RBG-45900, Cycle 11 Startup Report

Cycle 11 Startup Report

ML020430050
Person / Time
Site:	River Bend
Issue date:	01/21/2002
From:	King R Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G9.25.1.4, G9.5, RBF1-01-0280, RBG-45900
Download: ML020430050 (23)

Text

Entergy Operations, Inc.

River Bend Station "5485 U.S. Highway 61 P. 0. Box 220 St. Francisville, LA 70775 Tel 225 336 6225 Fax 225 635 5068 Rick J. King Director Nuclear Safety Assurance January 21, 2002 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

River Bend Station - Unit 1 Docket No. 50-458 Cycle 11 Startup Report File No.:

G9.5, G9.25.1.4 RBF1-01-0280 RBG-45900 Ladies and Gentlemen:

In accordance with River Bend Station (RBS) Technical Requirements Manual TR 5.6.8, enclosed is a Startup Report that provides a summary of the startup physics testing conducted on the Cycle 11 core reload and the second phase of the Power Uprate implementing the pressure increase.

If you have any questions, please contact Barry M. Burmeister at (225) 381-4148.

Sincerely, RJK/bmb Enclosure cc:

NRC Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011

LA Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. 0. Box 82215 Baton Rouge, LA 70884-2215 Attn: Prosanta Chowdhury Mr. David J. Wrona NRR Project Manager U.S. Nuclear Regulatory Commission Mail Stop OWFN/7D-1 Washington, DC 20555

ENCLOSURE River Bend Station Unit 1 Cycle 11 Startup Physics Test Summary And Power Uprate - Pressure Increase Phase

Enclosure Cycle 11 Startup Report Page 1 of 21 Startup Physics Overview River Bend Station (RBS) resumed commercial operation in Cycle 11 on October 10, 2001 following a Refueling/Maintenance Outage. The Cycle 11 reload consisted of replacing 123 GEl 1 and 77 GE8B General Electric fuel assemblies with 200 Atrium 10 Framatome fuel assemblies. The following startup tests were performed during Refueling Outage RF10 or while attaining full power after RF 10, and are summarized in this report:

1)

Core Loading Verification

2)

Control Rod Functional Testing

3)

Shutdown Margin Determination

4)

TIP Asymmetry In addition to the above startup physics tests, the startup test program included: Core Monitoring System Verification and Recirculation System Calibration, as well as other surveillance testing required by RBS Technical Specifications. The additional test results are available at the site on request.

Enclosure Cycle 11 Startup Report Page 2 of 21 Core Loading Verification Purpose Ensure each reactor fuel assembly is:

in its correct core location oriented properly seated properly in its support piece Criteria The reactor core is visually checked to verify conformance to the vendor supplied core loading pattern. Fuel assembly serial numbers, orientations, and core locations are recorded. A height check is performed to verify all assemblies are properly seated.

Results The as-loaded core was verified for proper fuel assembly serial numbers, locations, orientation and seating in accordance with the RBS Cycle 11 core loading pattern. There were no location or orientation deviations from the Cycle 11 core loading pattern.

The core verification procedure was successfully completed on October 5, 2001.

Control Rod Functional Testing Purpose Verify functionality of each control rod by:

performing normal withdrawals and insertions ensuring it is latched to its control rod drive assuring that it moves at design speeds without excessive friction

Enclosure Cycle i1 Startup Report Page 3 of 21 Criteria Testing of each control rod is performed to ensure proper operability. This testing includes coupling verification, friction testing where required, and scram time testing.

Results Friction testing was not required for any control rods based on the channel management analysis for BOC 11. Fifteen Control Rod Drive Mechanisms were replaced during the outage. They were scram time tested prior to startup during the vessel hydro in accordance with Technical Specification Surveillance Requirements 3.1.4.3 and 3.1.4.4 with satisfactory results.

The remaining control rods were scram time tested during the vessel hydro testing in accordance with RBS Technical Specification Surveillance Requirement 3.1.4.1. All of the control rod scram times were within the allowable limits.

A control rod coupling check was performed in accordance with RBS Technical Specification Surveillance Requirement 3.1.3.5 each time a control rod was fully withdrawn.

Enclosure Cycle 11 Startup Report Page 4 of 21 Shutdown Margin Determination / Reactivity Anomaly Check Purpose To ensure that:

"* the reactor can be made sub-critical from all operating conditions

"* the reactivity transients associated with postulated accident conditions are controllable within acceptable limits

"* the reactor will be maintained sufficiently sub-critical to preclude inadvertent critically in the shutdown condition Criteria The sub-critical demonstration verifies the reactor remains sub-critical with the analytically determined strongest worth control rod fully withdrawn.

The in-sequence rod withdrawal shutdown margin calculation begins by withdrawing control rods in their standard sequence until criticality is achieved. The shutdown margin of the core is determined from calculations based on the critical rod pattern, the reactor period and the moderator temperature.

Reactivity Anomaly verification is performed in accordance with Technical Specification Surveillance Requirement 3.1.2.1 after reaching equilibrium xenon concentrations at 100%

reactor power.

Results The sub-critical demonstration was performed on October 7, 2001.

The in-sequence critical shutdown margin surveillance procedure was completed on October 10, 2001. The shutdown margin (SDM) at the beginning-of-cycle (BOC) was calculated to be 2.5286 % delta k/k. The Cycle 10 "R" value is equal to 0.530% delta k/k. Therefore, the Cycle 11 minimum shutdown margin at the most reactive point in the fuel cycle is 1.9986 % delta k/k which is bounded by the RBS Technical Specification 3.1.1 requirement of 0.38% delta k/k.

Final steady state full power operation was achieved on October 22, 2001.

It was verified on October 22, 2001 that no reactivity anomaly was present by performance of Technical Specification Surveillance Requirement 3.1.2.1.

Enclosure Cycle 11 Startup Report Page 5 of 21 Tip Asymmetry Check Purpose To determine the reproducibility of the Traversing Incore Probe (TIP) system readings Criteria An asymmetry determination is performed as part of a detailed statistical uncertainty evaluation of the TIP System. A complete set of TIP data is obtained at steady state conditions while greater than 75% rated core thermal power.

The calculated x2 of the symmetric integral TIP measurement differences should not exceed the critical 2. (calculated j2 < critical x2)

Results The TIP reproducibility and symmetry uncertainty calculations were performed on November 12, 2001 at 100% core thermal power.

The critical X2is 29.14 and the calculated X2 is 15.31.

Enclosure Cycle II Startup Report Page 6 of 21 River Bend Station Reactor Dome Pressure Increase Startup Test Report Table of Contents Executive Summary 7

1. Purpose 8

2. Uprate Power Ascension Program Scope 8

2.1 Program Development 8

2.2 Prerequisites to Power Ascension Testing 9

2.3 Uprate Power Ascension Testing 10 2.4 Test Acceptance Criteria 11

3. Summary of Uprate Testing and Equipment Performance Results 14 3.1 Key Events 14 3.2 Testing and Equipment Performance Results 14 3.3 Exceptions 16

4. Application of the FSAR Initial Startup Test Program to the Power Uprate Project 16 4.1 General Discussion 16 4.2 Construction Tests and Equipment Demonstrations 17 4.3 Preoperational Tests and Operational Demonstrations 17 4.4 Startup Tests and Operational Demonstrations 18

Enclosure Cycle 11 Startup Report Page 7 of 21 Executive Summary This report is submitted to the Nuclear Regulatory Commission (NRC) in accordance with the requirements of the River Bend Station Technical Requirements Manual Section TR 5.6.8.

The Power Escalation Test Program, performed by Entergy Operations Inc. at River Bend Station, implements the testing and equipment performance monitoring commitments contained within Licensing Topical Report, "Generic Guidelines for General Electric BWR Generic Power Uprate," NEDC-31897P-A, Class III, May 1992 (LTR-1), and the letter from EOI to the USNRC dated August 1, 1999, "Request for License Amendment for Power Uprate Operation."

This report details the results of the second phase of the River Bend Station 5% Power Uprate.

The first phase of the River Bend Station Power Uprate Project was a zero reactor pressure increase (no increase in the reactor operating pressure) power uprate. This, second phase involved only an increase in reactor dome pressure, and no increase in core thermal power.

Consequently, there was no significant change in feedwater or steam mass flow as a result of this phase. The pressure increase was accomplished during the power ascension following Refueling Outage 10.

During the first phase of Power Uprate implementation, in October 2000, the Reactor Pressure Control System and the Reactor Level and Feedwater Control Systems were dynamically tested.

This was done to ensure that the increase in power and feedwater and steam mass flow rates did not adversely impact the proper function of these systems. During this second phase there were no significant increases in mass flow rates therefore, further dynamic testing of these systems was not performed.

Power ascension, including dome pressure increase began on October 11, 2001, and was completed October 25, 2001. Pressure was increased in 5 psi increments until the new maximum reactor dome pressure of 1070 psia was reached. After reaching the new maximum dome pressure, a Maximum Dependability Capacity Test was completed. The uprate power ascension test program was successfully completed with all acceptance criteria being satisfied. All equipment and system performance was as expected.

Enclosure Cycle 11 Startup Report Page 8 of 21 River Bend Station Uprate Power Ascension Startup Test Report

1. Purpose This Power Uprate Startup Test Report is submitted to the Nuclear Regulatory Commission pursuant to Section TR 5.6.8, Startup Report, which requires:

"* "A summary report of plant startup and power escalation shall be submitted following... (2)

Amendment to the license involving a planned increase in power level."...

"* "The report shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications."

"* "Any corrective actions that where required to obtain satisfactory operation shall also be described."

"* "Any additional specific details required in license conditions based on other commitments shall also be included in this report."

• "Startup reports shall be submitted within 90 days following completion of the startup test program...

2. Uprate Power Ascension Program Scope 2.1 Program Development The River Bend Station Power Ascension Test Program was developed in accordance with the generic guidelines provided in Licensing Topical Report (LTR) NEDC-31897P-A, Generic Guidelines for General Electric Boiling Water Reactor Power Uprate, the License Amendment Request including the Safety Analysis Report and various Power Uprate Project Task Reports. The power ascension test program included testing or equipment monitoring recommendations from many Task Reports. According to section 5.11.9 of NEDC-31897P-A, Power Uprate Testing, "Large transient tests (e.g., isolation) will not be required for uprates within 5% power. Initial plant testing and experience during plant operation is considered to be sufficient." Consequently no large transients were included within the River Bend Station Uprate Power Ascension Test Program, including this second, pressure increase phase.

Enclosure Cycle II Startup Report Page 9 of 21 The Uprate Power Ascension Test Program was developed to verify the following:

"* Plant systems and equipment affected by power uprate are operating within design limits.

"* Nuclear fuel thermal limits are maintained within expected margins.

"* The response of the main steam pressure control system is stable, with adequate control margin to allow for anticipated transients.

"* The response of the reactor water level control system is stable, with adequate control margin to allow for anticipated transients.

"* The response of the reactor core flow control system is stable and bi-stable core flow is within acceptable limits.

"* The feedwater heater drains and level control system is stable.

"* The MSR drains and level control system is stable.

"* Reliable system operation is maintained.

"* Radiation levels are acceptable and stable.

2.2 Prerequisites to Power/Pressure Ascension Testing Prior to the commencement of power/pressure ascension testing, the test procedure required the completion of numerous activities, which included:

"* The applicable plant operating procedures, administrative procedures, surveillance test procedures, calibration procedures, chemical and radiological procedures and other similar procedures were reviewed and revised as required.

"* Computer software programs were reviewed and revised as required to support the power uprate test program.

"* The applicable plant instrumentation setpoint changes or recalibrations were completed.

"* All plant modifications were reviewed to assure they were completed as required and had no exception which could affect the uprate test program.

"* Temporary Modifications logs and GL91-18 applicable degraded conditions were reviewed to assure there was no impact on the ability of the effected equipment to support uprate, and that uprate would not have an adverse impact on any existing degraded condition.

"* Baseline data was taken as required by the procedure, prior to exceeding the old maximum reactor dome pressure limit of 1040 psia.

"* Commitments which were the result of the Power Uprate Safety Analysis Report (SAR),

Power Uprate License Amendment, the NRC Power Uprate Safety Evaluation (SE), and actions resulting from Power Uprate project Task Report review, were verified as either closed, included in the power ascension program or evaluated as not impacting power ascension.

Enclosure Cycle 11 Startup Report Page 10 of 21 2.3 Uprate Power/Pressure Ascension Testing Power Ascension was performed in accordance with a River Bend Station Special Test Procedure. Operator Training and Infrequently Performed Test or Evolution (IPTE) briefings were completed prior commencing the pressure increase beyond the old limit of 1040 psia. Additionally, shift briefings were held for each Operations shift during the implementation period, to ensure adequate communication of current plant conditions and anticipated plant response.

Power ascension occurred such that pressure was increased in 5 psi increments, until 100%

core thermal power was achieved. Then pressure was increased, while holding power constant, until the new reactor dome pressure limit of 1070 psia was reached. During this power/pressure increase each increment included a period of data collection and evaluation.

Following each power increase, testing and equipment performance data was collected and evaluated in accordance with established acceptance criteria. At each incremental step in power ascension, the following activities were performed:

"* Core Thermal Performance data evaluated.

"* Reactor pressure control system stability, steam flows limit cycling, and variation in incremental regulation performance data evaluated.

"* Reactor water level control and the variation in incremental regulation performance data evaluated.

"* Electro-Hydraulic Control (EHC) System oil pressure to the Turbine control valve oscillation data evaluated.

"* Feedwater heater level control performance data evaluated.

"* MSR drain system level control performance was evaluated.

"* Bistable reactor recirculation flow data evaluated.

"* Reactor Recirculation Core flow / Drive flow relationship was evaluated.

"* A complete set of equipment performance data (e.g., control room readings, local readings, process computer, and Emergency Response Information System (ERIS) computer data) was collected, evaluated and predicted performance at the next power/pressure increment determined.

After pressure increase to 1070 psia was completed, a maximum dependable capacity test was performed.

Enclosure Cycle 11 Startup Report Page 11 of 21 2.4 Test Acceptance Criteria General Discussion The development of the power uprate test recommendations and acceptance criteria was based on the review of similar test programs performed at other plants, Chapter 14 of the River Bend Station Final Safety Analysis Report (FSAR), the outputs of the Uprated NSSS heat balance, and numerous RBS specific GENE Task Reports. The River Bend Station original Startup Test Program, Regulatory Guide 1.68 and LTR 31987 P-A were also used as inputs.

Following each step increase in reactor dome pressure, test data was evaluated against its performance acceptance criteria (i.e., design predictions or predictions which resulted from extrapolations of actual plant performance). If the test data satisfied the acceptance criteria then system and component performance were determined to comply with their design requirements.

Plant parameters during power ascension were evaluated with two levels of acceptance criteria. The criteria associated with safe and reliable plant operation are classified as Level

1. The criteria associated with performance expectations, either derived from design or actual performance history, are classified as Level 2. The following paragraphs describe the actions required to be taken if an individual criterion is not satisfied.

Level 1 Acceptance Criteria Level 1 acceptance criteria normally relate to the values of process variables assigned in the design of the plant, component systems or associated equipment. If a Level 1 test criterion is not satisfied, the plant must be placed in a hold condition that is judged to be satisfactory and safe, based upon prior testing. Plant operating or test procedures or the Technical Specifications may guide the decision on the direction to be taken. Tests consistent with this hold condition may be continued. Resolution of the problem must be immediately pursued by equipment adjustments or through engineering evaluation as appropriate.

Following resolution, the applicable test portion must be repeated to verify that the Level 1 requirement is satisfied. A description of the problem must be included in the report documenting successful completion of the test.

Level 1 acceptance criteria for power ascension included requirements that reactor feedwater flow, reactor water level, reactor pressure and other reactor systems are expected to exhibit stable full power operating characteristics. This Level 1 acceptance criterion of requiring all plant systems to exhibit normal high power level operating behavior (i.e.,

Enclosure Cycle 11 Startup Report Page 12 of 21 stable reactor water level control, and feedwater flow, with acceptable limit cycling if any) is to assure that that this testing can be performed with acceptable risk.

Level 2 Acceptance Criteria Equipment Performance If a Level 2 test criterion is not satisfied, plant operating or test plans would not necessarily be altered. The limits stated in this category are usually associated with expectations of system transient performance whose characteristics can be improved by equipment adjustments. An investigation of the related adjustments, as well as the measurement and analysis methods would be initiated.

If all Level 2 requirements in a test are ultimately met, there is no need to document a temporary failure in the test report; unless there is a lessons learned benefit involved.

Following resolution of temporary Level 2 test criterion failures, the applicable test portion must be repeated to verify that the Level 2 requirement is satisfied.

For the River Bend Station Power Uprate, specific Level 2 acceptance criteria were established as detailed in the following paragraphs.

EHC/Reactor Pressure Control The pressure increase phase of the River Bend power uprate resulted in no significant additional mass flow. Additionally, since the pressure increase resulted in increased main turbine control valve margin, due to the decreased specific volume of main steam, no dynamic testing on the reactor pressure control system was performed. The system was observed to ensure acceptable performance after each pressure increase increment.

Reactor Water Level and Feedwater (FW) Control The pressure increase phase of the River Bend power uprate resulted in no significant additional mass flow. Acceptable feedwater system dynamic response was verified during the first phase (power and flow increase phase) by performing system dynamic testing. No dynamic testing on the Feedwater and Reactor Level Control System was performed during this, second phase. The system was observed to ensure acceptable performance after each pressure increase increment.

Enclosure Cycle 11 Startup Report Page 13 of 21 Generator Stator Temperatures The maximum allowable RTD temperature limit is 168 degrees F. These temperatures are monitored by TAMARIS, an automatic temperature monitoring system located in the main control room. This system is equipped with alarming capabilities, should any temperatures approach the maximum temperature limit. No significant increases in generator temperatures were anticipated (gross generation was predicted to increase by only 2-4 Mwe) as a result of the pressure increase phase, and no control room TAMARIS alarms were experienced during the pressure increase phase.

Bistable Reactor Recirculation Flow The current plant limits associated with bistable flow were maintained.

Turbine Stop, Control and CIV Valve Testing The absolute power levels at which RBS Turbine Stop, Control and CIV tests will be performed was not changed for Power Uprate. The margins for neutron flux trip, and bypass valve open events were not affected, and these tests were not performed.

Enclosure Cycle 11 Startup Report Page 14 of 21

3. Summary of Uprate Testing and Equipment Performance Results 3.1 Key Events Power Ascension Chronological Sequence of Events No.

Event Description Date 1

Authorization granted to commence uprate power ascension testing 10-11-01 2

Place MSRs into Optimized configuration 10-13-01 3

Perform testing at 1025 psig Reactor Dome Pressure 10-13-01 4

Perform testing at 1030 psig Reactor Dome Pressure 10-13-01 5

Perform testing at 1035 psig Reactor Dome Pressure 10-13-01 6

Perform testing at 1040 psig Reactor Dome Pressure 10-13-01 7

Perform testing at 1045 psig Reactor Dome Pressure 10-13-01 8

Perform testing at 1050 psig Reactor Dome Pressure 10-13-01 9

Perform testing at 1055 psig Reactor Dome Pressure 10-13-01 10 Place Leading Edge Flowmeter into service and ascend to 3039 10-25-01 MWt, corrected.

11 Complete MDC test 11-28-01 3.2 Testing and Equipment Performance Results Control Systems Performance Results Control Systems most affected by uprate were monitored to assure acceptable performance and compliance with their specific Level I and 2 acceptance criteria. The following table summarizes these control systems.

Control System Performance Results Level 1 Level 2 Tuning No.

Control System Description Acceptance Acceptance Adjustments Criteria Criteria Required 1

Reactor Water Level Control System Satisfied Satisfied No 2

EHC and Reactor Pressure Control System Satisfied Satisfied No 3

Feedwater Heater Level Control Satisfied Satisfied No System 4

Rx. Recirculation and Bi-Stable Flow Satisfied Satisfied No

Enclosure Cycle 11 Startup Report Page 15 of 21 Equipment Performance Results The following systems and selected equipment most affected by uprate within these systems were closely monitored to assure that equipment performed as predicted and that they operated within their design requirements.

Equipment Performance Results Level 1 Level 2 Predicted No.

System Description

Acceptance Acceptance Performance Criteria Criteria Results 1

Condensate System Satisfied Satisfied Acceptable 2

Feedwater System Satisfied Satisfied Acceptable 3

Heater Drain System Satisfied Satisfied Acceptable 4

MSR Drain System Satisfied Satisfied Acceptable 5

Main Generator and Alternator Satisfied Satisfied Acceptable 6

Nuclear Boiler Satisfied Satisfied Acceptable 7

Reactor Recirculation System Satisfied Satisfied Acceptable 8

Main Turbine Satisfied Satisfied Acceptable 10 Main Transformer Satisfied Satisfied Acceptable 11 Stator Cooling System Satisfied Satisfied Acceptable 12 Isophase Bus Cooling Satisfied Satisfied Acceptable 13 TPCCW System (CCS)

Satisfied Satisfied Acceptable Reactor and Core Performance Results

1. Core thermal hydraulic parameters were verified to be within Technical Specification limits.

2. Margins to fuel thermal limits were verified to be acceptable.

3. Reactor Recirculation flow (drive flow) was in accordance with predictions for each power plateau.

4. Reactor operation was stable with no discernable change in reactor performance from pre uprate full power operating conditions. The core operated in a manner consistent with predicted expectations.

Enclosure Cycle 11 Startup Report Page 16 of 21 Radiation and Chemistry Results Radiation surveys of selected areas within the plant were performed before and after the pressure increase at 100% core thermal power. As predicted, no significant change in plant radiation levels were experienced as a result of the pressure increase.

Chemistry monitoring (reactor water, condensate water and off gas) continued throughout the uprate power/pressure ascension test program with no discemable change from prior full power operating conditions Net Gross Electrical Output Gain From Pressure Increase Phase of Power Uprate The net electrical output increased 3 MWE as a result of increasing reactor dome pressure from 1040 psia to 1070 psia, as indicated by MDC test results before and after the pressure increase. The pre-pressure increase value was obtained from an MDC test performed immediately after the first, flow-only phase of power uprate was implemented, in October, 2000.

3.3 Exceptions Equipment and Test Exceptions None. All Level 1 and 2 acceptance criteria were satisfied and equipment and system performance behaved in accordance with predicted expectations.

4. Application of the USAR Initial Startup Test Program to the Power Uprate Project 4.1 General Discussion The River Bend Station Safety Analysis Report section 10.4, Required Testing requires "This report will include... brief discussions as to why it was not necessary to repeat specific tests listed in USAR Section 14, during the power uprate test program." This section of the Uprate Startup Test addresses this requirement with respect to the Power Uprate Project. The USAR Section 14 addresses the River Bend Station initial startup test program. The initial startup test program was divided into three main parts. They are:

Construction test and Equipment Demonstrations, Preoperational and System Demonstrations, and Startup Tests and Operational Demonstrations. Each of these programs is discussed in the following paragraphs with respect to the River Bend Station Power Uprate Project.

Enclosure Cycle 11 Startup Report Page 17 of 21 4.2 Construction Tests and Equipment Demonstrations Construction tests (safety related) are those tests, which demonstrate that safety-related equipment meets functional operability requirements. These tests cover a wide variety of checks to assure that components are properly installed and adjusted according to manufacturers instructions, Architect Engineering drawings and specifications, satisfy code requirements comply with FSAR requirements, etc. They include but are not limited to test such as: hydrostatic pressure tests. Electrical megger tests, load tests, cleanliness inspections, rotational tests, alignment tests, etc.

Equipment demonstrations (non-safety-related) are those tests used to demonstrate that non-safety-related equipment meets functional operability performance requirements.

As applies to the Power Uprate, this category of test demonstration is conducted as part of the modification process. These tests, where required, are included in the modification (administrative procedure) package. Required Post Modification Tests for Power Uprate modifications were successfully completed as part of the modification closure process.

4.3 Preoperational Tests and Operational Demonstrations Preoperational test (safety-related) are those tests conducted prior to fuel loading to demonstrate that the plant has been properly designed and constructed, and that the safety related structures, systems and components meet safety-related performance requirements.

System demonstrations (non-safety-related) consist of those tests conducted to demonstrate that non-safety-related system and components function as required to meet normal plant operating requirements.

This category of test demonstration is conducted as part of the post modification testing process.

Power Uprate modifications were successfully completed as part of the modification closure process.

Enclosure Cycle 11 Startup Report Page 18 of 21 4.4 Startup Tests and Operational Demonstrations USAR Requirements Startup Tests are safety-related tests and consist of such activities as fuel loading, pre critical tests, critical and low power tests and power ascension tests that ensure fuel loading in a safe manner, confirm the design bases, demonstrate where practical that the plant is capable of withstanding the anticipated transients and postulated accidents, and ensure that the plant is safely brought to rated capacity and sustained power operation.

River Bend Station Power Uprate Startup Program Development The following method as described in the next two paragraphs was used in establishing uprate testing requirements.

The development of the power uprate test recommendations and acceptance criteria is based on the review of similar test programs performed at other plants, Chapter 14 of the River Bend Station USAR, the outputs of the NSSS heat balance and power flow map tasks, and the River Bend Station Startup Tests. From the total population of tests identified in the preceding programs, a set of tests were selected for further evaluation and incorporation into the River Bend Station uprate test program. The effect of the power uprate at River Bend Station on the operational parameters, performance characteristics and acceptance criteria of these tests were examined. If the test was potentially impacted by power uprate, it was then evaluated for applicability and inclusion within the River Bend Station Uprate Power Ascension Test Program. This evaluation resulted in a final set of test recommendations to be performed during the initial ascension and operation at full 105%

uprated power.

The recommendations are the result of a test selection process that is based upon a review of the original startup test program and changes resulting from the power uprate of the River Bend Station plant. The tests and equipment performance monitoring included in these recommendations fall into the following categories:

a. tests involving control systems with specific performance expectations assumed in the power uprate transient analyses and specific performance expectations for operational considerations,

b. tests affected by power uprate

c. tests required based on engineering judgement, and

d. performance monitoring of equipment impacted by power uprate

Enclosure Cycle 11 Startup Report Page 19 of 21 In general, most of these tests can be satisfied by completion of existing surveillance or functional tests, performance of instrumentation calibration and equipment setup, evaluation of the results of post modification testing, or through steady state data collection as part of normal system monitoring."

Transient Testing As applies to Power Uprate and allowed by the SAR, system transient and control system dynamic response testing to demonstrate acceptable system performance was performed as a part of the phase 1, flow-only Power Uprate Power Ascension Testing. All test data was reviewed to assure compliance with the acceptance criteria for power ascension testing for uprate affected equipment. No transient testing was performed during the phase 2, pressure increase portion of Power Uprate.

Comparison of Power Uprate Tests to USAR Power Ascension Tests As required by the SAR, the following Table addresses each of the initial power ascension tests and their applicability to the River Bend Station Uprate Power Ascension Test Program. Tests identified with a yes were incorporated in either phase 1 or phase 2 of the River Bend Station Uprate Test program unless credit was taken for another activity (i.e.,

surveillance test), that satisfies the requirement.

Enclosure Cycle 11 Startup Report Page 20 of 21 Results of FSAR Initial Startup Testing Evaluation For Inclusion In The Uprate Power Ascension Test Program Test No.

Required In Acceptance Start-up Power Ascension Test Description Uprate Test Criteria Procedure(l)

Same as FSAR 101 Chemical and Radiochemical Yes (2)

Yes 102 Radiation Measurements Yes Yes 103 Fuel Loading No NA 104 Full Core Shutdown Margin No NA 105 Control Rod Drive System No NA 106 SRM Performance and Control Rod Sequence No NA 107 Water Level Measurements No NA 108 Intermediate Range Monitor Performance No NA 109 Local Power Range Monitor Calibration No NA 110 Average Power Range Monitor Calibration Yes Yes 111 Process Computer Yes Yes 112 Reactor Core Isolation Cooling System No NA 113 Selected Process Temperatures Yes Yes 114 System Expansion No NA 115 Core Power Distribution Yes Yes 116 Core Performance Yes Yes 117 Steam Production Yes Yes 118 Core Power-Void Mode Response No NA 119 Pressure Regulator Yes Yes 120 Feedwater Control System Yes Yes 121 Turbine Valve Surveillance No NA 122 Main Steam Isolation Valves No NA 123 Relief Valves No NA 124 Turbine Stop Valve Trips and Generator Load Rejections No NA 125 Shutdown From Outside The Control Room No NA 126 Recirculation Flow Control System No NA 127 Recirculation System No NA 128 Loss Of Turbine Generator and Offsite Power No NA 129 Deleted NA NA 130 Vibration Measurements No NA 131 Deleted NA NA 132 Recirculation System Flow Calibrations No NA 133 Reactor Water Cleanup System No NA 134 Residual Heat Removal System No NA 135 Control Rod Sequence Exchange No NA 136 Drywell Piping Vibrations No NA 137 Off-Gas System No NA Note (1) From G.E.N.E. Task Report, Startup Test Recommendations, Testing Required (2) Credit Taken For Surveillance Monitoring Program