IR 05000425/1989022

Insp Rept 50-425/89-22 on 890612-16.No Violations Noted. Major Areas Inspected:Review of Completed Startup Tests & Closeout of Startup Test Insp Program

ML20245G584
Person / Time
Site:	Vogtle
Issue date:	06/25/1989
From:	Belisle G, Burnett P, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20245G574	List: IR 05000425/1989022 ML20245G571
References
50-425-89-22, NUDOCS 8906290186
Download: ML20245G584 (14)

Text

. _ - _ - - _ _ . _ _ _ _ _ . _ _ - - - - - . - - - . - . - -

- - - - - - - - - - - - - - - - - - - - - - - -

,M My .

A Nfo LUNITED ST TES

$$ +A '

NUCLEAR REGULATORY COMMISSION REGION il

-

' t [ *i ' .n g .-

,- MARIETTA STREET, N.W.

N 2r , 2 ATLANTA, GEORGI A 30323 4,

,

Report No.: 50-4?5/89-2 Licensee: 'Geor'gia Power Company P. O. Box 1295 Birmingham, AL-35201

Docket No.

50-425^ License No.: NPF-81'

. Facility Name: .Vogtle 2 Inspection' Conducted: June --16, 1989 Inspectors: h28UDate Signed

.

~Bu'rnett .

>>wW S-N $

'/ Date Signed

~

J. Zeiler Approved b C G. A. Belisle, Chief

/ f(

f

-

M'2-7h D;te Signed'

- Test Programs Section Engineering' Branch Division'of Reactor Safety SUMMAR Scope:

;This routine, unannounced- inspection addressed .the ' review of completed startup

' tests andEclosecut of the startup test inspection progra i Results:

' The licensee's performance and review of the Unit 2 startup tect progran are essentially' complete, except for issuance of the. completed Startup Test Rt. cor . 1The.; NRC's startup test inspection program for Unit 2 is also' essentiall,<

.

complete (paragraph 9) with' no adverse findings with. respect to the conduct of .

the program'or the test result This inspection identified twe items of concern about the conduct of future surveillance ' tests for both units. The first concern is that the ion chamber currents are being normalized to indicated nuclear instrument power ratherl than measured thermal power in the determination of the incore-excore

'

riuclear instrument correlation discussed in paragraph 3.b.

!

0906290186 890623 i PDR ADOCK 05000425 '-

4 Q PDC -

__ _ _ _ _ _ _ _ _ _ - - . _ _ _ _ _

.

..

.

..

.

'

The second concern is that the acceptance criterion in the test procedure for the measurement of feedwater flow is not conservative with respect to the accepted performance of installed plant surveillance instrumentation relative to the precision test instrumentation. Furthermore, the start-of-cycle surveillance test procecures used for both units did not have acceptance criteria for the performance of- installed plant surveillance instruments relative to the i precision test instruments to assure conservative indications of thermal power and feedwater flo This is discussed in paragraph No violations or deviations were identifie !

j

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

- _ _ - _ _ _

.

'

.

-

.

REPORT DETAILS Persons Contacted Licensee Employees

• ). G. Aufdenkampe, Technical Support Manager

• W. L. Burmeister, Operations Superintendent

• S. A. Bradley, Nuclear Generation Engineer

• C. L. Christiansen, Shift Supervisor

• C, L. Cross, Nuclear Procedures Supervisor

• R. Frederick, Quality Assurance Site Manager - Operations

• W. C. Gabbard, Senior Regulatory Specialist i

• W. F. Kitchens, Assistant General Manager, Operations and Maintenance

• A. L. Mosbaugh, Plant Support Manager

• A. G. Rickman, Reactor Engineering Supervisor Other licensee employees contacted included engineers and office personne Other Organizations C. B. Holland, Westinghouse

• C. Phoenix, Consul Tec 0. D. Hayes, Consul Tec NRC Resident Inspector

• R. F. Aiello, Resident Inspector

'* Attended exit interview on June 16, 198 Acronyms and initialisms used in Vogtle inspection reports are listed in j the last paragrap ! Power Reactivity Coefficient Measurements during Power Ascension (72576) l 2-6SC~01 (Revision 0), Power Coefficient Determination, was performed by i the licensee between April 20 ard May 22, 1989. The test results were  :

reviewed and approved by the general manager on June 2, 1989. The mea-

.

surements were performed by the licensee at reactor power levels of 30%, ,

'

50%, 75%, and 100%. Turbine load swings of approximately 40 Mwe were

'

initiated at each plateau agd the measured ratios of change in Tavg to the l l change in reactor power (C ) were compared ith the predicted ratios of '

l the Doppler-only power coefficient to ITC (C ). The predicted values of both variables were obtained by the licensee from WCAP-12098, The Nuclear

!

- _ _ _ _ - --_________--__a

___-

.

J '

.

-

,

l

Design and Core. Physics Characteristics of the Alvin W. Vogtle Nuclear Power Plant Cycle 1, and the inspectors confirmed that the correct values had been selected for use in the tests. The following table provides the results of this comparison at each power platea Predicted Inferred Predicted Power ITC CP C" ABS (C P +C") Doppler-0nly Deppler-0nly (%) - (pcm/ $ F) ( F/%) ( F/%) __( F/%) (pcm/%) (pcm/%)

30 -5.080 2.677 -2.308 0.369 -1 .6 50 -6.855 1.853 -1.738 0.115 -1 .7 75 -9.896 1.162 -1.249 0.087 -1 .5 100 -12.170 0.887 -0.828 0.059 -1 .8 Asshownforeachpowerglateau,theabsolutevalgeofthearithmeticsum of the measured ratio, C , .and predicted ratio, C , was less than 0.5 as prescribed by the test acceptance criterion. Also shown is a comparison of the inferred and predicted Doppler-only power coefficients as a func-

' tion of power level. The inspectgs obtained the inferred Doppler-only power coefficient by multiplying C by the predicted ITC. The results were in reasonable agreement with the predicted value No violations or deviations were identifie . Core Performance Measurements during Power Ascension (72578) i Power Distribution Limits Startup test procedure 2-6SE-02 (Revision 0), Incore Movable Detector and Thermocouple Mapping, was performed to obtain full core flux maps at each power plateau. The maps were evaluated by the licensee using the INCORE program provided by Westinghouse. The inspectors reviewed the program results and confirmed that the power distribution LCOs l listed below were satisfied at each plateau:

(1) TS 3.2.2 - Heat Flux Hot Channel Factor, 3 (2) TS 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and j (3) TS 3.2.4 - Quadrant Power Tilt Rati Incore-Excore Correlation of Nuclear Instruments 2-06SE-01 (Revision 0), Axial Flux Difference Instrument Calibration, was completed on June 9,1989, and the results accepted by the general manager on June 13, 1989. It used data obtained from the full core flux map at the 75% power plateau and from three quar-ter-core flux maps performed at that power level to evaluate PRNI chamber currents as a function of the measured incore axial offse The actual analysis of the data was performed using surveillance procedure 55003-C (Revision 3), Incore/Excore Detector Calibratio _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

-__ _ __

--_

.

L ~

.

-

F ,

i L

,

i In 55003-C, step 8.2.8 and data shaet 2 determine the equivalent full power detector current by dividing the measured current by the percent of full power indication of that PRNI channel, rather than' by the average thermal power measured during the flux ma Since the purpose of this exercise is to calibrate the PRNI channel, normaliz-ing the current to a suspect power measurement is inappropriat Furthermore, the reactor nay be at 1 different axial offset from that existing at the time the PRNIs were last normalized to thermal powe Using thermal-power-normalized chamber currents and a least-squares analysis spreadsheet with the microcomputer program SUPERCALC3, the inspector obtained correlations between currents and axial offsets that were consistently different from those reported by the license . The zero-offset currents were an average 1.34% less than licensee values. The top and bottom chamber slopes of current vs axial offset averaged 8.4% lesser magnitude and 9.4% greater magnitude, respec-tively, when calculated by the inspector. Typical results from the inspector's calculations are shown graphically in Attachment ,

At the exit interview, licensee management made a commitments to reevaluate the Unit 2 chamber correlations and to revise procedure 55003-C to normalize chamber currents to thermal power. The commit-ment will be tracked as inspector followup item 50-424,425/89-22-0 No violations or deviations were identifie . Generator Trip Power Ascension Test (72580)

2-700-02 (Revision 0), Plant Trip from 100% Power, was witnessed by the inspector when performed on May 20, 1989. This review of the completed procedure and associated documentation confirmed that all test acceptance criteria were . satisfied. In particular, the safety valves on the pressur-izer and steam generators did not lift; the reactor flux dropped to 15%

in less than 2 seconds; and the overall RTD response time in the RCS was less than 6.8 seconds (the measured value was 4.3 seconds). In addition, the pressurizer PORV did not lift. No plant performance anomalies were identified in the post trip review conducted immediately after the trip in accordance with procedure 10006-C (Revision 11), Post Trip Revie The test results were accepted by the general manager on June 1, 198 f No violations or deviations were identified, i Loss of Offsite Power Test (72582)

2-600-09, (Revision 1), Loss of Offsite Power at Greater Than 10% Power, was performed by the licensee on April 13, 1989. The test was witnessed by the inspectors and their observations are documented in NRC Inspection Report 50-425/89-16. Final approval of the test was completed by the general manager on May 15, 1989. The inspectors' review of the completed test package confirmed that all test acceptance criteria were met. Review

- _ _ . _ . _ - _ _ _ _ _ - _ _ - - _ _ - _ - - _ _ _ _ _ - _ -

_______

.

..

-

.

of the Proteus and ERF computer sequence of events data verified satisfac-tory' plant response during the tes No violations or deviations werr identifie j 6. Shutdown from Outside the Control Rocm Test (72583).

2-600-08 (Revision 1), Remote Shutdown Procedure, was witnessed by the inspectors when it was performed on April 11, 1989. Review of the com-pleted procedure confirmed that all of the acceptance criteria were satisfied, that control room personnel performed no activities beyond those permitted by attachment 10.1, Permissible Activities by Control Room Observers, and that test problems were limited to false indications by a few non-critical instruments and problems with computer displays and terminals. The completed test package was accepted by the general manager on April 28, 198 No violations or deviations were identifie I 7. Power Level Plateau Data Review (72600, 72608, 72616, 72624)

The following cumpleted startup procedures were reviewed: Automatic Control System Checkouts (1) 2-5SF07 (Revision 0), Startup Adjustment of the Reactor Control System, was performed at all test plateaus and was completed on May 19, 1989. All acceptance criteria were satisfie (2) 2-6AE01 (Revision 0), Auto Steam Generator Level- Control, was performed at all power test plateaus and was completed on May 26, 1989, and the results approved by the general manager on l June 1, 1989. All acceptance criteria were satisfie (3) 2-6AB01 (Pevision 0), Dynamic Automatic Steam Dump Control, was completed on April 4,1989, and all acceptance criteria were satisfie The results were accepted by the general manager on May 2, 198 Process Computer Checkout 2-6RJ01 (Revision 0), At-Power Inte comparison of Reactor Protection System Inputs and Plant Computer Outputs Test, was performed at the 30%, 50%, 75%, and 100% RTP power plateau The test was completed on May 16, 1989, and the results were accepted by the general manager on June 1, 198 The instruments compared included temperature sensors for the RCS cold and hot legs and average and delta tempera-tures, OTdT setpoints, RCS flow, pressurizer pressure and level, and L steam generator pressure and level. Instruments not meeting the j_ numerical acceptance criteria were recalibrates and retested success-I fully. All acceptance criteria were satisfie ,

I i

!

_ _ - _ _ _ _ ___- - >

__

.

.

-

.

5 Load Swings (1) 2-6SC-02 (Revision 1), Load Swing Tests, was performed at the 30%, 75%, and 100% RTP test plateaus. The tests consisted of 10% (of full power.) step decreases and increases in generator output demand at each plateau. The acceptance criteria included requirements that pressurizer PORVs and code safety valves and steam generator ARVs and codes safety valves not lift during any of the test transients. The test was completed on on May 20, 1989, with all acceptance criteria satisfied. The results were approved by the general manager un June 6,198 (2) 2-700-01 (Revision 0), Large Load Reduction, was performed on May 11, 1989. Power was reduced from 79 % to 24 %, as indicated on the PRNIs, and from 900 Mwe to 300 Mwe in less than a minut The plant was declared stable 18 minutes after the start of the transient. The following acceptance criteria were satisfied: the reactor did not trip, i the turbine did not trip, ii safety injection was not initiated, i pressurizer safety valves did not lift, and steam generator safety valves did not lif One acceptance criterion was not satisfied in that manual intervention with feedwater flow was required to prevent a trip on high steam generator level. In response to that test fail-ure, a TER and 0C-2-89-1065 were generated. The disposition of the DC was to retune the feedpump loop and then retest perfor-mance at the 100% tcs The 100% test was then deleted because it was not required by the FSAR. The retest requirement was evaluated by the licensee as being satisfied by the inadvertent 90% load reduction transient on May 14, 1989, during which the reactor did not tri It was noted in the inspection of that event (inspection report 50-425/89-20) that it was necessary to take manual control of feedwater flow to prevent a trip on low steam generator leve The results of this test were accepted by the general manager on June 1, 1989, Other Tests (1) 2-6AE-02 (Revision 0), Calibration of Steam and Feedwater Flow Instrumentation at Power, was performed at the 30%, 50%, 75%,

and 100% RTP test plateau It was completed on May 18, 1989, ano all test acceptance criteria were satisfie The test results were approved by the general manager on June 13, 198 l

.

,;

'

,

t ,

'

>

<

The inspectors questioned. the adequacy of acceptance criterion 9.3, which' accepted agreement of 2.5%.between. test =and plan feedwater flow instruments. During routine operation, the plant-feedwater. flow. instrumentation 11s the primary ' input to' the-calculation 'of thermal power for comparison with the license-

-limit. L This acceptance criterion creates a situation in which the required surveillance test could report thermal power 2.5%

lower' than. would be. indicated by better instrumentatio Hence,-

the surveillance test would not assure that the reactor was i operatednin. compliance with the license limit. An acceptance criterion that= standard plant feedwater flow instrumentation no indicate less' than the test instruments used in measuring -

- feedwater flow would contribute to assuring the . license limit of 3411 Mwth was not violate Currently, Unit 2 is operating with both feedwater flow and thermal power (see paragraph 8), as measured by installed plant instrumentation, at higher, more conservative, values than measured.by the. precision instrumentation. Hence this issue is not of immediate concern .for Unit 2i However,-precision heat'

balance and flow measurements are made at the start of every fuel cycle using surveillance procedure 54075-C (Revision 3),

Precision ' Heat Balanc That procedure has no acceptance criteria.

'

3 At the exit interview, licensee management made a commitment to assure that surveillance procedures for flow'and thermal power-require that plant instrumentation produce no less conservative results than do the precision instruments. This commitment will be.trackedlas inspector followup item 50-424,425/89-22-02.-

(2) 2-5RP-01, (Revision 0), RVLIS Final Calibration and Operational Checkout, was performed by the licensee during February 22.

through March 25, 1989. This test confirmed correct reactor l ,

vessel level indication during filling of the reactor vessel and -

with different combinations of reactor coolant pumps operatin Plant heatup data were also obtained to create plant specific curves of wide ' range' differential pressure versus reactor -

coolant temperature and primary system pressure. Test results met the acceptance criterion that reactor vessel level indica-tions are in agreement within $3% when comparing Train "A" with Train "B" readings. These readings differed by no more .than 1%

p' during the data gathering process. The test was accepted by the general manager on April 5, 1989.

l l

No violations or deviations were identified.

l LM EC _ - _ _ _

- - - -. --

7 Precision Heat Balance and Thermal Power Monitoring (61706)

2-55C-02 (Revision 0), Thermal Dower Measurement and Statepoint Data Collection, was performed by the licensee between April 17 and May 16, 198 Thermal power measurements were performed at 30, 50, 75, and 100%

RTP.- Data taking at each power plateau consisted of recording data at five minute intervals for fifteen minute These data were'then used to determine feedwater mass flow rates and enthalpys, steam enthalpys and flow rates, and the thermal power generated by the reactor cor During the initial data collection for each power plateau, problems were encountered from unstable measurements of plant parameters. During the 30% plateau measurement, the feedwater flow transmitters failed post-cal-ibration checks. The thermal power measurement was reperformed a day later after suitable flow transmitters were installed. At the 50% pla-teau, when control rods were placed in automatic as required by the procedure, the rods began stepping out.due to a mismatch between Tavg and Tref, and unstable plant conditions developed. The test was terminated and reperformed after stable plant conditions were met. During the 75%

plateau measurement, stability problems were again encountered when steam generator level varied greater than the 2% test requirement. A subsequent data review by the test director and operations superintendent determined

- that the 2.6% steam generator level change did not adversely affect the thermal power measurement and no retest was required. At the 100% pla-teau, the results of the calorimetric calculation indicated that the plant was .not at the proper power level when the data were taken. Nuclear instrumentation indicated 98.3% powe However, the test acceptance criterion was 100% +0, -1, and the 100% power level measurement was reperformed. The inspectors concluded that appropriate licensee action i was'taken~concerning test problems encountered during the tes ;

The microcomputer program TPDWR2, which is described in NUREG-1167 TPDWR2: Thermal Power Determination for Westinghouse Reactors, Version 2, was used by the inspectors to independently calculate thermal power at the 100% power plateau. This program was written as part of the NRC Indepen-dent Measurements Program. To customize a version for use at Vogtle, data on steam generator design features and pressurizer a.1d steam generator volumes were obtained from the FSAR, and plant drawings. The resultant heat balance data list is given in Attachment Some manipulation of plant data obtained from 2-550-02 was required by the inspectors. All pressures had to be changed from gauge to absolute and '

steam generator level had to be converted from narrow -ange level in percent to relative level in inches. The inspectors adjusted the total miscellaneous heat losses to match the value used in the licensee's calculations. Charging and letdown flows and temperatures were also given best estimate values since the licensee did not record these data. In the i

licensee's calculation, charging and letdown thermal power losses and

'

additions are included in the total heat loss term.

..

'

.

, - - - - - .

. - _ _ _ _ _ _ _

,

q

?. 76 a (.; ,- ,

G; g

,

,

q The results from TPDWR2 for the 100% power level are given in Attachmen The . agreement between TPDWR2. and the licensee's measurement was goo The ' licensee calculated the thermal power to be 3378.3 Mwth as compared

,

with 3372.2 obtained from TPDWR No violations 'or deviations were i antifie l

.

O ( Status 'of. Startup Test Program (72400)

Startup test procedure 2-600-13 (Revision 0), Power Ascension Test Sequence,

'

was essentially complete at the time of the inspection, but had not yet been reviewed and accepted by plant management. The review by the inspectors did not reveal any unfinished test ,

.The remaining inspection activities are described in inspector followup item 50-425/90-22-03: Complete inspection and review of thermal expansion tests and reactor internals vibrations tests by Septerrber 30, 198 .

' >

' The NRC Startup Test Program, Manual Chapter 2514, for Vogtle Unit 2 is close . Exit-Interview (30703)

The inspection scope and findings were summarized on June 16, 1989 with those persons indicated in paragraph I above. The inspectors described the ' areas inspected and discussed in detail the inspection finding l Dissenting- comments were not received from the licensee. Proprietary l information was reviewed, but is not contained in this report. The open  !

items listed below describe commitments made by licensee managemen ,

Inspector followup item 50-424 and 425/89-22-01: Revise procedure  ;

55003-C to normalize chamber currents to thermal power vice nuclear 1 instrument channel power indication paragraph 3.b).  ;

I Inspector followup item 50-424 and 425/89-22-02: Assure that surveil- !

lance procedures for flow and thermal power require that plant instrumentation produce no less conservative results than do the

. precision instruments (paragraph 7.d).

Since the following open item requires no action by the licensee, it was not discussed at the exit intervie l Inspector followup item 50-425/90-22-03: Complete inspection and review of thermal expansion tests and reactor internals vibrations ,!

tests by September 30, 1989 (paragraph 9).

L i

!

l

. - -

h y

.+

'

L L..

{sj Q -: . h

,

g; O .y V ,

l ii; p

11. Acronyms and Initialisms_Used in This Report

.

' ABS' - absolute value ARV_ - atmospheric relief valve DC . deficiency card

' '

ERF - emergency response facility FSAR - Final Safety Analysis Report-ITC.~ - isothermal temperature coefficient LCO - limiting condition for operation he ' -

megawatts - electrical h th - megawatts - thermal OTdT - over temperature delta temperature-

.pcm. - percent millirho, a. reactivity unit PORV - power operated relief valve -

PRNI - power range nuclear instrument RCS~ - reactor coolant system RTD resistence temperature detector RTP - rated thermal poi.er RVLIS- reactor vessel level : indication system Tavg.- average RCS temperature TER - test evaluation report Tref - reference'RCS temperature TS - _ technical specification Attachments:

1.- Typical Incore-Excore Correlation 2. -Heat. Balance Data

~ 3.- Heat Balance

!

Ii ar u;i . -

. *

-

<Ci "E '

.

,

. ,

s OoZgcZs"$k4Od$An^

-

c 2 3 3 9 4 9 4 0 0 4

- 0 0 2 - -

.-

-

I

-- -~

B f

N N C

,

'

1 i A6 X 4O I

A 4R

_

_

L - ' D E O a _

F - t F1 1, aE E

S T

aX nC ( dO

)

% LR e

- ,  ;

aE s

t C

qSO uR

- aR rE

1 i I I I I I I l ' i _

iil il i ' i *

1 l 'i,t ' ' e _

e I ,

_

• ? x O sL FA i

As IT Dt R D R e A e g T g T r

e A r e A O s

s i

o B

- s s

i o T

-

N n O n O T P B

O T

T

- T O

M T O

P

-

-

O M

lil  !

I I - -

- -_ __ *

l

,

ATTACHMENT 2

'

. ,- ,

EAT MMRCE DATA V0GTLE2 05-15-89 -

PLANTPARAMIfRS: .

REACTORCOOLANTSTSfD HTLECflVI IISUMf!05 PumpPower(HWeach) InsideSarfaceArea(egit) 5,000 Pusplificiency(1) 9 Beat Loss Coefficient (EfUe/hr eg it) 0.00 Preesurlier Inside Diaseter (inches) 8 NONREFLICflVI IISDM!l0N l STEAM GDERATORS InsideSurfaceArea(egft) 13,000 DeneInsideDiaseter(inches) 168.50 Thickness (inches) RiserOutsideDiameter(inches) 20.00 ThermalConductivityfBTUs/hrftI) 0.000 Number of Rieers 16 Holstne Carry-over (%) in A 0.000 LICHSD THEIBAL POUR (HWt) 3411 HolstureCarry-over(1)inB 0.000 HoletureCarry-over(X)in0 0.000 HoittureCarry-over(X)inD 0.000 DATA:

!

fIME 2335 FINE 2335 STEARGEHuf0RA STRHGENERATORB SteasPressure(psia) 101 SteasPressars(peia) 102 FeedvaterFlov(86lb/hr) 3.805 FeedvaterFlav(16lb/hr) 3.756 Feedvaterfesperature(F) 44 feedvaterfesperstare(F) 43 SarfaceBlovdown(g=) SurfaceBlowdown(sps) BottonBlovdown(sps) 7 BottonBlowdown(gps) 7 WaterLevel(inches) 51 WaterLevel(inches) 51 STEANGEHuf0RC STRE GIIDATOR D SteasPreesure(psia) 100 SteasPressere(psis) 101 FeenvaterFlov(16lb/hr) 3.806 Feedvater Flav (56 D/hr) 3.720 feedvaterfesperature(F) 44 feedvater fesiersture (F) 44 SurfaceBlowdown(gps) SurfaceBlowdown(gps) BottosBlowdown(gps) 7 BottonBlowdown(gps) 7 WaterLevel(inches) 51 WaterLevel(inches) 51 LEfDOWNLin CEARGibGLlH Flev(gps) 7 Flev(gps) 6 fesperature(F) 56 temperature (F) 54 PRESSURIZER REACTOR Presene (psia) 225 tare (F) 56 Water Level (inches) 34 fcold(F) 56 _ _ _ - _ 1

_ _ - __ _ _ _ _ _ _

'

... ATTACHl1ENT 3 -

1. .' .

HEAT BALANCE VOGTLE 2 05-15-89

'

DATA SET 1 OF 1 ENTHALPY FIDW POWER POWER 2335 hours0.027 days <br />0.649 hours <br />0.00386 weeks <br />8.884675e-4 months <br /> (BTUs/lb) (E8 lb/hr) (E9 BTUs/hr) (MWt)

STEAM GENERATOR A Steam 119 .777 4.501 Feedwater 41 .805 -1.598 Surface Blowdown 54 .00000 0.00000 Bottom Blowdown 48 .02827 0.01359

_______

Power Dissipated 2.9188 85 STEAM GENERATOR B Steam 119 .727 4.443 Feedwater 41 .758 -1.575 Surface Blowdown 54 .00000 0.00000 Bottom Blowdown 48 .02893 0.01389 Power Dissipated 2.8821 84 STEAM GENERATOR C Steam 119 .777 4.504 Feedwater 42 .808 -1.599 Surface Blowdown 54 .00000 0.00000 Bottom Blowdown 47 .02919 0.01400 .

Power Dissipated 2.9197 85 STEAN GENERATOR D

,

Steam 119 .891 4.401 Feedwater 42 .720 -1.583 Surface Blowdown 54 .00000 0.00000 Bottom Blowdown 48 .02887 0.01387 Power Dissipated 2.8513 83 l OTHER COMPONENTS Letdown Line 56 .02780 0.01558 Charging Line 53 .02488 -0.01328 Pumps -0.08031 Insulation Losses 0.00000 Power Dissipated -0.05804 -17.0 ;

REACTOH POWER 33 2

,