ML061800088

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Technical Specifications Change TS-431 Request for Extended Power Uprate Operation
ML061800088
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 05/16/2006
From: Crouch W
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Brown Eva, NRR/DLPM, 415-2315
References
TAC MC3812, TVA-BFN-TS-431
Download: ML061800088 (42)
v • d • e


Text

May 16, 2006 TVA-BFN-TS-431 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 No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE TS-431 - SUPPLEMENTAL RESPONSE REGARDING EXTENDED POWER UPRATE - TESTING PLANNED TO SATISFY DRAFT STANDARD REVIEW PLAN 14.2.1 (TAC NO. MC3812)

This letter provides TVA's supplemental responses to the NRC request for additional information regarding TVA's determination that the criteria in draft Standard Review Plan (SRP)

Section 14.2.1, Generic Guidelines for Extended Power Uprate (EPU) Testing Programs, is satisfied by TVA's planned testing programs.

On June 28, 2004 (Reference 1), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. contained a description of the Unit 1 EPU Start-up Test Program and justification for not performing large transient testing. The NRC's December 22, 2005 letter (Reference 2) requested additional information regarding TVA's test program for Unit 1. TVA and NRC met on April 20th to discuss the overall testing issue.

As part of the April 20th meeting, TVA was requested to submit the information presented at the meeting and a description of the program that will compare the results from Unit 1 restart testing with the previous operating experience on Units 2 and 3.

The requested information is provided in Enclosure I to this letter.

U.S. Nuclear Regulatory Commission Page 2 May 16, 2006 The SRP 14.2.1 acceptance criteria for a loss of offsite power, turbine trip or generator trip, or the isolation of all main steam lines will be satisfied as part of currently planned restart testing and activities. TVA will evaluate the results of the tests performed to satisfy the SRP acceptance, criteria against Technical Specification requirements, test acceptance criteria, and expected responses based on applicable analyses, Units 2 and 3 operating experience, and responses predicted by the BFN simulators. Based on this information, large transient testing would not provide additional useful information.

A summary of the commitments made in this letter is provided in .

TVA has determined that the additional information provided by this letter does not affect the no significant hazards considerations associated with the proposed TS changes. The proposed TS change still qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c) (9).

If you have any questions about this submittal, please contact me at (256) 729-2636. I declare under penalty of perjury that the foregoing is true and correct. Executed on May 16, 2006.

Sincerely, Original signed by:

illiam D. Crouch Manager of Licensing and Industry Affairs

References:

1. TVA letter, dated June 28, 2004, "Browns Ferry Nuclear Plant (BFN)*- Unit 1 - Proposed Technical Specifications (TS) Change TS-431 - Request for License Amendment -

Extended Power Uprate (EPU) Operation."

U.S. Nuclear Regulatory Commission Page 3 May 16, 2006

2. NRC letter, dated December 22, 2005, "Browns Ferry Nuclear Plant, Unit 1 - Request for Additional Information for Extended Power Uprate (TS-431) (TAC-No. MC3812)."

cc (Enclosures):

State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017

U.S. Nuclear Regulatory Commission Page 4 May 16, 2006 Enclosures cc (Enclosures):

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 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 Eva A. Brown, 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 5 May 16, 2006 DTL: SMK:BAB Enclosures cc (w/Enclosures):

B. M. Aukland, POB 2C-BFN M. Bajestani, NAB 1A-BFN A. S. Bhatnagar, LP 6A-C J. C. Fornicola, LP 6A-C D. F. Helms, BR 4T-C R. F. Marks, PAB 1C-BFN R. G. Jones, POB 2C-BFN B. J. O'Grady, PAB 1E-BFN K. W. Singer, LP 6A-C E. J. Vigluicci, ET IIA-K NSRB Support, LP 5M-C EDMS WT CA-K s:lic/submit/TechSpec/TS 431 RAI

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA)

BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATION (TS) CHANGE 431, SUPPLEMENTAL RESPONSE REGARDING EXTENDED POWER UPRATE -

TESTING PLANNED TO SATISFY DRAFT STANDARD REVIEW PLAN 14.2.1 I. BACKGROUND On June 28, 2004 (Reference 1), TVA requested a TS change to allow Unit 1 to operate at extended power uprate (EPU) conditions. Enclosure 8 contained a description of the Unit 1 EPU Start-up Test Program and justification for not performing large transient testing.

The NRC's December 22, 2005 letter (Reference 2) requested additional information regarding TVA's test program for Unit 1 and how Standard Review Plan (SRP) 14.2.1 criteria were addressed by the proposed EPU testing. TVA responded on March 7, 2006 (Reference 3).

Subsequent to that meeting, TVA prepared a summary of the process used by TVA in identifying the system testing requirements for the restart of each BFN unit, outlined how the Unit 1 testing is staged to support Unit 1 restart milestones, and compared the testing with regulatory guidance to ensure compliance.

TVA and NRC met on April 2Oth to discuss the overall testing issue. In that meeting, TVA was requested to submit the information presented at the meeting as well as a description of the program that will compare the results from Unit 1 restart testing with the previous operating experience on Units 2 and 3.

The requested information is contained in this enclosure.

II. REGULATORY GUIDANCE Guidelines for the recommended integrated plant testing for Extended Power Uprates is contained in Licensing Topical Report (NEDC-32424P-A, Appendix L), which states:

  • Large transient tests are not required for uprates of up to ten percent of rated power. The initial plant testing and plant operational experience are considered sufficient; El-1
  • The licensee should perform a Main Steam Isolation Valve (MSIV) closure test if the uprate is more than ten percent above previous recorded data or provide an explanation and justification in a licensing submittal for any different approach; and
  • The licensee should perform a Generator Load Reject test if the uprate is more than 15% above previous recorded data or provide an explanation and justification in a licensing submittal for any different approach.

Addition regulatory guidance is provided in draft Standard Review Plan (SRP) 14.2.1, Generic Guidelines for Extended Power Uprate Testing Programs. The SRP test acceptance criteria for a loss of offsite power, turbine trip or generator trip, and the closure of all Main Steam Isolation Valves (MSIVs) is reproduced in Table 1.

III. PROCESS FOR IDENTIFYING BFN SYSTEM TESTING REQUIREMENTS The safe shutdown requirements are identified in the BFN Safe Shutdown Analysis (SSA). The SSA is documented as a calculation for BFN Units 1, 2, and 3. The SSA is based on the Updated Final Safety Analysis Report (UFSAR) Chapter 14 requirements and identified the accidents, abnormal operational transients, and special events from which the plant must be able to achieve safe shutdown. The SSA also identified the system modes or functions required to achieve safe shutdown for the applicable operating states. A tabulation of the accidents, transients, and special events within the scope of the SSA is provided in Table 2.

For each system, a Baseline Test Requirement Document (BTRD) specifies the testing acceptance criteria for each mode or function. An example of the contents of the BTRDs for the Main Steam system is provided below:

SYSTEM AND MODE MODE DESCRIPTION ACCEPTANCE CRITERIA Main Provide Main Turbine trip Verify that position switches Steam on Turbine Stop Valve less transmit the correct signal to 001-01 than 90% open trip signal RPS when the corresponding to Reactor Protection Turbine Stop Valve are less than System (RPS). 90% open.

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SYSTEM AND MODE MODE DESCRIPTION ACCEPTANCE CRITERIA Main Close Main Steam Isolation 1. Verify each MSIV closes in 3 Steam Valves (MSIVs) on primary to 5 seconds upon receipt of 001-02 containment isolation a closure signal.

signal. 2. Verify each MSIV closes upon loss of both AC and DC solenoid power. Verify each valve closes on loss of air pressure.

3. Verify when only the 120VAC circuit or the 250VDC circuit are disconnected from the MSIV, the valve remains open.
4. Verify each MSIV accumulator leakage rate is within its design value.
5. Verify closure of either MSIV in one steam line transmits the correct signal to the Reactor Protection System.

Main Open Main Turbine bypass Test under Mode 047-05.

Steam valves on Turbine Control 001-05 System (047) turbine trip signal.

Main Open Safety Relief Valves 1. Verify actuation of Main Steam (SRVs) on high Reactor Control Room (MCR) manual 001-07 pressure to provide switches energize the Reactor Pressure Vessel correct SRV solenoid with (RPV) pressure relief the associated transfer switch in "normal".

2. Verify control of the non-Automatic Depressurization System (ADS) SRVs is not possible from the MCR when transfer switches are in "Emergency".
3. Verify SRV mechanical lift pressures.

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SYSTEM AND MODE MODE DESCRIPTION ACCEPTANCE CRITERIA Main Automatic opening of Verify the ADS SRVs will, as Steam Automatic Depressurization demonstrated by energizing the 001-08 System (ADS) Safety Relief solenoid, actuate automatically Valves (SRVs) upon upon receipt of the appropriate coincident signals for 2 combinations of each of the Core Spray pumps or 1 following signals:

Residual Heat Removal 1. Drywell pressure high (RHR) pumps running and (including ADS timers),

either Reactor Pressure and Vessel (RPV) low water 2. RPV water level (Level 1 level (Level 1), high and Level 3), and Drywell pressure and time 3. 1 Low Pressure Coolant delay, or Level 1 and high Injection (LPCI) pump or 2 Drywell pressure bypass Core Spray pumps running, time delay. and

4. High Drywell pressure bypass timer has elapsed.

No new system functions or modes were identified as a result of taking Unit 1 from the current licensed power level of Units 2 and 3 (105% of the Original Licensed Thermal Power level or 3,458 MWt) to EPU conditions (120% of the Original Licensed Thermal Power level or 3,952 MWt).

IV. UNIT 1 TESTING AND MONITORING The testing that will be performed as part of the restart of BFN Unit 1 can generally be categorized and sequenced as follows:

  • Post-Modification / Post-Maintenance Testing,
  • Testing Leading to Fuel Load,
  • Testing Leading to Start-up (Restart Testing),
  • Testing Leading to 100% Original Licensed Thermal Power (3,293 MWt) (Power Ascension Testing), and EPU Testing and Performance Monitoring.

Examples of these types of tests and how the testing is staged to support Unit 1 restart milestones is provided in Figure 1.

Additional details for each category is discussed below.

POST-MODIFICATION / POST-MAINTENANCE TESTING Post-Modification and Post-Maintenance testing is performed after the physical work for each individual Design Change Notice or Maintenance Work Order has been completed in the field.

Examples of this testing would include valve operability testing, system and component leak testing, and logic system El-4

functional testing. This-testing is required to be completed before the Design Change Notice or Maintenance Work Order is considered complete.

TESTING LEADING TO FUEL LOAD Prior to fuel load, the BFN restart testing program includes system testing and testing which is performed to demonstrate system operability in accordance with Technical Specification requirements for having fuel in the vessel (Mode 5). Examples of this testing would include Residual Heat Removal and Core Spray logic system functional testing.

TESTING LEADING TO START-UP (RESTART TESTING)

Restart testing includes additional system testing and testing which is performed to demonstrate system operability in accordance with Technical Specification requirements for start-up and power operations (Modes 2 and 1). Examples of this testing would include Recirculation flow control testing, control rod timing and testing, and primary containment leak rate testing.

A chronology of the development of the BFN Restart Test Program (RTP), from Unit 2 restart, Unit 3 restart, and now for Unit 1 is contained in Reference 4. The primary purpose of the RTP is to verify that systems are capable of meeting their safe shutdown requirements. Test procedures are developed to satisfy the required testing identified in the BTRDs and to verify that the acceptance criteria is met.

TESTING LEADING TO 100% ORIGINAL LICENSED THERMAL POWER (3,293 MWT) (POWER ASCENSION TESTING)

Power Ascension Testing includes additional Balance of Plant system testing. Examples of this testing would include shutdown margin testing, Reactor Feedwater Pump overspeed testing and tuning, and High Pressure Coolant Injection / Reactor Core Isolation Cooling cold quick start and vessel injection testing.

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EXTENDED POWER UPRATE TESTING AND PERFORMANCE MONITORING Extended Power Uprate testing includes additional Balance of Plant system testing. Examples of this testing would include Main Turbine, Condensate and Condensate Booster pump trip testing, and additional Reactor Feedwater Level Control system tuning.

Extended Power Uprate performance monitoring includes system and component vibration monitoring, moisture separator performance, and radiation measurements.

V. COMPLIANCE WITH REGULATORY GUIDANCE EXTENDED POWER UPRATE LICENSING TOPICAL REPORT In TVA's April 25, 2005 letter (Reference 5), TVA addressed the guidelines for the recommended integrated plant testing for Extended Power Uprates which are contained in the Licensing Topical Report (NEDC-32424P-A, Appendix L). TVA provided a comparison of the BFN initial startup tests to the planned testing of Unit 1 for EPU. This submittal concludes that large transient testing on Unit 1 is not warranted based on similarity of the units, the extensive Units 2 and 3 operating experience, and the currently planned Unit 1 restart testing.

STANDARD REVIEW PLAN 14.2.1 The key guidance contained in SRP can be summarized as follows:

" "(T)his review is intended to ensure that functions important to safety that rely on the integrated operation of multiple SSCs following an anticipated operational occurrence are adequately demonstrated ... "

" "The test program should be scheduled and sequenced to minimize the time untested functions important to safety are relied upon during operation ... "

S""If a function important to safety cannot be adequately tested by overlapping individual component- or system-level tests, the licensee should propose suitable system functional testing."

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" "If the licensee proposes not to perform a required transient test based on operating experience, a review should be conducted to determine the applicability of the operating experience to the specific plant configuration and test requirements."

" "The review scope can be limited to those functions important to safety associated with the anticipated operational occurrences described in Attachment 2 to this SRP."

TVA evaluated three of the operational occurrences described in to SRP 14.2.1:

  • Loss of Offsite Power,

The acceptance criteria specified in the SRP for each of these operational occurrences was reviewed against the SSA systems and modes and the associated BTRD testing to ensure the criteria would be satisfied by the planned Unit 1 restart testing.

Table 3 documents this comparison for the Loss of Offsite Power, Table 4 documents this comparison for the Turbine / Generator trip, and Table 5 documents this comparison for the isolation of all main steam lines. This review demonstrated that TVA will satisfy the SRP criteria for a loss of offsite power, turbine trip or generator trip, or the isolation of all main steam lines as part of its currently planned restart activities, without performing additional large transient testing.

The review of the SRP acceptance criteria supports TVA's position that the restart test program and planned EPU testing (References 4 and 5) will ensure that functions important to safety that rely on the integrated operation of multiple SSCs following-an anticipated operational occurrence are adequately demonstrated. The scheduling and sequencing of the testing, as shown in Figure 1, minimizes the time untested functions important to safety are relied upon during operation.

In summary, the SRP 14.2.1 acceptance criteria will be satisfied as part of currently planned restart testing and activities.

Large transient testing for a loss of offsite power, turbine trip or generator trip, or the isolation of all main steam lines would not provide additional useful information.

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VI. EVALUATION OF UNIT 1 RESTART TEST RESULTS SIMILARITY BETWEEN UNITS As part of the Unit 1 restart project, TVA is performing the same restart programs and implementing the same modifications that were previously completed on Units 2 and 3. In addition, TVA also incorporated the upgrades on the Units 2 and 3 five year plan into the scope of the Unit 1 restart project. As a result, Unit 1 will be operationally the same as Units 2 and 3.

SCOPE OF EVALUATION In order to ensure Unit 1 as a whole will respond operationally the same as Units 2 and 3 during steady state operation and transient events, the results of selected Unit 1 power ascension testing will be evaluated against applicable criteria and information.

The Unit 1 steady state tests to be benchmarked against Units 2 and 3 operating experience are:

" At between 15% to 20% power, before and after generator synchronization;

" At the power plateau, between 35% and 40% power; 0 At the power plateau, between 50% and 55% power;

" At the original Unit 1 licensed thermal power (100% - 3,293 MWt) plateau; and

" At the 105% power (3,458 MWt) plateau (Current Units 2 and 3 licensed thermal power level).

For example, before generator synchronization at between 15% and 20% power, the following parameters would be compared to the same parameters on Units 2 and/or 3 and any differences would be evaluated:

  • Core output (MWt),
  • Total steam flow,

" Number of bypass valves open,

" Reactor pressure, and

" Electrohydraulic control system header pressure.

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After generator synchronization, the following parameters would be compared:

  • Core output (MWt),
  • Total steam flow,

" Total Feedwater flow,

  • Reactor pressure,
  • Turbine control valve position,
  • Electrohydraulic control system header pressure, and
  • Generator output (MWe).

The Unit 1 power ascension transient tests to be benchmarked against Units 2 and 3 are:

" Reactor Feedwater level control system tuning;

" Recirculation flow control testing;

  • Electro-Hydraulic Control (EHC) pressure regulator tuning;

" Turbine valve testing (i.e., stop valves, control valves, and bypass valves);

" Full scram at greater than or equal to 30% power;

" HPCI and RCIC vessel injection tests.

For example, during safety relief valve cycling, the following parameters would be compared:

  • Core output (MWt),
  • Total steam flow,

" Total Feedwater flow,

" Number of bypass valves open,

" Reactor pressure,

" Electrohydraulic control system header pressure, and

  • Reactor water level.

These parameters would also be monitored with a safety relief valve open and compared to the changes identified during testing on Units 2 and/or 3.

The results of the power ascension transient tests will be compared to:

  • Unit 1 BTRD acceptance criteria,

" Unit 1 Technical Specification requirements,

  • Unit 1 Surveillance Instruction acceptance criteria, El-9

" Previous Units 2 and 3 restart or surveillance test results, and

  • Predicted responses from the BFN simulators.

Additional details for each category is discussed below.

UNIT 1 BTRD ACCEPTANCE CRITERIA As discussed above, the BTRDs specify the testing acceptance criteria for each system mode or function which ensures the plant is able to achieve safe shutdown. The testing and acceptance criteria for each BTRD required to satisfy the acceptance criteria for a loss of offsite power, turbine trip or generator trip, or the isolation of all main steam lines is provided in Tables 3 through 5. The results of the restart testing and the corresponding BTRD requirements are evaluated by the Joint Test Group prior to returning each system to service.

UNIT 1 TECHNICAL SPECIFICATION REQUIREMENTS The Unit 1 Technical Specifications contain requirements that are demonstrated by testing. Failure to meet Technical Specification acceptance criteria results in TVA taking the specified Required Actions.

UNIT 1 SURVEILLANCE INSTRUCTION ACCEPTANCE CRITERIA The failure to meet any acceptance criterion in the Surveillance Instruction results in a failure of the Surveillance Instruction, which would be evaluated as part of TVA's Corrective Action Program.

PREVIOUS UNITS 2 AND 3 OPERATING EXPERIENCE, AND PREDICTED RESPONSES FROM THE BFN SIMULATORS The goal for both BFN simulators is to ensure that no noticeable differences exist between the simulator control room and simulated systems, when evaluated against the control room and systems on the referenced unit. The software development, coupled with the scenario-based testing and the reconciliation of actual event data with the simulator modeling, ensure that the simulators will faithfully portray transient response, including turbine and generator trips. The results of the Unit 1 power ascension and transient testing shall be compared against the simulator response criteria specified in Sections 4.1.3 and 4.1.4 of ANSI/ANS-3.5-1998, Nuclear Power Plant Simulators for Use in Operator Training and Examination. The EI-1O

failure to meet any acceptance criterion in the standard would be evaluated as part of TVA's Corrective Action Program.

REVIEWS In accordance with the Master Refueling test Instruction (1-TI-319), the results of restart testing are required to be reviewed by the Joint Test Group. For each test plateau, a summary report is prepared by the Joint Test Group, reviewed by the Plant Operations Review Committee, and approved by the Plant Manager prior to proceeding to the next test plateau.

VII. CONCLUSION The SRP 14.2.1 acceptance criteria for a loss of offsite power, turbine trip or generator trip, or the isolation of all main steam lines will be satisfied as part of currently planned restart testing and activities. TVA will evaluate the results of the tests performed to satisfy the SRP acceptance criteria against Technical Specification requirements, test acceptance criteria, Units 2 and 3 operating experience, and responses predicted by the BFN simulators. Based on this information, large transient testing would not provide additional useful information.

VIII. REFERENCES

1. TVA letter, dated June 28, 2004, "Browns Ferry Nuclear Plant (BFN) - Unit 1 - Proposed Technical Specifications (TS) Change TS-431 - Request for License Amendment -

Extended Power Uprate (EPU) Operation."

2. NRC letter, dated December 22, 2005, "Browns Ferry Nuclear Plant, Unit 1 - Request for Additional Information for Extended Power Uprate (TS-431) (TAC No. MC3812)."
3. TVA letter, dated March 7, 2006, "Browns Ferry Nuclear Plant (BFN) - Unit 1 - Response to NRC Round 3 Requests for Additional Information Related to Technical Specifications (TS) Change No. TS-431 - Request for Extended Power Uprate Operation (TAC No. MC3812)."

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4. TVA letter, dated August 15, 2005. "Browns Ferry Nuclear Plant (BFN) - Unit 1 - Response to NRC Request for Additional Information Regarding the Restart Testing Program (TAC No. MC7208)."
5. TVA letter, dated April 25, 2005, "Browns Ferry Nuclear Plant (BFN) - Unit 1 - Response to NRC's Request for Additional Information Related to Technical Specifications (TS) Change No. TS-431 - Request for Extended Power Uprate Operation (TAC No. MC3812)."

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TABLE 1 STANDARD REVIEW PLAN 14.2.1 ACCEPTANCE CRITERIA Transient Test Rieferenlce ~ TyicaReordiPlan Inta TypicaITrasedu(ntTst cetneCiei n plcl cietAaye Dynamic response of plant RG 1.68, App A trip from steady state operation at greater Performance in accordance with design, including: 15.2.1 Turbine Trip to turbine trip 5.1.1 than 95% of RTP reactor coolant pumps do not trip (Turbine trip or generator IP 72580 initiation of the test by trip of the main trip) IP 72514 generator output breaker pressurizer spray valve opens and closes at the specified values recirculation system flow control mode must reactor pressure remains below the setpoint of the first safety be specified valves, pressurizer safety valves do not lift or weep pressurizer level within prescribed limits steam system power actuated pressure relief valve opens and closes at specified values reactor coolant pressure/temperature relationship remains within defined values steam generator level remains within prescribed limits, no flooding of the steam lines during the transient, no initiation of ECCS and MSIV isolation during the transient turbine bypass system operates to maintain specific pressure (plants with 100% bypass capability shall remain at power without scram during the transient) plants with select-rod-insertion shall maintain power without scram from recirculation pump overspeed or cold feedwater effect reactor protection system functions should be verified all safety and ECCS systems such as RPS, HPCI, diesel generators, and RCIC function without manual assistance if called upon Normal reactor cooling systems should maintain adequate cooling and prevent actuation of automatic depressurization system, even though relief valves may function to control pressure plant electrical loads (transferred as designed) turbine overspeed criteria met E1-13

Trasint es Rfernc Tyicl eactor Plant Initial TyiarnsetTs Accptnc Cieria and? ~ Aoplicale Accident Analyses

____________ ________ j ~ Condition Asoiae Fu~nctions Important to Safety) (SFRP Section)

Dynamic response of plant RG 1.68, App A Initial power level of 100% of Rtlý performance in accordance with design, including: 15.2.4 Main Steam Isolation to automatic closure of all 5.m.m Valve Closure (BWR) main steam isolation acceptance criteria include MSIV closure time valves IP 72510 Dynamic response of plant RG 1.68, App A 100% of RTP with electrical system aligned Performance in accordance with design, including: 15.2.6 Loss of Nonemergency for full load rejection 5.n.n for normal full-power operation and load AC Power to the Station rejection method should subject turbine to Automatic transfer of plant loads as designed, automatic start of Auxiliaries (Loss of Offsite Power IP 72517 maximum credible overspeed condition diesel generators, automatic load of diesel generators inthe Testing) specified sequence IP 72582 steady-state plant operations with greater than 10% generator output (IP 72517 & Reactor pressure remains below the first safety valve setting.

72582). Pressurized safety valves do not lift trip of the plant with breakers in specified All safety systems such as RPS, HPCI, diesel generators, and positions so that plant loads will be RCIC function without manual assistance transferred directly to the diesel generators following loss of house power Normal reactor cooling systems should maintain adequate core temperatures, and prevent actuation of the Automatic recirculation system flow control mode Depressurization System; however selected relief valves may specified function to control pressure Turbine bypass system operates to maintain specified pressure value Steam system power-actuated pressure relief valves open and close at specified value Pressurizer spray valves open and close at specified values.

Reactor coolant temperature/pressure relationship remains within prescribed values Pressurizer level is maintained within prescribed limits Steam generator level remains within prescribed limits E1-14

TABLE 2 SAFE SHUTDOWN ANALYSIS ACCIDENTS, TRANSIENTS, AND SPECIAL EVENTS

" Generator Trip

" Turbine-Generator Trip with Bypass Failure " Inadvertent Opening of All Bypass Valves

" Pressure Regulator-Failure - Closed " Pressure Regulator Failure - Open

" Isolation. of all Main Steam Lines " Loss of Offsite AC Power

  • Inadvertent Opening of a Safety/Relief Valve

" Loss of Feedwater Heater " Recirculation Pump Trip (One Pump Trip)

" Shutdown Cooling (RHRS) Malfunction

  • Recirculation Pump Trip (Two Pump Trip)

" Inadvertent Pump Start

  • Recirculation Pump Seizure

" Control Rod Withdrawal Error " Recirculation Flow Controller Failure Increasing Flow.

" Fuel Assembly Insertion " Startup of Idle Reactor Recirculation Pump

" Feedwater Controller Failure - Maximum Demand

  • Overpressure Protection (MSIV Closure -

Backup Scram)

" Control Rod Drop Accident

  • Pipe Break Inside Containment - Large Break
  • Rotated or Mis-located Bundle

" Intermediate Pipe Break Inside Containment " Flood

" Small Pipe Break Inside Containment " Low Reservoir Downstream Dam Failure

" Pipe Break Inside - Containment and Radiological

  • Tornado Effect

" Fuel Handling Accident

" Pipe Break Outside Containment

  • Fire

" Loss of Fuel Pool Cooling/Makeup El-15

TABLE 3 TESTING WHICH SATISFIES SRP 14.2.1 ACCEPTANCE CRITERIA FOR A LOSS OF OFFSITE POWER SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Automatic transfer of plant loads as designed, automatic start of diesel generators, automatic load of diesel generators in the specified sequence.

Criterion satisfied by:

SYSTEM AND MODE MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA Diesel Automatic and manual initiation Loss of voltage, degraded voltage and 1. Common Accident Signal and Pre-accident Generators of DGs to be available as an undervoltage test performed as part of Unit 2 Signal are tested in BTRD 057-05.

(DGs) on-site Standby AC source on: recovery. 2. Loss of voltage, loss of station power, 082-01 1. Common Accident degraded voltage, under voltage, and manual Signal; Automatic load sequencing of diesel start of DGs were tested during Unit 2 restart.

2. Pre-accident Signal; generator test performed as part of diesel
3. Loss of voltage / Loss of generator load acceptance test.

station power,

4. Degraded voltage and Both tests routinely required by Technical under voltage; and Specifications on a 24-month frequency.
5. Manual start.

480VAC Provide 480V Load Shed on Verifies loads are shed on loss of voltage on 1. Verify that load shedding will be initiated by 574-05 Degraded Voltage Condition Unit 1 480V Shutdown Boards 1A and 1B. undervoltage at each 480V Shutdown Board.

2. Verify by test that after a loss of voltage for Unit 2 480V Shutdown Board load shed greater than 5 seconds, the 480V Shutdown testing is currently performed on a 24-month Boards load shed.

frequency.

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SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

All safety systems such as Reactor Protection System, High Pressure Coolant Injection, diesel generators, and Reactor Core Isolation Cooling function without manual assistance.

Criterion satisfied by:

SYSTEM AND MODE MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA Reactor Provide automatic Scram and Functional testing and calibration of Turbine 1. The CRD System receives a Channel Al or Protection Scram Discharge Volume vent Control Valve low Electro-hydraulic Control A2 or a Channel B1 or B2 half Scram signal System (RPS) and drain valves isolation signal (EHC) pressure scram; from the RPS for all automatic scram 099-01 to Control Rod Drive (CRD) signals.

system (085) Functional testing and calibration of less than 2. Manual reset of a Scram is inhibited for a 90% Turbine Stop Valve open scram; and minimum of 10 seconds following a Scram.

3. System response is less than 50 msec from Reactor Protection System response of less opening of the sensor contacts up to and than 50 msec from opening of sensor contacts including opening of the trip actuator.

to opening of scram relay contacts.

Reactor Core Automatic RCIC initiation on RCIC logic system functional testing; 1. Verify that on RCIC initiation on RPV low Isolation Reactor Pressure Vessel (RPV) water level (Level 2), RCIC initiates from Cooling (RCIC) low water level (Level 2). RCIC flow testing at rated pressure; Standby Mode to Injection Mode and from 071-01 Automatic RCIC shutoff (if Pump Test Mode to Injection Mode.

operating) on RPV high water RCIC cold quick start at rated pressure; and 2. Verify initiation of RCIC automatic shutdown level (Level 8). Manual transfer logic on RPV high water level (Level 8).

of suction from Condensate RCIC injection into the vessel. 3. Verify the RCIC system can be manually Storage Tank to Suppression operated from the Main Control Room.

Pool. 4. Verify the RCIC suction path can be manually transferred from the Condensate Storage Tank (CST) to the Suppression Pool and that the RCIC suction valve interlocks function.

5. Verify steam line drain valves close when I_ RCIC steam supply valve opens.

E1-17

Criterion satisfied by (Cont.):

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Control Rod Provide Scram and close Scram Performs a full reactor scram at greater than 1. Verify a Scram signal from the Reactor Drive (CRD) Discharge Volume vent and 1% power; Protection System (RPS) will activate the 085-01 drain valves. Control Rod Drive (CRD) System to trip the Performs individual rod scram timing; and CRD Scram pilot valves and Scram Discharge Volume (SDV) vent and drain Performs Scram Discharge Volume vent and valves.

drain valve cycling and timing 2. Verify that due to trip of Scram pilot valves and SDV vent and drain pilots, all rods are inserted and SDV vent and drain valves close. (Above full scram must be from rated temperature and pressure at greater than 1% power.)

3. Verify Scram insertion time for all control rods.
4. Verify SDV vent and drain valves close on full scram in less than 60 sec.
5. Verify on loss of power to RPS, all rods are inserted and SDV vent and drain valves close automatically.
6. Verify indication of vent and drain valve position is provided in the Main Control Room and valves open to drain the SDV only after Scram reset.

Diesel Automatic and manual initiation Loss of voltage, degraded voltage and 1. Common Accident Signal and Pre-accident Generators of DGs to be available as an undervoltage test performed as part of Unit 2 Signal are tested (DGs) on-site Standby AC source on: recovery and routinely required by Technical in BTRD 057-05.

082-01 1. Common Accident Signal; Specifications. 2. Loss of voltage, loss of station power,

2. Pre-accident Signal; degraded voltage, under voltage, and
3. Loss of voltage ! Loss of manual start of DGs were tested during Unit station power; 2 restart.
4. Degraded voltage and under voltage; and 1 5. Manual start. I E1-18

Criterion satisfied by (Cont.):

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE High Pressure Automatic HPCI system initiation HPCI logic system functional testing; 1. Verify HPCI initiation on high Drywell Coolant on Reactor Pressure Vessel pressure or low RPV water level (Level 2).

Injection (RPV) low water level or high HPCI flow testing at rated pressure; Verify that HPCI initiates from standby mode (HPCI) drywell pressure signals. to injection mode and pump test mode the 073-01 Automatic HPCI system shutoff HPCI cold quick start at rated pressure; and injection mode.

(if operating) on reactor 2. Verify the steam line isolation valves close Feedwater system RPV high HPCI injection into the vessel, when the HPCI steam supply valve is open.

water level signal. Includes 3. Verify the CST suction valve opens when automatic transfer, ifneeded, the Suppression Pool suction valves are not from Condensate Storage Tank fully open.

(CST) to Suppression Pool on 4. Verify Suppression Pool suction valves low CST level or high automatically open and the CST suction Suppression Pool level. valve closes upon receipt of CST low level or high Suppression Pool level.

5. Verify the CST and Suppression Pool level instrumentation provides signals at the appropriate setpoints and level annunciation is provided to the Main Control Room.
6. Verify the Turbine Stop Valve closes on high RPV level (Level 8). Verify subsequent reactor low level (Level 2) will re-open the I_ Stop Valve.

EI-19

SRP 14.2.1 Acceptance Criterion: Perfotmanee in accordance with design, including:

" Normal reactor cooling systems should maintain adequate core temperatures, and prevent actuation of the Automatic Depressurization System; however selected relief valves may function to control pressure.

  • Reactor coolant temperature/pressure relationship remains within prescribed values.

" Steam system power-actuated pressure relief valves open and close at specified value.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Automatically open Safety Relief Performs calibration and functional check of Verify SRVs receive an open signal on high RPV 001-27 Valves (SRVs) on high Reactor SRV pressure switch opening logic, pressure.

pressure to provide Reactor Pressure Vessel (RPV) pressure relief.

Main Steam Open Safety Relief Valves Verifies lift pressure for all Safety Relief 1. Verify actuation of Main Control Room 001-07 (SRVs) on high Reactor Valves (SRVs). (MCR) manual switches energize the correct pressure to provide Reactor SRV solenoid with the associated transfer Pressure Vessel (RPV) pressure Verifies Automatic Depressurization System switch in "normal".

relief (ADS) and non-ADS SRV solenoids can by 2. Verify control of the non-Automatic manually cycled; Depressurization System (ADS) SRVs is not possible from the MCR when transfer Performs ADS logic system functional testing; switches are in "Emergency".

and 3. Verify SRV mechanical lift pressures.

Main Steam Automatic opening of Automatic Verify the ADS SRVs will, as demonstrated by 001-08 Depressurization System (ADS) Performs manual cycle of SRVs at greater energizing the solenoid, actuate automatically Safety Relief Valves (SRVs) than 935 psig. upon receipt of the appropriate combinations of upon coincident signals for two each of the following signals:

Core Spray pumps or one

  • Drywell pressure high (including ADS timers),

Residual Heat Removal (RHR) and pumps running and either , RPV water level (Level 1 and Level 3), and Reactor Pressure Vessel (RPV)

  • 1 Low Pressure Coolant Injection (LPCI) low water level (Level 1), high pump or 2 Core Spray pumps running, and Drywell pressure and time delay, 0 High Drywell pressure bypass timer has or Level 1 and high Drywell elapsed.

1 pressure bypass time delay. I I El-20

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Turbine bypass system operates to maintain specified pressure value.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Open Main Turbine bypass Verifies EHC logic functions to operate Test under Mode 047-05, below.

001-05 valves on Turbine Control bypass valves on a turbine trip System (047) turbine trip signal.

Turbine Control Provide hydraulic control to Verifies hydraulic control oil pressure after a 1. Verify hydraulic control is provided to open System open Main Steam turbine turbine trip opens bypass valves, and that the turbine bypass valves.

047-05 bypass valves on Turbine trip accumulators can hold valves open for a 2. Verify hydraulic pressure to keep all the signal. minimum of 10 seconds on loss of pumps; bypass valves open will be maintained for a and minimum of 10 seconds after a loss of the hydraulic supply or electrical power supply Verifies that after a turbine trip that bypass to the hydraulic unit.

valves open within 100 msec. of the start of 3. Verify, after a turbine trip, that the bypass Turbine Stop Valve closure, valves will start to open within 100 msec. of the start of the Main Steam stop valve Verifies hydraulic control oil pressure is closure.

provided to open the turbine bypass valves.

Main Steam Provide Main Steam line Calibration and functional check of EHC Main Verify 1-PT-001-16A and 16B provide control 001-24 pressure signal to Turbine Steam line and reactor pressure inputs to signals to the Turbine Control System (047).

Control System (047) for EHC logic operation of Main Steam turbine bypass valves.

E1-21

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

" Reactor pressure remains below the first safety valve setting. Pressurized safety valves do not lift.

" Pressurizer spray valves open and close at specified values.

  • Pressurizer level is maintained within prescribed limits.

Criteria not applicable to Boiling Water Reactors.

E1-22

TABLE 4 TESTING WHICH SATISFIES SRP 14.2.1 ACCEPTANCE CRITERIA FOR A TURBINE/GENERATOR TRIP SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

" Reactor coolant pumps do not trip.

" Plants with select-rod-insertion shall maintain power without scram from recirculation pump overspeed or cold feedwater effect.

Criterion not applicable to BFN since, by design, the Recirculation pumps trip upon receipt of a turbine trip or load rejection signal. BFN does not have the select-rod-insertion logic.

El-23

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

  • Steam system power actuated pressure relief valve opens and closes at specified values.

" Reactor coolant pressure/temperature relationship remains within defined values.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Automatically open Safety Relief Performs calibration and functional check of Verify SRVs receive an open signal on high RPV 001-27 Valves (SRVs) on high Reactor SRV pressure switch opening logic, pressure.

pressure to provide Reactor Pressure Vessel (RPV) pressure relief.

Main Steam Open Safety Relief Valves Verifies lift pressure for all Safety Relief 1. Verify actuation of Main Control Room 001-07 (SRVs) on high Reactor Valves (SRVs). (MCR) manual switches energize the correct pressure to provide Reactor SRV solenoid with the associated transfer Pressure Vessel (RPV) pressure Verifies Automatic Depressurization System switch in "normal".

relief (ADS) and non-ADS SRV solenoids can by 2. Verify control of the non-Automatic manually cycled; Depressurization System (ADS) SRVs is not possible from the MCR when transfer Performs ADS logic system functional testing; switches are in "Emergency".

and 3. Verify SRV mechanical lift pressures.

Main Steam Automatic opening of Automatic Verify the ADS SRVs will, as demonstrated by 001-08 Depressurization System (ADS) Performs manual cycle of SRVs at greater energizing the solenoid, actuate automatically Safety Relief Valves (SRVs) than 935 psig. upon receipt of the appropriate combinations of upon coincident signals for two each of the following signals:

Core Spray pumps or one 0 Drywell pressure high (including ADS timers),

Residual Heat Removal (RHR) and pumps running and either 0 RPV water level (Level I and Level 3), and Reactor Pressure Vessel (RPV) a 1 Low Pressure Coolant Injection (LPCI) low water level (Level 1), high pump or 2 Core Spray pumps running, and Drywell pressure and time delay,

  • High Drywell pressure bypass timer has or Level 1 and high Drywell elapsed.

pressure bypass time delay. I El-24

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

No flooding of the steam lines during the transient, no initiation of ECCS and MSIV isolation during the transient.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Feedwater Provide Reactor Pressure Verifies Reactor Feedwater Pump Turbines Verify High RPV water level (Level 8) signals are 003-11 Vessel (RPV) high water level and Main Turbines trip on high water level sent to Feedwater level control system for closure (Level 8) signal to Feedwater (Level 8); and of Main and Feedwater Turbine Steam Supply level control system for Main stop valves.

and Feedwater turbine trip. Performs Reactor Feedwater Pump Turbines Feedwater Energize solenoid to close Main and Main Turbines high level trip logic system Verify the trip signals from the Feedwater Level 003-18 Steam Feedwater turbine steam functional testing. control system and Main Condenser Low Vacuum supply valves on low vacuum or pressure switches energize the trip sorenoids for high Reactor Pressure Vessel the Feedwater Turbine stop valves.

(RPV) level (Level 8) signal.

Feedwater Provide Reactor Pressure Test under Mode 003-11.

Level Control Vessel (RPV) high water level 046-01 (Level 8) signal to energize Feedwater solenoid to close Feedwater turbine steam supply stop valve.

Low pressure Emergency Core Cooling Systems (ECCS), which initiate on high drywell pressure or low reactor vessel water level (Level 1) are not challenged. In addition to the above restart testing, TVA will perform Feedwater Level Control System tuning, turbine bypass valve, Reactor Feedwater Pump, Condensate Booster Pump, Condensate Pump trip, and turbine valve testing. TVA has confidence that these action will be adequate since the Feedwater Level Control System is the same on Unit 1 as Units 2 and 3. Units 2 and 3 have operating experience on turbine trips at 105% power and have had no issues with the response of the ECCS, have experienced no Main Steam Isolation Valve closures, and had no issues with reactor vessel water level control (high or low).

EI-25

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Turbine bypass system operates to maintain specific pressure (plants with 100% bypass capability shall remain at power without scram during the transient).

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Open Main Turbine bypass Verifies EHC logic functions to operate Test under Mode 047-05, below.

001-05 valves on Turbine Control bypass valves on a turbine trip System (047) turbine trip signal.

Turbine Control Provide hydraulic control to Verifies hydraulic control oil pressure after a 1. Verify hydraulic control is provided to open System open Main Steam turbine turbine trip opens bypass valves, and that the turbine bypass valves.

047-05 bypass valves on Turbine trip accumulators can hold valves open for a 2. Verify hydraulic pressure to keep all the signal. minimum of 10 seconds on loss of pumps; bypass valves open will be maintained for a and minimum of 10 seconds after a loss of the hydraulic supply or electrical power supply Verifies that after a turbine trip that bypass to the hydraulic unit.

valves open within 100 msec. of the start of 3. Verify, after a turbine trip, that the bypass Turbine Stop Valve closure, valves will start to open within 100 msec. of the start of the Main Steam stop valve Verifies hydraulic control oil pressure is closure.

provided to open the turbine bypass valves.

Main Steam Provide Main Steam line Calibration and functional check of EHC Main Verify 1-PT-001-16A and 16B provide control 001-24 pressure signal to Turbine Steam line and reactor pressure inputs to signals to the Turbine Control System (047).

Control System (047) for EHC logic operation of Main Steam turbine I bypass valves, I E1-26

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Reactor protection system functions should be verified.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Provide Main Turbine trip on Verifies calibration and functional testing of Verify that position switches transmit the correct 001-01 Turbine Stop Valve less than the Turbine Main Stop Valve limit switches signal to RPS when the corresponding Turbine 90% open trip signal to Reactor scram Stop Valve are less than 90% open.

Protection System (RPS).

Main Steam Provide greater than 26% open Verifies calibration and functional check of the 1. Verify trip signals are provided via the trip 001-23 turbine first stage pressure Turbine First Stage Pressure scram bypass units.

interlock signal to Reactor from a Main Turbine trip 2. Verify that each trip unit initiates the correct Protection System (RPS) fail trip signal (for each channel) tolthe RPS in safe logic. response to a real or simulated input to the associated sensor.

Control Rod Provide Scram and close Scram Performs a full reactor scram at greater than 1. Verify a Scram signal from the Reactor Drive (CRD) Discharge Volume vent and 1% power; Protection System (RPS) will activate the 085-01 drain valves. Control Rod Drive (CRD) System to trip the Performs individual rod scram timing; and CRD Scram pilot valves and Scram Discharge Volume (SDV) vent and drain Performs Scram Discharge Volume vent and valves.

drain valve cycling and timing 2. Verify that due to trip of Scram pilot valves and SDV vent and drain pilots, all rods are inserted and SDV vent and drain valves close. (Above full scram must be from rated temperature and pressure at greater than 1% power.)

3. Verify Scram insertion time for all control rods.
4. Verify SDV vent and drain valves close on full scram in less than 60 sec.
5. Verify on loss of power to RPS, all rods are inserted and SDV vent and drain valves close automatically.
6. Verify indication of vent and drain valve position is provided in the Main Control Room and valves open to drain the SDV only after Scram reset.

EI-27

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

All safety and ECCS systems such as RPS, HPCI, diesel generators, and RCIC function without manual assistance if called upon.

Criterion satisfied by:

SYSTEM AND MODE MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA Reactor Provide automatic Scram and Functional testing and calibration of Turbine 1. The CRD System receives a Channel Al or Protection Scram Discharge Volume vent Control Valve low Electro-hydraulic Control A2 or a Channel B1 or B2 half Scram signal System (RPS) and drain valves isolation signal (EHC) pressure scram; from the RPS for all automatic scram 099-01 to Control Rod Drive (CRD) signals.

system (085) Functional testing and calibration of less than 2. Manual reset of a Scram is inhibited for a 90% Turbine Stop Valve open scram; and minimum of 10 seconds following a Scram.

3. System response is less than 50 msec from Reactor Protection System response of less opening of the sensor contacts up to and than 50 msec from opening of sensor contacts including opening of the trip actuator.

to opening of scram relay contacts.

Reactor Core Automatic RCIC initiation on RCIC logic system functional testing; 1. Verify that on RCIC initiation on RPV low Isolation Reactor Pressure Vessel (RPV) water level (Level 2), RCIC initiates from Cooling (RCIC) low water level (Level 2). RCIC flow testing at rated pressure; Standby Mode to Injection Mode and from 071-01 Automatic RCIC shutoff (if Pump Test Mode to Injection Mode.

operating) on RPV high water RCIC cold quick start at rated pressure; and 2. Verify initiation of RCIC automatic shutdown level (Level 8). Manual transfer logic on RPV high water level (Level 8).

of suction from Condensate RCIC injection into the vessel. 3. Verify the RCIC system can be manually Storage Tank to Suppression operated from the Main Control Room.

Pool. 4. Verify the RCIC suction path can be manually transferred from the Condensate Storage Tank (CST) to the Suppression Pool and that the RCIC suction valve interlocks function.

5. Verify steam line drain valves close when I RCIC steam supply valve opens.

E1-28

Criterion satisfied by (Cont.):

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Control Rod Provide Scram and close Scram Performs a full reactor scram at greater than 1. Verify a Scram signal from the Reactor Drive (CRD) Discharge Volume vent and 1% power; Protection System (RPS) will activate the 085-01 drain valves. Control Rod Drive (CRD) System to trip the Performs individual rod scram timing; and CRD Scram pilot valves and Scram Discharge Volume (SDV) vent and drain Performs Scram Discharge Volume vent and valves.

drain valve cycling and timing 2. Verify that due to trip of Scram pilot valves and SDV vent and drain pilots, all rods are inserted and SDV vent and drain valves close. (Above full scram must be from rated temperature and pressure at greater than 1% power.)

3. Verify Scram insertion time for all control rods.
4. Verify SDV vent and drain valves close on full scram in less than 60 sec.
5. Verify on loss of power to RPS, all rods are inserted and SDV vent and drain valves close automatically.
6. Verify indication of vent and drain valve position is provided in the Main Control Room and valves open to drain the SDV only after Scram reset.

Diesel Automatic and manual initiation Loss of voltage, degraded voltage and 1. Common Accident Signal and Pre-accident Generators of DGs to be available as an undervoltage test performed as part of Unit 2 Signal are tested (DGs) on-site Standby AC source on: recovery and routinely required by Technical in BTRD 057-05.

082-01 6. Common Accident Signal; Specifications. 2. Loss of voltage, loss of station power,

7. Pre-accident Signal; degraded voltage, under voltage, and
8. Loss of voltage / Loss of manual start of DGs were tested during Unit station power; 2 restart.
9. Degraded voltage and under voltage; and
10. Manual start.

E1-29

Criterion satisfied by (Cont.):

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE High Pressure Automatic HPCI system initiation HPCI logic system functional testing; 1. Verify HPCI initiation on high Drywell Coolant on Reactor Pressure Vessel pressure or low RPV water level (Level 2).

Injection (RPV) low water level or high HPCI flow testing at rated pressure; Verify that HPCI initiates from standby mode (HPCI) drywell pressure signals. to injection mode and pump test mode the 073-01 Automatic HPCI system shutoff HPCI cold quick start at rated pressure; and injection mode.

(ifoperating) on reactor 2. Verify the steam line isolation valves close Feedwater system RPV high HPCI injection into the vessel. when the HPCI steam supply valve is open.

water level signal. Includes 3. Verify the CST suction valve opens when automatic transfer, if needed, the Suppression Pool suction valves are not from Condensate Storage Tank fully open.

(CST) to Suppression Pool on 4. Verify Suppression Pool suction valves low CST level or high automatically open and the CST suction Suppression Pool level, valve closes upon receipt of CST low level or high Suppression Pool level.

5. Verify the CST and Suppression Pool level instrumentation provides signals at the appropriate setpoints and level annunciation is provided to the Main Control Room.
6. Verify the Turbine Stop Valve closes on high RPV level (Level 8). Verify subsequent reactor low level (Level 2) will re-open the I_ Stop Valve.

EI-30

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Normal reactor cooling systems should maintain adequate cooling and prevent actuation of automatic depressurization system, even though relief valves may function to control pressure.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Automatically open Safety Relief Performs calibration and functional check of Verify SRVs receive an open signal on high RPV 001-27 Valves (SRVs) on high Reactor SRV pressure switch opening logic, pressure.

pressure to provide Reactor Pressure Vessel (RPV) pressure relief.

Main Steam Open Safety Relief Valves Verifies lift pressure for all Safety Relief 1. Verify actuation of Main Control Room 001-07 (SRVs) on high Reactor Valves (SRVs). (MCR) manual switches energize the correct pressure to provide Reactor SRV solenoid with the associated transfer Pressure Vessel (RPV) pressure Verifies Automatic Depressurization System switch in "normal".

relief (ADS) and non-ADS SRV solenoids can by 2. Verify control of the non-Automatic manually cycled; Depressurization System (ADS) SRVs is not possible from the MCR when transfer Performs ADS logic system functional testing; switches are in "Emergency".

and 3. Verify SRV mechanical lift pressures.

Main Steam Automatic opening of Automatic Verify the ADS SRVs will, as demonstrated by 001-08 Depressurization System (ADS) Performs manual cycle of SRVs at greater energizing the solenoid, actuate automatically Safety Relief Valves (SRVs) than 935 psig. upon receipt of the appropriate combinations of upon coincident signals for two each of the following signals:

Core Spray pumps or one 0 Drywell pressure high (including ADS timers),

Residual Heat Removal (RHR) and pumps running and either 0 RPV water level (Level 1 and Level 3), and Reactor Pressure Vessel (RPV)

  • 1 Low Pressure Coolant Injection (LPCI) low water level (Level 1), high pump or 2 Core Spray pumps running, and Drywell pressure and time delay, 0 High Drywell pressure bypass timer has or Level 1 and high Drywell elapsed.

I pressure bypass time delay. I E1-31

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Plant electrical loads (transferred as designed).

Criterion satisfied by:

SYSTEM AND MODE MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA Diesel Automatic and manual initiation Loss of voltage, degraded voltage and 1. Common Accident Signal and Pre-accident Generators of DGs to be available as an undervoltage test performed as part of Unit 2 Signal are tested in BTRD 057-05.

(DGs) on-site Standby AC source on: recovery. 2. Loss of voltage, loss of station power, 082-01 1. Common Accident Signal; degraded voltage, under voltage, and manual

2. Pre-accident Signal; Automatic load sequencing of diesel start of DGs were tested during Unit 2 restart.
3. Loss of voltage / Loss of generator test performed as part of diesel station power; generator load acceptance test.
4. Degraded voltage and under voltage; and Both tests routinely required by Technical
5. Manual start. Specifications on a 24-month frequency.

480VAC Provide 480V Load Shed on Verifies loads are shed on loss of voltage on 1. Verify that load shedding will be initiated by 574-05 Degraded .Voltage Condition Unit I 480V Shutdown Boards 1A and lB. undervoltage at each 480V Shutdown Board.

2. Verify by test that after a loss of voltage for Unit 2 480V Shutdown Board load shed greater than 5 seconds, the 480V Shutdown testing is currently performed on a 24-month Boards load shed.

I frequency.

E1-32

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

Turbine overspeed criteria met.

TVA will satisfy this criterion. Overspeed trip testing of the Main Turbine is performed as part of Power Ascension Testing. The BFN units directly trip the Main Turbine on a load reject signal. EPU has minimal impact on the store energy following a turbine trip.

EI-33

SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

" Pressurizer spray valve opens and closes at the specified values.

" Reactor pressure remains below the setpoint of the first safety valves, pressurizer safety valves do not lift or weep.

" Pressurizer level within prescribed limits.

" Steam generator level remains within prescribed limits.

Criteria not applicable to Boiling Water Reactors.

E1-34

TABLE 5 TESTING WHICH SATISFIES SRP 14.2.1 ACCEPTANCE CRITERIA FOR CLOSURE OF ALL MAIN STEAM ISOLATION VALVES SRP 14.2.1 Acceptance Criterion: Performance in accordance with design, including:

MSIV closing time.

Criterion satisfied by:

SYSTEM AND MODE DESCRIPTION TESTING DESCRIPTION ACCEPTANCE CRITERIA MODE Main Steam Close Main Steam Isolation 1. Verify each MSIV closes in 3 to 5 seconds 001-03 Valves (MSIVs) on primary upon receipt of a closure signaf.

containment isolation signal. 2. Verify each MSIV closes upon-loss of both AC and DC solenoid power. Verify each valve closes on loss of air pressure.

3. Verify when only the 120VAC circuit or the 25OVDC circuit are disconnected from the MSIV, the valve remains open.
4. Verify each MSIV accumulator leakage rate is within its design value.
5. Verify closure of either MSIV in one steam line transmits the correct signal to the I I_ I Reactor Protection System.

The MSIV closure time is checked and adjusted once per cycle in accordance with the In Service Testing program. Procedural requirements, which are based on operating experience, require the MSIV closure time to be set between four and five seconds while the MSIVs are cold to ensure the closing time will be greater than three seconds at rated temperature and pressure.

The change in closure time due to EPU conditions is negligible.

E1-35

4 Post /ModificationI Maintenance Testing

  • Verification of proper rotation and setting limits on valves
  • Motor-operated valve (MOV) setting and baseline diagnosl
  • Pump operability testing (verification of rated flow within sl 0 Valve operability testing 0 System flow balancing
  • System and component leak testing
  • Instrument channel calibration and functional testing 0 Protective relay calibration and functional testing
  • Logic system functional testing, E1-36 0 System run and confirmation that process parameters are

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY (TVA)

BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATION (TS) CHANGE 431, SUPPLEMENTAL RESPONSE REGARDING EXTENDED POWER UPRATE -

TESTING PLANNED TO SATISFY DRAFT STANDARD REVIEW PLAN 14.2.1

SUMMARY

OF COMMITMENTS TVA will evaluate the results of the steady state and transient tests listed in Section VI against Technical Specification requirements, test acceptance criteria, Units 2 and 3 operating experience, and responses predicted by the BFN simulators.

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