ML20199J663

Forwards Info Re Plant Responses to Staff Requests for Info on Section 3.3 of Plant Improved TS Submittal.Part of GE 24A5369,Rev 0,Class I, Supplemental Reload Licensing Rept for Plant Reload 14 Cycle 15, for Sept 1996 Encl

ML20199J663
Person / Time
Site:	Duane Arnold
Issue date:	01/29/1998
From:	Kelly G NRC (Affiliation Not Assigned)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20070L368	List: ML20199J663
References
NUDOCS 9802060006
Download: ML20199J663 (17)

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.p . 4 :__ UNITED STATES g [ NUCLEAR REGULATORY COMMISSION o, wAsHtNGToN D.C. 2026-4001

• • • • • January 29, 1998 MEMORANDUM TO: Information and Riscords Management Branch Office of Information Rescurces Management FROM: Glenn B. Kelly, Senior Project Manager , /

Project Directorate 3-3, DRPW, NRR u/

SUBJECT:

INFORMATION TO BE PLACED ON THE DUANE ARNOLD ENERGY CENTER (DAEC) DOCKET (50-331)

The enclosed information documents several faxes between the NRC staff and IES Utilities Inc.

regarding DAEC responses to staff requests for informatien on Section 3.3 of the DAEC imprwed technical specification submittal. Please place the enclosed information on the DAEC docket.

Enclosures:

as stated Docket Nos. 50-331 cc: Sec next page NRC MI@EEEE CDPV

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DF0 i 9802060006 980129 PDR ADOCK 05000331 P PDR '

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January 29, 1498 MEMORANDUM TO: Information and Records Management Branch Office of Information Resources Management FROM: Glenn B. Kelly, Senior Project Manager Project Directorate 3-3, DRPW, NRR .

SUBJECT:

INFORMATION TO BE PLACED ON THE DUANE ARNOLD ENERGY CENTER (DAEC) DOCKET (50-331)

The enclosed information documents several faxes between the NRC staff and IES Utilities Inc.

regarding DAEC responses to staff requests for information on Section 3.3 of the DAEC improved technical specification submittal. Please place the enclosed information on the DAEC docket.

Enclosures:

as stated Docket Nos. 50-331 cc: See next page DISTRIBUTION:

Docket File PUBLIC G. Grant, DRP, Rlli EAdensam (EGA1)

PD3-3 RF R. Savio G. itelly ACRS OGC P. Ray R. Laufer DOCUMENT NAME: G:\DUANEARN\97197COS.WPD To receive a copy of this document andcate in the box C= Copy w/o attachtnent/ enclosure E= Copy with attachment / enclosure N = No copy OFFICE PM:PD33, E NAME GKelly\ '

DATE 01/29 /d8 OFFICIAL RECORD COPY w _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _.-._.__.___-__._-m_ . . _ _ _ _ _ . _ . _ _ _ _

Dune AmeM Enerpy Comer 3277 DAECM.t Pek. M S3324

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UTILmES Fax Cover Sheet To: Rich Laufer t Company: US NRC Far Number: (301) 415-3061 From: Tony Browning Phone Number: (319) 851-7750

(

IES Far Number: (319) 851-7364 Number ofpages following this page: 4 DefaWme: January 14,1998 7:30 AM __

Rich, Attached are the UFSAR table on Diesel Generator loading sequence and simplified logic diagrams for the Emergency Bus Lo Ang Relays. His is a request from Carl Schulten and Ed Tomlinson for question 3.3.5.1 - 15. I don't know if we'll get to discuss this in this afternoon's call or not.

Any Questions, call me.Thanks.

Tony

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AnifskwusHs compacy l

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v/SAR/DAEC-1 I~E p Table S3II-O SINGLE DMMTDR LDADING MEWE AND M-LOSS 4F-COOLANT ACCIDENT PLUS IDSS OF OFFSITE POWER EinpsedTime Lead Total Valesse Recovery , Frequency RecoveryTaas C hemlaillmeios of lacrossent Imed (%of Tune to 90% (% ofrated) so 95% Rated Desuiption Evana (kW) (kW) red) Rased Psquency(sec)

I Voltags (sec)

OSEC LJCA ACCIDENT 3 SEC i LOW REACTOR WATER V LEVELOR H10H DRYWELL PAESSURE SIGNAL, EDG STARTSIGNAL 13SEC 100 100 EDG CLOSES 1D BUS, g SiONAL REAC1DR RECIRC AND CONTAINMENT OM ISOLATION, LPCI AND CORE SPRAYINAiCT g VI.LVES OPEN g 13 MCC1998 LOADS

$ 135 MCC IB34A,lB44A LOADS 345.1 MCC 1832 LOADS 114 MCC 1934 LOADS BH 711.I 73 1.2 96.7 .d.87 RIVER WATER SUPPLY PUMPSTAR1" 18 SEC 612 , 1323.I 80 1.3 97.2 3.98 COltESPRAY PUMPSTART ,'

28SEC CORE SPRAYINECT  !

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VALVE OPEN i _

T83-1 Revisis -1065

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Table 83-1

[- SINGLE DIESEL-GENERATOR LOADING SEQUENCE AND RESPONSE -

LOSS-OF-COOLANT ACCIDENT PLUS LOSS OF OPPSITE POWER 23SEC i 496 5819.1 to 1.3 f 97.2 3.91 PjRSTRHR(LPCI) PUMP r START 28SBC 496 2315.1 83 1.1 97.1 2.45 SECOND RHR (LPCI) PUMP -

START '

r 31SEC LPCI VALVESOPEN 43 SEC REACTOR RECIRC VALVES

. i CLOSE, .

COMPLETION OF LPCI sequENcs '

73 SEC 66 2381.1 - - - -

CONTROL BUILDING  !

h .

CHILLFR START 3

AAer 10 min' -162 2219.1 - - - -

, MCC LOAD REDUCTION

-496 1723.1 - - - -

- 3 TPJP RHk PUMP  !

470 2193.1 - - - --  !

PJtR SERVICE WA'I1!R PUMP t START 478

266~e.1 - - - -

! RRR SERVICE WATER PUMP l SJART i Q NOTE:

• Additionelloads may be applied at operator's discretion.

' If ths RWS pesop shd for auto start was running prictto the LOOP, the selected RWS penp will trip and will sessert 120 =a==d= l aber the diesel generator optput breaker closes.

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1 T8.3-2 Rep 12 - 10/95

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l LOW PRESSllRE COOLANT INJECTION (LPCI) SYSTEM REACTOR VESSEL

{ . WATER LEVEL o LOW-LOW-LOW (2.a)

LPCIINITIATION LOGIC "A" I

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.______.___.J______._____i Minirnum Channel Requireinents for LPCI Mode of RHR Initiation Capability in order to maintain the capability to initiate the LPCI Mode of RHR Loop 'A' c 1 the Reactor Water Level . Low. Low Low Trip Function, the following combination of channels is required to be Operable.

A channel in the tripped condition is comidered to be Operable.

LIS.4531 OR LIS-4532 AND LIS 4533 OR LIS.4534 TABLE 3.3.5.1 1 EMERGENCY CORE COOLING LFD ECCS-05

{ ELEMENTARY REFFRENCES SYSTEM INSTRUMENTATION REV.1, olvi7/96 l APED E21006 SHEETS 1,2 APED E11007 SHEETS 3.4,5,7,3,13,15,17 LOW PRESSURE COOLANTINJECTION (LPC11 SYSTEM SHEET 2 OF 3 REACTOR VESSEL WATER LEVEL. LOW LOW-LOW si r- _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _

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I CORE SPRAY SYSTEM DRYWELL PRESSURE - HIGH (1.b)

CORE SPRAY "A" ACTUATION LOGIC CORE SPRAY "B" A r - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - -'- - - - -- - - CTUATIO N LOG IC i

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r c_-----_...----------------------------------------------- >l Minimum Channst Requirements for Core Spray Initiation Capability in order to maintain the capability to initiate Core Spray System 'A' or Core Spray Cystem "B" on the Drywell Pressure - High Trip FunctJon. the following combination of channels is required to be Operable.

A channelin ik.e tr'pped condition is considered to t3 Operable.

PS 43108 OR PS-4311B AND PS 43128 OR PS 43133

.k TABLE 3.3.5.1-1 EMERGENCY CORE COOLING LED ECCS 02 ELEMENTARY REFERENCES SYSTEM INSTRUMENTATION REV.1. 08/17/96 APED-E21006 3HEETS 1. 2. 4 CORE SPRAY SYSTEM SHEET 2 OF2 DRYWELL PRESSURE. HIGH

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UTILITIES Fax Cover Sheet To: Rkh Laufer Company: US N'C FaxNumbert (301)415-3061 Froers: Tony Browning Phone Numbert (319) 851-7750 s s

/ES Far Number: (319) 851-7364 Number ofpages following this page: 11 Date/fime: January 14,1998 7:30 AM Rich, f

Attached is a package ofinformation regarthng the Feedwatsr/rurbine Trip instrumt.nts. The first part is a short write-up that I did to try and explain our position and to clear up what appurs to be a misunderstanding by the Staff m what our UFSAR says about this trip. In addition to UFSAR pages, l included a write-up faen GE's fuel tepical report (which is part of our licensing basis by reference la our UFSAR, TS and Core Operating Lhniu Report (COLR)) and the resultt of the Feedwater Controller Failure translen, out of our reload analysis for thh cycle's COLR.

As we discussed yesterday, we will try one more time to ruolvs this at our level, using this additionsi information for the Staff.

OtheMte, we'll need to appeal this assue to a higher nasnagemot level, as we will have great disculty incorporating this LCO i

into our ITS on out timetable. First, Engineering tells us, if they drop everything else thr,t they are doing, it will take them 4 manweeks to generata and verifv the setpoint calculetion for the Allowable Vahr for the calibration SR, as one does not currently d exist. Second, what safety basis do I use to write my BASES from, u our analysis shows it doesn't meet criteria 3 at the DAEC.

• Because the NRC said so" doesn't work either. Lastly, as tiils is a me;o restrictive change than the CTS, I think that we'd have

. trouble writing a convinchig M DOC that says that plant operation wouldn't be adversely affected, as we'd be adding LCO requiremen3 and surveillances on instruments ist we have not demonstrated improve p! sat safety or lower plant risk.

I{

Any Questions, call me. Thanks.

t' Tony AnIES kWustries Compeny

. 01/14/98 08:37 Osts all 7153 IES INDU$TRIES E662/b11 a

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- UFSAR/DAEC-1 7.7 CONTROL SYSTEMS NOT REQtBED FOR SAFETY ,

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! This section discusses control systems wh6se functions are not' easestial for the safety of the plant. These systems are the feedwater control system, the turbine-generator controls, the i

reactor manual control system, and the process computer system.

? 7.1 FEEDWATER SYSTEM CONTROL AND INSTRUMENT,ATION l

• \

c 7.7.1.1 Power Generation Objective i,

The power generation objective of the feedwater control system is to maintain a l preestablished water level in the reactor vesse! during normal plant operation.

7.7.1.2 Pcwr Generation Da=len Basin L

The feedwater control system regulates the feedwater oow (1) to maintain adequate water level in the reactor vessel according to the requirements of the system operators and (2) to l

prevent the exposure of the reactor core over the power range orthe reactor l

l 7.7.1.3 System Dascription

a

! During normal plant operation, the feedwater control system automatically regulates feedwater flow into the reactor vessel. The system can be manually operated. The feedwater control system includes the two main feedwater control valves and one feedwater startup control valve.

The feedwater flow control instrumentation measures the water level in the reactor vessel, h

' the feedwater flow rate into the reactor vesal, and the steam flow rate from the reactor vessel.

During normal operation, these three measurements are used in controlling feedwater flow.

The optimum reactor vessel water level is determined by the requirements of the steam -

separators. The separators limit water carry.over in the steam going to the turb*mes and limit steam crrry-under in water retuming to the core. The water level in the reactor vessel is normally maintained within 12 in. of the optimum level during normal operation. This control capability is achieved by comparing feedwater flow to the reactor vessel with the steam flow from the reactor vessel to provide an anticipatory level error signal. The feedwater flow is regulated by adjusting the feedwater contrcl v;.lves to deliver the required flow to the reactor vessel.

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7.7-1 Revision D - 4/97 l

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01/14/98 08:38 ft319 458 7193 -!ES-INDi$TRIES l'j0F97018 UFSAR/DAEC-1 ,' "

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7.7.1.3.1 Reactor Vessel Water Level Meamarement e... .. .. . .

Reacsor vessel water level is measured by three identical, independent sensing instmment loops (Figure 7.1-1). A differential-pressure transmitter senses the difference between the pressure causui by a constant reference column of water and the pressure caused by the variable height ofwater in the reactor vessel ne differendal pressure transraitter is instaded on lines that serve other systems (see Section 7.6.4). A total of three level differential pressure transmitters to each transmits a ievi signal to a level indicator and a level switch. Two of the level signals are selectable and the selected level signal is used to provide the level conn el function to the level controller. The selected signal also feeds a computer point, a level switch and a wder. The signal of the non selectable third level differential pressure transmitter only feeds an iridicator and a level switch. three pressure transmitters feed three reactor vessel pressure indicators, respecidj, in the control room. Signals from twe of the three pressure transmitters are selectable and the selected signal is fed to a recorder and a computer point in the control room.

The level signal from two of the three sensing systems can be selected by the operator as the signal to be used for feedwater flow control. The selected water level and the reactor vessel pressure signals are con'inually recorded in the control room.

7.7.1.3.2 Steam Flow Measurement

(

Steam flow is sensed at each main steam line flow restrictor by a differential-pressure k....

transmitter equipped with square root functions. Signals from these differential-pressure transmitters are added to provide a linear signal proportional to the total steam flow rate.

Individual steam line Dow signals are indicated in the control room. The total steam flow signal is used for feenwater flow control, and is also recorded in the control room.

7.7.1.3.3 Feedwater Flow Measurement Feedwater flow is sensed at a flow element in each feedwater line by differential-pressure transmitters. Each feedwater signal is linearized by square root converters. Then the individual mass flow signals are summed to provide a total m ass flow signal for the feedwater flow control l system. The total feedwate mass flow signalis also recorded in the control room.

In order to increase the reliability of feedwater flow indication, redundant flow measuring devices are installed on a local instrument rack in the tuttine building.

The feedwater Bow control system is a three-element control system. The three inputs are vessel level, feedwater flow and steam flow. The latter two constitute a flow mismatch that provides level e: Tor anticipation.

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l 7.7-2 Rsvisan13 N97 c

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ouives osias G als sol 71ss IES INDt!$ RIES %006/o12 UFSAP/DAEC-1 The setpoint of the master controller can be changed to a predetermined level by a manual push button at foe operators discretion, and the controller will control the .vactor level at that predetermined level automatically if the controller is in the Automatic Mode. 'Ihe setpoint transfer function will not be in effect when the controller is in Manual Mode. This function is primarily designed to provide a convenient setpoint change for the operator during the execution of IPOI S, immediate acticas in response to a SCRAh?.

ontinnal Autnmatie oneratinn A single-element concol signal (resetor vessel water level) can be used to replace the above three-elemem signal. In such cases, the operator switches the controller input to the "1 element" signal. The reactor level signal is fed to the level controller through a dynamic compm md a convetter and signal isolator. Reactor water level is then controlled by the reactor level signal in accorince with the controller setpoint.

,.,$[,. f Aurillarv Functinne Alarms are provided for high and low reactor water level and for high prt.ssure. A loss of power signal to the feedwater control valve, Manual / Auto (M/A) stations, and the feedwater startup control valve, or a loss of service air supply to the feedwater startup control valve, or a

( l loss of service air supply to the valves will cause the valves to lock up as is. Both pov.tr failure and low air pressure are annunciated. The feedwater str ap catrol valve has annunciation in the control room for 90% er greater open. This annunciation indicates that the startup valve is approaching maximum flow and that action should be taken to transfer to one of the feedwater control valves. The level control system provides interlocks and control functions to other systems. When one out of two reactor feed pumps is lost and coincident or subsequent low water level exists, recirculation flow is limited to within the power capabilities of one reactor feed pump (see Section 7.7.5). This action aids in avoiding a low. level scram by limiting the steaming rate. Reactor recirculation flow is elso limited on sustained low feedw.ater flow to ensure that adequate net positive suction head will be provided for the recirculation system.

l Two-out-of-three narrow range vessel water level signals at the Hi trip setpoint will cause the feed pumps to trip. Ccatrols to reset the trip are located on panel 1C05. The Reactor Feed Pumps (RFPs) High RPV Level Trip Defeat override may be used in support of the Emergency Operating Procedures (EOPs) in lieu ofjumpers and lifted leads. This defeat allows restoration of the feed pumps for flooding above the normal level either in support of RPV Flooding Contingency or the Primary Containment Flooding Contingency. He single key-lock switch has an amber light and individually annunciates on front panel 1C-14 when taken to override.

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l_ 7.7-4 Revision 13 -4/97 l _ _ _ _ _ _ _ ___ _ _ -_ __.

l ol/t ues_ ossas Vals est 71ss'--- M F WMJb I UFSAR/DAEC - 1

. 15.1 EVENTS RESULTING IN A REACTOR VESSEL WATIRTEMPERATilRE DECREASE Events that result directly in a teactor veuel water temperature decrease are those that either increase the flow of cold water to the vessel or reduce the temperature of water being delivered to the vessel. The three events that result in the most seve-e transients in this category are the following:

1. Feedwater controller failure - maximum demand ,,

2. Lou of a feedwater heare, r. ,

3. Inadvertent HPCI pump start.

15.1.1 FEEDWATER CONTP.0LLER FAILURE - MAXIMUM DEMAND ne failure of the fealwater controller in the direction ofincreased feedwater flow result in a moderator temperature decrease causing a reauor pow ;r increase through the effect of the negative reactivity void coef5cient.

{

The transient response of the plant to e failure of the feedwatec controller resulting in a demand for maximum feedwata flow is presented in Reference 3 of Sectirm 15.0.

The transient is initiated from the low end of the automatic recirculation flow coat ol range, which produces a more severs steam flow / feed flow r'tismatch and level transient than that resulting at other initial conditions. The feedwiter pumps are assumed to attain a maximum capability of 115% of rated flow.

Water level increases during the initial part of the transient. De high-water level turbine trip is not initisted until the sensed ' hasincreased to the preset value specified in the Technical Specification b.. nish water level trip also trips the high-pecasure coolant injection and reactor core isolation'dooling turV.nes and the feedwater

- pumps)..The turbine trip prevents excessive moisture c iism to the turbins. Scram

- l occurs consequentiatth'the turbine trip,~ liniiting the neutron flux peak and fuel thermal transient so that no fuel damage occurs.

The DAEC is analyzed for this event with the End4f-Cycle Racinulation Pump Trip (EOC-RPT) ed both Turbine Bypass Valves (IBV) outef-service. The thermal limit impact of these out of-service conditions, or any conibination thereof, are contained in the References 3 aui 39 of Section 15.0. The Anticipated Transient Without Scram Recirculation Pump Trip (ATWS.RPT) will assure power reduction with the EOC-RPT out of service. However, operation of ATWS RPT does not impact the fuel thermal limit I

( calculations for this event.

N.

15.1-1 Revision 13 - 5/97 l

et/14/ss '04:40 0010 0019100E 800 83KEE3D t#5Je

~

UFSAR/DAEC- 1 The turbine bypass valves open to limit the peak pressure in the steam line. The ~. .

- turbine bypass valves subsequently close, bringing the pressure in the reactor vesset .

under control during mactor shutdown.

15.1.2 LOSS OF A FEEDWATERHEATER A feedwater heater can be lost m at least two ways: (1) if the steam extraction line to the heater is shut, the heat supply to the heater is . ws.d, producing a gradual cooling of the feedwater; (2) a bypass line is usually provided so that the feedwater flow mn be passed around rather than through the heater. In either case, the ranctor vetsel aceives cooler feedwater, which produces an increase in core inlet subcOg. Because of the negative void reactivity coefficient, an increase in core power results.

Reference 3 of Section 15.0 describes the response of the plant to the '.oss of 100'F of the feedwater hating capability of the plant. This . rwts the maximum expected single heater (or group of heaters) which can be tripped or bypassed by a single

) event. The reactor is assumed to be at turbine-g sr design conditions on manual recirculation Baw control when the heater is lost. De feedwater flow delay time of apprai==-!y 25 sec between the heaters and the feedwater sparger is negW4 De plant %1d continw at steady-state conditions during this delay pe:iod. Neutron flux increases above the laitial value to produce turbine design steam Gow with the higher inle* suht!5 However, a significant amount of the energy from the subsequent

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neutron power incmass would be utill sd in heating the subcooled inlet water. Thereforw, ,

thermal power would inemase at a very moderate rste 'f power exceeded the normal full-power flow control line, the e r sr would be expected to insert control rods to return the power and flow to their nonnal range. If this were not done, the neutron flux could exceed the scram setpoint where a scram would occur. Because nuclear system pressure remains esstatially constant during this transient, the nuclear system process barrier is not threatened by high internal ra dra.

15.1.3 INADVERTENT HPCIPUMP START .

Severa1 systems art available for providing high pressure cold water to the vessel.

An inadvertent stat of the largest of these suglies, the 3,000 gym HPCI system, produces the most severe transient of this type.

^

The HPCI system is inadvertently started and cold wu'er is introduced at the

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feedwater spargers. hioction of this auxiliary supply of water with the feedwater flow

- will ause reactor weier levet to incisase. The feedwater controller will co=pa- by decreasing feedwater flew. Introduction of this cold water into the reactor will increase core inlet subcooling which reduces the void fkaction. The negnive vold reactivity coefficient will induce an increase in reactor pcwer. Reactor power will increase and

( may reach the high Dux scram setpoint. The power *meresse will generate a corresponding increase in steam flow, feedwater flow, and system pressure, i, O

15.1-2 Revision 13 - 5/97

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Page 15

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