ML19283A535

Annual Rept of Station Operation for 1964

ML19283A535
Person / Time
Site:	Dresden
Issue date:	01/26/1965
From:	COMMONWEALTH EDISON CO.
To:
References
NUDOCS 8008280580
Download: ML19283A535 (33)

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C01EONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION ANNUAL REPORT OF STATION OPERATION FOR THE YEAR 1964 January 26, 1965

DRESDEN NUCLEAR POWER STATION ANNUAL REPORT OF STATION OPERATION

References:

(1)

Letter to AEC dated February 26, 1964, requesting operation of Special Fuel Assembly SA-l.

(2)

Letter to AEC dated February 27, 1964, requesting authorization to receive SA-1.

(3)

Letter of February 28, 1964, to AEC requesting postponement of control drive and blade testir.g.

(4)

Telegram from AEC dated March 4, 1964, authorizing postponement of control drive and blade testing.

(5)

Letter to AEC dated March 10, 1964, requesting amendment to Byproduct Material License No. 12-5650-1.

(6)

Letter to AEC dated March 13, 1964, requesting amendment to Byproduct Material License 'No. 12-5650-1.

(7)

Letter to AEC dated March 13, 1964, requesting amendment to Special Nuclear Material License SNM-225.

(8)

Change No. 6 dated March 13, 1964, from AEC authorizing decrease in burnout ratio.

(9)

Letter from AEC dated March 13, 1964, re:

use of respiratory protective equipment.

(10)

Byproduct Material License No. 12-5650-3 dated, March 13, 1964.

(11) Amendment No. I dated March 20, 1964, to SNM-225.

(12)

Letter to AEC dated March 25, 1964, re:

heat flux limits for Type II fuel.

(13)

Letter to AEC dated March 26, 1964, re: modification of our February 26, 1964, request.

(14)

Change No. 5 dated March 27, 1964, from AEC - authorization to operate SA-1.

(15) Amendment Ho, 9 dated April 3, 1964, to Byproduct Material License 12-5550-1.

(16)

Change No. 7 dated April 9, 1964, from AEC - authorization to decrease frequency of control drive tests.

~

. (17)

Letter of April 24, 1964, to AEC requesting amendment to Special Nuclear Material Lecense No. SNM-638.

(18)

Letter of April 29, 1964, to AEC, re:

high radiation areas.

(19) Amendment No. 1, dated May 6,1964, to Special Nuclear Material License SNM-638.

(20)

Change No. 8, dated May 13, 1964, to License DPR-2, re:

change in heat flux limits for Type II fuel.

(21)

Letter of September 16, 1964, to AEC requesting general authorization to take credit for use of its respiratory protective equipment.

(22)

Letter of September 18, 1964, and attachments to AEC, re:

Proposed Revision of Technical Specification on Stack Limits and Monitoring.

(23)

Letter of October 1, 1964, authorizing increased flexibility in regard to high radiation area procedure and an increase in the number of " radiation" keys from one to two.

(24)

Letter of October 21, 1964, regarding modification of our September 18, 1964, request.

(25)

Letter of December 24, 1964, regarding operation of Dresden Reactor with 104 Type III-F fuel assemblies.

_3-I.

INTRODUCTION This third annual report is submitted in compliance with paragraph 3.c.(2) of License DPR-2, as amended, and covers operation of the licensed facility at Dresden Nuclear Power Station during the year of 1964.

II.

SUMMARY

OF OPERATIONS A.

Scope of Operations Operation of the plant continued from the preceeding year until March 4, 1964. At this time the plant was shut down for a weekend to conduct end-of-life physics tests.

The second partial reactor refueling outage was performed from April 12, to June 9.

During the last two weeks of this outage, the reactor containment vessel was successfully leak.: tested.

The plant continued operation to the end of the year with only five minor shutdowns as described below.

There were no major changes in facility design with the exception of a change in the off-gas instrumentation as described in Paragraph III-7.

B.

Shutdowns The plant was shut down seven times duri.g the year as indicated on Table I.

Four of these autages were forced outages: one resulting from a manual scram upon loss of primary feedwater pumps; one turbine trip due to high exhaust hood temperature upon start up; one to repair a leak in the 15th stage moisture removal line; and one to repair a leak in the 16th stage extraction line.

There were three planned outages: one to replace the off-gas filter; one outage for end-of-life physics tests; and one outage for refueling.

C.

L_,ad Restrictions c

Load restrictions imposed during the year are listed in Table 2.

The major portion of these restrictions were due to fuel depletion prior to refueling, and to minimize fuel failures during the last five months of the year.

III.

DISCUSSION A.

Operating Experience 1.

Generation The total reactor operating (critical) time during the year was 7,213.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, and the total power for the period was 138,688 MWD (theruml).

Gross electrical generation during the year was 1,037,511,250 KRH; net generation was 980,070,600 ENH.

As of December 31, 1964, the total gross

TABLE 1 REACIDR AND TURBINE CENERATOR SERVICE Reactor Turbine-Generator Operating Operating Period Period Operating On Off Hr.

Min.

Criticals On Off Hr.

Min.

Starts Conditions (3:56 AM) 10:26 PM 1534 26**

(2:48 FM) 10:08 PM 1534 08**

1 Power Operation. Shut-(11/12/63) 3/4/64 (11/12/63) 3/4/64 down for end-of-life nuclear testing.

11:23 FM 2:08 AM 25 43*

33 Temperature Coefficient 3/4/64 3/9/64 Tests.

549 39 5:53 AM 546 07 1

Power Operation.

2:21 AM 3/9/64 3/9/64 y

10:10 PM 286 05 9:43 FM 285 43 0

Refueling Outage.

4/12/64 4/12/64 10: 12 PM 10:15 PM 0

03 1

Zero Power; Steady 4/12/64 4/12/64 state Zenon critical.

0:55 AM 1:02 AM 0

07 1

Zero Power; head off; 5/13/64 5/13/64 critical.

10:23 FM 2:34 AM 24 12*

29 Temperature Coefficient 6/7/64 6/9/64 Tests.

2:34 AM 5:33 PM 182 59 6:45 AM 5:33 PM 178 48 1

Power Operation. Manual 6/9/64 6/16/64 6/9/64 6/16/64 scram, loss of primary feedwater pumps.

9:05 PM 10:07 PM 1

03 1

2:20 AM 2:23 AM 0 03 1

Start-up.

Turbine tripped 6/16/64 6/16/64 6/17/64 6/17/64 due to high exhaust hood temperature.

TABLE 1 (Continued)

REACTOR AND TURBINE GENERATOR SERVICE (Continued)

Reactor Turbine-Genera tor Operating Operating Period Period Operating On Off Hr.

Min.

Criticals On Off Hr.

Min.

Starts

_ Condition 10:12 PM 337 48 1

2:44 AM 333 16 1

Power Operation 6/16/64 6/17/64 0:36 AM 216 36 0:24 AM 216 24 Power Operation.

7/10/64 7/10/64 Shutdown to repair leak in 15th stage r.oisture removal line.

3:15 AM 476 45 1

11:53 AM 468 07 Power Operation.

7/12/64 7/12/64 1:27 AM 625 27 11:30 PM 623 30 Power Operation.

8/27/64 8/26/64 Shutdown to repair leaks in 16th stage extraction line and No. 3 gland steam drain ine.

10:21 PM 49 39 1

2:34 PM 33 26 1

Power Operation.

/29/64~

8/30/64 0:10 AM 264 10 11:31 PM 263 31 Power Operation.

9/12/64 9/11/64 Shutdown to replace off-gas filter.

09 AM 2638 51 1

2: 55 PM 2626 05 1

Power Operation.

/13/64 9/13/64 IAL 1964 7213 33 69 7109 08 7

Intermittent Operation From Midnight, January 1, 1964

. TABLE 2 LOAD RESTRICTIONS FOR 1964 Reduction From Maximum Date Capability of 210 MWe Condition January 8-16 10 Fuel depletion.

January 16-17 50 In-core calibration and fuel depletion.

January 17-30 10 Fuel depletion.

January 31 -

15 Fuel depletion.

February 2 February 3 20 Leaking fuel assembly detection test and fuel depletion.

February 3-29 30 (avg. for Fuel depletion.

month)

March 1-31 57 (avg. for Fuel depletion.

month)

April 1-12 77 (avg. for Fuel depletion.

month)

June 9-25 40 In-core calibration.

July 16 140 To manually close shut off valve on primary steam drum gauge glass.

July 24-31 4

High circulating water temperatura.

August 18 -

20 Minimize fuel failure.

September 7 September 8-24 50 Perform radiochemical tests and in-core calibration.

September 25 - 30 50 Minimize fuel failure.

~

October 1 -

65 Minimize fuel failure.

December 31

PLANT ELECTRICAL LOADING YEAR 1964 DRESDEN NUCLEAR POWER STATION KEY TO PIJMP SHUIDOWN (HotRS)

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. III.

DISCMSSION (Cont 'd)

A.

Operating Experience (Cont'd) 1.

Generation (Cont'd) generated since commencement of power operation (April 15, 1960) was 4,107,610,250 KWH.

On December 1, 1964, the gross electrical generation exceeded 4 billion KWH.

2.

Scrams a.

At 11:56 p.m. on March 4, the reactor scrammed from high flux on channels No. I and No. 4 during temperature coefficient reactivity tests conducted at rated temperature and maximum Xenon.

The reactor was critical.

There were 73 rods withdrawn.

The scram was caused by maloperation of the range selection switch.

b.

At 10:15 p.m. on March 7, a spurious trip signal occurred on channel No. 2.

The reactor was subcritical.

There were 25 rods withdrawn and rod F-3 was in the process of being withdrawn.

c.

At 10:28 p.m. on March 7, a spurious trip signal occurred on channel No. 2.

The reactor was subcritical with four rods withdrawn and rod H-8 was in the process of being withdrawn.

d.

At 5:33 p.m. on June 16, the reactor was manually scrammed.

The reactor was critical and operating at 518 MWt (159 MWe).

There wera 48 rods withdrawn, 10 rods at position 9, 5 rods at position 5, 2 rods at position 4, and 2 rods fully inserted.

The operators were in the process of returning "B" condensate pump to service after maintenance work on the suction bell diffuser cone was completed.

The pump suction valve had been opened three turns and the pump vent to the condenser was open. At 5:32 p.mt, "A" and "C" primary feedwater pumps tripped due to low suction pressure caused by excessive air-in leakage from "B" condensate pump.

The operators closed the suction valve to "B" condensate pump when they heard a loud noise from the primary feedwater pump area. The control room operator tried to restart "A" primary feedwater pump, and, at the same time, "B" primary feedwater pump which was on standby, attempted to start on automatic.

Neither pump held in.

The control room operator manually scrammed the reactor at 5:33 p.m., when primary feedwater flow could not be restored.

Scramming the reactor collapsed the steam voids and lowered the drum level, which, in turn, tripped the four reactor recirculating water pumps.

The control room operator opened the generator 138 KV main circuit breaker.

This operation placed the start-up protection relay in service which tripped.

Primary drum water level was restored by manually starting "A" primary feedwater pump.

When normal

. III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 4.

Control Rod Drives (Cont'd) b.

Inspection (Cont'd) functioning properly. Further attempts to withdraw the drive resulted in complete insertion of the drive to position "0".

Control rod drive C-3, Serial No. 1307, was selected for removal based on higher than average friction pressure and additionally because of operational difficulty in unlatching this drive when fully inserted.

During the outages, operational malfunctions were encoun-tered on the two drives in core position E-9 and C-4.

This occurred af ter removal and replacement of drives K-4 and C-3 which were selected for removal prior to plant shutdown.

Control rod drive Serial No. 1246, core position E-9, drifted out of the core on May 6, 1964, during reactor fuel loading operations.

The drive was exercised, stepped, and scrammed from position 0, but still drifted out of the core. The drive was scrammed from position 12 but drifted after exercising. Another full scram and further exercising on May 7, did not result in drifting, but it was decided to remove the drive for disassembly. On the same day, control rod drive, C-4, Serial No. 1309, could not be retracted.

System checks and increased operating pressure failed to move this drive and it was decided to remove this drive also. Table 3 gives a summary of the drives removed and inspected during the outage.

In summarizing the results of inspection of the removed four drives, the following items are noted:

1.

Despite different types of malfunctioning, all four driver were capable of being inserted.

ii. No specific trend was noted in the several failures noted in component parts iii. No cracks or defects were noted by the use of penetrant dye test.

The components tested are listed on Table 3.

5.

System Stability Tests a.

Recirculating Pump Trip On March 2, the unit was base loaded in preparation for preliminary end-of-life nuclear and stability testing.

On March 4, power testing was initiated by a load reduction in preparation for a recirculation pump trip and the valving off of its loop. One additional loop was slowly valved off and its pump tripped thereafter.

The dynamic system perfor-mance was thereby tested.

Heat balance and pressure drop IABLE 3 CONTROL ROD DRIVE INSPECTION Drive Removed Drive Repisced -

Cell No.

Serial Serial In Core Number Date Number Date Remarks C-3 1307 4/27/64 1264 4/27/64 No. 1 outer seal broken.

This was covered by foreign material between the outer seal and the inner tube surface.

K-4 1225 4/28/64 1279 4/28/64 Spirol pin fastening shart to anti-rotational roller failed.

This was apparently caused by a heavy blow during installation of this drive during major drive modifications in 1961.

E-9 1246 5/9/64 1284 5/9/64 Scratches were found on the idner surface of the shuttle piston and the piston ring area of the collet assembly.

The scratches were caused by foreign material between the two surfaces.

C-4 1309 5/11/64 1232' 5/11/64 No defects were found. A combi-nation of leaky seals and fric--

tion resistance in the collet assembly prevented rod withdrawal.

Components Inspected by Dye Penetrant Test Index Tube (5850288)

Piston Head Assembly (192C554)

Shuttle Piston (856B398)

Stop Piston (ll5A8600)

Collet Assembly (693C827)

Roller Mount Assembly (932C149) a.

Weld b.

Spud c.' Anti-rotational Roller d.

Guide Roller e.

Roller Housing Spring (lllA3298)

Spring Washer (145A5454)

Guide Plug (856B397)

Drive Housing Welds (5850289)

. III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 5.

System Stability Tests (Cont 'd) a.

Recirculating Pump Trip (Cont'd) data established the recirculation flow characteristics of the system under various pump and loop combinations.

The transient lLuits exhibited by station instrumentation and recorders subsequent to "A" recirculation pump trip are listed in Table 4 as to their magnitudes and likely coincidences.

fne gross electric power generation increased, dropped and partially recovered, settling out 4.5 percent below its initial value. The primary steam-flow dropped and held 11.0 percent below its initial value.

The secondary steam flow dropped, increased, and finally settled out 1.2 percent below its initial value.

The average in-core flux level dropped and partially recovered cettling out 22.0 percent below its initial level.

Reactor pressure and inlet water temperature dropped 2.0 percent.

Pressure drop across the reactor dropped 42 percent as a result of a 25 percent loss in recirculation flow. Approximately 80 per-cent loss in steam generation was experienced on steam generator "A", the unit whose pump was tripped. Steam generator "B" increased 15 percent; "C'.' and "D" experienced a 30 percent increase.

Reactor pressure was thereafter adjusted and the loop valved out of service.

The conditions existing prior to and subsequent to the valving out of the ".A" loop and the final step of valving out of the "D" loop are indicated in the following tabulation.

Initial Loop "A" Loop "D" Conditions Yalved Out Yalved Out Pumps in Service 3-3 2

Loops in Service 4

3 2

Power Level, MWe 127 118 91 6

PSF, 10 lbs/ hour 0.61 0.70 0.52 SSF, 106 lbs/ hour 1.09 0.97 0.67 Reactor Water Temperature, OF 499 502 505 Recirculation Flow, 106 lbs/ hour 29.0 21.7 15.2 Minimum burnout ratios were maintained in excess of 4.5 on all fuels throughout the test.

The recirculation flow was determined by heat balance technique.

TART.m 4 "A" RECIR NLATION PUMP TRIP TEST March 4, 1964 Final Percent Initial Conditions Transient conditions of Initial Number Pumps Operating 4

3 Power Level, HWe 136 143 126 130 4.5 0

PSF, 10 lbs/ hour 0.81 0.72

- 11.0 6

SSF, 10 lbs/ hour 1.19 1.07 1.23 1.17 1.5 RE #109, Average Level 49.5 38.5

- 22.0 RE #109. 3 Level 53.2 43.5 51.0 i

C Rasctor Pressure 1000 990 1.0 Water Temperature 'F 510 499 2.0 Core Pressure Drop Baise Cuage 12 7

- 42.0 Water Recirculation Flow 106 lbs/hr 29.0 21.7

- 25.0 Secondary Steam Generator Pressure, psig 540 450

- 17.0 AP 22 32

+ 45.0 0

Steam Flows 10 lbs/ hour A

0.300 0.005 0.060

- 80.0 B

0.300 0.390 0.346

+ 15.0 C

0.304 0.400 t 0.39d

+ 31.0 0.411 est.

D 0.284 0.383 0.370

+ 30.0 TOTAL 1.188 1.189 1.174 1.2 III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 5.

System Stability Tests (Cont'd) b.

Turbine Trip Tests 1.

Turbine Trip Test No.

1.

A turbine trip was initiated on March 4, from the following stated conditions af ter returr.ing all recirculating loops to service and reducing load in preparation for the trip.

MWe 68 6

PWF, 10 lbs/ hour 0.81 6

SWF, 10 lbs/ hour 0.12 Reactor Water Temperature, F

535 An initial increase in reactor pressure of 8 psi was followed by anl8 psi drop.

The primary steam flow dropped 20 percent, rose 33 percent, and dropped progressively thereaf ter due to rod inser-tion. A 15 percent increase in in-core readings was experienced during the pressure increase.

The pressure and flow transients experienced subsequent to the turbine trip are indicated in Table 5.

11.

Turbine Generator Trip Test No. 2 On April 12, the power level was reduced to 67 MWe via secondary steam generation in preparation for a turbine trip with all available rods completely withdrawn. The condition existing just prior to the trip and the transients experienced immediately thereaf ter are exhibited in Tables 6 and 7 respectively.

The turbine trig initiated a drop in primary steam flow of 0.17x10 pounds per hour from its initial rate of 0.75x106 pounds per hour.

The opening of the bypass valves due to the pressure rise of 6 psi resulted in a succeeding increase in flow of 0.42x106 pounds per hour.

The in-core fission chambers and out of core micro-microammeters experienced an 11 percent increase due to the partial collapse of "oids resulting from the pressure rise, and a 20 percent drop as the primary steam flow reached its peak, typical of all turbine trips conducted at this loading.

. TABLE 5 TURBINE TRIP TEST March 4, 1964 Initial Conditions Transient Conditions Power Level, MWe 68 0

0 0

Reactor Pressure, pai 1000 1008 992 982 0

Primary Steam Flow 10 lbs/ hour 0.81 0.65 1.08 0.78 0.44 In-core 20 23 Time 10:09 FM 10:19 FM TABLE 6

TRANSIENTS INITIATED VIA TURBINE TRIP Bypass Initial Turbine Valves Conditions Tripped Opened Pressure 1000 1008 990 0

PSF, 10 lbs/ hour 0.75 0.58 1.0 Micremicro Ammeter Average 42 46.8 22.5 In-Core Average, RE 105 32.7 36.3 26.6 TABLE 7 INCREMENTAL CHANGES INITIATED BY TURBINE TRIP VALVES IN TERMS OF INITIAL CONDITIONS Turbine Tripped _

Bypass Opened Pressure, PSI

+8

-10 0

PSF, 10 lbs/ hour

- 0.17

+ 0.25 Micromicro Ammeter Averages %

+11.5

-20.0 In-Core Averages, %

+11.0

-19.0 III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 5.

System Stability Tests (Cont'd) b.

Turbine Trip Tests (Cont'd) 11.

Turbine Cenerator Trip Test No. 2 (Cont'd)

Bypass flow was progressively reduced to zero and the minimum critical pattern attained was as exhibited in Figure 1.

The reactor was shut down thereafter by the progressive insertion of all rods and cooled in the normal manner in preparation for head removal.

iii.

End-of-Life Rod Pattern Figure 2 exhibits the operating rod pattern during end-of-life of Cycle No. 2.

6.

Control Rod Blades During periods of operation, control rod drives have been verified for blade following on a daily and weekly basis.

Monthly control rod worth tests were conducted until October, at which time these tests were terminated to eliminate local flux peaking and thereby minimize fuel failures.

During each start-up, control rod blade patterns for criticality have been predicted and all blade following verified.

7.

Changes in Facility Design On November 20, verbal authorization was received to replace the scintillation detector in the No. 1 off-gas system with an ionization chamber.

This installation achieved the follow-ing benefits:

a.

Reduced the radiation problem at of f-gas monitors by eliminating use of sampling pump and associated piping.

b.

Provide a more reliable system with more accurate readings and quicker response, c.

Provide a system less susceptible to saturation.

There were no other changes in plant design.

8.

In-Core Monitors During all periods of reactor power operation, the in-core monitoring system was in service and in the safety circuit within the requirements of License DPR-2.

. FIGURE 2 TURBINE TRIP PATTERN All Rods Out Except K-4 A B C D E F G H J K 10 9

8 7

MWe 67 6

PWF 0.660 SWF 0.150 5

Residual 0 Rods 4

7/ (-Stuck Rod 3

2 1

FIGURE 3 END-OF-LIFE, CYCLE #2 ROD PATTERN AND STATE CONDITIONS EXISTING DURING WIRE IRRADIATION A B C D E F G H J K 10 9

8 10 10 7

10 MWe 128 6

10 PWF, 106 lbs/ hour 0.32 6

SWF, 10 lbs/ hour 1.45 5

10 R,

F 500 T

4 10 7// (-Stuck Rod Resicual Rods 2 Rods and 4 Notches 3

10 10 2

1 III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 9.

Personnel Radiation Exposure Personnel exposures to radiation were within limits as specified in 10 CFR Part 20.

10.

Liquid Poison System The liquid poison system was operative at all times during the year.

The boron poison was sampled on March 6, May 5, and August 28.

There were no conditions which would indicate a loss of boron from the solution tank.

Boron concentrations in the reactor water remained low throughout the year, 11.

Radioactive Waste Disposal Release of radioactive liquid waste was accomplished in batch quantities at controlled release flow rates according to established procedures.

The contribution to the activity of dilution water was always maintained within the limits specified in the applicable Federal regulations.

The average contribution to the unidentified activity in the water utilized for radio-active liquid waste dilution during the year was calculated to be 0.118 uc/ml (11.8 uuc/ liter) compared to an average lLait of 1.00 uc/ml (100 uuc/ liter) for unidentified mixtures containing no radium 226 or radium 228 as specified in CFR Part 20.

Solid radioactive wastes were stored on-site pursuant to 3

License DPR-2.

A shipment consisting of 951 f t of dry radio-active waste with a total activity of 262.76 millicuries was sene. by truck to Nuclear Fuel Services, Inc., a subsidiary of W.

. Grace and Co., West Valley, New York, on September 11, 1964.

A second shipment of 968 f t3 of dry radioactive waste with an activity of 180.42 millicuries was also sent to Nuclear Fuel Services, Inc. by truck on September 15, 1964.

Concentration of noble fission products in the stack discharge to atmosphere was maintained well within license ILmits of 700,000 micro curies per second. The average activity release rate for the year while the plant was operating was approxi-mately 20,400 uc/second.

12.

Tests a.

Temperature Coefficient i'emperature coeffic.2nt reactivity measurements were coaducted at rated temperature and maximum Xenon on March 5.

Tempera-ture coef ficients reactivity measurements were also conducted during heating in a Xenon free state prior to the reattain-ment of rated power on the 9th of March.

III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 12.

Tests (Cont'd) a.

Temperature Coefficient (Cont'd)

The results obtained during temperature coefficient reactivity measurements, as plotted in Figure 4, indi-cate that a crossover point appears to be in the vicinity of 310 F and that all coefficients are posi-tive below that point and negative above.

The temper-ature coefficient at maximum Xenon was found to be approximately 75 percent of its Xenon free value at rated temperature.

b.

Fuel Assembly Leaker Detection Program A fuel assembly lenker detection program was initiated on January 9,1964, and conducted at a base load of 160 MWe with a constant rod pattern 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> after the load drop from 190 MWe to eliminate the effe~ cts of flux drif t due to Xenon transients following load changes.

Rods were selected and tested by group in such a manner as to minimize local effects of successive rod movements and to exhibit the lack of symmetry response should a rod either encompass a leaker or be in the vicinity of a leaker. The effects of rod movement on power gener-ation, steam flow, and in-core response were noted and recorded so as to provide data to aid in the analysis of stack and off-gas responses.

Many of the rods which exhibited leaker characteristics at 160 MWe exhibited the same or similar responses at 190 MWe, including one new leaker. The data obtained at 110 MWe did not exhibit the large leakers found at 160'MWe, but did exhibit two leaker locations found during previous testing.

The response to rod movement may be sensitive to power level. They certainly are sensitive to the background or field of activity due to power level, the strength of the rod and its region of control. Tests conducted at half power indicate that leakers may be found in the vicinity of B-6 and C-5 as originally found during the October, 1963, test program. The fuel assembly leak tests indicated nine cells suspected of containing leaking fuel assemblies, c.

End-of-Life Nuclear Testing Power generation was reduced from 128 MWe prior to the testing and calibration of control rods for power worth and control shape so as to increase primary steam flow since we had been limited to a minimum primary steam flow

FIGURE 4 END-OF-LIFE, CYCLE NO. 2 TEMPERATURE COEFFICIENTS OF REACTIVITY

3. 0 -

XENON FREE

=

N

~

E.

2.0

1. 0 -

Crossover ~310 F p

0.0 -

O Xenon Free v

1. 0 -

6

2. 0 -

i U

5 i

s

3. 0 -

g e

J f

x iF.

O g

4. 0 - y a

a e Maximum Xenon Point 5.0 -

6.0 -

o 7.0 -

O 8.0 -

i i

i 100 200 300 400 500 Reactor idater Temperature, F

- 22 '

III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 12.

Tests (Cont'd) c.

End-of-Life Nuclear Testing (Cont'd) of 300,000 pounds per hour at maximum secondary steam flows of 1,450,000 pounds per hour by the Genaral Electric Company.

The conditions existing pt: or to and after the reduction in power via the secondary steam are shown in the following tabulation.

Dual Cycle Characteristic Exhibited by a

_Raduction in Secondary Steam Generation 6

Power Level 10 lbs/ hour MWe

,_. wF SWF Remarks 128 0.320 1.45 Initial Conditions 116 0.380 1.17 Reduced Secondary Steam Generation

-12

+0.06

-0.28 Incremental Effect of Reducing Secondary Steam Generation The loss in capability due to the stuck rod K-4 (Paragraph III-4-b) was determined at reduced secondary steam flow by the insertion and withdrawal of its mate A-7.

The results obtained are indicated in the folicwing tabultstion.

Power Worth of K-4 in Terms of its Mate A-7 Power Level 106-lbs/ hour Control Rod Axial Position MWe PWF SWF A-7 12 116 0.38 1.17 0

109 0.32 1.17 Power Worth 12 Notches 7

0.06 0

Its power worth as exhibited is typical of the worth of peripheral rods at end-of-life.

The incremental and total power worth of a typical central control rod, E-8, was determined in both the normal end-of-life rod pattern pad after all rods were completely with--

drawn.

The results of such' calibrations are exhibited in Figure 5.

. FIGURE 5 CONTROL ROD E-8 INCREMENTAL AND TOTAL POWER WORTE CALIBRATION, MWe April 12, 1964

1. 0 -

0. 9 -

O

/

O

0. 8 -

0 0.7 8 Partials at 10 MWe = 17 8

PWF = 0.12x10 lbs/hourf

0. 6 -

e g

Y u

X All rods out

0. 5 -

MWe = 16 8

6 PWF = 0.15x10 lbs/hou.:

g 0.4 -

K o

0. 3 -

A 0.2 -

0.1 -

i s

a i

e i

i 0

1 2

3 4

5 6

7 8

9 10 11 12 Control Rod Position

. III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 12.

Tests (Cont'd) c.

End-of-Life Nuclear Testing (Cont'd)

Power level was reduced to 98 MWe. via secondary steam generation in preparation for the withdrawal of all remaining partial rods so as to determine the residual worth.

The results of such withdrawal is indicated in the following tabulation.

Residual Worth of 8 Partial Rods Power Level 106 lbs/ hour Partial Rod Position MWe PWF SWF_

10 98 0.40 0.87 11 102 0.44 0.87 12 108 0.51 0.87 The end-of-life loss in capability due to the necessity of maintaining axial shape control and licensed heat flux limits as exhibited by Figure 6 'is 10 MWe as indicated in the tabulation above.

d.

Fuel Sippine and Inspection On April 19, six days after the reactor was shut down for refueling, a program of sipping individual fuel assemblies to. locate defective fuel was initiated. A total of 151 fuel assemblies were sipped in the reactor from 3:45 p.m.

on April 18, through 4:10 p.m. on April 22.

In addition, 30 fuel assemblies ware sipped in the fuel building. The fuel building sipping was to recheck fuel assemblies which had shown a questionable result when sipped in the reactor.

The following tabulation identifies the 14 defective fuel assemblies located by the sipping technique and their core location.

The core map in Figure 6 shows the location of all fuels sipped. The results of the inspection are also shown in the tabulation on Jage 27.

~ FIGURE 6 EFFECT OT PARTIAL ROD POSITION ON IN-CORE MONITOR, RE 10J April 12, 1964

\\

N D-

• X Initini Conditions

\\N 8 Rods at Position 10 1 Rod at Zero, Stuck Rod

\\

We 98 6

PWF 0.40x10.lbs/ hour

\\

SWF 0.87x106 lbs/ hour 5

X C-3

\\

8 s\\

h B-O y-2 s\\N

\\

"x y

A-J@

98 We OfWe 108 We

//

//

25 5$

7$

10$

Per cent of Licensed Heat Flux

FIGURE 7 SIPPING RESULTS 26 25 24 R

23 R

R R

R F

R 10 22 R

R R

R R

R R

R 21 R

R R

i 9

20 R

8 R

R R

R R

R R

19 R

4 R

R R

12 R

8 18 R

R R

1 10 F

R R

F 17 R

R R

R F

R R

F R

R R

R R

7 16 R

R R

R R

14 1 R

R R

15 R

F F

R

~R F

R R

R R

R 6

14 R

F R

R R

R F

7 2

13 5 _.

R F

'R 5

R R

R R

R R

12 R

R R

F R

R R

R R

11 R

R 11 R

3 R

R R

R 4

10 R

R R

F R

R R

R R

R 09 3

R R

F 9

R R

6 R

R R

R 08 R

R R

F R

13 R

R R

07 R

F R

R 2

06 F

R R

R R

R R

F 05 R

R F

R R

R g

04 R

R R

03

_ _02 F

__ 01 l

A B

C D

E F

G H

J K

51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 R - Sipped in Reactor (Non-Leakers)

F - Sipped in Fuel Buildi'ng (Non-Leakers) 1 Leakers (See Text for Assembly Number)

. III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 12.

Tests (Cont'd) d.

Fuel Sipping and Ihspection (Cont'd)

Number Core Location Results of Inspection 1.

A-339 62 - 18 Corner rod, segments split 2.

C-93 69 - 14 UO2 rod, cracks in middle of assembly 3.

A-492 65 - 11 Corner rod, segments split 4.

A-185 61 - 19 Corner rod, corrosion penetration 5.

C-25 58 - 13 Corner and adjacent rod, 2 thoria rods cracked 6.

A-382 65 - 09 Not inspected 7.

A-223 68 - 14 No identified defects 8.

D-1 57 - 20 Multiple cracks in 3 thoria rods 9.

PF-12 61 - 09 No identified defects 10.

A-131 63 - 18 No identified defects 11.

C-97 58 - 11 No identified defects 12.

A-165 65 - 19 Corner rod, corrosion penetration

13. A-ll7 69 - 08 Corner rod, 3 cracks in segments 14.

C-30 65 - 16 No identified defects e.

Gamma Scanning A total of 102 assemblies were gamma scanned during the refueling outage as indicated in Figure 8.

All core fuels removed for examination and/or inspection were scanned along with a high percentage of the spent fuels.

The data has been analyzed by the General Electric Company and has pro-vided information on fuel performance in the form of rela-tive power, axial profile, and a measure of gross power distribution during the last month of operation.

A large fraction of the core was sampled.

The largest number of assemblies scanned occupied the northwest quad-rant of the core.

Symmetry was established by scanning fuel assemblies in other quadrants.

The behavior of various types of fuels with various exposures were deter-mined by measuring many groups of four associated with a given control blade.

The scans exhibit power distribution during the last month of operation and not exposure distribution during life.

The exposure distribut. ion was affected by the many power distri-butions experienced during operation.

No large differences have been found between the power distri-bution exhibited by the results of the gamma scans and the

. FIGURE 8 GAMMA SCANNED FUEL ASSEMBLIES 1964 REFUELING OUTAGE 26 25 24 23 10 X

22 X

21 X

X 9

20 X

X 19 X

X X

X 8

18 x

X X

X X

X

_ 17 X

X X

X X

X X

X X

7 16 X

X X

X X

X X

X X

X 15 X

X X

X X

X X

X X

X X

X X

6 14 X

X X

X X

X X

X X

X X

X X

X 13 5

X X

X 12 X

X X

X 11 X

X X

X X

X X

X X

X 4

10 X

X X

X X

X X 3 X

X X

X X

X 08 X

X X

X 07 X

X 2

06 X

05 X

X X

X g

04 03 02 01 A

B C

D E

F G

H J

K 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74

~

. III.

DISCUSSION (Cont'd)

A.

Operating Experience (Cont'd) 12.

Tests (Cont'd) e.

Gamma Scanning (Cont'd) outputs of the two computer codes used to compute end-of-Life conditions.

The standard deviation of differences between the measured and calculated values of radial power factor were found to be less than + 8 per cent.

The axial peak / average power ratios exhibited by the gamma scans were found to be about 1.28 on the average with a maximum of 1.35 characteristic of the relative flat axial power distributions exhibited by the end-of-Life wire irradiations.

f.

Sphere Leak Test A sphere [eak test at 20 psig gwaq. conducted. during the.

refueling outage to fulfill the' AEC requirement for contain-ment vessel leak. testing. The firstwas.conductedcon:

May 29, 1964, and resulted in the following measurable leaks:

Eauipment Location of Leak 1.

16' Equipment Opening Welded flange on inside of 16' door.

2.

Sphere Ventilation Inlet 3/4" test pipe plug between isolation valves.

3.

Sphere Ventilation Outlet 3/4" test pipe plug, sphere side.

4.

Fuel Transfer Tube (42")

Bolted cover plate and 2" drain valve.

The reactor enclosure was blown down and the four areas were repaired as follows:

1.

16' Equipment Opening Seal welded cracks.

2.

Sphere Ventilation Inlet Welded same.

3.

Sphere Ventilation Outlet Welded same.

5.

Fuel Transfer Tube (42")

Replaced cover plate gasket and replaced 2" drain valve.

The sphere was repressurized to 20 psig and the four areas which previously leaked were checked and satisfactorily passed the 20 psig 1 eak: test.

1 g.

Safety Valves We are changing the method of testing the primary drum safety valves. Our present method is to remove five of the ten safety valves at each reactor refueling outage in order to test the pressure at which each valve opens.

This test has been performed at one of our conventional plants to date, since-no facilities have been available at Dresden to produce the required test pressure (1205 - 1250 psig).

III.

DISCUSSION (Cont'd)

A.

_ Operating Experience (Cont'd) 12.

Tests (Cont'd) g.

Safety Valves (Cont'd)

In order to ship these five valves to our conventional plant, each valve had to be decontaminated prior to shipment.

In addition to this, one boiler at the con-ventional plant had to be removed from service over several weekends in order to test all five valves.

It usually took one day to test two valves.

The new relief valve test method consists of pressur-izing each valve with nitrogen gas to the blow off pressure.

Each valve is mounted on a flange identical to the primary drum flange.

The test flange is piped to an accumulator to provide a gas volume chamber.

Nitrogen gas bottle connections are provided to supply the necessary pressure.

The valve lifting pressure is adjusted if necessary to the prescribed value.

This test method is approved by the valve manufacturer (Manning, Maxwell and Moore) as well as the Travelers Insurance Company (Nelia Group) and State of Illinois Division of Boiler Inspection.

The use of this facility will eliminate the need for disassembly and decontamination and the possibility of damage in shipment off the site for testing.

h.

Air Locks All air locks, ventilating valves, and process isolation valves were tested periodically during the year as required by License DPR-2 and found to be within the licensed allowable leakage rate. All leaks found were repaired and made leak tight.

13 Aerovane The vind direction and velocity instrumentation was relocated at the south environs station to facilitate routine observation and mainte-nance. An additional advantage of this location is the immediate availability of the unit to the plant personnel in the unlikely event of an incident necessitating evacuation.

B.

Amendments to License DPR-2 Table.No. 8 lists the amendments to our license requirements and/or authorized during the year. Only correspondence pertaining to thes.e re-quests are tabulated in the references.

TABLJ B

SUMMARY

OF LICENSE AMENDMENTS REQUESTED DURUE 1964 Date Required Request Authorization Temporary authorization to take credit for July 30, 1963 March 13, 1964 use of respiratory protective equipment.

Authorization to change frequency of control August 29, 1963 April 9, 1964 rod drive and blade testing from quarterly to semi-annually, DPR License Change No. 7.

Authorization to decrease minimum burnout September 17, 1963 March 13,1964 ratio from 2.0 to 1.5.

DPR License Change No. 6.

Authorization to operate special fuel assembly Februtry 26, 1964 March 27, 1964 SA-1 in Dresden Reactor.

DPR License Change No. 5.

Authorization to receive, unload from cask, February 27, 1964 March 13, 1964 asserble and store byproduct material con-tained in SA-1 special fuel subassemblies.

License No. 12-5650-3.

Authorization to postpone required control February 28, 1964 March 4, 1964 rod drive testing until refueling shutdown.

Request to amend Byproduct Material License March 10, 1964 April 3, 1964 No. 12-5650-1 (H-65) to permit receipt of 50 curies of byproduct material in the form of metal test specimens to be placed in Dresden Reactor for additional irradiation.

Authorization to use plutonium beryllium March 13, 1964 March 20, 1964 neutron source.

License No. SNH-225, Amendment No. 1.

Request to amend Byproduct Material License March 13, 1964 April 3, 1964 No. 12-5650-1 (H-65) to authorize three cobalt 60 gamma ray sources of six curies each to be used for calibration of radiation detecting and measuring instruments.

Withdrawal of request to increase heat flux March 25, 1964 limits for Type II fuel.

Modification of request to amend Appendix A, March 26, 1964 May 13, 1964 DPR License to authorize an increase in heat flux limits. Change No. 8.

32 -

Date Required Request Authorization Request to revise Special Nuclear Material April 24, 1964 May 6, 1964 License No. SNM-638 ro oermit shipment of Fuel Type II and PF 1-7 in shipping cask "S-1" supplied by Stanray Corporation.

Authorization to increase flexibility in April 29, 1964 October 1, 1964 regard to high radiation area procedures and of an increase from one to two in the number of " radiation" keys in use.

General authorization to take credit for September 16, 1964 use of respiratory protective equipment.

Request to amend Appendix A, License DPR-2 September 18, 1964 regarding revising the specifications of maximtr.n permissible stack release rate to include I131 and halogens.

Modification of request to amend Appendix October 21, 1964 A, License DPR-2 regarding revising the specifications of maximum permissible stack release rate to include 1131 and halogens.

Request to amend License DPR-2 to permit December 24, 1964 operation with 104 Type III-F fuel assemblies with removable poison rod.