ML20207N053

Rev 0 to Palo Verde Nuclear Generating Station,Unit 2 Startup Rept

ML20207N053
Person / Time
Site:	Palo Verde
Issue date:	12/31/1986
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:
Shared Package
ML20207N048	List: ML20207N046 ML20207N053
References
NUDOCS 8701130419
Download: ML20207N053 (95)

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O O PALO VERDE NUCLEAR GENERATING STATION UNIT 2 STARTUP REPORT (Docket No. 50-529)

Revision 0 December 1986 8701130419 361219 PDR ADOCK 05000529 P PDR s

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O O PALO VERDE NUCLEAR GENERATING STATION _

UNIT 2 STARTUP REPORT TABLE OF CONTENTS SECTION TITLE PAGE Preface-------------------------------------------- 1 1.0 Introduction and Test Program Summary-------------- 2 1.1 Summary of Test Objectives by Test Phase------ 2 1.2 Chronology of Startup Testing:

Fuel Load through Low Power Physics Testing--- 4 1.3 Summary of Startup Test Results:

Fuel Load through Low Power Physics Testing--- 5 1.4 Chronology of Startup Testing:

Power Ascension Testing Phase----------------- 6 1.5 Summary of Power Ascension Testing------------ 8 2.0 Initial Fuel Loading------------------------------- 9 3.0 Postcore Hot Functional Testa--------------------- 14 3.1 Postcore Hot Functional Test Controlling Document------------------------- 14 3.2 Postcore Instrument Correlation-------------- 16 3.3 Postcore Reactor Coolant System Flow Measurement----------------------------- 17 3.4 Postcore Control Element Drive Mechanias Performance------------------------ 19

. 3.5 Postcore Reactor and Secondary Water Chemistry Data------------------------- 21 -

3.6 Postcore Pressurizer Spray Valve and Control Ad J ustments---------------------- 23 3.7 Postcore Reactor Coolant System Leak Rate Measurement------------------------ 25 3.8 Postcore Incore Instrumentation Test--------- 26 4.0 Initial Criticality------------------------------- 28 5.0 Low Power Physics Testing------------------------- 29 5.1 Low Power Biological Shield Survey Test------ 29 5.2 CEA Symmetry Test---------------------------- 31 5.3 Isothermal Temperature Coefficient Test------ 32 5.4 Shutdown and Regulating CEA Group Worth Test- 34 5.5 Differential Boron Worth Test---------------- 37 5.6 Critical Boron Concentration Test------------ 38

O O PALO VERDE NUCLEAR GENERATING STATION UNIT 2 STARTUP REPORT TABLE OF CONTENTS (cont'd)

SECTION TITLE PAGE 6.0 Power Ascension Testa----------------------------- 40 6.1 Variable Tavg (Isothermal Temperature

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Coefficient and Power Coefficient) Test------ 40 6.2 Unit Load Transient Test--------------------- 43 6.3 Control Systems Checkout Test---------------- 47 6.4 Reactor Coolant and Secondary Chemistry and Radiochemistry Test---------------------- 50 6.5 Unit Load Rejection Test--------------------- 53 ,

6.6 Loss of Offsite Power Test------------------- 55 6.7 Biological Shield Survey Test---------------- 58 6.8 Steady State Core Performance Test----------- 60

. 6.9 Intercomparison of PPS, Core Protection Calculator (CPC), and PMS Inputs------------- 68 6.10 Verification of CPC Power Distribution Related Constants Test----------------------- 70 6.10.1 Verification of CEA Shadowing Factors and Radial Peaking Factors------------------- 70 6.10.2 Verification of Temperature Shadowing Factors---------------------------- 73 6.10.3 Verification of Shape Annealing Matrix and Boundary Point Power Correlation Constants--- 75 6.11 Main and Emergency Feedwater Systems Test---- 79

,6.12 CPC Verification and COLSS Verification------ 84 6.13 Steam Bypass Valve Capacity Test------------- 88 -

6.14 Incore De.tector Tost------------------------- 89 G.15 Shutdown from Outside the Control Room Test-- 91 i

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O O 'v"$$ "" 2 STARTUP TEST REPORT PAGE 1 PREFACE The Palo Verde Nuclear Generating Station (PVNGS) is a three unit nuclear ,

power station located approximately 50 miles west of downtown Phoenix, Arizona. l PVNGS is owned by the Arizona Nuclear Power Project (ANPP), a consortium of  !

southwestern United States utilities. Arizona Public Service Company is the project manager of PVNGS for ANPP. l PVNGS Unit 2 (PVNGS 2) utilizes a System 80 pressurized water reactor nuclear stees supply system (NSSS) manufactured by Combustion Engineering, Inc.

(C-E). Systes 80 is C-E's standardized NSSS design and is described in the Combustion Engineering Standard Safety Analysis Report--Final Safety Analysis Report (CESSAR). PVNGS 2, the second System 80 NSSS to start operation, has a rated core thermal output of 3800 NWt, and a nominal not electric output of 1270 NWo. PVNGS 2 is a follow-on unit to PVNGS 1.

The objective of this report is to provide a summary description of the initial startup test program for PVNGS 2. This program consists of a series of tests which satisfy requirements of the Nuclear Regulatory Commission as detailed in the PVNGS Final Safety Analysis Report (FSAR). The FSAR references Chapter 14 of CESSAR, which incorporates the testing requirements of Regulatory Guide 1.68, Revision O. The test program summarized by this report consists of five phases:

1. Fuel Loading

2. Postcore Hot Functional Tests

3. Initial Criticality

4. Low Power Physics Testing

5. Power Ascension Testing The overall objectives of this test program are to:

a) Demonstrate that components and systems of the Nuclear Steam Supply l System (NSSS) operate in accordance with design requirements.

b) Demonstrate that the NSSS can be safely operated and that performance levels can be maintained in accordance with established safety requirements.

c) Confirm proper transient system operation and thereby verify that the NSSS can be brought to power as well as to shutdown condition in a controlled and safe manner.

d) Provide verification of core physica parameters and baseline performance data for use during normal plant operation.

This report describes the FSAR required testing from Fuel Loading through the Power Ascension Testing phase. Testing is listed and summarized by the applicable section of CESSAR. Note that the scope of testing for PVNGS 2 is reduced from that of PVNGS 1. This scope reduction results from both the elimination of those tests required only for the first-of-a-kind plant (described in CESSAR) and from exceptions taken to the CESSAR tests (described in the FSAR).

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O O PVNGS UNIT 2 STARTUP TEST REPORT PAGE 2 1.0 INTBgDMgTION AND_IEST PROGRAM _gUMMARY 1.1 Summerr_91_Innt_9haastirne by Inst _Ehste The initial startup test program described herein begins with Initial Fuel (ggding. This phase of the test program provides a systematic process for safely accomplishing fuel load. It also verifies that all fuel assemblies and installed sources are correctly located and oriented. Initial Fuel Loading is described in section 2.

E95tE9ts_H9t_Eunst19nsi Iggtg (HFT) follow Fuel Loading. .The objectives of these tests are to provide additional assurance that plant systems necessary for normal plant operation function as expected, and to obtain performance data on core related systems and components. Normal plant operating procedures, in so far as practical, are used to bring the plant from cold shutdown conditions (Operational Mode 5) to hot, zero power conditions (Operational Mode 3). The Postcore Hot Functional Tests provide the first measurements of N555 and secondary system performance with the core in place. Examples of systems tested under this phase are the control rod drive system, the reactor coolant system (RCS), and the incore neutron monitoring system. Examples of measurements include control rod drop times, reactor coolant system flow rate, flow coastdown following reactor coolant pump trips, and movable incore detector path lengths.

The Postcore Hot Functional Test phase is described in section 3.

Initigl_ Criticality follows the Postcore Hot Functional Tests. Initial criticality is performed at the normal hot zero power RCS conditions of 565 0F, 2250 psia. This phase of the test program assures a safe and controlled approach to criticality. Section 4 describes Initial Criticality.

Lgw Poygt_Ehygigs Testing (LPPT) immediately follows Initial Criticality, and is conducted with the reactor critical but producing no measurable heat.

Testing is performed at normal hot zero power conditions. This phase of testing consists of a series of measurements of selected core parameters, such as control rod worth, temperature coefficient of reactivity and soluble boron reactivity worth. These measurements serve to substantiate the safety analyses of the FSAR and the bases of the Technical Specifications relative to core behavior. The LPPT measurements also demonstrate that core characteristics are within expected limits and provide data for benchmarking the computer algorithms l

used for predicting core characteristics later in core life. Additionally, the LPPT phase includes the first measurements of radiation shielding by the biological shield. Section 5 describes these tests.

E9YSE_Asgggsion Testing (PAT), the longest phase of testing, follows LPPT.

This phase is structured to bring the reactor to full power,in stages, with testing performed at intermediate test plateaus" of approximately 20%, 50%, and 80k of full power, before final testing at full power. PAT demonstrates that

, the facility operates in accordance with its design during steady power 1.

operation and, to the extent that testing is practical, during anticipated transients.

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STARTUP TEST REPORT PAGE 3 Typically, a PAT test plateau begins with confirmation of the reactor power level by secondary heat balance, and calibration of the power instruments as needed. Next, initial plateau testing is performed while equilibrium xenon conditions are allowed to develop, after which time detailed physics testing is performed. Testing of the control systems are performed next. The following transient tests were performed for PVNGE 2:

• Loss of Offsite Power Test

= Unit Load Rejection Test Testing of the PAT phase is described in Section 6.

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PAGE 4 )

i 1.2 .ghtggolony of Etertup Testingi

.[331 Loed through Low Power Physisg_Iggling The approximate durations of the activities that took place from the start of fuel load to the completion of LPPT are shown below:

AgTIVIT_I DURATION DATES Fuel Load 6 days Dec. 11--Dec. 16, 1985 Post Fuel Load Checks and NSSS Assembly a 42 days Dec. 16--Jan. 27, 1986 Preparation for Mode 4 Entry (RCS Tavg > 210 0F) 43 days Jan. 27--Mar. 10, 1986 Postcore HFT 29 days Mar. 9--Apr. 6, 1986 Preparation for Initial Criticality 13 days Apr. 6--Apr. 18, 1986 Initial Criticality 1 day Apr. 18, 1986 Low Power Physics Testing 4 days Apr. 18--Apr. 21, 1986 132 days Dec. 11, 1985--Apr. 21, 1986

= Installation of reactor vessel head, control rod drive power cables, incore i detectors, etc.

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O O ev a= ==1T 2 STARTUP TEST REPORT PAGE 5 1.3 3MEtery of gigtigo Test Resgitg:

[ggLkggq_through Low Power Physigg TestiDS

[ggLLggd was completed in 6 days. There were no significant problems with the fuel loading, but there were minor problems with the post fuel load verification of the proper alignment of.the fuel (see Section 2.0). These problems were successfully resolved without undue delay.

E9figggg_H[T was completed in 29 days with no significant problems. The results of this test phase that are the most significant to power operations are:

• The perforasnee of the control rod drive system was excellent, and remains excellent throughout the early stages of the fuel cycle.

= The Reactor Coolant System (RCS) steady state flow rate wee 104.3* of the design volumetric flow. RCS steady state flow measurements taken during the power ascension test phase, both at zero power and at power conditions, showed the RCS flow to be higher than that measured at HFT.

• The Reactor Coolant Pump coastdown flow was measured to be slightly faster than that assumed in the safety analysis. Interim penalty factors were applied to the COLSS and CPC software. Additional coastdown data taken during the power ascension test phase was compared to the safety analysis criteria and was found to be acceptable without penalties.

Initi3 1 Criticality was completed with no significant problems. The measured critical soluble boron concentration was 1004.5 ppa versus e predicted value of 1015 ppa and was within the acceptance criteria.

Low Power _P_l}ysics_ Testing was completed in 4 days with no significant problems. All test results fell within their acceptance criteria, and all test results compared well with Unit 1 results.

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 6 1.4 ghtggglggr_of Shggtyg_Iggtiggi Power Ascensign Testing Phase The approximate durations of the various phases of Power Ascension Testing are tabulated below, and the power level history during this time period is shown on Figure 1-1.

IEgI_6gIlylIY DURATION DATEg Initial increase from zero to 20k power 30 days Apr. 22--May 21, 1986 First generation of electrical power -------

May 20, 1986 20k Power Test Plateau 15 days May 21--June 4, 1986 50k Power Test Plateau 82 days June 4--Aug. 24, 1986 80k Power Test Plateau 12 days Aug. 24--Sept. 4, 1986 100k Power Test Plateau 18 days Sept. 4--Sept. 21, 1986 153 days Apr. 22--Sept. 21, 1986 On September 22, 1986, based upon the successful completion of the Power Ascension Test program and of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous operation above 95k full power, Arizona Public Service Company declared the commencement of commercial operation of PVNGS 2.

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 7-FIGURE 1-1 FVNGS Uh17 2 POWER HISTCRY:

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 8 1.5 gysserY of Power A n sion Testing The Power Ascension Test phase was completed over a period of approximately 21 weeks. This test phase demonstrated that PVNGS 2 operates in accordance.with its design during steady state power operation and during the transient testa performed. The core performance test results compared satisfactorily with the predicted values. The transient tests were satisfactorily completed, although the Unit Load Re]ection Test had to be reperformed after the test method was changed (see Section 6.5). The Power Ascension Test results have verified the design models used for PVNGS and have confirmed that PVNGS 2 is constructed as designed and is similar to PVNGS 1.

Section 6.0 of this report describes the individual Power Ascension Tests that satisfy requirements described in the PVNGS FSAR. .

There were a total of nine reactor trips during Power Ascension Testing.

One of thes.a trips was a planned trip as part of Power Ascension Testing (Loss of Offsite Power, Section 6.7). One additional trip occurred inadvertently as a

. result of scheduled testing (Unit Load Re3ection Test, Section 6.6). The remaining trips were not directly related to any testing in progress.

There were four outages of between seven and eighteen days each, during the period from June 21 to August 15, 1986. Na3or work items accomplished during these outages include: condenser tube repair, control rod drive mechanism repair, Reactor Coolant Pump seal replacements, condensate system valve repair, and a repositioning of one excore neutron detector.

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PVNGS UNIT 2 STARTUP. TEST REPORT PAGE 9 2.0 INITIAL FUgL_LQADING (CESSAR Section 14.2.10.1)

IgST 08JECTl_Vg5 AND SUNNARY The governing procedure for the loading of the initial core into PVNGS 2 was 72IC-2RX01, " Initial Fuel Loading". The objective of this procedure was to provide a safe, organized plan for accomplishing the fuel loading. Fuel loading was conducted over the period of December 11 through December 16, 1985. During and after fuel load, several checks were performed to assure that the core loading was acceptable. These checks included verification of proper loading pattern, proper fuel assembly seating, and proper fuel assembly alignsent.

These checks were performed successfully and no problems were determined, with the exception that two fuel assemblies were out of tolerance on their 4

alignments. CE reviewed the data on these two assemblies and found them to be seceptable (see Test Results), and the procedure was successfully completed on December 19, 1985.

I((I DESCRIPTigN The initial core loading of PVNGS 2 was performed " dry"; that is, the refueling pool was dry except for the fuel transfer canal area, which was filled i with borated water to Just above the top of the fuel transfer tube. Before the start of fuel loading, this water was measured to have a boron concentration of 4121 ppa. The water level in the reactor vessel was maintained below the vessel flange, but above the top of the hot legs. This water was measured to have a boron concentration of approximately 2385 ppa at the beginning of fuel load.

One shutdown cooling loop was operated almost continuously during the fuel load evolution to ensure a uniform boron concentration throughout the Reactor Coolant System (RCS). Samples of the water were drawn from the reactor vessel and from the fuel transfer canal at least once each day to ensure that the boron concentration remained above the Technical Specification limit of 2150 ppa.

The core loading was initiated by the placement of the first of 241 fuel

. assemblies on the east side of the core area. This assembly contained a startup

! neutron source to provide a sufficient population of neutrons for suberitical multiplication monitoring. Succeeding assemblies were loaded in a sequence l which assured coupling of the assemblies with the source. In general, the fuel assemblies were loaded in north-south rows proceeding from the east to the west side of the core, as illustrated by Figure 2-1.

4 Monitoring of the suberitical status of the core was performed using four source range detectors: two temporary detectors, located in the reactor vessel; and the two permanently installed Startup Channel detectors, located outside the reactor vessel. Figure 2-2 shows the relative locations of the four detectors.

Each of the temporary detectors was moved once during fuel loading to maintain i proper monitoring of the core, and both were removed from the vessel prior to

. the loading of the final two fuel assemblies. After each fuel assembly was loaded, a series of neutron count rates were recorded from each of these detectors. This data was used to compute the inverse multiplication (1/M) for j the fuel assembly for each detector. Engineering personnel reviewed this information to ensure that the next fuel assembly could be loaded safely.

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PVNGS UNIT 2 STARTUp TEST REPORT PAGE 10 Figure 2-3 shows the inverse multiplication response of the temporary detectors, in their initial locations, for the first 25 assemblies loaded. After the first 7 assemblies, the core suberitical multiplication, as indicated by the 1/M response of the temporary detectors, stabilized and remained essentially the same for the remainder of the fuel loading.

During fuel sovement, personnel in the fuel building and in containment independently verified that each fuel assembly was transferred from its storage location to its core location in the prescribed sequence. After each essembly was lowered into the reactor vessel, the elevation of the fuel grapple was checked to ensure that the assembly was seated properly on the core support structure before the assembly was ungreppled.

Following the completion of fuel loading, the underwater television camera on the refueling machine was used to scan the serial numbers of the fuei assemblies to verify that each assembly was in its prescribed Cycle 1 location.

Furthermore, this scan verified that each assembly serial number was oriented to the plant north and ensured that both startup neutron sources were properly installed in the core. The performance of this scan was recorded on videotape.

A second scan was performed on all of the fuel assemblies using the underwater camera to ensure that the center of each assembly was aligned within an acceptable tolerance of the nominal centerline for that core location.

TEST RESULTS Fuel loading was completed on December 16, 1985. The fuel assemblies were verified to be properly loaded, seated, and oriented. Both startup neutron sources were verified to be properly loaded. Finally, scans of all of the assemblies verified that all but two of the assemblies (F-17 and L-16) were aligned within an acceptable tolerance of the nominal fuel centerlines. The difference between the fuel assembly centerline and the nominal centerline value for F-17 was 0.52 inches and for L-16 was 0.51 inches. C-E Windsor engineering reviewed the data and determined that since the differences of 0.52 and 0.51 '

j inches were very close to the Acceptance Criteria of 0.50 inches , that proper i

engagement of the upper end fitting with the Upper Guide structure would not be affected. 72IC-2RX01 was officially completed, with the results satisfactory, on December 19, 1985.

CONCLUSIONS The initial fuel loading of PVNGS Unit 2 was successfully accomplished in a safe and controlled manner, in accordance with the objectives and acceptance criteria of 72IC-2RX01.

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NOTES tee n~ umbers shown above in some core locations correspond to the total numoer of fuel assemblies in the reactor vessel after that location has been loaded. The first assembly loaded contained a neutron source and was located in position A-9. It was later relocated to position P-3, following the loading of Assimo1 7 193. Assembly 194 was then loaded into the " hole" left in position A-9.

j Assemblies 44, 195, 240, and 241 were used to fill the " holes" left after the i movement or removal of the temporary detoctors.

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 13 FIGURE 2-3 INVERSE NULTIPLICATION RESPONSE OF TEMPORARY DETECTORS PVNGS UNIT 2 INITIAL FUEL LOAD (First 25 Assemblies)

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O o- evacs u Ir 2 STARTUP TEST REPORT PAGE 14 3.0 POSTCORE__ HOT,, FUNCTIONAL TESTS (CESSAR Section 14.2.12.3) 3.1 Postcore Hot Functional Test Controlling Document (Section 14.2.12.3.1)

TEST _0BJECTIVES AND_

SUMMARY

The objectives of the Postcore Hot Functional Controlling Document, PVNGS procedure 72HF-2ZZO3, were:

(1) To demonstrate the proper integrated operation of the plant prieary, secondary, and auxiliary systems with fuel in the reactor vessel.

(2) To act as a sequencing / controlling document for the CESSAR required hot functional tests.

(3) To demonstrate that the plant can be brought from cold shutdown conditions (Mode 5) to hot standby conditions (Mode 3) using station operating procedures.

(4) To sequence / direct the initial performance of certain mode entry technical specifiestion surveillance procedures.

(5) To sequence / control the performance of Precore Hot Functional (Phase 1) carryover tests.

This procedure was performed over the period of March 9 through April 6, 1986.

During this time, the plant was brought from Operational Mode 5 to Operational Mode 3. Performance of this test successfully demonstrated the ir.negrated operation of the plant primary, secondary, and auxiliary systems during these Mode changes, thereby satisfying the test acceptance criterion.. Additionally, the individual testa controlled by 72HF-2ZZO3 were successfully performed.

IEST_ DESCRIPTION

. Testing commenced with Reactor Coolant System (RCS) at a temperature of approximately 200 CF and a pressure of 365 paia.. From this condition, the RCS was heated up and pressurized to 565 0F, 2250 psia using station operating procedures. During the heatup/ pressurization, conditions were stabilized at the direction of 72HF-22203 at five intermediate temperature / pressure plateaus to allow required testing to be performed. Table 3-1 lists the various i temperature / pressure plateaus at which testing was performed. The surveillance

requirements were verified as being satisfied prior to any changes in 1

operational mode. The plant was then maintained at 565 0F, 2250 psia in accordance with the prerequisites of procedure 72IC-2RX02, " Initial Criticality."

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O evacs u" r 2 STARTUP TEST REPORT PAGE 15 TEST RESULTS The plant was successfully heated and pressurized from cold shutdown to hot standby using normal plant operating procedures. The integrated operation of the primary, secondary and related auxiliary systems in accordance with CESSAR descriptions was verified by the heatup and pressurization. In order to replace several inadvertently damaged ultrasonic flow transceiver crystals, the RCS was cooled down to 215 CF on March 20 and returned to 340 CF on March 25, causing a 130 hour0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> delay in testing. The implementing test procedures sequenced by this procedure were properly performed and completed. One of the sequenced procedures had acceptance criteria that were not met. In procedure 73HF-2RC09, "RCS Flow Measurements", the 4 pump coastdown and RCS total flow did not meet the acceptance criteria. The final resolutions to these failures are detailed in section 3.3. The remaining individual test procedures acceptance criteria were properly met and documented in each procedure. Additionally, the carryover testing from the Preoperational Test Phase was satisfactorily completed.

CONCLUSIONS Proper integrated operation of the PVNGS 2 primary, secondary, and related auxiliary systems was successfully demonstrated during the Postcore Hot Functional Test. Therefore, these systems will functionally support power operation of the plant.

TABLE 3-1

~

i POSTCORE HOT FUNCTIONAL TEST PLATEAUS I I (Nominal Conditions) 1 I I j i RCS Temp RCS Press l l

l Date Time (CF) (psia) Mode i I I i 1 1 3/09/86 1305 195 365 5 I I 3/10/86 1930 280 380 4 1 I 3/11/86 0200 340 380 4 I I 3/26/86 0600 450 1100 3 1

! l 3/26/86 1130 450 1650 3 l l l 3/26/86 1640 500 2250 3 1 I 3/28/86 1338 565 2250 3 I I I l

l i

PVNGS UNIT 2

{

STARTUP TEST REPORT '

PAGE 16 3.2 Postcore Instrument Correlation (Section 14.2.12.3.2)

TE!T_ OBJECTIVE AND SUNNARY PVNGS procedure 73HF-22202, "Postcore Instrument Correlation," was performed over the period of March 10 to April 9, 1986 in Operational Modes 3 and 4. The objective of this test was to verify that the Main Control Room indications of selected plant parameters monitored by the Plant Monitoring System (PMS), Qualified Safety Parameter Display System (QSPDS), Plant Protection System (PPS), Core Protection Calculators (CPC) and Process Instru-ments were correct and consistent within acceptance criteria that were based on vendor and design accuracies. This ob J ective has been satisfactorily met.

TEST _ DESCRIPTION Data for this test was gathered at the following nominal test plateaus:

280 CF/380 psia 450 0F/1650 paia 340 CF/380 psia 500 CF/2250 paia 450 CF/1100 psia 565 CF/2250 paia Specified plant parameters that were displayed by more than one device were observed and the values recorded as simultaneously as possible. These j parameters included reactor coolant system (RCS) hot leg temperatures, RCS cold leg temperatures, core exit temperatures, pressurizer pressure, pressurizer level, steam generator pressures, steam generator levels, reactor coolant pump (RCP) differential pressures, steam generator differential pressures, RCP speeds, and auxiliary feedwater flows. The values recorded for each parameter were then cross-compared to verify that the various indications of that particular parameter were consistent and accurate within the specified acceptable agreement bands.

t TEST RESULTS l

The data recorded in this test wac sufficient and within established criteria. The Test Exception Reports (TERs) generated were succesafully ratested, with the exceptions of TERs 05 (RCB-TI-112CB) and 06 (AFA-FI-40B and AFA-FI-41B). Correlation of these instruments was performed in conjunction with test procedure 72PA-2SB01, "In*,ercomparison of PPS, CPC, and PMS Inputs (20x)",

(TER 05) and surveillance prc<edure 42ST-22210, " Post Accident Instrumentation Channel Checks", (TER 06) T.a TERs were generated primarily due to failed equipment rather tha, c t-or .olerance data.

At the conclusioe of w/ test, the selected parameters met the respective acceptance criteria with no outstanding Test Exceptions. Sufficient correlation was established to ensure that the indications observed were correct and consistent within the prescribed criteria.

CONCLUSION The accuracy and consistency of Control Room indications of selected plant parameters monitored by the PMC, QSPDS, PPS, CPCs, and process instruments were adequate to support plant power operation.

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 17 3.3 Post Core Reactor Coolant Systes Flow Measurement l (Section 14.2.12.3.3)  ;

TEST OBJECTIVE AND

SUMMARY

PVNGS procedure 73HF-2RC09, " Post-Core Reactor Coolant System Flow Measurements", was conducted over the period of March 27 through April 6, 1986 with the reactor at hot standby conditions (565 0F,2250 paia). The principle objectives of the test were as follows:

(1) To determine the postcore reactor coolant system (RCS) steady state flow rate and flow coastdown characteristics.

(2) To ed)ust Core Protection Calculator (CPC) and Core Operating Limita Supervisory Systes (COLSS) flow algorithm constants based on the measured steady state flow rate.

(3) To compare the measured loss of flow coastdown curve (4-pump trip) to that used in the CESSAR Final Safety Analysis.

The measured RCS flow rate (4 pump steady state operation) was 464,830 gpm and was within the acceptance criteria.

TEST _ DESCRIPTION For ten various steady state reactor coolant pump (RCP) configurations and one RCP flow coastdown, measurements were made of the RCP differential pressures (DPs), RCP speeds, RCS temperature, and RCS pressure. The measured RCP DP values were converted to values of head, and the corresponding pump flow rates were determined from the RCP performance curves relating pump head to flow rate. The total RCS flow rate for each pump configuration and coastdown was determined by summing the four individual RCP flow rates.

In addition, the total 4 pump steady state RCS flow rate was determined using the more accurate ultrasonic flow measurement (UFM) technique. Lithium i niobate crystals were mounted on each RCS hot leg to serve as ultrasonic signal l transmitters and receivers. The received signals were electronically processed

and used to analyze the fluid turbulence patterns as they possed successive crystal pairs. The mean transit time of the fluid between crystal pairs was determined using cross-correlation techniques, and the fluid flow rate was calculated as a function of the mean transit time, crystal spacings, and flow area. The RCS flow rate measured by UFM techniques was then used as the reference, or standard, flow rate for ad usting J the CPC and COLSS flow algorithm constants.

TEST RESULTS The four pump volumetric RCS flow rate determined by UFM techniques wes 464,830 spa, or 104.3% of the design flow rate of 445,600 gpm. This measured flow rate was not within the acceptance criteria of greater than or equal to 465,850 gpm and less than or equal to 501,800 gpm. However, full power flowrote measurements on PVNGS Unit i have revealed that the actual full power flowrate decrement is smaller than predicted. Taking partial credit for this reduced

~. - .

O evacs ==1T 2 STARTUP TEST REPORT PAGE 18

'flowrote decrement from zero power to full power, the NSSS vendor slightly relaxed the minimum flowrate acceptance criterion to 463,975 gpa. Therefore, the final UFN measured flowrote of 464,830 gpa is acceptable. Because the four pump volumetric RCS flow rate calculated from RCP differential pressure data did I not acceptably agree with the ultrasonic flow measurement result, revised RCP performance curves were provided by the NSSS vendor.

The four pump coastdown flow fractions were not conservative with respect i to the safety analysis assumptions. The faster coastdown has been postulated to have been attributable to an electrical braking effect and reduced pump j efficiencies. Interim resolution to this problem resulted in the NSSS. vendor

imposing a penalty factor of 4 percent on the CPC and COLSS to assure that the transients initiated (from 55 percent) up to 100 percent of rated thermal power remain within the safety analysis limits under the condition of faster than anticipated Reactor Coolant Pump coastdown. Specifically, the penalty was imposed to assure that the Loss of Flow (LOF) transient would be bounded by the existing safety analysis.

Credit was taken for more recent RCP coastdown data recorded at hot zero power which indicated essentially the same coastdown rate as that of pVNGS Unit i pumps without the electrical braking effect. This information superceded the earlier assessments of lower pump efficiencies for PVNGS Unit 2. Data from PVNGS Unit 1 indicated that the CPCs can provide a trip faster than the assumed

reactor trip during the LOF events (300 milliseconds versus 600 milliseconds).

Further analysis verified that this faster CPC generated trip can adequately accommodate the faster coastdown observed for PVNGS Unit 2's RCPs with the electrical braking effect. Therefore, the entire 4 percent COLSS/CPC flow coastdown penalty was ultimately removed.

993Gkg!1ggs I

The 4-pump steady state RCS flow rate is sufficient to provide proper cooling of the reactor core under power operation conditions. Additionally, the i measured 4 pump trip flow coastdown curve was retested and the ratest data was evaluated to be adequate to assure that the limiting event, the Loss of Flow transient, would be bounded by the safety analysis.

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l PVNGS UNIT 2 STARTUP TEST REPORT PAGE 19 l

3.4 Postcore Cc,ntrol Element Drive Mechanism Performance (Section 14.2.12.3.4) ,

i IggT_Q32[gTIVg3_AND

SUMMARY

(1) To demonstrate the proper operation of the control rod drive system (Control Element Drive Mechanisms, or CEDMs) including the control rods (Control Element Assemblies, or CEAs) under Hot Zero Power conditions. This objective was met by 73HF-2SF11, "CEDM Coil Testing 565 0F", conducted from March 29 through April 8, 1986 with the plant in Mode 3. The acceptance criteria were met by the successful movement of the CEAs and their respective CEDM coil current traces being normal.

(2) To verify the proper operation of the CEA position indicating system and alarms. This objective was met in Test 73HF-2SF02, " Post-Core CEDM l Performance", which was conducted from March 12 through March 15, 1986. The plant was in Mode 4 during the performance of this test. The acceptance criteria were met by verifying that the indicating systems provided the correct CEA position and that the alarms functioned per design.

(3) To measure CEA drop times. This ob]ective was met in Test 73HF-2SF08,

" Post-Core CEA Drop Time Test". The test was conducted from April 2 through I

April 5, 1986 with the RCS at 565 0F cnd all four RCPs running (Mode 3). The 4

acceptance criteria for CEA drop time (Technical Specification 3.1.3.4) was met by verifying that the CEAs dropped to 90k insertion in less than 4.0 seconds.

IIII_Dif9BIEII9N 23HE-2gE11: Each CEA was withdrawn individually to 120 inches (802 withdrawn) and then inserted to 7 inches. The CEA was then dropped by opening the individual CEA breaker. Current traces were taken while the CEA was being withdrawn and inserted. The CEA was verified to drop when the power was

! removed. Two minor problems were noted during the performance of the test; high noise level of the SCRs when in the OFF state, and intermittent negative grounds on two CEA coils. The CEAs perform adequately with the observed noise and are acceptable by the NSSS vendor. The grounded CEDMs can be operated in a normal manner and will be reworked at the next available outage when the reactor vessel head assembly is unstacked.

23HE-2EE92: Each CEA was withdrawn individually to its upper limit, then inserted to its lower limit, and then dropped by opening the individual CEA breaker. During this evolution, the following were verified:

1) The upper and lower electrical limits were set correctly

2) The upper, lower, and rod drop lamps were operational.

3) The Plant Computer minor and ma3or deviation alarms were set correctly.

4) The CEA Calculator (CEAC) deviation alarm was set correctly.

5) The CEA position indicated correctly on the CPC, CEAC, PMS, and CEA position CRT.

6) The CEA withdrawal and insertion drive speeds were correct (30 in/ min).

73HE-ggEgg: The CEAs were withdrawn by group to their upper limit and each CEA in that group was verified to be at its upper electrical limit. The CEAs were then dropped, one at a time, by opening the individual CEA breakers. The CEA position was recorded as it dropped, by monitoring the CEDM power and the Reed Switch Position Transmitter (RSPT) output. The recorded data for each CEA was reviewed, and the drop time for 90% insertion was calculated.

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O O evacs ""1r STARTUP TEST REPORT PAGE 20 TEST RESULTS The testing was completed for the CEDMs, with no outstanding test exceptions. The alarms, position lamps, and CEA position indicators were verified to respond within their assigned limits. The CEAs moved es required, and their withdrawal and insertion CEDM current traces were satisfactory. The drop time of each CEA to 904 insertion was less than 3 seconds, well within the allowed limit of 4.0 seconds.

CONCLUSION The testing of the CEDMCS proved that the system will operate as d'esigned, and will support plant power operation.

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O O avacs ==1r 2 STARTUP TEST REPORT PAGE 21 3.5 Postcore Reactor and Secondary Water Chemistry Data (Section 14.2.12.3.5)

TEST OBJECTIVE _AND

SUMMARY

pVNGS procedure 74HF-2SS01, "postcore HFT Chemistry Test," was performed (1) to demonstrate that proper water chemistry for the Reactor Coolant System (RCS) and Steam Generators can be maintained from ambient conditions to system operating conditions; and (2) to verify the adequacy of the prescribed sampling frequencies in establishing and maintaining proper chemistry control as well as detecting, and correcting, out-of-specification conditions in a timely manner.

Acceptability of the test was based on three criteria:

(1) The procedures for sample collection and analysis were adequate for RCS and Steam Generator chemistry control.

(2) The prescribed sampling frequencies were adequate for RCS and Steam Generator chemistry control.

(3) The analyses of water samples from the RCS and Steam Generators were capable of detecting deviations from the prescribed chemistry specifications in a timely manner.

Monitoring of the chemistry conditions per this procedure was initiated on March 9, 1986 and completed on April 5, 1986. The acceptance criteria were satisfied.

TEST DESCRIPTION Sampling and chemical analyses of the RCS and Steam Generators were performed using the appropriate plant operating procedures, as directed by 74HF-2SS01. Data was taken at the following test conditions:

Ambient conditions (<200 CF) 340 CF/500 psia 250 CF/380 psia 450 CF/1100 psia 280 CF/380 paia 565 CF/2250 paia At each of these plateaus, samples from the Reactor Coolant System and Steam Generators were taken, analyzed, and compared to the operating specifications provided in the test procedure.

TEST _RESULTS Ambient conditions (<200_OF) - The steam generators were in " Wet Lay-Up" with l _

Wet Lay-Up chemistry specifications being maintained. The RCS was in Mode 5 with RCS chemistry specifications being maintained.

250_OF/365_gsig,gigtggu - Both the RCS and steam generators met the Mode 4 specifications at this test plateau.

2g0C[f2gg_ psi 9 _21gtggu - The RCS and steam generators met the specifications for Mode 4 throughout this plateau.

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O O evMcS ==1T 2 STARTUP TEST REPORT PAGE 22 340_OF/380 gala _glatggu - This plateau lasted from March 11 until Merch 25, 1986. During this time frame the RCS hydrazine went out-of-specification low (less than 1-1/2 times the dissolved oxygen concentration) but was weived in ,

accordance with 74AC-92204, " Systems Chemistry Specifications", because the {

dissolved oxygen concentration was less than detectable at <0.005 ppa. On March i 19, 1986, the dissolved oxygen increased to out-of-specification high due to '

LPSI "A" loop being placed inservice. The condition was corrected by adding hydrazine. The RCS then remained in specification for the remainder of the test plateau.

During this plateau the steam generators had two (2) problems with -

out-of-specification conditions. Both were due to required chemical additions to the Auxiliary Feedwater system. The root cause of both problems was corrected by treating the Condensate Storage Tank to feedwater specifications such that chemical additions to the auxiliary feedwater system would not be required. The procedure for sampling and analysis proved to be acceptable to detect the changes in a timely manner so that corrective actions'could be taken i

in accordance with the procedures.

450 0F/1100 psig_glateau - The RCS chemistry was within specification for this test plateau. The steam generators had a low pH condition, which was corrected within six (6) hours, and therefore the procedures for controlling system chemistry were Judged to be acceptable. i j 500 0F/2250 psia glategu - The RCS chemistry was within specification for this test plateau. One of the steam generators bad a slightly high

- out-of-specification dissolved oxygen content. This condition was corrected in approximately twelve (12) h'ours; therefore, the procedures were Judged to be acceptable for detecting and correcting out-of-specification conditions in a timely manner.

565 0F/2250 psia plateau - The RCS and steam generators remained in specification during this plateau. The steam generators were noted te trending to an out-of-specification condition, but corrective actior e taken

to prevent an out-of-specification' condition.

99H96HSIGHa The test objectives for 74HF-2SS01 were satisfied in that overall proper

chemistry for the RCS and steam generators was maintained at system operating conditions. The prescibed sampling frequency was adequate to ensure proper i

chemistry control and out-of-specification conditions were detected in a timely manner.

P

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O O PVNGS UNIT 2 STARTUP TEST REPORT PAGE 23~

3.6 Postcore Pressurizer Spray Valve and Control Ad ustments 3 (Section 14.2.12.3.6)

IMI_9HEGIIYE3_M R_H MABI '

' Test 73HF-2RC10, " Pressurizer Spray Valve and Control Ad3 ustment", was performed to establish the proper settings for the continuous (bypass) pressurizer spray valves, to measure the rate at which pressure is reduced by maximum pressurizer spray, and to measure the maximum pressurization rate. The following acceptance criteria applied to the test: (1) the continuous spray flow was to be ad3 usted such that spray line temperature at the pressurizer inlet would be no more than 70 0F lower than the cold leg temperature; and,

(2) operation of both pressurizer spray valves together would reduce the pressurizer pressure at a rate equal to or greater than 1.06 psia per second from a nominal pressure of 2250 psia.

This test was performed on April 1, 1986, with the RCS at approximately 5650F, 2250 psia, and full flow conditions. The acceptance criteria were satisfied.

IEfI_DifQRIPTIQH Permanent plant temperature instrumentation on the cold legs was used to determine the average cold les temperature. Four strap-on thermocouples were located at various locations on the spray line check valve near the pressurizer.

An accurate measurement of the spray line fluid temperature was recorded during the Pre-Core Hot Functional Test utilizing an auxiliary thernowell in the bonnet of the spray line check valve. Comparing this data to the data recorded from

} the spring loaded thermocouples on the spray line check valve, the NSSS vendor supplied a temperature bias term to.be applied to one of the thermocouples in order to develop the proper temperature of the spray line fluid.

• Initial data was gathered to determine the temperature difference between the cold legs and the pressurizer spray line with both main spray valves closed and both continuous spray valves in the as found throttled position. The temperature difference was determined to be within the acceptance criteria.

Therefore no ad3 ustments to the continuous spray valves were made. Steady state conditions were re-established and data was taken to re-verify the acceptance criteria.

l The effectiveness of the pressurizer spray was measured by opening the main

spray volves, securing all pressurizer heaters, and recording pressurizer

pressure at a function of time to determine the depressurization rate. These

! measurements were performed with both main spray valves open, as well as with each valve opened individually. Following each depressurization, the main spray valves were closed, the pressurizer heaters were energized, and pressurizer

pressure was recorded as a function of time to determine the pressurization

! rate, for information only.

l i

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(} {} PVNGS UNIT 2 STARTUP TEST REPORT PAGE 24 l TEST,RESULTS The continuous spray valves were determined to be properly set to obtain a temperature difference between the cold legs and the spray line near the pressurizer of 70 0F. The depressurization rate obtained using both main spray valves was measured to be 1.50 paia per second. Therefore, the acceptance criteria for this test were satisfactorily met. Furthermore, the pressurization rate was measured to be 0.34 psia /sec using all pressurizer heaters, and 0.20 psia /see using all except the class powered pressurizer heaters.

The discrepancy between the measured depressurization rates on PVNGS Unit 1 (6.47 psia /sec) and PVNGS Unit 2 (1.50 psia /sec) has been attributed to the excessive leakage by the main spray valves in Unit 2. This is supported by the fact that Unit 2 generally operates with more pressurizer heaters energized than Unit 1.

CONCLUSIONS The testing required by CESSAR was completed and the acceptance er,iteria were satisfied.

O

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() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 25 3.7 Postcore Reactor Coolant System Leak Rate Measurement (Section 14.2.12.3.7)

TEST OBJECTIVE AND SUNNARY The purpose of this test was to measure the reactor coolant system (RCS) leakage at hot zero power conditions. Testing was performed using the PVNGS surveillance test procedure 42ST-2RCO2, "RCS Water Inventory Balance". In this test, " identified" and " unidentified" RCS leakage must be within the limits specified by Technical Specification 3.4.5.2; namely, that identified leakage shall not exceed 10 gpa and that unidentified leakage shall not exceed 1 gpa.

Testing was performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during Postcore Hot Functional Testing while the plant was in Mode 3. The test results were within allowable limits.

TEST _ DESCRIPTION -

The test is performed by measuring the changes in water inventory of the RCS and Chemical Volume Control System (CVCS) over a two hour interval. Changes in the levels of the Pressurizer, Volume Control Tank, Reactor Drain Tank, Equipment Drain Tank, and Safety Injection Tanks, as well as changes in the RCS temperature and pressure, are recorded and correlated to volume to determine the leakage rates in gallons per minute (gpa).

TEST RESULTS This surveillance test was performed six times during the course of the Posteore Hot Functional Testing, including three performances with the RCS at hot zero power conditions. Acceptable test results were obtained by each performance. A typical set of test results is illustrated by the April 5. 1986 performance, which measured identified and unidentified leakages of .063 gpm and .226 gpm, respectively. This particular measurement was conducted with'the RCS at 565 CF and 2243 psia.

CONCLUSION l The RCS leakage determined during postcore testing was well within the i limits specified by the Technical Specifications.

6

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

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' () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 26 3.8 Postcore Incore Instrumentation Test (Section 14.2.12.3.8)

T!!T_0BJECTIVES AND__

SUMMARY

PVNGS procedure 73HF-2RIO1, " Post-Core Movable Incore Instrumentation Test", was performed over the period of January 27 to May 20, 1986. The objectives and acceptance criteria of this test were as follows:

(1) To determine the Movable Incore Detector System (MICDS) path lengths using the detector drive systes encoder. Repetitive measurements of each path were to agree with the average length for the given path within 0.3 inch.

(2) To install the permanent plant movable incore detectors (fission chambers).

(3) To demonstrate proper computer control of the movable incore detectors.

(4) To verify proper operation of the MICDS at Hot Shutdown and Ho,t Standby conditions by accessing specified core locations in the manual control mode. This would also check changes in the physical fit of the detector in the path and the path growth due to temperature changes.

(5) To verify that leakage resistances for the Fixed Incore Detectors are at least 10 magohms at Hot Standby conditions.

TEST DESCRIPTION With the RCS at ambient conditions, the MICDS path lengths were measured by driving the dummy cable into each path using the Manual Control Box and recording the encoder readout from the Control Box. Three successive path length measurements were taken, then compared to the average length for each path.

Next, the permanent detectors (fission chambers) were installed with their associated cabling. The Manual Control Box was used to operate the system to measure the transfer times and drive rates for use as input to the computer -

control program. The computer control program was then tested to demonstrate its operability.

l j Following the heatup of the RCS to Hot Shutdown conditions, three of the path lengths were remeasured to obtain baseline data on tube growth due to temperature. These remeasurements were performed again at Hot Standby l conditions. Additionally, while at Hot Standby conditions (565 0F, 2250 paia nominal), the leakage resistance of each fixed incore detector was measured using a High Resistance Meter, to check for any abnormalities. The automatic test functions of the Fixed Incore Amplifier Bins (zero output; full scale output; insulation resistance) were also initiated and verified.

i IISI_BgsULIs The encoder path length measurements were performed using the dummy i detector and recorded as required. The repetitive measurements were within the 1 0.3 inch tolerance.

1

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 27 The computer operation of the MICDS was not successful. Prior to the test, the systes hardware had been modified per lessons learned in PVNGS Unit 1.

Extensive troubleshooting in Unit 2 has proved that the hardware modifications and Honeywell Plant Computer upgrade (45000 Model) were both needed. However, two more major problems, thus far, have been found in the computer hardware (I/O cards, interrupt cards). Removal of the capacitors to open-circuit the RC filters across the I/O cards appears to result in a much more reliable operation. The addition of a filter to the interrupt cards is believed to eliminate the extra pulses received by the cards, probably due to the use of high speed TTL logic encoder and the unfiltered interrupt inputs.

Troubleshooting on the system is on-going. Until the computer operation mode is s

successfully implemented, the MICDS may continue to be operated satisfactorily in the manual modo using the Manual Control Box.

The Fixed Incore Detector leakage resistances were well above the minimum acceptable value of 10 megaohns, indicating that the detectors and cabling are free from electrical grounds. Additionally, the automatic test functions of the Fixed Incore Amplifier Bins were successfully tested and demonstrated to operate per design.

99EfLHH19HE Although several problems were encountered during the performance of this test, the primary objectives and acceptance criteria as specified by CESSAR Chapter 14 were satisfied. That is:

(1) The leakage resistances of the fixed incore detectors were measured to be within design specifications.

(2) The ability of the MICDS to access the various paths was demonstrated using the Manual Control Box. Although the computer control mode was not

operable, the system may be operated satisfactorily manually until a fix of the computer mode is implemented.

(3) The path lengths of all movable incore paths were r easured using the manual control box and the dummy detectors. .

Thus, the fixed and movable incore detector systems were determined to be functional to the extent required to support plant power operation. It should be noted that the MICDS is not safety-related and that the Technical 7

Specification on incore detectors (Tochnical Specification 3.3.3.2) is not impacted by the operability status of the MICDS. In this Technical Specification, the MICDS is considered only as a backup to the fixed incore

detectors. The Technical Specification can be satisfied solely through the fixed incore detectors, even if the MICDS is declared inoperable, and plent operation in any of its operational modes will not be impacted. The manual control boxes can be used in lieu of the computer but with knowledge that

accuracy and speed will be degraded.

Although work is continuing to resolve the MICDS computer program problems, the test resulta met the required CESSAR acceptance criteria.

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PVNGS UNIT 2

~

STARTUP TEST REPORT PAGE 28 4.0 IglHAL CR_HighlII (CESSAR Section 14.2.10.2)

IIII_9Higgg AND SUNNARY Initial criticality for PVNGS 2 was achieved under test procedure 72IC-2RXO2, " Initial Criticality." The purpose of the procedure was to provide a safe, organized method for attaining the initial criticality of the PVNGS Unit 2 reactor and to verify that at least a one decade overlap existed between the startup excore detector channels and the log range of the safety encore detector channels.

The approach to criticality began at 0620 on April 18, 1986 and initial criticality was achieved at 1448 on April 18, 1986. The RCS boron concentration at criticality was measured at 1004.5 pps. An overlap greater than one decade between each startup channel and each log range of the safety channels was observed.

t TEST _ DESCRIPTION The approach to criticality began with the reactor coolant systes at

5650F, 2250 psia, approximately 1311 ppa boron and All Rods Out (ARO) with the exception of a single group of four rods (Regulating Group 5), which was 4 positioned at approximately 75 inches withdrawn (at midcore). The RCS boron

concentration was then reduced to achieve criticality, with Group 5 used to control the chain reaction.

Core response during the control rod group withdrawal and RCS dilution was monitored 1n the control room by observing the change in neutron count rate as indicated by the permanent source range nuclear instrumentation (startup channels). Neutron count rate was plotted as a function of control rod group i position and RCS boron concentration during the approach to criticality.

t Primary safety reliance was based on inverse count rate ratio (ICRR or 1/N) monitoring as an indication of the nearness and rate of approach to criticality.

i IEST RESULTS Initial criticality of the Unit 2 reactor was achieved in a safe and

, controlled manner as described above. The measured RCS boron concentration at

criticality, 2004.5 ppa, fell within the acceptance criteria of %5 ppa to 1065 ppa and differed from the predicted value of 1015 ppa by only 10.5 ppa. An

[ overlap greater than one decade was verified between each startup channel and the log range of the safety channels.

! $95S!:EE19dE f Satisfactory completion of this test demonstrated the validity of the core j physics predictions for initial criticality, as well as the adequacy and redundacy of plant instrumentation in monitoring the reactor in the source and low power ranges.

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 29 5.0 LOW POWER PHYSICS TESTS (CESSAR Section 14.2.12.4)

With the exception of the Low Power Biological Shield Survey Test (Section 14.2.12.4.1), all Low Power Physics Tests were performed as part of PVNGS procedure 72PY-2RX30, " Low Power Physics Test".

5.1 Low Power Biological Shield Survey Test (Section 14.2.12.4.1)

TEST OBJECTIVE AND SUNNNARY PVNGS procedure 75PA-2Z201, " Biological Shield Survey", was performed during the phases of Low Power Physics Testing (LPPT)-and Power Ascension Testing (PAT) to meet the following objectives:

(1) To measure radiation in accessible locations outside the biological shield.

(2) To obtain baseline radiation levels for comparison with future measurements of level build-up with plant operation.

1 Acceptance criteria for these measurements are based on predicted radiation levels for 100* power operation and are presented as maximum dose rates for five different access zones. Table 5-1 shows the applicable criteria and defines the access zones.

Baseline background data for this test was gathered on January 28, 1986 with the plant in Mode 5. Low power physica data was gathered on April 19, 1986 with the reactor critical and at Ox full power (FP) and the primary conditions of 565 0F, 2250 psia. The low power data met the acceptance criteria. ,

IESI.DESCHIEllON With the plant stabilized at the desired conditions, gamma and neutron radiation surveys were performed at selected locations in accessible areas outside the biological shield. These surveys were performed per the plant radiation survey procedure, and included general area surveys in rooms or areas as well as more detailed surveys around penetrations, shield plugs, and other areas where streaming could be occurring. A scan survey was also performed while the survey team was in transit between designated survey points. Surveys were performed in the Containment Building, Auxiliary Building, Main Steam Support Structure, Turbine Building, Fuel Building, Control Building, Decontamination and Laundry Facility, and at various site locations exterior to the plant.

1 O O evacs vazr 2 STARTUP TEST REPORT l

PAGE 30

)

IEST RE3ULTS i

Baseline data was gathered on January 28, 1986 in the accessible areas.

Comparison of this data to the acceptance criteria is not applicable. Low power physica data was gathered on April 19, 1986. The data gathered in the accessible plant areas during the low power physics surveys showed no increases above the baseline measurements.

CONCLUSION Reviews of the low power test results revealed no apparent deficiencies in the plant shielding. Sufficient baseline data was gathered for comparison with future measurements at higher power levels.

Table 5-1 RADIATIOh ZONE CLASSIFICATION 1 Zone i Dose Rate 1 Allowed Occupancy 1 I Designation I (aren/h) 1 (C.aign) l l 1 1 I 1 i I Less than 0.5 1 Uncontrolled, unlimited i 1 1 I access (plant personnel) i I I I I l l 2 1 0.5 to 2.5 1 Controlled, limited access,l

.l l l (40 h/wk to unlimited) i

~

I I I I l 3 1 2.5 to 15 1. Controlled, limited access I I i I (6 to 40 h/wk) I l I l i I 4 1 15 to 100 1 Controlled, limited access i I I I (1 to 6 h/wk) i 1 1 I I I 5 i Over 100 1 Normally inaccessible; i 1 l l access only as permitted byl l l I radiation protection i I I I personnel (1 h/wk) 1 I I ._

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O O evacs ==1r 2 STARTUP TEST REPORT PAGE 31 5.2 CEA Synsetry Test (Section 14.2.12.4.2)

TEST OBJECTIVE AND SUNNARY The ob3 ective of this test was to demonstrate that no core loading or control rod / fuel fabrication errors existed which would result in measurable control rod (Control Element Assembly, or CEA) worth asymmetries. This testing was performed from April 18 to April 20, 1986 with the reactor at hot zero power conditions (565 CF, 2250 paia). The data recorded during the symmetry test was analyzed to verify that the tsactivity worth of each CEA in a symmetric subgroup (of 4 CEAs total), relative to the avercge worth of a CEA in that same subgroup, was within 1.5 cents. The symmetric CEAs were determined to be within the specified tolerance.

IEST DE@GBIPTION .

In this test, the worth of each CEA was measured relative to the worth of the other CEAs within its symmetric subgroup. The technique used for the measurement involved the insertion of a " reference CEA" from each subgroup to its Lower Electrical Limit (LEL, or " full in" position) to establish a reference reactivity condition. The next specified CEA was then inserted to its LEL by trading its insertion with withdrawal of the reference CEA to its Upper Electrical Limit (UEL). The deviation of the resulting reactivity condition from the reference condition was recorded, and this CEA was then swapped to its UEL with insertion of the next CEA to its LEL. This process was repeated for all CEAs of the subgroup. The average deviation from the reference condition was then computed for the subgroup, and the deviation of each individual CEA was compared with this average. Differences which were no greater than i 1.5 cents were acceptable; however, differences greater than this limit may indicate either a misloading of fuel or fabrication errors in the fuel or CEAs.

! IEgI_BEgULIg Each CEA was checked within its symmetric subgroup as described above.

Data was recorded and analyzed per procedure and the symmetric CEAs were found to agree to within the acceptance criterion of + 1.5 cents of the symmetric CEA subgroup average, cggcLUgIggg Since the acceptance criterion for this test was satisfactorily met, it can be concluded that the fuel and CEAs were properly fabricated and the core was correctly loaded.

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O O "as "" 2 STARTUP TEST REPORT PAGE 32 5.3 Isothermal Temperature Coefficient Test (Section 14.2.12.4.3)

Ig[I_QH2[QIIV[ AND

SUMMARY

The Isothermal Temperature Coefficient (ITC) was measured twice during Low Power Physics Testing at different control rod (CEA) configurations. The conditions under which the ITCs were measured and the dates of performance are listed below:

1) 565 0F, 2250 paia, unrodded-----------April 20, 1986

2) 565 0F, 2250 pain, CEA Gps 5,4,3,2, and 1 inserted-------April 20, 1986 The measured ITC values were required to be within 1 0.3 x 10-4 delta-K/K/CF of their predicted values, a condition which was met during all the measurements. The moderator temperature coefficient (MTC) was determined from both of the ITCs measured, and verified to be in compliance with the Tech Spec limits (+0.22 x 10~4 to -2.80 x 10-4 delta-K/K/CF per Technical Specification 3.1.1.3) in each case.

IE!I_DESCRIPIION The ITC is defined as the the change in reactivity associated with a uniform change in the moderator and fuel temperature. To measure the ITC, the RCS temperature was changed approximately 5 0F at a rate of about 10-20 CF per hour, using the secondary Steam Bypass Control System. RCS temperature and core reactivity were recorded on a strip chart and, additionally, on an X-Y plotter. Following a short stabilization time at the new temperature, the RCS temperature was then returned to its initial value. Temperature and reactivity were again recorded during the transition.

The change in RCS temperature and the corresponding change in reactivity were obtained from the X-Y plots for both temperature swings and the ITC was obtained from the slope of the " beet-fit" line drawn through the data points.

The MTC was then determined by subtracting the predicted Fuel Temperature Coefficient from the measured ITC.

T[ST RESULT!

The measured ITCs agreed with the predicted values within the acceptable band. These results are summarized in Table 5-2. The MTCs derived from the ITCs were -0.27 x 10*4 and -1.15 x 10-4 at the unrndded (ARO) and rodded (Groups 5,4,3,2 and 1 inserted) conditions, respectively. Both of these values are within the aforementioned Technical Specification limits.

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 33 99HgkUgIggg The accuracy of the predicted isothermal temperature coefficients was verified by the measured values, all of which were within their acceptance criteria. Furthermore, the MTCa determined at hot, zero power conditions were in compliance with the Technical Specification limits.

TABLE 5-2 i ISOTHERMAL TEMPERATURE COEFFICIENTS I I (x10-4 delta-K/K/OF) i l 1.

I Diff. Acceptable i iConditions Predicted Measured (M-P) Dif(3__l l 1 1565 0F, 1 12250 psia i l I l--- unrodded -0.19 -0.43 -0.24 10.30 l I I l--- Grps i 1 5 to 1 i i inserted -1.32 -1.31 +0.01 10.30 l I__________________ ___ __________________________l e

I b

O O *a a= vart 2 STARTUP TEST REPORT PAGE 34 5.4 Shutdown and Regulating CEA Group Worth Test

-(Section 14.2.12.4.4)

IIII_9HECTIVE AND SUM 8431 The objectives of this test were as follows:

(1) To determine the individual group worths of the regulating control rod (Control Element Assembly, or CEA) groups.

(2) To determine the individual group worths of Shutdown Group B and of Shutdown Group A-1 (Shutdown Group A with the worst stuck rod out) if the total worth of the regulating groups failed to meet its acceptance criteria.

(3) To sum those measured group worths to demonstrate the adequacy of the shutdown margin.

Figure 5-1 shows the relative locations of the CEA groups in the PVNGS 2 core. The regulating CEA groups (5,4,3,2,1) were measured with the RCS at hot zero power (565 0F, 2250 psia) on April 20, 1986. The measured individual group worths were required to be within ! 10% of their predicted values, or 1 0.054 delta-K/K (whichever is larger) . The total worth of the regulating CEA groups was required to be within 3 10% of its predicted value.

IIII DEEEBIEIIQH A constant dilution of the boron concentration was initiated. Insertion of the desired CEA group was then performed in periodic, discrete steps, to offset the change in core reactivity from the boron dilution, and thus maintain power and reactivity within the desired control bands. Reactivity and power were recorded on a strip chart recorder. Insertion of the group continued until it reached its lower limit, et which time the dilution was secured or insertion of the next CEA group to be measured began. For the Group 1 worth measurement up to 102 inches withdrawn, a constant boration was initiated and the group was withdrawn from the core using the same general technique described above until it reached 102 inches withdrawn (see Test Results section below). Only one group was moved at a time, with no overlap between groups.

The data used to determine the CEA group worths was obtained from the reactivity strip charts. The reactivity change for each discrete group movement was determined and then summed over the length of the entire group to produce en j integral worth.

IIII_BEfHkIf The measured group worths are shown in Table 5-3. All of the CEA group worth measurements agreed with the predicted values within the allowed tolerances, including the total regulating group worth. During the worth j measurement of Group 1, CEA 1-59 slipped to the bottom of the core from j approximately 102 inches withdrawn. Power and reactivity were driven off scale i

(low) rendering further insertion data (below 102 inches) worthless. A Test Exception Report was generated to allow Group i to be withdrawn in the Manual

, Group mode (vice Manual Sequential), thereby gathering the necessary data which, when pieced together with the insertion date, would provide the integral worth.

PVNGS UllIT 2 STARTUP TEST REPORT PAGE 35 90!gbygig!!

The accuracy of the predicted CEA group worths was confirmed by the measured values. Furthermore, the total measured regulating group worth was in acceptable agreement with its prediction.

TABLE 5-3 I INDIVIDUAL CEA GROUP WORTHS I I (kdelta-K/X) i 1 l i Diff. Acceptable 1 l_Cgnditions Grou2 Pred. Meas. (M-P) Diff. I I I i 565 0F, I l_2259_Esis i I 5 -0.260 -0.276 -0.016 10.050 I I i l 4 -0.421 -0.452 -0.031 10.050 1 1 I I 3 -0.848 -0.792 +0.056 10085 I I I I 2 -0.994 -1.038 -0.044 10.099 I I I i 1 -1.276 -1.266 +0.010 10.128 I I I I Total 1 1 (5-1) -3.799 -3.824 -0.025 10.380 I l__----

________ .___ _____________________________I

'1

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O s ' V PVNGS UNIT 2 STARTUP TEST REPORT PAGE 36 FIGURE 5-1 RELATIVE CORE LOCATIONS OF CEA GROUPS i i . , . o S 3 S b w iu si i2 i3 ie it it if i.

, A 1 1 A it 10 2.  :: /J 14 at 26 a, 2s sw 40 ai 4 2 P 2 3 P, 2 4 3# 3J 34 JS 36 47 1. 39 40 di 42 43 ao 45 46 B B e B 3 B of 4e 49 to Si '

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96 98 3. 99 100 ici e0, '03 .04 ses iOb iO, 30. 30s iiO iii ii, 1 2 A 3 3 A B 1 ii4 ii. iin ii. ii, ... ii, iso i: 3 is ira is. its is. ist i:. its 3 3 5 Pi 5 3 3 i30 3. 'J i33 i34 i39 i3. 13, i34 ist ie0 14i ia2 esa saa is. is.

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.n - -. , e- - - . . --,- -

O avacs "" 2 STARTUP TEST REPORT PAGE 37 5.5 Differential Boron Worth Test (Section 14.2.12.4.5)

Ig!I_9_BJEgTIVE AND SUNN6BI The purpose of this test was to determine the differential boron worth.

The measurement was performed on April 21, 1986 with the RCS at 565 0F, 2250 psia. The measured differential boron worth was required to be within 1 10 pps/* delta-K/K of its predicted worth.

IEST DESgBIBIlgN The differential boron worth is defined as the change in boron concentration (in ppm) which would cause a it delta-K/K change in reactivity.

The data required to calculate the differential boron worth was obtained from the measurements of the CEA group wortha (see Section 5.4) and of the critical boron concentrations (see Section 5.6). The differential boron worth was determined simply by dividing the change in critical boron concentration in going from one CEA configuration to another, by the total reactivity worth of the CEA groups moved.

IgsI_BgsHLI!

The calculated differential boron worth agreed with the predicted value within 1 10 ppa /* delta-K/K. The comparison of the measured and predicted value is summarized in Table 5-4.

99H9kH!I9!I The differential boron worth met the acceptance criteria for the test..

i TABLE 5-4 1 _______________________________ ____________________________

l DIFFERENTIAL BORON WORTH I

I (pps/4 delta-k/k) I l________________________________________________________________l

! I Diff. Acceptable 1

IConditions Predicted Measured (M-p) Diff. I I i 1565 0F, -86.1 -81.09 +5.01 1 10.0 1 12250 paia i l________________________________________________________________l i

l s

O Pv GS ==1T 2 STARTUP TEST REPORT PAGE 38 i -

5.6 Critical Boron Concentration Test (Section 14.2.12.4.6) i I

IIII_9HIGIIYI_AND SHHABI i

The Critical Boron Concentration (CBC) was measured twice during Low Power Physics testing, at two different control rod (CEA) configurations. The conditions under which the measurements were performed and the dates of performance are listed below:

1) 565 0F, 2250 pais, unrodded------------April 20, 1986

. 2) 565 0F, 2250 psia,

CEA Gps 5,4,3,2 and 1 inserted---------April 20, 1986 The measured Critical Baron Concentrations were required to be within

50 ppa of their predicted values.

! IIII_DEf9BIPIlg!

a With the reactor critical at the desired CEA configuration and stabilized conditions, boron equilibrium was verified by observing that the reactivity drift was negligible. An RCS water sample was taken and analyzed for boron li content. The measured boron concentration was then corrected for the worth of l any CEA deviation from the position assumed for the prediction. This was done by inserting or withdrawing the deviating group to the assumed position and j measuring the reactivity associated with the move (i.e., the residual worth). ,

This residual worth was converted to ppa, using the differential boron worth, '

and used to appropriately adjust the measured concentration to provide the CBC for the same conditions assumed for the prediction.

j IE3I_BE8HLII , i t t l The measured Critical Boron Concentrations agreed with the predicted values within the acceptance criteria of : 50 ppa. Table 5-5 provides a summary of the measured and predicted values. ,

99E96H!I9H!

! l

The accuracies of the predicted Critical Boron Concentrations were l confirmed by the measured values, all of which were well within their acceptance

critieria.

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O av Gs ==1T 2 STARTUP TEST REPORT PAGE 39 TABLE 5-5 l CRITICAL BORON CONCENTRATIONS 1 (ppa) 1 I l_______ _ ______ __________ ____ ____ _ _ l 1 Diff. Acceptable i IConditions Predicted Measured (M-P) Diff. I I i 1565 0F, i 12250 psia 1 i 1 1--- unrodded 1025 1012 -13 150 l l 1 l--- Grps i I 5 to 1 1 I inserted 698 702 +4 150 I l________________ ______ _________________

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m O) U PVNGS UNIT 2 STARTUP TEST REPORT PAGE 40 6.0 POWER AECENSION Tg!Tg (CESSAR Section 14.2.12.5) 6.1 Variable Tavg (Isothermal Temperature Coefficient and Power Coefficient) Test (Section 14.2.12.5.2)

IEST__0BJEgIIVE_AND

SUMMARY

The objective of the Variable T oy test was to measure the Isothermal Temperature Coefficient (ITC) andthehowerCoefficient(PC) at the 50% and 100N power plateaus. Also, the Moderator Temperature Coefficient (MTC) was calculated from the ITC and was extrapolated to 100* to ensure compliance with Technical Specifications upon reaching the higher plateaus. Testing was accomplished using test procedures 72PA-2RX15 and 72PA-2RX50, " Variable Tay (Isothermal Temperature Coefficient and Power Coefficient) Test"(50%and$00x, respectively).

The measured ITC and MTC were required to be within 10 3x10-4 delta-K/K/0F and the measured PC within 10.2x10'4 delta-K/K/xPWR of their respective predicted values, a condition which was met for all tests.

Ig!I_DESCRIPTIgN This test involved measuring the reactivity changes which accompany changes in temperature and power. The Isothermal Temperature Coefficient (ITC) is a measure of the change in reactivity caused by a change in RCS average temperature, while the Power Coefficient (PC) is the change in reactivity (with RCS temperature constant) associated with a change in reactor power. The following measurement techniques were employed to determine the ITC and PC:

1) IIg_ Measurement _without_ggA_ Movement--The secondary steam loading was adjusted to establish a new core inlet temperature. The reactivity effects of the temperature change resulted in a new power being established. After a brief stabilization period for dato collection, the cycle was reversed by adjusting the secondary steam loading in the opposite direction to establish a new temperature and power. After another brief stabilization period for data collection, the secondary steam loading was readjusted and the core inlet temperature returned to its previous value, thus completing one temperature cycle. This cycling of temperature / power was repeated three times.

2) IIg_Hgasutggget_with_gg6_Hovegget--The secondary steam loading was adjusted to establish a new core inlet temperature. Core power was held essentially constant by compensating for the reactivity effects of the the temperature change with CEA group 5 movement. After a brief stabili stion period for data collection, the secondary steam loading was adjusted in the opposite direction and a new core inlet temperature established. Again, CEA group 5 movement was used to hold reactor power essentially constant.

After another brief stabilization period for data collection, the temperature cycle was completed by adjusting the secondary steam loading to return the core inlet temperature to its previous value, while CEA group 5 movement was used to hold reactor power essentially constant. This cycling of temperature with accompanying CEA movement was repeated three times. An optional fourth cycle was performed at 100%.

% rh (d V PVNGS UNIT 2 STARTUP TEST REPORT PAGE 41

3) Egwer_ggeffigient with CE6_Hovement--CEA group 5 movement was used to introduce a reactivity change which caused a subsequent power change. The average core coolant temperature (Tavg) was held essentially constant by ad]usting the secondary steam load to match reactor power. After a brief stabilization period for data collection, CEA group 5 was moved in the opposite direction (with secondary steam loading ad 3usted to keep the Tavg constant) and a new power was established. After data was collected, CEA group 5 was moved such that power was returned to its previous value, while the secondary steam loading was ad]usted to hold Tavg essentially constant, thereby completing the power cycle. This cycling of power and secondary steam loading was repeated three times. An optional fourth cycle was performed at 100% power After the data was collected, the ITC and PC were calculated using a reactivity balance which includes the reactivity effects of CEA group 5 movements, the change in average coolant temperature, and the the change in reactor power. The calculation was on iterative one which used the predicted ITC and PC as,the starting points and continued until successive iterations yielded agreement of

+0.005 x 10-4 for both the ITC and PC.

The MTC was calculated from the ITC by subtracting the predicted fuel temperature coefficient (FTC) as follows:

MTC = ITC - FTC At the 50% power plateau, the MTC was extrapolated to 100* power to ensure compliance with Technical Specifications (T.S.). The 100k power plateau test was performed at 962 power and the MTC was also extrapolated to 100k power to ensure T.S. compliance.

IESI_BESULIS The measured ITCs, MTCs and PCs agreed with the predicted values within the acco'ptable range at both power plateaus. These results are summarized in Table 6-1.

CQHCLg!!OHg The agreement of the predicted isothermal temperature coefficient, moderator temperature coefficient, and power coefficient with the measured values was verified for the two test plateaus. Furthermore, the MTC determined at each test plateau was verified to be in compliance with the Technical Specification limits.

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PVNGS UNIT 2

'STARTUP TEST REPORT PAGE 42 TABLE 6-1 l 1 I VARIABLE Tavg RESULTS 1 1

(x10-4 delta-K/KOF) ITC & NTC i I (x10-4 delta-K/K/kPWR) PC i i _ _ _ l 1 Power Boron (1) Diff. Accept.1 1.1Date) _____IEEm) Coef. Pred. Meas. (M-P) Diff. I I

I I 50% 709 ITC -0.601 -0.670 -0.069 10.3 1 1 (8/20/86) MTC -0.459 0

-0.528 -0.069 13 I I PC -1.150 -1.070 +0.080 +0.2 I I

i i 100* 626 ITC -0.814 -1.003 -0.189 10.3 I I (9/8/86- MTC -0.684 -0.872 -0.188 10.3 I I 9/9/86) PC -0.930 -0.787 +0.143 10.2 l I

I NOTES -- Diff. = Difference = Measured - Predicted

-- Accept. Diff. = Acceptable Difference Range (1) The measured values have been corrected to predicted boron concentrations of 790 at 50% and 699 at 100k

y PVNGS UNIT 2 VL STARTUP TEST REPORT PAGE 43 6.2 Unit Load Transient Test (Section 14.2.12.5.3)

IIII_982EGIIYg_g D SENNABY The Unit Load ~ Transient Tests were performed at the 50% and 100* power plateaus to demonstrate that both step and ramp load decreases and load increases of 12/ min and 1/24/ min can be performed at the design rates with key plant parameters remaining within design limits.

Testing was accomplished using procedure 73PA-22205 (50% test) on August 22, 1986 and procedure 73PA-22207 (1004 test) on September 10, 1986. At 504 power one parameter did not meet the acceptance criteria. At the 100m power, level all control systems responded as designed and key plant parameters remained within their required acceptance band.

IISI_pgSCBIEIIgg Unit load ramp decreases of cpproximately 52 per minute were performed by closing down slightly on the turbine control valves resulting in a mismatch between reactor power and turbine power. As the turbine load demand decreased, the Reactor Regulating System (RRS) detected a decrease in turbine first stage pressure and an associated decrease in the Turbine Load Index (TLI)-(i.e.,

Turbine Power). The RRS then calculated a power error term (based on the difference between the TLI and reactor power) and a reference temperature (T-ref; based on the TLI). The reference temperature was compared to the actual core average temperature (T-avg) to determine a temperature error.- When the summed error (temperature error plus power error) exceeded a specified value (ie, setpoint) the RRS instructed the CEDNCS to insert CEAs. CEA insertion continued until the summed error decreased below the setpoint.

The Steam Bypass Control System (SBCS), upon sensing the change in turbine load, opened steam bypass valves to relieve excess heat generation. The valve (s) remained open until insertion of the CEAs (by the RRS) reduced heat generation by the primary system. Once the mismatch between the primary and secondary' system power was eliminated, the valves resented. .

The decreasing power from the insertion of CEAs caused a reduction in core average temperature. Sensing the lower T-avg, the Pressurizer Level Control System (PLCS) lowered the pressurizer water level by increasing the letdown flow and decreasing the charging flow until it matched a programmed level based on the new reactor power.

The decrease in reactor power also caused a decrease in the mein steam flow. This decrease resulted in a steam flow / feed flow mismatch which is sensed by the Feedwater Control System (FWCS). To eliminate this mismatch, the FWCS decreased the feedwater pump speed and throttled back on the econceizer feedwater control valve until the steam flow and feed flow were approximately equal.

The 10% step decreases were also performed by closing down on the turbine control valves. In this case: however, the valves were closed at a faster rate than they were during the ramp decreases. Control system response to the 104 step changes were similar to those described for the ramp decreases except they occurred over a shorter period of time.

() () PVNGS UNIT 2 STARTUp TEST rep 0RT pAGE 44 TEST RESULTS During the performance of this test at the 50% power plateau, pressurizer level deviated from the minimum value defined by the acceptance criteria (see Tables 6-2 and 6-3). A Test Exception Report was issued to document this problem and to investigate potential solutions. Resolution of the TER required ratesting at the 100% power plateau and allowed power ascension based on the Reactor Regulating System being non-safety related and that the intent of the test was satisfied in that the reactor did not trip. The test at 1004 power successfully demonstrated that all control systems responded as designed and all monitored parameters remained within their acceptance band, hence no further testing was required.

Tables 6-2 and 6-3 summari=e the key plant parameters during a 5% per minute ramp from approximately 50% to 35% power and a 10% step change from approximately 35% to 25% power. As described earlier, pressurizer level did not remain within its acceptance band.

Tables 6-4 and 6-5 summarize the key plant parameters during a 5% per minute ramp from approximately 95% to 80% power and a.10% step change from approximatley 80% to 70% power. Evaluation of the data from this test showed that all parameters remained within their acceptance band and, therefore, that all control systems performed satisfactorily during design NSSS load changes CONCLUSION These testa successfully demonstrated that 10% step decreases and 5* per minute ramp decreases can be performed with the plant control systems maintaining key plant parameters within their design limits.

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PVNGS UNIT 2 l STARTUP TEST REPORT PAGE 45 TABLE 6-2 I i

. I

SUMMARY

OF TEST RESULTS (50%) I 1 5% PER MINUTE RAMP DECREASE TRANSIENT I I I I I Test Results 1 -Acceptance Criteria l i Parameter i Minimum Maximum i Minimum Maximum i I I I I l Pzr. Pressure 1 2217.5 2291.7 1 2150 2335 1 1 Pzr. Level I = 28.48 42.65 1 30.0 45.0 I I SG 1 Pressure i 1089.6 1156.7 1 1050 1180 1 I SG 2 Pressure i 1099.2 1167.8 1 1050 1180 I I SG 1 Level I 78.0 81.0 1 70.0 88.0 I I SG 2 Level I 75.0 81.0 1 70.0 88.0 1 I RCS T-avg I 569.6 580.5 1 567.0 587.0 1 I SG 1 T-hot 1 578.2 592.9 I 575.0 600.0 I I SG 2 T-hot i 579.3 594.0 1 575.0 600.0 1 I _ _

l _ _ _ _ _ _ . _

l ___ __ __ _ =1 TABLE-6-3

=

1 1 l-

SUMMARY

OF TEST RESULTS (50%) i I lok STEP DECREASE TRANSIENT I I I I I Test Results i Acceptance Criteria l l Parameter i Minimum Maximum i Minimum Maximum i I I I I i P:r Pressure 1 2224.8 2298.1 1 2150 2335 1 1 P:r Level i

• 28.8, 36.5 1 30.0 45.0 I I SG 1 Pressure i 1095.6 1150.7 1 1050 1180 1 I SG 2 Pressure i 1105.9 1161.8 l 1050 1180 1 I SG 1 Level I 73.0 80.0 1 70.0 88.0 1 I SG 2 Level I 74.0 80.0 1 70.0 88.0 I I RCS T-avg I 568.1 573.8 I 567.0 587.0 l I SG 1 T-hot i 575.2 581.8 1 575.0 600.0 1 I SG 2 T-hot 1 576.3 582.9 I 575.0 600.0 1 I__ _ _ _ _l -1 - _ _ _

l

• Out of tolerance results

, -- - - . . . - - - - _---s ,- .,

( PVNGS UNIT 2 STARTUP TEST REPORT PAGE 46 TABLE 6-4 I I I

SUMMARY

OF TEST RESULTS (100k) l I 52 PER MINUTE RAMP DECREASE I l__ _ _I

. I I Test Results l Acceptance Criterial 1 Parameter i Minimum Maximum i Minimum .. Maximum I I i i l l P r Pressure 1 2217 2286 1 2150 2335 I I Pzr Level I 45.6 54.5 1 40.0 60.0 I I SG 1 Pressure i 1064.4 1125.3 1 1020 1180 I .

I SG 2 Pressure i 1075.4 1136.2 1 1020 1180 1 I SG 1 Level I 77.0 81.5 1 70.0 88.0 1 I SG 2 Level I 76.6 81.0 1 70.0 88.0 I I RCS T-avg I 584.5 593.5 1 580.0 600.0 1 I SG 1 T-hot I 606.8 619.4 1 600.0 625.0 l I SG 2 T hot i 609.2 620.4 1 600.0 625.0 1 I I I I TABLE 6-5 1 I I

SUMMARY

OF TEST RESULTS (100%) I 1 10k STEP DECREASE I I I I I Test Results l Acceptance Criterial l_ _ Parameter i Minimum Msximum i Minimum Maximum I i l i I I P:r Pressure 1 2251 2292 1 2150 2335 I I P r Level I 40.9 48.5 1 40.0 60.0 l I SG 1 Pressure i 1070 1116 1 1020 1180 l I SG 2 Pressure i 1087 1127 l 1020 1180 l I SG 1 Level I 76.8 81.5 1 70.0 88.0 1 1 SG 2 Level 1 76.1 81.7 1 70.0 88.0 l I RCS T-avg i 582.0 586.0 1 580.0 600.0 1 i SG 1 T-hot i 601.2 609.4 I 600.0 625.0 I I SG 2 T-hot i 602.8 610.1 1 600.0 625.0 I l I l l__ _ _

i f

I l

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() ('~)' PVNGS UNIT 2 STARTUP TEST REPORT PAGE 47 6.3 Control Systems Checkout Test (Section 14.2.12.5.4)

Test Ob 2ective and_ Summary The Control Systems Checkout Tests were performed to demonstrate the satisfactory performance of the NSSS Control Systems during normal operations and under transient conditions. The control systems involved in these tests were:

Feedwater Control System (FWCS)

Steam Bypass Control System (SBCS)

Pressurizer Pressure Control System (PPCS)

Pressurizer Level Control System (PLCS)

Reactor Regulating System (RRS)

Reactor Power Cutback System (RPCS)

Control Element Drive Mechanism Control System (CEDMCS).

The specific ob]ectives of the tests were:

1) To demonstrate that the FWCS, SBCS, PPCS, PLCS, RRS and CEDMCS can control the plant within acceptable tolerances of their programmed values and to also demonstrate that the FWCS and RRS can control steam generator levels and Tavg, respectively, within acceptable limits following minor adjustments in control setpoints. The procedures which were used to test these control systems during normal operations and their dates of performance are listed below:

73PA-2SF01, " Control Systems Checkout at 20% Power" (May 23, 1986) 73PA-2SF04, " Control Systems Checkout at 50% Power".(August 21,1986) 73PA-2SF06, " Control Systems Checkout at 100% Power" (September 9, 1986)

IEEI.DESCBIPTIQN I2EA 2g[ git =SF04 &__SF06 Each of these Control System Checkout tests consisted of three parts:

1) A test of the Feedwater Control Systems (FWCS) ability to control steam generator level during steady state and minor transient conditions. The steady state pcrtion of the test afaply involved determining whether the FWCS maintained steam generator levels within 12% of the (steady state) level setpoint, while the transient portion involved both increasing and decreasing the level setpoint in ramp and step functions to determine whether the FWCS controlled steam generator levels to within +2x of the new level setpoint (after allowing for a brief stabilization period).

2) A test of the Reactor Regulating System's ability to control Tavg with respect to its control signal Tref. A mismatch was created between Tavg (Tavg#1, Tavg#2 and Tavg (average of Tavg#1 and Tavg#2) were all tested) and Tref by either varying Tavg with boration/ dilution or varying Tref with Main Turbine Power. The CEDMCS was then placed in the Auto Sequential mode to allow automatic RRS controlled rod movements. After an adequate time for temperature stabilization Tavg and Tref were required to be within 120F.

3) A test of the ability of all the control systems to function in an integrated manner. The control systems were all placed in the automatic mode while the plant was operating at steady state conditions. Data was taken over a thirty minute period to verify that the control systems maintained their respective parameters within the acceptable control band.

() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 48 Iest Results 23PA-2SF012 _2SF04 & 2SF06 The ability of the control systems to perform as designed during steady state operations and minor transients was verified by these tests. The ma3or test results are summarized in Table 6-6.

At the 20% and 1002 testing plateau, the Feedwater Control System, the Reactor Regulating System and the Integrated Checkout parameters fell within the acceptable limits as can be seen in Table 6-6.

At the 50x power plateau, the Feedwater Control System testing and Reactor Regulating System testing were all acceptable. However, during the Integrated Checkout portion, the aceptance criteria for the downcomer flow rates were not satisfied. Resolution to a Test Exception Report indicated that even though the downcomer flow rates were not as expected, the total flow rate was, and therefore system acceptability was declared.

CQNCLUSIONS These testa demonstrated that the control systems are capable of performing as designed under steady state, minor transient, or major transient conditions.

D

() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 49 TABLE 6-6 I I I DEVIATION FROM CONTROL SETPOINTS FOR I l CONTROL SYSTEM CHECKOUT TESTS I I Ace 99tgnee Criteria 73PA-2SF01 73PA-2SF04 73PA-1SF06 I I I Feed Water Control: 1 I I I Steady state S/G ivls. I i 12% from setpoint: 0.64 NA NA I I I I Transient S/G Ivls. I i 12% from setpoint: 0.774 1.26 1.1 1 I I -

1 Rgaetor Regulation: 1 I I I Tavg 120F from Tref. 1.72 1.48 1.82 I I I I Intggrgtgd_ Checkout: 1 I I I Tavg 12 0F of Tref: 1.85 1.55 0.84 I I I I #1 S/G 1evel 2 1% I i from setpoint: -2.0 to 0 1.5 -0.08 to 1.841 1 I I #2 S/G 1evel 1%

2 i I from setpoint: -2.0 to 0.3 1.0 -1.7 to-0.111 I I I Pzr. pressure 115 I

~

I psia from setpoint: 11.0 2.34 2.0 i l l I Pzr. level ilx I I from setpoint: 0.95 0.5 0.0 to 0.331 1 1 I #1 S/G pressure not I i .GT. 15 psia .GT. -28.0 -50.8 -51.0 1 I setpoint: 1 I I I #2 S/G pressure not I I .GT. 15 psia .GT. -22.0 -39.2 -40.0 l I setpoint: 1 1

l l

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 50 6.4 Reactor Coolant and Secondary Chemistry and Radiochemistry Test (Section 14.2.12.5.5)

IEgI_QB2ECTIVE_AND_

SUMMARY

The Reactor Coolant and Secondary Chemistry and Radiochemistry Test 74PA-2SS01 was performed at various power levels throughout the power ascension test program. The principal ob J ectives of the test were as follows.

(1) To conduct chemistry tests at various power levels with the intent of gathering corrosion data and determining activity buildup.

(2) To verify proper operation of the Process Radiation Monitor (PRM) and the Gas Stripper Effluent Monitor (GSEM).

(3) To verify the adequacy of sampling and analysis procedures and ensure proper chemistry control can be established and maintained.

(4) To verify that reactor coolant and secondary activity levels are maintained within the limits of the Technical Specifications.

Monitoring of plant chemistry during power ascension testing per 74PA-25501 was initiated on April 22, 1986 at Ox power and was completed on September 23, 1986 with the plant at 100* rated power.

IE!I_ DESCRIPTION Sampling and chemical analyses of the reactor coolant and secondary water systems were performed using the applicable plant operating procedures at various power levels during the power ascension. At each power level where chemistry testing was performed, samples from the reactor coolant system (RCS),

steam generators (SGs), feedwater system, condensate system, reactor makeup water tank (RMWT), and the refueling water tank (RWT) were analyzed and the results compared to the operating specifications. Out-of-specification conditions were corrected by initiating the applicable Chemistry Control Instruction. Proper operation of the Process Radiation Monitor (PRM) and the, Gas Stripper Effluent Monitor (GSEM) were to be verified by comparing PRM and GSEM readings to the laboratory analysis of grab samples which were representative of the fluid monitored by these systems.

TEST _REgykIg The key findings and major activities at each testing plateau are summarized below: .

gero Percent _ Power The.RCS, Steam Generators and makeup water supplies were all within specification for this plant condition.

3%_to 54 Power The primary systems were all within specification for this plateau and all of the secondary except for auxilliary feedwater cation conductivity was within specification. The high cation conductivity was from the makeup water (condensate storage tank) which was slightly high out of specification. This was deemed not to be a problem because the switchover to main feedwater would correct the condition.

6

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 51 lok __ Rated _Pgger The primary systems only had one item out of specification which was low hydrogen in the RCS. This condition was caused by the spraying in the pressurizer to equalize boron. The condition cleared up when the spraying was stopped. On the secondary side, everything was in specification with the exception of high oxygen in the feeedwater. The high oxygen was offset by maintaining one and one half times the normal amount of hydrazine.

20k_ Rated _ Power The primary system remained in specification throughout this plateau. The oxygen problem in the feedwater was solved during this plateau by tightening some valve packing and manway covers on the hotwells. An increase in the levels of cation conductivity, sulfate, chlorides, silica and sodium in the steam generators was experienced as new legs of the feedwater system and-feedwater drain systems were put into service for the first time. The parameters were brought back into specification by increasing the steam generator blowdown flow rate.

303_Bgtgd_Egygg The primary system remained in specification throughout this plateau. The steam generators were blown down continuously with increased flow rates at times to reduce cation conductivity, sodium, chlorides or sulfates as needed. Condensor tube leaks were experienced on two occasions and power was reduced to 40k to allow draining of hotwells for access to tubes. On each occasion, it was necessary to go to high rate blowdown on the steam generators.

Oxygen presented problems until an in-line loop instrument was found not to be filled.

80k Rated Pgyer The primary system remained in specification throughout the plateau. The steam generators were in continous blowdown with incrased flow rate to return sodium and sulfates to specification at times. The only significant problems occurred following reactor trips when it became necessary to recover from increased crud in the secondary side of the plant.

1003_Beted_Pgygt The primary and secondary systems were maintained within specification for this plateau.

Egdigtiga_Hggitgra The PRM was operable during this test with the following results:

- 1) At Ok and 5k power the activity level was too low to be able to .

compare.

2) AT the 10k and 20% power plateaus, a satisfactory comparison was made.

3) At 50% and above the monitor was off scale high. A design change was made to make the monitor less sensitive. The less sensitive monitor was not yet calibrated at the time of this writing but this portion of the procedure will be reperformed to verify the operation of the monitor.

D

,__ __ -_. -- -- - - - - - --r

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 52 CONCLUSIONS The adequacy of the sampling and analysis procedures and the ability to establish and maintain proper chemistry control was demonstrated throughout the power ascension test program. The RCS and secondary activity levels were maintained within the Technical Specification limits and increased as expected with increasing reactor power. The corrosion data gathered during the power ascension indicated no unusual or unexpected results with the exception of the high antimony levels as was experienced in Unit 1.

The proper operation of the GSEM was verified, and at lower power levels so was the PRM. At greater than 50% power, the PRM read offscale high. Additional testing on the monitor will be done after modification and recalibration is complete. The results of this testing will be addressed in a future supplement to this report.

4 0

() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 53 6.5 Unit Load Rejection (Section 14.2.12.3.7)

TEST OBJECTIVE _AND

SUMMARY

The primary objective of this test was to demonstrate that the Nuclear Steam Supply System (NSSS) can accomodate a 100% load rejection (1) without initiating a Reactor Protection System (RPS) signal or an Engineered Safety Features Actuation System (ESFAS) signal,and (2) without opening any primary or secondary safety valves. Additional objectives of the test were:

To assess the operation of the following control systems following a 100% load rejection:

Steam Bypass Control System (SBCS)

Feedwater Control System (FWCS)

Pressurizer Pressure Control System (PPCS)

Pressurizer Level Control System (PLCS) .

Reactor Power Cutback System (RPCS)

^

Control Element Drive Mechanism Control System (CEDMCS)

Turbine Rate Sensitive Power Load Unbalance Circuit (PLU).

Reactor Regulating System (RRS)

. Testing was accomplished using procedure 73PA-2MA01, " Unit Load Re3ection Test" and was successfully completed on September 13, 1986.

4 TEST DESCRIPTION This test will initiate a unit load rejection by opening the unit generator main output breakers while the plant was operating at essentialy 100% power. The loss of load will allow verification that SBCS, FWCS, PLU, RRS, PPCS, PLCS, CEDMCS and RPCS perform their designed functions.

When the load rejection is accomplished by opening the output breakers, the turbine will respond to the load rejection by initiating the Power Load Unbalance (FLU) circuit. This will momentarily close all turbine control valves to arrest terbine acceleratien, then restore normal speed.

As a result of the large reduction in steam flow, the SBCS will signal the RPCS to actuate on "Large Load Regeet" (LLR) and open all available steam bypass valves in quick open (Q.0) mode. The RPCS should drop CEA Group 5 into the reactor core to provide immediate power reduction. Following Q.0. activity the steam bypass valves should transfer to control on steam generator pressure in modulation and the RES will attempt to match primary to secondary power by inserting regulating groups to lower RCS average temperature until the Automatic Motion Inhibit (AMI) is achieved. The other NSSS control systems will function to maintain the plant within the programmed system operating parameters appropriate for the final power level achieved.

- - - , , - - , - . . . . , , . - - , , , -- - - , - - - - - - - - - - - - . , - - - - - e- , .- . - - - - -

fx

(_) (s) PVNGS UNIT 2 STARTUP TEST REPORT PAGE 54 IEgI,RESULIS This test was originally attempted on September 11, 1986 in conjunction with testing of the Fast Bus Transfer. A reactor trip occurred due to the fast bus transfer so the load rejection portion was repeated and successfully completed on September 13, 1986.

The acceptance criteria were fully satisfied. The Reactor Power Cutback System performed as designed (i.e. the reactor did not trip) no ESFAS signals were initiated and none of the primary or secondary safety valves were opened.

The plant parameters recorded during the sixty seconds following initiation of the transient and their comparison with the single value acceptance criteria supplied by the CESEC transient analysis code are provided in Table 6-7.

COggkUg10Hg The test demonstrated that the dSSS can sustain a 100% load rejection without a reactor trip, turbine trip, or a lifting of the primary or secondary safety valves. The control systems operated satisfactorily throughout the transient and data was collected to verify the CESEC predictions.

TABLE 6-7 I I I SINGLE VALUE ACCEPTANCE CRITERIA I I FOR 100* UNIT LOAD REJECTION I

i _ . __ __

i I . I I parameter Test Results Accentance Limit I I I I Max. Pressuricer Pressure (psia) 2318 2388 i 1 I l Min. Pressuricer Level (x) 50 29.4 I I I I Min. RCS Hot Leg #1 Tenp.(OF) 609 574 I I I I Min. RCS Hot Leg #2 Temp.(OF) 609 574 1 I I I Max. SG #1 pressure (psia) 1208 1242 I I I I Max. SG #2 pressure (psia) 1219 1242 I I - _

_ _ _ _ _ _ - - - - - - _ _ _ _ _ _ _l l

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() (T PVNGS UNIT 2 STARTUP TEST REPORT PAGE 55 6.6 Loss of Offsite Power Test (Section 14.2.12.5.9)

TEST OBJECTIVE _AND

SUMMARY

The principal ob]ective of the Loss of Offsite Power Test was to demonstrate that the reactor can be shutdown and maintained in a Hot Standby condition following the loss of all AC power. Testing was performed at approximately 50% power in accordance with PVNGS procedure 73PA-2NA01 following completion of testing at the 50% power plateau.

TEST DESCRIPTION Simulation of the loss of Main Generator and Offsite AC Power is accomplished by tripping the main turbine and then opening the offsite power supply breakers. The loss of offsite AC power will start the diesel generators, which should reach rated speed, voltage, and frequency eithin ten (10) seconds.

The Engineered Safety Feature sequencer will then load the Class 1E load groups. The RCP motors will stop on loss of non-class power and the reactor will trip on projected low DNBR.

The loss of power and resulting reactor trip will result in a rapid decrease in core flow and decrease in steam generator level and an increase in steam generator pressure. The essential feedpump will be utilized to provide feedwater flow to both steam generators and maintain adequate level to remove-decay heat.

The rise in steam generator pressure will be controlled by the operator action using the atmospheric steam dump valves since the steam bypass valves to the condenser will not be available. Emergency systems should maintain the reactor in hot standby under natural circulation flow conditions.

IEST_RESULTS Following the loss of offsite power initiated by a turbine trip at approximately 40% power, the reactor tripped on a Core Protection Calaulator l

(CPC) generated low Departure from Nucleate Boiling Ratio (DNBR) signal due to i the reactor coolant pumps (RCPs) coasting down. All CEAs fully inserted into the core and the ESF busses were sequentially re-energized by the diesel j generstors within 8 seconds.

l The first charging pump was manually started within one (1) minute of the l

trip followed by the charging suction being transferred to the Emergency Boration Path, then back to the VCT to preclude unneccessary boration of the

! RCS. Shutdown margin was verified to be greater then 6% delta k/k per the

! applicable surveillance procedure within one hour of the trip.

l Immediately following the trip, the critical safety parameters were monitored and stabilized by the control room operators. Verification of l

t

(m)- (]n PVNGS UNIT 2 STARTUP TEST REPORT PAGE 56 ,

l natural circulation was accomplished within 15 minutes after the trip. Primary parameters were monitored, and periodically recorded, to verify the fo11cwing:

(1) Hot leg temperatures were stable or decreasing; (2) Cold leg temperatures were close to and trending the steam generator saturation temperatures; (3) Core delta T was less than the full power delta T of 57 0F (i.e., power to flow ratio was less than 1);

(4) Hot leg RTDs and Core Exit Thermocouples were trending consistently.

Reactor vessel head and plenum subcooling were maintained at greater than 28 0F to preclude any loop void formation that could degrade natural circulation. The motor-driven emergency feedwater pump supplied flow to the steam generators which maintained their levels above 70% (wide range), thus providing an adequate heat sink for the primary system (the steam generators will provide an adequate heat sink if the U-tubes are at least one third covered, 0% wide range). The main steam safety valves and the atmospheric dump valves were utill:cd to release the transferred heat to the atmosphere.

Pressurizer level was maintained between 45% and 60%, thus remaining above the heater cutout level (26%) throughout the event. RCS pressure peaked at 2390 paia (concurrent with the low secondary heat removal and subsequent pressure increase to the MSSV setpoint) and reached a low of 2200 while natural circulation was in progress. Auxiliary spray was utilized twice (for approximately one minute each time) for primary pressure control. Careful monitoring of the secondary steam and feedwater flows precluded both RCS overcooling and initiation of safety injection (SIAS). Primary system integrity was maintained as per design throughout the loss of all AC power transient.

Included in Table 6-8 is a history of test events in chronological sequence.

CONCLUSIONS Following the loss of the main generator and all offsite power, the plant was stabilized and natural circulation was verified. Plant conditions were maintained in Hot Standby for approximately 45 minutes before offsito power was restored. Thus, the equipment, controls and instrumentation necessary to remove decay heat from the core using only emergency power supplies was demonstrated thereby satisfying all ob J ectives and acceptance criteria of the test.

1

(') [')'

'~

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 57 TABLE 6-8 SEQUENCE OF EVENTS Dgtg Ilmg Evggt 6/25/86 11:00:00 40% Full Power. Initial conditions verified 11:10:00 Turbine trip initiated.

11:10:03 Loss of Offsite Power initiated.

11:10:04 Generator Reverse Power Trip 11:10:04 2-E-NAN-S01 and 2-E-NAN-SO2 de-energized (Unit Auxiliary Transformer Output breakers open).

11:10:05 CPC channel C Low DNBR trip (90% RCP speed).

11:10:05 Reactor Trip Breakers open. All CEAs inserted.

11:10:06 4.160 Kv ESF Buses load shed on low voltage.

11:10:15 Diesel Generators started and class 1E load groups re-energized.

11:11:00 Started first charging pump.

11:13 Commenced Auxiliary Spray.

11:13:40 Operator starting to open Atmospheric Dump Valves.

11:14 Stopped Auxiliary Spray. Pressurizer pressure

'2325 psia 11:14 Main Steam Safety Valve opens.

11:14:41 ADVs open.

11:15 MSSV closed. Controlling RCS press, with PZR heaters 11:17 Controlling Steam Generator pressures with ADVs.

11:17:45 Feeding both Steam Generators with Auxiliary Feedwater Pump B.

11:25 Natural Circulation Declared. RCS stabilized at Hot Standby.

11:55 Offsite Power Restored.

12:52 Forced Circulation restored.

() -(-} PVNGS UNIT 2 STARTUP TEST REPORT PAGE 58 6.7 Biological Shield Survey Test (Section 14.2.12.5.10)

TEST OBJECTIVES AND__

SUMMARY

PVNGS procedure 75PA-22201, " Biological Shield Survey", was performed during Low Power Physics Testing (see Section 5.1) and also during the major test plateaus of the Power Ascension Test program-(20%, SON, 80%, and 100% full power). The principal objectives of this test were:

(1). To measure the radiation levels in accessible locations outside the biological shield; (2) To obtain baseline radiation levels for comparison nith future measurements of level buildup with plant operation; (3) To determine occupancy times for the measured areas during power operation.

Acceptance criteria for these measurements are based on predicted radiation levels for normal power operatic.) (100% full power) and are presented as ranges of dose rates for five different access zones. Table 6-9 shows the applicable criteria and defines the access zones.

All survey points met their applicable acceptance criteria.

TEST DESCRIPTION With the plant stabilized at the desired conditions, gamma and neutron radiation surveys were performed at selected locations in accessible areas outside of the biological shield. These surveys were performed per the plant radiation survey procedure, and included general area surveys in rooms or areas as well as more detailed surveys around penetrations, shield plugs, and other areas where streaming could occur.' A scan survey was also performed while the survey team was in transit between designated survey points. Surveys were

, performed in the Containment Building, Auxiliary Buildir.g, Main Stesa Support Structure, Turbine Building, Fuel Building, Control Building, and at various site locations exterior to the plant.

IEST RESULIS Baseline data was obtained on January 28, 1986 while the plant was in Mode 5 and again on April 19 during Low Power Physics Testing. These values were then used for comparison to the data taken during Power Ascension.

On April 22, 1986 a survey was completed while the plant was operating at 3% full pcwer (FP). All values were within the acceptable range for their designated zone.

On May 21, 1986 a survey was completed while the plant was operating at 204 FP. All values were within the acceptable range for their designated zone.

h

(' ) () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 59 On June 13, 1986 data was obtained while the plant was operating at approximately 50% FP and at 80% on August 31, 1986. Again all survey points met their acceptance criteria.

Data collection for the survey at 1004 FP was performed on September 5, 1986. All survey points were within the acceptance criteria. In addition the high radiation areas found in the Aux 1111ery Building of Unit 1 were not present in Unit 2, however, it is expected that these areas will exist when the shutdown cooling systems ~are used.

CONCLUSIONS All the data taken inside the Containment Building have met their acceptance criteria indicating the effectiveness of the Biological Shield. In addition, all other survey points within surrounding buildings were within their expected limits.

TABLE 6-9

~

I l 1 RADIATION ZONE CLASSIFICATION I I I l__ _ _ _ _ _ _ _

i I Zone i Dose Rate i Allowed Occupancy i I Designation I (arem/h) I _( Design) l I I I I i 1 1 Less than 0.5 1 Uncontrolled, unlimited i I I I access (plant personnel) l I I I I I 2 1 0.5 to 2.5 i Controlled, limited access, I I I I (40 h/wk to unlimited) I

.1 I I I I 3 1 2.5 to 15 i Controlled, limited access I

, I I I (6 to 40 h/wk) l I I I I I 4 1 15 to 100 I Controlled, limited eccess l I I I (1 to 6 h/wk) l I I I i l l 5 i Over 100 1.Normally inaccessible; I l l l l access only as permitted by I f I I I radiation protection l l l 1 I personnel (1 h/wk) l l __ i I l

I l

L

im (m) ()

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 60 6.8 Steady State Core Performance (Section 14.2.12.5.14)

TEST OBJECTIVE _AND

SUMMARY

The reactor core power distributions and core peaking factors were measured four times during power ascension testing at various power levels. These measurements were compared to predictions to confirm assumptions in the safety analysis and to verify expected core behavior. Measurements were performed at the 20%, 50%, 804, and 100% full power (FP) levels. The conditions of the power distribution measurements and the dates of performance are listed in Table 6-10.

The test acceptance criteria was satisfied if the root mean square (RMS) differences between measured and predicted power distributions were less than or equal to 3k and if the measured peaking factors were within 17.5% of their predicted values with the exception at the 20% power plateau where the values were 5% and 10% respectively. The acceptance criteria was satisfied for all measurements with exception noted in Table 6-11.

TEST _ DESCRIPTION Core power distributions and peaking factors were measured at steady state equilibrium conditions using fixed incore detector signals. The detector signals were recorded on magnetic tape using a plant computer snapshot function and then transferred to a main frame computer for further analysis. The incore detector analysis code CECOR was used to synthesize radial and axial power distributions from the fixed incore detector signals and to calculate core peaking factors from the synthesi=ed power distributions. The measured power distributions derived from the incore detector signals were compared to predicted distributions by calculating the root mean square difference between nodes. Core peaking factors were compared to predicted values on an individual basis.

TEST _RESULTS The measured and predicted core peaking factors and the RMS differences between measured and predicted power distributions are presented in Table 6-11.

l Additionally, Figures 6-1 and 6-2 show the axial and radial power distribution results and comparison with initial prediction from the 100* power test. The test acceptance criteria were satisfactorily met for all measurements with the exception of the 50% radial power distribution and the 100% radial and axial power distributions. These were found to be acceptable following a review which resulted in revised predicted values based on the difference in measured and predicted boron concentrations, Tavg and core average burnup. Figures 6-3 and 6-4 show the measured axial and radial power distribution results compared to the revised predictions at 1004 power.

CONCLygigN!

Since the acceptance criteria for this test were satisfactorily met, it can be concluded that the safety analysis assumptions concerning core peaking factors are valid and that the core is behaving as expected.

l l

l e

PVNGS UNIT 2 U.O '

STARTUP TEST REPORT PAGE 61 TABLE 6-10 I I l STEADY STATE CORE PERFORNANCE I I TEST CONDITIONS 1 I I I I I I I I I I 20k FP l 50% FP 1 80% FP l 1002 FP 1 l_ l i I I I I l l I I I I Performance 1 5-23-86 1 6-15-86 1 8-25-86 I 9- 5-86 I

. I Dates i I I I I I I I I I l l 1 I I I i 1 Actual l I I I i 1 Reactor Power i 20.78 I 50.81 1 80.12 1 99.8% 1 i i I I i i l l I I I I I RCS Teold i I I I I I (CF) 1 565.0 1 565.3 4 564.4 1 564.75 I l__ l I I I I l l I I I I I Primary i I I I I I Pressure 1 2300 psia 1 2263 psia 1 2245 psia i 2250 psia I i I I i I I i i l I I i i Boron i I I I I I Concentration 1 835 ppm I 735 ppa i 658 ppm I 622 ppm 1

-l I I I I I I  ! I I I I I CEA 1 l 1 I I I Position i Unredded i Unrodded i Unrodded i Unrodded i I --

I _

l _.I I .I I I I I I I I Core Average 1 27 MWD /T I 190 MWD /T I 739 MWD /T l 935 MWD /T I I Burnup I (1 EFPD) I (5 EFPD) I (19 EFPD) I (24 EFPD) I l -- . .I I .- .

I 3- -.I I Axial Shape i I I I I I Index (ASI)* I -0.007 i +0.020 1 +0.056 1 +0.078 I l-_=- - - - - - --I  ! -- -. . 3 I I (power in lower core half - power in upper core half)

• -- ASI = -----------------------------------------------------

total core power I

.,,,,y - - - . - - - . - . . , , , - , , , . - . . , _ .

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 62

. TABLE 6-11 (Part 1 of 2) 1 1 i STEADY STATE CORE PERFORMANCE I I TEST RESULTS I l_______ _ __ __________ . . . _ _ _ ____I I L I I l 20k FP_ TEST l l Peaking Factor i Measured i Predicted I

• 1 Accept i 1,E i Value 1 Value i Diff I Critoria_I I xy Core Planar Radial i I I i 1 Peching Factor i 1.39 1 1.46 I -4.8 1 <10k l l_I_

1 r Core Intgrtd Radial i I I I I Peabigg_ Factor i 1.37 I 1,43 1 -4.2 1 (10k I l_I i z Core Axial i I I I I

_1 1.27 1 1.27 1 0.0 1 (10k i l_I___

I q Core 3-DEeghing_Eggtgr 1 1 I I I l_____Eeakins_Eggtgr i 1.76 I 1.83 i -3.8 I <10k i l i I I i 50k FP TEST l l Peaking Factor i Measured i Predicted I k l Accept I l Value 1 Value i Diff I Criterig_l l_F l xy Core Planar Radial i I I I I Peghing_ Factor i 1.37 1 1.36 1 +0.7 I (7.5k i l_F I r Core Intgrtd Radial i 1 1 1, I 1 _13 36 1 1.35 1 +0.7 I (7.5k I l_F___ Core AxialEggh109_Eactgr_

i z i i l i i Peghing__ Factor 1 1.26 1 1.24 1 + 1. 6_ l <7.5k l 1.I l q Core 3-D i I l l l l_____Esaking_ Factor i 1.73 - __ i 1.68 i +3.0 1 <7.5k i l i I i l __ 80k FP TEST I 1 Peaking Factor i Measured 1 Predicted I k i Accept I 1 _Value__ l Value___1 Diff l_Criterig_1 l_I__y Core Planar Radial I x i I I I l l Peghing_Eggtgr_ 1 1.378 _I 1.36 i +1.3 I (7.5% __I I _I r Core Intgrtd Radial l I i l I i 12369 _1 1.34_ l +2.2__t <7.5k I l_F___

i z Core AxialEeghiDg_Eggtgr i l i i i Peghigg_Eggtgr i _1.285 _l_ 1.28 i +0.4 i <7.5k _t l_F 1 q Core 3-D 1 1 1 1 1 l__ Peghigg_Eggtgr i 1.778 1 1.71_ i +4.0 1 <7.5k I

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 63 TABLE 6-11 (Part 2 of 2)

I l

i STEADY STATE CORE PERFORMANCE I I

TEST RESULTS I I

I I

I I

I i 100k FP TEST I l Peaking Factor i Neasured i Predicted I k i Accept i 1 i Value i Value i Diff I Criteria i l'Ixy Core Planar Radial I i i 1

~~~~

l l Peakigg__ Factor l___1. 3 8__ 1 1.38 I 0.0 1 (7.5k i I _F r Core Intgrtd Radial i I I I I Peakin: ,Egetor 1 1.37 1 1.36 I +0.7 I (7.5k__l 1

l_Iz Core Axial 1 1 1 I i 1,I . Pecking Factor i 1.30 1 1.28 i +1.2 i <7.5% i i o Core 3-D I l l l l l_____ Peaking Factor 1 1.79 1.73 +3.5 i 1 1

<7.5k _l _

1 1

I RMS DIFFERENCES I i i l I (1) l (2) i I (3) 1 Acceptance i I I 20k I 50k 1 8Ck I 100k i Criteria I i i i i l I I l Radial Dist. I 3_.69k 1 2.97k_I 2.97k i 2,405% i _13k 1

-1 I l l l l l l Agigl_Dist. _I_3.47 l_2.51N I 2.39k i 1.80k i 13k __

l (1) -- Acceptance criteria at 20k was < 5k.

(2) -- Value prior to re-evaluation 0 50k was Radial: 3.164k and Axial:.2.677%

(3) -- Values prior to re-evaluation 0 100k were Radial: 3.258k and Axial: 3.031k.

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

_m , , - , - . . .

's v PVNGS UNIT 2 GN STARTUP TEST REPORT PAGE 64 FIGURE 6-1 RELATIVE AXIAL POWER DISTRIBUTION 100* FP TEST REl.ATIVE AXIAL PO'eB P' P. ........ rrrrrrrFrrN 8 858888888888888888888 j _i 1^ 4 : I 6 6 4 i i 1 e 6 4 4 4 6 SE .

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 65 FIGURE 6-2 RELATIVE RADIAL POWER DISTRIBUTION 100* FP TEST A B C D E F G H J K L M N P R S T IIII'lI .7 .. ..

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a o

r <x PVNGS UNIT 2 STARTUP TEST REPORT PAGE 66 FIGURE 6-3 RELATIVE AXIAL POWER DISTRIBUTION 100* FP TEST (REVISED PREDICTIONS)

IEl.ATIVE AXIAL. POWER p p. . . . . . . . . r r* r F F r* r r' F r* P 8 8588e8808885088888888

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PVNGS UNIT 2 STARTUP TEST REPORT PAGE 67 FIGURE 6-4 RELATIVE RADIAL POWER DISTRIBUTION 100* FP TEST (REVISED PREDICTIONS)

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PVCGS UNIT 2 STARTUP TEST REPORT PAGE 68 6.9 Intercomparison of PPS, Core Protection Calculator (CPC), and PHS Inputs (Section 14.2.12.5.15)

TEST _0BJECTIVES_AND SUMMAM The principsi obyective of the "Intercomparison" test was to verify that all Main Control Room indications of selected plant parameters monitored by the Plant Protection System (PPS), Core Protection Calculators (CPC), Plant Monitoring System (PMS) and the Main Control Board (MCB) process instruments were correct and consistent with acceptance criteria that were based on vendor and design accuracies.

Testing was accomplished using PVNGS procedures 72PA-2SB02 (20N power) on May 21, 1986, 72PA-2SB06 (50% power) on June 9, 1986, 72PA-2SB10 (80% power) on August 24, 1986 and 72PA-2SB15 (100k power) on September 5, 1986. All instruments were either within their required acceptance band or sufficient data was sent back to the vendor to Justify a broadening of the acceptance criteria.

IEgT_ DESCRIPTION PPS, CPC, PMS and MCB dats were collected (as simultaneously as possible from all available indications) for the following process intrumentation at the four major power ascension test plateaus:

(1) Reactor Coolant System (RCS) hot and cold leg temperatures (2) Reactor Coolant Pump (RCP) speeds and differential pressures (3) Pressuri=er pressure and level (4) Steam Generator pressure and level (5) Reactor Core differentisi pressure

- (6) Steam Generator differential pressure At all four power pisteaus the data from each instrument was. cross-compared to verify that the various indications of that particular persmeter were consistent and securate to within the specified acceptable tolerance bands.

IEgI_RE@ULIS Most of the instruments at each of the four major power pisteaus were found to be within the specified tolerance bands. Test Exception Reports were written to document those instruments thst did not meet their acceptance criteris.

These instruments were then recalibrsted and successfully retested at that particular power level.

O ev"os u"t: 2 STARTUP TEST REPORT PAGE 69 ggNgtUSIONS The accuracy and consistency of Control Room indications of selected plant.

parameters monitored by.the PPS, CPC, PMS, MCB, and process instruments were acceptable for power operation.

4 9

9 e

.._._,.__..s

f (j_ (9j PVNGS UNIT 2 STARTUP TEST REPORT PAGE 70 6.1C Verification of CPC Power Distribution Related Constants Test (Section 14.2.12.5.16)

The testing performed in accordance with this section of CESSAR verified the agreement of the various constants used by the Core Protection Calculators (CPCs) in the power distribution calculation with those values determined by measurement. These constants were the rod (CEA) shadowing factors, the planar radial peaking factors, the temperature shadowing factors, the shape annealing matrix, and the boundary point power correlation constants. The verification of these constants was performed in three major tests, as described in the following sections.

6.10.1 Verification of CEA Shadowing Factors and Radial Peaking Factors TEST OBJECTIVES AND__

SUMMARY

This test verified that the CEA shadowing factors and the radial peaking factors used in the CPC power distribution calculation are valid by comparing the measured values with the predicted values which are part of the CPC data base. If the measured and predicted values did not agree within the acceptance criteria, correction multipliers were ad3 usted in the CPCs to correct the predicted valuea. This test was performed on June 22, 1986 with the reactor at approximately 504 full power, using test procedure 72PA-2RX18, "CEA Shadowing Factor / Radial Peaking Factor".

TEST DESCRIPTION The test was initiated with the reactor unrodded (All Rods Out, or ARO) and at a stable power level and equilibrium xenon conditions. Baseline data was taken, including measurements of the excore detector responses, incore detector responses, and the secondary calorimetric power level. Several redded conf-igurations were then established while core power was held essentially constant using boron dilution or addition. These rodded configurations werei a Regulating group 5 fully inserted

• Regulating groups 4 and 5 fully inserted

• Partlength group P 75% inserted with 4 and 5 fully inserted

• Group P 754 inserted with only 5 fully inserted a Group P 754 inserted with no other rods inserted.

Data was recorded for each rodded configuration, including the excore detector responses, the incore detector responses, and the secondary calorimetric power.

Cs6_Shadgwigg,[agiggs--The CEA shadowing factors are used to account for the alteration in the neutron flux seen by the excore detectors when the control roda (Control Element Assemblies, or CEAs) are inserted, assuming no change in gross power level. A CEA shadowing factor (Fx ) for each condition was determined from the measured data using the general relation:

Da Psec(ARO) y . _______

x ___________

DARO Psec(R)

Where: Dp = Excore detector response with CEAs inserted (with part lengths inserted, use only middle detector response, all other cases, sum top, middle and bottom

! DARO = Excore detector response for the unrodded condition Psec(ARO) = Secondary calorimetric power for the unrodded condition

Psec(R) = Secondary calorimetric power with CEAs inserted

'U PVNGS UNIT 2 (f STARTUP TEST REPORT PAGE 71 The CEA shadowing factors determined from the measured data were then compared with the predicted values and new shadowing factor multipliers (ASM1) were calculated for input to the CPCs using the following equation.

Fx (calculated)

ASMi = -------------------

Fx(CPC Data Base) ggdig1_EggBigg_Egetggs--The radial peaking factors account for the change in overall radial power distribution caused by CEA insertion by ensuring that the most limiting radial peak for the existing CEA configuration is used in the CPC calculation. " Measured" radial peaking factors were determined from an analysis of the fixed incore detector data taken with the various CEA configurations using the code CECOR. The CECOR calculated radial peaking factor value (CECOR Fxy) for each of the six measured configurations (one unrodded case plus five rodded cases) was then compared with the predicted value used internally by the CPCs for the particular configuration. For each case in which the measured peaking factor was larger than the predicted peaking factor, the CPCs were ad 3 usted to increase the values used in the CPC power distribution calculation.

9 TEST RESULTS The CEA Shadowing Factors for all rod configurations were determined and are summarized in Table 6-12. New shadowing factor multipliers (ASM) were calculated for each condition and input to the CPCs.

The measured planar radial peaking factors were less than the predicted planar peaking factors for all CEA configurations. No ad 3 ustments were required and the test results are summarized in Table 6-13.

gQHC ggIgHg All CEA shadowing factors and radial peaking factors used by tne CPCs wsre verified to be accurate, or the appropriate ad 3 ustments were made to correct the CPC values to the measurement.

l

/

I l

l- .

t... . - _ ._ __ _ . - - . . - - . _ ,- _ _ _ _ - - . - --_ -

O O PvsGS U 1T 2 STARTUP TEST REPORT PAGE 72 TABLE 6-12 I I I MEASURED CEA SHADOWING FACTORS (CSF) i I I I Predicted CSFs Computed from Test Data i i CEA_Grouo/ Position CSF _ Expected Range CPC_A CPC B CPC C CPC D* I I I I 5 / Full in 1.104 1.071 to 1.137 1.1091 1.1047 1.1097 1.1175 I i i I 5 & 4 / Full in 1.038 1.007 to 1.069 0.9777 0.9826 0.9891 0.9992 1 1 I i 5 & 4 / Full in i I P / 75% in 1.054 $.023 to 1.085 0.9402 0.9317 0.9481 0.9874 1 1 -

1 I 5 / Full in i I P / 75x in 1.125 1.091 to 1.159 1.0776 1.0510 1.0742 1.1286 I I I I P / 75x in 1.011 0.981 to 1.041 0.8719 0.9523 0.9668 1.0085 I I l

• Detector D results are those that followed re-evaluation due to excore detector mispositioning (see Section 6.10.3)

TABLE 6-13 I I I MEASURED PLANAR RADIAL PEAKING FACTORS (ALL CPC CHANNELS) I

~

Predicted Measured I l CEA_Grou2/ Position Peaking _ Factor _ _ Peaking Factor I I I I (Unrodded) 1.39 1.3788 I I I i 5 / Full in 1.56 1.5241 I I I I 5 & 4 / Full in 1.68 1.5824 I I I I 5 & 4 / Full in i I P / 754 in 1.70 1.6756 I I I i 5 / Full in I l P / 754 in 1.56 1.5448 I I I I P / 75% in 1.40 1.3931 1 I

- =1

~' PVNGS UNIT 2 STARTUP TEST REPORT PAGE 73 6.10.2 Verification of Temperature Shadowing Factors TEST,,0BJECTIVEJ ND SUMMAM The purpose of this test was to verify the adequacy of the temperature shadowing factors used in the CPC power distribution calculation by measuring the decalibration of the excore safety channel detector responses associated with variation of the cold leg (reactor inlet) temperature. If the factors determined from the measured data differed from the predicted factors by more than 104, the measured values were installed in the CPCs. This test was performed at approximately 50% power on June 16, 1986 using test procedure 72PA-2RX22, " Temperature Decalibration Test (50%)".

TEST DESCRIPTION The temperature shadowing factors are used by the CPCs to ccupensate the calculated neutron flux power for changes in the reactor inlet temperature (Teold), since temperature variations in Tcold will be accompanied by density changes in the water which will affect neutron attenuation across the downcomer of the reactor vessel and, hence, the response of the excore detectors.

To obtain the data needed to determine the actual temperature shadowing factor Teold was increased about 3 degrees above the initial temperature

('563CF) in 1 degree increments, followed by a decrease of 12 degrees sgenerally in 1.5 degree steps), followed by a return to the initial temperature in 1.5 degree steps. The. temperature changes were made via boration/ dilution.

After a short stabilization period at each new value of Tcold, the following test data were recorded: excore detector responses, cold leg temperature, and secondary calorimetric power.

From each set of data, the ratio of the excore detector responses to the secondary calorimetric power was determined. Because the excore detector responses are affected by variations in cold leg temperature, but secondary calorimetric power is not, changes in this ratio are a direct indication of the impact on detector response caused by cold leg temperature variations. Each'of these ratios were then normalized to the ratio determined from the data taken at the initial Tcold, to correct for any actual variations in reactor power over the course of the testing. That is, at each different Teold (t),

(Excore detector response)t Dt RATI0t (Secondary calorimetric power)t Psec(t) and, Psec(563oF)

= X -----------

(NORMALIZED RATIO)t RATI0t D5630F The validity of using the smoothed secondary calorimetric power in the calculations was questioned due to the long wait periods following boration/ dilution -i.e. time for water mixing to occur and time for the secondary plant to respond to changes in the primary plant. In response to an evaluation request. CE-Windsor recommended the use of primary power (BDT) vice secondary power based on the afore mentioned reasoning.

A least squares fit of the normalized ratios versus Teold was then performed to produce the best estimate of the excore detector response variation as a function of Teold. The slope of the line resulting from this fit (i.e. the l change in excore response per OF change in Tcold) was the measured temperature shadowing factor.

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 74 IESI_BE!EI!

The temperature shadowing factor measured for each CPC channel was in acceptable agreemer.t with the predicted value. Table 6-14 summarizes the test results.

CONCLUSIONS The temperature shadowing factor measured for each CPC channel was in acceptable agreement with the predicted value. Therefore, the installed value used in the CPC calculation was satisfactory and no adjustments were necessary.

f TADLE 6-14 I I I MEASURED TEMPERATURE SHADOWING FACTORS (/0F) l I I I Accept. 1 I CPC channel Meas. Pred. kDiff*

• Diff I l - -- .

I A .0058 .0058 0.0 1 10 l l l l B .0053 .0058 -8.6 10 l I I l C .0056 .0058 -3.4 1 10 1 I I I D .0057 .0058 -1.7 1 10 I l -

1 (Meas - Pred)

• xDiff = -------------

X 100%

Pred

I

() () PVNGS UNIT 2 STARTUP TEST REPORT 1

PAGE 75 6.10.3 Vorification of Shape Annealing Matrix and Boundary Point Power Correlation Constants TEST OBJECTIVE _AND

SUMMARY

The Shape Annesling Matrix and Boundary Point Power Correlation Test was performed at the 20% and 50% full power (FP) pistesus. The ob]ective of the test performed at the 20% FP plateau was to verify that the installed CPC Shape Annealing Matrix (SAM) and Boundary Point Power Correlation Constants (BPPCCs) were suitable for power ascension to 50% FP. This was accomplished on May 23, 1986 using test procedure 72PA-2RXO4, " Shape Annealing Matrix", which performed a comparison of the measured average axial power distribution and the axial power distribution esiculated by each CPC channel to verify acceptable agreement between the two.

At the 50% FP pisteau, the ob]ective was to actually measure and install (if necessary) a new SAM and BPPCCs for each CPC channel. Testing was accomplished on June 23 through 25, 1986 and retested August 16 through 18, 1986 using test procedure 72PA-2RX19, " Shape Annealing Matrix (504)". The measured SAM and BPPCCs were compared to the installed CPC values for each CPC channel to determine the adequacy of the latter prior to power ascension above 50% power.

This comparison of measured and installed values did not meet the acceptance criteria for any of the CPC channels, necessitating the installation of the measured SAM and BPPCCs into the CPC data base.

TEST DESCRIPTION 20% FP_ Test--In this test, the reactor core average axial power distribution was measured with all rods withdrawn from the core and equilibrium xenon conditions established. A " snapshot" was recorded of the fixed incore detector responses concurrently with the recording of,the CPC calculated axial power distributions. The " measured" exial power distribution was determined from an analysis of the fixed incore detector responses using the code CECOR. The axial power distribution determined by CECOR was then compared to that calculated by the CPCs to verify that the root-mesn-square (RMS) error between the two was no greater than 52 If the RMS error exceeded 5%, the error between the measured and esiculated axial peaks ano axini shape indices would be further examined to determine whether the measured ano CPC csiculated values were in acceptable agreement. If this agreement was not verified, a measurement of the actual SAM /BPPCC values would be performed and these values installed in the CPCs before the reactor power was incressed above the 204 power level.

503_[P Test--The SAM /BPPCC constants are used by the CPCs to calculate an accurate axis 1 power profile from the excore detectors. In this test, the dats used to determine the SAM and BPPCCs were measured over a range of various axial power shapes to ensure that this data would be representative of the range of axial power distributions expected throughout Cycle 1. To accomplish this, a free running axial xenon oscillation was established. For the next thirty hours (approximately the length of one free xenon oscillation cycle) the excore detector responses for each CPC channel were recorded simultaneously with fixed incore detector responses at approximately fifteen minute intervals.

()

~

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 76 Each set of incore detector responses was processed using the code CECOR to provide a set of " measured" peripheral axial power distribution information. .A least squares analysis of the measured power distribution data from CECOR versus the corresponding excore detector data was then performed to determine the best set of SAM /BPPCC constants for relating measured excore detector responses to the true peripheral axial power distribution.

The results of the least squares analysis are subsequently used to compute a SAM " Test Matrix" value for each CPC channel which gives an indication of the acceptability of the SAM. Test matrix values in the range of 3.0 to 6.0 ensure that the design CPC power synthesis uncertainty factors are adequate and will result in conservative CPC Departure from Nucleate Boiling (DNBR) and Linear Power Density (LPD) calculations.

To determine whether the SAM values measured during the test needed to be installed into the CPC data base, the following criteria were used:

1) For each CPC channel, if the difference between the measured and predicted SAM is less than or equal to 2.0% for all elements, no adjusments are required.

2) For each CPC channel, if the difference is greater than 2.04, the SAM test matrix value shall be calculated. If the test matrix value is in the range of 3.0 to 6.0, all of the measured SAM elements shall be installed in the CPC data base.

To determine if the BPPCCs measured during the test needed to be inctalled into the CPC data base, the following criteria were used:

1) For each CPC channel if the difference between the measured and predicted BPPCC is less than or equal to 3.0% for each constant, no ad]ustment is required.

2) If the difference between the measured and predicted value is greater than 3.0%, the measured BPPCC shall be installed in the CPC data base.

TEST _RESULTS 20% FP_ Test--The RMS error b* tween the CECOR measured and the CPC calculated i exial power distribution was less than 5% for all CPC channels. Therefore, no further action nor any change to the CPC data base was required. The results are summarized in Table 6-15.

50% FP_ Test--The measured SAM and BPPCC values differed from the previously installed values by an amount that exceeded the acceptance criteria of 2.0% and 3.0% respectively. A subsequent review of the data revealed that detector D could have been mispositioned. Further investigation verified the misellignment and the detector was realigned during a following outage. The SAM test was repeated and again, the installed values failed to meet the acceptance criteria thereby requiring installation of the measured values. Additionally, during the data collection for the retest, some of the data stored on magnetic tape was found irretrievable. However, the obtained data was evaluated by.CE-Windsor to be acceptable for use since no adverse inpact would be placed on the CPC overall DNBR and LHR uncertainty factors. Consequently, the measured SAM and BPPCC values were installed in the CPC data base. As indicated by the results, the test matrix values for all CPC channels were within the 3.0 to 6.0 test acceptance range. The test results are summarized in Tables 6-16 and 6-17.

CONgLMSIQNS The SAM /BPPCC constants initially installed in the CPCs were satisfactory for operation up to the 50% power test plateau. New constants were determined and installed at 50% power.

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

,----m -v-, ,------,-,,-.----n~ , - - , - - , , - - -- - , . ,

~ . .. . - .. - - .

O O PvacS UN1T 2 STARTUP TEST REPORT .

PAGE 77 1

4 TABLE 6-15 1 1

, i RMS VAI.UES 20k SAM TEST l l' (x) l I I

. . I Acceptance i I CPC Chgggel RMS Value Criteria i i i i A 3.5593 5.0 1 I B 3.3613 5.0 l I C 3.7282 5.0 l I D 3.5861 5.0 1 I _ ----

I TABLE 6-16 i BOUNDARY POINT POWER CORRELATION CONSTANTS I l (ALL CPC CHANNELS) l I I I Parameter Original Installed Measured xDiff." l i I Value __ Value _l I I I BPPCC1 0.01389 0.01377 -0.857' I

i BPPCC2 0.08113 0.07802 -3.830 I I . BPPCC3 0.01433 0.01412 -1.470 1 i BPPCC4 0.08141 0.07883 -3.170 I l --- -

i Where, (BPPCi(measured) - BPPCCl(original installed))

xDiff = ----------------------------------------------

=100%

BPPCCi(original installed) for i = 1 to 4 l

~- - - - ~ ev=,--,,----,w,w-,,e

1 PVNGS UNIT 2 STARTUP TEST REPORT PAGE 78 j TABLE 6-17 I I I Shape Annealing Matrix Elements I l l I I I Channel A Channel B I l l I SAM _ i _ Original Meas. xDiff I Original Meas. kDiff 1 I Element i Installed Value i Installed Value i l_ l Value i Value i I I I I I S11 1 3.4995 3.6726 4.95 1 3.4943 3.8747 10.89 1 i S12 I -0.3993 -0.7288 82.52 1 -0.4052 -0.9828 142.55 I I 313 1 -0.2116 0.0327 -115.45 I -0.2029 0.1601 -178.9 I I S21 1 -0.7857 -0.5028 -36.01 1 -0.6600 -0.6487 -1.71 I I S22 l 4.5093 4.1278 -8.46 1 4.2663 4.2198 -1.09 I I S23 1 -0.8257 -0.6268 -24.09 I -0.6252 -0.5891 -5.77 I I S31 1 0.2861 -0.1699 -159.39 1 0.1656 -0.2260 -236.48 I I S32 1 -1.1100 -0.3990 -64.05 1 -0.8612 -0.2370 -72.48 I I S33 1 4.0371 3.5940 -10.98 1 3.8281 3.4290 -10.43 I I I I I I Test Matrix 3.6399 3.7333 I I Value- I l I I Channel C Channel D I I I I SAM l Original Meas. *Diff I Original Meas. kDiff I I Element i Installed Value i Installed Value I l_ _

l Value i Value ___ _ l_

i I I I i S11 1 3.5430 3.7539 5.95 1 3.6220 3.7164 2.61 1 1 S12 1 -0.3814 -0.8865 132.44 1 -0.5224 -0.8029 53.70 i i S13 1 -0.2287 0.1389 -160.7 l -0.1571 0.0815 -151.9 I I S21 1 -0.7587 -0.6150 -18.94 1 -0.8612 -0.6428 -25.36 I I S22 1 4.4281 4.3065 -2.75 1 4.6055 4.4004 -4.45 I I S23 1 -0.8265 -0.7228 -12.55 1 -0.8896 -0.8217 -7.63 I I S31 1 0.2157 -0.1389 -164.4 1 0.2392 -0.7360 -130.8 l l

l S32 1 -1.0467 -0.4200 -59.87 I -1.0832 -0.5975 -44.84 I I I S33 1 4.0552 3.5840 -11.62 1 4.0467 3.7402 -7.570 1 1 - - - - _ - _ - -I _ _ _ _ - - - _ _

l _ __________ _1 i Test Matrix 3.7446 3.7525 i 1____hin - -- ____ - - .

l Where, (Si3(measured) - Si)(original installed))

%Diff = ----------------------------------------- =1004 Si)(original installed) for i and 3 = 1 to 3 l

l D

, -- - , . - - - - - - - - - . < n.

l O ev"cs uarr 2 STARTUP TEST REPORT PAGE 79 6.11 Main and Emergency Feedwater Systems Test (Section 14.2.12.5.17)

TEST OBJECTIVE AND

SUMMARY

The primary objectives of this test were to verify the satisfactory operation of the Main and Emergency Feedwater Systems and also to verify the adequacy of the associated piping systems and supports.

Four test procedures were performed to evaluate the low power operation of the Feedwater Control System (FWCS), the downcomer-economi=er valve transfer which occurs at approximately 15% full power (FP) and the performance of the main feedwater pumps:

1) 73PA-2FWO1, "FWCS Test at 10% Power" evaluated the performance of the FWCS at a power level of 10% FP and was performed on May 19,1986.

2) 73PA-2FWO2, "FWCS Valve Transfer Checkout Test with Power Decreasing",

evaluated the transfer of the main feedwater flow from the economi=er to the downcomer during a decrease from 20% to 10% FP and was performed on May 20 and 21, 1986.

3) 73PA-2FWO3, "FWCS Valve Transfer Checkout Test with Power Increasing",

evaluated the transfer of the main feedwater flow from the downcomer to the economizer during an increase from 10% to 20% FP and was conducted on May 20, 1986.

4) 73PA-2FWO4, Feedwater System Operability" evaluated the performance of the main feedwater pumps by collecting data at each 10% power increment r10% to 1004 FP). This test also includes removal of one high pressure feedwater heater train from service while operating at 100% FP to determine if there is any plant capacity degradation.

To verify the adequacy of the piping systems and supports,-test procedure 73PA-22ZO3, " BOP Piping Dynamic Transient Test ", was performed. It consisted of two sections concerning the feedwater system:

1) Section 8.5 monitored the feedwater transfer from the downcomer to economizer during power increases and was performed on May 20, 1986.

2) Section 8.6 monitored the feedwater transfer from the economi er to downcomer during pos:er decreases and was performed on May 21, 1986.

TEST DESCRIPTION 73pA-2FWO1--This was a test of the FWCS's ability to maintain steam generator level within 15 % of the control setpoint during steady state and minor transient conditions. The steady state portion simply involved placing the FWCS in automatic and observing control of steam generator level vis 'a-vis the setpoint. The transient portion of the test involved both increasing and decreasing the level setpoint (approximately 54) in both ramp and step functions to determine whether the FWCS controlled steam generator levels to within 54 of the setpoint (after allowing for a brief stabili=stion period).

) ) PVNGS UNIT 2 STARTUP TEST REPORT .:

PAGE 80 '

Z3PA-2FWo2--This test evaluated the response to an automatic valve transfer from the economizer to downcomer feed systems at approximately 15% power for adherence to acceptance criteria. The decreasing power transient was accoaplished by leading the steam demand of the secondary plant and allowing the NSSS to follow. Data was taken taken during and after the transfer and was evaluated against specific acceptance criteria.

73PA-2FWO3--This test was performed in essentially the same manner as 73PA-2FWO2 described above, except that the transfer was from the downcomer to the economizer with reactor power increasing. Again, an automatic transfer was made and similar data were collected for comparison with acceptance criteria.

73PA-22293--To measure any loads that may have been imposed on the piping systems and restraints, thirteen load-sensing pins were installed at various hanger locations. Data was collected during the various evolutions previously mentioned and was evaluated against the acceptable loads calculated for each load pin. Also, a visual inspection of the piping, the supports and ad]oining structures was performed after the transient portions of the test.

73P6-2Fyg4--A data snapshot was taken on the plant computer at the start of a feedpump, after feedwater had been transferred to the feedpump from the Auxiliary Feedwater System, and at 10% power increments from 10% to 100% FP.

Single pump data was collected for each pump from 10% to 50% FP and dual pump data was collected above 50% FP.

TEST RESULTS Z3PA-2EW91--The acceptance criteria for the test were satisfied. The actual test results are compared with the acceptance criteria in Table 6-18.

Z3E6-2EW92--As can be seen in the test results listed in Table 6-19, steam generator pressure exceeded its acceptable limits following the automatic valve transfer. A Test Exception Report was generated and the results were declared acceptable based on- updated acceptance criteria provided by CE-Windsor. The relaxation of the acceptance criteria is allowable because of the minimal consequences of the additional pressure rise.

Z3P6-2FWO3--The Aceptance Criteria for the automatic valve transfer test were fully satisfied. The test results are listed in Table 6-20.

Z3PA-2FWO4--The feed pumps delivered water to the Steam Generators at the required flows and temperatures through all power levels. A review of all data indicated the Feedwater System to be performing as required.

23P6-22293--The observed loads were well below the maximum acceptable loads (which varied from 1,700 to 55,000 lbf) calculated for the specified 1 cad pins.

The loads observed were, in fact, at or below 100 lbf. The visual inspection revealed that no damage was sustained by the piping, the supports, or the adjoining structures. Thus, the acceptance criteria were satisfied.

{~}'

- ()' PVNGS UNIT 2 STARTUP TEST rep 0RT PAGE 81 CONCLUSIONS The ability to perform the downconer-economizer valve transfer automatically with no adverse impact on overall plant control was demonstrated by these tests. Also, the ability of the FWCS to perform as designed under a number of different circumstances was demonstrated. In addition, it was confirmed that the design and construction of the main and auxiliary feedwater systems and associated hangers are adequate to support any normally encountered operating modes without sustaining damage or apparent degradation.

TABLE 6-18 I I I FWCS CHECKOUT AT 10x POWER (73PA-2FWOl) l I TEST RESULTS I I I I Acceptance Criteria Maximum Deviation i I I I S/G 1evels do not deviate i I more than 5X froTA set pt. 1.13 I I for steady state conditions 1 1 1 i S/G 1evels do not deviate i I more than 1 5% from set pt. 1.12 I I for ramp set pt. changes i I I I S/G 1evels do not deviate i I more than 54 from set pt. 1,81 1 I for step set point changes I l_ i l

t i

l I

{

l l

i

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 82 TABLE 6-19 I I I FWCS VALVE TRANSFER--POWER DECREASE I I (73PA-2FWO2) TEST RESULTS I I .-

1 I ACCEPTANCE CRITERIA AUTOMATIC 1 l_ (deviation from initial value) Actual Value i I I I #1 Nuc power decreased -2.10 l I no sore than 4% 2 l I i #2 Nuc power decreased -2.40 l I no sore than 44 I i i i S/G #1 level increased less 7.0 l I than 30% l l- 1 I S/G #2 level increased less 10.0 1 I than 30k 1 1 1 I S/G #1 pressure decreased -47.0" I I less than 30 psia I I I i S/G #2 pressure decreased -47.0" I I less than 30 psia l I- 1 I RC cold leg temps, increased i I no more than 6 degrees 1 I Tc 1A: 3.9 I I Tc 1B: 2.4 I I Tc 2A: 3.3 I I Tc 2B: 4.0 I I -

1

• Allowable tolerance later changed to 50 psia by C-E resulting in this value being acceptable.

9

- - , , , - - - - - - - - ,,.n,.. . - -

-,c - - - - - - . - . ~ . . - - - -

O (] PVNGS UNIT 2 STARTUP TEST REPORT PAGE 83 TABLE 6-20 I I I FWCS VALVE TRANSFER--POWER INCREASE I I (73PA-2FWO3) TEST RESULTS I I _l.

I ACCEPTANCE CRITERIA AUTOMATIC 1 I (deviation from initial value) Actual Va5uel_

l i I #1 Nuc power increased less 3.70 1 I than 5% I I I i #2 Nuc power increased less 4.50 1 .

I than 5% 1 i i i S/G #1 level decreased less -11.2 I I than 40% i i i i S/G #2 level decreased less -16.7 I I than 40% 1 i l l S/G #1 pressure increased 48.0 I I no more than 50 psia I i i i S/G #2 pressure increased 48.0 I I no more than 50 paia i i 1

- I RC cold leg temps do not decrease i I more than 8 degreeu I I Tc 1A: -8.0 I ,

I Tc 1B: -3.6 I I Tc 2A: -7.7 I I Tc 23: -7.3 I i -- - -

-- 1 l

l l

l i

L

) PVNGS UNIT 2 STARTUP TEST REPORT PAGE 84 6.12 CPC Verification and COLSS Verification (Sections 14.2.12.5.18 and 14.2.12.5.20)

TEST 0BJECTIVE__AND_

SUMMARY

The objectives of these tests were to verify the calculations of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) performed by the Core Protection Calculators (CPCs) and the Core Operating Limit Supervisory System (COLSS), in addition to evaluating the effect of process instrument noise on the CPC system.

Testing was performed in accordance with the "COLSS/CPC Verification" test which was directed by PVNGS procedures 72PA-2SB02, at 20% full power (FP) on May 24, 1986; 72PA-2SB06, at 50% FP on June 16, 1986; 72PA-2SB11, at 80% FP on September 02, 1986; and 72PA-2SB16, at 100% FP on September 5, 1986. The results from each of these tests were satisfactory.

TEST _ DESCRIPTION The calculations performed by each CPC channel are verified by comparing the values of Local Power Density (LPD) and Departure from Nucleate Boiling Ratio (DNBR) recorded from each channel with the values calculated by the Combustion Engineering CPC FORTRAN simulator code CEDIPS. When provided with a known variation of input data recorded from each of the CPC channels, the CEDIPS code calculates a range of values for LPD and DNBR which would be expected to bound the actual values observed on each of the CPC operator display devices.

The CPC data used as input to CEDIPS is manually gathered from the CPC display devices as meximum and minimum values observed over a specified period of time and consists of: pressurizer pressure, RCP speeds, control rod positions, RCS cold and hot leg temperatures, and excore detector responses. The CEDIPS DNBR and LPD values are compared to those observed and recorded during the test. If the observed DNBR and LPD values are within the range of expected values, the functioning of each CPC channel is considered to be verified and process instrument noise has not affected the CPC operation.

As a further step in evaluating the effect of process instrument noise, input signals to the " noisiest" CPC channel and its analog outputs were recorded on FM tape (at 100% power only) over a period of approximately two hours. The determination of the " noisiest" CPC channel was accomplished by monitoring the maximum variation in the DNBR value calculated by each CPC channel at 1 minute intervals over a 10 minute period. This data was gathered for evaluation by the NSSS vendor and had no specific test acceptance criteria.

For the COLSS calculations, a statistical analysis of different sensor inputs measuring the same parameters was performed to ensure that the instruments were consistent and functioning properly. The statistical analysis is performed automatically on demand by the COLSS Sensor Deviation Statistical Routine which is executed on the Plant Monitoring System (PMS). Additionally, COLSS input and output values were collected via pMS data snapshots.

Following completion of testing at each test plateau, the COLSS statistical data, the COLSS input and output data snapshots, and the FM recorded CPC data (100N power only) were transmitted to the NSSS vendor for evaluation.

l

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 85 IESI_BESULTS The CPC calculated minimum / maximum DNBR and LPD values recorded during the test for the 20%, 50%, 80% and 1002 power plateaus are provided in Ta' ole 6-21 along with the CEDIPS calculated range of expected values. All of the CPC calculated DNBR and LPD values were bounded by the corresponding CEDIPS range of

~ values with the exception of the maximum DNBR value calculated by Channel C at 204, the minimum LPD value calculated by channel C at 50%, the maximum LPD calculated by channel A at 80% and 100%, the minimum DNBR value calculated by channels A,C and D at 80% and channel A at 1004 A Test Exception Report was written to perform a retest for all failed cases. The retest results, in all cases show that the CPC values for DNBR and LPD were satisfactorily bounded by the CEDIPS calculated range of values. These final results are those indicated in Table 6-21.

CONCLUSION The CPC and COLSS calculations of DNBR and LPD were satisfactorily confirmed at each of the major test plateaus.

e e

d 9

- - - - - , , . . , - , - , , . , . - . - _ . - - _ , . . , - - . - - . . - - . - . - - - - . _ . . . - - - - - .n .-

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-Q_ (J . PVNGS UNIT 2-STARTUP TEST REPORT PAGE 86 TABLE 6-21 (Part 1 of 2) 1 I I CPC/CEDIPS COMPARISON VS POWER LEVEL I I i 1_-- - - -- --

1 IPARAMETER CHANNEL A . CHANNEL B i

~ ~ ~

1204 FP TEST CEDIPS CPC CIDIPS CPC I

. I ~

I ILPD MAX 4.25402 4.1015 4.35187 4.0902 i l MIN 3.83537 4.0210 4.02511 4.0823 I I I IDNBR MAX 10.0014 9.2311 9.69534 9.2014 I I MIN 8.61693 9.0517 8.45415 9.1136 1 -

_I _ _ . _ _ _ = = _ _ _

I 1504 FP TEST I

-l I' ILPD MAX 7.95670 7.7100 8.07785 7.9130 1 I MIN 7.58250 7.6500 7.67928 7.7520 t i I IDNBR MAX 4.34481 4.1520 4.28044 '4.1240 I I MIN 4.00893 4.0730 3.92344 4.0860 I l . _ _ _ . __ - _ == ==-- _L 180% FP TEST I I_ _

i ILPD MAX 12.1906 12.058 12.3727 12.061 1 I MIN 11.6824 11.963 11.6823 11.941 1 I I

- IDNBR MAX 2.18855 2.1570 2.21607 2.1161 1 2.0980 1 I MIN 2.01965 2.0293 2.01961 L_ _ = - - - _ _ _ = - - - _- __=___ .=- = = = - -- - I 11004 FP TEST I 8

1_ -- - - - - - - - - - - - -

ILPD MAX 14.518 14.416 14.491 14.153 I I MIN 13.978 14.096 13.769 14.046 i

, I I IDNBR MAX 1.785 1.7377 1.7267 1.6359 I I MIN 1.576 1.6378 0.1413 1.5974 !

> ' - = = _ - - - - - - - - -

- _ = - - - - - - - - - _ _ _ - - - - - ---____1 1

h

() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 87 TABLE 6-21 (Part 2 of 2) i I I CPC/CEDIPS COMPARISON VS POWER LEVEL i I i 1 I IPARAMETER CHANNEL C - CHANNEL D I l=- - -1 120% FP TEST CEDIPS CPC CEDIPS CPC i l I ILPD MAX 4.55131 4.3202 4.07217 4.0564 I I MIN 4.21892 4.3129 4.00057 4.0469 I I I IDN3R MAX 9.24357 8.6831 9.66864 9.5397 I I MIN 8.11642 8.6238 9.31003 9.4774 I I -- - ____- _____ _ __ _

l 150% FP TEST I 1

l-ILPD MAX 8.25152 7.9914 7.96623 7.7336 I I MIN 7.86277 7.9011 7.64660 7.7005 I I I IDNBR MAX 4.27717 4.2043 4.32066 4.1821 1 I MIN 4.00272 4.1002 4.01040 4.1003 I 1

==______________I

,jg_ , ____ _ =_ - - -

1_= l ILPD MAX 12.2420 12.135 12.2707 12.1*2 I I MIN 11.8380 12.047 11.6523 11.893 1 I I

- IDNBR MAX 2.16354 2.1167 2.25534 2.1530 1 I MIN 2.06166 2.0822 2.06015 2.0983 1 I

= -=_ -

--_ . __ _ L 1100% FP TEST I I -- --

= __ 1 ILPD MAX 14.355 14.342 14.491 ----.406 14 I I MIN 14.068 14.168 13.772 14.010 1 I I IDNBR MAX 1.7089 1.6679 1.7373 1.6592 l I MIN 1.5745 1.5763 0.1426 1.5460 t I i 6

O) n

(_j PVNGS UNIT 2

(_

STARTUP TEST REPORT PAGE 88 6.13 Steam Bypass Valve Capacity Test (Section 14.2.12.5.19)

TEST OBJECTIVE AND

SUMMARY

The atmospheric steam dump and steam bypass control system valve capacity test 73PA-2SG01 was conducted on July 23, 1986 with the reactor initially stabilized at 304 power. The principal objectives of the test were as follows:

(1) To verify that the capacity of each of the Atmospheric Dump Valves upstream of the Msin Steam Isolation Valves (MSIVs) is prester than 6% and less than 11% of the total main steam flow to be encountered at full power conditions.

(2) To verify that the total capacity of the Steam Bypass Control System (SBCS) controlled turbine bypass valves is greater than 55% of the total full power steam flow rate.

These percentages are based upon a full power steam flow rate of 17,118,144

~

lba/hr at a steam generator pressure of 1070 paia. The measured capacity of each SBCS valve and ADV met the design criteria.

IE!T_ DESCRIPTION Stable plant conditions were established with the 4 ADVs and 8 SBCS valves closed and reactor power equal to turbine load. A baseline feedwater flow rate was determined. The capacity of each valve was measured individually by cycling the valve full open and then closed. As the valve position was cycled, reactor power and feedwater flow were adjusted to maintain the turbine load as steady as possible. The valve capacity was derived from the difference in feedwater flow rate with the one valve fully open and the baseline condition of all valves closed. This difference in feedwater flow rate was corrected for the difference between full power steam pressure and the test condition steam pressures.

IEST_REQU(T!

The capacity of each SBCS valve and each ADV was measured to be within the design criteria of greater than or equal to 6% and less than 11% of total full power steam flow rate. The minimum measured valve capacity was 9.904 and the maximum measured valve capacity was 10.604 The total capacity of the 8 SBCS valves was 82.20% and satisfied the minimum criteria of 55%.

99NCLUgIQHS The espacity of each SBCS valve and each ADV was measured to be within the design criteria and satisfied the safety analysis assumptions concerning the maximum capacity of a single valve.

~

PVNGS UNIT 2 STARTUP TEST REPORT PAGE 89 6.14 Incore Detector Test (Section 14.2.12.5.20)

IEgT.,0BJECTIVE_._ AND __ SUMM ARY Testing of the fixed incere detector system (FICDS) was performed in accordance with the "Incore Detector (Fixed) Test" which was directed by PVNGS procedures 72PA-2RIO1, at 20% full power (FP) on May 22, 1986; 72PA-2RIOS, at 504 FP on June 5, 1986; 72PA-2RIlo, at 80% FP on August 27, 1986; and 72PA-2RI15, at 1004 FP on September 5, 1986. The objectives of this testing were:

1) To verify the operability of the system via execution of automatic test functions on the Plant Monitoring System (PMS);

2) To record and review the fixed incore detector voltages to identify potential detector / amplifier failures;

3) To verify that the detector signals received at the input of the PMS were consistent with those measured at the output of the amplifiers:

4) To measure'the background voltages.

The operability of the FICDS was verified at each test plateau and all test objectives were satisfied.

IEST DESCRIPTION

[1ged_Incore Detector _Testigg--The operability of the FICDS was verified by

=

executing automatic test functions programmed into the PMS. The three test functions are:

1) Conversion of a zero current input to each amplifier to a zero voltage output to within 0.025 vde (2ERO OFFSET).

2) Conversion of a full scale input signal (10 microamps) to each amplifier to a full scale ouput (10 volts) to within 1 0.136 vde

. (AMPLIFIER GAIN),

3) Measurement of the cable leakage resistance in each detector and -

evaluation of the measured resistance values to a minimum acceptable value of 1000 k ohms.

The fixed incere detector test functions were executed via test pushbuttons located in each Fixed Incore Amplifier Bin. The PMS is programmed to evaluate the data and summarize the comparison of the measured values with the incorporated tolerances.

s

-~- ,

7 O

t,,,/ (O

,,7 PVNGS UNIT 2 STARTUP TEST REPORT PAGE 90 A set of fixed incore detector voltage signals and uncompensated flux signals were also obtained via the PMS and reviewed to verify that the signal levels were within an expected range for the appropriate power level and core location. This evaluation was performed by comparing signals from symmetric detectors and/or from detectors located in surrounding asserblies.

At 100% FP, raw detector voltage signal levels were measured at the amplifier assembly card test points (amplifier output) and compared to the signal read by the PMS to demonstrate that the voltages agree within acceptable levels (i.e., 1%). A measurement of the detector background signal contribution was also performed via the amplifier assembly card test points to verify that the actual background was equal to or lower than the background correction terms incorporated in the PMS data base.

TEST _RESULTS

[1ggd_lggogg_Dgiggtor Test--The zero and full scale amplifier signal checks were satisfactorily completed at each test plateau. However, PMS calculated cable leakage resistance values of less than the minimum acceptable limit of 1000 k ohns were obtained for seversi detectors at each test plateau. Following evaluation of the results, a "use-as-is" disposition was rendered based on continued monitoring of the detectors via weekly Technical Specification surveillance testing.

Measurements performed during 100% FP testing verified that the difference between the measured signal at the amplifier output and the signal read by the PMS were within the acceptance criteria limit of a 1%. In addition, measurement of the detector background signal levels showed the background to be less than the acceptance criteria limit of 5% of the flux signal from the fixed incere detector located at the core mid plane for the particular core location.

99E9LUSI95

. The operability of the fixed incore detector system has been verified at each of the test plateaus and all test objectives satisfied.

() () PVNGS UNIT 2 STARTUP TEST REPORT PAGE 91 6.15 Shutdown from Outside the Control Room Test (Section 14.2.12.S.8)

TEST OBJECTIVE AND_

SUMMARY

The objective of 73PA-2SF02, " Shutdown Outside the Control Room" was to demonstrate that the plant can be shutdown and maintained in a Hot Standby condition from outside the Control Room.

The acceptance criterion for this test was to perform a safe shutdown of the plant from outside the control room and maintain selected plant parameters within a specified range for at least 30 minutes using equipment that would normally be available only at the remote shutdown panel.

TEST _ DESCRIPTION The test is performed by utilizing a normal operating crew and a standby crew. The standby crew serves as Control Room observers and are to take action only if a problem that involves plant safety develops. The operating crew consists of the minimum shift complement as defined in Table 6.2-1 of the PVNGS Technical Specifications.

The operating crew performs the test by evacuating the Control Room and proceeding to the Remote Shutdown Panel. Once established, they initiate the switchgear panel. After the trip is verified, they establish control of the plant using equipment available at the Remote Shutdown Penel and maintain Hot Standby conditions for approximately one hour and fifteen minutes. Control of the plant is then transferred to the standby crew in the Control Room and the Remote Shutdown Panel is secured.

TEST REggkT!

. This test has not yet been performed. Following performance, results of the test will be presented in a supplement to this report.

/' A (N) L/ PVNGS UNIT 2

'5TARTUP 17.ST REPORT PAGE 92 ACKNOWLEDGEMEhTS The PVNGS Unit 2 Startup Report represents the efforts of several individuals. Section 1 was prepared by D.F. Hoppes. Sections 2,4 and 5 were prepared by G. Foster. Section 3 was prepared by M. Hulet. Section 5 was prepared by W. Ryder. The Report was edited by G. Foster. The authors and editor wish to recoanize and thank those people at PVNGS who contributed to the preparation of this report. Among them are:

W. Asbury R. Black E. Edmonds R. Einar L. Elliot T. Goetz T. Hall -

A. Herres P. Hoffspiegel B. Johnson I. Johnson P. Keller J. Moreland i W. Osmin T. Samuels H. Sattig B. Schumacher J. Taggart B. Wheelis e

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