ML20012E215

Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants,' for Peach Bottom.Response for Limerick Generating Station Will Be Provided by 900504

ML20012E215
Person / Time
Site:	Peach Bottom, Limerick
Issue date:	03/20/1990
From:	Helwig D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, NUDOCS 9004020057
Download: ML20012E215 (21)

Text

p;7 '10 CPR 50.54(f)

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PHILADELPHI A ELECTRIC COMPANY-NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBRCOK BLVD.

WAYNE, PA 19087 5691 (215) 640 6650 D AVID R. HELWlG -

vies raz.sount avet=== .an vies.

March 20, 1990-Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR-44 DPR-56 NPF-39 NPF-85 U.S. Nuclear Regulatory Commission

. Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Peach Bottom Atomic Power Station. Units 2 and 3 Limerick Generating Station, Units 1 and 2 Response to NRC Generic Letter 89-19, "haquest for Action

-Related to Resolution.of Unresolved Safety Issue A-47 Safety Implication.of Control Systems in LWR Nuclear Power Plants"

REFERENCES:

-(l) Letter from S. D. Floyd (BWPOG):to J. G. Par tlow ' (NRC) , - dated - F.:br uary 16, 1990 (2) Letter.from J. G. Partlow (NRC) to

.S. D. Floyd (BWROG), dated March 20, 1990 Gentlemen:

NRC Generic Letter 89-19 dated September 20, .'089, required licensees to provide, within 180 days of the date of the Generic Letter, a statement as to whether licensees will implement the

-recommendations provided in Enclosure 2 of the Generic Letter. The f Generic Letter also required licensees to provide a schedule for 1 implementation of the recommendations and the basis for the schedule Il L or to provide appropriate justification if the recommendations were not going to be implemented.

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L Attached 3s the Philadelphia Electric Company (PEco)

L response to the Generic Letter for Peach Bottom Atomic Power Station h I9004020057 900320 l PDR ADOCK 05000277 {h

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March 20, 1990 GL 89-19 Page 2-

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I -(PBAPS),10 nits 2, and 3.

-followed by PECO's response.

The,NRC recommendations are restated As outlined'in Reference 1, the Boiling Water Reactor Owners' Group (BWROG) is-addressing this issue on a generic-basis. Based

on our understanding of this work to date, we anticipate that the

.results will be. applicable to Limerick Generating Station (LGS)

Units 1 and D. 'PECo plans-to exercise the 45-day extension for

(! - 1 submittal of a' response as approved by the NRC in Reference 2. i Accordingly, the response'for LGS Units ll and 2 will be provided by May 4, 1990.

o If you have any questions, or require additional .

-information,,please contact us, Very truly yours, 3 '

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4 _ Attachments

-cc: :W. T~.. Russell, Administrator, Region I, USNRC J. J. Lyash, USNRC Senior Resident LInspector, PBAPS Kenny, USNRC Senior Resident: Inspector, LGS T. J.

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.,_ COMMONWEALTH OP PENNSYLVANIA a

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COUNTY OF,CHESTER  :

D. R '. Helwig, being first duly sworn, deposes and-says:

• That he'is Vice President of Philadelphia Electric Company, that he.has read.the response to Generic Letter No. 89-19, and knows.

the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

Vice Pres nt Subscribed and sworn to before me thisdO day of hM>e4 1990.

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NOTANIAL SEAL CAMAINE A MEICEZ Nctary Public j Tracyffrin Twp., Chester County ,

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Attachmsnt ic Peach Bottom Atomic Power Station, Units 2 and 3 Response to Generic Letter 89-19~

" Request for Action Related to Resolution of r Unresolved Safety Issue A-47 Safety Implication

• of Control Systems in LWR Nuclear Power Plants" NRC Recommendation-(1)(a)

'It'is recommended that all GE boiling-water-reactor (BWR) plant designs provide automatic reactor vessel overfill protection

'to mitigate main-feedwater (MFW) overfeed events. The design for i the overfill-protection-system should be sufficiently separate from the MFW control system to ensure'that the MFW pump will trip on a reactor high-water-level signal ~when required, even:if a loss'of power, a loss of ventilation, or a fire in the control portion of the-MFW control system should occur. Common-mode: failures-that-could disable overfill protection and the feedwater control system, '

but would still_ result in a feedwater pump. trip, are considered acceptable failure modes.

It is recommended'that plant designs with no automatic  :

-reactor vessel overfill protection be upgraded by providing a

-commercial-grade (or better) MFW isolation system actuated from at j least a 1-out-of-1' reactor vessel high-water-level system, or -

justify the design on some defined basis.

PECo Response The existing PBAPS' Reactor Feedwater Pump Turbine (RFPT) high level trip logic uses the feedwater control system to detect reactor vessel high level. There is redundant high level detection

, provided by the High Pressure Coolant Injection (HPCI) System. This H redundant HPCI high level trip provides adequate power supply independence, sensing line independence, separation of sensors, physical separation,~ indication and. alarming from the Feedwater Reactor. Level Control System. Details supporting this conclusion are provided in'the attached Design Analysis (Attachment 2 to this L

Generic Letter response). Therefore, it is concluded that the current design meets the recommendations of the Generic Letter, and no' plant modifications are necessary.

NRC Recommendation (1)(a) Continued, In addition, it is recommended that all plants reassess their operating procedures and operator training and modify them if necessary to ensure that the operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced pressure operation of the system.

[C' o Response

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.g Attcchm;nti Page 2 Peach Bottom Atomic Power Station does not have condensate booster pumps; however, the multi-stage condensate pumps installed t at PBAPS produce sufficient discharge pressure and capacity such-that a booster pump is not required. Our assessment assumes that '

the PBAPS condensate pumps perform similarly'to condensate booster pumps.

Reduced pressure operation, as it is discussed in the Generic Letter, is when Reactor Pressure Vessel (RPV) pressure is less than the shutoff head of the condensate pump, and, therefore, the condensate pumps can pump directly into the Reactor Pressure Vessel. .At PBAPS, the condensate pumps have a shutoff head of approximately 600 psi. Pressures of less-than 600 psi in the reactor vessel can only occur during the following three scenarios:

1) during a major plant transient, 2) startup or 3) shutdown. For each of these scenarios, PECo evaluated whether current operating 'j procedures and operator training were adequate to ensure that '

operators can prevent and mitigate reactor vessel overfill events due to condensate pumps during reduced pressure operation. Each scenario is addressed below:

Plant Transient

.The post scram scenario presents the most plausible condensate pump overfill possibility of the three scenarios in which condensate pumps are operated below 600 psi. In numerous industry scrams and several PBAPS scrams, pressure has dropped below 600' psi

-during the transient. The following two operating experience-documents were previously reviewnd by PECo and detail several e overfill events:

1. Information Notice (INN) 88-77: Inadvertent Reactor Vessel Overfill h

2. Institute of Nuclear Power Operations Significant. Event

' Report (SER) 5-88: Flooding Main Steam due to Improper Reactor Water Level Control l.

In response to these documents, PECo identified.and

) implemented the following corrective actions:

1

1. Revised existing Simulator Exercise Guides (SEG's) to

, incorporate some of the lessons learned from SER 5-88 and i-I' included a discussion of the possible consequences of filling-the main steam lines with water;

2. Included extensive training on level control during plant l- transients and in particular, during scram recovery in the six week simulator continuing training cycle that started 10/10/88;

c Attachment-1 Page 3

.3. Developed a Licensed Operator lesson plan (LOR 88-09A) for training the operators on various industry overfill events; and

4. Committed to revise the trip procedures to provide adequate direction to prevent vessel overfill from low pressure injection systems.

In January of 1989, Operations Section Performance Standard (OSPS)-4'was issued to evaluate Reactor Operator performance during a scram. OSPS-4 requires the operator to maintain level between 0 inches and 45 inches in order to receive an " excellent rating". An

" unsatisfactory rating" would result if RPV level reached 60 inches.

Programmatically, Reactor Operators are trained to trip the 'A' and

'B' RFPTs when level starts to recover following the scram. Level

. control is then established-via the 'C' RFP discharge valve bypass level' control valve. The feedpump discharge valves are subsequently shut. With the discharge valves shut, the only flow path to the vessel is through the 10 inch bypass valve. The rate of level increases due to condensate pumps is strictly limited. If level should increase, the following actions would sequentially occur.

a) The 'C' RPPT will auto trip at 45 inches (Note, the RPV high-low level alarm will already be present due to the level transient) b) In accordance with procedure OT-110, the operator will verify all RFPTs'are tripped, HPCI is tripped, and the RCIC ,

steam supply valve is shut.

c) If water level reaches 90 inches, the operator will close the MSIVs.

Procedure OT-110 will be revised by 7/20/90 to include specific guidance to the operator on what action to take on high water level if RPV pressure has decreased below condensate pump shutoff pressure.

-Startup During plant startup, one condensate pump is used to pump water from the hotwell back to the hotwell while RPV level is being maintained by matching Control Rod Drive (CRD) flow in with Reactor Water Cleanup (RWCU) flow being rejected (Step 5.16 of procedure GP- '

2, " Normal Plant Startup"). If additional feed is required to maintain level in the proper operating range, then makeup is

-provided by throttling startup recirculation valve MO-2(3)663 as necessary to raise pressure to inject condensate into the RPV. At 150 psi, RPV level control is maintained by valve AO-8(9)091 (Step 6.13 of procedure GP-2). Valve A0-8(9)091 i-s a ten inch bypass around the 'C' Reactor Feedwater Pump (RFP) discharge valve installed specifically for startup level control. It is controlled

4 s,-

Attschment 1

• Page 4 by the normal feedwater level control system and will automatically control-level in the normal band. t

. A failure of the feedwater level control system could <

result ~in-level increase or decrease depending on the type of ,

failure. However, an increase in level would be detected by the ,

operator by either visual indication on the11evel recorder or by a i high level alarm at 29 inches (Main Steam Lines are at 110 inches).

Procedure.OT-110, Reactor High Level, currently is entered only when a-high level alarm is received with the mode switch in "RUN". . In -

addition,-there is no clear direction if the cause of the high water level is from the condensate pumps. Therefore, OT-110 will be revised by 7/20/90 to provide guidance for overfill resulting from the condensate pumps, and to allow the procedure to be used in other modes whenever feedwater level control is in automatic. The procedure will also contain specific. guidance concerning a high ,

level condition during reduced RPV pressure conditions 1when '

condensate pumps-can feed the vessel. This will ensure the operators are'able to mitigate a RPV overfill event from condensate s pumps during startup conditions with a feedwater level control system component failure. ,

L Shutdown Procedure GP-3, " Normal Plant Shutdown", is the procedure used.to shutdown the plant.- The plant Operations group has determined through review that the possibility of an overfilluevent .

L

~ duefto reduced. pressure operation is.very remote. In addition, shutdowns generally use a planned reactor scram at about 30%-power, p at which time the Trip Procedures lare entered. As previously mentioned in the post scram discussion, the planned procedure changes, along with current operator training, will. adequately L address this concern.

To summarize, current operator training adequately '

addresses the issue of RPV overfill due to condensate pumps during l reduced RPV pressure operation. The following revisions to l

procedures will be made to ensure that operators can mitigate reactor vessel overfill events that may occur during reduced '

pressure operation.

ai OT-110 will be revised by 7/20/90 L

b) Trip procedures will be revised by 7/31/90 as part of the Rev. 4 upgrade NRC Recommendation (1)(b)

-It is recommended that plant procedures and Technical -

L Specif; t all BWR plants with main feedwater overfill rote > -

t m: ,

arovisions to verify periodically the operability 4" o v .> > i' p cte :4on and ensure that automatic overfill protection s-52.-'

i -

.c AttachmantII

-Page 5 to' mitigate main feedwater overfeed events is operable during power *

-operation.. The instrumentation should be deconstrated to be

• operable by the performance of t channel check, channel functional 1 testing, and channel calibration, including setpoint verification. '

The Technical Specifications should' include appropriate limiting-conditions'for. operation (LCOs). These Technical Specifications should be' commensurate with the requirements of existing plant *

' Technical Specifications for channels that initiate protective actions. Previously approved Technical Specifications for surveillance intervals and limiting conditions for operation (LCOs) for overfill protection are considered acceptable.

PECo Response As discussed in the attached Design Analysis, the Reactor Feedwater Pump Turbines can be tripped on reactor high level by either 1) activation of Feedwater Level Control relay 6A-K1 OR 2) actuation of HPCI relays 23A-K52 and 23A-K37. Presently, these specific relays and their associated contacts are not adequately addressed by Technical Specifications (TS). Therefore, an appropriate Technical Specification Change Request (TSCR) will be

-submitted to the NRC by January 31, 1991. Limited requirements,for L reactor level indication already exist in TS Tables 3.2.F and 4.2.F.

L

'Any necessary. revisions to these requirements to conform with the  ;

recommendations of the Generic Letter will be included in the TSCR discussed previ'ously. The frequency of the channel check and

' channel calibration will be consistent with other channels that 4 initiate protective actions. However, it is proposed that the- '

, frequency of the channel functional testing:will be once per operating cycle.- This is-less frequent than the current TS >

frequency of Logic System Functional Tests (LSFTs)'for other channels that initiate protective actions. Other LSFTs are on a once per six month or~once per three month frequency. PECo is considering:a TSCR to extend these LSFTs to a once per operating cycle frequency. The justification for extended channel functional i

Ltesting frequency is that testing relays requires temporary-circuit L -alterations (i.e., lifted leads, jumpers and booted relay contacts).

Each circuit alteration has an associated risk and possible error l

.that could result in circuit component damage or ultimately a reactor scram if performed during power operation.- Therefore,.the proposed channel functional testing for vessel overfill' protection provided by HPCI and Feedwater Level Control is consistent with the

' planned frequency of other channels which initiate protective actions. '

l As recommended on page 2 of the Generic Letter, short-term measures will be taken to provide periodic verification and testing L of the overfill protection system. Appropriate plant procedures will be developed or revised to implement the channel check, channel l1 ~functi...a testing and channel calibration as discussed above. This L will - cumpleted by prior to startup from the next refucling outage ,

l of Unit 2 (between cycles 8 and 9). '

l

e NUCLEAR ENGINEERING DESIGN ANALYSIS COVER SHEET DEPARTMENT ,

M-20594 Rw.12/89 DOCTYPE 1s2 uCD, EWR #P-51490 MOD DESCRIPTION: NRC Generic letter 89-19--Recuest for action related to resolution of unresolved safety issue A-47, " Safety Implication of Control Systems in i.WR' Nuclear Power Plants" pursuart to 10CFR50.54(F) ,

STATION: PBAPS UNIT. 2&3 .

PURPOSE OF ANALYSIS:

To review the Feedwater Control System to determine if the ,

NRC concerns raised in Generic letter 89-19 are valid for Peach Bottom APS Units 2 and 3. ,

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RESPUNSIBLE INDEPENDENT DEPTH OF REVIEW , .

CATE REVIEWER DATE EXHIBIT SEE ORGANIZATION ENGINEER 3441 DOCUMENTATION oc$hc%A b \ b 1.~_ *hk hW kl- l '

NIERMCING BRANCH N/A ,

INTERMCING BRANCH N/A , . , :. ,

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[ ATE: 4 BRANCH HEAD - ORIGtNATING ORGAN!ZATION REV.NO'g

'COFY1D: DAC (61B-5)

-T. J. Cabrey (63A-5)

M. C. Kray PAGE 1 OF 10 EXHIBIT 3.4-IV -

- 1/90

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Generic Letter. 89-19 Pac. of 10 w sypog

. .eric Letter 89-19'(reference 1) requests action related to

theiresolution of unresolved safety issue A-47, " Safety  ;

Implication 1of Control Systems.in LWR Nuclear Power Plan &s". In- i this-Generic Letter the NRC concluded that protection'should be

,provided for certain control system failures. In particular,  ;

protection should be provided for reactor vessel overfill- . The ,

purpose:of this' design analysis is to determine'if the feedwater control system provides adequate protection to. prevent an overfill event from.the feedwater control system. Specific features'to be analyzed are: power supply interdependence, u sharing of sensors between control and trip logic, separation of-

• ' control and-trip circuits, and designs for indication and alarm.  ;

b PURPOSE- _

.The purpose-of this dec1gn analysis is the review of the_ -

_l feedwater control system to determine ~if-the NRC concerns raised in Generic Letter.89-19 are valid for Peach Bottom Atomic Power Station;(PBAPS)' Units 2 and 3.

EEFERENCES

1. Generic Letter 89-19 Request For Action Related To -f Resolution Of Unresolved Safety Issue A-47 " Safety Implication Of Control Systems _In LWR Nuclear Power Plants"' Pursuant To 10CFR50.5S(f).

2. NUREG-1217 Evaluation Of Safety Implications of Control Systems-In LWR Nuclear-Power Plants - Technical F114 dings Related To USI A-47.

3. 6280-E-129 Rev 25 Electrical Schematic Diagram -

Reactor Feed Pump Control Scheme.

4. 6280-M1-S-25 Rev 46 Elementary Diagram - Feedwater Control System.

5. 6280-M1-S-36 Rev 62 Elementary Diagram - High Pressure Coolant Injection System, p 6. 6280-M1-S-65 Rev 86 Elementary Diagram - Residual Heat

! Removal System.

7. 6280-M-352 Rev 32 P&ID - Nuclear Boiler Vessel Instrumentation.

8 ', 6280-E-28 Rev 43 Single Line Diagram -

Instrumentation and Uninterruptible AC Systems Unit 2 and Common.

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P00ch~ Bottom APS EWR #P-51490 I# Generic Letter 89-19 Page 3 of'10

9. 6280-E-29 Rev 33 Single Line Diagram -

Instrumentation and Uninterruptible AC Systems-Unit 3.

10. 6280-E-3031 Rev 2 Electrical Schematic Diagram -

Panels 20C818 and 20C819 Power j Dintribution Y'

D$ SIGN _ REVIEW.

Refer to Figure #1 for a composite schematic of the' Reactor g  ;Feedwater Pump Turbines (RFPTs) trip logic.

TRIP LOGIC The reactor high-level trip of the RFPTs comes from the 6A-K1, 23A-K52, or 23A-K37 relays. These relays are arranged such that '

the 6A-K1 relay by itself.or the 23A-K52 AND 23A-K37 relays can trip the RFPTs on-high reactor level. All of these relays use

. NORMALLY OPEN, contacts that close to cause a trip signal. Upon  ;

closing to' complete the circuit the main trip solenoid valve (SV-12) for each RFPT energizes resulting in the RFPTs tripping.

The 6A-K1 relay is controlled from alarm units 6-121B'and 6-121C. '

- The' alarm units use NORMALLY CLOSED contacts which close on reactor high evel- .These alarm units are arranged such that both alarm units _must trip (on high level) in order to energize

- the 6A-K1 relay which-in turn will energize-the SV-12s trip solenoids. These alarm units receive their signal from the

~

reactorylevel. pressure compensation proportional amplifiers (LPAM-6-72s). ;These proportional amplifiers take a reactor level

- delta-pressure signal (LT-6-52) and adjusts this delta-pressure aignal with a. reactor pressure signal (PT-6-53) to account for-changes 11n water-density. The output of the proportional amplifiers is pressure compensated reactor level. This pressure compensated' level signal is then used as the controlled variable- ,

4 for the feedwater control system.

The 23A-K37 and 23A-K52 relays are controlled from the Foxboro reactor water. level compensation panels (C818 and C819). These

- panels consist'of Input / Output cards and a micro-processor (XAM-2-3-117s). The micro-processor receives inputs from wide range-(LT-2-3-72s) and fuel zone (LT-2-3-73s) reactor level

^ delta-pressure transmitters and a reactor pressure (PT-2-3-404s) transmitter. The micro-processor then performs pressure compensation on the two reactor level delta-pressure signals.

L The micro-processors also performs alarming and "tip functions on

' these pressure ccmpensated level signals. One of these trip

- functions is the t eactor high level trip signal (XS-2-3-2'.5s) to  !

l the 23A-K37-relays (similar for 23A-K52). The trip contact uses a NORMALLY OPEN contact which closes on a high reactor level.

Other outputs of the mirco-processor are used for indication

-(control room indicators and recorders) and Emergency Core

- Cooling System (ECCS) reactor level and pressure trip signals.

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? Page.4 of 10 POWER SOURCES Table #1 summarizes all the different systems and sub-syntems >

power. requirements. i t

The RFPT trip logic is powered from 125 VDC and requires power to be present in order to remotely trip the RFPTs. (Manual trip can always be performed locally at the RFPT front standard.) The trip solenoid (SV-12) energizec to trip.

.The feedwater control system is powered from Uninterruptable Power Supply (UPS). This UPS system receives power from either 480 VAC-(E12(3)4-R-C (2(3)OB36]) or from 125 VDC (Bus 'D').

Either of these power sources being available will make ,

continuous power available to the UPS and hence the feedwater control system. .A partial'(individual fuse or circuit breaker) or complete loss of UPS power to the feedwater control system  ;

could result in any of'the following: +

- 1. No impact to control system.

2. Partial loss of feedwater control system functions. This could' result in the lockup (loss of control. signal) of one or two RFPTs with one remaining available for control of '

reactor. level. One RFPT.for control is sufficient to control reactor level within tha range of the high and low l level' alarms.

3. A complete loss of all three (3) RFPTs for control (all '

RFPTs lockup). With all three RFPTs locked-up the fixed rate of feedwater make-up to the reactor could be either less than or greater than the. rate which is required to maintain reactor' level steady-state. In this case level would'either decrease or increase depending on when the lockup occurred. .

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4. If the power source to the GA ; ay coil was lost the

~feedwater control system could uc . vide any reactor high water level trip signals to the RFFJ crip l'ogic ,

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The High Pressure Coolant Injection-(HPCI) logic is powered from 125 VDC ('A' and 'B' Buses). A loss of either of these power sources will result in the inability of the HPCI system to E provide a complete (2-out-of-2) high level trip signal to the

.RFPT. trip logic.

The reactor water level compensation system (C818 and C819 L

panels) are powered from a dual power supply.(E/S-2-3-122s).

These-dual power supplies are powered from two independent sources. The C818 panel is powered from 120 VAC (2[3]OY35) and L 125 VDC (Bus 'C'), whereas the C819 panel is powered from 120 VAC (00YO3) and'125-VDC (Bus "D'). Either power source being K availabic to'one of these dual power supplies is sufficient to i - g

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POcch CottomtAPS EWR #P-51490

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Generic Letter 89-19

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. power its associated panel (C818 or C819). Loss of both power

-sources to one'of these danl power supplies will result in the

-associated panels from being unable to provide a high reactor

. level trip signal to the HPCI system and hence the RFPT trip logic.

TABLE 01 POWER SOURCES ASSOCIATED WITH THE RFPT C0fffROL AE HIGH LEVEL TRIP LOGIC ,

System / Subsystem Power source Cwments RFFT Tr1P Logic 125 VDC- 1. Requires pown- to remotely trip RFMs ISV-12A,B,C) Bus A,B.C 1

Feedwater Controls 120 VAC UPS 1. Requires power to control reactor level.

21310Y50 2. Loss of power to this portion results in High level trip signal f rom the fendwater control system and lockup of one or more of the RFFTs reedwater Logic 120 VAC UPS 1. Loss of this portion of UPS power results in the (6A-K1) El310Y50 inability to provide RFPTs high levd trip signal from the-6A-K1 relay.

HPCI Logic 125 VDC 1. Loss of power results in the inability to provide RFPTs '

(25A-K371 Bus 'A' high '.evel trip signal from the 23A-K37 relay.

HPCI Logic 125 VDC 1. Loss of power results in the inability to provide RFFTs .

(23A-K52)- Bus 'B' high level trip signal from the 23A-K52 relny.

Reactor Level 120 VAC 1. Power is provided via a dual power supply where either (XS-2-3-115C) 2(310Y35 power input available will maintain power to the components.

OR 2. Loss of both power sources will result in the inability _

125 VDC to provide RFPTs high level trip signal from XS-2-3-115C. - :

Bus 'C' i-I' Reactor Level 120 VAC 1. Power is provided win a duni power supply where either (KS-2-3-115DI 00YO3 power input available will maintain power to the components.

GR 2. Loss of both power sources will result in the-inability .

125 VDC to provide RFPTs high level trip signt ' from XS-2-3-115D.

Bus 'D' ch

!c P_HYSICAL SEPARATION The RFPT trip solenoid valve (SV-12) is located near the Reactor Feedwater Pump (RFP) bays on the turbine deck, 135' elevation.

The high. level trip logic circuit is power from the C06C panel in NA . . . _ _ - . _ - - _ . . _ . _ .

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• POcch Bottom APS EWR'#P-51490 Generic Letter =89-19 JPage 6=of 10  ;

panels'(C18 .C32, and C39) located in the cable spreading room.

From the cable' spreading room.the circuit routas out to the RFP baye (A,B,CC124).

I The majority of the feedwater control circuitry resides in the cable spreading room in panel C18. The balance of the.feedwater control circuitry-resides in the main control room panel (C06C).

The feedwater control system receives its reactor.and level  ;

signals from local field transmitters (LT-6-52s and PT-6-53s)  ;

which are located in the reactor building on instrument racks '

' C65A and C658, The feedwater control logic contact output is

-used in the RFFT. trip logic.

The HPCI logic is located in the cable spreading room in panels C32 and C39. The HPCI logic receives its input from two other panels also located in the cable spreading room, C818 and C819. -i The HPCI logic output contacts are used in the RFPT trip logic.

The reactor water level panels-(C818 and C819) reside in the cable spreading room. The reactor water level instrumentation receives its reactor level and prest,ure signals from local field transmitters'(LT-2-3-72s, LT-2-3-73s, and PT-2-3-404s) which are located in the reactor building on instrument racks C65A, C65B, C91A, and C91B. ' Power for the reactor water-level instrumentation comes from two other cable spreading room panels (C722A and C722B). Contact outputs from the reactor water level instrumentation are used in the HPCI logic in panels C32 and C39.

i ARRANGEMENT OF SENSORS ,

Figure #3 shows the arrangement of the level transmitters which

'are associated with the reactor high level trip logic. One common 11nstrument penetration (N12A) supplies steam to two (2) condensing chambers (2A and 3A). These. condensing chambers ,

condense steam into condensate to maintain a constant head of .!

water to the level. transmitters (LT-2-3-72C and LT-6-52C) l

. reference ports. Another instrument penetration (N11A) provides tivariable pressure signal to LT-6-52C. A separate instrument  ;

penetration (N16A) provides a different variable pressure signal to1LT-2-3-72C. These variable pressure signals consist of reactor steam dome pressure and the head of water above the <

instrument penetration (variable dependant upon reactor level).

This arrangement is then duplicated for the LT-2-3-72D and

. LT- 6-528 . transmi tte r s . Reactor level is inversely proportional to the differential pressure across the transmitter. That is, for a minimum differential pressure (P(ref) = P(var)) level is at its maximum value and for a maximum differential pressure (t,ref)

>>> P(var)) level is at its minimum value.

~

V

.De;ign Analyaic R;v. 2

~POcch Bottom APS EWR #P-51490 Generic Letter 89-19 Page 7 of 10

_ ANALYSIS E

POWER SUPPLY INTERDEPEND"i'C_E A single power failure that could result in the feedwater control system causing reactor level to rise would be a complete failure of the 120 VAC UPS power supply (assuming the RFPTs all lockup in such a way to result in a continuous net vessel level increase).

-The loss of 120 VAC UPS also results in the inability of the feedwater control system to provide a high reactor level trip signal to the RFPTs trip logic via the 6A-K1 relay. However, the reactor water level system and HPCI logic would remain available to provide the reactor high level trip signal to the RFPTs to stop the vessel level rise. Note that this transient would be a relatively slow transieht compared to others discussed later on

'in this analysis.

A loss of the 125 VDC Bus 'A' power source would result in the

' inability-to remotely trip the 'A' REPT ('A' SV-12 powered from 125 VDC Bus 'A') and the inability of the HPCI logic to provide a high level trip (23A-K37 logic powered from 125 VDC Bus 'A').

However, loss of this bus will have no effect on the feedwater  !

control systems ability to control reactor level. In addition, the feedwater control system would still be able to provide a high level trip signal to trip the 'B' and 'C' RFPTs.

A loss of the 125 VDC Bus 'B' power source would result in the inability to remotely trip the 'B' RFPT ('B' SV-12 powered from 125 VDC Bus 'B') and the inability of the HPCI logic ~to provide a-high level trip (23A-K52 logic powered from 125 VDC Bus 'B').

However, loss of this bus will have no effect on the feedwater control systems ability to control reactor level. In addition,  ;

the feedwater control system would still be able to provide a j high level trip signal to trip the 'A' and 'C' RFPTs. a A loss of the 125 VDC Bus 'C' power source would result in the inability to remotely trip the 'C' RFPT ('C' SV-12 powered from 1 125 VDC Bus 'C'). There are no other effects on either the control circuits or trip logics as a result of this loss of

. power. The only other system that receives power from this bus is the 'C' reactor water level instrumentation. However, since }

the reactor water level instrumentation is powered by a dual power supply. Loss of this power feed doesn't have any effect on  ;

the operation of this instrument.  !

A loss of any one of the 120 VAC 00YO3 or 2(3)0Y35 or the 125 VDC l Bus 'D' power sources does not have any effect on either the control circuits or trip. logic. The reactor water level instrumentation is powered by dual power supplies. Loss of any 1 single power feed to these power supplies doesn't have any effect on the operation of the instrumentation.

a

)

hD;;ign'Analy2iG-Ocv. 2'

POcch Bottom:AP3 UfR #P-51490  ;

e Generic Letter 89-19 Page 8:of-10 A: loss of both feeds (i.e. as a result of a fire in the C722A ,

panel) to either one of these_ dual power supplies will result in ,

that channel of the reactor water level instrumentation-inability- '

to provide a high. level trip signal to the HPCI logic. The .

L feedwater control system will not be affected by this power loss. .

Therefore, the feedwater control system will continue to maintain ,

level control and have the ability to provide a high level trip' .

' signal if required.

There are no other single power failures that by themselves would result in the feedwater control system causing reactor vessel-level to. rise.

SENSING LINE INTERDEPENDENCE A failure of a sensing line to any level transmitter could result in a false level indication. This section looks at single failures of sensing lines associated with the reactor high level trip logic. This section only looks at single failures for half of the sensing lines since the other half would have similiar results (refer to Figure #3). Failures can range from small1 leaks to complete breaks of the sensing.line.

1. Vessel Penetration N12A to Condensing ~ Chambers 2A.and 3A:

A failure of this line will result in.P(ref) decreasing which results in a false high level indication from LT-2-3-72C.and LT-6-520. If the feedwater control system-is controlling on the LT-6-52C the feedwater control-system would reduce feedwater flow'in response-

• to this false.high level indication. If the failure was large enough the false high level indication could cause a false high level channel trip signal from

, LT-2-3-72C and LT-6-520. -(Note this does'not result in '

L a complete trip. signal).

2. Condensing Chamber 2A to Level Transmitter LT-2-3-72C:

A failure of this line will result in a false high level-indication from LT-2-3-72C only. This-failure will have no effect on the feedwater control system.

If the failure was large enough the false indication could cause a falso high level channel trip signal from LT-2-3-72C. (Note this does not result in a complete '

trip signal). If the failure was large enough so that the leak becomes uncovered'the sensing line from penetration N12A to the condensing chambers 2A and 3A would also depressurize. This failure would be covered by (1) above.

3. Condensing Chamber 3A to Level Transmitter LT-6-52C:

Thiszfailure is similiar to the failure of the condensing chamber 2A to level trantn.tter LT-2-3-72C

7-

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$De31gn-AndlyaioR;v.2

% POcch Bottom APS EWR #P-514901

, 3 _ ,

i Generic Letter 89-19 Page-9 of 10-in (2) above. However, for.this failure if the feedwater control system is' controlling on LT-6-52C the-feedwater control system would reduce feedwater flow in response to a false high level-single.

4._ Vessel _ Penetration N11A to Level Transmitter LT-6-52C:

1

-A failure of this line will result in P(var) decreasing '

which results in a false low level indication from LT-6-52C. If the feedwater control system is controlling on LT-6-52C the feedwater control system would increase feedwater flow and actual level would rise. If actual reactor level reaches the high level trip setpoint the feedwater control s'/ stem will not be ,

able to initiate a high level trip siqual. However, >

the HPCI system will initiate a high level trip signal

[g via the LT-2-3-72C and LT-2-3-72D transmitters.

5. Vessel Penetration N16A to Level Transmitter LT-2-3-72C:

This failure is similiar to the failure for vessel-penetration N11A to level transmitter LT-6-52C in (4) above except the feedwater control system will not be

~

affected. ,

.Therefore, there is no potential for reactor overfill from any single failure of the sensing lines associated with the reactor ,

high level trip logic,

{ A SHARING OF SENSORS l

The feedwater control system does share sensors and compensation  ;

instrumentation between control and reactor high level trip to

~

the RFPTs functions. However, there is a redundant reactor high level trip signal _available from the HPCI logic. The redundant trip signnis come from diverse and independent sensors and pressure compensation instrumentation.

V ,

E SEPARATION OF FUNCTIONS The feedwater control system does share compensation  ;

instrumentation between' control and high level trip functions.

Therefore, there is-no separation within the feedwater control system between the control and trip functions. The panels that contain the instrumentation that makes up the redundant trip

<' signals reside in Panels C32, C39, C818, and C819. The C18 panel contains the'feedwater control system. Therefore, a fire in any one of the above listed panels will only affect one portion of the high level trip logic (feedwater or redundant trip signal). [bs The remaining trip channel will be available to provide a reactor high-level trip. signal if required. This arrangement provides sufficient separation between the redundant trip signals and the reactor' level control functions.

m i

FDacign Analycia Rev. 2 Pocch Bottom APS EWR OP-51490 Generic Letter 89-19 Page 10 of 10' RSS OF_VEN_TILAT.IOJJ The panels that house the majority of the reactor high level trip

. logic'and the feedwater control system reside in the plant cable spreading room. Normally the ventilatiog system maintains the cable spreading room at approximately 85 F. The equipment contaiged inLoss thicof room is designed to operate at temperatures up ventilation to the cable spreading room will to 120 F.

result in a control room trouble alarm and a slow temperature rise. Operator actign would then be taken to restore normal ventilation. At 110 F any one of 5 EHC cabinets temperature switch would alarm (the EHC cabinets are directly across an aisle ;2 from the feedwater control panel). At this point additional operator action would be taken. Based on the above it is expected that appropriate operator action cag be taken before the cable spreading room temperature exceeds 120 F.

INDICATION AND ALARMS Both the feedwater control system and the reactor water level instrumentation system provide control room indication of reactor level. Both systems provide indication on indicators and trend recorders. The feedwater control system reactor cate" level indication ranges from 0" to 460". The reactor water tevel instrumentation system indication ranges from -31.5" to 'AO". The RFPT high reactor level trip setpoint is at +45".

Both the feedwater control system and the HPCI logic provide high reactor level trip alarms in the control room. In addition, the feedwater control system provides a reactor high level alarm at 29".

MOD 1843 IMPACT Modification 1843 replaces the foodwater control system with a smart,-fault tolerant, Digital Feedwater Control System (DFCS).

The DFCS will replace the high reactor level trip logic to the 6A-K1 relay. The new logic will require both DFCS computers to call for a trip or be off-line taken twice (see figure 2). The power to the 6A-K1 relay is not effected by this modification.

This modification has no effect on the redundant' reactor high

' level trip signal provided by the HPCI and reactor water level instrumentation systems.

CONCLUSION Based on the above analysis the Peach Bottom APS Units 2 and 3

' meet the requirements of NRC Generic Letter 89-19. Specifically, there is adequate power supply independence, sensing line independence, separation of sensors, physical separation, indication, and alarming between the feedwater reactor level control system and the RFPTs high reactor level trip logics.

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