ML20096C369
| ML20096C369 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/31/1984 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20096C360 | List: |
| References | |
| 51-1121943-01, 51-1121943-1, NUDOCS 8409050250 | |
| Download: ML20096C369 (32) | |
Text
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51-1121943-01 AUGUST 1984 SAFETY PARAETER DISPLAY SYSTEM SAFETY ANALYSIS FOR ' DIE CRYSTAL RIVER NUQ. EAR GENERATION STATION UNIT 3
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FOR .
FLORIDA F0WER CDRPORATION BY The Babcock & Wilcox Company Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 8409050250 840830 DR ADOCK 05000
1 GNTENTS PAG.E EXE QJT I V E SUNY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l 1.0 I N TRODUCT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -1 1.1. Pu r po se . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2. B ack gr ou n d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 SPDS FUNCTIONS / PARA E TERS M0NITORED........................ 2-1 2.1. React i v i ty Coritrol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
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2.2. Reactor Core Cooling and Heat Removal f rom th e Pr i mary Sy stem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.3. Reactor Coolant System Integrity..................... 2-4 2.4. Rad i oact i v i ty Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.5. Containment Conditions............................... 2-5 3.0 APPL I CAB L E EV ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3 .1. I ntrod u ct i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -1 3.2. Excessi ve Feedw ater Event. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3 .3. Loss of Main Feedwater Event......................... 3-4 3.4. Steam Generator Tube Rupture ( SGTR) Event. . . . . . . . . . . . 3-5 3.5. Loss of Offsite Power (L00P).......;................. 3-6 3.6. Sm a l i Ste am L ea k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -7 3 .7. Loss of Coolant Accident...-.......................... 3-8 4.0 CON Q. U S l 0N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 5.0 RE F ER EN CE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 1
L!ST OF TAal_FM
. TABLE EAGE 2-1 Parameters Required to Monitor the Five SPDS l Safety Functions........................................... 2-6 l 1 l
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EXEWTIVE StateMtr s
The purpose of this document is to provide a written safety analysis for the Crystal River Unit 3 (CR-3) SPDS. This analysis describes the basis on which the selected parameters are suf ficier.t to assess the safety status of the plant with respect to fIve required f unctions of the SPDS for a wide range of events. This analysis is in response to a requirement for such a document in Section 4.2 of MJREG-0737, Supplement 1, " Requirements for Emergency Response Capability" (Generic 1.etter 82-33), dated ,
December 17, 1982.
The CR-3 SPDS consists of two color video monitors (CRTs) with associated control panel s. This system, located in the control room, allows the control room personnel to select fra a set of pre-programmed displays and also to select certain Information for display depending on the status of the plant. The displays, their format, and the parameters monitored are based on two major inputs: (1) NRC requirements for SPDS functions and (2) compatibility to ATOG concepts and requirements.
This document summarizes the NRC requirements for SPDS f unctions and how the CR-3 SPDS can be used to assess those functions. " Alert" signal s
- to control ro..w personnel alla them to assess the reactivity control, radi oactiv ity control and containment conditions functions. Pressure-temperature information on P-T displays alIows control room personnel to monitor for abnormal symptoms regarding subcooling and heat transfer for the reactor core cooling and heat transfer fra the primary system function. This same pressure-temperature Information allows the control roan personnel to monitor against key pressure-temperature limits for the reactor coolant system Integrity function.
The events analyzed for the ATOG program are the basis for the CR-3 SPDS.
They include a wide range of events of low to moderate frequency of occurr-ence. They are representative events which provide all the necessary symptoms for which the SPDS and the ATOG are designed to monitor and control. The key parameters chosen for the CR-3 SPDS to meet the above II e + , - , ,-,- - -,- r .,,---,,,,-.,--,----,,-%,y,--,.r_,.,, ,,,,,v,.,,m,-w, ,-, ,-..,-,-w-.---,,--,----,, --w,,w--,-r----
requirements are summarized in Table 2.1. of this report. Based on the Information provided in this report, it.Is demonstrated that the CR-3 SPDS provides control roon personnel with sufficient Information to enable them to determine the safety status of the plant for a wide range of abnormal and emergency conditions, in addition, it is demonstrated that the CR-3 SPDS provides sufficient information to be used in conjunction with ATOG-type procedures to detect abnormal symptons and to allow corrective actions' necessary to restore the control function or mitigate the consequenc-es of transients and accidents in a rapid and rol table manner.
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1.0 INm(EUCTION
- 1.1. Purname The purpose of this report is to provide a written safety analysis for the Crystal River Unit 3 '(CR-3) Safety Paraneter Displ ay System (SPDS) for describing the basis on which the selected parameters are suf ficient to assess the safety status of the plant for a wide range of events which include symptoms of severe accidents.
This report will summarize the NRC requirements for an SPDS and will demonstrate that the CR-3 SPDS meets those NRC requirements. In addition to these NRC requirements, another basis for the CR-3 SPDS is the B&W Owners Group Abnormal Transient Operating Guidelines (ATOG) program.
This report will stamarize that program, particularly the events analyzed, and demonstrate the capatibility of the CR-3 SPDS with ATOG. Finally, this report will identify the parameters selected for the CR-3 SPDS and the method of their presentation io control room personnel.
This report is in response to the requirement for such a document as contained in Section 4.2 of NUREG-0737, Supplement 1, " Req ui rements for Emergency Response Capabil ity" (Generic Letter No. d2-33), dated December 17, 1982.
1 1.7. Background As a result of the TMI-2 accident, the NRC issued an action pl an for items to be addressed in order to correct or improve the regulation and operation of nucl ear facilities. That plan was provided in NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Acci dent", dated
- May 1980. In the area of operational safety, it was concluded in every major study of the accident that insufficient attention had been given
- to ensuring compatibility between control roon personnel and the systems they are required to operate. .The variety and quantity of information displayed in the control room can of ten be overwhelming, especially during l transient operations or an accident. It has been determined that a concise di spl ay of those parameters necessary to assess the safety status of 1-1 l
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- --** -r-S'y e -e w-v a-ww,w-+% ,y--- - -,---wwww-r,-w- ww y % --,emw-,,-y, e m ysempe , w = w - r r w we w--o , e -- -=- v, e ve,w m -e - ww w w-== w w w -sp en e y- w,r yuse -r + wr e tsure e er ar- w e- or--- m
the plant would significantly aid control room personnel in determining plant status and diagnosing accidents. item 2 of Task I.D of NUREG-0660 established the requirements for a plant safety parameter display console to be installed in the control room.
In November 1900, the NRC issued NUREG-0737, " Clarification of TMI Action Plan Requirements." Concerning the requirements for a plant safety parameter display console, the NRC stated that issuance of NUREG-0696, " Functional Criteria for Emergency Response Facilities" would provide those requirements for an SPDS. NUREG-0696 was issued in February 1981. In December 1982, NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability (Generic Letter No. 82-33)" was issued. NUREG-0737, Supplement 1, describes the NRC requirements for SPDS. The CR-3 SPDS will be designed to those requirements. Section 5 of NUREG-0696 provides guidance for the design of the SFDS Including its purpose, location, size, display considerations and design criteria, such as conformance to Regulatory Guide 1.97 for
- the parameters chosen. These guidelines serve as a basis for the design requirements for the CR-3 SPDS.
The purpose of the CR-3 SPDS is to assist the control room personnel during abnormal and emergency conditions in the evaluation of the safety status of the plant and assessing whether abnormal conditions warrant corrective actions, to avoid a degraded core. _ Located in the control room, the SPDS will be continuously available during all modes of plant
, operation including cold or refueling shutdown, heatup and cooldown opera-tions, normal power operations, as welI as abnormal or emergency conditions.
The CR-3 SPDS video monitors are not seismically qualified and thus, the SPDS may not be available after certain seismic events. Plant parameters which Indicate very abnomal conditions such as inadequate core cooling (ICC) shall also be displayed. The CR-3 SPDS is designed to use a minimum number of displays and selected parameters, yet it will be able to concisely present to the operator information concerning the safety status of the folIowing functions as rt, quired by f(JREG-0737 and NUREG-06%:
F 1.-2 t
(1) Reactivity control (2) Reactor core cooling and heat removal fran the primary system (3) Reactor coolant system Integrity (4) Radioactivity control (5) Containment conditions.
Finally, the CR-3 SPDS is designed to be compatible with the control roan personnel's training and experience in normal and abnormal or emergency operati'on. Florida Power Corporation (FPC) is a member of the B&W Owners Group Operator Support Subecomittee which has worked to develop a program to respond to item I.C.1 of NUREG-0737, " Guidance for the Evaluation and Development' of Proeddures for Transients and Accidents". That program resul ted in the development of the Abnormal Transient Operating Guidelines (ATOG) which employs the concept of symptem-oriented rather than event-orient-ed procedures for handling abnormal or emergency conditions. The design of the displays for the CR-3 SPDS has incorporated the basic concepts f
of ATOG in order to present suf ficient information to control room personnel i
for monitoring the status of the plant during normal, transient and abnormal conditions, diagnosing symptoms and taking corrective actions to achieve stable plant conditions.
The CR-3 SPDS is designed to serve as an aid to the control room personnel and can - be used by them to quickly focus on certain key parameters.
However, the SPDS should not be used as the exclusive source of information to determine p! ant status or to monitor control actions. Other instrumenta-tion avaiIabte in the control room should also be used by controf room personnel to determine piant status.
The CR-3 SPDS will be located in the control room. The CR-3 SPDS wil l provide information to control room personnel via selectable di s pl ay s and will automatically di spl ay alert signals. The selectable displays include ihe Low-Range Pressuro-Temperature (P-T) displ ay, ATOG P-T displ ey, inadequate Core Cooling (ICC) display, two " normal" power operation displays and a four page alphanumeric display of several key parameters and safety systems actuation status. Control room personnel can select any available 1-3
. . - . . . . ri,-w..m n . - _ . . . , _ . . , .__,,,,.-.,_.-.,,._.....-,..--,,,,,-......,_.....--__..-e,#.--,. - .,- .-.--,,.e.,
di spl ay by a single pushbutton on the SPDS mntrol panel. If control rom personnel have selected either the Low-Range P-T di spl ay or the ATOG P-T displays, they can add any of several NT limit curves to the 1 base display by a single pushbutton for each limit curve. This feature allows the control room personnel to be able to monitor the plant status !
versus those P-T limits applicable to the current plant conditions or l
expected plant evolutions such as plant cooldown to cold shutdown cond!- I tions. ' Each control panel also contains toggle switches which al l ows control room personnel to select either Loop A or Loop 8 parameters for display on either CRT and to select spare parameter inputs. There is a feature to the CR-I SPDS which allows the control rom personnel to monitor the history of the RCS pressure-temperature relationshi p. Using a single pushbutton, control room personnel can demand or erase the history trace of the RCS pressure-temperature relationship. Finally, control room personnel can select the incore thermocouple temperature (average of the fIve highest thermocouples) to be displayed instead of T-hot on the two P-T displays.
The two " normal" power operation displays, " Normal" and " Flux / Imbalance" can be used by the control rom personnel during normal operations to monitor plant conditions against the Reactor Protection System (RFS)
P-T and Flux / Flow / imbalance limits. The Low-Range P-T and ATOG P-T displays can be monitored by control rom personnel during heatup, cooldown and normal operations and also during transients and accidents. The ATOG P-T dIspisy wiii be automatically displayed for each RCS Ioop upon roceipt of a reactor trip signal. This feature prevents control room personnel from having to manually switch d these displays in the first several seconds following the reactor trip, freeing them to perform their imediate post-tri p actions. The ICC displ ay can be monitored by control rom personnel during conditions of RCS saturation or superheat. The six " alert" signals will automatical ly f l ash on any selected display if conditions i warrant. The four page Alphanumeric display can be used by control rom personnel to obtain additional supporting information to the above displays and alerts.
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A detailed description of the CR-3 SPDS hardware, displays, alerts, and
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Input signals can be found in the document "CR-3 Safety Parameter Display System Functional Description", Dwg. #1147047. Algorithms for the display curve limits can be found in the document, " Crystal River Unit 3 Safety Parameter Display System (SPDS) Displays", Doc. ID #51-1121942, dated June 1984.
The SPDS displays have been Human Factors engineered to the extent that they present the necessary information to control room personnel in a manner which would allow them to carry out the required actions for abnormal transient operations described in ATOG. This includes displays which have a simple, uncluttered fonnat that can be called up by a single push button command. The displays are continuously updated and allow control room personnel to track trends for key variables. Additional information is presented to control roca personnel in order for them to assess the status of' the plant with respect to the five critical safety functions.
This information is presented in the form of Alerts in order that they not interfere to a significant extent with basic pressure-temperature displays required by ATOG. Further explanation cf these features is found in Section 2.0.
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2.0 SPDS FUNCTIONS /PARADETUt$ IOllTORED Section 1.2 describes the purpose of the SPDS. In order to meet that purpose, the SPDS must be designed to provide information to control rom personnel for all modes of normal operation fra refueling to power operation. It must also be designed to provide key Information to control rom personnel during abnormal and emergency conditions so that they can assess the safety status of the plant. In particular, it must provide sufficient Information to allow control room personnel to assess the safety status of the plant with respect to five major functions.
Each of the f ive safety status f unctions identified is Section 1.2 is discussed in detail below. Each section will include a brief description of what the safety function is followed by a discussion of how inf ormation concerning the safety function is provided to control com personnel by the CR-3 SPDS. In addition to the specific displays and alerts which are described in these discussions, it is important to note that a four-page 4
Alphanumeric display is available to provide support Information to the other displays and alerts.
The complete list of parameters required to monitor the five SPDS safety functions is provided in Tabl e 2.1. Also included in this table are the signal input string, range, the function it monitors, the means of presentation (display or alert signal) pl us a summary discussion. In addition to these parameters, there are many additional parameters available
! on the Alphanumeric display to aid control roon personnel in controlling the plant.
2.1. Rametiv ltv Contr ol l The first function is reactivity control . Fol l ow ing a reactor trip, or when conditions warrant a reactor trip, control room personnel need l l to know that the control rods have all inserted and that the reactor is subcritical. If not, they must begin taking required actions to achieve reactor subcriticalIty.
2-1 I
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On the CR-3 SPDS, the reactivity control function is monitored by a "Reactiv-Ity" alert logic. The parameters monitored for this alert are control rod group in-limit position, source range nuclear instrumentation flux level and reactor trip signal. These parameters are sufficient f or monitoring the reactivity control function since they provide the necessary Information to control rom personnel that control rods have Inserted upon reactor trip and that the reactor is subcritical at al l times other than when the reactor is already critical or where a planned reactor startup is underway.
, 2.2. Panetor- Cora Cool Ing and Heat Ramnval f rm the Primmev System
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The second function is reactor core cooling and heat removal from the primary system. This function will be discussed as two separate functions.
During all modes of operation and following reactor trip, control room personnel must have information to determine the thermal-hydraulic state of the reactor coolant. If the reactor coolant is in a liquid state and subcooled, control room personnel are assured that reactor coolant is available and capable of removing heat from the reactor core and transferr-Ing it to the steam generators. If subcooling is lost, these capabilities are in doubt and control rom personnel can begin taking actions to restore subcoolIng.
On the CR-3 SPDS, the reactor core cooling f unction is monitored in four ways. During normal operation, this f unction is monitored by the display of the RCS pressure-temperature status points with respect to the RPS l
pressure-tem perature limits which Inherently insures that the RCS is l'n a subcooled state since tiie RCS would have to pass through these trip paraneters before subcooled margin is lost. During post-trip operations, this f unction is monitored ~by the di spl ay of the RCS P-T status with respect to " expected" post-trip pressure and temperature Iimits as welI as the saturation line and the subcooled margin limit which accounts for possible errors for the pressure and temperature instrument strings.
, This Information is available to control room personnel on the ATOG Display.
During inadequate core cooling operations, this function is monitored
-2 2
j l by the display of the . RCS P-T status with respect to saturation and two fuel cladding temperature limits. This information is available to i control room personnel on the ICC Display. During cooldown operations, this function is monitored on the Low-Range P-T Displ ay in iIke manner i to the post-trip operations discussed above.
I in each of these situations, the parameters to be monitored are RCS pressure, l RCS hot l eg temperature (T-hot), RCS cold leg temperature (T-cold), and !
, everage incore thermocouple temperature. Wh'en displayed against saturation j l and subcooled margin, these are the only parameters that are r.ecesasry to determine the state of the RCS for monitoring the reactor core cooling f unction in these four situations.
In addition to reactor core cooling, heat removal from the primary system must be monitored. Control roca personnel must have Information which al l ow s them to monitor the heat transfer coupling between the reactor coolant syster. and the steam generators. Control room personnel can then begin taking corrective actions to restore that coupIIng l'f inadequate heat transfer is occurring. If excessive heat transf er is occurring,
-control room personnel can begin taking actions to restore the balance l
between the heat source (reactor core) and heat sink (steam generators).
On the CR-3 SPDS, the heat removal from the primary system function Is
- monitored by the ATOG P-T Display and the "EFW Alert". The RCS pressure-temp-erature status is displayed as a point and the steam generator saturation i temperature based on its pressure is displayed as a vertical line.
Five parameters (RCS pressure, T-hot, T-col d, everage incore thermocouple temperature and steam generator pressure), provide the symptans of inadequate or excessive heat transfer and allow verification of natural circul ation by control roan personnel. The CR-3 SPDS also provldes bar charts for steam generator startup and operate range level Indication to aid in a more rapid recognition of the problem and the "EFW Alert" which warns control roem personnel that a signal from the Emergency Feedwater instrument-atton and Control (EFIC) system is present to initiate EFW.
2-3 l
l - _ _ _ - _ . _ . _ _ _ _ _ _ _ . . _ _ _ _______
i 2.3. n W or Pnniant Svaten Intaaritv During all modes The third function is reactor coolant system Integrity.
of normal operation and during transients and accidents, control room personnel must be able to monitor the status of the reactor coolant system pressure-temperature relationship against several important pressure
- and pressure-temperature I imits to ensure the Integrity of the reactor coolant system against overpressurization, pressurized thermal shock or reactor vessel NOT. If Iimits are being approached or exceeded, control rom personnel should begin taking actions required to re-establish an acceptable state. Also, there are a number of pressure, temperature, and pressure-temperature' limits important to normal operation. Control ,
i roca personnel must have the necessary information to determine the status of the plant with respect to these limits and control the plant within these limits.
1 On the CR-3 SPDS, the reactor coolant system Integrity function is monitored using the following parameters: RCS pressure, T-hot and T-col d. These l
parameters are displayed as P-T status points against the following applic-able RCS P-T or pressure Iimits: RCS design pressure, DHRS design pressure, thermal shock Ilmit, heatup NDT and cooldown NDT limits. Operation of the plant within these limits will satisfy the reactor coolant system integrity function.
2.d. Radfanctivftv Control The fourth function is radioactivity control. During all modes of normal operation and especially during accident conditions, control rom personnel must be warned of the presence of high' radioactivity in the plant and monitor release paths to determine the extent of radioactivity releases f rom the plant. They can then take necessary actions to prevent or minimize such releases.
On the CR-3 SPDS, the radioactiv ity control function is monitored by
- the "Radioectivity Alert". The parameters to be monitored by this alert are the fourteen (14) radiation monitors identified in Table 2.1. These 2-4 !
i monitors were chosen because they monitor the major radiation release paths fra the plant or because they provide specific Information to aid control room personnel in determining the event or identifying the location of the radiation leak. More specific Information on these radiation monitors is found in the " Discussion" column of Table 2.1.
2.5. (bntainment conditlana The f inal function is containment conditions. The last fission product
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barrier to the environment is the reactor building. Control rom personnel must know the status of the reactor building in order to take necessary actions to ensure or restore its integrity during accident conditions.
On the CR-3 SPDS, the containment conditions f unction is monitored by three sets of alert logic. They are " Reactor Building Pressure Alert",
"Radi ation Al ert", and " Engineered Saf eguards Actuation Alert". Reactor building pressure and radiation are the two parameters which give the first Indication of a major accident, such as a LOCA or steam line break, inside containment. The " Reactor Building Pressure Alert" warns of
! Increasing pressure in the RB le advance of M isolation. The " Radiation Alert" contains inputs from thren radiation monitors located in the RS dome, M purge duct and RB vent duct. If the " Radiation Alert" is present, control room personnel can request the applicable page of the Alphanumeric display to determine if one of the R8 monitors is causing the alert.
If the " Engineered Safeguards Actuation Alert" is present, control room personnel can page the Alphanumeric display to determine if the RB isolation i signal is causing the alert.
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TABLE 2.1 l
PARAMETERS REQUIRED TO MONITOR THE FIVE SPDS SAFETY FUNCTIONS DISPLAY OR I ___,fARAMETER SIGNAL INPUT RANjE EUNCTION*** ALERT *8 ** BISCU1110N
- 1. T-Cald A&B NR RC54-TT2,-TT4 50 to 650*F 1.2, 1.3 2.1, 2.2, 2.4, Plotted and displayed with RCS pressure
, RC59-TT2,-TT4 2.9, 2.11 on P-T displays to monitor plant
' status against NDT, thereal shock, 4
fuel compression and RCI-NPSH limits, and to provide symptoes of excessive or inadequate heat transfer when j
used with S6 saturation toeparature.
l i See discussion of T-Cold A&B., in
! 2. T-Hat ALB NR RC4A-TTI -TT4 120 t o 920*F l.2, 1.3 2.1, 2.2, 2.4, i RC49-TTI,-TT4 2.11 addition, it is displayed against subcooling margin and saturation lients for symptoms of inadequate d core cooling. It is also used with incore Thereoccuple Temperature as an indication of natural circul-ation.
) 3. RCS P essure A&B NR RC3A-PT3, RC-158PT 0 to 2500 psig 1.2, 1.3 2.1, 2.2, 2.3, See discussion of T-Cold A & B.
RC39-PT3, RC-159PT 2.4, 2.9, 2.11 Also plotted with incore Thereo-l
' couple Temperature to monitor natural ciruciation or superheated conditions during ICC.
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- 4. RCS Pressure A&B RC-131-PT, O to 500 pssg 1.2, 1.3 2.1, 2.4, 2.9 Used as RCS Pressure signal during l low RCS Pressure operation, j Lou Range RC-131-PTI, O to 600 psig i RC-147PT, RC-14Bril 2-6 i
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t TABLE 2.1 (CONT'D)
DISFLAv OR
,, t ahint it r. a l utNL I NI d l b in NbE, , , , I UNC,1 tjeN,' ', * *, Al[h,l e s .,e e J.lSCUSSIDW
- 5. In core Thersocouple 19-aEH.-"6F 9- 2v00*f 1.2 2.1, 2.2, 2.3 Average value plotted with RCS Tracerature scalcu- -ofM. -l;F. -n96, Pressure on ICC curve during ICC lated average os 5 -120. -nJE, -65t. conditions and can be plotted with hegtest readings f roe -97M,-v5D.-n3F RCS Pressure on other P-T curves.
- 13 T/Est -950 ISee discussion of RCS Pressure, WRI.
- 6. OTSG Op. Range SFIA-LAM I/2 o to lool I.2, 1.3 2.1, 2.2, 2.3 Bisplayed on P-T displays primarily SPib-LAM is2 2.4 for namediate detection of estoselve feedsater and loss of feeduater events.
l 7. OTSG Su Range ALB SFI A LI4/5 u to 250 in. 1.2. 1.3 2.1, 2.2, 2.3 See discussion on OTSG Op. Range SPlb-Lites A&B. SU Range say be useful in 56 tube rupture identificatten, i
- 1
- 8. OTSG Pressure Att SP6A-PT 1/2 0 to 1200 psig 1.2 2.2, 2.4 Used as input to SS saturation SP6B-PT t/2 toeperature on Post-Trip ATOS display.
Used to display symptoes of excessive i heat transfer and inadequate heat j transfer uhen used usth T-Cold.
f i 2-7 i
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1 TABLE 2.1 (CONI Di DISFLAv OR
_ _. .P.A.R A.ME I E R , SIGNAL,)NFUT RANG [..._ F ut!L110N " *
- A((gJea.es g[gggggggg_ _
- 9. Sturce Pange NI A%B NI-Ol. NI-02 lu to 10* cps 1.1 2.4, 2.5 used in conjunction with reactor trap signal and CRG positions to alert the control race operator to high or unexpected neutron count rate. .
1
- 10. All Rods IN (Later) No/Ves 1.1 2.5 4
See discussion of source range NI A&B.
J ll. R2ector Trapped R Fatch Fnt hotyes 1.1 2.2, 2.4, 2.5, See discussion of suorce range NI 2.10 A&D, In addition, this signal aise anettates the automatic display a
the Post-Trip ATOG display for loops
. AtB on the two CRis and the RPS actuation alert.
- 12. Radiation Monitors t.4 2.4, 2.6 All the radiation monitors are used to sonator potential eator release paths. In addition, the individual sonitors are used to i a. Rb Dome FMG-19 's to "A le* F /hr a. Help discreannate between major i
LOCA and Steae Line Break.
, b. Rb Purge Duct RMA-1 19 to in* CFM b. Provides early indication of i
vent system radioactivity and possible unplanned release.
j c. Aux, Fuel Handling RMA-2 lu t o 19* CF M c. Saee a ; b.
! Duct
- d. Eb ver t Duct RMA-6 to to 10* CFn d. Provides early indication of RC5 lea 6 age.
2-8
TABLE 2.1 (CONT *DI DISPLAY OR PARAMETER _ %J EMsL_1MEEL RANGE Eyhi]J0Nes.o. ALERT ** *l DISCUSSIDW
- c. Condenser Wacuus EMA-12 to to lo* [PM e. Provides indication of stese Pues Discharge generator tube leakage.
- 6. Main Steam Line E Mb - 2*s la
- 3o IU' mRehr f. Ntil identify affected stese generator when a tube leak to present.
- g. Mann Steae Line FMb-26 lu
- to 16' mRshr g. Saee as f.
- h. Main Steae Line FMG-27 to-' to 10' mRshr h. Saee as f. *
. Main Steam Line EMG-28 10-' t o 10' mRihr t. Saee as f.
! 3. Primary Coolant RML-l 10 to 10* [PM j. Provides ear ly and: cation of fuel cladding failures.
W. Plant Discharge RML-2 10 to 10* CF M h. Provides early indication of abnormalttaes in the liquid effluent discharge flow path.
I. Turbine building EML-7 to to in* (FM l. Same as 6.
- o. [ontrol Ecoe Vent EMA-5 le to 16* Cf M e. Provides and: cation of degrading
. control roce environment prior to actuations occurring.
- n. Control Goon hMG-1 10-8 to lo* mR/hr n. Same as o.
I 13. Esactor Building BS90-P1.bS91-PT o to 200 psia 1.5 2.4,2.7 Used in Reactor Building Pressure alert to warn the operater of high pressure ,
conditions as a result of LOCA or steam leak, i
j 2-9
. IAF C 2.1 ICONT*D) 1 i
DISPLAY OR PARAMQ[R SjgNALjmPU_{ J,ANEL (UN{ T I ON ' 8 *
- E QTee.*' BISCUS$10ll
]
such that 2 out of 3 tripped in j any of the . groups will cause alert signal to uarn the control room operator of ES actuation.
- 15. EFIC Actuation glater) No/Yes 3.2
) 2.4. 2.9 Used in the EFN Actuation Alert l
to warn the operator that the EFIC signal to inst aate EFN is present.
1.0 SF'DS Function 2.0 O t sglav or Alert l 1.1 Reactivity Control 2.8 Lou Rangp P-T Display 1.2 Reactor Core Cooling und Heat Removal f rom Fr emary Systen 2.2 ATOG P-i Display j l.3 Reactor Coolant System Integrity 2. 3 Inadequate Core Cooling flCCI Display 1.4 Radiation Control 2.4 Alphanumeric Display
, 1.5 Containment Conditions 2.5 Reactivity Alert 4
i 2.6 Radiation Alert 2.7 Reactor Building Pressure Alert 2.8 Engineered Safeguards Actuation Alert
] 2.9 Energency Feedsator Actuation Alert i'
2.80 Reactor Trip 1RPS Actuation) Alert 2.11 Normal D splay 2-10 4
i o__ _.
3.0 APPLICABLE EVENTS 3 .1. IntraductIon in part, the CP 3 SPDS is based on ATOG. Control rom personnel must be provided with certain key parameters which they can monitor in conjunction with the ATOG-based emergency operating procedures. The key parameters are those necessary for the Identification of the three basic heat transfer symptoms (lack of subcooling, Inadequate heat tranfer or overheating, and exc'essive heat transfer or overcooling) and the special case of steam generator (SG) tube rupture. Once contr'ol room personnel identify one or more of these symptes, they can begin taking action to regain the control functions which' are not being controlled. The control functions
- associated with these symptoms are RC Inventory, RC pressure, SG Inventory, and SG pressure control.
The key parameters which are necessary for the identification of these four symptcas are also those necessary to assess the safety status of the plant with respect to two of the five required SPDS functions: reactor core cooling and heat transfer fra the primary system and reactor coolant f
- system Integrity which were discussed in Section 2.0. Additional parameters are provided in order to assess the safety status of tho' plant with respect to the remaining three required SPDS functions: reactivity control, radioactivity control and containment conditions. Expected control room personnel actions to verify that all control rods have Inserted and that
! measured neutron flux levels are decreasing are the basis f or. the two parameters necessary for the " Reactivity Alert". Expected control roem personnel actions to monitor Reactor Bullding pressure and radiation monitors are the basis f or the parameters necessary for the " Reactor -
Building Pressure" and " Radiation" alerts. Expected control rom personnel actions to determine an event and to monitor key radiation release paths j
durIng accidont assesment are the bases for the setoct!or of the fourtoen
! (14) Radiation Monitors used in the " Radiation Alert".
I in order to better understand the bases for the parameter selection f or the Ot-3 SPDS, it is important to understand the events that were analyzed 3-1
1 for developing the ATOG guidelines.
The AT0G guidelines have been written to cover a very large number of abnormal transient scenarios. This is an inherent benefit of a symptoo- i oriented approach to operation. However, specific accident events need to be analyzed in-depth to demonstrate that they can be handled using a symptem-oriented approach. These transients can include classic single initiating events as well as additional single or multiple f ailures.
For the ATOG development program, six Initiating events were identified and analyzed in detail. The six initiating events selected were:
- 1. Excessive feedwater
- 3. Steam generator tube rupture (SGTR)
- 4. Loss of off site power (LOOP)
- 5. Small steam leak
- 6. Small break loss of coolant accident (SBLOCA).
i These six events were chosen as representative transients, each of which would provide one or more of the basic symptoms on which ATOG is based.
Additionally, four specific criteria were used in selecting these six events:
- 2. Low probability events that can be confusing and difficult to recognize and mitigate (SGTR, SBLOCA).
- 3. The events cover a very large spectrum of conditions in the RCS (overheating, overcooling, loss of Inventory, loss of subcool-Ing).
- 4. Time exists for the operator to recognize and do something about the accident or event; therefore, guidelines are appropriate.
This would necessarily exclude large, rapid design basis events as such a large break LOCA.
3-2 l
. - . - - . . , - . . . , _ . , _ . . , _ _ _ , - , , . ___,_ ._ ____._._,_r___,,. , , _ _ . _ ,...,_,,,,_,,._-.-.,_,__.__..,m.,______-
For each of the six Initiating events chosen, a detailed event tree was developed to identify a main success path and major f ailure paths. Analyses using the TRAP code were perf ormed on the main success path and most single f ailure paths for the ATOG base plant, ANO-1. The ANO-1 plant design, eq ui pment, pertinent pl ant setpoints, etc. were then compared to CR-3. Transient Inf ormation Documents (TIDs) were then developed for use to modify the ANO-1 ATOG guidel ines to fit CR-3. Those documents applicable to CR-3 are Ref erences 10 through 16. Additional analyses have been performed for the Steam Generator Tubo Rupture event. Those analyses are docunented in Ref erence 17. The LOCA events analyses and recommendations were derived largely from those ef forts, which produced
~
the "Small Break Operating Guidelines".
A brief summary of each of the six initiating events is provided below.
Detailed discussions of these events are found In the Part 11 - Vol . 2 of Ref erence 5 and Reference 17. Those discussions include actual operating events which have occurred.
_3.2. Fve assive Feedw ater- Event The excessive f eedwater erent was analyzed for two cases. For the first case, the excessive feedwater addition was terminated by the ICS after reactor trip. The event was analyzed to >200 seconds at which time the RCS had stabilized in a hot shutdown condition. For the second case, the excessive f eedw ater addition to one SG was terminated by control room personnel approximately 5.5 minutes af ter event initiation. Prior to termination of feedwater, the overfed SG has filled, the pressurizer has emptied due to RCS contraction, subcooling margin has been lost and ESAS has started HPl and EFW. In addition to terminating main feedwater, control roon personnel must make three significant control actions during i
such an event. The first is to trip all operating RCPs on loss of subcooling margin. The second is to throttle EFW to obtain a gradual increase in SG level to prevent worsening the overcooling. The third is to throttle HPl once subcooling margin is restored. It should be noted that once the Emergency Feedwater Initiation and Control (EFIC) system is implemented 3-3
at Ot-3, control room personnel will only have to verify that main feedwater has been terminated and that emergency feedwater ficw is being controlled properly to achieve the desired SG level.
The pressure-temperature relationship response of the primary and secondary systems displayed on the Ot-3 ATOG P-T Display will provide control rom i personnel the symptom of excessive heat transfer. In addition, the two SG level bar charts will Indicate the excessive f eedwater addition.
This information will key control room personnel to take the required actions to terminate main f eedw ater and, to take the three additional control actions.
3.3. t_n== of Maf n Feedsater Evant The second event analyzed was the loss of main feedwater (LOFW) event.
This event was initiated by a trip of both main feedwater pumps f rom 1005 FP with anticipatory reactor trip occurring. In its early stages, this transient will look almost identical to a " normal" reactor tri p.
Control room personnel shoul d be able to identify LOFW by Indication of main feedwater pumps tripped or zero main feedwater flow rate at the feedwater control station, or rapidly falling SG levels on the SPDS display or existing control room Instrumentation. The only signifIcant control i
rom personnel action is to verify that EFW starts and is providing flow
! was also run with a number of failures out to approximately 10 minutes after reactor trip.
The first f ailure is f ailure of EFW to start. Based on the pressure tempere-ture relationship response displayed on the SPDS di s pl ay, control room personnel should recognize the symptem of inadequate primary to secondary heat transf er and take corrective actions to restore heat transf er.
The second f ailure is EFW overfeed. Control room personnel should recognize the sympts of excessive primary to secondary heat transf er and take corrective actions to throttle EFW. As was previously noted, once EFIC has been implemented at CR-3, control room personnel shout d only have 1 3-4 l
- - . . _ , , . . . - - , - , , . _ , . - .--,.~.---n.,- - - . . , . . . , . . - - - - - , , , , - - _ . . ~ - , . _ , , , - - . . - , , - -
to verify that EFW has initiated and Is being controlled to achieve the desired SG level. The third f ailure is steam leakage fra a stuck open SG steam safety valve or an open turbine bypass valve. Control room personnel should recognize the sympts of excessive primary to secondary
^
heat transfer, and by following plant procedures, they shoul d identify the steam leak and take corrective actions.
The pressure-temperature relationship response of the primary and secondary systems displayed on the CR-3 ATOG P-T Ols' play and the two SG level bar
! charts will provide control room personnel the symptom of inadequate heat transfer. This Information should key control room personnel to
', take the required actions.
3.d. St - Generater Tuha Runtura (SGTR) Event F
, A steam generator tube rupture (SGTR) is a loss of coolant accident (LOCA) through the secondary plant. It is an event which can contaminate the secondary plant and can lead to radiation releases if steam from the affected SG is released to the atmosphere. For these reasons, it is important that control rom personnel Identify the event and the af fected SG quickly.
J The analysis performed for ATOG began w(th a SGTR occurring at 100$ FP.
The initial primary to secondary leak rate was for a double-ended rupture of a single tube (approximately 400 gpm), in the first several minutes, control rom personnel will see RCS pressure and pressurizer level decreas-Ing. They wIII al so rocef ya steam iIne and condenser air ejector of f gas radiation alarms. Because the leak rate exceeds the MJ system capabil-Ity, control room personnel must quickly take action to Increase makeup
- flow or manually actuate HP! and terminate letdown flow in order to stabilize
. the RCS. Once the RCS pressure and pressurizer level are stable, the reactor can be shutdown in a controlled . manner. The controlled shutdown of the reactor should prevent lif ting the SG steam saf ety valves. Once ;
the reactor is shutdown, control rom personnel can cooldown and depressurize 3-5 J
i
(
m w --r-,r, ...~---,.r..r.--~~--,e,,-,-m.--,-wer. &,... , , _ _,. ww,--w-,w.
/
the RCS while maintaining minimum subcooled margin in order to minf alze the primary to secondary leak rate.
i However, If control room personnel take no action to increase makeup and because the leak rate exceeds M.! system capability, the reactor will l I
trip on Iow pressure. FoiiowIng the reactor trip, subcooling margin i is lost, _ requiring control room personnel to trip alI operating RCPs.
ESAS will actuate on low RCS pressure starting HPl and EFW, At approximately 11 minutes into the transient, the pressurizer will empty and the RCS wilI become saturated. In the next 4 to 5 minutes, control room personnel will throttle EFW to prevent overcooling. Af ter EFIC Implementation, they wilI verify proper EFW control. They will also throttle HPI to stabilize RCS pressure and pressurizer level following the re-establishment of subcooling margin. Once the plant has been stabilized with core decay heat being removed by the SGs via natural circulation, control rom personnel
- can restart RCPs and initiate plant cooldown and depressurization. During 1 the cooldown, controf room personnel shoutd malntain a minimum subcoolIng margin to keep RCS pressure and the primary to secondary leak rate low.
- Secondary system radiation alarms in conjunction with the pressure-tempere-ture relationship response of the primary and secondary systems plus the SG level bar charts on the CR-3 ATOG P-T Display will provide inf ormation ,
- to control room personnel to Identify the SGTR event and the af fected SG. They can then begin to take required actions to mitigate the consequenc-es of this event.
I 3.'5. Loma of Of f a t ta Power (LOOP) j- The analysis for this event begins at 1005 FP with the plant separation i fra the grid. During the reactor and turbine runback, the reactor trips i on high pressure. The diesel generators start on detection of undervoltage on the engineered safeguard buses and begin loading within approximately 10 seconds. Secondary steam pressure is controlled by the steam safetles
, 3-6 o
d i
~ ,~,,,---,,w---m,,,,---w-~_ m-.wm.-,-.,,,, _ _ , _ ,
_c-w._ ww.--
or atmospheric dump valves. Since the RCPs have tripped, natural circulation
,' will be established and EFW will start feeding to establish l evel in l the SGs. That level is reached approximately 8 minutes af ter grid separation I
occurs, and the RCS should be stabilized at normal hot shutdown conditions.
Af ter EFIC Implementation, the EFW flow will be autmatically controlled to ~ establish natural circulation level setpoint.
s !
Additional analyses were performed for a total loss of all power except ,
j station batteries. For a short period of time, the plant can be controlled at hot ' shutdown using EFW to the SGs to provide a heat sink. If volds begin to form in the RCS and natural circulation is Interrupted, control room personnel should fili the SGs to the required level to estabt Ish" the core bolling /SG condensing mode for decay heat removal. EFAC will
! automatically control SG levels to the LOCA setpoint once control room personnel selects that mode.
Every ef f ort must be made te get the diesel generators started. Since the pressurizar heaters and makeup pumps have lost power, control room personnel have no control of primary pressure or Inventory. The plant
+ cannot be cooled down because waar cannot be added to make up f or RCS
- contraction during cooldown. ', s The pressure-temperature relationship response of the primary and secondary systems displayed on the CR-3 ATOG P.-T Display will provide control room f personnel information on the adequacy of natural circulation and the sympts of inadequate heat transfer if natural circulation is Interrupted.
3.6. hal1 St- Laak -
The event analyzed was a f ailure of the turbine bypass system. This l f ailure results in a steam leak equivalent to approximately 15% of full
! power steam flow and occurs with an assumed maximum decay heat. In this i
event, the reactor trips on high flux in about 9 seconds f ollowing f ull '
opening of the turbine bypass system valves. The turbins trips and ICS
- runs back main feedwater. The turbine bypass valves on one SG remain 3-7 i
j
1 l
open. Approximately three minutes into the transient, the pressurizer drains due to excessive RCS contraction. ESAS Initiates due to low RCS l
pressure and starts HPI and EFW. As RCS pressure and pressurizer level are restored, control roca personnel begin throttling HPl. Approximately 5 minutes into the transient, control rom personnel identify the af fected SG based on symptoms of excessive heat transf er and decreasing steam pressure in the ef fected SG and allow it to boll dry. At approximately 8 minutes, control rom personnel stop HPI, realign to normal makeup / letdown mode and use the atmosphsric dump valves ort the unaf fected SG to stabilize the RCS.
Larger leaks, less decay heat or subsequent f ailures may cause a transient severe enough to cause loss of subcooling margin and possibly RCS satura-tion. Controf room personnel must f of Iow their procedures for ioss of subcooling margin. The stoam Ieak may also be Inside the reactor butIding (R8). If so, it is likely that an ESAS trip on high IE pressure will occur.
Steam leaks and other siellar events will exhibit the sympta of excessive .
heat transf er on the CR-3 ATOG P-T Display. Control rom personnel can begin taking the required actions to mitigate the consequences of this event. r 3.7. Lnu of rhalant Accident Th e Ioss of coolant accidont (LOCA) Is a compf ex and dif fIcuit accidont to handle for the following reasons:
(
i
- 1. A wide range of l'eak sizes is possible from small breaks in Instrument lines to large breaks in the RCS piping,
- 2. A wide range of break locations is possible and the RCS response can depend on break location,
! 3. Abnormal sy stem condition such as RCS saturation, Interrupted natural circulation, etc. may be a natural consequence of the event, t
' 3-8 1
l l
-1 i
j ,
- 4. Steam generator heat removal may be degraded,
- 5. Hot standby conditions may not be a safe, stable condition,
- 6. The reactor building environment will degrade due to increased 4
pressure, temperature, humidity, and radiation.
1LOCAs can be categorized by size. Large break LOCAs result from a major f ailure in the primary system pressure boundary which depressurizes the RCS re'pidly and almost completely. Since this is a design basis event for which extensive analysis is performed in the FSAR, it was not anclyzed i
= f or ATOG. SealI break LOCAs are Iess severe than ,1arge breaks. The RCS depressurIzation Is'much siover due to the Iower mass fIowrate out the break. The Emergency Core Cooling System (ECCSD should maintain adequate core cooling throughout the transient. For small breaks, control room _ personnel plays a vital role in minimizing the consequences of the occident. Small Isaks are events whees the loss of reactor coolant is within the capecity of the normal makeup system.
The primary objectives during a LOCA are to maintain core cooling, to cooldown and depressurize the RCS, and to establish a stable, long-term cooling mode. Core cooling is maintained by the proper operation of the ECCS which control roan personnel verify. In addition, they should make sure that the reactor building cooling is functioning properly and that long-term cooling is established once the RCS is cooled down and depressurized.
In order to determine actions required by control room personnel to monitor, verify or perf orm ranually, and to Identify the expected RCS and system transient response, five snall break LOCAs were analyzed as part of the development of the Small Break Operating Guidelines and were translated into the ATOG guidelines. The five events analyzed were the following:
- 1. Small breaks that are large enough to depressurize the RCS below secondary system pressure with feedwater available.
3-9
l 1
- 2. Smal l breaks which stabilize at approximately the secondary j
_ system pressure with feedwater available.
- 4. Saall breaks without primary to secondary heat transfer.
- 5. Shall breaks within the pressurizer steam space.
These ' events were analyzed in suf ficient detall to generate symptoms for identifying the type of break as well' as distinguishing a LOCA frca other events, especially overcooling events.
Reacto'r Building !Yessure and Radiation alerts on the CR-3 SPDS In conjunc- -
tion with the pressure-temperature relationship response of the primary and secondary systems on the CR-3 ATOG P-T Display wIII provide the informa-tion necessary for control room personnel to identify a LOCA event.
They can then begin to take the actions required by their procedures to control the plant and mitigate the consequences of the event.
e Additional analyses have been performed in order to generate the subcooling margin Iimits to be displayed on the two SPDS P-T displays and to generate the inadequate core cooling limits to be displayed on the SPDS ICC display.
Those limits and alogrithms are prr.vided in Ref erence 8. Additional limits such as RCP NPSH, heatup and cooldown NDT, DHRS, and fuel compression were al ready availabl e and no f urther analyses were performed. These limits and algorithms are also provided in Reference 8 and are available on the CR-3 SPDS P-T displ ays.
i 3-10 I
l 4.0 CONIX.USIONS The Crystal River Unit 3 SPDS has been designed to provide control roon personnel with suf ficient key Information to enable them to determine the safety status of the plant with respect to the five required functions discussed in Section 2.0. The CR-3 SPDS has also been designed to provide sufficient Information in a display format based on and compatible with the symptom-oriented guidelines developed in ATOG as discussed in Section 3.0. With this key Information and concise display format, control room personnel at CR-3 can monitor plant status, ' detect key symptoms of abnormal plant response and take corrective actions necessary to restore control f unction or mitigate the consequences of transients and accidents in a rapid and rollable manner.
l l
4-1 l
5.0 REFERENCES
- 1. MJREG-0737, Supplement 1, " Requirements for Emergency Response Capabil-ity" (Generic Letter No. 82-33), December 17, 1982.
l
- 2. NUREG-0660, "NRC Action Pl an Devel oped as a Result of the TMI-2 Accident," May 1980.
- 3. NUREG-0737, " Clarification of TMI Action Plan Requrlements," November 1980.
- 4. NilREG-0696, " Functional Criteria for Emergency Response Facilitles,"
February 1981.
- 5. , B&W Document 74-1126473, " Crystal River Nuclear Power Station Unit 3 Abnormal Transient Operating Guidelines, Part I, Part Il-Vol. 1, Part i I-Yol . 2," October 1,1982.
- 6. B&W Drawing No. 1147047, " Florida Power Corporation - Crystal River-3 Safety Parameter Display System Function Description,"
August 19, 1982.
- 7. 88W Document 51-1121942, " Crystal River Unit 3 Safety Parameter Display System (SPDS)," June 21, 1982. '
- 8. B&W Document NPGD-TM-414, Rev. 03, " TRAP 2- FORTRAN Program for Digital Simulation of the Transient Behavior of the OTSG and Associated RCS, Rev. J," May 1980.
- 9. 8&W Document 86-1124195-01, "CR-3 Excessive Feedwater Transient information Document," July 17, 1981.
- 11. B&W Document 86-1125293, "CR-3 Small Steam Line Break Event Transient Information Document," May 8, 1981.
- 13. B&W Document 85-1118041, "Results and Recommendations of the Main Success Path Analysis for SGTR ( ANO-1)," March 10,1980.
- 14. 84W Document 86-1118045, " Impact of an OTSG Tube Rupture wIth Concurrent Loop," April 29, 1980.
- 15. B&W Document 86-1?20490, "0TSG Tube Rupture Alternate Paths (ANO-1 ATOG)," August 8, 1980.
5-1 l
_ .- .--- - . _ . . . - . - . - . - _ _ , , . _ . . . . . , _ . ~ - . , - _ . . _ . - - - _ _ _ . - . _ _ _ , - . - -
IIEFBIENCES (cont'd)
- 16. B&W Document 74-1123094, " Operation Guidel ines f or Saall Breaks for Crystal River llI," January 12, 1981.
- 17. 84W Document 55-1149091-00, " Analytical Justification for the Treatment of RC Pumps Following Accident Conditions," February 1984.
) .
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l 5-2 I
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- - . . . . ._ . _ . _ _ . . . _ _ . . . . _ , , . . - - - _ . . . . _ . . _ .- _ . . . -- . . . . - _ _ _ _ , _ . . _ . - . . . _ . . . . . - . . - . - - - - . . . . - - . - . _ .