ML19263C229

Monthly Operating Rept for Jan 1979

ML19263C229
Person / Time
Site:	Rancho Seco
Issue date:	01/31/1979
From:	Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML19263C228	List: ML19263C227 ML19263C229
References
NUDOCS 7902130068
Download: ML19263C229 (19)

Text

AYERAGE DAILY UNIT P0hER LEVEL

/

DOC):ET NO. 50-312 e

gT 'CORTMN UNIT P,ncho Seco l' nit 1 I

Q\\TT i DATE 79-01-31 CO.\\1PLETED BY R. W.

Colombo TELEPilONE (916) 452-3211 h!ONT11 January lo7e DAY AVERAGE D A!LY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL

(>lWe Net)

(51we. Net) 894 893 37 2

757 jg g94 3

629 39 894 4

894 20 415 5

508 21 878 645 6

22 898 7

893 23 897 890 24 898 8

9 893 23 899 10 892 26 900 II 889 27 899-12 889 28 900 13 R91 29 898 14 892 30 850 15 892 3g 892 16 892 INSTRUCTIONS On this fermat. fist t!v ner.r;e dai!y c: it ; ower lac!in 51We Net for caeh day m the reporting month. Compute to the nearest wha?e m-:; nut:.

(9/77 )

790213006%

C'

DOCKET NO.

50-312 UNIT SilUTDOWNS AND POWER REDUCTIONS UNIT NAME R*cho hen Un4* 1 DATE 7 9-01'--31 REPORT MONTil J3"tary COMPLETED DY R.

W.

Colombo

'lELEPIIONE (91M 452-3211 E

E

.! E

}

$Yl Licensec

,E n,

{"3 Cause & Correceive No.

Date g

3g

,3 g 5 Event 7,i ?

o-Action to yE 5

.s ;p, c Report a nU

$U Prevent itecurrence H

g u

e o

1 79-01-02 F

6.8 A

3 N/A EG GENERA LOSS OF VITAL POWER BUS.lA DUE TO FAILURE OF THE A INVERTER SHUNT TRIP COIL.

REPAIRED INVERTER.

2 79-01-05 F

14.9 G

3 79-001/0lT-0 CA XXXXXX TECllNICIAN WORKING IN ICS CABINET INADVERTENTLY S110RTED ICS POWER SUPPLIES.

3 79-01-20 S

9.6 B

1 N/A HB XXXXXX MAINTENANCE SIIUTDOWN TO TORQUE FEEDWATER N0ZZLE ON "A" OTSG.

I 2

3 4

F: Forced Iteason:

Method:

Exhibit G. Instructions S: Scheduled A 13:uipment Failure (Explain) 1 Manual for Preparation of Data Il Maintenance of Test 2 Manual Scram.

Entry Sheets for Licensee C Itetucting 3 Automatic Scram.

Event Iteport (LElt) File (NMEG-D Itenularory itestriction 4-Other (Explain) 0161)

I -Oper.itor Training & License Examination -

F Administrative 5

' G.Oper.itional Er ror (Explain)

Exiiibit 1 - Same Source

('1/77)

Il-Ot he r (Explain)

/

OPERATING DATA REPORT DOCKET NO. _50-312 DATE 79-01-31 COMPLETED BY 9

U. Colombo TELEPliONE (916) 452-3211 OPERATING STATUS Rancho Seco Unit 1 Notes y, gg 3,,

2. Reportin; Period:

Jannarv 1979

3. Licensed Thermal Power IMWri:

2772 963

4. Namep? ate Ratin; s Cros M%e n:

5. Desi;n Ele:trical R tin;INet.'then:

41R

6. Slannum Denendable Cap 2 city (Gross MWe):

417

7. Maximum Dependable Capacity (Net Mbel:

873

8. If Changes Occur in Capacity Ratin;s (1: ems Number 3 Throu;h 7) Since Last Report. Gise Reasons:

N/A None

9. Power Lese! To Which Restr::*eJ. If Any thet MWe):

10. Reasons Fcr Restrictions. If Any:

N/A This Month Yr..to.Date Cumulative i1. Ilours In Reportin; Period 744 744 33,241

12. Nunner of liour, Reactor Was Critica!

726.9 726.9 19.797.7

13. Re:c!x Resene Shutdown llours 0

0 2.417.4

14. Ilours Ge. erator Oa Line 712.7 712.7 18,875.9

15. Unit Re,ene Shutdown Hours 0

0 10.8

16. Gross Thermal Ener;y Ge1er:ted (MWil) 1.948.322 1.948.32?

A7 091 679

17. Gross Electrical Energy Generated (MWi!:

625.882 625,882 15,888,682

18. Net E!ectrica! Ener;y Ger.erated IMWii) 594,134 594,134 14,993,024

19. Unit Senice Factor 95.8 95.8 56.8

20. U.,it Asai:abiH;y Factor 95.8 95.8 56.8 91.5 91.5 al./

21. Unit Caraeity Facter IUdn; MDC Nets

22. Unit Capacitt Factor (Us::u DER Nett 86.9 86.9 49.1

~

'J. Un:t forced Outa e Rate 2.9 2.9 37.6

24. Shutdouns ScheJaled Our Nest o %nths iI)pc. D.:te,and Utration of E2en t:

Maintenance Shutdown. February 24, approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

25. If Shut D..un \\t End Of Renart PerioJ.fitinuteJ Date of Start.:p:

N/A

26. Unitila Te t Status < Prior to Co:nn:erefal Operationi:

Forecast AchieuJ INITI \\ L CRl!!C \\LITY N/A N/A INITI \\ L I LEC 11'it i t 'i N/A N/A COMMERCI \\L 01'; a \\IION N/A N/A (0/77) e

NARRATIVE SU?CIARY OF PLANT OPERATIONS Date 1.1 Plant operating at 1J0% full power.

1-2 (2028)

Reactor trip due to loss of A Inverter.

(2331)

Reactor critical.

1-3 (0122)

Reactor trip due to inadvertently opening CRD breaker.

(0205)

Reactor critical.

(0317)

Paralleled Main Generator to the Grid (Closed OCB's)

(0821)

Reactor at 89% full power commenced 2-hr. hold.

(1225)

Reactor at 100% full power.

1-5 (1307)

Reactor trip due to short circuit in ICS.

1-6 (0226)

Reactor, critical.

(0400)

Paralleled Main Generator to the Grid (Closed OCB's).

(0738)

Reactor at 88% full power, commenced 2-hour hold.

(1200)

Reactor at 100% full power.

1-20 (0630)

Commenced reducing Reactor power to f'cilitate scheduled shutdown.

(0807)

Separated Main Generator from the Grid (Opened OCB's).

(0858)

Reactor at 10-8 amps.

(1504)

Commenced increasing Reactor pover.

(1745)

Paralleled Main Gc.erator to the Grid (closed OCB's).

(2058)

Reactor at 87% full power, commenced 2-hour hold.

1-21 (0535)

Reactor at 100% full power.

1-30 Reduced load to approximately 95% power due to removing A High Presr,ure feedwater heater train f rom service.

1-31 High Pressure Feedwater heater train "A" placed back in service.

Reactor at 100% full power.

PERSONNEL CllANCES REQUIRING REPORTING No personnel changes that require reporting in accordance with Technical Specifications Figure 6.9-2 were made in January, 1979.

MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1)

Functionally tested three Hydraulic Sway Suppressor (Snubbers) on DllR System following water hammer (LER-78-17).

2)

Repaired three damaged rigid pipe supports on DHR System following water hammer (LER-78-17).

SU5 MARY OF CHANGES MADE IN ACCORDANCE WITil 10 CFR 50.59(b)

Completed revis' to Rancho Seco Emergency Plan.

REFUELING INFORMATION REQUEST 1.

Name of Facility - Rancho Seco Unit 1 2.

Scheduled date for next refueling shutdown - February 1980.

3.

Scheduled date for restart following refueling - April 1980.

4.

Technical Specification change or other license amendment required -

a)

Change to Rod Index vs. Power Level Curve (TS 3. 5.2) b)

Change to Core Imbalance vs. Power Level Curve (TS 3.5.2) c)

Tilt Limits (TS 3.5.2) d)

Safety Equipment Testing (TS 3.3.3) 5.

Scheduled date(s) for submitting proposed licensing action - January 1980.

6.

Important licensing considerations associated with refueling - None.

7.

Number of fuel assemblies a)

In the core - 177 b)

In the Spent Fuel Pool - 122 spent.

8.

Present licensed spent fuel capacity - 579.

9.

Projected date of the last refueling that can be discharged to the Spent Fuel Pool - 1987.

r

SECTI0tt I - OVERVIEW Following the second refueling of Rancho Secto Unit #1, the startup test program was begun with initial criticality established at 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br /> on December 19, 1978.

Zero power physics testing commenced at that time and was successfully completed on December 21, 1978 at 0056 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />. As planned, the Zero power testing program was conducted at the iso-thermal Reactor Coolant temperature 532*F and below the power level commensurate with nuclear heat. Power escalation and testing was begun on December 21 and completed on Decenter 30 with the successful conclusion of the power imbalance detector correlation test.

The pc,wer escalation testing was donc at the three major power plateaus of 40%, 75% and 100% of full power.

The following descriptions of test data and results refer to the Cycle 3 Reload Report, BAW-1499, September 1978 testing committments and she District's flovember 15, 1978 response to the Cui;nissions 's November 7,1978 request for addi tional informat ion and commi ttment. Re fe re nce is made to that information rather than repeating i t here.

SECTION II - PRE-CRITICAL TEST

.1 Control Rod Trip Test Control rod trip time testing was done prior to establishing initial criticality and while maintaining refueling boron concentration. The conditions were, all four Reactor Coolant pumps running with the Reactor Coolant system established at 532 F and a pressure of 2155 PSIG.

All of the droppable control rods, which are assigned to Groups I through 7, were fully withdrawn, with Group 8 (APSR's) established at an inter-mediate position.

Using the nanual Reactor trip button to initiate the drop, all 61 control rods assigned to Groups I through 7 were dropped into the core f rom the full out position.

Drop time was deter-mined by using the plant computer and measuring the time from " trip" to three-fourths insertion. The fastest rod dropped in 1.175 seconds, and the slowest rod was at 1.224 seconds.

For acceptance, t he d rop t i me of Groups I through 7 had to be less than 1.66 seconds.

The nuasurement technique includes the control circuit and logic times in addition to the time for drop. All drop times were well below the acceptance criteria thus neating the Technical Specifications requirements for ful l-flo v drop t ime.

Confi rmation was made that the APSR's (Group 8) did not drop.

.2 Reactor Coolant Flow The four pump flow was determined for the hot zero power condition as being 415,700 GPM.

This can be compared to the Technical Speci fication value corresponding to 104.9% of the design flow o f 369,600 GPM. This meas u rene n t met the Cycle 3 performance requirement.

(2) c

3 Reactor Coolant Flow Coastdown From the four pump configuration described above, the reactor coolant pump determined to be the highest flow pump was tripped, and the total

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flow through the reactor core determined as a function of time. The acceptance criteria was applied to the before trip crror-reduced value, and it was determined that actual flow exceeded the minimum acceptable flow for the period of interest by a margin of approximately 15,000 GPM.

This test met the requirements for operation of Cycle 3.

(3) c

SECTION lli - ZERO POWER PHYSICS TESTING

.1 All Rods Out Boron Concentration The All Rods Out (ARO) Boron concentration was measured as described in the Cycle 3 Reload Report.

With control rod Group 8 at,<.5% withdrawn, the results were as follows:

Measured Vendor Prediction 1402.7 ppmB 1411 +100 ppmB The measured data is consistant with the prediction and meets all acce tance criteria.

.2 Boron Concentration at fiaximum Controlling Rod Group insertion Limit Meas u red Vendor Prediction 1022 ppmB 1032 +100 ppmB This measuremen t provides a second j ust cri tical Boron concentration neasurement corresponding to the predicted value. At the time of this me as u remen t, control rod Groups 5, 6 and 7 were fully inserted and control rod Group 8 positioned at 37.5% withdrawn. The measured data was con-sistant with predictions and nut all acceptance criteria.

3 Temperature Coef ficient of Reactivity at All Rods Out Boron licas u re d Vendor Predic t ion

-0.416x10-4 Ak/k/F"

-0.12x10-410.4x10-4 Ak/k/F (1394 ppmB)

(4) e

3 Temperature Coef ficient of Reactivity at All Rods Out Boron (Continued)

The val ue a t 1394 ppmB met the acceptance criteria which specifies the value shall not be more positive than +0.5x10-4 fk/k/F*.

1

.4 Temperature Coefficient of React ivi ty Correspondina to the Maximum Insert ion Boron Concent rat ion Measurement Meae.ed Vendor Prediction

-l

x10-4 Ak/k/F*

-0.997x10-4 0.4x10-'* Ak/k/ F" (1116 ppmB)

The acceptance criteria for this value is the sane as for the AR0 tempe rature coe f ficien t meas urement. This measurement met all criteria.

5 CRA Group Reactivity Worth Vendor Measured Worth Predicted To l e ra nce

%Ak/,k Worth, %Ak/k Allowed Group 5 1.217 1.24 115%

Group 6 0.904 0.87 115%

Group 7 1.505 1.49 115%

Total 3.626 3.60

+10%

I f the total neasured group worths were within 110% of the predicted value, further actions committed to in the November 15, 1978 le t ter would (5) e

5 CRA Group Reactivity Worth (Continued) not be requi red. This condi t ion was me t by these measurements. The shutdown margin calculations shown in the Cycic 3 reload report are substantiated by the above measurements and the excellent agreement between predicted and measured ARO Boron.

.6 Ejected Rod Worth Meas u remen t E rror Adjusted P redicted Ejected Rod

Worth, Tole rance Worth, %Ak/k

%Ak/k Allowed 0.773 0.64 120%

The ejected rod worth is determined for the configuration corresponding to the nuximum insertion condition allowed by Technical Specifications, namely, Groups 5, 6 and 7 fully inserted at zero powe r, wi th Group 8 at 37.5% WD.

From this configuration, the maximum worth " Ejected Rod,"

which is a rod in Group 6, was borated to full out and then swapped against Group 5 to return it to the fully inserted position as a second determination of its worth. Those two values were then adjusted for uncertainties, averaged, and are reported as the Error Adjusted value.

These results are consistent with the prediction and meet the absolute acceptance criteria of Technical Speci fications by being less than 1.0 ta k/k a t z e ro powe r.

Fu r t he rmo re, the worth of the three Group 6 rods symmetric with the measured ejected rod were determined by swapping them against Group 5 and using the calibrated worth of Group 5 over its interval to estimate the ejected rod worth. The non-error adjusted (6)

.6 Ejected Rod Worth Measurement (Continued)

Groi > 5 rod swap determined the highest worth of the three symmetric rods to be 0.669%t>k/k and the minimum measured at 0.608tak/k. These o

results are certainly within the margin of tolerance for the measurement and provide confirmation of power distribution symmetry and lack of core tilt for cycle 3 (7)

SECTION IV - POWER ESCALATION

.I Core Power Distribution Core power distributicas were taken and analyzed at the requisite Reactor power tes t plateaus of 40%, 75%, and 100% FP during cycle 3 power escalat ion. The purpose of these measurements was to verify that the minimum DNBR, maximum linear heat ra te, quad ran t power ti l t, power imbalance, and related power peaking factors would not exceed allowable limits.

In each case the measured variables were extrapo-lated to the over power trip setpoint for the next test plateau so as to assess the margin of conservatism prior to escalation. A s umma ry o f the tes t results follows:

o (8)

POWER DISTRIBUTION TEST KESULTS Measu red / Des i red Date of Data 12/22/78 12/23/78 12/26/78 Powe r level, %FP 40.3/40 72.8/75 99.1/100 Group 1-5, %WD 100/100 100/100 100/100 Group 6, %WD 100/100 100/100 100/100 Group 7, %WD 100.0/87.0 89.1/87.0 88.0/87.0 Group 8, %WD 35.5/38.5 20.0/32.0 11.0/26.0 Core Burnup, EFPD 0.3/2.0 0.83/3.0 3.6/4.0 Boron Concen tration, ppmB ll55/NA 1022/NA 923/NA Axial imbalance, %FP

-0.28/-0.58

-1.14/-0.63

+0.03/+0.84 Max incore Quadrant Power Tilt, %FP 0.83/<3.64 0.59/<3.64 0.61/<3.64 Minimum DNBR 8.84/>l.30 4.09/>l.30 2.65/>l.30 Worse Case LHR, KW/ft 5.01/<15.81 9.13/<17.15 12.84/<16.80 Max Radial Power Peak 1.377/1.37 1.353/1.35 1.340/1.34 Max Total Power Peak 1.619/1.63 1.608/1.55 1.690/1.59 Hax Peak at Core Grid / Level H-12/2 L-10/5 H-12/3 Equilibrium Xenon No Yes, 2D Yes, 3D Acceptable for Power Escalation Yes Yes Yes Extrapolations done to, %FP 85 105.5 105.5 Acceptance criteria which applies to the radial and total peaking factors is +5% and +7.5% respectively when compared to the predictions for the peak assembly at the 75% and 100% power plateaus. All acceptance criteria was net, and escalation based upon this test proved to be conservative.

The (9)

measured DNBR and linear heat rates verified that the Reactor Protective system setpoints provide protection for the core against exceeding transient CNBR and/or maximum linear heat rates assumed in the Safety Analysis and are suf ficient to protect against exceeding the Technical Specification LOCA heat rates.

.2 Power Imbalance Detector Correlation Test This test is performed to establish the relationship between the out-of-core nuclear inst rumentation and the full set of in-core sel f powered neut ron detectors as to how axial power imbalance is indicated.

Due to the effect of refueling on the neutron flux exiting the reactor, the out-o f-core indication of imbalance is expected to change.

Since the nature and magnitude of this change is not easily predicted, this test is initially performed at a l ow powe r l eve l and a ve ry con s e rva t i ve acceptance criteria imposed.

As a technique for revising the corre-lation conservatism to realistic values for the cycle, a retest at 75%FP may be done.

During this power escalation, the initial resul ts necessi tated regaining the Nuclear Instrumentation to increase the conservatism. Anytime regaining is done, a retest is required.

This retest was accomplished at 75%FP and all applicable acceptance criteria met.

The resul ts corresponding to the noximum and minimum imbalance conditions are shown here:

(10)

If1 BALANCE COND1T10f:a PARAMETERS OF INTEREST MAXIMUli MINIMUM

/

Power, %FP 74.85 74.82 i, balance, 2FP i

+1.07

-12.53 Maximum Tii t, '6 FP 0.51 0.54 DNBR, Minimum 4.02 4.32 Maximum LHR, Kw/f t 9.34 11.18 Extrapolated DNBR 1.34 1.33 E,vt rapolated LHR, Kw/f t 14.23 17.03 The test results met all acceptance criteria with minimum DNBR values at greater than 1.30.

Maximum linear heat rates remained below the center line fuel melt linear heat rate limit of 20.4 kw/f t when extrapolated to the error adjusted full power trip limit of 112%FP.

The nuclear instrementat ion correlation with in-core of fset* data is as follows:

(See Next Page)

• O f fset is defined as irrbalance divided by Power (11)

40%FP 75%FP NI Channel Di f ference Target Measured Di f fe rence Target Measured Ampl i fie r Correlation Correlation Amplifier Co r re l a t i on Correlation Gain Slope Slope Gain Slope Slope NI-5 3.70

>l.25 1.07 4.13

>l.15 1.255 NI-6 3.70

>1.25 1.03 4.13

>1.15 1.204 NI-7 3.70

>l.25 1.04 4.13

>1.15 1.243 NI-8 3.70

>1.25 1.06 4.13

>l.15 1.23i Accep-ta cc

>1.25 Re ga i n

_.1.15 Acceptable Required Cycle 3 tafety analysis assumes that the correlation slope is greater than or equal to 1.150 when above 75%FP. As the above data shows, this correlation criteria is satisfied on all protective channels, and the relationship between the incore and out-of-core instrumentation is shown to be conservative. At the same tire that this data was obtained, the relationship between the full set of incore instrumentation and those on the backup recorders was determined to rect its acceptance criteria. A gain factor of 4.13 is set into the Nuclear Instrumentation dif ferential amplifier circuitry for this cycle, and as shown here, is auequately con-servative. Thus the Reactor Protection System will protect the core from exceeding maximum linear heat rates cnd DNBR limits.

(12)

3 Power Doppler Coefficient of Reactivity From equilibrium conditions at near 100%FP, the power doppler coefficient was determined. The value obtained was -0.88x10 4 Ak/k/%FP.

The acceptance criteria for this parameter was that the value shall always be more negative than -0.55x10-4 Ak/k/%FP. This criteria is therefore satisfied.

.4 Moderator Temperature Coef ficient of Reactivity at Power The "at power" moderator temperature coefficient was measured as described in the Cycle 3 Reload Report, while operating the Reactor at equilibrium conditions and 100%FP.

Measurements determined the coefficient to be

-0.617x10~4 Ak/k/F* compared to a vendor predicted value of -1.35x10-4 Ak/k/F*. The acceptance criteria for this parameter is that it shall not be "posi;ive" for Reactor operations above 95%FP. This condition for operation is satisfied for Cycle 3 SECTION V - CONCLUSION The results of pre-cycle 3 testing and the conclusions summarized in this report demons t rate tha t Rancho Seco Uni t 1, Cycle 3, has been properly designed; and the unit caa be operated in a manner that will not endanger the health and safety of the public.

(13)