ML20009C067

Monthly Operating Rept for June 1981

ML20009C067
Person / Time
Site:	Rancho Seco
Issue date:	06/30/1981
From:	Colombo R SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20009C061	List: ML20009C060 ML20009C067
References
NUDOCS 8107200233
Download: ML20009C067 (14)

Text

. -.

~

t

SUMMARY

OF PLANT OPERATIONS 6-1 to 4

6-9 Reactor in cold shutdown (OTSG tube leak).

.6-10 Reactor in heat-up mode 6-11 Began deboration to criticality.

6-12 0655 Reactor critical.

0849 Began increasing power to 5%.

1235 Holding at 5% power.

1350~ Increasing power to 15%.

1958 Closed OCB's and began increasing power.

2015 Holding at 40% power.

6-13 1130 increasing power to 67%.

c 1525. Reactor at 67%, holding.

6-15 1337. Began increasing power to 70%.

-1610 Increasing power at 3% per hour.

1.

1900 Reactor at 75%, holding.

6-16 0015 Began increasing power at 3% per hour.

0845 Reducing power to 85% due to instability.

{

1015 increasing power to 87%.

i 1128 Reactor at 87%.

6-17 1509 Reactor / Turbine trip (Lost Main Feedwater Pump) 1948 Reactor in hot shut down.

6 6-18 0140 Began RCS cooldown for replacement of CRD stator.

2210 RCS in cold shut down.

6-19 0059 Began RCS heatup.

0511 Terminated RCS heatup due to CRD cooling water leak.

8107200233 810706 PDR ADOCK 05000312 R

PDR

SUMMARY

0" PLAhT OPERATIONS (Continued) 6-20 2020 Began pullino rods.

2159 Reactor Cri tical 2315 Reactor at 10% power.

6-21 0200 Closed OCB's.

0210 Reactor trip (lost main feedwater pump).

0354 Began pulling rods.

0416 Reactor critical.

0716 Closed OCB's.

Began increasing power to 40%.

1600 Reactor at 44%, increasing power to 50%.

6-22 Reactor at 50% power.

6-23 1147 Reactor trip (lost main feedwater pump) 1326 Began pulling rods.

1358 Reactor critical.

1656 Closed OCB's and began increasing power.

6-24 Reactor at 50% power.

6-25 Began inc.reasing power to 70% @ 3% per hour.

[

6-26 Reactor at 70% power.

6-27 to 6-30 Reactor at approximately 75% power.

PERSONNEL CHANGES-REQUIRING REPORTING No personnel changes that require reporting in accordance with Technical Specifications Figure 6.9.2 were made in June, 1981.

SUMMARY

OF CHANGES IN ACCORDANCE WITH 10 CFA 50.59(b) 1.

Added head rigging with clev ses to the reactor vessel head. The modification has been permanently attachet to the reactor vessel head to eliminate the critical path cutage time presently required for rigging, to lif t the vessel head for refueling.

2.

Removed ine relief valves on the makeup filters (PSV-23007 and PSV-23008) and on the seal return coolers (PSV-24017 and PSV-24018). This was a resasonable step to prevent leakage from systems that contain liquids or gases having large radioactive inventories af ter a serious transient or accident, as required in NUREG-0578, Section 2.1.6a.

Review has indicated that these valves can be removed without endangering the makeup fiiters, seal return coolers, or associated piping.

3 Removed leaky "B" feedwater cleanup block valve, FWS-022, and sealed the pipe with a cap.

The removed valve was not required by any operating or emergency procedures.

4.

Completed modi fications to allow installation of Movable Incore Detector System (MIDS) for monitoring performance of Axial Blanket Fuel Assemblies.

MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1.

Repai red loose valve linkage on reactor building personnel hatch outer equalizing valve.

2.

Rewound motor of reactor building emergency sump vcive.

SECTION 1 - OVERVIEW Following the fourth refueling of Rancho Seco Unit #1, the startup test program for Cycle 5 was begun with initial criticality established at 0715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br /> on May 4, 1981.

Zero power physics testing commenced at that time and was success-fully completed on May 9, 1981 at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br />. As planned, the Zero power testing program was conducted at the iso-thermal R..'ctor Coolant t empe ra t u re of 532 F, and below the power level commensurate witi, nuclear heat.

Power escalation was begun on May 9,1981 and testing has been cc.npleted at 40%

and 75% of full power. This final plateau being attained :n Ma/ 13, 1981.

As of June 30, 1981 the plant has not attained 10 of tusi power due to a return to cold shut down to repair the OTSG tube leak.

The following dTscriptions of test data and results refer to the Cycle 5 Reload Report, U AW-1667, March 1981 testing commi tments and the Di st rict 's May 7, 1981 response to the Commission's April 10, 1981 request for additional information. Reference is made to that information rather than repeating it here.

(1)

?

SECTION 11 - PRE-CRITICAL TEST Control Rod Trip Test

-Control rod trip time testing was done prior to establishing initial criti-cality 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 safety and regulating control rods, which are assigned _to Groups I through 7,-were fully withdrawn.

Group 8 (Axial Power Shaping Rods which do not drop)'were established at an intermediate position. Using the manuti Reactor trip button to initiate the drop, all' 61 droppable control rods were drapped into the

' core f rom the fully withdrawn position.

Drop time was determined by using the plant computer and measuring the time f rom " trip" to three-fourths insertion. The fastest rod dropped in 1.163 seconds, and the slowest rod was at 1.202 seconds.

For acceptance, the drop time of Groups 1 through 7

'had to be less than 1.66 seconds. The measurement technique.ncludes the control circuit and logic times in addition to_the rod travel time. All drop times were well below the acceptance criteria thus meeting the Technical Specifications requirements for full-flow drop time.

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

P (2)

~. __-

1 1

SECTION 111 - ZERO POWER PHYSICS TESTING

.1 All Rods Oat Boron Concentration The All Rods Out (AR0) Boron concentration was measured as described in the Cycle 5 Reload Report.

4 With control rod Group 8 at 37.5% withdrawn, the results were as follows:

Measured Vendor Prediction i

1201 ppmB' 1252 + 100 ppmB The measured data is consistant with'the prediction and meets all acceptance criteria.

.2 Boron Concentration at Maximum Controlling Rod Group Insertion Limit Measured Vendor Predict ion j

850 ppmB 868 + 100 ppmB This measurement provides e second just critical Boron concentration measure-ment-corresponding to a predicted value. At the time of this measurement,

. control rod Groups 5, 6, and 7 were fully inserted and control rod Group 8 positioned at 37.5% withdrawn. The measured data was consistant with predictions and met all acceptance criteria.

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

-0.40 x 10~

Ak/k/F

-0.42 x 10~

+ 0.4 x 10' at 1217 ppmB at 1217 ppmB The value at this boron concentration met the acceptance criteria of being within the predicted band.

.4 Moderator Coef ficient of Reactisity at All Pods Out Bo ror, The result at 1217 ppmB also met the acceptance criteria for Moderator Coefficient of Reactivity which specifies that, when corrected for fuel doppler effects, the value shall nct be more positive than +0.5 x 10-4 Ak/kF. The Moderator Coefficient of Reactivity was determined to be.-0.20 x 10-4 Ak/k/F.

i l

5 Temperature Coefficient of Reactivity Determined at the Maximum insertion Boron Concentration Measured Vendor Prediction

-1.38 x 10~

Ak/k/F

-1.132 x 10'

+ 0.4 x 10~

.at 852 ppmB at 852 ppmB

~

Ak/k/F 4

The acceptance criteria for this value is the same as for the ARO temperature coefficient measurement. This measurement met all criteria.

(3)

.6 - CRA Group Reactivity Worth Vendor Measured Worth Predicted Deviation Deviation

%Ak/k Worth, %Ak/,k Measu red Allowed Group 5 1.117.

1.096

-1.9%

f; 15%

Group 6-0.989 0.964

-2.51%

j; 15%

Group 7 1.389 1.422

-2.36 f;15%

Total 3.495 3.482

-0.37 f; IG%

The shutdown margin calculations shown in the Cycle 5 'eload Report are substantiated by the above measurements and the excellent agreement between predicted and measured ARO Boron.

7 Ejected Rod Worth Measurement Measured-Predicted Deviation Ejected

Worth, Measured Tolerance Worth, %Ak/k

%Ak/k Allowed 0.58 0.57

-1.72

+ 20%

The ejected rod worth is determined for the configuration corresponding to the maximum = insertion condition allowed by Technical Specifications, namely, Groups 5, 6, and 7 fully inserted at zero power, with Group 8 at 37.5% WD and all safety rods fully withdrawn.

From thi s. configuration, 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 results are consistent with the prediction and meet the absolute acceptance criteria of Technical Specifi-cations by being less than 1.0 %Ak/k'at zero power.

(4)

SECTION IV - POWER ESCALATION

.1 Core Power Distribution Core power distributions have been taken and analyzed at the nominal Reactor power test plateaus of 40% and 75% during-Cycle 5 power escalation. The purpose of these measurements was to verify that the minimum DNBR, maximum linear hear rate, quadrant power tilt, power imbalance, and'related power peaking factors would not exceed alloweble limits.

In each case the measured variables v tre extrapolated to the over-power trip setpoint for the next-test. plateau so as to assess the margin of conservatism prior to escalation.

A summary of the test results follows:

POWER DISTRIBUTION TEST RESULTS Measured / Desired ~

Date of Data 5/10/81 5/14/81 Power level, %FP 41.0/40 75.0/75 Core Burnup, EFPD 0.35/2.0 2.70/3.0 Group'1-5, %WD 100/100 100/100 Group 6, %WD 100/100 100/100 Group 7, %WD~

89.8/87.0 91.0/87.0 Group 8, %WD 34.9/35.3 27.0/28.2 Boron Concentration, ppmB 888/860 715/760 Axial Imbalance, %FP

-0.68/1.01 1.34/0.47 Max Core Quadrant Power Tilt, %FP 1.45/<3.40 1.35/<3.40 9.87/ 1.30 2.80/31.30 Minimum DNBR 3

Worst Case.LHR, Kw/ft 4.46/<20.4 9.03/<20.4 Max Radial Power Peak 1.264/1.282 1.279/1.267-Max Total Power Peak 1.465/1.520 1.520/1.472 Max Peak at Core Grid K-10/K-12 K-12/K-12 Max Peak in Fuel Batch Number 7/7 7/7 Equilibrium Xenon Yes/2D Yes/2D Acceptable for Power Escalation Yes Yes Extrapolations done to, %FP 92.1 112.0 (5)

Acceptance criteria which applies to the radial and total peaking fcctors is + 5% when compared to the predictions for the peak assembly at the 75%

power plateau. All acceptance criteria were met, and escalation based upon these results proved to be conservative. The measured DNBR and linear haat rates verified that the Reactor Protective system setpoints provide protection for the core against exceeding t,ransient DNBR and/or maximum linear heat rr es assumed in the Safety Analysis and are sufficient to protect agains, exceeding t'le limiting Technical Specification LOCA heat rates.

.2 Core Symmetry Test The core symmetry test was used during this cycle as-a method of verifying the symmetry of the core.

Previously the symmetric ejected rod method was used to provide this information, The core symmetry test uses the incore instrumentation to evaluate the quadrant tilt from 15% to 40% power. The acceptance criteria is that the tilt be less than 3.4%.

The maximum measured tilt in this. power interval was -1.73, well below the acceptance criteria maximum.

3 Power Imbalance Detector Correlation Test This test is performed to establish the relationship between the out-of-core nuclear instrumentation and the full set of incore self powered neutron detectors.

Both systems provide axial power imbalance data, with the incore system being the standard.

Due to the effect of afueling 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 relationship between the two systems is conservative.

Should it be desired to alter the out-of-core /incore relationship, regaining the out-of-core Nl dif ference ampli fier is requi red.

I During this power escalation, the initial results showed the out-of-core Nuclear instrumentation differencing amplifier had to be regained. Anytime regaining is done, a retest is required. This regaining and retest was accomplished at 40%FP and all applicable acceptance criteria met.

Cycle 5 safety analysis assumes that the correlation slope is greater than or equal to 1.15

'niis correlation criteria was satisfied on all protective channels, and the relationship between the incore and out-of-core instrumen-tation is shown to be conservative. At the same time that this data was obtained, the relationship between the full set of incore instrumentation and those on the backup recorders was also determined to meet its acceptance i

c ri te ria.

(6) l

SECll0N V -

SUMMARY

The startup test program was initiated on May 2, 1981. All tests at zero power were completed on May 9, ?981 and the power escalation to 100% was initiated. Testing at 40% and 75% was completed on May 15, 1981.

Escalation to 100% was delayed due to the OTSG tube repair.

The results of early Cycle 5 testing provided in this report demonstrates that Rancho Seco Unit 1, Cycle 5, has been properly designed; and that the unit can be operated in a manner that will not endager the health and safety of the public.

(7)

OPERATING DATA REPORT DOCKET NO.

50-312 DATE 81-06-30 COMPLETED BY R. W.

Colombo TELEPHONE (916) 452-3211 OPERATING STATUS Notes

1. Unit Name:

Rancho Seco Unit 1

2. Reporting Period:

June. 1981

3. Licensed Thermal Power 13thts:

2772

4. Nameplate Ratin; (Gro3s.\\1Wei:

963

5. Design Electrical Ratina INet 31Wei:

918

6. Sfaximum Dependable Capacity # Gross 31Wei:

917

7. Sfaximum Dependable Capacity (Net 31Wei:

873

8. If Changes Occur in Capacity Ratin;s (Items Numuer.5 Ihrough 7) Since Last Report. Gise Reasons:

N/A

9. Power Lese! To which Restricted. If Any (Net 31We):

N/A

10. Reasons Fer Restrictions,if Any:

N/A This Sionth Yr..to.Date Cumulative II. Hours in Reportin; Period 720 4.141 54.384

12. Number Of Hours Reactor Was Critical 365.9 1.320.1 32.816.3

13. Reactor Resene Shutdown Hours 0

0 4.469.6

14. Hours Generator On-Line 34?.8 1.232.2 _

31.477.8

15. Unit Resene Shutdown Hours 0

0 1.210.2

16. Gross Thermal Enerev Generated (31WH) 657.516 2.928.217

.79.900.472

17. Gross Electrical En[r;y Generated O!WH, 176.808 931.166 26,878.880

18. Net Electrica! Enerzv Generated 01hHi 155.552 865.235 25.391.360

19. Unit Servie: Factor 47.6 28.4 57.9

20. Unit Availability F::ctor 47.6 28.4 60.1

{

21. Unit Capacity Factor iL3m; 3tDC Net) 24.7 22.8 53.5

22. Unit Capacity Factor tC3ing DER Net) 21.5 21.7 50.9

23. Unit Forced Outage Rate 52 4 17.4 29.4 l

24. Shutdowns Sehedated Oser Next o.itonths T pc. Date,and Duration of Each>:

3 N/A

25. If Shut Down At End Of Report Perind. E3:imated Date of Startup; N/A i

26. Units in Test Status iPrior :o Commere:al Operation n Forecast Achiesed INITI AL CRITIC ALITY N/A N/A INITIAL ELECTRICITY N/A N/A CON 111ERCI AL OPLR \\ TION N/A N/A P3/77)

AVERAGE D AILY UNIT POWER LEVEL DOCKET NO.

50-312 UNIT Rancho Seco Unit 1 DATE 81-06-30 COMPLETED BY R. W. Colombo TELEPHONE (916) 452-321.1, MONTH June, 1981 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe NetI t MWe-Net) 1 0

37 354 2

o is n

3 0

19 0

4 n

20 0

5 0

21 144 6

0 22 306 7

0 23 245 g

0 24 1?R 9

0 25 426 10 0

26 538 11 0

27 565 12 31 28 980 13 395 29 988 14 510 30 606 15 910 3

16 621 l

I l

INSTRUCTIONS Or. this format. list the serage daily unit power lesel in MWe-Net for each day in the reportmg month. Compute to the nearest whole megawait.

l (9 /77 )

t

-~._

UNIT SifUTDOWNS AND POWER REDUCTIONS DOCKET NO.

50-317 UNIT NAME Rancho Seco Unit 1 DATE R1-nA-3n REPORT MONTil June, 1981 COMPLETEI) llY R. W.

Colombo TELEPilONE (91 M 4 47-1711 c

i?

"g j.E y Licensee y t, Cause & Corrective No.

Date g

3g

,gy&

Event 3, ?

pU Action to F

$5 3gc Report #

m0 Prevcnt Recurrence o

f u

O 2

81-06-01 F

284 A

1 81-026 01 T CC HTEXCH Continued shutdown due to OTSG tube leak.

3 81-06-17 F

82.9 A

3 N/A CF PUMPXX RC Pressure Trip (Loss of Main Feedwater Pump t

81-06-21 F

5.1 A

3 N/A CF PUMPXX RC Pressure Trip (Loss of Main Feedwater Pump) 5 81-06-23 F

5.2 A

3 N/A CF PUMPXX RC Pressure Trip (Loss of Main Feedwater Pump)

I 2

3 4

F: Forceu Reason:

Method:

Exbibit G Instructions S: Scheduhd A Isluipment Failure (Explain) 1 Manual foi Picparation of Data B Maintenance of Test 2 Manual Scram.

Entry Sheets fur Licensee C-Refueling 3 Automatic Scram.

Event Report (LER) File (NUREG.

D Regulatory Restriction 4-Other (Explain) 0161)

E Operator Training & Licens;; Examination F Administrative

'S G-Operational Error (Explain)

Exhibit I Same Source (9/77)

Il Other (Explain)

REFUELING INFORMATION REQUEST 1.

Name of Facility:

Rancho Seco Jnit 1 2.

Scheduled date for next refueling shutdown:

April, 1982 3.

Scheduled date for restart following refueling:

October. 1982 r

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) 5.

Scheduled date(s) for submitting proposed licensing action:

February, 1982 6.

Important licensing considerations associated with refueling:

None 7.

Number of fuel assemblies:

a)

In the core:

177 b)

In the Spent fuel Pool:

196 8.

Present licensed spent fuel capacity:

579 9.

Projected date of the last refueling that can be discharged to the Spent Fuel Poo'l:

1987

.