ML18283B549

Emergency Core Cooling Systems, Low Pressure Coolant Injection Modifications for Performance Improvement

ML18283B549
Person / Time
Site:	Browns Ferry
Issue date:	10/31/1977
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
Download: ML18283B549 (44)

Text

BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 2 EMERGENCY CORE COOLING SYSTEMS I'OW PRESSURE COOLANT INJECTION

'ODIFICATIONS FOR PERFORMANCE IMPROVEMENT OCTOBER 1977

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TABLE OF CONTENTS 1

2.0 0 INTRODUCTION BACKGROUND

................... Pacae 2

3 3 0 DISCUSSION 4 3.1 Accident Descri tion 4 3.2 Modification 5

3. 2. 1 Suction Line Break . 0 C 0 3.2.2 Dischar e Line Break 0 0 0 3.3 Model A lication . 9 3.4 Safet Anal sis 9

3. 4. 1 ui ment Ca abilit to Perform as Anal zed . 10 3.4.2 ui ment Interfaces 14 3.4.3 Functional Interface 0 0 ~ 0 0 17 3.4.5 'alit 3.4.4 Satisfaction of A ro riate S tandards Assurance and Control 17 18 4 0

SUMMARY

AND CONCLUSIONS....... 19

5.0 REFERENCES

. . . - 20

LIST OF TABLES Table Title ECCS Pump Configuration LIST OF ILLUSTRATIONS

~pi ate Title System Normal Operation System Mode of Operation During Unit 1 LOCA (Suction Line Break) No Failures System Mode of Operation During Unit 1 LOCA (Suction Line Break) RHR Injection Valve Failure System Mode of Operation During Unit 1 LOCA (Suction Line Break) Diesel Failure System Mode of Operation During Unit 1 LOCA (Suction Line Break) Battery Failure System Mode of Operation During Unit 1 LOCA (Suction Line Break) Opposite Unit Spurious Accident Signal 7 System Mode of Operation During Unit 1 LOCA (Discharge Line Break) No Failure System Mode of Operation During Unit 1 LOCA (Discharge Line Break) RHR Injection Valve Failure System Mode of Operation During Unit 1 LOCA (Discharge Line Break) Diesel Failure 10 System Mode of Operation During Unit 1 LOCA (Discharge Line Break) Battery Failure System'ode of Operation During Unit 1 LOCA (Discharge Line Break) Opposite Unit Spurious Accident Signal 12 Existing System Valve Bus Arrangement 13 Modified System Valve Bus Arrangement System Valve Control Power Arrangement 15 Modified System RHR Pump Divisional 11

'1 F

Priorities (recirculation pump discharge, LPCX injection, RHR pump minimum flow bypass

1 0 INTRODUCTION t

Browne Ferry Emergency Core Cooling System (ECCS) design and performance for Units 1 and 2 have been the subject of a recent review. 'his review led to a change in the system, which provided a significant reduction in the peak cladding temperature following a postulated recirculation line break. This reduction

'I in peak cladding temperature has been accomplished by elimination of the Low Pressure, Coolant Injection (LPCI) System recirculation loop selection and keeping the Residual Heat Removal (RHR) cross-tie valve closed. A report on this previous modification was submitted to the Nuclear Regulatory Commission in a letter from J.,E. Gilleland to Benard C. Rusche dated Fabruary 12, 1976.

Portions of that previous report are presented 'here to give a coherent description and safety analysis.

The proposed additional modification changes the power supply to the recirculation pump discharge valves, LPCI injection valves, the RHR pump minimum flow bypass valves, and RHR test isolation valves. .The change also modifies independent, valve AC power supplies to eliminate'RC concerns on paralleling of AC power supplies.

Major areas of discussion in this report include the proposed power supplies and a detailed safety analysis of the

'ndependent modification.

2.0 BACKGROUND

With the advent of the Interim Acceptance Criteria, it became

, advisable to consider the simultaneous occurrence of spraying and flooding to meet the stringent new temperature limit of 2300RF.

The thermal-hydraulic models were. refined to permit. an accurate

(

'alculation of coolant remaining in the vessel following the blowdown, and of spray coolant reaching the lower plenum after the boiloff which takes place as it passes through the active fuel region. These refinements permitted an accurate calculation of the flooding rate due to spray operation, and even with the new requirement of an active component failure anywhere in the

. ECCS, no jet pump BWR failed to meet the Interim Acceptance Criteria. ECCS modifications which might have been suggested by

.the new evaluation models were therefore unnecessary.

The final ECCS acceptance criteria adopted by the AEC are more conservative than the interim acceptance criteria. These new criteria reduce operating flexibility and could result in power level restrictions. To offset the effect of the new criteria, a modification has been added to Units l and 2 which takes advantage of the credit given for the flooding effect achieved through the availability of additional LPCI pumps under certain single-failure conditions. TVA commited to modify the power F

supply to the recirculation pump discharge valves, LPCI .injection valves, and RHR pump minimum flow bypass valves before return to power operation following the second refueling outage of the respective units.

3>> 0 DISCUSS ION.

3. 1 Accident Descri ion The Design Basis Accident (DBA), Loss-of-Coolant Accident (LOCA),

is one of several hypothesized events used to evaluate the ability of the plant to operate without undue hazard to the health and safety of the public.

The overall initial assumptions remain as described in Section 14.6.3. 1 of the FSAR:

The reactor is operating at the most severe condition at the time of the LOCA, which maximizes the parameter of interest:

primary containment response, fission product release, or core standby cooling system requirements.

A complete loss of normal AC power occurs simultaneously with

.the LOCA. This additional condition results in the longest delay. time for the core standby cooling systems to become operational.

The LOCA assumes that a recirculation loop pipeline is instantly severed. This results in the most rapid coolant loss and depressurization with coolant discharged from both ends of the break.

3.2 Modification Modification of the system requires the following hardware and

.wiring changes on Units 1 and 2:

The auto-transfer feature of, valve motive power for recirculation pump discharge, LPCI injection, RHR pump minimum flow bypass valves, and RHR test isolation valves will be isolated from redundant divisional power supplies by motor-generator (M/G) sets 3.2.1 Suction Line Break Effects Figure 2 illustrates operation of the modified system for a break in the recirculation pump suction line. The break location producing the highest peak cladding temperature is, as before, at the nozzle on the pressure vessel. The other side of the postulated "double-ended" break is fed through the recirculation loop by the jet pump no zles, whose small area limits flow to a low value and makes frictional losses negligible in the calculation.

The discharge valves of the recirculation loops will begin closing upon receipt of a permissive signal., The valves are capable of closing against a differential pressure of 200 psid.

To assure the recirculation system discharge valve is not required to close with a differential pressure greater than 200 psid, valve closures are delayed until reactor vessel pressure has decreased to less than 225 psig., By the time the

recirculation discharge valve has stroked sufficiently that it could present a flow-limiting restriction, the vessel pressure will have decayed below 200 psig.

Valve closure is therefore effected in about 62 seconds, of which 29 seconds represents the reactor vessel pressure permissive and 33 seconds the maximum valve closure time. The effect is isolation of the break from the LPCI system injection point.

Approximately 46 seconds after the break, the LPCI startup sequence .is complete and flow commences in both loops. Flow into the broken loop will not, reach its expected value for an additional 16 seconds, when the recirculation discharge valve has fully closed. The RHR pumps go nearly to full runout flow, as limited by the additional resistance in the pump discharge 'line, because each pair of pumps is. delivering flow to its own bank of jet pump nozzles rather than to one bank as would be the case. of loop selection logic. Additional resistance has been added to the RHR pump discharge lines. This replaces the resistance lost when, only one or two pumps are discharging into a system designed for three pump flow. The added resistance prevents insufficient Net Positive Suction Head (NPSH) in these modes of operation.

In analyzing the single failures for a suction line break, both AC and DC power failures are considered (see Figures 4 and 5).

For A8 power considerations the most significant single failure for the modified system is a 480-V Reactor Motor-Operated Valve (MOV) Board failure. This failure results in two RHR pumps

operating in one loop and four CS pumps operating in two CS systems.'

The most significant DC power single failure would be loss of a battery. For a suction line break this failure results in two RHR pumps operating in one loop, one RHR pump operating in the alternate loop, and two CS pumps operating in one CS system Table 1 shows the various pump combinations for postulated single failures.

The unique power arrangement at Browns Ferry Units 1 and 2 requires examination of an opposite unit spurious accident signal. For this event one RHR pump in each loop of each reactor and one core spray system (two pumps) plus all required valves are available.

The limiting single failure is that failure which results in the longest reflood time and consequently the highest peak cladding temperature (PCT). Sensitivity studies have been performed which demonstrate that a typical limiting failure in the modified system is the failure of the RHR injection valve in the unbroken loop. This failure results in four core spray pumps, two in each CS loop, and two RHR pumps in one loop providing ECCS flow to the core. This combination gives a longer reflooding time than one core spray system (two pumps) and one RHR pump in each loop which is available following an opposite unit spurious accident signal.

This is due in part to the effects of counter current flow limiting (CCFL) on the amount of the core spray flow available 1

for reflooding. The assumed occurrence of CCFL results in there

being only a slight improvement with four CS pumps when compared to two CS pumps.'dditionally, the two RHR pumps feeding into one loop deliver significantly less than twice the flow delivered by a single pump feeding each loop due to the system orificing effects. Thus the availability of one RHR pump in each loop for the alternate unit'spurious accident signal provides better reflood characteristics than two RHR pumps into one loop even when supplemented by two additional CS pumps.

3.2.2 Dischar e Line Break Effects Figure 7 illustrates the operation of the modified system with a break in the recirculation pump discharge .line.

When the RHR startup sequence is complete, the RHR flow in the broken loop is. lost through the break. With the modification, the worst-case single failures are failure during opening of the RHR injection valve opposite the break, failure during opening of the RHR pump minimum flow bypass valve serving the RHR pumps intended for injection into the unbroken loop, and failure of a 480-V Reactor Motor-Operated Valve (MOV) Board. Table 1 and Figures 8-11 show the pump combinations which result from the postulated single failures.

The suction line break remains the design basis accident for the modified system, but with a lower calculated peak cladding temperature.

A typical limiting single failure for the discharge line break is the RHR in)ection valve failure. This failure results in four core spray pumps available for core ref looding. This condition results in a longer reflood time than the opposite unit spurious accident signal in which two core spray and one RHR pumps are available for reflooding. As previously discussed one RHR pump provides faster reflooding and, consequently lower PCT than two additional CS pumps.

The present Browns Ferry Units 1 and 2 system utilizes two power supplies for the electrical distribution system providing power to the RHR valves. Figure 12 shows the arrangement of the buses and the valves fed from these buses. Figure 13 shows the modified system which isolates the auto-transfer feature from the electrical distribution system by use of motor-generator sets.

3.3 Model A lication The core heatup calculations are performed using the approved Appendix K emergency core cooling evaluation models.

3.4 'Safet Anal sis The proposed modification has been analyzed and evaluated to assure the changes do not introduce adverse effects to the overall plant. The areas evaluated are discussed in the balance of this section.

3.4.1 ui ment Ca abilit to Perform as Anal zed The major components of the proposed modification are unchanged, except for the power supplies to selected valves. Each major element is considered below:

3.4. 1. 1 Emer enc Diesel-Generators The proposed modification adds the running load of the M/G sets to the operating requirements of the diesel generators This addition does not adversely affect diesel operation and diesel loading remains within design specifications.

'I The operating modes of the RHR pumps were changed by the previous modification such that two pumps discharge to each injection header thereby changing the discharge flow characteristics from that previously established. Prior to reactor startup after the previous modification, flow tests were conducted to establish the pump discharge path characteristics from which pump flow curves were developed. This information was used to determine the additional resistance to be added on the dis'charge side 'of each pump to ensure satisfaction of pump Net Positive Suction Head (NPSH) requirements.

3.4.1.3 Control Circuitr All standards for engineered safeguards control equipment are maintained., Additional relays and wiring have been added to assure single-failure capability.

3.4.1.4 Recirculation Loo E ualizer Valve and RHR S stem Cross-Tie Valve Inadvertent opening of these valves could negate the RHR system injection when needed, therefore one equalizer valve and the cross-tie valve were closed and motive power removed by the previous modification. An annunciator was added to provide redundant, indication when the RHR cross-tie valve and/or equalizer valve are not fully closed.

3.4.1.5 Recirculation Pum Dischar e Valves Closure of the recirculation pump discharge valves is of importance to the proper application of the proposed modification. Three aspects of valve compatibility have been investigated:

3.4.1.5.1 Environment As reported in Section 5.2 of the Browns Ferry FSAR, the recirculation system valves are designed to operate under the environmental conditions associated with the DBA-LOCA.

3. 4. 1. 5. 2 Break Ef fects A study of the drywell geometry was performed prior to the previous modification to determine the effects of jet impingement resulting from a postulated recirculation line break. For the break effects study, breaks were assumed at,all terminals, branch lines, and at other locations based upon stress. Breaks were assumed at all locations where pressure plus dead load plus thermal plus earthquake stresses exceed 0.8(1.2$ +SA).

Additionally, in piping runs where no stresses occur in excess of

0. 8 (1. 2++@ ), a'inimum of two intermediate breaks were postulated based upon the highest total stresses combined as above.

For the suction line break, re-routing of cable has been provided, to ensure recirculation discharge valve operation.

Valve closure at the time of a discharge line break is not considered in the ECCS analysis. Also, closure of the discharge valve does not change the RHR system input capability during a discharge line break (See Figure 7).

3.4. 1.5.3 Valve Differential Pressure Recirculation pump discharge valve closure requires both a LOCA initiation signal and a decrease in reactor pressure to the permissive setting. With valve closure initiation delayed until reactor pressure has decayed to less than 225 psig (approximately 29 seconds) the differential pressure across the closed valve will always be less than the maximum 200 psid. The sensor and permissive circuitry are designed to satisfy all requirements for engin>ered safeguards control systems.

3.4 1.6 RHR Pum Minimum Flow B ss Valve RHR pump minimum flow bypass valves will,be provided'with redundant power supplies and control logic to assure maximum pump protection under postulated accident conditions.

3. 4. 1. 7 Batteries The proposed modification does not change any of the operating requirements on the station batteries.

3.4. 1.8 Motor-Generator M/G Sets

.Qualified motor-generator (M/G) sets will be used as isolation devices for the power feeds to the 480-V Reactor MOV boards with auto-tran'sfer feature (see Figure 13) . Although only one M/G set will normally supply power to the Reactor MOV board, both M/G sets will run at all times to assure readiness of 'the alternate M/G to accept load, if,required. M/G set's will be sized to accept valve loads at any time during the initiating event (62 secs.). Fire protection to the M/G sets will be evaluated and appropriate measures taken to assure adequate fire protection.

3,4. 1. 8. 1 Seismic ualification The operability of the motor-generator sets and all the appurtenances vital to their operation during and after a SSE is verified in accordance with IEEE 344 as applicable to the plant.

3.4. 1.9 RHR Test Isolation Valves he RHR test isolation valves will be provided with redundant power supplies and closing control logic to protect against those postulated occurrences which would leave the valve open and route an unacceptable amount of flooding water away from the core.

Redundant controls are wired. as specified in 3.4.2.3.

3.4.2 ui ment Interfaces The effects of the proposed change on the various operating modes of the equipment have been evaluated and found to be acceptable, as described below:

3.4.2. 1 Emer enc Diesel-Generators The proposed modification introduces no new or different interfaces for this equipment.

I 3.4.2.2 Valve Power Motor operated valve (MOV) boards for those valves necessary for automatic operation for LPCI injection (recirculation pump discharge, LPCI injection, RHR pump minimum flow bypass, and RHR test isolation) will interface with divisionalized 480 VAC shutdown boards with M/G sets as isolation devices. Two 480-V AC Reactor MOV boards will be divisionalized and supply power to those valves necessary for the RHR function (see Figure 13).

These valve boards will have an auto-transfer feature for redundant power but will be protected from the redundant .

divisional source by using M/G sets as isolation devices. The auto transfer feature will be eliminated from all valve boards not. protected by M/G sets.

3.4.2.3 Valve Motor Control To ensure that a malfunction in the individual valve controller does not couple back to the other valve control circuits, the redundant A and B circuits were provided separate relays and contacts in the logic panels on a previous modification. This separated, redundant arrangement has been applied to the RHR and recirculation system valves needed for operation as described.

System interfacing and protection as related to the valve motor control centers are unchanged.

3.4.2.4 DC Control Power As shown in Figure 14 and Browns Ferry FSAR Figure 8.6-3, 250 VDC from the station batteries provides control power to RHR logic panels. After the proposed modification the same equipment receives power from this source as in the original design.

Failure of any one station battery does not cause interaction's that exceed the limiting case for core cooling capabilities.

3.4.2.5 RHR Lo ic Panels To provide the necessary redundancy required on the previous modification, changes were made to the RHR logic panels. To preclude valve-to-valve interface, redundant and separate relays and contacts were provided for each RHR and recirculation system.

Each of the added redundant relays was provided full separation from all others by enclosure in a metal box. The wiring from redundant contacts between the two logic panels was provided separation by enclosure in flex conduit and termination in metal junction boxes. This logic scheme will be maintained in the new modification.

3..4.2.6 Motor-Generator (M/G) Sets Two M/G sets will interface between each board with the auto-transfer feature and shutdown boards. These M/G sets provide isolation of the boards from redundant sources.

3. 4. 3 Functional Interf ace The RHR system, as discussed in this report, performs as a short-term post-LOCA core cooling function. The system also provides a long-term containment cooling function'which is described in Sections 4. 8.6.2 and 14.6.3.3.2 of the FSAR. The

. effects of the proposed change to the core'ooling function on the containment cooling function were evaluated and found to be acceptable after modification as described below.

In analyzing single failures which might influence long-term suppression pool cooling, both AC and DC control and emergency power failures as well as component failures in the RHR and'RHRSW (cooling water) systems were considered. The worst case single failure (Reactor MOV Board lass) with the modified system still leaves two RHR heat exchangers, two RHR pumps, and two RHR Service Water pumps and associated valving available for suppression pool cooling. The suppression pool temperature versus time response for this combination of equipment is shown by curve C in FSAR Figure 14.6-12.

. 3.4.4 Satisfaction of A ro riate Standards The proposed modification directly affects as Engineered Safeguards System and has been designed to Class I system standards. The standards and guides which were applicable to the original design have been reviewed to assure the modified system design, equipment, and installation meet or exceed the qualifications of the unmodified system.

3. 4. 5 'alit Assurance and Control Quality assurance and control will be applied to this modification as detail'ed in Appendix D of the Browns Perry FSAR.

Appendix D incorporates the requirement of 10CPR50, Appendix B.

4.0

SUMMARY

AND CONCLUSIONS The proposed modification involves some physical changes to the plant to permit divisional isolation of the auto-transfer concept and adoption of the total system availability of the new design.

The analytical methods used reflect the most recent

'determinations of NRC staff and reactor suppliers for modeling the performance of Emergency Core Cooling Systems.

The application of the proposed modification adds to the overall capability of the plant to continue operation in a manner that ensures the health and safety of the public while providing benefit in the production of electrical energy.

-1 9-

5 0 REFERENCES

1. Interim Policy Statement, USAEC, dated June 19, 1971;

Subject:

AEC Adopted Interim Acceptance Criteria for Performance of, ECCS for Light-Water Power Reactors.

2. Letter from J., E. Gilleland (TVA) to Benard C. Rusche (NRC) dated February 12, 1976.

t TABLE 1 ECCS PUMP CONFIGURATION Suction Side Break Pum s Available++

No Failures 4 Core Spray, 2 RHR in each Loop Opposite Unit Spurious Accident 2 Core Spray, 1 RHR in each Loop Signal RHR Injection Valve Failure+ 4 Core Spray, 2 RHR in one Loop RHR Minimum Valve Failure+ 4 Core Spray, 2 RHR in one Loop Recirculation Discharge Valve 4 Core Spray, 2 RHR in one Loop Failure-Break Side* 'I 480 V Reactor MOV Board+ 4 Core Spray, 2 RHR in one Loop Diesel Failure 2 Core Spray, 2 RHR in one Loop, 1 RHR in other Loop Battery Failure 2 Core Spray, 2 RHR in one Loop, 1 RHR in other Loop Dischar e Side Break Pum s Available+4 No Failures 4 Core Spray, 2 RHR in one Loop RHR Injection Valve Failure~ 4 Core Spray RHR Minimum Flow Valve Failure+ 4 Core Spray Diesel Failure 2 Core Spray, 1 RHR Battery Failure 2 Core Spray, 1 RHR Opposite Unit Spurious Accident 2 Core Spray, 1 RHR Signal 480 V Reactor MOV Board+ 4 Core Spray

• Limiting Single Failure

+*In Unbroken Loop

DIV I DIV II D/G A 0/G 8 D/G C 0/G 0 lA lA 2A 1C l ~~~

1C 2C 2C 1

18 18 28 28 ID 10 2D 2D CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 DISCH SUCTION DISCH NOT RUNNING RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figure 1 System /t/ormal Operation

DIV I DIV II 0/G A 0/G 8 DIG C 0/G 0 1A 1A 2A 2A 1C 1C 2C 2C 18 18 28 28 1D 10 2D 2D L C g~< i CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI'A LPCI 8 BREAK DISCH SUCTION DISCH DISABLED OR NOT RUNNING RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figure 3 System Mode of Operation During Unit 1 LOCA (Suction Line Break) LPCI Injection Valve Failure

OIV I OIV II DIG A OiG 8 OlG C DiG D 1A t

0 4 q IAI I

2AI

~ 2A I 1C 'IC I 2CI 2C'

~ o-4 r 18 t IBI o

2BI 28l 10 ID I

20 ghC',

c $C,:. L C,'C:- sf"qi L C 20'ROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 BREAK D ISCH SUCTION 0 ISCH DISABLED OR NOT RUNNING RECIRC 8 RECIRC A RECIRC 8 RECIRC A Eigure System Mode of Operation During Vnit t LQCA (Suction Line BreakJ. No Eaj/ures

OIV I DIV II D/G A D/G 8 0/G C D/G 0 I 1 1A 1A 2A 2A 1C 1C 2C 2C 18 18 28 28 1D 10 2D 20 L C CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 BREAK OISCH SUCTION DISCH

,1.j:g DISABLED OR NOT RUNNING RECIRC 8 R ECIRC A RECIRC 8 RECIRC A Figure 5 system Node of Opert tron During Unit 1 LOCA fSuction Line Break J Battery Fa//ure

OIV I 0 IV I I 0/G A OlG 8 OIG C 0/G 0 1A 1A 2A 2A 1C 1C =

2C 2C 18 18 28 28 10 I M 10 20 20 I

L C -;;C'.~ ~4~, L C CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 BREAK 0ISCH SUCTION 0ISCH 0 ISA8 LEO 0 R NOT R UN N IN G RECIRC 8 RECIRC A RECIRC 8 RECIRC A figure 4 System Node of Operation Quring Unit 1 LOCA (Suction Line Breakj Qiesel failure

DIV I DIV II 0/G A 0/G 8 0/G C 0/G 0 1A lA 2A 2A 1C

~ mp-4~~

1C 2C 2C 1B 1B 2B 2B 1D I M 1D 2D 2D 1

g4a C c L C L C CROSSTIE CROSSTIE LPCI A LPCI B LPCI A LPC I B OISCH SUCTION DISCH

@j:;.'c. DISABLED NOT RUNNING OR NOT CONSIDERED IN RECIRC 8 RECIRC A ANALYSIS RECIRC 8 RECIRC A f Figure 7" Sjetem Node of Operation During Unit / LOCA (Dia/ra/ye Line Break),No Fai/um

OIV I 0 IV II DiG A 0/G 8 DiG C OiG D 1A 1AI 2AI 2A I 1C 1C I 2CI 2C 1B I o~

IB I l

2B I 2B I P

1D 10 I 20 I

20' L C C .

~ L C 4.'.-: L

+~

CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 BREAK 0ISCH SUCTION 0ISCH

~&+>?N "Xr';~, DISABLED OR NOT RUNNING RECIRC B RECIRC A RECIRC B RECIRC A Figure 6 $pstem Mode of Opere tion During Unit 1 LOCA (Suction Line Break J Oppoute Unit Spurt'aa Accident Signei

DIV I DIV II 0/G A 0/G 8 0/G C D/G 0 1A 1A 2A 2A IC 'C 2C 2C 18 18 28 28 1D 10 2D 2D I

CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 DISCH SUCTION D ISCH DISABLED, NOT RUNNING.

OR NOT CONSIDERED IN ANALYSIS RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figun.' - System Mode of Operation During Unit / LOCA (Discharge Line BraakJ Diesel Failure

DIV I DIV II 0/G A 0/G 8 0/6 C 0/G 0 1A 1A 2A 2A 1C 1C 2C 2C 18 18 28 28 10 10 20 2D C

CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPC I 8 DISCH SUCTION D ISCH DISABLED, NOT RUNNING OR NOT CONSIDERED IN ANALYSIS RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figure 8 System IVIode of Operation During Unit 1 LOCA (Discharge Line BreakJ LPCI Injection Valve Failure

DIV I = DIV II

~ D/G A 0/G,B DIG C 0/G D I -f.

1A'A '0 C,

I 1C 1C 2C 2C 18 18 28 28 I 1D 20 L L i C CROSSTIE CROSSTIE LPCI A ug LPCI 8 LPCI A LPCI 8 C

DISCH SUCTION DISCH DISABLE, NOT RUNNING OR NOT CONSIDERED IN ANALYSIS RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figure SYstem Mode of Operation During Unit 7 LOCA fDischarye Line BreakJ Opposite Unit Spurr'ous Accident Signal

0)V I DIV II 0/G A DIG 8 DIG C 0/G 0 1A I

1A 2A 2A I

1C l ~~&

-1C 2C 2C 7

18 18 28 28 1D

'I 1D 2D 20 C

CROSSTIE CROSSTIE LPCI A LPCI 8 LPCI A LPCI 8 0ISCH SUCTION OISCH

<<X~XX~~4 DISABLED, NOT RUNNING OR NOT CONSIDERED IN ANALYSIS RECIRC 8 RECIRC A RECIRC 8 RECIRC A Figure $0 SyatenI Node of Operation During Unit 7 LOCA (Discharge Line Breakf Battery Failure

O/G A DIG C OlG 8 DJG D Div 'I OIV II DIV I OIV II 4 kV SHNDII ED A'

'kV SSTDH

-'.HD.'C .

'O 4 kV SHTDN HD H NO 4 hV SHTDN HD D '0

'IA 1A; -. 2A 2A 18 - 18 28 28 1C 1C 2C 2C 10 1D 2D 2D L L C C L C C I.

)HC )Nc UNIT1 . NC NC VNIT2 4SOV SHTON 480V SHTON 480V SHTDN 480V 6HTON BD 1A BD 18 BO 2A BD 28

) NC )Nc 'NC )NC )HC )Bc )Nc

)Hc Nc .

IIO )NO

) )

Nc )NC 0 480V RX MOVBD 10 )

)NO480V RX MOV BO 1C 480V RX MOV BD 2D /

gPSOV RX MOV 80 2C

)NG)NG )Nc ) Nc )Hc )VC NC NC

) Nc )Nc 'C'C )bc )Hc 2458 10.16A 243A 0-258 2%58 '10.16A . 243A 10-258 2-538 10-25 A 245A lo-168 2438 10-25 A 245A 10 168 C IJ 0 z0 0 L

K~ O< I O IL 0 ~O IO O~

Qz z -- ELECTRICAL INTERLOCK CO I.

Figvre 32 Existing Syatem Valve 3us Arrangement

l. Valve closed and motive power removed.

D/G A D/G C D/G 8 O/6 O VXV SHTDX BD A No ) NRY SHTDN BD C No ) BD 8 blab DRY SHTDN No ) HXY SHTOX BD D XO }

J bbbb~: b bb b xc ) IIC )

%SOY SHTDH BD IA UNIT I OV SHTOH BD 8 NC) %80Y SHTDH BD 1A UNIT 2 HSOV SHTOH BD 28 KC ) IIC ) KC ) NC ) NC )Nc )Hc) No) N ) HC ) HC) HC ) NC) HC) HC) NC ) NC ) NC )

NC ) MY NC) xc ) RX RSDY xo )

RX MOV BD IA MOV BD 2A

)

b IH-l2A /75-25) X/G b

I 5-I2A (75-25)

Ho ) CORE SPRAY RSOV RX MOV BD IN).

IB X

) 2D-N No CORE SPRAY

%loY RX MOY Bo INJ 28 NC X/G 2D-A

)

X/G IE-A IO-l28 /75-53)

M/G IE-H M/0 2E-A .

-bIV-l28 /75-53l X/8 2E-N xo) COPE SPRAY IN3.'SOV xc) xo) CORE SPRAY INJ. NC)

RX MOV BO IC %SOY RX MOV SD 2C Nc NO CSOV RX MOY BD ID lIBOV RX MOV BO 20 FIG. IS MOD IF IED SYSTEM 2-538 (68-79) VALVE 2-538 (68-79) lo-3%A (?H-59) lo-f6A (7%-7) lo-25A (7%-53) l0-3%A,(?V-59) Io-16A (747) lo-25A (7%-53)

RECIRC DISCH RHR TEST ISOL RHR PMP: .LPCI IHJ. evs RECIRC DISCH RHR TEST ISOL RHR PMP LPCI INJECTION MINI-FLo ARRAHGEM HT MINI-FLO QSOV RX MOV BD IE HC NO )- %SOY RX MOV BD 2E NC)

UNITS I 4 2

)

2-53A (68 3) IO-3%8 (74-73) b Io-l68 (7%-30) Io-258 (? tI-67) K.2 53A (6S 3) Io-3%8 (7% 73) Io-168 (74-30) l0-258 (7% 67)

ReCIRC OISCH RHR TEST ISOL RHR PMP Mlxl-FLPf LPCI IHJECTIOH RECIRC DISCH. RHR TEST ISOL RHR PMo MINI FLOI LPCI INJ.

r NOTE 2 NOTES:

I LOGIC A I. THIS DIAGRAM IS FOR REPORT 250Y DC CLARIFICATION OIILY AHO IS POWER NOT IKTBDED TO REPRESENT THE BITIRE CONTROL SEE APPROPRIATE PS'RRANGEMEN<.

LOGIC A RNR ELEMENTAR. FOR COMPI.cTE LOCA WIRIHG LOCATIONS AND INTFRFACES.

LOW FLOW I20V AC DIV I FROM MOTIVE LOGIC A LOGIC A 2 GUARDED CIRCUITS OF ScPARATE RELAYS AHD CONTACTS IN LOG'lC LOGIC A I.OGIC A POWER RX PRESS RX PRESS PANELS - THE ONLY AREA WHERE RX PRESS RX PRESS <%50 <225 CIRCUITS INTERFACE

< N50 PSIG < 225 PSIG PSIG PSIG TEST TEST TEST PUMPS TEST TES TEST RELAYS ARE ENCASED IK SHEET

'UNNING l I STEEL. CABLING IS ROUTcD IN I'

I FLEX CCHDUIT: TEPAIHATIOHS I ARE IN METAL JUNCTION BOX.-

I I 3. ADMINISTRATIVE CONTROL I

L ON TEST FUNCTIONS.

LPCI INJ. RHR TEST REC IRC DISCN IIIN I-FLOW IO-25A l0-3UA 2-538 IO-IGA IO-258 IO-3%8 2-53A IO-IGB OPEN CLOSE CLOSE OPEN I

I I

I I

I TEST TEST TEST TEST TEST TEST PUMPS I PUNNING I LOGIC 8 LOGIC 8 LOGIC 8 LOGIC 8 PX PecSS RX PRESS RX PRESS RX PRcSS I < <50 PSIG < %50 PSIG < %50 PSIG < 225 PSIG I

120V AC LOW FLOW I

I FROM DIY Q I MOTIVE I

POWER I LOGIC 8 I

I LOCA I

I I

I I LOGIC 8 250Y DC I I I I POKER NOTE 2 FIGURE lq SYSTOI VALVE CONTROL POWER ARRAIIGEMENT

ir