '

4 PORC committee and no unreviewed safety-questions,1technicali specifi' cation changes or violations were involved with this change to the. facility.

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SM-3260.1

SOLENOID2VALVE REPLACEMENT: ' MECHANICAL

! The purpose-of this procedure is i for the. work associated

~

t with the solenoid-valve replacement modification..

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1 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 41 of 55 This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3260.2_ SOLENOID __ VALVE REPLACEMENT: EL_ECTRICAL The purpose of this procedure is for electrical work associated with the solenoid valve replacement modification.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3260.3 MODIFICATION OF PIPING AND SUPPORTS TO SAMPLING VALVE #955 The purpose of this procedure is for the necessary work required to modify piping and supports to sampling valve

1. 955 (loop B hot leg sample line control valve) .

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3319.1 CABLE INSTALLATION AND TERMINATION FOR THERMAL OVE_R_ LOAD RE_ LAY MOD.

The purpose of this procedure is to install cable, conduit, and supports for thermal overload relay modification.

Terminations will take place at MCClC, MCCID, MCC1A Aux. Panel, Aux. Building, DC Dist. Panel 1A, SI-Al rack, and the main control board.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

L_ W

y 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 42 of 55 SM-3435.1 1C SAFETY INJECTION PUMP CONT _ROL SCHEME MODIFICATI_O_N_

The purpose of this procedure is for modifying the 1C safety injection pump control scheme so as to improve pump avail-ability.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

_SM-3447.1 ELECTRIC _AL_INSTALLATI_ON OF_ "A" AND "B" ST_EAM_ ,

GENERATOR METAL IMPACT MONITOR MOD.

The purpose of this procedure is for the performance.of electrical work associated with the steam generator metal impact modification.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3447.3 STEAM _ GENERATOR - MOD. OF INS _ULATIO_N_ _ SUPPORT RINGS _F_OR I_N_STALLATION OF LOOSE PART MONITORS

! The purpose of this procedure is for modification of the steam generator -insulation support rings to allow installation of the sensors for the loose part monitors.

j This completed modification procedure-was reviewed by the PORC committee and no unreviewed safety questions,' technical specification changes or violations were -involved with this change to the facility.

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r 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 43 of 55 SM-3447.5 REPLACEMENT OF METAL IMPACT MONITOR SENSORS _

ON THE "A" AND "_B" ST_EAM GENERATOR The purpose of this procedure is for the replacement of the metal impact sensors up to the amplifiers along with the associated cable.

This completed modification procedure was reviewed by ' the PORC committee and_no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3447.5A MIMS: REPLACEMENT OF ACCELE_ROMETERS &'

HARDLI_N_E_ _C_ ABLES TEST _ PROCEDURE The purpose of this procedure is to verify that the replacement accelerometers installed.per this modification function properly in accordance with original design requirements.

This completed modification procedure was reviewed.by the PORC committee and no unreviewed safety-questions,' technical specification changes or. violations were. involved with this change to the facility.

SM-3571.1_ CCWST LEVEL _ INDICATION MODIFICATION The purpose of this-procedure is to 16 stall'a second means of indicating abnormal level in 'the component cooling water .

surge. tank.

This completed modification procedure was = reviewed by Lthe PORC committee and no unreviewed' safety-questions,_ technical specification changes or violations were involved with.

this change to the facility.

~ SM-3 572. l_-  : SODIUM HYDROXIDE TANK __S_UPPORT MODIFICATION The purpose of this procedure;is to - modify and -_ upg rade ._

. the sodium hydroxide tank' saddle _ supports including;the-l attachment'of one saddle support to.the' tank.

^

_ . _ _ _ . _ _ _ _ . _ _ _ . _ . _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ _ . _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ ____._._.____c _ _ _ _ _ _ _ _ _ _ _ _ _ _ _._._____.__m.__.1____._. _ _ _ _ _ .

J I

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted.-

Without Prior-Approval R. E. Ginna Nuclear Power Plant i Unit No. 1

Docket No. 50-244 o 3

Page 44 of 55 This completed modification procedure was reviewed by the

! PORC committee and no unreviewed safety questions, technical specification changes _or violations were involved with this change to the facility.

I SM-3580.1 SPRINKLER SYSTEM TIE-IN CONNECTION FOR-THE.

j S/G MOCKUP BUILDING The purpose of this . procedure is _ for .the tie-in connection of the S/G mockup building sprinkler system to_ the existing fire main. .

This completed modification procedure was reviewed by'the f PORC committee and no unrevi'ewed safety questions,~ technical j specification' changes or violatio'ns were involved with ,

this change to the facility.

l SM-3659.1 INSTALLATION OF SI' SIGNAL BYPASSES IN THE i RWST/ BAST SWITCH LOGICS i .

, The purpose of this procedure is.to perform the installation.

of the SI signal. bypasses in the rwat/ bast switching ' logic.

This completed modific'ation procedure was' reviewed by the-PORC committee-and no unreviewed safety questions,-.techni' cal'

specification changes or violations were-involved with.

l - this change to' the facility.

SM-3659.2 BORIC ACID TANK /RWST'SWITCHOVER TEST PROCEDURE:

~

. The purpose of this procedure . is; to ' verify that the modifi -

cations to the boric acid tank /rwat switchoverllogic function i correctly, in accordance.with.the design criteria.

This completed modification procedure was reviewed by. the

.PORC committee and-no unreviewed' safety questions',Jtechnical.

I '

specification ' changes .'or violations ;were involved with

- this change.to_the facility.s '

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f 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 45 of 55 SM-3784.1- TURBINE LIFFERENTIAL EXPANSION MODULE The purpose of this procedure is for the cutting of the main control board and the installation of the generator-differential expansion module.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions, technical specification changes or violations were involved with this change to the facility.

SM-3794.1 A& B DIESEL START SWITCH ALARM The purpose of this procedure is to install a circuit to annunciate the requirement for a diesel reset to the plant operators.in the control room and in each diesel generator room.

This completed modification procedure was reviewed by the PORC committee and no unreviewed safety questions,. technical specification changes or violations were involved with this change to the facility.

SM-3912.1 INSTALLATION OF THE FEEDWATER RTD'S The purpose of this procedure is for the installation of conduit and wiring for the feedwater RTD's.

~

This completed modification procedure was reviewed by the-

-PORC committee and no unreviewed safety questions,. technical specifi. cation changes or violations were: involved with this change to the facility.

-- -= :j

1983 Ann'ual Report of Facility Changes,

' Tests, and Experiments Conducted-

.Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No.: 1 Docket No. 50-244 Page 46 of 55 SPECIAL TESTS COMPLETED IN 1983 ST-81.1 DRUMMING OF WASTE EVAPORATOR BOTTOMS AND MISCELLANEOUS WASTE This analysis covers the'special test #ST 81-1 for the NUMANCO Drumming Unit...

Radio' active Liquid Waste System Leak The consequences.of this. event are not; increased by the' special' test because the new system will interface.with the liquid waste system in similar fashion'to the-~ existing 4 system. All' liquid waste-carrying components will withstand _-

maximum internal pressure and will_be secured to the-NUMANCO drum unit (anchored) and the liquid waste disposal system.

+

-They'will not be routed in proximity to safeguard-equipment-.

Fires i

~

j The proposed special t'est neither penetrat,es any' existing fire barriers nor does it affect any existing fire suppression ~

{ system. _ The . special test 'does not . increase any previously i determined. fire loadings.

- Flood.or. Storms '

4 The special test'neither affects nor is'affected by any

. flood or storm previously evaluated.

Earthquake The consequencesLof this event' fare not-increased by this _

' modification because the modification will be designed-

~

to withstand.a seismic-event.as defined'in'the Ginna Station; FSAR, using thelequivalent' static load _ method.

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1983 Annual Report of Facility Changes,

' Tests, and. Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit-No. 1 Docket No.-50-244 Page 47 of 55 Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the plant.have not been. reduced. The adequacy of. structures systems,.and components provided for the prevention of accidents have not been affected.

- ST-1444-82.2 BUS 18 UNDERVOLTAGE CABINET SYSTEM ENERGIZATION TEST This analysis covers the special tests for the connection of AC and DC sources to undervoltage cabinets.- This-will allow -operation of .the equipment prior -to connection .of the plant circuits to the undervoltage' tripping relays.

The undervoltage system is in the preoperational.-stage and no tripping relay has been connected tolthe plant system at this time. The AC and DC-power feeds to the cabinets

• are however connected and are within' :the scope . of - this-test.

Since proper fuse and breaker sizing and-cookdin'ation protection .

has been established,-faults and short circuits, due to workman error or s in improper installation, will not affect.

availability of AC or DC system.-

It has been determined, by this analysis, that: the margin

~

of' safety during normal l operation and transient conditions-during the test period have not-been affected.-

ST-1444-82.4 BUS 14 UNDERVOLTAGE CABINET - SYSTEM ENERGIZATION TEST < s

-.This analysis covers the specia1 tests ; forz the connection of AC and DC sources to ' undervoltage-' cabinets. This-will' l allow : operation of. the equipment - prior .to connectionof the_ plant 1 circuits.to the undervoltage-tripping relays.

~

The undervoltage system-is in'the preoperational stage?

and no tripping relay: has been. connected to the plant : system at this time. TheLAC and DC power feeds to the cabinets lare however connected and-are within the. scope of.this

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1983 Annual Report of Facility Changes,

' Tests, and-Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 48 of 55 Since proper fuse and breaker sizing and coordination protection has been established, faults and short circuits, due-to workman error or in improper installation, will not affect availability of AC'or DC system.

It has been determinedi by this analysis,-that the margin of safety during normal . operation and _ transient conditions during the test p'eriod have not been affected.

S T- 14 4 4 - 8 2_. _5, . BUS 16 UNDERVOLTAGE CABINET ~ SYSTEM ENERGIZATION TEST This analysis covers the special tests for the connection of AC and DC sources ' to undervoltage cabinets. ThisLwill allow operation of the equipment prior to connection of '

the plant circuits to-the undervoltage-tripping relays.

' The undervolt' age system 'is in-the preoperational' stage and no tripping relay has-been connected tocthe plant system at this time. -The AC and DC power feeds _to the cabinets '

are however connected and are within the cscope-_of-this-test.-

Since proper, fuse and breaker sizing and coordination protection has been established, faults and short circuits ,' . due :Lto workman error -or in -improper installation,- will notL affect availability of AC or-DC: system.

It - has been determined,- by this analysis, -that. the margin

. of safety during normal ' operation and transient ; conditions .

- during the: test period _have not-been'affected.

ST-3575-83.01- LOW-AMPLITUDE TESTING'OF'THE MAIN CONTROL BOARD

The: purpose of the proposed Special Test'Eis toidevelop- ' ~

the dynamic response l characteristics of the- three sections of the Main Control 1 Board at Ginna Station. . The' test;will be performed by URS/Blume and Associates by inducingi very'

- small:'vibrationsiat,known locations and. collecting tho' control board'sfresponse_to these vibrations at various-i grid locations. Accelerometers and'a' recorder will be; used to collect.and-store the' data._- The response data i

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J 1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval

! R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 49 of 55

~

will then be analyzed by Blume and the stresses due to a postulated. Safe Shutdown Earthquake (SSE) will be calculated r in the load bearing members of the MCB.

The Special Test is expected to last for approximately 80 - 100 days. Therefore, . the test equipment is considered temporary and none of the events analyzed in the Ginna FSAR need be evaluated due to this test. However,..a number of precautions will' be followed to prevent spurious operations of any system due to the-testing.

3 The precautions that will be taken during' the test are-outlined in the Special Test Criteria and are summarized

!= below:

a. The test equipment willibe arranged in the Control F Room so as to minimize any interferences with ' normal operation.

i b. -The accelerometers locations will-be selected so as l to minimize interferences with normal. operation..

j, In addition a cognizant individual will-be assigned

to assist in_ determining these locations. '

i

c. PORC will review.the URS/Blume test procedures.

[ It has, by thisianalysis, been de'. ermined thatiwith the-I above precautions the margins of safety during normali operations.

and transient conditions 'during the : test period. have noti f been affected.

ST-83-01 "A" AND "B" STEAM GENERATOR CHANNEL HEAD, i DILUTE CHEMICAL DECONTAMINATION This analysis' covers the special -test ST-83 -Ol' . Rev.10.

', for A-& B steam generator channel: head ' dilute chemical:

decontamination.

Calculat' ions reveal:that-if the RCS is-borated to 2460 ppm ' prior to : adding the . dilute -chemicair solution Lvolume

. equal to 2500 gallons, the Technical Specification on Contain-1 ment Integrity that s tates -. " Positive 1 reactivity . cha ng e s sha11 notLbe made by rod drive motion or boron dilution 1

'whenever - the containment integrity is- not' intact unless

& _ . the boron . concentration -is greater than 2000. ppm",- is assured.

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I 1983 Annual Report of Facility Changes, Tests, and Experiments' Conducted Without Prior Approval.

R. E. Ginna Nuclear Power Plant Unit No. 1 i Docket No. 50-244-Page 50 of 55 This event is analyzed for the potential to overpressurize the reactor coolant system. -The operability of the pressurizer PORV's or an RCS vent opening of greater than 1.1 square

inches ensures that the RCS will be protected :from pressure

, transients which could exceed the limits of Appendix G

to 10CFR Part 50 when one.or.more of the RCS cold legs

! are < 330 degrees F. Since the'special test will be done

~

during refueling shutdown this requirement can be satisfied in-all cases.

The doses which could be attributed to an accidental spill of resin would be-insignificant when compared to the 10CFR part 20 and 100 dose limits.

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Since this special test:is a temporary one conducted during j plant refueling shutdown, .no further evaluation is. required.

, Corrosion data studies on material compatibility of corrosion mechanisms like intergranular. attack on.304 densitized i stainless ~ steel was very slight :and -not ' considered to be l a cause for concern.

3 Corrosion data studies on other reactor materials . including -

Inconel, and Zircaloy were found to be acceptably low.1 i

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In addition, this process wil14 require isolation-at the S/G nozzles. Whatever means of isolation--is chosen the.

following characteristics of the reagents are:

i j 1) They are dilute, mainly organic. compounds.

2)

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They.are quickly decomposed a reactor operating temper =

atures.to innocuos compounds; carbon dioxide,. nitrogen, ammonia, water, potassium, oxygen and manganese.

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3) .They are very susceptible. to radiolytic decompos'ition

.to the same. innocuous, volatile compounds.

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4) :Those that are =not organic compounds =are applied in; a very. dilute solution, and>have been used in much'

more concentrated solutions with no deleterious effects. -

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1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 51 of 55

5) Control of the solvent . concentrations will be closely monitored by the vendor as well as our own laboratory analysis. The concentrations are controlled by small additions of reagents or removal by ion exchange equipment which may be rapidly placed in and out of service.

This system will allow close control of the process chemical concentrations.

Thus, even if isolation failed, and a quantity of reagent entered the primary circuit, it would be further diluted, and the constitutes would be thermally and radiolytically degraded to innocuous volatile compounds.

In addition this process has been used to decontaminate total systems with zircaloy clad fuel in the vessel since 1973 with no adverse affects.

Therefore, the margins of safety during normal operation and transient conditions anticipated during the life of the plant will nor be reduced. The adequacy of structures system and ccuponents provided for the prevention of accidents and for the mitigation of the consequences of accidents will be unchanged by 'the performance of this special test.

ST-83.02.1 HOT LEG TEMPERATURE STREAMING DATA COLLECTION PROCEDURE In order to' determine the reactor coolant flow measurement uncertainty associated with hot leg-temperature streaming, a test will be conducted during Cycle 13. Twelve thermocouples will be strapped around the. circumference of each hot leg adjacent to the RTD's. Measurements of the relative- temperature of the exterior of ' the pipe at zero power and at various.

levels and their- comparison to the -RTD measurements will provide an indication of the temperature difference between the RTD and average temperature of the' fluid..

The thermocouple instrumentation will not provide input to any control or protection system and will- be used only as a temporary monitoring system. The cable required and:

the existing (spare) containment' penetration used will not replace or alter any control or protection system func-tions.

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1983 Annual, Report of Facility Changes, )

Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 '

Page 52 of 55 RTD's 409A and 410A provide input to the R3 subcooling margin monitor and will have to be disconnected for the seasurement to_be taken. Only one RTD will be disconnected at any. time so that one. channel will remain operable.

The RTD will be reconnected after the specific measurement is taken.

Temporary cable and wire installed for the test _will comply with IEEE-383 flame test requirements, therefore there will be no increase in_ fire-loading.

The total -weight of the assembly on each loop will be less than 10 pounds. . This is negligible compared to the weight.

of the pipe and would not effect the integrity of the RCS for a seismic event. The assembly itself is not required to function during or after a seismic event.- Because of its low distributed mass it does not pose a threat as~a missile to surrounding components. The reference junctions will be installed in an instrument rack already existing inside containment.

The thermocouples, mounting brackets and strap ' re a fabricated .

from Type 304, 308 and 316 stainless steel, tharefore there-is no incompatibility of materials with the primary' system :

piping. Insulation removed to install the TC's will be replaced once installation is completed to minimize heat loss and any potential thermal-stresses between insulated and non insulated pipe.

Therefore, the margins of safety during_ normal operation and transient conditions. anticipated during the - lif e of-the plant will not be reduced. The adequacy of structures, systems _,-'and components provided for the prevention of accidents.and for'the mitigation of the consequences'of accidents will -be unchanged by .the - per formance of this special test.

l' ST-83-03 LITHIUM - ADDITION TEST PROGRAM REACTOR COOLANT SYSTEM This . analysis - will cover .the possible saf'ety concerns .of-

performing,an'EPRI/B

& W test associated-with the:EPRI

( -Radiation Control Program,~RP-825-01..

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1983 Annual Report of. Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Pla'nt Unit No. 1 Docket No. 50-244 Page 53 of 55 The test is designed to determine the optimum lioh concentration at various RCS boron concentrations for control of activated corrosion ~ products. At three RCS boron concentrations (550-700 ppm 300-400 ppm, and 100-200 ppm'b), the lithium.

concentration will be reduced to 0.2 ppm Li then increased in increments up t0 2.0 ppm Li. Samples will be taken after each incremental change to determine the effect of these changes on corrosion product solubilities.

The mixed-be'd demineralizers and-reactor coolant filter will be taken out of service for all three lithium- addition tests. The function of the mixed bed demineralizers and the RCS filter is to maintain reactor _ coolant purity.

The CVCS provides the required seal-. water flow for the reactor coolant ptmtp shaft seals. The two seal water injection filters collect particulate larger than five micron from the water supplied to the reactor coolant pump-seal.

When LiOH' concentrations are changed in the reactor coolant ~

system, it could possibly cause a. crud burst which could plug the seal water injection. filters causing a reduced.

reactor coolant water flow to the reactor coolant' pump seals and possible' loss of both reactor. coolant: pumps.

This accident has been analyzed for in section 14.'l.5 of the FSAR.

Ef-crud burst occurs, the seal water flow reduction would not be instantaneous,. so the operators _should: have time to place.a-mixed bed ~DI and the reactor'coolan- filter in service and then valve in 'the ' alternate seal injection -

filter. .

a A similar test was performed 1at the Rancho Seco Plant in: <

1977 and 1978. No large crud bursts were experienced so l the likelihood of this problem occurring is-.very.small. .)

The limit . for total suspended solids in the reactor coolant. l system:is 1.0 ppm. It is possible that for'a shert perioG ~ .i of timer during one or. all' these tests, this limit could- 1

-be exceeded. An-increase of total' suspended. solids could I cause . plugging of the seal water injection ' filters as stated

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1983 Annual Report of Facility changes, Tests, and Experiments.Condueled Without' Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 54 of 55 Therefore, the margins of safety during normal operation and transient conditions anticipated during the life.of the plant will not be reduced. The adequacy of structures, systems, and components provided for the prevention of accidents and for the mitigation of the consequences of accidents will be unchanged by the performance of this special test.

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1983' Annual Report of Facility Changes, Tests, and Experiments Conducted

'Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 55 of 55 CHALLENG_ES _'aD PRESSURIZER PORV'S AND SAFETY VALVES June 9, 1983 Both Pressurizer PORV's lifted-for approximately 1 minute.

System status was cold shutdown, overpressure protection system in service, performing Safety Injection System test RSSP-2.1. The charging pump failed to stop. Highest pressure indication 375 psig.

June.19, 1983

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< Pressurizer PORV 430 cycled open and shut at 2335 psig.

l. upon failure of pressure controller.431K. Reactor-ai. 15%

t power.

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ROGER W. KODER VICE PRESIDENT TELEPHONE tutCTRIC 6 STEAM $40 DUCTION A er E A C O CiE m 546:2700 May 21, 1984 Dr. Thomas E. Murley, Regional Administrator U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 l

Subject:

1983 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant, Unit No. 1 Docket No. 50-244

Dear Dr. Murley:

Transmitted herewith is the submittal of the Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval as required by 10 CFR 50.59 and Challenges to the Primary PORV 's and Safety Valves. This report is for the period of January 1, 1983 through December 31, 1983 inclusive.

I Vr Truly Yours, l u,__4A>r Ro W. Kober l

Attachment I 1

xc: Document Control Desk (1) s 4.e