ML20078R769

Rev 1 to Westinghouse Owners Group Emergency Response Guidelines Validation Program

ML20078R769
Person / Time
Site:	Seabrook
Issue date:	10/31/1983
From:	WESTINGHOUSE OPERATING PLANTS OWNERS GROUP
To:
Shared Package
ML20078R750	List: ML20078R748 ML20078R769
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-TM PROC-831031, NUDOCS 8311150285
Download: ML20078R769 (201)

Text

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Westinghouse Owners Group EMERGENCY RESPONSE GUIDELINES VALIDATION PROGRAM Westinghouse Owners Group l

(= SEABROOK STATION l

WESTINGHOUSE Nuclear Technology Division

\x -

/

x /

PROGRAM PLAN pnwn=g

WESTINGHOUSE OWNERS GROUP EMERGENCY RESPONSE GUIDELINES REVISION 1 VALIDATION PROGRAM PLAN l October 1983 l

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VALIDATION PROGRAM PLAN TABLE OF CONTENTS Topic Page 1.0 Introduction 5 2.0 Objectives 8 l

3.0 Validation Process 10 3.1 Preparation Phase 10 3.1.1 Scope of Validation 10 l 3.1.2 Validation Method 12 3.1.3 Validation Criteria 12 3.1.3.1 E0Ps - Operator 14 3.1.3.2 E0Ps - Control Room 14 3.1.3.3 E0Ps - Training 16 3.1.3.4 Training - Operators 16 3.1.3.5 Training - Control Room 16 3.1.3.6 Operator - Control Room 20 3.1.3.7 Procedure - specific criteria 20 3.1.4 Procedure Writing 21 3.1.4.1 Validation Test E0Ps 21 3.1.4.2 E0P Verification 24 3.1.5 Test Scenarios 24 3.1.6 Simulator Capability 27 3.1.7 Data Collection 32 3.1.7.1 Computer data 32 l 3.1.7.2 Videotape recordings 32 3.1.7.3 Transition flow charts 37 3.1.7.4 Debriefing sessions 37 3.1.8 Observation Teams 42 3.1.9 Test Crews 43 3.1.10 Training 43 1

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TAE'LE OF CONTENTS (Continued) l Topic Page 3.2 Assessment Phase 47 3.3 Resolution Phase 50 3.4 Documentation Phase 50 4.0 Application to Plant-Specific Validation 50 5.0 References 51 Appendix A Definitions A-1 Appendix B NUREG 0737, Item 10.1 B-1 Appendix C Acceptance Criteria For Individual Guidelines C-1 Appendix 0 Example Run Sheets D-1 Appendix E References ERGS and Seabrook Test E0Ps E-1 Appendix F Seabrook Plant Description F-1 Appendix G Example Lesson Plan G-1 2

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List of Tables No Title Page 1 Validation Program Schedule 11 2 Evaluation Criteria for E0P - Operator Validation 15 3 Evaluation Criteria for E0P - Control Room Validation 17 4 Evaluation Criteria for E0P - Training Validation 18 5 Evaluation Criteria for for Training - Operator Validation 19 6 Seabrook Validation Test E0P List 22 7 Major Seabrook/ ERG Reference Plant Differences 25 8 Validation Test Scenarios 28 9 ERG Rev. 1 Validation Test Parameters 33 10 Validation Training Schedule 45 11 Behavior Symptoms to Aid an Observer 48 3

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List of Figures No Title Page 1 Emergency Response Capability SYSTEM 13 Interfaces 2 Expected E0P Usage In Test Scenarios 30 3 Magnetic Tape Data Block 36 4 Control Room Video Coverage 38 5 Composite Video Image 39 6 Transition Flow Chart For E-Series E0Ps 40 7 Transition Flow Chart For F-Series E0Ps 41 4

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1.0Intrcduction On March 28, 1979, a commercial nuclear powered electrical generating plant.

. located at Three Mile Island, near Harrisburg, Pennsylvania, experienced the most severe accident in the history of commercial nuclear power. Although physical equipment damage in the plant was limited almost exclusively to the

-reactor core, and radiation release to the environs was well within current licencing requirements, the event _was viewed as indicative of a serious inability of in management of serious plant accidents.

A series of regulatory guidance and requirements followed, ranging in scope from improved instrumentation and diagnostic. tools, to control room staffing and offsite emergency support facilities. Utilities operating nuclear power plants joined together to form " owner's groups," each group having in common

'the generally uniform system layout and design philosophy of a common NSSS

. supplier. This similarity in plant design allowed groups of utilities to respond to the Three Mile Island regulatory requirements in a generic manner, resulting in reduced cost to member utilities, and simplifying the regulatory / review process by providing generic response.

One of the more significant items of regulation was NUREG-0737, "Classifi-cation of the TMI Action Plan Requirements (Nov 1980). Within that document, item I.C.1, calls for improved emergency operating procedures, and lists

.certain specific requirements of those procedures. (Appendix B) Generic

-letter 82-33 (supplement 1 to NUREG-0737) subsequently reaffirmed schedular requirements and encouraged prompt utility implementation.

The Westinghouse Owners Group (WOG) responded to this particular requirement by generating a comprehensive package of generic operating instructions for coping with plant emergencies. The package, referred to as the Emergency Response Guidelines (ERGS), provides operating instructions for two distinct types of situations: For those events which can be diagnosed by an un-ambiguous set of symptoms, specific guidelines have been developed to allow optimal plant recovery (Optimal Recovery Guideline - ORGs). For other events 5

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and/or malfunctions which are not amenable to diagnosis, a separate set of

. guidelines is provided whose' function is to maintain satisfied a limited set Jof Critical Safety Functions (Function Restoration Guidelines - FRGS). The set of' Critical Safety Functions was defined to be those necessary to protect the three primary boundaries preventing fission product release: The fuel matrix / cladding, primary system boundary, and containment building. The ORGs inherently contain instruction steps to maintain all of the Critical Safety Functions satisfied. However the FRGs only contain actions to restore and maintain the Critical Safety Functions satisfied, and contain limited guidance, if any, on recovering the plant.

In response to any protection system or safeguards system actuation, the Optimal Recovery Guidelines are entered first to obtain a diagnosis and perform the subsequent optimal recovery. Whether the event is diagnosed or not, the status of the Critical Safety Function (CSF) set is monitored using a corresponding set of Status Trees. Each tree consists of a series of binary decision points (branches) leading to a unique status condition for the CSF based on existing plant symptoms. Each unique status condition is color-coded to define the required action level, and all conditions other than " satisfied" provide a transition to the appropriate FRG. Details on status tree usage, action level priorities, and FRG implementation are contained in a separate users guide.

The ERGS are solidly based on detailed systems response and plant transient analyses, coupled with sound engineering judgement. They were subjected to detailed review by a WOG subcommittee consisting of operationally experienced members from several different utilities prior to issuance. However the initial set of ERGS, called " BASIC," was developed over an extended period of time (approximately 2.5 years). Over that period, additional analysis, experimental and operational data, and personnel changes, necessarily resulted in some differences between the later guidelines and earlier ones.

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In response to ava'ilable guidance at the time, the BASIC set of ERGS was subjected to a thorough validation test program in June of 1982. Details of that test program and the major conclusions are summarized in Reference 1. As

.a' consequence of that initial validation test program, and also to incorporate

-accumulated utility / vendor / industry input, a' revision to the ERGS was authorized by~the WOG. Revision 1 was from the outset extremely well documented.and laboriously reviewed to assure that' every potential change for BASIC.was duly addressed. As a result, the Revision 1 set is considered to be a much_ improved and better structured product than BASIC set.

~As'a' final demonstration of the effectiveness of Revision 1 to the ERGS,_the guidelines will be subjected to a comprehensive validation test program as described in the. remainder of this document. Whenever possible, the program will . follow appropriate guidelines available to the nuclear industry, and use

.previously published terminology and definitions. '(Appendix A) 7 0011V:1/101183

2.0 Objectives The originally stated program objectives were very brief and are paraphrased here:

1. Validate Revision 1 of the ERGS

2. Compare Revision I to BASIC in light of the validation results

3. Document the validation program in such a manner that it might be referenced by any plant writing E0Ps based on the ERGS These objectives have been expanded to better utilize all of the information being generated by the program, and also to allow correlation with the separate elements of System Validation as described in Reference 3.

It must be realized that the ERGS are generic in nature and contain no plant specific data other than general system characteristics of a reference plant.

In order to be tested, they must first be converted into Emergency Operating Procedures (EOPs). Since this conversion can be made with minimal perturbation to the structure, wording, logic, and usage of the ERGS, the validation results obtained can be claimed to be just as applicable to the ERGS as to the E0P set actually tested. Once written, the E0P set will need to be verified. Verification is the process which assures that the E0Ps are correctly written (as required by the E0P Writers Guide) and contain the correct plant-specific technical data (as obtained from plant design, license, or test data). Validation is the process which demonstrates that actions specified in the E0Ps can be followed by trained operators to manage emergency conditions in the plant.

This last definition mentions several additional aspects which are integral with the concept of E0P usage. First, there must be operators to perform the required actions. Then there must be training, so that the actionr ara performed as intended. And finally, there is the plant which is being controlled. For the operators, the plant exists (primarily) as it is monitored and operated from the control room.

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l The inseparability of these items during actual E0P usage in response to a plant emergency has resulted in the definition of SYSTEM validation: the overall evaluation of the man / machine system to determine that it works together to accomplish the desired results. The validation test program described below will exercise the operator / training /EOP/ control room SYSTEM for a broad range of emergency situations.

From the test data, conclusions about the validation of each SYSTEM element will be reached. For the purposes of this program, emphasis will be placed on the E0P (ERG) and training aspects of the SYSTEM.

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3.0 Validation Pr_ocess

~ 13 section will describe the various aspects of the actual validation test at  ; relates to the operator / training /EOP/ control room SYSTEM. The format is derived from on the guidance presented in Reference 3. Certain modifications have been made, however, to maintain the emphasis on ERGS as generic guidance, and to reduce the plant specific aspects. Table 1 presents the overall program schedule.

3.1 Preparation Phase This phase identifies the resources required for the program, selects the method to be used, develops test scenarios, and determines the extent and application of validation criteria.

3.1.1 Scope of Validation This test program will evaluate the effectiveness of the operator / training /

E0P/ control room SYSTEM in responding to a selected set of major plant transients. The E0Ps to be used will be based on, and closely resemble, the reference ERG Revision 1 set, and the training involved will be specially developed for the same Revision 1.

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lakle i Validation Program Schedule WOG Authorization 5-25-83 Preparation Phase 5-26-83 to 10-23-83 Operator Training 10-24-83 to 10-28-83 Validation Phase 10-31-83 to 11-4-83 Resolution Phase 11-5-83 to 12-31-83 Documentation Phase 12-15-83 to 1-15-84 l

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3.1.2 Validation Method The actual validations will be done on a plant-specific, full-scale control room simulator. A normal operating crew complement will use simulator-specific E0Ps to guide their actions in response to control room (plant / simulator) indications during major plant casualties. Detailed observations will be made of operator performance, procedure usage, and plant response, by a specially trained observation team.

3.1.3 Validation Criteria Since the stated objective of this SYSTEM validation is to test how well the separate elements work together to mitigate major plant transients, it follows that criteria are required for each element's interaction with each other.

This implies the following six areas of interaction:

1. E0Ps - Operator l

2. E0Ps - Control room

3. E0Ps - Training

4. Training - Operator

5. Training - Control room

6. Operator - Control room These interfaces are well-shown on the pyramidal structure shown in Figure 1 (taken from Reference 3). Furthermore, the effectiveness of the integral SYSTEM, with perhaps compounding, perhaps offsetting deficiencies, must have some criteria for acceptability. This final, global, objective is the most straightforward: The plant must be placed in a " safe," " stable" condition, regardless of imposed structural and equipment failures. " Safe" in this context means the reactor is adequately shut down ( subcritical) and cooled.

" Stable" means either steady-state, or changing ir, response to operator control (i.e., controlled cooldown). Satisfaction of this objective can be determined completely by monitoring process parameter trends (and other internal simulator computer variables).

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J l TRAINING !

( PLANT PROCEDURES Figure 1 Emergency Response Capability SYSTEM Interfaces 13 0011V:1/101183

Of the six areas of interaction within the SYSTEM, it should be noted that five concern E0Ps and training, which are of primary concern in this Program.

the >ixth involves interaction between the operator and the control room, which is of only secondary concern in this Program.

3.1.3.1 E0Ps-Operator The first interaction, E0Ps - ooerator, is concerned with the ability of the operator to use the E0Ps. This implies that the E0Ps properly convey what is intended at each step in a manner understood by the operator. This aspect of the E0Ps was referred to as " human factors" in the BASIC ERG validation program, and validity was indicated if the operator was able to " accomplish the required tasks in an orderly and ef ficient manner." Detailed criteria to identify any deficiencies in the E0Ps were presented in Reference 2, under the headings " level of detail" and "understandibility" and Reference 3. A composite list specifically for ERG - based E0Ps is presented in Table 2.

3.1.3.2 E0Ps - Control Room This interaction is concerned with the correct incorporation of the plant, as seen and controlled by the operators in the control room, into the E0Ps. In the BASIC validation, this aspect was part of " human factors" and required that " instrumentation and control devices which are referrenced actually exist in the control room and are correctly indentified." Specific criteria to identify deficiencies were pres.ented in Reference 2 under the headings

" plant compatibility" and " operator compatibility", and Reference 3. Although this Program is trying to de-emphasize the " plant" aspect, it is appropriate to demonstrate that plant-specific modifications can be properly made in ERG-based E0Ps without interfering with the basic structure and usage. A composite list of acceptance criteria for the ERG-based E0Ps is presented in Table 3.

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Table 2 Evaluation Criteria For E0P-Operator Validation o Did each step contain sufficient information?

o Were alternative actions explicit (use of "0R")?

o Were contingency actions sufficient (RNO)?

o Were procedures easily identified?

o Could procedure transitions be made correctly

- out of a procedure?

- into another procedure at the correct step?

- within a procedure?

o Was the organization of the E0P set understood?

o Was the organization within a procedure understood?

o Were CAUTIONS and NOTES recognized and understood?

o Were internal procedure loops performed correctly?

o Was the typeface easy to read?

o Was the two-column format easy to use?

o Were the LOGICAL statements understood (RNO)?

o Could the foldout page be accessed and used properly?

o Could Figures and Tables be read accurately?

o Were CSF Status Trees properly monitored and used to control procedure (FRG) implementation?

o Did CRT implementation of Status Trees conform to expected rules of usage?

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l l 3.1.3.3 E0Ps - Training This interaction is cencerned with the presentation and emphasis placed on the E0Ps in the operator training program. As part of this Validation Program, a special ERG training package was developed to respond to the critaria listed in Table 4. Details of the training program are described in a later section.

3.1.3.4 Training - Operator This interaction is concerned with the actual training given to the operators.

It assumes that the training procram materials already contain complete and accurate information on both the E0Ps and control room operation. The criteria for validating this interface are listed in Table 5, and were used to develop the initial ERG / operator training program lesson plans. For the purposes of this Program, it war asumed that the operators were already familiar with normal plant operations in the control room. Consideration of administrative, maintenance, and testing procedures is specifically excluded for the purposes of this Program.

3.1.3.5 Training - Control Room This interaction involves the complete coverage of plant operations from the control room in the operator training program. Since this aspect of the .,

operator training was assumed to be already complete, no special attempt to validate it will be made as part of this Program. However, deficiencies may be noted in SYSTEM performance due to some weakness in this area, and will be so

~

noted in the Program findings.

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Table 3 Evaluation Criteria For E0P-Control Room Validation o Were plant conditions (symptoms) used in the E0Ps readily available to the operators?

o Could the CSF Status Trees be properly monitored?

o Were plant conditions consistent with E0P assumptions?

o Were instruments and controls referenced by the E0Ps available in the control room?

o Were instruments able to be read to the accuracy required in the E0Ps?

o Could the required actions be performed by the control room crew (staffing)?

o Could proper step sequence be maintained (control room layout)7 o Did individual operators understand areas of cognizance in performing actions?

o Was communication to local operators available as required?

o Were the E0Ps physically usable in the Control Room?

o Could the E0Ps be readily distinguished from other plant procedures?

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Table 4 Evaluation Criteria for E0P-Training Validation o Are the following topics included in the Training program:

E0P structure?

E0P basis?

E0P usage as a set?

Individual procedure usage?

Use of NOTES and CAUTIONS?

Entry conditions?

Transitions?

Barrier Concept?

Critical Safety Functions?

Status Trees?

Status Tree usage?

Priorities of colors?

o Are individual guidelines discussed in detail?

o Is sufficient text material available for the operators to research any specific step?

o Is a control room simulator available to exercise the E0Ps?

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Table _5 Evaluation Criteria For Training-Operator Validation o Were lesson plans available for all topics listed under E0P-Training and Training-Control Room Validation criteria?

o Were knowledgable instructors available for all topics?

o Were all lessons presented in a timely fashion?

o Was a feedback mechanism (exam) employed to check operator understanding?

o Was a control room simulator employed to give real-time experience in E0P usage?

o Were all E0Ps presented to the operators in some form to assure familiarity?

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3.1.3.6 Operator - Control Room l

This interaction is commonly referred to as the man-machine interface. Again, this Program does not intend to make Human Engineering judgements about the adequacy of the control room used for the test. Seabrook Station has per-l formed a control room design review. The results of this review have been recorded and submitted to the NRC. All of the identified deficiencias have been evaluated for corrective changes. however, these changes have not yet been implemented on the main control board. If any deficiency in system performance is attributed to control room design, it will be so noted in the Program findings. These findings will first be screened with the present HEDs, and if not already identified, they will be added to the existing list and evalauated by the utility.

3.1.3.7 Procedure-Specific-Criteria Because the primary objective of this Program is validation of the ERGS, special criteria have been developed for each individual procedure expected to be exercised on the simulator. These criteria encompass the two areas "EOP-Operator." and "EOP-Control Room," and can be evaluated for an indi-vidual guideline by using recorded observations and process parameter data.

The entire set of procedure-specific acceptance criteria is presented in Appendix C.

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3.1.4 . Procedure Writing The effort to convert the ERG Revision I set into plant-specific E0Ps was performed by the Operations staff at Seabrook Station. Evaluation of plant-specific instrument settings and uncertainties were generated by the plant and utility engineering staff, while interpretation of required plant-specific additions were made in conjunction with the training staff.

3.1.4.1 Validation Test E0Ps l

A listing of the complete set of E0P titles is presented in Table 7. Every effort was made to keep the final E0Ps as unchanged from the reference ERGS as possible to facilitate transfer of validation results. Examples of two reference ERGS and the Seabrook Test E0Ps are presented in Appendix E. The procedure writing effort was facilitated by having a solid understanding of the BASIC version of the ERGS. Still, it is estimated that 2 man years of effort went into the present set of E0Ps.

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W Table 6 SEABROONVALIDATIONTESTE0PLIST E-0 Reactor Trip or Safety Injection ES-0.0 Rediagnosis ._

ES-0.1 Reactor Trip Response ES-0.2 Natural Circulation Cooldown .,

ES-0.3 Natural Circulation Cooldown With Steam Void in Vessel ', ,

(with RVLIS)

E-1 Less of Reactor or Secondary Coolant ES-1.1 SI Termination ES-1.2 Post-LOCA Cooldown and Depressurization ES-1.3 Transfer to Cold Leg Recirculation ES-1.4 Transfer to Hot Leg Recirculation E-2 Faulted Steam Generator Isolation -

E-3 Steam Generator Tube Rupture ES-3.1 Post-SGTR Cooldown Using Backfill ES-3.2 Post-SGTR Cooldown Using Blowdown ES-3.3 Post-SGTR Cooldown Using Steam Dump ECA-0.0 Loss of All AC Power ECA-0.1 Loss of All AC Power Recovery Without S.I. Required ECA-0.2 Loss of All AC Power Recovery With S.I. Required ECA-1.1 Loss of Emergency Coolant Recirculation ECA-1.2 LOCA Outside Containment ECA-2.1 Uncontrolled Depressurization of All Steam Generators ECA-3.1 SGTR With Loss Of Reactor Coolant-Subcooled Recovery Desired ECA-3.2 SGTR With Loss Of Reactor Coolant-Saturated Recovery Desired ECA-3.3 SGTR Without Pressurizer Pressure Control

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' ~- _ . _ _ _ _ . _ . _ _ _ _ _ . _ . _

s.

Table 6 (Cont)

.SEABROOK VALIDATION TEST E0P LIST F-0 The Critical Safety Function Status Trees FR-S.1 Response to Nuclear' Power Generation /ATWS FR-S.2 Response to Loss of Core Shutdown FR-C.1 Response to Inadequate Core Cooling FR-C.2 Response to Degraded Core Cooling FR-C.3 Response to Saturated Core Cooling Condition FR-H.1 Response to Loss of Secondary Heat Sink

-FR-H.2 Respense to Steam Generator Overpressure FR-H.3 Response to Steam Generator High Level FR-H.4 Response to Loss of Normal Steam Release Capabilities FR-H.5 Response to Steam Generator Low Level FR-P.1 Response to Imminent Pressurized Thermal Shock Conditions FR-P.2 Response to Anticipated Pre'ssurized Thermal Shock Conditions FR-Z.1 _ Response to High Containment Pressure FR-Z.2 Response to Containment Flooding FR-Z.3 Response to High Containment Radiation Level FR-I.1 Response to High Pressurizer Level

.FR-I.2 Response to Low Pressurizer Level FR-I.3 Response to Voids in Reactor Vessel l

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3.1.4.2 E0P Verification E0P Verification is defined as the comparative evaluation between the final E0Ps and their source documents. Detailed guidance for performing this evaluation is presented in References 3 and 4.

.The E0P Source documents are as follows:

o Seabrook Station Writers Guide for Emergency Operating Procedures o Emergency Response Guidelii.e Writers Guide, Revision 1

o. FSAR, Unit 1, Seabrook Station o Westinghouse Owners Group Emergency Response Guidelines - Revision 1 o Seabrook Plant Description Since the E0Ps were written to minimize differences from the ERGS, and the ERGS were extensively reviewed for consistency with the ERG Writers Guide, the format comparisons were limited to random checks.

Major emphasis during the verification process was on the addition of plant specific information, and changes required because of differences in plant systems from the " reference" plant. A summary of the major systems for Seabrook was compiled into a '"Seabrook Plant Description" which is included as Appendix F for reference. The detailed check against the ERGS was performed jointly by Seabrook and Westinghouse personnel, and any deviations were immediately'noted and corrected. A brief summary of major differences between the Seabrook plant and the " Reference" plant assumed in the ERGS is presented in Table 7.

13.1.5 Test Scenarios The general guidance provided in selecting event scenarios in Reference 3 was

. applied for this Program. The main objective of scenario selection was to exercise a major fraction of the E0P set. To accommodate and exercise the many branching paths within the set, variations of individual events were included. Events range from the most simple reactor trip, to multiple pressure boundary. failures with coincident multiple equipment / control failures. In all cases, a sequence of initiating events was constructed to arrive at the final plant configuration.

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.. 1

Table 7 MAJOR SEABROOK/ ERG REFERENCE PLANT DIFFERENCES Reference HP Plant Seabrook

1. AFW system has two motor-driven 1. EFW system has one motor-driven pumps and one turbine driven pump and one turbine driven pump. Each motor-driven pump Each pump is full capacity and feeds 2 SGs while the turbine feeds all 4 SGs. Startup feed l driven pump feeds all 4 SGs pump can feed through EFW lines.

2. PRZR PORVs are air-operated 2. PRZR PORVs are electrically operated

3. SG PORVs are air-operated 3. SG FORVs are electrically operated

4. Air-operated valves inside 4. Air-operated valves inside containment require air containment are powered by supply from outside containment containment air compressor

5. RCP thermal barriers cooled by CCW 5. Separate thermal barrier cooling system has heat exchangers cooled by PCCW

6. Single ultimate heat sink 6. Ultimate heat sink is either Atlantic Ocean or mechanical draft cooling towers 7 Spray additive tank feeds spray 7. Spray additive tank gravity pump discharge feeds to RWST

8. Safety grade containment fan coolers 8. Containment fan coolers t0T j

safety equipment 25 0011V:1/131183

1 Table 7(Cont}

r MAJOR SEABROOK/ ERG REFERENCE PLANT DIFFERENCES Reference HP Plant Seabrook

9. Containment recirculation sumps 9. Recirculaton sumps are are maintained empty. maintained full. Containment (water) level instrument is qualified l 10. SG narrow range level 10. SG narrow range and wide instruments are qualified range level instruments are (Wide range is not) qualified

11. RCS wide range pressure 11. Wide range RCS pressure transmitters are located , transmitters are located inside containment outside containment r

12. RVLIS has 3 ranges Full range, 12. RVLIS has 2 ranges. (does Dynamic head range, and Upper not have upper range) range.

13. Automatic switchover of AFW pump 13. Manual makeup to CST from suction or low CST level. Demineralized Water Storage Tank.

1 l'

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Details of the initial conditions, plant equipment status, failures to be initiated, and equipment to be made " inoperable" were summarized on individual

" Test Scenario Run Sheets" for each individual test scenario. Additional information about timing of sequential failures and incorporation of unrelated

" distractions" were also included on the Run Sheets.

i The initial list of test scenarios was developed without consideration of actual simulator capabilities. Later versions were changed to accommodate simulator upgrades. The final list of Test Scenarios is' presented as Table

8. Example Run Sheets are presented in Appendix D.

Since a stated objective of scenario selection was to exercise as many of the E0Ps as possible, a matrix showing expected precedure usage is presented as Figure 2. The FRGs designated as "Y" in the figure summary indicate a YELLOW action priority, which means that action is at the operators discretion.

Although the symptoms corresponding to entry conditions are expected to be exhibited frequently during the test scenarios, it cannot be stated, as in the case of the other E0Ps, that the operator is expected to use them.

3.1.6 Simulator Capability The Seabrook Control Room Simulator is an exact duplicate of the Seabrook Station Unit 1 Control Room. The entire simulator was supplied by the Link-Simulator division of Singer Corporation, and is located in the Training Center at the site in Seabrook, New Ham 3 shire. The system configuration includes 2 GOULD/SEL model 12/55 computers and a 300 Megabyte disc. Also coupled to the plant simulatisn is a MODCOMP CLASSIC computer with 2,300 Megabyte discs, which duplicates the functions of the plant process computer.

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(

Table 8 VALIDATION TEST SCENARIOS 1 Loss of offsite power - Reactor trip 1A Natural Circulation cooldown 2 Spurious-SI 3.Small LOCA (1000 gpm) 3A Post-LOCA cooldown l 4 Intermediate - size LOCA (5000 gpm) l~ 4A Post-LOCA cooldown 5 DBA LOCA - No RHR pumps 6 ICC 7 Small LOCA plus subsequent SGTR 8 Small LOCA - No EFW 9 Small LOCA - No HHSI - Return to critical 10 ' Secondary break outside containment

11' Secondary break - All MSIVs f ail to close 12 Secondary break plus ' subsequent secondary break

, 13 Secondary break - MSIV failure (all) plus LOCA 14 Secondary break - MSIV failure (all) plus SGTR 15 Secondary break inside containment plus LOCA 16 Secondary break plus SGTR in faulted SG 17 Secondary break in 3 SGs plus SGTR 18 Secondary break in 2 SGs p! .s-SGTR (1 intact)

1. ~

19 SGTR 19A Post SGTR cooldown using backfill 20 SGTRs in different SGs (subsequent) 28

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Table 8 (Cont)

VALIDATION TEST SCENARIOS 21 SGTR plus secondary break in non-ruptured SG 22 SGTR, loss of HHSI - Return '.o Critical 23 SG tube leak plus spurious SI 24 SGTR plus loss of EFW 25 SGTR plus secondary overpressure, all safety valves fail to close 26 Loss of all ac power ,

26A Loss of all ac power recovery plus SGTR 27 ATWS from full power 28 Loss of all feedwater power available 29 Loss of all feedwater - off site nower lost i

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TEST SCENARIO NUMBER (TABLE 8)

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[$-1.4 E-2 4K M M n au e nr n A N N x X XXXX A E5-3.2 g E5-3.3 LaJ ECA-0.0 4 8

ECA-0.1 X

k t-ECA-0.2

• ECA-1.1 x e ECA-l.2 o '

ECA-2.1 M rs K b [CA-3.1 K d 4 2 y ECA-3.2 w ECA-3.3 o

M 4 X a FR-$.I o FR-5.2 Y a N w

FR-C.1 X FR-C.2 A FR-C.3 Y FR-H.1 X X KK FR il.2 Y FR-H.3 Y X X X FR-H.S Y F R -It. 5 Y x xW # w -

K x A x M FR-P.)

FR-P.2 Y a F R-Z .1 4 WX FR i.2 rR-2.3 r FR-1.1 7 v

rR-i.2 FR-1.' T Figure 2 Expected E0P Usage in Test Scenarios 30 0011V:1/101183

For training purposes, a broad range of malfunctions is available for use by the instructors. For the major event initiators, for example, the instructor has the capability to input:

o. a single LOCA, variable to some maximum size, or 2 other smaller, fixed -

location LOCAs 1 o a single steamline break inside cont-inment o a single steamline break outside containment (upstream of the MSIV), and o a single tube rupture These failures are consistent with the licensing basis " single-failure events" historically required for training prior to the TMI event. However, to exercise the new E0Ps, and create the conditions required by the test scenarios, several changes were made in the software by Seabrook simulator support personnel.

o The maximum size of the variable-size LOCA was increased.

o Multiple steam generator failures (breaks inside and outside containment) can be input o MSIVs on any or all steamlines can fail to close.

o SG tube ruptures can be input in multiple generators, and the maximum leak size has been increased.

4 The simulation software includes calculations of an "RCS void-fraction,"

effectively a system inventory parameter, which is displayed as equivalent reactor vessel level on a CRT display.

Another computer routine handles evaluation of the Critical Safety Function Status Trees, using the real-time plant process parameters. A separate display routine allows monitoring of the Status Trees directly on one of the main control board CRTs.

The MODCOMP (plant process) computer serves both information base and alarm handling. functions. -The information section of_the computer has a graphics capabi'.ity that allows dynamic representation of all the major systems of the olanc as well as trending of individual points or groups of points on the 31 0011V:1/101183 s

- . __- . _ . - - - , . . . - . _ _ _ ,____m,_ . - , _ . - , _ _ - , . . -

control room CRTs. Both types of graphics are controlled by the operators 'via keyboard commands. The alarm function presents the operator with prioritized CRT displays of the off-normal events to facilitate interpretation of severity. l 3.1.7 Data Collection Several different types of data will be collected during the validation test to support the multiple conclusions which will result.

3.1.7.1 Computer data To preserve a record of plant behavior during each test scenario, a special software routine was written by Seabrook personnel to save a comprehensive list of process parameters and equipment status at 5 second intervals during each test run. The parameters are obtained from the simulator data " pool" and copied to available storage space on a 300 Megabyte disc. Following each test scenario, the same data is copied (off-line) to a magnetic tape. Data for each transient scenario will be kept on a separate tape. A listing of the parameters being recorded is presented in Table 9. An example of the tape format is shown in Figure 3.

3.1.7.2 Videotape recordings Two permanently installed video cameras are already available in the Seabrook control room, and will be used during the test. Both camera images are

-recorded simultaneously (split screen) on tape to effectively cover most of the control board at all times. A separate digital time is superimposed on the frame to provide the time reference. The simulator instructor controls the video recorder, digital time clock, and computer data recording. In this way, the computer clocks and video tape clock can be synchronized for each run. Figure 4 shows the approximate control room coverage by the two cameras. Figure 5 is a sketch of the video image composite.

32 0011V:1/101183

Table 9 ERG-REVISION 1 VALIDATION TEST PARAMETERS l

l System Parameter Description Number of Values _ Type

• NIS S.R. Count Rate 1 A S.R. Startup Rate 1 A I.R. Amps. 1 A I.R. Startup Rate 1 A )

Nuclear Power 4 A )

Pressurizer N.R. Pressure A 1

Level 1 A Relief Line Tempcrature 4 A Relief Line Flow 1 A Liquid Temperature 1 A Steam Temperature 1 A Boron Concentration 1 A Spray Flow 1 A Auxiliary Spray Flow 1 A PORV Position 2 D PORV Block Valve Position 2 D Heater Status 4 D Pressurizer Pressure 1 A Relief Tank Level 1 A Temperature 1 A RCS W.R. Pressure 1 A Auctioneered Tavg 1 A Hot Leg Temperature 4 A Cold Leg Temperature 4 A Loop Flow .

4 A

. Core Aug. Water Density 1 A

.. Boron Concentration 1 A Subcooling (from TCs) 1 A Core Exit TCs- 20 A Upper Range Vessel Level 1 A "WR" Vessel Level 1 A Core Decay Heat 1 A CVCS Makeup Flow to VCT 1 A Charging Flow 1 A.

Letdown Flow. 1 A Seal . Injection Flow 4 A Seal Leakoff Flow 4 A VCT Level. 1 A Emergency Borate Flow 1 A A = analog D = digital 33 0011V:1/101183

Table 9 (Cont)

ERG-REVISION i VALIDATION TEST PARAMETERS ,

F System Parameter Description Nur..ber of Values Type

• Sacor.dary SG Pressure 4 A SG Steam Flow. 4 A SG Main Feedwater Flow 4 A SG Emergency Feedwater Flow 4 A SG N.R. Level 4 A ,

7 SG W.R. Level 4 A SG PORV Position .

4 D Main Feedwater Isolation Valve Position 4 0 MSIV Position 4 0 MSIV Bypass Valve Position 4 0

-Steam Dump to Condenser-Demand 1 A .

Steam Header Pressure 1 A Steam Header Flow 1 A Feed Header Pressure 1 A '

Feed Header Flow 1 A Main Generator, MWe 1 A Condenser "A" Vacuum 1 A CST Level 1 A Main Feed Pump Status (Startup) 2 (1) D l Emergency Feed Pump Status 2 D ,

Condensate. Pump Status 3 D Radiation RCS Activity 1 A '

Monitoring SG Activity-Sample 4 -A Containment Activity 1 A SG Blowdown Activity 1 A Condenser Air Discharge 1 A Auxiliary Building Radiation 1 A  :

Stack: 1 A ECCS RWST Level 1 A BIT Flow to RCS 1 A SI Pump Flow 2 A RHR Pump Flow 2 A Containment Spray Pump Flow 2 A Accumulator Pressure 4 A Charging Pump Status 2 0 SI Pump Status 2 D RHR-Pump Status 2 0 PD Pump Status 1 0 Containment Recire. Sump Isolation Valve Status 2 D 34 0011/:1/101183

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

Table 9 (Cont)

ERG-REVISION 1 VALIDATION TEST PARAMETERS System Parameter Description Number of Values Type *

'PCCW RCP Thermal Barrier Flow 4 A RCP Flow 4 A RHR Heat Exchanger Flow 2 A Flow to SI Pumps 2 A Flow to CCP 2' A Flow to RHR Pumps. 2 A Containment' Pressure 1 A

. Temperature- 1 A Humidity 1 A Recirculation Sump Levels 2 A Normal. Sump Levels 2 A Hydrogen 1 A Fan Cooler Status 6 D Control Air Isolation Valve Status 1 D Control Air Pressure Inside Cont. 1 A Service Water Fan Cooler Water Flow 4 A l

FC Inlet Temperature 1 A FC Outlet Temperature 1 A CCW HX Flow 2 A CCW HX Temperatire In 1 A CCW HX Temperature Out 1 A

-Electrical- D/G Status. 2 0 Safeguards Bus Volts 2 A Safeguards' Bus Amps 2 A

' Service Bus Volts 4 A Break' RCS Leak Rate 4 A SGTR Leak Rate 4 A Steamline Break Flow 4 A Feedline Break Flow 4 A 35 0011V:1/101183

HOUR MINUTES SECONDS

-DATE ANALOG DATA POINTS (206) 8 y y o y 0 0 0 09/19/83 491 E+02 .491 E+02 .489h02 .491E+02 .489E+02 .489E+02 .592E+01

.595E+01 .595E+01 .592E+01 .833E+01 .777E+02 .781h02 .781h02

.777E+02 .829E+01 .740 E+00 .744 h00 .740E+03 .744E+03 .592E+01

.595 E+01 .000E+00 .100E+03 .131E<05 .200E+02 .151&O3 .305E+02

.443E+00 .443E+00 .443E+00 .269E-02 .513E+03 .512h03 .513E+03

. .513 h03 .513E+03 .513E+03 .513E+03 .512E+03 .513E+03 .513E+03

.513E+03 .513E+03 .513E+03 .513E+03 .513E+03 .513E+03 .513E+03

.513h03 .513E+03 .513E+03 .513E+03 2513E+03 .513E+03 .513E+03

.512E+03 .513 E+03 .513E+03 .512h03 .513E+03 .513h03 .512h03

.513E+03 .513E+03 .513E+03 .512h03 .513E+03 .512E+03 .512h03

.512E+03 .512E+03 .512h03 .512E+03 .512E+03 .512E+C3 .512E+03

.512h03 .512E+03 .512E+03 .512 h03 .512E+03 .512E+03 .512E+03

.512E+03 .512E+03 .512E+03 .512E+03 .512E+03 .512E+03 .000E+00

.000 E+00 .493E+02 .335E+00 .343E+00 .000 h00 .000 &OO .000 h00

.278E+00 .278E+00 .278E+00 .278E+00 .000E+00 .000E+00 .156E+04 *

.000E+00 .000E+00 .000E+00 .000E+00 .116E+03 .124E+03 .204E+01

.000E+00 .000E+00 .000 h00 .J00E+00 .000E+00 .000E+00 .000E+00

.000E+00 .152E+00 .160E+00 .163E+00 .157E+00 .204E+01 .546E-10

.316 E-07 .311E-07 .311E-07 .308E-07 .130E+04 .131E+04 487E+02 487E+02 .134E+00 .000E+00 .567E+03 .567E+03 .567E+03 .567E+03

.544E+02 .135E+03 .103E+03 .103h03 .103E+03 .103E+03 .170E+04

.513 E+03 .513E+03 .513E+03 .513E+03 .513E+03 .513E+03 .513E+03

.513E+03 .141E+04 .567E+03 .567E+03 .530E+03 .100E+03 .000E+00

.595E+02 .000 h00 .000E+00 .000E+00 .000 h00 .481 h02 .505E+02

.516 E+02 .496E+02 .000E+00 .000E+00 .000E+00 .000E+00 .481E+02

.505E+02 .516E+02 .496E+02 .745E+03 .507E+02 .860 h02 .745E+03

.498E+02 .858E+02 .745E+03 .500E+02 .859E'? .745E+03 .506E+02

.860E+02 .000E+00 .000E+00 .000E+00 .00C' J .000E+00 .000E+00

.154 & O1 .110E+02 .727E+03 .000E+00 .000E,00 .000 h00 .636E+03

.636 E+03 .636E+03 .636E+03 000000000000000000 N DIGITAL DATA POINTS (18)

Figure 3 Magnetic Tape Data Block 36 0011V:l/101183

3.1.7.3 Transition flow charts In order to track the operators' progress through the E0P network, transition

~

flow charts will be used. These charts are like large road maps, showing all planned transitions in the entire network. For each test scenario, flowcharts will-be marked with the " expected" path through the E0Ps. These charts will serve as the reference for the observation team when noting deviations from expected procedure usage. The charts will provide space for describing deviations, and also for recording elapsed time in the scenario. Figures 6 and 7 are reduced versions of the flow charts, showing the general E0P presentation. Figure 6 shows the ORG or E-series E0Ps, while Figure 7 shows the status trees and F-series E0Ps. These flowcharts will be generated specifically for the test E0Ps, and contain the wording for each high level operator action step within each procedure. Actual size of each flowchart is approximately 2 feet by 3 feet.

3,1.7.4 Debriefing sessions ,

immediately following each test scenario, the operating crew and observation

-team will meet in a classroom to discuss any noted deviations and E0P usage in general. The observation crew will use a special debriefing questionnaire as well as their annotated transition flow charts to evaluate deviations and document operator comments on EOP usage. Data from these sessicns will be written on the special debriefing questionnaires, and all discussions will be recorded on tape. Standard size audio cassettes will be used for this purpose; again, a separate tape will be used for each separate test scenario debriefing session.

37 0011V:1/101183

1 s

i f CCMPUTER ROOM -

CAMERA 1 -

CONTROL 1 CRT INSTRUCTORS BOARD CONSOLE (ALARMS)

SIDE 1 CRT ELECTRICAL D.G AIR 1 CRT SERVICE W ,7 (ALARMS) CIRC. W k / 's ,[i

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4, , COND.

p TURBINE /

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STEAM k /~

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\ OBSERVATION CONTROL \ AD

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CVCS j:'%

f, / '\

g HVAC SPRAY BIT CRT (ALARMS) n as SI TEST. L aSSt0w00wN

_. CONT. ISOL.

Figure 4 Control Room Videc Coverage 38 0011V:1/101183

CAMERA 1 l '

00-00 00:00:00:00' b

NUNC14yg 4 PCCW PRZR CRT

/ METERS METERS PORY PRT IIES 8I48/(s I gg@i cygs

• DI 41 PCCW RCP PRZR s /5

%# CVCs R@

CONTgot

/ o s

METERS 51Tgs E" i "d E* 10RB- CIRC SERVICE stH-gsW EBC A ER OCR1 RELIE gggg, WATER A;g M'

E I sitAM EBC g,g , i

- CAMERA 2 Figure 5 Composite Video Image 39 0011V:1/101183 L

i E-0 E-1 E-2 E-3 ES-00 ES-0] ES-02 E S-U. ES-1,2 ES-3] ES-32 ES-33 ES-03 ES-0A E S-1.3 ES-1A t.-v E-l E-J Series Series Series Foldout Foldout Foldout ECA- ECA- ECA- ECA- ECA- ECA-00 IJ 23 33 32 13 ECA-2] ECA-33 ECA-32 ECA-13 E - E - EC - Foldout Foldout Foldout Foldout Figure 6 Transition Flow Chart For E-Series E0Ps 40 0011V:1/101183

F-0.1 F-0.2 F-0.3 F-0.4 F-0.5 F-0.6 UEE TREE TREE TREE TREE TREE FR-5.1 FR-C.1 FR-H.1 FR-P.1 FR-Z.1 8

. =.

FR-C.2 FR-Z.2 na 1E E

FR-5.2 FR-C.3 FR-H.2 FR-P.2 FR-Z.3 FR-1.1 8

a d

i FR-H 4 FR.lf.3 FR H.S FR-1,3 FR-1.2 i

l t

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l Figure 7 Transition Flow Chart For F-Series E0Ps l l

l l

l 41 0011V:1/101183

3.1.8 Observation Teams Each test scena*io being performed in the control room will be observed by a specially selected and trained observation team. This team will be responsible for documenting all deviations from nominal (expected) performance observed during a scenario. A team will be made up of, at least, the following personnel.

o One member familiar with ERG development, power plant operations, and the training program developed for this test.

o One member familiar with ERG development and its analytical basis, o One member familiar with ERG development and skilled in human factors evaluations.

o One member from the Seabrook Operations staff, familiar with the control room, the operators, and the E0P development, o One member from the Seabrook training staff, familiar with simulator operations, the simulator control room, and the Seabrook operator training program.

Each individual on an observation team will be carefuly instructed in his duties as an observer. Special checklists stating t.he validation criteria will be provided as constant reminders of the expected nominal performance.

Team members will be briefed in the usage of the transition flow charts which will be available for each test scenario, and of the special debriefing questionnaire.

It will be the observation team's responsibility to record all observed deviations, real or suspected, discuss them subsequently with operators during the debriefing session, make an initial evaluation of the cause, and with the operators input, suggest possible resolutions.

42 0011V:1/101183

i I

3.1.9 Test Crews Two full crews of operators will be provided by Seabrook - Operations. Each crew will consist of a unit shift supervisor, senior control roon' operator, l

and control room operator, all normally in the control room, ar.d a shift superintendent, not normally in.the control room, but on shift with the operating crew. In the event of an emergency, the shift superintendent acts as the Shift Technical Advisor for the operating crew, and fulfills the assigned duties of that position. For the purposes of this test, the STA has the sole function of monitoring the CSF Status Trees.

All the operators on one crew have previous plant operating experience (although not t.t Seabrook). Both crews are familiar with the Seabrook control room and normal plant operations.

3.1.10 Training A special training program for the operating crews has been developed. This program uses the established validation criteria to assure that the proper interfaces with the E0Ps, control room, and operators are considered. Much of the program-structure closely follows the corresponding training program used in the BASIC Validation (Reference 1). Again, however, the training is compressed into a single week because of personnel and equipment constraints.

The training program covers basically two areas. Obviously, the new set of E0Ps are covered as thoroughly as time permits, both in the classroom and on i

the simulator. Special emphasis is placed on procedure usage, since strict adherence to proper usage will be requested. However, equally important to

( ' the success of the Program, is to make the operators fully aware of the overall validation being performed. For this reason, the training program covers such things as the observation team's responsibilities, transition flow

. chart usage, video tape recordings, and especially the debriefing sessions.

l The operators will see the full list of test scenarios, and will experience i many similar transients on the simulator during the training week.

i i

I 43 -

! 0011V:1/101183 l

1

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The operators will also be familiar with the full list of validation criteria being used by the observers / evaluators during and after the actual test. This.

should encourage them to point out any difficulties which they experience in 4

actually using the E0Ps.

Most importantly, the operators will be constantly reminded that their personal abilities are not being judged,. but that they are performing a valuable service in the overall Program implementation.

t Table 10 shows the daily schedule, both classroom and simulator, for the training program. Not obvious from the table is the fact that the simulator exercises will be gradually increased in scope to function exactly as intended during the test. By the last training day, the simulator exercises will be 1 identical to the conduct of the actual test scenarios, complete with data

-taking, observation teams, and debriefing sessions.

. Appendix G presents a sample Lesson Plan for E0P (ERG) Rules of Usage. During the training week, each operator will have his own complete copy of the E0Ps to aid familiarization. These E0Ps will be identical to the ones used during the_ test, so that continuity is maintained.

F 44 0011V:1/101183

. - . . . ~ . - . -

Table 10 -

Validation Training Schedule Classroom Simulator Day 1 Validation Test Program o Reactor Trip Structure of ERGS o Spurious SI Transition Flow Chart o Loss of All AC Power

. Rules of Usage - FORMAT Critical Safety Functions .

Status Trees Non-Accident E0Ps o E-0 series o ECA-0 series Day 2 ' Loss of Coolant Accident E0Ps o DBA LOCA o E-1 series (with switchover) o ECA-1 series o Small LOCA with Inadequate Core Cooling E0Ps cooldown o FR-C.1 o ICC (small LOCA

-o FR-H.1 without ECCS)

Day 3 Secondary Break E0Ps o Secondary break o E-2 outside containment o ECA-2.1 o Secondary break Function Restoration E0Ps inside containment o S-series (MSIVS fail to close) o P-series

• o ATWS

o'Z-series

-o I-series t

45 0011V:1/101183

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

+

Table 10 (Cont)

Validation Training Schedule Classroom Simulator Day 4 SG Tube Rupture E0Ps o Small SGTR with steam o E-3 series release cooldown i o Large SGTR with backfill

~

o ECA-3 series cooldown ,

i o SGTR with secondary break in ruptured SG.

Day 5 Function Restoration E0Ps (Full test process o H-series duplication) o C-series o Loss of all FW with 3 e loss of offsite power l l

Quiz on E0P usage o Surprise transients Final Procedure Check 1

+

1 l

o 1

46 l 0011V:1/101183  !

3.2 . Assessment Phase This phase comprises the actual simulator exercises of the test scenarios including E0P usage by the operators and will total approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of operation. Observation teams will be present and all validation data wil be recorded.

All personnel orientation for test participants will be complete prior to this phase, and all necessary materials will be on hanf. Selection of the test scenarios to be run each day will be made randomly. Similarly, personnel in each observation team will be rotated daily.

For each test scenario, the fcllowing sequence will be observed:

1. The simulator instructor will use the supplied Test Scenario Run Sheet to initialize the simulation and properly align the control board.

2. One full operating crew will enter the control roam and be briefed on current plant operating and equipment status, and be given operating instructions (i.e., raise or lower power, synchronize generator, start up

-the reactor, etc.).

3. The observation crew will take up positions in the simulator room observation booth, which provides them an unobstructed view of the entire

' control room. They will have Transition Flow Charts, appropriately pre-marked for the scenario, on which to record their observations and clock times. Also available in the observation booth will be several sets of E0Ps, lists of the validation criteria, and a list of " behavioral symptoms" to watch for (from Reference 3), shown here as Table 11. The booth will be equipped with a TV monitor showing the images being recorded by the two cameras, and the digital time; the monitor will also supply the audio track being recorded, so the observers can hear what is being said by the. operators.

47

'0011V:1/101183

- . _ - -= . , - _ _ -

TABLE 11 BEHAVIORAL SYMPTOMS TO AID AN OBSERVER COMMISSION TYPE ERROR INDICATORS o does not walk to correct area of control room on first try o does not look at correct display or does not look in correct direction on-first try o does'not touch correct control on first try 4 o does not set control to correct value on first try o performs an action not in the procedure o selects wrong procedures o selects too many procedures SEQUENCE TYPE ERROR INDICATOR o performs action out of sequence OMISSION TYPE ERROR INDICATORS o does not perform an action or step o allows a limit to be exceeded o fails to detect key signal o- fails to perform task within allotted time UNCERTAINTY INDICATORS o has to interpolate from charts, graphs, etc.

o has to re-read procedures.

O takes excessive time to read procedures o  ! takes excessive time to complete action o cannot remember what to do once procedures have been read o does not use procedures (when procedures are available is tentative, confused) o cannot find key information in procedures 48-0011V:1/101183

4. The simulator instructor will synchronize the digital clock on the video screen with the computer clock for this scenario (zero both), and then activate the simulation and the computer data recording program.

5. After a minute or two of " steady-state" operation, the. malfunctions specified on the Run Sheet for this test scenario will be input at the

-instructor's console.

~

6. The. operators will respond to indicated plant conditions using the E0Ps.

Observations and plant data are recorded.

7. At some appropriate time, determined by the observation team leader, the test scenario is terminated.

8.. Operators and observation team proceed to debriefing room, taking along the video tape and all observation notes. In the computer room, the

' stored plant data is copied to magnetic tape.

9.11n the debriefing session, the tape cassette recorder is turned on, and

-discussion of the scenario follows. A standard questionnaire is used, plus all noted deviations are discussed. Comments are recorded on forms provided for this' purpose. Copies of the E0Ps'are available in the debriefing room, plus a video player, if the tape needs to be replayed for reference. At the same time, a second operating crew and observation team are beginning the next test scenario in the control room.

10. All data for this scenario is cataloged by assigned number for future reference and analysis:

Magnetic tape

-Video tape Debriefing cassette Run sheet Transition flow charts Comment-sheets 49 0011V:1/101183

y, p .

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{ '

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4 11.,At\hispoint,theoperatingcrewandobservationteamarefreeuntil

't called for their next test scenario.

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• A separate c,lassroom in,the s trainingg bui ding will be equipped with a video i monitor (s) showint the' cor. trol room act) ity, for any non participating j -

.. 7 .

p observers. A cop /,0f thyyPs, a(rd ' am. apletransitionfjo tharts and s

(-,j questionnaires will be provided for reference.

?

. 7. ^

3.3 Resolution' Phase [

c F ,r 5' This phase will involve review ard vesolution of all noted deviations. It I

[ '

9111 include analysis of plant tr'ansient data / video tapes / debriefing

- \ recordings as npcessary to clarify the context of a deviation, and possibly to suggest or modify a resolution. Out of this phase will come statements of

!/ SYSTEM validity, and also guantitative validation results for each element in f,

the SYSTEM.

. [j is

'3. 4 Documentation Phase 4, f, .

>. .x. ,

Documentation for3 tbe Validation Program will include

i

'a

' ~

ofDiscussionofalltestscenariosasruh[onthesimulator ]

y . I, ' o Listing of all deviations and resolutions o Recommendations for improvements in trafning ,

, 8 s +

o Recommendation for improvements in Valida,  ; - tion (simulator) testing o Summary of the program , ,

[ ,,? I 4.0 Application to Plant-Specific Validation Tbis section will address application of the " reference" method of validation to be.used by any utility generating E0Ps based on the ERGS.

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5.0 -References

~

1. ' Summary Report - Emergency' Response Guidelines-Validation Program, sWCAP-10204, September 1982.

2. Emergency Operating Procedures Validation Guidelines, INP0-83-006, July, 1983.

.; 3. - 0 RAFT -' Component Verification _And System Validation Guideline, May 9, 1983 (NUTAC).

, 4. Emergency Operating Procedures Verification Guideline, INPO-83-004, March, 1983.

5. Westinghouse Owners Group Emergency Response Guidelines - Revision 1,-

XXX, 1983.

f T

i f l 51 0011V:1/101183 .

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APPENDIX A DEFINITIONS A-1 0011V:1/101183

p t-APPENDIX A DEFINITIONS LControl Room Simulator - Dynamic device which imitates functions of control room' hardware in real, fast, or slowed time.

Emergency Operating Procedures (EOPs) - Plant procedured directing operator actions necessary to mitigate consequences of transients and accidents that cause plant parameters to exceed reactor protection setpoints, engineered safety feature setpoints, or other appropriate technical limits.

Emergency Operating Procedure Guidelines (EPGs) - Guidelines that provide technical bases for the development of E0Ps.

Emergency Response Guidelines (ERG) - A complex and detailed network of generic emergency guidance for W plants.

Function Restoration Guideline (FRG} - Those sets of operator action steps which are specifically intended to respond to a Critical Safety Function challenge as determined by plant symptoms.

Operator-Plant-Procedure-Training System (System) - To address Emergency Response Capabilities (ERC) the system elements used to mitigate the consequences of an emergency condition are as follows:

o " operator" consists of the control room operating crew.

o " plant" consists of the plant as seen from its control room with its instruments and controls. It may either include or not include a Safety Parameter Display Systen, (SPDS).

o. " procedure" consists of the E0P set and supporting system operating procedures (EOP Network).

o " training" consists of the E0P training program.

A-2 0011V:1/101183

Mock-Up - Static device (e.g.,'3-D photos, 2-D photos, drawings) which portrays control room hardware and configuration.

Optimal Recovery Guideline (ORG) - Those sets of operator action steps in the ERG network which respond to a specific,-diagnosed event. Guidance is provided to recover the plant from the event in the most efficient manner.

' Paced Simulator Performance - Method of validation whereby actions are carried out by control room operating personnel in response to cues from simulated equipment in real, fast, or slowed time.

Plant Functions - Performance requirements and objectives of the plant design, such as core cooling, reactivity control, inventory control and electricity generation.

'Real Equipment - On-line, functional hardware contained in a nuclear power plant control room.

Real Performance - Method of validation whereby actions are carried out by control room operating per'sonnel in response to cues from functional on-line equipment-in real time.

Reference Validation - Method of validation whereby data developed in a common E0P validation program is referenced by similar plants.

(Critical) Safety Functions - A limited set of plant functions which, if maintained, will prevent core damage and/or radioactivity release to the environment. An activity which assures the integrity of the physical

' barriers against radiation release.

Source Documents - Documents or records upon which the System components are based.

Status Tree - Graphical device to quickly evaluate the condit. ion of a Critical Safety Function. Identifies off-normal conditions and the appropriate FRG for restoration of the function.

A-3 0011V:1/101183

l l

f l ' Symptoms - Displayed plant characteristics which directly or indirectly Lindicate plant status.

System Operational Correctness - A characteristic of the System which

. indicates the degree to which the components are compatible.

System Validation - The overall System (operator, control room, E0Ps and training) evaluation performed-to determine that the system components work together to accomplish the desired results.

l Table-Top - Method of validation whereby an operating crew explains their step-by-step actions during a proposed event scenario to an observer / review i team.

Verification - The evaluation performed to ensure consistency between any System element and its appropriate source documents.

Walk-Through - Method of validation whereby an operating crew conducts a step-by-step enactment of their actions during a proposed event scenario without carrying out the actual control functions.

Westinghouse Owners Group (WOG) - Organization of utilities which own nuclear power plants with Westinghouse-supplied Nuclear Steam Supply Systems.

Activities involve generic engineering, licensing, and operational issues relating to Westinghouse-designed nuclear units.

Writers Guide for E0Ps - A plant document that provides instructions for writing E0Ps, emphasizing the incorporation of good writing principles.

A-4 0011V:1/101183

Appendix B Item I.C.1 from .

NUREG 0737 (Clarification of TMI Action Plant Requirement, November, 1980)

B-1 0011V:1/101183

I.C.I. GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS Position In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nur. lear Reactor Regulation required licensees of operating plants, applicants for operating licenses andlicensees of plants under con-str -tion to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, including procedures for operating with natural circulation conditions, and to conduct operator retraining (see also item I A.2 1). Emergency procedures are required to be consister t with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months af ter emergency procedure guidelines were established however, some difficulty in completing.these requirements has been experienced. Clarification of the scope of the task and appropriate schedule revisions are being developed. In the course of review of these matters on Babcock and Wilcox (B&W)-designed plants, the staff will follow up on the bulletin and orders matters relating to analysis methods and results, as listed in NUREG-0660, Appendix C (see Table C.1, items 3, 4,16,18, 24, 25, 26, 27; Table C.2, items 4, 12, 17, 18, 19, 20; and Table C.3, items 6, 35, 37, 38, 39, 41, 47, 55, 57).

Changes to Previous Requirements and Guidance A. Modification to Clarification (1) Addresses owners' group and vendor submittals.

(2) References to task action plan items I.C.8 and I.C.9.

(3) Scope of procedures review is explained.

(4) Establishes configuration control of guidelines for emergency ,

procedures.

B-2 0011V:1/101183

B. Modification to Implementation j l

_(1) Deleted reference to NUREG-0578. Recommendation 2.1.9 for item I.C.I(a)2, inadequate core cooling. I Clarification The letters'of September 13 and 27, October 10 and 30, and November 9, 1979.

required that procedures.and operator training be developed for transients and accidents. The initiating events to be considered should include the events presented in the final safety analysis report (FSAR) loss of instrumentation buses, and natural phenomena such as earthquakes, floods, and tornadoes. The purpose of this paper is to clarify the requirements and add additional requirements for the reanalysis of transients and accidents and inadequate core cooling.

Based on staff reviews to date, there appear to be some recurring deficiencies in'the guidelines being developed Specifically, the staff has found a lack of justification for the approach used (i.e , symptom , event , or function-oriented) in developing diagnostic guidance for the operator and in procedural development. It has also been found that although the guidelines take implicit credit for ope' ration of many systems or components, they do not address the availability of these systems under expected plant conditions nor do they address corrective or' alternative actions that should be performed to mitigate the event should these systems or components fail.

The analyses conducted to date for guideline and procedure development contain insufficient information to assess the extent to which multiple failures are considered. NUREG-0578 concluded that the single-failure criterion was not considered appropriate for guideline development and called for the consideration of multiple failures and operator errors. Therefore, the analyses that support guideline and procedure development should consider the occurrences of multiple and consequential failures. In general, the sequence of events fer the transients and accidents and inadequate core cooling analyzed should postulate multiple failures such that, if the failures were unmitigated, conditions of inadequate core cooling would result.

B-3 0011V:1/101183

Examples of multiple failure even s include:

(1)~ Multiple tube ruptures in a single steam generator and tube rupture in more than one steam generator:

(2) Failure of. main and auxiliary feedwater; (3) Failure of high pressure reactor coolant makeup system:

(4) An anticipated transient without scram. (ATWS) event following a loss of offsite power, stuck-open relief valve or. safety / relief valve, or loss of main feedwater; and (5) . Operator errors of omission or commission The analyses should be carried out far enough into the event to assure that all relevant thermal / hydraulic /neutronic phenomena are identified (e.g., upper head voiding due to rapid cooldown, steam generator stratification). Failures and operator errors during the long-term cooldown period should also be addressed.

The analyses should support development of guidelines that define a logical transition from the emergency procedures into the inadequate core cooling procedure including the use of instrumentation to identify inadequate core cooling conditions. Rationale for this transition should be discussed.

Additional information that should be submitted includes:

1 (1) A detailed description of the methodology used to develop the guidelines (2) Associated control function diagrams, sequence-of-event diagrams, or others. if used; (3) The bases for multiple and consequential failure considerations; (4) -Supporting analysis, including a description of any computer codes used; and-(5) A description of the applicability of any generic results to plant-specific applications.

B-4 0011V:1/101183

Owners' group or vendor O bmittals may be referenced as appropriate to support this reanalysis. If owners' group or vendor submittals have already been forwarded to the staff for review, a brief description of the submittals and justification of their adequacy to support guideline development is all that is required.

-Pending staff approval of the revised analysis and guidelines, the staff will continue the pilot monitoring of emergency procedures described in Task Action Plan item I.C.8 (NUREG-0660). For PWRs, this will involve review of the loss of coolant, steam generator-tube rupture, loss of main feedwater, and inadequate core cooling procedures. The adequacy of each PWR vendor's guidelines will be identified to each NTOL during the emergency procedure review. Since the analysis and guidelines submitted by the General Electric Company (GE). owners' group that comply with the requirements stated above have been reviewed and approved for trial implementation on six plants with applications for operating licenses pending, the interim program for BWRs will consist of trial implementation on these six plants.

Following approval of analysis and guidelines and the pilot monitoring of emergency procedures, the staff will advise all licensees of the adequacy of the guidelines for application to their plants. Consideration will be given to human factors engineering and system operational characteristics, such as information transfer under stress, compatibility with operator training and control room design, the time required for comoonent and system response, clarity of procedural actions, and control-room personnel interactions. When

~

this determination has been made by the staff, a long-term plan for emergency procedure review, as described in task action plan item I.C.9, will be made available. At that time, the reviews currently being conducted on NTOLs under item I.C.8 will be discontinued, and the review required for applicants for operating licenses will be as described in the long-term plan. Depending on the information submitted to support development of emergency procedures for each reactor type or vendor, this transition may take place at different times. For example, if the GE guidelines are shown to be effective on the six B-5 0011V:1/101183

plants chosen for pilot monitoring, the long-term plan for BWRs may be complete in early -1981. Operating plants and applicants will then have the

-option'of implementing the long-term plan in a manner consistent with their operating schedule: provided they meet the final date required for implementation. .This may require a plant that was reviewed for an operating

license.under item I.C.8 to revise its emergency procedures again prior to the final implementation date for Item I.C.9. The extent to which the long-term

.. program will include review and approval of plant-specific procedures for operating plants has not been established. Our objective, however, is to minimize the amount of plant-specific procedure review and approval required.

The staff believes this objective can be acceptably accomplished by concentrating the staff review and approval'on generic guidelines. A key element in meeting this objective is the use of staff-approved generic guidelines and guideline revisions by' licensees to develop procedures. For this approach to.be effective, it'is imperative that, o'nce the staff has issued approval of a guideline, subsequent revisions of the guideline should not be implemented.by licensees until reviewed and approved by the staff. Any changes in plant-specific procedures based on unapproved guidelines could constitute an unreviewed safety issue under 10 CFR 50.59. Deviations from this approach on a plant-specific basis would be acceptable provided the basis is submitted by the licensee for staff review and approval. In this case, devir. ions from generic guidelines should not be implemented until staff approval is formally received in writing. Interim implementation of analysis and procedures for small-break loss-of-coolant accident and inadequate core cooling should remain on the schedule contained in 'mREG-0578, Recommendation 2.1.9.

Applicability ,

This requirement applies to all operating reactors and applicants for operating license.

B-6 0011V:1/101183

Implementation Reanalysis of transients and accidents and inadequate core cooling and prepar-ation of guidelines for development of emergency procedures should be completed and submitted to the NRC for review by January 1, 1981. The NRC staff will review the' analyses and guidelines and det' ermine their acceptability by July 1, 1981, and will issue guidance to' licensees on preparing emergency pro-

- cedures from the guidelines. Following NRC approval of the guidelines, licensees and applicants for operating licenses' issued prior to January 1, 1982, should revise and implement their emergency procedures at the first refueling outage after January 1, 1982. Applicants for operating licenses issued _ after January 1,1982 should implement the procedures prior to operation. This schedule supersedes the implementation schedule included in NUREG-0578, Recommandation 2.1.9 for item I.C.1(a)3, Reanalysis of Transients

- and Accidents. For those licensees and/or owners groups that will have difficulty'in attaining the January 1,-1981 due date for submittal of guidelines,'a comprehensive program plan, proposed schedule, and a detailed justification for all delays and problems shall be submitted in lieu of the guidelines.

Type of Review A preimplementation review of guidelines will be performed.

A preimplementation review of procedures will be performed.

Documentation Required See above, " Implementation."

f T_echnical Spucification Changes Required L Changes to technical specifications will not be required References (Deleted) >

l B-7 l 0011V:1/101183 l

[ l

J Appendix C Acceptance Criteria for Individual Guidelines 0040V:1/101183 C-1

E-0: ' Reactor Trip or Safety Injection

Purpose:

o To verify proper response of automatic protection systems following manual or automatic actuation of reactor trip or safety injection.

o To identify appropriate optimal recovery guideline for subsequent recovery.

Criteria for Technical Validation; o Optimal plant status can be monitored and maintained:

CSFs not unduly challenged Automatic actions and equipment verified operating as designed RCPs-tripped as necessary Low head SI pumps and diesel generators stopped when not required.

SG levels maintained l-l o An Appropriate Optimal Recovery Guideline Selected:

t i

Reactor trip without SI (ES-0.1)

' Loss of all AC power (ECA-0.0)

Secondary break (E-2)

SG tube rupture (E-3) l .- Loss of reactor coolant without tube rupture (E-1)

LOCA outside containment (ECA-1.2) o Appropriate Function Restoration Guideline Selected Before Monitoring of Status Trees Initiated-ATWS (FR-S.1)

Loss of Feedwater (FR-H.1) 0040V:1/101183 C-2

ES-0.1 Reactor Trip Response-

Purpose:

To provide.necessary instructions to stabilize and control the plant following

.a reactor trip without a safety injection.

Criteria For Technical Validation o ' Optimal Plant Status Can Be Monitored and Maintained

- ' . Plant stabilized at no-load conditions Necessary plant equipment checked and restoration attemptea if necessary

~

RCPs restarted if possible CSFs not unduly ch'allenged

-o Appropriate .Long-Term Action Selected Natural circulation cooldown.(ES-0.2)

Applicable normal plant procedure

. 0040V:1/101183 C-3

1 ES-0.2 Natural Circulation-Cooldown

Purpose:

To provide actions to-perform a natural circulation RCS cooldown and

-depressurization-to' cold. shutdown without upper-head void formation.

s Criteria For Technical Validation o L0ptimal Plant Status Can Be Monitored and Maintained RCP restarted if possible without-loss of RCS pressure control 4

Boration to Cold Shutdown Boration completed before cooldown I -

CRDM fans started if possible Cooldown under prescribed limits completed.

Depressurization completed without void formation CSFs not unduly challenged

! o Appropriate Long-Term Action Selected Completion.of this guideline for no void growth ES-0.3 or ES-0.4 if cooldown with void is necessary i

n i-l' J

4 0040V:1/101183-- C-4 i

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

ES-0.3' Natural Circulation Cooldown With Steam Void in Vessel (With RVLIS)

~

Purpose:

- To' provide actions to continue natural circulation cooldown and depressur-ization to cold shutdown under conditions that allow for the potential formation of a~ steam void in the vessel with a vessel without a vessel level system available to monitor void growth.

Criteria for Technical Validation 4

o Optimal Plant Status Can Be Monitored and Maintained:

RCP restarted if possible without loss of RCS pressure control Cooldown under prescribed limits completed Depressurization completed with void maintained within specified limits CSFs-not unduly challenged i -

J 0040V:1/101183 C-5

4 ES-0.4 Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)

Purpose:

To' provide actions to continue natural circulation cooldown and depressur-ization to cold shutdown under conditions that allow for the potential formation of a steam ' void in the vessel without a vessel . level system

, available to monitor void growth.

Criteria for Technical' Validation:

~

o Optimal Plant Status Can be Monitored and Maintained:

RCP restarted if possible without loss of RCS pressure control Cooldown under prescribed limits completed Depressurization completed with void maintained within specified limits CSFs not unduly challenged 1

- 0040V:1/101183 C-6

E-1: Loss of Reactor or Secondary Coolant-

Purpose:

o To maintain all critical safety functions satisfied o .To place the plant in an optimal condition following a loss of reactor or secondary coolant

' Criteria for Technical Validation:

o Optimal Plant Status Can IBe Monitored and Maintained CSFs not unduly challenged RCPs tripped as necessary SG 1evels maintained

- . Low head SI and containment spray stopped when not required o Coincident or Subsequent Failures can be Detected and Addressed in a

-Reasonable Time:

Secondary break (E-2)

SG_ tube rupture (E-3)

Loss of all AC power (ECA-0.0)

Critical safety function challenges (appropriate FRGs)

-- Loss of ECR capability (ECA-1.1) o Appropriate Long Term Recovery Method can be Selected and Performed:

SI termination (ES-1.1)

- Cooldown/depressurization to RHR (ES-1.2)

Emergency Coolant Recirculation '(ES-1.3 and ES-1.4) i 1

l 0040V:1/101183 C-7

ES-1.1 SI Termination

Purpose:

To provide instructions to terminate safety injection and stabilize plant conditions.

-Criteria for Technical Validation

. o Optimal Plant Status can be Monitored and Maintained:

- CSFs not unduly challenged

- SI pumps can be stopped and plant condition stabilized in acceptable time period

- Necessary plant equipment checked and restoration attempted if necessary

- RCPs restarted if possible-without loss of RCS pressure control o Appropriate Long Term Recovery Selected

- SI required-(E-1, ES-1.2)

- SI not required (appropriate plant procedure)

- o- Coincident or Subsequent Failures can be Detected and Addressed in a Reasonable Time Secondary break (E-2)

- SG tube rupture (E-3)

Loss of AC power (ECA-0.0)

CSF challenges (appropriate FRGs) 4 0040V:1/101183 C-8

o -

ES-1.2 Post LOCA Cooldown and Depressurization

-Purpose:

To provide actions to cool down and depressurize the RCS to cold shutdown conditions-following a loss of reactor. coolant inventory.

Criteria' for Technical Validation o Optimal. Plant Status can_be Monitored and Maintained:

CSFs not undully-challenged SI pumps can be stopped and plant depressurized to cold shutdown in acceptable time period

_Necessary plant equipment checked and restoration attempted if necessary SG levels maintained Cooldown completed within prescribed limits

-RCSP! restarted if possible

--SI reinitiation performed if required o Coincident'or Subsequent Failures can be detected and addressed in a Reasonable Time

-1 Secondary break (E-2)

- _SG tube rupture (E-3)

Loss of all AC power (ECA-0.0)

CSF challenges (appropriate FRG)

RWST inventory depletion (ES-1.3) 4 0040V:1/101183 C-9 L ,--.-

ES-1.3 Transfer to Cold Leg Recirculation Purpose To provide instructions for transferring the safety injection system and containment spray system to the recirculation mode.

Criteria for Technical Validation o Optimal system switchover can be completed o Loss of ECR identified and ECA-1.1 guideline implemented 0040V:1/101183 C-10

K

-E-2: Faulted Steam Generator Isolation

Purpose:

-.o To identify and isolate a faulted steam generator i

^ ~

" Criteria for Technical Validation:

~

- oIden'tification and Isolation of faulted SG can be. performed:

o Appropriate long term recovery method can be selected and performed:

- : Loss of reactor or secondary coolant alone or in combination (E-1)

SG tube ' rupture (E-3)

- -Uncontrolled.depressurization of all. steam generators (ECA-2.1) w s

i 4

4 s.

0040V:1/101183 C-11

E-3: Steam Generator Tube Rupture

Purpose:

o To stop primary-to-secondary leakage for a steam generator tube rupture event and determine'the proper recovery guideline

-Criteria for Technical Validation:

o 'All steam generators with failed tubes are identified o ' Steam flow from all ruptured steam generators is terminated or the operator is transitioned into ECA-3.1 to' minimize primary-to-secondary leakage o Primary-to-secondary leakage is terminated or sufficiently managed to control steam generator inventory, or the operator is transitioned to ECA-3.1 to minimize primary-to-secondary leakage

o. Recovery actions directed by the E-3 guideline should not result in an orange or red path challenge to any Critical Safety Function o RCS subcooling and pressurizer level are maintained greater than instrument uncertainties after SI flow is terminated.

-0040V:1/101183 C-12

I .

ES-3.1, ES-3.2, and ES-3.1, Post-SGTR Cooldown

Purpose:

o To cooldown and depressurize the RCS to cold shutdown conditions following a' steam generator tube rupture event.

o - To ' control RCS pressure and reactor coolant makeup flow to maintain indications of adequate coolant inventory and RCS subcooling, while minimizing primary-to-secondary leakage. ,

- Criteria for Technical Validation o Pressure in the RCS and ruptured steam generators and reactor coolant temperature should decrease.

of-Recovery actions should not cause water relief from any ruptured steam generator.

o -RCS subcooling and pressurizer level are maintained greater than instrument uncertainties.

o Subsequent loss of coolant events, including steam generator tube. failures, are detected and the operator is transitioned to the appropriate optimal recovery guideline.

0040V:1/101183- C-13

ECA-3.1 SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired ECA-3.2 SGTR With Loss of Reactor Coolant - Saturated Recovery Desired

Purpose:

o To cool and depressurize the RCS to cold shutdown conditions with an unisolatable loss of reactor coolant when symptoms of a SGTR are present Criteria for Technical Validation:

o Pressurizer level and RCS subcooling are maintained greater than instrument uncertainties, or core exit temperatures continue to decrease with RVLIS indicating level above the top of the core o RCS pressure and temperature continue to decrease toward cold shutdown conditions o Cycling of SI pumps is minimized o RCS pressure and reactor coolant makeup flow are controlled to limit RCS subcooling o The operator transitions to ECA-3.2 to further reduce reactor coolant leakage if makeup water supply is critically low or steam generator overfill is imminent.

0040V:1/101183 C-14

FR-S Series

Purpose:

o l- To respond to a challenge to the Suberiticality CSF o To respond to an ATWS event (FR-S.1)'

Criteria for Technical Validation:

o. Function Restoration Can be--Implemented Reactor trip checked Emergancy boration established, if possible

- Excessive RCS cooldown identified and addressed

- l0ther CSFs not unduly challenged

' o Appropriate optimal. recovery guideline selected for long-term recovery-

-Guideline and step in effect if CSF restored o Optimal plant status can be monitored and maintained for ATWS event.

l 0040V:1/101183 C-15

i +

-c FR-C Series-

Purpose:

To; restore core cooling after a challenge to the core cooling critical safety

-function LCriteria' for1 Technical Validation i

o Function restoration can be implemented

,7 1- <SI. flow established, if possible

-- /Cooldown/depressurization can be completed within prescribed 1.imits

RCPs. restarted,Lif necessary

- Other CSFs not unduly challenged

.o. Appropriate optimal recovery selected for long-term recovery .

t r

- Guideline and step in effect Loss of Reactor or Secondary Coolant (E-1) i i

i ';

i v

\

4 t

t

'r i

(

0040V:1/101183 C-16

FR-H Series

Purpose:

- .To respond.to a challenge to the Heat Sink CSF To respond to-a loss-of-all-feedwater event Criteria for Technical Validation o Function restoration can.be. implemented:

Source of secondary feed established, if possible .

Bleed and feed heat removal mode established if necessary

' Bleed and feed terminated properly af ter restoration of secondary heat sink Overpressure SG condition can be addressed properly High level or low level condition can be addressed properly 1 Other CSFs not unduly challenged o Appropriate optimal recovery can be selected for long-term recovery:

Guidel'ine and step in effect SI1 Termination (ES-1.1) o Optimal plant status can be monitored'and maintained for Loss-of-All-Feedwater Event s

0040V:1/101183 .C-17

FR-P. Series

Purpose:

To respond to a challenge to the integrity CSF  :

Criteria for Technical Validation:

o Function restoration can be implemented:

Excessive cooling can be identified and addressed properly Excessive pressurization can be identified and addressed proper 1y RCS depressurization completed within prescribed limits Other CSrs not unduly challenged o ' Appropriate optimal . recovery can be selected for long-term recovery

' Guideline and step in effect

. Any subsequent cooldown performed within prescribed limits.

0040V:1/101183 C-18

Appendix D Example Test Scenario Run Sheets D-1 0040V/1:101183

TEST SCENARIO NO. EXAMPLE-1 ERG REV. I VALIDATION PROGRAM TEST SCENARIO RUN SHEET TI'. LE : Spurious SI INITIAL ~

CONDITIONS: 100f. Power, Equilibrium, MOL I.C. No. 15 fr^*!ENCE OF MALFUNCTIONS DESCRIPTION No. OPTION

1. Inadvertent SIS 147' N/A SPECIAL INSTRUCTIONS FOR THIS SCENARIO

1. Inform operators on shift turnover that I&C performing tests in protection racks.

F D-2 0040V/1:101183

TEST SCENARIO NO. EXAMPLE 2 ERG REV. 1 VALIDATION PROGRAM ,

l TEL. SCENARIO RUN SHEET TITLE: ' Loss of All (High Pressure) Feedwater

~

. INITIAL .

. CONDITIONS: 100% Power, Equilibrium, EOL I.C. No. 17 SEQUENCE OF MALFUNCTIONS DESCRIPTION No. OPTION

1. Trip of Both MFW Pumps 014 N/A

2. Trip of S/U FWP 056 N/A 3 . Loss of Emergency FW 152 N/A SPECIAL INSTRUCTIONS FOR THIS SCENARIO

i. Inform operators at shift turnover that maintenance is being done on h^ steam packing exhauster (HX)

. 2. Tag out DG-IB (lube oli level was low, estimate 2 more hours to lift tag) l 4

D-3 0040V/1:101183

_______l

< m___1.____.______.____..___________

TEST SCENARIO NO. EXAMPLE 3 ERG REV. 1 VALIDATION PROGRAM TEST SCENARIO RUN SHEET TITLE: Loss of. All AC Powe,r - Recovery with SI Required INITIAL CONDITIONS: 15% Power, Ready to Synchronize I.C. NO. 9 SEQUENCE OF MALFUNCTIONS DESCRIPTION No. OPTION

1. Total loss of offsite power 114 N/A

2. EDG 1A Fails to Auto Start 118 N/A

3. EDG 1B Low Lube Oil Trip 119 N/A

4. RCS Leak - SG 3 Tube Rupture 025 100% (Ramp over 30 minutes)

SPECIAL INSTRUCTIONS FOR THIS SCENARIO

1. Shift turnover instructions - Synchronize and pick up 600 MWe over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

-2. Initiate loss-of-offsite power coincident with main breaker closure.

Diesels should not start

3. Start SGTR at +10 minutes. Restore one DG at +25 minutes. Restore second DG at +35 minutes.

D-4 0040V/1:101183

Appendix E Reference ERGS and Seabrook Test E0Ps 0040V/1:101183 l

I

I l

l SEABROOK TEST PROCEDURE E-0

~

0040V/1:101183

Cod 3: Symp ton /Ti tle: Prccedura Ns.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1-T 0-7 / 10/06/83

n. PUR10SE This procedure provides actions to verify proper response of the automatic protection systems following manual or automatic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate recovery procedure.

B. SYMPIDMS OR ENTRY CONDITIONS

1. Any symptom that requires a manual reactor trip listed in ATTACHMENT A, if one has not occurred.

2. The following are symptons of a reactor trip:

a. any reactor trip annunciator lit.

b. Rapid decrease in neutron level indicated by nuclear instrumen-tation.

c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.

3. Any symptom that requires a manual reactor trip and safety injection listed in ATTACHMENT B, if one has not occurred.

4. The following are symptoms of a reactor trip and safety injection:

a. Any SI annunciator or status lamp lit.

b. ECCS pumps in service.

c. Phase A isolation.

1 of 23

.-- _ _ - - = - - -

1 Code: Symptcc/Titist Prcesdura No.  ;

Ravision No./Date:

E-0 KEACTOR TRIP OR SAFETY INJECTION 05-1300 Rev. 1-7 0-T / 10/06/83 l GThP l - l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l NOTE e Steps 1 through 14 are IMMEDIATE ACTION steps.

e -Initiate monitoring of critical sr.fety function status trees at Step 27 OR if exiting from this procedure.

.1 Verify Reactor Trip: Manually trip reactor. I_F, reactor will NOT trip, THEN go e Rod bottom lights - LIT to FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS , Step 1.

o Reactor trip and bypass breakers - OPEN e Rod position indicators -

AT ZERO ON DRPI e Neutron flux - DECREASING J

2 Verify Turbine Trip:

a. All turbine stop valves - a. Manually trip turbine.

2 CLOSED 4

4 3 Verify Power To AC Emergency

' Busses:

a. AC emergency busses - a. Try to restore power to at AT LEAST ONE ENERGIZED least one ac emergency bus.

E power can NOT be re- g e E-5 Voltmeter stored to at least one ac-  !

emergency bus, THEN go to e E-6 Voltmeter ECA-0.0, LOSS OF ALL AC E0WER, Step 1. ,

b. AC emergency busses - b. Try to restore power to ALL ENERGIZED deenergized ac emergency busses.

2 of 23 1

)

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

Cod 31 Symp tom /Titis: Procsdura No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1-7 0-7 / 10/06/83 l STt P l l ACTION /EXPLCTED RESWHSE l l RESPONSE NOT OBTAINED l 4 Check If SI is Actuated: Check if Si is required.

e SI Annunciator Lit _

SI IS REQUIRED IF:

e SI Status Monitor Light e ECCS Pumps Running 1) PRZR PRESSURE < 1850 PSIC e Phase A Isolation e Auto Start of EDGs 2) CONTAINMENT PRESSURE > 4.3 PSIG

3) STCAMLINE PRESSURE < 585 PSIG IF SI is required, THEN manually actuate.

E SI is NOT required, THEN go to ES-0.1, REACTOR TRIP RESMNSE, Step 1.

5 Verif y FW Isolation: Manually close valves as necessary .

e Flow control valves -

CLOSED e Flow control bypass valves - CLOSED e FW isolation valves -

CLOSED 6 Verify Containment Isolation Phase A:

a. Phase A - ACTUATED a. Manually actuate Phase A.

b. Phase A valves - CLOSED AS b. Manually close valves as INDICATED BY STATUS PANEL necessary.

e TRAIN a e TRAIN B 7 Verify EFW Pumps Running:

a. MD pump - RUNNING a. Manually start pump.

b. Turbine-driven pump - b. Manually open steam supply RUNh1NG valves.

3 of 23

Code: Symptsc/Titlo: Procsdura No.

R3 vision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1-7 0-7 / 10/06/83 iBTEPl l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 8 Verify ECCS Pumps Running:  !!anually start pumps.

e CCPs - TRAIN A aND B e Si pumps - TRAIN A suiD B e RHR pumps - TRAIN A AND B 9 Verify PCCW Pumps - RUNNING: Manually start pumps. )

a. Loop A - ONE PUMP RUNNING

^

b. Loop B - ONE PUMP RUNNING

c. Thermal barrier cooling pumps - AT LEAST ONE PUMP RUNNING 10 Verify Ultimate Heat Sink Manually start pumps and align Operation: valves as necessary.

a. Train A - RUNNING

1) One SW pump

- OR -

2) One Cr pump AND CT fan in TA mode

b. Train B - RUNNING i

1) One SW pump j

- OR -

2) One CT pump AND CT fan in TA mode I

d 4 of 23

Cod 31 Symptor/Titist Procsdura No.

Revision No./Date:

E-U R: ACTOR TRIP OR SAFETY INJECTION OS-1300 Kev. 1-7 0-7 / 10/06/83 ISTEPl l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 11 Verify SW Cooling To DG Jacket Water Coolers:

a. Train A cooling established

1) SW-V16 - OPEN 1) OPEN SW-V16.

2) Flow indicated 2) Continue to Step lib.

b. Train B cooling established ,

1) SW-V18 - OPEN 1) OPEN SW-V18.

J

2) Flow indicated 2) Continue to Step 12.

12 Verify Containment Ventilation Close valves immediately.

Isolation:

a. Containment air purge valves -

CIASED INDICATED ON STATUS PANEL e CAP-V1 e CAP-V2 e CAP-V3 e CAP-V4

b. Containment on-line purge

, valves - CLOSED INDICATED ON STATUS PANEL e CDP-Vi e CDP-V2 e 00P-V3 e -CDP-V4 5 of 23

Code Symptc:/Titis: Procedure flo.

Revision flo./D. Le:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Re v. 1 -7 0-T/ 10/06/83 lMTSPl l ACTION / EXPECTED RESPONSE l l RESFUNSE NOT OBTAINED l 13 Check If Main Steamlines should Be Isolated:

a. Steam line isolation is a. Go to Step 14. i dEQUIRED IF:

e Any steamline - LESS THAN OR EQUAL TO 585 PSIG 1

I e - Containment pressure is -

GREATER ThAN OR equal TO 111-2 SETPOINT 4.3 PSIG

b. Verify MSIV AND MSIV bypass b. Manually close valves.

valves - CLOSED t

i i

6 i

6 of 23

Cod 2: Symptor/

Title:

Procedura No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Re v. 1-7 0-T/ 10/06/83 l STEP l l ACTION /EXPLt,iSD RESPONSE l l RESPONSE NOT OBTAINED l 14 Check Containment Pressure - g pressure has gone greater than itAS R&laINED LESS TilAN 11I-3 18 PSIG, THEN:

SETIVINT, 18 PSIC, BY PRESSURE RECORDING a. Verify containment spray initiated.

1) Train A in operation.

e CBS Pump A - RUNNING e CBS-Vil - OPEN e CBS Pump A discharge pressure - LESS THAN 310 PSIG e Miniflows - CLOSED

2) Train B in operation.

e CBS Pump B - RUNNING e CBS-Vl? - OPEN e CBS Pump B discharge pressure - LESS THAN 310 PSIG e Miniflows - CLOSED IF NOT initiated, THEN Enually initiate.

1) HOLD BOTH manual activate switches in a train AND place to ACTUATE.

b. Verify containment Phase B

('P' signal) valves actuate to proper position on status light panel. E status light panel does NOT indicate proper position, manually position valve s .

- c. Stop all RCPs.

i 7 of 23 l

l Cod t fymptoc/Titis: Procsdurs No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION 0S-1300 dev. 1-7 0-T/ 10/06/83 LSTEPl l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 15 Verify ECCS Flow:

a. CCP flow indicators - a. Manually start pumps and CliECK FOR FLOW Ti!ROUGil align valves. REFER TO BIT ATTACHMENT A, ECCS VALVE ALIGNMENT - CCP'VIA BIT TO RCS 00LD LEGS.

b. RCS pressure - LESS THAN b. Go to Step 16.

1550 PSIG e RC-PI-405 e RC-PI-403

c. SI pump flow indicators - c. Manually start pumps and CHECK FOR FLOW align valves. Refer to ATTACletENT B, ECCS VALVE e TRAIN A ALIGNMENT - SIP TO RCS COLD e TRAIN B LEGS.

d. RCS pressure - LESS THAN d. Go to Step 16.

200 PSIG

e. RHR pump flow indicators - e. Manually start pumps and CHECK FOR FIDW align valves. REFER TO ATTACliMENT C, ECCS VALVE e TRAIN A ALIGNMENT - RHR PUMP TO e TRAIN 8 RCS 00LD LEGS.

16 Verify EFW Flow - GREATER THAN Manually start pumps. E proper 470 GPM TOTAL COMBINED FLOW T0 flow can NOT be established, THEN

. AT LEAST TWO SGs go to FR-H.1, RESPONSE TO LOSS OF SECONDARY llEAT SINK, STEP 1.

i i

l i

I.

8 of 23

Cod 2: Sympton/Titlos Precsdura Na.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION 0S-1300 de v. 1-7 0-7/ 10/06/83 lbTEPl l ACTION / EXPECTED RES10NSE l l RESPONSE NOT OBTAINED l NOTE liigh EFW flow to a faulted SG will cause automatic closure of that SG's EFW flow control valves.

17 Verify EFW Valve Alignment - Manually align valves as PROPER EMERGENCY ALIGNMENT necessary.

REQUIRED SG VALVE NOMENCLATURE POSITION A FV-4214A, FLOW CONTROL O PEN FV-4214B, FLOW CONTROL OPEN b FV-4224A, FIDW CONTHOL OPEN FV-4224B, FLOW CONTROL OPEN C FV-4234A, FLOW CONTROL O PEN FV-4234B, FLOW CONTROL OPEN D FV-4244A, FWW CONTROL OPEN FV-4244B, FIDW CONTROL OPEN 9 of 23

Coder Sympt:I:/ Tit 10: Proc: dura No.

Revision No./Date:

E-O REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1-7 0-7 / 10/06/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 18 Verify ECCS Valve Alignment - Manually align valves as PROPER EMERGENCY ALIGNMENT necessary. ECCS valve align-INDICATED ON STATUS PANELS ment checklists are provided as ATTACletENTS C, D and E.

e TRAIN A e TRAIN B 19 Check RCS Teciperature - STABLE IF temperature less than 557'F AT OR TRENDING TO 557'F atid decreasing TiiEN:

e WR LDOP TEMPERATURE RECORDERS a. Stop dumping steam.

b. IF cooldown continues, T11EN throttle total EFW flow but not less than 470 GPM -

TOTAL 00MBINED FLOW. Main-tain WR level above top of SG U-tubes.

LEVEL A150VE SG U-TUBES ADVERSE CONTM NORMAL CONTM NARROW RANGE WIDE RANGE LEVEL GREATER LEVEL GREATER THAN 28% THAN 65%

c. IF cooldown continues, THEN close MSIVs AND MSIV bypass valve s .

E temperature greater than 557'F and increasing, THEN:

e Manually dump steam to l condenser

- OR - 4 o Manually dump steam with SG ASDVs.

1 10 of 23

Cod:: Sympton/Titis: Procsdura No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION 05-1300 Re v. 1 -7 0-T / 10/06/83 lSTEPl' l ACTION / EXPECTED RESNHSE l l RESMNSE NOT OBTAINED l 20 Check PRZR PORVs and Spray Valves:

a. WRVs - CIASED a. E PRZR pressure less than 2385 psig, THEN manually close FORVs. g any valve can NOT be closed THEN manually close its block valve . IF block valve can NOT be cEsed, TilEN go to E-1, LOSS OF REACTOR OR SECDNDARY CDOLANT, Step 1.

b. Normal PRZR spray valves - b. E PRZR pressure less than CLOSED 2260 psig, THEN manually close valves. IF valves can NOT be closed, THEN stop RCP(s) supplying failed spray valve (s).

e PCV-455A RC-P-1C e PCV-455B RC-P-1A

't NOTE Seal injection flow should be maintained to all RCPs.

21 Check If RCPs Should be Stopped:

, a. High Head ECCS Pumps - AT a. Go to Step 22.

LEAST ONE RUNNING e Centrifugal Charging Pump

- OR -

l e S1 Pump

b. RCP Trip Parameter - LESS b. Go to Step 22.

THAN 1375 PSIG IN RCS

c. Stop all RCPs 11 of 23

Cod:: Sympt r / Tit 10: Proctdura No.

Revision No./Date:

E-0 REACTOR TRIP UR SAFETY INJECTION 0S-1300 dev. 1-7 0-7 / 10/06/83 lSTEPl l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 22 Check If SGs Are Not Faulted:

a. Check pressures in all a. Go to E-2, FAULTED STEMi SCs - CENERAE R ISOLATION, Step 1.

o NO SG PRESSURE DECREASING IN AN UNCONTR01. LED MANNER e NO SG COMPLETELY DEPRESSURIZED 23 Check If SG Tubes Are Not Go to E-3, STEAM GENERATOR Ruptured: TUBE RdirrURE, Step 1.

e Condense r ef fluent radiation - NORMAL e Main steamline radia-tion - NORMAL 24 Check If RCS Is Intact: Go to E-1, IDSS OF REACER OR SECONDARY COOLANT, Step 1.

e Containment radiation -

NORMAL e Containment pressure -

NORMAL e Containme nt building level - NORMAL I i

I

~

12 of 23 L

Codd: Symptom /

Title:

Procsdure No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 ke y. 1-T 0T/ 10/06/83 I STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 25 - Check If ECCS Flow Should Be Reduced:

a. RCS subcooling based on a. DO NOT SIDP ECCS PUMPS. Go core exit TCs - GREATER to Step 27.

THAN 30*F

b. Secondary heat sink: b. IF, neither condition satis-fled, THEN DO NOT STOP e Total EFW flow to intact CENTRIFUGAL CHARGING PUMPS SGs - GREATER THAN 470 GPtt OR SI PUMPS. Go to Step 27.

TOTAL COMBINED FLOW WITH INTACT SG WR LEVELS ABOVE TO P O F U-TUBES , 65% IVR N0lulAL Q)NTAINMENT

- OR -

e NR level in at least one intact SG - GREATER TilAN 28% FOR ADVERSE CONTAIN-MENT

c. RCS pressure - STABLE OR c. DO NOT STOP ECCS PLMPS. Go INCREASING to Step 27.

d. PRZR level - GREATER TIIAN d. DO NOT STOP ECCS PUMPS. Try 5% to stabilize RCS pressure with normal spray. Return to Step 25a.

26 Go To ES-1.1, SI TERMINATION, Step i 27 Initiate Monitoring of Critical Safety Function Status Trees CAUTION CST makeup should commence as early as possible to avoid low inventory problems.

13 of 23

Codet Symptsc/Titls Proc: dure No.

Revision No./Date:

E-0 REACTOR TRIP UR SAFETY INJECTION OS-1300

.e v. 1-7 0 -T / 10/06/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 28 Check SG Levels:

a. NR level - GREATER a. Maintain total EFW flow THAN 5% greater than 470 GPM to intact SGs until NR level greater than 5% in at least one SG. DO NOT allow intact SG WR levH to decrease below top of U-tube s .

LEVEL ABOVE SG U-TUBES ADVERSE CONTH NORMAL ODNTM NARROW RANGE WIDE RANCE LEVEL GREATER LEVEL GREATER TilAN 28% THAN 65%

b. Control EFW flow to b. IF NR level in any SG maintain NR level - Entinues to increase in an BETWEEN 5% AND 50% uncontrolled ma.r.er, THEN go to E-3, STEAM GENERATOR TUBE RUPTURE, Step 1.

29 Check Secondary kadiation - Go to E-3, STEAM GENERATOR NORMAL USING RDMS: TUBE RUPTURE, Step 1.

e Main steamlines

- OR -

e Condenser ef fluent ,

I l

l 1

14 of 23

Cod 38 Symptom /Titl28 Procsdure Ns.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1-7 0-7 / 10/06/83

$ TEPl l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 30 Check Auxiliary Building Evaluate cause of abr.ormal Radiation - NORMAL USING conditions. I_F,the F cause is a RUNS loss of RCS inventory outside containment, TilEN go to ECA-1.2, LOCA OUTSIDE CONTAINMENT, Step 1.

31 Check PRT Conditions - Evaluate cause of abnormal NO M bsL conditions.

e FORV OR SAFETY valve o Pressurizer PORV's tailpipe temperature -

LESS THAN 140*F e RilR relief valves e Pressure - BETWEEN 2 PSIG e Letdown relief valve AND 4 PSIG e Seal return header relief i

e Lesel - BETWEEN 60% AND 86% valve o Temperature - LESS TilAN 120 *F CAUTION If offsite power is lost af ter SI reset, manual action may be required to restart safeguard equip me nt .

32 Reset SI 4

33 Reset Containment Isolation Phase 'a' AND Phase 'B' 15 of 23

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

. - . . . - - _ = - _ ._ - . _ - _

Code: Symptos/Titla: Procedura No.

Revision No./Date:

E-0 P.EACTOR TRIP OR SAFETY INJECTION . OS-1300 ae v. 1 -T o-r / 10/06/83

.1 STJP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l NOTE If a 14P has occurred, reset EPS-RHO.

34 Reestablish Instrument Air Supplies While Continuing With This Guideline:

a. Restart service air a. Open fire water cooling to compressors as follows: service air compressors and start at least one

1) Open SW isolation valves service air compressor.

to SCCW e SW-V4 e SW-V5 IF on CT, OPEN SW-V74 EID SW-V76, CLOSE SW-V75

2) Start ONE SCCW pump

3) Start at least one service air compressor

b. Restart containment air compressors as follows:

1) Open PCCW containment isolation valves TRAIN A, IDOP A TRAIN B, IDOP B CC-V16 8 CC-V175 CC-V57 CC-V176 CC-V122 CC-V257 j CC-V121 CC-V256  ;

2) Start one containment l air compressor i e' IA-C-4A (Loop A cooled) e IA-C-4B (Loop B cooled) t 16 of 23

Coda: Sympton/Titis: Proc: dura No.

Revision No./Date:

E-0 REACTOR* TRIP OR SAFETY INJECTION OS-1300 Rev. 1- T O-7/ 10/06/83 l STEP l l ACTION / EXPECTED RESIONSE l l RESPONSE NOT OBTAINED l CAUTION RCS pressure should be monitored. If RCS pressure drops below 200 PSIC, RER pumps must be manually restarted to supply water to RCS.

35 Check If RilR Pumps Should Be Stopped

a. Check RCS pressure:

1) Pressure - GREATER 1) Co to E-1, IDSS OF TilAN 200 PSIG REACTOR OR SE(I)NDARY COOLANT, Step 1.

2) Pressure - STABLE OR 2) Go to Step 36.

INCREASING

b. Stop RilR pumps and place in standby 36 Check If Emergency Diesel Generators Should Be Stopped:

a. Verify AC energency busses - a. Try to restore of fsite ENERGIZED BY UATs OR RATS power to AC emergency busses. IF offsite power can NOT be restored, TilEN RESET EPS - RMO.

1) Go to Step 37

b. Stop emergency dieael -

generators and reset for AUTO START

1) Stop diesel generator

2) Reset diesel generator 37 Return To Step 19

- END -

17 of 23

Gode: Symptom /

Title:

Procedure No.

Revision No./Date:

b-0 REACTOR TRIP UR SAFETY INJECTION 0S-1300 Re v. 1 -7 07/ 10/06/83 ATTACHMENT A The following are symptc ms that require a reactor trip, if one has not occurred:

FUNCTIONAL UNIT TRIP SETPOINT A. POWER RANGE NEUTRON FLUX:

1) 14W SETPOINT . . . . . . . . . . . 1 25% OF RTP

2) tlIGH SETPOINT . . . . . . . . . . I 109% OF RTP B. POWER RANGE, NEUTRON . . . . . . . . . 15% OF RTP WITil A FLUX, HIGli NSITIVE TIME CONSTANT > 1 RATE SECONDS C. POWER RANGE, NEUTRON . . . . . . . . . . 15% OF RTP WITH A FLUX, HIGH NEGATIVE TIME CONSTANT > 1 RATE SECONDS D. INTERMEDIATE RANGE, . . . . . . . . . . I 25% OF RTP NEUTRON FLUX E. SOUKGE RANGE, NEUTRON . . . . . . . . . 1 105 CPS FLUX F. OVERTEMPERATURE AT . . . . . . . . . . i 109.95% + PENALTIES G. OVERPOWER AT . . . . . . . . . . . . . i 109% - PENALTIES H. PRESSURIZER PRESSURE -- . . . . . . . . > 1945 PSIG LOW I. PRESSURIZER PRESSURE -- . . . . . . . . < 2385 PSIG i HIGH J. PRESSURIZER WATER .. . . . . . . . . . < 92% OF INSTRUMENT LEVEL--IIIGH 5' PAN K. LOSS OF FLOW . . .. . . . . . . . . . > 90% OF 140P DESIGN FLOW 18 of 23

Codes Sympten/Titlo: Procedure No.

Revision No./Date E-0 REACTUR TRIP OR SAFETY INJECTION OS-1300 Re v. 1 -T 0-T/ 10/06/83 ATTACHMENT A (cont.)

The following are symptoms that require a reactor trip, if one has not occurred:

FUNCTIONAL UNIT TRIP SETPOINT L. STEAM GENERATOR WATER . .. . . . . . . > 15% OF NARROW RANGE LEVEL - LOW-LOW RANCE INSTRUMENT SPAN M. UNDERVOLTAGE - REACTOR . . . . . . ... > 10,200 VOLTS AC 000LANT PUMPS N. UNDERFREQUENCY - . . . . . . . . . . . > 57.2 Hz REACTOR COOLANT PUMPS P. TURBINE TRIP

1) LOW TRIP SYSTEM . . . . . . . . . > 800 PSIG PRESSURE

2) TURBINE SIUP . . . . . . . . . ALL VALVES CLOSED VALVE CLOSURE SAFETY INJECTION .. . . . . . . . . . NA INPUT FROM ESF i

i f

1 i

19 of 23

Cod 3: Symp tor /Titis: Procsdura No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION 0S-1300 Re v. 1- T ,

0-7 / 10/06/83 s

aTTACID1ENT d i

The following .are symptoms that require a reactor trip and safety injec-tion,' if one has not occurred:

FUNCT10hAL UNIT S1 SET 101NT

A. PRESSURIZER PRESSURE - . . ...... < 1850 PSIG

~

WW '

B. CONTAINMENT PRESSURE - . . ...... > 4.3 PSIG HIGH C. STEAMLINE PRESSURE . . ..... .. < 585 PSIG WW

\

?

i 1

\

l

\.

f 20 of 23

l l Codes- Symp te:-/TitI21 Prc- dure No.

Revision No./Da to:

E-0 REACTOR TRIP OR ' SAFETY INJECTION 'OS-1300 de v. 1 - T 0-7/ 10/06/83 l

1 ATTACHMENT C ECCS VALVE ALIGNMENT - CCP VIA BIT TO RCS COLD LECS b-VALVE NOMENCLATURE POSITION CS-V142 CHARGING ISOL, CLOSED CS-V143 CHARGING ISOL. CLOSED CS-LCV-112B CVCT OITILET CLOSED CS-LCV-ll2C CVCT OLTILET CLOSED CS-LCV-112D RWST OUTLET OPEN CS-LCV-112E RWST OUTLET OPEN CS-V844 BIT INLET OPEN CS-V65 BIT INLET OPEN CS-V845 BIT INLET O PEN CS-V66 BIT INLET OPEN CS-V846 BIT BYPASS CIASED CS-V847 BIT BYPASS CLOSED CS-V165 BIT RECIRC. PUMP DISCHG. - CIDSED CS-V173 BIT RECIRC. ISOL. CLOSED US-V174 u1T KHCIRC. 1S01.. CIUSED SI-V138 BIT OUTLET TO RCS OPEN SI-V139 BIT OtfrLET TO RCS OPEN 21 of 23 3:

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

. Coda: Sympter/ Tit 1s: Procsdura No.

Revision No./Date:

-E-U REACTOR TRIP OR SAFETY INJECTION OS-1300 Rev. 1 -T 0-7 / 10/06/83 ATTACHMENT D

- ECCS VALVE ALIGNMENT - SIP -

TO RCS COLD LEGS VALVE NOMENCLATURE POSITION CBS-V47 SI PUMP a SUCTION FROM RWST O PEN Ciss-V49 SI PUMP A SUCTION FROM RWST OFEN SI-V90 S1 PUMP A MIN FLOW TO RWST OPEN SI-V102 SI TO Il0T LEGS CLOSED SI-V112 SI TO COLD LEGS OPEN SI-V114 SI 10 COLD LEGS OPEN CS-V460 SI PUMP A SUCTION CROSSOVER CLOSED CS-V461 SI PUMP A SUCTION CROSSOVER CLOSED CS-V475 S1 PGiP A SUCTION CROSSOVER O PEN CBS-V51 S1 PGiP B SUCTION FROM RWST OREN CBS-V53 SI PUMP B SUCTION FROM RWST OPEN SI-V89 SI PUMP B MIN FLOW TO RWST O PEN SI-V93 S1 PUMP A & B MIN FLOW TO RWST O PEN SI-Vill SI TO COLD LEGS O PEN SI-V77 SI TO liOT LEGS CIDSED 22 of 23 i

- . , - , - - , . - - - - _ . - . . - , ,_ n, -

Codes Symptoc/Titics Procsdura No.

Revision No./Date:

E-0 REACTOR TRIP OR SAFETY INJECTION OS-1300 Re v. 1-7 0-7/ 10/06/83 ATTACHMENT E

- ECCS VALVE ALIGNMENT - RIIR -

PGiP TO RCS COLD LECS ,

VALVE NOMENCLATURE POSITION ~

CBS-V8 CONTM SGtP TO RHR PG1P A & CBS PUMP A CLOSED CBS-V2 RWST TO CBS PWIP A & RHR PmlP A OPEN RC-V23 RHR SUCTION FROM RCS CLOSED RC-V88 RHR SUCTION FROM RCS CLOSED

.RC-V87 RHR SUCTION FROM RCS CWSED KC-V22 RdR SUCTION FROM RCS CWSED CBS-V5 RWST TO CBS PUMP B & RilR Palp B O PEN CBS-V14 (X)NTM SUMP TO RHR ' PUMP B & CBS PG1P u CLOSED RH-V36 RHR TRAIN B TO SUCTION OF SI Palp B CLOSED RH-V35 RHR TRAIN A TO SUCTION OF SI PLMP A CWSED RH-V21 RHR SYSTEM B TO 10T LEGS O PEN RH-V32 RHR A/B TO tiOT LEGS CLOSED RH-V26 RHR TRAIN B TO COLD LEGS OPEN RH-V22 RHR SYSTEM A TO IDT LEGS O PEN RH-V70 RHR A/B TO HOT LEGS CWSED RH-V14 RHR TRAIN A TO COLD LEGS O PEN i i

23 of 23

OPERATOR ACTION

SUMMARY

FOR E-O SERIES PROCEDURES l 1.- .kCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below occur:

~

a. CCPs or SI pumps - AT LEaST ONE RUNNING

b. RCP Trip Parameter - LESS TilAN 1375 PSIG

~

2.' SI ACTUATION CRITERIA

- Actuate SI .and go to' E-0, REACTOR TRIP OR SAFETY INJECTION,

-Step 1, if EITHER condition listed below occurs:

e RCS subcooling based on core exit TCs - LESS THAN 30*F e Pressurizer level - CANNOT BE MAINTAINED GREATER THAN 5% '

[(30)% FOR ADVERSE CONTAINMENT]

3. RED PATH

SUMMARY

a. SUBCRITICALITY - Nuclear power greater than 5%

b. 00RE COOLING '- Core exit TCs greater than 1200*F

- OR -

Core exit TCs greater than 700*F AND RVLIS full range less than 40%

with no RCPs running

c. HEAT SINK - SG narrow range level in all SGs less than 28%

AND total feedwater. flow less than 470 gpa ,

d. INTEGRITY - Cold leg temperature decrease greater than 100*F in 3

last 60 minutes AND RCS cold leg temperature less than 250*F

e. CONTAINMENT - Containment pressure greater than 52 PSIG t

4.- EFW SUPPLY Commence CST makeup as soon as possible to avoid low inventory p roblems .

WOG EMERGENCY RESPONSE GUIDELINE E-0 E-3 0040V/I:101183

SEP O 1 lge3 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 1

A. PURPOSE This guideline provides actions to verify proper response of the automatic protection systems following manual or automatic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate recovery guideline.

B. SYMPTOMS OR ENTRY CONDITIONS

1) The following are symptoms that require a reactor trip, if one has not occurred:

[ Enter plant specific setpoints and requirements]

2) The following are symptoms of a reactor trip:

.a. Any reactor

• rip annunciator lit.

b. Rapid decrease in neutron level indicated by nuclear instrumentation.

c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.

l

3) The following are symptoms that require a reactor trip and safety injection, if one has not occured:

[ Enter plant specific setpoints and requirements]

4) The following are symptoms of a reactor trip and safety injection:

a. Any SI annunciator lit.

b. SI pumps running:

[ Enter plant specific list]

63798:1/082983 1 of 14 ,

SEP 011983 i 1

- E-0~ REACTOR TRIP OR SAFETY lilJECTION HP-REV. 1 l FINAL NOTE: o Steps 1 through 14 are IMMEDIATE ACTION steps.

o Foldout page should be open.

.1 Verify Reactor Trip: Manually trip reactor. IF ,

reactor will NOT. trip, THEN go o Rod bottom lights - LIT to FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS, Step 1.

o Reactor trip and bypass breakers - OPEN o Rod position indicators -

AT.ZERO o Neutron flux - DECREASING 2 Verify Turbine Trip:

i

a. All turbine stop valves - a. Manually trip turbine.

CLOSED

-3 Verify Power To AC Emergency Busses:

a. AC emergency busses - a. Try to restore power to at AT LEAST ONE ENERGIZED least one ac emergency bus.

IF power can NE be restored to at least one ac emergency bus, THEN go to ECA-0.0, LOSS OF ALL AC POWER, Step 1.

b. AC emergency busses - b. Try to restore power to

-ALL ENERGIZED deenergized ac emergency busses.

4 Check If SI Is Actuated: ,

Check if SI is required. IF SI is required, THEN manually

[ Enter plant specific means]' actuate. IF SI is NOT required, THEN go to ES-0.1, REACTOR TRIP RESPONSE, Step 1.

f 6379B:1/082983 2 of 14

SEP 011963 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 5 Verify FW Isolation: Manually close valves as necessary.

o Flow control valves -

CLOSED o Flow control bypass valves - CLOSED o FW isolation valves -

CLOSED o SG blowdown isolation valves - CLOSED o SG sample isolation valves - CLOSED 6 Verify Containment Isolation Phase A:

a. Phase $"-ACTUATED a. Manually actuate Phase A.

b. Phase A valves - CLOSED b. Manually close valves.

7 Verify AFW Pumps Running:

a. MD pumps - RUNNING a. Manually start pumps.

b. Turbine-driven pump - b. Manually open steam supply RUNNING IF NECESSARY valves.

8 Verify SI Pumps Running: Manually start pumps.

o Charging /SI pumps -

RUNNING o High-head SI pumps - RUNNING o Low-head SI pumps - RUNNING 9 Verify CCW Pumps - RUNNING Manually start pumps.

t i

i 6379B:1/082983- 3 of 14

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 10 Verify Service Water Pumps - Manually start pumps.

RUNNING ,

11 Verify Containment Fan Coolers - Manually start fan coolers in RUNNING IN EMERGENCY MODE emergency mode.

12 Verify Containment Ventilation Isolation:

a. Dampers - CLOSED a. Manually close dampers.

[ Appropriate steps for verification of other essential equipment as required by the specific plant design should be placed after Step 12.]

I- 13 Check If Main Steamlines Should Be' Isolated:

a. [ Enter plant specific means a. Go to Step 14.

orsetpoints]

b. Verify main steamline b. Manually close valves.

isolation and bypass valves - CLOSED 14 Verify Containment Spray Not Required:

a. Containment Pressure - a. Perform the following:

HAS REMAINED LESS THAN (1) PSIG 1) Verify containment spray initiated. IF NOT, THEN manually initiate.

2) Verify containment isolation Phase B valves closed. IF NOT, THEN manually close valves.

3) Stop all RCPs.

63798:1/082983 4 of 14 W r --

-me-M/ m " "

7 e-ey- N._ c- F t - - - ' + - +v

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 15 Verify SI Flow:

a. Charging /SI pump flow a. Manually start pumps and indicators - CHECK FOR align valves.

FLOW

b. RCS pressure - LESS b. Go to Step 16.

THAN (2) PSIG [(3) PSIG FOR ADVERSE CONTAINMENT]

c. High-head SI pump c. Manually start pumps and flow indicators - align valves.

CHECK FOR FLOW

d. RCS pressure - LESS d. Go to Step 16.

THAN (4) PSIG [(5) PSIG FOR ADVERSE CONTAINMENT]

e. Low-head SI pump flow 2. Manually start pumps and indicators - CHECK align valves.

FOR FLOW 4

16 Verify AFW Flow - GREATER THAN Manually start pumps and align (6) GPM valves as necessary. IF AFW flow greater than (6) gpm can NOT be established, THEN go to FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, Step 1.

17 Verify AFW Valve Alignment - Manus'ly align valves as l

PROPER EMERGENCY ALIGNMENT necessary.

18 Verify SI Valve Alignment - Manually align valves as PROPER EMERGENCY ALIGNMENT necessary.

i -

6379B:1/082983 5 of 14

SEP 011963 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 19 Check RCS Average Temperature - IF temperature less than (7) F STABLE AT OR TRENDING TO (7)'F and decreasing, THEN:

a) Stop dumping steam.

b) IF cooldown continues, THEN control total feed flow. Maintain total feed flow greater than (6) gpm until narrow range level greater than (8)% [(9)% FOR ADVERSE CONTAINMENT] in at least one SG.

c) IF cooldown continues, THEN close main steamline isolation and bypass valves.

IF temperature greater than

. [7)*F and increasing, THEN:

o Dump steam to condenser.

-OR-o Dump steam using SG PORVs.

63798:1/082983 6 of 14

, - - . . . - - . ,,--+--e n ---n.. , . . , - - -----r., .,-.,w~ ,-.

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 4

20 Check PRZR PORVs And Spray Valves:

a. PORVs - CLOSED a. IF PRZR pressure less than (10) psig, THEN manually close PORVs. IF any valve can NOT be closed, THEN manually close its block valve. IF block valve can NOT be cTosed, THEN go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

b. Normal PRZR spray b. IF PRZR pressure less than valves - CLOSED {11)psig,THENmanually .

close valves. IF valves can NOT be closed, THEN stop RCP(s) supplying failed spray valve (s).

NOTE: Seal injection flow should be maintained to all RCPs.

21 Check If RCPs Should Be Stopped:

a. SI pumps - AT LEAST ONE a. Go to Step 22.

RUNNING o Charging /SI

-OR-o High-head SI

b. RCP Trip Parameter - b. Go to Step 22.

LESS THAN (12) [(13)

FORADVERSECONTAINMENT]

c; Stop all RCPs 6379B:1/082983 7 of 14

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 22 Check If SGs Are Not Faulted:

a. Check pressures in all a. Go to E-2, FAULTED STEAM

~

SGs - GENERATOR ISOLATION, Step 1.

o NO SG PRESSURE DECREASING IN AN UNCONTROLLED MANNER o NO SG COMPLETELY DEPRESSURIZED 23' Check If SG Tubes Are Not Go to E-3, STEAM GENERATOR Ruptured: TUBE RUPTURE, Step 1.

o Condenser air ejector radiation - NORMAL o SG blowdown radiation -

NORMAL 24 Check If RCS Is Intact: Go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

o Containment radiat-lon -

NORMAL o Containment pressure -

NORMAL o Containment recirculation sump level - NORMAL' l

6379B:1/082983 8 of 14

SEP 011963 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL

'25' Check If SI Flow Should Be '

Reduced:

a. RCS subcooling based on a. DO NOT STOP SI PUMPS. Go to core exit TCs - GREATER Step 27.

THAN (14)'F

b. Secondary heat sink: b.

IF neither condition satisfied, THEN 00 NOT STOP o Total feed flow to SGs - SI PUMPS. Go to Step 27.

GREATER THAN (6) GPM

-OR-o Narrow range level in at least one SG - GREATER THAN (8)%

c. RCS pressure - STABLE OR c. DO NOT STOP SI PUMPS. Go to INCREASING Step 27.

d. PRZR level - GREATER THAN d. DO NOT STOP SI PUMPS.

(15)% Try to stabilize RCS pressure with normal PRZR spray.

Return to Step 25a.

26 Go To ES-1.1, SI TERMINATION, Step 1 6379B:1/082983 9 of 14

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 27 Initiate Monitoring Of Critical Safety Function Status Tree,s CAUTION: Alternate water sources for AFW pumps will be necessary if CST level decreases to less than (16).

28 Check SG Levels:

a. Narrow range level - a. Maintain total feed GREATER THAN (8)% flow greater than (6) gpm until narrow range level greater than (8)% in at least one SG.

b. Control feed flow to maintain b. IF narrow range level in any narrow range level between SG continues to increase in (8)% and 50% an uncontrolled manner, THEN go to E-3, STEAM GENERATOR TUBE RUPTURE, Step 1.

29 Check Secondary Radiation - Go to E-3, STEAM GENERATOR TUBE NORMAL RUPTURE, Step 1.

[ Enter plant specific means]

30 Check Auxiliary Building Evaluate cause of abnormal Radiation - NORMAL conditions. IF the cause is a loss of RCS, inventory outside containment, THEN go to ECA-1.2, LOCA OUTSIDE CONTAINMENT, Step 1. ,

31 Check PRT Conditions - NORMAL Evaluate cause of abnormal

[

conditions.

63798:1/082983 10 of 14 t

. s .- . , - _ . - - _ . .- - ._ --

SEP 0 i 1983 E-0 RFACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL CAUTION: If offsite power is lost after SI reset, manual action may be required to restart safeguards squipment.

32 Reset SI 33 Reset Containment Isolation Phase A And Phase B 34 Establish Instrument Air To Start one air compressor and Containment establish instrument air to containment.

CAUTION: RCS pressure should be monitored. If RCS pressure decreases to less than (4) psig the low-head SI pumps must be manually restarted to supply water to the RCS.

35 Check If Low-Head SI Pumps Should Be Stopped:

a. Check RCS pressure:

1) Pressure - GREATER THAN 1) Go to E-1, LOSS OF REACTOR (4) PSIG OR SECONDARY COOLANT, Step 1.

2) Pressure - STABLE OR 2) Go to Step 36.

INCREASING

b. Stop low-head SI pumps and place in standby 63798:1/082983 11 of 14 ,

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL 36 Check If Diesel Generators Should Be Stopped:

a. Verify ac emergency busses - a. Try to restore offsite power ENERGIZED BY OFFSITE POWER to ac emergency busses. IF offsite power can NOT be restored,.THEN load the following equipment on ac i emergency busses:

[ Enter plant specific list]

b. Stop any unloaded diesel generator and place in standby 37 Return To Step 19

- END -

1 4

6379B:1/082983 12 of 14

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL FOOTNOTES (1) Enter plant specific containment pressure setpoint for spray actuation.

(2). Enter plant specific value for the shutoff head pressure of the high-head SI pumps, plus allowances for normal channel accuracy.

(3) Enter plant specific value for the shutoff head pressure of the high-head SI pumps, plus allowances for normal channel accuracy and post accident transmitter errors, not to exceed 2000 psig.

(4) Enter plant specific value for the shutoff head pressure of low-head SI pumps, plus allowances for normal channel accuracy.

(5) Enter plant specific value for the shutoff head pressure of the low-head SI pumps, plus allowances for normal channel accuracy and post accident transmitter errors.

(6) Enter the minimum safeguards AFW flow requirement for heat removal, plus allowances for normal channel accuracy (typically one MD AFW pump capacity at SG design pressure).

(7) Enter plant specific no-load temperature.

(8) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.

(9) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.

(10) Enter PRZR PORV pressure setpoint.

(11) Enter PRZR Spray pressure setpoint- .

-(12) Enter plant specific RCP trip parameter and setpoint, including allowances for normal channel accuracy. Refer to Generic Issues section of Executive Volume.

(13) Enter plant specific RCP trip parameter and setpoint, including allownaces for ncrmal channel accuracy and post accident transmitter errors. Refer to Generic Issues section of Executive Volume.

6379B:1/082983 13 of 14

SEP 011983 E-0 REACTOR TRIP OR SAFETY INJECTION HP-REV. 1 FINAL FOOTNOTES (Continued)

(14) Enter sum of temperature and pressure measurement system errors, including allowances for normal channel accuracies, translated into temperature using saturation tables.

(15) Enter plant specific value showing PRZR level just in range, including allowances for normal channel accuracy.

(16) Enter plant specific value corresponding to CST low level switchover setpoint in plant specific units.

'6379B:l/082983 14 of 14

SEABROOK TEST PROCEDURE FR-H.1 1

E-4 0040V/1:101183

C:de: Symptom /

Title:

Procedure No./

Revision No./Date:

FR-II .1 RESPONSE TO LOSS OF SECONDARY HEAT SINK OS-1353.1 Rev. 1-7 0 -7 / 10/11/83 A. PURPOSE This procedure provides actions to respond to a loss of secondary heat sink-in all steam generators.

B. SYMPTOMS OR ENTRY CONDITIONS This procedure is entered from:

1) E-0, REACTOR TRIP OR SAFETY INJECTION, Step 16, when minimum EFW flow is not verified.

2) F-0.3, HEAT SINK Critical Safety Function Status Tree on a RED condition.

l

! l i

i e

l of 16 v + -

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--w . . -. , , , . . , , -en < .,,, . , - , - ~ , -

C das Symptom /

Title:

Procedure No./

Revision No./Date:

F R-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK OS-1353.1 RIv. 1- T 0-7 / 10/11/R3 l_9 TSP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINEDl NOTE If ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, is in effect and total feed flow is less than 470 gpm due to operator action, then this procedure should not be performed.

CAUTION e If RCS temperature and pressure are increasing due to loss of secondary heat sink, RCPs should be tripped and Steps 10 through 16 should be imme-diately initiated for bleed and feed.

e Feed flow should not be reestablished to any faulted SG if a non-faulted SG is available.

1 Check If Secondary Heat Sink Is Required:

a. RCS pressure - GREATER THAN a. Go to E-1, LOSS OF REACTOR ANY NON-FAULTED SG PRESSURE OR SECONDARY COOLANT, Step 1.

b. RCS hot leg temperature - b. Try to place RHR System in GREATER THAN 350*F [320*F service yhile continuing FOR ADVERSE CONTAINMENT] in this guideline. Refer to OS-1013.03 and OS-1013.04, RHR TRAIN A and RHR TRAIN B STARTUP AND OPERATION.

Il[ adequate cooling with RHR System established, l

' THEN return to procedure and step in effect.

l l-I l

l l

2 of 16 l

l l

t

Csda Symptom /

Title:

Procedure No./

Revision No./Date:

FR-H.1 RESPONSE'To LOSS OF SECONDARY IIEAT SINK OS-1353.1 Rev.~1a7" 0*T / 10/11/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l Try To Establish EFW Flow To

~

2 At Least One SG:

! a. Check control room indi-cations for cause of EP4 fatture:

o CST level low e EFW pump power supply e EFW valve alignment e EFW pump failure

b. Try to restore EFW flow

c. Check total flow to c. Dispatch operator to SGs - GREATER THAN locally restore EFW flow.

470 GPM Check suction and discharge valve lineup. Go to Step 3.

d. Return to procedure and step in effect 3 Stop All RCPs 4 Check CCP Status - AT LEAST Go to Step 10 ONE AVAILABLE CAUTION If offsite power is lost after SI reset, manual action may be required to restart safeguards equipment.

4 3 of 16

II 4

1 C dat Symptom /

Title:

Procedura No./ l Revision No./Date: !

FR-il .1 RESPONSE TO LOSS OF SECONDARY llEAT "> INK OS-1353.1 R v.1-T 0-r/ 10/11/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINE0 l 5 Try To Establish SUFP Flow To At least One SC:

a. IF offsite power is NOT available, THEN power SUFP from Bus E5

h. Check CST inventory ade- h. Break condenser vacuum quate for SL/P operation on and take suction from lower suction. IF, adequate, condenser hotwell - CO-V145 THEN transfer to lower suc- OPEN.

tion - CD-V142 OPEN

c. Establish feed path c. Establish feed path through main feed header: through EF9 headers:

1) Reset FW isolation l 's Open FW-V163.

2) ' Place main feed regula- 2) Open FN-V156.

ting valve controllers in manual and close valves IF no feed is obtained, TH'EN go to Step 9.

3) Open FW isolation valves IF,no FW isolation valve can be opened, THEN go to Step 9

d. Start the SUFP and control d. Go to Step 7.

feed using the FW bypass e valves 4 of 16

'T Coda: Symptom /

Title:

Prcesdure Ns./

Revision No./Date:

FR-H.1 RESPONSE TO LDSS OF SECONDARY HEAT SINK OS-1353.1 Rev. 1-7 0-T / 10/11/83 l STEP l ~ \ ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 6 Check SG Levels:

a. WR level - ABOVE TOP OF a. IF feed flow to at least U-TUBES IN AT LEAST ONE Ee SG verified, THEN SG, 65% [95% FOR ADVERSE maintain flow to restore

! CONTAINMENT] SG level above U-tubes.

LEVEL ABOVC SG U-TUBES ADVERSE Q)NTM NORMAL CONTM NARROW RANGE WIDE RANGE LEVEL GREATER LEVEL GREATEE THAN 28% THAN 65%

4 IF feed flow NOT verified, TIIEN go to St@.

b. Return to procedure and 7 step in effect.

t l

I' l

l 5 of 16 d

a Cod 31 Sympter/Titist Procedure No./

Revision No./Date:

FR-H.1 RESPONSE TO UDSS OF SECONDARY HEAT SINK OS-1353.1 Re v. 1- T O- T / 10/11/83 1 STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l CAUTION Following block of automatic SI actuation, manual SI actuation may be required if conditions degrade.

7 Try To Establish Feed Flow From Condensate System:

a. Depressurize RCS to less than 530 PSIG:

1) Check letdown - IN 1) Use one PRZR PORV. IF SERVICE NOT, THEN use auxiliary spray. Go to Step 7 b.

2) Use auxiliary spray 2) Use one PRZR FORV.

b. Block SI signals:

o Low steamline pressure SI o Low PRZR pressure SI

c. Depressurize at least one SG to less than 530 PSIG:

1) Dump steam to condenser 1) Manually or locally dump at maximum rate steam using SG ASDVs. IF SG ASDVs NOT available, THEN go to Ttep 9.

d. Establish condensate flow:

1) Locally open FW-V103, SGFP bypass valve-

2) Close both SCFP discharge valves l

3) Close both SGFP recire valves f 4) Start at least one 4) Go to Step 9 condensate pump I 5) Close condensate header long- term recirc valve

6) Reset FW isolation 6) Establish flow through EFW header:

7) Open FW isolation valves I a) Open FW-V163

8) Control flow using FW bypass valves b) Open FW-V156 l

6 of 16 i

f

Cods: Symptor/

Title:

Precedura No./

Revision No./Date:

FR-H.1 RESPONSE 10 (DSS OF SECONDARY HEAT SINK OS-1353.1 Re v. 1-7 0-T/ 10/11/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 8 Check SG Levels:

a. WR level - ABOVE TOP a. IF feed flow to at least OF U-TUBES IN AT LEAST 'one SG verified, T!!EN ONE SG, 65% [95% FOR maintain flow to restore ADVEKSE CONTAINMENT] SG level above U-tubes.

LEVEL ABOVE SG U-TUBES ADVERSE (DNTM NORMAL 00NTM NARROW RANGE WIDE RANGE LEVEL GREATER LEVEL GREATEE THAN 28% THAN 65%

IF feed flow NOT verified, THEN go to Step 9.

b. - Return to procedure and step in effect 9 Check For Loss Of Secondary Heat Sink:

, a. SG WR level - LESS THAN a. Return to Step 1.

TOP OF U-TUBES WITH NO FEED FLOW AVAILABLE LEVEL ABOVE SG U-TUBES

• ADVERSE CONTM NORMAL 00NTM NARROW RANGE WIDE RANC,

i -LEVEL GREATER LEVEL GREATER THAN 28% THAN 65%

CAUTIct! Steps 10 through 16 must be performed quickly in order to establish RCS heat removal by RCS bleed and feed.

10 Actuate SI 7 of 16

C:das Syapton/Titto: Procedura No./

Revision 'lo./Da te:

FR-H.1 RESPO*ISE TO LOSS OF SECONDARY IIEAT SINK OS-1353.1 R:v. 1- 7 n-7 / :10/11/83 ,

l SiEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l c 11 Verify RCS Feed Path: Manually start pumps and align valves as necessary per

a. Check pur.ip status: Attachment A to establish feed path. 11r_ a feed path can NOT be established, THEN con-e CCPs - AT LEAST ONE tinue attempts to establish RUNNING feed flow. Return to Step 5.

- OR -

e SI pumps - AT LEAST ONE RUNNING ,

.h. Check valve alignment for operating pumps -

PROPER EMERGENCY ALIGN-MENT ON STATUS PANEL 12~ Reset SI 13 Reset Containment Isolation Phase A And Phase B 14 Establish RCS Bleed Path: ['

a. Verify power to PRZR PORV a. Restore' power to block i

block valves - AVAILABLE ~ valves.

h.. Verify PRZR PORV block b. Open block valves.

valves - ALL OPEN

c. Open all PRZR PORVs s

8 of 16

Cad 2: -Symptom /

Title:

Prcesdure No./

Revision No./Date; FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK OS-1353.1 Rev. 1-7 0 -7 / 10/11/83 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINEM 15 Verify Adequate RCS Bleed Path:

a. PRZR PORVs - AT LEAST TWO a. Perform the following:

OPEN

1) Depressurize at least one intact SG to atmos-pheric pressure using SG ASDV.

2) Align any available low pressure water source to the depressurized SC(s).

16 Maintain RCS Heat Removal:

e Maintain ECCS flow - CREATER THAN REQUIRED ON FICURE OS-1353.1-1 e Maintain PRZR PORVs - AT LEAST TWO OPEN CAUTION If RWST level reaches decreases to less than 23.5%, the ECCS system should he aligned for. cold leg recirculation using ES-1.3, TRANSFER TO COLD LEC' RECIRCULATION.

9 9 of 16

i l

l Cod] Symptom /

Title:

Procedure No./

Revision 'lo./Date:

i FR-M.1 RESPONSE TO LOSS OF SECONDARY IIEAT SINK OS-1153.1 R:v. 1-T 0-7 / 10/11/R1 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 17 Continue Attempts To Establish Secondary Meat Sink In At Least One SG:

o EFW flow e SUFP flow e Condensate flow IR Check For Adequate Secondary Heat Sinkt

a. SC WR level - ABOVE TOP a. Return to Sten 17.

OF U-TUBES IN AT LEAST ONE SG, 65% [95% FOR ADVERSE CONTAINMENT)

Check RCS Temperatures: Return to Step 17.

19 e Core exit TCs - DECREASING e RCS hot leR temperatures -

l l DECREASING l

l l

I

l. ,

l

]

i 10 of 16 l

i

Ctd2: Symptom /

Title:

Procsdure No./.

Revision No./Date: ,

FR-H.1 RESPONSE TO LOSS OF SECONDARY ftEAT SINK OS-1353.1 Rev. 1-T ~ 0 r / 10/11/83

__l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l NOTE e After stopping any ECCS pump, RCS pressure should be allowed to stahtitze before stopping another ECCS ptrup.

e The charging pumps and Si pumps should he stopped on alternate ECCS trains when possible.

20 Check If One CCP Should Be Stopped:

a. Two CCPs - RUNNING a. Go to Step 21.

b. Determine required RCS subcooling from table:

REQUIRED SUBC00 LING (*F)

W/RCPn WO/RCPs SI PUMP GTATUS NORMAL ADVERSE NORMAL ADVERSE CONTAINMENT CONTAINMENT CONTAINMEfff CONTAINMENT NONE RUNNING 121* 121* 135* 135' ONE RUNNING 68* 68* 75* 75*

1 TWO RUNNING 59' 59' 65* 65* >

c. RCS subcooling based on c. DO NOT STOP CC?. Go to core exit TCs - GREATER Step 23.

THAN REQUIRED SUBC00 LING

d. PRZR level - GREATER THAN d. DO NOT STOP CCP. Go to 5% [30% POR ADVERSE Step 23.

CONTAINMENT]

e. Stop one CCP 11 of 16

.. . . . . - _ -- . . . . . . - - ~ . ~ _ . _ . _ _ . - - . - . .

Cadet Sym7 tom / Tit 101 Procedure No./

Revision No./Datet FR-H . l _ - RESPONSE TO LOSS OF SECONDARY ftEAT SINK OS-1151.1 R:v. 1-T 0-T/ 10/11/81 i

l STEP l l ACTION / EXPECTED RESPONSE l l RESPnNSE NOT OBTAINED l P

21 Check If one SI Pump Should Be Stoppedt

a. Any SI pump - RUNNING a. Go to Step 22.

b. Determine required RCS subcooling from tablet w/RCP UO/RCP 1 SI PUMP ONE CCP 'NO CCP DNE CCP NO CCP STATUS NORMAL l ADVERSE NORMAL l ADVERSE NORMAL l ADVERSE NORMAL 1 ADVERSE j CONTAINMENT CONTAINMENT CONTAINMENT CONTAINMENT a

ONE RUNNING 236* 216' - -- 216* 216* -- -

190 RUNNING 70* 70* 104* 104* 77* 77* 117* 117*

?

c. RCS subcooling based on c. DO NOT STOP SI PtIMP. Go core exit TCs - GREATER to Step 23.

THAN REQUIRED S!!BC00 LING

d. PR;R level - GREATER THAN d. DO NOT STOP S1 PUMP. Go 5

, 5% [30% FOR ADVERSE to Step 23.

CONTAINMENT]

e. Stop one additional S!

pump f.- Return to Step 21a 12 of 16

_ _ _ _ _ . _ . . . . _ - . . ,_ - , _ _ _ . _ . _ _ _ , . _ _ , _ _ . _ . _ _ , _ _, _ _ = _ _ _ _ .

c - - ..

Code: Symptom /

Title:

Procedure No./

Revision No./Date:

FR-H.1 RESPONSE TO LOSS OF SECONDARY MEAT SINK OS-1153.1 Rev. 1.T 07/ 10/11/g3 l STEP l l ACTION / EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l 3

22 Check If Normal Charging Should Be Established:

4

a. Check the following: a. DO NOT ISOLATE BIT. Return to Step 20.

e . SI pumps - STOPPED e CCPs - ALL BUT ONE STOPPED

h. RCS subcooling based on b. DO NOT ISOLATE BIT. Go core exit TCs - CREATER to Step 23.

TilAN 100*F

c. PRZR level - CREATER titan c. DO NOT ISOLATE BIT. Go 5% [30% FOR ADVERSE to Step 23.

CONTAINNENT)

d. GotoStep2fb NOTE After closing a PRZR PORV, it may be necessary to wait for RCS pressure to inricase to permit stopping SI pumps in Steps 20 and 21.

21 Check PRZR PORV Status

a. PRZR PORVs - ANY OPEN a. Go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Sten 1.

! b. Close one PRZR PORV h. Close associated PORV block valve. IF block valve can NOT E closed, THEN go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

c. Return to Step 20 ,

i 11 of 16

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

Codes Symptom / Title Procedure No./

Revision No./Date:

FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK OS-1353.1

, Rev. 1-T 0 -7 / 10/11/81 l STBF l l ACTION / EXPECTED RESPONSE l -l RESPONSE NOT ORTAI'IEDl 24 Establish 60 CPM Charging

' Flow s.

Open charging line

• isolation valves

b. Establish 60 GPM charging flow using flow control valve

c. Ad just seal injection flow as necessary using HCV-lR2 25 Isolate RIT:

l. a. Close inlet isolation valves I h. Close outlet isolation valves

c. Check RIT bypass valves-l closed

-26 Check PRZR PORVs - ALL CLOSED Close PRZR PORVn. I,F,any valve can NOT he closed, THEN manually close its block valve.

4 27 Control Charging Flow To ,

Ma*ntain PRZR Level

~ 28 Go.To ES-1.1, SI TERMINATION, Step 11

- END -

1-14 of 16 4

i

-+-rwrw-, -e-%,-,v+g,- ,-v,-, ,c,,.,-,

I i

C: dst Sympton/Titlet Procedure No./

Revision No./Date:

a FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK OS-1353.1 Rev. 1 -7 0-T / 10/11/83 i

i l

ATTACHMENT A i

ECCS INJECTION ALICNMENT e CCPs VIA BIT TO RCS COLD LEGS VALVE NAME POSITION

. CS-LCV-112D RWST GUTLET OPEN CS-LCV-112E RWST GUTLET OPEN CS-v142 CHARCING ISOL. CLOSED CS-V143 CHAPCINC ISOL. CLOSED CS-LCV-112B CVCT OUTLET CLOSE9 CS-LCV-112C CVCT OUTLET CLOSFD CS-V844 BIT INLET OPEN CS-V65 BIT INLET OPEN CS-V845 BIT INLET OPEN CS-V66 BIT INLET OPEN CS-V846' BIT BYPASS CLOSED i

CS-VR47 BIT BYPASS CLOSED CS-V165 BIT RECIRC. PUMP DISCH. CLOSED

)

CS-V173 BIT RECIRC. ISOL. CLOSED CS-V174 BIT RECORC. ISOL. CLOSED SI-V138 BIT OUTLET TO RCS OPEN

.SI-V139 VIT GUTLET TO RCS OPEN t

15 of 16 l

4 a-, , - ,.- ,

Cod 31 Sympter/Titlos Procsduro Ns./

Revision No./Date:

FR-H.1 RESPONSE TO WSS OF SECONDARY HEAT SINK 0S-1353.1 Re v . 1 - T 0-T / 10/11/83 )

ATTACHMENT A (CONTINUED) e SI PUMPS TO RCS (DLD LECS VALVE NAME POSITION CES-V47 SI-P-6A SUCTION FROM RWST OPEN CBS-V49 SI-P-6A SUCTION FROM RWST OPEN SI-V 90 SI-P-6A MIN FWW TO RWST OPEN SI-V102 SI 10 HOT LEGS CWSED SI-V112 SI 10 COLD LEGS OPEN SI.-V114 SI TO (DLD LEGS O PEN CS-V460 SI-P-6A SUCTION CROSSOVER CW SED CS-V461 SI-P-6A SUCTION CROSSOVER CWSED CS-V475 SI-P-6A SUCTION CROSSOVER OPEN CBS-VS1 SI-P-6B SUCTION FROM RWST OPEN CBS-V53 SI-P-6B SUCTION FROM RWST OPEN SI-V89 SI-P-6B MIN FWW TO RWST OPEN SI-V93 SI-P-6A & B MIN FIDW TO RWST OPEN SI-Vill SI TO COLD LEGS O PEN SI-V77 SI 10 HOT LEGS CIDSED 16 of 16

COOE: SYMPTOD/ TITLE: '

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FR-H.1 RESPONSE TO LOSS OF * *' " " ' ** !

Os-1353.1 l REV.1 -7 SECONDARY HEAT SINK 007/09/22/83 l t

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FIGURE OS-1353.1-1 Required ECCS Flow for Core Bleed and Feed Cooling to Remove Decay Heat 700 E \

.$' N 3 500 o \s -

400 \

m x '

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300 J 200 \ ,

h  % '

N i

O p 100 ,-- .,

HR.- .I i 10 10 0 1000 neN. .6 60 600 6pOO 60p00 TIME AFTER SHUTDOWN NOTES: RCPs NOT RUNING ASSUMED 88*F ECCS EOL CONDITIONS

WOG EMERGENCY RESPONSE GUIDELINE FR-H.1 E-5 0040V/1:101183

SEP 01 I:0 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL A. PURPOSE This guideline provides actions to respond to a loss of secondary heat sink in all steam generators.

B. SYMPTOMS OR ENTRY CONDITIONS This guideline is entered from:

1) E-0, REACTOR TRIP OR SAFETY INJECTION, Step 16, when minimum AFW flow is not verified.

2) F-0.3, HEAT SINK Critical Safety Function Status Tree on a RED condition.

65858:1/082983 1 of 13

SEP 01 E:3 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL NOTE If ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, is in effect and total feed flow is less than (1) gpm due to operator action, this guideline should not be performed.

CAUTION o If parameter (2) [(3) FOR ADVERSE CONTAINMENT] is exceeded due to loss of secondary heat sink, RCPs should be tripped and Steps 10 through 16 should be immediately initiated for bleed and feed.

o Feed flow should not be reestablished to any faulted SG if a non-faulted SG is available.

1 Check If Secondary Heat Sink Is Required:

a. RCS pressure - GREATER THAN a. Go To E-1, LOSS OF REACTOR ANY NON FAULTED SG PRESSURE OR SECONDARY COOLANT, Step 1.

b. RCS temperature - b. Try to place RHR System in GREATER THAN (4)*F [(5)*F service while continuing FOR ADVERSE CONTAINMENT] in this guideline. Refer to

[ Enter plant specific procedure number and title].

IF adequate cooling with RHR System established, THEN return to guideline and step in effect.

i-I l

6585B:1/082983 2 of 13

SEP 0 1 N6)

FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 2 Try To Establish AFW Flow To At least One SG:

a. Check control room indications for cause of AFW failure:

o CST level o AFW pump power supply o AFW valve alignment

b. Try to restore AFW flow

c. Check total flow to c. Dispatch operator to locally SGs - GREATER restore AFW flow. Go to THAN (1) GPM Step 3.

d. Return to guideline and step in effect 3 Stop All RCPs 4 Check Charging /SI Pump Go to Step 10.

Status - AT LEAST ONE AVAILABLE CAUTION If offsite power is lost af ter SI reset, manual action may be required to restart safeguards equipment.

5 Try To Establish Main FW Flow To At least One SG:

a. Check Condensate System - a. Try to place Condensata IN SERVICE System in service. -IF NOT, THEN go to Step 9.

b. Check FW isolation valves - b. Perform the following:

OPEN

1) Reset SI if necessary.

2) Reset FW isolation.

3) Open FW isolation valves.

IF no FW isolation valve can be opened. THEN go to Step 9.

c. Establish Main FW flow: c. Go to Step 7.

[ Enter plant specific means]

6585B:1/082983 3 of 13

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SEP 0 1 lle)

FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 6 Check SG Levels:

a. Narrow range level in at a. -IF feed flow to at least least one SG - GREATER one SG verified, THEN THAN (6)% [(7)% FOR maintain flow to restore ADVERSECONTAINMENT] narrow range level to greater than (6)% [(7)% FOR ADVERSECONTAINMENT]. IF NOT varified, THEN go to Step 7.

b. Return.to guideline and step in effect CAUTION Following block of automatic SI actuation, manual SI actuation may be required if conditions degrade.

7 Try To Establish Feed Flow From Condensate System:

a. Depressurize RCS to less than (8) PSIG:

1) Check. letdown - IN SERVICE 1) Use one PRZR PORV. IF NOT, THEN use auxiliary spray. Go to Step 7b.

2) Use auxiliary spray 2) Use one PRZR PORV.

b. Block SI signals:

o Low Steam 11ne Pressure SI i

o Low PRZR l Pressure SI l c. Depressurize at least one l SG to less than (9) PSIG:

l l 1) Dump steam to condenser 1) Manually or locally dump at maximum rate steam from SGs:

o Use PORV.

-OR-l o [ Enter plant specific means].

l IF NOT, THEN go to Step 9.

I

d. Establish condensate flow: d. Go to Step 9.

[ Enter plant specific means]

l 65858:1,/082983 4 of 13 1

m

SEP 01-1983 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 8 Check SG Levels:

a. Narrow range level in at a. IF feed flow to at least least one SG - GREATER ine SG verified, THEN THAN (6)% [(7)% FOR maintain flow to restore ADVERSE CONTAINMENT] narrow range level to greater than (6)% [(7)% FOR ADVERSE CONTAINMENT]. IF E verified, THEN go to N

Step 9.

b. Return to guideline and step in effect 9 Check For Loss Of Secondary Heat Sink:

a. Parameter (2) [(3) FOR a. Return to Step 1.

ADVERSE CONTAINMENT] -

EXCEEDED CAUTION Steps 10 through 16 must be performed quickly in order to establish RCS heat removal by RCS bleed and feed.

10 Actuate SI 11 Verify RCS Feed Path: Manually start pumps and align valves as necessary to establish

a. Check pump status: feed path. IF a feed path can NOT be established, THEN continue o Charging /SI pumps - attempts to establish AFW flow.

AT LEAST RUNNING Return to Step 5.

-OR-o High-head SI pumps -

AT LEAST ONE RUNNING

b. Check valve alignment for operating pumps - PROPER EMERGENCY ALIGNMENT 12 Reset SI 13 Reset Containment Isolation Phase A And Phase B 6585B:1/082983 5 of 13

SEP 011963 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 14 . Establish Instrument Air To Start one air compressor Containment and establish instrument air to containment.

15 Establish RCS Bleed Path:

a. Verify power to PRZR PORV a. Restore power to block bicek valves - AVAILABLE valves.

b. Verify PRZR PORV block b. Open block valves.

valves - ALL OPEN

c. Open all PRZR PORVs 16 Verify Adequate RCS Bleed Path:

a. PRZR PORVs - AT LEAST TWO a. Perform the following:

OPEN

1) Open all RCS high point vents:

[ Enter plant specific list]

2) Depressurize at least one intact SG to atmospheric pressure using SG PORV.

3) Align any available low pressure water source to the depressurized SG(s).

17. Maintain RCS Heat Removal:

o Maintain SI flow o Maintain PRZR PORVs -

AT LEAST TWO OPEN 6585B:1/082983 6 of 13

SEP 0 i 1963 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL CAUTION If RWST level decreases to less than (10), the SI system should be aligned for cold leg recirculation using ES-1.3, TRANSFER TO COLD LEG RECIRCULATION.

18 Continue Attempts To Establish Secondary Heat Sink In At-Least One SG:

o AFW flow '

o Main FW flow o Condensate flow o Other low pressure flow 19 Check For Adequate Secondary Heat Sink:

a. Narrow range level in at a. Return to Step 18.

least one SG - GREATER THAN-(6)% [(7)% FOR ADVERSE CONTAINMENT]

20 Check RCS Temperatures: Return to Step 18.

o Core e it TCs - DECREASING o kCS hot leg temperatures -

DECREASING 6585B:1/082983 7 of 13

1 l

l SEP 011983 1

-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL NOTE o After stopping any SI pump, RCS pressure should be allowed to stabilize before stopping another SI pump.

o The charging /SI pumps and high-head SI pumps should be stopped on alternate ECCS trains when possible.

21 Check If One Charging /SI Pump Should Be Stopped:

a. Two charging /SI a. Go to S*tep 22.

pumps - RUNNING

b. Determine required RCS subcooling from table:

HIGH-HEAD RCS SUBC00 LING ('F)

SI PUMP DNE CHARGING /SI PUMP NO CHAliGING/SI PUMP STATUS RUNNING RUNNING NONE (11) F (11)*F RUNNING ONE (11)op (ii).p RUNNING RUNNING

c. RC5 subcooling based on c. DO NOT STOP CHARGING /SI core exit TCs - GREATER PUMP. Go to Step 24.

THAN REQUIRED SUBC00 LING

d. PRZR level - GREATER THAN d. DO NOT STOP CHARGING /SI (12)% [(13)% FOR ADVERSE PUMP. Go to Step 24.

CONTAINMENT]

e. Stop one charging /SI pump 6585B:1/082983 8 of 13

F SEP 011943 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 22 Check If One High-Head SI Pump Should Be Stopped:

a. Any high-head SI pump - a. Go to Step 23.

RUNNING-

b. Determine required RCS subcooling from table:

HIGH-HEAD RCS SUBC00 LING (*F)

PUMP ONE CHARGING /SI PUMP RUNNING NO CHARGING /SI PUMP RUNNING STATUS ANY RCP RUNNING N6 RCP RUNNING ANY RCP RUNNING ~~ NO RCP RUSNING ONE (11)*F (11)*F (11)*F (11)*F RUNNING TWO (11)*F (11)*F (11)*F (11) F RUNNING

c. RCS subcooling based on c. DO NOT STOP SI PUMP. Go to '

core exit TCs - GREATER Step 24.

THAN REQUIRED SUBC00 LING

d. PRZR level - GREATER THAN d. DO NOT STOP SI PUMP. Go (12)% [(13)% FOR ADVERSE to Step 24.

CONTAINMENT]

e. Stop one additional high-head SI pump

f. Return to Step 22a 6585B:1/082983 9 of 13

SEP 0 1 O FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 23 Check If Normal Charging Should Be Established:

a. Check the following: a. DO NOT ISOLATE BIT. Return to Step 21.

o High-head SI pumps -

STOPPED o Charging /SI pumps'-

ALL BUT ONE STOPPED

b. RCS subcooling based on b. DO NOT ISOLATE BIT. Go to core exit TCs - GREATER Step 24.

THAN (11)'F

c. PRZR level - GREATER THAN c. DO NOT ISOLATE BIT. Go to (12)% [(13)% FOR ADVERSE Step 24.

CONTAINMENT]

NOTE After closing a PRZR PORV, it may be necessary to wait for RCS pressure to increase to permit stopping SI pumps in Steps 21 and 22.

24 Check PRZR PORV Status:

a. PRZR PORVs - ANY OPEN a. Go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

b. Close one PRZR PORV b. Close associated PORV block valve. IF block valve can NOT be cTosed, THEN go to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

c. Return to Step 21 658SB:1/082983 10 of 13 ,

=. l

SEP 011983 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL 25 Establish (14) GPM Charging Flow:

a. Close charging line flow control valve

b. Open charging line hand control valve

c. Open charging line isolation valves

d. Establish (14) GPM charging flow using flow control valve 26 Isolate BIT:

a. Close inlet isolation valves

b. Close outlet isolation valves 27 Check PRZR PORVs - Close PRZR PORVs. IF any ALL CLOSED valve can NOT be closed, THEN manually close its block valve.

28 Control Charging Flow To Maintain PRZR Level.

29 Go To ES-1.1, SI TERMINATION, Step 11

- END -

t 6585B:1/082983 11 of 13 l

SEP 0 i 1963

-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL FOOTNOTES (1) Enter the minimum safeguards AFW flow requirement for heat removal, plus allowances for normal channel accuracy (typically one MD AFW pump capacity at SG design pressure.)

\

(2) Enter. plant specific paramenter and setpoint for diagnosing loss of secondary heat sink, including allowances for normal channel accuracy. Refer to background document.

(3) Enter plant specific parameter and setpoint for diagnosing loss of secondary heat sink, including allowances for normal channel accuracy and post accident transmitter errors.

(4) Enter plant specific temperature requirement, including allowances for normal channel accuracy, for placing RHR in service.

(5) Enter plant specific temperature requirement, including allowances for normal channel accuracy and post accident transmitter errors, for placing RHR in service.

(6) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.

(7) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy, post-accident transmitter errors, and reference leg process errors, not to exceed -

50%.

(8) Enter plant specific RCS pressure 50 psi below permissiv: to block SI.

(9) Enter plant specific shutoff head pressure of ccndensate pumps.

(10) Enter plant specific value corresponding to RWST switchover alarm in plant specific units.

6585B:1/082983 12 of 13

SEP 0 1 E::3 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK HP-REV. 1 FINAL FOOTNOTES (Continued)

(11) Enter plant specific subcooling criteria. Refer to "SI Reduction Sequence Evaluation" section of Executive Volume.

(12) Enter plant specific value showing PRZR level just in range, including allowances for. normal channel accuracy.

(13) Enter plant specific value showing PRZR level just in range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.

(14) Enter a charging flow rate comparable to normal charging /SI pump miniflow, i.e., 60 gpm.

6585B:1/082983 13 of 13

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Appendix F Seabrook Plant Description F-1 0040V/1:101183

EMERGENCY RESPONSE GUIDELINE SEABROOK PLANT DESCRIPTION TABLE OF CONTENTS SECTION PAGE

1. INTRODUCTION 1

2. PLANT SYSTEMS 2-17

3. ATTACHMENT A - CRITICAL SAFETY FUNCTION INSTRUMENTATION REQUIREMENTS

4. APPENDIX B - INSTRUMENTATION AND CONTROL REQUIREMENTS F-2 0040V/1:101183

1 Control and Protection Actuation Systems 2.1 Reactor Trip Actuation System 2.2 ESF Actuation System Instrumentation Systems 2.3 Nuclear Instrumentation System

' 2 . '4 Control Rod Instrumentation System 2.5 Radiation Monitoring System

.6 2 Containment Instrumentation System

' 2. 7 Critical Safety Function Monitoring System

- Process Control Systems 2.8 Reactor Coolant System 2.9 Emergency Core Cooling System 2.10 Residual Heat Removal System 2.11 Chemical and Volume Control System 2.12 Component Cooling Water System (Primary) 2.13 Service. Water System 2.14 Containment Spray System 2.15 Containment Atmosphere Control System 2.16 Main Steam System 2.17 Main Feedwater anc Condensate System

. 2.18 Emergency Feedwater System

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2.19 Steam Generator Blowdown System 2.20- Primary Sampling System and Post Accident Sampling 2.21 Spent Fuel Storage and Cooling System 2.22 Control Rod Drive Mechanism Cooling System 2.23- Control Rod Control System 2.24 Turbine Control System Support Systems

-2.25 Electrical Power System 2.26 Pneumatic Power System i

1 F-3 0040V/1:101183

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2.1 Reactor Trip Actuation System The reactor trip actuation system monitors specified process parameters and equipment status and actuates reactor trip if conditions exceed spe-cified limits. The reactor trip actuation system includes automatic actuations that occur concurrent with actuation of reactor trip.

The process parameters and equipment status monitored by the reactor trip actuation system are plant-specific. A reactor trip is automatically actuated if any condition exceeds its specified limit. Concurrent with actuation of reactor trip, a P-4 signal is generated and provides the following: .

o Turbine trip o Input signal for feedwater isolation o Input signal for SI block logic (concurrent with manual SI reset signal) o Steam dump control logic 2.2 Engineered Safeguards Features Actuation System The engineered safety features (ESF) actuation system monitors specified process parameters and actuates engineered safety features operation if conditions _ exceed specified limits. The ESF Actuation System consists of _

the following automatic actuation signals:

Safety Injection Signal The safety injection (SI) signal is the primary ESF actuation signal. It is automatically actuated on any of the following:

o Low pressurizer pressure o Low steamline pressure o High-1 containment pressure o Manual operator actuation F-4 0040V/1:101183

The following is automatically actuated by an SI signal:

o Reactor Trip o Feedwater isolation (Closure of feedwater isolation, flow control and bypass valves.)

o Emergency feedwater pump start (Start motor - driven and turbine-driven EFW '

- Pu'mp s) o Diesel generator start o Safety. injection system start j o Containment isolattor Phase A ("T" signal) o Containment ventilation isolation l

Containment Spray Signal '

The containment spray signal automatically actuates on any of the following:

o High - 3 containment pressure o Manual operator actuation The following plant equipment automatically actuates on a containment spray

-signal.

. - 3 o Containment spray pumps start-o Containment isolation phase B' valves close o' Containment spray system valves reposition for spray and chemical addition

.The containment spray signal actuation logic include's the following reset capability. \

go Manual reset.for containment spray system actuation signal

( l io Manual reset capability for containn.entiisolation phase B valve closure O f '

signal- s 4x f ., t 1

g, i

F-5 0040V/1:101183

Emergency Feedwater System The Emergency Feedwater System (EFW) consists of two pumps, one motor-driven and one turbine-driven. Either pump is full capacity and feeds all four steam generators. The system design provides for automatic isolation of a single line asociated with a faulted steam generator upon sensing high flow.

-Both EFW pumps automatically start on any of the following conditions:

o SI signal o Loss of power signal o' Low-Low icvel in any steam generator o Manual operator actuation Any EFW start signal also actuates isolation of steam generator blowdown.

-The EFW pumps are-not' used for plant startup and normal plant shutdowns. To accomplish.this task, a single Startup Feedwater Pump is provided. This pump receives an automatic start signal an loss of both main feedwater pumps only.

The Startup Feedwater Pump is also used as a backup to the EFW system. It can provide *feedwater to all steam generators through the main feedwater 1!nes or through the EFW lines.

Containment-Isolation Phase A Signal ("T" Signal) 7

'The containment isolation Phase A signal ' automatically isolates tion-essential containment penetrations to prevent or minimize the release of the radioactive material outside containment This signal automatically actuates on any of the following: 1 o SI signal s o Manual operator actuation of Phase A signal The containment isolction Phase A signal closes most valves in plant syste:ns that penetrate containment.e'The actuation logic includes separate reset" ,

, y capability for each actyation signal. , .

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i F-6 / .

0040V/1:101183 '

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Containment Isolation Phase B S'e al ("P" Signal)

The containment isolation phase B signal automatically isolates remaining containment penetrations to prevent the release of radioactive material outside containment. -The signal automatically actuates on any of the following:

o High - 3 containment pressure o Manual operator actuation of containment spray signal Y

The containment isolation Phase B' signal automatically closes the containment isolation valves in the component cooling water lines to the reactor coolant pumps. The actuation logic includes separate reset capability for each actuation signal.

Main Steamline Isolation Main steamline isolation is actuated on any of the following:

o HI - 2 containment pressure (same setpoint as HI-1) o Low steamline pressure in any steamline when pressuri7er pressure above P-11 o High-steam pressure rate in an:. main steamline when pressurizer pressure below P-11 o' Manual operator actuation The main steamline isolation logic includes the following reset capability; o Manual reset / block for low steamline pressure actuation signal (concurrent with a P-11 low pressurizer pressure permissive) o Automatic block for high steam pressure rate actuation signal when pressurizer pressure above P-11 se+. paint o Manual reset for main steamline isolation signal The following plant equipment automatically positions due to a main steamline isolation signal:

o- Main steamline isolation valves close o Main steamline isolation bypass valves close 1

F-7 0040V/1:101183-

Containment Ventilation Isolation.

The containment ventilation isolation signal automatically isolates containment ventilation penetrations to prevent the release of radioactive material outside contaiment. This signal automatically actuates on any of the following:

-.o SI_ signal o High containment radiation

'The containment ventilation isolation signal closes valves in the ventilation system. -The actuation logic includes separate reset capability for each actuation signal.

-Main Feedwater Isolation Signal The main feedwater isolation signal automatically isolates the main feedwater lines to prevent excessive cooling of the RCS and filling of the steam generators. The signal automatica1'ly actuates on any of the following input signals.

o SI' signal o High - High level (P-14 signal) in any steam generator

.o Reactor trip signal (P-4 coincidnet wil a low reactor coolant system T avg

. signal The feedwater isolation signal closes the following valves:

o Main feedwater isolation valves o' Main feedwater flow control valves

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o- Main feedwater bypass valves

'The main Feedwater isolation signal includes the following reset capability.

o Separate reset capability for the reactor trip signal coincident with RCS low T avg signal o Shared reset capability for the SI actuation signal. This capability requires reset of reactor trip signal (P-4) and reset of SI signal. Also SG high-high level condition cannot exist.

F-8 0040V/1:101183

2.3 _ Nuclear Instrumentation System The nuclear instrumentation system (NIS) monitors the reactivity state of the reactor core. It consists of instrumentation that monitors neutron flux and startup rate. Neutron flux is monitored over the source, intermediate and power range. Startup rate is monitored over the source and intermediate

~

range. The NIS includes a neutron flux recorder that can be switched to record different ranges. The source range neutron flux detectors automatic-ally re-energize when flux decreases below the source range high flux trip (P-6) setpoint following a reactor-trip, permitting the neutron flux recorder to be manually transferred to the source range scale. For post-accident harsh environment conditions, a qualified NIS consisting of wide-range neutron flux, flux rate and shutdown margin monitor is used.

2.4 Control Rod Instrumentation System The control rod instrumentation system monitors the position cf the reactor core control rods. It consists of control rod position and bottom light instrumentation.

2.5 Radiation Monitoring System

The radiation instrumentation system monitors the radiation levels in specified process systems and specified areas internal and external to the containment.

2.6 Containment Instrumentation System The containment instrumentation system monitors the environmental condition of the contaiment. It consists of containmerit pressure and temperature instrumentation, containment building level instrumentation and position indication for containment ventilation valves.

I F-9 0040V/1:101183

2.7 ' Critical Safety Function Monitoring System The. critical safety function monitoring system consists of instrumentation used to monitor plant variables. associated with the following critical safety function:

o Subcriticality o Core cooling o Heat sink o Reactor vessel integrity o Contaiment o RCS inventory This instrumentation system consists of post accident qualified instruments which display those variables necessary to monitor the critical safety

-functions. All variables are displayed on hard-wired indicators and/or recorders in the main control room. These same variables also input to the plant. computer system which provides automatic monitoring of the status of each critical safety function.

2.8 Reactor Coolant System The? reactor system (RCS) transfers heat from the reactor core to the main steam system or residual heat removal system and provides a barrier against the release of reactor coolant or radioactive-material to the containment enviroment.

The RCS consists of four identical heat transfer loops (connected in parallel to the reactor vessel) a pressurizer and a pressurizer relief tank. Flow from the RCS hot leg and cold leg eriter the bypass loop and returns via a common bypass header to the RCP suction. Each byp_as loop contains a hot leg and cold leg manifold that includes RTD temperature sensors used in plant control and protection. RTDs are also. installed in wells in each loop's hot and cold legs to provide accurate temperature measurement and indication while in natural circulation or RHR modet.

F-10 0040V/1:101183

The pressurizer is connected to the hot leg of one loop via the pressurizer surge line and the cold legs of two loops via the pressurizer spray lines..

The pressurizer has two power operated relief valves (with associated block valves), three code safety valves and heaters. RCS pressure is controlled by use of-the pressurizer where water and steam are maintained in equilibrium through use of the heaters, water spray and steam release.

~

The pressurizer PORVs and safety valves discharge to the pressurizer relief and (PRf) where steam discharge is condensed and cooled by mixing with water.

The PRT system has an external circulating pt.np and heat exchanger to assist in post-discharge cooldown.

The pressurizer PORVs are electrically operated and do not require pneumatic power sources. Each PORV is powered from independent, train associated power supplies.

A low temperature-overpressurization protection (LTOP) system is provided that operates the PORV's. Below an RCS temperature of 305 F, the PORV pressure setpoint is automatically decreased as a function of temperature.

Reactor coolant pumps (RCPs) are powered from 13.8 KV buses with two RCPs on each bus. An RCP that drives pressurizer spray is assigned to each bus.

.Non-condensible gas is vented from the RCS via a single train vessel head vent or the train associated pressurizer PORVs.

NtHES: Pressurizer PORVs and SG PORVs do not use air for operation.

Letdown and charging line valves are pneumatic valves but fail in safest position, i.e.;

o Letdown line isolation valves fail closed o Charging line isolation valves fail open o Excess letdown isolation valves fail closed o Auxiliary pressurizer spray valve fails closed o RCP #1 seal leakoff isolation valves fail open F-11 0040V/1:101183

Pneumatic power would be required to re-establish operability.

Air operated valves (A0Vs) inside containment are powered pneumatically from the containment air compressors. A0Vs outside containment are powered from the plant air compressors.

2.9 Emergency 2 ore Cooling System (ECCS)

The ECCS system provides coolant to the reactor coolant system and introduces.

negative. reactivity for events that require engineered safety features operation.

The-ECCS is designed to operate in three modes depending on plant transient:

o Cold leg injection mode (short term core cooling mode The Cold leg injection mode is defined as that period during which borated water is delivered from the refueling water storage tank (RWST) and accumulators to the RCS cold legs.

o Hot leg recirculation mode is defined as that period during which borated water is recirculated from the containment sump to the RCS cold legs.

The Hot leg recirculation mode is that period during which borated water is recirculated from the containment sump to both the RCS hot legs and RCS cold legs.

The ECCS consists of four major subsystems:

o Centrifugal Charging Pump Subsystem The charging subsystem consists of two centrifugal charging pumps and a boron injection tank. These pumps are part of the chemical and volume control system and provide charging and RCP seal injection flow during normal operation. Upon. receipt of an SI signal, these pumps are automatic-ally isolated from the normal charging function and aligned in the cold leg injection mode. In this mode the charging pumps take suction from the RWST and discharge through the BIT and RCP seals to all RCS cold legs. During recirculation modes, the charging pump's discharge remains through the BIT path to the RCS cold legs.

F-12 0040V/1:101183

I

.The discharge shutoff pressure of the charging pumps is-greater than RCS normal operating pressure.

The BIT contains 12 weight percent boric acid solution. During normal operation the BIT contents are isolated by parallel inlet and outlets sets of motor operated valves. The BIT is bypassed in the recirculaton mode to reduce radiation levels in the PAB.

o SI Subsystem The SI subsystem consists of two centrifugal SI pumps. These pumps are part.of the SI system and are aligned in the SI cold leg injection mode alignment. Upon receipt of an SI signal, the SI-pumps automatically start in the SI cold leg injection mode. In this mode the SI pumps take suction from the RWST and discharge to all RCS cold legs (through the accumulator discharge lines). During recirculation modes the SI pumps are aligned to take suction from the RHR pump discharge and discharge to all RCS cold legs or hot legs depending on recirculation mode.

The discharge shutoff pressure of the SI pump is approximately 1600 PSIG.

o- RHR Subsystem The RHR subsystem consists of two centrifugal pumps and two heat

,exchangers. These pumps and heat exchangers are part of the RHR system and provide normal plant shutdown heat removal. During plant operation, these pumps are aligned in the cold leg injection mode alignment. Upon receipt of an SI signal, the RHR pumps automatically start in the SI cold leg injection mode alignment. In this mode the RHR pumps take suction from RWST and discharge 'to all RCS cold legs (through the accumulator discharge lines). During recirculation modes, the RHR pumps are aligned to take suction from the containment recirculation sump and to discharge to the suction of the centrifugal charging pumps and SI pumps as well as to all RCS cold legs or two RCS hot legs depending on recirculation mode.

The discharge shutoff pressure of the RHR pumps is approximately 200 PSIG.

F-13 0040V/1:101183

The RHR heat exchangers are supplied with primary component, cooling water

-(PCCW) during ' recirculation modes.

o _ Accumulator Subsystem The accumulator subsystem consists of four accumulator tanks, each connecting to one RCS cold leg via an accumulator injection line. Each tank contains borated water and-is pressurized to a nominal 650 PSIG with a nitrogen cover gas. A single isolation valve is provided in each accumulator injection line and series vent valves are provided to vent the accumulators if necessary. During plant operation the injection isolation valves are open with power removed from the valve operators. The accumu-

,lators are available to deliver their contents to the RCS cold legs during the injection mode of any emergency transient that is accompanied by RCS depressurization below the accumulator pressure, i

2.10 -Residual Heat Removal System The residual heat removal (RHR) system removes residual heat from the reactor coolant system during plant shutdown operations at low reactor coolant system pressures.

The RHR system consists of two RHR pumps and RHR heat exchangers. The RHR system provides normal shutdown heat removal when RCS pressure and temperature are reduced to approximately 400 psig and 350 F. During normal shutdown heat removal operations, the IHR pump suction is aligned to the RCS hot legs and the RHR pump discharge is aligned to the RCS cold legs.

Portions of the RHR system also function as part of the ECCS. This shared l function-is described in Section 2.8.

i 2.11 Chemical and Volume Control System l

! The chemical and volume control system (CVCS) provides coolant to the reactor coolant system and provides core reactivity control for normal operation and l any event that does not require engineered safety features _peration.

I

F-14 l 0040V/1

101183

g l

l The CVCS consists of charging and letdown capability for control of RCS

.i nventory. . Letdown capability is provided by two letdown paths (letdown line and excess. letdown line). Charging capability is provided by three charging

. pumps (two centrifugal pumps that also function as ECCS pumps charging line and RCP seal, injection lines. The RCP seal injection lines supply _each RCP and provide RCP seal cooling.

Suction flow to the charging pumps is provided by the chemical volume control tank (CVCT) or by the refueling water storage tank (RWST). The charging pumps suction is normally aligned to the CVCT, but is automatically transferred to the RWST on the following:

o SI signal o CVCT low-low level signal The CVCS includes two redundant boric acid tanks and associated boric acid pumps. Each boric acid _ tank contains a four weight percent boric acid solution. This solution is fed to the suction of the charging pumps in metered quantities for core! reactivity control.

~

The CVCS also includes a boron thermal regeneration-system (BTRS) which is used to-control core reactivity.

Portions of the CVCS also function as-part of the SI system. This shared function is described in Sectin 2.8.

2.12 Primary Component Cooling Water System The primary component cooling ~ water (PCCW) system provides heat removal -from

- primary system processes and equipment.

. The system consists of two train associated loops, each of which serves one train of ECCS and containment spray related_ components and other plant related system processes and equipment used during normal operation. Each PCCW loop cools two reactor coolant pump motors.

F-15

.: 0040V/1:101183

In_ addition to the two train associated loops, the system includes a dedicated train associated thermal barrier cooling system that cools all four reactor coolant pump thermal barriers. This'substem consists of two heat exchangers.

Each heat' exchanger is cooled by one PCCW loop. Either thermal barrier cooling pump and either heat exchanger provides cooling for all four reactor

' coolant pump thermal ~ barriers.

2.13' Service Water System The service water (SW) system provides heat removal from the PCCW system and the emergency diesel generator' jacket cooling water system. Like the PCCW system, the SW system consists of two train associated loops, each serving its respective train associated heat loads. Either or both SW loops can cool secondary plant heat loads but are isolated on SI actuation.

. The ultimate ' heat sink can be either the Atlantic Ocean or the atmosphere.

When the Atlantic Ocean is the ultimate heat sink, each SW loop consists of-two redundant SW pumps per train, piping loops, heat exchangers and intake and

' discharge points for seawater. When the atmosphere is the ultimate heat sink, each SW loop consists of a mechanical draft cooling tower, cooling tower pump, heat exchangers and cooling tower fans.

-The normal heat sink is the Atlantic Ocean. Should the flow of seawater become restricted, the cooling tower is automatically placed in service.

2.14 Containment Spray System The containment spray system provides containment pressure suppression and airborne fission product removal for events that require engineered safety features actuation. Each train associated containment spray system consists of a containment spray pump, heat exchanger, spray header and a shared spray additive tank. The spray additive tank contains a twenty percent sodium hydorxide solution. This tank gravity feeds the RWST upon receipt of a Hi-3 ESF ' actuation signal . The RWST provides suction to all ECCS and containment spray pumps.

F-16 0040V/1:101183

l-2.15 Containment Atmosphere Control System

.The containment atmosphere control system provides containment heat removal

.and combustible l gas mixture control. It consists of the containment fan coolers, containment electric hydrogen recombiners and containment ventilation

-equipment that provide for mixing of.the containment atmosphere. During normal oerating conditions five of the six containment fan coolers operate to keep. internal containment temperature within environmental limits.

In accident conditions requiring containment spray, the function of heat removal is accomplished by spray system operation. Combustible gas is mixed by two train' associated air mixing fans. Recombination of H AND O p is 2

accomplished by two train associated electric recombiners inside the contain-ment. . Containment fan coolers are not used for any accident mitigation function. .

2.16 Main Steam System

'The main steam system provides controlled heat removal from the reactor coolant system via the steam genrators. It consists of separate main steamlines;from each steam generator that join-to a common steam header to the turbine generator / condenser. The steam generators can be isolated from the main steam header by main steamline isolation and bypass valves located in the individual main steamlines.

Main steam release capabiltiy is provided via the condenser steam dump system and the atmospheric steam release system. The condenser steam dump system

-uses the main steam header and steam dump valves to the condenser. The atmospheric steam dump system uses power operated. relief valves upstream of the main steam isolation and bypass valves to release steam to atmosphere.

The atmospheric steam dump valves are hydraulically operated and do not require plant air for operation.

Each main steam line contains ASME code safety valves for over pressure protection.

F-17 0040V/1:101183

Steam supply lines from the main steam lines to the turbine-driven EFW pump are provided from two steam generators. The steam supply lines include isolation valves for initiation and isolation of steam supply to the turbine-driven EFW pump.

2.17 Main Feedwater and Condensate System The main feedwater and' condensate system provides coolant to the secondary side of the steam generators during plant power operation. It consists of separate main feedwater lines to each steam generator that originate from a commmon main feedwater header. The steam generators can be isolated from the main feedwater header by feedwater flow control valves, bypass valves and isolation valves located in the individual main feedwater lines.

The condensate portion of the system consists of three fifty percent condensate pumps and two fifty percent heater drain pumps.

2.18. Emergency Feedwater System The emergency feedwater (EFW) system provides secondary coolant to the steam genrators for events that require engineered safety features operation or when the startup feedwater pump is inoperable.

The EFW system consists of two centrifugal pumps, one trubine-driven and one motor driven. Either or both pumps supply secondary makeup to all four steam

.genrators. Flow control valves allow the operator to throttle faulted steam generator, will terminate flow to meet the generic flow requirements.

Connections are made into the EFW discharge headers ' supply secondary coolant makeup from the startup feed pump.

The EFW pumps.e e always -aligned to the condensate storage tank (CST). Makeup to the CST is t.om the demineralized water storage tank (DWST) or the water treatment facility.

t F-18 0040V/1:101183

l The CST is a 400,000 gallon tank with 200,000 gallons dedicated for EFW. The DWST is sized at 200,000 gallons.

2.19 Steam Generator Blowdown System The steam generator blowdown system provides letdown from the secondary side of the steam generators. It consists of separate blowdown lines from each steam generator that terminate in the steam generator blowdown tank. The blowdown water is either cleaned up and recycled, discharged or processed in the blowdown evaporators should a steam generator tube leak develope.

Steam generator blowdown is automatically terminated when the EFW pumps are called upon to automatically start.

r 1.20 Primary Semp'11ng System The sampling system provides means fot sampling process systems. It con-sists of the sampling system equipment that can be used to sample the RCS and the containment recirculation sump.

2.21 Spent Fuel Storage and Cooling System -

The spent fuel storage and pool cooling system controle fuel storage post-tions to ensure a suberitical geometric configuration and provides heat removal to maintain stored fuel within specified temperature limits. It includes the level instrumentation for the spent fuel pool.

2.22 Control Rod Drive Mechanism Cooling System The control rod drive mechanism (CRDH) cooling system provides heat remo-val from the control rod drive mechanisms. It consists of the ventilation fans used to. circulate air around the control rod drive. mechanisms.

F-19 0040V/1:101183

l2.23 Control Rod Control System The control rod control system controls the positions of the control rods in the reactor core. It includes those controls used to manually insert control rods.

2.24 Turbine Control-System The-turbine control system controls the turbine generator. It includes those controls used to manually trip the turbine generator supplied by General Electric.

2.25 Electrical Power System The electrical power system provides AC and DC electircal power to equipment that requires electrical power to accomplish their functins. It consists of two independent off-site circuits, either of which can supply all on-site power requirements. The on-site emergency AC power supply is a two train system powered by separate diesel generators. The DC power system is a two train system. Each vital DC train consists of redundant batteries, battery charges and DC buses. Vital AC instrument power can be supplied by either AC from two sources or from the DC power supply inverters.

The emergency diesel generators automatically start on the following:

o SI signal (runs but does not connect to the bus unless loss of power occurs.).

o Loss of Power (as sensed by undervoltage relays.)

The diesel generators automatically energize their respective ac emergency buses if both off-site power circuits are unavailable. The following major loads'are sequenced on the associated buses in accordance with plant conditions as follows:

F-20 0040V/1:101183 i

r

a. Loss of. Power (LOP) without SI o Charging pumps o PCCW pumps

o. Service water pumps or cooling tower pumps / fans o RHR pumps 1f running prior to LOP

~

o Inverters and battery chargers o Emergency lighting o .HVAC systems o Air compressors (plant and' containment) o ' Boric acid transfer pumps o BIT recirculation pumps and heaters o' MOVs o Emergency feedwater pump (motor driven)

b. %ss of' Power '(LOP) with S o Charging pumps o SI pumps o 'RHR pumps o Containment spray pumps o PCCW pumps o Service water pumps or cooling tower pumps / fans o Inverters and battery chargers o' Emergency lighting o HVAC systems

-o Hydrogen mixing fans o Power receptacle for H rec mbiners 2

o Pressurizer heaters o Emergency feedwater pump l(motor driven) o MOVs-2.26 Pneumatic Power System The pneumatic power system supplies pneumatic power to equipment that requires pneumatic power to accomplish its functions. Equipment in this category includes:

o Condenser steam dump valves o Repostioning certain valves from their safety related position to their normal non-safety related position F-21 0040V/1:101183

a

%J ATTACHMENT A PROCEDURAL INSTRUMENTATION REQUIREMENTS FOR SEABROOK CRITICAL SAFETY. FUNCTION MONITORING AND STATUS TREE LOGIC INPUTS BASED IN PART ON THE WESTINGHOUSE OWNER'S GROUP GENERIC EMERGENCY RESPONSE GUIDELINES

'Rev. 00 June 1983 o

0040V:1/101183 F-22 l

THISLINFORMATION MEETS THE FOLLOWING CRITERIA FOR SEABROOK CRITICAL SAFETY

, FUNCTION MONITORING AND STATUS TREE LOGIC INPUTS:

o All variables-are hard wired t'o main control room indicators, recorders or displays for manual monitoring of critical safety functins. This provides on-the-spot monitoring ability should the Main. Plant Computer System or CRTs be off line.

o 7A11 variables. are hard wired to main control room indicators, recorders or displays to direct operator actions.

o All variables provide' input directly to the Critical Safety Function Status Tree Logics which direct Emergency Action Levels. '

o These variables are classified as Type A, Category I by Regulatory. Guide 1.97.

J l

c.

6

- 0040V:1/101183 F-23 3 - -- - . g ,e- + -,qe- - , - -~n --n-,-e,- -- , - -, - - - - ,,w, , - - - e, ,,, ,, -, r---,, ,m-- ,

~

3_.

VARIABLE UNITS. DISPLAY I SOBCRITICALITY:

Intermediate Range Flux Level  % Indicator

' Intermediate Range Flux Rate DPM Indicator

-Shutdown Margin Monitor- -CPS Indicator - of source range counts with-shutdown margin alarm CORE COOLING:

Core Exit Temperature *F Indicator - averaged T/C value of center and hot channel in

_ each quadrant RCS-Subcooling *F Indicator - computed from_RCS

' wide pressure and core exit temperature above-Reactor Vesse'l Level- %i Indicator Reactor -Vessel W Level  % Indicator HEAT SINK:

-Steam Generator Level - NR -  % Indicator Steam Generator Levelf- WR  % Indicator and recorder Steam Generator PSIG Indicator and recorder Emergency - Feedwater Flow : -GPM Indicator INTEGRITY:

RCS Cold Leg Temperature - WR F Indicator and recorder RCS: Pressure PSIG Indicators 3000 PSIG 0-700 PSIG CONTAINMENT:

Conta'inment Pressure PSIG Indicator and recorder Containment Building Level GALLONS ' Indicator and recorder Containment Post Accident Radiation R/HR Indicator

INVENTORY:

. Pressurizer Level  % Indicator 0040V:1/101183 F-24

e ATTACHMENT B PROCEDURAL INSTRUMENTATION AND CONTROL REQUIREMENTS FOR SEABROOK EMERGENCY RECOVERY PROCEDURES

~

BASED IN PART ON THE WESTINGHOUSE OWNER'S GROUP GENERIC EMERGENCY RESPONSE GUIDELINES Rev. 00 June 1983 b,.

0040V:1/101183' F-25

PROCEDURAL INSTRUMENTATION AND CONTROL REQUIREMENTS FOR THE SEABROOK EMERGENCY RECOVERY PROCEDURES I - Indication C - Control ITEM I C Reactor Trip Actuation System Reactor Trip and Bypass Breakers X X Reactor Trip' Signal X X Turbine Trip Signal X X ESF Actuation System SI Signal X X SI Signal Block X X Si Signal Reset X X Containment Isolation Phase A Signal Reset X X Feedwater. Isolation Signal Reset X X Containment Spray Signal X X Containment Spray Signal Reset X X Main Steamline Isolation Signal X X Tower Actur. ion Signal X X Nuclear Instrumentation System Intermediate Range Neutron Flux X Intermediate Range Flux Rate X Source Range Counts X Source Range Shutdown Margin Alarm X Containment Instrumentation System Containment Pressure X Containment-Temperature (averaged value) X Containment Building Level X Phase A Containment Isolation Valves X X Phase B Containment Isolation Valves X X Containment Ventilation Isolation Valves (CAP, COP) X X Containment Hydrogen Concentration X X Containment Building Level Recorder X 0040V:1/101183 F-26

_____ _ - __ _ _. l

l l

ITEM I C Reactor Coolant System 7~

RCS Wide Range Pressure X- -

Pressurizer Pressure X --

?~ .RCS Hot Leg Wide Range Temperature X -

'RCS Cold Leg Wide Range Temperature X -

RCS Average Temperature X -

Core Exit' Temperature (averaged _value) X -

Pressurizer Water Temperature X -

Pressurizer Level -X -

RCS'Subcooling ( F). X -

, Reactor-Vessel' Level Instrumentation System (RVLIS) X -

o Vessel Level o: Vessel W Pressure

~ Reactor Coolant Pumps X X Pressurizer PORVs X X Pressurizer PORV Block Valves X X Pressurizer spray Valves. X .X~

Reactor Vessel Vent Valves X X Pressurizer Heaters- X X Pressurizer Normal Spray X X Pressurizer Pressure Control X X lPRT Pressure X -

PRT Level X -

PRT Temperature X -

Relief Valve Tail-Pipe Temperatures X -

o -P0ressurizer PORVs o Letdown Line Pressurizer Safety Valve Tail Pipe Temperature X -

RCS Vent' Paths X X o RV Head Vent Valves

, o Pressurizer PORVs RCS Dilution Paths- X X o CVCS Makeup System o -BTRS Emergency Core Cooling System Accumulator Pressure X -

SI. Pump Discharge' Pressure X -

Boron Injection Tank (BIT) Temperature X -

Refueling Water Storage Tank (RWST) Level X -

. RWST-Level Recording- X -

CCP Flow to BIT X -

SI' Flow. X -

Throttle SI Flow X X 0040V:1/101183 F-27

ITEM I C Emergency Core Cooling System (Cont)

Throttle CCP Flow via BIT X X CCPs X X RHR Pumps X X SI Pumps X X All ECCs~Related Remotely Operated Valves X X (SI,RHR,CS)

Residual Heat Removal System RHR Flow. X -

RHR Temperature Control X X RHR-Suction Valves from rCS X X Chemical and Volume Control System VCT Level X -

' Boric Acid Tank Temperature X -

Charging Flow X -

RCP Seal Injection Flow X -

RCP Seal Leakoff Flow X CVCT Outlet Isolation Valves X -

Charging Line Isolation Valves X X-Charging Line Flow Control Valve (FCV-121) X X Charging Line Hand Control Valve (HCV-182) X X Pressurizer Auxiliary Spray Valve X X CCP Miniflow Valves X X CCP Suction Valves-from RWST X X Letdown Pressure Control Valve (PCV-131) X X RCP #1 Seal Leakoff Valys X X Letdown Flow X -

Letdown Isolation Valves X X Letdown Flow Control Valves (NCV-189, HCV-190) X X Excess Letdown Isolation Valves X X Excess Letdown Flow Control Valve (HCV-123) X X RCP Seal Injection Outside Containment Isolation-Valves X X Emergency Borate Valve (CS-V426) X Boric-Acid Pumps- X X CVCT Makeup Control System X X o RMW Pumps o_ Makeup Flow Control Valves and Controllers FCV-110A, 110B, 111A and 111B BTRS Master Control Switch 0040V:1/101183 F-28

ITEM I C Component Cooling Water System PCCW Pumps .

X X PCCW Temperature Control Valves X X Thermal Barrier Cooling Pumps X X RCP Thermal Barrier Outlet Valves X X PCCW Valves to Components and Systems X X o Letdown Heat Exchanger

o RHR Heat Exchangers o Containment Spray Heat Exchangers o Excess Letdown Heat Exchangers o Spent-Fuel Pool Heat Exchangers o PCCW Radiation Monitors PCCW Head Tank Level X X PCCW Flow to RCP Motors X Ultimate Heat Sink Service Water Pumps X X

-Cooling Tower Pumps .

X X All-Service Water Remote Operated-Valves X X Cooling Tower Fans X X Cooling Tower Level X Service Water Pump Bay Level .

X Flow Through Diesel Generator Jacket Water Heat Exchanger X

-Containment building Spray System Containment Spray Pumps X X Containment Spray Valves to Spray Rings X X Containment Spray Pump Discharge Pressure X Combustible Gas Control System

- Containment. Ventilation Iso.lation Valves (COP, CAP) X X -

Hydrogen Monitoring Systems X X Hydrogen Concentration X Hydrogen Recombiner Systems X X Hydrogen Mixing Fans X X.

Containment Enclosure System Containment Enclosure Emergency Exhaust Filter Systems X X Containment Enclosure Pressure X 0040V:1/101183 F-29

_ _ _ - _ - _ - - - _ _ _ _ _ - _ _ l

r ITEM I C Main Steam System Steam Generator Pressure X Steam Generator Narrow Range Level

~

X -

Steam Generator Wide Range Level X Steam Generator ASDVs X X Condenser Steam Dump Valves X X Main Steamline Isolation Valves X X Main Steamline Isolation-Bypass Valves X X Steam Supply Valves to Turbine Driven EFN Pump X X Turbine Stop' Valves X Condenser Steam Dump Controls (T,yg& Pressure Modes) X X Main Steamline Radiation X Main Feedwater and Condensate System Feedwater Flow Control Valves X X Feedwater Flow Control Bypass Valves X X Feedwater Isolation Valves X X Main Feedwater Flow X X Condensate Pump Discharge Header Pressure X -

Startup Feed pump X X Startup Feed Pump Discharge Valves (FW-V156 & V163) X X Emergency Feedweter System EFW Flow to Each SG X -

Condensate Storage Tank Level X -

Motor Driven EFW Pump X X Steam Driven EFW Pump X X

_.EFW Flow Control Valves X X Diesel Driven Fire Pumps (Emergency Makeup to CST) X X Steam Generator Blowdown System Steam Generator Blowdown Isolation Valves X X Electric Power System Emergency Bus Voltage (E-5 & E-6) X

-UAT Breakers X X RAT Breakers X X Diesel Generators X X DC Bus Voltage X -

DC Bus Current X -

Grid Voltagr X

'0040V:1/101183 F-30

i ITEM I C

' Air Systems Instrument Air Header Pressure X -

Containment Air. Header Pressure X -

Service Air Compressors X X Containment Air Compressors X X-

. Control' Rod Instrumentation System Control; Rod Position X -

Control Rod Bottom Lights X -

Radiation instrumentation System Process Monitors o SG Blowdown X -

o Steamlines (upstream or ASDVs) o Condenser Effluent o PCCW o Plant Vent Radiogas (low and high range)  ;

o . Plant Vent Air Particulate (low and high range) j o PAB Air Particulate o PAB Radiogas o Containment Enclosure Air Exhaust o Containment Enclosure Emergency Exhaust o PAB Misc. Ventilation

~

.o1 Air Intakes Area Monitors o RHR Vault 1 X -

o RHR Vault 2 o Charging Pump CS-P-2A o Charging pump CS-P-2B o -Charging Pump CS-P-128 o' CVCT Area o< Sampling Room o PAB Lower-level Turbine Control System Turbine Runback X -

~

-UO40V:1/101183 F-31

A

~

ITEM- I. C

~ ~

Control Rod Drive Mechanism Cooling System

' Control Rod Drive Mechanism Fans' X X

' Sampling System

.RCS Sampling .

X- X Containment Recirculation Sump Sampling -

X LSpent' Fuel Storage and Cooling System

' Spent Fuel PooliLevel- X i '

- 4)

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0040V:1/101183 F-32 -

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s Appendix C -

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LESSON OUTLINE NOTES AND REFERENCES 4.4.8 Purpose of Cautions o inform of potential hazards to personnel or equipment o advise on actions / transitions necessary due to plant condition changes 4.4.9 How to use Notes / Cautions o introduced by bold lettered NOTE or CAUTION o text extends across page o multiple items identified by bullet (o) o precede applicable step o applicable to entire guideline if precedes first operator step of guideline o NOTES / CAUTIONS preceding transition steps to other guidelines are applicable 4.4.10 Immediate Actions o can be performed without using written guideline o note advises of the immediate steps 4.4.11 Task Completion Requirements o need not be completed before proceeding unless specified o sufficient to begin task and know progression is satisfactory o allows efficient procedure use with lengthy steps 4.4.12 Transitions o different steps in the same guideline or o other guidelines o NOTES or CAUTIONS preceding step transitioned to are applicable G-4 0040V/1:101183

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4.4.15 Example of Guideline Usage (overall review) 1 TP 4.4.15 a c ;M. , g,.. ' :. ' . 6 .

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o six different trees 1 TP 4.5.1 a ..[% p nkh

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y g -, l LESSON OUTLINE NOTES AND REFERENCES o user should log color and instruction s and continue to next tree in sequence g o RED or 0 RANGE FRG entered, perform to completion unless pre-empted by a higher priority condition f 4.5.3 Red terminus encountered j h o immediately stop ORG in progress 2 o perform required FRG o if higher priority Red arises, suspend v [ lower Red and address higher Red g 4.5.4 Orange terminus encountered o monitor all remaining trees o if no Red encountered, suspenc,ORG and i y perform required FRG , o if Red condition or higher priority Orange [ condition arises, suspend Orange ) FRG and address Red or Orange condition { { 4.5.5 Yellow terminus encountered i o immediate operator action not required h o indicative of off-normal and/or -

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temporary cc'dition (may be restored to normal .; Q 'f v % c: Mi g by actions in progress) M -x ,. % . . .7

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e o Yellow condit:ons FRGs are implemented ([?,.] } b based on operator evaluation i ? . '"i 4.5.6 Green terminus encountered . F o indicative of normal plant conditions QJcwx Q. . +A i; (CSF of that tree not challenged) -h

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u y o no operator actions required 4 ., pf 'a. . [ 4.5.7 Tree Monitoring .. { 4.5.7.1 continuous if any status coded Red or ' k.jg j b Orange exist g,.#gWN? f/ .  % ~ .',54/ : 2 y _.

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LESSON OUTLINE NOTES AND REFERENCES . 4.5.7.2 no condition worse than yellow .. - o reduce monitoring frequency to 10-20 minutes intervals

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o dependent upon no significant change in plant status . . 4.5.7.3 Terminate tree monitoring o Rx protection system restored (trip breakers closed) o Engineering Safeguards System restored -- (SI reset) 4.5.8 Example of CSF status tree usage 1 TP 4.5.8 (subcriticality) 4.5.3.1 RED terminus encountered o immediate response required (no other higher priority than subcriticality (RED) .. o go to FRG FR-S.1 . o suspend whichever URG in proaress o after FRG completed return to guideline and step in effect 4.5.8.2 ORANGE terminus encountered  ; j o prompt response required 4 [.2 i i t cg i

  ;                 o go to FRG FR-S.1
                                                                                   }{p(I o continue monitoring status trees                              1..

o any RED terminus takes priority i over the subcriticality ORANGE 4.5.8.3 YELLOW terminus encountered L o action not required immediately o continue monitoring trees

   $       4.6      Use of ERG network in control roum                                      4. .,

4.6.1 Direct entry Use large trans- '- v  :..r o Rx trip or SI (E-0) ition charts for a:.t;gh {.Q ,. o Loss of all ac (ECA-0.0) this presentation l rd, /;d,yq.-

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LESSON OUTLINE NOTES AND REFERENCES 4.6.2 Entry into E-0 o remains in E-0 and directed by action step to monitor trees or o transition to other guideline, at which point trees are to be monitored 4.6.3 Tree monitoring and actions - 4.6.3.1 Use standard rules of usage 4.6.3.2 May be done by o control room operators o other member of shift assigned . to control room o dedicated computer routine (backup paper required if computer goes down) . 4.6.3.3 Operator in charge of recovery , o be immediately informed of RED or ORANGE priority status condition o be regularly advised of YELLOW and GREEN conditions 4.6.3.4 ORG actions o suspended if either RED or ORANGE condition detected o NO ORG actions are to be performed while CSF (RED or ORANGE condition) is being restored (unless specific 1ed by FRG) 4.6.3.5 CSF restored o normally return to guideline and l step in effect before FRG implemented l 0 at time's FRG require transitions to l ORG other than one in effect (due to l , conditions created by FRG) l ' l G-8 , 0040V/1:101183

LESSON OUTLINE NOTES AND REFERENCES 4.6.3.6 FRG to ORG ,. o operator judgement required Example: FR-H.1 to prevent inadvertent establishes alternate reinstatement of RED or ORANGE feed path to SGs. condition ES-0.1 requires main o ORGs optimal assuming safety feedline be isolated. equipment availt.ble Main feed or condensate o RED or ORANGE condition implies feed may be alternate .. seme safety equipment or function path in FR-H.1. The not available operator would not want o some ORG adjustments may be required to isolate the main feed line. . 4.6.3.7 Certain ECAs take precendence over FRGs Example: ECA-0.0 deals based on specific initiating events with loss of electrical (identified by NOTE at beginning of power to ac emergency 4. guidelines) busses. No safeguards M.m.e,4b. r.; y % equipment used to j{j#Q@.Q restore CSFs is oper- %ki;f.!l$( . able. Trees should be L(Nt g.+y ;. .: monitored for inform- .; e.- . ation only. ECA-0.0

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y[ T ,- provides best strategy S.flif for maintaining CSFs. 2 [_;"'i 1 4.6.4 ERG use ends 2j.b G o transition to normal plant f operating procedure ,[; ( d ;) / o transition to " appropriate" gS,.=

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procedure while on RHR at cold  ; shutdown conditions l,j'{;f,g$g [Qi . . . , . j. .

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o cold leg recirculation or hot hot leg recirculation with longer term recovery actions being analyzed G-9 0040V/1:101183

l l LESSON OUTLINE NOTES AND REFERENCES 4.7 ERGS Modes of Applicability o originally written for transients a- 1 TP 4.7 a-c

                                " hot" or "at power" conditions o transients would result in a protective function o guidance based upon availability of safety equipment defined in Tech Specs for Mode 1 and 2 operation 4.7.1                    Othee modes of operation o safety equipment defined for modes 1 and 2 may not be available o some detailed instructions in ERGS not applicable 4.7.2                    CSF trees o assumes mode 1 or 2 initial conditions o tree use may be expanded to other modes if it.+ent of each tree is understood
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