ML20084N086

Annual Rept of Facility Changes,Tests & Experiments Conducted W/O Prior Approval & Challenges to Primary & Secondary Sys PORVs & Safety Valves,1981

ML20084N086
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 8306020451
Download: ML20084N086 (22)

Text

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.. .a 1981 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS CONDUCTED WITHOUT PRIOR APPROVAL AND CHALLENGES TO THE PRIMARY AND SECONDARY SYSTEM PORV'S AND SAFETY VALVES 8306020451 830520 , sj ,

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i 1981 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 1 of 20 GSM-1 ADDITION TO AUXILIARY BUILDING A need has been identified for a_ standby auxiliary feedwater system ~at Ginna Station (GAI Repo rt' 1815). _Because space for

-this equipment is not available in existing plant structures,

'the existing: Auxiliary Building will be extended. -The Auxiliary.

Building Addition as shown on GAI Layout Drawing D-0E-001 i s- ~

a structure adjoining the south wall of the existing Auxiliary.

Building; For the purposes of this design criteria the Auxiliary-Building Addition is' considered to consist of two. independent structures as follows:

a.- Auxiliary Feedwater Pumphouse

-The Pumphouse is a' Seismic Class I concrete structure supported:by caissons. The building is presently one story high, but will be designed to accommodate the addition of a second story at)some future time.

For -the present design the second story is considered to be_a light' steel structure. The Pumphouse will contain major equipnent such as a 10,000 gallon condensate storage tank and standby auxiliary feedwater pumps, etc.

b. -Drum Storage Building (FUTURE)

The Drum Storage Building is a non-safety-related steel frame structure supported by spread footings.

The building is one story high and in located on'the east side of the Pumphouse. The Drum Storage Building will contain the cement silo and new drum storage area. The footings will be designed to span excavation

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required to serve the electric lines that run under the new footings. The Drum Storage Building shall have an easily excavated floor, such as asphalt' paving,

'to permit access to the buried electric line.

The Primary function of the Auxiliary Building Addition is to house:and provide a suitable operat'ing environment for the proposed standby Auxiliary Feedwater System. The function of the standby auxiliary feedwater system is to decrease the consequences of

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an accident, namely the consequences of a high energy break, which could conceivable damage the existing auxiliary feedwater

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1981'Annuni Report of:Fccility Changen, 2 Tests,_andJExperiments Conducted  !

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R.cE.'Ginna Nuclear Power Plant-Unit No. 1, j Docket No. 50-244 i Page 2 of 20 '

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system in the JInte'rmediate Building. During the construction. I of the foundation for the Auxiliary Building Addition, work _

must be-performed in the vicinity:of existing Class I utilities buried under the ground. These utilities consists. of the two 20"=._ diameter service _ water : lines , the 115 kV electrical' lines,-

and the 'electricalJduct bank for off-site power .and . centrol cables. Of the above mentioned utilities, the only one that could conceivably be damaged during the construction of the Auxiliary Building Addition 11s the service water pipe. The 115 kV lines-are enclosed in an oil static pipeline and are

- buried ' eight feet -balow grade and the electrical duct bank is under a concrete slab. In . the case ' of the service water- pipe, -

the reason the.new redundant return line was-installed before construction of.' the Auxiliary Building Addition 'is to provide -

redundancy-in. case the existing lines are damaged during con-ctruction. 'None of the above mentioned utilities'will be exposed while heavy construction lifts are being made. In addition ^, ,

the following safety precautions:will be taken during-caisson installation to protect'the Class I utilities:

1. Temporary casings will be installed around the upper portion of the two caissons nearest the-service water pipe thrust-block and the caisson nearest-the 115 kV line. These casings will be twisted into place to prevent loss of ground during.

installation. The casings in.the caissons ~ nearest the thrust block will be left in place.

2. The end of the service water thrust block has been coated : *

-in the field to assure that there i s clearance b e t w'e e'n it and the caisson.

i l 3.s An elevation monitoring point has'been established on the top of the service water thrust block to verify that no settlement takes place during construction.

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! 4. The caisson installation specification calls for the use of equipment which will not cause excessive vibrations. -

i 5. A monitoring point has been established for the new service 1 water redundant return line at the Auxiliary Building wall j

to verify that settlement does not occur at that' point.

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s ~ 1981 Annual Raport of Facility Changes, .

Tests, and Experiments Conducted- >

Without Prior Approval-R. E. Ginna Nuclear Power Plant Unit No. 1 Docket-No. 50-244 Page 3 of 20

6. The 115'kV line has been exposed and its location recorded

-to assure.that there is adequate clearance between it and

'the nearest caisson.

7. Other less critical known ? utilities have been located and -

.those that interfere with caisson construction have been relocated.

The building-and equipment electrical grounding grid will'be extended to include the Auxiliary Building Addition. Where construction interrupts this grid , . a temporary connection will be made until the pemanent ' connection has been made. The Auxiliary Building Addition' is considered to be isolated from the existing.

building as far as fire hazard is concerned and any_ openings cut-in:the existing wall'will be block _ed'with fire retarding material'. A hose station will' be established south of the building for construction fire protection. Fire extinguishers will also be placed at strategic. locations as required by' OSHA. During the construction of the'_ Auxiliary Building Addition, precautions

.'will be taken ~ to assure that a crane does not swing into_the existing Auxiliary Building. These precautions will consist of:

1. Safety Watch - A person knowledgeable about crane perfomance will supervise the crane operations and check lifting capaci--

ties, swing, position, etc.

2. Operator Indoctrination . Operators will be advised of the existing plant conditions and the critical nature of their work. ,

3. Safety Features of Cranes - Special attention will be _given to assure that contractor cranes meet OSHA requirements, for example: cable condition, posted load chart in cab, boom angle indicator, etc.

4.- Crane Position -~Whenever possible, cranes will be positioned at an angle to the existing building which would put the

. boom'somewhat parallel to the existing building duriug lifting and no loads will be carried over the existing building.

The cooling system- for the Auxiliary Building Addition is a redundant, Seismic Category I system. Electric power for it

,x 1981 Annual Raport of Facility Changes, Tests, and Experiments Conducted Without Prior Approval

'R. E.-Ginna-Nuclear Power Plant Unit'No."1 Docket No'.-l 50-244 -

Page 4 of 20 ccmes' fran the safety related Class lE electrical system.. Cooling water comes from an extension of_the safety related plant service

. water system into the Auxiliary Building Addition. .The heating-system for the Auxiliary Building. Addition-is not seismically classified and does not use Class 1 E power,:since its. operation-is not essential ' for . proper operation' of the standby auxiliary feedwater pumps. Failure of. the _ system's unit heater's supports would not damage any safety related equipment in the room.

After reviewing the documents stated.in the-Design Criteria, and'the. construction procedures as stated above, it'has_been

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determined that the proposed-modification does not involve an unreviewed safety question. Therefore, the proposed Auxiliary Building Addition is safe because:

'l. The probability of occurrence or the consequences of an accident or malfunction of equipment or structures important to safety previously evaluated in the safety analysis report-will not be increased.

-2. The possibility for. an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created.

While. referred to as an Auxiliary Building Additioni the new structure housing .the - standby auxiliary-feedwater pumps is ini reality a completely independent structure.- Therefore, a l l' normal and abnormal loads applied to-thisonew structure will-not be transmitted to any existing building.- The standby auxiliary feedwater pump building is'a~. Class I structure and as such'will be. designed to withstand the effects'of both the Operating Basis Earthquake and the Safe Shutdown Earthquake, and the effects of short term tornado loadings, including the impact of tornado-generated missiles. Therefore, this new Class I str6cture does not have a higher probability of failure than any of the ' existing Class I structures. Similarly, the Auxiliary Building - Addition cooling and heating system is completely independent of the cooling and heating system in the existing auxiliary building.

As such, there is no interaction between the new system and the existing system.

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

1981' Annual Raport of Facility Changen,.

Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant

--Unit No. 1 Docket No.-50-244 Page 5 of 20 GSM-16 STANDBY AUXILIARY FEEDWATER SYSTEM The. sole purpose of the standby auxiliary feedwater system is to provide auxiliary feedwater backup in the event of a high energy pipe break. It is not intended for other plant trips or. transients. The standby auxiliary feedwater system shall be . capable of being brought' into -service by operator action in the Main Control Room (MCR). The system will be activated _

by the operator if the existing auxiliary feedwater pumps , which start automatically, are 'not operative. The system flow diagram Lis shown on GAI Drawing D-302-071. Two Seismic Class I (as defined in the FSAR) sources of-water shall be available for emergency use by the standby auxiliary feedwater system via connections 'to both loops of the service water system. Service water is intended to be the cooling medium used in an actual emergency situation. In addition, a supply of condensate shall be available for periodic tests of the system. The standby auxiliary feedwater system shall include two motor driven pumps,

-each of which.will' deliver emergency feedwater to a separate steam generator. In addition, the system piping shall provide the capability for the pumps to deliver-water to either steam generator via a crossover with remotely operated valves between the two pump discharge lines.

The Plant Op_erations Review Committee performed a_ Safety Evaluation and determined there . were no . unreviewed safety questions .

however Technical Specification changes were required. ,

EWR-1837 COMPUTER ROOM CEILING, FIRE-DOORS, AND CONTROL ROOM KITCHEN This modification consists of replacing fourteen existing hollow metal doors, frames and hardware with fire door assemblies that carry the underwriters approved labels. This may be either an A or B label depending upon the fire hazard. This will fulfill our commitment to the NRC given in GAI Report No. 1936, 4.2-11, 4.3-3, 4.3-7, 4.3-8, 4.3-9, 4.4-4, 4.4-9, 4.4.5-2, 4.8-3, 4.9-2.

In areas.that are considered to be extreme fire hazards a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> protection class A fire door assembly is required. In areas where a severe fire hazard exist a 1-1/2 hour class B door assembly is required. All fire doors must be self-closing or close automatically in the event of a fire. All fire doors must be self-latching and obviously remain closed in case of fire. All hinge jambs and door leaves must have label attached.

1981' Annual Report of Focility Chcnges, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 6 of 20-Some doors have electric security locks. A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis.by the USNRC Regulatory guide, 1.70.

The events related 'to this modification are the Fire Event and Seismic Event. None of the existing hollow metal doors with fire rated door assemblies will be designed with no degradation in the function of the seismic systems and its components.-

In addition, failure of the doors in a seismic event shall not result in damage to safety related equipment and the s tipport s for the doors are designed'such that the doors will not damage safety related equipment as a result of 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-2462A REACTOR COOLANT PUMP OIL COLLECTION SYSTEM SEAL BYPASS LINE RELOCATION This modification consists of the relocation of the " B" Reactor-Coolant Pump #1 seal bypass line. This relocation is'necessary to permit the installation of the lower Level Control . oil coll;ection system enclosure. The installation of the enclosure is necessary to reduce the potential of an oil fire as described in Design Criteria EWR 2462. The purpose of the seal bypass connection is to proved a sufficient flow of water across the reactor coolant pump bearing to cool it until such time as a satisfactory flow rate is established across the No. 1 seal during pump startup.

The seal bypass connection also serves as a means to vent air out of the No. 1 seal cavity. A review has been made of all the events analyzed in the Ginna Station FSAR and the events requiring analysis by NRC Regulatory Guide 1.70. The events related to this modification are (1) earthquake, (2) Cnemical and volume Control System Mal f unc tions', (3) loss of reactor coolant flow, and (4 ) primary system pipe rupture. The modification is designed as Seismic Category I. Therefore, the consequences of an earthquake are not affected. The Chemical and Volume Control System is not being changed functionally since the seal injection water flow path remains the same. Therefore the mod-ification does not affect the probability of consequences o f a CVCS malfunction. The modification serves to inc'rease the reliability of the Reactor Coolant Pumps and since it decreases the probability of a RCP failure it wi:1 have no affect on the consequences of a loss of Reactor Coolant flow. The modified

1981: Annual kaport of Facility Changes, Tests,.and Experiments Conducted Without Prior Approval R. E..Ginna Nuclear. Power Plant Unit No. 1 Docket'No. 50-244 Page.7.of;20 piping ~ does not ~ increase the probability of a ' loss of coolant-accident since the modification will meet or exceed the criteria to which the present line was installed. The size of the piping (3/4") does, however, fall within~ the envelope of pipe breaks ~

which would not preclude a L aafe shutdown. . Therefore, the mod--

ification does not-change the consequences of a primary system -

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

-It-has also beenidetermined 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-2605 DIVERSE CONTAINMENT ISOLATION A letter from Harold Denton of the NRC, dated. October 30,=1979 requires in Section 2.1.4 Position'4 that: The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will:not result in .

the automatic reopening of containment isolation valves shall require deliberate operator action. And that: Resetting of containment 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 position. To further reduce.the likelihood inadvertent reopening

. o f. valves , a' system modification will be desinned to provide for individual resetting of all isolation valves to eliminate-any possibility of.an inadvertent opening. The modification will provide an additional reset requirement, for each individual valve, in the ' following ways (refer to sketch 1) Upon receiving -

a containment isolation signal existing relay C pickslup and opens normally closed contact C. This in turn removes power fran, and de-energizes 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 room. Upon pushing the PB from X & Y relays are energized. Additionally contacts from each. relay also close (Rl-X1 and Rl-Y1 for example) sealing in'the relays until another C.I. signal. Additionally contacts

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1981 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 20 from the relay air are used to light visual indicators behind the pushbutton so that relay actuation can be verified.from the control room. A third set of contacts are used to return the C.I. signal back to the existing logic, one X-Y pair for each existing C. I . contact. Relay contact pair Rl-X3, -Y3 showing

.the wiring for a normally ' closed contact and relay contact pair

-R2-X3, -Y3 showing the wiring for.a normally open contact.

A review has - been made of all events analyzed in the Ginna Station FSAR and ' the events requiring analysis by USNRC Regulator Guide 1.70. The events.related to this modification are: ( 1 ) --maj or and minor fires, (2) seismic events, (3) loss of all a.c. power to the station auxiliaries, (4) fuel handling accidents, (5) primary system pipe rupture, (6) events leading to high containment

-pressure, (7) radioactive release inside containment, and (8) inadvertent opening of a pressurizer safety or relief valve.

' The modification does not increase the possibility or impact i of a fire. Additional wiring and cable will be added in 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 thsre will be no increase of fire loading caused by this modification. 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 n Criteria, it is determined that this modification shall not affect the consequences of a loss of all a.c. power to the Station auxiliaries, in that it does not use a.c. power for control.

Because this modification is considered seismic category I, a seismic event 'will not impact this modification or cause this modification to impact any system in an adverse way. The remainder of the events listed in Section 3.1.all are effected by this modification in hat 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 existing plant equipnent. Section 20.0 requires that separations be maintained./ Section 21.0 requires that upon failure of

a. component that containment isolation is not degraded. Section 22.0 requires that the modification shall be tested prior to use. Section 6.0 requires compliance with IEEE' standards 336, 344, 383 and 384. The containment isolation function and initiating signals will be unchanged. The modification impacts only the reppening of containment isolation valves following a containment isolation signal reset. Deliberate operator action will now l l

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1981 Annual Rnport of Facility Changas, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 .,

Page 9 of 20 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.not been affected. .It has also been determined that the' adequacy of structures, systems, and components provided for-the prevention of accidents and the mitigation of the consequences of accidents have not been affected.

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

EWR-2608C ELECTRICAL PENETRATION INSTALLATION NUREG 0578 and NRC letter dated October 30, 1979 require that RG&E: Provide by January 1, 1981, two radiation monitor systems in containment. These new monitors will require special cables.

Spare NIS instrumentation cables shall be used and replacement cables for NIS use installed in a timely manner. To do this new electrical penetration assemblies shall be installed. This modification is for the installation of three new containment pene'trations to add the necessary coaxial cables and also provide spare instrumentation, control and power. penetrations that are needed. The 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. The events related to this modification are (1) major and minor fires, (2) seismic event, and (3) the spectrum of accidents inside of containment. The modification does not increase the possibility or impact of:a fire. Additional wiring and cable will be added in this modification, which could add to the fire loading of the plant. Therefore the Design Criteria

~c 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 modification. This modification has been classified seismic class I and electrical class'IE and therefore shall be designed to have no impact on the plant after an SSE. This modification is required by the Design Criteria to meet IEEE 317 and ASME BPVC III subsection NE and therefore is. qualified to not deteriorate and also to . fully function through the spectrum of accidents inside of containment . Therefore it shall have no impact on the plant during or after these accidents.

It has, there fo re , been determined that the margins of safety during normal operations and transient conditions anticipated i-

i ' 1981 ' Annual R3 port of Facility Changes, Tests, and Experiments Conducted Without Prior Approval

• • R. E. Ginna Nuclear' Power Plant.

Unit'No. 1-Docket No. 50-244-t Page 10 of 26 during. the life of the station have ' not been af fected. It has also been determined that the adequacy of structures, systems,

i. and componentsJprovided for the prevention of accidents and the mitigation of the' consequences of accidents have not been

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affected.

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

- or Technical Specification changes required.

EWR-2843 CONDENSATE STORAGE TANK INSTRUMENTATION The existing level indication and alarm circuitry for the Condensate Storage Tanks consists of.a single differential pressure transmitter, two remove vertical scale. level indicators, a high l ev.el alarm, and a low lev el alarm .Section X.4.3.2-1 of RG&E responses to NRC's recommendations-for Auxiliary Feedwater Systems dated November 28, 1979 and-May 22, 1980- rec. aire redundant level indica- -

tions and low level alarms in the conurol' room for the Condensate -

Storage tanks by January 1, 1981. This modification is intended to fulfill those requirements . The 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 Room. 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 be replaced with

a dual indicator having completely independent inputs.for each

!. transmitter.- The existing high and low' level alarms ~are input to a single annunciator window. These inputs will be replaced with the new redundant high and low level alarm circuitry. In addition,'both channels will also be monitored by the plant computer,: which has its own independent power supply. See attached P&ID 03021-353, Rev. 2. This design provides functional redundancy o f Condens' ate Storage Tank level' indication and high and-low

- level alarm all the way from the differential pressure transmitters to the indicators and alarms, including their' power supplies.

A review hs been made of all events analyzed in the Ginna FSAR

, and the events required by NRC Regulatory guide 1.70. The events related to this modification are: (1) earthquake, (2) loss of one D.C. System, (3) loss of A.C. power to station auxili-aries. The modification is not required to be seismically designed since.the condensate storage tanks are not seismically designed.

The modification will'be designed such that, in the event of an earthquake, it will not damage safety related equipment.

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1981 Annual Rsport of Fccility Changen, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 11 of 20 Each of the two level indication channels is powered by a separate D.C. bus, therefore, loss of one D.C. bus will not cause loss of condensed storage low level . alarm. Loss of A.C. power will have no effect on this modification. Therefore, the margins of safety during normal operations and transient conditions anticipated during the life of the plant are not decreased.

The structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents are not adversely affected and are adequate.

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

e EWR-2928 FIREPROOFING STEEL COLUMNS TURBINE LUBE OIL RESERVOIR AREA A GAI fire protection study determined that structural steel columns in the Turbine Tube oil reservoir area should be fire protected. The. extend of these modifications are outlined in the GAI study titled, " Fire Protection Study for Controlling Structural Failures Jeopardizing Cold Shutdown of Ginna Station,"

dated May 27, 1980, revised September 16, 1980. See attached sketch A for general arrangement. The fire protection as outlined in section 1.2 of this Design Criteria will allow Rochester Gas & Electric Co. to meet the Ginna Station Fire Protection Safety Evaluation Report (SER) ite.n 3.2.8 Docket No. 50-244, dated February 14, 1979 by the U.S. Nuclear Regulatory Commission.

The application of fireproofing to all structural steel columns within ten feet of the oil reservoir dike, in conjunction with automation of the existing manual fire suppression system will preclude structural failure during a fire. Automation of the existing manual fire suppression system is not included within the scope of this Design Criteria. The only event related to this modification is fire. The 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 dete rmined 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. The evaluation has been submitted to NRC, however the NRC has not confirmed that this modification is adequate to resolve SER item 3.2.8. The addition of fireproofing to columns within the ten foot area of the oil reservoir dikes will improve the fire safety of the

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1981 Annual Roport of Facility Chr.nges, Tests, and Experimento Conducted

^

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

Unit No. 1-Docket No. 50-244 Page 12 of 20 I_ 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 Plant Operations - Review Committee performed a Safety Evaluation and .ietermined there were no unreviewed safety questions

, or Technical Specification changes required.

EWR-2947 CONTROL BUILDING FLOODING

, The design will eliminate possible flooding of the two battery rooms caused by a - crack break in the service water line located 2

in the mechanical equipment room. A watertight wall will replace the door between the mechanical equipment room and battery room, therefore, flooding will be contained within the -mechanical equipment room. A gravity drain system will be installed between the turbine and control buildings to control the flood level in the mechanical equipment room. The concrete and the concrete block walls in the mechanical equipment room will contain flooding -

within this room. A drain system will-remove the water from-i the room to the turbine building. Within the drainnsystem a l fire . damper will maintain the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire rating of the concrete

. block wall and a pipe flapper will withstand the pressurization

(- o f the turbine hall. The drain system will remove the water

flow equal to the' water . flow from the postulated crack break. .

L 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. The events related to this modification are: -(1) a seismic event, (2) pipe break outside containment, (3) fire and'(4) flooding of the mechanical equipment room. The modification will be designed for a seismic event, turbine building pressurization i resulting from a high energy line pipe break, fire protection '

and flooding of the mechanical equipment room. In the event of turbine building. pressurization resulting from a high energy pipe break in the turbine building, and non-return valve has o boon incorporated in the drain system design to prevent pres- ~

l surization of the' mechanical equipment room. A fusible link actuated fire damper in the drain system will maintain the fire rating of the wall containing the drain system penetration.

Seismic design shall be in accordance with the seismic design F~ of the control building pressurization wall . The gravity drain system of the mechanical equipment room is designed to prevent level buildup of water from developing a hydrostatic pressure 4

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1981 Annuni Raport of Facility Changas, Tests, and Experiments Conducted

.Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 13 of 20 that could " fall" -the wall between the battery and mechanical equipment rooms. In the case of. turbine building pressurization.

and concurrent flooding- of the mechanical equipment . room,- a realistic pressure blast would not significantly hinder the~

- drain of service water.from the mechanical equipment room.

The proposed modification does'not change any assumptions in the safety analysis written in'the FSAR or its supplements.

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

- also been determined that the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the; consequences of accidents have not been affected.

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

EWR-3025 MOTOR DRIVEN AUXILIARY FEEDWATER PUMP DEFEAT CIRCUIT MODIFICATION The proposed design would modify existing circuitry installed on the Auxiliary Feedwater Pump Flow Controis under EWR 1431 to include a lockout tunction during periods of plant ~ startup-and shutdown. Specifically, the lockout switch that now isolates

- only the second auxiliary relay (MFPX2) which closes the control valve in the bypass loop, will be modified to lock out the primary relay as well. The existing lockout switches, labeled LO MPFX2 1A1 and LO MFPX2 1B1, will be wired in series with the feed pump 1A and 1B breaker back contacts (FPlA, FPlB). In addition a contact off the turbine auto stop-relay (63 XT 3) will be paralleled with the lockout contacts which will bypass the lockout once the turbine stop valve trip is latched. This feature can be used since the turbine stop valve is latched when at least one main feed pump is in service. A time delay is provided for opening this contact to assure an auxiliary feed pump start

' signal following a turbine trip. The proposed lockout feature will~not inhibit the automatic start signal to the motor driven reeJwater pumps in the event of a SI signal, a low-low steam genera *.or level. The need for this modification is due to the E

fact tiat when the plant is in either a startup or a shutdown mode, both main feedwater pump circuit breakers are opened which i

l i

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1981 Annual Rsport of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 14 of 20 generates an automatir: start signal _to the motor driven auxiliary feedwater pumps. The motor driven auxiliary feedwater pumps are used in both ON and OFF modes to adjust the steam generator ,

water level. During these conditions, the temperature of the primary side is above 350 degrees F.and existing Technical Speci-fications prevent the placement of these pumps in the pull to lock position, a position which~ renders them inoperable. Existing plant procedures require during startup or shutdown conditions, that the sliding link terminals on the main feedwater breakers be " opened" blocking the auto start cignal generated when both main feedwater pump breakers are opened. This allows the operator to use the motor driven auxiliary feedwater pumps and the bypass ~

loop installed as part of EWR 1431 without the presence.of the auto start signal. This is consistent with the technical specif-ications since the MDAFW Breakers are operational with the sliding link terminals opened and an auto start will occur on low-low steam generator terminal level or SI. This modification' eliminates the need for terminal block " opening" and " closing" during normal plant operating conditions. The basic function of the lockout switch is to inhibit the auto start signal to the motor driven auxiliary feedwater pumps generated only by the main feedpumps being taken out of service. In addition, the lockout prevents energizing the MFPX2 auxiliary relay. The purpose for introducing the shunt contact off tP turbine stop valve latch (auto stop) auxiliary relay is to e .rride the lockout during conditions when at least one main feed pump is operational. A time delay is provided to ensure that, during a condition that results in the loss of both main feed water pump breakers (e.g. loss of #11 transformer) combined with the lockout switch inadvertantly left in the defeat mode, the auto start signal to the motor driven auxiliary feedwater pump breakers is not blocked. This modirication will be accomplished by making wiring changes in existing control circuitry. A comparison of the safety bypass features of the proposed scheme to the existing scheme will be made and will serve to show the advantages of this modification:

Existing Control Schemes

a. During plant startup or shutdown conditions the LO MAFPX lockout switch must be placed in the " defeat" mode and the sliding link terminals of the main FW breaker placed in the "open" positions. The bypass loop can then be made operational and the auto start signal to the MDAFP is inhi-bited. The bypass loop is only closed by an SI signal.

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1981 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 15 of 20 The motor driven auxiliary pumps will receive an auto start signal on low-low steam generator level or on SI.

b. During plant conditions when at lest one n.ain feedwater pump breaker is closed, the lockout owitch can be placed in the " normal" position. The sliding link terminal can then i1 " closed". The bypass loop is still operaticnal and w be closed along with the initiation of the auto start nal to the MDAFP if a SI signal is received, or a low isv steam generator level signal is received on both main FW breakers trip.

Proposed Control Scheme

a. With the lockout switch wired in is new circuit location, the switch is still placed in the " defeat" mode during startup or shutdown conditions. However, opening the sliding link terminals to block the auto start signal to the MDAFP breakers will not be necessary. The bypass loop is made operational. An SI or low-low steam generator level signal will close the loop and initiate an auto start signal to he motor driven auxiliary feedwater pump breakers.

b. With the lockout switch in the " normal" mode and one main FW breaker closed, the bypass loop will be operational and will be isolated along with the initiation of an auto start signal on low-low steam generator level, SI or if both main FW breakers are opened or are tripped.

c. The primary advantage of the proposed scheme is that no terminal opening is required during start or shutdown condi-tions. The operator must, however, place the lockout switch in the normal mode to clear the annunciator. No signal will be inhibited when i is required to mitigate an accident.

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. Since this modification only invo2ves the relocation of an existing lockout switch, the only events that require evaluation are the loss of normal feedwater and fires and earthquake.

The effects of the loss of normal feedwater floor to the steam generators due to the introduction of a lockout of the auto start signal to the motor driven Auxiliary Feedwater Breakers has been analyzed. The existing system used a lockout switch

l'981 Annual Raport of. Facility Changea, Tests, and Experiments ' Conducted Without Prior Approval R. E .- --Ginna Nuclear Power Plant -

Unit.No. 1 Docket No. 50-244 Page 16 of 20 to inhibit an auxiliary relay thus allowing the . auxiliary bypass

. loop to'be made operational. Without 'the . lockout . feature, . the bypass loop would be locked . out during . plant startup, shutdown or any condition with both main feedwater purnps out of service. The

' lockout switch allows the bypass loop to be made operational and is required to be in the " defeat" position until one main

. feed pump breaker is closed. In addition, this modification

-will prevent n auto start signal to the MDAFP breakers during conditions when only one auxiliary pump :is required. This lockout, l

-in its new location,.will make the bypass loop operational during  :

startup or shutdown conditions. 'The defeat switch will not. l l inhibit an ' auto start signal to the MDAFP breakers due Tto low-low  !

e steam generator levels or an SI signal. When the defeat switch l is in the normal position, the auto start _ signal will also be l generated on-the loss of both m.ain feedwater pump breakers. -l j'

These automatic signals are consistent with the existing plant l operating practices. A contact off a turbine auto stop relay i will .bei paralleled with the. lockout contact. This feature will 4

autanatically inhibit the lockout whenever the turbine is -latched. -

The automatic inhibit and administrative control of the lockout i switch provide adequate assurance that the AFW start signal l . from_the feedwater pump breakers will be reinstated during power operation. The plant response'to a loss of normal'feedwater-l event during-power operation will remain unchanged. The loss l of feedwater analysis assumes that the transient begins' at low ' ow l steam generator level, a condition which, along with both feedwater f

pump breakers opening, produces both a' reactor-trip and a signal ,

to start the auxiliary feedwater pumps. The AFW automatic start-signals will . remain unaffected by, he m'odification during power

~

operation and.thus the analysis assumptions ~ and conclusions remain valid. A loss of feedwate_r event initiated after'after an orderly power reduction and defeat of the "feedwater pumps tripped' automatic start signal will.have a negligible impact on the plant because other automatic start signals will initiate AW flow. The criteria for the modification has imposed appropriate requirements on the modification so that neither a fire -or seismic event will have a deleterious effect upon the auxiliary feedwater aystem. Therefore, the margins of safety during normal operations

! and transient conditions anticipated during the life of the plant have not reduced. The adequacy of structures, systems, and components provided for the prevention of accidents and l

for the mitigation of the consequences of accidents'have not been affected.  ;

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_. _ . __ _ _ m 1981 Annual Raport of Facility-Changes, Tests, and Experiments Conducted-Without Prior. Approval R. E. Ginna Nuclear Power Plant Unit.No. 1 Docket.No. 50-244 Page 17 of 20 4

The Plant _Oper_ations Review Committee performed a Safety

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

TSR 79-09 DIESEL GENERATOR TACHOMETER INSTALLATICN' This modification involves the installation of a permanent' c=tmsha'ft driven tachometer for reading diesel generator rpm. The energency diesel generator overspeed rip setpoint is presently checked with a hand held tachometer. This method is not very accurate.

This modification does not change-(1) the assumptions in any safety analysis in the FSAR and its supplements (2) the probability of occurences and~(3)~the' consequences of an accident. None of the events analyzed in Ginna FSAR and listed ~in Tables I and'II of A-303, Preparation, Review and.Approva1 of~ Safety -

Analysis for Minor Modifications or Special Tests, will be affected by installation of this mod. The margins of safety during_ normal operation and transient conditions anticipated during the life of the station will be unchanged by insta,llation of this. modification. The adequacy of structures, systems and components provided for prevention of accidents and mitigation o f consequences of accidents is unchanged by. installation of i-this modification.

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

TSR 80-04 INSTALLATION OF ADDITIONAL ISOLATION VALVE ON DRAIN LINE FROM "AT CVCS HOLDUP TANK The 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. The above mentioned modification is to provide fo r transferring reactor

[ cavity: leakage which has accumulated in Sump "A: to the "A" l CVCS Holdup Tank for reprocessing. The additional isolation j valve is in the "A" CVCS HUT drain to the sump tank, between the existing elbow and the floor. A review has been made of all' events analyzed in the Ginna Station FSAR and NRC Regulatory Guide No. 1.70. None of these evens will be af fected by this

[ modification. Therefore, the margin of safety during normal-operations and transient conditions anticipated during the life l

1981 Annual Raport of Facility Changes, Tests, and' Experiments Conducted Without Prior Approval

-P. E. Ginna Nuclear Power Plant Unit No. 1-Docket No. 50-244 Page 18 of'20~

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

system, and components provided for the prevention of accidents and for 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.-

TSR 81-07 SAFETY INJECTION LOW FLOW INDICATIONS This design is for the installation of 'a low flow indication for the Sa f ety Inj ection System. It consists of installing-necessary pipes, valves, fittings flow indicator, hangers and anchors to provide low flow indication in the Safety Injection Recirculation .Line . The above modification is to provide for measurement of leakage past Safety Injection Check Valves.

A review of all events analyzed in the Ginna Station FSAR and events requiring analysis by USNRC Reg.-Guide 1.70 has been made. None of these events are related to the -ir.atallation o f Low Flow Indication on the Safety Inj ection Recirculation

~

Line. 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 structure, systems

=and components provided for the prevention ~ of accidents and for the mitigation of he consequences of accidents have not been affected.

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

ST 80-1 ACOUSTIC EMISSION MONITORING OF FEEDWATER ELBOWS This special test procedure is for the special test of the main feedwater elbows . It consists of attaching acoustic emission detectors to the feedwater elbows - inside containment of both steam generators and installing cable from the detectors to penetration CE-33 to a computer located in the Turbine Building.

This test is designed to provide continuous on-line monitoring o f the feedwater elbows and elbow to nozzle welds during plant start up and power changes.

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

r:

. 1 1981 Annecl Report of Facility Changea, Tests, and Experiments Conducted Without' Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 19 of 20

- Guide 1.70. The events related to the special test are:

1. Internal and-External Events - Fire, Flood, Storm or Earthquake and 2.- Spectrum of Postulated Steam and Feedwater Piping Breaks.

This first event includes major and minor fires, floods, storms or. earthquakes. The proposed special test includes installation of UL approved and flame retardant tested cable from the feedwater elbows to penetration CE-33. As such,- this test and cable instal-lation will not increase the consequences of. this event. For the second event considered, the proposed special test installations are attached only to the surface of the feedwater elbow and due to their weight (approximately 5.5 lbs.) versus the' weight of the feedwater elbow ( approximately 300 -lbs.) , these attachments

~

do not significantly increase the seismic loading of the feedwater pipe in this area. As such, the consequences of this event are not increased by this special test.

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

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

ST el-1 SPECIAL TEST OF NUMANCO DRUMMING UNIT This analysis covers the special test ST 81-1 for the NUMANCO Drumming Unit. This drumming unit is a self contained device for the solidification and drumming of waste evaporator bottoms generated during the waste evaporator operation. ' This device is skid mounted and will be anchored to the operating floor of the Auxiliary Building.

A review of the events in Tables I and II of A-303, Preparation, Review and Approval of Safety Analysis for Minor Modifications

~

or Special Tests, and the events requiring analysis per Regulatory Guide 1.70 has been made. The events related to this.special test are: 1. Radioactive liquid sy' stem leak or failure and

2. Internal and external events - fire, flood, storm or earthquake.

With regard to radioactive liquid waste system leak or failure, the consequences of this event are not increased by

' the special test'because the new system will interface with the liquid waste system in similar fashion to the existing system.

I , - .

1981 Annual Rr, port of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant Unit No. 1 Docket No. 50-244 Page 20 of 20 All liquid waste carrying components will withstand maximum internal pressure and will be secured to the NUMANCO drumming unit (anchored) and the liquid waste disposal system. They will not be routed in proximity to safeguard equipment. With regard to internal and external events - fire, floods, storms, or earthquake, the proposed special test neither penetrates any existing fire barriers nor does it af fect any existing fire suppression systems. The special test does not increase any previously determined fire loadings. The special test neither affects nor is affected by any flood or storm previously evaluated.

The consequences of an earthquake event are not increased by this special test because the installation will be designed to withstand a seismic event as defined in the Ginna Station FSAR using the equivalent static load method.

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

The Plant Operations Review Committee performed a Safety _

Evaluation and determined there were no unreviewed safety questions

~

or Technical Specification changes required.

l l CHALLENGES TO THE PRIMARY SYSTEM PORV'S AND SAFETY VALVES L

l In compliance with NUREG 0737 Commitments and Ginna Station

! procedure O-9.3, NRC Immediate Notification, no challenges to l the primary system PORV's and Safety Valves occurred during i 1981.

i l

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

In compliance with NUREG 0737 Commitments and Ginna Station procedurc O-9. 3, NRC Immediate Notification, no challenges to

the secondary system PORV's and Safety Valves occurred during

1981.

l 1

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, ,s, ROCHESTER GAS AND ELECTRIC CORPORATION

• 89 EAST AVENUE, ROCHESTER, N.Y.14649 <

JOHf4 E M AIE R Y E L E P** DN E v-._, pren.rs ne 4 , a corr *,s 546 2700 May 20, 1983 Mr. James M. Allan, Acting Regional 7dministrator U. S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406

Subject:

1981 Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval R. E. Ginna Nuclear Power Plant, Unit No. I Docket No. 50-244

Dear Mr. Allan:

Transmitted herewith is the submittal of the Annual Report of Facility Changes, Tests, and Experiments Conducted Without Prior Approval as required by 10 CFR 50.59 and Challenges to the Primary and Secondary PORV's and Safety Valves. This report is for the period of January 1, 1981 through December 31, 1981 inclusive.

Very Truly Yours,

., j b% ct John E. Maier Attachment xc: Document Control Desk (1) gY