ML20095A707

Forwards Response to NRC Safety Evaluation of Conformance W/ Station Blackout Rule (10CFR50.63),transmitted Via NRC 890418 ltr.Atmospheric-relief Valve Operation for Station Blackout Will Not Be Required Based on Design Basis

ML20095A707
Person / Time
Site:	Sequoyah
Issue date:	04/16/1992
From:	Joshua Wilson TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M68603, TAC-M68604, NUDOCS 9204200166
Download: ML20095A707 (49)

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April 16, 1992 U.S. Nuclear Regulatory Commission ATIN:

Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of

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Docket Nos. 50-327 Tennessee Valley Authority

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50-328 SEQUOYAH NUCudAR PLANT (SQN) - UNITS 1 AND 2 - RESPONSE TO NRC SAFETY EVALUATION (SE) ON THE CONFORMANCE WITH THE STATION BLACKOUT (SBO) RULE (10 CFR.50.63)

References:

1.

NRC letter to TVA dated Janaury l'4, 1992, " Station Blackout Analysis - Sequoyah Nuci+ar Plant, Units 1-and 2 (TAC NOS. M68603 and M68604)"

2.

TVA. letter to NRC dated April 18, 1989, "TVA's Station Blackout (SBO) Evaluation Results Pursuant to 10 CFR 50.63 for the Browns Ferry and Sequoyah Nuclear Planus" This letter provides the requested response to the Reference 1 NRC

-letter, which transmitted the SE regarding SQN's compliance to the SB0 Rule 10 CFR 50.63.

By Reference 1, NRC concluded that SQN Units 1 and 2 cannot be considered to.be in complete conformance with 10 CFR 50.63.

NRC requested that TVA submit a revised response for SQN no later than April.16, 1992. NRC requested that TVA address each of the staff's recommendations for SQN to resolve the identified nonconformances. The issue of SQN conformance to the SB0 rule remains open until the acceptable resolution of NRC's concerns are completed. contains TVA's response to each of the staff's recommendations in Reference 1. provides the commitments contained in this submittal.

The commitments contained in this letter supersede SQN Commitments Nos. 2 and 5 in Reference 2.

The other commitments in Reference 2 are not affected by this submittal.

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PDR. ADOCK 05000327

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v U.S. Nuclear Regulatory Commission Page 2 April 16, 1992 Please direct questions concerning this issue to Keith C. Weller at (615) 843-7527.

Sincerely,

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1 9 b(WLs IL. Wilson Enclosures cc (Enclosures):

Mr. D. E. LaBarge, Project Manager U.S.-Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike-Rockville, Maryland 20852 NRC. Resident' Inspector

-Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy. Tennessee 37379

-Mr.-B. A. Wilson, Project Chief U.S, Nuclear Regulatory Commission

_ Region II 101 Marietta Street, NW,-Suite 2900 Atlanta, Georgia 30323 N

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A ENCLOSURE 1 Response to NRC's' Station Blackout (SBO)

Analysis for Sequoyah Nuclear Plant (SQN)

The following_provides TVA's response to each of NRC's requests and/or recommendations contained in the January 14, 1992, letter and attached safety evaluation.

1.

NRC Recommendation:

"The licensee should submit the battery capacity calculation and-identify the loads that will be shed.

The battery capacity verification and any resulting modification or procedure changes should be included in the documentation supporting the SB0 submittals that is to be maintained by the licensee."

Response

The 125-volt (V) Vital batteries are evaluated for the 4-hour SB0 event-in Calculation SQN-SBO-001.

The 250-V station battery has been evaluated for the 4-hour event in Calculation SQN-CPS-031.

Both calculations consider loads to be shed during an SB0 event and these. loads are listed in Attachment 1.

These loads are manually removed and the ecticas are assumed to be taken at 30 minutes into the event (except-for the main turbine emergency bearing oil pump that requires removal at 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the SB0 event). These lists are preliminary in that revisions are planned as discussed below.

TVA will update these lists by December. 15, 1992, in conjunction with other changes discussed in this enclosure.

The final lists will reflect the loads that will be. maintained such that the operator will have the capability to monitor core conditions and to o

remove residual heat during the 4-hour SB0 event, TVA' evaluated the safe shutdown path available during an SB0 event and developed a list of components and instrumentation that are required to maintain the safe shutdown path utilizing the turbine-driven _ auxiliary feedwater (AFW) pump (see Attachment 3),

j-No components were shed that are necessary to support the shutdown j

path. _ Loads that are removed are not-required. The station batteries are necessary to provide. control-of the switchyard. -The loads removed from these batteries during the SB0 event.are not those required for access to offeite power at the end of_the event.

The loading information, including the load profiles, is in Attachments 2A and 2B for the vital batteries and station batteries,-

respectively.

Operations has' requested.that additional loads remain energized for an SB0 event that are powered by the vital batteries through the vital alternating-current (ac) power system. These loads include one train of reactor vessel level instrumentation and the interplant radio system. This udditional equipment-is not among the minimum

-set of components that are required to mitigate the SB0 event.

However, the Battery Capacity Calculation SQN-SBO-001 will be revised to_ account for the additional loading. Marked-up pages of the calculation have been included in Attachments 2A and 2B to reflect these additional loads.

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It should be noted that Battery Calculation CQN-SB0-001 is not in its final form as indicated by the commitments contained-in this submittal and therefore the_ entire battery calculation will not be submitted to NRC at this time.

Portions of interest are being provided in the' attachment.

The existing calculations are, however, available for NRC review upon request.

2.

NRC Recommendation:

"If the above cited modification is not made to the compressed air system, the licensee should perform a habitability assessment, including the lighting and communication equiptnent, for the areas in which operators need to be to operate the AkVs and the AFW flow

_ control valves."

-Response:

The_ operation of the turbine-driven, AFW pump level control valves has previously been evaluated as requiring a supplemental air supply

-in order to function-during the 4-hour Sb0 event. A new design using fail open valves with on-off control, which do not use a continuous-bleed controller, will be implemented. These valves will open at the initiation of the event and operate a limited number of times during the event. The flow -to and level in the 6 team generators (SGs) are monitored-from=the main controx.oem, w.'4-Ut flow magnitude is manually controlled by varying t..a turb'ne speed to control the SG 1evel. The carability for contro111ag the AFC

_ flow has been tested on the SQN simulator with acceptahlc results.

With this modification to the valves and the procedure revision to the control strategy for the turbine, supplementa' air.c the level control-valves will be reduced substantially from tne orfginal SB0 proposal such that compressed-air bottles or additional accumulators will notLbe required for SBO. No local hen"al clerator nction will be required. Since the mitigation of the S40 ever? will be proceduralized, there will be no additicune campensatory measures required.

Atmospheric-relief valve operation for SB0 will not be required because SQN's design basis is hot standby for safe-shutdown.

Because the rule requires the plant -to be capable of achieving and maintaining safe shutdown, hot standby was the objective. The strategy for the choser SB0 mitigation is based on the use of the safety-relief 1 valves that lift automatically at preset pressures.

The cooldown option using-the atmospheric-relief valves will be available and control'ed

.nually from a nonhostile environment if itLis necessary to. prevent damage to the reactor coolant pump (RCP) seals.

TVA has-the capabi~.ity to cool the reactor in the event of excessive leakage at the RCP seals. This capability _is proceduralized in Emergen, y Contingency Instruction (ECA) 0.0.

ECA 0.0 is a broader emergency procedure than that required for 10 CFR 50.63.

This emergency procedure considers the failure of an RCP seal and the required cooldown that is needed to reduce the outflow of reactor l

coolant by reducing system pressure.

The cooldown is performed by operating atmospheric-relief valves for SGs 1 and 4.

Hand operators (with extension rods) for these valves are located in the 480-V shutdown board rooms near the main control room.

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._3 telephones are available in these rooms for communication to the control room.

In addition, hand-held flashlights are available to.

the operators as well as permanently installed emergency and Appendix R lighting for travel to and operation of these valves.

3.

NRC Recommendation "The licensee shouldt

1) provide a detailed description of the 1

computer code used to perform the heat-up analyses; and 2) ensure that it has considered areas which house SB0 response equipment as areas of concern, including the switchgear room, cable spreading room, inverter room, etc."

Response

The Martin Marietta Interactive Thermal Analysis 9ystem (MITAS),

Version 2.0 (MITAS II), is an improved digital

...ter software system designed to solve the lumped parameter; e.g.,

resistor-capacitor thermal analogue network representations of the physical thermal systems using finite difference techniques.

The models developed to simulate the SB0 conditions assume one-dimensional-heat transfer through each heat flow path. Walls, floors, and ceilings were modeled as series of the diffusion nodes tied together by conductors. Diffusion nodes have the ability to store energy based on_their thermal capacitance. Thest % odes are initialized to normal maximum temperatures before starting the event based on assigned boundary conditions and are allowed to change during the transient. Thus, realistic temperature gradienta are formed through the walls, floors, and ceiling before and during-the event.

Room air'is also modeled as a diffusion node and is tied (via surface conductors) to the walls, floor, and ceiling surface nodes. The SB0 heat loads are applied _directly to the room air node. The forward dif ferencing subroutine is called on in the VARIABLES 1 section of the program to compute-the transient temperatures.

Equipnant required to operate during _ an SB0 has been evaluated and listed in Attachment 3 except as noted in the following paragraphs.

Calculation SQN-SQS2-0077 was issued to determine temperature transients during a 4-hour SBO. event for the main control room, 125-V battery and battery board rooms, 250-V battery and battery board rooms, penetration rooms, and pipe chases.- The conclusion of this calculation was that the equipment in the areas considered was l_

acceptable.

Additionally, the 480-V board rooms that contain the vital inverters should not be. subjected to temperatures above 110 degrees Fahrenheit (F) during'the 4-hour SB0 event. This engineering L

judgement is based en the temperatures measured in the vital battery rooms during the performance of Special. Test ST-7 for reactor coolant system natural circulation during initial plant start-up in which ac power was removed. The only ac power used during the special test was to supply the cooling associated with the RCP

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The temperature in the battery rooms did not increase during the 2-hour test. The battery rooms are adjacent to the 480-V board

. rooms,uand the doors on the battery rooms were opened frequently h.

such that air was exchanged with the battery rooms. - With ac power removed, there will not be any significant heat sources in the areas other than the vital inverters. The vital inverter procurement required them to be designed to operate continuously.in the range of 122 degrees F.

To confirm the engineering judgement, a transient heat balance calculation for the 480-V board rooms will be performed using a computer code with the results documented in Calculation SQN-SB0-001.

The cable spreading room, switchgear room, and turbine buildinF do not contain active equipment required to mitigate the SE0 event.

Since no active SB0 equipment is located in these areas, no analysis is required or occussary.

Even though the cable spreading and switchgear rooms have not been evaluated using a thermal transient analysis, Calculation TI-ECS "100FRSO, Appendix R, lleating, Ventilating, and Air-Conditioning Review," was issued to evaluate the loss of heating, ventilating, and air conditioning (HVAC) for areas required for safe shutdown of the plant.

The cable spreading and switchgear rooms are justified as not requiring ventilation for the 72-hour duration of an Appendix R event. A 4-hour SB0 condition duration with reduced cable heat loading and only emergency lighting is bounded by the TI-ECS-95 analysis.

4.

NRC Recommendation:

"In addition to the detailed description of the computer code discussed in Section 2.2.4 above, the licenree should provide the input parameters-(i.e., initial room temperature, heat loads, etc.)

for the staff to review. The licensee should also establish a procedure to ensure that the control room complex temperature during normal-power operation will not exceed the assumed initial temperature.used in the heat-up cu'.culation."

Response

Before the SBO, temperrtures (including boundary temperatures) were

~ssumed to be at theit t;rmal maximum value given on the applicable environmentalidata. drawings. These temperatures are representative of the maximum normal operating conditions that would occur during the summer months. A steady-state MITAS run was made under these conditions to initialize walls, floor, and ceiling diffusion node temperatures before the start of'the event.

At the start of the SBO, boundary temperatures were instantly increased to their design basis event (DBE) losslof coolant accident (LOCA) values-(again obtained from the environmental data drawings) with the exception of the spreading room that was set to its abnormal maximum temperature.

Use of the DBE LOCA temperatures is conservative since they are steady-state values that take many hours to attain.

These temperatures consider full DBE LOCA heat loads resulting from the safety-related and non-safety-related equipment and cables, many of which will not be energized during an SBO.

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Also, the-DBE LOCA' condition does not take credit for the. operation of any non-safety-related HVAC.

For the spreading room, a DBE LOCA 1

temperature was not listed for this space on Environmental Data Drnwing 47E235-25 as the room is not impacted by the DBE LOCA. As such, the abnormal maximum triperature has been chosen for the cable

-spreading room.

Periodic _ Instruction 0-PI-0PS-000-606.0 is an existing plant instruction that verifies ambient temperatures in critical spaces are within limits and records these temperatures once per shift.

Historical data documented by these reports was used as the basis for establishing the maximum and minimum normal temperatures shown on the Environmental-Data Drawings (47E235 series) that are used in the thermal transient analysis. This procedure contains limits for ambient temperatures and requires actions to notify appropriate-organizations to ensure correction of out-of-lir't conditions.

Since this procedure exists, there is no need to ntablish a new prograx.

5. -NRC Recommendation:

"In addition to the detailed description of the computer code discussed in Section 2.2.4 above, the licensee should provide the information'(see Appendix A to SAIC TER) as requested by the staff's consultant."

Science Applications International Corporation (CAIC) Request:

" Provide information that supports the west valve vault temperature studies bounding an SB0 event with the steam relief."

Response

The following is a summary of the evaluation for this response:

1.

Historical. data recorded during an HVAC failure with. maximum outdoor air temperatures while the plant was in Mode 1 operation listed a maximrm temperature in the west main steam valve vault (MSVV) of '163 degrees F.

2.

TVA'has evaluated the SB0 heat load in the west MSVV against the Modo 1 normal-heat load and concluded that the SB0 load is lower.

Therefore, the maximum temperature in an SB0 event would be-less-than 163 degrees F.

3.

. Extensive environmental response analyses considering main steam line and main feedwater line break temperature profiles in the west MSVV are the~ basis for equipment qualification of

'10 CFR 50.49 components necessary for safe shutdown.

4.

No operator entry into the west MSVV will be required during an SBO.

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The= west MSVV temperature exceeds 120 degrees F and contains components required to mitigate-an SB0 event.

This area is considered by IVA to be a dominant area of concern and is acceptable based on the evaluation contained in SQN-SB0-001 and the discussion above.

-The following provides additional details of this evaluation:

An SB0 event is initiated by a loss of of f sit e power to the nuclear station.with the unit operating at 100 percent power. Regulations and guidelines require event mitigation such that the core damage frequency, associated with an SB0 event, is reduced to approximately 10-per year for the average site.

Part of the event mitigation is to e1sure that the components located inside the building areas will remain nualified to perform their mitigative functions given the event conditions inside the containing building.

In order to ensure that the event mitigation equ3 y nt located within the west MSVV can withstand the environmer effects of an SBO, TVA proposes to utilize the plant-specific experience and the test data.

Comparison to the test data is allowable to provide reasonable assurance of the operability of equipment in accordance with the Nuclear Mancgement and Resources Council (NUMARC) guidelines (Appendix F. Section F.7).

In 1988, the temperature in the west MSVV was monitored with the unit in Mode 1.

During the monitoring period, the non-safety-related ventilation system failed causing temperatures in the valve vault to rise to 163 degrees F in localized areas. This occurred with an outside ambient temperature of 97 degrees F.

TVA then performed an assessment and determined that 10 CFR 50.49 equipment remained operable. Since the test was performed with the unit:in Mode 1, the heat sources _in the room bound those for the SB0 event. Hence, the temperature rise in the west M9VV because of the SB0 event will remain below 163 degrees F.

The following discussions will demonstrate this position.

The primary heat sources _in the west MSVV during Mode 1 operation

-are the main steam and main feedwater piping.

About 2,450 square feet (ft2) of heat transfer surface area is included in these two piping systems within-the west MSVV (main steam, 2,000 ft2; main feedwater, 450 ft2). Although insulated, the heal loss to the region through the-insulation is substantial. When ventilation terminates, the air becomes stagnant and heat becomes entrapped in the room.

Hence, the temperature rises upon failure of the ventilation system. Upon occurrence of an SBO, these two heat sources are reduced because_of valve actions but are partially offset by the addition of other heat sources.

7 The AFW system is-activated following reactor trip'and subsequently generates two.other sources of heat input to the west MSVV.

The turbine-driven AFV supply piping and the turbine-driven AW pump steam-exhaust vent line are also routed through the MSVV.

175ft{he turbine-driven AFW exhaust vent stack includes about of noninsulated piping exposed to the MSVV environment, and the AFW 2

supply contributes approximately 100 ft of noninsulated exposed surface area.

However, since the main steam isolation valves will isolate, the amount of exposed main steam piping will be reduced to 2

approximately 1,500 ft during the SBO. Note that the steam supply line of the turbine-driven AFW pump is normally charged with steam.

This heat source exists during Mode 1 operation and is included with the main steam piping area.

The exhaust piping of the turbine-driven AFW pump does not provide a large heat source to the west MSVV. After passing through the AFW pump turbine, the steam temperature is reduced to 228 degrees F.

This is the maximum temperature that the exhaust piping can reach.

Although noninsulatet, the amount of exposed surface area is relatively small and heat input to the environment is limited.

The temperature difference between the piping surface and MSVV ambient conditions-will be small and result in a low heat flux into the room. Any additional heat supplied by the AFW turbine exhaust -

piping will be offset by the AFW supply piping.

The supply sourc?

for the-AFW (condensate storage tank) is not expected to exceed temperatures of approximately 100 degrees F.

As air temperature rises above-100 degrees F, more heat will be absorbed by the AFW supply piping.

Following reactor trip, the steam system pressure rises.

The power-operated relief valves or the main steam safety valves (MSSVs) are reauired to dissipate the sensible heat of the primary coolant and the residual heat generated by the core.

The power-operated relief valves are assumed inoperable and the steam relief path is through_the MSSV via vent stacks.

Only one-of the five steam-safety valves per SG is required to maintain adequate heat removal; therefore,_two vent stacks will experience exposure to steam temperatures (one stqck each for Loops.1 and 4) in the west MSVV.

Approximately-.475 ft'-of noninsulated surface area is exposed to-the ambient atmosphere in the valve vault.

TVA-calculations have f ound that-because of throttling processes,

-the steam temperature will drop to about 380 degrees F after passing through_the MSSV. This is the maximum temperature that the vent stack will reach. Although noninsulated, the energy input will not cause a significant air temperature rise in the 4-hour SB0 event duration.

The relatively large volume of air in the valve vault will. require a much higher heat input te cause a significant_ rise in i

temperature. Also to be considered is the fact that during_the SBO, the main steam isolation valve will close and isolate approximately 2

500 ft of steam piping from steam flow.

This piping, along with thc main feedwater piping-(also isolated during the event), will begin-to cool and contribute less and less heat to the valve vault environment. This decrease in heat load, along with that of other equipment deenergized by the SB0 event, will adequately compensate for the MSSV vent stacks.

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TVA'has conducted extensive environmental response analysesLon the west MSVV, considering high energy line breaks. The maximwn temperature expected in this room for-a_ main steam line break, considering the effects of superheated steam, exceeds _425-degrees F

-for a short-duration. The west MSVV temperature profile for a feedwater-line break rises quickly t-300 degrees F-and slowly decreases to about 190 degrees F in tsur hours.

Present equipment qualification studies have demonstrated the ability of 10 CFR 50.49 componente inside the west MSVV to withstand these conditions.

Therefore, sufficient margin is available to ensure that 10 CFR 50.49 components in the west MSVV can also withstand the environmental conditions generated by an SBO.

Calculation SQN-SBO-001 will be revised to document the preceding engineering evaluation.

It should be noted that operator entry into this area will not be required following the change to the turbine-driven AFW valves.

6.

NRC Recommendation:

"The licensee should include a full description, including the nature and objectives, of any required modifications in the documentation supporting the SB0 submittals that is to be maintained."

Response

The only modifications required are the AFW valve changes previously E

descriF<d. This modification for Unit 2 is planned to be completed by restart following the Cycle 6 refueling outage. This is consistent with the 2-year commitments made in the original SB0 submittal dated April 18, 1989. The Unit 1 Cycle 6 refueling outage

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-is scheduled to begin in less than 10 monthsLof the-anticipated acceptance of TVA's'SB0 response by N3C. -This would not be sufficient time for the procurement and design of the new control system. Therefore, for Unit 1, TVA commits to completing thic modification by start-up following the Cycle 7 refueling outage scheduled for October 1994. This_ commitment might exceed two-years-from NRC's' approval-of.SQN's SB0 proposal.

4.

The SB0 supporting documentation for these modifications < and the SB0 strategy are maintained in a similar manner to other, commitments, statements, procedures, and descriptions.. This documentation includes, but is not limited'to, engineering calculations,' design change packages, training rosters, and 10 CFR 50.59 evaluations.

7.

NRC Recommendation:

"The licensee needs to list equipment that will be used to provide information and/or to support plant coping during an SB0 and should-verify that SB0 equipment is covered by an appropriate quality assurance (QA) program consistent with the guidance of RG 1.155, Appendix A.

Furthermore,.this verification should be documented as part of the package supporting the SB0 Rule Response."

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9 Ite m pon s e t The SED equipment that is safety telatou is t:1 cady required to Le in a QA operability pr% rani.

The min ) ' 's-lE dist ribution requir ed i

to provide offsite power to the safety-te.ated busses is equl'ed to be operable in accordance with Technical Speellication (18 ) 3. 8.1.1.

The 250-V station battaty is inspected petiodically and is scheduled for capacity testing every five years.

The condensate storage tank hat, a TS requir ement that the inventory of condensate be snaintaineo nbove 190,000 gallons. The other SB0 mitigation equigment in safety teinted.

For the above tensons. the equipment required f or coi.ing with SB') are reasonably assured opetability should the event occur.

SBO-required equipment is contained in Attachment 3.

T"A will establish an augmented QA program to be applied to components required for coping with the SB0 event that will be cousinunt with the guidance of flegulatory Gt.lh (PG) 1.155 Appendix A.

8.

NitC flecommendatiot;t "The licensee should prod de conf A rtnation and include the documentation supporting the SB0 submittals that a progt run it'neting as a tninimum the guidance of itG 1.155. Position 1.?, is in place or will be implemented."

Ile s pon se t The present reliabilit y program for the emergency diesel generator unit (ED';U) does no et the tequirements of itG 1.155; however, these procedures wil revised to incorporate requirements from itG 1.155 as to the Ei target ieliability and maintenance programs necessary to maintain ils reliability. The present reliability for the EDGUs as of March 1.1, 1992, is 99 perce.at for the averace of all four emergency onsite anpplies.

This is an improvement from those values concultted to f or SBO.

The data for the 1rst 20, 50, and 100 diesel starts ist EDOU_ _.. Twenty _ _Fitty._ _ Onc.llundred 1A 20 50 99 2A 19 49 99 1B 20 SO 99 PB 20 50 99 Data from 0-SI-OPS-082-007M This table 11sts the successful start and load-run attempts. The number of failures is the difference between the column heading and the number listed in the column f or a particular diesel generator (DG).

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1:eview of SAIC's technical evaluation repott (TER) identifled thice additional concerns not described in the safety evaluation.

The concetns are addtrssed belows 9

fiAIC Concern 1:

"The licensee needs to ensure that it has considered ileld flashing at the end of the four-hour SB0 event when determining the adequacy of the DG batter-capncity."

Responset The EDGt) butteries do not have capacity to supply contt el powes to the diesels for the entite duration of an SB0 event.

Since the failed diesel is, by definition, a contributing cause of the SBO event, it is are not consideted tequired for coping. At the onset of an SPO, the operators will send a team to the DG building to troubleshoot and attempt to repair the failed EDGlls.

The DG statting air system wisl perfotm a starting sequence and then lock out.

This first att.rt sequence depletes the norinal supply of starting airl and, until the troubic is found and corrected, the second starting air oupply is not connected.

It would be up to t he team that is sent to rnake the repairs to turn off the battecy if deen +ed necessar y.

In any event, only one rnore start sequence remains in the back-up starting air supply.

a f the engine is capable of being star'.ed, he voltage on the generator will build up without iield ilas.?.

This occurs because of residual inagnetism in the f '4.ld (totor) iron.

Since the delay of a few seconds is not critical at the end of the 4-hour event, flashing the ficld is not required.

The control power to the 6.9-kilovolt shutdown boards will be available at the time of the event to connect the emergency ac should it become available.

The DG batteries were discussed in the supplemental response to NitC dated April 5, 1990, in which it was stated that the battery did not have the capacity to cope the 'e-hour duration of an SB0 event, a

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SAIC Concern 21 "We did not receive any information on whett.er the licensee used any exclusion criteria in addition to those given in RG 1.155.

One value which cannot be excluded requires manual action if it needs to be closed during an 3B0 event."

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-11 Responst The exclusion criteria of N! MARC 87-00, nr. endorsed by RG 1.155, were used for containment isolation valves. After reviewing the TER discussion on the CVCS containment isoletion valve (not identifled),

it is believed that the discussion appli-t to Valve 62-63 on renetration X-44.

This valve is identifled in IVA Calculation SQN-SySP-0078, Revision 0, as a valve that requires mor.ual action for an SB0 event. The other CVCS isolation valves are also reviewed in this calculation.

11.

SAIC Concern 3:

"The licensee needs to include the inanual closure o.J this valve in nn appropriate procedure and ensute that the valve is accessible."

Responset SQN Nuclear Engineering perfo med an analysis of the mechanient containment isointion system to determine the capability to isolate g

containment within the guidelines of NUMARC 87-00 during an SB0 out of this event. Mechanical penetrations were reviewed analysis, six valves were identifled that would.equire manual operatien per NtHARC 87-00, Steps 2 and 3, in the event that containment isolation is needed during an SB0 event.

These six valves are located in " habitable areas" and would therefore be accessible for inanual closure if containmeni isointion were required during an SB0 event.

Pinnt operating procedures will be revised as connitted to in TVA's letter to NRC dated April 18, 1989, to incorporate necessary operator actions to accomplish closure and/or verification of closure of these valves in the event that

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containment isolation is required during an HDL event.

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ENCLOSURE 2 List of Commitments 1.

Calculation SQN-SBO-001 will be revised to include additional sta' ion blackout (880) equipment, to address heat-up of LSo areas where the f

vital inverters are located, to include the west main steam valve f

vault discussions, to update the emergency diesel generator c 9ain additional loads for St0 events reliability assessment, and t~

r by 1)ecember 15, 1992.

2.

TVA will install new fail-ort

. vel control valves on the turbine-driven auxiliary feedwater pump with on-off control and suf ficicat air supply for 4-hour remote operation during an SB0 event and implement required SB0 procedurea by restart from the Cycle 7 1

refueling outage for Unit 1.

3.

TVA will Jnntall new fail-open level control valves on the turbine-driven auxillary feedwater pump with on-off control and' suf ficient air supply for 4-hour reino*.* operation during an SB0 event and implement required.SB0 procedures by restart from the Cycle 6 refueling outage for Unit 2.

Notel Commitments 2'and 3 above supersede SQN's Commitments Nos. 2 and 5 in TVA's letter to NRC dated April 18, 1989.

4.

TVA will institute a quality assurance program for SB0 components that is governed by site proceduren and meets the requirements of Regulatory Guide (RC) 1.155, Appendix A, within one year after the issuance of a safety evaluation report (SER) by NRC on Sequoyah's SBO.

5. - TVA will revise the procedures for the eroergency diesel generator reliability progrum to incorporate requirements from RG 1.155, Position 'i.2, within one year after the issuance of an SER by NRC on Sequoyah's SBO.

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s ATTACHNENT 1 LIST OF LOADS THAT ARLREMOVED DURING STATION BLAc)( g Lists are excerpted from Electrical calculctions SQH-CPS-031 and SQN-SBO-001.

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i' ATTACHMENT 1 LIST OF LOADS REMOVED DURING SBO i

DC BUSSES i-i The following is a list-of loads that muut be removed from tho 125 V Vital Battery Board I within 30 minutos, in order te moot SB0 requirements.

Other Battery Board removed loads are similar.

Breaker Lgad Doncription 204 6.9 KV Shutdown Bd 1A-A Backup Bus Altornato Foodor 205.

-480 V Shutdown Bd 1Al-A Backup Bus Alternato Feodor 206 480 V Shutdown Bd 1A2-A Backup Bus Alternato Feodor 207 480 V Auxiliary Building Common Ucard Normal Foodor l

214 Rod Drivo PoWor Switchgear Bkr'lA 1-L-115A 215 Gas Wasto-Disposal Panol 0-L-2C 222 cas Analyzer 0-L-206 223 Unit 1 Romoto RCP Oil-Lovel 224 Soric Acid Evaporator Packago A 0-L-1A 301 6.9 KV Shutdown Bd 2A-A-Normal Bus Alternato Foodor 302 480 V Shutdown Bd 2Al-A Normal Bus Alternato Feeder 303 480 V Shutdown Bd 2A2-A Normal Bus A1tornato Fooder 304 6.9 KV Shutdown Bd 2Al-A Backup Bus Normal Feodor

{

305 480 V shutdown Ed 2Al-A Backup Bus Normal Feodor 306 480 V Shutdown Bd 2A2-A Backup Bus Normal Foedor 313 Auxiliary Relay Rack 1-R-54 315 Generator Auxiliarios Panel Annunciator 1-L-39 319 Unit 1 Reactor Trip SWGR Bypass Breaker BYA I

320 Unit 1 Reactor Trip SWGR Trip Bkr RTA.

328 Unit 1 Annunciator Panel 1-L-236

-Page 1 t

.c...

r ATTAC!! MENT 1 l

LIST OF LOADS REMOVED DURING SB0 DC BUSSES The followinq is a list of loads that must be removed from the 250 V Station Dattery System in order to meet SB0 requirements.

t 4

Erpnhgr Load Description Time Linit

(

401 TSC Invertor Turned off at 30 minutos into SBO ovent.

404 Main Turbine Emergency Boaring Turned off at 210 Oil Pump minutos into SB0 ovent.

525 Generator DC Seal Oil Pump Turnod off at 30 minutos into SB0 ovent.

527-MFPT Emorgency Bearing Turned off at 30 Oil Pump 1A minutes into SBO event.

528-Computer Invertor 1 Turned off at 30 minutos into SB0 ovent.

529 Preferred Inverter 1

-Turnod off at 30 minutos into SDO ovent.

530 MFPT Emergency Bearing Turned off at 30 Oil Pump 1B minutes into SBO

ovent, t

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Page 2

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

~..

a-ATTACllMEllT 1 LIST OF 19 ADS REMOVED DURIllG SBO VITAL AC POWER SYSTEM BKR VITAI INVERTER 1-1 SAFETY NO Load Description RELATED 7

ERCW & CllTMT RAD MOli 1-RE-90-106, -133 YES 12 RAD RATE METERS, PNL 0-M-12 Yl:S 13 RAD Moll. 0-RE-90-125 YES 17 PAS SOLENOID VALVES, 1-L-572/C YES 19 TOILET, LOCKER, & SPREAD RM ISOL DMPRS YES 20 DOP PROCESS I!1ST CO!1T RACK YES 22 AB STM ISOL VALVE FCV-12-82, PliL I-M-9 YES 23 CONT PURGE AIR EXil. RAD MON, 1-RE-90-130 YES 30 AUX DRYER TRAIN A YES 32 BORIC ACID TANK A IITR A-A CollT, 1-L-303 YES 33 AB GAS TRTM!1T Fall A-A MOD DMPR, 0-L-429 YES 34 BORIC ACID TA!1K C liTR A-A CollT, 0-L-306-YES 35 RAD MON 0-RE-90-205 YES 36 RCP 1 UV & UF RELAYS YES 37 PROCESS CollT GRP 1, P!fL 1-R-14 11 0 38 IllST DUS 1, PNL 0-M-27B 11 0 39 PLUGMOLO IllST BUS 1, P!1L 1-M-5 11 0 40-PLUGMOLD I!!ST BUS 1, PNL 1-M-6

!!O 41 INST DUS 1 & PIC-1-6A, -31A, 1-M-4 11 0 42 FIRE PMP 2A-A SEP RELAYS 11 0 43 AB GEli EXII Fall 1 A FLOW ColiT, 0-L-426 NO 46 PR-30-310 YES VITAL INVERTER 1-II 7

ERCW/CNTMNT RAD MON 0-RE-90-134,1-RE-90-112 YES 10 RB ISOL VLVE FCV-32-102A, -102B, JB2674 YES 11 AUX COMP B AUX BLDG ISOL VLV, FCV-32-85 YES 12 RAD RATE METERS, 0-M-12 YES 13 RAD MOli 0-RE-90-126 YES 17 PAS SOL VLVS, 1-M-10 YES 19 TOILET, LOCKER, SPRD RM ISOL DMPR, 1-R-78 YES 20 BOP PROC INST CollT RACK, 1-R-131 YES 22 AUX BLR STM ISOL VLV FCV-12-79, 1-M-9 YES 23 COliT PRGE AIR EXII MOli, 1-RE-90-131 YES 26 AUX RELAY RACK SEP AND AUX RELAYS, 1-R-78 YES 30 AUX DRYER TRAIN B YES 32 BORIC ACID T!iK A llTR B-B CONT, 1-L-304 YES 33 AUX BLDG GAS TRTMNT Fall B-B MOD DMPR 0-L-428 YES 34 BORIC ACID TANK C llTR B-B CO!!T, 0-L-305 YES 35 RAD MON 0-RE-90-206 YES 36 RCP 2 UV & UF RELAYS YES 37 PROCESS CONT GROUP 2, 1-R-17 11 0 38 INST BUS 2, 0-M-27B 11 0 39 PLUGMOLD INST BUS 2, 1-M-3

!!O 40 PLUGMOLD INST BUS 2, 1-M-6 NO-41 ACOUSTIC FLOW MON, 0-M-27A 11 0 42 FIRE PUMP 2B-B SEP RELAYS, JB3718

!!O 43 AUX GEN EXil Fall IB FIOW CONT, 0-L-427 NO Page 3 C:\\PFS\\BUZ\\SBO\\!1RCQS.AT1

. ~

ATTAcilMENT 1 LIST OF LOADS REMOVED DURIllG SB0 VITAL AC POWER SYSTEM DKR VITAL INVERTER 1-III SAFETY NO Load Description RELATED 7

RCP 3 UV & UP RELAYS YES 14 IllST BUS AllD XFMR PWR, 1-M-3 NO 15 AUX CollT PNL A INST BUS, 1-L-11A No 16 PROCESS CollT GRP 3, 1-R-20 NO 17 BOP PROC. INST CONT RACK, 1-R-126 No 10 C011 TROL ROOM DOORS SECURITY 14CK Ho 19 EGTS FILTER TRAIll A, 0-L-25 NO 22 AUX BLDG INST A BUS 1, 1-L-57 11 0 23 AUX RELAY RACK A BUS, 1-R-76 NO 24 AUX RELAY RACK C BUS, 1-R-76 NO 25 NSSS AUX RELAY RACK A BUS, 1-R-58 HO 26 AUX CONT PANEL A BUS, 1-L-10 NO 27 AUX RELAY RACK A BUS, 1-R-75 NO 28 SSPS CONT RM DEMUX, 1-M-22 13 0 29 AUX CONT PNL C RLY BUS, 1-L-10 NO 30 AUX CONT PNL A INST DUS, 1-L-10 NO 31 CONT AIR HDR A MOIST ALM, JB281 11 0 32 AUX RELAY RACK A BUS, 1-R-52 No 34 POST ACC M0!I PNL, 1-M-5 NO VITAL INVERTER 1-IV 7

RCP 4 UV & UF RELAYS YES i

14 INST BUS 4, 1-M-4 No 15 AUX CONT P!lL B INST BUS, 1-L-11B 11 0 16 PROCESS CONT GRP 4, 1-R-22 NO

-17 BOP PROCESS INST CONT RACK, 1-H-122 No 19 AUX BLDG INST BD BUS 2, 0-L-23 No 20 BOP PROCESS INST CONT RACK, 1-R-130 No 21 BACKUP CONTROL INST LOOPS No 22 AUX BLDG INST BD BUS 1, 0-L-283 No 23 AUX RELAY-RACK B BUS, 1-R-76 NO 24 NSSS AUX RELAY RACK C BUS, 1-R-58 NO 25 NSSS AUX RELAY RACK B BUS, 1-R-58 No 26 AUX RELAY RACK B BUS, 1-K-7S NO 27 AUX RELAY RACK C BUS, 1-R-75 NO 28 AUX CONT PNL B RELAY BUS, 1-L-10 NO 29 AUX CONT PNL B IllST DUS, 1-L-10 NO 31 CONT AIR NDR B MOIST ALARM, JB281 No 32 AUX RLY RCK B BUS, 1-R-72 No 33 AUX PLY RCK C BUS, 1-R-72 NO 34 POST ACC MON 2, 1-M-4 NO 36 LOCA H2 CNTMNT FLOW MON, 1-M-10 NO 39 FEED TO BKR 37, 38 No Page 4 C:\\PFS\\BUZ\\SBO\\NRCQS.ATI u.

l 6

ATTACliMENT 1 i

LIST OF LOADS REMOVED DURING SB0 i

VITAL AC POWER SYSTEM

)

BKR VITAL INVERTER 2-I SAFETY i

NO Load Doncription RELATED 14 INST BUS 1 & PIC-1-6A, -31A NO 15 PLUGMOLD INST BUS 1 PNL, 2-M-5 No 16 PLUGMOLD INST BUS 1 PHL, 2-M-6 11 0 17 PROCESS CONT GROUP PHL, 2-R-14 No 18 BOP PROC. INST CONT RACK, 2-R-126 NO 19 AUX CONT PNL A INST BUS, 2-L-11A NO 20 BOP PROC CONT INST RACK, 2-R-128 No 22 AUX BLDG INST A BUS, 2-L-57 HO 23 AUX RELAY RACK A BUS, 2-R-76 NO 24 AUX REIAY RACK C BUS, 2-R-76 No 25 NSSS AUX REIAY RACK A BUS, 2-R-50 No 26 AUX CONT PNL A RELAY BUS, 2-L-10

!!O 4

27 AUX REIAY RACK A BUS, 2-R-75 NO 28 SSPS CONT RM DEMUX, 2-M-22 NO 29 AUX CONT PNL A IllST BUS, 2-L-10 11 0 32 AUX RELAY RACK A BUS, 2-R-32 No 34 POST ACCIDENT MON 1, 2-M-5

!!O 36 LOCA H2 CNTMT FLOW MON, 2-M-10 NO VITAL INVERTER 2-II 7

RCP 2 UV & UP RELAYS l

YES 14 PROCESS CONT GROUP 2, 2-R-17 No 15 PLUGMOLD INST BUS 2, 2-M-3 NO 16 AUX CONT PNL B INST BUS, 2-L-11B 11 0 17 PLUGMOLD INST BUS 2, 2-M-6 No 18 BOP PROCESS INST CONT RACK, 2-R-122 No 20 BOP PROCESS INST CONT RACK, 2-R-130 11 0 22 AUX BLDG INST B BUS 1, 2-L-299 NO i

23 AUX REIAY RACK B BUS, 2-R-76 NO 24 HSSS AUX RELAY RACK C BUS, 2-R-58 No 25 NSSS AUX RELAY RACK B BUS, 1-R-58 NO 26 AUX RELAY RACK B BUS, 2-R-75 NO 27 AUX RELAY RACK C BUS, 2-R-7 NO 28-AUX CONT PNL B RELAY SUS, 2-L-10 No 30 AUX CONT PNL B INST EUS, 2-L-10 No 32 AUX RELAY RACK B BUS, 2-R-72 NO 33 AUX REIAY RACK C BUS, 2-R-72 NO 34 POST ACCIDENT MON, 2-M-4 NO 36 LOCA H2 CNTMT FLOW MON, 2-M-10 N0 48 NIS INSTRUMENT POWER NO Page 5 C:\\PFS\\BUZ\\SBo\\NRCQS.AT1

ATTACllMENT 1 LIST OF LOADS REMOVED DURING S!50 VITAL AC POWER.SUSTEM DKR VITAL INVERTER 2-III DATETY NO Load Description RELATED 7

CONT RAD MO!1 2-RE-90-106 YES 10 REACTOR BLDG ISOL VALVE, FCV-32-81A,-81B YES 12 RADIATION RATE METERS & 2-RI-90-106, 0-M-12 YES 13 RCP 3 UV & UP RELAYS YES 17 PAS SOLENOID VLVS, 2-M-10 YES 20 BOP PROCESS IllSTR CONT RACK, 2-R-128 YES 23 CONT PURGE AIR EXilAUST RAD MON YES 32 BORIC ACID TNK B llTR A-A CollT, 2-L-303 YES 37

. PROCESS CO'iT GRP 3, 2-R-20 NO 46 MAlli STEAM RADIATION MON. (2-RX-90-424)

DO 48 PWR RAliGE PEN RECORDER, 2-M-13 NO VITAL INVERTER 2-IV 7

CONTAINMENT RAD MON 2-RE-90-112 YES 8

CONTAINMENT ANNULUS DP, 0-M-27B YES 10 RB ISOL VLV, FCV-32-103A,-103B,-111A,-111B YES 12 RAD RATE MTRS & 2-RI-90-112, 0-M-12 YES 13 RCP 4 UV & UP RELAYS YES 17 PAS SOL VLVS 2-M-10, 2-L-572,'D YES 20 LOP PROCESS INST CONT RACK, 2-R-131 YES 32 BORIC ACID TNK B llEATER B-B CONT, 2-L-304 YES 37 PROCESS CO!JT GRP 4, 2-R-22 11 0 Page 6 C:\\PFS\\BUZ\\SBO\\NRCQS.ATI

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i ATTACHMENT 2A d

12$ VOLT VITAL BATTERY LOADING PROFILE t

i Excerpted from Electrical Calculation SQN-SBO-001.

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.ba c)?fi'fW APPENDIX A Reviewed by:d[h[e[g(E/yghyly ?

Prepared by:

0---

i VITAL BATTERY EVALUATION l i p

3{

LOAD DUTY CYCLE FOR VITAL BATTERY I (load currect in ampo) 898{

t

\\

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(429 amps) 429 296 (af ter stripping loads 4946 ampc) 278 1

30 240

~

(time in minutes)

FIGURE 1 The load tabulations result in the following loads and times for Vitha Battery I.

These load values will be used to verify that the Vital Batteries can meet the four hour duty requirement imposed by the plant / power system SB0 evaluation.

The duty cycle curve is then constructed from the load data.

~

From: t= 0 seconce to 5 accende-897.53 amps

-- Per IEEE 495, The t= 5 seconds to 1 minuto -

649.23 amps load current is assumed to be 898 t 1 minute to 30 minutes-428.91 ampc amps f or first min.

296 t= 30 minuten to 4-houre-270.;4 ampc j

8.0

SUMMARY

OE RESULTS The-vital battery system, with load u 'ipping, hac the capacity to supply the required loads for the required duty cycle with approximately 10 per cent excess capacity.

The voltage at the battery terminals at the end of the SB0 4-hour cischarge is 110.1 volte and the.

minimum voltage required for operability is 105.0 voltc.

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AH'ENDIX A q

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(JQ1 S UI NO WORVS HFTT, JJ[f f 4 9 5 s

ET' CELL TYPE NcK 2: 00 MINIMUM CELL VOLTAGE 1 ?!

MIN TEMPEkAIURE.ff F Totel*]

Load Change an Timo in Time to end Ampc/ pes Sub +

(Ampere)

Load Minuter (Minuter)

Total

_ __ m.mmm _ _-mw.-

.n = -

_mwm.m SECTION 1 l

lT=M13 Als 898 Als 899 M1= 1 1

160 5.t1

-r

---mmem m w= sex-m - n.

m- - - - &

SECTION 2 l

l 5,t1

{

1 Al= 89B Air 898 Mis 1 TsM1+M2= 30 102.1 8.79 A2= 429 A2-Alt -469 M2r 29 T=M2= 29 103.5

! -4.53 l

mra nnn===+--- - n...m --.. nn SFCTION 3 4.26 Ale 898

, Air 898 M1= 1 T=M1+M2+M3r 240-31.8 20.24 A2r 429 A2-Al=

469 M2= 29 T=M2+M3 239 31.9

-14.70 A 38: iW {g6 A3-A2= - M (33 M3 210 T=M3r 210 35.0

- r._.. ~mxx 6. n.._ ~ e.q-

~ --

SECTION 4 I

J-9.74-Als Mis T=M1+..+M4r l

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l I

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J.-.

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l l

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Positive Plates TEMPERATURE CORRECTION FACTOR: 1 11 AGING FACTOR = 1.25 NUMEER OF PLATE 3 AVAILALLE: ).1 13.51 HIGHEST PLATES REOUIRED X 1.11 X 1.25 =.:.

FiATES 2

l l

.. ~.

~

~^

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,p

.% g (, Lg.

C QI: ' C0 0C1

j. n [],j Q 9e;,_3 ]3 g.}3 g TABLE 1 AFFENDIX A v

As

.JM-.-. C.'. i :

_l},,.L.!d!2

g-*

CIRCU!T NAME CKT 0.00 10 0.05 to 1.00 70 30.e0 Tc LOAD BER 0.05 Leesi.00 mins 30 Mi% 240 M/na LHED s

6.9 KV SDBD 1Al-t 201 0.00 7s;.16 0.0h 0.00 NO NbNF 3.17 2.17 3.17 3.17 NO U.00 1.19 1.19 1.19 NO 1.50 0.00 0.00 0.00 NO 400V SCDD 1Al-A 20;l 0.94 0.94 0.94 0.94 NO NDNF 0.00 6.00 0.00 0.00 No 480V SDED 1A2-A 203 0.00 10.00 0.00 0.00, NO NBNF 1.1 *i 1.14 1.14 1.14 NO 6.9 KV SDBD 1A-A 204 1.96 0.00 0.00 0.00 YES BBAF 1.60 1.60 1.60 0.00 YES 0,00 0.64 0.64 0.00 YES 480 V SDBD 1Al-A 205 0.00 2.00 0.00 0.00 YEE BBAF 0.54 0.54 0.$4 0.00 YES 480V SDED 1A2-A 206 0.00 2.00 0.00 0.00 YES BEAT 0 54 0.54 0.54 0.00 YE2 480 V AUX ELDG 207 0.00 4.00 0.00 0.00 YES COMMON BOAED 10.00 0.00 0.00 0.00 YEr g,

1.6' 1.61 1.01 0.00 YEE i

W FUSE ASSEMBLY-COL A 210 11.19 11.19 11.19 11.19 Nv FUSE ASSEMBLY-COL B 211 4.40 4.48 4.48 4.4P NO FUSE ASSEMBLY-COL C 212 6.41 6.41 6.41 6.41 NO FUSE ASSEMBLY COL D 213 8.51 8.51 8.51 8.51 No R0D DRIVE FWR SWGR 214 0.04 0.04 0.04 0.09 YED GAS PANEL 0-L-2 215 0.95 0.95 0.95 0.00 YES RESPONSE !!ME TSTINC 221 0.00 0.00 0.00 0.00 NO GAS ANALY2ER 0-L-205 222 0.00 0.00 0.00 0.00 YES l

U1 RCP REM O!L LEVEL 223 2.24 2.24 2.24 3.00 YES 6.9 KV SDBD 2A A 301 0.00 78.16 0.00 0.00 YES NBAF 1.56 0.00 0.00 0.00 YES 3.21 3.21 3,21 0.00 YES 0.00 1.14 1.14 0.00 YES 480 V SDBD 2Al-A 302 0.00 6.00 0.00 0.00 YES 3

NBAF 0.96 0.d6 0.86 0.00 YES 4

480 V SDBD 2A2-A 303 1,18 1.18 1.18 0.00 YES NBAF 0.00 B.00 0.00 0.00 YES

~

~ ^ '

, p.3 TABLE 1 APPENDIX A Prer*r d. **k.

cl?

Ed*/?A __

_]

U l'

1.00~TCr2 d.Cns

-- l r ;

& Tir

• LOMJ- '7-"-l d* -

CIRCUIT NAME CKT 0.00 TO 0.05 TO Ef' BKR 0.05 Secs 1.00 minc 20 Mins 240 Manc SHED 6.9 KV SDBD 2A-A 304 1.96 0.00 0.00 0.00 YES BBNF 1.60 1.60 1.60 0.00 YES 0.00 0.00 0.00 0.00 YES 0.00 0.64 0.64 0.00 YES 490 V SDBD 2Al-A 305 0.00 2.00 0.00 0.00 YES BBNF 0.54 0.54 0.54 0.00 YES 480 V SDED 2A2-A 306 0.54 0.54 0.54 0,00 YES BBNF 0.00 2.00 0.00 0.00 YES l

DC BUS FILTER 300 0.08 0.08 0.08 0.0E NO j

1 AUX RELAY RACK 313 0.00 0.00 0.00 0.00 YES 1-R-54 GEN AUX ANN PKG 315 3.40 3.40 3.40 0.00 YES BORIC ACID PKGO L1A 319 0.71 0.71 0.71 0.00 YES BYPASS BKR BYA TR A 319 0.04 0.04 0.04 0.00 YES

{M RX TRIP BKR RTA 320 2.00 0.04 0.04 0.00 YES I

U1-AUX FD PMP Tt'Pb 321 0.00 12.00 0.00 0.00 NO 2.84 2.84 2.84 2.84 NO DC LIG CAB LD-1 325 550.00 96.00 96.00 9d.00 NO 9 9.$8

• VITAL INV 1-I 326 116.08 116.08 116.08

"". 9-No 60.84 #

VITAL INV 2-I 327 98.85 99.85 98.95

-54. 4 NO UNIT 1 ANN PANEL 328 50.00 56.00 56.00 0.00 YES 1-L-236 296.lf TOTALS 898.35 639.23 428.91 2.s..+

t These loads increased as a result of adding RVL / S and In 'lant VH F Raelio back to Wtal Inver ter 1.oad List.

M

... ~.

0 i

ATTACHMENT 2B i

250 VOLT STATION BATTERY LOAD PROFILE i

Excerpted from Electrical Calculation SQN-CPS-031.

i 3

I P

4 i

't Page - lG 2502 hj - C P S - C 31 em Table 1A Page I

of 2 y

\\

JfC lt*/t.7 l) 250 STATION BATTERY LOADINO CALCULATION Prepared by_

_/

EBDP Running for 3-hours Reviewed by 6,g

~/44 CIRCUIT NAME CKT 0.00 TO 10 s. TO 1.00 TO 30.00 TO 180 To 239 To Normal BKR 10 sces 1.00 mins 30 M. ins 186 Mine 239 Mins 240 Mins Current 205 V Turbine Bldg 201 21.74 21.74 21.74 21.74 21.74 21.74 21.74 Bd 1 Normal Feeder 255.00 25.00 l

r 205.V Turbine Bldg 202 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Bd 2 Alt. Feeder Elec Cont Bd Dist 203 35.00 35.00 35.00 35.00 35.00 35.00-35.00 Pn1 7 & B Not Fdr 168.00-24.00 Elec Cont'Bd Dist 303 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Pn1 6 Alt Fdr 0.00-TSC Inverter 401 145.45 145.45 145-.45 0.00 0.00 0.30 0.00 0.00 p-(g;py Turbine EBOP-Unit 1 404 0.00 0.00-283.64 283.64 0.00 0.00 0,.00 Normal Feeder _

_910.00 Turbine EBOP Unit 2 405 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Alternate Feeder 0.00 Electric _ Shop 409 0.00 0.00 0.00 0,00 0'.00 0.00 0.00' l

Test bench 480 V Water'5upply 501 0.61 0.61 0.61 0.61 0.61 0.61 0.61 Bd. Normal Feeder 5.00 5.00 480 V Svce Bldg Wtr 502 0.61-0.61 0.61 0.61 0.61 0.61 0.61 Normal Feeder 5.00 5.00 i-480 V Unit Bd 1A 507 1.50 1.50 1.50' 1.50 1.50 1.50 1.50 o

Normal-Feedor 20.00 480 V Unit Bd 2A 508 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Alternate Feeder 0.00 400 V Unit Bd 1B 509 1.76 1.'76 1.76 1.76 1.76 1,76 1.76

~

.; -s

_ Normal Feeder 30.00 480 V Unit Bd 2A 510 0,00 0.00 0.00 0.00 0.00 0.00 0.00 Alternate Feeder

., _ _ _ _ ____.--.~__--_ - -- - - -

Page jg,,A S Q hj - c P S - o 9 r

f Table 1A Page 2 of 2 250 STATION BATTERY LOADINO CALCULATION Prepared by

[

/

EB0p-Running for 3-hours Reviewed by_g M g

//gg i i

CIRCUIT NAME CKT 0.00 TO 10 s. TO 1.00 TO 30.00 TO 180 To 239 To Normal i

BXR 10 Sees 1.00 mina 30 Mins 180 Mine 239 Mins 240 Mins Current 250 Battery Bd 512 0.00 0.00-0.00 0.00 0.00 0.00 0.00 Bus Filter i

Turbino Trip 516 1.40' 1.40 1.40 1.40 1.40 1.40 1.40 Pus A Unit 1 0.00 Turbine Trip 519 1.40 1.40 1.40 1.40 1.40 1.40 1.40 Bus A Unit 2 0.00 t

Main Feed Pump Trb 523 2.06 2.06 2.06 2.06 2.06 2.06 2,06 A Trip Bus Unit 1 0.40 Main Teed Pump Trb 524 2.06 2.06 2.06 2.06 2.06 2.06 2,06 B Trip-Bus Unit 1 0.40 Gen DC Seal 011 Pmp-525 0.00 99.27 99.27' O.00 0.00 0.00.

0.00 Normal Feeder Unit 1 318.50

Ayn I.

Gen DC Seal 011 Pmp 526 0.00 0.00 0.00 0.00 0.00 0.00 0'.00 I

Alt. Teeder Unit 2 MTPT Emerg Brg 011 527 0.00 0.00 32.84 0.00 0.00 0.00 0.00 Pmp 1A-Unit 1 105.35 Computer Inverter 1 528 48.48 48.48 48.48 0.00 0.00 0.00 0.00 Preferred Inverter 1 529 58.18-58.18 58.18 0.00 0.00 0.00 0.00 MFPT Emerg Brg 011_ 530 0.00

.0.00

'32.84' O.00 0.00 0.00-0.00 Pmp 1B Unit 105.35

TOTALS 1122.55 1540.22

-768.83 351.78 68.24 127.14 68.14

.The motor amps are adnusted for average volta,,e over the duty cycle, Refer to Section 5.0.

C'

,_,_,.__,,_--.m.,__

PageIk Ch Table IB Sheet I

of 2

,e, 6

250 STATION BATTERY LOADINO CALCULATION SON-CPS-031 EBOP Running for 4-hours

)

J P JL tof g,; 7'4

\\

Prepared by

/

l Reviewed by 3r 6

/5h/27/fra i

CIRCUIT NAME CXT 0.00 70 10 a. TO 1.00 TO 30.00 TO 239 To Normal BXR 10 Sees 1.00 mins 30 Mins 239 Mina 240 Mins Current 205 V Turbine Bldg 201 21.74 21.74 21.74 21.74 21.74 21.74 Bd 1 Normal Feeder 255.00 25.00 205 V Turbine Bldg 1 202 0.00 0.00 0.00 0.00 0.00 0.00 Bd 2 Alt. Feeder Elec Cont-Bd Dist 203 35.00 35.00 35.00 35.00 35,00 35.00 Pn3 7 & B Nor Fdr 168.00 24.00 Elec Cent Bd Diat 303 0.00 O.00 0.00 0.00 0.00 0.00 Pn1 6 Alt Fde 0.00 TSC Inverter 1 401 145.45 145.45 145.45 0.00 0.00 0.00 g-~s

,gghi 0.00-

-(

)TurbineEBOPUnit1 404

-0.00 0.00 283.64 283.64 0.00 0.00 Normal Feeder 910.00 Turbine EBOP Unit 2 405 0.00-0.00

-0.00 0.00 0.00 0.00 Alternate Feeder 0.00

+

Electric Shop 409 0.00 0.00 0.00 0.00 0.00 0.00 Test Bench 480 V Water Supply-501 0.61' O.61 0.61 0.61

_0.61 0.61 Bd. Normal Feeder 5.00 5.00 480 V Svce Bldg Wtr 502 0.61 0.61 0.61 0.61

- 0.61 0.61 Normal Feeder 5.00 5.00 480 V Unit Bd 1A 507 1.50 1.50 1.50 1.50 1.50 1,50-Notmal-Feeder 20.00 400 V Unit Bd 2A 508 0.00 0.00' O.00 0.00 0.00 0.00 Alternate Feeder 0.00

.c

/

480 ? Unit Bd,1B 509 1.76 1.76 1.76 1.76 1.76 1.76 Normal Feeder 30.00 g w e,

y,--

p p g y-w- ---

e-<,4i-e-eev-,,pmm--,--,-+,,,v-,..we,,,-.y,-.we-.4my,

,wy,m, y

pww-w-w-*-.>

. +-

e v

wwi,-=rv4=

+c

---t e-mwew vrw-=-

Page O b f

i Table IB Sheet [h'of2 l

250 STATION BATTERY LOADING CALCULATION SON CPS-031 EBOP Running for 4-hours j

/ b[^.,7 [4C l

J Al Prepared by.

/.

/ t o / ?[f g Reviewed by_Af6 4 J

t CIRCU1T NAME CRT 0.00 TO 10 s. TO 1.00 TO 30.00 TO 239 to Normal BKR-10 Sees 1.00 mins 30 Mins 239 Mins 240 Mins Curron+

480 V Unit Bd 2A 510 0.00 0.00 0.00 0.00 0.00 0.00 Alternate Feeder 4

250 Battery Bd 512 0.00 0.00 0.00 0.00 0.00 0.00 Bus-Filter Turbine Trip 516

.1.40-1.40 1.40 1.40 1.40 1.40 Bus A Unit 1 0.00 Turbine Trip 519 1.40 1.40 1.40 1.40 1.40 1.40 t

Bus A Unit 2 0.00 t

Main Feed Pump Trb 523' 2.06 2.06 2.06' 2.06 2.06 2,06 A Trip Bus Unit 1-0.40

[

Hain Feed Pump Trb 524 2.06 2.06 2.06 2.06 2.06 2.06

\\

B Trip Bus Unit 1 0.40 Gen DC Seal Oil Pmp 525 0.00 90.27 99.27 0.00 0.00 0.00 Normal Feeder Unit 1 318.50

+

. Gen DC Seal 011 Pmp._

526 0.00 0.00 0.00 0.00 0.00 0.00 Alt. Feeder Unit 2 I

MFPT Emerg Brg 011 527 0.00 0.00 32.84 0,00 0.00 0.00 Pmp 1A Unit'1 105.35 Computer Inverter 1 528 48.48 48.48.-

48.48 0.00 0.00 0.00

- Preferred Inverter 1-529 58.18 58.18 58.18 0.00 0.00 0.00

MFPT Emerg Brg oil 530~

0.00 0 00 32.84 0.00 0.00 0.00 PmpH13 Unit 1-105.35 TOTALS 1122.55 1540.22 768.83-351.78 127.14 68.14

,- s.- The motor amps-are adjusted for average voltage ov3r the duty cycle.

l Refer to-Section 5.0.

i

-,,,,--.....m._,....-.s.

.-_,.._.-..m..,-_-,.---_.,..

_,_m

.._m,,.-a,..._,

.=.

a l'

SON-CPS *031 Page i

s Prepared by _ ft?

1.*/s;*//C J

/

Reviewed by

/,

.27 fd

/

?

PER LOAD IN CHANGE INDUR.

11ME RATE PLA7ES TOTAL ICD AMPS LOAD PER. TO END AMP /POS REQ'D FOR PER.

1 1540.22 1$40.22 1

1 160.00 9.63 9.63 13.36 2

1540.22 1540.22 1

20 97.00 15.88 768.83

-771,39 29 29 97.50

-7.91 7.97 11.05 3

1540.22 1540.22.

1 180 39.00 39.49 i

768.83

-771.39 29 179 39.00 -19.78 351.78

-417.06 150 150 43.00

-9.70 10.01 13.90 4

1540.22 1540.22 1 - ', 2 3 R 32.00 48.13 768.83 -771.39 29

.230.

32.00 -24.11-351.78 -417.06 150 209 33.00 -12.64-68.14 -283.64 59 59 73.00

-3.89 7.50

+

10.41 5

1540.22 1540.22 1

240 32.00 48.13 1

768.83 -771.39 29 239 32.00 -24.11 yr-ss ijrni 351.78 -417.06 150 210 33.00 -12.64

(

68.14

-283.64 59 60 73.00- -3.89 127.14 59.00 1

1 160.00 0.37 7.87 10.92 9

TOTAL NO. OF PLATELATES-------------

13.70 INCLUDES AGING AND AND TEMP. CORRECTION REMAINING DES!GN MGN'MAGIN t -------- - 0.75 i

Table 2A IEEE 485 Work Sheet E/bP on for 3-Houra 9^

i

< i,

-..,_v.,,,_,,

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

_,,.,,_v,..

e... -,

,#..,..r,-.__r.

,..2.,, _,,,..,,.. _.,.., _,,. -. _..,,,.,,.,

__._m..._________

k s

SON-CPS-031 Page JPY

/ _.__/t2/*

/p Prepared by Checked by

/ j_j 7//d l

/

IEEE 485 Work Sheet-EP,0P Running 4-hours PER LOAD IN CHANGE INDUR.

TIME RATE PLATES TOTAL l

10D AMPS LOAD PER. TO END AMP /PCS REQ'D TOR PER.

1 1540.22 1540.22 1.

1 160.00 9.63 9.63 13.36 2

1540.22 1540.22 1

30 97.00' -15.88 768.83 -771.39 29 29 97.50

-7.91 7.97 11.05 3

1540.22 1540.22 1

239 32.00 48.13-763.83 -771.39 29 238-32,00 -24.11" 351.78 -417.06 209-209-33.00 -12.64 11.39 15.80 4

1540.22 1540.22 1-240

_32.00 48,13 768.83

-771.39-29 239 32.00 -24.11 r

-351.78 -~417.06 209 210 33.00 12.64 127.14 -224.64 1

1 160.00

• 1.40 9.98 13.85

(%

a:h4 1

\\

TOTAL NO. OF PLATELATES-------------

15.80 INCLUDES AGING AND AND TEMP.' CORRECTION:

REMAINING DESIGN MARGIN IN 1 -----

-12.86 4

1 IEGE 4Ns W k Sheed (of 4

hour G30P cm I i l

fi-

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ATTACHMENT 3 COP'NG STRATEGY FOR A STATION BLACKOUT EVENT E

Excerpted from Electrical Calculation SON-SBO-001.

)

a 4

34

_ _.3

,o

~.

vQl:..-

Appendix J

/~ "_

_Date i."/ 3

.e:

prepared

~'%

Reviewed _s &

(-

Date_3fp/pv C

SQN COPING STRATEGY AND MINIHAL EQUIPMENT MECESS ARY FOR COPING WITH A STATION BLACKOUT 1.0 bcpose The purpose of this appendix is to provide the strategy and a list of the equipment needed to survive a 4-hour Statio'1 Blackout at Sequoyah Nuclear Plant.

2.0 Introduction The strategy to cope for four hours with the complete loss of all ac power at Sequoyah is presented in the Analycic and Results section of this appendix. The equipment list and he logic used in its compilation follows.

Additional information regarding the location of this equipment is provided in Table J-1.

3.0 Analyjic and Resulte - Corint Strateny 3.1 First phase time period - O to 30 minutes

)

a. Monitor reactor incore temperature and proccure. Use turbine driven auxiliary feedwater to maintain steam generator level and use steam generator safety relief valves to maintain steam generator pressure (of (non0B( Centr 0l Oi a ttn O 6 p h 6 fic, f't.d c t

\\/aN e.5).

b. If restoration of AC power is not likely within 30 minutes of initial loss of power:

V Vita l PC, 12.O V Jital Inverters (1) prepare to shed load on 125 Vig systemy; -Th c 2 5 0 '; DG' sysicm dre

=t req = re lead rheddir.g, a nd th e. 250 \\/ D C 5tation 6 ditch Shddown t

(2) prepare to remove loads on the 6.9 kv boards in preparation P

for regaining one or more emergency diesel generators.

lsvel Achieve (control of the appropriate steam generator 4 eve 4-(3) cerd rol val +e+ (within 15 minutes of station blackout).

3.2 Second phase time period - 30 minutes to restoration of AC power

a. Shed unnecessary loads from the 125 VDC system, 120 V Vitdl bC 6TE6" a nd 25 0 V t)C-Station Bettcry System.

b. Remove major loads from the 6.9 kv common and unit boards,

c. Continue manual control, if required, of the appropriate steam generator level ec-tral-;:1. =,

• The 250 VDC system is required for operation of the cwitchyard breakers.

w) 1119L

Appendix J e

SQ N EC'J 002 Prepared

"'n e.'a' Date. r>>-

a Reviewed./4yfv4 4 4 ~

Da t e Vn/p,

/

4" 3.0 Analysis and Recults - Copinr. Strateny (Continued) 3,3 Restoration of AC power by recovering offsite power.

Retucn to normal procedurcs.

3.4 Restoration of AC power by recovering two or more emergency diesel generators.

Return to normal loss of offsite power procedure.

I i

3.5 Restoration of AC power by recovering one diesel generator.

l SCO.PE : This scenario is for an event that is not required by the design basis of the plant and is not required by the SB0 rule.

It is presented here as an enhancement to the coping strategy,

a. Line ',

uch that adequate cooling water is available to s

cool the (;

2000.

b. E5c it ERCW at the ERCW/ condensate switch-over e,

vats tion of the depletion of the coadensato i nve,

c. Establ._

component cooling water pump, aligned to the ERCW train that is available,

d. Establish a centrifugal charging pump on cach unit.

The charging

'k%FEk pump on the 6.9 KV shutdown board arcociated with the operable diesel will roccive power normally; however, the opposite unit charging pump must be powered by the limited capacity intertie bus.

All the loads on the opposite unit shutdown board that is connected to the operable diesel generator nost be shed except for the centrifugal charging pump,

e. Establish 480v AC power to the vital battery charger that is associated with the operable dicscl generator,

f. Lighting and HVAC loads necessary to support the above equipment should be manually connectedg as necessa ry -

g. The 250 V DC station battery chargers can be manually connected to all diesels except 2B-b and will be used to control the switchyard circuit breakers when off site power is returned.

If 2B-B is the operable diesel, then power to the charger can be supplied by jumpering.

It ic felt that this strategy will be sufficient to reEain and maintain control given recovery of one diesel geocrator following an SBO event.

This configuration is only a stop gap measure that is to maintain core stability for about 2A-hours until off cite powcr is restored or until additional diesels are restored.

}.)}

1119L

S Q N 1 f30 C02 Appendix a prepared

'~7; $ E Date

'A Reviewed../2 #gph / M y,_Date 4.0 01alysis and Results - E_quj,pment List tAy 8 i

\\lC

.i.1 Steam Generator Only one steam get rator is required to remove heat froin the reactor coolant system under station blackout conditions (Reference 1).

(14oildown in all four steam generators is available for some time period immediately after reactor trip.) However, the steam generator used to remove heat f rom the RCS, i.e.,

the one that is fed by auxiliary feedwater, must also be one that is supplying steam to the turbino driven auxiliary feedwater pump.

Only steam generators 1 and 4 are capable of this.

Therefore, either steam generator 1 or steam generator 4 is considered neccesary for use during station blackout. Which of the two is used will be determined by the system alignment prior to the station blackout event.

4.2 125 VDC Vital Power System (inCIUdC5 Oial bO)

a. Provides control and instrumentation power primarily to turbine driven auxiliary feedwater pump,

b. Provides emergency lightint,.

c. Provides instrument power to monitor temperatures and pressures within the reactor coolant system and for level control for the AFW cystemx(and ic> moniior ihe, conclMcA Of th C-Ce r f., dnd R t.ut or Coo n a nt h stem p a r a m t.t e r $ ).

4.3 Steam Generator Safety Valves The operator should rely on the steam generator safety valves to dump steam from th9 steam generators and to transfer heat from the steam generators to the ultimate heat sink; in this case the atmosphere.

The steam generator safety valves require no operator action and no supporting systems to operate.

The tvsanoaily gera W dimos pheric. re(L'C{ yalVes 4 00 avail a blC., ii reguit td, fo r depr e % ur.

If steam generatoc 1 is used for heat removal:

(7,ing the RC,g, Unit 1 Unit 2 1-VLV-1-522 2-VLV-1-522 1-VLV-1-523 2-VLV-1-523 1-VLV-1 524 2-VLV-1-524 1 VLV-1-525 2-VLV-1-525 1-VLV-1-526 2-VLV-1-526 If steam gent.rator 4 is used for heat removal:

Unit,_1 Unit 2 1-VLV-1-527 2-VLV-1-527 1-VLV-1-528 2-VLV-1-528 1-VLV-1-529 2-VLV-1-529 1-VLV-1-530 2-VLV-1-530

}

1-VLV-1-531 2 VLV-1-531 f.)

p%

JJOTE:

SG -PGRVs will be available but are not required to rnet station blackout requirements.

g

SH1:7- %~I O;:

/03 SQ g,c g, Appendix J j

Prepared

^?'?' d ed Date Mw Reviewed _,/A b d DateJ)f/r+

.s 4.4 Auxiliary Feedwater System pump The operator wili use the turbine driven portion of the auxiliary feedwater system to cupply water from the condencate storage tanks to the appropriate steam generator.

AFW turbine driven pump (includes FCV-1-51, FCV-1-52, and associated controls).

For Unit 1, 125 VDC Battery Board 111 (normal feeder) is required.

For Unit 2 125 VDC Dattery Board I (normal feeder) is requit ed.

4.5 Eleam Supply to AFW Turbine Driven pump The operator will use these valves to supply steam to operate the turbine driven AFW pump.

If steam generator 1 is used for heat removal:

Unit 1 Unit 2 1-FCV-1-15 2-FCV-1-15 1-VLV-3-891 2-VLV-3-891 1-FCV-1-18 2-FCV-1-18 1-FCV-1-17 2-FCV-1-17 Tf steam generator 4 is used for heat removal:

i h

Unit 1 Unit 2

/-

i-FCV-1-16 2-FCV-1-16 1-VLV-3-892 2-VLV-3-892 l

1-FCV-1-18 2-FCV-1-18 l

1-FCV-1-17 2-FCV-1-17 4.6 Steam Gene ator Level Control Valves The AFW steam generator level control valves are used to control the flow of AFW to the steam generator. The level control valves from l

the turbir.c driven pump to steam generators 1 and 4 are air-operated l

valves.

Initially, they can be controlled remotely using the l

cona.rol air in their associated air accumulators. When the accumulators are depleted, the valves will close and must then be controlled manually or the quantity of control air reserve must be increased.

If steam generator 1 is used for heat removal:

Unit 1 Unit'2 l

1-VLV-3-869 - locked open 2-VLV-3-869 - locked open 1-LCV-3-174 2-LCV-3-174 1-VLV-3-873 2-VLV-3-873 1-VLV-3-877 - locked open 2-VLV-3-877 - locked open s

• 1-LT-3-174

• 2-LT-3-174

,\\

1119L

- - ~

3 , [ ',_ 3 $

G:

m3

$QN,5[.,.'-

-Appendix J O<,

Da t e___ ? < a / a Prepared

-F 1

/

keviewed J////51 dh Mate _3/szy g

4.6 Steam Generator Level Control Valvec (Continued)

If steam generator 4 is used for heat removal:

Unit 1 Unit 2 1-VLV.3-870 - locked open 2-VLV-3-870 - locked open 1-LCV-3-175 2-LCV-3-175 1.VLV-3-874 2-VLV-3-874 1-VLV-3-878 - locked open 2-VLV-3-878 - locked open

$1-LT-3-175

• 2-LT-3-175

• These level transmitters are required prior to establishment of manual control of the steam generator level control valves.

4.7 AFW Condensate Supply The condensate storage tanks are the water supply for the AF%'

system.

Either condensate storage tank A or B can be used to supply water to the AFW system.

If condensate storage tank A is used:

Unit 1 Unit 2 0-VLV-2-504 - locked open 0-VLV-2-504 - locked open 0-VLV-3-800 - locked open 0-VLV 800 - lecked open 1-VLV-3-809 - locked open 2-VLV-3-809 - locked open 1-VLV-3-810 - check valve 2-VLV-3-810 - check valve s

If condensate storage tank B is used:

Unit 1 Unit 2 0-VLV-2-505 - locked open 0-VLV-2-505 - locked open 0-VLV-3-800 - locked open 0-VLV-3-800 - locked open 1-VLV-3-f309 - locked open 2-VLV-3-809 - locked open 1-VLV-3-810 - check valve 2-VLV-3-810 - check valve 4.8 Instrumentation I

The operator is required to both monitor and control the RCS heat removal process during the period a station blackout exists.

To

(

monitor the heat removal process the operator needs to know the conditions within the RCS.

To control the process, he must control the steam generator since the steam generator is controlling the heat removal process.

To monitor the heat removal p acess:

l l

Incore Thermocouples - 60, 54, 44, 41 - providea the operator l

with an indication of the core temperature.

j pressurizer Pressure - pI-68-323 - provides the operator with the

)

RCS pressure.

)J l

1119L

f cy l

jo3 SQN.LLO.cc; Appendix J i

Prepared _,'

• E i f,,

• Date g '/'4 Reviewed Jh /ftG~___Datey/,,/if 4.8 Instrumentation (Continued) 4 These two parameterc allow the operator to determine if successful core cooling is being accomplished.

To control the heat removal process:

Steam generator level indicatien allows the operator to control APW flew to match the decay heat level and control the heat removal process.

Steam generator 1 (if used) IcVel:

Unit 1 Unit 2 1-LI-3-38 2-LI-3-38 1-LI-3-42 2-LI-3-42 Steam generator 4 (if used) level:

Unit 1 Unit _2_

1-L1-3 106 2-LI-3-106 1-LI-3-110 2-LI-3-110 Steam generator precoure indication allows the operator to verify that the heat removal process is occurring properly.

Steam generator 1 (if used) pressure:

Unit 1 Unit 2 1-pl+1-5 2-p1-1-5 Steam generator 4 (if used) pressure:

Unit 1 Unit 2 1-pl-1-30 2-PI-1-30 8

J p

1119L

$ HEE 7 90 0.r.

/C2.3 Appendix J Prepared

'V~4 # M/ /

Date * <' A SQ N i20 ;,);;

Reviewed Afl/> n-ff-Datcy/,7/f r C

5.0 References; 5.1 Westinghouse Owners Group Energency Response Cuidelines, Background Volume E-0, ECA-0, High Pressure Version, Revision 1, Westinghouse Elect ric Corporation, September 1,1983.

5.2 Sequoyah Nuticar Plant Phase I Probabilistic Risk Asser.cment, Draft Report, Ecliability and Performance, Tennessco Valley Authority, September 1988.

5.3 Emergency Contingency Instruction - Loss of all AC Power, ECA-0.0, R3.

5.4 Abnormal Operating Instruction - Lous of Offsite Power, A01-35, R11, 5.5 Emergency Plan Implementing Procedure, EPIP-o, R1.

5.6 Guidelines and Technical Bases for UUMARC Initiativos Addressing Station Blackout at Light Water Reactors, NUMARC 87-00, Nuclear hanogement and Resources Council Inc., October 1988.

5.7 Sequoych Nucicar Plant Master Components List.

5.8. TVA Drawings:

47W400-1E, R5 M

47W400-2E, R3 l i, 47W415-1, R11 47W420-55, RO 47W427-1, R20 4 7W427-2, R18 i

45N230, R13 l

l-l 6.0 conclusions No conclusions are made.

This appendix describes a coping strategy.

The system capabilities and capacitics are described elsewhere in this calculation.

)J Il19L

SQ N E EG-C..

5"2 2 E H

/

O

_ c.:

Appendin J Prepared w.. "," u, Date__,ec j's /L y f f ---Date g g g Reviewed A*

Table J.1 EQUIPMENT LOCATIONS Component Location

• 125 VDC Vital Battery I (0-BATB-249-QV) 749-A4Q 125 VDC Vital Battery Board I (0-BDG-249-KE) 734-A4Q 125 VDC Vital Battery II (0-BATB-249-QW) 749-A5Q 125 VDC Vital Battery Boat d II (0-BDG-249-KF) 734-A5Q 125 VDC Vital Battery III (0-BATB-250-QX) 749-A11Q 125 VDC Vital Battery Board III (0-BDG-250-KG) 734-A11Q 125 VDC Vital Battery IV (0-BATB-250-QY) 749-A12Q 125 VDC Vital Battery Board IV (0-BDG-250-KH) 734-A12Q 1-VLV-1-522 (2-VLV-1-522) 735-WMSVR 1-VLV-1-523 (2-VLV-1-523) 740-WMSVR 1-VLV-1-524 (2-VLV-1-524) 735-WMSVR 1-VLV-1-525 (2-VLV-1-525) 740-WMSVR 1-VLV-1-526 (2-VLV-1-526) 135-WMSVR 1-VLV-1-52 7 (2-VLV-1-52 7) 740-WMSVR 1-VLV-1-52S (2-VLV-1-528) 735-WMSVR 1-VLV-1-529 (2-VLV-1-529) 740-WMSVR jk j

1-VLV-1-530 (2-VLV-1-530) 735-WMSVR 1-VLV-1-531 (2-VLV-1-531) 740-WMSVR s

l Unit 1 Turbine Driven AFW Pump (1-FMP-3-142) 669-A1T l

Unit 2 Turbine Driven AFW Pump (?-PHP-3-142) 669-A15T 1-PCV-1-51 669-AIT 2-FCV-1-51 669-A15T 1-FCV-1-52 669-A1T 2-FCV-1-52 669-A157 l

1-FCV-1-15 (2-FCV-1-15) 725-WMSVR l

1-VLV-3-891 (2-VLV-3-891) 706-WMSVR l

1-FCV-1-16 (2-FCV-1-16) 725-WMSVR 1-VLV-3-892 (2-VLV-3-892) 706-WMSVR 1-l'CV-1-18 (2-FCV-1-18) 725-WMSVR 1-FCV-1-17 (2-FCV-1-17) 725-WMSVR 1-VLV-3-869 (2-VLV-3-869) 706-WMSVR 1-LCV-3-174 706-WVR i

2-LCV 174 723-WVR 1-VLV-3-873 (2-VLV-3-873) 706-WMSVR 1-VLV-3-8 7 7 (2-VLV-3-8 7 7) 706-WMaVR 1-VLV-3-870 (2-VLV-3-870) 706-WMSVR

• Locations are given as 'clevation-coordinates' or 'clevntion-room' where:

WMSVR = West Main Steam Valve Room WVR = West Vault Room

.a_)

1119L

NIY [ CO 0.'.

.I

?

Appendix J Prepared

'7 $ 'E$ /7 iso Date 44V4,,# t-Da te_?47/rr Reviewed 3

- /

Table J.1 (Continued)

EQUIPMENT LOCATIONS Component Location

• 1-LCV-3-175 706-WVR 2-LCV-3-175 723-WVR 1-VLV-3-874 (2-VLV-3-874) 706-WMSVR 1-VLV-3-878 (2-VLV-3-878) 706-WMSVR 0-VLV-3-800 690- A15Q 1-VLV-3-809 669-A1T 2-VLV-3-809 669-A15T 1-VLV-3-810 669-AIT 2-VLV-3-810 669-A15T 0-VLV-2-504 685-T15K 0-VLV-2-505 685-T15K 1-LT-3-174 697-A2 72o 30' 2-LT-3-174 699'2"-AZ 78o30' 1-LT-3-175 697-AZ 256030' 2-LT-3-175 712'9"-AZ 760 i

• Locations are given as 'clovation-coordinates' or 'clevation-room' where:

WMSVR = West Main Steam Valve Room WVR = West Vault Room The.

12Ov Wtal AC. Inveyttrs and Wtal AC Powcr

' Boa rd s aru. to be addtc( to ttss (t'st.

Inve,v tcr s are located in 4he Aux IblCl3 a f EL ~I49 in the l

48c v Board Rcoms.

The AC Powcc Ecards. a r e.-

loca ted ih f he. AssoCt a led Batt e ry hard rooms,

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j Key Input Parameters for Sensitivity Studies With I

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- Station Blackout Similarities

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REAT LOADS RCS ambient heat loss

.a modelled as a slab with RCS natural circulation cooldown

. pressurizer cooldown rate 4.152 F/hr U

RCS cooldown rate 7.92 F/hr upper compartment electrical heat loads cooler fan motors (3) 114,525 BTU /hr lighting 12,799 BTU /hr lower compartment electrical heat loads-CRDM fan motors (2; 426,033 BTU /hr lighting 13,311 BTU /hr primary concrete shield

-15,000 BTU /hr cooler fan motors (3) 458,460 BTU /hr y

RCP motor (4) 3,000,000 BTU /hr CRDM motor 280,000 BTU /hr RCS mass release n

Time (sec)

Mass (lbm/sec)

Energy (BTU /lbm) **

J

~

0.0 18.0 576.0 1550.0 14.0 576.0 1551.0 0.0 576.0

-3400.0 0.0 576.0 3401.0 14.0 567.0 3600.0 14.0 567.0 3601.0 0.0 567.0

-7000.0 0.0 567.0 7001.0 15.0 552.0 7200.0 15.0 552.0 14400.0 0.0 552.0 14401.0 16.0 523.0 i

corresponding estimated liquid enthalpy at upper head liquid temperatures COMPARTMENT VOLUMES AND INITIAL TEMPERATURES upper compartment 651,000 ft 117 cf 3

3 O

lower compartment 248,388 ft 189 F

3 U

dead ended compartments 129,900 ft 120 F

3 0

1 annulus 375,000 ft 108 F

3 ice compartment 110,521 ft 35 F

3 ice upper plenum 54,940 ft 42 F

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