ML060680583

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Response to NRC Round 3 Requests for Additional Information Related to Technical Specification Change No. TS-418 - Request for Extended Power Uprate Operation
ML060680583
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 03/07/2006
From: Crouch W
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC3743, TAC MC3744, TS-418, TVA-BFN-418
Download: ML060680583 (28)
v • d • e


Text

t TVA-BFN-418 March 7, 2006 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Mail Stop:

OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:

In the Matter of

)

Docket Nos. 50-260 Tennessee Valley Authority

)

50-296 BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 -

RESPONSE TO NRC ROUND 3 REQUESTS FOR ADDITIONAL INFORMATION RELATED TO TECHNICAL SPECIFICATIONS (TS) CHANGE NO. TS-418 -

REQUEST FOR EXTENDED POWER UPRATE OPERATION (TAC NOS. MC3743 AND MC3744)

This letter provides TVA's response to the NRC Staff's request for additional information, which was submitted to TVA by letter dated December 22, 2005 (ADAMS Accession No. ML053560177), in order to support review of the BFN Units 2 and 3 Extended Power Uprate (EPU) license amendment application.

TVA submitted the BFN Units 2 and 3 EPU application to the NRC by letter dated June 25, 2004 (ML041840301).

TVA supplemented that application by letters dated February 23, 2005 (ML050560337), April 25, 2005 (ML051170242),

June 6, 2005 (ML051640391), and February 28, 2006.

Enclosure 1 to this letter provides TVA's responses to the NRC requests. to this letter contains revised responses to five of the requests answered in TVA letter dated December 19, 2005 (ML053560186).

U.S. Nuclear Regulatory Commission Page 2 March 7, 2006 Some of the information in Enclosure 1 is proprietary to General Electric Nuclear Energy (GENE).

GENE requests that the proprietary information in the enclosure be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4),

10 CFR 2.390(a)(4), and 10 CFR 2.390(b)(1).

An affidavit supporting this request is included in Enclosure 1.

Additionally, some of the information in Enclosure 1 is proprietary to Framatone ANP (FANP).

FANP requests that the proprietary information in the enclosure be withheld from public disclosure.

An affidavit supporting this request is included in Enclosure 1.

A non-proprietary version of this response is contained in.

During preparation and final review of this submittal, a legacy error was discovered in the existing design calculation which determines the available Emergency Core Cooling System pump net positive suction head requirements.

The error has been documented in BFN's Corrective Action Program, and the calculation is presently being revised.

The effect of the error is small; however, it impacts numerical values that were provided in the original EPU submittal and in the February 28, 2006 submittal. Additionally, the error impacts values that are needed to respond to questions ACVB.17, ACVB.18, ACVB.26, and ACVB.32.

Therefore, the responses to these questions with the corrected information will be provided in a separate letter by March 24, 2006.

This issue was discussed with Margaret Chernoff on March 6, 2006.

U.S. Nuclear Regulatory Commission Page 3 March 7, 2006 There are no new regulatory commitments associated with this submittal.

If you have any questions concerning this letter, please contact me at (256) 729-2636.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 7th day of March, 2006.

Sincerely, q4A,. Li William D. Crouch Manager of Licensing and Industry Affairs cc:

See page 5.

U.S. Nuclear Regulatory Commission Page 4 March 7, 2006

Enclosures:

1. Response To December 22, 2005, NRC Round 3 Requests For Additional Information Related To Technical Specifications (TS) Change No. TS-418 -

Request For Extended Power Uprate Operation (Proprietary Version)

2.

Revised Responses To TVA Submittal Dated December 19,

2005, Related To Technical Specifications (TS) Change No. TS-418 -

Request For Extended Power Uprate Operation

3.

EPU Power Ascension Test Plan

4. May 23, 1975 -

Final Summary Report, Unit 2 Startup, Browns Ferry Nuclear Plant

5.

May 9, 1977 -

Final Summary Report, Unit 3 Startup, Browns Ferry Nuclear Plant

6.

RS-001 Revised Template Safety Evaluation

7.

Copies of Material Provided To U.S. Fish and Wildlife Service

8. Response To NRC December 22,
2005, Round 3 Requests For Additional Information Related To Technical Specifications (TS) Change No. TS-418 -

Request For Extended Power Uprate Operation (Non-Proprietary Version)

U.S. Nuclear Regulatory Commission Page 5 March 7, 2006 cc (w. Enclosures):

State Health Officer Alabama Department of Public Health RSA Tower -

Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Malcolm T. Widmann, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 Margaret Chernoff, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739

U.S. Nuclear Regulatory Commission Page 6 March 7, 2006 JEM:TLE:BAB Cc: (w/o Enclosures):

B. M. Aukland, POB 2C-BFN M. Bajestani, NAB lA-C A. S. Bhatnagar, LP 6A-C J. C. Fornicola, LP 6A-C R. G. Jones, POB 2C-BFN R. F. Marks, Jr., PAB 1A-BFN G. W. Morris, LP 4G-C B. J. O'Grady, PAB 1E-BFN K. W. Singer, LP 6A-C E. J. Vigluicci, ET 11A-K NSRB Support, LP 5M-C EDMS, WT CA-K, w. Enclosures S:lic/submit/subs/Response to EPU RAI 3 for U23.doc

I ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 REVISED RESPONSES TO TVA SUBMITTAL DATED DECEMBER 19, 2005 RELATED TO TECHNICAL SPECIFICATIONS (TS) CHANGE NO. TS-418 -

REQUEST FOR EXTENDED POWER UPRATE OPERATION NRC Request EMCB-A.1 Section 10.7, Plant Life, in Enclosure 4 of the June 25, 2004, submittal, identifies irradiation-assisted stress-corrosion cracking (IASCC) as a degradation mechanism influenced by increases in neutron fluence and reactor coolant flow.

This section indicates that the current inspection strategy for reactor internal components is expected to be adequate to manage any potential effects of EPU operating conditions.

Note 1 in Matrix 1 of Section 2.1 of RS-001, Revision 0 indicates that guidance on the neutron irradiation-related threshold for IASCC in boiling-water reactors (BWRs) is in Boiling-Water Reactor Vessel and Internals Program (BWRVIP) report BWRVIP-26.

The "Final License Renewal SER

[Safety Evaluation Report] for BWRVIP-26," dated December 7, 2000, states that the threshold fluence level for IASCC is 5 x 1020 n/cm2 (E > 1 MeV).

Identify the vessel internal components whose fluence, at the end of period of operation with the EPU operating conditions will exceed the threshold level and become susceptible to cracking due to IASCC.

For each vessel internals component that exceeds the IASCC threshold, either provide an analysis that demonstrates failure of the component will not result in the loss of the intended function of the reactor internals or identify the inspection program to be utilized to manage IASCC of the component.

Identify the scope, sample size, inspection method, frequency of examination and acceptance criteria for the inspection programs.

After review of the response to this request, the NRC informally noted "The staff has determined that a more detailed response to the original question is required regarding the top guide and core plate holddown bolts.

Because these two components exceed the threshold of 5x102 0 n/cm2, TVA is requested to identify the scope, sample size, inspection method, frequency of examination and acceptance criteria for the inspection programs of the top guide and core plate holddown bolts for BF, Units 1, 2, and 3.

E2-1

The staff requests that TVA provide these additional details as they are not provided in the BWRVIP documents."

TVA Reply to EMCB-A.1 The requested information is provided in Enclosure 1 of this letter by the reply to EMCB-A.4.

Additional information regarding the core plate holddown bolts is provided in Enclosure 1 of this letter by the reply to EMCB-A.3.

This response supplements the original response.

NRC Request SPLB-B.1 Discuss whether any administrative controls or fire protection responsibilities of plant personnel are affected by an increase in decay heat.

Also, address why an increase in decay heat will not result in an increase in the potential for a radiological release from a fire.

After review of the response to this request, the staff informally noted "Still needs to address why the EPU does not affect the elements of their fire protection program related to the fire protection responsibilities of plant personnel."

TVA Reply to SPLB-B.1 Administrative controls and fire protection responsibilities of plant personnel in the Technical Specifications, the Technical Requirements Manual, the Nuclear Quality Assurance Plan, and the Fire Protection Program were reviewed for affects associated with the increase in decay heat.

There are no administrative controls or fire protection responsibilities of plant personnel affected by an increase in decay heat associated with EPU.

As indicated by the results of the Appendix R analyses, all Appendix R acceptance criteria are met under EPU; therefore, there is no increase in the potential for a radiological release resulting from a fire.

This response replaces the original response.

E2-2

NRC Request SPLB-B.2 Section 6.7.1, of Enclosure 4 of the June 28, 2004, submittal states that:

... a plant-specific evaluation was performed to demonstrate safe shutdown capability in compliance with the requirements of 10 CFR 50 Appendix R assuming EPU conditions....

The results of the Appendix R evaluation for EPU provided in Table 6-5 demonstrate that fuel cladding integrity, reactor vessel integrity, and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions.

Upon reviewing Table 6-5, BFN Appendix R Fire Event Evaluation Results, the NRC staff was able to find references for all but the following values in the EPU submittal:

  • Cladding Heatup (PCT)), degrees F = 1428 (EPU)
  • Suppression Pool Bulk Temperature, degrees F = 227 (EPU),
  • 227 (Appendix R Criteria), including Note 3

13.6 (EPU)

Provide references, including appropriate extracts from the UFSAR, plant-specific Appendix R evaluation, etc., for these values in Table 6-5, including Note 4.

After review of the response to this request, the staff informally requested "If the referenced, but not provided, extracts from BFN Calculation MDN-0999-980113, App. R FP Evaluation, indicate the same PUSAR Table 6-5 values cited in the RAI for, then TVA should provide the extracts as copies or as quotes."

TVA Reply to SPLB-B.2 BFN calculation MDN0999980113, "Appendix R Fire Protection Evaluation," documents the EPU evaluation on compliance with the requirements of 10 CFR 50 Appendix R which was performed in Project Task Report T0611, "Appendix R Fire Protection." The limiting EPU PCT occurs for Case 1 and is presented in Project Task Report T0611 Section 3.3.1, "Key Results," Item 1 as 14280F. The EPU suppression pool bulk temperature is the same for Cases 1, 2, and 3 and is presented in Project Task Report T0611 Section 3.3.1, "Key Results," Items 4, 9, and 14 as 2270F.

The torus attached piping limit for EPU is the suppression pool bulk temperature of 2270 F and is the temperature used in the analyses for the torus attached piping for EPU Appendix R E2-3

conditions.

The primary containment pressure is the same for Cases 1, 2, and 3 and is presented in Project Task Report T0611 Section 3.3.1, "Key Results," Items 5, 10, and 15 as 13.6 psig.

This response replaces the original response.

NRC Request SPSB-A.12 Explain how the impact of increasing the ultimate heat sink temperature from 91 to 95 degrees F has been incorporated into the PRA.

Which PRA basic events are affected by this change?

Following an EPU PRA Audit in January of 2006, the NRC informally noted that this is a question for Units 2 and 3, but the answer is for Unit 1.

TVA Reply to SPSB-A.12 The previous reply provided to this request indicated that the response applied to Unit 1. The response is relevant to all three units and is corrected below.

The change in UHS temperature from 91 to 95 degrees F has no effect on the PRA model.

Engineering analysis has shown that systems and components perform their functions with the higher value.

The PRA does depend on MAAP analyses for some success criteria and post core damage behavior.

A MAAP model was developed and verified. An examination of the parameter file shows that parameter TWSW, the RHR (LPCI) heat exchangers service water inlet temperature, is set to 95 degrees Fahrenheit.

This response replaces the original response.

NRC Request SPSB-A.17 Address the questions in the SRP, Chapter 19, Table III-1 concerning low power and shutdown PRA.

Following an EPU PRA Audit in January of 2006, the NRC informally noted that TVA did not answer this question as requested.

TVA Reply to SPSB-A.17 Shutdown safety is maintained and monitored by compliance with work in accordance with the outage schedule/plan.

An assessment is performed of the outage schedule/plan implementation prior to the outage and, during the execution of the schedule/plan, anytime the outage schedule/plan is affected. These assessments E2-4

are performed using the EPRI sponsored program called Outage Risk Assessment Management (ORAM).

ORAM is a computer program that receives data from the scheduling software and performs deterministic risk assessments during reactor shutdowns and outages.

The implementation of the program is controlled by procedures and includes the ORAM software that takes the status (i.e., available, unavailable) of key plant equipment, evaluates the current/planned plant condition(s) against approved data models, and then produces an output of the relative level of safety/defense in depth of key shutdown functions:

  • Inventory Control,
  • Electrical Power Availability,
  • Reactivity Control,
  • Fuel Pool Cooling, and
  • Primary/Secondary Containment.

The program includes a structured approach to determine the effect of outage activities upon the key shutdown safety functions by assessing the following:

  • Identify key safety functions affected by the Structure, System, and Component (SSC) planned for removal from service
  • Consider the degree to which removing the SSC from service will affect the key safety functions
  • Consider degree of redundancy, duration of out-of-service condition, and appropriate compensatory measures, contingencies, or protective actions that could be taken if appropriate for the activity under consideration.

An integral part of an outage schedule/plan is the contingency plan.

This is an approved plan for compensatory actions:

  • To maintain Defense in Depth by alternate means when outage planning reveals that specific SSCs will not be available
  • To restore Defense in Depth when systems availability drops below previously established levels during the outage
  • To minimize the likelihood of the loss of key safety functions during higher risk evolutions The Shutdown Risk Assessment Program also includes a detailed review of the outage schedule/plan (including review of changes) by a multi-discipline team with extensive experience in the operation and maintenance activities at BFN.

This activity provides another level of assurance that shutdown safety issues E2-5

are addressed and all reasonable actions have been taken to minimize shutdown risk.

The review considers, for example:

  • Technical Specification Requirements
  • The degree of redundancy available for performance of the key safety functions served by out-of-service SSCs
  • The duration of the activity
  • The likelihood of an initiating event or accident that would require the performance of the affected safety function
  • The likelihood that the activity will increase the frequency of an initiating event requiring key safety functions
  • Component and system dependencies that are affected
  • Performance issues for the in service redundant SSCs
  • The risk impact of performing the maintenance during shutdown with respect to performing the maintenance at power Another important feature of the BFN shutdown risk program is the inherent flexibility that is provided by the structure of the program.

Calculations are prepared that provide BFN specific information into the program.

The calculations address any chances produced by the operating history of BFN prior to the outage, including power levels and durations.

These calculations provide input into functional parameters such as the availability of systems and support systems required to provide reactor vessel makeup water consistent with the decay heat generation load and availability of alternate sources of reactor vessel makeup water consistent with the decay heat generation rate.

This work also provides input regarding times associated with the reactor vessel and fuel pool boil down rates.

This information also provides insights for determining operator response times as an integral part of this pre-outage work.

Shutdown events include the following major categories:

  • Fuel assembly insertion
  • Inadvertent opening of a relief valve
  • E2-6

Total loss of off-site power Startup of idle recirculation pump

  • Accident Fuel-handling *
  • Special Event Loss of habitability of the control room
  • These postulated event impacts and associated mitigating SSCs (including operator actions) are not affected by the implementation of Extended Power Uprate (EPU).

BFN operation at a higher power level will not affect cool water effects, SSCs performance regarding operational capability, environmental heat load, or interfere with operator actions designed to assist with the mitigation of these postulated events.

For the remaining three events, BFN operation at EPU conditions will have a very minor affect regarding mitigation of the postulated events during shutdown conditions.

  • Total loss of off-site power
  • Loss of shutdown cooling For these events, EPU operation does not affect equipment reliability, availability, initiating event frequency, and mitigation approach including equipment utilized for mitigation.

The effect of EPU operation on the success criteria is similar to the effect on the at power PRA success criteria.

However, because the reactor has been shutdown for some period of time, the decay heat load is substantially lower than the at power values.

This situation results in boil down times that are much longer than the values associated with the at power conditions.

These conditions result in small or no changes in the success criteria of systems associated with mitigation of events postulated during shutdown conditions at BFN.

There is an effect on mitigation associated with decay heat load and the resulting effect on operation actions.

The BFN use of ORAM appropriately addresses this aspect regarding event mitigation by including calculations that are cycle specific and address previous operating power levels and associated durations.

These calculations provide input into functional parameters such as E2-7

the availability of systems and support systems required to provide reactor vessel makeup water consistent with the decay heat generation load and availability of alternate sources of reactor vessel makeup water consistent with the decay heat generation rate.

This information reflects operator action response times also.

The reduction for these operator action times due to EPU operation is shown to be less than 15%

(depending on the time after shutdown).

These small changes in already relatively lengthy operator response times result in negligible changes in human action probabilistic values.

BFN plans to continue the use of ORAM as a tool to provide for continued structured program associated with the outages schedule/plan.

Using this information, the following SRP questions are answered.

  • Does the application introduce new initiating events or change the frequencies of existing events?

No new initiating events or increased potential for initiating events during shutdown are postulated due to the proposed EPU.

  • Does the application affect the scheduling of outage activities?

No.

BFN operation at EPU conditions will not change the outage sequence of operations to accomplish shutdown activities.

Decay heat loads will increase due to EPU conditions but this minimal effect will be anticipated and appropriately planned for by the ORAM pre-outage schedule/plan.

  • Does the application affect the ability of the operator to respond to shutdown events?

No.

The BFN use of ORAM appropriately addresses this aspect regarding event mitigation by including calculations that are cycle specific and address previous operating power levels and associated durations.

These calculations provide input into functional parameters such as the availability of systems and support systems required to provide reactor vessel makeup water consistent with the decay heat generation load and availability of alternate sources of reactor vessel makeup water consistent with the decay heat generation rate.

This information reflects operator action response times E2-8

also.

The reduction for these operator action times due to EPU operation is shown to be less than 15% (depending on the time after shutdown).

These small changes in already relatively lengthy operator response times result in negligible changes in human action probabilistic values.

  • Does the application affect the reliability or availability of equipment used for shutdown conditions?

No.

For these events, EPU operation does not affect equipment reliability, availability, initiating event frequency, and mitigation approach including equipment utilized for mitigation.

  • Does the application affect the availability of equipment or instrumentation used for contingency plans?

No.

Consistent with the situation associated with the unaffected reliability or availability of equipment used for shutdown conditions, equipment and instrumentation reliability, availability, initiating event frequency, and mitigation approach associated with the contingency plan will not be effected by EPU operation.

This response replaces the original response.

E2-9

4 ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 EPU POWER ASCENSION TEST PLAN Table 1 contains a listing of the currently planned modifications necessary to support EPU that require testing during power ascension. A description of each activity and the planned testing is provided.

Required post modification testing that will be performed prior to power ascension in accordance with the plant design change process is also provided in Table 1. Modifications that are required for EPU that are not tested during power ascension are not listed.

Setpoint adjustments, including those required for Unit 1 due to the steam dome pressure increase, that are tested by standard plant procedures such as required Technical Specifications surveillance tests are not listed.

Table 2 contains a list of planned power ascension tests that are required to specifically address EPU implementation.

EPU testing performed by standard plant procedures as a part of normal startup testing are not listed.

Table 3 describes the BFN EPU Power Ascension Test Plan.

The modifications and testing activities in Tables 1, 2, and 3 represent the currently planned post modification tests and power ascension test activities.

Details of some testing activities may be modified based on further evaluation.

E3-1

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Tst Main Turbine

  • Replace HP Turbine
  • Turbine diaphragms and rotor balancing (if buckets required)
  • Replace HP Rotor/LP Overspeed test Rotors (Unit 1 only)

Control and stop

  • Replace springs, valve testing bonnets, washers, Relief valve bellows, & bolting on testing six cross-around relief bench testing valves to permit increased set pressure
  • Replace miter bend elbows in the condenser spray piping with long radius elbows to reduce back pressure Turbine
  • Modify the size of the
  • Monitor steam Sealing Steam steam seal unloader seal header valves and associated pressure piping to allow the of

~* Calibration o turbine sealing system the steam seal to accommodate the he presse larger steam flow contolle requirements

  • Inservice leak test E3-2

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Condensate Pumps

  • Replace 2 impellers in each of 3 pumps
  • Install 3 -

1250 hp motors

  • Recalibrate relay settings
  • Recalibrate/replace pump

& motor instrumentation

  • Modify HVAC ductwork
  • Verification of pump flow and head
  • Monitoring of pump and motor parameters (flow, pressure, temperatures, etc.)
  • Instrumentation calibration and functional testing
  • Condensate trip test Pump Condensate Booster Pumps
  • Replace 3 pumps
  • Install 3 -

3000 hp motors

  • Recalibrate relay settings
  • Recalibrate/replace pump

& motor instrumentation

  • Modify HVAC ductwork
  • Verification of pump flow and head
  • Monitoring of pump and motor parameters (flow, pressure, temperatures, etc.)
  • Instrumentation calibration and functional testing
  • Condensate Booster Pump trip test E3-3

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. : :.;: i Feedwater Replace 3 pumps Balancing Pumps and Turbines Recalibrate pump Overspeed instrumentation and testing control system for controls tuning increased flows at EPU condiions*

Verification of pump flow and

  • Replace turbine/pump head coupling Monitoring of
  • Replace turbine rotor, pump and turbine diaphragms and buckets parameters
  • Realiraterepace(flow, pressure,
  • Recatemperaturesc turbine instrumentation etc.)rtues
  • Instrumentation calibration and functional testing
  • Feedwater Pump trip test Moisture Change vanes and add Moisture removal Separators perforated plate on effectiveness moisture separators testing
  • Modify internal drains Inservice leak as needed test
  • Performance monitoring (flow, pressure, temperatures, etc.)

E3-4

Feedwater Heaters

  • Upgrade heater shell pressure certification
  • Replace level transmitters on FWHs 2 & 3
  • Performance monitoring (flow, pressure, temperatures, etc.)
  • Instrumentation calibration and functional testing
  • Inservice leak test 1,
  • Repair/replace 18 nozzles on FWHs 1, 2 & 3
  • Replace relief valves on FWHs 1, 2 & 3
  • Install new impingement plate & steam duct inside FWH 3
  • Reinforce / re-weld partition plates in FWHs pass all
  • Install 1 new vessel with valves & digital controls
  • Upgrade controls on 9 existing vessels to digital (Unit 2 only)
  • Install digital control on 9 existing vessels (Unit 1 only)
  • Replace valves for increased reliability
  • Control system functional testing
  • Initial installation startup test (flow, pressure, temperatures, etc.)

1 ___________________________

E3-5

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Steam Dryer

  • Modify dryer to ensure
  • Determine structural integrity at moisture EPU conditions carryover
  • Applicable Recirculation protection system instrumentation Pump Motors setpoints calibrations
  • Revise temperature Vibration monitoring setpoints monitoring
  • Assess additional heat Controls tuning load on plant 1-VAC &

and system cooling water systems operation during

  • Assess power cable vse yr voltage drop increase due to higher current
  • Revise pump/motor vibration monitoring setpoints
  • Re-rate pumps and motors for 120% power/105% core flow operating conditions E3-6

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I V I Main Generator System 1

Recalibrate/replace pressure regulators pressure switches and

  • Increase generator hydrogen to 75 psig to operate at increased loads
  • Field installation testing
  • Instrumentation calibration and functional testing
  • Monitoring of system parameters (voltage, amps, temperatures, etc.) during power ascension
  • Rewind generator stator and generator field (Unit 1 only)

Isolation

  • Modify Isolation Phase
  • Verification of Phase Bus Duct Bus Duct Cooling System system flow, Cooling to remove Bus Duct heat both air and under EPU conditions water Main Bank
  • Replace due to
  • Performance Transformers obsolescence issues.

monitoring The Unit 1, Unit 2, and Unit 1/2 spare transformers are in place and operating at this time.

The Unit 3 transformers are currently scheduled to be replaced in 2010 along with the installation of a dedicated spare Unit 3 transformer.

Vibration

  • Install temporary
  • Collect and Monitoring sensors based on ongoing analyze analyses vibration data
  • Conduct testing program on selected during power ascension E3-7

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Main Steam Replace MSIV poppets and

  • Performance Isolation modify operators (Unit 1 monitoring Valves only) as required to reduce differential pressure across MSIVs at EPU conditions
  • Install 2-inch MSIV stems as required due to increased stem forces caused by EPU MS flow increase EHC Software New program inputs &

Verification of logic for EPU conditions control functions

  • Turbine Valve setup
  • Controls Tuning Steam /

Increased flow rate to

  • Monitor to Feedwater accommodate increased ensure plant Normal Flow reactor thermal power remains within Rate Increase output anticipated operational limits Recirculation Increased required
  • Verification of Pump Flow Rate recirculation pump flow total core flow Increase rate required to achieve total core flow E3-8

STP 1 Chemical and Radiochemical Sampling and measurements selected power levels to determine 1) the chemical and radiochemical quality of reactor water and reactor feedwater and

2) gaseous release.

STP 2 Radiation Gamma dose rate measurements and Measurements where appropriate, neutron dose rate measurements at specific limiting locations throughout the plant to assess the impact of the uprate on actual plant area dose rates.

STP 10 IRM After the APRM calibration for Calibration EPU, the IRM gains will be adjusted as necessary to assure the IRM overlap with the APRMs.

This will be done during first controlled shutdown following APRM calibration for EPU.

STP 17 System Due to the 30 psi reactor (Unit 1 only)

Expansion pressure increase (and associated temperature increase), system expansion checks will be made for major equipment and piping in the nuclear steam supply system during heatup to assure components are free to move as designed and adjustments will be made as necessary for freedom of movement.

E3-9

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T~est f-Bye~t-Dsrition STP 19 Core Core performance parameters Performance (LHGR, APLHGR, and MCPR) will be calculated to verify they remain within limits as part of a careful, monitored approach to the EPU power level.

STP 20 Electrical Demonstrate that the plant net Output and electrical output and net heat Heat Rate rate requirements are satisfied.

STP 22 Pressure Evaluate pressure control system Regulator response to pressure setpoint testing.

STP 23 Feedwater Adjust the Feedwater Control Control System for acceptable reactor System water level control.

Demonstrate the capability to prevent a low reactor water level scram following the trip of a single condensate pump, condensate booster pump, or feedwater pump.

STP 92 Steam Determine steam separator-dryer Separator-moisture carryover.

Dryer E3-10

Main Turbine Table 1 x

I lx I I I I I I I I I I I I I Ixixixixix Turbine Sealing Table 1 Performance monitoring from 0% to EPU System Table 1 X

_Performancemonitoringfrom_0%_toEPU Condensate Table 1 X

Performance monitoring from 0% to EPU Pumps______________________________________

Booesater upsTable I X

Performance monitoring from 0% to EPU Feedwater Pumps and Table 1 X

Performance monitoring from 0% to EPU Turbines SMoisture Table x

x x x x x x Feedwater Table 1 X

x x

x x x x x

Heaters I___

Condensate Tal Demineralizers Table 1 X

Steam Dryer Table 1 X XXX X

Reactor Recirc TbeI XX X

Pump Motors Table 1 X

X X X X

Main Generator Table 1 X

Performance monitoring from 25% to EPU E3-11

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Isolation Phase Bus Duct Table 1 Performance monitoring from 25% to EPU Cooling Main Bank Tal Transformers Table 1 XX XX X

VibrationTal Monitoring Table 1 X

X XXXX X

Main Steam Tal Isolation Valves Table I x

x EHC Software Table 1 X

X X

X X

X X X X X X Steam/Feedwat er Normal Flow Table 1 x x x x Rate Recirculation Pump Flow Table l x

x I

x x

Rate Chemical and Tal2 Radiochemical Table2 XXXX X

RadiationTal2 Measurements Table 2 XXXX X

IRM Calibration Table 2 (Not a startup test. Will be performed during the first controlled shutdown following APRM calibration for

_E P U.)

E3-12

Ep.

Md Irs pEgt ePU TesT CONDIIN iECNQ 32X MW(OTP (AliowgrE e -

(Al1

+0%I E xpansion (Unit lTable 2 lX l X<<

P~eorfeormance Table 2

_X X XX X

OButput and TTable X

IL

<I X

XE Rregu lartor Table 2 X

X X

X X

X XX XX X Ceoentrol Seystem Table 2

_-X X XX X

Seepaarator-Dryer Table 2

___X X XX X

Line items may have multiple tests.

every power level indicated.

Each test will not necessarily be performed at E3-13

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