ML20134A030
| ML20134A030 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/29/1985 |
| From: | Wisenburg M HOUSTON LIGHTING & POWER CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| CON-#485-019, CON-#485-19 OL, ST-HL-AE-1480, NUDOCS 8511040069 | |
| Download: ML20134A030 (84) | |
Text
.
The Light COmpBM7 m><ium,ugamigmm miwx nm m>uumimxaa7mi mamsmii October 29, 1985 ST-HL-AE-1480 File No.: G9.17 Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Responses to DSER/FSAR Items Concerninn Chapter 7 on Safetv-Related Disolav Instrumentation
Dear Mr. Knighton:
The attachments enclosed provide STP's response to Draft Safety Evaluation Report (DSCR) or Final Safety Analysis Report (FSAR) items.
The item numbers listed below correspond to those assigned on STP's internal list of items for completion which includes open and confirmatory DSER items, STP FSAR open items and open NRC questions. This list was given to your Mr. N. Prasad Kadambi on October 8, 1985 by our Mr. M. E.
Powell.
The attachment includes mark-ups of FSAR pages which will be incorporated in a future FSAR amendment unless otherwise noted below.
The items which are attached to this letter are:
Attachment Item No.* Subject 1 F 7.5-1 Safety-Related Display F 7.5-21 Instrumentation F 7.5-22 1
10 h p
- Legend 6 D DSER Open Item C - DSER Confirmatory Item F - FSAR Open Item Q - FSAR Question Response Item l i
Ll/DSER/aav l
'Il
7 Houston Lighting & Power Company ST-HL-AE-1480 Fil,e No.: G9.17 Page 2 If you should have any questions concerning this matter, please contact Mr. Powell at-(713) 993-1328.
Very truly yours, l
M. R. Wiseqburg Manager,NuilearLicens[ing MEP/bl-Attachments: See above w
L1/DSER/aav ,
ST-HL-AE-1480 C File No.: G9.17 Page 3 cc:
Hugh L. Thompson, Jr., Director Brian E. Berwick, Esquire Division of Licensing Assistant Attorney General for Office of Nuclear Reactor Regulation the State of Texas U.S. Nuclear Regulatory Commission P.O. Box 12548, capitol Station Washington, DC 20555 Austin, TX 78711 Robert D. Martin Lanny A. Sinkin Regional Administrator, Region IV 3022 Porter Street, N.W. #304
' Nuclear Regulatory Commission Washington, DC 20008 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Oreste R. Pirfo,_ Esquire
. Hearing Attorney N. Prasad Kadambi, Project Manager Office of the Executive Legal Director
.U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Washington, DC 20555 Bethesda, MD 20814 Charles Bechhoefer, Esquire Claude E. Johnson Chairman, Atomic Safety &
Senior Resident Inspector /STP Licensing Board c/o U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission ' Washington, DC 20555 P.O. Box 910 Bay City, TX 77414 Dr. James C. Lamb, III 313 Woodhaven Road M.D. Schwarz, Jr., Esquire Chapel Hill, NC 27514 Baker & Botts One Shell Plaza
' Judge Frederick J.~Shon Houston, TX 77002 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission J.R. Newman, Esquire Washington, DC 20555 Newman & Holtzinger, P.C.
1615 L Street, N.W. Mr. Ray Goldstein, Esquire l Washington, DC 20036 1001 Vaughn Building 807 Brazos Director, Office of Inspection Austin, TX 78701
- and Enforcement.
'- U.S. Nuclear Regulatory Commission Citizens for Equitable Utilities, Inc.
Washington, DC 20555 .c/o Ms. Peggy Buchern Route 1, Box 1684 E.R. Brooks /R.L. Range Brazoria, TX 77422 Central Power & Light Company l
P.O. Box 2121 Docketing & Service Section l Corpus Christi, TX 78403 office of the Secretary U.S. Nuclear Regulatory Commission H.L. Peterson/G. Pokorny Washington, DC 20555 City of Austin (3 Copies)
P.O. Box 1088 Austin, TX 78767 Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission i J.B. Poston/A. vonRosenberg 1717 H Street l City Public Service Board Washington, DC 20555 j P.O. Box 1771
! San Antonio, TX 78296 l Revised 9/25/85 L1/DSER/aav I:
ATTACHMENT I STP FSAR ST-HL AE 14fD PAGE r OF gl TABLE 1.1-1-(Continued)
ACRONYMS USED~IN THE FSAR LPG liquid petroleum gas LPMS .
Loose Parts Monitoring System l39 LPZ low population zone ~
LRPT lead radiation protection technician 38 LRTS Liquid Radwaste Treatment System LSA low specific activity LTMD less than minimum detectable (concentration)
LVDT linear variable differential transformer LWPS Liquid Waste Processing System MAB Mechanical Auxiliary Building MCARS Main Condenser Air Removal System i l39 MCB main control board MCC motor control center MCR Main Cooling Reservoir l35 MDC moderator density coefficient MDWS Makeup Demineralized Water System MEAB Mechanical-Electrical Auxiliaries Building MEB Mechanical Engineering Branch Mg7 N eT.eerolo3mm) 53 stem g MFIV main feedwater isolation valves 41 MG motor generator MIL military standards MLW mean low water MOL . middle-of-life MOV . motor-operated valve ME: maximur permissible concentration 1.1 -11. Are nc --
~
ATTACHMENT I ST.HL AE- N P PAGE 7- OF 3)
Qusation 032.17 j State your conformance with Regulatory Guide 1.97 " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident." Justify any exception taken.
, Response As stated in Table 3.12-1, STP conforms to the intent of Regulatory Guide 1.97, Revision 2 (12/80)4eith..gt the guid; is ee: pplicati :; ST" I;; :c 40 it 4 71 --- t st i-- data A detailed discussion of post-accident monitoring instrumentation is presented in Section 7.5 and Appendix 7B.
0 1
Q&R 7.5-1 ^*d*! '
,=,
4 TABLE 3.12-1 (Continued)
REGUIATORT GUIDE NATRIK RECUI.ATORY CUTDE
,1uf. TITLE FSAR REFERENCE REVISION STATUS STATUS ON STF 1.97 Inst rumentation for Light-Water-Cooled Nuclear Power Table 7.1-1 Rev 2 (8/77) '5 See Note a 64 40 N Flsnts to Assess Plant conditions During and Figure 7.1-1 Pollowing an Accident Table 7.5-1 .
App. 7A App. 75 I.98 Assumptions Used for Evaluating the Fotentist WA See Note 1 Radiological Consequences of a Radioactive Offges System Failure in a Bolling Water Reactor 1.99 Ef f ects of Residual Elements on Fredicted Radiation 5.3.2.1 Rev 1 (4/77) D See Note 57 l33 Damage to Reactor Yessel Materists en 1.100 Seismic Qualification of Electric Equipment for Table 7.1-1 Rev 1 (8/77) .5 See Note' 3 l36 Q
,, Nuclear Power Plants Figure 7.1-1 m
- . 3.10.1 3.10.2.2 ,o )g k
3.10.2.2.1 43 3* q 3.10.2.2.2.2 3, 3.10.4.1 .O d)> I 1.101 Fmergency F1saning for Nuclear Power Flants 9 5.1.A pg 18 withdrawn See Note 37 l32 %
1.102 Flood Protection for Nuclear Power Fleets 3.4 Rev 1 (9/76) A k-4 3.8.4.2.3 33 % O_
l.103 root-Tensioned Frestressing Systems for Concrete 3.8.1.2.2 Rev 1 (10/76) A l45 Pentter Vessels and Containments 3.8.1.6.5.1 1.104 ovenbead Crane Band 11ag Systems for Nuclear Power 9.1.4.3.1.6 Rev 0 (2/76) FC 3 See Note 34 ly3 riants Table 9.1-3 1.10's Instrument Setpoints Table 7.1-1 Rev 1 (11/76) B See Note 3 Figure 7.1-1 ,
Note 28 l33 q 1.106 Therral Overload Protection for Electric Motors on 8.3.1.2.12 Rev 1 (3/77) A See Note 14 'I
~, Hot or-operated Valves 8.3.2.2.7 l18 l Q430.
u 130N 1.107 S'nlifications for Cer?nt Crouting for Frestressing NA See Note 11 l23 1rndone in Containment Structures t.in4 r.viedic Testing of Diesel Ceneratore Used as Onsite 8.3.1.1.4.7 Rev 1 (8-77) C See Note 40 l38 t lert ric Fever Systems at Nuclear Power Plante 8.3.1.2.10
- 1. t fiq r itrulat ions of Annust Doses to Man From Routine ll.A.1 Rev 0 (3/76) FC A Pelenses of Resctor Ef fluents for the Furpose of 12.4.2 Rev 1 (10/77) l32 t vnluating Compliance with 10CFR$0, Appendix I
ATTACHMENT 1 STP FSAR ST-HL-AE N80 PAGESt OF fl TABLE 3.12-1 (Cont'd.)
! REGULATORY GUIDE MATRIX NOTES -
If a work activity and contract is for a two-month period or less, 43 an audit is not necessary when a facility preaward audit has been conducted.
,: The QA program for operations will conform to the requirements of RG 1.94 45 Revision 1, with the same clarification:
- 55. Refer to Sections 3.7.4.1 and 3.7.4.2 for the discussion on seismic instrumentation.
- 56. Refer to Section 5.2.3.3.2 for Westinghouse alternate approach to RG 1.71. Also, refer to Section 10.3.6.2, for the BOP conformance to RG 1.71.
- 59. Revision 0 is utilized during the construction phase for RG 1.58, Posi-tions C.5, C.6, C.7, C.8, and C.10 of Rev. 1 are also utilized. 33 1
- 60. With respect to Section 3.1.2 of ANSI N45.2.3-1973 HL&P interprets the
) lighting level of 100 footcandles to be guidance. It is HL&P's normal practice that the lighting level for determining " metal clean" of acces-sible surfaces of piping and components is determined by the inspector.
Typically he uses a standard two-cell flashlight supplemented by other lighting as he deems necessary.
- 61. STP conforms to RG 1.140, with the exception that instrumentation is provided only to monitor and alarm pertinent pressure drops at critical points in the duct. Therefore, STP is in partial compliance with Position C.2.cc.
38
. 62. RG 1.1 as clarified by NUREG 75/087.
. 63. The basis for meeting the intent of RG 1.46 is the implementation of NRC Branch Technical Position (BTP) MEB 3-1, NRC BTP ASB 3-1, WCAP-8082-P-A, and WCAP-8172-A. Tables 3.6.1-2 and 3.6.1-3 provide a summary of the compliance with MEB 3-1 and ASB 3-1. -
40
- 64. Eweepti;;; w re -isic,n 2 cre identified in -TS".%. eefer;;;; :: ti;;;.--C"~
- 65. The QA program during operations will conform to the requirements of Revision 2.
- 66. The QA program for operations will conform to the requirements of 43 Position C.2 of Regulatory Guide 1.137 with the following clarification:
%e 7 514,i,I Tks. clise:=En ef GTP ca Nrmsace,. -le & G l.97 AM.2. is presukul i3 &ncks M.
4s e+fW in Appucks 78, ;ylmbb, s/ RC 1.97 reg uirM sa , ;4yaled w.'& Os.culvel coe,= cMy twisa a-cl a pufr< red g & k/ TAW Amendment 45 i Owas s Cray Fe% Acapua.Cujf,.12-24Dna% a.d emb w,*& & kW / & Ag,
___.____.-.___u___.__
ST-HL-AE- N30 PAGE G OF S t 7.5 Safety-Related Display Instrumentation 7.5.1 Post Accident Monitoring Instrumentation 7.5.1.1 Description. A task analysis was conducted to identify the appropriate variables and establish appropriate design bases and qualification criteria for, instrumentation employed by the operator for monitoring condi-tions in the Reactor Coolant System (RCS), the secondary heat removal system,
-and,the Containment, including Engineered Safety Features (ESF) and other 40 systems normally employed for attaining and maintaining a safe shntdown condition. The instrumentation is used by the operators to monitor the South Texas Plant throughout various opp ating conditions including anticipated operational occurrences and possaccident conditions. The analysis process )(
ensures that the information available to the operator following an accident is derived from specially designed and qualified instrumentation installed at the plant.
7.5.1.2 Analys *.s . The task analysis performed in response to Regulatory Guide (RG) 1.97 is described in Appendix 7B. Table 7.5-1 provides 40 a listing of the variables identified in the task analysis. In addition, the Q32.
table includes the following information on the STP instrumentation utilized 18 for each variable: (a) instrument range; (b) type and category (per the defi-nitions found in Appendix 7B); (c) environmental qualification; (d) seismic qualification; (e) number of channels available; (f) display device and loca-tion; (g) the schedule for implementation; (h) power supply; and (i) a state-ment of conformance to RC 1.97, Revision 2, or justification for deviations.
Seismic and environmental qualification3sare further discussed in Section$ 3.10 )(
and 3.11.
To assist in understanding the information provided in Table 7.5-1, the fol-lowing explanation of column headings is provided:
Variable: This column contains the RG 1.97 variable as defined in Appendix 7B.
Range / Status: This column contains the range of instruments used on STP for RG 1.97 purposes and a description of STP indications of valv,e position or 40 pump status. The ranges indicated meet or exceed the requirements described in Appendix 7B.
Type / Category: This column contains the types and categories applicable to each variable as defined in Appendix 7B.
Indicate,5 Environmental and Seismic Qualification: This column 3::::i5:: whether or not )(
the STP instrumentation is seismically or environaentally qualified. A "yes" in the Environmental Qualification column indicates that the channel is envi-ronmentally qualified to a level which meets or exceeds the requirements spec-ified in Appendix 7B for that variable.
Number of Channels: This je lumn contains the number of instrument channels available on STP for post 5 accident monitoring purposes. This column does not )(
take into account control room indication or recording capability. The number 7.5-1 Amendment 40
M e & MLMMLIN i j STP FSAR l ST HL-AE- Nfo l
'PAGEfo OF $ '
1
\
cf channale cveilcbla ments or cxceeds the rcquir m:nts in App ndix 7B, except for the cases in which justification for deviation from Appendix 7B is l provided in Table 7.5-1.
- g,jggg7}g' t
Control Room Indication: This column describes th control room indication '
cnd recording capability on STP for each variable. The control room indica- N tion and recording capability meets or exceed $the requirements described in )(
Appendix 7B. ,
4 Implementation Date: This column contains the STP schedule for implementing the RG.
Power Supply: This column describes the power supply which powers the STP instrumentation for each variable. The power supply provided meets or exceeds the requirements described in Appendix 7B.
cewter Eoc. Fen A* 5 Emergency Operations Facilit:. (.EGE) Indication: This column describe: the STP I Emergency Operations Facility indication capability for each variable. /
Ceniee p.ch Technical Support Center (TSC) Indication: This column d::crih:2 the STP X Technical Support Center indication capability for each variable.
Conformance: This co'umn provides a statement of the conformance to RG 1.97, Revision 2 or justification for deviation.
Further information concerning conformence to RG 1.97, Revision 2 is provided in Appendix 7B, which describes (a) the plant accident conditions under which 40 the instrumentation must be operable; (b) the selection criteria (Type A, B, C, D, or E); (c) the qualification criteria (Category 1, 2, or 3): (d) the design criteria (number of channels, power _ requirements, servicing requirements, etc); and (e) the processing and display criteria (accessibility, historical record, etc.).
The postd5ccident monitoring instrumentation consists of the instrumentation N identified in Table 7.5-1. The display systems for the post-accident monitoring instrumentation are identified in Table 7.5-1 and are further described in the sections identified below:
- 1. Qualified Display Processing System (QDPS) - Section 7.5.6.
- 2. Emergency Response Facilities Data Acquisition and Display System (ERFDADS) - Section 7.5.7.
- 3. Radiation Monitoring System (RMS) - Section 11.5. later N A
7.5.2 Reactor Trip System Display instrumentation for monitoring during normal operation in the Reactor Trip System is discussed in Sections 7.2 and 7.7.
7.5.3 Safe Shutdown Display instrumentation provided for monitoring safe shutdown during normal operations is discussed in Section 7.4. b 7.5-2 Amendment 40
a.n, ~ m h p
~
UDER$ f hlglAENfo' M $,' 7 k "% OP 'ADPs" - -
,k vd & L_
u
& d) Qaskfloa k ~
g mv %
(
athnLAs LA (ao ?s) '
l 4Hha L o r L M l]
sh autwmo
/
l
ATTACHMENII ST HL AE IW' STP FSAR PAGE t OF 3l r
. Visual indication (t(ough lampbox lights) that specific ESF equipment has g
-- 1.
been bypassed or deliberately rendered inoperable during normal plant operating modes. '
- 2. Annunciation to alert the operator that an ESF system or any of its support systems has been bypassed or deliberately rendered inoperable during normal plant operating modes.
The bypa,ss/ inoperable status indication subsystem continuously monitors the status of field' contacts and automatically indicates that a specific piece of ESF equipment has been bypassed or deliberately rendered inoperable. The fol-lowing conditions (as applicable) are automatically detected for each moni .
tored component of the ESF systems:
- 1. Loss of control power l
- 2. Control handswitch in pull-to-lock position l
- 3. Circuit breaker no't in operating position
- 4. Control transferred from the control room to a remote panel 5 .' Component not-in its proper aligned position The bypass / inoperable status indication is accomplished by lighting up the l 41 component level window. This indication also provides individual system level !
annunciation to alert the centrol room operator that an ESF system has been
. bypassed or rendered inoperable.
In accordance with RG 1.47, bypass or inoperable status indication is provided automatically for conditions which meet all three of the following guidelines:
- 1. The bypass or inoperable condition affects a system that is designed to automatically perform a safety-related function.
- 2. The bypass is utilized by plant personnel or the inoperable condition can reasonably be expected to occur more frequently than once per year and,-
- 3. The bypass or inoperable condition is expected to occur when the affected system is.normally required to be operable.
Deliberate manual actions which render ESF actuated components and devices inoperable (once a year or more frequently) are automatically displayed on a component level. -Active components not directly' actuated by ESF signal but rendered inoperative once a year or more frequently such that it compromises the safety functions of the ESF system are also automatically displayed on a component level to the control room operator.
Rendering a piece of ESF equipment inoperative through the use of features provided strictly for infrequent maintenance (less than once a year) is not automatically indicated. Such maintenance features may include manual valves provided for isolation of the equipment for repair and electrical cable con-nections, screw terminals or manual disconnects. The bypass / inoperable indi-cation of these conditions is manually initiated on an ESF system level.
l 7.5-4 Amendment 45 1
l H
ATTACHMENT i ST HL AE-N80 STP FSAR PAGE 9 OF PI The capability for initiating a manual bypass indication and larm is _I provided via a system level manual bypass switch to indicate the bypass /in-oper?ble condition to the operator for those components or conditions which ;g are not automatically monitored.
Manual bypass / inoperable indication may be ' set up or removed under administra-
-tive control.- The automatic indication feature of the ESF Status Monitoring System can not be removed by operator action. -
Bypssand[orstatusindicationonasystemlevelisprovidedforthefol-lowing safety-related systems:
~
- 1. Solid-State Protection System (bypass / inoperable only) 41 2 .~ Safety Injection System (SIS) (including RHR system components required
~P for accident mitigation or safe shutdown)
- 3. Containment Spray System (CSS)
- 4. Containment-Isolation Phase A
- 5. Containment Ventilation Isolation 41 Q3
- 7. Combustible Gas Control System (bypass / inoperable only)
- 8. Containment Heat Removal System
- 9. Fuel Handling Building (FHB) Iheating,3 ventilating, and Iaic nditioning X (HVAC) Exhaust Subsystem
- 10. Electrical Penetration Space HVAC System
- 11. Control Room Envelope and Electrical Auxiliary Building (EAB) Main Area HVAC System
- 12. Feedwater Isolation 41
.13.. Steam Line Isolation d 14. Auxiliary Feedwater System (AFWS) b( cggc.y rg is ,_ [g n i @ e., ,. Ed Af fmL t k v The following support systems activate bypass indication of all supported safety systems. listed above when they are bypassed or rendered inoperable:
2
- 1. _ Component Cooling Water System (CCWS)
- 2. Essential Cooling Water System (ECWS)
- 3. ESF Bus System (including the standby diesel generators and the ESF load sequencers) l 4. Essential Chilled Water System 7.5-4a Amendment 41
STP FSAR ATTACHMENLl ST-HL-AE NF PAGE(o 07 SI
- 5. Supportit.g HVAC equipment The ESF Status Monitoring System is not required to operate during or after a l4I , ,
design basis seismic event; however, the indicator light panels are mounted on the seismically designed and qualified control benchboard. The indicator panels are designed and have been type tested to prove their structural integ-rity. .
No credit is taken in the accident analyses of Chapter 15 for the ope.rability of the ESF Status Monitoring System. The system is not designed to safety-re-lated requirements. Interfaces with safety-grade equipment are through qual- 41 ified isolation devices, in accordance with IEEE 384 and RG 1.75. These iso-lation devices are part of the Emergency Response Facilities Data Acquisition and Display System (ERFDADS) (see Section 7.5.7).
7.5.5 M: = 1 Op ucti;n: M;nitnin;Y(Deleted) I TM; infomatien haa teen p a ... Tobl 7.5 [ )(
~7.5.6 Qualified Display Processing System 7.5.6.1 Descripti5n. The QDPS is an integrated system designed to perform the following functions:
- 1. Data acquisition and qualified displays for post-accident monitoring.
- 2. Safety grade control and position indication of several safety-related valves.
- 3. Data acquisition, display, and control to cddress the separation require-ments of the STP design approach to a control room (CR) or relay room (RR) fire.
4 Steam fenerator 1/arrov )(ange y'ater-} level compensation for the ef f ect of 40 y temperature changes in the reference leg fluid. I 5 ' "9 ' " 8 ' " "' ' "' " 8 "" ' 1 " " 4 " ' * * " * * " "" * " '" ' "' '* ") ,
7.5.6.1.1 System Description. The system functions are performed by '
several subsystems. These subsystems, though related, have sufficient inde-pendence such that the individual functions can be performed with maximum reliability and minimum unnecessary interaction between functions. A block diagram indicating the interconnections of the various QDPS subsystems as well as interfaces with other systems is provided in Figure 7.5.6-1.
7.5.6.1.1.1 Data Acquisition and Qualified Display for Post Accident Monitoring - The data acquisition and qualified display function is performed by a subsystem referred to as Plant Safety Monitoring System (PSMS). It is a modular and flexible general purpose system which performs the following func-tions:
- 1. Implements qualified monitoring channels to comply with post-accident jX monitoring gategory / equipment design and qualification criteria defined in Appendix 7B. ,i 7.5-4b Amendment 41
ATTACHMENI (
2 STP FSAR ST HLt t AE PAGE OF l@$
- 2. Provides safety grade signal processing for instrumentation to detect inadequate core cooling as defined in NUREG-0737 Item II.F.2. This includes signal processing for:
e Reactor vessel water level e Core exit temperature ,
(;gyg e. *fp*phk 78.lT,f.2 MAO & L.00 O m s' w .
- 3. IsolatefClass1Eandassociatedsignalstomakethemavailableto )(
non-Class 1E equipment including the Emergency Response Facilities Data Acquisition and Display System (ERFDADS) (see Section 7.5.7).
- 4. Providefconsolidated, unambiguous, human-factoreddisplaysofappropriate X arameters to address the requirements of paragraph 4.20 of IEEE 279-1971, e Figure 7.5.6-2 b6 Sche *% rgresenhh of synal processi q ^ g
@ Jispla3 cowsob ien .
The PSMS consists of four redundant, channelized,ClassdEdataacquisition T processors called remote processing units (RPUs). These RPUs send data to redundant data)ase processing innits (DPUs), which subsequently provide infor- Y mation to the operator via plasma e splay modules. A fifth, non-Class IE RPU (RPU N) provides data acquisition for non-Class 1E signals which are needed to complete logical graphic displays. The RPUs perform the engineering units L conversion, limit checks, and isolation or buffering required. The DPUs performhensor algorithms and auctioneering functionl en output the data {0 4 base to the plasma display modules. The plasma' display modules provide graphic and alphanumeric display pages containing comprehensive, human engi-neered display information. Display page selection is performed using a fune-tion keyboard for each display module.
The variables required in the PSMS database are categorized into three types:
- 1. Safety grade parameters required to address post-accident and safe shutdown monitoring requirements.
- 2. Variables identified for monitoring the minimum functions required to achieve safe shutdown under postulated fire conditions.
- 3. Parameters included for display consolidation on the main control panels.
7.5.6.1.1.2 Safety Grade Control and Position Indication of Safety-Related Valves - The safety grade valve control function is perform (d by a microprocessor based control syste. This consists of a set of Classs1E )(
equipment used to provide the'followin process enntrol functions:
Eagle 23
- 1. Closed-loop control and position indication for the steam generator (SG) power-operated relief valves (PORV).
I
- 2. Contact output signals for automatic control and position indication of auxiliary feedwater (fv) er- m l valves within upper and lower flow K limits. t i A):y/ Oou,throMe, .
Amendment 40
ATTACHMEh@T ST HL AE l' u STP FSAR ..fAGE & OFfl
- 3. Open-loop control and position indication for the reactor vessel head vent valves. . .
4-
- N C5 F M X Closect-leaIs.co eet & pe me\ mM. foSfnon Indit4% n for EW flo w C8 " **l ,
The SG FORV control equipment provides hardware to meet the requirements for x fullf valve control including transfer, without position change, of operation from the control room to the auxiliary shutdown panel. A separate transfer M cwitch selects the. active control station. Each control loop accepts the N cteam line pressure, valve position, and the setpoints as input variables and cutputs a 4-20 mA signal to control the valve. ,
Each auxiliary fee'dwater throttle vaJ1 e control loop accepts an input from a flow-transmitterandsuppliestwobigstableoutputsignals,lowandhigh X limit {,tothevalvecontroller. These signals maintain- - ___ , _ flow X within acceptable limits until manual control is assumed by the operator.
APr/
The reactor vessel head vent nerol loop accepts signal inputs from a pair of manual stations, one located in the control room and the other on the auxiliary shutdown panel. A separate transfer switch for each loop selects the active re m el station. X g 4- rnual 7.5.6.1.1.3 Data Acquisition, Display, and Control to Address Separation Requirements of the STP Design Approach to a CR or RR Fire - Signal buffering to meet fire protection isolation and separation requirements is achieved by using microprocessor based equipment, which provides interface with the NSSS process protection and control cabinets.
Field inputs for variables identified for monitoring the minimum functions 40 required to achieve safe shutdown following a CR or RR fire are routed to the QDPS auxiliary process cabinets (APCS). The signals are split into two inde-pendently buffered outputs. One of these outputs is routed to the process protection or control cabinets, and the other serves as an input to the RPU (see Figure 7.5.6-3). With this configuration, the QDPS displays of these parameters are available should any failure occur in the process protection or control cabinets or input and output cabling.
gt 7.5.6.1.1.4 Steam Generator Narrow Range 3 Level Compensatien and X Display - The gteam penerator )(aryg ifange Vat'e'r yevel 7ompensation system automaticallycompensatestheSG$levelsignalsfortheeffect of temperature Y changes in the reference leg fluid. This system serves to increase operating margin and to improve the accuracy of post-accident e level indications. With referenceleg-temperatureeg*pensationoftheS6[evelsignals,therequired 'd increaseinthelow-lows (*levelreactortripsetpoint to account for refer- X ence leg heat-up following a high energy line break inside containment is minimized. The compensation system is designed to limit the reference leg heatup error to 2 percent of the level instrument span. SG levels are dis-s played on the indicators. For X ry additional information, (j QPDG plasma displays refer to Section and 7.2. on main control pane water x
-Y m QOf5
/ 7.5.6.1.2 Equipment
Description:
The QDPS consists of the following equipment: four Class 1E APCS, two Class 1E dajabase processing units, eight Class IE plasma display units, three non-Class 61E demultiplexer units, and one X non-Class 1E RPU. Refer to Figure 7.5.6-1 for system configuration.
7.5-6 Amendment 40
j - . AI TACHMEMT 8 ST HL AE 14fD
,f.afaE 13 OF %l
$wsex.T s*
g }/a. eSen ' NVK k g c mt> con e7
- y/ The usa M &coolm +k(r&W c~pI p p,pJ pa,ak '
nM h J % y A .p,jattand A / A ---4 y i} k A =g n n.s , ,,::v.n. -rL- W ,,, L rs.
- 1=K4y & k = =,45 a,-M p:% a y-, ay' yx L th = = o.~6,/ ,,- k -is --::^ y ac asch ~ H . i
= % =.;Wm:_.
A & W ew,%4 u/~ C N f*J &/hp 4 fL- =- A s f = O L , a,,. M pm, ":: =Mf<4 Me y rw sW, & f a *
- A~
& & Mol4 n x l = ? n .~ A J Q u 4-204L c.g /, A -t4 w.
e .
- 5 9
- E
}
l l
M* *We _
- l PAGEI4 of gl
$dSELr- k z*
'F1 L RCS Ty RTO si 4 _
l (h) i 4
ATTACHMENU STP GAR ST HL AE IW PAGEg4 OF f!l 1.5.6.1.2.1 Auxiliary Process Cabinets - The four redundant APCS comply with IEEE 279-1971. Each channelized APC contains an RPU chassis, control system chassis, signal isolation / buffering equipment and associated DC power supplies for field inputs originating from this respective instrument channel.
Data is outputg via datalinks and individual analog signals as required. Each
) g ""t:rtalink is iiidependently buffered such that no fault on a datalink will de-P grade system. function beyond loss of data on that link. The APCS are located in four physically separated fire areas, such that no single fire will affect more than one APC. The APCS are powered from the four separate 120 vac vital instrument buses.
7.5.6.1.2.2 Database Processing Units (DPU) - The two redundant DPUs comply with IEEE 279-1971. Each DPU contains signal processing equipment, signal isolation / buffering equipment and the DC power supply. The DPUs re-ceive data inputs from each of the RPUsg nd transmit data outputs to the Class 1Eplasmadisplayunits,non-Class 1FhMUE,andotherdestinationsasneces- T sary. Each datalink is buffered such that no fault on a datalink will degrade system function beyond loss of information carried on that link. The DPUs are located in physically separated rooms with she separation group A and C APCS, and are powered by the separation group A and C f 120 vac vital instrument y buses, respectively. Seper tier group: ^.(end C ccrresp nd t instruncnt chan nc1; I and I" scapcctively. j % ] neget M
%ed As*
7.5.6.1.2.3 Plasma Displav Units - The eight plasma display units are grouped into two redundant sets of three display units each in the control room and the two redundant display units on the auxiliary shutdown panel. The 40 plasma display units conform to IEEE 279-1971. Each plasma display unit con-tains the microprocessor equipment and DC power supplykto receive data from gg, each DPU and generate graphic and alphanumeric display pages. A function keyboard attached to each display unitfallows operator selection of specific Y display pages. One redundant set of plasma display units is powered by the separation group A 120 vac vital instrument bus and the other set by separa-tion group C vital instrument bus. X CO 903 7.5.6.1.2.4 Demultiplexers - Two of the three DMUX units are located in the control room. The third DMUX unit is located in the auxiliary shutdown panel. The DMUX units are non-Class 1E devices which provide system outputs to drive analog panel meters and recorders. The units are seismically quali-fled in accordance with IEEE 344-1975 such that the recorder output will re-main functional following a seismic event. The DMUX units are powered from the non-Class 1E instrument bus backed-up by station batteries. X k 1;LO ya.c Mod 7.5.6.1.2.5 Remote Processing Unit N (RPU N) -
The single non-Class 1E RPU N provides data acquisition for certain non-Class 1E signals. The RFU is not required to function post-accident and is not redundant. RPU N is located in the relay roorthlectrical quxiliary guilding[EA glevation35kandis powered from the non-Class 1E instrument bus W
(, g m g A bat.ked up % sfcMon bodentS .
7.5.6.2 Analysis. Even though IEEE 279-1971 was not a design basis of the QDPS, an analysis was conducted to determine those criteria stated in the g standard that were met by the system design. The following sections discuss the applicability of the QDP,S to the respective sections of IEEE 279-1971. In performing this evaluation the functions performed by the QDPS are subdivided 7.5-7 Amendment 49
(
$Es#0E9h' PAGElip OF 8l SwseRt
- t. p To the. W tabase Processm3 U n ct 3, non - dass 1E DMV X tAniis aw A E R FDAps,
~
9 M
awalog outputs to conunha\ in h cdons awol re.c.orAc.rs aw A co wT ac t outputs to provihe- goa\i Re.A s taius iw formation .
4r (Sv - A r y . ~ ai ~ a. i., s,m A e. ,.,,,,)
m
- - - - + - , ---w --- -y -
ATTACHMENT l STP FSAR ST-HL.AE- MP PAGE Ft OF U hvoMO into the following subgroups: (a) steam generator water level compensation I
cystem (SGVLCS), (b) ESF qualified controllers (e.g., auxiliary feadwater (AFW regul:::r calve control), (c) qualified controllers utilized for [
ac eving a safe shutdown, and (d) post-accident monitoring displays.
References to the QDPS from a system level in the succeeding discussion indicates that all QDPS subsystems meet the stated requirement. Furthermore,
~
49 the applicability of the General Design Criteria (GDGs) are indicated below.
7.5.6.2.1 General Functional Requirement: This criteria only app'.ies to the SCVLCS -and the'ESF qualified controllers. Other functions do not automat-ically initiate appropriate protective action.
7.5.6.2.2 Single-Failure Criterion: The QDPS is designed to provide cdundant instrument channels for each safety-grade function as described in 40 Section 7.5.6.1. These redundant channels are electrically and physically independent. A single failure in the QDPS will not prevent proper response at the system level. The loss of power to any vital instrument bus will result ct most in loss of display from one channel. A failure modes and effects l49 enalysis has been performed and is presented in Table 7.5-4. The design meets the requirements of GDC 21, 22, and 23. 40 7.5.6.2.3 Quality of Components and Modules: The QDPS meets the 99 percent availability requirement defined in NUREG-0696 Section 1.5 under all pressure and temperature conditions exceeding cold shutdown conditions.
l 49 7.5.6.2.4 Equipment Qualification: The QDPS is seismically and environ- 40 centally qualified to IEEE 344-1975 and IEEE 323-1974, and meets the require-ments of GDC 2 and 4 with the exception of RPU N which performs non Class 1E functions. The DMUX units are seismically qualifiad. Equipment qualification is also discussed in Sections 3.10 and 3.11. 49 7.5.6.2.5 Channel Integrity: The QDPS is designed to operate during cccident conditions and maintain necessary functional capability and accuracy under extremes of conditions relating to environment, energy supply, malfunc-tions, and accidents. 40 7.5.6.2.6 Channel Independence: Channels that provide signals for the l 49
=ame function are electrically independent and physically separated to accom- i plish decoupling of the effects of unsafe environmental fectors, electric l 40 transients, and physical accident consequences. The system is designed to l 49 minimize the potential for interactions between channels during maintenance i 40 operations or in the event of channel malfunction. The QDPS features two redundant physically separated independent trains of display. The design cnsures that an initiating failure (short circuit, fault, etc.) in either a DPU or display unit will not result in the loss of both trains of DPUs and/or 49 display units. The design meets the requirements of GDC 22.
7.5.6.2.7 Control and Protection System Interaction: The only subsystem that is used for both protective and control functions is SGVLCS. Further-more, control grade signals are output from the post-acident monitoring QDPS subsystem.
7.5-8 Amendment 49
ATTACHMENT f STP FSAR ST-HL AE Me PAGED OF 8l In all cases the transmission of signals from the QDPS for control or use by other non-Class 1E devices is through qualified isolation devices which are part of the QDPS system. Faults, such as short circuits, open circuits, l ground, or the application of credible AC or DC fault potential at the output j of an isolation device, will not prevent the associated protection system 149 channel from m,eeting minimum performance requirements.
g ef**$ g Noise and isolation testing will be addressed in a WCAP,iv bs . b.itted ir tF- ,
,.. _.. .. . c ,noe This design meets the requirements of GDC 24 40 13erivation
": ! tier of System Inputs: To the maximum extent practica-7.5.6.2.8 ]49 /k ble, the QDPS inputs are derived from signals that are direct measures of the monitored variables. co, pow,gt 4yug, 40 7.5.6.2.9 Capability for Sensor Checks: The QDPS bas built-in diagnos-tics for checking the operational availability of each system nput sensor K during reactor operation. This is achieved by continuous scanning by micro-processor b'ased sensor data quality checks. A data quality is assigned to all rede channels of data input. The routine processes the redundant sensor X inputs and, when possible, returns a group value of the valid sensors for use in the upper level displays.
7.5.6.2.10 Capability I'or Test and Calibration: The SGWLCS and ESF qualifieo controllers have the capability for testing and calibration during reactor operation. The post-accident monitoring subsystem has the capability for checking the operational availability for each channel during reactor operation by cross checking between channels that bear a known relationship to each other. The safe shutdown qualified controllers are only required to be tested during scheduled station shutdowns. Refer to Section 7.2.2.2.3.10 for
! a description of the tesrf g of the protection loops. The design meets the 49 requirements of GDC 21.
7.5.6.2.11 Channel Bypass or Removal from Operation: The SGWLCS sub-system is designed to permit all channels, one at a time, to be maintained, ,
tested, or calibrated during power operation with no loss of. safety function.
The ESF qualified controllers are designed to permit all channels, one at a time, to be maintained, tested, or calibrated during power operation. Access to the cabinets for removing channels from service is administrative 1y con-trolled.
7.5.6.2.12 Operating Bypasses: There are no operating bypasses in QDPS. 40 7.5.6.2.13 Indication of Bypasses: If one or more channels of the ESF qualified controllers have been deliberately rendered inoperable, this fact 49 will be continuously indicated on the QDPS display. If one or more channels of the SCWLCS subsystem has been deliberately rendered inoperable in the QDPS hardware, the action will result in the partial trip of the respective chan-nel.
7.5-9 Amendment 49
ATTACHMEN STP FSAR ST-HL AE l@T I PAGEI4 OF 8l i
- 2. Data acquisition and signal processing for the ESF Status Monitoring l System. ,
1
- 3. Data acquisition and signal processing for other normal plant monitoring j systems including the plant annunciators and the plant computer. -
7.5.7.1.1 System
Description:
The ERFDADS functions are performed by {
several subsystems. Data acquisition is provided by multiplexers within the l ERFDADS, the-QDPS (see Section 7.5.6), the Meteorological (ME ) System and #
the Radiation Monitoring System (RMS) (see Section 11.5 ). The ERF com- /
puter-performs the required data processing. CRT devices provide display in the control room (CR), Technical Support Center (TSC) and the o A simpli- X fied interconnection diagram of the ERFDADS is shown in Figure 7.5.7- Q 7.5.7.1.1.1 Safety Parameter Display - The Safety Parameter Display System, as described in NUREG-0696 and NUREG-0737 Supplement 1, is implemented via the ERFDADS. The design of the ERFDADS is integrated with the implementa-tion of RG 1.97 (see Appendix 7B) and the Control Room Design Review (see Appendix 7A, Item I.D.1).
EOC.
The ERFDADS provides plant and environmental data to aid operators and manage-ment in the CR, TSC, and 0 to respond quickly to abnormal operating condi- N tions and mitigate the consequences of an accident. The ERFDADS functions during normal operations and emergencies to provide the following services: ,
[ 40
- 1. Provide plant and environmental data that is needed for the reactor oper- l ators to quickly assess the safety status of the plant. j.
- 2. Allow technical personnel access to comprehensive plant data, enabling them to assist operators without adding to the number of personnel in the control room.
- 4. Aid the operators in the detection of abnormal operating conditions.
- 5. Assist in the identification of the causes leading to any abnormal-ities.
- 6. Monitor plant response to corrective actions.
- 7. Providegroupingofparameterstoenhancetheoperators)abilitytoassess plant status quickly without surveying all control room displays.
- 8. Provide human factors engineered display formats (simple and consistent display patterns and coding).
- 9. Provide display information on a real time basis, along with validation of data and functional comparison capability.
7.5-10a Amendment 49
ATTACHMEN i ST-HL-AE- N STP FSAR PAGE lo OF fl
- 10. Provide display information on a real time basis for monitoring the RG 1.97 variables, as defined in Section 7.5.1 and Appendix 7B. These vari-ables are utilized to monitor the critical safety functions of:
Oberific4f;$ l e R::: d fi j -..m.1 ?
e Reactor coolant system inkpr :trih :: ::;l e Reacto'r coolant inventory :: : ;l c e Reactor core coolin8 49 e Heat sink maintenance e /
Pr' TJ . _;;;; gontainment environment The bases for the parameter selection are presented in Appendix 7B.
t
)
7.5-10b Amendment 49
ATTACHMENT i ST HL-AE-gfD8g PAGE S\
Table 7.5-1 identifies the specific parameters and indicates those available in the TSC and @ got, )(
7.5.7.1.1.2 Data Acquisition and Signal Processing for ESF Status Monitoring - The ERFDADS performs data acquisition and signal processing for ESF h Status Monitoring System. Input to the ESF Status Monitoring System $
Ts via demultiplexers within the FRFDADS. The ESF Status Monitoring System is described in Section 7.5.4. -
7.5.7.i.1.3 Data Acquisition and Signal Processing for Other Normal Plant Monitoring Systems - The ERFDADS performs data acquisition and signal X
processing annunciator [for and other normal the plant plant monitoring computer. Input tosystems includin the plan the plants annunciato via )(
[ERFDADSdemultiplexerj perconf o f )(
7.5.7.1.2 Equipment
Description:
prouw C orarol Cc4 b$ntTy 7.5.7.1.2.1 Multiplex es (MUX) - Non; Class multiplexers are utilized X for data acquisition from field inputs, 'switchjearpnd relay racks for input X into the ERF computer. The multiplexers are located in the glectrical x guxiliaryguildingswitchhearroomsA,BandConelevations10',35'and )(
60', 7 the relay room on elevation 35'; and the separation group D X distribution room on elevation 10'.
DMu d 7.5.7.1.2.2 Demultiplexer[(Non Class [E DMUXs are utilized to provide X input from the ERF computer to other ; plant monitoring systems including the ESF Monitoring System and the plant annunciators.
L (tatus, 0 7.5.7.1.2.3 ERF Computer - The ERF computer portion of the ERFDADS is located in the TSC (EAB, elevation 72') and provides data to the control room, TSC,andQdisplaydevicesandoffsitedata] links. Y E OC-TheERFcomputerreceivesdataconsispingoftheRG1.97definedanalogand Y digital variables and other va d La directly from the ERFDADS multiplexers.
QDPS, Meteorological System and RMS via redundant high speed data"9 1nks. X Su rple medati m Carmatim The ERF computer performs any data processing required beyond that performed by the data acquisition equipment. Redundant central processing units are provided with adequate memory capacity to support ERF data acquisition, management and transmission functions on a real time basis.
EOC.
7.5.7.1.2.4 Man Machine Interface - CRT display devices are located in the CR, TSC, and to present ERFDADS information to operators and management in a concise, easily intelligible format.
The primary SPDS display page is dx.insn: or :: 1: art m contrci rccr display )(
d-'~ > ~ e TSC ?ispicy devi::. This dispicy is available on all ERFDADS display devices. g gt e,,
7.5.7.1.2.5 HVAC Support - AC, with sufficient reliability to support /
the above ERFDADS availability requirements, discussed in Section 7.5.7.1.3, is provided in the TSC computer room. ERFDADS equipment located outside the TSC computer room is designed to function in the normal design environment for the areas in which the equipment is located.
7.5-11 Amendment 40
A 6 i ALHME l ST HL-AE- i STP FSAR PAGE yVOF }
f#
The TSC HVAC is further described in Section 9.4.1.
7.5.7.1.2.6 Power Supply - ERFDADS equipment including multiplexing X cquipment, the ERF computer and its peripherals, di lay devices, and //0ad printers. "hich 2re 1cc2ted eith!" " prer birck ,p#a rovided with/J20 vac X power from a dedicated non-Class IE uninterruptible power upply (UPS) capable of maintaining system operation for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and capable of maintaining system memory for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Normal AC power to the UPS is provided from a non-Class 1E dianel generator-backed bus. g$
- GOL ERFDADS equipment located within the s provided with reliable 120 vac I power from the r diesel generator-backed bus. X EOC-7.5.7.1 't System Operational Requirements: The ERFDADS data channels meet the 99 e . cent availability requirement defined in NUREG-0696 Section 1.5 under pressure and temperature conditions exceeding cold shutdown conditions.
The system meets an 80 percent availability requirement during plant cold shutdown condtions. N t
Data processed through ERFDADS is qualitatively corparable with other Post Accident Monitoring System, Radiation Monitoring System, and QDPS data displayed in the CR with respect to accuracy and response time.
7.5.7.2 Analysis. The ERFDADS design insures that any failure or
- alfunction of the ERFDADS equipment beyond the Class IE isolation devices does not compromise any safety-related equipment, components, or structures.
A verification and validation plan is provided for the ERFDADS software to demonstrate conformance with the functional requirements of NUREG-0696 and 40 NUREG-0737. This plan provides for an independent review of the system coftware.
Isolation and separation of Class IE signals is provided in accordance with RG 1.75. Inputs to the ERFDADS are isolated at the exit point of the isolation devices (see Figures 7.0.3-1 and 7.5.7-1). y 1.5.6 i This system is designed to meet the following criteria:
- 1. No single-point failure in any ERFDADS component has any effect on the plant operation. Any such failure is monitored in the CR. Redundant hardware is utilized when required to satisfy this requirement and to improve reliability.
- 2. Where redundant devices or assemblies are utilized, failure of one is detected and indicated to the ERF computer, and causes automatic transfer of functions to the other device or assembly without effect upon system performance.
- 3. On line diagnostic routines and transmission error checking provisions in the data network and host processora aid in maintaining validity of all data interchanges and in verification of the continuous functional integrity of system equipment.
t 7.5-12 Amendment 40
fl al o TABLE 2.5-1 j ho V ,
POST ACCIDENT MONITORINC INSTRUMENTATION 2ndice tpoh Sensor Eoc 'b x Qualif Icat ton -") Control Implemen- Sensor F#P TSC Conformance Type / Environ- M Number Room tation Power Indica- Indice- to RG 7 V etable Range / Status M mental Selante of Channels Date Supply tion , tion Rev. K Yes I r ;'- - @PS Core Load IE Tes bes Note b I RCS Fressure 0-3000 pats Al.Bl.B2 Ten (rt'e Range) Cl.r2.D2 I recorded 0-700*F Al.Bl.82 Yes Ten 1 per loop @PS Core load IE Tes Yes Conforan %
% tide Range 4 recorded A T
RQ tidi Range 0-700*P A1.Bl.52 Ten Yes I per loop @PS Core Load IE Tes Yes Conforms /
T s %I- 4 recorded g gand Wide Range Stess 0-100% of A1.Bl.52 Yes Ten 1 per steem ODPS Core Load IE Ten Tes Conforms C6nerator Level span D2 generator 4 recorded /(
WaTtf Tes Yes Conforms Marrow Range Steam 0-100% Al.Bl.B2 Yes Yes 4 per steam @PS Core Load IE Cenerato Level of span D2 generator / recorded )(
4W ii ptr 5f-Prscourire Level 0-1002 A1.Bl.D2 Ten Tes 4 ;- ;L g @PS Core Lead IE Tes Yes Conforms (
g of span I recorded )(
Containment 65
-5 to M peig A1.Bl B2 Tes Yes 4 g ;* ^ /@PS Core Lead IE Tes Yes Conforms 40 N in Pregnare Cl.C2.D2 2 recorded q r 3 Core load IE Yes Conforms Y Stsamline 0-1400 pelg A1.Bl .D2 Yes Yes -4=per loop @PS Yes )( ta I per loop N C Pensure recorded -
Yes @PS Core Load I F. Tes Yes Conforms E 0-1002 Al.Bl.D2 Ten R; fueling Water 3 ;- ,
2 meters X "Up Ststar.e Tan g str/ of span g K recorded Level Containment Water 0-609.000 gal Al.Bl.82 Tes Yes 3; - '-
@PS Core Load IE Tes Yes Conforms )( fl Level (Wide Range) (0-6 ft.) C2.D2 1 recorded hk ;
Containment Water Bottom of AI.52.C2 Tes Ten 2 --- '-- @PS Core Load . IE Tes Yes Conforms- / kr Level (Narrow Samp to Top D2 2 recorded 3 I
Range) of Sump ,
Austilary Peed- 0-100% of A1.Bl.D2 Ten Yes 3---'--- @PS Core Load IE Tes Tes Conforms 4 E water Storage span I recorded S Tae Level g 3 V 1414U Yes Note o ]
Austit'ary Feed- 9-t99t* A1.81.D2 Tes Yes 1 per loop PS Core Lead IE Tee )(
trter Flow ar -p== recorded y i 4/5T 4 .n,k3 o-7oo galp
TABLE 7.5-1 .
POST ACCIDENT MONITOR 1NC TNSTRtMENTAff0N (Continued) I' # *
, s.
EO C, 40 Sensor Qualification Control leptemen- Senent Jer" TSC Conformance Type / Environ- Number Room tation Power indica- indles- to RC t Range /S tatus Cateory mental Setaste of Channels Display Date Supply tion tion Rev.h.97 N Tartable 2 ;-- ' -* / QDPS Core Load IE Tes L'
Yes O
Note e y
1Rpren A1.Bl.B2 Yes Yes Kigh Range 2 meters
!'4 Containment 10 R/hr C2.E2 Radiation Level Comma 2 recorded Steam Cenerator '"_ !* A1.82.C Yes Yes 1per[ low- COPS Core Load IE Yes Yes Conforme 4 meters Blowdown Radia-tion Level g .1
[ down[ine 4 recorded IA NC C-Steam 1tne Radia- !" " *" A1.52.C2 Yes Tag 1 per stese QDPS Core Load IE Yes Yes Conforme tion Level E2 line 4 meters Y gp* L , p" 4 recorded
, gc .
100-2200*F A1.51.C1 Tea Tes 2 trains of QDP f tAcr a Core Load 1E Yes Yes Conforme Core Exit Temperature 25 thermo- H and ** d *e
- ' couples each, f4erage .# hathsf equally dis- h rent {
tributed recorded meroes corehg 40 grants)
RCS Subcooling 200*F sub- A1.51 Yes Yes 2 r- ' -
QOPS Core Load 1E Yes Yes Conforme )(
cooltas to 2 recorded 5044tP superheat K Meutron Flue I d te 2001 31.D2 Tee Yes 2 r- '-
QDPS Core Load IE Tee Yes Note r K (E; tended R uge) Full Power g K recorded Neutron Fler -0.5 to + 0 31.02 Yes Yes 2 -- ;' QDPS Core Load IE Tee Yes Note r )(
Secrtup' Rate dpa tT4 recorded ae Reactor Tessel Upper Core Bl.C2.D2 Yes Tee 2 ;:r ;'_- y ne=t%
QDPS fle Core Load 1E Tes Yes Conforme X water Level Support Plate gMd ,
y to top of vessel contatament Open/ Closed C2.D2 Yes Ten 1 per valve 1 pair of Core Load 1E Tee ,, Ye's Note c Itzhts per
{a Isolation Valve Sectus valve -
a containment 0-101 Bl.C1 Tee Yes 2 p;r %
Y 00PS Core Load 1E Yes Yes Conforme y
" 1 recorded ['
Hydrogen Concentration
$ Concentration ATTACHMENT /
ST-HL-AE- Inb PAGE -W OF &
Tant.E 7.5-1 POST ACCIDENT HONITORING INSTRUNFNTATION (Continued)
Sensor QualifIcatton Control Implemen- Sensor EOP YSC ~ Conformance Type / Environ- Number Room tation Power Indica- Indica- to RC 1.97, V rtabl9 Range / Status Catepry mental Setsmic of Channels Display Date Supply tion tioh Rev. 82
, Contral Rod Rods on D3 No No I per rod LFD Core Load N-lE Yes Yes Conforms Position Rottom Indtection (Note z)
Contcineent 0-180 pass C1.C2 Yes Yes 2 --- ' - # @PS Core Load IE Yes Yes Prx surs I recorded Conforms [
(E2 tended Range)
RCS Prscoure 0-3500 patg AI.RI Cl Yes Yes 2 -- '--* Y QDPS Core Load IE Yes Yes Note b (Extendid Range) 2 recorded K Primary Coolant N/A C3 No No 1 post CRT (ERF1) ADS) Core Load N/A Yes Yes Notes d, h Activity and accident Sampling sampling system W W
(Note a) C2,E2 Yes No 1 --- '- - Core Load Unit Vent "
Radici%m CRT (RNS) N-lE Yes Yes Conforms X m (Note a, v) N Lent 40 g Puel Kandling -6 -l N IWol0ay C2,E2 Yes Yes 2 --- '- " QDPS Core Load IE Yes Yes Conforms )(
Bldg. Radiation uC1/cc 2 meters g Levelk g g g 2 recorded T Adjacent ButIdtng 10hol0 C3 No No 5 -- ' - M CRT (RNS) Core Load N-lE Arve Radiation Level mR/hr Yes Yes No e / -[
Sits Environmental N/A C 3,E3 No No N/A Portable Core Load N-lE No No Cor. forms "U Radtstion Level Sampling (P.>rttble Monitoring) m Pramm rtser PORY Open/ Closed 52.D2 Yes Yes 1 per valve 1 pair of Core Load IE Yes Yes Conforms m Settus lights per valve
=
gm y
( Preseartzer PORY Open/ Closed D2 Yes Yes I per valve I pair of Core Load IE Yes Yes Conforms %_., ~
, Ricck Valve Status lights per valve
1-F J
d YARLE 7.5-1 POST ACCIDENT MONITORINC INSTRIMENTATION (Continued) t..
Sensor -
Control implemen- Sensor EOF TSC Conformance Qualtitcation Indica- Indica- to RC 1.97, Type / Environ-~ Numher Room tation Power Ranae/ Status Cateory mental Seismic of Channels Display Date Supply tion Itton - Rev . #2 V'rtable Yes I per valve ! Alarm Core Load N-lE Yes Yes Conforms Prssaurtzer Safety open/ Closed 82.D2 Yes -
CRT (ERFDADS)
Valve Status #
Yes Yes I per bank I pair of Core Load IE Yes Yes Note e Precourtser Hester Open/ Closed D2 lights per Brsaker Position bank
'--- Core Lead IE Yes Yes ' Conforms' X Prw oortaer 1700-2500 14 2 Yes Yes 4 -- QDPS
$ rtterdtd )t Pressure pets D2 No No 1 per pump I pair of Core" Load N-lE Yes Yes' Conferms RCP Status On/Off lights per pump No I per valve 1 ight per Core Load N-lE Yes Yes Conforms )L Pressurtter Spray Open/ Closed D2 No M V !ve Statu, alve D2 Yes Yes 1 --- '- F QDTS Core 1. cad IE Yes Yes Conforms y ,
Charging 0-500 m g Flow 40 y
Core Load M-lE Yes Ves Conforms X Litdown Flow 0-500 ppm. g D2 Yes Yes 1 pr ;'_ ^ l meter
- i vslume Contial 0-100% D2 Yes Yes 2 ;- ;'- # 1 meter Core Load IE Yes Yes Conforms X
of span k O(8 TcnkLevel)Ideitf D2 Yes isolation 1 per valve I pair of Core Load IE/4-lE Yes Yes Conforms CVCS Valve Status Open/ Closed (Note f)
Valves 11ghts per g valve by only
&r D2 Yes Yes ?, ;! I pair of Core Load IE Yes Yes Conforms K M --h Chcratng Pump on/ Oft (Note f) >' )= I 1pegwp lights per Status pump D, E Yes Yes 2 r--
< 1 pair of Core Load IE Yes Yes Conforms x %9 M
y Roric Acid On/Off D2 r *
(Note f) g0 g g'ge pom p lights per y Status %
Trrnsfer(Pump h Yes 1 per loop pump QDPS Core Load IE Yes Yes Conforms K RCP Seal 0-20 apa. - D2 Yes M 4 recorded (Note f) g Injection Flow e
E E
g -
YARI.F. 7.5-1 POSY ACCIDFNT MONITORING INSYRtMFNTAY10N (Continued) t..
Sensor Qualtitcation Control Implemen- Sensor EOF TS Conformance Indica , Indica- to Rc I 97 Type / t'nviron- Number Room tation Fower V'rfahle Ranae/ Status Cattory mental- Setssic of Channels Display Date _ Supply tion 'tlon ' Rev. #2 SC Atmospheric 0-1002 Open D2.E2 Yes Yes 1 per valve QDPS Core Load IE Yes Yes Conforms FOR7 Status I meter per valve .
Main Steamline Open/ Closed B2.D2 Yes Yes 1 per valve I pair of Core Load IE Yes Yes Conforms Isolation Valve lights per (Note f)
Stttua valve M2ia Stesaline Open/Clased R2.D2 Yes Yes I per valve 1 pair of Core Load IE Yes Yes Conforma RypIns Valve lights per (Note ()
Status valve SC Safety Valve Open/ Closed D2 E2 Yen Yes 1 per valve Alarm Core f. cad N-lE Yes Yes Conforms Status CRY (FRFDADS) .
M;ta Feedwater Open/ Closed D2 Yen Yes 1 per valve CRT (FRIY) ADS) Core Load IE Yes Yes Conforms )(
Centrol Valve (Not g . )
Stttus 40 ,
y Mata Feedwater Open/ Closed D2 Yes Yen 1 per valve CRT (FRFDADS) Core 1. cad IE Yes Yes Conforms / y y Cortrol Bypass (Note Q Q)
I y 3 Vcive Status g M2in Feedwater Open/ Cloned D2 Yes Yes I per valve 1 pair of ' Complete I F. Yes Yes Conforms isolation Valve lights per (Note f) M Status valve DHH Main Feedwater Open/ Closed D2 Yes Yes I per valve I pair of Complete IE Yes Yes Confor=*
(Note f)
X oiH Isolation '-
BypIss Valve lights per valve Mr -h Yg Status t drrig
,5.oxso,sy4, o m Mitn Feedwater 44405 D2 Yes Yes 3 per loop QCFS Core Load IE Yes Yes Conforms )( M 7C Flow ed-eyes I per loop recorded
, w(
SC Blowdown open/ Closed D2 Yes Yes I per valve I pair of Complete IE Yes Yes Conforms Irelation Valve lights per (Note f) valve
{Statua u
Open/ Closed 02 Yes Yes I per valve I pair of Complete IE Yes Yes Conforms
{SCBlowdown a Sample [ Isolation (Note ()
lights per j
{ValveStatus valve )(
O
YAPIF. 7.5 1 b's FnSY ACCIDENT MONITORINC INSTRt'MENTAYlON (Continued) al i.
Sensnr Qualification Control implemen- Sennor EOF TSC Conformance Type / Environ. Number Rons tation Power Indica- Indica- to RC 1.97 Vrriebte Rense/ States Catenry mental Setsste of Channeln Di= play Date Supply tion atton , R ev ., f2 L
M HMSI Flow 0-2000 gal / min D2 Yes Yen 2 per 51 6 metern Core Load N-I F. Yes Yes Conforms Y
pump (gut lag
.. w4 W LHSI Flow 0- al/ min D2 Yes Yen 2 per SI 6 meters Core Load N-I F. Yes Yes Conforms W M SO pump (hmtb s c.4 kg)t s- ogial/mu'a FCCS Accumulator 0-7 pafgN D2 Yes Yes 2 per tank 6 metern Care Load N-lE Yes Yes Conforms Fransure A stilary Teed- Open/ Closed D2 Yes Yes I per valve I pair of Core Load I F. Yen Yes Conforms water Valve Status lights per (Note f) valve Containment Spray 0-1001 D2 Yes Yes I per train 1 metern Complete N-I F. Yes Yen Conforms Finw of span 40 we C4ntainment Spray Open/ Closed D2 Yes Yen I per valve I pair of Complete IE Yes Yes Conforms y
. System Valve lights per (Note 1) .,,
[cmStatum valve g Containment Spray on/Off D2 Yes Yes I per pump i pair of Complete IE Yes Yes Conforma Fumgtatus lights per (Note f) X pump M met W.M Fan Cooler Yes Yen 1 per fan 1 A ara per Core Lead IE/M-lE Yes Yes Note a N 3-4 in, water gf,D2 Differential On/Off fan Pretnure/Ststun
. train QDf$ hh-4 CCW Pump Discharge 0-150 pelg D2 Yes Yen 1 per headma ' .- Core Load IE Yes Yes Conforms -4 M --
Frannure (Note 1) Or*h r
V9l% d >riI r g.,
containment Vent- Dren/ Closed D2 Yes Yen I per dampas. I pair of Core Lned IE Yes Yes Conforma Y i1.1ttonammmar.Yehpg 1tahts per X Om Statun 50 dampee-Yelv t t rain nors X MM xe CCW Header %-250'r D2 Yen ten 1 per headas '--n Core Load IF. Yes Yes Jonforms X -
Yamperature QDFS N CCW Surge Yan 0-100% of Core Load IE Yes Yes Conforms npan D2 Yes Yen I per tank cowpartment (Note f) [
{l.evel gg a
s
YARLF 7.5-1 POST ACCIDENT W1NTYORING INSYRifMFNYAYION (Continued)
Sensnr Qualtflestion Control Implemen- Sengor F0F TSC Conformance Type / Fnviron- Number Room tation Power I r.d tc a- Indien- to RC l.97, V rt'ble Ranae/ Status Caterry mental Seismic of Channels Dis la Dare Supply tion t t'on Rev. #2 affr*rrietsgs) n 40 Yes Yen I per ESF "
Core Load IE Yes Yes Conforma CCW Flow to F.SF 0- N D2 Compon;nta y for M-component,heuc/e composenT[4(ader 1 y C(w IE Yes Yes Conforms CCW Vzlve Status Open/ Closed D2 Yes Yen I per valve I pair of Core 1.oad lights per *
(Note f) valve Scy s repMt s.1]m IE Yes Yes Conforms
'n :_ ' " ' - 0- fov' D2 Yes Yes 1 per major ODPS Core Load
- ,_ Flow opam. con 6penenT/ FSF component (Note f)
W sse t e temenT5 h ead# Conforms
--'-! C!n Open/ Closed D2 Yes Yen 1 per valve 1 pair of Complete IF/N-lE Yes Yes 44a=&aa. Ifghts per (Note f)
Q. ^ = ECW Valva Status
"'s valve er-W mete v-ESF Fnyironnent Temperature Yemperature D2 Yes Yes I per ESF 1 alarm Core Load N-lE Yes Yen Conforma above metroint component / (Note f) cubicle EsF CvMf.le,. 40 a Yes Yes 1 per ESF 1 pair of Core Load IE Yes Yes Conforms Fsu/ Cooler Fan Stopped / D2 y
component / lights per (Note f) M 6 Status Running cubicle item $
f,,,
e Yes Yes I per ime 1 meter or Core 1.oad IE/N-IF. Yes Yes Conforms Sttndby Power Bus Specific D2 tnd Emergency alarm for Power Source each power Statua nource gg
> -t-4 Othrr Safety- Component D2 Yes Yen 1 per source 1 meter or Core 1.nad IE/N-]E Yes Yes Conforms (Note y)
Oi-4 Reittsd Energy Specific alarm for I"*
Sourcen each power .I source K Yes Yes 1 per heat QDPS Core Load IE Yes Yes
- Conforms RHR Heat Exchanger 50-400'F D2 Dischtrge enchanger 3 recorded 3,0 Temperature g Yes Yen 1 per DMP Core 1.oad IE Yes Yes Conforms RitR Flow %N D2 QDPS span train 3 metern k
E.
2 E
I s
m TARIE 7.5 1 POST ACCIDFNT MONITORINC INSTRUMENTATION (Continued)
Sengor Oualification .
Control Implemen- Sensor EOF TSC Conformance Type / Fnviron- Number Room tation Power Indica- Indica- to RC 1.97 Vr r is hle Ranae/ Status Cateory mental Setsste of Channels Display Date Supply tion t' ton Rev. f2 Open/ Closed D2 Yes Ye I per valve I pair of Core Load IF N-lF. Yen Yes Conforms RHR Valve Status (Note f) is lights per tion Valves valve Iv)
QOr$
Rxctar Trip open/ Closed D2 Yes Yes I per brenner 1 pair ,of Complete IE Yes Yes Conforms Brsaker Position ll Rh ts per (Note f) breaber Open/ Closed D2 Yes No I per valve 1 pair of Complete N-lE Yes Yes Conforma Turbine Covernor Valve Position lights per (Notes f.z) valve Open/ Closed D2 Yes No I per valve 1 pair of Complete N-IF. Yes Yen Conforms Turbine Stop Valve Position lights per (Notes f.s) valve 40 Yes Yen 1 per pump 1 pair of Core Load IE Yes Yes Conforms Noter-Driven On/Off D2 U
Auxtllary Feed- Ilghts per (Note f) '
w witer Pump Status pump p.mp dialmy, g o r$ 3
%00 f 5s g Yes I sasMaa. Ccre 1.oad IE Yes Yes Conforms $
w Arztllary Feed- 0-56MN>-eye D2 Yes I meter.
wIter Turbine Open/ Closed assad pots 5prt I pair of (Note f)
Pump Status indicator. lights per 1 per steam valve inlet valve St Pump Status On/Off D2 Yes Yen 1 per pump 1 pair of Complete IE Yes Yes Conforms h -4 (Note f) lights per Pump tjo St Valve Statum Open/ Closed D2 Yes Yen 1 per valve 1 pair of Complete IE Yes Yes Conforms .E J
Ilghts per (Note () O,3 CZ valve ,
y ':4 S Yen Yen 1 per pump 1 pair of Complete IE Yes Yes Conforma -
F-asential Cooling On/Off D2 a,
rs Water Pump Statum !!Rh tm per (Note f) pump f
on/Off D2 Yes Yen 1 per pump 1 pair of Complete IE Yes Yes Conforan gCCW Pump Featus (Note f) lights per pump l
m TARI.E 7.5-1 POST ACCIDENT MONITORING INSTRUMENTATION (Continued) r.,
Sennor Qualification Control Implemen- Sensor EOP - TSC Conformance Type / Environ-
- Number Room tation Power Indica- Indica- to RC 1.97 Vsrlable Ranae/ Status Cateory mental Seismic of Channels Display Date Supply tion tion Jtev. f2 RHR Pump Status On/Off D2 Yes Yes 1 per pump 1 pair of Complete IE Yes Yes Conforms lights per (Note f) pump a(TAten St Actuation On/Off D2 Yes Yes I per p&ame. I Alarm Core Load IE Yes Yes Conforme y Status -tygm aftu @ a Castsineent Iso- On/Off D2 'Yes Yes I per p&ame. 1 Alarm Core I. cad IE Yes Yes Conforms
!stion Actuatton tyain K Status Control Room 10 I
E3 #
to No No 1 -- ;' CRT (RMS) Core Load N-lE Yes Yes Note 1 RsdistionLfAh 10{mR/hrhete) 10~ to 10'I E2 Yes Yes 2 p w M QPDS Core Load IE Yes Yes Conforms X uC1/cc(tdal(gg,} 2 meters 2 recorded )f Access Area 10' to E3 Yes No I per CRT (RMS) Core Load N-lE Yes Yes Note 1 RIdiation 10 R/hr designated 40 p area y Condenser vacuum 10~ to E3 Yes No I 7-- '-
CRT (RMS) Core load N-lE Yes Yes Note n 5
~ rump Radiation 10 uCl/cc $
Level Concentration from Lfquid Pathways -
g(f3y
> -4 SC Blowdown 10 to E2 Yes No I per plant CRT (RMS) Core Load N-lE Yes Yes Note t 10 uCt/cc d>O C[I cond. Polish 10 to E2 Yes No 1 per plant CRT (RMS) Core Lnad N-lE Yes Yes Note t N m 10' uC1/cc 1
-h
. O2 ri C M
2 1.fquid Radwaste 10 to E2 Yes No i per plant CRT (RMS) Core Load N-lE Yes Yes Note t 30 g
10'y uCi/cc '_
e E. TCB Drafn 10 to E7 Yes Nc, I per plant CRT (RMS) Core Load N-1E Yes Yes Note t
~I W 10 uCf /cc _
f.
8 l
m TABt.E 7.5-1 POST ACCIDENT MONITORING INSTRUMfNTATION (Continued)
Sensor Qualification Control implemen- Sensor EOF TSC Conformance Number Room tation Power Indica- Indica- to RC 1.97, Type / Environ-Cateory mental Sstemic of Channele Diaslay Date Supply tion tion Rev. #2 i V rf<ble Ranae/ Status Effluent Fath Flow Rate / Status __ --
No No 1 per plant CRT (RMS) Core Load N-1E Tes Yes Note q SC Blowdown 0-1001 E3 Flow of span l
Tes 1 per valve 1 pair of Core Load N-1E Yes Yes Notes q, w Valve Status Open/ Closed E2 No Ifghts per 1
valve No 1 per plant CRT (RMS) Core Load M-1E Tee Yes Note q cond. Polish 0-100% E3 No Flow of span 1 per valve 1 pair of Core Load M-1E Tes Yes Notes q, w valve Status Open/ Closed E2 Yes No j
lights per valve _
40 No 1 7--
- CRT (RMS) Core Load N-1E Yes Yes Note q Liquid Redweste 0-1001 E3 No Flow of span Tes no 1 per valve 1 pair of Core Load N-1E Tes Yes Notes q. w h 4 ? Valve Status Open/ Closed E2 ,
lights per y valve g
, y No 1 per plant CRT (RMS) Core Load N-1E Yes Yes Notes q TCB Drain 0-1001 E3 No Flow of span Valve Status Open/ Closed E2 Yes No 1 per plant CRT (ERFDADS) Core Lead M-1E Teo Ten Notes q. w gy
> ~f-4 EtCOADs Unit vent 0-100% E2 Yee No 1 per ;' ' CRT p Core Load M-1E Tes Yes Note w QI"i j C~o 3
Flow of span DI Condenser 0-100% E3 No No 1 --- *--* # M (RMS) Core Load N-1E Tes Yes Note v NE O
vac e Pump of span g
k, Pump Status on/off E2 Yes No he Pomp 1 ; - '--* # CRT (ERFDADS) Core Load N-1E Tee Yes Notes v. w V
[*
['
No 15p ;?r' CRT (ERFDADS) Core Load M-1E Teo Yes Note u Netsorological 4/4- E3 No P.r..et.ra < _a
- 2. sm) 1
_ . o .
7
- O C t
t
?
TABLE 7.$-1 I POST ACCIDENT WONITORINC INSTRt19tENTAT10N (Continued) .
O Sensor ggc Qualification Control implemen- Sensor Ber TSC Conformance Typef Environ- Number ' Room tation Powar Indica- Indice- to RC 1.97, V*rt'ble Renae/ Status Cateory mental Setemic of Chas.nela Display Date Supply tion tion Rev. M h
Contsineent Sump N/A E3 No No 1 post CRT (ERFDADS) Core Load M-1E Yes Yes Notes d. h I cnd Atmospheric accident Sampling sampling sys-4 Lem 6 .- p:_..; '
Roric Acid Tank --
Note 3 Charging Flow C;.ntainment -- Note 1 Atmo:pheric Temperature f
l Accumulator -- Note j 40 Tx k Level i
Containment -- Note k Sump Water Temperature .
7 N u H ut Removal by Note m ,
I
++ th3 Containment ll.
F a Heat Removal "N ty tem Emergency -- -- -- Note y Vsat11stion Damper Position O
- m >
)>OI t i
E .
Om
- uZ ,-
oo-4 M-i 1
2 3
8 l
I
(
ATTACHME I STP FSAR ST-HL-AE-PAGE 4) M TABLE 7.5-1 (Continued) g NOTES (10 -11
- a. Noble Gas: 10 to 1/ce, Particulate: 10 to10/.lCi/ce, 2
X Halogens *: 10 to 1/cc.
To cover the required range of yarticulates and halogens, a cbmbination of on-line detection and grab sample capability with onsite analysis is em-ployed. These monitors are environmentally qualified, but not seismically qualified, since they are attached to a non-seismic system.
- b. .RCS Pressure - one qualified channel of wide range RCS pressure and two qualified channels of extended range RCS pressure are used to monitor RCS Pressure for STP. y
- c. Containment Isolation Valve Status - TP has identified inst mentation that is necessary to assess the pro ess of accomplishing or aintaining critical safety functions. The er tical safety functions efined are equivalent to those utilize'd in t Westinghouse Owners oup Emergency Response Guidelines, i.e. , R::: tin it~ C:ntr:1, RCS Pr :: r Cent :1, Reactor Coolant Inventory C:6:1, Reactor Core Cooling, Heat Sink f Maintenance, and R: $ r Containment Environment. Containment isolation f(
valve status is not a critical safety function. However, the containment (solation valve positions were designated variables for monitoring the actual gross breach of the containment and are therefore qualified to
_ . _ , category 2 criteria. 40 )(
- d. The STP Pos ident Sampling System is sufficient for obtaining samples N to perform detailed analysis of RCS coolant, containment sump, and con-tainment atmospheric activity. Offline measurement systems are considered Category 3 variables.
- e. Pressurizer Heater Status - RG 1.97, Rev. 2 specified that heater current was the preferred parameter for determining heater status. For STP, heater breaker position was selected for determining pressurizer. heater status due to hardware considerations. Breaker position provides adequate indication to the operator to ensure that the two pressurizer heater banks powered from the Class lE busses are operable.
- f. The study performed on STP indicated that these' parameters were in:1:f:3 seeAnd y in the minimum set of parameters necessary to monitor the performance of:
- 1. Plant safety systems employed for mitigating the consequences of an accident and subsequent plant recovery to attain a safe shutdown condition, including verification of the automatic actuation of safety systems.
- 2. Systems normally employed for attaining a cold shutdown condition,
- g. Boric Acid Tank Charging Flow - For monitoring the performance of the
} Emergency _ Core Cooling System (ECCS), STP has designated Refueling Water Storage (RWST) Tank Level, High ead Safety Injection (HHSI) Flow, f(
7.5-24 Amendment 40
r.. _ ,_....;_ _ _ _ _ --_
ATTACHME I ST-HL-AE- i STP FSAR PAGE % 0 TABLE 7.5-1 (Continued)
~
NOTES Low Head ' Safety (LHSI) Injection Flow, Containment Water Level, and ECCS [
Valve Status. Since the ECCS does not take suction from the Boric Acid Tank (BAT), the Boric Acid Charging Flow was not designated a~ key vari-able. If the operator uses the BAT for boration following an accident, normal charging flow and RCS sampling is used to demonstrate that the RCS '
is being borated.
- h. Data entry is via manual keyboard.
- i. Conte.inment Atmospheric Temperature - The WOG Emergency Response Guidelines (ERG) do not require the operator to take an action hat would result in adverse consequences if the Containment temperature was iedt tr- F pvw.by,,eeting an erroneous value. As such, the Containment temperature indi- y ,
v cation is considered a D3 parameter and is not specifically identified on this listing. -
- j. Accumulator pressure indication and valve position indication for the accumulator discharge isolation valves and accumulator vent valves provide adequate status of the accumulators.
- k. Containment sump water temperature indication is not utilized by the operator to take corrective action. Other parameters were designated as 40 O
- STP type D variables to demonstrate that the Safety Injection System (SIS) is operating properly when taking suction from the Contabnent sump.
- 1. Conforms to RG 1.97, Rev. 3.
- m. Heat removal by the Containment Heat Removal System (CHRS) - Other para-meters were designated as STP type D variables to demonstrate that the containment heat removal systems are operating properly. These include the following:
e Containment Spray Flow
! e Containment Spray System (CSS) valve status e Containment Pressure e Containment Water Level e Containment Spray Pump Status e Reactor Containment Fan Cooler (RCFC) Differential Pressure e RCFC Status l n. Condenser Vacuum Pump Radiation Monitor - This parameter is considered to be a backup variable for the measurement of secondary side radiation.
l 7.5-25 Amendment 40
p-. . a .o _ . s.w - a -- - - - _ . . .q r- --
ATTACHM i ST-HL AE- 0 STP FSAR PAGE % OF 9l
, . TABLE 7.5-1 (Continued)
- - NOTES Main steamline radiation monitors are adequate to provide primary indica-
-tion of this information. The condenser vacuum pump radiation monitor is anyironmentally qualified, but not seismically qualified, since it is attached to a non-seismic system,
- o. The STP design utilizes our physically separate auxiliary feedwater lines. The four Class lE transmitters provide the redundancy required. U The requirement is to ensure flow to at least one intact steam generator post-accident. The required redundancy with a four loop plant is provided by one channel per oop. SG 4 4
tom kN % (o- p-d is dbgel , J;qWide kpd vi"Range
-tt- Gprovides DPS.
a diverse backup.
- p. Emergency Ventilation Damper Position - As an alternate to monitadng ventilation damper position, STP monitors radiogas, radioparticulate, j,g ---- End/or radiM concentrations at various locations in the plant which '/,
,F I provide information concerning the status of the ventilation system.
These parameters include:
e Area radiation in locations which contain, or could contain, sig-nificant quantities of radioactive material e Unit vent radiogas concentration e Radiogas concentration discharged from non-headered vents e Environs radiation e Fuel handling building vent radiation e Effluent path flow rate
- q. Effluent Path Flow Rate / Status - Variables which provide the operator with information to estimate the magnitude of release of radioactive materials through identified pathways. Valve status is the primary variable and flow rate is a backup variable.
8-
- r. Neutron Flux - No diverse variable is required since the failure of one channel vill not cause the operator to violate the required safety fun-etion.
- s. Two Cont,inment high range monitors meet the requirements of a Type A variable. These monitors are Class IE, redundant and fully qualified to %
Category I requirements. Six area monitors are located throughout Containment with the range of 0.1 to 10,000 mR/hr that provide additional monitoring over this range. In addition, the off-scale high readings of the low range monitors provide some information to resolve ambiguity above this range. These two qualified high range radiation monitors also satis-fy the requirements of NUREG-0737.
'i 7.5-26 Amendment 40
f
= {
ATTACHMENT I ST-HL AE- l'IsD PAGE M OF TI STP FSAR TABLE 7.5-1 (Continued)
NOTES
- t. The study performed on STP indicated that these parameters were included in the minimum set of parameters necessary to monitor for release of radioactivity via liquid effluent pathways. These monitors are environ-mintally qualified, but not seismically qualified since they are attached to non-seismically qualified systems.
- u. Meets requirements of RG 1.23.
- v. For the purpose of radiological release calculations, the conservative ,
ass'umption of maximum flow will be utilized. Actual flow indication serves as a backup parameter and is designated Category 3.
- w. These Category 2 sensors are environmentally, but not seismically qual- 40 ified, since they are attached to a non-seismic system.
- x. Rod position indication is provided in the CR via the digital rod position indication system light emitting diode (LED) display.
- y. Instrument loops on Class 1E systems are qualified up to and including channel isolation devices.
- z. These Category 2 sensors are environmentally and seismically qualified;
( however, they are installed in a non-seismic system and are therefore not listed as seismically qualified instruments. They are installed using mountingssimilartothoseusedforcomparablgseismicallyqualified equipment.
X' I
l 7.5-27 Amendment 40 l
/ r bg cM d TARLE 7.5-2 (f &% lytvfW ' 6& E rc, MAIN CONTROL BOARD INDICATORS AND/OR RECORDERS AVAll.ABLE TO THE OPERATOR CONDITION IV EVLNTS No. of Channels Accuracy Parameter Available Required Range Required _
Indicator / Recorder P'urp e
- 1. Containment 2 1 0-Il5% of design +10% of full Both channels indicated; onttor post-LOCA Containment Press A pressure scale ! recorded conditions N.
- 3. Refuteing Water ' - 1 0-100% of span +3% of level Both channels indic d; Ensure that water is flowing Storage Tank 'N. upan 1 recorded to the Safety Injection Sys-Water Level 's tem after a LOCA; determine
~ when to shift from injection to recirculation mode
- 3. Steam Generator 1/SG * +7 to -5 ft from 110% of level All iannels indicated; Detect SG tube rupture; mont-Water Level om nominal full span ** c . nnels used f or cont rol tor SG water level following (narrow range) 10 level are recorded a steam line break
- 1/SG 17 to -4 t from 120% of level All channels recorded Detect SG tube rupture; mon-Water Level nominal fut nad spsn tor SG water level following (wide range) level a steam line break
- 5. Steam Line 2/ steam 1/ steam 0-1,300 psig -
full scale Both channels indicated; Monitor steam line pressures y Pressure line line 1/ steam line recorded following SG tube rupture or
]
L steam line break O E
- 6. Pressurtzer 2 1 Entire distane Indicate that Both channels indicated; Indicate that water has re- $
Water Level between tap - level is some- 1 recorded turned to the pressurizer where between f ollowing cooldown af ter SG 0 and 100% of tube rupture or steam line span *** break
- 7. Containment 2 1 102 to 100 mR/hr 15% Bot h chans s indicated; Monttor and indicate Hadiation Level I recorded radiation in containment h [.-4 O
, CO H. Containment 2 2 O to 8% 15% Both indicated; I orded Monitor and indicate hydro- f 'k1 112 Concentration gen concent ration in Con- %K tainment OE n .tZ
, , M 3%
- One level channel per ' SG either wide or narrow range) with at least two wide-range channels for the plant. *
h *** Indicated er level should be above the pressurizer heaters and below 100% of span (apprustmately 25% of span).
b O
l' N
C oIimZ -
=C 1 o
, s n ui i a
- s. l
- sit Aa" >Nomo-o ni en et s o s n ai gt a
g w e
a er ee t t t t h a nc at ic cd eu d n dicin n oe n i y i ck a t t d et d i s n vn nvu aii aia t th r cr r c x e ai oae s
o t oa t o iin p tn e nd o r o h oai
- J. M rt M rt f
d d e e d
r w o
N c c O e e T r r A
R r 1 I E e %
P d ; ;
O r d d o e e E c t t H e a a T R c c
/ i i o r d n
d T o n t i i E a L c h h B i t t A d o o L n b B
- I I A s V
) A d
e S u R n E i D t R n O h o C S C E T
( W N E yd 2 V c e
~ E a r
$. pDV ri uu 7 Nk c q t t A c e o o E N AR t t I S > t R R i A O T T T Is A
C Ni>
I e c c D C c I
N /
t
.'c
/
t C nC D p i p R
A 61 S1 O - - - -
B 00 o0 1 1 t1 L e ,
O g ax xx R n T a I1 11 N R O
C N
I N A
H d ' .
se l r ei s nu nq i 1 ae hR C
f
-.,s oe
.l b oa NLi 2 2 a
l'f
~ -
d l
_' l i
u B nl e
v gi v oe N me oL o
R n i
nt e ll di n aL r o nt o e li ani t ot Het e ra V ai L m ti l a nd egd r oa un e a CR Fi M P
g . 0 9 1 N - Ea! " g
e- - = = u- u - = - - - c- - -
+- -- . - =m . , e m s. .
m
, ATTACHMEyl )
. . ST-HL-AE- I .60 PACF Llo OF @
5
!! i )
z.5 5 ..- .. .- .. .. .- ..
g rf .j . ! .y.
l .- , !S !
)
li.
s cl- fi-ca et- et. !!-
es
. n6 44 4 o n +9 ( n
. f ..
l g
/l k E
[
/ // / / "
/;
i E
1 i - =
i 8 . n . . o 4 I I _ _
I -
I i
~
~
t o o o P. /
o // J\ d/
/ / ii!
a-.
- / E5I
/
(]
y, g ,
, [/ ,
o of n - " "
ql!i a3 n
a'll dE. b 3 ig',t oE t<
a a sw: g <l I g, (E
- g -j =- ag<.I." r y 3 , 3y -;
!!*i i. h l 52 / . ri ! &:
0!!:il , ;
g U I 15 1 2!!.A l:
?!!g$5 i31: d3I: 5 4/ EE Ik i
!$$5 boonn . annono nonu s'. , unn ,
o I o "
o j o !.
. . .... g . . _ .
g, , . .... g . .... g
- :. e :
. : . I.I = = . g - :
I I t- : j g 2 T J3j E I I TJ is e Te ) TJ e
v a n n o o o o 3 Er : Ee E U i
i.
EI S EI j.
BI 3I a El
.. ..I 3
3 " [4 3 "I
- f ,. _{
J SOUTH TEXAS PROJECT UNITS 1 & 2 1)2 P23
- dv Ix
's k i 1%
5 Qh d
$ ?'!3--
QUALIFIED DISPLAY PROCESSING
) w }U 5, * $ 4 d #c E*
- o a SYSTEM BLOCK DIAGRAM d $ w s g -s k 4 *>
M
- 4 il 4 cc i & Figure 7.5.6-1 Amendment 40
X S S) & R 4 .
MO T U P Y DE I
P Y
0 NTJ GUP S
o"c )
P REL I I
T
(
t t W iS"(T t
f WP iSN(T Y
s" t
i5 Y
T
(
F F l F l r T- l
' _:~l 5
l
)
T
- gg &x & a D
gg 1Ug 0 LUk )0 t e /Tf oS t
L E
5 8 g O
S M (/
1
, S i
I 4 C.
R g, N/F 4
A G g g
E/O H
C W
S o
g g . ME
= H A /</
= CE AH L
)
S P R S Y P P S
A D I O8 T
(
O O S T
,U P
N I
S S P P d I DC G O C
M l S
Y A
L P
S t
O C f '
0 t M
4 4^
, - l p R Es a T E j
/ F O
L F
- PSUT QE J
tMY OS C
= '
_ Il
. ,i
- _ 4 _
_ TI T R
S Y ,N- X N
O S. uOl N" I
T ET
-. y UFM O "" NU
_ A AP L N N 8 S S Lu/ M.
P SI p N'U eE D U T
O IS M (t M O D" A"
LM P O D A O R 5 C T R
- S R
- ( E T
- RU S
T U
P M
- " RP N
O
" I C
@OC Z qm*>(D T O o" c. m O H ZQ4 ( e 9 N 8
m$ ny0c n oz m cf Eim3oUEmQgz cEQMhs gfa w 8
r w
L STP FSAR ATTACHMEN I Appsndix 7B h4 0 RI --l TABLE OF CONTENTS Chapter 7 Section ,
1.pg.andix 7B Page 7B.1 DISCUSSION ,
7B.1-1
.7B.l.1 PLANNED VERSUS UNPLANNED OPERATOR ACTIONS 7B.1-2 7B.1.2 Variable Types 7B.1-2 4 f,, 7B.1.3 Design and Qualification Criteria 7B.1-3 ie W 4 /
s**b/ "?" 7B.2 DEFINITION OF VARIABLE TYPES 7B.2-1 7B.2.1 DEFINITIONS 7B.2-1 7B.2.1.1 Design Basis Accident Events 7B.2-1 7B.2.1.2 Safe Shutdown (Hot Standby) 7B.2-1 7B.2.1.3 Cold Shutdown 7B.2-1 7B.2.1.4 Controlled Condition 7B.2-1 7B.2.1.5 Critical Safety Functions 7B.2-1 7B.2.1.6 Immediately Accessible Information 7B.2-1 7B.2.1.7 Primary Information 7B.2-1 40 7B.2.1.8- Key Variables 7B.2-2 7B.2.1.9 Backup Information 7B.2-2 7B.2.2 Variable Functions 7B.2-2 7B.2.2.1 Type A 7B.2-2 7B.2.2.2 Type B 7B.2-2 7B.2.2.3 Type C 7B.2-2 7B.2.2.4 Type D 7B.2-3 7B.2.2.5 Type E' 7B.2-3 7B.3 Criteria 7B.3-1 7B.3.1 General Requirements 7B.3-1 7B.3.2 -Equipment Design and Qualification Criteria 73.3-1 7B.3.2.1 Design and Qualification Criteria - Category 1 7B.3-1 7B.3.2.1.1 Selection Criteria - Category 1 7B.3-1 7B.3.2.1.2 Qualification Criteria - Category 1 7B.3-1 7B.3.2.1.3 Design Criteria - Category 1 7B.3-1 7B.3.2.1.4 Information Processing and Display Interface 7B.3-2 Criteria - Category 1 7B.3.2.2 Design and Qualification Criteria - Category 2 7B.3-3 7B.3.2.2.1 Selection Criteria - Category 2 7B.3-3 7B.3.2.2.2 Qualification Criteria - Category 2 7B.3-3 7B.3.2.2.3 Design Criteria -- Category 2 7B.3-3 7B.3.2.2.4 Information Processing and Display, Interface 7B.3-4 Criteria - Category 2 7B.3.2.3 Design and Qualification Criteria - Category 3 7B.3-4 7B.3.2.3.1 Selection Criteria - Category 3 7B.3-4 7B.3.2.3.2 Qualification Criteria - Category 3 7B.3-4 7B.3.2.3.3 ' Design Criteria - Category 3 7B.3-5 7B.3.2.3.4 Information Processing and Display, Interface 7B.3-5 4
Criteria - Category 3 7B.3.3 Extended Range Instrumentation Qualification 7B.3-5 Criteria I
7B-1 Amendment 40 ;
I
T e
STP FSAR ATTACHMENT I App 2ndix 7B ST HL AE- 14fD PAGE WS OF #1 TABLE OF CONIENTS (Continued)
Chapter 7 Appendix 7B Page Stetion Tcble 7B.3-1 Summary of Selection Criteria for 7B.3-7 Category 1,2,3 versus Type A,B,C D.E .
Tcble 7B.3-2 Summary of Design, Qualification, and 7B.3-8 Interface Requirements TYPE A VARIABLES 7B.4-1 7B.4 7B.4-1 7B.
4.1 INTRODUCTION
7B.4-2 Table 7B.4-1 Summary of Type A Variables 7B.5-1 73.5 TYPE B VARIABLES 7B.5-1 7B.
5.1 INTRODUCTION
7B.5-2 Table 7B.5.1 Summary of Type B Variables TYPE C VARIABLES 7B.6-1 40 7B.6 7B.6-1 7B.6.1 Introduction 7B.6-2 Table 73.6-1 Summary of Type C Variables TYPE D VARIABLE 7B.7-1 7B.7 7B.7-1 7B.7.1 Introduction 7B.7-3 Table 7B.7-1 Summary of D 7ariables 7B.8-1 7B.8 TYPE E VARIABLES 7B.8-1 7B.8.1 Introduction 7B.8-2 Table 7B.8-1 Summary of Type E Variables Table 7B.9-1 !!) f V ri:51: : d C:te; crier - Cr : -!
" fer re: Lieting of Veri:ble: ef Typ: ^ " C,D,E r!!b C2te; cry
!,' 3 l-l 7".10 1 Table 7B.be-E NUREG-0737 Conformance Nova to pay 73.i belo 73,g. 3 7B-il Amendrent 40
E STP FSAR i Apptndix 7B ATTACHMg@g ST HL AE-oAGE 44.0iF 7B.1 DISCUSSION An analysis was conducted to develop a response to Regulatory Guide (RC) 1.97, Rev. 2. This analysis identified the appropriate variables and established appropriate design bases and qualification criteria for instrumentation employed by the control room operr. tor during and following an accident.
This design basis establishes the key and preferred backup variables to be monitored by the control room operating staff of the South Texas Project (STP) X
^
following the initiation of an s.ccident. The design basis recognizes the variables essential to the ontrol room staff up to the time other ,(mergency A reg genst7 0 . . c ir; [acilities are manned as well as the information essential to X the control room staff in subrequently controlling the plant and proceeding to safe shutdown conditions. Also included, to aid the system designer,are cri- X teria for determining the requirements for the instruments used to mEnitor these variables.
The selection of variables was integrated with the Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGS) in accordance with the guidance on integration of energency response capability elements outlined in NUREG-0737 Supplement 1 (See Appendix 7A, Item S.3).
This was accomplished by performing a task analysis based upon the WOG ERCS to identify those variables necessary for implementation of the guidelines. The 40 Optimal Recovery Guidelines (ORGs) were reviewed to determine those Type A Q32.
variables necessary to (a) perform diagnosis, (b) take preplanned manually 18 controlled actions and (c) take actions necessary to reach and maintain a con-trolled condition. The Critical Safety Function (CSF) Status Trees were reviewed to determine those Type B variables necessary for the operator to determine if a Functional Restoration Guideline (FRG) should be implemented.
Furthermore, the FRGs were reviewed to determine those Type B variables neces-sary to assess the process of accomplishing or maintaining CSFs, i.e., sub-criticality, reactor core cooling, heat sink maintenance, RCS integrity, con-tainment environment and RCS inventory. The ERGS were also reviewed to deter-mine those Type D variables necessary for (a) monitoring those plant safety systems employed for mitigating the consequences o' an accident and subsequent plant recovery and (b) other systems normally employed for attaining a cold shutdown condition. Finally, the ERGS were reviewed to determine those Type E variables necessary to (a) determine the accessability of areas at the plant following an accident and (b) continually assess the release of radioactive materials due to the accident.
Utilization of this task analysis process ensures that the plant information utilized by the plant operators following an accident to implement the STP Emergency Operating Procedures (EOPs) is obtained from specially designed and qualified instrumentation as defined in this design basis.
The WOG ERGS, the results of the Control Room Design Review (see Appendix 7A, Item I.D.1), and the interpretation of RG 1.97 Revision 2, as described in this Appendix, will be used to develop STP E0Ps that are human factored, function oriented, and integrated with the plant design.
7B.1-1 Amendment 43 l 1
J
ATTACHM(m0NT I STP FSAR ST HL-AE. ,
PAGE 45 OF 8l Apptndix 7B The de*_ ailed methodology for the handling of displays was addressed g the design of the Qualified Display Processing System (QDPS) and in conjuption [ =
with the Control Room Design Review programs to address NUREG-0696 and NUREG-0700 (See Appendix 7A, Item S.5). Section 7B.3 describes interface criteria which must be satisfied for the display methodology to meet the in-tent of RG 1.47 Revision 2 and this design basis.
7B.1.1 Planned Versus Unplanned Operator Actiors 40 Q32.
The plant safety analyses and evaluations define the design basis accident 16 event scenarios for which preplanned operator actions are required. Accident monitoring instrumentation is necessary to permit the operator to take required actions to address these analyzed situations. However, instrumen-tation is also necessary for unplanned situations (i.e., to ensure that, should plant conditions evolve differently than predicted bf the safety anal-yses, the operator has sufficient information to monitor ti.e course of the event). Additional instrumentation is also needed to indicate to the operator whether the integrity of the fuel clad, the Reactor Coo 1. ant System (RCS) prec-sure boundary, or the reactor containment has degraded beyond the prescribed limits defined as a result of the plant safety analyses and other evaluations.
Such additionni requirements are considered by this design basis.
7B.I.2 Variables Types Five classifications of variables have been identified. Operator manual ac-tions identified in the operating procedures, associated with design basis accident events, are preplanned. Those variables that provide information needed by the operator to perform these manual actions are designated Type A.
The basis for selecting Type A variables is given in Section 7B.2.2.1.
Those variables needed to assess that the plant critical safety fuictions are being accomplished or maintained, as identified in the plant safety analysis 40 and other evaluations, are designated Type B.
Variables used to monitor for the significant breach or the potential signif-icant breach of fuel clad, the RCS pressure boundary, or the reactor contain-ment, are designated Type C. Type C variables used to monitor the potential '
breach of containment have an arbitrarily-determined, extended range. The extended range is chosen to minimize the probability of instrument saturation even if conditions exceed those predicted by the safety analysis. The response characteristics of Type C information display channels allow the con-trol room operator to detect cenditions indicative of significant failure of any of the three fission product barriers or the potential for significant failure of these barriers. Although variables selected to fulfill Type C functions may rapidly approach the values that indicate an actual significant failure, it is the final steady-state value reached that is important.
Therefore, a high degree of accuracy and a rapid response time are not neces-sary for Type C information display channels.
Those varinbles needed to assess the operation of individual safety systems 1 and systems normally used to attain cold shutdown are designated Type D.
7B.1-2 Amendment 40 i
r
--ATTACHMENI i
- STP FSAR ST-HL AE- N30 Appandix 7B PAGE tilo 0F fl l
The variables that are required for use in determining the magnitude of ;
f release and continually assessing any releases of radioactive materials are designated Type E.
-The five classifications are not mutually exclusive in that a given variable (or instrument) may be included in one or more types. The cross-referencing y ef[ariableto[ypeisgiven.inTable 7 3. 9- 1. '*- '7,,5~- 1
-Tab.le'18. -l ' identifies the instruments utilized at STP which address the 7".j^
/
recommendations of both NUREG-0737 and RG 1.97 Revision 2. The instruments identified meet the intent of the guidance provided in NUREG-0737.
78.1.3 Design and Qualification Criteria Three categories of design and qual!fication criteria have been identified.
The differentiation is made in order that an importance of information hier-archy can be recognized in specifying post-accident monitoring instrumen-tation. Category 1 instrumentation has the highest pedigree and should be 40 utilized for primary information which the operator should use for preplanned manual actions and determining the state of the plant. Category 2 and 3 in- ,
struments are of lesser importance in determining ,of the plant and do not l require the same level of operational assurance. \,4,4 i M
The primary di erences between category requirements are in qualification, application of single failure criterion, power supply, and display require- /
ments. Category I requires seismic and environmental qualification, the application of a single failure criterion, utilization of emergency standby
{ power, and an immediately accessible display. Category 2 requires quali-fication commensurate with the required function but does not require the single failure criterion, emergency standby power, or an immediately accessible display. Category 2 requires, in effect, a rigorous performance verification for a single instrument channel. Category 3 does not require qualification, single failure criterion, emergency standby power, or an
-immediately accessible display.
(
7B.1-3 Amendment 40
L ATTACHMENT l ST HL-AE- 14i% N
), 7
,PAGE e OF BT 1 gg.,
f06' ( '/
g6yg (cg 0 STP FSAR % bit WGV#d p/ g Appendix 7B p 7 f./g -/ fo 7 O l'Y d o#
f po" b
S$ (**
),O' g.l TABLE 7B.10-1
~
Y Y O g W hotpen cese h NUREG-0737Cd//dAWff y Applicable Section of NUREG-0737 o Varia'A e'-
ay SJety Valve.
- h. II.D.3 Pressurizer PORVA Status )(
ment O7 II.F.1 Attach 4 Containment Pressure (Extended Range) h II.F.1 Attach
,g Containment Water LevellWdt N*T.
ava Nau w Racp')
Y h II.F.1 ~ Attach 6 Containment H 2 Concentration II.F.2 Core Exit Temperature klater Reactor Vessel3 Level Y RCS Subcooling 40 h I.D.2 Emergency Response Facilities y
Data Acquisition and Display h I.E.1.2 ment Auxiliary Feedwater Flow II.F.1 Attach 3 Containment h Radiation A
^ 9(iigh Ran W nent II.F.1 Attach 2 Unit Vent Sample
,,g c.naense. Vacuum Tap b @c. lA II.F.1 Attachpl Unit Vent A cticit; X ex~
Condense 44. Vacuum Pump Ef fim. . X Me4* Steamline Radiation y ECCS and Other Systems Valve O// II.K.1.5 Status 1-9 7B. M Amendment 40
^ ATTA.CHMENT I ST-HL-AE- IWD PAGE 8 OF 85 l 3.fP FSAR Appendix 7B 7B.2 DEFINITION OF VARIABLE TYPES
[
7B.2.1 Definitions I 78.2.1.1 Design Basis Accident Events. Pesign basis accident events are those events, any one of which.may occur du ing che lifetime of a particular plant, and those events not expected to occur A , postulated because their consequences would include the potential for release of significant amounts of radioactive gaseous, liquid, or particulate matet'ai to the environment.
j Excluded are those events (defined as " normal" and " anticipated operational l occurrences" in 10CFR50) expected to occur more frequently than once during the lifetime of a particular plant. The limiting accidents that were used to l determine instrument functions are: 1) LOCA, 2) Steamline Break, 3) Feedwater Line Break, and 4) Steam Generator Tube Rupture.
7B.2.1.2 Safe Shutdown (Hot Standby). The state of the plant in which
, the reactor is suberitical such that K is less than or equal to 0.99 and io' the RCS temperature is greater than or*kual to 350*F.
7B.2.1.3 Cold Shutdown. The state of the plant in which the reactor is suberitical such that K is less than or equal to 0.99, the RCS temperature islessthan200*F,and*beRCSpressureislessthanorequalto10CFR50 Appendix G limits.
78.2.1.4 Controlled Condition. The condition that is achieved when the 40 plant has been stabilized using the ORG D the recovery procedures are being /
( 1mplemented and the critical safety functions are being accomplished or main-taintd by the control room operator.
7B.2.1.5 Critical Safety Functions. Those safety functions that are es-sential to prevent a direct and immediate threat to the health and safety of <
the public. These are the accomplishing or maintaining of:
- 1. suberiticality
- 2. reactor core cooling
- 3. heat sink maintenance o
- 4. RCS integrity
- 5. containment environment
- 6. RCS inventory 7B.2.1.6 Inunediately Accessible Information. Information that is visu-ally available to the control room operator, or is accessible through the exe-cution of the EOPs.
7B.2.1.7 Primary Information. Information that is essential for the direct accomplishment of the preplanned mantal actions specified in the ERCS; it does not include those variables that are associated with contingency
( actions.
l 7B.2-1 Amendreent 40 l
ATTACHMENT I ST-HL AE-14So AGE IM OF 8\
STP FSAR Appendix 7B 7B.2.1.8 Key Variables. Those variables which provide the most direct measure of the information required.
7B.2.1.9 Back Information. Back!informationisthatinformation,madeup [
cf additional variables beyond those classified as key, that provides cupplemental and/or confirmatory information to the operator. Backup vari-cbles do not provide an indication which is as reliable or complete as that provided by. primary' variables, and they should not be relied upon as the sole cource of information. Those backup variables which should be first consulted by the operator are designated as preferred backup variables.
7B.2.2 Variable Functions The accident monitoring variables and information display channels are those that are required to enable the control room operating staff to perform the functions defined by Types A, B, C, D, and E below.
7B.2.2.1 Type A. Type A variables are those variables that provide the primary information required to permit the control room operating staff to:
o Perform the diagnosis specified in WOG ERGS e Take the specified preplanned manually controlled actions for which 40 r.o automatic control is provided, that are required for safety systems to accomplish their safety function in order to recover from the Design Basis Accident (DBA)[ vent,and e Reach and maintain a safe shutdown condition.
'The verification of the actuation of safety systems have been excluded from the definition of Type A. The variables which provide this verification are included in the definition of Type D.
Variables in Type A are restricted to preplanned actions for EBA events.
7B.2.2.2 Type B. Type B variables are those variables that provide to the control room operating staff information to assess the process of accomp-lishing or maintaining critical safety functions, i.e., suberiticality, reac-tor core cooling, heat sink maintenance, RCS integrity, containment environ-
- g menthRCSinventory. The WOG contingency guidelines which go beyond the [
design basis were reviewed for additional variables which may be utilized as variable types B, C, D, and E.
7B.2.2.3 Type C. Type C variables are those variables that provide the control room operating staff information (1) to mor.itor the extent to which variables which indicate the potential for causing a significant breach of a fission product barrier have exceeded the design basis values and (2) that the fuel clad, the reactor coolant system pressure boundary (RCPB) or the reactor containment may have been subject to significant breach. Excluded are those associated with monitoring of radiological release from the plant which are included in Type E. ,
7B.2-2 Amendment 40
r .
ATTACHMENT I ST-HL AE- WID STP FSAR PAGE 40 OF 3)
Appendix 7B Type.C variables used to monitor the potential for breach of a fission product
( barrier have an arbitrarily-determined, extended range. The extended range is chosen to minimize the probability of instrument saturation even if conditions exceed those predicted by the safety analyses.
7B.2.2.4 Type D. Type D. variables are those variables that provide to the control room operating staff sufficient information to monitor the performance of:
- 1. Plant safety systems employed for mitigating the consequences of an acci-dent and subsequent plant recovery to attain.a safe shutdown condition (These include verification of the automatic actuation of safety systems).
40 A
- 2. Systems normally employed for attaining a cold shutdown condition.
7B.2.2.5 Type E.s Type E variables are those variables that provide to the control room operating staff information to:
- 1. Monitorthehabitdbilityof/controlroom, h
- 2. Monitor plant aret.s where access may be required to service equipment necessary to monitor the progress of or mitigate the consequences of an accident.
- 3. Estimate the magnitude of release of radioactive materials through identi-fied pathways, and continually assess such releases, and
(- 4. Monitor radiation levels and radioactivity in the environment surrounding the plant (via portable monitors).
l l
1 7B.2-3 Amend:sent 40
F ATTACHMENT I STP FSAR Appsndix 7B ST PAGEHLt;IAE 0F Nfo@
7B.3 CRITERIA 7B.3.1 General Requirements The following design and qualification criteria are applied to instrumentation for Type A,"B, C, D and E variables. These are summarized in Tables 7B.3-1 and 7B.3-2. ,
7B . 3". 2 Equipment Design and Qualification Criteria.
The qualification _ requirements of the Type A, B, C, D, and E accident moni-
.toring instrumentation are subdivided into three categories (1, 2, 3).
Descriptions of the three categories are given below.- Table 7B.3-2 briefly summarizes the design and qualification requirements of the three designated categories.
7B.3.2.1 Design and Qualification Criteria - Category 1.
7B.3.2.1.1 Selection Criteria - Category 1: The selection criteria for Category 1 variables have been subdivided according to the variable type. For Type A, those. key variables used for diagnosis or providing infornation for necessary operator action have been. designated Category 1. For fype B, those X key variables which are used for monitoring the process of accomplishing or maintaining critical safety functions have been designated Category 1. For Type C, those key variables which are used for monitoring the potential for 40 breach of a fission product barrier have been designated-Category 1.
as amassed 7B.3.2.1.2 Qualification Criteria - tegory 1: The instrumentation is environmentally and seismically qualifiedy n .;;;rd_r.;; sin FC.'? Sections N 3.11 and 3.10, respectively. Instrumentation continues to read within the required accuracy following but not necessarily during a seismic event. At least one instrumentation channetfis qualified from sensor to display. For %
the balance of instrumentation channels, qualification applies up to and including the channel isolation device. (Refer to Section 7B.3.3 in regard to extended range instrumentation qualification).
7B.3.2.1.3 Design Criteria - Category 1:
- 1. No single failure within either the accident monitoring instrumentation, its auxiliary supporting features, or its power sources, concurrent with the failures that are a condition of or result from a specific accident, prevents the operator from being presented the required information.
Where failure of one accident monitoring channel results in information ambiguity (e.g., the redundant displays disagree), additional information is provided to allow the~ operator to analyze the actual conditions in the plant. This is accomplished by providing additional independent channels of information of the same variable (addition of an identical channel), or by providing independent channels which monitor different variables which bear known relationships to the multiple channels (addition of a diverse channel (s)). Redundant or diverse channels are electrically irdependent and physically separated from each other, to the extent practicable with train separation, and from equipment classified as mensafety-related in accordance with RG 1.75.
W 7B.3-1 Amendment 40
ATTACHMENT l ST-HL AE- lL 70 STP FSAR PAGE 6pO M Appsndix 7B s
For situations such as isolation valves in series, the intent is generally ,
to verify the isolation function. In such a situation a single indication on each valve is sufficient to satisfy the single failure criterion if _
those indications are from different trains (i.e., unambiguous indication of isolation). .
If ambiguity does not result from failure of the channel, then a third redundant or di. verse channel is not required.
' 2. The instrumantation is energized from station emergency standby power sources, battery backed where momentary interruption is not tolerable, as rr--i'M i- RG 1. 32.
&<es %
The out-of-service interval is based on normal Technical Specification 3.
requirements on out-of-service for the system it serves where applicable or where specified by other requirements.
- 4. Servicing, testing, and calibration programs are specified to maintain the capability of the monitoring instrumentation. For those instruments where '
the required interval between testing is less than the normal time inter-val between generating station shutdown, a capability for testing during l N power operation is provided. $ f40
- 5. Whenever means for removing channels from service are included in the design, the design facilitates administrative control of the access N
y such removal means.
- 6. The design facilitates administrative control of the access to setpoint adjustments, module calibration adjustments, and test points.
i
- 7. The monitoring instrumentation design utilizes human-factored displays to minimize indications potentially confusing to the operator. - i
- 8. The instrumentation is designed to facilitate the recognition, location, a replacement, repair, or adjustment of malfunctioning components or l modules.
! 9. To the extent practicable, monitoring instrumentation inputs are from i sensors that directly measure the desired variables. An indirect measure-ment is made only when it can be shown by analysis to provide unambiguous information.
- 10. Periodic checking, testing, calibration, and calibration verificatico are in accordance with the applicable portions of RG 1.116.
J
- 11. The range selected for the instrumentation encompasses the expected oper-ating range of the variable being menitored to the extent that saturation ;
does not negate the required action of the instrument in accordance with !
the applicable portions of RG 1.105.
73.3.2.1.4 Information Processing and Display Interface i
[ Criteria - Category 1: The interface criteria specified here I Provide ( requirements implemented in the establishment of the design basis for N. 8 processing and displaying of the information.
7B.3-2 Amendment 40
ATTACHME !
/
ST-HL-AE- I STP FSAR PAGE G3 OF $
Appendix 7B
- 1. The operator has immediate access to the information from redundant or di-( verse channels in the units familiar to the operator (e.g.,'for a temper-aturereadinggegreesnotvolts). Were two or more instruments are needed to cover a particular range, overlapping of instrument spans are
)(
provided.
^
A y 2. .. hi:t:rie:1 ::;e d ;f ; :.:nir Of ;;; i..e6 aaien ;t;.;; 1 fer --9 jr : a ..sieL1. 1. -ointained. A recorded pre-event history for these .
channels is required for a minimum of one hour and continuous recording of these channels is required following an accident until such time as con-tinuous recording of such information is no longer deemed necessary. This recording is available when required, but need not be immediately acces-sible. One hour was selected based on a representative slow transient p -
wnich is oun -by this time requirement. A one-half inch equivalent break area Loss-o -Coolant Accident (LOCA) was selected since trip occurs at ap-4 proximately fifty minutes after break initiation. Were direct and imme-diate trend or transient information is essential for operator information or action, the recording,is immediately accessible.
1 7B.3.2.2 Design and Qualification Criteria - Category 2.
7B.3.2.2.1 Selection Criteria - Category 2: The selection criteria for Category 2 variables are subdivided according to the variable type. For Types A, B, and C, those variables which provide preferred backup information are
- designated Category 2. For Type D, those k variables that are used for non-
, itoring the performance of safety system e designated Category 2. For 40 )
Type E, those key variables to be monitored for use in determining the mag-
' '{- nitude of the release of radioactive materials and for continuously assessing such release designated Category 2. Ml 7B.3.2.2.2 Qualification Criteria - Category 2: Category 2 instrumen-tation is qualified from the sensor up to and including the channel isolation device for at least the environment *(seismic and/or environmental) in which it must operate to serve its intended. function. Instrumentation associated with those safety-related systems that are required to operate following a Safe o
Shutdown Earthquake (SSE) to mitigate a consequential plant incident are seis ,
mically qualified in accordance with IEEEf344-1975. Category 2 instrumenta-
[
tion is environmentally qualified in accordance with IEEE 323-1974. X h
- 7B.3.2.2.3 Design Criteria - Category 2:
4
- 1. Category 2 instrumentation associated with those safety-related systems
, that are required to operate following an SSE to mitigate a consequential plant incident are energized from -c-!--icelly ;;1144ed power source,
, not necessarily the en'ergency standby power, which is battery-backed where momentary interruption is not tolerable. Geheew6ee-sh: in:: n .:;;i;; 1; crrr;;ie ' frc: c 'hi;;hly reltati; en :10: ;=:: ::::::, nrt 20 :::::117 :h;
-- :-- y -*- '57 ;rrr, --hich i: Sett:ry 5::h:f d.;;; ::::;;:ry int:; ;g tic i: :: ::1;;;ti;. yak
- 2. The out-of-service interval is based on normal Technical Specification j [ requirements on out-of-service for the system it serves where applicable
\ or where specified by other requirements.
i i
2
- 73.3-3 Amendment 40
ATTACHMENT l
, TN56X 7' A ST HL-AE-14fD
.EAGE 64 OF 8%
p, s ga f,A p ,,tg Pmm 67 sk (GDPQ yr.Aks i J;u-% A~ pl ~~ dayly b Aich A w 4al 1. -
y u L g h % nyai & . D.'ylay ./ g% 1.
m-in -a. .-ad;Aly m Ma. 4 A-ymk a~~s;yel t
-' p A ba % a & m N G 0f 5 p w s., dispI y n Vs. 1Asaa w;.htw w:II k.d.yt y a k . l2 y n y:.,,Ju- %4 e s . m r , , o, a , w . u e x dcy I Q ~4-p/~ diyly ab ;, -p . syle ~ s+
o ~pahJte. A" h4 n. Ahd s, pal M d 6 J reAud t w &:% alJek re.%.1k- ya<h ./'
t k. L w L d valid 4 1,v calecLL. Leb:da.A w una- w - a.,rn;lds. ~ fox awfdyt y
,,;g
- 3. L addah. 4 -flu GDPS p /wn d;yfq p huel ~du.s p a s da m L m . eli x n a u L;sk;uA <ued -fw at IM ~ CAM of uch g% d.
m;ds. 1%u.- ra.<.e dw.s aa Acakd ~ -h- %hv I <ow.
~% -a a:n ~.t'l ~ p. 7s.s->
1 1
ATTACHMEN I STP FSAR ST HL-AE Di 0 Appendix 7B PAGE % OF El 7B.3.2.3.3 Design Criteria - Category 3:
- 1. Servicing, testing, and calibration programs are specified to maintain the capability of the monitoring instrumentation. For those instruments where the required interval between testing is less than the normal time inter-val between generating station shutdowns, a capability for testing during power operation is provided. .
- 2. "Whenever means for removing channels from service are included in the
- design, the design facilitates administrative control of the access to such removal means.
- 3. The design facilitates administrative control of the access to setpoint adjustments, module calibration adjustments, and test points.
- 4. The monitoring instrumentation utilizes human-factored displays to mini-mize indications potentially confusing to the operator.
- 5. The instrumentation is designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.
- 6. To the extent practicable, monitoring instrumentation inputs are from sensors that directly measure the desired variables. 'An indirect measure-4 ment is made only when it can be shown by analysis to provide unnabiguous information.
7B.3.2.3.4 Information Processing Display, Interface Criteria -
Category 3: The interface criteria sTecified here provide 40 requirements considered in the establishment of the design basis for proces-sing and displaying of the information.
- The instrumentation signal is, as a minimum, processed for display on demand.
Recording requirements are variable specific and are determined on a case-by-case basis.
7B.3.3 Extended Range Instrumentation Qualification Criteria The qualification environment for extended rang: information display channel I
components are based on the design basis accident events, except the assumed maximum of the value of the monitored variable is the value equal to the spec-ified maximum range for the variable. The monitored variable is assumed to approach this eak by extrapolating the most severe initial ramp associated with the DBA ents. The decay for this' variable is considered proportional to the decay or this variable associated with the DBA vents. No additional qualificationmarginneedstobeaddedtotheextended[rangevariable. The [(
)
environmentc1 envelopes, except that pertsining to the variable measured by the information display channel, are those associated witi the DBA The environmental qualification requirement for extended range equ paent does ents. K not account-for steady-state elevated levels that may occur'in other environ-
. mental parameters associated with the extended range variable. For example, a sensor measuring containment pressure must be qualified for the measured 4 - (- process variable range (i.e., 3 times design pressure for concrete 1
7B.3-5 Amendment 40
- _ . _ _ _ ~ . , _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . . _ _ _ . _ . _ , _ - . _ _ . - . , _ . .
~
I ATTACHMENT i STP FSAR ST HL-AE WJD PAGE 4L, OF 8l Appendix 7B containments), but the corresponding ambient temperature is not mechanis-tically linked to that pressure. Rather, the ambient temperature value is the bounding value for design basis accident events analyzed in Chapter 15.;f ;;. o y . The extended range requirement is to ensure that the equipment will 40 continue to provide information if conditions degrade beyond those postulated in the safety analysis. Since extended variable ranges are nonsechanistically determined, extension of associated parameter levels is not justifiable-and is therefora not required.
i 7B.3-6 Amendment 40
N Table 7B.3-1 Summary of Selection Criteria CATEGORY 2 CATEGORY 3 TYPE CATECORY 1 Variables which provide None ,
A KEY variables that are used '
for diagnosis or providing PREFERRED BACKUP information.
information for necessary operator action. ~
Variables which provide Variables which provido B KEY variables that are used BACKUP information.
for monitoring the process of PREFERRED BACKUP information.
accomplishing or maintaining ^
critical safety functions.
Variables which provide Variables which provide u C KEY variables that are used BACKUP information. 40 m for monitoring the potential PREFERRED BACKUP information. Y F
Y breach of.a' fission product g
" barrier.
N KEY variables which are used Variables which provide PREFERRED D None for monitoring the performance BACKUP information rM ' crr cred N of plant systems used to attain for$o#"itoringtheperformancenf n /
a controlled plant condition or plant systems used to attain a controlled plant condition or a a safe shutdown condition.
safe shutdown condition.
KEY variables to be monitored Variables to be monitored which E None provide PREFERRED EACKUP informa- MD for use in determining the magnitude of the release of tion for use in determining the y radioactive materials and magnitude of the release of radio- g i for continuously assessing active materials and for continu- _smg
$ such releases, ously assessing such releases. g
{E 9 .,_
8 1
ATTACHMENT
. A 0 SYP FSAR gg Table 75.3-2 Summary of Design, Qualification, and Interface Requirements 1
(
Category 1_
Y hCategory 2 a Category 3_
Qualification ,
Environmental Yes As appropriate U) No ,
1 e
Seismic Yes As appropriate Q) No Design Single Failure Yes No No Power Supply Emergency Reliable As Required Standby 40 l l
Channel out of Service Technical Technical No Specifications Specifications Testability Yes Yes As Required l Interface Minimum Immediately Des.and Demand Indication Accessible Recording Yes As Required As Required Qualdi A 55 cram ( .
f-
= i " " "' " T " '
J l o . . .. yeTA W/'i ca6/t Pr o 5vo" t oCFR50 I R.<g u e*h fly,dv8 kr- ( L)
Qa$c adnM& b A D1 (q
ap fh ' '
& ckAmm.L ' '
~
fdbw l .
4 W m4M pau1 ykt t w cc udua -)u%L.
Z.) M b MMA CLC 0 5 ") k
.gu a crua h i rttl & Tf*" % JYT*
T
! rd E m d atte m - Amendment 40 75.3-8
STP FSAR '-ATTACHMENT I Appandix 7B AE FifD ST PAGEHL69 OF R\
7B.4 TYPE A VARIABLES 7B.4.1 Introduction Type Ag[ariables are defined in Section 7B.2.2.1. They are the variables /(
which provide primary information required to permit the control room operating staff to: .
- 1. JPerform'the diagnosis specified in the WOG ERGS
- 2. Takespecifiedpreplannedmanuallycontrolledactionsfforwhichnoauto- /(
matic control is provided, that are required for safety systems to accom- 40 plish their safety function to recover from the DBAp K'v ent (Verification of f(
actuation of safety systems is excluded from Type A and is included as Type D);
- 3. Reach and maintain a safe shutdown condition O Key Type A variables have been designated Category 1. These are the variables which provide the most direct measure of the information required.
No Type Aj V'ariables have been designated Category 2 or 3. )(
The Type A variables are listed in Table 7B.4'1.
. i 1
l i
7B.4-1 Amendment 40
STP FSAR ATTACHMEN I Appindix 7B ST HL-AE. m b PAGE (,0 OF M TABLE 7B.4 1 hotore cas g y SW1 ht Thf A V)(f//jff)
C ieseri /
- 1. RCS Pressure (Wide Range) Al
- 2. Hot Leg Reactor Coolant Temperature (Wide Range T ) Al
/
- 3. Cold Leg Reactor Coolant Temperatur5 (Wide Range Tcold} ^
W4Ter
- 4. Wide Range Steam Generator g Level Al X (130ft f
- 5. Narrow Range Steam Generator3 Level Al Y U)e tt f
- 6. Pressurizerg level Al 5
- 7. Pri : 7 2 ::::: Containment Pressure Al X
- 8. Steamline Pressure Al
- 9. Refueling Water Sterage Tank (RWST)4L evel Al )(
- 10. Containment Water Level (Wide Range) Al 40
- 11. Containment Water Level (Narrow Range) A1 (AFp(ST's kJa4er
- 12. Auxiliary Feedwater Storage Tank Level 3
Al )(
- 13. Auxiliary Feedwater Flow Al l 14. High Range Containment Radiation Level Al
- 15. RCS Pr - re (Extended Range) Al LNd X
- 16. Steam b ...cator Blowdown Radiation M: nit:r- Al Lv er
- 17. Steamline Rediation ."enn.e.\ Al /
- 18. Core Exit Temperature Al
- 19. RCS Subcooling Al n
7B.4-2 Amendment 40
=
STP FSAR ATTACHMENT I ST-HL-AE- NfD Appzndix 7B PAGE lA OF $
7B.5 TYPE B VARIABLES 7B.5.1 Introduction Type B variables are defined in Section 7B.2.2.2. They are the variables that provide to,the control room operating staff information to assess the process of accomplishing or maintaining critical safety functions, i.e.,
~
1..Suberiticality
- 2. Reactor Core Cooling
- 3. Heat Sink Maintenance Reac.or d
Coelant S 3 mm 40 4.gCS) Integrity Y
- 5. Containment Environment Reo& Cooknt $3 stem 6 4(RCS) Inventory (
Variableswhichprovidethemostdirectindication(i.e.,[f/ variables)to X assess each of the 6 critical safety functions have been designated Category
- 1. Preferred backup variables have been designated Category 2. These are listed in Table 7B.5 gl. All other backup variables have been designated y Category 3. g l
7B.5-1 Amendment 40
AppIndix 7B ST-HL-AE- 1430 PAGE WOF 8\
Table 7B.5fl X Summary of Type B Variables Categ ory K Subcriticality . Key: a. Neutron Flux (Extended Range) B1
- b. Neutron Flux Startup Rate B1
~
Preferred a. Wide Range T B2 N Backup: b. Wide Range gold Reactor Core Cooling Key: a. Core Exit Temperature B1
- c. RCS Subcooling B1
- d. AFS evel B1 Y
- , e. RWS Level w 4tr B1 }
Preferred a. Wide Range T B2 /
Backup: b. Wide Range B2 X
- c. RCS Pressur TN) B2 Y E et Heat Sink Maintenance Key: a. Narrow Ra Level B1 /
- b. k'ide Rang %-Level B1 y
- c. Auxiliary Feedwater Flow B1 g T d MfrT- Level B1 )r M e. Steamline Pressure
. B1
- f. Core Exit Temperature B1
- g. Wide Range T B1 %
~h. Wide Range od Preferred a. Main Steamline Isolation B2 Backup: Valve Status
- b. Main Steamline L,vlaim. # B2 X Bypass Valve Status Reactor Coolant System Key: a. RCSPressure(W[ B1 Y (RO} Integrity b. RCS Pressure (Extended Range) B1 y Preferred a. Containment Pressure B2 Backup: b. High Range Containment Radiation Level v B2
- c. NarrowRangeSGhve4u l B2 #
GG d. 9fe-Blowdown Radiation Level B2 Y Steamline Radiation Level B2
- f. Pressurizer PORV Status B2
- g. Pressurizer Safety Valve Status B2 7B.5-2 Amendment 40
ATTACHMENT I ST-HL-AE LWD STP FSAR PAGEla OF 8l Appendix 7B l
Table 7B.5 (Continued) y Summary of Type B Variables j Coiegoc1 )
?. ;; :: s/ Key: a. Containment Pressure- B1 1
Containment Environment b. High Range Containment.
, Radiation Level r e. B1
- c. Containment Water Level ( W B1 Y
- d. Containment Hydrogen B1 Concentration 0
Preferred None Backup:
%W #
Reactor Coolant Key: a. Pressurizer, Level B1 Sysi:em Inventory b. Reactor Vessel Water Level B1 Y A aleRaqL-Preferred Containment Water Level ( B2 X QO) a.
Backup: b. Containment Water Level (NA-) B2 y
- c. Wide Range Steam Generator B2 A Narrow Ea ng g, tueW 7B.5-3 Amendment 40
ATTACHMEb i ST-HL-AE F 7B.6 TYPE C VARIABLES 7B.6.1 Introduction Type C variables are defined in Section 7B.2.2.3. Basically, they are the variables that provide to the control room operating staff information to mon-itor the potential for breach or actual significant breach of:
- 1. Fuel Clad; (R6) -
y
- 2. Rpactor oolant System Boundary; 3 or
- 3. "can s. Containment Boundary.
(Variables associated with monitoring of radiological release from the plant are included in Type $ N Those Type C key variables which provide the most direct measure of the h/ATENTIALforbreachofoneofthe3.fissionproductboundarieshavebeendes- /
ignated Category 1. Backup information indicating potential for breach is designated Category 2. Variables which indicate actual breach have been des-ignated as preferred backup information and have been designated Category [
- 2. All other backup variables have been designated Category 3.
Table 7B.6-1 summarizes the selection of Type C variables.
7B.6-1 Amendment 40
gh T Mi h*I k p ld HL 0 p*
STP FSAR PAGE(16 0 RI '/ [ g yy j) p VW App;ndix 7u hU hA\p ,
b TABLE 7B.6 I E Summary of Type C Variables
- c. cw" # Sr POTENTIAL FOR BREACH C. hoe t ACTUAL BREACH g O*D 8# ' ' y IN-CORE ,
Key: e Exg Temperature Backup: RCSSampling(C3f- ? )
FUEL CLAD:
(Cl i, Backup:
";;ferr;d Reactor
/
IVessel Water Level (C2)
RCS BOUNDARY Key: RCS Pressure (Cl)--+ preferredBackup: RCS Pressure b (Extended Range) 4 W (Wide Range) )r (C2) )
RCS Pressure (Cl) --> Containment M p (Wide Range) Pressure (C2)1
}
ContainmentWatep >
tad. R* *H Lev (C2)
Containment Waterp- J M0 N** D3t. Levep (C2)- J >
Steamline Rndi- p't ;
ation Level (C2)I 40 SG (W e Blovaovn ,
Radiation Level #
(C2)
High Range Containment Radi
,ation Level (C2)
CONTAINMENT Key: Containment Ecessure }PraferredBackup:'UnitVentRadi- )
(Extended Range) (Cl) t A Ption Level (C2) b0UNDARY g gyp A us T' Fuel Handlin .,
ContainmentJressure (Wido
-3, (Cl)---+ Buildin adiation Level (C2)
Codm**JHydrogen Concentration Containment Iso-(Cl)- ) lation Valve Status (C2)
Containment Pres- 3 sure (Extended '
Range) (C2)--- -- ~~
7B.6-2 Amendment 43
STP FSAR ATTACHMENT l ST-HL AE Ngo Appzndix 7B PAGE bb OFKi TABLE 7B.6-Q(Continued)
Summary of Type C Variables en W Cyfrppi POTENTIAL FOR BREACH 44, ACTUAL BREACH -
CONTAINMENT Backup: Site Environmental 40 BOUNDARY (Cont'd) Radiatfo g(Portable ked r Monitoring}(C3)P Adjacent Building d Area Radiation l Level (C3)-
7B.6-3 Amendment 43 e
ATTACHMEN4 ST.HL-AE- N3 i STP FSAR PAGE b10F 31 7B.7 TYPE D VARIABLES 7B.7.1 Introduction Type D variables are defined in Section 7B.2.2.4. Basically, they are those variables that provide sufficient information to the control room operating staff to monitor the performance of: .
- 1. JPlant safety systems employed for mitigating the consequences of an acci-dent and subsequent plant recovery to attain a safe shutdown condition, including verification of the automatic actuation of safety systems; and
- 2. Other systems normally employed for attaining a cold shutdown condition.
bcaS t.
Typed /$/variablesaredesignatedCategory2. Preferred backup information )(
is designated Type D Category 3.
The following systems have been identified as requiring Type D information to be monitored:
- 1. Pressuriner Leve1 and Pressure Control (assess status of RCS following return to normal pressure and*1evel ebntrol under certain post-accident-conditions) 40
- 2. Chemical and Volume Control System (CVCS) (normally employed for attaining a safe shutdown under certain post-accident conditions)
- 3. Secondary Pressure and Level Control (employed for restoring / maintaining a secondary heat sink under post-accident conditions)
- 5. Auxiliary feedwater II T
- 6. Containment Systems
- 7. Component tooling water (CCW) )(
s
- 8. Essential Cooling Water (ECW) 63577m
- 9. Residual Heat Removal (RHR)3(normally employed for attaining a cold hl shutdown condition) -
- 10. Heating, ventilation, and air conditioning (HVAC) if required for Engineered Safety Features operation
- 11. Electric power to vital safety systems
- 12. Verification of automatic actuation of safety systems t
. Table 7B.7-1 lists the key variables identified for each system listed above and specifies the seismic and environmental qualification for each variable.
7B.7-1 Amendment 40
ATTACHMENT (
ST.HL-AE fl480 STP FSAR PAGE 61 OF 8l Appendix 7B For purposes of specifying seismic qualification for Type D Category 2 vari-ables, it is assumed that a seismic event and a break in Category I piping will not occur concurrentiv. As a result, the limitinc event is an unisolated (Jsingle failure of a main steam ,'_y J Iation valve (MSIV)1) break in non-Category F main steam pipin f lustrumet. cation associated with the safety systems which 40 are required to mitigate and monitor this event should be seismically qual-ified instrumentation. Similarly, the environmental qualification for Type D Category 1. variables depends on whether the instrumentation is subject to a high-energy line break (HELB) when required to provide information.
7B.7-2 Amendment 40
c ..
TABLE 7B 7-1 SoemmeH obType D Variables ,
Y System Designation a r i ab le '- * -"- * -
- i - Seismic Environmental fategari /t
- 1. PressurpzerLeveland Pressurizer PORV 5tatus Ye s HELB D2 PressurI. nee Control Pressurizer PORV Block Valve Status Yes HELB D2 Pressurizer Safety Valve Status Yes HELB D2 Pressurizer Spray Valve Status No Ambient D2 Pressurizer Heater Breaker Position No Ambient D2 PressurizerAl evel a w c Yes HELB D2 Y Reactor Vessel Water Level Yes HELB D2 RCS Pressure ( g eo Je 43 g, Yes HELB D2 y Pressurizer Pressure Yes HELB D2 RCP Status No Ambient D2
- 2. CVCS Charging Flow No Ambient D2 Letdown Flow No Ambient D2
- y. g VCT Level No Ambient D2 y Ro Seal Injection Flow No Ambient D2 40 )I h d, Yes (Isolation Valve Status 5 valves only) Ambient D2 $
Charging Pump Status Yes Ambient D2 BAT Pump Status Yes Ambient D2
- 3. Secondary Pressure and SCrsf& PORV Status Yes HELB* D2 )(
Level Control Main Steamline Isolation Valve Status Yes HELB* D2 Main Steamline Bypass Valve Status
$6 6/G-Safety Valve Status Yes Yes HELB*
HELB*
D2 D2 )/
[3 m
Steamline Pressure Yes HELB* D2 r>I MFW Control Valve Status Yes HELB* D2 g MFW Control Bypass Valve Status Yes HELB* D2 n z y MFW Isolation Valve Status Yes HELB*
D2 oo g MFW Isolation Bypass Valve Status Yes HELB* D2 )(
S
$ *These systems must be qualified to the worst case environment in which they must function (including HELBs inside and outside containment).
i
TABLE 7B.7-1 (Continued)
Susamert of Type D ariables ,.,
System Designation M Variable !-.M i.......ta '...
. Seismic Environmental
- 3. Secondary Pressure MFW Flow No Ambient D2 and Level Control (Cont'd) Auxilipry Feedwater Flow Yes HELB* D2 6G S/t.)Svel (""' ...? (""'(Wdeba9# 8@7 )Yes HELB D2 X SG t/C-Blowdown Isolation Valve Status Yes * * ' ..; N Et,8#
D2 X SG S/C-Blowdown Sample Isolation Valve I Status Yes Amh& ear $6L6' D2 MIer #
- 4. ECCS RWST evel Yes Ambient D2 Teee4.HHSI Flow Yes HELB** D2 X 15es.e1.LHSI Flow (QSyge JMrimoggo HELB** D2 )r Containment Water LeveT 4 (Z) onu (Z)- No HELB D2 y Pump and Valve Status Yes llELB** D2 M Accumulator Pressure Yes HELB D2 40 m
eg I. 5. Auxiliary Feedwater Auxiliary Feedwater Flow. Yes HELB* D2 U
Pump and Valve Status y, tats- Yes HELBf D2 $
Auxiliary Feedwater Storage TankgLevel Yes Ambient D2 X
- 6. Containment Systems Containment Spray Flow w ,g,q,a,gu, % % ,No HELB** D2 Containment Water Level % =} ono doa) No HELB D2 X SpraySystegValveStatus No HELB** D2 j( i
, Fan Cooler Differential Pressure / z Status No HELB D2 yg3 f'g*pa Containment Pressure No HELB D2 ph>g m
Containment Isolation Valve Status Yes HELB* D2 Containment Ventilation Gemper Status Yes HELB* D2 y j$$
Nahl4 Mm o$
a g T~
,,. *These systems must be qualified to the worst case environment in which they must function (including HELBs inside and outside containment).
- These systems may see radiation from components in the recirculation path.
TABLE 7B.7-1 (Continued)
So....y e/ Type D Mariables y System Designation Variable ' m tr-;:ntrt'-- Seismic Environmental (s try ,4 %
- 7. EfMCompohani bohnj Pump Discharge Pressure Yes Ambient D2
(.uder Header Temperature Yes Ambient D2 Surge Tank Leve W ater Yes Ambient D2 Flow to ES omponents Yes llELB* D2 Pump and Valve Status Yes Ambient ///gt$ D2 %
- 8. Essential Cooling Water Flow io GSF Comgewnts Yes Ambient D2 Y Syntr Pump and Valve Status Yes MF4,A - O dent D2 K
- 9. RHR System lleat Exchanger Discharge Temperature No HELB** D2 Flow No llELB** D2 Pump and Valve Status No llELB** D2 Yes HELB D2 X RCS Pressure ( $ gg,j,g ,qg,, a
- 10. HVAC Environment 'SF "I~empey,'tuf t Yes HELB** D2 W
- 11. Electric Pswer
.sr c.. .. eser 51=ws Standby Power and Emergency Source yes wa s *
- o2. Kl]px Status Yes Ambient D2 Other Safety-Related Energy Sources Yes Ambient D2 Reactor Trip Breaker Position Yes HELB D2 N
- 12. Verification of Auto-matic Actuation of Turbine Covernor Valve Position No No D2 Safety Systems Turbine Stop Valve Position No No D2 Kg Auxiliary Feedwater Pump Status ciH
$' (turbine) Yes HELBi D2 Y *d y Auxiliary Feedwater Pump Status (motor driven) Yes HELB g D2 X o m SI Pump and Valve Status Yes a-h * ,41 ELE D2 X g3 9-E -
- These systems must be qualified to the worst case environment in which they must function (including IIELBs inside and outside containment).
- These systems may see radiation from components in the recirculation path.
l
TABLE 7B.7-1 (Continued)
So-1
- AType D d variables ,.
System Designation ariable !...,a......; ;i. ! Seismic Environmental Ca%P3ai
- 12. Verification of Auto- CCW Pump and Valve Status Yes /bekMT[ELB D2 matic Actuation of ECW Pump and Valve Status Yes i+Pt:fr AmEIPot D2 Safety Systems (Cont'd) Containment Spray Pump and Valve bratus Yes "" t en: #Et5 d D2 '
Neutron Flux (Extended Range) Yes HELB U D2 Neutron Flux Startup Fate Yes HELB F D2 Containment Isolation Valve Status y, Joe. Yes HELB* D2 Containment Ventilation % ' Status Yes HELB* D2 RCB Fan Cooler Differential Pressure /
Status No HELB D2 SI Actuation Status Yes Ambient D2 Containment Isola on Actuation Status Yes Ambient D2 Control Rod Position (Backup) No Ambient D3 m 40 h&rcn%n i S {
4 i
)
N#4 -
j rn c)%>H l
g d i>
m a rme S
a
?z9w -4
\ n c_a d
- These systems must be qualified to the worst case environment in which they must function (including HELBs inside and outside containment).
- These systems may see radiation from components in the recirculation path.
i
ATTACHMENT I STP FSAR ST-HL AE N" Appindix 7B PAGE 13 OF 3\
7B.8 TYPE E VARIABLES 7B.8.1 Introduction u.
Type E variables are defined in Section 7B.2.2.5. They are these variables that provide the control room operating staff with information to:
- 1. Monitor,the habitability of control room,
- 2. Monitor plant areas where access may be required to service equipment necessary to monitor or mitigate the consequence8 cf an accident, 40 )(
- 3. Ectinate the magnitude of release of radioactive materials through iden':1- }
-fied pathways and tenhd3 assess such releasesj awd )(
- 4. Mor.itor radiation levels and radioactivity in the environment surrounding the plant (via portable monitors). A A A Key Type E variables are qualified to Category 2 requirements. Preferred backup Type E variables are qualified to Category 3 requirements. Table f 7B.8- lists the Type E variables.
7b.8-1 Amendment 40
ATTACHMENT i STP FSAR ST-HL-AE PAGE 1:4 OF l@M Appendix 7B TABLE 7B.8g C S/1%(d dF TYPE E VMIABLES
- 1. Control Rcom Habitability 0# Y Control Room Radiation Lev e.\ E2 X
- 2. Post Accident Access 6 AreaRadiatio)n Post Accident Sampling Station E3*
Operational Support Center E3*
Cers 'tt f y Emergency Operations F::ili:3 E3*
Unit Vent Monitoring Station E3*
- 3. Release Pathways High Range Containment Radiation Level E2 Steamline Radiation Level & Relief Valve Status E2 Vent Unit Vent Radi.-no nrrrtivi:3 Level andp Flow E2 X li Condenser Vacuum Pump Radiation Level & Flowrate b '=
gt. ,, ,,c. 4 ,eu d a . Pu y w p stdvs E3 y 2, hf Exknus T y FHBg Vent Radiation Le.oe,I n E2 Containment Sump & Atmospheric Sampling E3 S:::: rencr;;;; Radi :ic; Level i Elcud r Fler Ecte -fd Y Status >.
Co..dar.;;;; Polich Fediatier le"-1 1 71 ~3 Da*-/Firr 50 ; ; E3 >
Valut.,
Liquid Radwaste Radiation Level & F1. E:::/F1: 1 Status E2 X Li j ui 4 bJ cuesTe Flow Ret E3 TCE Drai; R;dictier Lavc.1 i ils S : /rls St : ; E2 ' )
- Category 3 per Regulatory Guide 1.97, Revision 3 )
7B.8-2 Amendment 40
ATTACHMENT I STP FSAR ST-HL-AE 1460 Appendix 7B PAGE 76 0F M TABLE 7B.8-1 (Continued)
].awer cass S)O!ARY df TfW E VJ0tf8tts
- 4. Site Environmental Radiation Level Ooi'18"1 Area Monitors (Portable) -
E3*
~
Meteorological Parameters E3*
40
~
- Category 3 per Regulatory Guide 1.97, Revision 3 7B.8-3 Amendment 40 1 1
ATTACHMENT /
ST-HL-AE- N30 STP FSAR PAGE % OF M AppIndix 7B TM LE Dsit.EM SECTION 7B.9
~'
IF8' T% M __7 E*I Table 7B.9-1
-g y- - -- . r ne aio. - > ce r ;,,-4..
trv-irdir; E:ler*4-- rf r 2) .-
ar'iable Type and Category Type Type Type Ty e Type A B C E RCS Pressu e (Wide Range) 16[ 1,2 1,i/ 2 /
Wide Range T h t 1/ 1,$
//
Wide Range T cold 1 / 1,V Wide Range S/G Level 1/ 2/ 2-Narrow Range S/G Level 1- 1,I. 2/
Pressurizer Level / 1 #' 2<
~
Containment Pressure 1- 1,2- ~1,I- 2- 40 Steamline Pressure 1- 1/ 2 RWST Level 1- 1 2<
Containment Water Level (WR) 1,I' 2 2 Containment Water Level ) 1. 2 2. 2-Auxiliary Feedwater S rage Tank Level 1- <' 2-Auxiliary Feedwate Flow 1- 1. 2 High Range Cont ament Radiation Level 1. l,2- 2- 2.
S/G Blowdown adiation Level 1 2v 2 -
Steamline adiation Level 1. 2, 2 2
Core E t Temperature 1 1/ 1 ,-
RCS ubcooling 1. 1
' utron Flux (Extended Range) 1 . 2 Neutron Flux Startup Rate. 1 2-i
$ ,$0' l
i 7B.9-1 Amendment 40 I
ATTACHME l )
ST-HL AE-l P STP FSAR PAGE 11 OF R\
Appindix 7B Table 7B.9-1 (Continued) [
Summary of Variables and Categories (Excluding Selection of D-3)
Variable Type and Cate cry Type Type Type Type Type A B C D E Containment Hydrogen neentration 1/ 1 Reactor Vessel Water Le 1 1./ ,/ 2-Main Steamline Isolation V ve Status 2 - ?' 2 Main Steamline Bypass Valve S atus 2- 2.
Control Rod Position Ind. 3f Pressurizer Pressure 2 .
RCP Status 2 ,.
40 Pressurizer Spray Valve Status 2' Containment Pressure (Extended Range) 1,2 '
RCS Pressure (Extended Range) 1 l 1-Containment Isolation Valve Status 21 ~ 2-Unit Vent Radiation Level 27 26 RCS Sampling (Primary Coolant etivity) 3.
~
Fuel Handling Bldg. Radiati Level 2' 2. ,
Adjacent Building Area Ra ircion Level 3-Site Environmental Rad tion Level 3 3-Pressurizer PORV Val e Status 2. 2 .-
Pressurizer PORV ock Valve Status .
Pressurizer Saf ty Valve Status 2y 2 Pressurizer ' ater Breaker Position 2/
Charging S stem Flow 2 ['
n 78.9-2 Amentiment 40
. ATTACHMENT /
ST-HL-AE. It/fo STP FSAR PAGE12 OF F1 Appendix 78 7;ble 7B.9-1 (Continued) e pg 8 Summary of Variables and Categories (Excluding Selection of D-3)
Variable Type and C egory Type Type Typ Type Type A B C D E Letdown Flow 2s VCT Level 2-CVCS Valve Status _
2 *:'
RCP Seal Injection Flow 2 S/G Atmospheric.PORV Status 2 -c' c_ '-
S/G Safety Valve Status 2 -
Main Feedwat;c Control Valve Status 2-Main F/* .ontrol Bypass Valve Status 2c- 40 Main F/W Irolation Valve Status 2 .; > '
Main Feedwater Flow - 2-S/G Blowdown Isolation Valve Stat s 2 S/G Blowdown Sample Isolation Ive Status 2
'.uarging Pump Status 26 Main Feedvater Isolation alve Bypass Valve Status 2 5-Auxiliary Feedwater solation Valve Status "7, . _ , ,. 2 Containment Venti tion Damper Status Total HHSI Flow 2..
Total LHSI F v . 2 ECCS Valve' Status 2-ECCS Ac mulator Pressure 2 7B.9-3 Amendment 40
,s --n-- --- , - - - , _ , - - -
-.Q - - , - - - -
ATTACHMENT I STP FSAR ST-HL.AE IV9 Appsndix 7B PAGE *hl OF 91 Table 7B.9-1 (Continued) 1 (O
Summary of Variables and Categories 90 (Excluding Selection of D-3)
Variable Type and Cate ory ,
.I '
Type Type Type Type Type A B C D E Aux F/W Valve Status 2[
AUX F/W Motor Driven Pu Status 2.
Aux F/W Turbine Driven Pu Status 2' Turbine Stop Valve Position 2 Turbine Covernor Valve Position 2 s' Containment Spray Flow 2 .'J Containment Spray Pump Status 2 6' 40 Containment Spray System Valve Status 2</
RCB Fan Cooler Differential Pressure / Stat 2#
CCW Pump Discharge Pressuru 2.
CCW Header Temperature 2 CCW Surge Tank Level 2 6.-
CCW Flow to ESF Components 2 '"
CCW Valve Status 2..>
CCW Pump Status 2' RHR Pump Status 2.?
Essential Cooling Wat System Flow .'
Essential Cooling W er System Valve Status 2' Essential Coolin ater System Pump Status 2(
RHR Heat-excha er Discharge Temp 2a RHR Flow 2/'
7B.9-4 Amendment 40
ATTACHMENT I ST HL-AE-14dO STP FSAR PAGE 90 OF 8l Appendix 7B a(CL- ,
Table 7B.9-1 (Continued) 8-Summary of Variables and Categories (Excluding Selection of D-3)
Variable Type and C egory Type Type Ty Type Type A B C D E RHR Valve Stat 2<
ESF Environment 2-SI Actuation Status 2.'
SI Pump Status 2- s '
SI Valve Status - _
2 Containment Isolation Actuat n Status 2 ~#'
Boric Acid Transfer Pump Status 2 Standby Power and Emergency Power our 2 40 Status Other Safety Related Energy Source 2.
tb Reactor Trip Breaker Position 2.
Control Room Radiation -- 2 Acce , Area Radiation 3
, FHJ Vent Radiation 2 ~
Meteorological Param ers 3-
- ~
Condenser Vacuum P p Effluent Radiogas 3 Concentration Effluent Path ow Rate
^ ~ ~
Steam G erator Blowdown 3 Cond sate Polish , ,
3
+
L uid Radsaste 3-7B.9-5 Amendment 40
r-ATTACHMEfi (
ST-HL-AE- 14 STP FSAR PAGE 91 OF l Appendix 7B Table 7B.9-1 (Continued) e ( & re.-
Summary of Variables and Categories (Excluding Selection of D-3) f Variable Type and Cate ry .
~
Type Type Type Type Type A B C D E TGB Drain . 3' Unit Vent 2 >>
Condenser Vacuum Pump 3 .. .'
Concentration From Liquid Pat ays ,
2- 40 Steam Generator Blowdown Condensate Polish .:2 Liquid Radwaste 2.gr TGB Drain 2 Effluent Pathway Status Steam Cenerator Blowdewn Valve Sta s 2 Condensate Polishing Valve Stat 2 Liquid Redwaste Valve Status 2' TGB Drain Valve Status 2 Condenser Vacuum Pump St tus 2 i
Containment Sump & Atmosp ric Sampling 3 .'
l 1 -
l 1
h 7B.9-6 Amendment 40
. - _ _ . , _ _ . - . - . . _ . , .- - . . .