ML20209G166
| ML20209G166 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/31/1986 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20209G134 | List: |
| References | |
| 47-1139815-01, 47-1139815-1, TAC-56525, NUDOCS 8702050291 | |
| Download: ML20209G166 (31) | |
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i 47-1139815-01 December 1986 SAFETf PARAMETER DISPLAY SYST5M SAFETY ANALYSIS FOR THE RANCHO SECO NUCLEAP. GENERATING STATION, UNIT 1 c
for Sacramento Municipal Utility District by The Babcock & Wilcox Company Utility Power Generation Division P. O. Scx 10935 Lynchburg, Virginia 24506-0935 8702050291 870112
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v CONTENTS Page EXECUTIVE SIMMARY..........................
iff 1.
INTRODUCTION 1-1 1.1.
P u rp o se...........................
1-1 1.2.
Background
1-1 2.
SPOS FUNCTIONS /PARAETERS MONITORED................
2-1 2.1.
Introduction 2-1 2.2.
Reactivity Control 2-1 2.3.
Reactor Core Cooling and Heat Removal F rom th e P r i ma ry Sy st em...................
2-2 2.4.
Reactor Coolant System Integrity 2-3 2.5.
Rad ioactiv ity Control....................
2-4 2.6.
Containment Conditions 2-4 2.7.
S um ma ry..... '......................
2-5 2.8.
RGl. 97 Pa rameters......................
2-5 3.
APPLICABLE EVENTS........__._.................
3-1 3.1.' ' Introduction 3-1 3.2.
Excessive Feedwate r Event..................
3-3 3.3.
Loss of Main Feedwater Event 3-3 3.4.
Steam Generator Tube Rupture (SGTR) Event.
3-4 3.5.
Loss of Offsite Power (LOOP) 3-5 3.6.
Small Steam Leak 3-6 3.7.
Loss of Coolant Accident 3-7 i
4 CO NCLU S ION S. *...........................
4-1 5-1 l
S.
REFERENCES t-l l
l List of Tables IAb.la Page 2-1 Parameters Required to Monitor the Five SPDS Safety Functions 2-6 a.
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EXECUTIVE SLNMARY
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The purpose of this document is to provide a written safety analysis for the Rancho Seco SPDS.
This analysis describes the basis on which the selected paramete rs are sufficient to assess the safety status of the pl ant with respect to five requi red functions 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 NUREG 0737, Su ppl eme nt 1,
" Requirements for Emergency Response Capability".1~
The Rancho Seco SPDS consists of two color video monitors (CRTs) with associated control panels.
This system, located in the. control room, allows the control room operator to select from 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 fu nctions and (2) compatibility to ATOG concepts and requirements.
This document summarizes the NRC requirements for SPDS functions and how the Rancho Seco SPDS can be used to assess those functions.
" Alert" signals to the control roan operator allow him to assess the
- reactivity control" and
" containment conditions" and " radioactivity control" fu nctions.
Pres sure-temperature information on P-T displays allows the control roon operator to monitor for abnormal symptans regarding subcooling and heat transfer for the
" reactor core cooling and heat transfer from the primary system" fu nction.
One " alert" signal plus this same pressure-temperature information allows the control room operator to monitor against key pressure-temperature limits for the " reactor coolant system integrity" function.
The events analyzed for the B&W Owners Group ATOG program are the basis for the Rancho Seco SPDS.
They include a wide range of events of low to moderate frequency of occurrence and were chosen becau se they encompass al 1 the 1.,
necessary symptoms for which the SPDS and the ATOG are designed to monitor and
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control.
The parameters chosen for the Rancho Seco SPDS to meet the above requirenents are summarized in Table 2-1.
. Based -on the information provided.in.this report, it.is demonstrated..thatethe Rancho Seco SPDS provider the control room operators with sufficient information to e.nable them to determine the safety status of the plant for a aide range of abnormal and eme rge ncy conditions.
In addition, it is demonstrated that the Rancho Seco SPDS provides sufficient information to be used in conjunction with ATOG-type procedures to detect abnormal symptoms and to al l ow corrective actions designed to restore the control function or mitigate the consequences of transients and accidents in a rapid and reliable manne r.
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1.
INTRODUCTION i
1.1.
Pu rnose The purpose of this report is to provide a written safety analysis for the
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Rancho Seco Safety Parameter Display System (SPDS) describing the basis on which the selected parameters are sufficient to assess the s'afety status of the plant for a wide range of events which include symptoms of severe accidents.
This rep ort will summa rize the NRC requi rements for an SPOS and will dImonstrate that the Rancho Seco SPDS meets those NRC requirements.
In addition to these NRC requirements, another basis for the Rancho Seco SPDS is the B&W Owners Group Abnormal Transient Operating Guidelines (ATOG) program.
This report will summarize that program, particularly the events analyzed, and demonstrate the compatibility of the Rancho Seco SPOS with ATOG.
Finall y, this report will identify the parameters selected for the Rancho Seco SPDS and l,
the meth6d of their presentation to the operator.
This report is in response to the requirement for such a document as contained in Section 5.2 of NUREG 0737, Supplement 1,
" Requirements for Emergency R:sponse Capability" (Generic Letter No. 82-33), dated December 17, 1982.1 1.2.
Back crou nd As a result of the TMI-2 accident, the um issued an action plan for items to bs addressed-In order to correct or imps ove the regulation and operation of nuclear facilities.
That plan was provided in NUREG 0660, "NRC Action Plan 01veloped as a Result of the TMI-2 Accident", dated May 1980.2 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 the reactor operator and the systems he is required to operate.
The variety and j
quantity of information displayed in the control room can eften be cverwhelming, especially during transient operations or an accident. ~ It has b0en determines 1 that a concise display of those parameters necessary to assess the safety status of the p,lant would significantly aid the reactor operator in l-1
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datermintag plant status and diagnosing accidents.
Item 2 of Task I.D of NWEG 0660 established the requirements for a plant safety parameter display console to be installed in the control rom.
In November 1980, the NRC issued NUREG 0737, " Clarification of TMI Action P1 tnt Requirements."3 Conce rning the requi rements for a plant safety parameter display console, the NRC stated that issuance of NLREG 0696,
" Functional Criteria for Emergency Response Facilities" would. provide those requirements for an SPDS.4
' UREG 0696 was issued in February 1981.
N NUREG 0696 describes the requirements for all the systems and facilities which cake up the licensed eme rgency response facil ities.
Section 5 provides guidance for the design of the SPDS including its purpose, location, size, display considerations and design criteria.
These guidelines serve as a basis fer the design requirements for the Rancho Seco SPDS.
The purpose of the Rancho Seco SPDS is to assist the control roan operator during normal, abnormal and emergency conditions in the evaluation of ' the sa fety status of the plant and assessing whether abnormal or emergency conditicns warrant corrective actions to avoid a degraded corn.
Located in the control roan, the SPDS will be conti,nuously available during all modes of plant o# ration including cold or refueling shutdown, heatup and cooldown oparations, normal power ope rations, as well as abn ormal or emergency canditions.
The Rancho Seco SPDS, including video monitors is seismically qualified to satisfy the requirements of RG1.97 in order that the SPDS may be used for monitoring and displaying the RG1.97 Category 1 variables.
Plant l
parameters which indicate very abnormal conditions such as inadequate core cooling (ICC) shall also be displayed.
The Rancho Seco 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 following functions:
(1)
Reactivity control.
(2)
Reactor core cooling and heat removal from the primary system.
(3)
Reactor coolant system integrity.
(4)
Radioactivity control.
(5)
Containment conditions.
1-2
g Finally, the Rancho Seco SPDS is designed to be compatible with the control room operator's training and experience in normal and abnormal or emergency oparation.
SMUD is a member of the B&W Owners Group Operator Support
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" Committee "which 'has worked "to ' develop"a program to respond'to^ Item #T.C.T'cf o
NLREG 0737, " Guidance for the Evaluation ard Development of Procedures for Transients and Accidents."
That program resulted in the development of the Abnorma'l Transient Operating Guidelines (ATOG) which employs the concept of symptom-oriented rather then event-oriented procedures for handling abnormal or emergency conditions.
The design of the displays for the Rancho Seco SPDS has. incorporated the basic concepts of ATOG such that sufficient information is provided to the control room operators to monitor the status of the plant during normal, transient and abnormal conditions, diagnose symptoms and take corrective actions to achieve stable plant conditions.
Ths Rancho Secc SPDS is designed to serve as an aio to the contral room operators and will be used by them to quickly focus on certain key parameters.
Horever, the SPDS will not be used as the exclusive source of information to datermine plant status or to monitor control actions.
Other instrumentation in the control rean will also be used by the opera. tor to detennine plant statu s.
Tho Rancho Seco SPDS will be located in the control roan.
The control panels (one for each CRT) will be mounted in the operators console.
The CRTs will be mounted vertically in a rack next to the vertical control panels to the right of the operators console.
The supporting SPDS equipment is located in the i
equipment room to the right of the control room.
The Rancho Seco SPDS will l
provide information to the operator via selectable displays and automatically displayed alert signals.
The selectable displays. include the Low Temperature f
Prossure-Temperature (P-T) display, Post-Trip (ATOG) P-T display, two l
Inadequate Core Cooling (ICC) displays, two " normal" power operation
- displays and a five-page -alphanumeric displap of selected parameters and containment isolation valves' status.
The control room operator can select any available display by a single pushbutton on the SPDS control panel.
If the control room operator has selected either the Low Temperature P-T cisplay or the Post-Trip i.
(ATOG) P-T display, he can add any of several P-T l'mit curves to the base f
display by a single pushbutton for each limit, curve.
This feature allows the
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control room operator to be able to monitor the plant status versus those P-T 1-3 e
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licits applicable to the current plant conditions or expected plant evolutions such as plant cooldown to cold shutdown conditions,.
Each control panel also c3ntains a toggle switch which allows the operator to select either Loop A or Lecp B parameters for display on either CRT.
There~1s also a~ feature to the RIncho Seco SPDS which allows the operator to m>nitor the history of the RCS pressure-temperature relationship.
Using a single pushbutton, the control room operator can demand or erase the history trace of the RCS pressure-temperature relationship.
Finally, the control room operator can select the incore thermocouple tenperature (average of the five highest thermocouples) to b3 displayed instead of T-hot on the two P-T displays or at the same time as T-hot.
Tho two " normal" power operation displays, " Normal" and " Flux / Flow / Imbalance" cIn be u sed by the control room operator during normal operatiens to monitor plant cenditions against the Reactor Protectio n System (.lPS ) P-T and flux / flow / imbalance limits.
The Low Temperature P-T and ATOG P-T displays can b3 monitored by the control room operator during heatup, cooldown and normal operations and also during transients and accide.its.
Die two ICC displays can b3 monitored by the control room operator during conditions of RCS saturation or superheat, for example.
The seven " alert" signals will automatically flash on any selected display if conditions warrant.
The five-page al phanume ric display provides backup, confirmatory information for the parameters used in the displays and the " alert" logics.
The alphanumeric displays also provides information on a number of other key parameters that are important to the control room operator for the operation of the plant during both normal and abnormal or emergency conditions.
A detailed description of the Rancho Seco SPDS hardware, displays, alerts, and input signal list can be found in the document " Rancho Seco Safety Parameter Dis pl ay System Functions Description".7 Algorithms for the display curve i
limits can be found in the ' document, "SMUD Safety Parameter Display System (SPDS)."8 1
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s 2.
SPOS FUNCTIONS / PARAMETERS MONITORED 2.1 Introduction In Section 1.2 above, the purpose of the SPOS was stated.
In order to meet that purpose, the SPOS must. be designed to provide information to the control room ope rator - for all modes of norn1 operation from refueling to power operation.
It must also be designed to provide key information to the control room operator during abnormal and emergency conditions so that he can assess the safety status of the plant.
In particular, it must provide sufficient information to allow the control room operator to assest., the safety status of tae plant with respect to five major functions.
2.2 Reactivity Control The fi rs t function is reactivity centml.
During normal ope ration, the control room operator needs to know that reactor power and imbalance are being properly controlled.
Following a reactor trip, or when conditions warrant a reactor krip, the control room operator needs to know that the control rods have all inserted and that the reactor is subcritical.
If not, he must begin taking required actions to achieve reactor subcriticality.
On the Rancho Seco SPOS, the reactivity control function is monitored by one enormal" (Flux / Flow / Imbalance) display and the " Reactivity" alert logic.
The parameters to be monitored for the " normal" display are power range nuclear instrumentation power l evel and imbalance, control rod g roup in-limit position, source range nuclear instrumention count rate and reactor trip signal for the " Reactivity" alert.
These parameters are sufficient to monitor the reactivity control function since they insure that reactor power and imbalance are being controlled within the RPS limits and that control rods 1
have in se rted upon reactor trip.
.They also insure that the reactor is sufficiently subcritical at all times other than those where the reactor is already critical or where a reactor startup is underway.
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2.3.
Reactor Core Coolina and Heat R val From the Primarv Svstem 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, the control room operator must have information to determine the thermal-hydraulic state of the reactor,
crolant.
If the reactor coolant is in a liquid state and subcooled, the control room operator is assured that reactor coolant is available and capable of remov ing heat from the reactor core and transferring it to the steam generators.
If subcooling is lost, these capabilities are in doubt and the control room operator can begin taking actions to restore subcooling.
On the Rancho Seco SPOS, the reactor core cooling function is monitored in five ways.
During normal ope ration, this function is monitored by the
" normal" display of the RCS pressure-temperature, status poirts with respect to the RPS pressure-temperature limits which inherently insures that the RCS is in a subcooled state. During post-trip operations, this function is monitored by the display of the RCS P-T status with, respect to " expected" post-trip pressure and temperature l im its as well as the saturation I fne and the va ri able subcooling margin limit which takes into account the instrument string errors for the pressure and teinperature.
This information is available to the control room operator on the Post-Trip (ATOG) Display.
During inadequate core cooling operations, this function is monitored by the display of.the RCS F-T statu s with respect to saturation and two fue' cladding temperature limits.
This in formation is available to the control mem operator on the ICC display.
During cooldown operations, this function is monitored on the Low Range P-T Display in like manner to the post-trip operations discussed above.
In each of these situations, the parameters to be monitored are RCS pressure, RCS hot leg temperatyre (T-hot), RCS cold leg temperature (T-cold), and RCS ave rage incore thermocouple temperature.
When displayed against saturation and variable subcooling margin, these are the only paramete rs that are necessary to determine the state of the RCS for monitoring the reactor core cooling function in these four uttuations.
The final means of monitoring the reactor core cooling function occurs once the RCS is being cdoled by the Decay Heat Removal System (OHRS).
The "CHRS 2-2
Low Flow" alert logic is used to detect less than required DHRS flow rates.
The parameters to be monitored for this alert are RCS pressure, T-cold, DHRS 1 cop flow rates, and DHR3 drop-line valves' and LPI control valves' positions.
In addition to reactor core cooling, heat removal from the primary system must be monitored.
The control room operator must have information which allows him to monitor the heat transfer coupling between the reactor coolant system and the steam generators.
The control room operator can begin taking corrective actions to restore that coupling if inadequate heat transfer is occurring.
If excessive heat transfer is occurring, the control mom operator can begin taking actions to restore the balance between the heat sou rce (reactor core) and heat sink (steam generators).
Cn the Rancho Seco SPDS, the heat removal from the primary system function is monitored by the Post-Trip (ATOG) P-T, low range P-T and ICC displays.
The RCS pressure-temperature status is displayed as a point.on *all three displays and the steam generator saturation temperature based on its pressure is displayed as a vertical bar on the Post-Trip (ATOG) P-T display only.
With th e'se four pa ramete rs (RCS pressure, T-hot, T-cold and steam generator pressure), the control roon operator is provided the symptes of inadequate or cxcessive heat transfer.
The Rancho Seco SPDS P-T displays also provides bar charts for steam generator startup and EFIC low range and EFIC full range icvel indication to aid -Jie operator in recognition of possible causes for the symptom and also to verify that OTSG 1evels are being properly controlled by tho MFW and AFW systems.
2.4.
Reactor Coolant System Intecrity The third function is roactor coolant system integrity.
During all modes of ncrmal operation and during transients and accidents, the control r6cm operator must be able to monitor the status of the reactor coolant system pressure-temperature rela tion sh ip against several important pressure and pres sure-temperature limits to ensure the integrity of the reactor coolant system against overpressurization, pressurized thermal shock limits and heatup or cooldown NOT limits.
If limits are being approached or exceeded, the c:ntrol room operator can begin taking actions requi ed to re-estabitsh an
" acceptable state.
Also, there are a number of pressure, tempe ratu re, and l
pressure-temperature limits important to normal, operation..
The control mun 2-3
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Cperator must have the necessary information to determine the statu.* of the
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plant with respect to thesis limits and control the plant within these limits.
On the Rancho Seco SPOS, the reactor coolant system integrity function is
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monitored by the "?ressurizer valve" alert and also by the use of the following parameters:
RCS pressure, T-hot and T-cold.
These paramocers are displayed as P-T status points against the following applicable RCS P-T or pressure limits: RCS design pressure, DHRS design pressure, pressurized thermal shock limit, heatup NOT and cooldown NOT limits.
Operation of the plant within these limits will insure the integrity of the reactor coolant system.
In addition, the status of the two pressurizer code safety valves and the PORV are monitored.
The " pressurizer valve" alert will alert the control rocm operator that one or more of these valves is open and thus, the integrity cf the RCS is in.1oubt.
23.
Radioactivity Control The fourth function is radioactivity control.
During all modes of normal cperation and especially during accident conditions, the control room operator must be aware of the presence of high radioactivity throughout the plant and monitor possible release paths to_ determine the extent of radioactivity releases
- from the plant.
He can then take necessary actions to prevent or einimize such releases.
The input parameters to this alert are containment rad i atio n, main steam line radiation, reactor building stack ef fluent radiation, auxilia ry buil ding' stack effluent radiation and rad waste area ef fluent radiation levels.
2.6.
Containment Conditions The fin a.1 fu nctio n is containment conditions.
The last fission product barrier to the envi mnment is the containment building.
The control room operator needs information as to the status of the containment building in
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crder to take necessary actions to ensure or restore its integrity during accident conditions.
On the Rancho Seco SPOS, the containment conditions function is monitored by fcur sets of aler: logic.
They are the " radioactivity" alert which monitors containment radiation, " containment pres sure" ale rt, " containment hydrogen concentration" alerts, and " containment isolation" alert.
Containment pressure and radiation are the two parameters which should give the control 2-4
roan operator the first indication of a major accident, such as a LOCA or 7
steam line break, inside containment.
Should the accident evolve to such a state that containment isolation is necessary, the control. room operator will b3 alerted to a failure to completely isolate containment for a level 1 SFA'S.
Should the accident evolve to such a state that hydrogen is generated by the c:re, the control room operator will be alerted such that he can begin taking neces sa ry actions to reduce the concentration in containment.
The input pa ramete rs to these alerts are containment radiation level, containment p r es sure, containment hydrogen concentration, SFAS channel 1A and 18 trip status, plus the position of 33 containment isolation valves.
2.7.
Summa rv The complete list of parameters requi red to monitor the five SPDS safety fu nctions is provided in Table 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), plus a summary discussion.
2.8.
R.G.1.97 Pa rameters In addition to monitoring the parameters required to assess the five SPDS functions which.have been discussed above, the Rancho Seco SPDS also is used to monitor the RGl.97 Category 1 variables.
Several of these ( f.e., power range nuclear instrumentation, RCS pressure, EFIC full range OTSG 1evel, EFIC OTSG pressure, containment pressure, incore thermocouple tempe ratu re, centainment radiation, containment hydrogen concentration and containment isolation valves status) are required to monitor the five SPDS functions for Rancho Seco.
The remaining RG1.97 variables that are monitored by the Rancho Seco SPDS are the following:
1.
RCS hot leg level - alphanumeric display, RCS inventory alert, ICC display.
normal displays, P-T displays, al phanume ric 2.
Pressurizer level display.
3.
AFW flow - alphanumeric display.
4.
Candensats storage tank level - normal displays, P-T displays, CST la"el alert and alphanumeric display, w
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Table 2-1 Parameters Reout red to Monitor the Five SPDS Safety Functions
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2.0)
Display Parameter Sional inout Ranoe Fu nct ion (1.0) or alert Discussion 6
1.
Source range X100001, X100002 10~1 to 10 cps 1.1 2.4,2.5 Used l'n conjunction with reactor tr NI A & B signal and CRG 2-7, positions to alet the control room operator to high, unexpected neutron count rate post-reactor trip.
2.
Power range X100401, X100402 0 to 1255 FP 1.1 2.4,2.10 Plotted and displayed with Power NI A,B,C & D X100403, X100404 Range Imbalance on normal display.
Used to monitor and verify the propi control of Power and Imbalance vs ti RPS Power / Imbalance Flow trip y
I envelope during normal power os elevation.
This parameter is also,
s RG1.97 Category 1 variable at Ranch
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Power range XI-004018,XI-00402B ~-62.5 to 1
1.1 2.10
. See discussion for Power'Rmge NI A l
imbalance XI-004048,XI-004058
+62.5% FP B, C & D.
NI-5,6,7 & 8
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T-cold A & B TE-21025C,TE-21023B 50 to 650 F
- 1. 2', 1. 3 '
2.1,2.2, Plotted and displayed:with RCS TE-210240,TE-21024B 2.4,2.9, prcssure on P-T cli;ves,to monitor 2.10, plant, status against1 RPS, NDT.and
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' thermal shock limits, and' to provid l-i N
symptoms of excessivo or inadeguatc 7
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heat transfer when used with SG saturation temperature.
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S' e' discu'ssion f'oh T-cold A & B. ' Il 5.
T-hot A & B TY-210310,TY-21032C 120 to 920"F 1.2, 1.3 2.1,2.2 e
2.3,2.4 ad d ition', it is displayed against 2.10 subcooling margin fand saturation limits for symptom of inadequate core cooling.
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Function (1.0) Display (2.0)
Pa rameter Signal input Range or alert Discussion 6.
RCS pressure PT-21051, PT-21050 0 to 3000 psig 1.2, 1.3 2.1,2.2, See discussion for T-cold A & B.
A&B WR 2.3,2.4, Also plotted with incore thermo-2.9,2.10 couple temperature to monitor natura circulation or superheated condittor during ICC.
This parameter is a RGl.97 Category 1 variable at Ranchc Seco.
7.
RCS pressure PT-21043, PT-21042 0 to 2500 psig 1.2, 1.3 2.1,2.2, See discussion for RCS pressure "WR' AaB NR 2.4,2.9,2.10 except for RGl.97.
8.
RCS pressure PT-21261 0 to 400 psig '
1,.2, 1.3 2.1,2.4, used as RCS pressure signal during A&B, Low R I
2.9 low RCS pressure operation and DilRS switchover.
9.
OTSG 1evel SU LT-20503D,LT-20504D 0 to 250 in.
1.2 2.1,2.2, Displayed on normal and P-T displays rin0e AAB 2.3,2.4, primarily for inmediate detection 01 2.10 excessive feedwater and loss of feedwater events.
SU range may be useful in SG tube rupture ide nti fication.
10.
EFIC low LT20505A,LT20505B 6 to 156 in.
1.2 2.1,2.2, Used on the P-T and ICC displays to range level, LT20506A,LT205068 2.3.2.4 provide information to aid the OTSG A&B control room operator in recognittor of excessive or inadequate heat transfer and that SG 1evels are beir properly controlled by the MFW and AFW systems.
11.
EFIC full LT20507A,LT205078 6 to 619 in.
1.2 2.1,2.2, See discussion for EFIC low range range level, LT20508A,LT20508B 2.3,2.4 level, SG A&B.
This parameter is 0fSG AaB also a RGl.97 Category 1 variable a1 Rancho Seco.
T bla 2-1 (Cont'd)
Function (1.0) Display (2.0)
Pa rameter Signal inout Range or alert Discussion 12.
EFIC OTSG PT20545A,PT20545B 0 to 1200 psig 1.2 2.2,2.4 Used as input to SG saturation pressure A&B PT20546A,PT20546B tmperature on Post-Trip ATOG display.
Used to display symptoms excessive heat transfer and inadequate heat transfer when used i
with T-col d.
This parameter is als a RGl.97 Category 1 variable at Rancho Seco.
13.
Containment PT-536068
-5 to 35 psig 1.4 2.4,2.6 Used in containmont pressure alert y
pressure, NR signal to alert the operator to hig pressure co iditions as a result of g
LOCA or steam leak, l.
Containment PT-53621, PT-53622
-5 to 180 psig 1.4 2.4,2.6 See discussion for Containment j
This parameter is al a RGl.97 Category 1 variable at Rancho Seco.
j 15.
Decay heat FT-26003, FT-26004 0 to 5000 gpm 1.2 2.4,2.9 Used in conjunction with DHRS valve flow AaB status to alert the control room operator to low flow conditions and potential core cooling problems during DHR System operations.
l 16.
Incore 16 selected T/Cs 100 - 2300 F 1.2 2.1,2.2, The 16 incore thermocouples are Thermocouple 2.3,2.4 found in Reference 7.
The average j
(calculated value of the five highest is plotte average of 5 with RCS pressure on ICC curve duri highest readings)
ICC conditions and can be plotted j
with RCS pressure on other P-T l
curves.
(See discussion of RCS pressure, WR).
This parameter is j
also a RGl.97 Category 1 variable a e
hA Rancho Seco.
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r Tcblo 2-1 (Cont'd)
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Function (1.0) Display (2.0)
Pa rameter Signal inout Range or alert Discussion 8
- 17. Containment RP-15049,RP-15050 1 to 10 R/hr 1.5
- 2. 4,2. 7 Used in radioactivity alert to aler radiation 1&2 the control rom operator that high radioactivity level is present in tl containment building.
This paramet' is also a RG1.97 Category 1 variabli at Rancho Seco.
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- 18. Main Steam R15047, R15048 1 to 10 mR/hr 1.5
- 2. 4, 2. 7 Used in the Radioactivity alert to line radiation alert the control mom opcrator to Adts the presence of high radioactivity
{
1evels in an ef fluent release path.
1 13 19.
Reactor R15044 10 to 10 p01/sec 1.5 2.4,2.7 See discussion for Main steam line building radiation A&B.
stack ef fluent l
20.
Auxili ary R15045 10 to 10 p C1/sec 1.5 2.4,2.7 See discussion for Main steam line 1
13 j
building radiation A&B.
stack ef fluent 1
13 21.
Rad waste R15546A 10 to 10 p C1/sec 1.5 2.4,2.7 See discussion for Main steam line area effluent radiation A&B.
22.
DHRS, Suction HV-20001,HV-20002 NCLO/CLSD 1.2 2.4,2.9 See discussion for decay heat flow valves status A&B.
23.
LPI Injection SFV-26005,SFV-26006 HOP /0 PEN 1.2 2.9 See discussion for decay heat flow valves sta tus A&B.
24.
Reactor trip 86/RT Normal / trip 1.1 2.2.2.4, See discussion for source range NI sta tu s 2.5 A&B.
In addition, this signal also triggers the. automatic display of tt Post--Trip ATOG display for loops A&E on the two CRTs.
l t
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I Table 2-1 (Cont'd)
Display (2.0)
Parameter Signal inout Range Fu nct io n(1. 0) or alert Discussion 25.
CRG 2-7 Z9130 Not in/in 1.1 2.4,2.5 See discusstor. of source range position NI A&B.
i 2f..
SFAS channel G9021, G9025 Normal / trip 1.4 2.4,2.8 Used in the containment isolation ll.&lB trip alert to alert the control room statu s operator to the improper positionin!
of one or more containment isolattoi valves following a level 1 SFAS i
w initiation.
The control room y;
operator can determine which valve (:
from either existing control room I
{
indications or the' alphanumeric display.
I
- 27. Containment 33 valves are NCLO/CLSD 1.4 2.4,2.8 The 33 valves monitored are found ti I
isolation monitored reference 7 and are used in valve status conjunction with a level 1 SFAS to provide an alert signal to the control room operator if they are n<
all in the correct SFAS position.
This parameter is,also a RGl.97 Category 1 variable at Rancho Seco.
28.
Pressurizer PSV-21507,PSV-21506 CLSO/NCLO 1.3 2.4,2.11 Used in the pressurizer valve alert code safety to alert the operator that one valves 1&2 or more of the pressurtzer valves 1:
open.
l 29.
Pressurizer PSV-21511 CLSD/NCLO 1.3 2.4,2.11 See discussion for pressurtzer code l
PORV safety valves.
i
t I
Ttblo 2-1 (Cont'd) 1 l
Display (2.0) i 5
Pa'rameter Sional inout Ranoe Fu nct io n(1.0) or alert Discussion i
l 30 Containment AES3811, AE53812 0 to 10%
1.4 2.4,2.12 Used in the containn.ent H l
H2 concentration concentration alert to albrt the control room operator that high concentration of H2 exists insido containment.
This paramoter is als a RG1.97 Category 1 variable at Rancho Seco.
I
!w 1.0 SPDS Function
} i e
1.1 Reactivity control I
l 1.2 Reactor core cooling and heat renoval from primary systan 1.3 Reactor coolant system integrity 1.4 Containment Integrity 1.5 Radioactivity control 2.0 Disolav or Alert 2.1 Low Temperature P-T display 2.7 Radioactivity alert 2.2 Post-Trip ATOG P-T display 2.8 Containment isolation alert 2.3 ICC display 2.9 DHRS alert 2.4 Alphanumeric display 2.10 Normal display 2.5' Reactivity aiert 2.11 Pressurizer Valve alert 2.6 Containment pressure alert 2.12 Containment H2 concentration alert 4
e 9
3.
APPLICABLE EVENTS 3.1.
Int rodu ction-In part, the Rancho Seco SPDS is based on ATOG.
The control roan operator.
-must be provided with certain parameters which he can monitor in conjunction with the ATOG-based emergency operating procedures.
These parameters are those necessa ry for the identification of the three basic heat. transfer symptans (lack of subcool ing, inadequate heat transfer or overheating, and excessive heat tran sfer or ove rcool ing) and the special case of steam ganerator (SG) tube rupturo.
Once the control room operator identifies one or more of these symptoms, he 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 / temperature, SG inventory,
)
and SG pressure control.
The parameters wh ich are necessa ry for the identi.f f cation of these four symptoms. 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 from the primary system, reactor coolant system integrity as l
ras discussed in Section 2.0.
Additional pararieters are provided in order to assess the safety status of the plant with respect to the remaining three requi red SPDS functions: reactivity control, radioactivity control and containment conditions.
Expected control naam operator actions to verify that al l control rods have inserted and that measured neutron. flux levels are l
decreasing are the basis for the two parameters necessary for the " Reactivity" al e rt.
Expected control room operator actions to monitor containment
~
radiation and pressure values are. the basis for the parameters necessary for the " Containment" alerts.
Expected control roan operator actions to monitor rad ioactiv ity source and release paths are the basis for the paramete rs necessary for the " Radioactivity" alert.
In order to better understand the btses for the parameter selection for the Rancho Seco SPDS, it is important to understand the events that were analyzed -for developing the ATOG c'aidelines.
e 3-1 E
.l
Tho ATOG guidelines have been written to cover a very large number of probable abnormal transient-scenarios.
This is an inherent benefit of a symptan-
~
cricnted ap proach.
These transients can include classic single initiating evsnts as well as ' additional sin'gle or multiple'fa'ilures.
For the ATOG development program, six initiating events were identified and analyzed in detail.
The six initiating events selected were:
1.
Excessive feedwater 2.
Loss of main feedwater (LOFW) 3.
Steam generator tube rupture (SGTR) 4.
Loss of of fsite power (LOOP) 5.
Small steam leak 6.
Small break loss of coolant accident (SBLOCA).
These six events were chosen as representative for providing the b:1 sic symptoms on which ATOG is based.
Additionally, four specific criteria were used in selecting these six events:
1.
Moderate frequency events in which operator action is expected (LOFW, Excessive FW, LOOP).
2.
Low probability events that can be confusing in recognition and mitigation (SGTR, SBLOCA).
3.
The events cover a very large spectrum of conditions in the RCS (overheating, overcooling, loss of inventory, loss of subcooling).
4.
Time exists for the operator to recognize and do sonething about the accident or event; therefore, guidelines are appropriate.-
This would necessarily exclude large, rapid FSAR events such as LCCA and steam line break.
l For each of the six initiating events chosen, a detailed event tree was dcveloped to identify a main success path and major failure paths.
Analyses 9
using the TRAP code were performed on the main success path and most single fa il ure paths for the ATOG base pl ant, ANO-1.
The ANO-1 plant design, squipment, pertinent plant setpoints, etc. was then compared to Rancho Seco.
Transient Information Documents (TIDs) were then developed for use to modify the ANO-1 ATOG guidelines to fit Rancho Seco.
Those documents applicable to Rancho' Seco are refe rences 10 th rough 16.
The LOCA event analyses and recommendations were derived largely from those ef forts which produced the "Small Break Operating Guidelines."17 h
6
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3-2 s.
t-
e A brief summary of each of the six initiating events is provided below.
Osta11ed discussions of these events are found in the " Rancho Seco Abnormal
. Transient Operating Guidelines, Part II - Vol. 2."5 Those discussions include actual operating events which have occurred.
3.2.
Exces sive Feedwater Event The excessive feedwater event was analyzed for two cases.
For the first case, the excessive feedwater addition was tenninated 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 feedwater addition to one SG was terminated by the control rom operator approximately 5.5 minutes after event initiation.
Prior to termination of feedwater, the overfed SG has filled, the pressurizer has emptied due to RCS contraction, subcooling nargin has been lost and SFAS has started HPI and EFW.
In addition to terminating main feedwater, the control room operator must make tih ree significant control actions.
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 tihe overcooling.
The thi rd is to throttle HPI once subcooling margin is restored.
It should be noted that with the Emergency Feedwater Initiation antr Control (EFIC) system implemented at Rancho Seco, the control rom operator will only have to verify that main feedwate r has been terminated and that emergency feedwater flow is being controlled properly to achieve the desired SG level.
l The pressure-temperature relationship response of the primary and secondary systems displayed on the Rtncho Seco Post-Trip P-T Display will provide the control room operator the symptom of excessive heat transfer.
In addition, the SG 1evel bar charts will indicate the excessive feedwater addition.
This information will key the control room operator to take the required actions to terminate main feedwater and to take the three additional control actions.
3.3.
Loss of Main Feedwater Event The second event analyzed was the loss of main feedwater (LOFW) event.
This ev:nt was initiated by a trip of both main feedwater pumps from 100% FP with anticipatory reactor trip occi.rring.
In its early stages, this transient will look almost identical to a " normal" reactor trip.
The control rocm operator should be able to identify LOFW by indication of ma1n feedwater pumps tripped P
or' zero main feedwater ficw ~ rate at the feedwate'r centrol station, or rapidly l
iy 3-3 L
falling SG levels on the SPDS display.
The only significant operator action is to verify that EFW starts and is providing flow to the SGs.
If not, he should take manual contml of EFW.
This event was also run with a number of failures out to approximately 10 minutes after reactor trip.
The fi rst failure is failure of EFW to start.
Based on the pressure-temperature relationship response displayed on the SPDS display, the control mcm operator will recognize the symptom of inadequate primary to secondary heat transfer and take corrective actions to restore heat transfe r.
The second failure is EFW overfeed.
The control mom operator will recognize the symptom of excessive primary to secondary heat transfer and take corrective actions to throttle EFW.
As was previously noted, with EFIC implemented at i
Rancho Seco, the control room operator should only have to verify that EFW has init ated and is being controlled to achieve the desired SG level.
The third failure is steam leakage from a stuck open SG. steam safety valve or an open turbine bypass valve.
The control mom operator will recognize the sympton of excessive primary to secondary heat transfer, and by following his procedures, he will identify the steam leak and take corrective actions.
The pressure-temperature relationship.-cesponse of the primary and secondary systems ' displayed on the Rancho Seco Post-Trip P-T Display and the SG level bar charts will provide the control room operator the sympton of inadequate hsat transfer.
This information will key the control mom operator to take the requi red actions.
l 3.4.
Steam Generatsr Tube Ruoture (SGTR) Event A steam generator tube rupture (SGTR) is a loss of coolant accident (LOCA) l th rough the secondary plant.
It is a serious event which contaminates the secondary plant and can lead to significant radiation relea' es if steam from s
the affected SG is released to the atmosphere.
Since the leak cannot be isol' ted until the plant is cooled dowri and depressurized and the primary a
system loops have been drained, it is important that there be a timely identification of the event and the affected SG by the control' room operator.
l The analysis performad for ATOG began with a SGTR occurring at 100% FP.
The initial primary to secondary leak rate was approximately 400 gpm.
In the first seven to eight minutes, the control room operator will sea RCS pressure and. pressurizer level decr e'as ing.
He will al so receive steam lice and
~
3-4
a e
condenser air ejector of f gas radiation alarms.
Because the leak rate exceeds MU system capability, the reactor will trip on low pressure.
If the leak rates are within the capability of the MJ system, the control roan operator trould begin a rapid, controlled power reduction to zero power before tripping the reactor.
This action prevents 11 fting SG steam safety valves.
Following the reactor trip, subcooling margin is los, requiring the control room operator to trip all operating RCPs.
SFAS will actuate on low RCS' pressure starting HPI and EFW.
At ap p roximatel y 11 minutes into the transient, the pressurizer will empty and the RCS will become saturated.
In the next 4 to 5 minutes, the control room operator will throttle EFW to prevent overcooling.
With EFIC implementation, he will verify proper EFW control.
He will also thruttle FPI to stabilize RCS pressure and pressurizer level following the re-estabitshment of subcooling margin.
Once the plant has been stabilized with core decay heat being remosed by the SGs via natural ci rculatio n, the control room operator can restart RCPs and initiate plant cooldown and depressurization.
During the cooldown, the control room operator will maintain a minimum subcooling margin to keep RCS pressure and the primary to seconda ry leak rate low.
Main steam line radiation al a rms lrom the " Radioactivity" al e rt, in conjunction with the pressure-temperature relationship response of the prima ry and secondary systems plus the SG level bar charts (on the Rancho Seco Post-Trip P-T Display) will -provide information to the control room operator to identi fy the SGTR event and the affected SG.
He can then begin to take required actions.
3.5.
The analysis for this event begins at 100% FP with the plant separation from the grid.
During the reactor and turbine runback, the reactor trips on high pres sure.
The diesel generators start on detection of undervoltage on the two engineered safeguard buses and begin loading within approximately 10 seconds.
Secondary steam pressure is controlled by the steam safeties or atmospheric dump val ves.
Since the RCPs have trip ped, natural ci rculation will be established and EFW will start feeding to establish level in the SGs.
That level is reached approximately 8 minutes.after grid separation occurs, and the RCS should be stabilizing at normal hot shutdown conditions.
With EFIC implementation, the EFW flow will be automatically controlled to establish 3-5 g.
a
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natu ral ci rculation level setpoint at a rate not to exceed 8 inches per min ute.
Additional : anal yses. were performed > for a : total loss of. al1 power x except 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 voids begin to form in the RCS and natural ci rculation is interru pted, the control roon operator can fill the SGs to ECCS setpoint to establish the core boiling /SG condensing mode for decay heat removal.
EFIC will autmatically control SG 1evels to the ECCS setpoint once the control room operator demands that mode.
Every ef fort must be made to get the diesel generators started.
Since the pressurizer heate rs and makeup p umps have lost power, the control mom operr. tor has no control of primary pressure or inventory.
The plant cannot be cocled down because water cannot be added to makeup for RCS contraction during cool down.
The pressure-temperature relationship response of the primary and seconda ry systems displayed on the Rancho Seco Post-Trip P-T Display will provide the operator information on the adequacy of natural circulation and the symptcm of inadequate heat transfer if natural cJ rr.ulation is intermpted.
3.6.
Small Steam Leak The event analyzed was a failure of the turbine bypass system.
This failure results in a steam leak equivalent to approximately 15% of full power steam flow and occurs with an assumed maximum decay heat.
In this eve nt, the raactor trips on high flux in about 9 seconds following full opening of the turbine bypass system valves.
The turbine trips and ICS runs back main fsedwater.
The turbine bypass valves on one SG remain open.
Ap proximately three minutes into the tran'sient, the pressurizer drains due to excessive RCS contraction.
SFAS initiates due to low RCS pressure and starts HPI and EFW.
As RCS pressure and pressurizer level are restored, the control rom operator begins th rottl ing HPI.
Ap proximately 5 minutes into the tran sient, the operator ide nti fies the affected SG based on symptoms of excessive heat transfer and decreasing steam pressure in the affected SG and allows it to boil dry.
At approximately 3, minutes, the control room operator stops HPI, O
3-6
u v
realigns to normal makeup / letdown mode and uses the atmospheric dump valves on the unaffected SG to stabilize the RCS.
Larger leaks, less decay heat or subsequent failures may cause a transient s: vere enough to cause loss of subcooling margin and possibly RCS saturation.
The control room operator must follow his procedures for loss of subcooling ma rgin.,The steam leak may also be inside the reactor building (RB).
If so, it is likely that an SF.AS trip on high RB pressure will occur.
Steam leaks and other similar events will exhibit the sympton of excessive hrat transfer on the Rancho Seco Post-Trip P-T Display.
The control mom operator can begin taking the required actions.
3.7.
Loss of Coolant Accident Tha loss of coolan t accident (LOCA) is a complex and difficult accident to handle for the following reasons:
1.
A wide range of leak sizes is possible fiom small b rea.< s 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 system condition such as RCS saturation, interrupted natural
,cimulation, etc. may be a riafGral consequence of the event.
4.
Steam generator heat removal may be degraded.
5.
Hot standby conditions may not be a safe, stab 'e condition.
6.
The reactor building env i ronment can degrade due to increased i
pressure, temperature, humidity, and radf ation.
I LOCAs can be categorized by size.
Large break LOCAs result from a major failure in the primary system pressure boundary which depressurizes the RCS rapidly and almost completely.
Since this is a design basis event for which extensive analysis is performed in the FSAR, it was not analyzed for ATOG.
Saml1 break LOCAs are less severe than large bi eaks.
The RCS depressurization is much slower due to the lower mass flowrate out the break.
The Emergency Core Cooling Systen (ECCS) will maintain adequate core cooling throughout the tran sient.
For small breaks, the control mom operator plays a vital role in minimizing the consequences of. the accident.
They have been analyzed in great detail.
Small leakr are events where the loss of reactor coolant is within the capacity of the normal makeup system.
3-7
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n-
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e.
The primary objectives during a LOCA are to maintain core coolir.g, 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 the control room operator veriffes.
In addition, the control room op9rator must make sure that the reactor building cooling is functier.ing properly and that p
long-term cooling is established once the RCS is cooled down and d: pressurized.
1 In order-to determine actio ns requi red by the control roan operator to L
monitor, veri fy or perform manually, and to identify the expected RCS and system transient. response, five small break LOCAs were analyzed as part of the 7
Sm al l Break Operating Guidelines and were translated into the ATOG guidelines.17 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.
2.
Small breaks which stabilize at ap proximately the secondary system pressure with feedwater available.
3.
Small breaks which may repressurize the RCS in a saturated condition with feedwater available.
]
- 4., Small breaks without primary To secondary heat transfer.
5.
Small breaks within the pressurizer steam space.
These events were analyzed in sufficient detail to generate symptons for identifying type and location of the break as well as distinguishing a LOCA from other events, especially overcooling events.
Containment radiation and pressure alerts on the Rancho Seco SPDS in conjunction with the pressure-temperature relationship rt.sponse of the primary 1
and seconda ry systems on the SPDS Post-Trip P-T Display will provide the information necessary for the operator to identify a LOCA event.
He can then begin to take the required actions.
Additional analyses have been performed in order to generate the subcooling margin limits to be displayed on the three SPDS P-T displays and to generate N
the inadequate core cooling limits to be displayed on the SPDS ICC display.
- Those limits and algorithms are provided in reference 8.
Additional limits l
such as RCP NPSH, heatup and cooldown NDT, DHRS, fuel-in-compression were cs 3-8
.e..
already available and no further analyses were performed.
These limits and algorithms,are also provided in reference 8 and are available on the SPOS P-T displays.
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e i
3-9
4.
CONCLUSIONS t-Tha Rancho Seco SPOS has been shown to be designed to provide the control rom operators with sufficient key information to enable them to determine the safety sta tu s of the plant with respect to the five required functions discussed in Section 2.0.
The Rancho Seco SPDS has been shown to be designed to provide sufficient information in a display format based on and compatible with the sympte-oriented guidelines developed ~ in ATOG as discussed in Section 3.0.
With this key information and concise display format, the control room operators at Rancho Seco can monitor plant status, detect symptoms of abnormal plant response and take corrective actions designed to restors control function or mitigate the consequences of transients.and accidents in a rapid and reliable manner.
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w 5.
REFERENCES 1.
NUREG 0737, Supplement 1,
" Requirements for Eme rge ncy Response Capability" (Generic Letter No. 82-33), December 17, 1982.
2.
NUREG 0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident", May,1980.
3.
NUREG 0737, " Clarification of TMI Action Plan Requirements", Ncvembe r, 1980.
4.
NUREG 0696, " Functional Criteria for Emergency Response Facilities",
Feb rua ry, 1981.
5.
B&W Document 74-1127469, " Rancho Seco Nuclear Power Station Abnormal Transient Operating Guidelines, Part I, Part II-Vol.1, Part I -Vol.
2",
October 8,1982.
, Ipstrumentation for Light-Water-Cooled 6.
Regulatory Guide 1.97, Rev. 02, Nucie'ar Power Plants to Assess Plant and Environs Conditions During and Following an Accident", December,1980.
7.
B&W Drawing No. 1134042, " Rancho Seco Safety Paramter Display System Function Description".
8.
B&W Document 51-1121938, "SMUD Safety Parameter Display System (SPDS)".
9.
B&W Document NPGD-TM-414, Rev. 03, " TRAP 2 - FORTRAN Program for Digital l
Simulation of the Transient Behavior of the OTSG and Associated RCS, Rev.
l J", May, 1980, 1
10.
B&W Document 86-1127307, "ATOG Transient Information Document, Excessive Main Feedwater, Rancho Seco", September 3,1981.
11.
B&W Document 86-1127426, " Rancho Seco ATOG LOFW TID", August 20, 1981.
12.
B&W Document 86-1127102, "SMUD ATOG SSLB Transient Information Documentd, August 17, 1981.
13.
B&W Document 86-1126639, " Rancho Seco ATUG LOOP TID", October 8,1981.
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14.
B&W Document 86-1118041, "Results and Recommendations of the Main Success r
Path Analysis for SGTR ( ANO-1)", March 10,1980.
15.
B&W Document 86-1118045, " Impact of an OTSG Tube Ruoture with Concurrent r
LOOP", April 29,1980.
16.
B&W Document 86-1120490, "0TSG Tube Rupture Alternate Paths (ANO-1, f.
ATOG)", August 8, 1980.
17.
B&W Docume nt 74-1122501, " Operating guidelines for Small Breaks for f
Oconee 1,
2, and 3, Three Mile Island I and 2, Rancho Seco 1, and Arkansas 1", December 8,1980.
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