ML20197J190

Power Uprate Startup Test Rept for Nov 1992-Mar 1993

ML20197J190
Person / Time
Site:	Fermi
Issue date:	03/31/1993
From:	DETROIT EDISON CO.
To:
Shared Package
ML20197J186	List: ML20197J190 NRC-93-0072, Forwards Power Uprate Startup Test Rept, Which Provides Results of Test Program Following Increase in Licensed Power Level for Fermi 2 Authorized by Amend 87 in Operating License.Program Was Completed at 98% of Rated Power
References
NUDOCS 9801020107
Download: ML20197J190 (34)

Text

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o POWER UPRATE STARTUP TEST REPORT NOVEMBER 1992 TO MARCH 1993

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'4 Table of Contents 1.0 Introduction 1.1 Purpose 1.2 Test Report Format ,

1.3 Plant Description 1.4 Startup Test Program Description  ;

1.5 Test criteria  ;

' l .6 . References  !

2.0 General Test Program Information 2.1 Chronoic;y of Major Events 3.0 Test R(aults Summary 3.1 Chemical and Radiochemical 3.2 Radiation Measuremenu 3.3 Local Power Range Monitor (LPRM) Calibration 3.4 Average Power Range Monitor (APRM) Calibration 3.5 Pncess Computer  ;

3.6 Reactor Core Isolation Cooling (RCIC) i 3.7 - High Pressure Coolant Injection (HPOI) ,

3.8 Core Perfonnance l 3.9 Pressure Regulator 3.10 Feedwater System 1.11 Turbine Valve Surveillance 4 3.12 Recirculation System Flow Calibration 3.13 Steady State Data Collection 3.14 Main Steamline Radiatior Measurement 3.15 Loose Parts Monitoring System 6 e t

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. 1.0 latroduction 1.1 Purpose

'Ihe purpose of this Stanup Ten Repon is to provide a summary of the test resuhs obtained during stanup testing completed following implemeniation of the

- Power Uprate (104.2 % of original rated power). 'Ihe test objectives and criteria are described in the Power Uprate Safety Analysis which was reviewed and approved by the Nuclear Regulator Commission (NRC). In addition, the original Stanup Test Program as descrited in the Updated Final Safety Analysis Report I (UFSAR) Section 14.1.4.8 wa', evaluated er tests which were applicable for the .

uprated power condition. Sequence of Events:SOE 92-01 contains the test procedures used to meet these test requirements. This repon is submitted as - l required per Technical Specification 6.9. In addition, this test augments an  ;

extensive post refueling outage test program. l Included in this test report are the descriptions of results obtained dering the test l program and any corrective measures that were requirtd to obtain satisfactory 1 operation.-

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'Ihe stanup test program confirmed that the small change in power has little effect  ;

on process water chemistry corulitions and background radiation. T5 chemical l and radiochemical analysis results were essentially unchanged when compared to  ;

h'. rical data. 'Ihe small change in background dose rates at uprated conditions f we<e masked by the small variation in survey instrument positioning or performance. - l Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection  ;

(HPCI) were automstically actuated in the test mode at uprated power conditions.

The system performed, as expected, by achieving rated flow and pressure well within design time requirements, while maintain sufficient margin to the overspeed trip.

The Pressure Control and Feedwater systems functioned, as expected, in a well  :

behaved and stable manner efter introduction of chan, es to the control systems during testing.  ;

Steady state data were collected at each of the Test Condition and evaluated. ,

Coneems identified during this review were addressed before continuing to the l next Test Condition. In general, the data was consistent with expectations.

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1.1 Purpose (continuation)

The analysis of the Drywell noise level data collected at uprated conditions concluded that the I.cose Pans Monhodng system sensitivity would not be compromised and a change to the noise monitonng setpoint would not be necessr.ry. , i

, l This test was originally intet.ded to include power ascension to 100% Core

' thermal Power (CTP), however, this was not possible and 98% CTP was the maximum achieved power level. Between 98% and 100% power the Turbine Control Valves (1tVs), which were modified dudng the Refueling Outage, began modulating excessivdy and the average Turbine Control Valve position was larger than expected. This behavior prevented any funher attempts to increase power ,

above 98%. In anticipation of operation at 98% CTP for the balance of the fuel cycle, the startup tests odginally planned for 100% were evaluated and applicable tests were performed at 98% power.

The Turbine Manufacturer has begun model testing in order to better understand the reason for the Turbine Control Valve flow limitation. Several proposed design i modifications have been beodel tested and a finalized design is forthcoming.

Modifications are planned for the next refueling outage. A supplemental test program will be perfonned at 100% CTP and a test report will be submitted following completion of the test program.

In conclusion, the Startup Test program-confirmed that the nsll increase in power has little effect on reactor and safety system performai ' radiological and chemistry parameters were within expected limits.

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1.2 Test Report Fonnat Sections 1.0 and 2.0 of this report provides general information about the Ferti 2 plant and the testing program. Section 3.0 provides a basic description of the testing performed along wi'h a summary of the results and analysis obtained from each test. Each test summary is divided into three subsections covering the pulpose, test criteria and results of each test.

1.3 Plant Descriptlon ne Fermi 2 Nuclear Power Plant is located in Frenchtown Township, Monroe County, Michigan, ne Nuclear Steam Supply System consists of a General Electric Boiling Water Reactor (BWR) 4 nuclear reactor rated at 3430 Megawatts thermal (MWt) (104.2 % of original rated power), coupled to an English Electric Turbine Generator rated at 1203 Megawatts electric (MWe) (105% of original-rated power). The plant has a Mark I contalnment with a torodial suppiession

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1.4 Power Uprote Test Program Description Operating at power uprate conditions increased Reactor Dome pressure by 25 psi and increased feedwater and steam flow. Systems and components were evaluated to determine which would be affected by operation at uprated power conditions.

The necessary test requirements ,and the following test objectives were established:

1. To perform tests directed towards deruonstrating proper :ystem performance of High Pressure Coolant Injection system, Reactor Core Isolation Cooling system, Feedwater Level Control system and Pressure Control system.

2. To collect steady state data such that Balance of Plant (BOP) performance, Core Thermal Limits and those systems and components which were identified as potentially limiting are monitored and can be projected for uprated power conditions before the previous power level is exceeded.

3. To verify proper Main Steamline Radiation Monitor setpoints for higher background radiation conditions.

4. To verify proper Loose Parts Monitoring setpoints for higher ambient noise levels inside the Drywell.

5. To verify proper Area Tcmperature setpoints for higher ambient area tempenture conditions.

Table 1 1 shows a general overall view of the test prognm which includes test categories and should be considered in conjunction with Figure 1-1 which describes the various test conditions.

The Power Uprate Test Program began with plant heatup and continued through power ascension to 98% CTP.100% CTP could not be reached due to Turbine Control Valve flow limitation.

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O 1.5 Test Criteria level 1 - A Level I criterion nonnally relates to the value of a process variable assigned it the design of the plant, component, systems, or asmciated equipment, if a Level I criterion is not satisfied, the plant will he placed in a suitable hold condition until resolution is obtained, Testing compatible with this hold condition may be continued. Following resolution, the applicable test must be repeated to verify that the requirements of the Level I criterion are now satisfied.

level 2 - A level 2 criterion is aswelated with expectations related to the performance of the systems. If a Level 2 criterion is not satisfied, operating and testing plans would not necessarily be thered.

1.6 References

1. Power Uprate Safety Analysis 92150

2. Updated Final Safety Analysis Report UFSAR Section 14.1.4.8

3. Sequence of Events SOE 92-01 Power Uprate Startup Test 6

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Table 11 TEST CATEGORIES TEST CONDITIONS A B C D E* 'i

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Chemical and Radiochemical X Radiation Measurements X X LPRM Calibration X X APRM Calibration X X X X X Process Coinputer X X RCIC X X HPCI X Core Performance X X X X Pressure Regulator X X X  ;

Foodwater System X X X Turbine Valve Surveillance Various >

Recirculation System Flow  ;

Calibation - X  :

Steady State Data Collection X X X X X i Loose Parts Monitoring X X Main Steamline Radiation Monitors X i Test Condition A < 86% Power Test Condition B = 86% 87% Power Test Condition C = 95% 96% Power Test Condition D = 97% 98% Power '

Test Condition E = 99%.100% Power Because of the h. ability to reach 100% power, Test Condition E testing was performed where applicable in Test Condition D.

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2.0 General Test Program Infonnation 2.1 Chrwoology of Major Events Plant enters Operational Condition 2 11-4-92

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Reactor declared critical 11592 Completed HPCI/RCIC low pressure test Il 6 92 Synchronized the Turbine Generator 11-7 92 Completed Test Condition "A" (< 86% CTP) Testing 11 10-92 11P Probe "E" failed Il 10-92 Entered Test Condition "B" (86% CTP) Il-?0-92 Completed Test Condition "B" (86% CTP) Testing Il 11-92 t

Entered Test Condition "C" (96% CTP) 11-11 92 Replaced TIP Probe "E" Il 14-92  ;

Completed Test Condition "C" testing Il 18-92 Entered Test Condition "D" (98% CTP) Il-18 92 Reactor Manual Scram due to loss of feedwater flow 11-18-92 Reactor Startup following reactor scram 11-20-92 Reduced power to repair a steam leak on the south reactor feedwater pump 11 28-92 Completed Test Condition "D" (98% CTP) testing 12-1-92 Increase in reactor power from 98% to 100%

(Test Condition "B") was unsuccessful due io an l apparent Turbine Control Valve flow limitation 12-1-92  ;

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2.1 Chronology of Major Events Reactor Recirculation pump runback occurved as the result of an extraction steam line rupture 12-1 92 Plant shutdown to repair the ruptured extraction steam line 12592  ;

Reactor stanup following repairs 12 13 92 i Reactor power was reduced and the Nonh  ;

Condenser pump taken out of service due to high motor vibration 12 25 92 -

Returned to 98% CTP following replacement of the Nonh Condenser Pump Motor 1-9 93

• Ihe #3 Turbine Control Valve failed closed ,

and reactor power was reduced to compensate for reduced turbine now 1-22-93 Returned to 98% CTP following Unitized Actuator repairs 1-23 93 Reactor Shutdown to repair condenser tube leaks 21093 Reactor stanup following repairs 2-13 93 On Site Safety Review Organization (OSRO) approval of changes to the Stanup testing program for long term operation at 98% CTP 2-16-93 Reactor Automatic Scmm due to the loss of Condenser vacuum 2-19 93 Reactor Stanup following loss of Condenser vacuum 2-20-93 Completed the test program and obtained OSRO approval of test results 3 23-93 10 1

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3.0 Test Results Su== mary 3.1 Cha= leal and Radlochensteal l 3.1.1 Purpose l De purpose of this test was to collect and analyze chemical and radiochemical samples to ensure that power uprate can be maintained _

while complying with Technical Specification and the Fuel Warranty chemistry specifications.

3.1.2 Criteria level 1 Chemical and radiochemical parameters identified in the Technical i Specifications and Fuel warranty sliall be met.

Level 2 l Primary loop chemistry and radiochemistry quality shall be within the Action Level 1 guidelines of the Chemistry Specifications. ,

3.1.3 Results .

Chemical and radiochemical samples were taken while operating at 98%

CTP, ne chemical and radiochemical analysis results for uprated conditions were essentially the same when compared to historical values.

All Level 1 and Level 2 test criteria were met excep for Condensate Demineralizer Effluent Dissolved Oxygen levels. De actual dissolved oxygen level was 13 Pans Per Billion (ppb), which is outside the specified ,

level 1 limit of between 20 and 200 ppb. This condition is not new or unknown, low dismived oxygen levels were observed during the original Stanup Test Program and have been continually monitored. As the resul.

of a study conducted in 1991, oxygen levels between 10 and 20 ppb are accepted as having no detrimental effects on corrosion rates or other parameters of concern. De lower dissolved oxygen level is the basis for the :urrent operating limits.

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3.2 Radiation Measurunent 3.2.1 Purpose ne purpose of this test was to collect baseline area radiation levels and some selected effluent radiation levels for.uprated power condition and to ensure that ;

postings are in accordance with the requirements of 10CFR20, ' Standards for Protection Against Radiation".

3.2.2 Criteria Level 1 The radiological postings meet the requirement .0CFR20.

3.2.3 Results Radiatlan measurements were taken in the fonn of process and area radiation '

monitor data and site surveys. De measurements were taken at 96% CTP and 98% CTP All areas were posted in accordance with the requirements of 10CFR20.

He Main steamline radiation monitor readings increased by an average of 6.8%

between Test Co.1dition C (96% CTP) and D (98% CTP). During normal steady state conditions radiation reading can differ by as much as 5% over the test period and therefore, the increase in background radiation levels are within the expected range. The increase in main steam line dose rate did not result in a measurable increase in general area hallway dose rates in the Reactor Building or Turbine Building. Dose rates measured at room entrances and at penetrations ,

affected by live steam or neutron radiation showed both increases and decreases between 96% CTP and 98% CTP. This indicates that the small increase in general area dose rates were occasionally masked by small variations in survey instrument positioning or perfonnance.

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. i 3.3 LPRM Calibration  !

3.3.1 Purpose ne purpose of this test was to document and correctly sequence the perfonnance f of a routine IERM calibration surveillance for the Power Uprate Test Program ,

in order to assure that core thermal limits were correctly calculated prior to  ;

operation at uprated power and before any transient testing was performed using reactor power scram avoidance as an acceptance criteria. his did not replace the Technical Specification routine calibration requirements but was used to augment  ;

the program for testing purposes.  !

3.3.2 - Criteria Level _1 LPRMs are calibrated and meets the acceptance criteria of surveillance test procedure 54.000.05,*LPRM Calibration Process Computer Determination".  ;

3.3.3 Results ,

he LPRMs were calibrated in Test Condition "C" (95%) for preparatic.n to enter  ;

uprated power condition and transient testing in Test Condition "D" (98 % CTP). l It was also intended to repeat this test in Test Condition "E" (100% CTP) but the Turbine Control Valve performance precluded the need for this test.  ;

i All LPR'J channels were satisfactorily calibrated. During calibration, an

- obstruction was detected in the Transverse In-core Probe (TIP) machine "E" channel 4 which prevented further probe movement. De obstruction was later determined to be a detached ' IIP detector.

Fermi 2 is equipped with the 3 D Monicore computer system. This system utilizes .

three dimensional neutron diffusion equations and is capable of calculating and substituting axial flux profiles for calibration of suspicious or unusable TIP channels. Using TIP data for all channels, 3 D Monicore calculated the axial flux :

profile used to complete the calibration of LPRMs associated with TIP machine l "E" channel 4. ,

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3.4 APRM Calibration j 3.4.1 Purpose  !

The purpose of this test was to document and correctly sequence the performance of a routine APRM calibration surveillance for the Power Uprate Test Program +

in order to assure that core power was correctly calculated prior to operation at uprated power and before any transient testing was performed using reactor power scram avoidance as an exapeare criteria. This did not replace the Technical Specification routine calibration requirements but was used to augment the program for testing purposes.

3.3.2 Criteria i Level 1 i APRMs are calibrated and meets the acceptance criteria of sutveillance test procedure 54.000,06, "APRM Calibration". - ,

3.3.3 Results ,

During power ascension APRM calibrations were performed in each Test Condition to assure accurate power readings were used to evaluate transient testing. As a result of APRM calibration, minor adjustments were necessary for some of the APRM at each test conditions.

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3.5 Process Cosaputer  !

. 3.5.1 Purpose ne purpose of this test was to document and correctly sequence the performance of the permanent plant procedure used to verify the correct Process Computer cycle data bank, his test provided additional assurance that core thermal limits were accurately monitored prior to operation at high power levels. In addition, this test verified that the Process Computer was properly condgured following the first full 1.PRM calibration.

3.5.2 Criteria Level 1 De Process Computer cycle data bank is verified and meets the acceptance-criteria of test procedure 53.000.07, " Process Computer Cycle Data Bank Verification".

The Process Computer was properly conDgured and meets the acceptance criteria of test procedure 53.000.07, " Process Computer Cycle Data Bank Verification".

3.5.3 Results The Process Computer contains cycle specific data which is changed at the beginni'ig of each fuel cycle. The Process Computer cycle data bank was loade<!

and veritial prior to operation above 25% CTP. In addition, after the first high power TIP calibratloc, the Process Computer was verified to be configured for high power conditions.

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P 3.6 RCIC 3.6.1 rurpose RCIC system is designed to operate over a wide range of steam pressure. At Power Uprate conditions Reactor Dome pressure was increased by 25 psi and therefore, the required operating range of RCIC increased by 25 psi. De purpose of this test was to demonstrate adequate system performanae following installation of the steam by pass valve and operation at uprated power conditions. ,

3.6.2 Criteria Level 1 ne RCIC pump must be able to deliver 640 GPM,100% rated flow, w.u a pump discharge pressure between 100 and 250 psi above reactor pressure.

Pressure and rated flow must be reached within 50 seconds from initiation on automatic start at any reactor pressure between 150 psig and rated pressure, ne RCIC turbine shall not trip or isolate during the automatic starts.

Level 2 To provide a margin to overspeed trip and isolation, the first and subsequent speed peaks on the transient start shall not exceed the rated speed of the RCIC turbine by more than 5 %.

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3.6.3 Results The RCIC system tune up runs were conducted during plant heatup and an  :

automatic actuation test was performed at 97.9 % CTP. 'Ihe newly installed steam by pass valve functioned as expected.

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At 945 psig, RCIC was auto started several times to confinn settings associated .

with the steam by pass valve and governor modification. Following the tune-up runs, the RCIC system was capable of supplying 640 GPM at a discharge  ;

pressure of 1120 psig within 31.5 seconds. No isolation occurred. The RCIC turbine speed response was well behaved and only exhibited one speed peak at 4659 RPM, which is less than the maximum acceptable speed of 4777 RPM (5 %

above rated speed).

At 97.9% CTP, with the Reactor Pressure Vessel (RPV) pressure at 1027 psig, RCIC was automatically actuated from a cold ambient condition. The RCIC system was capable of supplying 660 GPM at a discharge pressure of 1160 psig within 36 seconds. This meets the. required minimum Dowrate of 640 GPM with a discharge pressure between 100 and 250 psi greater than the Dome pressure within 50 seconds. No isolation occurre.d. The RCIC turbine speed response was well behaved and only exhibited one speed peak at 4539 RPM, which is less than the maximum acceptable speed of 4777 RPM.

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i 3.7 HPCI 3.7.I Purpose HPCI system is designed to operate over a wide range of steam pressure. At  !

Power Uprate conditions Reactor Dome pressure was increased by 25 psi and i

therefore, the required operating range of HPCI increased by 25 psi The purpose of this test was to demonstrate adequate system performance for power uprate power conditions.

3.7.2 Criteria Level 1 J

l The HPCI pump must be able to deliver 5200 GPM,100% rated flow, with a l pump discharge pressure between 100 and 250 psi above reactor pressura.

Pressure and rated flow must be reached within 29 seconds from initiation un automatic start at any reactor pressure between 150 psig and rated pressure.

1 The HPCI turbine shall not trip or isolate during auto start.

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'Ihe margin to avoid the overspeed trip shall be at least 10% of the nominal overspeed trip setpoint of 5000 RPM during auto starts of the HPCI system 1.7.3 Results During the first attempt to auto start HPCI, the governor system failed to function due to a loss of instrumentation power caused by a failed power supply dropping resistor. 'Ihe failed component was replaced and a second attempt to auto start HPCI was successful.

At 97.3% CTP, with the RPV pressure at 1027 psig, HPCI was automatically started. The HPCI system was able to supply 5300 GPM at a discharge pressure of 1140 psig within 18.1 seconds. This meets the required minimum flowrate of 5200 GPM at between 100 and 250 psi greater than Dome pressure within 29 seconds. No isolation occurred. The HPCI turbine speed response was well behaved and only exhibited one speed peak at 4023 RPM, which is less than the maximum acceptable speed of 4500 RPM (10% below rated speed).

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C 3.8 Core Performance 3.8.1 Purpow he purpose of this test was to document and correctly requence the performance of a routine core thermal limits verification for the Power Uprate Test Program in order to assure that core thermal limit margin was maintained prior to operation at uprated power and before any transient testing was performed using reactor power scram avoidance as an acceptance criteria. His did not replace the l Technical Specification routine surveillance requirements but was used to augment ,

the program for testing purposes. l 3.8.2 Criteria Levei1 Core thermal limits are within the required limits specified in 54.000.07, " Core Performance Parameter Check" 3.8.3 Results The Maximum Fraction of Limiting Critical Power Ratio (MFLCPR), Maximum Fraction of Limiting Power Density (MFLPD) and Maximum Average Planar Linear Heat Generation Rate Ratio (MAPRAT) were calculated in Test Conditions "B", "C" and "D" in preparation for transient testing. These ratios are the fraction of calculated values to the limiting values. A ratio less than 1.000 would be within limits. As expected, the MFLCPR, MFLPD and MAPRAT were all less than 0.900.

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3.9 Preuure Regulator 3.9.1 Purpose The purpose of this test was to determine the optirnum settings of the Pressure  !

Control System by analysis of transients induced in the reactor by means of the Pressure Regulators. In addition, the test demonstrated the takeover capability of I the backup Pressure Regulator on failure of the controlling Pressure Regulator and set spacing between the setpoints at an appropriate value.

3.9.2 Criteria .

1 Level 1 The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to Pressure Regulator changes described as follows:

a. 10 psid step change on Pressure Regulator No. I while Turbine Control Valves are in control of pressure and the Bypass Valves are closed,

b. 10 psid step change on Pressure Regulator No. 2 while Turbine Control Valves are in control of pressure and the Bypass Valves are closed,

c. Failover of Pressure Regulator No. 2 to Pressure Regulator No, I while Tarbine Control Valves are in control of pressure and the Bypass Valves are closed.

d. Failover of Pressure Regulator No, I to Pressure Regulator No. 2 while Turbine Control Valves are in control of pressure and the Bypass Valves are closed.

c. 10 psid step change on Pressure Regulator No. 2 while Tuttine Bypass Valves are in control of pressure.

f. 10 psid step change on Pressure Regulator No. I while Turbine Bypass Valves are in control of pressure.

g. Fallover of Pressure Regulator No, I to Pressure Regulator No. 2

+ while Bypass Valves 4re in control of pressure.

h. Fallover of Pressure Regulator No. 2 to Pressure Regulator No. I while Bypass Valves are in control of pressure.

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3.9.2 Crkeria level 2 i

The decay ratio must be less than 0.25 for each process variable that exhibits oscillatory response to Pressure Regulator changes described as follows:  ;

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a. 10 psid step change on Pressure Regulator No. I while Turbine .i Control Valves are in control of pressure and the Bypass Valves  !

are closed.

b. 10 psid step chanFe ressure Regulator No. 2 while Turbine Control Valves ars control of pressure and the Bypass Valves  !

are closed.

c. Failover of Pressure Regulator No. 2 to Presse c Regulator No. I while Turbine Control Valves are in control of pressure and the Bypass Valves are closed.  :

d. Fallover of Pressure Regulator No. I to Pressure Regulator No. 2 l while Turbine Control Valves are in control of pressure and the Bypass Valves are closed.

e. 10 psid step change on Pressure Regulator No. 2 while Turbine '

Bypass Valves are in control of pressure.

f. 10 psid step change on Pressure Regulator No. I while Turbine

. Bypass Valves are in control of pressure. .

g. Failover of Pressure Regulator No. I to Pressure Regulater M. 2 while Bypass Valves are in control of pressure,

h. Failover of Pressure Regulator No. 2 to Pressure Regulator No. I while Bypass Valves are in control of pressure.

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3.9.2 Critoria (continuation)

The pressure control deadband shall be small enough for steady state limit cycles, If any, to produce turbine steam flow variations no larger than 0.5 % of ruled flow

(< = 1.0% peak to peak). 'Ihis criteria is satisfied by utilizing the generator -

output equivalent to steam flow. ,

During simulated failure of the controlling Pressure Regulator, if the setpoint of the backup Pressure Regulator is optimally set, the backup pre sure regulator shall control the transient so that the peak neutron flux and peak vessel pressure i

remains below the scram setting by 7.5% and 10 psi, respectively.

After a pressure wtpoint adjustment, the time between the setpoint change and the '

occurrence of ths ,aressure peak shall be 10 seconds or less.

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i 3.9.3 Results ,

The Pressure Regulator Test was performed in Test Condition "C", %% CTP (100% of original sted power) and in Test Condition "D",98% CTP.

In Test Condition "C', %% CTP, the *B" pressure transmitter required recalibration in order to reduce the difference in values between the two '

regulators. This adjustment was necessary because the difference was too great to allow proper swap over to the backup Pressure Regulator.  ;

For the 10 psi down and up step changes performed in this test, all process ,

variables were highly damped and no decay ratio were found to exceed 0.25, #

satisfying both Level 1 and Level 2 criteria, An analysis of the collected data showed a maximum peak to peak MWe variation of 9 MWe which was less than the 12 MWe acceptance criterion (1.0%

equivalent steam flow).

During the Pressure Regulator fallover to the backup Regulator, the neutron peak flux was between 101% and 103% which was less than the 110.5% limit. The reactor peak pressure was between 1034 psig and 1036 psig which was less than 1083 psig limit. The responses were highly damped and well behaved with decay ratios less than 0.25 which meets both Level I and Level 2 acceptance criteria.

An analysis of the step changes shows that peak reactor pressure occurred between 4 ano 7 seconds after the step initiation, thus satisfying the acceptance criterion of less than or equal to 10 seconds.

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3.9.3 Resuks (continuation)

In Test Condition *D" attempts were made to reduce steady state nrbine Control-Valve modulation through dynamic setting adjustments. However, this effort did not resuk in sufficient improvement to justify a change. De dynamic settings were returned to the original values. During the 10 psi step changes the Bypass Valves momentarily lifted and rescated while testing the No.1 Pressure Regulator. De Bypass Valves remained seated while testing the No. 2 Pressure Regulator. This behavior was acceptable but did cause some concern over l adequate Turbine Control Valve capacity. This concern was later confirmed while attempting to increase reactor power to 100%.

For the 10 psi down and up step changes performed in this test, all process variables were highly damped and no decay ratios were found to exceed 0.25, satisfying both level 1 and 1.evel 2 criteria.

An analysis of the collected data showed a maximum peak to peak MWe variation of 11.33 MWe which was less than the 12 MWe acceptance criterion.

During tne pressure regulator failover to the backup regulator, the neutron peak flux was 103.7 % and 105.3% which was less than the 110.5 % limit. De reactor peak pressure was between 1035 and 1037 psig which was less than the 1083 psig limit, ne responses were highly damped and well behaved with decay mtios less .

than 0.25 which meets both Ixvel I and level 2 acceptance criteria.

An analysis of the step changes shows that peak reactor pressure occurred between 4.1 and 8.1 seconds after the step initiation, thus satisfying the acceptance criterion of less than or equal to 10 seconds.

T g l.

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. . .- .- -, - = . . _ _ - . . . - - -

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3.10 Feedwater System 3.10.1 Purpose The purpose of this test was to verify the adequacy of the Feedwater level Control System for power uprate conditions. nis test demonstrated  ;

adequate. system performance fellowing setpoint step changes while [

operating in nroe Element Mode and Single Element Mode of operation.

3.10.2 Criteria ,

level 1 The restonse of any level related variable to any test input changes, or i disturbance, must not diverge during the setpoint changes, that is, the-Decay Ratios are less than or equal to 1.0.

level 2 he Level Control System related variables may contain oscillatory response. However, during setpoint changes the Decay Ratios for each controlled mode of response must be less than or equal to 0.25.

3.10.3 Results ,

De Feedwater level Control System test was performed in Test Condition "C", 96% CTP (100% of original rate power) and in Test Condition "D",98% CTP.

In Test Condition "C",5 inch down and up setpoint step changes were made in Three Element Mode and Single Element Mode of operations. All level control related variables were highly damped and no decay ratios exceeded 0.25.

In Test Condition "D",5 inch down and up setpoint step changes were made in Three Element Mode and Single Element Mode of operations.

Likewise, all level control related variables were highly damped and no decay ratios exceeded 0.25.

i 25 i

= . - - .

y 3.11 Turbine Valve Surveillance 3.11.1 Purpose ne purpose of this ten was to detennine the maximum power level at which the periodic testing of the Main Arbine Control Valves (TCV) could be performed with acceptable system response for power uprate conditions.

3.11.2 Criteria level 1 l None Level 2

a. Peak neutron flux must be at least 7.5% below the scram trip setting.

b. Peak vessel pressure must remain at least 10 psi below the high pressure scram setting,

c. Peak steam flow in the high flow line must re 9ain at least 10%

below the high flow isolation trip setting,

d. All Turbine Control Valves must be less than 100% open at all times when another Tuitine Control Valve is stroking or is closed,

c. Peak heat flux must remain at least 5.0% below its scram trip setting.

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3.11.3 Results

'the Turbine Control Valves (TCVs) were modified during the refueling outage at the recommendation of the Turbine Manufacturer. '!he modification was intended to ensure that sufficient flow capacity would be available for uprated power conditions. By increasing the existing valve capacity, the TCV surveillance could also be performed at a higher power level, ne TCVs were routinely tested at 90% CTP before power uprate or 86% CTP. He TCVs were satisfactorily tested at 88% CTP, approximately 2 % above the previous limit. All level 2 test criteria were -

met at this point, and from a nuclear safety standpolrt tids test could have been performed at a higher power level. However because of the problems encountered with the Turbine Control Valw '.t is prudent not to attempt to continue te. sting at higher powers.

At 88% CTP, the peak neutron flux was 91.3%, which is less than the 110.5% limit, ne peak vessel pressure was 1018.4 psig, which is iess ,

i than the 1083 psig limit, ne peak steam line flow was 3.5 Mlbm/hr, which is less than the 4.57 Mlbm/hr limit. The TCV maximum lift was  :

88.2 % which was less than 100 % opening. The peak heat flux was 87.5 %

which was less than the 94.6% limit.

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.s, 4

3.12 Recirculation System Nw Calibration 3.12.1 Purpose

'Ihe maximum core flow for the power uprate condition was maintained at 105% flow. Increased core pwer for the same core flow affected the Recirculation Pump speed / core flow relationship. 'Ibe purpose of this test was to document and conectly sequence the performtnce of 56.000.02, l

" Core Nw Calibratloa" and 54.000.20 " Reactor Recirculation System MG Set Scoop Tube Pcultioner Operability Test".

3.12.2 Criteria I.evel 2

'Ihe core flow calibration is completed and meets the acceptance criteria ,

of the test procedure 56.000.02, " Core Nw Calibration".  ;

'Ihe Reactor Recirculation Pump speed limits are set and meets the l acceptance criteria of the surveillance test procedure 54.000.20," Reactor l Recirculation System MG Set Scoop Tube Positlow Operability Test".

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4 sn --p - . . ,,m.,,,w,.,,---.--.-.-_e y,---+ - , , . . , -_ ,.,.,,.-,.-,w.n-cy,, ,..y,,,.y.m.s-- .,,-,-u,,,.,-., , ,,e-, , g

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4-3.12.3 Results Core Flow Calibration The core flow calibration was a prerequisite for cetting the Reactor Recirculation Motor-Generator Set speed stops. The core flow calibration was performed at 98% CTP, During this test, it wa*

necessary to recalibrate jet pamps 1-10 loop "B" flow indicator, .

The total rated drive flow was determined to be 30.877 Mlbm/hr with an M Ratio of 2.237. The new Recirculation Flow Unit and loop Summer Composite Gain Adjustment Factors ( CGAPs) were determined.

Reactor Ru:ltculation Speed Stop Eadngs At 98 % CTP, the Reactor Recirculation pump speed stop setpoints were detennined. Minoy adjustments were made to the mechanical and electrical speed stops for both Reactor Recirculation Motor-Generator sets.

l.

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sh '

3.13 Steady State Data Collection

-- 3.13.1 - Purpose h purpose of this test was to :

' ~

1. Collect and evaluate steady state plant data for BOP systems.

2. Monitor potentially limiting systems or parameters which were chosen based on experience and engineering evaluation.

3. Perfonn core thermal limit projections to provide additional assurance that core thermal limit margin would be maintained for uprate power conditions.

3.13.2 Criteria Level 1 Average Drywell temperature is less than 145 F.

De Main Steam Tunnel Area Temperature is less than 200 F.

The Turbine Building Steam hnnel Area Temperature is less than 200 F.

Reactor Power Stability is confirmed by verifying that APRM noise level is less than 10% peak to peak and LPRM noise level is less than 30% peak to peak.

Turbine Control Valve positions at 100% CTP are greater than 75% open.

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.< b' 3.13.2 Criteria 4

I.evel 2 Projected Average Daywell Temperature at 100% CTP will be less 1 . thar 145 F. .

Projected Main Steam Tunnel Area Temp =tures at 100% CTP will be less than 200 F.

Projected 'Ibrbine Building Steam Tunnel Area Temperatures at 100% CTP v!!! be less than 200 F.

Projected Mdn Steam line Radiation Monitor readings at 100%

CTP will be greater than 1000 mrem /hr for Detectors A,B and C and 867 mn:m/hr for Detector D.

Projected Isophase Bus Duct Temperatures at 100% CTF will be less than 125 C.

Projceted Heater Drain Pump Flow at 100% CTP will be between 4.29 and 4.37 Mlbm/hr.

No excessive Recirculation Pump vibration Projected Turbine Control Valve Positions at 100% CTP will be greater than 88% open.

Projected values of MAPRAT, MFLPD and MFLCPR are less than 0.970.

Ievel None The Net Generation and Unit Heatrate are within the expected range.

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3.13.3 Results i A. BOP performance Net Generation and Gross Heat Rate values were within the expected ranges for uprated power conditions and therefore met the acceptance criteria. In addition, approximately 70 BOP perfonnance parameters were collected and evaluated for Test Conditions "A", "B" (86% CTP), "C" (96% CTP), and "D" (98%

CTP). The parameters were consistent with projection of historical values.

B. Potentially limiting or marginal system / components Parameters were recorded and projected during power ascension.

All Level I criteria were met and projections made for Level 2 criteria were acceptable except for Turbine Control Valves performance and Heater Drains Flow.

The manufacturer recommended modification of the TCVs to assure adequate flow margin. TCV lift vs. flow characteristics, which are used in the Tbrbine trip transient analysis, were changed by this modification. This test was intended to confirm the minimum valve lift at 100% CTP. Early observation up to 96%

CTP confirmed predicted characteristics. At 98% CTP, divergence from the predicted curve was observed and served as a precaution for funher power increases. When attempting to increase from 98% CTP to 100% CTP excessive TCV lift and large valve modulation (10% peak to peak) was observed at 99% CTP which precluded funher power increase. This problem remains under investigation.

Heater Drains system flows were projected to 100% CTP, The projected value was 4.09 Mlbm/hr which is less than the Heat Balance value of 4.31 Mlbm/hr. Although the actual flowrate is lower than the Heat Balance, the actual flowrate is consistent when compared to the projections of historical flowrate values and therefore is considered satisfactory.

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g8 '

3.13.3 Results (continuation)

B. Potentially limiting or marginal system / components (continuation)

Although area temperatures were satisfactory, this test was not performed during the most limiting meteorological conditions and must be qualified. Area temperatures are dependent on seasonal variations with the more limiting condition being the summer months. Programs are in place which continue to monitor area temperatures, equipir. cat and room coolers and will identify these conditions, when and if they exist, so that appropriate action can be taken.

C. Core Thermal Limits Fermi 2 is equipped with the Genen! Electric (GE) 3-D Monicore system which is capable of predicting core thermal limits. The system was used to predict MAPRAT, MFLPD and MFLCPR before operating at uprated conditions. Core thermal limits were projected in Test Condition "C" (96% CTP) and in Test Condition "D" (98 % CTP) to the next Test Condition. The projected thermal limits were all less than 0.900, which meets the Level 2 acceptance criteria.

?

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m N'

S 3.14 Main Steamline Radiation Measurement 3.14.1 Purpose ne purpose of this test was to verify that the current Main Steam Line Radiation Monitor (MSLRM) setpoints are acceptable for power uprate conditions.

3.14.2 Criteria I.evel 1 The MSLRM nominal trip setpoint (NTSP) is less than or equal to 3.6 times the full rated power background radiation levels.

Level 2 The MSLRM nominal trip setpoint (NTSP) is less than or equal to 3.0 times the full rated power background radiation levels.

3.14.3 Results The background readings for MSLRM "A","B","C" and 'D"_ wen-1.07E3 mr/hr, 1.08E3 mr/hr, 1.15E3 mr/hr and 0.932E3 mr/hr, respectively. Three times the background readings for MSLRM "A","B","C" and "D" are 3.21E3 mr/hr, 3.24E3 mr/hr, 3.45E3 mr/hr and 2.79E3 mr/hr, respectively. The current NTSPs for MSLRM "A",

"B", 'C" and "D" are 3.0E3 mr/hr, 3.0E3 mr/hr, 3.0E3 mr/hr and 2.6E3 mr/hr, respectively, which are less than three tiraes the background readings and therefore, meets the Level 2 acceptance criteria.

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