Text

Operation Rept 1,Initial Startup & Test Operations of Yankee Reactor for Period 600709-610129
	ML20031C314
Person / Time
Site:	Yankee Rowe
Issue date:	02/13/1961
From:	Coe R; YANKEE ATOMIC ELECTRIC CO.
To:	;
References
	NUDOCS 8110070049
	Download: ML20031C314 (59)
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'd OPERATION REPOPT NO.1 l

I Initial Startup and Test Operations of the Yankee Reactor s:

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July 9,1%0 -- January 29,1%1 l

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t Submitted in accordance with Facility License OFR-3, as amended, d'

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Attention: Division of Licensing and Regulation m

Oear Sirs:

In accordance with Facility Lfconse No. DFR-3, as amended, n

we are sutenitting 43 copies of our Operations Report No. I covering Initial Startup and Test Operations of the Yankee reactcr for the

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period July 9th,1960 to January 29th,1%1.

Respectfully submitted, 6

YANKEE ATailC ELECTRIC COEPANY i

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I n t rod u c t i on ----------------------------------

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S umma ry ---- --------------- ---- --------- -------

2 - 13 Q

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Pe riod and Act ivitie s ---------------------

2 N

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

2 4;

3.

Cold and Hot Control Rod Scras Tests ---------

2 44

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In it ia l C riti ca l ity ----------------------

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Low Power Phys ic s Testing -------------------

3 6.

Power Opera tion Te s ting --------------------

3 ---8 kB mo 7.

500 Hour Test Run at 392 W Thers.a1 ---------

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Ope ra t ing Sta ti s t i c s -----------------------

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C h e mi s try ---------------------------------

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10. He a l th Phy s i c s ----------------------------

10 W

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11. Sy stems and Components -----------------------

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12. Decign Changes --------------------------------

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13. Opera ting Procedure Change s -------------------

12 14 Plutonium Buildup Experimental Procedure ------

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Rea ctor Core Loading and As sembly --------------------

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IV.

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

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Te s ts a nd W ea suremen t s ------------------------------

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

17 3

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Purpos e a nd Scope ----------------------- -

17 s

b.

Control Rod Drive and Plant Scram Tests ---

17 M

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c.

Control Rod and Boron Worth Determinatiuns at Low and Operating Temperatures --------

18 he d.

Temperature, Pressure and Flow

>i Coe f ficient Detentina tions ----------------

18 e.

Main Coolant System Heating Rate e2 De t ermina t i on s -------------------------

19 f.

Nuclear Instrumentation Response to Asymmetric Control Rod Positioning --------

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Tcsts cnd Nasurements (cont'd)

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In : Mal Power Opera tion Tests ----------------

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Purpese and Scope ------------- - --

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Power Coefficient and Loss of Load y

Trs n s i en t Te s t --- -----------------------

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Nuclear instrumentation Power Calibration -

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Reactivity vs Fission Product Level h

Following Power Level Changes -------------

21 g;3 e.

Biological Shielding Effectiveness Test ---

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. m f3 Instrumentation and Control Response ----

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Emergency Cooling by Natural Circulation --

22 3.

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

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C h em i s try ------------- ----------------------

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Prima ry Cool an t ---------------------------

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Wa te r R ad ioa c t ivi ty ----------------------

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W a s t e Di spo s a l ---------------------------

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Seconda. y Dater Chemistry ---------------

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He a l & Phys i c s -----------------------------

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'N Radiation Levels id Mr Contamination Levels

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Waste Disposal: Liquid, Solid, Gaseous ----

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Sys tems and Componen ts ------------------------------

30--47

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In-Cor c Instrumentation System --------------

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M a i n Cool a n t Sy s t em ---------------------------

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Pressure Control and Relief System ----------

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

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

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Puri f i ca t ion Sys tem ------------------------

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Component Cool ing Sys tem ---------------------

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

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

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

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

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12. Primary Plant Vent and Drain System ----------

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a SECTICW PAT

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Systems cnd Components (cont'd)

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13. Safety Injection System -------

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14. Reactor Control System -----

15. Nuclear Instrumentation and Reactor h'

39 Protection Syster ---------

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16. Radiation Monitoring System ---

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17. Vapor Container Atmosphere Control Systems ----

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18. Fuel Handling System ------ -

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19. Main and Auxiliary Steam Systems

20. Condensate and Feedwater System ---------------

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21. Circulating Water System ------

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22. Crapressed Air System ---

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23. Electrical System -----

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24. R e a c tor Ve s s el -------------

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25. Radlation Shielding ----

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26. Turb i n e Ge n e ra tor ---------------

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D e s i gn C ha ng e s --------------------------- --------

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VIII.

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

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INTRCDlCTION k

H siix all This report is submitted in compliance with paragrap' 3.C.(2)

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e cf Facility License No. DPR-3, as amende' l

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Yankee Atoric Electric Company's power station at Rowe, Massa-lf chusetts, has been engaged in initial core loading and startup operations I'A f r a period beginning July 9,1960, and ending January 29,1%1, six months r

citer issuance of the Provisional License.

Thic operation was carried out

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y as authorized by Interim Facility Licente No. DPR-3, dated July 9,1960,

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i cnd by Amendment No. I to that license, dated July 29,1%0. Amendment No.1

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cllowed reactor operation at pwer levels not to exceed 392 megawatts thermal.

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The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at 392 m thermal began at 3:25 P.M. on January

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17,1%1 and, at the end of the six months reporting period, midnight January i

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

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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, t

I 1%1, 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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II. E N ARY 1

period and Activitics

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O This report covers the six month period from date of Provisional License issuance, July 29,1%0 to January 29,1%1. During the period, core loading and attendant reactor assembly, initial criticality, low-power physics testing and the initial power operation and testing were cocpleted. The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at 392 F thermal was begun on January 17, just prior to the expira-tion of the six month reporting period, and completed on February 6. This

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operation is reported in Section V. - Tests and Measurements, paragraph 3.'

2.

Reactor Core Loadino and Assembly Previous to July 9,1%0, the reactor vessel internals required for core loading i >d been assembled, cleaned and installed.n the reactor vessel.

k The first of the 1.ur 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 e

instrumentation difficulties and an operating license restriction. Loading of 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)

I 3.

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

Electrical grounds appeared on one or more 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 resa.inged onto control rod mechanisms which had formerly held grounded coils, allcwing completion of the full withdrawal and drop tests of 23 out oT 24 control rods.

Investigation of th( inoperative 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-y 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.

f 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 e

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mechanisms. All the total rod drop times, including release time, drop time l

and dashpot closure time were measured within the specified two seconds.

3 Variations in drop times for the same mechanism and between different mechan-f 1

isms were small and could be explained by slight dimensional variations and differing temperature, pressure and flow conditions.

(See Section V. - Tests and Measurements, paragraph 1.b.)

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4-Initial Criticalit; 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 avas carried out in accordance with written procedu.es. The low-power physics testing program, as outlined in Section SOSE of the license application, thus began in earnest on August 19 at 8:19 P.E.

(See Section IV. - Initial Criticality).

t 5.

Lev Power Physict Testinc Although the initial startup program was ready to proceed with the low power physics testing phase, it was not until September 4 that tests other than cold control rod drop tests were begun. During the low power tests, the reactor was operated at pwer levels not in excess of 5 megawatts themal.

The period of September 4 to 16 saw collection of data at low reactor tempera-tures, free which the reactivity worths of control rods and boric acid were j

calculated. Temperature, pressure and flow coefficients with increasing reactor temperature were determined, and Main Coolant System Heating Rate Deteminations were made frat September 17-22.

Tests of Control Rod and Boron North at Opera-ting Temperature and Instrumentation Response to Atyxetric 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 power testing program not being completed until November 8,1%0.

Collection of supplemental data for several of the tests previously perfomed, 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-ficulths 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 1

the stea generators.

(See Section VI. - Systems and Components)

Redu: tion of the data produced during this phase of startup shows that values for Control Rod and Soron Wcrth werc in good agreement with the calculated values.

The measured temperature and pressure ccefficient values,

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with nc boron in the main coolant, were also approximately as preducted, whereas N

coefficients measured with high boron concentrations were more negative than y

analytical work predictec.

There were no detectable flow coef ficients at low ly reactor powers. Results of Main Coolant Heat Rate Deteminations showp that l,

with four main coolant pumps operating, the heatup rate ranged from,19 F per s,

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 asymetric control rod positioning showed that detection of resulting flux tilts was possible. Further i

testing, however, during power operation showed that nuclear instrumentation was clso sensitive to variations in coolant temperature and xenon condition in the

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core. As a result, those additional factors complicate absolute calibration of the nuclear instrumentation for flux tilt datection.

(See Section V. Tests and u(

Measurements, paragraphs 1.c, d, e and f) 9 6.

Power Operation Testinc r.'.

Initial power operation and testing began November 10,1%0 at 9

2:37 A.M. when the turbine-generator was phased with the New England inter-3 connected transmission system. Plant operation and performance of tests during 3

this phase of startup followed, in general, the schedule outlined in Section Mi 503F of the license a 3 -

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The perfctmanco cf th;s Ests showed that th2 recetcr pinnt cs cxtremely st ble under all operating cnd tr:nsicnt crnditions (120 MX gross clcctric, nominal). Limiting operational conditions, shown by transient tests,

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were reached in the secondary plant before they were reached in the reactor plant. Power coe7ticient measurements taken during th uleg generator load changes show a value of -3.3 x 10-g majority of the sc dk/k/MW themal 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 MZ gross electric plant loads. No plant limitations U

were exceeded during these tests. Collection of data, showing relationship betwec. pimary and secondary systems, showed excellent agreement with heat balance data.

It has been detemined that measurement of average reactor output is possible to approximately iSL 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 coefficieat during the load reduction. The calculated values of reactivity loss due to xenon are in reasonable aoreement with experimentally detemined values. Performance of Biological Shield Effectiveness tests at 15, 30 and 60

gross electric, revealed higher than anticipated neutron levels throughout the site. Neutron levels at 60 MW gross electric load were reduced by t

a factor of 8 with installation of temporary shielding over the top of th( annulus between the reactor vessel and the neutror shield tank. Later installation of 2

pemanent masonite shielding showed that neutron levels were reduced by a factor of 19 over the unshielded condition. The highest neutron level at 120 W gross electricis7.0 mrem /hr. Plant control was found to be extremely stable; the inherent negative temperature coefficient proving to be an effective stabilizing influence.

It wi3 also found that the reactor regulated on temperature control with no rod motion required for small load variations. The automatic rod con-a trol systet has proven to be more stable than anticipated, holding main coolant h

temperature within the control band during normal generator load changes and steady power operation. The tests of emergency cooling by natural circulation P

were perfomed from 60 and 120 VW gross electric. These showed that adequate cooling was provided to the core under conditions simulated in the test.

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

This period included 15 plant shutdowns; two for reasons c' scheduled nuclear data collection and i3 due to turbine-generator testing and operation 4"

difficulties. A summary and chronology of the shutdowns follow, with further 1

information on systems and component modification performed appearing in Sec-p tion VI - Systems and Components.

L 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 k

duration occurred at the conclusion of b:

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 n

tests. Upon manually tripping the unit, high moisture separator levels occurred.

K The turbine overspeed trip tests were also perfomed during this period.

Y 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 l'

casing temperature in addition to continued high water level problems in the turbine moisture separators. During the shutdown period, a one inch vent line S

was installed from each moisture separator to the feedwater heater drain re-M ceiver. Installation of these lines was made in an effort to eliminate the 7

hi@ meisture seperator levels before proceeding to the scheduled 50 hour5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months -

OfL _30 m o.rou. electric ocm run-- - -

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F Shutdown No. 3 11/11/60 This scheduled shutdown cf 110 hours0.00127 days 0.0306 hours 1.818783e-4 weeks 4.1855e-5 months v-dureticn occurred et the conclusion k.

of the 30 m gross electric run. Turbine-generator shutdown was accomplished 1

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by the Loss of 1.oad Transient Tcst detailed in Section 503F1 of the lhense

,7 application. During the transient test, high turbine moisture separator levels y

again anpeared. Accordingly, the previously installed 1" moisture separator-Q drainreceiverventlinewasremovedandreplacedwithaIf"ventline. The f,rst of the several 503F6 tests, Variation of Reactivity Due to Change in E

Fission Product Level Following Reactor Shutdown, was also performed.

E Shutdcwn No. 4 11/16/60 This shutdown period of 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months and

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24 minutes was the result of an inability to hold a constant load on the turbine-generator unit. Shutdown was orderly,-

lJ with later investigation showing that No.1 Turbine Control Valve Servo-motor

,7 operation was erratic. The shutdown period was, therefore, consumed with fab-4 rication and installation of the necessary new parts for this piece of equipment.

h Shutdown No. 5 11/22/60 This I hour and 39 minute shutdown

'y occurred during the scheduled 70 hour8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months -

B 60 W gross electric power r n.

This shutdown was initiated by a turbine trip

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due to a high moisture separator water level indication. A series of coinci-i dences involving the feedwater heater drain system operation and a critical y{

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

y Shutdown No. 6 11/23/60 This schedulad shutdown period, at the M,

conclusion of the 60 W 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 t low flow reactor scra.t and turbine trip test, accomolished by tripping one gl>

of 'bree operating main coolant pumps. Startup Procedure 503F7, Emergency T

Cooling by Natural Circulation, followed the transient test. Meawrement of xenon decay, cs detailed in 503F6, took place throughout the remainder of the 7

shutdown period.

s.f During the 60 m gross electric power operation, Biological Shield K

Effectiveness Tests continued to be made as called for by Startup Procedure h

503F4. These tests disclosed that in certain plar.t areas, fast neutron radia-Ji tion was at levels in excess of calculated values, although still well below acceptable levels. Durinc the shutdown period, the reactor was operated at a

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low thermal power level while detailed surveys were performed. Fast na. tron Q

streaming was found to exist in an upward direction from the annulus be. ween M

the reactor vessel and the inner cylindrical wall of the neutron shield tank.

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The situation was remedied during the shutdowa period by placing temporary M

shielding in this area in the form of 46 water-filled drums arranged in a D

double layer above the offending annulus.

k 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 g

occurreo only one-hs1f hour after the 4

start of power generation. Shutdown was initiated by the primary system flow g

instrumentation. Rearrangement of loop flow instrumentation components, re-sulting from a loop flow instrument being out of service for maintenance pur-4 poses, resulted in this spurious reactor scram and turbine trip. Completion of flow instrumentation checking and maintenance was thus completed during the shutdown peri (d before proceeding to the next power level of operation at 90 W s

gross electric.

3 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

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vibration indications at No. I turbine bearing. Pref.ous to the vibration trouble, generator loading had been help at approxiwately 90 W gross electric 3a_ - _ ______~-.- -

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fcr some 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months . In cn offcrt to d2t:rmina th2 cause cf the vibratioc, vcrious alsctrical loads wer2 pieced on the unit whilo changing lubricating W

bb cil temperature. N3 r;peating conditions could be found betwe:n thes2 fcetors cnd the turbine vibration.

In an attempt to correct this difficulty it was M

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decided.o raise the turbine high pressure casing slightly, by the addition of a 7 til 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 An previously installed shims froc each of the four corners of the turbine high M

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 O

mada during this shutdown since a major inspection was planned for the follow-ing day.

rite S5utdown 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 butage.. was y

th2 r2sult of continued shaf t vibration. Various analyses indicated that the h

solution to the problem in all probability lay in a bearing modification. As bf a result, turbine bearings Nos.1, 2 and 3 were removed from the unit and the W

n cos::~. machining operations performed. No. 4 bearing was also removed, in-k

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

In addition to th2 above bearing modification, clearances on turbine oil and steam seals c2ro increased.

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Shutdown No. 11 12/13/60 - This 15 minute separation from the elec-trical transmission system was accident-4 ally caused by the turbine manufacturer's representative while testing the over-speed trip test oil pressure. During the test, the turbine shaf t raowd a suf ficient iQ c=unt in an axial direction to cause a high thrust bearing pressure to be devel-fi cped, resulting in the turbine trip.

The previous power run also proved that the bearing modifications b

h:d not solved the vibration difficulty.

In the continuing search tir the answer, h

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

Q Shutdown No. 3.2 12/15/60 - While investigations of control valve h

opening sequence and their effect on p

shaf t vibrations at various loads was being explored, a sudden load drop from Q

60 to 10 m gross electric was experienced. This occurred during the adjust-ment of the No. I cor. trol valve gag, resulting in the ' arbine manuf acturer's

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representative manually tripping the turbine. The No. I turbine throttle valve,

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howsvar, did not go fully closed and thus did not initiate a reactor scram. The 6

latter operation was perfortned 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, p.

[Q repair and assembly of the faulty No.1 throttle valve. Upon disassembly it

%y cas discovered that relative motion between the valve disc and stem was restricted l

preventing proper poppet valve operation. The temporary repair of this valve t

allowed further turbine test work to proceed.

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During a vapor container inspection tour, a leak was discovered at q

l cn instrument nozzle in the shell of the No. 1 Steam-generator. This leak dic-7 l tated that No. I loop be isolated and drained. A load limit of 70 E gross olectric was imposed upon the plant during the resulting 3 loop operation while L

l cxperimental data for this mode of operation was obtained.

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Shutdown No. 13 12/18/60 - This shutdown cf 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 6 cinut s Qg r2sulted from a sudden Icss cf conden:cr sq 1,acuum fallowing cn cxr.sssiva indicated sh2f t vibration at N). I turbine bearing.

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1.ccd ca: reduced from approximately 50 MH to 18 LM, and then increased to 40 ist gross electric. The vibration reappeared at No. I bearing resulting in a decision k

to shut down the unit. This was carried out in an orderly manner with a control h

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

@4 Inv;stigation of the turbine condenser system and the low pressure turbine casings 3

i rev:aled no serious leaks.

j y Shutdown No. Q 12/19/60 - This shutdown of 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months and 35 minutes I4 @

modificStions to the No. 2 turbine throttle valve as uas done to No. I valve.

' b was required in order to make the same hi f[Q The condition of the valve was found t be similar to the No.1 valve.

N As a result of further nalysis of the shaf t vibration and the indi-cated movement of the turbine shaft withir the bearings, a 65 MW gross electric

.'f licitation was placed on the plant in ord.er to insure proper lubricating oil flow yn into csrtain of the turbine bearings.

, g Shutdown No. 15 1/2/61 Analysis of ekcessiveiturbine' shaft vibration continued, resulting in a decision being

Qjh, made cn December 22 to shut down the plant and make the necessary modifications to tha high pressure turbine section. During this same period a permanent repair ML was mad 2 to the turbine throttle valves.

(See Section VI - Systems and Components, Iparagraph - Turbine-Generator) k.

Finalizing of the design for permanent additional shielding to be 31 aced on top of the annulus between the reactor vessel and the inner cylindrical h

wall of the neutron shield tank during operation was accomplished. Fabrication g

end installation of the metal cravered masonite blocks, which make up this addi-g Gienal shielding, was also completed.

g up The control rod drive air cooling system was modified. This con-b h.:i 31sted of replacement of the transition ductwork between the supply duct and 2h2 control rod cooling housing, removal of the existing square distributor ducts

[j betcnn control rod mechanisms, and installation of orificed distribution ducts g

and filler pieces between control rod coils.

d C3 In addition to the above, the following work was perform.ed:

4 (1)

Installation of a pemanent cable tray for support of the in-Q l

core instrumentation storage tubing. The storage tubing has Q

h[:\\

been provided as a replacement for the original storage drums, thus allowing "in-line" retraction and storage of the withdrawn d

flux wires.

{k E

(2)

Installation of equipment to provide for voltage reduction on l

the control rod stationary gripper coils.

h (3)

Installation of equipment to provide for loss of power channel p@

g power supply protection.

g?

(4)

Inst:f t1ation of equipment to provide for further dropped rod protection.

g frhe duration of this shutdown was 338 hours0.00391 days 0.0939 hours 5.588624e-4 weeks 1.28609e-4 months and 10 minutes.

Es lAl lim,

.-_.L n.

h in Three snutdowns cccurred during the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run ct the 392 m th;r.nal pow 2r lev 2l.

Tuo of these wera cf a plcnned nature, while the third w

was an unscheduled turbine trip. A sumary and chronology of these shutdowns follow:

/g Shutdown No. 16 1/22/R - This scheduled shutdown of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months N

and 22 minutes duration occurred in 9

crder to perform Startup Procedure 503F7 - Emergency Cooling by Natural Circu-lation at the 392 E thermal level. Inspection of the vapor container coulp-

[N ment revealed thermal insulation blistering on the No. 2 steam-generator outlet M

steam piping. It was decided to return the plant to full load operation, with Y

another inspectior scheduled in approximately one week.

Id e

Shutdown No. 17 1/25/61 This shutdown occurred as a result of Q

construction forces accidentally con-A tacing a moisture seperator high water level switch. Closing of this switch Q

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.

-h; Shutdown jio. 1_8 1/28/61

[{

The plant was shut down in order to maxe a scheduled vapor container in-j spection. Thermal insulation on No. 2 steam-generator outlet piping remained y,

blistered, resulting in removal of a small pcrtion of this covering outside of W

the secondary shield wall. This was removed in order to check the pipe for a 6

possible leak. Subsequent inspection found no such leak. The blistering was K

apparently due to moisture left in the pipe covering during installation. Dur-y.

ation of this shutdown was I hour and 59 minutes.

n

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l; h;i.

q, 7-500 Hour Run at 392 VK Th'rmal

,y The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months licensed full power level run at 392

thermal began N

at 3:25 P.R. on January 17 and was cocpleted at 12 noon on February 8,1%1.

This operation was carried out as detailed in Yankee's Provisional License and M

paragraph D.6. of the Technical Specifications.

M,

'h The 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run resulted in approximately 514 hours0.00595 days 0.143 hours 8.498677e-4 weeks 1.95577e-4 months of turbine-A

q gen: rator operation since loads other than full load level were experienced pp during the pericd. Three shutdowns occurred, two of a planned nature (One Vd fcr Emergency Cooling by Natural Circulation Test and one for a vapor container ly inspection), the third being due to an accidental turbine trip. Operation of the plant during the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run at the nominal 392 W therrnal rating produced

!h 60,000,000 gross kilowatthours.

jih

.p The reactor exhibited excellent stability and good response an:t qM control characteristics as well as smooth trouble-free operation during the f &

operating period. Collection of and analysis of in-core instrumentation data N

continued at regular intervals throuchout the period, with results showing isp actuel core thermal performance values well within the predicted values.

T;W Turbine-generator performance was also er:ellent. The turbine shaft indicated jN css;ntially vibration-free operation while the turbine control equipment con-

$r tinued to exhibit excellent response with rapid predictable operation. The

. D) cisctrical transmission system experienced major load disturbances due to a

$1 s;vrre winter storm during which plant control systems behaved in a most excel-65 Icnt manner.

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

g

..n 8-Operatino Statistics M-q g[

The initzal power operation program which began on November 10,1%0 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 29,1%1.

This "i

cn2rgy was produced by 992 hours0.0115 days 0.276 hours 0.00164 weeks 3.77456e-4 months of turbine-generator operation. Power levels cf operation for both reactor and electrical generator for the above period may K

be summarized as follows:

g je Gross We Reactor Wt H_oltrs Operation p

0-20 0

'81

'41 kII 20 - 30 81 - 112 14 I %

40 - 50 141 - 168 60

' f3 30 - 40 112 - 141 45 50 - 60 168 - 198 52 ETF 60 - 70 198 - 227 (37 f5 70 - 80 227 - 258 45 D!

80 - 00 258 - 208 29 D

90 - 100 288 - 321 51 W

100 - 110 321 - 354 3

Ifi l

110 - 120 354 - 392 272

%k Initia? criticality, low power physics testing anci power operation k

testing through the reporting date, has resulted in Yankee's reactor being main-Jq tained. critical for 2042 hours0.0236 days 0.567 hours 0.00338 weeks 7.76981e-4 months . The primary systgm has been maintained at its pg n:rmal 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 p'

including, however, pre ure loading system operation.

' l:,i l

P

9-Chemistn p

N' Chemical evaluation of the plcnt systems has procreded sinco Fchruary 1%0. During the core loading operation, the priman system boron concentration 7

was 1600 ppc, with a typical analysis showing a pH of 5.4, a conductivity of 9.8 gd cicro ahos and chlorides of less than 01 ppe. From this perfod, until boron removal, nickel concentration increased to 0.80 ppm while manganese reached a g

value of 0.50 pps. This high manganese concentration was due to an incident it involving the inadvertent injection of boric acid from a tempora n injection jg system which contained carbon steel and its associated corrosion products. After boric acid removal no corrosion products were found chemically within the limits

%g of detection (approximately 0.1 ppn in most cases).

Reference primary coolant conditions have been maintained, except df fer 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.

'l The primary coolant crud level has remained well within the 2 ppe specification, x -

cctual values varying between 0.1 to 0.5 p;n. Primary coolant radioactivity q

Icvels and nuclides present are considerg/t/ml to 8 x 10-g/t/c1 during opera-normal for opera ion to date. Gross IT specific activity has ranged from 2 x 10 M

tion to 120 ME gross electric. Specific activites of corrosion product nuclides D

ore tabulated in Section V, paragraph 4.

M' M)

The secondary water system specifications have been maintained with-N out difficulty with steam generator chemistry being controlled with s blowdown f

rate of approximately 0.1%. No primary to secondary leakage has occurred to y

date in this system.

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

!U$

10.

Health Physics p

Gama radiation levels throughout the plant have been well below d2 sign estimates. The highest radiation level encountered at equipment not in 4'

thevaporcontainerwas1500mr/hr,thisbeingincontactwiththeinletpiping n

to the Purification Pumps. The highest radiation levels encountered in the vapor container shortly af ter reactor shutdown were in the loop cubicles. These y

care found to reach about 150 mr/hr, except levels of 450-700 mr/h;wereexp/hr eri-f(p cnced on contact with main coolant piping. Radiation levels were belorr 2 mr in the radiochemistry Sample Cubicle, except while obtaining main coolant crud j

samples at which time levels rose to 200 mr/hr. The radiation level over the ion exchange pit mached 20 mr/hr with no water in the pit.

In other generally eccessible plant areas, gama radiation levels have been close to background IcVels. Neutron flux levels gutside 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-tially Contaminated Area and nor.e in the Clean Area. Highest level showed 160,000 dpc at a primary system leak area and a contamination of 10--40,000 dpm/sq.f t. was found in the working area during the leak repair. Other local-ized cases of contamination were less than 500 dpc/sq.ft. In working areas.

Near.y all beta-gamma radiation exposures were at the lower limit of detectability of film badges. The highest monthly exposure was 80 mr and l

the highest accumulated exposure July - December,1%0 was 150 mr.

i Q-

___n._

m-~~~~~"~~~~~

~ ~ * * * ~ ' ' ' ~ ~ "

~'~~~~#^

21" lL q

( S" A t:tal cf 2465/c cf cctivity in cxc ss ci str;a= background was

h relcased during Fcbru::ry 1960 - J nusry 29,1%1. All ralccs;s, aftGr dilu-tim, were well below the MPC for a mixture of unidentified isotopes.

I Thirteen 55 gallon drums containing 0.7/t per drum were shipped h/[

cff-site for disposal. At the conclusion of a study of gaseous iodine carry-p over in the waste disposal evaporator, ten 55 gallon drums containing evaporator bottoms mixed with concrete and each containing approximately 0.15 mc of I-131 i g,-

were also shipped to disposal. During the period Dececher 1,1%0 through 6

J:nuary 29,1%1 606/c of radioactive gas was released in 7945 cubic feet

[k cf gas from waste disposal. Release concentrations were well belor KPC. (See

{t:{

Section V. - Tests and Measurements, paragraph 5.)

i$

11 Systems and Components p

IC The nawly developed plant systems and components performed well i y wh;n considering their novel features of design. The more conventional features, By however, of a number of components, such as gasket closures and electrical coils, did develop trouble, resulting ir. program delays. A brief summary of the instal-w 10 tion and evaluation of a number of systems and their componentt is given in UA Section VI. - Systems and Ccmconents.

Included are sumaries of the following:

L'(

O

[1.[

In-Core Instrumentation System 8

Main Coolant System ff Pressure Control and Relief System 4

Charging and Volume Control System I

Chemical Shutdown System Purification System Q

Component Cooling System

. C Primary Plant Corrosion Control System yl Primary Plant Sampling System

];

i Waste Disposal System p

Shutdown Cooling System

%p Primary Plant Vent and Drain System Safety Injection System Reactor Control System

' 4p Nuclear Instrumentation and Reactor Control System

.u(,

Radiation Monitoring System Vapor Container Atmosphere Control Systems 4

Fuel Handling System R

Main and Auxiliary Steam System y

Condensate and Feedwater System p

Circulating Water System 5

Compressed Air Systems O

Electrical System N

Reactor Vessel h

Radiation Shielding 3

Turbine-Generator s

L v 12.

Design Changes The teros and conditions of Yankee's Provisional License and re-

, p.

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 k!ff a

}

reporting period. Th2 summary is crranged in two cicssificctions:

A. - Changes j

in the Secondary Plent, cnd B. - Changes C:nsidered to Af fcct the Primary Plcnt.

Cicssific tion A lists fiva d: sign changas while Classification B details 22 h

such changes.

13. Operatino Procedure Changes

]

The preparation and filing of Operating, Emergency and Maintenance Instructions, in Volume II - Part B of Yankee's License Application, was intended I

to show that operation and maintenance problems had been examined for general F

workability. This effort was of intertcediate detail and was intended to act

(

as a guide to plant operation.

Initial plant operation revealed that certain 9

revisions in procedures were necessary in order to correct for page to page incon-B sistencies and to present further detail or clarification of a particular opera-g tion. The preparation and issuance of certain procedure changes has, therefore, continued on a day-to-day basis. These changes have been made only to the de-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

[

kanual which serves to keep the plant eperation on a saft. and current status.

1 In this connection, all the Objectives, Conditions, and Precautions set forth in the license application have been observed and, in fact, additional Conditions y

and Precautions have been added in the detailed revised Instructional Sections M

of certain Operating Procedures.

[

z 14.

Plutonium Buildup Experimental Procedure E

IU A test procedure (Test Procedure 509F1) for measuring the power I]

coefficient during core life has been prepared and may be found in Section VIII h

of this report. This procedure is basically the same as Startup Procedure 503F1 Q

as filed in Volume II - Part B. of the license application.

An A

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 procedure consists of measuring the reacti'rity change corresponding to a power level by:

lli (3) Compensating for the reactivit=, change by allowing the average main coclant i#

terperature to change and (2) Compensating for the reactivity change by moving l

control rods.

t The reactivity change is calculated using the equilibriutr. change in average main coolant temperature obtained in step (1) with an experimentally k

mea ured 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,

.l these reactivity changes are corrected for variations in menon poisoning, main k

coolant pressure changes and small load fluctuations.

The change in reactor power is obtained from the change in gross electric generation using experimentally obtained calorimetric data. The power coefficient is then obtained by dividing the reactivity change by the corres-l ponding change in reactor power level. This procedure for measuring the power coefficient was adopted because it provides the best way of controlling the many variables associated with this measurement, because it does not violate plant L

operational limitations and because it conforms closely with the normal opera-ting procedures used by the operators and hence is well understood by tnem.

}

l (See Section VIII. - Plutonium Buildup Experimental Procedure.)

(

l

li]

Th2 t:st procedure (Tcst Procedure 509E1) fer messuring the mod-($

cratcr tcaper:ture coef ficient during ccro life has also be:n pr: pared cnd may

'f be found in Section VIII of this report. This procedure is v:ry siciler to i

Startup Procedures 503E5 and 503E8 used during initial core testing.

[

Basically, the moderator temperature coefficient measurement consists cf varying the average main coolant temperature while observing the corresponding

,(

chrnge in the reactor startup rate measured at low power levels. The change in f

'[

startup rate is converted to a reactivity change and this is divided by the change in moderator temperature to obtain the moderator temperature cxfficient.

J, The procedure is modified slightly from those used during initial tcsting to simplify the experimental instructions now that the operators are l

fcalliar with the basic technique involved and to improve the accuracy and reli-ability of the data obtained.

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- 6 3

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s

III. REACTm CGE LOADING AND ASSEMBl.Y On date of Interic License issue, July 9,1960, reactor vassal internals required for core loading had been assen. bled, 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-ing level with 1670 ppm borated water and the temporary nuclear instrumentation readout equipment was installed in the vapor container. Accumulation of back-ground count data was also begun. July 13 saw the special BF neutron detectors installed in their portable thimbles and the thimbles placed $n 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 day the initial core loading check-of f list 503E-1 was initiated. The temporary nuclear instrumentation indi-cated, however, a base count rate too low to be entirely satisfactory. Modifica-tions were made to detector thimbles and a higher, acceptable base count was then achieved.

On July 15 actual core loading began with the placing of the first shie rod and follower. Recording of hourly data on flow through the shut-down cooling system and temperature and boron concentration of the mair. coolant were started. The next day an operational test of the temporary boric acid injection system was performed, this system providing for protection during 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 due to the necessary re-positioning of the installed source vane from Position No. I to Ko. 4.

Source vane re-positioning was required to accomplish the correct location of the larger depth dimensioned vane in its proper core position.

Core loading progressed slowly due to various occurrences. Major A license among these was that of loading the first few assemblies into the core.

referenced loading precaution stated that, "if at any tire during core loading, the extrapolated value for critical size of the core is less than twice the number of fuel assemblies in the core, the loading operation must be 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-blics were added.

It was later concluded that the increased counts were due to a geometric ef fect between the source, the fuel and the temporary nuclear detec-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 pcrmitted core loading to proceed in compliance with all license requirements. As a re-sult of the above changts in loading procedure, the tot 31 core loading time was approtimately doubled.

i Another important area causing delay was malfunctioning of the fuel handling system equipment. The resulting delay due to this equipment consumed nearly 16% of the total 270 hour0.00313 days 0.075 hours 4.464286e-4 weeks 1.02735e-4 months core loading time.

(See Section VI - Systems and Components) Electrical interference on the temporary nuclear instrumenta-tion readout equipment also caused delay. Contactors controlling operation of I;

the fuel chute carriage induced high noise levels in the counting channels so that its operation had to be restricted te times when counting was not in progress.

The restrictions t>us placed on movement of fuel chute carriage delayed the load-ing process.

1 b

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+ e.aame w w w m

w t.,_c%,

i

g On July 26 c:re loading was complcted with the installation cf the l

76th fuel assembly and the three remaining P;-Be sourcss. The f:llowing day saw the boreted r :ctcr catsr diluted ta 1600 ppo cnd cn cxt:nsive check-out begun of the normal plant source range instrumentation. Certain modifications i

in this equipment were found to be nr.cessary. These changes were made result-ing in satisfactory count levels being obtained.

(See Section VI - Systems and Components)

T On July 30 the temporary boric acid injection system was dismantled, the in-core instrumentation structure was moved into the vapor contained and the F

upper core support plate and barrel assembly were installed in the reactor.

During the following two days, the in-core instrumentation structure was placed in the reactor and the flux wire thimbles and thermocouples inserted into their p

raspective positions. Assembly of reactor internals continued with installation cf the 24 contro) rod drive shaf ts; slight adjustment of several coupling fingers cas required in order to allow proper joining of the orive shaf ts to the control rods.

^

By August 3 installation of shim port plugs, guide tube holdown plate and 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 eas carried forward on eliminating noise froc the normal plant source 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 52 vessel head studs being installed and tightened by the 13th. Connection of 7

in-core instrumentation flux wire thimbles and control rod drive power and f(

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 g

aire drive control equipment. The initial core loading and reactor assembly cas thus complete on August 14.

U

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l 16.

IV-INITIAL CRITICALITY On August 19 preparations for the initial approach to criticality were complete. At approximately 6:30 P.M. a banked rod approach was begun, with criticality being achieved at 8:19 P.M.

Criticality was attained with rod Safety Group No. 6 withdrawn to 40" and other groups withdrawn to 36.25".

The extrapolated banked rod position was 36", showing excellent agreement with the predicted position of 42".

The normal plant nuclear instrumentation per-formed in a satisfactory manner throughout the approach with accurate predic-tion of the critical rod position being possible during the last 3" of banked rod movement.

The primary system conditions for this gnitial banked rod criticality were:

boron - 1650 ppm, temperaturer% 100 F, and pressure 225 psig.

The initial inverse multiplication plots produced curves convex in shape which resulted in expressed concern.

It has been four.d, however, that as boron concentrations are reduced and compensating control rod insertions made, curves gradually shif t to straight lines and then to concave curves upon application of programmed control rod withdrawals. As a result of an approach employing an asymmetric control rod configuration, there is evidence that the shape of the inverse multiplication curve is influenced by reactor boron con-centration and the position of control rods in the core.

1 2

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W TESTS AND MEASUREMENTS g

't Y

1.

Initial Low-Power Nuclear Core Tests jtp N

dim a.

Purpose and Scopes h

The low power physics test program was begun on August 14, with eg performance of cold rod drop tests. On September 4 the low power physics testing program, other than the above cold rod drops, was started with reactor i

operation at power levels not in excess of 5 m thermal. These tests were carried out in accordance with paragraph C. of the Technical Specifications y.

as referenced in paragraph C.(2) (a) of Yankee License DPR-3.

Tests were p

complete on November 8,1%0.

i,h b.

Control Rod Drive and Plant Scram Tests (503E2) - C_old Scram Tests:

m These tests were performed essentially as planned without serious

.J difficulty. Allrodswereproppedatleasttwicewhilethemaincoolantsystem

-Q was at 150-250 psig and 100 F with a flow of approximately 1000 gpe. All rods M

dropped and dashpots closed in approximately the design time of 1.2 and 0.1 seconds, respectively, for cold drops. Drop times varied from 0.89 to 1.27 g4 seconds with the majority falling in approximately 1.15 seconds. For a given F

rod drop, time varied up to 10% in consecutive drops. Circuit firing ties h;p were as follows:

y Millisec. to open

.g.

Input Sionals Rod Breaker

@h Manual Scram Buttons 22 26 Hi-Flux Channels 252 290 Hi-Startup Rate Channels 1%

233 Loss of Pressure Signals 599 609

}

Loss of Flow Signals 737 885 I

The circuit firing times for the conventional turbine-generator sensing de-j vices were also checked and found to be well within acceptable values.

w Th Control Rod Drive and Plant Scram Tests (503E2) - Hot Scram Tests:

One rod was dropped thirty times, a second rod was dropped fifteen g

times, the next four were dropped five times and the remainder wore dropped g'

two times each with the main coolant at approximately operating conditions of temperature, pressure and flow. All scram times vicre less than two seconds.

y The time interval from opening of the scram breaker to the attaining of the 1

fully inserted position was 1.725 - 1.835 seconds for the rod dropped thirty d

times. During this test the scram time decreased with the number of drops.

ie (Approximat ely 0.1 seconds). Rod scram times varied from 1.558 to 1.956 for Ek all scram tests made at operating conditions. The results indicated that the Ms scram tice is definitely longer for rods nearer the edges of the core. With ik no coolant flow, scram times appeared to have no relation to position in the l

core. It appears that cross flow in the upper plenum affects dropping time.

t Drop tests made with two pumps in operation resulted in drop times as low as i

1.417 seconds.

rt d

.____.___-_a

c-Contr~1 Rod and Bcron Drth Determinations at Los and Operatino Temperatures - (503E4 cnd 503E7) h-2 y

Experiment:1 rcsults cf thes2 tests were in good agreement cith i$f calculated values presented in Volume I, Part B. of the license application.

9 The fellowing tabulation compares the experimental arxf calculated values for h

the esid clean and hot clean reactors

(..

. r.

Celd. Clean Core Predicted Measured LT Fir Excess reactivity 16.5% d k/k 15.7 i 1.5% d k/k

'O

.i t Control rod worth with M

1150 ppo coron in main coolant 11.8%dk/k.

11.6 1 0.4% d k/k hy R:aetivity with 1150 ppm boron ard

/di, all control rods inserted

-6.1% A k/k

-5.8 1 0.8% d k/k h.

'[d;)

I:

B:ron concentration required for 5% d k/k minimum thutdown 950 ppm 1150 ppm Boron worth (in measured range

-5 of 1100 to 1900 ppm)

-9 x 10 k/k

-7.8 1 0.4 x 10 dk/k N

J{n Per ppe per ppe jot.CleanCore Y,

Excess reactivity 11.2% d k/k 11.6 1 0.5% d k/k Control rod worth with M

0 ppa boron in main coolant 16.0%dk/k 15.4 1 0.5% d k/k h5 R; activity with 0 ppm boron l ?;M and all control rods inserted

-4.8% d k/k

-3.5 1 0.5% dk/k y

Boron worth (in measured range h

-5

~0 of 0 - 1700 ppm)

-7.3 x 10 dk/k

-6.8 1 0.2 x 10 dk/k gi per ppe per ppa g

g d-Ternoerature. Pressure and Flow Coefficient Determinations - (503E5

~fi an E8)

Moderator Temperature Coefficient (at 2000 psig) g#

In general, the moderator temperature coefficient measured with no g

boren in the main coolant was apprcximately as predicted, whereas the coefficient B

$j l measured with high boron was nore negative than the analytical predictions. The l cxperimenta?. values compare with calculated values in Volume I, Dart B of the ri lliccnsaapplication,asfollows:

gg Temperature Coefficlent k

F

}gg Boron Temperature Idi (ppm)

(OF)

Predicted Experimental

-1175 68

-0.15 x 10

-0.5 1 0.1 x 10

?

-4

-4 1050 250

-0.69 x 10

-0.7 1 0.1 x 10

,r

-4 4

1050 514

-2.15 x 10

-1.6 1 0.3 x 10

-4

~ k, 0

514

-2.9 x 10

-3.1 1 0.7 x 10 hy.N e

g?

W:;

"g

i 19.

l Pressure Coafficient (at 20C0 osf o)

'M%

3e pressure coef ficient, measured with no tsoron in the main coolant f

systst was fr. good agreement with the predicteo coefficient values.

The experi-cental values compare with calculated values in Volume I, Part B. of the license cpplication, as follows:

}

b Pressure Coefficient (dk/k)

Q B:ron Temperature osi (Ppe)

(3F)

Predicted Experimental hh High 514

+2.0 x 10

+1.0 1 0.5 x 10 h

(with 950 ppa B)

(with 1300 ppa B)

r 0

514

+2.6 x 10-6

+2.7 + 0.7 x 10-6 s

fjow Coef ficient h

There is no detectable flow coefficient at Icw powers.

n4 Mair, Coolant System Heating Rate Deter 1cinations - (503E9) e.

Operating with four main coolant pumps, the main coolant heatup k

rato cas determined to be 19 F/hr and 14 F/hr at 4700 0

0 F and 5400 Syst:m heat losses due to convection and conduction at operating temperatureF, respectively.

FA are cpproximately 3 x 106 Btu /hr. With the clean core, pump heat was not suf-6 ficicnt to hold the main coolant temperature up when the turbine generator k

auxiliaries were operating.

It was necessary to discontinue steam flow to the air ejector and to stop steam generator blow down periodically.

{

Apparent heat Icssas varied considerably during thase tests, primarily because of changing

.f canditions of valve stec leakoffs and steam traps. Having improved these con-

' ditiens and with core decay heat, it is now possible to maintain the main coolant K

P tsmperature without reactor power when the turbine auxiliaries are operating.

f.

Nuclear Instr'nentation Recoonse To Asvurnetric Control Rod Positionino -

(503E10)

It was detersined that the nuclear instrumentation responds to rod S.

motien in any portion of the core when the reactor is shut down and that the instrumentation responds to both radial and axial changes in com flux or core QT

.powsr distributions when the reactor is critical. The nuclear instrumentation

'i

, responded to flux tilts and asymmetries, but further tests during power opera-

,tien showed that it was sensitive to variations in coolant temperature, contro'.

irod position and xenon condition in the core. These factors complicate absolt.te lcolibration of the nuclear instrumentaticn for flux tilt detection.

lfcund that other instrinentation provided a more reliable check on power and It was F

flux tilts and this is diccussed i.Section V, paragraph 2.c.

Durir:g early power testing, an unexplainea core pcwer non-synrnetry was detected by the loop temperature detectors, the core exit thermocouples and r

the nuclear instrumentation.

Investigation revealed that in re-connecting cablas to the mechanisms, those supplying mechanisms No. 8 and No.12 had been inter-shanged.

Rod #12 being used with the other three rods in group 4, which was the

$cntrollir.f_ group at the time, pr2sented an asspicetrical control rod pattern and ithus caused a flux tilt.

The cable interchange was corrected during Shutdown

-. 9 (12/2/60). -The control rod mechanism nurbering system, which wac faulty h'utureorrorsofthiskind.nd c:used this error, has since been modified to a positive scheme to

+

I

,Lw 2 ~'

. _ _ - - ~ ~ ~ '

._ _ --- --- - - - ~ ~ ~ ~

in

'h 2,

Initial poner Operation Tests ih

!a4 a-Purpose and Scope:

iN These tests were performed to establish the performance t.haracter-h istics of the plant when operating at power. The reactor plant was t.hown to be g{

cxtt:mely stable under all operating and transient conditions. Transient tests e

showed that limiting operational conditions were reached in the secondary.

Q plcnt before they were reached in the primary plant.

Tests have been completed

'We by step increases not exceeding 30 W e up to 120 W e gross output.

k,W b.

Power Coef ficient cnd Less of Load Transient Test - (503F1) h.

Power coefficient measurements have been made during the majority N

cf the scheduled generator load changes so a great deal of experimental data is availab ower levels up to 392

thermal. The power coefficient is g

-3.3 x 10 ge at p/W thermal i 1.7 x 10-5 dk/k determined from data gathered during Q

th2 first 500 equivalent full er hours of operation. This compares with an h

cnalytical value of -3.3 x 10- M k/W thermal. This excellent agreement is h

probably fortuitous when one considers the uncertainty of the values.

[g, The loss of load transient test was performed with the plant load y{'s at 30 R electric and 60 W electric gross output. The generator load was re-duced to :pproxiaately 3 E electric in seconds with no scram on the reactor g

cr turbine and with the rods on manual control. When this test was performed g

at 30 W electric it was found that the water level in turbine moisture separa-W tcrs increased above the normal control points. The moisture separator piping t s modified and the test performed at 60 W electric. No plant lhitations Q

wera exceeded during this transient. The core outlet temperature reached 542 F i

main coolant pressure reached 2330 psigt and secondary plant plant pressure in-cr:ased to a maximum of 850 psig. This test will not be done from power levels ab;ve 60 We to avoid operation of secondary plant safety valves and because the E

data already obtained is adequate to use in extrapolating to the conditions that h

would exist af ter a drop from full power of 120 We.

Loss of load large enough p

to cause operation of the secondary plant safety valves (approximatelf 90 We) jiq could cause a plant scram because of high water level in the moisture separators.

In the event that reactor scram did net occur for loss of load from 120 W e, the R@

primary and secondary relief and safety valves would operates however, no damage

.k could occur to equipment in either the primary or secondary plant. Unnecessary cperation of safety valves is undesiraba because of the increased maintenance Q

rcquired on these valves and because of the possibility of failure of the valves g

to reseat leak tight.

r.2 c.

Nuclear Instrumentation power Calibration - (503F2 and 503E10)

%q M

Data was obtained on the relationship between primary and secondary gg plcnt power levels up to approximately 98 W gross. biectric which corresponded pr to approximately 315 W themal output from the reactor. Data obtained to date M

his shown excellent agreement with the heat balance data presented in Section 200 h

cf Volume I, Part B. of the license application. The plant is slightly more of ficient than calculated and is able tc prM sce slightly over 120 W electric h

gross with a thermal output of 392 W from the reactor.

It has been found that d

th? average reactor output can be measured to approximately 1 5% and that the nuclear instrumentation can be calibrated against the gross electrical output

,d M(fp cith a resulting error of no more than f SL Therefore, the nuclear instrumenta-tion will continue to be calibrated using the gross electrical output as the 1

measure of average reactor power.

F5h Uit y

_.~

- ~

r m

g Several methods have been found useful fcr ch;cking power asyneetrics 17 the ccre, including tha nuclocr Instrument tion, tha loop tesparature det:cters,

,.j i

ccre crit wat:r thermocouples and ccro flux cires. The prefctred method is the lR[

un cf loop temperature detectors 3 ince these are less subjtet to short tent variations than the nuclear instrumentation and because the loop temperatures are

,n the best measure of the average power developed on a region of the core.

It is possible to detect power asymmetries of approximately + 3% using these methods.

I ip d.

Reactivity vs Fission Product Level Followino Power Level Chances -

h (503F3 and 503F6) op The variation in reactivity due to transient fission product such as 86 x non following power level changes has been recorded throughout the power opers-4 E

tions to date and has included observation of the reactivity change following

%j chutdown of the secondary plant. The Yankee reactor will be able to override pe:k x non after turbine generator shutdown throughout core life and return to 3

full power for approximately three-quarters pf core life, since the reactivity I,-

lest due to peak xenon is less than the reactivity gained from the power coeffi-P cicnt during the load reduction. A return to approximately 25% of full power R

cill be possible near the end of core life. Tht experimentally determined values U

cf th? reactivity loss due to xenon are in reasonable agreement with the calcu-lot;d values.

g e-Biolooical Shieldino Effectivenest - (503F4)

,h y

Radiation surveys made at 15, 30 and 60 W gross electric revealed

'p highsr than anticipated neutron levels throughout the site. The source of the g

neutron radiation was found to be located at the top of the neutron shield tank.

7-It was due to a streaming path of intermediate energy neutrons throug. the space ig betwe:n the reactor vessel and the top of the neutron shield tank. Measurements bf, made at the streaming location, using a " Double Moderator Neutron Dosimeter" f,

indicated an average neutron energy of 0.3 mev. Following are neutron levels M

indicated at 60 E gross electric load at vapor container and general plant area Mj locaticns Q

Top of Vapor Container 57 mrem /hr Equator of Vapor Container 4 mrem /hr e

Below Vapor Container 0.4 mrem /hr y

Plant Of fices and Laboratories 0.4 mrem /hr j

J Temporary shielding consisting of two rows of 55 gallon drums filled q

cith cater, which was installed over the walk plate above the neutron shield d

tank, reduced the neutron level by a factor of about 6.

Neutron levels at the N

l vapor container and general plant area af ter the addition of the temporary shield-

[

j ing aro listed below for a reactor power level of 90 W gross electric.

g l

Top of Vapor Container 18 mrem /hr N

Equator of Vapor Container 1 mrom/hr b

Below Napor Container 0.2 mrom/hr M

Plant Of fices and Laboratories 0.1 mrem /hr D

l Ft>rmanent shielding of masonite encased in aluminum, 22 inches thick s

let tha streaming location and 12 inches thick over the remainder of the neutron ishiold tank has been installed. Neutron levels at the vapor container and general lplent area af ter the cddition of the permanent masonite shielding are listed below:

}

6,0_ W e 90 s e

_120 We i

Top of Vapor Container 3 mrem /hr 4 mrem /hr 7 mrem /hr Equator nf Vapor Container 0.2arem/hr 0.3 mrem /hr 0.6 mrem /hr j

l Below Vapor Container 0.04 mrem /hr 0.1 mrem /hr 0.1 mrem /hr h

Plant Offices and Laboratories 0.05 mrem /hr

<0.05 mrem /hr

@.04 mrem /hr

. ; Mi s

- - ~

~ ~-~ -~ ~ ~ -- ~ ~- -~ A $ A

.-.--c._.~

n.-~~

. ~ - - -

l

22.

3s Gansa radicti:n Icv 21s through auxilirry shiolding in the control room,p/hr.rir.ary auxiliary building, and waste disposal building did not exceed O.05 mr There were no indicatf or.s of significant streaming around pipes, i

plugs, or void spaces in shielding through which radiation might leak.

f.

Instrumentation and Control Response - (503F5) hi This test procedure was used to check the behavior of the plant control circo s under both normal steady state and transient conditions. The plcnt control sas found to be extremely stable even with low plant steam load <..

)t

[

As cxpected, the inherent negative temperature coefficient control proved an of fcetive stabiliz'ng influer.ce.

It was found that the reactor regulated on t moerature control with no rod motion required for small load variations at power levels as low as 1% of full power. The automatic rod control system has proven more stable than anticipated and has held the main coolant temperature J

cithin the contrcl band during normal generator load changes and steady power

{

operation, insluding transient and stable xenon conditions.

g.

a Emeroency Coolino bY Natural Circulation - (503F7) c.

This test was performed from 60 and 120 W gross electric loads.

It cas deterrined that adequate cooling was provided to the core under the con-ditiens simulated in the test. During the initial transient following scram, N

tra main coolant pumps running on turbine-generator inertia provided adequate cooling to safely remove the stored heat in the fuel. After all pumps had come i

to rost, natural circulation was established as exp'cted and provided adequate m

ccoling for removal of the decay heat generated in the core.

i During the performan.;e of the test from the 120 W gross electric k

Ictd level, however, the temperature transient was more severe than predicted.

I This test was carried out by manually tripping two main coolant pumps causing a rsactor scram and turbine trip when the main coolant flow decreased to 60L y

g As indicated above, the two other pumps remained in cperation for a period of E

1.5 minutes after initiation of scram, running on turbine-generator inertia for d

that length of time. The decrease in main coolant average temperature imwdi-

.j ately following reactor scrcm was greater than had been gnticipated; the averagg 3

maig coolant temperature dropping from the normal of $14 F to the range of 470 --

480 F.

Analysis of a = ore severe transiant, loss of 485 W thernal load with m

accompanying reactor scrac,, shown graphically on pag"e 402:11 of the license appli-N.

cttien,. predicted a temperature reduction of or.ly 10.

It has since been deter-I mined, however, that the analyses made for the license application did not take

(

into account the significant effect of relatively cold feedwater being pumped

{

into the steam generators during the first few seconds of the transient. This cdditional water compensates for the falling water level caused by the increase in pressure which follows a sharp reduction in steam demand.

Following the test the temperature transient and its implication on plcnt operation was evaluated. Projection of the test data to the most severe tr:nsient of this type possible under the present license,(that is, loss of 392 W thrreal load with reactor scram, all four main coolant pumps and both boiler ferd pumps remaining in operation and the highest worth rod stuck), and a pre-liminary analysis of these data indicated that there was no possibility of the reactor regaining criticality due to temperature decrease se long as 50% of equilibrium xenon is present in the core. As a result, power operation was i,

allowed to proceed but with the restriction that the plant was not to l>e taken ab*.va 60 W gross electric, corresponding to 198 W thermal, unless at least s

j j

50% of equilibrium xenon was present. This xenon restriction continues in force.

n

~--"-

" ~ ~ ~ ~ '

" ' ~ ~ ~ ~ ~

m.

..a.

eo i -

f,,

t r_

e

,.i s

', A furth2r, more d tailed cnalysis cf this tr nsient, wheroin cccount

j. O.}

wae taken of the effect of cold feedwaiter entering the secondary side of the

{ %.

t lct:am-seaergtors,wasmade.' This analysis was based on experimental information obtched dur,jng plant startip tests.

It was' assumed that the plant was operating l

i Og initielly at 392 m thermal, or 120 m gross electric, all main coolant pumps

%(

were run.ning and that the boiler feed put p line-up was normal. Reactor scram cnd turbir,9 trip occurred after wnich all main coolant and boiler feed pumps hk continuedtgoperste. Th( analysis shows that a minimum main coolant tempera-pij.

ture of 468 F would,be reached. Considering the present core condition, with Q

no credit taken for equilib:ium renon, but a reactivity worth of 0.5% for equili-i

' briue: samarium, a worth'of 0.4% for depletion due to burnup and the highest worth

}

cf a stuck rod, it is ' concluded that gritL;ality would not occur until the main dr coolant. temperature had fallen to 425 F.

The reactor then, at the peak of the T

transient, is shown to be cub-critical by at least 1% without the worth of the equilibrium xenon,;ind at least 4% sub-critical when the equilibrium xenon worth is included. Although this set of circumstances cannot, of course, be duplicated 5

at the plant for, test purposes, the calculational procedure by which this minimum M

temperature was developed has been checked against recorded plant data resulting t

i from a 60 m gross electric loss of load with continued main coolant pap and W,

'nboilcr fead pumg3 peration. The calculated minimum main coolar.t temperature for f(

o

this case is 10 acn than the temperature actually experienced, indicating that q

the cnalysis is colservative in this respect.

4, The possibility of restarting one of the main coolant pumps after a gy low flor scram, which conceivably could introduce a slug of cold water $n the 44 cera, is ef fectively prevented by a pump restarting interlock arrangecent. This g.

interlock prevents restarting any pump unless the cold leg stop valve is closed g

'and the 5" bypass valve is open. Once closed, the cgld leg stop valve cannot 9

,be opened until the temperature is matched within 20 F.

b

?a

@ct N

YiVe k,1 i

$ii sy

.v, Yi

'9.k

' f f

f

_s

]ts, i

_ _. _.+.

.-~~~

--~ ~ ~ ~ ~ ~-"-~~-

~"

~~ ~~*""**~^'*" ~ ~'

3.

500 Hour Run At 392 MX Thermal (f

Q Under the terms of Yankee's Provisional License and paragraph D6 l $

cf the Technical Specifi:ations, Operation at Full Power, Yankee is required E

to operate the reactor plant for not less than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months , as continuously as possible, at a level of 392 MW thermal.

g N

l At the conclusion of Shutdown No.15, details of which were re-l ported in S?ction II, paragraph 6., the plant was re-phased to the electrical h

transmis, ion system on January 16,1%1 and load increased to 60 W gross b

olectric. Operation at this load level was carefully watched for approximately h

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , results showing that the former turbine shaft vibration had been re-p$

duced to approximately It mils. Electrical load was then increased to 90 W gross electric and maintained at this level while a number of flux wire irradi-ations wera performed. Analysis of these flux wire runs were moct favorable, p:

r2sulting in a power level increase to 105 MW gross electric. Two additional

. d in-core flux wire runs were made at this power level, the analysis indicating M

that load could safely be increased to the licensed full power rating of 392 W h

th2 mal, or approximately 120 MW gross electric, without exceeding heat flux a

and temperature limitations. Full load was attained at 3:25 P.M. on January j

17, 1961. On this date, the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months full power run was started.

p During the succeeding two days, a reduced load rod interchange t:st and full load biological shield ef fectiveness tests were performed. Col-4 Icction of ard analysis of in-core instrumentation data at the licensed full d

power leve' continued at regular intervals throughout the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months running B

period. The results of this work have continued to indicate that all actual y

core thermal perf ormance valves are well w ithin the predicted values. The hot T

channel factors experienced to date have been found to be 10--20% lower as com-G pared with the design hot channel factors which were used in arriving at the 4

392 MW themal rating, During the entire 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months run the reactor exhibited D

cxcellent stability and good response and control characteristics, as well as M

smooth and trouble-free operation. The in-core instrumentation system also cp: rated in a reliable and consistent manner.

Perf ormance of the turbine-9:nerator proved it to oe essentially vibration free.

Its control equipment centinued to exhibit excellent response and was capable of rapid predictable f

operation. This was evidenced during the severe winter storms of early Febru-E

{g d

ery whan many transmission and generating unit interruptions occurred resulting

)

in major syster load and frequency variations Plant control systems handled th?se variations in t most 7xcellent manner.

Mc m

Primary and secondary systems chemistry, continued to be maintained y

all within their specifications. A complete report may be found in Section V.-

kt Tests and Measurements, paragraph 4.

Survevs of neutron levels at various e

lccations around the vapor container, plant offices and laboratories also proved O

to be very satisfactory, Although the primary system radiation levels have risen, cs expected, they also remain quite satisfactory. Results of operation, with re-spect to health physics and shielding at the licensed full power level, may be q

fcund in Secticr. V. - Tests and Measurements, paragraphs 2.e, and 5.

h

{ +:

During the 50C hour runnir.g period, three shutdowns occurred, two E

of a planned nature, the third being due to an accidental turbine trip and y

reactor sc-am.

(See Section II. - paragraph 6.)

id(

i p

f

r s-wl Conclusion cf the 500 h:ur run occurred on Fcbruary 8 at 12 noon.

[

Th2 operation cf the plant during this period ct the nominal 392 E th:rmal I t rating produced 60,500,000 gross kilowatthours.

f; l' q It is intended that in the near future a report will be submitted rj to the Division of Licensing and Regulation containing data, calculations and j

cccident analysis demonstrating that operation at some increased power level, g

n;t in excess of 485 W themal, can be accomplished within the limits imposed

.h by the Technical Specifications.

h

' E

' 6 n.

5 5

4-Chemistrv d

' g C

Chemical evaluation of plant systems has proceeded since February,

b 1%0.

From that time until June the major activity was the cleaning of equip (- E ment and auxiliary systems. Water storage facilities were filled; carry-over cf r2 sin froc the water treatment ion exchangers made it necessary to dischange b

ater from primary water storage.

Inspection of ion exchanger internalt resulted 6

in rminor maintenance being perfomed on one of the units. A filter cartridge g

downstream of the exchange unit was also installed. Np further difficulties have g

be:n experienced with the water storage facilities.

($

The primary system hot flush was perfomed in June. Oxygen was scavenged from the water with hydrazine. Thirteen hours were required to decrease K

the cxygen concentration to less than 0.2 ppm. As experience was gained, subse-quent oxygen removal stages have been accomplished in from four to six hours.

' r Total solids fluctuated between 2 and 10 ppm and conductivity between 30 and 9 t,

micro mhos. No difficulty was experienced in maintaining the chloride concen-d traticn to values of less than 0.2 ppm.

$b In July the primary system was borated to 1600 ppe boron.

During h

the core loading a typical borated water analysis showed pH 5.4, conductivity

. f 9.8 micro chos and chlorides less than 0.1 ppm. From this period until boron y

removal, nickel concentration increased to 0.80 ppm. At one time, immediately af ter a rod drop test series, the nickel concentration was observed to be 1.12 ppe. Manganese concentration was surprisingly high during this same period, A

racching a value of 0.50 ppm. However, this was traced to an incident involving E

the inadvert,ent injection of boric acid from a jury-rigged shutdown system that i

ccntained carbon steel and its assor.iated corrosion products. After the boric f

ccid removal process in September, no corrosion products have been found chemi-1 cally within the limits of detection,--about 0.1 ppm in most cases.

D n _- Primary Coolant From the start of power operation, reference water conditions were maintained, except for minor deviations. primary coolant pH values have fluc-q tuated between 7.2 and 9.0.

Since hydrazine is added to the coolant as an cxygen scavenger following cold sh~utdown periods, it is expected that a hign pH

. [(

condition will exist after plant startup due to ammonia formation. In addition, primary grade make-up water is aerated which introduces a source of nitrogen g

into the system which, when coupled with hydrogen and the neutron flux, reacts p

p ee am-&,=%--

iAM-3 MC4

'e'-wM-'-*1

iw-b4+'.-am**-a-w-m99 NW sS

'P m4WundedAMomeme t lem.J-WA's=W- ' W h%H'#4 *P w'!.ush Jr. $M %.s% d5Eh>D- %

+4 4

h

~ ~ (%*)O in th2 cme mann2r as hydrotLu cnd results in a high pH. Operation und2r th:s cond'.tions is of sh:rt duration cod is not harmful to thm system.

The crud level in the ptimary coolant has varied between 0.1 ppm and 0.5 ppm. This is considerably lower than specification, 2 ppm. and may well be due to the frequent operation at high pH values.

The pritrary coolant rurification system operated satisfac'.orily.

The anion ion exchanger effcctively removed residual boric acid from tre system during the boric acid removal o >eration. Mixed bed ion exchangers are perictm-ing as designed.

The Neutian Shield Tank contains chronated water. There is no apparent separation of constituents ir the tank. A typical analysis from the top and bottom of the tank shows pH 8.7 and 9.23 chromate 365 pre and 340 p,x::,

respectively.

b.

Water Hadiocherristry Ra Jioactivity level and the nuclides present in the prinary coolant are considered normal for the present operation. The major radionuclide p esent in the water is the 2.5 hr Manganese -5.

primary coolant has varied from 2 x 10'g/t/ml during low power testing toThe gross specifi 8 x 10-2 long-liv /'c/ml at 120 la gross electric operation.

Specific activities of the ed corrosion product nuclides are presented in the tabulation below, b

Crud Analysis dis / min /mg EFPH+

Date Fe-59 Co-60 Co-58 Cr-51 Mn-54 4

56 11/21/60 4.1x10 5.1x10 8.7x10 2.0x10 3

3 5

167 12/14/60 6.2x10 3.0x10 7.0x10' 6.9x10 1.2x10 o Equivalent Full Power Hours Radiciodine Analysis [c/ml Date I-131 I-133 1-135 8/21/60 3.9x10-5 12'/2'/60 2.3x10-D 2.3x10 7.0x10 12/13/60 3 3.7x10-5" 12/14/60 5.2x10-5 12/16/60 14.9x10'5"

[

12/26/60 6.8x10-5 1/23/61 6.5x10-5 l

- System Sampled After Reactor Scram Radioactive Gas Analysis f c/cc (gas)

Date A-41 Kr-87 Kr Kr Xe Xe 11

/60 3.2x1 T4 2.0x10-4 1.9x10~4 3.9x10" 11

/60 2.5x10-3 4.4x10-3 9.8x10~5 6.5x10-5 2.0x10-4 1

/60 1.3x10-3 g,$xto-5 7.5x10-5 1.3x10-3 4.7x10-5 1.3x10-4 12

/60 1.1x10-3 1.6x10~4 4.7x10~5 2.9x10-4 12/2/60 6.6x10-3 5.5x10-4 4.3x10-3 1.4x10-4 9.5x.104 12/29/60 8.4x10-3 8.8x10-3 3.9x10-4 1.7x10-3 1/3/60 #

7.3r10-5 1.8x10-3 1.5x10'4 5.9x10-4

Reactor Shutdown l

i l

k

- - - -..~.- - -

^-

- i l

N3 cerr:sicn products cf the control rod cbscrber mat; rials hav' been d2tected chemically er rrdiochemically, which indicat:s that the control l

rod diffusion bonded nickel clad is still present. Small quantities of fission product radioactivity has been detected in the water and gases of the prirary

j coolant.

It is not certain at this time whether this activity is due to tratap Uranium in the system or to microscopic fuel defects.

'l Argon -41 is present in the dissolved gases of the primary coolant.

This nuclide results froc small quantity of target argon present as an impurity in the aerated primary grade water. This is a transient conditon that should be lossened as primary grede water becomes de-aerated and system make-up require-ments are reduced.

Ion exchanger decor.i. amination factor averaged approximately 110 prior to January 25.

This was due to the fact that the sample lines from the

(

inlet and outlet of the ion exchanger terminated in a comon header.

Outlet I

s=ples picked up deposited radioactivity in this header which was reflected in a lower decontamination factor. Modification of the outlet sample line re-sulted in an ion exchanger decontamination factor of approximately 600.

c.

Waste Disposal

^

The stripper-evaporator system has processed 21,000 gallons of primary plant water. Excellent decontamination was achieved with th: evaporator g

distillate samples counting background. However, the recovered water was not raturned to primary water storage since the presence of hydrazine in the feed solution resulted in the formation of amonia in the electrolysis process with subsequent carry-over tc the distillate. The effect of ppm concentrations of cmonia in primary grade water is under study.

r d.

Secondarv Water Chemistry i

No dif'iculty has been experienced in maintaining secondary water specifications. Oxygen control presented no problems. Chlende concentra-tions were high during the initial steam blow through the meisture separators and it was necessary to empty and refill the secondary system. Steam generator chemistry has since been controlled with approximately 0.1% blowdown rate.

No leakage has been detected in feedwater heaters, de-aerating con-denser or primary to secondary in the steam generators. The low solids treat-ment using hydrazine and morpholine appears satisfactory and its use will be centinued.

3_ _. _ _ _. -

/

in ii!

5 H alth_ Physics

+

Radiatior. Levels Gama radiation levels throughout the plant have been well below

{'

design estimates. No doubt this is due, as yet, to only slight build-up in the deposition of activated corrosion products. The hiqhest radiation level encountered at equipment not in the vapor container war 100mr/hroncontact with a purification pump and 1500 mr/hr in contact with the inlct pine at the pump. Radiation levels at the Low Dressure Surge Tank and the Low Pressure Surge Tank Cooler reached 300 mr/hr and 60 mr/hr respectively.

The highest radiation levels encountered in the vapor container cubicles. These were founc shortly af ter reactor shutdown were in the loop /hr were experienced on contact toreachabout150mr/nr. Levels of 450--700 mr with main coolant piping.

At the Telefler in-core intrumentation area for storage of irrtdi-ated cables, the maximum radiation levels were 100-150mr/hr.

Increase in the rador, concentration in the vapor container was noted af ter plant warm-up to cperating temperature. The radon concentration was well below maximum permis-sible concentrations and therefore not a personnel exposure problem.

In the Radiochemistry Sample Cubicle, general radiation levels were below 2 mr/hr except while obtaining a main coolant crud sample when release of disselved radioactive gases, chiefly Argon -41, on depressurization of the main e

coolant has produced locally levels of 200 mr/hr inside the sample hood. Release concentrations of these gases from the primary vent stack were well below madmum permissible concentraF ons.

Theradiat.onlevelovertheionexchangestoregepitwas20mr/hr with no water in the pit. Elsewhere in the generally accessible plant areas, gama radiation levels have been close to background levels.

Neutron flux levels outside the vapor container required the instal-lation uf additional shielding. This is described in the Biological Shield Effectiveness Test. (See Section V. - Tests and Measurements, paragraph 2.e.)

Contaminatior. Levels There have been only a few instances of contamination in the Poten-tially Contaminated Area and none in the Clean Area. The highest contamination was found at a leak in #1 steam generator. Smears of the leak area over a 12" length of 1" pipe showed 160,000 dpm. Contamination of 10--40,000dpm/ft2 was

~

found in the working area during repair oftthe leak.

Localizec cases of contamination in the Primary Auxiliary Building, WasteDisposalguilding,RadiochemistryLaboratoryandSampleCubiclewereless t

^:

than500dpm/ft in working areas.

s 9

Radiatior Exposures l

{

Nearly all monthly beta-gamma radiation exposures f or Yankee and Contractor personnel were at the lower limit of detectability of the film badge of 10-20 rrr. The highest monthly exposure was 80 mr, and the highest

,j accumulated exposure, July - December,1%0, was 150 mr. All r.eutron exposures, i

except one, were less than 60 mrom/ month, which is the lower limit of detect-

}

ability for a neutron film badge. One neutron exposure of 70 mree was re-S ceived in conjt.nction with installation of four 100 curie Po-Be reactor startup j

sources, a __.

29.

Mi' Waste Disposal:

}l Liould A total of 562,060 gallons of low level radioactive liquid waste yh containing 2465/c of activity in excess of the background of the stream was h

released frce February,1%0 through January 29,1%1 Tht: orioin of most f

of this waste was from floor dra5ns in the Primary Auxiliary Building, Waste 6;i, Disposal Building and Radiochemistry Laboratory drains. All releases after dilution with condenser cooling water were well below the MC for a mixture ff.{

of unidentified isotopes.

i Thirteen 55 gallon drums containing evaporator bottoms with 0.7/c

,f per drum were shipped off-site for disposal. This waste arose from reducing boron concentration in main coolant for power testing.

Solid

%i Ten 55 gallon drums containing evaporator bottoms mixed with con-k crete were shipped off-site for disposal. Each drum contained about 0.15 ac T:t of I-131 which was used for study of gaseous iodine carry-over in the evaporator.

@3 i

7 Gaseoyi f

During December,1%0 and through January 29, 1%1, 606 g>

active gas, chiefly Xenon-133 was released in 7945 cubic feet of ga/t of radio-s from the d

Caste Disposal System. The release was made through the primary vent stack g4 thile being diluted with air from the primary auxiliary building ventilatien g;

fan. Release concentrations were well below WC.

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VI. SYSTEMS AND COlpONENTS 31$&

1.

In-Core Instrumentation System 3v The In-Core Instrumentation System was added to the plant design

[4) after all other design and a large part of construction were complete. There-fera, by necessity it had to be adapted to existing conditions. This required h

some cxtension of existing technology and experience. The resulting system consists of two types of measuring devices; 22 in-core flux wires and 27 core

.p cutlet water thermocouples. Equipment is provided to remotely insert and with-

?

dro the flexible wires, measure and plot the neutron flux seen by each wire while in the core.

Individual wire flux plots are then combined into ' Flux-

,6 maps". Thermocouple measurements are also recorded thus aiding in the prepara-tion of integrated power outputs at positions throughout the core.

o In field erection, two errors were found in dimensions of the in-x ctre instrumentation structure. One was easily corrected, in the field, by ha cutting off a protruding edge of the egg-crate assembly. Tha cacond dimen-E3 sional error was that the shim rod adapter holes through the reactor head were k

too small to receive tho in-core instrumentation thimble risers. This dimen-ly sicnal error resulted in minct damage to the instrumentation structure and the sh2cring of fcur small assembly pins during installation of the vessel head.

All pins were recovered and new ones installed. The adapter holes in the

?

rzactor head were bored to a larger size, in the field, with a resulting delay cf tbout eight days.

During pressure testing, the special mechanical seal closures on the thimble risers developed leaks. These were a result of in$orrect dimen-sien details and leakage af ter thermal cycling to normal operating pressures.

p In view of the access difficulties for routine re-tightening af ter thermal 9

cycling, and in view of the sensitive area shared by the electrical coils, y

cables and thimble riser seals, it was decided to seal weld the closures.

In operational testing, the original design of the core instrumenta-ticn flux wires exhibited twisting and kinking problans. The barrel storage of the retracted wires had to be abandoned. The field tnen modified the system to an "in-line" retraction of the withdrawn wire. This prevented the wire from hya kinking but interfered with the Selsyn motor position indication used in the h,%

criginal design. Attepts by the field to accomplish accurate position indi-NE caticn without wire kinking resulted in modification of the position indication 3

system using micro-switghes at certain important positions along the path of a T

oire r These modifications, along With increased size of the drive motors, made thm flux wire system conceptually operational. Difficulties continued, however, oith the operation of titanium flux wires, which could not be fabricated to tolerances close enough to allow free and reliable operation through the helical drive wheels. As a result,. carbon steel flux wires we.re installed. The flux wire operational dif ficulties caused about three weeks delay in the startup program g

prior to December 1,1%C.

I e

The exit core thermocouples of the core instrumentation have per-formed very well. The27therecoupleshavecontinuedtoperformsatisfactorigy cnd it is believed that coolant temperatures are being measured closer than 12 F f

cf cctual temperatures.

Kk M

y 2.

Main Ceolant Systee Construction of this system proceeded rapidly eith components Deing

'i easily placed due to advantages of the elevated vapor container. As an example, the 100 ton stear generators were taken from railroad cars to final positions I

in six hours.

l Hydrostatic testing of this system showed no weld leaks, although g

several gasketed flange leaks were detected. Hot flush testing showed almost a

no foreign material in the system but the hot flush strainers which were used

' k were of too fine a mesh, resulting in iron oxide clogging as corrosion increased eith temperature.

In operational testing, with the core in place, the system y

performed according to, or slightly better than, design with respect to flow, j

pressure drops, temperatures, and heat losses. Complete venting of the main g

coolant system could not be achieved without water circulation due to a captive p

air mass in the vertical steam generators. Pump operation was possible and en-g traireent of the air mass was easily achieved, collection occurring under the y

reactor vessel head from which venting was possible. Draining an isolated loop should be an infrequent occurrences however, w,and refilling of g

hen it..is done, air

,i removal may have to be made by chemical means. In early operational testing, y

the entrained air masses under the reactor head were advantageous as surge i g chambers to pressure transients caused by pump starting and heating. Finally, g

whan complete air removal had been achieved, the system became a "hard water" system, and rapid pressure transients occurred, sometimes positive and some-times negative. Therefore, "hard water" operation of pumps is closely watched

, y 4

cith pressurizer steam bubble operation used wherever possible.

h During the hydrostatic test, there appeared to be no leakage across 1

the primary-secondary tubing of the four vertical steam generators. Later power y

testing confirmed this, with no radioactive transfer to the secondary system.

During hydrostatic and operational testing, the primary side gasketed manhole Ip closures leaked. The primary side closures were seal welded after several R

attempts had failed to make the gasketed closures leak tight. The secondary d,

side gasteted closures, which operate at a maximum of 800 psig vs the primary b

side 2000 psig operation, were made leak tight after two or three attempts.

d Further leakage on each steam generator secondary side occurred from four screwed k

test plugs left in by the manufacturer. These leaks were eventually located under T

the thermal insulation and the screwed plugs were replaced with welded plugs.

On December 16, during a routine inspection in the vapor container, a leak was fi-discovered in the take off nozzle to the No. I loop flow DP cell. No. I loop h

sas isolated and later preliminary examination showed the leak to be between d

the stainless steel clad of the steam generator and the stainless nozzle weld.

This leak was repaired externally during the period of turbine modification by application of the manufacturer's recomended preheating and welding procedures.

The heat transfer characteristics of the steam generators, at power levels en-

- y countered to date, are better than design. Little fouling, however, has occurred

k as yet. The steam quality exiting the steam generators appears to be about 3/4 to lj%, but some inaccuracy in these early deteminations exist.

S The four main coolant pumps have operated at flow and various main g

coolant pressures and temperatures for approximately 4000 hours0.0463 days 1.111 hours 0.00661 weeks 0.00152 months through the re-

j iorting date. No bearirig, vibration, or low winding resistance to ground problems F/

rere encounted. Considerable trouble was encountered with the -leakage of.

h the gesketed closures.

It was found that it was impossible to make the originally 3rovided nickel gaskets leak tight.

In replacing the nickel gaskets with copper l

1 jaskets, it was discovered that the flanges of the pumps had been deformed about t

WJd

__ _ a ~ ~ - - - - - - - - -~-~ ~ ~ - ~ - ~ ~ ~

' 5;

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h.

0.012", cno re-facing cf the flangrs c:s r: quired. Pump removal and re-fccing M

gperations resulted in cbout a two we:k delay in test operations. The r: place-7 ment copper gaskets were leak tight when originally installed.

After cycling the plant from ambient to temperature to ambient, the copper gasketed joints ih le k:d.

Relaxation of the bolt tightening on the flanges was observed to occur kyr{

cith cach cycle to high temperature and cooldown.

It was found possible to operate the pumps without leakage by using reduced pressures while at low temp-R i craturas and not applying the full 2000 psig pressure until the SOf F tempera-h tura t'as reacheo.

Seal welding has not been required by using this mode of l operation.

The majority opinion on the pump gasket leakage is, that a complex W

combination of thermal expansion forces both linear and in bending are causing th2 gasket flange bolts to exert higher than yield forces on the flange faces h@

cs temperature increases, if bolts are tightened to the manufacturer's recom-mended stretch in the cold condition.

[5 The eight main coolant stop valves and the four main coolant check v31vcs showed no leaks at their flanges in either hydrostatic or operational h

t sting. Stem and packing leakage on the stop valves has been very low in the k

open-backseated condition, but high and erratic in the closed condition.

The parallel-disc step valves hold high pressures across their discs with very low Ic:k:gs. At lower pressure differentials, leakage has a tendency to increase.

M Difficulties in opening the stop valves were encountered under certain pressure Q

conditions due to a trapped high pressure condition in the valve bonnet cavity.

D'y varying pressures on one side or the other of the two parallel discs, or by h4

< rre Esing Limitorque motor slugging, valve opening has always been possible.

Main soolant check valve slam noise was slight during hydrostatic and flush testing y

ith:ut the core in place when one of four operating pumps was stoppcd.

W

However,

,f ith the core installed, slamming became more noticeable upon stopping one of hr:e cperating main coolant pumps. The slaming, upon stopping one of two S

i%

sperating pumps was loud but not felt to be a source of possible damage.

There-df cre, the operating procedures, which detailed the closing of stop valves before Check valve backflow, W

.tcpping pumps, were justified.

en hot standby with a stopped pump, was designed for 100 GFLto keep an isolated loop g

fcund to be insufficient and a hot standby loop must be started intermittently in This flow has been Q

erder to keep its temperature at the desired level even with the steam non-return 6

falvesclosed. The 5" by-pass and I" loop relief valves have operated satisfac-4

<orily.

g I

p The process instrumentation of the main c)olant system has performed Q-reasonably well.

Accuracy of the instrumentation is within expectations, and

@y intenance has not been excessive.

The use of the steam generator pressure jrop for flow readings has worked satisfactorily.

The narrow range loop resist-y Q

ace thermometers are capable of showing core flux tilts as small as +5% of full Loop pressure instrumentation has been satisfactory except for occasional p

ower.

$f Drctic readings on the readout indicators.

This does not appear to be a sensing Bement difficulty since alam: and scrats are not actuated.

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7 11 M

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s 3.

Pressur* Control and Relief System f

Compon:nts cf this system wero casily placed end th:re c:re no Mi field crection or welding difficulties.

In hot flush operational testing, y#2

,cithsut the core in place, the two self actuated safety valves failed to reseat j

i properly after testing for proper lifting pressure. Field readjustment of the d

h(!

valvas by the manufacturer's representative resulted in acceptable pressure blowdown (approximately 150 psig) on the third test. Minor leakage and valve 1

l cicsing difficulties were experienced with operation of the solenoid relief P

v31v2 cnd the solenoid surge spray valve. These values were later maintained l cith sstisfactory results. During power transient testing (including a 60 W M

M.h gross electric load drop without reactor scram) surge spray valve operation h s be:n sufficiently offective so that no solenoid relief or safety valve operation has been required. The circulation spray valve has sufficient capacity 8[f to provide long term equilibrium conditions of pressure, temperature and boron conc:ntration at about one-fif th of its full valve travel. Heater consumption O

cvsragesapproximately65kwh/ hour,butnoattempttooptimizecirculationspray d{j his y2t been made. No electrical heaters have failed. Safety valve quenching in th2 Low Pressure Surge Tank worked well during the safety valve tests.

y

-a Process instrumentation of the Pressure Control and Relief System d

has performed satisfactorily. The water phase thermocouple is, however, located M

too low in the pressurizer. During periods of little circulation spray, water fi temperatures above the heaters are not reflected very closely. Operating pro-y cedur:s for overpressure on the main coolant system mere, therefore, modified Qi to cecount for this condition.

it, t

.4.

Charging and Volume Control System

%N Charging pumps were used as hydrostatic test pumps for many of the y,

'high pressure hydrostatic tests, operating up to 3435 psig pressure without V

difficulty. During core loading, the charging pumps were subjected to an ex-Q tended period (approximately one month) of operation. This consisted of recir-culating 160 F - 12% boric acid solution. This duty resulted in the failure

'4 ef coveral Armco iron pump suction flange gaskets. These gaskets were replaced

?

cith Teflon gaskets. One of the three pump plungers has been replaced and some 4

plunger glands have been repacked. These pumps have performed well, considering ji

, tha rather severe duty they have been subjected to.

y l

The other components of the system such as the Feed and Bleed Heat h

Exch ngar, the Pressure Reducing Orifices, the Low '>ressure Surge Tank, the Low g

Pressuro Surge Tank Cooling Pump, the Low Pressur-2rge Tank Cooler and the y

Low Pressure Surge Tank Make-up Pumps have perfozr.

very eatisfactorily. The ad~

jmoter-operatedvalvesinthesystemgaveconsidera' Mouble, both with the h

Imoter operators and within the valve body itself.

a.. motor operators had not M

.been prcperly locked, by the valve supplier, into the assembled condition. Valve 4

discs came loose from the stems because of improper locking techniques.

I During low power physics testing, the motor-operated root valve on

'l thm bleed line failed to close properly, requiring a complete plant cooldown to make repairs., Fortunately, the core had no decay heat and plant activity 1cv31s were low. Repair of the detached disc was accomplished by draining the cater in the reactor vessel to an elevation just above the core. The low power physics testing also demonstrated that the isolated loop demineralized water a

fill line manual valve should be motorized for greater safety in boron dilution A

operations.

M

_ _ _. _ _ _ - - - ~. - -, - -

34 kd The charging and volume crntr:1 system has met its function:1 r2-M quirements. Water charging cnd removal,.ictegrated into pr;ssurizer level b

control by use of the variable speed charging pumps has worked well, with 3

pressurizer level being maintained, with the exception of reactor pcwer tran-sicnts, to within ff". The boric acid dilution, the chemical addition and the y

M hydrogen gas addition functions of the system have also performed satisfactorily.

Pr:ssurizer cooldown, by use of the cross-connection by-pass to the pressurizer M

spray, has been found to be very necessary during a com; 2ete plant cooldown.

A socket weld leak developed in this line after approximately 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months at h

pr;ssure and temperature resulting in a complete plant cooldown, since isola-h tion for repair by freezing was not possible. The leak was attributed to a faulty weld with a small fissure which allowed minute leakage to cut a path h

cver this period of time.

g}

The system's level, pressure, and temperature instrumentation has h

perforced very satisfactorily. The charging and bleed flow indicators and in-S ttgrators have not, however, performed to their intended design functiong n:mely, to assist the operator in determining main coolant leakage. One DP jN csil has been relocated to improve flow readings but resulting indicated flows Q

da not balance.

A complete flow balance is probably impossible since slight rf Icakage of main coolant into drains through valves, etc.

(presently about

$q 1 cubic foot per hour) is beyond accuracy of the flowmeters. The small leaks W

that have developed in the high pressure systems within the vapor container hava thus far been discovered by routine inspection. Observation of the remote 2

rolative humidity indicator has shown that detection of leaks may be possible fM{

by this method. The vapor container compartment sound pickup system has not been effective in detecting small leaks.

,{$

%W

?1 V[

t 5.

Chemical Shutdown System

%b This system provides boric acid for cold shutdown of the reactor

/p and for partial removal of boric acid by ion exchange.

In cooldown of the gd reactor to ambient temperature, cooling rates are limited to 50 F per hour.

Therefore, under normal operating conditions, the chemical shutdown system lg W

need only be a slow acting control system and is not intended for emergency

((V transient conditions.

i i

During early operational testing, this system developed a partial l

stcppage of its connecting line to the charging pump suction header. This was due to the fact that the 12% boric acid solution is insoluble below approxi-g mately 130 F resulting in partial solidification in the connecting line. Addi-b tional heat tracing was added to the connecting pipe line which resulted in satisfactory pumping operation during the periods of 12% boric acid injection.

W Operating procedt.:res must be diligently applied, however, in order to maintain

'N the concentrated mix at desired conditions and to insure that cool feed lines, Q

through which the 12% solution has circulated, are thoroughly flushed.

p1 System equipment, such as the Beric Acid Mix Tank, the agitator, transfer pump, chemical shutdown ion exchangers and connections, piping, valves p ;

y+

and process instrumentation has performed satisfactorily. No deborating capsules hava, as yet, been exhausted or disconnected from the system. Only one complete d:borating run has, however, been made since core loading.

.. ~. - -

il

i 35.

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$N 6.

Pp ifI-.

x. Sys t*m gT FM This system has performed well. Minor leakage difficulties have Ik N

cccurred with certain valves. New valves are on order to replace those where

[Q lark-tightness is icportant to operation and maintenance. The system embodies c 1000 GPM Low Pressure Surge Tank Pump. Operation of this pump has not, how-cvar, been required since the two 100 GPE canned Purification Pumps have been

g c: pable of meeting dilution cooling needs. The two Purification pumps have E

operated perfectly with no maintenance required to date. The resin capsules D

hav2 performed well.

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

Y?

Tha disconnecting and' sealing features of the system's resin capsules have not f

be:n used as yet, since no resins have become completely exhausted. The ion

{

cxchange pit has, to date, oeen operated dry without water cover over the cap-M sules. This has made possible study of activity buildup of the capsules while 3

in s;rvice-The system process control was found to require several minor Sh modifications.

"h p

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(Oe 7.

Component Cool:no S < stem This system has performed satisfactorily. Temperature through the Ik system have bgen low because the raw water supplied to the heat exchangers is v2ry cold (34 F - 60 F). No cooling capacity problems are evident. Water condit'oning of the circulating component cooling water has not been a problem.

Some dif ferences in temperature are evident in the static water of the neutron m

shield tank which is serviced by vertical heat exchangers of the component cgol-(g ing system, At the top, temperatures gre being measured in the range of 120 F,

y while the bottom is in the range of 80 F.

No leakage has occurred either into 4

(from the main coolant system) or out of the component cooling system. No activ-b ity has been detected in the system by the radiation monitor.

It has been found d;sirable to add anti-sweat insulation to the piping of this system because of g

low cperating temperatures and high atmospheric humidity.

y y

. r Ikr 8

Primary Flant Corrosion Control Syster

[

i This system has also performed satisfactorily.

(See Section V. -

[}

I't l Tests and Measurements, paragraph 4)

The concept of pressure letdown on'the

-(g main coolant bleed water (2000 to 15 psig) and resulting maintenance of 25cc/

liter of hydrogen in the main coolant has proven to be very acceptable. During early operation, nitrogen was used as the blanket gas in the Low Pressure Surge f

Tank. This was done in order to avoid any hydrogen problems which might result g

from pipe line failure or leaksin systems not completely tested. After comple-k tion cf early testing and low power physics testing, the hydrogen corrosion b

centrol blanket gas was injected, M

, mM w

Ln 9

Primary P1 ant'Samplino System Qx 147 The location of the various sample lines have been found to be more id[.

thin edequate for chemical control of the main coolant. The location of the samplo cubicle has posed no difficulty.

.f'

_i t

36.

P Errstic boron gnalysis tas cxperienced toward the end of the lor power physics test program when boroa concentraticas were in the 50 to 100 ppm j

It was later found that sampled low boron water was r'cking up addi-range.

j tional boron that had deposited on the Low Pressure Sample cooler piping.

3 A number of minor piping modifications were made in and adjacent to the sampling hood. These included changes to allow for proper sample sink drain-cge, addition of needle valves so that better regulation of sample flows could p

be accomplished, modified sample lead-in piping and installation of a crud probe.

k The sample line froc, the ion exchanger effluent header terminated at a comon simple header which had been found to contain sufficient deposited activity to g

?

affect the ion exchanger effluent sample. This sample line has been modified i

in order to bypass any contaminated lines.

yg The sampling system is now adequate, but further imp.ovement is 6

possible by use of (" tubing throughout, rather than }" pipe. This reduction h

would reduce quantities of fluid necessary for flushing to a representative sample.

(Q

10. Radioactive Waste Disposal Systee The workmanship in the fabrication of the Waste Disposal plant was excellent.

Weld leakage and piping misplacement was at a minimum.

The equip-

}

ment associated with the various tanks has worked well and the movement and storage of the various waters has proceeded without difficulty.

3 The waste gas T

system was purged and backfilled with nitrogen with only minor difficulties.

The gas system has been on automatic control for approximately eight months and has operated satisfactorily.

The stripper-evaporator process has been the major source of trouble in the Waste Disposal System. During the operation of the electrode evaporator, stoichiometric quantities of hydrogen and oxygen from the electrolysis of water are released into the steam phase. A portion of the steam was used for stripper stripping steam. It was not realized for some time that gases were being gener-eted in the evaporator therefore, high oxygen analysis of stripper effluent seemed to indicate poor stripper efficiency. Consequently, the stripper tower l

ens disassembled and the major portion of the sieve tray area was blanked.

7 When stripper effluent gas content did not improve, further investigation revealed gis producing characteristics of the electrode evaporator.

q t,

Gas production in the evaporator raised many questions of importance.

A testing program was evolved and temporary lines installed to handle the evapo-t rotor off-gas. The steam line from the evaporator to the stripper was blanked and building steam was installed in its place. This imediately improved the stripper effluent and solved tne problem of charging volumes of hydrogen and oxygen into the waste gas system. By maintaining a vent on the distillate accum-j ulator, oxygen in the steam distillate was reduced to 0.5 ppm. Corrosion speci-mens have been installed to determine if there is a possibility that stress cor-l resion is present in the evaporator shell. No information has been obtained as yet from these specimens. A temporary activated charcoal bed was installed in the off-gas line when a test involving tracer radiolodine in the evaporator bottoms indicated the carry-over of small amounts cf iodine.

o The charcoal bed M

removed all detectable amounts of the radiolodine. Permanent design o' the

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l

37.

L

(

cvaporater eff-gas system is cocpleta cnd includ;s a bed cf activated charcoal.

O In addition, the system will be piped to the stack through the suction of the S

primsry auxiliary building fan.

(

r The cement feeding arrangemect for the drunning station was not h

satis factory. The cement bin location will be changed and feeder equipment i

r instelled.

s i b l

The waste disposal system has not yet made any long process runs due to the above listed difficulties. However, the changes that have and will l h be made to the waste disposal process should provide an efficient, workable system.

$r h

11 Shutdown Coolino System h

r This system has worked well as a process system, but has experienced i

cin:r component troubles. The flow and heat removal capacity of the system ex-L c;cds the design by approximately 50L Several valves were found to be not suf-i ficisntly leaktight across the discs to allow nomal operating flexlbility.

i ThIse valves were declared basically deficient after two attempts had been made y

to rapair and reseat them. Replacement valves were ordered from another supplier.

K The Shutdown Cooling Pump tripped on overload after approximately one month of operation. Disassembly showed a defective bearing and pump misalignment. A new

' r bearing was installed. During the repair, the I.ow Pressure Su:tje Tank Cooling

' [

Pump was used for shutdown cooling service.

F.

N 5

C 0

12-Primary Plant Vent and Drain Syster N

f This system has operated satisfactorily.

Isolated loop venting has j

not, however, been completely effective but pump operation has not beer. affected.

k ph Draining has been satisfactory. The use of the drain system and a l

charging pump for controlling isolated loop heatup or cooldown has resulted in l

rather rapid pressure transients. Thus far, isolated loop operation has been more frequent than anticipated and censideration is being given to modifying the j

v nt and drain system and sampling system in such a way as to provide for a smooth 3

i throttling arrangement for effecting isolated loop water expansion, contraction, rj l

and dilution of boron.

t L

There are many small manual capped valves in the vent and drain is system. The caps are screwed on and are sealed with an "O" ring seal. Packing V

leakage on these valves was sufficient in early testing to fill some of the hand l

tight caps with borated water resulting in thread corrosion and seizing of the t

caps. Operation and maintenance surveillance then became difficu.tt. A decision J

tas made to remove all caps in order to obtain better maintenance surveillance.

f A change in packing material was also made resulting in approximately zero pack-ing leakage or the valves. Caps are now used for emergency only; that is, caps

I cre placed on valves if some unusual valve packing leakage occurs and repacking

. j ct temperature and pressure is not possible due to location of the particular valv2 in the system. Two such instances have occurred.

g

. _. a -

l 38.

During construction consitreblo difficulty as encountered in making the N;utrcn Shield Tcnk Icck tight. Plcnt d2 sign had provided two Neutron Shield Tank leak monitor telltale drain lines, one from inside the Neutron Shield Tank and one from outside of the tank. These drains proved to be most valuable in the final determinations of the Neutron Shield Tank leak tightness.

13. Sa fety Injection Syster l

This system has operated satisfactorily on initial tests, and in subsequent periodic tests. Some vibration problems were experit.ced on the Safety Injection Pumps, but realignment improved conditions. The pumps are cctually high capacity pumps and when throttled for recirculation er test, their noise level and vibration is greater than is realized :t flow capacity.

14 Reactor Control Syster This system arrived at the site as part of the nuclear control board and as such had been partially wired and tested at the factory. The interconnecting wiring between the temperature detectors, rods in and out relays and the servo control was trouble-free. The control is normally used under steady state conditions and has performed very satisfactorily. When temperature adjustment is required, the control initiates proper rod action without hunting er overshoot.

Several difficulties were encountered in the installation and opera-tion of the control rod drive mechanisms. The position indicating coil tubes cere undersized resulting in a great deal of extra time and effort in removing these coil stacks when the need arose. New, larger ID tubes have been installed.

It was found that during the initial operation of the drives, use of the pulldown

coil, af ter the rod had bottomed, could cause locking of the latch mechanisms.

One drive mechanism pressure tube was removed and the mechanism fingers released by hand manipulation of the drive shaft. Because of tolerances and spring char-ceteristics within the mechanism, the original operating procedure called for continued run-in of the rod to insure that all rods attain the same bottoming position. Revised operating procedures now provide that rods-in motion cease when the rods reach the full-in position. Under this condition, the most that the rod positions can vary is one latch position.

Coil troubles were experienced on several occasions. Several coils became grounded. This was attributed to moisture, resulting possibly from a l

poor control rod venting arrangement or to a leak in one of the control rod mechanism seal welds. All coils were removed, dried, re-varnished and rein-stalled. A stationary gripper and lif t coil also failed. Examination revealed Q

that the failed coil had grounded and the retaining coils were all operating above normal temperature. The air cooling system has been modified, and a t

modified coil cesign is now in manufacture.

I n._ --

One centrol rod driva mechanism had been essembled incorrrctly l

cnd on occasions would not drive-in properly. This mechanism was removed and returned to the manufacturer for proper spacing of the parts. The de-fcctive mechanism was capable of scraming at all times. Several rods have ll had erratic operation during their run-in, although all have scramed properly.

l This was corrected by changing the programing cam switch settings slightly.

The primary and secondary rod position indications have performed vary well. Initially, the secondary position odometers were occasionally 3

fcund to be off by a 1/8" step. This error was corrected when it was dis-1 covered to be due to the manner in which the odometer was reset. The secondary position indicators provide a reliable and accurate measure of rod position.

It has been found from operating experience that the brake settings en the cam switch motor drives need to be adjusted on a regular maintenance I

schedule to insure proper stopping of the cam switch on completion of a drive cycle.

IS. Nuclear Instrumentation and Reactor Protection System This system has, for the most part, performed well. Prior to initial criticality several problems were encountered with the source range instrumentation. The original Westinghouse BF e unters, type WL 6307, were 3

not sensitive enough tc register source counts through heavily borated water.

These detectors were replaced with WL 7087 BF s which increased the sensi-3 tivity by a factor of 10. With this type BF, the counters registered approxi-mately seven counts per second.

Inearlyop$ ration,noisepickuppresented some difficulty.

Connecting the outer cable shield to the BF detector housing andinsulatingthedetectorhousingfromgroundreducedthisp$oblemconsiderably.

Placing capacitors across contacts found to be causing noise and reducing the discriminator voltage setting on the BF 's essentially eliminated the interference.

Cableconnectorshavecausedminordiff$cultyinseveralinstances. Replacing or reassembly of connectors have cleared them for operation.

One CIC, af ter a short operation period, began to give intertnittent, erratic signals. Examination of the detector showed that it had apparently been da2 aged during installation. The three auxiliary meters on the CIC detector leads, located in the Nuclear Instrument Cabinet, did not cover the full power range. To correct this condition a 15,000 ohm potentiometer was placed across j

the meter enabling them to be adjusted to span the full power range. The three

)

Nuclear Control Board powr range meters were too sensitive and would have gone off scale at about 80% power.

One magamp winding was, therefore, removed from the circuit to decrease the sensitivity and allow the meter to cover the full rcnge. To date, the scrac set points have shown no appreciable drift.

A slight zero drif t in level occurs on the source range and intermediate range I

instrument, but this is corrected in going through the pre-startup check-off list.

A circuit designed to afford protection in case of a rod drop or l

loss of signal or power supply to power range meters has been incorporated into the power range nuclear channels. This modification consisted of replacing the three power range indicating instruments with three meters having adjustable, j

low alarm contacts. The contacts alam on a decrease in flux and will energize t

c relay to cut back the turbine load limit device to 78% of full load.

The

l e.

[

loss of valtage er Ices of signal cill cau:e ths metsrs to go to_zaro, whils a rod drop will cause an instantaneous flux deprension. In either of these circucatan:es, automatic load cutback will occur, if above 78% load, and an f

annunciator window will indicate the condition. White lights supervising the circuit between each control rod scram breaker trip coil and the turbine generator permissive relay have been added on the nuclear control board to pro-

[

vide indication that the turbine-reactor trip circuit is ready to receive a i

trip signal. The scram breaker indicating lights were changed to monitor the rircuit'between the three manual scram push buttons and the breaker trip coils l

to provide indication that a manual scram circuit is always available.

V Several sultches were moved from the Nuclear Instrument cabinet '

l to the nuclear control board for better operating convenience and many items T

were mod fied slightly for improved operation and accessability. These changes i

are listed under Section VII. - Design Changes.

5 f.

16 Radiation Monitorina System This system was initially calibrated by exposing the detectors to known amounts of radioactivity. Results of this calibration showed the detectors t

to be at least as sensitive as had been predicted.

?

During the initial stages of operation, noise pickup by tne electronic

[

circuits gave a number of false readings. This noise was created by the mechan-ical failure alarm on the air particulate detector setting up a pulse when oper-

[

ating near its set point. A capacitor installed across the input of each channel j

corrected this problem. The natural background increases with rain, possibly due l

to an increase in radon activity or fallout in the rain. As a result the alars

[

set point on.the stack detector had to be increased to compensate for this change in natural background.

i f

Depending on conditions in the vapor container, the air particulate f

detector gave readings approximately equal to maximum permissible concentrations

[

for unidentified isotopes. This was attributed to radon daughter decay coming j

from the thermal insulating materials. A thin aluminum foil was placed over the

(

scintillation crystal to eliminate the alpha pickup by the detector, thus reduc-S ing the high background reading. Starting the air pump f.ir air flow through the I

air particulate monitor created a temporary overload on the AC instrument vital f

bus. The power supply te this pump was changed to a higher capacity bus. The four steam ganerator blowdown moritors located in the sample cubicle respond to 3f direct radiation from the sample lines while sampling primary water. This is

{

presently under study to determine whether additional shielding or relocation of the detectors is the best solution. During sampling periods the stack moni-tor reads high due to the release of A-41.

A scintillation type monitor has r

L been purchased for continuous sampling of the incinerator stack. This monitor c111 read-out on the radiation monitoring panel.

The portaole and semi-portable monitoring equipment has had many e

{

oinor difficulties. The majority of the troubles have been due to rough hand-i 3

[

ling of the equipment and have been cured by education in the proper handling i

j cf the equipment. A gamma guard monitor has been installed in the corridor of i

the primary auxiliary building leading to the charging pump cubicles. This is-to alert the nuclear auxiliary operator to radiation levels in this area before entering the area for his normal inspection tour.

---~

t L,

m, _ _

41.

g t'

ird As an item of interest, on the first day of the hunting season, Y

an outside contractor's employee set off the portal conitor clam when enter-N ing the gatehouse. Examination found the source of radiation to be originat-ing from a radium dial compass. Further investigation showed the compass to

= 2 haveagrossbeta-gannaactivitywiththecaseopenof17.5mr/hrand8mr/hr

)

with the case closed. The owner turned the compass over to Health Physics for y

disposal.

I n

17.

Vapor Container Atmosphere Control Systems uQ The ventilation system to outside atmosphere is used only when K

the reactor is sub-criticag and the main coolant pressure and temperature are below 300 psig and 200 F, respectively. Ventilation requirements, to p

u date, have consisted only of supply of fresh air for maintenance activities.

Radioactive dilution purging of the internal vapor container atmosphere has y

not been necessary since air activity levels have remained low.

?

Q)

The four fan units, providing recirculation, heating and cooling of the vapor container atmosphere, have been operating satisfactorily.

H h

The leakage testing features, whereby the vapor container is operated at a slight positive pressure, have not yet been evaluated. The h

number of necessary entries into the vapor container for minor maintenance and the season of the year have made leakage rate detemination difficult.

p#

The increasingly colder meather has caused temperatures near the outer skin of the sphere to drop, resulting in slight negative pressures occurring at 9,

night. Evaluation of vapor container leakage rate systems will be attempted PI when stabilized conditions are eventually realized.

-dv1 Operational tests of the internal air filtration fans have been k

made. Air contamination levels have been low so that no filtration has been W

required.

d e4

.t8

18. Fuel Handlino System 1

4 This system is rather complex due to the elevated vapor container construction and to the uncertainties of viewing and positioning fuel and core 4

internals through forty-foot depths of water. Fortunately, first core loading

'[

was accomplished without water shielding. Late delivery of certain system com-g ponents, limited operational check-out time to a minimum. Minor adjustments

(

and furthat component checking required access to the normally underwater equip-j{

ment.

i The new fuel storaoa vault was found to be rather small, particu-larly when used for new fusi inspection. Humidity and water dripping prcblems 3

occurred in the vault. Heat and louvers were installed in order to correct 4

this condition.

5

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._.--~~- -~ ~ -..--+- _.-~._ _ - _ -.-.---- -

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I gd:p Failure of the spent fuel trcnsfer pit manipulator crar.e bevel I$$

driva gears occurred. The jccking mechanism pume had several failures of 4pig prossure switches and problets with excessive noise ar.d erratic pressure regu-

$3 1ction.

Several modifications were made to the puep assembly. The spent fuel stcraga r:ck was found to have incorrect dimensions for storage of control rods y

O rnd shins.

This error was due to improperhandling of superseded drawings. The y

positicr.ing of the spent fuel pit manipulator crane by the target system was y

unsatisfactory.

Icprovements will be made before undemater handling is attempted.

pv ph The winch equipment, which raises and lowers the fuel chute carriage h

car thr: ugh the fuel chute, gave trouble throughout core loading.' The speed dE contrc11er was modified and operation was improved, but did not meet specifica-The springs, pressure switches and other parts of the winch gave consid-

[d tion.

er:ble trouble and contributed to erratic performance. Nevertheless, handling UT of fuel by the winct, cable, chute and carriage system appears to be sound in h

c':nce pt.

d, The shield tank cavity manipulator crane performed satisfactorily o

with the exception of failure of several bearings on the main boom. Trolley

,g%@

and boom speeds and controlability were satisfactory. Use of the pointer and grid system was found to be generally satisfactory. While the underwater tele-risicn camera was not used for the first core loading, television may not be ncessary as a fine position indicator.

gg The universal handling tool was found to be improperly assembled.

.h

[t ras necessary to correct this condition before proceeding to load fuel.

storage racks, target systez.s and control rod assembly fix+ures require improve-g ant before underwater fueling can proceed smoothly.

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l In spite of the many system component d'fficulties experienced php.

luring initial core loading, no fuel elements or control rods were jarred, Q

iropped or in any way bumped or darraged. This is attributed to the basic fail 3fe features incorporated in the system.

w L%

95 ks M

9 Main and Auxiliary Steam Systems Erection of the main and auxiliary sys tems steam piping was carried l9}

o completion with the exception of the final connections to the steam generators, i Q hise connections were not made until cleaning had been accomplished by using h

h2 boiler feed piping and a tempority connection to the condenser.

N 6e The systems were then given a hot alkaline flush and an acid wash y using the condensate pumps. During this wash, the safety valves and boiler M

eed pumps were blanked off. The piping was then drained and rinsed and left Q

ith en internal nitrogen blanket. This was maintained except during final h

annection stages.

M

,p Upon completing the connection of the main steam piping to the G

, 0;.;t

=am generators, a hydrostatic test was performed revealing a crack in the g

>nnectirg pipe weld at No. 2 steam-generator. A complete review of the weld-ig procedures and the metallurgy involved was then undertaken resulting in welds being ground out, rewelded, stress-relieved and X-rayed.

ig 1 fcur No

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'fccts tere noted when the system was again hydrostatically tested.

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V Safety valves core cdjusted at a convrntional steam generating slant under supervision of the manufacturer's rapresentative. A Yankee engineer I

-nd an insurance company represintative oitnessed the test.

t The main steam stop and non-return valves were thought to be defect-l0 vs when apparent leakage was experienced across the seats with steam pressure I

cting in the forward direction. Manufacturer's instruction in their operation eve proven valves to be leak tight.

b The steam supply line to the steam jet air ejectors had etcessive iressure drop when the primary plant was at partial cool-down. This condition

[

as corrected by increasing the pipe size from 2" to 3".

Small steam leaks have developed, with time at temperature, at

-alorimeter connections and small valves along the main steam header.

Isola-icn and repair was possible, for the most part, by closing of non-return valves, hereby preventing reactor cooldown. All other fittings and components of the team system have been satisfactory. Examination of the throttle valve strainers

{

nd high pressure turbine indicate that the cleaning procedures teere adequate.

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'O.

Condensate and Feedwater System These systems were chemically cleaned during the same period of

' sin and Auxiliary Steam Systems cleaning. Strainer., were provided in suction ines in order to prevent pump damage from large debrit,. Condensate and Boiler eed pump motors were disassembled and cleaned. while Nos. 2 and 3 Boiler Feed 4

> umps required coupling adjustment to reduce vibration. These systems have y;

iperated satisfactorily in all respects to date.

j

[

h 21. Circulatina Water System h.

5 The capacity of the circulating water pumps (and electric motoi

l oad) is slightly greater than design. This condition does not appear to be F

,erious, so pump removal and inspection have been deferred until the next ex-

_ ended plant shutdown. The condenser water box air removal equipment operated

-orrectly, but points of air removal and draiaage were modified in order to increase efficiency. Several pipeline sizes were increated to reduce pressure r

l osses.

t The concrete portion of the discharge piping also required modifi-

stion. A catisfactory vacuum could not be obtained due to air leakage at ioints along the concrete section of the line. This further increased the re-z ired pu:::p power since head recovery was not possible under these conditions.

u

-he concrete pipe joints were regrouted but air leakage was still experienced.

=he grout was removed from the joints and the steel rings were welded.

(Joint ietail included a metal ring and rubber gasket.) This reduced air leakage but lid not entirely eliminate it.

A complete steel liner was then installed nside the concrete pipe resulting in satisfactory performance.

The rotating screens have operated most satisfactorily.

Tw 3d 7-

__ ___:_ji;f

~ _ _ _.. _ _

_-._._.__,___m_

R2. Compressed Air System gy The compressed air systec, composed cf th( instrument air End k;

servic2 air systems, saw an early completion date and was used during the final stages of construction. System cleaning was accocplished by the blow-M@i ing cf all lines.

The instrument air system has operated trouble free since November cf 1959. A minor revision was made in their controls, however, which provided 6

f:r a choice of the standby compressor. The service air compressor, operating

_j#

St high2r than normal temperatures, experienced valve fouling due to the in-ifE crassed cylinder lubrication temperature. Upon inspection, the piston was found to be cracked and the cylinder liner was scored. A new head of improved 3

d; sign,as well as other new parts required to rebuild the compressor, were sup-it plied by the manufacturer. The discharge piping to the cooler was also modified y

froc a 4" to a 6" line, thereby providing a greater surge volume. Since comple-V

'tien of this work, the nation service air compressor operation has been satisfactory.

M 4?

4 ~M hv id 09 23 Electrical Systet jg no

}Mg The basic design concept of this system has proven to be most satis-fcctcry and system cperation during the initial period has been excellent. No difficulties of any consequence have beeri experienced with the generator, main

.p

@5d!

transformer, station service transformers and in-plant station switchgear equip-ment. The arrangement of the station service system including the connections to external sources of supply and the arrangement of duplicate or complimentary M

equipment on various station busses has also proven to be very satisfactory.

The inverter equipment has functioned reliably, maintaining the design regulated 1%

oltage. AC vital bus operation has also been satisfactory. Minct revisions fy sera made on several electrical controls - (See Section VII for listing of p

these revisions): The control and protective features have, however, been g

fcund to be adequate.

fg 0m Wh 24. Reactor Vessel Installation of the reactor vessel was very rapid due to the W{g (f,

elsvated vapor container construction of the plant. A direct lift was made from the railroad car, positioned under the vapor container, to its final g

locatien. The reactor vessel and main coolant system were hydrostatically field tested at 3,435 psig three times in order to prove welds and other parts d

)f tha system. No leakage has ever been observed from the main closure (self pg energized dcuble "O" rings). Leakage may be determined by the use of the bsi bakag2 telltales. Seal welding has not been necessary although the seal reld ring has been placed on the vessel. Field closure of the vessel has Q

>een ccepletely made twice.

Q~

q Two areat, of dif ficulty have been experienced with the closure Ltuds. Difficulty occurred with removal of the heac studs upon completion

$3 pf tha hot flush test. This removal was made to install the core and vessel ip Me WS AohN Rb 4s sM - M& 4 i

wt_$ms.m ee Te

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nt2rnals. Considarrble forca cas required in crd2r to remove the studs and

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ylling and defomation of the silver-plated threads, particularly at the top Df the threaded stud, was observed. Most studs were returned to the manufac-i lurar fer reworking and replating of the threads. The manufacturer claimed 7

lhet poor field cleaning of stud holes was responsible for the galling. This j

Is disputed, however, and appears to be inconsisteat with the area of maximum

$amage to the threads. The thread lubricant was questioned and, as a result, p

Zs ch;nged. The stud hole cleaning solution residual was felt to be compatible 4';

bith the thread lubricant. The point to be made is that two makings of the blesureforcoldtestingresultedinnothreadproblems,whilegallingoccurred a*

bitsr th2 vessel had been heated to operating temperature and then cooled down.

k i

Stud interference occurred when the normal vessel bolt-up was made pithth?corereactorinternalsandinternalalignmentpinsinplace.

The V

pntarnal alignment pins caused a slight rotation of the vessel head counter-

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clockwiso from previous bolting and syneetrical hole positions. As a result, e

, t bectme impossible to insert several of the studs due to an intefering

, -b i

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should2r on the studs. This was corrected by remachining of the stud shoulder.

The stud hole plugs (used for keeping stud holes clean and preventing them from E

fillingwithwaterwhenrefuel*'g)werefoundtobedifficulttouseandwere nodified by machining.

t

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t T

t 25.

Radiation Shieldinq

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[d Gama radiation surveys made outside the vapor container have demon-rated the adequacy of the primary and secondary concrete shields which surround e rocctor plant. There was no indication of radial radiation streaming, and nce no indication of voids in the concrete shielding. Gama levels at working H

etaticns and intemittently manned locations at a reactor power of 120 Ime are f#

eonsiderablybelowthedesignvaluesof0.75mr/hrand2mr/hr,respectively.

.Ti Neutron flux levels outside the vapor container required the instal-R(

Baticn of mascnite shleiding over the neutron shield tank to correct neutron iy otreaming through the space between the reactor vessel and the top of the P

lcieutron shield tank.

(See Section V. - Tests and Measurements, paragraph 2.e.)

Dose rates through the auxiliary shielding in the control room, 4

grimary auxiliary building and waste disposal building have been close to back-

}

grcund levels. Dose rates through the concrete and water shields for the fuel R

Gr:ndling facilities will be determined at the time of refueling.

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Th2 ceneratcr cnd cxcitsr hava perfcrmed perfcctly up t's the 120 m gross clcctric load. The turbin2 cnd throttle valv:s hava, how:v;r, cxpericnced Q

cifficultics.

g y

During loss of load transient tests at 15 and 30 m gross electric, (2

high moisture separator water levels were experienced due to non-existance of

, (4 th? moisture separator-heater drain receiver vent lines and then due to the

$@f small size of the subsequently installed lines. These vent lines were reinstalled es 6" lines and check valves were added in the drain lines from the moisture t

s parator to the drain receiver. Performance of the loss of load transient test frac 60 W gross electric showed that build-up of moisture separator water level

@@E to a high level had been corrected and that a ramp unloading rate of approxi-D l

mately 10 W/ min. was possible without high level signal and scram.

l we hcg While performing tests in the 30 to 40 W gross electric range, governor regulation became increasingly worse. Shutdown of the unit was orderly, h#

olth later investigation showing that No. I control valve servo-motor operation b

tas erratic. The servo was disassembled, stem cleaned and a newly fabriccted bushing installed.

P#

7 The scheduled 90 W grost electric run experienced turbine diffi-culties which later caused some twelve days delay in the testing program. After Cry cpproximately 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months of operation at 90 m, excessive turbine shaf t vi~ oration W

indication appeared at the No.'l bearing. It had been noted that the shaft i

tas riding approximately 20 mils higher than expected at No. 1 bearing.

In an cffert to correct the vibration, it was decided to raise the turbine high pres-D19 l

sura casing 0.007" at each corner. Return to operation showed no improvement M

in shaf t vibration. Various analyses indicated that an " oil whip" might be lifting the shaft. As a result, Nos.1, 2 and 3 bearings were reoved and l yi the necessary circumferential grooves machined.

No. 4 bearing was also re-

,b l

moved, inspected, and reinstalled since indications showed no modification was p%

n:cessary. The following power run proved that the most recent modification g

h:d not solved the vibration difficulty.

Investigation of control valve open-Q ing sequence and their effect.on shaf t vibration at various loadt was next l

carried out.

It was concluded that the shaf t was being lifted by opening the

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tro lower control valves first, to the extent that rubbing eventually occurred.

g l

K While the investigations of control valve opening sequence and Q

their effect on shaft vibration at various loads was being explored, a sudden g

lo:d drop from 60 to 10 m gross electric took place. The turbine was manually g

tripped by the manufacturer's representative, but the No. I turbine throttle pe v21va did not go fully closed. Subsequent disassembly showed that the poppet M

valve had remained partially open. This was caused by restricted motion be-y' twe:n the valve disc and stem. A teeporary repair was made allowing further n.

turbine test work to proceed.

The manufacturer's analysis of the excessive turbine shaft vibra-

' d g

tien antinued. It was concluded that a " steam lif t" was occurring as a result cf partial admission of steam to a lower quadrant of the turbine, and the fact ytN that insufficient space had been provided in the desion to allow the steam that Mih had passed through the impulse stage to distribute around the full 360 of the rotor before entering the r eaction blading. As a recult of this restricted space, there was a large unbalanced pressure across the rotor resulting in a

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force which lifted the rotor.

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A maximuw loading cf 65 W gross c1;ctric uas cstablished cnd l 44 i

continuous operation was maintained f r cpproximat21y cleven d:ys. In carly kA; I

J:nu:ry, the turbine was disa:sembled f r modification. Brisfly, this con-

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,1 2

sisted of a nachining operation on the two high pressure cylinder blade rings which support the first eight stages of fixed reaction blading. This provided 4

the necessary increased voltue to allow proper steam flows into the reaction g

l blading.

w3 u( ;p-g I

During this same period of turbine modification, a permanent repair

!?

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was made to the turbine throttle valves. This consisted of sh. inking a locked-1 in, case-hardened bushing into the valve disc, followed by installation of a I G%

'j podified valve ster. of improved design.

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VII. DESIGN CHANGE StBetARY For purposes of clarity, the design change sumary is arranged into two classifications as folicws:

A.

Changes to the Secondary Plant.

B.

Changes considered to affect the Nuclear Plant.

& Five changes of importance were made in the secondary plant.

I L Installation of a steel liner in the circulating water concrete discharge line. This change was necessary because of the in-g ability to meet the vacuum requirements as specified in the design specifications. Primarily because of the pbrosity of concrete pipe

l.

and the fact that a portion of the pipe is located above the pres-

-?

sure gradient, air was being pulled through the pipe, thus reducing

(

the vacuum.

3 e

L Increasing the vent line from the top of the turbine moisture separator to the heater drain receiver from 1" to 6" and instal-1ation of check valves in the moisture separator discharge lines.

I These changes were needed because of unwanted turbine trips on a Y

large reduction of turbine load. With a greater than 2% reduction y

in turbine load a greater pressure existed temporarily in the No.2

(

feedwater heater and heater drain tank than in the moisture separator, y

causing the water in the heater drain tank to rise into the discharge p

line from thp moisture separator and trip the moisture separator high

.t level trip.

l 1 Nachined an if" x 1/8" groove circumferentially in the bottom of

[

the lower half of Numbers 1, 2 and 3 turbine bearings. This h

change was made to redace the bearing oil film in an effort to reduce (g

turbine vibration.

l 4,. Changed the design of the throttle valve stems to eliminate valve t

disc b.ng up.

_j.

h Modifie J the steam passages between the Rateau and First Reaction

[

Stages on the high pressure turbine to eliminate " steam lift" p

and shaf t vibration at the higher loads.

f..

L The changes listed under this section affect the nuclear plant but are con-f sidered to be of such a nature as not to have required an amendment to the j

Facility License.

I L Changed the range on the nuclear SLR meter from -1 to +10 DPM F

to

.2 to +2 DPM. Since the operator should not exceed 2 DPN

[

SUR, this change was made to make the full scale of the meter i

correspond to the nonnal operating range.

{

L Changes to the Nuclear Recorder:

(a) Converted chart speed to a two-speed motor with chart speeds of two inches per hour and 120 inches per hour, selection

.s of speed accomplished from the front of the instrument.

q (b) Converted indicating scale to a dual range scale marked l-10-100 two cycle log and 0-150 linear.

_b 1

~

(c) Changed the range selection on each pen to allow dacade

[

operation while recording source and intermediate rangs h

channels.

g (d) Added a flux level contact alarm on each pen. These

=

changes were made to give the operator a more understand-k able trace on the recorder and to further alert the operator

,h that he is exceeding inaximum desired power level during the g

many hours of startup' test period.

r L

3 L Recorded the wide range pressurizer level, wide range pressurizer pressure, and loop No. 2 wide range temperature on the process recorder. This is to provide records for test purposes and to be used for analysis during transients.

,4_,.

Removed the handles from the individual rod disconnect switches

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on the back of the control board so that Gese switches can be

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cperated only with a master switch handle. This is to insure that tampering with these switches cannot occur without the operator's knowledge.

J r

h Installed a continuous hydrogen monitor in the waste gas header.

g i

This installation provides added safety and fire prevention pro-g F

tet, tion.

It is felt that continuous monitoring is a more reason-t i

able approach to determining hydrogen concentrations than a

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periodic laboratory analysis.

,Q2 Installed dropped rod protection and loss of power to nuclear

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power range channels protection.

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E F

L Changed the power supply to the nuclear and process recorders E

to the instrument AC vital bus. This change eliminated a noise r

probler in the recorder.

8.

Changed free the type WL 6307 BF detectors to type W1. 7087 BF 3

3 to get added sensitivity.

y, Added a relative humidity indicator to measure RH in the vapor i

container. This data is necessary for calculating air leakge.

k F

,la., Relocated vapor container pressure transmitter to the vapor con-5 tainer side of the trip valve. 'This allows continuous indication

-E of vapor container pressure in the event of trip valve closure.

^*

11. Added a voltage meter and transfer switch to nuclear control board to allow transferring from AC vital bus to emergency power r

supply in case of loss of power to the vital bus.

[

,lh Changed reactor ;wrzissive relay light to be on above 15 kNe.

This gives a positive indication that the scrat circuits are ready to accept low flow and low pressure scram signals.

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13. Modified source rcnge instrumentation to percit a lower dis-I criniutor voltage cotting to reduce naise.

14.

Installed a disabling switch and superviscry light on generator field relay for use during shutJ.w periods of turbine generator.

This is to prevent accidental tripp?ng of main line breakers.

Modified input circuit to the radiation monitoring system to 15.

reduce noise pickup.

& Installed an electronic system for monitoring sounds in the vapor container.

& Added additional shielding cver the neutron shield tank to reduce neutron level during high power operation.

F

& Several changes were made to the control rod drive coil stacks.

The ID of the pcsition indicating coil stack tubes was increased to allow easier removal of the stacks. Resistors were added to the power supply circuit of the stationary gripper coils to re-duce the voltage on the coils and thus reduce the heating of the coils.

The air cooling around the coils was modified to get better heat removal on all coils.

7

& W n te Disposal design changes -- The steam from the evaporator to the gas stripper has been blanked r,ff and building steam is now being supplied to the gas stripper. The operation of the electric evaporator generated large amounts of hydrogen and

. oxygen. The above change resulted in improved stripper of fluent and solved the problem of charging volumes of hydrogen and oxygen r

into the waste gas system.

20.

Converted rod power supply scram breaker lights to monitor the thrae manual scram pshbuttons and the breaker trip coll.

21. Added auxiliary relays to prevent voltage relay for 2400 V tie breaker transfer scheme from becoming inoperative due to a seal-in circuit.

22.

Added ground fault detection equipment on 2-126 line so as to increase sensitivity of line relaying.

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~ ~ ~ ~ ~ ~ ~ ~

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SECTICN VIII. - PLUTONIlN BUILD-UP EXPERIMENTAL pROGDURES The following test procedures will be used in determining the of fects of Pu build-up at intervals during core life.

Test Procedure No. 509F1 Continuino Nuclear Plant Power Operational Tests Power Coefficient Weasurement I.

Objective To detemine the power coefficient of the reactor at various tfees in core life.

II. Conditions 1.

All norn.al plant operation conditions have been satisfied.

2.

Check-off list and load change schedule prepared and dis-cussed with Operations Personnel.

3.

Experimental Data Sheet available.

III.

Precautions 1.

The power levels used shall be limited to those within the normal operating range of the plant.

2.

The reactor control rods shall be maintained on manua)'

operation throughout the procedure.

3.

The pressurizer pressure and level shall not be allowed to drop below the lower operational limits.

IV.

Check-Off List Normal operational check-of f lists will be used by the operator (s).

These check-of f lists will be modified when necessary to allow, establishment of special conditions required for the test.

V.

Instructions A.

Experimental Method The reactivity effect of stepwise power level changes will be measured in terms of moderator temperature change and control rod position changes which are then converted into reactivity change to obtain the power coefficient.

B.

Data Required 1.

Cold leg and hot leg temperatures for each operating main coolant loop and average main coolant temperature.

2.

Main coolant pressure.

3.

Gross electrical output of the turbine generator.

4.

In-Core thermocouple teuperatures.

5.

Control rod position.

6.

Time at which above data are taken and time checks on recorders at periodic intervals.

C.

Special Instrumentation Recuirements None required.

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D.

Experimental Procedures I

1.

Establish some steady state plant load condition as requested in load change schedule.

2.

When main coolant loop temperatures and reactor power are stable, reccrd data in Section V.,B.

3.

Change the load on the plant as requested in load s

change schedule. Do not move control rods during this step but silow main coolant average temperature to change to maintain reactor critical.

4.

When r.:in coolant loop temperatures and reactor power are stable, record data in Section V.,B.

5.

Nove control rods to return average main coolant teep-ereture to value observed in. Step V.,

D.2.

6.

When main coolant loop temperature and reactor power are stable, record data in Section V.,B.

7.

Repeat steps V.D.,3 through V.D.,6 until final load

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conditions are established as requested in the load change schedule.

Test Procedure No. 509El Continuino Nuclear Plant Zero Power Tests Moderator Temperature Coefficient peasurement I.

Obiective To measure the moderator temperature coefficient of reactivity at various times in core life.

II. Conditions 1.

Normal operating check lists for the system required during this procedure have been completed.

2.

Test procedure and schedule discussed with Operations Personnel.

III. Precautions 1.

Criticality should be anticipated at any time when control i

rods are being withdrawn.

i 2.

Any plant changes which would produce a sudden lowering of 0

the reactor coolant temperature (of the o: der of 10 F) must be avoided while the reactor is approaching criticality, i

or is critical.

3.

Startup rates greater than one decade per minute should be avoided.

4.

Available shutdown reactivity must not be reduced below a minimum 2% d k/k by a reducticn of temperature as detbruined by analysis and experiment.

}

5.

Power level scram switch should be set to LOW position and i

power level channels set for single channel scram protection.

AW.. -

53.

4wM IV. Check-Cff List 4g Eor.a1 operating che:k-off lists used or codified as required to Q

establish desired conditions for this test procedure.

.R V.

Instructions W

A.

Experimental Method The moderator temperature coefficient is measured at various temperatures as outlined in the test schedule by changing the average main coolant tenperature and observing the resultant changes in startup rates which are then converted into changes in core reactivity.

2 B. _ Data Required y

1.

In-Core thermocouple;terperatures.

2.

Stable startup rate.

3.

Selected main coolant loop temperatures.

I 4.

Main coolant pressure.

5.

Control rod positions.

M 6.

Time at which data are taken and tice checks on recorder

charts, r

C.

Special Instrumentation Recuirements 1.

Linear amplifier and recorder for compensated ion chamber f

readout.

Note: The main coolant temperature may be changed by ramp

/

increases or ramp decreases or by st3p increases or

}

decreases.

There seems to be no particular advan-i tage to any one nethod, so all are included here.

The method to be used will be selected at the time, based or, overail test scheduling considerations.

5 (Refers also to paragraphs V.-D, E, F, and G.)

g D.

Experimental procedure for R mp Temperature Increase

,?

1.

Establish rod position to obtain e,1 d/m startup rate and i

record data listed in V.-B.

2.

Level off reactor at a power level less than 1 let themal and detemine the just critical rod position. Record data 1isted in V.-B.

~

3.

Reduce flux level to <v2 decades cbove source level.

1 4.

Measure startup rate with rods at position of step D.-l.

/

Record data listed in V.-B.

5.

I.evel eff' reactor ' power and detemine'just" critical rod f

positibn. Record data listed in V.-B.

I

,i 6.

Repeat steps D.-3, D.-4, and D.-5 until the reactor goes sub-critical at this rod position of step D.-l.

7.

Repeat steps D.-1 to D.-6 until the scheduled temperature range has been covered.

E.

Experimental procedure for Ramp *=mperature Decrease 1.

Level off reactor at a power level less tnan 1 MW thermal rp and detemine the just critical rod position. Record data

,;2 listed in V.-B.

2.

Reduce flux level to #2 decades above source level.

3.

Measure startup rate with rods at position a' cep E.-l.

S Record data listed in V.-B.

9, (cont'd.)

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E.

Excerp ental Procedure for Ramp Tm perature Dec_rease (cont'd.)

4.

Level off reactor power and determine just critical rod position. Record data listed in V.-B.

5.

Repeat steps E.-2, E.-3 and E.-4 until the startup rate is al d/m with the rods at the position of stop E.-l.

6.

Repeat step E.-l through E.-5 until the schedule tempera-ture range has been covered.

F.

E,xperimental Procedure for Step Temperature In rease 1.

Establish stable temperature condition in reactor.

2.

Establish rod position to obtain ~ 1 d/m startup rate.

Record data listed in V.-B.

3.

Level of f reactor at a poner level less than 1 is thereal and determine the just critical rod position. Record data listed in V.-B.

4.

Increase temperature several degrees F and establish new stable temperature conditions.

5.

Measure startup rate with rods at position of step F.-2.

Record data listed in V.-B.

6.

Level off reactor power level and obtain just critical rod position. Record data listed in V.-B.

7.

Repeat steps F.-l thro" i F.-6 until the scheduled tempera-ture roage has been covered.

G.

, Experimental Procedure for Step Temperature Decrease 1.

Establish stable temperature c.ondition in reactor.

2.

Level off reactor at a power level less than 1 IM thermal and determine the just critical rod position. Record data listed in V.-B.

3.

Decrease temperature several degrees F and establish new stable tecperature conditions.

i 4.

Measure SUR with rods at position of step G.-2.

Record data listed in V.-B.

5.

Level off reactor power and ob.ain just critical rod position.

6.

Repeat steps G.-3, G.-4, and G.-5.

(Use data obtained in previous step G.-5 as reference rod position for new step G.-4.)

Continue this sequence until the scheduled temperature range has been covered.

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