ML20135G589
| ML20135G589 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/03/1985 |
| From: | Brooks B, Joseph Kelly, Newton D BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20135G587 | List: |
| References | |
| 74-1152414, 74-1152414-R, 74-1152414-R00, NUDOCS 8509190481 | |
| Download: ML20135G589 (332) | |
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r BWNP 20004 2 (9 84) l BABCOCK & WILCOX NUCLEAR POWER DivlSION TECNNICAL DOCUMENT !
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EMERGENCY OPERATING PROCEDURES i TECHNICAL BASES DOCUMENT j
! t i 74 - 1152414 - 00
- Dec.10 - Serial No., Revision No.
for i
The B&W Owners Group
! Operator Support Comittee i
1 l Ar*ansas Power & Light Company
( Duke Power Company
- Florida Power Corporation l General Public Utilities Nuclear Sacramento Municipal Utility District I
Toledo Edison Company Tennessee Valley Authority Washington Public Power '
' i O 8509190481 850911 l
DR ADOCK O y7 '
PAGt 1
BwNP 20005 2 (9 84)
BABCOCK & WILCOX NUCLEAR POWER Olvisl0N NUdttR RECORD OF REVISION 74-11s2414-00 REV. NO. CNANGE SECT / PARA. DESCRIPTION / CHANGE AUTHORIZATION
- 00 Original Issue PREPARED BY m . j DATE 5 2.1 F T REVIEWED BY , DATE & 30 4 REVIEWED BY / DATE REVIEWED BY DATE [ d (jv /
APPROVED BY DATE 9/ /f f 1
DATE: 7-23-85
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BASCOCK & walCOX NUCLEAR power DIVISION NU SE TECHICAL DOCUMEllT
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Table of Contents Pace I. Introduction I.0 A. Purpose of This Document I.A-1 B. Scope of This Document I.B-1 II. Symptom Approach to Abnormal Transient Diaanosis and Mitication II.0 A. Philosophy of Symptom Approach II.A-1 B. P-T Relationship to Monitor Symptoms II.B-1 C. Five Control Functions to Regulate Symptoms II.C l III. Diagnosis and Mitiaation III.0 A. General Approach Overview / Entry Conditions III.A-1 B. Loss of Adequate Subcooling Margin III.B-1 C. Lack of Adequate Primary to Secondary
[
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Heat Transfer Excessive Primary to Secondaty Heat Transfer III.C-1 III.D-1
\' E. Steam Generator Tube Rupture III.E-1 F. Inadequate Core Cooling III.F-1 G. Cooldown Methods III.G-1 IV. Ecuinment Operation IV.0 A. RC Pumps IV.A-1 B. HPI/LPI/DHRS/CF Operation IV.B-1 C. MFW/AFW System Operation IV.C-1 D. Incore Thermocouples IV.D-1 E. High Point Vents IV.E-1 F. Containment Systems IV.F-1 G. Reactor Vessel Pressure / Temperature Limits IV.G-1 V. Specific Rules V.0 A. HPI/LPI Specific Rules V.A-1 B. MFW/ATW Specific Rules V.B-1 C. RCP Specific Rules V.C-1 VI. References VI.0
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List of Ficures Ficures II.B-1 Normal P-T Trace Following a Reactor Trip II.B-2 Inadequate Subcooling Margin II.B-3 Typical Overheating P-T Trace II.B-4 Typical Overcooling P-T Trace III.A-1 Vital System Status Verification Flowchart III.B-1 Lack of Adequate Subcooling Margin Flowchart III.C-1 Lack of Heat Transfer Flowchart III.D-1 Excessive Heat Transfer Flowchart III.E-1 SGTR Functional Flow Diagram III.F-1 Core Exit Fluid Temperature for Inadequate Core Cooling III.F-2 Inadequate Core Cooling Flowchart III.G-1 Cooldown Logic Diagram III.G-2 Plant Status vs. Cooldown Concerns III.G-3 Head Fluid Temperature Response While Active Vent is Open (Primary Pressure - 2200 psia)
III.G-4 Head Fluid Temperature Response While Active
'N Vent is Open (Primary Pressure - 1600 psia)
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) III.G-5 Head Fluid Temperature Response While Active Vent is Open (Primary Pressure - 1000 psia)
III.G-6 Head Fluid Temperature Response While Active Vent is Open (Primary Pressure - 400 psia)
IV.B-1 HPI Throttling Limit (for High Flow Line) -
for DB-1 IV.G-1 Subcooling Margin Limit and Reactor Vessel RC Pressure / Temperature Limits List of Tables Tables IV.G-1 RC Temperature Measuring L ice for Determining RV Pressure Temperature Conditions A
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SABCOCK & WILCOX NUCLEAR POWER DIVISION NUMSER r~x 74-1152414-00
/ TECHNICAL DOCUMENT N ,i List of Acronyms / Abbreviations ADV -
Atmospheric Dump Valve AFW -
Auxiliary or Emergency Feedwater ANO-1 -
Arkansas Nuclear One Unit 1 ATOG -
Abnormal Transient Operating Guidelines ATWS -
Anticipated Transient Without Scram BWST -
Borated Water Storage Tank CF -
Core Flood CFT -
Core Flood Tank CR-3 -
Crystal River Unit 3 DH -
Decay Heat DHR -
Decay Heat Removal DHRS -
Decay Heat Removal System ECC -
Emergency Core Cooling
<~ ECCS -
Emergency core Cooling System g )EFW Emergency Feedwater
^'
EOP -
Emergency Operating Procedures ERV -
Electromatic Relief Valve FA -
Fuel Assemblies FW -
Feedwater gpm -
Gallons per minute HPI -
High Pressure Injection HPV -
High Point Vent I/C -
Incore ICC -
Inadequate Core Cooling ICS -
Integrated Control System IST -
Integrated System Tests LCO -
Limiting Condition for Operation LOCA -
Loss of Coolant Accident LOFW -
Loss of Offsite Power LPI -
Low Pressure Injection p
s
)MFW -
Main Feedwater y_j MU -
Makeup DATE: 7-23-85 PAGE 111
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SADCO!K & WILCOX NUCLEAR power DIVISION NUMBER 74-1152414-oo TECHNICAL DOCUMENT List of Acronyms / Abbreviations (Cont'd)
MSSV -
Nil-Ductility Transition NNI -
Non-Nuclear Instrumentation NPSH -
Net Positive Suction Head NSS -
Nuclear Steam Supply NSSS -
Nuclear Steam Supply System ONS 1 -
Oconee Nuclear Station Unit 1 ONS 2 -
Oconee Nuclear Station Unit 2 ONS 3 -
Oconee Nuclear Station Unit 3 OTSG -
Once Through Steam Generator PORV -
Pressurizer Power or Pilot Operated Relief Valve PSIG -
Pounds per Square Inch Guage j P-T -
Pressure versus Temperature PTS -
Pressurized Thermal Shock Pzr -
Pressurizer RB -
Reactor Building or Containment RBS -
Reactor Building Spray RC -
Reactor Coolant Pump
- RCS -
Reactor Protection System RTD -
Resistance Temperature Detector RV -
Reactor Vessel SBLOCA -
Small Break Loss of Coolant Accident SCM -
Subcooling Margin SER -
Safety Evaluation Report SFAS -
Safety Features Actuation System SFRCS -
Steam Feed Rupture Control System SG -
Steam Generator Tube Rupture SPND -
Self Powered Neutron Detector T,y, -
Reactor Coolant Average Temperature TAP -
Transient Assessment Program i
7-23-85 IV DATE: PAGE
BWNP 20007 3 (9 84)
SASCOCK & WitCOE NUCLEAR POWER DIVISION NUM8tt 74-11s2414-o0 f' TECHNICAL DOCUMENT List of Acronyms / Abbreviations (Cont'd)
TBD -
Emergency Operating Procedure Technical Bases Document TBV -
Turbine Bypass Valve T/C -
Thermocouple Tc -
Reactor Cold Leg Temperature !
T -
Reactor Cold Leg Temperature cold T -
Core Temperature hot Th -
Reactor Hot Leg Temperature T -
Reactor Hot Leg Temperature hat ,
TRACC -
Tube Rupture Alternate Control Critoria ;
TMI-l -
Three Mile Island Unit 1 T sat -
Saturation Temperaturs >
WNP-1 -
Washington Nuclear Projects Unit 1 l
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SABCOCK ( WILCOX NUMSER NUCLEAR POWER DivtSION 74-1152414-oo l TECHNICAL DOCUMENT t
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i i PART I INTRODUCTION i
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SASCOCK & WILCOX NUmtER NUCLEAR POWER DIVISION
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TECHNICAL DOCUMENT Part I Introduction This document was developed primarily to establish a technical bases format that provides an efficient vehicle for document i maintenance and periodic updates to address new issues and opera-
! ticnal methods on a generic bases.
i BACKGROUND mm ttee although separate entities within the W Owners l Group, are pursuing two important programs that are interre-lated. The Operator Support Committee is tasked with addressing j the open issues in the ATOG Safety Evaluation Report (SER), some of
. which require a basis from new analyses. The Analysis Committee is l
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- , managing the development of an. Integral System Test (IST) Program j to gather data on Reactor Coolant System (RCS) transient phenomena
, which will be used for benchmarking analysis codes. Should the benchmarking result in substantive code modifications, there is a potential for modifications to the guidelines provided in the ATOG program.
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This document provides a mechanism for presenting the current +
) analytical bases for abnormal transient response and the current f recommended methods for diagnosing and mitigating the consequences of abnormal transients at B&W plants. ,
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Chaoter I.A Purpose of This Document There are four main purposes for developing this Technical Bases Document (TBD). These purposes are summarized as follows:
PURPOSE #1 - To provide the bases for operator actions for miti-gating abnormal transients using plant symptoms.
WHY - To assist utilities in maintaining emergency operating procedures (EOP).
HOW - By describing the basic heat transfer symptoms and control functions used to diagnose and mitigate abnormal transients.
- By giving operating guidance for key equipment and systems used for core cooling based on:
- a. equipment design limitations and
- b. expected system performance.
- By giving guidance for restoring stable plant conditions.
- By giving guidance for diagnosing symptoms.
- By providing the bases for recommended guidance.
PURPOSE #2 - To provide a consistent technical bases for operation of nuclear plants with B&W aupplied NSS systems.
WHY - To facilitate regulatory review.
- To provide a common ground for utilities to exchange opera-ting experience and ideas.
HOW - By providing a document applicable to all nuclear plants with B&W supplied NSS Systems.
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BABCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION 74-11s2414-oo TECHNICAL DOCUMENT l
PURPOSE #3 - To provide an efficient vehiclo for document mainten-ance.
Why - So that the document will not require changing every time small plant modifications are made. l
- So that timely changes can be made.
- So that changes can be made economically.
- So that the document will be kept up to date and therefore have high credibility.
- To address new issues and operating methods (e.g., IST results, ATOG SER open issues, TAP results, etc.).
HOW - By providing one generic document applicable to all nuclear plants with B&W NSS systems.
- By making a "high level" document. The document discussions will avoid plant specific design detail. Document discus-sions will tend to be in terms of operational functions which are common among all plants with B&W NSS systems.
- The document format is designed so that revised pages can be inserted with minimum perturbation to the remainder of the document.
- Related topics tend to be discussed in one localized section of the document.
PURPOSE #4 - To consolidate related information.
WHY - So that material is easily referenced.
- So that related facts can be easily drawn from the document.
- So that related facts are not overlooked.
HOW - By formatting concisely and rigidly.
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SABCOCK & WitCON NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 1naoter I.B Scope of This Document The Technical Bases Document (TBD) addresses operations related to core cooling and reactor building integrity using the B&W supplied Nuclear Steam Supply System and systems which interface directly with it by using symptoms of abnormal heat transfer to diagnose and mitigate transients. This document is designed to provide the technical bases for this theme.
The TBD is written for the writer of plant specific emergency operating procedures (EOPs) and deals with individual subject matter. It does not interrelate all operating procedures and operator actions. The TBD user will have to apply the subject matter in the appropriate place in the user's EOPs. In other the material is being provided as " textbook information. "
m)words, The TBD user must be familiar with the plant specific EOPs and plant design details because the TBD is written on a functional level and does not provide plant specific detail. Unless super-ceded by guidance in this document, normal limits and precautions always apply.
The TBD is divided into six parts. Each part addresses a specific subject matter associated with symptom oriented diagnosis and
! mitigation. Although duplication of subject matter occurs between parts, the user of the document must read and understand each
- previous part before continuing with any subsequent parts. The l contents of Parts II - VI are summarized on the following pages.
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BONP 30007 3 (9-84)
SABCOCK & WILCOX NUMBER NUCLEAR POWER DIVl$10N TECHNICAL DOCUMENT l Part II Part II, entitled " Symptom Approach to Diagnosis and Mitiga-tion", provides a general discussion of the symptom approach. The specific questions addressed in Part II are:
- Why use symptoms to detect abnormal transients?
- Why were certain symptoms chosen?
- What are the basic priorities should more than one symptom be applicable to a given situation and what are the bases for these priorities?
- How are the chosen symptoms monitored?
- What causes the chosen symptoms to occur?
- What functions should the operator control to counter the symptoms?
The information provided in Part II provides an overview of the details provided in subsequent parts.
Part III O Part III, entitled " Diagnosis and Mitigation", is designed to discuss the basic sequence required to diagnose and mitigate trans-ients. The specific question addressed in Part III are:
- What is the overall sequence for diagnosing and mitigating abnormal transients and the bases for the sequence?
- What is the general sequence used in response to each symptom and the bases for the sequence?
- What are the basic priorities should more than one action be applicable to a given situation and what are the bases for these priorities?
- What are the general sequences for treating inadequate core cooling, steam generator tube rupture and plant abnormal cooldown?
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- What are the general guidances for entering and exiting sequences?
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- What are the bases for the specific actions used in a sequence?
Part IV Part IV, entitled " Equipment Operation", is designed to explain operational methods of key systems and equipment used in diagnosing and mitigating transients. Parts II and III view diagnosis and mitigation from an overall plant behavior. Part IV focuses on specific systems and equipment. This section:
- Discusses design limits which can be challenged in the course of mitigating transients and discusses how to prevent exceeding the limits.
- Discusses when and why certain systems and components need to be operated in a special way.
- Explains methods to maximize usefulness of the systems
)
\j and components.
i Part V Part V summarizes specific rules associated with diagnosis and mitigation. These are singled out and put into this section for emphasis.
Part VI Part VI is a listing of the references used in development of this document.
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Chaoter II.A Philosophy of Symotom Anoroach 1.0 Introduction This chapter describes the symptom approach, why the symptom approach should be used ant -hat the symptoms are. The symptom approach avoids drawnacks inherent in the event oriented approach for trancient diagnosis and mitigation.
The event oriented approach uses mitigating procedures which are tailored for each initiating event. Inherent complica-tions with event oriented procedures surfaced in the after-math of the March 1979 accident at the Three Mile Island Unit 2 Nuclear Plant. The complications are:
a) Writing a procedure to address every conceivable m initiating event is impractical. In addition every possible initiating event cannot be defined; some will
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be overlooked. Consequently, for some events there will be no procedure.
b) Operators will have to try to follow more than one procedure simultaneously if multiple failures occur.
Procedures may not be structured to facilitate simul-taneous use.
c) The operator must immediately and correctly diagnose *:he initiating event. The type of event is not always immediately apparent, especially if multiple failures occur. Consequently, the operator may lose valuable time following wrong procedures. In addition, once the error is disecvered the operator will need to transfer from the wrong procedure to the correct procedure.
Immediate decisions will need to be made without the aid l of a procedure as to which steps should be reversed and which should not when transferring to the new procedure.
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I ( A new type of procedure was implemented to avoid these
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complications. The procedure uses the symptom approach.
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l B AB COCK & WILCOX NUMBE NUCLEAR POWER DIVISION TECHNICAL DOCUMENT This approach requires the operator to monitor directly observable symptoms of upsets in heat transfer from the core to the SG, then mitigate the abnormal transient by restabil-izing the same or an alternate heat transfer process.
2.0 Stable core heat removal is the basic coal of transient mitication.
The normal method of core cooling is by transferring core heat to the RC, then transferring the heat from the RC to the secondary steam system via the SGs. If the rate of heat transfer from one medium to the next is equal, then stable heat transfer exists. When this process is disrupted a transient occurs.
The symptom approach uses an upset in heat transfer symptom as a symptom of an abnormal transient. Consequently, to mitigate an abnormal transient the operator must correct or circumvent the upset. Knowing the cause of the upset or whether it is caused by one failure or a combination of failures is not important to the ability to mitigate the abnormal transient.
i 3.0 Three symptoms of upsets in heat transfer are used The three symptoms of upsets in heat transfer are:
a) lack of adequate subcooling margin b) inadequate primary to secondary heat transfer c) excessive primary to secondary heat transfer 3.1 Lack of adecuate subcoolina marain:
The core transfers heat to the RC. As long as the core is
( covered with RC, sufficient heat transfer from the core will l occur to keep the core adequately cooled. As long as the RC is subcooled the core will be covered. Therefore, the operator should assume that if a sufficient margin to saturation does not exist, the core has a potential of not
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SINP 20007 3 (9-84) [
BA8 COCK & WRCCX ;
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NUCLEAR POWER DIVISION '
74-1152414-00 TECHNICAL DOCUMENT being adequately cooled. The operator should take appro-priate actions to assure adequate core cooling, including
- restoring the margin to subcooling. (Refer to Chapter
, III-B, Loss of Subcooling Margin.)
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3.2 Inadeauate/ excessive crimarv to secondary heat transfer. !
i If too little or too much heat is transferred to the SGs l the heat transfer process is upset. The RC transfers heat l to the secondary system via the SG. If'not enough heat is j i
transferred the RC will increase in temperature. If too j i
much heat is transferred the RC will decrease in tempera- l ture. Therefore, if the operator detects uncontrolled l heatup or cooldown symptoms he should take the appropriate l actions to stop the overheating or overcooling.
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The corrective actions to counter the loss of subcooling l margin, inadequate heat transfer and excessive heat transfer f are designed so that the upset in heat transfer can be j corrected independent of the initiating event. Conse- l
- l. quantly, the heat transfer process can be stabilized without j correcting the failure or failures. Once the heat transfer j i process is stabilized core cooling is assured and time is !
$ available for identifying and correcting the failure. I I
l l The three symptoms have features (discussed in Chapter !
II.B) making them readily recognizable. Consequently, f
- the operator should be able to quickly ascertain the upset and commence mitigating actions. l l
i 3.3 Priority of Svantons j
- Treatment for a loss of SCM has the highest priority of the {
three symptoms of upsets in heat transfer because as long as the RCS remains subcooled, adequate core cooling is as- {
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' mured. Once core cooling is assured, the next concern is to I t
control primary to secondary heat transfer. ;
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S ABCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION TECHNICAL DOCUMENT Therefore, treatment of a lack of primary to secondary heat transfer, along with its counterpart, excessive primary to secondary heat transfer are second priority symptoms.
4.0 Steam Generator Tube Ruoture A steam generator tube rupture (SGTR) is an event that the operator should identify and treat as a specific event in addition to treating the symptoms. An exception is made in treating the SGTR as an event in addition to upsets in heat transfer because the SGTR has unique indications allowing it to be easily identified and because by treating the SGTR also as an event, the operator can significantly reduce radiation release, improve core cooling and minimize waste water management problems. Consequently, this event is addressed separately in Chapter III.E.
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i Chanter II.B P-T Relationshin to Monitor Symot' oms 1.0 Introduction This chapter discusses how symptoms are monitored.
The main tool used to monitor symptoms of changes in primary ,
to secondary heat transfer is the Pressure-Temperature (P-T) l relationship as shown in Figure II.B-1. Changes in heat i transfer are observed by monitoring variable pressure and temperature measurements. These pressure and temperature
! variables are: ;
I a) Reactor Coolant Hot Leg Temperatura (each loop) ,
j b) Reactor Coolant Pressure, c) Reactor Coolant Cold Leg Temperature (each loop),
% d) Incore Thermocouple Temperature, e) Steam Generator Pressure (each SG).' l
! l In addition, the relationship of these variable measurements
! to fixed limits associated with the variable pressure and temperature measurements are also monitored. - These fixed l limits are:
- a) Saturation Line, i
b) RC Subcooling Margin Limit,
- c) Post-Trip Window.
By monitoring these variable parameters 'and limits, their relationship can be readily observed allowing a quick transient diagnosis.
l Continual monitoring of the P-T variables provides a real time relationship of the variable measurements which demonstrates both the present plant conditions and the
- trend 'n i plant conditions. Knowing the present plant I conditions enables the operator-to take actions applicable
! to the existing conditions. The trending information is 23-D PAGE DATE:
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BASCOCK & WILCOX NU" I NUCLEAR POWER DIVISION TECHNICAL DOCUMENT valuable in diagnosing the transient, predicting what actions may need to be taken, judging whether or not the mitigating actions are working, and determining when the plant has reached a stable condition.
2.0 Fixed P-T RelationshiD 2.1 Saturation Line The saturation line indicates the pressure and temperature combination where water changes to steam and vice versa.
This is applicable to both primary and secondary conditions.
2.2 Subcoolina Marcin Limit The subcooling margin (SCM) limit presents pressure and temperature combinations which are more subcooled than the saturation line. The intent of this limit is to assure that the RC is subcooled. The extra subcooling (i.e., the area between the saturation line and the SCM limit) is chosen based on the ability to accurately measure the RC pressure and temperature (instrument errors) and for pressure and temperature variations from the point of measurement (e.g.,
the elevation head). If the SCM limit is violated, the assumption should be made that the RC is no longer subcooled even though it may be. This limit applies only to RC conditions.
2.3 Post-Trio Window The post-trip window encloses an area of the P-T relation-ship where the RC pressure and temperature combination will normally stay after a trip. The minimum RC temperature and pressure boundaries of the window were compiled from a review of actual reactor trips and computer simulations.
The normal post-trip cooling of the RC by the secondary system should not cause the RC pressure and temperature to go outside the post-trip window.
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BABCOCK & wlLCOR NUCLEAR POWER DIVISION NU"8EE !
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) TECHNICAL 00CNNENT The maximum RC pressure is based on the pressurizer power operated relief valve (PORV) pressure setpoint. Following a reactor trip the RC pressure will not normally increase to f the PORV setpoint. However, if it does rise to the PORV setpoint, the relief valve should prevent the RC pressure j from exceeding its setpaint.
The upper temperature limit is based on the high temperature j reactor trip setpoint and/or the SCM limit. A normal reactor trip will result in an overall decrease in RC temperature.
3.0 Variable P-T Relationshio The variable parameters are used to provide the following information.
3.1 Recoanizina Abnormal Transient Conditions Durina Normal j' Post-Trio Transient Conditions During power operation, the primary to secondary heat transfer process is a stable process. The process is
- upset when the reactor trips, causing a transient condition l as a normal transition occurs to a post trip stable heat i transfer condition. This normal transition could mask a i simultaneous abnormal transient. However, these conditions can be recognized during the post trip transient.
- i l The shape of the RC P-T trace for a normal transient l following reactor trip from power operation when the
- pre-trip Ta ve is above post-trip-Tave is similar to the one !
- shown in Figure II.B-1. The magnitude of the transient [
becomes smaller as the initial power level decreases. The
{
l dip of the curve is due to cooldown of the RCS to near Tsat l of the SGs for the turbine bypass valve (TBV) setpoint.
- j. The cooldown results in coolant shrinkage, which results in i a pressurizer outsurge and pressure reduction. After the [
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SABCOCK & wlLCOX NUM8tk NUCLEAR POWER OtVISION 74-1152414-oo TECHNICAL DOCUMENT RCS reaches a temperature slightly above Tsat of the SGs, the RC will repressurize and stabilize due to the MU pumps partially 2.efilling the pressurizer and to energizing of the pressurizer heaters. Depending on prior power history, the low point of the dip will have different values, but the characteristic shape of the trace will remain the same when the reactor trips when the RC Tave is greater than the post-trip Tave-If the transient is normal, the RC temperature vs. RC pressure relationship will stabilize at one of two locations inside the post-trip window depending on RCP status. When the RCPs are off Tcold will be essentially the same as the SG temperature but Thot will be greater. The value of Thot will depend on the amount of decay heat. When the RCPs are ,
running, Thot and Tcold will stabilize within a few degrees of each other after trip and T cold will be within a few degrees above the SG temperature.
In the secondary system, the steam pressure will initially rise to the TBV or MSSV setpoint and steam temperature will decrease to saturation temperature. l L
A P-T plot for a normal trip is shown in Figure II.B-1.
Some parameter changes associated with a normal trip are as follows:
- 1) Hot and cold leg temperatures stabilize in 2-3 minutes.
- 2) RC pressure stabilizes in 5 to 6 minutes.
- 3) Tcold will be nearly equal to saturated steam tempera-ture indicating that RC is transferring heat to the steam generators.
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- 4) Steam pressure stabilizes in 2 to 3 minutes.
- 5) RC SCM increases.
3.1.1 Post-Trio Window The post-trip window is an aid for determining if systems are operating correctly after a trip. If the RC temperature and pressure path goes outside the post trip window then an abnormal transient is probably occurring. The post trip window is not an absolute determinant as to whether or not an abnormal transient is occurring. However, if the pressure vs temperature plot goes outside the post-trip window the transient is likely abnormal and the operator should proceed with abnormal transient mitigation proce-dures.
Conversely, if the RC pressure and temperature plot remains inside this window or if the transient path goes outside this window slightly but returns, then the transient is likely going as expected and the core cooling with SG heat transfer is correct. The operator need only verify proper plant conditions exist to identify any abnormal transient 4 which may be masked by the normal plant RC pressure and temperature behavior or which needs to be acted on before the RC pressure and temperature conditions go outside the post trip window. If the plant conditions are not as expected, the operator should take appropriate corrective actions. One such transient which combines both of these situations is the excessive MFW transient. Excessive MFW will cause an overcooling trend. However, this trend is masked by the post-trip RC cooling characteristics which initially seem to be normal. In addition, the excessive MFW i can rapidly overfill the SGs requiring operator actions to I
prevent overfilling the SG or verifying automatic actions l
.\'
have occurred while the RC pressure and temperature rela-tionship is still inside the post-trip window.
-2 N
- DATE: PAGE
BWNP40007 3 (9 84)
SASCOCK & WILCOM NUM E NUCLEAR POWER DIVISION TECHNICAL DOCUMENT An abnormal transient may also be indicated by secondary system steam pressure and steam saturation temperature.
Generally, if steam pressure falls below the saturation pressure corresponding to the lower temperature of the post-trip window after trip, a failure has occurred. The SG pressure should stabilize at the TBV setpoint.
3.2 Identifyina Uesets in Heat Transfer The P-T relationship is used to detect the three main symptoms of an upset in the heat transfer process. These symptoms are:
- Loss of Subcooling Margin,
- Lack of Adequate Primary to Secondary Heat Transfer, and
- Excessive Primary to Secondary Heat Transfer.
Loss of subcooling margin determines the status of the heat transfer medium (the RC) which is used to transfer the heat to the SGs and to remove heat from the core.
Lack of adequate primary to secondary heat transfer and
( excessive primary to secondary heat transfer are mutually l
l exclusive, i.e., the RCS cannot be overcooled and overheated l at the same time. However, lack of SCM can occur simultane-ously with either of these, l
3.2.1 Indications of a Loss of Subcoolina Marcin The RCS P-T relationship will clearly indicate when a loss of subcooling margin (SCM) occurs (see Figure II.B-2). The SCM is lost when the RCS P-T relationship becomes less subcooled than the loss of SCM limit.
If a loss of SCM occurs without a concurrent indication of excessive or inadequate primary to secondary heat transfer, the P-T relationship will show a decrease in RC pressure
~
- DATE: PAGE
BWNP 20007 3 (9 84)
BASCOCK & WILCOE NUMSt
( NUCLEAR POWER DIVI $lON TECHNICAL DOCUMENT with little or no change in RC temperature and secondary saturation temperature. This symptom would be caused by a ,
loss of RC inventory or pressure control. i
. t l !
A loss of SCM can be concurrent with and caused by inade- ;
l quate or excessive primary to secondary heat transfer. The i l
i j secondary steam saturation temperature can be used as j j supporting information to determine what has caused the lack f of SCM. If the cause was excessive heat transfer, then the f RCS T eold will have el sely foll wed the secondary steam j saturation temperature as it decreased. In the event of I inadequate primary to secondary heat transfer, the SG and j j RCS will have become uncoupled with the secondary steam j saturation temperature slowly decreasing and the RCS !
- repressurizing and beginning to heat up.
1 2
3.2.2 Indication of Inadeauata Primary to Secondary heat Transfer
- If only inadequate heat transfer occurs, the P-T relation- ,
ship will show both the RC pressure and temperature increas- l ing. The resulting swell will increase the pressurizer
! level causing the RC pressure to increase. The pressurizer
- spray will try to offset the pressure increase. However,
the rate at which this equipment :nn reduce RC pressure is
! limited. Consequently, the faster the heatup the larger the
- pressure increase per degree of heatup. When the RC l l pressure becomes limited by the PORV or code safeties, the
! SCM will be lost as the RCS oontinues to heatup. A typical overheating trend is shown in Figure II.B-3.
)
l The P-T relationship will depend on initial conditions. The possible RCS initial conditions during a loss of heat transfer are as follows:
A. RCB'Subcoolad with acDs On When a lack of heat transfer begins, the SG Tsat and RC temperature will begin to diverge with the RCS P-T l t __ _ _ i 1-n-so 1
m ,
_ _ _.__ _ _ _ _ _ __ __ ____ _ _ __ _ _ _ _ .___ _ _j
BWNP.20007 3 (9-84)
SABCOCK & WILCOX NUCLEAR POWER DIVISION NUM E TECHNICAL DOCUMENT increasing in temperature and pressure and the SG Tsat decreasing. The hot and cold leg RTDs along with incore T/C will begin to heatup. Indications of FW flow, SG 1evel, and steam and FW system valve positions may be used to confirm a loss of heat transfor.
B. RCS Subcooled With RCPs Off If natural circulation has been established before a loss of heat transfer occurs, then the secondary Tsat will almost coincide with the primary Tcold. Thot is expected to be about 50F higher than T cold within a few minutes after reactor trip. This delta T will decrease on the order of about 10F per each 1% decrease in decay heat. Once the lack of heat transfer occurs the incore T/C temperature indication will begin increasing causing a higher than expected core delta T between incore T/cs and Tcold. Other control room indications of FW flow, SG level, and steam and FW system valve positions may also confirm the reason for a loss of natural circula-tion.
The best single indication of a loss of natural circula-tion flow when the RC is subcooled is a divergence developing between the incore T/C and T When the hot.
flow is lost, the incore T/C will begin a continual increase toward saturation. The rate will depend on the amount of decay heat. T hot indications may also increase but can actually decrease and begin to converge with T cold. In any case, T hot will n t increase as rapidly as the incore T/cs and the two indications will diverge. Another indication of loss of natural circula-tion is a "decoupling" between Tcold and SG pressWe (secondary Tsat). If Tcold ceases to follow SG pres-sure, then natural circulation flow has been lost.
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SASCOCK & wtLCOR N 8" NUCLEAR POWER DIVISION ,
% TECHNICAL DOCUMENT i
C. RCS is Saturated ,
1 With the RCS saturated, the best-indication of a loss of i natural circulation flow or interruption of boiler condenser cooling is a trend of incore T/C temperature j l
vs. RC pressure increasing away from the SG Tsat al ng the saturation curve. The divergence between Thot and the incore T/cs may not develop significantly. Heat i l
transfer can also be lost due to the RCS P-T decreasing along the saturation curve below the SG P-T due to a l I4CA.
! 3.2.3 Indication of Excessive Primary to Secondary Heat Transfer If only excessive heat transfer occurs, the P-T relationship will show both the RC pressure and temperature decreasing.
The resulting RC contraction will lower the pressurizer level causing the RC pressure to decrease. RC MU and pressurizer heaters will attempt to correct the RC pressure drop. However, the rate at which this equipment recovers RC l pressure is limited. Consequently, the higher the cooling
} rate, the larger the pressure drop per degree of cooling.
l The RC will remain subcooled unless the overcooling-empties j the pressurizar. If the pressurizer empties, the RC will approach saturated conditions.
f L
The P-T relationship is the quickest and most accurate means I
of determining that an overcooling is occurring. A typical overcooling trend is shown in Figure II.B-4. Immediately after reactor trip, T hot and Teold "Ill C""V**9* ""d l
i approach SG saturation temperature. This will occur if the i RCPs are running. If the RCPs are not running and the RC is subcooled, T and Tcold will not come together; rather, a hot ,
temperature difference will develop across the core which is !
necessary for- natural circulation of the RC. The primary j T and T cold P-T relationship will continue to decrease in hot N~'" ##* "
DATE: -PAGE
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SASCOCK & WitCOX l NUM E j NUCLEAR POWER DIVISION TECHNICAL DOCUMENT ,
both pressure and temperature. The most positive indication that the transient that is occurring is an overcooling rather than a LOCA is the fact that the secondary Tsat " 1 be decreasing rapidly. The primary T cold will be following I close behind SG temperature indicating that the primary and secondary are closely coupled in the overcooling. With an excessive overcooling the primary P-T relationship will begin to decrease in pressure very rapidly after the pressurizer surge line is drained and the RCS goes into saturation. The RCS P-T relationship will continue to :
follow the secondary Tsat al ng the saturation curve.
I 3.3 Indication of Primarv to Secondary Couolina )
When primary to secondary coupling exists with either forced )
or natural circulation, T cold will be about equal to the saturation temperature in the SGs and Thot will be about equal to the incore T/Cs. If the RCPs are lost, Thot " I increase as necessary to develop the driving head required for flow (by developing a density difference between T hot and Tcold). If the loss of RCPs occurs at a high enough RC pressure and decay heat, this may result in lifting the PORV even though a Joss of primary to secondary heat transfer has not occurred.
The best indications that subcooled natural circulation has started are the coupling between T cold and the SG pressure (saturation temperature), the relationship of Thot and the incore T/Cs, and the temperature difference between Thot and Tcold. The relationship of T cold vs. steam pressure should f remain a few degrees to the right of the saturation curve, the incore T/C temperature indication should track Thot within approximately 10F, and the temperature difference between Thot and Tcold should be approximately 5GF to 60F (maximum decay heat). When steam pressure is changed, T cold should follow but a time delay in T cold response (on the DATE: PAGE
l 8%NP 20007 3 (9 84)
SABCOCK & WILCOE NUCLEAR POWER DIVISION NU"#EE 74-11s2414-00 i TECHNICAL DOCUMENT order of a few minutes) can be expected due to low loop flow rates.
l The best indication that saturated natural circulation has l
{ started, is coupling between Tcold and the SG pressure l
! (saturated temperature) . The relationship between Thot and 6 the incore T/cs and the temperature difference between T hot and T cold re n t good indications of sat urated natural l circulation flow.
I The incore T/C temperature indication will track T hot during I saturated natural circulation flow. However, if the RC is saturated, T will be similar to the incore T/C tempera-hot
- ture indication even if natural circulation flow does not i exist.
The temperature difference between Thot and T cold can vary l
from about 50F to 0F depending on how much of the core heat is transferred to the RC as latent heat of vaporization.
Natural circulation flow will regulate itself. That is, as j
j the heat source (decay heat) decreases, the delta T (Thot -
Tcold) will decrease, and there will be less driving head available; therefore, flow will decrease.
t If only one SG is operating during natural circulation only T
hot in the operating loop will indicate core outlet
{
temperature; Tcold on the operating SG will be approximately equal to Tsat in the operating SG; Tcold in the isolated l
j SG may ngj; be equal to Tsat in the isolated SG; it will l probably be colder due to ambient losses and due to cooler
! injection water (peal injection, MU, HPI); (T hot -Tcold) "
the operating SG may be 10F higher than the 50F to 60F delta T expected with two operating SGs.
l 7-23-85 "*
DATE: PAGE I
,,,,-.m- .,%.-- ..,_...-._,,,,__.-,%__-.,,,em-y. _ .., ,.,my , ,- , , , _ .
i i
O !
l Figure 11.8-1 NORMAL P-T TRACE FOLLOWING A REACTOR TRIP !
l i
i
[
TCOLO !
t o TH0T POST TRIP ;
d '
WINDOW i
I as b TURBINE l U BYPASS g U i y __
lSETPOINT 2
SATURATION S EAM SUBC00 LING !
PRESSURE !
MARGIN LIMIT LIMIT i i
I
(
Temperature l l
i
(
D00. NO. 74-1152414-00 t
l
O i
l Figure ll.B-2 INADEQUATE SUSC00 LING MARGIN POST TRIP WINDOW TH0T SUBC00 LED
, REGION t
g T l SUPERHEAT
, BP '
I
$ SETPOINT
~
STEAM 7 ~~~
PRESSURE LIMIT , SATURATION ;
SUBC00 LING MARGIN LIMIT Temperature 1
- O D00. NO. 74-1152414-00
l Figure ll.B-3 TYPICAL OVERHEATING P-T TRACE I
l POST TRIP r TH0T WINDOW l
l E I E TURBINE l E BYPASS l
A
- n. SETPOINT 7
O _ 7 _ _ _ __ _ _
STEAM SATURATION PRESSURE LIMIT SUB000 LING MARGIN LIMIT l
l Temperature l l i l
l l
l \
ooc. no. 74-iis2 Sin-oo
, ,_,_-..m ._ r _ _ , . -.....,__.g., _ . . , ., ----,w_,.n,ge.m .m c%w, , . . , .
O Figure 11.8-4 TYPICAL OVERC00 LING P-T TRACE i
POST TRIP - THT WINDOW I
b TURBINE I
3 BYPASS 2 SETPOINT O S 1 E ,.
PRESSURE SATURATION LIMIT SUBC00 LING NARGIN LIMIT Temperature
)
I D00. NO. 74-11524i4-00 i
0 .
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BWNP 20007 3 (9 84)
SASCOCK & wlLCOR "00" NUCLEAR POWER DIVISION 74-11s2414-o0 TECNNICAL DOCUMENT f
Chanter II.C Five Control Functions to Reaulate Svantoms a 1.o Introduction The heat transfer process is controlled by five control functions. This chapter describes the five control functions and how they affect the heat transfer process.
The five control functions are a) Reactivity Control b) Reactor Coolant Inventory Control
- c) Reactor Coolant Pressure Control d) Steam Generator Pressure Control i
e) Steam Generator Inventory Control 1.1 Umsets in the Heat Transfer Process are caused and Corrected
^
By Chances in the Five control Functions The normal method of removing heat from the core is to. -
transfer the core heat to the RC and then transfer the heat l from the RC to the secondary system. Abnormal transients occur when this heat transfer process is upset. This heat transfer process is controlled by the five control func-i tions.
l Abnormal transients are caused when one or more of the control functions are out of control. The symptoms of abnormal heat transfer are.used to determine which function ,
or functions are out of control. Each of the control functions affect the heat transfer process in a specific way. When a failed control function is identified the [
operator can attempt to regain control of'the function. If
! he can regain control he will be able to restabilize the normal heat transfer process. In the event the failure ,
cannot be corrected, operating the remaining control ;
functions in an off-normal mode may be required in order to compensate for the failed control function.
l DATE: 7~23-85 pAgg II.C-1 m.,.,,,,em
BWNP 20007 3 (9 84)
S ABCOCK O WitCcX NUCLEAR POWER OlVISION NUM B t t 74-1152414-oo TECHNICAL. DOCUMENT The failures may result in a loss of the normal method of heat transfer. The normal method relies on heat transfer to the SG for core heat removal. If SG heat transfer cannot be established a backup heat transfer process is available.
This process consists of transferring core heat to the RC , then transferring the high enthalpy RC to the RB. The flow path to the RB can be any opening in the RCS but the opening has to be large enough to remove the core energy.
Of course, this requires adding low enthalpy water to the RCS. This process is called HPI cooling and is discussed in Chapter IV.B.
A thorough understanding of how each of the five control functions affect the heat transfer process will:
A. Enable operators to manipulate equipment to regain heat transfer control without knowing exactly which equipment has failed. Different equipment may affect the same control function. It is of less importance which equipment is used and of more importance that the control function is being addressed. Consequently, proper heat transfer can be restored more quickly and accurately than if the operator had to identify a specific equipment failure.
B. Allow easier identification of what equipment has failed and, by doing so, isolate, remove from service, or repair the equipment.
C. Provide an understanding of the outcome of an action.
The operator actions will have some consequence on the heat transfer process. Consequently, knowledge of heat transfer process will allow judgments to be made about the general effects of operator actions.
O DATE: 7-23-85 PAGE II.C-2
SWNP 20007 3 (9 84) 8ASCOCK & WILCOR '
OU NUCLEAR POWER DivtS40N 74-118*414~
TECNNICAL DOCullENT 2.0 Details of How the Five Control Functions Influence Heat Transfer The following discussion explains how each of the control functions influence the heat transfer process.
2.1 Reactivity Control Reactivity control is normally automatic either by ICS rod control or by reactor trip. Reactor trip lowers the core heat output to the decay heat level. The operator must .
verify rod insertion and decreasing reactor power to ensure l the reactivity control systems function properly. After l the trip no more heat transfer control can be achieved by ;
use of the rods; unless the rods did not insert. If after j manual trip, reactor power is not decreasing (total failure to trip) the operator should begin boron addition and
! attempt a manual rod insertion, while performing local breaker trips to attempt to de-energize the control rod drive mechanisms. If more than one rod remains stuck out, i the operator should begin boration to increase the shutdown margin. t i
2.2 Reactor Coolant Inventory control l RC heat transfer and pressure can be affected by changes in
( the RC inventory (the volume of subcooled fluid in the :
l reactor coolant system) . The volume is affected by changes ,
t in mass and density of the RC. !
l The mass of RC can be varied by: LOCAs, changes in HPI or l MU flow, RCP seal injection, seal return, and letdown.
l Several ways also exist to vary the density of the reactor l coolant. Changes in 'the rate of heat transfer from the RC to the SG can cause the RC to cooldown when the SGs remove too much heat (low steam pressure, too much FW) causing density to increase and thus RC volume to decrease; 7-23-85 PAGE II.C-3 DATE:
l
i l
BCNP40007 3 (9-84)
B ASCOCK & WitCOX NUMBER NUCLEAR POWER DIVistON 74-11s2414-oo TECHNICAL DOCUMENT or the RC can heat up when the SGs don't remove enough heat (not enough FW) causing density to decrease and thus RC volume to increase.
Regardless of the cause, the changes in inventory in the RCS have two effects:
A. A loss of mass can affect the ability of the RC to transport heat from the core to SGs. If the RCPs are not running, steam can collect in the hot legs and possibly block natural circulation.
If the volume of liquid RC continues to decrease and the core is covered mostly by steam, then the core will retain the heat and heat up. Fuel failures can result if this situation is not corrected. Conversely, if the core is kept covered it will be adequately cooled.
B. A change of volume of subcooled RC can affect the pressurizer level, limiting the ability of the pressuri-zer to provide pressure control of the RCS.
The operator has indications to determine if RC inventory is sufficient for core cooling. Pressurizer level is an accurate measure of RC inventory only when the rest of the RC is subcooled (except for a rare condition when free non-condensible gas may exist in the loops; this condition will likely exist only after fuel failures have occurred) . This includes subcooling in the volume under the RV head and in the top of the hot leg pipes. These volumes could be saturated even though Thot, T cold and the incore T/cs indicate subcooled water temperatures. The other measure is the incore T/cs; if these read subcooled or saturation temperature, then enough mass exists in the RV to cover and cool the core. The incore T/cs cannot be used to indicate actual RC inventory. In addition to these indications, hot 7-23-85 II.C-4 DATE: PAGE
i B7NP-20007 3 (9 84)
SASCOCK & WRCOM NUCLEAR power DIVISION NumStB 74-11s2414-oo
! TECHNICAL BOCUMENT i l l
leg level measurements, if available, can be used when no
) 'RCPs are running and the HPVs, if installed, are closed. RV head level measurement, if available, can be used for checking inventory in the RV head. Restrictions may be imposed on the use of the level measurement instrumentation, e.g., all RCPs off. However, the conditions are plant dependent.
, 2.3 Reactor Coolant Pressure Control RC pressure control is required to keep the RC subcooled so that the coolant is in the best state to transfer heat from the core to the SGs. Usually, RCS pressure control is
} provided by the pressurizer. Use of pressurizer heaters and spray is the normal method of controlling RCS pressure when l a steam and water interface exists in the pressurizar. The purpose'of the heaters is to maintain the reactor coolant in
, a subcooled condition by maintaining RCS pressure greater than saturation pressure; the spray retards pressure
- I increases to limit operation of the PORV and safety valves.
! Neither the heaters nor spray have enough capacity to prevent large abrupt pressure changes, but they can moderate small changes. As a backup to normal spray, auxiliary spray from HPI or LPI may be available. The pressurizer vent or PORV can also be used to reduce pressure.
t i
- RCS pressure control by the pressurizer can be lost in four ways
A. Draining the Pressurizar If the pressuriser level drops sufficiently to uncover the heaters, the heaters cannot provide pressure control because no water is available to be boiled-by the heaters and the heater power will be terminated' by the pressuriser low level heater interlock.
l 7-23-85 PAGE II.C DATE:
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- TECHNICAL DOCUMENT 74-1152414-oo ,
l B. Fillina the Pressurizer Spray depressurizes the RC by condensing the steam in the pressurizer. If the pressurizer fills with water, the spray cannot be effective for depressurizing because the steam space is lost.
l C. RCS Voids The RCS pressure can decrease to the saturation pressure of the highest temperature in the system which could be the hot legs or the RV head region. If voids develop in one or more of these locations, they will hinder normal pressure control by the pressurizer.
D. Failure of Sorav and/or Heaters A failure of the spray or heaters in the pressurizer control system can also cause a loss of pressure control. If the spray fails open and cannot be closed, the system will depressurize. Depressurization may also occur if the heaters fail in the "off" mode. Failure of the spray in the "off" mode will limit the ability to depressurize. If the heaters fail "on", pressure increases will not occur because the spray will operate to provide a balance. However, if the spray is inoperable the heaters can cause the system to pressur-ize and cause coolant (steam) to be lost as long as liquid covers the heaters. When only steam covers the heaters they will no longer raise pressure. If the heaters fail "on" when they are uncovered, no liquid exists to cool them and they will burn out.
2.4 Steam Generator Pressure Control Heat transfer from the RC to the SG goes to both the steam
! and water in the SG. After reactor trip, the steam and FW in the SG are saturated and changes of steam pressure will cause a direct change in the saturation temperature of the DATE: 7-23-85 PAGE II.C-6
. . .- _.- _ _ - - - = - .
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! TECHNICAL DOCUMENT _
j steam and water. When the steam pressure is lowered, the heat transfer from the RC to the SG increases because the steam and water in the SG becomes a colder heat sink causing more heat to flow away from the reactor coolant. Two reasons combine to create the colder heat sinks first, the saturation temperature of the steam and water is reduced by
! lowering the steam pressure. Second, reducing the SG pressure causes the rate of boiloff to increase. The increased boiloff requires more FW flow to be added to maintain level. The FN inlet temperature is colder than the water already in the 80 thus its addition contributes to the colder heat sink.
Steam pressure can be lowered in two ways:
A. By releasing steam faster than it is produced (opening .
condenser dump valves, steam line break, opening ADV, etc.)
B. By spraying cold AFW into the steam space causing more steam to be condensed than is being formed.
conversely, steam pressure can be increased when below the main steam safety valve setpoints
- A. By releasing steam at a slower rate than it is produced
) (closing ADV's, condenser dump valves etc.).
j B. By reducing the spray of cold AFW into the steam space causing less steam to be condensed than is being j formed.
.l i
steam pressure cannot increase above the M88V setpoint because the safety valves will release steam at the same rate it is produced. Furthermore, several minutes after reactor trip, steam pressure will not increase above ADV or TSV pressure setpoints during automatic pressure control. l DATE: 7-23-85 PAGE II.C-7
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SABCOCK & WILCOX NUMS E R NUCLEAR POWER DlilSION '
74-11s2414-oo TECHNICAL DOCUMENT 2.5 Steam Generator Inventory Control The SG inventory is controlled to maintain a minimum inventory for a heat sink, to prevent overfilling the SG and for varying the effective elevation of the SG heat sink.
Heat transfer from the RC is transmitted to both the steam and the water in the secondary side of the SGs. However, most of the heat transfer occurs just below the water steam interface (in the nucleate boiling region) where the heat is used to change the saturated water to steam.
The AFW enters near the top of the SG. This FW flow will increase heat transfer from the primary to the secon-dary side as the FW flow is heated and turned to steam.
Since the AFW flow enters the top of the SG, the flow will have the effect of raising the SG thermal center higher in the SG. As the flow rate is decreased to zero the thermal center will decrease until it reaches the SG water level.
Consequently, the SG thermal center can be changed by changing the water level and by adding FW through the AFW nozzles. Refer to Chapter IV.C on controlling FW.
FW is cooler than the SG temperature. Therefore, if FW is added to the SG faster than the primary system can heat up the FW, the overall temperature of the SG will de-crease. This will result in the RC temperature decreasing if primary to secondary heat transfer exists.
AFW can have a proportionately largci cooling effect on RC for the same flow rate than MFW because of two effects:
a) it is colder (T inlet AFW is less), and !
b) it has a steam pressure reduction effect that MFW does ,
not have because it is injected into the steam space at !
the top of the tube bundle. (This pressure reduction DATE: 7-23-85 PAGE II.C-8
BTlNP 20007 3 (9 84)
BABCOCK & WitCOX NUMBER NUCLEAR POWER OlVISION 74-1152414-oo TECHNICAL DOCUMENT effect will also occur to a lesser extent with MFW if it is added through the AFW nozzles.)
The operator should ensure that the rate of FW addition is controlled properly to maintain the SG inventory. Level measurements in the SG (after trip) give a good indication of the SG inventory for control.
v l
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l DATE: 7-23-85 PAGE II.C-9
SWNP 20007 3 (9 84)
, SABCOCK & WitCOX
. NUCLEAR POWER DIVISION 4
74-1152414-00 )
TECHNICAL 00COMENT f
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Part III Diaanosis and Mitication .
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DATE: 7-23-85 PAGE III.0
9%NP 20007 3 (9 84)
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\
Part III Diacnosis and Mitication l
l
1.0 INTRODUCTION
l 1.1 Purpose The chapters in this part of the TBD provide generic Emergency Operating Procedure (EOP) technical bases for plant transients. The purpose of these technical bases is to provide sufficient information regarding the expected j NSSS response during plant transients such that the user cr.n maintain plant specific procedures for diagnosis and mitigation of plant transients.
, 1.2 Format (g Each of the chapters in this part is divided into three sections:
Section 1.0 " Introduction,".
Section 2.0, " Diagnosis and Mitigation" (General Operator Actions" in Chapter III.G).
Section 3.0, " Technical Bases".
The " Introduction" contains a discussion of the purpose of the chapter, particular concerns and objectives and, where applicable, possible causes of the symptom.
At the end of each chapter a flowchart of the recommended logic for mitigating the transient is provided. The flowcharts are written on a functional level to focus upon the function that is to be accomplished not the mechanics and the equipment necessary to perform that task. Each block on the flowchart contains two numbers. The upper numbers correspond to the subsections contained in the 7-23-85 PAGE III.1 DATE:
BWNP 20007 3 (9-84)
BASCCCK & WitCc2 NUMBER NUCLEAR POWER OiVISION 74-11s2414-oo TECHNICAL DOCUMENT respective " Diagnosis and Mitigation" section. This section provides a brief discussion of each action and decision point on the flowchart. The bottom numbers on the flowchart correspond to the subsections contained in the respective
" Technical Bases" section. This section provides the expanded, detailed explanation of and bases for the action covered in Section 2.0.
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DATE: 7-23-85 PAGE III'2
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- sASCOCK & WnCOR I 00" NUCLEAR POWER DIVI $lON 74-11s2414-oo l TECNNICAL BOCUMENT i
- i Chanter III.A General Annroach Overview / Entry Conditions 2
1.o INTRODUCTION l
1.1 Purpose This chapter provides a recommended overall' approach to ;
j diagnosis and mitigation of transients. The entry
'I conditions are identified and a logical flowpath from entry l conditions to stable plant conditions is provided for the ;
! scope of these technical bases. !
l f The flowpath includes checking for the symptoms of upsets in j heat transfer (Chapters III.B, III.C, and III.D), SGTR
) (Chapter III'.E) , proper control of the five control func-l tions (Chapter II.C) , availability of power sources and
- verification of automatic actuations. These elements are !
combined in the flowchart to provide one acceptable overall
! procedure network for mitigation of abnormal ~ transients.
I
- Details on each element are provided in the referenced 1-
{ chapters. ,
1.2 concerns and Obiectives r f 1.2.1 Concerns l Several concerns exist which must be' addressed during
! the mitigation of abnormal transients and in obtaining i stable post-trip conditions. These' concerns include the
{
j following:
l A. Anticinated Transient Without Scram (ATWS) I
! ATWS could occur due to a failure of the RPS to initiate l l
a reactor trip signal upon one of the reactor trip l
1 parameters reaching its trip limit or the control and safety rods failing to insert once the "RPS trip signal is given. The operator must recognize and react to any of the reactor trip parameters exceeding its limit ;
without causing a reactor trip. ,
DATE: 7-23-85 PAGE III*A-1
BINP 20007 3 (9 84)
SASCOCK & WitCOX NUMet NUCLEAR POWER OlVISION TECHNICAL DOCUMENT B. Steam Generator Tube Ruoture (SGTR)
A SGTR is a particular type LOCA which requires special handling to ensure that an unisolable steam leak does not occur which would increase the offsite radiation release. Once an SGTR is verified, a controlled reactor shutdown should be initiated to prevent lifting of the MSSVs. Any time an MSSV lifts, it has the potential for failing open (or failing to reseat completely) resulting in an uncontrolled radioactive release to the environment. This can be prevented by reducing the reactor power to a low enough level (within the capacity of the TBV) such that the secondary side pressure will not spike to the point of lifting any MSSVs when the reactor is tripped.
C. Automatic Action Failure Several automatic actions are required to place the plant in a controlled post-trip stable condition. These actions include main turbine stop valve shut, all site loads transferred as necessary to maintain electrical power, MFW flow running back to establish low SG level, and other automatic systems which may have been actuated post-trip.
D. Loss of Offsite Power (IDOP)
During a loss of all offsite power, it is necessary to rely completely upon the emergency power supply which has limited capacity. This results in a loss of RCPs requiring natural circulation of RC to be established as i well as the loss of other equipment that is normally I used after a reactor trip. )
l O
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BWNP 20007 3 (9 84) i SASCOCK & WILCOX NUMS88 NUCLEAR POWER DIVISION 74-1152414-oo l TECNNICAL DOCUMENT l
E. Unsats in Heat Transfer Symptoms
- It is necessary to check for upsets in heat transfer following a reactor trip since a reactor trip creates a i perturbation in heat transfer.
4 4 F. Loss of Instrumentation and Control Power i The operator must verify the operability of certain l power supplies to assure important plant parameters can be monitored and power is available to important j control devices. These devices include pumps, valves, I etc., which are needed to safely control the plant and I mitigate abnormal transients. ;
I i l.2.2 Obiectives I r
j The objective of the diagnosis and mitigation of transients j is to address the above concerns in a manner applicable to I j as broad a number of conditions as practical while mini-j mixing the impact of misdiagnosis.
I l 2.0 DIAGNOSIS AND MITIGATION The flowchart of Figure III. A-1 should be used in conjunc-tion with the following discussion. The numbered subsec- ;
I tions of Section 2.0, correspond to the upper numbers in the blocks on Figure III.A-1.
i I
j 2.1 Entry conditions i i 2.1.1 canditians triatina for n==ater Trin j
! The emergency procedures must be used by the operator l
- anytime a condition exists for reactor trip. This includes l
) the case where a reactor trip has occurred and also the case l where the conditions exist for reactor trip but the reactor l i
i trip has not occurred.
l l i l i
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2.1.2 Reactor Shutdown Recuired for SGTR In the event of a SGTR, a forced reactor shutdown will be required. If a SGTR caused a reactor trip or an upset in heat transfer or the SGTR resulted from these conditions, the operator would first perform Steps 2.2 through 2.6 and then treat the SGTR in 2.7. A SGTR is a type of LOCA which requires special handling.
2.1.3 Uosets in Heat Transfer Whenever symptoms of upsets in heat transfer are encountered above cold shutdown, the applicable parts of the guidelines, if any, should be followed to correct the heat transfer condition.
2.2 Reactivity Controlled? (Detailed discussion Section 3.1)
If reactor power is not decreasing on the intermediate range NI detectors, the operator should attempt to maintain MFW to the SG while manually shutting down the reactor.
The reactor is shutdown by borating the RCS and simultan-eously attempting to insert the control rods. Rod insertion should be accomplished by manually tripping the reactor and tripping the breakers to remove power from the control rod drives. If power cannot be removed, the operator should try to drive the control rods into the core.
l l If reactor power is decreasing on the intermediate range l but all of the safety or control rods have not fallen 1
l into the core, the operator must begin borating as necessary J to achieve an acceptable shutdown margin.
l l
l 2.3 . secondary Inventerv and Pressure Controlled? (Detailed discussion Section 3.2)
It is necessary for the reactor operator to ensure that both steam flow and FW flow are controlled immediately after reactor shutdown or trip to match reactor decay heat 1
7-23-85 PAGE III'A"4 DATE:
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load. Ensuring that turbine stop valves have closed and l FW runback has occurred are two of the more important actions which the operator must verify. If applicable, the
! l
! operator should also verify that the appropriate automatic actions have occurred any time the automatic steam line j break and AFW control system has been actuated. j l
l 2.4 Primary Inventory and Pressure Controlled? (Detailed discussion Section 3.3) i Immediately'after reactor trip or shutdown, the operator should verify that MU and letdown are properly control-led. He should also verify that all automatic actions have ,
occurred if the emergency safeguard system has been actuated l (e.g., HPI initiated).
I !
2.5 Plant Electrical Power Controlled? l l
j Immediately after reactor trip or shutdown, the operator
{
j must verify that plant electrical power is being properly j j controlled. This includes verifying that a) the output l
- breakers have tripped, b) instrument power is on (in the ;
4 r i event that a loss of instrument power has occurred, it is !
l necessary for the operator to manually stabilize and control !
the plant using known valid instrumentation and controls while attempting to restore instrument power as quickly as
- possible), and c) ensure that plant electrical loads are }
! being properly maintained. In the event of a loss of i t
! offsite power (LOOP), the operator must ensure that the l l
emergency power supply-has been started and that essential l' i
power is being supplied. Offsite power must be restored as quickly as possible.
1 i l 2.6 Svantoms of Unsets in Heat Transfer Exist? !
If a lack of subcooling margin exists, the operator must l f
! take appropriate actions as detailed in Chapter III.B. [
j If a lack of adequate heat transfer exists, the operator i
i DATE: 7-23-85 PAGE III.A-5 !
- i. 6
KWNP 20007 3 (9 84)
B ABCOCK & WitCOX NUMBER NUCLEAR POWER OlvislON 74-1152414-oo TECHNICAL DOCUMENT must take appropriate actions as detailed in Chapter III.C. If excessive heat transfer exists, the operator must take appropriate actions as detailed in Chapter III.D.
2.7 SGTR? )
This separate check for SGTR covers cases where the SGTR causes or is the result of a reactor trip or upset in heat transfer. An SGTR can be recognized by steam line radiation monitors, condenser air ejector monitors, SG level increases, mismatches in FW flow, as well as secondary system chemistry. Concurrent LOCA symptoms of decreasing RCS pressure, inventory, pressurizer level, and increased makeup flow will provide confirmation that a SGTR has occurred. However, the LOCA symptoms of increasing reactor building radiation, pressure and temperature may not be present. Refer to Chapter III.E for details on how to recognize and treat a SGTR.
2.8 Verify Stable Plant Conditions Verification of stable plant conditions includes checks for any problems which have occurred but may not have shown up as upsets in heat transfer symptoms (e.g., small steam line break or SBI4CA) . A plant cooldown or prepara- l tion to restart is at the discretion of the station manage-ment. The operator must begin a continuous monitoring of heat transfer and take appropriate action should any of the symptoms of upsets in heat transfer appear.
3.0 TECHNICAL BASES FOR DIAGNOSIS AND MITIGATION The flowchart of Figure III. A-1 should be used in conjunc-tion with the following discussion. The numbered subsec-tions of Section 3.0 correspond to the bottom numbers in the blocks on Figure III.A-1.
DATE: 7-23-85 PAGE III.A-6
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BABCOCK & WILCOX gw NUCLEAR POWER DivtSION 74-1152414-oo
( TECHNICAL DOCUMENT 3.1 Reactivity Controlled?
The reactor must be immediately shutdown. This step provides an acceptable reactor shutdown margin. It deter-mines whether or not an ATWS or a stuck rod has occurred and then directs the operator to take the appropriate action.
The bases for the action is as follows:
A. Reactor Power Not Decreasina on the Intermediate Rancre The reactor can be generating more heat than the emergency feedwater system can remove. Therefore, the operator should attempt to maintain operation of the main feedwater system to remove adequate heat to prevent overpressurizing the RCS. The reactor power is not decreasing on the intermediate range; therefore, the reactor has not been shut down and there has been a v failure of all or most of the control and safety rods to insert into the reactor core. Consequently, the operator should immediately attempt to shut down the reactor by the alternate methods available including RCS boration. The operator should also take action to try to trip the rods into the core. This includes actions such as manually tripping the reactor and, if l breakers are available in the control room that can interrupt power to the control rod drives tripping these breakers. If the rods have still failed to trip into the reactor, the operator should begin to drive the control and safety rods into the reactor core. Once the control and safety rods are successfully tripped into the core,or sufficient boric acid has been added to provide an adequate shutdown margin, the reactor will be shut down.
B. All Safety and control Rods are not at the In Limit C/ Even if the reactor power is decreasing on the inter-mediate range, it is possible more than one safety or DATE: 7-23-85 PAGE III.A-7
BONP.20007 3 (9 84)
SASCOCK & wtLCOK =
NUMS E R NUCLEAR POWER OlvlSION TECHNICAL DOCUMENT 74-1152414-oo control rod has failed to trip into the core. If this has occurred, it is necessary to begin adding boric acid to the RCS to achieve an adequate shutdown margin. The immediate need to shut down the reactor has been satisfied. The operator is concerned with assuring an adequate shutdown margin is achieved.
3.2 Secondary Inventory and Pressure Controlled?
As soon as the reactor is shut down or tripped, secon-dary inventory and pressure must be carefully checked to ensure that the secondary system is operating as designed for proper RCS heat removal. These checks include the fol-lowing:
A. To Ensure that Steam Flow is Controlled l Too much steam removal will cause the RC to over- i cool.
This includes actions such as:
- 1. Verify turbine stop valves have shut,
- 2. Verify the secondary pressure is controlling at the proper setpoint,
- 3. Verify auxiliary steam flow is controlled properly,
- 4. Verify minor steam leaks are not indicated such as leaking MSSVs. It is necessary to carefully l check and compare SG pressures and levels. In
, many cases, MSSVs have been ressated by reducing SG pressure, in which case more drastic measures such as those outlined in Section III.D are not required.
l B. To Ensure That FW Flow is Controlled This includes actions such as:
- 1. Verify FW has run back. A failure of HFW to run back is one transient that requires quick oporator evaluation and action to prevent the possibly severe consequences of water spillover into the steam lines. If the reactor has been operating at full DATE: 7-23-85 PAGE III.A-8
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TECHICAL DOCUMEllT 74-1152414-oo
(
power. and then trips without a MFW runback, it is possible that the steam lines can be flooded within
- one minute. To prevent steam line flooding, it may j be necessary for the operator to trip or verify tripped the running MFW pumps. He should then start 4 or verify AFW starts and then control AFW. (Chapter V.B)
! 2. FW must be ensured to both SGs after a reactor trip. This includes verifying that the cross-over l valve' opens on some plant designs.
C. To Verify That Automatic Actions Have Occurred 1
Large secondary side transients at some plants may result in actuation of the automatic steam line break and FW control systems. If the system is actuated, it is necessary for the operator to verify that all of the automatic actions have occurred and to take any manual
, actions necessary to control the systems and stabilize
! the plant.
3.3 Primary Inventory and Pressure controlled?
l
- Immediately after reactor trip, it is necessary for the l reactor operator to ensure that the RCS inventory and pressure is being properly controlled. This includes
- verifying proper MU and letdown flow. Increased MU flow may i
be necessary to ensure pressurizer level does not decrease l abnormally due to the RCS T,y, change during normal post-trip cooldown. ECC systems may have been actuated immedi-l
- ately after reacter trip. If these systems are actuated, it i is necessary for the operator to verify the automatic l
l actions and to take any manual actions necessary to control the systems.
DATE: 7-23-85 PAGE III.A-9
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\
TECHNICAL DOCUMENT Chaeter III.B Lack of Adecuate Subcoolina Marcin
1.0 INTRODUCTION
The purpose of this chapter is to provide the technical bases for actions taken on loss of subcooling margin (SCM) .
These bases are applicable whenever a loss of SCM occurs 3
above cold shutdown and have the highest priority of the three symptoms of upsets in heat transfer, i.e., actions for loss of SCM are performed prior to actions for either lack of heat transfer or excessive heat transfer. In addition, this chapter provides guidance and technical bases for a saturated cooldown to LPI/DHR operation with either SG heat
- removal or cooling by LOCA and HPI flow. All other cooldown I methods are discussed in Chapter III.E (tube rupturos) and Chapter III.G (cooldown methods).
)
! J 1.1 Concerns and Obiectives Durina a Lack of Adecuate Subcoolina Marcin 1.1.1 Concerns As long as the RCS remains subcooled, adequate core cooling ,
is assured. As soon as a lack of adequate SCM occurs actions must be taken to ensure adequate core cooling. For this reason the lack of SCM has top priority requiring treatment ahead of other abnormal heat transfer symptoms or SGTR. The specific concerns during a lack of adequate subcooling margin are as follows:
A. A potential threat to core cooling exists, B. A saturated RCS can create voids in the hot legs which could impede heat transfer to the secondary side, C. The potential exists for a LOCA having occurred, D. Possible entry into ICC conditions.
v 1
i DATE: 7-23-85 PAGE III*B-1
i BCNP 30007 3 (9 84)
B A BCOCK & WRCOX NUCLEAR POWER DIVISION <
74-lls2414-oo TECHNICAL DOCUMENT l 1.1.2 Obiectives The objectives to be considered during the treatment of a lack of subcooling margin are (listed in order of relative priority):
A. Maintain Adeauate Core coolina - Adequate core cooling always has first priority. If assurance of a subcooled RCS has been lost, it is necessary to take actions to ensure that the core remains adequately cooled. These actions include tripping (or verify tripped) RCPs, ensuring adequate HPI/LPI flow, and maintaining primary to secondary heat transfer. If these actions are taken and the equipment operates as designed, ICC conditions will be prevented.
B. Restore Subcoolina Marcin - Provided that HPI/LPI is operating, subcooling margin should be restored within about 10 minutes unless a LOCA has occurred. Once subcooling margin is restored, the operator must carefully control RCS pressure to prevent exceeding PTS limits (if applicable - refer to Chapter IV.G or Techni-cal Specification Pressure Vs. Temperature Limits).
C. Ensure ProDer Secondary Control - While the core can be adequately cooled by using HPI or LPI cooling, primary to secondary heat transfer is preferred. During a loss of adequate SCM, the SG level must be raised to the loss of SCM setpoint (defined in Chapter IV.C) .
However, the rate at which the level is raised should be controlled as necessary to prevent excessive heat removal, especially if the loss of SCM was caused by excessive heat transfer.
1.2 Causes A loss of SCM may occur for several reasons. They are:
A. LOCA - Large break IOCAs will result in a sustained loss ,
of adequate SCM. Since RC inventory is being rapidly DATE: 7-23-85 PAGE III.B-2
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SABCOCK & WILCOX NUM8tt NUCLEAR POWER OlvlS10N
D) TECHNICAL DOCUMENT lost during a large break LOCA, it is especially important that HPI/LPI flow is assured. A loss of adequate SCM can also occur due to a SBLOCA; however, I subcooling may be restored by HPI flow.
B. Prolonaed Excessive Overcoolina -A large steamline break or an extensive overfeed of a SG is required to create the excessive overcooling which will result in a loss of SCM. For most steamline breaks, the automatic actuation of HPI will maintain the SCM; however, if HPI fails or HPI flow is not sufficient, a relatively small steamline break or failed-open steam valve would result in a loss of adequate SCM once the pressurizer was drained. Chapter III.D discusses overcooling and its mitigation in detail.
C. Prolonaed Loss of Heat Transfer - A prolonged loss of primary to secondary heat transfer could result in a v loss of adequate SCM. Chapter III.C discusses loss of heat transfer in detail.
D. Failures of the RC Pressure Control System - Failures such as a pressure transmitter failing high could cause the pressurizer spray and PORVs to open and the presn'tr-izer heaters to turn off.
2.0 DIAGNOSIS AND MITIGATIOl{
The flowchart of Figure III.B-1 should be used in conjunc-tion with the following discussion. The numbered subsec-tions of Section 2.0 correspond to the upper numbers in the blocks on Figure III.B-1.
2.1 Identification of a Lack of Adecuate Subcoolina Marcin (Detailed discussion in Section 3.1)
The RCS P-T relationship will clearly indicate when a loss (N )
of adequate SCM occurs. (Refer to Chapter II.B)
\ /
%./
DATE: 7-23-85 PAGE III.B-3
BWNP-30007 3 (9 84)
B A5 COCK & WILCOE NUCLEAR POWER OlVi$lON NUMBER 74-11s2414-oo TECHNICAL DOCUMENT 2.2 Trio RCPs All RCPs must be tripped immediately upon a loss of adequate SCM. Refer to Chapter IV.A for details about RCP operation.
2.3 Control RCS Inventory (Detailed discussion Section 3.2)
Control of RCS inventory requires that A. HPI/LPI flow must be maximized into the RCS, B. All possible leaks which are isolable should be iso-lated.
2.4 Maintain Proper SG Levels (Detailed discussion in Section 3.3)
While a loss of SCM exists, the SG levels must be controlled at the loss of SCM setpoint in each SG that can hold pressure. However, the SG fill rate to the setpoint should be controlled per Chapter IV.C.
2.5 Subcoolina Marcin Reestablished?
Further actions will depend on whether or not subcooling is regained. Very small LOCAs and termination of excessive overcooling will allow SCM to be regained; large break LOCAs, total loss of FW combined with partial loss of HPI and extended overcoolings may take longer to recover SCM.
It is possible for the operator to reach this point in the procedure quickly and SCM may not have been regained but will be restored within a few minutes. If not, the operator should continue to monitor the further steps that are required if SCM is not regained. However, as soon as SCM is regained, the operator should return the plant to as near normal as possible. When the CCM is regained, the actions required by Block 2.6 (Reestablish Normal Plant Control) should be followed. If, however, the RCS remains i
DATE: 7-23-85 PAGE III.B-4 l
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. SABCOCK & WILCOR NUCLEAR POWER DIVISION 1
74-1152414-oo TECHNICAL OSCUMENT i
saturated, then the actions beginning with 2.7 (Super-heated?) should be followed.
Reestablish Normal Plant Control (Detailed discussion in 2.6 !
} Section 3.4)
As soon as the adequate SCM is restored, the operator should f l
throttle HPI to ensure that RCS pressure-temperature limits l
)
i are not oxceeded. The operator should also restart RCPs (this includes restoring RCP services) and reestablish pressurizer spray if possible. If secondary side heat
) i transfer is available, one RCP per loop should be started. 1 If secondary side heat transfer is not available, only one RCP should be started - see Chapter IV.A for instruction or RCP restart with voids. In the event of a lack of primary ,
- to secondary heat transfer, the operator should take actions l to 1 store heat transfer. Other symptoms, especially .
! relat.ive to SGTR, which is also a LOCA, should be moni-
- tored. In the event that a cooldown is required, refer to
' Chapter III.G for plant cooldown guidelines. Caution should be taken in reopening the pressurizer spray block valve since this could conceivably reopen the LOCA (if the l j
pressurizer spray block valve had isolated the leak).
l 2.7 Superheated? l The loss of SCM is a possible prelude to ICC. Should the l l
l RCS continue to heat up, becoming superheated, the operator should take actions to mitigate ICC conditions. Refer to i i
i Chapter III.F for a detailed discussion of the required ;
actions for ICC conditions. ;
2.8 Reat Transfer in Both SGs (Detailed discussion in Section 3.5) i Further actions are dependent on whether or not primary to secondary heat transfer exists in either SG. If primary to l l
l t
DATE: 7-23-85 PAGE III.B-5 ;
'-,--n-.-- , . - - , . - , - -ne~..ev--e,~~-,--av- -
r~r
BVINP 20007 3 (9 84) sASCOCK & wlLCOE NUM$tt NUCLEAR POWER OlVISION 74-11s2414-oo TECHNICAL DOCUMENT secondary heat transfer exists in at least one SG, then a saturated cooldown can be performed while attempts are made to restore heat transfer to both SGs, if necessary (Section 2.9 of this Chapter) . If, however, heat transfer does not exist to either SG, then further actions depend on the RCS response (Sect',on 2.lo of this Chapter) .
2.9 Saturated Cooldown With SG(s) (Detailed discussion in Section 3.6)
Since the RCS is saturated, HPI flow should be continued at the maximum required flow rate (two HPI pumps if avail-able). Also, secondary side pressure will have to be controlled carefully to continue the cooldown as the RCS cools and depressurizes along the saturation line. Con-tinued saturated natural circulation cooldown will be evident by the incore T/C temperature decreasing as SG saturation temperature decreases due to decreasing secondary side pressure. Some core heat removal may be through latent heat of vaporization of the RC. Consequently, it is possible that little core delta T exists. However, indi-cated delta T may be larger due to the influence of HPI on the T cold RTDs. If applicable, efforts should continue to restore heat transfer to the idle SG.
It is highly probable that, with heat transfer to at least one SG and full HPI flow, sustained saturation of the RCS is due to a small break LOCA. However, this is a relatively short-term condition as the RCS will evolve to one of two states.
If the RCS is in natural circulation and cooling, RC pressure will decrease along the saturation curve. As RC pressure decreases, the break flow will decrease and HPI l
DATE: 7-23-85 PAGE III.B-6 l
SWNP 20007 3 (9 84)
BADCOCK & WitCOE NUCLEAR POWER DIVillON 74-1152414-oo TECHNICAL BOCUMENT flow will increase. In this case SCM should eventually be restored without a loss of heat transfer.
If, however, the RCS is actually in boiler-condenser cooling, then heat transfer to the SGs may be cyclic and heat transfer will probably be lost before SCM is restored.
This is because, as the PCS refills, the condensing surface in the SG tube region will be lost.
Saturated cooldown with the SG(s) is discussed in detail in Section 3.6. The loop in Figure III.B-2 from 2.12 (condi-tions not yet established for LPI/DHR cooling) back to 2.5 (check for adequate SCM) covers the possible evolutions discussed above and in Section 3.6. Actions to restore heat transfer to an idle SG are discussed in Chapter III.C.
2.lo RCS Continues to Cool and Denrassurize (Detailed discussion in Section 3.5)
Heat transfer will have been lost by one of three situations l occurring: a) the SBLOCA has depressurized the RCS below the secondary side pressure (RC temperature below SG temperature), b) a loss of FW has occurred, c) steam voids have collected in the hot leg terminating natural circula-tion (this includes cyclic boiler-condenser cooling) . For the first case, core cooling is being provided by the break flow and HPI (Section 2.11 of this Chapter) .
For the latter two cases, FW flow and/or heat transfer must be restored. Methods of restoring heat transfer are detailed in Chapter III.C, " Lack of Heat Transfer." A detailed discussion of this decision point is provided in Section 3.5. -
DATE: 7-23-85 PAGE III.B-7
C"rNP-2OOO7 3 (9 84) 3 AsCOCK & witCOR NUCl[ AR POWER OlvlSION NUMS[s 74-11s2414-oo TECHNICAL DOCUMENT 2.11 Cooldown on Break /HPI Flow (Detailed discussion in Section 3.7)
If the small break I4CA has resulted in the RCS continually cooling and depressurizing below the SG temperatures and pressures, then the SGs may no longer be required for heat removal. The combination of break flow and HPI is providing adequate core cooling and may continue to do so until the transition to LPI or DHR cooling can be made. The loop on Figure III.B-2 from 2.12 (conditions not yet established for LPI/DHR cooling) back to 2.5 (check for adequate SCM) covers a subsequent change of state, i.e., SCM restored or RC pressure and temperature no longer decreasing. A detailed discussion is provided in Section 3.7.
2.12 Conditions Established for Transition to LPI/DHR This decision. point, an6 the loop back to 2.5 (check for adequate SCM), are provided to address the possibility of a change in plant conditions before LPI or LPI/DHR cooling can be initiated. For example, if the RCS is in boiler-conden-ser cooling, the progression on the flowchart would be through block 2.9 (saturated cooldown with SG(s)) . If the RCS subsequently cycles out of boiler-condenser cooling (Section 3.6), indicating a loss of heat transfer, then the loop back to block 2.5 would provide for a progression through block 2.10 to Chapter III.C for actions to restore heat transfer.
l l 2.13 Initiate LPI/DHR Coolina (Detailed discussion in Section 3.8)
When the RCS cools and depressurizes to within the LPI operating range, the transition should be made to LPI or LPI/DHR cooling, including other long-term cooling actions l such as prevention of boron precipitation in the core region. This is discussed in detail in Section 3.8.
DATE: 7-23-85 PAGE III.B-8
BWNP 20007 3 (9 84) ;
)' SABCOCK & wiLCOR NUMtta l NUCLEAR POWER DIVISION 74-11s2414-oo ;
TECHNICAL DOCUMENT 1
3.0 LACK OF ADEQUATE SUBCOOLING MARGIN TECHNICAL BASES
! The flowchart of Figure III.B-1 should be used in conjunc- j i tion with the following discussion. The numbered subsec- i i
tions of Section 3.0, correspond to the bottom numbers in ;
- the blocks on Figure III.B-1.
i l I 3.1 Identification of a Lack of Adecuate Subcoolina Marcin f The P-T relationship provides the fastest and most obvious j indication that a lack of adequate SCM has occurred. Refer f to Chapter II.B for a discussion of the loss of SCM curve. f l
j 3.2 Control RCS Inventory i I
Immediately upon loss of adequate SCM, HPI/MU must be !
initiated and maximum flow into the RCS ensured. Refer to
' ~
Chapter IV.B for a discussion on" maximizing HPI flow.
i HPI will be started automatically by the emergency safe-guards system when RCS pressure decreases below the safe-guards system actuation setpoint. Increasing RB pressure
- will also actuate emergency safeguards which initiates HPI.
j once automatic actions occur, they must be verified and j maximum HPI flow into the RCS ensured.- It-is necessary to ensure maximum HPI flow to ensure adequate core cooling as well as restoring SCM as quickly as possible.
l The lack of SCM can'ba caused by the following:
A. Fag _essive Overcoolina Has occurred In the event of an overcooling transient, maximum HPI flow only has to make up for RC contraction. Conse-quantly, SCM will be regained rapidly, and HPI flow must then be controlled to prevent excessive repressurization and possibly exceeding RCS pressure-temperature limits.
l l
DATE: 7-23-85 PAGE III.B-9
ScNP 30007 3 (9-84)
B ABCOCK & witCOX NUMtte NUCLEAR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT B. A LOCA Has occurred In the event of a LOCA, maximum HPI is required to replace the RC inventory that is being lost out the break. If the LOCA is a large break LOCA (break area approximately equal to or greater than a 10 inch diameter hole), the HPI will be augmented by LPI and CFTs. However, some small LOCAs will require cooling solely by HPI (i.e., will not initially depressurize the RCS below the CFT and LPI actuation pressures).
Consequently, HPI must be maximized to replace the RC inventory that is being lost until SCM is restored.
All isolable leaks should be isolated, if possible.
There are several possible leaks in the RCS which may be isolated by closing certain valves. These may include:
- 1. The PORV
- 2. PORV block valve
- 3. Pressurizer spray block valve and spray valve
- 4. RC letdown valves
- 5. Pressurizer vent valves
- 6. Pressurizer sample valves
- 7. Hot leg high point vent valves
- 8. Reactor vessel head vent valves C. A LOFW Has Occurred In the event of a LOFW, RC inventory will be lost after the RCS pressurizes to the PORV setpoint. HPI flow is needed to replace this inventory and thereby cool the Core.
l
! 3.3 Maintain Proper SG Levels Whenever a loss of SCM occurs, the SG level in the pressur-ized SG(s) must be controlled at the " loss of SCM" set-DATE: 7-23-85 PAGE III.B-10 l
'l 8%NP 20007 3 (9-84)
SABCOCK & WILCOX NUCLEAR POWER DIVISION NumttR
' 74-lls2414-oo TECNNICAL BOCUMENT l
point. Details on SG level requirements are discussed in Chapter IV.C.
! The lack of SCM can be caused by the following: i A. Excessive Overcoolina Has Occurred
- In the event of a steam leak, the SG with the steam leak should not be fed until the leak has been terminated.
If a steam leak has occurred, attempting to raise the
. SG water level in the SG with the steam leak would only prolong the excessive heat transfer. However, the level in the SG without the steam leak must be increased to
! the " loss of SCM" setpoint as soon as the SCM is lost.
If the steam leak can be isolated and heat transfer to the previously leaking SG restored, then its level must also be increased to the " loss of SCM" setpoint. SG pressure should be controlled to limit RCS reheat and swell for RV pressure-temperature limitations.
B. A LOCA Has Occurred Certain size LOCAs when only one HPI pump is available require that SG levels be increased to the " loss of SCM" l
setpoint. However, since the size of the LOCA cannot be measured and to allow for a subsequent loss of one of l two HPI pumps, the SG water level should be increased for all IDCAs, except for the larger LOCA's-which cause the RC to continually depressurize below the SG pres-sure. These larger LOCA's do not require heat removal by the SGs. For smaller LOCAs, increasing SG levels helps to ensure that saturated natural circulation will continue. However, in the event that steam voids form in the hot leg and block natural circulation, the I
establishment of high SG levels will allow boiler condenser cooling to occur. In addition to raising SG s levels, it may be necessary to reduce secondary side DATE: 7-23-85 PAGE III.B-11
ScNP 30007 3 (9 84)
SASCOCK & WILCOX NUCLEAR POWER OlVISION NUMtit 74-11s2414-oo TECHNICAL DOCuldENT pressure to maintain heat transfer. Some medium size breaks can cause the RCS to depressurize and saturate below the secondary system steam pressure. This will result in the SG becoming a heat source because the temperature of the secondary inventory will be higher than that of the RC. If the RCS continues to cool and depressurize, SG heat removal may not be necessary.
However, if the RCS pressure stabilizes at a pressure below SG pressure, it may be necessary to also reduce secondary side pressure. This is discussed in detail in Chapter III.C.
Boiler condenser cooling occurs when RC is boiled in the reactor core forming steam (removing core heat) which then flows through the hot leg piping to the SG where it condenses in the SG tubes. The condensed water then returns to the core by the cold leg piping. For the condensed water to flow back into the reactor core, the RC water level in the SG must be above the elevation of the RCP internal lip. This will provide the driving force to allow the water in the cold leg pipe to flow up and over the RCP discharge into the reactor core.
It is necessary to increase the SG level to above the RC water level in the steam generator tubes to provide a condensing surface where the RC steam can be condensed on the surface inside the tubes. This tube surface has l to be large enough to remove all of the latent heat of steam at the expected RC steam flow rate. To ensure that the condensing surface above the RC water level in the SG tubes is high enough, the SG water level must be l raised to the " loss of the SCM" setpoint. It is necessary to ensure full HPI flow to assure that the i DATE: 7-23-85 PAGE III.B-12
Bf!NP 20007 3 (9 84) sAscoca a wncox NUCLEAR POWER DIVISION TECNNICAL IOCUMENT reactor core remains covered (full HPI flow must be ensured anyway as long as there is a lack of SCM).
If FW is sprayed into the SG through the AFW nozzles at or above the minimum required flowrate (Chapter IV.C) the l effective condensing surface of the SG tubes is higher than l the " loss of SCM" level. This is because the AFW nozzle spray will be cooling the tube surface above the " loss of I SCM" level. i
' i Primary to secondary heat transfer is verified by ensuring that the RCS and secondary are coupled. Incore T/C tempera-ture shou 1.d decrease toward the SG saturation temperature as SG pressure is reduced.
! If a larger LOCA occurs, and RCS pressure continually l j decreases below the SG pressure, then secondary side heat transfer will no longer be needed and further core cooling will be supplied through HPI/LPI/CFT.
C. Imma of Feedwater Has occurred j If the loss of SCM was caused by total loss of FW, then l the operator may not be able to restore FW and raise SG
( levels. But after he has established HPI cooling, he should continue his efforts to restore FN. Once FW is l
restored than the appropriate SG level should be established " loss of 8CK" if still saturated,-" natural circulation" if now subcooled as indicated by the incore T/Cs. Loop voids may still exist when the core is subcooled but the " natural circulation" setpoint is adequate because a) a transition to boiler-condenser cooling will not occur with the core subcooled, b) the level is adequate to restore heat transfer when the RCPs are bumped or the HPVs are used to eliminate voids, and DATE: 7-23-85 PAGE III.B-13
BWNP40007 3 (9-84)
S ABCOCK & WILCOE NUMBER I NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 74-11s2414-oo. .
c) when heat transfer is restored, the level in the SG will swell as its liquid inventory heats up.
i 3.4 Reestablish Normal Plant Control As soon as SCM is regained, HPI will probably have to be throttled to prevent exceeding RC pressure-temperature limits. Also, RCPs should be started which will aid in establishing primary to secondary heat transfer and will also provide better mixing of the cold HPI flow to alleviate any PTS concern. If secondary side heat transfer is not possible (i.e., continued loss of FW), then only one RCP should be started. This one RCP will provide thermal mixing. If a RCP cannot be restarted and primary to secondary heat transfer does not exist after SCM is re-stored, it is probably due to a lack of FW or steam voids in the hot leg blocking natural circulation. The operator then must take actions to restore heat transfer from the primary to the secondary side (see Chapter III.C for discussion on restoring heat transfer). Failing this, HPI cooling must be continued or started.
After this symptom of lack of adequate SCM is trsated, attention should be turred to the other symptoms of upsets in heat transfer. If the adequate SCM is restored during the treatment for inadequate heat transfer, then the operator can stop the treatment of lack of adequate SCM.
For example, if the SG level is being raised to the " loss of SCM" setpoint, the level increase can be stopped. Since a SGTR and could have caused RCS depressurization to the point of lack of adequate SCM, this symptom should.be closely monitored. A discussion relative to SGTR is detailed in Chapter III.E.
O DATE: 7-23-85 PAGE III B~14
BWNP 20007 3 (9 84)
SASCOCK & WILCOX NUM8it g NUCLEAR POWER DIVISION 74-1152414-oo ;
TECHNICAL DOCUMENT Reestablishing pressurizer spray is a concern. The possi-bility exists that a LOCA occurred between the pressurizer spray block valve and the pressurizer spray valve. If so, reopening the pressurizar spray block valve to re-establish pressurizer spray would result in re-establishing the LOCA.
Consequently, the pressurizer spray block valve should be opened and RCS pressure monitored for a while to ensure that RCS depressurization does not occur. If the depressuriza-tion does occur, the LOCA has been identified. The pressur-izer spray block valve must be reclosed and further control of RC pressure will be by opening and closing of pressurizer vent /PORV.
3.5 Heat Transfer in One or Both SGs Further actions depend on whether or not heat transfer
] exists in one or both SGs. Previous actions have been made to keep the core covered with water which allows the core to transfer heat to the surrounding coolant. The next step is to remove the heat from the surrounding coolant. If heat transfer exists in one or both SGs, then a saturated cooldown can continue (see Section 3.6) . If heat transfer does not exist in either SG, then further action depends on the response of the RCS.
If the RCS continues to cool and depressurize below the SG pressures and temperatures, then the break size and HPI flow are sufficient to provide core cooling. The SGs are not required while this condition exists but the SGs should be made available in case they are needed later; e.g., FW restored, if applicable. The cooldown should continue on break /HPI flow (see Section 3.7).
If, however, the RCS pressure and temperature (incores)
\s .
stabilize or begin to increase along the saturation curve, DATE: 7-23-85 PAGE III.B-15 1
I
ScNP 20007 3 (9 84) 1 S ASCOCK & witCO:'
NUMBER NUCLEAR POWER DIVISION 74-11s2414-oo TECHNICAL DOCUMENT then SG cooling should be restored. Methods to restore primary to secondary heat transfer or, if necessary, initiate HPI cooling are described in Chapter III.C.
At this point, the RCS is still saturated. If heat transfer exists, it is by saturated natural circulation or boiler condenser cooling (see Section 3.3 of this chapter for a discussion of boiler condenser cooling) . Since the RCS is saturated, it will be more difficult to determine whether or not natural circulation or boiler condenser cooling mode exists. The best indication of a loss of natural circula-tion flow when the RCS is saturated or loss of boiler condenser cooling is a trend of incore T/C temperature vs. RCS pressure increasing up the saturation curve.
Natural circulation flow can be lost due to low thermal centers in the SGs or blockage due to voids in the RCS. It is possible that the RCS will be in natural circulation for a while, then natural circulation can be lost if the RCS is losing inventory. With a loss of RCS inventory, steam will form in the hot legs and eventually stop the saturated natural circulation flow. If the RCS continues to lose inventory boiler condenser cooling should occur as long as the SG(s) are available, i.e.,'high enough level and/or minimum required AFW flowrate (Chapter IV.C).
Boiler condenser cooling can be lost due to a loss of od2quate condensing surface in the SG tubes. This can occur due to a decrease in effective thermal center on the secondary side due to a low liquid level and insufficient AFW flow or due to an increase in RC level inside the tubes. Boiler condenser cooling may be cyclic in nature because, as the RC steam is condensed, RC pressure will decrease toward SG pressure causing leak flow to decrease DATE: 7-23-85 PAGE III.B-16
BWNP 20007 3 (9 84)
B ABCOCK & WILCOX NUMBER
~s NUCLEAR POWER DIVISION 74-11s2414-oo
[V') TECHNICAL DOCUMENT and HPI flow increase. This may allow the RCS to refill which will reduce the available condensing surface inside the SG tubes. If the heat transfer to the SGs is lost, RC temperature and pressure will begin to increase causing the leak flow to increase and HPI flow to decrease. This will, in turn, cause a net loss in RC inventory which will restore the condensing surface. This self-regulating process may occur with only fluctuations in the amount of heat transfer or it may result in periodic losses of heat transfer. In either case, boiler condenser cooling should continue without further operator action required to restore heat transfer. However, the operator may take the actions discussed in Chapter III.C to aid restoration of heat transfer.
Boiler condenser cooling may also be partially lost due to a (N continued net loss of RC inventory while boiler condenser cooling exists. Boiler condenser cooling provides or aids core cooling in two ways. First, the condensation of the RC steam in the tube region removes heat from the RCS and, by reducing RC pressure, allows higher HPI flow which also provides cooling. Second, by maintaining a high enough RC liquid level in the SG tubes, coolant is forced over the RCP internal lip where it can flow into the core region. If the RC inventory continues to decrease even with boiler conden-ser cooling, the level required to force coolant over the RCP internal lip will be lost. However, the condensing surface inside the SG tubes will continue to grow and thus the heat removal obtained by condensing RC steam will l continue. This case is essentially a combination of saturated cooldown with the SGs (3.6) and cooldown on break /HPI flow (3.7).
A I
O) 7-23-85 PAGE III.B-17 DATE:
BWNP 3OOO7 3 (9-84)
BABCOCK & witCOX NUntatt NUCLEAR POWER DIVI $lON 74-11s2414-oo TECHNICAL DOCUMENT During a saturated cooldown, using either the SGs or break /HPI flow or a combination of both, it is possible for the RC pressure to appear to stabilize when the CFTs begin to discharge. If the net loss of RC inventory is relatively small, the CFTs will " float" on the RCS. The RCS should continue to cool and depressurize but possibly at a slower rate. If the RCS begins to reheat, the operator should attempt to increase heat removal by the SGs.
3.6 Saturated Cooldown Usina the SG(s)
A saturated cooldown is performed essentially the same as a normal natural circulation cooldown with a subcooled RCS but with two major differences. First, full HPI flow from two HPI pumps if available must be maintained while the RCS is saturated. This HPI flow can also provide substantial core cooling thus the operator does not have total control of the RCS cooldown rate with the SGs.
Second, a saturated cooldown with the SGs is inherently unstable. In order to maintain SG heat removal with a saturated RCS, a mass and energy balance must exist. If the RCS is initially in saturated natural circulation, core decay heat may exceed the heat removal rate. The core ,
boiling may eventually result in sufficient steam formation to block natural circulation flow. This may eventually result in boiler condenser cooling or RCS refill if core boiling is suppressed by HPI cooling through the break before boiler condenser cooling begins.
Once a mass and energy balance is established, further decreases in core decay heat or increases in SG heat removal will offset this balance. If the RCS is in boiler condenser cooling, heat' transfer to the SGs may be lost due to RCS refill as described in Section 3.5. If the RCS is in
- DATE
- 7-23-85 PAGE III.B-lE l
l
BWNP 20007 3 (9-84)
BASCOCK & wiLCOK NUCLEAR POWER OlVISION NUM8tt 74-11s2414-oo TECHNICAL DOCUMENT saturated natural circulation, the decrease in core decay heat or increase in SG heat removal will result in RCS
, cooling and depressurization. This will decrease the break i flow while increasing the HPI flow and should allow restora-tion of the SCM while maintaining primary to secondary heat i transfer.
While the RCS is saturated, the CFT isolation valves must remain open until sufficient LPI flow exists (Chapter j IV.B). However, once SCM is restored the CFT isolation I valves should be closed when conditions permit. HPI/LPI piggyback operation may be required during the cooldown l
(Chapter IV.B).
3.7 Cooldown on Break /HPI Flow If the break size is large enough, the RCS will continue to i
cool and depressurize below the temperature and pressure of the SG(s). This size break, coupled with HPI flow, will provide adequate core cooling and use of- the SG(s) may not be necessary. However, the SG(s) should be made available in the event that the RCS does begin to reheat. ,
The only operator actions required for breaks of this size are to ensure full HPI flow and that other injection sources
- (CFTs and LPI) are available. RC pressure may tend to stabilize when the CFTs begin to discharge but, as long as 1 '
the RCS does not begin to reheat, actions to restore SG l
cooling are not required.
HPI/LPI piggyback operation may be required during the I
cooldown (Chapter IV.B).
U DATE: 7-23-85 PAGE III.B-19
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BABCOCK & WILCOX ]
NUCLEAR POWER DIVISION 1 74-lls2414-oo TECHNICAL DOCUMENT 3.8 LPI/DHR Coolina Proper operation of the LPI system should be verified as ;
l soon as LPI is actuated whether manually or automatically.
This includes verification of proper valve lineup and verification or performance of LPI suction switchover to the sump from the BWST when required and HPI/LPI piggyback lineup if required (Chapter IV.B).
When the RCS pressure decreases to the LPI discharge pressure, the operator should verify LPI flow. In addition, the operator may terminate HPI/LPI piggyback operation, if applicable, and close the CFT isolation valves when the LPI flow is adequate as defined in Chapter IV.B.
While the RCS is saturated, both LPI trains should remain operating in the injection mode (the one exception is for SGTR, discussed in Chapter III.E). However, the large capacity of the LPI system may restore SCM. If SCM is restored, then one LPI train may be switched to the DHR mcde (suction from the drop line) while the other LPI train remains in the injection mode to compensate for the break flow. If SCM is subsequently lost, both LPI trains must be placed back in the injection mode (suction from the BWST or containment sump).
Actions should be taken to prevent boron precipitation in the core region within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA. These actions involve establishing alternate flowpaths to prevent excessive boron accumulation in the core region. These flowpaths are plant specific thus site procedures should be referenced.
The LPI coolant should be sampled periodically for boron concentration and pH after suction is transferred to the l
l DATE: 7-23-85 PAGE III.B-20
_ _. . - .-. . -. . - . . . - - ~~- .-. . _ . =. - .- _
d BWNP 20007 3 (9 84)
SABCOCK & WitCOE NUMSER j NUCLEAR POWER DIVISION 74-11s2414-w TECHNICAL DOCUNENT sump. If necessary, the appropriate chemicals should be added to maintain plant specific boron concentrations and pH l levels (refer to plant specific procedures) . In addition, l
if the boron concentration is low, the operator should l
verify the proper lineup to prevent boron precipitation
- discussed above and check for possible leaks (e.g., cooling water) in the RB that may be diluting the sump water.
i t i
e
^
I l
J l
4 l
I i
l \
t 1
l .
l DATE: 7-23-85 PAGE III.B-21 !
i I -- - - - . , - _ . _ _ _ , _ _ , _ _ _ , _ _ , _ _ , . , , _ _ . , . _ _ , . , . _ _ _ _ , _ _ . , _ __,..__ _ _ _ , , _ ]
Figura III.8-1 LACK 0F ADEQUATE SUBC00 LING i III.B MARGIN FLOWCHART H
2.1 l IDENTIFICATION OF
- A LACK OF ADE-QUATE St.BC00.ING
- MARGIN d 3.1 0 0 -
2.2 l TRIP ALL RCPs S W REATED x YES IV.A 1i U
TAKE ICC ACTIONS 2.3 l CONTRQ. RCS NO INVENTORY 3.2 .
U 2.4 l TRANSFER IN MAINTAIN PR(FER SG LEVELS 33
\-[ E 10 ES 2.\5 2.9 l i CONTINUES TO M
PESSURIZE SU8C00 INGN NO MARGIN RESTORED ;
SATWATED C00LDONN BELOW SGs U WITH SGs E STORE E AT 3.6/III.C YES III.C p 2.11 l 2.6 l COOLDONN ON BREAKAPI REESTABLISH t0RMAL FLOR PLANT CONTRQ.
3.4 3.T/IV.B COPOITIONS ESTABLISHED FOR TRANSITION TO LPI/DFE l
g 25 '
2.13 l INITIATE LPI/DHR-COOLING 00C. NO. 74-1152414-00 3 8'IV 8 b
BT!NP 20007 3 (9-84) eascoes a witcox Numtts NUCLEAR POWER DIVISION 74-1152414-oo TECNNICAL DOCUMENT Chanter III.C Lack of Adeauate Primary to Secondarv Heat Transfer I
1.0 INTRODUCTION
The purpose of this chapter is to provide technical bases )
i for restoring adequate primary to secondary heat trans- l fer. These bases are applicable whenever a symptom of inadequate primary to secondary heat transfer exists. The i i
RCS may be subcooled or saturated.
i
{ 1.1 Concerns and Obiectives 1.1.1 Concerns >
l A lack of primary to secondary heat transfer is the second j priority symptom (along with'its counterpart, excessive !
primary to secondary heat transfer) which requires treatment l as soon as adequate SCM is assured or treated. While a backup source of cooling, HPI cooling, is available, its use will degrade the RB environment. Consequently, heat l transfer from the primary to the secondary side should be
! restored as quickly as possible. However, core cooling has priority and HPI cooling must be used if SG cooling is not available. Concerns during a lack of primary to secondary ;
heat transfer include the following:
A. A Lack of Adecuate Core Coolina
- The normal method of core cooling does not exist.
Maintaining adequate core cooling is most important.
J '
B. Extended Loss of Feedwater with Inadeauate HPI. This can result in ICC.
C. Increasina RC Pressure. This can minimize HPI flow into i the RCS.
Voids in the Hot Lea LooDs. Voids can prevent l D.
\ restoration of natural circulation.
E. Reactor Coolant System Pressure-TemDerature Limits.
1 I
P-T limits could be violated during HPI cooling and DATE: 7-23-85 PAGE III.C-1
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BONP 20007 3 (9-84) sAscOCK & WILCOX i NUCLEAR POWER DIVISION NUMsER j 74-11s2414-oo ;
TECHNICAL DOCUMENT l
repressurization of the RCS without proper operator 1
action. This includes the technical specification RCS l
P-T limits and PTS limits if applicable. l F. HPI coolina HPI cooling results in relessing large quantities of RC j to the RB. If the BWST becomes depleted, recirculation l from the RB sump will be required.
G. Addina FW to dry SG Adding FW to a dry SG can cause unanalyzed thermal stresses.
1.1.2 Obiectives The objectives which must be considered during a lack of heat transfer include:
A. Restorina and Maintainina Adecuate Core Coolina Adequate core cooling is the primary objective. Cooling with the SGs is the preferred method, but HPI cooling will be used if necessary.
B. Restore and Maintain Primary to Secondary Heat Transfer Core cooling via primary to secondary heat transfer is the preferred method of core cooling.
C. Maintain Subcoolina Marcin SCM may be lost due to an extended LOFW, excessive overcooling, or a LOCA. This chapter covers the impact of a loss of SCM on efforts to restore primary to secondary heat transfer. The details for restoring SCM are provided in Chapter III.B.
D. Prevent Exceedina PTS Limit A prolonged loss of primary to secondary heat transfer will require HPI cooling. Due to the possibility of cold HPI t. rater entering the RV and creating thermal shock to the inside of the RV wall, it is important that HPI and RCS pressure be carefully controlled to prevent exceeding the PTS limit (if applicable). The PTS limit is described in Chapter IV.G. .
DATE: 7-23-85 PAGE III.C-2
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BASCOCK & WILCOX NUMSER NUCLEAR POWER DIVISION
[ TECHNICAL DOCUMENT 74-11s2414-u l
1.2 Causes A lack of adequate primary to secondary heat transfer can occur due to one or more of the following: '
A. Loss of Feedwater i A LOFW can occur for several different reasons. For example, FW pumps can stop, valves can be inadvertently closed, or the source of FW may be depleted. A LOFW would also include the case where, although some FW inventory remains in the SG, it is not sufficient to maintain heat transfer between the primary and secondary sides. This could occur if the SG level was not raised to the setpoint when necessary to establish boiler condenser cooling or failure to raise SG level to the setpoint when required to maintain natural circulation.
B. Loss of Primary Side Looo Flow O A loss of RC flow could occur for several reasons. For
( example, c LOCA can occur, requiring RCPs to be stopped due to loss of SCM, depleting the RC inventory to the point where steam voids form in the hot legs, thus stopping any natural circulation flow. Prolonged exces-sive overcooling can also create voids which could impede natural circulation. A prolonged LOFW could result in steam voids blocking the hot legs which could
( prevent natural circulation from occurring once FW is l restored. A loss of heat transfer due to voids in l
the hot legs would initially be treated through the lack of adequate SCM symptom. Once the lack of adequate SCM
- actions were taken, this section provides follow-on actions to restore heat transfer.
C. RC Temperature Decreasina Below Secondary Side-Temnera-ture l A LOCA can saturate the RCS and cause the RC pressure to go below the SG pressure. This will cause a loss of l
Q primary to secondary heat transfer. If the RCS con-l DATE: 7-23-85 PAGE III.C-3
BWNP 20007 3 (9 84)
B ABCOCK & WILCO K NUCLEAR POWER DIVISION NUMBER 74-11s2414-o0 TECHNICAL DOCUMENT tinues to cool and depressurize below the SG tempora-tures and pressures, then the break flow and HPI are providing adequate core cooling. This case is covered in Chapter III.B.
2.0 Diacnosis and Mitication The flowchart of Figure III.C-1 should be used in conjunc-tion with the following discussion. The numbered subsec-tions of section 2.0 correspond to the upper numbers in the blocks of Figure III.C-1.
2.1 Identification of Lack of Adecuate Primary to Secondary Heat Transfer (Detailed discussion in Section 3.1)
The P-T relationship provides the most obvious indication that a lack of primary to secondary heat transfer has occurred. Refer to Chapter II.B.
2.2 HPI Coolina Reauired? (Detailed discussion in Section 3.2)
The operator continually monitors RCS and SG conditions to determine if HPI cooling is required. HPI cooling is required for the following situations:
A. Feedwater Available If the RCS is saturated and no RCPs are operable, then HPI cooling must be established if primary to secondary heat transfer is not established after making the SG a heat sink per Section 2.7 (i.e. the SG saturation temperature is about 50 0F lower than the incore T/Cs and the SG water level is raised to the loss of SCM setpoint.)
B. Feedwater not Available With FW not available, HPI cooling will eventually be required. HPI cooling must not be intentionally delayed beyond the point when RC pressure reaches the PORV 7-23-85 III.C-4 DATE: PAGE
B7fNP 20007 3 (9-84)
SABCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION 74-1152414-oo
[\)
%/
TECHNICAL DOCUMENT setpoint or the first PORV automatic lift. (For Davis-Besse only - HPI cooling, as described in Section 3.3.B, must be established immediately after loss of heat transfer if FW is not available).
2.3 Establish HPI Coolina (Detailed discussion in Section 3.3)
First, two HPI pumps are started or verified operating with suction from the BWST. (The HPI pumps may have previously been started on a loss of SCM.) The operator can open the PORV immediately or he can let the increase in RC pressure open the PORV. If no HPI pumps start, do not open the PORV. If flow from only one HPI pump starts, then the PORV must be opened. (For Davis-Besse only - the actions for establishing HPI cooling and for opening the PORV are dependent on makeup flow and RCS temperature as
} described in Section 3.3.B).
Qj When establishing HPI cooling, the PTS limit may become applicable. Consequently, the RC pressure may have to be controlled to prevent exceeding the PTS limit. This may require opening the PORV when the RC pressure reaches the PTS limit rather than allowing RC pressure to increase to the PORV,open setpoint. Refer to Chapter IV.G for details on how to prevent exceeding the PTS limit.
RCS heat input may be minimized by reducing the number of running RCPs. If possible, at least one RCP should be run to promote thermal mixing of the HPI in the RC.
Refer to Chapter IV.B for HPI equipment operation and Chapter V.A for HPI Specific Rules. Refer to Chapter IV.G for a discussion regarding PTS and RV P-T limits during HPI f '3 cooling. Refer to Chapter III.G for cooldown guidelines (j' during HPI cooling.
l DATE: 7-23-85 PAGE III.C-5 i
BcNP 30007 3 (9-84)
B ABCOCK & WILCOX NUMBER NUCLEAR POWER OlVISION 74-1152414-o0 TECHNICAL DOCUMENT 2.4 FW Flow Exists?
The operator must now determine whether or not FW flow exists before taking further actions. There are three possibilities: heat transfer may have been lost because of a LOFW, primary side steam voids blocking natural circulation (decreasing RCS inventory) or RC pressure is below SG pressure. Control room indications should be monitored closely to determine if adequate FW flow and SG levels exist. Alt irnate channels of instrumentation should be checked to guard against instrumentation error.
2.5 Take Actions to Restore FW Flow (Detailed discussion in Section 3.4)
Initiate or verify automatic initiation of AFW. If AFW is initiated, then establish the appropriate SG level while preventing excessive overcooling and depressurization of the primary by controlling AFW flow.
If AFW is not available, then attempt to restore MFW. If neither AFW nor MFW is available, then feed the SG(s) with an alternate source of FW if available.
2.6 FW Flow Restored?
While attempting to restore FW, the operator should continu-ally determine if HPI cooling is required. If FW is restored before HPI cooling is required, then actions should be taken to restore heat transfer.
2.7 Take Actions to Restore Primary to Secondary Heat Transfer (Detailed discussion in Section 3.5)
First, the SG pressure is reduced to create a RC to SG l
temperature difference of about 50F and the SG level is increased to the appropriate level as determined by RCS conditions, i.e., saturated or subcooled and RCPs on or off. Refer to Chapter IV.C.
DATE: 7-23-85 PAGE III.C-6
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BASCOCK & WILCOX NUCLEAR POWER DIVISION TECHNICAL DOCUMENT
(
Second, stimulate primary to secondary heat transfer depending on the plant condition as follows:
A. If the RC is subcooled with an oDerable RCP, start a RCP. (Refer to Chapter IV.A for RCP restart criteria)
B. If the RC is subcooled and no operable RCP
- 1. Open the RC hot leg HPVs and maintain or increase RC '
pressure.
C. If the RC is saturated with an operable RCP
- 1. Bump one RCP about every 15 minutes.
D. If the RC is saturated and no oDerable RCP, no addi-tional actions are available other than the initial actions to establish the 50F primary to secondary delta T and the appropriate SG level.
2.8 Primary to Secondary Heat Transfer Restored?
I While attempting to restore primary to secondary heat transfer, the operator should continually determine if HPI l cooling is required. If heat transfer is restored before l HPI cooling is required, then controlled decay heat removal
{ with the SG(s) should be established.
The P-T relationship provides the easiest method for recognizing when heat transfer is reestablished. Refer to Chapter II.B for a-discussion of P-T relationship.
l l
DATE: 7-23-85 PAGE III.C-7 I
CONP 20007 3 (9 84)
B Aa COCK & WILCOX NUMBit NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT 2.9 Establish Controlled Decay Heat Removal With SG(s)
Once heat transfer has been restored, it is necessary to establish controlled decay heat removal. If the RCS is subcooled, it may be possible to return to near normal conditions. This would require recovering from HPI cooling, if initiated, by closing the PORV and HPVs, if open, and controlling HPI (HPI/HU) flow. The primary to secondary heat transfer rate should be controlled by adjusting the TBVs or ADVs to maintain the RC temperature at the present value or to begin a controlled cooldown. If HPI cooling has been initiated, there may or may not be a bubble in the pressurizer. If possible, a bubble may be drawn in the pressurizer.
If the RCS is saturated, it will be necessary to begin a cooldown with a saturated RCS. First, however, it will be necessary to recover from HPI cooling and to carefully control the primary to secondary heat transfer rate. With a saturated RCS, a small break LOCI. or an extended LOFW could have occurred. A saturated cooldown with the SGs can be performed while HPI maintains the RCS inventory. Refer to
- Chapter III.B for discussion of saturated cooldown with SGs.
If heat transfer is occurring in only one SG, then actions should continue to attempt to restore heat transfer to the other SG, if possible. There may be other overriding concerns where it will not be possible to reestablish heat transfer to the other SG (e.g., unisolable steam line break or SGTR having occurred in that SG). These are discussed in Chapters III.D and III.E.
! Refer to Chapter III.G for plant cooldown methods.
O DATE: 7-23-85 PAGE III.C-8
BWNP 20007 3 (9 84) l SASCOCK & WILCOX
'I" i ' NUCLEAR power DIVISION 74-1152414-o0 TECNNICAL DOCUMENT 3.0 LACK OF HEAT TRANSFER TECHNICAL BASES Note: The numbered subsections of Section 3.0 correspond i
to the bottom numbers of the appropriate blocks on Figura III.C.
3.1 Identification of a Lack of Heat Transfer The P-T relationship will indicate a lack of heat transfer. !
Refer to Chapter II.B for a discussion of the P-T !
relationship.
3.2 HPI Coolina Reauired Primary to secondary heat transfer.is the primary method for 4 cooling the core. Should this method fail, core cooling can i be provided by HPI cooling. However, HPI cooling must be l established in a timely manner depending on RC and SG conditions to assure that adequate core cooling will be provided. :
l If while attempting to restore FW the RC pressure increases 4 to the PORV setpoint, the RV P-T limit, or the PTS limit, then HPI cooling should be established.
A. Feedwater Available If heat transfer does not start after establishing the
! SG as a heat sink (i.e., increasing the SG level to the loss of SCM setpoint and reducing the SG pressure to create an incore T/C to SG temperature difference of about 50F), then if the RC is saturated and no RCPs are available HPI cooling must be started because no i
subsequent actions can be made to stimulate primary to secondary heat transfer.
! B. Feedwater not available 1
D For a total LOFW, HPI cooling will be required to keep )
f i
the core covered and adequately cooled. There are two I important aspects of HPI cooling. First, operator l
7-23-85 PAGE III*C~9 l DATE:
BONP 30007 3 (9 84)
B A S COCK & WILCOX NUMBER NUCLEAR POWER DIVISION 74-1152414-o0 TECHNICAL DOCUMENT initiation is required. There are no autt.matic systems to initiate HPI cooling. Consequently, the operator needs to continually monitor conditions for when HPI cooling should be started.
Second, HPI cooling does not initially match decay heat. Therefore, it must be started early enough to slow RC inventory depletion enough so that HPI cooling will match decay heat before the core is uncovered.
Consequently, the HPI addition must be started as soon as RC inventory starts being lost, i.e., when the PORV open setpoint is reached or the first automatic PORV lift, whichever occurs first. (Davis-Besse only -
makeup addition should start as soon as a total I,OFW occurs as described in Section 3.3.B).
In additior the rate of heat removal can be increased by opening the PORV. T is will decrease the RC pressure allowing more HPI flow. Refer to Chapter IV.B, Section 2.2.3 for a discussion of when to open the PORV.
If no HPI/MU pumps start, then the operator must not open the PORV because there is no inventory addition available.
3.3 Establish HPI Coolina A. Establishment of HPI cooling is accomplished as follows (except Davis-Besse):
- 1. Two HPI pumps are started and verified to be operating at full flow for the existing RC pressure.
- 2. If flow can be achieved from only one HPI pump, the PORV i must be opened and left open. This will reduce RC pressure and maximize the available flow from the
! degraded HPI system.
l l
7-23-85 PAGE III.C-10 DATE:
1
BWNP 20007 3 (9 84) 8ABCOCK & WILCOX NUMBER m NUCLE 4R POWER DivlSION S 74-1152414-o0 TECHNICAL DOCUMENT L
NOTE: Analyses have been performed that indicate the core will not uncover, and thus be adequately cooled, with only one HPI pump and without opening the PORV. However, considering the collapsed liquid level can approach the top of the core and that the HPI system is already degraded, the PORV must be opened to provide additional assurance of adequate core cooling.
- 3. If full flow can be achieved from two HPI pumps, the PORV may be opened or allowed to open automatically when RC pressure increases to the PORV setpoint. Opening the PORV will reduce RC pressure and allow greater HPI flow thus providing additional core cooling. However, adequate core cooling can be achieved with two HPI pumps f if the PORV is not opened.
V
- 4. If the PORV is not opened, HPI flow must Dot be throt-tied until incore T/cs begin to cooldown Ansi SCM exists. This supercedes the HPI throttling criteria in Chapter IV.B based solely on SCM since, as stated in 3 above, adequate core cooling is assured as long as two HPI pumps are run at full capacity. Once the incore T/Cs indicate decreasing core temperature, the normal HPI throttling criteria (Chapter IV.B) apply.
- 5. One RCP should be left running as long as SCM is maintained to provide thermal mixing. Only one RCP should be used to limit the additional heat input.
- 6. Turn off all pressurizer heaters.
Refer to Chapter IV.B for additional considerations during v HPI cooling.
l DATE: 7-23-85 PAGE III.C-11 i
CCNP 30007 3 (9 84',
EMBCOCK & WILCOX NUCLEAR POWER DIVISION NUMBER 74-11s2414-00 TECHNICAL DOCUMENT B. Establishment of HPI cooling for Davis-Besse is accom-plished as follows:
- 1. As soon as a total LOFW is identified, start the second MU pump (two MU pumps operating) . Fully open the MU control valve and switch MU pump suction to the BWST.
- 2. If only one MU pump is available, then:
- b. Open the HPVs and pressurizer vent
- 3. If both MU pumps are operating, then efforts to restore FW flow and primary to secondary heat transfer may continue until core exit temperature (or hot leg temperature if a RCP is operating) reaches 600F. At 600F the actions listed in 2a-c above must be taken.
NOTE: Analyses have been performed that indicate the l core will not uncover, and thus be adequately cooled, with two MU pumps started and the PORV and HPVs opened at 20 minutes following a total LOFW from 102% of full power. At 20 minutes, the RCS was saturated at about 660F. Thus, with two MU pumps operating, taking the above actions at
~
l 600F will provide additional assurance of core l cooling and should saturate the RCS at a pressure below the shutoff head of the HPI pumps in piggyback operation, providing more injection flow than assumed in the analyses. If only one MU pump is available, it is important to reduce RC pressure as far as possible to maximize available injection flow. Therefore, the actions DATE: 7-23-85 PAGE III.C-12 1 _ _
BWNP 20007 3 (9 84)
SABCOCK & WILCOR NUM8E8 NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 74-1152414-00 of item 2 are taken immediately rather than wait i until the RC temperature reaches 600F.
4
- 4. One RCP should be left running as long as SCM is main-tained to provide thermal mixing. One RCP should be
< used to limit the additional heat input.
- 5. Turn off all pressurizer heaters.
Refer to Chapter IV.B for additional considerations during HPI cooling.
l 3.4 Take Actions to Restore FW Flow If FW flow is not available, the operator must take actions to restore FW flow. The operator should first initiate or verify automatic initiation of AFW. AFW should immediately be throttled as necessary to establish the appropriate SG level and prevent an excessive overcooling and depressuriza-tion of the primary system. Refer to Chapter IV.B for details of AFW throttling. If AFW is not available, it may be possible to regain MFW (for example, if an ICS failure had occurred closing a MFW valve or tripping a MFW pump, it may be possible to take manual control and restore MFW) .
The cause of failure should be examined during the attempt to restore MFW or initiate AFW. Possible causes include j lack of a source of FW, FW pumps failing, and FW valves ,
failing closed. The determination of the cause of failure of AFW or MFW should continue in parallel with restoring i
core cooling immediately even if HPI cooling is required. ;
t 1 If a concurrent loss of both AFW and MFW has occurred, l I
then it may be possible to feed the SGs with alternate j sources of FW using alternate pumps. However, specific {
guidance'on this would be determined beforehand.
DATE: 7-23-85 PAGE III.C-13 !
l
_ . . _ --._ _ ._. . _ _ _ _ , . , _ . _ _ _ _ . - , _ _ _ _ - - _ _ , _ _ . , _ _ _ ~
BWNP 20007 3 (9-84)
BASCOCK & wlLCOX NUCLEAR POWER DIVISION OIII TECHNICAL DOCUMENT 74-1152414-oo In the event that one or both SGs boil dry before AFW or MFW can be restored, FW must be carefully reinitiated into a dry SG. This guidance is provided in Chapter IV.C.
3.5 Take Actions to Restore Heat Transfer With FW available the operator can take actions to restore primary to secondary heat transfer.
The first actions are to establish proper SG level and pressure. This will create the necessary heat sink for RCS heat removal. The SG level which must be established depends upon the RCS conditions e.g., if the RCS is satur-ated, then the SG level must be raised to the loss of SCM setpoint so that boiler-condenser cooling can occur.
(Refer to Chapter IV.C for a discussion of the proper SG levels). Verification of the proper SG level being estab-lished and maintained is vital. SG level instrumentation should be monitored to confirm the proper SG level.
Alternate channels may be checked along with the differ-ent ranges. Caution must be taken in comparing SG levels
, among the different ranges. Some instrumentation may require temperature compensation before they can be used as a valid check of the SG level. If the RC is saturated AFW is preferred over MFW for raising the SG level because AFW instantaneously raises the SG thermal center above that required for boiler condenser cooling. Also in lieu of
- building a SG level when level cannot be increased e.g. ,
after a steamline break, AFW flow can be continuously l added. The SG pressure which must be established depends on .
1 the RCS temperature. The SG pressure must be reduced as j l necessary to create about a 50F temperature difference between the incore T/C and SG. The SG temperature will be determined by the saturation temperature for the existing SG pressure. If primary to secondary heat transfer was lost !
because of RC pressure continually decreasing below the 1
i DATE: 7-23-85 PAGE III.C-14
B7tNP-20007-3 (9-84)
BASCOCK A WitCOX NUMBER gg NUCLEAR POWER DIVISION 74-1152414-oo !
( ') TECHNICAL DOCUMENT G/ l secondary pressure, then cooling can be provided by LOCA/HPI flow (see chapter III.B). If, however, RC pressure stabili-zes below SG pressure, then decreasing the secondary pressure is necessary to enable heat transfer to be re-stored. While lowering secondary pressure, it is important to monitor SG levels very closely. FW flow must be in-creased once heat transfer is established to maintain the appropriate SG level. Provided the RCS hot leg lcops are water solid, heat transfer should be restored once secondary pressure is sufficiently reduced. After establishing the SG as a heat sink the operator will perform actions which stimulate heat transfer. The actions will be different depending on whether the RC is subcooled and whether any RCPs are operable. These actions are as follows:
/h
/ \ A. RC Subcooled With An Operable RCP I
( '
\m/ The RCP should be started. This will establish primary to secondary heat transfer. However, if the SCM is lost after starting the RCP, the pump should be tripped. To limit the chance of losing the SCM when the RCP is started, the operator should increase the pressurizer level and SCM, as discussed in Chapter IV.A, prior to starting the RCP. If only one SG has FW available, the first RCP started should be in that loop if possible.
B. RC Subcooled and No Operable RCP The operator should vent the hot leg by opening the hot leg HPVs to remove any steam or gas which may be blocking natural circulation flow of the RC. This use of the HPVs would only be advantageous if the incore T/Cs are subcooled. While the core region is saturated, the HPVs cannot remove a sufficient amount of steam to G
/
) allow refilling of the hot legs. If all of the steam (n,j' voids can be relieved thrt ugh the HPVs before the RCS l becomes saturated or the PTS limit is reached, then it DATE: 7-23-85 PAGE III.C-15
BcNP 20007 3 (9-84)
S A B CO CK & walCOE NUCLEAR POWER DIVISION NUMBER TECHNICAL DOCUMENT 74-11s2414-00 may be possible to restore heat transfer. The hot leg level measurement indication, if available, might be used to confirm that the steam void is being elim-inated. However, the level measurement may not be correct while the HPVs are open and for a while after closing them. The SG pressure should be further lowered to create about a 100F temperature difference between the incore T/C and SG.
The SG thermal center should be raised by simultane-ously using two approaches: increasing the SG level to the loss of SCM setpoint and adding water to the SG using AFW (AFW enters the SG at a high elevation) . If the RC is subcooled and no RCPs are operable, raising the SG level to the loss of SCM level using AFW may be desirable to get the SG thermal center as high as possible to stimulate RC flow. A steam void in the loop may penetrate below the loss of SCM setpoint but not the natural circulation setpoint. Increasing the SG water level and using AFW will condense scme of the steam void. Also the water in the cold leg may have cooled considerably (due to seal injection, etc.) re-quiring a higher SG thermal center to move the cold slug of water over the RCP lip and into the RV.
I C. RC Saturated With An Operable RCP While trying to stimulate primary to secondary heat transfer, the RC pressure:may be reduced by opening the !
1 PORV whenever the RC pressure increased to the PORV i setpoint. This will allow increased HPI flow for core !
cooling while attempting to establish primary to !
secondary heat transfer. However, the PORV should be closed when the RC pressure decreases to the point where the temperature difference between the incore T/C and SG is about 50F. This is to maintain the SG as a heat DATE: 7-23-85 PAGE III.C-16 I
BWNP 20007 3 (9-84)
SABCOCK & WILCOX
- I" f NUCLEAR POWER OlVISION (g) TECHNICAL DOCUMENT 74-11s2414-00 Ad sink. In addition, the PORV should be closed during RCP bumps to prevent loss of RC loop inventory and to ensure that the pressurizer inventory is available to compen-sate for the collapse of RC loop voids.
To stimulate primary to secondary heat transfer, the operator should first bump RCPs. (Refer to Chapter IV.A for details on bumping a RCP) . If loucring secondary pressure did not restore heat transfer, then steam voids (or gas) must be blocking the primary hot leg loops. It 4.s necessary to eliminate the9e voids before natural circulation can be established. For this reason, it is important to maintain as much HPI flow as possible. Bumping an RCP requires starting and running the RCP until the motor current drops back to normal,
[ , then turning it off. (The breaker may be damaged if the
( l
(./ RCP is tripped while passing a high current) . Once an RCP is bumped, the voids in the hot leg will be flushed into the SGs where condensation of steam will occur.
The condensation of the steam voids will cause RCS pressure to decrease. The amount of RCS pressure decrease will depend upon the volume of steam voids in the RCS. RCS pressure response depends upon initial conditions. This response is discussed in Chapter IV. A.
One RCP bump should be made about every fifteen min-utes. If only one SG has FW, attempt to bump RCPs only in that loop. The RCP NPSH can be violated when bumping RCPs under these conditions. However, do not override any RCP starting interlocks. (TVA only - the automatic RCP trip function will need to be defeated). If the RCP bumps have not started heat transfer, then further lower the SG pressure to create a RC to SG temperature p
/ h difference of about 100F.
\ )
x_-
DATE: 7-23-85 PAGE III.C-17
BWNP 20007 3 (9-84)
BASCOCK & WitCOX NUMBER NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 74-11s2414-oo After further lowering the SG pressure, the RCP bumps should be continued until primary to secondary heat transfer is restored or a decision is made to establish HPI cooling.
D. RC Saturated and No Operable RCP The operator can do nothing additional to stimulate primary to secondary heat transfer. Therefore, HPI cooling should be established if primary to secondary heat transfer does not occur after the SG has been established as a heat sink.
O O
DATE: 7-23-85 PAGE III.C-18
Figure III.C-1 LACK 0F ADEQUATE PRIMARY TO III.c SECONDARY HEAT TRANSFER FLOWCHART il 2.1 l IDENTIFICATION OF LACK OF ADEQUATE PRIMARY TO SECCNO-ARY TAT TRANSFER 3.1 if 2.
2.3 l ESTABLISH W I YES W I COOLING COOLING REQUIRED 3.3,IIIG,Ive,IvG PO 2.A FN FLOR EXI 2.5 ]
TAKE ACTIONS TD
'l RESTORE FM FLOW U
2.7 l 3'd TAKE ACTIONS TO RESTORE ADEQUATE PRIMARY TO SECONO.
ARY HEAT TRANSFER ,,
If FW FLOW RESTORED r ADQUATE PRIMARY TO SECOPCARY NO HEAT TRANSFER YES RESTORED e S it ;
YES if 2.9 l ESTABLISH CONTROLLED OH REMOVAL WITH SGs 8,IIID.IIIE.I m 00C. NO. 74-1152414 00 ,
~
k
._ _ ._. _ m _ _ . _ . _ . . _ _ _ . . _
5 8%NP-20007 3 (9 04)
[ i SASCOCK & WILCOR j.
NUMSER NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 74-lls2414-oo j i t t
(
Chanter III.D l -
Excessive Primary to Secondary Heat Transfer ;
l
- 1.o INTRODUCTION )
- The purpose of this chapter is to provide technical bases l for operator actions to terminate excessive primary to {
j secondary heat transfer from the RC to the SG(s) . These !
o
}
bases are applicable whenever a symptom of excessive primary [
! to secondary heat transfer exists. The RC can be either f
! saturated or subcooled. l l
l.1 Concerns and Obiectives ;
1.1.1 Concerns j j Excessive primary to secondary. heat transfer is always j
- _ caused by a failure in the control of secondary side f l parameters. This failure manifests itself as a loss of
} steam pressure or excessive steam or FW flow or perhaps a
- combination of both. While a momentary overcooling is only ;
l troublesome, an extended overcooling is a severe shock to !
i k the plant and requires quick and effective action by the ;
i t
- operator to mitigate the transient. There are several [
i
- concerns as follows
- !
i !
l f i A. Loss of Pressurizer Level l l t I An extended overcooling can result in a loss of i I i
! pressurizer level. This, in turn, causes a loss of RC r i pressure control. An extended overcooling can empty the surge line which results in a saturated RCS.
I i
B. Saturated RCS With Extended Overcoolina A large steam line break or extended overcooling (i.e. ,
l j continued FW with small steam line break) can result in I a saturated RCS. This requires treatment of the top priority symptom which is a lack of adequate SCM. After I
! DATE: 7-23-85 PAGE III.D-1 l
BWNP 20007 3 (9 84)
SABCOCK & WILCOX NUCLEAR POWER DivlSION NUMBER 74-1152414-00 TECHNICAL DOCUMENT treating the lack of adequate SCM (e.g., tripping RCPs and initiating HPI), the excessive overcooling should be treated.
C. Possible SG Damace A rapid overcooling can result in severe SG damage from tube vibration due to high steam / feed flow as well as thermal shock to the SG. The potential exists for an SGTR to occur from the overcooling. If the steam line break is unisolable it will be necessary to boil the SG dry. Tube-to-shell delta T limits are important and reintroducing FW to a dry SG must be performed carefully (See IV.C).
D. Pressurized Thermal Shock (If applicable)
With an extended overcooling, thermal shock becomes a concern if the overcooling results in cooling the RCS below 500F at a rate greater than 100F/hr. Then the RCS must be prevented from repressurizing to ensure that PTS limits are not violated. This requires action on the part of the operator to control RC pressure and tempera-ture.
1.1.2 Obiectives The main objective for the operator is to terminate the l overcooling transient as quickly as possible. This will minimize the contraction of the RCS and minimize the potential for exceeding PTS limits as well as minimizing possible SG damage.
l If the RCS does cool below 500F at a rate greater than 100 F/hr it will be necessary to maintain the RC P-T within PTS limits. The operator must carefully control HPI to prevent RCS repressurization.
DATE: 7-23-85 PAGE III.D-2
B"tNP-20007 3 (9-84)
BASCOCK & wtLCOX m NUCLEAR POWER DivlSION 74-11s2414-00
[V) TECHNICAL DOCUMENT Once the overcooling is terminated it is necessary to restore control of decay heat removal.
1.2 Causes In general, an overcooling is caused by either excessive steam flow (failure of SG pressure control low) or excessive FW (failure of FW inventory control high).
Overcoolings can result from failed open TBV's, ADV's, MSSV's, or steam line breaks. Overcooling can also occur from excessive MFW or AFW.
2.0 DIAGNOSIS AND MITIGATION The flowchart of Figure III.D-1 should be used in conjunc-tion wit.h the following discussion. The numbered subsec-O t l tions of Section 2.0 correspond to the upper numbers in the (j blocks of Figure III.D-1.
2.1 Identification of Excessive Heat Transfer (Detailed discussion in Section 3.1)
An overcooling transient is best identified by the P-T relationship (Refer to Chapter II.B) .
2.2 Control Pressurizer Level It is inportant to control pressurizer level during an overcooling transient. Increased MU flow and even HPI flow should be utilized to prevent pressurizer level from going off scale low. If the pressurizer surge line is emptied, the RCS will become saturated. Early initiation of additional MU or HPI will help to prevent a saturated RCS condition or minimize the time the RCS is saturated. However, with an overcooling transient in progress, careful control of HPI or MU flow is required to ensure the PTS limit is not
\g violated.
DATE: 7-23-85 PAGE III.D-3
BWNP-20007 3 (9-84)
SABCoCK & WILCOX NUCLEAR POWER DIVISION NUMBER 74-1152414-o0 TECHNICAL DOCUMENT 2.3 Overcoolina SG ADDarent? (Detailed discussion in Section 3.2)
SG and RC loop parameters should be compared between SGs to determine which SG is causing the overcooling. If the overcooling is small it may be difficult to determine which SG is causing the overcooling. It is pocsible that both SGs could be causing the overcooling.
2.4 Isolate Both SGs If the overcooling SG is not apparent or if both SGs are causing the overcooling, then it may be necessary to isolate both SGs. Isolating a SG means to stop all FW flow (MFW and AFW) and steam flow (close TBVs, ADVs, steam supply to FW pumps, etc.). If a large overcooling exists due to exces-sive MFW such that steam line flooding is imminent, then it is necessary to trip both of the running MFW pumps. If the MFW overfill is not as severe then it should be adequate to close the appropriate FW valves. Steam line break control systems may close all or some of the FW and steam valves.
In this case, it is only necessary for the operator to verify that these automatic actions have terminated the overcooling transient and, if necessary, close additional valves.
2.5 Isolate the Overcoolina SG If the overcooling SG has been identified then that SG should be isolated. Isolating a SG means to stop all FW flow (MFW and AFW) and steam flow (e.g., close TBVs, ADVs, steam supply to FW pumps, MSIVs etc.) If the steam lines are in danger of being flooded, then it may be necessary to trip the running MFW pumps. FW flow should be maintained to the unaffected SG and cooling stabilized using the unaffec-l ted SG.
1 O DATE: 7-23-85 PAGE III.D-4
l 87tNP-20007 3 (9 84)
' SADCOCE & WILCOX MumSER NUCLEAR POWER DIVISION 74-11s2414-oo ;
TECHNICAL DOCNNENT i
Isolation of a SG or both SGs should always follow a logical progression of increasingly more drastic attempts to isolate the SG. For example, if the overcooling is not severe it may be possible to close both the TBVs and ADVs as well as the auxiliary steam valves in an attempt to isolate the SG. If this does not work then, for those plants which have main steam isolation valves, the main steam isolation i valve should then be closed. However, in a more serious l overcooling it may be necessary to close the main steam isolation valve first, in an attempt to isolate the steam leak. Provisions must also be made for frequent occurrence of MSSVs leaking immediately after trip. In most cases a l leaking MSSV can be reseated by simply lowering SG pressure un' the leaking valve ressats, then normal SG pressure can be stored.
O2.6 SG Pressures and Levels Have Stabilized (Detailed discussion in Section 3.3)
The required actions from this point depend upon determining whether or not SG levels and pressures have been stabil-ized. If both SG pressures and levels have stabilized (or SG pressures begin increasing) then the overcooling trans-ient has been terminated. The cause of the overcooling was either an isolable steam leak (failed open TBV, ADV or steam line break downstream of the isolated valves) or excessive feedwater which has now been terminated. On the other hand,1f SG pressures and levels are not stabilized then one or both SGs have an unisolable steam line leak.
2.7 Reestablish Heat Transfer to One or Both SG(s) (Detailed discussion in Section 3.4)
The following operator actions will be required:
A. If FW was isolated, FW must be. reintroduced to the available SG(s).
DATE: 7-23-85 PAGE III.D-5
BONP-30007 3 (9-84)
S ABCOCK & WILCOX NUCLEAR POWER DIVISION NUMBER 74-115241e-o0 TECHNICAL DOCUMENT B. TBVs or ADVs must be utilized to stabilize primary to secondary heat transfer.
C. Reheat and swell of the RCS should be prevented to maintain the RCS within pressure limits and prevent filling the pressurizer.
D. Maintaining pressure limits may also require throttling HPI and even opening the pressurizer vent /PORV to limit repressurization.
2.8 Trickle Feed One or Both SG(s) or HPI Coolina (Detailed discussion in Section 3.5)
If pressure and level cannot be restored to either SG, it may be possible to trickle feed one or both SGs. Careful control of AFW flow rates is necessary to maintain RCS cooldown limits. AFW should be used for feeding a dry SG.
However, a SG with a steam line break inside the RB should not be trickle fed if the steam release to the RB is determined to be inappropriate. HPI cooling must be used if trickle feeding is not possible.
2.9 Verify Stable Plant Conditions As soon as the overcooling is terminated and heat transfer has been reestablished, it is necessary to verify stable plant conditions. The operator should continue to monitor for all the symptoms of upsets in heat transfer. For example, immediately after an overcooling has occurred, the operator should check for a lack of adequate SCM. A prolonged overcool'ing can result in a loss of adequate SCM, in which case the operator would have to take actions. See Chapter III.B for details of treatment for a loss of adequate SCM. It is also necessary to check for a SGTR.
Excessive overcooling could result in damage to the SG tubes, thereby causing a tube leak. Verification that l a SGTR has not occurred may require steam line radiation monitoring in the case of an unisolable steam line break.
l DAT5: 7-23-85 PAGE III.D-6
h B7.'NP 20007 3 (9-84)
J 5ADCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION 74-1152414-00 TECHICAL DOCUMElli If the break is upstream of the steam line monitors, the steam line monitors may not give an indication that a SGTR had occurred.
3.0 TECHNICAL BASIS The flowchart of Figure III.D-1 should be used in conjunc-tion with the following discussion. The numbered subsec-tions of section 3.0 correspond to the bottom numbers of the appropriate blocks on Figure III.D-1.
3.1 Identification of Excessive Heat Transfer The P-T relationship is the quickest and most accurate means of determining that an overcooling is occurring. The characteristics of an " overcooling" type transient are discussed in Chapter II.B.
3.2 Overcoolina SG ADDarent?
Identification of the overcooling SG will allow early isolation of the overcooling SG while not disturbing FW flow to the non-overcooling SG. This is the preferred method for mitigating this type of transient.
To identify the overcooling SG, first compare SG and RC loop parameters. The SG associated with the lowest and fastest decreasing T cold is the overcooling SG. It is possible for excessive MFW in one SG to appear as a steam leak in the non-overfed SG because the non-overfed SG will be losing pressure faster than the SG with excessive MFW. This is because excessive MFW quickly raises SG level in the overfed SG which compresses steam in the top of the SG causing the overfed SG to maintain pressure. At the same time, however, excessive MFW will be rapidly overcooling the RCS. This l causes reverse heat transfer in the non-overfed SG and will ,
i cause the non-overfed SG to lose pressure. Consequently, the pressure reduction in the non-overfed SG, may at first DATE: 7-23-85 PAGE III.D-7 l
BCNP 20007-3 (9 84)
B ASCOCK & wlLCOX NUMBER NUCLEAR POWER DIVISION 74-11524I4-oo TECHNICAL DOCUMENT appear to be caused by a steam leak on the non-overfed SG; !
therefore the operator must also check MFW flow rates and ;
levels to insure that excessive main feedwater is not occurring. Stopping the excessive MFW flow will end this phenomenon. For some plants if a large overcooling occur- I red, the automatic systems for steam line break or overfill may be actuated. If an automatic system is actuated the operator must verify that all of the actions have occurred as designed. The most important verification after actua-tion of an automatic system is to ensure that the right SG has been isolated and that the overcooling has been termin-ated.
3.3 SG Pressures and Levels Have Stabilized Whether or not SG pressures and levels have stabilized will indicate if the overcooling was due to an isolable or nonisolable steam leak or excessive FW (MFW and AFW). If SG pressures and levels have stabilized, then the overcooling was isolable and the overcooling has now been terminated by stopping FW flow and closing steam valves. If SG pressures and levels continue to rapidly decrease then the leak is nonisolable. The overcooling SG(s) must then be allowed to boil dry.
3.4 Reestablish Heat Transfer to One or Both SGs Since the overcooling has been terminated and pressure and level has stabilized in at least one SG, heat transfer can be reestablished to the good SG(s) . At this point it is possible that only one SG was isolated. In this case, heat transfer has been maintained in the unaffected SG. FW can then be carefully reintroduced to the isolated SG if stable and its steam valves manually controlled. If both SGs have been isolated then heat transfer must be carefully restored to the good SG(s).
i DATE: 7-23-85 PAGE III.D-8
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At this point, the cause for the overcooling may not have l yet been determined. consequently, attempting to l reestablish heat transfer in the isolated SG(s) may reinitiate the overcooling transient. For example, if the cause of overcooling was a failed ICS controller to the 1
TBVs, then when the operator attempts to restore automatic control of the TBVs the overcooling would be reinitiated.
In this case the operator should restore manual control of the TBVs and maintain manual control until the problem with the automatic control could be determined and fixed.
I Immediately after termination of a severe overcooling transient, it is necessary to reestablish controlled heat transfer. If the RCS is cooled at greater than the allow-able cooldown rate to below 500F, the PTS limits (if
! applicable) must be maintained. Observing PTS limits is i
necessary, even if momentary opening of the pressurizer vent /PORV is required to maintain the RCS pressure. The SG pressure should be controlled to prevent reheat and swell of the RC. HPI must also be carefully throttled to prevent repressurization of the RCS. Refer to Chapter IV.G for discussion of pressure limits.
3.5 Trickle Feed SG(s) or HPI Coolina If restoration of pressure and level is not possible in both SGs, then it may be possible to trickle feed one or both SGs to perform the cooldown. This would be necessary in the l case of an unisolable steam line break on both SGs. The most likely reason for this would be a failed MSSV on each steam line.
If it is decided to perform the cooldown by using trickle feeding, it will be necessary to control the rate of AFW O) addition to the SGs to maintain RCS cooldown limits. The FW flow rate should be adjusted to get the desired cooldown DATE: 7-23-85 PAGE III.D-9
BWN? 20007 3 (9-84)
SABCOCK G WILCCX NUMBES NUCLEAR POWER O! VISION 74-1152414-oo TECHNICAL DOCUMENT rate. Once the proper flow rate is attained, this flow rate O will have to be gradually decreased in order to match the exponentially decreasing decay heat. Refer to Chapter IV.C for control of feedwater to steam generators that cannot hold pressure.
If trickle feeding is not possible, HPI cooling must be started (refer to Chapter III.C). The quantity of available FW should also be considered before deciding on trickle feeding. Since the FW is being irretrievably lost through the unisolable break, it may be necessary, at some point in the cooldown, to switch to a lower quality FW or HPI cooling.
O I
l O
DATE: 7-23-85 PAGE III.D-10
Figure 111.0-1 EXCESSIVE PRIMARY TO ,
SECONDARY HEAT TRANSFER FLOWCHART III.D 0
2.1 l IDENTIFICATION OF EXCESSIVE PRIMARY TO SECOPCARY HEAT TRANSFER 3.1 2.2 l CONTROL PZR LEVEL 2.
OVERC00 LING 2,4 l 60 ISG. ATE W ARENT BOTH SGs v YES 11 2.5 l ISOLATE OVERC00 LING SG
__ l' H
SG LEVEL AND PRESSURES FAVE STABILIZED NO IN (NE OR BOTH r SGs MY YES 9 g 2.8 l 2.7 l TRICKLE FEED ONE REESTABLISH PRI. OR BOTH SGs OR MARY TO SECO OARY WI MING HEAT TRANSFER IN ONE OR BOTH SGs 3.4 3.5
=
2.9 l VERIFY STABLE PLANT COPOITIONS 000. NO. 74-1152414 J
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_ _ . _ _ _ _ . . _ _ _ _ ~ . . _ ._ ._ _ . _ _ _
EWNP 20007 3 (9-84) j SABCOCK & WKCOM NUCLEAR POWER DIVISION N"8EE 74-11s2414 00 TECHNICAL BOCUMENT i .
Chanter III.E 4
Steam Generator Tube Ruoture 1
1.0 INTRODUCTION
i This chapter provides the technical bases for the guide-lines to mitigate steam generator tube ruptures (SGTR).
The purpose of these technical bases is to provide ;
+
sufficient information regarding the expected NSSS l
j response during SGTR transients such that, when coupled l with the related technical bases in other chapters i
referenced herein, the user can develop plant specific j procedures for diagnosis and mitigation of SGTRs.
i 1.1 Concerns and Obiectives Relative to SGTRs 1.1.1 Concerns l A SGTR is a break of the RCS pressure boundary which is i the second of the three barriers to fission product release. This provides a path through the. third barrier j (reactor building) via the steam lines. Specifically, the concerns associated with SGTRs are:
l A. The loss of RC (a SGTR is essentially a SBLOCA) with I the potential threat for interruption of core cooling.
B. The potential for radiation release outside the containment. '
f C. Management of contaminated secondary water.
l D. Depletion of BWST inventory with no borated water accumulation in the RB sump, i.e., sump recirculation, l if needed, may not be available.
l
- 1.1.2 Obiectives i The objectives to be considered during the mitigation of l i SGTRs are (listed in order of relative priority)
l DATE: 7-23-85 PAGE III-E-1
BWNP 20007 3 (9-84)
BARCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION n -1152 m -oo TECHNICAL DOCUMENT A. Maintain core coolina - Core cooling is always a top priority, however, with one, and possibly two, barriers to the release of fission products already breached, clad integrity becomes even more important.
B. Minimize radiation release to the atmosphere - The intent is to steam the affected SG(s) as much as possible to aid the RCS cooldown, but limit the releases to less than pre-determined limits.
C. Minimize the intecrated tube leakaae - This objective consists of two basic parts; minimize the leak flow rate and minimize the duration of the transient before the leak flow is terminated. The purpose of this objective is also two-fold; a lower integrated leakage minimizes storage problems associated with the contaminated secondary water and minimizes the potential for BWST depletion. In addition, the accomplishment of this objective directly aids the objective of minimizing radiation release in 1.1.2.B above.
NOTE: Reference is made throughout this chapter to "affected SG," "most affected SG," and "least affected SG." These terms are used to denote, respectively, the SG with a SGTR (when only one SG has a SGTR) the SG with the largest leak rate when both SGs have SGTRs and the SG with the smallest leak rate when both SGs have SGTRs. i l
2.0 DIAGNOSIS AND MITIGATION OF SGTR This section provides a brief description of the recom-mended logic to be used in dealing with SGTRs from the initial diagnosis through plant cooldown and depres-O DATE: 7-23-85 PAGE III-E-2
EWNP 20007 3 (9 84)
SASCOCK & WILCOM NUCLEAR POWER DIVISION NUMBER 74-11s2414-oo hTECHNICAL DOCUMENT surization to decay heat removal system (DHRS) operation when the leak flow can be terminated.
Figure III.E-1, "SGTR Functional Flow Diagram," provides a basic action / decision logic chart for mitigating SGTRs. l This chart is intended to provide major decision points and major function level actions. In addition, the chart is organized to present a recommended logic and priority l in dealing with the different aspects and complications that can occur during a SGTR. However, the order that items appear in the flow chart is not intended to repre-sent the only logical sequencc. The primary purpose of the flow chart is to identify the major aspects of SGTRs in some logical order. The user should address these aspects in plant specific procedures in the order which best fits the specific plant capabilities.
A loop is provided in the flow chart from 2.6.1 (when l
1 plant conditions are not yet established for DHRS opera-tion) back to 2.2.1 (subcooling margin status). The purpose of this loop is to signify continuous surveillance ;
of key plant conditions during the cooldown that may a) l require changing cooldown methods (e.g., approaching an !
l alternate control criterion, 2.4.1) or b) permit changing !
to an alternate cooldown method (e.g., RCP restart
[
l criteria satisfied, 2.5.2). l l I l I The numbered subsections in the remainder of Section 2.0 l t
j correspond to the upper numbers in the blocks on Figure :
III.E-1.
I t ;
I DATE: 7-23-85 PAGE III-E-3 l i l l .
BONP-30007 3 (9-84)
BABCOCK & witcom NUCLEAR POWER OlVISION NMH TECHNICAL DOCUMENT 74-11s2414-oo 2.1 Identification of SGTR and Pla_nt Shutdown 2.1.1 Identification The most rapid means of identifying that a SGTR has occurred while at power is by secondary plant radia-tion monitors (e.g. the condenser air ejector, vacuum pump exhausts or main steam line radiation monitors. Main steam line monitors may also identify the affected, or most affected, SG.
Backup methods include primary leak rate calculations, SBLOCA symptoms (except changes in RB environment), SG samples, reduction in the steam outlet temperature, and anomalies in FW flows and/or SG levels after reactor shutdown. Methods for identifying a SGTR and the affected SG are discussed in 3.1.
2.1.2 Plant runback and Reactor Shutdown If possible, run back the plant as quickly as possible, but in a controlled manner, to well within the turbine bypass system capacity before tripping the reactor to prevent lifting of the MSSVs. This includes cases where maximum MU or HPI flow is required to keep up with the l tube leak and is intended to minimize atmospheric releases and avoid the possibility of a stuck open MSSV. Plant runback is discussed in 3.2 and 3.4.1.1.
2.2 Subcoolina Marain and Primarv to Secondary Heat Transfer l
2.2.1 Subcoolina Marain Verification of SCM is the first major decision point in the flow chart for several reasons:
- a. If the tube leakage is large enough to cause a loss of adequate SCM (which requires complete failure of several tubes), the loss of adequate SCM will occur early in the transient.
l DATE: 7-23-85 PAGE III-E-4 l
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NUCLEAR POWER DIVISION 74-1152414-oo
. TECHNICAL DOCUMENT
- b. On any loss of adequate SCM, it must be assumed that a
- r i SBLOCA exists which requires certain prompt actions, l l discussed in 2.2.2. t
, c. Maintaining SCM is very important in maintaining l optimum control of a SGTR transient and performing an l expeditious cooldown.
l The flow chart continuously loops back to this decision
! point for the duration of the cooldown until the DHRS is in operation. This signifies a constant surveillance of .
SCM and other key plant conditions during the performance j of any subsequent portion of the flow chart.
l ,
The content and definition of SCM is discussed in Chapter 4
II.B. During the cooldown it is highly desirable to
. maintain RCS pressure and temperature close to, but above, f the minimum SCM. This minimizes the differential pressure between the RCS and the affected SG(s) , thus minimizing i the tube leak flow rate.
[ 2.2.2 Loss of Adecuate Subcoolina Marain ;
on loss of adequate SCM,-the operator must assume a SBLOCA exists and perform the following actions: i
- a. Trip all RCPs. l
l c. Begin (or verify) raising SG level (s) to the loss of :
! SCM setpoint. (SG level setpoints are defined in I
Chapter IV.C.
i These actions, and their bases, are discussed in detail in
! Section III.B. Actions a and b above must always be
- performed on loss of adequate SCM. However, control of SG l levels may differ slightly for SGTRs. Normally, a loss of l
t DATE: 7-23-85 PAGE III-E-5 !
l
EWNP 20007 3 (9 84)
BABCOCK & WILCOX N W SER NUCLEAR POWER DIVISION 74-lIs2414-oo TECHNICAL DOCUMENT adequate SCM requires raising the level in both SGs to the loss of SCM setpoint; however, this may cause subse-quent inventory control problems in the affected (leaking)
SG. Therefore, if full HPI flow from two HPI pumps exists and filling of the affected SG is not desired at this time, only raise the level in the unaffected SG (or least affected SG, if both SGs have tube leaks) to the loss of SCM setpoint. Do not intentionally raise the level in the affected, or most affected, SG. Instead, continually injeci sufficient AFW to maintain natural circulation flow while allowing the level to increase to the loss of SCM setpoint due to the tube leak flow.
If, however, full HPI flow from two HPI pumps cannot be achieved or maintained or the affected SG will be allowed to fill, then raise the level in both SGs to the loss of SCM setpoint.
If SCM is restored at any time during the subsequent cooldown, restart RCPs (if all other conditions for RCP restart exist) and allow the unaffected SG level to boil down to the normal forced flow setpoint, or normal natural circulation setpoint if the RCPs were not restarted. In ,
addition, throttle HPI flow as necessary to maintain I pressurizer level and maintain RCS pressure close to, but l above the SCM and, if applicable, below the pressurized l thermal shock (PTS) limit (Chapter IV.G).
2.2.3 Primary to Secondary Heat Transfer The second major decision point in the flow chart is verification of controlled primary to secondary heat transfer. A major goal in mitigating SGTRs is to achieve an orderly but expeditious cooldown to cold shutdown conditions. This goal is best achieved with forced flow ,
l I
DATE: 7-23-85 PAGE III-E-6 1
h BYNP 20007 3 (9 84) j- BASCOCK & WitCOE N W stb f NUCLEAR POWER DIVISION
' 74-11s241*oo TECNNICAL 90COMENT i
and controlled primary to secondary heat transfer to both l
SGs.
j Methods of verifying heat transfer and recognizing a loss of primary to secondary heat transfer are discussed in ,
! Chapter III.C. Methods of recognizing and mitigating excessive primary to secondary heat transfer are discussed ,
I in Chapter III.D.
i
! 2.2.4 Loss of Controlled Heat Transfer If primary to secondary heat transfer does not exist to either SG, restore primary to secondary heat transfer to at least one SG as soon as possible, even if it is the
! affected or most affected SG.
Possible causes for loss of SG primary to secondary heat l
- transfer and methods of restoring heat transfer are j discussed in Chapter III.C. Possible causes for excessive t
pr3. mary to secondary heat transfer and methods of re-storing controlled heat transfer are discussed in Chapter III.D. In addition, special considerations for steam leaks concurrent with SGTRs are discussed in Section 3.8
! of this chapter.
i i
- 2.2.5 Restoration of Heat Transfer As soon as controlled primary to secondary heat transfer is restored to one SG, begin cooldown while attempting to l
restore controlled primary to secondary heat transfer to j the other SG.
1 :
I If heat transfer was lost and cannot be restored to either l l SG before RCS pressure reaches the PORV setpoint, initiate ;
i HPI cooling (2.4.6)._ Recognition of heat transfer [
restoration is _ also discussed in Chapter III.C. Chapter !
i !
- -r l DATE
- 7-23-85' PAGE- III-E-7 l
l l
BONP 20007 3 (9 84) l l
BABCOCK & WILCOK NUMBER NUCLEAR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT III.D provides methods for terminating excessive heat transfer transients and restoring controlled cooling including control methods for unisolable steam leaks. j Unisolable steam leaks impact SGTR transients as discussed j in Section 3.8 of this chapter. l 2.3 Initial Cooldown and Ootional SG Isolation 2.3.1 Cooldown and Deoressurization to Below MSSV Setooint The intent of the initial cooldown is to prevent / minimize lifting of the MSSVs by reducing primary hot leg tempera- _
ture below the saturation temperature corresponding to the lowest MSSV setpoint (requires cooldown to approxi-mately 540F for 177 FA plants or 560F for 205 FA plants).
Reduce RC pressure below the lowest MSSV setpoint to preclude lifting of the MSSys in the event the affected SG(s) filled solid. In order to maintain SCM, this requires further cooldown.
Use the normal cooldown rate limit or the emergency cooldown rate limit depending upon conditions and limits provided in Section 3.3 of this chapter.
The normal cooldown rate limit is 100F/hr (1.67F/ min) for 177 FA plants and 50F/hr (.83F/ min) for 205 FA plants.
The emergency cooldown rate limit is 240F/hr (4F/ min) for 177 FA plants and 268F/hr (4.47F/ min) for 205 FA plants.
2.3.2 Decision to Isolate SG At this point, the affected SG may be isolated (i.e.,
termination of feeding and steaming) and the cooldown continued on one SG. However, based on existing plant conditions, it may be desirable to continue the cooldown on both SGs.
7-23-85 PAGE III-E-8 DATE:
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SAsCOCK & wlLCOX NUCLEAR POWER DIVISION NUMett 74-1152414-oo l f TECillllCAL DOCUMElli
's Section 3.3.2, discusses these options and their relative benefits.
This decision point also covers the option to return a i previously isolated SG to service by the loop back through i
! the from 2.6.1 to 2.2.1.
2.3.3 Isolation of the Affected SG once the decision has been made to isolate the affected SG, two options exist for further control of the isolated SG. The SG may remain isolated and allowed to fill or f periodic feeding, steaming and/or draining may be per-formed.
If it is desired to fill the SG, then proceed to 2.4.3.
If, however, periodic feeding, steaming and/or draining is desired, then return to the main path at 2.4.1. Periodic unisolation and steaming, feeding and/or draining of the affected SG may be desired to:
, a. Prevent the SG level from exceeding a predetermined high level limit if subsequent steaming may be desired or if RC pressure is above the MSSV setpoint (requires draining and/or steaming).
- b. Maintain a minimum SG level to promote shell cooling (requires feeding).
- c. Periodically induce natural circulation in the idle ;
RC loop to prevent void formation (requires s, teaming). i i
The relative benefits of these options and the bases for their respective actions are discussed in section 3.3.3. ,
i i > I
! \
DATE: 7-23-85 PAGE III-E-9.
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SABCOCK & wlLCOM NUCLEAR POWER DIVI $l0N NUMttR 74-1152414-0o TECHNICAL DOCUMENT 2.4 Limits on SG Cooldown/ Alternate Methods 2.4.1 SGTR Alternate Control Criteria At this point in the flow chart cooldown is being accom- i plished with either both SGs or on one SG with periodic cooling provided by the affected SG. There are, however, limitations on continued steaming, even periodically, of both SGs.
Three criteria are recommended for use in determining when I an alternate cooldown method should be used. These are designated Tube Rupture Alternate Control Criteria (TRACC) and are based on the following considerations:
- a. Radiation releases approaching predetermined limits.
- b. BWST level approaching a predetermined low limit.
- c. SG filling due to tube leakage despite steaming to achieve the maximum allowable cooldown rate.
Those criteria and their bases are explained in detail in i section 3.4. I 2.4.2 Decision on Use of SG Drains If a TRACC limit is reached, there are two options to be considered: the affected, or most affected, SG can be isolated and allowed to fill or SG drains, if available, may be used to prolong the availability of the SG for steaming.
There are relative benefits for both options that should be considered in making this decision. The benefits and l
! limitations on the use of drains are discussed in section 3.5. The advantages and disadvantages of filling the SG are discussed in sections 3.3.2.1 and 3.3.3.2.
1 01 DATE: 7-23-85 PAGE III-E-10
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SA4 COCK & wtLCOX NUCLEAR POWER DIVI $l0N NUMett 74-1152414-oo TECHNICAL DOCUMENT 2.4.3 Filline of the Isolated SG If the decision was made not to provide for further steaming of the affected SG (or most affected SG if both have tube leaks) in 2.3.3 or 2.4.2, or if use of the drains will not prevent reaching TRACC (2.4.9) then the SG is isolated here and allowed to fill.
As the SG approaches a solid condition, it is important to maintain RC pressure below the MSSV lift pressure since the solid SG will pressurize to RC pressure. Section 3.6 discusses special considerations for isolation and filling of the SG.
2.4.4 Cooldown on One SG Without TRACC If only one SG had a tube leak and was isolated (2.4.3),
cooldown on the remaining SG (if available) should be possible without reaching TRACC.
If both SGs had tube leaks, isolating of the most affected SG (2.4.3) m allow continued cooldown on one SG without reaching TRACC. Therefore, closely monitor the parameters listed in 2.4.1 for TRACC.
2.4.5 Isolation of Remainino SG If cooling down on one SG does not prevent reaching TRACC (because both SGs have tube leaks), then isolate the remaining SG, allowing it to fill, and initiate HPI cooling (2.4.6).
2.4.6 Initiation of HPI Cooling If neither SG can provide heat transfer, initiate HPI cooling.
l s
I DATE: 7-23-85 PAGE III-E-11 1
BWNP 20007 3 (9-80)
SABCOCK & WILCOK NUCLEAR power DIVISION NUM8tt 74-1152414-oo TECHNICAL DOCUMENT If HPI cooling is required due to a lack of adequate primary to secondary heat transfer (e.g., total loss of feedwater) from 2.2.5, then continue attempts to restore heat transfer to at least one SG (covered by eventual loop back through 2.2.3).
If HPI cooling is required due to intentional isolation of both SGs (TRACC, 2.4.5) then HPI cooling will probably be required until DHRS transition unless there is reason to believe that resumed SG operation will not violate TRACC (e.g., pressure and temperature subsequently decreasing to within SG drain capability). This is covered by eventual loop back through 2.4.1.
Methods of initiating and controlling HPI cooling are discussed in Chapter III.G.
2.4.7 Actions to Prevent Liftina MSSVs If HPI cooling is required due to intentional isolation of both SGs (TRACC, 2.4.5), maintain RC pressure below the MSSV setpoint since the SGs may fill Polid.
If HPI throttling (with adequate SCM) will not keep RC pressure below the MSSV setpoint, perform one or more of the following as necessary:
- a. Open letdown line.
- b. Open RCS HPVs.
- d. Open turbine bypass valves.
- e. Open atmospheric dump valves.
These actions and their bases are discussed in Section 3.7 of this chapter.
O DATE: 7-23-85 PAGE III-E-12
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B ASCOCK & WILCon i NU"'II NUCLEAR POWER DliISION f 74-1152414-oo OTECHNICAL DOCUMENT l
- o 2.4.8 Oceration l
t t
The intent of using the SG drains is either to prolong use l of the SGs for the cooldown or to prevent filling of an f isolated SG. Use of the drains to allow continued ,
steaming may be especially effective if the limiting TRACC !
is continued SG fill at the maximum allowable cooldown f rate. Using the drains may also be effective for the other TRACC within the limitations discussed in section l 3.5. !
i 2.4.9 Continued Cooldown/Drainina Without TRACC !
Once SG draining has been established, reevaluate the approach to TRACC to confirm effectiveness of the drains and take the appropriate path as shown. Refer to Section [
3.4 of this chapter for a discussion of TRACC and their {
bases.
t 2.5 Continued Cooldown Usina SG(s) !
2.5.1 Reactor Coolant Pump Status Forced circulation cooldown is preferable to natural j circulation, especially during SGTRs where an expeditious {
- cooldown is desired. I i
Forced circulation eliminates stagnant hot spots that can ,
I
- occur during natural circulation, and thus avoids the t I
complications and cooldown delays that result from void j formation in the RCS. In addition, forced circulation I provides pressurizer spray flow which optimizes RCS pressure control. RCP operation also results in a lower (
RCS differential temperature which allows a lower primary !
I to secondary differential pressure thus minimizing the !
tube leak flow rate. !
d l l 1
% l l l l 0 l
l DATE: 7-23-85 PAGE III-E-13 l w- . . . . _ - . - . - _ . . .. .- . . - . . . . _ - . - . .
ScNP-30007 3 (9 84)
I SABCOCK & WitCOE f NUMBER NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT Finally, forced circulation allows for a faster overall cooldown to DHRS operation thus minimizing the integrated tube leak flow and radiation releases.
2.5.2 Criteria for RCP Restart If the RCPs are not operating, attempt to satisfy the criteria for restarting the RCPs as soon as possible while continuing to cooldown with natural circulation.
The criteria for RCP restart are provided in Chapter IV.A.
Observe RCP NPSH requirements, which may require RC pressure somewhat higher than SCM, during the cooldown when RCPs are operating or restarted.
2.5.3 RCP Restart As soon as the criteria for RCP operation (Chapter IV.A) are satisfied, restart RCPs. The number and selection of the RCPs to be restarted depends on plant conditions.
Running one RCP in each loop balances heat transfer, but other RCP combinations may be desired for higher spray flow capability. (Refer to Chapter IV. A for precautions and recommendations to be considered before RCP restart).
2.5.4 Natural Circulation Cooldown If tube leak conditions are relatively stable (i.e., small leak flow, TRACC not imminent), it may be preferable to maintain existing temperatures and pressures and just remove decay heat with natural circulation until RCPs become available. If, however, continued cooldown in warranted, or RCPs will not be available for an extended period of time, then proceed with a natural circulation cooldown (Chapter III.G).
O DATE: 7-23-85 PAGE III-E-14
e i 1 i 0"4NP 20007 3 (9 St) i eAecocn s wacon 1 I NUCLEAR POWER DIVISION i
74-1152414-oo
! TECMICAL IOCONElli i
2.5.5 Forced Circulation Cooldown l- Continue forced circulation cooldown to DHRS cut-in condi-
! tions (refer to Chapter III.G for discussion and limits).
i 1
2.5.6 Pressurizer Bubble Status
{ ,
The existance or absence of a pressurizer steam bubble !
l I j dictates RCS pressure control methods during the cooldown g (Chapter III.G). f l
I t
2.5.7 Restoration of Pressurizer Bubble l l RCS pressure control with a steam bubble is preferred. If l conditions permit, draw a steam bubble in the pressurizer l (the method for drawing a steam bubble is provided in ;
} Chapter III.G). l
[
P 1 2.5.8 Solid Plant Pressure Control {
j Control of RCS pressure by MU/HPI flow and letdown /- l l leak flow is discussed in Chapter III.G. Draw a pressur-l izer steam bubble, however, should conditions subsequently permit (covered by loop back through 2.5.7).
{
}
2.5.9 Enhanced Idle Imon and RV Head Coolina f
Actions to enhance RV head cooling are only applicable !
during natural circulation when the RV head region is relatively stagnant. Similarly, actions to enhance idle >
RC loop cooling are only applicable during single loop i natural circulation.
$ l i
The objectives in performing these actions are to reduce I thermal stresses on the RV head and to prevent void forma-f 4
tion in the RV head or idle RC loop. Void formation will hinder RCS depressurization in that the void acts like a i pressurizer. This will result in a prolonged cooldown and l higher integrated tube leakage.
t 7-23-85 PAGE III-E-15 DATE:
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B%NP-20007 3 (9 84)
S ABCOCK & WILCOX NUMetR NUCLEAR P0wtR DivislON 74-1152414-oo TECHNICAL DOCUMENT Methods to enhance RV head and idle RC loop cooling are discussed in Chapter III.G.
2.5.lo Void Mitication Actions to mitigate voids are also only applicable during natural circulation cooldown. If the actions to enhance idle loop /RV head cooling are unavailable or unsuccessful and a void does form, hold the cooldown and eliminate the void. Recognition and mitigation of voids is discussed in Chapter III.G. ,
2.6 Transition to Decav Heat Removal System Operation 2.6.1 Plant Conditions continue the cooldown with either the SG(s) or HPI until RCS conditions allow DHRS operation. In addition to the normal required conditions of pressure and temperature, a special case may exist for SGTRs if the RCS is saturated (2.6.2).
A loop is provided back to 2.2.1 when DERS conditions have not yet been attained. The purpose of this loop is to signify continued attempts to establish the most " normal" cooldown mode possible. The required plant conditions for DHRS operation are provided in Chapter III.G.
2.6.2 Initiation of DHRS ODeration Place the DHRS into operation when RCS conditions permit and cooldown and depressurize to cold shutdown. At cold shutdown drain the RCS to an elevation that will terminate or limit the tube leak flow (below the elevation of the SG steam nozzles).
If the RCS is saturated, initiate DHRS operation if the only unidentified PCS leakage is through the SG tubes.
DATE: 7-23-85 PAGE III-E-16 l
l
BINP 20007 3 (9 84) 4 BASCOCK & WILCOM !
5 NUCLEAR PowfR OlvlS40N 74-11s2414-oo
- TECNNICAL BSCWWENT !
\ !
l The reason is that there should be a sufficient RCS liquid i
- elevation to provide adequate NPSH for the DH pump even !
j with.the RCS saturated. However, when initiating DHRS
- operation with a saturated RCS, monitor the suction i pressure of the DHR pump and trip the pump immediately if ,
l NPSH is lost. The other DHR pump should be aligned for
) continued injection (with the HPI pump if necessary) to i makeup for the tube leak. ;
Specifics on DHRS operation are provided in Chapter III.G. [
! 3.o SGTR TECHNICAL BASES I 1
j This section provides a more detailed discussion of the ;
j items unique to SGTRs. Items that are not unique to SGTRs !
I are covered elsewhere in the referenced technical bases, ,
) and augmented. as necessary with aspects to be considered !
I for the case of SGTRs in Section 2'T0. It should be noted that the discussions in 3.1 through 3.7 assume that ;
j unisolable steam leaks do not exist. The impact of j j unisolable steam leaks on SGTR mitigation is discussed in f
] 3.8. The numbered subsections in section 3.0 correspond '
to the botton numbers in most of the blocks on Figure III.E-1. !
i I
3.1 Identification of SGTR and the Affected SG l; When a SGTR occurs, it is important to diagnose the event i and identify the affected SG(s) since subsequent actions [
f j depend on this information. The initial indications of a l I SGTR will usually be either radiation readings / alarms on f the steam lines and/or condenser or an unidentified RCS inventory loss.
I l
- Radiation alarms on the steam lines or condenser are positive indications of a SGTR but may only occur while at f power. If radiation monitors are unavailable or ineffec- !
t DATE: 7-23-85 PAGE III-E-17 l
BWNP 30007 3 (9 84)
S A BCOCK & WILCOX NUMBER NUCLEAR PCWER DivtSION 74-1152414-oo TECHNICAL DOCUMENT tive due to plant conditions, then the first indication may be anomalies in RCS inventory control indicating an unidentified loss of reactor coolant (e.g., makeup flow greater than letdown and seal return flows while maintain-ing a constant pressurizer level). Depending on the amount of tube leakage, indications of a SBLOCA may exist without an accompanying change in RB parameters.
A SGTR may cause the steam exit temperature to be ab-normally low. Tube leaks high in the tube bundle may cause a reduction in the steam exit temperature of up to 9F for a double-ended rupture of a single tube at full power.
If the reactor is shutdown and FW is on automatic SG level control, anomalies may also exist in FW flow rates and/or SG levels (e.g., both SGs maintained at the correct level but one with a significantly lower FW flow rate or an SG level increasing above the level setpoint without FW flow). These indications will not exist while at power since the SGs are not on level control and differences in FW flow rates would be insignificant.
In any case, if a SGTR is suspected, draw SG samples as soon as possible to verify a SGTR exists and to identify the affected SG(s). Coolant activity introduced into one SG through a SGTR can quickly contaminate both SGs through the FW sys'em. Therefore, sample the SGs for boron since boron will normally be contained in the affected SG(s).
3.2 Plant Runback and Reactor Shutdown If a SGTR occurs while at power, it is highly desirable to run the plant back, at least to well within turbine bypass system capacity, before tripping the reactor. The intent DATE: 7-23-85 PAGE III-E-10
8%NP 20007 3 (9-84)
BASCOCK 4 wsLCOX
- NUCLEAR POWER DIVISION 74-11"*'l'-
TECNNICAL DOCONENT is to prevent lifting of the MSSVs on the affected SG(s) for two reasons: 1) any lifting of the MSSVs provides a path for radiation release directly to the atmosphere and 1
1
- 2) less likely, an MSSV may fail to ressat, thus prolong-ing the release to atmosphere and complicating control of the plant cooldown.
l
! When reactor power is within turbine bypass capacity and below the anticipatory trip setpoint for turbine trip, unload and trip the turbine, then trip the reactor.
1 Attempt a controlled plant runback even if HPI (MU) flow j from the BWST is required to keep up with the tube leakage. The added boron from the BWST will reduce reactor power and may induce some RCS cooling if a power 4
mismatch develops.
4 If a reactor trip is unavoidable, perform normal post-trip actions and stabilize the plant as soon as possible to i permit continued cooldown.
3.3 Cooldown and Ootional SG Isolation 3.3.1 Initial Cooldown/ Limits i once the reactor is shutdown, commence cooldown with both f SGs to reduce RC hot leg temperature below 540F for 177 FA
! plants and 560F for 205 FA plants. At this temperature,
, setpoint, thus preventing an MSSV from lifting on an isolated SG. Above this temperature T hot, steaming both
! SGs with turbine bypass (or atmospheric dump valves if necessary) should maintain steam pressure below the MSSV i
setpoint.
i 6 s
DATE: 7-23-85 PAGE III-E-19 l
i
B'%NP-20007 3 (9 84)
B ASCOCK & WILCOX PeUCLEAR power Divist0N 74-1152414-00 TECHNICAL DOCUMENT If the SG is to be isolated, this isolation may be performed anytime after the initial cooldown to 540F Thot for 177 FA plants or 560F T hot f r 205 FA plants (unless an SG is filling - refer to 3.3.1) and continue the RCS cooldown in accordance with the rate limits given in 3.3.1 and depressurize the RCS (while maintaining SCM) to below the lowest MSSV setpoint. This prevents lifting the MSSV should the isolated SG fill and pressurize to RCS pres-sure.
The cooldown should be performed within normal cooldown limits as follows:
- a. RCS cooldown rate $ 100F/hr (1.67F/ min.) for 177 FA plants or i 50F/hr (.83F/ min) for 205 FA plents,
- c. Observe fuel pin in compression limits (except us noted in 3.3.1.3).
- d. Observe RCP NPSH limits In addition, except when fuel pins are in compression or RCP NPSH limits are applicable and are more rostrictivo,
~BCS pressure should be maintained close to, but above, the minimum SCM to minimize RCS-SG differential prosauro. It should be noted, however, that loss of the RCPs will cause T and RC pressure to increase during the transition to hot natural circulation, which could lead to a loss of SCM.
Thorofore, it may be desirable to maintain additional subcooling, especially during periods of potential grid instability (e.g., lightning storms) . If adequate SCM is maintained during the cooldown, isolate the CFTs when l conditions permit. Do not, however, isolate the CFTs if I SCM cannot be maintained.
I DATE: 7-23-85 PAGE III-E-20 l
BINP-20007 3 (9 84)
SABCOCK & wnCOX NUCLEAR POWER Divi $10N NUutta 74-1152414-00 OTECHNICAL 00CullENT V These limits may vary depending on the plant conditions.
Each one is discussed separately in sections 3.3.1.1, 3.3.1.2, and 3.3.1.3.
3.3.1.1 Emercency Cooldown Rate The cooldown rate may be increased to a maximum of 240 F/hr (4F/ min) for 177 FA plants or 268F/hr (4.47F/ min) for 205 FA plants down to 500F Thot NI
- a. the affected SG level (s) is/are increasing rapidly (several large tube leaks) and carry-over could occur before 500F Thot at the normal cooldown rate gI
- b. radiation release rates are projected to reach the integrated TRACC limit (as defined in 3.4.1) before 500F T hat at the normal cooldown rate (again, several large tubs leaks would probably have to occur) .
O The typical plant design allows for 40 cycles for 177 FA plants (80 cycles for 205 FA plants) of an emergency cooldown to 500F T hot at 240F/hr for 177 FA plants j (268F/hr for 205 FA plants) and this rate is allowed for any SGTR event. However, it is recommended that the use of this emergency cooldown rate be limited to the two ,
cases noted because the faster rate of cooldown may I
increase the tube-to-shell delta T and may increase the potential for voiding the RV head region during natural circulation. The tube-to-shell delta T is still limited to a maximum allowable value as discussed in 3.3.1.2.
The emergency cooldown rate 13 recommended for the two cases noted because several large SGTRs and/or a rela-tively high percentage of failed fuel already exist.
j In these cases, it is more important to prevent liquid 1 discharge through the MSSVs and limit the duration of
( high radiation release rates. For the case of impending DATE: 7-23-85 PAGE III-E-21
8 NP-20007 3 (9 84)
B A8 COCK & WRCOX NUM8it NUCLEAR POWER DIVISION 74-1152 m -00 TECHNICAL DOCUMENT carryover, the affected SG (or most affected SG if both have SGTRs) should be steamed as much as possible to achieve the cooldown rate in order to limit the rate of level increase. The limit to be used to determine if carry-over is imminent is plant specific, but a sugc,ssted basis for the limit is provided in 3.6.1.1.b.
Usually, a SG can be isolated after Thot has been reduced to 540F for 177 FA plants or 560F for 205 FA plants.
However, if a SG is filling, T hot should be further reduced before isolation. The additional cooling must be sufficient to allow the RC pressure to be reduced below the MSSV setpoint (plus adequate pressure margin 100 psi) while maintaining adequate RC SCM. This will allow the isolated SG to fill and pressurize to the RC pressure without lifting a MSSV.
For the case of high radiation release rates, the unaf-facted SG (or least affected SG if both have SGTRs) should be steamed as much as possible to achieve the cooldown rate in order to limit the integrated release before the SG(s) can be isolated. If both cases apply, then prevention of carry-over takes precedence. It is especially important to prevent liquid discharge through the MSSVs with high RC activity.
3.3.1.2 Tube-to-Shell Delta T For 177 FA plants, the normal tube-to-shell delta T limit for cooldowns is 100F (tubes colder) and during an emergency cooldown as described in 3.3.1.1, this limit may be increased to 150F. For 205 FA plants, the tube-to-shell delta T limit is"160F during normal and emergency cooldowns. Methods to control tube-to-shell delta T are discussed in Chapter III.G.
DATE: 7-23-85 PAGE III-E-22
CWNP-20007 3 (9 84)
SASCOCK & WILCOR NUa4 Sit NUCLEAR POWER DIVl$10N 74-1152414-00 TECNNICAL BOCUMENT This relaxation is allowed to facilitate an emergency cooldown should it be required. However, two important points should be considered:
- a. Anytime tube-to-shell delta T exceeds 100F for 177 FA plants or 160F for 205 FA plants, a post-transient stress evaluation will be required.
- b. Higher tube-to-shell delta Ts will increase the tensile stresses on the tubes and may lead to higher leak flows. Indications of this occurring have been observed during actual tube leak transients.
Therefore, some judgment is required befora a decision is made to increase tube-to-shell delta T. Normally, it is recommended that tube-to-shell delta T be kept much loiver than the normal cooldown delta T limit if at all pos-
) sible. However, there may be cases where an increase in delta T is necessary to accommodate an expeditious cooldown which may be accomplished with little or no risk (e.g., decision has already been made to isolate the affected SG and allow it to fill, thus increases in leak flow rate may not significantly impact the transient).
3.3.1.3 Fuel Pin In ComDression Limits The fuel pin in compression limit is designed to prevent zirc hydriding of the fuel cladding in the radial direc-tion which can weaken the cladding. The impact of violating this limit during a cooldown is not well known, but it is possible that a single violation could lead to clad failure during the subsequent heatup. For this reason, the limit curve contains considerable conser-vatism. It is also believed that some hydriding formed during the single violation would, in a sense, be annealed during a subsequent normal heatup and pressurization DATE: 7-23-85 PAGE III-E-23
CWNP-20007 3 (9-84)
B A BCOCK & wlLCOR NUCLEAR POWER DIVISION NUM8tt TECHNICAL DOCUMENT 74-1152414-oo (i.e., little or no residual effects) . However, a rapid heatup/ pressurization following violation of the limit could induce cladding failure.
In any case, no clad failure is expected during the cooldown and depressurization in which the limit is violated. For this reason, violation of the curve during the emergency cooldown is allowed.
The impact of observing this limit during SGTR transients is that RCS pressure must be maintained much greater than the SCM curve down to a T f aPProximately 415F (forced hot circulation; lower limit of 380F in a natural circula-tion). Below approximately 425F cladding temperature zirc hydriding is not a concern. The T hot temperature limits allow for differences between coolant and cladding temperatures during forced or natural circulation.
Maintaining a higher pressure results in a larger differ-ential pressure between the RCS and the SG, thus in-creasing tube leak flow. If this increased pressure and leak flow cannot be accommodated, then the fuel pin in compression limit may be violated; otherwise the limit should be observed. Actual limits are plant specific.
Two conditions that can develop during SGTRs definitely dictate violation of the fuel pin in compression limit:
- a. If the affected SG has been isolated and will be allowed to fill, maintain RCS pressure lower than the lowest MSSV setpoint. If the MSSV were to lift, liquid RC would be discharged directly to the atmos-phere with a much higher probability of MSSV failure to rescat. MSSV failure to reseat while discharging liquid has occurred.
DATE: 7-2J-85 PAGE III-E-24
CWNP 20007-3 (9 84)
SABCOCK & WILCOX NWHR NUCLEAR P0wtR DivlSION 74-1152414-00 g TECHICAL DOCUMEllT C b. If the pressurized thermal shock (PTS) limit is in
'effect (refer to Chapter IV.G), it supersedes the fuel pin in compression limit (the two limits are mutually exclusive).
Any violation of the fuel pin in compression limit will require a post transient evaluation to determine the extent of zirc hydriding, if any, and any subsequent corrective actions.
3.3.1.4 Summary of Limits Durina Cooldown The following limits should be observed, if at all possible:
- a. Above 500F, the cooldown rate limit is 240F/hr (4F/ min) for 177 FA plants and 268F/hr (.83F/ min) for 205 FA plants,
- b. Below 500F, the cooldown rate limit is 100F/hr (1. 67F/ min) for 177 FA plants and 50F/hr (4.47F/ min)
, for 205 FA plants.
- c. MSSVs should not be allowed to lift, especially if the affected SG is full (requires RCS pressure below MSSV set pressure).
- d. PTS limit, if applicable, has priority over conflict-i ing curves (e.g., fuel pin in compression).
- e. Site emergency radiation limits should not be exceed- l ed.
i f. Tube-to-shell delta T should not exceed 150F for 177 FA plants or 160F for 205 FA plants, with the' require-ment that any increase above 100F for 177 FA plants or i 160F for 205 FA plants will require a post-transient l evaluation.
! g. Fuel pin in compression limit should be observed, with the requirement any violation wjl1 require a post-transient evaluation.
\
DATE: 7-23-85 PAGE III-E-25 l
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BABCOCK & WILCOE NUCLEAR POWER DivtS40N NUM8tt 74-11s2414-oo TECHNICAL DOCUMENT 3.3.2 Considerations for Ootional Isolation of the Affected SG Large or multiple SGTRs may dictate that continued steaming of the affected, or most affected, SG is not feasible. These conditions are covered by use of the TRACC as discussed in section 3.4.
However, smaller tube leaks that do not present immediate inventory control or radiation release problems may be handled by one of two methods. The affected SG may be isolated and the cooldown continued on the remaining SG or the cooldown may continue on both SGs within the TRACC limitations.
The decision on which method to use is largely situation dependent and therefore these guidelines provide relative advantages and disadvantages of each method. These relative merits can then be considered by the user in determining the appropriate method.
These same relative merits should also be considered before returning a previously isolated SG to service.
It should be noted that each discussion on advantages / dis-advantages is with respect to the other option in that block only. Thus, all the advantages and disadvantages of a given path need to be combined for consideration against an optional path.
i 3.3.2.1 SG Isolation This option involves termination of feeding and steaming i
of the affected, or most affected, SG. The SG may be left
( isolated or fed and/or steamed periodically as discussed in section 3.3.3.
1 DATE: 7-23-85 PAGE III-E-26
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- arme.2 coo 7 3 (s se) t
! e4ecoca a witcon i muctEAa rowen omseon """*"
TECNNICAL 80COMENT ,
i Advantaaes
- 1. Termination of steaming will terminate radiation ,
l releases if only one SG has a tube leak. If.both SGs f l have tube leaks, the radiation release rate will be !
significantly reduced by termination of steaming from !
l l the most affected SG.
l 2. Early isolation of the SG will minimize contamination ,
) of the secondary plant.
i 3. Refer to related advantages in section 3.3.3.2 with ,
I respect to filling the SG.
i l Disadvantages f
) 1. The isolated SG becomes a heat source and cooldown on f I the remaining SG takes longer. [
j 2. If in natural circulation, the loop with the isolated (
l SG becomes stagnant. The loop may void, hindering RC l
) pressure control, and thus further delaying the l l cooldown. f
- 3. Refer to related disadvantages in section 3.3.3.1 f
j regarding periodic steaming of an isolated SG and in l section 3.3.3.2 with respect to filling the SG.
t i j 3.3.2.2 continued Steamina l This option involves continued steaming of both SGs all l the way down to DHRS cut-in conditions unless a TRACC (
limit is reached.
Advantages j 1. Two loop cooldown allows the fastest possible cool- l l down. It is desirable, with a SGTR, to minimize the I
time that the RCS is hot and pressurized.
- 2. Two loop cooldown minimizes potential problems with tube-to-shell differential temperatures and idle RC
! loop voids.
r l
I i DATE: 7-23-85 PAGE III-E-27 !
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BINP-20007 3 (9 84 SABCOCK & WILCOX UI NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT
- 3. Two loop cooldown is a " normal" plant configuration that is familiar to the operator.
- 5. Continued steaming of both SGs will avoid the poten-tial problems noted under Disadvantages in Section 3.3.3.2.
Disadvantaces ,
- 1. Although the TRACC limits ensure unacceptable radia-tion releases do not occur, this method results in higher integrated releases than if the SG were isolated and allowed to fill, if only one SG has a tube leak. If both SGs have tube leaks, this method may result in higher integrated releases depending on relative leak sizes.
- 2. It is possible to pursuo this path.and still end with the SG isolated and full (2.4.3) with the net result of higher integrated releases and greater depletion of the BWST (although probably at a lower RC pressure and temperature).
3.3.3 Control of the Isolated SG As stated in 2.3.3, two options exist for control of the isolated SG: periodic feeding, steaming and/or draining or filling of the SG. Considerations for use of drains are covered in 3.5 and 3.6.3.
3.3.3.1 Periodic Feedinc and/or Steamina This option involves temporary unisolation of the SG and feeding and/or steaming as required to accomplish a given objective. The SG is then isolated again until further needed.
DATE: 7-23-8F* PAGE III-E-28 l _ _ _
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SAleCOCK & Wl(COM I * * ' " '
NUCLUJt POWER OlVISION 74-115'414-
! TECNNICAL OSCENENT l
! Advantages
- 1. Periodic steaming may prevent the SG level from
]
j exceeding a predetermined high level. This is useful if subsequent steaming is desired (to prevent carry-over) or if RC pressure is above the MSSV setpoint (to f
i, prevent lifting of the safeties on a solid SG) . A l suggested basis for a high level limit is provided in 3.6.1.1.b.
l 2. Periodic feeding may be desired to maintain a minimum level in the SG to promote shall cooling (should only l
be necessary for small tube leaks). ,
f
! 3. Periodic steaming (and feeding as necessary) will
)1 promote idle RC loop flow and cooling during natural
) i circulation.
NOTE: Periodic steaming should not be attempted if the level in the isolated SG, accounting for i instrument errors, is above the elevation j corresponding tot a) the bottom of the steam nozzles for 177 FA plants. Imakage by the level tap in the steam annulus can begin to fill the steam i
{ lines before water spills over the
! shroud.
i i b) the top of the SG shroud for 205 FA ,
) 3 plants.
l
! Disadvantages f 1. Gas buildup in the SG while isolated can result in i higher release rates during periodic steaming.
Although the integrated releases may still be less than with continued steaming of the SG, the higher
) rates during release may be more limiting, depending on the radiation limits and monitoring methods being l ,
i used.
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B ABCOCK & wlLCOI NUa48 t R NUCLEAR POWER DIVISION 74-11s2414-o0 TECHNICAL DOCUMENT
- 2. Monitoring of level and pressure in the isolated SG and the actions required for periodic feeding and steaming places a greater burden on the operator.
3.3.3.2 Fillina of the Isolated SG This option involves allowing the affected SG to fill due to the tube leakage after isolation. The timing of this option should be such that the SG will not fill solid before RC pressure is below the MSSV setpoint. The SG will fill at approximately one inch per minute for every 30 gpm cf leak flow with no steaming or feeding. This option should also not be used until RC pressure and temperature are within boundaries established by the user in the analyses performed to demonstrate steam line integrity.
Advantaces
- 1. Assuming leak integrity on the secondary side, filling the SG will terminate the tube leak flow, thus limiting the demand on the BWST.
- 2. Filling the SG reduces the radiological impact of the SGTR. Contamination of the secondary plant is minimized, releases to the atmosphere from steaming are terminated (assuming no subsequent lifting of the MSSVs on the full SG) , and, if drains are not subse-quently needed, impact on the environment at the drain location is minimal.
- 3. Although the SG is not being steamed, the existence of a high level or full SG should aid control of the tube-to-shell differential temperature.
- 4. In line with item 1 above, if the leak flow is terminated, the cooldown can be slowed as necessary to aid control of SG shell, RV head and idle RC loop cooling during natural circulation, although idle RC loop cooling methods are restricted.
l DATE: 7-23-85 PAGE III-E-30
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[] TECHNICAL DOCUMENT
\v /
Disadvantaces
- 1. Although the flow chart allows for subsequent return to service by the loop back through 2.3.2, the decision to fill the SG should assume that the SG will not be available for the- remainder of the cooldown.
Once the steam lines have begun to fill, use of the drains may not ensure drainage of the steam lines, depending on plant configuration.
- 2. The isolated, full SG will become a heat source, increasing the burden on the remaining SG and slowing the cooldown. If the tube leak flow has terminated this may not be a problem. If, however, the other SG also has a tube leak or if the steam lines on the isolated.SG do not maintain leak integrity, then the integrated release during the remainder of the cooldown can become higher than if both SGs were steamed due to the longer cooldown time required.
- 3. Subsequent steam or FW control problems on the remaining SG may require transition to HPI cooling that would not have been necessary had the isolated SG been available for heat removal. HPI cooling with a solid SG poses special problems as discussed in Section 3.7.
- 4. In line with item 2 above, options for idle RC loop cooling during natural circulation cooldowns (discus-sed in Chapter III.G) are severely limited with a fill SG. Forced circulation cooldowns may be slowed due to the tensile tube-to-shell delta T limit.
3.4 Bases for Alternate control Methods These guidelines have been structured to allow steaming of the SG(s) with a SGTR as far into the cooldown as possible
(
v j to provide better overall plant control. The basic DATE: 7-23-85 PAGE III'E"31
BWNP-20007 3 (9 84)
BABCOCK & WILCOX II NUCLEAR POWER DIVISION 74-11s2414-o0 TECHNICAL DOCUMENT philosophy of these guidelines is to maintain as normal a plant configuration for as long as possible in order to expedite the cooldown to DHRS operation and minimize potential for development of complications. However, there are limitations on continued steaming of a SG with a SGTR. These limitations consider the real overriding concerns of SGTR transients that dictate the use of alternate methods.
These limits have been designated SGTR Alternate Control Criteria (TRACC) and are as follows:
- a. Radiation release rates approaching predetermined limits.
- b. BWST level approaching a predetermined low limit.
- c. SG filling due to tube leakage despite steaming to achieve the maximum allowable cooldown rate.
The bases for each of these criteria are discussed separately in 3.4.1, 3.4.2, and 3.4.3.
3.4.1 Radiation Release As stated in 1.3.2, one of the major objectives in mitigating SGTRs is to minimize the amount of radiation released to the environment. This guideline achieves this j objective as follows:
- a. Performing a plant runback rather than tripping the reactor.
- b. Performing actions as necessary to minimize the
- potential for uncontrolled releases.
- c. Providing an overall limit on integrated doses at the site exclusion area boundary (EAB).
l These facets of radiation control are discussed in the following subsections.
l O
l DATE: 7-23-85 PAGE III-E-32
CWNP-20007 3 (9 84)
S ASCOCE & witCCR NUMSER NUCLEAR POWER DIVISION 74-1152414-00 Os 3.4.1.1 TECHNICAL DOCUMENT Plant Runback This guideline strongly recommends performing a controlled plant runback as opposed to a reactor trip for two radiological considerations:
- a. A reactor trip will result in steam release directly to the atmosphere through the MSSVs and ADVs. A plant i
runback will direct the radioactive steam to the condenser. Although all of the noble gases will still
! be released, virtually all of the iodine that leaked into the SG will be contained in the secondary system. Since studies show that the thyroid doses,
! which are due to iodine, are limiting for SGTRs, this will greatly reduce the radiological consequences of
$ the event. In addition, the noble gases will be released through the stack which will provide better dispersion.
t
.d j Even though the runback will take longer to reach the j same post-trip conditions, calculations show that
! the integrated thyroid done during the runback will be on the order of 200 times less than if the plant were tripped from full power. Whole body doses will increase slightly due to the runback, but even for
.05% failed fuel the integrated whole body dose should still be less than 3 mrem. In addition, these calculations show that allowing the steam safeties to lift for only 30 seconds due to a trip from full power j will contribute approximately 75% of the integrated thyroid dose for the entire transient assuming a
{ subsequent 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown using the condenser and
! the SG with the SGTR (66% for a lift duration of only
! 20 seconds). Thus a runback will significantly reduce j the radiological consequences.
i DATE: 7-23-85 PAGE III-E-33
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TECHNICAL DOCUMENT 74-1152414-oo
- b. There is a small but real possibility that a MSSV can fail to reseat after lifting. A failure to resent on a SG with a SGTR will result in uncontrolled releases since no practical method exists to terminate the leak flow or steam flow. Therefore, the prudent approach is to prevent challenges to the MSSVs during a SGTR event. As can be seen from the calculations described in a. above, a failure of a MSSV to rescat for even a short duration will significantly increase the integrated thyroid dose.
There may be instances when a reactor trip is unavoidable, but following the actions outlined in section 3.2 should prevent MSSV lifts for most SGTR transients.
3.4.1.2 Uncontrolled Releases In addition to the plant runback before reactor shutdown, there are two other specific conditions when the guideline recommends actions to minimize the potential for uncon-trolled releases through the MSSVs on the affected SG.
Again, the intent is to prevent the initial lift of the MSSVs even if use of the ADVs is required, since the ADVs can be isolated should they fail. The two conditions are:
A. The SG should not be allowed to overfill into the steam lines while RC pressure is greater than the lowest MSSV set pressure. This condition would cause the MSSV to open as the steam line fills with water and the secondary and primary pressures equal-ize. The subsequent liquid relief through the MSSV would increase the probability of a failure to rescat.
B. If a transition to HPI cooling is required, it may not be possible to maintain RC pressure below the lowest DATE: 7-23-85 PAGE. III-E-34
l EWNP-20007 3 (9 84)
S ASCOCK & WitCOX NUMSER NUCLEAR POWER OtVISION 74-11s2414-0o
/9 TECHNICAL DOCUMENT O set MSSV lift pressure following normal HPI cooling procedures. If the affected SG is also full or filling, then additional actions are necessary to prevent lifting the MSSVs. This is discussed in more detail in section 3.7.
3.4.1.3 Intearated Dose Limits SGTR events impose a unique problem on the guideline and procedure writers as well as the plant staff. There is an apparent conflict of interests between steaming the affect 6d SG and minimizing the amount of radiation released to the environment.
The ultimate goal of a SGTR procedure is to achieve cold shutdown on the decay heat removal system and termination of the tube leak flow. The most expedient and control-O lable method to achieve this state is by cooldown using both steam generators. However, this is in apparent conflict with the goal to minimize the release of radia-tion. One method to minimize the release of radiation is to isolate the SG as soon as conditions permit. This action (isolation of the SG), at best, places the plant ~in an abnormal condition and lengthens the time required to complete the cooldown. More importantly, this guideline is designed to cover SGTRs in both SGs. Isolation of both SGs requires a transition to HPI cooling which, although adequate, is certainly a less desirable mode of core cooling.
The philosophy of this guideline, in' light of this apparent conflict, is to attempt to strike a reasonable balance between controlled cooldown of the plant and what can be considered acceptable levels of radiation release.
Previous guidance on acceptable radiation release for t ,.
DATE: 7-23-85 PAGE III-E-35
OWNP.20007 3 (9 84) I SASCOCK & witCOE NUCLEAR POWER DIVI $lON 74-1152414-00 TECHNICAL DOCUMENT SGTRs has been limited for FSAR requirements for the design basis event. At present, these limits are inter-preted to be 10% of 10CFR100 doses during the first two hours at the EAB and for 30 days at the low population zone (LPZ).
However, the design basis SGTR is a double-ended rupture of a single tube while this guideline covers multiple ruptures in both SGs. In addition, plant cooldown cannot be accomplished in two hours. Thus the question becomes what is an acceptable dose for an event that has larger total tube leak flows and cooldown times than covered by the design basis case.
To answer this question, integrated dose calculations were performed for a spectrum of tubo leak flows and RC activity levels. These calculations were performed using both conservative assumptions as in the FSAR and realistic assumptions. The conclusion reached from this effort is as follows:
The overall integrated dose limit for all tube leak rates and cooldown times should be no more than 10% of 10CFR100 doses. In other words, the dose limits, regardless of the number of ruptured tubes should be no more than 30R to the thyroid and 2.5R to the whole body at the EAB for the duration of the cooldown. The user should establish monitoring methods and appropriate limits to ensure these integrated doses limits are not exceeded.
i 1
This limit is consistent with current design basis requirements for SGTRs that approximate the design )
l DATE: 7-23-85 PAGE III-E-36
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\
basis case and is considered to be conservative for SGTRs beyond the design basis. For tube leaks smaller than the design basis case, realistic calculations indicate that expected doses will be much lower than these limits even when the entire cooldown is performed steaming directly to the atmosphere.
3.4.2 BWST Level The RC lost through the SGTR to a steaming SG is not recoverable for reinjection to the RCS. This fact will not normally present a problem during single SGTRs with an expeditious cooldown to DHRS conditions since the integrated leakage is much less than the available BWST inventory.
However, very large leak rates or delays in establishing g DHRS operation may lead to depletion of the BWST inven-tory. Assuming a backup source for the BWST is not available, then BWST depletion must be terminated at some pre-determined low level.
It is suggested that the low level limit be based on sufficient remaining BWST inventory to:
- a. Fill both SGs out to the first steam line isolation valve,
- b. Fill the RB sump (via HPI cooling through the PORV) to the minimum level required for sump recirculation, and
- c. Account for all applicable instrument errors.
The necessity to isolate and fill both SGs is unlikely, but it is possible and therefore should be accounted for.
In addition, plant procedures may call for filling, or aiding the fill, of a SG by the use of FW (refer to i
O 3.6) but the BWST low level TRACC limit should be based on
\
U, i
t DATE: 7-23-85 PAGE III-E-37
BONP-20007-3 (9-84)
B AB COCK & witCOX NUCLEAR POWER DIVISION NUMSER v4-lis2414-oo TECHNICAL DOCUMENT the assumption that FW is not available and the entire fill must be accomplished using BWST inventory.
3.4.3 SG Overfill This criterion is intended to prevent carry-over from a steaming SG. Carry-over in a steaming SG could damage the open steam valves (TBVs or ADVs) and result in steam leaks. In addition, it is preferable to prevent filling a SG, even if not steaming, since once the SG is filled it may no longer be available should it subsequently be needed (e.g., use of SG drains to restore a steaming lcvel in the SG may not effectively drain the steam lines) . As the cooldown progresses, it may be necessary to reduce or terminate FW flow to, and increase steaming from, the affected SG to limit the SG level increase. Reduce steaming and feeding of the good SG to maintain the cooldown rate within limits.
However, for relatively large SGTRs or SGTRs in both SGs, these actions may not prevent overfilling of one or both SGs. Therefore, a high SG 1evel limit should be deter-mined that will prevent carry-over into the steam lines, accounting for instrument errors.
3.4.4 ADolication of TRACC Note that all three TRACC are somewhat rate dependent.
Allow for the rate at which the limit is being approached and the estimated time required to implement an alternate control method in the preparation of plant procedures. ,
Note too that, in actual practice, either none of these l limits will be reached (for a relatively small SGTR and expeditious cooldown) or all may be reached (for a DATE: 7-23-85 PAGE III-E-38
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- V TECHNICAL DOCUMENT i
relatively large SGTR and/or significant delays in the
- cooldown).
For the latter case, any one of the three TRACC may be ,
limiting, depending on the actual plant conditions, i l
j therefore all three (or their equivalent) should be j included in the procedures. .
l i
When a limit is being approached, an alternate course of action is to use SG drains if available (refer to 3.5) .
~
i This allows for reduction of the level increase rate I and/or the steaming rate of the affected SG(s) so that the j use of the SGs for cooldown may continue without vio- '
i lating TRACC. Carefully monitor the approach to TRACC j during and after the transition to SG draining to assess i the effectiveness of the drains.
If the drains are. unavailable, are not used, or are otherwise ineffective, isolate the affected SG and allow f it to fill (3.6). If SGTRs exist in both SGs, isolate ;
l the SG with the larger leak rate and monitor the approach !
to TRACC during continued steaming of the other SG. If {
l TRACC is still expected to be violated due to continued l steaming of the remaining SG (both have tube leaks), i isolate the remaining SG and initiate HPI cooling (refer
- to 3.7).
l i ;
- 3.5 Considerations for Use of SG Drains i i The intent in using SG drains is either to prevent or i j delay the necessity of isolating the affected SG(s) by reducing the required steaming rate or to prevent filling of an' isolated SG. It is definitely desirable to Aelay ,
fill of the SG until RCS pressure is below the MSSV f i s s
- setpoint. [
I i l
DATE: 7-23-85 PAGE III-E-39 i i
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S A B COCK & WILCOX NUCLEAR POWER OtytSION 74-lls2414-oo TECHNICAL DOCUMENT This discussion assumes that a readily accessible high energy SG drain system is available during a SGTR. When developing procedures consider the plant specific SG drain system design limitations, which will not be discussed here.
There are, however, generic considerations for limiting the use of SG drains depending on the limiting TRACC. If approaching the low BWST level TRACC limit, for example, use of the drains will not prevent reaching the limit.
But, if the drain flow path is such that the drained coolant is available for reinjection by HPI (e.g., drains to the reactor building sump), then use of the drains may negate the low BWST level TRACC limit. In this case, continue steaming and draining of the affected SG(s) until normal low level limits are reached in the BWST then switch over to sump suction and HPI piggyback operation (unless, of course, another TRACC is reached). This method is considered a,cceptable since it involves a large tube leak flow rate and therefore virtually all of the fluid drained from the SG(s) will be RC. However, provisions should exist to cample the drained fluid and add boron as necessary.
If radiation release rate is the limiting TACC, use of the SG drains reduces the steaming rate required on the j affected SG(s) and thus may reduce the radiation release rate depending on drain location. However, in this case j the drains should not be used if the site boundary doce rates would increase due to the storage location for the l
contaminated fluid.
1 1
O DATE: 7-23-85 PAGE III-E-40
B%NP 20007 3 (9-84)
BARCOCK & wlLCOX NUMSIE NUCLEAR POWER DIVISION 74-1152414-oo O TECHNICAL DOCUMENT O If the limiting TRACC is high SG level, use of the drains should permit continued steaming of the SG(s) until another TRACC become limiting.
3.6 SG Isolation If the SG drains are unavailable, are not used, or are ,
otherwise incapable of preventing violation of a TRACC, isolate the affected, or most affected, SG (or second SG, in block 2.4.5) by closing all steam, feed, and drain 4
lines to that SG. If the SG is isolated due to a TRACC limit, the tube leak is probably very large and therefore the SG will probably fill.
Depending on' plant conditions, there may be situations
- where it is desirable to temporarily unisolate a steam,
- s feed, or drain line as follows
3.6.1 Steam Lines
~
- 1. Chapter III.G describes periodic feeding and steaming of an otherwise isolated SG to enhance idle loop cooling during natural circulation. These methods may also be used for SGTRs provided:
- a. additional steaming, at least periodically, will not result in unacceptable radiation releases
- b. the level in the isolated SG is low enough to allow steaming without inducing carry-over. A suggested level limit for 177 FA plants is that equivalent to the elevation of the bottom of the steam line, accounting for full range level instrument errors. Water will enter the steam annulus via the penetration for the operate range upper level sensing tap, but should not induce carry-over at low steam flow rates if the level does not reach the steam line.
i 7-23-85 PAGE III-E-41 DATE:
l
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BWNP 20007 3 (9 84)
S AS COCK & WILCOX NUMBER NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT These two restrictions should apply if the sole motivation for steaming is to enhance idle RC loop cooling during natural circulation.
- 2. It may be desirable to open TBVs or ADVs to prevent lifting of MSSVs even if excessive radiation releases and/or carry-over will occur. This is especially true if the MSSVs would pass liquid, since failure to reseat becomes more probable. A TBV or ADV could also fail to reseat under this condition but should be used if remotely operated block valves are available to isolate the stuck open valve.
- 3. Use of the steam lines might be considered to convert the SG to a water / water heat exchanger. This method of cooling involves filling the SG, steam line, and turbine bypass system and recirculating FW to provide cooling. This method should be considered if other cooling methods (steaming, HPI or DHRS) are unavail-able or, due to extenuating circumstances, less desirable. This method should only be considered if analysis indicates that the plant's secondary system piping and components can withstand the stresses and maintain structural integrity. If this method is used, flow control should be maintained using FW control valves to prevent / minimize cycling of TBVs.
3.6.2 Feed Lines
- 1. Unisolation of the feed lines may be desirable to enhance idle RC loop cooling as discussed in Item 1 of 3.6.1 and in Chapter III.G.
- 2. Periodic use of AFW spray may be beneficial in reducing steam pressure in an isolated SG to prevent lifting of the MSSVs if the level in the isolated SG is below the AFW nozzles.
O r
DATE: 7-23-85 PAGE III-E-42
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BABCOCK & WILCOX i NUm6ER l NUCLEAR POWER DIVISION 74-11s2414-oo j p TECHNICAL DOCUMENT I
- 3. If the isolated SG is expected to fill due to the tube !
leakage, it may be desirable to augment the filling
, with FW. Use of FW for filling will:
- a. provide additional RCS cooling during the fill
- b. reduce the depletion of the BWST inventory
- c. dilute the RC in the steam lines to reduce radiation levels.
However, filling of the SG should not be expedited in this manner while RCS pressure is at or above the MSSV setpoint.
- 4. Use of FW might be considered in another unusual case of converting the SG to a water / Water heat exchanger.
In this case, the secondary side flow path would be injection by FW and removal by the SG drains. As in Item 3 of 3.6.1, this method should only be considered if other cooling methods are unavailable and if plant conditions are within design capability of the SG drains. This method requires careful control of secondary system pressure since it is essentially solid. In addition, this method should only be used if:
- a. the fluid drained from the SG will aqt be used for reinjection to the RCS (due to the lower boron concentration),
- c. precautions are taken to guard against potential backflow of FW into the RCS through the SGTR (e.g., increased sample frequency, higher RC pressure). '
i 3.6.3 SG Drcins
- 1. Unicolation of SG drains may be desired to prevent ,
fill of an isolated SG if:
DATE: 7-23-85 PAGE III-E-43 l
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B ABCOCK L witCOX NUCLEAR PObER DivlSION 74-11s2414-oo TECHNICAL DOCUMENT
- b. subsequent steaming of the isolated SG is antici-pated.
- 2. Use of the SG drains may be desired to aid depressuri-zation of the isolated SG and the RCS during HPI cooling (see 3.7).
- 3. Use of the SG drains may be desired to establish a water / water heat exchanges as described in Item 4 under 3.6.2.
If the SG is isolated because of approach to a TRACC, any subsequent actions to unisolate a steam, feed, or drain line should consider the potential impact on the TRACC.
There may be instances where violation of a TRACC, at least temporarily, is warranted to prevent potentially worse situations. For example, as stated in Item 2 under 3.6.1, it may be preferable to open an ADV on a full, isolated SG even though excessive radiation release could occur to prevent the potential of uncontrolled release through a stuck-open MSSV (assuming remote-operated block valves are available to isolate a stuck-open atmospheric dump valve).
i 3.7 SGTR considerations Durinc HPI Cooline Methods to initiate and control HPI cooling are provided in Chapter III.G. However, HPI cooling concurrent with a SGTR may pose special problems which are discussed here.
Criteria for control of HPI (MU) (Chapter IV.B) dictate fill HPI flow from two HPI (MU) pumps when adequate SCM does not exist. When adequate SCM is established, then DATE: 7-23-85 PAGE III-E-44
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SABCOCK & witC0K !
NUCLEAR power DIVI $lON NUsttS !
v4-11s2414-oo i TECHNICAL DOCUMENT 1
l l HPI (MU) flow can be throttled to limit RCS pressure and cooldown rate while maintaining SCM. -
4, However, following the HPI (NU) criteria may result in RCS r pressure greater than MSSV setpoint. If a SG has been f ,
isolated and allowed to fill, its pressure will follow RCS pressure, thus presenting the risk of lifting an MSSV and passing liquid.
If adequate SCM does not exist, the criteria for full HPI (MU) flow from two HPI (MU) pumps takes precedence; I
therefore, throttling of HPI (MU) flow to limit or reduce
) RCS pressure cannot be performed until adequate SCM exists. However, if this requirement could result in l
i lif ting of an MSSV on a solid SG, the following actions may be taken in an attempt to limit RCS pressure increase:
{
l a. If available, reestablish letdown flow.
- b. If available, open HPVs on the hot legs, pressurizer,
} and RV head. This will augment the relief through the I PORV and may provide enough additional flow to limit or reduce RCS pressure below the MSSV setpoint.
- c. If available, open the drains on the isolated SG.
This action essentially performs the same function as l
l use of the HPVs.
- d. If neither of the above are available or sufficiently i effective, open t'.te TBVs and/or ADVs on the isolated l SG. Again, this. essentially. performs the same
, function of increasing relief capacity from the RCS.
t 3.8 Innact of Unisolable Steam Leaks j Unisolable steam. leaks concurrent with SGTPs limit
! the flexibility the operator otherwise has in controlling ,
cooldown rates, SG inventories, and radiation release
! rates. The extent of the impact is dependent on the {
DATE: 7-23-85 PAGE III-E-45
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02NP 20007 3 (9 84)
SASCOCK & wiLCOX NUMSit NUCLEAR POWER OlVISION 74-lls2414-oo TECHNICAL. DOCUllENT number, size, and location of the steam leaks. Chapter III.D discusses methods to control cooldown rates and SG inventories (without SGTRs) with unisolable steam leaks.
This section discusses the impact of concurrent SGTRs on control of SG inventories and radiation release rates.
The basic philosophy presented in Chapter III.D for mitigation of unisolable steam leaks is to attempt continued cooling on both SGs and to isolate a SG only if the cooldown cannot be controlled or if the steam leak is inside the RB. The same basic philosophy should be followed in mitigating unisolable steam leaks concurrent with SGTRs, which may impact the decision to isolate a SG. However, there is a major difference. In Chapter III.D, a steam leak can be terminated, if necessary by isolating FW to the SG and allowing it to boil day. If the SG also has a SGTR, however, the steam leak will continue due to boiloff of the tube leakage. Since continued steaming through the leak is unavoidable, intentionally steam the SG for the following cases:
- a. If the steam leak is outside the RB steam the SG to the condenser, if available. This will reduce the steamflow through the leak to atmosphere and thus limit the radiation release.
s
- b. If the steam leak is inside the RB, steam the SG to the condenser, if available, only while steaming the SG is necessary for other considerations. When steaming of the SG is no longer required, the SG should be isolated (all steam and feed lir.es closed) and allowed to steam to the RB through the leak.
- c. As an option, SG drains may be used, if available, to reduce the steaming rate through the unisolable steam leak. This will not prevent steam release, but may decrease the rate of release if accumulated tube 7-23-85 PAGE III-E-46 DATE:
l
CWNP-20007 3 (9-84)
S ABCOCK & w LCOM NUMBER NUCLEAR POWER DivlSION 74-1152414-oo
' TECHNICAL DOCUMENT leakage can be drained before it boils off. This will only be effective for small steam leaks and/or large
, SGTRs; for example, a steam leak size the equivalent of a stuck opon MSSV with a single SGTR will result in virtually no accumulated leakage; therefore, the drains would be ineffective.
i i
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8%NP 20007 3 (9-84) ;
B ASCOCK & WILCOX ,
NUCLEAR POWER DIVISION NUm8tt .
74-1152414-oo OTECHNICAL DOCUMENT l U l Chanter III.F ;
Inadeauate Core Coolina f 1.o INTRODUCTION 7 This chapter provides the Technical Bases for the guidelines to mitigate an inadequate core cooling condition. An I inadequate core cooling condition exists whenever the incore !
thermocouples indicate a superheated-temperature. [
1.1 Concerns and Obiectives Durina ICC Conditions '
i 1.1.1 Concerns Inadequate core cooling is not expected, as long as the guidelines are followed. However, any transient can !
- progress into ICC conditions, provided enough equipment ;
failures occur. As soon as the RCS is superheated, adequate i
, core cooling can no longer be assured. Consequently, ;
actions must be taken to restore the RCS to at least I
- saturated conditions as quickly as possible. The specific I I
concerns during ICC are as follows:
i A. Possible fuel damage.
! B. Production of non-condensable gases.
- i C. Degraded RB environment. f D. Possible radiation releases to the atmosphere. ( ,
E. Equipment damage. ,
F. RB Integrity. [
G. Clad-water reaction becoming a dominant heat source. !
- i 1.1.2 Obiectives [
The objectives to be considered during the treatment of ICC i
- conditions are as follows. ;
A. Restore adeauate core coolina - The primary concern is j to restore the RCS to at least saturated conditions. j The core is adequately cooled while the RCS is !
saturated.
DATE: 7-23-85 PAGE III.F-1 f 1
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BABCOCK & witCOI NUCLEAR POWER DivlSION NUMSit 74-1152414-oo TECHNICAL DOCUMENT B. Estimate the severity of the ICC condition -
The actions taken to mitigate ICC depend upon the severity of the ICC condition. The more severe the ICC is, the more drastic the actions. If clad damage and/or fuel melting is imminent, then running an RCP to destruction or operating a SG such that its reuse may be unlikely is justified. However, if the ICC condition is not as severe, then less drastic measures should be taken which will preserve the future integrity of the RCS and associated equipment.
C. Eliminate non-condensible cas in the RCS - If the ICC condition is severe enough, then non-condensible gases will be produced. If primary to secondary heat transfer is to be restored, it is necessary to eliminate the non-condensable gases from the RCS to allow for RC natural circulation or boiler condenser flow.
1.2 Causes The events which will cause ICC have a low probability of occurrence. Some examples of where ICC conditions could develop, provided the event lasts for a long enough time are:
A. SBLOCA with a total failure of the HPI system.
B. Total loss of feedwater (both MFW and AFW) with a concurrent total failure of the HPI system.
C. A total loss of power including all diesel generators with a failure of the steam-driven AFW pump to run (even if the steam driven AFW pump runs, an extended total loss of power may eventually lead to degraded RCP seal performance. Also, without power to the HPI pumps, ICC conditions would eventually occur).
D. During a specific size SBIOCA tripping the RCPs at a time when the RC void fraction is 70% or greater.
[
l
( DATE: 7-23-8s PAGE III.F-2
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.NUCLEAR POWER Divi$lON 74-1152414-oo TECHNICAL DOCUMENT l
1 This would only occur if the RCPs had not been tripped i
! upon loss of SCM.
l 2.0 DIAGNOSIS AND MITIGATION The flowchart of Figure III.F-2 should be used in conjunc-tion with the following discussion. The numbered subsec- 7
{
tions of Section 2.0 correspond to the upper numbers in the !
I blocks of Figure III.F-2.
1 l
}
2.1 Identification of ICC (Detailed discussion in Section 3.1)
! The RCS P-T relationship will indicate when ICC conditions occur (see Figure III.F-1) . As soon as the combination of
, RCS pressure and incore T/C temperature exceed the satura-tion curve then superheated conditions exist.
An instrumentation failure could inaccurately signal the i onset of ICC conditions. For example, a failed pressure ,
transmitter could cause an indication of ICC conditions l j occurring. This points out the need for the operator to I check not only alternate instrumentation channels, but also to verify the expected plant response with other !
instrumentation. The axact state of the RCS should be determined as closely as possible, so that incorrect actions l
j would not be taken. Some actions required by severe ICC l
! conditions could cause problems where none actually existed i before. ,
i j It is also possible that an instrumentation failure could I indicate that the RCS is being adequately cooled, while, in l I fact, it is in ICC. The prevention of this is the same, j that is, to verify alternate instrumentation channels, as !
l j well as cross checking with other instruments to verify the I expected plant response during such conditions. These other instruments, such as RV head level and hot leg level 1
i 7-23-85 PAGE III.F-3
! DATE:
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BA5 COCK & WitCOX NUMBER NUCLEAR POWER Olvl$10N 74-1152414-00 TECHNICAL DOCUMENT instruments if available, have plant specific characteris-tics and usage.
2.2 Injtiate HPI/LPI/CFT The first action which should be taken during the onset of ICC conditions is to ensure maximum HPI and LPI flow into the RCS. Also, the CFT block valves should be verified open at this time. Decreasing CFT levels should be verified as soon as the RCS pressure decreases below the CFT actuation point. Ensuring maximum HPI and LPI flow is especially important because most postulated scenarios for the occurrence of ICC involve a failure of HPI/LPI. If RCS pressure increases to the PORV setpoint, open the PORV and reduce RC pressure to about 100 psi above SG pressure, then reclose the PORV.
Refer to Chapter IV.B for a discussion on maximizing HPI/LPI flow.
2.3 Take Actions to Increase Primary to Secondary Heat Transfer (Detailed discussion in Section 3.2) i SG levels should be increased to the loss of SCM setpoint.
I FW flow must be assured and SG pressure lowered to achieve a secondary Tsat of about 100F lower than Tsat for existing l
RCS pressure. This temperature differential must be main-l tained as the RCS depressurizes until primary to secondary heat transfer is established.
2.4 RCS is in which Recion of ICC Ficure III.F-17 (Detailed discussion in Section 3.3)
If the RCS P-T point has returned to Region One, then the previous actions were successful and a normal cooldown can begin.
7-23-85 PAGE III.F-4 DATE:
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s TECHNICAL DOCUMENT i
I If the RCS P-T' point is still in Region Two , then the .
) operator should continue to ensure maximum HPI/LPI/CFT flow l l- and continue to induce primary to secondary heat transfer. :
s~
l f
If ICC conditions worsen into Region Three, then more !
}; drastic actions are needed to restore adequate core cool- fc ing. Refer to Section 2.5. j j If the RCS P-T point reaches Region Four, then serious ICC l conditions exist. Drastic actions are required to minimize l
) major core damage. Refer to Section 2.8. l 1 l 2.5 Take Rection Three Actions to Increase Heat Transfer From Reactor Core to Reactor Coolant (Detailed discussion in i
} Section 3.4)
I
! s The RCS P-T' relationship is - in Region Three. One RCP per
- loop should be started, if possible. However, do not r l override normal RCP starting interlocks. However, once a l RCP is s t a r t e'd , it should not be tripped for other than l i
j motor electrical faults. Refer to Chapter IV.A. j 3
1 j All HPV valves should be opened to relieve noncondensable !
i i gases from the RCS. Further instructions on the use'of HPVs j are detailed'in Chapter IV.E.
, HPVs should. remain open at least until RCS becomes subcooled ,
or LPI cooling is established. j l
4 r
!, 2.6 Take Actions to Further Increase Primary to Secondary Heat ;
! Transfer .
i' operating SGs should be depressurized as quickly as possible l j to achieve a 100F delta T between secondary Tsat and RC Tsat j for the existing RC pressure. ' Actions must continue to .
l
- j. !
- t DATE: 7-23-85' PAGE III.F-5 }
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ScND 20007 3 (9 84)
B ABCOCK & witCOM NUCLEAR POWER DIVl$10N 74-1152414-oo TECHNICAL DOCUMENT for Davis-Besse) setpoint and FW flow should be ensured.
(Refer to Chapter IV.C for SG level requiremsnts).
2.7 Primary to Secondary Heat Transfer Established?
After the previous actions of 2.5 and 2.6 have been taken, primary to secondary heat transfer should be checked. If heat transfer has not been reestablished then it will be necessary to continue with HPI cooling. If primary to secondary heat transfer has been established, then it will be possible to cool the RCS with SGs, using HPI/MU to replace RCS inventory.
If ICC conditions worsen, such that the RCS P-T point enters Region Four, then further, more drastic actions are required. Refer to Section 2.8.
2.8 Take Rection Four Actions to Maximize Heat Transfer From Reactor Core to Reactor Coolant (Detailed discussion in Section 3.5)
The RCS P-T relationship is in Region Four. Severe ICC conditions exist. All RCPs should be started if available.
If necessary, all starting interlocks should be defeated.
The overload trip circuit should not be defeated.
The RCS should be depressurized until LPI restores core cooling. All HPVs, PORV, and PORV block valves, etc.,
should be opened.
2.9 Take Actions to Maximize Primary to Secondary Heat Transfer (Detailed discussion in Section 3.5)
Because of the severe ICC conditions which are occurring, l
the operating SG(s) should be depressurized as quickly and l
as far as possible.
DATE: 7-23-85 PAGE III.F-6
BWNP-200013 (9 84)
S ABCOCK & wlLCOM NU"8EE NUCLEAR POWER OtVISION ,
( ) TECHNICAL DOCUMENT
\d i 2.10 Incore T/C Return to Saturation? <
As long as the RCS P-T relationship remains in ICC, the previous actions of 2.8 and 2.9 should be continued. l However, as soon as the incore T/Cs return to saturation it l will be possible to continue cooldown as normally as possible. i F
i 2.11 Continue Lona Term Cooling The RCS must remain depressurized to ensure maximum LPI j flow. Transferring LPI suction to the RB sump when the BWST l t
reaches the sump switchover point is necessary,if not j already done. It may also be possible to use the decay heat [
t removal system, depending on whether or not subcooling is j reestablished. LPI operation is detailed in Chapter IV.B.
[
Necessary actions to control RB containment systems are i
) detailed in Chapter IV.F. Because of the nature of an ICC event the radiation levels in the RC can be exceptionally ;
high. Consequently, appropriate radiation precautions should be invoked.
i 3.0 TECHNICAL B)L_ES f The flowchart of Figure III.F-2 should be used in conjunc- l tion with the following discussion. The numbered subsec- !
tions of section 3.0 correspond to the bottom numbers of the !
appropriate blocks on Figure III.F-2. !
3.1 Identification of ICC Conditions i The RCS P-T relationship will clearly indicate when ICC !
conditions occur. As soon as the RCS P-T point reaches the j adequate SCM cumre the incore T/C readings should be used to {
determine the actual conditions of the reactor core. As l
soon as the RC pressure and incore T/C temperature combina-l tion exceeds the saturation curve, then superheated condi-l DATE: 7-23-8s PAGE III.F-7 f
BONP-2OOO7 3 (9 80)
SABCoCK & wlLCOK NUmsta NUCLEAR POWER OlVISION 74-11s2414- w TECHNICAL DOCUMENT tions are possible in the RCS. The operator must now take action as though the RCS were superheated in an attempt to restore the RCS to aaturated conditions. However, due to instrumentation errors, it is possible that although the RCS P/T combination is slightly to the right of the saturation curve, the RCS is indeed only saturated. This is not a problem with the recommended actions, since most of the recommended actions are to continue the same actions that are taken in the event of a loss of adequate SCM. In addition, the other actions required have been reviewed to assure that these actions do not cause further problems if the RCS in only saturated. The fact that the RCS is saturated rather than superheated can be verified by noting that the incore T/Cs temperature moves parallel to the saturation curve. If ICC conditions actually exist the RC P-T would continue to trend into the ICC region away from the saturation curve. See Figure III.F-1.
The response time of the temperature detectors must be considered when determining if the RC has become super-heated. For a rapid decrease in RC temperature, the hot and cold leg temperature detectors will indicate a higher temperature than actually exists, due to the response time of the detectors. For example, a large IOCA will cause the RC pressure to rapidly decrease to the saturation pressure. The RC hot leg temperature will also decrease, however, the indicated hot leg temperature vill change slower than the actual temperature causing a superheat temperature indication while actually the RCS is only saturated. This is one of the reasons that incore T/C temperature is recommended, rather than using the hot or cold leg RTD indications. The primary reason for using the T/cs is because they respond quicker than the hot and cold leg temperature detectors and will indicate actual RCS DATE: 7-23-85 PAGE III.F-8
B7 NP 2000L3 (9-84)
S ABCOCK & WILCOX NUMttt NUCt( AR POWER OlVISION
{g I ECHNICAL T DOCUMENT 74-1152414-0o V
- conditions more accurately. Another cause for a false ICC indication is the increased RB pressure following a LOCA.
When the absolute pressure of the RC is determined from the RC gage pressure the effect of the increased RB pressure is generally not accounted for. The hot and cold leg tempera-ture detectors can become disassociated with core outlet temperature when the temperature becomes superheated because of a lack of RC loop flow. If there is any doubt of the actual RCS conditions, ICC conditions should be assumed and appropriate actions taken. As stated before in the discussion on instrumentation errors, the required ICC actions in Region Two are about the same as those required during a lack of adequate subcooling margin. It is only after the RCS has proceeded into Region Three or Four that more drastic actions must be taken.
N 3.2 Take Actions to Increase Primary to Secondary Heat Transfer At the onset of ICC conditions the operator must attempt to restore primary to secondary heat transfer, if at all ;
possible. Actions to restore primary to secondary heat transfer include raising SG levels to the maximum limit ;
allowed. To accomplish this it is necessary to ensure adequate FW flow. Also, SG pressures should be lowered in an attempt to induce heat transfer. The pressure should be lowered until primary to secondary heat transfer is l restored or secondary Tsat is about 100F lower than T sat f#
l the existing RC pressure. This differential must be ;
l maintained as the RCS depressurizes until heat transfer is restored, then SG pressure will be controlled to maintain [
I the desired cooldown rate. These actions are similar to those for a loss of SCM in case the ICC guidelines are inadvertently implemented. t O .
DATE: 7-23-85 PAGE III.F-9
BWNP 20007 3 (4 84)
B A B COCK & wlLCOK NUMBtB NUCLEAR POWER OlVISION 74-11s2414-oo TEC!!NICAL DOCUMENT 3.3 RCS is in Which Recion of ICC Ficure III.F-l?
After taking initial actions to mitigate ICC conditions, the operator must now determine how severe the ICC conditions are before taking further actions. This enables the operator to take actions which may damage plant equipment depending on the severity of the ICC conditions. If the RCS P-T point (as determined by RCS pressure and the incore T/C readings) is in Region One then the RCS has returned to a saturated or even subcooled condition because of the previous actions. At this point a cooldown can proceed with the RCS in a saturated or subcooled condition. It may be necessary, at first, to perform a saturated cooldewn with the SGs. On the other hand, it may be necessary to cooldown with HPI/LPI/CFTs alone, depending upon the condition of the RCS. Refer to Chapter III.G for details of cooldown and Chapter III.B for details of cooldown with a small LOCA.
If the RCS is still superheated and in Region Two of ICC O
Figure III.F-1, then ICC conditions still exist, but it is not serious enough to cause immediate core camage. In this case, the operator should maintain HPI/LPI flow and continue to control primary to secondary heat transfer until the RCS returns to a saturated condition. As soon as the RCS does return to a saturated condition then cooldown may proceed.
If the RCS P-T point reaches Region Three, then cladding temperature in the high power regions of the core may be 1400F or higher. Above this temperature there is a chance for rupture of the fuel rod cladding material. The clad-water reaction begins to produce hydrogen which will collect in the reactor' coolant loops and may escape to the reactor building. Region Three, then, is the the onset of very serious ICC conditions. Conditions are serious enough DATE: 7-23-85 PAGE III.F-10
._ --- - --. ... ..- - - -... - ~ _ - _. - -. - . . . . . . - . . . . _ . - . . - _ . _ .
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1 SABCOCK & wnCOR NMER l NUCLEAR POWER DivlSION 74-1152414-0o ;
L TECNNICAL DOCUMENT :
! 1
- to warrant abnormal use of plant equipment. However, I t
precautions in operating plant equipment should still be j l
taken in an attempt to maintain future integrity of the l equipment. Necessary actions to mitigate ICC conditions in Region Three are detailed beginning with Section 2.5 and the
{
corresponding bases in Section 3.4. r i
i i
I If the RCS P-T point reaches Region Four, the cladding !
temperatures in high power regions of the core may be 1800F l or higher. This is a very serious condition. ICC l
! conditions are now such that significant amounts of hydrogen i i
i are being formed and core damage may be unavoidable. !
t Extreme measures are warranted to prevent major core j damage. Some plant equipment may be sacrificed in an effort !
to save the reactor core. Necessary operator actions in this ICC condition are detailed in Section 3.5.
The following principles apply to actions in all regions: j
- 1. In checking each system (e.g., HPI, LPI, AFW) the i operator confirms that the system is operating correctly i and at maximum capacity. He attempts to correct any
! malfunctions in the system, if practical and possible.
} 2. If while taking actions in one region, the RCS i
! conditions enter a more severe ICC region, the operator should transfer to guidelines for the more severe ICC :
I region. !
l 3. If any time the RCS conditions enter a less severe ICC t
! region, the operator should not revert to the actions l for the less severe ICC region. The reason being that i the action which caused the conditions to improve may be ;
l-i cancelled causing the ICC conditions to worsen again. l For example, when Region Three is entered, the operator is required to start two RCPs. This action may cause l
! conditions to improve with a return to Region Two. The DATE: 7-23-85 PAGE III.F-11 j
(
CWNP.20007 3 (9 84)
S ABCOCK c witCSX NUCLEAR POWER DIVISION NUMett 74-1152414-oo TECHNICAL DOCUMENT operator should not make any further Region Three actions but should continue with the actions already made until the RC becomes subcooled or saturated. If the RCPs are tripped when the conditions return to Region Two, the conditions could go back to Region Three and this time the RCPs may not start. If the intent of a group of actions for a given ICC region has been satisfied then further actions may not be neces-sary. The operator should watch the system response to each action being made so that he can judge whether not the intent has been satisfied; e.g., the incore T/C temperature is returning to saturated conditions.
3.4 Take Recion Three Actions to Increase Heat Transfer From Reactor Core to Reactor Coolant The RCS P-T point has reached Region Three. More drastic actions are now justified to restore adequate core cooling.
If possible one RCP per loop should be started to pump any water which may be in the SG into the reactor core, even though adequate NPSH is not available. However, ICC conditions are not so severe as to justify permanently damaging RCPs. Consequently, normal RCP interlocks on cooling water and seal injection, as well as overload should not be overridden at this time. However, once a RCP is started, it should not be tripped for other than moter electrical faults unless another RCP 13 started in the same RC loop. If starting the RCP does not help cool the core, the RC will heat up into ICC region 4. Region 4 allows bypassing the RCP start interlocks (except motor electric faults) and permanently damaging the RCPs if necessary.
Consequently, if the RC is heating up into region 4, there is no reason to trip the RCP. If the RC remains in region 3 or returns to region 2, then running the RCP may be helping cool the core. Therefore, a RCP should not be tripped.
l l
DATE: 7-23-85 PAGE III.F-12
., ~ . . . ~ . .--.n-. . ~ - .- . . . - . . . . - - ~ - - - - . _ - _ _ - . . - . . . - - - -
i BWNP 20007 3 (9 84)
, BABCOCK & WILCOE NUd4BER
, NUCLEAR POWER OlVI$lON 74-1152414-00 '
l TECNNICAL BOCUMENT When the RCS P-T point is in Region Three, clad temperature l in some regions of-the core may be 1400F or higher. 7 i Hydrogen gas, along with other non-condensable gases are
! being produced. To prevent these non-condensable gases from
! collecting in the high points of the RCS, all HPV valves !
l -should be opened. Refer to Chapter IV.E for further !
i !
instruction on HPV operation. ;
3.5 Take Recien Four Actions to Maximize Heat Transfer from Reactor Core to Reactor Coolant '
The RCS P-T points have entered into Region Four. Cladding temperatures in the high power regions of the reactor core may be 1800F or higher. Significant amounts,of noncondens- ;
7
! able gas are being formed and core damage may be unavoid- !
)t able. Plant equ'ipment should now be sacrificed in order to -- >
save the reactor core. Clad-water reaction becomes a
~
significant heat source and;the core must be cooled to stop this reaction.
i To cool the reactor core with superheated RC it is necessary {
to start as many RCPs as possible to try and move all water trapped in the cold legs to the cere. All interlocks should i be defeated, if necessary,.to start RCPs. However, the overload trip circuit should not be defeated. If a RCP trips on overload, then RCP failure is imminent and may have
! already occurred. It is best to save the RCP in hopes that !
it could be restarted later. Operation of the'RCP, from the !
l time the overload trip circuit would have tripped the RCP !
i until failure of the RCP, would provida negligible cooling i to the RCS. ,
i I l
The RCS should be depressurized as quickly as possible :
j to achieve CF and LPI cooling. To achieve quick RCS l l depressurization the operating SG(s) should be depressurized ;
I i !
1 L l DATE: 7-23-85 PAGE III.F-13 j 3 )
I BcNP 20007 3 (9 84) l B ABCOCK & wtLCOE NUMBit NUCLEAR POWER DIVISION 74-lls2414-oo TECHNICAL DOCUMENT as quickly and as far as possible. However, the minimum steam pressure should not be decreased below that necessary to power the steam driven AFW pump, unless auxiliary steam is being used to power the pump. Depressurizing the SG(s) may cause permanent damage to the SG but no rupture of the RCS pressure boundary is expected. However, Region Four ICC conditions warrant this action. All possible RCS valves that vent to the RB should be opened, along with dump-to-sump valves if available. Also, all HPV valves should be opened if not already opened. Opening all HPVs at this point will also relieve noncondensable gases. Further use of HPVs are detailed in Chapter IV.E.
To depressurize the RCS low enough to dump the CFTs and start LPI may require boiling the core dry. As the core drys out and pressure decreases below the CFT pressure, the CFT will begin adding water to the RCS. This will remove core heat but may also increase RC pressure as steam is formed, therefore core cooling will by cyclic until the core is cooled below saturation temperature for the CFT prec-sure. LPI flow will not begin until the RCS is depressur-ized below LPI pump discharge pressure. This may also be cyclic if the core has not cooled enough.
O l
DATE: 7-23-85 PAGc III.F-14 i
1
g_.
i Figure Ill.F-l CORE EXIT FLUID TEMPERATURE FOR INADEQUATE CORE C0OLING 2600 2400 -
SUPERHEAT REGION 2200 -
. i TCLAD > 1400F 2000 -
s 1800 -
g REGICN 1 REGION 2 REGION 3 REGION i 4 i 1600 -
1 2
S 1400 -
- E l
O y, E
1200 -
1000 - C E
800 - 4
, 't i
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200 400 500 600 700 800 900 1000 1100 1200 1300 Core Exit Tnermocouple Temperature (F) k i
> I D0C. NO. 74-1152414-00
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IDENTIFICATION 0F ICC 3.1 Il 2.2 [
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IV.8 k, \
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2.3 l T=E ACTIONS TO INCREASE PRIMARY TO SECOPOAHY K AT TRANSFER 3.2
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\r 2.s l 2.5 1 TAKE REGION FQJR TAKE REGION THEE ACTI@ S TO MAXI- ACTIONS TO IN-MIZE MAT TRANSFER CREASE MAT TRANS-FRO 4 Ra COE TO RC FER FRO 4 Ra CORE TO RC 3.5 3,,
if I' 2.9 l 2.6 - l TAKE ACTIONS TO TME ACTImS 70 t MAXIMIZE PRIMARY TO FtmTT R INCREASE 30'XNCARY MAT PRIMARY TO SECOPO.
mA W ER ARY E AT TRANSFER 3.5 IV.C If it E 'I I/C PRIMARY TO T RETimN TO K GION 4 N0 PO, .y /Cs SECONDARY M AT s Q f ,$AftPATION TRANSFER
, . . \ ESTA8LIS40 l
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,.t, 00C., NO. T41152414-00 %
k'
BWNP 20007 3 (9-84)
B ASCOCK & WILCOX NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT Chanter III.G Cooldown Methods
1.0 INTRODUCTION
This chapter provides technical bases for plant cooldown.
The purpose of these technical bases is to provide suffi-cient information regarding the expected NSSS performance during cooldown transients and provide guidance for the operation of key systems and equipment such that the user can develop plant specific procedures for cooldown.
Although normal and near normal conditions are considered, the main purpose is to provide technical bases for cooldown under abnormal conditions. Guidance for cooldown with a SBLOCA or with a tube rupture is provided in Chapters III.B and III.E, respectively.
G 1.1 Concerns and Obiectives 1.1.1 Concerns Several concerns are created because of various equipment failures which are assumed in this chapter. The major concerns are as follows:
A. Cooling the liquid volume under the RV head during natural circulation cooldown.
2 B. Cooling the RC in an idle RCS loop during a natural i circulation cooldown with only one operable SG.
} C. SG tube-to-shell compressive and tensile stress limits when cooling down with only one operable SG.
q D. Controlling RC pressure when cooling down with a liquid water filled pressurizer.
7-23-85 PAGE III.G-1 DATE:
BINP 20007 3 (9 84)
S ASCOCK & WlLCOR NUMBER NUCLEAR POWER OlvlS10N 74-1152414-oo TECHNICAL DOCHENT E. Reducing the RC pressure and temperature low enough to initiate DHRS operation before depleting existing cooling water inventories.
1.1.2 Obiective The objective is to cooldown and depressurize the RCS as quickly as possible to the RCS conditions which allow DHRS operation. This is done without violating equipment design limits and with any combination of the following equipment availability:
A. with or without an operable RCP B. with or without a pressurizer steam bubble C. With two, one or no operable SGs.
Furthermore, the objective assumes the cooldown follows any upset in heat transfer. However, cooldown following a LOCA and SGTR are discussed in Chapters III.B and III.E respec-tively.
2.o GENERAL OPERATOR ACTIONS This section provides a brief description of the recommended logic to be used during a plant cooldown from the initial decision to cooldown to decay heat removal system opera-tion. Figure III.G-1, "Cooldown Logic Diagram," provides a basic action / decision logic chart for detelmining the appropriate cooldown method.
A loop is provided in the logic diagram from 2.7.1 (when plant conditions are not yet established for DHRS operation) back to 2.2.1 (subcooling margin (SCM) status). The purpose of this loop is to signify continuous surveillance of key plant conditions during the cooldown that may a) require changing cooldown methods (e.g., lous of reactor coolant pumps (RCPs)) or b) permit changing to a preferred cooldown method (e.g., heat transfer restored to both steam gener-ators (SGs)).
DATE: 7-23-85 PAGE III.G-2
. _ _ _ _ . - . . ~ - - - . , _ . . _ . _ . _ _ . _ . . _ _ . - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ , _ . _ _ _ _ _ , _ _ _ _ _ _ _ _
j 0%NP 20007 3 (9 SS) l i BASCOCE a witCOE
""0I NUCLEAR POWER DivtS60N 74-11s2414-oo
! TECNNICAL 80CONENT 1
i
- The logic diagram is structured such that positive responses ;
}
to normal cooldown criteria lead down the far left vertical l j path. Any abnormal conditions that af fect the cooldown i method result in branches to the right. !
! (
2.1 Determination of Cooldown Reauirement [
l The decision to cool down the plant should be based on an +
} evaluation of specific plant conditions weighing both l 4
present and future plant safety and control at hot standby
- against' cooling down with abnormal plant conditions. The ;
i recommended procedure is to restore the plant to as near j normal conditions as possible prior to cooldown unless j
! compelling reasons exist to continue the cooldown. The I
! reactor must be shut down within applicable reactivity
! limits prior to starting cooldown operations. Care must be !
l taken to maintain the shutdown margin during cooldown. l i .
This chapter of the technical bases assumes the decision to i cooldown has been made and neither a SG tube leak nor a IACA exists. Cooldown with a LOCA or SGTR is described in [
Chapter III.B or III.E, respectively. However, this chapter !
does describe the impact of a simultaneous SGTR. l i
I 2.2 Subcoolina Marcin and Primary to Secondary Heat Transfer ,
2.2.1 Subcoolina Marain f
l Verification of subcooling margin (SCM) is the first major- [
i decision point because it determines available equipment and !
4 .
heat transfer characteristics: !
, t l A. Heat removal from the core and transport to the steam ;
- generator is most easily accosplished when the fluid is subcooled.
)
B. Loss of SCM requires certain prompt actions such as the securing of the RCPs.
C. Maintaining SCM is very important in maintaining optimum l Plant control and performing an expeditious cooldown. l DATE: 7-23-85 PAGE III.G-3
l BONP-20007 3 (9-84) eAscoca a wtLCOR NUCLEAR POWER DIVISION NUM8 t t 74-1152414-oo TECHNICAL DOCUMENT The logic diagram continuously loops back to this decision point for the duration of the cooldown until the DHRS is in operation. This signifies a constant surveillance of SCM and other key plant conditions during the performance of the remaining cooldown. The concept and definition of SCM is discussed in Chapter II.B.
2.2.2 Loss of Subcoolina Marcin on loss of SCM, the operator must perform or verify the following actions:
A. Trip all RCPs.
B. Initiate full HPI (MU) flow with two HPI (MU) pumps from the BWST.
C. Begin raising SG levels to the loss of SCM setpoint.
(Chapter IV.C discusses raising the SG level.)
Further discussion of actions (and their bases) requirod upon loss of SCM is provided in Chapter III.B. Since this chapter does not assume a SGTR or LOCA, the loss of SCM is assumed to be a temporary condition (e.g., overcooling) that can be corrected and SCM restored. Chapter III.B covers cooldown with a sustained loss of SCM.
When SCM is restored, restart RCPs (if all other conditions for RCP restart exist) and shift SG level control to the appropriate setpoint. In addition, throttle HPI (MU) flow as necessary (Chapter IV.B) to maintain pressurizer level and minimum SCM and, if applicable, maintain RCS pressure below the pressurized thermal shock (PTS) limit (Chapter IV.G).
2.2.3 Primary to Secondary Heat Transfer The second major decision point in the logic diagram in verification of controlled heat transfer in both SGs. The preferred cooldown method is a normal cooldown with forced DATE: 7-23-85 PAGE III*C"4
{
~. _ . - - _ _ _ _ - - . . . . . - . _ . . .
9 ;
? e BINP-20007 3 (9 84)
BASCOCK & wlLCOR NU"#EE NUCLEAR POWER Divi $10N 74-11s2414-oo 4
TECHNICAL DOCUMENT
. flow and controlled heat transfer to both SGs. Methods of
- verifying heat transfer and recognizing inadequate or excessive primary to secondary heat transfer are discussed in Chapter II.B. Methods of mitigating inadequate or excessive primary to secondary heat transfer are discussed in Chapters III.C and III.D respectively. If controlled heat transfer exists in both SGs, then proceed down the left l vertical path of the logic diagram (toward a normal cool-down). ,
1 r
2.2.4 Loss of Controlled Primary to Secondary Heat Transfer If controlled primary to secondary heat transfer does not exist in either SG, restore controlled heat transfer in at least one SG (and preferably in both SGs) as soon as poscible. Possible causes for inadequate primary to ;
secondary heat transfer and methods of restoring primary to ,
l secondary heat transfer are discussed in Chapter III.C.
I Possible causes for excessive primary to secondary heat transfer and methods of restoring controlled primary to i
secondary heat transfer are discussed in Chapter III.D.
2.2.5 Restoration of Primary to Secondary Heat Transfer to at Least One SG If controlled primary to secondary heat transfer is restored
! to at least one SG, return to the far left vertical path.
Subsequently, a decision point is used to separate cooldowns
(
with one SG and cooldowns with two SGs. If primary to secondary heat transfer was lost and cannot be restored to either SG following the guidance provided in Chapter III.C, ,
l then HPI/MU cooling will be required (2.2.7).
i j 2.2.6 Number of Steam Generators Operatiner (Detailed discussion in l Section 3.5) i -
If the operator is controlling the feeding and steaming of !
i, both SGs, the cooldown can continue in a normal mode '
i i 7-23-85 PAGE III.G-5 DATE:
. __ _ _ .,_.. _ _ _ . _ _ _ . . _ . _ _ _ _ _ _ ~ . . _ _ _ _ . . . _ , . _ _ . _ _ , . . . _ . . _ _ , . . _ _ , . . . _ _ _ . _ . . . _ _ _ _ _ _
CNP 20007 3 (9 84)
S ABCOCK & wtLCOR NUM4t t NUCLEAR POWER OlvlSION 74-11s2414-oo TECHNICAL DOCUMENT (depending on RCP and pressurizer status). If the operator can control heat removal in only one SG, continue attempts to restore primary to secondary heat transfer in the other SG. Section 3.5 of this chapter discusses the special considerations of single SG cooldowns.
2.2.7 HPI Coolitig (Detailed discussion in Section 3.9)
If neither SG can provido primary to secondary heat trans-fer, initiate HPI cooling per Chapter IV.B. HPI cooling will be required until DHRS initiation unless SG operation can be restored. The operator should continue efforts to restore primary to secondary heat transfer in the SG.
Restoration of controlled primary to secondary heat transfer to either SG is covered by the eventual loop back through 2.2.1. Methods of initiating and controlling HPI cooling are described in section 3.9 of this chapter.
2.3 Reactor Coolant System Flow 2.3.1 Reactor Coolant Pumo Status (Detailed discussion in Section 3.2)
RCP operation is preferred since it may provide pressurizer spray, eliminates RV head and idle loop cooling concerns, alleviates thermal shock concerns (if HPI cooling is being used), mixes the RCS (providing better use of RC temperature measurements), and allows for faster cooldown. Section 3.2 discusses alternate depressurization methods for RCP combinations which do not provide sufficient pressurizer spray flow.
2.3.2 Criteria for RCP Restart If the RCPs are not operating, attempt to satisfy the criteria for restarting the RCPs as soon as possible. The criteria for RCP restart are provided in Chapter IV.A.
DATE: 7-23-85 PAGE III.G-6
BWNP 20007 3 (9-84)
S ASCOCK & wtLCOX NUM8tt NUCLEAR POWER DIVISION 74-1152414-oo
/'3 TECHNICAL DOCUMENT
\
\v' Reactor Coolant Pumo Restart 2.3.3 Restart RCPs when the criteria for RCP operation are satisfied. The number and selection of the RCPs to be restarted depends on plant conditions. Running one RCP in each loop balances heat transfer, but other pump combina-tions may produce higher spray flow. Section 3.9.1 dis-cusses RCP operation considerations applicable for HPI cooling. Refer to Chapter IV.A for precautions and recom-mondations to be considered before RCP restart and for expected system response to RCP restart.
- 2. 4 Reactor Coolant System Pressure Control 2.4.1 Pressurizer Status (Detailed discussion in Section 3.4)
Normal RCS pressure control is achieved with a pressurizer steam bubble, spray and heater operation, and with the rest of the RCS solid and subcooled. If a pressurizer steam b3 )
t bubble does not exist and cannot be drawn, then solid plant pressure control will be required. If a steam bubble does exist but heaters and/or spray capability is lost, solid plant pressure control may be required (if the pressurizer cannot be maintained as the hottest region in the RCS) .
Maintenance and restoration of a pressurizer steam bubble and transition to solid plant pressure control are discussed in section 3.4.
If normal pressure control is hindered due to voids in the hot legs or RV head, actions should be taken to elimin-ate the voids (see Sections 3.7 and 3.8).
2.4.2 Solid Plant Pressure Control (Detailed discussion in Section
, 3.4) l During solid plant operation, primary pressure is controlled n by primary inventory controls (e.g. MU/HPI and letdown /
/ s i ( ) seal bleed off/ leakage flow). Solid plant pressuro control i s -
'~ is not the preferred path and therefore attempts should be DATE: 7-23-85 PAGE III.G-7 l
8 NP-20007 3 (9-84)
SAtcocg a watcom NUCLE AR POWER OlvlSION "
74-1152414-oo TECHNICAL DOCUMENT made to restore pressurizer steam bubble control whenever possible. Solid plant pressure control is discussed in section 3.4.
2.5 Cooldown and Depressurization 2.5.1 Normal Cooldown (Detailed discussion in Sections 3.1 and 3.2)
A normal plant cooldown is defined as one where adequate SCM is maintained while utilizing forced flow, two steam generators, and normal RC pressure control with a pressur-izer steam bubble. This chapter provides guidance for plant cooldown when one or more of these do not exist, except for sustained saturation which is covered in Chapter III.B. The purpose of the dashed box on the logic diagram (Figure III.G-1) is to signify coverage of normal plant cooldown by site specific procedures (i.e., not covered by this chap-ter). In addition, the inclusion of this box in the logic diagram signifies that the site specific procedures should have decision points to return to cooldown procedures for abnormal plant conditions should the plant status change during the cooldown (e.g., loss of RCPs).
2.5.2 Abnormal Forced Circulation Cooldown (Detailed discussion in Sections 3.1 and 3.2)
An abnormal forced circulation cooldown is defined as one with forced flow combined with one inoperable SG and/or solid plant pressure control.
For forced circulation with one SG inoperable, SG shell cooling must be considered by the operator in maintaining tube-to-shell delta T limits (see section 3.6). Solid plant pressure control is discussed in Section 3.4. The operator should attempt to restore equipment such that a normal cooldown can be initiated.
DATE: 7-23-85 PAGE III*C-8
BONP 200013 (9 84)
BASCOCK & witCOX NUMSE R NUCLEAR POWER OlvlSION 74-1152414-o0 h TECHNICAL DOCUMENT 0 2.5.3 Natural Circulation Cooldown (Detailed discussion in Sections 3.1, 3.2 and 3.3)
A natural circulation cooldown may be required following a loss of RCPs. It is preferable to maintain existing RCS conditions and just remove decay heat with natural circula-tion until the RCPs become available. This is because of specir.1 considerations applicable during natural circulation cooldowns: lack of normal pressurizar spray, RV head cooling (2.6.2), idle loop cooling (2.6.3), and SG shell cooling ( 2. 6.1) . Should it become necessary to perform a natural circulation cooldown, these considerations must be addressed.
2.6 Soecial Considerations for Abnormal Cooldowns 2.6.1 SG Shell Coolina (Detailed discussion in Section 3.6) n Actions to increase SG shall cooling are performed to allow single loop cooldowns to proceed at an appreciable rate without violating tube-to-shell temperature differential limits. Methods of enhancing SG shall cooling are discussed in Section 3.6.
2.6.2 Reactor Vessel Head Coolina (Detailed discussion in Section 3.7)
Actions to increase RV head cooling are performed to allow natural circulation cooldowns to proceed at an appreciable rate without causing head void formation. Cooldowns with a void in the upper head can proceed at 50F/hr without causing excessive head stresses. Methods for enhancing RV head cooling, and recognizing and' eliminating head voids, are discussed in Section 3.7.
2.6.3 Idle Loon Coolina (Detailed discussion in Section 3.8)
A Actions to increase idle loop cooling are performed to allow single loop natural circulation cooldowns to proceed at an appreciable rate without causing idle loop void formation.
7-23-85 PAGE III.G-9 DATE:
BONP 30007 3 (9 84)
B ABCOCK & WitCOK NUCLEAR POWER Divi $10N 74-1152414-00 TECHNICAL. DOCUMENT Methods for enhancing idle loop cooling, and recognizing and eliminating loop voids, are discussed in Section 3.8.
2.7 Decay Heat Removal System Operation 2.7.1 Plant Conditions (Detailed discussion in Section 3.10)
Continue plant cooldown with either SG(s) or HPI cooling until conditions allow DHRS operation. A loop is provided back to 2.2.1 when DHRS conditions have not yet been attained. The purpose of this loop is described in Section 2.0. Required conditions of DHRS operation are provided in Section 3.10 of this chapter.
2.7.2 DHRS Initiation (Detailed discussion in Section 3.10)
When plant conditions for DHRS operation have been met, place the DHRS into operation and continue the cooldown per the applicable site procedures to cold shutdown. Continued RCP operation will aid cooling of the RV head and loops but operation may be restricted because of RCP requirements.
Transition to DHRS operation is described in Section 3.10.
3.0 ABNORMAL COOLDOWN TECHNICAL BASES This section provides a detailed discussion of abnormal cooldowns and the technical bases for operator actions described in Figure III.G-1.
3.1 Cooldown Limits and Considerations Normal and abnormal cooldowns must observe the following constraints. (Shown in parentheses are the plant parameters that must be monitored continuously to ensure these critoria are not violated).
e Cooldown rate limit.
(Average primary temperature vs. time) e Subcooling margin.
(Primary pressure and hot leg or incore temperature)
DATE: 7-23-85 PAGE III.G-10
s:NP.20oor.3
erecoca a wacom NU"0IE NUCLEAR POWER DIVI $10N 74-1152414-oo TECHNICAL DOCUMENT e Reactor vessel P-T and pressurized thermal shock (PTS) limits.
(Primary pressure and cold leg or incore temperature) e Pump NPSH limits. (when applicable)
(Primary pressure and temperature) e Tube-to-shell delta T limits.
(Primary and SG shall temperatures) e Fuel-in-compression limits.
(Primary pressure and Thot t RCP status) e Shutdown margin.
(Boron concentration measured periodically)
Normal cooldowns are preferred because they can be better controlled and because faster cooldown rates can be ach-inved. When the normal cooldown criteria are not satisfied, additional cooldown concerns arise. Figure III.G-2 is a O, matrix of plant status vs. cooldown concerns. These cooldown concerns, along with others, are listed below along with the technical bases section in which they are addres-sed.
e Solid Plant Pressure Control (3.4).
e SG Shell Cooling (3.6).
e Reactor Vessel Head Cooling (3.7).
e Idle RC Loop Cooling (3.8).
o HPI Cooling (3.9).
It should be emphasized that attempts should be made to return plant status to the normal cooldown path whenever possible.
3.1.1 Ircact of Tube Ruoture on Plant Cooldown Steam generator tube ruptures (SGTR) complicate plant cooldowns and will usually result in abnormal cooldowns. To minimize offsite releases, expeditious cooldowns may be required. Therefore, the operator may not be able to prevent RV head or loop voiding. However, void formation 7-23-85 PAGE III'G"II DATE:
CwNP-20007-3 (9 84)
S ABCOCK & wlLCOE NUM8 t t NUCL( AR POWER DIVl510N 74-1152414-oo TECHNICAL DOCUMENT will hinder RCS depressurization because the void acts like a pressurizor. This will result in a prolonged cooldown and may result in higher integrated tubo leakage. SGTRs may also result in single loop natural circulation cooldowns or HPI cooldowns because of loss of SCM and SG isolation.
Chaptor III.E contains a discussion of SGTR concerns.
3.2 Alternate Depressurization Methods If pressurizar spray is unavailable (e.g. loss of RCPs) and solid plant pressure control is not required, the following methods may be used for primary depressurization. Note: On some plants, certain RCP combinations result in little or no spray flow. For these cases the following mothods may be used to augmont primary depressurization.
3.2.1 Hich Pressure Auxiliarv Sorav (if available)
High pressure auxiliary spray is a backup to normal prosaur-izer spray. Plant pressure control methods for high pressure auxiliary spray should be the same as normal spray though depressurization rates may be lower than those of normal spray. Should voids form in the RV head or an idlo loop, achievablo depressurization raton will decrease.
Actions may be taken to oliminato any voids that form. (soo Sections 3.7, 3.8).
High pressura auxiliary spray may be used if the temporaturo differential betwoon the auxiliary spray nozzle and the spray water is not greator than the limit (soo Plant Limits and Procautions). Onco initiated, a continuous minimum flow should be maintained to limit thormal cycles. Also, adoquato llPI pump racirculation flow should be maintained to provido the uinimum required IIPI pump flow. The lotdown flow rato should be adjusted if necessary to control prosaurizor lovel.
DATE: 7-23-05 PAGE III.G-12
C"ZNP 20007 3 (9 84)
BASCOCE & WitCOR NUCLE AR PO*tR DevlSION Numsta 74-1152414-00
() TECHNICAL DOCUMENT J
3.2.2 Ventinct Depressurization rates with the PORV will be larger than l with the high pressure auxiliary spray. At operating pressures, PORV deprosnurization rates with a pressurizer steam bubble are 180 to 240 psi / min, while pressurizer vont depressurization rates are expected to be much less. Using )
the PORV or pressurizer vont has the characteristic of removing coolant from the system, which must be made up (in addition to the contraction experienced during normal cooldowns). The operator should monitor quench tank instrumentation to provent unintentionally blowing the rupturo disk on the tank. Sufficient dischargo of fluid will degrado the environment of the RB.
g- During venting, SCM and prosaurizer level should be moni-( ) tored. Primary inventory and prosauro controls (makeup /IIPI I and letdown) should be adjusted to provent loss of SCH and to maintain pronaurizer level on scale. Allowing the RCS to I becomo saturated will lead to void formation in the RV head 1
or hot logo which will tend to slow the deprensurization. l If voids form, pressurizar venting should be stopped if the voids are to be oliminated. (soo Sectionn 3.7 and 3.8).
3.2.3 Ambient Losses l
If other depressurization methods are not unod, the primary '
proosure will decrease slowly due to ambient loonos. l Deproonurization ratos dopond on proneurizar inventory and l pronaurizar innulation (an-built); a nominal value is 0.5 i pol / min. Doprosaurization duo to ambiont lonnon is not normally unod for cooldowns but may becomo important in maintaining pronouro at hot standby if heatern are unavail-n able. To minimize the pronouro reduction due to ambient i
( )
lonnon the proasurizar level should be maintained no low no practical sinco the heat transfor rato in the steam region DATE: 7-23-05 PAGE III'G"13
BZNP-20007-3 (9-84)
BABCOCK & WILCOX NUCLEAR P0wtR DmSt0N 74-1152414-oo TECHNICAL DOCUMENT is considerably lower then the heat transfer rate in the liquid region.
3.2.4 Lowerina the Pressurizer Level and Re filliner It with Subcooled Water The temperature inside the pressurizer determines the pressurizer pressure. If the temperature is reduced the pressurizer pressure will also be reduced. The temperature can be reduced by draining the pressurizer then refilling the pressurizer with colder water. The pressurizer tempera-ture will decrease to a now, lower equilibrium temperature.
Overall, the pressurizer pressure will decrease from an initial saturation pressure to a new saturation pressure.
However, the decrease in pressure will not be steady and continuous. The pressure will decrease as the pressurizer is drained then increase as the pressurizer is refilled then decrease as the pressurizer cools.
3.2.5 Imoact of Steam Generator Tube Ruoturu Steam generator tube ruptures remove primary inventory and therefore tend to depressurize the RCS. The tube rupture (s) will cause pressurizer level to decrease. HU or HPI flow should be increased to maintain pressurizer level.
This will help stabilize RC pressure.
3.3 Natural Circulation Cooldown 3.3.1 Natural Circulation Cooldown Concernq Natural circulation occurs in the RCS if the RCPs are not operating and certain conditions are satisfied (e.g., heat cource at a lower elevation than heat sink, solid loop). It is preferable to remove decay heat with natural circulation and maintain existing primary conditions until the RCPs become available. The difficulties in performing a natural circulation cooldown ares e R'/ head cooling DATE: 7-23-85 PAGE III.G-14
BON?-20007 3 (9 84)
S ABCOCK & WILCOx Numtta NUCLEAR POWER DIVISION 74-11s2414-oo
[)';
q TECHNICAL DOCUMENT e Idle RC loop and SG shell cooling e Primary pressure control e Longer cooldown time (FW requirements) e Possible thermal shock (if on HPI cooling)
During natural circulation, the RV head fluid remains stagnant and cools very slowly (unless the plant has a passive vent system). To prevent head void formation head cooling must be enhanced or the cooldown rate must be limited (see Section 3.7 for a more detailed discussion of RV head cooling). A similar situation occurs in the hot leg during single loop natural circulation cooldowns. The idle loop of the RCS stagnates and cools slowly, unless action is taken to enhance idle loop cooling (see Section 3.8 for more detailed discussion of idle loop cooling) . Also, SG shell cooling dif ficulties may occur during single loop natural
(] circulation cooldowns (see Section 3.6 for a more detailed
( ) discussion of SG shell cooling) . Normal pressurizer upray is unavailable following a loss of RCPs. Therefore, alternate primary depressurization methods are needed during natural circulation (see section 3.2).
Because of lower achievable cooldown rates and the previ-ously mentioned difficulties, longer cooldown times may result before reaching DHRS initiation for natural circula-tion cooldowns. Longer cooldown times result in larger FW requirements if steam is being vented to the atmosphere.
3.3.2 Natural Circulation Cooldown Initiation and Verification To initiate a natural circulation cooldown when adequate subcooling margin exists, the water level in each SG should be raised to the natural circulation level setpoint. If the RCS is saturated, raise the level to the loss of SCM
,] setpoint. The TBVs (condenser dump valves or ADVs) should l
's ! be opened and controlled to limit the cooldown rate. Enough
~
boron should be in the reactor coolant system to ensure DATE: 7-23-85 PAGE III.G-15
l I
BZhP 20007 3 (9 84)
SASCOCK & wlLCOE NUM8th NUCLEAR POWER OtVl$10N 74-1152414-oo TECHNICAL DOCUMENT preservation of shutdown margin should fluid in regions such as the pressurizer, idle loop or RV head mix with the remainder of the RCS and subsequently enter the core.
Once natural circulation has been initiated and verified, RV head cooling considerations should be addressed (see Section 3.7). For the case of single loop natural circula-tion cooldowns, idle loop cooling and shell cooling diffi-culties should be addressed (see Section 3.8 and 3.6).
Verification of natural circulation and recognition of a loss of natural circulation are discussed in Chapter II.B.
3.3.3 Imoact of Steam Generator Tube Ruoture Natural circulation cooldowns Vith SGTRs may result in an idle loop because of SG isolation or a 1C,op with intermit-tent flow because of periodic steaming. In either case, one loop may be cooling much more slowly than the other and voids will result if saturated conditions occur in the loop that is cooling much more slowly. Discussions of the prevention and elimination of loop voiding are contained in Section 3.8.
3.4 Solid Pressurizer Operation Plant operation with a pressurizer steam bubble is prefer-able to solid pressurizer operation. With a pressurizer steam bubble, primary pressure can be increased using the pressurizer heaters and decreased with pressurizer spray (if RCPs are operating) or with one of the alternato depressuri-zation methods described in section 3.2. With a solid pressurizer, primary pressure can be increased or decreased using primary inventory controls (makeup /HPI and letdown).
However, because of water's low compressibility, small changes in inventory or temperature result in large pressure
! responses. Primary pressure control is thus more censitive with a solid plant than with a pressurizer steam bubble.
DATE: 7-23-85 PAGE III.G-16
BWN7 20007 3 (9 84)
SASCOCK &' wlLCOM NURStB {
NUCLEAR POWER DIVISION 74-lls2414-oo j TECHNICAL BOCUMENT one of the . operator's objectives is to prevent pressurizer l t
safety valve challenge. - During solid plant operation this l objective has added importance because of an increase !
potential of failing the safety valve due to passing water.
Under some conditions, it may be difficult to control primary pressure with a steam bubble in the pressurizer.
For instance, ambient heat losses from the pressurizer may j prevent maintenance of a steam bubble if heaters are not 5 available. It may then be easier to control pressure with a solid pressurizer than with a pressurizer steam bubble. The pressurizer should be filled slowly to avoid an excessive pressure increase when the steam bubble completely col-lapses. It is pref.zrable to use the pressurizer vent instead.of the PORV to eliminate the steam bubble since the f vent has a smaller relief rate. To fill the pressurizer solid, MU/HPI flow must exceed pressurizer vent and letdown flow. A full pressurizer is indicated by a sudden change in i rate of pressure increase. As the vent is closed, MU should f be throttled.to stop the pressure increase. MU and letdown f
can then be used to control RCS pressure during the cool-down.
l During solid plant operation sudden large pressure increases {
or decreases may occur when . tripping or- restarting reactor l coolant pumps (see chapter IV.A for RCS response to RCP l trip / restart during solid plant operation). ! l 3.4.1 Incact of Stean Generator Tube Ruoture For solid pressurizer operation, SGTRs represent another {
outflow path that must be considered along with the other inventory controls: MU, letdown, HPI, PORV, vents. Also,
-rapid pressure changes that can occur during operation with a solid pressurizer can affect the leak rate of the SGTR and the pressure in the SG if the SG is also filled.
DATE: 7-23-85 PAGE III'0-17
_ _ _s
BWNP 20007 3 (9 84)
B ASCOCK & w LCOM j NUCLEAR power DIVl510N NUMatt n -lls2 m -oo TECHNICAL DOCUMENT 3.5 Number of Steam Generators Operatina It is preferable to ccoldown with both SGs controlling primary to secondary heat transfer. Two cooldown concerns exist when only one SG is available: SG shell cooling (forced and natural circulation) and idle RC loop cooling (natural circulation). SG shell cooling concerns are discussed in 3.6 and idle loop cooling concerns are discus-sed in 3.8.
From a heat removal standpoint proper SG operation with a subcooled primary system has the following characteristics:
e T cold approximately equal to SG T sat e Automatic or operator controlled FW maintains constant level at the appropriate setpoint.
e Steam pressure is controllable.
Single SG operation implies that one SG has been isolated because of a lack of heat transfer (III.C), excessive heat transfer (III.B) (e.g., unisolable steam leak), or Steam Generator Tube Rupture (III.E). The isolated SG will not possess all of the above characteristics.
During the cooldown, controlled heat transfer should be restored to an idle SG whenever possible to help alleviate the shell cooling and idle loop concerns.
3.5.1 Impact of Steam Generator Tube Ruptures SGTRs may result in SG isolation or limited steaming in one l
l or both SGs.
t 3.6 SG Shell Temperature ConcerAs SG shell cooling concerns arise when one SG is isolated, for both forced and natural circulation cooldowns. In dry idle SGs, the shell is no longer cooled by steam and FW flow but DATE: 7-23-85 PAGE III.G-18
+
B2NP 2OOG7 3 (9 84)
B ABCOCK & witCOE NUMBER NUCLEAR POWER OlVl$10N 74-1152414-00 OTECHNICAL DOCUMENT V rather by ambient losses. Limits pertinent to cooldowns and shell cooling are:
e Normal tensile tube-to-shell Delta T (tube colder) < 100F (177 FA Plants) 160F (205 FA Plants) o Compressive tube-to-shell Delta T (shell colder) < 60F (177 FA Plants) 60F (205 FA Plants) e Emergency Tensile Tube-to-Shell Delta T (tube colder)
< 150F (177 FA Plants) 160F (205 FA Plants)
During forced circulation cooldowns with one dry, isolated SG the associated RC loop is cooled via the unisolated SG. During natural circulation cooldowns with one dry, isolated SG, the loop is cooled using idle loop cooling {
methods (see Section 3.8). In both cases, the shell cools due to ambient losses unless other shell cooling actions are
) performed. The cooldown should be controlled such that:
Average SG Shell Temperature - Tcold < Tensile Tube-to-Shell Delta T ADA ,
l T hot
- Average SG Shell Temperature < compressive Tube-to-Shell Delta T These formulas ensure a conservative definition of the !
temperature difference.
Seal injection flow may influence the Tcold reading in an sat f a Pressurized SG is more idle loop. In this case, T (
, representative of the tube temperature than Tcold, assuming l the idle loop SG is pressurized. [
l I If the tubes are cooling faster than the shell, then the cooldown rate will have to be decreased or shell cooling ;
, increased to prevent violation of the tensile tube-to-shell delta T limit. If the shall of the isolated SG is cooling DATE: 7-23-85 PAGE III+G-19
82NP-20007 3 (9 84)
B ASCOCK & wlLCOE NUM8tt NUCL[AR POWER Olvf$t0N 74-1152414-0o TECHNICAL DOCUMENT faster than the tubes, increasing the cooldown rate using the operable SG, if possible, may provent violation of the compressive tube-to-shell delta T limit.
The SG shell will cool slowly via ambient losses. If it is cooling too slowly the operator can supplement the ambient losses by providing a source of water to produce steam for heat transfer between the shell and the tubes. This is more offective in the 205FA plants since the steam traverses most of the length of the SG shell. To accomplish this, some source of FW must be available. Introduction of AFW is preferred. Some of this FW will be flashed into steam in a hot SG but some will also tend to run down the tubes and cool them. Introducing FW at a sufficiently small rato may provido an adequate supply of water to produce steam for shell cooling while a minimum number of tubes are impacted over a minimum length. The key aspect is that the estab-lishment of steaming providos a thermal coupling between the tubes and the shell. If the operator observes that small rates of AFW injection to that idle SG are causing the tensi~0 . tube-to-shell temperaturo difference on the SG to approach the limits without cooling the SG shell, ho should stop AFW injection to the idle SG.
Steam may also be produced to cool the SG shall by introduc-ing MFW. In the 177FA plants the flow through the MFW nozzles helps ccol the lower portion of the shell. However, introduction of )!FW into a dry SG may bo undesirable from the standpoint of the thermal stresses which develop in the lower tubeshoot region. A stress analysis will be required to assess the possibility of having damaged the SG if this procedure is followed. If MFW is used (i.e., AFW is not available) a small but continuous flow in desirable to minimize thermal cycling in the lower tuboshoot region.
7-23-85 PAGE III.G-20 DATE:
BINP 20007 3 (9 84)
S ADCOCK & wnCOR NUM$tt NUCLEAR POWER OlvtSION 74-115'414-C'iTECHNICAL DOCUMENT
?
)
Additional SG shell cooling will be provided if a level can be established in the idle SG.
During single loop forced flow cooldown the idle SG tube temperature will follow the temperature in the active loop.
If the shell temperature cools too slowly and the tensile ,
limits in the idle SG are approached, the cooldown rate can be slowed or stopped while the shell cools. Another option in single loop operation with forced flow is to try to 1 establish heat removal with the idle SG before continuing
] with the plant cooldown. However, while attempting this, the idle SG shell will cool due to ambient losses while the i,
tubes remain hot and there is a potential for approaching l the compressive shell/ tube temperature limits. In that case i
the operator must cool the primary fluid by steaming the (
l active SG.
' 5 V
In single loop natural circulation the idle loop SG is not directly influenced by steaming the active SG. The idle I loop shell will cool by ambient losses while the active loop j temperature drops in accordance Vith the cooldown. During this period the idle loop temperature remains high. The
! operator may have to cool the idle loop to prevent the
! occurrence of saturated conditions as well as to maintain the tube-to-shell temperature difference below the compres-
! sive limits (shell cooler). To do this the operator induces natural circulation in the idle loop by injecting AFW and I
steaming the idle SG. In addition to the cooling provided by the CG, some of the cooler active loop fluid is circu-lated into the idle loop. As a result, if the active loop has been cooled too far below the idle loop shell tempera-ture, excessively cold water will be introduced into thn
, idle loop. Actions to cool an idle loop should be performed l before the active loop temperature decreases below the idle SG shall temperature by an amount (100F for 177 FA plants, 7-23-85 PAGE III.G-21 DATE:
i
B2NP 20007-3 (9 80)
BASCOCK & wlLCOE NUCLEAR power Olvl510N Numbia 74-1152414-00 TECHNICAL DOCUMENT 160F for 205 FA plants) that could result in excessive tensile stress in the idle SG. Further discussion of cooling an idle loop is provided in section 3.8.
3.6.2 Impact of Steam Generator Tube Ruotures SG tube-to-shell limits are loss of a plant control problem when a lovel exists in the SG. SGTRs may result in SG isolation. In this caso, if the loop is idle, the SG may have a substantial inventory and the shell will cool more slowly than a dry depressurized SG. However, the shall below the water level should cool at about the same rato as the tubos. Oponing the drains (if available) may also cause the shell and tubes to cool. If conditions allow, as discussed in Chaptor III.E, stoaming the isolated SG will induce shall cooling.
3.7 Reactor vessel Head Coolina ConcQEDA During natural circulation, the RV head fluid romains stagnant and does not communicato with the rest of the RCS (unless the plant is equipped with a passivo vont sys'.om) . Thus, the RV head fluid cools slowly (4F/hr to 5F/hr) . The impact on the cooldown is as follows:
e The cooldown rato may have to be slowed if RV head void formation in to be provented without venting the RV head.
e Head void formation will slow the depressurization and thus the cooldown.
3.7.1 Head Void Prevent 1QD RV head void formation can bo procluded by controlling the cooldown and deprosaurization auch that the RV head fluid tamparaturo is loss than the primary saturation tempora-tura. RV Hoad fluid temperaturo may be obtained directly from a temperaturo moanuromont (if available) or inferrod from predicted plant specific head flitid cooldown ratos.
DATE: 7-23-85 pAcg III.G-22
B;NP 20007 3 (9 84) 6 ASCOCE & witCOE NUCLE AR power DivlSION NUM8th 74-1152414-oo J TECHNICAL DOCUMENT Fer plants without RV head vents, RV head ambient cooldown ratas are about 4F/hr to 5F/hr. (NOTE: GPUN has performed independent analysos, not verified by D&W, that indicate these cooldown rates can be as high as 8F/hr to 30F/hr depending on RC cooldown rates.) To avoid steam formation, slow cooldowns are required. Periodic bumping of the RCPs will help koop the temperature in the RV head lower.
Higher cooldown ratos can be achieved with passivo vont systems (flow path betwoon RV head and the hot log or steam generator plonum). Flow through those linos occurs duo to density differences betwoon the RV head fluid and the dischargo (hot log or SG plonum) fluid. There are two parallel paths betwoon the RV and the upper portion of the hot log. The normal path through the hot log will have most
,-m but not all of the flow. Thoroforo, faster cooldowns can bo
/ s
( ) performed without void formation. In situations where flow
through the passivo vont system cannot occur (e.g., voida or stagnant conditions in dischargo hot log), the RC loop !!PV must bo opened to initiato flow through the passive head vont lino.
Somo plants are equipped with activo vont systems in which operator action is required to provido a flow path from the RC head to a suitablo dischargo (e.g. quench tank) . Itoad fluid temperaturo responno during venting in dopondent on primary pressure, temperature, RV head temperatura, and venting configuration (minimum vont flow aron).
A samplo outlino of a method for cooling the RV head using an activo vont is provided below, e Dotormino head fluid temperatura oither from direct
,m monsuroment or estimato from previously calculated
( ) valuo. If venting has not boon performod, estimato RV hand fluid tamparaturo as the highest incoro T/C DATE: 7-23-05 pggg III.G-23
CT NP 20007 3 (9 84)
S ABCOCK & WILCOR NUM8 t B NUCL( AR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT reading since the RCPs were last operating adjusted for RV head ambient heat lossos as described above.
e Cool the primary until a 50F temperature differential exists betwoon the head fluid temperature and incore T/C. Maintain at least 20F SCM in the RV head.
It is proforable to maintain maximum SCM sinco the venting rato increason with primary prosauro and the head fluid cooldown rato increason with venting ratos, e When a 50F head fluid-incoro T/C delta T in achievod, halt the primary cooldown and maintain constant primary temperature with natural circulation continuing. Open the RV head vent. Maintain RC prosauro and pronsurizar level by increasing MU. Vent for 15 to 30 minutos (if possible) and uno Figuros III.G III.G-6 to antimate the RV head fluid temperaturo ronponso. The fastent RV head fluid cooldown raton occur in the first 20 to 30 minuten.
e Close the RV head vont and throttle HU to maintain primary prennure and pronourizar level.
e This venting procoon in repeated during the cooldown.
3.7.2 RV Head Void RecoonitiSD RV head void formation may be detectable using RV head lovel measuromonto, if availablo. In addition, formation of head voids can usually be annociated withs e Opposito trending betwoon RC prennure and pronour-izar and/or HU tank level (pronsurizar and/or MU tank lovel incroanon with RC prennuro decreano). Note: Thin in also an indication of RC loop void formation.
Thorofore, hot log level monnuromonta may be unoful in indicating RV head void formation.
e Difficulty in reducing prennure after void formation.
e RV head fluid tamparature (if availablo or from previoun antimaton) or pannivo RV hnad vont line temperatura equal to primary naturation temperaturo.
DATE: 7-23-05 PAGE III.G-24
j BINP 20007 3 (9 84)
SASCOCE & witCo t NUCLE AR power Divtsl0N "
. S TECHNICAL DOCUMENT 74-1152414-oo
( )
m-If natural circulation flow can be verified in both loops, it may be assumed that any void forming is a RV head void.
Note that, unless vented, the RV head will be the hottost region in the primary loop and thus more likely to void first.
3.7.3 Head Void Elimination Elimination of the RV head void should facilitato the depressurization during cooldown after RV head void forma-tion occurs. As previously mentioned, head voids tend to i sicw the depressurization, and thus the cooldown, sinco it
! acts as a second pressurizer. RV head voids can be elimin-
! ated by:
e Venting.
e Ambient Hoat Loss Induced Condensation.
t ) e RCP Restart Increasing the void sizo due to deprousurization should be provented after the vont is opened. To accomplish this, MU l
flow chould be increased to componnate for venting. The RC pressure is to be maintained or slightly increased through regulation of the size of the pressurizar steam bubblo.
Successful venting has occurred when primary pressure and pressurizar lovel increase ouddenly. This lovel increano will occur even if a void oxisto in a hot log.
RV head loval moacuromonto should only be used to indicato trendo and may not suffico to datormino if the void has boon completely eliminated. The RV head lovel monouromonto (if availablo) may not bo accurate while venting la in progrens.
1
/
s RV head void elimination due to ambient heat loss induced
( l condannation is a slow procono and may require oxtremely
(_
DATE: 7-23-85 PAGE III.C-25
SwNP-20007 3 (9 84)
B ASCOCK & wlLCOR NUCLEAR POWER Divi $l0N NUM$tt 74-11s2414-0o TECHNICAL DOCUMENT long cooldown times. RCP restart with a RV head void is addressed in Chapter IV.A.
3.7.4 Cooldown With Voided Head The RCS can be cooled and depressurized with a RV head bubble. The depretsurization rates achievable using the PORV will depend on the volume of steam in the system.
Depressurization will only be possible with the PORV while a steam bubble exists in the pressurizar. The exact reduction of the depressurization rate will depend on the sizes of the RV head bubble and the pressurizar bubble. As the cooldown progresses the RV head bubble will slowly expand. When the PORV is opened the bubble will expand rapidly and if largo enough will expand into the hot leg. For natural circula-tion flowratos as low as 3 percent, and temperatures in the hot leg within 10F of saturation, a steam void escaping into the hot log will be condonned long before it reachos the highest portion of the hot leg. A 50F/hr cooldown with a 600F steam void in the RV upper head is acceptable from an operations as well as a stress analysis standpoint. The flow regime expected at the RV outlet is one of bubbly flow as opposed to slug flow. This indicates that the condensa-tion will not be a violent one. During this modo of expansion of the RV head void (steam escaping into the hot log), the RV head void will not hindor depressurization.
3.7.5 Imoact of Steam Gonorator Tubo Runtures If a fast depressurization and natural circulation cooldown is performed for SGTRa, RV head void formation is likely.
Actions consistent with SGTR management may be taken to oliminate the RV head void and prevent subscquent RV head void formation whenever possible. (see Chaptor III.E).
O DATE: 7-23-85 PAGE III.G-26
.. . . - . . . . ... , .- . . - - - . - - _ . . . . . - ~ ..
s s BWNP 20007 3 (9 84)
BASCOCK & WILCOX l NUCLEAR power Div;$t0N ' ' * * "
74-lls24I4-oo TECHNICAL DOCUMENT a 3.8 Idle Loon Coolina During sibgle loop natural circulation cooldowns, fluid in ,
the idle loop remains stagnant and does not thermiilly.
communicate'with the rest of the RCS. It thus cools very i slowly due to ambient losses. This has the following .
- e potential in;gact on the plant cooldown: '
1 e To avoid idle loop void fcrm,ation the plant cooldown rates may have to be reduced. .
I e The formation of loop voids will slow the depressuriza- i tion and cooldown of the plant.
- e Unacceptable compressive tube-to-r. hell temperature ,
differences in the idle SG could result if the idle loop y ,
i is not cooled while the SG shall is cooling to ambient.
3.8.1 Idle RC Loon Void Prevention t Thetemperatureofthefluid,in$heidleRCloopmaybekept !
below saturation temperNture through eIther periodih f injection of AFW .to the SG in the idle RC loop or the !
bumping of a RCP in the active RC loop. The hot fluid in j the idle RC loop is replaced with fluid closer to thd active l RC loop temperature by the bumping of a RUP. The criteria j for bumping a RCP are contained in Chap.ter IV.A.
i
! Using AFW to induce idle RC loop cooling involves period-i
)
ically injecting AFW to the SG in the; idle RC loop and then observing the primary temperature response.- Since natural.
circulation will start slowly, several minutes may elapse
-before the hot leg RTD begins to change. ,
l .
I The SG in the idle loop may be; bottled up or dry. AFW (or ;
I
( MFW diverted through the AFW header) should be fed (at less than 700.gpm) to a dry-SG whenever the following conditions exist:
e Idle RC loop. tube-to-shell temperature differential (shell' hotter than the SG tubes) is increasing to within i
DATE: 7-23-85 PAGE- III.G-27
~.-m - 7 .--. _ - ,_,c,
, h
BcNP40007 3 (9 84) 8ASCOCK & wtLCOR NUMBER NUCLEAR POWER OtVISION 74-1152414-00 TECHNICAL DOCUMENT 10F of the tensile tube-to-shell temperature differen-tial limit. It is expected that this action will momentarily increase the temperature differential before the proper thermal coupling is established between the shell and the tubes. A more complete discussion of shell/ tube temperature differences is contained in Section 3.6.
e The saturation temperature during the cooldown is reduced to within 10F of the idle loop hot leg tempera-ture.
e The active loop temperature is reduced more than 50F below the idle loop temperature.
The initial injection of AFW should be two to three minutes in duration. The maximum flowrate of 700 gpm is based on engineering judgement. A sufficient flow to induce natural circulation is desired while preventing excessive cooling of the tubes. The minimum flow required to induce natural circulation flow is plant and situation dependent and has not been determined by analysis. Seven hundred gpm should
[ be sufficient based on engineering evaluations of available l data.
l l
This process promotes natural circulation in a void free loop. Idle RC loop hot leg and cold leg temperatures and tube-to-shell temperature differential should be observed for the next several minutes. Subsequent injections of AFW I may be necessary to continue the cooling of the idle RC loop l
if the active loop cooling continues. The operator should l
be able to maintain a sufficient inventory of water for l
stoaning the idle RC loop SG with small injections of FW once natural circulation is started. The small flowrates will minimize tube impingement and direct contact cooling of the SG tubes although some tubes will always be cooled by AFW flow. The goal is to establish thermal coupling between DATE: 7-23-85 PAGE III.G-28 s
CYtNP 20007 3 (9 84)
BABCOCK & WitCOX NUCLEAR POWER DIVISION NUMBER 74-11s2414-oo CTTECHNICAL DOCUMENT the tubes and the SG shell through use of the FW. If cyclic idle RC loop cooling operations are performed, a very small rate of FW should be continuously fed to the SG between cooling cycles to avoid repeatedly thermally shocking the AFW nozzles. If this process does not result in natural circulation in the idle loop, available loop instrumentation should be checked for the presence of loop voids. If voids are present they should be eliminated before trying to induce natural circulation in the idle loop.
3.8.2 Idle Loon Void Recoanition and Elimination The formation of idle loop voids can be detected by some of the same means used to determine if a RV head void exists.
These are discussed in section 3.7.2 and include opposite trending of pressurizer pressure and pressurizer level and difficulty in reducing pressure with the PORV or pressur-izer. A RC loop void may be determined from the trend of loop level measurements. If the hot leg temperature becomes equal to the saturation temperature, void formation may be imminent if it has not already occurred. An unexpected increase in pressurizer level is also a symptom of loop void formation. However, this is inconclusive by itself since it is also a characteristic of RV head voiding.
Three means of void elimination are:
e venting through the HPVs, e bumping a RCP, e allowing the void to condense through ambient heat losses.
The conditions prior to recognition of the void in the idle loop may indicate which of the means to eliminate the void is preferable. For example, the use of HPVs may not be V desirable if no release to the containment has occurred DATE: 7-23-85 PAGE III.G-29
__ _ ~
BONP 20007 3 (9-84)
S AB COCK & wlLCOE NUCLEAR power OlvlS10N NUMatt 74-1152414-oo TECHNICAL DOCUMENT prior to the venting. If vents are used in this case and the vents are routed to the quench tank, it may be desir-able to avoid exceeding the capacity of the quench tank. In such a case bumping a RCP may be preferred. However, if a SBLOCA has occurred, venting will not drastically alter conditions in the containment. If the plant is recovering from a period of ICC, the nature of the void may not be known (i.e., it may be composed of non-cendensible gases) .
In that case venting the void is preferable to bumping a RCP.
The venting process for eliminating loop voids is similar to the RV head void elimination. After the vent is opened, MU or HPI should be increased to balance letdown and vent flow and maintain pressurizer level. This should maintain RC pressure. Hot leg temperature indications are not suffi-cient to confirm void elimination since the saturated level in a stagnant RC loop may be above the elevation of the hot leg RTD. Loop level indications (if available) may be used to determine if the void is becoming smaller. If the instrument tap is common with the HPV, the level indication may be inaccurate with the HPV open.
RCP bump or restart may also eliminate the loop void. The steam voia size can be approximated by the volume associ-ated with the pressurizer level increase that occurred during void formation. Loop level measurement may only indicate the approximate loop void volume and there may also be a RV head void present. Before bumping a RCP, primary
, pressure and pressurizer level should be restored (if necessary) to values indicated immediately following void formation. This step is necessary if the automatic pressur-izer level control (makeup and letdown) acted in response to the void formation. This will prevent the pressurizer from draining upon RCP bump. The RCP bump should mix cooler DATE: 7-23-85 PAGE III.G-30
BWNP-20007 3 (9 84)
SAtcoCK & wlLCoA
] NOCLEAR POWER DivlSION 74-1152414-oo TECHICAL DOCUMENT fluid with voids and cause voids to condense. RCS response to RCP bump with loop voids is discussed in Chapter IV. A.
System pressure control is more difficult during a RCP bump than it is during venting.
The idle loop void may also be condensed over a long period
, of time due to ambient heat losses. This method is slow and may not be feasible due to compressive tube-to-shell delta T ,
limits (i.e., shell cooldown rate > idle loop cooldown rate l
to ambient). {
3.8.3 ImDact of Steam Generator Tube Ruoture SGTRs may result in isolation of a SG. Opening the con- ;
denser dump valves (if available) will result in idle loop !
cooling. Periodic steaming of the isolated SG will also i promote natural circulation in this loop but also results in
! offsite releases. SGTR management criteria (Chapter III.E) i may preclude the use of the methods mentioned to eliminate ,
voids. [
t 3.9 HPI Coolinct [
HPI cooling is the least preferred cooldown method discussed I l in this chapter. Even with RCPs running, HPI cooling has l l
the following problems: l e RB contamination. f j
e Potential for recirculation from RB sump if long term l
- HPI cooling is required. l l e Primary cooldown rate and pressure control difficulty. l t >
i r l i In addition, HPI cooling without RCPs operating poses !
, RV head and idle loop cooling problems, as well as PTS l problems. Because of these concerns, HPI cooling should be f
used to cooldown the plant only if no other means exist. l i
l Primarily due to the PTS concern one RCP should be operated )
if possible during HPI cooling. To minimize heat addition I l
DATE: 7-23-85 PAGE III.G-31 j
l
t BWNP20007 3 (9 84)
SASCOCK & witCOM NUM8tt NUCLEAR POWER DIVISION 74-11s2414-oo TECHNICAL DOCUMENT to the primary system, it is desirable to operate no more than one RCP during HPI cooling.
HPI cooling requires one or more HPI pumps to inject coolant into the RCS, a source of borated water (BWST), and a flow path to the RB or quench tank.
The letdown flow paths to the reactor building are one or more of the following:
o PORV e Pressurizer Vent e HPVs The vent paths alone may not be of sufficient size for HPI heat removal at high decay heat rates. Specific actions to establish HPI cooling are provided in Chapter IV.B.
3.9.1 Transition to and Control of HPI Coolina It is preferable to maintain SCM during the transition to and subsequent control of HPI cooling. Tb.is allows continu-ous operation of the RCPs which prevents formation of hot stagnant regions in the RCS. Running RCPs also alleviates the PTS concern associated with HPI cooling and enables better use of primary loop temperature measurements. The ability to maintain SCM depends upon the decay heat level and amount of SG cooling achieved post-trip. Lower decay heat levels are more conducive to maintaining SCM.
Pressurizer spray and/or the pressurizer vent or PORV can be used to slowly condense or vent the pressurizer staam bubble while HPI increases the pressurizer level. This may allow subsequent opening of the PORV without losing the SCM. Opening the PORV may still lead to a loss of SCM. If l
it doesn't, HPI should be throttled as necessary to limit l
the cooldown rate to within the normal limit while maintain-l DATE: 7-23-8s PAGE III.G-32 1
. . _ , - . . . . . _ ~ . - - - _ . . - . . - _ . .. . . . ~ . ~ . - - - . . - - . - - . . . . . - . . - . . _ . - . _ . . - -
l
. 8YtNP-2OOO7 3 (9-84)
SASCOCK & WILCOX NUMStB NUCLEAR POWER OlvlS10N 74-1152414-oo
] ' TECHNICAL DOCUMENT l
ing SCM. Once SCM is regained, HPI flow rate and the RCS vent paths may be adjusted to control the cooldown rate as j
} long as PTS and the SCM limit are not violated.
During the cooldown, the following additional plant limits should be monitored and observed (if possible) . The incore T/cs should be used in ensuring adequate core cooling and i monitoring these limits since the loop temperatures may be t
- significantly different (especially without RCPs). ,
, e Tube-to-shell Delta Ts.
i If SCM is being maintained and a tensile tube-to-shell delta l T limit is approached, the operator may use methods des-
- cribed in Section 3.~6 to enhance shell cooling. As the primary heat input decreases, it may be possible to use the l pressurizer vents, HPVs and/or letdown instead of the PORV
! to conticol RC cooldown rate. In addition, since the RC is water solid, these vent paths can be used in conjunction j with HPI to control RC pressure.
1 3.9.2 Restoration of Steam Generator Coolina ,
l If heat transfer.can be restored during the cooldown in one l or both SGs, the transition from HPI to SG cooling should be j made. The RC temperature distribution following HPI I
cooling, if all RCPs are off, may make it difficult to restore natural circulation. Bumping a RCP may help. SG
, inventories should be restored to their appropriate levels l
3 (forced or natural circulation). Commence steaming the '
j SGs. As cooldown rate and SCM increase, slowly decrease HPI flow to control' the cooldown while maintaining SCM and RCS l l
inventory. Close the PORV and/or vents after heat transfer ;
-has been established through the SGs. Throttle HPI and i
open letdown as the PORV and/or vents are closed to prevent l
DATE: 7-23-85 PAGE III.G-33 i, _ _ _ _ _ _ _ _ _ _ - - - - - . . . . _ - - . - _ , _ . . - . _ _ _ _ . , _ _ . - . - _ . _ - , , , _ . - - . . - , , _ . . .
BWNP 20007 3 (9 84)
B ABCOCK & wlLCOE NUMBER NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT exceeding RCS pressure / temperature limits. Re-establish a pressurizer bubble if possible. If a bubble can't be established use alternate primary pressure control as described in 3.2. Continue the cooldown to DHRS conditions using SG cooling.
NOTE: Also see discussion of HPI operation in Chapter IV.B.
3.9.3 Impact of Steam Generator Tube Ruoture SGTRs result in a depletion of primary inventory until the leak flow is stopped. Therefore, it is desirable to restore heat transfer to the SG to enable an expeditious cooldown or reduce RC pressure by throttling HPI to minimize loss of coolant. Inventory may also be preserved if the SG drain discharge is recoverable for reinjection to the RCS.
In addition, HPI cooling with a SGTR may impose RC pressure control problems. Refer to Chapter III.E for a detailed discussion.
3.10 DHRS Transition When the RCS is cooled and depressurized to the DHRS initiation temperature and pressure, start flow through the decay heat cooler bypass valves and then slowly initiate decay heat cooler flow to prevent thermal shock to the RV.
To reduce the low pressure operating time of the RCPs and improve RCP seal life, turn off the RCPs during DHRS operation. If a faster cooldown is required, the RCPs can be run if the operating limits are met. The RCPs operated l should be those which provide the best combination of
! pressurizer spray and pressure range for DHRS operation. In
, each case, the objective in the cooldown is to decrease the 1
I entire RCS temperature, including the RV head and RCS loops.
If there is evidence of insufficient cooling in the RCS l
loops, steam the SGs concurrently with operation of the DATE: 7-23-85 PAGE III.G-34
B'f!NP-20007 3 (9 84)
B A B COCK & wiLCox NUMats NUCLEAR POWER Divi $lON 7_ '
74-1152414-00
(' ) TECHNICAL DOCUMENT i
y,/
DHRS. Keep cooling via the DHRS (DHRS return temperature equal to T c) while steaming the SGs to prevent interrupting natural circulation. When the TBVs are wide open, increase DHRS cooling. Depressurize at this time using auxiliary spray, if available. Reset the pressurizer level control frequently to slowly increase pressurizer level e.fter starting the DHRS so that the auxiliary spray flow does not cause a pressurizer outsurge during the remainder of the depressurization. Control the DHRS cooldown and depressuri-zation to prevent void formation in the RCS loops. RV head cooling and RCS loop and RV head void elimination methods are used as necessary while in the DHRS mode. Note that RV head and RC loop level instrumentation may not be useful during forced flow conditions.
to DHRS
,[~} 3.10.1 Transition From HPI Coolinct
/ When the RCS pressure and temperature during the LPI/HPI operation permit DHRS operation, align one train of the LPI for DHRS operation, leaving the other LPI train aligned for LPI/HPI cooling. Throttle HPI while maintaining RCS subcooling. When normal flow in the DHRS mode is estab-lished, close the PORV, and/or vents, stop LPI/HPI and establish MU centrol.
Balance MU and letdown to control RCS pressure with the pressurizer solid. Energize pressurizer heaters to form a bubble in the pressurizer or continue solid plant operation until RCS pressure temperature conditions permit injecting a nitrogen bubble into the pressurizer. Cooldown methods after the DHRS is in operation are the same as those described in Section 3.10.
79
,,r DATE: 7-23-85 PAGE III.G-35
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DOCU _
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PAGn PU _ LED i
- A\ O . sso mo w NO. OF PAGES !
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O HARD COPY FILED AT: PDR CF OTHER O BETTER COPY REQUESTED ON /_ /
. O PAGE 100 LARGE TO RLM.
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2 FILMED ON APERTURE CARD NO. 3509/90 Ve/-d.s
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\
J Figure III.G-2 PLANT STATUS VS C00LD0WN CONCERNS C00LD0ts CONCERNS PW GENEPATOR PRE 55U212Et C00LDCe SOLIO PLANT RV 10LE ST ATU S STATUS ST ATU S M)DE PL AN T $N ELL PR ES5URE HEAD LOOP d RCS LOOP FLOW LIMtif C00 Ling CC# TROL COOLING C00L lu G I
BUBBLE FotCEO y g 3g, 0 guesLE FORCEO I 1 0
- 0. E ,0
...E ,0 . _ . .
NO OUB8LE F0 tCED I I I l
t LOOP M gg, .8KE NATURAL I I 2 LOOP Putes no eussLE asyn,gt x x x 0FF l LOOP usett NATURAL X X X X onE u 1 LOOP no susett uATURAL X X I I I 1
PUv's On no 56 pg,g , , ,
j (nriC00 Ling) no suseLE
- 8* u0nE x x x DOC NO. 74-1152414-00 J
__..m._ __ . _ _ _ _ . _ . . _ _. . . . . . - _ . _ _ . _ _ _ _ . _ . . . -_ . . _ _ . . . _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ . _ _ _ _ . _ _ . . . _ _ . . .
1 t
e !-
. I i l I
i I ;
Figure Ill.G-3 HEAD FLUID TEMPERATURE RESPONSE WHILE ACTIVE VENT j IS OPEN (PRIMARY PRESSURE - 2200 PSI A) l;
'0 i i i i i
.l i
i 0.8 -
a J
I ,
i
= 3. 4 f x 104 FT2 0.6 - -g '
i-6g T hf = 600*F
- c j O ,
- 0. 4 _ A = Ven t A re a .
T hf = 500*F
] } g T hf = Head Fluid Temperature Prior to Venting A = 7.67 x10~4 FT2 a _
> c T; = Incore Teeperature
. 0.2 -
Thf(t) = Head Fluid Temperature as a T hf = 600'F Function of Time T* g = m *F 0.0 i e i e i 10 20 30 40 50 60 {
Time, minutes .
t s
l D0C.po. 74-1152414-00 l
i l
I t
l l
6 4
\
i 4
I l
l I
i i
2 T
l i
1
- Figure til.G-4 HEAD FLUID TEMPER ATURE RESPON3E WHILE ACTIVE VENT is OPEN (PRIMARY PRESSURE - 1600 PSI A) 1.0
, , i i 1 i
i -
i
)
! 0.8 - ~
j S T'hf = 600*F
! N I s -A = 3.41x10'4 FT2-N 0.6 - ~
T*y = 500'F j
8
-- T'hf = 400*F
! 3 '
A = Vent Area j
T*h t = 600*F
- ~
, T hf = Head Fluid Temperature O Prior to Venting a L- A = 7. 67 x 10-4 FT2<
T; = Incore Temperature 0.2 -
T hf = 500*F Thf(t) = Hesd Fluid Temperature as a T* g = 400'F s i Function of Time 0.0 e i i e 1 10 20 30 40 50 60 Time, minu tes j
00C. No. 74 1152414-00 I
l i
l
. . . ~_. _ _ _ . _ _ _ _ _ _ _ . . _ _ _ . . _ . _ _ _ - _ _ . . _ . . _ _ _ . . . _ _ . . _ _ _ _ _ _ . _ _ _ . . _ _ - . _ . . _ _ - _ _ _ _ - _ - - _ _ . . _ .
i i
l I !
i
! i
) '
lI i :
1 i
a 1
4 i
- 1 1
1 1
I i
I l Figure 111. G-5 HEAD FLUID TEMPERATURE RESPONSE WHILE ACTIVE VENT l 15 OPEN (PRIMARY PRESSURE - 1000 PSI A)
'0 i i i i j i 0.8 - -. A = 3. 4 t x 10*" FT2 l
! I'g = 500'F !
~
0.6 -
4 -
T'h t = 300* F
- i' 3 ' A = Vent Area A = 7.67x10-4 FT 2
- I 0*4 - ~
4 # *e- T',9 = Head Fluia Temperature T'h f = 500*F j u Prior to Venting ,
o -
Ty = Incore Temperature T* g = 300'F l 0.2 -
l Thf(t) = Heas Flula Teeperature as a f
! Function of Time i
j ;
1 1
0.0 e i i I I 10 20 30 40 50 60 Time, minutes l
! 00C. NO. 74-1152414-00 i t I
l 1 !
l i
i
} ,
i 1
1
,n--r-r-,-. em --wn,m- -n- w-,w.,.- r--.w, .wm>., ,-,----w ..e. -_, _ _n ,,,m wy w nww.w-m,ww--
. . ~ _ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ . _ . . _ _ . _ . . . _ . _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ . .
e l
4 '
i i
i l
i i
t I !
f I 4
i Figure Ill.G-6 HEAD FLUID TEMPERATURE RESPONSE WHILE ACTIVE VENT 13 OPEN (PRIMARY PRES $URE - 400 PSI A)
'0 i i
{ i i i i 1
4
- A = 3.41m10*4 FT2 !
l i ,
i 0.8 - -
T*hf = 400*F [
T'hf a 300*F !
0.6 - ~
j g i T hf = 400*F I 1 C A = Vent Area l E .
- T ,, - 300 r j .
- 0. 4 - T*n g , = Head Fluia Temperature -
, e5 7- Prior to Venting i #
- rA = 7.67x10*4 FT2
- Tg = Incore Temperature 1
) 0. 2
- Ihf(t) = Head Fluid Temperature as a _
,1 Function of Time
}
i l !
1 0.0 l i I e i
! 10 20 30 40 50 60 l
! line, minutes !
, t
! i I
D0C. H0. 74-1152414-00 i
i I
I l i
I I
i I
t 1
I '
1 l
b i
4 SWNP 20007 3 (9-84) j I
] BADCOCK & WitCOM NUMBER
( NUCLEAR POWER DivtSION 74-1152414-oo l TECMICAL DOCUMEllT t
i i
i i
i i
I i l l
i l i i :
l l
1 l
! l Part IV l Eauinnent Oneration f i
2 i
j s
4 -
{
I I I I
l DATE: 7-23-85 PAGE IV.0
i BWNP 20007 3 (9 84)
BASCOCK & wlLCOK NUCLEAR POWER DIVISION 74-1152414-00 OTECHICAL DOCUMENT
/
4 Chanter IV.A Reactor Coolant Pumpg i
1.O INTRODUCTION The following guidelines for RCP operation provide guidance on when to trip, restart or bump RCPs. Unless superceded by guidance in this document, normal RCP limits and precau-tions always apply.
2.0 RC PUMP TRIP GUIDELINES i The following guidelines state when and how many RCPs should be tripped.
2.1 Loss of Subcoolino Marcin When the adequate SCM is lost, all RCPs must be tripped immediately. Because of the importance of tripping RCPs immediately upon loss of SCM, this action must be followed immediately, regardless of other circumstances. However, i
there is one exception to this. If the RCPs are not tripped immediately when the SCM is lost, (i.e., within two minutes) then reduce the number of operating RCPs to one in each 13op.
The reason for tripping all RCPs is as follows.
- r Analysis has determined that an early trip of RCPs is required to prevent possible core damage. This analysis used conservative Appendix K assumptions with the objective of meeting the requirements of 10CFR50.46. Using conserva-tive Appendix K assumptions, it was shown that RCPs must be tripped within two minutes after losing SCM to prevent the RCS from evolving to a high enough void faction such G that the core would be uncovered if the RCPs were tripped at l
[' a later time. This analysis showed that continued RCP l DATE: 7-23-85 PAGE IV.A-1 i . _ _ _
BCNP 20007 3 (9-84)
B ABCOCK & WILCOX NUCLEAR POWER DIVISION NUMBER 74-1152414-oo TECHNICAL DOCUMENT operation could allow the RCS to evolve to a void fraction of 70% or greater if a certain range of break sizes were present. If the RCPs were tripped when the void fraction was 70% or greater, core uncovery would occur. Since RCP trip later in time cannot be absolutely prevented, it is necessary to trip RCPs before the RCS void fraction could increase to 70%. Once RCPs are tripped, the rate of loss of RCS inventory is reduced to the point where HPI (along with SGs in some cases) can keep the core covered.
2.2 Exceptions to RCP Trio Criterion It is important to trip RCPs to minimize RC inventory loss as discussed in Section 2.1 and there are times when the RCPs would normally be tripped to prevent mechanical damage. However, there are two cases when the RCPs must not be tripped: (1) If the RCPs were not tripped immediately upon the loss of SCM, then they must be operated even though-RCP damage is possible; (2) If severe ICC conditions exist, the RCPs must be operated in an attempt to restore adequate core cooling.
If all the RCPs are not tripped immediately upon loss of SCM (i.e., within two minutes) then it is possible, for a certain size small break, for a high void fraction to evolve in the RCS after that time. Tripping all the pumps at this later point in time could cause core uncovery and ICC.
Consequently, the operator must make sure that cooling water i and seal injection are supplied to the running RCPs to l
l prevent RCP damage. These services must be maintained for several hours. To prevent mechanical damage to all the RCPs only two RCPs, one in each loop, should be operated.
If they fail, the two pumps which were idle should be started even if mechanical damage is again likely.
l 1
l DATE: 7-23-85 PAGE IV.A-2
I BWNP-20007 3 (9-84)
BAtcOCK & witCOR NUMBER NUCLEAR POWER DIVISION 74-1152414-0o i , TECHNICAL DOCUMENT If severe ICC conditions exist, the need to restore core cooling takes precedence over damage to the RCPs as dis- i cussed in Sections 3.5 and 3.6.
2.3 HPI Coolinc During HPI cooling the number of operating RCPs should be reduced to one. This will reduce the amount of RCP heat input to the RC. The operating RCP will provide thermal and chemical mixing of the RC. This is especially important for PTS limits. Refer to Chapter IV.G for a discussion of PTS limits. Operating the RCP allows normal RC loop T cold to be used for maintaining RCS pressure temperature limits.
3.0 RCP RESTART OR BUMP There are several conditions where it may be beneficial to Q restart or bump RCPs. RCP operation provides the following:
A. Helps to promote thermal and chemical mixing and better temperature indication.
j B. Helps to restart natural circulation (mitigate voiding in hot. legs).
C. Restarts in the right combinations may restore pressur-izer spray which enables better control of RCS pressure.
D. Allows faster cooldown.
E. Helps to restore core cooling during ICC conditions.
i RCP operation is especially important for PTS concerns.
Refer to Chapter IV.G for a discussion on maintaining the PTS limit.
If a SGTR has occurred, then restarting RCPs provides optimum control over the RCS cooldown. Refer to Chapter O III.E for a detailed discussion.
7-23-85 PAGE IV.A-3 DATE:
C'.YNP 20007 3 (9 84)
BASCOCK & WILCOX NUMBER NUCLEAR POWER OtVISION 74-1152414-00 TECHNICAL DOCUMENT The decision of whether or not to restart or bump RCPs depends upon the condition of the RCS. If the decision is made to restart or bump RCPs, then the required steps again depend upon the condition of the RCS. Eight such RCS conditions have been identified. When and how to restart or bump the RCPs will be discussed for each RCS condition as follows:
3.1 SCM with natural circulation existina As long as the RCS has the minimum adequate SCM vith natural circulation existing then the RCPs may be started and run.
When the first RCP is started, the RC pressure and pressur-izer level should decrease as Th approaches T c. As each subsequent RCP is started, one at a time, a slight pressure spike may occur in the RCS. This pressure spike will be around 20 to 50 psi and is noticeable in RCS pressure as well as pressurizer level. However, RCS pressure should stabilize quickly at the previous value.
3.2 SCM with no natural circulation In this case a lack of primary to secondary heat transfer exists, which is due to either a lack of SG heat sink or steam voids in the hot legs blocking natural circulation.
If the SG heat sink exists, i.e., FW flow exists to at least one SG and the SG temperature is below the RC temperature, then the RCP can be restarted.
Restarting a RCP should cause any steam voids which have collected in the hot legs to be circulated into the SG where they will be condensed on the inside surface of the SG tubes. This condensation of the hot leg voids is expected to cause a pressure drop in the RCS. The magnitude of this-pressure drop is dependent upon the size of the steam void.
DATE: 7-23-85 PAGE IV.A-4 l
. m_ ___.
b F
l BWNP-20007 3 (9 84) i
+
SABCOCK & WILCOE <
NUmStB
~ NUCLEAR POWER DIVIsl0N 74-1152414-00 TECHNICAL DOCUMENT ;
l(
1 .
.The-pressure drop may be as much as 500 to 600 psig. If l j possible, prior to RCP restart the RCS pressure and tempera-ture should be established such that there is a 600 psig difference between.the existing pressure and the saturation ;
} pressure for-the existing temperature. This may not be possible if the PTS limit is in effect. . Analysis for a l 177 FA plant has shown that condensation of voids upon l l- restart of RCPs causes a pressure drop of 70 to 120 PSI a
- with only 70 cubic feet of steam trapped in the top of ;
l.
the hot legs. This same analysis also showed that pressur- ]
! izer level initially decreased 4 to 8 feet. Therefore, the i J :
j pressurizer level should.be increased to reduce the possi- j i bility of uncovering the pressurizer heaters. At this !'
I '
4 point, however, it is still beneficial to restart a RCP even l if it is expected that the RCS will depressurize to a saturated condition. Condensing the steam voids will reduce t l RC pressure and allow more HPI flow into the RCS. As soon
4 i f The RCP may be started and left running if the RCS SCM is {
! not lost. ;
, t i '
, During HPI cooling, one RCP should be started. Starting one l
RCP will promote. fluid mixing and reduce the thermal stress }
!- on the RV. However, the PTS limit must still be observed.
~
! 3.3 The RCS is saturated with desired cooldown rate
- As long as the RCS is saturated with suf ficient natural !
circulation, boiler condenser cooling, or. break-HPI flow to
. provide . the - desired cooldown rate the RCPs should not be restarted.
I DATE: 7-23-85 ~ PAGE IV.A !
.2._____ - _ . _ ..._. . _ _ . _ . _ , . _ . , _ . . _ _ _ _ _ . _ _ _ . _ _ _ _ . _ . _ _ . _ _ _ . _ _ .
BWNP-20007 3 (9-84)
B ABCOCK & WitCOR NUmstt NUCLEAR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT 3.4 The RCS is saturated without desired cooldown rate
( If FW flow is available and one SG is available as a heat sink then steam voiding probably exists in the hot legs such that natural circulation is prevented. RCP bumps may be used to restart natural circulation. Bumping an RCP means to start the RCP, wait until the starting current drops off (about 10 seconds), and then trip it. The starting current should drop back to the run current before tripping the pump breaker to prevent possibly damaging the breaker. One available RCP should be bumped every 15 minutes but do not exceed RCP motor starting duty limits. The 15 minutes should be adequate time for natural circulation to develop after a RCP bump.
If saturated natural circulation is initiated, then, the RCS cooldown can be performed by gradually lowering steam pressure along with RCS pressure. If natural circulation is not initiated, the SG pressure should be lowered again until the SG temperature is 100F colder than the incore T/C temperature. RCP bumps should continue until HPI cooling or primary to secondary heat transfer is established.
Bumping a RCP in this condition will have the same effect on RCS pressure as previously mentioned under Section 3.2.
If steam voids exist and are condensed in the SG, then a rather large pressure drop in the RCS is expected. With l
continued HPI flow the RCS will repressurize. Immediately after each RCP bump, the SG pressure should be lowered again, if necessary, to maintain the 100F differential between secondary T s at and the incore T/C temperature. As soon as natural circulation b, Tins, SG pressure should be controlled as necessary to maintain the desired cooldown rate.
O DATE: 7-23-85 PAGE IV.A-6
ECNP-20007 3 (9 44)
BABCOCK & WILCOX NUMSER NUCLEAR POWER Olvisl0N 74-11s2414-00 (9
\j TECHNICAL DOCUMENT 3.5 RCS is suoerheated with clad temoerature creater than 1400F At temperatures between 1400F and 1800F some clad oxidation is expected. This ICC condition is severe enough that it justifies starting and running one RCP in each loop in an attempt to restore adequate core cooling. Therefore, start and run one RCP in each loop, provided the RCP operating limits permit RCP operation. Refer to Chapter III.F ,
for additional discussion.
3.6 RCS suoerheated with clad temoerature creater than 1800F Severe clad oxidation will begin at temperatures greater than 1800F. All RCP interlocks may be bypassed with the exception of the motor electrical protection trips. All RCPs should be started and run even if pump damage can occur. This action may move any water from all four cold i .
leg pipes and SGs to the RV. The RCP motor electrical protection trips should not be bypassed, because if these faults exist then in general RCP operation would not last long enough to make a difference to core cooling. Refer to Chapter III.F for additional discussion.
3.7 RCS subcooled with indications of RV head voids If RCS pressure does not respond to pressurizar actions after verifying subcooled natural circulation, then it should be assumed that an RV head void exists. The creation of an RV head void would have been verified by a sudden increase in pressurizer or MU tank level while lowering RCS pressure during cooldown. If the RCS pressure had been decreased quickly below the saturation pressure for the initial temperature in the RV head, then the presence of an RV head void is likely. ,
!O Once the presence of a RV head void has been determined, it is recommended to use the RV head vent to purge the steam 7-23-85 PAGE 1Y'A~7 DATE:
BWNP-20007 3 (9 84)
B ABCOCK & WILC01 ,
NUM8tt l NUCLEAR PCWER DIVI $10N 74-11s2414-00 TECHNICAL DOCUMENT from the RV head and provide some cooling to the RV head.
However, if RV head vents are unavailable, or it is not desirable to open the RV head vent due to discharging steam to the RB, then a RCP should be started to remove the RV head bubble. Before starting the RCP, the pressurizer level should be increased to about 75% full scale and the SCM increased to about 70F subcooled. After increasing pressur-izer level, the pressurizer should be allowed to come to saturated conditions before starting the RCP.
After the SCM and pressurizer level increase and the RCS reaches equilibrium, then one RCP should be restarted after initiating MU or HPI. The RCP restart should not cause a loss of SCM. However, if adequate SCM is lost, the RCP must be tripped and full HPI flow should be established to restore SCM. A sudden RCS pressure decrease of 200 psi or more should be expected upon restarting a RCP with a void in the upper RV head region. This sudden RCS depressuriza-tion is due to the steam condensing under the RV head as cooler water is forced into the RV head region. Even if it is expected that attempting to run a RCP will cause the RCS to depressurize to a saturated condition, it is still desirable to start a RCP to regain RCS pressure control.
Condensing the steam voids will allow more HPI flow into the RCS, and will reduce the size of the steam void. HPI flow eventually will again restore SCM. Once the SCM is re-stored, another RCP may be restarted after reestablishing the previously required conditions. A few repetitions of this process should condense all of the steam and the RCS will remain subcooled and the RCP may be left running. Each RCP should not be started more frequently than motor starting limits allow.
O 7-23-85 PAGE IV.A-8 DATE:
0";NP 2000T3 (9 84)
B ASCOCK & wlLCOX
,- NUCLE AR POWER DivtSich NUMBER 74-1152414-00
( ; TECHNICAL DOCUMENT 3.8 RCS Water Solid When a RCP is started with the SG(s) available, Thot and T
cold will approach the SG saturation temperature. This will cause the RC pressure to either increase or decrease depending on whether the weighted average RC temperature was higher or lower than the SG saturation temperature prior to starting the RCP. The magnitude of the pressure change will depend on the magnitude of the change in the weighted average RC temperature. The rate of RCS pressure change is dependent on the rate of heat transfer to the SG(s) . This depends on the SG water inventory. The rate is faster for larger SG water inventories.
For example, when a RCP is started while in natural circula-tion, T hot will decrease toward T cold. Tcold "ill # ""1" i
/7 t slightly above the SG saturation temperature. The resulting y/ decrease in the weighted average RC temperature will cause the RC pressure to decrease. Pressurizer pressure will decrease approximately 100 to 200 psi if a RCP is started in a 177 raised loop plant while in natural circulation with 4%
decay heat and SG level at the natural circulation setpoint (this total pressure change includes the pressure increase due to the pump head). For this case, the pressure decrease will take about 11 seconds.
The RC pressure should be adjusted prior to starting a RCP to avoid opening the pressurizer relief valves, if the RC pressure increases, or to avoid losing the SCM (or RCP NPSH) if the RC pressure decreases. When in natural circulation, the RC pressure should be increased at least 100 pai above the SCM limits before starting a RCP.
[] During natural circulation the initial weighted RC average
) temperature can be determined and appropriate adjustments DATE: 7-23-85 PAGE IV.A-9
B;NP 20007 3 (9 84)
BABCOCK & WILCOX NUM4tR NUCLEAR POWER Divi $lON 74-1152414-oo TECHNICAL DOCUMENT can be made in RC pressure before starting a RCP. However, following idle loop operation or HPI cooling, determining the weighted average RC temperature can be more difficult because of more extreme variations in RC temperature distributions. This makes measuring and averaging the temperature difficult. These modes of operation can also cause the weighted average RC temperature to be signifi-cantly different from the SG saturation temperature which can cause significant pressure changes when a RCP is started. Consequently, making adjustments in RC pressure prior to RCP restart requires more consideration when recovering from idle loop operation or HPI cooling.
If a RCP is started without a SG available, heat transfer to the SG will not occur. Consequently, the weighted average RC temperature will change very little since the resulting RC pressure changes will be due only to the RCP head.
O DATE: 7-23-85 PAGE IV.A-lo
81'NP 200071 (9 84)
SABCOCK & WILCOX Numtt8 NUCLEAR POWER DIVl$10N gTECMCR DOCUMW Chaoter IV.B HPI/LPI/DNRS/CF ODeration
1.0 INTRODUCTION
This chapter discusses the technical bases for the guidelines that use the following systems for transient mitigation:
A. High Pressure Injection System (HPI) which is dis-cussed in Subsections 2.A and 2.B (Note: due to significant design differences, two subsections are provided; 2.A is generic except Davis-Besse and 2.B is Davis-Besse specific).
B. Low Pressure Injection System (LPI) which is discussed in Subsection 3.0.
C. Decay Heat Removal System (DHRS) which is discussed in Subsection 4.0.
D. Core Flood Tank System (CFT) which is discussed in Subsection 5.0.
In addition to discussing technical bases associated with specific systems, this chapter also discusses some general technical bases which involve more than one of these systems. These technical bases include the following:
E. Boron precipitation which is described in Subsection 6.0.
F. HPI/LPI piggyback operation which is described in Subsection 7.0.
2.A HPI SYSTEM OPERATION (for all plants except Davis-Besse)
The HPI system is used to:
I A. Makeup for lost RCS inventory due to a LOCA.
B. Makeup for RC contraction due to excessive cooling of 4
the RCS.
C. Provide a backup method of core cooling when the SGs do not provide adequate heat transfer.
D. Provide boration of RCS.
3~'#
- DATE: PAGE l
I 1
LONP 20007 3 (9 84) 8ABCOCK & WILCOX NUMtta NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT The HPI flow rate to the RCS should be started, maximized, throttled or stopped depending on certain existing conditions. The following sections will discuss when, how and why each of these actions are to be made. In addi-tion, each action will be stated as mandatory (must) or desirable (should}. If a conflict exists between a mandatory and a desirable action, the mandatory action takes priority.
2.A.1 Definitions The actions to control HPI flowrate are based on certain conditions. Some of these conditions need to be defined in order to correctly understand when and how to control the HPI flow. These conditions are defined below:
A. Loss of Adeauate Subcoolina Marcin The adequate subcooling margin (SCM) is considered lost whenever the RC pressure and temperature rela-tionship is below the RC SCM limit as measured at the location or locations of concern.
B. Adecuate Subcoolina Marcin Exists The adequate SCM exists whenever the RC pressure and temperature relationship is equal to or more subcooled than the SCM limit as measured at the location or locations of concern.
C. Normal Makeup Capacity Normal MU capacity is defined as the maximum expected water addition to the RCS through the MU line with the letdown line isolated. This amount will vary with RC pressure.
D. Feedwater Available to a Steam Generator FW is available if FW flow rate is adequate to meet all SG level and FW flow requirements.
7-23-85 PAGE IV.B-2 DATE:
! l
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B'INP-20C07 3 (9 84)
E sAncocx a witcox s NUMBR NUCLEAR POWER OlVtSION
(\ j) TECHNICAL DGCUMENT '
E. HPI Coolinc This is a method of removing core heat by adding low enthalpy HPI fluid (HPI/MU applies to Davis-Besse only) to the RCS while ~ removing high enthalpy fluid from the RCS to the RB. This method of cooling is used4when inadequate primary to secondary heat transfer exists. <
2.A.2 Initiatina HPI HPI is' initiated by starting two HPI pumps taking suction '
from the BWST'and pumping to the RCS through all available injection nozzles. HPI should be manually started if automatic initiation has not already occurred.
i 2.A.2.1 HPI MUST be INITIATED whenever loss of adecuate SCM occurs.
g When the SCM is lost, HPI flow to the RCS is required.
2.A.2.2 HPI flow SHOULD be ESTABLISHED durina reactor shutdown with a SGTR if normal MU cannot maintain desired pressur-izer level.
This action is taken as necessary to maintain a sufficient
. pressurizer water volume so that should a reactor trip j occur during the shutdown, the associated sudden decrease in the RC temperature will not cause a loss of adequate SCM due to the decrease in the pressurizer level.
) 2.A.2.3 When SG heat transfer is not adecuate and FW is not avail-able to either SG then HPI must be initiated. If two HPI oumo flow cannot be achieved then the PORV must be opened.
If FW cannot be supplied to either SG, then HPI if started soon enough, can provide adequate decay heat removal.
This has been determined by an analysis using 10CFR50 v Appendix K input assumptions. The analysis shows that
- ~ ~" PAGE DATE:
s E"'NP 20007 3 (9-84)
S ABCOCK & witCOM NUMBER NUCLEAR POWER OlvlSION 74-11'2414-TECHNICAL DOCUMENT during a complete loss of FW event, the core can be cooled meeting the criteria of 10CFR50.46 if two HPI pumps are started within twenty minutes and with mass and energy removal by the pressurizer safety valves (i.e. the pressurizer PORV remains closed and the RC pressure remains near the pressurizer safety valve setpoint.) An additional analysis using 1.0 ANS(1971) decay heat and the remaining 10CFR50 Appendix K input assumptions demon-strated that one HPI pump is sufficient to cool the core.
Consequently, the guidelines state HPI must be started if FW is not available. However, this action must be made soon enough (i.e. before too much RC inventory is lost).
The 20 minutes in the above analysis occurs some time after the RCS starts losing RC inventory out the pressur-izer safety valves. Therefore, the guidelines state to start HPI when or before the RCS pressure reaches the PORV open setpoint or the first automatic PORV lift since that is when RC inventory will start being lost. Keying the action to RC pressure also avoids the use of time as an action criteria.
In addition to starting HPI, the PORV may be opened so that HPI flow will match decay heat as soon as possible.
Opening the PORV will reduce RC pressure causing increased HPI flow. If the PORV is not opened the RCS may saturate and the RCS water level may decrease below the pressurizer surge line nozzle. With maximum decay heat, not opening the PORV will require a relatively long time before the flow from two HPI pumps matches decay heat (approximately l
6100 seconds) and the RCS water inventory starts to l
gradually refill. The SCM may not be restored until after the RCS is refilled to the PORV. If the PORV is not opened, HPI flow must not be throttled until incore T/Cs begin to cooldown and SCM exists. If the PORV is opened, DATE: PAGE
8WNP 20007 3 (9 84)
BASCOCK & wlLCOR Numttt NUCLEAR POWER DIVISION n -11s2414-oo DTECHNICAL DOCUMENT (V the SCM should be lost only momentarily and the RCS inventory will be maintained. If two HPI pump flow cannot be achieved then the PORV must be opened to provide the maximum HPI cooling possible.
Other concerns besides core cooling must be considered.
These concerns are discussed below:
- 1. Reactor Vessel P-T Limit The RV P-T limit is always applicable. However, since the plant has been cooling down, the limit becomes more restrictive. The RC pressure must not exceed the RV P-T limit. Therefore, if the RC pressure increases to the RV P-T limit, the PORV should be opened and HPI cooling started to limit the pressure increase. After the pressurizer fills with RC, the HPI flow should be throttled as necessary to try and keep the RC pressure below the RV P-T limit. Refer to Chapter IV.G for information on the RV P-T limit.
, before starting HPI.
Refer to Chapter IV.G for information on the PTS Limit.
i i
! 3. Steam Generator Tube Ruoture If a SGTR results in an isolated solid SG, RC pressure must be kept below the low MSSV setpoint if at all possible. With HPI cooling this is normally accom-v' plished by throttling HPI flow while maintaining
~
7-23-85 PAGE DATE:
1 BWNP 30007 3 (9 80)
B A BCOCK & WILCOX NU; LEAR POWER OlvlSION "
74-1152414-oo TECHNICAL DOCUMENT adequate SCM. If, however, the RCS is saturated, HPI must not be throttled. In this case, additional relief paths must be opened. These include HPVs, letdown, and drains on the isolated SG if available.
Refer to Chapter III.E for more details on HPI cooling with a SGTR.
2.A.3 Maximizina HPI Flowrate 2.A.3.1 Whenever the SCM is lost. maximum HPI flow MUST be provided to the RCS.
A. Maximum HPI flow is established to provide maximum core heat removal. HPI will provide heat removal from the core by continual addition of low enthalpy water to the RCS. This requires a concurrent break or opening in the RCS for removal of the high enthalpy RC to allow addition of the low enthalpy HPI.
B. Maximum HPI flow is required to restore the SCM as quickly as possible. As long as the SCM exists the core is assured of being covered making the core adequately cooled. Therefore, it is important to reestablish the SCM as quickly as possible.
C. Maximum HPI flow is required to provide subcooled RC for primary to secondary heat transfer. If the SGs are available for heat removal, then adding water to the RCS will replenish the heat transfer medium for primary to secondary heat transfer.
D. Maximum HPI is achieved by operating two HPI pumps and balancing the HPI flow. The intent of balancing the HPI flow is to address such failures as a break in the HPI injection line or a closed HPI injection valve.
O DATE: 7-23-85 PAGE IV.B-6
87NP 20007 3 (9 84)
SABCOCK & wlLCom NUM$tt NUCLEAR POWER DIVISION
" ~ 11"" 14 ~
OTECHNICAL DOCUMENT t These failures will cause imbalances in the HPI flow with the result that the HPI to the RV may not be as large as possible. For example, if an HPI line break exist, the broken line may have a much higher flow rate than in each of the unbroken lines. If the flow only in the broken line is throttled more flow will go through each of the other lines to the RCS and less HPI water will be lost out the broken line. The intent of balancing the flow is to increase the total flow reaching the RCS and not to try to make the flow 4
through each flow path exactly equal. Balancing may or may not be inherent in the HPI system design by use of crossconnected injection lines, venturi flow ,
nozzles, orifices, preset valve positions etc.
I E. When attempting to maximize the HPI flow rate the HPI t pump flow rate should not be allowed to exceed the I
i maximum allowed pump flow rate. The HPI system design may inherently prevent excessive pump flow.
F. When using Th or Tc as an indication of loss of SCM, the corresponding loop must have loop flow. This is to avoid requiring maximum HPI when adequate core cooling exists. This can occur in two situations.
First, during one loop natural circulation, the operating loop can provide adequate core cooling while the idle loop can saturate as the RC is depressurized .
- during the cooldown.
Second, during HPI cooling, both RCS loops can
! saturate while the HPI is providing adequate core i cooling as indicated by the incore T/Cs. l h
o ,
~
- DATE: PAGE
)
02NP 200013 (9-84) l SABCOCK & WILCOX NUMSt NUCLEAR POWER OlVISION TECHNICAL DOCUMENT 2.A.4 Throttlina HPI Flow Throttling means to reduce the HPI flow rate below the maximum flow rate. This can be done by regulating HPI flow valve positions and/or stopping an HPI pump and/or HPI recirculation flow. In general, HPI flow may be throttled anytime adequate SCM exists as indicated by the incore T/Cs. HPI flow must not be throttled when the RC SCM is lost.
When the HPI flow is throttled, the pump flow rate should not be throttled below the minimum allowed pump flow rate. The HPI recirculation lines may be available to provide the minimum required HPI flow rate. If the recirculation lines discharge to the MU tank, the MU tank inventory will need to be controlled. This is especially true if HPI/LPI " piggyback" operation is in progress.
When throttling HPI flow to control RC pressure, care should be taken not to allow the RC pressure to drop below:
A. The SCM limit.
B. RCP NPSH requirements if a RCP is operating.
C. Fuel pin in compression limit if in effect.
If a SGTR exists, special considerations are in effect.
Maintaining the primary to secondary system pressure differential as low as possible may be desirable to reduce primary to secondary leak rate. Refer to Chapter III.E.
2.A.4.1 The HPI flow SHOULD be THROTTLED to keep the Dressurizer level near the normal operatina level setooint when the SCM exists.
A. If the pressurizer level is too high the RCS is susceptible to rapid pressure increases which can cause undesirable opening of the pressurizer relief DATE: PAGE
...s_ . - -_ .- ____m _ _ - _ _ _ _ - .__---_._m _ _- .. _ _ _ _ _ . . _ _ . _. _ _ _ - _, _
O!NP 20007-3 (9 84)
SABCOCK & WILCOE NW4tt NUCLEAR POWER DtVISION
"~1* * **~
- ] TECHNICAL DOCUMENT valves and relief of two-phase and subcooled water.
I If the pressurizar water level is too low, the RCS is I susceptible to large, rapid decreases in pressure i which can cause a loss of SCM. Also, the level should ,
be high enough for pressurizer heater operation. The level should account for possible instrument errors 4 including elevated RB temperatures as applicable.
B. If a leak exists in the pressurizer, maintaining a i pressurizar level may not be possible and the pressur-
! izer will fill solid if the RC is being kept sub-cooled.
t 4 C. Throttling HPI for the reasons discussed is expected
! during a SBLOCA or after an overcooling transient; e.g., the SBIDCA can initially be larger than the MU
)
j system capacity causing the pressurizer level to l - drop. HPI will be started with a flow rate greater i
j than the SBIDCA leak rate causing the pressurizer to I refill. The SBIDCA can also initially be larger than i the HPI capacity. However, when the RC pressure
! decreases the HPI flow rate will increase causing the i
j pressurizer to fill.
,! D. If the pressurizar drains due to overcooling, HPI can l rapidly refill the pressurizer once the overcooling l stops. HPI flow will need to be reduced significantly or stopped.
E. Continued filling will cause unnecessary valve 4
operation and fluid release through the pressurizer relief valves. Too much fluid release can overfill i
f and overpressurize the pressurizar relief (or quench) i tank.
I i
_.m
BONP-20007 3 (9 80)
SA8 COCK & witCOX NUM8th NUCLEAR POWER DIVISION TECMCR DOCUMW F. If core cooling is provided by HPI cooling, the pressurizer level cannot be maintained. In this situation, HPI is throttled to prevent over pressur-izing the RCS and to limit the cooldown rate only if SCM is maintained.
2.A.4.2 HPI flow MUST be THROTTLED to Drevent overpressurizina the RCS when the SCM exists by keepina the RC pressure below the NDT or PTS limit as acolicable. (See Chapter IV.Gl A. If a pressurizer steam bubble exists rapid filling of the pressurizer and the resulting pressurizer steam bubble compression can cause the RC pressure to increase. Increasing the RC temperature will also cause the pressurizer pressure to increase.
B. If the pressurizer is full of water (water solid) either because the pressurizer steam bubble cannot be maintained (i.e., pressurizer heaters are inoperable or the pressurizer has a small leak) or HPI cooling is in progress, the RC pressure is increased when the HPI volume flow rate going into the RCS exceeds the RC volume flow rate leaving the RCS or when the RC temperature is increasing.
C. This action applies if the SCM exists as measured by the incore T/C. Th and Te can still indicate a loss of SCM. This accounts for the possibility of an area of the RCS not having the SCM established while the core is being adequately cooled.
l 2.A.5 Stoopina HPI Flow 2.A.5.1 HPI flow SHOULD be STOPPED and normal MU flow control l started if the RC leak rate or contraction rate is within l the normal makeuo flow capacity and an adeauate SCM exists.
/ -4 J -U !) 1V.U-1U
__ _ . _ . . _ . __ . ~ . .
9WNP-20007 3 (9-84) eascocx a wncox N 8 NUCLEAR POWER DIVISION _
TECHNICAL DOCUMENT l
l A. When stopping HPI while a leak exists in the RCS, the MU pumps must be able to take suction from the BWST. i However, if the BWST is empty, the MU pumps must be
- able to take auction from the LPI discharge.
B. Normal MU flow control is preferred because the system 4
provides automatic volume control.
1 C. The RC leak rates or contraction rate should be ;
verified to be within normal MU. flow capacity as indicated by pressurizer level.
j D. Normal letdown and use of the MU tank may also be started along with normal MU flow control depending i on existing plant conditions; e.g., if the RC radia-tion level is high then letdown should be isolated.
I 2.A.5.2 HPI SHOULD be STOPPED if LPI has been flowina in each line
- for 20 minutes at a rate in excess of the minimum recuired j
) LPI flow rate listed below.
1 Plant Min. rec. LPI flow 205 FA NSS 1500 gpm 177 FA NSS (Except ANO-1) 1000 gpm ANO-1 (2 LPI pumps on) 2630 gpm i AND-1 (1 LPI pump on) 3020 gpm (total)
A. This condition is applicable to a large LOCA when the RCS depressurizes enough to allow the LPI to flow into the RV. Since LPI will provide emergency injection at a much greater rate than the HPI, the HPI can be i
stopped.
i
~#" *"~11 DATE: PAGE
1 EWNP 20007-3 (9 84)
BABCOCK & WILCOX NUM8tt NUCLEAR POWER DIVISION TECHNICAL DOCUMENT l B. The 20 minute delay provides reasonable assurance that the primary system will not repressurize and result in a loss of LPI flow. The minimum required LPI flow rate is used to ensure that the injection flow can remove decay heat after the HPI is stopped.
C. The LPI flow rate is required in each line to assure that at least the minimum required LPI flow is reaching the RV in the event that a break exists in one of the LPI/CF lines which could prevent LPI water from reaching the RV through one of the lines.
D. Discontinuing HPI flow prevents the added operation of lining up the HPI suction from the BWST to the LPI discharge when the BWST becomes empty. It also prevents pumping RB emergency sump water through the HPI pumps. This limits the transporting of the RB sump water which can have higher than normal radiation and debris levels. The debris can also cause in-creased wear on the HPI pump seals.
E. This guideline has priority over the requirement to provide maximum HPI if the SCM is lost (Section
- 1. 3 .1) . Consequently, if the SCM is lost but the requirements of this guideline are met then HPI can be stopped.
2.B HPI/MU SYSTEMS OPERATION (Davis-Besse only)
The HPI/MU systems are used to:
A. Makeup for lost RCS inventory due to a LOCA.
B. Makeup for RC contraction due to excessive cooling of the RCS.
C. Provide a backup method of core cooling when the SGs do not provide adequate heat transfer.
D. Provide boration of RCS.
l DATE: PAGE
SWNP-200003 (9 84)
SASCOCK & WILCOX NUMBER NUCLEAR POWER DmSION 74-11s2414-oo f' TECHNICAL DOCUMENT :
\
- The HPI/MU flow rates to the RCS should be started, maximized, throttled or stopped depending on certain existing conditions. The following sections will discuss when, how and why each of these actions are to be made.
In addition, each action will be stated as mandatory (must) or desirable (should). If a conflict exists between a mandatory and a desirable action, the mandatory action takes priority.
2.B.1 Definitions The actions to control HPI/MU flowrates are based on certain conditions. Some of these conditions need to be defined in order to correctly understand when and how to ,
control the HPI/MU flow. These conditions are defined below:
- A. Loss of Adecuate Subcoolina Marcin The adequate subcooling margin (SCM) is considered i lost whenever the RC pressure and temperature rela-tionship is below the RC SCM limit as measured at the location or locations of concern.
l B. Adecuate Subcoolina Marcin Exists The adequate SCM exists whenever the RC pressure and j temperature relationship is equal to or more subcooled than the SCM limit as measured at the location or locations of concern.
l C. Normal Makeuo Canacity i
Normal MU capacity is defined as the maximum expected water addition to the RCS through the MU line with the letdown line isolated. This amount will vary with RC pressure.
D. Feedwater Available to a Steam Generator FW is available if FW flow rate is adequate to meet all SG level and FW flow requirements.
E. HPI/MU Coolina 7-23-85 PAGE IV.B-13 ;
DATE:
OWNP.20007 3 (9 84)
BABCOCK & WILCOX NUCLEAR POWER DIVISION NUMBER TECHNICAL DOCUMENT E. HPI/MU Coolina This is a method of removing core heat by adding low enthalpy HPI and/or MU fluid to the RCS while removing high enthalpy fluid from the RCS to the RB. This method of cooling is used when inadequate primary to secondary heat transfer exists.
2.B.2 Initiatina HPI/MU Initiation of full MU flow is accomplished by starting the second MU pump (two MU pumps operating), fully opening the MU control valve, and switching the MU pump suction to the BWST. Full MU flow is used when RCS pressure is greater than the 1650 psig SFAS trip setpoint or when necessary to supplement HPI flow.
Initiation of fall HPI flow is accomplished by verifying two HPI pumps start at the 1650 psig SFAS setpoint (or manual start if necessary) and verifying the injection valves are open (or manually opening if necessary) . In addition, under certain conditions initiation of HPI may involve lining up the HPI pumps in piggyback operation on two LPI pumps. Whenever HPI is in piggyback operation and LPI flow to the RCS begins, HPI must be realigned to the BWST if the BWST is still in use. This provides greater total injection flow and is required by ECCS analyses.
2.B.2.1 HPI/MU MUST be INITIATED Whenever Loss of Adecuate SCM Occurs.
When the SCM is lost, full MU flow to the RCS is required and, if RCS pressure is less than the 1650 psig SFAS setpoint, initiation of full HPI flow must be verified and, if necessary, accomplished manually. If the RCS saturates above 1650 psig, the PORV, HPVs, and pressurizer vent should be opened and two HPI pumps should be started in piggyback operation on two LPI pumps.
DATE: PAGE
i c NP.roooi 3 (9 a4) ;
SABCOCK & witCOM Numett -
NUCLEAR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT 2.B.2.2 When SG heat transfer is not adeauate and FW is not avail- !
i able to either SG then HPI/MU murt be initiated. I If FW cannot be supplied to either SG, then HPI/MU if !
started soon enough, can provide adequate decay heat f removal. This has been determined by analysis. The !
analysis shows that during a complete loss of FW event, j
! the core can be cooled if two MU pumps are started within {
l twenty minutes and with mass and energy removal by the {
PORV, HPVs, and pressurizar vent. Consequently, the l guidelines state full MU must be started if FW is not j
^
available. However, this action must be made soon enough (i.e. before too much RC inventory is lost). !
i Therefore, the guidelines state to start MU as soon as the f total I4FW is identified. f t
s In addition to starting MU, the PORV, HPVs, and pressur- [
izer vent must be opened and HPI started in piggyback f l
operation if incore T/C temperature (or T hot if a RCP is j operating) reaches 600F. Opening the PORV at 600F will j reduce RC pressure causing increased MU flow and the RCS to saturate at a pressure below the HPI shutoff head when operating in piggyback. If the PORV is not opened in time j to saturate the RCS well within the piggybacked HPI range i it will require a relatively long time before the flow from two MU pumps matches decay heat (approximately one hour) and the RCS water inventory starts to gradually refill. The SCM may not be restored until after the RCS is refilled to the PORV. With high decay heat, opening the PORV at 600F may still result in the RCS reheating and repressurizing along the saturation curve above the shutoff head of the HPI pumps while piggybacked (at about 623F). If this occurs, adequate core cooling will still b
a 7-23-85 PAGE IV.B-15 DATE:
C'::NP-20007-3 (9 84)
S ABCOCK & wlLCOR NUM88 NUCLEAR POWER DIVISION TECMCR DOCUMM be provided but it will again require a relatively long time to match decay heat. If two MU pump flow cannot be achieved, then the PORV, HPVs, and pressurizer vont must be opened and HPI started in piggyback operation to provide the maximum HPI cooling possible, i.e., do not wait until 600F.
Other concerns besides core cooling must be considered.
These concerns are discussed below:
- 1. Reactor Vessel P-T Limit The RV P-T limit is always applicable. However, since the plant has boon cooling down, the limit becomes more rostrictive. The RC pressure must not exceed the RV P-T limit. Thorofore, if the RC pressure increases to the RV P-T limit, the PORV should be opened and HPI cooling started to limit the pressure increase. After the prosaurizer fills with RC, the HPI flow should .50 throttled as necessary to try and koop the RC pressure below the RV P-T limit. Refer to Chapter IV.G for information on the RV P-T limit.
- 2. Steam Generator Tube Ruoturo If a SGTR results in an isolated solid SG, RC pressure must be kept below the low MSSV sotpoint if at all possible. With HPI cooling this is normally accom-plished by throttling HPI flow while maintaining adequate SCM. If, however, the RCS is saturated, HPI must not be throttled. In this caso, additional relief paths must be opened. Those include HPVs (if not already open), lotdown, and drains on the isolated SG if available. Refer to Chaptor III.E for moro details on HPI cooling with a SGTR.
O 7-23-05
- DATE: PAGE
. - . . - . . - _, - - _ _ . . . - - ~ - _ _ - _ - . . - - ...
1 8%NP 20007 3 (9 84)
BASCOCK & wlLCom NUatttR NUCLEAR POWER DIVISION 74-11s2414-oo TEClllllCAL BOCONElli i
2.B.3 Maximizina HPI Flowrate k 2.B.3.1 Whenever the SCM is lost. maximum HPI/MU flow MUST be f crovided to the RCS.
4 A. Maximum HPI/MU flow is established to provide maximum j
core heat removal. HPI/MU will provide heat removal from the core by continual addition of low enthalpy l
i water to the RCS. This requires a concurrent break or
- opening in the RCS for removal of the high enthalpy RC 1 to allow addition of the low enthalpy HPI/MU. ,
i B. Maximum HPI/MU flow is required to restore the SCM as ;
t quickly as possible. As long as the SCM exists the core is assured of being covered making the core 4
adequately cooled. Therefore, it is important to i I reestablish the SCM as quickly as possible.
q C. Maximum HPI/MU flow is required to provide subcooled RC for primary to secondary heat transfer. If the SGs are available for heat removal, then adding water to i the RCS will replenish the heat transfer medium for i primary to secondary heat transfer.
D. Maximum HPI/MU is achieved by operating two makeup pumps with suction from the BWST and the makeup
- control valve fully open and two HPI pumps and
! balancing the HPI flow. If RCS pressure is greater i
j than 1650 psig, the HPI pumps should be placed in
These failures will cause imbalances in the HPI flow J
]
with the result that the HPI to the RV may not be as l
! large as possible. For example, if an HPI line break i
7-23-83 PAGE IV.B-17
! DATE
_ . , . . , . . - _ _ _ , , . _ _ . . , . _ . _ _- ,-4-.,_,--,_,m ..-,.%.,_.-
. -. ,,-= ,..., ,v ,.,._.--,,,,,,.m.__,.,,,._,m.i.,-.v,e _%,, ~%wm,- e- w --,.,.-__.-w----
BINP-2OOO7 3 (9 84)
B ABCOCK & WILCOX NUMBtB NUCLEAR POwtR Olvi$l0N nCumCR DOCUMENT exist, the broken line may have a much higher flow rate than in each of the unbroken lines. If the flow only in the broken line is throttled more flow will go through each of the other lines to the RCS and loss HPI water will be lost out the broken line. The intent of balancing the flow is to increase the total flow reaching the RCS and not to try to make the flow through each flow path exactly equal. Balancing HPI flow is accomplished by adjusting the two lines on each pump such that the higher flow is less than 1.5 times the lower flow. Throttle only the high flow line and do not throttle it below the value on Figuro IV.B-1 (note: Figure IV.B-1 shows the throttling limit for HPI directly from the BWST, i.e., not while in piggyback operation).
E. When attempting to maximize the HPI flow rate the HPI pump flow rate should not be allowed to exceed the maximum allowed pump flow rate of 950 gpm. This should only be a concern during piggyback operation.
F. When using Th or Te as an indication of loss of SCM, the corresponding loop must have loop flow. This in to avoid requiring maximum HPI when adequato core cooling exists. This can occur in two situations.
First, during one loop natural circulation, tho operating loop can provido adequato coro cooling while the idle loop can naturato as the RC is depressurized during the cooldown.
Second, during HPI cooling, both RCS loops can saturate while the HPI is providing adequato core cooling as indicated by the incore T/cs.
O
~" "*
DATE: PAGE l
L
i I swNr.20oo7 3 (s e4) !
i sascoca a witcon MsEE !-
NUCLEAR P0wtR DMS60N 74-1152414-00 TECNNICAL BOCONENT !
r f 2.B.4 Throttlina HPI/MU Flow f j Throttling means to reduce the HPI/MU flow rate below the f maximum flow rate. This can be done by regulating HPI/MU
!, f i
flow valve positions and/or stopping an HPI or MU pump j and/or HPI recirculation' flow. In general, HPI/MU flow j j may be throttled anytime adequate SCM exists as indicated i by.the incore T/Cs. HPI/MU flow must not be throttled i when the RO SCM is lost.
1 l When the HPI flow is throttled, the pump flow rate should l
not be throttled below the minimum allowed pump flow
! rate of 35 gym when the pump recirculation valve is [
closed. The HPI recirculation lines may be available to (
j provide the minimum required HPI flow rate.
t j i When throttling HPI/MU flow to control RC pressure, care j i should be taken not to allow the RC pressure to drop l { '
1 below:
l I l
A. The SCM limit.
j B. RCP NPSH requirements if a RCP is operating.
l C. Fuel pin in compression limit if in effect.
! If a SGTR exists, special considerations are in effect.
Maintaining the primary to secondary system pressure i
differential as low as possible may be desirable to reduce primary to secondary leak rate. Refer to Chapter III.E. l f '
4 ,.
2.B.4.1 The HPI/MU flow BHOULD be THROTTLED to keen the Dressur-izer level near the normal operatina level setnoint when f the scM exists. ,
I A. If the pressuriser level is too high the RCS is !
I j susceptible to rapid pressure increases which can !
i j
cause undesirable opening of the pressurizer relief valves and relief of two-phase and subcooled water.
f
! I i
DATE: PAGE i- - - . - . - _ _ - --. - - - - _ .. - -. - _ _
I BWNP-20007 3 (9 84)
SABCOCK & wlLCO2 NUMbt NUCLEAR POWER OlvlSION TECHNICAL DOCUMENT If the pressurizer water level is too low, the RCS is -
susceptible to large, rapid decreases in pressure which can cause a loss of SCM. Also, the level should be high enough for pressurizer heater operation. The level should account for possible instrument errors including elevated RB temperatures as applicable.
B. If a leak exists in the pressurizer, =sintaining a pressurizer level may not be possible and the pressur-izer will fill solid if the RC is being kept sub-cooled.
C. Throttling HPI/MU for the reasons discussed is expected during a SBLOCA or after an overcooling transient; e.g., the SBLOCA can initially be larger than the MU system capacity causing the pressurizer level to drop. HPI will be started with a flow rate greater than the SBI4CA leak rate causing the pressur-izer to refill. The SBLOCA can also initially be larger than the combined HPI/MU capacity. However, when the RC pressure decreases the HPI/MU flow rate will increase causing the pressurizer to fill.
D. If the pressurizer drains due to overcooling, HPI/MU can rapidly refill the pressurizer once the over-cooling stops. HPI/MU flow will need to b9 reduced i
l significantly or HPI stopped.
l E. Continued filling will cause unnecessary valvo l operation and fluid release through the pressurizer relief valves. Too much fluid release can overfill and overpressurize the pressurizer relief (or quench) tank.
O 7-23-85 PAGE DATE:
9%NP-20007-3 (9-81)
BASCOCK & wiLCOX NU f NUCLEAR POWER DIVISION TECHNICAL DOCUMENT F. If core cooling is provided by HPI/MU cooling, the pressurizar level cannot be maintained. In this situation, HPI/MU is throttled to prevent over pressurizing the RCS and to limit the cooldown rate only if SCM is maintained.
2.B.4.2 HPI/MU flow MUST be THROTTLED to Drevent overoressurizina the RCS when the SCM exists by keeDina the RC Dressure below the NDT limit. (See Chanter IV.G)
A. If a pressurizer steam bubble exists rapid filling of the pressurizer and the resulting pressurizer i steam bubble compression can cause the RC pressure to increase. Increasing the RC temperature will also cause the pressurizer pressure to increase, s B. If the pressurizer is full of water (water solid) either because the pressurizer steam bubble cannot be
- maintained (i.e., pressurizer heaters are inoperable or the pressurizer has a small leak) or HPI/MU cooling is in progress, the RC pressure is increased when the HPI/MU volume flow rate going into the RCS exceeds the RC volume flow rate leaving the RCS or when the RC temperature is increasing.
C. This action applies if the SCM exists as measured by the incore T/C. Th and Tc can still indicate a loss of SCM. This accounts for the possibility of an area of the RCS not having the SCM established while the core is being adequately cooled.
2.B.5 Stocoina HPI Flow 2.B.5.1 HPI flow SHOULD be STOPPED and normal MU flow control started if the RC leak rate or contraction rate is within l the normal makeuo flow capacity and an adeauate SCM 3
\ exists.
DATE: PAGE
l BWNP 30007 3 (9 84)
SASCOCK & WILCOI NUCLEAR POWER DIVI $f0N NUMSER 74-11s2414-00 TECHNICAL DOCUMENT A. When stopping HPI while a leak exists in the RCS, the MU pumps must be able to take suction from the BWST.
B. Normal MU flow control is preferred because the system provides automatic volume control.
C. The RC leak rates or contraction rate should be verified to be within normal MU flow capacity as indicated by pressurizer level.
D. Normal letdown and use of the MU tank may also be started along with normal MU flow control depending on existing plant conditions; e.g., if the RC radia-tion level is high then letdown should be isolated.
2.B.5.2 HPI SHOULD be STOPPED if LPI has been flowina for 20 minutes at a rate in excess of 1000 com/line.
A. This condition is applicable to a large 14CA when the RCS depressurizes enough to allow the LPI to flow into the RV. Since LPI will provide emergency injection at a much greater rate than the HPI, the HPI can be stopped.
B. The 20 minute delay provides reasonable assurance that the primary system will not repressurize and result in a loss of LPI flow. The minimum required LPI flow rate is used to ensure that the injection flow can remove decay heat after the HPI is stopped.
C. The LPI flow rate is required in each line to assure that at least the minimum required LPI flow is reaching the RV in the event that a break exists in l
DATE: 7-23-85 PAGE IV.B-22
BWNP 20007-3 (9 82)
B ASCOCK & wlLCOE NUMBE NUCLEAR POWER D4VIS10N TECHNICAL DOCUMENT one of the LPI/CF lines which could prevent LPI water from reaching the RV through one of the lines.
D. Discontinuing HPI flow prevents the added operation of lining up the HPI suction from the BWST to the LPI discharge when the BWST becomes empty. It also prevents pumping RB emergency sump water through the HPI pumps. This limits the transporting of the RB sump water which can have higher than normal radiation and debris levels. The debris can also cause in-creased wear on the HPI pump seals.
E. This guideline has priority over the requirement to provide maximum HPI if the SCM is lost (Section 1.3.1). Consequently, if the SCM is lost but the requirements of this guideline are met then HPI can be stopped.
3.0 L.P.I SYSTEM OPERATION In general, the LPI system is used to:
A. Makeup for lost RCS inventory due to a LOCA.
B. Provida post LOCA core cooling.
C. Provide RB emergency sump water to the HPI suction.
l 3.1 LPI MUST be initiated whenever any of the LPI initiation I
setooints are reached.
The setpoints are those used by the safety features actua-tion system.
LPI is initiated by starting two LPI pumps taking suction from the BWST first, or from the RB emergency sump as applicable, and pumping to the RCS.
3.2 The LPI suction MUST be chanced from the BWST to the RB eneroency sumo when switchover conditions are met. The O' BWST water inventory remaining should provide sufficient
/-ea-as A v . 5-2.i
BINP 20007 3 (9 84)
SABCOCK a wnCOE NUMBtB NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT time to make a transition to the RB cmorgency sump to prevent losing LPI suction water, provido adequate LPI NPSH and prevent air entrainment in LPI flow.
3.3 Loss of LPI Recirculation Ability The ability to recirculate water from the RB emergency sump can be lost during recirculation or when trying to start recirculation. If this ability is lost, the cause should be determined and flow started as soon as pos-siblo. There are two general causes: 1) loss of sump water and 2) loss of both flow paths from the sump.
3.3.1 Loss of Sumo Water A SGTR can causo a loss of sump water - A SGTR is a LOCA; however, the water leaking from the RCS does not accum-ulate in the RB sump unless it is being drained from the SG to the sump. Instead, it leaks out of the RB via the secondary system while the SG is steamed and accumulatos in the SG after it is isolated. If the BWST is depleted and the time has come to transfer to the RB sump, thoro will be no sump water. This situation should not occur if the SG with the SGTR is allowed to fill and the tubo leak flow is terminated or if the SG is drained to the sump.
However, if the SG must be steamed, or otherwiso the leak flow is not terminated and not drained to sump, BWST depletion may occur.
I 3.3.2 Loss of Both Flow Paths From Sumo Both sump flow paths can be lost if both the RB omorgency sump inletc becomo clogged with debris or if valvos in both lines fail to open. If the motor operated valvos fail to operate remotely, then local manual oporation of the valves should be attempted to open at least one of tho f
valves. Local attempts to open thoso valvon may not be possible because of high radiation lovels.
7-23-85 PAGE IV.B-24 DATE:
CNP 200013 (9 84)
SABCOCK & WsLCOR NUanttR NUCLEAR POWER Olvist0N
" -11**414~
OTECHNjCAL DOCUMENT V If the cause of a loss of sump water or flow path cannot be corrected, the operator should attempt to cool the
- reactor core with the DHRS. In this situation, starting DHRS without adequate SCM is permissible. This method of i cooling will be successful if the cooling water can flow through the RCS without leaking to the RB. To accomplish this, the RCS water must be 1) subcooled to prevent steaming out the break, 2) below the break elevation to stop RCS water from continually leaking to the RB, and 3) high enough to prevent a vortex formation as water is drawn into the DHRS suction pipe. (This would be the same elevation as required for normal DHRS operation when draining the RCS.) When the RCS pressure drops below the ,
DHRS design pressure, DHRS operation can be initiated and the RCS water level will drop to the break location. If the break location is high enough, vortex formation in the
1 During LPI cooling, water will be lost out the break until the water level in the RB increases to the elevation equal l
to the water level required in the RCS for DHRS operation and above the break elevation. This may require borated water in addition to that contained in the BWST. These additional sources, if available, should be pumped to the RV for core cooling until either a) the sump recirculation can be established or b) in some plant configurations, the RB is flooded to the RCS level needed for DHRS operation and above the break elevation so that the DHRS can be put 4 into operation. If the RB needs to be flooded, the
\
operator should prepare for equipment and instrument DATE:
7~ " PAGE
BTNP-20007 3 (9 84)
SASCOCK & WILCOE Humsta NUCLEAR POWER Divl510N 74-1152414-oo TECHNICAL DOCUMENT failures due to water submergence. For examplo, the DHRS suction line valves should be opened before they are submerged because, after they are submerged, they may not be operable.
If available, the hot log level measuromonts should be used as an aid to indicato if an adoquate water level exists above the DHRS suction nozzle.
Onco DHRS operation is established, the RCS water tempora-ture should be hold constant to provent volumetric water changos which cause the RCS water level to fluctuato. In addition, the PCS water temperature should be hold below the boiling point to provent water loss by steam flowing out the break.
4.O D119S SYSTEM OPERATION 4.1 DHRS Onoration W1th a Saturated RC When the RC is saturated, the DHRS should not be placed into operation without careful monitoring of D!lRS rump cavitation because the liquid level above the DHRS drop line is not known. Without adequato lovel above the DHRS drop line the DHRS pun.ps could cavitato.
Specific conditions for starting the DHRS when the RCS is naturated are taken if a SGTR oxists (Refer to Chaptor III.E) or if LPI racirculation ability in lost (Refer to Section 3.3).
4.2 DHRS Onoration With a Subcooled RC and a LOC 4 4.2.1 Two LPI Pumns Onorable When the RC is subcooled, and a LOCA exists, one LPI pump must bo kept in the LPI mode of operation. The other LPI pump should be put in the DHRS modo of oporation (Notes RB spray system design may require stopping a spray pump if 7-23-85 IV.D-26 DATE: PAGE l
S!N720007-3 (9 e4) l 4ASCOCK & wtLCoE Numete NUCLEA4 Powf R DivlSION 74-1152414-o0 TECHICAL DOCUNE11T !
U on). Consequently, when the DNRS operating conditions are reached during a plant cooldown, one DHRS train should be started. HPI will continue to make up for lost RCS inven-tory. The LPI pump that was left in the LPI mode will be o able to provide water to the HPI if the BWST empties and will provide makeup for lost RC inventory when the RCS pressure reaches LPI operating conditions. i 4.2.2 One LPI Punn Operable If a SBLOCA exists and only one LPI pump is operational, f l it should be left in the LPI mode and not used in the DHRS
! mode. The LPI provides both RC inventory MU and core
! cooling. The DHRS provides only core cooling. ;
I 4.3 DHRS Oceration With a Subcooled RC and No IDCA Usually the DHRS is started as soon as possible. However, if the RC is highly radioactive, it may be decided to ;
continue using the SG(s) for heat removal rather than ,
spreading the very radioactive RC to other fluid systems !
outside of the RB.
5.0 ggtE FLOOD TANK OPERATION 5.1 The Core Flood Tank Isolation Valves should be closed :
durina cooldown with a loss of SCM when the LPI is flowinq ;
in each LPI line at a rate in excess of the flow rate i listed below. [
PLANT FLOW _ RATE 205 FA NSS 1500 gpa [
177 FA NSS (Except ANO-1) 1000 gym f
AU0-1 (2 LPI pumps on) 2630 gym ANO-1 (1 LPI pump on) 3020 gym (Total)
Leaving the CFT isolation valves open during a saturated -
cooldown will not interrupt core cooling and, therefore closing them is not mandatory. During a large break LOCA, the core flow will be so turbulent that adequate core 7-23-85 PAGE IV.B-27 l DATE:
b
BWN32000M (9 84)
SAtCOCK & wlLCOX NUMllt NUCLEAR POWER Divl510N TECMCAL DOCUMENT heat transfer will exist even with entrained nitrogon.
The nitrogen may accumulato in the SG(s) or RC loops; howevnr, the SGs are not used for any heat removal. The RC saturation temperature will rapidly decrease to a value too low for the SGs to be of any uso in removing heat. In addition, the RCS blowdown and nitrogon release will occur too rapidly for the operator to isolate the CFT isolation valves. During a SDLOCA, the decay heat is low, the nitrogon addition rato is slow and the core is submerged so that adequate core heat removal exista. However, closing the CFT isolation valvos to provent nitrogen addition in donirable. Ono reason for not releasing nitrogen during a SDLOCA in to koop the Sco available for heat removal. The nitrogon can accumulato in the SG(a) reducing SG heat transfer and in the RCS loop blocking natural circulation flow. Another roanon in to reduce wanto gas management.
Ideally, during a SDLOCA the CFT isolation valvoo should be closed just before the CFTs are empty an datormined by CFT level or RC proosure instrumentation. !!owever, this instrumentation should not be unod for this purpoon because it may be incorrect due to the adverso RD onviron-mont resulting from the SDLOCA. The LPI flow rato in dopondent on RC pronsure. The LPI flow instrumentation is not affected by the RD onvironment. The RC prosaura must decreano below approximately 100 poig before the CFTo empty. LPI will be pumping the required flow rato before the RC pronouro decreason to 100 psig. The critorion for stopping IIPI is banod on the LPI pumping the required flow rato. Consequently the critorion for stopping flPI can be applied to cloning the CFT inolation valven.
!!owever, the critorion for atopping itPI pumpn also ntipu-laton that the required LPI flow han oxinted for 20 min-DATE: PAGE
CCNP 200074 (9 84)
SASCOCK & witCOR NU#tt G NUCLE AR POWER DIVIsl0N 74-1152414-00 ,
(\ TECHNICAL DOCUMENT utes. The 20 minutes is not applicable to closing the CFT isolation valves.
5.2 The Core Flood Tank isolation valves should be closed durina subcooled cooldown before the reactor coolant Dressure decreases to the Core Flood Tank Dressure.
If the RCS is subcooled, the core is being adequately cooled and CFT water is not needed for core cooling. The CFT isolation valves should be closed so that the CFTs do not interfere with RCS depressurization. If the CFT isolation valves are open, the CFT nitrogen cover gas will inhibit RCS depressurization.
6.0 H9RQN PRECIPITATION Within twenty four hours after a LOCA, actions should be taken to preclude the possibility of boron precipitation.
/ With a large hole in certain areas of the RCS, the reactor can, acting like a dirtiller, boil off almost pure steam and leave impurities (mostly boron) to concentrate in the RV. If enough boron accumulates, core flow blockage might occur. To limit the boron concentration, the operator should follow the plant procedure for preventing boron precipitation.
7.0 HPI/LPI " PIGGYBACK" OPERATION l 7.1 If LPI flow in each iniection line for at least 20 minutes has not been in excess of the followina flow rates before the BWST level reaches the reactor buildina emeraency sumo switchover level metooint, then HPI and LPI must be coerated in the HPI/LPI olacyback mode.
l I s l
7-23-85 PAGE IV.B-29 DATE:
BWNP-20007 3 (9 84)
SASCOCK & WILCOE
" E NUCLEAR P0wtR OfvlS10N TECumCR DOCUMENT PLANT LPI FLOW 205 FA NSS 1500 gpm ,,
177 FA NSS (Except ANO-1) 1000 gpm ANO-1 (2 LPI pumps on) 2630 gpm ANO-1 (1 LPI pump on) 3020 gpm (Total)
LPI is not providing the required flowrate for stopping HPI (Refer to Section 2.5.2) . Therefore, HPI must con-tinue. However, the BWST is depleted and the HPI system cannot take suction directly from the RB emergency sump.
The LPI system must take suction from the RB emergency sump and discharge to the HPI system suction. This system line up is entitled "HPI/LPI piggyback operation."
In addition to providing a flow path from the RB sump, LPI also provides the required NPSH for the HPI when recircu-lating RB emergency sump water.
7.2 When operatina in the HPI/LPI "Dicavback" mode, one HPI numn may be secured erovided both LPI Dumos are Drovidina suction water for the runnina HPI Dumo and the runnina HPI numo iniects into all four iniection lines.
This action would prevent the poor quality RB sump water from degrading both HPI pumps. The second HPI pump can be stopped even if the SCM does not exist. This is contrary to the SCM rule but is permissible in this situation because the decay heat has decreased substantially. In order for the BWST to empty before the RCS depressurizes to the LPI operating pressure, the LOCA must be small.
This means that, before the BWST emptics, a relatively long period of time will have elapsed, during which time the decay heat will have appreciably decreased.
O 23-85 IV.D-30 DATE: PAGE
4 f
FIGURE IV.R-1 HPI TPROTTLI'!G LI' TIT (FDP r 4 s !
HIGH. FLOW LIflE) FOR DB-1 2
, i t
i i i
1
} I i !
j 1600 -
l 4
- 1400 -
[
!' t I
r j 1200 -
t
,, j i i l 5 1000 -
m,.
l as i
3
! 800 - i i .,
- - I
- o. l u, ,
c il a:
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' I
?
400 i c l i
' i i
1 200 - i il ;
i- l
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l ' ' '
i 0 !
> i
, 0 100 200 300 400 i l
l' HPI Flow, gpm !
' t
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i
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I
_ __ _ =_-._-. _ _ _ _ _ _ _ _ .
BWNP 20007 3 (9 84)
S ABCOCK & WitCOM NUCLEAR POWER OlVISION NUMSER 74-1152414-o0 TECHNICAL DOCUMENT Chapter IV.C MFW/AFW System Ooeration
1.0 INTRODUCTION
Steam generator inventory control is one of the five control functions discussed in Chapter II.C. The SG inventory is controlled by AFW and MFW. This chapter t
will discuss special considerations associated with the operation of AFW and MFW.
In general, FW should be supplied only to SGs which can hold pressure (i.e., no significant unisolable steam leak). However, exceptions do exist and will be discussed in Section 5.0.
For the SGs that can hold pressure, this chapter will discuss:
a) SG water level requirements. (Subsection 3.0) b) Excessive FW flow. (Subsection 4.1) c) Initiating AFW flow. (Subsection 4.2) d) Maximizing FW flow. (Subsection 4.3) e) Throttling FW flow. (Subsection 4.4) 2.0 DEFINITIONS 2.1 Steam Generator That can Hold Pressure This is a SG without an unisolable steam leak or with a very small unisolable steam leak. A very small steam leak is one which itself cannot remove more energy than is being transferred to the SG; i.e., the steam leak cannot cause the RC to decrease in temperature.
2.2 Loss of Subcoolina Marcin Refer to Chapter IV.B. ,
.[- '
G l 7-23-85 PAGE IV.C-1 DATE:
C'2NP 20007 3 (9 84)
BABCOCK & WitCOX NUCL[AR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT 2.3 SG LEVEL SETPOINT DEFINITIONS The following SG level setpoint definitions are used in this document:
A. Forced Flow Setooint -
This setpoint is used when RCPs are operating and the RCS has adequate SCH.
However, this setpoint may be different depending on whether the MFW system or the AFW system is supply-ing FW.
B. Natural Circulation Setooint - This setpoint is used when no RCPs are on and the RCS has adequate SCM.
C. Loss of Subcoolina Marain Setooint -
This setpoint is used when adequate SCM does not exist.
D. ICC Setooint - This setpoint applies only to Davis-Besse and is used when the RC temperature and pressure is in region 3 or 4 of Figure III.F-1.
E. Shutdown - SG Overfill SetDoint - The SG level must be kept below this value to prevent water from entering the steam lines after a reactor trip.
F. AFW Start SG Level - This is the SG water level at which the AFW system, starts following a loss of HFW.
3.o STEAM GENERATOR WATER LEVEL This section discusses the bases for SG level setpoints. Setpoint values are plant specific. In general, the SG water level is maintained only high enough for adequate primary to secondary heat transfer to limit unnecessary filling of the steam generator.
7-23-85 PAGE IV.C-2 DATE:
87tNP 20007 3 (9-84)
S ABCOCK & WitCOX NU"8II NUCLEAR POWER DIVISION
' 74-1152414-oo O TECHNICAL DOCUMENT The required SG water levels when two SGs are operating
' When only one are discussed in the following sections.
SG is operating in natural circulation, raising the water level slightly higher may be beneficial. This action increases the heat removal ability of the one operating SG.
The rate at which the SG level setpoints are achieved is determined by the FW flow rate guidelines discussed in Section 4.0 3.1 SG Level With Ooeratina RCP(s)
When usina MFW or AFW and at least one RCP is on and SCM 4
exists, the SG level should be controlled at or above a
the forced flow setuoint.
This level is sufficient for removing core and RCP heat when the RC is subcooled and forced circulation of RC exists between the core and the SG.
3.2 SG Natural Circulation Setuoint If all the RCPs are deenercized and the RC SCM exists, the SG level must be controlled at or above the natural
- circulation setuoint.
The RC-will have to flow between the core and the SGs by natural circulation. This requires a higher SG level than does forced-RC flow. The SG level must be high ;
enough to create a SG heat sink thermal center suffi-ciently above the core thermal center to induce adequate natural circulation of the RC.
3.3 SG Loss of Subcoolina Marcin
- If the RC SCM is lost, the SG level must be controlled
\, at or above the loss of subcoolina marcin setooint.
DATE: 7-23-85 PAGE IV.C-3
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l BCNP-20007 3 (9 84)
BASCOCK & WILCOX NUCLEAR POWER DIVISION NUMBER 74-1152414-oo TECHNICAL DOCUMENT This SG level is even higher than the natural circula-tion setpoint. When the RC SCM is lost, the RC has a potential of being saturated. Therefore, the SG water level is raised to the loss of SCM setpoint. This setpoint has been determined for a saturated RCS, with steam in the hot leg pipes for the case when core cooling is assisted by boiler condenser cooling. The core heats the surrounding water creating steam which flows through the hot leg pipes to the SGs where it is condensed. The resulting pool of water in the SG tubes must be higher than the elevation of the RCP internal spill-over so that the cold leg water will flow to the RV. For this to happen the SG condensing surface has to be higher than the RCP internal spill-over. Also the condensing surface of the SG tubes must be adequate, combined with HPI cooling, to remove all the heat being generated by the core. The elevation of the AFW nozzles is high enough to provide the required con-densing surface. The level setpoint is set high enough to provide the required condensing surface during periods of no AFW flow.
3.4 If ICC conditions exist with an indicated fuel clad temperature creater than 1400F the SG 1evels should be raised to the ICC level setooint. (only applicable to Davis-Besse)
The SG water level should be raised to the maximum level possible without causing water to enter the steam lines or losing SG level measurement or causing SG overfill protection system actuation; i.e., SFRCS trip on high SG level. This will provide the greatest SG condensing surface.
O DATE: 7-23-85 PAGE IV.C-4
\
f SWNP.20007 3 (9 84) ;
SASCOCK & WILCOX 08EE i NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT ,
l 4.0 FEEDWATER CONTROL TO STEAM GENERATORS THAT CAN HOLD l
PRESSURE This section applies only to SGs that can hold ;
I pressure. Special considerations for SG(s) that cannot hold p'ressure are given in Section 5.0. ,
i j
The FW flow rate should be controlled to increase and
! decrease the SG level to obtain the required setpoint.
i I Guidelines are provided for controlling FW flow. Some 7 guidelines are mandatory (must) while others are j desirable (should). If a mandatory and a desirable ;
} guideline conflict, the mandatory guideline has
{ priority. ;
i j
! Some utilities may be able to feed MFW to the AFW l nozzles. For these plants, MFW flow through the AFW nozzles can be substituted for AFW. flow in the
, guidelines.
I If a SG becomes dry, AFW rather than MFW should be added to the SG to reestablish level and a RCP in that loop ;
{ should be on to limit overcooling the SG tubes. Once a l level has been established, MFW may be used.
' I Adding AFW to a dry SG without RC loop flow has not been !
analyzed. Without RC flow through the SG tubes, the AFW f may be able to overcool the tubes creating excessive l tube-to-shell delta T. Also adding MFW to a dry SG has l
- not been analyzed. If either of these situations occur, l it must be evaluated for excessive stresses before f returning to service. !
i i 4
i
.DATE: 7-23-85 PAGE IV.C-5 l
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BWNP 20007 3 (9 84)
S A BCOCK & wlLCOX NUMatt NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT 4.1 Excessive FW Excessive FW is an uncontrolled condition and must be stopped. It occurs when the FW addition causes the RC to overcool or the SG level to increase above the desired setpoint which, if not stopped, will cause the mass of water in the SGs to increase until water is carried over into the steam lines.
4.1.1 Excessive MFW 4.1.1.1 When the reactor is shutdown. the MFW flow must be controlled to prevent the SG level from exceedina the shutdown-SG overfill setooint.
When the reactor is shutdown the water level in the SGs must be kept below the main steam outlet nozzles while steaming so that the steam lines do not start filling with water. Note: For the 177 FA plants the SG upper baffle has instrumentation holes which will allow water to flow through the baffle. Consequently, water can start to fill the space between the SG baffle and the SG chell without having to flow over the top of the baffle to fill this space.
Excessive MFW is undesirable for several reasons.
It can cause the RCS to overcool, water release through the MSSVs, and the elimination of the SGs as a heat sink because they should not be steamed once they have filled.
As soon as the excessive MFW transient is identi-fled, actions should be taken to quickly stop it because excessive MFW can fill the SGs quickly; e.g. , water can start entering the steam lines in as little as 1 minute after a reactor trip if MFW flow is not reduced.
i l
7-23-85 PAGE IV.C-6 DATE:
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BT!NP 20007 3 (9 84)
S ABCOCK & WILCOX OIII NUCLEAR POWER DIVISION 74-1152414-0o TECHNICAL DOCUMENT As the SG water volume approaches the shutdown-SG overfill setpoint, the MFW pumps must be tripped to assure the SGs do not overfill. However, if the excessive MFW is discovered before the shutdown-SG overfill setpoint is reached, then the operator should try to stop the excessive MFW by closing the MFW isolation valves.
't Tripping the MFW pumps will immediately stop MFW flow. However, this action causes all MFW flow to be discontinued. If only one SG is being overfed, it is preferred to stop MFW to only the overfed SG. By
. closing MFW isolation valves instead of tripping the MFW
! pumps the MFW flow can be continued to the other SG.
Whenever FW is stopped to both SGs, FW flow (either MFW or AFW) should be restarted before both SGs are dry.
O 4
4 .1.2 Excessive AFW 4.1.2.1 When the reactor is shutdown, the AFW flow must be stooped to the overfillina SG before the SG level l
l reaches the shutdown - SG overfill setooint.
i Excessive AFW can also cause SG overfill. Excessive AFW can cause significantly more overcooling for the same excess flow as MFW because the AFW sprays into the SG steam sphce rather than entering the SG water space
, and the AFW water is colder than MFW. However, AFW cannot fill a SG as fast as MFW flow because the AFW I
system has less flow capacity.
f If excessive AFW occurs, the operator should try to i regulate the AFW valves or pump speed to stop the overfeed and, if unsuccessful, stop the pumps and batch feed the SGs, e.g., starting and stopping AFW pumps f
as needed.
I l
7-23-85 PAGE 'IV.C-7 DATE:
BONP.20007 3 (9 84)
S ASCOCK & WitCOX NUCLEAR POWER DivlSION 74-1152414-oo TECHNICAL DOCUMENT 4.2 Initiatina AFW 4.2.1 AFW must be started whenever MFW flow to both SGs is disrupted causina the SG water level to decrease to the AFW start SG level setooint.
This is to provide continued primary to secondary heat transfer.
4.2.2 AFW must be started whenever there is inadeauate crimary to secondary heat transfer or whenever the SG 1evel must be controlled at the loss of subcoolina marain setooint. This is because the AFW enters the SG above all operating setpoints. Consequently, when AFW starts flowing the SG thermal center is rapidly raised higher than required without waiting for the actual SG water level to be raised to the desired setpoint.
4.2.3 AFW must be started or MFW rerouted to the AFW nozzles when forced RC flow is lost.
This is to provide a smooth transition to natural circulation by raising the SG thermal centers.
4.3 Maximizina AFW 4.3.1 Whenever there is inadeauate crimary to secondary heat transfer, AFW flow rate should be maximized without causina the SG oressure to ao more than 100 osi below the SG oressure control settina, until the recuired SG level setooint is reached or adeauate crimary to secondary heat transfer is established.
This is done to establish primary to secondary heat 1
I transfer as quickly as possible without overcooling the SG while heat transfer does not exist.
O DATE: 7-23-85 PAGE IV.C-8
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BWNP-20007 3 (9 84) 8ASCOCK & witCOR I
' NUMSit NUCLEAR POWER OtVISION 74-11s2414-oo TECHNICAL DOCUMENT
{
4.4 Throttlina AFW i
4.4.1 If AFW flow should be added. it should be controlled so l i
that a continuous AFW flow is orovided and the SG level I
y never decreases with an overall oroaression toward the ,
I f recuired setooint.-
l Except when a minimum AFW flow must be provided for a l loss of SCM combined with a loss of primary to secondary l heat transfer (per Section 4.4.3), the SG level can be l held constant as long as the AFW spray flow is removing i adequate heat. If flow is not sufficient to remove all
. the decay heat, the SG level will start to decrease.
3 A continuous AFW flow will place the SG thermal center above all required-SG 1evel setpoints. The continuous I
flow will also reduce thermal cycles on the AFW spray j nozzle.
- i. ?
i j Holding the SG level constant instead of raising the SG
$ level would be desirable during an overcooling. If the i SG level is increased , the overcooling situation would i
, become worse; For certain SGTR situations the SG level ;
1
- does not have to be raised with AFW. The tube leakage !
will raise the level. Thus. throttling AFW will delay the time when the SG will need to be steamed or drained for preventing the SG from exceeding' the SGTR overfill j limits. Refer to Chapter III.E for more information f i about SGTR and SG level control with a SGTR.
i 4.4.2 The AFW flow rate should be throttled to Drevent the SG cressure from dronnina more than 100 osi below the I desired SG oressure control settina. t i Throttling AFW is done -to prevent overcooling the RCS.
j The AFW can cause significant RCS overcooling because j the'AFW is significantly colder than SG fluid and it is f f
I sprayed into the SG steam space. 'However,.the AFW flow i
(
Iv. C-, .
- om. 2-u-o s ,,
._.-,_.,,,y., y . _ . , , , , , , . , , , , _ , , , , , _ _ , .
BCNP-20007 3 (9-84)
BAaCOCK & WILCOX NUMBER NUCLEAR POWER DIVISION TECMCR DOCUMENT must not be thrcttled below any flow rate required by section 4.4.3.
4.4.3 Whenever adeauate SCM is lost and incore T/cs are not decreasina. AFW must be controlled to Drovide at least the minimum flow rate listed below to each SG until the loss of subcoolina setDoint is reached.
This is to assure adequate SG heat removal for a SBLOCA.
205 NSS systems - 500 gpm ANO-1, CR-3 - 125 gpm Rancho Seco - 150 gpm TMI-1 - 100 gpm Oconee and Davis Besse - maximum possible AFW flow 5.0 INVENTORY CONTROL OF STEAM GENERATORS THAT CANNOT HOLD PRESSURE When FW is introduced to a depressurized SG, AFW is preferred over MFW. If AFW is not available and FW must be added to the SG then MFW can be used. In either case, FW must be added slowly and continuously to prevent RC overcooling and excessive thermal stress of the SG. A stress evaluation of the SG may have to be made before plant restart. The following will discuss when and how to add FW to a SG that cannot hold pres-sure..
5.1 Recuired FW Flow Rates to SG(s) When Neither Can Hold Pressure.
A total loss of steam pressure control exists whenever an unisolable steam leak exists in both SGs. If this condition exists, the operator should perform the following (A stress evaluation of the SGs will have to be made before plant restart if the following actions are taken):
7-23-85 .C-10 DATE: PAGE
BWNP 20007 3 (9 84)
B ABCOCL & wtLCOM l NUm8tt NUCLE AR POWER Divt$10N 74-11s2414-0o O TECHNICAL DOCUMENT b
- 1. Attemot trickle feedina both SGs -
Maintain primary to secondary heat transfer by supplying AFW to both SGs at a limited rate while attempting to stop the steam leak on at least one SG (e.g.,
manually isolate stuck open ADV, gag shut MSSV, etc.). Automatic FW control system will probably need to be bypassed to allow AFW flow to both SGs and to allow trickle feeding. One RCP per loop should be operating to limit overcooling the SG tubes.
- 2. Trickle feedina one SG -
If one steam leak is inside the RB then that SG should have all FW flow stopped to prevent steaming
'N flow to one SG in order to aid in repairs or if manual heat removal with the two SGs cannot be controlled. If one SG needs to be isolated then isolate FW to one SG and allow it to boil dry while continuing restricted AFW flow to the other SG and attempting to maintain controlled decay heat removal. Continue trying to stop the leak on at least one SG. If a forced cooldown situation exists, such as during a SGTR, the operator may have to feed the other SG to maintain tube-to-shell delta T or RC loop flow if the RCPs are stopped.
HPI cooling is used so that the water collecting in
[ the RB emergency sump is borated. SG operation l
i (x
l with a steam leak in the RB will result in non-bor-7-23-85 PAGE IV.C-ll DATE:
ScNP 20007 3 (9-84)
S AB COCK & WILCOR NUCLEAR POWER DIVISION NUM8tt 74-1152414-oo TECHNICAL DOCUMENT ated water collecting in the RB emergency sump.
Core cooling with recirculation from the RB emergency sump may become necessary.
5.2 Feedina a Dry SG When the Other SG Can Hold Pressure Normally if only one SG can hold pressure the other SG should not be fed. It should be allowed to boil dry.
However, certain circumstances may require feeding the SG with the unisolable steam leak. These are as follows: l t
- 1. Natural Circulation Cooldown - If a natural I
circulation cooldown is required, some SG cooling may be required to circulate the relatively hot water in the idle RC loop. If it is not cir-culated, a steam bubble will form in the hot leg as the RC is depressurized and will hinder further RCS depressurization. SG cooling should be established by adding AFW. Refer to chapter III.G.
l
- 2. Excessive SG Tube-to-Shell Delta T -
If a forced circulation cooldown is required, the cooldown rate of the RCS may be faster than the dry SG shell cooldown rate. In order not to violate the SG tube-to-shell temperature difference limit, some FW should be added to the SG to assist in cooling the SG shell or the cooldown rate may be reduced.
For the 205FA SG, almost the entire shell is exposed to steam flow such that AFW should be able to cool the shell. However, for the 177FA SGs, only about half the shell is exposed to steam flow. The lower shell is not exposed to steam flow. However, the lower shell is exposed to MFW flow. Therefore, MFW flow may cool the lower SG DATE: 7-23-85 PAGE IV.C-12
. . . _ _ _ . . . _ . . _ _ _ = _ . ..m__,_ _ . . . _ _ _ _ _ _ _ . . _ __..___. - . _ _ __._ - ______ _. _
8%NP 20007 3 (9 84)
S ABCOCK & WitCOR ,
NUmSEE NUCLEAR POWER OlVISION 74-1152414-oo TECHNICAL DOCUMENT shell. Also, MFW flow can cause thermal stress of l the lower tube sheet; this also has not been l f
analyzed. Consequently _MFW and AFW must be l r
added slowly.
I I
i i
i i
i f
i l
t t
6 I
I f
[
7-23-85 PAGE IV.C-13 DATE:
I
' BYNP.20007 3 (9 84)
SAtCOCK & WILCOX NUCLEAR POWER DlylSION NUmste 74-1152414-00
] TECHNICAL DOCUMENT i Chapter IV.D Incore Thermocoucle
1.0 INTRODUCTION
There is one T/C in each self powered neutron detector (SPND) string. This T/C senses the core exit temperature.
Because of their close proximity to the reactor fuel, they 1
are the best method for determining the temperature of the fuel when the RC is saturated. Also, Th and Tc RTDs may not give valid indications during saturated conditions.
During normal operation, a spread of up to 50F is expected from the outer peripheral incore T/cs to the center of the core incore T/cs. The shape of the temperature profile will parallel the profile of the flux across the core during y normal operation. After the reactor is tripped, the 4
incore T/Cs should all read within about 10F of each other.
, 2.0 USES OF INCORE THERMOCOUPLES 2.1 Detect Core Uncoverina 2.1.1 Thermocouoles vs. RTDs Normally the hot leg RTDs will be used to determine the core exit temperature. The hot leg RTDs are more accurate than the individual incore T/Cs. However, because of their location, the hot leg RTDs are not a true indication of core temperature if the RC is not circulating or if there has been a loss of SCM (RCS is saturated). As soon as RC flow I ceases the RTDs will not provide as accurate an indication ,
l of reactor core temperatures. Furthermore, if steam voids occur such that the hot leg RTDs are enveloped in steam, they are even less accurate as a temperature indication.
l The heat transfer from the steam to the RTD is poor. There
' is also the possibility of a hot leg break during saturated i RCS conditions, whereby, the hot lag RTD would not come in DATE: 7-23-85 PAGE IV.D-1
BONP 20007 3 $84)
B ABCOCK & WitCOX NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT contact with water or steam which is actually flowing across the core. The incore T/Cs are the only valid core tempera-ture measurement without RC circulation and should be the main indication when SCM does not exist (i.e., saturated conditions). Therefore, as soon as the SCM is lost, the incore T/cs should be used for determining the reactor core temperature. Only after the RCS has become subcooled and natural circulation or forced circulation has been reestab-lished should the RTDs again be relied on as accurate indication of core temperature.
2.1.2 Determination of Incore T/C Temoerature Because of the normal gradients of temperature across the core, T/C readings should be averaged to determine the actual conditions in the RCS. As soon as the SCM is lost, an average T/C reading should be used to determine core outlet temperature. The method of averaging is plant specific.
2.2 Indicate Marcin to PTS Limit Although the PTS limit is colder than the SCM, there are some cases where the incore T/Cs must be used to indicate the RCS temperature to compare to the PTS limit rather than relying on the RTDs. Any time there is no natural circula-tion or forced circulation in the RCS, the incore T/Cs should be compared to the PTS limit. If, for example, a cold leg or hot leg break existed upstream of the RTDs, then the flow of the RC would not cross the RTDs. However, a sufficient SCM could develop with this type of flow, so that thermal shock would be a concern. In this case HPI flow l would have to be throttled to prevent exceeding the PTS limit after comparing RCS pressure with the average incore j T/C temperature. The P/T point would be used to determine l when to throttle HPI flow to maintain SCM within the PTS limit.
DATE: 7-23-85 PAGE IV.D-2 l
BWNP 2000's 3 (9-84)
BABCOCK & walCOX NUCLEAR power OlvlS10N NUMSER
' 74-1152414-oo
' TECHNICAL DOCUMENT 2.3 Indication of Natural Circulation When the RCS is subcooled the relationship between the hot leg RTD reading and the incore T/C temperature is a good indication of natural circulation. The hot leg RTD indica-tion should be within 10F of the incore T/C reading when subcooled natural circulation is occurring.
If the RCS is saturated, the relationship between the hot leg RTDs and incore T/Cs cannot be relied upon to give a good indication of saturated natural circulation. When the RCS is saturated, the hot leg RTDs may track the incore T/Cs even when natural circulation does not exist. A divergence between the hot leg RTDs and incore T/Cs may indicate a loss of natural circulation, but the hot leg RTDs cannot be used to confirm natural circulation. Therefore, when the RCS is saturated incore T/Cs should be used to confirm natural O circulation as discussed in Chapter II.B.
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l DATE: 7-23-85 PAGE IV.D-3 [
___ ,_.__. . . _ _ . . _ _ _ _ _ . . _ ~ _ . - _ -_ . _ - _ _ . . . .__ _ ._._-. _ __ _ . _ __ _ _ _ _ _ , . . ._
-. .. .. - . - . ~ . _ - . .- .- .- --. - . .. -. . - - . - - ~ -
87JNP 20007 3 (9 84) i B ASCOCK & WRCOM NUCLEAR POWER DIVISION NUMBER TECHICAL DOCUMENT 74-11s2414-00 ,
Chanter IV.E l
, Hioh Point Vents i i
, 1.0 , INTRODUCTION
. Depending on the plant design, high point vents (HPVs) can
, be located at the top of both hot legs, on top of the RV head, and also on top of the pressurizer. Some of these HPVs are connected by drain lines to the quench tank, however, other HPVs discharge into the RB. The point where f
l they discharge will determine whether or not they should be
! used for certain conditions. This chapter discusses the uses of the HPVs.
i' 2.0 USE OF HIGH POINT VENTS (HPV) '
2.1 Open HPVs Durir.a ICC Conditions l Use of the HPVs during the ICC condition can contribute significantly to restoring primary to secondary heat r transfer, thereby regaining adequate core cooling. Once the [
l fuel clad temperature increases to greater than 1400F, clad i i
n oxidation begins which produces hydrogen gas and other l noncondensables. These noncondensable gases will collect 3n the high points of the system, RV head and pressurizer. .I f !
l
! sufficient quantities of noncondensable gases collect in ths l hot legs, natural circulation flow can be stopped. With the
- presence of noncondensable gases in the hot leg, boiler i ;
l condenser cooling could be greatly impeded. Bumping a RCP f I would be ineffective in this case, because once the RCP was i stopped, the noncondensables would again collect.at the top ,
l of the hot leg, preventing natural circulation. It is ,
I therefore necessary to vent these noncondensables out of i l the primary system to allow a restoration of natural i circulation / boiler condenser cooling. As soon as the clad i
reaches a temperature of greater than 1400F (shown on Figure l III.F-1 Regions 3 and 4) all HPVs should be open.
DATE: 7-23-85 PAGE IV.E-1 ,
BcNP.20007-3 (9 84)
B ABCOCK & WatCOX NUCLEAR POWER DIVISION NUMBit ,
74-1152414-o0 TECHNICAL DOCUMENT 2.2 Reclosina HPVs Upon Return From ICC If adequate SCM is regained, all HPVs may be reclosed. Once SCM is regained all of the noncondensable gas production will have ceased. While there may be noncondensable gases left in the RCS, they should be entrained in the RC and not impede natural circulation flow. However as the RCS is depressurized, these gases will come out of solution. If natural circulation is lost at a later time, and cannot be restarted with RCP bump, or if boiler-condenser cooling is lost, it may be necessary to reopen the HPVs to remove any noncondensable gases left in the system. Opening HPVs at this time can also help to eliminate steam trapped in the top of the hot legs and thereby restore natural circulation.
If the RCS is brought back to saturation, then the RCS should be cooled and depressurized until the decay heat removal system is in operation. After the decay heat removal system is operating and RC pressure is decreased to less than 150 psig, then noncondensable gas production will have ceased. Consequently, it is allowable to close the HPVs once the decay heat removal system is operating with RC pressure less than 140 psig.
2.3 Other Uses of HPVs There are several other uses of the HPVs which may be viable, depending on the severity of the situation, as well as the discharge points of the HPVs. Each of these situa-tions and the appropriate use of the HPVs is discussed in the following:
l A. Control of RCS Pressure The pressurizer HPVs have a lower capacity than the PORV. As long as adequate pressure control can be maintained, the pressurizer HPV should be used for DATE: 7-23-85 PAGE IV.E-2
i I BYtNP.20007 3 (9 84)
BA8 COCK & witCOX NUMSER NUCLEAR POWER DivtSION 74-1152414-oo TECHNICAL DOCUMENT controlling RCS pressure in situations where neither sufficient pressurizer spray or auxiliary pressurizer spray is available.
i It is also possible that all HPVs could be opened to i
! be especially advantageous where low head HPI pumps are used. It may be necessary to open HPVs in an attempt i
to reduce RCS pressure when the PTS limit is in danger of being exceeded during HPI cooling. '
I Pressure reduction in the RCS is also vital during HPI 4
cooling with a solid SG which has a tube rupture in it.
. Opening the HPVs will augment RC depressurization during i this situation. Refer to Chapter III.E for. further l details of HPV use and control during this situation.
B. Refill of a Voided Hot Leo If the RCS had been saturated and steam voids collected in the hot leg impeding natural circulation flow, the I HPVs could be used to restore natural circulation, provided the core has been restored to a subcooled l condition. The HPVs could be opened in the top of the hot legs to allow the trapped steam voids to escape.
This will allow HPI flow to refill the hot. legs as long ,
! as RC pressure'is maintained. It may be preferable to
! use RCP bumps when possible, as opposed.to opening the HPVs, because most HPV designs discharge into the RB and the RCP bump will remove the steam voids faster by sweeping them into the subcooled liquid where they collapse.
i I
7-23-85 PAGE IV.E-3 f DATE:
BWNP-20007 3 (C-84)
SASCOCK & WILCOX -
NUM8tt NUCLEAR POWER OlVISION 74-11s2414-oo TECHNICAL DOCUMENT C. Provide for RV Head Coolina Durina Natural Circulation During a natural circulation cooldown, the RV head is expected to cool very slowly. If the natural circula-tion cooldown and depressurization is too rapid, steam voids may be formed in the RV head. These voids would be indicated by a loss of RCS pressure control and a sudden increase in pressurizer level. The RV head vent may be opened to provide some RV head cooling or to relieve steam if necessary. Refer to Chapter III.G.
O O
DATE: 7-23-85 PAGE IV.E-4
' BWNP 20007 3 (9-84)
S A SCOCK & witCOE NUmHS NUCLEAR POWER OlVi$ ION 74-1152414-oo TECHICAL DOCUMElli Chacter IV.F i Containment Systems 4
1.0 INTRODUCTION
The containment systems are used for RB control. RB control 8
consists of two basic objectives:
A. Limit leakage from the RB (RB isolation control). ,
B. Control the RB environment (RB internal environment j control).
1 The systems available for RB control and their designs can '
- vary significantly among plants. This chapter will discuss
- some situations which can be created by the NSS which need
! to be considered in RB control. The chapter will not discuss details of systems operations.
s 2.0 REACTOR BUILDING ISOLATION CONTROL i RB isolation control involves reducing and controlling r leakage through the RB penetrations following a diagnosis of an abnormal condition. The control actions can include:
- 2) Selectively isolating and unisolating penetrations as needed for abnormal transient operations.
- 3) Operating equipment in a secondary containment or auxiliary building to monitor and control -leakage from
- the RB.
Specific parameters can be monitored that may indicate the need for full or partial RB isolation. These parameters may
, include:
- High RB radiation.
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High RB pressure.
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7-23-85 PAGE IV.F-1 DATE:
C"JNP-20007 3 (9 84) sAsCOCK & wlLCox NUCLEAR POWER DIVl$ ION NUMatt 74-1152414-o0 TECHNICAL DOCUMENT
- Low RCS pressure.
Each of the three conditions listed above relate to a potential release of radioactive material as follows:
During abnormal transients the operator may need to use some of the fluid system penetrations to help maintain core cooling and to control the RB environment. The RB isolation valves may have to be selectively operated as choices are made between the need for RB isolation and fluid system operation. The decision to open the valves should be accompanied by a judgement of possible consequences (e.g.,
, the penetration path cannot be reclosed).
f f
l cooling water systems to the RB can provide a path for radiation release from the RB space. This is unlikely since the cooling water systems do not normally carry radioactive fluid and many of the systems operate at a pressure which is usually higher than can occur in the RB.
3.0 RB INTERNAL ENVIRONMENT CONTROL After an abnormal transient, the environment inside the RB can become harsh enough to cause failures or degradation of equipment.
O DATE: 7-23-85 PAGE IV.F-2
BINP 20007 3 (9 84) j S A B C0CK & wlLCOX NUantit NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT ,
The RB environment needs to be controlled to reduce the possibility of failures of equipment inside the RB. RB contrcl includes controlling the following to bring them within acceptable limits:
A. RB pressure and temperature.
B. RB hydrogen concentration. l C. RB sump chemistry. t l
D. RB radiation.
3.1 RB Pressure and Temoerature RB temperature and pressure are cot 1 pled together for a LOCA so that if the temperature is reduced the pressure will also be reduced and vice versa.
' As the LOCA releases steam and water, this mixture heats up the RB atmosphere. The decreasing mixture temperature and increasing air temperature will come to equilibrium at a common atmospheric temperature. This temperature will determine the RB pressure. The pressure will be the sum of the partial pressures of air and steam.
J k
Hydrogen burning or a steam line break can also cause containment pressure and temperatura excursions. ;
The RB pressure and temperature must be reduced to:
i A. Prevent exceeding the failure pressure and design temperature of the RB.
B. Reduce the driving force for RB leakage.
C. Prevent equipment damage.
i The number of emergency coolers and sprays does not signifi- l cantly affect the peak RB pressure for large LOCAs in which f the peak pressure is reached soon after this equipment is i
7-23-85 PAGE IV.F-3 DATE:
l
OcNP 20007 3 (9 84)
B ASCOCK & wlLCOE NUM6tt NUCLEAR POWER Divl510N 74-1152414-oo TECHNICAL DOCUMENT actuated. On the other hand the peak pressure for a SBLOCA is very dependent upon the amount of cooling equipment in operation.
Large increases in RB temperature and pressure may not occur from SBLOCAs such as stuck open PORV. However, radiation can be released to the RB. Also, the RCS pressure may drop slowly causing automatic initiation of RB spray and coolers based on RC pressure to be delayed.
Normal operation of some systems may delay or prevent these parameters from reaching setpoints which actuate safety equipment. The RB coolers may provide sufficient cooling (in the normal mode) during a SBLOCA or secondary side line break inside the RB such that only a small RB pressure or temperature transient will occur. Consequently certain systems may be isolated on a radiation signal. The amount of energy being released to the RB may be reduced by:
- 1) Maintaining core cooling to keep the core covered.
- 3) Increase the heat removal by the LPI cooler (if in
! operation).
- 4) Increase the heat removal by the SGs (if in operation).
Containment temperature can affect the operation of some l equipment; consequently, if a high RB temperature is reached, the RB coolers should be started or verified i operating to bring the temperature back within limits.
I i 3.1.1 Pressure and Temperature Control With Eauipment Failures The primary method of reducing the RB pressure and tempera-ture abnormal conditions is with the RB coolers in the 7-23-85 PAGE IV.F-4 DATE:
BWNP 20007 3 (9 84)
SABCOCK & WILCOX NUM8tt q NUCLEAR POWER DivlSION 74-1152414-o0 TECHNICAL DOCUMENT emergency mode. If this equipment fails to keep the RB pressure below limits, backup cooling with the RB spray system could be used.
Failure of both containment coolers and sprays is not i
likely; however, were this to occur the effects on RB pressure and temperature for a LOCA would depend on the ability to remove heat through the SGs and, for a steam line break, would depend on the ability to stop FW to the broken SG.
3.2 Reactor Buildina Hydrocen Concentration The sources of hydrogen following a LOCA will result in a time dependent build-up of hydrogen in the containment until J control measures are taken. The sources are: a) radiolytic releases from the core and the RB sump, b) release from j galvanized metals in the RB, c) release from zinc primer paint in RB, d) release from corrosion of aluminum, e) release from hydrogen dissolved in RCS, and f) release from any zirconium-water reaction in the core.
The extent of the Zr-water reaction is controlled by the ECCS performance. Consequently, if the core is adequately covered and cooled, the Zr-water reaction will be an insignificant source of hydrogen.
i Possibilities exist for high local concentration or strati-fication. Flammability is dependent on the concentration and the concentration depends on how well the reactor building atmosphere is mixed.
Hydrogen burning will increase RB temperature and pressure.
The most apparent feature of hydrogen burning is a sudden RB pressure spike with a return to a pressure only nominally 7-23-85 PAGE IV.F-5
+
DATE:
BWNP 20007 3 (9 84) e ASCOCK & WILCOx NUM8tt nth t L AR POWER OlVi$lON 74-11s2414-oo TECHNICAL DOCUMENT higher than before the burn. The SG and RCS pressure indications will dip by about the same amount as the RB increases during the hydrogen burn since the low pressure side of these transmitters are vented to the containment.
The hydrogen concentration should be controlled as possible to either prevent or limit burning to prevent possible equipment damage within the RB.
After estimates of hydrogen concentrations have been made measures may be taken to control as possible any excessive concentrations.
The first method of hydrogen control in the RB is by mixing the RB atmosphere to prevent stratification at high RB elevations and any local concentration. This creates a more homogeneous mixture for sampling and for purging the hydrogen from the RB and reduces high product concentra-tions.
The second method of hydrogen control is reduction of RB hydrogen. Hydrogen recombining and purging are two methods for reducing hydrogen. The RB atmosphere activity and purge filter efficiency must be considered before purging hydrogen to the site atnosphere to stay within release rate limits.
3.3 Reactor Buildina Emeraency Sumo Chemistry The RB emergency sump water boron concentration following a LOCA should not be diluted. This is to assure the recircu-lation water has sufficient boron to maintain the core subcritical. Boron dilution can be caused if a fluid system leaks non-borated water to the RB sump such as a FW leak or a leak in the service water. If this occurs the source of non-borated water should be isolated. Boron addition may be DATE: 7-23-85 PAGE IV.F-6
CrNP 20007 3 (9 81)
S ABCOCK & WILCOX NUMett NUCLEAR POWER DIVISION 74-1152414-oo
,Q TECHNICAL DOCUMENT V an option. Boron precipitation in the RV can also lower boron concentration in the RB sump. Appropriate actions should be taken to prevent boron precipitation in the RV. Refer to Chapter IV.B for instructions on preventing boron precipitation.
3.4 Reactor Buildina Radiation The amount of radioactivity in the RB atmosphere or sump following an abnormal transient is dependent upon many factors such as the percent of defective fuel pins, the power history, time in core life, amount of RC leaked to the
, containment. A reduction in the amount of radioactive i material released to the RB atmosphere helps to limit the amount of radioactive material that can be released offsite.
. 3.5 Reactor Buildina Sumo Level
\
The best method of detecting an improper RB water or ,
emergency sump level is by level instrumentation. The level of the BWST after ECC injection is also an indication of nump level unless a SGTR has occurred. Detection of a high RB water level by other than level instrumentation is more difficult and may not be confirmed until submerged equipment
) is damaged. High RB water level should not be a problem
! unless more water than that added from the BWST has been added to the RB. If this additional water was non-borated, then there will be boron dilution in the RB sump.
The water level should be maintained high enough for LPI or RB spray recirculation flow (NPSH), but not so high that it submerges equipment important to core heat removal or RB control. If the water level is at the low level limit, more borated water should be added. The operator should also l s check for a SGTR or some other possible source of water less
)
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DATE: 7-23-85 PAGE IV.F-7
BWNP 20007 3 (0-84)
SABCOCK & wtLCOX NUMSER NUCLEAR POWER DivlSION 74-1152414-oo TECHNICAL DOCUMENT such as inadvertent pumping or draining from the sump or a break in LPI or spray recirculation line.
If the water level becomes too high, consideration of lowering it should be mada depending on the core cooling method being used (sump recirculation or DHR) . The amount of radiation in the sump must also be considered before attempting to remove any of the sump water.
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DATE: 7-23-85 PAGE IV.F-8 l
bj '
8;NP 20007 3 (9 84)
BASCOCK & WILCOM NUCLEAR POWER DIVISION 7'-11414- '
TECHICAL DOCUMEllT ha Chanter IV.G Reactor Vessel Pressure /Tennerature Limits i ^
{ 1.0. INTRODUCTION i The fracture toughness, which is a measure of resistance to j- brittle fracture of the RV, will vary during plant opera-j tion. The three most influencial factors causing it to vary l are:
, 1) RV temperature - fracture toughness will decrease with i decreasing temperature.
i 2) RV load rates - fracture toughness will decrease with increasing load rates. The load rates are primarily due to pressure inside the RV (RC pressure) and to stresses from thermal gradients across the RV.
l 3) Neutron irradiation - fracture toughness will decrease
{ with irradiation.
i i In recognition of these varying factors, all RVs are l l operated within restrictions imposed by technical specifice-
! tions. These restrictions assure that the RV will not be subjected to a combination of RV temperature, pressure, thermal stress and neutron irradiation which could cause
- brittle fracture of the RV if'significant flaws existed in
! the RV material.
j i The effects of thermal stress is accounted for by limiting
- the heatup and cooldown rates of the RCS.
- The effects of RV temperature and stress due to RV pressure are accounted for by limiting.RC pressure for different RC temperatures.
l l
DATE: 7-23-85 PAGE IV.G-1 l
l _ .__ _ . - - _ . _ . _ _ . _ _ _ . ._ _ _ .___ ___.____ _
BCNP-20007 3 (9 84)
BABCOCK & WILCOX NUM8tt NUCLEAR POWER DIVISION 74-1152414-oo TECHNICAL DOCUMENT The effect of neutron irradiation on the fracture toughness is accounted for by revising the Technical Specification limits as the RV irradiation accumulates.
2.0 PRESSURIZED THERMAL SHOCK LIMIT Abnormal transients can cause the maximum cooldown limit of the RV to be exceeded. If the cooldown limit is exceeded, the thermal stress will be greater than that assumed in developing the RC pressure and temperature limit of the Technical Specification.
Other RC pressure and temperature limit called the pressur-ized thermal shock (PTS) pressure-temperature limit is to be used if the RV cooldown rate limit is exceeded during an abnormal transient.
Two basic types of transients exist which can cause the cooldown rate to be exceeded. The first is an excessive heat transfer transient. In this transient the SG removes an excessive amount of heat causing all the RCS to cooldown faster than is assumed in normal P-T limits.
The second type is due to HPI cooling without the RCPs on so that HPI flow does not mix well with the RC. In this transient the HPI fluid enters the RV and flows down the side of the RV causing the RV wall to cool faster than is assumed in normal P-T limits.
The second type also causes a related concern. HPI cooling requires sufficient HPI flow to remove core heat. This flow rate and the PORV orifice size establishes the RC pressure.
Simultaneously, the HPI flow rate and water temperature establishes the RV metal temperature. The RC pressure and RV temperature cannot be independently controlled; i.e., the 7-23-85 PAGE IV.G-2 DATE:
l I
- 1 BTNP 20007 3 (9 84)
\
SABCOCK & wtLCOX '
NURSER l NUCLEAR POWER DIVISION 74-1152414 700 j E
TECHNICAL DOCUMENT , ,
i HPI flow rate required for core cooling will establish both !
i the RC pressure and RV temperature. The resulting condi- '
tions must be within the allowable RV pressure-temperature }
limits. This concern was considered in developing the f PTS pressure-temperature limit. j 1 The RV internal wall temperature cannot be measured.
1 Therefore, the RC temperature adjacent to the RV wall is
! used in the analyses to determine RV thermal stress and l i
1 temperature.. Consequer.tly, operating limits can be.made j based on RC temperature. . k i
In the first type of PTS transient, the RC is being circu-
! lated so that the cold leg temperature detectors can be used j i as a temperature-indication of the fluid next to'the RV f wall. ;
i i l In the second type of PTS transient, the cooling is local- l
! ized so the cold leg temperature detectors cannot be used as j a temperature indication of the fluid next to the RV wall. -l The incore T/cs must be used, the assumption being the HPI flow will mix into a homogeneous fluid in the RV lower planum then pass through the core to the incore T/Cs. The fluid temperature increase when passing through the core has been determined and conservatively factored into the PTS f curve. !
I i
The PTS pressure-temperature limit must be used if any valid RC temperature (see Table IV.G-1) is below 500F AHQ: l l a) The RC cooldown rate has exceeded the cooldown rate l
{ limit specified in the Technical Specification 9B l I
b) All RCPs are off and HPI flow exists through any HPI l
, line to the RV (not the normal MU line). l
)
i i DATE: 7-23-85 PAGE IV.G-3
C"JNP 20007 3 (9 84)
BABCOCK & WitCOX NUMBER NUCLEAR POWER DIVISION 74-11s2414-oo TECHNICAL DOCUMENT In addition, if the PTS pressure temperature limit must be used then:
a) Prevent any significant heatup or repressurization once the SCM is restored.
b) The PTS limit takes priority over the fuel pin in compression limit.
c) The PTS limit remains in effect for the remainder of the cooldown, even if the cause (e.g., overcooling) has been corrected.
The PTS pressure-temperature limit is depicted in Figure IV.G-1. The instrumentation to use to measure RC tempera-ture is given in Table IV.G-1.
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O DATE: 7-23-85 PAGE IV.G-4
f I
i l Figt.re IV.G-1 SUBC00 LING MARGIN LIMIT AND REACTOR VESSEL RC PRESSURE-TEMPERATURE LIMITS i
2400 RC TEMP = 500F 2200 _ g i
2000 -
7 I800 .
! i
/
1600 - f l RC TEMP = TSAT -100F
! _= i400 _
- I g 1200 f
3 /
E lo00 _
TYPICAL LIMIT j FROM TECH. SPEC.
" LIMIT 800 - l f
/ l 600 -
f I
400 - f SATURATION i
[RCPRESS=250PflG 1 A _-
f 200 - e ~~
f I
- I I t g 100 200 300 400 500 600 700 MCTES: RCTemperature(*F)
- 2. Limits shown do not include instrument error, 000.N0. 74-1152414-00 i
1 l
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. - - - . - - , _ , . , . - . . - . _ - - . , , - . , _ , . . _ . , _ . . - - . , _ . _ - _ - . . _ . . . - . _ _ . . ~ . _ _ _ _ . _ _ . . .
l l
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l Table IV.G-1 RC TEMPERATURE MEASURING DEVICE FOR DETERMINING ,
l RV PRESSURE TEMPERATURE CONDITIONS l l
i RCP RCP OFF ON NAT' L NO NAT'L ,
CIRC CIRC i HPI INCORE INCORE ON T/C T/C HPI Tc Tc 0FF
' /
i I
j NOTES: TABLE NOT APPLICABLE DURING ICC OR LPl/DHRS OPERATION r 1
I 1
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i l 000. NO. 74-1152414-00 i
l
8%NP 20007 3 (9-84)
OABCOCK & wlLCOX NUCLEAR POWER DIVISION 74-1152414-00 TECHICAL DOCUMENT i
f 1
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i, Part V l i Specific Rules '
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! DATE: 7-23-85 PAGE V.0 !
1 4
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l SWNP-20007 3 (9 84) l BASCOCK & WILCCX O
NUCLEAR power DIVISION 4 TECHNICAL DOCUMENT .,,
Chanter V.A HPI/LPI Soecific Rules 1.A.0 HPI Soecific Rules (Applicable to' all plants except Davis-Besse) 1.A.1 HPI must be INITIATED whenever loss of adequate SCM occurs (Refer to Chapter IV.B).
. 1.A.2 When SG heat transfer is not adequate and FW is not avail-able to either SG then HPI must be INITIATED. If two HPI pump flow cannot be achieved, then the PORV must be opened.
- (Refer to Chapter IV.B).
1 1.A.3 Whenever the SCM is lost MAXIMUM HPI flow must be provided to the RCS. (Refer to Chapter IV.B) .
1.A.4 HPI flow must be THROTTLED to prevent over pressurizing the RCS when the SCM exists by keeping the RC pressure below the appropriate limit:
a) the RV pressure-temperature limit given in the plant technical specifications or b) the PTS limit. (Refer to Chapter IV.B).
1.B.O HPI Soecific Rules (Applicable to Davis-Besse) l 1.B.1 HPI/MU must be INITIATED whenever loss of adequate SCM f occurs. (Refer to Chapter IV.B).
l
I The PORV must be opened when incore T/C temperature reaches 600F. (Refer to Chapter IV.B). [
1.B.3 HPI/MU flow must be MAXIMIZED whenever the SCM is lost.
(Refer to Chapter IV.B). ;
i DATE: 7-23-85 PAGE ' V. A-1 !
I t
BWNP 20007 3 (9 84)
BASCOCK & WILCOX NUCLEAR POWER DIVISION NUMSER TECHNICAL DOCUMENT 1.B.4 HPI/MU flow must be THROTTLED to prevent overpressurizing the RCS when the SCM exists by keeping the RC pressure below the RV pressure-temperature limit given in the plant technical specifications. (Refer to Chapter IV.B).
2.0 LPI SDecific Rules 2.1 LPI must be INITIATED whenever any of the LPI initiation setpoints are reached. (Refer to Chapter IV.B) 2.2. LPI suction must be SWITCHED from the BWST to the reactor building emergency sump when switchover conditions are met. (Refer to Chapter IV.B).
O s
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, DATE: 7-23-85 PAGE V.A-2 l
I
i SwNP 20007 3 (9 84)
BADCOCK & WILCox NUCLEAR POWER DIVISION 74-1152414-00 TECHNICAL DOCUMENT .. -
~ '
Chaoter V.B MFW/AFW Soecific Rule j 1. If the RC SCM is lost, the SG level must be CONTROLLED at or above the loss of SCM setpoint (Refer to Chapter IV.C). !
- 3. When the reactor is shutdown, the MFW flow must be CON- !
TROLLED to prevent the SG level from exceeding the shutdown (
- SG overfill setpoint (Refer to Chapter IV.C). [
- O l
disrupted causing the SG water level to decrease to the AFW l l start SG level setpoint (Refer to Chapter IV.C). [
- 6. AFW must be STARTED whenever there is inadequata primary to f
secondary heat transfer or whenever the SG level must be l controlled at the loss of SCM setpoint (Refer to Chapter !
- IV.C). ,
i
i
- 8. If AFW flow should be added, it must be CONTROLLED so that a f
decreases with an overall progression toward the required !
setpoint (Refer to Chapter IV.C). !
\ i DATE: 7-23-85 PAGE ' V .B-l' [
i
1 I
BCNP 20007 3 (9 80) sAncocs a wncox NUMSit NUCLEAR POWER DIVISION 74-1152414-o0 ,
TECHNICAL DOCUMENT
O f
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DATE: 7-23-85 PAGE V.0-2
SWNP-20007 3 (9 84) l sAeCoCK & wtLCom NUCLEAR POWER DIVISION N St l
TECHICAL DOCUMENT Chanter V.C RCP Specific Rule
- 1. When the adequate SCM is lost, all RCPs must be tripped immediately.
Exception - If thes RCPs are not tripped within two minutes after losing adequate SCM, then reduce the number of operating RCPs to one in each loop (Refer to Chapter IV.A).
l DATE: 7-23-85 PAGE V.C-1
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B'%NF 20007 3 (9 84) i SABCOCK & wlLCOE NUCLE AR POWER DIVISION NM88 74-1152414-oo
! TECHICAL DOCUNENT i
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! PART VI i
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3%NP 20007 3 (9 84)
BASCOCK & WILCOE MuttR ;
! NUCLEAR POWER OlVISION i
74-1152414-oo -
TECHICAL DOCUMENT k
I PART VI l REFERENCES j i The following list of references represents information (calcula- ;
,! tional Packages, Reports, Transient Information Documents, Analyti- f
} cal Input Summaries) created by Babcock & Wilcox during the j f performance of the Abnormal Transient Operating Guidelines, Plant Cooldown Guidelines and Multiple Steam Generator Tube Rupture ;
f Guidelines contracts. Each reference provides information on specific questions that were being answered for each contract, but [
f i
t a one-for-one correlation between the information contained in this !
- Technical Bases document and the references by themselves cannot be j made. Engineering judgement was used in making the transition {
! between the strict, detailed analyses and the guidance provided in l i ATOG and the Technical Bases Document. Furthermore, material in !
! some of the documents is of more theoretical than of practical i interest for various reason (e.g., overly conservative assumptions, j subsequent design changes, etc.). [
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CWNP 20007 3 (9 84)
S ABCOCK & WILCOX NUCLEAR POWER Divisl0N 74 1152414-00 TECHNICR DOCUMENT References 1.0 General Plant Comparisons 1.1 Document No. 51-1123827-0, " Comparison of DB-1 and ANO-1 for ATOG," P. R. Boylan. Comparison of select plant parameters significant to ATOG between ANO-1 and DB-1. Facilitated modifying the ANO-1 FOAK ATOG document to reflect the DB-1 plant. Technical bases for comparison provided by list of references contain-ed in this document.
1.2 Document No. 51-1121907-0, "ATOG Comparison Between ANO-1 and CR-3 Nuclear Stations," E. A. Hiltunew.
Same as Reference 1.1 above except comparison between ANO-1 and Crystal River-3 plants.
1.3 Document No. 32-1121199-0, "ATOG ANO-1/TMI-1 Systems Comparison," P. R. Boylan. Same as Reference 1.1 above except comparison is between ANO-1 and TMI-1 plants.
I 1.4 Document No. 32-1120675-01, "Conaparison of ANO-1 and I ONS-III for ATOG," E. Hiltunew, M. Benak, E. Kuhr, 9/17/81. Function: similar to Reference 1.1 above except applies to Oconee 1, 2, and 3 plants.
1.5 Document No. 32-1106885-01, "ATOG ANO-Rancho Seco Systems Comparison," L. Rudy, D. Newton, 11/18/83.
Function: similar to Reference 1.1 above except applies to Rancho Seco plants.
2.0 Excessive Main Feedwater Event 2.1 Document Nc. 32-1106880-0, " Excessive FW - Success Path with ESAS - ANO Path," K. J. Vavarok. Function:
i supporting document for ATOG Part I guidance and Part II discussions dealing with detection and mitigation l of an excessive FW event on the ANO-1 plant.
2.2 Document No. 86-1117945-0, " Excessive FW - Event Description and Analytical Output to ANO - Main Success Path with ESAS," K. J. Vavarek. Function:
similar to that of Reference 2.1 above.
2.3 Document No. 86-1118690-0, "ANO - Excessive FW - Main Success Path - Path 1," K. J. Vavarok. Function:
similar to that of Reference 2.1 above.
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SABCOCK & WMCOM O
NUCLEAR POWER DIVISION ,
i TECHNICAL BOCullENT .= a n 4,
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! 2.4 Document No. 32-1106881-0, " Excessive FW - Success i Path without ESAS - Path 1 ANO," K. J. Vavarek.
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. Function: similar to that of Reference 2.1 a'bove.
2.5 Document No. 86-1119078-0, "ANO - Excessive MFW with MSSV Failed Open," P. R. Boylan. Function: similar l
j- to that of Reference 2.1 above.
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j 2.6 Document No. 86-1119443-0, "ATOG/ANO Excessive Main 1
Feedwater with No Emergency FW," K. J. Vavarek. Func-tion: similar to that of Reference 2.1 above. ;
q 2.7_ Document No. 86-1119075-0, "ATOG/ANO Excessive Main Feedwater with Excessive Emergency FW," K. J.
!j Vavarek. Functions similar to that of Reference 2.1 above.
i 1 2.8 Document No. 86-1119073-0, " Excessive FW with Turbine l
Bypass Failure," K. J. Vavrek. Function: similar to i
! that of Reference 2.1 above.
4 j 2.9 Document No. 86-1119074-3, "ANO - Excessive Main (FW) '
4 Success Path with Excessive Makeup," K. J. Vavarek.
Function: similar to that of Reference 2.1 above. '
I l 2.10 Document No. 32-1106882-0, " Excess. FW - Turbine
- Bypass' System Fails open - ANO Path 2," K. J. ,.
Vavarak. Function:
similar to that of Reference 2.1 l above. ,
h 1 2.11 D2cument No. 86-1125690-0,-"DB-1 Excessive Main
} Feedwater Transient Information Document," L. J.
I Rudy. Functions similar to that of Reference 2.1 l above except for Davis Besse-1 plant.
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j 2.12 Document No. 86-1124195-0, "CR-3 Excessive Main j Feedwater Transient Information Document," L.'J.
j Rudy. Functions similar to that of Reference 2.1 j above except for Crystal River-3 plant.
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) 2.13 Document No. 86-1123876-0, "TMI-l Excessive Main Feed-1 water Transient Information Document," L. J. Rudy.
I Function: similar to that of Reference 2.1 above l except for Three Mile Island Unit 1.
2.14 document No. 86-1127307-01, "ATOG Transient Info. -
Doc., Excessive Main FW, Rancho Seco," L. J. Rudy.
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! Functions similar to Reference 2.1 above except for '
Rancho Seco plant.
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SASCOCK & WILCOX NUCLEAR POWER DivlSION ucmen nocuum g
2.15 Document No. 86-1132609-0, " Midland-2 ATOG - Ex MFW TID," R. B. Brownell. Function: similar to Reference 2.1 above except for Midland Unit 2.
2.16 Document No. 86-1122759-0, "Oconee TID for Excessive Main Feedwater," K. J. Vavarek. Function: similar to Reference 2.1 above except for Oconee Nuclear Station Unit 1.
2.17 Document No. 32-1106884-0, "Oco Exc. MFW w/o Emerg SFGDS ACT S FW TRP - Hi OTSG LVL," K. J. Vavarok.
Function: similar to Reference 2.1 above except for Oconee Nuclear Station Unit 1.
2.18 Document No. 79-1100902-0, "ANO EX FW Event Tree,"
K. J. Vavarek. Function: provides logical evaluation and confirmation of detection and mitigation techni-ques for an excess FW event on the ANO-1 plant.
Serves as a basis for the ATOG Part I guidelines and for the Part II discussion dealing with excess FW events.
2.19 Document No. 79-1121461-0, "SMUD EMFW Event Tree,"
K. J. Vavarek. Function: same as for Reference 2.18 above except for Rancho Seco Plant.
2.20 Document No. 79-1120034-0, "Oconee - EMFW Event Tree,"
K. J. Vavarck. Function: same as for Reference 2.18 above except for Oconee plant.
2.21 Document No. 79-1121312-0, " Davis Besse - EMFW Event Trees," K. J. Vavarek. Function: same as Reference 2.18 above except for Davis Besse-1 Plant.
2.22 Document No. 79-1121415-00, " Crystal River - EMFW Event Tree," L. J. Rudy. Function: same as for Reference 2.18 above except for Crystal River-3 plant.
2.23 Document (B&W Dwa) No. 1121221F-00, "TMI-l Excess Main Feedwater Event Tree." Function: same as for Reference 2.18 above except for TMI-l plant.
2.24 Document No. 79-ll28213F-00, " Excess MFW Event Tree -
Midland Unit 1." function: same as for Reference 2.18 above acept for Midland Units 1 and 2.
2.25 Document No.~79-1120137-0, " Excessive Main Feedwater Event Tree (THI) , " L. J. Rudy. Functions same as for Reference 2.18 above except for TMI-l plant.
DATE: 7-23-85 PAGE VI-4
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B?!NP 20007 3 (9 84) f
, BASCOCK & wtLCOX , , , ,. i NUCLEAR POWER DIVISION 4 1152414-00~
l TECHNICAL DOCUMENT u ye: w 2.26 Document No's. 2-1094608-0 throuah 2-1094613-0,.
] " Excessive Feedwater Safety Sequence Diagram," " ,
I supplied for customer-(ANO-1) by EDS Nuclear.
I contents: customer supplied information' relating to sequences of equipment operation and plant response :
{ following an excessive feedwater event. Applicable to l 4 ANO-1 plant. !
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! 2.27 Document No's. 2-1094615-0 throuch 2-1094619-0, l
" Excessive Feedwater Safety Sequence Diagram," }
supplied for customer (Duke Power. company) by EDS [
i Nuclear. Contents: similar to Reference 2.26 above l~
. except applicable to Oconee Nuclear Station.
i 2.28 Document No's. 2-1094666-0 throuah 2-1094670-0, t
! " Excessive Feedwater Safety Sequence Diagram," i i
supplied for customer (TMI-1) by EDS Nuclear. !
Contents: same as Reference 2.26 above except !
) applicable to TMI-1 plant.
t l 2.29 Document No's. 2-1094736-0 throuah 2-1094739-0, i " Excessive Feedwater Safety Sequence Diagram," ,
l supplied for customer (DB-1) by EDS Nuclear. Con- [
- tents: same as Reference 2.26'above except applicable ;
to Davis Besse-1 Nuclear Station.
2.30 Document No ss. 2-1094859-0 throuch 2-1094863-0,
. EDS Nuclear. Contents: same as Reference 2.26 above l I
except applicable to Crystal River-3 Nuclear Plant.
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! 2.31 Document No. 2-1094964B-00, "SMUD Rancho Seco Unit 1
- Control Logic Diagram." Contents
- similar to l Reference 2.26 above except applicable to Rancho Seco
! Nuclear Plant.
1 2.32 Document No. 32-1119324-00, "ANO ATOG Test for TRAP l Steam Pressure Response," P. R. Boylan. Function: l
! similar to Reference 2.1 above. ;
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i 2.33 Document No. 86-1125714-0, "OTSG Level Curves for DB-1 l l Excessive Main Feedwater." Function: similar to i
{ Reference 3 1 above except for Davis Besse-1 Plant.
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[WNP 20007 3 (9-84)
S ABCOCK & witCOx NUMtit NUCLEAR POWER Divisl0N 74-1152414-00 TECHNICAL DOCUMENT 3.0 Loss of Main Feedwater Event 3.1 Document No. 32-1106474-0, "ANO-1 ATOG LOFW-TBS Failure," J. M. Knoll. Function: supporting document for ATOG Part I guidance and Part II discussions dealing with detection and mitigation of a loss of feedwater event in the ANO-1 plant.
3.2 Document No. 32-1106476-0, "ANO-ATOG LOFW Path 6 (Excessive EFW)," P. R. Boyland. Function: similar to Reference 3.1 above.
3.3 Document No. 32-1105475-0, "ANO-ATOG Loss of Maia FW/ Failure of EFW," P. R. Boyland. Function: similar to Reference 3.1 above.
3.4 Document No. 32-1106473-0, " Loss of MFW Transient Where MSSV Fails to Reseat," J. M. Knoll. Function:
similar to Reference 3.1 above.
3.5 Document No. 86-1119380-0, " Transmittal: ANO-1 ATOG LOFW MSSV Fails Open, NO ESAS," J. M. Knoll. Func-tion: similar to Reference 3.1 above.
3.6 Document No. 86-1119483-0, " Transmittal: ANO-1 ATOG LOFW PZR Spray Fails On," J. M. Knoll. Function:
similar to Reference 3.1 above.
3.7 Document No. 32-1128491-2, "TVA Loss of Feedwater Main Success Path TRAP Analysis," R. L. Bright. Function:
Similar to Reference 3.1 above except for TVA Belle-fonte 1 and 2 plants.
3.8 Document No. 32-1130818-0, "TVA ATOG Total Loss of Feedwater Analysis, "R. L. Bright. Function: Similar to Reference 3.1 above except for TVA Bellefonte 1 and 2 plants.
3.9- Document No. 32-1131486-0, "TVA ATOG LOFW Excessive AVW," R. L. Bright. Function: Similar to Reference 3.1 above except for TVA Bellefonte 1 and 2 plants.
3.10 Document No. 86-1131779-0, "TVA ATOG Total LOFW Path,"
R. L. Bright. Function: Similar to Reference 3.1 above except for TVA Bellefonte 1 and 2 plants.
3.11 Document No. 86-1131930-1, "TVA ATOG LOFW with Excessive AFW (Transient Information Document) TID,"
R. L. Bright. Function:
Similar to Reference 3.1 above except for TVA Bellefonte 1 and 2 plants.
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""0IE l NUCLEAR POWER OtVISION 74-1152414-00 l TECHNICAL DOCUMENT u.
l s, Document No. 51-1151135-0, "TVA ATOG Analysis; LOFW; l 3.12 Delete (Reactor Trip) RT on MFW Pump Trip," R. W.-
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! Moore. Function: similar to Reference 3.1 above except for TVA Bellefonte 1 and 2 plants.
3.13 pg_cument No. 86-1134278-1, "TVA ATOG LOFW w/ Loss of 4
Primary Inventory Control (Low), L. J. Rudy. Func-
! tion: similar to Reference 3.1 above except for TVA '
4 Bellefonte'l and 2 plants.
3.14 Document No. 86-1132897-1, "TVA ATOG Loss of Main ,
l Feedwater w/ Low Secondary Pressure," L. J. Rudy.
j~ Function: similar to Reference 3.1 above except for i TVA Bellefonte 1 and 2 plants.
3.15 Document No. 86-1131779-1, "TVA ATOG Total LOFW Path,: L. J. Rudy. Function: similar to Reference
- 3.1 above except for TVA Bellefonte 1 and 2 plants.
3.16 Document No. 86-1130793-1, "TVA ATOG LOFW Main Success.
j Path," L. J. Rudy. Function: similar to Reference i 3.1 above except for TVA Bellefonte 1 and 2 plants. t 4 3.17 Document No. 86-1134279-1, "TVA ATOG LOFW with Loss of Offsite Power," L. J. Rudy. Function: similar to Reference 3.1 above except for TVA Bellefonto 1 and 2 plants.
3.18 Document No. 86-1131527-0, " Midland-2 LOFW (Transient J Information Document) TID," B. L. Bowman. Function:
.j similar to Reference 3.1 above except for Midland i
Units 1 and 2.
3.19 Document No. 32-1123961-0, "ATOG (Davis Besse) DB-1 i LOFW and Failure of AFW TRAP Analysis," P. R. Boylan.
Function: similar to Reference 3.1 above except for
! Davis Besse (DB) -1 plant.
1 3.20 Document No. 86-1123962-1, "DB-1 ATOG Transient Information Document for LOFW," P. R. Boylan.
I Function: similar to Reference 3.1 above except for i DB-1 plant.
3 3.21 Document No. 32-1146244-0, " Supply System LOFW/ LOOP i (Loss of Feedwater/ Loss of Offsite Power) ATOG !
Analysis," L. J. Rudy. Function: similar to Refer-i ence 3.1 above except for Supply System Plants WNP-1
' and WNP-4. Also applicable to portions of ATOG j dealing with loss of offsite power events.
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BONP-30007 3 (9 84)
BABCOCK & wlLCOE NUCLEAR POWER DIVI $10N NUMBER 74-1152414-00 TECHNICAL DOCUMENT 3.22 Document No. 86-1146245-0, "SS LOFW/ LOOP ATOG Trans-ient Information Document," L. J. Rudy. Function:
same as Reference 3.21 above.
3.23 Document No. 86-1125515-0, "CR-3 (Crystal River-3)
Loss of Feedwater Transient Information Document," R.
B. Brownell. Function: similar to Reference 3.1 above except applicable to CR-3 plant.
3.24 Document No. 86-1117655-0, "ATOG LOFW - Main Success Path," J. M. Knoll. Function: similar to Reference 3.1 above except generically applicable to B&W's 177 FA Lowered Loop Plants.
3.25 Document No. 32-1123291-0, "ATOG Analysis for TNI-1 LOFW Event, TBS Failed Open," R. B. Brownell.
Function: similar to Reference 3.1 above except applicable to TMI-1 plant.
3.26 Document No. 79-1100937-0, "ANO LOMFW Event Tree,"
J. M. Knoll. Function: provides logical evaluation and confirmation of detection and mitigation techni-quos for a loss of feedwater event in the ANO-1 plant. Serves as a basis for the ATOG Part I guide-lines and for the ATOG Part II discussions dealing with loss of feedwater events.
3.27 Document No. 79-ll213 8Q-Q, "SMUD ' LOMFW Event Tree, "
P. R. Boylan. Function: similar to Reference 3.26 above except applicable to the Rancho Seco plant.
l 3.28 Document H_o. 79-1126191-0, "TVA - LOMFW Event Tree."
Function: similar to Reference 3.26 above except applicable to the TVA Bellefonte 1 and 2 plants.
3.29 Document No. 79-1120030-0, "Oconee - LOMFW Event
! Tree," P. R. Boylan. Function: similar to Reference 3.26 above except applicable to the Oconee 1, 2, and 3 plants.
3.30 Document No. 79-1121281-0, " Davis Besse - LOMFW Event Tree," P. R. Boylan. Function: similar to Reference 3.26 above except applicable to the Davis Besse-1 plant.
3.31 Document No. 79-1121345-0, " Crystal River - LOFW Event Tree," R. B. Brownell. Function: similar to Refer-ence 3.26 above except applicable to the Crystal River-3 plant.
O 7-23-85 VI-8 DATE: PAGE
i 8 NP-20007 3 (9 34) i SABCOCK & WitCOR N W $tt NUCLEAR POWER DIVl$lON 74-1152414-00 j TECHICAL 00CUMEllT 3.32 B&W Drawina No. 1120125F-0, "Three Mile Island Unit il One (TMI-1) Loss of Main Feedwater Event Tree," P. R.
1 Boylan. Function: similar to Reference 3.26 above except applicable to the TMI-l plant.
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! 3.33 B&W Drawina No. 79-1128211-0, " Midland Plant Units 1 and 2 Loss of Feedwater Event Tree." Function:
similar to Reference 3.26 above except applicable to the TMI-1 plant.
i 3.34 Document No. 2-1148492-1, "SMUD Auxiliary Feedwater System SAD (System Auxiliary Diagram)," provided for customer by EDS Nuclear Function: similar to Refer-ence 3.27 above.
3.35 Document No. 86-1118379-2, "EFW Reliability for ANO-4 1," W. Weaver. Function: similar to Reference 3.1 above.
3.36 Document No's. 2-109589-0 throuch 2-1094594-0, : Loss l
i of Feedwater Safety Sequence Diagrams," supplied for
! customer by EDS Nuclear. Contents: customer supplied j information relating to sequence of equipment opera- i a tion and plant response following a loss of feedwater event. Applicable to ANO-1 plant.
5 3.37 Document No. 2-1094638-0, " Loss of Feedwatar Safety Sequence Diagrams," supplied for customer by EDd Nuclear. Contents: similar to Reference 3.36 above. Applicable to Oconee 1, 2, and 3 plants.
1 3.38 Document No. 2-1094712-0, " Emergency FW System i Auxiliary Diagram," supplied for customer by EDS j Nuclear. Contents: similar to Reference 3.36 above.
Applicable to Oconee 1, 2, and 3 plants.
i l 3.39 Document No. 2-1094679-0, " Emergency Feedwater System Auxiliary Diagram," supplied for customer by EDS Nuc-lear. Contents: similar to Reference 3.36 aobve.
Applicable to THI-1 plant.
3.40 Document No's. 2-1094687-0 throuah 2-1094691-0, " Loss of Normal Feedwater Safety Sequence Diagram," supplied for customer by EDS Nuclear. Applicable to TMI-l plant. Contents: similar to Reference 3.36 above.
! 3.41 Document No's. 2-1094720-0 throuah 2-1094723-0, " Loss of Feedwater Safety Sequence Diagram," supplied for customer by EDS Nuclear. Applicable to Davis Besse-1 plant. Contents: similar to Reference 3.36 above.
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BINP 20007 3 (9 84) 8 ASCOCK & wnCOE NUMBit NUCLE AR P0wtR OtVISION 74-1152414-00 TECHNICAL DOCUMENT 3.42 Document No. 2-1094821-0, " Auxiliary Foodwater System Auxiliary Diagram," supplied for the customer by EDS Nuclear. Contents: similar to Reference 3.1 above.
Applicable to Davis Basso-1 plant.
3.43 Document No's. 2-1094840-0 throuch 2-1094845-0, " Loss of Foodwater Safety Sequence Diagram," supplied for the customer by EDS Nuclear. Contents: similar to Refrence 3.36 above. Applicable to the Crystal River-3 plant.
3.44 Document No. 2-1094873-0, " Emergency Foodwater System Auxiliary Diagram," supplied for the customer by EDS Nuclear. Contents: similar to Referenco 3.1 above.
Applicable to the Crystal River-3 plant.
3.45 Document No's. 2-1094965-0 throuch 2-1094967-0, " Loss of Focdwater Sfoty Sequence Diagram," supplied for the customer by EDS Nuclor. Contents: similar to Reference 3.36 above. Applicable to the Rar.cho Seco plant.
3.46 Document No. 86-1118379-2, "EFW Rollability for ON-1 Nucl Gen. Station, Units 1, 2, 3," W. Weavor.
Functions similar to Reference 3.1 above. Applicable to Oconeo 1, 2, and 3 plants.
3.47 Document No. 2-1107643-0, " Diagrammatic Layout of FW System." Contents: similar to Reference 3.36 above.
3.48 Document No. 86-1122498-00, " Transient Information Document for Loss of Foodwater Event Oconeo Nuclear Station Unit III," R. B. Brownell, M. E. Newlin, 12/12/80. Function: similar to Reference 3.1 above. Applicable to Oconeo 1, 2, and 3 plants.
3.49 Document No. 86-1127426-01, " Rancho Seco ATOG LOFW TID," L. Rudy, D. Newton, 11/22/83. Function:
similar to Reference 3.1 above. Applicable to Rancho Seco plant.
3.50 Documpnt No. 32-1130653-0, " Loss of Foodwater - Coro Cooling," J. Seals, W. Bloomfield, 1/12/82. Func-tion: similar to Reference 3.1 above. Specifically, used to verify that operator actions specified in guidelines will result in adequato cooling following a LOFW. Gonorically applicable to all B&W plants.
O 7-23-85 '~
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/~~'s TECHNICAL DOCUMENT
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3.51 Document No. 32-1103583-00, " Loss of Feedwter Acci-dent," M. G. Gharakham, J. R. Dieterman, 8/8/79.
Function: similar to Reference 3.1 above. Specifi-cally, provides verification that HPI flow will provide adequate core cooling (prevent core uncover-ing) during a total loss of feedwater if HPI gr AFW is actuated within 20 minutes after the event starts.
3.52 Document No. 32-1127513-01, " Davis Besse-1 Loss of All Feedwater Transient with Startu-up Feedwater Pump,"
T. Geer. Functions similar to Reference 3.1 above; specifically, relates to operator actions and start-up feedwater pump performance during a total loss of feedwater event. Applicable to the Davis Besse-1 plant.
3.53 Document No. 32-1100479-00, "TMI-2 LOFW - CADDS Evaluation," R. Vosburgh, 5/9/79. Deals with bench-marking of CADDS code against data from 3.28.79 TMI-2 transient.
3.54 Document No. 32-1102465-0, " Complete Loss of Foodwater Transient," M. Haghi, 8.29.01. Functions similar to e ~x Reference 3.1 abover applicable to Davis Besse-1
( )/ plant.
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4.0 Loss of Offsite Power Event 4.1 Document No._32-1106955-1, "ATOG - PZR Pressure Response During LOOP (Loss of Offsite Power)/ Main Success Path," M. E. Newlin. Functions supporting document for ATOG Part I guidance and Part II discus-sions related to detection and mitigation of a loss of offsite power event. Applicable to the ANO-1 plant.
4.2 Document No. 32-1106954-0, " Loss of Offsite Power /Ex-cessive AFW - ATOG," M. E. Newlin. Function: similar to Reference 4.1 above. Applicable to the ANO-1 plant.
4.3 Document No. 32-1106953-0, " Loss of Offsite/Onsite Power - No EFW - ATOG,2 M. E. Newlin. Functions similar to Reference 4.1 above. Applicable to the ANO-1 plant.
4.4 Document No. 32-1106949-0, " Main Success Path, Loss of Offstie Power," M. E. Newlin. functions similar to Reference 4.1 above. Applicable to the ANO-1 plant.
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82NP 20007 3 (9 84)
S ASCCCK O W3(CCR NUMS t B NUCLEAR POWER DivtSION 74-1152414-00 TECHNWAL DOCUMENT 4.5 Document No. 86-1119369-0, "ATOG Excessive AFW/ LOOP,"
M. E. Newlin. function: similar to Reference 4.1 above. Applicable to the ANO-1 plant.
4.6 Document No. 86-1119367-0, "ATOG Loss of Off-site Power Exc. Makeup with EFW,2 M. E. Newlin. Function:
similar to Reference 4.1 above. Applicable to the ANO-1 plant.
4.7 Document No. 86-1119255-0, " Main Success Path: Loss of Offsite Power - ATOG,: M. E. Newlin. Function:
Similar to Reference 4.1 above. Applicable to ANO-1 plant.
4.8 Document No. 86-1119582-0, "ATOG LOOP /NO CCS (Diosol Generator) , " M. E. Nowlin. Function: similar to Reference 4.1 above. Applicable to ANO l plant.
4.9 Document No. 86-1118159-1, "ANO Loss of Onsite/Offsite
- No EFW ATOG," M. E. Newlin. Function: similar to Reference 4.1 above. Applicable to ANO-1 plant.
4.10 Document No. 86-1131520-0, "CPCo Midland 2 LOOP transient Information Document," R. S. Talley.
Functions similar to Hoforence 4.1 above. Applicable to Midland Units 1 and 2.
4.11 Document No. 86-1125976-1, "CR-III ATOG LOOP TID,"
M. E. Newling. Function: similar to Referenco 4.1 above. Applicable to Crystal River-3 plant.
4.12 Document No. 32-1120510-0, "TMI-l Loss of Offsite/On-site PWR: 1 MSSV Failed Opon on S/G-B," E. A.
Hiltunow. Function: similar to Reference 4.1 above.
Applicable to TMI-1 plant.
4.13 Document No. 86-1124178-0, "ATOG Loss of Offsito Power Transient Information Document," M. E. Newlin. Func-tion: similar to Reference 4.1 above. Generically applicable to all B&W 177 FA loworod loop plants.
4.14 Document No. 86-1123921-0, "ATOG TID: Loss of Offsito/Onsite Power, THI-1," E. A. Hiltunow.
Functions similar to Referenco 4.1 above. Applicable to THI-l plant.
4.15 Document No. 86-1122497-0, "ATOG TID / Loss of Of fsite Power," M. E. Newlin. Function: similar to Reference 4.1 above. Applicable to Oconeo 1, 2 and 3 plants.
O 05 VA-12 DATE: PAGE 1
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sAscoca a witcos 7
muctue pown omsson
{
4 TECNNICAL ISC5NENT I
4.16 Document No. 86-1125435-0, " Analytical Results for I
TMI-l Loss of Offsite/Onsite Power. Function:
similar to Reference 4.1 above. Applicable to TMI-1 l
. plant. j
, f 4.17 Memo. M. E. Newlin to Distribution, " Analytical Input l Summary for Loss of Offsite/Onsite Power at Rancho l Seco," 10/27/80, File No. NSS-11/T3.4. Discusses IDOP I and comparison of plant responses between ANO-1 and i Rancho Seco. Concludes that ANO-1 and Rancho Seco ,
{
systems are similar and that no specific LOOP analyses f l
are required for Rancho Seco. See Reference 4.1 !
above. i I' !
) 4.18 Document No. 86-1124178-0, "ATOG Loss.of Offsite Power l l
Transient Information Document," M.-E. Newlin. Func- [
l tion: similar to Reference 4.1 above. Applicable to i l Davis-Besse 1 plant.
i 4.19 Document No. 79-1100949-0, "ANO-IDOP Event Tree,"
l
- G. P. Bennett. Functions provides logical evaluation i
- and confirmation of detection and mitigation techni- i i ques for a loss of offsite power event at the ANO-1 !
plant. Serves as a basis for the ATOG Part I guidance !
f
]. and Part II discussions dealing with loss of offsite
! power.
4.20 Document Nc. 79-1121480-0, "SMUD - ICOP Event Tree," ;
M. W. Newlin. Similar to Reference 4.19 above except }
applicable to Rancho Seco plant. I i
l 4.21 Document No. 79-1120026-A, "Oconee - ICOP Event Tree," [
! M. E. Newlin. .Similar to Rference 4.19 above except l applicable to Oconee 1, 2 and 3 plants.
4.22 Document No. 79-1120153-0, "ATOG/IDOP/ Event Tree,"
i E. A. Hiltunaw. Similar to Reference 4.19 above i except applicable to TMI-1' plant.
l
! 4.23 Document No. 79-1121253-0, " Davis-Besse - IDOP Event l Tree," M. E. Newlin. Similar to Reference 4.19 above
! except applicable to Davis-Besse 1 plant.
i
! 4.24 Document No. 79-1121357-0, " Crystal River - I4OP Event l ,
Tree," E. A. Hiltunew. Similar to Reference 4.19 ,
above except applicable to Crystal River-3 plant. }
o 4.25 Document No.'79-1128212-00, " Midland Plant Units 1 and
! 2, Loss of Offsite Power Event Tree." Similar to
! Reference 4.19 above except applicable to Midland 1 ;
4 and 2 plants. i
)
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BWNP-200074 (9-84)
BASCCCE a WILCoX I
NUCLEAR POWER DIVISION TECHNICAL DOCUMENT 4.26 Document Nos. 2-1094602-0 throuch 2-1094607-0, " Loss of Offsite A/C Power Sequence Diagram," supplied for customer by EDS Nuclear. Contents: Customer supplied information relating to sequence of equipment failure and plant response following a loss of offsite power event. Applicable to ANO-1 plant.
4.27 Document Nos. 2-1094813-0 throuch 2-1094818-0, " Loss of Offsite Power Safety Sequence Diagram." Similar to Reference 4.26 above. Applicable to TMI-l plant.
4.29 Document Nos. 2-1094724-0 throuah 2-1094729-0, " Loss of Offsite AC Power Safety Sequence Diagram." Similar to Reference 4.26 above. Applicable to Davis-Besse 1 plant.
4.30 Document Nos. 2-1094846-0 throuch 2-1094851-0, " Loss of Offsite AC Power Safety Sequence Diagram." Similar to Reference 4.26 above. Applicable to Crystal River-3 plant.
4.31 Document No. 2-1094970-0, "SMUD - LOOP SSD." Similar to Reference 4.26 above. Applicable to Rancho Seco plant.
4.32 Document No. 86-1126639-01, " Rancho Seco ATOG LOOP TID," L. Rudy, 10/8/81. Function: similar to Reference 4.1 above. Applicable to Rancho Seco plant.
4.33 Document No. 86-1137718-00, "B&W Plant Transient Prediction," J. C. Seals, N. K. Savani, 11/2/82, Contract No. 582-7198. Benchmarking of LOCA analysis l code against ANO-1 plant loss of offsite power event, l 6/24/80. Indicated that code could accurately predict phenomena of LOOP event. Increased confidence in LOCA-associated trends predicted by code and in the use of code results as a basis for ATOG.
4.34 Document No. 86-1125356-0, "P-T Plots - TMI-1 ATOG -
ANO-1 LOOP Data Plotted per THI-1 Format," E. A.
Hiltunew. Function: Loss of Offsite Power plant response data plotted on a simulated P-T display; illustrated the plant response to a LOOP event as it would be displayed in the control room. A basis for the P-T diagram in the LOOP discussion. Applicable to ANO-1 and TMI-1 plants.
4.35 Document No. 86-1126639 ,QR, " Rancho Seco ATOG LOOP TID," E. Hiltunew. Function: similar to Reference 4.1 above. Applicable to Rancho Seco plant.
~
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EWNP 20007 3 (9-84)
SASCOCK & witcom NUM$tt NUCLEAR POWER DIVISION 74-1152414-00
/
V; TECHNICAL DOCUMENT 5.0 Small ,cteam Line Break Event 5.1 Document No. 32-1106466-00, "SSLB-Event Description of Success Path #1," R. J. Schomaker. Function:
Supporting document for ATOG Part I gyldelines and Part II discussions related to detection and mitiga-tion of a small steam line break. Applicable to the ANO-1 plant.
5.2 Document No. 86-1117538-00, "SSLB-Event Description for Success Pasth #1," R. J. Schomaker. Function:
similar to Reference 5.1 above. Applicable to ANO-1 plant.
5.3 Document No. 86-1118050-00, "DYSID vs.15UUP Comparison fcr ATOG SSLB Success Path 1," M. S. Kai. Function:
similar to Reference 5.1 above. Applicable to ANO-1 plant.
5.4 Document No. 86-1106760-00, " Power Train Analysis of Steam Line Break," J. M. Knoll. Function: similar to Reference 5.1 above. Applicable to ANO-1 plant.
) 5.5 Document No. 86-1123200-00, " Transient Information s- / Document for Small Steam Line Break Events at ONS-III," R. H. Ellison. Function: similar to Reference 5.1 above. Applicable to Oconee 1, 2, and 3 plants.
5.6 Document No. 86-1123784-00, "TMI-l ATOG Transient Information Document for SSLB," J. S. Schwenn. Func-tion: similar to Reference 5.1 above. Applicable to TMI-l plant.
5.7 Document No. 51-1120507-00, " Crystal River III ATOG Data Base," M. Lockey. Function: compilation of available information for the ATOG Data Base - Crystal River III.
5.8 Document No. 86-1125293-00, "CR-3 Small Steam Lino Break Event Transient Information Document," J. S. -
Schwenn. Function: similar to Reference 5.1 above.
Applicable to Crystal River-3 plant.
5.9 Document No. 86-1127102-01, "ATOG SMUD SSLB Transient Information Document," L. J. Rudy. Function: similar to Reference 5.1 above. Applicable to Rancho Seco plant.
O 7-23-85 VI-15 DATE: PAGE l
BWNP 20007 3 (9-84)
BABCOCK Q wlLCCX EU NUCLEAR POWER OlvlSION 74-1152414-00 TECHNICAL DOCUMENT 5.10 Document No. 86-11125543-00, "DB-1 ATOG SSLB TID,"
R. H. Ellison. function: similar to Reference 5.1 above. Applicable to Davis-Bess-1 plant.
5.11 Document Nos. 32-1137021-00 throuch -02, "CPC SLB DNB Analyses...," M. V. Parece. Function: similar to Reference 5.1 above. Applicable to Midland 1 and 2 plants.
5.12 Document No. 32-1118994-0. "ANO-1 ATOG SSLB Inside Containment," R. J. Schomaker. Function: similar to Reference 5.1 above. Applicable to ANO-1 plant.
5.13 Document No. 32-1131202-0, "ATOG TVA Small Steam Line Break - Stuck Open MSSV," B. L. Bowman. Function:
similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.14 Document No. 32-1131766-01, "ATOG TVA Small Steam Line Break - 2.5 ft2 Unisolable Leak," B. L. Bowman. Func-tion: similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.15 Document No. 86-1131200-01, "ATOG TVA Small Steam Line Break - Main Success Path," L. J. Rudy. Function:
similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.16 Document No. 86-1131201-01, "ATOG TVA Small Steam Leak Break - Stuck Open MSSV," L. J. Rudy. Function:
similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.17 Document No. 86-1131767-02, "ATOG TVA Small Steam Line Break," L. J. Rudy. Function: similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.18 Document No. 86-1132572-01, "ATOG TVA Small Steam Line Break - Loss of Primary Inventory (LOW)," L. J. Rudy.
Function: similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.19 Document No. 86-1132573-02, "ATOG TVA Small Steam Line Break - Loss of Primary Inventory (LOW)," L. J. Rudy.
Function: similar to Reference 5.1 above. Applicable to Bellefonte 1 and 2 plants.
5.20 Document No. 86-1132574-01, "ATOG TVA Small Steam Line Break - Loss of Secondary Inventory Control," L. J.
Rudy. Function: similar to Reference 5.1 above.
Applicable to Bellefonte 1 and 2 plants.
7-23-85 -6 DATE: PAGE 1
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I j s!NP 20007 3 (9 84)
I
- sAscoca a witcom N O NUCLEAR POWER DivlSION TECHNICAL DOCUMENT ;
i 5.21 Document No. 32-1130515-01, "ATOG TVA Small Steam Line l Break - Main Success Path," J. S. Muransky. Func- i tion: similar to Reference 5.1 above Applicable to Bellefonte 1 and 2 plants. ,
e l 5.22 Document Nos. 2-1094989-0 throuah 2-1154992-0, "SMUD -
l Small Steam Line Break Safety Sequence Diagram," .
supplied for customer by EDS Nuclear. Contents:
i customer supplied information relating to sequence of 1
. equipment failure and plant responses following a small steam line break. Applicable to Rancho Seco '
j plant.
5.23 Document Nos. 2-1094583-0 throuch 2-1094588-0, "Small !
f Steam Line Break Safety Sequence Diagram," supplied i
{ for customer by EDS Nuclear. Contents: similar to J Reference 5.22 above. Applicable to the ANO-1 plant.
4 ,
- 5.24 Document Nos. 2-1094620-00 throuah 2-1094625-00, )
"Small Steam Line Break - Safety Sequence _ Diagram,"
. supplied for customer by EDS Nuclear. Contents:
i similar to Reference 5.22 above. Applicable to the j oconee 1, 2 and 3 plants. l 4
5.25 Document Nos. 2-1094671-00 throuch 2-1094676-00, i "Small Steam Line Break Safety Sequence Diagram,"
! supplied for customer by EDS Nuclear. Contents:
similar to Reference 5.22 above. Applicable to TMI-l .
]
plant.
l 5.26 Document Nos. 2-1094740-00 throuah 2-1094744-00,
- "Small Steam Line Break Safety Sequence Diagram," ,
supplied for customer by EDS Nuclear. Contents:
similar to Reference 5.22 above. Applicable to Davis .
Besse-1 plant. l
-l l 5.27 Document Nos. 2-1094864 throuah 2-1094870-00, *
"Small Steam Line Break Safety Sequence Diagram,"
supplied for customer by EDS Nuclear. Contents:
i similar to Reference 5.22 above. Applicable to Crystal River-3 plant.-
i 5.28 D2sument No. 79-1121462-0, "SMUD Small Steam Line '
t Break Event Tree," J. S. Schwenn. Function: provides I logical evaluation and confirmation of detection and mitigation techniques for a small steam line break; serves as a basis.for guidance and discussions on this j type of event. Applicable to Rancho Seco plant.
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B ASCOCK a w1tCOE NUMBit NUCL[AR POWER DivlSION 74-1152414-00 TECHNICAL DOCUMENT 5.29 Document No. 79-1126281-0, "TVA Small Steam Line Break Event Tree." Function: similar to Reference 5.28 above. Applicable to Bellefonte 1 and 2 plants.
5.30 Document No. 79-1100954-00, "ANO - Small Steam Line Break Event Tree," R. J. Schomaker. Function:
similar to Reference 5.28 above. Applicable to ANO-1 plants.
5.31 Document No. 79-1121381-00, " Davis Besse Small Steam Line Break Event Tree," R. H. Ellison. Function:
similar to Reference 5.28 above. Applicable to Davis Besse-1 plant.
5.32 Document No. 79-1121454-00, "Small Steam Line Break Event Tree," R. H. Ellison. Function: similar to Reference 5.28 above. Applicable to Crystal River-3 plant.
5.33 B&W Drawina No. 1120031F-01, "Oconee III Small Steam Line Break Event Tree." Function: similar to Refer-ence 5.28 above. Applicable to Oconee 1, 2 and 3 plants.
5.34 B&W Drawina No. 1120141F-00, "TMI-1 Small Steam Line Break Event Tree," Function: similar to Reference 5.28 above. Applicable to THI-1 plant.
5.35 Document No. 79-1127643-00, "CPCo SSLB Event Tree for ATOG," J. S. Schwenn. Function: similar to Reference 5.28 above. Applicable to Midland 1 and 2 plants.
5.36 Document No. 86-1131368-00, "CPCo Small Steam Line Break Event Transient Information Document," J. S. -
Schwenn. Function: similar to Reference 5.1 above.
Applicable to Midland 1 and 2 plants.
5.37 Doguient No. 32-11391.76-00, "WPPSS ATOG Small Steam l Line Break - 1 Ft2 Break," M. V. Parece. Function:
similar to Reference 5.1 above. Applicable to WNP 1 and 4 plants.
6.0 Steam Generator Tube Rurture Event 6.1 Document No. 32-1107031-00: "ANO-1 SGTR Transient Analysis (ATOG),: M. Liebmann, 3/10/80. Function:
Analysis of ANO-1 plant response to SGTR using a digital thermal hydraulics code. Generically applic-able to all B&W plants.
~
85 PAGE DATE:
BWN? 2000T3 (9 84)
SABCOCK & watCOE NUM8tR NUCLEAR POWER DivlSION 74-1152414-00 -
TECHNICAL DOCUMENT ON 6.2 Document No. 86-1118041-00: "ANO-1 SGTR Transient Analysis (ATOG)," M. Liebmann. Function: Summary document for Reference 6.1 above. f 6.3 Document No. 32-1118044-00: "ANO-1 OTSG Tube Rupture w/ LOOP Analysis (ATOG) ," M. Liebmann. Function:
Identical to Reference 6.1 except with loss-of-offsite power coincident with rupture. Applicable to all B&W plants.
6.4 Document No. 51-1148397-00: "SGTR Mitigation in 177 FA Plants," M. V. Parece, 6/26/84. Function:
Engineering evaluation document providing guidance and discussions concerning the detection and mitigation of SGTR. Generically applicable to all 177 plants.
6.5 Document No. 32-1148117-00: "177 FA SGTR Best Estimate Analysis," M. A. Haghi, M. V. Parece.
Function: Computer analysis of SGTR on ANO-1 plant design to show that operator actions as per ATOG preclude loss-of-subcooling margin. Generically applicable to all B&W 177-FA lowered loop plants.
p 6.6 Document No. 86-1148118-00: "177 FA SGTR Best Estimate Analysis," M. V. Parece. Function: Summary i
document for Reference 6.5 above.
6.7 Document No. 32-1150650-0,Q: " Davis-Besse REDBL5 SGTR Analysis," D. W. throckmorton. Function: Identical to Reference 6.5 above but applicable to TED's Davis-Besse plant only.
, 6.8 Document No. 77-1152840-00: "Best Estimate Steam Generator Single Double-Ended Tube Rupture Analysis."
Function: Summary report for Reference 6.7.
6.9 Document No. 51-1146263-00: " Tube Rupture and Solid '
Plant Operation Simulator Runs," B. L. Bowman. Func-tion: Document typical plant response to SGTR and solid plant using OFR simulator. Generically applic-able to all B&W plants.
6.10 Document No. 32-1119841-00: "OFR-1 Simulator Runs -
SGTR Baseline Case - Force Circulation; Steam Genertor Tube Rupture / Natural Circulation Cooldown; Tube Leak in "A" Generator and Open Steam Valve in "B" Gener-ator," T. A. Daniels. Function: Similar to Reference 6.9 above. Applicable to all B&W plants.
\
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B"?NP 20007 3 (9 84) l BASCOCK & WILCOX NUCLEAR POWER DIVISION NUmsta TECHNICAL DOCUMENT 6.11 Document No. 12-1148447-00: " Analysis of Plant Response and Environmental Consequences During SGTR at the Ginna Power Plant." Function: Westinghouse analysis of Ginna SGTR event. General system response and " lessons learned" are applicable to all B&W plants.
6.12 Document No. 32-9857-00: " Safety Assessment of SG Tube Leakage at the Oconee Nuclear Power Station,"
M. V. Bonaca. Function: similar to Reference 6.11 above, except performed for Oconee SGTR event.
Applicable to all B&W plants.
6.13 Document No. 77-1147486-02: " Assessment of SGL Tube Rupture in 1 OTSG w/small Tube Leak in Other," M. V.
Parece. Function: Similar to Reference 6.4 above.
Generically applicable to the B&W 205 FA plants.
6.14 Document No. 77-1147487-01: "An Evaluation of Plant control During MSGTR in Both OTSGs," M. V. Parece, function: Similar to Reference 6.4 above. Gener-ically applicable to the B&W 205 FA plants.
6.15 Document No. 79-1100960-00: "ANO - SGTR Event Tree,"
M. Liebmann. Function: Provides logical evaluation and confirmation of detection and mitigation techni-ques for a SG tube rupture event; serves as a basis for guidance and discussions on this type of event.
applicable to ANO-1 plant.
6.16 Document No. 79-1122865-00: "SMUD - SGTR Event Tree,"
T. A. Daniels. function: Similar to Reference 6.15 above. Applicable to Rancho Seco plant.
6.17 Document No. 79-1126268-00: "TVA - SGTR Event Tree,"
Function: Similar to Reference 6.15 above. Applic-able to Bellefonte 1 and 2 plants.
6.18 Document No. 79-1128179-00: " Consumers - SGTR Event Tree," R. Enzinna. function: Similar to Reference 6.15 above. Applicable to Midland 1 and 2 plants.
6.19 Document No. 79-1120056-QQ: "Oconee - SGTR Event Tree," T. A. Daniels. Function: Similar to Reference 6.15 above. Applicable to Oconee 1, 2, and 3 plants.
6.20 Document No. 79-1121390-00: " Davis-Besse - SGTR l
Event Tree,"~T. A. Daniels. Function: Similar to Reference 6.15 above. Applicable to Davis-Besse plant.
'~ ~
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DATE: PAGE
BWNP 20007 3 (944)
SASCOCK & witCOX NUMSEE NUCLEAR POWER DIVI $10N s 74-1152414-00 ,
) TECHNICAL DOCUMENT 6.21 Document No. 79-1121366-00: " Crystal River - SGTR Event Tree," L. B. Wimmer. Function: Similar to Reference 6.15 above. Applicable to Crystal River-3 plant.
6.22 Document No. 79-1120137-00: " Steam Generator Tube Rupture Event Tree," W. M. Herwig. Function: Similar to Reference 6.15 above. Applicable to TMI-1 plant.
6.23 Document No's. 02-1094595-00. 02-1094596-00,_,
02-1094597-01, 02-1094598-01. 02-1094599-01, 02-1094600-00. 02-1094601-01: "SG Tube Rupture Safety Sequence Diagram," supplied for customers by EDS Nuclear and B&W. Contents: customer-supplied information relating to sequence of equipment failure and plant responses following a SG tube rupture.
Applicable to the ANO-1 plant.
6.24 Document No's. 02-1094626-00 throuch 02-1094631-00:
"SG tube Rupture Safety Sequence Diagram," supplied for customers by EDS Nuclear. Contents: Similar to Reference 6.23 above. Applicable to Oconee 1, 2, and 3 plants.
6.25 Document No's. 02-1094660-00 throuch 02-1094665-00:
"SG Tube Rupture Safety Sequence Diagram," supplied for customers by EDS Nuclear. Contents: Similar to Reference 6.23 above. Applicable to TMI-1 plant.
6.26 Document No's. 02-1094730-00 throuch 02-1094735-00:
"SG tube Rupture Safety Sequence Diagram," supplied for customer by EDS Nuclear. Contents: Similar to Reference 6.23 above. Applicable to Davis-Besse Unit 1.
6.27 Document No's. 02-1094852-00 throuch 02-1094858-00:
"SG tube rupture Safety Sequence Diagram," supplied for customer by EDS Nuclear. Contents: Similar to Reference 6.23 above. Applicable to Crystal River-3 plant.
6.28 Macument_Hp. 32-1119841-00: "TVA ATOG Steam Generator Tuba Rupture IEOTEG Overfill Analysis," T. A.
Daniels. Function: RELAP5 digital computer code analysis of IEOTSG secondary fill rates vs. subcooling margin during SGTR. Applicable to 205-FA plants.
N 85
DATE: PAGE
l BONP 20007 3 (9 84)
SABCOCK & WILCOX NUMllR NUCLEAR POWER DIVISION TECMCR DOCUMENT 6.29 Topical Report BAW-1801: " Single Loop Natural circulation Cooldown," C. W. Tally, 8/82. Function:
Digital computer code analysis of single loop natural circulation cooldown on lowered loop 177 FA plant.
Generically applicable to all B&W lowered loop 177-FA.
6.30 Document No. 32-1153343-00: "SGTR Dose Calcs for TRACC Limits," M. V. Parece. Function: Thyroid and whole-body radiation dose calculations are performed to show that off-site radiation doses are acceptable during SGTR. Applicable to all B&W 177-FA plants.
6.31 Document No. 86-1153344-00: "SGTR Dose Calcs for TRACC Limits,: M. V. Parece. Function: Summary document for Reference 6.30 above.
6.32 Document No's. 02-1094976-00. 026350805-00.
02-1148673-00. 02-10949'9-01. 02 -1094980-00, 02-1094981-01: "SG tube Rupture Safety Sequence Dia-gram," supplied for customer by EDS Nuclear and B&W.
Contents: Similar to Reference 6.23 above. Applic-able to Rancho Seco plant.
7.0 Loss of Coolant Accidents 7.1 Topical Report BAW-10102, Rev. 2, December 1975, "ECCS Evaluation of B&W's 205 FA NSS." Extensive Large Break LOCA and limited Small Break LOCA Analyses of 205 FA B&W Plants. Provided system response data for various LOCA situations.
7.2 Document No. 74-1122501-03, " Operator Guidelines for Small Breaks for Oconee 1, 2, and 3; Three Mile Island 1 and 2; Rancho Seco; and Arkansas 1," M. M. Horne, R. G. McAndrew, 4/6/83. Generic small break LOCA guidelines for the plants listed. These guidelines deal with SBLOCA recognition and mitigation. Guide-l lines were updated and adopted as necessary to reflect
! ATOG philosophy.
l 7.3 Document NE. 79-1141741-00, "TVA Small Break LOCA i Zvent Tree (Dwg. $1132194)," G. E. Anderson, L. R. -
Cartin, 4/7/83, Contract No. 600-5250/620-0015.
l Provided logical evaluation and confirmation of detections and mitigation techniques for a range of small break LOCAs on 205 FA (TVA) plants.
7.4 Document No.' 79-1142152-00, " Supply System Small Break I LOCA Event Tree (Dwg. #1134962)," G. Anderson. Func-l tion: similar to Reference 7.3 above; applicable to WNP-1 and -4 plants.
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CWNP 20007-3 (9 84)
BABCOCK & WILCOX NUCLEAR POWER OlVI$lON 7'~11514- '
- TECHICAL DOCOMElli s
7.5 Document No. 86-1148388-00, " Letter, J.'H. Taylor j (B&W) to S. A. Varga (NRC), dated 7/18/78," J. Seals, i 1/16/84. Summarizes results of design basis analyses i of a series of intermediate-size SBLOCAs on the B&W I i
177 FA plants, Also defines the approximate largest '
! applicable to all B&W plants.
I '
7.6 Document No. 86-1126621-00, "LOCA: RCS Fluid Dis .
' charge to Interconnected Systems," G. E. Anderson.
function: similar to Reference 7.3 above; specifi-
3 specifically to WNP-1 and -4; generically to all B&W plants. l 7.7 Document No. 86-1120656-00, "Small Break Guidelines for High Point Vents in NSS." Function: prescribes l
- guidelines for use of RCS high pint vents to remove
-residual voids after subcooling has been restroad to a .
saturated RCS. Generically applicable to all B&W I plants.
7.8 Document No. 32-3124757-00, "205 FA HPI Line Break' with 36-Foot AFW Level and No Cavitating Venturis,"
! J. C. Seals. Function: establishes emergency j secondary level at 36 feet for the B&W 205 FA plants.
7.9 Document-No. 32-1117414-00, " Steam Generator Level j Adequacy in Reflux Mode," G. E. Anderson. Function:
' 7.10 Evaluation of Transient Behavicr'and Small Reactor Coolant SysteB. Breaks in the 177 Fuel Assembly Plant, j Babcock and Wilcox, S/7/79. Function: Documents the ,
f benchmarking of B&W analytical codes against tne TMI-2 l accident (3/28/79). Also documents results cf I
analyses of small LOCA phenomena, with.and without i l
feedwater systems failures and delays, and.for -l feedwater failures without accompanying small LOCA.
Directly applicable to B&W's 177 Fuel Assembly plants. Used as input in program to define general l
plant transient' responses and mitigative actions.
i s.s l
1 7-23-85 VI-23 DATE: PAGE i
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BONP40007 3 (9 84)
BASCOCK & vnLCoX NUM8tt NUCLEAR POWER DIVISION ucmen nocuum 7.11 Document No. 32-1107008-00, " Sequential Auxiliary g
Feedwater with CRAFT 2," M. G. Gharakhani, J. A.
Randazzo, 3/14/80. Function: supporting technical documentation for Reference 7.9 above. Verified that analytical methods accurately predict an actual plant transient involving sequential starting of feedwater pumps. Serves as a basis for defined operator actions for mitigation of primary and secondary pressure and inventory control transients.
7.12 Document No. 32-1117685-00, " Response to Item 2B of 8/20/79 Ross Letter," S. A. Kellogg, M. G. Gharakhani,.
2/12/80. Function: supplemental supporting technical documentation for Reference 7.9 above. Provides analytical verification that B&W's 177 FA plant FCCS is adequate to protect core.
7.13 Document No. 12-1121718-00, "Respone to GPUN letter, dated August 28, 1980 (TMI-1/E1203)," N. K. Savani, R. C. Jones, 10/18/80. Deals with requirement for early RC pump trip during a SBLOCA to ensare compli-eance with 10CFR50.46. Verifies that Rc pump trip based on loss of RCS subcooling margin would allow 10CFR50.46 compliance. Serves as a basis for rule requireing RC pump trip whenever RCS subcooling is lost.
7.14 Document No. 32-1102025-00, " Simulation of 3/28/79 TMI-2 Transient," M. G. Gharakhan, N. H. Shah, 11/27/79. Function: a basis for the portion of Reference 7.9 above which deals with computer simula-tions of the 3/28/79 TMI-2 accident.
7.15 Document No. 86-1103943-00, "Model Documentation for Masy 7, 1979 Report, entitled, ' Evaluation of Trans-ient Behavior and Small Reacter coolant System Breaks in the 177 Fuel Assembly Plant, May 7, 1979',"
W. Bloomfield, 9/7/79. Function: supporting document for Reference 7.9 above.
7.16 Document No. 32-1153310-0, " Extension of PC Pump Trip
- Manual Action Time," E. P. Menard, 11/15/84. con-l firmed
- 1) need for RC pump trip early in some SBLOC\
l transients, 2) the existence of a reasonable amount of l time (10 minutes) available for operator action to i
trip RC pumps, based on realistic analytical assump-tions. Applicable to all B&W plants.
l O 7-23-83 VI-24 DATE: PAGE
BWNP 20007 3 (9-84) l SASCOCK & WILCOE NUMSER NUCLEAR POWER DIVISION 74-1152414-00, TECHNICAL DOCUMENT 7.17 Document No. 86-1131176-0, " Phase III of.RCP Trip Program," J. C. Seals, W. Bloomfield, 2/8/82.
Function: analytical. confirmation of the need for early RC pump trip, during certain sizes of SBLOCA, on the B&W 205 FA plants.
i 7.18 Document No. 51-1132119-03, " Guidance for Post-LOCA i Tripping of Ractor Coolant Pumps," G. E. Anderson, l L. R. Cartin, 8/2/82. Function: provides analytical ,
confirmation that automatic RC pump trip cirteria for !
B&W's 205 FA plants will ensure 10CRR50.46 conformance for SBLOCA situations. ,
s 7.19 Document No. 86-1149090-0, " Task 3: Part 1 RC Pump Restart with Solid System," M. A. Rinkel. Function:-
- portion of analytical basis for. solid RCS cooldown
I 7.20 Document No. 86-1149318-0, " Task 3: Part 2 ATOG RC i Pump Restart with RV Head Void,: M. A. Rinkel. r Function: portion of analytical basis for cooldown t
. preocedures with RCS voids and for RC pump restart >
guidelines. A basis.for warning statement concerning
- sudden RCS pressure decreases following RC pump l
l restart with voids present. Applicable to all B&W plants.
/
q j 7.21 Document No. 32-1149087-0, " Task 4: Depressuriza-l tion," M.'A. Rinkel. Function: part of analytical ,
basis for guidance on plant.depressurization following reactor trip. applicable to all B&W plants.
7.22 Document No. 77-1150445-00, " Evaluation of Minimum EFW
,1 Requirements Following Small Break LOCA," Function: ;
! to analytically determine the EFW requirements '
l following a SBLOCA with only one HPI pump.
i 8.D Reactor Buildina Control t
> 8.1 Document No. 32-1117950-01, "ANO-1 ATOG RB Pressure l Calc . . ., " G. Shukla, 5/7/80. Function: defines l
containment pressure and temperature response for high '
i energy discharges (various rates) within containment. '
Verifies adequacy of safety system actuation setpoints l in accident mitigation. Applicability: specifically j to ANO-1; generically to all B&W plants l
i 7-23-85 PAGE VI-25
! DATE:
BWNP 20007 3 (9 84)
S ABCOCK & WILCOX NUMtta NUCLEAR POWER OlVISION 74-1;;2414-00 TECHNICAL DOCUMENT 8.2 Document No. 86-1118047-00, "ANO-1 ATOG RB Pressure Calc.," G. Shukla, 3/28/80. Supplements the analysis contained in Reference 8.1 above.
8.3 Document No. 86-1118911-00, "ANO-1 ATOG RB Pressure Calc. Additional Analysis," G. Shukla, 5/6/80.
Supplements the analyses contained in Reference 8.1 above.
8.4 Document No. 86-1125975-00, "RB Spray Setpoint in TVA FSAR." Verified adequacy of RB cooling / spray system to maintain conditions within design limits during LOCA conditions with 4 psic high pressure actuation setpoint. Applicable to Bellefonte 1 and 4 plants.
8.5 Document No. 86-1103119-0, " Containment Response to a Small Break LOA," K. C. Shich, 7.10.79. Function:
analytical basis for defining pressure and temperature transient responses to high energy discharges within containment. Specifically applicable to Midland plants; generically applicable to all B&W plants.
Description of plant and RB response to a spectrum of LOCA. Applicable to B&W's 177 FA Lowered Loop plants.
8.7 Topical Report BAW 10105, "ECCS Evaluation of B&W's 177 FA Raised Loop NSS," W. Bloomfield, 6/75. Similar to Reference 8.6 above; applicable to the Davis Besse-1 plant.
8.8 Document No. 2-1094994-0, "SMUD - Reactor BLDG Spray System auxiliary Diagram," supplied for customers by EDS Nuclear. Contents: technical information
! essential to defining containment transient responses
. (pressure, temperature, radiation, gas level in-creases), equipment operation (e.g., automatic isolation, cooler and spray operation), and mitigaticn actions (e.g., manual equipment operation, purging, sampling). Applicable to Rancho Seco plant.
B.9 Document No. 2-10R1999-0, "SMUD - containment Isola-tion System auxiliary Diagram," supplied for customer l by EDS Nuclear. Contents: similar to Reference 8.8 above. Applicable to Rancho Seco plant.
DATE: 7-23-85 PAGE VI-26
BWNP 20007 3 (9-84)
BASCoCK & WRCOX NUMBER NUCLEAR POWER DIVISION 74-1152414-00.,
ggg ggg 8.10 Document No. 2-1094678-0, "RB Spray System Auxiliary Diagram," supplied to customer by EDS Nuclear.
Contents: similar to Reference 8.8 above. Applicable to TMI-l plant.
]
8.11 Document No. 2-1094682-0, "RB Emergency Cooling System Auxiliary Diagram," supplied for customers by EDS Nuclear. Contents: similar to Reference 8.8 above.
Applicable to TMI-1 plant.
8.12 Document No. 2-1094701-0, " Spray System Auxiliary !
Diagram," supplied for customer by EDS Nuclear.
Contents: similar to Reference 8.8 above.
' Applicable g
to ANO-1 plant.
a j 8.13 Document No. 2-1094705-0, "RB Emergency. cooling System *
' Auxiliary Diagram," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8 above.
Applicable to ANO-1 plant.
8.14 Document No. 2-1094706-0, "RB Isolation System Auxiliary Digram," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8 above.
, N Applicable to ANO-1 plant.
j
8.15 pocument No. 2-1094711-0, "RB Spray System Auxiliary
- Diagram," supplied for customers by EDS Nuclear, i Contents: similar to Reference 8.8 above. Applicable to Oconee 1, 2, and 3 plants.
\ 8.16 Document No. 2-1094715-0, "RB Emergency Cooling System i Auxiliary Diagram," supplied for customers by EDS Nuclear. Contents: similar to Reference 8.8 above. l Applicable to Oconee 1, 2, and 3 plants.
8.17 Document No. 2-1094715-0, "RB Isolation System Auxiliary Diagram," supplied for customer by.EDS Nuclear. Contents: similar to Reference 8.8 above. '
Applicable to Oconee 1, 2, and 3 plants.
8.18 D2cument No. 2-109482.9-2, " Containment Spray System
/
- Auxiliary Diagram," supplied for customer by EDS {
Nuclear. Contents:- similar to Reference 8.8 above.
Applicable to Davis Besse plant.
I 8.19 Document No. 2-1094624-0, " Containment in cob [1cy l
System Auxiliary Diagram," supplied for customer (by EDS Nuclear. Contents: similar to Reference 8.8 l
above. Applicable to Davis-Besse plant.
' i 7-23-85 PAGE -YI~27 -
DATE:
l
BCh? 2WJ ) ') 84>
4Atrocr 4 witc o
- eP.i f Ak eqwp p ra ivigtnN NUM$tt 74-1152414-00 TECHNICAL DOCUMENT 8.20 Document No. 2-1094825-0, " Containment Vessel Isola-tion System Auxiliary Diagram," supplied for Customer by EDS Nuclear. Contents: similar to Reference 8.8 above. Applicable to Davis Besse plant.
8.21 Document No. 2-1094872-0, " Reactor Building Spray System Auxiliary Diagram," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8 above. Applicable to Crystal River-3 plant.
8.22 Document No. 2-1094876-0, "RB Emergency Cooling System auxiliary Diagram," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8 above.
Applicable to Crystal River-3 plant.
8.23 Document No's. 2-1094877-0 and 2-1094879-0, "RB Isolation System Auxiliary Diagram," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8 above. Applicable to Crystal River-3 plant.
8.24 Document No. 70-1150827-0, "TVA containment Event Tree," N. Goulding. Function: provides logical evaluation and confirmation of detection and mitiga-tion techniques for transients and/or equipment failures within the containment. Applicable to the Bellefonte 1 and 2 plants.
8.25 Document No. 2-1094998-0, "SMUD - containment Coolers (System auxiliary Diagram) ," supplied for customer by EDS Nuclear. Contents: similar to Reference 8.8
- above. Applicable to Rancho Seco' plant.
8.26 Document No. 2-1094683-0, "RB Isolation System Auxiliary Diagram," supplied for customer by EDS l Nuclear. Contents: similar to Reference 8.8 above.
l Applicable to TMI-1 plant.
l 9.0 Inadeauate Core Coolina l
9.1 Document No. 32-1105235-00, " Correlation of Clad Temperature to Core Exit Fluid Temperature on Differ-l ent Power Shapes," J. A. Randazzo, R. L. Lowe, l
November 1980. Determination of relationship between exit fluid temperature for 177 FA plants. Relation-ship is utilized in the discussion of recognition and mitigation of the stages of ICC.
9.2 Document No. 32-1126993-00, "Incore Thermocouple Utilization of 205 FA Plants," J. A. Randazzo, W. L. Bloomfield, 9/2/81. Same as Reference 9.1 above except for 205 FA plants.
DATE: 7-23-85 PAGE VI-28
- - - - - . . -- - , ~ . - .- _- . - - _ . - . -.
l ' I 87tNP 20007 3 (9 84) )
~eascoci a wucom O
NUCLEAR POWER DWISION ,
i TECHNH3AL DOCUMENT ,
j 9.3 Document No. 51-1155643-00, ."The Basis for Inadequate Core Cooling Operating Guidelines for 205 FA Plants," l i D. Mulvihill, 5/84. Discussion and explanation of prescribed actions for recognizing and responding to
'. core uncovering conditions. Applicability: speci-fically to Bellefonte 1 and 2; generically to all B&W l plants. l l
9.4 Document No. 79-1143529F-00, "205 FA ICC ATOG Event l Tree," D. Mulvihill, J. R. Paljug, 2/28/84. Provided logical evaluation and verification of detection and I q
mitigation techniques for dealing with and recovering l
- from Anadequate core cooling conditions. A basis for i
the ICC discussion for the 205 FA plants. Generically j applicable to all B&W plants.
j 10.0 Transient Recoanition. Mitication Control 10.1 Document No. 86-1125426-0, " Makeup Line and HPI flow .
4 Rates vs. Pressure for NSS-ll - ATOG Prog.," J. W. !
?
Marchant. Function: part of analytical' basis for j evaluation of methods and effectiveness of operation for equipment vital to maintenance, during transients, i' of control over the five fundamental functions of:
, i l 1. Reactivity Control
- 2. RCS Pressure Control !
- 3. RCS Inventory Control t i
- 4. Secondary Pressure Control I
- 5. Secondary Inventory Control Applicable to Rancho seco plant. !
4 j 10.2 Document No. 32-1106961-01, " Calculations of OTSG Levels for ATOG," M. W. Newlin. Function: similar to Reference 10.1 above. Generically applicable to all ;
i
! 10.3 pagpment No. 32-1150199-00, " Minimum AFW Flowrates/SG
- Level Requirements for SBLOCA," D. Mulvihill.
! Function: simialr to Referenca 10.1 above. Gener-
! ically applicable to all B&W plants. Specifies SG emergency level requirements (e.g., 95% level for 177 l FA Lowered Loop plants) during possible LOCA situa-tions.
! 10.4 Document No. 86-1132874-0, "TMI-l SBLOCA Analysis for
! Thermal Shock Evaluations." Function: part of
( analytical basis for defining Rll pressure-temperature thermal stress limits, and similar to RGfarence 10.1 above. Applicable to TMI-1 plant.
i 4a os v2-is
, p .
BONP 20007 3 (9 84)
BABCOCK & witCgx NUCLEAR POWER DIVISION NUM8tt TECHNICAL DOCUMENT 10.5 Document No. 77-1152846-00, " Stress Analysis of the Reactor Vessel Closure Region for a Natural Circula-tion Cooldown Transient," July 1984. Function:
similar to Reference 10.4 above. Generically applic-able to all B&W 177 FA plants.
10.6 Document No. 32-1128155-00, " Thermal Shock Duke,"
A. Shaw, W. Bloomfield, 10/28/81, Contract No.
582-7218. Provided additional technical bases for existing thermal shock analyses. Confirmed the need for operator action to prevent thermal shock, as prescribed by guidelines for HPI flow control and actions with RC pumps.
10.7 Topical Report BAW-1648, " Thermal Mechanical Report -
Effect of HPI on Vessel Integrity for Small Break LOCA Event with Extended Loss of Feedwater," November 1980. Discussed potential for and possible conditions leading to RV thermal stress. Explained need for, and nature of, operator actions needed to prevent RV brittle fracture. Provides a basis for discussions and guidance onHPI flow control, RC pump operation, and cooldown limits, particularly during LOFW condi-tions.
10.8 Document No. 86-2496-01, "Std 205 Hydraulic Report,"
J. M. Knoll, R. B. Park, 7.20.79. system flow distributions for 1 , 2, 3- and 4-pump operation.
Used throughout ATOG for defining transient systems
, responses and consequences of operator action during l partial pump operator conditions. Primarily for 205 l FA plants; results and trends can be extrapolated to
! 177 FA plants.
10.9 Document No. 32-1132533-00, "Two_ Phase Flow Pump Model Evaluatien," M. Boston, N. K. Savani, 4/9/82.
Evaluation of performance of upgraded two-phase flow l pump model in B&W accident evaluation ccdes. Bench-l marked weil with test data. Used to define expected RCP behavior during two-phase flow situations; this information was used as a basis for pump-related guidance and discussion for all B&W plants.
l 10.10 Document Nq. 32-1128018-00, "Non-Equilibrium Pressur-izer Evaluation," M. Bostoni, J. R. Paljug, 10.26.81.
Confirmed accuracy of pressurizer models used in transient analysis codes. Code predictions about transient pressurizer responses are a basis for related discussion in guidelines.
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RWNP 20007 3 (9 84) i BASCOCK & WILCOX
- NUCLEAR POWER DIVl$10N TECHNICAL OOCUMENT -
1 1 10.11 Document No. 32-1127045-00, " Evaluation of PORV and PSV's Fluid Inlet Conditions," D. A. Doss, W. L.
Bloomfield, 6/30/81. Determined the capacity and effectiveness of pressure relief valves in the pressurizer (the PORV and safety. valves) in relieving sufficient mass and energy to limit RCS pressure t
increases. Pressure limits assumed in previous analyses of SBLOCAs, LOFW and other events, were valid based on relief capacities of valves as confirmed by i these analyses.
! 10.12 Document No. 86-1125060-0, "TMI-1 P-T Diagrams for ATOG," L. J. Rudy. Function overcooling event' plotted on a-simulated P-T display; illustrates plant response i'
as it would be displayed in the control room. A basis for the P-T diagrams in the overcoolign (excess MFW) and other sections. Applicable to the THI-1 plant. ,
i l 10.13 Document No. 86-1125130-0, "TMI-l P-T Plots and l
{ Tabular Data for ATOG," R. B. Brownell. Function: l
] illustrates the ability of the ATOG P-T display to l i visually communicate to the operator present condi- )
q tions and trends. A basis for the P-T diagrams used i , to illustrate various events.
i 10.14 Document No. 12-1134300-01, "Non-condensible Gas
- Source Model," R. H. Smith. FunctAon
- similar to i Reference 10.1 above.. Also specifically pertinent to high point vent operation and recommended operator
- actions during cyclic boiler-condensor conditions.
Generically applicable to all B&W plants.
i 10.15 Document No. 32-1128016-00, "RCS Liquid Volume versus [
Liquid Level." G. E.. Anderson. Function: similar to i Reference 10.1 above; specifically serves as analy-
! tical basis for guidance for RV and hot leg level j mesurement systems during LOCA (saturated RCS) and ICC
- conditions in the Bellefonte 1 and 2 plants.
1 10.16 Document No. 2-1094995-0, "SMUD -Auxiliary Feedwater i System Auxiliary Diagram," supplied for customer by i EDS Nuclear Function: similar to Reference 10.1 i above. Applicable to Rancho Seco plant.- l I
i 10.17 Document No. 2-1094995-0, "SMUD - Low Pressure Injection System Auxiliary Diasgram." Similar to
-Reference 10.16 above.
l 10.18 Document No. 2-1095001-0, "SMUD - Turbine Bypass and :
Atmospheric Dump." Similar to Reference 10.16 above. !
i 7~23~' ##~##
DATE: PAGE l
+
BONP 20007 3 (9 84)
SABCOCK & WitCOK NUMBtk NUCLEAR POWER OlVISION TECHNICAL DOCUMENT 10.19 Document No. 2-1095001-0, "SMUD - Reactor Coolant Pressure Control System Auxiliary Diagram." Similar to Reference 10.16 above.
10.20 Document No. 2-1095000-0, "SMUD - Turbine Control System auxiliary Diagram." Similar to Reference 10.16 above.
10.21 Document No. 2-1094819-0, "HPI System Auxiliary Diagram." Similar to Reference 10.16 above. Appli-cable to the Davis Besse plant.
10.22 Document No. 2-1094680-0, " Chemical Addition System auxiliary Diagram," supplied for customer by EDS Nuclear. Function: a basis for the functions described in Reference 10.1 above plus corrosion control. Applicable to TMI-l plant.
10.23 Document No. 2-1094681-0, " Low Pressure Injection System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to TMI-1.
10.24 Document No. 2-1094686-0, " Turbine Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to TMI-1.
l 10.25 Document No. 2-1094685-0, " Turbine Bypass Atmospheric Dump system Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to TMI-1.
10.26 Document No. 2-1094686-0, "RC Pressure Control System Auxiliary Diagram," supplied for customer by EDS Nucir>ar.
, Function: similar to Reference 10.1 above.
Applicable to THI-1.
10.27 Document No. 2-1094677-0, "High Pressure Injection System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above, plus provides a basis for reccamended operator actions in the event of an HPI line break. Applicable to TMI-1.
10.28 Document No. 2-1094700-0, "HPI System Auxiliary Diagram," supplied for customers by EDS Nuclear.
Function: similar to Reference 10.27 above. Applic-able to ANO-1.
O
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BWNP 20007 3 (9-84) sAtCoCK & WKCoA NUMBER NUCLEAR POWER DIVISION
+
P TEcucat accoutui
\
10.29 Document No. 2-1094703-0, " Chemical Addition System Auxiliary Diagram," supplied for customer by EDS I Nuclear. Function: similar to Reference 10.22 i above. Applicable to ANO-1. l 1
10.30 Document No. 2-1094704-0, " Low Pressure Injection i
System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to ANO-1.
10.31 Document No. 2-1094707-0, " Electro Hydraulic Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to ANO-1.
10.31 Document No. 2-1094707-0, ' Electro Hydraulic Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to ANO-1.
- 10.32 Document No. 2-1094708-0, " Turbine Bypass System 1
Auxiliary Diagram," supplied for customer by EDS C Nuclear. Function: similar to Reference 10.1 above.
Applicable to ANO-1.
I
! 10.32 Document No. 2-1094708-0, " Turbine Bypass System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to ANO-1.
10.33 Document No. 2-1094709-0, "RC Pressure Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to ANO-1.
10.34 Document No. 2-1094710-0, "HPI System Auxiliary Diagram," supplied for customer by EDS Nuclear.
Function: similar to Reference 10.1 abova. Applic-able to Oconee 1, 2, and 3.
10.35 Document No. 2-1094713-0, " Chemical Add system Auxiliary Diagram," supplied for customer by EDS Nuclear. ~ Function: similar to Reference 10.22 above. Applicable to Oconee 1, 2, and 3.
10.36 Document No.- 2-1094714-0, " Low Pressure Injection
- System Auxiliary Diagram," supplied for customer by ,
EDS Nuclear. Function: similar to Reference 10.1 I above. Applicable to Oconee 1, 2, and 3.
i
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DATE: PAGE
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CWNP 20007 3 (9 84) )
i i SASCOCK & WILCOX l NUMSER NUCLEAR POWER DIVISION I TECHNICAL DOCUMENT l
10.37 Document No. 2-1094717, " Turbine Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
i Applicable to Oconee 1, 2, and 3.
10.38 Document No. 2-1094718-0, " Turbine Bypass Atmospheric Dump System Auxiliary Diagram," supplied for customers by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to Oconee 1, 2, and 3.
10.39 Document No. 2-1094719-0, " Pressure Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to Oconee 1, 2, and 3.
10.40 Document No. 2-1094822-0, " Chemical Addition System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.22 above. Applicable to Davis Besse plant.
10.41 Document No. 2-1094823-0, " Low Pressure Injection System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to Davis Besse plant.
10.42 Document No. 2-1094826-0, " Turbine Control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to Davis Besse plant.
10.43 Document No. 2-1094827-0, " Turbine Bypass System Auxiliary Diagram," supplied fcr customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to Davis Besse plant.
! 10.44 Document No. 2-1094828-0, "RC Fres. Con. Sys. PZR, HTR
& Spray System Auxiliary Diagram," supplied for customer by EDS Nuclear. Funct. ion: similar to Reference 10.1 above. Applicable to Davis Besse plant.
l 10-15 pocument No. 2-1094830-0, " Turbine Control System
! Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to Crystal River-3.
l l 10.46 Document No. 2-1094839-0, " Turbine Bypass and Atmos-pheric Dump System Auxiliary Diagram," supplied for customer by EDS Nuclear. function: similar to Reference 10.1 above. Applicable to Crystal River-3.
i-za-ca vA-34
, p
87NP-20007 3 (9-84)
B ARCOCK & w LCOE NU"#fE NUCLEAR POWER DIVISION 74-1152414-00
[~N t
D)' TECHICAL DOCullENT 10.47 Document No. 2-1094871-0, "High Pressure Injection System Auxiliary Diagrari,W supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above. Applicable to Crystal River-3.
10.48 Document No. 2-1094871-Q, " Chemical Add. System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.22 above. Applicable to Crystal River-3.
10.49 Rocument No. 2-1094875-0, " Low Pressure Injection System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1.
Applicable to Crystal River-3.
10.50 Document No. 2-1094878-0, "RC Pressure control System Auxiliary Diagram," supplied for customer by EDS Nuclear. Function: similar to Reference 10.1 above.
Applicable to Crystal River-3.
10.51 Document No. 51-1140889-0, "PS&C Suggested TRAP Base ,
Deck for TVA ATOG," P. R. Boylan. Function: documen-tation of base computer code deck to be used in
's analyses of ATOG transients. Applicable to Bellefonte i 1 and 2 plants.
10.52 Document No. 67-1002380-0, " Limits and Precautions." - i Defines limits of normal and transient operation, e.g., cooldown ratio, SG tube to shell differential .
temperatures, RC pump operating requirements. I Applicable to Crystal River-3.
i 10.53 Document No. 67-1003781-02, " Limits and Precautions." '
i Similar to Reference 10.52; applicble to Bellefonte 1 and 2 plants.
10.54 pocument No. 67-1005821-0, " Limits and Precautions."
Similar to Reference 10.52; applicable to Midland 1 [
and 2.
10.55 Document No. 67-1118131-0, " Limits and Precautions."
Similar to Reference 10.52; applicable to WNP 1 and 4. :
10.56 Document No. DP-1101-01, " Plant Operating Limits and :
Precautions for Duke Power company Oconee Unit I." l Similar to Reference 10.52; applicable to Oconee I.
10.57 Document No. DP-1101-01, " Plant Operating Limits and em s Precautions for Duke Power company Oconee Unit II."
Similar to Reference 10.52; applicable to Oconee II.
~
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DATE: PAGE
BWNP 20007 3 (9-84)
B AB COCK & witCOX NUM0tt NUCLEAR POWER DIVISION n-11s2414-00 TECHNICAL DOCUMENT 10.58 Document No. DP-1101-01, " Limits and Precautions for Duke Power Company Oconee Unit III," similar to Reference 10.52; applicable to Oconee III.
10.59 Document No. DP-1101-01, " Metropolitan Edison company Three Mile Island Unit One Plant Limits and Precau-tions," similar to Reference 10.52; applicable to TMI-1.
10.60 Document No. DP-1101-02, " Plant Set Points for Arkansas Power and Light Nuclear One," similar to Reference 10.52; applicable to ANO-1.
10.61 Document No. DP-1101-01, " Sacramento Municipal Utility District Rancho Seco Unit 1 Plant Limits and Precau-tions," similar to Reference 10.52; applicable to Rancho Seco.
10.62 Document No. DP-1101-01, " Limits and Precautions for Toledo Edison Company Davis Besse 1," similar to Reference 10.52; applicable to Davis Besse 1.
O l
e 7-23-85 ~#
DATE: PAGE