ML17298A625

Response to Required Actions Based on Generic Implications of Salem ATWS Events (Generic Ltr 83-28).

ML17298A625
Person / Time
Site:	Palo Verde
Issue date:	11/30/1983
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:
Shared Package
ML17298A623	List: ML17298A622 ML17298A625
References
GL-83-28, NUDOCS 8311070440
Download: ML17298A625 (88)

Text

PALO VERDE NUCLEAR GENERATING STATION RESPONSE TO REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF THE SALEM ATWS EVENTS (NRC GENERIC LETTER 83-28)

NOVEMBER 1983 8311070440 83f 103 PDR ADOCK 05000528 PDR I'RIZONA PUBLlC SERVICE COMPAN PROJECT MANAGER AND OPERA

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INTRODUCTION This report provides the Palo Verde Nuclear Generating Station response to "Required Actions Based on Generic Implications of the Salem ATWS Events", Generic Letter 83-28 dated July 8, 1983.

Each action is addressed, as it applies to the Palo Verde Nuclear Generating Station design and operation, in the manner specified in the documentation requirements of Generic Letter 83-28. The NRC positions presented herein are from Generic Letter 83-28. The PVNGS response to an action may be segmented to correlate with the format of the NRC position.

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TABLE OF CONTENTS PAGE

1. POST-TRIP REVIEW 1.1 Program Description and Procedure 1.1-1 1.2 Data and Information Capability 1.2-1

2. EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE 2.1 Reactor Trip System Components 2.1-1 2.2 Programs for All Safety-Related Components 2.2"1

3. MAINTENANCE TESTING 3.1 Reactor Trip System Components 3.1-1 3.2 All Other Safety-Related Components 3.2-1

4. REACTOR TRIP SYSTEM RELIABILITY 4.1 Vendor-Related Modifications 4. 1-1 4.2 Preventive Maintenance and Surveillance Program for Reactor:Trip Breakers 4.2-1 4.3 Automatic Actuation of Shunt Trip Attachment on Westinghouse and B&W Plants 4.3-1 4.4 Improvements in Maintenance and Test Procedures for B6W Plants 4.4-1 4.5 System Functional Testing 4.5-1

5. APPENDICES A Sequence of Events Parameter List A-1 B Sequence of Events Display (Typical) B-1 C Chart Recorder List C-1 D ERFDADS Transient Data Point List D-l E ERFDADS Transient Data Plot (Typical) E-l

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REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ASS EVENTS 1.1 POST-TRIP REVIEV (PROGRAM DESCRIPTION AND PROCEDURE)

Position Licensees and applicants shall describe their program for ensuring that unscheduled reactor shutdowns are analyzed and that a determination is made that the plant can be restarted safely. A report describing the program for review and analysis of such unscheduled reactor shutdowns should include, as a minimum:

1. The criteria fo determining the acceptability of restart.

2. The responsibilities and authorities of personnel who will perform the review and analysis of these events.

3. The necessary qualifications and training for the responsible personnel.

4. The sources of plant information necessary to conduct the review and analysis. The sources of information should include the measures and equipment that provide the necessary detail and type of information to reconstruct the event accurately and in sufficient detail for proper understanding.

(See Action 1.2)

5. The methods and criteria for comparing the event information with known or expected plant behavior 'e.g., that safety-related equipment. operates as required by the Technical Specifications or other 'performance specifications related to the safety function).

6. The criteria for determining the need for independent assessment of an event (e.g., a case in which the cause of the event cannot be positively identified, a competent group such as the Plant Operations Review Committee, will be consulted prior to authorizing restart) and guidelines on the preservation of physical evidence (both hardware and software) to support ind pendent analysis of the event.

7. Items 1 through 6 above are considered to be the basis for the establishment of a systematic method to assess unscheduled reactor shutdowns. The systematic safety assessment procedures compiled from the above items, which are to be used in conducting the evaluation, should be in the report.

PVNGS RESPONSE:

1. Criteria for determining restart will be contained and explained in a station manual procedure. Both procedures will be approved by 12/1/83.

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I ~ I 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

RESPONSE CONT.

2. A Station Manual procedure will specify responsibilities and authority for performing the post-trip review and the analysis.

This procedure will be approved by December 1, 1983.

3. Personnel performing the post-trip review and analysis are trained in accordance with guidance provided in ANSI/ANS 3.1-1978 as endorsed by Regulatory Guide 1.8 and INPO Document "Nuclear Power

-Plant Shift Technical Advisor Recommendations for Position Description, Qualifi ations, Education and Training", GPG"Ol, Ref.

1, 4/28/81.

The individual involved in immediate review of the Post-Trip Review Report will be licensed as a Senior Reactor Operator.

Personnel involved in any independent reviews required are members of the PVNGS plant staff qualified in accordance with ANSI/ANS 3.1-1978 as endorsed by Regulatory Guide 1.8.

Personnel involved in approval of restart are members of the PVNGS Plant Staff qualified to the: level specified in ANSI/ANS 3.1-1978 and are members of APS management.

4. Sources of plant information 'used in the Post Trip Review will be described and referred to in a PVNGS Station Manual procedure.

This procedure will be approved by December 1, 1983.

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5. Criteria and methods for comparison of event information with the known or expected plant behavior will be contained in a PVNGS Station Manual procedure. ;This procedure will be approved by 12/1/83.

6. Criteria for determining need for independent review, as well as guidelines for the preservation of physical evidence to support this review, will be contained in PVNGS Station Manual Procedures.

These procedures will be approved by December 1, 1983.

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1.2 POST-TRIP REVIEW DATA AND INFORMATION CAPABILITY Position Licensees and applicants shall have or planned a capability to record, recall and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns prior to restart and for ascertaining the proper functioning of safety-related equipment.

Adequate data and information shall be provided to correctly

'diagnose the cause o unscheduled reactor shutdowns and the proper functioning of safe y-related equipment during these events using systematic safety assessment procedures (Action 1.1). The data and information shall be displayed in a form that permits ease of assimilation and analysis by persons trained in the use of systematic safety assessment procedures.

A report shall be prepared which describes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown. The report shall describe as a minimum:

1. Capability for assessing sequence of events (on-off indications)

1. Brief description of equipment (e.g., plant computer, dedicated computers,'strip chart)

2. Parameters monitored!

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3. Time discrimination between events I

4. Format for displaying data and information

5. Capability for retention of data and information

6. Power source(s) (e.g., Class IE, non-Class IE, non interruptable)

2. Capability for assessing the time history of analog variables needed to determine the cause of unscheduled reactor shutdowns, and the functioning of safety-related equipment.

1, Brief description of equipment (e.g., plant computer, dedicated computer, strip charts)

2. Parameters monitored, sampling rate, and basis for selecting parameters and sampling rate

3. Duration of time history (minutes before trip and minutes after trip) 1.2-1

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1.2 POST TRIP REVIEW POSITION CONT.

4. Format for displaying data including scale (readability) of time histories

5. Capabilty for retention of data, information and physical evidence (both hardware and software)

6. Power source(s) (e.g., Class IE, non-Class IE, non interruptable)

3. Other data and information provided to assess the cause of unscheduled reactor shutdowns.

4. Schedule for any planned changes to existing data and information capability.

PVNGS RESPONSE:

Capability for Assessin Sequence of Events:

1. The Plant Computer is equipped with a Sequence Of Events (SOE)

Log. The SOE Log is initiated when any one of the plant parameters that are monitored changes state.

2. The SOE monitors those parameters closely related to the tripping of the reactor and turbine-generator and the actuation of safety-related equipment. Appendix A lists those parameters monitored by the SOE log.

3. Changes of parameter state are recorded on the SOE Log in the order of occurrence, with a resolution of one millisecond.

'I 4~ Appendix B provides the display format of the SOE Log.

5. The Plant Computer will record these events until a period of 15 seconds has elapsed 'in which no changes of state have been detected. At the end of this time period all events recorded are considered one log and is automatically printed out.

Space is allocated on bulk memory for up to four SGZ Logs awaiting printout. The retention of the SOE Log's hard copy printout is addressed in the procedure referred to in the PVNGS Response to Action 1.1, Post Trip Review.

6. The power source for the Plant Computer is the Non-Class IE 120 VAC Uninterruptible Power System.

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g s 1.2 POST-TRIP REVIEW RESPONSE CONT.

2. Ca abilit for assessin the time histor of analog variables:

1. A combination of Class and Non-Class IE Chart recorders, and the Transient Data File from our Emergency Response Facilities Data Acquisition and Display System (ERFDADS) can be utilized to assess the cause of unscheduled reactor shutdowns, and the functioning of safety-related equipment.

2. The parameters( monitored by the chart recorders listed in Appendix C. Ea(ch of these parameters is continually monitored in order to produce a graph of the value vs. time. The chart speed is 20mm per hour.

The parameters monitored by the ERFDADS Transient Data File are listed in Appendix D. The Transient Data File has the capability of monitoring up to 70 parameters at a rate of 10 times per second (0.1 second resolution).

3. Each chart on the Control Room chart recorder is capable of displaying up to 30 days worth of data. The ERFDADS Transient Data File is capable of storing up to 160 minutes worth of data.

4. Each chart on the Control Room chart recorder is 100mm wide and can hold about 30; days of charts. Appendix E is an example of how a plot generated from the Transient Data File would look. The maximum number of points that can be plotted along the X-axis is 60.

5. The retention of both 'chart recorders and ERFDADS transient data plots are discussed in the procedures referred to in the PVNGS response to Action 1.1, Post-Trip Review.

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1.2 POST-REVIEW RESPONSE CONT.

6. The classification for the power supply for each of the recorders is indicated in Appendix C.

The power supply for each unit's Data Acquisition System (DAS) is 120V Non-Class IE instrument AC. The power supply for the ERFDADS Technical Support Center Computer System (TSCCS) is a Non-Class IE Uninterruptible Power Supply (UPS).

3. Other data and in ormation provided to assess the cause of unscheduled reactor shutdown Other data and information that will be utilized in assessing the cause of unscheduled reactor shutdowns may come from the Unit Log, Control Room Log, Shift Technical Advisor (STA) Log, statements of operators and other personnel (as required), and other operating logs (as required). The use of these sources will be discussed in the procedures discussed in the PVNGS response to Action 1.1, Post-Trip Review.

4, Schedule for any lanned changes to existin data and information PVNGS Design Change Package ..(DCP) RK-012 will include Channels C and D Reactor Trip Circuit Breaker Position on the SOE recorder.

This DCP will be implemented Prior to fuel load, 1.2-4

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

Position Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement. In addition, for these components, licensees and applicants shall

.establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life /of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these components should be contacted and an interface established.

Where vendors can not be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reactor trip system reliability. The vendor interface program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system .of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgement for receipt of technical mailings. The program shall, also define the interface and division of responsibilities among the licensees and the nuclear and non-nuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of and applicable instructions for maintenance work are provided.

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2.1 EQUIPMENT CLASSIFICATION & VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)

RESPONSES PVNGS RESPONSE:

The following type components are identified as components whose function is required to trip the reactor:

Reactor Trip Breakers S/G Pressure Transmitters Pressurizer Pressure Transmitters S/G Level Tfansmitters Excore Detectors RCP Speed Sensor CEA Position Indication (Position Transmitters only)

RTD's for Thot and Tcold (for CPC inPut)

Containment Pressure Tranmitters CPC Cabinet PPS Cabinet Foxboro Cabinet SPS Cabinet Note: Included with the above type components are interconnecting cables, I/E converters, indicators, and recorders.

All of the above components are Class IE and are safety-related.

These components are identified as safety-related (Class IE) on electrical elementary diagrams, in the instrument index, and are classified as safety-related in accordance with a PVNGS Station Manual procedure concerning the identification and classification of safety-related parts.

Maintenance activities and work orders involving . safety-related components are designated as such on associated documents. This information is in the instrument index per quality classification procedure in the Station Manual.

P Replacement parts are procured as safety-related in accordance with Station Manual procedures. The quality classification of an item for procurement purposes is in accordance with the instrument index and the Station Manual procedure on quality classifications.

The PVNGS Maintenance Department has initiated a procedure change to require that the latest available revision of developmental and implementing references have been used in the development of all procedures, and that this is verified prior to procedure approval and implementation. Verification will be accomplished on a procedure review check list. This procedural change will be incorporated by December 1, 1983.

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2.1 RESPONSE CONT.

The PVNGS Maintenance Department has initiated a procedure change to require all revision or addenda to Technical and Instruction Manuals, received by PVNGS, be routed through the Maintenance Engineering Department prior to their being incorporated into the applicable manual. This front-end review will ensure that any change to a maintenance, test or design of safety-related equipment is implemented, or where deviations are necessary, Justification is provided. This information will be kept in the applicable manual, and changes to Test and Maintenance Procedures made as required.

This procedural chan e will be incorporated by December 1, 1983.

The PVNGS Maintenance Department initiated a procedural change to require that the Maintenance Engineering Department review all test and maintenance procedures during their bi-annual review. This review process will ensure that revisions and changes in vendor recommendations are incorporated in Test and Maintenance procedures on a regular basis. It will also ensure that where vendors cannot be identified, have gone out of business, or will not supply the information, adequate attention is paid to safety-related equipment and associated procedures to compensate for lack of vendor backup.

The vendor of the reactor trip system components, Combustion Engineering (CE), was contacted by letter, dated September 9, 1983. CE responded on October ll, 1983, providing confirmation to our request for establishing: an interface program. CE's Nuclear Power Systems division will'ontinue to provide updates to the vendor documentation already 'provided. These updates may take the form of Bulletins to all owners of CE designed NSSS's, or plant specific transmittals.

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

Position Licensees and applicants shall submit, for staff review, a description of their programs for safety-related* equipment classification and vendor interface as described below:

1. For equipment classification, licensees and applicants shall describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functi ns are identified as safety-related on documents, procedures,'nd information handling systems used in the plant to control safety-related activities, including maintenance, work orders and replacement parts. This description shall include:

1. The criteria for identifying components as safety-related within systems currently classified as safety-related.

This shall not be interpreted to require changes in safety classification at the system level.

2. A description of the information handling system used to identify safety-related components (e.g., computerized equipment list) and the methods used for its development and validation.

3. A description of ttfe process by which station personnel use this information handling system to determine that an

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activity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined ;in the introduction to 10 CFR 50, Appendix B, apply to safety-related components.

4. A description of the management controls utilized to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed.

5. A demonstration that appropriate design verification and qualification testing is specified for procurement of safety-related components. The specifications shall include qualification testing for expected safety service conditions and provide support for the licensees'eceipt of testing documentation to support the limits of life recommended by the supplier.

• Safety-related structures, systems, and components are those that are relied upon to remain functional during and following design basis events to ensure: (1) the integrity of the reactor coolant boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10 CFR Part 100.

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I l I I I g I g I 2.2 Position Cont.

6. Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components. Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures, systems, and components, im ortant to safet required by GDC-1 (defined in 10 CFR Part 50, Appendix A, "General Design Criteria, Introduction" ).

2. For vendor interfac , licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information for safety-related components is complete, current and controlled throughout the life of their plants, and appropriately referenced or incorporated in plant instructions and procedures.

Vendors of safety-related equipment should be contacted and an interface established. Where vendors cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GDC-1).: The program shall be closely coupled with action 2.2.1 above (equipment qualification). The program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive ':feedback with vendors for mailings containing technical informafion. This could be accomplished by licensee acknowledgement for receipt of technical mailings. It shall also define the interface and division of responsibilities among the licensee and the 'nuclear and non-nuclear divisions of their vendors that provide aervice on safety-related equipment to assure that requisite control of and applicable instruction for maintenance work on safety-related equipment are provided.

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PVNGS RESPONSE:

l. EQUIPMENT CLASSIFICATION

1. Components are classified as safety-related in accordance with a PVNGS Station Manual procedure. The procedure complies with Reg.

'Guide 1.26, Rev. 1 and Reg. Guide 1.29, Rev. 1 to determine safety-related components. The specific identification of components as being safety-related is via small flags on P&ID's and via a quality classification. code in the plant Instrument Index, Equipment Index, and Valve List.

The classification rocedure is currently undergoing revision to provide additional guidance in identifying safety-related components and the basis for such classifications. Specifically, the procedure will detail the basis for classifying a component as Class IE.

The procedure will use as a basis for classification the following documents:

Reg. Guide 1.26, Rev. 1 Reg. Guide 1.29, Rev. 3 Reg. Guide 1.30, Revi 0 Reg. Guide 1.97, Rev. 2 FSAR Table 3~2 1 Additionally, the classification/reclassification of components will require an engineering !evaluation per ANSI N45.2.11-1974 to document design criteria review. The bases, i.e. safety-related functions, Quality Group A, E, C, Class IE, Fire Protection, and/or Seismicity, shall be reviewe6 to include consequential failure of interconnecting or parent components.

will be The revised procedure approved prior to December 1, 1983.

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APS uses two information handling systems to identify safety-related components. One was developed by Bechtel at the beginning of the PVNGS project and the other by APS to support the PVNGS operation activities. Bechtel's system produces the Instrument Index, Equipment Index, and Valve List. This list and the indexes have been maintained by Bechtel in accordance with their internal procedures. APS has used these indexes and list in the form of controlled design drawings. Arrangements have been made for the turnover of the control of the Bechtel system's data base to APS Records Management. This turnover is scheduled to be completed by January 1984. The indexes provide a three digit alphanumeric quality classification. This classification identifies whether an item i's safety-related, whether it is seismically qualified and whether an item is ASME Section III, Class IE, or designed to a non-safety classed industry standard.

The second information handling system is an APS interactive data retrieval system. Its data base is similar to the Bechtel data base but contains more information to facilitate it's use by plant operators, maintenance personnel, and plant engineers. The equipment identification and quality classification stored in the data base are extracted from the Bechtel system data bases. The APS data base is currently being developed by a contractor in accordance with the contractors QA program and under APS supervision. The data base formulation is approximately 60%

complete, reviews performed by the contractor and PVNGS Engineering department will verify accuracy. The data base can be accessed from numerous computer terminals on-site and off-site. All data base information is retrievable in hard-copy form.

PVNGS personnel will perform a random sample check to verify accuracy of the data base information. This audit will validate the initial data base. Subsequent to completion of the data base, no changes to the quality,: classification code field will be permitted without authorization from the PVNGS Engineering department. Authorization will only be granted to changes in the quality classification data field subsequent to a quality classification review performed in accordance with a Station Manual procedure. Changes will be approved by the PVNGS Engineering department, the QA department and plant management.

Periodically, a listing wi'1 be requested for items which have undergone a change in the quality classification data field. This list generated by the computer will be compared to the authorized changes. The number of individuals authorized to edit the data base will be minimized and each individual authorized to edit the data base will be approved by management. The comparison cited above will validate all changes to the data base.

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3. The process by which personnel use the PVNGS data to determine if a component is safety-related is covered in the response to item 2.2.1.2.

The identification of which components are safety-related is covered by a Station Manual procedure. This procedure is applicable to all procedures, including those involving procurement, design and work control. This ensures that parts procurement of safety-related components is identified and appropriately controlled under the operational phase PVNGS QA program.

This component class fication procedure is also referenced by work control procedures and departmental procedures that identify which maintenance activities and procedures are safety-related. Those tasks and procedures subsequently identified as safety-related will therefore require that plant management and Nuclear Safety Reviews are performed.

PVNGS design control procedures also reference this procedure to ensure that an appropriate quality classification is applied to site controlled and off-site controlled designs. Both on-site and off-site design control procedures are in accordance with ANSI N45.2.11-74. All APS design control procedures are in conformance with ANSI N45.2.11-74.

Design Drawing development: is controlled by APS's Nuclear Engineering Department. All quality related drawings shall be marked with a "Q" in the lower right hand corner. Additionally, all P&ID's shall show safety-related boundaxies via small flags.

All safety-related Class IE,drawings shall be marked as such per IEEE-494-1974.

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4. In accordance with the requirements of ANSI N18.7-76, Paragraph 4.5, PVNGS Operations QA/QC has established an audit program which includes the performance of audits to verify compliance with and effectiveness of the design and material control programs. Annual.

audits of the design material control process are performed.

Procedure requirements covering the audit process include sections of a Corporate Procedures Manual, a Department Instruction, and an audit schedule.

In accordance with the requirements of the Operations Quality Assurance Criteria Manual, Section 4.3.3, PVNGS Operations QA/QC

-reviews all quality related purchase requisitions and purchase orders initiated at PVNGS. Procedural requirements covering this review include a station manual procedure and a Department Instruction.

These audits and reviews provide assurance that:

1) changes to the design of safety related items are justified and subject to design control measures commensurate with those applied to the original design.

2) technical and quality assurance requirements specified in procurement document for spare/replacement items are the same as or at least equivalent to the requirements specified for the original item.

3) in those cases wher'e the technical requirements for the original item must. be modified, such modification is handled in accordance with the design control process.

4) purchased items and lservices comply with the requirements of the procurement documents.

5) materials, parts, and components are identified and controlled throughout all phases of production, assembly, shipping, receipt, i storage, installation, inspection, test, maintenance, modification, and use.

6) only items which satisfy specific requirements set forth in codes, standards, procurement documents, and other specifications, are accepted for installation or use.

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5. Procurement of safety-related equipment is in accordance with Nuclear Engineering Department procedures. These procedures require as a minimum that procurement technical specifications meet the requirements of the PVNGS FSAR, PVNGS Design Criteria Manual, and meet or exceed the original design requirements for electrical equipment procurement in accordance with IEEE-323-74.

Design verifications within APS are required per ANSI N45.2.11-74 as noted in the response to Section 2.2.1.3. Design verification performed by equipment suppliers (who are on APS'pproved Vendor List) are performed in accordance with the vendor's QA program.

Also required is th]t the vendor be apprised of the documentation turnover package req'uired to satisfy APS QA requirements. Inasmuch as the specifications must at least meet the original plant requirements, the packages developed will comply with IEEE 323-74 and NUREG-0737.

In accordance with the requirements of ANSI N18.7-76, paragraph 4.5, PVNGS Operations QA/QC has established an audit program which includes the performance of audits to verify compliance with the procurement program. Annual audits of the procurement process are performed. Procedure requirements covering the audit process include a portion of a corporate quality assurance manual, an internal instruction and a PVNGS Operations QA/QC Department audit schedule.

In accordance with the requirement of current Operations Quality Assurance criteria manuals, 'VNGS Operations QA/QC reviews all quality-related purchase requisitions and purchase orders initiated at the site. Procedural requirements covering this review include a station manual procedure and an Operations QA/QC internal instruction.

The audits and reviews discussed above will provide assurance that quality-related items and services are procured properly.

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6. The program and criteria for identifying and classifying safety-related equipment is detailed within a PVNGS Station Manual procedure as noted in the response to Section 2.2.1.1, 2.2.1.2, 2.2.1.3. Other items which are important to safety are identified in an appendix to a procedure found in the PVNGS Station Manual.

This appendix covers those additional items covered under the operations phase of the PVNGS QA program and as identified in the Section 3.2 of the PVNGS FSAR, (e.g. Fire Protection, Radwaste, Containment HVAC, Reactor Vessel head lift rig, radiation protection equipment, etc.)

2. "VENDOR INTERFACE APS is a member of Nuclear Utility Task Action Committee (NUTAC) formed for the specific purpose of defining either an appropriate vendor interface program or an acceptable alternative to item 2.2.2.

This NUTAC is meeting under the auspices of INPO's Analysis and Engineering Industry Review Group with the support of INPO.

On September 8, 1983, we submitted a request for extension to respond to item 2.2.2, After completion of the NUTAC's efforts, projected to be March 1, 1983, we will submit our program description to the NRC 60 days after, the completion of the NUTAC's efforts.

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

Position The following actions are applicable to post-maintenance testing:

1. Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post~aintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety function& before being returned to service.

2. Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

3. Licensees and applicants shall identify, if applicable, any post~aintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety. 'ppropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval. (Note that action 4.5 discusses on-line system functional testing.)

PVNGS RESPONSE:

A review of all Reactor 'rip System test and maintenance Technical procedures, which are presently approved, and Specifications indicates that post~aintenance testing to assure operability is included. All procedures still in draft will have retest requirements included, as required. All test and maintenance procedures for Reactor Trip System Components are scheduled to be written and approved by January 1, 1984.

2. A review of all approved Test and Maintenance procedures and Technical Specifications indicates that the latest vendor and engineering recommendations available to PVNGS have been incorporated. Any deviations from vendor and/or engineering recommendations have been reviewed by the vendor and found to be satisfactory.

3. A comprehensive review of post-maintenance test requirements in existing Technical Specifications has revealed no requirements that would degrade rather than enhance safety.

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

Position The following actions are applicable to post-maintenance testing:

Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to assure that post~aintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equi ment is capable of performing its safety functions befor being returned to service.

2. Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications where required.

3. Licensees and applicants shall identify, if applicable, any post~aintenance test requirements in existing Technical Specifications which age perceived to degrade rather than enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.

PVNGS RESPONSE:

l. A review of all safety-related equipment test and maintenance procedures, which are presently approved,

-. and Technical Specifications indicate's that post~aintenance testing to assure operability is in'eluded. All procedures still in draft will have retest requirements included, as required.

2. A review of all approved test and maintenance procedures and Technical Specifications for safety-related equipment indicates that the latest vendor and engineering recommendations available to PVNGS have been incorporated.

All procedures still in draft will meet the same requirements.

3. A comprehensive review of post-maintenance test requirements in existing Technical Specifications has revealed no requirements that would degrade rather than enhance safety.

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4.1 REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED MODIFICATIONS)

Position All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either: (1) each modification has, in fact, been implemented; or (2) a written evaluation of the technical reasons for not implementing a modification exists.

For example, the modifications recommended by Westinghouse in NCD-Elec-18 for the DB-50 breakers and a March 31, 1983, letter for

-the DS-416 breakers shall be implemented or a justification for not implementing shall b] made available. Modifications not previously made shall be inc6rporated or a written evaluation shall be provided.

PVNGS RESPONSE:

A review of the latest vendor recommendations and station work documents indicates that all vendor modifications to reactor trip breakers available to PVNGS as of the date of this response, have been implemented.

To ensure that all future changes to reactor trip breakers, either in maintenance, testing or 'design which are recommended by the vendor are incorporated, .or when not incorporated, proper Justification is provided, the station procedure governing control of Technical/Instruction Manuals at PVNGS is being revised to include a front-end revie4, by the Maintenance Engineering Department, of all changes/revisions prior to their incorporation.

Scheduled completion date of .this procedure revision is December 1, 1983.

In addition, the Nuclear Engineering Department reviews all vendor recommended modifications.

J APS Records Management files these recommendations with controlled vendor data, and distributes to the PVNGS Maintenance Department, to assure they have been reviewed as discussed above.

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

Position Licensees and applicants shall describe their preventative maintenance and surveillance program to ensure reliable reactor trip breaker operation. The program shall include the following:

A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment suppl er.

2. Trending of parameters affecting operation and measured during testing to forecast degradation of operability.

3. Life testing of the breakers (including the trip attachments) on an acceptable sample size.

4. Periodic replacement of breakers or components consistent with demonstrated life cycles.

PVNGS RESPONSE:

l. The preventive maintenance program for reactor trip breakers is being developed. This program should be approved for implementation by December 1; 1983. The surveillance program for reactor trip breakers is in. development, with procedure approval anticipated by December 1, 1983, and implementation scheduled prior to initial criticality.

2. The Maintenance Engineering Department has the responsibility for review of Reactor Trip Breaker Maintenance and Surveillance results for the purpose of forecasting operability degradation by trending measured parameters. This is required by station procedures.

3. The equipment qualification life testing program of both the General Electric Model AKR-30 and Westinghouse Type DS-206 curcuit breakers are addressed in the Reactor Trip Switchgear System Equipment Environmental Qualification Program No. 14273-DCE-3516, 12/16/80. This Program, thus far, includes the System Operations Cycling Qualification Report, System Environmental Qualification Report, Seismic Qualification Report, Aging Analysis Report and an Age Conditioning Procedure 4.2-1

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The environmental, cyclic, and seismic testing completed in early 1980 demonstrated the environmental compatability and seismic integrity of the reactor trip switchgear. The seismic portion of this program was accepted by APS and assigned status 1. However, aging was not specifically addressed in this phase.

In accordance with the Environmental Qualification Program Document, if the initial Qualification Program does not include age preconditioning prior to the equipment DBE, an aging analysis shall be performed in two phases. ,Phase 1 identifies all age degradable components which have seismic or environmental related failure

'mechanisms, which could affect equipment operability. Phase 2 includes the assi nment of qualified lives and recommended replacement intervals for these components, based on prior type testing and sound engineering judgement. Concurrent with these activities, an age conditioning procedure is generated which is to be used if prior test experience is not available and/or should the results of the age analysis be inconclusive. In this case, the full test program, including cycling and seismic, would be repeated. Phase 1 age analysis activities and the age conditioning procedure have been completed, the forecasted completion date of Phase 2 is January, 1984.

The results of this program are being factored into our Qualification Maintenance Pr'ocedure (QMP), specifically defining maintenance and part replacement intervals along with surveillance activities identified in the qualification program. This will ensure equipment operability integrity.

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4. The demonstrated life of life limiting components for the Reactor Trip Switchgear System (RTSS}, is addressed in the Aging Analysis Report, discussed in section 4.2.3, and the Supplemental Aging Analysis Report (SAAR). The SAAR defines specific part replacement intervals of all life limiting components in the RTSS. These life limiting components are replaced according to the QMP. The QMP divides the organizational responsibilities and interfaces to ensure that, at any point in time, equipment installed in PVNGS, which is required to be qualified, is qualified. The equipment requiring scheduled replacement is identified in the Equipment Qualification Listing (EQL), established and maintained by PVNGS Nuclear Engineering. PVNGS Nuclear Operations is then responsible for identification and tracking of equipment, using the EQL, approaching the end of it's qualified life, and replacement of such equipment, or its life limiting component prior to, or at its end of qualified life. All equipment which is life limited is addressed in this manner.

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4.3 REACTOR TRIP SYSTEM RELIABILITY (AUTOMATIC ACTUATION OF SHUNT TRIP ATTACHMENT FOR WESTINGHOUSE AND B&W PLANTS Postion Westinghouse and B&W reactors shall be modified by providing automatic reactor trip system actuation of the breaker shunt trip attachments. The shunt trip attachment shall be considered safety related (Class IE).

PVNGS RESPONSE:

Not applicable to P GS Units; which are CE plants.

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4.4 REACTOR TRIP SYSTEM RELIABILITY (IMPROVEMENTS IN MAINTENANCE AND TEST PROCEDURES FOR B&W PLANTS)

Position Licensees and applicants with B&W reactors shall apply safety-related maintenance and test procedures to the diverse reactor trip feature provided by interrupting power to control rods through the silicon controlled rectifiers.

-This action shall n t be interpreted to required hardware changes or additional envi onmental or seismic qualification of these components.

PVNGS RESPONSE:

Not applicable to PVNGS Units, which are CE plants.

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4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

Position On-line functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be performed on all plants.

1. The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&W (see

.Action 4.3 above) and CE plants; the circuitry used for power interruption width the silicon controlled rectifiers on B&W plants (see Act'ion 4.4 above); and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

2. Plants not currently designed to permit periodic on-line testing shall justify not making modifications to permit such testing. Alternatives to on-line testing proposed by licensees will be considered where special circumstances exist and where the objective of high reliability can be met in another way.

3. Existing intervals .for 'on-line functinal testing required by Technical Specifications. shall be reviewed to determine that the intervals are consistent with achieving high reactor trip system availability when: accounting for considerations such as:

l. uncertainties in component failure rates

. 2. uncertainty in common mode failure rates

3. reduced redundancy during testing

4. operator errors during testing 5, component "wear"out" caused by the testing Licensees currently not performing periodic on-line testing shall determine appropriate test intervals as described above. Changes to existing required intervals for on-line testing shall be justified by information on the sensitivity of reactor trip system availability to parameters such as the test intervals, component failure rates, and common mode failure rates.

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PVNGS RESPONSE:

1. Testing of the Undervoltage and shunt trip features shall be covered by station procedures in accordance with the requirements of the PVNGS Technical Specifications. The design of the system permit and the applicable procedures require that such testing shall be accomplished while the plant is on-line. These procedures should be approved by January 1, 1984.

2. The PVNGS design allows periodic on-line functional testing of the

-reactor trip breaker's diverse trip features.

3. A review of on-lin functional testing intervals required by the PVNGS Technical Specifications will be performed by APS. A program to conduct this review is being developed by the Nuclear Engineering Department's Nuclear Safety and Licensing group as part of the PVNGS reliability program. At this time, we do not propose any changes to the test intervals required by the Technical Specifications.

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APPENDIX A

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Parameters Monitored b the Se uence of Events Recorder Exciter Protective Trip Exciter Voltage'egulator Trip Main Transformer 525 KV Line Differential Trip Main Transformer Sudden Pressure Trip Auxiliary Transformer Phase/Ground Differential Trip Auxiliary Transformer Sudden Pressure Trip Auxiliary Transformer Phase(Neutralizer Overcurrent Trip Generator Unit DifferentialI Trip Generator Differential-1 Trip Generator Differential-2 Trip Generator Stator Ground Trip Generator Stator Ground Backup Trip Generator Loss of Field Trip Generator Negative Sequence Trip Generator Underfrequency Trip Generator Reverse Power Trip Generator Backup Volts/Hz Trip in Turbine Sequential Trip Generator Backup Distance Trip 525YV Breaker Fail to Trip Generator Coastdown Trip Generator Post-Coastdown Trip Generator Backup Protective Trip Circulating Water Pump A Trip Circulating Water Pump B Trip Circulating Water Pump C Trip Circulating Water Pump D Trip Channel A Trip Circuit Breaker Position Channel B Trip Circuit Breaker Position Channel C Trip Circuit Breaker Position*

Channel D Trip Circuit Breaker Position*

• PVNGS Design Change Package RK-012 will include these points on SOE recorder.

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Parameters Monitored b the Se uence i

of Events Recorder CEDM Power Bus Undervoltage 1 CEDM Power Bus Undervoltage 2 CEDM Power Bus Undervoltage 3 CEDM Power Bus Undervoltage 4 Recirculation Actuation Signal A Manual Actuation Recirculation Actuation Signal B Manual Actuation Recirculation Actuation Signal C Manual Actuation Recirculation Actuation Si al D Manual Actuation Feedwater Pump Turbine A Low Vaccuum Feedwater Pump Turbine A Active Thrust Bearing Wear Detector Feedwater Pump Turbine A Inactive Thrust Bearing Wear Detector Feedwater Pump Turbine B Low Vaccuum Feedwater Pumo Turbine B Active Thrust Bearing Wear Detector Feedwater Pump Turbine B Inactive Thrust Bearing Wear Detector Feedwater Pump Turbine A Hydraulic Control Pressure Trip Feedwater Pump A Bearing Oil Low Pressure Feedwater Pump Turbine B Hydraulic Control Pressure Trip Feedwater Pump B Bearing Oil Low Pressure Electro Hydraulic Control Low Fluid !Pressure Trip Main Steaming Rate High Level Trip Loss of Both Speed Signals or Bacl:uP Operation Speed Trip Mechanical Overspeed Trip Generator Reverse Power Trip Permissive Turbine Supervisory Instrumentation High Vibration Trip Exhaust Hood High Temperature Trip Condenser Low Vaccuum Trip Master Turbine Trip Fast Closing of Isolation Valves Commanded Upper Thrust Bearing Wear Detector Trip Lower Thrust Bearing Wear Detector Trip Emergency Trip System Low Pressure Trip No Electrical Trip Valve Pressure Trip Loss of Stator Coolant Bearing Oil Low Pressure Trip Shaft Pump Low Discharge Pressure Trip A-2

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) Parameters Monitored b the Se uence of Events Recorder No Electrical Trip Solenoid Valve Pressure Trip Loss of Electro Hydraulic Control 24VDC Power Loss of Electro Hydraulic Control 125 VDC Power Turbine Supervisory Instrumentation Trip Voltage Turbine Supervisory Instrumentation Trip Low Voltage Supply Generator/Reactor Initiated Trip Manual Turbine Trip Handle Status Feedwater Pump Turbine A Governor Control Power Supply Feedwater Pump Turbine B G vernor Control Power Supply Heater Drain Pump A Trip Heater Drain Pump B Trip Low Steam Generator=1 Level Channel A Low Steam Generator 1 Level Channel B Low Steam Generator 1 Level Channel C Low Steam Generator 1 Level Channel D High-High Containment Pressure Charm'el A High-High Containment Pressure Channel B High-High Containment Pressure Channel C High-High Containment Pressure Chan@el D Low Refuel'ing Hater Tank Level Channel A Low Refueling Hater Tank Level Chanqel B Low Refueling Hater Tank Level Channel C Low Refueling Mater Tank Level Channel D Steam Generator 1 Greater Than Steam Generator 2 Pressure Channel A Steam Generator 1 Greater Than Steam Generator 2 Pressure Channel B Steam Generator 1 Greater Than Steam Generator 2 Pressure Channel C Steam Generator 1 Greater Than Steam Generator 2 Pres ure Channel D Steam Generator 2 Greater Than Steam Generator 1 Pressure Channel A Steam Generator 2 Greater Than Steam Generator 1 Pressure Channel B Steam Generator 2 Greater Than Steam Generator 1 Pressure Channel C Steam Generator 2 Greater Than Steam Generator 1 Pressure Channel D Low Steam Generator 2 Level Channel A Low Steam Generator 2 Level Channel B Low Steam Generator 2 Level Channel C Low Steam Generator 2 Level Channel D A-3

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Parameters Monitored b the Se uence of Events Recorder High Log Power Channel A High Log Power Channel B High Log Power Channel C High Log Power Channel D Variable Over Power Channel A Variable Over Power Channel B Variable Over Power Channel C Variable Over Power Channel D High Local Power Density Cilannel A 1

High Local Power Density Channel B High Local Power Density Channel C High Local Power Density Channel D Low DNBR Channel A Low DNBR Channel B Low DNBR Channel C Low DNBR CHannel D Main Steam Isolation Signal A Manual; Actuation Main Steam Isolation Signal A Manual Actuation'- Remote Shutdown Panel Main Steam Isolation Signal B Manual Actuation Main Steam Isolation Signal B Manual Actuation Remote Shutdown Panel Main"Steam Isolation Signal C Manua3 Actuation 1fain Steam Isolation Signal C Manua1 Actuation Remote Shutdown Panel Main Steam Isolation Signal D Manual Actuation Main Steam Isolation Signal D Manual Actuation Remote Shutdown Panel Containment Spray Actuation Signal A Manual Actuation Containment Spray Actuation Signal B Manual Actuation Containment Spray Actuation Signal C Manual Actuation Containment Spray Actuation Signal D:fanual Actuation Auxiliary Feedwater Actuation Signal-1 A Manual Actuation Auxiliary Feedwater Actuation Signal-1 B Manual Actuation Auxiliary Feedwater Actuation Signal-1 C ".fanual Actuation Auxiliary Feedwater Actuation Signal-1 D Manual Actuation Auxiliary Feedwater Actuation Signal-2 A Manual Actuation Auxiliary Feedwater Actuation Signal-2 B Manual Actuation Auxiliary Feedwater Actuation Signal-2 C i~fanual Actuation Auxiliary Feedwater Actuation Signal-2 D Manual Actuation A-4

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i Ii I Parameters >fonitored b the Se uence of Events Recorder Low Steam Generator 2 Pressure Channel A Low Steam Generator 2 Pressure Channel B Low Steam Generator 2 Pressure Channel C Low Steam Generator 2 Pressure Channel D High Steam Generator 1 Level Channel A High Steam Generator 1 Level Channel B High Steam Generator 1 Level Channel C High Steam Generator 1 Leve( Channel D Low Steam Generator 1 Pressure Channel A Low Steam Generator 1 Pressure Channel B Low Steam Generator 1 Pressure Channel C Low Steam Generator 1 Pressure Channel D Low Steam Generator 2 Level Channel A Low Steam Generator 2 Level Channel B Low Steam Generator 2 Level Channel C Low Steam Generator 2 Level Channel D fligh Steam Generator 2 Level Channel A High Steam Generator 2 bevel Channel' High Steam Generator 2 Level Channel; C High Steam Generator 2, Level Channel D High Pressurizer Pressure Channel A; High Pressurizer Pressure Channel B l High Pressurizer Pressure Channel C Pressurizer Pressure Channel D 'igh Low Pressurizer Pressure Channel A Low Pressurizer Pressure Channel B Low Pressurizer Pressure Channel C Low Pressurizer Pressure Channel D High Containment Pressure Channel A High Containment Pressure Channel B High Containment Pressure Channel C High Containment Pressure Channel D Low Steam Generator 1 Level Channel A Low Steam Generator 1 Level Channel B Low Steam Generator 1 Level Channel C Low Steam Generator 1 Level Channel D A-5

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Parameters Monitored b the Sequence of Events Recorder Low Steam Generator 1 Reactor Coolant Flow Channel A Low Steam Generator 1 Reactor Coolant Flow Channel B Low Steam Generator 1 Reactor Coolant Flow Channel C Low Steam Generator 1 Reactor Coolant Flow Channel D Low Steam Generator 2 Reactor Coolant Flow Channel A Low Steam Generator 2 Reactor Coolant Flow Channel B Low Steam Generator 2 Reactor Coolant Flow Channel C Low Steam Generator 2 Reac or Coolant Flow Channel D Condensate Pump A Trip Condensate Pump B Trip Condensate Pump C Trip Remote Manual Reactor Protection System Channel A Remote Manual Reactor Protection System Channel B Remote Manual Reactor Protection System Channel C Remote Manual Reactor Protection System Channel D Safety Injection Actuation Signal A.Manual Actuation Safety Injection Actuation Signal B .Manual Actuation Safety Injection Actuation Signal C'Manual Actuation Safety Injection Actuation Signal D!Manual Actuation Containment Isolation Actuation Signal A Manual Actuation Containment Isolation Actuation Signal B Manual Actuation Containment Isolation Actuation Signal C Manual Actuation Containment Isolation Actuation Signal D Manual Actuation A-6

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APPENDIX 3 I

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sEonEHz n~vEHIs~nn (ShfiPLE)

SEquErtCE OF EVENTS RECORD, TtlE FIRST EVENT OCCURRED AT ll:04:31.000 POINT ID DESCRIPTlotl'"

" "EVENT TINE AFTER FIRST EVENT Y1062 Ctl 1 STtl GEN 2E24A IIIGll LVL TRIP 0.000 SECONDS Y1074 CII 1 SG 2E24tt P > SG 2E24A P TRIP 0.604 SECOHDS Y0350 tlFWPR DRVR 2A ACT TIIR DRG TltlP 1.414 SECONDS

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APPENDXX C

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'I A ~ ' B CHART RECORDERS NAME BOARD DESIGNATION*

Shutdown Cooling Heat Exchanger fjl inlet/outlet B02 temperature Loop 1 Hot Leg temp. /Loop 1A Cold Leg temp. B02 Loop 2 Hot Leg temp. /Loop 2A Cold Leg temp. B02 Reactor Vessel Level/Satu ation Margin B02 Pressurizer level B02 Saturation Margin Core Exit Thermocouple/ B02 Core Exit temperature Pressurizer Pressure B02 Cl SG /ll Pressure/SG //2 Pressure B02 SG /ll level/SG //2 level B02 Plenum level/Reactor head B02 Containment Hydrogen/Containment water I I

level B02 RWT level/CST level B02 Containment Pressure - B02 Shutdown Cooling Heat Exchanger f32 j.nlet/ B02 outlet temperature Gas Stripper Activity B03 N Reactor Makeup Mater Tank flow/setpoint B03 N Boric Acid flow to VCT B03 N Boron Concentration B03 N Coolant Activity B03 N Pressurizer level/setpoint B04 Pressurizer Pressure X/Y B04 Loop 1B Cold Leg temp/Loop 2B Cold Leg Temp. (WR) B04 'N Loop 1 Hot Leg temp/Loop 2 Hot Leg temp. (NR) 804

• C Class IE Power Source N Non-Class IE Power Source C-1

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N&fE BOARD DESIGNATION Log Power Channel A/Log Power Channel B B04 N Log Power Channel C/Log Power Channel D B04 N T g/T B04 Startup Channel 1/Startup Channel 2 B04 N Calibrated/Excore linear Power Channel A B05 Calibrated/Excore linear Po er Channel B B05 Calibrated/Excore linear Power Channel C B05 Calibrated/Excore linear Power Channel D B05 Hotwell level 1C/2C B05 SG 81 Downcomer flow B06 N SG 81 Steam flow/feed flow B06 N SG /tl level 1/level 2 B06 SG f/2 Downcomer flow B06 N SG f32 Steam flow/feed flow B06 N SG //2 level 1/level 2 B06 N Condenser A/Condenser B vacuum B06 N Condenser C vacuum N B06'07 Blowdown Flash Tank Discharge flow/BFT Pressure N C-2

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APPENDIX D "

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ERFDADS TRANSIENT DATA POINTS Auxiliary Feedwater Flow to Steam Generator 1 Auxiliary Feedwater Flow to Steam Generator 2 RC Cold Leg 1A Temperature RC Cold Leg 1B Temperature Charging Flow to Regenerative Heat Exchanger RC Hot Leg to Steam Generator 1 Temperature RC Cold, Leg 2A Temp rature RC Cold Leg 2B Temperature Containment Pressure A Containment Pressure B RC Hot Leg to Steam Generator 2 Temperature Pressurizer Pressure Condenser A Shell Pressure Condenser B Shell Pressure Condenser C Shell Pressure First Stage Steam Pressure First Stage Pressure to REacgor I

Regulating System Core Loop 1A Differential Pressure Core Loop 1B Differential Pressure Core Loop 2A Differential Pressure Core Loop 2B Differential Pressure Letdown Flow Pressurizer Level (Wide Range)

Steam Generator 1 Total Feedwater Steam Generator 2 Total Feedwater 4.16kv Bus S03 Voltage 4.16kv Bus S04 Voltage Steam Generator 1 Primary Loop Delta Pressure Steam Generator 2 Primary Loop Delta Pressure Linear/Calibrated Linear Power A Linear/Calibrated Linear Power B

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APPENDIX E go C

PTID 1127 UNIT 1 120 lpp 80 4 ~

60 gp 20 0

0.0 1.0 2.0 3.0 4.0 5.0 TIME (SEC) 12/20/1982 10: 0: 0 DATE/TIME 12/20/1982 10: 0: 5 TRANSIENT PARAMETER DISPLAY ERFDADS TRANSIENT. DATA PLOT'(TYPICAL).',

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