Text

Emergency Operating Procedures,Reactor Operator Simulator Training
	ML20215M874
Person / Time
Site:	Vogtle
Issue date:	10/27/1986
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20215M870	List: ML20215M869; ML20215M874; ML20215M877; ML20215M881; ML20215M887; ML20215M890; ML20215M892;
References
	PROC-861027, NUDOCS 8611030395
	Download: ML20215M874 (109)
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3y E0P COURSE INTRODUCTION LICENSED OPERATOR CLUSTER 37 TITLE: Emergency Operating Procedures PHASE: Reactor Operator Simulator Training I. OVERVIEW Following any automatic reactor trip the operator enters the emergency operating procedures. In most emergency situations, the control of the plant is no longer determined by the operator and the automatic control systems. The operator needs to develop an understanding of plant characteristics in these off-normal states in order to regain some measure of control and put the plant in stable, safe condition. Vogtle E0Ps are based on observable plant " symptoms" and systematically enables the operator to regain control of the situation.

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II. TOPICS TO BE COVERED The Vogtle E0Ps are based on the Westinghouse Owners Group guidelines. As such, there are three levels of emergency procedures which are covered by these lessons:

, b) , 1. Optimal Recovery - directed actions are based on an assumed V cause of the event

2. Emergency Contingency Actions - these respond to less probable initiating events.

3. Function Recovery - these focus on protecting plant radiaion barriers regardless of the initiating cause.

Additionally several abnormal operating procedures which have their basic impact on the integrated plant have been included in these lessons.

! The materials have been separated into seven basic types of plant accidents:

l 1. Non-accidents - inadvertant operation of safety systems

2. Natural Circulation Cooling

3. Loss of Coolant Events (both primary and secondary)

4. Steam Generator Tube Ruptures j 5. Support System Failures j 6. Post-accident Monitoring Concerns

7. Function Recovery Procedures 4 This first volume deals only with the first category above.

8611030395 DR 861027 ADOCK 05000424 PDR O

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- r p9f III. GENERAL OBJECTIVES OF THE E0P TRAINING Specific objectives are stated for each section of the E0P training program. The general objectives are to develop the following knowledges and abilities of the reactor operator trainee:

1. Ability to recognize abnormal indications for system operating parameters, which are entry-level conditions for emergency and selected abnormal operating procedures.

2. Ability to perform those actions, without reference to procedure for all casualties which require immediate operation of system components or controls.

3. Ability to execute procedural steps of the abnormal and emergency operating procedures.

4. Ability to use plant computers to obtain and evaluate parametric information on system and component status.

5. Knowledge of the bases of the E0Ps relating to operational limits'.

6. Ability to locate, explain, and apply all limits and precautions.

7. Ability to recognize indications for system operating parameters which are entry level conditions for tech specs.

8. Knowledge of system status criteria which require the notification of plant supervisors or off-plant personnel.

9. Ability to take actions called for in the Facility Emergency Plan.

Objectives have also been included to satisfy licensing commitments and to include lessons learned from events occurring in the nuclear industry.

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IV. TRAINING MATERIALS - E0P TEXT VOLUME I. Introduction to the EOPS

1. Introduction to the E0Ps

2. Format of the E0Ps

3. Plant Response to a Reactor Trip s,

4. Operator Response to a Reactor Trip

5. Plant Response to an Inadvertant SI

6. Immediate Manual Actions Following an SI

7. Evaluation of Plant Conditions following an SI

8. Termination of an Inadvertant SI R

9. Rediagnosis of Plant Conditions After Entering an E0P VOLUME II. LOSS OF FORCED FLOW

1. Respond to Partial Loss of Flow between P-7 & P-8 '

2. Plant Response to Natural Circulation

3. Operator Response to a Loss of RCP Capability R

4. Performing a Natural Cire Cooldown with a Void (with RVLIS)

R

5. Performing a Nat. Cire. Cooldown with Void but without' RVLIS t

4 VOLUME III. LOSS OF COOLANT ACCIDENTS

1. Plant Response to Loss of Reactor Coolant Accidents 1

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2.

3.

4.

Responding to an RCS Leak Responding to a Small or Medium Break LOCA Performing the Post-LOCA Cooldown

5. Responding to a Large Break LOCA

6. Plant Response to Loss of Secondary Coolant

7. Responding to a Steam or Feedline Break R

8. Responding to a Loss of Emergency Coolant Recirculation R

9. Adding Makeup from the Spent Fuel Pool VOLUME IV. STEAM GENERATOR TUBE RUPTURES R

1. Respond to Abnormal Secondary System Chemistry

2. Plant Response to a Steam Generator Tube Rupture

3. Respond to a Steam Generator Tube Rupture

4. Use Steam Dumps to Cooldv<c luptured Steam Generator

5. Performing a Post-SGTS Poole. 3na using Backfill Method R

6. _ Responding to a SG74

t;n loincident LOCA

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7. Responding to a SGTL siti.e, Pressurizer Pressure Control

(*) asterisi refers to a simulator performance module.

(R) refers to module done normally in requal and not in basic program.

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< o o S VOLUME V. SUPPORT SYSTEM FAILURES

1. Respond to Loss of Instrument Air R

2. Establish Charging and Letdown without Instr. Air

3. Respond to Loss of IE Electrical Bus

4. Plant Response to a Loss of All AC

5. Respond to Loss of All AC R

6. Recover from Loss of All AC VOLUME VI. POST-ACCIDENT MONITORING

1. Plant Safety Monitoring Sytem & RVLIS

2. Safety Parameter Display System

3. Emergency Plan Overview

4. Responding to High RCS Activity

5. Containment and Radiation Hazards R

6. Responding to Abnormal Containment Pressure R

7. Responding to a Realease of Chlorine Gas

8. Use of Instrumentation to Detect Core Damage VOLUME VII. FUNCTIONAL RECOVERY PROCEDURES

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1. Monitoring of the Critical Safety Functions (CSFSTs)

2. Plant Response to a Anticipated Transient Without Trip

3. Operator Response to a Failure of the RPS

4. Plant Response to Inadequate Core Cooling

5. Operator Response to Inadequate Core Cooling

6. Plant Response to a Loss of Secondary Heat Sink

7. Responding to a Loss of Secondary Heat Sink

8. Pressurized Thermal Shock R

9. Responding to an Imminent PTS Condition R

10. Responding to a High or Low Pressurizer Level R

11. Responding to Voids in the Reactor Vessel O

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O V. STUDENT EVALUATION There is a written test associated with each of the seven sections of the E0P training materials. The questions are part of the Vogtle exam bank and conform to procedural standards for question and test construction and use.

There are performance tests on each of the identified procedures.

Each performance-based chapter includes the test objective co be signed off by the simulator instructor following successful demonstration of the skills being practiced. A final performance evaluation period is given to verify proper accident recognition and response.

The general procedure for the performance exam is as follows:

You will be asked to perform the task specified in the Performance Obj ective. The Performance Test will be conducted in the Plant Vogtle simulator and an instructor will determine if your performance j demonstrates that you could complete this task during an actual event. ~

The Performance Test of one chapter may be conducted in conjunction with performance tests for other chapters. In the simulator exercises and the performance test you must meet the following criteria in addition to those specified in the objective. All immediate operator actions must be performed without reference to the procedure. Any " Response Not O' Obtained" actions must be carried out as appropriate. All procedure cautions and notes must be observed. Any continuous action steps must be performed when appropriate. The foldout page must be monitored. The emergency procedure actions must be followed in sequence. Transitions to other procedures or transfers to con-sequential steps must be performed according to the direction of the step in effect.

Before going to take the p'rformance tests do the following:

1. Review the referenced procedures. Pay attention to all procedure cautions, notes, and attachments. If you have any questions, refer to the Step Description section of the associated chapter or ask your instructor.

2. Take the text to the simulator when you can get in without interfering with other training. Be sure you can locate all necessary instruments and perform the steps of the E0P quickly.

3. See the schedule or the instructor for the practice simulator exercises and the performance test.

4. During the simulator practice session, follow the procedures as the instructor talks through the response to the event. If you have any l trouble following the actions of your instructor, ask the instructor to l stop and explain.

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5. Your instructor will observe you as you respond to each even. The i O, task is satisfactorily completed when the criteria listed in the Performance Objective have been met.

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6. Have your instructor sign and date the Performance Test for each type of event after you have demonstrated mastery of the Performance Obj ective.

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[ INTRODUCTION TO VOLUME VII

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LICENSED OPERATOR CLUSTER 37 TITLE: Emergency Operating Procedures - Function Recovery Procedures PRASE: Reactor Operator Simulator Training I. OVERVIEW Several emegergency procedures are designated as Function Recovery Procedures and are used only when there is a serious challenge to one of the plant barriers to radiation release. They are designed to recover each of the six critical safety functions as described in volume I.

II. TOPICS TO BE COVERED

1. Use of the Function Restoration Procedures

2. Anticipated Transient Without Trip

3. Respond to Failure of the RPS to Trip When Required

4. Inadequate Core Cooling (TMI2 Scenario)

5. Respond to Inadequate Core Cooling -

6 Loss of Secondary Heat Sink

7. Respond to Loss of Secondary Heat Sink

8. Pressurized Thermal Shock

9. Respond to Pressurized Thermal Shock

10. Respond to High Pressurizer Level Respond to Voids in the Reactor Vessel

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11.

(*) asterisk refers to a simulator performance module.

III. PERFORMANCE OBJECTIVES FOR FUNCTION RECOVERY PROCEDURES

3. RESPOND TO FAILURE OF THE REACTOR PROTECTION SYSTEM TO TRIP Given that a reactor trip is required but has not occurred either automatically or as a result of operating both manual trip switches and manually opening supply feed breakers to the control rod drive motor generator sets, add negative reactivity to the core. The reactor must be suberitical:

1. Power range channels must be less than 5 percent.

2. Intermediate range channels must indicate negative startup rate.

5. RESPOND TO INADEQUATE CORE COOLING CONDITION Given that either

1. The core exit thermocouple temperatures are greater than 1200* F, or

2. RCS subcooling is less than 31* F, no RCPs are running, the core exit thermocouple temperatures are greater than 700* F, and the RVLIS full range indication is less than 39 percent, restore core cooling.

Adequate core cooling must be established as defined by steps 5-7, O-s step 16, or step 23 of Procedure 19221.

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7. RESPOND TO A LOSS OF SECONDARY HEAT SINK Given that a loss of secondary heat sink has been identified from a heat sink status tree evaluation (all narrow range steam generator levels are less than 27 percent, and the total available feedwater flow to the steam generators is less than 550 gpm), or that an impending loss of heat sink has been idenrified from Procedure 19000 (step 18, total feed flow is less than 550 gpm), restore and/or maintain adequate secondary heat removal capability.

Establish RCS bleed and feed heat removal if adequate secondary heat removal cannot be restored or maintained. The restored secondary heat sink must be verified. The RCS bleed and feed must be terminated.

9. RESPOND TO PRESSURIZED THERMAL SHOCK CONDITION Given an imminent pressurized thermal shock condition detected by any of the three possible paths from the Integrity status tree, limit or prevent any potential flaw growth. Depressurize the RCS while maintaining the required subcooling and pressurizer level.

Stable normal operating conditions must be established and any need for a " soak" must be identified. -

10. RESPOND TO HI PRESSURIZER LEVEL Given that the inventory status tree evaluation has identified that the pressurizer level is greater than 92 percent and RVLIS indicates that the upper head is full, draw a pressurizer steam bubble so that normal pressurizer pressure control is possible.

The pressurizer level must be less than 92 percent.

11. RESPOND TO VOIDS IN THE REACTOR VESSEL Given that the RVLIS indicates the upper head of the reactor vessel is not full, eliminate reactor vessel voids. The RVLIS must indicate that the upper head of the reactor vessel is full. The PRZR level must be stable.

IV. STUDENT EVALUATION See the Course Introduction for more detail. As in all the modules there are performance tests on each of the identified procedures. Each performance-based chapter includes the test objective to be signed off by the simulator instructor following successful demonstration of the skills being practiced. A final performance evaluation period is given to verify proper accident recognition and response.

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'9 SCHEDULE

TRAINING DAY SELF-STUDY / DISCUSSION SIMULATOR

1. Introduction to the E0Ps Reactor Trip Reactor Trip

2. Procedure 19000 Inadvertant SI SI Termination

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3. Natural Circulation Natural Circulation

, 4. Loss of Vital DC Loss of Vital DC Loss of Instrument Air Loss of Instrument Air Safety Grade Charging

5. Exam on sections 1 & II. Performance Evaluation LOCA Demonstration .

6. LOCAs LOCAs 1

, 7. Secondary Breaks Steamline/Feedline break

8. Loss of All AC Blackout Loss of IE busses Loss of IE busses

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j 9. ATWT ATWI Loss of Heat Sink Loss of Heat Sink l ////////////////////////////////////////////////////////////////

10. Exam on sections III & VI Performance Evaluation SPDS practice

11. Steam Generator Tube Rupture SGTR l 12. EPI.Ps SGTR Cooldown

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13. Pressurized Thermal Shock PTS

14. Inadequate Core Cooling ICC

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15. Exam on sections IV, V, VII Final Performance Evaluation.

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O INTENTIONAL BLANK O

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, . i t p CHAPTER 1 USE OF THE FUNCTION RESTORATION PROCEDURES A

After completing this chapter you will be able to:

I l 1. Describe the purpose of the Function Restoration Procedures.

2. List in order the six Critical Safety Functions from memory.

I 3. Identify the order of priority of challenges to the critical safety functions. That is, tell exactly which of several possible procedures

the operator should be following for a given set of plant conditions.

4. Describe how CSFST status is used in determining Emergency Plan event

! clascification.

The concept of the Function Restoration Procedures was presented in volume i I of this text. The first chapter described the six critical safety functions (CSFs) and how both Optimal Recovery Procedures and the Function Restoration Procedures were designed to maintain the CSFs. Maintaining the CSFs as indicated by a given set of parameters ensures that the barriers to radiation

() release are not violated. This chapter will present further detail in how the Optimal Recovery Procedures and Function Recovery Procedures work together to guide the operator in responding to any type of plant casualty.

1. Maintaining Critical Safety Functi ns l *3 1

f Under normal operating conditions, all CSFs in a full set are continuously i

satisfied with ample margin. The NSSS control systems, augmented by operator response to annunciator alarms and backed by plant technical specifications,

, serve to insure that small departures from normal operating conditions are rectified before a challenge to the CSFs develops. Under other circumstances,

, which are less likely to occur and usually involve equipment functional l failures, the plant protection systems automatically act to block potential challenges to the CSFs and to reinforce the protection of the fuel rod and RCS i pressure boundary barriers. Specifically, the protection system:

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2.

Stops nuclear power generation by initiating reactor trip.

Stabilizes RCS temperature, pressure and inventory by initiating a lh turbine trip, main feed isolation and steamline isolation, as appropriate.

3. Insures the availability of a secondary heat sink by starting auxiliary feedwater flow and enabling the condenser dump system.

4. Prevents overpressurization in the primary and secondary systems by opening the pressurizer and steamline safety valves, as necessary.

Action is required only to insure that the automatic protection systems are functioning as intended and, depending on the actual cause of the reactor trip, to initiate recovery operations.

In rare but potentially hazardous situations in which either the barrier-has failed (a LOCA or steam generator tube rupture) or an essential function is jeopardized or lost (for example, a secondary system break or a station blackout), the engineered safeguards system (ESF) is activated to insure that CSFs are maintained to protect surviving barriers. The ESF duplicates all of lf the safety functions provided by the plant protection system and broadens the barrier protection processes by automatically:

1. Starting the emergency diesel generators.

2. Initiating safety injection.

3. Isolating all nonessential containment penetrations.

4. Actuating containment spray, if appropriate.

Concurrently, operator response is invoked through the optimal recovery procedures to: -

1. Verify that the automatic systems are functioning.

2. Identify the accident.

3. Restore or replace lost essential functions.

4. Restore the plant to operating conditions as expeditiously as possible, when appropriate.

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F F However, for multiple event / multiple failure scenarios that go beyond the

%j design basis of the ESF and the scope of optimal recovery procedures, the operator is provided with a means of directly monitoring the CSFs and guidance for restoring CSFs which might be in jeopardy. In this way, a last line of defense is established against the potential release of radioactive materials due to barrier failure.

Monitoring a CSF involves checking for an appropriate set of plant parameters. These parameters are then compared with previously selected reference values in a logical array called a status tree. The combination of existing parameters defines a unique path through the tree, and also a unique

" status" of the respective CSF. If the CSF status is not satisfied, the operator is directed to an appropriate function restoration procedure (FRP) for instructions intended to restore the CSF to a satisfactory status.

2. Status Tree Format

) Each status tree consists of binary decision points that check conditions v' in the plant related to fixed, reference criteria. The decision points require the user to decide only whether a condition does or does not exist, or if a certain process parameter limit is or is not exceeded. Each possible response at a decision point leads to either another decision point or a terminus (end point). A terminus summarizes the CSF status for the particular combination of decisions leading to it. Each terminus consists of a color-coded symbol representing the degree of challenge to that CSF. The line extending from the last decision point to the terminus is also color coded to convey the same i information. Immediately adjacent to each terminus is an instruction which directs the operator-to the appropriate FRP is the CSF is not completely l satisfied.

The plant parameters that define the state of each CSF are identified on the associated status tree. Typically, only a few parameters are required to identify the status of a CSF. This limited set of parameters must be evaluated l to determine the CSF status. Each status tree has a single entry point and several exit points (termini) depending on the parameters that define the CSF status. Each pass through a status tree will yield only one exit point.

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3. Priorities in Response to CSF Challenges An order of priorities has been established for the CSFs. Because the first barrier to fission product release is the fuel matrix / cladding, the CSFs related to this barrier are given highest priority. Challenges to this barrier can come from inside and outside the barrier. The internal challenge comes from excessive core heat production resulting from fission power production; normal decay heat production is considered in safeguards system design. Core heat production in excess of safeguard systems core heat removal capability is the most severe challenge to the fuel matrix / cladding barrier. If the core is at power, the energy production represents a potential additional significant challenge to the other barriers which may also be challenged or failed.

Consequently, suberiticality is the highest priority function. The external challenges to the fuel matrix / cladding barrier come from inadequate decay heat removal due to either inadequate reactor coolant or secondary coolant. Even though the reactor core is shutdown, failure to remove the thermal energy from decay heat production can rapidly lead to sufficiently high core temperatures to fail the first barrier. Core Coolant and Heat Sink are the second and third priority CSFs.

The second barrier to fission product release is the RCS pressure boundary.

Although challenges can come from the outside and the inside, only internal threats have been considered in eatablishing priorities for CSFs, because only they can be addressed by the operator. Potential internal threats due to excessive core heat production and inadequate core heat removal are addressed through the Suberiticality, Core Cooling and Heat Sink CSFs. The remaining internal threat to the RCS pressure boundary results from a reactor vessel pressurized thermal shock condition. Such a challenge can result from thermal stresses acting on a radiation embrittled reactor vessel in a low temperature reactor coolant condition. Thus, RCS Integrity is the fourth priority CSF.

The third barrier, containment, is similar to the second barrier in that only internal threats have been considered in establishing priorities for CSFs.

Containment is the fifth priority CSF.

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r a The sixth priority CSF is RCS Inventory. This CSF is actually a subset of the Core Cooling CSF, but has been considered separately in the construction of the status tree and the establishment of priorities for challenges. This CSF addresses situations in which RCS inventory is adequate to satisfy the Core Cooling CSF, but not within nominal operational limits. The challenges associated with the RCS Inventory are the lowest priority of all CSFs.

Thus, the CSFs in order of priority are:

1. Suberiticality

2. Core Cooling

3. Heat Sink

4. Integrity

5. Containment -

6. Inventory Each pass through a status tree produces a single terminus (exit) based on the status of the CSF. Priorities have been established for the termini based on the severity of the challenge. The four status conditions are:

1. Extreme challenge (red) ... solid path.

2. Severe challenge (orange) ... dashed path.

3. Not satisfied (yellow) ... dotted path.

4. Satisfied (green) ... outlined path.

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The color coding and path codes are used to immediately inform the operator that a CSF is in jeopardy, and to indicate the severity of the challenge. The relationship between color, path code, symbol code, and status / response is shown in Figure 1. The action an operator takes in response to a CSF challenge is related to the severity of the challenge. Each terminus symbol which is not green (satisfied) is annotated with the instruction "Go to ... (the appropriate function restoration procedure.)"

FIGURE 1 STATUS TREE PRIORITY INDENTIFICATION

, Color Line Code Symbol Code Status / Response g

Red '~

The critical safety function is under extreme challenge;immediate operator action is required.

Orange EMME .

The critical safety function is under severe i challenge; prompt operator action is required.

Yellow SSSSSS The critical safety function i conditon is off - normal.

Operator action may be

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taken.

Green i i The critical safety function is satisfied. No operator action is needed.

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In summary, the priority of operator action is fixed by the physical Q arrangement of the trees. Each tree contains multiple termini, each of which represents a possible current status of that CSF. Each terminus (and preceding branch) is color coded, reflecting the urgency of the condition, and each also indicates the appropriate procedures to be used. For the entire set of trees, priority of operator action is:

First do any Red (extreme challenge), in tree order Then, if no reds, do Oranges (severe challenges), in tree order Then, if no reds or oranges, do Yellows (not satisfied), in tree order.

For example, a Red condition for Core Cooling is of higher priority than a Red condition for Containment; Core Cooling is the second priority CSF, while Containment is the fifth priority CSF. However, the Red condition for -

Containment is higher priority than any Orange condition, because of the order of colors.

4. Status Tree Usage Status trees provide a mechanism that coordinates event-related recovery and function-related restoration. Start status tree monitoring when the symptoms of an emergency transient result in transition out of Procedure 19000, Reactor Trip and Safety Injection," or when so instructed in that procedure. Monitor the status trees on the Safety Parameter Display System (SPDS). Should the SPDS fail or SPDS accuracy cannot be verified, manually monitor the status trees using Procedure 19200. If transition out of Procedure 19000 is not made due to lack of appropriate symptoms, Procedure 19000 gives explicit instruction-to monitor the status trees while remaining in that procedure (see step 28 of proc. 19000). Once the status trees are being monitored, the following " rules of usage" apply:

1. Monitor the status trees continuously in order of CSF priority.

2. If an extreme challenge is diagnosed, immediately stop optimal recovery and initiate function restoration to restore the CSF under n

v extreme challenge.

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l 4 i SUBCRITICALITY - F-0.1 -

GO TO 19211-1 19 11 1 1

E -

NO N POWER RANGE E LESS THAN 5% 5 GOTO

- ** 19212-1 l

YES e

e INTERMEDIATE NO RANGE SUR INTERMEDIATE NO MORE RANGE SUR NEGATIVE g

l ZERO OR THAN 0.2 DPM YES NEGATIVE YES CSF SAT NO SOURCE RANGE ENERGlZED l YES GO TO I e8 19212-1 l _ e e

l NO SOURCE RANGE SUR ZERO OR I NEGATIVE YES l

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3. If a severe challenge is diagnosed, continue to check the status of all CSFs.- Then stop optimal recovery and initiate function restoration to restore the highest priority CSF under severe challenge.

4. If a not satisfied condition is diagnosed, continue monitoring the status trees. It is the operator's prerogative to continue optimal recovery or to initiate function restoration to restore the affected CSF.

5. If a higher priority challenge is diagnosed during function restoration to address an extreme or severe challenge, terminate on-going response and initiate function restoration to address the higher priority CSF challenge.

-The CSF status trees provide systematic methods to explicitly determine the statuses of CSFs. None require operator action other than monitoring a limited set of plant parameters and comparing them to reference values within the trees.

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Since extreme conditions are postulated to force entry into the FRPs, the values references are appropriate for both normal and adverse containment conditions.

The status tree F-0.1, Suberiticality is the highest priority CSF and, as ,

such is always entered first any time tree monitoring is initiated. This tree will be described as an example of status tree interpretation. It is shown in Figure 2.

Because this tree is monitoring the reactivity state of the core, the parameters being evaluated are those of characterizing neutron (leakage) flux behavior as measured by the excore nuclear instrumentation system (NIS). An adequately shutdown core typically exhibits a randomly fluctuating count rate on the source range instruments. For purpose of this tree, the core is considered shut down (suberiticality satisfied) whenever the level of subcritical multiplication is steady or decreasing in the source range (zero or negative startup rate).

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SUBCRITICALITY - F-0.1 GO TO 19211-1 FIGURE 2 m3aeg GOTO NO POWER RANGE E LESS THAN 5% 3 GOTO YES

- 19212-1 INTERMEDIATE NO ANGE SUR INTERMEDIATE NO RANGE SUR - NEGATIVE ZERO OR THAN 4.2 DPM YES NEGATIVE YES CSF SAT NO SOURCE RANGE ENERGlZED l

..@aizei NO SOURCE RANGE SUR ZERO OR NEGATIVE YES rev. 1 v}

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Suberiticality Blocks l Power range less than 5%: Following a reactor trip nuclear power promptly f drops to only a few percent of nominal, and then decays away to a level I

some 8 decades less. Decay levels resulting from radioactive fission j product decay are never more than a few percent of nominal power and also 1

decrease in time. Safeguards heat removal systems are sized to remove only decay heat and not significant core power. The 5% power level was chosen j because it is clearly readable on the power range meters. Nuclear power l above 5% in a core intended to be shut down is considered an extreme challenge to the fuel clad / matrix barrier and a Red priority is warranted.

l The appropriate Procedure for function restoration is 19211-1, " Response to i

Nuclear Power Generation."

l' Intermediate range SUR zero or negative: At this point, p'we range flux has been determined to be not significant, so no extreme challence exists.

However, a positive startup rate (SUR) in the intermedicea range will shortly lead to power production if operator action is not taken, since no j inherent feedback mechanisms exist below the point of adding heat. A 4

positive SUR is considered a severe challenge to the safety function and an Orange priority is warranted. The appropriate Procedure for function j response is 19211-1, " Response to Nuclear Po.ser Generation."

Source range energized: Use this decision point to determine if further i evaluation should be directed at the source range flux behavior, or back at l the intermediate range channel indication.

l Intermediate range SUR more negativo than -0.2 dpm: Normally following reactor trip, intermediate range flux decays at a constant -0.3 dpm. A i rate decrease less negative than -0.2 dpm, such as +0.2 dpm, represents a not satisfied condition and a Yellow priority is warranted. The I appropriate Procedure for function restoration is 19212-1, " Response to

{ Loss of Core Shutdown." If the rate of decrease is more negative than -0.2 i dpm, then the CSF is satisfied.

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l SUBCRITICALITY - F-0.1 -

GO TO 19211-1 FIGURE 2 m33mg GOTO NO POWER RANGE -

LESS THAN 5% GOTO YES

** 19212-1 O

l .

INTERMEDIATE NO NO ANGE SUR INTERMEDIATE O RANGESUR NEGATIVE ZERO OR THAN 4.2 DPM YES NEGATIVE YES CSF SAT NO SOURCE RANGE ENERGlZED l YES

..@ ano, .

NO SOURCE RANGE SUR ZERO OR NEGATIVE YES P3v VII-1-12

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Source ranse SUR zero or negative: Normally following reactor trip, neutron flux decreases into the source range and stays there. Typically, source range count rate fluctuates, and does not exhibit any sustained increasing trend. Such a trend, as indicated by a positive SUR, is considered a not satisfied condition and a yellow priority is warranted.

The appropriate Procedure for function restoration is 19212-1, Response to Loss of Core Shutdown." If source range SUR is zero or negative the CSF is satisfied.

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} 5. Emergency Event Classification t

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The CSFSTs represent a refinement in the process used to diagnose and respond to plant transients. A similar approach is used for the VEGP event 1

i classification system. Decision trees constructed around a symptomatic versus

! event-specific approach are drawn to guide event class determination.

i

, As discussed in volume I the event classification method used by VEGP consists of two phases, Phase 1 addresses the immediate conditions present and yields a " snapshot" classification which describes the plant condition as it exists. The accidents discussed in this text have been described in terms of i challenges to radiation release barriers. These barriers are assessed by means of the CSFSTs and are logically used in both the first and second phase of event classification.

j During the second phase of event classification, the CSFSTs provide a l systematic means to assess the degree of hazard to the remaining fission product 4

barriers. A review of the CSFSTs employing information from the plant process monitoring equipment ~ enables the TSC staff to advise the emergency director as

! to the magnitude of hazard to the remaining fission product barriers.

l The CSFTs also highlight the most significant hazards by color coding such 1

i that responses by emergency personnel may be ordered appropriately according to their respective priorities.

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O FIGURE 3 FUEL CLADDING INTEGRITY SUSCRIIICALITV V CSFST

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RED OR ORANGE CORE COOLING V ICSFST .

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RED OR ORANGE FROM ATT. 3 H NCAT SINN Y ST

_ ORIf ORIP OR1r FROM ATT. 3 N FUEL CLAODING 15 BREACHE0 OR CHALLENCEO OR4 L DR

1. CVC5 LE TOOWN NONITOR (RC-48000) 0FF SCALE-MIGN. ---1>

2. CONTAINNENT AIN05PNERE PARTICUL ATE, IODINE OR RAD 10 GAS OR NONITORS WILL INCREASE SIGNIFICANTLV (i.e.,, NORE TNAN 1000 TIMES)

REACTOR DUC TO NIGN RA0!OACIIVE CONCENIRAll0NS IN IIC (RC25A2A, COOL 8 ANOANT WHICN NORNALLY LE AMS INTO CONI AINNENT C RESPECTIVELT).

OR _

3. LOS5 OF CL A00 LNG A5 A FISSION PRODUCT S ARRIER RESUL VERY MICH RADIATION INTEM581Y

T5 IN LOCATIONS MNERC REACTOR C00LANI NOREEXISIS,ll INAN 1000.e., TINES) A T 15 PROCESSED, OR LE AMS.

A.

POST-ACCIDENT AND PROVIDES AN SAMPLING VERIFIES TPE EXISTENCE OF M FAILED FUEL 10 !!NES TECHNICAL SPECIFICA!!0N LIMIT 5.INOICAI!ON OF COOL ANT l

I O

VII-1-14 rev. 1

i

/

/9 An Alert, Site Area, or General Emergency is determined by first evaluating

)

the status of the six CSFs. The CSFST results are used in conjunction with plant radiation monitor readings to evaluate the integrity of the three barriers.

For example, the Suberiticality CSFST is used as one test of challenge to the Fuel Cladding integrity as shown in Figure 3. These decision trees are part of the Emergency Plan Classification procedure 91001. Once the number of challenged barriers is determined then the emergency class is determined; one barrier = ALERT; two barriers = SITE AREA; three barriers = GENERAL.

Should the emergency response staff and management be unable to obtain or analyze barrier status during operation or transient response, it is necessary that affected barriers be assumed to be challenged. Conditions that could have this effect might include loss of de power, extended loss of ac power, control room evacuation, loss of control of the plant, etc. Other events could be postulated that might make it impossible to monitor a single barrier or combinations of two barriers. Procedures specify that if the status of any barrier cannot be determined it is considered challenged and the emergency classification is to be made per 91001.

Natural phenomena and other hazards such as tornado, floods, airline crashes, etc., are classified per 91001 as Notification of Unusual Events. It is recognized that these events could have a wide spectrum of effects on the plant. An airline crash, depending on the specific conditions, might warrant classification as an Alert, Site Area, or even General Emergency, depending on the size, location, and compounding events (such as a fire). Should any of these events have, or have the patential for, direct or indirect effects on a barrier, the classification will be made based on 91001.

t l

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! VII-1-15 rev. I r

l l \

O INTENTIONAL BLANK O

O VII-1-16 rev. 1

(

t j (A) SELF-CHECK

1. The Function Restoration Procedures are:

a. similar in format to the Optimal Recovery Procedures

b. entered by using CSF status trees

c. used only when Optimal Recovery Procedures are unable to maintain the CSFs

d. all of the above.

2. List the six critical safety functions in order of priority.

3. Put the following CSF paths in order of priority. -

a. orange on core cooling

b. red on heat sink

c. orange on suberiticality

d. yellow on inventory

e. red on integrity

f. green on containment

4. The reactor was tripped on Lo-Lo steam generator level h hour ago. The operator is in procedure 19001, reactor trip recovery when the source range read a start-up rate of a fairly solid .3 dpm. What other procedure, if any, should the operator go to? The intermediate range is off-scale low.

I L

l l

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Vll-1-17 rev. 1

s

ANSWERS TO SELF-CHECK

1. d

'2. subcriticality, core cooling, heat sink, integrity, containment, inventory

3. b,e,c,a,d,f

4. He could go to procedure 19212, Respond to loss of suberiticality.

However, this is a yellow condition. He may remain in procedure 19001 until it is convenient to go to the steps of the FRP or until conditions worsen such as the intermediate range coming on scale and reading the same startup rate.

O O

VII-1-18 rev. 1

</

CHAPTER 4 INADEQUATE CORE COOLING After reading this chapter you will be able to:

1. Describe the inadequate core cooling' event that took place at Three Mile Island Unit 2 to include the following:

a. The initiating event

b. Two instances where operators did not follow the applicable procedure

c. The cause of the inadequate core cooling condition

d. Two actions that if taken could have prevented the inadequate core cooling condition

e. The cause of containment pressure and aux. building radiation problems

f. A comparison between the actual implementation of the TMI emergency, plan and how the Vogtle plan would be implemented.

g. Three or more changes that have been implemented at Vogtle because of ;

studies regarding the TMI-2 accident.

h. How the core condition was eventually stabilized.

2. Explain why the secondary system was ineffective in removing heat from the THI-2 core during this event.

3. Explain why the high pressure safety injection was ineffective in removing heat from the TMI-2 core.

4. Describe the effect of RCP operation during the TMI-2 event.

1 I

j 5. Describe the effect on nuclear instrumentation of reduced water level in

the core.

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i iO VII-4-1 rev. I

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Figure 1 THREE MILE ISLAND CourA - EuTetoo a SAFETY M VE ATMOSPMRIC e-VENT VALVE ELECTROMAnC LV

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VII-4-2 rev. 1

j

s Three Mile Island Unit Two is a B&W two-loop PWR with once-through steam generators. There are two reactor coolant pumps in each loop. Otherwise the system is much like that at Vogtle. One group

studying the event concluded that operational errors appeared to be a significant factor in the accident.

They claimed that that, overall, actions taken by the operators during the first ,

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months of the event worsened the situation. The core was damaged because adequate cooling was not provided to remove decay heat. The operators

, apparently did not understand that they were not providing adequate cooling.

They were confused by the phenomenon of high pressurizer level and low reactor coolant system pressure which they had not experienced before. The procedures used did not recognize that this phenomenon could occur and therefore did not provide adequate guidance.

1. Basic TMI-2 Event Log The reactor was at 97% of 2772 MWt power with the B&W Integrated Control System in full automatic. RCS pressure was normal at 2155 and boron

('~'

concentration was at 1030 ppa. Pressurizer relief discharge header temperatures

( were high enough to give periodic high alarms (70' above maximum allowable operating temperature). See Figure 1 for a general la/out.

! For approximately 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months prior to 0400 March 28, 1979 TMI-2 operating l personnel had been attempting to transfer spent resin from an isolated condensate polisher unit to the resin regeneration system. The resin was apparently clogged in the outlet of the polisher and the operators were injecting a water and air mixture into the polisher to break up the clogged resin. The air system is isolated only by a check valve while performing the operation and water can leak into the service and instrument air system through the check valve. Water entering the instrument air system can restrict air flow l 1

The conclusion was drawn by Jasper L. Tew of the Technical Assessment Task Force. The event summary used here is taken from that group's " Technical Staff O Analysis Report on Summary Sequence of Events," published October, 1979.

VII-4-3 rev. 1

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VII-4-4 rev. 1 l

ID to the control valves for the polisher outlet valves causing them to shut. This is the most probeble cause of the outlet valve closure and loss of feedwater.

A few seconds before 0400:37 March 28, 1979 the condensate polisher outlet valves shut causing the Condensate Booster Pumps to trip due to low suction pressure which caused the Main Feedwater Pumps to trip from low suction pressure at 0400:37. (Unit 2 Emergency procedures 2202.2-2 " Loss of Main Feedwater to Both OTSG's, Section 2.B states in Manual Action 1. "If loss of FW is due to loss of both feed pumps: a. Trip the Reactor." The operator did not manually trip the reactor, although it tripped on high pressure 8 seconds later.

0400:37- Both Main Feed Pumps trip (figure 1) automatically tripping the (00:00:00) main turbine. Three Emergency Feed Pumps start automatically.

With the reactor still operating the primary coolant began to heat up because the turbine was no longer extracting heat from^

the system.

0400:40- Reactor coolant system pressure increases to 2255 psig (figure 2)

(00:00:03) opening the PORV as designed. The pressure in the Reactor Coolant drain tank began to increase.

0400:45- Pressure reached 2355 psig and the reactor tripped on a high (00:00:08) pressure signal as designed (figure 2). After the reactor tripped the plant began to cool down due to heat rejection through the steam generator relief valves, which had lif ted, and the turbine bypass valve, decreasing the plant pressure (figure 3).

0400:49- Reactor coolant system pressure was reduced to 2205 psig, where (00:00:12) the PORV should have closed (the PORV did not close)(figure 2).

The expected insurge of reactor coolant into the pressurizer peaked at about 260 inches and began to decrease. The operators, as specified in the operating procedures, stopped letdown flow and started another makeup pump (IA) to c~npensate for the expected pressurizer out-surge as the plant continued to cooldown.

O VII-4-5 rev. 1

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VII-4-6 rev. 1

's 0401:07 PORV and one pressurizer safety valve high outlet temperature (00:00:30) alarm alarm were received (temperatures were 239.5*F and 203.5'F respectively). The operators were aware that the PORV had lifted but thought the valve had closed because the valve position 4

indicator light was extinguished. This indicator only indicated that power was applied to the pilot solenoid for the PORV and did

, not indicate valve stem position. The high temperature was assumed to be a result of the temporary opening of the PORV and

an existing leak in either the PORV or a code safety valve.

j Although the TMI-2 procedures indicate the PORV will open on a severe transient (Abnormal Procedure 2203-2.2 Turbine Trip, j Section 2.0 automatic action A.3 states " Pressurizer Pilot Operated Relief Valve Open"). None of the operating procedures required the operators to make positive checks to ensure the PORV and closed after an increasing pressure transient that approached or exceeded the PORV setpoint.

0401:07- Both steam generators reached the water level control setpoint of (00:00:30) 30 inches, (figure 4), where the Emergency Feedwater Control

, through Valves EF-V-11 A and 11 B opened. No water was added to the 0401:10- steam generators because the downstream block valves EF-V-12A and (00:00:33) 12B were closed. The operators were not aware that the block valves were closed.

0401:25- With two makeup pumps (1A and IB) running the rate of pressurizer (00:00:48) level decrease was reduced and af ter reaching a minimum of about 160 inches it began to increase (figure 2).

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0401:37- A second pressurizer safety valve high outlet temperature alarm (00:01:00) was received. The indicated temperature was 294.5*F. The safety valve outlet temperature increase was probably due to the hot reactor coolant being discharged through the PORV which increased

, the temperature of the safety valve outlet piping.

l 0402:22- Both steam generators boiled dry and effective heat transfer from

, (00:01:45) the reactor coolant system to the secondary system stopped.

1 (Figures 3 and 4).

0402:38- The open PORV continued to reduce a reactor coolant pressure to i

(00:02:01) 1640 psig where Engineered Safeguards Features (ESF) for High Pressure Injection (HPI) activated. ESF Actuation automatically stopped makeup pump 1B, started makeup pump 1C (makeup pump 1A was started previously at 0400:49) and fully opened the makeup valves providing a total injection flow rate of about 1000 gpm to the reactor coolant system.

0403:50- The HPI portion of the Engineered Safety Features was bypassed.

(00:03:13) (TMI-2 Emergency Procedure 2202-1.3 Loss of Reactor Coolant /

pressure requires the operator to " Bypass" the Engineered Safety Features and throttle the valves to prevent pump runout.) Note:

" Bypass" only returns HPI to manual control-this action does not change any valve settings (see action at 0405:15).

O VII-4-7 rev. 1

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VII-4-8 rev. I

a o p 0403:50- Reactor coolant drain tank relief lifts at about 122 psig.

V (00:03:13) 0404:03- Reactor coolant drain tank high temperature alarm occurred.

(00:03:26) 0404:05- Pressurizer high coolant alarm occurred (260 inches) (figure 2).

(00:03:28) Note: Over 100 alarms occurred during the first few minutes of the accident.

Note: TMI-2 operating procedure 2103-1.3 " Pressurizer Operation" - requires the operator to maintain level between 45 and 385 inches. In addition, it states, "the pressurizer /RC System must not be filled with coolant to solid conditions (400 inches) at any time except as required for system Hydrostatic tests."

0405:15- Operator stopped makeup pump IC and throttled the high pressure (00:04:38) injection valves. (Operators had previously bypassed HPI at 0403:50)

Note: TMI-2 Alarm Procedure 2201-13 Alarm 13.A2 (Engineered Safeguards Features Actuation) states that the cause for the ESFA alarm actuation (other than test or channel failure) is "LOCA" and requires followup action with TMI-2 Emergency Procedure 2202.1-3 " Loss of Reactor Coolant / Reactor System Pressure."

Prior to Automatic initiation of HPI at about 0402 the plant response to the turbine trip appeared to be normal and the operators probably had no reason to suspect a Reactor Coolant system leak.

After the automatic initiation of HPI the rapidly decreasing reactor coolant pressure with constant reactor coolant system temperature was an unambiguous symptom that coolant was leaking from the system.

Subsequent to overriding and reducing the HPI flow the operators increased letdown flow to its maximum value (about 160 gpm) in response to high pressurizer level, which further exacerbated the. loss of coolant.

buringtheperiodHPIwasactivatedfromabouttwominutes to about four minutes the reactimeter traces indicate there was no net heatup of the reactor coolant system (figure 6) indicating that the plant had achieved a heat rejection rate equal to the decay heat and reactor coolant pump heat input.

0406:07- The indicated reactor coolant system hot leg temperature and (00:05:30) pressure reached saturation conditions (Figure 5). As steam bubbles formed in the Reactor Coolant System they took control of j the plant pressure increasing the pressurizer level as the bubble expanded.

l VII-4-9 rev. I

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TMI 2 Loss of Coolant Accident 3/28179 N Reuctor Buildie g Temperaturn and Pressure *

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. s 0406:28- Pressurizer level indication was off scale high (greater than 400 C, (00:05:51) inches)(Figure 2) 0408:06- Reactor building sump pump 2A started automatically at a water (00:07:29) level of 38 inches.

0408:37- The operators discovered that the Emergency Feedwater Block (00:08:00) Valves EF-V-12A and 12B were shut and began opening the valves.

Addition of the cold feedwater to the steam generators sub-cooled the reactor coolant system over the next 15 minutes and the system pressure followed saturation temperature (figures 5 and 6). Since the pressurizer could not regain control of plant pressure, due to flow out of the open PORV, it appears that the eight minute delay in providing feedwater to the steam generators did not materially affect the outcome of the accident.

0410:56- Reactor building sump pump 2B started automatically at a water (00:10:19) level of 53 inches.

0411:25- Reactor building sump high level alarm occurred. This alarm is (00:10:48) one of the symptoms of a loss of coolant shown in TMI-2 Procedure 2202-1.3 Loss of Reactor Coolant / Reactor System Pressure.

0415:25- The Reactor Coolant Drain Tank (RCDT) rupture disc failed as (00:14:48) designed when pressure increased to about 191 psi. The Reactor i Building ambient temperature began to increase rapidly as a result of released steam (figure 7).

0415:27- Reactor Coolant pump alarms occurred. Reactor Coolant System (00:14:50) pressure was about 1275 psig and the temperature was about 570'F at this time (figures 5 and 6). These conditions are very close to the lower limit for operating the reactor coolant pumps. The pumps were apparently vibrating due to the voids being formed in the reactor coolant.

0420:37- The out-of-core Neutron Instrument Flux levels on the source (00:20:00) range began to increase. (figure 8). The Reactor Coolant System t contained significant steam voids at this time and the source

, range nuclear instruments located outside the reactor vessel and measuring the radiation levels as attenuated by any water in the i reactor, were responding to this decrease in density.

0423:21- The steam generator A water level reached 30 inches and could (00:22:44) have been used for heat transfer. However, the turbine bypass valve control was set to automatically control the steam generator's pressure at a value about equal to saturation pressure of the primary coolant. Therefore, the valve was not effectively used to remove heat from the Reactor Coolant System.

0425:35- The operators requested PORV outlet temperature. The PORV outlet (00:24:58) temperature was 285.4*F.

VII-4-11 rev. 1 J

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Trace During the TMI-2 Incident l

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VII-4-12 l

0427:03-

. O (00:26:26)

Plant status information requested by the operator was printed out by the utility typewriter:

i Reactor coolant loop A hot leg temperature- 551.9'F Reactor coolant loop B hot leg temperature- 550.9*F Reactor coolant loop A cold leg temperature- 541.1*F Reactor coolant loop A cold leg temperature- 547.0*F Reactor coolant loop B cold leg temperature- 547.0*F Reactor coolant loop B cold leg temperature- 546.8'F Reactor coolant loop A pressure- 1040 psig Reactor coolant loop B pressure- 1043 psig 0430:00- Reactor building temperature and pressure were increasing (00:29:33) rapidly. The operators responded by starting the reactor building emergency cooling booster pumps and switching all 5 reactor building cooling fans to high speed. The rate of pressure increase in the reactor building slowed down as a result of these actions (figure 7).

0433:13- In-core thermocouple 10 R read greater than 700*F which is the (00:32:36) highest reading the computer software was programmed to record.

4 The significance of this reading is still not understood since the core was covered and being cooled at this time.

0438:47- Both Reactor Building sump pumps were stopped. Based on the run

-- (00:38:10) time and the pumping capacity of these pumps. They could have to transferred as much as 8100 gallons of water out of the reactor 0438:48 building. The pumps were apparently aligned to discharge to the (00:38:11) auxiliary building sump tank (which had a failed rupture disc) instead of the miscellaneous waste hold up tank (the level of 4 this tank did not change during the March 28, 1979 operations).

The sump pump, by procedural guidance, could be aligned to either tank, i

0440:37- The source range out-of-core nuclear instruments continued to (00:40:00) show an increasing count rate (Figure 8) due to the continuing l through decrease in reactor coolant density.

0500:37 (01:00:00)

Increases in the Reactor Building background radiation level were shown on the Reactor Intermediate Closed Cooling System Letdown Monitor (1C-R-1092).

Reactor coolant flow had decreased from a normal rate of about 69 million pounds per hour to less than 50 million pounds per hour.

0514:00- Reactor coolant pumps 1B and 2B were stopped because of the (01:13:23) vibration readings and because the plant conditions (temperature (approx) and pressure) were outside the specified range for pump operation.

O VII-4-13 rev. 1

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0514:00- The out-of-core nuclear instruments, both source and intermediate

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s_ , (01:15:23) range, increased their readings as the reactor coolant density to continued to decrease (Figure 8).

0541:00 (01:40:23)

The PORV discharge line temperature remained at about 283*F.

Reactor coolant flow continued to decrease. There were some momentary indications of steam flow from steam generator A. The feed flow rate to steam generator B was increased.

Steam generator A boiled dry and a few minutes later feedwater flow to steam generator A was increased, and was apparently used effectively for a few minutes to remove heat from reactor coolant loop A (Figures 9).

0515:00- Intermediate closed Cooling System Radiation Monitor (1C-R-1092)

(01:44:23) began increasing from 3500 counts per minute. The monitor through reached its alarm point of 5000 counts per minute at 0518.

0518:00 (01:17:25) 0518:00- The reactor building air particulate monitor HP-R-227 (P) reached (01:17:23) its alarm point of 50,000 counts per minute. Due to the fact that the reactor coolant system had been below the required pressure conditions for fuel rod compression for some time and f- the core temperature was increasing above normal (at least one f -

core exit thermocouple was reading off-scale high at 0433:03) it

\; is inferred that these radiation monitor readings indicate that fuel cladding was being ruptured mechanically by internal pressure. The ruptures at this time were probably small but did allow some of the fission gases accumulated in the fuel-to-cladding gap to escape into the reactor coolant system.

0541:22- Reactor coolant pumps 1A and 2A were stopped. Forced cooling of (01:40:45) the core was terminated.

(approx)

The source and intermediate range nuclear instruments decreased significantly as the cooler water being held up in coolant loop A hot leg fell back into the core, temporarily increasing the coolant density in the core.

Reactor coolant loop A hot and cold leg temperature both decreased for about 12 minutes. Then the hot leg tenperatures

' indicated in the control room began to rise rapidly going off scale high (greater than 620*F) within 38 minutes. The loop A cold leg temperature continued to decrease slowly over the next hour (Figure 9). Reactor coolant loop B hot leg temperature continued to decrease until about 0605 at which time it began to increase rapidly and the indicated temperature in the control room went off scale high (greater than 620*F) within about 25 minutes. The loop B cold leg temperature continued to decrease O (Figure 9).

VII-4-15 rev. 1 i

Data, which was not available to the operators in the control room, indicates that superheated conditions existed in both the A and B loop hot legs from about 0615 to about 1430.

About 2 minutes after the reactor coolant pumps stopped, the out-of-core nuclear instruments, both source and intermediate range, began to increase rapidly, indicating boil off of the reactor vessel inventory.

Reactor system pressure at the time the A loop reactor coolant pumps were stopped was about 1000 psi and it continued to decrease rapidly.

0602:00- Analysis of a reactor coolant sample showed the gross beta-gamma (02:01:23) activity to be 4 micro curies per milliliter which is about 10 (approx) times the normal expected reading. This sample is a further indication that some mechanical damage to the fuel cladding had been sustained and fission products had been released into the reactor coolant from the fuel to cladding gap space.

0615:00- The self powered neutron detectors (SPND's) installed in the core (02:14:23) began responding to high temperatures indicating that the water (approx) level in the reactor vessel was below the top of the active core.

The response of the SPND's is consistent with the sharp rise in the reactor coolant loop hot leg temperatures which started at about 0615. (Figure 9) 0622:37- The PORV Block Valve was shut. Reactor system pressure began to (02:22:00) increase and the reactor building pressure began to decrease (Figure 7) indicating that the PORV was the source of coolant leakage from the system.

0624:00- The Reactor Building Air Particulate Sample Monitor (H0-R-227 (02:23:23) (P)) reached its alarm point of 50,000 counts per minute for the j second time.

0626:00- The area monitor in the reactor building on the 347 ft. level j (02:25:23) reachec' its alarm point of 50 mr/hr.

(approx) l 0630:00- General radiation levels in the auxiliary building increased and l (02:24:23) ranged from about 10 mr/hr to more than 5 R/hr at the to purification valve room door.

0700:00-(03:59:23) 0643:00- Analysis of a reactor coolant sample taken at this time showed (02:42:23) gross beta-gamma activity of 140 microcuries per milliliter. The area monitor for Unit One Sample Room (RM-G3), which contained the Unit 2 sample lines, reached its alarm set point of 2.5 mr/hr.

0648:00- Unit 1 Hot Machine Shop area monitor RM-G4 reached the alarm (02:47:23) set point of 2.5 mr/hr. A survey of the reactor coolant sample VII-4-16 rev. 1

. e line running through this area read 1.5 R/hr.

Summary: The reactor vessel inventory continued to boil off after the reactor coolant pumps were stopped at 0541 and by 0615 there is evidence of super heated steam in the coolant loop hot legs. Subsequent to closing the PORV block valve at about 0622 no significant heat was removed from the core until the block valve was again opened at 0712.

By 0700 the temperature of the het leg was at least 750'F in the B loop. The A loop temperature was about 775'F.

No significant change in make up flow to the loops is evident until almost 0720, one hour after the PORV was closed.

Response of the SPND's at 0648 indicates that the reactor vessel water level may have been 8 to 9 feet below the top of the active core. Radiation monitoring instruments and reactor coolant sample analyses had previously indicated that some mechanical damage (ruptured cladding from internal pressure) to the core was

occurring about 0602. .

By 0650 the high radiation levels indicated by the radiation monitoring instruments indicate that severe core damage was taking place.

I It is unclear why the operators, or engineers and supervisors who were V present, did not immediately start high pressure injection when the PORV block valve was closed and plant pressure began to increase indicating that the PORV was the source of the leak.

0648:23- The operators managed to get the condensate system to function (02:47:50) automatically by 0650. The difficulty with this system was found to to be a broken air line which supplies operating air to the valve 0655:37- air operator. The valve was then opened manually and the system (02:55:00) began to control the condenser hot well level automatically. The operating air line was apparently broken during the transient since the hot well level control was functioning normally prior to the turbine trip.

Af ter jumpering interlocks in the control circuits Reactor Coolant Pump 2B was started at 0654:46 and allowed to run for 19 minutes.

The reactor out-of-core nuclear instruments showed a sharp decrease in level as the colder water trapped in the reactor

]

coolant loop cold legs and the B steam generator was transferred 4

into the reactor vessel.

! Radiation level increases and alarms in several areas of the plant including Reactor Building Atmospheric Sample monitors and the Hot Machine Shop Area Radiation Monitor led the Shift O Supervisor and the Unit 2 Superintendent, Technical Support to decide to declare a Site Emergency at approximately 0656.

VII-4-17 rev. 1

a .

1 D. Declaration of Emergency and Stabilization of the Plant (Time 0656 to approximately 1950) 0656:00- A Site Emergency was declared.

(02:55:23)

(approx) 0656:00- Radiation levels continued to increase in the Reactor Building, (02:55:23) the Auxiliary Building and in the Fuel Handling Building.

to 0700:00-0705:00- The TMI Station Superintendent arrived in the Unit 2 Control Room (03:04:23) and assumed the role of Emergency Director.

0705:00- Radiation levels throughout the plant continued to increase. The (03:04:25) PORV block Valve was opened at 0713 and closed at 0717. High to pressure injection was manually initiated at 0720.

0724:00 (03:23:23) 0724:00- A General Emergency was declared by the TMI Station (03:23:23)

Superintendent. The radiation monitor in the dome of the Reactor Building had reached a reading of 8R/5r which is a specified condition in the TMI-2 Emergency Plan that requires declaration of a General Emergency.

0724:00- The operation of plant systems and components during this period (03:23:23) are summarized as follows:

to 1950:00- A. From 0712 to 1108 a combination of high pressure injection (15:49:23) flow into the loop and flow out of the PORV was the principal means of cooling the core. Based on the out-of-core nuclear instrument readings the reactor vessel inventory appears to have been recovered to a level above the active core by about 1100.

B. Starting at about 1140 a prolonged depressurization of the reactor coolant system began, with a relatively low high pressure injection flow, which may have resulted in some core uncovery as indicated by the out-of-core nuclear instruments.

C. Even though substantial quantities of steam were discharged into the reactor building through the open PORV until it was isolated at 0622, the operators precluded a reactor building pressure rise to 4 psig (the actuation set point for Engineered Safeguards Features Actuation for reactor building isolation) by manipulation of the reactor building ventilation system. However, a prolonged discharge out of the PORV which started at 0740 caused the first reactor building isolation to occur at 0756.

VII-4-18 rev. 1

i I

1

. , 1

'~' D. A pressure spike of at least 28 psig occurred in the reactor ,

building at about 1350. The pressure spike was apparently the result of a hydrogen burn caused by flammable concentrations of hydrogen which were generated by the zirconium / water reactions that took place during the times the core was uncovered and overheated.

E. Continued operation of the letdown system and other systems such as the waste gas decay system after the core was damaged contributed significantly to the escape of radiation to the environment. A leak in the waste gas header system was an important factor in this.

F. Plant repressurization was started at 1508 and at 1950 forced circulation was reestablished when Reactor Coolant Pump 1A was started and run continuously placing the reactor in a stable cooling mode.

2. Decay Heat Removal During the TMI-2 Accident t

During the period of time between trip of the main turbine and the reactor trip at 8 seconds into the accident some of the heat generated by the reactor

! operating at power was removed by (1) the turbine bypass line discharging to the I

() condenser (2) the main steam safety valves in the secondary loop opening and discharging steam to the atmosphere and (3) the (PORV) on the pressurizer which opened at about 3 seconds into the transient.

After the reactor tripped 8 seconds into the transient, the steam generators began to cooldown the reactor coolant and the main steam relief valves closed. The turbine bypass system continued to remove heat until the i steam generators boiled dry at about 1 minute 45 seconds into the accident. The steam generators boiled dry because the emergency feedwater block valves were left shut improperly, The open PORV continued to reject heat and mass from the reactor coolant system until flow through it was stopped after about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes by closing the block valves down stream of the relief valve.

From 1 minute 45 seconds to 2 minutes the discharge of the PORV, together l with modest makeup flow, probably on the order of 300 gpm, was the heat removal method and was clearly not sufficient as the plant continued to heat up (Figure C::)

VII-4-19 rev. 1

9 3 Iligh Pressure Injecticn was initiated automatically at about 2 minutes, delivering on the order of 1000 gallons per minute flow to the reactor coolant system. This high injection flow rate in conjunction with the continuing flow out of the open relief valve removed an amount of heat equal to the decay heat at that time. Figure 3 shows that the reactor coolant temperature rise stopped af ter $1gh Pressure Injection commenced. It then leveled off and was essentially constant when high pressure injection was terminated at about 4 minutes and 38 seconds.

Terminating High Pressure Injection at 4 minutes and 38 seconds upset the equilibrium of this decay heat removal mode and the reactor coolant system started to heat up (Figure 3). The plant continued to heat up until feedwater was added to the steam generators after opening the emergency feedwater block valves, at about 8 minutes. Between 8 minutes and 30 minutes the reactor coolant system temperature was reduced from about 597'F to 550*F by use of the steam generators (Figure 9) and the open PORV. During this period the makeup flow rate was very low, probably less than 100 gpm. Letdown flow of about 160 gpilons per minute was started at about 4 minutes 38 seconds (to reduce pressurizer level) further increasing the rate of coolant loss from the reactor coolant system.

The reactor coolant system pressure continued to decrease due to flow out the open relief valve while the loop was being subcooled and reached 1100 psig at 18 minutes. When feedwater was rapidly added to the steam generators i

starting at about 18 minutes pressure was reduced to about 1050 psig but returned 1100 psig with a few minutes as the added feedwater heated up.

A fairly constant heat balance was maintained from about 18 minutes until the B loop reactor coolant pumps were stopped. During this period the A and B steam generators, together with the boil of f through the open relief valve, were removing essentially all the decay heat generated by the core and the heat input of the reactor coolant pumps (Figure 9).

The B loop reactor coolant pumps were stopped at I hour and 14 minutes and both the A & B loop temperatures began to increase (Figure 9) indicating that VII-4-20 rev. 1 i

4 the heat balance was upset by the sharply reduced forced circulation in the O' reactor coolant system.

The atmospheric dump valves were opened on the secondary side of the A Steam Generator at about I hour and 31 minutes as evidenced by sharply decreasing steam generator pressure. The combination of increased feed flow and possible opening of the B loop atmospheric dump valves, at I hour and 14 minutes appears to account for the rapidly decreasing B steam generator pressure. It is concluded that there was flow in both loops until about I hour and 31 minutes when the B steam generator appears to have been isolated and the A steam generator boiled dry. The conclusion is based on the fact that the average 4emperatures and the differential temperature across the steam generators-were essentially equal in both loops. For average loop temperatures and delta temperatures across the steam generators to be equal each generator must have .

been removing equal amounts of heat.

The temperature in both loops had been reduced to about 530*F just prior to i

(' stopping the A loop reactor coolant pumps at I hour and 41 minutes. The reactor coolant system pressure followed the decrease in saturation temperature indicating that this heat removal mode was capable of subcooling the system (Figure 9).

When the A loop reactor coolant pumps were stopped, followed closely by a rapid feeding of the A steam generator at about I hour and 42 minutes, the A and B Icop hot leg temperatures began to diverge.

The B loop hot leg temperatures began to increase and then stabilized for a l

few minutes. The A loop hot leg temperature decreased until I hour and 52 minutes at which time it began to increase rapidly from 530*F and was greater than 800* by about 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The B loop temperature began to decrease at about 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 55 minutes and reached 620*F at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes. It then began l

to increase rapidly and reached about 790*F by about 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 15 minutes (Figure 9).

i O

l VII-4-21 rev. 1

_ _ _ - _ _ - _ . _ _ _ _ ~__ _ . . _ . _ _ . _ . _ ...____ ._ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ ___ _ -_ _ _ _ _ _ _

The extremely rapid beat up of the reactor coolant loop hot legs after the last reactor coolant pumps were stopped indicates that any fluid that had been in the loops collapsed leaving only steam in the hot legs which achieved superheated conditions within a few minutes. In order for superheated steam to be present a heat source with a temperature greater than saturation temperature had to be available to heat the steam, thus it is concluded that a portion of the active core was exposed (uncovered) between I hour and 55 minutes and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

After flow in the loops was stopped at I hour and 40 minutes boiling off the water inventory in the reactor vessal was apparently an effective heat removal method until loss of mass began to expose (uncover) the active core.

Steam flow with its low heat transfer coefficient, apparently was inadequate to remove the heat generated in the exposed fuel; since the fuel began a rapid heat up during this period. This mode continued until the PORV was shut at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes.

While it is not possible to show the' precise water level in the core, from I hour and 30 minutes to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes, the level can be inferred by use of data from the out-of-core neutron detectors and the response of the Self Powered Neutron Detectors (SPND'S) located at various elevations in the core as indicators of what parts of the core were covered at various time. These instruments indicate that the water level at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes could have been as low as 8 to 9 feet below the top of the active core.

l No effective cooling occurred between 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 22 minutes (Block valve shut) and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 19 minutes when high pressure injection was started and maintained for 18 minutes. This was a period that produced large quantities of hydrogen from the zirconium-water reaction at high temperature and significant damage to the core.

l

( Sustained High Pressure Injection Flow was started at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 26 minutes and maintained until about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and 4 minutes. The core appears to have been recovered by about 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 30 minutes as indicated by the out-of-core nuclear instrument readings. Heat removal during this period was by l (1) letdown flow which is believed to have been close to maximum level of 160 VII-4-22 rev. 1

gpm, (2) periodic opening of the PORV (open 5 minutes starting at 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 12 minutes, open 98 minutes starting at 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 40 minutes, periodic cycling (about 30 cycles) between 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and 43 minutes and 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and 38 minutes),

(3) and some steam refluxing to the steam generators.

Reactor coolant pump 2B was restarted at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 54 minutes and some of the water trapped in the steam generator B and loop B cold legs was returned to the reactor vessel.

Temperature measurements from the core exit thermocouples readings obtained between 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and 30 minutes show at least 9 temperatures above 2000*F with the highest being 2580*F and several above 1000*F.

During this period high pressure injection flow into the reactor vessel and out through the loop A hog leg then via the surge line out the pressurizer when the PORV is open is the cooling flow path.

The continuous depressurization of the reactor coolant system, in an attempt to dump the core flood tanks (accumulators) into the system, which was started at about 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and 38 minutes may have uncovered the core again as evidenced by the increasing out-of-core neutron instruments. The High Pressure Injection flow during this period was very low and cooling was accomplished by boil off of the core inventory.

Since the conditions of this depressurization and possible uncovery is similar to the earlier uncovery it must be assumed that conditions for zirconium-water reactions also existed during this period. The continuous depressurization allowed the hydrogen generated by the zirconium water reaction to be expelled into the reactor building. This new hydrogen combined with any hydrogen generated and expelled during the earlier uncovery reached a level sufficient to support combustion. A 28 lb pressure spike shown on the reactor building pressure instrument at about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and 50 minutes probably indicates a burn of the hydrogen in the reactor building atmosphere (Figure 7).

O O

VII-4-23 rev. 1

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VII-4-24 rev. 1

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() The depressurization attempt was terminated at 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months and 8 minutes with the closure of the PORV Block Valve. For the next 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 35 minutes there was very little heat removed from the system., The PORV Block Valve was opened for two periods of about 10 minutes each, there was no forced circulation flow and high pressure injection flow was sporadic until about 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months and 23 minutes.

1 It can be inferred from the actions during this period that substantial continued core heat up occurred at least until the system was finally filled and reactor coolant pump 1A was started and remained running at 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months and 50 minutes (Figure 10).

After the 1A main coolant pump was started a slow cooling trend was later established and the plant was placed in the natural (no pumps running) ,

4 circulation mode of cooling on 27 April 1979.

i THE RESULTS OF INADEQUATE CORE COOLING In the years since the accident occurred, the damage to the core has been

(} .

carefully evaluated. Due to the high radiation of the core materials, this inspection has progressed mainly by using remotely operated devices. The full extent and type of damage will take many years to assess. *0n March 11, 1986 i the manager of the Department of Energy's TMI-2 accident evaluation program met with the NRC to report conclusions drawn as of that date. The manager, Don McPherson, estimated that as much as 70% of the core may have melted but determining exactly how much of the core became molten is difficult because of the mixture of melted metals DOE found in the containment vessel. DOE is certain that there has been some molten UO2 but because there is a mixture of UO2 and zirconium dioxide, as well as a mixture of those two in solution, figuring out just how much of the core was damaged is more complex.

Based on its findings thus far, DOE has also determined, among other things, that there is a 30% void in the TMI-2 core as well as localized regions of oxidized and molten stainless steel. The top of the existing core is made up of " prior molten fuel" and " fully oxidized zircaloy," McPherson said. Below that is a hard layer of rubble, about 1.75 meters thick, composed of some unreconstructed fuel, cinders, and previously molten oxides. This hard layer

(

INSIDE NRC.; March 31, 1986 VII-4-25 rev. 1

sits above a portion of the core made up of as yet unknown matter. Below the g

"relatively undamaged" core support assembly sits "from 10% to 20% of the original core laying in the lower plenum."

"A slurry of the core mixture flowed down from within the core into the lower plenum through the holes in the core support structure," McPherson said.

The " relocated" core material could not be cooled by the 2 feet of water then remaining at the bottom of the TMI-2 core, he added. "At that point... it grew in magnitude as far as the liquified portion was concerned and either melted itself through the crust (that had formed) or dissolved itself through any UO2 that was in the lower part of the crust."

The question of whether the 2 feet of water then remaining in the TMI-2 core saved the plant from a complete meltdown is as yet undetermined but there is some speculation that it played a crucial, positive role. NRC commissioner, Frederick Bernthal said, "The message...at least to me, is how much a saving .

grace is even a small amount of water that remains in the bottom of the vessel.

I think it's fair to say this was a core melt accident. There has been an inclination not to call it that, but it is quite clear now that it was."

Bernthal then asked if it was fair to say that the 2 feet of core remaining underwater and perhaps the bottom (support) structures remaining underwater as well formed the boundary between the next step of " core on the floor". The DOE representative replied that the data available was sufficient to confirm or refute that conclusion. Both Bernthal and DOE acting assistant secretary for nuclear energy expressed the commitment to see that the opportunity to get that data never occurs. This commitment can only be implemented through trained operators and effective emergency operating procedures.

l l

I l

l l

l VII-4-26 rev. 1

l i

a

l CHAPTER 5: Respond to Inadequate Core Cooling O

PERFORMANCE OBJECTIVE Given that either

1. The core exit thermocouple temperatures are greater than 1200* F, or

2. RCS subcooling is less than 31' F, no RCPs are running, the core exit thermocouple temperatures are greater than 700* F, and the RVLIS full range indication is less than 39 percent, restore core cooling.

Adequate core cooling must be established as defined by steps 5-7, step 16, or step 23 of Procedure 19221.

ENABLING OBJECTIVES

1. Describe the three basic strategies used to reestablish core cooling in procedure 19221

, REQUIRED MATERIALS

1. Plant Vogtle Procedure 19221, " Response to Inadequate Core Cooling" l

N O

4 l VII-5-1 rev. 1

O CORE COOLING - F-0.2 GOTC 19221-1 Figure 1 19221-1

=

NO CORE EXIT RyuS

+ TCsLESS - No FULL RANGE THAN 12004 -

l YES GREATER l THAN 39 %

YES NC

' CORE EXIT "

GO TO l TCsLESS 19222-1 THAN 7004 M

GO TO No " 19222-1 AT LEAST ONFAC7 -

RUNNING gygg NO YES FULL AANGE GREATER THAN 39%

ygg RCS GO TO SuSC00UNG No 19223-1 l MONITER. _

"IA YES 28'F GO TO

- , 19222-1 RVUS OYNAMIC' HEAD MANGE NO ,

GREATER TNAN 44 %-4 RCP -

30%-3RCP 20 % -2 RCP M 13 % -1 RCP GO TO 19223-1 CW O

SET VII-5-2 rev. I

THE CORE COOLING CRITICAL SAFETY FUNCTION STATUS TREE O

The core cooling status tree monitors the state of core fuel clad heat removal based on RCS pressure, core exit temperature, RCP status and reactor vessel level. The CSF is considered to be satisfied if subcooling is indicated at the core exit. It is the second highest priority status tree and leads to any of 3 FRP's as shown in Figure 1.

The most serious challenge to the CSF is an indication of inadequate core cooling (ICC). An ICC condition is defined as a high temperature state in the core which has exceeded design basis accident acceptance criteria and where operator action is needed to prevent core damage from occurring.

Extensive analysis of design basis events (such as LOCAs) have been performed and safeguard systems have been designed to insure that no unacceptable level .

of core damage will occur for design basis events. If equipment failure or multiple events occur and result in the design basis assumptions being exceeded, it is possible that conditions can exceed those predicted in design basis analysis. However, if symptoms indicate that this is occurring, O alternative methods can be attempted to restore core cooling. The use of these methods should be restricted, because they:

1. Are beyond the original design basis of equipment.

2. Could j eopardize other CSFs.

Two symptoms of inadequate core cooling have been defined in this tree -

one using core exit thermocouples and reactor vessel level, and the other using core exit thermocouples alone. Either indicates an extreme challenge to the fuel clad / matrix barrier and a red priority is warranted.

If an ICC condition has not been reached, but a degraded core cooling condition as defined in this tree exists, then there are still actions to be taken to respond to the challenge. In most cases, sLailar actions are already provided in the ORGs, but they are repeated in the FRPs to insure that the proper priority is given to these actions. Therefore, any condition VII-5-3 rev. 1

O CORE COOLING - F'0.2 GOTO 19221-1 Figure 1

_ gg g 19221-1 NO COREEXIT TCsLESS - RVUS NO THAN 12004 FUU. RANGE GREATER YES THAN 3 9 %

YES 4

NO COREIIIT

GO TO '

TCsLESS - 19222-1 THAN7909 m go 7o 19222-AT LIAST QNIRCA -

RUNNING gyug NO YES FUU. RANGE GREATER THAN 39%

ygg RCS GO TO SUSC00UNG NC 19223-1 MONITER, GREA M THAN Yg5 28'F GO TO

, 19222-1

=

RVUS OYNAMIC NEAD RANGg NO ,

GREATIR THAN 4 4 %-4 RCP -

30%-3RCP 20 % -2 RCP M 13 % -1 RCP GO TO 19223-1 CW O

SET VII-5-4 rev. 1

- s i

. . l symptomatic of either an inadequate or degraded core cooling condition has

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'~# been given a red or orange priority indicating an extreme or severe challenge to the safety function. If RCS subcooling is not indicated, then the RCS may be saturated. Because this is not a normal condition for core cooling and may be due to inadequate RCS inventory possibly requiring SI flow, the function is considered not satisfied and a yellow priority is warranted.

Saturated conditions in the RCS are expected during some events and adequate core cooling should be maintained if SS is operational.

Core Cooling Blocks Core exit TCs less than 1200 degrees F: Analyses of ICC scenarios show that core exit temperature greater than 1200 degrees F is a satisfactory criterion for basing extreme operator action. At least five theimocouples should read greater than the criterion temperature. Use the following criteria to determine which thermocouples to monitor:

1. At least one thermocouple should be located as close as possible to

(/}

\- the geometric center of the core.

2. The other thermocouples should be located at least one per quadrant over the highest assemblies in each quadrant. Exclude the outer two rows of assemblies, because they can receive significant cooling from SG drainage due to refluxing. Select the thermocouples at each refueling to insure that the highest power assemblies are always being used.

Five thermocoup1es have been chosen to allow for thermocouples failing high. This temperature indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core. An extreme challenge to the fuel matrix / clad barrier is imminent and a red priority is warranted. The appropriate procedure for functicnal response is 19221, " Response to Inadequate Core Cooling." If core exit thermocouples l

are less than 1200 degrees F, then check subsequent blocks for other I

(T

% ,)

conditions for the safety function.

I VII-5-5 rev. I l

l l

NOTE: Do not use hot leg temperature indications to determine an ICC condition. Analyses have shown that hot leg temperature reacts significantly slower than core exit temperature when the core is uncovered. Water draining from the SGs to the core can affect the hot leg temperature indication. Also, hot leg temperature at best indicates the average core temperature, while core exit thermocouples indicate a localized exit temperature above the hottest regions of the core.

RCS subcooling greater than 31 degrees F: If core exit subcooling is less than 31 degrees F, maintain SI flow to the RCS to provide inventory makeup; the Core Cooling CSF is not satisfied. Subsequent blocks check for inadequate or degraded core cooling conditions. If RCS subcooling is greater than 31 degrees F, the CSF is satisfied.

At least one RCP running: If any RCP is running, then use the dynamic head range of RVLIS to assess core cooling conditions. If no RCP is running, use the RVLIS full range.

Core exit TCs less than 700 degrees F: Use the criteria listed in the 1200 degrees F block for selecting thermocouples. If at least five core exit thermocouples indicate greater than 700 degrees F, superheat at the core exit is indicated. An ICC condition will exist if, in the next block, RVLIS indicates less than 3.5 feet collapsed liquid level in the core. If core exit thermocouples indicate less than 700 degrees F, then an ICC condition does not exist and the subsequent RVLIS check will assess whether a degraded core cooling condition has been reached.

RVLIS full range greater than 39% (core exit temperatures greater than 700 degrees F): If RVLIS full range is less than 39% (the value which is 3-1/2 feet above the bottom of active fuel in core with zero void fraction, plus uncertainties), the core is uncovered and an ICC condition has been reached. A red priority is warranted and the appropriate procedure for functional response is 19221, " Response to Inadequate Core Cooling." If RVLIS full range is greater than 39%, than a degraded core cooling condition exists, because the core exit temperatures are greater than 700 degrees F l

l VII-5-6 rev. 1 i

from the previous block. An orange priority is warranted and the appropriate procedure for functional response is 19222 " Response to Degraded Core Cooling."

RVLIS full range greater than 39% (core exit temperatures less than 700 degrees F): If RVLIS full range temperature is less than 39%, then the core is uncovered, but since core exit temperature is less than 700 degrees F, an ICC condition has not been reached. A degraded core cooling condition exists; an orange priority is warranted and the appropriate procedure for

functional response is 19222, " Response to Degraded Core Cooling." If RVLIS full range is greater than 39%, then only a saturated core cooling condition exists; a yellow priority is warranted and the appropriate procedure for functional response is 19223, " Response to Saturated Core Cooling."

RVLIS dynamic head range greater than 44% - 4 RCP; 30% - 3 RCP; 20% - 2%

RCP; 13% - 1 RCP: If an RCP is operating, then even under a highly voided RCS condition the core exit thermocouples can be expected to indicate saturated temperatures. This block checks for RCS voiding less than 50%

which, if RCPS are subsequently stopped, would insure the core would initially be kept covered and adequately cooled. If RVLIS indication of greater than 50% has been reached, do not subsequently trip RCPs, even if RCP trip criteria has not been exceeded or all normal support conditions are not i available. If RVLIS dynamic head range is less than the specified percentages, then a degraded core cooling condition exists and an orange priority is warranted; the appropriate procedure for functional response is 19222 " Response of Degraded Core Cooling". If RVLIS dynamic head range is

! greater than the specified percentages, then a saturated core cooling ,

condition exists and a yellow priority is warranted; the appropriate I Procedure for functional response is 19223, " Response to Saturated Core Cooling." .

I i

l O

VII-5-7 rev. 1 l

+

lhlITI ATE WI-HEAD Eccs No CORE y Cox.

CONDIDOMS ;- hhTkM IM%00 SATism V ppogg Har y AND slap si4 K lu EFFECT NO BLE e

V g DCPRfiE00ZE .

SEcoWDARY sr No CORE Cott CouDITOM S A coyDrpous IHW0) E ATiS ICD 7

NO ES hV4 S1 ART RPs IF TCs

> l200 #

V vewr RCS e l

> CouDiTbkS An5p TS

%f%f LOCA PEotCDoEi 19oio -1 Src p 12.

RLL21VAA Rev.0:12/05 Figt:re 1 Response to inadequate core cooling VII-5-8 rev. 1

c O CORE COOLING FUNCTION RECOVERY PROCEDURES The objective of the recovery / restoration technique incorporated into procedure 19221 (FR-C.1) is directed toward preventing substantial core damage to inadequate core cooling as detected by the core cooling status tree red condition. There are three subsections to this procedure:

Establish safety injection flow to the RCS.

Rapidly depressurize SGs to depressurize RCS. This allows injection of accumulators and low head safety injection, RHR.

Start RCPs and open all RCS vent paths to containment. Forced flow .

will move all available fluid to the core and low RCS pressure should aid low-head ECCS injection.

'(~'\ , Each of these steps is more serious than the last. After trying each of b the first two the operator must evaluate its effectiveness before trying the more extreme measures.

A flow diagram of the procedure is shown in Figure 2. Note the three major actions.

You must evaluate conditions for improved core cooling and increasing vessel inventory before you perform the next action in the sequence. If core cooling is restored through high-pressure safety injection, return to the procedure and the step in effect. If core cooling is not restored, continue with this procedure; once core cooling has been restored and adequate makeup flow has been established and verified, transfer to Procedure 19010. " Loss of Reactor or Secondary Coolant," step 12.

Indications of inadequate core cooling require prompt operator action.

Inadequate core cooling is caused by a substantial loss of primary coolant that results in a partially or fully uncovered core. Without adequate heat VII-5-9 rev. 1

l removal, core decay energy will cause the fuel temperatures to rise. Unless core cooling is promptly restored, severe fuel damage will occur.

The most effective method of re-covering the core and restoring adequate core cooling is to reestablish high pressure safety injection. If some form of high pressure injection cannot be established or is ineffective, take actions to reduce the RCS pressure ao that the SI accumulators and the RHR pumps will inject fluid. Analyses have shown that a rapid secondary depressurization is the most effective method of reducing the RCS pressure. If secondary depressurization is not possible, or primary-to-secondary heat transfer is significantly degraded, start the RCPs. The RCPs will provide forced two-phase flow through the core and temporarily improve core cooling until some form of makeup flow to the RCS can be established.

The recovery techniques applied in this procedure were developed from transient analyses. The expected system response to each of these techniques is described below.

REINITIATING HIGH PRESSURE SAFETY INJECTION The introduction of subcooled safety injection into the highly voided RCS will cause steam condensation in the cold legs. This condensation effect will increase steam flow throughout the RCS. Superheated steam forced out of the core may initially cause the core exit TC temperatures to increase. As the vessel begins to refill, heat transfer from the fuel vill cause the fluid entering the core to boil vigorously. This will create a frothy two-phase mixture that will eventually re-cover the entire core. This two-phase mixture will also cause the core exit TC temperatures to quickly decrease to the saturation temperature of the RCS. The RVLIS indication, which is an indirect measure of vessel level based on local pressure differences within the vessel, may fluctuate as the core is re-covered; however, the general trend in RVLIS indication should increase as the vessel is refilled.

The trends in core exit TC temperatures and indicated vessel level determine the appropriate actions. Use the trend in RVLIS indication to determine 1

1 VII-5-10 rev. 1 l

i whether safety injection is effectively restoring RCS inventory. If this

(

indication is increasing, further action may not be necessary. A RVLIS full l range indication greater than 3.5 feet above the bottom of the active fuel indicates that the RCS inventory has been restored; this allows'a return to i the procedure and step in effect. Use the trend in core exit TC temperatures to determine whether safety injection is effectively restoring core cooling.

If these temperatures are decreasing, further action may not be necessary.

Core exit TC temperatures less than 700* F indicate that core cooling has been restored; this allows a return to the procedure and step in effect.

i RAPIDLY DEPRESSURIZING THE SECONDARY SIDE i

i If high pressure safety injection does not restore adequate core cooling, conduct a rapid steam generator depressurization. A rapid secondary depressurization will increase primary-to-secondary heat transfer and condense the steam in the primary side of the steam generator U-tubes. When the condensation rate exceeds the steam generation rate in the primary side,

, the RCS will begin to depressurize. As the RCS pressure drops, voiding of

, the water in the lower plenum and downcomer will partially re-cover the core I

with a two-phase frothy mixture. The continued RCS depressurization will eventually cause SI accumulator injection and temporary core recovery.

Check the hot leg temperatures to determine whether the steam generator depressurization is reducing the RCS pressure. The ho. leg temperatures may initially rise as superheated steam in the core is forced out by the advancing froth, but should quickly decrease to saturation and continue to i decrease as the RCS depressurizes.

Isolate the SI accumulators when the RCS pressure reaches 250 psig; this will prevent nitrogen injection into the RCS. Use RCS hot leg temperatures of I less than 400* F or intact steam generator pressure of less than 125 psig to .

I determine when the SI accumulators should be isolated.

After the SI accumulators have been isolated, depressurize the secondary side to atmospheric pressure. The RCS pressure should follow the secondary 1

i i VII-5-11 rev. 1

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

C

+

lh/ITIATE 41-HEAD Ecc.s

v

yo CoM eg D y C M DITio WS %w IM%0J sAtip p YE6 PRocCDuE HEAR > AND sEp S4K lu EFFECT NO BIE e

vw EEPRES500ZE -

SEtowoARY v No GRE Corr CouDITc4 S '

ConDrDous

'"M,mJ W uis iED NO WS EV+

START EPs IF Tcs

> l 200

V Ve7 RCS yo I

C4R

> CouciT1oWS TS anspe vsr LOCA PRoCCDOEi 19oio-1 Sm g 12.

RLL21VAA Rev.012/05 Figure 1 Response to inadequate core cooling VII-5-12 rev. I

/'"N pressure until the RHR pumps begin to inject. Once RHR flow has been t i

'- / established and the core is completely re-covered, adequate core cooling has been restored and preparations for long-term plant recovery can be started.

RESTARTING THE RCPs AND OPENING THE PRESSURIZER PORVS If neither high pressure injection nor rapid steam generator depressurization restores adequate core cooling, starting the RCPs will provide forced two-phase flow through the core and temporarily improve core cooling. The core exit TC temperatures should rapidly decrease and the RVLIS dynamic head range indication should rapidly increase as the RCPs force a steam / water mixture through the core. Analysis has shown that, with secondary heat sink available, the RCPs will maintain core cooling as long as they continue to run. However, note that a degraded core cooling condition still exists.

The RCPs cannot be expected to run indefinitely under highly voided RCS conditions. Establish a source of makeup water to the RCS to restore

(}

\- /

adequate long-term cooling. Reduce the RCS pressure so that the SI accumulators and/or the RHR pumps will inject.

Continue attempts to depressurize the steam generators or to establish the secondary heat sink; however, if the core exit TC temperatures remain above 1200* F and all available RCPs are running, the only other option is to effectively enlarge the hole in the RCS to reduce RCS pressure. This may be achieved by opening all available RCS vent paths to containment (pressurizer PORVs, head vents, etc.).

Note that venting the RCS to containment reduces RCS inventory and is not as effective as steam generator depressurization in reducing RCS pressure. Some form of low-pressure flow to the RCS must be established as soon as possible.

VII-5-13 rev. 1

OTHER CORE COOLING PROCEDURES Procedure 19222 (FR-C.2) is entered from the CSFSTs on an orange priority. The major actions to be performed in 19222, " Response to Degraded Core Cooling" are the first two listed above for 19221. Reinitiation of high pressure injection is always the most effective method to restore RCS inventory. If this is not possible the operator then depressurizes the secondary system to allow the accumulators and RHR system to inject.

However, unlike the rapid depressurization used in the red priority 19221, the steaming rate in 19222 is controlled. This procedure limits the RCS cooldown rate to a maximum of 100*F/hr. This is continued until RHR flow to the RCS has been established and verified.

Procedure 19223, " Response to Saturated Core Cooling" is entered, based on operator judgement, from the CSFST, on either of two yellow conditions.

This procedure is an aid to establishing safety injection as necessary to maintain minimum RCS subcooling then return to the procedure and step in effect. A saturated RCS condition does not necessarily imply that core cooling is in jeopardy, but rather that conditions are appropriate to take action to prevent any further degradation of core cooling. The actions of 19222 and 19223 are similar to that of 19221 and normal operating procedures and will not be discussed in the step description section.

l I

l l

O VII-5-14 rev. 1

1 4

TASK: Respond to Inadequate Core Cooling SELF-TEST I l Answer the following questions as completely as possible.

i

1. List, in sequence, the actions that should be performed to restore core

$ cooling.

a.

J

b. i t

c.

j 2. The most effective method of re-covering the core is to .

3. What is the purpose of the rapid secondary-side depressurization? .

l

4. What is the effect of starting the RCPs?

4 1

VII-5-15 rev. 1

- ~ - - - ~ - - -n._ - - , - - - --_wa-, --.---v--,,-m-,n-w, men,_w..- . , - _m- ,-rw,,,,w,,- ww,

?

ANSWERS

1. a. Reestablish high-pressure safety injection.

b. Rapidly depressurize the secondary side.

c. Restart the RCPs and/or open the pressurizer PORVs.

2. Reestablish high-pressure safety injection

3. To reduce the RCS pressure so that the SI accumulators and the RHR pumps will inject fluid

4. Forced two-phase flow will temporarily improve core cooling.

O O

VII-5-16 rev. 1

TASK: Respond to Inadequate Core Cooling

[GD PERFORMANCE GUIDE The following actions are required to respond to an inadequate core cooling condition.

I. Turn to Procedure 19211.

II. Try to establish ECCS flow to the RCS and check for adequate core cooling. (procedure steps 1-7)

III. Rapidly depressurize the steam generators to depressurice the RCS; check for adequate core cooling. (procedure steps 8-16)

IV. Start the RCPs and open all RCS vent paths to containment; check for adequate core cooling. (procedure steps 18-22) ,

V. Continue II-IV until adequate core cooling is established. (procedure step 23)

() You vill use the specific steps of Procedure 19211 to respond to an inadequate core cooling condition. Before you complete the Self-Test I, read through the Step Description section for an explanation of the procedure steps. Obtain a copy of the most recent revision of this procedure and refer to it on a step-by~ step basis as your read through the Description section.

i _

VII-5-17 rev. 1

TASK: Respond to Inadequate Core Cooling STEP DESCRIPTION CAUTION: If the RWST level lowers to less than 45 percent, transfer to cold leg recirculation.

CAUTION: The RHR seal coolers and the RHR heat exchangers are cooled by CCW. If the RCS pressure is above the shutoff head of the RHR pumps, runninh thase pumps for longer than 30 minutes without CCW flow may damage the pumps.

OVERVIEW (procedure steps 1-7): Properly align the ECCS valves, start the ECCS pumps, and check for flow through the ECCS lines into the RCS. Check .

the core exit TCs and the appropriate RVLIS indication to determine whether the safety injection is effectively restoring core cooling and vessel inventory.

-- TRY TO ESTABLISH ECCS FLOW (steps 1-4):

PROCEDURE STEP 1 ACTION: Verify the proper emergency ECCS valve alignment.

INTENT: Using attachments A, B, and C, verify the alignment and manually align the valves as necessary. These attachments are the same ones used in the other emergency procedures.

PROCEDLRE STEP 2 ACTION: Evaluate the ECCS flows.

INTENT: This step is used to identify the existence of a degraded ECCS system. If CCP flow cannot be verified, establish high-pressure injection flow to the RCS by starting the positive displacement charging pump.

VII-5-18 rev. 1

PROCEDURE STEP 3 ACTION: Ensure that RCP support conditions are available.

INTENT: The RCPs may be started in later steps, even if all of the support conditions are not available.

If RCP support conditions are not available, begin actions to reestablish these conditions; these actions should not delay other steps to restore core cooling.

PROCEDURE STEP 4 ACTION: Ensure that the SI accumulator isolation valves are open.

INTENT: Accumulator injection may be required to re-cover the core. It is assumed that complete accumulator water injection has not already occurred.

O -

-- CHECK FOR ADEQUATE CORE COOLING (steps 5-7):

PROCEDURE STEP 5 ACTION: Check the core exit TC temperatures.

INTENT: If the core exit TC temperatures have not been reduced to less than 1200* F through safety injection, immediately perform the alternative actions in this procedure to restore core cooling (beginning with procedure step 8).

If the core exit TC temperatures are less than 1200* F, use procedure steps 6 and 7 to check for adequate core cooling.

O VII-5-19 rev. 1

1 PROCEDURE STEP 6 ACTION: Use the RVLIS full range indication to determine whether safety injection is effectively restoring RCS inventory.

INTENT: If the RVLIS full range indication is above 39 percent, return to the procedure and the step in effect. If the indication is below 39 percent, but rising, return to step 1 and repeat the initial procedure steps until the RVLIS full range indication is greater than 39 percent.

If the RVLIS full range indication is less than 39 percent and not increasing, use procedure step 7 to determine whether an inadequate core cooling condition still exists.

NOTE: A RVLIS full range indication of 39 percent is the value that represents a level of 3.5 feet above the bottom of the active fuel in the core, with zero void fraction, plus uncertainties.

O t

O VII-5-20 rev. 1 l

)

PROCEDURE STEP 7 ACTION: Check the core exit thermocouple temperatures to determine if an inadequate core cooling condition still exists.

INTENT: If these temperatures are less than 700* F, transfer to the

, procedure and step in effect.

If the core exit TC temperatures are greater than 700* but decreasing, return to step 1 and repeat the initial procedure steps.

If the core exit TC temperatures are greater than 700* F and not decreasing, and the RVLIS indication is low (procedure step 6), immediately perform the alternative actions in this procedure to restore core cooling (procedure step ,

8).

- NOTE: The checks in procedure steps 5, 6, and 7 are to determine whether core cooling is adequate.

If the core exit thermocouple temperatures are greater than or equal to 1200*

F, OR if the RVLIS indication is less than 39 percent and not increasing AND the core exit thermocouple temperatures are greater than or equal to 700* F and not decreasing, core cooling is not adequate.

If core cooling is inadequate, it is necessary to inject the accumulators in order to restore a portion of the RCS inventory. Depressurize the steam generators to effect an RCS depressurization. This will allow the accumulators to inject, temporarily providing core cooling. After the accumulators inject, RCS pressure should be low enough to allow the RHR pumps to inject ECCS flow (if these pumps are available). Without a secondary heat sink (and if core exit thermocouple temperatures are greater than 1200* F),

you may have to start an RCP to help cool the core.

VII-5-21 rev. 1

If core cooling is adequate, it is unnecessary and undesirable to inject the accumulators.

Do not perform the following steps unless it is determined that these actions are absolutely necessary to cool the core. However, if the inadequate core cooling condition still exists, these steps should be performed quickly.

OVERVIEW (procedure steps 8-16):

Rapidly cool down and depressurize the RCS by dumping steam or opening the steam generator ARVs while maintaining adequate feedwater to the steam generators. The SI accumulators must be isolated once the steam generators have been depressurized to 125 psig or the RCS has been depressurized to 250 psig (indicated by RCS hot leg temperatures less than 400* F). Continue the RCS cooldown and depressurization until RHR flow into the RCS has been established and verified. Check the core exit TC temperatures and the RVLIS indication to determine whether accumulator and/or RHR safety injection is' effectively restoring core cooling and vessel inventory.

PROCEDURE STEP 8 ACTION: Check for an excessive containment hydrogen concentration by obtaining a current hydrogen sample. Verify that the containment hydrogen analyzers are in service.

INTENT: Depending upon the degree of the hydrogen concentration, turn on the l hydrogen recombiners or notify the plant engineering staff to determine I additional recovery actions, while continuing with this procedure.

I 1

When an inadequate core cooling condition has occurred, the containment j

hydrogen concentration may be as much as 10 to 12 volume percent, depending I upon the amount of metal / water reaction (to produce hydrogen) that has occurred in the core. Obtain the current hydrogen sample co ascertain the flammability / combustibility of the gases in the containment. Note that in l

VII-5-22 rev. 1

order to have flammable hydrogen concentrations, an inadequate core cooling

[V3 situation must have already existed.

The flammability of the hydrogen mixture is determined with respect to the possible rise in containment pressure. If the hydrogen mixture is between 0.5 volume percent and 6.0 volume percent in dry air, either no hydrogen burn is possible or a limited burn may occur (with no significant pressure rise).

In this case, start the hydrogen recombiner system to slowly reduce the containment hydrogen concentration.

NOTE: Operation of the hydrogen recombiners may cause a rise in containment pressure; the recombiners can act as an ignition source.

If the hydrogen concentration is less than 0.5 volume percent in dry air, a ,

flammable situation is not imminent; continue with Procedure 19221. If the concentration is greater than 6.0 volume percent in dry air, immediately notify the plant engineering staff. Consult the staff before starting the

T recombiners, and continue with this procedure.

(O All hydrogen measurements are referenced to coocentrations in dry air (even though the actual containment environment may contain significant steam concentrations) to be conservative and because most hydrogen measurement systems remove moisture from the sample and approximate dry air conditions.

NOTE: Because obtaining a hydrogen concentration measurement may take j some time, continue to perform the alternative actions for restoring core cooling'in this procedure.

PROCEDURE STEP 9 ACTION: Check the steam generator levels.

l INTENT: Check for adequate feed flow or steam generator inventory for secondary heat sink requirements. If the inadequate core cooling symptoms were caused by a loss of secondary heat sink (inventory less than 4 percent

(}

VIl-5-23 rev. 1

in the narrow range indication and total feed flow of less than 550 gpm), in combination with a loss of high-pressure safety injection, perform procedure step 18. In steps 9-17, attempts are made to cool the core using the secondary heat sink. If the heat sink is not adequate, bypass all of these steps and start the RCPs (step 18). The actions begun in step 18 will provide temporarily improved core cooling until either feedwater or safety injection is restored.

NOTE: This is a continuous action step; the steam generator levels should be monitored throughout the rest of the procedure.

CAUTION: If the CST level drops below 15 percent, alternate water sources for the AFW pumps will be necessary.

CAUTION: A faulted or ruptured steam generator should only be used if no intact steam generator is available. Depressurizing a ruptured steam generator may create a. path to the atmosphere for release of radioactive

- materials. In addition, a faulted steam generator has probably already depressurized.

PROCEDURE STEP 10 ACTION: Try to terminate the loss of RCS inventory through the RCS vent paths.

INTENT: In particular, check the pressurizer PORVs and block valses in addition to other RCS vent paths.

To ensure the operability of the pressurizer PORV block valves, verify that power is available to them. Close the pressurizer PORVs in case a valve is stuck open and not detected. Leave at least one block valve open so that at least one PORV will be available for pressure excursions in the RCS (caused by degraded conditions) and to prevent a challenge of the pressurizer safety valves.

O VII-5-24 rev. 1

-- i O

v PROCEDURE STEP 11 ACTION: Depressurize all intact steam generators to 125 psig.

INTENT: The rapid secondary depressurization has been shown to be the most effective way to reduce RCS pressure; RCS pressure must be reduced for the SI accumulators and the RHR pumps to inject.

The RCS should not be depressurized below 250 psig without first isolating the SI accumulators. This is to prevent nitrogen injection into the RCS.

This is accomplished by stopping the secondary depressurization at a steam generator pressure of 125 psig or when the RCS hot leg temperatures fall below 400* F (indicating RCS pressure less than 250 psig). When pressure in the secondary side is at 125 psig, the RCS pressure should be greater than or ,

equal to 250 psig.

NOTE: It will be difficult to maintain the steam generator levels during the rapid depressurization. The steam generator tubes may be C partially uncovered if the steam removal rate exceeds the maximum feedwater addition rate. This is an anticipated result of the rapid steam generator depressurization. Using the maximum available feed flow, try to keep the steam generator tubes covered to maximize the primary- to-secondary heat transfer.

PROCEDURE STEP 12 ACTION: When the RCS hot leg temperature criteria are satisfied, isolate the SI accumulators. -

INTENT: This prevents nitrogen injection into the RCS (two RTDs are used to ensure that one RTD is not giving an erroneous reading). Nitrogen could collect in the high places and either produce a "hard" pressurizer bubble or cause gas binding and reduced heat transfer in the steam generator U-tubes.

If it is necessary to vent the nitrogen, open the vent lines and continue

() with this procedure.

VII-5-25 rev. I t

The hot leg temperature of 400* F is used so that the RCS saturation pressure exceeds the accumulator pressure after the accumulator water has been discharged.

PROCEDURE STEP 13 ACTION: Stop all RCPs.

INTENT: In preparation for the subsequent depressurization of the steam generators to atmospheric pressure, stop all RCPs. The RCPs should be secured because of

1. The anticipated loss of number 1 seal requirements (less than 200 psid)

2. The possible loss of minimum NPSH

3. The possibility of flashing in the RCS, which could cause slug flow .

through the RCS, causing mechanical damage to the RCPs PROCEDURE STEP 14 ACTION: Continue the steam generator depressurization.

INTENT: With continued steam generator depressurization, the RCS pressure should follow secondary pressure until the shutoff head of the RHR pumps is reached. Low-pressure safety injection should begin to refill the RCS.

PROCEDURE STEP 15 ACTION: Verify the ECCS flow.

INTENT: If no ECCS flow is present, continue trying to establish ECCS flow (start the positive displacement pump if no other sources are available) by any possible means.

PROCEDURE STEP 16 ACTION: Check for adequate core cooling.

VII-5-26 rev. 1

. -_ ._. . _ . .A

p INTENT: Before this procedure can be exited, the core exit TC temperatures V must be less than 1200* F, at least two RCS hot leg temperatures must be less than 350' F (to ensure that RCS pressure is less than the shutoff head of the RHR pumps), and the RVLIS full range indication must be greater than 62 percent. When the above conditions have been met and ECCS flow or other makeup flow has been established, core cooling has been restored. Note that these conditions are more stringent than earlier transition conditions because the RCS should now be fully depressurized.

OVERVIEW (procedure steps 18-22):

Restart the RCPs and open all RCS vent paths to containment, while continuing efforts to establish makeup flow to the RCS and to restore the secondary heat sink and/or steam generator depressurization capability.

PROCEDURE STEP 18 ACTION: Determine whether the RCPs should be started.

INTENT: The actions of this step may provide temporary core cooling until some form of makeup flow to the RCS is established or one of the items below is restored.

NOTE: This step will be entered if:

1. The steam generators cannot be depressurized

2. Steam generator depressurization was not effective in restoring adequate core cooling

3. The secondary heat sink was lost Start RCPs one at'a time until the core exit TC temperatures are less than 1200* F. The RCPs should force two-phase flow through the core, temporarily keeping it cool. Even single-phase forced steam flow will cool the core for some time provided the RCPs can be kept running and a heat sink is available.

O VII-5-27 rev. 1

If restarting RCPs does not decrease the core exit TC temperatures below 1200* F, open the pressurizer PORVs. This may help reduce the RCS pressure enough to cause low-pressure (or intermediate-pressure) safety injection flow. If the core exit TC temperatures remain above 1,200* F after all of the pressurizer PORVs and block valves are opened, open all other RCS vent paths to containment to reduce RCS pressure.

NOTE: RCPs may be required to temporarily cool the core under highly voided RCS conditions. RCPs should be started when required even if all normal start-up conditions have not been met. Failure to start the RCPs when required coulu cause core damage.

Start the minimum number of RCPo necessary to provide temporary core cooling.

In this situation, it is undesirable to run all RCPs unless absolutely .

necessary. The RCPs may be required at a later time for core cooling, and they should r.ot be run unnecessarily. Depending upon the circumstances, one RCP is likely to provide as much core cooling as two or more pumps running would provide. In addition, running RCPs under these adverse conditions will likely make them unusable at a later time. Seal package failure is a possibility and would provide another inventory loss path (from an undesirable location, low in the cold leg).

It is unlikely that conditions that require starting an RCP will be reached without a lack of RCS inventory. With the lack of an effective secondary heat sink, the RCS vent pathe, are the heat sink. The loss of RCS inventory through these vent paths will compound the inventory problem.

ACTIONS SHOULD CONCU_RRENTLY BE TAKEN TO RESTORE RCS INVENTORY (ECCS FLOW)

PROCEDURE STEP 19 ACTION: If the depressurization of the steam generators from the control l room is unsuccessful, have an operator try to locally depressuri::e them.

l VII-5-28 rev. I t

PROCEDURE STEP 20 ACTION: Determine whether the SI accumulators should be isolated.

INTENT: The SI accumulators will be allowed to inject their entire water inventory if necessary. The accumulator isolation criterion, intermittent RHR pump flow, will ensure that safety injection flow to the RCS is established before accumulator isolation occurs. Some nitrogen may be injected as a result of this criterion because the shutoff head pressure of the RHR pumps is less than 250 psig.

PROCEDURE STEP 21 ACTION: Determine whether the RCPs should be stopped. .

INTENT: If the required conditions are satisfied (at least two RCS hot leg temperatures are less than 350' F and at least intermittent RHR pump flow is eqtablished), the RCPs are no longer needed for cooling and can be stopped.

PROCEDURE STEP 22 ACTION: Verify the ECCS flow.

INTENT: The objective of the previous steps was to reduce the RCS pressure below the shutoff head of the RHR pumps. This should cause low pressure safety injection. This step verifies ECCS flow to the RCS.

PR_0CEDURE STEP 23 j ACTION: Check for adequate core cooling.

INTENT: See procedure step 16. If core cooling is adequate at this point, transfer to step 12 of Procedure 19010 (discussed in volume III) to check the overall plant status with respect to radioactivity leakage and availability of equipment needed for long-term plant recovery.

If adequate core cooling has not been restored, return to procedure step 18.

I VII-5-29 rev. 1 l

l

TASK: Respond to Inadequate Core Cooling SELF-TEST II Before proceeding to the Task Practice section, answer the following questions as completely as possible.

1. If safety injection (established in procedure steps 1-4) does not lower the core exit TC temperatures to less than 1200* F, you should I. Check the RVLIS indication to be sure it is greater than 39 percent.

II. Return to procedure step 1 to verify the proper ECCS flow.

III. Check for a secondary heat sink,

a. I

b. I and II

c. II

d. III

2. After the steps to establish safety injection (1-4) have been completed and the core exit TC temperatures have been verified at less than 1200*

F, what would your next action be under each of the following conditions?

a. RVLIS indication is greater than 39 percent

b. RVLIS indication is less than 39 percent and rising c.RVLIS indication is less than 39 percent and not rising

3. The steam generator depressurization should be stopped at 125 psig.

This should correspond to an RCS pressure of and RCS hot leg temperatures of .

l

4. Before the steam generators are depressurized to atmospheric pressure, you should perform the following:

l l a.

I b.

l O

l l

VII-5-30 rev. I

5. Depressurizing the steam generators to atmospheric pressure will allow

\ the to inject into the RCS.

6. After the steam generators are depressurized to atmospheric pressure, if the core exit TC temperatures are less than 1200* F, what are the criteria for exiting Procedure 192217 a.

b.

! 7. It may be necessary to immediately restart the RCPs, even if RCP support conditions cannot be established, in which of the following cases:

l a. RVLIS indicated level is less than 39 percent.

t b. The steam generators cannot be depressurized.

c. The steam generator depressurization did not restore core cooling.

d. The secondary heat sink is lost.

8. If all RCPs are running and the core exit TC temperatures are still l'

above 1200* F, you should .

4 i

i 1

1 i

i l

i 1

1 4

l l~

VII-5-31 rev. 1

ANSWERS

1. d. III

2. a. Return to the procedure and step in effect.

b. Return to step 1 to verify proper ECCS flow.

c. Check the core exit TC temperatures to see if they are less than 700* F or lowering.

3. 250 psig, 400* F

4. a. Isolate the SI accumulators

b. Secure the RCPs

5. RHR pumps

6. a. At least two RCS hot leg temperatures less than 350' F

b. RVLIS full range indication greater than 62 percent

7. b. The steam generators cannot be depressurized .

c. The steam generator depressurization did not restore core cooling

d. The secondary heat sink is lost

8. open the pressurizer PORVs and block valves i

O i

l VII-5-32 rev. 1

/' TASK: Respond to Inadequate Core Cooling Name Date Social Security No. Trainee No.

PERFORMANCE OBJECTIVE Given that either

1. The core exit thermocouple temperatures are greater than 1200* F or

2. RCS subcooling is less than 31' F, no RCPs are running, the core exit thermocouple temperatures are greater than 700* F, and the RVLIS full range indication is less than 39 percent, restore core cooling.

Adequate core cooling must be established as defined by steps 5-7, step 16, or step 23 of Procedure 19221.

Os VERIFICATION OF SATISFACTORY COMPLETION o Completed Performance Test satisfactorily o Turned in Instructional Unit Feedback Sheet Instructor's Signature Date O

VII-5-33 rev. 1

O INTENTIONAL BLANK

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VII-5-34 rev. 1

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CHAPTER 11: Respond to Voids in Reactor Vessel PERFORMANCE OBJECTIVE i

Given that the RVLIS indicates the upper head of the reactor vessel is not full, eliminate reactor vessel voids. The RVLIS must indicate that the upper head of the reactor vessel is full. The PRZR level must be stable.

ENABLING OBJECTIVES

1. State the venting termination criteria of procedure 19263.

2. Explain the concern if an RCP is stopped during the venting operation.

REQUIRED MATERIALS Plant Vogtle Procedure 19263, " Response to Voids in Reactor Vessel" .

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VII-11-1 rev. 1 l

INTRODUCTION A rapid RCS cooldown/depressurization may result in the vessel head fluid reaching saturation. Whenever saturation conditions exist in the vessel head, the potential exists for the formation of steam (condensable) voids in the reactor vessel. Complete SI accumulator tank discharge into the RCS may result in nitrogen voids forming in the upper head. Inadequate core cooling may create hydrogen voids in the upper head. These gas voids are called "noncondensable" voids.

If voids are detected in the reactor vessel, use Plant Vogtle Procedure 19263, " Response to Voids in the Reactor Vessel," to reduce and eliminate these voids.

ENTRY CONDITIONS Plant Vogtle Procedure 19263 should not be implemented if

1. steam voids are detected, but RCS depressurization must continue. Refer to riant Vogtle Procedure 19004, " Natural Circulation Cooldown with Steam Void in Vessel (without RVLIS)" (discussed in volume II of this text)

2. the primary purpose is to mitigate an inadequate core cooling condition.

Refer to Plant Vogtle Procedure 19221, " Response to Inadequate Core Cooling," (chapter 5) for defense of the core cooling critical safety function.

3. a controlled natural circulation is in progress and a void in the reactor vessel upper head is expected. Refer to Plant Vogtle Procedure 19003 " Conduct Natural Circulation and Depressurization with Steam Void in Vessel (with RVLIS)" (volume II) .

Entry to Plant Vogtle Procedure 19263 vill occur based upon operator judgment from a YELLOW on the inventory critical safety function status tree if the PRZR level is at or above normal and the reactor vessel is less than full.

See Plant Vogtle Procedure 19200, " Critical Safety Function Status Trees" for specific entry conditions for Plant Vogtle Procedure 19263. In general, a RVLIS indicating less than a full upper head is the primary means of VII-11-2 rev. 1 1

('~] determining whether voids exist. Indirect indications of voids in the RCS

\- / (not necessarily located in the reactor vessel head) are

1. Abnormal PRZR level response to RCS pressure changes. The level may decrease during a RCS pressurization because of void compression. The level may rise rapidly during a spraying or any other depressurization because of void expansion or generation.

2. Reactor vessel head temperatures equal to or greater than saturation temperature. Steam bubble formation in the reactor vessel head is likely.

3. Complete SI accumulator tank discharge or inadequate core cooling.

Noncondensable voids in the RCS are likely.

MAJOR ACTIONS Do not attempt to eliminate voids until the bulk of the RCS is stable and subcooled with SI terminated. Then attempt to condense the voids by pressurizing the RCS; if the voids are composed of steam this action should eliminate the voids. However, if pressurization fails to eliminate the voids and noncondensable voids are not suspected,-attempt to start one RCP.

O Starting one RCP will force water into the upper head to cool and condense a superheated (hard) steam void. Next, check the RVLIS and the PRZR level. If the voids have been eliminated, exit the procedures. If the voids remain, prepare to vent the reactor vessel. Continue venting until the voids have been eliminated or the venting criteria have been exceeded. When the voids have been eliminated, exit Plant Vogtle Procedure 19263 and return to the procedure and step in effect.

ADDITIONAL EXIT CONDITIONS l

Exit Plant Vogtle Procedure 19263 and return to the procedure and step in effect when ANY of the following conditions exist:

l

1. SI has not been terminated.

2. RCS repressurization eliminates the void and PRZR level is stable.

3. RCP restart eliminates the voids and PRZR level is stable.

4. Voids are vented and PRZR level is stable.

O VII-11-3 rev. I 1

NOTE: This instructional unit assumes that the plant's reactor vessel liquid inventory system (RVLIS) is availnble. If RVLIS is unavailable, secondary indications must be used to detect void formation:

1. Increasing PRZR level

2. Increasing or erratic neutron flux reading

3. Increasing or varying core exit thermocouple reading Refer to volume II for further explanation of secondary indications.

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VII-11-4 rev. I

fg SELF-TEST I O

1. Whenever saturation conditions exist in the reactor vessel upper head, the potential exists for the formation of steam voids in the reactor vessel.

a. True

b. False

2. What type of void can be formed in the upper head of the reactor vessel from complete SI accumulator tank discharge?

3. What condition can lead to hydrogen void formation in the upper head of the reactor vessel?

4. For each of the conditions listed below, indicate whether or not Plant

-~g Vogtle Procedure 19263 should be implemented. Circle "Y" if it SHOULD

')

\,, be implemented, "N" if it SHOULD NOT be implemented.

Y N a. RVLIS indicates the upper head of the reactor vessel is full; RCS depressurization must continue.

Y N b. A controlled natural circulation cooldown is in progress and a void in the reactor vessel upper head is expected.

Y N c. The primary purpose is to mitigate an inadequate core cooling condition.

S. For each of the conditions listed below, indicate whether or not Plant Vogtle Procedure 19263 should be exited. Circle "Y" if it SHOULD be exited, "N" if it SHOULD NOT be exited.

Y N a. SI has not been terminated.

Y N b. RCS repressurization eliminates the void; PRZR level is stable.

Y N c. RCP restart fails to eliminate the void; PRZR level is stable.

Y N d. Voids are vented; PRZR level is stable.

VII-11-5 rev. 1

1 i

ANSWERS

1. True

2. Nitrogen

3. Inadequate core cooling

4. a. N

b. N

c. N

5. a. Y

b. Y

c. N

d. Y O

a O

VII-11-6 rev. 1 J

/N TASK: Respond to Voids in Reactor Vessel

, k_s PERFORMANCE GUIDE The following actions are required to respond to voids in the reactor vessel.

I. Turn to Plant Vogtle Procedure 19263.

II. Establish charging flow and letdown. (procedure steps 1-3)

III. Stabilize RCS conditions. (procedure step 4)

IV. Repressurize RCS to condense the voids. (procedure steps 5-8)

V. If noncondensable voids are suspected, go to VI. If condensable voids are suspected, start one RCP to .

try to condense voids. (procedure steps 9-11)

VI. Make preparations for venting the reactor vessel (procedure steps 12-18)

VJI. Vent the reactor vessel. If venting is successful, return to procedure and step in effect. If unsuccessful, return to VI. (procedure steps 19-20)

VIII. Return to the procedure and step in effect when RVLIS indicates the upper head of the reactor vessel is full. (procedure steps 21-22)

You will use the specific steps of Plant Vogtle Procedure 19263 to perform this task. Before you complete Self-Test I, read through the Step Description section for an explanation of the procedure steps. Obtain a copy of the most recent revision of this procedure to refer to on a step-by-step basis as you read through the description section.

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V VII-11-7 rev. 1

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STEP DESCRIPTION Turn to Plant Vogtle Procedure 19263.

OVERVIEW (procedure steps 1-3): Establish charging and letdown to control PRZR level in future steps.

PROCEDURE STEP 1 ACTION: Check whether SI has been terminated.

INTENT: If SI is still operating, more critical plant conditions must be addressed before attempting to eliminate voids. Return to the procedure and ,

step in effect.

PROCEDURE STEP 2 ACTION: Check whether charging flow has been established.

INTENT: Establish charging flow to establish stable conditions and to control PR2R level in subsequent steps. Though it is expected that one CCP will be running at this time, check the status of CCP operation. Charging and letdown flow probably have already been established.

If charging pumps are running, check that RCP seal injection is at least 8 I gpm. If charging pumps are not running, then seal injection is lost, and the i only remaining source of seal cooling is ACCW. If ACCW flow to the RCP thermal barrier is also lost (zero reading), then the seal is assumed to be hot. Rather than initiate the slow process of carefully reestablishing seal l cooling at this time, isolate seal injection to the affected RCP before starting one CCP. If instrument air is not available, verify that instrunear

{ air pressure is normal. If instrument air pressure is not normal, initiate l

l Plant Vogtle Procedure 13710 " Service Air System," Section 4.1, " Service '

O VII-11-8 rev. 1

e

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~N Air System Startup". Then open instrument air to containment isolation valve HV-9378.

Verify that charging flow has been established. If it has not, establish 60 gpm flow and control the rate as necessary to maintain PRZR level.

NOTE:

Miniflow should be verified prior to reducing charging flow below 60 gpm in subsequent steps. If 60 gpm cannot be established, then there will be no cooling for the regenerative heat exchanger, and normal letdown should not be established.

Establish excess letdown to provide a bleed path to aid in controlling PRZR level, because RCP seal injection was already established. Initiate Plant Vogtle Procedure 13008, "CVCS Excess Letdown," Section 4.1, then go to procedure step 4.

PROCEDURE STEP 3 C

ACTION: Check that letdown is in service.

INTENT: Normal letdown provides a controlled mechanism for offsetting volume additions through the charging system. It also supplies high temperature RCS water for heating the normal charging flow in the regenerative heat exchanger. Therefore, to establish stable conditions and control PRZR level in subsequent steps, letdown should be established. If excessive activity levels in the RCS are suspected, evaluate the consequences of establishing letdown prior to taking this action. If letdown is not in service, establish letdown by initiating 13006 "CVCS Startup and Normal Operation," Section 4.4,

" Establishing Normal Letdown Flowpath". If letdown cannot be established, then establish excess letdown by initiating Plant Vogtle Procedure 13008, "CVCS Excess Letdown," Section 4.1.

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VII-11-9 rev. 1

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OVERVIEW (procedure step 4): Stabilize RCS conditions before attempting to eliminate the voids.

PROCEDURE STEP 4 ACTION: Establish stable RCS conditions.

INTENT: Stabilize RCS conditions to detarmine the extent of the void. In addition, changes in pressure, charging flow, and reactor coolant pressure may condense a steam void before you are prepared for it, resulting in a rapid PLZR level decrease and possible SI initiation. If the PRZR level has not stabilized between 40 and 60 percent, control charging and letdown as necessary. If the RCS pressure is not stable, energize PRZR heaters and use normal PR2R spray as necessary. If normal spray is not available and letdown is in service, use auxiliary spray. If RCS hot leg temperatures are not stable, dump steam as necessary.

OVERVIEW (procedure steps 5-8): Increase RCS pressure by 50 psi to condense steam voids by increasing the RCS pressure above the void's fluid saturation point. As the void volume shrinks or condenses, the PR2R level will drop.

Use charging and letdown to maintain the PRZR level above the PRZR heaters.

Monitor RVLIS for indication of the effectiveness of this action. If the voids are primarily condensable, then the total void volume will decrease with increasing pressure, and the PRZR level will drop.

PROCEDURE STEP 5 ACTION: Check whether all RCPs are stopped.

INTENT: If RCPs are running, condensable steam voids should not be present in the vessel head. Go to procedure step 12 for eliminating noncondensable voids.

PROCEDURE STEP 6 ACTION: Check whether RCS pressure should be raised.

VII-11-10 rev. 1

r

/~'3 INTENT: Before increasing RCS pressure, two conditions must be met:

-- 1. Assure that at least a 100 psi margin exists between the RCS pressure and the technical specification cooldown pressure limitation determined for the RCS temperature.

2. Verify that the RCS pressure is less than 1,836 psig, the pressure at which further increases will do nothing to condense the void.

If these conditions are met, energize PRZR heaters to raise RCS pressure 50 psi. If the conditions for pressure increase cannot be met, go to procedure step 9 and initiate the next void elimination strategy, starting the RCPs.

Condensing of the void may be a slow process. If an upward trend is witnessed on the RVLIS, pressurization should be maintained until the head is refilled or the upward trend stops. Some vessel level increase may'also be detected even with noncondensable voids, because of the compression of gas.

PROCEDURE STEP 7 N- ACTION: Control charging and letdown to maintain PRZR level greater than 21 percent (50 percent for adverse containment).

INTENT: This is the indicated level at which the PRZR heaters are reliably covered. This is a programmed level for 0 percent power (CCP or PDP can control level in AUTO here). During the RCS pressure increase of procedure step 6, the condensable portion of the void may condense and the reactor level may rise. This will cause the PRZR level to drop. A balance between charging and letdown can be maintained so that any decrease in the PRZR level would be a supplementary indication (to RVLIS) that the void is shrinking.

PROCEDURE STEP 8 ACTION: Check RVLIS upper range indication.

INTENT: As the RCS pressure is increased, the void should condense and the f- reactor vessel should increase. The increase will be slight if only l

N- compression of the voids is occurring. If condensation occurs, the level VII-11-11 rev. 1

r O e will rise slowly, but pronouncedly. The first verification that a void is condensing is evidence that the vessel level is rising toward the full indication. If the level is not rising or is asymptomatically approaching a level less than full, go to procedure step 9. If the reactor vessel indicates full, the void has been condensed; return to the procedure and step in effect. If the RCS level is increasing but has not reached a full condition (98 percent), return to procedure step 6 and attempt a further RCS pressure increase.

OVERVIEW (procedure steps 9-11): Try to start one RCP, if the repressurization has failed to eliminate the voids. RCP operation will force cooling flow into the upper head from the downcomer and should condense any steam in the upper head. By limiting RCP operation to one pump, the potential for sweeping any noncondensable gases into the rest of the RCS is minimized. Monitor RVLIS for indications of the effectiveness of this action.

PROCEDURE STEP 9 ACTION: Try to start one RCP.

INTENT: The approach in this procedure step assumes that the voids consist almost entirely of steam. If, for any reason, a large part of the voids is suspected to be noncondensable gases, procedure step 9 should be bypassed in favor of direct vessel venting. This ceurse of action would retain the noncondensables in the upper head for venting.

The potential for degradation in RCP seal performance increases with increasing temperatures above 300* F. If RCP seal cooling has been lost for only a few minutes, the inventory of cold water in the seal area should prevent excessive seal heatup. If seal cooling has been previously lost for a significant period of time, seal and/or bearing damage may occur. The potential nonuniform sealing surfaces and seal crud blockage that may exist prior to an RCP start can aggravate bearing and seal damage once the RCP is started. Following restoration of seal cooling, conduct a complete RCP status evaluation before restarting the affected RCP.

VII-11-12 rev. 1

l 4 .

(}

%d RCPs should not be started prior to a status evaluation unless an extreme (red) or severe (orange) CSF challenge is diagnosed. Under such a CSF challenge, the Rules of Usage apply and an RCP should be started if so instructed in the associated FRP; potential damage to the RCP is an acceptable consequence. This is not the case if you are in Plant Vogtle Procedure 19263; the conditions of Attachment B should be met.

There are PRZR connections to RCS hot leg 4 via the surge line and to RCS cold legs 1 and 4 via the spray lines. Single pump operation in the loop that provides the best spray (loop 4 then loop 1) is preferred to obtain normal PRZR spray capability.

Prior to this step, attempts to repressurize the RCS and condense the upper head void have been made. Because at this point the upper head void is still .

present, the next action is to start an RCP and force cooling flow through the upper head spray nozzles. This will condense any steam voids present.

By starting only one RCP, any redistribution of noncondensables through the RCS will be minimized. Requirements on 2RZR level and RCS subcooling listed O'- in procedure step 9 should be met prior to starting an RCP. If these conditions cannot be met, DO NOT start any RCP, but proceed to procedure step

12. The void could condense rapidly during pump startup, drawing liquid from the pressurizer and reducing reactor coolant subcooling. If pressurizer inventory is not sufficient, level may decrease off span. Local flashing of reactor coolant could occur if RCS subcooling is not adequate. These conditions would require SI reinitiation.

PROCEDURE STEPS 10-11 ACTION: Check RVLIS indication. If RVLIS indicates that the upper head is full, go to step 21, bypassing the venting operation.

INTENT: Though a true vessel level indication will not be available with a pump running, the RVLIS dynamic head range provides enough information to determine whether starting an RCP was successful in eliminating the voids.

If the RVLIS indicates a full vessel, the void has been eliminated. Bypass VII-11-13 rev. 1

the venting steps, stabilize PRZR level, and return to the procedure and step in effect. If the RVLIS indicates that the the reactor vessel is not full, proceed to procedure step 12 to prepare for venting.

OVERVIEW (procedure steps 12-18): If an RCP cannot be started or if RCP operation fails to eliminate the voids, the voids probably consist of noncondensable gases. Prepare containment for venting, calculate vent time limits, and review vent termination criteria. Stabilize RCS conditions prior to venting.

PROCEDURE STEP 12 ACTION: Obtain a containment hydrogen concentration measurement.

INTENT: Obtain a containment hydrogen concentration measurement to determine an allowable venting period. Though this information is not needed until procedure step 17, this action is performed now because of the time required to obtain a hydrogen concentration reading.

PROCEDURE STEP 13 ACTION: Check whether low PRZR pressure SI can be blocked.

INTENT: Venting the vessel will decrease the RCS pressure. Thus, it is necessary to block the low PRZR pressure SI signal so that SI will not initiate during venting. Though it is expected that SI has been initiated and terminated prior to entering Procedure 19263, check the low PRZR pressure SI signal to ensure that it is blocked. Even though the signal is blocked, check RCS conditions continuously for any degradation that would warrant l manual SI actuation.

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PROCEDURE STEP 14 1

ACTION: Record the RCS pressure prior to the start of the venting operation.

O VII-11-14 rev. 1 J

(~'T INTENT: This value provides a benchmark from which to begin each venting

\

N--) operation. It is important with each venting operation to keep RCS pressure from decreasing into a region where additional steam void generation could begin. By restoring the RCS pressure before each venting, additional steam void formation is minimized.

PROCEDURE STEP 15 ACTION: Stabilize the RCS.

INTENT: Stabilize the RCS prior to venting so that any change in RCS conditions will only be due to venting. Several of the RCS parameters that are established in this procedure step are monitored in procedure step 18.

PROCEDURE STEP 16 ACTION: Prepare containment for reactor vessel venting.

('~h

'-- INTENT: The venting operation will result in RCS gases being vented to the PRT. Therefore, align the waste gas system to receive any gases vented to the PRT. Also, PRT rupture disks may rupture if PRT pressure becomes great enough, causing a release of gases to containment. Therefore, isolate the containment puree and exhaust system to prevent the release of any radioactive gases to the environment. Start all available containment air l circulation equipment to prevent hydrogen from forming a potentially explosive local gas pocket, t

l PROCEDURE STEP 17.

l ACTION: Determine maximum allowable venting time.

INTENT: Verify that containment hydrogen concentration is less than 3 percent in dry air. This limit ensures a reasonable margin against a potentially explosive hydrogen concentration forming inside containment. The lower the

/N initial hydrogen concentration, the longer the venting operation can VII-11-15 rev. 1

l I

continue. Attachment D provides the methodology for determining venting time. A brief description of the basis for this calculation follows.

Decemine a maximum allowable time period for venting to limit the containment hydrogen concentration:

Attachment step 1: Determine the total containment air volume in cubic feet at STP conditions by correcting for temperature and pressure difference:

Cont. air volume (STP) = (cont. volume ft 3) x (cont. air pressure **/

14.7 psia) x (492'R*/ cont. temp.)

Temperature in 'R (* F + 460) ** If pressure is less than 14.7 psia, use actual value. If pressure is greater than 14.7 psia, use 14.7 psia as a _

conservative measure.

Attachment step 2: Determine the volume of hydrogen concentration that

- can be safely added to the containment atmosphere. Allow sufficient time for the air circulation equipment to mix the containment atmosphere prior to sampling to determine a representative concentration.

Calculate the maximum volume of hydrogen that can be vented to limit the containment hydrogen to less than 3 percent volume.

Attachment step 3: Determine the hydrogen flow rate as a function of RCS pressure from a curve that plots RCS pressure vs. hydrogen flow rate (procedure Figure 1). The curve was generated from a calculation that determined the flow rate of hydrogen at various RCS pressures through a 3/8-inch orificed line. The calculation assumes pure hydrogen, which is conservative because the gaseous void in the vessel will probably be a mixture of gases including steam.

Attachment step 4: The venting time will be the volume allowable from attachment step 2 divided by the expected flow rate from attachment step 3.

O VII-11-16 rev. I

, =s e

. .. e PROCEDURE STEP 18 ACTION: Review reactor vessel venting termination criteria.

INTENT: Review the five vent criteria before venting, and closely monitor the instrumentation for these criteria during venting:

1. RCS subcooling: This criterion requires head vent termination if minimum RCS subcooling is reached. This is consistent with the SI reinitiation

! criterion.

2. PRZR level: This criterion requires that the PRZR level remains above the top of the heaters, because use of the PR2R heaters is the preferred means of maintaining RCS pressure, and because letdown will not isolate at this level.

3. RCS pressure: This criterion limits the allowable RCS pressure decrease to 200 psi. This value is chosen to maintain a relatively stable RCS as weil as provide additional margin against subcooling loss. In addition, 200 psi is readable on RCS pressure instrumentation.

( 4,. Venting time: This criterion ensures that the 3 percent by volume hydrogen limit is not reached. This time criterion is conservative, because it is based upon the assumption that the voids are composed completely of hydrogen (see procedure step 17).

5. RVLIS: This criterion stops the venting when the RVLIS indicates that the reactor vessel is full (the void is eliminated). The applicable RVLIS indication whether determine if the reactor vessel is full depends upon the status of the RCPs.

Any of the vent criteria listed above could be reached in a short time span.

Venting should be stopped if any of the vent termination criteria are exceeded.

OVERVIEW (procedure steps 19-20): Check the RVLIS after the venting to i determine whether any voids remain in the reactor vessel. If voids still l

l remain, repeat venting after restabilizing RCS conditions.

O VII-11-17 rev. 1 l

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PROCEDURE STEP 19 ACTION: Vent the reactor vessel.

INTENT: To vent the reactor vessel, open a vent path and monitor the vent termination criteria. Continue venting until any of the vent termination criteria is reached. Use only one of the two redundant vent paths. Open both isolation valves in the selected vent path to initiate venting. If one or both valves in the selected path cannot be opened, close the valves and then open both valves in the other vent path. Close both valves when any one of the criteria listed in procedure step 18 is exceeded. Because actual venting time depends upon RCS pressure as well as void composition and size, it may be necessary to terminate venting with voids remaining in the reactor vessel. ,

If any RCPs stop during venting, continue venting while establishing natural circulation flow until a venting termination criterion is exceeded. This action minimizes the amount of gas that could collect in the steam generator U-tubes. During a subsequent RCS depressurization, any voids existing in the steam generator will expand, which may result in anomalous PRZR level readings and degradation of steam generator heat transfer.

PROCEDURE STEP 20 ACTION: Check if RVLIS indicates that the upper head is full.

INTENT: If reactor vessel venting was terminated for any criterion other than a full indication on_RVLIS, check the reactor vessel level to determine whether the void has been eliminated. If the void still exists, repressurize the RCS and initiate another venting cycle. Note that the proper RVLIS indication of a full reactor vessel depends upon the status of the RCPs.

PROCEDURE STEPS 21-22 ACTION: When the void has been eliminated, check that the PRZR level is stable and then return to the procedure and step in effect.

VII-il-18 rev. 1

_ . _ _ J

r

. s' e b) TASK: Respond to Voids in Reactor Vessel SELF-TEST II Before proceeding to the Task Practice, answer the following questions as completely as possible.

1. Which of the following is NOT a venting termination criterion?

a. PRZR level

b. Venting time

c. RCS subcooling

d. Cooldown rate

e. RVLIS

f. RCS pressure .

2. If voids in the reactor vessel are suspected to consist of noncondensable gases, one RCP should be started, r a. True

b. False

3. The RVLIS dynamic head range provides enough information to determine whether starting an RCP was successful in eliminating voids.

a. True

b. False

4. If any RCPs stop during venting, venting should continue while establishing natural circulation flow until a venting termination criterion is exceeded,

a. True

b. False

5. If the voids are primarily noncondensable, then the total void volume will markedly decrease with increasing pressure, and the PRZR level will drop,

a. True

b. False VII-11-19 rev. 1

ANSWERS

1. d. Cooldown rate

2. b. False. If noncondensable gases are suspected, RCPs should NOT be started.

3. a. True

4. a. True

5. b. False. If the voids are primarily CONDENSABLE, then the total void volume will decrease. (If the voids are noncondensable, there may be a slight decrease in void volume due to compression.)

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VII-11-20 rev. 1

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/; - TASK: Respond to Voids in Reactor Vessel i ?

t/

Name Date Social Security No. Trainee No.

PERFORMANCE OBJECTIVE Given that the RVLIS indicates that the upper head of the reactor vessel is not full, eliminate reactor vessel voids. The RVLIS must indicate that the upper head of the reactor vessel is full. The PRZR level must be stable.

VERIFICATION OF SATISFACTORY COMPLETION o Completed Performance Test satisfactorily o Turned in Instructional Unit Feedback Sheet Instructor's Signature Date f

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VII-11-21 rev. 1

7 INTENTIONAL BLANK .

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i VII-11-22 rev. 1 l

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ML20215M874

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