TXX-4071, Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Station Procedure EGT-706 Will Be Revised to Include Checking Indications of Breaker Operability

Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Station Procedure EGT-706 Will Be Revised to Include Checking Indications of Breaker Operability

ML20081K589
Person / Time
Site:	Comanche Peak
Issue date:	11/03/1983
From:	Clements B TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
GL-83-28, TAC-R00171, TAC-R00172, TAC-R171, TAC-R172, TXX-4071, NUDOCS 8311100183
Download: ML20081K589 (59)

Text

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L g # TXX-4071 TEXAS UTILITIES GENERATING COMPANY mu nava mwen.nau_urexis mo, File # 10140 BILLY R. CLEMENTS

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November 3,1983 Mr. D. G. Eisenhut, Director U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Licensing Washington, D.C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446 RESPONSE TO GENERIC LETTER 83-28

Dear Mr. Eisenhut:

Enclosed are forty copies of the CPSES response to NRC Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events".

Respectfully submitted,

/ fbY  % ub BRC:grr Affidavit Enclosure i

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8311100183 831103 DR ADOCK 05000445 PDR

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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

, TEXAS UTILITIES GENERATING COMPANY ) Docket Nos. 50-445

) 50-446 (Comanche Peak Steam Electric )

Station, Units 1 and 2) )

AFFIDAVIT B. R. Clements being duly sworn, hereby deposes and says that he is Vice President, Nuclear of Texas Utilities Generating Company, the ,

Applicant herein; that he is duly authorized to sign and file with the Nuclear Regulatory Commission this response to NRC Generic Letter 83-28 with respect to the captioned facilities; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.

0Yks$

B. R. Clements Vice President, Nuclear STATE OF TEXAS )

) ss COUNTY OF DALLAS )

Subscribed and sworn to before me, a Notary Public in and for k d d , on thisd day of 41 Cw-nddA.) ,198 r

YIW Notary Public My commission expires A. 17 ,19[

e CPSES RESPONSE TO NRC GENERIC LETTER 83-28

" REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS" November 7, 1983 s

CPSES l 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

A description of our program to ensure a safe restart fol;lcwing an unscheduled reactor shutdown is provided below:

1. The criteria for determining the acceptability to restart will be included in Integrated Plant Operating Procedure IP0-002A, Plant Startup from Hot Standby to Minimum Load.

This criteria will require the satisfactory completion of Operations Department Administration Procedure ODA-108, Post Trip Review Evaluation, in its entirety, or the performance of Section 1 of ODA-108 and the verification that selected plant parameters have responded properly if the cause of the shutdown has clearly been determined. Both of the above also requires the authorization of the Operations Superintendent or his designated alternate for restart.

2. The responsibilities and authorities of personnel who will perform the review and analysis of these events are clearly defined in the following procedures:

- ODA-101, " Operations Department Organization and Responsibilities"

- ODA-102, " Shift Complement, Responsibilities and Authorities"

- 00A-108, " Post Trip Review Evaluation"

3. The necessary qualifications and training for the responsible personnel are defined in the following procedures:

- TRA-201, " Operations Department Training"

- TRA-203, "NRC Licensed Operator Replacement Training Requirements" l

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CPSES

- TRA-204, " Licensed Operator Requalification Training

Program" 4, 5 & 6.

Operations Department Administration Procedure ODA-108, Post r Trip Evaluation, describes the steps required to perform a

! systematic safety assessment of unscheduled reactor shutdowns. The review process is divided into eight sections ,

with section 1 being performed by personnel on shift at the time of the event and the remaining sections being performed by an Operations Department engineer. The following is a brief description of each section of the review process:

Section 1, " Identification of Accident", contains preliminary information provided by on-shift personnel such as apparent cause of trip, description of event, and activities in progress prior to trip.

Section 2, " Preliminary Investigation", lists selected plant parameters just prior to the trip obtained from control board recorders and operator interviews.

Section 3, " Surveillance Testing and Maintenance Activities in Progress", will list any equipment out of service or

! activities in progress due to maintenance or testing.

Section 4, " Plant Response", will contain information regarding actuation of safety systems and their proper f operation.

Section 5, " Trip Review", will contain information obtained from the review of applicable trend recorders and typewriter l printouts.

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Section 6, " Trip Investigation", will contain information necessary to arrive at a definite conclusion as to the cause of the trip and shall include such items as equipment or controls operated to stabilize the plant, radiation monitors alarmed, minimum and maximum values of selected plant

• parameters during the transient, and any deficiencies noted on system response or component operation. -

Section 7, " Preliminary Safety Assessment", will contain a review of specific plant parameter response such as RCS Pressure, Cooldown Rate, RCS Inventory, and Pressurizer Level.

Section 8, "The Conclusion", will summarize all information obtained and arrive at a conclusion for the trip and a recommendation for restart. If a recommendation for restart cannot be made, provision for assigning other engineers or a committee for followup action is provided.

7. CPSES Procedure ODA-108, " Post Trip Review Evaluation" (Attachment 6), establishes a systematic method to be used to assess unscheduled reactor shutdowns.

1.2 POST-TRIP REVIEW (DATA AND INFORMATION CAPABILITY)

This section describes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown.

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1. Capaoility for Assessing Sequer.ce of Events (On-Off Indications)

1. System Hardware / Software l

i The principal unit of this system is a modified Westinghouse 2500 series general purpose digital computer located in a single cabinet in the plant computer room.

CPSES This cabinet contains the memory boards, central processing unit (CPU), fixed head mass memory discs, and operator's panel. A summary of computer charactefi stics is shown below:

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Word Size: 16-bit Core Memory Size: 64K words (Expansion capability of 2048K words)

Mass Memory Size: 1024K words Memory Cycle Time: 750-950 nanosec Disc Assess Time: 8.7 millisec The Operator's Panel is an entry and display device, located within the Mass Memory Cabinet which allows the operator to perform such functions as boot-strapping binary programs into the system; halting, resetting and starting the computer; stepping through the instructions of a program; displaying and/or altering the registers and displaying and/or altering specific memory locations.

This device is primarily used by computer maintenance personnel but may be used by the Plant Operator on special occasions to shutdown or restart the P-2500 computer system.

Also, located in the computer room are the following hardware components:

Input /0utput Cabinets - Cabinets which receive

information from plant inputs and pass signals to the CPU l

for processing.

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CPSES Paper Tape Equipment - A combination paper tape punch and paper tape reader. The reader receives data or program instructions from paper tape and passes it to the CPU for transfer to storage. The punch receives data fro'm the CPU and transfers it to a paper tape.

t Programmer's Console - An input / output device which allows a programmer to enter programs, access memory locations, determine program status, give commands and activate or disable programs.

Card Equipment - A separate card reader and card punch.

These peripherals perform the same function as the paper tape equipment except they use computer cards rather than paper tape.

Lineprinter - 'A 300 line per minute printer which is used for long printouts which would consume much time if allowed to print out on the control room typers.

In the control room, the following hardware is available to the plant operator for program initiation, interrogation, alarm, and display:

Analog Trend Recorders - Two two-pen recorders for obtaining trend of plant inputs.

Operator's Console - This unit is the plant operator's means of communicating with the P-2500 computer. The Operator's Console is divided into four sections: an alarm printer, a trend printer, a log printer, and control module.

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I CPSES Alarm Printer The alarm printer prints out messages that fall ihto four general categories:

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a. Alarms These result automatically when a plant sensor exceeds a preset limit or when a calculation based on plant inputs exceeds a preset limit.

b. Operator Confirmation Messages These result when a pushbutton is used on the Operator's Console control module keyboard.

c. Conditional and Value Reviews These result when the operator has selected the alarm printer as the output printer for a review.

d. Demand Value Printout These result when the operator requests the instantaneous value of an analog.

An alarm bell is actuated when an alarm is printed on the alarm printer.

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CPSES Trend Printer The trend printer prints out messages that fall iAto three general categories:

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a. Periodic Trends Operator selected addressable values will be printed periodically or when requested via the keyboard.

b. Nuclear Program Trends The values of certain quantities collected over a period of time will be printed automatically when initiated by various nuclear monitoring functions.

c. Conditional and Value Reviews These result when the operator has selected the trend printer as the output printer for a review.

Log Printer The log printer uses a pre-printed log sheet to present the important plant parameters collected and calculated by the computer automatically on the hour and whea requested by the operator.

l Control Module l

l The control module is a CRT with an associated alphanumerical and coded function keyboard used by the plant operator to prepare and initiate available computer programs. The display is designated as the Operator CRT.

Also available are indicating lights which display console failure status. Depending on the options i

CPSES integrated into the computer system, the indicating rows are located vertically on either side of the keyboard face or are grouped into horizontal rows on a sep'erate indicating panel. An ALARM ACKNOWLEDGE function key allows the operator to re-enable the audible alarm and reset (turn off) the indicating light.

The software associated with the P-2500 computer system consists of program instructions and stored data. The computer language used to write software is normally FORTRAN IV or Westinghouse Assembly language, which are in turn converted to P-2500 machine language by an internal c.ompiler. The computer uses binary notation internally for processing and numerical computation. For numerical loading, display and printout the computer may use either binary, decimal, or hexadecimal notation dependent on the procedure or input / output device selected.

2. Parameters Monitored Refer to Attachment 1.

3. Time Between Events The order shown in the sequence of events print-out is accurate for events occurring four or more milliseconds apart.

4. Format For Displaying Data And Information Refer to Attachment 4

CPSES

5. Capability For Retention Of Data And Information

a. Plant procedure requires retention of .

sequence-of-events printout as a permanent plant record, t

b. Printout begins automatically when 50 register changes occur or 1 minute after any one register change occurs.

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c. Data is retained upon a loss o' power to the computer and may be accessed upon a resti ration of power to the computer.

6. Power Sources Although the P-2500 is not considered a Class 1E component, it is powered as a non-interruptable load from a Class IE electrical bus.

2. Capability for Accessing the Time History of Analog Variables Needed to Determine the Cause of Unscheduled Reactor Shutdowns, and the Functioning of Safety-Related Equipment

1. System Hardware / Software

Refer to Section 1.2.1.1, above.

2. Parameters Monitored, Sampling Rate and Basis for Selecting Parameters and Sampling Rate Refer to Attachment 2.

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CPSES The list of analog variables monitored for the post trip analysis and the frequency and duration for monitoring them were supplied by the NSSS vendor (Westinghouse).

This information is standardized for Westinghouse four loop plants and is based upon (1) past history of operating plants and (2) an engineering evaluation to determine what information will be required or desirable for determining the cause of an unplanned reactor trip.

3. Duration of Time History Refer to Attachment 2.

4 Format for Displaying Data and Information Refer to Attachment 5.

5. Capability for Retention of Data, Information, and Physical Evidence

a. Upon power loss to computer, information is maintained in memory through an independent DC power supply and can be retrieved upon restoration of power to the computer.

b. Plant procedures require restoration of post-mortem trend analysis (post trip) printout as permanent records for the life of the plant.

c. Printout begins autom3tically three minutes after a trip.

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CPSES

6. Power Source Refer to Section 1.2.1.6, above. .

3. Other Data And Information r

No other data and information is required to assess the cause of unscheduled reactor shutdowns.

4. Schedule For Planned Changes No changes are planned to existing data and information capability.

2.1 E0VIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)

In letter TXX-4041 dated September 6,1983 to Mr. D. G. Eisenhut from Mr. H. C. Schmidt, an extension of time was requested to respond to Sections 2.1 and 2.2. The Nuclear Utility Task Action Committee (NUTAC) was formed September 1,1983 by 42 utilities and is sponsored by the Industry Review Group (IRG) of the Institute of Nuclear Power Operations (INPO). The specific purpose of NUTAC is to define an appropriate vendor interface program. Being a participant in NUTAC as well as the INP0 SEE-IN l and NPRDS programs, we will concurrently evaluate in more detail j our plant's existing capabilities for interfacing with vendors.

After completion of NUTAC's efforts, we will submit our program l description to the NRC by February 29, 1984. -

2.2 E0VIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAMS FOR ALL SAFETY-RELATED COMPONENTS)

Refer to Section 2.1, above.

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CPSES 3.1 POST-MAINTENANCE TESTING (REACTOR TRIP SYSTEM COMPONENTS)

1. Testing and maintenance procedures were reviewed to determine if post-maintenance operability testing of safety related

, components in the reactor trip system is required to be conducted and adequately demonstrates that the equipment is capable of performing its safety functions. Adequate post-maintenance testing requirements are well defined in Instrument and Control procedures for returning to service.

These procedures specify retest requirements for corrective maintenance actually performed. Post-maintenance testing requirements are also defined during the prework review of Maintenance Action Requests (MAR) and are verified during the post work MAR review.

2. CPSES procedures contain the current testing guidance for the reactor trip system with one exception. Testing to verify the state of the P-4 interlock has not yet been incorporated into plant procedures because the hardware changes to implement corrective action for this undetectable failure problem have only recently been installed. Plant procedures will be revised before fuel load to incorporate this test.

Any additional test guidance which arises from the L'estinghouse Owner's Group life cycle testing and evaluation on DS-416 breakers will be evaluated and incorporated as necessary in CPSES procedures.

3. The Westinghouse Owner's Group (WOG) is conducting a study of the maintenance and testing results of DS-416 breakers and has performed an analysts of maintenance and testing practices which may degrade reactor safety (WCAP-10271).

l CPSES endorses WCAP-10271 and will submit proposed changes to Technical Specifications when the WOG review of the DS-416 breakers is completed.

CPSES 3.2 POST-MAINTENANCE TESTING (ALL OTHER SAFETY-RELATED COMPONENTS)

1. Test and Maintenance procedures were reviewed to dets,rmine whether post-maintenance operability testing of components in other safety related systems is required to be conducted and adequately demonstrates that affected equipment is capable of performing its safety-related functions. A review of each maintenance action to determine post-maintenance testing is required by existing plant procedures. To insure that the testing properly demonstrates the operability of affected equipment, an index of component tag number versus potential testing requirements will be developed. Tnis list will be used by the Shift Supervisor and the performing maintenance organization (Maintenance Services or Instrument & Controls) to determine the testing requirements for specific maintenance.

2. CPSES Test and Maintenance procedures are still in development. Those procedures which have already been written were thoroughly reviewed for testing requirements and technical accuracy and consideration was given to vendor and engineering recommendations. In addition, CPSES has developed a Master Surveillance Test List to ensure all Technical Specification testing and maintenance requirements are properly addressed.

3. CPSES Technical Specifications are under review in preparation for licensing. CPSES has been-reviewing the technical specifications for testing or maintenance which could degrade rather than enhance safety. No testing requirements in this category are outstanding at this time.

CPSES 4.1 REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED M0[Qf CATIONS)

Following the recommendation of Westinghouse, CPSES will ,replace the existing under voltage trip attachments on all reactor trip breakers as soon as these are made available by Westinghouse.

(Ref: Westinghouse letter No. NS-EPR-2753 dated April 21,1983.)

This is the only modification identified for the DS-416 breakers.

i 4.2 REACTOR TRIP SYSTEM RELIABILITY (PREVENTATIVE MAINTENANCE AND SURVEILLANCE FOR REACTOR TRIP BREAKERS)

1. Electrical Maintenance has a Maintenance Instruction, EMI-302, "480V Air Circuit Breaker Inspection", which addresses the inspection, testing and adjustments of the Westinghouse CS-416 circuit breaker. Also, Results Engineering has a procedure, EGT-706A, " Engineered Safeguards System Safety Injection Actuation Testing", which tests the undervoltage and shunt trip coils. Westinghouse is currently proposing a comprehensive maintenance program for these breakers. Maintenance Services Engineering will review Westinghouse's recommended maintenance for these breakers and incorporate the steps that are necessary to ensure the l

operability of the reactor trip breakers into our current maintenance program.

2. The Technical Specifications for CPSES Unit I do not currently require maintenance activities on these breakers and CPSES Procedure EMI-302, discussed above, coes not require data to be trended. However, this procedure does have acceptance criteria to which the parameters measured during testing are compared. Maintenance Services Engineering will review Westinghouse recommendations and evaluate the value of trending parameters.

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CPSES

3. The Westinghouse Owner's Group is sponsoring a Westinghouse program to conduct life cycle testing of the reactor trip breaker type installed at CPSES. The results of thig effort are expected during the second quarter of 1984 4 Refer to Section 4.2.3, above.

4.3 REACTOR TRIP SYSTEM RELIABILITY (AUTOMATIC ACTUATION OF SHUNT TRIP ATTACHMENT FOR WESTINGHOUSE AND B&W PLANTS)

CPSES will implement the auto shunt trip feature in reactor trip circuit breakers as outlined in the Generic Design submitted by the Westinghouse Owner's Group on June 14, 1983. The plant specific information for CPSES as identified in the NRC September 20, 1983 SER on Generic Auto Shunt Trip Design is described below.

1. " Provide the electrical schematic / elementary diagrams for the reactor trip and bypass breakers showing the undervoltage and shunt coil actuation circuits as well as the breaker control (e.g. closing) circuits and circuits providing breaker status information/ alarms to the control room."

The electrical schematics / elementary diagrams incorporating auto shunt trip actuation are awaited from Westinghouse.

Tentative date of availability is middle of November 1983.

2. " Identify the power sources for the shunt trip coils. Verify that they are Class 1E and that all components providing power to the shunt trip circuitry are Class 1E and that any faults within non-Class 1E circuitry will not degrade the shunt trip function. Describe the annunciation / indication provided in the control room upon loss of power to the shunt l trip circuits. Also describe the overvoltage protection l

l CPSES and/or alarms provided to prevent or alert the operator (s) to an overvoltage condition that could affect both the UV coil and the parallel shunt trip actuation relay."

Power Supply e

Shunt trip coils in the reactor trip breakers are powered from Clast IE station batteries via 125 vdc switch boards and distribution panels for both the redundant trains. Switch boards, distribution panels, circuit breakers and cabling are Class 1E. Class 1E battery chargers (two per train) are powered from separate Class IE power sources.

Since the Class 1E circuitry provided to the shunt trip is separated from non-Class 1E circuitry as addressed in the FSAR, credible faults within non-Class IE circuitry will not degrade the shunt trip function.

Indication Existing indications on the Main Control Board for breaker operation are red and green position lights. These lights are powered from the same fused 125 vdc supply used for closing and shunt tripping the circuit breakers. The green light being on indicates that the breaker is open and power is available for closing and tripping the breaker. The red light indicates that the breaker is closed. Since the red light is connected in series with the shunt trip coil and an l

i "a" auxiliary contact, the red light also indicates that power is available to the shunt trip coil. This provides an indication that the shunt trip coil is ready to perform its function when required.

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CPSES Overvoltage Protection CPSES employs a solid state protection system. The overvoltage protection 1;; set at;115% of the nominal koltage of 48 vdc. Circuit malfunctiond resulting in overvoltage conditions will result ir. failsafe consequences of load removal including UV coil and parallel shunt trip actuation relay which will trip the breaker. Components in the added shunt trip circuitry have been selected based on their ability to perform their intended function up to a voltage as high as 115% of the nominal voltage.

3. " Verify that thh relays added for the automatic shunt trip function are within the capacity of their associated power supplies and thit the relay contacts are adequately sized to accomplish the thunt trip function. If the added relays are I

other than the Fotter and Brumfield MDR series relays (P/N 2383A38 or P/N 955655) recommended by Westinghouse, provide a description of the relays and their design specifications."

CPSES intends to use Potter and Brumfield relay P/N 955655 specified in the generic design.

4 " Verify that the shunt trip attachments and associated circuitry are/will be seismically qualified (i.e., be demonstrated to be operable during and after a seismic event) in accordance with the provisions of Regulatory Guide 1.100, Revision 1 which endorses IEEE Standard 344, and that all non-safety related circuitry / components in physical proximity to or associated with the automatic shunt trip function will not degrade this function during or after a seismic event."

The Westinghouse Owner's Group (WOG) is working with

, Westinghouse to obtain seismic qualification of shunt trip 1

CPSES attachments. CPSES will ensure that seismic qualification is provided for the added circuitry. As described in Paragraph 5.1.4 of the WOG Generic Design Package, the DS-416 reactor trip switchgear including drawout element is environdentally and seismically qualified in accordance with IEEE-323-1974

, and 344-1975. Non-class 1E circuitry is physically separated from Class 1E circuits.

5. " Describe the physical separation provided between the circuits used to manually initiate the shunt trip attachments of the redundant reactor trip breakers. If physical separation is not maintained between these circuits, demonstrate that faults within these circuits can not degrade both redundant trains."

Physical separation is maintained between redundant trains in the main control board, reactor trip switchgear, and reactor protection logic for shunt trip circuitry. Ways that this separation is maintained include:

1. Dual section manual reactor trip switches with metal barriers between redundant train switch decks.

2. Shunt trip attachments interposing relays and their associated terminal blocks mounted in separate metal enclosures.

3. Reactor protection logic outputs for energizing the shunt trip interposing relays are housed in existing separate metal enclosures.

4. Field cabling from the main control board and reactor protection logic to redundant Train A and Train B reactor trip switchgear are routed as Train A and Train B circuits.

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CPSES

5. Coil to contact isolation within reactor trip switchgear, ,

6. " Verify that the circuitry used to implement the automatic shunt trip function is Class 1E (safety related), and; that the procurement, installation, operation, testing, and maintenance of this circuitry will be in accordance with the t

quality assurance criteria set forth in Appendix B to 10 CFR Part 50."

When implemented, Class 1E criteria will be maintained for the added circuitry. Installation, operation, testing and maintenance of this circuitry will be in accordance with the existing quality assurance procedures which satisfy the requirements of Appendix B to 10 CFR Part 50.

7. " Verify that the components used to accomplish the automatic shunt trip function are designed for the environment where they are located."

The reactor trip switchgear is located in the Safeguards building. The environmental conditions are the same as enveloped by Table 1 of the Westinghouse Owner's Group Generic design package under "outside of containment-ventilated." The shunt trip mechanism will be qualified for such environment.

8. " Verify that the test procedure used to determine reactor trip breaker operability will also demonstrate proper operation of the associated control room indication / annunciation.

Station Procedure EGT-706 will be revised to include checking indications for operability. Red and green lights are provided on the main control board to indicate breaker position. A red light indicates the breaker is in the closed i

CPSES position and a green light indicates the breaker is in the open position.

4.4 I REACTOR TRIP SYSTEM RELIABILITY (IMPROVEMENTS IN MAINTENANCE AND TEST PROCEDURES FOR B&W PLANTS) v Since CPSES is of Westinghouse design, this section is not applicable, 4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

1. CPSES currently intends to implement the generic design modification as proposed by the Westinghouse Owner's Group and accepted by the Office of Nuclear Reactor Regulation (Letter from Mr. D. G. Eisenhut to Mr. J. J. Shepard dated August 10,1983), to provide automatic reactor trip system actuation of the breaker shunt trip attachments.

On-line functional testing of the reactor trip system, including independent testing of the shunt trip coil and undervoltage trip attachment, will be performed per revised station surveillance procedures.

2. Periodic, on-line testing will be provided by implementation of the new design discussed in Section 4.5.1, above.

3. CPSES endorses changes to the standard technical specifications recommended and justified by WCAP-10271 (and its Supplement #1), " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System."

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Attachment 1: Digital Parameters Included in the P-2500 Sequence of Events Report The following digital parameters are monitored for the Sequence of Events Report that is printed on the process computer's alarm printer any time

, that one of the parameters changes status. The printout occurs when the

~ number of parameter changes reaches 50, or after 1 minute has elapsed from the first parameter change. Abbreviations in this list are defined in Attachment 3.

RCL LO F & P8 CAUS RE RCL LO F & P7 CAUS RE TURB CH 1 GEN ST CKT PW OUT LO F TURE CH 2 GEN ST CKT PW OUT LO F TURB CH 1 GEN RT CKT PW OUT LO F TURB CH 2 GEN RT CKT PW OUT LO F TURB CH 1 GEN BUSH PH A PW LO F TURB CH 2 GEN BUSH PH A PW LO F TURE CH 1 GEN BUSH PH B PW LO F TURB CH 2 GEN BUSH PH B PW LO F TURE CH 1 GEN BUSH PH C PW LO F TURB CH 2 GEN BUSH PH C PW LO F STM CEN 1 LO LO L 4 PART RE STM GEN 1 LO LO L CAUS RE STM GEN 2 LO LO L 4 PART RE STM GEN 2 LO LO L CAUS RE STM GEN 3 LO LO L 4 PART RE STM GEN 3 LO LO L CAUS RE STEM GEN 4 LO LO L 4 PART RE STM GEN 4 LO LO L CAUS RE PRESSURIZER HI L & P7 CAUS RE TURE CH 1 MSR 1A HI L TURB CH 1 MSR 1B HI L TURE CH 2 MSR A HI L TURE CH 2 MSR B HI L TURB CH 1 PW TANK LO L TURB CH 2 PW TANK LO L l TURB CH I GEN TERM BOX WTR HI L ,.

TURE CH 2 GEN TERM BOX WTR HI L l

Paga 2 of 4 PWR RNG CHAN HI Q HI SP CAUS RE ;

PWR RNG CHAN HI Q LO SP CAUS RE INTERM RNG HI Q CAUS RE

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PWR RNG CHAN HI Q RATE CAUS RE SOURCE RNG HI Q CAUSE RE PRESSURIZER HI P CAUS RE PRESSURIZER LO P + P7 CAUSE RE PRESSURIZER LO P SI CAUSE RE SG 1 LO STM LINE P CAUSE RE SG 2 LO STM LINE P CAUSE RE SG 3 LO STM LINE P CAUSE RE SG 4 LO STM LINE P CAUSE RE CONTAINM HI P SI CAUSE RE TURB CH 1 COND LO VACULH TURB CH 2 COND LO VACULH TURB CH I LUBE OIL LO P TURB CH 2 LUBE OIL LO P CCW HX 01 OUTL P CCW HX 02 OUTL P TURB CH 1 CONTR REM MANUAL TURB CH 2 CONTR REM MANUAL RCL OVERTEMP N16 CAUS RE RCL OVERPWR N16 CAUSE RE TURB CH I GEN PW OUT HI T TURB CH 2 GEN PW OUT HI T RCP UNDER VOLT + P7 CAUSE RE REAC MANUAL TR 1 CAUS RE REAC MANUAL TR 2 CAUSE RE REAC MAIN TR BKR A REAC MAIN TR BKR B REAC AUX TR BKR A REAC AUX TR BKR B RCP UNDER FREQ + P7 CAUSE RE UNIT ON LINE TIE SF6 E3 BKR UNIT ON LINE TIE SF6 W3 BKR l

Page 3 of 4 TURB TRIP + P7 CAUS RE -

E SFTY INJ SET MANUAL 1 CAUS RE SFTY INJ SET MANUAL 2 CAUS RE TURB CH I GEN PW PUMP HI VIBR TURB CH 2 GEN PW PLYP HI VIBR IURB CH 1 GEN LOCKOUT RELAY TURB CH 2 GEN LOCKOUT RELAY TURB CH 1 GEN FIELD TOTAL LOSS TURB CH 2 GEN FIELD PARTIAL LOSS TURB CH 1 GEN PW TRITIUM HI CONT TURB CH 2 GEN PW TRITILH HI CONT TURB CH 1 T-G BRG HI VIBR TURB CH 2 T-G BRG HI VIBR TURB CH 1 SG HI HI L OR SIS TURB CH 2 SG HI HI L OR SIS TURB CH 1 REACTOR TURB CH 2 REACTOR SG FW PMP A TURB SG FW PMP B TURB COND PMP 01 BKR COND P9P 02 BKR HTR DRN PMP 01 BKR HTR DRN PMP 02 BKR l CIRC WTR PMP 01 BKR CIRC WIR PMP 02 BKR CIRC kTR PMP 03 BKR CIRCWhRPMP04BKR TP COOL kTR MAIN PMP BKR AUX BOILER DSL GEN FUEL OIL TR PMP l-A BKR DSL GEN FUEL OIL TR PMP l-B BKR DSL GEN FUEL DIL TR PMP 2-A BKR DSL GEN FUEL OIL TR PMP 2-B BKR DSL GEN 1 DSL GEN 2

Prg2 4 of 4 9

SSW PMP 01 BKR .

SSW PMP 02 BKR CCW PMP 01 BKR t

CCW PMP 02 BKR RCTR MAKE UP WTR MAIN PMP BKR RCTR MAKE UP kTR S/B PMP BKR

. . = ._.

Attachment 2: Analog Parameters Included in the P-2500 Post-Trip Report Theparameterslistedbelowareincludedintheprintoutontbeprocess computer line printer that is produced three minutes after a reactor trip or operator request. A value for each parameter is printed at 10-second

, intervals for the period from two minutes prior to the trip / request time until three minutes following the trip / request time. Abbreviations in this list are defined in Attachment 3.

(The parameters denoted by an asterisk are also included in a printout at 2.5 second intervals for the period from 10 seconds prior to the trip / request time until 10 seconds following the trip / request time.)

RCP4 SEAL INJECTION WTR F RCP3 SEAL INJECTION WTR F RCP2 SEAL INJECTION WTR F RCPI SEAL INJECTION WTR F RCLI 1 F RCL1 2 F RCLI 3 F STM GEN 1 FEED WTR IN 1 F STM GEN 1 FEED kiR IN 2 F STM GEN 1 STM OUT 1 F STM GEN 1 STM OUT 2 F RCL2 1 F RCL2 2 F RCL2 3 F STM GEN 2 FEED WTR IN 1 F STM GEN 2 FEED WTR IN 2 F l STM GEN 2 STM OUT 1 F STM GEN 2 STM OUT 2 F RCL3 1 F RCL3 2 F RCL3 3 F STM GEN 3 FEED WTR IN 1 F STM GEN 3 FEED hTR IN 2 F STM GEN 3 STM OUT 1 F SuGM3SMOW2F m

Psgo 2 of 8 RCL4 1 F ,-

RCL4 2 F RCL4 3 F

?

STM GEN 4 FEED WR IN 1 F STM GEN 4 FEED WTR IN 2 F STM GEN 4 STM OUT 1 F STM GEN 4 STM OUT 2 F TPCW PMP DISCH HDR TOTAL F T-G ROTOR PRIM WR INLET F SSW PMP 01 DISCH TO AUX BLDG F SSW PMP C2 DISCH TO AUX BLDG F STM GEN 1 NAR RNG 1 L STM GEN 1 NAR RNG 2 L STM GEN 1 NAR RNC 3 L STM GEN 1 NAR RNG 4 L STM GEN 1 WIDE RNG 1 L STM GEN 2 NAR RNG 1 L STM GEN 2 NAR RNG 2 L STM GEN 2 NAR RNG 3 L STM CEN 2 NAR RNG 4 L STM GEN 2 WIDE RNG L STM GEN 3 NAR RNG 1 L STM GEN 3 NAR RNG 2 L STM GEN 3 NAR RNG 3 L STM GEN 3 NAR RNG 4 L t

SIM GEN 3 WIDE RNG L STM GEN 4 NAR RNG 1 L STM GEN 4 NAR RNG 2 L STM GEN 4 NAR RNG 3 L STM GEN 4 NAR RNG 4 L STM GEN 4 WIDE RNG L PRESSURIZER 1 L PRESSURIZER 2 L PRESSURIZER 3 L PRESSURIZER LVL CONTROL SP

. i

Paga 3 of 8 CONDENSATE STORAGE TANK L SOURCE RNG DETECTOR 1 LOG Q SOURCE RNG DETECTOR 2 LOG Q

, INTERM RNG HI DETECTOR 1 LOG Q INTERM RNG HI DETECTOR 2 LOG Q Pk'R RNG CH 1 TOP DET Q (QUAD 4)

Pk'R RNG CH 1 BOT DET Q (QUAD 4)

Pk'R RNG CH 2 TOP DET Q (QUAD 2)

Pk'R RNG CH 2 B0T DET Q (QUAD 2)

Pk'R RNG CH 3 TOP DET Q (QUAD 1)

Pk'R RNG CH 3 BOT DET Q (QUAD 1)

Pk'R RNG CH 4 TOP DET Q (QUAD 3)

Pk'R RNG CH 4 BOT DET Q (QUAD 3)

Pk'R RNG CHANN% 1 Q (QUAD 4)

• Pk'R RNG CHANNEL 2 Q (QUAD 2)

• Pk'R RNG CHANNEL 3 Q (QUAD 1)

• Pk'R RNG CHANNEL 4 Q (QUAD 3)

• TB FIRST STAGE 1 P

• TB FIRST STAGE 2 P

• STM GEN 1 STM OUT 1 P STM GEN 1 STM OUT 2 P STM GEN 1 STM OUT 3 P STM GEN 2 STM OUT 1 P STM GEN 2 STM OUT 2 P STM GEN 2 STM OUT 3 P STM GEN 3 STM OUT 1 P STM GEN 3 STM OUT 2 P STM GEN 3 STM OUT 3 P STM CEN 4 STM OUT 1 P STM GEN 4 STM OUT 2 P STM GEN 4 STM OUT 3 P PRESSURIZER 1 P PRESSURIZER 2 P PRESSURIZER 3 P PRESSURIZER 4 P u__ __

Pags 4 of 8 CONTAINMENT 1 P  ;

CONTAIhMENT 2 P CONTAINMENT 3 P HEATER 1A SHELL P HEATER IB SHELL P HEATER 2A SHELL P HEATER 2B SHELL P HEATER 3A SHELL P HEATER 3B SHELL P HEATER 4A SHELL P HEATER 4B SHELL P HEATER SA SHELL P

HEATER SB SHELL P HEATER 6A SHELL P HEATER 6B SHELL P CONDENSATE PMP 01/02 DISCH P COND A SHELL ABS P COND B SHELL ABS P HP TURB EXH P LP TURB INLET STM LINE 1 P LP TURB INLET STM LINE 2 P MAIN STM LOOP 1 AT STOP VLV P MAIN STM LOOP 2 AT STOP VLV P MAIN STM LOOP 3 AT STOP VLV P MAIN STM LOOP 4 AT STOP VLV P CCW PMP 01 DISCH P CCW PMP 02 DISCH AUX COND A SHELL ABS P AUX COND B SHELL ABS P SG FW PMP A DISCH P SG FW PHP A DISCH P SSW PMP 01 DISCH P i

SSW PMP 02 DISCH P GENERATOR GROSS MW

• l

Pags 5 of 8 AUX TRANSFORMER 1 UT MW .

GENERATOR MVAR IN CORE T JUNCTION BOX A 1 T

• t IN CORE T JUNCTION BOX A 2 T

• IN CORE T JUNCTION BOX B 1 T

• IN CORE T JUNCTION BOX B 2 T

• RCP1 SEAL WIR 1 OUT

• RCP2 SEAL WTR 1 OUT

• RCP3 SEAL WR 1 OUT

• RCP4 SEAL WIR 1 OUT

• RCPI KIR STATOR WINDING T

• RCPI MTR UPPER RADIAL BRG T

• RCPI MIR UPPER THRUST BRG T

• RCPI MTR LOWER RADIAL BRG T

• RCPI MTR LOWER THRUST BRG T

• RCPI LOWER SEAL kiR BRG T

• RCL2 1 TAVG RCL2 WIDE RNG COLD LEG T STM GEN 2 FEED WIR IN T RCL2 WIDE RNG HOT LEG T RCL3 1 TAVG RCL3 WIDE RNG COLD LEG T STM GEN 3 FEED kiR IN T RCL3 WIDE RNG HOT LEG T RCL4 1 TAVG RCL4 WIDE RNG COLD LEG T STM GEN 4 FEED kiR IN T RCL4 WIDE RNG HOT LEG T PRESSURIZER STM T RC TREF

• RCL HIGHEST TAVG (AUCTIONEER)

'RCL1 DT - N16 POWER RCLI OVERPWR N16 SP RCL 1 OVERTEMP N16 SP I

i

P2gs 6 cf 8 RCL2 DT - N16 POWER .

s RCL 2 OVERPOWER N16 SP RCL 2 OVERTEMP N16 SP

?

RCL3 DT - N16 POWER RCL 3 OVERPWR N16 SP RCL 3 OVERTEMP N16 SP RCL4 DT - N16 POWER RCL 4 OVERPWR N16 SP RCL 4 OVERTEMP N16 SP RCL HIGHEST N16 (AUCTIONEER)

T-G HP CASING TOP FRONT T T-G HP CASING BOTTOM FRONT T T-G HP CASING TOP REAR T T-G HP CASING BOTTOM REAR T T-G LUBE OIL CLR IN T T-G LUBE OIL CLR OUT T COND HTR DRN PMP DISCH T LP TURB INLET STM LINE 1 T LP TURB INLET STM LINE 2 T LP TURE INLET STM LINE 3 T LP TURB INIST STM LINE 4 T MAIN STM LOOP 1 AT STOP VLV T MAIN STM LOOP 2 AT STOP VLV T MAIN STM LOOP 3 AT STOP VLV T MAIN STM LOOP 4 AT STOP VLV T T-G COLD GAS CLR A T T-G COLD CAS CLR B T l T-G COLD GAS CLR C T T-G COLD GAS CLR D T T-G HOT CAS CLR B AND C T T-G HOT GAS CLR A AND D T T-G HOT AIR RECT WHEEL T T-G HOT AIR MAIN EXCITER T T-G COLD CAS EXCITER T T-G SEAL OIL CLR OUT H2 T

Psg9 7 of 8 T-G SEAL OIL CLR OUT AIR T .

T-G SEAL OIL DRN TURB END T T-G SEAL OIL DRN EXC END T T-G UPSTREAM PRIM WTR T T-G STATOR PRIM kTR OUTL T T-G ROTOR PRIM WTR OUTLET T GEN IG WIND PH A HOT SPOT X T GEN IG WIND PH B HOT SPOT Y T GEN IG WIND PH C HOT SPOT Z T TRANSF XSTI WIND X HOT SPOT T TRANSF XST1 WIND Y HOT SPOT T TRANSF XST1 OIL T TRANSF XST2 WIND X HOT SPOT T TRANSF XST2 WIND Y HOT SPOT T TRANSF XST2 OIL T TRANSF IUT WIND X HOT SPOT T TRANSF IUT WIND & HOT SPOT T TRANSF IUT OIL T TRANSF 1MII WIND HOT SPOT T TRANSF IMIl OIL T TRANSF IMT2 WIND HOT SPOT T TRANSF 1MT2 OIL T FP A INBOARD RADIAL BRG T FP A INBOARD RADIAL BRG T FP A OU1 BOARD RADIAL BRG T FP A A/F THR BRt; T l

FP A I/F THR BRG T FP A RETURN LUBE OIL T FP A OUTER CASE MTL TOP T FP B HP BRG T FP B LP BRG T FP B INBOARD RADIAL BRG T FP B OUTBOARD RADIAL BRG T FP B A/F THR BRG T FP B I/F THR BRG T

I d

Pagm 8 of 8 I

'i FP B RETURN LUBE OIL T -

FP B OUTER CASE TOP MTL T FP PMP COMMON DISCH T t

GEN IG PH AB VOLTAGE GEN IC PH BC VOLTAGE GEN IG PH CA VOLTAGE T-G PRIM hTR OXYGEN CONTEhT GEN IG PH A CURRENT GEN IG PH B CURREhT GEN IG PH C CURRENT k

c '^"

I I F o

f-

[

Attachment 3: Definitions of Abbreviations Abbreviations Definitions A/F -

Active Face ABS -

Absolute

, AUX -

Auxiliary BKR -

Breaker B0T - Bottom BRG -

Bearing BUSH -

Bushing CAUS RE -

Causes Reactor Trip CCW - Component Cooling Water CH -

Chanel CHAN -

Chanel CIRC - Circulating CKT - Circuit CLR - Cooler COND -

Condensor CONT - Content CONTAINM -

Containment CONTR - Control COOL -

Cooling DET - Detector DISCH -

Discharge DRN - Drain DSL -

Diesel DT - Temperature Difference (Delta T) i EXC - Exciter EXH - Exhaust F - Flow FREQ -

Frequency FW -

Feedwater i CEN -

Generator 1

EDR -

Header HI - High HP -

High Pressure 1

s

, Pcga 2 of 4

~ + ~ . . , Abbreviations _ Definitions HTR -

Heater

, ' HX -

Heat Exchanger

, H2 - Hydrogen I/F -

Inactive Face IN -

Inlet INJ -

Injection INTERM -

Intermediate L - Level LO -

Low LOG -

Logrithmic LP -

Low Pressure m LVL -

Level s MSR - Moisture Separator and Reheater MIL -

Metal MTR - Motor MVAR -

Megavar JGi -

Megawatts NAR -

Narrow N16 -

Nitrogen 16

, - OUT -

Outlet g OUTL -

Outlet OVERPWR -

Overpower OVERTEFT -

Overtemperature P -

Pressure

. PART RE -

k' A signal necessary but not sufficient to cause a reactor trip.

PH -

Phase PMP -

Pump

.. PRIM -

Primary PW - Primary Water s

PWR -

Power P7

- At Power Permissive Switch s .

P8 -

3-Loop Flow Permissive Switch

, Q -

Power

('

)

Pags 3 of 4 l

Abbreviations Definitions 1

QUAD -

Quadrant of Core RC - Reactor Coolant t RCL - Reactor Coolant loop RCP -

Reactor Coolant Pump RCTR -

Reactor REAC - Reactor RECT -

Rectifier REM -

Remote RNG -

Range RT -

Rotor S/B - Standby SG - Steam Generator SI -

Safety Injection SIS -

Safety Injection Signal SP -

Setpoint SSW - Site Service Water ST -

Station STM -

Steam T -

Temperature T-G - Turbine Generator TB - Turbine TERM - Terminal THR - Thrust TP -

Turbine Plant TPCW -

Turbine Plant Cooling Water TR -

Trip or Transfer TRANSF - Transformer TURB -

Turbine UT -

Unit Transformer VIBR -

Vibration t VLV -

Valve t

j VOLT -

Voltage I

WIND -

Windings i

l

. Pcga 4 cf 4 Abbreviations Definitions WIR -

Water t

i i

l l

d

~ .,

m m

f =

Y

.N I!?

XX:XX SEQUENCE OF EVENTS RECORD FIRST EVENT A*'H10 M20 S40

• • SYSTEM CAPACITY EXCEEDED **

p ** EVENTS WITH OUESTIONABLE TIMING OR SEQUENCING ARE INDICATED BY ASTERISKS

• j j Y2803D 4160 V SFDG. BUS SB- UV RELAY LOW *C 0 g Y2803D 4160 V SFDG. BUS SB UV RELAY NORMAL

• C 1

{ Y2804D 4160 V SFDG. BUS SB UV RELAY NORMAL

• C 1 m Y2805D 4160 V SFDG. BUS SB UV RELAY NORMAL

• C 1 3 p Y2805D 4160 V SFDG. BUS SB UV RELAY LOW C 15 C

Y2804D 4160 V SFDG. BUS SB UV RELAY d

w L  %

f Y2803D 4160 V SFDG. BUS SB UV RELAY LOW LOW C 15 C 15 5

e, - 8

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[ Y2803D 4160 V SFDG. BUS SB UV RELAY NORMAL C 25 Y2802D '4160 V SFDG. BUS SA UV RELAY NORMAL C 25 93 Y2801D 4160 V SFDG. BUS SA UV RELAY NORMAL C 25 XX:XX END OF SEQUENCL OF EVENTS RECORD R

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• 161 St #

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I Figure 3-11a. Post Trip TBX/TCX 16055/16048 3 -14 TPSI66

-- ATTACHMENT 6 COMANCHE PEAK STEAM ELECTRIC STATION OPERATIONS DEPARTMENT ADMINISTRATION MANUAL D -

/-: b*oik '

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~~ E $ l l,g ,.??*n eR-

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1. i. e Nll.J. ,3, $y N.[l' f .~ ,.i' , . . . .

%f e POST TRIP REVIEW EVALUATION PROCEDURE NO. ODA-108 REVISION NO 0 SUBMITTED BY:

DATE:

OPERATIONS SUPERINTENDENT APPROVED BY:

DATE:

MANAGER, PLANT OPERATIONS

CPSES ISSUE DATE

• PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 2 0F 18 1.0 Purpose This procedure describes the steps necessary for completing the Post Trip Review evaluation.

J.0 Applicability This procedure applies to all unplanned manual or automatic actuations of the Reactor Protection System or planned actuation where abnormal events or malfunctions have occurred. This procedure becomes effective after initial fuel load.

3.0 Definitions None i

4.0 Instructions Note: This report may not substitute for a problem report. This evaluation should be attached as a supplement to the problem report.

I 4.1. The " Identification of Event" Section 1 of Attachment 1 shall contain all preliminary information regarding the trip. It shall be completed as follows by an STA or an individual designated by the Shift Supervisor.

4.1.1 List the time and date the trip occurred.

4.1.2 Explain the apparent cause of the trip.

Note: This will not necessarily agree with the final conclusion of the report.

4.1.3 The " Description of Transient", shall include equipment that started during the trip and the operators response to i

the incident.

4.1.4 List all activities that were in progress prior to the trip.

Example:

1) Diluting

2) Borating

3) Starting a pump i

1 i

l 1

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

. CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 I

- t

\

POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 3 0F 18 4.1.5 List all personnel present during the trip. As a minimum the shift complement should be indicated. .

4.2 Section II, " Preliminary Investigation" will list parameters in the plant just prior to the trip. The information can be e obtained from the individual control board recorders.Section II and all subsequent sections shall be completed by an Operations Department engineer.

4.3 Section III, " Surveillance Testing or Maintenance" shall include any equipment that was out-of-service due to Maintenance activities or any surveillance testing activities. The following should be given.

4.3.1 List the system or equipment that was affected.

4.3.2 List maintenance or testing activities in progress.

4.3.3 Explain in general what was being performed at the time of the trip.

Example:

1) Bypassing a channel

2) Jumpering leads 4.3.4 List all clearances that were performed within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the trip.

4.4 Section IV, " Plant Response" will contain all information regarding any actuations of safety systems and any failures that resulted. This section should be completed as follows:

4.4.1 List if the trip was automatic or manual. If automatic, give the signal that tripped the reactor.

4.4.2 If an ECCS actuation occurred, then list flows and -

isolation times for each portion of the ECCS.

4.4.3 If any equipment failed to operate correctly, it should be stated at the end of this section.

4.4.4 In the remaining portions of this section, explain if any pressure reliefs occurred and if the response was normal.

Also, explain if any of the listed parameters were abnormal.

Note: Normal refers to equipment actuating and resetting at the correct setpoints.

)

. CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 ,

. l POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 4 0F 18 4.5 Section V, " Trip Review" will contain a check off portion and fill in portion for recorders and typewriter printouts. It shall be completed as follows:

4.5.1 In the first part list any abnormal findings pertaining to y the sequence of events, trend, alarm summary or RMS typewriter printouts.

4.5.2 List all recorders reviewed and explain any findings.

4.5.3 As a minimum, the following recorders should be reviewed:

e NR-45 e Steam Generator Level i

e Steam Flows e Tavg e Tc e Pressurizer Pressure e Pressurizer Level e Seal Recorders for each RCP Note: A copy of all recorders reviewed and typewriter printouts should be attached to the report.

4.6 Section VI, " Trip Investigation" shall contain any information that might be helpful in ar' riving at a definite conclusion as to the cause of the trip. The following information should be included in this section:

4.6.1 All information obtained in Sections 1 through 5 shall be reconstructed in a chronological description of the event and then compared to the sequence of events printout to determine a clear cause of the trip.

4.6.2 Where appropriate the reconstruction shall be compared with similar transients described in the FSAR or Previous Post Trip Review Evaluations.for similar Trips.

4.6.3 List all controls that were manipulated after the trip in order to stabilize the transient. &

4.6.4 If any radiation monitors alarmed, give an explanation of what was done to correct the situation.

4.6.5 List the maximum and minimum values for the parameters listed in this part.

Note: The information can be obtained from the respective recorders.

' CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 5 0F 18 4.6.6 List all deficiencies noted on system responses or particular components failing to move to demanded positions.

4.7 Section VII, " Preliminary Safety Assessment" shall contain specific parameters that pertain to plant safety. The following shall be included:

e RCS Pressure e RCS Cooldown rate e RCS Inventory e Pressurizer Level 4.8 Section VIII, "The Conclusion" shall summarize all information obtained and arrive at a conclusive reason for the trip. If there is any recommendation for a follow-up investigation on a particular component or system it should be noted in this section.

4.9 After completion, the Operations Department Engineer shall sign and date.

4.10 The Shift Supervisor shall then with the Operations Department Engineer review the entire evaluation.

4.11 This review shall be completed to ensure the following criteria j has been met:

4.11.1 The report was properly prepared and evaluated.

4.11.2 The cause of the event has been clearly identified and corrected.

i 4.11.3 All required equipment and systems functioned as designed during the event and recovery or any abnormalities have been corrected as required by Technical Specifications.

4.12 If the above criteria is met, and the shift supervisor concurs with the evaluation, he shall sign and date it and contact the Operations Superintendent and recomuend a start-up authorization.

4.13 The Operations Superintendent after consulting with the shift supervisor shall if satisfied give approval for start-up.

4.14 If all of the above criteria is not met or the Shift Supervisor does not agree with the Operations Department Engineer's evaluation of the trip, then approval for start-up shall not be recommended.

4.15 All reasons for not tecommending start-up shall be documented in section 8 of Attach 2ent 1.

i I

CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 6 0F 18  ;

I 4.16 The Operations Engineer shall then be notified so that a special committee or another Operations Department Engineed can be l selected to do an additional investigation.  !

4.17 Following completion of the evaluation the report shall be t

forwarded to the Operations Engineer who shall be responsible for the following:

4.17.1 Maintain a log of all Post Trip Evaluations by date and unit.

4.17.2 Review the report and assign a number to it. The number should be the current year followed by a sequential number for each evaluation.

4.17.3 Assign the appropriate personnel to any follow-up analysis recommended on the report.

4.17.4 Ensure all follow-up analysis are adequately resolved in a timely manner.

4.17.5 Evaluate the report to determine lessons learned for operator and/or plant staff training, procedure changes or plant modifications.

4.18 The completed Post Trip Evaluation form shall be maintained in accordance with ODA-104, " Operations Department Document l

Control."

5.0 References 5.1 NRC Generic Letter 83-28 5.2 (INPO OP-211 Post Trip Reviews) 6.0 Attachments CPSES Post Trip Review Evaluation I

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CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 7 OF 18 ATTACHMENT 1 ,

PAGE 1 0F 12

• CPSES

, POST TRIP REVIEW EVALUATION OPFRATIONS DEPARTMENT DATE:

NUMBER:

Unit:

I. Identification of Event A) Time of Trip: Date:

B) Apparent cause of transient:

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Description of transient.(include any observed abnormal responses):

1 D) Activities in progress prior to trip:

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ODA-108-1

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. CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 8 0F 18 ATTACHMENT 1 PAGE 2 0F 12 -

Shift Supervisor Assistant Shift Supervisor Reactor Operator ~

Relief Reactor Operator Shift Technical Advisor Other Personnel Completed by: Date: Time:

Reviewed by: Date: Time:

Shift Supervisor II. Preliminary Investigation Initial Conditions:

a) Reactor Power b) RCS Pressure c) Turbine Load d) MF Pumps Operating e) Rods (Auto / Manual) f) Rod position g) Pzr Pressure Control (Auto / Man) h) Pzr Level Control (Auto / Man)

1) Chg. Pump In Service LOOP (1) (2) j) Temperature Tc (3) (4) k) Temperature Tavg

1) Steam Gen. Level m) Stm Flow / Feed Flow Deviation n) F/W Vivs (Auto / Man) o) Bypass FW Valves (Auto / Man) p) #1 Seal Leak Off q) #2 Seal Leak Off ODA-108-la t______ _-

' CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 9 0F 18 ATTACHMENT 1 PAGE 3 0F 12 Remarks / Comments:

t III. Surveillance Testing or Maintenance Activities in Progress System Testing / Maintenance Remarks 1

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CPSES ISSUE DATE

[ OPERATIONS DEPARTMENT ADMINISTRATION MANUAL PROCEDURE NO.

ODA-108

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POST TRIP REVIEW EVALUATION REVISION NO. O

, PAGE 10 0F 18 ATTACHMENT 1 PAGE 4 0F 12 -

Remarks / Comments:

t IV. Plant Response A) Reactor Protection System Actuation (Auto / Manual)

First Out Annunciator Actuation Time (Typewriter)

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B) ECCS Actuation (Yes/No)

(Auto / Manual)

(Actuation Signal)

1) Charging Pumps Actuation Flow _

Isolation Time

2) Safety Injection Pumps Actuation Flow Isolation Time

3) RER Pumps Actuation Flow Isolation Time ~~

4) Containment Spray Actuation Flow Isolation Time

5) Accumulators Discharge (Yes/No)

6) Diesels Start (Yes/No)

7) Diesel Bkr Closed (Yes/No)

ODA-108-1c 1

CPSES ISSUE DATE PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 11 0F 18 i

ATTACHMENT 1 PAGE 5 0F 12 i l

Did all equipment perform properly? Were all flows proper?

e (Yes/No)

If no, explain

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C) Did any Pzr safeties lift? (Yes*/No)

D) Did PORV's open? (Yes*/No) l E) Did Main Steam Safeties Lift? (Yes*/No)

! F) Did safeties reseat properly? (Yes/No*)

G) Did Steam Dumps function properly? (Yes/No*)

H) Did S/G Atmospherics open? (Yes*/No)

• If question was answered in this manner, give a full explanation:

ODA-108-1d l

CPSES ISSUE DATE PROCEDURE NC.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 i --

POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 12 0F 1R ATTACIDiENT 1 i

PAGE 6 0F 12  ; '

I) Control Systems t

1) Turbine Runback (Yes/No)

Reactor Power from to  ; Time from to

2) Turbine Trip (Auto / Manual)

First Out Annunciator

3) RCS Pressure Control

) a) Was heater response normal? (Yes/No) b) Was spray response normal? (Yes/No) c) Was level response normal? (Yes/No)

If no, explain:

4) I&C Systems Auto Response a) S/G Pressure Control Normal (Yes/No) b) S/G Level Control Normal (Yes/No)

If no, explain:

V. Trip Review The following has bee.n reviewed:

A) Trend Typewriter B) Alarm Summary Typewriter C) Sequence of Events Typewriter D) RMS Typewriter ODA-108-le I

CPSES ISSUE DATE

. PROCEDURE NO.

OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 i .

POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 13 0F 18 ATTACRMENT 1 PAGE 7 0F 12  ;

, Attach copy of printouts and explain any findings:

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Recorders Reviewed:

Recorder Findings I I i

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I Remarks /Comunents:

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CPSES ISSUE DATE PROCEDURE NO.

, OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO, O PAGE 14 0F 18 ATTACHMENT 1 PAGE 8 0F 12 ,

VI. Trip Investigation

, A) Chronological Description of Event Time Event Description Event compared with FSAR TRANSIENT Page Number Previous Trip on /

Date Time B) Manual Actions Were any controls taken from auto to manual? (If yes, give reason and explain)

ODA-108-Ig

, . CPSES ISSUE DATE PROCEDURE NO.

. OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 15 0F 18 ATTACHMENT 1 PAGE 9 0F 12 .

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VI. Tr% Investigation t

j , C) Response to Radiation monitor alarms.

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1) RCS Pressure

2) Pzr Level

3) Subcooling Margin

4) Tc

5) Stm Gen Pressure

6) Stm Gen Level If available, attach transient plots of pertinent plant parameters.

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OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 16 0F 18 ATTACHMENT 1 PAGE 10 0F 12  ;

E) Identification of systems with inadequate performance (discuss the nature of the deficiency).

t F) Discussion of any unexpected aspect of transient behavior.

VII. Preliminary Safety Assessment A) RCS pressure remained above setpoint for Automatic SI Activation. (Yes/No)

B) RCS pressure remained below setpoint for pressurizer code safety valve actuation. (Yes/No)

C) RCS temperature decrease was less than Tech Spec limit (100

• F/Hr. ) . (Yes/No)

D) Reactor Coolant was contained within the primary RCS and PRT. (Yes/No)

E)

Indicated Pressurizer level remained on scale.

(Yes/No)

ODA-108 I l _ -. ,-

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. OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 17 0F 18 ATTACHMENT 1 PAGE 11 0F 12  ;

If any answer is no, explain:

t VIII. Conclusion A) Trip

Conclusion:

B) Recommend Start-Up Yes [ ] No [ ]

Completed by: Date Time Engineer Reviewed by: -

Date Time Shift Supervisor Approval For Start-Up Granted by: Date Time Operations Supt.

ODA-108-lj

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CPSES ISSUE DATE PROCEDURE NO.

? OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-108 o

POST TRIP REVIEW EVALUATION REVISION NO. O PAGE 18 0F 18 ATTACHMENT 1 PAGE 12 0F 12 .

REPORT REVIEW t

Recommendation for need for any follow-up analysis Reviewed By: Date: Time:

OPERATIONS ENGINEER l

ODA-108-Ik

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