ML20039E069

Forwards Annual Rept of Facility Changes,Tests & Experiments Conducted W/O Prior Approval for 1980

ML20039E069
Person / Time
Site:	Ginna
Issue date:	12/28/1981
From:	Maier J ROCHESTER GAS & ELECTRIC CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 8201060492
Download: ML20039E069 (34)

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• 89 EAST AVENUE, ROCHESTER, N.Y.14649 JOHN E.

MAIER vnsmo c vitt prest 0LNT apt a coot Tia 546 2700 December 2R, imi Mr. Ronald C. Haynes U. S. Nuclear Berulatory Cornission Office of Inspection and Fnforcement Region I 631 Park Avenue King of Prussia, Pennsylvania 19406

Subject:

Annual Report of Facility Changs, Tests, and Expericents Conducted Without Prior Approval.

R. E. Cinna Nuclear Power Plant, Unit No.1 Docket No. 50-244

Dear Mr. Haynes:

Transmitted herewith is the resubmittal of the Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval as required by '10 CPR 50 59 This report is for the period of January 1,1TO through December 31, 1980 inclusive.

Very truly yours,

<--/3+7 NM John E. Maier JDi/Imt Enc.

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- U.S. Muclear Regulatory Comission M,,

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80-45

'ISR 80-04, Chemical Addition to' Auxiliary Feedwater Bypass.. T.

Meyer presented _ Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this mdification.

We design is for the modification of the "A" CVCS Holdup Tank Drain Line. It consists of adding an isolation valve on the drain line downstream of the tee for valve 1071.

he above nentioned nodification is to provide for transferring reactor cavity leakge which has accumulated in Sump "A" to the "A" CVCS Holdup Tank for reprocessirg.

The additional isolation valve is in the "A" CVCS HUr drain line to the sump tank, between the existing elbow aM the floor. A review has been made of all events analyzed in the Ginaa Station FSAR aM NRC Regulatory Guide 1.70.

None of these events will be affected by this modification.

W erefore, the margins of safety _ durire normal operations and transient mMitions anticipated durire the life of the plant have not been reduced.

We Mequacy of structures, systens and couponents provided for the prevention of accidents and for the mitigatien of the consequences of accidents have mt been affected.

We Committee performed a Safety Evaluation aM determined there were to unreviewed safety questions or Technical Specification changes required.

%is item IS LOMPLETE.80-389 EWR-2447, Reactor Vessel Head Vent System.

T.

Meyer presented Revision 0 of the Safety Analysis and Revision 0 to the Design Criteria for this nodification.

As part of the requirements of NUREG 0578, a reactor vessel hea3 vent system shall be designed aM installed prior to January 1, 1981.

We purpose of this system is to provide a neans for rapid ventire of the reactor vessel head for postulated incidents of degraded emergency core c]oling or impaired natural circulation.

We nececsary mechanical nodifications to the reactor coolant system are shown. in schematic form on Westirghouse Dwg. 2654C03.

Catalytic Dwg.

B-15604 provides a basic one line electrical drawing for the' reacto-vessel head vent isolation valves.

We primary function of the Reactor Vessel Hea3 Venting System (RVHVS) is to vent. non-condensable gases or steam frcm the reactor vessel to ensure adequate emergency core coolire capability.

The system functions to vent the reactor vessel hea3 at any system pressure greater than containment pressure.

We RVHVS may be operated in conjunction with other methods of degassire the reactor vessel head such as pressurizer spray or letdown /degassirg.

We reactor vessel head vent system is shown on Engineerirg Flow Diagran 2654C03.

%e system consists of redundant flow paths connected to the 3/4 inch reactor vessel' vent pipe downstream of manual valve 500.

Each flow path shall consist of two solenoid operated glove

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valves in series powered fran vital busses. We valves in each flow path shall.be pwered by the. same. vital bus and shall be fail closed, active valves.

We arrangement of failure nodes and-electrical train assignments allows the system to meet the sirgle active failure criteria for both initiating and terminating reactor head venting.

We reactor vessel head vent system'shall discharge locally to the refueling cavity, a well ventilated area of the containment capable of acccmiodating :the discharge of -gases or water.

'Ihe RVHVS 'shall be designed to preclude the initiation of safety injection.if a break occurs downstream of the interconnection -

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with the 3/4 inch normal vent pipe.

Ead remotely operated ' flow path shall have a flow restrictor to limit the blowdown from a break.

3c downstream of the restrictor to within the dargirg capacity of-two t

multispeed positive displacement charging pumps.

The flow restrictor is the ANSI Safety Class 1 to 2 boundary.

We existirg.

reactor vessel vent pipe is-3/4" Sch. 80S and terminates _ at a blind flange.

We renotely operated vent piping shall be austenitic

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stainless steel 3/4" Sch. 160S.

We existirg '. vent pipirg shall-terminate with a blind flarge.

We RVHVS isolation valves -shall meet the following design / performance requirements:-

(a). Fluid -

reactor a3olant, (b) Design pressure rating - 2500 psig, (c) Design temperature - 680*F, (d) Solenoid operated, active valves,.(e)

Section III, 1974, Class 2, and (f) Radiation Qualified to 1x10 rads pst accident.

Wis safety evaluation has been performed, consistent with the requirements of 10CFR50.59, to ensure that the plant nodifications required to incorporate a reactor vessel, head vent systerr. do not involve an unreviewed safety question..

he existing reactor vessel head vent pipe is used only for shutdown -

venting of the~ reactor vessel.

It functions as a reactor coolant pressure boundary during plant operation arr3 performs no accident mitigation function.

The reactor vessel. head vent system modifications add piping and rerotely operated valves to the existing head vent.

We reactor vessel head vent system functions as a reactor coolant pressure boundary durirg plant operation.

We modified head vent system will be designed to perform an accedent mitigation

function, which is to discharge accumulated noncondensable gases or steam fran the reactor vessel for postulated i

incidents of degraded emergency core cooling or impaired ' natural circulation.- We design criteria for the reactor vessel hea3 vent system has been reviewed and found.to satisfy the criteria applicable to safety-related systens - (including safety class criteria, ASME and ANSI (bde requirements, seismic design criteria, sirgle failure criteria, and pst-accident operability criteria).

The accident analyses in the Ginna Station Final Facility Description arrl Safety Analysis Report have been reviewed with

PAGE 3 PORC ITEM #

respect to the reactor vessel hem vent system audification.

In conjunction with the Ginna Station accident analyses, the following sections of the design criteria were specifically considerai:

Section 1.3 Performance Requirements, Section 21.0 Failure Effects Requirements, and Section 26.0 Fire Protection Requirements.

None of the accident analyses will be deleteriously affected by the addition of the RVHVS design has included orifices so that ruptures in the added pipes will result in inventory loss rates smaller than that which can be provided by two diarging pumps.

Sus, the likelihood of a loss of c3olant accident is not increased.

Consideration has been given in the cbsign to use of the vent s; stem.

S e. vents have been orificed to that relief of pure hydrogen for ten minutes will not result in a hom)geneous hydrogen concentration greater than 4%.

Operator action will be required to initiate and terminate venting aM to determine containnent hydrajen concentration (redundant hydrogen nonitors are to be installed under a separate EWR).

Controlled, intennittent venting can take place using release periods which are less than ten minutes so that containment hydrogen c>ncentrations can be nonitored.

Procedures which define the conditions under which the vents should be used, as well as the conditions under which the vents should not be used, will be developed.

We design daaracteristics of the vent systen and appropriate operating procedures will enhance the core cooling an3 cantainment integrity safety functions.

Installation of the RVHVS will orovide greater operational flexiility for resp)Minj to LOCA e)Mitions.

We Committee performed a Safety Waluation an3 determined there were n) unreviewed safety questions or Technical Specification chan3es required.

Eis iten IS COMPLETE.80-452 EWR-2605, Diverse Containment Isolation.

T.

Meyer preseated Revision 0 of tne Safety Analysis aM Revision 1 to the Design Criteria for this nodification.

A letter fron Harold Denton of the NRC, dated October 30, 1979, requires in Section 2.1.4 Position 4 that:

We design of control systems for automatic containment isolation valves shall be such that resettin3 the isolation signal will not result in the automatic reopening of containment isolation valves.

Reopening of containment isolation valves shall require deliberate operator action.

AM that:

Resetting of c)ntainnent isolation signals shall not result in the automatic loss of containment isolation.

At Ginna Station certain valves reopen upon reset of the containment isolation or containment ventilation isolation if their controllers are set in the open p3sition.

'Ib further reduce the likelihood of inMvertent reopening of valves, a

PAGE 4 IORC

.IEi system nodification will be designed to provide for individual resetting of all isolation valves to eliminate any pssibility of an inadvertent opening.

We nodification will provide an addit.ional reset requirement, for each irdividual valve, in the followirg way:

(refer to sketch 1).

Upon receiving a containment isolation signal existing relay C picks up and opens normally closed contact C.

%is in turn renoves pwer from, aM deenettjizes the new relays.

When containment isolation is reset contact C will reclose. For each X-Y pair of new relays there is a pushbutton in the control rom. Upon pushing the PB the X & Y relays are energized.

Additionally contacts fran each relay also close (R1-X1 ard R1-Y1 for example) sealiry in the relays until another C.I.

signal.

Additionally contacts frm the relay pair are used to light visual indicators behind the pushbutton so that relay a:tuation can be verified frm the mntrol rom.

A third set of contacts are used to return the C.I.

signal back to the existing logic, one X-Y pair for each existin3 C.I. contact.

Relay contact pair R1-X3, -Y3 showirr3 the wiring for a rormally closed contact ard relay.

A review has been made cf all events analyzed in the Ginna Station FSAR and the events requirire analysis by USNRC Regulatory Guide 1.70.

We events related to this nodification are:

(1) major and minor fires, (2) seismic events, (3) loss of all A.C.

pwer to the station auxiliaries, (4) fuel handling accidents, (5) primary system pipe rupture, (6) events leMing to high containment pressure, (7) radioactive release inside containment, and (8) inadvertent opening of a pressurizer safety or relief valve. W e nodification does not increase the possibility or impact of a fire. Additional wiring ard cable will be Mded in this nodification, which oauld Md to the fire loading of the plant.

Berefore, 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 loMing caused bj' this nodification.

Section 26.2 of the Design Criteria provides requirements to preserve any silicone foam fire stop or seal that need to be penetrated.

Per these requirements of the Design Criteria, _ it is determined that this nodification shall not affect the consequences of a loss of all A.C.

pwer to the Station auxiliaries, in that it does not use A.C.

power for control.

Because this nodification is considered seismic Category I,

a seismic event will not impact this nodification or cause this modification to impact any system 1. an Myerse way.

We remairder of the events listed in Section 3.1 all are effected by this nodification in that-without correct operation of containment isolation, an uncontrolled release of radioactive substances could occur.

Section 18.0 of the Design Criteria requires the new equipment to have the same or better operating ability as the

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PAGE 5 PORC ITEM i existing plant. equipment. Section 20.0 requires that separations be maintained.

Section 21.0 requires that upon failure of a cmponent that containment isolation is not degraded.

Section 22.0 requires that the modification shall be tested priot ' to use.

Section 6.0 requires cmpliance with IEEE stardards 336, 344, 383 aM 384. We containment isolation function aM initiatirg signals will be uncharged.

We nodification impacts only ' the reopening of containment isolation valves followirg a containment isolation signal reset.

Deliberate operator action will trx be required.to open each valve.

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

%e Cmmittee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification charges required.

his item IS COMPLETE.80-487 TSR 79-15, Evacuation Lights / Sound Powered Phones.

T.

Meyer presented Revision 0 of the bafety Analysis and Revision 0 of the Evacuation li<3 t wiring from h

Design Criteria for this nodification.

the charging pump room is routed alongside cable trays 161, 103, 102 and then enters conduit to -penetrate the operating level floor near the end of MCC1C. his conduit and one fran the sound p)wered ptone jack on the end of MCC1C enter A junction box and these two cables then run in a sirgle conduit through an existire sleeve in the Aux.

Building Addition wall to another junction box.

Wis box will be the location for the sound powered phone jack and the junction box for power to the Auxiliary Buildits addition red beacon lights.

This nodification is required since the plant evacuation alarm is presently inaudible in the Aux. Building addition area when the standby Aux. Feedwater Pumps are in operation.- Communications for testire the standby Aux.- Feedwater Pumps is the plant pagity systen.

A sound powered phone jack is needed in this area to ensure reliable ccmnunications.

We junction box near MCC1C will accmodate a 3/4" conduit going towards the evacuatica light for the ~ charging pump rocm' and a 3/4" conduit to the sound powered phone jack on the eM of MCC1C.

A 3/4" conduit fran this junction ~ box routes these signals to the Aux.

Building Addition junction box where a sound powered jack is to be installed.

A 1/2" conduit is routed out of this box to a red beacon light and the conduit continues fran here to the next red beacon light.

We events in Tables I ard II of

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PORC ITEM #

l l

procedure A303 which involve a release of radioactivity are unaffected by this nodification since they nuntion soundirg the plant evacuation alarm in certain instances but cb not spell out where the individual alarms are located.

W is nodification has no affect on the actions to be taken for any of these events.

Therefore, the margins of safety durirg normal operations ard transient corditions anticipated durirg the life of the plant have not been reduced.

We Mequacy of structures, systems and ccmponents provided for the prevention of accidents ard for the mitigation of the mnsequences of accidents have not been affected.

The Committee performed a Safety Evaluation and determined there were no unresolved safety questions or Technical Specification changes required.

Eis item IS COMPLETE.80-513 TSR 79-01, Main Feedwater Piping Drain Connections.

T.

Meyer presented Revision 1 of the Safety Analysis ard Revision 1 of the Design Criteria for this modification.

Wis design is for tha modification of the main feedwater lines inside containment.

It consists of Mding drain lines aml valves to the low points of each steam generator feedwater line to facilitate drainirg of the feedwater piping inside mntainment.

Functions he above mentioned nodifications are designed to prcuide drainiry of the main feedwater piping inside mntainment when required for maintenance or in-service inspection. A review has been made of all events analyzed in the Ginna Station FSAR ard the events requirirg analysis by NRC Regulatory Guide 1.70.

We events related to this modification are:

i 1.

Internal and External Events - Fire, Flood, Storm, or Earthquake.

2.

Spectrum of postulated steam ard feedwater pipiry breaks.

We first event considered is internal ard external events such as major and minor fires, floods, storms, or earthquakes.

We valves ard piping associated with this nodification are within the seismic portion of the feedwater system ard is designed as Seismic Category I.

As such, the consequences of this event are not increased; and l

the capability of this nodification to perform its interded function l

is not reduced by this event.

l

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We second event considered is the spectrum of postulated steam and feedwater piping breaks.

We drain valves ard piping are protected from the pipe whip effects of. a' main steam or feedwater pipe bread by virtue of the fact that double-eMed rupture of these pipes are -

precluded by the In-service Program. A failure of the new piping is within the bounds of a full diameter feedwater line break as previously analyzed.

As such, the' consequences of this event are not increased, and the capability of this nodification to perform its interded-function is mt reduced by this event.

W erefore, the margins of safety during normal operations ard transient corditions anticipated during the life of the plant have not been reduced. We adequacy of structures, systems and camponents provided for the prevention of accidents and for the mitigation of the consequences of accidents have mt been affected.

We Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Tedinical Specification changes required. mis iten _IS COMPLETE.80-515 NR-2608C, Electrical Penetration Installation.

T. Meyer presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

NUREG 0578 and NRC letter dated October 30, 1979, require that BG&E:

provide by January 1,1981, two radiation nonitor systems in containment.

Rese new nonitors will require special cables. Spare NIS instrumentation cables shall be used and replacement cables ' for NIS use installed in a timely manner.

Tb do this new electrical penetration assemblies shall be installed.

Wis nodification is for the installation of three new containment penetrationc to a3d the necessary coaxial cables ard also provide spare instrumentation, control ard pwer penetrations that are needed.

We new penetrations shall be CE-30, CE-31, and CE-34.

A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Regulatory Guide 1.70.

We events related to this nodification are (1) major-ard minor fire, (2) seismic event, and -(3) the spectrum of accidents-inside of containment.

We nodification does not increase the possibility or impact of a fire.

Additional: wiring ard cable.will be a3ded in this' nodification, which could add to the fire loadirg of the plant. %erefore, the Design Criteria requires that all such cable neet the IEEE 383-1974 flame" test requirements.

Because of this there will be no. increase of fire loading caused by this modification. %is nodification has been classified seismic Class I and electrical Class IE and therefore shall be designed to have no impact on the plant during or after SSE.

Wis nodification is required by the Design Criteria to neet IEEE 317 ard ASME BPVC III

PAGE 8 PORC ITEM i subsection NE and therefore is qualified to not deteriorate and also to fully function through the spectrum of accidents inside of containment. % erefore, it shall have no impact on the plant duriry or af ter these accidents.

It has, therefore, been determined that the margins of safety during normal operations aM transient conditions anticipated during the life of the station have not been a-ected.

It has also been determined that the Mequacy of structures, systems, and camponents provided for the prevention of accidents and the mitigation of the consequences of accidents have not been affected.

We Committee performed a Safety Evaluation ard determined there were ro unreviewed safety questions or Tednical Specification darges required.

%is item IS COMPLETE.80-519 EWR-2607B, Containment Sump Level Indication.

T. Meyer presented Revision 0 of the Safety Analysis ard Revision 1 of the Design Criteria for this ndification.

We purpose of this nodification is to provide irdication of containment normal sump level (i.e., sump

'A').

Specifically this nodification involves the installation of level transmitters that will indicate water level inside containment fran the bottom of the 'A' sump (elevation 205')- to the basement floor (elevation 235').

%is nodification is intended to fulfill the requirements of the Three Mile Island Iessons Learned Task Force (Gection 2.1.9).

We proposed nodification is schematically stown on RG&E drawing 21439-341, Rev. O.

We proposed system is mde up of. two loops, each loop drives a vertical scale iMicator in the control roan. IE-2039 and LE-2044 are renote, differential pressure elements, which cense the hydrostatic head developed by the weight of fluid above the element.

This force is transmitted electronically to LT-2039 and LT-2044 which darge the electronic signal to a proportional 4-20 mAdc signal.

his 4-20 mA signal is transmitted to the Foxboro SPEC 200; analog control system conponents which perform the following functions: LY-2039A aM LY-2044A act as current to voltage converters aM. input isoltion devices.

%ey act to change the 4-20 mAdc signal into a 0-10Vdc which is used by the Foxboro SPEC 200 system.

LY-2039D ard LY-2044D are analog output devices which isolate the internal analog signal in the SPEC 200 system fran the external analog loco.

LI-2039 and LI-2044 are vertical scale indicators,nounted in the control roon, calibrated to rem out sump depth in feet of water We system to be installed is to meet the purposes of Harold Dentun's October 30, 1979, letter, although it will not be qualified to stardards requested by the staff.

Nevertheless, failure of the system will not jeopardize existing safety related equipment ard the system will serve a useful i

PAGE 9 PORC ITEM i function by nonitoring A sump level to the height diere B sump begins to fill.

A review.has been nude of all events analyzed in the Ginna Station FSAR and the events requirirg analysis by (ENRC Regulatory Guide 1.70.

.For this nodification the only event requiring analysis is fire.

Section 26.0 of the abwe referenced Design Criteria requires use of qualified flame retardant cable insulation. Werefore, there is no increase in the probability of a fire and re-analysis of a fire and 'its pastulated effects as a-result of this nodification is not required. Werefore, the margins of safety during normal opertions and transient conditions anticipated during the life of the plant have tot been reduced. We adequacy of structures, systems, and components provided for the prevention of accidents ard for the mitigation of the mnsequences of accidents have not been affected.

We Oammittee performed a Safety Evaluation ard detennined there were no unreviewed safety.

questions or Technical Specification changes required. W is it s IS COMPLETE.80-564 TSR 80-02, PCP Vibration Monitoring.

T. Meyer presented Revision 0-of the Safety Analysis ard Revision 0 of the Design Criteria for this nodification.

We nodification involves installation of two seismic pickups on the Reactor Coolant Pump notor frame and two proximity probes on the notor shaft in the area where it couples to the pump.

Signals to vibration nonitorirg equipment in the control rom will be routed through existirs unused cabling to penetrations CE2 and CE4 outside containment. -

New cablity will be routed in containment frm these penetrations to the probes. t is cabling can be routed in existing unused conduit and non-safety related cable trays.

Two velocity seismic pickups per pump feed into a velocity to. desplacement converter which provides this~ signal to one RVXY vibration nonitor in the control room.

To proximitors per pump feed another RVXY vibration nonitor in the control roon.

Wese vibration pickups also provide a signal 'o a relay rack ard if the signal exceeds a certain level due to excessive vibration, an annunciator becomes energized. Wie equipment is all powered by one rack nounted ' power supply at rack BMS-1.

None of the events analyzed in Ginna FSAR and listed in tables.1 ard 2 of A-303 will be affected by the installation of this nodification. We nodification does not change (1) the assumptions in any safety analysis -in the FSAR (2) the probability 'of occurrence of an accident (3) the consequences of an accident.

Werefore, the margins of safety durity rormal operations and transient conditions anticipated durirg the life of the plant have not been reduced.

We aiequacy of structures, systems and cmponents provided for the prevention. of accidents and for the mitigation of the consequences of accidents

y,a PAGE 10 PORC ITEM #

have rot been affected. %e Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Tectraical Specification danges required. W is item IS COMPLETE.80-807 TSR 79-14, Condensate Storage Tank.

T. Meyer presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

Wis design is for the nodification of the condensate storage tanks.

It consists of lowering the werflow pipe below the level of the diaphragm boltire argle, renoving plug frm 2"

condensate drain above diaphragm bolting angle.

We atxne mentioned nodifications are designed to Irevent water accumulation above the diaphragm ard to prevent pieces of diaphragm frm being drawn into the pumps. A review has been made of all events analyzed in the Ginna Station FSAR and the events requirirg analysis by ?RC Regualtory Guide 1.70 ard to events were found relatirg to these nodifications.

% erefore, the margins of safety durirg normal operations ard transient corditions anticipated durirg the life of the plant have not been reduced.

%e zdequacy of structures, systes and cmponents prwided for the prevention of accidents ard for the mitigation of the consequences of accidents have not been affected.

We Committee performed a Safety Evaluation ard determined there were ro unreviewed safety questions or Tednical Specification &arges required. 2 is item IS COMPLETE.80-916 TSR 80-09, Replacement of Bowl Assenblies on Fire Pumps.

G. Larizza presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

Wis nodification involves replacement of the 16M bowl for the fire pumps with vendor imprwed bowl 16MCF.

We above mentioned nodification is to enable the fire pumps to supply adequate flow at proper discharge head to emply with FSAR requirements.

A review has been nade of all events analyzed in the Ginna Station FSAR ard the events requiring analysis by IRC Regualtory Guide 1.70.

%e only event related to this modifica lon is internal ard external / fire,

flood, storm, or earthquake.

Werefore, the margins of safety durirg normal operations and transient corditions anticipated during the life of the plant have not been reduced..

%e adequacy of structures, systems ard components provided for the prevention of accidents ard for the mitigation of the consequences of accidents have rot been affected.

We Committee performed a Safety Evaluation ard determined there were no unreviewed safety questions or Te&nical l

t PAGE 11 PORC ITEM #

Specification danges required.

Wis iten IS COMPLETE.

80-1038 TSR 79-09, Diesel Generator Tachometers.

T.

Meyer presented Revision 0 to the Safety Analysis aM Revision 0 to the Design Criteria for this nodification.

This nodification involves the installation of a permanent camshaft driven tachometer for reMirg D/G rpm.

The energency D/G overspeed trip setpoint is presently checked with a hard held tachometer.

This nethod is not very accurate. The tachometer generator consists of a camshaft driven AC generator, the frequency of the generated voltage being proportional to camshaf t speed.

This variable frequency voltage is fed into a saturable core transformer; the magnitude of the output voltaje fran this transformer is proportional to the frequency of the input voltage of the generator.

The transformer box also contains a recti ler bridge which converts this AC signal to DC for the DC indicator calibrated in RPM.

This nodification does not dange (1) the assumptions in any safety analysis in the FSAR and its supplements (2) the probability of occurrences and (3) the consequences of an accident.

None of the events analyzed in Ginna FSAR and listed in Tables I ard II of A-303 will be affected by installation of this nodification. Therefore, the margins of safety during normal operations ard transient conditions anticipated durire the life of the plant have not been reduced.

The adequacy of structures, systems ard omnponents provided for the prevention of accidents aM for the mitigation of the consequences of accidents have not been affected. The Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required. This item IS COMPLETE.

80-1057 IMR-2607A, Containment High Pressure Indication.

T. Meyer presented Revision 0 to the Safety Analysis ard Revision 0 of the Design Criteria for this nodification. The purpose of this nodification is to provide a wider range irdication of containment high pressure.

Specifically this nodification involves the upgra3ing of existirg containment high pressure (0-90 psig) transmitters ard nuin control board pressure indicators.

This requirement is in accordance with reference 2.4 which states that

" measurement aM irdication capability shall include three times design pressure of the containment for concrete...".

The. design basis accident pressure load for the Ginna containment is 60 psig (FSAR Section 5.1.2-6a).

The proposed nodification is schematically shown on RG&E drawirg

PAGE 12 PORC ITEM #

03021-230, Revision 0.

We proposed nodification consists of replacing each of the three existirg pressure transmitters (PT-946,

-948 and -950) with wider range transmitters.

In addition, existirg main control board indicators will be nodified to correctly irdicate the wider span of containment pressure.

Transmitters PT-946, -948, and -950 measure containment pressure (via tubing penetrations) ard convert that measurement to a 10-50mAdc signal. We pressure signal generated by the pressure transmitter ard the power supply (PQ) is fed to both a duplex alarm ard an isolation amplifier.

The duplex alarm (PC), (which contains two separate, a3justable alarm points) charge state when the input exceeds either alarm setpoint.

Rese state changes are fed to the logic for steant w isolation ard containment spray.

Se isolation amplifier signal cping to the indicator nounted on the main control board.

We proposed nodification to the containment pressure nonitorirg system is intended to fulfill the commitment made to the NRC in reference 2.4 above, which is in response to the NRC's " Discussion of the Stort Term Lessons Learned Requirements" (October 30, 1979). A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Regulatory Guide 1.70.

For this modification the events requiring analysis are:

seismic event and those events which result in pressurization of containment which in turn cause isolation of the main steam lines ard/or possible initiation of containment spray.

Section 8.0 of the above referenced Design Criteria requires seismic qualification of the new transmitters and analysis to show the installation of these transmitters does not impose any new significant loading on the existire transmitter supports.

Werefore, the nargins of safety during a seismic event have not been reduced. As shown on reference 2.3 each of the transmitters replaced by this nodification feed a dual bistable device.

his bistable has two outputs which independently change state when the input exceeds each output's setpoint. One output feeds to the steamline isolation logic, while the other to the containment spray logic. Any event, (such as IDCA) which results in containment pressurization will potentially result in actuation of these bistables.

We analysis required by Section 10.0 of the Design Criteria is inten3ed to stow the effect of this modification on the bistable setpoints.

Specifically this new analysis will denonstrate that these new setpoints and any new instrument errors will result in actuation of this system that was within the tolerances established by the original accident analysis.

In zddition, Section 22.0 requires testing to verify proper actuation of this portion of the plant protection system.

Therefore, the margin of safety for those events which result in containment pressurization ard rely on nain steam line isolation

PAGE 13 PORC :

ITEM #

and/or initiation of contalment spray has rot been reduced by this operations and transient conditions' gins of ~ safety - during normal modification.

Werefore, the mar anticipated ~ during the life of the plant have not been reduced.

We : Mequacy of structures, systems,' and camponents provided for the prevention of accidents and for the mitigation' of the consequences of accidents have not been affected.

We Cmmittee perfonned a Safety Evaluation and determined there were no mresolved safety questions or Technical Specification diarges required.

Wis ' iten IS COMPLETE.

80-1203 EWR-2843, Condensate Storage Tank Instrumentation.

.T.

Meyer presented Revision 0 of - the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

We..existig level indication and alarm circuitry for the Condensate Storage Tanks consists of a single differential pressure transmitter, two renote vertical scale level indicators, a high level alarm, and a low level alarm.Section X.4.3.2-1 of BG&E responses to NRC's recommendations for Auxiliary Feedwater Systems dated November 28, 1979 and May 22, 1980 require redundant level indications aM low. level alarms in the control rom for the Condensate Storage Tanks-by January 1, 1981.

This nodification is. intended to fulfill those requirements.

We design will consist of two new transmitters, each with their own power supplies and ' associated high and low level alarm circuitry located ~ in separate Foxboro racks in the Relay Iban.

Each Foxboro rack has two independent power sources, one of which is a battery.

The existing single level indicator on the main control board will

.ith a dual indicator having empletely independent be' replaced w

inputs for each transmitter. We existing high an3 low level alarms are input to a single annunciator window.

Rese inputs will~ be replaced with the new redundant high and low level alaan circuitry.

In addition, both channels will also be nonitored by the plant computer, which has its own independent Inwer supply.

See attached P&ID. 03021-353, Revision 2.

Wis design provided functional redundancy 'of Condensate Storage Tank level indication and high ard low level alarm all. the way 'fran the differential pressure transmitters ' to the indicators and alarms, including their power-supplies.

A review has been made of all events analyzed in the Ginna PSAR and the events required ' by NRC Eegulatory Guide 1.70.

The' events related to this nodification are:

(1) ~ earthquake, (2) loss.-of one D.C.

system, (3) -loss of A.C.

power. to station auxiliaries.

We nodification is not required to be seismically designed. since the condensate storage-tanks. are not seismically designed.

the nodification will be designed such that, in the event

PEE 14 PORC ITEM #

of an earthquake, it will not damage safety related equipment. Each of.the two level indication diannels is powered by a separate D.C.

bus, therefore, loss of one D.C.

bus will not cause loss of' condensate storage tank level indication. High and low level alarms are annunciated on the plant conputer and main control board annunciator, e3 of which is powered by a separate D.C.

bus, therefore, loss of one D.C. bus will not cause loss 'of condensate storage low level alarm. Ioss of A.C. Iower will have not effect on this nodification.

terefore, the margins of saety during normal operations and transient conditions anticipated during the life of '

the plant are not decreased.

We structures, systens, - ard components provided for the prevention of accidents ard the mitigation of the consequences of accidents are not adversely affectd ard are adequate.

We Committee performed a Safety Evaluation and determined there were_ no unreviewed safety questions or Technical Sp[ecification charges required.

%is item IS, COMPLETE.

80-1321 DiR-1837, One-hour Rated Fire Barrier Between Kitchen Area and Control Room at Ginna Station.

W.

Hunt -r presented Revision 0 of the Safety Analysis and Revision 0 to the Design Criteria for this nodification.

We existing control room kitchen area is to be rebuilt with UL approved fire rated walls, ceiling ard door as per SER F14, 1979.

We licensee is to. provide a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire--

barrier between the kitchen area and the control roan. A review has been made of all events analyzed in the Ginna Station FSAR an3 the events requiring analysis by the USNRC Regulatory Guide,1.70.

%e-events related to this nodification are the Fire Event and Seismic Event.

None of the existing follow metal doors carry a U.L. label, all of which will be replaced by an "A" label ' door assembly.

this door will improve the fire safety of this building. %e replacement of the existirg hollow metal doors with. fire rated door assemblies will be designed with no degradation in the function of_ the seismic system and its components.

In addition, failure of the doors in a seismic event shall not result in damage to safety related equipment.

%e supports for the d] ors are designed such that the doors will not damage safety related equipment as a result - of a seismic event.

We Committee performed a Safety Evaluation-and determined there were ro mreviewed safety questions or Ted2nical Specification changes required.

his item IS COMPLETE.

PAGE 15 PORC ITEM #

80-1350 EWR-2720, Backflow Protection - Prevent Spread of Oil at Ginna.

T.

Meyer presented Revision 0 to the Safety Analysis and Revision 0 to the Design Criteria for this nodification. Within the Station there are areas that contain quantities of canbustible liquids.

the output frcxn this Design Criteria will prevent the spre d of flammable liquids via the underfloor drainage systems into Mjacent areas, where systems or equipment required for safe shutdown are located.

Eis nodification is required by paragraph 3.2.6 of the Fire Protection Safety Evaluation Report prepared by the USNRC dated 2/14/79 and RG&E response dated 9/28/79. W e areas involved in this nodification are as follows:

(a) All drain boxes in the Basement Floor (elevation 235'-8") of the Containment Building are included -

reference drawing D327-008, Revision VII, (b) W e drain box between the Charging Pumps in the Basement Floor of the Auxiliary Building aM the 6 inch diameter pipe connecting it to the sump pit at elevation 219'-0" - reference drawing D327-006, Revision VIII, (c)

We four floor drains in the rorthwest corner of the Intermediate Building at elevation 253'-6" between column lines 7 and H -

reference drawing D327-010, Revision V, (d) The drain box in the floor of the Mechanical Equipment Roan at elevation 253'-6" in the Control Building - reference drawing 0311-003, Revision XIII, (e)

The drain boxes in the Diesel Generator Roans 1A and 1B at elevation 253'-6" - reference drawing 33013-468 and D311-003, Revision XIII, (f) The drain box in the Turbine Building about 4 feet west of the 6 line just north of the F line - reference drawing D311-003, Revision XIII.

Within the existing floor drain boxes and/or connecting piping system, mechanical devices will be inserted to prevent spills in one area from backing up ard exiting at an adjacent floor drain box.

Reference General Arrangement Drawings 03021-354 and 355.

Where necessary, curbs may be required to confine the spread of a spill and reduce the extent of this nodification. A review has been made of all the events analyzed in the Ginna Station FSAR and the events requiring analysis by the USNRC Regulatory Guide 1.70.

We only event requiring analysis for the subject nodifications is a seismic occurrence.

Most of the WRK required by this nodification is located at or below floor level.

Rus no safety related equipment will be threatened by parts of this nodification falling upon it.

%e remainder of the WRK that may threaten Mjacent safety related equipment will be designed as Seismic Category I, for example:

We Inrtion of the WRK located in ard above the Residual Heat Renoval Pump Pit will be designed to resist the effects of the postulated seismic ground notion.

Werefore, there will be no adverse consequences caused by a design basis seismic event due to any WRK installed under this nodification.

We margins of safety during normal operations and transient conditions anticipated during the life of the station will not be reduced by installation of this

PAGE 16 PORC ITEM #

r modification.

We adequacy of structures, systems ard cmponents provided for the prevention of accidents _ and - the mitigation of the consequences of-accidents will not be reduced by installation of this modification.

We Ccunmittee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification darges required.

his item IS COMPLETE.

90-1379 SM-2846, Block Wall Modification.

T. Meyer presented Revision 0 of the Safety Analysis and Revision 0 to the Design Criteria for this -

modification.

We design.- will incorporate necessary structural modifications. to existirg concrete block walls shown on IE&E drawirgs 33013-971, 972, and 973.

Modifications will be required where analysis of the existing block wall indicates failure to meet the criteria.

We criteria used in this progra correspond -to the

" original Ginna FSAR" (Appendix A).

%is pecgran is a direct result of NRC Inspection ard Enforcement Bulletin 80-11.

We fmetion of the design will be to verify that all block walls "as built" or "as

~

nodified" fulfill the " original Ginna FSAR".

A review has been made of all events analyzed in the Ginna Station FSAR ard the events requirirg analysis by.NRC Regulatory Guide 1.70.

. We only event related to this nodification is a seismic event.

Were are to consequences of the proposed nodifications from an earthquake since the nodifications are seismically designed, Class I, according to

.the original FSAR.

An earthquake does not affect the safety function of the modification since it is designed for - a seismic event.

It has, therefore, been determined that the margins of safety during normal operations and transient conditions anticipated durirs 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 ard the' mitigation of the consequences of accidents have not been affected.

The Comnittee performed a Safety Evaluation and determined there were no unreviewed safety questions or Te&nical Specification chanjes required.

Wis item IS COMPLETE.

80-1380

.SM-2608A, High Range Effluent Monitor.

T. -Meyer presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

NUREG 0737 requires that noble. gas effluent 5

~

u Ci/cc be installed on monitors with an upper rarge capacity of 10 the plant vent and contairunent vent exhaust stacks at Ginna Station.

5

'(a) Noble gas effluent nonitors with an upper rarse capacity of 10

PAGE 17 PORC ITEM #

u Ci/cc Sc-133) are considered to be practical aM should be installed in all operating plants.,

(b)

Noble gas effluent monitoring shall be provided for the total range of concentration extending frcm normal condition (AIARA) concentrations to a maximum of 105 u Ci/cc (Xe-133).

Multiple nonitors are considered to be necessary to cover the ranges of interest.

Each effluent nonitor will continuously sanple the air in the vent and analyze it for particulate,

iodine, and noble gas concentrations.

Control terminals in the Control Boom aM Technical Support Center will provide the automatic logging function aM the operator-to-system interface.

A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by WNRC '

Regulatory Guide 1.70._

'Ihe events related to this nodification are:

( 1) _ major and minor fires, (2) a seismic event, (3) a loss of coolant accident, (4) a fuel handling accident, and (5) a high energy line break.

New wiring and cable will be required for this modification, which could a3d 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 iacrease of fire loading caused by this nodification.

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

Cables for this modification.will be installed per IEEE-384 and isolated at the power source with appropriate isolation devices. No vital equipment cables will be used in this nodification which have not been reviewed under a fire protection safe shutdown analysis.

Seismic F'

qualification of the nonitors is not required in accordance with reference 2.3 because none of the events for which the nonitors are required to operate are postulatM to be caused by a seismic event.

The primary purpose for installation of - the high rarge effluent monitors is to quantify potential releases following a LOCA 'with extensive fuel damage or a fuel handling accident.

For-these types of accidents tne nonitors will not be extnsed to a harsh environnent and therefore will remain operational. A IDCA with fuel damage or_ a fuel handling accident could cause the contairment vessel to became a radiation source to the effluent nonitors.

Over the 40 year life including the dose from a LOCA determined using source terms identified in references 2.4 and 2.5, is calculated - to be 110 rad.

This dose is negligible with respect to thresholds for material degradation in this system.

'Iberefore, this nodification will not be adversely affected by a loss of coolant accident or fuel handling accident.

A high energy line break in the ' intermediate building could subject the effluent nonitors to an environment of high l

PAGE 18 PORC ITEM #

tenperature (215'F) and humidity.

High energy line breaks are limited to relatively small break sizes in the intermediate buildiry by the Augnented Inservice Inspection Program. High rarge radiation releases are not anticipated during this situation.

Alternate techniques for radiological release assessment are available mder existing plant programs for routing aM emergency nonitorirg. Wese would include deployment of portable instrumentation for smple collection and radiation measurement to determine in-plant aM offsite radiological levels.

Releases from samall high energy line breaks outside containment may be dominated by the break. flow and not the releases through the plant or contaiment vents.

It has, therefore, been determined that the margins of safety durirg 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 componentc provided for the prevention of. accidents and the mitigation of the consequences of accidents have not been affected.

We Ommittee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification charges required.

Wis iten IS COMPLETE.

80-1407 SM 80-2147, Containment 480 Volt _ Distribution Panel.

T.

Meyer presented - Revision 0 of the Safety Analysis and Revision 1 of.the Design Criteria for this nodification. At the present time there'is only one 480 volt power panel existirg in Contaiment, with no spare breakers left.

Additional 480 volt feeders for welding equipment and future nodifications inside the Containment Vessel ~ are required.

Wis cbsign is intended to fulfill those requirements.

We design will consist of a new 480 volt power panel installed on the sane floor elevation as the existiry Containment p3wer panel, between two steel columns near the elevator.

We panel requires a 225 amp feed which will be obtained from notor control center IA through renetration CE-34.

Since there are no 225 amp breakers in MCC-1A, a modification kit will be purchased to change cnnpartments 4F and 4H -

into a single 225 amp breaker compartment.

A review has been made of all events analyzed in the Ginna FSAR and the events required by NRC Regulatory Guide 1.70.

%e events related to this nodification are:

(1) major and minor fires, and (2) a seismic event..New wiring and cable will be required for this nodification, which could add to the fire loadirg of the plant.

Werefore, 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 nodification.

We Design Criteria

u PAGE 19

-popc-ITEM #

1

. requires that the 480 volt power panel be located and _ installed.such that it will;not increase the impact of a seismic event. Werefore, the margins.of safety during. normal opertions and transient conditions anticipated during the ~ life of the plant are not decreased. We structures, systems, and camponents provided for the

~

prevention of accidents and the mitigation of the consequences of accidents are not a3versely affected and are adequate.

We Comittee performed a Safety. Evaluation and determined there were no -

unreviewed safety questions or Technical Specification darges -

required.

his item IS COMPLETE.

80-1462 TSR 79-15, Evacuation Light / Sound Powered Phone Installation.

T.

Meyer presented Revision 0 of the Safety Analysis ard Revision 1.to the Design Criteria for this nodification.

%erefore, the margins of safety during normal operatiom and transient conditions anticipated during the life of the plant:have not been reduced. We adequacy 'of structures, systems and camponents _ provided for - the prevention of accidents an3 for the mitigation of the consequences of accidents have not been affected.

We Comnit. tee ' performed a Safety Evaluation and_ determined there were no tnreviewed safety questions or Technical Specificat; ion darges required.

Wis iten IS COMPLETE.

80-1540 EWR-2831, Anchor Bolt Modification of Electrical Assembly.

T. Meyer 4

presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

We purpose-of ~ this engineering and construction effort is to evaluate and nodify the l

andorages of selected existing electrical assemblies so that continued assurance of their capability to withstand a safe shutdown i

earthquake can be denonstrated.

Wis effort is in response to a request from the NRC (Reference 2.1) and culminates the overall l

evaluation of existing andors as addressed in the Seismic Evaluation Program, reference 2.2.

Specifically this nodification project is phase III of IE&E's _ Seismic Action Plan. Other phases of the Action Plan determine which assemblies have andorages requiring

+~

modifications or anchors which are inaccessible for testing.

All such assemblies will be nodified with ' new - anchors using current -

criteria and taking no credit for existing andors. _ All an&orages, bolts and structural hardware will be designed and installed such that the assemblies will be sdely secured durirg a seismic event (SSE).

A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by NRC Regulatory Guide 1.70.

We only event related to this nodification is that all

.c

. u.

~

~

PAGE 20 PORC ITEM #

electrical assemblies be able to withstand a safe shutdown eartinuake.

We safe shutdown earthquake (SSE) event' has been res _ewed and it was determined that all proposed anchorages must be consistent with IEEE-344, 1975, reference 2.4.

We proposed anchorages shall be designed and installed such that the structural integrity of the assemblies shall be maintained when subjected to seis:nic accelerations acting simultaneously in the vertical and horizontal planes, thus insuring that the nodified assemblies, themselves, do not became a missile during an SSE.-

he addition to the new anchorage system will not affect the normal. operating parameters of any plant system.

Werefore, the margins of safety during normal operations and transient corx11tions anticipated durirg the life of the plant have not been reduced.

We 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. We Cmmittee performed a Safety Evaluation and determined there were no' unreviewed safety questions. or Technical Specification changes required.- h is item IS COMPLETE.

80-1549 EWR-2462, Reactor Coolant Punp Oil Collection System.

T.

Meyer presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

Wis nodification consists of a system of drip pans, splash guards, and enclosures for oil collection to reduce. the potential for fire resulting frm oil contacting and ignitirg on. hot reactor coolant system cmmnents.

Drain piping will be run frm seven collection points on the RCP to a new oil collection storage tank. His nodification is in response to the USNRC Fire Protection Safety Evaluation Report, Section 3.1.39, "RCP Lube Oil Collection System." A drip pan will be placed

-under the oil-fill and drain valve to collect any leaks. frm the valve.. Drain provisions will be provided.

A drip pan will be provided below the notor lower oil pot.

%is pan will collect oil r

l' that may seap down the shaft from the lower oil pot due to gasket l

leakage, overflow caused by lower oil pot cooling coil leakage, or spillage when the' lower oil pot is topped off. A segmented circular cyclinder arranged above the pan will collect oil 'hich is thrown by centrigual force from the notor/ pump flange coupling.

A catch basing will be provided around the notor shaft below the lower oil pot and will -collect any oil that may be thrown frm the shaf t by its rotation.

Wis catch basing will extend out frm the notor to I -

include the area below the lower oil pot level detector so that leakage frm the level detector flange is contained.

A drip pan, placed under the upper oil pot level detector, will collect any oil l;

PAGE 21 PORC ITEM #

leaking frm the level detector fittings.

We oil _ cooler has a nunber of flanged connections which could be the sources of potential oil leaks.

An enclosure will be installed around the oil cooler and oil cooler piping to isolate any leakiry oil fran hot camponents of the reactor coolant-pump.

We oil life system pressurizes oil to 2500 PSI during startup of the notor which provides the initial oil film on the thrust bearing shoes. A leak in this system could result in oil being sprayed on hot system conponents.

We oil lift system enclosure will isolate the high-pressure oil canponents-from the local environment.

An oil collection drain and storage system will be installed to collect spill oil fran the seven collection points described in sections 1.2.1 to 1. 2. 6.

Oil from the collection points will be run to a comnon collection tank, that will be gravity fed. A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC ' Regulatory Guide 1.70.

%e events related to this nodification are:

(1) the Ioss of Reactor Coolant

Flow, (2) the Fire Evcnt, and (3) the Seismic Event.

We installation of an _ oil collection system on the Reactor Coolant Pumps in no way increases the probability of. a Ioss of Reactor Coolant Flow event.

We systen is d) signed to not-create any interference or to diarge the operating ' ability of the Reactor Coolant system, as required by the Design Criteria.

The installation of the oil collection systems shall reduce the probability of a Fire Event resultirg fran free oil igniting on hot Reactor Coolant Components.

We installed oil callection systen shall be supported and restrained s) as to naintain its integrity during a seismic event, as required by the Design Criteria.

We supports shall be designed to ASME Section III, Subsection NP.

It has, therefore, been determined that the margins of safety durirg normal operations and transient canditions anticipated during the life of the plant have not been affected.

It has also been determined that the adequacy of structures, systems, and canponents provided for the prevention of accidents ard the mitigation of the consequences -of accidents have not been affected.

We Conmittee performed a Safety Evaluation ard determined there were no unreviewed safety questions or Tedinical Specification dianges required.

Eis item IS COMPLETE.

80-1562 EWR-2603, Direct Indication of Valve Position - Pressurizer Safety Valve Position Indication.

T.

Meyer presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this-modification.

We purpose -of this nodification is to provide

PAGE 22 PORC ITEM #

irdication aM alarming, in the Control Rom, of pressurizer safety valve position. This nodification satisfies the requirments of the tree Mile Islard Lessons Learned Task Force (Section 2.1.3.a).

This nodification is functionally shown on BG&E drawirg 03021-225 Revision 0 and consists of the followirg:

Linear variable displacement transducers (LVDr's), (tag nos. ZT-434 ard ZT-435) are connected directly to each pressurizer safety valve's stem.

Any valve stem notion will be imnediately translated to a charge in output voltage of the LVDr. Tag numbers ZY-434 ard ZY-435 represent an integral power supply, ard calibration unit for tach LVDr. Wese units also contain a bistable device which provides a contact state change when a preset setpoint value of displacement is exceeded.

This is used to generate an alarm on the Main Control Board, (tag nos. ZAH-434 and ZAH-435). Actual valve stem position is displayed in the Control Rom via indicators ZI-434 and ZI-435.

Drawirg 03021-225 Revision 0 also schematically shows the existirg system which can determine safety valve openirs and/or valve leakage. %is system consists of: resistance temperature detectors (RID's) TE-436 and TE-437, resistance to current converters 1M-436 and TM-437, bistables TC-436 and TC-437, aM temperature irdicators TI-436 and TI-4 37.

On high safety valve outlet temperature, the bistables change state ard actuate their respective alarm (TAH-436 or TAH-437) on the main control board.

A review has been made of all events analyzed in the Ginna Station FSAR ard the events requiring analysis by USNRC Regulatory Guide 1.70.

For this nodification the only events requiring analysis are fire and seismic. Section 26.0 of the above referenced Design Criteria requires use of qualified flame retardant cable insulation.

%erefore, there is no increase in the probability of a fire or re-analysis of a fire aM its postulated effects as a result of this nodification.

The pressurizer code safety valves are classified seismic Category I.

Section 10.0 of the above Design Criteria stipulates that the Mdition of the LVDr's to safety valves will not degrade their ability to perform their interded function under those conditions for which they are designed. Therefore, the ability of these valves to function durity a seismic event has rot been degraded.

Therefore, the nurgins of safety during normal opertions ard transient conditions anticipated during the life of-the plant have not been reduced. The Mequacy of structures, systems, and cmponents provided for the prevention of accidents and for the mitigation of the consequences of accidents have tot been affected. The Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes rcquired. This its IS COMPLETE.

PAGE 23 PORC ITEM #

80-1618 NR-2421, Hydrogen Supply to VCT and Hydrogen Recombiner.

T. Meyer presented Revision 2 of the Safety Analysis ard Revision 3 of the Design Criteria for this nodification.

% is nodification consists of building a small storage building detached fran the Auxiliary Building for the storage of the necessary hydrogen for the volume control tank ard hydrogen recambiner.

A hydrogen line will be run fran this building into the Auxiliary Building ard will be tied into the existing hydrogen line to the volume 03ntrol tank in the vicinity of the volume control tank.

A hydrogen line will be run fran the new storage building into the Auxiliary Building and will be tied into the existing hydrogen line to the hydrogen reconbiner.

This line will be inguarded pipe inside the Auxiliary Building as the existing line is.

%e hydrogen supply will be retoved fran the Nitrogen Storage Building and the hydrogen lines fran the Nitrogen Storage Building to the volume control tank area will be retoved.

This nodification will retove the ptential of a hydrogen fire in.

safety related ares of the Auxiliary Building.

Sufficient nitrogen will be available for purging purposes.

Wis nodification is in resp)nse to the Fire Protection ard Safety Report (USNRC), Section 3.1.48, " Hydrogen Piping". We proposed storage building will house the hydrogen for the volume control tank and the hydrogen reconbiner.

We proposed hydrogen lines will carry the hydrogen from the storage area to the volume c>ntrol tank and the hydrogen recombiner.

A review has been made of all events. analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Regulatory Guide 1.70.

We events related to this nodification are (1) the Chemical and Volume Control Systen Malfunction, (2) the Fire Event, and (3) the Seismic Event.

We relocation of the hydrogen supply and installation of new hydrogen lines will N based on the original design flows and pressures.

Wis is a requirement of the Design Criteria.

Reduced flow or pressure will not happen as a result of this nodification ard hence the Chenical and Volume Control System Malfunction patential will not be afferted.

By relocating the hydrogen supply to an area that has a fire barrier between it and the plant, the probability of a fire damaging equipnent in the Auxiliary Building has been decreased.

We reloaction of the hydrogen lines removes them fran an area' that contains safety related cables and hence the potential of a hydrogen leak burning these cables.

% is nodification will enhance the fire safety of the plant.

Based on USNRC Regulatory Guide 1.29 and the.

RG&E Ginna Station Quality Assurance Manual, Appendix A,

che hydrogen line to the volume control tank upstream of the last closed valve (this nodification) is to be non-seismic.

We portion of the hydrogen line to the hydrogen recombiner that is to be relocated I

PAGE 24 PORC ITEM #

inside the Auxiliary Building in this nodification shall be designed to be non-seismic.

All hydrogen piping inside the Auxiliary Building in this nodification shall be designed to maintain structural integrity during a seismic event.

We nodifications outside the building will be designed to be non-seismic.

We existirs hydrogen system for the hydrogen recombiner is non-seismic ard the entire system will be addressed in a future review.

his installatin will not degrale the existing systems.

We storage building housing the hydrogen shall be constructed so that it does not degrade the adjacent Stardby Auxiliary Feedwater Buildirg in a seismic event.

We.nodification done by this Design Criteria will not degrade any seismic structures or systens.

It has, therefore, been determined that the margins of safety during nonnal operations and transient conditions anticipated during the life of the plant-have not been affected.

It has also been determined that the adequacy of structures, systems, and canponents provided for the prevention of accidents and the mitigation of the consequences of accidents have not been affected.

We Canmittee perfonned a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification charges required. Wis iten IS COMPLETE.

80-1659 TSR-80-10, Rewire W-2 Type Co. trol Switches.

W.

Faustoferri presented Revision 0 of the Safety Analysis and Revision 0 of the Design Criteria for this nodification.

% is nodification involves rewiring the control switches of safety related equipment to provide for constant vicual nonitoring of the neutral contacts of the control switches in the auto start circuit.

Wis visual nonitoring will be thru the green light for the individual equipment on' the control board.

tis nodification is required as corrective action for a problem with Westirghouse W-2 type control switches identified in IE Bulletin 80-20.

We neutral switch contact (after trip -

after close "o" contact) which is in series with the energency auto start circuit contact will be wired to the green light on the control board.

We "cto" contact which is presently in series with the green indicating light will be eliminated fran the circuit.

None of the design basis -events listed in Tables I and II of A-303 changes (1) the assumptions in any safety analysis in the FSAR and its supplements (2) the probability of an accident occurring and (3).

the consequences of an accident.

Werefore, the margins of safety during normal operations and transient conditions anticipated durirg the life of the plant have not been reduced.

We adequacy of structures, systems and conponents provided for the prevention of accidents and for the mitigation of the c>nsequences of accidents

PAGE 25 PORC I'IEM #

have mt been affected.. We Committee performed a Safety Evaluation and. determined there were no unreviewed safety questions or Technical Specification changes required. Wis it s IS COMPLETE.

80-1677 EWR-2950, Containment Isolation and Spray Reset.

W.

Faustoferri presented Revision 0 of the Safety Analysis ard Revision 0 of the Design Criteria for this nodification.

We NRC Staff Interim Position on Purging, transmitted to RG&E on October 23,1979 requires that Cbntainment Ventilation Isolation (CVI) signals be segregated so that at least one automatic signal be operable when others are blocked, reset, or overridden.

An NRC Staff letter dated 10/31/80 requires that this be implemented by 1/1/81 or the purge valves must be left closed except durim cold or refueling shutdown.

We Staff has requested that these nodifications also be implemented on the Containment Isolation (CI) and Containment Spray (CS) actuation systems.

Sections 1.2,1., and 1.4 briefly describe the existig actuation and reset logic for CVI, CI, and CS at Ginna Station. In the existing design CVI may be actuated manually at the main control boata or automatically by either containment high radiation or safety injection signals.

If CVI is reset following an automatic actuation signal, both the remaining autmatic isolation signal and manual actuation will be blocked as long as the original signal persists. In the existing design CI may be actuated manually at the main control board or automatically by a safety injection signal.

If CI is reset following safety injection, manual actuation will be blocked as long a the safety injection. signal persists.

In the existing design CS may be actuated manually at the main control board or automatically on containment high pressure. If CS is reset following a high pressure signal with the high pressure signal still present, manual actuation will be blocked as long as the high pressure signal persists.

We purpose of this nodification is to renove the blocking mechanism in the actuation ani reset logic of the CVI, CI, and CS systems m that any actuating signal, manual or automatic, will cause the related system to perform its safety function at any time.

%e troposed neditication will remove the blocking logic by disconnecting the RI relay 'in each train of the CVI, CI, and CS systems (refer to attached schematic of CVI system).

An important. consequence of deleting the RI relay is that any automatic actuating signal must now be cleared before the related system can be reset.

23refore, if CVI was initiated by a safety injection signal or a high radiation signal that signal must first be cleared before CVI can be reset.

Manual actuation does not affect the resetting procedure. for any of the nodified systems. A

PAGE 26 PORC ITEM #

review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Guide 1.70.

21s design involves only wiring changes with no addition or removal of equipment.

We response of the plant to all transients will be unaffected.. %e capability of the plant to_ respond to new energency conditions during recovery fran a transient will be enhanced.

Therefore, none of the analyzed events will be affected by the proposed nodification.

It has, therefore, been deteunined that the margins of safety during normal operations ard transient coMitions anticipated durire the life of the station have not been affected.

It has also been determined that the Mequacy of structures, systems, and camponents prcnided for the prevention of accidents ard the mitigation of the consequences of accidents have rot been affected.

We Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification charges required.

Eis iten IS COMPLETE.

'80-1701 EWR-2463, Fire Dampers.

G. Meier presented Revision 0 of the Safety Analysis and Revision 1 of the Design Criteria for this modification.

Wis nodification is in response to the USNRC _ Fire Protection Safety Evaluation Report, Ibcket Ib. 50-244, Sections 3.1.9, 3.1.37,

" Fire Dampers,"

ard 4.9.2,

" Ventilation Duct Penetrations."

his nodification involves the installation of fire dampers in air conditioning and ventilation duct penetrations at seven (7) fire barrier locations. _h e oeneral locations of the fire dampers are as follows:

At the duct penetration between the southeast wall of the Intermediate Building ard the Auxiliary _

Building aM about 30 feet above the Auxiliary Buildirg Operatirg floor.

(FD-27)

At the duct penetration between the. south Tell of the Intermediate Building and the west end of the north wall of the Auxiliary - Building.

(FD-26)

At the duct penetration between' the water treatment area of the Service Building aM the Intermediate Building.

(FD-28) At the duct penetration above the Sample Roan in the Intennediate Building and the Chemical Laboratory in the Service Building.

(FD-29 ard FD-30)

At the ' duct penetration between the Air Sampling Roan in the Intermediate Building ard the south wall of the Turbine Building.

(FD-36) At the duct penetrations between-the Relay Ibom in the Control Complex and _ the Computer Roan.

(FD-31, 32, 33, 34, and 35)_ At the duct' penetrations to the charcoal filter above the spent fuel pit, through -the north wall of-the Auxiliary Building to the Intermediate Building.

(FD 2-10, 13, 15-24, and 37)

The above fire dampers shall have a fire ratirg of.1 1/2 hours except when the postulated fire severity is greater than 1 1/2 o

---

.------a---a m,.x

PAGE 27 PORC I1EM #

~

hours.

In this case the fire dampers will be rated at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Spaces between the ducts and the. barriers they penetrate will.be sealed to provide a fire resistance rating cemensurate with the fire hazards on either side of the penetration. Due to deficiencies in the existing roughing filter enclosure mer the Spent Fuel Pit, a new enclosure and roughing filter will be installed with the new fire dampers.

We roughing filter and enclosure shall be designed to allow for ease of maintenance of the sarcoal filters.

We general arrangement of this nodification is shown on BG&E Drawirg No. 30321-361.

We function of all of _ the fire. dampers is to provide a fire rated barrier between adjacent roans or buildings, at-duct penetrations, when activated by a fusible link.

We fmetion of the - roughing filter, located over the spent fuel pit, is to renove airborne particles, so as to prolong the life of the darcoal filters. A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USimC Regulatory Guide 1.70.

We events related to this modification are (1) the Fire Dient, (2) the Seismic Event, aM (3) the Fuel Handling Accident.

For this nodification, three events have been analyzed for effects on safe plant operation. We first event considered-is the " Fire Event".

As required by the Design Criteria, in the event of fire the installed. fire dampers shall cause delay. in the spread of, heat and snoke through HVAC ductwork and their penetrations for a prescribed period of time. - Each fire damper will be automatically activated by a fusible link.

We fire dampers shall have a 1 1/2 hour fire rating except where the postulated fire severity is greater than 1 1/2 hours.

In this case the fire dartpers will be rated at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

We spaces between the ducts ard the barriers they penetrate will oe sealed to provide a fire resistance ratirg commensurate with the fire hazards on either side of the penetration, as required by the Design Criteria.

We secoM e/ent considered is the " Seismic Event."

Tne installed fire dampers ard the immediately adjoining ducts shall have supports designed to withstand the seismic loads described in the Desibn Criteria.

We fire dampers shall be so supported so as to not degrade any safety related _ systems or _ components, during a seismic event, as a result of a loss of the support. We fire damper supports and the roughing filter frame, over the spent fuel pit shall be designed as Seismic Category

'I, as defined ~ in ASME Section III, Subsection NF, as required by the Design Criteria. The third event considered in this analysis is the " Fuel HaMling Accident."

In the event of a fuel haMling accident the darcoal and roughing filters shall be responsible for renoval of contaminents fran the intake air.

We installation of a new fire damper, roughing filters, and their frane shall not degrade the performance of the darcoal filter below the n

4

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.m.

..ai m

.a.

m.

PAGE 28 PORC ITEM #

effectiveness level defined by the Design Criteria aM the Ginna Station Provisional Operatiry License, ?b. DPR-18, Appendix A,

Section 4.11.1.

In mnclusion, the nargins of safety duriro normal operation aM transient mMitions anticipated during the life of the plant have not been reduced.

It has also been determined that the adequacy of structures, systems, and canponents provided for the prevention of accidents aM the mitigation of the consequences of accidents have not been affected.

We Conmittee performed a Safety Evaluation ard determined there were ru unreviewed safety questions or Technical Specification charges required. %is iten IS COMPLETE.

80-1707 EWR-1832B, Fire Signaling System.

W. Faustoferri presented Revision 1 of the Safety Analysis ard Revision 1 of the Design Criteria for this nodification.

A new fire signalire systen will be installed.

This system will provide the electrical functions described in Sections 1.2, 1.3, 1.4, 1.5, 1.6, 1.7, 1.8, 1.9, ard 1.10.

Equipment to be installed includes the following:

Equipment Location Function Satellite Station "A" Relay Rxxn Plant Fire Protection Satellite Station "B" Control Roan Relay Foan Protection Satellite Station "C Relay Iban Valve Supervision d

Fire Control Panels Control Roan Annunciator / Control Sensitivity Test Plant for detectors (reTote)

Panels Smoke Detectors Plant Fire Detection Heat Detectors Plant Fire Detection Enclosures Plant Terminal, Pull Boxes Manual Pull Stations Plant Fire Suppression Bells, Ibrns, Lamps Plant Alarm, Trouble Valve bbnitor Plant Valve Supervision Switches ReTote Relays Plant Alarm or Control Retote Panel Turbine Bldg.

Contairrnent Detection Included in the scope of these Design Criteria is the nountire of this equipment, the routing ard nountire of conduit, and all wirity associated with the new systen and nodifications to the existirg systems.

Also included is the wiring of fire suppression systen electrical devices, installed under other jobs, to the new fire signaling system.

We interface requirenents for these jobs are itenized in Section 10. We new fire signalire systen canplies with the following items in the NRC Safety Evaluation Report (SER):

PAGE 29 PORC I'IEM #

SER item D.C.

Section Description a) 3.1.1 18.14-

" Fire Detection Systems" b) 3.1.2 10.5

" Water Suppression Systems" c)

'3.3.3 10.6

" Gas Suppression Systens"'

d) 3.1.9 10 7

" Fire Dampers" e) 3.1.12 10.1.2

" Fire Alarms" f) 3.1.15 10.8

" Relay Ibom Water Suppression" g) 3.1.19 10.9

" Diesel Roon Water Suppression" h) 3.1.37 10.10-

" Fire Dampers" i) 3.1.42 1.2

" Signaling Circuits Supervision" j) 3.1.43 22.4

" Detector Testing

k) 3.2.8 10.11

" Exposed Structural Steel" 1) 5.1.6 18.14

" Containment Fire Detector" Wese criteria also include requirements for the following nodifications requested by the plant:

a) Former EWR #1692 request: new detection zone scope:

see 1.5.18 b) Former EWR #2158 request: New air duct detector scope:

see 1.10.2 c) Verbal Request request: upgrade temperature nonitor scope:

see 1.9.5b t

The new fire. signaling system will provide electrical supervision, receive fire alarm signals framt hese fire detection systens, annunciate such conditions to the plant operator in a way conducive

.to immediate action, and provide all automatic ~ actions required for the particular fire alarm condition. - We following exist in fire detection and suppression systems will be converted for supervision and control by the new fire signaling system.

Wis conplies with SER item 3.1.42,

" Signaling Circuits. Supervision."

J i

m

i PAGE 30 M

ITEM #

Condenser Pit Existing New System Nunber 1

S24 Detection Type Heat Heat j

Suppression Type Manual Deluge Manual Deluge Turbine Oil Reservoir Existing New System Number 2

S27 Detection Type

!! eat Heat Suppression Type Manual Deluge Automatic Deluge Seal Oil Unit Existing New.

System Number 3

S25 Detection Type Heat Heat Suppression Type Manual Deluge Manual Deluge 1A Diesel Generator Existing New System Number 4

S12 Detection Type Heat Heat Suppression Type Manual Deluge Pre-Action 1B Diesel Generator Existing New System Number 5'

S13 Detection Type Heat Heat Suppression Type Manual Deluge Pre-Action Aux. EWP Oil Reser.

Existing New System Number 6

S14 Detection Type Heat Heat Suppression Type Manual Deluge Manual Deluge Oil Storage Roczn

~ Existing

~New System Number 7.

S16 Detection Type Pneumatic Pneumatic Suppression Type

. Automatic Deluge Automatic Deluge l-

+)

__J L

,s PAGE 31 i

PORC ITEM #

Main Transfonner Existing New System Number 8

S20 Detection Type Pneumatic Pneumatic Suppression Type Automatic Deluge Automatic Deluge

1. 11 Unit Aux.

Transfonner Existing New System Number 9

S21 Detection Type Pneumatic Pneumatic Suppression Automatic Deluge Automatic Deluge

1. 12A Station Aux.

Transformer Existing New System t: umber 10A S22 D. action Type Pneumatic Pneumatic Suppression Type Automatic Deluge Automatic Deluge

1. 128 Standby Sta.

Aux. Transformer Existing New System Number 10B S23 Detection Type Pneumatic Pneumatic Suppression Type Automatic Deluge Automatic Deluge Cable Tunnel Existing New System Nuitber 11 S05 Detection Type Heat Smoke Suppression Type Manual Deluge Automatic Deluge Turbine Island Existing New System Number 12 S26 Detection Type Ibne None Suppression Type Sprinklers Sprinklers Service Building Existing New System Number 13 S19 Detection Type Ibne None Suppression Type Sprinklers Sprinklers

~

,a PAGE 32 PORC ITEM #

Computer Rom Existing New-System Number None S07 Detection Type Heat Smoke Suppression Type Manual Halon Automatic Halon Pelay Room Existing New System Number None S08 Detection Type Heat Snoke Suppression Type Manual Halon Automatic Halon 1G Charcoal Filter Existing New System Number 14 S02 Detection Type Heat Heat Suppression Type Automatic Deluge Automatic Deluge The systems described in Section 1.0 have been evaluated ard found acceptable on the basis of conformance with the documents listed in Sections 2.1 through 2.10.

A review has been made of the events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Regulatory Guide 1.70.

We events - related to this modification are the fires analyzed in GAI Report ~#1936, ard the seismic event.

We fire detection and suppression systems are non-safety related and are not designed to function during or after an earthquake.

Where the nodifications involve Seismic Category I structures, the nodifications will not degrade the seismic capability of the structure. The seismically iMuced failure of the installations will not reduce the capability of any safety related system to perform its function.

Noncmbustible aM fire resistant materials will be used whereever possible in this nodification. All new wiring used for field routing will be IEEE-383 qualified.

We failure or inadvertent operation of the fire suppression system shall not incapacitate safety related systems or canponents.

We assumptions ard conclusions of existing analyses are undarged. No new types of events are postulated.

%erefore, the nargins of safety during normal operations ard transient conditions anticipated during the life of the station are undanged.

We zdequacies of structures, systems and emponents ' provided for the prevention of accidents aM the mitigation of the consequences of accidents are undanged.

We Committee performed a Safety Graluation aM determined there were no unreviewed safety questions or Te&nical Specification darges required.

This its. IS COMPLETE.

4

.,a'.,

PAGE 33 PORC ITEM #

80-1708 EMR-2028, ' Fireproofing of Steel Columns.

W. Faustoferri presented Revision A of the Safety Analysis and Revision A of the Design Criteria for this nodification.

A GAI fire protection study determined that structural steel columns in the Turbine lube oil reservoir area should be fire protected.

We extent of these nodifications are outlined in the Gr.I study titled, " Fire Protection Study-for Controlling Structural Failures Jeopardizing Cold Shutdown of Ginna Station," dated May 27, 1980, revised Septenber 16, 1980.

See attached sketch A for general arrangement.

We fire protection as ' outlined in Section 1.2 of *his Design Criteria wil allow Rochester Gas and Electric to neet u.v Ginna Station Fire Protection Safety Evaluation Report (SER) iten 3.2.8 Docket No. 50-244, dated February 14, 1979 by the U.S. Nuclear Regulatory Cmmission.

We application of fireproofing to all structural steel columns within ten feet of the oil reservoir dike, in conjunction with automation t

of the ~ existing manual fire suppression system will preclude structural failure during a fire. Automation of the existing manual fire suppression system is rot included within the scop of this Design Criteria.

We only event related to this nodification is fire.

We evaluation reported in Reference 2.2 assessed the impact of fires on the structural integrity of plant structures.

Based on existing fire loadings, it was determined that protection was required only in the area of the Turbine lube oil reservoir. It was determined that protection as described in Section 1,

was sufficient.

We evaluation has been submitted to NRC, however the NRC has not confirmed that this nodification is adequate to resolve SER item 3.2.8.

%e aidition of fireproofing to colunns within the ten foot area of the oil reservoir dikes will improve the fire safety of the plant.

Bis nodification is passive ard not designed to perform any safey function, therefore, safety during normal operations and transient corditions anticipated during the life of the plant will not be affected.

We Cmmittee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required. E is item IS COMPLETE.

s