Text

Startup Rept, for 831115-841031
	ML20101U229
Person / Time
Site:	Diablo Canyon
Issue date:	10/31/1984
From:	Norem M, Shiffer J, Womack L; PACIFIC GAS & ELECTRIC CO.
To:	Knighton G; Office of Nuclear Reactor Regulation
References
	CON-#185-374 DCL-85-031, DCL-85-31, OL, NUDOCS 8502070192
	Download: ML20101U229 (87)
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DIABLO CANYON NUCLEAR POWER PLANT l

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STARTUP REPORT i

LICENSE NUMBER DPR-80

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PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT UNIT 1 STARTUP REPORT TO THE UNITED STATES NUCLEAR REGULATORY C0ft11SSION LICENSE NUMBER DPR-80 4

FOR THE PERIOD NOVEMBER 15, 1983 THROU6H OCTOBER 31, 1984 9

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PREPARED BY:

M. J. ANGUS C. E. BARBEHENN B. A. LOCONTE T. S. OHARA P. 6. SARAFIAN R. A. SAVARD REVIEWED BY:

M. J. ANGUS S. 6. BANTON W. E. COLEY M. N. NOREM C. 6. RA0 L. F. WOMACK APPROVED FOR ISSUE:

h.M.h M. N. NOREM GENERAL CONSTRUCTION LEAD STARTUP SUPERVISOR J

L. F. WOMACK NUCLEAR PLANT OPERATIONS I

ENGINEERING MANAGER

DIABLO CANYON NUCLEAR POWER PLANT UNIT 1 STARTUF REPORT Contents Pete No.

Summary................................. 1 1.0 Fuel Loadi g Program..........................

2 Summary..............................

2 Initial Core Loading (Frerequisites and Periodic Checkoffs)....

3 Initial Fuel Loading........................ 4 2.0 Fre-Critical _ Test Prograa........................ 11 Summary..............................

11 In-Core Moveable Detectors......'............... 12 Rod Mechanism Timing..................'...... 13 Rod Drop Time Measurements...................... 14 Rod Position Indication System................... 23 Rod Control System Operational Test................

24 Pressuriser Spray and Heater Capacity and Continuous Flow Setting.

25 RTD Bypass Loop Flow Measurements.................

28 Incore Thermocouples and RCS RTD Cr,oss Calibration......... 29 RCS Flow Measurement......................... 30 RCS Flow Coastdown........,................. 32 Fressuriser Safety Valve Loop seal Temperature.Frofile....... 35 3.0 ' Initial Criticality and Zero Power Physics Test Prograa......... 37 S umma ry..............................

3 7 Initial Criticality........................

38 puclear Design Checks.............*..........

39 Rod and Boron Worth Nessurements.................. 48 4 '

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l DIABLO CANYON NUCLEAR POWER PLANT UNIT 1 STARTUP REPORT Contents Page No.

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RCCA Pseudo Ejection at Zero Power................. 56 Minimum Shutdown Margin Verification and Stuck Rod Worth Measurement.

...........59 4.0 Turbine Driven AFW Pump Endurance Test.................

61 5.0 Special Low Power Tes t Program.....................

62 Summary..............................

62 Natural circulation........................

63 Natural Circulation With Loss of Pressurizer Heaters........ 66 Natural Circulation at Reduced Pressure..............

66 NaturalCirculationWithSimulatedLossofOff-siteACPower.... 69 Effect of Steam Generator Isolation on Natural Circulation..... 70 Cooldown Capability of the Charging and Letdown System....... 75 Simulated Loss of All on-Site and Off-Site AC Power........

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DIABLO CANYON NUCLEAR F0WER PLANT UNIT 1 STARTUF REPORT List of Figures Page No.

Fig. 1 Diablo Canyon Power Plant Unit 1 Core Map............. 8 Fig. 2 Detector N31 ICER Flot......................

9 Fig. 3 Detector N32 ICRR Flot......................10 Fig. 4 S am ple A+B Trace.........................

16 Fig. 5 Sample A and B Trace.......................

17 Fig. 6 Combined (A+3) and (A AND B) Traces................

18 Fig. 7 Control Rod Drop Time - Cold, No Flow

..............19 Fig. 3 Control Rod Drop Time - Cold, Full Flow.............. 20 Fig. 9 Control Rod Drop Time - Hot, No Flow..........

21 Fig. 10 Control Rod Drop Time - Hot, Full Flow..............

22 Fig. 11 Fressuriser Spray Ef fectiveness.................. 26 Fig. 12 Fressuriser Heater Ef fectiveness.................

27 Fig. 13 Inverse Core Flow Curve...................... 33 Fig. 14 RCS Fl ow Coas td own........................

34 F1's. 15 Pressuriser Safety Valve Loop Seal Piping............. 36 Fig. 16 Core Average Radial Power Distribution - All Rods Out.

45 Fig. 17 Core Average Radial Power Distribution - Control Bank D Inserted.. 46 Fig. 18 Rod Withdrawal Limits To Maintain Negative Moderator Temperature Coefficient.

..............,........... 47,

Fig. 19 Control Bank D Differential and Integral Worth..........

49 Fig. 20 Control Bank C Differential and Integral Worth..........

50 Fig. 21

' Control Bank B Dif ferential and Integral worth..........

51 Fig. 22 Control Bank A Differential and Integral Worth..........

52 Fig. 23 Control Banks With Overlap Dif ferential and Integral Worth....

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DIABLO CANYON NUCLEAR F0WER FLANT UNIT l'STARTUP REPORT List of Fiaures Page No.

(Continued)

Fig. 24 Measured FDHN (Fra and Foot Pseudo RCCA Ejection)......... 58 Fig. 25 Reactor Coolant Loop Temperature on Natural Circulation...... 65 Fig. 26 Pressuriser Pressure vs. Time................... 68 Fig. 27 Steam Generator Isolation During Natural Circulation.......

72 Fig. 28 RCL Temperatures During Steam Generator Isolation......... 75 Fig. 29 Steam Generator Pressures During Steam Generator Isolation....

74 Fig. 30 CVCS Cooldown Curve........................ 76 4

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DIABLO CANTON NUCLEAR POWER PLANT UNIT 1 STARTUP REPORT List of Tables Pase No.

Table 1 Rod @ rop Times for Various Plant Conditions.. '.......... 15 I

Table 2 RTD Bypass Loop Plows........,................ 28 Table 3 Reactor Coolant Loop Flows....................

31 Table 4 Pressuriser Safety Valve Loop Seal Temperatures.......... 35 Table 5 Measured vs. Predicted Boron End loint Concentrations....... 42 i

Table 6 Measured vs. Predicted Isothermal Temperature Coefficient and Derived Moderator Temperature Coef ficient.

43 Table 7 Powe r Di s t ribu t ion Re s ul t s....................

44 Table 8 Measured vs. Predicted Control Bank Worth............. 54 Table 9 Measured vs. Predicted Average Boron Worth............

55 Table 10 Power Distribution Results (Pre and Post Pseudo RCCA Ejection)..

57 Table 11 cvCS Cooldown capability.....................

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SID9tARY The Diablo Canyon Power Plant Unit 1 Startup Program activities included in this report are divided into the following sections:

1.0 Puel Loading Program 2.0 Pre-Critical Test Program 3.0 Initial Criticality and Zero Power Physics Test Prograa 4.0 Turbine Driven APW Pump Endurance Test 5.0 Special Low Power Test Program (Natural Circulation)

Unit 1 fuel loading was performed during the period November 15-24, 1983. After the initial core loading of 193 fuel assemblies was completed, two fuel assemblies

.had to be removed in order to replace their rod cluster control assemblies, and the fuel assemblies then were reloaded. Except for this delay, no major prob-less were encountered.

The Pre-Critical Test Program was performed between December 4,1983 and April 27, 1984. Cold System Tests included Rod Mechanisa Tiaing and Rod Drop Time tests under no flow and full flow conditions. Hot System Tests included Rod Control System tests, Digital Rod Position Indication tests, Rod Mechanism Timing, Rod Drop Time tests under no flow and full flow conditions, Pressuriser Spray and Heater Capacity tests, RTD Bypass Loop Flow tests, Incore Thermocouple /RCS RTD Cross Calibrations, RCS Plow Measurement and RCS Coastdown tests. Results were acceptable a

and no major equipment problems or delays were encountered. The primary reasons for the long duration of pre-critical testing were licensing hearings, regulatory

- reviews and equipment problems.

Initial Criticality and Zero Power Physics testing were conducted from April 28 i

to May 6, 1984. -All tests were completed satisfactorily, and no major probleas were encountered. The all-rods-out sero power moderator temperature coefficient was slightly positive, requiring administrative limits to be placed on control rod withdrawal. These limits remained in place throughout the remainder of the low power testing program.

Following Zero Power Physics testing, a Turbine Driven Auxiliary Peedwater Pump-endurance test was performed with the reactor at low power. Next, a series of special low power tests was conducted with the reactor in natural circulation conditions. The first series of tests was conducted from May 19 to May 23,1984.

Portions of the natural circulation test sequence were repeated on July 24 to satisfy previously incomplete operator' training requirements.

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e 1.0 FUEL LOADING PROGRAM Summary The initial core loading for Diablo Canyon Nuclear Power Plant Uuit 1 consisted of two phases. During the first phase, all of the 193 fuel assemblies were loaded into the reactor vessel within the period November 15-20, 1983. During the second phase, two fuel assemblies A-35 and A-40, were removed to the Unit 1 Fuel Randling Building for replacement of their rod cluster control assemblies (RCCAs) which failed the drag test. These two fuel assemblies were then reinserted into the reactor vessel, and the second phase was comple-ted within a single eight hour shift on November 24, 1983. Except for the re-placement of the two RCCAs, no significant difficulties or delays from equipment problems were encountered. The first phase of the core loading proceeded rela-tively smoothly with only a few minor equipment problems and short suspensions of operations (less than a single eight hour shift).

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1.1 OF,B-8D: INITIAL CORE LOADING (PREREQUISITES AND PERIODIC CHECK 0FFS)

TEST OILTECTIVE Operating Procedure 8-8D providedLa checklist of prerequisites for Unit 1 fuel load along with their scheduling and frequency requirements.

TEST DESCRIPTION OF B-8D included actual prerequisites for fuel load such as procedures and tests, periodic tests to be completed during fuel load, valve lineup checklists, and chemistry sampling requirements and data sheets.

TEST RESUITS Preparations began several weeks ahead of the projected fuel load date and were signed off as each ites was completed. Periodic tests were repeated as necessary and signed off.

By retaining signed copies of all prerequisite tests, a complete record of requirements was assembled and verified. This helped to ensure that all open items were completed in an orderly fashion. No major problems or delays were encountered.

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1.2 INITIAL FUEL LOADINC i

Operations Fuel loading operations commenced at.approximately 1545 hours0.0179 days 0.429 hours 0.00255 weeks 5.878725e-4 months on Tuesday, November 15, 1983. Operations were perfomed in accordance with plant Operating Procedure OF B-8D, Supplement I, and commenced with the grappling of i

the first, fuel assembly in the Unit 1 Fuel Handling Building (FHB).

The core loading map and core loading sequence that were used had been provided to Pacific Gas and Electric Company (PGandE) by Westinghouse, the NSSS vendor (see Figure I for core map).

Prior to being loaded into the core, the fuel assemblies had been wrapped in polyethylene sheaths and dry stored in the FHB spent fuel storage racks and were arranged in the order of loading. Each fuel assembly consisted of a 17 x 17 square array of aircaloy-clad fuel rods with an active fuel length of twelve feet and one of three fuel enrichments (corresponding to assembly i

number prefixes A, B and C).

Fuel assemblies were placed sequentially into the fuel transfer mechanism and transferred under water along the partially flooded refueling canal into the Containment Building. Assemblies were then grappled and transferred dry to the reactor vessel and lowered into the partially filled vessel.

Boron concentration in the vessel was maintained between 2000 and 2100 pga.

Opera-i tions were conducted 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day with three shif ts of personnel, and person-l nel at major work stations were rotated every four hours. All operations were conducted by PGandE personnel with Westinghouse representatives on hand to provide t

technical support. Fuel loading continued with only minor equipment problems and interruptions until initial loading of all 193 fuel assemblies was completed at approximately 2307 hours0.0267 days 0.641 hours 0.00381 weeks 8.778135e-4 months on Decembe~r 20, 1983. This corresponds to an average of 1 1/2 assemblies per hour including all interruptions.

i Prior to core loading, the two permanent Source Range Nuclear Instrumentation channels N31 and N32 read,only about 0.01 and 0.03 counts per second (cys).

t After core loading, the count rate had increased to about 7.78 and 10.01 cps respectively. This corresponded to a signal-to-noise ratio of about 778 and 334, well above the required number of 2 for initial criticality.

Inverse Count Rate Ratio (ICRR) plots for the fuel load for channels N31 and N32 are shown in Figures 2 and 3.

Three other temporary neutron detectors were also used to continously monitor the neutron count rate.

These were lowered into the core and were secured with safety lines.

Since the refueling pool was dry (except for some water in the transfer canal area), the top of the reactor vessel was available for personnel access for observation and positioning of the temporary detectors.

After initial loading of the core, a videotape recording was made of the upper nossle serial numbers on the fuel assemblies to verify proper loading.

-Burnable poison and-thimble plug insert numbers were also verified at this time.

An entire row of fuel assemblies was scanned in one pass by movement of the manipulator crane trolley along the bridge and scanning the camera from side to side to record both the fuel assembly number and the insert number.

Af ter all initial preparations and equipment adjustments were ande, videotaping wea completed in about four hours.

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Following initial core loading, reactor upper internels were installed and t

RCCAs were latched with little or no problems.

However, during RCCA drag through the fuel assembly guide tuben, it was found tha J

assemblies A-35 and A-40 exceeded the drag specifications.

tions with Westinghouse representatives, it was decided to replace the RCCAs After consulta-b Assemblies A-35 and A-40 were unloaded from the core connencing at approx mately 0125 hours0.00145 days 0.0347 hours 2.066799e-4 weeks 4.75625e-5 months on November 24, 1983.

ferred to the FRB, and their RCCAs were replaced with Unit I spares.The finger visibly bowed. tion of the replaced RCCAs showed numerous vertical sc Examina-l new fuel inspections, however the RCCf. had passed drag t ne fuel load.

into the core, and the operation was completed at approx November 24.

Drag tests were completed successfully on the following day.

Problems

.Several minor equipment and related problems caused short delays during Unit fuel loading and are summarized below.

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Fuel Assemblies - In order to compensate for dimensional tolerances i

of some fuel assemblies, manual positioning of the manipulator crane plate alistunent pins.was used successfully to mate the assemblies with the The polyethylene sheath that was placed over the fuel assemblies in the jammed between fuel assembly B-35 and the spent fu An overload indication on the spent fuel pool crane caused operations to be suspended while the sheath was stripped off the fuel assembly by hand.

and operations were resumed.A visual inspection of B-35 revealed no physica 2.

Manipulator Crane - Manipulator crane hoist action at slow speeds was somewhat irregular and intermittent and caused momentary over-load indications.

alleviating the problem. Lubrication of pulleys with water was successful in Manipulator operation caused severe electrical disturbances on various circuits around the refueling cavity.

Noise spikes were induced in the temporary neutron detectors that were placed in the vessel-for fuel loading.

The manipulator crane had to be moved from the vessel area and lef t parked at the extreme end of the refueling cavity near the upender mechanism prior to taking neutron count rate datE.

This slowed the over-all fuel loading operation considerably and added some 5-10 minutes per fuel assembly.

is planned for future fuel loadings.A more isol,sted powe'r source for the n 1

The Selsyn indicator on the manipulator crane was slightly out of

. adjustment in certain core locations at the start of fuel loading

. (approsimately 1/8 inch) and appeared to drift further out of adjust-ment es the fuel loading progressed. The indicator was used for 5

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Fuel Transfer Mechanism - The carriage drive air actor became frozen causing operations to be suspended.

and lubricated and returned to service.The air actor was removed, cleaned i'

A Design Change Notice was on the motor and prevent water from leaking into th 4.

Temporary Neutron Detectors - Three temporary detectors were us fuel loading to help monitor subcritical multiplication.

probleme interfered with operation of these detectors.

Several minor viously, operation of the manipulator crane caused electrical spikes and false indications on the detectors, requiring delays in fuel loading to; permit count rates to be determined.

I Minor physical contact with temporary detector cables that were laid around the vessel flange canned false indications on these detectors Indications included spikes and abrupt continuous changes in count rate.

Detector readings had to be renormalized.

detector close to a hot leg nozzle resulted in flow induced vibration Placement of one of the detector and cable.

ed from this motion.

A falso high count rate indication result-cable were repositioned.The signal stabilized when the detector and 5.

Source Range Nuclear Instruments - The two permanently installed source range detectors activated seseral spurious High Flux at Shutdown alaras with resulting suspension of operations and assen-blage of personnel in Containment at the personnel hatch.

of several of the spurious alarna were traced to electrical inter-The causes ferences caused by activation of the Unit 1 Code Call System, oper-ation of the Containment Polar Crane and operation of the Movable Incore Detector System.

are being evaluated.

Design changes to alleviate these problems 6.

Underwater Lights - The nominal 1000 watt model had to have several bulbs replaced with lower wattage floodlight bulbs to alleviate a problem with short bulb life.

t The other style of light, a 750 watt model that could not be employed i

during fuel load, had a plexiglass lens cover that was loosely retained by four cap screws.

The cap screws are being replaced with longer screws to prevent loss of the cover into the RCS under high flow L

conditions.. This has occurred at other plar.ts.

i Another probles encountered with moving underwater lights was motion of the cables disturbing the temporary neutron detector cables and caus-ing false count rate indications as discussed previously.

placement-became an important factor in ensuring efficient operations.

Cable 7.

Diesel Generators - Because Unit 1 ' fuel loading was conducted with

.one operable diesel, DC 1-2, several minor problems with this diesel caused fue1 ~1oading to be suspended temporarily.

Problems included F

Low Turbo-Start Air Pressure and Low Fuel Oil Frising Tank Level 6

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. Licensee Event Report-LER 83-27).

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Instrument Power Inverters - Inverter 1-2 was removed from service for testing. This caused a momentary interruption of power to a Source Range Nuclear Instrument during the load transfer. Fuel loading was i

suslended temporarily during this period to allow a functional test

. of the instrument to be performed after the load transfer.

9.

Manipulator Crane Auxiliary Hoist - While using the manipulator crane

, auxiliary hoist for latching and handling control rods, it was noticed that the hoist had not been load tested within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to use for performing RCCA drag tests (LER-83-31).

10.

Containment Equipment Hatch - Subsequent to fuel loading and while in Mode 5, an NRC Inspector noted a visible gap at the top of the con-tainment equipment hatch. The hatch had been held in place by four (4) bolts during fuel loading, as required by Technical Specifications; however, the bolt placement and torque had been inadequate. The equip-ment hatch was secured with additional bolting (LER-83-28).

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2.0 PRE-CRITIcAt TIMT PR0t: RAM Summary This phase of the Startup Program was divided into three sections.

Cold.irsten Tests RCS Heatup-Hot System Tests Cold System Tests were performed after initial fuel load, reactor assembly and RCS filling and venting. The tests that were done during this phase were Incore Moveable Detector Checkout, Phase 1 of Rod Drive Mechanism Timing, and i-Phase 1 of Rod Drop Time Measurements. These were performed during the period from December 4,1983 to December 10, 1983.

~

Mode transition surveillance testing began af ter approval for heatup was granted. RCS Heatup phase started on February 20, 1984 and was completed on March 8,1984. RCS Heatup included four taperature plateaus: 250 deg. F, 340 deg. F, 450 deg. F, and 547 deg. F.

i At each plateau :

1)

Piping walkdowns were performed and interference problems were fiand or evaluated before leaving the plateau.

2)

Data was taken for Reactor Vessel Level Indication System (RTLIS).

4 3)

Inoore thermooouples and RCS RTD oross calibration was performed (except at 450 deg. F plateau).

Hot System Tests were performed with the RCS at rated temperature and pressure.

The tests included Rod Drop Time measurments, Pressuriser Spray and Heater Effectiveness, RCS Flow Measurments, RCS Plow Coastdown, and RfD Bypass Loop Flow Measurements. These tests started on March 25, 1984 and were empleted on April 27, 1984.

Major delays during the prescribed test program were as follows:-

Doomber 10, 1983 to January 25, 1984:

Waiting for approval to heatup.

March 9,1984 to March 24, 1984:

Replace RCP 1-3 motor.

L

- April 7,1984 to April 25, 1984:

RCP 1-4 seal repairs, Pressuciter Spray Valve repair and waiting for p

issuance of the Low Fower Test License.

11

^

~

7 2.1-Test Procedure No. 38.5 - In-core Moveable Detectors.

TEST OBJECTIVE The purpose of this test was to functionally check the operation of the In-Core

.i Moveable Detector System.

TEST DESCRIPTIOff This procedure was a comprehensive functional test of the In-Core Moveable Detector System. Using a dummy cable, operation of all five and ten path transfer devices was checked. The dummy cable was also used to verify path length measurements. In addi-tion, 'all alarms and indicator' lights were checked for proper actuation. Finally, leak detection and gas purge systems related to the moveable detectors were tested.

TEST RESULTS All acceptance criteria were met and the system was ready for standard flux mapping at the conclusion of the test.

The main problem involved sticking of the dummy cable due to excessive friction at various points along the detector paths. The solution consisted of cleaning all thimble tubes, and modifying and aligning the isolation valve rack.

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2.2 Test Procedure.No. 36.1 - Rod Mechanism Timing TEST OBJECTIVE The purpose of this test was to operationally check the cycler timing for each control rod drive mechanism (CRDM) with a rod control cluster assembly (RCCA) attached under both cold and hot plant conditions.

TEST DESCRIPTION Timing was accomplished by monitoring the lift coil, movable gripper coil and l

stationary gripper coil currents with an oscillograph. Microphones were placed on the top cap of each rod travel housing and their sound signals were monitored with their respective mechanism current traces. These traces were used to verify proper latch operation in conjunction with the lift, mov'able gripper and station-ary gripper coil current traces.

Cold System Testing was performed from December 5, 1983 to December 9, 1983 at approximately 350 psig and 110 deg. F.

Because the Digital Rod Position Indication (DRPI) system had not been declared operable, Technical Specification Special Test Exception 3.10.5 was declared in effect. Each rod was individually operated to verify mechanism timing, brought to 228 steps out of the core to verify DRPI and then dropped to perform rod drop test (ref. T.P. 36.3). This sequence was repeated until all the rods were tested and Special Test Exception 3.10.5 was no longer neces-sary for mechanism timing tests.

Hot System Testing began on March 26, 1984 with the RCS at approximately 547 deg. F and 2235 psig. Because the DRPI was now declared operable it was possible to test the mechanisms using standard testing techniques (by pulling one bank up and testing one mechanism at a time).. This part of the test program went smoothly with no problems.

TEST RESULTS The traces for each mechanism were evaluated immediately following their performance and were determined to be satisfactory.

Listed below are some of the problems encountered and their associated resolutions during. the performance of Cold System Testing:

1)

DRPI indication problems / replaced encoder cards 2)

Blown stationary fuses / replaced fuses 3)

Step counter malfunction / cleaned step counters 3

4)

Rod E-10 would not move / lift pole connector pin at the mechanisa had retracted into its housing and was repeired 5)

Rod D-12 DRPI probise/ loose electrical connector pins, from the DRPI

-coils at the head area were repaired 13 m~

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2.3 Test Procedure No. 36.3 - Rod Drop Time Measurements TEST OBJECTIVE The purpose of this test was to perform the following:

1)

Measure the drop time of all control rods under four different con-ditions;_ cold no flow, cold full flow, hot no flow and hot full flow.

Under each of the conditions, obtain a separate rod drop trace for a combined data coil signal ("A+B" trace) and an individual data coil signal ("A&B" trace).

2)

Repeat the rod drop test on the rods with the slowest and fastest drop times ("A+B". tracee only) under all of the above mentioned conditions a minimum of six times.

3)

Demonstrate that the system meets the requirements of Technical Spec-ification 3.1.3.4 which states that the individual full length (shut-down and control) rod drop time from the fully withdrawn position shall be <2.2 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry with Tavs.)_541 deg. F and all reactor coolant pumps running.

TEST DESCRIPTION All measurements were made using a high speed visicorder to record the change in mechanism stationary gripper voltage, the output of the Digital Rod Position Indication (DRPI). data coils and the output of the microphones on the top cap I

of the mechanism housings. From the traces thus obtained, it was possible to measure the rod drop time from the loss of stationary gripper coil voltage R

to entry into the dashpot region as well as the time to reach the bottom of the

-dashpot. Figures 4 and 5 are examples of the traces obtained.

Since the DRFI could not be declared operable until it was functionally demonstrated l

that the DRFI system correctly' tracked rod position, it was necessary to perform the cold no flow ("A+B" traces). rod drop test in conjunction with the rod mechanism timing

-test (see T.F. 36.1 - Rod Mechanism Timing Test). This portion of the test was con-ducted from December 5,1983 to December 9,1983 with several equipment problems delay--

ing the test program (see T.P. 36.1).. The cold no flow portion of the test was then repeated for the "A&B" traces.' The remainder of the test was performed using a method that. allowed the "A+B" and "A&B" traces to be taken on one trace. (Figure 6)

~

. Listed below are the remaining rod drop test plant conditions and their performance

-dates:-

~~

350 pois/110 deg. F December 10, 1983

- Cold No Flow Hot full Flow 2235 peig/J 547 deg. F March 27, 1964 e

~

Hot No Flow 2235 pois/547 deg. F March 28, 1984 l

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i TEST RESULTS Figures 7 through 10 show the rod drop times for the four plant conditions and Table 1 lists the core average, and slowest and fastest drop times. The traces for each rod drop were evaluated soon af ter perforgance and were satisfactcry. All rod drop times were well below the Technical Specification Requirements of 2.2 seconds from initiation of event to dashpot entry.

Table 1 Rod Drop Times (Sec.) for Various Plant Conditions Core Slowest Fastest Standard Plant Conditions Average Rod Rod Deviation Cold Shutdown - No Flow 1.178/1.680 1.207/1.722 1.150/1.650

+0.013/0.017 Cold Shutdown - Full Flow 1.435/2.059 1.527/2.183 1.363/1.967

+0.032/0.041 Hot Standby - No Flow 1.114/1.542 1.146/1.563 1.080/1.500

_+0.014/0.019 Hot Standby - Full Flow 1.304/1.788 1.352/1.843 1.266/1.745

+0 016/0 024 Times indicated are: Initiation of event to dashpot entry / initiation of event to bottom of dashpot.

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l-2.4 Test Procedure No. 36.5 - Digital Rod Position Indication System TEST OBJECTIVE l'

The purpose of this test was to verify that the Digital Rod Position Indication (DRPI). System satisfactorily performed the required indication and alarm functions for each individual RCCA under Hot Standby conditions.

TEST DESCRIPTION On March 27, 1984, with the plant in Hot Standby conditions, the control rod system i

was operated and Rod Bottom LEDs (Light Emitting Diodes) indication and proper agreement of rod position between the step counters DRPI, P-250 computer, and pulse-to-analog (control banks only) systems were verified.

TEST RESULTS All DRPI, pulse-to-analog and step counter readings agreed exceptionally well.

However, the P-250 printout did not agree between rod positions of 96 steps and 168 steps out. This was determined to be a software problem in the P-250 pro-gram. The problem was corrected and further testing verified that the P-250 computer tracked the rod position, properly. Rod bottom LED indication operated satisfactorily.

9 i

I e

e 23

2.5 Test Procedure No. 36.6 - Rod Control Systea Operational Test TEST OBJECTIVE The purpose of this test was to verify the proper operation of the Rod Control System.

TEST DESCRIPTION With the plant at Hot Standby conditions, the control rod system was operated to verify the proper functioning of the following:

1)

Rod movement status. lights.

2)

Rod position indication systems.

3)

Rod speed indicator.

4)

DC hold supply cabinet.

5)

Bank overlap.

6)

" Rod Botton" and " Rods at Botton" alaras.

During the bank overlap test, rod control was in manual and the overlap settings were lowered from their normal values to preclude excessive rod withdrawal.

TEST RESULTS All rod control system functions performed as expected.

L i

i-L 24 i

2.6 - Test Procedure No. 7.10 - Pressuriser Spray and Heater Capacity and Con 61uuuus Flow Setting TEST OBJECTIVE This test had three objectives:

1)

To establish the continuous pressurizer spray flow rate by i

adjusting the spray flow bypass valves.

2)

To determine pressurizer spray effectiveness.

3)

To determine pressuriser heater effectiveness.

^

i i

TEST DESCRIPTION For the continuous spray setting, the plant was initially stabilized in Hot Stand-by conditions with the spray flow bypass valves (valves 8050 and 8051) 3/4 turn open. Each spray valve was then adjusted to obtain the minimum possible continuous spray flowrate while maintaining a pressurizer to spray line temperature difference

. less than 200 deg. F and a spray line temperature above the low temperature set-point (500 deg. F +5 deg. F). The resulting valve positions represented the final settings.

To initiate the pressurizer spray effectiveness portion of*this test, the plant was stabilized in Hot Standby conditions and all pressurizer heaters were de-energized.

Next, both normal spray valves were fully opened to cause a rapid depressurization.

The pressure transient response (i.e., pressure vs. time as measured on a strip chart recorder) was then compared to the acceptance criteria.

The final section of this test was intended to verify pressurizer heater effec-

.tiveness. With the plant in stable Hot Standby conditions and both normal spray valves closed, all pressuriser heaters were energized.to their maximum capacity.

The pressure transient response, as measured by a strip chart recorder, was then compared to the acceptance criteria.

TEST RESULTS The-pressuriser continuous spray flow bypass valves were set as follows:

Loop 1: Valve 8050: 1/2 turn open, Loop 2: Valve 8051:

1 turn open.

During the adjustment of valve 8051, its valve stem broke; valve 8051 was subse-quently replaced.

i Pressurizar spray effectiveness was determined to be approximately -150 psi / minute.

Thi's rate was well within limits, sa shown by Figure 11.

L Pressuriser heater effectiveness was determined to be approximately 20 psi / minute.

l Again,'the transient response was well within the limits, as shown in Figure 12.

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2.7 Tes't Procedure No. 7.3 - Resistance Temperature Detector Bypass Loop Flow Measurements TEST OBJECTIVE The objective of this test was to verify transport times and alares in the Resistance Temperature Detector (RTD) bypass loops for Hot Standby conditions af ter core loading.

TEST DESCRIPTION RTD bypass loop total flow, hot leg flow, and cold leg flow were measured for each reactor coolant loop. These measured values were compared to calculated minimum flow rates necessary to achieve the design reactor coolant transport time (i.e. 1.0 second) for each RTD. In addition, the RTD low flow alaras were

. set and verified.

TEST RESULTS

.During the conduct of the flow measurements, restricting orifices (RO) were installed in each of the four cold leg bypass loops to balance the flows and to' reduce the total bypass flows to within the flow indicator range.

RTD bypass loop low flow alara setpoints were set and checked to trip within l-

+3 spa.of 90% of the total sensured loop RTD bypass flow rate. RTD hot leg and cold leg bypass loop flows were significantly greater than the minimum required flows, thus ensuring acceptable reactor coolant transport times for each RTD. Final results are shown in Table 2.

i Table 2

+

RTD Bypass' Loop Flows i:

Reactor Cold Leg Coolant-

. Bypass Loop Cold Leg Flow (spm)

Hot Les Flow (spa)

Total Flow Loop RO/ Size Minimus Actual Minimus Actual (spa) 1-1

'10-406/0.77" 56.0 120.3 63.6 138.8 259.0 1-2 R0-407/0.739" 70.3 132.7 61.2 136.3 269.0

'1-3 10-408/0.745" 70.1 121.8 64.7 133.2 255.0 1-4.

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2.8 STP R-27/TB-8403: Incore Thermocouples and RCS RTD Cross Calibration TEST OBJECTIVE The combination of Surveillance Test Procedure R-27 and Temporary Procedure TB-8403 provided a means to calibrate the incore thermocouples using the RCS loop RTDs as a reference over the temperature range.250 - 547 deg. F.

The procedures also allowed cross calibration of the RTDs themselves.

4 1

TEST DESCRIPTION

-Temporary procedure TB-8403 was performed at 250 deg. F and 340 deg. F.

This test consisted of establishing a stable, full-flow, isothermal temperature in the RCS using the Residual Heat Removal System and recording spare narrow range RTD resistance and incore thermocouple temperature and millivolt readings. Ther-nocouple temperature readings were obtained from the output of the P-250 process computer, front panel of the two Post Accident Monitors (PAMs), the Emergency Response Facility Data System, and the Subcooled Margin Monitor. Millivoit readings were recorded at the input terminal boards to the PAM panels.

Surveillance Test Procedure R-27 was performed at 547 deg. F.

This test consisted of establishing stable, full-flow ' isothermal RCS temperature using condenser steam dumps and simultaneously recording all RTD resistance readings in one RCS loop along with all thermocouple temperatures and millivolt readings and steam generator pressures. In order to read operating RTD resistances, those RTDs had to be taken out of service.

Because of Technical Specification requirements, only the RTDs in a single loop were removed from service'at any time and measured. Therefore, the test was repeated for each remaining RCS loop after restoring the previous loop to ser-vice and re-establishing isothermal temperature in the RCS by operating the steam dump system in pressure control mode.

TEST RESULTS Thermocouple readings.at the various panel readouts initially were uncalibrated and therefore contained substantial errors. After calibration, errors generally were reduced to acceptable levels (12 deg. at 547 deg. F).

.RTD readings were consistent and only a few required sas11 temperature corrections, all less than il deg. at 547 deg. F.

All but about 10% of the RTDs met the +O.3 deg. F accuracy specification for narrow range RTDs at 547 deg. F with no correction.

' The P-250 computer thermocouple readings were inconsistent as a result of exces-sive temperature sensitivity of the isolation amplifiers. A design change had L

been initiated previously to solve this problem but had not been completed at the time of the test.

L e

9 29 i

- 2.9 Test Procedure No. 7.5 - RCS Flow Measurement TEST OBJECTIVES The primary objective of this test was to calculate steady state Reactor Coolant System (RCS) flow at pre-critical conditions. Addition 41 data, to serve as base-line information for an undamaged core, were also collected.

~

1 TEST DESCRIPTION Loop flow instrumentation consisted of three elbow tap differential pressure transmitters on each of the four reactor coolant loops. In order to dampen flow oscillations, snubbers were temporarily installed on these loop flow i

transmitters.

Initial ~ conditions for the RCS flow measurement required steady state Hot Standby conditions with all four reactor coolant pumps operating. With the RCS

. stable, flow transmitter output was recorded for a ten minute period. The voltage readings from each elbow tap flow transmitter were averaged and converted to a differential pressure based on calibration data. The pressures were then converted to reactor coolant loop flow through the use of a. Westinghouse supplied curve.

i RCS baseline data were collected for various operating pump configurations to

- serve as a reference to which future data could be compared in the event that assessments of possible core damage become necessary.

TEST RESULTS The total RCS flow rate was 390,000 gpa. The individual loop flow rates were all within +3% of the average and all acceptance criteria were met. Table 3 provides the details of the results.

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' Loop Flow - 97,500-x100 97,500 ACCEPTANCE CRITERIA 1t 1.

Flow rate for each loop within 5% of average.

2.

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3.

At Hot Standby, total RCS flowrate 1 90%'of 363,000 GPM.

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2.10 Test _Frocedure No. 7.6 - Reactor Coolant Systen Flow Coastdown TEST OBJECTIVE The main objective of this test was to measure changes in the reactor coolant flow rate resulting from trips of various reactor coolant pump (RCP) breakers.

Delay times associated with these trips were also determined.

TEST DESCRIPTION Four coastdowns were analysed:

1)

Four pumps operating initially, two pumps coasting down (2/4),

2)

Four pumps operating initially, four pumps coasting down (4/4),

3)

Three pumps operating initially, two pumps coasting down (2/3),

4)

Three pumps operating initially, three pumps coasting down (3/3).

In each case, the pumps coasting down were tripped within 100 asec of one-another under Hot Standby conditions. The resulting coastdowns, i.e., flow as a function of time, were compared to coastdowns in the FSAR.

TEST RESULTS For the 4/4 coastdown, the rate at which actual flow changed was evaluated through the slope of the inverse core flow curve, as shown in Figure 13. This curve was compared to the FSAR inverse core flow curve in the time range of 3 to 10 seconds.

Although the slope of the actual curve was greater than the slope of the FSAR curve,'the results were evaluated by Westinghouse and determined to be acceptable.

The actual inverse flow curve was also used to determine flow sensor delay. - (Flow sensor delay is defined as the time at which the best straight line approximation to the. inverse flow curve drawn in the 4/4 coastdown, between.three and ten seconds, intersects the inverse flow value of 1.0).

For all four coastdowns, the actual flow, corrected for flow sensor delay, was

. compared to the flow in the FSAR. Results are shown in Figure 14. To be con-servative, the FSAR curve must lie below (i.e., show a more rapid reduction in 4

coolant flow) the actual curve. However, due to conservative testing procedures, actual flow curves typically lie slightly below the FSAR curves for Westinghouse plants. The results were evaluated and declared acceptable by Westinghouse.

Data-from the 2/4 coastdown was used to calculate the low flow time delay, the undervoltage trip delay time, and the under' frequency trip delay tiee. All three

-parameters met their respective Acceptance Criteria (A.C.).

The low flow time: delay, defined as the time from beginning of coastdown until rod action, was calculated to be 1.44 seconds (A.C. of 13.06 seconds). The undervoltage

. trip ~ delay time,~ defined as the difference between the time undervoltage trip

~

conditions are reached and the time the rods are free to fall, was calculated to

'be 0.08'second-(A.C.~of fl.2 seconds). The under frequency trip delay time.. defined as the difference between the time underfrequency trip conditions are reached and

-the time the rods are free to fall, was calculated to be 0.17 second (A.C. of 10.6 second).

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.ww.

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m --

2.11 Test Procedure No. 7.9 - Pressurizer Safety Valve Loop Seal Temperature Profile TEST OBJECTIVE The objective of this test was to verify acceptable performance of the insulation on the pressurizer safety valve loop seals and show compliance with NU"?.G-0737.

TEST DESCRIPTION This test involved temperature measurements associated with the modified insu-lation on the pressurizer safety valve loop seals. As shown in Figure 15, ther-aocouples wsre located on the inside of the insulation of each loop seal and on'the body of each safety valve. Data were collected over a thirty minute period with the primary system at Hot Standby conditions.

Safety valve temperatures were required to be less than 350 deg. F while the loop seal pipe temperatures were required to be greater than 260 deg. F.

TEST RESULTS Initial temperature measurements yielded satisfactory results for the valve temperatures, but loop seal piping temperatures were too low. Additional insulation was installed and existing insulation was modified to eliminate the excessive heat loss. After these corrections, all acceptance criteria were met as shown in Table 4.

Table 4 Pressurizer Safety Valve Loop Seal Temperatures Pressurizer Safety Valve Acceptance Criteria 8010A 80108 8010 (deg. F)

(deg. F)

Loop Seal Piping Temp.

>260 328 322 316.

Safety valve Temp.

<350 148 184 171 L

35

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LOOP SEAL DIA8LO CANYON POWER PLANT - UNIT 1

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PRESSURIZER SAFETY VALVE LOOP SEAL PIPING Figure 15 s,.

36

-.c-

.m t

g.

3.0 INITIAL CRITICALITY AND ZERO POWER PHYSICS TEST PROGRAM Summary This portion of the Startup Program consisted of Initial Criticality and Zero Power 1

- physics Testing. The approach to criticality started on April 28, 1984 and the

low power physics measurements were completed on May 6, 1984. No major problems were encountered during the conduct of these tests.

Initial criticality was achieved on April 29, 1984 at 0007, hours. Next, nuclear design checks were performed by measuring parameters including:

Critical boron concentrations Isothermal temperature coefficients Control rod bank reactivity worths Zero power neutron flux distributions

. Boron reactivity worths These parameters were determined at' nominal all-rods-out conditions as well as for various control bank configurations.

Additional physics testing included a pseudo rod ejection and a minimum shutdown margin verification. For the pseudo ejection, an individual control rod was with-drawn in order to obtain the, flux distribution ~and ejected reactivity worth.

Finally, adequate shutdown margin was verified by measuring the reactivity worth of the shutdown banks and the worth of the nost reactive stuck rod.

The physics testing was completed in a timely manner and verified that the zero-power physics characteristics of the reactor core are consistent with design.

4 4

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3.1 Test Procedure No. 41.2 - Initial Criticality TEST ORJECTIVE.

- The objectives of this procedure were to 1) achieve criticality, 2) increase reactor

. power. to the point of adding heat, 3) establish the sero power test range, and 4) p verify proper operation of the reactivity computer.

1 TEST DESCRIPTION Initial conditions were established with the shutdown banks fully withdrawn, 1

control banks fully inserted, boron concentration at 1895 ppa, RCS temperature at 547 deg. F, and RCS pressure at 2235 psig.

The control banks were withdrawn in 50 step intervals until Control Bank D reached 160 steps. An inverse count rate ratio (ICRR) was taken at each interval.

During the control rod withdrawal, the ICRR dropped from 1.0 to approximately 0.85.

Dilution to criticality was then commenced at approximately 1000 pen /hr. Again, an ICRR was tracked and plotted. When the ICRR reached 0.2, the dilution was

. stopped to allow RCS a1xing. At 0007 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months on April 29, 1984 criticality was

. achieved.

Rods were pulled to obtain a positive startup rate and power increased to 1x10-8 amp on the intermediate range. Power was then stabilised and reference initial criticality data taken.

[

Following recording of the initial criticality data, power was increased toward the point of adding heat (POAH). The POAH was~found and the approach repeated three times to ensure data reliability. From the POAH data (1x10-6 amp as i

I

-indicated on the reactivity computer) the zero power test range (ZPTR) was established. This was set from 1x10d to 1x10-7 amp on the reactivity computer.

-Reactor power was then reduced to the lower end of the ZPTR in preparation for the reactivity computer checkout. A 35 pcm positive reactivity addition was made 'and the reactor doubling time recorded. The results were checked against Westinghouse design criteria and found to be satisfactory.

TEST RESULTS All parameters measured during this testing fell within the Acceptance Criteria provided by Westinghouse. Critical boron concentration was measured at 1335 ppa.

-The estimated critical condition was 1300 ppa. The difference was well within the design data allowance.

The*P0AR was measured at 1.03x10-6 amp on the intermediate range.

Recording the same data for'esch of the three approaches to the POAH verified the value was indeed correct and repeatable.

-The last test to verify proper cperation of the reactivity computer indicated proper response for reactivity changes. All test cases fell within the +4% Acceptance

. Criteria provided by Westinghouse. This test was repeated on a nominal daily basis to ensure continued proper operation of the reactivity computer throughout testing.

38

a.

a 3.2 Test Procedure No. 41.3 - Nuclear Design Checks TEST OBJECTIVE The objective of this test was to measure the Boron Endpoint, Isothermal Temperature Coefficient and the Zero Power Neutron Flux Distribution and compare with predicted values.

~

TEST' DESCRIPTION At various control rod configurations, measurements were made to determine the Boron Endpoint, the Isothermal Temperature Coefficient, and the Zero Power Neutron Flux Distribution.

1 Boron End Point Measurement This measurement was performed to determine the boron concentration at which the reactor-would be just critical at the control rod endpoint configuration.

The endpoint configurations at which this was performed were

1)

All rods out. (ARO)

',2)

Control bank D fully inserted.

3)-

Control banks D and C fully inserted.

4)

Control banks D. C, and B fully inserted.

5).Controi banks D, C E, and A fully inserted.

6)

Shutdown bank D, and all control banks fully inserted.

7) _ Shutdown banks D and C, and all control banks fully inserted.

3)

All control banks fully inserted less the most reactive rod controi

~

cluster assembly.

~

f This measurement was performed with the reactor just critical and within 60 pea of the endpoint configuration. The critical RCS boron concentration was de-termined. The controlling bank was then withd'rawn/ inserted to the endpoint con-

-figuration and the reactivity change measured. The corresponding critical boron endpoint concentration was determined using the following equation:

[

(Cg)end = (Cg)j,c, - [Ap/(Boron Worth)]

L Uhere:

i

-(Cg)e'ad = Critical boron endpoint concentration.

l (Cg)j,c, = Measured just critical boron. concentration at beginning of measurement

(-

. Ap = The reactivity change by bank insertion / withdrawal to endpoint configuration.

i.

Boron Worth.= The reactivity change per unit boron concentration change as spec-jj ified by the nuclear design report.

l i

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-Isothermal Temperature Coefficient Measurement This measurement determined the reactivity change due to the overall temperature l change of the core'. This measurement was performed at the following endpoint configurations:

1)

All rods out.

2)

Control bank D fully inserted.

3)

' Control bank D and C fully inserted.

4)

Control bank D, C, and B fully inserted. *

~5)

Control bank D. C, B, and A fully inserted.

with the output from the reactivity computer and an average RCL Tavg signal con-nected to an x-y recorder, the RCS was gradually cooled approximately 5 deg. F using the steam dump system aan then reheated to the no-load Tavg. The slope gen-

-ersted on the x-y recorder was then taken to be the isothermal temperature coefficient.

Another parameter of interest, the moderator temperature coefficient, was then deter-eined from the relationship:

a iso = anod + adoppler I

.. who re:

a iso - Isothermal Temperature Coefficient a nod = Moderator Temperature Coefficient a doppler = Doppler (Fuel) Temperature Coefficient (from Nuclear Design Report)

Zero Power Flux Distribution In order to verify the correct fuel loading pattern and to verify design calcu-lations, low power testing included two flux distribution measurements: the first

-with all rods out and the second with Control Bank D almost fully inserted. The core average temperature was maintained at approximately 547 deg. F and neutron flux was maintained just above the nominal sero power physics test range and just below~the point of adding nuclear heat. The core average radial power distribution is shown in Figures 16 and 17 for the two cases.

The Movable Detector Flux Mapping System was used to collect data from the 56 fuel assemblies with instrument paths., Due to small detector currents during i

aero power testing, the movable detector system required a special set up for each detector consisting of a high quality power supply and a Keithly Picosameter for signal input to the flux trace recorders and the P-250 computer.

The. collected data (i.e., the P-250 output) were then input to the INCORE computer code, which expands the measured information to a detailed three-dimensional 3

i full-core power distribution.

l; 40 L

3. 7.. -.. -.

=

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TEST RESULTS i

' Boron Endpoint Measurement The _ results of the Boron Endpoint Measurements are shown in Table 5.

The measured

-values agreed very.well with predicted values.

/

IsothermalTenkrtureCoefficientMeasurement The results of the Isothermal Temperature Coefficient measurements are suramarized in Table 6.

It was determined that the moderator temperature coefficient was positive at the ARO endpoint configuration. Rod withdrawal limits were established using an i

extrapolation technique on the isothermal temperature coefficient data of the ARO and Control Bank D fully inserted endpoints. The rod withdrawal limits are a function of boron concentration and power level as shown in Figure 18. They shall remain in effect until sufficient core burnup has occurred such that the critical boron i

_ concentration is reduced to the point where the moderator temperature coefficient n

is negative. -(The Technical Specifications require only that the moderator temp-erature coefficient be more negative than 0.0 pen /deg.F).

- Zero Power Flux Distribution Both flux distribution measurements yielded results close to expectations and well within the acceptance criteria.. The core average axial power distribution was close to a cosine shape while the unrodded radial distribution was reason-

ably flat with the peak assemblies closer to the core periphery than the center.

_8ee Figure 16 for relative assembly powers. Insertion of the Control Bank D caused a slight increase in flux paaking, as shown in Figure 17. The radial dis-tribution was also characterized by a small, but acceptable, flux tilt. Peak-ing factors are summarized in Table 7.

'One of the six detectors failed'(detector C) during the ARO Flux Mapping Inves-

~

-tigations determined that a high resistance short had developed in the detector cabling between the drive unit and the detector. Detector C remained out of service for the remainder of Zero-Power physics testing and Low-Power testing.

t This doubled the amount of time required to obtain a full core flux map.

4

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41 i.

i T__.,.-.._....__..-~.--~-:-----

Table 5 Measured Versus Predicted Boron Endpoint Concentrations Criticai Boron Concentration

, Endpoint tonfiguration Actual Predicted (ppa)

(ppa)

ARO 1344 1310 1 50 CD in 1254 1255 1 15.3 CD CC in 1160 1162 1 15.8 CD,CC,CB in 1083 1081 1 13.4 CD CC.CB,CA in 967 971 1 16.8 CD.CC,CB,CA SDD in 874 876 1 13.4 CD,CC CB,CA.SDD.SDC, in 789 791 1 12.9 ARI, N-1 619 611 1 50 e

F t

+

h 5

1 4

4 i

42

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Table 6 Measured Versus Fredicted Isothermal Temperature Coefficient and Derived Moderator Temperature Coefficient r

Derived

Isothermal Temperature Coefficient Moderator Endpoint

( a lso)

Temperature Configuration Measured Fredicted Coefficient (pca/deg. F)

(pca/deg. F)

(a mod)

(pca/deg. F)

ARO

.77

-1.7 13.0

+1.09 CD in

-2.25

-2.5 13.0

-0.39 CD,CC in

-4.75

-4.3 13.0

-2.89 CD,CC,CB in

-6.58

-6.4 13.0

-4.72 CD,CC,CB.CA in

-6.78

-7.9 13.0

-6.92 u

i

-1.86 pca/deg.F

From the Design Report, a RZP

=

doppler a nod a iso - a doppler

=

= (a iso + 1.86) pen /deg. F 4

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Table 7 POWER DISTRIBUTION RESULTS

-(

UNRODDED RODDED

-ITEM FLUX MAP FLUX MAP (All Rods Out)

(Control Bank D In)

CONDITIONS-temperature 547 deg. F 547 deg. F boron conc.

1342 PPM 1255 PPM

' power 0%

0%

^

burnup 0 MWD /MTU 0 MWD /NTU DATE April 30, 1984 May 1, 1984 ROD CONFIGURATION Bank D G 218 steps Bank D 9 13 steps Bank C 9 228 steps Bank C 9 228 steps F

- measured value 1.438 1.619 AH

- location

D12-IR D12-FE

- acceptance criteria 1.40 110%

1.58 110%

T~

F

- measured value 2.362 2.639 g

I

location

D12-IR 9 77" D12-FE 9 77"

-F

- measured value 1.522 1.m -

2 l

QUADRANT. TILT-measured value 1.004 1.008

- acceptance criteria -

<1.020

- by quadrant 1.000 1.002

.998 1.005

.995 1.003

.994 1.003

Assembly locations (i.e., D12) as shown in Figure 1.

Pin location within assembly (i.e., IH) based on 17x17 matrix ranging from AA to QQ.

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-1.5

.4 1.0

.9

.7

.4

.599 1.057 1.142 1.076 L.165 1.10B 1.118 1.066 1.106 1.108 1.145 1.0M 1.142 1.J36

.587 11 1.9 1.9 1.9

.'1

.2

.9

-2.0

-3.2

-3.0

.9 t.0 2.0 2.0

.2

.916.M7 1.332 l.093 1.183 1.116 1.142 1.100 1.153 1.103 1.339 1.014

.915 II 1.7 1.4 f.1 L.7 1.5

-2.0

-3.3

-3.4

-1.0 2.6 "2.7 3.2 1.6

.526 1.102

.998 L.134 1.1M 1.130 1.0B1 1.134 1.100 1.150 1.000 1.119

.537 13 1.1 1.1 1.5 l.3

.1

-2.1

-3.4

-1.7

.3 2.7 2.5 2.7 s.2

.526

.913 l.053 1.012 1.056 1.020 1.055 1.039 1.062

.917

.533 g

1.1 1.3 l.5

.1

-1.8

-1.7

.1 2.8 2.4 1.8 f.5

.Mt

.Ess

.stl.739

.sst.70s

.502 15

.7

.8

.6

.4 2.0 2.9 2.4 agem R

P R

95 L

E J'

R S

F E

B C

5 A

DIABLO CANTON POWER PLANT - UNIT 1 CORE AVERAGE RADIAL POWER DISTRIBUTION - ALL RODS OUT Figure 16 45

u

, 7 1-1 SE1ATIVE AS$DSLT j

4.

PSER (Pt)

F' M ASUSED Pt - EXPECTED P1 I I" (IPECTED P1

.391

.405

.585

.599 591

.'405

.402 I.

-4.4

-4.1

-2.0

.2

-1.0

-3.9

-3.8

.563

.899

.761

.430

.791

.899

.7H

.439

.805

.8M

.564 2

2.2 2.2

-4.3 4.5

-2.2

.6

-1.2

-2.4

.8 1.9 E.3

.563 1.200 1.052 1.019

.921 1.096 1.127 1.111

.978 1.095 1.054 1.203

.566 3

2.1 2.1 1.7

-4.3

-1.2

.3

.1

.6

.6 1.9 r.3 2.7

.899 1.055 1.478 1.240 1.350 2.292 1.366 1.297 1.322 1.217 1.487 1.066

.911 4

~

2.1 2.0 2.0 2.9 2.9

.3

.1

.8 1.0 2.6 3.0 3.5

.416

.009 1.095 1.214 1.428 1.401 1.437 1.383 1.436 1.393 1.420 1.241 1.093

.812

.400

.6 3

.7

.8 2.3 2.2 1.8 1.8 1.8 1.6 1.7 3.0

.5 0.0

-4.3

.406

.437

.962 1.304 1.381 1.413 1.235 1.302 1.220 1.387 1.354 1.347

.990

.451

.404

-3.8

-3.0

-2.1

.5

.7 1.3 1.0 1.0

.3

.6

-1.3 2.8

.7

.3

-4.3

.57 9

.792 1.092 1.278 1.403 1.225 1.164

.M5 1.154 1.206 1.407 1.310 1.112

.806

.588 7 -3.0

-2.1

-1.6

-1.4

.6

.2

.7

.6

.1

-1.4

.3 1.1

.2

.3

-1.5 g

.582

.887 1.103 1.347 1.339 1.285

.958

.573

.951 1.283 1.358 1.375 1.136

.908

.5M

-2.6

-1.9

-1.1

-1.4

.3

.1

.8 s4 0.0

.9 1.1

.5 0.0

.581

.792 1.091 1.280 1.389 1.205 1.140

.949 1.165 1.233 1.431 1.306 1.121

.802

.596

-2.7

-2.0

-1.7

-1.2

-1.6

-1.5

-1.4

-1.0

.8

.8 1.4

.8 1.0

.2

.415

.441

.953 1.313 1.371 1.381 1.199 1.271 1.217 1.404 1.389 1.329

.999

.444

.417 10 -1.7

-2.0

-3.0

.1

-1.0

-2.0

-1.4

.5

.6 1.3 1.3 1.6

-1.4

-1.1 l

.417

.810 1.085 1.208 1.405 1.367 1.399 1.331 1.388 1.366 1.429 1.233 1.117

.807

.415 I

.4 ~

.3

.2

.3

.6

.3

.9

-2.0

-1.7

.4 2.3 2.4 2.7

.6

.6 i*

.883 1.043 1.47 5 1.228 1.336 1.283 1.330 1.265 1.305 1.238 1.490 1.074

.896 r -

12

.3

.8 1.9 2.0 1.9

-1.0

-2.3

-2.4

.5,

2.8 2.9 3.8' 1.7 l

.554 1.182 1.044 1.103

.M4 1.090 1.089 1.067

.943 1.091 1.067 1.216

.572

.5

.9 1.5

.1

-1.8

-3.1

-3.9

-4.0

.3-3.1 3.4 3.8

.554

.883

.813

.445

.788

.881

.700

.425

.801

.903.549 34

.5

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-5.5

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.412

.413

.585

.586

.577

.399

.412

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-2.1

-2.0

-1.9

-3.3

-5.4

-1.4 g

i Os m R

P N

N L

K J..

N G

F E

D C

B A

DIABLO CANYON POWER PLANT - UNIT 1 CORE AVERAGE RADIAL POWER DISTRIBUTION - CONTROL BANK D INSERTED v

r Figure 17 e

9 46

1

~7 I

I I

I j

1 i

DIABL O CANYON POWER PLANT.

j UNIT I CYCLE 1 j

. ROD WITH0RAWAL LIMIT 8 REQUIRED j

l TO MAINTAIN MODERATOR TEMPERATURE' o

nem t

COEFFICIENT (MTC) LESS THAN ZERO.

i h

+

~

4

\\

f h

(

i s

yh h N

"E * *,

n s

I p

g i

h p

M N

==

g,-

N N a.

E, t

=

I.

N W

\\

~~

_a A

C u

atma 40E.

s 3e 8

N NN 7

ses 1

u

==

'""==

- a w

'wllll:::::

m g

N

===--

==

me e e l T

aos E

A essa

,es

.a l

m E8 l

M-N

=

l I.:

m :

in we see ne saan causen eAn e a

. centest ens c i

i a

i i

~

ear se IN 18e let

.coutest east a.

STEPS WITimitAlfil 3M tas as utileRARIRL LintT5 W MINTAIR NEGATIVE IEBEMTOR TDFERATURE COEFFICIENT

' Figure 18

c 3.3 Test Procedure No. 41.4 & 41.5 - Rod and Boron Worth Measurements TEST OBJECTIVE The objective of these tests was to determine the reactivity worth of each control bd'ak, total reactivity worth of control banks with normal 100 step overlap and the average boron re, activity worth.

s

- TEST' DESCRIPTION

. Individual Control Bank Worth With' all control rods withdrawn, a reactor coolant system boron dilution was es-tablished. The control banks were inserted to compensate for the resulting reactivity gain. The sequence of individual control bank insertion was Control Bank D, Control Bank C, Control Bank B, then Control Bank A.

The reactivity changes were recorded using the reactivity computer. The data obtained was used to develop integral and differential bank worth curves. Figures 19 through' 22 show these curves.

Control Bank Worth With Normal Bank Overlap With all shutdown banks withdrawn and all control banks inserted, a reactor cool-ant system boration was established. The control banks were withdrawn to compensate for.the resulting reactivity insertion.

The integral and differential worth curves are shown in Figure 23.

Boron Worth

- The average borc/a reactivity worth was determined using the reactivity data ob-l, tained during the control bank worth seasurements and the change in boron end-point concentration from the ARO configuration to the all control banks inserted

[

configuration.

l' l

TEC* RESULTS l

The measured values agreed well with predictions as can be seen in Tables 8 and 9.

1 O

-f p.

s 6

E

.a 48

[

.t i

j

+

F F

F a

a U

T i

i F

R s

s N

p p

/m

s e

4 D

s n

s

/

e n

5 s

p e 7 9

u l

a 4

0 m

5 5

i i 's 8

0 0

0 r

n u6 6

3 3

n a

9 v a, t

2 2

2 1

r P

4 4

e2 2

2 o

s 1

4 e

2 2

2 M

E 5

5 r2 2

u t

0 s

o a

O l

n P

e r

u e

i r

e s

e l

l o.

p s

r t

a I

mt el o

i a

e s r a C

d B

r T i l P i l e

A e t a t a m

C n s C mA cw S i n S i n g i

t C

C o 0 O a C D c e C ni C ni v 8

R ss S C cC C aF A I F R I F A

" P.

ts t.

e I

U" 7 o T

1 2

3 4

5 atg e"E 0

0 8

8 e.

8 s.

8 0'

0 0

0.

0 4

2 0

8 6

4 2

8 8

6 4

2 0

H 2

2 2

1 1

1 TR O

W 1

L 0

A 2 T R

I G

2 N

E U

T N

0 I

0 T D

l 2 N N

o A

A

,o L

9 P

L 1

I l-0 A

8 R I

l

'l

!i

{i P

R Il

}

!'1 l

e

'l1 1

E T

r

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i'l l

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g l

0 E

F Il iill 6 N F

l 1.

l

,1 O

F 1

'lI

,i lll llJ..

N D

Y I

A 0 C D

4 1

O K

L N

B A

A B

i 0

I 2 D LO 1

R m'

T N

0 O

l i

0 C

i 1

l q'

80

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l 0

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/

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1l Ilj i

5 7i, -

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g e

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2 3

1 1

3 2"D iE8

5 E

t

)i 1lI

!l!:i' jlji!

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+

o l.

Mutt:

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j

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41.4 2000 :}l.

n lii 24.0

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e

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>=

g 1600.lB li11]

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m

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illi!!j a

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4gg 4a0 Inittet 2235 psis Il:ll l : illg1!"ipn n!:.. li!::!:!![jn::::; ;;n "} n :::.l!!!H!ii;4il! I! l i! jIj

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g'ggPlI 5'! j!hjiii1i111jilijiiii il I lilillh,.I i ii! rii iii lili ;i!j iiii!iiiiiiji ij!:

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5.

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200 l

- !!!!!!!!il liiiiiii 0

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l monun: ! !ilisi n:: i g

mil t

I-0.0 0

20 40 60 80 100 120 140

.160 180 200 220

.it DIAOLO CANYbit POWER PUWIT - UIIIT 1 COIITROL BAIIK C DIFFERENTIAL AND INTEGRAL WORTH Figure 20

t b

O 2488

!I' I!

l l

'l l ll[

l l

ll lil 24.0 I

I 'I q

i l

t l

Unit 1

f'l 220s

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22.0 4:,

.I' ll l g.

T* F*8 4I'4 i.

ls

,l I..,

- i n i,

- e est.:

vue 2e to.e

l

i i

ipl-Test caindittenes

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~

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[ geen.

g 16.o 9 soA 22e

$ lese (.

. k t

a ses 228 S

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4 ;

/i l

14.0 sec -

22e

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l l

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. :, j g

b I

5 I"

l l

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cA 22e 12ee 1

12.0 g 3

ce styring o

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less 10.e p j

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us I

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p i,

a 2

Power I,oret:

, d

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i

/

y

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% RTP ese

...u i

[

N 3.

RCS Temperatures 4ee I

<..i;,

4.o Initiat 545.4

'y Final 546.9

'r

' i 2ee 2.s 4.

RCS Pressures l

"-i fattist 2235 psia

.i e

1 e.e pinet 2235 psia 0-20 de 60 80 les 120 140 160 Iso 200 220 3*

8*

8P8 DIABLO CANYON POWER PLANT - UNIT 1 9

leID/Ilftf CONTROL BANK B DIFFERENTIAL AND INTEGRAL WORTH Figure 21

2400 l'I,!

];!

II l

24.8

lj~ll{il l

tinitt 1

' II I

l I

I i I

l gH li j

0288 EA 4j j.

ji I

I

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,l

[lj 22.8 T. P.:

41.4 i,

q u

g I

{

j l lll Date:

5/3/54 2000 % r.1 h <.4I l f 4

20.0

)E

$.li h*'

Il lll Test Conditlems:

- 9 t..

lli i

e i

i 4

i 8

18.0 1

RCC Senk Festtlems:

- gegg l)~

Son 220 i

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a i

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3 l

8llq

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i:i

~

3 i

12.0 l CA newan.

.8 A,;l, 4

_ 120e f

o' p

E CS a

j j i

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CC a

1' j

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't 10.0 5

sees Co o

i

,.ll l

s H

gi RCCA m/A l

l ese

- d.e g:

2 hver I.evel:

iii l

i i

a l

s i

e

% RTP ese

- -i q

[

4.0 4

f 5

[.

3 RCSTemperhures g

p!II f

h Initial 547.5

F i

i 400

' li.

d 9

n +

4.0 I

I I/

5 Final 547.2

F l

.i

.in j-1 4

RCS Pressures 200 2'. 0 4

u Initist 2232 psig

_i

{

j g

g j

Finst 2245 psia 0.0 j

e 20 40 60 80 100 120 140 160 1A0 200 220 5

Arg. Core Burmup:

DIABLO CANYON POWER P p MT. UNIT 1 a

Def0/9fru CONTROL BANK A DIFFERENTIAL ~ AND INTEGRAL WORTH Figure 22 i

i

an.e See:

--'J" g

tinitt 1

88*8 T. P.:

41.5 gge; Date 5/6/84 i

tess i

as.e io;

,4'

'l R--

i

-d Test Condittenes l

I !

2 o.

C m

m i

3e.e 1.

RCCBankFostthens:

3 F

9 i,

-ll.l n.

e a

h SOA 228 g nase.

i :

.1.

1s,e I

a SDE 228 I-s l

5 I

30C -

228

[

o, i, p ysees.

i ;

l4.e

~

300 228 E!

.,.c

, g a

j l-Q CA Movino i

a o s '-

i.

Ia.e l II

. [.

l*

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[.

l i

vi i '

d CC Movina I

toes

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i n

j

('d.,

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l g,,,,

p i

i l

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. li l

P i

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t o

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% RTP lass [

i J-g

/

l

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l 3.

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1 ese L"

- h-l '- c'N 4.s inisset 547.2

'f, J '.a.

h l'l

\\

4 I

'r.

Final 547.0

p

-1

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I.

e,e 4

SCS Pressures i

g,;

L h,.

I l

Inittag 2235 psis ll H

i 2

u 1

a

'i,-.

e,e P!nal 2235 pois I

=

s. Avs. core seway l

,=

'4,,

6 s

s

=== * '

A

'a=

o nuarnre NABb CkNYON b ER kANT b NIT *Y

i I

CONTROL BANKS WITH OVERLAP DIFFERENTIAL AND INTEGRAL WORTH 4

1

w._.-

Table 8 Measured versus Predicted control Bank Worth

- Control Bank Measured Worth Predicted Worth (pca)

(pca)

CD 937 918 1 92 CC 1031 978 i 98 CB 8b5 795 1 80 CA 1250 1195 1 120 Total 4023 Within +4% of the Total Control Banks Measured Worth of Individual Banks in overlap 3987 pcm 4023 1 161 pea O

e e

f L

4 I

+

6 9

9 e

E 54 3

6

-,,.,n,en-v-----

I Table 9 Measured Versus Predicted Average Boron Worth Control Bank Bank Worth C

Boron Worth g

~

(Pes)

(ppm)

(pen / ppa)

.CD

-937 89.5

-10.47 CC

-1031 94.0

-10.97 i

CB

-805 77.5

-10.39

'CA

-1250 115.5

-10.83 1

- CD+CC+CB+CA

-4023 376.5

-10.69

- CD,CC,CB,CA in

-3987 376.5

-10.59 overlap Average

-10.64 Predicted

-10.4 +1.0 w

i e

i a

r 4

l

+

t i-e

+

e i

i

~

e l

55

3.4 Test Procedure.No. 41.6 - Rod Cluster Control Assembly (RCCA) Pseudo Ejection At Zero Power JTEST OBJECTIVE The objective of this test was to simulate ejection of the most reactive control rod (506) with control banks at the sero power insertion limit. Integral worth of Rod 506 was measured and compared with design values. Core power, distribution was determined to verify that hot channel factors were within predicted values and FSAR limits.

TEST DESCRIPTION Reference conditions were established with the reactor critical in the sero power test range (10-8 to 10-7 amp). Control rods were positioned at the sero power rod insertion limit and Control Bank D adjusted to reference conditions as shown below:

S hutd own Ba nks A, B, C, D..................................... 2 2 8 s te ps Control Bank A................................................ 228 steps Control Bank B................................................ 183 steps Control Bank C................................................

55 steps Control Bank D.................................................

5 steps A baseline incore flux map was obtained and analysed.

Lif t coil disconnect switches for all rods in Control Bank D except rod 506 were opened and a 300 pen /hr continuous boron addition commenced. Criticality was maintained by withdrawal of the ' ejected rod', 806. During withdrawal, integral rod worth for 506 was measured. With B06 near the full out position, boron addition was stopped. 306 was then fully withdrawn and reactor control shifted to Bank B.

The ejected rod incore flux acp was then taken and analysed.

Following the flux asp, a dilution was commenced and Rod 506 was realigned with the other rods in Control Bank D.

During the dilution, additional rod worth measurements for 506 were taken. Lif t coil disconnect switches were then closed and normal rod configuration maintained.

l l

TEST RESULTS i

l The reactivity worth seasured during withdrawal of control rod 506 was 275 pcm.

The measured value plus 10% for uncertainty was 303 pca. This was considerably l

less than the design value of 480 f48 pen but well within the FSAR limitation of less than 785 pcm. The rod worth seasurement taken during the dilution phase resulted in a measured value of 275 pen. Due to the excellent repeatability of the two measurements, the reedits are considered to be accurate. Westinghouse was informed of the discrepancy between predicted and measured worth and accepted the results. Since the rod worth is lower than predicted, no safety review criteria have been violated.

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Summaries of the power distributions are shown in Figure 24 and Table 10. The

'ejectedrod'fluxmapshowedtheheatfluxhotchannelfactor,Ff,tobe5.63-.

This is well within the FSAR acceptance criteria limit of less than or equal to 13.

TABLE 10 Power Distribution Results (Pre and Post Pseudo RCCA Ejection)

PRE-FJECTED POST-EJECTED ITEM FLUX MAP FLUX MAP l

CONDITIONS - temperature

~545 deg. F

- boron conc.

1200 ppa 1200 ppe

- power 0%

0%

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- burnup 0 MWD /NTU 0 MWD /NTU DATE May 5, 1984 May 5, 1984 l.

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- measured value 1.683 3.357 AH

- location

E06-IA B05-AQ T

F

- measured value 2.974 5.633 E

- location

M12-LE 9 36" B07-AA 9 55" QUADRANT TILT - measured value 1.004 1.660 i

- by quadrant 0.997 1.003 0.708 1.624 1

i 0.999 1.002 0.603 1.066 l

l l

Assembly locations (i.e., E06) as shown in Figure 1 l

Fin location with'in assembly (i.e., IA) based on 17 x 17 matrix ranging from AA to QQ.

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DIABLO CANYON POWER PLANT - UNIT 1 E ASURED FDM (PRE AND POST PSEUD 0 RCCA EJECTION)

Pre Qection

Figure 24 Post Q ection s

58

3.5 Test Procedure No. 41.7 - Minimum Shutdown Margin Verification and Stuck Rod Worth Measurement TEST OBJECTIVE.

The objectives _of this test were to 1) measure the reactivity worth of the Shutdown Banks, 2) measure the critical boron concentration with all full length rods inserted and the most reactive rod (r10) fully withdrawn, and

3) measure the reactivity worth of the most reactive rod.

TEST DESCRIPTION The test began with the reactor critical in the sero power test range with all control banks inserted and all shutdown banks withdrawn. A 500 pen /hr RCS di-lution was then initiated. Individual shutdown banks were inserted to compen-sete for the dilution and their reactivity worths measured.

i Upon completion of the worth measurements for Shutdown Bank D, preparations were i

ande to enter the Technical Specification Special Test Exception for minimum t

shutdown margin. This required demonstration of tripability from at least 50% withdrawn of each full length rod not fully inserted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the shutdown margin to less than 1.6% Ak/k. Therefore, with Shutdown Banka A, B and C fully withdrawn, the reactor was tripped. Shutdown Bank D was then withdrawn to greater than 114 steps and tripped. These actions met the requirements for entry into Test Exception 3.10.1.

The reactor was then returned to criticality with all rods inserted except Shutdown Banks A. B and C fully withdrawn.

RCS boron dilution was then commenced and Shutdown Bank C rod worth was measured.

"The dilution was stopped.

Shutdown Bank D (the bank containing the most reactive rod, F10) was then pulled to 5 steps. Lif t coil disconnect switches were opened for all rods on Shutdown Ba'ak D with the exception of F10.

L t

While maintaining criticality, Shutdown Bank B was exchanged with Rod F10. As Shutdown Bank B reached the fully inserted condition, the exchange with F10 was continued using Shutdown Bank A.

When Rod F10 reached its fully withdrawn condition l

a dilution was commenced to allow the insertion of the remainder of Shutdown Bank A.

At this point the reactor was critical with all rods inserted'vith the exception of l-the most reactive rod, F10, and shutdown Bank A within 20 to 24. steps from the

~ bottom. This was considered to be the design All-Rods-In N-1 configuration. The boron endpoint was obtained for this condition.

l' Once the boron endpoint was obtained, the reactivity computer was rescaled so as to observe the reactivity insertion associated with tripping F10. Once ready, the stationary gripper coil fuses for Rod F10 were pulled, dropping F10

- into the core.

s The reactor trip breakers were opened and the RCS borated until conditions were reached to achieve criticality with all shutdown banks withdrawn.

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59 L

7 TEST RESULTS No abnormalities were observed during this testing. The measured integral rod worth for Shutdown Bank D of 948 pen was slightly outside the design value of 854 pen 110%.. Westinghouse later reevaluated the worth of Shutdown Bsak D and pro-vided a new design.value of 941 pcm.

The measured idtegral rod worth for Shutdown Bank C of 905 pea agreed well with the design value of 892 pea 110%.

The critical boron concentration for All-Rode-In Minus F10 configuration was measured at 62714 ppa. This agreed well with the design value of 611 150 ppa.

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4.0 Test Procedure No. 3.7 Addendua 3 - Turbine Driven Auxiliary Feedwater Pump Endurance Test 7 OBJECTIVE l

The objectiv/ of this test was to demonstrate the reliability of the Turbine Driven Auxiliary Feedwater Pump (AFW Pump 1-1) by operating the pump for an extended period of time to comply with the requirements of NUREG-0737.

i..

TEST DESCRIPTION With the reactor at approximately 4.4% power, sufficient to support rated flow from the Turbine Driven Auxiliary Feedwater Pump (1-1), AFW Pump 1-1 was run in the mini-mun recirculation mode. Af ter an initial inspection of all components, the pump was 4

aligned to supply water to the Steam Generators and return it to the Condensate Storage Tank via the Condensate Reject Loop from the hotwell. Feedwater flow to the Steam Generators was established and the system was run at rated flow for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . During this endurance run, selected data such as pump head, pump flow, yturbine bearing temperatures, pump bearing temperature, vibration readings, pump room temperature and humidity were monitored. Fusp flow was allowed to remain above rated flow of 880 CPM during the test. At the end of the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months run flow was adjusted to rated flow and all the parameters recorded to ensure pump perfor-mance had not degraded. Af ter 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the pump was shuPdown and pump temperatures were allowed to cool to within 20 deg. F of their initial values. Following the cool-down the pump was restarted and run for one hour at rated flow.

1 TEST RESULTS Auxiliary Feedwater Pump 1-1 operated satisfactorily and all acceptance criteria were met. During the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months run, pump flow remained above its 880 CFM minimus, pump suction pressure varied between 16.9 and 21.5 psig (>_10.14 psig required),

and pump differential pressure remained very close to 1400 psid (>1253.4 psid required). Bearing temperatures and vibration readings were within limits for both the AFW pump and turbine.

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4 5.0' llPECIAL LOW POWER TEST PROCRAM 4

I SUMMAAT A series of Special Low Power Tests, conducted to provide supplementary techni-cal informatioh~ and operator training, were performed at the Diablo Canyon Nuclear Power Plant, Unit I between May 14-23, 1984. These tests, as listed below, fulfilled certain post-TMI action items of NUREG-0737 and consisted of the seven tests outlined in Supplement 10 of the Diablo Canyon Safety Evaluation Report (SER) dated August 1980 - and evaluated in detail in Supplement 14 of the SER dated 4

April 1981.

Procedure Test i

No.

Title i

44.1 1.1' Natural Circulation 44.1 1.2 Natural circulation with Loss of Pressuriser Nesters 44.1 1.3 Natural Circulation at Reduced Pressure 44.1 1.4 Natural Circulation with Simulated Loss of Off-Site AC Power

- 44'.1 1.5 Effect of Steam Generator Isolation on Natural Circulation 44.2 Cooldown capability of the Charging and Letdown System 44.3 Simulated Loss of All On-Site and Off-Site Power The last two tests were conducted prior to establishing natural circulation condi-tions and T.F. 44.2 was repeated for operator training. Of the five natural circu-lation tests,1.1 through 1.4 were conducted in succession and then repeated for further operator training. Test 1.5 was completed once at the conclusion of the

, test program.

Results of the Natural Circulation Testing demonstrated the stability and cool-ing capability of Diablo Canyon Unit I under natural circulation conditions.

The program also demonstrated the ability of the licensed operators to establish, maintain, and recover from natural circulation in an orderly fashion.

Startup Test Procedure No. 44.1 was the governing document for the five natural circulation tests, Numbers 1.1 through 1.5.

The procedure outlined each test and gave the prerequisites, initial conditions, and detailed instructions for establishing," amintaining, and recovering from the various configurations or conditions. It also provided a detailed procedure for temporary modifications to plant safety systees that were required in order to operate the reactor at low power without reactor coolant pumps (RCPs) running. Safety aspects of the procedure were evaluated in Supplement 14 of the Diablo Canyon SER.

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5.1 Test Procedure No. 44.1 Test 1.1 - Natural Circulation TEST OBJECTIVE The objective of this test was to establish, maintain and recover from natural circulation co'nditions while at low power.

TEST DESCRIPTION With the reactor at about 3% power, the reactor coolant system (RCS) pressure at a reduced value of approximately 2135 psig, and steam being dumped to the condenser, all four reactor coolant pumps were tripped in rapid succession.

Pressuriser spray valves were placed in manual and opened, and one power oper-ated relief valve controller was temporarily placed in the closed position to prevent its opening during the expected pressure rise.

Temporary modifications had previously been made to plant safety systems to pre-vent automatic initiation of safety injection due to the off-normal operating conditions encountered during natural circulation. Plant conditions were moni-tored and recorded during the subsequent stabilization of plant parameters.

Adjustments were made to RCP seal water flow rate, charging flow rate and auxiliary feedwater flow rate to maintain stable conditions. When all plant parameters had stabilized and training was completed. one of the other five natural circulation tests was begun or the plant was returned to forced circulation.

Recovery from natural circulation was achieved frca stable natural circulation conditions with pressuriser heaters operating. The controlling bank of control rods was fully inserted, and the RCPs were restarted one at a time.

Pressurizer spray valves were realigned for normal operation and the plant was stabilized at normal operating conditions prior to withdrawing the controlling bank to a critical configuration and returning the reactor. to about 3% power.

TEST RESULTS Results were consistent with the response of other Westinghouse pressurized water reactors. Following the trip of the RCPs, loop cold leg temperatures dropped a -

few degrees while hot les temperatures rose roughly 35-40 deg. F, as shown in Figure 25.

RCS loop temperatures stabilised within about 15 minutes. Calculated loop A T(Thot Teoid) never approached the limit of 65 deg. F maximum. The hotter RCS temperatures, combined with a slightly negative temperature coefficient of reactivity, caused a small reduction in power. This was compensated for by the operators manually withdrawing control rods to maintain constant power. The entire test was conducted with Control Bank D rods at about 150-190 steps withdrawn.

Pre'eauriser pressure increased during the initial transient as expected and peaked at about 2195 psig af ter about 7-8 minutes, as shown in Figure 26. Pressure slowly decreased. thereaf ter without any requirement for initiating auxiliary spray.

Pres-suriser level required approximately 8 minutes to reach a peak of 33.5% from an ini-tial level of 25%.

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4 Core exit temperatures, as monitored by thermocouples, stabilized and showed no evidence of hot spots in the upper plenus or head region. Temperatures remair.ed below the limit of 610 deg. F maximum. Subcooling margin in the reactor vessel stabilized at more than 60 deg. F below saturation.

The full core flux asp was obtained using movable incore detectors with the plant stabilizad in natural circulation. Results were consistent with plant conditions, and peaking fac~t' ors all were acceptable. There was no appreciable flux tilt nor evidence of poor mixing or hot spots.

Recovery from natural circulation was achieved with only the expected minor per-turbations in plant pressures and levels, well within the capability of automatic control systems.

Test 1.1 demonstrated that licensed plant operators could establish, maintain and recover from stable natural circulation conditions with the reactor operat-ing at low power. The planning and execution of these evolutions provided ample opportunity for training all operators in the core cooling capability of Diablo Canyon Unit I without the benefit of reactor coolant pumps.

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~5.2 Test Procedure No. 44.1 Test 1.2 - Natural Circulation With Loss of Pressurizer Heaters Test 1.3 - Natural Circulation at Reduced Pressure TEST OBJECTIVES' These procedures had a number of objectives:

1)

To determine the rate of decrease of margin to saturation, 2)

To verify recovery of margin through cooldown and make-up, 3)

To verify operation and accuracy of the subcooling margin monitor, 4)

To verify that changes in margin will not affect natural circulation, provided natural circulation exists, 5)

To determine the eff'ect of auxiliary spray on pressuriser pressure, 6)

To determine the position of the normal spray valves to obtain a sufficient back pressure for the auxiliary spray valves.

TEST DESCRIPTION Natural Circulation with Loss of Pressuriser Heaters and Natural Circulation at

= Reduced Pressure tests were performed concurrently. The description and results given below apply.to both tests.

Initial conditions required that the reactor be critical at approximately 3% power under natural circulation conditions. A transient was then initiated by deenergizing the-pressuriser heaters. Reactor coolant system pressure was monitored to determine the rate of depressurization and resulting -loss of margin to saturation.

l The auxiliary pressuriser spray valves and, as required, the normal spray valves were employed to accelerate the rate of depressurization.. Core exit thermocouples were monitored to determine core flow distribution. The operation of the subcooling margin l'

monitor was verified through hand calculations.. Finally, prior to reaching saturation conditions, primary system charging flow was used to recover margin prior to restoring the pressuriser heaters to service.

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TEST RESUI.TS l

Upon de-energizing the pressurizer heaters, the plant responded as expected with a steady decrease in primary system pressure. Over a period of approximately three hours, pressure dropped from approximately 2235 psig to approximately 2020 psig, a rate of 72 psi /hr (corresponding to approximately a 4 deg. F/hr of decrease of margin to saturation). When auxiliary spray was introduced, the depressurization rate increased to roughly 600 psi /hr. Consequently, pressure control through the use of the normal spray valves was not necessary. The transient was permitted to continue to an RCS pressure of 1710 peig (a subcooling margin of 29 deg. F) before-charging flow was successfully used to demonstrate the ability to restore margin.

Restoration was then completed by using the pressurizer heaters. Fressurizer pressure for the entire transient is illustrated in Figure 26. The primary system subcooling margin monitor was found to be both reliable and accurate; hand calculations showed excellent agreement with the monitor readings throughout the depressurization.

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5.3 ' Test Procedure No. 44.1 Test 1.4 - Natural Circulation With Simulated Loss of Off-Site AC Power TEST OBJECTIVE l

71:e objective 6f this test was to verify that natural circulation cooling can be adequately maintained following a simulated loss of off-site AC power.

TEST DESCRIPTION i

Initial conditions were achieved with the reactor critical and operating at approxi,mately 1% power under natural circulation conditions. Loss of off-site power to the vital 4160 vac buses was.then simulated by opening the main. supply breaker, 52-HG-15, from the main control board. That initiating event caused the diesel generators to automatically start and pickup loads. Natural circulation i

conditions were maintained and monitored. Core exit thermocouples were used to t'

monitor core flow distribution.

l

' TEST RESULTS The plant response to the s'inulated power loss proceeded as expected. All diesel 1

generators automatically started and picked up associated loads. Steam generator levels were adequately controlled through the use of the auxiliary feedwater system.

Fressurizer pressure was controlled with vitally powered heater groups, while 1

pressurizer level was maintained using normal letdown and charging. Finally, steady state natural circulation conditions were maintained without any problems.

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- 5.4 Test Procedure No. 44.1 Test 1.5 - Effect of Steam Generator Isolation on Natural Circulation TEST OBJECTIVE Th'ere were two objectives associated with this test.

1)

Monitor the effects of isolating two steam generators while on natural circulation.

2)

Determine the effect of reduced reactor coolant system average

, temperature (Tavg) on the nuclear instrumentation system (NIS).

The first objective was to ensure the plant would respond.as designed should one or more steam generators require isolation while the RCS was in natural circulation.

This configuration would be required in the event of a steam generator tube rupture.

Primary concerns centered on,1) amintaining natural circulation cooling in the non-affected loops, 2) correct RCS parameter response in the affected loops and

3) control of the pressure on the secondary side of the isolated steam generators.

The second objective was to observe and quantify the change in coupling between the reactor core and NIS with reduced RCS Tavg. With c'ooler water in the reactor vessel downconer region, fewer neutrons are sensed 'by the NIS. This results in

. indicated reactor power being less than actual power. This effect required seasurement prior to commencement of natural circulation with reduced Tavg to establish limits for indicated' reactor power.

TEST DESCRIPTION Initial'ecaditions required the reactor to be critical at 3% power with all reactor coolant pumps running. RCS temperature and pressure were' maintained at 547 deg. F and 2235 pois respectively.

' At ~ the outset, RCS pressure was reduced to 2000 psig to minimize the potential for exceeding the 1600 paid limitation across the steam gecerator tubes. Subsequent

. steps would reduce the secondary side steam pressure to approximately 735 psig.

- If normal operating pressure were maintained, an approximate 1500 psi differential across the steam generator tubes would result. The reduced RCS pressure thus effected an increased margin to the 1600 paid limit.

. A 5 deg. F/ step cooldown was then initiated from 547 deg. F to $10 deg. F.

At each

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plateau, RCS parameters were stabilized at the core delta-T corresponding to 3%

reactor power.' RCS temperatures and'NIS data were then recorded. Though the reactor power!1evel was maintained at a constant 3%, at each temperature plateau, NIS indicated power decreased at a rate of 0.036% per 1.0 deg. F.

This number ~

was, consistent with Westinghouse predictions. At the end of the cooldown, all data was ec11ected and reduced. ~From the information gathered, a plot was gener-ated and provided for the operators showing NIS indicated power level vs temp-erature for the 5% licensed power limitation.

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...,_.2 While maintaining RCS temperature at 510 deg. F and pressure at 2000 psig, reactor power was reduced to 1%. Once stable, initial conditions for entry into natural circulation were verified and RCPs secured. Natural circulation was then estab-lished and stabilized.

Steam Generator 1-3 was then isolated by closing the main steam isolation and bypass valves, securing auxiliary feedwater, and isolating steam generator blow-

~

down. As expected Teoid for loop 1-3 began to increase toward Thot resulting in a corresponding increase in Steam Generator 1-3 pressure. Parameters gradually stabilized until the delta-T across the isolated steam generator decreased to less than 10 deg. F, as shown in Figure 27.

After the 10 deg. F stabilization criteria for the first steam generator was set, the second steam generator,1-1, was isolated. Again, plant parameters responded as expected with no anomalies noted.

Following the required stabilization period, the steam generators were unisolated in the reverse order that they were isolated; that is, Steam Generator 1-1 was returned to service first followed by Steam Generator 1-3.

This was done by equal-izing across and opening the main steam isolation valve and restoring auxiliary feedwater. The entire transient is covered by Figures 28 and 29.

i TEST RESULTS The entire evolution progressed in a slow and controlled manner. At no time were any saomalous plant responses or conditions observed. By using the steam dump system, operating personnel were able to control RCS temperatures and thus control isolated steam generator pressures.

Test 1.5 demonstrated that natural circulation could be maintained following isolation of one or more steam generators. All plant responses were consistent with expectations and results obtained at similar Westinghouse pressurized water reactors. Operating personnel demonstrated their ability to initiate, maintain and recover from a steam generator isolation during natural circulation conditions.

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DIABLO CANYON POWER PLANT - UNIT 1 RCL TEMPERATURES DURING STEAM GENERATOR ISOLATION Figure 28 73

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5.5 Test Frocedure No. 44.2 - Cooldown Capability of the Charging and Letdown System TEST OBJECTIVE.

The objective of this test was to determine the capability of the chemical and volume control system (CVCS) charging and letdown system to cooldown the reactor

' coolant system (RCS).

TEST DESCRIPTION With the RCS in Mode 3 (RCS temperature of approximately 540 deg. F and RCS' pressure of approximately 2235 psig) three of the four reactor coolant pumps were tripped and all steam generators were isolated. The cooldown capability of the CVCS charging and letdown system was determined from the hot and cold leg temperatures of the active loop at maximum (120 spa) and minimum letdown flow (45 spa). In addition, core exit thermocouples were monitored to assess core flow distribution.

TEST RESULTS Plant response consisted of an average cooldown rate of about 2.1 deg. F/hr with maximum letdown flow and an average heatup rate of about 0.2 deg. F/hr with min-inum flow.

Results from the initial performance of the test are shown in Figure 30. Addi-tional test performances were done for operator training, and average results are summarized in Table 11.

j Table 11 CVCS Cooldown Capability l

l Maximum Flow Minimum Flow Conditions Conditions (spa)

(spa) i Target Charging / Letdown Flow Rate 120 45 Actual Charging Flow Rate 125.6 53.6 Actual Letdown Flow Rate 116.6 44.8 Cooldown Rate 2.1 deg. F/hr l

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5.6 Test Procedure No. 44.3 - Simulated Loss of All On-Site and Off-Site AC Power i

i TEST OBJECTIVE The objective of this test was to demonstrate the ability to maintain Hot Standby conditions under simulated loss of all on-site and off-site AC power.

TEST DESCRIPTION The' initial ~ conditions for the test were 1) reactor shutdown and the RCS in Hot Standby conditions (i.e., reactor coolant system at approximately 2235 psig and 547 deg. F), 2) all four reactor coolant pumps running, and 3) steam generator levels and steam pressure being maintained with auxiliary feedwater and steam dump respectively. Selected equipment were then tripped or realigned to simulate loss of all AC power. The final initiating event involved tripping the RCP bus under-voltage relays which automatically started the turbine driven auxiliary feedwater Pump.

The reactor coolant system was maintained in Hot Standby conditions for two hours using atmospheric dump valves and controlling the steam generator levels by man-

'ual control of the turbine driven auxiliary feedwater pump and/or its discharge level control valves.

TEST RESULTS i

Flant response was satisfactory in that the turbine driven auxiliary feedwater j

pump started automatically and the operators were able to successfully maintain steam generator levels within normal operating range during the two hcur period.

Reactor coolant loop average temperatures were all maintained at or below 551 deg. F.

Voltages for all vital batteries remained above the minimum allowable value of 109 Vdc.

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PACIFIC GAS AND E LE C T RI C C O M PANY EbW3 l

77 BEALE STREET. SAN FRANCISCO, CALIFORNI A 94106. (415) 781-4211. TWX 910 372 6587

"",* ; *O"""

January 29, 1985 IsUCLEAR POWER GENERATION PGandE Letter No.: DCL-85-031 Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555 Re: Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit.1 Startup Test Report

Dear Mr. Knighton:

As required by the Operating License for Unit 1 (Section 6.9 of the Technical Specifications), the Startup Test Report for the period from fuel load to completion of special low power testing is transmitted herewith.

Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

Sincerely, Enclosure

.cc:

.J. B. Martin H. E. Schierling Service List Ra" * "

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PGandE Lettar No.: DCL-85-031 ENCLOSURE DIABLO CANYON UNIT 1 STARTUP REPORT 0098S/0025K t-- -

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DCL-85-031, Startup Rept, for 831115-841031

Contents

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DCL-85-031, Startup Rept, for 831115-841031

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