Text

Insp Repts 50-369/87-42 & 50-370/87-42 on 871215-18. Violations Noted.Major Areas Inspected:Thermal Power Monitoring & Nuclear Instrument Calibr & Operability.Heat Balance Data Sheets Encl
	ML20149E209
Person / Time
Site:	McGuire, Mcguire
Issue date:	01/14/1988
From:	Burnett P, Jape F; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20149E192	List: ML20149E190; ML20149E199; ML20149E209;
References
	50-369-87-42, 50-370-87-42, NUDOCS 8802100403
	Download: ML20149E209 (13)
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UNITED STATES

_ f,f* 'o. NUCLEAR REGULATORY COMMISSION

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101 MARtETTA STREET, N.W.

s ATLANTA, G EORGI A 30323

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Report Nos.: 50-369/87-42 and 50-370/87-42

Licensee: Duke Power Company

-422 South' Church Street

Charlotte, NC 28242

Docket Nos.: 50-369 and 50-370 License Nos.: NPF-9 and NPF-17

Facility Name: McGuire 1 and 2

Inspection Conducted: December 15 - 18, 1987

Inspector: h d , E /v /- / v-gg

Date Signed

[,v- P. T. Burne/t'

Approved by: M V

d

Date Signed

F. Jape, Section Chief r

Engineering Branch

Division of Reactor Safety

SUMMARY

Scope: This routine, announced inspection addressed the areas of therrnal power

monitoring and nuclear instrument calibration and operability.

Results: One violation was identified - failure to make a timely report -

paragraph 6.

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REPORT DETAILS

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1. Persons Contacted

Licensee Employees

N. Atherton, Compliance

R. Banner, Compliance

R. Browder, Engineer, Operations

M. Case

D. Ethington, Engineer, Ccmpliance

G. Gilbert, Test Engineer

B. Hamilton, Technical Services Superintendent

M. Kitlan,* Reactor Engineer

W. Kronenwetter, General Office, Design Engineerir.g

L. Kunka, Engineer, Reactor Group

S. LeRoy, General Office, Licensing

T. McConnell, Station Manager

E. McCraw, Compliance Engineer

M. Mustian, Engineer, Instrumentation and Electrical

M. Nazar, Engineer, Performance

M. Sample, Superintendent of Integrated Scheduling

D. Smith, Performance Test Engineer

J. Snyder, Performance Engineer

W. Suslick, Engineer, Test Group

R. Travis, Operations Superintendent

Other licensee employees contacted included engineers, operators, and

c.ffice personnel .

NRC Resident Inspectors

W. Orders, Senior Resident Inspector

D. Nelson, Resident Inspector

Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized on December 18, 1987,

with those persons indicated in paragraph 1 above. The inspector de-

scribed the areas inspected and discussed in detail the inspection find-

ings. Dissenting coments were not received from the licensee.

Proprietary information is not contained in this report. The inspection

findings included:

a. Violaticn 369/87-42-01: Failure to make a required report within the

allotted time - paragraph 6.

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b. Licensee identified violation 369/87-42-02: Thennal power averaged

over eight hours exceeded 100% RTP on severa1' occasions during the

period May 20 to June 11, 1987 - paragraph 6.

3. Licensee Action on Previous Enforcement Matters

This subject was not addressed in the inspection.

4. Unresolved Items

No unresolved items were identified during this inspection.

5. ' Calibration of Nuclear Instrument Systems (61705)

When Unit I returned to power on November 15, 1987 following a refueling

outage, the bottom chamber of PRNI N43 was found to be producing about 50%

more current than predicted; while the other seven chambers had outputs

close to predicted currents. Channel N43 was then placed in trip pending

review of the situation by the licensea and Westinghouse, the manufacturer

of both the chamber and the NSSS.

The reviewers concluded there were no known failure modes of the chamber,

which would lead to an increased current. The most likely cause of the

high current, then, was the addition of moderator, water, to the instru-

ment thimble surrounding the hermetically sealed canister containing the

neutron-detecting chambers. Since virtually all of the neutrons leaking

from the reactor ate epi-thermal, ar.d the chambers are most sensitive to

thermal neutrons; thermalization of the neutrons would lead to increased

current. That the upper chamber did not show an increased current indi-

cated that the thimble was only partially filled with water.

The channel was returned to service and calibrated to eliminate any

quadrant power tilt to permit raising power above 50% RTP. Between 50 and

80% RTP, all PRNI channels were calibrated against the incore flux maps to

obtain consistent axial offset. As operation continued, current from the

lower N43 chanber decreased relative to the other chambers, increasing the

N43 A0 relative to that of the other channels. This was judged to be the

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result of water draining or evaporating from the thimble.

. Technical Specifications require an incore-excore calibration every

90 days. Technical Specification Table 4.3-1 requires agreement between

incore and excore A0s of 3%. The licensee established a criterion to

perform the incore-excore calibration whenever the N43 A0 changed by 3%

relative to the other three channels. Three recalibration have been

reev ired since the initial one to satisfy the criterion. The licensee's

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trending of the current indicates it is decreasing linearly at the rate of

0.33%/ day. Other PRNI parameters being trended include excore tilt from

the lower detectors, excore axial offset, and lower and upper excore

powers.

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With one exception, the contribution of a PRNI channel to control and

safety systems is based on the total channel, sum of both chambers, out

put. For any channel, adjusting the output to agree with calorimetric

power is easily and routinely done whenever the absolute difference

approaches 2%. The over temperature delta temperature circuit trip

setpoint, however, does include a tenn dependent on A0. A sensitivity

study performed by Westinghouse showed that the lower chamber current

could decrease by 50% without requiring an adjustment in the circuit

constant.

The inspector concluded that channel N43 was operable and that the licens-

ee's trending and recalibration activities were adequate to maintain

operability or to promptly detect failure.

In the course of reviewing the incore-excore recalibration activities, the

inspector learned that there is an incore quadrant power tilt of cpproxi-

mately two percent. That tilt was not reflected by the excore PRNIs when

they were recalibrated. They showed no tilt. It was not clear that this

action is permitted by the Technical Specification requirement that power

be reduced when a quadrant power tilt measured from the PRNIs exceeds two

percent. This procedure was discussed with members of the Reactor Systems

Branch in NRR. They confirmed that when an incore power measurement shows

the hot channel factors to be within limits in the presence of a tilt the

PRNIs may be recalibrated to show no tilt. The purpose of the tilt limit

determined by the PRNIs is to monitor core changes between incore power

maps.

Unit I tripped on December 28, 1987. During that brief outage between

five and six gallons of water were drained from the N43 thimble, and the

entire instrument canister was replaced. Subsequent operation of N43

appears to be normal.

No violations or deviations were identified.

6. Statistical Analysis of Power History (61706)

References:

(1) NRC Memorandum, "DISCUSSION OF LICENSED POWER LEVEL," dated

August 22,1980, from E. L. Jordan, Assistant Director for

Technical Programs, Division of Reactor Operations Inspection,

Office of Inspection and Enforcement, to multiple addressees.

(2) McGuire Nuclear Station Memorandum to File, "Statistical Analy-

sis Showing Unit 1 Did Not Violate 100% Rate (sic) Thennal Power

Between May 20 and June 11, 1987, Revision 1."

(3) Standard Mathematical Tables, Tenth Edition, Chemical Rubber

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Publishing Company, Cleveland, Ohio (1954), pp 239-244.

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(4) Acheson J. Duncan, Quality Control and Industrial Statistics,

Richard D. Irwin, Inc, Homewood, Illinois (1974), pp 580-584 and

709-711.

The criteria for determining significant overpower operation in excess of

the license limit were established in reference 1. One criterion is that

thermal power averaged over an eight-hour shift exceeds 100% of RTP. In a

review initiated on June 11, 1987, the licensee identified occurrences of

eight-hour averages exceeding 100% RTP during the period May 20 to that

date. The largest average identified was 100.18%.

The thermal power monitoring program on the OAC performs a new calculation

every minute and updates the power level displayed on the operators screen

with that frequency. However, at the time of this event, the results were

saved to computer storage only once per hour, with the hourly printout of

operating parameters. The stored value was the most recent vne-minute

calculation, not the average over the hour. The licensee's initial

evaluation was based on the assumption that the recorded observation

represented the average power of the preceding hour. (This is connon

practice in most surveillance-related sampling activities.)

The subsequent corrective action was excellent. (That much of it had been

planned earlier as a routine operating improvement is also to the licens-

ee's credit.) The programming of the OAC was changed to save the

minute-by-minute thermal power calculations to calculate and display the

running average power for the preceding one and eight-hour periods. The

program modifications also provided computer-generated alarms when any one

of the operational limits discussed in reference 1 is approached. Discus-

sions with operators in the control room confirmed they had become more

sensitive to the need to monitor and respond to average power.

Prior to issuing a LER, the licensee performed additional analysis of

their power history and determined, to their satisfaction, that they had

not averaged over 100% RTP over the entire period in question. This

conclusion was based on a statistical analysis performed by one of the

operations engineers. His work is summarized in raference 2.

The essence the analysis was that in an eight-hour period 480 calculations

of power would be perfonned, but on'y eight retained, and those eight did

not give a good estimate of the average of the 480. Therefore, an accu-

rate eight-hour-average power could not be obtained fram eight data, but

could be represented by a larger number of data collected over a longer

period of time. To that end, the engineer reviewed the 576 power-level

data retained over the period of interest. He discarded all observations

less than 98.5% RTP as not being representative of full power operation

and created sequential eight-hour averages of the remaining 546 readings

as well as an overall mean. He then calculated the standard deviation of

the eight-hour averages and a standard deviation of of the eight-hour

mean. His results for the mean of eight-hour averaus and standard

deviation of that mean were 99.91 and 0.0215% RTP, respective?y. Since

the mean plus three standard deviations of that mean did not exceed 100%

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RTP; he concluded that there was 99% confidence that in the average

eight-hour period power did not exceed the limit.

Although accepted by licensee management as valid, the argument is weak

from several aspects. First, the regulatory issue is not what the average

power was over the period of interest or the confidence that can be placed

in that average. The issue is whether in any eight-hour period power

averaged above 100% RTP, and that issue is not addressed in the licensee's

analysis. Second, from a statistical stand point, there is an assumption

that power, except for periods of reduced power operation, was held at a

constant value by the automatic rod control system, and that any varia-

tions observed about that constant value are from random variations of

process variables. However, no analysis of variance or test of the mean

was performed.

The inspector obtained the licensee's power data file for this period on a

micro floppy disk, and converted the file to a format compatible with the

micro computer program SUPERCALC3. He then used that program for inde-

pendent statistical analyses of the licensee's power history data. Review

of the data in sequence shr,wed there were five periods of apparent full

power operation separated by periods of reduced-power operation. The five

periods contained in order 32, 201, 178, 92, and 33 power records. The

lowest recorded power in any set exceeded 98.9% RTP. Assuming the data

were truly random, a mean and variance were calculated for each set of

power data. The means of the sets were compared in pairs using the

Students' T test. In tnree of the seven pairings considered, the means

were significantly different at the 95% confidence level and two pairings

were different at the 99% confidence. level. These results challenge the

assumption of constant power over all subperiods of the total period of

interest. The variances were analyzed in pairs using the F test. In one

of four pairings considered, there was a significant difference in vari-

ances. This challenges the assumption of constant variability among the

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data. The T test and F test and their tables are given in reference 3.

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Bartlett's test of homogeneity of variances is discussed in reference 4.

It provides a means of testing the variances of the five sets as a group.

The results were that there was less than 1% probability that all sets

have a common variability. The text of the reference notes that the test

is based upon an assumption of random variation of the data within each

! group of data. If this is not true, a finding of significant difference

in variability may reflect a departure from normality rather than hetero-

geneity of variance.

If the data are not normally distributed, then no statistical test is

valid for the analysis of the data, and no valid conclusions may be drawn

if the tests are performed. Qualitatively, this lack of normality is

demonstrated from calculating the running eight-hour-average power level.

This calculation showed that there were 146 occurrences of power averaging

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over 100% RTP over the inmediately preceding eight-hour period. Many of

these periods were sequential with cluster sizes of 22, 20, 17, 16, 13, 9,

9, 7, and 6. A cluster size of 22 means that the average of 29 sequential

power records obtained with hourly frequency exceeded 100% RTP,

Finally, a chi-squared test was performed on the 536 data as a whole as a

test of nonnality. The result was there was less than five percent

probability that a normal distribution would have produced as large a

statistic. The test results are summarized in Table 1 below. This test

was also performed for a sequence of 96 one-minute-interval data obtained

form Unit 2 with a similar result. This result does not impugn the

validity of the average of such data being representative of a one or

eight-hour period of reactor operation. It does suggest that the

minute-to-minute variations in power are systematic rather than rindom.

The inspector concluded there was no statistically valid means of refuting

the straight-forward interpretation of the reactor power data and the

conclusion that power had averaged over 100% RTP for an eight-hour period.

The analyses discussed above were performed by the inspector subsequent to

the on-site portion of the inspection and the exit interview. At the exit

interview, the licensee responded to the inspector's qualitative concerns

about the validity of their statistical analysis by volunteering to issue

an LER. The failure to 'ssue a timely LER has been identified as a

violation of 10 CFR 50.73.a(2) (VIO 369/87-42-01). The extended overpower

operation has been identified as a licensee identified violation (LIV

369/87-42-02). No requirement for additional corrective action to pre-

clude further overpower operation has been identified.

Table 1: Chi-Squared Test of Normality

Class Actual Theoretical

Upper Frecuency Frequency (F-f)2

Limit (F) (f) f

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99.548 30 24.9 1.034

99.663 28 39.9 3.565

99.778 60 71.6 1.882

99.893 101 99.6 .020

100.008 128 107.3 3.990

100.121 98 88.2 1.095

100.236 50 58.7 1.302

100.351 29 29.6 .014

infinity 12 16.1 1.035

............... 3:.......... 3'.............

SUMS = 536 536.0 13.938

For 6 degrees of freedom, the probability, P(6,13.9), that

a random sample gives no better fit is 5%. Hence, the

hypothesis of nornality is re.jected. I

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. 7. Independent Analysis of Thermal Power (61706)

a. References

(1) NUREG-1167. TPDWR2: Thermal Power Determination for Westinghouse

Reactors, Version 2,

(2) -McGuire Nuclear Station FSAR, Chapter 5,

(3) Westinghouse Technical Manual 1440-C247, Pressurizer Instruc-

tions ... ,

(4) Westinohouse Te:hnical Manual 1440-C250, Vertical Steam Genera-

tor ?.n ruction: ... ,

(5) 0AC Manual:

(a) Thermal Outputs Calculation, Section 3.2.10, and

(b) Thermal Outputs Calculation Dump, Section 3.2.14.

b. Parameter and Cata Acquisition

The micro computer program TPDWR2, developed by the NRC's Indepen-

dent Measurements Program for analysis of licensee thennal power data

is described in Reference (1). In order to customize the program for

use at FcGuire, plant specific physical and performance parameters

were obtained from references (2) to (5). Those parameters are given

on page 1 of attachment 3 along with typical input data for the

calculations described below.

Using the plant computer to log the input data for TPDWR2 assures

better numerical resolution and contemporaneousness of the data than

manual collection from MCB indicators can provide. All of the

necessary data were obtained using edits frem computer point identi-

fication tables 1, 4, 10, and 11. The points were printed out with

one-minute frequency for over an hour. The inspector selected six

sets of data for analysis in pairs separated by about 30 minutes.

Most of the data were not in a form that could be input directly to

TPDWR2. Data sources, with the loop A computer points used in the

examples of loop-specific parameters, and the required manipulations

are described below:

!

S/G pressure = pressure (psig, A1107) + atmospheric pressure (P0117)

! FW flow (Mlb/hr) + average (P1412, P1413)

FW temperature = A0454 (no manipulation necessary)

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BD ficw (gpm, S/G conditions) = A0652(lb/hr)*0.002514

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'S/G level (inches) = A1059(%)*2.33-+ 394

LD flow e A0764 (no manipulation necessary)

LD temperature = A1088 (loop C cold leg)

CHG flow (gpm) = A0758 - 32gpm (flow to the seals does not return

enthalpy from the regenerative heat

exchanger)

CHG temperature = A0758 (regenerative heat exchanger outlet

temperature)

PZR pressure (psia) = P1389

PZR level (inches) = A0976(%) * 5.205 + 25.75

NC average temperature = P1461 (no manipulation required)

NC average cold leg temperature = average (A1064, A1076, A1088)

A SUPERCALC3 spreadsheet was used to perform all of the necessary

calculations and to organize the results in an order best suited for

input to TPDWR2.

c. TPDWR2 Calculational Results

The comparison between TPDWR2 and the licensee's calculations of

power was very good, with TPDWR2 consistently the lower by 0.2 to

0.6% RTP. Typical results for TPDWR2 are given on pages 2 and 3 of

attachment 2. Some of the differences in results may come from the

calculation of blowdown enthalpy. TPDWR2 uses an average of steam

generator saturation conditionr. and feedwater conditions to calculate

enthalpy for bottom BD flow.

The licensee's value of 75% net efficiency for the reactor coolant

pumps was used in the TPDWR2 calculations, hence that quantity does

not contribute to the differences in results. However, 75% is the

lowest value of pump efficiency the inspector has seen clained by any

licensee, and, if in error, would slightly penalize power generation.

The licensee's calculational method is acceptable as prograrned. The

differences noted, if real, would lead to over estimating thermal

power rather than ur. der estimating it.

No violations or deviacions were identified.

Attachments:

1. List of Acronyms and Initialisms Used in This Report

2. TPDWR2 Parameters, Data, and Results

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ATTACHMENT 1

LIST 0F ACRONYMS AND INITIALISMS

- AO - axial offset, the difference in percent'pewer between the upper and

Iower chambers of a PRNI channel

BD - blowdown (water removeii from S/Gs as part of a continuous cleaning

process)

CHG - charging or makeup water

FSAR- Final Safety Analysis Report

FW - feedwater

LD - ietdown (water removed from the primary system its part of a purification

process)

LER - licensee event report (required by 10 CFR 50.73)

MCB - main control board

NC - nuclear coolant (system)

NSSS- nuclear steam supply system

0AC - operator aid computer ( .he plant computer)

PRNI- power range nuclear instrument

PZR - pressurizer

RTP - rated thermal power

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'ATTACHMgNT2,

HEAT BALANCE CATA

Mc6UIRE 2

12-15-87

PLANT PARAMETERS:-

FEACTOR COOLANT SYSTEM FEFLECilVE INSULAi!0N

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Pump Power (MW each) . 5.2 Inside Surface Area (sq ft) .

15,958

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.Puep Efficiency (I) 74.7 Heat Loss Coefficient (BTUs/hr sq f t) .55.00

Pressurizer Inside Diaseter. finches) 84.0

N0hREFLECTIVE INSULATION

STEAM GENERATORS . Inside Suriace Area (sq f t) !!,575

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Dose inside Diaseter (inches) 168.50 Thickness (inches) - 4.0

Riser Outside Diaseter (inches) 21.00 Thereal Conductivity (BTUs/hr it F) 0.035

4 Number of Risers 12

Moisture Carry-over (1) in A 0.100 LICENSED THERMAL F0WER (MWt) 3411-

'

Moisture Carry-over (Il in 9 0.100

Moisture Carry-over (U in C 0.100

Moisture Carry-over (I) in D ' O.100

DATA: SET I SET 2 SET 1 SET 2

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TIME 1742 1811 TIME 1742 1811

STEAM SENERATOR A STEAM GENERATOR B

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Steae Prassure (psial 1000.2 999.6 Steas Fresnre (psia) 996.3 996.3

I

Feedvater Flow (E6 lb/hr) 3.889 3.906 Feedsater Flou (E6 lb/hr) 3.722 3.779

Feedsater Temperature (F) 438.7 438.8 Feedsater Temperature (H 438.0 438.0 1

Surface Blewdown (gps) 0.0 0.0 Suriace Bloudown (gps) 0. 0 ' O.0

Bettca Bloedown (gpal 151.0 150.1 Botten Blondenn (gpal 148.6 142.1 :

Water Level (inches) 558.0 556.4 Water Level (inches) 548.7 548.7 [

4

STEAM SENERATOR C STEAh EENERATOR D

Steae Pressure (psia) 1001.5 1001.2 Steas Fressure (psia) 998.6 998.6

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Feedmater Flcu (E6 lb/hr) 3.911 3.902 Feedsater Flce (E6 lb/hr) 3.708 3.716

Fredeater Temperature (F) 438.1 438.3 Feedsater feeperature (F) 439.0 4*8.7

)

L Suriace Blcadoun (gpa) 0.0 0.0 Suriace Blendcun (gps) 0.0 0.0

l- Botton Bhudenn (gps) 162.5 165.0 Bettes 81cedean (g;s) 138.4 144.0

Water Level (inthes) 545.7 545.5 Water Level (inches) 545.9 546.1

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LETDC W LINE CHAR 81N6 LIXE

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Fica (gps) 82.8 83.1 Fles (gse) 60.5 60.8

iesperature (F) 559.5 559.4 Teaperature (F) 510.1 510. 'e

i

PEES!JRl!ER F.EACTOR

Pressure (psia) ?267.0 2266.0 i ave (F) 559.0 587.9 ;

j kater Level (inches) 357.3 3 7.8 i cold (F) 559.3 55'.3 t

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Attachinent'2 ' -i2~

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a HEAT BALANCE

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.McGUIRE 2

12-15-87-

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DATA SET 1 OF 2 ENTilALPY. , FLOW POWER POWER

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1742 hours0.0202 days 0.484 hours 0.00288 weeks 6.62831e-4 months (BTUs/lb) (E6 lb/hr) (E9 BTUs/hr) (MWt)

Y STEAM GENERATOR A

Steam 1192.3 3.831 4.567

Feedwater 418.0 -3.889 -1.626

Surface Blowdown 542.6 0.00000 0.00000

Bottom Blowdown 478.0 0.06005 0.02870

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Power Dissipated 2.9703 869.9

STEAM GENERATOR B

Steam 1192.4 3.663 4.368

Feedwater 417.2 -3.722 -1.553

Surface Blowdown .542.0 0.00000 0.00000

, Bottom Blowdown 477.3 0.05913 0.02822

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Power Dissipated 2.8429 832.6

STEAM GENERATOR C

Steam 1192.2 3.847 4.586

Feudwater 417.4 -3.911 -1.632

Surface Blowdown 542.8 0.00000 0.00000

Bottom Blowdown 477.7 0.06464 0.03088

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Power Dissipated 2.9847 874.1

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STEAM GENERATOR D

Steam 1192.3 3.653 4.355

Feedwater 418.3 -3.708 -1.551 .

! Surface Blowdown 542.3 0.00000 0.00000 ;

Bottom Blowdown 478.1 0.05504 0.02631 ;

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Power Dissipated 2.8303 828.9

OTHER COMPONENTS :

Letdowa Line 559.0 0.03072 0.,,717

< Charging Line 499.2 -0.02389 -0.01193

Pressurizer 637.8 0.00010 0.00012

, Pumps -0.05296

Insulation Losses 0.00147

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Power Dissipated -0.04613 -13.5

__=_ _ :

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REACTOR POWER 3392.1

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.. Attachment 2' 3 .

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HEAT BALANCE-

McGUIRE 2

1: ,

12-15-87

DATA SET 2 OF 2 ENTHALPY FLOW. POWER POWER

1811 hours0.021 days 0.503 hours 0.00299 weeks 6.890855e-4 months (BTUs/lb) (E6 lb/hr) (E9 BTus/hr)- (MWt ) -

STEAM GENERATOR A

Steam 1192.3 3.848 4.588

Feedwater 418.1 -3.906 .t.633

Surfa'ee Blowdown 542.5- 0.00000 0.00000

Bottom Blowdown 478.0 0.05969 0.0285i

. . _ _ _ _ _ . .

Power Dissipated 2.9833 873.7.

i

. STEAM GENERATOR B

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Steam 1192.4 3.722 4.439

i Feedwater 417.2 -3.779 -1.577

I Surface Blowdown 542.0 0.00000 0.00000

Bottom Blowdown 477.3 0.05655 0.02699 >

Power Dissip0ted 2.8890 846.1-

STEAM GENERATOR C

Steam 1192.2 3.837 4.574

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Feedwater 417.6 -3.902 -1.629

Surface Blowdown 542.7 0.00000 0.00000

Doktom Slowdown 477.8 0.06563 0.03136

..______

Power Diusipated 2.9762 871.6

STEAM GENERATOR D

Steam 1192.3 3.658 4.362

Feedwater 41L 0 -3.716 -1.553

Surface Blewdown 542.3 0.00000 0.,00000

Bottom Blowdown 477.9 0.05727 0.02737

_______

Pows. Dissipated 2.8361 830.6

OTHER COMPONENTS

Letdown Line 558.9 0.03083 0.01723

Charging Line 499.3 -0.02401 -0.01199

Pressurizer 637.5 0.00018 0.00012

Pumps -0.05296

Inse.lation Losses 0.00147

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Power Dissipated -0.04613 -13.5

______

REACTOR FDWER 3408.6

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ML20149E209

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