ML20081M342

Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Response Addresses Status of Current Conformance W/Positions in Generic Ltr

ML20081M342
Person / Time
Site:	Point Beach
Issue date:	11/07/1983
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, Eisenhut D Office of Nuclear Reactor Regulation
References
GL-83-28, TAC-52786, TAC-52787, TAC-53186, TAC-53187, TAC-54096, TAC-54097, NUDOCS 8311170222
Download: ML20081M342 (22)

Text

t ntsconsm Electnc mencourany 231 W. MICHIGAN, P.o. BOX 2046, MILWAUKEE. WI 53201 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U.S. NUCLEAR REGULATORY COMMISSION Washington, D.C. 20555 Attention: Mr. D. G. Eisenhut, Director Division of Licensing Gentlemen:

DOCKETS 50-266 and 50-301 9ESPONSE TO GENERIC LETTER 83-28 REQUIRED'AClIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On July 11, 1983, we received generic letter 83-28 entitled " Required Actions Based on Generic Implications of Salem ATWS Events" which was dated July 8, 1983. Generic letter 83-28, and the enclosure to the letter, presented the Commission's requirements for intermediate-term actions to be taken by licensees as a result of the NRC review of the Salem anticipated transient without scram (ATWS) events. These actions were characterized as addressing i issues related to reactor trip system reliability and general management

! capability. However, many of the actions discussed in the enclosure were related to all safety-related systems within the plant.

l l Licensees were required pursuant to 10CFR 50.54(F) to furnish a l

report on the status of current conformance with the positions contained in generic letter 83-28 and to present plans and schedules for any improvements for conformance with the positions. The attachment to this letter provides this information for the Point Beach Nuclear Plant Units 1 and 2 in an item by item format. You will note that as suggested in your letter, our actions i regarding some of these items are dependent upon the results of our participa-tion in several Owners' Group activities.

l 8311170222 831107

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Mr. H. R. Denton -

2- November 7, 1983 We trust the attached discussions are responsive to your action requirements. Should you have any questions regarding our proposed activities or the schedules for accomplishment of those activities, please let me know.

Very truly yours, fg$

C.W. Fay Vice President-Nuclear Power Enclosures Copies to NRC Resident Inspector Subscribpd and sworn to before me this 8Hs day of November 1983.

Notary Public, State of Wisconsin.

My Commission exp h [5 pers awe /I.

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O RESPONSE TO NRC GENERIC LETTER 83-28 POINT BEACH NUCLEAR PLANT l

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1.1 Post-Trip Review

' Activities currently performed at PBNP which constitute post-trip review of unplanned reactor shutdowns are being reevaluated. .It is anticipated that modifications to our current practices will be instituted that will result in a post-trip review procedure reflecting the guidelines of the INPO draft good practice OP-211 entitled, " Post-Trip Review." The modified procedure will be in place by February 29, 1984.

A description of current practices associated with post-trip review follows:

1.1.1 Criteria for Restart NOTE: RESTART MEANS ATTAINMENT OF CRITICALITY PER PLANT OPERATING PROCEDURES.

1.1.1-1 The cause of the trip must be known and have been corrected. If these conditions are not met, a Manager's Supervisory Staff review of the event is necessary -

before any restart activities are resumed.

1.1.1-2 Equipment which did not function properly during and following the transient must be repaired or a determi-nation made by the Duty Shift Superintendent and Duty &

Call _ Superintendent that the equipment is not required by Technical Specifications er r,ecessary for safe and reliable operation subsequent to restart.

1.1.1 ~ Equipment'which is necessary for critical operation under Technical Specifications is verified operable by a precriticality checklist.

1.1.1-4 Permission to restart must be obtained from the Duty &

Call. Superintendent.

1.1.2 Responsibilities and authorities of personnel who perform the review:

1.1.2-1 The Duty Shift Superintendent is responsible for the safe operation of PBNP Units 1-and 2. He has the authority to maintain either unit in a shutdown condition until he is satisfied that it is safe to change the condition of the unit. He.is assisted in making necessary decisions of this nature by the Duty &

Call Superintendent and Duty Technical Advisor.

1.1.2-2 The Duty _& Call Superintendent is responsible to act as an advisor to the Duty Shift Superintendent on matters relating to safe operation of PBNP Units 1 and 2. He L

Q may make a determination that further evaluation by the Manager's Supervisory Staff is necessary before the condition of a shutdown unit is changed.

1.1.2-3 The Duty Technical Advisor is responsible to provide to the Duty Shift Superintendent and Duty & Call Superintendent a diagn'osis of plant response during off-normal events.

L 1.1.3 Necessary Training and Qualifications l 1.1.3-1 The Duty Shift Superintendent must have, as a minimum, 5 years power plan't experience of which one year is in a nuclear power plant and hold a Senior Reactor Operator's j License. It should be noted that there are 8 Duty Shift Superintendents employed and each has more than 13 i years' experience working at PBNP in an operating crew.

l 1.1.3-2 The Duty & Call Superintendent is an indiv.idual l appointed by the Manager --PBNP, taking into

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1. Past SRO license and experience.

2. Current SRO license at PBNP.

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3. Professional degree and extensive nuclear power l

plant experience.

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( 4. Member of Manager's Supervisory Staff as defined in Technical Specifications.

It should be noted that currently there are 5 Duty &

Call Superintendents, 4 currently hold and one has held a.SRO license at PBNP. All have greater than 4.5 years working experience at PBNP.

1.1.3-3 The Duty Technical Advisor must have a bachelor's degree in an engineering or physical science discipline and 18 months of nuclear power plant experience of which 12 months experience must be at PBNP as well as success-

-fully completing the Duty Technical Advisor training program. It should be noted that there are currently 15 Duty Technical Advisors who meet these requirements.

1.1.4 Sources of Plant Information Necessary for Review Following any unscheduled reactor shutdown, a post-trip data package is put together for review by the operating crew. It consists of the following:

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I l'. Strip chart recordings for primary system parameters, t

2. Strip chart recordings for secondary system parameters, 7

3. Sequence of event listing from plant process control,

4. Analog trend post-trip review from plant process computer,

5. Operator observation of alarms (including first outs),

instrumentation displays including recorders, equipment status lights as well as first hand knowledge of equipment which has not functioned as required by emergency operating procedures or other plant operating requirements.

1.1.5 Methods and Criteria for Comparing Event Information With Expected Plant Response From the sources of information listad in 1.1.4 above, an evaluation of the transient is made by the Duty Shift Superintendent and the Duty Technical Advisor. This evaluation

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' must satisfy the restart criteria outlined in Section 1.1.1 above before any decision to restart can be made. The Duty Shift Superintendent. reviews his decision with the Duty & Call Superintendent who will concur if he believes the' restart i criteria are' met. The Duty & Call Superintendent will contact the Manager-PBNP, if he is available, to inform him of the

, circumstances and any decision regarding restart.

1.1.6 Criteria for Determining Need for Independent Assessment The determination that an independent assessment of the event by the Manager's Supervisory Staff can occur at three levels in the event review process:

1. The Duty Shift Superintendent is not satisfied that the restart criteria of 1.1.1 above are met.

2. The Duty & Call Superintendent is not satisfied that the restart criteria of 1.1.1 above are met.

3. The Manager - PBNP, if available, is not satisfied that the restart criteria are met.

', The physical evidence necessary to support an independent analysis i

of an event in which the cause is not immediately obvious is retained in accordance with plant practice and procedure.

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1.1.7 Systematic Method to Assess Unscheduled Reactor Shutdowns Although 'it is clear that current plant. practices do provide a systematic method for the assessment of unscheduled reactor shutdown, the present practice will be reviewed as stated above.

This-review, taking into consideration the draft INPO good practice

" Post-Trip Review", will be completed by February 29, 1984.

1.2 ' Post-Trip Review Data and Information Capability 1.2.1 Sequence of Events Capability 1.2.1-1 Description of Equipment This capability is provided by the plant computer system (a Westinghouse P-250 computer). There are currently separate computers for e;.ch unit.

1.2.1-2 Parameters Monitored A list of the sequence of events parameters monitored by the current computer system is provided in Attachment A.

The same parameters are monitored for both units.

1.2.1-3 Time Discrimination The time discrimination between events on the current computer system is 1/60 of a second, (16-2/3 msec).

1.2.1-4 Format Sequence of events data is output to the alarm type-writer. The output starts out with a line listing the time in hours, minutes and seconds of the first event. Each event is then output with its address, point description, status (open or close) and the cycle count (in 60ths of a second) since the first event. The data is output after either 25 events have occurred or after one minute has elapsed.

1.2.1-5 Data Retention The computer can collect data for a second " sequence of events" while the first is being output. When printout of the second sequence begins, data collection resumes for the next sequence. Once data is output to the alarm typewriter, it is no longer retained by the computer.

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1.2.1-6 Power Sources The Unit 1 computer power supply is 2831 (a nonvital motor control center supplied from Unit 2 803 safeguards 480v bus) and the Unit 2 computer power supply is 1831 (a nonvital motor control center supplied from Unit 1 B03 safeguards 480v bus).

1.2.2 Time History of Analog Variables (Post-Trip Review) 1.2.2-1 Equipment Description This capability is provided by the plant computer system. There are currently separate computers for each unit.

1.2.2-2 Parameters Monitored A list of the post-trip review parameters monitored by the current computer system is provided in Attachment B.

There is a small group sampled every 2 seconds and another larger group sampled every 8 seconds. Parameters which indicate power mismatched between the primary and secondary system were selected for the 2-second group.

In addition to the critical monitored parameters, several other parameters are monitored in the plant which provide more detailed information on these safety systems and other plant systems necessary to take the plant to a safe shutdown condition.

1.2.2-3 Duration of Time History l

l l The parameters in the 2-second group are collected for i approximately 8 seconds before the trip and 8 seconds after the trip. The parameters in the 8-second group are collected for approximately 2 minutes before the trip and 3 minutes after the trip.

1.2.2-4 Format The post-trip review data is output to a typewriter with l a column for each parameter. The pretrip data starts l off the column and is separated from the post-trip data by a line listing the trip time. The actual time of data collection for sub groups of parameters is included with each line of data.

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1.2.2-5 Data Retention Data collection is suspended while the printout takes place. Once data is output to the typewriter, it is no longer retained by the computer.

1.2.2-6 Power Sources See 1.2.1-6 above.

1.2.3 The following is a brief summary of the information that is available in the control room and is utilized in the evaluation of plant behavior during unplanned reactor shutdowns:

1. Rod position indicators,

2. Reactor trip breaker status lights,

3. Nuclear Power indicators and recorders including overpower and delta flux recorders,

4. AT and AT setpoints indicators and recorders,

5. . Pressurizer pressure indicators and recorder,

6. RCS wide range pressure recorder,

'7. RCS T hot and T cold temperature recorder,

8. Pressurizer level indicators and recorder,

9. Pressurizer level program recorder,

10. RCS T,yg and T ref recorder and indicators,

11. Various pressurizer system temperature indicators,

12. Various CVCS system parameter indicators and recorders including boration and dilution flow recorders,

13. First-out alarms for reactor trips, safeguards actuation and turbine trips,

14. Reactor protection and safeguards system bistable status lights,

15. Steam generator. narrow range level indicators and wide range level recorders,

16. Steam. flow / feed flow indicators and recorders,

17. -Steam generator pressure indicators,

18. Turbine parameter recorders,

19. Condenser pressure,

20. Status indication of various valves and pumps associated with important systems including safeguards,

21. Four computer trend recorders, one dedicated to subcooling and one to VCT level, the other two are operator selectable.

22. Various other system alarms including safety system channel alerts.

1.2.4 Planned Changes to Existing Data and Information Capability The present plant computers will be replaced by an updated computer system monitoring both units. This sytem should be in operation in 1985.

This computer system will. consist of redundant computers with automatic pickup if the operating computer fails, as well as have battery backed instrument bus power supplies.

The sequence of events monitoring capability will be changed as follows:

1. The list of parameters monitored will be expanded. A list of the additional points is Attachment C.

2. The t,ime discrimination will be improved to 5 milliseconds.

3. The maximum number of events per sequence will be enlarged to 100 and the capacity for up to 10 events will be provided.

4. The sequence of events reports will be outputted to a printer and stored on a disc and/or magtape for later reprinting, if desired.

The post-trip review capability will be changed as follows:

1. The current 2-second group will be sampled at one-second intervals for 48 seconds before and after the trip.

2. The current 8-second group will be sampled at five-second intervals for 4 minutes before and after the trip.

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3. The post-trip review report will be outputted to a printer and stored on disc and/or magtape for later reprinting, if desired.

2.1 Equipment Classification and Vendor Interface An ',nitial review has confirmed that those safety-related components necessary to assure a reactor trip are included within the quality assurance scope at Point Beach Nuclear Plant. To further ensure the proper quality assurance coverage of safety-related systems and components, a detailed intensive review of the drawings used to categorize the quality assurance scope has been initiated. This review, scheduled to be completed by February 29, 1984, will include all aspects of those safety-related systems and components including those necessary to provide a reactor trip function.

A program designed to control vendor technical information has existed at Point Beach since initial construction. This program includes the distribution of vendor technical information to the appropriate personnel so that this information can be utilized and incorporated in plant procedures, as necessary. Also included in this program is a control process for vendor technical manuals. Although this program undergoes continuing review in the normal course of business, a specific review will be conducted to ensure all aspects of this program function efficiently. This specific review will be conducted in concert with other INP0 sponsored efforts in the control and dissemination of vendor information.

Wisconsin Electric is a participating member of the INPO Nuclear Utility Task Action Committee (NUTAC) which has been chartered to study and make recommendations concerning vendor interface in response to Section 2.2.2 of this letter. (See Section 2.2.2 for further detail.) We would expect that the recommendations of the NUTAC will envelop vendors that provide components whose function is to trip the reactor. Also, since these components fall under the NSSS scope of l supply, we believe vendor feedback and communication mechanisms l already exist through the NSSS vendor.

2.2 Equipment Classification and Vendor Interface 2.2.1 Equipment classification and the administration of the program is accomplished through existing quality assurance procedures.

The Quality Assurance and Reliability Manual (Point Beach Nuclear Plant QA Volume II) contains a rigorous listing of those systems and major components which are defined as l QA-scope (includes those systems and components considered to be safety-related). The QA-scope boundaries of the l

systems and components are then further delineated in drawings, lists and tables to clearly establish the scope  ;

and extent of the equipment classification. Specific, more detailed classifications such as environmental equipment j qualification or containment isolation functions are l included in the lists and tables.

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-The information in QA Volume II pertaining to equipment classification is distributed and is available to all Nuclear Power Department (NPD) personnel. This information is used to establish applicability of the Quality Assurance Program to equipment and services pertaining to activities such as procurement, maintenance, modifications and design. Documents, such as purchase requisitions, maintenance and modification requests, maintenance procedures, etc. , are clearly marked to indicate their QA applicability. These documents receive appropriate QA reviews to assure inclusien of all pertinent technical and quality assurance requirements. Items or activities determined to be QA-scope are subjected to rigorous controls as defined in each NPD organization's quality assur-ance and administrative procedures.

Compliance with procedures and the described information handling system is verified through technical audits by the on-site audit group and through overall QA Program audits by the Quality Assurance Division (off-site organization).

2.2.2 As indicated in the response to Section 2.1, Wisconsin Electric is a participating member of the INPO NUTAC regarding Section 2.2.2. The NUTAC is formulating recommendations and guidance to be used by utilities in responding to the specific issues in this section. Wisconsin Electric will evaluate these recom-mendations when they become available and at that time will establish a specific response to this section. The NUTAC has established a goal of February 1,1984 for completion of this task. In concert with this schedule, we would expect to have a more complete response to this section by April 1, 1984.

3.1 Post-Maintenance Testing (Reactor Trip System Components) c 3.1.1 Post-Maintenance Operability Testing Review It has been a practice at P3NP to verify that instrumentation and control devices (including breakers) associated with the reactor trip system are verified to be operable in accordance with Technical Specification requirements prior to placing the equipment in service after maintenance. The verification is accomplished utilizing test requirements contained in periodic test procedures and calibration procedures employed to satisfy

' the operability verification of this equipment required in Technical Specifications.

It has also been determined that no administrative requirement exists which requires this verification to be done. Therefore, maintenance request procedures will be modified prior to February 29, 1984, to include such a requirement for post-maintenance testing of equipment.

3.1.2 Vendor & Engineering Recommendations Although it is believed that the testing currently performed on reactor trip system equipment includes appropriate guidance concerning vendor and engineering recommendations, a review of test procedures, technical manuals and other equipment information will be completed and any findings incorporated into appropriate test procedures by February 29, 1984.

3.1. 3 Post-Maintenance Test Requirements Which Degrade Safety No post-maintenance test requirements have been identified in Technical Specifications that degrade safety.

3.2 Post-Maintenance Testing (All Other Safety-Related Components) 3.2.1 Post-Maintenance Operability Testing Review The safety-related equipment discussed in this section is divided into the following general categories; pumps and valves, piping, instrumentation and controls, and breakers.

In addition to periodic surveillance testing, it has been our practice to perform post-maintenance testing on safety-related pumps and valves. This testing is. performed in accordance with the guidelines of ASME Code Section XI.

This post-maintenance testing is performed by completing applicable portions of periodic tests employed to monitor the inservice performance of these pumps and valves. The same acceptance criteria used in evaluating performance during periodic testing is used in evaluating performance after maintenance.

Following repairs to piping in safety-related systems' post-maintenance testing as delineated by the applicable codes or standards is performed. This testing may include various methods such as dye penetrant, visual, radiographic, hydrostatic, or other nondestructive test methods.

In addition to the instrumentation and control system testing mentioned in a previous section, it has been our practice that safety-related instrumentation and control devices be verified operable in accordance with Technical Specifica-tion prior to placing the equipment in service after mainten-ance. The verification is accomplished utilizing test

, requirements contained in periodic test procedures and cali-l bration procedures employed to satisfy the operability verification of this equipment required by Technical Specifica-tions.

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The' post-maintenance testing described above is adequate

'to demonstrate that safety-related pumps,; valves,. piping systems and' instrumentation;and controls are capable of 1 performing their safety functions following repairs. It

'has also been determined that, although.this testing is being performed, adequate. administrative controls do not

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exist.to require that.this. verification be done. .Therefore, in place administrative controls, such as the Maintenance' .

Request Procedure, will'be modified prior to February 29, 1984, to include such a requirement for post-maintenance testing

, of safety-related equipment.

The adequacy of post-maintenance testing of safety-related i breakers is still under investigation. In addit. ion.to checking proper protective tripping setpoints,. breakers used e to supply power to safety-related pumps or fans are presently

, tested by-verifying that the applicable pump or fan can be~

. started and stopped. . Safety-related bus breakers are also post-maintenance cycled to verify their operation. .The i' investigation to determine if this testing is adequate will.be completed prior to April 30,_1984.

3.2.2 Vendor and Engineering Recommendations Although it is believed that the testing currently performed on the majority of safety-related equipment includes appropriate guidance concerning vendor and engineering recommendations, a review of test procedures, technical manuals and other equip-ment information will be completed and any findings incorporated into appropriate test procedures by November 1, 1985.

j 3.2.3 Post-Maintenance Test Requirements Which Degrade Safety i

Although, to date, no post-maintenance test requirements have been identified in Technical Specifications that degrade safety, some may be identified as a result of the detailed review discussed above. You will be informed if any are identified.

4.1 Reactor Trip System Reliability (Vendor-Related Modifications)

The, Licensee has reviewed and verified that the reactor trip system vendor-related modifications applicable to the Point Beach Nuclear Plant, specifically modifications in accordance with Westinghouse document NCD-ELEC-18 for the DB-50 switchgear, have been implemented.

4.2 Reactor Trip Preventive Maintenance-(Existing) 4.2.1 Reactor trip and bypass breakers are inspected annually during ea::h refueling. The inspection is made per the DIS-50 breaker technical manual. The UV trip attachment is cleaned (if

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necessary) and. lubricated with silicone spray. A discussion of-Licensee's maintenance program for reactor trip breakers was also.provided in our response.to IE Bulletin No. 83-01 i

dated March 3, 1983.

4.2.2 The dropout voltage of the UV trip attachment is checke'd

'at each refueling inspection and is recorded. This infor-mation is compared with prev A measurements in order to detect signs of. degradation whict may warrant more involved checks or replacement.'

.4.~2.3 'We.have no. records of formal life testing; however, Westing-t house letter NS-EPR-2737 to Mr. H. Denton of the NRC, dated March 22, 1983,' makes reference to testing done on UV trip attachments in 1972. The letter states that a UV trip device 4

"was successfully tested more than 8000 operations without malfunctions." It is doubtful .that our reactor trip breaker

.would experience that many operations in the life of the plant. . Life cycle testing of the shunt trip attachment and

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3' the.undervoltage trip attachment of the reactor trip switch-gear is presently being conducted by Westinghouse for the Westinghouse Owners Group. This program is aimed toward establishing the service' life of these devices and sub-stantiating periodic test requirements with proper mainten-ance. The results of this program may be. factored into maintenance, replacement and qualification programs if necessary. The test program is scheduled for completion in the.second quarter of 1984.

- 4.2.4 We have no knowledge of any periodic replacement of breakers or components required or recommended by Westinghouse.

P The conclusion of the program discussed in 4.2.3 may result in periodic replacement criteria.

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4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment for Westinghouse and B&W Plants)

Wisconsin Electric is working closely with the Westinghouse

Owners Group and Westinghouse in developing a design modifica-l tion relating to reactor trip system reliability. The design modification will incorporate'both the requirements for automatic actuation of the shunt trip attachment and the capability t for the on-line surveillance requirement. A detailed generic design package for incorporation of an automatic shunt trip feature was developed under the sponsorship on the Westinghouse Owners Group and submitted to the NRC in Mr. J. J. Sheppard's letter dated

.. June 14, 1983.

This generic design package for an automatic shunt trip modifica-

tion included the design basis, functional requirements, conceptual design and discussion of the conformance to safety criteria. The

. design of the system includes hard-wired component installation provisions for on-line surveillance testing that independently

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verifies by manual means the operability of the UVTA and the automatic shunt trip on the main reactor trip breakers.

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The NRC has issued a safety evaluation report (SER) on this generic design as an attachment to Mr. Darrell G. Eisenhut's letter to Mr. Sheppard dated August 10, 1983. The SER concluded that the conceptual design was generally acceptable and noted several exceptions. The SER included a listing of plant-specific

-information required by the NRC to perform the plant specific reviews.

Wisconsin Electric is presently reviewing this generic design modification and coordinating with Westinghouse in the develop-ment of a plant specific design. We are committed to incorpora-tion of an automatic shunt trip actuation modification and intend to incorporate the provisions for on-line testability. It is anticipated that the details for a Point Beach specific modifica-tion will be finalized over the next six months. At that time we will prepare an ammendment to this submittal to provide a report describing the modification and provide the additional information required by the NRC's SER of the generic design package. We expect to submit this report by May 1984. Assuming NRC approval of this plant-specific modification by August 1, 1984, we would expect to install the modification on each Point Beach unit during the next unit refueling shutdown. These are presently scheduled for fall 1984 for Unit 2 and spring 1985 for Unit 1.

4.4 Not Applicable To Point Beach Nuclear Plant 4.5 Reactor Trip System Reliability 4.5.1 Diverse Trip Features The current design of the reactor trip system does not include a shunt coil trip of the reactor trip breakers when an auto-matic reactor trip signal is generated. Periodic testing of all automatic reactor trip functions can be accomplished on line. Modifications will be made to add a shunt coil trip as well as the capability for on-line independent testing of each trip device. The schedule for accomplishing this modification is outlined in Section 4.3.

4.5.2 On-Line Testing Design The reactor protection system at PBNP is designed for on-line testing of its components except for the manual trip func-tion from the main control board. This portion of the reactor trip system is tested prior to startup each refueling shut-down. The manual reactor trip actuates both the shunt and undervoltage coil trip devices. To accomplish independent l testing of these devices without use of jumpers, lifted leads l

or pulled fuses, modifications to this equipment will be made in conjunction with modifications described in Section 4.5.1 above.

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4.5.3 Reactor Trip System Performance fhe reactor trip system testing interval based on operating

' experience.is considered adequatc. Throughout the approximate ten year history of PBNP, the reactor trip system has required only minor maintenance due to nonconservative component failures and has resulted in very few incidents due to operator error.

WCAP-10271 submitted by the Westinghouse Owners Group to the

.NRC in January 1983 documents a more formal evaluation of the impact on reactor protection system (RPS) unavailability of current and extended surveillance intervals. Supplement 1 to WCAP-10271 which was submitted in September 1983 is an extension of the evaluation and provides a discussion of.

component wearout caused by testing. The conclusion of WCAP-10271 and Supplement 1 is that less frequent testing of RPS components is warranted and will result in an improvement in overall plant . safety and equipment reliability.

We shall follow the NRC's review and disposition of these documents and will consider adjustments to our testing intervals based on the resolution of the issues discussed in these documents.

ATTACHMENT A SEQUENCE OF EVENTS PARAMETERS P-250 COMPUTER Computer Address Description F0403D Loop "A" Low Flow 2/3 Trip F0423D Loop "B" Low Flow 2/3 Trip LO406D Steam Generator "A" Low-Low Level 2/3 Trip LO426D Steam Generator "B" Low-Low Level 2/3 Trip LO483D Pressurizer High Level 2/3 Trip N0005D PR High Power 2/4 Trip N0010D Low Power Range 2/4 Trip N0020D IR 35 High Flux Trip Train "A" N0021D IR 36 High Flux Trip Train "B" N0030D SR 31 High Flux Trip Train "A" N0031D SR 32 High Flux Trip Train "B" P0399D Turbine AST,0il Low 2/3 Trip P0403D "A" Steam Pressure Low SI 2/3 Trip PO423D "B" Steam Pressure Low SI 2/3 Trip PO483D Pressurizer High Pressure 2/3 Trip PO488D Pressurizer Low Pressure 2/4 Trip P1003D Containment High Pressure SI 2/3 Trip T0498D- Overtemperature ATsp1 2/4 Trip T0499D Overpower ATsp2 2/4 Trip

V0324D RCP Bus Undervolt 2/4 Trip Y0004D Manual Reactor Trip ,

Y0005D Manual Reactor Trip Y0006D Reactor Trip Breaker "A" Y0007D Reactor Trip Breaker "B" Y0026D Reactor Bypass Breaker "A" Y0027D Reactor Bypass Breaker "B" Y0335D Unit On-Line output Breaker YO394D Turbine Stop Valve Closed 2/2 Trip YO400D Reactor Coolant Pump "A" Breaker - Open - Trip YO401D Steam Generator "A" Low-Level SF/FF Mismatch Trip YO420D Reactor Coolant Pump "B" Breaker - Open - Trip Y0421D Steam Generator "B" Low-Level SF/FF Mismatch Trip

, .YO480D Pressurizer Low Pressure SI Trip YO920D Manual SI Trip Train "A" YO921D Manual SI Trip Train "B"

POST TRIP REVIEW P-250 COMPUTER Eight Second Group (continued)

Computer Address Description Units F0405C Steam Generator "A" Steam Flow F464 And F0406C Steam Generator "A" Steam Flow F465 KBH F0423C 5 team Generator "B" Feedwater Flow F476 KBH F0424C Steam Generator "B" Feedwater Flow F477 KBH

'F0425C Steam Generator "B" Steam Flow F474 KBH  ;.

F0426C Steam Generator "B" Steam Flow F475 KBH~

PO400A Steam Generator "A" Steam Pressure PT-468 psig P0401A Steam Generator "A" Steam Pressure PT-469 psig P0402A Steam Generator "A" Steam Pressure PT-482 psig P0420A Steam Generator "B" Steam Pressure PT-478 psig P0421A Steam Generator "B" Steam Pressure PT-479 psig P0422A Steam Generator "B" Steam Pressure PT-483 psig l T0400A RCLA 1 Tavg T40lW Deg F T0401A' RCLA 2 Tavg T402W Deg F T0403A RCLA 1 AT T405P Deg F T0404A RCLA 2 AT T406P Deg F T0407A RCLA Overpower ATspi T40lS Deg F T0408A RCLA Overpower ATsp2 T402S Deg F T0410A RCLA Overtemp ATsp1 T405D Deg F T0411A RCLA Overtemp ATsp2 T406D Deg F T0420A RCLB 1 Tavg T403W Deg F T0421A RCLB 2 Tavg T404W Deg F T0423A RCLB l'AT T407P Deg F T0424A RCLB 2 AT T408P Deg F T0427A RCLB Overpower ATsp1 T403S Deg F T0428A RCLB Overpower ATsp2 T404S Deg F

, T0430A RCLB Overtemp ATspl T407D Deg F i

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POST TRIP REVIEW P-250 COMPUTER Eight Second Group (Continued)

Computer Address Description Units T0431A RCLB Overtemp ATsp2 T408D Deq F LO480A Pressurizer Level LT426 PC LO481A Pressurizer Level LT427 PC LO482A Pressurizer Level LT428 PC LO483A Pressurizer Level Cont Setpoint LC428F PC PO480A Pressurizer Pressure PT-429 psig PO481A Pressurizer Pressure PT-430 . psig PO482A Pressurizer Pressure PT-431 psig P0483A Pressurizer Pressure PT-449 psig T0406A RCLA Cold Leg Temp Deg F T0426A RCLB Cold Leg Temp Deg F T0497A RC average AT T405S Deg F T0499A RC Auctioneered Tavg T401F Deg F LO400A Steam Generator "A" Narrow Range Level LT461 PC LO401A Steam Generator "A" Narrow Range Level LT462 PC LO402A Steam Generator "A" Narrow Range Level LT463 PC LO403A Steam Generator "A" Wide Range Level LT460 PC LO420A Steam Generator "B" Narrow Range Level

! LT471 PC LO421A Steam Generator "B" Narrow Range Level LT472 PC LO422A Steam Generator _"B" Narrow Range Level LT473 PC LO423A Steam Generator "B" Wide Range Level LT470 PC P1000A Containment Pressure FT-945 psig P1001A Containment Pressure PT-947 psig P1002A Containment Pressure PT-949 psig T0481A Pressurizer Steam Temp T425 Deg F

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ATTACHMENT B POST TRIP REVIEW P-250 COMPUTER Two'Second Group Computer Address Description Units N4100P Power Range Channel N41 Power' Level PC N4200P- - Power Range Channel N42 Power Level PC-N4300P Power Range Channel N43 Power Level PC N4400P Power Range Channel N44 Power Level PC P0398A Turbine First Stage Pressure Channel 1 PT-485 . psig P0399A' . Turbine First Stage Pressure Channel 2 PT-486 psig Q0340A Unit Generator Gross MW MW T0496A RC Tref T401Y Deg F Eight Second Group Computer

. Address Description Units-N0031C' Source Range Channel 1 Count Rate DKCS

.N0032C Source Range Channel 2 Count Rate DKCS N0035C ' Intermediate Range Channel 1 Power Level MCAMP N0036C Intermediate Range Channel 2 Power Level MCAMP N0041A Power Range 1 Top Detector Flux N41A Volts N0041B Power Range 1 Bot Detector Flux N41B Volts N0042A Power Range 2 Top Detector Flux N42A Volts N0042B Power Range'2 Bot Detector Flux N42B Volts N0043A' Power Range 3 Top Detector Flux N43A Volts N00438 Power Range 3 Bot Detector-Flux N438 Volts N0044A - Power Range 4 Top Detector Flux N44A Volts N0044B Power Range 4 Bot Detector Flux N448 Volts ,

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N4100P Power Range Channel N41 Power Level PC N4200P Power Range Channel N42 Power Level PC

.N4300P . Power Range Channel N43 Power Level. PC t N4400P. Power Range Channel N44 Power Level PC P0398A Turbine First Stage Pressure Channel 1 PT-485 psig P0399A' Turbine First Stage Pressure Channel 2 i PT-486 psig i

QO340A Unit Generator Gross MW MW

! T0496A RC Tref 'T401Y Deg F l F0403C Steam Generator "A" Feedwater Flow-F466 KBH

.F0404C Steam Generator '"A" Feedwater Flow

i. F467 KBH ,

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ATTACHMENT C SEQUENCE OF EVENTS PARAFETERS ADDITIONAL ON NEW COMPUTER Computer Address Description ICDROP RPI Rod Drop Indication INCRDROP NIS Rod Drop 1PC484A Low Vacuum Steam Dump Interlock 1PCV430 Pressurizer PORV 1PCV431C Pressurizer PORV 1PCV434 Pressurizer Safety Valve IPCV434R Pressurizer Safety Valve IPCV435 Pressurizer Safety Valve IPCV435R Pressurizer Safety Valve IRECFANA Containment Recirc Fan W1A1 Breaker 1RECFANB Containment Recirc Fan W1B1 Breaker 1RECFANC Containment Recirc Fan WlCl Breaker 1RECFAND Containment Recirc Fan W1D1 Breaker 1RHRP10A RHR Pump P10A Breaker 1RHRP10B RHR Pump P10B Breaker ITBSVACL Turbine Stop Valve "A" Close l ITBSVBCL Turbine Stop Valve "B" Close i

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