Text

Operation Rept 2 for Feb 1961
	ML19351E019
Person / Time
Site:	Yankee Rowe
Issue date:	03/08/1961
From:	; YANKEE ATOMIC ELECTRIC CO.
To:	;
References
	NUDOCS 8011250435
	Download: ML19351E019 (11)
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YANKEE NUCLEAR POWER STATION OPERATION REPORT NO. 2 for the period January 30, 1961 ----- February 28, 1961 O

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The following repsrt covers the operation of the Yankee Atomic Electric Company plant at Rowe, Massachusetts, from the conclusion of the si.x months report period (January 29,1960) through February 28, 1961.

For the period of January 30, 1961 through February 13, 1961 the plant operated continuously at a level of 1pproximately 120 M4 gross electrical i

output. On February 8, 1961 the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at 3c2 M# thermal was completed.

Commencing February 14 and continuing through February 23, a series of-outages occurred as a result of excessive valve stem leak-off flow from shutdown cocling and primary loop by-pass valves. Corrective meatares were taken and the plant-once again is operating at the presently license 1 full power level.

A side effect of the increased stem leak-off flow was a temperature increase in the Primary Drain Collecting Tank. Subsequent boiling in the tank h-

caused several small puffs to be released through the waste disposal system blanket gas loop seal. Monitoring of the area, however, indicated no detectable increase in radioactivity.

An unusually severe period of cold weather at the beginning of Feb-ruary resulted in freezing of various lines in the waste disposal area.

There were 6 plant shutdowns during this report period. Following p

is a listing and description of the shutdowns, the numbering of which follows Q

sequentially those indicated in the six months report.

Shutdown No. 19 2/14/61 - A 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months 43 minute outoga resulting frora excessive stem leakage from shutdown cooling valve SC-MOV-553.

Immediate symptoms were an increase in temper ~

ature in the Primary Drain Collecting Tank and popping of the safety valve in the stem leak off header. Entry into the vapor container led to detection of the de-

-O fective valve. The valve packing was tightened and this action appeared to reduce leakage to an acceptable level.

Shutdown No. 20 2/14/61

- A 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 49 minute outage resulting'once again from excessive stem leakage from shutdown cooling valve SC-MOV-553. As happened earlier on the same day, a rise in temperature in the Primary Drain Collecting Tank was noted, together with an increase in main coolant system leakage. The vapor container was entered and an attempt was made to repack the valve. Inability to attain sufficient backseating tightness to repack the valve under pressure led to an alternate solution of plug-ging the leak off line from the valve at the break union. The packing was also taken up aca'i and the secondary packing appeared tight enough to prevent further leakage.

Shutdown No. 21 2/19/61

- A 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months 23 minute outage for control rod interchange and valve inspection inside the vapor container. The condition of the primary loop bypass valves and shutdown cooling valves indicated that cooldown of the primary plant and repack-ing of the valves would-be required.

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g iShutdown Nor 22;.2/19/61 A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 55 minute' outage resulting

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scram occurred-after the generator had been~ phased and load was being increased

to.obtain noise level data on normal plant neutron flux channels for a repre-sent tive Af the: B & W company. The cause of the scram is being investigated.

Shut'down No. 23 2/19/61 An outage of 91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months , 41 minutes for

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2/23/61

_ valve maintenance inside the vapor container. Borating and depressurizing of the' primary system were necessary since valve backseating was not sufficiently. tight. Considerable difficulty was experienced in repacking the valve's due.primarily to the valve design and con-struction (all valves concerned were supplied by the'same manufacturer). Upon.

return ~of the. primary system to pressure.and temperature, stem leak-off flow re-mained high.. It'was then decided to cap the leak-off lines on the two shutdown cooling valves,'the four primary loop by-pass valves and the four safety injection.

valves..The tightness of the packing above the leak-off point appeared satisfac-tory.. While maintenance measures to this point.are considered adequate on_a temp-orary. basis,'he problems of stem leakage and shortcomings in-the valve design have

_ been referred to the manufacturer for solution.

Shutdown No.'24 '2/28/61 - A 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 50 minute. outage to inspect

-valves in the vapor container which had been. repaired at the previous ~ shutdown. The above-leak-off packing on those valves whose leak-off lines had been capped continues to remain tight.

On February 14, 1961 at approximately 2:00 A.M. the Waste Disposal System blanket gas water seal was blown by overpressure. Several small puffs of blanket ga's were' released through the loop seal. While the volume of gas dis-charged is indeterminate, monitoring _of the area indicated no detectable activity.

The Nactor nad just'been reduced in power level from 120 MWe to O MWe at the time d the release.

This occurrence was caused _by an increase in valve stem leak-off flow from a shutdown cooling valve.

Increased flow from the leak-off' header to the Primary. Drain Collecting Tank elevated the temperature in the tank to the point -

where boiling occurred. The byerpressure, caused by boiling, was transmitted to h

the waste disposal blanket gas system.

To prevent recurrence.of such an incident, it has been established Lthat' plant: operators will close the leak-off header discharge line to the Primary Drain _ collecting Tank if'the tank temperature increases because of large valve stem leakage. I.eak-off flow then will pass to the low pressure surge tank thrnugh the leak-off header relief. valve. A temperature indicator has been instc R.a in the leak-off. header and a radiation monitor installed in the incinerator stack.

Plant' Maintenance and Modifications

.. Extremely cold weather in the latter part of January and'the begin-ning of February resulted in freezing of.several outdoor lines. Maintenance personnel.were engaged during this period in thawing operations and increasing line tracer-heating capability.

As-described in the tabulation of outages, numerous difficulties were encountered :in the maintenance operations.on certain valves. While a tempo-

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. rary modific'ation'has been made -- namely, capping the stem leak-off lines and-b

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relying on the tightness of the above-leak-off packing, a decision on a permanent t

soluticn has been deferred pending the manufacturer's recommendations.

Leakage from the pressurizer solenoid relief valve has been noted and the valve will be repaired at the next maintenance outage.

A defective relay which had rendered a core flux wire drive inop-erative has been replaced.

Repair and adjustment of position indicating micro switches on the core flux wires in the vapor container were accomplished during the plant outages.

In addition to the above, normal plant and instrumentation mainten-ance procedures were carried out during this report period.

plant Chemistry

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Some difficulty has been experienced during recent weeks with high carryover from the waste disposal evaporator. Since the evaporator effluent has not been of suf ficient purity for reuse in the primary system, a temporary in-crease in liquid waste has been evident.

A total of 144,360 gallons of liquid were processed through the waste disposal evapora tor in February. Seventeen, 55 gallon drums of evaporator bottoms with a total activity of 13 milli curies were stored.

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Because of inability to recirculate evaporator effluent, make-up feed to the primary system during much of this period consisted of demineralized water.

The primary system was deborated following Shutdown No. 23.

The anien column used for this operation performed satisfactorily.

The primary water system operated within design specifications during (3

this period with a purification rate of 10 gpm. No difficulty was encountered in

's,1 maintaining the secondary system within design specifications.

Routine prittary and secondary system chemical analysis and treatment wereca3riedoutduringtaigperiod.

primary system specific activity ranged from 9 x 10 /c/ml to 1.2 x 1C, A/ml before boration while ef ter boration the activ-ity level was approximately 5.5 x 10 V c/ml.

A typical primary system crud analysis before boration i.'icated:

9 Fe 4.65 x 10 dpm/mg 60 5

Co 2.69 x 10 dpm/mg 8

Co 5.44 x 10 dpm/mg Cr 6.3 x 10 dpm/mg Mn #

3.49 x 10 dpm/mg Crud Level

= 0.026 ppm 131 1.16 x 10~ /if/ml.

Iodine present in primary water

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typical radioactive gas analysis indicated:

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-2 A-41 1.72 x 10 vt/cc j

Kr-85m 9.3 x 10

/"c/cc Xe-133 5.6 x 10

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-3 Xe-135 2.42 x 10

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E was dis-During January liquid waste with a total activity of 298/

charged while in February, the total activity of discharged liquid waste was 1607

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At all times, the concentration was well below the maximum permissible con-centration.

v In January waste gas with a total activity of 363j 'c was discharged, the concentrations being well below the maximum permissible. No scheduled dis-(;

charge of waste gas took place during February.

V Seventeen barrels of solid wastes with a total activity of 13 milli curies were drummed during this report period. No barrels were shipped.

Surveys of contamination were made in the vapor container in con-junction with valve maintenance. Activitylevelsof10,000DPM/ft2 were noted in the mmediate area.

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A small area in the vicinity of the stem leak-off header safety valve was surveyed following popping of the valve. An activity level of 6000 DPM/ft2 was indicated and decontamination was ef fected without dif ficulty.

No cases of overexposure were noted during this report period.

The regular monthly station safety meeting was conducted in February and an in-plant first aid course has been initiated.

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v Reactor Plant Performance The noise on normal plant neutron flux channels was measured at vary--

ing power levels by representatives of the B & W company. The primary purpose of this measurement was to check equipment to be used on the N. S. Savannah.

Several flux wire irradiations and analyses have been made during this report period.

In addition, routine plant data analysis continues.

Turbino plant Performanco Studies are being conducted to determine secondary plant thermo-dynamic performance.

Specific measures are also being taken to improve steam quality measurements.

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A training course-ior plant personnel providing a review of reactor theory and other subjects pertinent to the A.E.C. operator licensing examination commenced during February.

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Plant Operations i

l Included in this-report are copies of the plant operating summaries i

for November and December,1960, and for January and February,1961. These will j

serve to provide a continuous record of operating statistics from initial power generation in November to the preseat report period.

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YANKEE ATO4IC ELECTRIC CCMPANY 3000-Gross Generation - - - --

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OPERATING St254ARY I

MONTH:

November 1960 Month to Date Gross Generation KWH 15',250,800 15,250,800 i

Station Service (While Gen. inc. Losses)

KWH 2,203,084 2,203,084 Net Generation KWH 13,047,716

'13,047,716 Station Service -- Percent 14.4%

14.4%

Station Service (While Not Gen. inc. Losses)

KWH 1,545,411 1,545,411 Ave. Gen. -- For Month

( 720 Hrs.)

Net KW 18,122 O

Ave. Gen. -- Running

( 255.13 Hrs.)

Net KW 51,141 Equivalent Reactors Hours at 392 WIT HRS 143.56 143.56

'Use Factor -- Percent 25%

'25%-

Circulating Water Temperature -- F Max. F Min. F O

Control Rod Position at Month End --

Equilibrium and Full Power.(at 90 EfE Gross)

Group 1 -- Rod Out Inches -------

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0" Group 3 --

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vo-7/8" Group 5 --

0" Group 6 --

90" Times Critical 36 175 Hours Critical 532.87 1146.35 Times Scrammed 7

94-Net Plant Efficiency -- Percent 23.2%

Steam Flow - Lbs.

Lbs. Steam / Net.KWH Accurate reading unavailable during initial operation.

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YANKEE ATOMIC ELECTRIC COAPANY

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SUMMARY

MONTH:

December 1960 Month to Date Gross Generation-KWH 23,634,100 38,884,900 Station Service (While Gen. inc. Losses)

KWH 2,994,709 5,197,793 Net Generation KWH 20,639,391-33,687,107 Station Service.-- Percent 12.7%

13.4%

Station Service (While Not Gen. inc. Losses)

KWH 2,182,485 3,727,896 Ave. Gen. -- For Month.-

( 744 Hrs.)

Net KW 27,741

. Ave. Gen. -- Running

( 388 95 Hrs.),

Net KW 53,064 Equivalent Reactors Hours at 392 M#T -

HRS 200.30 343.86 Use-Factor -- Percent 26%

26%

Circulating Water Temperature --

F.

Max.

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-40 Min. F 34 Control Rod Position at Month End --

Equilibrium and Full Power (at 65 MNE Gross)

Group 1 -- Rod Out Inches -------

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A 0"

66-3/4"

. Group 4 --

r Group 5 --

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. Group 6;--

90" Times Critical 22 197

. Hours Critical 558.73' 1705.08 Times Scrammed.

12 106

~ Net' Plant: Efficiency -- Percent

~26.3%

-Steam-Flow

.Lbs.

Lbs.' Steam / Net KWH.

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Accurate reading unavailable during initial operation.

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YANKEE ATC841C ELECTRIC COMPANY OPERATING

SUMMARY

MONTH:

January 1961 Month to Date Gross Generation KWH 43,130,200 82,015,100 Station Service (While Gen. inc. Losses)

KWH 3,667,325 8,865,118-Net Generation KWH 39,462,875 73,149,982 Statim Service -- Percent 8.5%

10.8%

Station Service (While Not Gen. inc. Losses)

KWH 1,011,499 4,739,395 Ave. Gen. -- For Month

( 744 Hrs.)

Net KW 53,041 Ave. Gen. -- Running

( 395.63 Hrs.)

Net KW 99,747 Equivalent Reactors Hours at 392 W T HRS 361 99 705.85 Use Factor -- Percent 48%

34%

Circulating Water Temperature -- F Max. "F 40 Min. F 35 O

Control Rod Position at Month End a Equilibrium and Full Power (at 120 W E Gross)

Group 1 -- Rod Out Inches -------

84" Group 2 --

0" Group 3 --

90" Group 4 --

O 0"

Group 5 --

0" Group _6 --

90" Times Critical 13 210 Hours Critical 484.25 2189.33 Times Scrammed 9

115 Net Plant Efficiency -- Percent 27.8%

Steam Flow - Lbs.

Lbs. Steam /NetKWH Accurate reading unavailable during initial operation.

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OPERATING'

SUMMARY

MONTH:

February 1961 Month to Date Gross Generation KWH 64,480,700 146,495,800 Station Service (While Gen.'inc. Losses)

MNH 5,642,182 14,507,300 Net Generation NWH 58,838,518 131,988,500 Station Service'-- Percent 8.8%

9.9%-

Station Service. (While Not Gen. inc. Losses)

KWH 533,151 5,272,546

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Ave. Gen. -- For Month

( 672 Hrs.)

Net KW 87,557 Ave. Gen. -- Running

( 558.63 Hrs.).

Net KW 105,326 Equivalent Reactors Hours-at 392 MWT HRS 538.31 1,244.16 Use Factor -- Percent 80%

46%

Circulating Water Temperature -- F Max. F 42 Min. F 34 O

Control Rod Position at Month End --

Equilibrium and Full Power-(at 120 MWE' Gross)

Group 1 -- Rod Out Inches -------

0" Group 2 --

90" 17}"

Group 3 --

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Group 4 --

90" Group 5 --

0"

' Group 6 --

90" Times Critical 1

211 Hours Critical.

581.30 2770.63 Times' Scrammed 1

116 Net Plant Efficiency -- Percent 27.9%

Steam Flow - Lbs.

754,092,000

' Lbs. Steam / Net MNH '-

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YANKEE f;UCLEAR POWER STATION i

l OPERATION REPORT NO. 1 Initial Startup and Test Operations of the Yankee Reactor for tle period July 9, 1960 -- January 29, 1961 i

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Submitted in accordance with Facility Lisense DPR-3, as amended, i

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YANKEE ATO.iIC ELECTRIC CO.tPANY Boston Massachusetts February 13, 1961

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CONTENTS SECTION PAGE I.

~ Introduction ~-----------------------------------------

1

. II.

Summary -- -------------------------------------------~

2--13

1. -Period and Activit'ies -------------------------

2 2.

Reactor Core Loading and Assembly -------------

2

3. - Cold and Hot Control Rod Scram Tests ----------

2 4.

Initial Criticality ---------------------------

3 5.

Low Power Physics Testing ---------------------

3 6.

Power ~ 0pera tion Te sting -----------------------

3---8 j-A 7.

500 Hour Test Run at 392 MW Thermal -----------

9 k-8.

Operating Statistics --------------------------

9 9.

Chemistry -------------------------------------

L 10.

-10.

Health Physics --------------------------------

10

11. Systems and Components ------------------------

11 i

12. Design Changes --------------------------------

11

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13.

Operating Procedure Changes -------------------

12 14.

Plutonium Buildup Experimental Procedure ------

12 III.

Reactor Core Loading and Assembly --------------------

14 s

IV.

Initial Criticality ----------------------------------

16 V.

Tests and Measurements -------------------------------

17--29 1.

Initial Low Power Nuclear Core -Tests ----------

17 a.

Purpose and Scope -------------------------

17

b. -Control Rod Drive and Plant Scram Tests ----

.17 c.

Control Rod and Boron Worth Determinations a t-Low and Operating Temperatures ---------

18 d.' Temperature, Pressure and Flow Coefficient Determinations ----------------

18 e.

Main Coolant System Heating Rate Determinations ----------------------------

19 f.- Nuclear Instrumentation Response to

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. Asymmetric Control Rod. Positioning.--------

19 e

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4 SECTION PAGE V.

Tests and Measurements (cont'd)

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2.

Initial Power Operation Tests -----------------

20 a.

Pu rpo s e a nd Sc op e ------- ------------------

20 b.

Power Coefficient and Loss of Load Transient Test ----------------------------

20 c.

Nuclear Instrumentation Power Calibration -

20 d.

Reactivity vs Fission Product Level Following Power Level Changes -------------

21 e.

Biological Shielding Effectiveness Test ---

21 f.

Instrumeritation and Control Response ------

22 g.

Emergency Cooling by Natural Circulation --

22 3.

500 Hour Test Run at 392 WW Thermal -----------

24 4.

Chemistry -------------------------------------

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a.

Prima ry Co o l a n t ---------------------------

25 b.

Wa te r R a d i oa c t ivi ty -----------------------

26 c.

Waste Disposal ----------------------------

27 d.

Secondary Wa ter Chemistry -----------------

27 5.

Health Physics --------------------------------

28 O

88ei tien tevete Contamination Levels Radiation Exposures Waste Disposal: Liquid, Solid, Gaseous ----

29 VI.

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Systems and Components -------------------------------

30--47

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In-Core Instrumentation System ----------------

30 2.

Main Coolant System ---------------------------

31 3.

Pressure Control and Relie f System ------------

33 4.

Charging and Volume Control System ------------

33 5.

Chemical Shutdown System ----------------------

34 6.

Purification System ---------------------------

35 7.

Component Cooling System ----------------------

35 8.

Primary Plant Corrosion Control System --------

35 9.

Primary Plant Sampling System -----------------

35

10. Radioactive Waste Disposal System -------------

36

11. Shutdown Cooling System -----------------------

37 M(x) 12.

Primary Plant Vent and Drain System -----------

37

F 4

SECTION PAGE

~'S VI.

Systems and Components (cont'd)

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13.

Safety Injection System -----------------------

38 14.

Reactor Control System ------------------------

38 15.

Nuclear Instrumentation and Reactor 39 Protection System -----------------------------

16. Radiation Monitoring System -------------------

40 17.

Vapor Container Atmosphere Control Systems ----

41 18.

Fuel Handling System --------------------------

41 19, Main and Auxiliary Steam Systems --------------

42

20. Cond;asate and Feedwater System ---------------

43

21. C ircula t ing Wa ter Sys tem ----------------------

43 44 22.

Compressed Air System -------------------------

44

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23.

Electrical System -----------------------------

U 24.

Reactor Vessel -------------------------------

44

25. Radiation Shielding ---------------------------

45 26.

Turbine Generator -----------------------------

46 VII.

D e s i g n C h a n g e s ------------- ------ ------------ --------

48--50

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VIII.

Plutonium Buildup Experimental Procedures ------------

51--54 Q;'

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LA I.

INTRODUCTION This report is submitted in compliance with paragraph 3.C.(2) of Facility License No. DPR-3, as amended.

Yankee Atomic Electric Company's power station at Rowe, Massa-chusetts, has been engaged in initial core loading and startup operations for a period beginning July 9, 1960, and ending January 29, 1961, six months after issuance of the. Provisional License.

This operation was carried out r

as at.'horized by Interim Facility License No. DPR-3, dated July 9,1960,

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and by Amendment No. 1 to that license, dated July 29, 1960. Amendment No.1 allowed reactor operation at power levels not to exceed 392 megawatts thermal.

The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at 392 MW thermal began at 3:25 P.M. on January 17, 1961 and, at the end of the six months reporting period, midnight January n

29th, operation had continued with only minor interruptions for a total of U

approximately 280 hours0.00324 days 0.0778 hours 4.62963e-4 weeks 1.0654e-4 months . The run was completed at 12 noon on February 8,.

1961, with a total gross generation of approximately 60,500,000 kilowatt-hours. A factual report of this test operation is included herein.

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2.

II.

SUMMARY

1.

Period and Activities O( T Th*s report covers the six month period from date of Provisional License issuance, July 29, 1960 to January 29, 1961. During the period, core loading and attendant reactor assembly, initial criticality, low-power physics testing and'the initial power operation and testing wcre completed. The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at 392 MW thermal was begun on January 17, just prior to the expira-tion of the six month reporting period, and completed on February. 8. This operation is reported in Section V. - Tests and Measurements, paragraph 3.

2.

Reactor Core Loadino and Assembly previous to July 9, 1960, the reactor vessel internals required for core loading had been assembled, cleaned and installed in the reactor vessel.

The first of the four neutron sources was installed on July 13 with the first fuel assembly being loaded on July 16.

Core loading progressed slowly due to various mechanical difficulties with fuel handling equipment, temporary nuclear f'} _

instrumentation difficulties and an operating license restriction. Loading of

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the 76 fuel assemblies, 24 control rods and 8 shim rods was completed on July 26.

The period from completion of loading to installation of the reactor vessel head on August 10 was consumed with nuclear instrumentation check-out, assembly of the upper reactor internals, and installation of the in-core instrumentation structure. The installation of the vessel head studs was completed by August 12, with control rod. drive power and position indicating coils and cables con-nected by August 14.

(See Section III. - Reactor Core Loading and Assembly)

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3.

Cold & Hot Control Rod Scram Tests The check-of f list detailing the Control Rod Drive and Cold Plant Scram Tests was initiated on August 14.

Electrical grounds appeared on one or

... ore of the operating coils on eight control rod mechanisms, while one control rod drive could not be moved from the fully inserted position. Operating coil stacks free of grounds were rearranged onto control rod mechanisms which had formerly held grounded coils, allowing completion of the full withdrawal and s,

drop tests of 23 out of 24 control rods.

Investigation of fLe inoperative s

control rod drive revealed that a procedure of first driving a control rod inward to assure bottoming against a stop had, in this case, resulted in lock-ing of the mechanism. The revised procedure provides for bottoming by gravity only. On August 18 the cold control rod drop tests were completed; the' total drop times from initiation of scram to full in position being well within the calculated two seconds.

On September 23, after the reactor had reached normal operating temperature and pressure, the hot control rod drop tests were begun.

These tests consisted of 30 full height withdrawals and scrams for one mechanism followed by multiple drops,-but not less than two each for the remaining 23 mechanisms. All the total rod drop times, including release time, drop time and dashpot closure time were measured within the specified two seconds.

Variations in drop times for the same mechanism and between different mechan-isms were small and could be explained by slight dimensional variations and f~

dif fering temperature, pressure and flow conditions.

(See Section V. - Tests

(,7 and Measurements, paragraph 1.b.)

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4.

Initial Criticality Preparations for the initial approach to criticality were complete on August 19.

The initial approach, executed by a banked rod configuration, required approximately two hours, was smooth, and was carried out in accordance with writte p20cedures. The low-power physics testing program, as outlined in

%ction 503E of the licens 3 application, thus began in earnest on August 19 at 8:19 p.M.

(See section IV. - Initial Criticality).

5.

Low power Physics Testi_ng Although the Initial startup program was ready to proceed with the low power physics testing phase, it was r.ot until September 4 that tests other than cold control rod drop tests were begun. During the low power tests, the reactor was operated at power levels not in excess of 5 megawatts thermal.

The period of September 4 to 16 saw collection of data at low reactor tempera-tures, from which the reactivity worths of control rods and boric acid were calculated. Temperature, pressure and flow coefficients with increasing reactor temperature were determined, and Main Coolant System Heating Rate Determinations

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were made from September 17-22.

Tests of Control Rod and Boron Worth at Opera-ting Temperature and Instrumentation Response to Asymmetric Control Rod Posi-tioning were begun on September 24.

Leaks and minor mechanical troubles were experienced from time to time on both the primary and secondary plant, resulting in the low pcwer testing program not being completed until November 8, 1960.

Collection of supplemental data for several of the tests previously performed, as well as measurement of temperature, pressure. and flow coefficients with de-creasing reactor temperature were, however, continued as plant operating condi-tions permitted.

Major delays during the 503E tests were caused by water leakage dif-ficulties at the main coolant pump bolted flanges, malfunction of the in-core instrumentation system, and repair of test plug welds located in the shell of the steam generators.

(See Section VI. - Systems and Components)

Reduction of the data produced during this phase of startup shows that values for Control Rod and Boron Worth were in good agreement with the calculated values. The measured temperature and pressure coefficient values, g!j with no boron in the main coolant, were also approximately as preducted, whereas coefficients measured with high boron concentrations were more negative than analytical work predicted. There were no detectable flow coefficients at low reactor powers. Results of Main Coolant Heat Rate Determinations showgd that with four gain coolant pumps operating, the heatup rate ranged from 19 F per g

g hour to 14 F per hour as main coolant temperature changed from 470 Fgo540 F.

The system heat losses were determined to be approximately 3.0 x 10 BTU per hour. Tests of nuclear instrumentation response to asymmetric %ntrol rod positioning showed that detection of resulting flux tilts was possible. Further testing, however, during power operation showed that nuclear instrumentation was also sensitive to variations in coolant temperature and xenon condition in the core. As a result, these additional factors complicate absolute calibration of the nuclear instrumentation for flux tilt detectio6.

(See Section V. Tests and Measurements, paragraphs 1.c, d, e and f) 6.

Power Ooeration Testinq Initial power operation and testing began November 10, $960 at 2:37 A.M. wnen the turbine-generator was phased with the New England inter-l f

t connected transmission system. Plant operation and performance of tests during this phase of startup followed, in general, the schedule outlined in Section 503F of the license application.

4.

s The performance of these tests showed that the reactor plant was extremely stable under all operating'and transient conditions (120 MW gross y

(

electric, nominal). Limiting operational ~ conditions, shown by transient tests, were reached in the secondary plant before they were reached in the reactor uleg generator load changes show a value.of -3.3 x 10~g major plant. power coefficient measurements taken during th dk/k/MWthermal 11.7 x 10-for the first 500 equivalent full power hours of operation which compares very favorably with the analytical value. Loss of load transient tests were performed at 30 and 60 MV gross electric plant loads. No plant limitations were exceeded during these tests. Collection of data, showing. relationship between primary and secondary systems, showed excellent agreement with heat balance data. It has been determined that measurement of average reactor output is possible to approximately 15%. The nuclear instrumentation is calibrated using gross electric output as a measure of average reactor power. Data ob-tained, relative to variation in reactivity due to transient fission products-following power level changes and changes following turbine-generator shutdown, shows that the Yankee reactor will be able to override peak xenon throughout core' life since reactivity lost due to peak xenon is less than reactivity gained from the power coefficient during the load reduction. The calculated values of O

reactivity loss due to xenon are in reasonable agreement with experimentally V

determined values. Performance of Biological Shield Effectiveness tests at 15, 30 and 60 EV gross electric, revealed higher than anticipated neutron levels throughout the site. Neutron levels at 60 MW gross electric load were reduced by a factor of 8 with installation of temporary shielding over the top of the annulus between the reactor vessel and the neutron shield tank. Later installation of permanent masonite shielding showed that neutron levels were reduced by a factor of 19 over the unshielded condition. The highest neutron level at 120 fM gross O

electric is 7.0 mrem /hr. Plant control was found to be extremely stable; the V

inherent negative temperature coefficient proving to be an effective stabilizing influence.

It was also found that the reactor regulated on temperature control with no rod motion required for small load variations. The automatic rod con-trol system has proven to be more stable than anticipated, holding main coolant temperature within the control band during normal generator load changes and steady power operation. The tests of emergency cooling by natural circulation were performed from 60 and 120 fM gross electric. These showed that adequate p

cooling was providad to the core under conditions simulated in the test. * (See U

Section V. - Tests and Measurements, paragraphs 2b, c, d, e, f and g)

This period included 15 plant shutdowns; twc for reasons of scheduled nuclear data collection and 13 due to turbine-generator testing and operation difficulties. A summary and chronology of the shutdowns follow, with further information on systems and component modification performed appearing in Sec-tion VI - Systems and Components.

Shutdown No. I 11/10/60 - This shutdown of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 26 minutes duration occurred at the conclusion of a 15 MW gross electric loss of load trip test. Although this loss of load test was not detailed in Startup Procedure 503F1, the turbine manufacturer requested that it be done before proceeding to the scheduled higher loss of load transient tests. Upon manually tripping the unit, high moisture separator levels occurred.

The turbine overspeed trip tests were also performed during this period.

Shutdown No. 2 11/10/60 This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 29 minute shutdown was caused by a high, low pressure turbine f')

casing temperature in addition to continued high water level problems in the U

turbine moisture separators. During the shutdown period, a one inch vent line was installed from each moisture separator to the feedwater heater drain re-ceiver.

Installation of these lines was made in an effort to eliminate the high moisture seperator_ levels before proceeding to the scheduled 50 hour5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months -

30 MW gross electric power run.

4 5.

' Shutdown No. 3 11/11/60 - This scheduled shutdown of 110 hours0.00127 days 0.0306 hours 1.818783e-4 weeks 4.1855e-5 months duration occurred at the conclusion of the 30 W1 gross electric run. Turbine-generator shutdown was accomplished

(~)

by the Loss of Load Transient Test detailed in Section 503F1 of the license application. During the transient test, high turbine moisture separator levels again appeared. Accordingly, the previously installed 1" moisture separator-drain receiver vent line was removed and replaced with a 11" vent line. The first of the several 503F6 tests, Variation of Reactivity Due to Change in Fission Product Level Following Reactor Shutdown, was also performed.

Shutdown No. 4 11/16/60 - This shuthwn period of 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months and 24 minut o sas the result of an inability to hold a constant load on the turbine-generator unit. Shutdown was orderly,-

with later investigation showing that No.1 Turbine Control Valve Servo-motor operation was erratic. The shutdown period was, therefore, consumed with fab-rication and installation of the necessary new parts for this piece of equipment.

, Shutdown No. 5 11/22/60 - This 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 39 minute shutdown occurred during the scheduled 70 hour8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months -

60 MW gross electric power run. This shutdown was initiated by a turbine trip due to _ a high moisture separator water level indication. A series of coinci-p) dences involving the feedwater heater drain system operation and a critical

(

location of the moisture separator high level switch caused a false high level indication, resulting in a spurious turbine trip and reactor scram.

Shutdown No. 6 11/23/60 - This scheduled shutdown period, at the conclusion of the 60 MW gross electric run, resulted in an 83 hour9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months outage. Shutdown was initiated by the performance of a low flow reactor scram and turbine trip test, accomplished by tripping one

(,")

of three operating main coolant pumps. Startup Procedure 503F7, Emergency Cooling by Natural Circulation, followed the transient test. Measurement of xenon decay, as detailed in 503F6, took place throughout the remainder of the shutdcwn period.

During the 60 WI gross electric power operation, Biological Shield Effectiveness Tests continued to be made as called for by Startup Procedure 503F4. These tests disclosed that in certain plant areas, fast neutron radia-tion was at levels in excess of calculated values, although still well below t

)

acceptable levels. During the shutdown period, the reactor was operated at a n

low thermal power level while detailed surveys were performed. Fast neutron streaming was found to exist in an upward direction from the annulus between the reactor vessel and the inner cylindrical wall of the neutron shield tank.

The situation was remedied during the shutdown period by placing temporary shielding in this area in the form of 46 water-filled drums arranged in a double layer above the offending annulus.

Shutdown No. 7 11/26/60 - This shutdown of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 54 minutes occurred only one-half hour after the start of power generation. Shutdown was initiated by the primary system flow instrumentation. Rearrangemen+ of loop flow instrumentation components, re-sulting f rom a loop flow instn cent being. out of service for maintenance pur-poses, resulted in this spurious reactor scram and turbine trip. Completion of flow instrumentation checking and maintenance was thus completed during the shutdown period before proceeding to the next power level of operation at 90 MW gross electric.

Shutdown No. 8 11/30/60 - This shutdown of 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months and 45 minutes resulted from excessive turbine shaft vibration indications at No. 1 turbine bearing. Previous to the vibration trouble, generator loading had been held at approximately 90 MW gross electric

6.

for some 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months .

In an effort to determine the cause of the vibration,

-various electrical loads were placed on the unit while changing lubricating 7_s j

)

oil temperature. No repeating conditions could be found between these factors and the turbine vibration.

In an attempt to correct this difficulty it was decided to raise the turbine high pressure casing slightly, by the addition of a 7 mil shim to each corner of the casing.

Shutdown No. 9 12/2/60 - This shutdown of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and 40 minutes was for the purpose of removing the previously installed shims from each of the four corners of the turbine high pressure casing. Results of the completed power run had indicated no improve-ment in the shaft vibration. No further attempt to correct the condition was-made during this shutdown since a major inspection was planned for the follow-ing day.

Shutdown No. 10 12/3/60 - As indicated in Shutdown-No.9 of 12/2/60, this 244 hour0.00282 days 0.0678 hours 4.034392e-4 weeks 9.2842e-5 months and 28 minute butagea was the result of continued shaft vibration. Various analyses indicated that the solution to the problem in all probability lay in a bearing modification. As a result, turbine bearings Nos.1, 2 and 3 were removed from the unit and the

(;

necessary machining operations performed.

No. 4 bearing was also removed, in-

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spected and reinstalled. The inspection of this bearing indicated that its operation was satisfactory and therefore required no modification.

In addition to the above bearing modification, clearances on turbine oil and steam seals were increased.

Shutdown No. 11 12/13/60 - This 15 minute separation from the elec-trical transmission system was accident-ally caused by the turbine manufacturer's representative while testing the over-()

speed trip test oil pressure. During the test, the turbine shaft movala sufficient amount in an axial direction to cause a high thrust bearing pressure to be devel-oped, resulting in the turbine trip.

The previous power run also proved that the bearing modifications had not solved the vibration difficulty.

In the continuing search for the answer,.

consideration was now being given to another possible solution; that of a revised opening sequence for the turbine control valves.

()

Shutdown No. 12 12/15/60 - While investigations of control valve

'-'4 opening sequence and their effect on shaf t vibrations at various loads was being explored, a sudden load drop from 60 to 10 MW gross electric was experienceJ. This occurred during the adjust-ment of the No. 1 control valve gag, resulting in the turbine manufacturer's representative manually tripping the turbine. The No. 1 turbine throttle valve, however, did not go fully closed and thus did not initiate a reactor scram. The latter operation was performed manually.

This 65 hour7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months and 55 minute shutdown was occupied with disassembly, repair and assembly of the faulty No. 1 throttle valve. Upon disassembly it was discovered that relative motion between the valve disc and stem was restricted preventing proper poppet valve operation. The temporary repair of this valve allowed further turbine test work to procaed.

During a vapor container inspection tour, a leak was discovered at an instrument nozzle in the shell of the No. 1 Steam-generator. This leak dic-tated that No. 1 loop be isolated ard drained. A load limit of 70 MW gross

(']'

electric was imposed upon the plant caring the resulting 3 loop operation while experimental data for this mode of operction was obtained.

i

8.

Three shutdowns occurred during the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at the 392 MW thermal power level. Two of these were of a planned nature, while the third was an unscheduled turbine trip. A summary and chronology of these shutdowns follow:

Shutdown No. 16 1/22/61

- This scheduled shutdown of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 22 minutes duration occurred in order to perform Startup Fracedure 503F7 - Emergency Cooling by Natural Circu-

~

lation at the 392 MW thermal level.

Inspection of the vapor container equip-ment revealed thermal insulation blistering on the No. 2 steam-generator outlet steam piping.

It was decided to return the plant to full load operation, with another inspection scheduled in approximately one week.

Shutdown No. 17 1/25/61 - This shutdown occurred as a result of construction forces accidentally con-tacing a moisture seperator high water level switch. Closing of this switch caused a turbine trip which in turn caused a reactor scram. This resulted in a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 45 minute seperation from the interconnected electrical system.

Shutdown No. 18 1/28/61

- The plar.L was shut down in order to make a scheduled vapor container in-spection. Thermal insulation on No. 2 steam-generator outlet piping remained blistered, resulting in removal of a small portion of this covering outside of the secondary shield wall. This was removed in order to check t'ie pipe for a possible leak. Subsequent inspection found no such leak. The blistering was apparently due to moisture left in the pipe covering during installation. Dur-ation of this shutdown was 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 59 minutes.

(

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D

. st 7,

Shutdown tio. 13 12/18/60 - This shutdown of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 6 minutes resulted from a sudden loss of condenser n-()

vacuum following an excessive indicated shaft vibration at No. 1 turbine bearing.

r Load wts reduced ~from approximately 50 MW to 18 MW, and then increased to 40 MW gross electric. The vibration reappeared at No. 1 bearing resulting in a decision to shut down the unit. This was carried out in an orderly manner with a control i

room throttle valve trip being applied after generator load had reached zero.

Investigation of the turbine condenser system and the low pressure turbine casings revealed no serious leaks.

Shutdown No. 14 17/19/60 - This shutdown of 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months and 35 minutes was required in order to make the same modifications to the No. 2 turbine throttle valve as was done to No,1 valve.

The condition of the vai.ve was found to be similar to the No.1 valve.

As a result of further analysis of the shaf t vibration and the indi-cated movement of the turbine shaft within the bearings, a 65 MW gross electric limitation was pixed on the plant in order to insure proper lubricating oil flow -

into certain of the turbine bearings.

Analysis of ekcessivesturbine' shaft vip. M on Shutdown No. 15 1/2/61 p/

continued, resulting in a decision bein;

, (

made on December 22 to shut down the plant and make the necessary modifications to the high pressure turbine - section. Dur.ng this same period a permanent repair was made to the turbine throttle valves.

(See Section VI - Systems and Components, paragraph - Turbine-Generator)

Finalizing of the design for permanent additional shielding to be placed on top of the annulus between the reactor vessel and the inner cylindrical wall of the neutron shield tank during operation was accomplished. Fabrication

/

and installation of the metal covered masonite blocks, which make up this addi-tional shielding, was also completed.

l The control rod drive air cooling system was modified. This con-sisted of replacement of the transition ductwork between the supply duct and the control rod cooling housing, removal of the existing square distributor ducts between control rod mechanisms, and in' *11ation of orificed distribution ducts l

and filler pieces between control roc coils.

3 i

' (m)

In addition to the above, the following work was performed:

(1)

Installation of a ermanent cable tray for support c' the in-core instrumentati on s torage tubing. The storage tuoing has been provided as a replacement for the original storage drums, thus allowing "in,ine" retraction and storage of the withdrawn flux wires.

(2) Installation of equipment to provide for voltage reduction on the control rod stationary gripper coils.

(3) Installation of equipment to provide for loss of power channel power supply protection.

(4)

Installation of equipment to provide for further dropped rod protection.

The duration of this shutdowri was 338 hours0.00391 days 0.0939 hours 5.588624e-4 weeks 1.28609e-4 months and 18 +*nutes.

f3 V

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g, s

l 7.' 500' Hour Run af 392 MW Thermal '

h;4; The:500' hour'11 censed full { power level run at-392 MW thermal began at 3:25..P.M. on; January 17.and was completed at 12 noon on February 8,'1961.

iThisLoperation was carried out as' detailed:in Yankee's Provisional Licenserand-f paragraph D.6 jof the Technical Specifications.

~ The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months ' run resultec' in approximately 514 hours0.00595 days 0.143 hours 8.498677e-4 weeks 1.95577e-4 months of turbine-generator operation: sin'ce? loads other than full load level were experienced

during the period. Three shutdowns occtcred, two nf a planned nature (One for' Emergency Cooling by-Natural Circulation Test and one for a vapor container inspe: tion), ' the ' third being ' duet to an accidental t'urbine trip.- Operation of-tbs. iant'.during' the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months .' run at' the nominal 392 MW ' thermal rating produced 60,000,000 gross kilowatthours.

The reactor e'xhibited excellent ~ stability and good response and '

control characteristics as.well as smooth' trouble-free operation.during the operating period. Collection of and analysis ~ of in-core instrumentation data continuec' at regular intervals.throughout the period, with results showing actual coreethermal; performance. values well within the predicted values.

O-rerdine-9ener ter verrer nce w e e1s< exce11ent. The terdine sw rt inoic tee essentially vibration-free operation while the turbine control equipment con-tinued to exhibit excellent response with rapid predictable operation. The electrical transmission system experienced major' load disturbances due to a severe winter storm during which plant control systems behaved ~ in a most' excel--

lent manner.. (See-Section V. --Tests and Measurements,~ paragraph 3.)

~

h 8,_

Op< ratino Statistics The' initial power operation program which began on November 10, 1960

' I has resulted in Yankee producing 76,300,500 gross kilowatthours, or 67,904,745 net kilowatthours of power at the 115;KV bus through January 29, 1961.-

This energy was produced by 992 hours0.0115 days 0.276 hours 0.00164 weeks 3.77456e-4 months of turbine-generator operation. ' Power levels of operation for both reactor and electrical generator for the above period may be summarized as'follows:

Gross MWe Reactor MWt Hours Operation 0:

-20 0 - !81

'41 20 30 81 - 112 14

'30 - 40 112 - 141 45 40 - 50 141 - 168 60 50 - 60 168 - 198

'>2 60 - 70 198 - 227 4 37 70 - 80 227 - 258 46 80 - 90 258 - 288 29 90 - 100 288 - 321 51 100 - 110 321 - 354 3

110 - 120 354 392 272

. Initial criticality, low power physics testing and power operation testing through the reporting date, has resulted in Yankee's reactor being main-tained critical for.2042. hours. The primary systgm has been maintained at'its

- -Q normal' operating conditions of 2,000 psig and 514 -F for 2440 hours0.0282 days 0.678 hours 0.00403 weeks 9.2842e-4 months , this period A/

including, however,;precore-loading system operation.

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9.

Chemistry

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Fb Chemical evaluation of the plant systems has proceeded since February 1960. During the core loading operation, the primary system boron concentration was 1600 ppm, with a typical analysis showing a pH of 5.4, a conductivity of 9.8 micro mhos and chlorides of less than 0.1 ppm. From this peri'od, until boron removal, nickel concentration increased to 0.80 ppm while manganese reached a value of 0.50 ppm. This high manganese concentration was due to an incident involving the inadvertent injection of boric acid from a temporary injection system which contained carbon steel and its associated corrosion procucts.- After boric acid removal no corrosion products were found chemically within the limits of detection (approximately 0.1 ppm in most cases).

Reference primary ciolant con'ditions have been maintained, except for minor deviations, since the start of power operation. The pH values have fluctuated between 7.2 and 9.0, the higher pH condition existing after startups.

The primary coolant crud level has remained well within the 2 ppm specification, actual values varying between 0.1 to 0.5 ppm. Primary coolant radioactivity normal for opera Gross levels and nuclides present are considereg/c/ml to 8 x 10 gion to date.fc/mlduringop

~

f) specific activity has ranged from 2 x 10 V

tion to 120 MW gross electric. Specific activites of corrosion product nuclides are tabulated in Section V, paragraph 4.

The seconJary water system specifications have been maintained with-out difficulty with steam generator chemistry being controlled with a blowdown rate of approximately 0.1%.

No primary to secondary leakage has occurred to date in this system.

(See Sectior V. - Tests and Measurements, paragraph 4.)

nU 10.

Health Physics Gsma radiation levels throughout the plant have been well below design estimates. The highest radiation level encountered at equipment not in the vapor container was 1500 mr/hr, this being in contact with the inlet piping to the Purification Pumps. The highest radiation levels encountered in the vapor container shortly after reactor shutdown were in the loop cubicles. These

[I were found to reach about 150 mr/hr, except levels of 450-700 mr/hrwereexp/hr eri-enced on contact with main coolant piping. Radiation levels were below 2 mr in the radiochemistry Sartple Cubicle, except while obtaining main coolant crud samples at which time levels rose to 200 mr/hr. The radiation level over the ion exchange pit reached 20 mr/hr with no water in the pit.

In other generally accessible plant areas, gamma radiation levels have been close to background levels. Neutron flux levels outside the vapor container required installation of temporary ' shielding, which was later replaced with permanent shielding. (See Section V. - Tests and Measurements, paragraph 2.e.)

There have been only a few instances of contamination in the Poten-tiillv Contaminated Area and none in the Clean Area. Highest level showed dpm at a primary system leak area and a contamination of 10--40,000 160,C s

dpm/sq.ft. was found in the working area during the leak repair. Other local-ized cases of contamination were less than 500 dpm/sq.ft. in working areas.

Nearly all beta-gamma radiation exposures were at the lower limit A

of detectability of film badges. The highest monthly exposure was 80 mr and Li the highest accumulated exposure July - December, 1960 was 150 mr.

7 e.

11, i

i A total of 2465,/ c of activity in excess of stream background was

T released during February 1960 - January 29, 1961.. All releases, after dilu-

\\/

tico, were well below the MPC for a mixture of unidentified isotopes.

Thirteen 55 gallon drums'containing 0.7,j 8 per drum were shipped 4

off-site for disposal. At the conclusion of a study of gaseous iodine carry-over in the waste disposal evaporetor, ten 55 gallon drums containing evaporator bottoms mixed with concrete and each containing approximately 0.15 mc of I-131 were also shipped to disposal. During the period December 1, 1960 through 606/ c of radioactive gas was released in 7945 cubic feet F

January 29, 1961 of gas from waste disposal. Release concentrations were well below MPC. (See Section V. - Tests and Measurements paragraph 5.)

11.

Systems and Components The newly developed plant systems and components performed well when considering their novel features of design. The.more conventional features, however, of a number of components, such as gasket closures and electrical coils, did de"S op trouble, resulting in program delays. A brief summary of the instal-gS

(_" )

lation ano evaluation of a number of systems and their components is given in Section VI.

Systems and Components.

Included are summaries of the following:

In-Core Instrumentation System Main Coolant System pressure Control and Relief System Charging and Volume Control System Chemical Shutdown System 7\\

(,)

Purification System Component Cooling System

~

Primary Plant Corrosion Control System Primary Plant Sampling System Waste Disposal System Shutdown Cooling System Primary Plant Vent and Drain System Safety Injection System

(_,)

Reactor Control System Nuclear Instrumentation and Reactor Control System

'~-

Radiation Monitoring System Vapor Container Atmosphere Control Systems Fuel Handling System Main and Auxiliary Steam System Condensate and Feedwater System Circulating Water System Compressed Air Systems Electrical System Reactor Vessel Radiation Shielding Turbine-Generator 12.

Desian Chancos

(~)

The terms and conditions of Yankee's Provisional License and re-(_/

lated Technical Specifications allow certain design changes to be made in the facility without the filing of a license amendment. A Summary of Design Changes may be found in Section VII which includes the changes made during this

.m 3_

, 7 7, -

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T12.

L.'

s areporting ' period. The summary is arranged -in two. classifications:

A. - Changes l T(N.

in the Secon'dary Plant,' and.B.

. Changes Considered :to Af fect the Primary Plant.

M Classification. A' lists five design: changes'while Classification B details 22

(

such changes.-

c l'3. Operatino procedure Chances h

The. preparation and filing of Operating, Emergency and Maintenance L

Instructior,. in' Volume II - Part B of Yankee's. License ' Application, was intended :

to show tha,. operation and maintenance; problems had been examined for general workability.;.This. effort was~of intermediate detail and was intended to act as a guide to"plantioperation.

Initial plant operation revealed thatycertain.

revisions in procedures were necessary in' order to correct for page to page incon-sistencies' and to present; further: detai1~ or clarificatiion of 'a particular opera-

-tion. The preparation and issuance of certain procedure _ changes has, therefore, l-continued on. a day-to-day basis. These ch6nges have been made only to;the de-l tailed step-by-step Instructional Section'(Section V) of the Operating Procedures.-

' Issuance _of such changes has resulted in the compilation of a Plant Operating l

Manual which serves to keep the plant operation on a safe and current status.

In this connection,.all the Objectives, Conditions, and Precautions set forth in

~

j i

the' license application have been observed and, in fact, additional Conditions and Precautions have been added in the detailed revised Instructional. Sections

'of certain Operating Procedures.

l4.

Plutonium Buildup Experimental procedure

-A test procedure (Test Procedure 509F1) for measuring the power coefficient during core life has been prepared and may be found in Section VIII-of this report. This procedure is basically the same as,Startup Procedure 503F1 as filed in Volume II - Part B. of the license application ^.

This. procedure will be used to measure the power coefficient at intervals.not exceeding 2,000 equivalent full power hours which represents gen-eration of approximately 240,000 megawatt hours of electricity. The p, tdure-

~%

consists-of measuring'the reactivity change corresponding to a power level by:

- (O (1) Compensating for-the reactivity change by allowing the average main coolant' temperature to-change and (2) Compensating for'the 'eactivity change by. moving control rods.

The reactivity change is calculated using !.he equilibrium change in average main coolant temperature obtained in step (1) with an experimentally.

measured value'of the moderator temperature coefficient. The reactivity change is also calculated using the change in equilibrium rod position obtained in step;(2) with experimentally measured control rod worth data. When. required, these' reactivity changes are corrected for variations in xenon poisoning, main coolant-pressure changes and small load fluctuations.

The change'in' reactor power is obtained from the change in gross electric generation-using-experimentally obtained calc,rimetric data. The power coefficient ~1s then obtained by dividing-the reactivity change'by the corres-

.ponding change in reactor power level. This procedure for measuring the power I

.~

coefficient'was' adopted because:it provides the best way of controlling the many

- variables' associated.with this measurement, because it does not violate plant operational limitations and because it corforms closely with the normal opera-ting procedures used by.the: operators and hence is well understood by them.-

(See Section VIII.' ; Plutonium' Buildup Experimental Procedure.)

I'

#l L

^

.2.

p 2

The test procedure (Test Procedure 509El) for measuring the mod--

(

)

erator temperature coeffic ant during core life has also been prepared and may be found in Section VIII of this report. This procedure is very similar to Startup Procedures 503E5 and 503E8 used during initial core testing.

~

Basically, the modarator temperature coefficient measurement consists of varying the average main coolant temperature while observing the corresponding change in the reactor startup rato measured at low power levels. The. change in startup rate is converted to a reactivity change and this is divided by the change in moderator temperature to obtain the moderator temperature coefficient.

The procedure is modified slightly from those used during initial testing to simplify the experimental instructions now that the operators are familiar with the basic technique involved and to improve the accuracy and reli-ability of the data obtained.

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-III. : REACTOR CORE LOADING AND ASSEMBLY 7;

S' 10n date of' Interim. License' issue, July 9, 1960, reactor vessel internals required ~for core' loading had been assembled, cleaned.and' installed

. within the reactor vessel.. During.the following two days, the: 52 studs were -

~

removed from the vessel flange. On July 12 the. vessel was filled to core load -

-3ng level with 1670 ppm borated water and the. temporary nuclear instrumentation readout equipment was installed in the vapor container. Accumulation of bcck-

_L ground count data was also begun. July 13 saw the special BF neutron detectors 3

' installed in their portable. thimbles and the thimbles placed in their planned locations. The first-of four of the ~Po-Be sources was also assembled to a source vane.and placed in source position No. 1..The next oay the initia.1. core loading check-off list 503E-1.was initiated. 'The temporary. nuclear instrumentation indi-cated, however, a base count rate too low to be entirely satisfactory. Modifica-

~

y 4

tions were made to detector thimbles-and a higher, acceptable base count was then achieved.

On July 15 actual core lading began with the placing of the first-

.' shim rod and follower.. Recording of hourly data on flow through the shut _ down

.p-V:

cooling system and temperature and boron concentration of the main coolant were started.- h i..a 4 an operational test of the temporary boric acid injection system was performed, this system providing for protection

ring the core' loading procedure. This was followed with the insertion of two control rods and the first

~ fuel assembly. Revision _of the established' core loading sequence was' required j-due to the necessary re-positioning of the installed source vane from Position No.'l-to No.~4.. Source vane re-positioning was require'd to accomplish the correct location of the larger depth dimensioned. vane in its proper core position.

L:.O.

' Core loading progressed slowly due to various occurrences.' Major j.

among these was that of. loading the first few assemblies into the core. A license referenced loading-precaution ~ stated that, "if at any time during core loading, e

the extrapolated value for' critical size of the core is less than twice'the number.

of fuel assemblies in the core, the 1oading operation mustMe suspended."- In order that this precaution be met, the originally conceived slab loading plan required revision, since.large increases in count rates occurred as fuel'assem -

blies were added.. It was later concluded that the increased counts were due' to -

+'

a geometric effect between the source, the fuel and the temporary nuclear de_tec-tors,'and this effect masked the true neutron multiplication due to the addition of fuel. Selective positioning of' temporary instrumentation and review and re-vision of the loading sequence minimized these geometric effects and permitted core loading to proceed-in compliance with all' license requirements. As a re-I sult of.the'above changes in: loading procedure, the total. core loading time was a-approximately doubled. ~

~

i LAn.other important area causing delay was malfunctioning.of the fuel l

handling system equipment. The resulting delay due to this equipment consumed nearly 16% of the total 270.. hour core loading time.. (See Section VI. - Systems j

and Components) ' Electrical interference on the temporary nuclear instrumenta-j-

' tion readout equipment also caused delay. Contactors controlling operation of the fue1~ chute-carriage induced high noise leve1s in the counting channels-so

~

3-that its operation had -to be restricted 6 times when counting was not in progress.

i

The restrictions thus placed on movement of fuel chute -carriage delayed the load-ing process.'

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C 15.

On July 26 core ],ading was completed with the installation of the 76th fuel assembly and the.three remaining Po-Be sources. The following day saw the borated reactor water diluted to 1600 ppm and an extensive check-out egun of the normal plant source range instrumentation. Certain modifications S

in this equipment were found to be necessary. These chan;;: were made result-ing in satisfactory count levels being obtained.

(See Section VI - Systems and Compaents)

On July 30 the temporary boric acid injection system was dismantled, the in-core instrumentation structure was me" into the vapor contained and the upper core support plate and barrel assemblj were installed in the reactor.

During the following two days, the in-core instrumentation structure was placed in the rea cor and the flux wire thimbles and thermocouples inserted into their respective positions. Assembly of reactor internals continued with installation of the 24 control rod drive shaf ts; slight adjustment of ceveral coupling iingers was required in order to allow proper joining cf the dri,e shafts to the control rods.

By August 3 installation cf shim port plugs, guide tube holdown plate

~. ring, and new "O" ring head gaskets had been completed allowing installation of the reactor vessel head. Upon lowering the head onto the vessel, interference developed between two of the in-core instrumentation thermocouple penetrations columns and the vessel head ports through which they had to pass.

(See Section VI - Systems and Components)

As a result, it became necessary to disassemble the upper reactor internals to correct minor damage to the in-core instrumenta-tion structure. Reassembly of all reactor internals was complete by August 8.

During the period from completion of core loading to August 9 work i

was carried forward on eliminating noise from the normal plant so'urce range instrumentation channels.

(See Section VI - Systems and Components)

This work resulted in virtual elimination of the noise problem and a very satisfactory indication of source level counts.

The reactor vessel head was installed on August 10, with the $2 vessel head studs being installed and tightened by the 13th. Connection of in-core instrumentation flux wire thimbles and control rod drive power and position indicating cables occurred the following day. It was also found neces-sary to take further steps in order to eliminate noise in the source range in-strumentation, caused by operation of the control rod drive contactors and flux wire drive control equipment. The initial core loading and reactor assembly was thus complete on August 14.

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_

ML19351E019

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