ML20084N094

Annual Rept of Facility Changes,Tests & Experiments Conducted W/O Prior Approval & Challenges to Primary & Secondary Sys PORVs & Safety Valves,1982

ML20084N094
Person / Time
Site:	Ginna
Issue date:	05/20/1983
From:	Maier J ROCHESTER GAS & ELECTRIC CORP.
To:	Allan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 8306020453
Download: ML20084N094 (32)

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1982 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS CONDUCTED WITHOUT PRIOR APPROVAL AND CHALLENGES TO THE PRIMARY AND SECONDARY SYSTEM i PORV'S AND SAFETY VALVES t

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Without Prior Approval

- R. E. Ginna Nuclear Power Plant-

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[ { Page 1 of 30 EWR-1024 MAIN STEAM POWER OPERATED RELIEF VALVE REPLACEMENT' The main steam power operated relicA valves, (PORh's) provided .

controllad venting of the main steam system for st.artup, physics testing, and cooldown of the plant. In addition.the PORV's

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provide initial pressure relief of the main steam , system,-i.e. the' PORV's are tripped full open on high steam pressure prior to' opening of the main - steam \ safety valves . The existing PdRV's have balanced piston operators with a vo1Ume tank connected s to each operator. ,ThevEanks provide the s air.pressur.e,needed to keep' the valves 'clysed in the event tge"iristyt'enent air : supply is' lost. If the instrument air supply,is Ntost, existihg volume tanks wilj be depleted if the main steam shigh pressure switch ^;~

opens the PORV's more'than_three' times. At that timb theJORV.'s will fail open result'ipg'in an excessive reactor cooldgwnfraf.e . y V '

In accordance with currdnt' Westinghous.3 PWR design philosophy w y

the PORV's should fail c16 sed on loss ofJ air.

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This would include loss'of air due.to either a seismic event or loss of offsite s

power. Also the .PORV's must be capable of localsmanual control for plant'c'ooldown. Remote manual and automatic c o n t r o'i' o f the PORV's isanot re3R ired for cooldown 'or any other plant safety i function.~ . In . o'rd er-to ' a s s ure - a " fail-closed" mode, new air operated, spring _to close relief valves will be installed including handwheels for manual operation The existingsvalve' controls will be used=to operate the new PORV's. To provide-added plant-protection, .two non-seismic nitrogen qupply systemq will' be installed insthe Turbine Building toTpermit operatlon g of the PORV's i;n the event that loss - of of fsite' power 6chgses loss of

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the ~ instirument' %ir system. The design of the ; hew PORV's shall .

consider the -possibility of reducing the pois,e genegated during Y -

operation of the PORV's. Special valve trim ~ and/,or noise r, educing ,

equipment will; be (hdded if economically poss1ble. s \The planned w

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modifications hre shown schematically on RG&E'dra' wing' 3301348?9. ,

The new PORV's$3410 and 3411 provide pressure control of f.Jt p , 4 main steamfsystem during start-up, physics testing, 'and cooldown when the ^ steam dump valves to the condenser are not available.

These valves also provide initial ov.erprbssure" protection 4of, ,

-the steam system,1 thus preventing unneeded operation of thi mai'n steam safety valves.; The new PORV's ' are ddsigned ' to failm $ .

closed in the event-either3 the control signals or the air /nitrog,en y m o

.supplyLis lost. The eight main steam mechanical relief va[ves A provide the required pressure rblief. capacity for overpr6shure protection of'the main stgam system.. Existing valves 3506' ands

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3507. . permit' isolation of the PORV's for maintenance and repair.

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1982 Annual Report of Facility Changes,

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

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p Existing pressure controllers PC-468 and PC-478 and hand controllers HC-468 and HC-478 will be used to provide automatic and remote manual control of the PORV's for steam pressure control. New poaltioners on the PORV's will accept the pneumatic control

.; .' signals from the existing I/P transducers. Existing pressure switches PS-2093 and PS-2092 will be used to energize and open S ', new col,enold'v51ves 3410s and 3411S on high steam pressure.

This will cause the PORV's to trip full open for pressure relief.

i Position switches on the PORV's will energize the existing position i ,. indicating 1,ightsyt.a the control room. The high pressure nitrogen dN bottles provide an a'uxi'liary gas ~ supply for' operation of the i

(' PORV's during loss of instrument air. Valves and check valves are provided to permit isolation and replacement of each bottle.

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Pressure reducing valvos_,. pressure indicators and relief valves permit control of the low. pressure nitrogen supply to the PORV's.

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The control pressure of the nitrogen system is set below the

. normal air supplyspressure to prevent using nitrogen when instrument

's u air is available',6 Cbect valves prevent low pressure nitrogen m from leaking out-to the instrument air system if the nitrogen system becomes active. Drain traps will be provided on the MT dischargeLpiping to prlvent a '6uildup of condensate or rain water on (the .outl'et side of the PORV's.

% s, , A, revie'w haspbeen' reade of the events requiring review by .

d \ Regulatory Guide .1*~ 70. AThd bain steam PORV's are not required i do ' func':Lon to mitigntc -tpe' effects of design basis events.

Tide, main, steam safety valvess provide relief capacity equal to the stenm; generation tatie at maximum calculated conditions and q x mitigate' those events requiring relief capacity. No other analyses v 'are a'f f e c t'e d by this modification. The inlet piping to the

PORV's is not mod,1fico(.j Re discharge piping is not considered

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' high - energy {biping du(Nto.jits infrequent use. The effects of

.' ~ , high energy pipe' breaks are tpherefore. unchanged.

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1. sic Therefore, based;upon the above Unalyses, it has been determined

$jlkt: (a) the margins of safety during~ normal operations and

}4 V tr gusient ' conditions anticipatediduring the life of the stac. ion

Q re not reduced and (b)'the structures, systems and components

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mprovided for the-prevention of'acdidents and the mitigation

a.nofbthe consequences of accidents are adequate.

N ' #[ , Th e . P,13nt'dperations Review . C6nmittee performed a Safety hi V f' Evaluation and determined there were 'no u_nreviewed' safety questions y_ o'r Technical Specification changes required.

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$ /1982 Annual Report of Facility Changes, ,

1. ' Tests, and Experimenta Conducted e i Without'# Prior Approval j i '

R .' E . Ginna Nuclear Power Plant Unit No. 1

• d, '

- / # .? Docket No. 50-244

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'EWR-1832C FIRf1 RETARDANT COATING-CABLE TUNNEL i . /

Electrical cable's penetr[. ting the three existing Cable  ;

Tunnel barriers will be coviredi within isix linear feet of the barriers by an approved flame ' rete.rdant coating. 'Ihis modification is necessary to meet J a commitment made to the NRC on page 17

, of the June 4, 1980 submittal. This commitment was made subsequent

/ to the " Fire Protection Evaluation," and supersedes the previous j requirement to provide 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> , rated fire barriers at these loca-

'A tions. The 6/4/80 commitment readsfas follows: " RG&E has reevaluated '

the conceptual requirement'fdr 3 hour-rated fire barriers at ,

the Cable Tunnel Entrances ." "RG&E?has committed to the conversion '

of the water deluge jsy' stem in ,the = Cable Tunnel to automatic '

actuation by early-warning (smoke) detectors. In addition,

/ .l, RG&E has committed to l'nstall automatic water deluge! systems 4

i at the 3. entrances to the Cable Tunnel in the Intermediate, Control, and Auxiliary Buildings. These systems will also be actuated by early varning smoke detectors."The configuration of the Cable Tunnel entrances precludes installation of a rated' fire barrier. It is proposed that smoke? barrier,s

be installed at the 3 entrances to the Cable Tunnel' to contain any smoke or combustion products and that an approv'e'd .fl em e retardant cable coating be applied to cable trays within 6 feet ,

of the smoke barriers.  ?

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RG&E believes the proposed design will provide adequate -

control and confinement of a fire originating on either side of a smoke barrier." (a) A review has been -made of all events analyzed ~in the Ginna FSAR. This modification does not incdea^se ,

the probability of occurrence of any of these events./ The eventa '

affected by this modification are seismic and fire. The flame

retardant coating shall not cause tray s,uppo-rt/and' limits to be' exceeded. Therefore the seismic capabi}ity of the loaded trays will not be- degraded. The maximum coating thickness will not have a significant effect on cdble ampacity.- Therefore, i it will not significantly increase,the;probabil'ity of. fire caused l by insulation failure. This modification does not' create the possibility of an accident or malfunction of a different type than any evaluated previously. This modification does not reduce the . margin of safety as definod in any technical specification.

, = (b)' Based on the analysis performed in (a) the following has ,

been determined. The margins of safety during normal operations and. transient conditions anticipated during the life of the l station have not been reduced. The coating, When applied in accordance with approved procedures, will not significantly

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o 1982 Annual Report of Pacility Changes,

. Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1

i. Docket No. 50-244 Page 4 of 30 1

, reduce the - current carrying - capacity of the cable to which'it is applied. The adequacy'of structures, systems, and components provided or the prevention of accidents and the mitigation of

the consequences of accidents have not been affected.

L- The Plant Operations Review Committee performed a Safety >

Evaluation and determined there were no unreviewed safet r questions or Technical Specification changes required. l j

EWR-1836 TURBINE BUILDING WALL PRESSURIZATION,

! SECURITY AND FIRE PROTECTION ,

.i Should there be a rupture of e'ither.a 26" diameter Feed-Water Line or a 12." Main Steam;Line, or a crack break in.the 36" diameter Main Steam Line, there would be a resultant temperature and pressure build up-inside the Turbine' room will-. damage the existing structures between_the Turbine room and.the Control room, the Relay room, the Battery rooms, the Diesel Generator rooms, and the Air Handling room. To prevent damage to these existing structures and to enable a safe and' orderly shut down o f the plant, we would propose installation of the following.

structures: A steel diaphragm will be. erected at the south

! side of.the Turbine Building adjacent to the Control Building i

from elevation 253'-6" to 308'-8". The diaphragm will consist of horizontal steel beams spanning between the existing columns i to provide support for vertical corrugated steel panels. The i diaphragm will protect the' controls ,- Control room staff, Relay room, Battery rooms and Air Handling room from the collapse

!~ of the existing wall due to pressure resulting from the rupture

' of either the 20" Feed Water Line (below the operating floor) or a 12" Main Steam Line or a crack break in the 36" diameter Main Steam Line. A similar diaphragm will be erected on the

. north side of the Turbine Building adjacent to the Diesel- Generator rooms f om elevation 253 '-6" to 280 '-0" to protect the Emergency Diesel Generators and their controls from the collapse of the existing wall due to the same pipe breaks. The diaphragms will protect the existing wall between the Turbine room and Control room, Relay room, Battery rooms, Air Handling room and Diesel Generator rooms' from collapse and will also prevent any adverse effects of heat and humidity in these enclosures. Existing doorways will be protected by blast and pressure resistant doors.

All existing penetrations through the existing walls will be sealed against pressurized steam leakage. The steel diaphragms shall;be designed to sustain the design basis accidents (i.e. either

1982 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 5 of 30 a rupture of the 20" diameter Feed Water Line or a 12" Main Steam Line, or a crack break in the 36" diameter Main Steam Line) which develop pressure and temperature resulting from these pipe breaks. This modification shall also meet structural requiranents of 10 CFR Part 73 Section 73.55, and the fire orotection requirements of GAI Report No. 1936 or Appendix A to the Branch Technical Position APCSB 9 5-1. Full diameter breaks of the 36" main steam line are not postulated because the probability of such breaks is reduced to acceptably low values by the Inservice Inspection Program for High Energy Lines. The design basis is a crack break in the 36" main steam line, a full diameter break in a 20" feedwater line or a full diameter break in a 12" main steam line. The addition of passive pressure sheilding will not affect the normal operating parameters of the systems.

The diaphragms will be designed to allow passage through of all piping systems. The diaphragm design is adequate for all design basis loads and events.

The plant safety margins are in no way diminished by the diaphragms. Plant safety is improved by the additional protection o f fered by the diaphragms. Double sets of doors will not be provided. However, the fraction of the time that the doors to the Control building or the Diesel Generator rooms are open is small so that the probability of a pipe break' during that period of time is acceptably small. The modification will comply with the fire protection requirements. The new wall between the Turbine building and the Control room will provide the required protection of the Control Room. Existing walls between the Turbine building and the Relay room, Battery rooms, Air Handling rocm, and Diesel Generator rooms meet the established requirements.

The pressure shielding steel diaphragm is being designed as Seismic Category I and the seismic capability of the Control building and the Diesel Generator rooms will be maintained.

The margins of safety during normal operations and transient conditions anticipated during the life of 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 and the mitigation of the consequences of accidents have been degraded.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

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i 1982 . Annual Report of Facility Changes, Tests, and Experiments Conducted

, Without Prior 7pproval R. E. Ginna Nuclear Power Plant-Unit No. 1 Docket No.'50-244 Page 6 of - 30 .

EWR-2462 REACTOR COOLANT PUMP OIL COLLECTION SYSTEM This. modification consists of a system of drip pans, splash guards,-and enclosures for oil collection to reduce the potential for fire resulting from oil contacting and igniting on hot reactor coolant. system components. Drain piping will- be run from seven collection points on the RCP to a new oil collection storage .

tank. This modification is in response to the USNRC Fire Protection Safety Evaluation Report, Section 3 1.39, "RCP Lube Oil Collection i Systems." A drip pan will be ' placed under the oil fill and l drain valve to collect any leaks from the valve. Drain provisions i

will be provided. A drip pan will be provided below' the motor .

-oil pot. This pan will collect oil _ that may sap down the shaft from the lower oil pot due to gasket leakage,- overflow caused by lower. oil pot cooling coil leakage, or. spillage when the

lower oil pot is topped off. A segmented. circular cylinder

arranged above the -pan will collect ' oil which is thrown by centri--

fugal force from the motor / pump flange coupling . 'A' catch ba' sin

! will be.provided around the motor shaft below the-lower. oil

. pot and will collect any oil that may be thrown from the shaft l by its rotation. ,This catch basin will' extend out f rom ' .the motor to include the area below the. lower oil pot level . detector .

so that leakage from level detector flange is contained. .A drip pan, placed under the upper oil pot level detector, will collect any oil leaking from the level detector fittings. The l~ oil cooler has a number of ~ flanged connections which could be I the sources of potential oil leaks. An enclosure will .be installed i around the oil cooler and oil cooler piping-to isolate any leaking oil from hot components of the reactor coolant pump. The oil

' lift system pressurizes oil 2500 PSI during startup of- the motor ,

i lwhich provides the ir.itial oil film on the thrust bearing shoes.

1 A leak in this system could result- in oil being sprayed on hot L system components. The oil lift system enclosure will isolate l the high-pressure oil components from the local environment.

i .An oil collection drain storage system will be installed to collect spill oil from the collection points. Oil from the collection points will be run to a common collection tank, that l- will be gravity fed.

t-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. The events related to this modification are: 1) the' Loss of Reactor coolant Flow, 2) the Fire Event, and 3) l the Seismic Event. - The installation of an oil collection system j .on the Reactor Coolant Pumps in no way increases the probability 9

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1982 Annual Report of Facility Changee, Tests, and Experiments Conducted Without Prior Appro/al R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 7 of 30.

of a Loss of Reactor Coolant Flow event. The system is designed to not create any interference or to change 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 resulting from free oil igniting on hot Reactor Coolant Components. The installed oil collection system shall be supported and restrained so as to maintain its integrity during a seismic event, as required by the Design Criteria. The supports shall be designed to ASME Section III, Subsection NF. It has, therefore, been determined that the margins of safety during normal operations and transient conditions anticipated during the life of the plant have 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.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

, EWR-2604A DETECTION OF INADEOUATE CORE COOLING RCS I The purpose of this modification is to increase the range of the existing saturation margin monitoring system and provide a redundant wide range reactor coolant system pressure transmitter.

The proposed modification is shown schematically on RG&E drawing 10904-92 Rev.l. The modification to the existing saturation margin monitoring system consists of the following: Installation of PT-420A and obtaining RCS wide range pressure from it (PT-420A) and PT-420, which replaces existing narrow range pressure inputs (PM-429C and PM-430-A1). Conversion of the existing analog instrumentation to cover the range of 200-700 degrees F and 0-3000 psig. To accomplish this and still maintain system accuracy the analog instrumentation is split into two sections; (within each channel or rack); a wide range (0-3000 psig) and narrow (1750-2500 psig).The narrow range section will be used during normal operation, whereas the wide range section will be used during startup. Transfer from one section to another will be completely automatic and " b um pl e s s " (to the extent practical).

Outside of these changes, the system will function in a manner similar to that of the existing system. That is the pressure signal will be received and converted to a voltage signal (PY420A and PY420) this signal is then fed to a function generator (PY420A-1

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1982 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 30 and PY420-1) which converts RCS pressure to-equivalent saturation temperature. This saturation _ temperature (TY410-4) is compared

, to RCS hot-leg temperature and the difference representing the' subcooling margin. .This subcooling margin, from the wide range section, is sent to a switching module, which based on RCS hot '

leg temperature . selects which subcooling margin 'is to be _ used. -

The output of the switching unit is sent to be used. The output of th'e switching unit is sent to a bistable device, which alarms when the margin is 40 degrees F, and an output isolation amplifier

, which transmits a 4-20 ma de signal to an- indicator in the Control Room.

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 only events requiring analysis are fire, seismic and the spectrum of the loss of coolant-

. accident. The Design Criteria requires use of qualified flame l retardant cable insulati6n. Therefore, there is no increase -

'i n the' probability of fire or re-analysis of a fire and its

! postulated-effects as a result of this modification. The Design Criteria requires seismic mounting of the new pressure transmitter i and any additional conduit supports installed inside containment.

It is also required that the instrument tubing and its supports-be ' designed to withstand the SSE. Therefore, _the consequences of a postulated seismic event will not be~ changed as a result of this modification. Other portions of the system need not l be seismic category 1 in accordance with Harold Dentons's USNRC l letter dated,0ctober 30, 1979. Since the new pressure transmit -

I ter requires a . fluid path to the RCS, the potential for failure

! of this tubing connection exists; and hence a LOCA event. However, l given the relatively small size of the tubing, .the consequences i

of such a failure are minimal when corr. pared to the consequences of the previously analyzed large IDCA's. 'Iherefore, the consequences

! of a failure of this tubing will not change the results and consequence of previously analyzed LOCA's. This modification has been reviewed to ensure that failure of any electrical cable I 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. Cables for this modification will be installed per IEEE-384 and will be isolated at the Foxboro racks with qualified isolation device. No vital equipment cables l will be used in this modification which have . not been reviewed under a fire protection safe shutdown analysis. Equipment inside containment is being qualified to withstand accident conditions and is therefore assured of proper functioning when required.

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1982 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 9 of 30-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, system, and components provided for the prevention of accidents and-for the nitigation of the consequences of accidents have been affected.

-The Plant-Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

EWR-2607D CV TRANSMITTERS The purpose of the instrument transmitter upgrade program is to provide direct replacements, which have been environmentally qualified in accordance with current. standards, for the existing transmitters used in safety related applications. Foxboro N-E10 Series transmitters will be used for this purpose. The existing qualification for N-E10 Series transmitters is in accordance with IEEE 323-1971 and IEEE 344-1971. ' An ongoing cooperative program is.now underway to qualify th'ese transmitters in accordance with IEEE 323-1974 and IEEE 344-1975 as endorsed and supplemented by NUREG 0586. This test program 'is described in criteria docu-ments. 'The program is scheduled for completion in October 1981.

Foxboro 600 Series transmitters now installed in Ginna~ Safety related systems have been qualified for pressure and temperature effects of accidents , but not long-term radiation. They will perform the safety functions for which'they were intended, which are. completed-within a short-time after the accident. They will not reliably perform the long-term post accident monitoring ,

function required by the DOR Guidelines. Qualifications to these criteria provides the necessary compliance.with the " DOR Guidelines" as provided for in the October 30, 1980 RG&E submittal concerning environmental qualification of electric equipment.

These criteria are intended to provide guidance for replacing existing installed transmitters by functionally identical transmit-ters with greater tolerance for harsh environmental conditions.

Any mod.ifications to safety related instrumentation system designs shall be addressed by separate Criteria. The proposed replacement o f ' exis ting transmitters in safety-related systems subject to

-- harsh environments is intended to meet, in part, the Technical Specifications referenced in Paragraph 1.1.0.

'A review as been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC Regulatory i

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l9'82 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 30 p Guide 1.70. The events. requiring analysis are the LOCA,.HELB,and the SSE. The ef fects of LOCA/HELB events are addressed in the

Design Criteria. The required temperature, pressure, and radiation

exposure are in the Design Criteria. The test program envelopes are in' the Design Criteria. The test program in all cases envelopes the required values. The composite RRS (required- response spectrum)-

J for all transmitter locations is given by Figure 1 of the Design ,

Cr ite ria'. In.all cases.the TRS (transient response spectrum) -

envelopes the RRS. The Design Criteria provides direction for ,

assuring. that. the transmitter supportingL structure will remain 3

intact _during and af ter the SSE. Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the plant have not been reduced. _The adequacy-

i. - of structures, systems, and components provided for the prevention

! of accidents and for the mitigation of the consequences of accidents

have not been affected.

l The Plant Operations Review Committee performed a Safety

Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

EWR-2921 STEAM DISCHARGE RADIATION MONITOR i

1- Item II.F1, Attachment 1 of NUREG 0737' requires that' noble l gas effluent monitors be installed with an extended range designed

! to function during accident conditions as well as during normal operating-conditions. EWR 2921_ addresses effluents rele_ased

through the main steam safety valves and PORV's, turbine driven l auxiliary feedwater pump (TDAFWP) discharge, and' condenser air-

! ' ejector exhaust. Main steam activity in .each steam line will

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be monitored upstream of the code safeties, PORV's and TDAFWP-supply lines with adjacent steamline detectors located in the Intermediate Building. The outputs of the steamline monitors and the t open/close status of the safeties, PORV's, and TDAFWP supply valves will be logged on strip chart recorders in the Control Room. Any radioactive releases can then be quantified by integrating the product of steam activity and the exhaust flow rate with respect to time. An air sampling unit with the required range and sensitivity will be installed on the air ejector exhaust stack near the existing R-15 unit in the Turbine

. Building. By programming the appropriate conversion factors, ,

the sampler. control' console can print out release rates in micro-curies per second in the Control Room.

A review has been made of all events analyzed in the Ginna

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1982 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 11 of 30 FSAR and the events required by NRC Regulatory Guide 1.70.

The events related to this modification are (1) . major and minor fires (2) a seismic event,.(3) high energy line breaks, (4)

a. loss of coolant accident and (5) steam generator tube ruptures.

New wiring.and cable will.be. required for this modification, 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 I be no increase of fire loading caused.by this modification.

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 event. . Position switches will be mounted on the main steam . code safeties for indication of valve open/close status. This will require drilling holer in the valve' cap assembly. Since the valve cap is not part of the pressure retaining boundary of the valve, and performs

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no safety function, the installation of these switches will not degrade the seismic capability of the valves. All other equipment shall be installed such that it will not cause the failure of any safety related equipment during or after a seismic

• event. The. primary purpose for installation of the extended range main steamline monitor is to quantify potential releases following a large LOCA with core damaged. For other high energy line' breaks, such as steam or feedwater line breaks, the monitor may be useful but is not required. Thus, no environmental qual .

ification for harsh environments is required. For breaks inside containment the monitors will be exposed to no adverse environment and should remain operational. Large high energy line breaks in the intermediate building are precluded by the existing augmented inservice inspection program. Small breaks may occur, which could affect the monitors. However, uncertainties in quantifying  :

releases from the ruptured steam generator will outweigh un-certainties in the releases from the intact steam generator

, through the relief and safety valves or turbine driven auxiliary-feedwater pump. Sampling of the intact steam generator will provide acceptable quantification of releases and environmental l qualification of the main steamline monitors is not required.

A large LOCA with extensive fuel damage could cause the containment

, vessel to become a radiation source to the steamline monitors.

I

'Over-the 40 year life of the plant the total dose accumulated by the detectors, including the dose from a large LOCA is calculated i

to be 110 rem. This dose is considered negligible. Therefore,

' this modifica tion will not be adversely affected by a loss of coolant accident. Steam generator tube ruptures will not subject i <

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1982 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 30

~

the monitors to a harsh environment and thus.the monitors will be useful in quantifying releases during these events.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

EWR-3118 MOISTURE CARRYOVER MODIFICATION This modification consists of the following changes to the steam generator moisture separators: 1) Modify the mid-deck plate.to provide adequate deck-relief. 2) Provide improved i drainage for the upper tier of the' secondary separators. 3)

Add steam-water deflectors above the orifice rings. 4) Add three-sided steam chimneys to each of the deck

plate hatches.

5) The modification to the mid-deck plate will allow for proper

water drainage from the deck and also allow steam to be vented from below the deck plate. The addition of upper tier drains 1

will augment the existing drainage capacity to prevent overflow

, o f the existing drains and thus increased moisture carryover.

l The addition of steam-water de flectors above - the orifice rings

! will reduce the vertical momentum of the steam-water jet exiting the orifice. This will help reduce the impingement of the moisture laden jet onto the face of secondary separators. The addition of three-sided steam chimneys to each deck plate hatch will -

l. create a relief for steam vented below the mid-deck plate.

Since it extends below the deck plate it will deflect the steam l~

jet exiting the primary separator tangential nozzle.

A review has been made of.all the events analyzed in the Ginna FSAR and the events requiring analysis by NRC Regulatory Guide 1.70. The events related to this modification are-(l) earthquake (2) postulated steam piping break and (3) steam generator

( tube rupture. There are no consequences of the proposed modification

i. from an earthquake since it is designed as Seismic Category L I. The modification is not designed to perform any safety function.

l Therefore,-an earthquake does not affect-the safety function

! of the modification. The proposed modification does not change

(- -the consequences previously analyzed of a postulated steam piping

.. break . The only affect of the modification is to slightly increase

! the steam flow pressure drop which would tend to slightly decrease

-steam flows during a postulated break. Since steam piping break

~

analyses assume the worst conditions, either dry-saturated steam

,. or 96%; moisture, the modification does not change the assumptions L

used. The proposed modification does not change the consequences l-

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1982 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 13 of 30 i

! previously analyzed of a steam generator tube rupture. .The

' modification is not designed to perform any safety function. There-fore, neither of these events can' affect a safety function of i' the modification. It.has, therefore, been determined that the margins of safety during normal operations and transient conditions l_ 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 i

and the mitigation of the' consequences of accidents have not

. been affected.

i The Plant Operations Review Committee performed a Safety 1

Evaluation and determined there were no unreviewed safety questions l- or Technical Specification changes required.

EWR-3130 REACTO'R COOLANT PRESSURE INDICATION t

Reactor coolant system pressure at Ginna Station is indicatel i by one wide range (0-3000 psig) and four narrow range (1700-2500 psig) instrumentation channels. Wide range pressure transmitter PT-420 is powered from -Instrument Bus ID 'ria Twinco MQ-400.

A loss of of fsite power will therefore re 'ti t in loss of RCS pressure indication below 1700 psig. To pt Ide uninterrupted RCS wide range pressure indication in the event of loss of offsite -

power, PT-420A will be utilized to provide an input to a. spare

, pen on the_ Main Control Board pressurizer pressure recorder.

! PT-420A receives its power from. Instrument Bus 1C and will remain j operational during loss of offsite power. This modification L will~ require the installation of a shielded instrumentation cable from Foxboro Spec. 200 Rack #2 to the Main Control Board.

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

! 1.70. .The events related to this modification are (1) major and minor fires and (2) a seismic event. New wiring and cable i will be. required for this modification, 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 requirenents.

Because of this. there will be no increase of fire loading caused 4

by this modification. This modification is designated not Seismic Category.I, however, any new cable and conduit shall be installed such that'it will not impact any safety related systems during '

, a seismic event. This modification has been reviewed to ensure

. that failure of the recorder or electrical cable installed as

~

a part of.this modification will not result in disabling of

r e=l ,.

~1982 Annual Report of Fac'ility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket'No. 50-244 Page 14 of 30 vi.tal equipment needed to safely shut down the plant during postulated fires or a seismic event.

The Plant Operations Review Committee - performed a Safety Evaluation and determined there were no unreviewed safety questions -

or Technical Specification changes required.

EWR-3323 REDUCTION OF FIRE -HAZARD IN THE VICINITY OF THE

. DIESEL DRIVEN FIRE PUMP AND OIL TANK IN THE SCREEN HOUSE ~

.This modification involves existing curb surrounding the area' occupied by the Diesel' Driven Fire Pump and - the Diesel Fuel Supply Tank. In order' to , assure rapid and certain removal of. split diesel fuel from the area, a second (redundant) floor drain will be-installed and the existing concrete floor surface will be . regraded to provide positive pitch to both floor drains . .

, The. top of the existing curb will be raised to maintain the volumetric capacity within the curbed area to 11% of the contents

- .of the diesel supply tank. A letter dated February 6, 1981

frcan the United States Nuclear Regulatory Comission (NRC) requires l- protection of unprotected structural steel to assure structural l integrity in the event of a fire in the area of the four service

. water pumps.' The only source of a fire large enough to threaten structural integrity is the diesel fuel. By assuring prompt.

I ' removal of any diesel fuel spilled by any rupture by. any rupture ef the tank of piping, the size and duration of any possible fire will not be sufficient to threaten structural integrity of the building superstructure. The proposedJadded drain,-and improved drainage along .with the ' existing buried holding tank will . provide a passive first line of defense to assure structural

' integrity. in the event of a fire and/or spill of the diesel fuel in the Screen House. The defense indepth consists of an existing automatic early detection system, an existing automatic suppression system, and a hose line for fire brigade use.

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. The event related to this modification are the Seismic Event and the Fire Event. A letter dated February 6, 1981 from - the USNRC requires protection of unprotected structural steel to assure structural integrity in the event of a fire in the' area of the four service water pumps. However, the only

source of a fire large enough to threaten structural integrity is the diesel fuel. The existing buried holding tank has a capacity of at least l110% of the voltane of the diesel fuel tank.

,_a._ 2_-.,_..~.__._._.____.._._.___.____._.. _ . .. _ _ .. ~ _ _

o' .'

1982 Annual Report of Facility Changes, T e s t's , and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 15 of 30 By assuring prompt removal of any diesel fuel spilled by any rupture of the tank or piping, the size and duration of any possible fire will not be sufficient to threaten structural integrity of the building superstructure. The addition of a second ( redundant) floor drain will speed and/or assure removal of spilt oil and thus improve the fire safety of the plant.

Regarding (pitching) the existing concrete floor will also speed removal and eliminate any standing puddles of spilt oil and thus improve the fire safety of the plant. This modification is passive and not designed to perform any safety function, therefore, safety during normal operations and transient conditions anticipated during the life of the plant will not be affected.

The structural integrity of the curb and the concrete fill will be maintained during a seismic event. Thus the adequacy of structures, systems and components provided for the prevention of accidents and the mitigation of the consequences of accidents will not be reduced.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specificacion changes required.

TSR 80-20 INVERTER TROUBLE ANNUNCIATOR There is presently no annunciator for transfer of instrument buses lA and 1C from inverter power (normal feed) to the alter-nate feed from MCC'S 1C and 1D. The transfer is now indicated by a computer printout directing the operator to investigate trouble with the inverter. This situation is satisfactory for normal "on line" low computer information output periods.

But during "off line" and computer shutdown periods inverter transfers are often overlooked. This modification involves the addition of a relay to monitor transfer of instrument buses lA and 1C to their alternative seeurce of power and provides for annunciation and computer printout for any inverter trouble. A reset pushbutton will be installed to clear the alarm state once the problem has been corrected. Presently the common alarm point for the inverter ties into the plant computer and signals any off normal condition with the inverter.

In the proposed scheme the common alarm point for the inverter (the normally closed contact with the inverter operating) is in a series with the DC relay RL-6 (the RL-7 for B inverter annunciator) which will be energized when the inverter is operating and with an aux contact from relay RL-6 (RL-7) which

0* .i*

1982 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 30 is used as a seal-in contact. Relay RL-6 ~(RL-7) aux contacts will be used for the computer inputc, for the control room annunciator and for the seal-in contact. in parallel with the reset pushbutton. The annunciator and computer relay aux contacts will close if the inverter alarm circuit loses power. The pushbutton contact is normally open with the inverter operating and is depressed after the inverter trips and after the alarm condition clears to reset the alarm.

Depressing the pushbutton energizes the.RL-6 (RL-7) relay coil and seals.the relay coil in by the seal in contact in parallel with the pushbutton. A IN-4007 diode is installed across the RL-6 (RL-7) relay to protect the coil when it is deenergized.

None of the . events analyzed in Ginna FSAR and listed.in tables 1 and 2 of A-303, Preparation, Review and Approval of Safety Analysis for Minor Modifications or Special Tests, will be affected by the installation of this modification. This modification does not change, (1) the assumptions in any safety -

analysis in the FSAR, . (2) the probability of occurrence of accident,-

(3) the consequences of an- accident. The margins of safety during normal operation 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 prevention of accidents and for mitigation of the consequences of accidents is unchanged.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

TSR 80-21 PERMANENT FEED FOR CONTROL ROOM DATA TERMINAL-This modification involves installation of an 8 pair #20 cable from Westinghouse radiation monitoring cabinet located on the East . wall of the control room to the computer room leased line modem equipmenc. The data terminal in the control room is presently connected to the leased line modem in the computer room through unused spares in the #2 Incore cabinet. The 8 pair #20 cable provides a link to the communication line for the data terminal equipnent in the control room.

None o f the events analyzed in Ginna FSAR and listed in

-tables 1 and 2 of procedure A-303, Preparation, Review and Approval

-- o f Sa fety Analysis for Minor Modifications or Special Tests, will be affected by the installation of this modification.

. 1

..: . .. l 1982 Annual Report of Facility Changes, Tests, and Experiments Conducted l Without Prior Approval  !

R. E. Ginna Nuclear' Power Plant Unit No. 1 Docket No.-50-244 Page 17 of 30 This modification does not change (1) the assumptions in any f safety. analysis in the FSAR (2) the probability. of occurrence of an accident (3) the consequences of an accident. The. margins

- of ' safety during - normal operation and during transient conditions anticipated during the life of the station will be unchanged by the . installation of this modification. 'Ihe adequacy of structures

! systems and components provided for the prevention of accidents and for the mitigation of consequences of accidents is unchanged by installation of this modification.

.The Plant Operations Review Committee performed n Safety Evaluation and determined there were no unreviewed safety questions j or Technical Specification changes required.

TSR 80-23 MORE-RELIABLE 120 VAC FEEDER'FOR THE WESTINGHOUSE INSTRUMENT PANEL IN THE CONTROL ROOM j .

This modification involves providing a more reliable 110 .

VAC Power Feed to the Westinghouse Instrument Cabinet in the-Control Room. This cabinet provides indication of RCP Oil Reservoir. ~

Level, Pressurizer Safety Valve position and houses the- Containment Isolation Relay. Status Panels. The containment isolation relay.

status panels are feed from instranent Bus A, Breaker 15, along with some secondary plant in s tr um'en t s . The AC feed is used

.for illtanination of the relay status lights. .The RCP Oil Reservoir level indication and pressurizer safety valve position indication are presently supplied with 110 VAC from a 120 volt wall plate near the Westinghouse instrument panel which is supplied from

. Breaker #14 of the control room lighting panel . This feed from the. lighting panel will be rertoved and a feed from instrument Bus B. Breaker 16 will be routed to feed the Westinghouse instrument cabinet. (RCP oil reservoir indicator power supply and PZR safety valve position power supply) . The existing feed to the

containment isolation. panels from instrument Bus A will remain as is but the secondary plant instruments will be moved to the feed from instrument Bus B, Breaker 16. The instrument Bus B power source will be used to supply the Westinghouse-instrument cabinet with 120 VAC power. The Westinghouse instrument cabinet provides an indication of pressurize'r safety valve position, RCP Oil Reservoir level and status of the containment isolation

, relays.

The incorporation of this modification will not affect any of the events listed in Tables 'I and II of Ginna Procedure A-303,' Preparation, Review and Approval of Safety Analysis for 1

i

. -. - . _ . , ,.. - ...,. , , - . . - - , - . . . . . . . - , .---..--.w

OO' a a 1982 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 18 of 30 Minor Modifications or Special Tests. Therefore the proposed 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 (3) The consequences of an accident.

In addition none of these events affects any design safety functions of the proposed modification. Based upon these evaluations 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 the accidents and for the mitigation of consequences of accidents is unchanged by the installation of this modification.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

TSR 81-03 INSTALLATION OF RCS PRESSURE INDICATOR BEHIND MCB This modification involves installation of a RCS pressure indicator behind the MCB in the vicinity of the control station for the overpressure protection system and the reactor vessel head vent system. The modification is necessary to provide the operator with an indicator of RCS pressure within line j of sight if these control stations. A new pressure indicator

(0-3000 psi) will be mounted behind the MCB. This indicator will be continuously energized.

None of the events.in Tables I and II of procedure A-303, Preparation, Review and Approval of Safety Analysis for Minor Modifications or Special Tests, changes (1) The assumptions in any safety in the FSAR and its supplements (2) The probability of occurrence of an accident (3) The consequences of an accident.

In addition none of the events of tables I and II affects any design safety functions of the proposed modification. Based upon these evaluations: a. The margins of safety during normal operations and 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 l for the prevention of accidents and for mitigation of the l consequences of accidents will be unchanged by the installation of this modification.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

, - . _ . , - ~ .- ,- - .

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s 1982 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 19 of 30 TSR 82-01 REPLACEMENT OF HYDROGEN TEMPERATURE MONITORING RECORDER FOR THE TURBINE / GENERATOR This modification involves replacing the existing turbine generator hydrogen cooling temperature monitoring recorder in the rear of the main control board with a newer model recorder.

The new recorder is required because (1) it is compatible with

! the data logger:being installed in the ~ RTD control cabinet in the turbine _ building intermediate level during the 1982 outage.

And'(2) the new system is capable of monitoring the additional generator RTD's installed during the 1981 outage. The new recorder has a 0-10 VDC input from the data logger whereas the old recorder input came directly from'the RTDs.

The recorder is used for i plotting selected outputs from the data logger.

None of.the events listed in Tables I and II 'o f A-3 03, j Preparation, Review and Approval of Safety Analysis for Minor

Modifications or Special Tests, are related to this modification.

Therefore, this modification does 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 new recorder has no control or safety function

! so will not be affected by any of the design basis events listed in Tables I and II o f A-303, Preparation, Review and Approval of Safety Analysis for Minor Modifications or Special Tests.

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 i

for the prevention of accidents and for the mitigation of the consequences of an, accident will be unchanged by the installation of this modification.

The Plant Operations Review Committee performed a Safety-Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

l-l ST-82.1 HOLDING CURRENT CHECK OF GOULD J13 SERIES RELAYS IN THE CONTAINMENT ISOLATION RELAY PANELS AND SPARES IN THE STOCKROOM

-A special test is to be performed on each J13, DC relay currently in use on the Containment Isolation and the Undervoltage Relay Racks, to assure that they meet their original design requirements. Specifically, the relay coils are to be tested

e.* ,,

i 1982 Annual Report of Facility. Changes, Tests, and Experiments Conducted Without Prior Approval 1

R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 20 of 30 to insure that they are within tolerance. The relays will be

" insit_u". te sted using their actual de voltage source. The test

-will be conducted when the equipment is removed from : service. That is, the Containment Isolation racks will be tested during the 1982 Refueling shutdown while the Undervoltage Racks will 'be tested. prior to being placed in service.

. There are no identifiable events requiring analysis due to the performance of this special testing program, since~all the . testing will be done with the plant at cold shutdown. There is, . however, a requirement to insure that the equipment function properly after it is returned to service. Assurance of proper

. equipnent operability shall be obtained prior to it being returned-to service- by performing either an acceptance test' or a visual inspection. The acceptance test will only be required if the

, existing control wiring is disconnected and reconnected in the performance of the test. A visual inspection can be performed, in lieu of an acceptance test, providing no wiring changes are made. The visual inspection shall, however, insure that no test leads are inadvertently left in the rack and/or that all fuses are properly replaced. The acceptance test or visual inspection must be performed after all the individual J13 relays

have been tested.

l Therefore, it has, by this analysis, been determined that the margins of safety during normal operations and transient conditions during the test period have not been affected.

The-Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

1 I

l ST-82.3 "B" STEAM GENERATOR GAS SAMPLING AND PURGING ST-82.3.1 "B" STEAM GENERATOR GAS PURGING This safety analysis involves the Special Test ST-82.3 l and ST-82.3.1 for the purge of the " B" steam generator. The L events related.to this special test are (1) radioactive gas waste systen leak or failure and (2) internal and external events.

~

The temporary piping is routed totally within containment so any leaking gas would be directed to the purge system, which l is also where the piping from the head vent eductor is routed I to. Ventilation monitors would shutdown the containment purge L system if gaseous activity approached ~10% of the Tech Spec I limits. A solenoid valve with remote operation is provided i

l

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1982 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 21 of 30 for remote shutdown of the purge piping. Also the purge system is routed through HEPA and charcoal filters prior to going to the CV vent. Since this special test is a temporary one and conducted during plant cold shutdown, no seismic evaluation is required.-

F 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 i 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.

L The Plant Operations Review Committee reviewed .this special test safety analysis,- determined it was the most practical means of controlled release of the gas- space in the "B" steam generator including discussion of the alternative method of release to the closed containment and piping to the gas decay tanks and recommended. approval.

t ST-82.4 SAMPLING REACTOR HEAD VENT EFFLUENT AND VENTING TO PRT This safety analysis involves the Special Test ST-82.4 for the sampling and venting of the reactor head vent. The events related to this special test are (1) radioactive ~ gas waste system leak or failure and (2) internal and external events.

-The temporary piping is' routed. totally within containment _so any leaking gas would be directed to the purge system, which l is also.wher_e the piping from the head vent eductor is routed to.- Ventilation monitors would shutdown the containment purge

-system if gaseous activity approached ~10% of the Tech Spec limits. Also the purge system is routed through HEPA and charcoal filters prior to going to the CV vent. Since this special test is a. temporary one and conducted during plant cold shutdown, no seismic evaluation is required.

l .There fore - the margins of . safety during normal operation l~ 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 l for the mitigation of the consequences of accidents will be '

L unchanged by the performance of this special test.

The Plant Operations Review Committee reviewed this special

[ test safety analysis, determined there were no unreviewed safety questions or Tech Spec changes required.

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r . , _ _ _ - - . . , . . . , _ _ _ _ _ _ . . _ . _ _ _ _ - _ _ _ _ . _ . . . . _ . . - - - . _ . _ _ _ _ _ __

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1982 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 30 ST-82.5 SPECIAL TEST OF T AVE REDUCTION 4

The purpose of this analysis is.to determine whether a i proposed test of reducing T ave by 15 degrees F while maintaining '

I system pressure at 2235 psig would be bounded by previous accident analysis and thus does not present any unreviewed safety questions..

- It has :been postulated that decreasing primary. system. temperature

_will decrease the rate of corrosion of steam generator tubes.

However, it is not clear what the effect of a 15 degrees .F reduction on the primary system temperature will have on the electrical output of the-plant. .To determine this'effect, .a test will be conducted where T ave is reduced by 15 degrees F while maintaining the. normal system pressure of 2235 psig. 'Ihe test will incorporate manual rod con h ol but no changes to.any reactor trip, safety i injection or other setpoints that would affect safeguards actuation.

This safety analysis is separated into three parts.

I - a. A determination that the resulting~ core neutronics parameters are bounded by previous analysis.

b. A review of all events analyzed in the Ginna FSAR

to determine the af fect of the temperature reduction l

upon results.

c. A review of potential differences in ECCS performance. l l-

! Exxon Nuclear calculated the difference in key neutronics parameters for Cycle 12 assuming a 15 degrees F reduction in.T. ave and system pressure reduced to 2000 psia. While the test conditions-l do not include a reduction in pressure, these calculations provide

! a good estimate of parameters under test conditions because

- the principle effect is due to the temperature reduction. A reduction in temperature results in an increase in the critical

boron concentration by about 20 ppm and a slightly more positive s

moderator temperature coef ficient by about ' 2 pcm/ degree F.

L . With a mid cycle 12 MTC in the range' of -15 to -20 pcm/ degree

~ F, the MTC will remain well within the bounds of the transient analysis. Also indicated is that the temperature reductions

[ ' will change the doppler coefficient and boron worth at BOC and

, EOC slightly, but again they will remain well within previously assumed values. Total.-rod worth increases ~at BOC by 60 pcm and decreases at EOC by 42 pcm. The net effect at mid cycle i would be negligible. Based upon these minimal changes to the i neutronics parameters, it is concluded that the previous assumptions g, e y,.,-. s w =-- =,,,,e,re-ne-ym--g-pems ,mn.-,-,,,ewear-*+=rw-+ern ves-s+rt<--------iw e -+e -'* *w=.=*eerw-*-e *ee- *--t'---*'-+**-+-w-w "-e-----* ** -me--w-

1982 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 30 used in the transient analysis are bounding.

A review of all events analyzed in the Ginna FSAR and subsequent referenced safety analyses has been made. Listed below is a compilation of a11 ' events .that could be affected by the change in. primary system conditions and the reactor trip function which ~

provides core and coolant boundary protection. Please note that only those transients which are assumed to be initiated from a hot full power condition are being considered. Operation of the reactor at other than a hot fu'.1 power, reduced temperature condition is not being considered for this test.

a. Rod Withdrawl - Fast Nuclear Over Power

- Slow Over Temperature T

• b. Loss of Flow Low Flow i.

• c. Locked Rotor Low Flow

• d. Loss of Load High Przr. Pressure

e. Loss of Feedwater Low S/G Level or Steam Flow-Feed Flow mismatch

f. Excessive Heat Removal Low Przr. Pressure l due to Feedwater Temp.

j decrease i g. Excessive Load Increase Low Przr. Pressure

h. Steam Generator Tube Rupture Safety Injection

• i. Steam Line Break at Power Safety Injection -

• Limiting events for DNB In general, a lower T ave will result in a greater margin to DNB. Those events limited by a trip function not related to system temperature should not be affected adversely by the test.

This would include events b through g. There are other events where the impacts are negligible. The water mass transfer from the primary to secondary during a steam generator tube rupture would increase due to the larger pressure differential. However, the increased pressure difference would only be approximately 90 psi as compared to a normal pressure dif ference of about 1500 psi. Since the mass transfer is dependent upon the square

4 1982 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 Pa Je 24 of. 30 root of the pressure difference the effect_ of the increase'd differential is negligible. The steam line break at power is bounded by the event at HZP. The larger steam generator water inventory at HZP condition being the principal reason. Also-the 15 degrees F reduction results in a more positive MTC which for cooldown transient from HFP makes the consequences less

severe. The remaining events are the fast and slow rod withdrawl

at power. These events are terminated by the nuclear overpower j and .overtemperature delta T trip . functions respectively. The nuclear overpower trip can be affected by the temperature reduction because the higher density water decreases the. neutron fl ux at the detector for any power level. To insure that the difference between detected and actual nuclear power is minimized , the t

, procedures for the test provide for secondary side calometrics L and adjustment of indicated nuclear power while the temperature is being reduced. This procedure, along with the DNB benefit from the lower temperature will insure that the analyzed. event is bounding. Slow rod withdrawl is terminated by the overtemperature.

delta -T trip function. - This is a relatively slow acting transient

, with the trip occurring at-17.7 seconds and a MDNBR of 1 73

! when initiated from normal temperature and pressure. The reduced [

temperature would change progress of the event to some degree, I however the overtemperature delta T trip function will protect

' the core from DNB where the event is slow as compared to piping delays (4 second s) . Combined with a lower initial T ave this insures that the~ analyzed event is bounding.

The impact of the reduced temperature upon the ECCS/LOCA evaluation has been reviewed. In this review, the Exxon Nuclear ECCS/.LOCA-analysis at a 50 degrees F reduction in' T ave and a system pressure of 2000 psia was relied upon to provide the rationale as to why the acceptance criteria of 10 CFR 50.46 would be met. The-Exxon Nuclear report provides results for ECCS/LOCA calculations at nominal (573.5 degrees F and 2250 psia) and reduced (523.5 degrees F and 2000 psia) conditions.

The lower temperature case is limiting because the controlling phenomena are the LOCA blowdown pressure transient (determined by saturation conditions) and the initial primary fluid energy.

The resulting effect is an increase in the amount of water lost and a lower containment backpressure. Because saturation pressure and fluid enthalpy vary smoothly and continuously between the two conditions analyzed, the two results bound the results expected from operating conditions in between. For this reason, a LOCA initiated from operating conditions reflecting a 15 degrees F reduction in T ave would meet the acceptance criteria of 10

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1 1982 Annual Report of Facility Changes, Tests, and Experiments Conducted '

Without Prior Approva1

R..E. Ginna Nuclear Power Plant l .

Unit No. l' j

[ Docket No. 50-244  :

Page - 25 of 30

! CFR 50.46.

I . There fore, the review of the FSAR transients and .the ECCS/LOCA i- analysis concludes that.the proposed test would result in no i

unreviewed safety questions.

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

i i ST-1444-82.1 BUS 17 UND;;RVOLTAGE CABINET SYSTEM ENERGIZATION TEST ST-1444-82.2 BUS 18 UNDERVOLTAGE CABINET 4 SYSTEM ENERGIZATION TEST ST-1444-82.4 BUS 14 UNDERVOLTAGE CABINET

! SYSTEM ENERGIZATION TEST 4

i ST-1444-82.5' BUS.16 UNDERVOLTAGE CABINET

] . SYSTEM ENERGIZATION TEST This analysis covers the special tests for the connection of AC and DC sources to the undervoltage cabinets. This will allow operation of the equipment prior to connection of the plant circuits-to the undervoltage tripping relays.

. 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. The only events related to this special-test and to be reviewed is the loss of the AC or DC sources listed belows

a. Instrument Bus lA

b. Instrinnent bus 1C

c. Screen House DC -Panel 1A

d. Screen House DC Panel 1B

e. Screen House Lighting Panel

f. Main DC Distribution Panel 1A

g. Main DC Distribution Panel 1B

-h. Auxiliary Building DC Distribution Panel 1A

1. Auxiliary Building DC Distribution Panel 1B

- j. . Auxiliary Building Lighting Panel AB-2

. k. Auxiliary Building Lighting Panel AB-3 The undervoltage system is in the preoperational stage and no o tripping relay-has been connected to the plant system at this

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1982 Annual Report of Facility Changes,  !

, Tests, and Experiments Conducted

Without Prior Approval i

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

, Docket No. 50-244 Page 26 of '30 -

1 time. The AC and DC power feeds to the cabinets are however, connected and are within the scope of this test. . Since proper i

fuse and breaker sizing and coordination protection has been-established, faults and short circuits, due to workman error i or.in improper installation, will not' affect availability - o f l the AC or DC system.

i

.. Therefore it has been determined by this. analysis that 1

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3 the margin of safety during normal operation and transient conditions F during the test period have not been affected.

! The Plant Operations Review Committee performed a Safety

Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

{

ST-1444-82. 7 '- BUS 14 UNDERVOLTAGE CABINET l

GOULD J13 SERIES. RELAYS

~

ST-1444-82.8 BUS.16 UNDERVOLTAGE CABINET GOULD J13 SERIES RELAYS i ST-1444-82.9 BUS 17 UNDERVOLTAGE CABINET

! GOULD J13 SERIES RELAYS h ST-1444-82.10 BUS 18 UNDERVOLTAGE CABINET GOULD J13 SERIES RELAYS A special test is to be performed on each J13, DC relay

[ ' currently in. use on the Containment Isolation and the Undervoltage t-Relay Racks, to assure that they meet their original design requirements. Specifically, the-relay coils ~are to b'e tested  !

to insure that they are within tolerance. . The relays will be f

~

"insitu" tested using their actual dc voltage source. The test' will be. conducted when the eq'uipment is_ removed from service.

That is, the Containment Isolation racks will be tested during i

the.1982 Refueling shutdown while the Undervoltage Racks will be tested prior to being placed in service. ~

! There are no identifiable events requiring analysis due

, to the performance of this special testing ~ program, since all l the testing will be done with the plant at cold shutdown. There is, however, a requirement to insure that the equipnent function

. properly after it is returned to service. Assurance of proper equipnent operability shall be obtained prior to' it being returned i

to' service by performing either an acceptance test or a visual

- inspection. The acceptance test will only be required if the t

o3v .. y r -- e.-,, y . ,%.. ,# ,g, w.,*m,,w,. ,ei.,,,,,..#, ,.-,. .,m-, ----._~r- ,--r.- - - . , - - - . . - ,,-,.-w,.---.,,,.---..4 --.-.s+.-

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, 1982 Annual Report of Facility Changes, t Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 '

Docket No. 50-244 Page 27 of 30 existing control wiring is disconnected and reconnected in the 4

performance of the test. A visual inspection can be performed, in lieu of an acceptance test, providing no wiring changes are made. The visual inspection shall, however, insure.that no test leads are inadvertently left in the rack and/or that all.

fuses are properly replaced. The acceptance test or visual inspection must be performed after all the individual J13 relays i have been tested.

, - Therefore, it has, by this analysis, been determined that the margins of safety during normal operations and transient ,

-conditions during the test period have not been affected.

~

The Plant Operations Review Committee performed a Safety Evaluation and determined there were no unreviewed safety questions or Technical Specification changes required.

ST-1444-82.12 BUS-16 UNDERVOLTAGE CABINET ACCEPTANCE TEC.T (FUNCTIONAL)

ST-1444-82.13 BUS 17 UNDERVOLTAGE CABINET ACCEPTANCE TEST (FUNCTIONAL)

ST-1444-82.14 BUS 14 UNDERVOLTAGE CABINET

_ ACCEPTANCE TEST (FUNCTIONAL) l ST-1444-82.15 BUS 18 UNDERVOLTAGE CABINET t ACCEPTANCE TEST (FUNCTIONAL) f l' ST-1444-82.16 BUS 17 UNDERVOLTAGE CABINET FUNCTIONAL TEST l

ST-1444-82.17 BUS 16 UNDERVOLTAGE CAUINET FUNCTIONAL TEST I

ST-1444-82.18 BUS 18 UNDERVOLTAGE CABINET p FUNCTIONAL TEST ST-1444-82.19 BUS 14 UNDERVOLTAGE CABINET i

FUNCTIONAL TEST I- In 1975 the NRC sta f f requested that RG&E assess Ginna Stations susceptibility of its safety-related electrical equipment

[ to (1) a sustained degraded voltage condition at the offsite l

. s '. , ,

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1982. 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 28 of 30 1 and onsite power sources and (2) interaction between the of fsite

~

and onsite emergency power sources. This undervoltage modification-is the result of this request. That is, a second . level of under-  :

voltage protection will be added to existing loss of voltage system in compliance with the NRC . staff position. This-second

• level system will detect a degraded voltage condition and correct it by placing the safety related loads onto the onsite diesel  !

generators. In addition, the existing loss ~ of . voltage system will be upgraded so as to reduce its susceptibility to single. .

component failures which results in a system failure. Greater  !

redundancy will be provided through coincident logic in compliance

l. with the NRC staff position.

A review has been made of all events analyzed in- the Ginna ,

Station FSAR and the. events requiring analysis by NRC Regulatory I Guide 1.70. The events relating to -this undervoltage modification f ares (1) loss of AC power to station' auxiliaries,-(2) events o f . fire , flood, storm or earthquake, (3) loss of one DC systemz i

and (4) major and minor fires. The first event that has been.

analyzed for this modification are the consequences of a complete

. loss of all AC power and/or a degraded l AC power source, .hile w the reactor is at power. The second level. modification is designed

!' to detect and correct a degraded offsite system . voltage. A degraded voltage is that voltage at which some safety related equipment =

could malfunction and yet be sufficiently high to prevent the loss of voltage system from functioning. The degraded. voltage

levels are defined a letter to the NRC from Mr. L. D. White dated July 21, 1977. The second level system will consist of 1

primary and backup undervoltage monitors and auxiliary relays on each of the four safeguards buses. Once added to the existing loss of voltage-system, both systems will afford protection

. against all voltage abnormalities. The use of coincident logic will upgrade the existing loss of voltage systems tolerance

, o f single component failure by providing greater redundancy.

Thus the consequences of future safeguards equipment " lock outs" are greatly reduced. The load sequencers are presently initiated j' by the undervoltage relays on all four safeguards buses. With

] the addition o f the second level relays, the reliability of 1

-the sequencer start logic using only buses 14 and 16 will be

, at least- equal to that of the existing system using all four buses.- This logic simplification is acceptable because the service water. pumps will be started whether or not an SI signal is present. This insures that the diesel generators will have adequate cooling. Seismic category I standards will be imposed

• on. components of this modification thus mitigating the consequences 4

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s ,s 1982 Annual Report of Facility Changes, N ' Tests, and Experiments Conducted T. 3

. _ _ _ Without Prior Approval R. E. Ginne We'.: lear Power Pla.it . -

, Unit No. 10 ,

w . Docket No F50-244 8

,;- ' . :n A,

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Page 29 of 30

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of eckulpment failure durinror-after a seismic event.

The d.c. cour ,

trol voltage for eachitrain will be modified so that it .is f6d' = ' ,

frors a preferred hattery ' source and will une the second lE battery -

system as a bachup source. Presently each train has ity-primary and, backup und,ervoltage systems fed from a,aj nglo- bate ury system. '

'the primary and proposed backup de sourc,e will be-fedithrough and isolated by a throwover contactor assembly. Th6lcontactor is designed so that a' failure of any component or dystem' will not. cause the two dc. sources to be parallele.d. In bddition

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proper fuse coordina'tioni ensures that either system is not made'.

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'C inoperable due to a fault on undervoltage entiipment. The use '

of throwover contactors in the Class 17,480 volt' swi,tchg ea r -

hns'bden evaluated and a report issued to thF USNRC which was foun'd acceptable. ' The proposed throwover corjtyctors willJunction -

in -the same manner thus ensuring improved operebility of each und 9 rv ol't ag e system as well as the entire E.3F A S-( eng ine n r'ed ~

safety,Jcatures actuation system) system. In add 1 tion, it will insurci116 proved _ system operability during the loss of the preferred de source! Chrrently the buses will operate s on loss of de control power since the throwover will supply the'~ backup de source.

The underyoltace systems that monitor the lE 480 volt buses 1 will not operate when de control power is lost as presently '

con figured . 'Thg. criginal deggn intended was to maintain dedicated- ,

de sources to'each system and maintain de separation. This separation can '.s_tlll' be ma'iritained through a th,rowover contactor ,%

y T'1e- modification sdoes not increase and give improved operability.

the possibility'or impa'e t, o f- a' Il re . Additio ria l wiring and y

cable will be adopd inthis. modification,<which_could add to x' the fire loadi ng .~o f ths ; plant . ' Therefore the. Design Criteria requires that all such;gabl.e,me Qft the IESE 383-1974? flame' test N requirements. Because\of s thi.s,there will be no, increase of fire loading caused by thi'sJmodification. j Trte: Design Criteria ,

provides requirements togpreserve any silicon'e foam fire stop or seal that need to be penetrated.s Both the second level modif-ication and the upgraded 'locs of voltage systeni will be designed to meet IEEE 279-1971 and'IEEE 384-1974 to'the extent practical given present plant configuration. Cable routing'and equipment layout for this modification will be analyzed for physical separation acceptability in accordance with correspondence regarding the Fire Protection Safety Evaluation Report.

Therefore, based upon the above analyses, it has been detennined that (a) the margins of safety during normal operations and transient conditions anticipated during the life of the station are not reduced and (b) the structures, systems and components

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Tests, and Experiments Conducted r

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• R. E. Ginna Nuclear Power Plant

.f# A, . Unit.No. 1

[,. ],j Docket No. 50-244

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.N 'provided "fo'r~thef y$evention of acc'idents and the mitigation

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of the copa~equenceu~ of accidents are adequate.

ThexPlant Operations Peview Committee reviewed this special k  ?. test ..na fe ty sanelysis, determined there were no unreviewed safety *

' questions o'r Tech Spec changes required. .

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CHALLENGES TO THE \ PRIMARY SYSTEM PORV'S AND SAFETY VALVES g-- ,e l',Ny Tr. compilduce with "NUREG 0737 Commitments and Ginna Station '

4* cx k'{%.f"F'to k

j 2procedure lhe7 primary system" PORV's occurred during 1982. On January 5, .1982, following O-973 the -Stc,am NRC Imediate Generator Tube Rupture Dotification, Incident, . the fo kg

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the pressurizerfpower-opereted relief valve PCV-430 was opened g ym: ,at 1007:30 hours 'for approximately ' 3 second ,, at 1007:49. hours R, N1' *N \7for seconds,approximately and at 1009:10 6 seconds, at 1008:44 hours for.approximately hours for less than 5 seconds. Two h,3 h .h2),second= after an attempt to close the PORV following'the M ' ,7ppening at 1009:10 hours, the PORV reopened and block valve '

"'-MOV-516 was completely closed approximately 40 seconds later.

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i CHALLENGES 1TO?THE SECONDARY SYSTEM PORV'S AND SAFETY VALVES

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,~ In, com'pliance with NUREG 0737 Commitments and Ginna Station p ' Q p~rocedure 0-9.3,' NRC Imediate Notification, the following challenges 5,3 Lp M, cro ithe , secondary system Safety Valves occurred during 1982.

, f "~ i "On' January 25, 1982, following the Steam Generator Tube Rupture 6%_ TI6cident,"there was indication that the B Steam Generator safety l

" N tvalve' lif' 2d at 1019 hours0.0118 days <br />0.283 hours <br />0.00168 weeks <br />3.877295e-4 months <br />, at 1028 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91154e-4 months <br />, and at 1937 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.370285e-4 months <br />.

' At 1040 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.9572e-4 months <br />, the A Steam Generator power operated relief valve was throttled open. At 1137 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.326285e-4 months <br /> there was again an indication

. [ ;that the B. Steam Generator safety valve lifted. '

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ROCHE;TER GAS AND ELECTRIC CORPORATION

• 89 EAST AVENUE, ROCHESTER, N.Y.14649 *d

.)OHN E. PAAIER YttteMONE v.c. m.ws.nc *=racootna 546 2700 May 20, 1983 Mr. James M. Allan, Acting Regional Administrator U. S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia,- PA 19406

Subject:

1982 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 Mr. Allan:

Transmitted herewith is the submittal of the Annual Report of-Facility Changes, Tests, and Experiments Conducted Without -

l Prior Approval as required by 10 CFR 50.59 and Challenges to

! the Primary and Secondary PORV's and Safety Valves. This report l

I

! is for the period of January 1, 1982 through December 31, 1982

. inclusive.

Very Truly Yours, D k hlMab J

Jo n E. Maier Attachment xc: Document Control Desk (1) m l

2~C2-F

_ . --