ML20086S725
| ML20086S725 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/06/1982 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20086S719 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 74-1125531, 74-1125531-00, NUDOCS 8403050114 | |
| Download: ML20086S725 (245) | |
Text
{{#Wiki_filter:_ _ - - _ - _ _ _ _ _ _ _ - _ _ _ BWNP-20004 (6-76) BABCOCK & WILCOX uveaa.- ,o m ,a m u n n o,v,se n TECilNICAL DOCUMENT kk S pge'4 DAVIS BESSE NUCLEAR POWER STATION UNIT UNIT I ABNORMAL TRANSIENT OPERATING GUIDELINES PART II VOLUME 1 74-1125531-00 Doc. ID - Serial No., Revision No. for < TOLEDO EDISON COMPANY by BABCOCK & WILCOX O THIS DOCUMENT WAS PREPARED FOR TOLEDO EDISON COMPANY UNDER MASTER SERVICE CONTRACT NO. 582-7151 (B&W No. 582-7108). ANY USE OF THE INFORMATION CONTAINED HEREIN OTHER THAN UNDER THE EXPRESS CONDITIONS OF SAID CONTRACT IS EXPRESSLY PROHIBITED WITHOUT THE WRITTEN PERMISSION OF THE BABCOCK & WILCOX COMPANY. 8403050114 840301 DRADOCK05000g PAGE 1
BWNP-20007 (6-76) BABCOCK & WILCOX .... NuoeAs powea oeNBATON DIVISION TECHNICAL DOCUMENT 74-1125531-00 ATOG GUIDELINES PART II , TAB NAME VOLUME 1. FUNDAMENTALS OF REACTOR CONTROL FOR ABNORMAL TRANSIENTS Int roduc t ion Chapter A - Basic Heat Transfer Heat Transfer Addendum A - Subcooled, Saturated, A-Subcooling Superheated Water Addendum B - Natural Circulation B-Natural Circulation Chapter B - Use of P-T Diagram P-T Diagram Chapter C - Abnormal Transient Diagnosis Diagnosis and
& Mitigation Mitigation Chapter D - Backup Cooling Methods Backup Cooling Chapter E - Best Methods for Equipment Equipment Operation Operation Chapter F - Post Transient Stability Stability Da te rmination Chapter G - Fundamentals of Reactor RB Control Building Control Chapter H - Use of the Guidelines Use of ATOG f.
I DATE: 7-6-82
BWNP-20007 (6-76) } BABCOCK & WILCOX suusen l NUCLEAR POWER GENERATION Olvi$lON 7'- 25531-00 ECHNICAL DOCUMENT VOLUME 1 FUNDAMENTALS OF REACTOR CONTROL FOR ABNORMAL TRANSIENTS List of Figures Figure 1 Fundamental Methods of Heat Transfer Control Figure 2 Illustration of Parameters Contribucing to Natural Circulation Driving Head
,- Figure 3 Transition to Natural Circulation Using AFW
[ ) Figure 4 DELETED
\'# / Figure 5 DELETED Figure 6 DELETED Figure 7 DELETED Figure 8 Power Operation P-T Diagram Figure 9 Post Trip P-T Diagram Figure 10 Typical Response of Major Plant Parameters Following a Reactor Trip Figure 11 Overheating Transient (Pre-trip)
Figure 12 Overcooling Transient (Pre-trip) Figure 13 Overpressure Transient (Pre-trip) Figure 14 Depressurization Transient (Pre-trip) Figure 15 Loss of Main Feedwater [, \, Figure 16 Small Steam Leak in One SG (TBV Fails Open) (" ,
) Figure 17 Excessive Feedwater Figure 18 Small LOCA in Pressurizer Steam Space 3
Figure 19 Small LOCA in RCS Water Space Figure 20a General Plant Accident Mitigation Figure 20b Excessive Primary to Secondary Heat Transfer Figure 20c Loss of Primary to Secondary Heat Transfer Figure 20d Inadequate Subcooling Margin Figure 21 Transient Mitigation Approach Figure 22 Overcooling Diagnosis Chart Figure 23 Overheating Diagnosis Chart Figure 24a Backup Cooling by MU/HPI for Loss of All Feedwater (No Operator Action) CN Figure 24b Backup Cooling by MU/HPI for Loss of All Feedwater ( ) (With Operator Action) (./ Figure 25 RC Pre ssure/ Temperature Limits Figure 26 Illustration of Loss of Natural Circulation Due to Buildup of Steam in the Reactor Coolant System Fiugre 27a Illustration of Boiler-Condenser Cooling Figure 27b Boiler-Condenaer Cooling Figure 28 Loss of Boiler-Condenser Cooling - System Refill by MU/HPI Figure 29 Core Exit Fluid Temperature for Inadequate Core Cooling Fiugre 30a hPI Throttling Limit (For High HPI Line During HPI Operatio, With Only One HPI Pump) CN Figure 30b HPI Control Logic Figure 31a Cooldown on One Steam Generator (Other Generator Pressurized (p/ ) with SGTR) DATE: PAGE 3 7-6-82
BWNP-20007 (6-76) SABCOCK & WILCOX NUMB ER NUCLEAR POWta OtNEATION DIVISION 74-1125531-00 TECHICAL DOCUMEllT Figure 31b Cooldown on One Steam Generator.(Steam Pressure Not Controlled) Figure 32 Reactor Building Temperature vs. Pressure Trend for Large LOCA Figure 33 Hydrogen Generatior from Zr-H 2O Reaction Figure 34 Reactor Building Hydrogen Concentration Following a I#rge LOCA Figure 35 Estimate of Total Hydrogen Volume in RCS of IMI-2 Figure 36 Reactor Building Radiation Monitor Response Following 2-25-80 Loss of Non-Nuclear Instrumene.ation with HPI Cooling Event TMI-2 Dome Monitor Readings Figure 37 Figure 38 Reactor Building Temperature vs. Pressure Trend for Hydrogen Explosion Figure 39 Reactor Building Temperature vs. Pressure Trends for Steam Line Break Figure 40 General Approach for Control of the Reactor Building DATE: 7-6-82 PAGE 3-1
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BWNP-20007 (6-76) BASCOCK & WILCOX ., n NUCLEAR POwet OtNWATION OfVISION f 7'-1125531-oo TECHICAL DOCUMEllT VOLUME 1 FUNDAMENTALS OF REACTOR CONTROL FOR ABNORMAL TRANSIENTS List of Tables Table 1 How Failures Af fecting Heat Transfer Can Af fect Reactor Ope ration Table 2 Standard Post Trip Actions a Table 3 Actions to Correct Fast Transients Table 4a How to Dif ferentiate a LOCA from Other Transients Table 4b Symptoms for LOCA's That Can be Located or Isolated Table 5 Rules for RC Pump Trips Table 6 RC Pump Restart Guidelines Table 7 Summary of Reactor Building Isolation Control Actions Table G Summary of Reactor Building Internal Environnent Centrol Actions Table 9 Penetrations of Non-Safety Fluid Systems Used in Power Ope ration DATE: 7-6-82
L BWNP-20007 (6-76) SABCOCK &-Wl'. l ... Nucteam rowen oeNBA T TEC HICAL 00% 74-1125531-00 ATOG CUIDELINES PART II
% -INTRODUCTION The 3-28-79 accident at the Three Mile Island Nuclear Power Generating Plant \v has caused the Nuc lear Industry's perspective of emergency operation to ch ange . That accident was difficult for the plant operators to handle be-cause several things were happening C .c e . Loss of Main reedwater, Loss of Emergency Feedwater, and Small Break LOCA occurred at the same time. An in-correct interpretation of pressurizer level misled the operator to think the core was covered when it was not. The operator acted on that misleading information and core cooling was stopped when he shut down Emergency High Pressure Injection and the Reactor Coolant Pumps. The combination of mul-tiple failures and incorrect interpretation of information are the main fac-tors which have caused a new perspective of emergency ope ra t ion to be developed.
In the past emergency procedures and operator training concentrated on single event accidents. But accidents do not usually happen with only single fail-
\' ures; several things of ten go wrong at the same time. These guidelines have been developed so that an operator can understand what has gone wrong, he can circumvent failures and keep the core cool with the available equipment.
When failures of equipment occur, they frequently cause a change in the heat t t rans fe r from che core to the steam ge ne ra to rs . When the reactor is b DATE: PAGE 7-6-82 4
BWNP-20007 (6-76) BABCOCK & WILCOX NucLEAs POWis GENERATCN OlVISCN 74- 25531-00 TECHNICAL DOCUMENT operating normally, all tha heat produced by the core is being removed by the steam gene rators ; primary and secondary system pressures, tempe ra tur e s , and levels are stable. Heat transfer is balanced. Any transient will c ause an upset in the heat transfer from the core to the steam generators and the main objective of emergency procedures is to restore and maintain adequate core cooling. Hest transfer will be affected in different ways depending on what aquipment has operated incorrectly. When the heat transfer changes, the ef fects will show up in primary and secondary system pressures, temperatures, and levels. Pressures and temperatures are parameters from which three basic symptoms of improper heat t rans fe r can be de rived and used to discover what has gone wrong. These guidelines ill use those heat transfer symptoms as the source of information for the operator action. Recognition of just three basic heat transfer symptoms will give the knowledge needed so the operator can restore and maintain adequate core cooling. Correlation Between Part I and Part II Use of symptom-oriented procedures involves a different approach to plant control and requires a shift in emphasis in operator training. The symptom approach of Part I will work, regardless of the event. By training on the ATOG approach, the operator will have a thorough understanding of heat transfer, plant control, and the various options available for controlled core cooling when systems and equipment fail. Part II of ATOG was written to provide the basis for this understanding with the intent that it be used as part of the operator training program. O DATE: PAGE 7-6-82 5
BWNP-20007 (6-76) , BABCOCK & WILCOX Nuaa:En NUCLEAR POWER GENERATON DIVISION 7'- 25331-00 TECHNICAL DOCUMENT Part II, Volume 1 Volume 1, " Fundamentals of Reactor Control for Abnormal Transients," provides the basic background necessary for understanding heat transfer and builds on this info rma t ion to enable the operator to recognize abnormal conditions when they develop and take the appropriate actions to correct them. Volume 1 covers primarily in forma t ion regarding the heat trans fe r process, including s ubcooling and natural circulation. Volume 1 also shows how to use the P-T diagram and this knowledge of heat trans fe r to diagnose abnormal transients and mitigate them. The preferred method of core coolir.g is with controlled primary to seconda ry heat transfer and many abnormal transients involve restoring a balance to 9 this heat removal path. However, this is not always possible; therefore, Volume I also discusses core cooling methods when the steam generators are not ava ilable. Volume 1 also covers operational methods for key sys t ems, ( fe ed wa t e r , HPI, etc.) and equipment for various conditions, provides guidance on verification of pla nt stability, discusses the fundamentals of reactor building control, G and provides instructions on the use of the guidelines. Volume I contains a considerable amount of information and should be studied periodically for optimum comprehension and retention. Three major points should be kept in mind when reading Volume 1: O DATE: PAGE 7-6-82 6
BWNP-20007 (6-76) BABCOCr & WILCOX Nuussa NUCLEAR POWtN GENT 5tATION DIVISION 74-ii2n 31-oo TECHNICAL DOCUMENT
- 1. Understanding heat transfer is essential.
- 2. Relationship of symptoms and control functions.
- 3. Dif ferentiating between actions which are rules and those which are guidelines.
Heat Transfer One aspect of plant control and the use of ATOG cannot be ove rs t res sed : the import ance of understanding heat transfer and primary pressure-temperature relationships. A thorough grasp of the heat trans fer process and P-T relationships will enable the operator to: recognize symptoms of abnormal heat trans fer , evaluate plant response to corrective actions implement backup cooling methods when needed Although virtually any event or combination of events could conccivably oc-cur, they all present the common threat of disrupting core cooling. Thus, the major thrust of ATOG is to maintain sor.e form of controlled core cooling, whether it be by the steam generators or ECC systems. Simply put, understand-ing heat transfer allows recognition of symptoms of abnormal transients. Recognition of symptoms allows implementation of the appropriate sections of Part I. Implementa t ion of the appropriate sections of Part I and verifica-tion of plant res po ns e allows stabilization and restoration of controlled core cooling. O DATE: pAGE 7
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BWNP-20007 (6-76) BABCOCK & WILCOX Num En NUCLEAR POWER GENE 2ATION DIVl560N 74- 2ssai-00 TECHNICAL DOCUMENT Symptoms and Control Functions Section III of Part I of the guidelines is divided into four main sect ions to address the basic heat transfer symptoms (lack of subcooling margin, over-heating, and overcooling) and the special case of SG tube rupture. Part II discusses the importance of " control functions". These control functions are:
- 1. RC inventory
- 2. RC pressure
- 3. SG inventory
- 4. SG pressure A fifth control func t io n, reactivity, is also important in heat trans fer considerations. However, reactivity is quickly controlled by automatic G reactor trip, manual reactor trip, and/or emergency boration.
If control of one of these four functions is loet, it will impact primary to secondary heat trans fer and become evident au an abnormal heat t rans fe r symp-tom. For example, a loss of SG inventory control low (loss of feedwater) will result in a loss of primary to secondary heat trans fe r (overheating). Conversely, a loss of SG inventory control high (too much feedwater) will 9 result in excessive primary to secondary heat trans fer (overcooling). When a symptom appears, one or more of these functions are not being con-trolled properly. Regaining control of these four functions will res to re controlled core cooling. O DATE: PAGE 7-6-82 8
EWNP-20007 (6-76) BABCOCK & WILCOX Numsta NUCt.6AR POWER GENERATION Divi 5 TON TECHNICAL DOCUMENT 74-1125531-o0 Rules and Guidelines Volume i provides guidance on the operation of systems and equipment for many conditions and various events. When an action must always be taken for the conditions s peci fied it is called a rule and is enclosed in a box for emphasis. For example, the RC pumps must always be tripped whenever the subcooled margin is Icst; therefore, it becomes the RC pump trip rule. Whenever specified act ions are recommended, but not always mandatory, they are considered guidelines. For example, Table 6 in Volume 1 provides guidelines for RC pump opcration for different plant conditions. ATOG was designed for maximum flexibility in order to address the spectrum of conceivable transients. Therefore, rules have been kept to the minimum necessary. The user should remember, however, that the guidelines are also important and should be follewed whenever they are applicable and feasible. Pert II, Volume 2 Volume 2, " Discussion of Selected Transients," prevides detailed coverage of six specific initiating events. Although the ATOG concept is a breck from the traditional event oriented approach, Volume 2 was structured in the' event oriented approach to meet the following objectives:
- 1. Validate the ATOG Concept Most operators involved in the initial implementation of the ATOG concept will be experienced with use of event-oriented procedures O
"^"' "~" '^" '
i l BWNP-20007 (6-76) BABCOCK & WILCOX Numeen NUCitAR POWER GENERADON DIVI 510N 74-ii25531-00 TECHNICAL DOCUMENT and may understandably be resistant to a dif ferent approach. Volume 1 discusses ATOG in a genaral ove rview manner and, standing alone, may not fully promote user confidence in the concept. Therefore, Volume 2 is provided to give examples of representative
+ events and how the use of ATOG will lead to successful mitigation.
Although the event is given, the diagnosis and mitigation is written with the assumption that the operator (in the example) is unaware of the specific cause. In addition, the discussion on each transient demonstrates suc-cessful mitigation of events compounded by other failures using the same basic ATOG procedure. This highlights the relative simplicity of using a single, comprehensive procedure as opposed to several discrete procedures. The transients depicted in Volume 2 are derived from more realistic analyses ' than previously used for design bases accident analyses. Thus these transient discussions should give the operator a better feel for how the plant would actually respond should similar conditions occur. Where available, actual plant data from representative transients is used.
- 2. Amplify Volume 1 The structure of Volume 2 provides a ready vehic le fo r conveying more detciled information about transient types (e.g., overcooling)
DATE: 7-6-82 PAGE 10
BWNP-20007 (6-76) ,
- f. ABCOCK & WILCOX Numset NUCLEAR POWER GENERATION DivlSION 74-1125531-o0 TECHNICAL. DOCUMENT and peculiarities and complexities of specific events. This is especially true for the appendices covering SG tube rupture and small break LOCA. These two events are unique in that they cannot be quickly terminated and stabilized. They impact many facets of plant operation and their mitigation is highly dependent on specific conditions at the time of occurrence (including the size of the ]
leak). Cons eq uen t ly, considerably more event specific in fo rmat ion is provided in these two appendices. The entire purpose of these guidelines is to give an overview on reactor transients, their diagnosis and control, so transients as severe as the Three Mile Island accident will be prevented. Because transients will not follow a planned course, anything can h ap pe n. Consequently, a symptom-oriented approach is necessary to ensure transient control. These guidelines should provide enough background and understanding so that no matter what happens, the operator will have sufficient understanding to correctly respond to the transient using the principles of heat transfer control. O DATE: 7-6-82 PAGE 11
BWNP-20007 (6-76) BASCOCK & WILCOX NUMBER MUCLEAR POwet GEN 8BATION OlV15 TON 74- t i 25s31-00 . TEClllllCAL DOCUMEllT CHAPTER A BASIC HEAT TRANSFER Introduction and Summary This chapter of the guidelines gives the basic principles of heat trans fe r that are important for removing heat from the core so that it can be properly The chapter is divided into three parts: 1) " Basic Heat Transfer", ( . cooled.
- 2) Addenaum A "Subcooled, Saturated, Superheated Water", and 3) Addendem B
- " Natural Circulation". Addendum A and Addendum B give information on two general subjects. The part on " Basic Heat Transfer" covers two related top ics : 1) the general process for heat removal through the steam generators, and'2) the ways the operator can control that heat trans fe r .
The preferred way to protect the core and prevent fuel failure is to control the rate of heat removal by transferring core heat to the steam genera tor s . Other ways to protect the core do exist; they are covered in a later chapter called " Backup Cooling Methods". To control core heat removal with the steam generator the operator should balance the heat generated by the core with the heat removal through the steam generators. This section will show the fundamentals of heat trans fer control and how the operator applies these fundamentals to get balanced heat removal. DATE: PAG 2~ 7-6-82 12
l i BWNP-20007 (6-76) BABCOCK & WILCOX uu ,,,, NUc1 EAR POwta GENER AftON DIVislON 74-1125531-00 T_ECHNICAL DOCUMENT Heat Transfer Equations The path for heat flow from the core to the cteam generator is: Core Heat ) reactor coolant Reactor coolant heat J Steam generator water and steam The steam generator then releases the heat either to the atmosphere or to the condenser. The concepts of heat sinks and heat sources are useful. For the first heat transfer path the core is the heat source for the reactor coolant and the reactor coolant is the heat sink. When the plant is tripped the reactor coolant pump heat becomes a significant heat source. For the second heat trans fer path the reactor coolant is the heat source and the steam generator water and steam is the heat sink. The atmosphere and the condenser are heat sinks for the steam from the steam generator. In some unusual cases the reactor coolant can be colder than the steam generator fluid; then the steam generator is a heat source which passes heat to the reactor coolant sink. Two " kinds" of heat can be trans ferred to the steam generators:
- 1. Generated Heat-which includes RC pump work and nuclear heat which is the heat made within the core by the fission process; it includes decay heat O
7-6-82 PAGE 13 DATE: l
BWNP-20007 (6-76) BABCOCK & WILCOX - sumen wucteAn rowse oewenAnow oivision 74-1125531-o0 s TECHNICAL DOCUMENT
- 2. Stored Heat - which is the heat of the metal parts of the reactor coolant system and of the reactor coolant When the reactor is operating at steady state and heat removal is balanced, the steam generators will remove the nuclear heat and RC pump heat as it is generated and reactor coolant t emperatures will not ch ange . In other words, the stored heat will stay the same.
If the steam generators remove more heat than the core and RC pumps are creating, then they will remove both generated heat and stored heat; reactor coolant temperatures will drop. Normal cooldown is a condition
^ when both generated heat and stored heat are being removed within a specified rste; this is a contro41ed condition. If the condition is ab-normal or not controlled then it would be called overcooling and corrective actions would have to be taken to bring it under control.
On the other hand , if the steam generators remove less heat th ar, the core is creating, then the nuclear heat will increase the amount of reactor coolant stored heat; reactor coolant temperatures will increase. Heatup from 0% to 25% power illustrates a controlled example where the stored heat of the reactor coolant is increased by heat addition from the core nuclear heat and the reactor coolant pumps. If a condition exists where the reactor coolant temperatures increase abnormally it is called overheating; correct ive actions would have to be t aken to bring overheating under control. 14 DATE: 7-6-82 pAGE
BWNP-20007 (6-76) , BABCOCK & WILCOX Nuusta HUCLEAR POWit GtHEaATCH DIV!5CN 7 - 2s531-00 TECHNICAL DOCUMENT Equations can be used to describe the heat transfer path from the core to the stean generators. When the heat transfer is balanced: Equation 1)hcore= reactor coolant for the heat transfer path frora the core to the reactor coolant and Equation 2) hreactor coolant = h steam generator fluid for the heat transfer path from the reactor coolant to the steam and water in the secondary side of the steam generators h is heat rate - units are BTU /hr. When heat transfer is balanced all the way from the core to the steam generator Equation 1 equals Equation 2. But when heat trans fer becomes unbalanced they will not be equal. Inter rupt ions of the heat transfer path can happen when the reactor coolant is not a good heat sink for the core (Qcore f kreactor coolant); or when the steam generator fluid is not a good heat sink for the reactor coolant (hreactor coolant 8hsteam generator fluid)- The unbalanced condition of concern for core heat trans fe r to the reactor coolant is when there is not enough heat transfer from the core to the reactor coolant. This can only happen when the reactor is no DATE: PAGE 7-6-82 15
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWSE GENERATION OlvtS60N 74- 123331-o0 TEClllllCAL DOCUMENT longer subcooled; for example, when the core is partly covered by water and partly by steam or covered completely by steam; then Qcore f 9 reactor coolant. When this happens not enough nuclear heat can be transferred from the core to the reactor coolant and the core will heat up. The stored heat of the fuel clad will increase which will result in increased fuel pin temperatures. When the steam generator heat flow path becomes unbalanced then the steam generator fluid will remove too ==h er too little heat from the reactor coolant and cause an overcooling or overheating condition. When this happe ns during a transient, Qreactor coolant will increase or decrease depending on the heat removal by the secondary side. The reactor coolant temperatures will change in order that temperature (thermal) equilibrium can be rc -es tablished between the primary and secondary side fluids. To show the ef fect s Equations 1 and 2 can be written to add temperature terms: Equation 1 (Qcore " 9rc) can be written as: Equation la Qcore " M rcCprc(Th-Tc ) where: M rc = reactor coolant system mass flow rate (1bm/hr) Cpre = specific heat capacity of the reactor coolant (BTU /lbm-F) Th = core outlet temperature (F) Tc = core inlet temperature (F) Equation 2 (dec
- Qsg) can be expanded as follows:
s s DATE: PAGE 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Huanta HUCLEAR FOwEt GENEitATION (AYl510N 7- 23531-00 TECHNICAL DOCUMENT Equation 2a: Q 88 = UA A T where: U = overall heat transfer coef ficient
' Ecu/hr-f t2 _p)
A = total area of heat trans fer surf ace (f t2) AT = temperature dif ferential across the heat transfer boundary (F) Overall heat transfer coef ficient is dependent on many factors. including the iluid conditions (primarily density and flowrate) on both sides of the boundary and the properties of the boundary (primarily the thic kne ss and thermal conductivity of the barrier and oxide layers). For th is discussion m can assume that the properties of the boundary (steam generator tube walls) remain constant and therefore can be ignored. The secondary side of the steam generator has three different regions along the tub e bundle during power operation: nucleate boiling, film boiling, and su pe rh ea t . Each region has a di f ferent coefficient (U), surface area (A), and temperature differential across the tube wall (A T) . The nuc leate boiling region has the highest U of the three and ac c oun t s for approximately 70 to 85% of the total heat trans fe r into the steam generator over the power range. The heat transfer coefficient decreases by a factor of 3 to 10 in the film boiling region and again by another factor of 3 to 10 in the superheat region. The heat trans fe r surface areas and AT's involved for each of the three regions vary over the power range with the two boiling regions account-ing for an increasingly higher percentage of the total heat trans fer DATE: PAGE 7-6-S2 17
BWNP-20007 (6-76) BABCOCK & WILCOX Numen NUCLEAR POWER GENERAtlON OlVISION 74-1123331-0o TECNNICAL DOCuldENT with increasing power levels. Thus, to determine the ef fects of trans-ients on secondary heat removal during power operation, the effects in each of the three regions along the tube bundle must be studied. However, fo r the purposes of these guidelines, we are primarily concerned with control of heat removal by the steam generato rs 9fter a reactor trip. After trip the steam ge nerators tre at saturation conditions with two basic regions, saturated water and saturated steam. Almost all of the heat transfer occurs in the water region and most of the heat trans fe r in the water region occurs in the nucleate boiling portion below the s tea.n/wat er interface. Saturated water is absorbing the latent heat of vaporization and the nucleate boiling provides a much s higher heat transfer coefficient (U). Below this level the water is saturated with a considerably Ic.ter heat transfer coef ficient, although this heat trans fer coef ficient is still much higher than exists in the steam space. Very little- heat tr ans fe r occurs in the steam space (primary side temperature can be considered equal to Thot throughout the steam space). Even though the area is large, the heat transfer coefficient is small due to low steam flow rates and low density with respect to the water region. During forced circulation the AT across the tube walls in the steam space -is also very small as That is close to Tsat-of the steam. The AT is larger between Thot and T sat during natural circulation but the heat transfer coefficient is even smaller due to the lower primary k flowrates.
' 18 DATE: 7-6-82 PAGE
l BWNP-20007 (6-76) BABCOCK & WILCOX NumsEn NUCLEAR POWER GENERATION DIYl5 ION 74-ii 25531-Oo TECHNICAL DOCUMENT The primary factors affecting heat transfer in the water region are sur-face area and the AT between the primary and secondary sides. Su-face area is increased by increasing feedwater flow to raise level. The pri-mary increase in area t ake s place in the saturated water region. Even though most of the heat transfer occurs in the nucleate boiling region, overall heat t rans fe r is increased because the area of the steam space (with a very small heat transfer coefficient) is decreased and replaced by area in the saturated water region (with a relat ively much larger heat transfer coefficient). The major method of af fecting primary to secondary A T is on the second-ary side by varying steam pressure. When steam pressure is dev s r ad (e.g., by opening turbine bypass valves) saturation temperature also decreases which increases the AT across the tube wall. The higher A T causes heat trans fe r (Qsg) to increase thus cooling the prima ry side. Of these two factors (surface area and AT), AT the major factor. Af-fects due to surface area are significant only while the surface area is changing. Once a constant water level is maintained, changes in heat transfer are primarily due to changes in A T. Heat transfer can be increased significantly by injecting auxiliary feed-water which enters the SG through the uppe r nozzles. The increase in heat transfer is due to chree factors. First, and most significant, the spray of feedwater into the steam space reduces steam pressure similar O 7-6-82 PAGE 19 DATE:
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BABCOCK & WILCOX " NUCLEAR POWER GENERATION OlvisiON 74-1125531-00 , TECHNicht DOCUMENT to the action of pressurizer spray. This reduces the saturation tempera-ture wh ich increases heat t rans fer as described previously. Second, , where water contacts the tube surfaces in the steam space the heat L I
' transfer coefficient is increased, essentially replacing steam area with 1 i ' water area as in the case of raising steam generator level. Third, A
auxiliary feedwater will have a greater cooling effect than main feedwater due to its colder temperature, t [ Assuming a minimum adequate level is maintaincd in the steam generators, r I variations in steam pressure will have a greater ef fect on heat trans fer than variations in level. The best method to decrease heat transfer is j' to close the turbine bypass valves and allow the . steam generator N pressure to increase. Allowing steam generator level to decrease will not have an appreciable ef fect on heat ~ transfer until the level becomes i inadequate (too ~ 1ow for maintaining natural circulation or virtually dry i with forced circulation). I i In summary, the operator can control- primary to secondary heat transfer af ter reactor trip by controlling two major parameters on the secondary
' side (assuming the capability of the reactor coolant to transport core heat to the steam generators remains intact). The operator can increase f heat transfer by reducing steam pressure or by raising steam generator level. He can decrease heat transfer by allowing the steam generator
! pressure to increase. f. 20 DATE.: 7-6-82 PAGE t__.--~,,_._ > _-,,,_,,._,,,,_.,,_,__.._,_,__,___,,,_,____..,.....,.,.__,,,.__,..,___,,,_,_,_,,,,,_m,m._..,,,._,_....~.
BWNP-20007 (6-76) BABCOCK & WILCOX Nuuset NUctEAR POWER GENERATION DIVI $loN 74-ii2ss31-00 TECHNICAL. DOCUMENT FOOTNOTE Equations la and 2a have been simplified to show the general heat trans-fer process. To be complete additional heat transfer terms would have to be included. All of the vtcer that flows through the reactor coolant system loops does not flow through the core and get all the way to the steam generators. Some flow is let down to the makeup system, some goes to the pressurizer spray and there is some " leakage" through spaces in the internals. This amount of flow is small and it has been ignored for these equations. Also, all the heat of the core does not go to the stean ge nerators ; some of it is lost through the piping to the reactor building or through the letdown water. But this amount of heat is small compared to the total amount and it has been neglected. Heat is also added by reactor coolant pumps (as in plant heatup to power operation), but it is small compared to core heat when the reactor is at power (but the reactor coolant pumps are a large heat source af ter trip or at low power). Control of heat transfer requires control of all the parameters in these l two equations. Some are fixed by design or properties of fluids; the re-mainder can be influenced by the operator. The general methods of heat I transfer control are to be discussed next. Control of Heat Transfer The preferred way of removing heat from the core is to transfer the heat to the reactor coolant and then transfer the reactor coolant heat to the secondary fluid in the steam generators. Steam generator heat removal O PAGE 2I DATE: 7-6-82 l l
. . _ - - - . . . . - ._ - .- - . _ -=-
BWNP-20007 (6-76) BABCOCK & WILCOX m un wucteam rowse oewmanon oms.ou 7'-i i 2333i-oo TECHNICAL DOCUMENT is controlled by adjusting steam pressure and feedwater. To keep the s c ore-t o-s tean generator heat transfer in balance the heat removat rate from the steam generators must be equal to the heat generation rate of the core. In order te balance the heat removal two very basic condi-tions must be satisfied: 1) There must be enough liquid reactor coolant
/
in the vessel and piping to trans fe r the heat to the steam generators, J. and 2) the steam generator pressure and level (feedwater flow rate) must I be balanced at the correct heat removal rate. Figure 1 illustrates these sundamental methods of heat transfer control. Figure 1 shows the controls that the operator can use to change heat , transfer. i The five fundamental functions of heat transfer control are:
- Reactivity control (core heat output control) i
- Reactor coolant pressure control
- Reactor coolant inventory control ;
i - Stean generator pressure control
- Steam generator inventory control f
i When an abnonnal transient occurs, one or more of these five ft.ne t ions will be out of control. It is the operator's job to determine which are out of control, and to make corrections to restore the right heat transfer balance so the core heat can be removed by the steam generators. 4 "A M M
BWNP-20007 (6-76) BABCOCK & WILCOX NumsEn NJCLEAR POWER GENERATsON OlVISION 74-1125531-00 TECHNICAL DOCUMENT
- 1. Reactivity control - Reactivity control is usually taken care of automatically by ICS rod control or by reactor trip. Reactor trip lowers the core heat output to the decay heat level.
- 2. Reactor Coolant Inventory Control - The link between the core and the steam generator is the reactor coolant. It is tLe fluid which transports the heat. To do its job best the coolant should be in a liquid state, that is, subcooled. (Discussion of sub-cooling is given in Addendum A.)
- 3. Reactor Coolant Pressure Control - The reactor coolant system is pressurized to keep the reactor coolant in a liquid state.
- 4. Steam Generator Inventory Control - The reactor coolant trans fers its heat to the water and the steam in the secondary side of the steam generator. The water-steam inventory is the heat transfer fluid which removes the heat from the reactor coolant. In order for it to remove heat at the correct rate the amcunt of fluid (steam generator level) and its flow rate through the steam generator (feedwater flow) must be centro 11ed.
- 5. Steam Generator Pressure Control - The water temperature of the reactor coolant is best controlled by controlling the pressure of the steam generator. In combination with reactor coolant pres-sure control, steam generator pressure control will maintain the reactor coolant in a subcooled liquid state.
Each one of these control functions will be discussed individually as they relate to heat trans fer. pAGE 23 DATE: 7-6-82
. - ~ - - . - - .. ._ _ . . .. .. .. . . - ... - - . _
4' BWNP-20007 (6-76) BABCOCK & WILCOX NumEn NUGEAR POWER GENERADoN DIYl510N TECHNICAL DOCUMENT 74-1125531-00 l t Steam Generator Pressure Control a Heat transfer from the reactor coolant to the steam generators goes to both the steam . and water in the generator. After reactor trip the steam and feedwater in the generator are saturated, and changes of } \ - steam pressure will cause a direct change in the saturaticn temperature of the steam and of the feedwater. A review of the saturated water and i
- steam properties will show how much the steam and water temperature are changed by increasing and decreasing steam pressure. There are situa-tions where the operator controls the steam pressure by manually in-
! creasing or decreasing steam pressure using the turbine bypass valves i or the atmospheric vent valves. When the steam pressure is lowered the v heat trana fer from the reactor coolant to the steam generator increases because the steam and water in the steam ~ generator become a colder heat sink causing more heat to flow away from the reactor coolant. Two I ' reasons - combine to create the colder heat sink: first, the saturation temperature of the steam and water is reduced by lowering the steam pressure which causes the rate of boilof f to increase. The increased boiloff takes away more heat. Second, the increased boilof f requires
% more feedwater flow to be added to maintain level. The fecdwater inlet temperature is colder thea the water already in the steam generator and so its addition contributes to the colder heat sink. Because a colder secondary heat sink exists the primary side temperature will drop as heat is trans ferred.
l I % l l l 7-6-82 PAGE 24 DATE: l
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BWNP-20007 (6-76) BABCOCK & WILCOX III NUCitAR POWtt GENERATION DIVl3 TON 7'-112333i-00 TECHNICAL DOCUMENT Steam pressure can be lowered in two ways:
- By opening the stean line and releasing stean (turbine bypass valves, steam line break, atmospheric veet valves, steam to AFW pump turbine).
- By spraying cold Auxiliary Feedwater into the steam space and con-densing it. This is similar to the way pressurizer pressure is reduced by the pressurizer spray.
Steam pressure can also increase; but normally it will only increase from the operating condition to the reactor trip condition where it will be limited by the steam safeties or by the turbine bypass valves, so the affect on reactor coolant temperature is small. But if steam pressure is low because of a failure, for example a steam line break, the change in reactor coolant temperature could be much larger. When the steam break is isolated the reactor coolant adds heat to the gene-rator and causes the steam pres sur e to increase. The operator can limit the increase in reactor coolant tempe rature under these condi-tions by lowering the steam pressure with AVVs and keeping steam pressure low. Steam Generator Inventory Control Heat t rans fe r from the reactor coolant goes to both the steam and the feedwater in the secondary side of the steem generators. When changes of feed water flow or steam pressure occur the volumes occupied by the steau or water will change and the heat transfer will change. For ex-ample , when the volume of water increases, it occupies space formerly occu pied by stean, so the volume of steam has to decrease. This DATE: PAGE 25 7-6-82
1 i, BWNP-20007 (6-76)
- BABCOCK & WILCOX NUMBEk NUCLEAR POWER OfNERADON Olvi5 TON 7'- 23331-00
- TECNNICAL DOCUMENT changes the relative amounts of OTSG tube surface area covered by water and stean. Because water has a greater heat capacity than steam does, it is a better heat sink for heat transfer from the reactor coolant than steam is. Simply stated there are more pounds of water in a cubic l
foot to absorb heat than there are of steam. If the water inventory in-A g creases then the generator will become a better heat sink for the reac-1 i tor coolant, but if the water inventory decreases or is lost the genera-tor will lose some or all of its ability to absorb heat from the reactor coolant. For example, after trip when the core heat is nearly constant, if the ! w water level in the steam generator is raised rapidly without changing steau pr es s ur e , the reactor coolant tempera ture will drop and stay low unt il the feedwater addition reaches a new level and that level is held. Once the new level is fixed the reactor coolant will reheat and I temperatures will return close to their former values. l l This cooling ef feet of feedwater is caused by the inlet feedwater i temperature which is colder than the general temperature of the bulk of the fluid in the steam generator. The inlet feedwater temperature allows a colder heat sink to be established in the steam generator. i l i The stean generator level can, however, be increased slowly after trip 4 without a large drop of reactor coolant tempera ture by controlling the ' rate of addition of feedwater. , w I ( DATE: PAGE 7-6-82 26 v-- m- v , ,
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BWNP-20007 (6-76) BABCOCK & WILCOX Numsen NUctt*R Powta GENERATION DIVISION TECHNICAL DOCUMENT 74-1125531-00 Too much inventory can also be the result of overfeeding with the Aux-iliary Feedwater System. Even though its flow rate is lower, Auxiliary Feadwater will have a proportionally larger cooling effect on reactor coolant than main feedwater because: a) it comes on when the reactor is tripped and core heat is lowest, b) it is colder (Tinlet feedwater is less), and c) it has a steam pressure reduction effect that main feedwater does not normally have. On the other hand, if steam generator inventory is too low (insuffi-cient feedwater or loss of feedwater can lower the water level), the reduced heat sink will not allow the reactor coolant to transfer all of its heat to the steam generator. When the steam generator's heat sink is reduced, the reactor coolant must retain more of the core heat and it will heat up. For example, if all feedwater is lost, the water in the generator will boil away and only steam will remain to remove heat. But because the steam does not have enough heat capacity, the reactor coolant must re-tain the core heat and the reactor coolant temperatures will increase. When all feedwater is lost the reactor coolant pressure will increase to the PORV setpoint and the reactor coolant will eventually become O
~t DATE 7-6-82 PAGE 27
BWNP-20007 (6-76) , BABCOCK & WILCOX ER NUCLEAR POWER GENERATION DIVl510N 74-1125531-00 g TECHNICAL DOCUMENT saturated as the core continues to add heat. The steam remaining in the generator will flow out through the steam lines and steam pressure will drop; loss of the steam eliminates the heat sink of the steam generators altogether, l \ Finally, another part of steam generator inventory control is feedwater t empe rat ure. The heat sink of the generators will be af fected by an ! abnormally low feedwater temperature. A reduction of feedwater heating i steam or loss of a feedwater heater will cause reactor coolant tempera-tuce to decrease. Usually ICS operation will stabilize the plant, but i the decreased feed temperature will cause a change in the heat sink and an increase of heat transfer from the reactor coolant. The operator should ensure the rate of feedwater addition is controlled properly to maintain the steam generator inventory. Level measurements in the steam generator downcomer give a good indication of the steam generator inventory for control. Reactor Coolant Inventory Control
- Reactor coolant heat t rans fer can be af fected by changes in the amount t i
I of inass of fluid in the reactor coolant system or by changes in the den-l i i sity of the reactor coolant. Several ways exist to vary the mass of reactor coolant: LOCA or small ,. l break, and changes in HPI or makeup, RC pump seal injection, seal PAGE 28 ! DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCLEAR Powtt GENERATION DIVI $loN TECHNICAL. DOCUMENT 74-1125531-00 return, and letd3wn. Several ways also exist to vary the density of the reactor coolant. As shown by the previous discussions of steam generator pressure and inventory control, changes of the rate of heat trans fer from the reactor coolant to the steam generator can cause the reactor coolant to cool down when the steam generators remove too much heat (low steam pressure, too much feedwater); or the reactor coolant can heat up when the steam generators don't remove enough heat (not enough fe edwa t e r ) . These effects cause density changes in the reactor coolant; t.he coolant contracts or expands accordingly. Regardless of the cause, changes in invento ry in the reactor coolant system have two effects:
- 1) A loss of mass can affect the ability of the reactor coolant to transport heat from the core to the steam generators. If the RC pumps are not running steam can collect in the hot legs and block natural circulation. When circulation stops and heat transport stops then the steam generator temperature will not " set" the temperature of the reactor coolant; Tcold will not ch ange when Tsat-SG changes.
l If the mass of the reactor coolant system continues to decrease and the cow is mostly covered by steam, the mass of the RC will not provide a sufficient heat sink and the core will retain the heat and heatup. Fuel failures can result if this situation is not corrected. DATE: 7-6-82 PAGE 29 l
BWNP-20007 (6-76) SABCOCK & WILCOX NI y NUCLEAR POWER GENERATION DIVISION 74-1125531-00 TECHNICAL DOCUMENT 4
- 2) A change of mass orr dens ity can a f fect the ability of the pres-surizer to provide pressure control of the reactor coolant system i
(this will be discussed next under Reactor Coolant Pressure i Control). Operator control of reactor coolant inventory requires the ability to N I balance mass increases or decreases by adding water with makeup or ECCS I or removing mass with the letdown. Control of reactor coolant dens ity changes requires control of the steam generator pressure and inventory. t The inventory of the reactor coolant system cannot be measured direct-ly. But the operator has two indications to determine if the inventory is su f ficient fo r core cooling. Pressurizer level is an accurate mea-sure of the inventory when the rest of the reactor coolant is subcooled (except fo r a rare possibility when free hydrogen gas may exist in the loops; this condition will likely exist only after fuel failures caused by uncove ring of the core). The other measure is the incore thermocouples; if these read subcooled or saturation temperature then enough mass exists in the reactor vessel to cover and cool the core. But the incore thermocouples will not show if the loops are full. Reactor Coolant Pressure Control Keactor coolant pressure control is required to keep the reactor cool-ant subcooled so the coolant is in the best state to trans fe r the heat
\
DATE: 7-6-82 30 l_ __. - _ _ _ , _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,
BWNP-20007 (6-76) BABCOCK & WILCOX Nuxut NUCLEAR POWER GENERATION DIVISION TECHNICAL DOCUMENT 74-1125531-00 from the core to the steam generators. For all cases of reactor opera-tion except LOCA's, RCS pressure control is provided by the pressuri-zer. (Reactor coolant pressure control is different for LOCA's and small breaks than for other plant conditions. It is discussed in de-tail in Appendix F.) Use of pressurizer heaters and spray is the usual way of increasing and decreasing RCS pressure when a steam and water in-terface exists in the pressurizer. The purpose of the heaters is to maintain the reactor coolant in a subcooled condition; the spray re-tards pressure increases to limit operation of the pressurizer relief and safety valves. Neither the heaters nor spray have enough capacity to prevent large abrupt pressure changes, but they can moderate small ch ange s . As a backup the PORV can be used to reduce pressure but it is not as desirable to use as the spray because it relieves to the pres-surizer quench tank, cons eq ue ntly , frequent use of the PORV can cause the quench tank rupture disk to blow. RCS pressure control by the pressurizer can be lost in two ways: l i i
- 1) The steam-water interface in the presurizer can be lost i
either by draining the pressurizer or if the pressurizer l l fills solid with water l
- 2) The heaters and spray can fail.
l l l Each of these is discussed below: 1 PAGE 31 DATE: 7-6-82
BWNP-20007 (6-76) t SABCOCK & WILCOX " l NUCLEAR POWER GENERATION OtVISION
- t i 25 531-00 l TECHNICAL DOCUMENT i
Draining the Pressurizer: l If the pressurizer level drops suf ficiently to uncover the heaters , the i heaters cannot provide pressure control because no water is available to be boiled by the heaters to create steam. If the pressurizer drains
,comple tely, RCS pressure will then be controlled by the highest fluid l
/ temperature in the system.
l s/ When the pres sur ize r drains the reactor coolant system pressure will l ! decrease to the saturation conditions corresponding 'to the hottest i point in the system, which could initially be the hot leg containing the pressurizer surge line, the other hot leg, an the core outlet. In e f fect , the hottest point becomes surge volume. V Filling the Pressurizer: Spray depressurizes the reactor coolant system by condensing the steam in the pressurizer. If the pressurizer fills with water, the spray can-not be ef fective for depressurizing because the steam space is lost. When the pressurizer fills, the reactor coolant system may or may not s lose subcooling and become saturated depending on what caused it to fill. If the filling was caused by makeup and the steam generator is still removing heat, then the RCS will stay subcooled because the make-up pumps will cause the pressure to stay high and the steam generator i DATE: PAGE 7-6-82 32 l
BWNP-20007 (6-76) BABCOCK & WILCOX Numset NUCLEAR POWER GENERATION DIVl510N 74-1125531-00 TECHNICAL DOCUMENT will keep the temperature controlled. If the filling was caused by heatup and swell because the steam generators were not removing enough heat, then the system may become saturated because the heat from the core will go only into the reactor coolant and not out the steam genera to rs. When the pressurizer fills, either because of heating the reactor cool-ant or because of too much makeup, the water will be lost through the PORV and pres sur ize r safety valves. This loss is cons idered to be a LOCA, even if the action was deliberately done. The operator should not confuse the above losses of RCS pressure control with the normal pressure responses during plant transients. Any transient wh ich rapidly decreases pre s surize r level will cause a concurrent drop in RCS pressure. The rapid decrease in level (except for a LOCA) is caused by a temperature decrease in the RCS. The drop in RCS pressure is caused by the expansion of the pressurizer steam space. Addi t ional coolant will flash to steam in the pressurizer as the steam volume expands, which will reduce the rate of pressure de-crease. This flashing of steam, however, has a cooling ef feet on the remaining coolant in the pressurizer since the latent heat of vapora-tion is required to convert saturated water to steam. These transients are those rapid enough so that the pressurizer heaters can not keep up with this cooling ef fect. O DATE: 7-6-82 33
BWNP-20007 (6-76) BABCOCK & WILCOX Numen NUCLEAR POWN OtNBATION DIVISION 74- 12553i-00 TECMICAL DOCUMEllT l Any transient which rapidly increases pressurizer level will cause :s concurrent rise in RCS pressure initially. This rise in RCS pressure is usually followed by a . decrease in RCS pressere unless the pressuri-zer heaters can maintain pressure. The rapid increase in level is usually caused by a temperature increase in the RCS. The initial rise in RCS pressure is caused by the compression of the steam volume in the pressurizer, since the insurge acts like a piston. Three things happen i to reduce this initial pressure increase: The spray effects start to show up; some of the steam condenses on the water surface (releasing the latent heat of condensation and the cooler water which entered the bottom of the pressurizer starts to mix with the hotter water, wh ich in turn, increases the condensinr, of steau. (In addition the spray bypass k -flow contributes some cooling af ter main spray stops, which is normally balanced by the pressurizer heaters under steady state conditions). This cooling of the pressurizer may exceed the capacity of the pressuri-zer heaters to maintain pressure, so that the initial increase in RCS j pressure is followed by a large decrease in RCS pressure even though s pres sur ize r level is now stable. The operator can anticipate this above situation and manually activate the pressurizer heaters. The operator should also limit the amount of pressurizer refill to 100 inches. By limiting the amount of water in the pressurizer the heaters will be able to heat the water quicker. DATE: PAGE 34 7-6-82 l
BWNP-20007 (6-76) BABCOCK & WILCOX " " NUCLEAR POWER GENERATON DIVISION 74- 125531-oo TECHNICAL DOCUMENT Failure of Heaters and Spray: A failure of the spray and heaters in the pres surizer control system can also cause a loss of pressure control. If the spray fails and can-not be turned off the system will depressurize. Depressurization may also occur if the heaters fail in the "of f" mode . The reverse is not true; failure of the spray in the "o f f" mode will only limit the ab ility to depressurize. Unless soue thing else happens to the plant, pres s ur e increases and decreases will not occur. If the heaters fail "on" pressure increases will not occur because the spray will operate to provide a balance. However, if the spray is not working then the heaters can cause the system to pressurize and cause coolant (steam) to be lost through the PORV and the pres surizer safety valves; subcooling will not be lost as long as water covers the heaters. When only steam c ove rs the hea ters they will no longer raise pressure and subcooling can gradually drop. If the heatecs fail "on" when they are uncovered, no water exists to cool them and they will burn out. Reactivity Control Reactivity control is usually taken care of au tomat ically by ICS rod i l control or by reactor trip. Reactor trip lowers the core heat output to the decay heat level. The operato r must verify rod insertion and l l decreasing reactor power to ensure the reactivity control systems func-tion properly. After the trip no more heat trans fe r control can be ! achieved by use of the rods, unless the rods did not fully insert. If 1 one or more rods are stuck out after trip the operator should manually l l DATE: PAGE 35
BWNP-20007 (6-76)
, SABCOCK & WILCOX " O j
7'- 12n31-00 C ~ NUCLEAR POWEA GENERADON OfVI$10N TECHNICAL DOCUMENT 1 ( If one or more rods reraain stuck out the operator should trip them. begin emergency boration and a reactivity balance calculation should be performed to ensure a shutdown margin in excess of 1% Ak/k is achieved. Summary s The preceding discussion introduced the concept of reactor-steam genera-tor heat transfer and the balance that heat transfer must have. When an imbalance of heat trans fer occurs, its ef fects will often'be trans-mitted throughout the steam and reactor coolant systems. The purpose a ~ of unde rs tanding heat transfer is to understand its ef fects so the operator can step in and diagnose what has gone wrong and correct it. E An understanding of the major influences on reactor-steam generator heat transfer control (reactor coolant system inventory control, reactor coolant pressure control, s tean generator pressure control, steam genera tor inventory control, and reactivity control) will allow the~ operator to focus on achieving controlled heat trans fer and stable f plant conditions without necessitating the ide nt ifica t ion of specific failures. Thus, an understanding of the principles of heat transfer and the control methods permits a more direct and ef ficient approach to N abnormal transient diagnosis and correction. 1 i The ef fect s of changing one of the controls will nearly always cause changes in other parts of the system and there fore will require other controls to be changed to balance heat trans fe r. The controls are interdependent because they affect total heat transfer from core to s s tean outlet. t DATE: PAGE 7-6-82 I , . _ _ _ _ . . _ _ - - ._,_, -, _ __._._. _ _ _ _.,.__ _ _ _.._-._ _ __.__.,_._ ._._ _ _ .____.... _ .__ _
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWER GENERATION OlVI$10N
"- 25531-00 TECHNICAL DOCUMENT Core cooling with the steam generator can occur as long as two things exist:
The reactor coolant can transport the heat. The best way to do this is with subcooled liquid. Reactor Coolant In-ventory and Pressure Control contribute towards this. The heat removal is controlled by the steam generator. Stean Generator Inventory Control and Pressure Control aid this. Usually an abnonnal transient will be caused by a failure of one or core of the heat transfer controls. The understanding of the control influences allows the operator the freedom of two approaches to abnormal transient correction:
- 1. He can treat the symptoms by manipulating equipment to regain heat O
transfer control without knowing exactly which equipment has failed. Consequently, proper heat transfer can be restored quic k-er and more accurately than if the operator had to hunt for the equipment failure. In some instances, treating the symptoms will also uncover the failed equipment.
- 2. He can use these control failures as symptoms of poor heat trans-O fer to discover the equipment that has failed and by doing so, isolate it, remove it from service, or repair the equipment.
Understanding the influence each of these controls has on overall heat trans fer will also give an understanding of what the outcome of an DATE: PAGE
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER j NUCLEAR POWER GENWAHON OfVI510N 74-1125531-00 ! TECHICAL DOCUMEllT action is. All operator actions will have some cansequence to heat 7 trans fe r and a knowledge of the heat trans fer will allow judgements to j i l be made about the general effects. i 1 ' 1 Table 1 is a summary of the previous discussion. Like all summaries, l material has been condensed. When that happens, information has been i lef t out. The table should be used only to provide an overview. , P f The next section builds on the information about heat transfer and I extends those principles into a disciplined approach to accident diagnosis and recovery. f I i i DATE: PAGE 7-6-82 38
Figure 1 FUNDAMENTAL METHODS OF HEAT i TRANSFER CONTROL i SPRAY PORV STEAM
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) j HEATERS STEAM GENERATOR PRESSURE CONTROL REACTIVIT)
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BWNP-20007 (6-76) BABCOCK & WILCOX "" NUCLEAR POWER GENERAHON DIVISION 74-1125531-00 O TECHNICAL DOCUMENT U ADDENDUM A (SifBCOOLED, SATURATED, SUPERHEATED WATER) The state (solid, liquid or gas) of the water in the reactor coolant system or the stean system is determined by the pressure and temperature conditions which exist. The terms subcooled, saturated, and supe rheated are normally i used within operating procedures. These terms mean the following: b S ubcooled : Water can exist only in the liquid state. Saturated: If heat is added to subcooled water a temperat ure , for the existing pressura, will be reached where the water can exist cither as a liquid or as a gas (steam). At this point, the liquid is called satuated water and the gas is called saturated steam. The liquid and steam s states both can exist at this temperature and pressure. Heat must be added to saturated water to change it to saturated steam. Heat must also be removed from satu-rated steam to change it to saturated water. The heat required to make the change is called the latent heat of vaporization for heat added and the latent heat of con- - densation for heat removed. Superheated: Water can exist only in a gaseous (or steam state). , This phase can be dis tinguished from saturated condi-tions because the temperature will be higher than the saturation temperature for the existing pressure. DATE: PAGE 7-6-82 39
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWER GENERATION Olvl510N 7'- 12ss3 t-oo TECHNICAL DOCUMENT The normal state af the steam coming out of the steam generator is superheated during power operation and saturated af ter trip. The state of the reactor coolant can be determined by watching the RCS pressure and temperature on a pressure-temperature diagram (see below): SUBC00 LED d SATURATION LINE P SUPERHEATED T r P-T condicions which are to the left and above the saturation line are in the s ub cooled region (or state), and P-T conditions to the right and below the saturation line are in the superheated region (or state). Subcooling O Subcooled conditions are maintained in the reactor coolant system (except pressurizer) during normal operation. During a reactor transient it is desirable to maintain the reactor coolant subcooled. When subcooled : O DATE: 7-6-82 PAGE 40
EUNP-20007 (6-76) BASCOCK & WILCOX Numeen NUCLEAR POwte OtNORATION OlVISION
' 74- 2n31-00 TECHNICAL DOCUMENT
- 1. The primary loop s are solid water and a water level is present within the pressurizer. (Assuming no more condensable gas voids.)
- 2. The pres sur ize r water level is a true measurement of RCS inventory.
The inventory can be controlled by regulation of pressurizer level by the MU system and letdown. (NOTE: A very special case can exist t when the reactor coolant is subcooled and a water level is in the pres-surizer but the loops are nor full. In that case pressurizer level is not a true measurement of inventory. That condition is when there is
- a large amount of free gases in the loop. The gases will be mostly H, 2 that probably have been created after a large amount of fuel failure. Since this would be an uncommon event, reliance on pres sur ize r level is usually ecceptable when the reactor coolant is F
subcooled . )
- 3. The reactor coolant is liquid and is ideal for heat removal from the core and heat transport to the steam generator by either forced or natural circulation.
- 4. RC pressure can be maintained by the pressurizer and can be regulated by using normal procedures and equipment (spray and heaters).
..- 5. RC tempe ra tur e can be controlled by the seconda ry system (with fe ed-
\ water available) by adjusting feedwater flow and steam pressure. -
1 Subcooling should be checked in all parts of the loop especially when natural circulation is removing heat. The operator should check Th ot and Teold in both loops and the core exit thermocouples. Anytime the reactor coolant gets too close to being saturated, the makeup or HPI system should be turned on s V full. DATE: PAGE 7-6-82 41
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCLEAR POWER GENERATION DIVI 540N 74- 12553 t-00 TECHNICAL DOCUMENT The reactar coolant is too close to being sa tur a t ed if it exceeds a subcool-ing margin limit wh ich is a limit about 20 to 50F subcooled. This subcooling margin is explained in more detail in Chapter B. Subcooling Rule: When ever the reactor coolant sub-cooling margin is lost:
- Two makeup pumps should be run at full MU system capacity taking suction from the BWST AND
- Two HPI pumps should be run at full HPI system capacity when the reactor coolant pressure is < the low RC pres-sure SFAS actuation setpoint (1650 psig).
Saturation A loss of subcooling can happen when the pres sur izer drains or when filled solid (if the pres sur ize r is water solid and cooling is by the steam genera to rs or MU/HPI cooling is greater than core heating then the Reactor Coolant can stay subc ooled) . A loss of subcooling can be caused by an over-heating or overcooling transient or a loss of reactor coolant. Saturated l l l c ondi t ions can exist in isolated pockets of the loop (i.e., within one or both hot i<g pipes or the reactor vessel head but not in the cold leg pipes) or within the system as a whole, as would be the case during a major LOCA. Th e r e fo re , temperatures should be checked in the hot legs of both loops. l l When the RCS is saturated: l 1. The reactor coolant temperature and pressure will not show whether the l l saturated fluid is liquid or gas (steam). s DATE: PA E 42 7-6-82
BWNP-20007 (6-76) 1-SABCOCK & WitCOX Nuune NUCLEAR POWEs GENE 4 TION OlV1580N
' 74-1 2s531-o0 TECMICAL DOCUMENT _
! 2. Voids (steam bubbles or pockets) can exist within the primary system.
'3. The pres sur ize r water level indication is not a true measurement of reactor coolant inventory.
i '4. If the RC pumps are off a loss of natural circulation may occur because steam voids can form at the top of the hot leg and block water flow.
- s. Normal. pressure control by the pressurizer has been lost. The RCS hot leg loops, which have a steam bthble at the top, now work as a surge volume . RC pressure will be controlled by the highes t fluid tempe ra-ture in the system. The amount of steam can change because of steam condensation by the steam generators, by addition of cold HPI water, or by loss of steam generator heat removal.
I Under ideal conditions subcooling should exist in all parts of the reactor coolant loop to be able to transport heat from the core to the steam gene r-ators. However, given the proper conditions, the steam generators can remove i l heat when the reactor coolant is saturated. For all events, except a LOCA or a total loss of secondary fluid, saturated conditions should be a temporary e f fec t . For example, if inteam generator ove rcooli ng causes the Pressurizer to drain, saturation will occur, but MU or HPI will restcre the reactor coolant to a subcooled state. , i DATE: PAGE 43 I 7-6-82 i I. , , . . - - - . - , , - . . - - _ _ _ . . _ _ __
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen WCLEAR POWit GENERATION DIVI 510N TECHNICAL DOCUMENT 74-1i 2 s 33i-00 Superheating Superheated reactor coolant conditions are to the right and below the satu-ration line of the P-T diagram. Superheated steam results when the core is uncovered. Heat from the core is passed to the steam and its temperature rises above saturation. When the reactor coolant is superheated the core is cooled by steam. Steam cannot remove enough heat to prevent the core and clad from heating up. Fuel failure may result. Supe rhea ted steam ind icate s Inadequate Core Cooling (ICC). When the RCS is supe rhe a t ed the only accurate measure of tempe rature is the incore thermocouples, and they should be used along with hot leg pressure to determine the amount of superheating . Superheat nule: The inadequate core cooling O i procedure must be used anytime superheated l , conditions exist in the RCS. See " Backup l Cooling Methods" section for a discussion of Inadequate Core Cooling. O l l O DATE: PAGE 7-6-82 44 1 l
1 BWNP-20007 (6-76) SABCOCK & WILCOX Numun NUCLEAR POWER GENERATION DIVISION 74- 125531-oo TECHNICAL DOCUMENT ADDENDUM B NATURAL CIRCULATION When the reactor coolan*. pumps are tripped forced circulation is lost and an ~ alternate anthod of removing core decay heat must be found. The preferred method is to transport this heat to the steam generators by natural cir::u la-l tion of the reactor coolant. Natural circulation is possible as long as the following requirements are met: 1) a heat source is available to prod uce ~ warm (low density) water; 2) a heat sink is available to produce cold (high f d$nsity) water; 3) a flow path (loop) is available connecting the warm and cold water;. and 4) the cold water is at a higher elevation than the warm j water. Requirements 1, 2 and 3 are met by the following: decay heat in the core is the heat source, water on . the secondary side of the steam generators provides a heat sink, and the hot and cold legs connect the two. Requirement 4, "the cold water is above the warm water," involve s a concept called ther-mal center. In reality heat is transferred continuously as the water moves up through the core and again as it moves down through the steam generator. The thermal center is the point in the core or the steam generator where the
- primary water is at average temperature. It can be used to represent the entire column of water in its " average" conditions.
Thermal Center Definition
- 1. Core thennal center: That elevation in the core which the coolant may be considered to go from Tcold to Thot-l 1
DATE: 7-6-82 PAGE j 45
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCLEAR POWit GENERATION DivaSION
- n 2 5 5 3 i-00 TECHNICAL DOCUMEMT
- 2. Steam gene ra tor thermal center: That elevation in the steam generator 0 at wh ich the reactor coolant may be considered to go from Thot to Tcold-Requirement 4 for natural circulation can be met if the thennal centes of the steam gene ra tor is at a higher elevation than the thermal center of the core.
This will put the " average" cold water above the " average" hot water, the cold water (more dense) will sink, the hot water (less dense) will rise and there will be circulation. The rate of natural circulation (gpm) depends on the following things: e The friction (resistance to flow) of the piping and components around the primary loops: this is determined when the plant is designed and built; the operator has no control over it. e The strength of the heat source: this depends on the available decay heat, which is a function of past power history and time since the reactor trip. It will, of course, decrease with time after trip. The ope ra tor has no control of this after trip except to make sure the reactor is shut down so that the only heat input is decay heat, e The strength of the heat sink: the colder the heat sink is, the more it will be able to cool the primary coolant passing through the steam generator. This will make the water more dense and the natural circulation flowrat e will increase. The operator can make the heat O DATE: 7-6-82 46
BWNP-20007 (6-76) BABCOCK & WILCCX wumun wuctua nrven oewswoon oevision l ,D TECHICAL DOCUMENT 74- 123331-00 sink colder by 1) lowering secondary steam pressure (opening the tur-I bine bypass valves or AVVs more), this vill lower secondary saturation temperature which will increase heat transfer across the tubes; or
- 2) lower feedwater temperature this will increase the heat trans fer i
across the tubes by praviding a larger primary to secondary AT.
\ e Di f ference in he igh t between the core thermal center and the steam generator thermal center: As the dif ference in height becomes larger l
the natural circulation flow will increase. The core the rmal center l ! is fixed, but the operator can control the steam generator thermal center by two methods: 1) most of the heat trans fe r occurs in the ! violent boiling area just below the established secondary side water level 3 Therefo re, the operator can raise the thermal center by l i raising the steam generator water level; 2) the operator can add AFW. This injects feedwater high in the generator and thereby raise the average heigh t (the rmal center) of heat removal. This only works l while AFW is being added. If AFW is stopped, the thermal center will move back down to just below the water level. In summary, the natural circulation flowrate can be changed by changing the difference in temperature (density) between the hot water and the cold water or, changing the di f ference in height between the core thermal center and steam generator thermal center. This can be expressed in equation form as: 1 l l N DATE: 7-6-82 PAGE
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen NUcit AE POwlt GENteATION OtVISION 74- n 2 s s 31-Oo TECHNICAL DOCUMENT where: Pdriving head = available driving head for natural circulation hegg ' distance between core thermal center and steam generator thermal center (ef fective heigi.t) pc = density of cold water at steam generator thermal center P H = density of hot water at core thermal center This is shown graphically in Figure 2. Natural Circulation (All Other Conditions Normal) When the reactor coolant pumps are tripped the ope ra tor should check two th ings to make sure natural circulation is being initiated properly. First he should make sure the reactor coolant remains subcooled . If it does not he should make every e f fo r t to restore subcooling (the methods for doing this are discussed in the accident mitigation chapter of these guidelines). Second , he should make sure that AFW is feeding both SGs so that the thermal center is being raised in both steam generators. Normally automa t ic equipment will start AFW and maintain a steam generator level of 40" on the SU range of each steam generator (or 93" if SFAS level 2 actuates) when the RC pumps trip. The operator should monitor this process while keeping the following in mind: l I e As long as AFW is flowing at suf ficiently high rates into the top of l the gene rator it is not necessary to get a level in the generator to h ave natural circulation. If the heat source (decay heat) is high l ! DATE: PAGE ! 7-6-82 48
, - - . - - - _ - . - . - - _ . - . - . - -.- - - .~.-. -- ---
l ' BWNP-20007 (6-76) s t BASCOCK & WILCOX NUMStR , nucteam powes oewenaren omsen i 1 J 7'-l i 23s 3 i-00 TECHNICAL 00COMENT i l enough, the AW may come in and boil right of f and go out as steam. I } This is receptable; the thermal center is high and natural circulation ' I - j will develop. j l s If AW is not ' available natural circulation can be initiated using ; i main feed wa ter. Again, the level should be raised to 40" on the Sif ' ' range when the RC pumps are tripped. . ; A e If MW is to be used for natural circulation either 1) the required , f , i
- s tean generator level should be established before the RC pumps are I tripped or 2) the steam generator level should be established first by AW.
l Figure 3 shows how RCS t einpe ra ture and pressure, and steam generator tem-I i ' ] perature and pressure will vary during the transition to natural circula-i tion using AW. Approximate times for the transient are also included. , f The times are approximate because the rate of recovery of the steam pres- . sure depends on the amount of decay heat available. When steady state is i t reached , the cold leg temperatures (Teold) will be just about equal to the saturation tempera t ure in the steam generators. The hot leg and incore ! thennocou ple temperatures will increase as necessary to develop the driv- > i i 1 ing head required for flow (by developing a density change between Th and Tc ). The best measur es to use to see if natural circulation has started i i are when the incore ' T/C temperature follows the steam generator tempe ra- ! i ture and coupling between T and the steam generator t empe ra tur e . These
; e f are checked by lowering the secondary pressure and verifing that the reac-tor coolant incore thermocouple temperatures decrease and the seconda ry 1
I < l 4 i DATE: 7-6-82 49
. - - ____.____--,m-_.-,,-__,,_ _ - . _ _ _ _ . _ - . _ , _ . _ _ -
BWNP-20007 (6-76) BABCOCK & WILCOX I NUCLEAR POWEa GENERAh0N DIVISION 74- 123s31-00 TECHNICAL. DOCUMENT p res sur e Tsat and T e become approximately equal and change together. Confirmation of natural circulation can be seen by the relationship of Th and the incore th e rmo c ou pl e s , and the temperature difference between Th and T. When the incore the rmocouple s , T h, and Tc are subc ooled , they c should fo llow s t ean generator T sat when it ch ange s ; the temperature dif ference between Th and Te should be 50-60F (maximum decay heat with both SG 1evels at 93" on the SU range) and an average of the five highest incore ;he nnacou ple s should track Th within approximately 10F. However, if T only e is subcooled and Th and the incores are s atur cted , the core AT will be small (because much of the core heat is transfered to the ceactor coolant as latent heat of vaporization) and the incores and Th will read the same since both Th and the incores are s a tur a t ed . Once natural circulation is established the operator must ensure feedwa t er is available to replace the steinn generator water being boiled of f removing decay heat and to maintain the RCS subcooled. Transition to natural circulation using MFW will look ve ry similar. The major dif ference would be slightly less primary system cooldown (with the same feedwater flow rates) while the SG levels are being established due to higher temperature fe ed wa t e r . Natural circulation flow will regulate itself. That is, as the heat source (decay heat) dies down the core AT (T h - T) c will go down and l there will be less driving head available; the re fo re , flow will go down. l Natural Circulation - Abnormal Operation l The discussion so far concerned expected or normal natural circulation conditions. That is, the RCS is subcooled, the level in both steam l E DATE: 7-6-82 50
I BWNP-20007 (6-76) ' BASCOCK & WILCOX "umu wuosas even oewmanon omsion t 74- 12533i-00 TECHNICAL DOCUMENT i e genera tors is 40" on the startup range and both steam generators are being steamed. This sect ion ~ will discuss off normal condi t ions : 1) natural i j circulation with one OTSG, 2) natural circulation with saturated RCS, and 2
- 3) recognition of loss of natural circulation,
, r One OTSG F There may be t ime s when an operator does not want to stean a generator (OT5G tube leak) or cannot steam a generator (steam line break and iso- - t ! lated generator is dry). If he is also in natural circulation he can . 1 expect the following: P 4 Only the Thot in the operating loop will indicate core outlet temperature; Teoid on the operating generator will be equal to Tsat in the operat ing , + { steam generator; eT old in the isolated generator will not be equal to Tsat i
.in the isolated generator, it will probably be much colder being influenc-j ed by seal injection water temperature coming into the idle pumps; (Th-Tc )
on the operating steam' generator may be 410F higher than the 50-60F AT i
- expected with two operating SGs and the level in the operating steam gen-i 1
! erator may have to be raised above 40" on the SU range to maintain ade-l quate natural circulation flow. S tead y state operation under these i conditions is stable and safe. Plant cooldown, however, is complicated becau'se the cooldown of the loop with the isola ted stean generator will 4 lag behind the steaming steam generator. If there is water in the iso-l l lated stean generator it will become a heat source instead of a heat sink. In' fact, the i sola ted generator may add enough heat to cause the reactor 4 DATE: PAGE f 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUCitAR POWER GENERADON DIVISION 74- u 235 31-00 TECHNICAL DOCUMENT coolant in its hot leg to flash to steam. If this happens, that For leg will act as a surge volume and slow down the depressurization during cooldown. This will also slow down the cooldown rate. The operator must carefully watch subcooling in both loops under these conditions and make sure adequate margin is maintained by regulating the rate of cooldown with ste2m pressure control of the operating steam generator. Natural Circulation with a Saturated RCS A subcooled reactor coolant system is the desired 5. tate, however, natural circulation can remove core heat when the RCS is saturated. As long as the four requireme nt s for natural circulation are met, heat will be removed from the core and transferred to the steam generator. The problem with saturated natural circulation is that the operator doesn't know how much of the reactor coolant is steam and how much is water (see discussion of saturation in Addendum A). If the RCS is losing inventory steam will fo rm in the hot legs and eventually stop natural circulation flow (this is i a violation for the requirement that a flow path exists connecting the hot l water and the cold water). This could also be violated by a large col-1 lection of non-condensible gases in top of the hot legs, however, such a collection would probably exist only following a core uncovery. At that point the operator would be using inadequate core cooling procedures. l Another fo nn of natural circulation could still exist under these con-ditions called boiler-condenser cooling (boiling in the core and con-( densing in the s t ean generator) but it requires a higher steam generator 3 DATE: ^ 7-6-82 52
4 EWNP-20007 (6-76) i i BABCOCK & WILCOX ..., wucteas town oewmation oivision 74-1125531-00 !, TECMICAL 00CUMEllT i f level (93" on startup range). This method is discussed in detail in the i Backup Cooling Methods chapter of these guidelines. l 1 The point to remember is that primary inventory (mass) is unknown under , i saturated conditions and therefore, every ef fort should be made to keep i. ! the RCS s ubcooled or if subcooling is lost, to return it to a subcooled , l l ' condition. Recognition of Loss of Natural Circulation i + A loss of natural circulation can occur for various reasons and several in-i dications may be available. If the RCS is subcooled , a loss of natural circulation flow is more than likely a result of inadequate heat removal l i by the steam ge ne rato rs . The thermal center in the stean generators may j L be too low. At low decay heat levels or during single loop cooldown, the I SG Levels may ha*e to be raised above 40" on the startup range to induce or matntain natural circulation flow. When natural circulation flow ex-ists, Thet and the incore thermocouples will track together within 1 10F ll l ( although there will be some time lag due to long loop transport times). l ( _. In addition, T cold and T sat of the SG should track together. l The best single indication of a loss of natural circulation flow when the ' RCS is subcooled is a d ivergence developing between the incore thermo- i cou ple s and Thot. When the flow is lost, the incore thermocouples will begin a continual increase toward saturation. The rate will depend on the decay heat level. Th ot indications may also increase but can actually l i I i DATE: PAGE 7-6-82 53
BWNP-20007 (6-76) BABCOCK & WILCOX " " NUCitAt POWER GENISATION DIVISION 7'- 12s33i-00 TECHNICAL DOCUMENT decrease and begin to conve rge with Tcold. In any case, Thot will nt increase as rapidly as the incore thermocouples and the two indications will diverge. Another indication of loss of natural circulation is a "de-oupling" between T sat in the SG and Tcold. If Tcold ceases to follow T sat natural circu!ation flow is lost. When the RCS is saturated and natural circulation flow is lost, this di-vergence may not develop significantly. The best indication of a loss of natural circulation flow when the RCS is saturated is a trend of incore thermocou ple temperature v re . RCS pressure increasing away from the SG Tur either along the saturation curve or into the superheat region. Flow can be lost due to low thermal centers in the SG's or blockage due to voids in the RCS. When saturated, SG levels should be maintained at 93" on the startup range and full HPI flow should exist. If voids exist in the RCS , it is possible that boiler-condenser cooling was in progress. As the RCS refills, cooling in this manner is expected to be lost when the RCS liquid level increases above the SG tubes. However, in this case cooling should be restored by continued refill and by following the actions specified in the Lack of Heat Transfer Section III.B of Part I (RCP bumps, reducing SG ' pressure, etc.). O DATE: PAGE 7-6-82 54
y F
^
Figu,*e 2 ILLUSTRATION OF PARAMETERS CONTRIBUTING TO NATURAL CIRCULATION DRIVING HEAD ()
.=
NY COLD FLUID COLUMN THERMAL CENTER FOR HEAT REMOVAL HOT FLUID 40" LEVEL COLUMN n
/ )
h er. W - pl - t A Il
/y \ THERMAL CENTER FOR HEAT ADDITION APariving nead = her, % -ph) )
3 g ; _ 1 l ' '. , P, 1 00 !4 3i
r FAure3 TRANSITION TO NATURAL CIRCULATION USING AFW i 2600 -- . POST TRIP 2400 - WINDOW s , 2200 - p , od l4 l [__4_ihI _ 2000 - a l
; SUBC00 LED REGION S 3 SUPERHEAT !
g 1800 - REGION l
- o. .
w 3 1600 - a 3 1400 - j M , END PolNT-POST TRIP WITH FORCED CIRCULATION (T Moi E _ 8 T COLO) anD FOR NATURAL STEAN PRESSURE LIMIT - CIRCULATION (TCOLD) _2 o 1000 - g 0 . MORMAL OPERATING P0lMT-POWER OPE?ATION (T H0T
=
E - SATURATION I EMD P0lNT-P0$T TRIP WITH 600 - [-- u_ j N A TUCIRCULATION I R A L (T HOT I SUBC00 LED 400 - MARGIN LINE i i i i I O 500 550 300 650 700 400 450 Reactor CO0lant and Steam Outlet Temperature, F Reference Time Points (_ Seconds) Remarks 1 0 RCP's trip; reoctor trip. 2 40 Tcold reaches maximum value. 3 6S RCS pressure at ninimum value; recovery of RCS pressure begins. 4 240 SG 1evel boils down to AFW setpoint. II25531 00 4
.- - . _ _- --. - . _ _ = .. ..
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWE6 GENGEATION OtVISCN 74-1125531-00 TECHNICAL DOCUMENT CHAPTER B USE OF THE P-T DIAGRAM Introduction The previous chapter provided the fundamentals of reactor heat trans fer con-trol and also presented information about natural circulation, subcooling, saturation and supe rhea t ing . These basics are the background information needed to diagnose transients and follow through with the correct operator actions. This chapter builds on that information. The foundation for abnormal transient diagnosis and operator action is the reactor coolant pres sur e-t empe ra ture diagram (P-T) which is used to show how [ changes of heat transfer af fect plant operation. On this diagram are plotted reactor coolant temperatures, reactor coolant pressure and secondary system 4 temperature pressure. The reactor coolant temperatures are both plotted ver- r i sus reactor coolant pressure as dots moving with time on the P-T diagram. 1 Secondary steam pressure and steam saturated temperature are plotted as lines t on the P-T diagram. In addition to these variable parameters, the P-T dia- t 1 gram has fixed limits displayed so that the variable parameter can be moni-tored in relation to these fixed limits. Examples of reactor coolant system t pressure and temperature response for normal trips are shown in this chapter; the res ponse is also shown for a few ' selected abnormal events. These examples will show the dif ference between transients in which all systems and equipment function properly and those which have several failures. 1 j- DATE: PAGE 55 7-6-82
BWNP-20007 (6-76) 8ABCOCK & WILCOX NUMBER NUCitAR POWER GENERAtlON DIVISloN 74- t i 23331-oo TECHNICAL DOCUMENT The P-T diagram is used to identify a transient " type". There are two general " types" of transients which cause the core to steam generator heat transfer to be abnormal: overheating (inadequate heat transfer), and over-cooling (exces s ive heat trans fer) . Changes of the amount of subcooling can also occur for a number of reasons. It can be simultaneous with over-cooling or undercooling or it can be independent of them. A normal trans-ient does not have " overcooling", " overheating" or a loss of RC subcool-ing. The P-T diagram can be used to find out in general what may be wrong and can be used to narrow down the number of possible failures. Observing the P-T di ag ran is the first step for abnormal transient diagnosis; the second step is to observe a few pertinent parameters associated with the
" type" of transient to narrow down the possible failures.
The P-T diagram will be used to monitor actions taken by the operator to O see if they are producing the right ef fects. When equipment failures cannot be found or cannot be fixed the P-T diagram can be used to follow the ef fect s of operator corrections as the plant is steered toward the best possible condition. The diagram may also be used to ensure the plant has stabilized after a transient has been terminated. O DATE: PAGE 36 7-6-82
=" - - ,-+-a n.- , - - - _ _ _ _ _ _
BWNP-20007 (6-76) BABCOCK & WILCOX NUMSIR NUQtAR POWER GENERATION Divi $lON 74-1125531-00 . g TECMICAL DOCUMENT i f Description of the P-T Diagram ' Figure 8 shows t'ne P-T diagram with information pertinent to normal power operation. The features of plant power operation that this diagram shows include the saturation line which applies to both primary and secondary water and steam conditiota. Above the saturation line is the subcooled ! water region; below it is the superheated steam region. The reactor coolant information displayed also shows the RPS trip enve-lope. Two small windows show the expected nonnal reactor 100% power opera-tion po int. One point is based on That leg; and the other on T eoid leg
- The size of the window is based on an expected approximate instrument
( t error and also an allowance from the desired setting due to ICS control of l minor plant variations. Actual " normal" power operation could be anywhere within this window and be acceptable. Steam generator outlet pressure is shown as a line crossing the saturation line, and stean generator outlet temperature is also shown. The point where these . two lines cross in the superheat region is the " normal" steam ou tlet operating point at power. The amount of superheat is shown as the dif ference between the saturation temperature (where the steam pressure line meets the saturation curve) and the steam generator outlet terepe ra-I ture. The amount of superheat will change when the power level changes. i l (Note: In an actual P-T display, superheat will be shown only if steau tempe rature is measured. If steam temperature is calcul&ted from steam i prissure, the P-T diagram will always show saturation tempe rature even at j (N I \ power.) DATE: 7-6-82 PAGE 57
BWNP-20007 (6-76) 3ABCOCK & WILCOX " E NUCLEAR POWER GENEAAiloN DIV8510N 7'-i t 2 s5 31-o0 TECHNICAL DOCUMENT Figure 9 snows a P-T diagram for post-trip conditions. Most of the fe a-tures of Figure 8 are also shown on Figure 9. The impo rt ant difference between Figures 8 and 9 is a line that shows the subcooling margin from the saturation curve and the post-trip window. This subcooling margin line is to be used only to gauge the condition of the reactor coolant and not the steam gene rator fluid. Because the reactor coolant conditions a round the loop can be different and because the conditions can be di f-ferent fran one loop to the other this line must be compared to reactor coolant pressures and temperatures in the hot and cold legs of both loops. The s ubc ooling margin was chosen based on the ability to accurately mea-sure the reactor coolant temperatures and pressures (instrument errors ) during degraded reactor building environmental conditions (LOCA or SLB). It also includes an extra 5F to allow for temperature variations from the point of measureunt in the system. This will give assurance that the reactor coolant ic truly subcooled when above the subcooling margin line and that it has the ability to move the heat from the core to the generator. If the subcooling margin is lost, the assumption should be made that s ub-cooling has teen lost (ie., the RCS is at saturation). The subcooling rule that was given in Addendum A should be invoked (it is repeated here): Subcooling Rule Whenever the reactor coolant subcooling margin is lost:
- Two makeup pumps should be run at full MU capacity taking suc tion from the BWST AND
- Two HPI pumps should be run at full HPI system capacity when the reactor coolant presure is below the low RC pressure SFAS actuation setpoint (1650 psig).
DATE: 7-6-82 PAGE 58
BWNP-20007 (6-76) BABCOCK & WILCOX Numeen NUCLEAR POWER GENERAftON DivlSION
- TECHNICAL DOCUMENT 74-1125531-00 The P-T diagram can also be used to monitor and control MU and HPI and RC pump operation. When MU or HPI is initiated it can be throttled only when the subcooling margin is regained. In general, if the RC pumps have been t ripped they can be restarted anytime the subcooling margin is regained I
m and OTSG level exists (i.e., heat sink available). Exact details of HP1 and RC pump control are given in the chapter called "Best Methods for Equipment Operation". Pigure 9 also shows a " post t rip" operating window. This window has been
- drawn to show diere the reactor coolant pressure and temperature should end up after reactor and turbine trip. The size of the window has been l -
compiled from a review of several actual reactor trips (plus computer si-mulations) with and without equipment failures; its size is not exact and it is possible for a trip (with minor failures) to end slightly outside < t
- the window and still have a stable plant. Some judgement will have to be f #
l applied. However, this window gives a good "first" basis for determining if the plant is operating correctly af ter a trip. If the reactor coolant system pressure and temperature move out s ide the window af ter trip and do not return in a fairly short time (about 2 to 3 minutes) then an abnormal transient is in progress and operator corrective actions are needed. A review of other plant readouts may be required to find out the exact cause. After the corrective actions have been taken the plant will be stabilized and the stable point can be inside or outside of the window (Criteria for pla nt stability are given in the chapter entitled, " Post Accident Stability Determination".) DATE: PAGE 7-6-82 59
BWNP-20007 (6-76) BABCOCK & WILCOX N SER NUCLEAR POWER GENERATION DIVI $10N 7'-i 23331-00 TECHNICAL DOCUMENT An abnormal transient is also indicated by the steam pressute and steam saturation temperature lines. Generally if steam pressure falls belov 960 psig af ter trip, some failure has occurred and the operator should begin a diagnosis of the plant. The " steam pressure limit" on the pos t-trip win-dow is 960 psig. A steam temperature of 547F corresponds to 960 psig, therefore, if steWn terspe rature is lower than 342F after a trip an abnor-mal condition is indicated. Also a loss of reactor coolant to steam ge ne ra to r heat t rans fe r may be noted whenc T does not follow T sat in the steam generator. The "pos t trip window" shows two end points: One is for Thot during natu-ral circulation and the other is for Tcold during natural circulation and both Tc and T hduring forced circulation. When the RC pumps are off Teold will be nearly the same as steam temperature but T hot will be greater. will depend on the decay heat level. The other end The value of Thot point shows forc ed circulation. When the reactor coolant pumps are run-ning Thot and Teold will be almost the same after trip and both will be ' Nearly every trip will almost the same temperature as stemn temperature. end at either the forced or the natural circulation point if all equipment ( operates correctly and no equipment failures have happened. If some minor equipment failures have occurred (a leaky steam safety valve for example) the end point will be somewhere else inside the post-trip window. The post trip window is a good gauge for determining if sys tems are oper-ating correctly after a trip. If the reactor coolant temperature and pres-sure path stay inside this window or if the transient path goes out s ide DATE: 60 7-6-82
. -. _ . . _ _ . .._ .=__ - _ __ _ _ _
n BWNP-20007 (6-76) BABCOCK & WILCOX " NUcLtAa POWER GENEGADON DIVISION 74- 12s53i-00 TECHICAL DOCUMENT ) this window slightly but returns, then the transient is going as expected and the core cooling with stean generator heat trans fer is correct. How-ever, severe exce s s ive feedwater transients must be discovered before the ; transient path goes outside this window. This will be discussed in more de tail later. If the reactor pressure a r.d temperature are moving away ( from thia window and do not return, then an abnormal transient is in pro-gress and correct ive actions for abnormal transients should be imple-mented. These corrective actions are directed toward restoring control of reac to r-s t em generator heat trans fe r which is the pre fe rred method for core cooling. ; Successful transient mitigation can end with reactor tempe rature and pres-
-sure inside the window, but the plant can be stabilized outside the win-dow. In some cases it is desirable to achieve stability out side thie window. ,
Figure 9 also shows s tean pressure; as illustraced its value is at the 960 psig "lowe r" steam pressure limit. After trip, steam pressure will nor-
% mally be approximately 1015 psig. Steam tempe rature is als3 shown. After trip the steam temperature should decrease to the steam generator satura-tion tempe ra tur e (approximately 548F) which is set by the steam generator pressure of 1015 psig. (Note: In an actual P-T display, steam tempe ra-tures will e1 ways be shown at saturation te"perature if the steam tempe ra-t tures are calculated fror :..eam pressure rather than measured.)
O DATE: 7-6-82 61 ,
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCLEAR POWER GENERATION DIYlSION 74-ii2ssai-oo TECHNICAL DOCUMENT Steam pressure and temperature are very important parameters to review to de t e rmi ne if the plant is working correctly after trip. These two para-me te rs in combination with reactor coolant pressure and t em pe ra t ure , will show if the sec onda ry s id e is: 1) removing the right amount of heat from the reactor coolant, and 2) indicate if the reactor coolant it trans-porting the core heat to the steun generator so the steam generator can remove the heat. It is important to note that other parameters that are not displayed on the P-T diagram must also be checked to ensure proper pri-mary to secondary heat t ra ns fe r . For example, excessive main feedwater will not initially cause no t ic eable steam generator pressure or tempe r a-ture reduction. By the time excessive feedwater causes the transient path to leave the po s t-t r ip windou, the s t e ran ge nera to r will be overfilled. The re fo re , riain fe edwate r flowrates and SG levels must be checked very early following a reactor trip. Heat Transfer Characteristics Shown by the P-T Diagram This section will show examples of various transients on the P-T diagram. Both normal and abnormal transients are shown fo r c om pa r i so n . The trans-ients to be illustrated include:
- A normal reactor-turbine trip with no failures
- Transients that show the ef fects of equipment failures before trip
- Transients that show the ef fects of single and multiple equipment failures after trip.
O DATE: 7-6-82 62
__ . , _ . _ __ _ - . _ . _ _ . _ _ _ _ ~ . _ _ _ . . _ . _ . BWNP-20007 (6-76) BARCOCK & WILCOX " i NucteAn rowse oeNeRADON DIY15 TON ~ 74-1125531-00 O TECHICAL DOCUMENT These examples are used to show how reactor coolant system pressure and tempe ra tur e and stean pressure change when different f a ilures cause i changes in heat transfer. , l P-T Transient - Normal Trip Figure 10 shows the typical response of both primary and secondary plant para:<a ters following a reactor trip. Individual important parameters are shown as well as the P-T diagram. The shape of the reactor coolant P-T characteristic path frr,m power operation (above 25%) to hot zero power is always like this unless an abnormal transient is in progr es s. The " dip" of the curve is due to cooldown of the RCS to near Tsat f the steam gene-t N rators for the turbine bypass valve (TBV) setpoint. The cooldown results in coolant shrinkage which causes a pressurizer outsurge and pressure re-duction. After the RCS reaches a temperature slightly above T sat f the SG's, the reactor coolant will repressurize and stabilize due to the make-up pumps pa rt ially refilling the pressurizer and to energizing the pressurizer heaters. Depending on prior power ope rat ing conditions the low point of the " dip" will have different values, but the characteristic shape will always exist when the reactor trips from full power. When the 5
% plant trips the stean pressure will initially rise to the safety valve setpoint and then quick.ly settle out at the post trip turbine bypass valve se tpoint and stean temperature will fall from the superheated condition to saturation temperature (if steam temperature on the P-T diagram is derived from steam pressure, saturation temperature will always be shown).
a DATE: 7-6-82 PAGE 63
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCtEAR POWER GENERATION DIVISION 74-ti2553t-00 TECHNICAL DOCUMENT A similar P-T characteristic shape can also be seen for some abnormal O transients, es pe c i al ly those that are caused by secondary side overcool-ing. On the other hand, small LOCA's which depressurize the RCS " slowly" will not show the characteristic repressurization upturn (unless they are very small leaks or they are isolated). Individual parameters are shown, in Figure 10, ve rs u s time to show the approximate time for s t abilizat ion. Since stabilization takes a certain amount of time the overcooling charac-t e ri s t ic can mask failures that would not show up while the " overcooling" trend exists. Since overcooling can be caused by too much feedwater or low steam pr e s s ur e , one of the immediate post trip operator actions includes a review of the steam pressure, FW flow, and steam generator level to as sur e that the trip is " normal" and not combined with an over-cooling transient. Indications of a nonnal trip as shown by the P-T diagram include:
- 1. Hot and cold leg temperatures will stabilize in 2-3 minutes.
- 2. Reactor coolant pressure will stabilize in 5 to 6 minutes.
- 3. Tc old will be nearly equal to saturated steam temperature in-dicating that reactor coolant is transferring heat to the steam generators.
- 4. Stean pressure will stabilize in 2 to 3 minutes.
- 5. Reactor coolant subcooled margin will increase.
P-T Characteristics - Abnormal Transients - Before Trip Although many transients will go so fast that operator action be fore trip is unlikely, the changes in displayed parameters prior to trip can provide clues as to the type of transient (overheating, overcooling, etc.). When the reactor trips the trend of the accident can be covered up by the P-T DATE: 64 7-6-82
r BWNP-20007 (6-76) BABCOCK & WILCOX " NUCttAR POWts GtNERAftON OlVi$lON 74-1125531-o0 TECHNICAL DOCUMENT change caused by the cooling ef fects of the trip so the characteristics
- that occur in the short time before trip can help identify the trend.
Operator action in response to a change from the normal position in the P-T window may be possible, and trip may be avoided, but usually trans-i ients will happen too fast for the operator actions to be successful. Nevertheless, some of the indications before trip will help to de termine what may be occurring. Figures 11, 12, 13,and 14 show pre-trip movements on the P-T diagram. Steam pressure and RC temperature and pressure will respond differently depending on the cause. The events represented by these curves are: O Figure 11 - Overheating Transient Figure 12 - Overcooling Transient Figure 13 - Overpressure Transient Figure 14 - Depressurization Transient P-T Characteristics - Abnormal Transients - Af ter Trip
\ Figures 15, 16, 17, 18, and 19 show examples of transients which may occur because of failures either on the primary or secondary side. These exam-ples show transients which end as expected and also go past the expected point because of additional f a ilure s . Those transients which are cor-rected properly will follow the expected course and will end up in the "pos t t r ip wir.dow" near the normal post trip end point. When the path s
DATE: PAGE 65 7-6-82
.~_ -. _ _.__ -_ _ ..__-.._. _ _ _ ,..__._- _ _ ._ ._ _ _ ._ ,.-__,-.. _ _ _
\
BWNP-20007 (6-76) BABCOCK & WILCOX wuma NUCLEAR POWER GENERATION DIVl510N TECHNICAL DOCUMENT 74-i i 2 s s3 i-oo goes outside the wind ow , the transient is defined as abnormal and the direction reactor coolant pressure and temperature move toward can be clas-sified as overheating or overcooling. In combination with overheating or ove rcooling the reactor coolant temperature and pressure path can also move toward more or less subcooling. These t re nd s , ove rh e a t ing , overcooling, and loss of subcooling, are the O first indications to check in transient diagnosis and correction. In the case of overcooling, which can be masked by the normal pos t-trip response, excessive main feedwater must be checked very early in the transient using other pa rame t e rs such as MFW flow and SC levels, which are not shown on the P-T diagram. An abno nnal t rans ient will show dif ferent characteristics depending on the O fa ilures that may have occurred. Some characteristics of RC pressure and tempe ra tur e and o f s t eam pressure that show undesired heat trans fer on the P-T diagram are:
- 1. Reactor cool, ant subcooled margin is lost
- The trend may be caused by overheating or overcoolit.g.
- The trend may be caused by loss of reactor coolant.
1
- Subcooling will be lost for all except the very ,
smallest breaks. O DATE: 7-6-82 PAGE 66
BWNP-20007 (6-76) BA8 COCK & WILCOX NUCLEAR POWtt GENGEAllON Olvl510N 7'-i 2333i-00 A TECHNICAL DOCUMENT
- 2. Steam pressure is much lower than normal A value of 960 psig has been established as a low limit similar to the
" post t r ip" window for the Reactor Coolant P-T. If steam pressure drops below this limit after trip, then an abnormal condition may exist. A corresponding value of 542F has also been chosen for satu-rated steam temperature low limit.
- Stean pres sur e may be low because of a f ailure in the steam lines.
Overcooling will result. Subcooling may or may not be lost.
- Stean pressure may be low because of a loss of all feedwater. Over-heating will result. Subcooling will eventually be lost.
g - Stean pressure may be low because a large amount of reactor cool-ant has been lost and cannot pass core heat to the feedwater in the generator to create steam. Large LOCA's can cause this or an Inade-quate Core Cooling (ICC) situation can cause this. Both LOCA and ICC are discussed in detail as separate topics later.
- Steam pressure may be low due to excessive AFW. Overcooling will re-sult and subcooling may be lost.
s 3. Steam generator saturation temperature and Tcold do not correspond (not coupled) (Lack of primary to secondary heat transfer)
- When Teoid does not change when Ts at-SG changes, then heat trans fe r
! from the reactor coolant to the steam generator is interrupted. Natural circulation has probably stopped when this occurs and the l reactor coolant may heat up. The reactor coolant condition can be t DATE: PAGE g l 7-6-82 l
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWIR GENERATION DIVl$lON 74-112 5531-w TECHNICAL. DOCUMENT .. suocooled or saturated. If the reactor coolant is superheated, natural circulation has been lost. The transient s used as examples are: Figure 15 - Loss of Main Feedwater
- 15a) Shows Loss of Main Feedwater with AFW actuated. The important feature of this transient is that the main feedwat er heat sink is quickly replaced with an AFW heat sink; the trend look s similar to a normal reactor trip.
- 15b) Shows Loss of Main feedwater with AFW delayed. Important fea-tures of 15b) are: 1) loss of steam pressure, and 2) the reactor coolant heats up and would eventually saturate at 2500 psi. This is an indication of lack of primary to secondary
! heat t rans fer. l Figure 16 - Small Steam Line Leak 1 i Shows break (TBV fails open) caused by a failed open TBV that is l l terminated by isolating main steam and fe edwa t e r . The important l feature is that the reactor coolant was overcooled be fo re l isolation. j Figure 17 - Excessive Feedwater (to one SG) l
- This transient is shown to be corrected by ICS operation and j
look s similar to a normal trip. Were the transient to con-tinue, water could enter the steam lines and cause damage but DATE: PAGE 7-M g
BWNF-20007 (6-76) BABCOCK & WILCOX " NucttAt POWER GENERATION Olvi$lON 74- t i 2 s s u-00 (~') TECHNICAL DOCUMENT V the amount of damage and its effects are not known. The RCS would overcool to sattrated conditions (i.e., drain the pressurizer) by the time water entered the steam lines. Figure 18 - Small Break LOCA in the Pressurizer Steam Space ( ) - The impo r t ant feature of this transient is that water will b/ flow into the pressurizer from the reactor coolant loops. Although the pressarizer will show a level it is not a good indication of reactor coolant inventory when the reactor coolant is saturated.
- 18a) Shows a LOCA with the break isolated after the accident starts. Refill and repressurization of the reactor coolant I \
\ / system will allow a normal cooldown with a pressurizer bubble.
- 18b) Shows a LOCA that is not isolated. Subcooling does not return as quickly although the entire reactor coolant system fills with watet. Cooldown after this accident will be with a pressurizer full of water.
N Figure 19 - Small Break LOCA in the RCS Loop Water Space b' - This transient is different from Figure 18 because the pressurizer does not fill with water from the loops as a result of the break. 3 P \ / v DATE: 69 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Huusen NUCttAR POWit GENERATION DIVISION 7'-i t ssai-00 TECHNICAL DOCUMENT
- i9a) Shows a small break with AFW used to remove heat.
- 19b) Shows the same break with AFW delayed. The ef fect of the heat transfer to the steam generators can be seen by comparing the RC pressures with and without AFW. With no AFW the RC system pressurizes to 2500 psi. At this pressure the leak rate is h ighes t; use of the steam ge nera to r helps to reduce RC pres-sure, thereby, reducing the leak flow and allowing HPI flow fo r covering the core. 19b also shows that steam pressure is lost because no steam generator inventory exists to create stean.
O O O DATE: 7-6-82
u 1 i Figure 8 POWEP OPERATION P-T DIAGRAM A 2600 RPS TRIP ENVELOPE l HIGH 2400 - l TEMP HIGH PRESSURE TRIP l TRIP 2200 - l T T COLD H0T 2000 - 1 2 y LOW PRESSURE TRIP / ' l VARIABLE LOW PRESSURE TRIP 1800 - gg a gW
!! 1600 -
S 5 SUPERHEAT j 1400 - SUBC00 LED g REGION
$ 1200 -
O ! SUPERHEAT o ." l REGION
- 1000 - E I t SG PRESSURE 800 -
600 - STEAM GENERATOR NORMAL OPERATING POINT e sc *s@ OPERATING POINT POWER OPERATION (THOT) 400 - 1 I i i I 400 450 500 550 600 650 700 Reactor Coolant and Steam Outlet Temperature F s l I 1 l l 7.!-i125533 -00 ; 4 i
Figure 9 POST TRIP P-T DIAGRAM 2600 POST TRIP 2400 - WINDOW 2200 - B F---'Q y la L _ _. _ l ta - 2000 - SUPERHEAT g SU8000LE0 r
- REGION 3 REGION y 1800 -
a-C
- 1600 -
5 . s
" O 1400 -
J' S - END POINT POST TRIP WITH-1200 - f HUED CIRCUL Ail 0N : T H0T STEAN PRESSURE g ,, , &TCOL O) AND FOR NATURAL 1000 - CIRCULAil0N fTCOLO) h NORMAL OPERATING POINT PO S 800 - hOPERAil0NiTH0T) E
- SATURAil0N r- 7
[ND POINT POST TRIP WiiH 600 -
'_jNATURALCIRCULAil0 t nit H0T)
- SUBC00 LING 400 -
MARGIN LINE O L f f I 8 400 450 500 550 600 650 700 Reactor Coolant ana. Steam Outlet Temperature, F 74-1I25531-00 ' J
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l Figure 11 OVERHEATING TRAi4SIENT (PRE-TRIP) e 2600 l RPS TRIP ENVELOPE I HIGH 2400 - l TEMP on HIGH PRESSURE TRIP TRIP l E /
' ~ I i = 1 COLO J HOT j 1 U 2000 - '
g LOW PRESSURE 1 RIP / l VARIABLE LOW PRESSURE TRIP 1800 - lg a
- 2 1600 -
lS SUPERHEAT g 1400 - SUBC00 LED g _ REGION b 1200 - ; SUPERHEAT j ,T l REGION O 1000 - E ! SG PRESSURE O 800 - 600 - s@ STEAM GENERATOR NORMAL OPERATING POINT OPERATING POINT Q4 POWER OPERATION (TH0T) 400 - 1 I I I I 400 450 500 550 600 650 700 Reactor Coolant ana Steam Outlet Temperature. F Plot shows an increase in both RC pressure and Yhot. A slight increase in superheat and steam pressure is also possible. Possible Causes Possible Alarms e Decrease in or loss of main feedwater e High - RC pressure e ICS malfunction causing steam pressure e High - Pressurizer level increase (turbine valves closing) e Low - MFW pump flow e Low - SG level e High main steam temperature e High main steam pressure .. 74-1125531-00 _ ,
Figure 12 OVERC00 LING TRANSIENT (PRE-TRIP) 2600 1 RPS TRIP ENVELOPE HIGH 2400 - l TEMP on HIGH PRESSURE TRIP l TRIP
"'" - 1 8
= 1 COL 0_AH0T 1 i / " 2000 - , LOW PRESSURE TRIP / l VARIABLE LOW PRESSURE TRIP 1800 - lg a
- g 1600 - M SUPERHEAT g 1400 -
SUBC00 LED g REGION 1200 - ; ! SUPERHEAT g " REGION _ l " a
" l 1000 -
SG PRESSURE O B00 - 1, 600 - g STEAM GENERATOR NORMAL OPERATING POINT OPERATING POINT 400 - 9s8 POWER OPERATION (TH0T) e i i I I 550 600 650 700 400 450 500 Reactor Co:lant ana Steam Outlet Temperature. F Plot shows a decrease in both RC pressure and T A drop in superheat and SG pressure is also possible depending on tggt.event. Possible Causes Possible Alarms e Excessive Feedwater e Low RC Pressure e Decrease in Feedwater e Low Pressurizer Level Temperature o High Makeup Flow e Decrease in Steam Generator Pressure (steam leaks) e High Steam Generator Level e Low Main Steam Pressure o 7 tt - 1 1 2 5 5 3 1 - 0 0 l
Figure 13 OVERPRESSURE TRANSlENT (PRE-TRIP)' 2600 I RPS TRIP ENVELOPE I HIGH 2400 - on HIGH PRESSURE TRIP l TRIP E I
. 2200 -
= COLD HOT g 2000 -
VARIABLE LOW I* LOW PRE 3SURE TRIP PRESSURE TRIP 1800 - lg; E
=
2 1600 - LO
" ga SUPERHEAT g 1400 -
SUBC00 LEO g REGION
! SUPE 2 HEAT 1200 -
O a g - l REGION u 6
" l 3 1000 -
g SG PRESSURE O 800 - 600 STEAM GENERATOR
- s@ NORMAL OPERATING POINT sd OPERATING POINT POWER OPERATION (TH0T) 400 -
i t 1 I i 550 600 650 700 400 450 500 Reactor Coolant anu Steam Gutlet Temperature. F Plot shows an increase in RC pressure with little or no change in Thot* Possible Causes Possible Alarms l l e Too much makeup e High reactor coolant oressure e Insufficient letdown e High pressurizer level l e Pressurizer heater misoperation e High makeup flow l l l 74-1125531-00 l
Figure 14 DEPRESSURTZATI0ii TRANSIENT (PRF-TRIP)
=
2600 RPS TRIP ENVELOPE k gigg 2400 - l TEMP ma HIGH PRESSURE TRIP l TRIP I l 2200 M T g TCOLO H0T t C 2000 - E
- LOW PRESSURE TRIP / l VARIABLE LOW PRESSURE TRIP 1800 -
g 3 a g:e 2 1600 - , g SUPERHEAT G j 1400 - SUBC00 LEO REGION
= -.
" g ! SUPERHEAT 1200 -
J' l REGION E I a 1000 -
;; SG PRF.SSURE O
" 800 -
600 STEAM GENERATOR NORMAL OPERATING POINT s@ g@ OPERATING POINT POWER OPERATION (TH0T) 400 - l
' I I I 1 600 650 700 400 450 500 550 i
Reactor Coolant and Steam Outlet Temperature, F Plot shows a decrease in RC pressure with little or no change in Thot* Possible Causes Possible Alarms
.sss of makeup; too much letdown e Low RC pressure o Pressurizer spray misoperation e Low pressurizer level e Small LOCA e Makeup pump trip e Low seal injection flow e High or low makeup tank level ' ' ~
l l': :1!.:i i - o o i
I 1 e Fioure 15a. LOSSOFMAINFEED'.:A)h AF2 STARTED WITHI Reference Time Points (Seconds] RemarC 1 0 Accident starts. Main Fee 1-2 0-40 RCS P&T decrease due to log Typical post trip response with a corresponding decreg level. 2 40 AFW delivered to both steas level controlled at low let 2-3 40-225 Systen stabilizes with dect the steam generators. RC 9 the saturation temperature sure, and pressurizer level' makeup. 3-4 225-600 Presst.rizer level restored RC pressure restored by thG 4 600 STABLE PLANT C0?.DIT10NS. AP R"URE CARD l Figure 15b. t LOSS OF MAIN AND EMER Reference Time Points (Seconds) Pemark 1-2 0-40 Sane as Figure 15a. 2 40 AFW fails; no feedwa'ar is< steam generator. 2-3 40-50 System continues to exhibit trend; feedwater inventory l 3 104 Steam generators are dry. O 3-4 104-283 RCS reheats and repressuria to secondary heat transfero increases. l l 4 285 PORV lifts and cycles to c@ 4-5 283-880 PORY controls RCS pressureo to bestup; the primary sysG conditions. 5 880 Pressurizer fills with wat stay at the FORY or safety saturated primary conditio2 is left in an abnormal coni be established to maintain boils out of the vessel. 74-1125531-00 J
ho Avdl-bla On j
' APerture Card Figura 15 LOSS OF MAIN FEEDWATER {
2600 2400 - 81N008
, 2200 -
r- - ;
- R WITH d g 2000 -
SU8C00LE0
" SUPERHEAT 3 REGION 2 REGION
$ 1800 a.
4ttr pumps trip. , of fission powcr. overcooling trend) E
- O in pressurtzer 1400 -
" E END PolNT POST TRIP stTH E FORCED CIRCUL ATION (TH0T gentrators, with 1200 - STEAM PRESSURE m &T COLD) AND FOR NATURAL Lluli 'I (no SFAS). e m CIRCULATIGN (TCOLO) ! heat remova through a 1000 -
r-l'mperature approaches or the secondary us- [. -- NORMAL OPERATING POINT-POWER OPERATION (TgOT) increases becau' )f g 800 - E SATURATION r-- I END POINT POST TRIP WITH md steady w. . normal 600 - b_]NATURALCIRCULATION(THOT) pressurtzer heaters. SU8C00 LEO 400 - MARGIN LINE I I I I I O 550 600 650 100 400 450 500 Reactor Coolant and Steam Outlet Temperature. F 2600 POST TRIP 4- 5 2400 - 2 elND0s iENCY FEEDWATER \ 2200 -
--3 t : "' "
. 2000 -
SUSC00tED " '" 3 REGION 2 kilvered to either 3 1000 - n. Post trip overcooling E 1600 - s betog boiled off. j e a 1400 - [
's dua to loss of primary END P0lNT POST TRIP flTH Pressurtzer level also j 1200 -
STEAu PRESSURE F0*cED CIRCUL ATION a TH0T
; &TCOLD) AND FOR NATURAL giggy) itrol RCS pressure. 2 - CIRCULATION (TN) g 1000 , , _ ',,,_,1_,,,__,,_,,,,,__
RC temperature continues - NORM AL OPER ATING POINT-POWER m is approaching saturated _ 0PERATION (TH0T) I 8 IN END POINT POST TRIP BITH
. RC pressure would -
'alve setpoint and 600 L_J NATUR AL CIRCUL Ail 0N f THOT)
Iwould develop. Plant tion; AFW and/or HP! must - Sil8C00 LED ore cooling tefore water 400 utRgin ting t i e i n 0 52 600 650 100 400 42 500 Reactor Cor int ana Steam Outlet Temperature. F
'I 8403050114 - 0 3 .I o
Figure 16 SMALL STEAM LINE LEAK IN ONE SG (TBV FAILS OPEN) 2600 POST TRIP
,,400 slN00s x .
2200 - C '- -- - ' ; . I ' ' - - ~ k TEllP IN 4 3 2 SUPERHEAT g 2000 - AFFECTED SUBC00 LEO - REGION sg 3 REGION y 1800 - n. 1600 - a 3 e 3 1400 - [ SG WITHOUT LEAK g . END POINT POST TRIP WITH ] E 1200 - STEAM PRES $URE FOMCED CIRCUL ATION I TH0T i a ", - i p/ l M " COLO 1000 - i3 g C 800 - b< >3 1 i NORMAL OPERATING POINT P0aER OPERATION tTH0T) " [4 M " SATURATION r- ~ 9 END POINT-POST TRIP WITH 600 - / sg'EITH LEAK {__j NATURAL CIRCUL ATION r THOT)
- SUBC00 LED 400 MARGIN LINE i i i e i O
500 550 600 650 100 400 450 Reactor Coolant ano Steam Outlet Temperature, F Reference Time Points (Seconds) Remarks 1 0 TBV fails open. 1-2 0-13.5 Increase of steam flow causes slight reduction in Tave, ICS attempts to keep Tave up by pulling rods. 2 13.5 Reactor trip. 2-3 13.5-100 RCS P&T drop due to loss of fission power and excessive primary to secondary heat transfer. 3 100 SFAS actuation; HPI initiated. 3-4 100-300 Pressurizer level indication goes off scale low. 4 300 SFRCS actuates on low SG pressure. 4-5 > 300 RC pressure and temperature start increasing, pressurizer level regained, and operator con-trols steam pressure in steam ger.erators. I 74-1125531-00
Figure 17 EXCESSIVE FEEDWATER j 2600 POST TRIP 2400
- WINDOW 2200 - 1 4 F--i u___;
j y 1 2 SUPERHEAT
. 2000 -
SUBCOOLED REGION
; REGION a.
1;30 - k# 3 em 5 1600 - a 5 1400 - $ E 1200 - STE AN PRESSURE
@ENDPOINTPOST1RIPelTH FOMCEO CIRCUL ATION i THOT t'4 gTCOLD) AND FOR NATUR AL 5 q Lluli l [ CIRCUL ATION f Igggg) i u
iOOD . _ u _ _ .i _. -- ._ _ NORMAL OPER ATING POINT.P0eER
; y OPERATION iTygy)
- 800 -
[- SATURAil0N {-] END POINT. POST TRIP WITH
'; NA1 URAL CIRCUL ATION e TH0TI 600
- SU3C00 LEO 400 -
MARGIN LINE 450 00 400 Reactor Coolant ano Steam Outlet Temocrature, F Reference Time Remarks Points (Seconds)_ 1 0 With the plant operating at 100% power, a failure of the MFW pump controller al2ows pump overspeed. Excessive feedwater addition begins. 1-2 0-60 Slight overcooling of RCS occurs due to excessive festwater addition. ICS pulls rods to compensate for reduction of T , but rod withdrawal is limited by the high flux $ Niter to 103%. 2 60 Manual reactor trip. 60-200 RC P&T decreases due to loss of fission power and 2-3 higher than normal secondary inventory. The ICS initiates a feedwater runback and the MFW addition s tops . Pressurizer level decreases because of reactor coolant contraction. 3 200 Minimum pressurizer level reached. 34 >200 Normal system pressure restored *by operation of MU system and pressurizer heaters. Primary system is lef t in .: stable, hot shutdown condition. 74-1125531-00 '
l [ Figure 18_a. STUCK OPEN PORY W1T TO CLOSE BLOCK YA Reference Time Points QeconM @ 1 0 PORV assumed to open. 1-2 0-60 Pressure drops due to out of PORV. Little a occurs. An insurge of suriz-.r occurs and an will be observed. 2 60 Reactor trip on low RC 2-J 60-125 RCS P&T decrease due i primary to lecondary t drops. MU can't keep pressure drops. 3 125 Subcooling margin los1 i starts HPI. AFW actug , controlled to 40" on f actuates. 4 185 Hot leg saturates and occurs. Operator acti PORY block valve. LO( 4-5 185-600 HP! in combination wit i on the generator tube ( voids within primary f subcooled state and r< increases. ' Subcooling margin esti 5 600 and restarts RCP's. 5-6 600-700 Operator stops HPl ani 6 700 STABLE PLANT CONDITIO' AE m ci l Figure 18b. STUCr OPEN PORY ] j Reference Tine 1 Points deconds) 1-4 0-185 Same as Figure 18a ex leak continues. I I 4-5 185-400 RCS is in two-phase P&T conditions decre j stabilize at about 1 increasing as pressuf 5 400 Pressurizer level i changes from steam t 5-6 400-1000 Primary system remai 1200 psia with tha core boil-off cnd st collapsed. 6 >1000 Operatcr initiates p1 to place plant in a g break a return to a f during cooldown. Sol quired thereafter unl 711-1125531-00 1
Figure 18 SMALL LOCA IN PRESSURIZER STEAM SPACE jOPERATORACTION WE AT s 3 MIN. POST TRIP 2400 - 8tN00s
' 2200 -
t-_-._ E ischarge of pressurizer steam g 2000 - SUBC00 LEO 2
=
no change of RC temperature REGION SUPERHEAT rsactor coolant into pres- " REGION nerease in pressurizer level 1800 - k +
\
3 pressure. ; 1600 - 5 loss of fissicn power with g *- { 4 t transfer. Pressurizer level 3 1400
=
u with tha leak, and RC " END POINT POST TRIP WITH 1200 - h FORCED CIRCULATION g (T
; STEAN PRESSURE j optrator trips RC pumps and g id &TCOLD) AND FOR hATUR L <ed with leval sutomatically a t(Nit m g 1000 O \ / CIRCULATION (TCOLO) brtup range, 93* when $fAS u - - - L __
NORMAL OPERATING POINT-insurge irto the pressurizer U a 800 - POWER OPERATION (THOT)
- is assum d to isolate the E r7 END POINT POST TRIP WITH is isolated.
- SATURAil0N I
- d NATURAL CIRCULAil0N H0T)
(T 800 cond;nsation of the RC steam Izads to collapse of steam . SU8C00tED . stem. System returns to a 400 - ressurtres as pressurizer level NARGIN tlNE i f I I e O tilshed; operator throttles HPI 550 600 650 700 400 450 500 Reactor Coolant and Steam Outlet Temperature. F restarts normal MU and letdown. 4, 2600 POST TRIP 2400 -
,,,00, 4 l hum 2200 -
_ht
$UPERNEAT g 2000 -
'2 SUIC00 LED REGION 3 REGION (NO ISOLATIO*t) f 1000 -
3 5 1500 - emarks 1400 - 4 pt PORY is not isolated and I
" END P0lhi POST TRIP elin FORCED CIRCUL Afl0N i TMOT ' 1200 -
STEla PRES $uRE I along saturation line and ; ' tTCOL O) AND FOR NATUR AL h'turalcirculationcondition.
' psia. Prsssurizer level is 2;r sttam space depletes.
=
g 1000 - q g g,9 ] 3 - _ _ __ _ _ _ _ [ CIRCULAfl0N rTCOLD' o cat 2s full scale; leak flow ; O NORual DPER ATING POINT P0eER a stsam-water mixture.
~
000 - Q OPERail0N rTwof) E SATUR AT10m r-i
' stable at approximately ' ' END P0lki POST TRIF o b itg saturated. HPI exceeds 500 ~ LJ NATURAL CIRCULAT10m rf,gg3 voids are slowly being ,
- SUICCOLED 400 s4# GIN tlNE gnt cooldown and depressurization
,Jfa condition. For this size hbcooled state would be expected 0 i e a i i 5 50 600 6 50 100 d 400 4 50 500 {ss-wattr cooldown would be re-
/ PORY is isolated. Re.; tor Coolant ana Steam Outlet Temperaure, F Also Avallable Om APerture Card 8403050114 - Y - s.
f 2 Figure 19a. 0.01 FT BREAK AT PUMP DISC 0Q TO BOTH STEA'i GENERATORS (U Reference Time points Eeconds] Rs 1 0 LOCA occurs; break is e at discharge of R,, pump 1-2 0-50 Pressure drops due to break; pressurizer leve 2 50 Reactor trip on low RC-2-3 50-90 RCS P&T drops due to It mary to secondary heat overcooling trend resui up with leak, pressuri scale low and pressure 3 90 Subcooling margin lost; starts HPI. 3-4 90-120 AFW actuated and is au - SFAS actuates. 4 120 Pressurizer drains and-4-5 120-600 RCS in two-phase natura crease along saturatica mately 1225 psi. 5 600 Two-phase natural circu' top of hot leg prevents generator so steam genG 5-6 600-1500 RCS repressurizes becar to reheat the reactor tion curve. Steam bub in size; condensation i possible. Pressurizer 6 1500 Boiler condenser cooli-level is low enough to steam generator tubes. 6-7 >1500 Pressurizer level decr< uration curve and will - psi. 0;>erator should i pressurization to recos A3ERTURE; CAD I 2 Figure 19b. 0.01 FT BREAK AT PUMP DI FOR 20 MINUTES (LOCA I Referen,e Time Point', 15econdsl 8 1-4 0-120 Same as Figure 19a. es water has occurred. 4-5 120-1200 Steam generators boil tion line if core coo ( C711ng. Pressurizer f will slowly drop once 5 1200 Operator restores AFW started to both steam is restored. 5-6 >1200 With ATW on, boiler c RCS P&T drops along s at approximately 1200 zero indication becau Operator 5 ould initi zation to recover pla 74-1125531-00
K)E WITH Also Avrikhic On w TN RCS t'ATERATJ OPERATim SPACE) Aperture Caril Figure 19 SMALL LOCA IN RCS WATER SPACE 1rks 2600 Jivaltnt to 1.35 in 00 hole POST TRIP 24h - RIN008 s 1 rase of reactor coolant out dicreases. 2200 - Psssure.
=
{
]
E F---]J L-- J i
~ SWH E A T 3 of fission power with pri- 4 SUBC00 LED M G'0* ?ansftr; gin:ral post trip 5 REGION 2 ->
- 3. Bxause MU can't keep E I ION ~
P IIvel indication goes off - Eps. [ 3 Eperator trips RCP's and 1600 -
, 6 e $tically raised to 93" when 1400 -
g 4 5 lt b.g saturates. 1200 - STE Au PRESSURE h"FORCE END P0lNT 0 GRCUL POST T ATION iTHOT circulation mode. P&T de- 5 Ll'lIN g COL O) AND FOR NATURAL Iurva and stabilize at approxi-3 \ / CIRCUL ATION (TCOL D) 3 1000 '- -{}- - 1 - - --- - _
; NORM AL DPER4flNG POINT.P0sER Dtiin stops; steam bubble in O 800 - S OPERAll0N tig) liquid carry over to steam a Dtor canr,ot remove heat. E ' SAluRATION END POINT-POST TRIP flTN 600 -
t_; N ATURat CIRCUL Ail 0N ( THOT) gallheatfromcoreisgoing pl ant. RCS stays on the satura- y - SU8C00 LEO 9 in hot leg is slowly increasirg RCS steam on tubes is not yet 4N ~ NARGIN tlNE
;; vel is int wsing, 0 i i i sstablished; RCS hot leg water 400 450 500 5 50 600 650 100 Diow RCS steam to condense on Reactor Cooiant ana Steam Outlet Temberature. F Ms. RCS P&T de:rease along sat- )cbiliza at approximately 1200 itiata a plant cooldown and de-
? plant. 2600 POST TRIP
,,g ,
5 tlN00 22H - p___
- L___;
2HO - SUPERMEAT 4 SUSC00 LED _
~
2- , 3 REGICN { _
~
MRGE '.llTH AFW DELAYED lRCS WAT"R SPACE) ;
- 1600 -
O a ._" 3 5 f 1400 - 4 END PolNT POST TRIP tlin Sept that a total loss of feed-E 6 F0fitE0 CIRCutAfl0N afM07 1200 - STE Aa PRESSURE y; P&T increase along satura- E AICOLO) AND FOR NATURAL g cannot be satisfied by MU/HP! I
- Ligig] gigCUL Afl0h t iCOLD zl is incrsesing. Steam pressure 1000 -
nytntory is boiled of f.
" ~
.~ II~~3~~~-~~~~~~- n 40Ru AL OPER AflNG POINT PCeER
; S stem operation. AFW flow is O 800 - @ CPERATION ifM0f8 erators and steam pressure
* 'f r'
" ' 5810RAflom END POINT. POST TRIP st TH
~
5
- NATUR AL CIRCm Afl0N iinof' t The
$ensercoolingisstarted. ration curve and will stabilize ' /- $UBC00 LEO kl. Pressurizer level drops to 400 - BARGIN LINE of coolant contraction. a plant cooldown and depressuri-O i i i i i 500 550 600 650 700 40 0 4 50 Reactor Coolant and Steam Outlet feeueratuie. F 840 3 0 5 0114 _or ~~
BWNP-20007 (6-76) BABCOCK & WILCOX NUA48 ER NUCitAR POWER GENERATION DIVISION 74-1125531-00
)TECHICAL DOCUMElli CHAPTER C ABNORMAL TRANSIENT DIAGNOSIS AND MITIGATION Introduction This chapter shows how an abnormal transient can be diagnosed and mitigated using the information provided by the P-T diagrara and the concepts on heat transfer discussed in the previous chapters. A simplified flow chart of the
\
approach to be used to diagnose and mitigate an abnormal transient is provided in Figure 20A " General Plant Accident Mi t ig a t ion" . It is broken down into a few separate steps although these steps will " blend" together into one continuous process in actual practice. The Abnormal Transient Operating Guidelines are implemented whenever an automatic reactor trip occurs (although the operator may have manually tripped the reactor after recognizing something was wrong) or a forced shutdown is necessary. The guidelines are provided in Part I. They list the appropriate ope ra to r actions necessary to mitigate an abnormal ' transient. They follow the approach outlined in Figure 20A. The guidelines incorporate the following features.
/ 1. Use of the P-T diagram which provides a constant feedback to the opera-
%/
tor on his success or failure af ter taking each step in Part I. This dia-gram should be checked frequently to make sure thing s are progressing as expected. It will thus give the operator early indications of subsequent failures that are delayed after the initial event, or multiple failures that were covered by the predominant event and d idn ' t appear until that k- - one was corrected. DATE: 7-6-82 PAGE 71
BWNP-20007 (6-76) BABCOCK & WILCOX Nuussa NUCLEAR POWER GENERATON OlVi$lON 7'- 2ss31-oo TECHNICAL DOCUMENT
- 2. The guidelines are constructed such that the operator makes an attempt to correct the problem wi th a given piece of equipment or system (e.g., AFW to correct loss of main feedwater). If that fails he is instructed to go on to the next available system. This failure of the AFW system is not given priority attention in the body of Part I, protection of the core is.
- 3. The ope ra to r is given frequent present plant status "(STATUS)" aid s N
throughout the procedure to help him maintain proper orientation. s
- 4. If new symptoms appe a r he is instructed to recycle (go to the section that treats that symptom) to the appropriate part of the procedure.
Immediate Actions The first block in Figure 20A is the "Immediate Actions" block. The immedi-ate actions should be completed in the first 2-3 minutes. The first action to be made is to detennine if a reactor trip has occurred or plant conditions a forced shutdown exist. If a reactor trip has occurred the opera-requiring tor should manually trip the reactor and turbine, then proceed to the next post trip step of the ATOG procedure which is " Vital Systems Status Verifica-tions". However, if plant conditions warrant a forced shutdown, the operator should initiate the appropriate shutdown procedure. If a reactor trip should oc cui- during the fo rced shutdown operations, the pos t-t r ip ATOG procedures should be implemented. O DATE: 72 7-6-82
BWNP-20007 (6-76) SABCOCK & WILCOX NUMBER NUCLEAR POWfs GENEGATION Ofvl510N 74-1125531-00
,Q TECHNICAL DOCUMENT Vital Systems Status Verification The next major block on Figure 20A is the Vital Systems Status Verifications.
This section requires reviewing specific plant status items including the P-T diagram to determine if they are behaving as they should for a r.ormal reactor trip. If the specific plant status items cannot be verified as performing as expected, the operator should perform the speci fied remedial actions. The procedures provide specific remedial actions for each plant status item which cannot be ve ri fied . The plant status items which are checked first are the normal automatic post trip functions which control core reactivity, primary and secondary inventory, and primary and secondary pressure. Next the operator must verify the operability of certain power supplies to (j assure the importan': plant parameters can be monitored and the power is avail-able to the important control devices. These include pumps, valves, etc., which are needed to safely control the plant and to mitigate abnormal trans-ients. The plant cannot be controlled or abnormal transients mitigated if the control devices do not work. Nor can- the plant be safely controlled if the ins trumentat ion parameters to be controlled are not available for diag-
} nosing the plant status.
Next the operator must check to see if any safe ty system has been initiated and if so that they are operating properly. For example, when the operator notes the presence of an SFRCS alarm he will be directed to verify the actuation of SFRCS controlled equipment in accordance to a Table in Part I. t v) DATE: 73 7-6-82
BWP-20007 (6-76) BABCOCK & WILCOX N ER NUCLEAR POWER GENERADON DIVl510N TECHNICAL DOCUMENT 74-1125531-00 The table is divided into two major sections, one for Channel 1 and one for Channel 2. Each of the sections is further divided into two subsections for l full and half SFRCS trips. To use the table, the operator must be aware of what trip or combination of trips exists. A half trip condition for the respective channel would direct the operator to check the equipment listed in the half trip box. Full trips are handled in a similar manner where the box for single full trip is select-ed and equipment is verified. However, if a combination of full trips exist, then the operator must select the furthest left column for which a trip condi-tion is present. This is due to the " priority" of the Low Steam Line Pres-sure trip over all others and the limited equipment actuated on Loss of Four Reactor Coolant Pumps. For example on Actuation Channel 1, if both a Steam Line On Low Pressure and OTSG level trip occurs, the SFRCS activated equip-ment will align for the Steam Line One Low Pressure trip. This will be veri-fied by the operator who will realize that Steam Line On Low Pressure repre-sents the furthest lef t column and verify the equipment with that list. I 1 l l l Next, the operator needs to check the P-T diagram for a loss of subcooling l margin, " overheating" or " overcooling" conditions. Table 2 summarizes ge ne r-ally the plant status items to be checked and the actions to be taken if the l items are not as they should be. The P-T diagram is the foundation for transient diagnosis and for the actions to correct abnormal transients. DATE: 7-6-82 PAGE 74 l t
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER oENERATION DIYl$ ION n-i 25531-00
. TECHNICAL DOCUMENT When the P-T diagram is checked, the reactor P-T should stay within the "po s t-t r ip" window and steam pressure should stay above the steam pressure limit of 960 psig. If the plant responds so that these limits are not ex-ceeded then the transient is going "as expected". If the plant does not "go
, as expected" the P-T characteristic should be checked to find out the " type" t
j of abnormal transient so that proper corrective actions can be made to re-store the reactor-steam generator heat trans fer. Finally, the operator must check to see if there are any steam generator tube ruptures. If the Vital Systems Status Verification shows everything is alright, then the plant is in a stable subcooled condition with proper primary to secondary heat transfer and no primary boundary failures. Further action will be at
. the discretion of the station management.
j If one or more items are not alright, the ATOG procedure will direct the oper-ator to make a remedial action. The remedial action may be a finite action such as closing a valve or a " Followup action". l Followup Action As shown on Figure 20A when the P-T diagram indicates a " loss of subcooling margin", " overcooling", or " overheating" exists the operator must determine which one of the three conditions exist, and then start the appropriate fo l-lowup actions procedure. These followup actions are first directed at l l re-e s tablishing the correct amount of steam generator cooling. If it cannot m l DATE: PAGE 7-6-82 75
1 l BWNP-20007 (6-76) l BABCOCK & WILCOX NUMBER NUCLEAR POWER GENERATION OlVi$ ION 74- 2n31-oo , TECHNICAL DOCUMENT l l I be re-established, then backup cooling methods are to be implemented. (The f backup cooling methods are discussed in a separate chapter of Part II). l l l Once steam generator heat transfer or backup cooling has been established, 1 the plant should be brough t to a stable condit ion fo r plant cooldown. The 1 l condition may be hot or cold depending on the circumstances; a return to the j 1 1 l
"pos t-trip" window is not required for plant stability. I 1
)
l Cooldown Procedures l Once a s t able plant condi t ion is re ach ed , the plant is cooled down using one ; of several cooldown procedures depending on the existing stable plant condi- ) 1 tions such as a solid water system or saturated RCS. The ATOG procedures i state which cooldown procedure to use. This is discussed in the Post-Acci-dent Stability Chapter of ATOG. Inadequate Core Cooling In the event that neither steam generator cooling or backup cooling is estab-lished, " Inadequate Core Cooling" will occur. This topic is discussed in the Backup Cooling Chapter. i O l DATE: 7-6-82 PAGE 76
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWER GENERATION OtVl510N 74-1125531-00 p TECHNICAL DOCUMENT Abnormal Transient Diagnosis and Treatment Although the type of transient may have become evident during the first 2 or 3 minutes after trip, plant monitoring is required to make sure that the transient is going as expected. Generally, after 2 or 3 minutes the plant wilI begin to stabilize within the " Post Trip Window" (examples of this were 1 given in the P-T Diagram Chapter). Actions have already been taken to iden-tify and handle the "f as t" transients and the systems which should be opera-ting have been checked to make sure that they are working correctly. Further plant monitoring should begin. At this stage the ef fort should now be to make sure that the plant stabilizes as it should. To do this the P-T diagram J J is kept under surveillance. If reactor coolant pressure and temperature sta-bilize within the P-T post trip window, and steam pressure is above the low stean pres sur e limit, the transient is probably not abnormal and a quick check of the following should be made to ensure system and equipment para-meters are within expected values: Heat Transfer Balance Indicators
- P-T diagram (for RC presure and temperature and subcooling and secondary saturation temperature)
- Pressurizer Level
- Steam generator level and pressure Equipment Status and Operation (depending on what was started);
- Makeup /HPI flow rates and pump status
- Main or auxiliary feedwater flow rates and pump status
\
\
DATE: PAGE 77 7-6-82
~._ _ . - _ _ . . . . - ~ _ _ _ _ _ . _ _ _ _ _ - , _ _ _ . _ _ . _ _ _ _ .___ ._--._ _ _ -,
BWNP-20007 (6-76) BABCOCK & WILCOX uumsta NUCLEAR POWER GENERATION OtVi$loN 74- 12353i-00 TECHNICAL DOCUMENT RC pump operation including cooling water and seal inj ec t ion Position of important valves (letdown, PORV, main steam isolation valves, feedwater isolation and control valves, pressurizer spray valve) Reactor Building isolation and cooling systems Power supplies (AC and DC) Once these rev iews are completed a more thorough check can be conducted O and a decision made to determine if the plant is stable. (See the Chap-ter on "Pos t Trans ient Stability Determination" .) But if the first review of the P-T indicates that the reactor coolant pr es sur e and temperature are not going to remain within the post-trip win-dow (or return to it), or that steam pressure is below the steam pressure limit, then somethim; is wrong with heat trans fer and co rrect ive actions are required to bring the heat transfer into balance. Diagnosis There are three general symptoms of abnormal transients: e Excessive Primary to Secondary Heat Transfer " overcooling". e Not enough Primary to Secondary Heat Transfer " overheating". 1 e Loss of Reactor Coolant Subcooling Margin A loss of reactor coolant subcooling margin can also occur and can be combined with or caused by either " overheating" or " overcooling". O 1 DATE: PAGE 78 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER GENERATION DPISION
~
75 i 12 s s 31-00 t' ) \ / TECHNICAL DOCUMENT s_ - The characteristics of an " overcooling" type transient are shown on a P-T diagram in Figure 20B. The figure shows Thot and Tcold c ming together as the reactor coolant reaches an isothermal condition following reactor trip because the reactor coolant pumps are running. If no RC pumps are r un n ing , and the RC is subcooled T hot and Tcold will not come together
/ \
( ) because a tempe ra ture dif ference will develop across the core which
,U creates the the rmal driving fo rc e for natural circulation of the reactor coolant.
The exc es s ive heat removal by the steam generators will cause the average reactor coolant temperature to go down. As the temperature goes down the 73 reactor coolant will contract causing the pressurizer level to go down. I i
\ ,
If the effect cannot be offset by increased makeup flow to the RCS the pr es sur ize r level will go down which will cause the RCS pressure to also go down. If the pressurizer empties, the RC pressure will decrease toward the saturation pressure cauaing a " loss of subcooling margin". The o ve rc ooling is caused by a tempera ture decrease on the secondary side [ m
\ of the steam generators. Since the secondary steam is saturated after
( v' reactor trip (no supe rhea t ) the steam pressure must also be decreasing. Figure 20B shows this decrease in seconda ry pres sur e and temperature. The steam pressure must go below 960 psig after RC trip for the RC temperature to go out of the post trip window on the low temperature side. The steam genera to r pressure be fore reactor trip would have been I r i i \ ,/ DATE: PAGE 7-6-82 79
BWNP-20007 (6-76) BABCOCK & WILCOX N W BER NUCitAR POWER GENERATION DIVISION 7'-1125531-00 TECHNICAL DOCUMENT around 920 psig but superheated. Following turbine trip the steam pressure will go up to the safety valve setpoint and then decrease down to the turbine bypass valve setpoint. The ch a rac t e r is t ic s of an "ove rh e a t ing" type tr ans ient are shown in Fig-ure 20C. The figure shows the Thot - RC pres sur e trace. Initially the RC pressure and temperature decrease after a reactor trip while the existing SG inve ntory boils away. However, as the primary to secondary heat transfer is lost the RC average temperature increases. As the temperature increases, the RC expands causing the pressurizer level to increase and RC pressure to go up. This pressure increase will continue until stopped by the pressurizer PORV or safety valves opening. However, the RC temperature will continue to increase. This will lead to a " loss of subcooling margin" . Without heat being trans fe rred to the secondary s id e , the secondary side 1 steam will gradually cool which will also cause the SG pressure to decrease. 1 A direct " loss of RC Subcooling" without a simultaneous " overcooling" or l " overheating" caused by a secondary system malfunction is shown in Figure l 20D. This type loss of subcooling margin could be caused by a loss of RC l inventory (LOCA). This loss in inventory will cause the RC pressure to 1 fall but without the large decrease in RC temperature associated with the
" overcooling" type transient.
PAGE 80 DATE: 7-6-82
BWNP-20007 (6-76)
' BABCOCK & WILCOX NUMB ER NUCLEAt POwen GENESADON Olvl510N 74-i i 2333 i-00 TECHNICAi. DOCUMENT Mitigation The path for correction is charted and shown on Figure 21, " Accident Mi t i-gation Approach". The chart keys on the three general characteristics displayed on the P-T diagram: loss of subcooling, overcooling (too much steam generator heat transfer) and overheating (insuf ficient steam genera-1 tor heat t rans fe r) . The chart ties together a wide variety of informa-tion for corrective actions. With the exception of LOCA and the steam generator tube ru p tur e which is a special LOCA, cor rect ive actions for all abnormal transients are discussed in this section. LOCA and SGTR are discussed separately in considerable detail in Appendix F and Appendix C of these guidelines. Specific information for mitigation of overcooling and ove rheat ing transients are provided in Figures 22 and 23 respectively.
i Corrective Actions for Overcooling (too much steam generator heat transfer) Figure 22 shows the correct ive actions to be taken for ove rcooling . The chart is largely sel f-explana tory so only a brief discussion will be given. Information provided by the chart will not be repeated. Overcooling is always caused by failures on the secondary side. The usual sources of failare are low steam pressure or excessive main or aux-iliary feedwater or by combinations of high feedwater and low pressure. l The P-T diagram shown is typical for a more severe case of overcooling; l 0 usually excessive feedwater alone (unless severe and not terminated l (s , within 2-4 minutes) or small reductions in stean pressure will not cause DATE: PAGE 81
- 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NuusEn NUCLEAR POWER GENERAflON DIVI 510N 74- 125531-00 TECHNICAL DOCUMENT loss of s ubcoo ling . But the general trend shown by the P-T diagram is characteristic of overcooling. Some LOCA's can also cause a loss of steam pressure because the RCS will depressuri::e, cool and draw heat away from the steam generators; this will be temporary for small breaks. Once this overcooling trend is exhibited, checks should be made on steam pr es sur e , steam generator level, loop Tcold temperatures, and main or auxiliary feedwat er flow. If the cause is obvious, then actions to isolare the cause should be taken. If the subcooling margin is lost during an overcooling transient, the subcooli ng rule should be followed. When the subcooling margin is lost the RC pumps should be tripped and not restarted until subcooling is restored. j Overcooling transients induced by steam pressure and/or feedwa t e r control l l failures on only one SG may be obvious when the Teold tempe ratures are l cocpared. If the magnitude of the overcooling is signi f ic ant , Tcold in the af fected loop will always lead Tcold in the loop with the good SG (i.e., Tcold in the af fected loop will be lower). Detection of overcooling by low steam pressure and determination of which generator has failed can be done by two methods . The first way is to stop all feedwa t e r when a level exists in both generators (if level O DATE: PAGE 7-6-82 82
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POWER GENERATION DIVISION 7'- 12n31-00 gTECHICAL DOCUMENT i O exists only in one generator the one without level is likely to be failed). When feedwa t er is isolated to both generators , the level should fall at a faster rate in the failed generatot. Detection h possible be fore both genera tors boil dry and feedwater should be restored to the
" good" gene ra to r before it dries out. A characteristic feature of steam leaks is that the steam generator with the low pressure will transfer the heat from the RCS and lower the RCS temperature; the " good" ge nera to r f
will not. Consequently the RCS temperature can fall below the tempe ra-ture of the " good" generator. When this occurs the pressure in the good generator will drop below the TBV setpoint, the TBVs will close, and stean generator level will drop slowly or not at all. Consequently, one s tea:a generator will retain level, so stopping all feedwater to both generators for a short du' ration is not dange rou s . However, when one steam generator - boils' dry, then the remsining generator will begin to t rans fe r heat and level will . drop. Feedwater must be restored both SGs become dry. The second (but less reliebie) way to identify a leaking generator is to ! compare the rate of drop of steam pressure in both generators. The l failed gene ra to r will permit steam pressure to fall faster than the good generator will. A dif ferential pressure between the two steam generators of abou t 100 psi, with the failed generator lower, will show the correct one to isolate.. The 100 psi differential will show up rapidly for large leak s and slower for small leaks. A steam le.ak of about 5% total flow i (about equal to one main steam safety valve stuck open) will show this DATE: 83 7-6-82
l BWNP-20007 (6-76) BABCOCK & WILCOX NumsEn NUCLEAR POWER GENERATION DIYl$ ION 7'- 23331-00 TECHNICAL DOCUMENT trend within 3 to 5 minutes. However, this magnitude of pressure differ-O ential may exist only for a short duration. The re fo re , comparison of level changes in isolated SG's is preferred, if the affected SG is not obvious. Correct ive actions for exces s ive feedwater are shown on the overcooling diagnosis chart (Figure 22) and discussed in Appendix A. i l l Corrective actions for low feedwater temperature are also shown on Figure f 1 22. i l O O l 1 O PAGE 84 DATE: 7-6-82
i BWNP-20007 (6-76) : BABCOCK & WILCOX su sen NUCLEAa POWee GENteATION DIVISION 7'- n 2s sa i-00 ; c TECHNICAL DOCUMENT ' ~ l ! Corrective Actions for Overheating (not enough steam generator heat . transfer) i , ' Figure 23 shows the corrective actions to take when the reactor coolant j cannot transfer heat to the steam generators. Reactor-to-steam generator heat t rans fe r is not coupled; Te and SG T sat do not track together. In general there are three causes of insuf ficient heat transfer: e There is no inventory in the steam generato r to rece ive the heat (loss of all feedwater) e The reactor coolant cannot trans port the heat to the ge ne ra to r because there is insufficient inventory (LOCA) e Circulation has stopped (forced and natural) Natural circulation can be temporarily interrupted because of reactor coolant contraction after a severe overcooling transient. A long inter-ruption would not be expected since HPI will refill the system and natu-ral circulation will normally restart. Loss of natural circulation would I be expected for most LOCA's or for an extended loss of feedwater. There-fore, to restore natural circulation for either of these, the failure must be corrected ( feedwater restored) and subcooling should be restored (to ensure the best natural circulation). Since loss of natural circu-lation heat transfer is the most probable of the overheating conditions, h DATE: PAGE 85 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numsta NUCttAn POWER GENttATION DIVISION 74-1125531-00 TECHNICAL DOCUMENT the overheating" corrective actions include restoration of heat transfer as an integral part of the action. Loss of all Feedwater Ove rhe a t ing when all feedwa t er is lost can take dif ferent paths depending on the decay heat level, when feedwater was lost and whether llPI was op- ! e ra t ing be fo re it was lost. The Figure 23 P-T diagram illustrates a total loss of all feedwater immediately after reactor trip from full power, with MU/ilPI cooling started when the operator recognized the loss of pr ima ry to secondary heat trans fer and loss of all f e ed wa t e r . MU/llPI cooling is di sc u s sed in Chapter E. Regardless of the path, the loss of all feed wate r from power operation will exhibit two clear characteristics on the P-T diagram: e After the no rmal post-trip cooldown the RCS will begin to reheat and repressurize. The RC pressure will reach and stop at the PORV or pressurizer safety valve setpoint. The RC tempe ra ture l 1 l will continue to increase toward saturation temperature. o Stem pressure and steam temperature will drop because there is l no feedwater. The co r rec t ive actions for this transient are to attempt to restore feed-O water; failing to do so, MU/IIPI cooling should be started when primary to secondary heat t rans fer is lost. Two MU pumps should be started and run at full MU system capacity taking suction from the BWST then the PORV i l O l I i i DATE: PAGE I 7-6-82 86
a BWNP-20007 (6-76) i BABCOCK & WILCOX u,u ,, ; NUCitAs POWts OtNWAflON olVISION 75 ii2 n 3i-00 1 TECHECAL DOCUMENT l l i manually opened (and verify HPI actuation if the low RC pressure SFAS se t po int is re ach ed ) . Without this action, the loss of all feed water transient will result in' losing primary inventory out of the PORV (or pres sur ize r safety valves) for about 30 or 40 minutes before subcooling ; margin is lost at which time the loss of subcooling margin rule will be
! impl eme nted . Howeve r, waiting for a loss of subcooling margir. could lead {
x ! i t c' conditions of core uncovery. There fore , the operator should start MU/HPI cooling when primary to secondary heat transfer is lost to enhance ! core protection rather than waiting until the subcooling margin is lost i , i and implementing the subcooling rule. I i When MU/HPI cooling (with the PORV open) is started under these circum-
~
stances, all but one RC pump should be tripped. This will reduce the heat input to the RC. One RC pump should continue to run as long as pos-sible to maintain forced core cooling. When the subcooling margin is i lost all the RC pumps must be tripped. L l Continued operation without feedwater and with MU/HPI cooling will allow subcooling to be restored if the heat capacity of the MU/HPI flow becomes suf ficient ~ to remove the decay heat when the pressurizer becomes water i solid. If the subcooling margin is restored the MU/HPI flow may be
' throt tled . A reactor coolant pump should also be restarted at this time.
1 DATE: 7-6-82 PAGE 87
BWNP-20007 (6-76) BABCOCK & WILCOX " " NUCLEAR POWit OtNERAt4ON DIVISION 74- u 2ss3i-00 TECHNICAL DOCUMENT LOCA The other condition in which heat transfer to the steam generators can be interrupted is during a LOCA. LOCA is discussed in great detail in Appen-dix F and will not be covered here in detail . However, this section will show how LOCA's are to be identified and will show how to locate those that can be isolated. Although some small b reak s will allow the reactor coolant to trans fe r heat to the steam generators, some will not. The most significant charac-teristic that shows poor heat trans fer is an increase of Incore T/C t empe ra ture along the saturation line away from seconda ry Ts at when a steam generator level exists. Incore T/C temperature is increasing be-cause the reactor coolant is absorbing the core heat and not passing it to the generators. Steam pre s sur e and steam generator saturation t empe ra tur e will gradually drop because little or no heat is being absorbed. Figure 13 of Appendix F (LOCA) shows these P-T characteristics. i Figure 23 indicates that LOCA's can cause poor heat transfer and includes three references for supporting actions: O DATE: PAGE 88 7-6-82
i-i BWNP-20007 (6-76)
.SASCOCK & WILCOX wu een !
l NWCLEAR POWEs 04NGsATION DWi$10N I 74-1125531-00 TECHNICAL DOCUMENT ! (1) Table 4a shows how to distinguish LOCA's from other transients. (2) Table 4b shows symptoms for LOCA's that can be located and ( { shows wh ich equipment to use for isolation for those that can r
; be isolated.
j (3) Appendix F shows the corrective actions fo r LOC A's. , Pressurizer Control System Failures i
- j. Two failures of the pressurizer controls can occur that can change RC
- pressure. These are not serious events because they are slow, but if l they are left without correction, plant control can become more i.
! dif ficult . I l f Failure of pres sur ize r heaters (on) with no spray operation will cause ! I
- the RC pressure to increase to the PORV setpoint at a constant reactor i
f coolant temperature. (If the spray is operating it will stop the heater
! pressure increase.) Steam will be released to the quench tank until the i
j heaters are turned off. The normal makeup control will cont inue to add l reactor coolant. The first source of makeup water will be the makeup 1 i I tank. However, on low makeup tank level the MU pump section will switch to the BWST. _ Although this is a very slow transient and should be easy l to correct (manual cutof f or power disconnect) if it is left unattended, [ it can cause rupture of the quench tank rupture disk. ! i ! A spray failure (on) will . cause a pres sure decrease at a constant RC temperature until the reactor coolant becomes saturated. This may be
- corrected by blocking spray flow.
L DATE: PAGE [ 7-6-82 89 f i
1 l BWNP-20007 (6-76) 8ABCOCK & WILCOX NUMBER NUCtEAR POWER GENERATION DIVI $lON 74-1125531-o0 TECHNICAL DOCUMENT Neither failure is considered serious because there is ample time for O correction. Pres sur e and pressurizer level ala rms will sound be fo re these lead to serious problems. O l i i f l O 1 I O O I DATE: 7-6-82 PAGE 90
)
i Fioure 20a GENERAL PLANT ACCIDENT MITIGATI pN IMWEDIATE ACTIONS (2-3 N!NUTES) I REACTOR TRIP l MANUALLY TRIP REACTOR i AND TURBINE l 1 V VITAL SYST N STATUS VERIFICAfl0N P T DIAGRAN CHECK FOR I PLANT STATUS CHECK LACK OF SU8C00 LING l PER'ORM REMEDI AL OVERHEATING lACTIONIFREQUIRED OVERC00 LING l l 1 i I YES I NO l 4 i FOLLOW UP ACil0N
- FAILURE DIAGNOSIS AND CORRECT 10N -
STEAN GENERATOR HEAT TRANSFER RESTORED C00LDOWN PROCEDURE i ,. PLANT tTABiliZAil0N N0 i YES - ai I CORE COOLING BY BACKUP METH005 i I
- NO - YES l g l INADEQUATE CORE COOLING j V Gul0ELINES l
l t 7d-1125531-00 I l l. I.
Figure 20b EXCESSIVE PRIMARY TO SECONDARY ' HEAT TRANSFER 2600 OST TRIP 2400 - WINDOW , , 2200 - g p-- gq g R L__; ) - g 2000 - SUBC00 LED l 3 REGION SUPERHEAT $ REGION n. 1800 -
/
- 3 1600 -
/
$ TURBINE BYPASS 5 1400 - PRESSURE SETPOINT ,_O [ MS SAFETY g . END POINT POST TRIP WITH E _ VALVE SETPOINT q , FORCE 0 CIRCUL Ail 0N (TH0T g j &TCOLO) AND FOR NATURAL F* CIRCULAil0N (TCOLOI
~~-~~ ~~~~~
. S NORMAL OPERATING POINT-800 -
STEAM g POWER OPERAil0N (TH0T) = i , PRESSURE E LIM;T . SATURAil0N r- - i END POINT-POST TRIP WITH l _ _j NATURAL CIRCUL Ail 0N (TH0T I SUBC00 LED 400 - MARGIN LINE I I I I I C 550 600 650 700 400 450 500 Reactor Coolant and Steam Outlet Temperatiire F 74-1I2553I-00 f
7 Figure 20c LOSS OF PRIMARY TO SECONDARY o HEAT TRANSFER 2600 2400 - POST TRIP [ WINDOW
- PRESSURIZER SAFETY VALVE r _ _ __ ,
2200 -- ' _ i . SETPOINT ! L _ __ _ a
.; 2000 -
SUBC00 LED . 5
*', REGION
* - REGION 1800 O
~
1600 - E - 5
=
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" FORCE 0 CIRCULATION (TH0T 1200 p.
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- NORilAL OPERATING POINT-b
" 800 -
3 p STEAll PRESSURE POWER OPERATION (TH0T) LIFT *
" r-1 END POINT-POST TRIP WITH SATURATION 1
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SUBC00 LED 400 - MARGIN LINE I I I I I 500 550 600 650 700 400 450 Reactor Coolant and Steam Outlet Temoerature F 74-1125531-00
, e
I Figure 20d INADEQUATE SUBC00 LING MARGIN l 2600 POST TRIP
'400 -
WIN 000 , 2200 - p--- q g h L___; - p 2000 - SUBC00 LED REGION SUPERHEAT 3 1800 - REGION j [
- o 5 1600 -
3, [ g TURBINE BYPASS !
# 1400 - PRESSURE SETPOINT g m e-END POINT-POST TRIP WITH E
MS SAFETY VALVE SETPOINT 3 l200 - FORCE 0 CIRCULATION (TH0T g j &TCOLD) AND FOR NATURAL E =. 7 CIRCULAT10N (TCOLD) 1000 - .i 3 800
/
[
- NORMAL OPERATING POINT-POWER OPERATION (TH0T) 3 - STEAM E 8 -
SATURATION r- 7 END POINT POST TRIP WITH T
' I 600 - L _j NATURAL CIRCUL ATION (TH0T SUBC00 LED 400 -
MARGIN LINE Gi i i l i I 490 450 500 550 600 650 700 Reactor Coolant and Steam Outlet Temperature. F 74-1125531-00 e i
J 9 1
, I
# inst 0l Af f Atil0h1 t $1 E $icilC410F PAR 1 ' -
IRIP RI AC100 TRIP f uRIlkt
'I III'h ' I" yl14L $11f19114tul AkB l$lt $ltit04 VlltFICAfl04 t$tt $ttfloh 110F PART l) utnE RAPIB $1 Alu$ Cnt CR C5 $f $tte$ 1 TRIP RC Pg U$t0 FOR 2 shCRE A$t Pr Rt AC111111 C04til0L ptet TRA41
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= $
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$14fl04 EA4AGER If 9
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\ 74-1125531-00 i
r
s I Aho lvailable On Aperture Car <l Figure 21 TRANSIENT MITIGATION APPROACH ,. i [ If If Dotis6 sAssig FOLLCeu? ACTIC45 FOR SGTR 11 A 0F PART 1) l$lt SECflom til 0 0F PAR 1 f ) P5 i BI41ei!E OFF11TE RAnl A1104 leART TO SEC0hCARf g[t[33[$ PRivEnt OPEhlhG e5$Vs
' RECUCE $ffle FIED FLCe CF ,,,Et APERTU Rf 00CE PRleARY TO SEC0hDARY LE AA=GE n hf 0RT gE POS5IBLE RC LEAR 2 C00LOCs4 OEFORE east (spi!
AC e&TER lhv(hTORT WI & LPI) [ TE5 I 40 (Rei h4T E S l VES If If g Il ]L
'I
) PL Ahi C00LCCs4 CN001E CORRECT C00LCCe4 e00E AND C00Loce4 THE Ptahl h0RBAL C00LCCe4 (StE CP 102 PART I)
OR C00LDCe4 AFTER L ARGE LOCA sHE4 q CF TAMES NAvt lePilED 5!0P QVE AC00 Ling ESTAgtl$n PRisART C00 Ling ;5EE CP-101 PART 1) ($tt $ECTION Ill C OF PART I) (SEE SECTION 118 8 0F PART I) OR 1 14CRE A$( PRleART 10 SECCh0ARY C20LOCs4 etTM RCS SATURafE0 I REDUCI PRinART TO 5ECONCARf MENT TRANSFER C00L lhG eMN $6s RE50VinG ME AT
. $TCP FEE 0 EATER OvtRFito RAI5E SG LEVEL (SEE CP-103 PART I)
OR
$fCP SG STI As LEAN RE0uCE SG PRE 55uRE 2 IF SG lhVtNTORY NOT AvAILAstt C00L0ceu elfw RCS SATURATED 2 INCR!a5E RC IWWE4f 0RT TO
$ FART su MP1 C00 Ling elTM su HPI COOLING COBFE45 ATE FOR C0hiRACflon (SEE CP IO4 PART ])
OR C00LOCs4 elin RCS ENTER 50iiO 1RAN$114T TERelNATED TRat$ LENT TEReinATED ($lt CP-105 PART I) OR N0 l TES VIS l h6 C00LOCet FOLLCeinG ICC (Lit CP 106 PART ]) U V U 8 40 3 0 5 0114 -d fo !
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l ATTEMPT TO RESTORE FEE 08ATEh START MU/HPi COOLING DHEN PRIMARY TO SECONDARY HEAT TRANSFER IS LOST STOP ALL BUT UNE RC PUMP TO REDUCE THE REACTOR C00LAF.T HEAT LOAD OTHER Al S:E E01 TRIP ALL RC PUMPS ON LOSS OF FOR HPl SUBC00 LING MARGIN TRIP U SATURATED CONDITIONS WILL EXIST UNTil 04 FEE 0 WATER IS RESTORED OR MU/HPl FLOW MATCHES DECAY HEAT (HOURS) SUBC00 LING MARGIN RESTORED BY OTHER f MU/hPI COOLING SEE EOL FOR HPI UNE RC PUMP SHOULD BE RESIARIED WHEN SUBC00 LING MARGIN IS REGAINED. A MOMENTARY LOSS OF SUBC00 LING MARGIN MAY OCCUR ON RESTART t s 7I4-1125531-00 I i
I Also Avail:ble On Ap;rture Dd Figure 23 OVERHEATlhG DIAGNOSIS CHART GENERAL OVERHEATING OCCURS WHEN THE REACTOR COOLANT CANNOT TRANSPORT THE CORE HEAT TO THE STEAM GENERATORS FOR HEAT REMOVAL. NATU9AL CIRCULATION WILL NORMALLY BE LOST FOR AN EXTENDED TIME (VS RRIEFLY INTERRUPTED). T H0T WILL BE SATURATED SINCE THE STEAM GE!;[RATOR CANNOT REMOVE HEAT, STEAM PRESSURE AND T 3At SG elLL DECREASE. GENERALLY ONLY it0 CONDITIONS WILL PERMIT OVERHEATING: t0CA's ANO LOSS OF ALL FEE 0 TATER. LOCA SEE FIGURE F-13, APPENulx F. "LOCA" IN PART 11-2 DISCUSSION OF SELECTED TRANSIENTS" FOR: A - P-T DIAGHAM CHARACTERISTICS PERENCES: B - CORRECTIVE ACTIONS PMENT OPERATION CHAPTER COOLING AND RC PUMP , SEE IIBLE 4A "HOW TO OlFFERENTIATE A LOCA FROM OTHER l TRANSIENTS" SEE TABLE 48 "$YMPTOMS FOR LOCA'S THAT CAN BE LOCATED OR ISOLATE 0" }FERENCES: tPMENT OPERATION CHAPTEF THROTTLING AND RC PUMP T! APERTURE CARD 8403050114 -. o ? '
Plant status verification
- 1. Reactivity contro11ea
- Reactor power decreasing
- All rods on sotton.
- 2. Secondary pressure controlled
- All main turbine stop valves shut and aan steam valves closed
- 3. RC inventory controlled
- Letdown throagh block crifice only
- letcown to MU tank
- 4. rW inventory controlled
- Feedwater flow provided
- Feedssater flow not excessive (FAST TR
- 5. Instrur entation and control available
- Vital CC and AC power, erergency stan busses and instrur,ent air
- 6. 5FAS actuation T
A3E b-C
- 7. 5FRCS actuation
- 8. Loss of cf fsite power (LOOP)
- 9. Steanline or condenser high radiation alarm dae to SG tute leak (FAST TPRiSIENT)
- 10. P-T diagram check Overneating Overcooling Lack of subcooling Mrgin 74-1125531-00
Mso Available On Aperture Card Table 2 STANDARD POST-TRIP ACTIONS pperator action required P3 sis for action e Emergency borate Af ter reactor and turbine trip, the operator should ensure that the fission process is shutdown. The simplest method is to ensure the rods are fully inserted; if not, the operator should manual'y trip both the turbine and reactor and ensure that a rapid decrease in neutron flus has occurred. Com-pensation for a stuck rod will have to be by boration to maintain a subcrit-Ital margin whca the plant is stabilized or plart cooldown is required. all e Close M51ds If stean pressure is allowed to drop below farbine bypass setpoint, the EC e Trip Tu"Line from front standard e Close block orifice bypass valve When a reactor trip occurs, T will decrease due to the loss of core fission power, and an outsurge from t$, pressurizer uill be caused by the contraction o' the reactor coolant. The MU control valve will open to increase MU in response to a decrease in pressurizer level. To minimize the potential fer a less of pressurizer level and/or indication, the ooer: tor should manually isolate the letdca typass of block orifice. It is not necessary to isv oe
" normal' letdown.
e Align letdown to M') tank Makeup should be directed to MU tank for makeup to RC5. 1 e initiate AFW feedwater should be added to maintain steam cenerator heat removal. N5!E4T) e Trip Mfd pumps Too much feedwater will lead to overcooling of the RCS and water in the steam lines. Water in steam lines could lead to s.eam line ruptures. by e Make use of what is available while trying instrumentation is needed to determine what the plant's condition is to deter-to re-estabitsh power and air sueplies. mine if corrective measures are neeaed and i' the correctiv s actions are ef-fective. Plant controls are needed to te able to correct any adverse plant conditiens. o Confim that tne HPI and LPI are started When an SFAS actuation occurs, the opera ;or should assure that at least cne e Verify trat at least one tu in in each ECCS train is operetive (one pump on) and tnat flow is present. At this point, is on (pump on). If not, try to start ECCS HP!/tPI flow balancing is not required, but can be done later. manually, e Verify, ty review of ECCS flow indication, that flow exists in the injection lines. e Confir'n containment isolation
"* - e Confirm containment cooling systems start
= e Trtp RC pumps on loss of subcooling margin
= e If containment isolation has stopped coh ing water to the RC pumps, either reinstate or trip pumps.
e Verify that appropriate MSIVs and MFSVs 5fRCS design function is to protect the com equences of simultaneous tilow-are closed. down of tioth steam generatcrs due to a stean line break. Upon detection e Verify t' tat at least one ATW pump is on of a 5;.eam line break, SFRCS automatically 'nitiates action to: and flow is delivered when 53 level is 1. Close both main steam isolation valves below the appropriate setpoint. AFW should 2. Close both main feedwater stop valves arts W MW sys and Ms W gm paMW en r to ain d to t od generator. The actions would isolate a steam line break downstream of the MSIVs 4.;d re-store SG pressure control. They also minimize the amount of RCS overcooling and release to the CV for steam end feedwater line breaks inside the CV. The operator should confirm that the above actions are complete by examination of valve position indicators and ATP status anG flow. AFW injection should te isolated from the generator with low pressure (if the STPCs isn!ation did not re nare pressure) to prevent overcooling, Verify that at least one emergency diesel Upon loss of normal and standby power sources, the two engireered safeguard e generator starts and automatic load busse', are energized, each powered by its respective diesel generator. Bus segaencing is completed. If only one load hedding, bus transfer to the oteset generators, and pickup of ci itical has started manually, start the other, loads is automatic. When a loss of power occurs, the operator should ensure that at least one diesel generator starts ar.d that Icading sequencing is completed and he should try to start the otter. (See ATOG quidelines Part II, Section 2. " toss of Of fsite Power," for details about equipment which is automatically loaded on the diesel and for Equipment which must be manually started. e Start imediate cooldown and depressuriza- Reactor coolant has flow path to outside environment via secondary system. Must depressurize RCS to stop tube leak. tion of RCS e Regain % beat removal Provide stable core cooling, Provide stable core cooling. o Reduce % heat removal Provide stable core cooling. e Replenish RC inventory 8 40 3 0 5 0114 ~/d - ,
Taele 3 ACTIONS TO CORRECT FAST TRAllSIENTS d ..
'h f
PLANT STATUS INDICATORS OPERATOR ACTION REQUIRED SYMPTOMS BASIS FOR ACTION .Y
- 1. Excessive feedwater Irenediately following reactor trip examine High SG Level flow Excessive MFW is the addition of water to the SG levels for excessive feedwater. High FW Flow SG at a rate faster than it can be boiled off.
It can result in overcooling of the RCS and e If SG 1evel is "high" and MFW flow is water spillage into the steam lines must be i still on, stop further MFW addition. avoided. Overfill of the SG can occur very rapidly because of the large flow capacity of e Allow SG 1evel to (ecrease to appropriate the MFW system; this is especially true follow-setpoint; then resume FW addition by ing a reactor trip. Following a reactor trip j the operator should assure that MFW runs back,
- manual control of MFW or If a FW ficrv remains high and SG 1evel is increasing, MFW should be controlled (trip
- AFW addition if MFW has been isolated. pumps).
NOTE: Ensure AFW starts if MFWP's are tripped. l 2. Steam generator Confim radiation monitor reading supports tube rupture Steam line or con- Steam generator tube leaks or ruptures are alarm; start an immediate cooldown and de- denser vacuum pump LOCA's which result in contamination of the pressurization of the RCS. The cooldown radiation alarm main system and offsite release. To should continue to cold shutdown. minimize the offsite release, a complete
- Note: Radiation cooldown and depressurization of the RCS is monitors may indicate a required to reduce the primary to secondary
] tube rupture leakage and to prevent unnecessary discharge to
- prior to the atmosphere through the steam generators.
j reactor trip, Since this is a LOCA, HP! must be kept on if
; subcooling is lost but this will keep reactor ,
j coolant pressure high and continue the leak. I Cooldown is required to lower RC pressure to stop the leak. 1 i j l 1 4 74-1125531-00
Table 4_a_ HOW TO DIFFERENTIATE A LOCA FROM OTHER TRANSIENTS Unique Characteristics of LOCA's e Rapid system depressurization to saturated conditions with little or no change of reactor coolant temperature (characteristic of all but the very smallest breaks) e Sustained saturation (HPI does not : return the reactor to a sub-ccoled state within 5-10 minutes after actuation) e Containment radiation (only for breaks in containment) NOTE: A steam or f eed line leak inside containment will cause high pressure, temperature and humidity but will not cause high radiation unless there is a SGTR. e 6 ':e an pressure, feed flow and steam generator level do not indicate overcooling (this helps to differentiate LOCA's from overcooling transients) e High steam line radiation alarms (tube leaks only) e Low le td own storage tank level (in the absence of all of the above, this indicates a leak outside the containment) NOTE: LOCA's CAN BE DIFFICULT TO DETECT, ESPECIALLY IF THE BREAKS ARE SMALL. THEY CAN OCCUR INSIDE THE CONTAINMENT 1 AND STEAM GENERATOR TUBE LEAKS ARE LOCA'S. IF THERE IS ANY DOUBT THAT AN ACCIDENT IS A LOCA, ASSUME THAT IT IS AND TAKE APPROPRIATE LOCA ACTIONS UNTIL CLEARLY PROVEN CTHERWISE. 74'11.'25531-00
Tante 4n STuPT0eS FOR LCCA'S THAT CAk' BE LOCATED OR ISOLATED This chart will aid in locating some breaks; all breaks cannot be located. Some breaks which can be located can also be isolated and the LOCA can be stopped. It may be difficult to distinguish small steam line leaks inside containment from LOCAs; building environ-ment will change for both and the steam pressure will not always be low. However, a LOCA will change building radiation levels. The reactor will repressurize and regain full subcooling with a steam line break. Symptoms for LOCAs that can be isolated Symptoms for LOCAs that cannot be isolated (Symptoms or alarms most likely to show location are underlined) (Symptoms or alams most likely to show location are underifned) Failure Locating symptoms !solating hardware Makeup and purification - Low makeup tank level Steamgeneratortube(s) - High steam line radiation system outside contain- - High CCW radiation Letdownvalveup-(,) stream of coolers - High steam generator level ment and letdown cool- - High CCW surge tank level ers (For breaks in letdown Pressurizer safety valves - Relief line flow alam cooler) - High quench tank level *
- Local sump levels, radia. - High quench tank temperature tion alarins Seal return Ifne and - 1.ow makeup tank level Seal return i o seal return cooler - High CCW radiation lation valve a HPI fnjection Ifne break - Flow italance between injection lines outside containment - High CCW surge tank level (High flow will be through broken line)
(For breaks in seal re-turn cooler) RC pump seal failure - High seal return temperature (s350oF)
- Local sump levels, radic- Combined with:
tion alams Low stage and upper stage pressures are equal and high Pilot operated - Relief line flow alam PORV isolation valve RCS instrtsnentation lines relief valve - High quench tank level *
- liigh quench tank tempera- - Pressurizer level - False high or low level reading tur_e - Pressures - False low pressure
- RC flow - False high or low flow compared with known pump operation Makeup-letdown feal- - High makeup tank level Footnotes: (a) If makeup tank drains, assure operating makeup ance (this is not a - Clean waste receiver tank Letdown valve (ascgntrol pump takes suction from BWST.
break, but is a loss - Makeup flow rate (+) seal (b) Loss of decay heat removal emergency procedure of coolant) injection flow (-) letdown should be implemented. flow Decay heat removal line - High or low decay heat re- Decay heat letdown break outside contain- moval flow drop ifne valve (b) ment (decay heat 'a- - Low pump suction pres. moval system in opera- - Local sump and local radia-tion-plant is cooled tion alarms down) Decay heat cooler tube - High CCW surge tank level Cooler isolation leak (decay heat removal - High CCW radiation valves system in operation - plant is cooled down)
*Will only be good when the quench tank rupture disk is good.
M-1125531-00
BWNP-20007 (6-76) SABCOCK & WILCOX NUMSER NUCLEAR POWfe GENERAflON DIVISION 74-u 2553i-00 TECHNICAL DOCUMENT CHAPTER D BACKUP COOLING METHODS The normal method of core cooling, when the RCS pressure / temperature con-g ditions are above those for DHRS operat ion, is by trans ferring heat from the core to the steam generators using subcooled reactor coolant. How-ever, if subcooled heat trans fer via the steam generators is lost two other means of core cooling are available. This section will discuss these backup cooling methods and explcin their uses and limitations. The two methods are:
- 1. MU/HPI Cooling
- 2. Boiler-Condenser Cooling HPI cooling is when the core is cooled by MU/HPI flow passing through the core (removing core heat) and out a break or open PORV to the containment vessel. This method is used whenever feedwater is not available.
Boiler-Condenser Cooling is when reactor coolant is boiled in the core to form stean (removing core heat) which flows through the hot leg pipe to the steam generator where it is condensed. The condensed water returns to the ' core via the cold leg pipe. In addition, this chapter also dis-cusses how to restore natural circulation heat removal if lost and what to do if core cooling is lost causing inadequate core cooling. DATE: PAGE 9g i
BWNP-20007 (6-76) BA8 COCK & WILCOX NUCLEAR POWit GENitATION DIVISION 74- 12ss31-00 TECHNICAL. DOCUMENT Backup Cooling by MU/HPI Cooling ( for Loss of All Feedwater) A c omple te loss of feedwa ter is not a likely event, but it can occur because of multiple equipment failures or ope ra to r error If primary to secondary heat removal is lost because of the loss of f eedwa ter , it may be pos s ible to cool the core by the MU/HPI system until fe edwater is re-stored. The core is kept covered and cooled by MU/HPI. The ccre energy is removed by the reactor coolant and released to the reactor building, which serves as the heat sink instead of the steam generator. A total loss of all feedwater without operator action is illustrated and discussed in Figure 24a. A total loss of all feedwater with anticipatory trip and appropriate operation is discussed below. The P-T response is shown in Figure 24b.
- 1. With the plant at power, a loss of main feedwater would result in a reactor trip (anticipatory trip on steam to feedwater AP or RPS actuation on high RCS pres sur e ) . A loss of all FW could also occur during hot shutdown or plant heatup/cooldown.
- 2. If auxiliary feedwater also fails including the startup fe ed-water pump, the secondary side of the steam generators will boil dry and the RCS will then heat up due to decay heat.
O DATE: PAGE 92 7-6-82
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- 3. RC pressure will increase as the steam space in the pressurizer is compressed due to the insurge of reactor coolant. Steam /wa-ter relief out the PORV or the pressurizer safety valves will i also occur. Subcooling may initially increase to the increased 1 pressure, but it will later decrease towards saturation, i
e i NOTE: The rate at which the above occurs will dpend upon the initial inventory in the steam genera to rs and the core t' decay heat level. For example , the RCS heatup rate may , be as high as 4F/ min with high decay heat or as low as IF/ min with low decay heat following boil off of the SG d inventory. The need to rapidly recognize loss of primary to seconda ry cooling and get the plant into MU/HPI cooling is imperative. If it is delayed too i long, the RCS saturation pressure reached when the PORV is opened will be above the shutoff head of the HPI pumps and HPI will not be avr.ilable for , L MU/HPI cooling, i % The event can be recognized by observation of steam generator level and equipment status checks or through use of the P-T diagram. The operator !. should make every attenpt to regain feedwat er to at least one stean generator. This includes main feedwater, auxiliary feedwater and startup l l feedwa ter. If feedwa ter cannot be regained before primary to secondary (- heat trans fer is lost he should manually start two MU pumps, open and DATE: PAGE 7-6-82 93
T BUNP-20007 (6-76) BABCOCK & WILCOX " NUCtfAR POWER GENERAtlON DIVl510N 7'-2125331-00 TECHNICAL DOCUMENT leave open the PORV and verify HPI actuation if the SFAS setpoint is r each ed . HPI flow should be balanced to give maximum flow possible. When MU/HPI is started the RCS should eventually go to a water solid con-dition (subcooled) with RC pressure controlled by a combination of the MU/HPI pump head rise and the relief capability of the PORV . The numbe r of running reactor coolant pumps should be reduced to one to reduce the heat input to the RC. The time to become water solid depends on the co. e decay heat level and the number of operating MU/HPI pumps. If core decay heat is high, the reactor coolant will saturate and its pressure will rise along the saturation curve (Figure 24b). The running RCP should be tripped when subcooling margin is lost. The water inventory in the RCS will drop because of core boiling and relief out the PORV, until the amount of water added by MU/HPI can take away all the decay heat of the core. Until that time water from the RCS and MU/HPI water together are needed to remove core heat. If the heat remove capacity of the MU/HPI water can take out all the decay heat, MU/HPI is said to " match" decay heat. When MU/HPI matches decay heat the liquid inventory in the RCS will stabilize. If MU/HPI heat removal capacity begins to exceed decay heat generation, RCS inventory will start to increase, steam will be condensed and the system will slowly go to a water-solid ( subcooled ) l state (Figure 24b). If tha core decay heat level is low at the time feedaater is lost, MU/HPI may be enough to take out all the heat and the RCS may remain subcooled (water solid) until feeedwater is restored. O l PAGE DATE. - 94 7-6-82
BWNP-20007 (6-76) i SABCOCK & WILCOX NUMSER NUctfAR POWER GENERATION DIYl510N 74-i i2353 i-00 S TECHNICAL DOCUMENT i When all feedwate r is lost it is very important that two MU and llPI pumps _ i e be run un t il sub cooling exists. With only one HP1 pump for core hear , removal, the RCS pressure may raise above the shutoff head of the llPI pump and the core may not be able to be cooled . l When subcooled conditions (based en both incore thermocouples and hot leg l l RTD's) are reached , the operator should start one reactor coolant pump i and throttle HPI flow to maintain the reactor coolant subcooled but ! l within equipment design limits such as the Reactor Vessel P-T limit of l l Figure 25, "RCS Pressure / Temperature Limits". In summary, MU/HPI cooling should be initiated if secondary heat removal is lost. It is not a normal operating mode for many reasons; three examples are:
- 1. The PORV will pass large amounts of solid and two phase flow since ' the valve was designed for steam flow it may fail to close when feedwater is reins tated requiring closing the PORV isolation valve to stop PORV flow.
l
- 2. Long-t e nn operation (subcooled solid water operation) must be closely monitored to prevent exceeding equipment design limits (Reactor Vessel P-T limit).
i 3. The degraded containment vessel environment may cause failure or bad readings of instrumentation. DATE: 7-6-82 95 t
._ _ _ ~ . - _ - . . _ _ _ _ _ _ . _ . _ . _ _ . . _ _ _ _ _ _ _ _ _ . _ _ - - - _ - _ _ _ _ _ _ - . . _ . _ . _ _ _ . _ . _ _ _ -
BWWP-20007 (6-76) BASCOCK & WILCOX NUM8ER NUCitAR POWit GENERAhoN DIV1540N 74-1125531-00 TECHNICAL DOCUMENT Consequently, secondary cooling should be restored as quickly as pos sible so that normal primary to secondary heat trans fer can be resumed . Boiler-Condenser Cooling A subcooled reactor coolant system is the desired state, however, core heat can be removed when the RCS is saturated. The problem with satu-rated na tur al circulation is that the operator doesn't know how much of the reactor coolant is steam and how much is water (see discussion of sat-uration in Adde ndum A of Chapter A). If the RCS is losing invento ry steam will form in the hot legs and eventually stop natural circulation flow (this is a violation of the subcooled natural circulation require-ment that a flow path exists connecting the hot water and the cold water. This could also be violated by a large collection of non-conde ns ib le ! gases in the top of the hot legs, however, such a collection is expected only following a core uncovery. At that point the operator would be using inadequate core cooling procedures). l Another fo rm of core heat removal could still exist under these condi-tions called core boiler-condenser cooling (boiling in the core and con-densing in the stean generator). Figure 27a illustrates boiler-condenser cooling. In this mode of cool-ing, water is boiled in the core. The resulting steam flows through the hot leg pipe to the s t ean generator where the stean is condensed in the O DATE: PAGE 7-6-82 96
d BWNP-20007 (6-76) i BABCOCK & WILCOX NuctEAR POWte GENERATION OfVl51oM 74-i i 2533 i-00 m TECHICAL 00CMENT
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cool steam generator tubes. The condensed water flows back to the reac-tor core. In order to flow back to the core the elevation of the RC water level in the steam generator must be above the elevation of the RC , pump di scharge pipe. This will allow the water in the cold leg pipe to flow up and over the RC pump discharge into the core. The RC stean entering the SG is condensed on the ins ide tubes surface be-tween the RC and SG water level. This tube surface has to be large enough to remove all the latent heat of the s tean at the expected RC steam flow rate. The amount of condensing surface is determined by the j secondary side water level. The steam generator tubes will condense the RC steam if the cooler feedwater is on the outside of the tubes. To make I the condens ing surface above the required RC water level high enough the steam generator water level must be at 93 inches on the startup range. I The problem with boiler-condenser cooling is the operator does not know the RC water level in the steam generator. The re fore , to assure enough water exists the operator should have both MU and HPI pumps on at full I capacity. If a LOCA occurs and the size is such that both stean generator and HPI i cooling are needed to remove the core heat but subcooled natural circula-l l - tion is lost, the RC will heat up and repressurize causing the pre s sur i-i zer safety valves to open and relieve RC to the containment vessel. This , l
- DATE
- 7-6-82 PAGE 97 I
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BWNP-20007 (6-76) BABCOCK & WILCOX I NUGEAR POWER GENERATION OlVl510N 74-t i2 ss ai-oo TECHNICAL DOCUMENT will continue until the RC water level drops below the steam generator feedwa te r level to expose su f ficient conde ns ing surface to the RC steam. With the feedwater level at 93 inches on the startup range, the RCS will inherently go into boiler-condenser cooling as the RC level in the steam generators boils down below the feedwa ter level to expose the condensing tube sur f ace. HPI flow will be needed to makeup fo r RC lost out the break when the RC pressure drops due to the steam generator condensing. If auxiliary feedwa te r is sprayed into the steam generator the ef fect ive condensing surface of the steam generator tubes is much higher than the 93 inches on the startup range (loss of subcooling margin level) because the AFW spray will be cooling the tube surface above this level. How-ever, if the main feedwater is used to add water to the steam generators, the feed water enters the bottom of the steam generator so that the only condensing surf ace is below the 93 inch level . If boiler-condenser cooling exists Thot should be equal to primary T sat and the incore thermal couple temperature, andcT old will be equal to the SG saturation temperature. The point to remember is that primary inventory (mass) is unknown under saturated conditions and therefore, every ef fort should be made to keep the RCS subcooled . O DATE: PAGE 98 7-6-82
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BABCOCK &-WILCOX- N"' NUCLEAR POwte OtNERATION DIVI $lON 74-1125531-00 (q >TECHNICAL DOCUMENT us Restoration of Natural Circulation When RC pumps are off, heat is removed from the reactor core by natural circulation as di scu s sed in Addendum B. Some abnormal trans ient s can lead to a loss of natural circulation but methods exist to restore it if it is lost. The intent of this section is to highlight the recovery i measures and to give an understanding of why certain actions are recom-me nded and when they are to be taken. Brief discussions of other issues on natural circulation are also provided. A loss of natural circulation can occur fo- two reasons, which are: Reason 1. Insufficient secondary inventory control (i.e., not
,. enough feedwater)
Reason 2. Formation of steam voids within the hot leg which are of sufficient volume to block water flow to the steam gene-l rator (i.e., not enough reactor coolant). N The previous section addressed system operation when natural circulation is lost due to insu f ficient feedwater (Reason 1). The only way to re-cover natural circulation under that condition is to restore feedwat er . i Void formation (Reason 2) is more complicated because the reactor coolant I l system can operate dif ferently depending on what has happened. The two principal events which lead to void format ion are overcooling transients and loss of coolant accidents. For these transients, voids are formed in the following manner: DATE: 99 7-6-82
BWP-20007 (6-76) BABCOCK & WitCOX Numsta NUctEAR POWER GENERATION Dri;$10N 74-il25sn-00 TECHNICAL DOCUMENT Overcooling: Too much primary to secondary heat transfer c au t.ca a drop of RCS temperature which causes a contraction of fluid inventory, a decrease in reactor coolant pressure, and a loss of pressurizer liquid. Some of the steam in the pressurizer flows into the RC piping and collects in the hot legs. Because the RC pressure drops, some of the reactor coolant may flash and cause void format ion in the hot legs. LOCA: A LOCA results in a loss of RC inventory and a reduced RC pressure. Voids are formed directly as a result of loss of RC inventory and also because of flashing of the reactor coolant as RC pressure drops. The RC tempera-ture does not drop as much as it would fo r an over-cooling event. Figure 26 illustrates the buildup of steam void s and the format ion of a steam bubble in the upper hot leg piping. The size of the steam bubble will depend on the rate of system over-cooling or loss of inventory versus the rate at which HEI adds water to the RCS to refill it. If HPI flow is large compared to the rate of RC inventory loss no steam bubble will form at all and natural circulation will not be lost. O DATE: 7-6-82 PAGE 100
BWNP-20007 (6-76) BA8 COCK & WILCOX EE NUCLEAR POWER GENERATION DIVISION 7e n 2 ss31-00
- TECHNICAL DOCUMENT If a steam bubble does form its size has a direct ef fect on primary to secondary heat t rans fe r. If the bubble is big enou gh such that the hot leg level is at or below the secondary side feedwater level then steam can be conde nsed within the steam generator tubes and the steam genera-tors can still remove a large amount of decay heat. This is a boiling 4
( mode of natural circulation (boiler-condenser cooling) and is illustrated in Figure 27a. Boiler-condenser cooling is an expected small break LOCA condition. If the steam bubble is smaller and steam cannot be condensed i l in the steam genera to r tubes (see Figure 28) then primary to secondary heat trans fer will be much lower. In this condi t ion the RCS may heat up and might repressurize. Several examples of transient conditions that could get into this mode of operation are: 4
- 1. Small LOCA's where HPI can match the leak rate (at reduced RCS pressures) and refill the RCS.
- 2. A severe overcooling event (e.g., major steam line break) in combination with d 31ayed actuation of HPI.
- 3. A total loss of feedwater, where AFW is restarted af ter the RCS is in a highly voided condition and the HPI is refilling the RCS.
Figure 27b shows boiler-condenser cooling (boiling in the reactor vessel and condensing in the steam generator tubes) with a saturated hot leg and RCS pressure near the steam generator pressure. For this condition, it is import ant to ensure that 1) SG level is at 93 inches on the startup range (to allow the condensed reactor coolant to flow over the cold leg DATE: 101 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR Powf t GENERATION DIVISION 74-1125531-00 TECHNICAL DOCUMENT pump clevation and into the core) and, 2) HP1 is on at full capacity (two pumps). A check of containment vessel pressure and tem pe ra ture condi-tions should also be made to see if the cause is a LOCA. If LOCA condi-tions are indicated, an immediate plant cooldown at design rates (100F/hr) should be inititated. The P-T diagram should be monitored to see if subcooled natural circulation returns or boiler-condenser cooling is lost. If natural circulation has been lost and stean cannot be condensed in the steam gene ra tors (that is, the steam bubble is in the top of the hot leg pipe, "c andy c ane" , but not low enough to be in the steam generator tube region) the RCS will repressurize. This mode of operation will be indi-c a ted (see Figure 28) by saturated hot leg conditions with Reactor Cool-ant pressure above the steam generator pressure (SG pressure may be d rop ping due to lack of primary to secondary heat trans fe r) . As indi-c at ed in Figure 28 the same actions identified for boiler-condenser cooling apply to this operating mode. In this mode, the HPI is refilling the RCS. During refill, steam in the upper region of the hot leg piping will be compressed and/or conde nsed as the water level in the loops and steam generator rises. In some cases (i.e., low decay heat with both HPI pumps on) subcooling and natural circulation will occur with minor in-creases in RC pre s s ure . Under other circumstances, it may be difficult to fully condense the steam in the hot leg and restore natural circula-tion. Figure 28 shows the actions to take to rectart natural circulation O DATE: 102 7-6-82
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BWNP-20007 (6-76) BABCOCK & WILCOX NuusER NUCLEAR POWER GENERAfloN DIVIStoN 74-1125531-00 Cs TECHNICAL DOCUMENT
\ l u if the stean generator can be used as a heat sink and the RC pumps are available for restart (see the RC pump restart guidelines in the " Equip-ment Ope r a t ion" chapter). If the RC pumps are available, pump bumps (short run times of 10 seconds duration) are allowed. This momentary use of fo rc ed circulation tries to force reactor coolant stean conde ns at ion A
[ ]/ by mixing steam with liquid reactor coolant and by moving the steam into
\ \_ / Use of the PORV to reduce RCS the generator tubes where it can condense.
pressure and increase HPI flow is also allowed ( separately or in conj unc-t ion with RCP operation). To be e f fec t ive the s tean generator must be a heat sink; the steam generator saturation temperature selected is somewhat arbi tra ry ; it was chosen to ensure a strong tempe ra ture grad ient for condensation. When the pumps are bumped and steam is condensed the f,,e I! RCS pressure will drop as much as several hundred psi. HPI flow will w-increase to help refill of the voids. If natural circulation starts the RC pressure will stay low; if natural circulation does not start the RCS will repressurize and another bump can be used about 15 minutes later (see the pump restart guidelines). Finally, a LOCA of a certain size could depressurize the RCS below the l l ; l , j stean generator pr es sur e before it se t tles to an equilibrium with HP I l (HPI will automatically start because this size LOCA will drop pressure below the SFAS s e t po in t ) . If this happens the operator should lower the steam gene ra tor pressure (u.ing the TBVs or AVVs) until the saturation t empe rature on the secondary side is 50F below the primary saruration
,,m This will ensure the steam generators are heat sinks. / \, t empe rature .
( ) C' Other actions are the same as discussed above for boiler-condenser cool-ing (Figure 27b). DATE: PAGE 7-6-82 103
BWNP-20007 (6-76) BABCOCK & WILCOX " EI NUCttAR POWER GENERADON OtVIStoN 74-1125531-00 TECHNICA'. DOCUMENT Inadequate Core Cooling (ICC) The first obje c t ive of the operator during any sbnormal transient is to keep the core cooled. As discussed in Addendum A, core cooling is taking place wh eneve r the reactor coolant is in a subcooled or saturated state and the core is covered. If the reactor coolant becomes superheated, the core has been uncovered and is not being adequately cooled; that is, decay heat is not being removed fast enough and the temperature of the fuel and cladding are increasing. This, in turn, causes the reactor coolant to heat up, flash to steam, and become superheated. Inadequate core cooling is not expected as long as the ATOG are followed . However, any transient can become an inadequate core cooling event if enou gh equipment f ailures happen. These events have a low probability of occurrence. Some examples where ICC conditions could develop are:
- 1. Small LOCA with a total failure of the HPI system.
- 2. Total loss of feedwater (both main, AFW and SUFW) with a total
( f ailure of the MU and HPI system.
- 3. A total loss of power including both diesel generators with a failure of the stean turbine-driven AFW pump to run.
- 4. During a small break, tripping the RC pumps at a time period when the RC void fraction is about 70% or greater rather than immediately tripping the RC pump on loss of subcooling margin as required.
DATE: PAGE 7-6-82 104
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLEAR Powta GENERAT'c)N OlVl5 TON 7'-i 25531-o0 TECHNICAL DOCUMENT The intent of the Inadequate Core Cooling (ICC) guidelines is:
- 1. To' allow the operator to identify when core cooling is inadequate and provide corrective actions. .
'2. To provide the operator with a way to es t imate the severity of the accident.
- 3. To ident i fy those systems which are vital so that the operator's attention will be focused on these items in his attempts to re-establish core cooling.
- 4. To identify some known alternative actions to try to correct or minimize the consequences of the accident until normal cooling can be re-established. These actions are based on the severity of the aceident.
ICC is indicated wh en the reactor coolant pres sure and temperature enter the superheat region. The response time of the temperature detectors must be considered when de termining if the reactor coolant has become l superheated . For a rapid decrease in RC temperature the hot and cold leg g tempe rature detectors will indicate a higher temperature than actually exists due to the response time of the detectors. For example: A large LOCA will cause the RC pressure to rapidly decrease to the saturation p res sur e . The RC hot leg temperature will also decrease. However, the indicated hot leg temperature will not drop as quickly as the actual temperature. This will give an indicated superheated condition while the actual RC condition is saturated. DATE: PAGE 105 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER GENftATION OfvtSION 74- t i 23331-00 TECHNICAL DOCUMENT The incore the rnocouple will respond quicker than the hot and cold leg t empe ra tur e detectors. Therefore, if the loop tempe ra tur e de tecto rs indicate superheated conditions. The existance of superheat should be ve ri fied by the incore thermocouples before proceed ing to the ICC guide-lines. This condition can occur with or without forced circulation. If the RC pumps are operating supe rhea t ed condi t ions imply that the reactor coolant is nearly all steam (see Figure 24a-Time IV). That ic, the liquid in the RCS has been lost, due to a leak in the primary system or boiled off out the safety valve (assuming the PORV is isolated) by decay heat, and the stean mass left within the system is not enou gh to remove core heat even though it is circulated by the RC pumps. When the RC pumps are off, core cooling is accomplished by keeping the core cove red with a s t e am-wa t er mixture. If not enough cooling water (MU and HPI) is s uppl ied to make up for losses, the core will become uncovered and the core exit fluid temperature will become superheated (see Figure 24a - Time IV.) Supe rhea t ed t em pe r a tur e s , as indicated by the core exit the nnoc cu ple s , are ICC symptoms. (NOTE: Incore Thermocouples are tb- only valid tem-pe ra tur e me asur eme n t when the RC is i.o t circulating). These indicators can also be used to estimate how serious the situation is. Analyses have been pe rfo rmed wh ich show the relationship be tween core exit steam t emperature and fuel cladding temperature for various RC pressures (see Figure 29). This figure gives the rollowing information: O DATE: 7-6-82 PAGE 106
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- 1. When the RCS P&T conditions are superheated but to the left of l
curve 1 on Figure 29, and TCC condition exists but it is not serious enough to cause core damage.
- 2. If the RCS P&T conditions reach or exceed Curve 1 on Figure 29, the 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. A chem-
) ical reaction between the cladding and the water at high tem-peratures also occurs and will add heat to the fuel rods in-c reasing the chances of fuel failure. The clad-water react ion also causes free hydrogen formation which collects in the h reactor coolant loops and may escape to the containment vessel. The accumulation of hydrogen in the RC loops can also block natural circulation when water is added to the RCS. If RCS P&T conditions reach or exceed Curve 2 of Figure 29, the 3 l 3. cladding temperatures in the high power regions of the core may be 1800F or higher. This is a very serious condition. At this level of ICC, significant amounts of hydrogen are being formed, I and core damage may be unavoidable. Extreme measures are war-l ranted to prevent major core damage. If an ICC condition develops, the operators should try to get equipment working to supply water to the reactor and/or stean generator. The general strategy during ICC should be as follows : DATE: PAGE 107 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX III NuctEAn rowEn GENEnADON DIVISION 76-112ss31-00 TECHNICAL DOCUMENT
- 1. To check vital equipment e All available MU/HPI should be on with flow into RCS. (For a total loss of feedwa te r , the HPI must be manually started.
SFAS will not be actuated automatically since RCS pressure does not decrease). e FW should be flowing to at least one stean generator with level at 93 inches on the startup range level instrumentation.
- 2. Start any backup equipment tu correct for problems found in vital equipment check, e Start standby FW/HPI pumps, e Take suction from any available borated water source.
e Start SU FW pump if AFW and MFW are not ope ra t ing . e Start any backup pumps thich can supply water to the steam generator if MFW, AFW and SUFW are not operating.
- 3. Minimize the consequences of the event if conditions degrade, o Start an RC pump to circulate primary system fluid (water or steam) through the core. This action will make available water trapped in the lower region of the reactor vessel and the loops for core cooling (see Figure 24a - TIME IV) and provide im-proved heat trans fer due to forc ed conve c t ion wh ich will pro-vide additional time to restore emergency injection, e At t enpt to decrease RC pressure by opening the PORV in order to increase the rate of available high pressure injection and open the high point vents to remove nonconde ns ib le gases ac cumula ted in the hot leg pipes.
DATE: PAGE 108 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCLEAR POWit GENERAftON DIYl510N 74-112s331-00 (G TECHNICAL DOCUMENT l
'%)
e If secondary cooling is available, decrease primary pressure by dec reas ing SG pressure. This action is directed at making the core flood tanks and LPI system available to restore core cooling.
/
rN 6 !(,/ l In general, the ICC strategy depends on operator action to locate and cor-rect the cause of low RCS inventory or to take alternate actions to make backup sources of cooling water available. Some of the actions ide n t i-fied above can be detrimental to major components, and others carry a cer-tain amount of risk, but keeping the core cooled is the first priority. For example, an emergency cooldown/depressurization of the system may icpose high thennal stresses on the SG internals; this action can be {'] (/ / shown to be acceptable but it challenges the design to its limit. A second example would be the restart of one or more RC pumps. This action carries some risk because later on a pump trip may leave the RCS with less water than be fo r e . These risks are small compared to those wh ich could happen with extensive core damage. Because the severity of the ICC condition can be es t ima ted (by using Figure 29), the appropriate ac t ions
, ,~
l' } have been picked so that the risk of the action is small compared to the \ /
'~
consequences if the action is not taken. These actions are outlined below and are based on where the RC pressure-incore thermocouple tempera-ture (P-T/C) combination corresponds to the curves of Figure 29. If the P-T/C combination is between the saturation curve and Curve 1 [ ; i., ) superheated conditions exist and the operator should: w, DATE: PAGE 7-6-82 109
BWNP-20007 (6-76) BABCOCK & WILCOX Nu= E n NUCttAR POWER GENERATION DIVl510N 74- 125331-00 TECHNICAL. DOCUMENT
- 1. Verify emergency cooling water is being injected through all HPI nozzles into the RCS,
- 2. Initiate any additional sources of cooling water available such as the standby makeup pump,
- 3. If st ean generator level is not at 93" on the startup range, raise level to the 93 inch level,
- 4. If the desired s tean generator level cannot be ach ieved , actuate any additional available sources of feedwater.
- 5. Es t ab li sh 100F/hr. cooldown of RCS via steam generator pressure control until secondary steam saturation temperature is 100F below the incore thennocouple temperature.
6 Open core flooding line isolation valves if previously isolated.
- 7. If RC pres sur e increases to 2260 psig, open the pressurizer PORV to reduce RC pressure and reclose PORV when RC pressure falls to 25 to 60 psi above the secondary pressure.
'Ihe se actions are directed toward depressurization of the RCS to a pres sur e at wh ich the ECCS water input exceeds core s tec:n generation.
The alignment of other sources of cooling water is the recognition that the injection of the HPI system alone is not sufficient to exceed core boil off. O PAGE 110 DATE: 7-6-82
.~ . . - .-
1
^
BWNP-20007 (6-76) i BABCOCK & WILCOX NUCitAt PCWER GENERATION OlV1510N 7'-il23531-00 TECHillCAL DOCUMENT ! If the P-T/C combination is between Curve 1 and Curve 2 of Figure 29, the operator should do the following: e
- 1. Start one RC pump in each loop; do not defeat RC ' pump interlocks.
- 2. Depres sur ize the steam generator as rapidly a pos s ible to 400 psig or as far as necessary to achieve a 100F decrease in secondary satu-ration t em pe ra tur e , but not below the steam pressure necessary for the AFW pump. turbine to deliver AFW.
- 3. Immediately continue the plant cooldown by maintaining a 100F/hr 1
cooldown rate until the sccondary saturation tempe ra ture is low enough to achieve a 150 psig RC pressure.
- 4. Open the power operated relief valve (PORV), as necessary, to vent i
noncondens ible ga se rs and to aid RCS depressurization and refill e l' i until the RCS is 50F subcooled or RCS pressure decreases to 150 psig.
- 5. Open all high point vents to vent noncondensible gases. ,
- 6. Raise SG level to 95% on the operating range, r
i l 1 The operator action in starting the RC pumps will provide forc ed flow g core cooling to reduce the fuel cladding temperatures.
\,.
The rapid depressurization of the steam pressures will help to depressur-ize the primary system to the point where the core flood tanks will actuate. Stopping the depressurization at 400 psig (or as far as neces-sary to achieve a reduction in secondary saturation tempe ra ture of 100F) m will maintain the OTSG tube to shell temperature difference within the s design limit. The continued cooldown to 150 psig will reduce the primary DATE: PAGE g 7-6-82
. _ _ _ - . __.- _ ..-_ _ _ _ _ _ _.---. _ . _ _ _ ___.._ ___._._ ~ . . _
BWNP-20007 (6-76) BABCOCK & WILCOX "# 0 I" NUCLEAR POWEa GENERATION DIVI $10N 74-1 25531-00 TECHNICAL DOCUMENT system pressure to the point where the Low Pressure Injection System can su ppl y cooling. The opening of the PORV will also help to depressurize the primary system. Tha PORV should be closed when the primary pressure is wi th in 50 psi of the secondary pressure and then should be used only as necessary to nmintain the primary system pressure at no greater than This method of operation 50 psi above the secondary system pressure. will minimize the loss of water from the primary system through the PORV. The high point vents should not be closed until the RCS is 50F subcooled or RCS pressure decreases to 150 psig. If the P-T/C combination is to the right of Curve 2 of Figure 29, the operator should: Depr es sur ize the steam generators as rapidly as pos s ib le down to low O 1. pressure while ensuring sufficient steam pressure remains in the steam generators to operate the turbine driven AFW pumps. (If the AFW pump is being supplied with steam from the auxiliary boiler or the SU FW pump is suppling FW, then depressurize the steam generators to as low a pressure as possible.) Start the remaining RC pumps. Defeat starting interlocks; do not 2. defeat the overload trip circuit.
- 3. Open the PORV and leave it open.
l The goal of these actions is to depressurize the RCS to a level where the core flooding tank s will fully discharge and the LPI system can be actuated, thus providing prompt core recovery. 1 PAGE DATE: 112 7-6-82
.. -... .-. .- _~ - .- . _ _ . _ . _ . - - - _ - _
j.
- j. BWNP-20007 (6-76)
'- CABCOCK & WILCOX NUMBER NUCLEAR POWER GENtRATION DIVISION " 74- 125531-00 TECHICAL DOCUMENT 4 After reaching Curve 2, significant core damage may have occurred wh ich j will add significant radioactive contaminants to the reactor coolant and via the PORV to the containment vessel. Special cooldown precautions need to be followed to contain these contaminants and hydrogen. ; i d i I l 1 i~ i f i t IG i ( ! i l i a j . 1 I DATE: PAGE 113 ! 7-6-82 . ymyy p _yqmnv _ , _ _ _ _ , _ , _ _ _ _ w wmi
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- 1. LOSS OF FEEDwATER.
- 2. RE ACTOR TRIP On NIGH RCS PRES $URE (OR ANTICIP ATOR'. T
- 3. PRESSURIZER LEVEL DECREASES (NORM AL RESP 0s5E FOLL0wl REACTCR TRIP) THEN INCREASES BECAUSE OF 4U ADDITION REnEAT OF RE ACTOR COOL AN T.
- 4. SECONDARY SIDE B0ILS DRY.
- 5. Afw DOES NOT START.
- 6. RC PRES;)RE SHowS NORMAL POST TRIP oESPonSE THER luCREASES AS PRESSURIZER LEVEL ll RESTORED.
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TIME IV (RC PUMPS On) Eno C0noltion FOR TOTAL L0ss 0F Fw wiinoui Mu/ net ACT SYSTEM wout0 COMPLETELY WOID ($ TEAM OnLY la RCS) TEMPERATURE WILL l# CREASE AND CaubE SUPERuf ATED FORM. In ADEQUATE CORE C00LluG Con 3lfl0N11X157. I I i 74-1125531-00
Figure 24a BACKUP COOLING BY MU/HPI FOR LOSS OF ALL FEEDWATER (NO OPERATOR ACTION)
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TIME II TIME fil
- 1. RC HEATS UP DUE TO LOSS OF SECONDARf MEAT Sint AND 1. RCS GOES SATURATED AT - 2500 PSI 6.
EIPANDS lhTO PRES $utilER.
, g I, 2 RC PatSSURE INCREASES TO PORv SETP0ini. Tut CORE.
PRES $UtilER STE AN IS EJECTED OUT OF PORV. S. PRIMARY SYSTEM G0ES v4TEt SOLID (STILL SUSC00 LED) AT All0VND THE LOOP (SEE ASOVE).
~2500 PSIG. CouTINUED ME ATUP OF F8. ACTOR COOLANT
- IF RC PUMPS ARE OFF. THE STEAM WILL SEPARATE FROM CAUSES WATER RELIEF OUT OF PORf OR SAFETY VALVES.
THE RE ACTOR C00Lani AnD COLLECT 14 TME UPPER REGION
- 4. RC TEMPER ATURE l$ SL0wlY APPRO ACHING SATURATED AT Ry AnD HOT LEG.
CONDITIONS e 2500 PSIG. TI
"~~9 IPERT0RE 4 q C/RD
-= % % .
h, LEGEND PRIM ARY SIDE SECONDIRf SIDE WATER
'4 WATER
\
STEAR h TWO PHASE L10ulo
*.:, m . N STEAM
. s N
x N
@h 5 TIME IV (RC PLMPS OFF) l08 - EnD ConD4Tl3n FOR TOTAL LOSS OF FW WITMOUT MU/MP1 ACTUATI0n.
WITM RC PW85 0FF EARLY IN THE EVEnf. REACTOR COOLANT CAN BE TRAPPED 14 LOWER REGIONS OF LOOP AND REACTOR VESSEL. CORE WILL ME ATUP WMER MIITURE LEVEL DROPS BELow TOP 0F FUEL. IR ADEOU ATE CORE C00Llue wlLL EXIST. Also Xvallable On 8403050114 -f/ \ Aperture CarEl
l l i , Figure 24b BACKUP C0OLING BY HPI FOR LOSS l 0F ALL FEEDWATER (WITH OPERATOR ACTION) 2600 POSI TRIP 2400 stN00s 2200 - . e, B -&
. 2000 - j $ SUSC00 LED [ .
O REGION SUPERN[AT 0 - 2 g RE GION
$ 1600 -
5 ' 5 I 1400 s 5 to, .o.s. ..os. .e., e..a p tr , 3200 - SIE Au PRESSURE h'osciocincuta'ionano 8 ceto) see natuN g Listi . ciecutation ii g,to, seemat optestist, Point.poste
; opgestion (t,,,,)
800
- bll C,
E - SATURAil0N l-~ ~l t o o Po i n t-Pos t retP sein 600 - b_)''C'"'' I'soi l SUBCOOLED 400 - MARGIN LINE 0 ' e i i n 500 550 600 650 700 400 450 Reactor Coolant and Steam Outlet Temperature. F Reference Time Points hqqrgh)_ Remarks 1-2 3 Reactor tripped or anticipatory loss of feedwater. Normal post-trip cooldown and depressurization in progress. AFW doeso not initiate. 2 100 Steam generators dry. 3 288 Operator diagnoses loss of heat transfer. initiates full PU flow and opens PORV. 3-4 344 RC goes saturated. Operator trips RCP's on loss of subcooling margin and verifies auto start of HPI on low RC pressure and balances HPI flow. Pressurizer level going off scale high. 4 361 PORV is flowirg steam and water. 5 _ MU/HPI has matched decay heat and the RCS returns to a subcooled condition. The operator starts one RCP when subcooled margin is regained and begins throttling of HPI to depressurize RCS. 74-1125531~00 s
Figure 25 RC PRESSURE / TEMPERATURE LIMITS (6 EFPY) 2400 I /
/
2200 I I j I --
/
2000 . 1800 l r 1600 , I 7
)
g (TYPICAL) ;
;, II0llNAL C00LD0tel *:, 1400 j REACTOR VESSEL P-T LINii [ = Fil0N TECHNICAL SPECIFICATION a 1200 N f o 1000 =
I l 800 UNACCEPTABLE q l 3 l IIARGIN I 600 - f l UNACCEPTABLE 400 g 200
/ s,' /
/.
/
/
l - 100 200 300 400 500 600 Reactor Vessel Temperature (*F) 74-1125531-00 e
m _-~ _ ~ m - . - - ~-==-- - - - - - - - - - - - - -- - - - - - - - - - - - d 7 i d -o@
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- - -1 i i s- 4
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TIME I
- 1. DECREASING DRESSURIZER LEVEL BECAUSE OF LOSS OF 1.
REACTOR COOLANT (LOCA) OR CONTRACTION OF REACTOR COOLANT (OVERC00 LING).
- 2. REACTOR TRIP DN L0n REACTOR COOLANT PRESSURE. 2.
E
- 3. SFAS ACTUATION ON LOW REACTOR COOLANT PRESSURE.
- 3. 5
- W I ACTUATION p
- RC PUneS TRIPPED BY OPERATCR (ON LOSS OF SUBCCOLING MARGIN)
- AFn ACTUATION (ON RC PtNP TRIP) 4- 2
- 5. I a
1'
' 74-1125531-00 1
7 j l Figure 26 ILLUSTRATION OF LOSS OF NATURAL 1 CIRCULATION DUE TO BUILOUP OF j STEAM IN THE REACTOR COOLANT SYSTEM QO O O 0 0o o 9 0,0n 0 9 0 oo O Y OI o, , > o . N
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0
- 0 N
bsd 0 (md 0 , o 0 0 6 m
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o, [\ ) A c: ^ a3
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O O o ,o N/
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/ N a7 TIME II TIME III ESSWIZER LIQUID VOLUME IS LOST; STEAM FROM 1. STEAM SEPARATES IN LPPER REGIONS OF LOOP.
ESSURIZER CAN ENTER RCS LOOPS nHEN THE RCS 3RESSURIZES. 2. 2-PHASE NATURAL CIRCll. ATICN STOPS. ACTOR COOLANT PRESSURE DROPS TO A VALUE ABOUT 3. SIZE OF THE STEAM BUBBLE DEPENDS ON SEVERITY OF luAL TO STEAM GENERATOR PRESSWE. ACCIDENT, FPI FLOARATE, AND HPI STARTtP TIK. EAM FORMS IN HDT LEG BECAUSE OF ACCUMULATION OF iESSURI1ER STEAM AND BECAJSE OF FLASHING OF REACTOR (1. ANT . STEAM IS IN THE FORM OF BUBBLES nITHIN RCS. PHASE NATURAL CIRCli.ATION NILL OCCUR - BOILING MAY CUR IN CORE. LEGEND A ARY S M SUFFICIENT STEAM IS CREATE 0, IT WILL START TO LLECT IN UPPER REGION OF LOOP BECAUSE STEAM CAN nATER / WATER SE AT A FASTER VELOCITY THAN WATER. o TWO PHASE LICUID STEAM
@ STEAM i
i Also Available On ! Aperture Card 8403050114 -/A !
Figure 27a ILLUSTRATION OF B0ILER - CONDEN3ER COOLING e STEAM () AFW () y ADDITIONAL CONDENSING V SURFACE IF AFW SPRAYING ON S.G. TUBES 93" LEVEL w CONDENSING SURFACE MFW + ~
' ' ~ ~
o oo o,6 ,o ]o o WATER LEVEL AB0VE o *o ELEVATION OF PUNP K*..*/ Ol' CHARGE X
/..s' x 74-1125531-00
Figure 27b B0lLER-CONDENSER COOLING , 2600 2400 - POST TRIP WlN00W 2200 -- ___ h I._____jh 2000 SUBC00 LED
- e. REGION
$ 1800 - f SUPERHEAT REGION y 1600 -
- n. _
O E 1400 - ~ M . u I
- I -
T g 1200 - T COLD H0T U g = 1000 SG PRESSURE
/
j i
- 7. -'3 END POINT POST TRIP WITH FORCED
___ _ _ h CIRCUL ATION
. _ T __ ._ - 2 800 -
STEAM p NORMAL OPERATING FDINT-POWER OPERATION 600 - - PRES RE
/
L. SUBC00 LING MARGIN [-] END POINT-POST TRIP WITN NATURAL 400 - . CI RCUL ATION' I f f l l 400 450 500 550 600 650 700 Reactor Coolant Temperature F NOTES ON P-T DIAGRAM
- 1. RC pressure is slightly higher than steam generator pressure.
- 2. T is equal to T sat for existing RC pressure.
hot
- 3. T is equal to T sat for existing steam pressure.
cold OPERATOR ACTION REQUIRED
- 1. Turn MU/HPI on to highest flow rate.
- 2. Turn AFW on and raise steam generator level to 93 inches on startup range.
- 3. Start plant cooldown at 100F/hr.
- 4. Monitor plant conditions for a loss of boiler-condenser cooling or a return to subcooled natural circulation (subcooling).
74"1125531-00
Figure 28 LOSS OF BOILER-CONDENSER COOLING , SYSTEM REFILL BY MU/HPI m,gg LEGEND y PRIMARY SIDE SECONDARY SIDE
-.-t y o o WATER h WATER
[o 3 o { TWO PHASE LIQUID STEAM o 7 o o { STEAM
._.= : ,
-i 0 n
NOTE: MU/NPI IS REFitLING THE PRIMARY SYSTEM. THE STEAM GENERATOR CANNOT REMOVE DECAY HEAT BECAUSE THE HOT _ 93" ON SU LEVEL LEs iS rua or STEAM Ano Ftos is noCm. NOTES ON P-T DIAGRAM 8* - Pelt telP einen 1. T H0T IS EQUAL TO T SAT
, tree -
FOR EXISTING RC PRESSURE. i I [_~_~_]l 2. j
~
g" - - - - - e ,["' RC PRESSURE WILL INCREASE ABOVE STEAM GENERATOR I 'm - PRESSURE AND CAN G0 AS I im A
' HIGH AS THE PRESSURIZER s ::
- SAFETY VALVE SETPOINT 5 "" [ t e esias n si toi, siin
! encia ciecetsvin ,,,,, (2500 PSIG).
Im 'este
- stem msnies tesi:
I a 'cas.,e 3.
! 'er c,ee , nam u,,,see
, , =arest STEAM PRESSURE MAY OROP 3 em 1
i _---{-_ i
/_ nessai arreari c e ,,, BECAUSE HEAT TRANSFER FROM
* - ** * *" " D " " w r ' REACTOR COOLANT IS LOW-j l sainavies , - . . , in p.,,, ,,,, ,,,, ,,,,
stem 4.
- PMESSURE Ljnatensiciecetaissaet,,, T COLD MAY OROP DUE TO HPl SueCootiNG MARGIN OR RC PUMP SEAL INJECil0N l m _
BUT WILL G0 UP WHEN RC l m [w d sn m
- a. .o. . c.. i .,ii ne s i e = e. o . i : =.. . . .. . t sw see PUMP BUMPED.
- OPERATOR ACTION REQUIRED
- i. SAME AS BOILER CONDENSER.
2. ESTABLISH THE STEAM GENERATORS AS A HEAT SINK (TSAT - SG SHOULO BE ABOUT 50F LESS TH THE INCORE THERMOCOUPLE TEMPERATURE). 4 OPEN PORV O BUMP ONE RC PUMP 74~1125531-00 , 4
L Figure 29 CORE EXfT FLUID TEMPERATURE F0F, INADEQUATE CORE C0OLING 2600 RC 2400 - SUPERHEATED RC SUBC00 LED 2200 2000 - 1800 - Q REGION 1 REGION 2 REGION 3 i 1600 - 5 y 1400 - S w I 1200 - g T ' a o CLAD o 1000 - C REGION 4 800 - 4 ' s b 600 - [ [e 400 - 200 ' ' ' ' 8 I I 400 500 600 700 800 900 1000 1100 1200 1300 Core Exit Tnermocouple Temperature (F) 74--1125531-00 "
j' -
)
BWNP-20007 (6-76) l I BABCOCK & WILCOX suma 74-1125531-00 TECHNICAL DOCUMENT CHAPTER E t BEST METHODS FOR EQUIPMENT OPERATION i During an abnormal transient the operator has to perform several actions l to control dif ferent systems. This section will show the best ways to do 4 the following things: ! e Start and Stop RC Pumps e Throttle or Stop HPI i e- Control or Stop Auxiliary Feedwater e Stop Main Feedwater l o Use the Incore Thermocouples e Cooldown with One Generator out of Service e Use of liigh Point Vents e Loss of LPI Recirculation Ability i RC Pumps During the course of an abnormal transient the RC pumps may need to be stopped and at a later time may be restarted depending on the kind of transient and the conditions of the reactor coolant system. In general the reasons for stopping the pumps are: e To prevent pump damage e To prevent possible core damage if a small break LOCA occurs e To decrease heat input into the RCS if SGs not available. l DATE: ^ 7-6-82 114
' ' " *~--a ,u. O O' THIS PAGE INTENTIONALLY LEFT BLANK O. O O: i l l 1
1 p BWNP-20007 (6-76) BASCOCK & WILCOX Numen NUCLEAR POwta GENERATION DIVl510N 7e n255n-oo j TECHillCAL DOCUMEllT In general the reasons a pump should be restarted are: e To start natural circulation if it has stopped by giving sub-cooled or saturated RC a " boost" around the loop. e To allow a faster rate of cooldown and RCS depressurization by forcing RC through the steam generato r and providing l pressurizer spray flow. j e To provide core cooling if the core has become uncovered (inade-quate core cooling) by forcing more flow through the core. e Prevent exceeding reactor vessel pressure tempe rature limits when the reactor coolant is subcooled without circulation and with IIPI flow by mixing HPI fluid with RC in the RV downcomer
~
( and providing an indication of RV downcomer temperature. This section vill show the rules and guidelines for stopping and re-starting RC pumps. i 1~
- RC Pump Trip j- RC pumps must be tripped during a small break LOCA if the subcooling mar-I
[ gin is lost to prevent core damage. If the pumps are kept running they will force s tean and water by the break; because the water is forced by [ the break more reactor coolant mass will be lost out the b reak than if I the pumps were not running. This can cause insuf ficient liquid in the l RCS. If the pumps are tripped later in the transient, when insufficient l' liquid remains in the RCS, the steam and water rc aining in the vessel { i and loops ' will separate (steam will collect in the high points and water i DATE: PAGE 115 f 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCitAt FOwla otNtaAflON DIVI 510N 7'- 2 ss31-oo TECHNICAL DOCUMENT will collect in the low points) and not enough water will collect in the O vessel leaving the core uncovered so that the core will not be adequately cooled. Core damage can result. Analyses show that a later pump trip can be dange rou s , but an early pump trip is safe. However, as long as the pumps continue to run the core will be cooled by the steam and water mixture circulating through the core. The RC pumps must be tripped upon loss of subcooling margin. A siginifi-cant loss of coolant accident will nearly always cause a loss of subcooling margin. Other transients, such as severe overcooling or loss of all fe edwa t er can also cause loss of subcooling margin. Because the effects of failure to immediately trip RC pumps during a LOCA can be very serious, the operator should trip the pumps on the loss of subcooling margin without trying to find out the cause first. To avo id core damage and RC pump damage, the following rule and associated note is given: RC PUMP TRIP RULE The RC pumps shall be tripped immediately whenever the subcooling margin is lost. O DATE: PAGE 116 7-6-82 l l
. .= . - _ , - _ _ . -- . -. - . -
BWNP-20007 (6-76) BASCOCK & WILCOX HUcttAR POwta GENERATION DIVI $10N 7'- t i 2 55 3 t-00 TECHICAL DOCUMENT . NOTE: IT IS ABSOLUTELY MANDATORY TO TRIP THE RC PUMPS IMMEDIATELY BUT IF l THE PUMPS ARE NOT TRIPPED IMMEDIATELY (I.E. , WITHIN TWO MINUTES) WHEN THE SURC00 LING MARGIN IS LOST IT IS MANDATORY THAT THEY SHOULD NOT BE TRIPPED AT A LATER TIME. THE OPERATOR MUST MAKE SURE THAT COOLING WATER AND SEAL 3-1 INJ ECTION ARE WORKING TO PREVENT PUMP DAMAGE. THESE SE RVICES MUST BE i MAINTAINED FOR SEVERAL HOURS. IF MECHANICAL DAMAGE TO THE PUMPS IS LIKELY, THEN W O PUMPS (ONE IN EACH LOOP) SHOULD BE STOPPED. THE WO REMAINING PUMPS MUST BE KEPT RUNNING. IF THEY FAIL THE TWO PUMPS WHICH ! WERE STOPPED SHOULD BE STARTED EVEN IF MECHANICAL DAMAGE IS LIKELY. THE OPERATOR MUST ALSO TRY TO GET AS MUCH HPI FLOW INTO THE RCS AS POSSIBLE. 4 The RC pumps can be tripped to prevent mechanical damage in all cases except the one noted above and during severe ICC, Mechanical damage is not expected to cause safety problems unless total seal failure occurs. It is desirable to trip the pumps to prevent mechanical damage in case they must be restarted at a later time. Prese rving the pumps for long-term cooling or cooldown is desirable, and it is recommended that they be shut down if high vibration or loss of cooling services occurs, j
\, Limits on continued pump operation are given in the "NSS Limits and i
l Precaut ions" PP1101-01. These limits apply to normal conditions and should also be applied to emergency conditions except where and when exceptions are taken by ATOG. O i DATE: 7-6-82 117
BWNP-20007 (6-76) BABCOCK & WILCOX NUCitAt POwta GENERATION OlVl310N 7'-ii23331-00 TECHNICAL DOCUMENT Table 3, " Rules for RC Pump Tr ips" summarizes these requireme nt s . In-cluded in this table are the limits on pump operation due to failures of cooling water and seal injection. These limits are shown because con-t ainne n t isolation can af fect cooling water. When the RC pumps are tripped to prevent mechanical damage, auxiliary feedwa t e r will be automatically started and the s tean generator level will be maintained at 40 inches on the startup range. The operator should make sure that natural circulation starts. If the pumps are tripped on loss of subcooling margin, natural circulation may or may not start depending on the amount of steam in the RCS. Nevertheless, a check on natural circulation is desired. Actions to establish natural circu-lation when the pumps are tripped because of subcooling margin are given in the pump restart guidelines which follows. l O O
^
DATE: 7-6-82 118
BWNP-20007 (6-76) BABCOCK & WILCOX NM E" NUCLEAR POWER GENERATION DIVISIC A 7e n 2 s s 31-00
/~';] \
TECHNICAL DOCUMENT v RC Pump Restart Core cooling and plant control are best if the RC pumps are running. Pumps can be restarted after trip if the reactor coolant conditions are right. The r e fo re , to complement the RC Pump Trip Rule given previously, conditions p, when the pumps can be restarted are given. These conditions cover both LOCA / (v, and no n-LOC A evente and have been carefully chosen so that a pump restart followed shortly afterwards by an inadvertent trip will prevent fuel damage for small breaks. Restart of the RC pumps is desirable for several reasons: 7~ e To mix reactor vesel downcomer fluid and provide a good reactor vessel l ( ') tempe r ature indication via the RC loop tempe r at ure detectors d ur ing HPI injection with no or low natural circulation, e If natural circulation was lost, the pumps can be used to restart it. e If the plant must be cooled down and depressurized , the RC pumps will permit use of the pressurizer spray, e Cooldown will be faster with forced circulation and the decay heat removal system can be placed in operation before the BWST is depicted.
,m o If severe In adequa t e Core Cooling (ICC) conditions exist the RC pumps
/ i must be restarted,
- i. l w/
The major rea son for restarting the RC pumps is to increase the rate of heat transfer from the core to the steam generators; or if natural circulation has stopped and there is no heat t rans fer from the core to the steam ge ne rato rs n (
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DATE: 7-6-82 119
BWNP-20007 (6-76) BABCOCK & WILCOX Nu stR NUCLEAR POWER GENERA 160N DIVI $ ION 74- n 25 531-00 TECHNICAL DOCUMENT then a pump bump will help to restart natural circulat ion. Bec au se the pur-pose of restarting the pumps is to increase core-to-steam generator heat transfer, it is necessary that the steam generator be available for heat re-moval. The s te au generator will remove heat if: 1) the steam generator satu-ration tempe ra t ure is lower than the RCS incore thermocouple temperature - a 50 F tempe ratui e di f fe rence is a good rule of thumb to use, and 2) the steam ge ne rator is fed with f e ed wa t e r. It is preferable to start the pumps in the loop with the operating steam generator if only one is in service, and it is best to start the pumps in the loop with the pressurizer spray if possible. Since it is pr e f e rable to keep the pumps operable, a pump restart is not desired if mechanical damage can result. One RCP should be restarted and run when the reactor coolant is subcooled and the core is being cooled by HPI cooling and no circulation exists. For this unusual situation, which can be caused by a prolonged loss of all feedwa t e r , one RC pump may be run even though there is no steam generator cooling. l If subcooling However, it should not be run if mechanical damage can occur. is lost the RC pump >hould be stopped. l l The RC pump is operated to mix the fluid in the reactor vessel downcomer and l to circulate suter through the RCS loops so that the loop temperature indica-tors (T h and T) c can provide a good indication of reactor vessel downcomer t empe r a ture . This tempe rature indication is used to monitor reactor vessel pressure temperature in conjunction with the limits of Figure 25. O PAGE 7-6-82 1
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCilAR POwta GENERAt4ON DIVl510N 74- t i 2 s s 3 t-oo O TECHNICAL DOCUMENT C i
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/
Inadequate Core Cooling is a condition where the reactor coolant is super-heated. This is a condition when core dwaage could occur. For this condition, exceptions are taken: 1) RC pumps can be restarted if the steam ge ne ra to rs are not available, and 2) If severe ICC conditions exist, RC pumps must be restarted even if mechanical damage can occur to the RCPs. For all [sh ( / other cases of pump restart, mechanical damage should be avoided.
%J When the RC pumps are restarted the operator should expect to see pressure ch ange s in the RCS as follows:.
e If the reactor coolant is subcooled and the pressurizer is filled g solid , an ab rupt rise or drop in pressure could occur. fi A
/ e If the reactor coolant is subcooled with a near no rmal pressurizer b
level, almost no change should occur, e If the reactor coolant is two phase and saturated, a large pressure drop could occur when the 1.e a t removal rate of the steam generator increases, fA, Table 6, "RC Pump Restart Guidelines", shows the conditions when the pumps r a can be restarted. The table is divided into three parts: subcooled, satu-rated, and su pe rh e a t ed . Guidelines for restart of the pumps in the subcooled and saturated conditions are dependent on the existence of liquid or two phase natural ci rcu la t ion. Generally, if natural circulation does not exist the RC pumps are " bumped" to try to start natural circulation; if N
\t/
DATE: 7-6-82 PAGE 121
BWNP-20007 (6-76) BABCOCK & WILCOX " I NUCLEAR POWEa GENERADON DIYislON e 74-i i 25 s a i-00 TECHNICAL. DOCUMENT natural circulation does start then that is a good indication that a large amount of water is in the RCS. " Bump" means to start a pump observe the start current drops off then run it for 10 seconds, then turn it off. When a RC pump is " bumped" it will cause hot reactor coolant in the vessel and hot leg to move into the steam generator; and will cause cold water in the s t ean generator to move into the reactor vessel. This will establish communication between the thermal centers and initiate natural circulation (if enough water is in the RCS). When the RCS is saturated the " bump" may or may not start circ.1ation, but it will help to depressurize the RCS by condensing reactor coolant steam in the genera to rs (if the SG tempe rature is below the RC temperature) and allow more HPI to flow into the system. The " bumps" are used only every 15 minutes because: 1) that will limit the O liquid flow out of a break and 2) it will take some time for natural circulation to develop and stabilize and 3) allow time for the motor stator to cool. Between " bumps" the development of natural circulation should be checked. Table 6 shows two columns when the RCS is saturated - one with primary to secondary heat transfer and one without. Both show that HPI is on. Natural circulation pos itively exists when the steam generator Tsat controls the incore the nnoccu pl e temperature; if Tsat is ch anged the incore th e nno-couple t emperature will follow. If the incore thermocouple t empe ra t ure does DATE: PAGE 122 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX suosin NUCLEAR POWER GENERATCN OlviStON 7'-i 2ss3 t-00 TECHNICAL DOCUMENT not change when T sat c h a nge s , the steam generators are not coupled to the reactor coolant systen. Extended saturation with the steam generators available as a heat sink can exist only because of a LOCA. For the condition with no natural circulation, the operator is directed to perform several " bumps". If after four bumps natural circulation does not start, and one hour has elapsed since the reactor trip, then one RC pump should be run for cooldown as long as at least one OTSG is available as a heat sink. Natural circulation will not start when there is not enough water in the RCS. The reasons for allowing RC pump operation in this case, (wh ich goes against other requirements that do not pe rmit RC pump operation when the subc ooled margin is lost), are that the RCS should be depressurized and placed on the decay heat removal system before the BWST runs dry (to avoid llPI recirculation from the sump and reduce RC pressure below the llPI shutoff h ead ) and that the several " bumps" have consumed an hour since reactor trip. The one hour time limit has allowed the decay heat load to drop suf ficiently so that the IIP l system is now capable of adding enough water to make up the flow out of the break and still remove all of the heat. There is no chance for insu f ficient liquid in the core when the RC pumps are run at this time when the IIPI system is working. DATE: 7-6-82 PAGE 123
BWNP-20007 (6-76) j BABCOCK & WILCOX ,vy,,, NUCLE AR POWER GENERATCN DIVI 5 ION 74-1125331-00 TECHNICAL DOCUMENT MU and HPI Control Although the Makeup System is primarily used for no rmal makeup, the makeup (MU) system can be used with the HPI system for eme r gency injection of borated water to make up for lost inventory from a small b r e ak . These systems may also be actuated for other reasons associated with abnormal t r ans ient:5. The operator will have to control the flow rate in dif ferent ways as the primary system conditions change. Flow control actions include selec t ing either HPI or MU system operation (or a combination of both systems) and regulating the flow into the RCS. The general control actions are:
- i. Maximize the injection flow for small break
- 2. Balance the flow between the HPI injection nozzles.
- 3. Throt tl e the flow to prevent runout and cavitation of the HPI pumps at low pressure
- 4. Throttle or stop the flow to prevent filling the pressurizer solid when the RCS is subcooled (except during MU/HPI cooling as described in
" Backup Cooling Methods").
- 5. Stop the HPI and MU system when the LPI system is operating
- 6. Throttle the HPI and MU to prevent exceeding the reactor vessel pressure temperature limits of Figure 25.
Each one of these topies will be addressed, and the best ways for handling it. will be de sc ribed . The discussion will be divided into two sections: Maximizing MU and llPI Flow and Throttling HPI and MU Flow. i O l PAGE 124 DATE: 7-6-82 l
i BWNP-20007 (6-76) i BABCOCK & WILCOX wuyait NUCLEAR POWit GENERADON DIVISION 74-i i 2n 3 i-00 TECNNICAL DOCUMENT Maximizing MU and HPI Flow SUBCOOLING RULE: Whenever the reactor coolant subcooling margin is lost:
- e Two makeup pumps should be run at full MU system capacity taking suction from the BWST e Two HPI pumps should be run at full HPI system capacity when 'e RC pressure is < the low RC pressure SFAS actuation setpoint.
(1650 psig). When the MU or HPI systems are started for either of these two conditions, i their purpose is to remove decay heat either by "once through cooling" or by replacing RC to ensure heat trans fer to the stean generator. "Once through cooling" occurs when the MU and/or HPI water passes through the core, picks up heat, and exits through a break. The break could be a stuck open pressurizer s a fe ty valve or PORV. "MU/HPI Cooling" as discussed earlier is also "o nc e through cooling"; however, the term "MU/HPI cooling" is used in connection with a loss of primary to secondary heat transfer where the opera-tor starts full MU , opens the PORV and verifies HPI actuation if RC pressure i , drops below the low RC pressure SFAS actuation setpoint. In order to be most e f fect ive the flow to the core must be the greatest amount pos s ib le. There-
> fore, both HPI and MU pumps are desired. HPI flow rate at the HPI injection
< g nozzles should be balanced to ensure that the greatest amount of pumped flow enters the core. However, if both the HPI and MU are on, the flow rate through the HPI injection nozzle used for normal makeup will be the sum of the HPI flow and the makeup flow. When the HPI system is automatically g actuated the operator should check flow, the flow in that line should be
,, throttled to be within 1.5 times the low flow nozzle but not below the limit of Figure 30a.
PAGE DATE: , 125
BWP-20007 (6-76) SA8 COCK & WILCOX j NUM B E R NUCttAR POWit GENitATION OtVi$10N 74-1125531-00 TECHNICAL DOCUMENT Throttling MU and HPI Flow 0 After being started the MU or HPI must be run at full capacity until the reactor coolant system conditions (given below) allow it to be terminated. Guideline 1 for HPI termination or throttling: HPI operation may be terminated if the LPI system has been started and has been flowing at a rate in excess of 1000 gpm in each injection line for 20 minutes. This condition is applicable to a large LOCA when the RCS depressurizes enough to allow the LPI to flow into the reactor vessel. Since LPI will provide emergency injection at a much greater flow rate than the HPI, HPI can be stopped. The 20-minute delay is used to make sure that the primary system will not repr es sur ize and result in a loss of LPI flow. The minimum flow requirement of 1000 gpm is used to make sure that the injection flow can remove decay heat with no loss of reactor vessel water inventory af ter HPI is stopped. 1000 gpm flow to each injection line is required to make sure at least 1000 gpm gets into the RCS. A possible break in one of the two LPI/CFr l lines would allow LPI water to be lost out the break and not reach the i reactor vessel. i l O DATE: PAGE 126 7-6-82
BUNP-20007 (6-76) BABCOCK & WILCOX NUF)Et HUCitAt power GENERAlloN DIVISION 7'-i i 2 n 3 i-00 (V~') TECHNICAL DOCUMENT Guideline 2 for MU and llPI termination or throttling: e MU ano HPI may be throttled any time the reactor coolant subcooling margin is restored. e llPI may be terminated any time the reactor coolant subcooling margin is re-stored and pr es sur ize r level is above 100 inches and increasing. Normal f g
)
t makeup should be restarted if stopped. The one exception to this guideline is the case where core cooling is provided solely by MU/HPI. In this case HPI can be throttled when the subcooling margin is restored but should not be stopped until secondary heat removal is es t ab li shed , even though the pressurizer could be solid. These guidelines appl y to both LOCA and non-LOCA transients and are intended l I / g\ l kv j' to limit the amount of water going into the RCS so that the pressurizer will not fill solid. The pressurizer can fill for two reasons:
- Continued MU and HPI injection
- Reheat and swell of the reactor coolant af ter an overcooling transient has been stopped, i /7 Although the core will be covered and safe if MU or HPI is not throttled, it
! / a 1 \ I If MU water were allowed to flow through the pressu-V' is desirable to do so. rizer valves, the plant conditions could get worse. Continued flow through the valves could fill the quench tank and cause the rupture disc to fail rele as ing water to the containment vessel, or the pre s sur izer valves could i ! fail to reclose and a LOCA would result. Also if the reactor coolant is very I
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7-6-82 PAGE 127 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX "" I " NUCLEAR POWER GENERATION DIVISION 7'- n 2 s sai-00 TECHNICAL DOCUMENT cold the repressurization caused by MU or HPI flow could cause a violation of the Reactor Vesel P-T limits. If there is any doubt about throttling then don't do it. An overcooling transient causes the recctor coolant to shrink. If MU or HPI is started additional water is added to the RCS. When the overcooling is s top ped the core heat will cause the reactor coolant and the added water to swell. It can expand enough to fill the pressurizer. In order to limit the amount of filling the MU and HPI can be throttled when the reactor coolant subcooling margin is restored and the HPI can be stopped when the subcooling margin is restored and a low pressurizer level indication of 100 inches is reached. The 100 inch pressurizer level indication was chosen because it is high enou gh above the pressurizer heater cut-off level of 40 inch e s to compensate for level instrument error due to high containment vessel l temperature so that a pressurizer water level exists for heater operation. l If the overcooling was severe, throttling MU and HPI alone may not be enough to prevent the pres sur ize r from filling; th ere for e , the reheat of reactor coolant must also be limited. This can be done by lowering the secondary systen stean s atur a t ion temperature to below the RC cold leg tempe ra tur e . The operator can monitor the effects of stean pr es sur e by monitoring Teold and pressurizer level and control steam pressure as necessary. The operator should be ca re f ul not to lower steam pressure too much or the pressurizer will drain. Throttling MU and termination of HPI and lowering steam pressure will keep the P-T from returning to the " post trip window". However, this is an acceptable end point if the system is stable. DATE: PAGE 7-6-82 128
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i BWNP-20007 (6-76) l BABCOCK & WILCOX I NUCLEAR POWER GENERATION OlVISION 74-1125531-00 , TECHICAL DOCUMENT For many reactor events that use the MU and HPI systems to return subcooling, these conditions will be restored in the firs t several minutes. When the i reactor coolant subcooling margin is established the following general procedure to control RCS inventory should be followed-l J i l l f After RC Subcooling is Regained
- 1. Avoid too much subcooling (high RCS pressure). There is a tendency to '
think that if " adequate" subcooling margin is good, then 200F subcooling must be better. There are transients such as a steam generator tube rup-ture where the higher RCS pressure makes the outcome worse (large leak 4 rates). The refo re, the operator should throttle MU and HPI and begin to
- l. stabilize RCS pressure as soon as the subcooling margin is regained.
) 2. If the pressurizer level is offscale low, maintain MU and HPI flow but i reduce the amount of flow that is being added to the RCS.
e If both HPI and MU are on reduce HPI flow first. l l 3. If pressurizer level is on scale but below 100 inches, attempt to stabi-lize pressurizer level, e Maintain MU and HPI at the reduced flow rate if pressurizer level and x subcooling margin stabilizes (i.e., HPI is matching a leak). e If pressurizer level continues to increase above 100 inches, control HPI per Item 4 below (except during HPI cooling as described in
" Backup Cooling Methods") .
s DATE: 7-6-82 PAGE 129 4
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BWP-20007 (6-76) BABCOCK & WILCox "#
- NUCLEAR POWit GENERATION DIVt510N 74- M 25 531-00 TECHNICAL DOCUMENT
- 4. If pressurizer level is increasing and greater than 100 inches (indi-cated), stop HPI and start normal makeup and letdown.
e Monitor reactor coolant subcooling. Be prepared to reinstate the sub-cooling rule if the subcooling margin is lost. e Reset SFAS and realigin the HPI system af ter HPI has been stopped (if RCS pressure is high enough). e Monitor pressurizer level. Be prepared to restart HPI if level drops below 50 inches.
- 5. If the pr es sur ize r level is increasing rapidly, it may also be necessary to open the turbine bypass valves to decrease steam pressure to prevent the RCS from going water solid.
Guideline 3 for HPI termination or throttling: e If action is necessary to balance HPI flow, throttle the high flow line to balance injection flow but not below the limit in Figure 30A. If one train is unavailable only one train remains with two flow injection paths. If one of the two flow injection paths has a higher flow the operator will not know whether the high flow rate is due to a break in the high flow line or a blockage in the other. Therefore, to as sur e adequate flow to the core, the operator should not throttle the high flow line below the flow value of Figure 30A in case the high flow rate is due to the other path being blocked. Remember, if both HPI and MU are on, the flow rate through the HPI inj ec t ion nozzel used for normal makeup will be the sum of hte HPI flow and MU flow. DATE: PA E 130 7-6-82
BWNP-20007 (6-76)- i . BABCOCK & WILCOX " l NuCttAA POWft OtNERATION OlVISION 74-n 2 ss31-00 TECMICAL DOCUMENT . I L , If MU is on and the operating train is the one with the normal MU line, then the HU flow has to be considered when balancing the injection flow. The flow . through the injection line without the normal makeup line should be equal to the sum of the HPI flow and MU flow through the other injection line. e , Guideline 4 for HPI termination or throttling e The HPI low flow limit is 35 gpm. Total pump flow should not be throttled below this limit. Pump overheating and damage can occur at very low flows. The total flow is a i combination of the recirculation flow and the injected flow. There fore , as t . long as the recirculation line to the BWST is open the HPI valve can be shut i because the recirculation flow will be greater than 35 gpm. Guideline 5 for MU and HPI termination or throttling: i- e MU and HPI should be throttled to prevent exceeding the reactor vessel pressure temperature limit of Figure 25 during "once through cooling" or "MU/HPI cooling". 5 l-During "once through cooling" or MU/HPI cooling", one RC pump should be started so that the RV downcomer temperature can be determined while the RC is subcooled , i i
-) ATE : PAGE 131 7-6-82 c , , -,- - < ---- - ,,~ ,,, _ .n,-n- ,,,,,an- . , , , , , - _ _ , , , - - , , , , . - , ,- .,,,,,an,_.,n__w__-,,_,-,
BWNP-20007 (6-76) BABCOCK & WILCOX N W 8Et Nuctern rowen otNanAflON Dm$lON TECHNICAL DOCUMENT 74-112s531-o0 Without a RC pump on and HPI on, with or without natural circulation the cold leg temperature detectors cannot measure the RV downcomer temperature because the RC loop flow and HPT flow mix downstream of the tempe ra ture detector. The ratio of the HPI flow to RC loop flow can be sub s t ant ial . Consequently, the resulting mixed temperature of the two fluids will be substantially lower than the cold leg t empe ra ture indication. The ratio of the two flows will vary with the size of break in the reactor coolant pressure boundary. The larger the break the more the HPI flow and the less the reactor coolant flow. Consequently, without a RC pump on the amount of subcooling should be mini-mized because the RC downcomer temperature cannot be determined. During "once through cooling" or "MU/HPI cooling", throttling the MU and HPI flow is the only method for gradually reducin; RCS pressure. Also, the rate of cooldown is dependent on MU and HPI flow rate. Th e r e fore , careful and con-sistent throttling of MU and HPI flow is required. Without feedwater the time to depressurize will depend on the amount of decay heat and, if a LOCA exists, on the size of the break. The MU and HPI flow will be pe r forming two functions. It will be maintaining system pressure which will be a function of the MU and HPI pump head and the choked flow out the PORV or break. It will also be removing decay heat from the core. The amoun t- of decay heat will determine the amount of MU and HPI flow needed and the flow will establish the RC pressure. Consequently, as the decay heat level decreases the flow can be throttled back which will cause the RC pres-sure to reduce. DATE: PAGE 7-6-82 132
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BWNP-20007 (6-76) SABCOCK & WILCOX II NUCLEAR POWit GENERATION DIYl560N 74- H 25 531-00 TECHNICAL DOCUMENT ! During plant cooldown a situation may occur where the RC pressure cannot be
' educed by throttling HPI flow. This can be caused by hot water flashing to steam in either of the RC hot leg 180 degree elbows (due to no flow in one or both loops) or in the pressurizer (PORV closed).
RC pump operation ( for steam in the loop) or opening the pressurizer PORV ( for stean in the pressurizer) is required to eliminate the problem. These guidelines are summarized in Figure 30b. Feedwater Control Abnormal transient operation with main or auxiliary feedwater requires 4 special attention to feedwater control. Failures can cause too much water to be added. Exces s ive main feedwater addition can fill the steam lines with water (steam lines may fail) and cause undesirable overcooling, especially if feedwater heating is lost and cold water is added to the steam generator. Excessive auxiliary feeu wa te r can have the same general effects, but it will cause a more severe cooldown (for the same flowrate) because of the greater ! O steam pressure reduction ef fect due to the colder water and where it enters. Both excessive main and auxiliary feedweter may require that quick actions be taken to .stop it. This section will recommend the best methods for manual control. Main Feedwater Overfill For very rapid MWF overfill conditions the operator should immediately trip s both MFW pumps. In the event of a slower overfill transient, attempt to i DATE: PAGE 7-6-82 33 y- 9y, .e r --- ,- ---9, - e- -,w.,--,-4,-y ,,v.-,,.,,we,e--,_,----,,-.y--,----,,---w,.e,-,,,sw-,--.vwr,- -,...,---+~c, - . - , - , . . . . . -
BWNP-20007 (6-76) BABCOCK & WILCOX HUOtAs POWit GENERATION Divi $lON 74- t i 2553 i-00 TECHNICAL DOCUMENT manually control the feedwa ter valves on the overfilling steam generator with the hand / auto station. This may not work if the controls to the valves have failed, so be prepared to quickly trip the main feedwat er pumps. This is the quickest and surest ne thod of stopping the ove r f ill . It is also the preferred nwthod if the OTSG is overfilling rapidly. This will stop all feedwater to both generators (it will be a loss of fe ed wa t e r , but since the genera to rs have a large inventory the heatup ef fects will be delayed). This action can be take n if both generators have exc es ive feedwa te r or the other actions do not work. Flow should be monitored in all cases; it will show the ef fects of corrective actirn faster than level. The corrective actions must be taken within 2-3 minutes to prevent steam generator overfill (water level at the top of the shroud) with large excessive MFW flow rates. If all main feedwater has been s top ped , the operator should make sure auxiliary feedwa t e r starts so it can start to inject when the generator water level boils down to the automatic s e t po in t . Auxiliary Feedwater Overfill To stop auxiliary feedwater from filling one stean generator: e Put the pump in manual control and try reducing the pump speed. e At tem pt to close the auxiliary FW isolation valve to high level /high flow steam generator. To stop auxiliary feedwa te r from filling both steam generators (this condi-tion may happen if power supplies are lost to the speed changer motor. DATE: 134 7-6-82
ai. , - 2-m _-a.*Aa. . - 4# m __ ,_y _ - - . - - J --4 m L- - 1s:. I BWNP-20007 (6-76) i SABCOCK & WILCOX NUCLEAR P0wte GENEGATION DIVISION 74- " 25531-00 TECHNICAL OOCUMENT e When steam generator level ic high, reduce the AFW pump speed, allow I the stean generator level to drop and restart one pump. Then increase pump speed as necesssary to maintain approrpiate level. e If no control of AFW can be obtained the pumps can be isolated and MU/HPI cooling can be attempted. This method is not desirable.
- . There fore , every attempt should be made to maintain AFW to at leas t j one genertstor, even if the operation is not steady.
I i FW Throttling (MFW and EFW) When the AFW sys t en is actuated and no SFAS level 2 signal occurs, the SG level will be controlled at 40" on the SU range. This level is below the j normal operating level so the SG level will have to boil down before the AFW flows to the SG. I i Anytime AFW is operating and SFAS level 2 actuates, the AFW automatically in- ' creases steam generator level to 93 inches on the startup range. The 93" level will permit primary coolant steam condensation during boiler-condenser cooling in case a small break LOCA has occurred. Anyti e the sub-cooling margin is lost the level should be raised to 93 inches. This will need to be done manually if an SFAS level'2 does not exist. If the subcooled i margin is regained while the level is increasing then the level increase does not need to be ;ontinued to the 93 inch level, but must be controlled at 40" ! on the s tartup range if the RC pumps are not running. s DATE: PAGE 135 i 7-6-82 i
BWi4P-20007 (6-76) BABCOCK & WILCOX NUM8 E R NUCitAt POWER GENERATION DIVI $lON 74- t i 25 5 3 i-00 TECHNICAL. DOCUMENT Steam Generator Level Rule Anytime the sub c oo li ng margin is lost, levels in the operable steam generators must be raised to 93" on the startup range using AFW flow. Exception: If the loss of subcooling margin was due to a loss of secondary steam pressure control, do not attempt to raise level in the af fected stean generator (s) until steam pressure control is regained. If the AFW system were to increase SG level above the 93" SG level at full flow an overcooling of the RCS can result because AFW flow through the upper nozzles injects water into the steam space of the generator. As the flow s pr ay s into the stean space it causes steau condensation and a reduction of steam pressure; when the level increases the inventory accumulation is a colder heat s ink than is needed to balance decay heat. The combination of the steam pressure reduction and the colder heat sink causes the overcooling. Full flow of AFW is not needed but contineous flow is. A continuous addition of FW into the steam space will cause the normal center for natrual circulation to be high in the reactor, and contineous addition will cause primary condensation. However, if natural circulation is lost ( for l l l example., due to a delay in AFW actuation or an interruption in MFW flow) the AFW flow should not be reduced until 93" on SU range is reached. Steam. Generator Level Setpoints:
- 35" on the startup range when one or more RC pumps are operating.
I PAGE l DATE: 7-6-82 136
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BWNP-20007 (6-76) . BABCOCK & WILCOX NUMBER t r i NUCLEAR POWEA GENERATION OlvillON 74-ii23531-00
- TECHNICAL DOCUMENT Steam Generator Level Setpoints:
l
- 35" on the startup range when one or more RC pumps are operating.
40 inches on the startup range with two steam generators (it may be nec es sary to raise the level higher than 40 inches if only one steam generator is working) when no RC pumps are operating.
- 93 inches on the startup range when the subcooling margin is lost or
~
SFAS level 2 actuates. I : ( 1 Use of the Incore Thermocouples The incore thennocou ple s can be used for a variety of purposes. Information about the incores is given in dif ferent chapters. The following summarizes that information:
- 1. They are used to detect core uncovering. They are the most valid ;
- indication of core cooling. If the incore thermocouples clearly indicate supe rh eated conditions, then the actions to counter Inadequate Core Cooling should be taken.
- 2. They provide. indication of natural circulation. Thot should read
[] within 10F of the incore thermocouples when the plant is subcooled
- and solid water natural circulaiton is occu rring . When the reactor coolant is saturated, the incore thermocouples do not provide as good i an indication of natural circulation.
- 3. They are the only valid indication of core outlet conditions when no l 1 circulation exists.
i
'DATE: PAGE 7-6-82
BWEP-20007 (6-76) BABCOCK & WILCOX NuusER HUCtEAR POWER GENERATON DIV1510N 74-1125531-00 TECHNICAL DOCUMENT Cooldown with One Steam Generator Out of Service Attempting to cool the plant down using one " good" steam generator can cause exc e s s ive the rmal stresses in the other " bad" steam generator if it is dry and the c ooldown rate is large. Although one steam generator can remove the decay heat and the stored heat needed to cooldown, the cooldown process will be slow because the dry steam generator shell cannot be quickly cooled. During normal cooldown the shell of each generator is cooled by liquid in the lower part and by steam in the upper part. When the shell is not cooled and the t ub es are cooled by reactor coolant, the tube s can get much colder than the shell causing them to contract relative to the shell. But because the tubesheets hold the tub es in a fixed positioa and the shell does not shrink the t ubes go into tension. If they get cold enough the tension stresses will be greater than the yield stress and they will pe rmanently stretch. If the tube s are cracked, flawed, or thinned they may fail. Consequently, limits are placed on the tube-to-shell A T. For normal cooldown this limit has been conservatively set at 100F. Iloweve r , in an emergency situation when cooldown is abs o lu t ely required the limit has been relaxed to 15CF A T, with the understanding that any transient which results in exceeding the design AT limit of 100F requires specific stress evaluation to de termine SG tube integrity. Cooldown with one generator at the highes t rate of cooldown should not be done unless it is absolutely necessary. The choices to be made prior to cooldown are: PAGE DATE.
- 7-6-82 138 )
i 1
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BWNP-20007 (6-76) i BASCOCK & WILCOX NUCitAA POwet GENERADON DIVI $lON
- 74- n 2 n a i-00
' TECMICAL DOCUMENT 4
e Stay at stable hot conditions until the generator is repaired and I returned to service. e Cooldown at a slow rate so that the tube-to-shell tempe ra ture limit does not exceed the " normal" AT of 100F, e Cooldown at a more rapid rate, but do not allow the tube-to-shell } temperature limit to exceed the " emergency"AT of 150F, The need to cooldown can be established only after a review of the plant status. There are a limited number of reasons why cooldown may be required; l these include: , - II)CA - small or intermediate break LOCAs will require cooldown so that the primary system can be depressurized. Depressurization will [ slow down or stop the leak rate. Tube leaks or ruptures especially require depressurization to stop the leakage into the steam 4 generator. l
- l. -
BWST Draining - in conjunction with LOCA, it is desirable to have the i f plant completely cooled down to avoid recirculation from the sump using the HPI system. For tube leaks which do not return water to the sump it is absolutely required to have the plant depressurized
- before the BWST drains.
- Condensate tank draining - to avoid using backup serv ic e water with poor water chemistry in the steam generator, it is desirable to have l
P DATE: PAGE 7-6-82 139
BWP-20007 (6-76) BABCOCK & WILCOX Nuusta NucttAR POWER otNtaAfaON DIVISION 7'- n 2 553 t-00 TECHNICAL DOCUMENT the pl an t on the decay heat removal system before the condensate storage tank is drained.
- Other accidents - most accidents will not require cooldown for mitigation, so the plant can be placed in hot shutdown wh ile the
" bad" stean generator i t repaired.
However, some situations, such as fires, may have left the plant so badly damaged that a decision to cool down is necessary to avoid un-known side effccts. In order to cool the plant down with one generator out of service it will be necessary to add water to t'. ; " bad" generator so the shell can be c ooled . If the generator is completely dry, the shell will cool only by heat loss through the insulation to the reactor building; the average shell cooldown rate will be low (around 3-5F per hour). Water addition to the generator will allow the shell to cool faster and the rate will depend on whether a water level can be maintained. If water can accumulate and cover the lower part of the shell, the average rate of shell cooldown will be abou t 20F/hr. But if a water level cannot be maintained and the shell is mostly cooled by steam, then the average rate of shell cooldown will be around 10F/hr. Since the rate of shell cooldown is greatest when water is in contact with it, the preferred way to add water is with the main feed wa ter system. However, the main feedwa ter flow rate must be DATE: PAGE 7-6-82 140
4-I BWNP-20007 (6-76) e BABCOCK &.WILCOX numeu ,
- wucteam powes osmaanow oivneow
! 7'- t i 2 n 31-00 : TECHNICAL DOCUMENT carefully controlled so the tubes do not "o ve rc oo l" . The cooldown l- ! rate of the plant will be limited by the cooldown rate of the shell i j and the cooldown limit is based on the tube-to-shell AT limit. When i i L loop flow exists the tube-to-shell limit can be calculated by i averaging the five shell thermocouples and subtracting the reactor f- coolant average temperature, f , However, in some cases T ave might repres,ent the average tube tempera-1 . k ture; these cases can occur if the hot leg is steam bound and no } ! circulation is occurring. If Thot is increasing but Teold is con- ;
- i. , !
4 stant or decreasing due to HPI then Teole should be used rather than j Th ot-t
- i To illustrate. a plant cooldown, two examples are given. Both of the examples assume that a tube leak has occurred. The re fore , the plant t
? must be cooled down; it cannot stay at hot condi t ions . The firs t , t example shows a tube leak with a generator that can hold pressure i but cannot 'be steamed. In this case a water level can be achieved
- to cool the lower part of the shell. The eecond example shows a i tube leak with a generator that cannot hold pressure (a failed steam i ,
I safety valve could do this). In this case a water level cannot be maintained because of constant steaming (in fact, a water level i l ! could be maintained if the main feedwater system were allowed to i j operate at high capacity, but the tubes would cool down extremely ' i fast .and the tube-to-shell temperature limit would be violated). The examples are illustrated in Figure 31a and 31b. 4 h j [ I DATE: 7-6-82 PAGE
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BWNP-20007 (6-76) BABCOCK & WILCOX
- * " * * " " " * " " **" 74-1125531-00 TECHNICAL DOCUMENT Both examples follow the recommended procedure for tube leaks. That is, the plant is runback, depressurized and cooled rapidly from 550F (Tave) to 500F (Tave) at a rate 240F/hr. After that the RCS is cooled down and depressurized at % 100F/hr until the " emergency" tube-to-shell limit of 150F is reached. At that time the cooldown is slowed and follows the cooldown rate of the shell and the tube-to-shell A T is the controlling limit. When the plant first reaches the " emergency" tube-to-shell temperature limit of 150F, the RCS pressure will be around 400 to 450 psig and the tube leak cate will be lowered.
The procedure shown in Figure 31a with a generator that can hold pres-sure, is to add water to the generato - ' when the f irs t stage of cooldown is comple t ed . (i. e. , at 500F RCS Tave). A water level should be main-tained at the appropriate level so t he. Icwer portion of the shell is cooled by wa t e r. Stean will be created in the generator and will help cool the upper - shell . Main feedwater is preferred, but AFW will be adequate; both must be centrolled to prevent overcooling. The cooldown after 150F emergency tube-to-shell limit is reached will be about 20F/hr.
,The proc edure to ha used when steam pressure cannot be maintained is O
shown in Figure 31h. The RCS should be cooled down at 100F/hr while slowly adding main feedwater (if pos s ib le ) or auxiliary fe edwa t e r until the "eme rgency" tube-to-shell tempe rature limit is reached. [SFRCS will need to be overridden in order to . add FW to the dry SG and the operator DATE: 142 7-6-82
a BWNP-20007 (6-76)
- BABCOCK & WILCOX NUMBER NUCLEAR POWPR GENERAflON DIVl560N 1 74- i2n31-00 TECH 111 CAL 00CUMEllT will have to check for a steam leak into the containment vessel. Also, wheneve r feedwater is added to a hot depressurized SG during a transient, the transient will require a s pecific stress evaluation to de te rmi ne SC tub e integrity.] When that limit is reached the cooldown a
rate should be slowed so as not to exceed the emergency tube-to-shell I 1 temperature limit. The rate of feedwater flow should be around 100 gpm, but actual flow rate will be dictated by the circumstances. A contin-uous low flow rate is desired rather than an interrupted " batch" feeding i rate. Main feedwater will be dif ficult to control at this low flow rate and it may be necessary to use the " bypass" flow valve around the control valves. N For the case when the SG can hold pressure but cannot be steamed, if the reactor coolant pumps are not operating or have been shut off sometime during the cooldown sequence, natural circulation will not occur in the loop with the generator out of service. If the reactor coolant in that loop is at a high temperature when the RCS depressurization begins, it may flash to steam. The steam will collect in the candy cane and that loop will become a surge volume. The system pressure will " hang up" at that pressure, preventing further depressurization. G In some cases not much can be done to prevent this except slowing the rate of cooldown. If possible, one or more RC pumps should be run, at least periodically, to preclude stean voids - collecting in the idle loop. If steam does form, and any reactor coolant pump can be " bumped", it will help to mix
} the fluid so cooldown can continue. If a reactor coolant pump cannot be J
i I DATE: PAGE 7-b-52 143
BWNP-20007 (6-76) BABCOCK & WILCOX NUCitAR POWER GENERAflON OlVISION 74-1 2 n31-00 TECHNICAL DOCUMENT started, then an alternate method to stimulate circulation and cool the stagnant water can be obtained by spraying AFW on the tubes. If neither AFW nor RC pumps can be used the cooldown rate will have to be slowed. These actions should be performed only for situations, such as tube ruptures, where it is necessary to continue the cooldown. If continued c ooldown isn't required, the plant should be held stable with continued natural circulation in the one loop to remove decay heat unt il the voids in the idle loop are collapsed due to losses to ambient and/or HPI or unt il the idle steam generator is returned to service (i.e., the prob lem requiring single steam generator cooldown is corrected). Use of High Point Vents During Inadequate Core Cooling Following a lose of coolant accident (LOCA), it is neces s ary to remove the decay heat from the core to prevent cladding damage. Core heat re-moval is ac c anpli shed by supplying cooling water to the core. If the supply of cooling water to the core is decreased or interrupted, a lower mixture level in the core will result. If the uppe r region of the core uncovers, cooling for these regions will be by heat transfer to super-heat steau, wh ich is a " po o r" heat transfer process. Cont inued opera-tion in this mode may result in elevated fuel cladding temperatures with subsequent core damage and possible hydrogen generation due to Zr - H 2 O rea c t ion . O DATE: 144 7-6-82
BWNP-20007 (6-76) BASCOCK & WILCOX NUMBER Nuctema powea oemenATON OlVISION 74-ll2s 53i-00 TECHICAL DOCUMENT During inadequate core cooling, significant hydrogen generation due to metal-water reaction begins when a cladding temperature of 1800F is at-tained. The re fore , if the operator even has indications that the fuel cladding temperature is at or above 1400F (Figure 29), he should open the high point vents at the hot leg and at the pressurizer. This is a precau t ionary action to prevent noncondens ible gases, which are being formed in the core, from accumulating within the steam generator tubes. The s tean generators are expected to be utilized within the inadequa te core cooling procedure in order to depressurize the primary system allowing sub sequent actuation of the core flooding and/or low pressore injection systems. There fore , concentration of noncondensible gases within the steam generator tubes should be minimized in order not to degrade the steam generator heat removal process. When the subcooled margin is regained or the RCS depressurizes to the point where the LPI system can be operated, then noncondensible gas pro-duction due to core damage has ceased. The operator can then close the high point vents and continue a " normal" small break cooldown. Loss of LPI Recirculation Ability E ! The ability to recirculate water from the reactor building emergency sump can be lost during recirculation or when trying to start recirc-ulation. If this ability is lost, the cause should be determined and v DATE: PAGE 7-6-82 145
i BUNP-20007 (6-76) BABCOCK & WILCOX HUCLEAR POWER GENERAIBON DIV1510N 74- u 25 sa i-00 TECHNICAL DOCUMENT flow started as soon as possible. There are two general causes, 1) loss O of sump water, and 2) loss of both flow paths from the sump. Loss of sump water can occur because of: A. A steam generator tube rupture - A steam generator tube rupcure is a LOCA; however, the water which leaks from the RCS does not accumu-late in the reactor building sump. Ins te ad , it leaks out of the reactor building via the secondary system. When the BWST is depleted and t ime has come to transfer to the containment vessel sump, there will be no sump water. This transient must be mitigated as explained elsewhere in these guidelines to prevent depleating the BWST. B. Diluted sump water - Diluted sump water caused by a leak from a non-borated water source such as the service water system or the feedwater sys t en will make the sump water unusable if the boron concentration becomes too low. The diluted water should not be added to the reactor core because it could allow the reactor to go critical. This would make the core cooling problem worse. This problem must be corrected by adding borated water to the recirculation flow and by stopping the dilution process, e.g., if a feedwater line is adding water to the reactor building, the flow to the bad generator should be stopped. Both sump flow paths can be lost if both the contairraent vessel emer-gency aump inlets become clogged or if both the sump valves fail to DATE: PAGE 146 7-6-82
BWNP-20007 (6-76) i BABCOCK & WILCOX 'I" NUCLEAR POwea GENtAATION OtVISION 7'-ii25531-00 s TECHNICAL DOCUMENT i open. If both of the sump inlets becomes clogged, it may be possitle to back flush the line to clear away any debris that might be cove ring at l least one of the sump line inlets. If both the sump motor opera ted valves fail to operate remotely, then local manual operation of the valves should be attempted to open at least one of the valves. However, (v the problem may not be electrical, e.g., the valve stem or disk may be binding, and manual operation of the hand wheel may not work, but some kind of mechanical leverage may be able to open the valves. Local at-tempts to open these valves may not be possible because of high radia-tion levels. If the cause of a loss of sump water or flow path cannot be corrected, ( . the ope rator should attempt to cool the reactor core with the DHRS. This method of cooling will be succes s ful if the cooling water can flow through the RCS without leaking to the reactor building. 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 reactor building, and 3) high enough to prevent a vortex formation as water is drawn into the DHRS suction pipe (this would be h the s ane elevation as required for normal DHRS operation when draining the RCS). The RCS water is brought subcooled by increasing the LPI cooling as much as possible. When the RCS pressure drops below the DHRS design pressure, DHRS operation can be initiated and the RCS water level vill drop to the break location. The operator should try to start the DHRS even if the RCS is saturated. This is contrary to instructions t DATE: 7-6-82 147
BWNP-20007 (6-76) BABCOCK & WILCOX Nuy Ee NUCLEAR POWER GENERAllON DIVISION 74-1 i 2 s531-o0 TECHNICAL DOCUMENT given in Appendix F for smell break cooldown. Appendix F states normal DHRS cannot be started when the RCS is saturated because the DHRS pumps may cavitate. The difference is because plant conditions are not the same. In Appendix F, LPI recirculation is assumed ava ilab le . However, in this scenario LPI recirculation is not available causing a more severe situation wh ich allows more severe actions. If the the break location is high enough, vortex formation in the DHRS drop line will not occur. However, if the break elevation is too low, then a vortex will form in the DHRS letdown line and LPI cooling will need to be continued. Vortex fomation can be detected by noting LPI pump cavitation and should be continually monitored if RCS water level is unknown especially when initiating the DHRS with a saturated RCS. Because of the possibility of vortex formation the transition from LPI to DHRS cooling should be made cautiously by switching one LPI train at l a time from LPI to DHRS mode of operation and watch for indications of pump cavitation e.g. pump vibrations and low flow. l During LPI cooling, water will continually be lost out the break until the water level in the reactor buil' ding increases to the elevation equal l to the water level required in the RCS for DHRS ope ra t ion . This will require more water than that contained in the BWST. The re fo r e , sources O DATE: PAGE 7-6-82 l
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWN GENNATION DIVISION 74- 25531-00 TECHICAL DOCUMENT of borated water in addition to the BWST water will need to be used for core cooling. These additional sources will need to be pumped to the i 4 reactor vesssl for core cooling until the sump recirculation can be es-4 tablished or the RB is flooded to the RCS level needed for DHRS opera-i tion so that the DHRS can be put into operation. If the containment ( vessel needs to be flooded, the operator should prepare for equipment and instrument failures due to water submergence. For example, the DHRS suction line valves should be opened before the containment vessel water level is increased. The increased and hydraulic pressure of the increased water level could cause leakage into the sealed DHRS valve 4 cavity. Af ter they are submerged they may not operate. Water may need to be added periodically to makeup for leakage from the RCS and DHRS. i During DHRS operation the RCS water temperature should be held corstant I to prevent volumetric water changes which cause the RCS water level to fluctuate and the RCS water temperature should be held below the boiling point to prevent water loss by steam flowing out the break, t N i 4 i s DATE: PAGE 149 7-6-82
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- RC PUMPS ON OR OFF (PROBABLY ON)
STEAM GENERATOR "lSOLATE0" BUT HAS INVENTORY OF WATER ADDED AFTER THE INITIAL DEPRESSURIZATION TO 500 F. 74'-1125531-00 "
T Figure 31b C00LDOWN ON ONE STEAM GENERATOR (STEAM PRESSURE NOT CONTROLLED) 2000
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- STEAM LEAK IN GENERATOR; STEAM PRESSURE IS AMBIENT SLOW FEEDING OF GENERATOR BEGINS AT ~ 500F - IT IS DOUBTFUL THAT LEVEL WILL LJILO FOR SEVERAL HOURS: LOWER SHELL WILL NOT BE COVERED FOR SOME TIME. SHELL COOLING IS BY STEAM CONDENSATION 74-1I255:11 00 i
+
1 Table 5 RULES FOR RC PUMP TRIPS , RULE REASON
- 1. The RC pumps shall be tripped Precludes the potential for un-immediately whenever subcooling covering the core (ICC) during a margin is lost. small loss of coolant accident due to a pump trip when the amount of water in th? RCS is low.
- 2. *If component cooling water to Pump trip precludes motor failure the RC pump motor is lost and and minimizes the chance of a fire the pumps are running, the RC inside containment due to lack of pumps must be tripped if cool- cooling water to the RC pump ing water is not restored motors.
within 4.5 minutes or the bearing temperature reaches 185F or stator temperature reaches 311F.
- 3. *If seal injection and component Pump trip reduces damage to the cooling water are lost to a RC pump seals and the chance of a pump for a period of 90 seconds, LOCA. Injection and/or CCW can the pump (s) must be tripped in- be reinitiated by following pump mediately and the seal return line manufacturer's instructions. If closed within 90 seconds. possible, an engineering assess-ment should be performed before restarting since seal failure could occur due to a high tem-perature in the seal cavity.
- These rules do not apply if the pumps were not tripped immediately af ter the subcooling margin was lost. The operator should try to restore co_oling water.
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BWNP-20007 (6-76) SABCOCK & WILCOX NUM8tt , NUCitAA POWEA OtNWAflON DIVISION TECNNICAL DOCUMENT 74-1125531-00 ! CHAPTER F l POST TRANSIENT STABILITY DETERMINATION i h I' To determine if the transient has been brought under control four general 4 areas must be checked.
- 1. Reactivity Control - The reactor must have a suberitical margin 1
of at leest 1% Ak/k.
- 2. Core Heat Removal Control 5. 3re must be covered and cooled; f the heat removal rate is equal _ -
slightly greater than the core heat generation rate. l
- 3. Radiation Release Control - Release to of fsite is terminated.
- 4. Plant Equipment - is Operating Correctly - Equipment to maintain
( the plant safe and stable is operating and within design duty; equipment failures have been bypassed, isolated or repaired. ! Several things around the plant must be checked to make sure these four gene-ral rules are being met. The following basic check list defines the more im-portant items. The list is divided into two cases. Case I applies to LOCAs which can be stopped by complete isolation of the leak and to all other trans-ients. Case II applies to LOCAs which cannot be isolated. The dif ference i between the two parts is simple: a reactor leak that cannot be stopped is a transient that cannot be positively te rmina ted . However, a leak can be re-f duced to the smallest amount possible and become stable for "long-term cooling". Stean generator heat removal can be used for some small leaks but i DATE: PAGE 150 ! 7-6-82 l~ .. __._ _ __ _ _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ . - - _ _
e- w .+ a a s - - - - - , - g l 1 l O l l ( O THIS PAGE INTENTIONALLY LEFT BLANK P O O i i i O
BWNP-20007 (6-76) BABCOCK & WILCOX " NUCttAR POWEA GENERATION DIVISIDN 7'-ti23sai-00
' TECHNICAL DOCUMENT '
e MU/HPI must be kept running to keep the reactor coolant s ubc oo led . S ubcoo l-ing can be regained fo r some very small break sizes at a time when the decay heat decreases and MU/HPI is able to refill the RCS loops and add water to the pressurize r. 6 Case I - All transients (including LOCA'a which can be isolated)
- 1. Reactor coolant pressure and temperature are preferably within the "po s t-t r ip window" of the P-T diagram; however, pressure and tempe rature may be anywhere on the P-T diagram within a region bounded by: a) Reactor Vessel P-T limits, b) the subcooling margin, c) an RC pressure upper limit equal to the PORV se t point minus 100 psi, d) fuel pin compression limits and e) RCP NPSH requirements, if i
- \ applicable. Subcooling will exist in the hot and cold legs of both loops.
l 2. The "long term" trend of reactor coolant pressure and temperature is 4 constant or slowly decreasing with time. "Short-term" fluctuations of t empe rature and pressure are small and can be attributed to periodic operations of other equipment ( pres surizer heaters, spray, or feedwater).
- 3. Pressurizer level is within the indicated range.
- 4. If forced circulation exists (RC pumps on) then reactor coolant T ave 1
is about equal to the saturation temperature of the water in the stean generator (or generators) that is removing the heat. [ 4 T i . DATE: PAGE 151 6-82
. . - - . . ._. , . . - - , - - . . . - - - - _ _ - - - . _ _ - . - - ~ . - . . . - - . . - - - - .
BWNP-20007 (6-76) BABCOCK & WILCOX N BER NUCLEAR POWER GENftATON DIVISION TECHNICAL DOCUMENT 74-1125531-00
- 5. If natural circulation exists, Tcold leg will be about equal to the 0
saturation temperature of the water in the steam generator (or gene-rators) that is removing the heat. The di f ference be tween incore thermoc ou pl e s and Thot in the operating loop (or loops) will track wi thin 10F. If only one generator is removing heat the other reac-tor loop is subcooled.
- 6. Stean generator (or generators) level will be at the correct set-point (either natural or forced circulation setpoint) and will be s t e ady.
- 7. Stean generator pressure is steady and is below the safety valve opening setpoint.
- 8. The core is at least 1% Ak/k subc rit ic al on rods and boron. If more than one rod did not fully insert the core is at 1% Ak/k sub-critical on bor on alone (assuming all rods out).
- 9. If the accident caused water to enter the containment vessel (reac-tor or stean generator) and the contaiment vessel environment was increased, it will now be ceduced to near normal levels. Pressure will be close to atmospheric pressure; average containment vessel tempe rature will be near prior operating temperature; relative humidity will be about 100%.
- 10. If rad ioac t ive water leaks occu rred in auxiliary building those areas will be setled and the spillage either trapped or drained to storage tanks.
O DATE: PAGE 7-6-82 l
i i BWNP-20007 (6-76) SABCOCK & WILCOX "M" wucteAn Powea oewenAnow omsion
- TECHNICAL DOCUMENT 74-t i 25 sai-00 4
- 11. The component failure (or failures) wh ich caused the transient is known. It has been bypassed, isolated, repaired, or other-
! wise handled so that it no longer compromises plant safety.
- 12. Components which support plant safety are operating within their design limits (examples: pumps are operating away from the minimum shutof f flow and have adequate NPSil, throt tle valves are near the proper opening, ciectric motors are in the normal service range, electronic equipment is environmentally r
j protected). If a component is operat ing off design and future f ailure is pos s ible , then redundant or alternate equipment is on standby and ready to replace the equipment which might fail.
- 13. Stored wa ter (c onde ns a te storage tank, BWST) is adequate for long term use or alternates are readily available.
i b
- 14. Instrumentation to monitor plant pe rfo rmanc e is ope ra t i .sg cor-rectly. Potential failures of critical instrumentation have been identified and alternate instrumentation is available.
Case II - LOCAs which cannot be isolated NOTE: With the exception of steam generator tube leaks, all reactor i _ coolant leaks outside the reactor building can be isolated. Al-though a tub e leak is " ins id e" the containnent vessel a direct I path outside the containment vessel exists through the steam l lines, a i
, DATd; 7-6-82 PAGE 153 ~ . . ,. _ _ _ _ ._.. _ _ _ _... _ _.__ _ __ _ _ _ _ . . _ _ _ - . . _ _ _ . , . _ . , ~ . - . . . . . - .. .
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLEAR POWER GENERATION DIVI $lON TECHNICAL DOCUMENT 74-112s331-00 Many of the criteria of Case I app y to this part except that the reac-l 0 tor coolant will not always regain the subcoo led margin and operating conditions that depend on subcooling will not apply. The very smallest reactor coolant leaks may allow the reactor coolant system to repres-surize (because of continued MU and High Pressure Injection) and some amount of sub c ooli ng may be regained, but it is not likely that the subcooling marg in can be reinstated Consequently, the criteria for LOCA stability does not include the subcooling margin. Also because subcool-ing may not exist the hot legs may have steam binding and natural circu-lation may not exist; therefore, the criteria do not include natural circulation requirements (however, it can exist for very small breaks and could be ch ecked ) . A reactor-steam generator heat trans fer balance cannot usually be accomplished because of saturated (or near saturated) conditions wh ich may not permit the reactor coolant to move the heat from the core to the s t e ata generator, but some heat transfer to the s t ean ge ne rato r is po s s ib le for small b re ak s . The steam generator i operating level should be at the 93 inch startup level for small breaks l to permit condensation of primary side steam. Pre s sur iz er level cannot be relied upon if saturation exists. l The most impo r t ant criterion for LOCA is to keep the core covered. This condition is confirmed by readings of the incore thermocouples and the hot leg RTD's; both should show that the reactor coolant is saturated (or even subcooled) but not superheated. O 7-6-82 154
BWNP-20007 (6-76) 1 BABCOCK & WILCOX NucttAt POwta GENtaATION Olvi$ ION 74-i i 23 s31-Oo TECHNICAL ' DOCUMENT The continued loss of coolant from a LOCA will not permit the accident 1 to be truly te rmi na t ed , but the leak rate can be minimized. Lowering i RCS pressure is the best way to lower the leak rate. This can be done by loss through the leak, by opening the PORV, or by lowering secondary side pressure. Long term loss of coolant when the RCS is depressurized occurs in two ways: 1) steaming out of the leak because of continued s 4 boiling, and 2) water loss because the head of water is above the break and water will "run" out of it. The rate of leak age will depend on the system pressure, the decay heat level (which causes boiling), and the elevation of the leak (a leak high in the system will have a lower flow i rate than a leak low in the system). The leak rate will also depend on m the hole size. The criteria fo r stability is that the leak rate is as low as pos sible and that the flow into the core keeps it covered. It may take a very long t ime to recover from some LOCA's and during that time there will be two general stages when the leak rate diminishes. The first s tage is when the reactor coolant system is depressurized to containment pressure (big breaks will depressurize rapidly, smaller breaks will take longer); 4 d the second stage is when the core heat drops so that it cannot boil the water in the reactor vessel. Steaming will stop at that time (which may be as long as several months after the accident). Unt il the water in the vessel becomes subcooled (incore thermocouples will read less than ' S 212F), the plant must' be operated by injecting containment sump water (a in the recirculation mode or by continuing to inject fresh borated water DATE: 7-6-82 155
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLEAR POWER GENERAisON Divi $lON 74- 123331-00 TECHNICAL DOCUMENT from other sources. When the vessel water becomes subcooled the opera-O tor has the option to t rans fe r one train of LPI to the decay heat re-moval mode and keeping the other train on sump recirculation. The reaso n one train is left on recirculation is that it will keep water above the hot leg suction for decay heat removal. Decay heat removal has the adva nt age of rapid RCS cooldown, but it must be carefully moni-tored to make sure the decay heat pump does not lose suction (or it will fail). Because the leak utay continue a long time until the decay heat system is placed in service, an arbitrary definition of stability is given. The following criteria define post-LOCA long term stability:
- 1. The core is covered. Incore thermocouple readings show saturated or subcooled reactor coolant.
- 2. ECCR inj ection is in the "long term cooling" mode . Long term cooling exists when the ECCS is operating with recirculation from the containment vessel emergency sump. (NOTE: A deci-sion may have been made not to transfer but to bring in back-up water to re fill the BWST. Nevertheless, if recirculation could have been started, "long term cooling" is considered to have s tarted) .
- 3. The reactor coolant system is depressurized to near atmospher-
. ic pressure so that the leak rate is as low as pos s ible. The LPI system is used to cool the core. (NOTE: If the break size did not permit depressurization be fore the BWST was DATE: PAGE 7-6-82 136
__ . __ __._. ________ _ =_. . . _ . _. _ ___ __ BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen NUCttAa POWts GENisATON DIVI $lON TECHICAL DOCUMEllT 74-1125531-00 i empty, and llPI " piggyback" recirculation had to be used while further depressurization took place the plant is not considered to be stable until the pressure and leak rate are ; as low as possible).
- 4. Stean generator level is at 93" on the startup range and is steady.
- 5. Reactor coolant pumps are off (operation of RC pumps could move water past the break and increase the leak rate).
- 6. The following criteria from the previous part also apply
Numbers 7, 8, 9, 10, 11, 12, 13, 14. l
- 7. For the special case of stean generator tube leaks (LOCA's):
! a) Feedwater (main and auxiliary) has been stopped to the bad generator. b) Stean created by boiling the RCS leakage is directed to the condenser (if it is operating) when SG 1evel reaches 95% on operate range. c) The plant is on decay heat removal or standby backup borated water sources are available to replenish BWST inventory , f i i i b 4
.DATE: .
PAGE 7-6-82 157
k BWNP-20007 (6-76) SABCOCK & WILCOX NUMttR NUCLEAa POwte GENERAnoN OfVISION 74-t i 2 s s31-00 <! TECMICAL DOCUMENT ' t CHAPTER H i USE OF THE GUIDELINES
-Philosophy of Part I Organization Part I is designed for use following any reactor trip or forced shutdown.
Its primary purpose is to maintain core cooling and ensure plant stability. A reactor trip, depending on the cause and initial plant conditions, can result in demands on various systems and components (MSSV's, TBV's, AFW, etc.). These demands, coupled with the cause of the trip or forced shutd own , . are occ urrenc es that have a higher probability for abnormal conditions to
! develop.
i When equipment or system failures occur resulting in an abnormal plant N res ponse following a trip, it is not so important to immediately identify the cause as it is to restore stable, controlled conditions. Once the plant has ) i i been s t abilized , then time exists for failure identification and the decision f for future operations (i.e., return to power, remain at existing conditions, or begin controlled cooldown). The main thrust of ATOG and certainly the i most impo rtant as pect of dealing with any transient or accident is to maintain adequate core cooling. The most expeditious and positive approach N to accomplish this objective is to recognize abnormal conditions when they develop and take appropriate actions to restore stability. i I
~
Part I of ATOG contains four basic sections that flow in a logical sequence based on this philosophy. The four sections, in order, are: f I O DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NuusEn NUCLEAR POWER GENERAtlON DIYlSloN 74-1125531-00 TEPHNICAL DOCUMENT
- a. Section 1: Immediate Actions
- b. Section II: Vital System Status Verification
- c. Section III ( A-D): Follow-up Actions for Key Symptoms
- d. Cooldown Procedures (including Inadequate Core Cooling)
A two-level fo rma t is used in Section III and in the cooldown proced ures (including inadequate core cooling). The main steps or headings are de-signed to provide suf ficient information fo r the supe rvi so r or experi-enced operator. The sub-steps provide additional detailed information for the operator actually performing the actions as well as a step by step proced ure for the less experienced operator. This approach is de-s igned to allow the more experienced operator to follow the procedure more quickly by not requiring him to read detailed in format ion he is al ready familiar with. The flowcharts for Section III are provided as a visual aid to supplement the text. They do not replace the text, since the detailed info rma t ion of the sub-s t ep s is not included in the flow-charts. Only the main steps are represented on the flowcharts. The immediate actions in Section I are those actions taken af ter every reactor trip regardless of cause (e.g., manually tripping the reactor and turbine). Once the immediate actions have been performed, the next priority is to verify that the reactor is shutdown and that key systems and equipment are available and functioning properly (e.g., NNI/ICS power, turbine stop valves shut, e tc . ) . These items are included in O DATE: 7-6-82 PAGE 220
' - BWNP-20007 (6-76) s i
BABCOCK & WILCCX I sv su
- wucteam rowee oeweemow omsca -
74-1125531-00 TECHNICAL DOCUMENT - s ( Section II. g 0ne abnormal transient '(excessive MFW), should it occur, can require prompt recognition and respoase by the opera' tor to prevent the possi-
' bly severe co'nsequences of water spillover into the steam lines. There fo re ,
one of the first checks in Section II is the verification that MFW flow has runback. The important items to ch.ck early are reactor /r_rbine trip and MFW flow status. No particular importance should . be placed on the sequencing of V i the other verificatien steps as'long a t. all of them are performed except for the verification'of
- the four main symptoms discussed below.
Section II also includes obtaining a status of the four main symptoms of abnormal transients lie three basic heat trans fe r symptoms and the special
'T '
case for SG; tpbe rupture). Hodever,. operator training should emphasize
~
l c ontinuou s surveillance, for indicatione of of f-normal condi t ions . Four main
~
t . I symptoms ar'e: i
- 1. Lack of adequi.te subcooled margin
- 2. Lack of primary tc secondary hesc transfer (overheating)
- 3. Excessive primary to second1ry heat transfer (overcooling)
- 4. Indications of steam generhtm tube rupture 1
, 1 i
N Recognition of these symptoms is covered' in more detail in the "P-T Diagram" and "Abnornial Transient ,Diagnoeis a;:d 'ititigation" chapters of this volume. Should any of these symptoms occur, Section II relerences the operator to the t l appropriate folloa.' up action' in Section III of Part 1. l i s x
\ Philosophy on the Use of Section III s The order that the symptoms areE list'ed in Section II corres po nd s with the w ,s. ,
order of the' respective follow-up actions in Section III and is based on the DATE: " 7-6-82
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BWi4P-20007 (6-76) BABCOCK & WILCOX wumsen l WJCLEAR POWER GENERATION DIVI $10N 7'- 23331-00 TECHNICAL DOCUMENT relative priorities of the corrective actions. Lack of adequate subcooling margin is of top priority because core cooling cannot be assured until certain actions are performed. If a lack of adequate subcooling margin occurs, tripping the RC pumps and initiating full MU/HPI flow must be pe r-formed quickly. While ICC is the most severe condition of the RCS, it fol-tows after a lack of adeq ua te subcooling margin and the ope rato r would be directed to the ICC section if it should occur. However, if the actions in the lack of adequate subccoling margin can be performed, ICC will not occur. Lack of heat transfer and excessive heat trans fer are both second priority symptoms. These two symptoms are equal in priority because they are mutually exclusive conditions. However, excessive heat trans fe r will require the quickest resporse by the operators. The SG Tube Rupture is the last priority symptom. The order of the priority for these symptoms can be understood by focu s ing on the obje ctive s for treating any abnormal transient, which is to maintain adequate core cooling and minimize radiation release. For example, even though a SGTR occurs which can release large amounts of radiation if it is not quickly treated, the lack of adequate subcooling margin and the lack I i of heat t rans fe r take precedence. This is true because, if adequate core cooling is not maintained, the radiation release would be much greater. O Again, even though it is important to diagnose and treat a steam generator tube rupture as soon as possible, termination of a conc ur ren
- overcooling transi.ent t ake s precedence. This is true because:
O DATE: PAGE 222 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX " NVCLEAR POWER GENERADON DWi$lON [ 's TECHNICAL DOCUMENT 74-1125531- @ r
\v/ 1. The overcooling could be the result of a steam leak to atmosphere on the stean generator with the tub e rupture, resulting in higher of f-site releases.
- 2. The ove rc ooli ng increases the tensile stresses on the steam genera-7 tor tubes which could result in a larger leak size.
/ \ kv ) 3. The contraction of the RCS liquid volume due to the overcooling, especially when compounded by the inventory loss through the tube leak, could result in draining the pressurizer and saturation of the RC loops. Subsequer.t voiding in the loops can sigrificantly delay the cooldown and thus lead to increased of f site releases. This discussion of priorities is g',ven to show the logic behind the develop- [q v
\
I Lv / ment of Part I. The important point to remember is that the operator should always be alert for the presence of these symptoms and should always proceed directly to the appropriate section for follow-up act ions for a symptom with-out necessarily waiting to see if the " higher priority" symptom develops. This constant operator sutveillance should continue during and after the stabilization of a transient. Some symptoms can mask the presence of others. , '^'] For example, an overcooling transient can mask the presence of a small LOCA. x /
'w/ However, once the overcooling transient is terminated, the small LOCA should quickly become evident. Abnormal transients can occur at any time and will not always oblige by beginning immediately after a reactor trip. There fore ,
continuous surveillance is warranted. [ One symptom, lack of adequate subcooling margin, always requires immediate
\ /
(j attention. The operator must trip reactor coolant pumps and initiate MU/HP1 DATE. PAGE 223
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCitAR POWit GENERATION OlVISION 7'-i l 2 n 3 i-00 O TECHNICAL. DOCUMENT flow immediately whenever the subcooled margin is lost regardless of wh ich part of Section III he is currently following. For this reason, the other parts of Section III will either a) reference the o pera tor to Section III.A or b) reiterate the required actions given in Section III.A for loss of subcooling margin. It can be seen that these symptoms can occur in various combinations and vir-O tually at any t ime during a transient or aubsequent cooldown. Thus it would be very difficult to write a comprehensvive procedure that will lead the operat,r through any possible sequence of events. Where it is known that multiple symptoms ate more likely to occur the guidelines will specifically address the possibility. However, to ensure maximum cover age this approach is su ppl eme nt ed by a combination of operator training and procedural guide-lines, Recognition of the four basic symptoms snd following the appropriate actions of Section III will provide plant stabilization for all single events regardless of cause. Operator training is the key in the recognition and understanding of the four basic symptoms. When events become more complex, either due to additional failures or due to an event progressing to the point of inducing other symptoms (e.g., loss of subcooling margin), operator training is again the key in successful mitigation. The procedural guide-lines contain all the necessary information, but the operator must know when to implement the appropriate sections. He does this by careful surveillance of the pla nt conditions during and following a transient, recognizing the symptoms whenever they occur and going to the appropriate section. In this O DATE: PAGE 224 7-6-82
BWNP-20007 (6-76) L BABCOCX & WILCOX Numten NUCitAt POWit GtNetAilON DIV1510N 74- n 2353 t-00 ( TECMICAL DOCUMElli t respect ATOG is somewhat similar to event-oriented procedures in that the l l operator must recognize a condition and react to it. The di f ference is that the symptom will always be evident when the failure occurs. The cause 4 (event) will not always be evident. Also, there are just four symptoms to recognize as opposed to numerous events, t I j- In addition to recognizing a symptom and implementing the appropriate part of Section III, the operato,r must also know when to trans fe r between parts of Section III ( for multiple symptoms or incorrect diagnosis) and when to , t e nninate his actions if the problem is corrected. Some basic instructions 3 for the use of Part I can be summarized as follows:
- l. Priorities between the parts of Section III:
{
- a. Loss of subcooling margin always requires immediate atten-
- tion regardless of which part of Section III is being followed.
I b. Lack of heat t rans fe r or excessive heat trans fer must be cor-rected before, or at least concurrently, with actions for SGTR.
- c. Exces sive heat transfer (overcooling transients) must always be l terminated as soon as possible, s
- 2. Follow the appropriate part of Section III for the dominant symptom, unless a " higher priority" symptom (in item 1 above) appears, in which case recycle to the part of Section III fo r the higher priority symptom.
- 3. If a reactor trip occu rs during a forc ed shutd own , recycle to i
! Section 1. L DATE: 7-6-82 PAGE 225
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sta NUCtf AR POWit GINttAtlON Divi $lON 7'- 125s31-oo TECHNICAL DOCUMENT
- 4. If a major change in equipment status occurs during the performance of a part of Section III or subsequent cooldown, carry out the appro-priate actions of Section II (i.e., loss of NNI/ ICS' power, loss of offaite powe r , safeguards actuation, etc.). This can be accom-plished in parallel with Section III.
- 3. If it is discovered that an incorrect diagnosis has been made and:
- a. no other symptom exists (e.g., inadvertent entry into Section III because an overshoot after trip misinterpreted as over-cooling), then stabilize plant (e.g., restore heat transfer; don't exit III.C immediately after isolating both SG's) and recycle to Section II; or
- b. a dif ferent symptom exists, then recycle to the apy. apriate part of Section III.
- 6. If, during the performance of follow-up act ions in Section III, the cause of the transient becomes evident and is corrected, then hold at that point and allow the plant to stabilize while checking fo r other symptoms / problems. Similarly, if the intent of a group of actions is satisfied, then continuation of those actions may not be necesary (e.g., HPI flow can be throttled and SG levels do not need to be raised to the high level once subcooling margin is restored).
- 7. All normal limits and precautions are applicable during the perform-ance of Part I unless specifically superceded by the ATOG procedure (e.g., the use of pump bumps regardless of NPSH requirements when s atur a t ed with SG level). Whenever a step appe ars in Part I that O
DATE: PA E 226 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUM8 t t NUCtf At POWER GENERATION Divi $10N
^
7'- u 2 5331-00 ( 'jTECHNICAL DOCUMENT
\ /
u-supersedes a normal limit or precaution, it has been carefully con-s id e red and deened acceptable for plant conditions existing at the point in the procedure (e.g., violation of fuel pin compression limits during a large SGTR). One should not infer, however, that
,- since it is acceptable at one point in the procedure then it is ,I \
s (,/ always acceptable to violate the limit or precaution. The guidelines may tell the operator to completely isolate both SG's (steam and fe edwa t e r) . This is because Part I is designed to restore plant sta-bility assuming worst case accidents, which, in the case of overcooling, would be an unisolable stean line break. If the operator does not know the
,o cause of the overcooling or even which SG is causing the overcooling then the fi +
( (,/ ' proper response h to isolate both SG's to stop the transient. With the transient terminated he should then be able to conitor secondary conditions and isolate the problem to one SG so that controlled decay heat removal can be restored using the unaffected SG. However, if he should discover the cause of the transient wh ile he is isolating the SG's (e.g., s tuc k open TBV's) and isolation of this cause does indeed stop the overcooling, then l
/ \ there is no re a son for him to complete the isolation of both SG's. If he
- i. I understands the purpose of the steps he is taking in following the guide-1 l lines, then he will understand when that intent is satisfied and may be able to exit the procedure at that point rather than arbitrarily following it
.h rou gh to completion. Of course, should he fail to realize that isolation of the TBV's terminated the transient, continuing through the guidelines,
,m l 7 \
l
!1 /
'O DATE: PAGE
, 7-6-82 227 I
BWNP-20007 (6-76) , 8A8 COCK & WILCOX NUCLEAR POWER GENERAhoN DIVISION 74-t i 2 553 i-00 TECHNICAL. DOCUMENT including isolation of both SG's for this example, will not cause any sig-nificant problem; it is merely unnecessary. Continuing through the procedure will lead to restoration of core cooling using the enaf fected SG(s), i i Objectives of Section III In order to promote understanding of the procedural guidelines, this section will address each subsection of Part I, Section III in terms of what the plant conditions are, what are the possible causes, and why the operator is directed to take the specified actions (i.e., what the actions are intended to accompli sh) . The re fe renced section of Part I should be followed while reading this section. Section III.A, Follow-up Actions for Treatment of Lack of Adequate Subcooling Margin Whenever plant conditions reach or exceed the subcooling margin curve the assumption is that the RCS is saturated. The RCS can beccme saturated as a result of three basic causes:
- i. loss of coolant inventory (LOCA) ii. ove rcooling that results in suf ficient coolant contraction to drain the pressurizer, or iii. prolonged loss of heat transfer that allows the RCS to overheat to 1
! saturation at high pressure (this cause would be recognized as lack l . of heat transfer before loss of subcooling and handled in accordance with Section III.B). O l I DATE: PAGE 7-6-82 228
BWNP-20007 (6-76) BABCOCK & WILCOX ' NUCtf AR Powta GENERATION DIYl$10N 74- t i 2 s s 3 t-oo (~'iTECHNICAL DOCUMENT
\
v Y The primary objectives of this section are to 1) restore subcooling margin and 2) maintain or restore core cooling. To accomplish these objectives, MU and itP1 must be initiated and either secondary heat transfer or MU/HPI cooling must be established.
,-~. .
I s
\
I \
,/ Two actions are always required whenever the subcooling margin is lost:
- 1. trip all reactor coolant pumps and
- 2. initiate full,1 alanced MU/HPI flow.
These actions are necessary in the event the loss of subcooling margin is due to a small break LOCA. Tripping the RC pumps must be done immediately fol-O lowing the loss of subcooling margin to minimize inventory loss if a small ( )
\. / b reak exists. In addition, if the ?oss of subcooling margin was the result of an overcooling transient, MU/HPI will compensate for the coolant contrac-tion and restore sub c ooling . Re-establishing controlled secondary heat removal should then be possible. If the loss of subcooling margin was due to a total loss of fe edwa t e r (main and auxiliary) then full MU/HPI will be needed to establish MU/HPI cooling (in Section III.B). In this case sub-
/m's cooling should also be restored but it may take some additional time until ( ) the cooling capacity of the MU/HPI flow exceeds the core heat generation rate. Raising SG levels should also be done in the event of a small break LOCA. High SG levels will allow condensation of steam voids in the RCS side of the m
- 8
\ ]
v' DATE: PAGE 229 7-6-82 ,
BWNP-20007 (6-76) BABCOCK & WILCOX NuctEAR Powta GENEaATION OlvistON 7'-i t 2353i-no TECHNICAL. DOCUMENT upper tube region to establish boiler-condenser cooling. If the transient O was initiated by total loss of feedwater then the operator may not be able to restore feed wate r and raise SG levels but after he has established MU/HPI cooling he should c ont inue ef forts to restore f e ed wa t e r . If the transient indicates overcooling then he should not raise level in the af fected SG(s) until the ove rcooli ng is corrected. This is to prevent further uncontrolled plant cooldown. Actions are also included to isolate possible causes for the loss of primary pressure. At this point the operator has full MU/HPI flow, the RC pumps are off, he is raising or attempting to raise SG levels, and he has isolated possible causes of the loss of sub c ooling . Further actions will be determined by the plant re s ponse to the actions already taken. The subcooling margin may be restored by MU/HPI or the plant may remain at saturation. In either case, primary to secondary heat transfer may or may not exist. Subcooling Margin Restored ! If the sub cooling margin is restored, the operator should restart reactor coolant pumpe. This will aid in establishing primary to secondary heat removal if it does not already exist. l l If primary to secondary heat transfer does not exist even after the sub-I cooling margin is restored, it is probably due to a lack of feedwater or a 1 blockage of RC flow due to steam voids in the hot legs. The operator will 1 O PAGE 230 l DATE: 7-6-82 l
BWNP-20007 (6-76) BABCOCK & WILCOX Nu srR NUCttAR Powit GENERATION DIVISION 74-i i 2 n 3 i-00 (~) TECHNICAL DOCUMENT
\ /
v proceed to Section III.B which will provide for restoration of feed wa t e r and primary to secondary heat transfer or establishment of MU/HP1 cooling. If excessive heat transfer exists, the operator will proceed to Section p_.s III.C. The operator should also be alert for indications of a small break l (refer to Appendix F in Volume 2) as a small break could exist that is within the capacity of the MU/HPI system. If controlled primary to secondary heat trans fer exists then the operator should regulate feedwa t er flow to establish SG levels at the appropriate set-point (dependent upon whether RC pumps were re s t a rt ed ) . The operator should
, also control steam pressure to prevent RCS reheating and swell. If the RCS / }
(v / were allowed to reheat, the added inventory from MU/HPI could result in a large pressure increase and possibly a full pressurizer. Subcooling Margin not Restored If full MU/HPI flow does not restore the subcooling margin even with heat l [ trans fer to the SC's, it is a LOCA. If heat transfer does not exist or
/^x exists in only one SG, the operator will proceed to Section III.B to attempt
)
\ /
v' restoration of heat trans fe r to both SG's. "h e prolonged period at satura-l t ion may be due to a total loss of feedwater or a small break. In either 1 case, restoration of feedwa te r flow in III.B will aid primary cooling. If, l however, the CFT's begin to empty, a large break exists and primary to secon-dary heat transfer cannot be regained. In thic case the operator will go to
,,.m '
cP-101 for long tenn cooling following a major LOCA. (,
%_/
DATE: PAGE 231 7-6-82
BWNP-20007 (6-76) 8ABCOCK & WILCOX Num En NUCttAR POwlt GENIPATION OlVi$10N 7'-112553i-00 TECHNICAL DOCUMENT NOTE: Wheneve r adequate subcooling margin does not exist, the operator should be alert to indications of superheat in the RCS (The incore the nno cou pl e s read higher than saturation temperature for the existing RCS pressure). If indications of superheat occur, the operator should proceed to the Inadequate Core Cooling (ICC) guidelines. Summary O The bases for Section III. A can be summarized as follows: Symptom: RCS pressure-temperature to the right and/or below the subcooling margin curve. Problems: a) Possible LOCA b) Void formation in RCS at saturation can interrupt core cooling Objectives: a) Restore subcooling margin b) Maintain or restore core cooling:
- i. preferably with SG's ii. with MU/HPI cooling (af ter transfer to Section III.B) if SG cooling unavailable Key Points: a) Be alert for indications of ICC b) Lack of subcooled margin can severely hamr* r primary to secondary heat transfer c) MU/HPI cooling, is not a stable long-term cooling mode. Cooling with one or both SG's must be restored as soon as possible.
(MU/HPI cooling is discussed in more detail in the " Backup l Cooling Methods" chapter). DATE: 7-6-82 PAGE 232 F
= .- - .. - _ _ _ _ _ - . _ . . .- -. -
BWNP-20007 (6-76) BABCOCK & WILCOX NUM8(R Nuctern powee oewenAnON DWISION TECMICAL DOCUMENT 74-ii25531-00 Section III.B: Follow-up Actions for Treatment of Lack of Primary to Secondary Heat Transfer in Either OTSG i i , If adequate subcooling margin exists, the most likely cause for lack of pri-mary to secondary heat transfer is no heat sink (loss of fe ed wa t e r ) . The operator will take steps to restore feedwa te r. If he cannot restore feed- i water he will establieh MU/HPI cooling and he should reduce the number of run- ; ning RC pumps to one to minimize heat input to the RCS. This will provide adequate core cooling while he continues e f fort s to restore teedwater to at , t least one SG. i Lack of Adequate Suocooling Margin If, however, adequate subcooling margin does not exist, then the lack of heat S transfer could be due to no heat sink (loss of feedwater)' and/or no reactor coolant flow (hot leg void ing) . The primary object ive is to restore core cooling. The pre ferred method of core cooling is with primary to secondary heat transfer. i If the operator has feedwater flow and a level in at least one SG , then the g lack of heat transfer is due to lack of RC flow. If the CFT's are emptying, a major LOCA has occurred and there is no benefit in res to ring primary to secondary heat transfer. In this case the operator will proceed to the procedure for long term cooling following a major LOCA. If a major LOCA has not occurred, he will attempt to induce natural circulation flow by raising his SG 1evel(s0 and lowering SG pressure. If this f a ils , and RCP's are l DATE: ' 233 7-6-82
BWEP-20007 (6-76) BABCOCK & WILCOX Num Em NUCLEAR PoWit GENERAf40N OlVl5loN 74- 125531-00 TECHNICAL DOCUMENT operable, he will attempt to induce natural circulation flow by bumping an RC pump. With the SG ava ilab le as a heat sink, bumping an RC pump will forc e steam voids in the RCS into the SG tubes where they can be collapsed. Fifteen minutes should be altotted between suc c es s ive pump bumps to allow natural circulation fl ow to start. If natural circulation flow is still not established af ter all operable RCP's have been bumped and one hour has passed since the reactor trip, a pump should be started and run, if possible, in a loop with the SG available as a heat sink. The hour limitation is based on allowing decay heat to decrease to a level that the HPI flow can accommodate so that additional inventory loss through the break due to forc ed flow is no longer a serious concern. If the RCP ' s are not operable or FW is not available, he must cool the core with MU/HPI and he will go to CP-104 for MU/HPI cooling. If he is successful in establishing primary to secondary heat trans fe r then he will go to the appropriate cooldown procedure depending on the degree of subcooling. If the subcooling margin is not restored then a small break probably exists. Summary The bases for Section III.B can be summarized as follows: j Symptoms: a) With subcooling margin, symptoma are those indicative of loss of feedwater:
- i. RCS reheating and repressurizing af ter normal post-trip cooldown.
O DATE: 7-6-82 234
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUCLEAR POwet GENERATION OlVi$10N TECHICAL DOCUMENT 74-1125531-00 Section III.B: Follow-up Actions for Treatment of Lack of Primary to 1 Secondary Heat Transfer in Either OTSG If adequate subcooling margin exists, the most likely cause for lack of pri-mary to secondary heat t rans fe r is no heat sink (loss of fe edwa t e r) . The operator will take steps to restore feedwa ter . If he cannot restore feed-water he will establish MU/HPI cooling and he should reduce the number of run-ning RC pumps to one to minimize heat input to the RCS. This will provide adequate core cooling wh ile he continues e f fort s to restore feedwat e r to at least one SG. Lack of Adequate Subcooling Margin
-. If, however, adequate subcooling margin does not exist, then the lack of heat
( transfer could be due to no heat sink (loss of feedwater)' and/or no reactor coolant flow (hot leg voiding). The primary objective is to restore core i cooling. The preferred method of core cooling is with primary to secondary heat trans fer. If the operator has feedwa ter flow and a level in at least one SG , then the S lack of heat transfer is due to lack of RC flow. If the CFT's are emptying,
)
b a major LOCA has oc cu rred and there is no benefit in restoring primary to secondary heat transfer. In this case the operator will proceed to the procedure for long term cooling following a major LOCA. If a major LOCA has not ~ occurred, he will attempt to induce natural circulation flow by raising his SG 1evel(s0 and lowering SG pressure. If this f a ils , and RCP's are i 1 PAGE I DATE: 233 7-6-82 , i _ . ..-..-.,_.~...,----,___..._,_.....,...,,_..,_,-,...-._._,,.__.---....,-_.,-._._m,..m,,, ,....r.,
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCitAR PoWit GENERAfloN Olvl5loN 74-t 2333i-00 TECHNICAL DOCUMENT operable, he will attempt to induce natural circulation flow by bumping an RC pump. With the SG available as a heat sink, bumping an RC pump will forc e steam voids in the RCS into the SG tubes where they can be collapsed. Fifteen minutes should be allotted between suc ces s ive pump bumps to allow natural circulation flow to start. If natural circulation flow is still not established af ter all operable RCP's have been bumped and one hour has passed since the reactor trip, a pump should be started and run, if possible, in a loop with the SG available as a heat sink. The hour limitation is based on allowing decay heat to decrease to a level that the HPI flow can accommodate so that additional inventory loss through the break due to forced flow is no longer a serious concern. If the RCP 's are not operable or FW is not available, he must cool the core with MU/HPI and he will go to CP-104 for MU/HPI cooling. If he is successful in establishing primary to secondary heat trans fer then he will go to the appropriate cooldown procedure depending on the degree of subcooling. If the l subcooling margin is not restored then a small break probably exists. 1 l Summary ! The bases for Section III.B can be summarized as follows: Symptoms: a) With subcooling margin, symptoms are those indicative of loss of feedwater:
- i. RCS reheating and repressurizing af ter normal post-trip l cooldown.
l DATE: # l 7-6-82 23.4
4 ' BRNP-20007 (6-76)
~
i BABCOCK & WILCOX NUM8tt i NUCLEAR POWER GENga TTON DIV15pON ' 74- i25531-00 ; TECHNICAL DOCUMENT ii. Low or non-existent SG levels and feedwater flowra t e s , f i
' b) With lack of adequate subcooling margin, symptoms j could be as above for loss of feedwater and/or small break ,
I symptoms (see Appendix F in Volume 2). Problems: a) Lack of core cooling l l 1 i b) Extended loss of feedwater will lead to saturation and i leop voiding-necessitates MU/RPI cooling. c) Possible LOCA. A Objectives: a) Maintain or restore subcooling margin. i b) Restore core cooling, preferably using the SG's. Key Points: a) Lack of heat transfer with adequate subcooling is more than likely due to total loss of feedwater or insuf ficient feedwater to induce natural circulation, b) Eve ry e f fort must be made to restore primary to secondary heat transfer (unless a Major LOCA occurred). i i 1 i i 'l DATE: PAGE 235 7-6-82
~. . - , . , . . , _ _ - . , . _ _ , _ , - - _ . - _ . . _ _ _ _ _ . . . . - . _ . _ . . _ _ _ . _ . _ _ _ _ . _ _
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBit NUCLEAR POWER GENERAflON DIVISION 74-t i2 ss31-00 TECHNICAL DOCUMENT Section III.C: Followup Actions for Treatment of Too Much Primary to Secondary Heat Transfer Exc e s sive heat transfer is always caused by a f ailure in the control of se-condary side par ame t e r s , resulting in a loss of steam pressure or excessive feedwater fl ow or a combination of both. The overcooling places large ther-mal stresses on the RCS piping and components and on the steam generators. It can also le ad to saturation of the primary sys t en if the RCS contraction is large e aough to drain the pressurizer. The primary objective of this section is to te rmina t e the overcooling transient and then to restore con-trolled decay heat removal. MU is increased if pressurizer level is low and RCS pressure is decreasing in an e f fort to prevent drainage of the pressurizer and loss of subcooling mar-gin. If the SG level > 95% the MFW pumps will be t r ipped . If the SG level is not excessive the operator will then check to see if the SG causing the ove rcooli ng can be identified. The best method to identify the affected SG is to compare Tcold temperatures. The loop with a significantly lower Teold is the loop with the af fected SG. However, Teold temperatures can be fairly close together even when only one SG is causing the overcooling (refer to the discussion on overcooling transients in the " Diagnosis and Mitiga tion" chapter). Since the primary objective is to first terminate the transient, both s tean generators should be isolatei if there is any doubt which SG is affected. In either case (one or both SG's isolated), the af fected SG will either: DATE: PAGE 236 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numet NUCLEAR POWER GENERATON OlvlSioN 7'-i i 2 333 i-oo (^'sTECHNICAL DOCUMENT
\ !
b/ a) stabilize in level and pressure if the overcooling was due to excessive feedwater or isolable steam leak, or b) cont inue to lose pressure and level due to an unisolable steam leak. If a), then controlled decay heat removal can be restored using both SG's / \ (being careful not to unisolate a steam leak). If b), then the SG with the ( ,
)
v steam leak should be allowed to boil dry while decay heat removal is established with the intact SG. Once controlled decay heat removal is establisheu, the operator should check for indications of a tube rupture (since the tubes have been stressed by the ove rc ooling) and verify adequate subcooling margin exists. If the affected
,a, SG was returned to service, then the check for tube rupture can be made using
( ) k./ s tean line monitors. If the af fected SG was left isolated, then the operator should check for indications of a small break LOCA and continued cooling of the RCS by the isolated SG (due to boilof f of tube leakage). It should be noted that identification of a small break LOCA that is not a tube rupture (e.g. RCP seals) does not preclude the possibility that a tube ruptura also
,- exists. The identi fied small break LOCA may mask the presence of a tube
('( m ,,/ s i rutpure. The import ant point is to ensure that the unisolable steam leak is not a source of radioactivity release that is unknown to the operator. If adequate subcooling margin does not exist he should proceed to Section III. A. Wheneve r an overcooling transient has been t e rm in a ted , the operator should
/m / h hold RCS tempera tures at the existing values. If the RCS were allowed to \ J v reheat, the added inve nto ry from MU and HPI could result in a large pressure increase and possibly a full pressurizer.
DATE: PA E 237 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nuuien NUcttA2 POWER GENERATION OlVI$lON 7'- t i 2 s s 3 i-00 TECHNICAL DOCUMENT Summary l The bases for Section 111.C can be summarized as follows: Symptoms: a) Decreasing Tcolds and/or SG pressures significantly below the normal post-trip cooldown. b) Possibly high SG level (s) and feedwater flowrate(s). c) Low RCS pressure and pressurizer level. Problens: a) Thermal stresses on RCS components and SG's (tubes). b) Possible RCS saturation due to pressurizer drainage, c) Exc es sive feedwater flow could result in water carryover into the main steam lines. Obje ct ives: a) Pr eve nt loss of subcooling margin due to RCS contraction and drainage of the pressurizer. b) Terminate the overcooling, c) Re-establish controlled primary to secondary heat t rans fe r. d) Stop SG overfill. Key Points: a) Comparison of loop Tcold temperatures is best method for identifying af fected SG before SG isolation. b) Comparison of SG 1evels and pressures is best method for identifying affected SG after both SG's are isolated, ( c) Severe overcooling can induce tube lerAs. d) The RCS should not be allowed to reheat after the
. transient is terminated, e) Unisolable steam leaks require boiling the affected SG dry to stop the overcooling.
l l l DATE: PAGE 238 7-6-82
a BWNP-20007 (6-76) l BABCOCK & WILCOX " NUCLEAR Powet GENERATION OtVi$lON 74-1125531-00
') TECHNICAL DOCUMENT Section III.D: - Followup Actions for OTSC Tube Rupture Several concerns exist whenever indications of a steam generator tube rupture become evident. In addition to being a LOCA, the primary inventory lost through the tube leak cannot be recovered for sump recirculation as it would for other LOCA's. Thus, it is important to cooldown and stop the tube leak \ before makeup capacity (BWST) is lost. However, it is also important to mini-mize offsite releases. Therefore, if at all possible, it is desirable to pe r-form a controlled power runback and reactor shutdown at a power level less than the capacity of the TBV's rather than trip the reactor at high power.
This prevents lif ting of the safety valves on the SG with the tube leak which helps reduce overall releases to the atmosphere. Thus the primary objectives
% in mitigating a tube rupture are to minimize offsite releases and total tube "Q 1eakage by performing an orderly but expedient shutdown and cooldown. Open-ing the TBV's be fore tripping the turbine and reactor (when power is less than total TBV capacity) will prevent lifting of the main steam safe ty valves.
Reactor Trip If a reactor trip should occur or be required because the tube leakage ex-
\
ceeds MU capacity, then it is important to ensure proper plant re s ponse , par-ticularly with respect to stean and feedwater control. If a loss of subcool-l ing margin occurs the RC pumps must be tripped and full MU/HPI initiated. 1 These actions are required for the same reason as in any small break but, in addition, it is very important with a tube rupture to prevent void formation
\
d DATE:
- A 7-6-82 239
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCitAR POWER GENERATION OtVi$ ION
"~"*"~
TECHNICAL DOCUMENT. in the hot legs which could block coolant flow and severely hamper the cool-down rate. It is also very important to te rmina te exc es s ive heat trans fe r should it occur. Uncontrolled cooldown could also result in void formation in the hot legs and could overstress the SG tubes resulting in larger leak rates. Thus, if excessive heat trcnsfer is indicated, the operator is directed to Section III.C to correct the condition before continuing in Section III.D. During the period that the operator is performing the shutdown or stabilizing the plant af ter a trip, a survey of the steam lines should be made to verify which SG has the tube rupture. It is desirable to isolate the af fected SG as sos t as possible after primary temperature is low enough to preclude lifting of the main stean safety valves. Cooldown Methods Two basic cooldown methods are provided for tube ruptures, designated here as
" normal" and "eme r ge nc y" . The method to be used is determined by the tube leak rate and the existing plant status. The dif ferences between the methods are as follows:
Normal Emergency
- 1) Cooldown rate 100F/hr 240F/hr to 500F
- 2) Tube /shell A T limit 100F 150F
- 3) Fuel pin compression limit Applies May be violated, if neces s ary O
DATE: . PAGE 240 m a um
BWNP-20007 (6-76) SASCOCK & WILCOX OII NUCttAA POwge GENIAATION OWISCN 74- t i 2553 i-00 p TECHICAL 00CUMEllT V The factors determining use of the " normal" cooldown method are:
- 1) Tube Irak rate within capacity of makeup system (minimal rate of BWST depletion).
- 2) Condenser available
- 3) RC pumps available All three conditions must exist to use the normal cooldown. If any one condi-tion is not met, the emergency cooldown method must be used. If, during the performance of the emergency cooldown, all three conditions are satisfied then the normal cooldown can be used. Conversely, if during the performance of the nonnal cooldown, any one of the conditions is no longe r satisfied, then the cooldown must be switched to the emergency method. In either case, the SG with the tub e rupture should be isolated as soon as it is identified and Thot is < 540F ( to prevent lifting the steam safeties on the SG with the tube leak).
Loss of Offsite Power /RC Pumps Not Running A loss of of fsite power can significantly impact the mitigation of a tube rup-ture and therefore power should be restored as quickly as possible. While power is unavailable a natural circulation cooldown will be required. It
}
will be necessary . to periodically steam the SG with the tube leak during a natural circulation cooldown to maintain loop circulation and avoid hot leg flashing in that loop. Void fo rma tion in the hot leg would hamper the cool-down due to the inability to depressurize the RCS (the hot leg would act as a surge volume). With the condenser not available, stemning of the af fected SG would have to be done directly to atmosphere thus increasing offsite releases. DATE: 7-6-82 24i
BWNP-20007 (6-76) BABCOCK & WILCOX wussen NVQEAR POWit GENERATION DIVISION 74- u 25531-00 TECHNICAL DOCUMENT In addition, bAile RC pumps are not available, RCS pressure reduction must be accomplished using the pressurizer relief. This is a less desirable method since it repeatedly challenges the relief valve, results in additional inven-tory loss, and may degrade the reactor building envirornne n t . The cyclic operation be tween minimum subcooling margin and 30F more subcooled than the subcooling margin is a compromise between the need to limit the cycles on the relief valve and the need to maintain as low a pressure as possible to mini-mize the tube ickage rate. RC Pumps Running With RC pumps available, steaming of the affected SG can be reduced to the minimum necessary to keep the steam pressure less than 1025 psig and the level less than 95% on the operate range. This is required to prevent lif ting of the steam safeties and atmospheric vent valves (additional release to atmosphere) and to prevent water spillover into the stemn lines. Forced circulation will pr even t the formation of steam voids in the idle (non-s teaning) loop. RCS pressure control is better with RC pumps and spray available. The re fo re , spray should be used as necessary to omintain the RCS pressure as close as possible to the minimum subcooling margin to minimize the leak rate. Continued Cooldown and Isolation of the Affected SG The plant is not stable after a tube rupture until the tube leakage has been l stopped. This will require cooldown and depressurization and DHRS operation l DATE: 7-6-82 PAGE 242
. . . --. - - - - . . - . . . . --- - _. - _ . _ . -. - _ _ . _ - - _ =
$ BWNP-20007 (6-76) l , BABCOCK & WILCOX NUMBER f MUCLEA: POwes 04NBATION DIVl560N 74-i125531-00 i i- TECHNICAL DOCUMENT to the point whe re the RCS can be drained to below the elevation of the tube leak. The re fo re, the cooldown should progress as expediently as possible.
=
t
% i Once the af fected SG has been isolated, it should only be fed and/or steamed !
as nec es sa ry to maintain steam pressure < 1000 psig and level < 95% on the , operate range and to maintain the tube /shell T within the limits. Excessive tube / shell T could result in a higher leak rate due to the increased tensile stress on the failed tube. In additioa, as previously stated, steaming of ! the isolated SG may be required during a natural circulation cooldown to pre-i i vent void formation in the hot leg. i i i Summary , Symptoms: a) high radiation in steam lines and/or condenser vacuum system b) LOCA symptoms (decreasing pressure, unaccountable RCS in-ventory loss, etc .) l Problems: a) SG tube rupture /LOCA b) unrecoverable RCS inventory loss (i.e., not available for 4 sump recirculation) c) offsite releases l I Objectives: a) minimize offsite releases ' i b)~ tenninate leakage before BWST depletion i i c) maintain core cooling / expedient cooldown and depressuriza-tion DATE: PAGE 7-6-82 243
. . _ . _ _ _ . - _ _ . ~ . . . . . . _ _ _ _ _ . _ . . _ _ . _ . _ . _ _ . . . _ . _ _ _ _ _ ._...,_ _.._ .-,... _ _
BWNP-20007 (6-76) BABCOCK & WilCOX yv ,,,, NUCitAR POWER GENERAlloN DivtSION 7'-ii23331-00 TECHNICAL DOCUMENT Key Points: a) transient (leakage) not terminated until RCS cooled, depressurized, and drained below tube leak elevation b) LOOP / natural circulation cooldown will probably result in highe r offsite releases due to greater need to steam the af fected SG to the atmosphere c) natural circulation cooldown necessitates use of the PORV to reduce RCS pressure Cooldown Procedures / Inadequate Core Cooling The objective of these guidelines is to umintain adequate core cooling by ter-minating transients and stabilizing the plants with controlled decay heat re-moval. Once' stable conditions are achieved, further plant cooldown can be accomplished by existing plant procedures. However, the end conditions at stabilization following the execution of the guidelines will not necessarily coincide with the entry conditions for plant cooldown procedures. There fo re , procedures are provided in Part I to accomplish the transition fran the guide-lines to the plant procedures. Five cooldown procedures are provided to cover the five possible end conditions of the guidelines:
- 1) Cooldown following a large LOCA i
- 2) Normal cooldown i
l 3) Saturated cooldown with primary to secondary heat transfer
- 4) MU/ilPI cooling 5 )- Solid plant cooidown/ recovery from solid plant.
1 O DATE: 7-6-82 PAGE 244
BWNP-20007 (6-76) BABCOCK & WILCOX NUMSER NUCLEAR POWER GENetATION OlVi$lON
' 7'- 12553i-00
' TECHNICAL DOCUMENT .
A sixth procedure is provided for the special case of Inadequate Core Cooling (ICC). The philosophy and the objectives of the actions for ICC are dis-4 cussed in detail in the " Backup Cooling Methods" chapter in Volume 1 of Part II. r At the end of the section containing the five cooldown procedures (immedi-ately before the ICC section) are specific rules. These rules are provided in a separate section to avoid repetition throughout Section III of the guide-lines. These rules apply wherever they are referenced in Section III. Specific rules are provided to cover:
- 1) Initiation of HPI
! N 2) HPI flow control l
- 3) Auxiliary Feedwater Control methods l
l
- 4) SG 1evel setpoint. ,
In addition, four figures are provided at the back of Part I for easy refer-ence during the use of the guidelines. These figures provide: I ! 1) HPI flow vs. RCS pressure
- 2) RCS pressure-temperature limits 3)~ Core exit fluid temperature for ICC
- 4) HPI throttling limit Operator Aids In addition to the guidelines in Part I and the training material in Part II, two other developments of the ATOG program can provide significant assistance 1
i V DATE: AGE 7-6-82 245
BWNP-20007 (6-76) BABCOCK & WILCOX Numste NUcLIAA POWER GENT 9ATION DIV1510N 7'- 125531-oo TECHNICAL DOCUMENT to the operator during the mitigation of abnormal transients: the pressure-tempe ratur e (P-T) display and the System Auxiliary Diagrams (SADs). The ATOG user should provide for full utilization of these aids in the implementation of the guidelines. P-T Display The info rma t ion required to identify and track the symptoms discussed previ-ously is already available in the control room. However, it consists of dis-crete displays for reactor coolant system hot and cold leg tempe ra tur e s , reactor coolant system pressure and steam generator pressure. This format requires mental integra t ion on the part of the operato r to quickly assess plant cooling status using individual displays and the steam tables. Thus, the problen is how these variables could best be displayed in real time t give the operator a simple and logical method of monitoring the symptoms of interest. The solution developed in ATOG was the use of a P-T display on a cathode ray tube (CRT). The display continuously shows the primary sub-cooling margin and the dynamic relationship of the primary to secondary heat transfer. The particulars of the display format and identification of the symptoms is discussed in detail in the "P-T Diagram" chapter. This sectio will discuss the various functions the display can perform to aid the operator. Symptom Identification The primary purpose for developing the P-T dispicy is to provide the means for the operator to monitor the plant response following a reactor trip o DATE: 246 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX se,,,, suctean rowen oewenArow oivision 3,3 g g g g g i
/ MTECHNICAL DOCUMEMT 4 \ /
V during a forc ed shutdown to verify normal response and to quickly identify abno rmal re s pons e should it occur. The "P-T Diagram" chapter de sc ribe s the normal post-trip cooldown of the RCS and the various trends that can develop wh en the response is abnormal . With the P-T display the operator can quickly recognize the loss of subcooled margin, lack of heat trans fer or excessive 7 (\w heat t rans fer and proceed directly to the appropriate section of Part I to
/
)
restore plant stability and core cooling. Response Verification The P-T display also provides positive feedback to the operator on the re-s ponse of the plant to his actions. For example, after the loss of sub-cooling margin, the operator can easily determine the e f fec t ivene s s of full l ,cx ( \
) IIPI flow by monitoring the P-T display and determine when to throttle HPI V
flow if the subcooling margin is restored. Controlling Within Limits Certain operations require that the operator control the primary system within specified P-T regions that can readily be displayed on the P-T dia-em gram. For example, the guidelines for a natural circulation cooldown with a i Q ,/ tube rupture require maintaining RCS pressure between the subcooling margin line and 30F more subcooled than the subcooling margin line. If the operator has the capability to select such a curve for display on the CRT (the sub-cooling margin line is already displayed) then he has a sim ple , convenient format for monitoring the cooldown and ensuring compliance with the limits, [m \ z Many other examples exists, : inclu' ding uses during normal plant cooldowns and i l
'd heatups (e.g., fuel pin compression limits, NDT limits, etc.).
DATE: PAGE 7-6-82 247
BWNP-20007 (6-76) BABCOCK & WILCOX y ,,, NUCLEAR POWER GENERADON DIVISION 7'-t 23s31-oo TECHNICAL DOCUMENT Power Operation The CRT format is readily adaptable for displaying plant status informa t ion during normal power opdration. Some examples for such usage are the reactor protection sys t en pressure-temperature trip envelope (shown in the "P-T Dia-gram" chapter) and power imbalance envelope. However, if the ATOG CRT is used for these di s play s , they should be of a sec onda ry nature with the CRT automatically reverting to the ATOG P-T display on reactor trip. Backup for the CRT It can readily be seen that the availability of a P-T display improves the flow of in forma t ion to the operator and enhances the use of Part I. E f fort should be made in the implementation of CRT displays to provide high reliability. Iloweve r , control room personnel should allow for the possibility that the CRT displays are unavailable when needed. Provisions should be made to facili-tate hand plotting of the parameters on a P-T diagram similar to the diagrams depicted herein. Hand plotting is quick enough to provide data which can be used for plant control. The format of the diagram for hand plotting (with l the saturation and subcooled margin lines pre-drawn) would allow for trend 1 diagnosis and still be a significant improvement over mental assimilation of discrete data displays. I System Auxiliary Diagrams System Auxiliary Diagr am s (SADs) were developed in the ATOG program to ide n-tify supporting aystems essential to the operation of systems having direct l 248 DATE: PAGE 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUMB ER NUCitAR POWER GtHERATION DIV!5 TON 74-t i 23 s 31-00
/A~'s TECHNICAL DOCUMENT \ l w/
input to plant r e s pon s e . They also identify instrumentation required to verify proper operation of supporting syctems. The SADs serve as a useful aid in the event that a critical system fails. For example, the operator ty be required to initiate HPl. In the highly ( ( j) unlikely event that a total loss of HPI occurs, the associated SAD can be used as a rapid t roub le shoot ing aid to restore HPI operation. The SAD fo r itPI shows llPI in the center and various arrows pointing toward HPI wh ich ident ify everything required to make HPI initiation s uc c e s s f ul . Pump power supplies, required cooling / lube oil sources, major inline valve positions, ve nt ila t ion cooling, etc., are all identified along with available instru-mentation to verify proper operation of the HPI system. Only those items [] (
\v) that are within the operators ability to control and can be accomplished quickly are included. Corrections that are longer term (e.g., replacing a pump impeller) are omitted .
Since t roub leshoot ing will be performed by roving operators or maintenance groups, the SADs are packaged separately es opposed to being contained in the A / T ATOG volumes. However, the appropriate SADs are reference in Part I where i ! applicable. Station management will determine the availability and use of the SADs. SADs have been developed for the following systems and components:
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DATE: PAGE 249 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POs/tt GENERATCH DIVIS:ON 7'-1125531-00 TECHNICAL DOCUMENT
- 1. Main feedwater system (loss of flow).
- 2. Auxiliary Feedwater System (loss of fiow).
- 3. Steam line components (loss of steam pressure),
- a. Turbine bypass valves
- b. Main steam safety vaives
- c. Atmospheric vent valves
- d. Turbine controls
- 4. ECCS Systms ( f ailure to deliver water).
- a. Makeup
- b. IIPI
- c. LPI
- 5. Contaiment cooling systems (f ailure to depressurize containment)
- a. Containment spray
- b. Contaiment coolers
- 6. Containment isolation (failure to isolate).
- 7. Boron addition (inability to add boron).
- 8. Component s for RC pressure control
- a. Pressurizer heaters
- b. Pressurizer spray O
DATE: 7-6-82}}