ML20086S729
| ML20086S729 | |
| 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 8403050118 | |
| Download: ML20086S729 (230) | |
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ABNORMAL TRANSIENT OPERATING GUIDELINES PART II VOLUME 2 74-1125531-00 Doc. ID - Serial No., Revision No. 4 for TOLEDO EDISON COMPANY by BABCOCK & WILCOX 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.
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BWNP-20007 (6-76) 1 BABCOCK & WILCOX Nu sen l NUCLEAR POWER GENER Af TON DIVISION 74-ti2ss3i-00 TECHNICAL DOCUMENT ( ATOG GUIDELINES PART II VOLUME 2. DISCUSSION OF SELECTED TRANSIENTS Tab INTRODUCTION O A. Excessive Main Feedwater Excessive MFW
- 3. Loss of Main Feedwater Loss of FW C. Steam Generator Tube Rupture SGTR D. Loss of Offsite Power LOOP E. Small Steam Leak Steam Leak F. Loss of Coolant Accidents LOCA v
C DATE: 7-6-82 PAGE 2
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sta NUCLEAR POWEn GENERATION DIVISION
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(3 TECHNICAL DOCUMENT (N.- / VOLUME 2 DISCUSSION OF SELECTED TRANSIENTS LIST OF FIGURES FIGURE NUMBER TITLE A-1 Excessive Feedwater Terminated by ICS A-2 Excessive Feedwater Not Terminated by ICS
' A-3 Actual Plant Response During an Excessive Main Fecdwater Transient A-4 Actual Plant Response During an Excessive Main Feedwater Transient A-5 Typical Excessive Feedwater Transient A-6 Time Relationship of Key Parameters A-7 Excessive Feedwater Logic Diagram in / \ '\ _,
s B-1 Loss of Feedwater (Initial Stage) B-2 Loss of Feedwater (Middle Stage) B-3 Loss of Feedwater (Final Stage) B-4 Time Relationship of Loss of Main Feedwater with Failures B-5 Loss of Feedwater Logic Diagram C-1 Steam Generator Tube Leaks - Operator Action [,_s% Outline
\ ) ' - ' C-2 Typical Steam Res po ns e fo r Large SGTR Which Results in a Reactor Trip C-3 Steam Generator Tube Rupture Logic Diagram C-4 RCS Cooldown Using the PORV D-1 Loss of Offsite Power - 2/22/75 Loop A RC Temperatures vs. Time l
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> D-2 Loss of Offsite Power - 2/22/75 i 'x_ / Loop B RC Temperatures vs. Time l
PAGE 3 D ATE :7-6-82
. . . - . - . -.. - . - - _ _ = _ - - . . . _.
, BWNP-20007 (6-76) BABCOCK & WILCOX NumER NUctEAR POWER GENERATION olvisiON TECHNICAL DOCUMENT 74-i125531-o0 FIGURE NUMBER TITLE 1 D-3 Loss of Offsite Power - 2/22/75 Loop A Pressure vs. Time D-4 Loss of Offsite Power - 2/22/75 OTSG Full Range Level vs. Time { i i D-5 Loss of Offsite Power - 2/22/75 j- OTSG Steam Pressure vs. Time < D-6 Loss of Offsite Power - 2/22/75 i D-7 Typical Parameter Trends for LOOP f D-8 Typcial Parameter Trends for LOOP / Loss of Secondary Inventory Control (Lcw) { D-9 Loss of Offsite Power Logic Diagram i D-10 Typical Loss of Offsite Power P-T Response ; i. E-1 Small Steam Leak 4 E-2 Typical Small Steam Leak Transient t E-3 Time Relationship of Key Parameters ] E-4 Small Steam Leak Logic Diagram l l ! F-1 Fluid History,During a Large Break i F-2 System Response for Large Break in Cold Leg Piping { ' , .F-3 System Response for Large Break in Hot Leg Piping i F-4 Typical System Response During Flow Circulation Phase of a Small Break 4 F-Sa System Response for Small Break Which Continually 4' Depressurizes the RCS (0.5 ft2 Break in Cold Leg Pipe) ! i F-5b_ System Res pons e for Small Break Which Continually Depressurizes the RCS (0.1 s ft2 Cold Leg Break) 4 1 i DATE: 7-6-82 PAGE 3-1
BWNP-20007 (6-76) BABCOCK & WILCOX Numset NUCLEAR POWER GtNetATION DIVI $10N 74-i i 2333 i-00
-TECHNICAL DOCUMENT FIGURE NUMBERS TITLE F-6 System Response for Small Break vmich Stabilizes At Secondary Side Pressure.
F-7 System Response for Small Break Which Repressurizes in a Saturated State F-8 System Response for Small Break Which Depressurizes the RCS Without Feedwater F-9 System Response for Small Break That Stabilizes at High RCS Pressure Without Feedwater F-10 System Response for Small Break That Repressurizes If Feedwater Is Lost F-ll System Response for Small Break Within Pressurizer Steam Space F-12 Post LOCA Corrective Action for Lack of Adequate Subcooled Margin 9 F-13 Post LOCA Corrective Action for Lack of Primary to Secondary Heat Trans fer DATE: 7-6-82 PAGE 3-2
BWNP-20007 (6-76) 7
! BABCOCK & WILCOX NUCLEAR POWER GENERAYlON OlvistoN 7'- 12553 t-00 l TECHNICAL DOCUMENT i !
l VOLUME 2 DISCUSSION OF SELECTED TRANSIENTS t LIST OF TABLES l i Table Number TITLE C-1 Ways to Detect a SGTR ! l C-2 Deleted i i C-3 Ef fects of Failures on Steam Generator Tube Leak t l_ Control D-1 Summary of Major Component Loadings During LOOP [ i l l F-1 General Post-LOCA Corrective Action to Maintain ' l Core Cooling l l -F-2 How to Distinguish LOCA's from Other Transients I l F-3 Summary of General LOCA Symptoms l
-F-4 Symptoms for LOCA's That Can Be Located or Isolated F-5 Summary of Long Term Cooling Actions ,
i i l I I l i : 1 I i l I i DATE: 7-6-82 PAGE 3-3 ;! . _ _ - , . ~ , ,
i BWNP-20007 (6-76) SABCOCK & WILCOX NUMBER NUCLEAR POWER G;NERATION DIVISION 74-1125531-00 TECHNICAL DOCUMENT DISCUSSION OF SELECTED TRANSIENTS INTRODUCTION To develop the ATOG approach to integrated plant transient control six initiating events were studied. The results of these studies are given as examples:
- 1. Excessive Main Feedwater
- 2. Loss of Feedwater
- 3. Steam Generator Tube Leaks f
- 4. Loss of A/C Power
- 5. Small Steam Line Break
- 6. Loss of Coolant Accident N
l-These events show how the procedures of Part I are applied to specific trans-ients and they amplify the general guidance given by the first volume of l Part II. After reviewing these individual transients in detail, the opera-t tor should be able to see that regardless of (1) the initial event, or l (2) whether or not he can immediately identify the initial event, or (3) how i l many additional failures occur, he can keep the core and the plant in a safe condition by following Part I of these guidelines. i Each appendix , discusses the planc lesign response to its initiating event and, if available, gives an example of an actual transient. The discussion points out what operator actions are required and how Part I directs the operator to those actions. PAGE 4 DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numate NUCit AR POWit GEN ERAfloN DIVISION M-n2ss31-00 TECHNICAL DOCUMENT , Next is a review of the initial event compounded with other equipment fail-ures. These are generally broken down into the failures of the fundamental methods of heat transfer control that are important to that event. Again the appropriate operator action fo r that particular sequence of failures are given. Part 1 is referenced so that the operator can see how the basic pro-cedure covers many mult iple failures. Each appendix contains a logic diagram which is a summary sheet for the transient being discussed. It is a simplified event tree wh ich has been modified to show how correct operator actions will influence the outcome of the transient. The central vertical block diagram is the initiating event without additional failures. The failure paths branch out to the right and the lef t . The details on the diagram show identifying symptoms including P-T, the corrective actions, and the limits to be considered for each addi-tional failure that might occur. References are made to the appropriate parts of Part I and Part II. Nearly every kind of plant condition is covered either in the initiating event path or in the branches. Even if the initiating event is different from the examples, the principles to be used are the same, and therefore these examples illustrate how various other plant conditions can be handled. Plots of varioits parameter trends are also provided. The times can only be approximate because they vary with such things as decay heat, initial power level and the size of the leak involved. However, they will give the opera-tor a feel for the timing involved in these transients. 5 DATl: 7-6-82 PAGE
2 i BWNP-20007 (6-76) o 4 BABCOCK & WILCOX , NUCLEAR POwta GENtaATION DIVB510N
, 74-1125531-00 i- TECHNICAL DOCUMENT DISCUSSION OF SELECTED TRANSIENTS ! INTRODUCTION 1
To develop the - ATOG approach to integrated plant transient control six , initiating events were studied. The results of these studies are given as , i l examp le s :
- l. Excessive Main Feedwater
- 2. Loss of Feedwater
- 3. Steam Generator Tube Leaks
! .4 . Loss of A/C Power i . 5. Small Steam Line Break
- 6. Loss of Coolant Accident i
i These events show how the procedures of Part I are applied to specific trans-l ll ients ani- they amplify the general guidance given by the first volume of Part II. After reviewing these individual transients in detail, the opera-tor should be able to see that regardless of (1) the initial event, or (2) whether or not he can immediately identify the initial event, or (3) how many additional failures occur, he can keep the core and the plant in a safe condition by following Part I of these guidelines. Each appendix discusses the plant design response to its initiating event and, if ' available, _ gives an example of an actual transient. The discussion points out what operator actions are required and how Part I directs the m operator to those actions. 4 DATE: i-6-82 PAGE
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sen NUCLEAR Powta genit AitoN DivlS10N 7'-i 2ss31-oo TECHNICAL DOCUMENT Next is a review of the initial event compounded with other equipment fail-ures. These are generally broken down into the failures of the fundamental methods of heat transfer control that are important to that event. Again the ppropriate operator action for that pa rt ic ula r sequence of failures are given. Part I is referenced so that the operator can see how the basic pro-cedure covers many multiple failures. Each appendix contains a logic diagrare which is a summary sheet for the transient being discussed. It is a simpli fied event tree wh ich has been modified to show how correct operator actions will influence the outcome of the transient. The central ve rt ical block diagram is the initiating event without additional failures. The failure paths branch out to the right and the left. The details on the diagram show ident ifying symptoms including P-T, the corrective actions, and the limits to be considered for each addi-tional failure that might occur. References are made to the appropriate parts of Part I and Part II. Nearly every kind of plant condition is covered either in the initiating event path or in the branches. Even if the l 1 initiating event is different fron the examples, the principles to be used are the same, and therefore these examples illustrate how various other l plant conditions can be handled. l l Plots of various parameter trends are also provided. The times can only be approximate because they vary with such things as decay heat, initial power level and the size of the le ak involved. However, they will give the opera-tor a feel for the timing involved in these transients. 5 ' DATE: 7-6-82 PAGE
BWNP-20007 (6-76) J BABCOCK & WILCOX NUCLEAR POWER GENERAflON OlVI$lON 7'- u 2333t-00 TECMICAL 00CUMElli , The operator should give special attention to the sections on LOCA and steam generator tube rupture because very de tailed information has been prepared for these two events. l , + All transients discussed in this volume start with the reactor critical, the plant at 100% full power and all parameters operating within their normal ; 1 > bands. However, the thermodynamic principles apply anytime the RCS is full and pressurized. There fore , the guidelines are applicable all the way to i cold shutdown. t \ 4 i i i r DATE: 7-6-82 PAGE 6
BWNP-20007 (6-76) BABCOCK & WILCOX l HuusEn NUCLEAR POWER GENERATION Division 74-ii23s3i-00
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N' APPENDIX A EXCESSIVE MAIN FEEDWATER 1.0 GENERAL TRANSIENT DESCRIPTION Excessive main feedwater is a failure to control secondary inventory.
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} It is an overcooling transient that results in too much primary to l
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secondary heat transfer. Excessive main feedwater is defined as the sustained addition of more water to the steam generator than can be boiled off by the available core heat to make supe rheat ed steam. This mismatch between heat source and heat sink will cause the steam generator level to rise and will cool the reactor coolant down. The severity and rapidity of the (vs ) transient will vary with the size of the mismatch. Under worst case conditions (i.e., maximum mismatch and failure of the automatic MFW pump trip on high SG 1evel) the excessive main feedwater flow must be terminated within two minutes to prevent water spillage into the steam lines. Thus, this is a transient that may require fast operator
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As the reactor coolant temperature decreases, the RCS water volume will shrink, dropping pressurizer level. This in turn causes RCS pres-sure to drop. In the secondary side, while at power, the excessive feedwater will cause a loss of superheat and may cause a slight reduc-
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tion of steam pressure. A reactor trip may occur on low RC pressure i !
'K - or high flux. If a trip occurs and the excessive feedwater continues, I
Appendix A, Page A-1 l DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numain NUCttAt POWER GENEGATION Division 74- 2353 i-oo TECHNICAL DOCUMENT the mismatch will be much larger (less core heat) and the steam gene-rator fill rate and RCS cooldown rate will increase. If the shrinkage of the RCS water volume is su f ficient to drain the pressurizer, the RCS will rapidly approach saturation conditions and SFAS will actuate. A loss of subcooling margin will require that the RC pumps be tripped and AFW be started. However, AFW will not flow unless SG level is less than 93" on the startup range. If the SG level in either SG is less than 93" AFW will flow to it. If the loss of subcooling margin is caused only by the overcooling it will be t empo ra ry . When the sub-cooling margin is restored HPI should be throttled and the RC pumps can be restarted. If the exc ea s ive addition of feedwater to the steam generator is not stopped, water will spill into the steam lines. The ability of the s t ean system to maintain its integrity with water spillage is not known; the re fore it is very important that the excessive feedwate r transient is te nni na ted be fore spillage occurs. In addition, it is highly desirable to stop the RCS cooldown be fore the pressurizer is drained. This will significantly reduce the magnitude of the trans-ient and number of operator actions required, as well as limit chal-lenge s of protection systems and allow for quicker recovery to s t able plant conditions. In general, main feedwater overfeed can happen in three ways: O DAM: 7-6-82 Appendix A, Page A-2
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER GENERAfiON Olvl510N
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( '/; TECHNICAL DOCUMENT U A f ailure of the Feedwater Control System to run back af ter 1. reactor trip.
- 2. An operator error of feedwater control while in manual.
- 3. Equipment failure wher. the plant is in automatic operation.
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Exc es sive feedwater can occur at any time the main feedwater system is (w' / in operation. The plant may be tripped or at power. The steam gene rator will fill at dif ferent rates depending on what the plant power level is when the high flow begins. The rate of fill of the generator will be greater when the reactor is at low powe r (or trip-ped) than at high power. The overcooling effects on the reactor coolant systen will be greater at low power. The reasons are that, at i l [ n) / low power, less core heat exists to boil off the additional feedwate r and the feedwater system (valves and pumps) has a lot of capacity left to overmatch the low reactor power. At full load, the valves and pumpa are near full capacity and cannot open much more to increase feed flow. l l
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Because the ef fect s of exc es s ive feedwater are dif ferent across the s /
\d power range and because it can be caused by different failures, the rate of the Reactor Coolant System res ponse will be dif ferent depend-ing on what has h appe ned . But all excessive feedwater additions will look similar.
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DATE: 7-6-82 Appendix A, Page A-3
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLEAR POWit GENERADON DivislON 74-1125531-00 TECHNICAL DOCUMENT The P-T curve and sequence of events shown in Figure A-1 depict a typical main feedwater transient that is terminated by the ICS before water enters the steam lines or the pressurizer is drained. The tran-sient shown is also applicable if terminated early by the operator. The P-T curve and sequence of events shown in Figure A-2 depict an ex-ces s ive main fe edwa ter transient that is not te rmina ted before water enters the steamlines or before the pressurizer is drained. The trans-ient is initiated by a reactor trip from 100% power with a failure of feedwater to runback on a steam generator. Several important points should be noted regarding this transient:
- The af fected steam generator can fill very rapidly, in this case three minutes af ter the reactor trip. Thus, if the failure caus-ing the excessive feedwater condition is not corrected by the ICS or by SFRCS actuation the operator has little time avellable to prevent spillage into the steam lines.
l - The operator is required to trip the RC pumps. AFW will control SG 1evels to 40" (93" on SFAS level 2 actuation) on the startup range. If the subcooling margin is lost without a SFAS level 2 actuation, the operator must manually raise the SG 1evel to 93" on l the SU range . The reduction in steam pres sure in the non - ove r fed SG may eventually result in SFRCS actuation, which finally termi-l nates the excessive MFW flow to the overfed SG by closing the MFW stop valves. DATE- 7-6-82 Appendix A, Page A-4
BWNP-20007 (6-76) , BABCOCK & WILCOX Nu sen NUCtfAR POWER GENERATION OlVisiON 74-1i2s331-00 TECHNICAL DOCUMENT
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Excessive MFW flow to one SG was allowed to continue for almost two minutes after the steam generator was full. Thus a signiti-cant ' quantity of water was spilled into the steam lines with the potential for severe consequences. Since the consequences are not
- known, this discussion does not describe those effects.
( - Even though the MU flow was increased and SFAS started HPI before the pressurizer was empty, the HP1 flow was not sufficient to over-come the shrinkage due to the cooldown and the pressurizer was drained.
- Once the overcooling transient has been terminated, the RCS will reheat and the water volume will swell. Since a large quantity of cold HPI water was added to the RCS, the operator must act to pre-vent the pressurizer from going solid. This will be discussed in more detail later.
Actual Plant Excessive Feedwater
- j. On November 16, 1980, Davis-Besse Unit I was operating at steady-state 74% rated thermal power (RTP) with the Integrated Control System (ICS) in the full automatic, mode. At 0318 hours, a failure in the ICS caused an OTSG overfeed transient. The plant operators placed the ICS in manual control (feedwater and reactor demand in hand) and stabi-lized the plant with low reactor coolant temperatures and a large AT c (difference between reactor coolant system (RCS) cold leg tempera-ture). Once the plant was stable, the control room operators pro-ceeded to restore normal plant parameter conditions over a period of about 30 minues.
Appendix A, Page A-5 DATE: 7-6-82
r BWNP-20007 (6-76) BABCOCK & WILCOX Nu.sER NUCLE AR POwth GENERAflON DivtStoN 74-1125531-00 TECHNICAL DOCUMENT During the transient one steam generator remained within the allowable operating levels while the other steam generator overfilled to approxi-mately 400 inches indicated on the full range level instrumentation. The water level was off scale (greater than 100%) on the operating range level instrumentation for 2.6 minutes. The reactor did not trip during the transient. ! Following the transient, the plant was maintained in manual control (reactor demand, A T c , both feedwater demands and main and startup con-trol valves in hand) for about ten hours while the cause of the trans-ient was investigated and isolated. Normal plant ope rat ion was re-l sumed at 1303 hours when the ICS was returned to full automatic control. I I Sequence of Events l The following sequence was prepared from the alarm printer, plant l l logs, and plant personnel interviews. l 03:18:40 ICS REACTOR PWR LIMITED BY FW 03:18:49 ICS UNIT MASTER IN TRACKING 03:19:11 RC PRZR PRESS LOW 2102 l 03:19:14 RC MU FLOW HIGH RC PRZR LVL HI LO 03:19:20 RC CLG LOOP 1 VS 2 TGEMP DIFF 5.304 03:19:24 ICS SG 2 BTU LIMIT 03:19:35 EHC High LOAD LIMIT 03:19:30 CRD MANUAL MODE i 7-6-82 Appendix A, Page A-6 D ATli -
BWNP-20007 (6-76) BABCOCK & WILCOX NumseR NUCLE AR POWER GENERATION DIVl$10N 74-112553i-00 [ i/ TECHNICAL. DOCUMENT i x-03:19:57 RC PRZR AVG LVL LO 151.4 03:20:02 SG 1 OPERATE LVL HIGH 03:20:11 SG 1 OUTR STM PRESS (925.0) HIGH 03:20:24 ICS HI LOAD LIMIT 03:20:35 RC LOOP 2 HLG PRESS NORMAL 03:02:36 SG 1 FULL RANG LVL HIGH
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l i 03:21:08 SG 1 MN STM TO RC HLG DIFF TEMP
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' s' 03:25:34 RPS CH 2 CH TRIP RPS CH 2 HI FLUX TRIP 03:28:58 EHC HIGH LOAD LIMIT Initiating Event The fe edwat er tran9ient was initiated when either the open limit switch on the loop 1 main feedwater block valve, field relay 33X, or
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( ) ICS relay 86-MBV-B failed. During plant startup when the main feed-xv water block valve is less than 20% open, ICS relay 86-MBV-B causes the startup feedwater flow signal from loop I to be sent to an integral and differential circuit which corrects the temperature compensated main fr dwater flow signal from loop 1. When the main feedwater block valve is more than 20% open, 86-MBV-B energizes, blocking the startup gy feedwat er flow signal and sends the temperature compensated main feed-I i
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, ,/ water flow signal itself into the integral and differential circuit which correct s the feedwater flow signal. Just prior to the transient the reactor power was 74% RTP and main feedwater flow was about 4.2 x 106 lbm/hr to each steam generator. Startup feedwater flow indication on loop I was probably about 1x 106 lbm/hr. Since the main feedwater n block valve was open, the transfer switch (ICS relay 86-MBV-B) was l
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~' sending the 4.2 x 106 lbm/hr main feedwater flow signal on loop 1 into l
the circuit which corrects the feedwater flow. When the ICS or ICS 7-6-82 Appendix A, Page A-7 DATE:
1 BWi4P-20007 (6-76) BABCOCK & WILCOX Nu sta NUCLE Aa POWEa GENERATION OtVl$loN 74- 12ss31-00 TECHNICAL DOCUMENT input signal failed, the transfer switch on loop 1 (ICS relay 86-MBV-B) switched inputs to the feedater flow correction circuit from the 4.2 x 106 lbm/hr main feedwater flow signal to the ] x 106 lbm/hr startup feedwater flow signal. The result was that the zero correc-tion to the t empe rature compensated main feedwater flow suddenly became a larger negative correction and the " corrected" flow for loop 1 sent to the rest of the ICS was significantly lower than the actual flow. Since the " corrected" flow was less than the feedwater demand on loop 1, the ICS responded by opening the loop 1 main feedwater con-trol valve SP6B to admit more feedwater to OTSG 1. Feedwater flow to OTSG 1 increased to greater than 7.0 x 106 lbn/hr (uppe r limit on the main feedwater flow indication) and remained above 7.0 x 106 lbm/hr for 0.5 minutes. Prior to the feedwater transient OTSG 1 operate level ind ic at ion was 37.4%. During the transient the steam generator i level went off scale at 100% on the operate range for 2.6 minutes. (The Shift Supe rvi so r 's log indicates the maximum level reached was 400 inches indicated on the full range level instrumentation.) Transient and Post-Transient Plant Response The plant performed as would be expected for this initiating event. As excessive fe edwat er flow was fed to OTSG 1, the feedwater flow to OTSG 2 decreased. The result was an overcooling of the reactor cool-ant (Tave) c'e c reas ed from 582.2F to 569.8F and a large imbalance in the heat removal by the two steam generators which resulted in a delta Tccf approximately 30F (maximum). 7-6-82 Appendix A, Page A-8 DATE:
BWNP-20007 (6-76) i BABCOCK & WILCOX NuusEn NUGEAR 70WEa GENesATION Olvi$ ION 7'- 12533i-00 (r \ m/) TECHNICAL. DOCUMENT v l Early in the t ransient, just after the steam gene ra to r i operate level ! was pegged high, the control room ope ra tor took manual control of the reactor demand and feed wa t e r control valves and was able to stabilize the plant without a reactor trip.
'\g) Strip charts monitoring the ICS signals indic a t ed the ICS performed its intended function and would have reduced feedwa ter flow rapidly as a result of the OTSG operating range high level signal if the ICS feed-water control stations had remained in automatic control.
The following Figures A-3 and A-4 show actual plant data from the above transient. Note the large disparity between levels and feed-(N ( ) (_/ ' water flowrates fo r the two steam generators. This magnitude of mis-match should, and did, facilitate rapid recognition and response by the operator. On 3/18/77, an exc es sive feedwa t er transient occurred at an ope ra ting plant. The transient was not serious and the plant ended in a good [n} \ condition because the operator recognized what was going on and took control quickly and the excessive feedwa ter was terminated automatic-ally by the ICS af ter reactor trip. The steam generator did not fill and spill water into the steam lines; that is the most impo rtant limit fo r this transient. The operator also turned IIPI on and prevented the pressurizer from draining and that is another impor tant limit. Be-A
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cause the cooldown was stopped early in this t rans ie nt , the ope ra to r (v ) DATE: 7-6-82 Appendix A, Page A-9
BWNP-20007 (6-76) EABCOCK & WILCOX Numset NUCLEAR POWER GENERATION DIVISION TECHNICAL DOCUMENT 74-1125531-00 did not have to take fast action to prevent filling the pre s sur ize r so lid due to reactor coolant reheating. After trip, the HPI flow was throttled. The plant was operating at 75% in a power esc ala t ion sequence. Be-c ause of the power escalation sequence, the overpower trip setpoint was set at 8 5% . The overfeeding started because main feedwate r pump "A" failed and went to full speed. The plant data shows main feedwa ter abruptly increasing to full flow on gene ra tor "A" and, within about 30 to 45 seconds, its a f fec t s appe a r in other parameters: "A" generator level goes up, T ave drops (because of the increaae heat transfer), pressurizer level drops (because of shrinkage due to lowered Tave), and RC pressure lowers (because of the lowered pre s sur ize r level). When T ave began to drop, ( the ICS pulled rods to try and keep temperature stedy. The operator l sensed the ch ange in plant conditions almost immediately and quickly l d iagnos ed the problem and - took the appropriate action. He tried to l He did this by manually reducing the feed wa ter cut back feed flow. demand; this did not work (more about this later). He then put rod control into manual to try to raise T,ye, and he also l started HPI to stabilize the pressurizer level. The reactor then O DATE: 7-6-82 Appendix A, Page A-10
BWNP-20007 (6-76) BASCOCK & WILCOX NUMBER NWCLEAR POWES GENERAh0N DIVISCN 74-1125531-00 TECHICAL DOCUMENT tr ipped at 85% power because of the low overpower trip setting. When the reactor tripped, the ICS switched into track and controlled the feedwater on level; because of the high steam generator level, the "A" , main feedwater valve closed stopping the transient (the "A" start-up [ _ flow also closed down) and th e "B" generator startup and main valves 6 k controlled to maintain the OTSG level after trip. When the plant t r i pped , the large inventory of cold water in the "A" generato r ,
. coolded down the reactor coolant considerably (illustrating the effect on core heat removal due to the boil of f of the extra inventory in the gene ra tor ) . The reactor-ateam generator heat transfer interplay is shown well by this example. After the trip, the reactor pressure l
l increased and the pressurizer level increased; this is largely due to HPI. i Steam pressure made a very slight reduction (%8 psi) (before trip) because of the cooling and condensing effect of the excessive feedwater on the steam in the . generator. This illustrates the magnitude of steam pressure loss to be expected because of excessive feedwa te r; a greater loss would indicate an additional failure that l the operator would have to correct. The "A" pressure is lower than "B" after trip mostly because of one mis-set safety valve; the water in the generator would have some pressure reduction effect, but it would be small. The lower steam pressure did have some ef fect on overcooling, but because it was only about 100 psi low, the ef feet was small. l l DATE: Appendix A, Page A-11 7-6-82
-, . . , . . . . _ _ . . _ _A._.__._._.__...____,.__..__.____._
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen NUCLEAR POWER GENEAAflON OlVISION 76- 125531-00 TECHNICAL DOCUMENT This steam pressure loss was mostly an inconvenience; if it had been greater (abmt 200 psi), the pressurizer could have drained. When the transient was over, the "A" steam generator was about half full. It took about two minutes to increase the level about 200 inches; most of the filling took place before trip when a high reactor power (75% to 85%) was available to boil the water off. The steam generator level increase was about 150" during this time (or about 3/4 of the total increase). If the main feedwat er failure had occurred after trip, the rate of level rise would have been much faster. Excessive feedwater is probably the one transient the operator must react to faster than any other transient. The steam gene ra tor can fill and water can spill into the steam lines in as little as 3 to 4 minutes (after trip). The operator must act fast to stop feed wate r i and the equipment he chooses to use is impo rt a nt . He shculd under-l l stand that the equipment he elects to use may be the component that failed caasing excessive feedwater and there fore may not re s pond . Thus he must be prepared to switch to an alternate device if necessary to terminate feed flow. Wnen the operator tried to correct this t r ans ie nt , he made the right choice of action: cut back feedwater. But the equipment he used did not r e s pond . He used the ICS feedwater demand to try to run back the feed pump and valves. A post trip review showed that the feed pump controller had failed. It was essentially dead, and no signal would DATE: Appendix A, Page A-12 7-6-82
BWNP-20007 (6-76) BASCOCK & WILCOX NUMSER Nucteam Powea oewmATON OMSON 74-i i 2533 i-00 TECNNICAL DOCUMENT have made it re s po nd . Exc es s ive main feedwat e r is a complex accident which can be caused by ebout 20 dif ferent equipment failures (operator error with the feedwater control in manual can also happen). The ICS 4 can have several failures. The accident can be too fast to try to figure out wh at failure has occurred. The re fo re , the best way to V correct a very fast overfill is to use the direct controls to trip both main feedwater pumps. Adequate time will be available to regain MFW or AFW to at lest one SG. Tripping both MFW pumps is the fastest method to stop exc es sive MFW flow and should prevent water entering the steam lines for even the most severe MFW transient. However, for much slower fill rates, or if a pump should fail to stop, the operator should isolate feedwater to the SG with the high, increasing level. l 4 The above transient had a large disparity between levels and feedwater ,l ) flowrates fo r the two steam generators. This magnitude of mismatch should, and did, facilitate rapid recognition and response by the
- O l
( ( ope ra to r. 2.0 OPERATOR ACTIONS
SUMMARY
l Immediate Actions Attempt to centrol the MFW control valves if the overfill is slow l and the af fected generator is obvious. I DATE: 7-6-82 Appendix A, Page A-13
BWP-20007 (6-76) , BABCOCK & WILCOX Nu sen NUCLEAR POWER GENERATION DIVISION 7'-i 25531-00 TECHNICAL DOCUMENT Trip run ning MFW pumps to stop fast overfills or if SG level greater than 95% or pressurizer drops below 50 inches. Start AFW rand verify operation;
- Increase MU flow if pressurizer level is less than 100" and RCS pressure is decreasing.
- Follow remainder of Part I, Section III.C, Excessive Heat Trans fer.
Identifying Symptoms
- Exc es s ive feedwater is an overcooling transient as shown in Figure A-5.
Other identifying symptoms to distinguish excessive fe edwate r from other overcooling transients are: High stean generator level
- High main feedwater flow.
NOTE: Rapid e;cc e s s ive feedwa ter trans ien t s, e.g., large MFW flowrate after reactor trip, will result in the pressurizer being in a near-drained condition by the time Th/ Pressure exceed the over-cooling boundary. Drainage and RCS saturation will occur very quickly after the overcooling boundary is exceeded (dotted path in Figure A-5) and water will enter the steam lines. The re fo re, it is very important that the operator recognize the overcooling transient before the ove rcooling boundary is exceeded by checking MFW flowrates anu SG levels. He should di sc ove r that exc es sive feedwater is in progress in Step 5.0 of Part I, Section II, " Vital System Status Verification". DATE: Appendix A, Page A-14 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER NUctEAR POWER GENERATION DIVisON 7en2ss2t-00 () TECHNICAL DOCUMENT i \. .r'
/
The previous section discussed three of the many possible examples for excessive main feedwater transients. However, it can be seen from that discussion that the primary transient of concern is the rapid filling of a steam generator that is not automatically terminated early by the IC S . SFRCS actuation will occur too late to prevent
-s
/ xt t
('% ) water in the steam lines and drainage of the pressurizer and the low _/ steam pressure SFRCS signal may occur in the non-overfed SG causing AFW to feed the overfed SG. Therefore, such a transient requires rapid response by the operator. In addition, the operator must exer-cise caution whenever feedvater is in manual cont.rol to prevent large feedwater mi sma tch es from developing. His toric ally , ove rs ight s while ex in nanual control have been a significant contributor to the frequency i
/ s \ l N /
of excessive feedwater tranaients. i This section will discuss how operator actions in accordance with Part I will terminate the overcooling transient and provide recovery to stable shutdown conditions. The assumed transient will be similar to the second transient discussed in the previous section. A reactor l [ trip occurs ( for whatever reasons) and the MFW control valve to a SG
/
'" f ails open.
Af ter performing the immediate actions of Part I, Section I, the opera-tor will verify vital system status in accordance with Section II. Step 6.0 of Section II requirer, the operator to verify that feedwater
,a I \
\
/
DATE: 7-6-82 Appendix A, Page A-15
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sen NUCLEAR POWie OtNERAflON OfVl560N 7'- 2n3i-00 TECHN! CAL DOCUMENT l l has runback. Ile should check steam genera to r levels and feedwater flow rates and note that level and flow of one SG are high. The cor-rect ive action noted is to trip the running feedwater pumps and verify proper operation of the AFW system. These actions will terminate the excessive feedwater and return the plant to stable conditions. Iloweve r , for illustative purposes it is assur.ed the operator fails to ! note the abnormal conditions at this time. Step 14.0 of Section II re-quires the operator to verify that primary to secondary heat transfer is not exces s ive . The operator should note by RCS response on the P-T curve that an overcooling transient is in progress and there fore pri-mary to secondary heat transfer is excessive. He does not need to con-cern himself at this time whether the overcooling is due to excessive feedwater or loss of steam pressure control. The procedure directs him to Section III.C of Part I. Step 1.0 of III.C requires the operator tc increase MU flow. While full MU flow will not maintain pressurizer level during this magnitude of shrinkage due to overcooling, it may slowdown the rate of pressuri-zer level decrease and provide additional time for the operator to ter-minate the overcooling transient before the pressurizer drains (see Figure A-6). Step 6.0, SFRCS actuation, will ef fect ively isolate the steam genera-tors from most failures that would cause overcooling and indeed, by DATE: 7-6-82 . Appendix A. Page A-16
. - . . . _ ~ , - - - _ . - . . . . _ _ _ - - - . . . _ _ _ . _ - - - _- - . -_- . . . - . - -
BWNP-20007 (6-76) BASCOCK & WILCOX NUaAttR wuctema powee oewmatow omsow 74-n 25331-00 TECMICAL DOCUMENT I isolating MFW in this particular overcooling transient will be terminated. 4 Maintaining RCS tenperature at the present value by adjusting the TBV or AVVs setpoint will prevent filling the pressurizer solid due to RCS I \ reheat ing and swell and is especially important since RCS inventory has been increased due to MU and HPI. However, HPI is still in pro-gress and must be throttled when the Subcooling Rule is satis fied . Establishing AFW flow to maintain OTSG 1evels will restore stable primary to secondary heat transfer. b If the operator follows Ge guidelines in an expeditious manner and performs the actions such that MFW is isolated within two to three I minutes following reactor trip, he will probably prevent water en-tering the steam lines and drainage of the pressurizer. In fact, the
- RCS will probably stay within the post trip window and recovery to i'
stable shutdown conditions can be quickly achieved. If, however, the cooldown is allowed to continue to the point of pressurizer drainage, the RCS will rapidly approach saturation conditions and further opera-v
) tor actions and precautions are in order.
When the subcooling margin is lost the operator will trip the RC pumps, verify AFW initiates, and control SG level to 93" on SU range. i l l' i DATE. Appendix A, Page A-17 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX NUM4tR NUCitAR POWit GENERAflON DIVl5 TON 7'+-n 2 5 53 t-00 TECHNICAL DOCUMENT When the overcocling transient te rmina tes , MU and HPI flow will re-cover pressurizer level and the pressurizer heaters will recover RCS pr e s s ur e . The operator must again respond to present reactor coolant swelling due to reheating, from filling the pressurizer. He should perfonn the following actions: Control the SG pressure to a value near the co r res ponding satura-tien pressure for the .xisting cold leg temperature. This will limit RCS heatup and thus limit the resultant swell of the RCS in-ventory. (If SFRCS has actuated, the MSIV's and AVV's will shut; the re fo re the operator must block the AVV close signal and control steam pressure with the AVV's.) 3.0 EXCESSIVE MAIN FEEDWATER WITH OTHER PLANT FAILURES Introduction The previous section de sc r ibed exc es s ive main fe edwa t er in general, but did not discuss other f ailures that might also happen at the same t ime . This section will show what symptoms to look for when other equipment fails and will show what steps the operator should take to restore the heat t rans f e r from the core to the steam generators. The event that was chosen for simulation starts with the reactor at 100% power; a failure in the feedwater system allows main feed to run away in one generator; automatic ICS corrective acticn to control main feed-water does not happen; the plant is tripped automatically on high f'.ux or low RC pressure, and excessive feedwater continues. All the data that is shown starts from the time of plant trip. DATE: Appendix A, Page .A-18 7-6-82
N
, BWNP-20007 (6-76) . BABCOCK & WILCOX NUMSit NUCLEAa POWEs OsNORAh0N OM$ ION )
74-t i 25531-oo :
; ^ TECHICAL DOCUMEllT Remember that all feedwater transients will not start from high power.
They may look dif ferent than the examples used. The reason for these , examples is to provide understanding, so close study of the ef fects is required. Branch Discussion Figure A-7 has separate f ailure branches for loss of reactor inventory control (high and low), loss of secondary inventory control (high and low), and loss of secondary pressure control. Significant failures in RC pressure control, such as those due to overcooling or excassive MU, are adequately covered by these branches ; therefore separate branches i specifically for loss of RC pressure control are not shown. Minor failures, such as loss of pressurizer hea t e rs , are discussed at the end of the main transient path. This section will discuss each of 4 i these additional failure branches and illustrate how operator actions in accordance with the procedures in Part I and with the " Equipment Ope ra t ion" chapter in Part II, will restore proper control of the para-meter in question.
- These branches are structured to address the particular function fail-ure in question even though the excessive feedwate r transient may still be in progress. However, since some additional failures result in further overcooling of the RCS, actions to correct such failures may also correct the excessive feedwater condition. In any case, it should be understood that Figure A-7 and this discussion are provided l
l DATE: 7-6-82 Appendix A, Page A-19
BWNP-20007 (6-76) BABCOCK & WILCOX ,,n NUCLEAR POwta GENfaAftoN OfVl5 TON 74-1125531-00 TECHNICAL DOCUMENT as tools to promote familiarity with expected plant re s po ns e s . An-other valuable tool to facilitate operator recognition and identifica-tion of overcooling transients is Figure 22, "tfercooling Diagnosis Ch a r t ," in Part II, Section I.C. The operator should become familiar with this chart. Figure A-6 is provided to show key distinguishing parameters for exces-O sive main feedwater that have a time dependency important to the opera-tor in identifying both the type and severity of transient. The para-meter plots ahow typical responses to a large excessive main feedwater transient. Arrows, where used, show the e f fect of other failures and operator actions on the time relationship. O One iten of particular note on Figuire A-6 is the ef fect of large ex-c es s iv e feedwa t e r transients on steam pressure in the unaffected gene-ra t c,r . The pressure is reduced because the primary system has been cooled so rapid ly that the unaf fected SG becomes , tempo ra rily , a heat i source and loses heat (and thus pressure) to the primary system. The
- pressure in the overfed SG does not show a pres sure decrease because l
the presure reduction affect caused by RCS cooling of the steam is of f-set by the presure increasing af fect of the rising SG level. i Loss of Reactor Inventory Control (High) A loss of reactor inventory control (high) exists whenever makeup or i HPI flows are exc es s ive causing the pressurizer to fill and overpres-surizing the RCS for the existing plant conditions. Severe excessive DATE: 7-6-82 Appendix A, Page A-20
d BWNP-20007 (6-76) SABCOCK & WILCOX NUMB ER NUCLEAs POWit OtNERAilON DWillON
^ 74-112ss3t-00 TECHICAL DOCUMEllT feedwater transients involving large mismatches between feed flow and l primary heat input will result in RCS cooldowns and shrinkages that I
cannot be compensated for by full HPI and MU flow. Thus, while the ex-cessive feedwater transient is in progress, pressurizer level will con-i cinue to drop. Howev e r , for smaller feedwater transients, and when ! O the excessive feedwater has been terminated, full HPI and MU flow will f more than compensate for the coolant shrinkage and result in a rapid 9 increase in pressurizer level and RUS pressure. Rapid operator re-sponse may be required to prevent a solid pressurizer and RCS overpres-surization. The HPI will not pressurize the RCS above N1700 psig but the MU will. The operator should perform the following actions to restore proper RCS inventory and pressure control:
- 1) Throttle HPI as soon as the subcooled margin is restored and RCS pressure is increasing.
- 2) If the RCS is reheating and thus swelling, lower the TBV (or AVV 4
if the MSIV's are shut) setpoint to a value near the corresponding saturation pressure for the existing cold leg tempe ra ture . This g ! will stop the RCS heatup and swell. If desired, the operator can l then gradually increase the se tpoint to allow a gradual heatup while contro11in-c pressurizer level.
- 3) When pressurizer level returns on-scale (with the RCS ebove the :
subcooled margin) and is increasing above 100", the operator should terminate HPI and realign for normal makeup / letdown
)
operation. DATE: 7-6-82 Appendix A, Page A-21
\
B'JNP-20007 (6-76) BABCOCK & WILCOX Nuesta NUCttAt POWER GENERATION OlVl510N 74-n2ssai-00 TECHNICAL DOCUMENT NOTE: Throu ghou t Part I the operator is required to throttle HPI as soon as the subcooling margin is restored and to reduce TBV or AVV se t po int s to maintain RC temperature. Thus, adhering to these guidelines will prevent a loss of RCS inventory control. This is discussed in more de t ail in Part II, Section I.E.,
" Equipment Operation."
Loss of Reactor Inventory Control (Low) A loss of reactor inventory control (low) exists unenever makeup flow is insu f ficient to overcome a primary leak rate or the coolant contrac-tion rate, resulting in drainage of the pressurizer. As stated pre-viously, full makeup flow will be insu f ficient to maintain pressurizer level during severe excessive main feedwa te r transients, but will re-fill and repressurize the RCS once the overcooling is teminated. Too little makeup or HPI flow, while undesirable, is not a major con-cern for this particular transient. If the overcooling is terminated be fore the pressurizer empties, the RCS will reheat and the resultant sweli will restore pressurizer level. If the ove rcooling continues, SFAS will actuate and HPI will initiate. It is extremely unlikely that at least one HPI pump will not start; however, should that occur the RCS will lose subcooling mat gin. The opera tor will trip the RC pumps and AFW will start. O DATE* Appendix A, Page A-22 7-6-82
BWNP-20007 (6-76) SABCOCK & WILCOX Numeen NUCLtAA POWte OtNERAffoN Olvt$10N 74-n 2 n a t-00 l 3 TECHillCAL DOCUMENT Control of AFW to attain and maintain 93" level on the startup range will provide adequate core cooling while the problem with HPI is being corrected. 1 Following the actions specified in III. A of Part I will restore pri-I mary system inventory control and subcooled margin. 2 Loss of Secondary Steam Pressure Control
- A loss of steam pressure control exists whenever one or both steam generators und a rgo a pressure reduction significantly below the TBV reseat setpoint. It is an overcooling transient and will look similar (on the P-T diagram) to an excessive feedwat e r transient. It can be caused by an unplanned steam flow through stuck open valves or a pipe break. Improper AFW flow control could also result in a SG pressure reduction because of the cold AFW spraying into the SG steam space.
l The operator will isolate both SG's and then monitor their respective } 1evels and pressures. If both SG's stabilize, indicating the steam l . leakage path is downstream of the MSIV's, AFW will be supplied to both l l SGs. If only one SG stabilizes AW will be supplied to that SG for DH r removal and the broken SG will boil dry. It should be noted that the overcooling caused by the excessive feed-water coupled with the overcooling due to loss of secondary pressure s 1 DATE: 7-6-82 Appendix A, Page .A-23
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLEAR POWtt OtNEEAiloN DIVISION TECHNICAL DOCUMENT 74-n 2ss31-oo control may be top rapid and too severe to prevent pressurizer drain-age and s9turation of the RCS. Figure A-6 shows the impact of exces-s ive feedwater overcooling compounded by overcooling due to loss of steam pressure control. The curves for RCS pressure and pressurizer level will shift to the left, i.e., RC pressure reduction and drainage of the pressurizer will occur faster. When the overcooling transient is t e rmina t ed , the operator must react to prevent overpressurization of the RCS and possible violation of the Reactor Vessel P-T limits. Loss of Secondary Inventory Control (High) A loss of secondary inventory control (high) due to excessive AFW, is very similar to the main initiating event covered in this section (excessive main feedwater). However, there are basic difference in plant re s po ns e . Exces s ive AFW can occur simulatneous with excessive MFW or separately after excessive MFW is stopped. Exc es s ive auxilia ry feedwater will cause depressurization of the a f-fected SG to a much larger extent than excessive main fe edwa t e r . This is due primarily to the condensing action introduced by spraying AFW l in near the top of the tube bundle (into the stema space) and due to the lower AFW temperatures. AFW flow can cause overcooling of the primary system if it raises SG level above the control setpoint. O DATE: 7-6-82 Appendix A, Page A-24
1 i BWNP-20007 (6-76) BABCOCK & WILCOX NUMBEe [ wuctema powee oewesAnow o.visiog 74- u 25531-00 TECHNICAL DOCUMENT t Should AFW flow control fail, the operator should recognize the over-cooling as well as high AFW flow with the SG level higher than the ap-propriate se t po in t. Following the actions in Part I, Section III.C. for excessive primary to secondary heat transfer will terminate the exces sive AFW. The operator should not re s to:, e AFW to the generator with high level until the failure causing excessive AFW has been iden-s tified and corrected. Restoration of AFW to the " good" generator will provide DH removal . The operator should align the AFW system to allow feeding of the good generator with both AFW pumps. Figure A-6 shows the impact of excessive main feedwater overcooling compounded by overcooling due to excessive AFW. The curves for RCS (q) pressure and pressurizer level will shift to the left, i.e., pressure y/ reduction and drainage of the pressurizer will occur faster. l Loss of Secondary Inventory Control (Low) l A loss of secondary inventory cer. trol (low) exists whenever too little feedwa ter is being supplied to the stean generators resulting in too little primary to secondary heat transfer and overhe' ating of the RCS. This is an unlikely event since the initial condition was excessive l-l main feedwater with too much primary to secondary heat transfer. In any case, should a total loss of both main and auxiliary feedwater sub-sequently occur, the operator will have more time available for correc-tive actions due to the initial SG inventory increase caused by the excessive main feedwater transient. A detailed discussion is provided in Appendix B, " Loss of Main Feedwater." DATE: Appendix A, Page A-25 7-6-82
s Figure A-1 EXCESSIVE FEEDWATER TERMINATED' BY ICS 2600 POST TRIP ~
~ ~'
2400 - glN00s - - - - - j l 2200 - !
\ l4 r- - - ~ 1 *
? 5 L. ..; I S
2000 - SUBC00 LED I 2 - SUPERHEAT REGION l j REGION 3 y 1800 - i l [ 5 1600 - 5 0
*- i 3.- 1400 -
u, I a "' END POINT POST TRIP WITH
= AFFECTED SG E -
FORCEO CIRCUL Ail 0N # TH0T 1200 STE AN PRESSURE - tTCOLD) AND FOR NATURAL l
'I' E, 3,4 CIRCULATION tiCOLD3
= iOOO -
<n___ __ .
l NORMAL OPERAllNG POINT POWER [= OPERAil0N ITH0T) g 800 - 2 2 STEAN PRESSURE SATURAil0N f- ~ 'l END POINT POST TRIP WITH
- LIMIT > '_
t _ j NATURAL CIRCULATION eTH0T3 600
/- SUBC00 LED 400 MARGIN LINE 400 450 500 Reactor Coolant ana Steam Outlet Temperature. F Reference Time Points (Seconds) Remarks 1 0 Excessive feedwater addition begins.
1-2 0-60 Slight overcooling of RCS occurs due to excessive feedwater addition. ICS pulls rods to compensate for reduction of Tave-2 60 Reactor trip on high flux or low pressure. (Note: Depending on severity and power history, a reactor trip may or may not occur.) 2-3 60-200 RC P&T decrease due to loss of fission power and higher than normal secondary inventory. The ICS initiates a feedwater runback and the MFW addition stops. Pressurizer level decreases because of reactor coolant contraction. 3 200 Minimum pressurizer level reached. l 3-4 >200 Normal system pressure restored by RCS reheating, operation of MU system, and pressurizer heaters. Primary system is lef t in a stable, hot shutdown condition. *
. f , J - l 1 2 5 3 ,9 l -00 i I
5
i Figure A-2 EXCESSIVE FEE 0 WATER NOT TERMINATEC BY ICS 2600 POST TRIP 440u - e W00s 2700 - i p---)3 t _ _. ; DI g 7300 - SUPERMEAT SUBCODLfD
#E GION j
% REGION C 7 l 1800 -
3
.~. i 4 1 5 1600 -
a - E a
~
OVERFE0 SG E
, UNAFFECTE0 SG ENO POINT POST TRIP elfH
% fonLEO CIRCUt All0N . i s0; 1700 -
STEAM PRESSURE B3 1ICOLO AND FOR NATURAL 5 5 uli I Cut afl0N T C0t 0 ' u 5 1000 . - - - - - - L.---- - --/
- ,, < , 6
/ U NORM AL CPER AilNG POINT P0ef R f 800 -3 5 m
hGPERAft0N*TH01' L SATURAil0N END PolNT POST TRIP Wi1H 600 - 5 di 't_ j NATUR AL CIRCUL ATION T i H0i'
- iUBC00tfD 400 MARGih LINE O i i L i n 400 450 500 553 600 650 '00 Reactor Coolant and Steam Outlet Temperature.*F ReferencS Time Points _ Mconds) Remarks 1 0 Reactor trip from 1005. MFW begins to run back but MFW control valve for one SG fails open causing over-feed of the SG.
.2 60 Operater senses rapidly decreasing pressurizer level and RCS pressure and increases MU flow.
3 180 Overfed SG full. 4 250 Pressurizer empty. RCS rapidly approaches saturation. On loss of subcooling margin the operator trips RC pumps and AFW starts. 5 330 AFW flow and RCS cooling of non-overfed SG causes steam pressure to decrease and initiate SFRCS. Subsequent closure of MFSVs will stop the MFW flow and teminate the excessive MW. 6 410 The excessive MFW has terminated and the AFW is con-trolling SG 1evel at 93" on the startup range. Refill of the pressurizer by Hp! has begun to repressurize the RCS and subcooling is regained. 71-112E53I-00 '
Figure A-3 ACTUAL PLANT RESPONSE DURING AN ' EXCESSIVE MAIN FEEDWATER TRANSIENT (RCS AND FEED FLOW PARAMETERS) 24.000 - 26.000 23.500 - _- 24.000 -
=
g ~ PRESSURIZER S O 23.000 - 0" 22.000 - LEVEL
\
~ 5 3 20.000 A \
22.500 -
= _.
22.000 - j18.000 - S - a
" -- e n.
21.500 -
- - 16.000 -
=
E $ 21.000 - t 14.000 - 20.500 - 12.000 - RC PRESSURE 20.000 - 10.000 ' ' - - - ' ' ' 40 40.800 41.280 41.760 42.240 42.720 I Time. seconds (x103 ) 70.000 g . 63.750 - l SGI g 57.500 - 2 a 51.250 k 5 w
'45.000 -
- , : _ ^:,-^^ ,
g 38.750 - G SG2 32.500 - 26.250 - 1 20.000 ' 40.800 41.280 41.760 42.240 42.720 Time, seconas (x 103) 74-.1-1255:11-00 " L.__ l
Figure A-4 ACTUAL PLANT RESPONSE DURING Arl' EXCESSIVE MAIN FEEDWATER TRANSIErlT (STEAM GENERATOR PARAMETERS) 10.000 n _- 9.750 -
= 9.500 -
< Sg ;
9.250 - 9.000 - H. h
$^n $:l:^ %
= 8.750 -
SG 2 5
= 8.500 -
E 8.250 - 8.000 ' ' ' 40.800 41.200 41.760 42.240 42.720 3 Time, seconos (10 ) 10.000 -- 9.000 - S" 8.000 - , y SG 1
$ 7.000 5 6.000 -
a 5 5.000 __ _ --- d j SG 2
= 4.000 -- -' J S
j 3.000 - o y 2.000 - 1.000 - 0.000 - 40.800 41.200 41.760 42 240 42 270 Time. seconos (x103) 7 . - 1 1 2 Ti.5:1 1 - 0 0 j
r Figure A-5 TYPICAL EXCESSIVE FEEDWATER 5 TRANSIENT 2600 2400 - T POST TRIP h 2200 - WINDOW r- - - 0 L___ { { 2000 - 2 SUBC00 LEO - 1800 - REGION ,# SUPERHEAT REGION 5 1600 -
/ ,_
g a O 1400 - l l [' [c 1200 - l STEAM PRESSURE ] END POINT-POST TRIP WITH FORCEO j
$ 1000
_ LIMIT 3_ .__/ CIRCULATION (TH0T &TCOLD) AND FOR NATURAL ClRCULAT10N (TC aLO) 800 - 3 @p g NORMAL OPERATING POINT-POWER 600 8 S3 OPERATION (TH0T) { I l END POINT-POST TRIP WITH 400 A G N L NE '-- NATURAL CIRCUL ATION (TH0T) l I e i i 450 500 550 600 650 700 400 Reactor Coolant And Steam Outlet Temperature-F e *9
Figure A-6 TIME RELATIONSHIP 0F KEY PARAMETERS EXCESSIVE MRI TRANSIENT n200 TRANSIENT TERMINATED 5 TRANSIENT BEFORE *
- TERMIN ATE D
.- PRES $UtilER
- BEFORE
- DRAINS
- PRES $URIZER 120 ORAINS 2000 - d 2
7 m
~
a ' d
/ =
a l : / U
~
LOSS OF STEAM
/
PRESS. CONTROL OR + INCREASE " - LOS$ OF STEAM PRESS. CONTROL OR + INCREASE EXCESS AfW My C EXCESS AFW gg j ORAINED t i 1 1000 e i 0 1 2 3 0 1 2 3 Time, Minutes Time, Minutes 1800,- OVERF H SG. 600 - h -
)
L - - o N UNAFFECTED SG
= (00E TO OVER
," - - - =
- COOLED PRIMARY)
~
400 -
\
~
TRANSIENT .
\
$ TERMINATED i J a BEFORE PRESSURIZER
=
" /
" DRAINS p 900 -
200 -
! E
, i i 800-0 0 1 2 3 2 3 ri,,, ,,,,,,, Time. Minutes 741-1125531-00 t
I l l I
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_A_PPENDIX B LOSS OF MAIN FEEIMATER 1.0 CENERAL TRANSIENT DESCRIPTION Loss of main feedwater is a failure to control secondary inventory. (q)'
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xs Loss of Main Feedwater (LOFW) is not be itself a severe transient. The plant was designed to respond automatically to this event so that the important plant parameters, like RC pressure and temperature, will stay within acceptable limits. l l However, when main feedwater is lost, some important backup systems [~x ( ) l \m./ like the Auxiliary Feedwater System or startup feedwater pump are l l called into play. If these backup systems fail to function cor-rectly, a much more severe transient may start. The important plant 1 parameters may go outside their limits and the operator will have to step in to control them. So even though loss of main feedwater is not by itself a bad transient, it can be the starting point for some os severe abnormal transient.
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This section will discuss wh at can cause a loss of feedwater and how the plant will behave, m f
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V 7-6-82 Appendix B, Page B-1 DATli.
BWNP-20007 (6-76) BABCOCK & WILCOX uvon, NUCLE AR POWit GENERATION DIVlilON TECHNICAL. DOCUMENT 74-1223531-00 Causes for a Loss of Main Feedwater A loss of MFW flow actuates SFRCS on high steam to feedwater differen-tial pressure. An ant ic ipato ry reactor trip occurs when SFRCS actu-ates. The anticipatory trip helps reduce the severity of the trans-ient because it quickly drops the production of heat in the reactor trips on high RC pressure. LOFW can also be caused by inad ve rt ent closure of the main feedwater control valves on the main feedwater isolation valves. If these fail-ures occur, they may not result in an anticipatory trip. However, they are much less likely than a main feedwater pump trip. The most common cause for a loss of main feedwater is a trip of the main feed-water pumps. So fo r the remainder of this discussion, we will assume that the LOFW has been caused by a main feedwater pump trip and that an anticipatory reactor trip also occurs. Plant Behavior Following a LOFW i For the first few seconds following a LOFW, RC pressure, pressurizer level and steam pressure all " spike" upwards. This happens because within a second after the main feedwater pumps trip, the reactor and turbine also trip, and the effects of the turbine trip are seen more quickly than the reactor trip. The turbine trip causes steam pres-1 sure to increase and reduces primary to secondary heat removal. The l reactor coolant gets hotter and expands slightly causing an increase in pressure and pressurizer level. On the P-T diagram of Figure B-1, 1 7-6-82 Aerendix B, Page B-2 D ATli .
BWNP-20007 (6-76) BABCOCK & WILCOX uumiin NUCLEAR POWtt OtNWATION OtviltON 74- 12n31-00 3 TECilNICAL DOCUMENT this initial spike is shown (greatly exagge ra ted) to indicate its direction. In actual practice the spike is so small and so quick that the operator probably won' t see it. (Note: Only for the case where there is no anticipatory trip does the s pike become signifi-cant. In that case RC pressure will rise to 2300 psig before causing s a reactor trip). Quickly the ef fects of the reactor trip become dominant. The loss of power generation in the core ends the overheating spike. As the stean ge ne rato r inventory is boiled away, heat is removed from the secondary side faster than it is being supplied by decay heat in the core and the RCS cools. This part of the transient takes much longe r and will probably be the first system behavior observed by the opera-tor. The P-T diagram of Figure B-2 shows the trend that he will see. Reactor coolant pressure, temperature and pressurizer level will fall as a resul t of the RCS cooling down. Pressurizer heaters will come on and makeup flow will automatically increase. Stean generator levels in both generators will continue to drop as
' the inventory is boiled away. At the same time, steam pressure will peak as the main stean safety valves lif t.
Because the reactor trip and turbine trip occur so quickly af ter the main feedwater pumps trip, the transient up to this point will look b' almost identical to a normal reactor trip on the P-T diagram. For DATE: 7-6-82 - Appendix B, Page B-3
BUNP-20007 (6-76) 1 BABCOCK & WILCOX Nueste NUCLE AR POWit GENER ATION DIVl510N 74-ii23331- w TECHNICAL DOCUMENT pos i t ive identification of a LOFW, the operator must rely on other in-dications such as the main feedwater pumps have tripped or the auxil-inry feedwater pumps h ave been turned on by SFRCS. Main feedwater flow rate is a fast indication which also should be checked. The auxili ary feedwater pumps are actuated early in the transient, but no AFW flow will occur until the steam generator levels have dropped to the low level setpoint. When this occurs, the AFW flow will be automatically controlled to maintain a nearly constant level and the heat removed from the steam generator will closely match the core decay heat. When this happens, RC tempe rst ure will remain nearly constant. Pressurizer level will slowly increase in response to the additional makeup flow, and operation of the pressurizer heaters will cause RC pressure to go up. This represents the final stage of the LOFW transient as the system approaches its desired post-trip cond i t ior.a . The P-T diagram of Figure B-3 shows the ex-pected system behavior during this period. 2.0 OPERATION ACTIONS
SUMMARY
Major Operation Actions l
- Confirm auxiliary feedwater system starts and is providing proper steam generator water level.
i Verify proper SFRCS actuation. 1 O i l Apr.endix B, Page B-4 f DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nu.s t a NUcttAt POWER GtNta ATION DIVI 560N 74-1125531-00 TECHNICAL DOCUMENT Identifying Symptoms Main Feedwater Flow Rate Low
- Main Feedwater Pumps Tripped Alarm
- AFW controlling OTSG levels NOTE: The RC pressure and temperature behavior is almost identical to a " normal" reactor trip.
If the plant responds normally, there is no need for the operator to take any immediate actions other than his normal post-trip response (Part I, Sect ions I and II). If the loss of main feedwater is com-pounded by other failures, Sections II and III of Part I will help the operator identify these situations and t ake the appropriate actions. 3.0 LOSS OF MAIN FEDWATER WITH OTHER PLANT FAILURES Introduction The previous section describes the loss of main feedwater without additional plant failures. The plant is designed to automatically handle the simple loss of main feedwater events without immediate operator action. However, a number of other plant failures can also occur at the same time which will compound the LOFW event and in-crease the complexity of the transient. These compound events re-quire operation recognition and correct ive action to mitigate the DATE: 7-6-82 Appendix B, Page B-5
BWNP-20007 (6-76) BABCOCK & WILCOX NU AA 6 E R NUCL( AR POWER GENERAflON DIVISION 74-u 2s s a i-00 TECHNICAL DOCUMENT transient. This section will show wh at symptoms to look for when other equipment fails and will show what steps the ope rato r should take to restore the heat trans fer from the core to the steam generators. There are three significant failures wh ich may compound the loss of main feedwater event. These are:
- Loss of Secondary Inventory Control (Low)
- Loss of Secondary Inventory Control (High)
- Loss of Secondary Pressure Control These events are shown on the Loss of Feedwater Logic Diagram (Figure B-5) and are discussed separately below.
1 j Loss of Secondary Inventory Control (Low) Loss of niain feedwater is already a loss of secondary inventory con-trol. If the LOW is compounded by a failure of the AW system, the stean generators will dry out and a loss of heat transfer will result. The operator can recognize the total loss of feedwater by the lack of both MW and AW flow indication on the flowmeters and by the low steam generator level which will be decreasing below 35 inches on the startup range. Secondary steam pressure will decrease below the TBV setpoint of 1015 psig once the steam generator dries out and can no DATE. Appendix B, Page B-6 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX uu,o,, wucteAn Powta GENetADOd Olvi$lON 74-1125531-00 (
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\"'/ longer produc e e nough steam to hold the se t po int pressure. Also, by following the steps in Part I, Sections I and II, the operator will identify that inadequate primary to secondary heat transfer exists at Step 13 of Part I, Section II. This step directs him to Section III.B of Part I where instructions for es t ab li shing HPI cooling, and p
(%/ ) for restoring proper primary to secondary heat transfer including u s ing the startup feedwater pump are given. MU/HPI cooling should be started as soon as possible when the loss of heat transfer is noted (Tcold decou ple s from steam gene ra to r satur a t ion tempe ra ture and steam pressure drops). Also at that time all but one RC pump should be t ripped; if the subcooling margin is lost the remaining RC pump should be stopped. The operator should also continually try to re- [\ \ store primary to secondary heat transfer,
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w Figure B-4 is provided to show how RC hot leg temperature and RC pres-sure typically change as a function of tima. On a loss of main feed-water with no AFW both RC temperature and pressure respond initially like a normal LOFW trip but then within minutes both parame t ers in-(~~N\ crease due to the inadequate primary to secondary heat transfer. j
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Loss of Secondary Inventory Control (High) The loss of main feedwater may be compounded by a failure in the AFW system res ul t ing in AFW ove r fe ed . This is an ove rcooling event be-cause too much primary to secondary heat removal occurs; however, it n
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[ ) is di f fe rent from the exces s ive feedwater event of Appendix A because (j DATE: 7-6-82 . Appendix B, Page B-7
BWEP-20007 (6-76) BABCOCK & WILCOX NUCLEAR Powie GEN!aADON Divl500N 74-1125531-00 TECHNICAL DOCUMENT l main feedwater has tripped off and cannot be c au s ing the ove r fe ed . AFW will fill the SGs slower than MFW. Ilowever, because the AFW sprays into the SG s t ean space it will cause a quicker decrease in SG pressure than MFW will. Exce s s ive auxiliary feedwater can be recognized by the following symptoms: liigh stean generator level in one or both generators Continuous AFW feed flow indication above the correct level setpoint in one or both generators
- Falling steam generator pressure in the " bad" (overfed) OTSG By following the instructions in Part I, Sections I and II, the opera-tor will identify the excessive primary to secondary heat transfer at Step 14.0 of Section II. This step directs him to Part I, Section III.C. Following those actions will terminate the runaway AFW. A more de t ailed discussion of Loss of Seconda ry Inventory Control (Iligh) is provided in Appendix A.
Figure B-4 shows the behavior of RC hot leg temperature and RC pres-sure versus time for the LOFW event compounded by excessive AFW. S tean pres sur e will decrease in both OTSG's and RC cold leg tempera-ture will follow stean generator Tsat throughout the transient. O DATE: Appendix B, Page B-8 7-6-82
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SASCOCK & WILCOX NUM O tt NWCLEAR POwes OsNOBATON DM560N 74- t i 2553 i-00 TECHNICAL DOCUMENT Loss of Secondary Pressure Control (Low) i A loss of main feedwater event may also be compounded by a loss of secondary pres sur e control. The loss of seconda ry pressure control I can be caused by such things as a turbine bypass valve or a main steam safety valve failing in an open position. Any of these initi-j O ating causes will result in an overcooling transient. However, over-cooling will only last as long as the SG inventory lasts. Therefore, 1 I loss of secondary pressure control is a significant failure to the loss of main feedwa te r event only if feedwat er flow is restarted to the SGs, e.g. , AFW or SU feedwater pump operation, s Symptoms of a loss of secondary pressure control caused by steam leak-I age through an open safety or turbine bypass valve are: (1) rapidly falling steam generator pressure in both OTSG's with the " bad" genera-tor pressure falling more rapidly; after a few minutes the " bad" gene-rator pressure may be as much as 200 psi lower than the good genera-tor pressure, and (2) low steam generator level in the " bad" generator, s The operator should follow the instructions in Part I, Section I and II. Step 14 in Section II will identify the overcooling event and direct him to Section III.C which provides the explicit instructions i for identifying and dealing with the steam leakage. DATE: 7-6-82 Appendix B, Page B-9 l - - _-_ __. . _ _ __ _ _ _ _ _ _ _
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBER seucttAR POWER GlNERATION DIVI $1c N 7'-i l 2 n 31-00 TECHNICAL DOCUMENT These steps. include isolating the " bad" OTSG and if the leak cannot be isalated then allowing the " bad" OTSC to boil dry. With a loss of secondary pressure control caused by a stuck open tur-bine bypass valve or main steam safety valve, the reactor coolant hot leg tanperature and pressure decrease very rapidly. The time depen-dence of these parameters for a stuck open MSSV is shown in Figure B-4. O I O O DATE: 7-6-82 Appendix B, Page B-10
Figure B-1 LOSS OF FEEDWATER (INITIAL STAGE) 2600 2400 POST TRIP 3 g 2200 - 00s
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h $UBC00 LED
! 2000
- REGION y -
SUPERHEAT k 1000 _ REGION E 1600 - o g a ' 3 1400 - c, = . [ 1200 ~ STEAM PRESSURE LIMIT 3 ENO POINT-POST TRIP ElTH FORCEO $ 1000 -
% CIRCUL ATION (T H0T &TCOLO) ANO FOR g a NATURAL CIRCUL ATiON t T C0t0 3 5 800 -
+
E m 600 - 5*[ ' NORM AL OPERATING POINT POWER OPERATION (TH0T) 400
-- SUSC00 LING MARGIN LINE
[' _j CIRCUL ATION (TH0T) I f i f f 400 450 500 550 600 650 700 Reactor Coolant Ana Steam Outlet Temperature-F 7 l - l l 2 Ti.5 3 I OO
Figure B-2 LOSS OF FEEDWATER (MIDDLE STAGE) 2 2600- , 2400 Post Trip
- Winaos 2200 r___
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[ SUPERHEAT SUBCOOLED E 2000 REGION 3 REGION - i U g 1800 -
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E 1600 5 >= R 5 y 1400 - o E 1200
- STE AM PRES $URE -
3 J END POINT POST TRIP WITH FORCEO g 1000 -
) LIMIT / CIRCut ATION ( TH0T &TCOL O) AND FOR N ATURAL CIRCUL Ail 0N ( TCOL D 8 0 800 -
" . F NORM AL OPERATING POINT P09ER 600 Q*[ { OPERAil0N (TH0T) r- 1 END i
POINT-POST TRIP WITH NATURAL L SUBC00 LING MARGIN i 400 t.lRCUL Ail 0N ( TH0y) LINE '-J t i i i I 400 450 500 550 600 650, 700 R e'a t t a t Coolant Ann Steam Outlet Tempeiature-F 9 7t- il25531-00 J. 1
Figure B-1 LOSS OF FEEDWATER (FINAL STAGE) L 2600 , OSI IRIP 2400 - -- WIND 0s i g c _:m / y 2000 - SUBCOOL10 /
- d REGION SUPERHEAI REGION 1800 -
U h 1000 - o _ . i 1400 - N
- END POINI POSI TRIP glIH
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% f 0 hcl 0 CIRCUL All0N ilH01 E
1200 - SitAM PRESSURE LIMil h&I COLD, AND FOR NATURAL CIRCULAll0N ilC010 3 o o 1000 - 3
- NORM Al 0PE R AllNG POINI 800 -
y Pust R OPE H AI10h iIH0I' E
- SaluRAll0N r7 [NO POINI POSI IRIP silH 600 -
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$00 550 600 650 100 400 450 Re.it. lo s Coo l .in t antl S t e.im Ou t l e t tempeiatute f 74-1125531-00 1
Fioure B-4 TIME RELATIONSHIP 0F LOSS OF > i1Alti FEEDWATER WITH FAILURES 610 HDT LEG TEMPERATURE 600 - m J s 590 0 g 580 - 3 m 3 570 - NO AFW
- NORMAL LOF#
560 - 550 -
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EXCESSIVE AFW 540 - MSSV F AILED OPEN 530 - 520 0 1 2 3 4 5 6 7 8 9 10 Time, minutes 2300 - 2200 - NO AFW
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RC PRESSURE 2100
=
l0 2000 - E E 1900 NORMAL LOFW
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1800 - 1700 - EbESSIVE AW MSSV FAILE0 OPEN
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APPENDIX C STEAM GENERATOR TUBE RUPTURE 1.0 INTPODUCTION A steam generator tube rupture (3GTR) is a loss-of-coolant accident ( r
\m/ (LOCA) that allows reactor coolant to leak into the secondary side of the once through steam generators (OTSG) where it is released into the steam plant. A SGTR is a serious accident; it contaminates the secon-dary plant and can lead to significant offsite doses if steam from the affected steam generator (s) is released to the environment. It can also have the complications associated with a normal small break LOCA
[] (see Appendix F). ( i
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A SGTR is a loss of integrity of the steam generator tubes. It can be a small leak of one tube or failures of more than one tube. OTSG tube f ailt.re can be caused by corrosion (due to bad water chemistry), exces-sive thermal or hydraulic loadings during severe plant transients, or mechanical wear due to foreign object s in the primary or secondary sys- [A) s x,. /
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tem. Tube failure can occur by itself; it can also be a consequence of a severe plant transient. The leak rate during a S';T R can range from a few gpm to several hundred gpm. Some of the major factors which influence the leak rate of reactor coolant into the steam generators are:
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%d 7-6-82 Appendix C, Page C-1 DATE:
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- 1. The number of defective tubes;
- 2. The size (break area) of the tube failure (s);
- 3. The pres s ur e and temperature condi t ions in the primary and secondary systems.
Small tub e ruptures with le ak rates <20 gpm are more frequently en-countered than large tube ruptures. However, larger leak rates may occur. For a complete severance of one OTSG tube, a leak rate of approximately 400 gpm at normal steam pressure and temperatore condi-tions would be expected. Geveral large tube ruptures with leak rates of several hundred gpm, have occurred at commercial nuclear plants. The leak from a failed tube cannot be isolated until the plant is com-pletely cooled and depressurized and the primary system loops have been drained . Hot standby is not a safe end condition for a SGTR. Maintaining an RCS subcooling margin will prevent the equalization of the primary and af fected OTSG pressures. The primary to secondary ele-vation dif ference will also prevent termination of the tube leak until the RCS is drained down to the tube leak elevation. The re fo re , a plant cooldown and depressurization is required as soon as possible for transient mitigation. Higher RCS pre s sur es result in higher tube leak flows and ultimately higher of fsite doses, so early detection and diagnosis of this accident is very important. O DATE: 7-6-82 Appendix C, Page C-2
BWNP-20007 (6-76) BABCOCK & WILCOX yyy , ,, NUCLEAR #OWER GENERATION DIVI $lON 74-ii2553i-oo O TECHNICAL DOCUMENT
\N jl The RCS must be at cold shutdown before the BWST is depleted. Recirc-ulation from the sump is not possible; injection water will be lost through the broken tube. Therefore, RCS cooldown and depressurization must be initiated as soon as safely possible.
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( j Since a tube le ak is a small break LOCA, the general procedures for LOCA correction must be also followed (see Appendix F). Some modifi-cations to the LOCA rules are req ui red for this unique accident:
- 1) the leak rate through the tube will increase as subcooling margin <
increases; therefore, a minimum subcooling margin should be main-tained and; 2) delays in cooldown and depressurization must be
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Cooldown and depressurization could be delayed if failures occur in plant systems during a SGTR. If possible, the additional failures should be corrected before final cooldown and depressurization to coid j conditions. However, to minimize delays in the RCS cooldown and de-pressurization, the failures may have to be corrected during the
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cooldown.
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! x.d This appendix will show characteristics of and correct ive actions for large and small tube leaks, with and without additional failures. l l l I
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7-6-82 Appendix C, Page C-3
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a BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR Powtt otNUAfeoN DIVISION 74- n 2 s s31-00 TECHNICAL. DOCUMENT 2.0 GENERAL OPERATOR ACTIONS The more important stages in the t rea tment of steam generator tube leaks are:
- 1. Diagnose that a SGTR is in progress and ident i fy wh ich steam generator has the leak.
- 2. Bring the plant to a stable, zero power condition by:
o Performing a plant runback, or o Pe r forming the immediate actions and status verification of Part I if a reactor trip occurs.
- 3. As quick as possible cooldown and depressurize the RCS so that the RCS pressure is below the main steam safety valve setpoint.
4 Once the RCS temperature is below 540F, isolate feedwater to the generator with the tube leak. Isolate all steam extraction lines from the steam generator with the tube leak except the turbine bypass line (do not shut the MISV be-suse inte rmit tant steaming i to the condenser may be necessary). Leak size of 5-20 gpm will l probably not fill the steam ge ne rato r before plant cooldown is completed if the RCPs are running. Slightly larger leaks can be accommodated depending on the t ime it takes to completely cool-down. Isolation of the generator will help prevent contamination l l of the secondary plant and eliminate costly cleanup ef forts. For large leaks and small leaks without RCP operation, secondary 1 plant contanination will occur; the generator with the leak will have to be steamed periodically to prevent overfilling and to l l l l DATE: 7-6-82 . Appendtx C, Page C-4
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circulate the RC in the loop. The operator . should make every ef-i fort to limit the degree of contamination in the balance of plant without inhibit ing mitigation of the SGTR event. This includes isolation of the auxiliary steam supply, gland steam supply, the steam su pplies to the MFW and AFW pump turbines, the second stage 4 reheaters, and any steam line drains or traps from the affected generator.
- 5. If additional component failures have occurred which will a) limit plant depressurization with the spray or PORV, or b) prevent the use of both generators, repair or bypass the l
failures if possible.
- 6. Cooldown and depressurize the plant to cold shutdown condit ions I
while maintaining the reactor coolant subcooled and minimizing offsite releases. I ! These steps are shown in a block diagram, Figure C-1. The text fo l-I lows the outline of this diagram. The first part of the text shows ac-tions for tub e leaks without other failures; ac t ions to take if other l
- failures occur with the tube rupture are discussed in Section 9.0 of i
this appendix. ! During these stages, the tube leak must be treated to correct for the [ l LOCA and also to limit the radioactive steam release. Since LOCA's l l 1 DATE: 7-6-82 Appendix C, Page C-5
f BWNP-20007 (6-76) SABCOCK & WILCOX suusta NUCtt AR POWER GENERATION DIVISION 7'- n 2333 i-on TECHNICAL DOCUMENT are discussed at length in Appendix F of these guidelines, the follow-ing discussion will be focused on treatment of the tube rupture as a unique accident and wil1 only discuss the LOCA as it relates to plant control. Figure C-3, wh ich is at the end of this appendix, summarizes the gene-ral operator actions for a SGTR. Sheet 1 of Figure C-3 summarizes those act ions req ui red to identify the accident and to bring the plant to a stable zero power condit ion and Sheet 2 addresses plant cooldown. Tab le C-3 outlines actions to m nimize offsite releases for SGTR's i with ad d it ion al eq uipment failures. These figures should be studied as a supplement to the following sections. 3.0 EVENT DETECTION The first stage for correction of steam generator tube leak s is prompt detection and determination of wh ich generator has the leak. Tab le C-1 summarizes the ways a SGTR and the af fected steam generator can be detected. The Conden;3r Vacuum System discharge radiation monitors and steam line radiation moni to rs are the best and most timely indicators of a SGTR. The other SGTR symptoms (LOCA or asymmetric high water level in one generator) are best used as back up detection methods only for O DATE- 7-6-82 Annendix C, Page C-6
BWNP-20007 (6-76) SABCOCK & WILCOX NUMBER NucteAs powes oeNERADON Otvi$lON 74- 12ss31-00 TECHNICAL DOCUMENT cases where the plant's radiation detection equipment is inope ra t ive . Figure C-2 shows a P-T diagram and sequence of events for a large tube leak without operator intervention prior to trip. As expected, the tube leak looks like a small break LOCA (see Appendix F). 4.0 PLANT CONTROL AT POWER FOLLOWING A SGTR The second stage for correcting a SGTR is to stabilize reactor coolant (RC) pressure and pressurizer level so the plant may be run back with-out tripping. Tube leaks will cause a decrease in RC pressure, pressu-rize r level, and makeup tank level. The makeup (MU) system will auto-mateially increase MU flow, or stabilize pressurizer level and the pressurizer heaters will come on to restore RC pressure. Tube leaks larger than about 200 gpm will probably require starting a second makeup pump. Leaks greater than 300-350 gpm will cause a decrease in RC pressure which causes a reactor trip and automatic actuation of the HPI. Letdown should be isolated , and MU pump suction should be switched to the BWST. \ Davis Besse is equipped with low head HPI pumps which are ineffective until the RCS pressure goes below the HPI pump shutoff head of about 1650 psi. 'If two makeup pumps cannot match the leak rate, SFAS will actuate, the reactor will trip on low RCS pressure, the pressurizer s may drain, and subcooling margin will be lost. The RC pumps must be tripped when subcooling margin is lost. ATE: 7-6-82 Appendix C, Page C-7
BWNP-20007 (6-76) BABCOCK & WILCOX NuusER NUCLEAR POWER GENERAflON OtVISION 74-li2ss31-00 TECHNICAL DOCUMENT 5.0 PLANT CONTROL TO HOT ZERO POWER (See Sheet 1 of Figure C-3) The third stage for correcting a SGTR is to bring the plant to hot standby (unless the le ak is large enough to depressurize the RCS to a low pressure and automatically trip the plant). It is pre fe r ab le to run power back as much as po s s ib le , even if a reactor trip is imminent to avoid or limit lifting the main s t e ar, safety valves and releasing radioactive steam to the environment. Tube leaks can occur alone or with other plant failures. If other plant failures occur at the same time, it may be difficult to estab-lish hot standby and to proceed with cooldown and depressurization. Generally, the best course to fo llow, if possible, is to fix the other failures wh ile taking actions to cool the plant down. This section will show how to es t ab li sh hot standby and prepare fo r cooldown with tube leaks only. Treatment of other failures will be given in Section 9.0. Action should be initiated to stabilize pressur. r level and RCS pres-sure wh ile conducting a plant runback (at 5% pe. - minute) to low power without tripping. RCS inventory should be closely watched during the plant runback and subsequent cooldown. Upon reaching a low power level where the turbine bypass capacity is sufficient to avoid lifting the steam safeties (roughly 20% full power), the plant can be tripped as fo llows : 7-6-82 Appendix C, Page C-8 DATE:
BUNP-20007 (6-76) SABCOCK & WILCOX NUCLEAR POWtt GENERAftON olvi$lON 74- 125s31-00 TECNNICAL DOCUMENT ~
- 1. Ensure that the turbine throttle pressure limiter is in auto-matic.
- 2. Place the TBV's in manual and open them to unload the turbine.
- 3. Unioad the turbine generators and trip the turbine.
- 4. Trip the reactor.
- 5. Imnediately place the TBV's back in automatic with appropriate pressure setpoints or control header pressure in manual (at operator discretion).
The plant is now at hot zero power. The'next step is to begin a nor-mal plant cooldown. If pressurizer level control is lost and a rea: tor trip occurs, there is a chance that the pressurizer will drain, SFAS will actuate on low RCS pr e s sur e , and the reactor coolant subcooled margin will be lost. If this happens, the operator must:
- 1. Ensure full HPI flow.
- 2. Trip the RC pumps immediately following the loss of subcooled margin.
- 3. Ensure that AFW starts.
- 4. Throttle HPI once the reactor coolant subcooled margin is re-gained and maintain pressurizer level at 100 inches or greater,
- s. Bring the plant to a hot stable condition so that a plant cool-down may be initiated.
\ The plant is now at hot zero power. The next step is to begin a cooldown and depressurization. s DATE: 7-6-82 Appendix C, Page C-9
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUCLE AR pow GENERATION DIVISION 74-ii2ssn -00 TECHNICAL. DOCUMENT 6.0 INITI AL C00LDOWN AND DEPRESSURIZATION Once the plant has been brought to a hot zero power condition by a run-back or has been stabilized following a reactor trip, a cooldown and depressurization of the RCS should be started. A minimum subcooled margin should be maintained during this tooldown. This action is required to minimize the offsite dose because:
- 1. By reducing RCS pressure to near the secondary systec pressure, the primary to secondary leakage will decrease and;
- 2. By decreasing RCS temperatute and steam pressure, rad ioact ive steam release through the stream safety valves is less likely.
The objective of the initial cooldown is to bring the RCS subcooled temperature to a value (<540F) that corresponds to a saturation pres-sure which is below the main steam safety valve setpoint. This action will prevent inadvertent lifting of the main steam safety valves and the resulting radiation releases. Below 540F the SG with the tube rup-ture can be isolated without lifting the MSSVs. The selection of the cooldown method to be used ( normal or emergency) is discussed in Section 8.0. The plant cooldown and depressurization should be continutd to cold shutdown conditions. The primary-to-secondary leak will not stop completely until the loops are drained. l 7-6-82 Appendix C, Page C-10 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX sumeEn NUCLE At POWER GENERATION Divi 5 TON n 7'-i i 2 s s 31-00 / N. TECHNICAL DOCUMENT \
%d
/
Prior to and during the cooldown and depressurization pe riod , the operator should check the status of several things so that he can decide on the best method for cooldown:
- 1. Estimate leak size by maintaining constant pressurizer level.
T] / Normal makeup will be adequate for a small tube leak. A large L/ ' tube le ak will require initiation of a second makeup pump (cool-ing the RCS will require aditional makeup for contract ion).
- 2. Verify the operability of the condenser and all TBV's.
- 3. Verify the availability of both OTSG's. Assure both can be fed and are able to maintain pressure. If not, refer to SGTR with other failures section.
[] ) 4 Verify operation of RC pumps for use of pressurizer sprays. If \
'i / subcooling has been lost and the pumps ha.- been tripped, restart one pump in the spray loop as soon as the subcooling margin is re-gained.
7.0 ISOLATION OF AFFECTED SG j#~% ( ) During cooldown and depressurization of the plant, isolation of the af- \
%j/
fected steam generator is recommended once the reactor coolant tempera-ture decreases below 340F. Below this t empe ratur e , steam pressure will be below the main steam sa fe ty valve lift setpoints. Isolation is defined as: CN I )
\ '
w, 7-6-82 Appendix C, Page C-Il DATE:
l BWNP-20007 (6-76) BABCOCK & WILCOX " " NUCLEAR POWER GENERADON Olvl110N 7'-i t 2 ss 3 t-00 TECHNICAL DOCUMENT
- 1. Stopping MFW and AFW flow to the af fected s t e am generator; except as needed to maintain the low level limit to promote natural c i rcula t ion to maintain adequate subcooling margin or to maintain tube to shell AT requirements.
- 2. Stopping s t eaning of the affected steam generator except as necessary to promote natural circulation or maintainin the steam ge ne ra to r pressure below 1000 psig and the level below 95% on the operate range.
- 3. Swi tch ing both the msin and auxiliary feed pump cnd other s te am supplies to the unaf fected SG.
The reactor coolant wh ich leaks into the s tean generator eventually may need to be steamed to limit water accumulation in the generator. Steaming must be done when 95% ca the operator range is reached. Steaming may also be needed to maintain natural circulation so that a reasonable rate of RCS cooldown to cold shutdown can be ach i eved . The decision to steam should be based on the actual situation; RCS coo!- ing , stean generator level, tube-to shell AT, and the need to add feed-water for natural circulation will help make this decision. Feedwater should be isolated to the af fected SG until the leak size is known. Small leaks will not provide enough water to the gernator to maintain a minimum level; the re fo re , periodic fe eding will be required to keep the shell cool and rate of cooldown constant. Large leaks will provide enough water without feedwat e r addition. Feedwater may be initially required to establish and maintain natural circulation in DATE: Appendix C, Page C-12 7-6-82
BWNP-20007 (o~/6) BASCOCK & WILCOX NUMBER NucteAn Powea oeNenADON OrdSCN TECHNICAL DOCUMENT 74-t i2s s3i-00 the loop with the tube leak. A decision to add feedwat e r must be o based on the actual plant situation; RCS cooling, steam Fenerator level, t ub e-t o-shel l AT, and natural circulation conditions will help make this decision. ( If the plant RC pumps are off, feedwater addition and steaming will be required from both loops during cooldown and depressurization to main-tain loop flow in both loops. This will prevent stagnet hot reactor coolant in either loop which will go saturated in the " candy cane" of the hot leg and forming a steam bubble as the RCS is depressurized.
- When this steam bubble will act as a prc'surizer and inhibit further plant cooldown and depressurization.
Although the intent of isolation is to " bottle up" the af fected SG and ? to use it as a s to rage tank for the additional reactor coolant that i. escapes during the remainder of the plant cooldown and depressuriza-l tion, total isolation is not always pos s ib le . For large tube leaks, l l steaming may be required; feedwater may have to be added for some l l other situation. Level control in the stean generator and a high rate
- of RCS cooldown will provide a lower radioactive release for large leak s than a slow cooldown. A fast rate of cooldown will limit the I
total water that leaks into the steam generator, thus limiting offsite radiation releases. A timely cooldown will also allow the plant to be l brought to a cold, depressurized condition before the BWST is drained. l Therefo re, a fast as allowed and sustained rate of cooldown should u DATE: 7-6-82 Appendix C, Page C-13
BWNP-20007 (6-76) BABCOCK & WILCOX wu.e a NucttAn POwea otutaATION 01VISIOte 7'- " 25531-00 TECHNICAL DOCUMENT take priority over isolation of the OTSG for large tube leaks. For smaller tube leaks where the transient is well under control (i.e., low rate of SG level increase, adequate '5WST inve nto ry , etc.) the SG with the tube rupture should not be steamed for the sole purpose of achieving a 100F/hr cooldown. Specific isolation actions for various plant conditions are addressed O below (additional failures that may occur are covered in Section 9.0): A. Small Leaks with itC Pumps Running Small SG tube leaks are events where the loss of reactor coolant is wi th in the capacity of the no rmal makeup system at system operating pressure with both MU pumps on and letd~.m I : : '. - * ,. . . For this condition, the reactor coolant pumps will force circula-tion through the isolated loop so it will not stagnate and flash; a cont inuou s uninterru pted cooldown with the una f fected gene rato r is mandatory. The cooldown rate should be close to normal, the tube-to-shell temperature dif ference should be monitored and main-tained within no rmal limits (100F) to avoid excessive tube stress. Feed addition should be maintained at the low level limits to keep the lower shell covered. A continuous depressu-rization of the RCS is required to maintain a minimum subcooling margin wh ile proceeding with RCS cooldown. If heat removal does become interrupted, a path to the condenser from the a f fected OTSG should be opened to prevent radioact ive release through the main steam safety valves or AVVs of the isolated generator. DATE: Appendix C, Page C-14 7-6-82 _ _ _ _ _ _ _ _ I
BWNP-20007 (6-76) BABCOCK & WILCOX NuusER NUCLE AR POWER GENT 2AllON DIVIStoN C' s TECHNICAL. DOCUMENT 74-ti2ss31-00 ( ! 1 <
\_/
Intermittent feeding to the isolated generator may be required to maintain a minimum water level if a minor tube leak occurs. A le ak of this size may not be large enough to fill the generator before the couldown and depressurization are completed. The leak rate and the steam generator level should be nnnitored at all i ! ( ,/ Limes. If the level increases or the leak rate changes, the af-fected generator may have to be steamed (preferably to the conden-ser). Steaming of the af fected OTSG may also be desired at lower RC temperatures to maintain a reasonable rate of cooling depen-ding on existing plant conditions (e.g., low BWST inventory). C00LDOWN MO DE 1 should be used for the entire transient if the gx condenser and RCP's are available (see Figure C-3, Sheet 2).
/ a \ /
mj B. Small Leaks with RCP's Of f Intermittent or minimal continuous feeding and steaming may be re-quired to maintain natural circulation in the affected loop and tc prevent stagnation and flashing in the RCS. The cooldown rate should be close to 100F/hr, but it may have to be lowered so that o / h makeup can keep up with the leak rate and RCS contraction. Pres- 's ! surizer level should be maintained. The tube-to-shell tempera-ture dif fe rence should be maintained within the eme rge ncy limits (150F). Continuous feeding and steaming of the af fected OTSG (at low flow rates) may be required to maintain natural circulation cooling and the tube-to-shell limits. Cooldown Mode 2 should be
/
- l. n')
Lj used for the duration of transient (See Figure C-3, Sheet 2). DATE: 7-6-82 Appendix C, Page C-15
BWNP-20007 (6-76) BABCOCK & WILCOX NUh.3 E R NUC1EAa POWER GENERATION DIVISION 76-1325531-00 TECHNICAL DOCUMENT C. Large Leaks / Ruptures with RCP's On or Off 0 If the tube leak is large enough, the af fected OTSG will contrib-ute significantly to RCS cooling. The best approach is to steam the af fected OTSG at a rate adequate to maintain a constant level. The rate of cooling (up to 240F/hr above 500F and up to 100F/hr below 500F) can then be controlled by steaming the unaf fected OTSG. If the affected OTSG begins to fill, steaming should be increased to limit the water level to 95% on the operate range. If this happens then steaming of the una f fected SG should be decreased, as necessary, to maintain the total RCS cooldown rate at 100F/hr (below 500F). Once the 95% level is reached, continuous steaming of the leaking OTSG, without feedwater addition, will be required to maintain a constant rate of RCS cooling and to prevent overfill. If isola-t ion of feedwater and adequate s tcaning of the af fected OTSG are achieved early in the transient, the dif ficulties associated with OTSG overfill can be avoided. Toward the latter stages of the cooldown, the affected OTSG may begin to fill; the reduced core heat rejection will be unable to boil off the leaking reactor coolant. If this occurs, most of the heat reaoval should be transferred to the affected OTSG. Intermittent feeding and steaming of the unaffected OTSG may be O DATE: 7-6-82 hppendix C, Page C-16
BWdP-20007 (6-76) NUMBER BABCOCK & WILCOX _NUCLEAD POWER GENERATION OtVislON 74-1125531-00
'CHNICAL DOCUMENT t r ipped ) ,
required to maintain natural circulation (if RCP's are up to 100F/hr cooling, and the emergency tube-to-shell limit. Cooldown Mode 2 should be used. (See Figure C-3, Sheet 2).
,['~'}
e i 8.0 COOLDOWN AND DEPRESSURIZATION TO COLD SHUTDOWN
?
L./ The final step for mitigating tube leaks is to bring the plant to a completely depressurized condition. The tube leak rate will be the lowest when the RCS pressure is approximately equal to the steam gene-r'a to r pressure. To achieve this condition, the RCS must be depres-suried so that the decay heat removal system can be started. ('m) ( /
'L/ The RCS sub cooling requirements will not allow equalization of RCS and secondary system pressures while the steam generator is removing heat.
When the decay heat removal system is started for heat removal, the steam pressure may be allowed to remain slightly higher than the RCS pr es sur e . Reverse leakage will occur, but the effect will be minimal l l if this condition is no*. maintained for extended periods of time. l ,,-~. I \ f
\ ! 'w./ Although the leak flow will be lowest when the RCS is placed on the decay heat removal system and the steam pressure is increased by OTSG re-steau isolation, further cooldown and depressurization will be quired. Since the elevation of the hot leg " candy cane" is always higher than the tube leak, elevation head will cause the leak to con- / , y
( ) tinue. Special actions for handling the leak after the decay heat < v systen is placed in operation will be described in Section 11.0. Appendix C, Page C-17 DATE: 7-6-82
BWMP-20007 (6-76) BABCOCK & WILCOX " HUcLEAR POWER GENERATION olva5loN 74- 2ss31-u0 TECHNICAL DOCUMENT Control of plant depessurization is needed to bring the plant to cold shutdown conditions; methods of depressurization are described in Section 11.0. The approach chosen for cooldown to the decay heat removal system will depend on the conditions of the plant at the start of cooldown. To de t e rmine the mode of cooldown the leak size, condenser availability, RC pump availability, and other factors must lie known. The following sub s ect ion describes the various factors to be cons id-ered for final cooldown: 8.1 Selection of Plant Cooldowno Mode Figure C-3 shows the three modes of plant cooldown to be selec-ted. There are two basic cooldown rates shown in this figure:
" normal" and "eme rgency" .
A " Normal" cooldown is defined by: For the entin. cooldown, a close-to-normal cooldown rate is used (<100F/hr); Tube-to-shell temperature limits do not exceed 100F; Fuel pin compression limits are not violated deliberately; An " Emergency" cooldown is defined by: Tube-to-shell temperature limits do not exceed 150F DATE: 7-6-82 Appendix C, Page C-18
BWNP-20007 (6-76) SABCOCK & WILCOX NUMBER Nucteas Powes oeNenAnoM DmSION 7'-1123s31-00 TECHNICAL DOCUMENT 4 Fuel pin compression limits may be deliberately violated The RC temperature is reduced to 300F as fast as pos s ib le 4 (not to exceed 240F/hr). After 500F is reached the cooldown rate should be close to 100F/hr . , as long as that rate can be held. Steaming of both OTSG's will be required at lower RCS temperatures. The cooldown rate should be reduced when the tube-to-shell w temperature limit approaches 150F, unless there is danger of
- running out of BWST inventory.
Selection of mode of cooldown is primarily based on three condi-tions: leak s iz e , RC pump status, and availability of the conden-I ser. The three conditions .take into account the severity of the - event. If the leak is large or the condenser or the RC pumps are s i not available, o f f-site releases will be large; the emergency approach is chosen. Small leaks with the condense r and RCP's i available will not allow a significant off-site release, so a nor-mal approach is selected. 1 Mode 1 (near " normal" cooldown) should be selected when
- a. The condenser is available, and;
- b. The tube leak is small, and; l c. the RC Pumps are operating.
Note that all three conditions must be satisfied to use the "nor-mal" mode. s DATE: 7-6-82 Appendix C, Page C-19
BWEP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER GtHttATION DIVI $1ON
- 7'- 12ss3i-00 TECHNICAL DOCUMENT Mode 1 is a "n o rm a l" cooldown except that the af fect ed SG is isolated ( pe r Section 7.0) when the reactor coolant tempe r a ture (Thot) is reduced to less than 340F. It should be used for SGTR's where the leak rate (including RC contraction during cool-down) is within the capacity of the MU system and no additional equipment failures have occurred. For this mode, all normal plant equipment or systems (condenser, RC pumps, pres sur ize r sprays, etc.) must be available. The plant should be cooled with-in normal plant cooldown limits (fuel pin compre s s ion limits, 100F, t ub e-to-she l l AT's, e tc . ) . This cooldown mode will most likely be preceded by a plant runback. This cooldown mode is the most likely condition which will be faced, based on tube failure histories with the OTSG.
The entire cooldown cr.n be carried out by using the unaf fected genera-tor. To cooldown with one generator under these conditions the fol-lowing apply:
- 1. RC Pumps will keep circul.at ion in both 1 cops; if all RC Pumps were off, the reactor coolant in the isolated generator would flash to steau and prevent cooldown and depressurization. The RC Pumps also allow spray depressurization, wh ich is more control-lable than PORV depressurization.
- 2. The leak should be small enough and the cooldovn is fast enough with one generator so that the water accumulation in the isolated generator will not build up and require steaming to lower the UA
- 7-6-82 Appendix C, Page C-20
BWNP-20007 (6-76) SABCOCK & WILCOX Nuusu OWCLEAR POWER GENNATION OtVI$40N
"- 2 s s31-00 ' TECHNICAL DOCUMENT water level. The generator may ins tead be used to store the leaking reactor coolant. The level should still be closely moni-tored, and steaming should be started if the water level approaches 95% on the operate range.
p 3. Continuous heat removal by the single generator will be required. ( j\ \ The single generator will not only remove heat from the reactor coolant and core but also from the seconda ry fluid in the iso-lated generate r, which will act as a heat source. At lower RC temperatures, the additional heat removal requirements on the good OTSG may result in a significant reduction in the rate of RCS cooling. If the rate of cooling is significantly reduced or interrupted, the isolated generator TBV's should be opened to in- % creaso RCS heat removal. This will prevent ove rfill of the af-fected OTSG.
- 4. Depre s suriza t ion should be controlled so that it closely follows the cooldown. The reactor coolant should be maintained close to the subcooled margin to keep the leak rate as small as possible.
- s. Feedwater to the ruptured generator should be isolated after a normal . level is established. The feedwater will be cooled as
'" ) heat is removed from the generator and shell heat will reduce in the region contacted by the water in the generator. Steam gene-rator tube-to-shell A T should be monitored and normal limits main-t a ined . The rate of cooldown should be controlled to maintain normal limits. O DATE: 7-6-82 Appendix C, Page C-21
BWNP-20007 (6-76) BABCOCK & WlLCOX NumerR NUCLE AR POWER GENERATION DIVISION 74-ii2ss31-00 TECHNICAL DOCUMENT The " emergency" mode is selected when:
- a. the condenser is not available, or;
- b. the leak is large, or;
- c. the RCP's are not available.
Note that if any one of these three conditions exist, the " emergency" mode must be used. Figure C-3 shows emergency n. odes for conditions wh en one generator or both generators are available. Cooldown mode 2 (shown in Figure C-3) should be used for a SGTR " hen llPI must be used to compensate for the leak rate, and when both steam generators are available and removing heat when the plant cooldown is initlated. This cooldown mode may be preceded by a reactor runback or by an automatic reactor trip (if the SGTR is large or other equip-ment failures occur). The general actions used for this cooldown mode are as fo llows :
- 1. Once the plant has been rapidly cooled down (<240F/hr) and depres-surized to about 500F and 1000 psig, pressurizer level has been stabilized, and isolation of the af fected SG been performed:
e If the condenser is not available, the AVVs must be control-led manually to continue the cooldown. If the condenser is available, the TBV's should be used. O l 7-6-82 Appendix C, Page C-22 DATE:
BWNP-20007 (6-76) BASCOCK & WILCOX *II NUCLEAR POWtt GSNERADON DivlS40N 74-1125s31-00 i TECHNICAL DOCUMENT e If the 'RC . Pumps were tripped on of subcooling margin los g" % they should be restarted if the subcooling margin is i i rega ined; the pump in the spray loop should be started first
-to make the pressurizer spray. available (See Pump - Restart Criteria in Part II " Equipment Ope ra t ion") . If the RC Pumps V
4 g cannot be restarted, the PORV willi have to be used for ( de pres sur iza t ion.
- 2. From 500F and 1000 psig, continue the plant cooldown and depres-e t
'surization at up to 100F/hr with' the unaf fected or both OTSGs as required to maintain cooling, until the decay heat system can be started.
I e Maintain minimum subcooling to minimize the leak, o Monitor the water level in the isolated SG and initiate steau- '
~
ing if the water level rises above 95% on the operator range. 4
'e If the
- 1eak does not picvide enough water to keep the re-l quired . niinimum level (na tur al e- circulation <1evel, forced circ-I ,
ulation level etc.), feed the rupturef generator as required, e If the plant is in natural circulation, both generators will have to be fed and steamed as, neces sary to maintain natural circulation in both loop .. N e Monitor t ub e-t e-shell limits and slow the plant cooldown to avoid exceeding a AT of 150F (See Part II, Vol. 1 Chapter E "Cooldown With One Steam Generator Out of Service"). s lt DATE* Appendix C, Page C-23 7-6-82 ,
BWNP-20007 (6-76) BABCOCK & WILCOX Numsta NUtt[ AR POWER GENER AtloN DIV151oN 74-ti23s31-00 TECHNICAL DOCUMENT
- 3. After the decay heat removal system has been started, continue to cooldown and depressurize to cold shutdown so the loops can be drained and repairs started.
The above cooldown procedure is an emergency measure. Its goal is to get the plant to a depressurized state as quickly as possible in order to stop the leak and minimize offsite releases. In this mode, fuel compression limits do not apply. Mode 2 can be applied when all plant systems, are available, or under more degraded conditions such as a loss of offsite power. Figure C-3 outlines this cooldown mode; it shows the ways the plant can be controlled and identifies monitoring and corrective actions which are unique to a SGTR. Mode 3 cooldown is identical to Mode 2 except that only one steam gene-rator is ava ilab le fo r hest removal when the plant cooldown is started. As indicated in Figure C-3, this could result from equipment failures that cause a loss of feedwater to one SG or a loss of steam pressure control to one SG. The failed steam generator may or may not contain the tube ru ptur e . The actions required for Mode 3 are the same as Mode 2 except that isolation of the affected SG may not be possible. For Mode 3, significant offsite releases can occur under some failure modes. Figure C-3 outlines this third cooldown mode and the circumstances when large offsite doses could occur. O 7-6-82 Appendix C, Page C-24 DATE:
BWNP-20007 (6-76) SABCOCK & WILCOX wumste NUCLtAR POWit GENttATON DIVI $lON ('%J
\
')/ TECHNICAL DOCUMENT 74-1225331-00 Since the Mode 3 "one generator cooldown" can only come about because of failures, it will be discussed in more detail in the following sections which address failures.
f^'N A summary of the SGTR conditions and appropriate cooldown nxade s ( } \ ( ,/ follows: SCTR Conditions Cooldown Comments Normal Small Lee'A, RCP's On Mode 1 Isolate Affected OTSG (Cond.nser Available) Emergency ( n. Mode 2 Isolate Affected OTSG (L/ ) Small Leak RCP's Off or condenser unavailable Large Leak RCP's On Mode 2 Isolate Affected OTSG Large Leak RCP's Off Mode 2 Isolate Af fected OTSG depressurized with PORV One Generator Cooldown Mode 3 Refer to Section 9.0 l l es\ 9.0 PLANT CONTROL FOR TUBE LEAKS WITil OTHER FAILURES i / i l
\ !
'wJ Plant equipment failures during a tube leak transient can make plant l control very di f ficult . Some failures will make it difficult to es-l I tablish a stable condition for cooldown; some will cause the cooldown rate to be much slower than desired. Some failures can cause the
~s off-site radioactivi ty releases to be large. Some failures can be re-1
[ } (U/ paired or bypassed and some cannot. Failures that can make plant con-trol difficult will be of two kinds: 7-6-82 Appendix C, Page C-25 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX suustR Nyctt As PowtR GENERATON DIVIStoN 74-ii2ss31-00 TECHNICAL DOCUMENT Those that limit the heat removal of the steam generator (s) (these failures will limit the ability to establish hot standby and will limit the cooldown rate).
- Those that limit the ability to depressurize the RCS (these fail-ures will limit the depressurization rate and may require the cooldown rate be limited in order to stay within NIYr limits).
Table C-3 s un.na r i z e s the major failures and the actions required to correct them. The general approach for correcting other equipment failures that occur in addition to the tube leak is as follows: e Sinultaneously:
- a. Rapidly depressurize and cool the reactor coolant system (if possible) and bring the system to the point where the RCS pressure is below the setpoint of the steam safety valves.
Moat failures will not prevent this step and may even cause the depressurization,
- b. Diagnos e the additional failures in accordance with the prin-ciples shown elsewhere in these guideliner.
e Correct the failure prior to or during the final cooldown to cold l shutdown. Do not delay the cooldown even if the failure limits the rate of RCS cooling and depressurization. De t ails of the effects of and correctione for failures are shown in Tables C-2 and C-3. O 7-6-82 Appendi,. C, Fage C-26
\TE:
l
EWNP-20007 (6-76) BABCOCK & WILCOX NMR i NUQEAR POwtB GEN tATION 91V1510N 74-i i 2553i-00 ; _ 'ECHNICAL DOCUMENT If the failures cannot be corrected to enable both generators to e be used, or the RCS cannot be deprescurized with use of the pres-surizer spr ay or the PORV, the cooldown will be di f ficult . In general these kinds of f ailures are limited to:
- a. Failure of a main steam safety valve (open). This will not
>r] 'y, allow a " normal" rate of feedwater addition because excessive RCS cooling rates will result. Turbine bypass failures may be corrected by isolating the failed valve (s) and do not ap-ply; MSSV failures cannot be isolated. The plant must be cooled down with the remaining generator even if it contains the leaking t ub e . Of f-s ite releases of radiation may be high if the open steam ialve is in the same generator as the tube
(]
'v leak.
low steam pressure, closing both SFRCS will actuate on MSIV ' s . The operator should mar.ually spen the MSIV in the good SG. A' During cooldown, the generator with the steam failure (low / \ J pressure) must be continuously fed with sufficient flow (s100 gpm) to keep the tube-to-shell temperature difference wir.hin limirs; preferably through the main feedwater nozzles since it will cool the lower shell better. If the generator is com-plet ely depre ssurized , extreme care must be exercised when
,m / \ A flow path through the warmup valve can be used to (j'.
1 feeding. Dm: Appendix C, Page C-27 7-6-82
BWNP-20007 (6-76) SABCOCK & WILCOX NUmsta NUC1tAS POWit GENERAllON DIVl560N 74-1 25sa t-00 TECHNICAL DOCUMENT ach ieve low feed flow rates (100 gpm), if the RCP's are trip-ped, it is impor t ant to provide some feeding of the depressu-rized generator to establish or maintain natural circulation. If the steam failure is in the affected (ruptured) generator and the tube leak flow is large, the tube leak will usually provide adequate cooling. Unlike most other transients which do not recommend feeding a depressurized generator , a SGTR re-quires a fast RCS cooldown to terminate the transient; the re fo re, careful feeding of a depressurized generator is allowed. Steau leaks will most likely be caused by valve failures and other breaks outside containment rather than piping f ailures (TBS, MSSV's, etc.). If the stean leak occurs inside contain-ment, it will likely be a piping failure in a steam, feed, or drain line. If the steam failure is on the s te _en generator with the tube leak and is inside the containment, the reactor l coolant from the tube leak will be returned to the contain-ment sump. Although adverse containment conditions may re-sult, the of f-site releases of radiation will be reduced af ter the RC temperature reaches 540F. l ( b. A loss of all feedwater to ene OTSG will nor.nally require cooldown on the remaining generator. The rate of RCS cooling must be slow to prevent violation of the tube-to-shell temper-ature limits on the OTSG without fe ed water . If the RCP's are i DATE: 7-6-82 Appendix C, Page C-28 L
BWNP-20007 (6-76) BABCOCK & WILCOX NUMBit NUCitAR POWER GENf9 AtlON OtvlSION . j 74-i 23531-00 TECHNICAL DOCUMENT i tripped, .the rate of cooling will be even slower; reactor i coolant will not naturally circulate in the idle loop. The ' stagnant loop must be closely monitored for saturation in the hot leg. Releases off-site will be significant because of the slow rate of RC3 cooling and depressurization. In addi-f tion. the SFRCS SG low level or steam-to-feedwater pressure Only the di f fe rential function will have closed the MSIVs. AVVs will be available for cooling until SFRCS has reset and t i the MSIVs are manually reopened. Every ef fort should be made j to realign feedwater to both generators through the AFW cross-connect s. If realignment cannot be made, plans should be made to replenish the BWST inventory to avoid loss of injec-tion water. If feedwater is lost to the ruptured OTSG and large le ak flows are present, the leak will provide enough j i feedwater to maintain cooling of the af fected OTSC. I c. Pressurizer valve failures (f ailed open or just leaking) will require constant HPI to maintain subcooling. Failures in the pressurizer steam space will cause HPI to fill the pressu-rizer. RCS depressurization from this solid water condition will require throttling of HPI without loss of subcooling. The BWST inventory should be carefully monitored when this mode of depressurization is utilized for extended periods of 1 t t1me. I l Appendix C, Page C-29 DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nuuste NucLEAa POwta GENteATCH olVl510N 74-t i 2 s s 31-00 TECHNICAL DOCUMENT _
- d. Total loss of feedwatec (MFW and AFW) will require a solid water cooldown without the benefit of secondary heat removal.
Atte pts should be made to obtain feedwa t e r from all avail-able sour c e s . For cases of this type requiring excessive cooldown t ime s , BWST invento ry should be carefully monitored. RPI inventory drawn from the BWST will flow into the affected OTSG . It will not be returned to the reactor building sump and, the re fore , cannot be reclaimed. Backup water supplies for the BWST may be nec es s ary and should be prepared. Of f-site radioactive releases may be very significant. 10.0 LIMITS FOR RCS C00LDOWN O The following is a summary of the basic limi::s which the operator should be aware of while proceeding with the RCS cooldown for all SGTR's :
- 1. Fuel pin compression limits apply to the case of small SG leaks with RC Pumps running and the condenser available; they may be violated for other cases.
- 2. The normal tub e-t o-sh ell temperature limit (100F) applies to the case of small leaks with RC Pumps running and the condenser ava ilable . The emergency tube to shell temperature limit (150F) applies to all other situations, except:
1 O l I DATE: 7-6-82 Appendix C, Page C-30 l
BWNP-20007 (6-76) BABCOCK & WILCOX wu ste NUcitAs POWta GENitATION Olvi$lON 7'-ti23531-00 TECHNICAL DOCUMENT The emergency limit may be violated (at management discre-tion) if Icss of core cooling injection water is imminent (BWST draining) . The steam generators may be damaged if the emergency limits are violated; core cooiing, however, takes priority. 1 i
- 3. The cooldown rate will be near " normal" for small le ak s with the condenser available and RC pumps running. The cooldown rate will be 100F/hr, except where tube-to-shell tempe ra tur e limits re-strict cooldown rate. Overcooling should be avoided and the pres-surizer should not be allowed to drain. If the plant is on natu-
[' ral circulation, the cooldown rate should be reduced to avoid RCS I (
\ / voiding (the recognition of void formation is discussed in Part II, Volume 1 of these guidelines). Cooling with both OTSG's will be required at least periodically to maintain natural circula-tion. For large leaks or small leaks without the condenser avail-able or the RC pumps running the cooldown rate will be 1240F/hr when the RC temperature is above 50CF and fl00F/hr below 500F.
[n} \ /
- 4. The minimum subcooling margin should be maintained. Excessive subcooling should be avoided to keep a minimun leak rate.
- 5. The plant should be completely cooled down and depressurized be-fore the BWST is drained. Simply placing the decay heat removal system into operation is not adequate.
f I s I
\j 9 ATE: 7-6-82 Appendix C, Page C-31 l
BWNP-20007 (6-76) SABCOCK & WILCOX " 8 wucteAa rowse ceNenAfiON OlVISION 74-u 255 31-00 TECHNICAL DOCUMENT
- 6. It is desirable to pl ac e the plant on the decay heat removal system be fore the condensate storage tank is depleted and poor quality backup water is injected into the steam generator by AFW.
- 7. When RCP's are running do not violate their NPSti requirements.
11.0 SPECIAL TOPICS RELATED TO C00LDOWN 11.1 Ways to Depressurize the RCS Two ways are available to depressurize the primary system. The first way is to use the pressurizer spray; the second is to open the PORV. Use of the pressurizer spray is pre fe rred since PORV operation can result in a quench ta nk rupture disk ru p t u re , ab-normal reactor building pressures and tempera ture s , and addi-t ional RB c le anup ef forts. Pres s ur ize r spray also provides better, constant control in comparison to cyclic PORV operation. Therefore, RCP operation should be maintained whenever possible. The pressurizer spray is dependent on RC pump operation. Conse-que ntly , if the RC pumps ha been tripped, an RC pump should be restarted as soon as the restart criteria are satisfied (See RC pump Restart Criteria in " Equipment Ope r a t ion") . An early restart of the RC pumps will allow the pressurizer spray to be used and will limit the need to add additional feedwate r to the af fected OTSG to sustain natural circulation. The pe r formance of the pump motor and seal cavity cooling sys t ems should also be l O l DATE: Appendix C, Page C-32 7-6-82 l
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen NUGk AR POWit GENEAAtlON DIV1510N 7'-ti23s3i-oO TECHNICAL DOCUMENT I periodically monitored to ensure that the pumps can be continu-t ly run until the decay heat removal system can be placed into operation. a. i In the event the pressurizer spray is of f or the RC pumps cannot be restarted, the pressurizer PORV should be used to depressurize < the plant. In ge neral , it will not be possible to open the PORV and maintain pressurizer level and pressure control. Instead, cyclic (open-shut) operation will be necessary. The best way to l j maintain plant control is to continue the RCS cooldown and depres-surize with the PORV in a sawtooth manner between the subcooling margin - line, as shown on the P-T diagram of Figure C-4, and a temperature around 30F cooler than the subcooled margin line. When the PORV is opened, pressurizer conditions should be closely monitored. As the pressure drops, some of the water in the. pres-
. surizer will flash to steam which would cause level to decrease.
However, the expansion of the RCS liquid volume may have a greater affect causing the pressurizer level to increase. Makeup and le tdown flows should be adjusted as necessary to maintain an P adequate pressurizer water inventory. 11.2 Placing the Decay Heat Removal System into Operation Once the plant has been cooled down and depressurized, the decay heat removal system (DHRS) should be used for final cooldown and r e 7-6-82 Appendix C, Page C-33 i DATE:
.-,- - --,-- - ~ ~ - - . - - - 1
BWNP-20007 (6-76) BABCOCK & WILCOX ~ Zet, HUCtfAR POwta CINERATION OtVa$10N 7'- 12n 31-00 TECHNICAL DOCUMENT depressurization. Special actions must be taken fo r tube leak s when the decay heat system is actuated to continue SGTR mitiga-tion. Special actions must also be taken to place the DilR system into operation during a natural circulation cooldown. The most important items are (niso refer to Appendix F):
- 1. As a general rule, if the RCS is saturated (which will only occur if several tubes are failed), it may not be po s s ible to place the DHR into operation. The operator will not know the liquid level above the DHR drop line; a loss of decay heat suction would cause pump cavitation. However, if it is def-initely known that no LOCA other than a tube leak exists, the decay heat removal system may be placed into operation. The reason is that the leak location is such that liquid will be trapped in the lower parts of the hot leg and the RCP inter-nal " lip" will trap water between the pump and the vessel.
When the RCS is saturated and the decay heat system is en-gaged, the decay heat pumps must be locally monitored for cavitation. The pumps must be shut down quickly after loss l ef suction to prevent cavitation damages. l t
- 2. One decay heat pump should be aligned in the decay heat re-moval mode. The other LPI pump should be placed in the injec-tion mode from the BWST or the HPI pump should be placed in
" piggyback" on the decay heat pump for injection. Injection flow rate needs will determine the choice.
DATE: 7-6-82 Appendix C, Page C-34
. - - ..... - _-. - .. - _. _ . . . ~ . - - . - ._ - _- - _- - _
J BWNP-20007 (6-76) I' BA8 COCK & WILCOX Numet NucttAt PoWit GENERATCN DIYl510N 74-n 2553i-00
- TECHNICAL DOCUMENT
- 3. Due to the location of the decay heat removal drop line and i return line, the system may not adequately cool the higher j po rt ions of the reactor coolant loop, expecially if the plant has been cooled down on natural ciruciation. Cooling of the 4
i lower portions of the loop and continued RCS depressurization can permit reactor coolant flashing in the high parts of the ' hot leg; this will inhibit further depressurization. This l +
.erfect will be more pronounced if a natural circulation cool- i I
I- down rather than a forced circulation was used. To avoid this possibility, the plant should be operated as follows: i Decay heat removal system startup after forced circulation I cooldown: The following options may be available: 5
- Continue RC pump operation until the hot leg temperature is lower than 212F. .
- After the DHRS is placed into operation, stop the RC i l
pumps, monitor hot leg temperatures for saturation, and i control the cooldown rate to avoid saturation. If satu-ration occurs, " bump" a pump. A slight pressure rise may
\
occur when the pump is " bumped".
)r Actions During DHRS Operation:
- Iso, ste the ' steam generator with the tube leak at the steam and feed lines. Steam presure may rise for a short i
i 6-82 Appendix C, Page C-35 . DATl!: 4
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BWNP-20007 (6-76) BABCOCK & WILCOX Nueta NUCit At POWit GENER AfiON DivistON 7'- n 2 s s 3 i-00 TECHNICAL. DOCUMENT time, but thereafter will drop since the reactor coolant temperature will be lower than the steam generator liquid t empe ra t ur e . Lack of heat transfer into the generator will not permit steam pressure to build up.
- Add nitrogen to the isolated steam generator to provide a slight overpressure. Vnen the RCS is completely depres-surized, the nitrogen pressure should be about 2-3 psi greater than RCS pressure. The operator should watch for nitrogen le akage in the RCS if the SG 1evel is below the tube leak elevation.
- Do not isolate the unaf fected generator; it may have to be used for heat removal if the '!CS re pres s urir.es and l
l reheats.
- Since the hot leg is at a higher elevation than the steam l
generator, a small amount of primary coolant may cofitinue to le ak into the generator af ter the RCS is depressurized. 1.iq uid may be allowed to accumulate in the isolated gene-ator as long as continued heat removal occurs and reheat and repressurization of the RCS is avoided. It may be de s irab le to manually drain liquid accumulation to avoid completely filling the af fect ed generator (600 inches on the wide range level) if a radioact ive waste reservoir is available. Draining should not be per formed above the design pressure of the waste systems. O DATE: 7-6-82 Appendix C, Page C-36
BWNP-20007 (6-76) BABCOCK & Wi!.COX NUM8ER NUCLEAR POWER GENERAT60N OfVISION 74-1125531-00 s
) TECHNICAL DOCUMENT Decay heat removal startup after natural circulation cooldown:
In ge ne r al , the steps are similar to the forced circulation l method; however, it is not likely that the hot leg reactor l O coolant can be cooled to 212F using the steam ge nerato rs . l t N Consequently, hot leg flashing is more likely. To avoid hot l i leg flashing: !
- Just before placing the DilRS into operation " bump" a pump, l if possible, or; f
- When the decay heat system is placed into operation, cen-trol the rate of cooldown and further depressurization to w avoid hot leg saturation.
j l I j I s .. 1 ! 'O l I i l Appendix C, Page C-37 DATE. 7-6-82
4l 1 i Table C-1 WAYS TO DETECT A SGTR
- 1. Abnormal Radiation Level in The secondary plant radiation monitors give Steam Plant the best indication of a primary to secondary leak. They are effective when the SGTR is e High readings and alarm the only accident or when a SGTR occurs along of vacuum discharge system with another accident. Both the steam line radiation monitor, and vacuum discharge system radiation monitors will indicate a SGTR; the individual steam o High readings and alarm line monitors will show which steam generator of main steam line contains the leak. Typically, a vacuum dis-radiation monitors. charge radiation alarm and an alarm on one of the steam line monitors will sound if a SGTR occurs.
NOTE: A local frisk of the steam lines using portable equipment will also identify the affected SG.
- 2. LOCA Symptoms .
The LOCA symptoms depend on the size of the f SGTR (leak rate) and may not show up immedi-e Decreasing RCS Pressure ately if the leak rate is within the capacity of the normal MU system. These sumptoms are e Dec. easing Pressurizer Level best utilized as confirmatory indicators of e High MU Flow (or HPI initiation) a SGTR. e Decreasing or Low Makeup lank Level NOTE: A SGTR can be distinguished from a normal small break LOCA in that nortnal CV conditions (Pressure, temperature, and radiation) should exist unless additional equipment failures occur.
- 3. Asymmetric SG Conditions Once the plant is tripped, asymmetric OTSG water levels may develop. When feedwater is stopped e Increasing water level with and water level is at or above the appropriate zero feedwater addition, control setpoint, the water level in the affected SG may continue to rise because of the primary to secondary leakage. This method of detection should be used only as a backup because it is effective only for very large leak rates. For small leaks, asyninetric water level conditions will not develop; the reactor coolant will simply boil off as if it were normal feedwater. The differences in indicated feedwater flows between the unaffected and affected 0TSG will not be significant enough to detect.
e High acttvity levels or The affected SG can also be identified by drawing boron concentrations in a SG water sample. The affected SG will contain the secondary water some activity and boron due to the presence of inventory, reactor coolant. a 74-11'5531-00 2
3 Figure C-2 TYPICf!L SYSTEM RESPONSE FOR LARGF SGTR WHICH RESULTS IN ' A RF,,CTOR TRIP 2600 2400 - POST TRIP _ 2200 - G L _ _ _m; i 4 3 h ~ $USC00 LED REGl0N SUPERHEAT
, REGION j 1600 . 5 3 M 3 1400 -
g E
- 4 1200 -
; $TEAN PRESSURE Lluli j* \ END PolNT POST TWIP WITH FORCEO 3 1000 -
\ CIRCUL ATION (Tgy&TCOLD) AND FOR g NATURAL CIRCUL Ail 0N (TCOLD) ll 000 -
Eg NOREAL OPERATING PolNT P0WER
$40 - l @[ gOPERAil0N(THOT)
- L. SUSC00 LED WARGIN [ END POINT POST TRIP WITH 400 .
NATURAL CIRCULAil0N (TH0T) e i i t t aig 450 500 550 600 650 700 Peactor Coelant And Steam Outlet Temperature F Reference Time Points (Second_s.1 Remarks 1 0 A SGTR occurs from 1001 FP. (Initial leak rate 400 gpm) 1-2 0-440 RC pressure and pressurizer level slowly drop. e Pressurizer heaters come on; e High til flow alarm occurs. Steam line and condenser radiation alarms sound. 2 440 Pressuriror level falls below 40 inches; pressurizer heaters turn off automatically. 2-3 440-660 RC pressure and pressurizer level continue to drop because the leak rate exceeds MU. 3 660 Reactor trips on variable low pressure (PRCS 4000 psig) 3-4 660-680 The turbine trips; MFW runs back; the reactor coolant subcooled margin is lost (RC pump trip required); SFAS actuated on low RC pressure; HPI starts. 4 680 The pressurizer completely drains and the reactor coolant is saturated. l 4-$ 680-900 The RC pumps are tripped and AFW starts. The operator throttles ATW to prevent overcooling. The reaccor coolant subcooled wrgin is reestablished, and the operator throt*1es hPI to stabilize pressurizer level and system repressurization.
>5 >900 The plant is stable with decay heat being removed by natural circulation. The RC pumps can be restarted and a plant cooldown and depressurization can be initiated.
A minimum subcooled margin is maintained to keep RC pressure and leak rate low. 74-1125531-00 r
O e < i e I DOCUMENT ~
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r Figure C-4 RCS C00LD0WN USING THE PORV 2600 POST TRIP 2400 - WINDOW . , 2200 - g p --g g 0 L___ JW g 2000 - SUBC00iED y REGl0N SUPERHE AT r REGION 5 1800 - a. cd l ' 1600 - PORY CLOSED 0 w 1400 -
/ g~
- PORY OPEN END POINT POST TRIP WITH
% / FORCE 0 CIRCULAil0N (THOT 1200 -
STEAN PRESSURE - 7 &TCOLD) AND FOR NATURAL E LINii / e
/ CIRCULATION (TCOLO) g 1000 -
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/ A NORWAL OPERATING POINT-5 800 SUBC00 LED g / POWER OPERAil0N (TH0T)
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" 9 9
r Figure C-1 STEAM GENERATOR TUBE LEAKS - OPERATOR ACTION OUTLINE EVENT 10ENilFIC A' AN SGTR StuPf0u3 AFFECTED OTSG SECilCN 3 0 TABLE C 1 A00lil0NAL FAILURES 5 SPECIAL CASi$ stTH FAILURES CCRRECT IF POS$18tt l SECil0N 9.0 i M TABLE C 3 I PLANT CONTROL AT PC#ER hCRE Ai! WU SECil0N STA81LilE RC PRESSURE 4.0 I, PRES $URilER LEVEL
- ISOLATE LETDOWN I f REACTOR PLANT RUNBACK 10 hlP TRIP OR PREVENT LlfilNG MSSV'S SECil0N 5 0 SECil0N 5.0 If REACTOR TRIP i
f I f If FOR A $sALL TUBE LEAR RITH RC PUNPS RUNNING AND THE RAPT 0 000t0Ces N OR RITHOUT RC PUNPS FIGURE C-3 CONDENSER OPERATING OE PPE SSUR ilail0N
- APIO C00LOORN AND SHEET I PROCEED TO COLD SHUTCOWN DEPRES$URilsTION TD -
WITH A NORMAL RATE OF 500F AND 1000 PSI A C00t00sN gAINTAIN RCS SUBC00 LING II ISOLATE STEAM AND FE[0Lik[$ j f 0F AffECi[0 GENERATOR. REEP STEAM LINES OPEN AND U 15CLATE FEE 0 EATER. FEED 6EEOsATER AND $1E Au GENERATOR AS NEEDED T0: ISCLAfl0% AS SPECIFICAitt SECilCN i 0
- NAINTAlh OTSG LEVEL REQUl1ED
- h4TURAL CIRCULAll0N .p[ palp F Agggy As pg$$ gt6
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t ErFECT Ce t'R LE A< cwa 1 af PLW r00 Lor-m ctp , F AILUFt ) QTSG PRESS'M T4 STE AM PRES $UPE CE CRE AM AILL CAJE E.t ACTOR T8V F AILS TLN CORANT Cm'# ACTION IN CtMBINATIN .N M TR LEM . Tw!S C010 CAJSE PRESS $12ER CHAIN!% VER* SMALL Leasts DLE TO betWF5 TT DESCATING TIGr%Y el,L Had LITTLE EFFECT T4 TM TO CEPaESSLPIZE T4 PE4FATOR TO AMUT 600 PSIG AILL BE NFEt(ENT ON T4 MPER OF b4LVES 08TN . SEVERAL COMB! NATIONS OF FAILURES ARE POSSIBLE A. A SI'RE 4 ALwl CAN FAIL TO'CLOSE AFTER OPENING, STEAM PRESME (IN Cat W4RATOR) AILL CRtF 70 AR[MO 600 PSIG 15 TO 20 uI'A>TES AFTER TRIP.
- 9. A CON' DOL SIGhAL ERROR CAN CAUSE ALL VALvtS IN OP( OR ECTH STE AM LMS TO F AIL OPEN PRE SSLFE IN (PE Cm 80'N MTRATORS AILL F ALL TC 600 PSIG 3 TO 12 MIMJTIS AFTER TRIP.
T4 EFFECT ON M AMJUNT T RC5 CONTRACTION A$ PNSSLPIZER ORAINI% aILL DEPEW Ch T4 MMER OF VALVES CPEN. SET 4R FEEEmATER IS STTPED OR C[hTIMf D. APO M APOLAT OF *ey41 FL0n SUBC0aING MARGIN AILL POT BE LOST FGt SINGLE bALVE
#ENI%$ IF WI IS STARTED BUT IT MAY BE LOST IF AL T4 T@BM BYPASS b4LVES FAIL CPL.4. IF M SUBC0al%
MARGIN 15 LOST REC 0 DRY AILL OCCLR *EN WI IS AELE TO mEFILL T4 SYS'EM APO ESTORE PESSUp!2ER LEVEL. IF WI IS ON. '4 $UBC0aING MARGIN a!LL BE RECOVERED (IF IT eA$ LOST) APO T){ A*0LhT OF SUBCOOL1% #ILL INCRE ASE . T>E TUM LFAN tJa AILL ET LARER #fN EACTOR Cor1 ANT TEnfTRATiPE APO PESSLFE ITREAMS LOLP VCIOS CAN FORM BECAUSE & T'( STEAM PRESSt.FE FAILURE. T4 T'*E LEM a!LL ALSO TE@ TO EDUCE %IBC0CLI*G. M 512E OF M voIOS FCPMD eILL 2PDO ON M NLMEE.9 0F WALVES OPEN. M SIZE CF M TLAE LEM APO WI FLOR. NATURAL CIRCLLATION MAY E LOST IF VCICS IN M LOOP ARE LARGE ESO KN A*O VCI35 MAY FORM IN T4 EACT@ VESSEL 4A0 FOR T4 LARMST STEAM PESSLPE LOSSES. IF VCIOS ARE FORED IN T4 44C APO TT REACTOR COCLANT PUWS ARE NOT OPERATING, T4 STEAM BJBBLE WILL LIMIT T4 COCLoonN. 11 MILL *OT PEHMIT DEPRESSURIZATION. WI ALOE CAreOT ELIMINATE M VOIOS. MAT etJSY E Eh0NED BY T){ STEAM ETRATOR TO CCWLETELY ELIMINATE VCIOS. WI, 4MEwtR. d!LL NOUCE T4 vCLJ( OF T4 WCICS. LARGE STEAM LEMS NILL CAUSE AN UNCONTAOLLED COOLDCAN. C00LDCm eTTH TEN Tav'S RILL RE3; IRE THAT EIMR T4 VALVE 5 EE EPAIRED SO THAT T4Y CAN BE U5EO FOR A CONTROLLED C00LD[mN. OR THAT FEEDRATER E \TRY CAEFLLLY CONTRCLLED TO A DEPRESSLPIZED ETRAT@ . REACTOR COOLANT OVERC0 CLING MAY v11 ATE M TUEE TO-54LL AT LIMIT $ APO PLAG M TUBES IN TENSION. TMIS MAY ITREASE M TWE LEAK RATE IF T4 TWE CRACM IS PERPDOICLL.Jt TO M CENTERLIT OF T4 TUBE. 74-1125531-00 t
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C* EN liff 9TO, fd 3f tat F E f DRATE R TO Te( G11) G oaRAfgp Ard BV $TO'PI% f ff DRA t[ R TQ b( ((I'54 5MillD Sil Ass GM NAfuR STL81 Lift T)(. M Acf0R CIII. ANT #f ?H WI, C(MRI RJp( AT M T9W MAY f>(N Ef PitSt@t0 '3 LIMICE AS DJTL INED MtM P8455tPilaflON ty CrMMa t!% WI MO 8Y CthTMILI% STEAM (1DUE CF he MSIv mouS T4 *ATH 70 f>4 CUPet NSE R P5(Wpf RITH M 18v &s T>( G M RATJfs T>df DIA S M)T HA4 EvtN IF T4 CfN!a N'4R 15 MIT AVAIL Aa f af M $f ART Or & D*. $TI AM Lt AM . JIA DLMN ULIC Di Avie 11 MAY E setiT3HI D Af A L AttR Tint 70 (X TDeMM m( be A b4 L E AM !5 T480LAA4 T4 19v OR b4 94a A D Et Ust D ID( M VtR Aval, AM E g Af ft +T ft) MAMdLLY (lora M itiv' t IN t.( rA'd RAfrse 57RCS MAf If BVPA5'lD Af ESO PSIG THis n]LL Pnf vf NT T)4 Msiv LUt? CIfM M STI AM (f an (MJYE 1F b( CONTM15 HAVE tmM CLOGl%, teAINT AINi% A PATM to M Cop 43Nsfe. f EEDAATER FilLID THIS MAY MJT nrsak) IF YMIS ACT10N HFlitM S 5ft AM VO L'l 80 #4454 ALLV 15LI AT[3 AF ff R 8vPAS$!% 5f RC5, f'HE 5tM. , D(N b4 $Af f TV v4LW $ MAY f( {LIMIpenT[D Ag M (111. V!5Lett v ( AAM!st f>( T 8v ' 5 CLOSE bt va' vt1 teAP4ALLV i!N M HMdet [L (m CLCfd nf Etn M WALvt1 A5 AP8=sHJSIAft' (N1 M R ANT HA5 al[N $1A61L!llD. C04HICTIthS CAN tt MAit 70 PthMit CCIIIIJnN bi 66 btO FCm Ctn(1)nN e:LL CIPoo (N m(De R TQ Sif AM Lt AM CAN tf ChtRt1LfD APO Aft L OtPtPO (m aHICH G 'd MTOR HA5 ft( hAf Lf AM . Uf f0f4 C011XW4 !$ StGM, T6( TBy sCJ$f Elfte R BE M ST0f4 0 TO 5tas!CE OR MAMJak.LT CMTN[lttD T0 sitsf0f( D( Tav't TO 5tRv!Lt. ATTEPT Mea;AL CtwTfat. atTH ni HAPO/ AUTO $fif UN, IF TH15 tri; M)T hjRR inO 90!CE5 ENIST . 4 8%#Altf CONTIEL Di 79# in utR v4LvES nITH M HAtivne ELS. ,,, _ IF 4 f a0 Cf Pd RAf0ft CattIwo IS TO 6f UsE O 15t t *M4( l' Ato *estE 2", F IGL8( C.3), Of VAL vt IN ( AO* L, a \ v ,,. G4 RATOR 9(110 ft tf40 M RIManING v4Lvis stut0 K CLDst.D Lht!L M LAST PAR 1 or cturanN mg N p( $ tt AM GMIATOR AT IS sMAL1. IF G( G MRATOR 15 TO IIE (fiD tiert 3 0F F19pE C 3), M Tev'S IN DE *0tI1D* CAMRATOR 9n1D 8t Mars;4LLY CONTPKILED By b( Hedw(ELS T't VALv($ IN THE OM R (ANE RATLWI 94110 BE CLD'AD CLPING CIIttXps mITH OPd EWATOR DE DG SK.[NCY hAf TQ 9(LL LIMIY5 9G10 ti MAIN 1 AN D APO G M.RAf@ 1hr ENT'lHY CONTal, SenLD BE IN
.eCGJf0ANG WITH M 94 E'JL AAfDR C0ll!1#4 APMIDACH
[($CP!310 Ine Pt af 3ftP4NT art *ATICN* QW'TER JF ATOG PARf .. 9 IF M TBv'l CAteOT BL U'40, Th( ROCM VALvts SnA D BE Mania 4 Lv UAfatifD AS & se$TITUTE FOR T,4 Tsv's . It Etn(2 0F M SE CASES. M G *d 4ATOR a!TH M ft4E LE AM . *10 O(Y Q Eft Ap4010 LIMIT THE aAfin LEWL APO M vALM stetc 04 Y U 04M O AS M CrisARY TO 4tp M aAT[R LEvil LON Ca100mt SQLD If aITH M 84NrRAf0R 9]DfLT b( ftff LI AM
#4 P4 vt R P055111E .
IF AT ANY TIM DLRi% M C011XinN M mATIR Livf L OF **. Af f fCTf D Gpf PATUR Atht%Of 5 951 tm b4 CFfRATE RAP'.d kT AL/4 0F . M ftfi Li AM . Def G M Raf04 tMT If STI Ap4 D TO RtDLIE M LEvil . Ft trmAfiR 941LLD BE Lite 1TED TO M GM RA'OR mITH M ftsi Lf AM 70 4v010 FILLING DE GM RATOR. EMID4 #4ft# LfWL 9nLD BE edINTAIM D TO MEEP T4 ft4( YO 9(LL af IN LIMITS APO TO MAINT AIN NAfuRAL CIRC 41 Af!ON IF M OvtRC011NG MA$ CAU5f D v0105 To r&et IN Di LCFPS b( RC Puf'$ 94ttD HAvt illlN TRIf+tD D( 70 SY$itM SaftPAT!aN M WI SYSTIM $*n10 It STAMTtD TO at Fill M #Cs APO TO RFSTON 5Lafa11NG. AFTE3 EArtil]NG 15 NESillfD. $ftApe Rat (5 MAv ST!U. tutif IN M if*11 t GPS OR IN bt it AD. V~l3$ CAN lif MlMIMD 87 GIR ING WITH D( 5f t AM G *d RATOR$ IF be 5s1 TIM I$ SLKOR TD, bt 1C POF$ CAN GE h(5 TARTED APO Em TO *aAM4* D( $tt AM makt INTO DE CT[ Ae4 QNERAfDR #44 If C AN EI C041N'dD 8Y 171888 Q4 MA70n att1.e; IN 4 Amt.r ta.L CASE $ FaR tLsn Li An$. b( WI SYSTIM #1LL REST 3IE Secott 1%.14TtitATION #!LL CONT!*al na.Y IF Lines gutST IN Mvf NAN W hAE , OR IF ANJM A LOCA bd$ ef TAN IN h( 5f C hiME *4 L Y RFt(tTE CAS* 5 DE " POP F1E ST ART CP! fERI A* JF M
'f JL;1P*(NT O'tWATIW CHA*ff R OF AfDG PART II WD BE USED I
w wm on 1%perture Card 8403050118 - o
I e . 4 f al.a.*L {.UE ".T <w t g k OTsc eaEssa i T4 Er*EC'5 CPtN&% T4 { M&!N STEsp haFE TV CPv4 f ME C' valve FAILS CPEN 5' Eau saFETT a5 L')G AS F. DE *9E 55URIZE CONTI41. SINCE T4 57-
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N w1pEO T-C31_ Jinn nest TUBE LE as . IF T4 TyEE , M F AILE3 54 AILL GE vtRY CCNTI415 AM 5F4CS MILL a< CLOSIt 74 : TO T4 GE4ma OTSG PMs5M T4 SI'AIFI341 RACIDACTIVE STEa avv FAIL CPEN AT*eJ5F49E IF T' 24RATC# A5 T4 IF M C3&TNSlp BE U5E3 FOR C011 SF SIGH 1FI3NTO aFTER T4 OT4R 3FaC5 AILL ACTbs IT M COaINSEs R.0CMED etN T4 R.CERING 5FRCS e as. LOR M TBv'$ R*4ESSJ3E 4 FAILUES RE541TI F AILED Q.OSED PRCIALE M Fh T9v
- a 1 ALL '9W'S Oh
- 2. CPE YBV FAIL CLO5LPE & M T M COCE*EER 4%
& SETrad#Y iaF CONTE 3_/LIwIT ST CLD5uRE & aLL T COLLD (CE5SITAT C00L3DeM . IF Tw WFECTE3 SG. MIL DLPIW T4 LChG AILL EStLY IN a GECafi M avy' CDesRED TO T4 F4ILURE & O( '
COOL 3DnN CAPa8]L edvE awtf CAPat CCKL3DnN EMRE m#TJit . OTSG P4E55JIE S. IF T4 79V'S AE Fa7tuRE$ FFECT FAILED CLOSE3 T*E ABILITT TO C avi (INOPERAT!vE SERI3USLT AFFECT T9V' S ) CLOSED. plant CC BE Ih!TIa'ED es*I CDCITIJN THE F IF CORRECTId AC I. M Lian Rat ImENTGi? #1 LOSS & *I CCPCITI3N
- 2. DVE8F ILLIC a.TITION TO 10 SECOCam FAILURE OF CNE &
CPE SG C3L2m USED F3R PLANT G CDN'AWINaTIh aK IF T4 CPERaTIVE FUWTING & M ( >LE5s uTE w
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74-1125531-00
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f Table C-3 EFFECTS OF FAILURES ON STEAM m m a m u -- e a m 'om w w m i w Sum Suttv vu ralI% a f al gF gg tBy mt FAILED M m am na r 0m a a* t s GENERATOR TUBE LEAK CONTROL y vt t+4 S te am tr as a s wt Iv0 !% 0 7",v
, Put 5TM 1 ) i s'f > f f na f a nes '% og a qs 3
..v Cmi . m c , . ma m,,,_,,,,_,,,, " ' ' " Dafts U'RT!sek i 70 f*4 F n ;aan g '9 M g g pa top al% 's ste am Le ar SHEET 2 0F 4 c eG ra tuR . Ovt W.') t ts, a h L PRrr 5(3 m:YH a are q w Pats au 1ers as out 14 3
!% M 1 Juldtf NT fft Pat! h* C,se!T p afgG papf gg hP5'F '.?!d a',?I'pc F"p vtgas s oa a q pt ss cpeq c 8 CDe natrp en arm % agtt gg C4 f.aftp f3 9t t/4 C F Upl I' h4 Y Of l'A A5 OUTL I'd 3 IN C f W **f i m M 1.
- M*f M;f W T>g ryd eItu m g, y ," 5 *dt 40 V M D a'e] IMatt 3 Filid&M4 TO M Sft am U *e wa frP eug at AC T [7e TO IV1 a tt 15 *6d an IS !% Ts4 game gi ng AaftA 45 M '4 55a#v T4 bre S SI9eal 4f;f Rf eenae Ativ E Ds.D IF 4a%4. filma't p a[Y3 t!'N TO T4 E ff wat W . R'.CIC/f'IvE WILf a'A S
.rks as t (M; a5 tit i1M Lf a, 3P8f 5J!l!3 Sff am C4 ss aaTP 15 *ta!4I3 70
"'I*** % CI'FJda t h8' I% *** T LO# a% TO **aI4f aI% T4 FT Ca*G Rafff' Sftans I'Jul 7C Sb4 LL af a] THIN Llwits alp!% C(g11mps 74 AM CDs LC3 STlam P'E WM - 'J uri sinment s>ua a at 50 et a tx ut3 to M rd ed navan
!v a@ Cuf f!% W F 4 L fit;*AM 4 LfJ 3 FOR C/11 Des TO PSE vt %f thea%43 Iv1&TI'Je 4.5 D f!TM T4 tiaaIE Saf f ff. TM Sq aw W w aa?tp [5 ra enr 55#1?f J IF ass #vy Fi!LUML 15 ina r f4 ttsme:TIvt atfl yri f t,p a F AILE3 ',.rth aW am T4
' tast G wif asta Llgt; tty fD las( 45 FGI Te( 1SV 5 { WE OT% PsE WM 3 } #f fte ry( TL3 LFan 15 !*e f>( lane I tt >T I'N CtNYie'4m av!Lagh tly. IF T4 CJNf *(JR 15 aballam E , M fBv'S mud
;l MJT avi!Lalf. avv. se6f gf t.r43 a% ?>( F AATv siv SM110 lE RfDED i mme M CGl'7JAN # aft AILL , , g,q g,
+ 381# If Ga Y Gd eti 15 V'lJ D E PWW D e 4 I AW. C; n isrwPs elTM M avv thL TAPE &% (sTRt4LY LChr, t ON LC2 C;te AaTLp P4 WM ?!4 and) M C01DL4Pe Ra'E n!LL BE SLanLa af T4 t%
!5 avalt as f , 9 mJ5 5s4&a D SE OF T4 L31 DL*IN D av';ID RMe!% CLJ OF tspr,f EATER 60 P5]G ALaese 15 Rf C([WD ( Rail 0F LOSS IS CE*t41NY ON T4 Y,fiF Llam Ra't j i L Pro vt ut se51v CLMWE , 4% GP (fd UF SACM# #4'i# Fl* C'.PitN5afE T4 Ratt OF 1 G U5f D FWI Cf1L DI*pd CE L Al a4 GF T4 is#57 aPO CWINWE LEVEL SPOAD SE neA4 TONED APO Citf' APED TO T4 Raft Cf Cy] Cteh IF [f A*"t 4R5 Twaf f4 Sf0ptD na'IR SutD:25 GILL CE*LI Tt ,
7 4% nt 1%5f a!E STEml% FJR CD'I D(vee alt 6e pKFH G4 RATURS Evt's THrtme4 nacIaTIJs a Lia5E5 shi OCC#
. I't CLOS M CF f*C f Bv's CaN IF M 4RIllfY TO Ot**' 5f tw 10 T4 Loran %sEn Is (OST ,, ,
IE C084 ITICMS 88UM N OR Bufw SM ae M M eaTys, T4 aw'5 CA% at , , Uf fLIIC FOR plamt Cort.aosi% st atrJ v5t Cy ng Ygy's si, (m DTM %e f ait CLCf4J #!LL 88INIuIZE JFS!1E MLla'A5 4% T4 [.31;ImN CLLyd a PE RILX), M (Pt.RAICp Mu 3 Af ft*f TO RE T fee T4 YBv ' 5
^
all PsEvt%f5 SYtam ;h.me' TO Plac(5 mi PELIag-( Ch u5t 1 4Tftff arg;q en&%AL CCh%1 Fame T4 Ctmetu Y v4L G 5 a% a We 70 80 " we POE 59 F4. 2 Ud ta M 5 tat'Ja CF M ate straLY FWim ANO , fgy's CN w STr ase ca se eaTOR ifi & M aW'e FGI Plan 1 !* M &&M aC TI h5 00 M.T af 5?JE TBv's rrtnaf1,'le.. Lvf 5 K.Af fif LFi(3 LN f>f Pl a%f COLJCAN L6IE T4 avv'5 a5 MCE55aAv SMRAD fpF5]ft 4Lia5f 5 #!LL (KLp JE Sf ap'TJ 1% T4 (q% frew FLP M CAM e>(at M e ins, CLlaM E fue M ano 'Bi'5 a4E tiO",t3 Ge 14 AFFECYt3 SG IT IS &?vI5aa f . tnan a PL a%Y ColareN PtnIro 10 L0tativ 0FT% usI% M v.4 eenate>< rLS. tPE Ter aid & Li55 CaPLCITV IF STf amIt 15 n[gilsIED TO P5E vi%f % FLDX'I%. Tw15 !!v ' 5 aCTI% 15 aIW3 al w!%IMIZIM; M rJF514 ettta#45.
# M'43 0 %f If*.14 Pla867 II 5. M as eenI%I% T3v'S Y 70 PETI (vas M fie *Midiv FCJ 5ttam MweafGI T;st WCnarl4 apa7 aWiflopent STI w psE WFE COV'Ir3 (IE , T4 ASILITY TO upastLT '( OPl9at!t1pe (7 T64 anv(5), N JULE S rt an P81t S5iM i 15 Of 514a% [ Im a:TM Xweg M *t a%Y CIlt at G ia na t ps . 54113 M OR SC% Cs T4 ans At CCD4 1 54410 G3f41;v-5 pail I*anwa'Ivt . LOCat esagaL CRANI N sitam PerssM i Taps ;V of t Ct311% se61 #ILL Rt 8E Ju1M O. I T4 CME %sta 15 ava!Laat If n 15 vt R' % CJ p .y 504 a 15 95 5T 70 aTTr*1 10 enag A.Ly Crt eaTE 75f itv 5 t Ltm!% (Destataartl may (vtm.vf 5140 M y AME & MIGP4 # CAPac!YV ham 754 avse at IJss ar( upsstrC[5M1 MI4 USE mI;L wINIuttE avstalC STEau REita5E r!LL CCNTI'4E . AM' tha51 IF M( T94'5 CAPNOT 8E (P010 on nf CDtJthSER 15 < SLfJL Y yCFla5E , T4 [vt %% '8M avaliaft (, s'Y aps 9q SSgp[ eg,j$y gg ggygg g gg L3 Alv i15I4, THE AW fag!LITY en&T L[a0 70 ame ICC - M AF FIC ff 3 5G A% #4 9 4
( STI 488 LIO. M TO **I'84#' f an.4 Came()T BE avoIQ 0 v ma V TILL PE Fist USE & a IF f54 af F EC'ID SG atis? lE 1&ses, MI5313 54ais PL 4%f C9F s!'E #f,EiW S AtiL Offut 24v 15 (M THE to,artictrD SG MCff 0 SC CeeCY lt '91vt%40 I s4 aCTIgpa !$ f ap.IN 50 Nat
.MPAC CAN C2 CIDeMA LID Ehildre on 8403050118 ~g6
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f' l i A FAIL;M EFrF T (ps TJIF L g OTSG IwEmCtry I. A F AILJE WI) CITRATPS CAN We OdRFEED CAJ4 RCS Od90 occur . T4 ECESSIK FE P913R TO T4. 5Ta IF ECE551VE FEE eITw T4 TLEE LE, STEA*tD TO JJ5T
#mNGE DURI% T RELEASES RILL C OTSG IN(4TCt.Y 2. M EFFECTS CF AF A @.tRFEEJ (W e As Wa OVERFEEJ 5TCPPED)
IF T4 OvtWEll MAY STCP. e4 TJ414 OR!vER RATE MAY 2" AM OF STE HDRF.VER MN 1 T4 AF# PJe e'
$TE4W PWTIl IF DMSSIVE Al OROP BELon 600 M SFRCS SYSTI MAIM A@ AJRILi E4RATOR5.
A DEPRE55#12A CabeINATION AI SFRCS 4AY 4CT'
- IF SFRCS ACTUA ETH GENERA (P Fe AEIID WI OT5G INVENTORY 3. FW 9OLD BE PR0vil
@T Av&!LAELE CDRE L3S$ OF AFe Me WI C00 LIM. G T0 . oT5.., -mD I. M .
Aft 4 GJIDELMS SFRC5 WILL ACPJATE GE4RATORS. SFRCS RESTOND. M ffE MIPEM M usive STATIJe i I I 74-1125531-00
F w ,e,. n u m , n - .A, , ,n ,, am Cffe NI Aeg.) Pg. ANT (ryg lg mN iyng " 'I ,e A; t [ % T; 44. ; (4 & O'M A (g , Ofl .'.y mN m a 55m .e. ,2. we o um n.r m5 ,, 5 m SM , m.A., . m Taole C 3 EFFECTS OF FAILURES ON STEAM
, ....,.3 M e,A og5 m n A,m> Am.a e . Am. . see > 114e 15 IW ' ,,c5,u, R ,AI m - n se .- n.usn A.o ITwix GENERATOR TUBE LEAK CONTROL
- w. . n 5- m Aams ,
of R AllTI'h MJ5T fiF 5 fJo*f 3 1 AT'l*F'f TO O Ck M FiisAA'l R C0e'F41 vni vt 5 y 4 UJ A 'XAse IF h( QW W Itt I5 %(4 JJ CIM:T gat ,q u 15 .ma n M m oa,
***'aa""-*2-"'"* SHEET 3 0F 4 2 IF M C. pew 1 v4L VE 5 3) M'T 'JPTRATE . TRI*
T*( f s(tS$1vf (Evt L est at Ce t'A Os f*( rythATING 50 R(5 5 STI AuI% *t RI'O . #AL!'.ADI A 5 IF M (1'W 4F !s t 15 F a5' M wA m F f pA MM DIM P)*5 CAN f( S TT'"t 3
. vinIrv u n PtRAt!pg IF '4 PRE s2 pile # MAS ORAINED. eu AT Pt M157 at USLO fu REFEE LiVfL 70 t!w!f MLIAMS A% TO L! wit saff e ENTE #1% INTO T4 Sf t au LIM S . M G 4 RATP a!*H M Tift LE AM 94R.LD RE 5'f 44.D TO MLL# D( AATf R L(VFL JtAf BELOW 9% LN ft(
CFT.RA? ]% RAME , IN'T 9W[f T[%f ME&efl% e44V BE HE114L7 70 ma!NTAIN D( ea'tF, t Evf L 45 D4 Lf AN MPeT[*e15 TO AUD AA4R CAGl% M COUL.Xm . IF M ErJ 55!d FEEC*ATER
#AS IN f4 M me R&fDR AITHtRJT h4 TM LE Aw I f MA Y BE 5?t A4 0 TD h( app 4tP#! A'F AATE9 LEvFs FOR CPERAT!aN (Rt PMS *Ge' OR 'G F ") . % ANT Cort <f_ee 9013 If tsy 74 "Gf11)" 5't Am MM #4708
'O Ptaw!T CmLO,mN at a ni &M* eat 1 E nATE , a&TE R es.sf !E A0iT3 70 b4 STEAM My eafW AIN M LEam 70 MAIN'a!N Aas ACCEPf as.E fdf TO vt LL A7 IF nf OvtRFEED IS N THat Gi'd RafCit, bf 4E SM.n A 0 (N % PROLEtt 5!4 [ A L ARM I4vt NT08' HA5 ACLL841=TED H M vt R 15 ALL FE E,ma'tA 11
,,I 5tFLD TO O e. M 4 9Af 0R APO IT fCILS OPr. D( *# A # A TE R LE vf L *eJST RE ADDL3 TO ALI Os A Rf t9haet_f RATE CF CD13tm j j IF to e CAPfLT BE us(3 s e CAN BF STA#TED.
4 OWRFilo APE MEN Df 5Ame CtM9E CTI4 ACT!ws FOR u# OW NF(f 3 Asg etLH M SA4 A5 , (0133 INvt NTOrt 1) . FOR es e . MV A4 k9ee&R!!EC tt SE g , 1 < ,,,.
- 1. QD5E D( up ik1 AT[Cre v4LvE5 TO F( DVERFILLIM STE se Gad RA'OR
- 2. RL ALIGN T4 u n P)+'5 70 FIED DE TERaf!% $7E&at . n,.e I% CENTIM A S. Sit 4M P'(Asif tm " "A " N M "#
THIS OLDoe.', De um f tLCfus 5. IF M 15nL ATION v4L d5 Emed;T BE CL0stD D4 N D( mat SLG2 ULwe &% P'881D FLfm 0ROSSCVCR v4LvE Mar BL Q.040, #O D( IS1 ATED 'i , THIS EFFECT DEPf 44 04 M PM CAN BE i#IP*ED TLAPFfC IN De STE A4 L IM 5. ( E AL(55 FEEfmATEM D ST&#LD. ONLE twi OvtRFEf 3 CrJCITIJN Hai 5TCPPLJ D( MtK RA TOR 11 N594 greeAL SPtIO 45 b( AND be F# 5751E4 Mt.IST fE M TJad 0 T0 sEwv!M FOR N II1I #. UIA DimN IF STE Aas estSSURE MAS CECRE AWD TO 612 PSIG. e C4J.E5 M STEus PRF55@t TO VRC5 #1 '. HavE ACTLATE3. 74 GENE Raf0R Maf ed<t &L50 "$10 IN Os OR Obe G 4 RAftFS, GCILEO Ok
- N Gn1R To REstunt 14 uw Raf0R ANo n(
w 1D t ecf t,' ATE uo !$r1 ATE tuffH FW SV5 flu 40 SERVICE, DE 5FRC5 SIGPeAL MAY BE ELOCKCD arf ttflarLA T3 (M Op ETH eO A RATER LEAL IN D4 M*dRATORS RE EStaa !MD THI5 t A LY , AF 2 EL G e AILL NUT C8M5E RILL '5EMCE FDR 341L CML ING IN D( "AFTECTED" MNE AATOR. WO A ChfMA LID FLOW IN D( "GIE" gegRATCs4 FOR Cat (1spe fW W THIS A417sf HM Vf R. N h THI C013m Cf *elt$5uR!/ATION. f1 tNf 3 M51 C!RLise514M E5. i s !.80 5775 ALL FLEDeafER TO
- JAPI QIlING es4f at St 4R TED te M SU F3 P M*
gD ev D( TJ F1 PJe IF p a 15 5U re P# OR etJAPI CM1% (SEE saCx# Cart!NG 1 at;57 eE T111% t A7 C eenNT AMD Sv INITIATED TO 4AINTAIN CDRL QXLING RITH A TOTAL iO55 or ID4 NET PC) eAL4f t CD11% 15 FEED 4ATLR, h( SG'S WILL SDIL Drf AND 3*eL53pI2E gp( AP an!%" $ECT13e OF M CM11NG CR MTENTI:Pt OF A SG #ATER LEVEL WAY OCCUR IN M ArrtCTED SG DE T3 D( PRI4AfaY TO SEQ)CARY [E& GAGE FtECmATER SW10 nr PESTPED TO AT LEAST &( 5G AS Sf1N A5 1051% De MsIVs
- ETH POSSIR E , A TI41Y 45TORAT!W OF PRIIeARY TO SECONCaRY 4ILL C5t.T IF FIE0nATER 15 4 AT ThaN5FER 15 NLQ5$ARY BECau',E >F! COllING PLActs a BIG
'OR WILL , '441L vf A MAvt 70 MMMO ON BR5T INkTNTORY SING Pett OF M BR5T INVENTDRY let D4 CINTEL 'EIN CHe#O 15 +0T RECOVEPA&E (IE , LOST TO M 5FQDOARY PL#et THROUGH M LE4p), DE PU'ENTI AL F38 & LD510F rDRE CJ11NG CAPABILITY NLREA5e.s rRFCF*!%ATEuv RITH D( TM PERI 30 THAT
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" GENERATOR TUBE LEAK CONTROL
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BWNP-20007 (6-76) BABCOCK & WILCOX wu ete > NUCLE AR POWER GENERAtlON otvtSION 7'-ii2333i-00 TECHNICAL DOCUMENT APPENDIX D LOSS OF OFFSITE POWER i.0 GENERAL T_RANSIENT DESCRIPTION Loss of of f site power is a failure to provide power to the plant aux-iliaries from an offsite source. It is not an " abnormal" transient or a failure to control the plant unless coincident failures oc.:ur. . This transient is initiated by the plant separating from the grid due to a grid upset. The unit auxiliaries are powered by the main genera-tor through the auxiliary trans fo rme r during both normal operation and grid separation. Therefore, power is still supplied to all plant auxiliaries and the ICS (by design) begins to run the plant back to 25% power. Since the turbine was supplying 100% power and " house loads" are only about 4% full power, the turbine protection system must partially close the turbine throttle valves to prevent ove rs pe ed. This results in a sudden increase in secondary pressure and temperature and conse-quently an increase in primary pressure and temperature. The steam I produced during the runback will be dumped to the turbine through the turbine throttle valves, to the condenser through the turbine bypass valves, and to the atmosphere through the atmospheric vent valves (AVV). 7-6-82 Appendix D, Page D-1 D ATli :
BWNP-20007 (6-76) No. Et BABCOCK & WILCOX NUCEE AR POWER GENER Af TON DIVISION 74-ii2ss,i-na TECHNICAI. DOCUMENT With the present pressurizer powe r operated relief valve setpoint (2400 psig) and h igh pressure trip setpoint (2300 osig), the plant will trip during the runback on high RC pressure. The plant auxil-iaries will switch automatically from the auxiliary transformer to startup transfer wh ich now has no power. Therefore, all power is lost to the plant auxiliaries except those components loaded on the 125V DC battery banks. The Safety Features Actuation System (SFAS) monitors vo lt age on the two engineering safeguard buses C1 and DI. Upon detection of under-vo lt age on these two buses, SFAS will start the two diesel generators DGl and DC2. Approximately 10 seconds after starting, the diesel generators will begin to accept loads. After the diesel loading is complete the operator should start a makeup pump and re-e s t ab li sh makeup and RCP seal injection and ensure the instrument air compressor has started. Because the reactor coolant pumps have lost power natural circulation must be used for decay heat removal. To establish natural circula-tion both the steam driven auxiliary feedwater pumps are automat-ically started. Water from the AFW flow will be automctically direc-
- ted to the steam ge ne rat ors . This AFW is desirable when establishing natural circiulation because it enters at the top of the OTSG and will provide a higher thermal center for heat removal.
Appendix D, Page D-2 DATE: 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numset NUCttAR POWER GENERAtlON DIVISION I, 74-ll2ss31-00
\
'lTECHNICAl. DOCUMENT f
L/ After c LOOP and consequent main feedwater pump trip, the SFRCS will actuate on h igh steam to feedwater differential pressure and close the main steam isolation valves (MSIV) and atmospheric vant valves (AVV). Thereafter, the only steam dump will be to the atmosphere
} through the main steam safety valves, y[ )
L ./ Davis Besse ! requires at c'; level o' 40 inches on the startup range to establish and maintain natural circulation of the reactor coolant. This is below the OTSG level when the plant is at full power opera-tion. Cons eq ue nt ly, after a LOOP the AFW will not supply feedwater j until the OTSG 1evel boils down to the 40 inch level. l
/c\ \ /
l > Knen the OTSG IcVel reaches its setpoint, the AFW will maintain this I level. Steam pressure and temperature will be controlled by the 1 steam safety valves. Primary pressure will increase as makeup flow l raises pressurizer level to ita normal setpoint and pressurizer heaters heat the pressurizer water. With the secondary pressure at the sfacty valve setpoint and pressurizer level at its normal set-J'%
/ \
; point, primary to secondary heat transfer will again be equal and the
\ _ ,-
plant will be at steady state. Actual Plant Loss of Offsite Power On February 22, 1975, Arkansas Nuclear One, Unit I was operating at 100% power when a storm blew down a 500 kv transmission line near 7- i ( Little Rock, Arkansas. This resulted in a loss of of fsite power. I l Appendix D, Page D-3 DATE: 7-6-82
BW14P-200G7 (6-76) BABCOCK & WILCOX Nu sta MUCLEAR POWER GENERATION DIVI 510N 74- 12ss3t-00 TECHNICAL DOCUMENT The reactor and consequently, the turbine generator tripped due to under voltage on the control rod drive breakers. The reactor did not trip on high RC pressure as would be expected. The control rod drives are normally powered from startup trans former #1 and make a t rans fe r to the auxiliary fransformer upon LOOP. During th is trip, there apparently was a momentary delay in this t rans fer and the con-trol rods fell into the core. Buses H-1, H-2 and A-1 trans fe rred to startup transfer #1 which had lost power due to the LOOP. But A-2 trans ferred to startup transformer #2. Following the reactor trip, the reactor coolant and main feedwater pumps tripped placing the plant in a natural circulation mode. The diesel generator GDI started automatically and emergency feedwater and makeup flow was initiated in about 30 seconds after the LOOP. After approximately 4-1/2 minutes, bus H1, H2 and Al were switched to startup trans former #2 which was powered and one RC pump was started in each loop. 1 Figures D-1 to D-5 show the plant data. Figures D-4 and D-5 show the OTSG full range level and steam pressure versus time. Since OTSG l level was slowly decreasing, the EFW flowrate was probably very small. Although the OTSG level is to be raised to 50% on the operate range after a LOOP, the failure to do so in this case did not ad-I versely affect the transient for the first five minutes. However, failure to raise the level for an extended period of time would have led to possible loss of natural circulation. DATE: 7-6-82 Appendix D, Page D-4
BWNP-20007 (6-76) BABCOCK & WILCOX Nu.ste NUCLEAR POWit GENERATION DIVISION 74-1125531- M TECHNICAL DOCUMENT The slow oscillations in steam pressure are probably due to the steam safety valves lifting and blowing down. Steam pressure leveled off at about 1010 psig which is the setpoint of the MADV 's. Figure D-1 and D-2 show the hot and cold leg temperatures versus time for loops A and B. The initial decrease in RC temperature after trip is followed by a slow rise in the not leg temperature. The hot leg temperature should be approximately 2 0-40F higher than the cold leg when natural circulation is developed. The cold leg temperature will approach the saturation temperature of the steam pressure. Figure D-3 shows the loop A RC pressure versus time after trip. After an initial - decrease due to T ave decrease following a reactor trip, the pressure begins to increase. This is caused by the injec-tion of makeup and the reactor coolant expansion due to the increas-ing hot leg temperature. Figure D-6 is a P-T diagram of the ANO LOOP of February 22, 1975. This plot is similar to but not exactly the same as Figure D-10. ' Since the plant tripped on loss of power to the control rod drives, there was no initial spike in pressure. Figure D-10 shows a temporary increase in both pressure and tempera-ture before OTSG level reaches 5 0%. This increase is due to the in-crease ' in the AT req ui red to establish natural circulation. The V 7-6-82 Appendix D, Page D-5 DATE: _ _ _ ___-_ __ l
BWNP-20007 (6-76) BABCOCK & WILCOX sumien NUCLEAR Powta GENEaATION Olvl5loN 7'-ti2ss3t-00 TECHICAL DOCUMENT plant data for the February 22, 1975 LOOP does not show a similar trend. This difference is due to the long time interval used between plant data points. The following switching problem transient, while not a LOOP, ex-hibited some problems associated with a LOOP. On November 29, 1977, Davis Besse I while at 40% FP, the reactor tripped on RPS Channels 2 and 4 high flux. Initially it was assumed t' n e trip was caused by a dropped rad, since the operators observed an in limit light on one of tre group 7 rods jast seconds prior to the trip. It was later found to be caused by not removing some jumpers in the startup reactimeter patch board wh ich connected reactor demand and feedwa ter demand to two control rod rela t ive position indicators. The jumpers caused reactor demand and feedvater demand to go to 62%. The ICS increased feedwater flow and pulled rods to go to 62% FP. The reactor tripped on the overpower trip setpoint of 50% FP. The reactor trip tripped tho turbine. At TECo, there is a 30 second delay after a turbine trip to automat- ! ically trip the main breakers. This allows time for the automa.ic throw over of the house load to of f site power. The operators concerned about monitoring the generator noticed that the main output breakers had not tripped, therefore, manually tripped them prior to the auto throw over of the house load. With the main breake rs tripped, the generator starts to coast down allowing the DATE: 7-6-82 Appendix D, Page D-6
BWNP-20007 (6-76) CABCOCK & WILCOX wumsta NUCitAR POWER GENttADON OlVi$lON 74- 25s31-00
^ ECHNICAL DOCUMENT house load to drif t o- of phase with offsite power and prevents auto throw over of the house load. The diesels started. However, one of l
the two diesels tripped out on overspeed. In the next few minutes, the following events occurred: [] a. Had a N15 seconds station blackout. ( ) (j b. Tripped RCPs circulating water pumps and the makeup pump.
- c. Maintained natural circulation for about 15 minutes. Had a maxi-mum AT core of 52F.
- d. A cold leg temperature drop of approvimately 55F in 3 minutes.
- e. RC pressure drop of 410 psi in 7-1/2 minutes.
The following events are noteworthy: (-]) i
\v/ a. All RC pumps tripped and were off for a minimum of 14-1/2 min-utes. In a little more than a minute and a half, the seal re-turn valves closed on all four RC pumps,
- b. All during this transient, a minimum of two CC water pumps were in operation. However, the plant lost all AC power for about 7 seconds during the early stages of this transient.
[.m '
- c. The alarm printout shows seal injection flow to the RC pumps was N '~ 'l; very low-to-none for a period of 5-1/2 minutes. However, the MU pump was only off for 2 minutes and 57 seconds.
- d. Some dif ficulty was experienced with the component cooling water cooler on the No. 1-2 RC pump that prevented it from starting.
This condition was corrected by realigning valves and pump No. x
/ \
2-1 was started. ( ) kj DATE: Appendix D, Page D-7 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numin NUC1 tat POWER GENERATION DIVISION 74-1 2ss3 t-00 TECHNICAL DOCUMENT l
- e. Two (2) pumps started af ter 14 minutes 27 accends 2 and 2-2.
2-1 and 1-1 were started af ter clearing a CC water flow problem with 2-1 pump.
- f. Two out of the 3 pumps were running at all times. However, there were reported CC water flow problem to 2-1 pump.
- g. Seal return flow was off for 13 minutes, 25 seconds. Based on seal return valve being closed.
- h. A makeup pump was off for 2 minutes, 57 seconds.
- i. On the next day all RCPs were staging normally.
2.0 OPERATION ACTIONS
SUMMARY
Major Operation Actions
- Verify both diesel generators have started and are loading.
- Verify SFRCS initiated and AW has started to bo th OTSG's and OTSG 1evel is being controlled at 40 irches on the startup range.
Verify a MU pump starts, establish makeup and RCP seal injec-tion.
- Assure emergency instrument air compressor starts.
- Control SG pressure with the AVV's to keep SG pressure below the MSSV setpoint.
Identifying Symptoms
- A loss of offaite power is a f ailure to provide power to the plant auxiliaries from an offsite source. It is not a failure DATE: Appendix D, Page D-8 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX wuuset NucttAt POWit GENERATCH DIVISION 7'-i 23331-00 _ 'ECHNICAL DOCilMENT to control the plant unless additional failures occur. A typ-ical P-T plot for a loss of offsite power is shown in Figure 0-10.
- Other identifying symptoms to help distinguish a LOOP from other transients are:
(~] I /
\d - RC and MFW pump trip
- Loss of voltage indication on 13.8 KV A and B buses.
- Trip of condensate and circulating water pumps
- Automatic start of diesel generators Discussion f^3 This section will discuss how actions in accordance with Part I of i i
\j ATOG will help the operator recognize a loss of offsite power and ob-tain a stable shutdown af ter it occurs.
An operator using the guidelines f, the event described would first perform the immediate actions of Part I, Section I then go to Section II, vital system status verification. The first verification which [n} \
\j
/
might require action is to verify that letdown is flowing through the block orifice only. Step 6.0 requires the operator to verify that feedwater has runback. However, both MFW pumps will trip after a LOOP and AFW will be automatically initiated. Ther e fore , the opera-tor does not need to take any remedial action, m
\
t i DATE: 7-6-82 Appendix D, Page D-0
BWNP-20007 (6-76) BABCOCK & WILCOX NUM BER NUCLEAR POWER GENERATION OlvtSION
'" * " 3 ' ~
TECHNICAL DOCUMENT The remedial actions (the initiation of AFW flow to af fected OTSG) should have been done automatically. Step 9.0 is the c ruc i al step for diagnosis of a LOOP. The station auxiliaries should be shifted from the auxiliary to the startup trans-former after every reacter trip. If voltage is not regained on all buses after this transfer, a LOOP has occurred. The remedial action to be taken for Step 9.0 are:
- a. Start or verify the auto start and loading of the diesel genera-tors (automatic);
- b. Control OTSG operate level to 40 inches on startup range
- c. Ensure at least one makeup pump starts (if no SFAS level 3 actuation) and provide seal injection to the RCS.
- d. Obtain AVV steam generator pressure control to keep SG pressure below the mainstream safety valve setpoint,
- e. Easure the service water and component cooling, water systems start.
- d. Ensure the station air system starts (if no SFAS actuation).
O 3.0 LOSS OF OFFSITE AC POWER WITH OTHER PLANT FAILURES Introduction The previous section described loss of offsite power with no coinci-dent failures. This section will show what symptoms indicate other equipment has failed and what steps the operator should take to cor-rect the heat transfer from the core to the steam generators. DATE: 7-6-82 Appendix D, Page D-10
BWNP-20007 (6-76) BABCOCK & WILCOX Nu-s te NucLt At POWER GENsdATION DIVISION 74-1125531-o0 (^')
\
TECHNICAL l DOCUMENT
\ /
In addition to a pos s ib le failure of one of the major control func-tions discussed throughout these guidelines, both diesel generators can fail to start after a LOOP. While this is not one of these con-trol functions, it does result in loss of RC Inventory and Pressure m control.
/
x t d' Branch Discussion The loss of Offsite Power Logic Diagram (Figure D-9) has separate failure branches for loss of all power except batteries, loss of reac-tor inventory control (high and low), loss of secondary inventory con-trol (high and low), loss of primary pressure control, and loss of
/ N secondary pressure control. This section will discuss each of these i ( )
l \
\_./
/
additional failure branches and illustrate how operatnr actions in accordance with procedures in Part I will restore proper control of the parameter in question. I 1 Figures D-7 and D-8 show typical primary and secondary pressure, and 1 l pressurizer and OTSG full range level trends for each of the failure branches. ( )
\m/'
l Loss of Reactor Inventory Control (Low) A loss of reactor inventory (low) exists whenever makeup or HPI is in-sufficient to overcome primary leak rate. The most probable primary leak is a failure to stop letdown and start a MU pump after a LOOP.
/
This will lead to a low RC pressure which will initiate SFAS. SFAS
)
7-6-82 Appendix D, Page D-ll DATE:
BWP-20007 (6-76) BA8 COCK & WILCOX Nuuset NUCLEAR POWER GENERATION Divls ON 74-i i 2 n 3 i-00 TECHNICAL DOCUMENT will isolate the letdown flow. The CC pumps will start automatically following a LOOP thus providing cooling to the letdown coolers so that the high temperature interlock will not close the letdown isola-tion valve. If a MU pump f ails to start, the pres sur ize r may drain even if let-down is isolated. This is especially likely if excessive AFW occurs anel overcools the primary. Also without seal injection supplied by the MU pump some seal failures are expected within 30 minutes. The re fo re, seal injection must be restored as quickly as possible. In either of the above cases (excessive letdown or inadequate make-up), the symptoms will be the same, decreasing pressurizer level and pressure with eventual draining of the pressurizer. The primary pres-sure will then quickly approach the saturation pressure of the pri-mary temperature. As long as primary temperature has been maintained below 609F (which has a saturation pressure of 1650 psig), SFAS will actuate HPI. If SFAS is actuated the AFW system will control OTSG level to 93 inches instead of 40 inch e s . The AFW flow is injected into the top (or steam space) of the OTSG's. Inis AFW flow will quench on condense the steam. As a result the OTSG steam pressure may decrease slightly. If subcooling is lost the operator must use Section III.A. of Part I to regain inventory control. DATE: 7-6-82 Appendix D, Page D-12
BWNP-20007 (6-76) BABCOCK & WILCOX ~umsta NUCllAR Powta oENEG ATION DIVIStoN 74-i t 2n3i-00 O TECHNICAL DOCUMENT ( Loss of Reactor Inventory Cont .. A loss of reactor inventory contrc! (high) can result from excessive makeup flow or HPI. These can happen by:
- 1. Pressurizer level is controlled by varying makeup using the makeup control valve; therefore, excessive makeup can occur by
[ k the failure of the makeup control valve (in an open pcsition) or failure 'of inputs into the valve (i.e., erroneous pressurizer level measurements).
- 2. If makeup flow or HPI is initiated, the operate. is required to throttle the flow when subcooling margin is regained. If he does not, excessive HPI can occur. However, the HPI pumps have A a max. pump head < 1700 psig; consequently, the HPI system cannot
~.
cause a loss of RC inventory control high when the RC system is above 1700 psig. As stated below, an operator using ATOG, Part I, Section II only, may not directly diagnose excessive ' makeup. However, a slowly increasing pressurizer level will indicate the problem. It should be emphasized that excessive makeup is an extremely slow transient and there should be ample time for the operator to recognize and correct the problem. Several indications, such as high makeup flow and pressurizer level alarms, are available to assist him. If HPI has been started due to loss of subcooling margin, Section III.A of Part I will direct .he operator to throttle the HPI once subcooling margin has been regained. 7-6-82 Appendix D, Page D-13 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCtE AR Powlt GENERATION CIVl5loN 74-1125531-00 TECHNICAL. DOCUMENT Figrue D-7 shows typical parameter trends for loss of reactor inven-tory control (high). The actions the operator can take to regain control of the reactor in-ventory are as follows:
- 1. Throttle or stops makeur using inline valves.
- 2. Increase letdown to maximum by opening both the letdown cooler isolation valve and bypass valve.
Caution: Verify cooling water flow to the letdown coolers.
- 3. Trip the makeup pumps.
Caution: The makeup pumps should only be tripped af ter all other actions have failed to stop the excessive makeup. Tripping the makeup pum;s makes RCP seal injection unavailable. Loss of Secondary Inventory Control (Low) A loss of secondary inventory control (low) exists whenever there is too little AW being injected into the steam generators to remove the decay heat output of the core. This can result from such things as:
- 1. The AW pumps have both failed to start.
- 2. AW isolation valves have f ailed to open.
The symptoms will be typical of a loss of primary to secondary heat transfer. OTSG 1evel will decrease with eventual dry out of both O I D ATli : 7-6-82 Appendix D, Page D-14
. . - . __ - , _ - -=
BWNP-20007 (6-76) i; BABCOCK & WILCOX Nu see NUCLE Ae POwta GENERADON DIVISION 7'-1125531-o0 i TECHNICAL DOCUMENT OTSG's. RC t empe rature will increase along with a rise in pressu- > rizer level and pressure. (See - Figure D-8.) Some :C flow may con-tinue even af ter the . OTSG's are dry (i.e., not removing any heat). Tne increasing hot leg temperature maintains a density difference ' between cold and hot legs. However, this is not stable natural cir- . culation nor does.this flow adequately cool the core for long periods of time, i I At Step 13.0 of Part I, Section II, the operator will see that pri-i mary to secondary heat transfer has been lost. Step 13.0 will direct him to follow Section III B for loss of heat transfer. Primary to 1 secondary cooling will be regained or the plant will be placed in ! MU/HPI cooling and brought to a stable shutdown.
- Loss of Secondary Inventory Control (High) l A loss of secondary inventory (high) exists whenever significantly
. more. auxiliary feedwater is being injected into one or both steam generators than is . required to remove the existing decay heat. The symptoms ' and corrective actions for excessive AW are discussed in Appendix A and will not be repeated here. However, it should be em-i 4 phasized that draining of the pressurizer due to overcooling may re-
-sult in void formation in the primary. This may lead to an accumu-3 lation of steam at the top of each hot leg and a stopping of natural circulation cooling. Chapter D of Part II, Volume i discusses n
motbods .of recovery from this condition. ( i 4 7-6-82 Appendix D, Page D-15 DATli : l
. _ ~ . _ - _ _ . _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
BWNP-20007 (6-76) BABCOCK & WILCOX i NuestR HUcit AR POWER GtNERAfloN DIV15toH 74-1125331-00 TECHNICAL. DOCUMENT Loss of Primary Pressure Control (Low) During a LOOP, a loss of primary pressure control (low) can be the result of: (1) Power being lost to all pressurizer heaters. (2) The pressurizer draining. The essential pressurizer heaters (252 KW) are powered after a LOOP by the diesels. These heaters are designed to compensate only for ambient heat losses. If all heaters are lost, the steam bubble in the pressurizer will cool and condense. Eventually the bubble will collapse, the pressurizer will fill and primary pressure control will be lost. An operator using Part I, Section II will not diagnose the loss of heaters until subcooling margin is lost. However, this is a very slow process and the operator should have ample time to diagnose the event using the P-T display. The required operator actions are to regain the pressurizer heaters or increase makeup to compensate for the collapsing steam bubble or cool the RCS to maintain subcooling margin. The pressurizer may drain on an overcooling event (excessive AFW or stuck open steam line valve). If it does, control of primary pres-sure will be lost. RCS pressure will then approach the saturation 9 DATE: 7-6-82 Appendix D, Page D-16
BWNP-20007 (6-76) BABCOCK & WILCOX N u nen NUCLE AR pow (R GENERAfloN OlV15 TON 74-ti2ss31-00 TECHNICAL DOCUMENT pressure.of the primary tenperature. As tong as the primary tempera-o ture is iess than 609F (saturation pressure equals 1650 psig), SFAS will actuate HPI wh ich will add water to the RCS to compensate for the RC shrinkage due to cooling. Sinca the HPI pumps have cut off head of <1700 psig the makeup pumps should be stated to increase RC
\ pressure above 1700 psig.
An operator using Part I, Section II will diagnose the overcooling at Step 14.0. Section III.C will then instruct him on how to regain sub-cooling margin (if it was lost) and how to t ake steps to stop the overcooling and terminate the transient. g v Loss of Steam Pressure Control (Low) A loss of steam pressure control exists whenever one or both steam generators undergo' a pressure reduction significantly below the safety valve reseating setpoint. This is an overcooling transient and will look similar (on the P-T curve) to the small steam line
' break transient discussed in Appendix E. There is nothing unique O
about loss of this control function coupled with a loss of offsite power; there fore , it should be identified and treated as discussed in Appendix E. Loss of All Power Except Batteries A loss of all power (or station blackout) exists wher both die.~1 gen-Thereafter, the only power ( erators fail to start after a LOOP. Appendix D, Page D-17 DATE: 7-6-82 l
-- _ _ _ _ _ _ _ _ _ _ . _ __ _ \
BWNP-20007 (6-76) BABCOCK & WILCOX Numsen j NUCLEAR PowfR GENT 4AftON Devl5loN 74-1125531-00 TECHNICAL DOCUMENT _ available to the plant auxiliaries are the two 125V DC buses MCCl and MCC2. This result is a loss of power to the following major l components.
- One AFW train (the other train has DC powered valves)
- Instrument air compressor PRZR heaters HPI pumps Component cooling water pump
- Service water pumps MU pumps This event is one of the most serious that can happen to a pressu-rized water reactor. The main feedwater pumps have tripped placing the plant into a loss of feedwater transient. The llPI pumps are not powered and thus llPI cooling also cannot be initiated. The only means of core cooling is the one AFW train with DC operated valves.
Ilowever, the other AFW train with AC operated valves can be manually operated. This situation cannot be raaint ained indefinitely because of the lack of primary inventory control and limited secondary water supply. Therefore, the operator should get at least one diesel gene-rator running so that makeup can be remotely initiated and l l controlled. Since the pressurizer heaters and makeup pumps have lost powe r , the operator has no control of primary pressure or inventory. Any voids } DATE: 7-6-82 Appendix D, Page D-18
BWNP-20007 (6-76) BA8 COCK & WILCOX Numsta NUCLEAR POWit GENERATION DIVISION 74-1 2 ss31-oo
'ECHNICAL DOCUMENT that form in the p r ima ry will tend to accumulate in the hot leg and obstruct natural circulation. Therefore, the plant may wind up in the reflux boiling mode. The operator should monitor the subcoolir.g margin and, if necessary, raise the steam generator levels to 96" on the startup range.
n\ V An operator using the guidelines for a LOOP with no diesel generato.s would first perform the inmediate actions of Part I, Section I acd verify vital system status in Section II, At Step 9.0 he will be directed to verify start of the diesel generators and of MU. These are the two most important steps he can take to regain control of the fg plant. In the long run, offsite power must be regained or a diesel ( \
\,~/ ) generator started before primary inventory control is lost.
k j\ O I h LJ' DATE: 7-6-82 Appendix D, Page D-19
s f Figure D-1 LOSS OF 0FFSITE POWER - 2/22/75 LOOP A RC TEMPERATURE VS. TIME 16 610 -
~
Reactor Trip T' Hot leg 600
- -- -- C o l a i.e g 590 -
w
~ ' .e 580 9"
570 - _ 560
/ ^ 'N ""
g N _- t 550 1.0 2.0 3.0 4.0 O Time Af ter Reactor Trip, Ilinutes i i
- \
71-1I25531-00 i
Figure D-2 LOSS OF OFFSITE POWER - 2/22/75 > LOOP B RC TEMPERATilRE VS. TIME t J 6 610 - a--e Reactor Trip 600 l Hot leg 590
--- Cold leg w
E 580 - a b. R 570 560 -
/ N -
./ %y
%__,-------,s"~~~, \
j 550 ' ' ' I 0 1.0 2,0 3.0 4.0 Time After Reactor Trip, Hinutes l I 74- - 1125531-00 ., 4
\'
Figure D-3 LOSS OF 0FFSITE POWER - 2/22/75 LOOP A PRESSURE VS. TIME f il 2200 - Reactor Trip 2150 - r 2100 -
= /
E 2050 - 5. y 2000 - 1950 - 1900 ' ' ' W 0 1.0 2.0 3.0 4.0 Time After Reactor Trip, Ilinutes 7 4 - 112 55 31 - o g t
1 l Figure D-4 LOSS OF 0FFSITE POWER - 2/22/75 OTSG FULL RANGE LEVEL VS. TIME n
~
275 -k\_ Reactor Trip 250 ,- ,,
~
225 A Loop 200 - 175 - ~~~ 8 Loop f 150 - E a 125 - 100 - 75 - N s
's 50 -
~~~ - - . _ _ _ _ _
25 ' ' ' '
?
0 1.0 2.0 3.0 4.0 Time After Reactor Trip, Minutes - 71 - 1125531-00
l Figure D-5 LOSS OF 0FFSITE POWER - 2/22/75 OTSG STEAM PRESSURE VS. TIME o 1020 - m f,,__ 1010 N
/
Reactor Trip / 1000 -
/
A / N / 990 -
/ g
~ / /
980 - N /
/ s
.' / v 970 A LOOD
- a I
= 900 -
/ ---- B Loop 5 /
y 950 -
/
E 940 - I p u I 930 - f
/
920 f 910 -- ! 900~~~J ' ' ' ' - 0 1.0 2.0 3.0 4.0 Time Af ter Reactor Trip, Minutes
/.1-1125531-00 1
)
.. _ _ _ ._. _ .__ _ _ . _ . . - . , . . . -~ _ . _ - -
Figure D-6 LOSS OF OFFSITE POWER - 2/22/75 2000 2400 -
, POST TRIP
{ 2200 -
#1N005 g ,r - -,- ,
u _ . _ 2000 - 7, SUPERHEAT ) SUBC00 LED f = REGION j 2 1800 - REGION / \ - g I I q 1600 - o _ . E , '. $ 1400 - w
;; a g 1200 ~
f STEAM PRESSURE LIMIT
~ END POINT POST TRIP WITH FORCED g
1000 --------- h--
~~ j HOT Ek0d # #
0 NATURAL CIRCULATION .TCOLD) f 800 g 8 5 $.(TH0T) NORMAL OPERATING POINT-POWER
.3 600 -
@ r- ,END POINT POST TRIP #1TH NATURAL l.jCIRCULATION(T l 400 - H0T)
SUBC00 LING NARGIN LINE I 1 f I f f 400 450 500 550 600 650 700 Reactor Coolant and Steam Outlet Temperature. F l l ) 74-1125531-00 t
r Figure D-7 TVPICAL PARAMETER TRENDS FOR LOOP 2500 - i J e
*1000 1100YINHNA E
e J s
~
/ 0 y, /C ,
j 900 -
" M 2000 -
c> 1 e i 800 I i 1 15 0 5 10 15 0 5 10 Time, Minutes Time, Minutes e, v' E
~
~
A - PRIMARY INVENTORY CONTROLLED ~ B - LCSS OF INVENTORY CONTROL (HIGH) f C - LOSS OF INVENTORY CONTROL (LOW) h 80 -
- 200 q y
e [~ < f m 60 -
- a. //C 3 E -{
= 100 - 40 _
% C" 20
." m a
i i i i 5 0 I i M 0
,0 5 10 15 E O 5 10 15 Time, Minutes Time, Minutes 7 -t - 1 1 2 5 5 3 1 - 0 0 ,
i
I Figure D-8 TYPICAL PARAMETER TRENDS FOR LOOP / LOSS OF SECONDARY INVENTORY , CONTROL (LOW) 1200[ 2500 - g PORY CYCLING E ~ e
$ 1000 =
52300 OTSG DRYOUT _ a E E 800 32l00
] "
E U U 600 ' ' ' " 1900 i i i 0 5 10 15 0 5 10 15 l Time, minutes Time, minutes 5 400 - - _- 5
- e
" ~
300 - 100 3 80 5 " I 200 60 , I a 5 . 40 3 100 [ 20 s ' ' ' o L S 0 0 5 10 15 E O 5 10 15 Time, minutes Time, minutes 74-1125531-00 ,
DOCUMENT ~
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Figure D-10 TYPICAL LOSS OF OFFSITE POWER P-T RESPONSE 2600 POST TRIP j 2400 - elN008 x '2 2f I 2200 - a [1 r-- 4- pg /> T 5 L- - J
- 4
$UPERHEAT 2000 -
J SUBC10 LED . REGION j
= 1800 -
, REGION . a 3 1600 - ,_
.r'
! 1400 - M SS SAFETY VALVE SETPOINT END POINT-POST TRIP WITH FORCEO CIRCUL AT6ON a THOT
$ 1730 - STEAM PRESSURE g
- LIMIT &ICOL O) AND FOR NATUR AL 3 1000 - H # # #
g -- g NORMAL OPERATING POINT POWER 800 - OPERAil0N fTH0T) F~' 2 SATURAil0N
' END POINT-POST TRIP WITN 600 - l t- j NATURAL CIRCUL Ail 0N (TH0T)
- SUBC00 LEO 400 MARGIN LINE O M0 450 500 52 600 6%
400 Reactor Coolant ana Steam Outlet Temperature, F Reference Time Points (Seconds) Rc.na rks 1 0 Plant separates from grid. 2 2 Reactor trips on high pressure. 2-3 2-60 Plant cools down as steam is relieved to atmosphere by steam safety valves. Diesel generators begin loading within 10 seconds after reactor trip. Table D-1 lists major components loaded on the diesel generators. 3-4 60-200 Core AT increases to develop required temperature difference for natural circulation. 4-* 200 -
- OTSG boils down to 40 inches on the startup range and levels are maintained by AFW. RCS is being cooled by natural circulation.
5 74-1125531-00
?
r Table D-1 SUMARY Of MAJOR COMP 0NENT LOADINGS ON DIESEL GENERATORS. DURING A LOOP (Sheet 1 of 2) Summary of Major Component Loadings en Diesel Generators During A t.00P A. Auto Loaded - Without SFAS Actuation
- 1. Component Cooling water pumps
- 2. Service water pumps
- 3. 480 V Unit Substations El and F1 (Emergency instrument air compressor)
B. Aute Loaded - With SFAS Actuation Nominal Size Ea. Number Loading
- HP Operated Sequence
- 1. Jockey fire pump 5 1 1
- 2. DG room ventilating ~ fans 7 1/2 2 1
- 3. ECCS sump pumps 5.9 kW 4 max 1
- 4. Essential rectifiers 14 kW 2 1
- 5. Battery chargers 86 kW 2 1
- 6. Emergency control room, auxiliary building lighting 16 kW All 1
- 7. Miscellaneous MOVS - -
All
- 8. Low-voltage switchgear room vent fan 2 1 1
- 9. Comp. cooling pug room vent 7 1/2 1 1
- 10. boric acid heat tracing 112.5 kW 1 1
- 11. Auxiliary feed pump vent van 7 1/2 1 1
- 12. Emergency ventilation system fans 15 1 1
- 13. Component cooling water pumps 400 1 1
- 14. High-pressure injection pump 600 1 2
- 15. Decay heat pump 400 1 3
- 16. Service water pump 600 1 4
- 17. Containment spray pump 200 1 5
- 18. Containment air cooler fans 150/40** 1 5
- 19. ECCS room cooler fans 7 1/2 2 5
- 20. Service water pump strainer 1 1 1
- 21. Control room emergency system condensing unit 10 1 1
- 22. Battery room vent fan 1 1 1
- 23. Emergency diesel generator immersion heater 15 kW 1 1
- 24. Emergency DG soak-back pump 1 1 1
- 25. Diesel oil pump house lighting 9 kW 1 1
- 26. Diesel oil pump house unit heater 4 kW 1 1
- 27. Boric acid tank heaters 7 1/2 2 1
- 28. Boric acid tank room space heaters 40.5 kW 1 1
- 29. Diesel oil pump house sump pumps 1/2 1 1
- 30. Makeup pumps main oil pumps 1/2 1 1
- 31. Makeup pumps auxiliary gear lub.
oil pumps 1.0 1 1 7 4 - 11 e 3 5 31 - 0 0
?
Table D-1 SUM 4ARY OF MAJOR COMPONENT , LOADINGS ON DIESEL GENERATOP.S DURINGALOOP(Sheet 2of2) C. Components which can be manually loaded or automatically loaded after automatic sequencing is co;.pleted. J hominal Size Ea. Number HP Operated
- 1. Control room emergency ventilation system fan 5 1
- 2. Turbine bearing oil lift pump 5 6
- 3. Turning gear 60 1
- 4. Containment recirculation fans 50 1
- 5. Hydrogen dilution blower 20 1
- 6. Control room emergency system standby condensing unit fan 7 1/2 1
- 7. Diesel oil transfer pump 3 1
- 8. Emergency diesel generator air conpressor. 7 1/2 1
- 9. Control room emergency vent. system vacuum pump 1.5 1
- 10. Containment vessel vent. system vacuum pump 1.5 1
- 11. Emergency instrument air dryer 150 VA
- 12. Turbine turning gear oil pump 30 1
- 13. Emergency instrument air compressor 40 1 The diesel gen 2rator is loaded in five sequential steps, beginning with the closing of the diesel breaker after the machine has reached rated voltage and frequency. The fifth and last step will be initiated approximately 30 seconds after the engine start signal is received.
Two-speed motors, 40 hp, during emergency diesel generator operation following LOCA-i 74--II25531-00
BWNP-20007 (6-76) 9ASCOCK & WILCOX uunu ~ NUCLfAa POwte otNesAllON OtVIslON 7'-t i 2553 i-oo TECHNICAL DOCUMENT APPENDIX E SMALL STEAM LEAK i.o GENERAL TRANSIENT DESCRIPTION i A small steam leak is a failure to control secondary pressure. It is an overcooling transient that results from too much prinary to 4 secondary heat transfer. A small steam leak is defined as loss of steam up to about 25% of full flow. The largest leak assumed is that due to equipment failure (failed open turbine bypass valves). Larger steam le ak s could exist only with piping failures; these are not as likely as component fail-
.ures. However, the basic principles and sequence of events to be dis-cussed also apply to larger breaks.
~ The steau leak causes a power mismatch between the heat source (the core) and the heat sink (the steam generators) and will cool the , reactor coolant down. The ICS will pull rods to increase core power and maincain T ave. If initially the plant is ope rat ing at much less tha.s full power , the plant will stabilize with reactor power greater than megawatt demand by an amount equal to the size of the steam leak. Depending on the leak location and size the operator may be able to a locate and isolate the leak without tripping. For example, if the leak results from a stuck open TBV valve, the mass flow through the leak goes to the condenser and the plant could operate indefinitely. A The only indication will be the reactor-turbine power n.ismatch. An ; l l DATE: 6-82 Appendix E Page E-1 ,
~ - -- -
~ . . . _ . . . . _ _ _ _ _ _ . __ , _,,__ _ _ _ ,__,
BWNP-20007 (6-76) BABCOCK & WILCOX Nu=sER MJCLEAR POWER GENERATON clVISION 74- t i2 s s 3 t-oo TECHNICAL DOCUMENT atmospheric leak will require makeup to prevent the hotwell from drain-ing. Hotwell level and conde ns ate level as well as visual inspection will indicate the leak. If, however, the leak is inside containment, an SFAS signal on high RC pressure may occur. Small steam leaks without a reactor trip do not result in rapid over-cooling and the reactor coolant temperature drop will not be severe. If no additional failures occur and the leak is isolated by the MSIV's, the plant may recover from a leak as large as 25% automat-ically without losing the subcooling margin. HPI will be actuated by SFAS and (combined with MU and pressurizer heaters) will recover RCS pressure, although tempo rary drainage of the pressurizer may occur. SFRCS actuates on low steam generator pr e s s ur e . SFRCS actuation will isolate the MSIV's, AVV's and MFW and will start AFW. The leak will be isolated if downstream of MSIV's. After the leak is isolated by SFRCS, the operator would have to take menual control of HPI, MU and the AVV's to limit the presurizer refill by limiting the reactor coolant water addition and swell while maintaining RC subcooling ( margin. i l l The following transient example is based on a leak equivalent to 15% l full flow and occurs with maximum decay heat. A larger leak or lower l decay heat levels would result in a faster cooldown rate and could 1 cause RCS saturation. As an example, refer to Figure 16 in Part II, Volume 1, showing the expected response to a larger steam leak. DATE: 7-6-82 Appendix E, Page E-2 l l
BWNP-20007 (6-76) SABCOCK & WILCOX NUCLEAR POwta GENERAflON DIYl5aON 74- u 2 s s 31-00 3 TECHNICAL 00CUMENT
! If the leak is upstream of the MISIV's, SFRCS actuation will not iso-late the leak from the broken SG but will initiate AFW flow to the l good steam generator. SFRCS will also isolate MFW to the broken SG and allow it to boil dry. Other failures may occur which will require operator action. These are discussed in Section 3.0 of this appendix.
! The operator should become familiar Figure 22 " Overcooling Diagnosis i Chart", in Part II, " Diagnosis and Mitigation". The P-T curve and sequence of events shown in Figure E-1 depict a typ-ical small steam leak transient (stuck open TBV's) that is terminated automatically. The transient shown is also applicable if terminated by the operator. I l v 2.0 OPERATOR ACTIONS
SUMMARY
l Immediate Actions l
- If the overcooling SG is apparent close the MSIV's, AVV and control MFW of the overcooling SG. If leak continues actuate SFRCS.
- Start additional MU if pressurizer level is less than 100" and RCS O\ pressure is decreasing.
- If SFRCS actuates and both generators repressurize, feed both generators; if only one repressurizes SFRCS did not isolate the leak. SFRCS should stop AFW to the generator that did not repres-surize and allow it to boil dry.
Verify HPI actuation if low RC pressure SFAS actuates. v - Follow remainder of Part I, Section III.C. DATE: Appendix E, Page E-3 7-6-82
. . . - . _ _ - . _ _ - - . _ . _ _ . _ . _ _ _ _ . _ _ . . , . _ . ~ _ . . _ _ _ _ _ . _ , _ _ _ _ _ _ _ _ _ . _ . - _ _ _ _ _ - . _ . - _ .
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sen NUCLEA9 POWER GENftATON OlVl510N 7'- t i 2 s s 3 i-00 TECHNICAL DOCUMENT identifying Symptoms Smal t s tean le ak is an overcooling transient as shown in Figure E-2: Other key parameters to distinguish small steam leaks f rom overcooling transients due to excessive feedwater are: Decreasing level and pressure in one or both SG's without feed-water addition * (steam pressure will be lower than the 960 psig steam pressure limit) Near normal or low SG levels with feedwater addition
- NOTE: Excess MFW can cause a slightly reduced steam pressure in the unaffected SG; excess AFW can cause a lerge reduction of steam pressure in the affected SG; however, in both cases the SG with reduced pressure should have steady or increasing level indicating that inventory is not being lost.
This section will present an example of a small steam leak and discuss how operator actions in accordance with Part I will term ina te the transient and provide recovery to stable shutdown conditions. The assumed transient will be the same shown in the previous section, i.e., failed open TBV's from 100% full power. l l A failure of all TBV's in the open pos it ion results in ai: proximately 25% excess steam flow. The reactor will trip and high flux within about eighteen seconds, the turbine will trip, and the ICS will run O DATE: 7-6-82 Appendix E, Page E-4
BWNP-20007 (6-76) SABCOCK & WhLCOX " uucteAs powen GENeaAh0N DM560N 74-1125531-00 TECHICAL DOCUMENT {' back MFW. After the operator performs the immediate actions in Part I, Section I, he .till verify vital system status in accordance with I Section II. Steam pressure in both steam gene ra to rs will decrease I fairly quickly (70 seconds) below the normal post-trip pressure boun-dary of 960 psig. In Step 14.0 of Section II the operator should note that primary to secondary heat trans fer is exce s s ive by observing lower than normal steam pressures and the overcooling trend indicated by Th re s po ns e on the P-T curve. (Refer to the P-T curve, Figure E-2.) The procedure directs the operator to Section III.C. Step 1.0 in III.C requires the orerator to increase MU if pressurizer level goes below 100 inches. For this transient pressurizer level s will go below 100 inches (see Figure E-3, " Time Relationship of Key Parametere") after trip, thus the operator will initiate full MU flow.
-If low RC pressure SFAS actuates full HPI flow will more than compensate for the reactor coolant contraction rate for this size leak; t he refo re pressurizer refill and RCS repressurization by the MU and HPI systems may begin to occur even be fore the steam leak is isolated. However, a temporary loss of indicated pressurizer level
! , may be expected. Subsequent steps will effectively isolate the steam generators from 4 most failures that would cause overcooling. If the overcooling SG is - not apparent, the operation will actuate SFRCS. Suc c es s ful completion
)
v DATE: 7-6-82 Appendix E, Page E-5
BWNP .?0007 (6-76) 8ASCOCK & WILCOX NUM8tR NUCLEAR POWER GENERATION DIVISION n- t i2 m i-oo TECHNICAL. DOCUMENT of this step will terminate the overcooling transient in thi case by isolating the leak. Since the TBV's are failed open on both SG's he would have to actuate SFRCS to close both MSIV's to isolate the leak. This action will initiate AFW to both SG's. If this action is pe r fo rmed be fore the subcooling margin is lost, the RC pumps will remain on and AFW will be establishing the low level se t po int . However, if the subcooling margin is lost it will quickly be regained once the stean leak is isolated. Once the steam leak and overcooling has been terminated, the oeprator must perform a few actions to regain RCS inventory and pressure con-trol. At this stage in the transient he still has MU and/or HPI flow and the RCS is slowly reheating and will continue to reheat until steam pressure increases to the MSSV setooint. The actions he should take are given in the procedure and are as follows:
- Maintain RCS temperature at the present value by adjusting the AVV's on both SG's. Closure of the MSIV's has isolated the TBV's.
1 1 Since the MU and HPI flow have increased the RCS inventory, allow-1 ing the RCS to reheat to norme.1 post trip conditions would result in a much highe r pressurizer level and possibly a full pres surize r.
- Throttle HPI when the subcooling margin is restored by using one HPI pump and one injection line ( pr e ferably the normal makeup I
O DATE: 7-6-82 Appendix E, Page E-6
BWNP-20007 (6-76) BABCOCK &- WILCOX "# " " Nucean powen GeNeBAHON Om$ ION v.- t i 2 n 31-00 TECHillCAL DOCUMENT nozzle with the thermal sleeve). In this particular transient, the subcooling n.argin was not lost and RCS pressure began to in-crease be fore the leak was isolated. The re fo re , the operator . could have throttled HPI as soon as RCS pressure began to increase.
- When pres sur ize r level returns on-scale low and is 100" and in-creasing (and the RCS is above the subcooling margin) the operator i
should terminate HPI and realign for normal makeup /le tdown i operation. The transient of most concern for the operator is one where successful operation of automatic plant features will not isolate the leak. That j transient is an unisolatable leak in one SG upstream of the MSIV. HPI actuation by SFAS would still occur and SFRCS actuation would ensure the leak is isolated from the good steam generator and would isolate MFW to the '? bad" SG. SFRCS would initiate AFW flow only to the " good" SG. Blowdown of the " bad" SG would continue the overcooling until the SG became dry. Stable shutdown conditions would be returned by con- , trolling DH removal with the good SG and AFW. s. 3.0 SMALL STEAM LEAK WITH OTHER PLANT FAILURES Introduction The previous seccion described small steam leaks in general, but did not discuss other failures that might also happen at the same time. This section will show what symptoms to look for when other equipment g DATE: 7-6-82 Appendix E, Page E-7
BWNP-20007 (6-76) BABCOCK & WILCOX NumsER NUCLE AR POWER GENERATION DIVIStoN 74- 2ss3i-00 TECHNICAL DOCUMENT fails and will show what steps the operator should take to restore the heat transfer from the core to the steam generators. Figure E-4, "Small Steam Leak Logic Di agr am , condenses the event tree into the main transient path and branches fo r the major failures that can occur in addition to the main transient. The P-T diagrams are shown in Figure E-4. The effects on other parameters is shown in Figure E-3, " Time Relationship of Key Parameters". The event that was chosen fo r simulation starts with the reactor it power. A small steam le ak occurs that may or may not result in a reactor trip on high flux. If a trip daes not occur, the plant will stabilize with a power mismatch with the possible conditions given in Note 1 on Figure E-4. The main path assumes a reactor trip does oc-cur. All the data that is shown starts from the time of reactor trip. The failures are shown to occur when a component has to operate (for example, efter a reactor trip the ICS is shown to automatically run i i back MFW; the failure shown is that MFW does not run back). It must be remembered that all steam leaks will not look exactly like O l the examples used. The primary purpose of these examples is to pro-l mote operator familiarity with the symptoms and effects of steam le ak s . DATE: 7-6-82 Appendix E, Page E-8
0 BWNP-20007 (6-76) BABCOCK & WILCOX wumeen
- ' NuoeAn Powea seNenATION Olvi$lON 74-1125531-00 gTECHNICAL DOCUMENT J Branch Discussion J
Figure E-4 has separate failure branches for loss of reactor inventory
- control (high and low), loss of secondary inventory control (high and low), and loss of secondary pressure control. (Minor failures, such
- A as toas of pressurizer heaters, have relatively little impact on the d overall transient. The operator has time to diagnose and correct such failures; the re fore they are not shown.) Another valuable tool to facilitate operator recognition and identification of overcooling transients is Figure 22, " Overcooling Diagnosis Chart", in Part II, Section III.C., " Diagnosis and Mitigation". The operator should become familiar with this chart.
s Figure E-3 shows key distinguishing parameters for small steam leaks that have a time dependency important to the operator in identifying both the type and severity of transient. The parameter plots show typical responses to a steam leak equivalent to 25% full flow with automatic HPI and SFRCS actuations. Arrows, where used, show the i e f fect of other failures and of operator actions on the time rela-tionship. Loss of Reactor Inventory Control (High) A loss of reactor inventory control (high) exists whenever makeup flow is exces s ive causing the pressurizer to fill and overpressurizing the RCS for the existing plant conditions. Steam leaks ranging in size up
~
to approxima tely 25% of full flow, assuming maximum decay heat, i I DATE: 7-6-82 Appendix E, Page E-9
BWidP-20007 (6-76) SABCOCK & WILCOX Nuuset NUCLEAR POWit GENERATION O4 VISION 7'- n 2 s s 31-00 TECHNICAL DOCUMENT probably will not cause RCS contraction rates greater than the full capacity of the HPI system af ter pressurizer empties. Therefore, pres-surizer refill and RCS repressurization by the MU and HPI should begin to occur even before the steam leak is isolated (refer to Figure E-1 and sequence of events in Section 1.0 of this appendix). Larger steam leaks or lower decay heat levels smuld result in a faster cooldown rate and could cause RCS saturation even with full MU and HPI flow (refer to Figure 16 in Part II, Volume 1, to see the expected response to a larger steam leak). Full HPI flow without MU will not result in a rapid increase in RCS pressure because of the relatively low head HPI pumps. However, operator response will be required to prevent a solid pressurizer and RCS overpressurization because of RCS reheat or MU control failure. Loss of Reactor Inventory Control (Low) l l A loss of reactor inventory control (low) exists whenever makeup or l HPI flow is insu f ficient to overcome the primary leak rate or (as in this case) the coolant contraction rate, resultin; in drainage of the pres sur ize r. Too little makeup is not a major concern for th is particular trans-ient. If the overcooling is terminated be fo re the pressurizer empties, the RCS will reheat and the resultant swell will restore pressurizer level. If the overcooling continues, SFAS will actuate and HPI will automatically intiate. DATE: 7-6-82 Appendix E, Page E-10
BWNP-20007 (6-76) l BASCOCK & WILCOX ,, n , NUCLEAR POwte GENERATION DIVl540N 74-1 2 n 3i-oo TECHNICAL DOCUMENT
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' Following the actions specified in III.A of Part I will restore primary system inventory control and subcooled margin. 1 i Loss of Secondary Inventory Control (High) ) L A loss of secondary inventory control (high) exists whenever signifi-l i ( cantly more feedwater (main or auxiliary) is being injected into one I or both steam generators than is required by existing plant condi-i tions. It is an overcooling transient and re s ponse on the P-T curve I is very similar to that of a small steam leak. Excessive main feed-l ( water can be distinguished from small stean leaks by high MFW flow- l rates and high SG 1evels. Large feedwater mismatches can cause more severe ef fects on the RCS than small steam leaks and must be corrected j much faster. A detailed discussion of excessive main feedwater is contained in Appendix A. 1 Excessive auxiliary feedwater can cause depressurization of the af-fected SG to a much larger extent than excessive main feedwater. This is due primarily to the condensing action introduced by spraying AFW in near the top of the tube bundle (into the steam space) and due to the lower AFW temperatures. Should the FW ficw be excessive, the ope rato r should recognize the overcooling as well as high FW flow. Following the actions in Part I, Section III.C. for excessive primary to secondary heat t rans fe r will terminate the runaway FW. Step 6.0 of III.C requires the operator to trip.the MFW pumps, i DATE: 6-82 Appendix E. Page E-11
. _ _ - _ . . _ - ~ _ . _ _ , _ . ._ .__ _ _ . _ _ _ _ _ _ _ . - _ . . _ . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___
BWNP-20007 (6-76) BABCOCK & WILCOX " III NUCLEAR POWER GENERATION DIVl510N 74- 12ss31-oO TECHNICAL DOCUMENT Figure E-3 shows the impact of small steam leak overcooling campounded by overcooling due to excessive feedwater. The curves for RCS pres-sure and pressurizer level will shif t to the le f t , i.e., pressure re-duc t ion and drainage of the pressurizer will occur faster. Exces s ive FW will also cause the steam pressure curve to shift to the left. O Loss of Sec.ondary Inventory Control (Low) A loss of secondary inventory control (low) exists whenever too little f eedwa te r is be ing eupplied to the steam generators resulting in too little primary to secondary heat trans fer and overheating of the RCS. The operator should recognize this condition by low feedwater flow rates and low SG 1evels. to isolate O The operator may be required , by the procedures in Part I, all feedwate r in order to determine which SG is leaking. However, he will also, by procedure, immediately restore feedwater to the good SG as soon as the broken SG is identified and isolated. A detailed discussion concerning too little primary to secondary heat transfer is provided in Appendix B, " Loss of Main Feedwater". Total Loss of Steam Pressure Control A loss of steam pressure control exists whenever an unisolable steam leak exists in both steam generators. If this condition exists the operator should perform the following: 1 DATE: Appendix E, Page E-12 7-6-82
BWNP-20007 (6-76) SABCOCK & WILCOX womsen NUCLEAR POWtt GENERADON DIVISION 7'-1125531-00 TECNNICAL DOCUMENT 4
- 1. Maintain primary to secondary heat transfer by supplying feedwater to both SG's at a very limited rate while attempting to stop the steam leak on at least one SG (e.g., manually isolate stuck open
' AVV , gag shut MSSV, etc.). If heat removal is exces s ive with two SG's or steam flow needs to be stopped from a SG (dite to location of stean leak, SGTR, to aid repairs, etc.) then,
- 2. I solate FW to one SG and allow it to boil dry while continuing res trict ive feed to the other SG and attempting to maintain level and controlled DM removal. Continue trying to stop leak on at least one SG. If controlled DH removal with one SG is not pos-sible, then,
- 3. Initiate MU/HPI cooling, allow both SG's to boil dry.
Continually try to repair steam leak of at least one SG. When the 4. on one SG then use the repaired SG to stean leak is repaired restore SG DH removal, I i k l DATE: Appendix E, Page E-13 l 7-6-82 f - - __ . _ _ _
L 1 Figure E-1 SMALL STEAM LEAK r 2500 2400 - POST TRIP 5lN005
, 2200 - ; a r-- ri
- if L__.
g 2000 - SUBC00 LED 2 a , REGLON SUPE RHE AT f 1800 - REGION Iis00 -
=
/'
; i Ag -
2 1400 - L /
~5 [
2 ENO POINT POST TRIP WITH 1200 - STEAu PRESSURE h FORCEO CIRCULAfl0N HOT (T E22 &T ) AND FOR NATURAL g 1000 ; _ ,,___t___ ___ CIR Ut TION (TCOLO) 800 - POWER OPERATION (T g gy) E 5 i j SATURAfl0N r- - 1 ENO POINT POST TRIP WITH 600 t _] NATURAL CIRCULATION (THOT) SU8C9 LED 400 - MARDIN LINE O I I I I I 400 450 500 550 600 650 700 Reactor Coolant and Steam Outlet Temperature, F Reference Time Points (Seconds) Remarks 1 0 All TBV's fail open from 100% full power. 2 14 Reactor trip on high flux, turbine trip. ICS runs tack MFW, makeup valve goes wide open.
; 60 TBV's from only one SG close.
4 100 Pressurizer drained; SFAS trip on low pressure. Starts HPI. 5 s300 SFRCS actuates on low SG pressure isolating steam leak. MFA and initiating AFJ. 74 I i *. r,r,3 1 - 0 0
1 Figure E-2 TYPICAL SMALL STEAM LEAK TRANSIEN" 2600 2400 ~ POST TRIP WINDOW
- 2200 -
A r-- , e 0 u___ t 2000 SUBC00 LED 3 - REGION . SUPERHEAT S REGION 1800 - C 3 1600 - e - O y g 1400 - O 1200 - 3 STEANPRESSyRE END POINT-POST TRIP WITH FORCED E 1000 - LIMIT _\ __ _ _ , _ _ CIRCULAil0N (TH0T &TCOLD) AND g FOR NATURAL CIRCULATION (TCOLO) U 800 - 3 NORMAL OPERATING POINT-POWER 600 - OPERATION (TH0T)
'r- i END POINT-POST TRIP WITH NATURAL 400 -
SUBC00 LED l j CIRCULATION (TH0T) NARGIN LINE 1 I i i I 400 450 500 550 600 650 700 l Reactor Coolant And Steam Outlet Temperature-F 741-1I"5531-00 l l
7 Figure E-3 TIME RELATIONSHIP OF KEY ' PARAMETERS 200 2501 -
=
En$ [ 120 i2000 -
?
o N 2 40 g 1500 h EXCESS 5 EXCESS NFW OR 5 NF'l OR 1 I i DRAINED 0 1000 0 2 4 6 2 4 0 i 0 Time, minutes Time, minutes 1100 REHEAT TO NSSV SETPOINT
. I WITHOUT MANUAL AVV g 900 -
CONTROLg l ! .i 5 EXCESS - l AFW l E MANUAL AVV CONTROL 10 g 700 - PREVENT RCS REHEAT SFRCS ISOLATES LEAK t i I 500 2 4 6 O I. Time, minutes 1 [ l 1 74-1125531-00
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d BWNP-20007 (6-76) SABCOCK & WILCOX NUM4it NUCLEAR P0wte OtNERATION OtYlStON 74- 12ss31-00 TTECHNICAL DOCUMENT . t APPENDIX F I LOSS OF COOLA!!T ACCIDENTS
1.0 INTRODUCTION
A loss of coolant accident (LOCA) is defined as any event during which reactor coolant, escapes due to loss of integrity of the primary cool-ant system. A LOCA can occur at any time (power operation, cooldown, ' heatup, cold shutdown) and can be either an initiating event or a i re sul t of another accident. Usually LOCA's release reactor coolant to i the contaminant environment; however, there are some LOCA's which can 4 release reactor coolant to the secondary plant (steam generator tube l leaks) or to the auxiliary building (leaks in the MU, letdown, or decay heat removal systems).
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LOCA's directly attack one of the five fundamental principles of reac-tor operation: "RCS Inventory Control". The loss of reactor coolant I reduces the plant's ability to transport the core's heat to the steam generators. If the lost reactor coolant is not replaced, core cooling will be lost and fuel damage may occur. The Emergency Core Cooling System (ECCS) consisting of the High Presure Injection (HPI), Low
\ Pressure Injection (LPI), and Core Flooding Tanks are used to supply emergency core cooling water to prevent fuel damage. The auxiliary feedwater system is also used for core heat removal and RCS depressurization for some sizes of small break LOCA's.
l DATE: 7-6-82 Appendix F, Pgae F-1
I BWEP-20007 (6-76) BABCOCK & WILCOX " ' NUCLtAR POWit GENERAT60N DIVl5lON 74-1125531-00 TECHNICAL DOCUMENT The Makeup pumps assist in adding RCS inventory for small breaks where the RC pressure remains above the shut of f head of the liPI pumps. A LOCA is a unique accident. Some of the major factors which " single out" the LOCA from other abnormal transients are as follows: A. A wide range of leak sizes is possibig: O The rate at which reactor coolant is lost from the primary system depends on the break size through which the reactor coolant can
,' ass. LOCA's can range from small leak rates (leaking pressurizer relief or safety valves, leaking RC pump seals, cracks or breaks in instrumentation lines connected to the primary system) to large losses of reactor coolant due to cracks in or a severance of the RCS piping or its attachments (letdown line, HPI lines, decay heat drop line).
P, - Abnormal system conditions are a natural consequence of the event: When reactor coolant is lost at a rate greater than it is re-placed, the reactor coolant will become saturated, a loss of nor-mal natural circulation (assuming the RCP's are tripped) is inevi-table and boiler-condensing may be the only means by which the steam generators can remove energy from the reactor coolant. For many LOCA's these conditions are temporary and may be prevented by operator action to isolate the leak. For a LOCA which cannot be isolated, the evolution of the transient to abnormal conditions is unavoidable. DATE 7-6-82 Appendix F Page F-2
BWNP-20007 (6-76) CASCOCK & WILCOX PApCLEAR POWtt GENERAftON Olvi$lON i 74-1 2ss31-00 p TECHNICAL DOCUMENT
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C. Steam generator heat removal may be degraded: i The transport of core heat to the steam generators may be severely reduced depending upon the size of the LOCA. Steam generator heat removal is needed for some small breaks because the break alone < does not remove enough core heat. If the break is large enough the steam generator is not needed as a heat sink. In some large breaks the RCS depressurizes below steam pressure, and the steam f gene rator can even be a heat source if the temperature of the secondary inventory is higher than that of the reactor coolant. D. Maintenance of hot standby is not a safe end condition in most cases: Some LOC A' s can be isolated and the loss of reactor coolant stop-
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ped. Hot shutdown is a safe condition. However, when isolation is not possible the loss of reactor coolant cannot be stopped. Then the RCS must be depressurized to reduce the leak rete, to in-crease the ECCS flowrate, to engage the LPI system and to estab-f lish long term cooling. It is also desirable to depressurize be-fore the BWST is empty to avoid HPI piggyback operation. Depressurization is also desirable to engage the LPI system so the stean generators do not have to be used with service water as feed-water when the condensate storage tank empties. Depressurization \ is required for the tube leaks to limit the of fsite doses and to V DATE: 7-6-82 Appendix F, Page F-3
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCLEAR POWER GENERATION DivlSION 74- t i 2 n 31-00 TECHNICAL DOCUMENT avo id emptying the BWST into the secondary system (tube leakage is not re tur ned to the contaiment sump and recirculation is not pos s ib le ) . E. The containment environment can degrade: If the break is inside the containment, reactor coolant water and stem will collect in the reactor building. The pressure, t empe ra-titre and radiation levels in the reactor building will increase, and use of emergency building cooling systems may be necessary. In additien to maintenance of core cooling, the operator must also control the reactor building environment to prevent or limit off-site radiation releases. Control of the environment is also de-sired to prevent equipment damage. Containment isolation is re-quired to limit the offsite doses. LOC A's can be complicated events because of the wide range of sys-tem conditions that can happen. The plant's safety systems, espec-ially the ECCS, are capable of safely limitind the LOCA effects. Core cooling can be maintained through use of the ECCS, but opera-tor actions and other equipment can be used to enhance the perform-ance of the ECCS and to achieve better results. The rema in ing portions of this appendix will describe the response of the RCS for the full spectrum of LOCA's, the general approach to accident mitigation and establishment of long term cooling, how the ECCS and other equipment are best used, and also outline con-trol limits which apply to plant re c ove ry . DATE: 7 -t. -82 Appendix F, Page F-4
BWNP-20007 (6-76) BABCOCK & WILCOX NUCLEAR POWER GENERATION DIVISION 74-1125531-00 ( s ) TECHNICAL DOCUMENT 2.0 LOCa COCNEPTS Lost, of coolant accidents are classified in three categories:
- 1. Large Break
- 2. Small Break
/\ ; 3. Small Leak L./ These categories are used because the response of the RCS is di f fer-ent. Each category will be discussed separately. Large Breaks A large break is a major failure of the primary system pressure boun-
/ dary which depressurizes the reactor coolant system rapidly and almost \ /
b' completely. This is a designer's event that es t ab lishes the s iz e of the core flood tanks, the size of the LPI flow rate and the size of the containment cooling systems (sprays and coolers). A large break LOCA has never occurred at a commercial nuclear power plant. As a rule of thumb, a LOCA with a break area greater than that of a
/N l ') 10-inch diameter hole is a large break. For breaks of this size or \ /
xj larger, the plant transient can be broken down into three distinct phases: blowdown, refill, and reflood. Blowdown: Blowdown is a rapid depressurization of the primary system as reactor coolant escapes in large quantities. Large 3 N4 r pressure gradients are created in the primary system and (v-) 7-6-82 Appendix F, Page F-5 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX Numsen NUCLE AR Powet GENERATION D;V5 ton 74-1125s31-00 TECHNICAL DOCUMENT _ low pressure :wnes are created at or near the break. These pressure gradients disrupt the normal flow patterns within the primary system, and the reactor coolant will be drawn to the break (low pressure zone) from alI regions of the primary system. The reactor coolant will saturate and ch ange to steam due to system depressurization and core heat removal. Blowdown ends wh en the primary system and the containment pressures are the same; the loss of reactor coolant stops. From start to finish, the blowdown period can be from 20 to 200 seconds depending upon the break size and location. At the end of blowdown, the reactor coolant system will contain very little water and will be mostly steam. During blowdown, RC pressure drops rapidly and falls below the secondary side pressure within a few seconds. The steam generators become a heat source as opposed to a heat sink because the tempersture of the reactor coolant will be less than the secondary feedwater temperature. The steam generators remove heat only as long as the primary temperature is above the secondary temperature. Steam pressure will hang up initially and then slowly decrease as energy from the secondary is slowly trans fe rred to the primary system. 7-6-82 Appendix F, Page F-6 DATE. l l
BWNP-20007 (6-76) SABCOCK & WILCox NUMBEP Nuct!AR Powet GENERATION OlVi$10N 125531-00 p TECHNICAL DOCUMENT 7'-
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Blowdown causes a reactor trip on low RC pressure, and the fission process is shutdown to the decay heat level. But, because large quantities of steam are produced, the core will heat up; the fuel and cladding temperature will rise until cooling systems recover the core with water. During blowdown the HPI and LPI systems are actuated, and the core flood tanks discharge when the primary system pressure drops below 600 psig. These emergency core cool-4 ing systems have little or no ir. pact on the blowdown phase of a large breale. The blowdown transient is too fast, and the cooling water that does enter the primary system is
" swept out" the break along with the reactor coolant. The i
k ECCS ' systems restore core cooling during the second and l third phase of a large break. The blowdown also has a severe ef fect on the reactor build-ing environment. Rapid increase in the contaimnent pres-sure, temperature, and radiation levels occur. The emer-
% gency coolers and sprays are started by the SFAS on high containment pressure and water will accumulate in the reac- f l
tor building emergency sump. The reactor building becomes l the heat sink for the energy (primarily RC enthalpy) origi- . nally contained in the primary systen. J k 1 l l l DATE** 7-6-82 Appendix F, Page F-7 ' _ . ~ ._ _ , . __._______ _._ ____._._______ __
BWNP-20007 (6-76) BABCOCK & WILCOX NumsEn NUCLEA9 POWER GENERATION DIVISION 74-1125531-00 TECHNICAL DOCUMENT Figure F-1 shows the primary system response during the blowdown transient of a large break. Refill: At the end of blowdown the primary system is at a low pressure, and little or no water is left in it (see end condition in Figure F-1). However, the CFT's, the LPI and the HPI will be adding water to the system. The emergency inject ion water eatering tha system accumulates and begins to refill the reactor vessel. Re-fill is defined as the time required to refill the reactor vessel lower head up to bottom of the fuel (see end of refill in Figure F-1). This is a sho rt time (10-12 seconds) and for smaller breaks a refill period may not occur if enough water remains in the reac-tor vessel. The refill phase of a large break cannot be detected by any plant instrumentation. As shown in Figure F-1, no water exists in the core during refill and the core will heat up until ECCS water en-ters the core. When the core is uncovered the hot zircaloy clad chemically reacts with the oxygen of the steam to fo rra zire oxide and release hydrogen. Hydrogen may accumulate in the loop or it l l may be vented through the break or the high point vents to the l containment. l l Reflood: The last phase of a large break is called reflood. During re-flood, water from the ECCS refloods the RV up to the elevation of 7-6-82 Appendix F, Page F-8 DATE: i
.___....____m t
BWNP-20007 (6-76) SABCOCK & WILCOX NUCLEAR POwta OtNORATION OtvlSION 74-" 2 n 31-oo TECHillCAL DOCUMEllT the inlet nozzles and refills the system up to the elevation of the break. This is a slow process because a lot of the emergency cooling water is boiled of f as it enters the core since decay heat 2 levels are s till high. l
- Within 5 to 10 minutes the core will'be recovered with water and ,
the fuel clad temperatures reduced to within a few degrees of the
+
i saturation temperature (the RCS pressure will be approximately equal to the RB pressure and the core exit thermocouples should i i show approximately saturated temperature for these conditions), i i Core cooling thereafter is maintained by continuously adding injection water from the low pressure injection system so that the i core remains covered by water. The CFT's will empty, and the HPI system can be secured (see Gu'delines for HPI termination in "Best Methods for Equipment Operation"). Figure F-1 shows the reflood phase. Two examples are given: one for a large break at a high location in the system and one for a
~.
t 4 g large break low in the system. ,*
- During the reflood phase, additicnal steam is created by
l a) boiling of emergency injection water in the core. b) boiling of emergency injection water due to heat from the hot metal components, and l i DATE: Appendix F, Page F-9 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusen NUCLEAR POWtd GENERATCN DIVISION 74-1125531-00 TECHNICAL. DOCUMENT c) boiling of emergency injection due to heat transferred from the secondary side of the steam ge r.e ra t o rs (this is a source of additional steam only if primary side water re-enters the tube region of the steam generators). This steam is vented out the break to the reactor building. The - reactor building pressure and temperature conditions may rise dur-ing the early stages of reflood. When decay heat and the heat from the hot metal and steam generators drop, the steam telease rate will decrease and the reactor building pressure and tempera-ture will decrease as the sprays and coolers can remove more heat than is being released. Figures F-2 and F-3 give a summary of the expected response (P-T dia-gram) of the RCS during large hot and cold leg b re ak s . The diagrams also outline the sequence of events. At the end of blowdown, the reac-tor coolant is in a saturcted condition a: or nest the containment pressure. Thereafter, the core outlet thermocouples should be used to check the core outlet temperature for saturation. At the end of the blowdown the core outlet thermocouples may indicate superheated condi-tions since the core may be completely void of water. As the ECCS re-floods the reactor vessel, the fuel rods vill be quenched and the core outlet temperature should return to saturation. O 7-6-82 APPendix F, Page F-10 DATE:
l BWNP-20007 (6-76) BABCOCK & WILCOX ** i wuonna rewee oewenArion oivision 7e n 2n31-00 TECHNICAL DOCUMENT the inlet nozzles and refills the system up to the elevation of the break. This is a slow process becaase a lot of the emergency t cooling water is boiled of f as it enters the core since decay heat levels are s till high. Within 5 to 10 siinutes the core will be recovered with water and the fuel clad temteratures reduced to within a few degrees of the saturation temperature (the RCS pressure will be approximately equal to the RB pressure and the core exit thermocouples should show approximately saturated temperature for these conditions). 2 Core cooling thereafter is maintained by continuously adding injection water from the low pressure inj ect ion system so that the core remains covered by water. The CFT's will empty, and the HPI , system can be secured (see Guidelines for HPI termination in "Best Methods for Equipment Operation"). P ! Figure F-1 shows the reflood phase. Two examples are given: one for a large break at a high location in the system and one for a N larga break low in the system. j-l During the reflood phase, additional steam is created by: a) boiling of emergency injection water in the core. b) boiling of emergency injection water due to heat from the hot metal components, and 4 7-6-82 Appendix'F, Page F-9
BWNP-20007 (6-76) BABCOCK & WILCOX wu m e < NUCitAR POWER G(Nit ATION DIVI $ ION TECHNICAL DOCUMENT 74-ti2ss3t-00 ; c) boiling of emergency injection due to heat transferred from the secondary side of the steam generators (this is a source of additional steam only if primary side water re-enters the tube region of the steam generators). This steam is vented out the break to the reactor building. The reactor building pressure and temperature conditions may rise dur-ing the early stages of reflood. When decay heat and the heat from the hot metal and steam generators drop, the steam release rate will decrease and the reactor building pressure and tempera-tore will decrease as the sprays and coolers can remove more heat than is being released. Figures F-2 and F-3 give a summary of the expected response (P-T dia-gram) of the RCS during large hot and cold leg breaks. The diagrams also outline the sequence of events. At the end of blowdown, the reac-tor coolant is in a saturated condition at or near the containment pressure. Thereafter, the core outlet thermocouples should be used to check the core outlet temperature for saturation. At the end of the blowdown the core outlet thermocouples may indicate superheated condi-tions since the core may be completely vcid of water. As the ECCS re-floods the reactor vessel, the fuel rods will be quenched and the core outlet temperature should return to saturation. O 1 7-6-82 Appendix F, Page F-10 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX NuuiER NUCLEAR POWER GEN 5RATON DIVISION 74-t i 2ss3i-oo (~')
'R /
TECHNICAL DOCUMENT Small Breaks The second category of LOCA's are small breaks. Small breaks are r >t as severe as large breaks. The depressurization is much slower, the nass is lost at the lower flowrate, and the core cooling can be main-f'N\ tained throughout the accident as long as the ECCS is operating. For ( ( small breaks the operator can play a vital role in minimizing the con-sequences of the accident. As a rule of thumb, small breaks are between small le ak s and large breaks: Small Leak Large Break p Breaks for which the (Area approximately (
\s) leak rate is within equal to or greater the capacity of the < SMALL_ BREAK < than that of a MU system 10 in, diameter hole).
Small breaks include cracks in the primary system piping, breaks of O i lines attached to the RCS (HPI lines, letdown line, spary line), and j s v) failed open pressurizer relief or safety valves. The reference hole sizes given above are not exact boundaries for small breaks. They are given only as physical reference points to acquire a feel for the spread in break size fo r this category of LOCA's. The response of the RCS during a LOCA is actually a better way to gauge whether a s pe c i- [h
'g I fied LOCA is a small leak, small break, or a large break. '% )
7-6-82 Appendix F, Page F-Il DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusee NUCLE AR POWER genit ATION DIV15loN 7'-il2ss31-oo TECHNICAL. DOCUMENT i A small bre ak transient is characterized by a slow depressurization relative to large breaks. Flow conditions in the RCS change gradually and smoothly, and temperature and pressure gradients between different places in the primary system tend to be small. Rather than the dist-inct blowdown, refill, and reflood phases associated with large breaks, small breaks have a pe riod of relatively high loop flow fol-lowed by a period of relatively low or no loop flow. The no loop flow cond.tions are a semi-stable period where circulation stops and the steam separates from the water. When the steam and water separate the reactor coolant system is in a " boiling pot" mode. Figure F-4 shows a small break, including P-T responses and sequence of events. Early in the transient reactor coolant will circulate from the core to the steam generator, and the steam generator will remove heat. This part of the transient is called the " flow circulation" phase. Flow can circulate by natural or forced circulation. For very ( small breaks natural circulation (normal or two phase) will not be l lost, and the primary system will return to a subcooled state. For l "la rge r" small breaks, the circulation flow phase will end almost im-mediately after the RC pumps are tripped. For all small breaks, immed-l intely after the RC pumps are tripped. For all small breaks, core cooling is maintained during the " flow circulation" phase, i After the " flow circulation" phase of a small break, the reactor cool-ant " settles out". The water falls by gravity and collects in the 7-6-82 Appendix F, Page F-12 DATE:
- - - _ _ _ - _ - - _ -- _. _ = . _ . - _ - . . - . - - _ .
BWNP-20007 (6-76) BABCOCK & WILCOX Numsen [ j NUCLE AR POWit GENERATION DIVISION 74-112s531-00 ; TECHNICAL DOCUMENT { lower regions, and the steam bubbles up through the water and collects in the high points of the system. A " boiling pot" water level will j exist that will vary depending on: 1
, j i e Break Size e Break Location i
i i . e Primary to Secondary Heat Transfer (steam generator cooling) e Number HPI/LPI Pumps Operating i e Decay Heat Levels 1 These variab le s will cause the response ch aract eris t ics of the RCS to
- change in different ways af ter the reactor coolant " settles" into the s
" boiling pot". The effects of these variables will be discussed next.
To illustrate the effects of break size, steam generator cooling and t the other small break variables, the following five specific examples will be considered: , i
- 1. Small breaks large enough to depressurize the RCS
- 2. Small breaks which stabilize at approximately secondary side 1
pressure i
- 3. Small breaks which may repressurize in a saturated condition t
- 4. Small breaks without primary to secondary heat transfer
- 5. Small breaks within the pressurizer steam space These discussions will also highlight the ef fects of break size on the l' " flow circulation" phase of a small break.
7-6-82 Appendix F, Page F-13 DATE:
.- . _ _ _ _ _ _ . _ . _ . _ _ _._ ___ _ _ _ ] E _ _ _ _ _ __,.. _ _ _ _ . _ __ __ _ _ _ _ _
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sta NQ((( At POwta GENER ATION Divi 5loN 74-1125531-00 TECHNICAL DOCUMENT Small Breaks Large Enough to Depressurize the RCS At the "large" end of the small bre ak spectrum, the breaks release enough mass and energy from the system such that the primaiy system will continually depressitrize to a stable low pressure condition. For these breaks the " flow circulation" period is over very rapidly. RCS pressure will also drop very rapid below the s eco n'd r ay side pressure. Therefore, fo r these breaks primary to secondary heat t rans fe r does not play a vital role in the accident. During the " boiling pot" period, the water levels in the system will quickly drop to the eleva-tion of the break. Thereaf ter the water levels within the system will decrease until emergency cooling water flowrate exceeds the rate of water boiloff in the core. The net amount of water in the system will then increase slowly, and the water levels in the primary system will gradually build back up to the break elevation. For small breaks of this size the primary system ' sill never refill water above the break. Figures F-Sa and F-Sb show two examples of the system P-T response up to the time where the lowest system water levels occur. These figures show that these breaks depend upon the llPI and CFT's to maintain ade-quate water in the system to keep the core covered. When the system finally depressurizes, LPI increase the water level up to the break elevation. The steam generators provide very little cooling for this size of bre ak and are not very important for accident mitigation. O 7-6-82 Appendix F, Page F-14 D ATl! .
BWNP-20007 (6-76) BABCOCK & WILCOX NuusEn NUCtf AR power GENER ATION C.'Vi$loN 74-i 2ss31-00
/'~' x TECHNICAL. DOCUMENT
( ) N,_,/ Small Breaks Which Stabilize At Approximately Seconda y Side Pressure When the break size is smaller than that of the previous example, the leak rate and consequently the energy removal rate of the break de-creases (the bre ak removes energy when reactor coolant escapes). ,f ] These smaller breaks cannct remove energy from the primary system fast
- /
L/ enough to allow a continuous RCS depressurization. Following the
" flow circulation" portion of the transient, the system will stabilize at or near the secondary side pressure. For these breaks, the steam generators can remove a significant amount of the core's decay heat by condensing primary system steam within the tube region. Steam gene-rator operation is important fo r these break sizes. To prcmote reac-tor coolant steam condensation (commonly called reflux boiling) and to g']!
i reduce RCS pressure, the SG water levels should be increased with AFW C/ to 93" on the startup range. This will occur automatically when the RC pumps are tripped on loss of subcooling margin and SFAS actuates. The system response for small breaks within this size range is illus-trated in Figure F-6. As shown in Figure F-6, the " flow circulation period" is short-lived, and the system stabilizes slightly above steam [ generator pressure. Because the RCS pressure remains high, HPI is the only means by which emergency cooling water can be supplied to ensure cooling until the system depressurizes further. Since the genera-core tors are removing heat from the RCS by condensing steam, another impor-tant operator action will be to begin a plant cooldown and depressuri-e zation by decreasing steam pressure. The primary system will ( s), 7-6-82 Appendix F, Page F-13 DATE:
BUNP-20007 (6-76) BABCOCK & WILCOX NuusEn NUCLEAR POWER GENERATION Division 7'-t i 2553 i-oo TECHNICAL DOCUMENT depressurize because more steam will be condensed. As system pressure is decreased the ECCS system will flow at a higher ra e, and a refill of the system necomes more likely. During the refill, RCS presure may stabilize or increase slightly (while still saturated) as the RCS water level covers the conder. sing surface in the steam ge ne ra to rs . For this break size range, a refill and repressurization is not likely until the RCS pressure has oeen lowered to the LPI pressure. Repressurization of the RCS wh en the RC is in a saturated state will be more likely for the next category of breaks in the small break range. Small Breaks Which May Repressurize in a Saturated Condition For the break sizes smaller than the two previous examples the loss of reactor coolant mass and energy is sufficient to initially depressu-rize the RCS and results in enough steam formation to interrupt loop circutation. When the flow stops and the reactor coolant " settles out", the RCS may gradually repressurize. The repressurization occurs because the break alone is not large enough to depressurize tne RCS and the reactor coolant steam in the hot leg cannot be condensed because the reactor coolant water level is higher than the SG tubes. The primary system can repressurize as high as the pressurizer safety valve setpoint. Once enough RCS inventory has been lost and the reac-tor coolant water level rops below the AFW injection level, reactor coolant steam will condense on the SG tubes and the reactor coolant 7-6-82 Appendix F, Page F-16 DATE.
BWNP-20007 (6-76) BASCOCK & WILCOX u.3 ,,, NUCLEAR POWee GtHERATION OlVtSION 7'- u 23 5 31-00 T TECHICAL DOCUMENT pressure will drop to the steam generator pressure. Figure F-7 illus- l l trates this repressurization behavior as it would be observed on a P-T l l diagram. During the repressurization phase, pressurizer level may re-l turn on scale, as shown in Figure F-7, because the reactor coolant is heating up and expanding. For more severe cases, the pressurizer i
\ level may go off scale high. Because the reactor coolant is satu-rated, the HPI and MU should not be throttled. Both HPI pumps and hoth MU pumps should be kept on full, and the PORV should be opened if itCE pressure increases above the PORV pressure setpoint.
Summary m The previous three examples highlighted the importance of small break size and ateam generator cooling and showed how these two variable in-terplayed to change the reactor coolant system pressure. The reactor l coolant system pressure is one of the most important factors in small bresk control. When the reacataor system pressure is high the leak flow will be high and HPI flowrate will be low. The steam generator is valuable because it condenses reactor coolart steam and lowers the RCS pressure. To get the best e f fect steam generator level should be
}
! raised and AFW flow continuously run until the steam generator level I is at 93" on the startup range. Although decay heat level and HPI flow rate were not specifically ad-dressed for their ef fects, their influence is important. As the decay heat level lowers with time (or if it was low at the start) the heat l DATE: 7-6-82 Appendix F, Page F-17
BWNP-20007 (6-76) BABCOCK & WILCOX Numien NUClt AR POWER GENER ATioN DIVl5lGN 74-112ss31-oo TECHNICAL DOCUMENT input to the reactor coolant is reduced and the HPI flow can more easily " match" decay heat. Pressure will fall quicker with low decay heat levels than with high levels. HPI flowrate is important because it wilI keep the core covered and cooled when two HPI pumps are r.a n-ning. One pump will take much longer to match decay heat and if the system pressure is high the flowrate will be very low or none; steam generator cooling is used to reduce the RCS pressure to get more HPI. The next three examplec will repeat the first three, except that eteam generator cooling is not available. The final example will illustrate the importance of break location. Small Break Without Primary to Secondary Heat Transfer Steam generator heat transfer becomes more and more important as the break size gets smal le r . This is due to the inability of the break to remove decay heat by release of reactor coolant. Without primary to secondary heat trans fer, the RCS is more likely to repressurize or to depressurize more slowly. Figures F-8 to F-10 show system responses fo r three di f fe rent break sizes. These break sizes are similar to those previously illustrated; except without steam gene-rator cooling. Figure F-8 shows the size where the break is large enough to continu-ally decrease RCS pressure. For these breaks, which can depressurize 7-6-82 Appendix F, Page F-18 DATE:
l BWNP-20007 (6-76) l SABCOCK & WILCOX i l
- NUCLEAR POWee GSNERAilON oIVl5lON 7'- n 2 n n-00 C'sTECHillCAL DOCUMENT l b the RCS below the normal secondary side pressure, the loss of secon-l
- dary inventory has little or no ef fect on the transient. Core cooling J is maintained by the ECCS. Becauae these breaks are fairly large, pressurizer level would rapidly fall and remain of f-scale low. Figure F-9 shows a smaller break. The RCS quickly depressurizes; the HPI is actuated; and the RC pumps are tripped when the reactor coolant subcooled margin is losts. Pressurizer level also goes off-scale low. l When flow circulation stops and the steam and water separate, the RCS < presaure hangs up. RC pressure would normally drop to the secondary i s ide pressure if primary to secondary heat trans fer were available. If feedwater is restored and primary to secondary heat transfer es- ' y tablished, the RCS pressure would drop and boile r-condense r cooling would start. The PORV should be opened to reduce RC pressure. ' Figure F-10 shows a transient where HPI fails to start. The results are similar to a transient where HPI does not fail but the HPI flow is small because the ' leak is small or the RC pressure does not go below the HPI pump cutoff head. The HPI flow is not large enough to cool s the core. The refo re , the RCS repressurizes when the secondary inven-tory is boiled of f. It is possible for the RCS to repressurize to the i pres sure se tpo ints of the PORV or the pressurizer safety valves and t I for pressurizer level to go off-scale high. For any break size without secondary inventory, the operator should start MU/HPI cooling i DATE: 7-6-82 Appendix F, Page F-19 L _.- _- . _ . _ _ _ . . _ _ _ _ _ . ~ -_
BWNP-20007 (6-76) BABCOCK & WILCOX Nu sta NucteAn Powen GENERATION DIVillON 74-1125531-00 TECHNICAL DOCUMENT as soon as primary to secondary heat transfer is lost. The RC pumps should also be tripped if the RC subcooling margin is lost. Figure l i F-10 shows a transient where no ope ra tor actions are taken for the f its t 20 minutes. After 20 minutes, Figure F-10 chows the ef fect s of restoring feedwate r and starting HPI. The RCS depressurization causes an ou t s ur ge from the presrurizer. A complete loss of pressurizer level may result. Thereafter, the system response if similar to that of a small break with primary to secondary heat trans fer. Small Breaks Within the Pressurizer St am Space For small breaks within the pressurizer, the system pr e s s ure-t empe ra-ture re s po ns e will be similar to that discussed previously. The in i-- tial de pre s s ur i za t ion will be faster because steam, rather than water, is released (steam has a higher speci fic volume and enthalpy than water does). The big change for pressurizer breaks is the response of pres sur ize r level. Figure F-ll shows a P-T diagram and transient pres-sure and pressurizer level histories for a stuck open PORV. As shown in Figure F-ll, pressurizer level initially rises; this is due to the pressure reduction in the pressurizer and an in surge into the pres-surizer from the RCS prior to reactor trip. Once the reactor trips, the reactor coolant contraction (normal post trip re s ponse's causes a decrease in pres sur ize r level. Because RC pressure is dropping, flashing occurs in the hot leg ( sa t ura t ion ) , and the steam bubbles expa nd and force water into the pres s ur iz er . The pressurizer subsequently fills and will remain full for the remainder of the DATE* Appeniix F, Page F-20 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX Numien NUCLEAR POwta GENERATION OlVISION 7eii2ss31-00 (~');
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TECHNICAL. DOCUMENT transient. In the long term, the pressurizer may contain a two phase mixture, and the indicated level may show that the pressurizer is only partly full. This peculiar behavior is the main reason that pressu-rizer level c'annot be relied on to measure recctor inventory untit the reactor coolant is subcooled. If the b reak in the pressurizer is 73 [ ) (/ f airly small and can't be isolated, the HPI can refill the system and the pres sur ize r can be water solid and subc ooled . This would be the expected res ponse for a stuck open PORV without closing the PORV isolation valve. For these cases, attempts to draw a pressurizer steam bubble will f ail, and a solid water cooldown will be necessary.
~ Small Leaks I
)
\v/ Small leaks are events where the loss of reactor coolant is within the capacity of the normal makeup system at system operating pressure with both MU pumps on and letdown isolated. Several pos sible occurrences which could fall in this category are leaking pressurizer safety or relief valves, leaking RC pump seals or tiny cracks in the primary syetem piping or attachments (instrument lines, RTD hotwells, or aux-The category would correspond to approximately
,[ )1/
iliary systen lines).
\V a 3/8 inch or less diameter hole in the primary system.
When the leakage rate is very small, these is no impact on the primary or secondary system. These small-small leaks are discovered due to the necessity to add water to the makeup Lank more frequently than ( ' normal. The action required is in the Technical Specifications. ( j
\, /
DAH: 7-6-82 Appendix F, Page -21
BWNP-20007 (6-76) BABCOCK & WILCOX " ' NUcit AR POWit GENERATION DIV1510N 74-1125531-00 TECHNICAL DOCUMENT At the upper e nd of the leak rate for small leaks, more direct symp-toms may be observed, such as
- 1. Excessive MU Flow
- 2. Decreasing pressurizer level and pressure
- 3. Reactor trip These symptoms would likely exist when the leak is matched by the MU system. This transient is very slow, but pressurizer low level and pressure and makeu p tank low level alarms show the imbalance in reac-tor inventory control. There is time to increase the makeup rate to the RCS by starting the second MU pump and to stop letdown. Subcool-ing should not be lost; but if a reactor trip occurs, the pressurizer might drain if level had initially dropped and HPI will be started.
The plant can be cooled down in a near normal manner and as the system is cooled down the RCS pressure will drop. This will decrease the le ak rate and increase the makeup capability. Thus the system pres-l sure will be easier to control at lower pressures. However, if a small leak in the pressurizer steam space occurs the steam release rate is greater than the steam production capacity of the pressurizer l heaters. If so, the pressurizer will fill and a solid water cooldown l l will be necessary. 3.0 POST LOCA - PLANT CONTROL The primary objectives during a LOCA are to maintain core cooling, to cooldown and depressurize, and to establish a stable long-term cooling 7-6-82 Appendix F, Page F-22 DATE:
BWNP-20007 (6-76) BABCOCK & WILCOX Numen ) NUCLEAA POWER GENEAAftoN DIYl$10N 74- 2ss31-00 P TECMICAL DOCUMENT mode. Maintenance of core cooling is the first priority and is achieved through operation of the ECCS, Maintenance of core cooling will limit fuel damage and thereby limit radiation release. With essentially no actions other than verifying I the rods fully inserted following a reactor trip and tripping the reactor coolant pumps on loss of subcooling, the core can be cooled by ensuring that the ECCS works when required and can be run continuously until their safety function is satisfied and that for small break the AFW system operates. In addition to ensuring core cooling, the opera- l i. tot must make sure that containment cooling is working and that l l long-term cooling must be established. The general things to be done l i after a LOCA are: 4 Core Cooling i
- 1. Actuate HPI on loss of subcooled margin (if not automatically l
actuated).
- 2. Trip RC pumps on loss of subcooled margin and verify AFW starts.
l
- 3. Verify automatic start and injection flow rate of LPI and HPI af ter SFAS.
- 4. Balance LPI and HPI as necessary.
- 5. Verify that Core Flood Tanks release when RC pressure dropc below i
f 600 psig. l V l DATE: 7-6-82 Appendix F, Page F-23 l _ . - . . , _ ~ , _ _ . ~ . , . . , _ _ , _ , . _ _ . _ , _ . , _ , . _ _ _ . _ . . _ . _ _ _ . _ _ _ _ _ _ . . _ . _ , . _ . . . _ _ _ . _ _ _ _ . _ _ . _ _ _
BtJNP-20007 (6-76) BABCOCK & WILCOX wu,,ite NUCttAR POWER GENERAfsON OlVISION 74- t i 2 n 31-oo TECHNICAL DOCUMENT
- 6. Locate and , if possible, isolate the break (check for steam genera-tor tube leaks and leaks inside or outside contaiment).
- 7. Control the SG steam pressure and raise the steam generator level to 93" on the startup range if SFAS level 2 has actuated or the subcooling margin is lost (<axcept for the " bad" generator for tube ruptures) for small breaks, Containment Control
- 8. Refer to Chapter G in Volume 1 of Part II.
Long Term Cooling
- 9. Switch the s uc t ion of the LPI to the containment emergency sump when the BWST low level is reached. If the RCS is not depres-surized enough to use LPI, place the HP1 in piggyback on the LPI and use recirculation from the sump for injection
- 10. For large br eak s in the sump recirculation mode, stop one LPI train and leave one running so long as the core exit thermocouples indicate the core is cooled. This will prese rve one pump in case it is needed later. Continue to monitor LPI flow.
- 11. Take actions to prevent boron precipitation.
For large breaks, maintenance of core cooling leads directly to a stable long-term cooling mode because the break is large enou gh to depressurize the RCS and secondary pressure and inventory control have little or no ef fect on the course of the accident. O DATE: 7-6-82 Appendix F, Page F-24
BWNP-20007 (6-76) BABCOCK & WILCOX Nuuset NUClf AR power GENERATION DIVI $loN 7'-:i2ss31-00 ( _ l TECHNICAL DOCUMENT
\ /
rs For small breaks, core cooling can be maintained using two HPI pumps. The core will be cooled, but reactor coolant will continue to be lost in large quantities as long as the RCS is at high pressure. Depres-surization and cooldown using the steam generator is necessary to de-
/ crease the leak and to allow for a potential refill of the primary
\
'}/
'A > system. Depres sur iza t ion to cold conditions is essential fo r tube leaks because all the BWST water is lost out of the leak and cannot be recirculated from the containment sump; the plant must be placed on the decay heat removal system before the BWST drains.
In the following sections, the general approach to accident mitigation [' (core cooling) and establishment of long-term cooling using the Abnor-
\ )
'v' mal Transient Operating Guidelines will be outlined for loss-of-coolant accidents. the full spectrum of LOCA's will be addres-sed with the exception of the small le ak s which would typically never result in loss of reactor coolant subcooled margin. For small le aks ,
i l except in pressurizer steam space, the loss of reactor coolant can be 1 matched by the normal makeup system and plant control is essentially p-no different than any other abnormal transient which results in a (v } reactor trip followed by a plant cooldown and depressurization except makeup water comes from the BWST and the plant should be cooled down before the BWST empties. Small leaks in the steam space will le ad to a solid water cooldown if
/' }
( ) the steam break cannot be compensated for by the pressurizer heaters.
%J DATE: 7-6-82 Appendix F, Page F-25
BWP-20007 (6-76) BABCOCK & WILCOX wu-stt NUCtt At Powtt genit AfloN DIVlhioN 74-i 12533 i-0o TECHNICAL DOCUMENT Ceneral Overview Table F-1 outlines the general approach to post-LOCA plant control from a core cooling standpoint. This sequential breakdown of operator actions has many of the same features described in the accident miti-gation chapter for non-LOCA's except that actions to cool down the plant and to establish long-term cooling that are unique to LOCA have been included. If a LOCA can be located and isolated, however, the plant can be stabilized at or near a hot standby condition since the loss of reactor coolant has been stopped. Although not shown in Table F-1, the containment environment must also be controlled; both short and long term actions from a containment integrity standpoint will be discussed. O Imn.ediate Actions and Vital System Status Checks The operator actions required during the first 2-3 minutes of a LOCA are identical to those for any abnormal event. These actions include 1 the immediate actions and verification that systems are working pro-perly as outlined in the Accident Mitigation Chapter. For LOC A' s , some of the required actions are of special significance:
- 1. Verification that SFAS starts HPI and LPI and that containment cooling and isolation have occurred as required
- 2. Trip RC pumps if the reactor coolant subcooled margin is lost.
The SFAS initiates and aligns the HPI and LPI systems for emergency injection, start s emergency containment cooling systems (sprays and l l 7-6-82 APPendix F, Page F-26 DATE: l
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( < kj fan coolers) for high pressure containment conditions, and isolates non-vital containment pe net rat ions to limit offsite dose releases. Tripping the RC Pumps, as explained in the chapter on Fest Methods of Equipment Operation, is a preventive action specifically fo r a small ,/ break. If the RC pumps are tripped on loss of reactor coolant sub-
' cooling margin, enough reactor coolant will be retained within the primary system (not lost out the break) and HPI will maintain the core cove red and cooled. Both loss of subcooling margin and SFAS actuation are alarmed (visually and audibly) in the control room to alert the operator of the plant's status. By verifying that the SFAS actuated systems automatically start (or by applying corrective actions in the ,r~~N event of failure in SFAS actuated systems) and tripping the RC pcmps, l I 'LM the operator ensures:
- 1. Adequate core cooling for the vast majority of possible LOCA's.
- 2. Containment integrity so that the offsite doses will be within acceptable limits.
Monitoring
,m
[ ') During and immediately following the immediate actions, the P-T and
\ /
%J other parameters should be monitored to determine if the abnormal transient i:. a LOCA and not some other accident. Many overcooling accidents will look like small breaks, and all small break LOCA's will not look the same. In some cases, a LOCA can only be determined by showing that some other accident is not underway. Usually the acci-
) dents that can be eliminated readily are overcooling accidents; these
% ./
7-6-82 Appendix F, Page F-27 DATE:
BWNP-20007 (6-76) BAPCOCK & WILCOX NuusER XUCLtu POWER GENERATION DIVl$10N 74-1125531-00 TECHNICAL DOCUMENT can often be eliminated by reviewing steam pressure, secondary water level, and feed flows. The operator should get a " fe e l" for LOCA's by comparing the examples of this section with other transients. LOCA's do have some unique characteristics; these are shown in Table F-2. (This is Tabie 4a from the " Abnormal Transient Diagnosis and Mitigation" chapter, repeated here for convenient reference.) For most LOCA's the general symptoms shown in Table F-2 will be appar-ent, and other indications will supply additional evidence that a LOCA is in progress. Table F-3 scopes the characteristics of a wide range of LOCA bre ak sizes and summarizes the other evidence to be used to diagnose that a LOC A is occurring. Table F-3 includes both the pres-sure-temperature response characteristics and other event or plant symptoms for the comple te spect rum of LOCA's. Some LOCA's can be iso-
' lated. Table F-4 gives symptoms for LOCA's that can be located and isolated. (This is Table 4b from the " Abnormal Transient Diagnosis and Mitigation" chapter repeated here for convenient reference.)
Some LOCA's (such as large breaks) have such distinctive ch aract eris-tics that a quick diagnosis is assured. For small LOCA's the event may not be properly diagnosed for some time into the event, but core cooling is assured so long as the ECCS systems are flowing fully. The symptoms in Table F-2, F-3 and F-4 should be studied because some LOCA's can be isolated and the loss of reactor coolant stopped. DATE: 7-6-87 Appendix F, Page F-28
BWNP-20007 (6-76) , BABCOCK & WILCOX Numset NUCLEAR POWER GENERAfloH DIV15 ton 74-112553i-00 tO TECHNICAL DOCUMENT Corective Action for LOCA's - Introduction , As outlined in the " Diagnosis and Mitigation chapter", the actions to be t aken in response to abnormal plant symptoms are aimed at restoring and controlling primary to secondary heat transfer or starting backup cooling methods. For LOCA's, in addition to restoring primary to b secondary heat t rans fer an RCS cooldown and depressurization must be l started. The plant is not in a safe condition until it is depres-I ! surized and the ECCS systems are aligned for long-term cooling (unless I the break is isolated). For large breaks the plant will depressurize i quickly and only act ions to es t ab lish and maintain long-term cooling are reuqired.
\
/ The three symptoms for which operator actions are based during any ab-normal event are:
- 1. Lack of Adequate Subcooled Margin
- 2. Lack of Primary to Secondary Heat Transfer (overheating)
- 3. Too much Primary to Secondary Heat Transfer (overcooling) l During a LCCA all of the above symptoms can exist at some time during r
the transient. Lack of adequate subcooling and/or lack of primary to secondary heat transfer are the expected symptoms, because the loss of reactor coolant will result in saturated conditions and impede the transport of core heat to the steam generators. The following sec-tions will show how to control the RCS for LOCA, control containment ! cooling and ensure isolation, and establish long-term cooling. , 7-6-82 Appendix F, Page F-29 , DATE: l
i l BWIiP-20007 (6-76) BABCOCK & WILCOX Numan NUCtf An POwta GENEnATION D:VISIO N TECHNICAL DOCUMENT 74-1125531-00 Corrective Actions for LOCA's - Loss of Subcooling Margin Figure F-12 outlines the symptoms of and gener al actions for a lack of subcooling margin fo r a LOCA. Tripping the RC Pumps is a preventive action for small breaks as discussed previously, and manual HPI actua-tion se rve s as a backup to the normal start feature provided by the f SFAS. Balancing the HPI flows is a general action to be taken follow- ) ing HPI actuation at any time, but the reason for balancing is only l important for LOCA's. A break in an HPI line at the RCS nozzle or in the cold leg near the RCS nozzle would allow a large flow rate in the line; adequate HPI flow would not reach the reactor vessel. Balancing will allow adeq ua te flow to reach and to cool the core from the other injection lines. The chapter on Best Methods outlines specific ways to balance HPI flows. The third action in Figure F-12 is raising the steam generator water level to 93 inches on the startup range. Con-trolling the water level high in the steam generator is required for small LOCA when the pumps are tripped because it aids reflux boiling (boiling in the vessel and conde ns ing in the steam generator) for i decay heat removal and can help to establish natural circulation. The last action, which is to attempt to locate and isolate the break, should be taken whenever a LOCA is suspected. Typically, the PORV and block valve should be routinely closed under these conditions as a pre-cautionary measure. In addition, Table F-4 also outlines symptoms of other specific LOCA's wh ich can be located and shows how they can be isolated. By performing the four actions given in Figure F-12, core O DATE: 7-6-82 Appendix F, Page F-30
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCLEAR POWER GENER ATioN DIVISION 74-i l 2s s 31-00 (~') TECHNICAI. DOCUMENT i v/ cooling is assured for che f ull spectrum of small breaks. If the reac tor coolant returns to a subcooled state due to break isolation or sys-tem refill by HPI, the plant can be cooled down in a near normal manner if no other abnormal plant symptoms exist. If saturated condi- / . tions perstst with heet transfer te the steam generators (see example i ]
\d P-T diagram on Figure F-12), additional evidence to support a LOCA diagnosis should be gathered (see Tables F-2 and F-3)- With evidence to confirm or suspect a LOCA, an immediate cooldown of the system to establish a stable long-term cooling mode should be started. If lack of primary to secondary heat transfer (overheating) exists, these conditions must be treated to restore use of the steam generators for f^'s plant cooldown, f ) 'J Corrective Actions for LOCA's with Lack of Primary-to-Secondary Heat Transfer Figure F-13 outlines the symptoms of and general correct ive actions for a lack of primary to secondary heat trans fe r (overheating) during a LOCA. Actions are shown for two types of overheating transients.
7 (\ ) The first type (right-hand portion of Figure F-13) is overheating due J to a loss of secondary inve ntory (no feedwater), and the corrective action is to use MU/HPI cooling (open PORV with two HPI pumps on) un-til fe edwat e r is restored. The second type is overheating due to reverse heat transfer (RC pressure drops below secondary pressure) or due to loss of steam conde ns ing surface in the steam ge nerators as the
/ s.
\
I primary system refills. For reverse heat transfer symptoms to exist, t., )
%_/
7-6-82 Appendix F, Page F-31 DATE:
bWNP-20007 (6-76) BABCOCK & WILCOX uumsra NUCLEAR POWER GENERATION OlvisiON 74-li2ss3t-00 TECHNICAL DOCUMENT a f airly "large" small break must be in progress because the break is For these LOCA's the primary large enough to depressurize the plant. If and secondary systems can be recoupled by lowering steam pressure. RC pressure fo llows steam pressure, small break cooldown procedures should be followed to establish long-term cooling. For plant symptoms wh ich imply a lack of primary tc secondary heat t rans fe r due to the inability to condense steam in the generators, methods are available ta st imul a te a return to " normal" natural circulation. The method used to start natural circulation is to lower steam saturation tem-pe ratur e abou t 50F below core exit thermocouple temperature and bump an RC pump (see pump restart criteria in the " Equipment O pe ra t ion" Wen the pump is bumped RCS pressure will drop to secondary chapter). pressure as primary steam is swept into the condenses in the steam generators. As a result of the reduc t ion in primary system pressure, the HPI can add more water to the primary system and the system may refill and establish natural circulation. If the primary and secon-dary sys t ems do not recouple after four succes s ive pump bumps, steam temperature should be dropped until it is 100F below the core exit the nnocou ple temperature and a RC pump should be started and run a continuously. After these actions, a cooldown can be started to establish a stable long-term cooling mode. If the RC pumps are not available, the PORV can be opened (HPI cooling) to minimize the RCS pres sur e increase and to increase HPI flow when lack of primary to secondary heat transfer symptoms exist. O DATE: Appendix F, Page F-32 7-6-82
BWNP-20007 (6-76) BABCOCK & WILCOX wo sen NUCLEAR POwte GENT 4ATION DIYtSION
' 74-ii2ss31-o0 TECHNICAL DOCUMENT t
Corrective Actions for LOCA's with Overcooling Symptoms i LOCA's are not overcooling events. Nevertheless, the P-T response of 4 RCS during a LOCA can res emb le a severe ove rcoo li ng event (too much primary to secondary heat t rans fer) initially because the reactor cool-ant pressure and temperature will drop to staturation and steam pres-sure can also be low. Low steam pressure in both steam generators is pos s ib le (without equipment failures) because LOCA's can block the pri-mary to secondary heat trans fer process. This similarity between 1 non-LOCA and LOCA events will be short-lived because the non-LOCA over-cooling event will return to a subcooled condition fairly rapidly , I after HPI and makeup are actuated. Figure 22 in the Abnormal Transient Diagnosis and Mitigation ch apte r outlines diagnostic techniques and corrective actions for too much pri- { mary to secondary heat t rans fe r. Corrective actions are identified for high steam generator level (too much feedwater) and low steam pres-sure conditions. These actions. are also appropriate for LOCA's because equipment failures can occur in conjunction with a LOCA to cause excessive heat trans fer, even when the reactor coolant is satu-l rated. If steam pressure or feedwater control in one steam generator is - lost, that steam generator can be isolated and allowed to boil dry with no loss of core cooling capability. i , i i i DATE: 7-6-82 Appendix F, Page F-33
BV:iP-20007 (6-76) BABCOCK & WILCOX sumstR HUCLEAR POWER GENERA' TON DivaSION 74-12 2s s3i-on TECHNICAL DOCUMENT Cooldown for Smal1 Breaks The act ions identified in Figures F-12 and F-13 of this ap pe nd i x and Figure 22 of the Abnormal Transient Diagnosis and Mitigation chapter allow the operator to maintain the plant in a safe condition; but since the plant is still at high pressure, the leak will continue (un-less it has been i so la t ed ) . Cooldown and depressurization is re-quired. At this point into the accident, the conditions can be one of the following:
- 1. Isolated LOCA with the plant at hot standby ( forced or natural circulation).
- 2. Small break with HPI "on" and
- a. Saturated coo lant conditions with primary to secondary heat transfer (RC pumps "on", restarted when treating lack of primary to secondary heat trans fer, or "of f"); or
- b. Subcooled coolant conditions with primary to secondary heat trans fer (RC pumos "on" or "of f"); or
- c. Saturated coolant conditions (RC pumps "off") with no pri-mary to secondary heat trans fer (HPI cooling).
- 3. Large break with the primary system completely depresuTized.
Cooldown and depressurization from any of these conditions is possible with or without steam generator heat t rans fe r. Steam generator heat t rans fe r will permit a better cooldown and depressurization, but the plant can be depressurized without it. Large breaks do not require de-pressurization, but long term cooling must be established. The fol-lowing section will discuss cooldown from each of these conditions. DATE: 7-6-82 Appendix F, Page F-34 0
t BWNP-20007 (6-76) BABCOCK & WILCOX Num:en NUCLEAR POwta GENEAAtlON OtVl$loN 74-112533i-00 TECHNICAL. DOCUMENT i Small Break Cooldown (With Primary to Secondary Heat Transfer)
- a. Subcooled Cooldown L
If the reactor coolant; has a subcaoled margin, the plant can be cooled i in ' a near normal manner to get on the decay heat removal system for ung-term cooling. The RC . Pumps should be started (see RC Pump re-i
- 9 h start criteria) if the plant is in a subcooled natural circulation con-dition on at least . one generato r (loop with spray line prefered).
i- With RC Pumps running and the. reactor coolant subcooled, the steam gen-erator water leve! s can also be cont rolled at the no rmal low level
-limits. If the LOCA has been isolated, a pressurizer bubble can be
' drawn and ' cooldown can proceed normally. If the LOCA is not isolated {; and not in the pressurizer, a pressurizer bubble can be drawn using the pressurizer heaters. The HPI injection system must be throttled i i~ to match the leak and to maintain the reactor coolant subcooled as the 4 . plant is cooled and depressurized. i l If the LOCA is in the pressurizer, a solid water cooldown is necessary * [- . because RC pressure is controlled by the discharge pressure of the HPI pumps. To maintain . plant control, HPI flow must be reduced slowly to reduce RCC pressure and coordinated with RCS temperature control (by the secondary system) to maintain desirable cooldown limits, t
.If 'a pressurizer level exists -the cooldown which causes RC contraction f'
I may be limited by the capacity -to makeup to the RCS with a LOCA. j i I 7-6-82 l Appendix F, Page F-35 i' _DATE. l
BWNP-20007 (6-76) BABCOCK & WILCOX NuyseR NUCLEAR POwtR GENERAflON DIY15 TON 74- 2 n 31-00 TECHNICAL DOCUMENT Tigh t control of HPI is necessary at low RC temperatures at avoid a sudden flow increase and subsequent inc rease in system pressure. For plant cooldown (wi th or without a pressurizer s t e a.a bubble) the depletion rate of BWST should be monitored. Makeup to the BWST should be started when level reaches ten feet. If low BWST levels occur, the l LPI pumps must be aligned to t ake suc t ic a from the contairunent cnergency sump and the HPI-LPI systems aligned in the piggyback mode (HPI pumps takes suc t ion from discharge of the LPI pumps) so as to maintain continuous RCS inventory control. The CFT's may also be iso-lated if the plant is cooled down in a subcooled state.
- b. Saturated Cooldown For some small b r e ak s , a plant cooldown with saturated reactor coolant conditions will be necessary. In this mode, the steam pres sur e controls the reactor coolant temperatur e and RCS pres-l sure. Pressurizer bubble control is not possible when the RCS is saturated. When steam pressure is reduced, the RCS pressure will l
- drop as long as primary steam can be condensed in the steam gene-rator tube region. For saturated cooldown, a high steam generator level is required and the dPI cannot be throttled as leng as the subcooling margin is lost. The RCS may refill and re-establish subcooling during saturated plant cooldown because (1) the leak flow decreases and HPI flow increases as RCS pressure is reduced, and (2) decay heat is slowly dropping. If the RC Pumps are off, the need to apply co r rect ive action for loss of primary to A 7-6-82 Appendix F, Page F-36 l
1
BWNP-20007 (6-76) SABCOCK & WILCOX wuusta NUCLEAR POwta GENT #ADON OtVISION 7'-1i23331-0o TECHNICAL DOCUMENT i secondary heat transfer (see left side of Figure F-13) may again be necessary. BWST level should be monitored and if it get low, the ECCS should be realigned to draw from the containment emergency sump to maintain continuous HPI inj ec t ion. If the kCS remains in a saturated condition, the CFT isolation valves should s not be closed unless the tank's water volume has been depleted. The plant cooldown and depressurization should be continued to acquire approximately 150 psig to establish long-term cooling. When 150 psig is reached the normal decay heat removal system can be engaged if the reactor coolant is subcooled. The normal decay heat removal system cannot be engaged if the reactor coolant is saturated,
\ because water level may not exist at the hot leg suction nozzle. When subcooled with the normal decay heat removal system started up, con-tinued makeup either from the HPI or from the LPI (aligned to the con-i tainment emergency sump or BWST) will probably be required. If the i LPI is used for makeup, then one string will have to be aligned to the containnent emergency sump or BWST and the other aligned in the "nor-1 ma " decay heat removal mode. Cooling in this configuration is re-( quired until the reactor coolant temperature is below 212F. At this time the additional makeup can be stopped if the break location is above the hot leg suction of the dwcay heat drop line. If it is below this level or it is not known where the break is, makeup should con-tinue.
)
sj DATE: 7-6-82 Appendix F, Page F-37 l l L - -t - . - - . , , .
BWNP-20007 (6-76) BABCOCK & WILCOX Numsta NUCLEAR POWER GENERAilON Olvl510N 74-ti2ss31-00 TECHNICAL DOCUMENT Small Break Cooldown (Without Primary to Secondary lleat Transfer. HPI Cooling) If a LOCA occurs and primary to secondary heat trans fe r is not pos-s ib le due to a total loss of feedwater, IIPI cooling is used to main-tain the core c oo led . Generally, as discussed in the Backup Cooling Methods chapter, this is not a pre fer re d mtthod of plant control and alternative ac t ions to restore primary to secondary heat trans fe r should be started at once. Should long-term reliance on HPI cooling be necessary, the operator's primary responsibilities wh ile the system is saturated are:
- 1. To open the PORV and leave it open.
- 2. To monitor the performance of the HPI system and realign the HPI in piggyback with LPI to draw from the containment emer-gency sump on recirculation if the BWST runs out.
While saturated conditions and no primary to secondary heat t r ans fe r exists, it is not possible to make the plant depressurize. repending on the break size .ind location, the number of HPI pumps operating, and the plant decay heat level, some different system trends that can oc-l cur are:
- 1. A slow but continuous depressurization as decay heat drops and the HPI is able to remove more energy.
- 2. Repressurization to the PORV setpoint followed by opening the PORV which leads to 1. above.
O l l 7-6-82 Appendix F, Page F-38 DATl! 1
BWNP-20007 (6-76) 8ABCOCK & WILCOX Nuusee NUCLEAR POWit GENERATION Divi $loN 74-li2ss31-oo TECHNICAL DOCUMENT
- 3. Either of the above followed by re-establishment of subcooled conditions (some small breaks will neve r allow a system refill and re-establishment of subcooled conditions).
Should the RCS return to a subcooled condition, HPI can then be throttled to slowly depressurize the plant as the reactor coolant temperature drops. S HPI cooling will eventually (may be many days) allow es t ab li shme nt of
- _a long-term cooling mode where the DHR System (normal or LPI mode) can i
be used . to keep the core cooled. This mode of core cooling (HPI cool-ing) should be maintained only until alternate means of cooling are possible. I Long Term Cooling Long-term cooling is defined as the time af ter a LOCA where the Dec - Heat Removal System, either in a normal or emergency mode (LPI), is operating and can be used for core heat removal. The _ duration of long-term cooling is the period between the onset of long-term and the end of core cooling requirements by the ECC system. The end of core-cooling requirements is the time when the core is removed from the reactor vessel or other pe rmanent means are used for core heat re-moval. The exact duration of long-term cooling will vary depending on l several factors, including the size of the break and the radiation l release. For the worst case LOCA (i.e., a large break), the duration l l- . l DATE: 7-6-82 Appendix F, Page F-39
BWNP-20007 (6-76) BABCOCK & WILCOX Nuusta NUCit AR POWER GENER ATION DIVI $lON 74-;125531-00 TECHNICAL DOCUMENT of the long-term cooling period may vary from one month to a maximum of one year depending upon the result ing accident consequences. For large breaks, long-term cooling could begin as soon as 30 minutes after the break occurs. For smalI breaks, long-term cooling condi-tions may take days to achieve depending on how fast the plant can be cooled and depressurized. T ab le F-5 presents a summary of actions required to establish and main-tain l ong-t e rm cooling following a loss-of-coolant accident. Table F-5 includes both core cooling and containment related act ions. To Establish Long-Term Cooling To establish long-term cooling af ter a small break, the decay heat re-moval system can be aligned with one train in the decay heat removal mode (subcooled only) and the other train in the LPI mode (possibly with HPI piggyback from the containment emergency sump or from the BWST). If the system is saturated, recirculation from the containment emergency sump in the LPI mode (possibly with HPI piggyback) estab-lishes long-t e r:n cooling. To es t ab li sh long-term cooling for large breaks, the LPI system is placed in the recirculation mode from the e ntainment emergency sump. l l l 1 Boron Precipitation Within twenty-four hours after a LOCA, actions should be taken to pre-clude the possibility of boron precipitation. With a large hole in 7-6-82 Appendix F, Page F-40 DATE: 1
4 BWNP-20007 (6-76) 4' BABCOCK & WILCOX NumsEn NUCLEAR POWER GENERATION Divi $10N 74-112553i-00 l TECHNICAL DOCUMENT certain areas of the RCS, the reactor can, acting like a distiller, boil of f almost. pure steam and leave impurities (mostly boron) to con-centrate in the vessel. If enough boron accumulates, core flow block-age might occur. To limit the ooron concentration, the operator should follow the plant procedures for preventing boron precipitation. i Preserving the LPI System for Long Term i If a large break LOCA has occurred, or if a small break LOCA has oc-4 curred that is in a location that does not permit use of the decay 4 heat system in the normal mode ( for core cooling or prevention of boron precipitation), -it may be desirable to take one decay heat train [ out of service to preserve it- for the future (so it can be placed back in use if the operating train develops problems). One decay heat 1 train can be removed from service safely if sump recirculction is in progress and the LPI flow in the opposite train is equal to or greater l than 1000 gpm. t, 1 ! 4 e i l i 1 1 l DATE: 7-6-82 Appendix F, Page F-41
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Figure F-2 SYSTEM RESPONSE FOR LARGE BREAK IN COLD LEG PIPING , Aho Xvailable On Aperture Caril 2600 NOTE: FOR A L ARGE BRE AR STE Au PRES $URE & 2'00 ~ TEMPERATURE 81LL SL0fLT OROP AFTER REFERENCE PolNT 6. POST TRIP 2200 - U. r- T'- d i 5 2000 - SU8C00 LEO ' '2
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=
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f I DOUBLE ENDED HOT LEG BRE AR REFERENCE fitE PolNTS (SECONOS) RENARAS 1 00 LOCA OCCURS 81TH THE PL ANT INiil ALLT AT 100s POBER 2 <1 0 RCS PRESSURE OROP110 RPS TRIP SETPolNT AND A REACTOR TRIP 151 3 <1 0 SUBC00 LING uaRGIN 15 LOST 4 <l 0 SFAS ACTUATED ON LOW RC PRESSURE. HPl AND LPl SILL BEC0ef CPERAllVE WITHIN 35 SECONOS 5 <1 0 THE REACTOR COOLANT SATURATES IN THE HOT LEGS 6 40 RCS PRESSURE CROPS BEL 0e SECONDARY SIDE PRESSURE; THE SG'S BE0i A HEAT SOURCE INSTEAD OF A MEAT SINE. i 90 RCS PRESSURE OROPS 70 600 PSIG AND BORATED WATER FR0u iME CFis ' TO ENTER THE REACTOR VESSEL 8 18 THE ENO 0F BLC9000N 15 RE ACHED AS THE RCS AND CONTAINMENT EQUAL PRESSURE (~ 20 40 PSIGL THE RCS is ESSENil ALLY VDIO (LitTLE E WATER EllSIS) ANO THE CORE CUTLET THER50C00PLES WOULO INDICATE SUPERHEATED CONDITIONS 9 19 ENO CF REFILL THE RV HAS BEEN REFILLEO UP TO THE 80iT0m 0F T" CORE AT THl3 ilut THE NPI AND LPI SYSTEMS ARE FULLV OPERAllt 10 30 THE CFT WATER V3tuuE 13 OEPLETED 11 350 THE CORE IS RECOVERED Si SATER AND IME FUEL'S TEllPERATURE EICUF TEReinATED THE CORE Elli THERe0 COUPLES RETURN TO SATURATED C1 ( 2500 2000 Y s 1500 \
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= 1000 N m 9 N 500 0
0 2 4 6 Ti I i 74-1125531-00 l
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~-
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. TEMPERATURE WILL SLLfLY DEvr AFTER REFERENCE POINT 6)
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0.5 FT2 BREAR IN COLO LEC PIPE REFERENCE TluE POINTS ESECONOS) REutR81 1 0 LOCA OCCURS elTH REACTOR INITI ALLY AT poser; 8 REAR IS EQUlvaLENT TO 9.5 INCH Ol AGETER HOLE. 12 02 PRESSURE DROPS OUE TO RELEASE OF REACTOR COOLANT CUT THE OREAA. 2 0.2 APS TRIP SETPolNT 15 REACHEO: A REACTOR TRIP 15 INITIATED. 2-3 0.2 0.5 OREAR 15 LARGE ENOUGH TO DEPRESSURIZE THE RCS TO SATURATED CON 0lil0NS 8EFORE PRESSURIZER EBPTIES. IITHIN O.5 SEC. SFAS IS ACTUATED AND THE RC PUuPS SHCULO BE TRIPPE0 8ECAUSE SU8C00 LING l$ LOST. 3 0.5 HOT LEG SATURATES. 34 0.5-40 RC PUuPS ARE TRIPPE0 AND AFW IS ACTUATE 0. THE #CS CONTINUES TO DEPRESSURIZE IN A SATURATED STATE ( THE HPl 8EGINS TO DELIVER FLOW TO THE PCS. 4 40 THE " FLOW CIRCULATION PER100 ENDS. AND THE PRiuARY S10E PRESSURE AND TERPERATURE DROP BELOE THE PRESSURE AND TERPERATURE OF THE SG's. THE SG's 8ECOSE A HEAT SOURCE. 45 40-110 80TH THE I:31 AND COLO LEG ARE SATUR ATE 0. ANO RC PRES $URI CONTINufS TO CR0P. 5 110 RC PRES $uRE HAS CROPPE0 T0 600 PSIG AND T 2 CFT's 8EGIN-ADDING 84TER TO RE ACTOR VESSEL. 56 110 200 RCS PRES $URE CONTINUES TO OROP. 5 200 RCS PRESSURE IS 8EL0s 200 PS!G AND THE LPI PuuPS BEGIN TO DEllVER P OW TO THE RE ACIOR VESSEL. T >200 THE ACS 8EGINS TO REfili 2ACK TO THE ELEVATION OF THE 3 REAR. CORE COOLING elLL CE BAINTAINE010 LONG AS LPl 15 CONTINUE 0. 2500 l
~
2000 l EiiOO l i E 1000 - i s i l - 500 l , 0 . . 0 50 100 ' 1,... A. t e ! u 74-1125531-00 1 A l
Fig'Jre F-Sa SYSTEM RESPONSE FOR SMALL BREAK WHICH CONTINUALLY 2 DEPRESSURIZES THE RCS (0.5 FT BREAK IN COLD LEG PIPE) 2600 NOTE' $ TEAR PRESSURE AND TEuPERATURE SOULO SL0tLY OROP AFTER REFERENCE POINT 4 2400 (REVERSE NEAT TRANSFER 900LO SE INDICATED) POST TRIP L_d,1
=
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{ REGION 3 1800 SUBC00 LED REGION g C APPROIlutif
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F 1 ENO POINT-POST TRIP 91TH TURAll0N . g,jNAfuRALCIRCULAil0N(THOI) n,0 - t I I 200 300 400 500 600 700 Reactor Coolant Ana steam Outlet temperature. F APBTUR-CA1) 200 s i_
= , 00 -
i 3 0 - \ % -. ISO 200 250 0 2 4 3 0 to 12 14 r Time, sec 1 8403050118 ff Also Xvailable on Aperture Card
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{ 0.1 FT2 COLD LEG BREAR REFE*ENCE TiuE POINTS (SECON05) RENARKS 1 0 LOCA OCCURS WITH REACTOR INITI ALLY AT P0tER E00lVALENT TO 4.3 INCH DI AMETER HOLE. 12 0 -5. 0 RCS PRESSURE AND PRESSURIZER LEVEL DROP 00E OT REACTOR C00LANT 00T THE BREAK. 2 5.0 RPS TRIP SETPOINT IS REACHE0; REJCTOR TRIPS 23 5-18 RCS DEPRESSURIZES TO SATURATED CCN0lil0NS; ACTUATED CN LOW RC PRESSUPE; AND THE PRESSL 3 18 PRESSURIZER IS COMPLETELY EMPTY AND THE HOT 34 18-140 THE RCS CCNTINVES TO DEPRESSUR12t. THE RC AFW IS ACTUATED; AND 2 PHASE NATURAL CIRCUL a 140 THE " FLOW CIRCULATION" PHASE ENDS; THE RCS CCNDENSER COOLING. 45 140 210 THE RCS CONTINUES TO DEPRESSURIZE. STEAM P TO AFW INJECT 10N AND REDUCED PRINARf TO SEC 5 210 PRIMARY PRES $URE CRDPS BELOW THE SECONDARY ARE NOW A HEAT SOURCE. 5-6 210 910 RCS PRESSURE CONTINUES TO OROP. 6 910 RCS PRESSURE HAS DROPPED TO 600 PSIG AND TH ADDING WATER TO THE REACTOR VESSEL. T > 910 THE RCS 81LL CONTINUE TO DEPRESSURIZE SL0tt SYSTEM BECOMES OPERATIVE. 2500 2000 -
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E g 1500 - E 5 m g 1000 - N 500 - 0 0 fi o 400 EDI f Time. 1 j 74-1125531-00
Figure F-50 SYSTEM RESPONSE FOR SMALL BREAR WHICH CONTlHUALLY 2 I DEPRESSURIZES THE RCS (0.1 FT COLD LEG BREAK) l 2500 P0$7 TRIP
, BREAK 15 2200 - N 3 r---'Oi TO LOSS ? NOTE STEAR PRESSURE AND TEMPERATURE \ 3 t _ _ _ 21 i 80ULO SL0fti DROP AFTER REFERENCE '"
E 12 2000 - POINT 5. REVERSE HE AT TR ANSFE0 REGION I- WOULO BE INDICATED. y - 1800 - - 5FAS j$ r I Il2ER EQPilES. , 1400 - SU8 COMED 3 LEG 15 SATURATED. o REGION g
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I 2 0.04 FT COLO LEG BREAR { REFERENCE TluE P0lNTS (SECONDS) RENARKS 1 0 LOCA CCCURS SITH REACTOR INITIALLV AT I EQUIVALENT TO 2.7 INCH DIANETER HOLE. 1-2 0-15 RCS PRESSURE AND PRESSURilER LEVEL OR01 0F RE ACTOR COOL ANT. 2 15 PRES $URE DROPS TO RPS SETPOINT; THE RE. 2-3 15 23 RCS PRESSURE COMithutS TO DROP AND SUBI l$ LOST. RC PUuPS ARE TRIPPED AND AFD 3 23 SFAS ACTUATED CN LOW RC PRESSURE. SG ' INCREASED TO 93" ON SU RANGE. 34 23-30 THE PRESSURilER EWPTIES AND THE HOT LE-APPROACH SATURATION. 4 30 NOT LEG SATUR ATES. 4-5 30 340 HEAT CONTINUES TO BE TRANSPORTED FRC4 ' BY Tf0-PH ASE NATUR AL CIRCUL Afl0N. RC2 TEMPERATURE ARE SL0 SLY APPROACHING THE SATURATION CON 0lil0NS. 5 340 THE FLOW CIRCUL Ail 0N PH ASE EN0;; THE 5 CONDENSER COOLING MODE. 6 >340 THE RCS STA81LilES AT OR NEAR SECONDAR' THE RCS elLL RENAIN IN THIS CON 0lil0N AND THEN BEGIN A SLCW C00LOCsN AND DED HEAT OROPS. NOTE: OPERATOR ACTION TO C00LDOWN THE SE INITI ATED TO ACHIEVE A LONG ' 2500 2000 -
.?
1500 s 5 1000 500 e n p 0 200 400 f 74-1125531-00
e l Figure F-6 SYSTEM RESPONSE FOR SMALL BREAK WHICA STABILIZES AT SECONDARY SIDE PRESSURE OVER BREAK l$ 2000 POST TRIP OUE TO INE LOSS sinoce r. 2200 - N y CTOR TRIPS. E \ a 7__ e___ _ .b 3 l00 LING NaRGIN 20 0 -
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$ 200 T
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O 01 FT2 8 REAR AT P O? Ol'
\
,' OEFECENCE TluE j POINTS (SECONOS REunCE:
1 0 LOC A OCCURS fl!H NE ACIOR INITI ALLY Al P01 TC 135 INCH Di&uE TER HOLE l2 0-50 RCS PRESSURE AND PRESSURilER LEVEL OROP 0 COOLANT 2 50 PRESSURE ORCPS TO RPS TRIP SETPOINT. THE 2-3 50-80 RCS PRESSURE CONilNUES TO OROP; THE REACT MARGIN IS LOST (PuuP TRIP REQUIRED) 3 80 SFAS ACTUAll0N ON LOW RCS PRESSURE. 34 B0-100 HOT LEG APPROACHES SATURATED CON 0lilCNS A THE RC PUuPS ARE TRIPPED AND AFR 15 ACTUA 4 100 THE HOT LEGS ARE SATURATED 45 100 600 RCS is 14 if0 PHASE NATURAL CIRCULATION. 5 600 THE " FLOW CIRCUL Ail 0N" PHASE STOPS, A LOS SECONDART HEAT TRANSFER CCCURS BECAUSE ST IN SG TUBES IS NOT POSSIBLE 56 600 1250 RCS REPRESSURilES BECAUSE NO CCRE HEAT CA GENERATORS. STE AN BUBBLE IN HOI LEG PIPI' SilE RE ACTOR COOL ANT EIP ANDS INTO THE F REPRESSURilES 6 1500 STEAM CCNDENSATION STARTS IN SG IUBES 6-7 >1500 00lLER CONDENSER COOLING IS ESTABLISH'0; PRES $URE AND TEMPERATURE SLOWL1 DECREASE SATURATION CONDlil0NS 2200 2000
.?
1800 5 g 1600 1400 1200 - 0 1 5 7I4-1125531-00 4
auGr Figure F-7 SYSTEM RESPONSE FOR SMALL BREAR WHICH REPRESSURIZES R, sRiu is t0uivau=f IN A SATURATED STATE 2600 i TO TMt LOSS OF RE ACTOR POST TRIP 2400 - 914000 [ [ ACTOR TRIPS 1200 - l 3 COOL ANT SU8C00t thG , 3 [_ [l
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SUSC00 tid '2 SUP E Rut 4i 3 REGION RIGlom E 14:0 -
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,7 BE REBOVED DY STEAU
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40REat OP[R ATING Point.P091R h CP(Rail 04 iTMOT) 6 II St0tiY GR09lNG .14 E $4IURail0h r7 !$$UR12ER AS IME $f$f[u ' ' ' '* 8 P'
- 600 -
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- $UIC00tE0 400 e4RGIN litt 0 THE PRIDARY SYSitu 9 i i , g ,
J THE SECONDART S!0E 400 42 500 550 600 6 50 700 Reactor Coolant ano $tems Outlet Temposature. F A 3ER- U RE CAD 200 i E
$ 100 0
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500 1000 1500 2000 2500 3000 200 600 1000 I400 1800 2200 2600 3000 Time. sec r .e. see Also Available On I Aperture Caril 84090=011o -/g j
a-I 5 0.07 F1 2 gaEAR WITHOUT AFs REFERENCE TInf REMARR$' PolNTS (SECONOS) 1 0 LOCA OCCURS WliH RE ACTOR INiil ALLY [ IS EQUIVALENT II 3.6 INCH OIAuffER i 12 Q.IO RCS PRESSURE AND PRESSURIZER LEVEL 1 0F REACTOR COOLANT. 2 10 RCS PRESSURE OROPS TO RP TRIP SETPI TRIPS. 23 10 24 AS RCS PRESSURE DROPS. THE REACTOR ( NARGIN 15 LOST (RC PuuP TRIP 15 RE0t ACTUATE 0 ON LOW RCS PRESSURE. 3 24 THE HST LEG SATURATES. 34 24 150 THE RC3 SL0elv OEPRES$URIZES: RESIDOAL FEEDeATER 15 SL0eLY 80lLED 4 150 SG COOLING IS ESSENilALLv LOST. 4-5 150 1200 THE RCS CONTINUES TO SL0eLY DEPRESSL 8 REAR CAN REMOVE DECAY MEAT BY RELEA to THE CONTAlhuENT VESSEL. 5 1200 RCS PRESSURE HAS DROPPED 10 600 PSIE 100 #ATER TO THE RE ACTOR VESSEL. 56 >1200 THE RCS CONilNUES TO DEPRESSURIZE si PROGRESS. 2500 l 2000 2 E 1500 ! E U 1000 500 - 0 i O 500
, 74-1125531-00
-4 f-
Figure F-8 SYSTEM RESPONSE FOR SMALL BREAK WHICH DEPRESSURIZES THE RCS WITHOUT FEEDWATER 2500 POS' TRIP 2400 - B18:008 2200 -
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< p 5 LED $UPERMEAT 60P OJE TO LOSS 2- RE GION O 1000 -
NY: i:.E REACTOR I i/ f 00L ANT $UBC00 LEO g 0 - o STEAM PRESSURE AND 7 RE03 A 3 $FAs is , 3 1400 - iluPERATURE SOULO $L0f tf g
- ~
OROP AFTER REFERENCE PolNT4 [ E 1200 - STEns PRES $uRE / M[ , thC Polhi POST TRIP sitH (j F0wtEO CIRCut afl0h . T n0T OFF. E
- Lluif 4 COLO 440 FOR mafuRAL 1000 f CiRCULAfloh iT COL D '
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- fMOT' E L SAfunAft0h F7 i;;E CFT BEClk! TO -
E40 PolNT-POST TRIP titw 600 L '_jnatuRAL CIRCUL Afl04 e f u MOT 8 6
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0 m i i i a 400 4 50 500 5 50 60 0 6 50 100 2eactor Coolant ano steam Outlet Temperature. F T A3 EFU E CAD t i i i i i 1000 1520 2000 2500 3000
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PRESSURE 3 60 HOT LEG SATURATES 3-4 60-250 THE RCS STA81LilES IN PRESSURE AND THE S B0il CRY 4 250 SG COOLING 15 ESSENTI ALLY LOST 5 >250 THE RCS MANGS UP IN PRESSURE AS THE BREA CONithutti Y OEPRESSURIZE THE SYSTEM TH HPl COOLING THE PL ANT IILL SLORLY COOL DECAY MEAT DROPS 2500 2000
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Figure F-9 SYSTEM RESPONSE FOR SMALL BREAK THAT STABILIZES p AT HIGH RCS PRESSURE WITHOUT FEEDWATER WER. BREAK l$ EQUlv& LENT 2500 3UE TO LOSS OF RE ACTOR POST f#iP 2'00 ~ 914009
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=
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Fi gu rs F- 10 SYSTEM RESPONSE FOR SMALL BREAK , ,in Gi THAT REPRESSURIZES IF FEEDWATER 8 IS LOST 2600 40 f f $ffA5 PRE 1500f & T[uPIRifyR[ O 5 83ULO ORCP Af f[R $G $ 2409 - ORIES OUT Sftan Ptf $( z 8 0 REC 0vtR tHtN Are Af PCe[R gatan it (Culv4Lihi I RESTORIO 2200 - r-Y e .. __-'$ 3 I OROP Out f 0 tnt LO55 CF RE ACICa 2000 - S Plantaf
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RCs PRES 5URE AFR STARTS AND CONT 3 185 THE MOT LEG SaiyRATEs 3-4 115 490 RCS 15 in TRO-FMatt hafuRAL CIRCUS C0hDITIONS DECREASE ALONG THE Saft 1200 PSI. PRE 550RilER LEVEL $ ARE DEPLEft0 4 490 f>E PRE 55URilER FILLS 45 >490 Sf5TEu sfABit ZE5 AT ABout 1200 P5 CIRCUL AflC415 NEVER LOST HPl R_ TO A $LBC00 LED 51 ATE fifM00f ist C00LDCR4 ROULO BE REQulRED t i , 74-1125531-00 A
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Figure F-12 POST LOCA CORRECTIVE ACTION FOR LACK OF ADEOUATE SUBC00 LEO IIARGIN tata ce Acggyett t#ECTfD M57 LOCA P T CDPCITION RIT*e SuSTAtto u ryRatIr.pe & PeliuRY 70 SECOPN HAT , SAKIIL10 emRcIN t"'Wl8 1886 e re.. -
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- 2. SEE FIGURE F.6 FOR A TYPICAL LOCA ERAWLE.
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Table F-1 GENERAL POST-LOCA CORRECTIVE ACTION TO MAINTAIN CORE COOLING LOCA m3 IM0!afE 4C713s5 v!?tL $r$fEu Status C4Cm3 tr toeITFDG
- 1. EVIER PLAsef SveFTGs5 v!A P.T ESPGeX ase OTitR Demissef PLAsef vaAIAa.ES
- 2. ICENTIFY THa7 A LOCA IS Ds PR0rJIES$
.i COUECTIW ACTIGI6 70 ASOWL TRAN53ENT PLANT SitFTOsl LACM OF a0E3aaTE LACK OF PRDenRY TO 700 94JCH PRDenRY To earrn Ds enamGDe ECGCAAT MT TRANSFEA SEC3CAAT MT TbN5FEA ATTDFT TO LOCATE & ISCLATE TK LOCA ISOLATED Luta O!G LOCA NDN.!$3Afang
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l l 74-1125531-00 k
1 l Table F-2 HOW TO DISTINGUISH LOCA'S 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 subcooled state within 5-10 minutes after actuation). ! l e Containment radiation (only for breaks in containment) Note: A steam or feed line leak inside containment will cause high pressure, temperature and humidity but will not cause high radiation. l e Steam pressure, feed flow and steam generator level do not indicate overcooling ! (this helps to differentiate LOCA's from overcooling transients). o High steam line radiation alarms (tube leaks only). e Low makeup 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 OR OUTSIDE. 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 OTHERWISE. THE GENERAL ACTIONS INCLUDE HPI ACTUATION RC PUMP l TRIP, AND C00LD0WN TO COLD CONDITIONS. 74-11'26531-00 " l l
Table F-3
SUMMARY
OF GENERAL LOCA SYMPTOMS A. INITIAL P-T DIAGRAM CHARACTERISTICS rios
/
":,P % cc..m
- sustcotts 'aet a. Rapid system depressuriza-5* isee 33,,,,,,,
tion to saturated reactor Msion n g,,, 3
- y, coolant conditions.
! i .=
a 8 b. Sustained saturation (i.e.,
] Un HPI does not return the
}sM nessun two niet.ecst 1:4 eife scoct reactor coolant to a sub-i ine :.___.___\._... atec Q cascusfie* (f.orros i r emainaat
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cinCaarincooled ts, a.,a state within 5-10 j " 5' 3 g .. . , #3 0 seisate mosset < t,,,t>cPttafION F014f PCett
- i. tutttSLES - 9 gne Petut Post 's p site 888 '*
Eatin tint L natuant CIACuta iCm ten 0 T) deg . 500 554 lag 550 M8 Besster restant b . steam Butlet tesseratere f B. POS$1BLE EVENTS OR OTHER PLANT LEVEL SYMPTOMS (EXTRACTED FROM LOCA CONCEPTS SECTION) SMALL SMALL BREAK LARGE LEAK Small Large BREAK
- 1. Excessive Phkeup X X - -
- 2. Decreasing Pressurizer Level and Pressure *** X X X X
- 3. Reactor Trip X X X X X X X 4 SFAS Actuation (Low RC Pressure) ,
- 5. Loss of Subcooled Margin t X X
- 6. Lack of Primary to Secondary Heat Transfer X (System Repressurization along Saturation Curve)
- 7. Reverse Primary to Secondary Heat Transfer X X
- 8. Rapid Depletion of CFTs X 9 Rapid Drop in RCS Pressure to where LPI X System becomes Operative X X
- 10. gg Increase in Containment Pressure and Tempera-
- 11. Increasing Containment Radiation Levels
- X X X X
- 12. Inadequate Core Cooling Symptoms ** X
*Dagraded containment conditions can occur for other events such as steam or feedwater line breaks inside containment. These non-LOCA events would not have high containment radiation levels. High containment radiation levels are thus a good indicator that a LOCA is in progress.
**For large breaks. the core exit thermocouples can indicate superheated coolant conditions fromThis approximately the end of blowdown to up to 10 minutes into the reflood portion of the event.
is an expected condition. Because the RC pressure is reduced to where the LP! system is fully operational, actions specified for ICC are not required. The core's temperature excursion will be ter,nitiated when CFT and HPI watar refloods the reactor vessel.
*** Break in pressurizer space will cause level to increase.
I 75-112553'-40 ,
Taels F-4 STWT05$ FC] LOCAs THAT CAN BE LOCATE 0 OR ISOLATE 0 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 ifne 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 4 3l2 (Symptoms or alanns most likely to show location are underlined) (Symptoms or alarms most likely to show location are underlined) kl Failure Locating symptoms 16 Isolating hardware O Makeup and purification - Low makeup tank level Letdown valve up-system outside contain- Steam generator tube (s) - High steam line radiation
- High CCW radiation stream of coolers (,)
ment and letdown cool- - High CCW surge tank level - High steam generator level ers (For breaks in letdown Pressurizer safety valves - Reifef line flow alarm cooler) - High quench tank level *
- Local sump levels, radia- - High quench tank temperature tion alarms Seal return line and - Low makeup tank level seal return cooler - High CCW radiation Sealreturnio}
lation valve a HPI injection line break outside containment - High CCW surge tank level - Flow imbalance between injection lines TFor breaks in seal re- (High flow will be through broken line) turn cooler) RC pump seal failure - High seal return temperature (s350 F)
- Local sump levels, radia-tion alanns Combined with:
Low stage and upper stage pressures are Pilot operated equal and high
- Relief line flow alarm PORV isolation valve RCS instrumentation lines relief valve - High quench tank level *
- High quench tank tempera- - Pressurizer level - False high or low level reading ture - Pressures - Talse low pressure
- RC flow - False high or low flow compared with known pienp operation Makeup-letdown imbal- - High makeup tank level F
ance (this is not a Letdown c9ntrol Footnotes: (a) If makeup tank drains, assure operating makeup
- Clean waste receiver tank valve (aJ pump takes suction from BWST.
break, but is a loss - Makeup flow rate (+) seal (b) Loss of decay heat rennval 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 line valve (b) ment (decay heat re- - 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 leak (decay heat removal Cooler isolation
- Righ CCW radiation valves system in operation -
plant is cooled down)
*A111 only be good when the quench tank rupture disk is good.
74-1125531-00
u e i Table F-5
SUMMARY
OF LONG TERM COOLING / ACTIONS r ACTIONS TO MAINTAIN C0_RE_ COOLING
- 1. ECCS Alignments
- a. For saturated reactor coolant conditions (RCS pressure < 150 psig):
e When BWST reaches low level limit, transfer suction of LPI pumps to containment sump. e If LPI flow in each train is > 1000 gpm for 20 minutes, HPI can be stopped. If LPI flow in each train can be maintained > 1000 gpm with one LPI pump, the other LPI pump can be stopped to preserve it for future use. e If neither or only one LPI train flow is > 1000 gpm, run two HPI pumps in the piggyback mode.
- b. For subcooled reactor coolant conditions (RCS temperature < 280F):
e If both LPI pumps are operative, place one LPI train in normal decay heat removal mode. Cool down to 100F with decay heat coolers. e If only one LPI pump is operative, maintain SG cooling. Control RCS inventory using HPI with suction from BWST or in piggyback mode. Start normal decay heat removal when second LPI pump becomes available. l 7.1-11e3531-oo {
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