ML19320B987

Monthly Operating Rept for June 1980

ML19320B987
Person / Time
Site:	Rancho Seco
Issue date:	07/07/1980
From:	Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML19320B984	List: ML19320B983 ML19320B987
References
NUDOCS 8007150592
Download: ML19320B987 (3)

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_Q' NARRATIVE SDDIARY OF PLANT OPERATIONS 6-1 to

. Continuous reactor operation at s96%.

Performed routine 6-30 surveillance and preventive maintenance items.

MAJOR ITE'!S OF SAFETY-RELATED MAINTENANCE 1.- Replaced and tested Control Rod Drive undervoltage trip delay.

SIDDIARY OF CHANGES MADE IN ACCORDANCE WITH 10 CFR 50.59(b)

'1.

Added a key switch & annunciator in Control R om for bypassing control grade reactor trip on loss of feedwater.

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2..' Provided " enclosure around nuclear service transformer X43A.

3.

Modification of NNI.

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a UNIT S!IUTOOVINS AND PCWER REDUCTIONS.

DOCKETNO,

' 50-31?

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UNIT NAME __ R:mcho Secc. cl, DATE _ 80-06-30

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, REPORT MONT!!'. Tune ~1980 COMPLETrl) ny R tt c.31 cese

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1 E1.[PilONE _ *116_- W>--T> 1 1-e.

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Date M.

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NO OUTAGES OR SIGNIFICNE POWER REDUCTION (GRFXEit *0i,W203 R

' TOR 'THE PRECEDING 24 Houas) MIS toNm.

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F: Forced

' S: Scheduled Reason:

3 A Eip:ipnierit Failure (Exp!ain)

Method:

4 11-M.ilnter.ani.e ni Tc.st l M must Exhibit G. instructions C. Refueling 2-M mual Scram.

fin Preparation of D.its Entry Sheets Ihr I.icensee I).Hegulatory Restriction 3-Antomatic Scram.

E Operain Trainierg & License Examinailon 4 Other(Explain)

Eveni Report (LI:RI File (NUREG-0161)

F. A.iniinir.t ra tive (V/77)

(i Ope:ational !!rror (Explain) 11041 r(thplain)

S Eshibit I. 5.une Source g

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OPERATING DATA REPORT 50-312 DOCKET NO. -

D ATE _ 80-06-30 CO3!PLETED P,Y R.

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T'ELEP110NE 91 A-32-3211 OPERATING STATUS.

1. Unit Name:

Rancho Seco #1 Notes

2. Reportin; Period:

June 1980

3. Licensed T hennal Power f.\\ twit:

2772

4. Nameplate Rating iGross 31We):

963

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5. Design Electrical Ratin;INet 3!We):

918

6. Maximum Dependable Capacity (Gress 31We):

o"

7. Maximum. Dependable Capacity (Net MWe):

U3

, 8. If Changes Occurin Capacity Ratings titems Number 3 Through 7) Since Last Report. Gise Reason vn

'- 9. Power tvvel To Which Restricted. If Any (Net 5tWe):

N/A

- 10.'Reasoris For Restrictions.If Any:

M/A

. 'Uds Month Yr. to.Date Cumularise

!!. Hours !n Reporting Period 720

12. Number Of Hours Reactor Was Critical 720 4,367 45,624 1,514 27,587.5

..13. Reactor Reserve Shutdown Hours 0

0 3,975.1

. - : 14. Hours Generator On.Line -

720 1,429.6 26,370.9

15. Unit Reierre Shutdown Hours

' 16.' Gros's Tlierraal Ener;y Generated t MWH) 0 0

1,210.2 1,902,171 3,587,222 66,620,920 i '.! ross Electrical Ener;y CencrateJ (StWH) 638.818 1,196,626 22,461,102 l7 C

18. Net Ekc:rical Energy Generated (MWH) 607,076 -

1,12.4,94.5 41,4a+,oa4

19. Unit Senice Factor 100 32.7 58.5

20. Uriit Asa'ilability Factor 100 32,7 bl.z

21. Unit Cap.ieity Fae:or IU>in; MCC Net 96.6_

29.5' 53.3

22. Unit Capacity Faetor(Unia; DER Net) 91.8

23. Unit Forecd Out.:;e Rate 28.0 50.7 n

5.3*

31.0*

24. Shutdown > Seaeduled Oser Next 6 3fonths (T)pe'. Date.and Duratioh of Eachi:

vh

25. If Shut Down At End Of Report Period; Estimated Date of Startup:

26. Unit > ln Tot Status IPsior to Co'mmercial Optrationi:

Foreca>r Achiesed INITI\\L CRITICALITY

_N/A'

_ N/A INITIAI. El FCT RICIT Y n

n CO\\lMLRCIAL CPTR \\ fl0N

• These figures reflect a correction made to May 1980. report.

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AVERAGE DAILY UNIfi OhER LEVEL N

DOCKET NO.

50-312 Rancho Seco #1 g.!T DATE-80-06-30 m

CO*.!PL[TED !!Y R. 11. Color.ho TELEPIIONE 916-452-3211 June 1900

.\\lONTH

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DAY AVERAGE DA!LY P0hER LEVEL DAY AVERAGE DAILY POWER LEVEL (51We Net) 15fWe-Neil 850 17 R67 2

m60 18

- 871

869 3

863

-- 19

• 862 20 A77 4

861 23 873 5

872 871 c

22 871 7

23 8'70 873 24 8

9 R7n

,25 869 8f3 10 863 26 869 jg 872 27 868 12 872 28 871 13 29 870 14 A71 30 869

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'868 35 3g 869 16 INSTRUCTIONS m

On this tiirinat, list the aver.:e daily imii power'!ae!in SIWE Net for each day in the reporting month. Compute to the nearest whole nwpwatt.

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4 PITUELI:'c I:TOTJSTIC:: ?.EQUEST Rancho Seco Unit 1 1._

trace of racility:

2.

Scheduled date for next refueling shutdown:

July 1981

'3.

Scheduled date for restart fo11cwing refueling:

Septenber 1981 4: Technical Specification change or other license amend: cat required:

a) Change to Rod Indcr. vs. Power Level Curve (TS 3.5.2)

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b) Change to Core 'Id. balance vs. Pouer Level Curve (TS 3.5.2)

. c). Tilt Limits (TS 3.5.2) d) Safety Equipment Testing (TS 3.3.3)

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5.

Scheduled date(s) for subnitting propcsed ' licensing ' action:

uav 1981

b. Impertent licensing consideratione. associated with refuel'ing:

None

7. ' Ku.nber of fuel assemblies:

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a) In the core:

177 b) In the Spent Fuel Pool:

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S..Present licensed spent' fuel capacity:

579 9.

Projected date of the last refueling that cad be discharged to the Spent Fuel Fool:

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Cycle 4 Power Distribution Comparison

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In the District's letter to Mr. Robert W. Reid, dated February 27,.1980, we committed to perform Power Distribution Comparisons through Cycle 4 as a result of this being our first reload core utilizing Lumped Burnable Poisons.

Power Distribution Analyses at the beginning of Cycle 4, and 25' EFPD, were accomplished per the techniques and criteria speci-ficd in the Power Escalation Test program.

The results of that program are included elsewhere in this month's report. Addi,r.ionally,

B&W has performed an RMS analysis on the 25 EFPD power distribution.

The value determined was 0.024,8 which compares f avorably with the requirement that it be less than 0.0731.

As' of the end of June, apprcximately 43 EFPD had been accumulated, hence the data for RMS comparison specified for 50 EFPD has not been done.

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SECTION I - OVERVIEW-Following the third refueling of Rancho Seco Unit #1, the startup test program for Cycle 4 was begun with initial criticality established at

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0510 hours0.0059 days <br />0.142 hours <br />8.43254e-4 weeks <br />1.94055e-4 months <br /> on May 9,1980.. Zero power physics testing commenced at that time and was successfully completed on May -10,1980 at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />. As planned, the Zero power testing program was conducted at the iso-thermal Reactor Coolant temperature of 531*F, and below the power level ccmmensurate with nuclear heat. Power escalation was begun on May 10, 1980 and testing was done at th ree major power plateaus of 40%, 75% and 96% of full power. This final plateau being attained on May 18, 1980.

As of June 30, 1980 the plant has not attained 100% of' full power due to a sel f-imposed restriction to insure inadvertant Power / Flow trips would not occur. See sections on Reactor Coolant Flow and Flow Coastdown testing.

Tests intended at full power were completed at 96% full power on June 16, 1980.

The following descriptions of test data and resul ts refer to the Cycle 4 Reload Report, BAW-1560, August 1979 testing commitments and the District's February 19, 1980 and February 27, 1980 responses to the Comraission's

. February 11,'1980 request for additional information and commi ttent.

Re fe rence is made to that information rather than repeating it here.

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SECTION 11 - PRE-CRITICAL TEST

.I Control Rod Trip Test Control rod trip time testing was donc prior to estabiishing initial criticality and whil'e 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 dreppable ::ntr:! red:,,th!:h are assigned to Oroups ! through 7, were fully withdrawn.

Group 8 (Axial Power Shaping Rods which do not drop) were established at an intermediate posi tion.

Using the manual Reactor trip button to initiate the drop, all 61 droppable control rods were dropped into the core f rom the fully withdrawn position.

Drop time was determined by using the plant computer and measuring the time from

" trip" to three-fourths insertion.

The fastest rod dropped in 1.176 seconds, and the slcwest red was at 1.238 seconds.

For acceptance, the drop time of Groups 1 through 7 had to be less than 1.66 seconds. The measurement technique includes the control circuit and logic times in addition to tL rod travel time. All drop times were well below the acceptance criteria thus meeting the Technical Speci fications require-h' ments. for-full-flow drop time. Confirmation was made that the APSR's (Group. 8) did not drop.

.2 Reactor Coolant Flow The steady state four pump flow was determined for the hot zero power condition as being 404,820fGPM. This can be compared to the Technical

. Specification minimum acceptable value of 387,600 GPM.

The maximum

- flow is estabdIshed based on core lif t criteria. 'This upper limit for (2) e m_a e

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'Peactor Contant F?cw (Continued)

Cycle 4 Is 419,600 GPM.

These measurements met the Cycle 4 perfornance requi remen ts. This test was perforced at BOC-4 to ve'rl fy performance

following the installation of $2 Lumped Burnable Poison Assemblies in fuel which was unrodded during Cycle 3 testing. Correcting 80C-3 data to BOC-4 conditions shows an apparent reduction of 6356 GPM, or about 1.6%.

Analysis had expected the effect on core bypass flow to be a decrease from 10.4% to 8.3% of total flow. These measurements show the effect has been properly anticipated and the results to be acceptable.

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 flow through the reactor core determined as a function of time. The acceptance criteria was applied to the before trip conservative error-reduced value.

It was determined that the. actual coastdown transient flow exceeded the minimum acceptable flow for the period of interest by a marg.in in excess of 4000 GPM. This test met the requirements for operation of Cycle 4.

A ser'nd ' feature of this test was to veri fy that the time delay assumed in the flow coastdown safety analysis was not exceeded. This time' delay Is due to the use of hydraulle snubbers in the sense lines to the flow signal aP transmitters. A conservative one-second delay had been assumed.

The test compared the rate of flow coastdown between one channel, free of snubbers,'and the three remaining Reactor Protection Flow channels, whose (3) a m

Reactor Coolant Flow Coastdown (Continued)'

3 snubbers had been set to provide a slightly less damped signal than previously. The comparison showed the delay to range between 0.65 and 0.73 seconds, well wlthin the assumed interval.

As a result.of this snubber position, the noise seen on the flow signals has caused Internitten.t drops in the conservatively set Power / Flow RPS Trip signal to the point that the trip signal could be generated at as low as 100.8% full power, down from its Technical Speci fication upper limit of 105.0% full power.

For this reasco, power has been limited,to a nominal

.96%. fu11 power while analysis in support of a licensing action is under-taken. Measured flow is approximately 109 5% of design flow, hence such an analysis. Is in order.

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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 4 Reload Report.

With control rod Group 8 at 37.5% withdrawn, the resul ts were as "ollows:

Measured Vendor Prediction 1361.85_ ppm 8 1368fjooppms The measured data is consistant with the prediction and meets all acceptance cri teria.

2 Bordn Concentration at Maxinu, Cor trolli,a Rod Group Insertion Li,i t Measured Vendor Prediction 1012 ppmB 1004 fj00 ppmB This measurement provides a second just critical Boron concentration

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neasurement corresponding to a predicted value. At the tine of this

' measurement, control rod Groups 5, 6 and 7 were fully inserted and control rod Goup 8 ' positioned at 37.5% withdrawn. The. measured data was con-sistant with predict!' ns an'd met all acceptance criteria.

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Temperature Coefficient of Reactivity at All Rods Out Boron Measured Vendor Prediction

-0.231x10"" ak/k/F'

^ -0,29x10~"+0.4x10~"

at:1359 ppmB-at 1359 ppmB' The value at' this boron concentration met the acceptance criteria of being within.the predicted band.

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.4 Moderator Coefficient of Reactivity at All Rods Out Boron The result at 1359 ppmB also' met the acceptance cri teria for Moderator Coefficient of Reactivity which specifies that, when corrected for fuel doppler ef fects, the value shall not be nore positive than

+0.5x10-4 SkisF*.

The Moderator Ccef ficient of Reactivity was deter-mined to be -0.03x10-4 Ak/k/F'.

.5 Temperature Coef ficient of Reactivi ty Dete rmined at the Maximum Insertion Boron Concentration Measured Vendor Prediction

-0 918x10-4 Ak/k/F*

-0.95x10-4 0.4x10~4ak/k/F*

at 1012 ppmB at 1012 ppm 3 The acceptance criteria for this value is the same as for the ARO temperature coef ficient neasurement. This measurement met all cri teria.

.6 CRA Grouc Reactivity Worth Vendor Measured Worth Predicted Deviation Deviation tak/k Worth, tak/,k Measured Allowed Group 5 0.866 0.98

-13.16 115%

Group 6 0.825 0.87

-5.45 115%

Group 7 1.389 1,46

-5.11 115%

Total 3.080 3 31

-7.47 As the measured total. group worth was wi thin +10% of the predicted value, further actions commi tted to in the District's February 19, 1980 letter were not required. The shutdown margin calculations shown in the Cycle 4 Reload Report are substantiated by the above measurements and the excellent agreement between predicted and measured ARO Boron.

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-E}ected Rod Worth Measurement-4 Error Adjusted Measured Predicted Ejected Rod Ejected

Worth, Tolerance Worth, tak/k Worth, tak/k

%ak/k Allowed 0.L.

0.7805 0.76

+20%

The ejected rod worth is determined for the configuration corresponding

o the maximum insertion condition allowed by Technical Specifications,

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namely, Groups 5, 6 and 7 fully inserted at zero power, with Group S at 37.5% WD and all safety rods fully withdrawn.

From this configu-ration, the maximum worth " Ejected Rod," which is a rod in Group 7, 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. These two values were then averaged, and are reported as the Measured value.

These resul ts are consistent with the prediction and meet the absolute acceptance criteria of _ Technical Specifications by being less than 1.0 %ak/k at zero power. Furthermore, the worth of the three Group 7 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 worths-ranged from a high of 0.794%Ak/k to the minimum measured at 0.758%Ak/k. These results are certainly within the margin of tolerance for the measurement technique and provide an early confirmation of power distribu'tlen symmetry and lack of power tilt for Cycle 4.

Sabsequent observations during power escalation confirm the tilt free nature of this Co re.

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SECTION IV - POWER ESCALATION 1-Core Power Distribution Core' power distributions were taken and analyzed at the nominal Reactor. power test plateaus of 40%,- 75%, and 96tFP during Cycle 4 power escalation. The purpose of these measurements was to veri fy that the ' minimum DN8R, maximum linear heat rate, quadrant power til t, power imbalance, and related power peaking factors would not exceed allowable limi ts.

In each case the measured variables were extrapo-

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lated to.the over power trip setpoint for the next test plateuu so as' to assess the margin of conservatism prior to escalation. A summary of the test results follows:

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POWER DISTRiililTl011 TEST RESULTS Measured / Desired Date of Data 5/13/P0 5/16/80 5/19/80 6/10/80

Powe r -leve l, %FP 41.0/40 74.7/75 95.5/100 95.5/100 Core Burn'up, EFPD 1.0/2.0 1.92/3.0 5.0/4.0 23.6/25.0

-Group l-5, %WD 100/100 100/100 100/100 100/100 Group 6, %WD 100/100 100/100 i00/100 100/100

'Gr' up 7, %WD 87.1/87.0 87.0/87.0 98/87.0 98.9/87.0 o

Group 8, %WD.

25.9/25 0 23.0/22.0 23.0/19.0 22.0/19.0 Boron concentration, ppmB 1021/1000 933/880 866/862 789/811 Axial' Imbalance, %FP

-2.13/-0.29

-1.76/-0.07

-0.67/-2.95

-3.67/-3.07 Max incore Quadrant Power Tilt, %FP 0.61/ <3.64 0.58/<3.64 0.57/ <3.64 0.48/<3.64 Minimum DNBR.

8.54/>l.30 4.26/>l.30 3.30/>l.30 3.30/>l.30 Worse Case LHR, Kw/ f t

4. 85/ <20.4 8.73/ <20.4 11.09/<20.4 11.09/<20.4 Max Radial Power Peak 1.281/l.306 1.276/1.291 1.264/1.285 1.285/1.289 Max Total Powe r Peak 1.515/1.529 1.510/1.525 1.456/1.528

,1.493/1.527 Max Peak at Core Grid L-13/H-li L-13/il-11 L-13/H-11 L-13/H-13 Max Peak in Fuel Batch Number 6/6 6/6 6/6 6/6 Equilibri um Xenon Yes, 2D Yes, 2D Yes, 3D (es, 3D Acceptable for Power Escalation Yes Yes Yes Yes Extrapolations done to, %FP 91.5 112.0 112.0 112.0

Power Distribution Tes.t Results (Continued)

Acceptance criteria which applies to the radial and total peaking factors is +5% and +7.5% respectively when compared to the pre-dictions for the peak assembly at the 75% and 100% power plateaus. All acceptance criteria was met,.and escalation based upon these results proved to be conservative. The measured DNBR and linear heat rates veri fied that the Reactor Protective system setpoints provide protection for the core against exceeding transient DNBR and/or maximum linear heat rates assumed in the Safety Analysis and are sufficient to protect against exceeding the limiting Technical Specification LOCA heat rates.

.2 Power Imbalance Detector Correlation Tes t This test is. performed to establish the relationship between the out-of-core nuc! car ir.strumentatica and the full set of incere self-powered neutron detectors.

Both systems provide axial power imbalance data, with the incore system b' lng the standard.

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. Due to the effect of refuei?ng on the neutron flux exiting the reactor, the out-of-core indication of imbalance is expected to change.

Since the nature and magnitude of this change is not easily predicted, this test is performed at a low power level to establish that the relation-

' ship between the two systems is conservative. Should i t be desired to l

al ter the out-of-core /inco're relationship, regaining the out-of-core Ni

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di f ference ampli fier is :requi red.

During this power escalatic, the. ini tial resul ts showed the out-of-core Nuclear instrumentation. to : e very conservative.,, Anytime regaining is (10) ct.

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.2 Power imbalance Detector Correlation Test'(Continued) done, a retest is required. This regaining and. retest was accomplished at 90%FP' and all applicable acceptance criteria met.

The results corresponding to the maximum and mininum imbalance conditions are shown here:

s 40%FP s90%FP (Retest)

Ni Channel Di f fe rence Target Measured Di ffe rence Target Measured Anpli fie r Correlation Cor rela tion

/mplifier Correlation Correlation Gain Slope Slope Gain Slope Slope HI-5 4.13

>1.15 1J5 3.54

>l.15 1.266 til -6 4.13 3J.15 1.52 3.54 2).15 1.217 l

MI-7 4.13

>l.15 1.55 3.5h 2).15 1.260 i

NI-8 4.13

>1.15 1.54 3 5'4

?_1.15 1.232 Accep-

)1,15

>J.15 Acceptable Cri-Met tance teria Cycle 4 safety analysis assumes that the correlation s1. ope is greater than or equal to 1.150. As the above data shows, this correla tion -

criteria is satisfied on all protective' channels, and the relationship between the incore and out-of-core instrumentation is shown to be conse rva ti ve.

At the same time that this-data was obtained, the relationship between _the full set of incore instrumentation and those on the backup ; ecorders was also de,termined to heet lts acceptance criteria.

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.2 Power'I mbalance Detector 'Correlat ion Test (Co 't I nued)

A gain factor of 4.13 set into the Nuclear. Instru mntation dif ferential ampli fier ci rcul e' - " r this cycle was determined to be exessively con-se rvat i ve.

Thus the gains were subsequently reset to 3.54 and an opera-tional transient used to induce the imbalance changes necessary to demonstrate acceptable correlation. The resul ts are shown above, thus Cycle 4 will operate,with these values.

3 Power Doppler Coefficient of Reactivity From equilibrium conditions at near 96%FP, the power doppler coefficient was determined. The value obtained was -1.68x10-4 Ak/k/%FP. The 1

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

4 Moderator Temperature Coefficient of Reactivity at Power T'he "at power" moderator temperature coefficient was measured as described in the Cycle 4 Reload Report, while operating the Reactor at equilibrium conditions' and near 96%FP. Measurements determined the coefficient to be

-0 99x10-4 ak/k/F compared to a vendor predicted value.of -1.44x10'4 Ak/k/F*. The acceptance criteria for this parameter is that it shall not be " positive" for Reac, tor operations above 95%FP. This condition for operation is satisfied -forLCycle 4.

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SECTION V - SUWARY

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The District's letter of. February 27, 1980 commi tted to a follow-on program of ' core power distribution review, analysis, and reporting due to the unique fuel management scheme.(LBP in Reload Fuel) being utilized. This program involves 37.! analy:Is ccch SC EFPD to' determine their cbli' ty t:

predict LSP behavior in a relcad. Requisite reports ull t be included in the L

monthly plant performance report to the NRC.

- Since this startup program was completed at 96%FP, the increment to 100%FP and the associated 100%FP Power Distribution Analysis will be reported in a monthly report.

The ' final test.in this program was the Reactivity Coefficients at. Power measurements reported above. Those tests were completed on June 16, 1980, hence this report is. due submission within 45 days, or by July 31, 1980.

a The. resul ts of early Cycle 4 testing provided in this report

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demonstrate that Rancho-Seco Unit 1, Cycle 4, has been properly designed; and that the unit can be operated in'a manner' t' hat will not endanger the health

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and' safety off the public.

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