ML20084F427
| ML20084F427 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/24/1984 |
| From: | GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML20084F418 | List: |
| References | |
| NUDOCS 8405040139 | |
| Download: ML20084F427 (84) | |
Text
C SAFETY PARAMETER DISPLAY SYSTEM ACCIDENT MONITORING INSTRUMENTATION DATABASE IDENTIFICATION STUDY FOR RIVER BEND STATION UNIT 1 l
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GULF STATES UTILITIES COMPANY POST OFFICL BOX 2431 OA1ON ROUGE. LOUISIANA 70H21 ARLA CODE SO4 767 1802 April 24, 1984 RBG- 17,668 File Nos. G9.5, G9.33.4, G9.8.6.2 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Denton:
River Bend Station - Unit 1 Docket No. 50-458 In Gulf States Utilities Company (GSU) August 31, 1983 submittal concerning Generic Letter 82-33, " Supplement 1 to NUREG-0737-Requirements for Emergency Response Capability," a l Safety Parameter Display System (SPDS) implementation plan and
. operability schedule (item 4. 2.b) was provided based on the currently projected River Bend Station (RBS) fuel load date of April 1985. The schedule identified the SPDS safety analysis report to be submitted in May 1984. In support of the RBS Safety Evaluation Report issuance, the RBS SPDS safety analysis report is attached for Nuclear Regulatory Commission Staff review to demonstrate compliance with Generic Letter 82-33 item 4.2.a.
Additional submittals addressing Detailed Control Room Design Review, Regulatory Guide 1.97, Revision 3 and Emergency Operating Procedures Generation Package will follow.
Sincerely, J. E. Booker Manager-Engineering Nuclear Fuel & Licensing f River Bend Nuclear Group JEB/WJR/ /je Attachment i
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SAFETY PARAMETER DISPLAY SYSTEM ACCIDENT MONITORING INSTRUMENTATION DATABASE IDENTIFICATION STUDY FOR RIVER BEND STATION UNIT 1 TABLE OF CONTENTS Page Introduction..................................................... 1 0bjective........................................................ 1 Methodology...................................................... 1 Findings......................................................... 2 Conclusions...................................................... 3 Figure 1 Implementation Methodology..................................... 4 Table 1 Legend......................................................... 5 Composite List of Accident Monitoring Variables................ 6 Notes......................................................... 20 Table 2 Accident Monitoring Instrumentation Study Findings............ 26 References...................................................... 49-Appendix A Typical BWR Functional Event Tree.............................A-1 Anticipated Transient Event Tree..............................A-2 Operator Action Event Tree....................................A-3 Small LOCA Event Tree.........................................A-6 Operator Action Event Tree....................................A-7 Large LOCA Event Tree.........................................A-8 Operator Action Event Tree....................................A-9 Summary of Variables Identified in Sequence Evaluations......A-10
' Appendix B Emergency Operating Procedure Variable List...................B-1 Notes.........................................................B-2 Appendix C......................................................C-1 Appendix D Variable Category Analysis....................................D-1 Table D-1.....................................................D-3
-Table D-2.....................................................D-7
i INTRODUCTION The accident at Three Mile Island Unit 2 has served to focus Industry attention on the need for adequate instrumentation and human-factored displays for plant operators to follow and help mitigate the consequences of various plant transients. The NRC Staff and the Industry have commissioned several studies (References 1 through 5) to identify the subject instrumentation. River Bend Station (RBS) has reviewed the existing literature on this subject and has within this study documented a listing of needed ,
instrumentation which is specific to the plant. This report establishes a listing of RBS-specific plant variables which is used as a basis upon which to compare the Safety Parameter Display System (SPDS) database and the inventory of main control room instrumentation to be reviewed by the Detailed Control Room Design Review (DCRDR) study.
OBJECTIVE The objective of this study is to validate an RBS-specific listing of parameters that will be available for monitoring to furnish control room operators with sufficient information to mitigate or limit the consequences of abnormal and accident events.
Additionally, the study endeavors to describe the basis upon which RBS believes that the listing is necessary and sufficient to assess the safety status of the plant. The listing delineates specific instruments which are used to monitor each identified parameter.
It is intended that this study will serve as the licensing basis for demonstrating compliance with the guidance provided by the NRC Staff in Generic Letter 82-33 Item 4.2.a.
METHODOLOGY The validation of an RBS-specific listing of accident monitoring variables required an action plan depicted in Figure 1 and described below to insure a necessary and sufficient list of variables:
- 1) Several event tree analysis studies (References 1 and 2) were reviewed for information and a tabulation of variables to be monitored was developed.
Additionally, the RBS emergency operating procedures (EOP's) were reviewed to define specific instrumentation needs for implementing required operator actions defined therein.
- 2) The two lists generated (Appendix A'and B) were merged with the variable list identified in Table 2 of Regulatory Guide 1.97, Revision 3 to form a composite variable list (Table 1) which would encompass the monitoring of a large number of possible events.
1
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- 3) RBS-specific information such as instrument identification numbers and location of benchboard displays was added to the !
list compilation.
1 A separate review (Appendix D) was performed to verify the l categorization of each variable identified in the composite list in terms of.importance of its function. This data was entered into the list.
- 4) The composite variable list (Table 1) was compared to the SPDS database of monitored variables to insure its adequacy-based on the study results.
Further, the plant design was reviewed against Table 1 to identif; opesi.fic deviations from regulatory direction. While every effort s ss made to identify RBS deviations, there still remains equipment qualification-related work to confirm required accuracies, operability times, and response times for the
- subject instruments identified in Table 1. Therefore, RBS will continue to appraise the existing plant design for equipment
, qualification-related anomolies as part of the RBS equipment qualification program.
4
- FINDINGS j RBS has identified several differences in the plant design with the-
- regulatory direction provided by R.G. 1.97, Revision 3 as well as
- the results of this study. Table 2 provides a formal listing of_the findings together with an evaluation of each. Specific design revisions, if required, and an Laplementation schedule for same i
shall be developed and documented as a part of the DCRDR Summary
{ Report to be submitted to the NRC Staff for review during October, 1984.
{. The findings were generated as a result of a sequential review of each variable in the composite variable list against regulatory and technical guidance which RBS has committed to comply with. The review criteria consisted of, but were not limited to, the following:
3 .1) The variable was reviewed to insure that the necessary i
information was transmitted to the SPDS database.
- 2) The implamentation of the variable (i.e., proper range,-
acceptable power supply, redundancy as requ' ired, etc.) was checked against R.G. 1.97, Revision 3 criteria for agreement.
I
- 3) Additionally, the implementation of the variable was reviewed against the study results which include variables identified by-the event tree analysis and the emergency operating procedures.
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l' CONCLUSIONS The existing design of RBS accident monitoring instrumentation and the completeness of the SPDS database in conjunction with the resolution of the findings identified herein are deemed sufficient j to safely operate the plant given current design bases.
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l FIGURE 1 ACCIDENT MONITORING INSTRUMENTATION STUDY IMPLEMENTATION METHODOLOGY EVENT TREE REG. GUIDE 1.97 ANALYSIS " REVISION 3 R D ES k k VAR VAR VAR LIST LIST LIST e e e Y
CATEGORIZATION ^^
STUDY ARIA E . MNGE
,\ LIST . LOCATION
. INSTRUMENT LD
. SEQUENCE NO. I
. SPDS SIG ID P
LICENSING AND REVIE TECHNICAL -
LIST FOR REQUIREMENTS SCREPA ES DCRDR l
1r RESOLUTION OF DISCREPAN-CIES 4
TABLE 1 LEGEND SEQUENCE NO. - Each distinct variable is identified by a unique sequence number for reference purposes.
SOURCE - An "X" is placed in the appropriate column if a variable was identified by the particular method shown.
INSTRUMENT ID - This is the alphanumeric designator assigned by Stone &
Webster Engineering Corporation (SWEC) for the sending instrument (s).
CATEGORY - This is the category designation assigned by GSU Engineering from its independent review delineated in Appendix D. The numbering convention definitions are the same as those used in R.G. 1.97, Revision 3.
MCR DISPLAY BENCHBOARD - The SWEC identification number is shown for the panel (s) where the instrument channel display is located.
SPDS SIG ID - This alphanumeric number is the unique identifier assigned by General Electric (GE) to identify the variable monitored by the SPDS database. GE is furnishing the RBS SPDS.
RANGE - The current design range of each listed instrument is given.
DIVERSE BACKUP - This column identifies any variable (s) which serves as a diverse backup for the present variable. The diverse backup variable is shown by referencing its sequence number as indicated in Table 1.
REFERENCE NOTES - Table 1 is furnished with a REFERENCE NOTES section at the end of the table for inclusion of technical comments.
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TABLE 1 COMPOSTTE t.IST Ol' ACCIDENT MONITORING VARIABLES d
)o SOURCE u
o ha g a. N MCR DISPLAY SPDS h .
2 8 h "
- VARIABLE e INSTRUMENT ID BENCH 80ARD SIG ID RANCE X 1821*LTN044C.D 1 IN13*P601 521EA001 209 - 409 in (Fuel Zone) 1 1 Beactor Vessel Water Level X X 007 521 - 581 in (Narrow)
- PTN062A B 1H13*P680
, *LTN081C C33EA024 361 - 581 in (Wide Range)
- LTN027 027 521 - 921 in (Shutdown) 521 - 701 in (Upset)
IC33*LTN017
- LTN004A,B,C 1413*P680 B21EA008 0-1200 psig 2 Beactor vessel Pressure ' X X X 2C33*PTN005 1
- PTN008A 850-1050 psig a 1821*PTN062A,3 1H13*P601 013 0-1500 psig C33EA028 033 .
ICNS*PT2A,B 1 1H13*P808 CMSPY024 0-30 psis.
3 Drywell Pressure X X X 025 4 Containment /Drywell 0 - 10% Hydrogen (Narrow) 2 Rydrogen Concentration X X X ICMS*AT25A,B 1 IH13*P808 CMSYY001 002 0 -'30% Hydrogen (Wide) minus 18 - 4 ft measured 3
$ Suppression Pool Level X X X -ICMS* LIT 23A,B 1 IN13*P808 CMSLY028 029 from normal pool level W+ _. _ _ _ _ __
.- -. . . .-a....-~. - . _ _ _ _ . . - _ - . . . . - _.. _ , _ _ .- . ~ _ _ . . . . - -
TABLE 1 COMPOSITE I.iST or ACCIDE!!T M0!!ITORI!!G VARIABLES d
4 SOURCE y v - n a
=
8 w %
" ww M n. d Q MCR DISPLAY SPDS b(;
VARIABLE E O N INSTRUMENT ID BENCH 80ARD SIG ID RANCE 6 Suppression Pool Temperature X X X ICMS*TT24A,C.E 1 1H13*P808 CMSTY005 0-200*F
.I 4 *TT248,D,F
- TT24G,J-
- TT24H,K 018
- TT40A,C 1
- TT40B,D 7 Control Rod Position X X X 1813-D124 3 IH13*P680 C11EC004 Full In/ Discrete Inter- 4 (Typical for 145 Rods) Channel A & B mediate Positions / Full Out
'8 LPCS Flow X X X 1E21*FTN003 2 1H13*P601 E21EA001 0-7000 gpa 9 Condensate Storage Tank X X X ICNS-LT110 3 IH13*P680 CNHLY006 Top to Botton 20 Level 10 SLCS Tank Level X X X IC41*LTN001 2 IH13*P601 C41EA002 0-5000 gal 4 Il SRV Position X X X ISW*ZE10 2 1Hl3*P601 B21EC042 Full closed / Intermediate / 19 A-N, J-N, P-R Full Open 060
_ _ _ - . _ _ _ _ . _ _ _ _ . _ - _ _ _ _ _ ___-_ - - _ _ _ _ _ --__m_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
TABLE 1 COMPOSITC LIST Or ACCIDENT MONITORING VARIABLES 4
d SOURCE u H k 5 ha u p. y MCR DISPLAY SPDS h VARIABLE E O e INSTRUMENT ID BENCHBOARD SIG ID RANGE f
12 Primary Containment 5 CD Radiation X X X 1 IH13*P879
- P878 Primary Containment 1RMS*RE16A,B 1.0 10 R 10 33 ; 10 {hrC1/cc Atmosphere *RE111
- II Drywell Atmosphere 1RMS*RE112 10 - 10'I C1/cc IRMS*RE20A,B 1.0 - 10 R/hr Drywell Area 1.0 - 10 nr/hr
. Drywell Personnel Airlock 1RMS-RE138 5
Containment Effluent X X X 1RMS*RE125 2 10*I - 10 C1/cc 6 13 Radioactivity *RESA,B 14 Radiation Level in Circu- 21 lating Primary Coolant X 15 Drywell Atmosphere 29 Temperature X X -ICMS*RTD41A,B,C,D 1 1H13*P808 CMSTYO26 0 446 F 027 16 Containment /Drywell 7 Oxygen Concentratica X X 1
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TABLE 1
! COMPOSITC !.f ST or ACCIDDIT HO!!ITORI!!G VARIABLES T
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A 4 NCR DISPLAf SPDS
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- VARIABLE e INSTRUMENT ID BCNCHBOARD SIG ID RANGE i
17 Drywell Equipment and Floor 5 - 25 spa O Drain Sump Water Level X X 1E31*LTN093 2 1H13*P632 DFRLY001
- LTN094 005 l
- LTN095 DERLYOO3 l
- LTN096
- LIN097 I 18 Neutron Flux X X X '1 IEt3*P680 C51EA003 3
1 - 120 percent full power 8 LPRM Detectors 1C51*JEN011
- JEN012
- *JEN013
- JEN014 ,4 IC51*JEN002A 010 5 X 10 - 10.0 percent IRN Detectors
- JENC028 021 full power ,
- JEN002C
- JEN002D
- JEN002E 024
- JEN002F
- JEN002C
- JEN002H SRM Detectors- 3C51*JEN001A,B,C D 10"# - 10~ percent full power t
t _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ -.
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TABLE 1 I
l COMPOSITC f.1ST Or ACCIDE!JT M0!!ITORit!G VARIABLES i
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" SPDS VARIABLE E! INSTRUMENT ID
" MCR DISPLAY '$*
BENCHBOARD SIG ID RANCE i
19 Main Steam I,ine Radiation ID17*REN003A,B,C,D g
o X 3 1H13*P680 1.0 - 10 ar/hr 9 20 Status of HPCS DC X 1H13*P601 3 Available/ Unavailable 10 21 RER HX Inlet / Outlet Temperatures X X 1E12* TEN 004A,B 2 1H13*P601 E12EA124 0* - 600 F
- TEN 002A,8
- TEN 027A,B
- TEN 005A,8 129
- TEN 003A,B
. IRHS*RTD47A,B f 22 Containmer.t Isolation Valve Positions X X 1 1H13*P863 Open/ Intermediate / Closed 11
- P601
- P870 23 RPV Baron Concentration X X 3 12 (grab) 24 RHR Flow X X IE12*FTN015A,B,C 2 IH13*P601 E12EA005 0 - 8,000 gym i
007 i
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. . - ~ . - . . . - . -. _ . , . - . . . , . _. _ . - _ - . . -. - . --- - . - -._ --_---_.- - .- -
TABLE 1 COMPOSITC !.IST OF ACCIDEt4T MO!!!TORIt3G VARIABt.ES l
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h U HCR DISPLAY SPDS b'd l
VARIABLE M 8 e INSTRUMEllT ID BENCH 80ARD SIG ID RANCE 25 MSIV Positions X X 1821*A0VF022A,B,C,D 1 Ilf13*P601 B21EC070 Open/ Closed 13 g" *A0VF028A,B C.D 074 076 077 0 79 080 082 083 085 086 088 089
[ 091 092
- 26 RCIC Flow X X 1E51*FTN003 2 Illl3*P601 ES1EA005 0-800 gpa 006 6
27 Feedwater Flow X X 3 IH13*P680 C33EA019 0 - 8 X 10 lbs/hr Pump A 3C33-FTN002A 020 Pump B -FTN0028 i
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TABLE 1 Cor:POSITE f.fET or ACCIDDIT Mot!!TOPflir. VAPTART.ES
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" 12 tid VARIABLE lE " #*
e, IIISTRUME!!T ID ItElicl! BOARD SIG ID RAtlCE t 28 HPCS Flow X X 1E22*FTN005 2 1H13*P601 E22EA001 0-7000 gym g 006 9-29 Area Radiation X X 3 1RMS- 14 ,
DSPL230 i
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30 Airborne Radioactivity X X 1RMS*RE125 2 1RMS- 10 -
10 Ci/cc 15 Releases *HE5A,8 ~DSPL230
-RE6A 31 Core Temperature X X 16 32 Suppression Pool Nydrogen/
0xygen Concentration X 3 27 33 Containment Water Level X ~3 II 34 Containment Atmosphere c 0 Temperature X X ICMS*RTD42A-G,J 3 '1H13*P808 CMSTYO28 0 - 200 F 037
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TABIC 1 COMPOSITE I.1ST Or ACCi!>EllT f10!!!TORiflG VARIAB!.ES
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O eh MCR DISPfAy SPDS VARIABI.C IE " g *U INSTRUMENT ID DENCll80APD SIG ID RAtlCC 35 Primary containment Pressure X X X 1 s
G Annulus Differential 1HVR*PDT60A thru F HVRPY222 25 223 Drywell Differential ICPfS*PDT29A,B 1H13*P808 CMSPY019 minus 15 - 30 psid 020 Absolute , ICMS*PT4A,B 1Hl3*P808 0 - 75 psia 36 RCIC Turbine Speed X 1E51*PC002-1 3 1H13*P601 E51EA014 0 - 6000 rpm "37 Reactor Cooldown Rate X 1821-N029A,B 3 1H13*P614 B21EA022 0 - 600*F 26
-NO30A,B
-N050A,8 38 Turbine Stop Valve Positions X IC71*ZSN006A-H 3 1H13*P870 N32EA001 Open/ Closed 9 002 003
> 004 39 Turbine Control Valve Positions X 11tSS-HYVCVI 3 1H13*P870 N32EA005 Open/ Closed 9
-NYVCV4 008
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f' TADI.C 1 Cot 4POSITC t.TST OF ACCIDE!47 Hot!!TORiflG VARIABI.ES l b SOURCC y u > u h S h d O
u NN N 0- D" MCR DISPLAY SPDS 8-g VARIABLE 2 8 eM INSTRUffENT ID BENCHBOARD SIG ID rat:CC 48 Steam Flow to RCIC p Turbine X IE51-PTN007 3 IH13*P601 E51EA007 0-1500 psig w
49 RifR Key Valve Positions X 3 till3*P601 - E12EC004 Open/ Closed 18 006 008 i
0I2 014 i 016 018 50 LPCS Key Valve Positions X 3 Illl3*P601 Open/ Closed 18 l
51 Recirculation Line Flow X IB33-FTN0llA,B 3 !!!!3*P680 B33EA021 0 - 40,000 spe 024 025 52 flain Steam Line Tiow X IB21-FTN003A.B,C,D 3 lill3*P680 0 - 4 X 10 lbm/hr 3
l 53 circulating Water Flow I ICWS-PDT101 3 till32P680 20 54 SSW Discharge Pressure / X X - ISW PT50A,8 2 Illl3kF870 ShTFY004 0 - 150 psig Flow *FT60A,B 005 0 - 18,000 gpa 006 l
- - - - _ _ _ _ - . _ _ _ _ - _ _ = _ - - _ _ .-
TABLE 1 COMPOSITE 1.IST OF ACCI!iE!!T Hor 3! TORI!3G VARIABLES d
] SOURCE M
g ,
- 5 5 8
- 8 ,
O "
n.
50 N SPDS b$
2 O eh
" MCR DISPLAY VARIABLE INSTRUMENT ID BENCHBOARD SIG ID rat 3CE 55 SLCS Discharge Pressure X X IC41*PTN004 2 IH13*P601 C41EA003 0-1800 psig 3g u
cp . 56 Condensate Pump Discharge Pressure X ICNM-PT105 3 Illl3*P680 CNMPY0IO 0-800 psig 57 Cumulative Boron Injected X IC41*LTN001 3 IH13*P601 0-5000 gal 30 58 HPCS Key Valve' Positions X 3 IH13*P601 Open/ Closed 18 59 SSW Xey Valve Positions - X 3 !!!!3*P8 70 Open/ Closed 18 60 RCIC Key Valve Positions X 3 IH13*P601 Open/ Closed 18 61 Liquid Effluent Radioactivity X 3 RHR HX Service Water IRMS*BEISA,B IH13*P879 10~ to 10~ C1/cc Cooling Tower Blowdo m -RE108 *P878 Liquid Radwaste Effluent -RE107 62 SSW Temperature to ESF Components X 'ISWP*TT31A,B 2 Ill!3*P870 SVPTY017 0-125 F 018
TABLE 1 CortPOSTTC f.iST Ol' ACCIff!!T t10!!!TORIt!G VARIADf.f:S u SOURCC
!! $ u
@ E 8 !!
O
" n.
de E O h MCR DISPLAY Ll'US VARIABLE e "
It3STRUMDIT ID DENCtf BO AR D SIG ID "*
RAllCC 63 tieteorology X 3 Wind Direction 25 O' - 360
$ Wind Speed 0 - 100 mph Atmospheric Stability Estimate minus 10 - 20,F 64 Accident Crab Sampling X 3 12 Primary Coolant and Sump
. Cross Activity
. Camma Spectrum
. Baron Content
. Chloride Content
. Dissolved Ilydrogen or Total Gas
. Dissolved Oxygen
. pli Containment Air
. !!ydrogen Content
. Oxygen Content
. Camma Spectrum
1 TABLE 1 COMPOSITC LIST Or ACCIDENT HONITORING VARIABLES t
d SOURCC u h $ N E We N S 8 u uu u "- b 'f M o. d h MCR DISPLAY SPD0 VARIABLE E O N INSTRUMENT ID BENCl! BOARD SIG JD RANCC Top to bottom of tank 25 65 Liquid Radwaste Tank Levels X ILWS-LT13A,B,C 3 k' -LT8A B.C.D
- -LT26A,B
-LT521A,B,C,D
-LT320
-LT24A,B 66 Emergency Ventilation. Open/ Closed 23 2 1H13*P863 ffVRBX001 Damper Positions X 004 009 C10 HVWBX001 24 67 Status of Standby Power X 2 IH13*P808
- P877 e
____..-____.__m___E .
. . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ .. _-._ _ _ _ ..._ _. _ . .m . _ .____ . ._ ,__...___..._m ._--__..-m. ._-..,m- _ . - _ _ . .
TABLE 1 Cor4POSITC f.IST or ACCIDENT HONITORING YARIABLES i
f r
l
- 6 d
SOURCE s.a 0 U $$
$ E. O du 8
VARIABLE liI O N " "* '
INSTRUMENT ID BENCHBOARD SIG ID RANCE s
t.o 68 Room Temperatures for Detection of Leakage from Containment Breach X X IE31* TEN 001A,8 3 1H13*P632 50 to 350 F 28
- TEN 004A,B
- TEN 015A,B,C,D
- TEN 017A,B,C,D
,
- TEN 018A,B
- TENO3tA,B,C,D
- TENO34A,B
- TENO37A,8
- TEN 040A,B '
- TEN 043A,B
- TEN 046A,8
-
- TEN 049A,8
- TEN 052A,8
- TEN 055A,B O
6 m
TABLE 1 NOTES GENERAL NOTES
- 1. The design and qualification criteria defined by Regulatory Guide 1.97, Revision 3 for the instrument categories defined therein are used as definitions within this study. For example, if Table 1 of this study identifies a variable as being Category 1 then RBS instrumentation and displays used to measure same either presently meet or will meet the regulatory guide criteria for Category 1 unless otherwise determined by l the DCRDR Summary Report.
- 2. Table 1 lists additional instrumentation for monitoring the subject variables which may not meet all the guidance provided by R.G. 1.97, Revision 3. These instruments are listed because it is anticipated that for a large group of events they will be available for use by control room operators in addition to those instruments qualified for use in accordance with R.G.
1.97, Revision 3.
TECHNICAL NOTES
- 1. The approximate overall height of the RPV measured from the bottom of the vessel to the top of head flange is 842 inches.
The following elevation points are provided for reference:
REFERENCE POINT ELEVATION (inches)
Vessel Bottom 0.0 Top of Active Fuel 358.6 Bottom of Dryer 535.6 Steam Line Nozzle 636.5 l
- 2. RBS utilizes a thermal conductivity type devits for hydrogen level detection. Continuous air samples for each channel are drawn from the containment or drywell areas at the discretion l of main control room operators.
- 3. The normal level of the suppression pool is at approximately.90 feet above MSL. The depth of the pool is approximately 20 feet.
- 4. Control rod position is determined by the actuation of magnetic reed switches mounted at approximately three-inch intervals along the control rod length.
20
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TABLE 1 NOTES
- 5. A digital radiation monitoring system is utilized for monitoring radiation levels throughout the plant. A CRT display is located in the main control room for ready use by plant operators. Class 1E readouts are located on vertical boards for all safety-related instrumentation in addition to the CRT display.
- 6. FSAR Sections 11.5 and 12.3 identify additional plant instrumentation which the main control room operators may use in diagnosing primary and secondary containment effluent releases.
- 7. RBS does not provide for on-line monitoring of primary containment air for oxygen levels. The station design does provide for this variable via other means as shown in Note #12.
- 8. The neutron monitoring system is furnished power via zero time outage motor generators which are used for suppression of transients of the supply voltage. Class 1E backup power is provided for maintenance and unusual events which might render the system inoperative.
- 9. The emergency operating procedures direct the plant operators to determine the cause of an automatic scram. Variables identified with this note allow the operator to ascertain this information. The subject variable is annunciated in 'the main control room. No direct indication of the variable is presented unless otherwise identified by the stipulation of a j display located on a benchboard.
- 10. The implementation of R.G. 1.47 as well as maintenance annunciation (e.g. trouble alarms) are deemed sufficient to l ascertain the operational status of the HPCS DG for plant '
operators.
- 11. The valve position of all containment isolation valves is monitored on various benchboards in the main control room. 1 Additionally, the SPDS database contains this information.
The redundancy requirement for Category 1 is met by system design in that each containment penetration has two isolation valves (inboard and outboard). The position indication for each valve does not in itself meet Category 1 requirements for redundancy.
- 12. Analysis of grab samples obtained from a Post Accident Sampling System (PASS) at RBS is provided on-site by the following methodologies:
21
1
TABLE 1 NOTES '
-s PRIMARY COOLANT AND SUMP 5 2
. Gross Activity NaI Detector -
. Gamma Spectrum HPGe Detector
. Boron Content Plasma Jet Spectrophotcmeter y
. Chloride Content Ion Gas Chromatagraph Dissolved Hydrogen or Total Gas Ion Gas Chromatagraph i Dissolved Oxygen Ion Gas Chromatagraph pH pH Electrode Probe -
CONTAINMENT AIR
. Hydrogen Content Derived by measuring partial
. Oxygen Content pressures at PASS panel --
Gamma Spectrum HPGe Detector '
Particulates Particulate Filter Iodine ]
Silver Zeolite Filter M Noble Gases 15 cc Gas Sample Vial 3 N
q
- 13. The limit switches furnished with each MSIV are not -31 individually labelled. The instrument identification numbers -
shown in Table 1 are those for the MSIV's. '
]j 4
The subject variable constitutes a special case of the G containment isolation valve position variable. ;
a
- 14. Consult Section 12.3 and Table 12.3-1 of the RBS FSAR for a description of plant areas which are monitored for radiation 4 3
levels in excess of nominal values experienced during plant ,
operation.
g Area radiation monitors are used primarily to determine %
x y personnel accessibility into areas containing equipment.
important to plant safety. Where direct radiation sources are k not present the monitors will provide a general indication of k containment integrity.
description of instrumentation for monitoring airborne g g radioactivity releases. =
- 16. RBS does not presently monitor the core teroperature directly as do PWR's owing to basic design differences between PWR's and D BWR's. Further, the state-of-the-art has not made available .i g instrumentation which is capable of reliably measuring core temperatures in a BWR.
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TABLE 1 NOTES Reference 6 has demonstrated the fact that as long as the core of a BWR remains covered with water, core temperatures will not exceed fuel rod limits.
- 17. RBS emergency operating procedures have identificJ the need for a containment water level indicator to be used to ascertain the extent of containment flooding which might occur owing to several scenarios. The plant does not presently provide for _'
dedicated indication of this variable in the main control room except for measurement of suppression pool water level.
- 18. The positions of all key motor-operated valves are provided on the benchboard from which the system is operated.
All key valve positions necessary for system operation are monitored for the SPDS database. .
- 19. The design of RBS Unit 1 utilizes acoustical monitors for position indication. Additionally, the plant has thermocouples installed in the discharge line to establish safety / relief valve (SRV) actuation. However, the slow cooldown rate of the discharge line piping obviates the use of these instruments as a primary means of detecting SRV position.
The acoustical monitors detect two phase as well as single-phase flow for SRV discharges.
- 20. No dedicated, on-line readout is provided for this instrument -
channel. However, the data may be accessed through the process computer and displayed via CRT on the principal plant control console 1H13*P680.
- 21. The state-of-the-ar* does not provide for reliably measuring this variable as an on-line, Category 1 device. The use of radiation monitors for known effluent pathways and periodic v sampling of the primary coolant will suf fice for detection of fuel rod failures. _
References 7 and 8 delineate further the RBS Unit 1 position on :-
monitoring of this variable.
7-
- 22. SRV position, tailpipe thermocouples, and the suppression pool temperature are all monitored at RBS. These parameters will ~ '
give ample indication of SRV discharge line flow when-used in conjunction with RPV pressure indicators.
s
- 23. The position of all safety-related damper / valve positions is indicated on benchboard 1H13*P863 which is dedicated to HVAC control. .
23 '
TABLE 1 NOTES '-
a=
Selected damper and valve positions are monitored by the SPDS _2 database to insure primary and secondary boundaries for _[
redioactive releases remain intact. c_,-
- 24. The status of the electrical distribution system is monitored on vertical board 1H13*P808 and by the SPDS database to a point ;4 of resolution that the SPDS is capable of generating RBS one- i=
line diagrams.
Other safety-related sources of motive power such as air __
pressure are monitored and annunciated in the main control room ,
also. Z2 The following is a partial listing of instruments furnished at RBS Unit 1 for monitoring the subject variable: __,
Description Instrument ID (1) Air Supply ILSV*PT9A,B i~
==_
(2) Standby DG Voltage V-1EGSA07 ._
B07 ---
1E22*VR611 (3) Standby DG Current A-1EGSA07 t-B07 1E22*AR607 (4) Standby DG Power W-1EGSA07 P-B07 s
1E22*WR609 _
(5) Standby DG VAR-1EGSA07 Reactive Power
"[?
B07 -
1E22*VARR608 (6) Standby DG F-1EGSA07 --
Frequency B07 5 1E22*FREQR612 ___
m (7) Standby 4.16kV V-1EGSA08 .-
Bus Voltage B08 e 1E22*VR610 --
(8) Standby 4.16kV A-1ENSA07 Incoming Breaker '5_
807 Current IE22*AR619 '-F
-M-24 -
TABLE 1 NOTES Description Instrument ID (9) Standby Switchgear V-1ENBA03 125 VDC Bus Voltage B03 1E22*VR618
- 25. The present design of River Bend Station does not provide for direct or indirect readout of the subject variable in the main control room.
- 26. The instrument readout is a timed, strip chart, multi pen recorder. Operators ascertain the cooldown rate visually by observing chart slopes. The plant is also provided with a permanent record for later analysis as required.
- 27. RBS will provide for grab sampling of the suppression pool water inventory. On-site chemistry facilities are equipped to measure the subject variables in a timely fashion commensurate with the needs of MCR operators.
- 28. The following plant areas are monitored for the subject variable:
Main Steam Line Pipe Tunnel RHR Equipment Areas RCIC Equipment Areas RWCU Equipment Areas
- 29. Heatup calculations performed for RBS Unit 1 predict a maximum drywell temperature not to exceed 320 F. Therefore, the existing instrumentation has sufficient range to function as an information source during accident or abnormal conditions.
The range specified is that of the sending instrument. The actual recording range will be determined at a later time prior to fuel load.
- 30. RBS E0P's instruct the operator to ascertain this variable by measuring the SLCS tank level drop thus inferring the quantity of sodium pentaborate solution which is injected into the vessel.
- 31. RBS does not measure SLCS flow directly owing to sensing instrument problems arising from sodium pentaborate contamination of moving parts. Measurement of SLCS pump discharge header pressure is deemed ~ sufficient to ascertain SLCS flow when used in conjunction with other variables such as SLCS tank level.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS l
1 SEQUENCE NO: 1 VARIABLE: Reactor Vessel Water Level FINDING: The present design of RBS Unit I does not provide Category 1 instrumentation over the fuel zone range. Further, the variable and reference leg instrument lines inside the drywell are subject to flashing and boiling which could lead to erroneous water level readings.
TECHNICAL EVALUATION i
' Reference No. 6 of this study explains the effects of flashing and boiling on water level instrument lines inside the drywell. The NRC Staff has reviewed this reference and concurs with its findings.
Inadequate core cooling concerns dictate that the fuel zone range water level instrumentation be Category 1.
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REFERENCES:
RBS FSAR Section 7.5.1.1.2 Appendix 1A, Item II.F.2 RBS Response to NRC Question 421.014 l DESIGN CONSIDERATIONS River Bend has presently undertaken the following design revisions which shall be complete prior to fuel l load:
(1) Upgrade the fuel zone instrumentation to Category 1.
l (2) Furnish a high drywell temperature alarm in the main control room.
(3) Reduce the reference leg vertical drop inside the drywell to a minimum.
( (4) Relocate restriction orifices as close to the drywell penetration as possible.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS (5) Train operators to recognize possible anomolies in water level readings.
(6) Revise the SPDS database as required to insure that all Category 1 instrument information is monitored, analyzed and displayed.
Consult the DCRDR Summary Report for specific design revisions to RBS Unit I and an implementation schedule for same.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS
. SEQUENCE NO: 6 !
VARIABLE: Suppression Pool Temperature FINDING: The plant design does not provide for a Category I display of the bulk pool temperature on either 1H13*P808 or 1H13*P601.
TECHNICAL EVALUATION The controls for the safety / relief valves are located on 1H13*P601. The operator must base decisions on whether to manually operate on discrete data only available at 1H13*P808. This condition does not appear to be desirable from a human factors viewpoint because (1) the bulk pool temperature must be computed by g hand, and (2) the vertical board which contains this information is not located in close proximity to where the information is needed.
REFERENCE:
NUREG-0783 " Suppression Pool Temperature Limits for BWR Containments" DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for the RBS resolution of the subject finding. ,
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS ,
SEQUENCE NO: 9 VARIABLE: Condensate Storage Tank (CST) Water Level FINDING: A direct readout display for this variable is not provided in the main control room except through use of the process computer and SPDS databases and display CRT's.
~
TECHNICAL EVALUATION The majority of RBS instrumentation providing this type of information is provided with direct reading displays in the main control room. It is expected that a seismic disturbance would render the process and SPDS computers out of service thus temporarily precluding information from this variable.
e e
DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for a discussion of the DCRDR review and resolution of this finding.
l TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 10 VARIABLE: SLCS Tank Level FINDING: The present instrumentation furnished to RBS does not meet the equipment qualification guidelines set forth by R.G. 1.97, Revision 3.
TECHNICAL EVALUATION The operator will require the subject instrumentation infrequently if ever. However, the information derived from this instrument will allow the operator to ascertain proper system function. The need for this instrument to meet Category 2 criteria is highly dependent upon the resolution of the neutron monitoring system (NMS) finding. If RBS upgrades its NMS to meet Category I criteria then this variable o need not confora to Category 2 criteria but may remain as purchased without modification insofar as accident monitoring considerations are concerned. .
DESIGN CONSIDERATIONS The DCRDR shall evaluate this finding in light of other control room improvements and make a final determination which will be documented in the DCRDR Summary Report.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 12 VARIABLE: Primary Containment Radiation FINDING: Radiation level data is almost solely obtained by the digital radiation monitoring system (DRMS) and stored in its database. However, none of this data is presently available for use by the SPDS database.
l TECHNICAL EVALUATION The DRMS is a mini-computer based system with CRT and direct readouts in the main control room (MCR).
All safety-related instruments have dedicated as well as CRT readouts in the main control room. The w close proximity of the DRMS and SPDS displays in the MCR, TSC, and EOF enhances the operators use of both databases for information retrieval during those events when both might be required. This fact and the difficulty of having the DRMS database communicate with the SPDS database has led RBS to leave both systems functionally isolated insofar as information retrieval is concerned.
This discussion is typical for other variables which are radiation level related.
DESIGN CONSIDERATIONS RBS deems the existing plant design to be in compliance with the intent of NUREG-0696 and related documents insofar as not having radiation level information in the SPDS database.
l i TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS l
SEQUENCE NO: 14 VARIABLE: Radiation Level in Circulating Primary Coolant FINDING: River Bend Station does not presently incorporate in its design any instrumentation or displays to monitor this variable on-line in a real-time mode.
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TECHNICAL EVALUATION The usefulness of the information obtained by monitoring the radioactivity concentration or radiation level in the circulating primary coolant, in terms of helping the operator in his efforts to prevent and mitigate accidents, has not been substantiated. The critical actions that must be taken to prevent and a mitigate a gross breach of fuel cladding are (1) shut down the reactor and (2) maintain water level.
Monitoring the subject variable will have no influence on either of these actions. The purpose of this monitor falls in the category of (1) information that the barriers to release of radioactive material are being challenged and (2) identification of degraded conditions and their magnitude, so the operator can take actions that are available to mitigate the consequences. Additional operator actions to mitigate the consequences of fuel barriers being challenged, other than those based on R.G. 1.97 Type A and B variables, have not been identified.
Regulatory Guide 1.97, Revision 3 specifies measurement of the radioactivity of the circulating primary coolant as the key variable in monitoring fuel cladding status during isolation of the NSSS. The words
" circulating primary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows past the core. A basic criterion for a valid measurement of the specified variable is that the coolant being monitored is in active contact with the fuel (i.e. flowing past the failed fuel). Monitoring the active coolant or a sample thereof is the dominant consideration. The RBS post-accident sampling system (PASS) provides a representative sample which can be monitored.
The subject of concern in the RG 1.97, Revision 3 position is assumed to be an isolated NSSS that is shutdown. This assumption is justified as current monitors in the condenser off gas and main steam lines provide reliable and accurate information on the status of fuel cladding when the plant is not isolated.
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l TABLE 2 l
l ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS l
l Further, the PASS will provide an accurate status of coolant radioactivity, and hence cladding status, once the PASS is activated. In the interim between NSSS isolation and operation of the PASS, monitoring of the primary containment radiation and containment hydrogen will provide information on the status of the fuel cladding.
DESIGN CONSIDERATIONS RBS proposes no design changes at the present time based on the above technical evaluation.
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TABLE 2 l ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS 1
SEQUENCE NO: I6 VARIABLE: Containment /Drywell Oxygen Concentration FINDING: The design of RBS does not provide for instrumentation to continuously monitor the containment and drywell for oxygen concentration.
TECHNICAL EVALUATION The containment atmosphere of RBS will not be inerted. Therefore, the percent volume of oxygen prior to any plant transient will be known. The large volume of the containment air mass insures that the volumetric rate of change of oxygen is sufficiently slow to permit testing of grab samples to determine the amount of oxygen inside the containment. This variable is not considered to be dominant to measuring hydrogen levels and as a consequence it serves no immediate information needs to the plant operators for g mitigation of abnormal events.
The oxygen concentration inside the drywell is also a known quantity pursuant to the above. Therefore, the oxygen deflagration overpressure limit is already exceeded during the initial stages of a hydrogen-generating event and the hydrogen concentration becomes the controlling factor to be measured to ascertain the approach to and the exceeding of the hydrogen deflagration overpressure limit.
DESIGN CONSIDERATIONS No design revisions are contemplated by RBS at the present time.
i TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS l SEQUENCE NO: 17 VARIABLE: Drywell Drain Sump Water Level l
FINDING: RBS has furnished Category 2 instrumentation to monitor this variable. However, Regulatory Guide 1.97 specifies the use of Category I instruments.
TECHNICAL EVALUATION The RBS Mark III containment has two drywell drain sumps. One drain is the equipment drain sump which collects identified leakage. The other is the floor drain sump which collects unidentified leakage.
, o Although the level of the drain sumps can be a direct indication of breach of the reactor coolant system l Pressure boundary, the indication is not unambiguous, because there is water in those sumps during normal i
l operation. Other RBS instrumentation that indicates leakage in the drywell is:
l (1) Drywell pressure (2) Drywell temperature (3) Primary containment atmosphere radiation level I
(4) Drywell radiation level l
While the drywell sump level signal does not automatically initiate safety-related systems, the plant i operators will take manual actions based on this variable in accordance with Section 3/4.4.3 of the RBS i
Technical Specifications. The surveillance requirements of the Technical Specifications are sufficiently stringent to allow the use of a Category 2 device for this application.
Regulatory Guide 1.97, Revision 3 requires instrumentation to function during and after an accident. The drywell sump systems are deliberately isolated at the primary containment penetration upon receipt of a I
l TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS LOCA signal to establish containment integrity. This fact renders the drywell sump level signal irrelevant. Therefore, by design, this instrumentation serves no useful accident monitoring function.
The RBS emergency operating procedures use RPV level and drywell pressure as entry conditions for the ;
level control. A small line break will cause the drywell pressure to increase before a noticeable increase in the sump level. Therefore, the drywell sumps will provide a lagging versus early indication of a leak.
1 l DESIGN CONSIDERATIONS i
l l RBS considers its Category 2 instrumentation to be adequate.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 18 VARIABLE: Neutron Flux FINDING: Regulatory Guide 1.97, Revision 3 identifies neutron flux as a key variable and classifies it as Category 1. RBS does not furnish the source range (SRM) and intermediate range (IRM) neutron detector drive positioning modules with Class IE power. Further, while the backup power supply for the neutron monitoring system is Class IE, the power distribution breaker panels are not.
TECHNICAL EVALUATION The need to measure neutron flux in an accurate and reliable fashion is necessary for at least one or w
" more accident sequences. The plant design provides for Class IE or equivalent equipment except as stated above. The large number of detectors in the core, or capable of being inserted into the core, insures that random single failures will not render the system inoperable should it be required to verify shutdown or an approach to criticality. However, common mode failures (i.e. detector drive positioning modules) could render the SRM and IRM systems inoperable for certain design basis events.
DESIGN CONSIDERATIONS Consult the DCRDR Summaary Report for a discussion of this finding and the RBS resolution of same.
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ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS l SEQtENCE NO: 19 VARIABLE: Main Steam Line Radiation FINDING: The EOP's require the operator to identify the initiating event for a reactor scram. Several variables including the subject variable will produce a scram of the reactor protection system
( (RPS). The SPDS database does not contain the status information provided by this variable.
TECIDfICAL EVALUATION All of the variables capable of initiating a reactor scram are monitored by the SPDS with the exception of the subject variable and scram discharge volume level (Sequence No. 42). Since the SPDS will be used g by other than control room personnel a binary status indication of these variables within the SPDS i database would be of use to TSC and EOF personnel.
l l DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for a determination in regards to this finding.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 20 VARIABLE: Status of HPCS DG FINDING: The variable while displayed in the control room is not contained within the SPDS database.
TECHNICAL EVALUATION This variable was identified via the operator action event trees of Appendix A to allow the operator to diagnose the reason for failure of the NPCS. However, other variables such as the bus supply voltage will allow persoael to make a determination of the failure mode if they are relying upon the SPDS for j information.
l DESIGN CONSIDERATIONS RBS believes the SPDS database to be sufficiently complete so as not to require monitoring of this variable. No design revisions are anticipated as a result of this finding.
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 33 VARIABLE: Containment k'ater Level FINDING: The design of RBS does not provide for main control room instrumentation which covers the range ;
required by this variable.
)
TECHNICAL EVALUATION The suppression pool level monitoring instrumentation will be sufficient for most all of the operators
- formation needs in measuring this variable. However, this instrumentation does not measure beyond
.eproximately the 94 ft. elevation. Measurement of this variable will require an upper range of approximately 120 ft. elevation. Consequently, local pressure indicators IE12*PIR002A,B,C are considered to be adequate for measuring this variable over its entire range. These pressure indicators provide NPSH readings for RHR Pumps A, B, and C. If suction is taken from the suppression pool, static suction head readings will provide an acceptable indication of the containment water level by correlation.
DESIGN CONSIDERATIONS RBS anticipates no further design revisions to implement this variable. Consult the DCRDR Summary Report for further discussion of this finding.
TAELE 2 i
3 ACCIDENT MONITORING INSTRINENTATION STUDY FINDINGS 1 I
i SEQtENCE NO: 35 i VARIABLE: Primary Containment Annulus Differential Pressure l FIMING: Direct readout of this variable is not available in the main control room or the SPDS database.
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TECMICAI. EVALUATION i The RBS emergency operating procedures provide for depressurizing the containment structure into the l aamulus region to maintain containment integrity. Therefore, the integrity of the shield wall could be challessed unless the main control room operators are provided with the subject information so the rate ;
of depressurization may be controlled.
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DESIGN CONSIDERATIONS Consalt the DCRDR Summary Report for a discussion of this finding and resolution of same pursuant to the !
DCEDR review. ;
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TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQtENCE NO: 37 VARIARE: Reactor Cooldown Rate FINDING: RBS has performed calculations to allow for an accelerated RPV cooldown rate of 200 F/hr during accident conditions. The instrumentation, displays, and power supplies for this variable are all non-Class 1E. Therefore, the subject instrumentation cannot be expected to remain operational under accident conditions. Also, the SPDS database does not monitor all of the RPV temperatures available thus precluding any decisions involving reactor cooldown rate based on information from the SPDS.
TECENICAL EVALUATION g The operator must be provided with accurate information as to the cooldown rate duri.lg accident conditions to insure that thermal stresses on the RPV do not create a reactor pressure coolant boundary failure.
DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for a detailed review and discussion of this finding to include the RBS l
resolution.
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. - - . . .. ~- - , ~ . - - - . . _ m-______-._
C TABLE 2 ACCIDENT MONITORING INSTRUMEhTATION STUDY FINDINGS SEQUENCE NO: 41 VARIABLE: MSIV Leakage Control System Pressure FINDING: This variable is identified by R.G.1.97, Revision 3 but does not appear in the SPDS database pursuant to guidance provided by NUREG-0696.
TECHNICAL EVALUATION The emergency operating procedures do not make direct use of this variable. Consequently, this data has not been introduced into the SPDS database for this reason.
u Personnel performing off-site dose assessments may require this information to insure that MSIV leakage is properly taken into account for dose calculations. This information is available via communications with the MCR operators in the unlikely event it is ever needed.
DESIGN CONSIDERATIONS There exist no immediate plans for RBS to monitor the subject variable for the SPDS. The DCRDR shall review the subject finding and a final resolution of same will be given in the DCRDR Summary Report.
TABLE 2 ACCIDENT MONITORING INSTRtMENTATION STUDY FINDINGS SEQENCE NO: 52 VARIABLE: Main Steam Line Flow FINDING: The subject variable is not monitored by the SPDS database.
TECNNICAL EVALUATION This variable is monitored for diagnostic purposes to identify the initiating event. Also, .t is used to indicate the availability of a restored power conversion system. The lack of this data in the SPDS database serves to limit its usefulness by not furnishing SPDS users with a database that is necessary and sufficient.
N DESIGN CONSIDERATIONS The DCRDR shall review the need for this variable in the SPDS database and document its conclusions and
, rec - adations in the DCRDR Summary Report.
I THE 2 l
! ACCIDENT MONITORING INSTRIRfENTATION STUDY FINDINGS l
1 i
l SEQUENCE NO: 53 VARIABIE: Circulating Water Flow FINDING: The SPDS database does not contain information from this variable.
The present design of RBS provides for this variable to be displayed via the process computer. The process computer, however, does not display a flow value but rather a pressure value.
TECENICAL EVALUATION l Consult the discussion for the study findings for main steam line flow (Sequence No. 52).
Busan factor principles would seem to dictate that the process computer software be revised to accommodate a flow value algorithm so that a flow value may be displayed for operator use.
DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for resolution of the subject finding.
TABLE 2 ACCIDENT MONITORING INSTRIRIENTATION STUDY FINDINGS SEQtENCE NO: 63 VARIABLE: Meteorology i FINDING: The present design of RBS Unit I does not display meteorological data in the main control room or !
furnish same to the SPDS database.
TECHNICAL EVALUATION The need for this information during the initial stages of an accident event is evident. The lack of :
design is due to uncertainties in regards to interfacing existing equipment with the main control room. '
e CD DESIGN CONSIDERATIONS Consult the DCRDR Summary Report for the RBS resolution of this finding. !
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I APPENDIX A A!TTICIPATED TRANSIENT EVENT TREE ANTICIPATED REACTOR S/R VALVE S/R VALVE TEEDWATER RCIC OR LP RifR & SSW SEQUENCE TRANSIENT SUBCRITICAL OPEN RICLOSE (T)
HPCS ECCS CR PCS (C) (M) (P) (Q) (U)
I (V) (W)
W -
T Q _
SUCCESS P N N M U
7 Nw Q W if g Y
> TAILCRE o b _. '
V --
TQW
- Q TP W
TPW
[ M U
C _
W 7pqv E w v w T U E TPQW V
TPQUV T
TIf
'C
' TC
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l TABLE 2 ACCIDENT MONITORING INSTRUMENTATION STUDY FINDINGS SEQUENCE NO: 65 VARIABE: Liquid Radwaste Tank Levels FINDING: Level indication either direct or indirect for liquid radwaste tanks is not furnished in the main '
control room or the SPDS database.
TECHNICAL EVALUATION Monitoring of this variable is directly attributable to specific conditions which occurred during the :
TMI-2 accident. Several radwaste tanks overflowed as a result of excessive accumulation of water from the containment building. The design of RBS Unit I precludes a similar condition from existing because e containment isolation valves will not automatically reopen once they are closed by an isolation signal.
Operator action is required to reset and reopen the subject valves. Therefore, RBS does not monitor this variable in the main control room. This variable is monitored locally should the need arise for this information.
i DESIGN CONSIDERATIONS RBS envisions no revisions to the plant design but will review the subject finding further as a part of the DCRDR. Consult the DCRDR Summary Report for the RBS resolution of the subject finding.
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_ _ . -. , ,./ - . . .
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n TABLE 2 ACCIDENT MONITORING INSTRtRENTATION STUDY FINDINGS
- rP m NO: 68 VARIABII: Room Temperatures for Detection of Leakage from Containment Breach FIEING: The SPDS database does not monitor this variable.
TEC WICAL EVALUATION It is not obvious what use SPDS viewers in the TSC and EOF would make of the subject information. The hemefit/ cost ratio for introducing the subject data into the SPDS database is low. The control room indications appear to address the plant needs.
r ao DESIGN CONSIDERATIONS RBS does not anticipate any design revisions to monitor this variable for the SPDS. However, final resolution of the finding will be made .via the DCRDR Summary Report.
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4 REFERENCES * ~
- 1. NUREG/CR-2100, " Boiling Water Reactor Status Monitoring During Accident Conditions", EG&G Idaho, April 1981. . k :f JE'J
- 2. NUREG/CR-1440, " Light Water Reactor Status Monitoring During Accident Conditions", EG&G Idaho, June 1980. ,'
- 3. WASH-1400, " Reactor Safety Study: An Assessment of Accident
'i Risks in U.S. Commercial Nuclear Power Plants", U.S. Nuclear lis' ;
Regulatory Commission, October 1975.
9
- 4. Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs' ~^
Conditions During and Following an Accident", U.S. Nuclear Regulatory Commission, May 1983.
- 5. NSAC/21, " Fundamental Safety Parameter Set for Boiling Water Reactors", December 1980.
1
- 6. SLI-8218, " Inadequate Core Cooling Detection in Boiling Water Reactors", Sol Levy Incorporated, November 1982. J.j . ;
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,Y
- 7. "BWR Owners' Group Position on NRC Regulatory Guide 1.97 Revision 2", BWR Owners' Group, July 1982.
- 8. .'.' Evaluations of Certain Instrumentation' Requirements Specified in NRC Regulatory Guide 1.97,-Revision 2", BWR Owners' Group, i February 1982.
J'
- 9. " Emergency Procedure Guidelines", BWR Owners' Group, November 1983.
- 10. " River Bend Station Final Safety Analysis Report", Gulf States Utilities Company, January 1984 (Amendment 11).
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APPENDIX A OPERATOR ACTION EVENT TREE
- ANTICIPATED TRANSIENT FOLLOWED BY LOSS Or DECAY HEAT REMOVAL (TW)
INITIATING OPERATOR VERITIES TRANSIENT OPERATOR SECURES SCRAM, MSIV CLOSURE, OPERATOR DETECTS RifR FAILURE DIAG-PCS FAILURE EVENT-LOSS TW TRIP, RCIC AND HPCS AND MANUALLY TAILURE OF RilR NOSED, REPAIRED CONTROLS RCIC DIAGNOSED, RE-Cr CollDENSER t(PCS INITIATION, AND AND OPERATOR RE-VACUUH PAIRED.'AND SATLTY/ RELIEF VALVE STORES USE OF RHR OPERATOR RE-ACTUATION STORES USE OF PCS SAFE SUCCESS SHUTDOWN d k SATE SHUTDOWN
/a II TAILURE BEGRADATION Or CORE DECRADATION or CORE 5-;
. SUCCESS ASSUMED-
. NOTE:
- Loss of condenser vacuum is a typical
.y Initiating event which was chosen SUCCESS owing to its high frequency of ASSUMED occurrence.
l
APPENDIX A OPERATOR ACTION EVENT TREE ANTICIPATED TRANSIENT COMBINED WITH FAILURE OF HPCS, RCIC, AND LP ECCS (TQUV)
LOSS OF REACTOR SCRAM, MSIV. HPCS RESTORE TI W MANUAL LP RESTORE RESTORE RHR/SSW TEEDWATER CLOSURI OR AT HIGH PRES _SURE DEPPIS- ECCS LP ECCS TW . OR PCS S/R VALVE OPERATION RCIC (HFCS, RCIC, TW) SURIZATION SAFE SHUTDOWN SUCCESS d b IAILURE CORE ASSUMED DEGRADATION E
1I FAILURE
> CORE h DEGRADATION SAFC SHWMW CORE DEGRADATION SAFE SHUTDOWN
~'
8 D ,RADATION D
CORE DEGRADATION CORE DEGRADATION e
l.
. 6 e
APPENDIX A OPERATOR ACTION EVENT TREE
, A!!TICIPATED TRANSIENT COMBINED WITH FAILUPE OF AUTOMATIC SHUTDOWN SYSTEMS (TC)
TRANSIENT RAPID OPERATOR ENSURES OPERATOR MAKES OPERATOR ENSURES INITIATING AUTOMATIC RAPID RPT C REACTOR SUB- LONG TERM HEAT EVENT SHUTDOWN INITIATION Of* HPCS CRITICAL 6 AVOIDS TRANS!'t.R TO CONTAINMENT OVER- ENVIRONMENT PRESSURE n
CORE 3 DEGRADATION hv SIECESS J L CORE DEGRADATION lI
, TAILURE CORE DEGPADATION 9
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APPDIDIX A OPERATOR ACTION EVENT TREE SMALLLOCAWITH'TAILURE0[ECCS I SMALL ELECTRIC POWER HICH , MANUAL ADS DEPRESSURIZE LPCS/ CONDENSATE LONG BREAK' .RPS PRESSURE CONTROL OF TW VIA MAIN STEAM LPCI TERM SYSTEM.
(Sg ) UNTMT CONTROL CORE SPRAT RESTORE HPCS HPCS, RCIC, OR RESTORE COOLING RWCU LPCI/LPCS SAFE S!!UTDOWN EVENTUAL CONTAINMENT FAILURE LEADING TO CORE DEGRADATION TAILUREA,SSUMED f FAILUREgSUMED ee SATE SHUTDOWN EVENTUAL CONTAINMENT TAILURE I.EADING TO CORE'DECRADATION
- CORE DEGRA9ATION ca SUCCESS ASSUf1ED SAFE SHUTDOWN SUCCESS d EVENTUAL CORE DECRADATION "r:
7 SUCCESS ASSU!IED TAILbRE CORE DECRADATION 1 .
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m APPENDIX A
SUMMARY
OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS BWR Hajor Purpose for Indicated Accident Sequence Pleasured Variable TW TQUV SE AE TC 3
CONTROL ROD POSITIONS Verification Verification Verification Verification Indication of of scram of scram of scram of scram failure of au-tomatic scram and success /
failure of manual inser-tion of rods NEUTRON FLUX Verification Verification Verification of scram of scram Verification Indication of May be unreliable under severe voiding of scram of scram failure to conditions.
scram and de-
), termination of effect of man-
[') ual shutdown actions VESSEL MATER LEVEL Indication of Indication of Indication of Indication of Indication of initiating initiating initiating initiating initiating event event event event event Indication of Indication of Indication of Indication of Indication of itPCS, RCIC, !!PCS, RCIC, !!PCS, RCIC, IECI, LPCS etfectiveness
~
LPCI, LPCS, LPCI, LPCS, ITCI, ITCS, operation of IIPCS & RCIC FW, or Con- FW, or Con- FW, or Con-densate Pump densate Pump densate Pump operation operation operatic..
REACTOR PRESSURE Indication of Indication of Indication of Indication of Indication of auto responses auto responses auto responses initiating auto responses to initiating to initiating to initiating event to initiating event event event event l
~
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APPENDIX A
SUMMARY
OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS 8'JR M.ijor Purpose for Indicated Accident Sequence Measured Variable W S TC COMMENTS REACTOR PRESSURE (cont) Indication of Indication of Indication of Guidance for Indication of S/R valve S/R valve S/R valve switchover to failure to operation, operation operation shutdown cool- auto scram ing Cuidance during Guidance durinC Cuidance during Indication of depressurizatiot depressurizatto o depressurization RPT process process process Indication of Indacation of Indication of Indication of success / failure ability to ability to ability to' of manual use condensate use condensate use condensate shutdown 3, pumps pump pumps e
>' Guidance for
' Indicationo'f Indication of Indication of switchover to restored PCS, restored PCS, restored PCS, , shutdown cool-RHR RHR RHR ing Guidance for Cuidance for Guidance for switchover switchover switchover for shutdown for shutdown for shutdown cooling cooling cooling CORE TEMPERATURE Indication of Indication of ' Indication of Indication of Indication of Other parameters such as radiation approach to approach to approach to approach to core damage approach to level and hydrogen concentration core damage core damage core damage core damage or exceeding or exceeding could be used to indicate core damage or exceeding or exceeding or exceeding has occurred.
fuel tempera- fuel tempera- fuel tempera- fuel tempera- fuel tempera-ture limits ture limits River Bend Station monitors tura limits ture limits ture limits reactor water level as an indication of approach to core damage.
I l
l . _ _ _ . _ . . . - _ . . . . . . . . . . . - .
_ -m APPENDIX A
SUMMARY
OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS BWR Ksjor Purpose for Indicated Accident Sequence Measured Variable "
TW TQUV SE g
AE TC RECIRCULATION LINE - - - -
Indication of Not included in Reg. Cuide 1.97 FLOW auto or manual recirc. pump trip PRIMARY CONTAINMENT Indication of Indication of Indication of Indication of Indication of Containment Temperature and Humidity (DRYWELL) PRESSURE effect of RHR effect of RHR initiating initiating effect of RHR can also be used to indicate LOCA.
contoinment containment event event containment cooling and cooling and cooling and maintenance maintenance maintenance of containment of containment of containment
$" integrity integrity integrity P
"O PRIMARY CONTAINMENT Indication of Indication of Indication of Indication of Indication of Provides a backup indication of a RADIATION LEVEL core damage core damage core damage core damage core damage LOCA inside containment.
SUPPRESSION POOL Indication of Indication of Guidance in Indication of Indication of TEMPERATURE system re- system re- depressuriza- effectiveness auto responses sponses to sponses to tion process of RHR cooling to initiating '
initiating initiating event event event Indication of -
RHR pool cool- Indication of Guidance in Cuidance in ing effective- failure toe depressuriza- depressuriza- ness scram tion process tion process Indication of Indication of Indiention of -
RHR failure RilR failure effectiveness of Ri!R pool cooling F
r n
APPENDIX A
SUMMARY
OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS BWR Major Purpose for Indicated Accident Sequence Measured Variable TW SE TQUV AE COMMENTS g TC SUPPRESSION POOL LEVEL Indication of Indication of Indication of auto responses auto responses auto responses Indication of to initiating to initiating auto responses to initiating event event to initiating event event Guidance for Cuidance for Cuidance for switchover of switchover of Cuidance for HPCS & RCIC switchover of switchover of HPCS & RCIC HPCS & RCIC
'from conden- from conden- from conden-HPCS & RCIC sate storage sate storage from conden-sate storage
), tank tank tank sate storage a tank
' ha S/R VALVE POSITION Indication of
'# Indication of Indication of auto responses auto responses ADS failure Indication of to inittacing to initiating auto response event event to initiator Cuidance in depressuriza-Guidance in Cuidance in tion process Indication of depressuriza- depressuriza- failure to tion process tion process scram SRV DISCHARCE LINE FLOW Indication of Indication of Indication of auto responses auto responses ADS failure Indication of SRV position, ta11 pipe thermocouples,.
to initiating auto response event to initiating' to initiator and the suppression pool temperature
' event Cuidance.in . Indication of are all monitored at River Bend depressuriza- Station. These parameters will give failure to tion process , scram ample indication of the SRV discharge line flow.
Guidance in Guidance in depressuriza - depressuriza-tion process - tion process
_ _ . . _ . _ _ . . = _ _ . _ _ _ _ . _ . _ _ _ . _
APPENDIX A
, SUtttARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS Hajor Purpose for Indicated Accident Sequence Measured Variable TW TQUV SE AE TC 3
MAIN STEAM LIisE FLOW - Indication of Indication of Indication of Indication of Indication of Not included in Reg. Cuide 1.97.
auto response initiating initiating initiating initiating to initiating event event event event event Indication of Indication of Indication of depressurizatfor availability.
availability through m. sin of FW of restored steam line PCS to condenser Cuidance in depressurization through PCS.
j, s POSITION OF CONTAINMENT -
Guidance for Indication of Indication of -
in ISOLATION VALVES ' depressurization auto responso auto response through' tu initiating to initinting aux!Itary event event systems TURBINE BYPASS VALVES Indication of Indication of Indication of - -
Not included in Reg. Cuide 1.97.
POSITION initiating availability , availability event. of PCS of PCS Indication of- Indication of- Indication of availability depressurizatio depressurization of restored to main- to main PCS condenser condenser CONDENSER lt0TWELL. Indication of . Indication of Indication of - -
Not included in Reg. Cuide 1.97.
LEVEL availability- availability availability of repaired of FW for. of condensate PCS~ makeup' pumps for makeup J
__w
_e APPENDIX A, SUF2tARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS H.njor Purpose for Indicated Accident Sequence Measured Variable TV TQUV SE g
AE TC FEEDWATER FLOW OR PUMP Indication of Indication of Indication of Indication of Indication of DISCliARGE PRESSURE auto response initiating FW availability auto response initiating to initiating event for makeup after to initiating event event initiating event Indication of event Indication of FW makeup PCS restoration FEEDWATER CONTROLLER Indication of Indication of Indication of Indication of Indication of Not included in Reg. Cuide 1.97.
POSITION auto response initiating FW availability auto response initiating to initiating event for makeup after to initiating event If event initiating event ka Indication of ev:nt 4#
Indication of FW makeup PCS restoration POWER AVAILABLE TO Indication of Indication of Indication of Indication of Indication of Not included in Reg. Cuide 1.97.
FEEDWATER PUMPS auto response - initiating FW availability auto response initiating to initiating event for makeup aftet to initiating event event initiating event Indication of event Indication of FW makeup PCS restoration MSIV's POSITION Indication of Indication of Indication of Indication of Indication of Not included in Reg. Cuide 1.97.
auto response auto response initiating initiating initiating to initiating to initiating event event event event event Indication of Indication of Indication of FW availability availability ability to de-of restored PCS pressurire through con-denser t
APPENDIX A SUtetARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS IS'R Hajor Purpose for Indicated Accident Sequence Hessured Variable TW TQUV SE y
AE TC CONDENSER PRESSURE . Indication of -
Guidance in - -
Not included in Reg. Guide 1.97.
initiating establishing event PCS as heat sink Indication of availability of repaired PCS CONDENSATE PUMP Indication of Indication of Indication of - -
Not included in Reg. Cuide 1.97.
DISCHARCE PRESSURE. ava11ab!11ty availability availability of condensate of condensate of condensate
. :p pumps for pumps for pumps for 4- makeup makeup makeup tn CONDENSATE STORACE Indication of Indication of Indication of -
Indication of TANK LEVEL available available . available available water supply water supply water supply water supply for HPCS and for HPCS, RCIC. for HPCS, RCIC, for HPCS and RCIC and condenser and condenser RCIC hotwell for hotwell for Cuidance in FW makeup IN makeup Cuidance in switchover switchover to
-to suppres- Guidance in Cuidance in suppression sion pool switchover to switchover to pool suction suction for suppression suppression for RCIC, RCIC, HPCS pool suction pool suction HPCS for RCIC and for RCIC and HPCS HPCS
l APPENDIX A SUttLARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS Hajor Purpose for Indicated Accident Sequence tieasured Variable TV SE AE ~ TC TQUV I STEAM JET AIR EJECT 9k Indication and Indication and - - - Not included in Reg. Cuide 1.97.
LINE FLOW diagnosis of diagnosis of initiating initiating event event Indication of Indication of availability availability of repaired of repaired PCS PCS CIRCULATING WATER Diagnosis of Indication of Indication of - Indication of Not included in Reg. Guide 1.97.
SYSTEM FLOW OR PUMP initiating availability availability availability 7
H
- DISC!!ARCE PRESSURE event of condenser of condenser of condenser as heat sink as heat sink as heat sink Determination of availability of repaired PCS HPCS FLOW OR PUMP Indication of Indication of Indication of -
Indication of
. DISC 1 FARCE PRESSURE availability availability availability availability of system to of system to of system to of system to maintain level maintain level maintain level maintain level llPCS KEY VALVE Indication of Indication of Indication of -
Indication of Not included in Reg. Cuide 1.97.
POSITIONS safety system safety system :afety system safety system status or status or status or status or diagno: sis of diagnosin of diagnosis of diagnosis of system failure system failure system failure system failure A
APPEt! DIX A SL7CtARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS BWR Hajor Purpose for Indicated Accident Sequence
!*easured Variable TW TC TQUV S,E AE HPCS DILSEL OPERA- Diagnosis of Diagnosis of Diagnosis of . Diagnosis of Not included in Reg. Cuide 1.97.
TIONAL STATUS system failure system failure system failure system failure or guidance in or guidance in or guidance in or guidance in system system system system operation operation operation operation RCIC FLOW OR PUMP Indication of Indication of Indication of -
Indication of DISCilARCE PRESSURE availability availability availability availability of system to of system to of system to of system to maintain Icvel maintain level maintain level maintain level RCIC KEY VALVE Indication of Indication of Indication of -
Indication of Not included in Reg. Cuide 1.97, f
ho POSITIONS safety system safety systen safety system safety system status or status or status or status or diagnosis of diagnosis of diagnosis of diagnosis of system failure system failure system failure system failure STEAM FLOW TO RCIC Diagnosis of Diagnosis of Diagnosis of -
Diagnosis of Not included in Reg. Cuide 1.97.
TL1tBINE system failure system failure system failure system failure or guidance in or guidance in or guidance in or guidance in system opera- system opera- system opera- system opera-tion tion tion tion RI!R FLO'J OR PUMP Indication of Indication of Indication of Indication of -
DIScitARCE PRESSURE system avail . system avail- system avail- system avail-ability to ability to ability to. ability to maintain level maintain level maintain level maintain level l-
[
r
-. . . . . ~_
APPENDIX A SUt2tARY OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS BWR tujor Purpose for Indicated Accident Sequence Measured Variable w V S AE TC COMMENTS RER KEY VALVE Indication of Indication of Indication of Indication of -
Not included in Reg. Cuide 1.97, POSITIONS system avail- system avail- system avail- system avail-ability to ability to- ability to ability to maintain level maintain level maintain level maintain level Diagnosis of Diagnosis of Diagnosis of Diagnosis of system failure system failure system failure system failure Cuidance in cuidance in Guidance in Cuidance in system re- system re- system re- system re-configuration configuration configuration, configuration j, LPCS FLOW OR PCMP Indication of Indication of Indication of Indication of -
to DISCHARGE PRESSURE system avail , system avail- system avail- system avail-ability to ability to ability to ability to maintain level maintain level maintain level maintain level LPCS KEY VALVE Indication of Indication of Indication of Indication of -
Not included in Reg. Cuide 1.97.
POSITIONS system avail- systen avail- system avail-ability to
- system ability ability to ability to to maintain maintain level maintain level maintain level level Diagnosis of Diagnosis of Diagnosis of Diagnosis of system failure system failure system failure system failure
APPENHlX A SUf#t\PY or VARIABLES IDE! T!rf rD I:1 SEQUENCE EVALUATIONS B"R ttijor Purpose for Indicated Accident Sequence Measured Variabi ""
TW AE TC TQUV S,E SSW FLOW OR PUMP Indication of Indication of Indication of Indication of Indication of Not included in Reg. Guide 1.97.
DISC 11ARGE PRESSURE systen avail- system avall- system avail- system avail- system avall-ability for ab!11ty for ability for ability for ability for beat renoval heat ren oval beat removal heat removal heat removal through RHR through RHR through RHR through RHR through RIIR heat exchanger: heat exchangere beat exchanger > heat ex- heat exchangers changers SSW KEY VALVE Indication of Indication of Indication of Indication of Indication of Not included in Reg. Cuide 1.97 POSITIONS system status system status system status system status system status and diagnosis and diagnosis and diagnosis and diagnosis and diagnosis of R!fR failure of RHR failure of RIIR failure of RHR failure of RHR failure ca C3 Indication of Indication of Indication of Indication of RHR !! EAT EXCitANGERS Indication of INLET AND OUTLET effectiveness effectiveness effectiveness effectiveness effectiveness TEMPERATURES of RHR cooling of RUR cooling of RHR cooling of RHR coolinr of RHR cooling and diagnosis and diagnosis and diagnosis and diagnosis and diagnosis of failure of failure of f ailure of failure of failure ECCS PUMP ROOM Ir.Jication of Indication of Indication of Indication of Indication of TEMPERATURE system per- system per- system per- system per- system per-formance and formance and formance and formance and formance and diagnosis of diagnosis of diagnosis of diagnosis of diagnosis of ECCS pump ELCS pump ECCS pump ECCS pump ECCS pump failure failure failure failure failure SLCS FLOW OR PUMP - - - -
Indication of DISCIIAEGE PRESSURE accomplishment of system function.
SLCS EXPLOSIVE VALVE - - - - Indication of Not included in Reg. Cuide }.97.
POSITION system status l
_.-e -_ _ m_- - .--~m* -*=
APPENDIX A SLWfAAT OF VARIABLES IDENTIFIED IN SEQUENCE EVALUATIONS Hajor Purpose for Indicated Accident. Sequence Measured Variable TV TQUV SE AE TC g
SLCS TANK LEVEL Indication of SLCS avail-ability and operation RPV BORON CONCENTRATION - - - -
Indication of Could be useful backup under accident effectiveness conditions which make neutron flux of SLCS monitors less reliable p HYDROCEN CONCENTRATION Indication of Indication of Indication of Indication of Indication of s fuel damage fuel demage fuel damage fuel damage fuel damage o
CONTAINHENT EFTLIIENT Indication of Indication of Indication of Indication of Indication of RADIOACTIVITT Primary primary primary primary primary containment containment containment containment containment failure or failure or failure or failure or failure or leakage leakage leakage leakage leakage PRIMARY CONTAINMENT -
Indication of Indication of - -
Not included in Reg. Guide 1.97.
TEMPERATURE cause of initiating initiating event event
l APPENDIX B EMERGENCY OPERATING PROCEDURE VARIABLE LIST l
l l
Variable Name Comments and Notes Reactor vessel level Note 1 Reactor vessel pressure Note 1 i Drywell pressure Notes 1 and 2 l MSIV isolation signal Note 1 l SCRAM initiating signal when reactor greater Note 1 l than 3*. power:
1 Neutron flux (APRMS and IRMS)
Main steam line radiation MSlV closure Turbine stop valve closure Turbine control valve fast closure SCRAM discharge volume level Loss of condenser vacuum Suppression pool temperature Note 2 Drywell temperature Note 2 Containment temperature Note 2 Suppression pool level Note 2 Containment /drywell H2 concentration Notes 2 and 5 Annulus differential pressure Note 3 Room temperatures in safety related equipment areas Note 3 HVAC HX dhT in safety related areas Note 3 Area radiation levels Note 3 Floor drain sump levels Note 3 Containment water level Note 3 Identified offsite radioactivity release rates Note 4 RHR pump flow Note 5 LPCS pump flow Note 5 Condensate storage tank water level Note 5 RCIC turbine speed Note 5 Reactor cooldown rate Note 5 l
SLC tank level Note 5 SRV position Note 5 Control rod position Note 5 Cumulative boron injected Notes 5 and 6 Suppression pool H2 and 02 concentrations Note 5 Containment /Drywell pressure Note 5 Neutron flux Note 1 B-1
l Notes for Appendix B i
Note 1 - Variable is monitored for reactor pressure vessel control.
Note 2 - Variable is monitored for primary containment control.
Note 3 - Variable is monitored for secondary containment control.
Note 4 - Variable is monitored for radioactivity release control.
Note 5 - Variable is listed in the E0P procedures and has not been required for monitoring for control of any of the functions listed in Notes 1 through 4.
Note 6 - Derived from SLC tank level only.
l B-2
APPENDIX C This study utilizes the content of USNRC Regulatory Guide 1.97, Revision 3 for the purposes of this study. The following Type A variables which are specific to River Bend Station and not defined within the guide are presented below for information.
VARIABLE (1) Reactor Coolant System (RCS) Pressure (2) Containment /Drywell Hydrogen Concentration (3) Suppression Pool Water Temperature 1
4 l
l C-1
T APPENDIX D VARIABLE CATEGORY ANALYSIS INTRODUCTION The purpose of this analysis is to categorize the variables defined in Appendix A and B. The guidelines depicted in R.G. 1.97, Rev. 3 and plant specific information obtained on systems and components were used to determine these categories. The variables categorized in this analysis are selected from a wide range of significant accident event sequences.
Tables D-1 and D-2 of this appendix show the instrumentation categories for the RBS design in comparison with the regulatory guide categories. A note of explanation follows when RBS categories are less stringent as compared with regulatory categories or when regulatory categories are not available.
METHODOLOGY The methodology used in the analysis to categorize additional variables is in accordance with the regulatory position of R.G.
1.97, Rev. 3. The analysis was carried out in two steps.
The first step was to determine the variable type. Based on the variable type, the second step to determine the instrumentation category was performed. Tables 1 and 2 of this appendix show the outcome of this analysis with explanatory notes as applicable.
Variables Type Analysis This part of the analysis is based on defining the variables types. The five variable types are stated for reference as follows:
a) Type A - variables that provide primary information needed to permit the control room operating personnel to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.
b) Type B - variables that provide information to indicate whether plant safety functions are being accomplished.
c) Type C - variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product release.
D-1
APPENDIX D VARIABLE CATEGORY ANALYSIS d) Type D - variables that provide information to indicate the operation of individual safety systems and other systems important to safety, e) Type E - variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and for continuously assessing such releases.
Knowing the plant specific purpose of each variable, the type is selected pursuant to the above definitions.
Instrumentation Category Analysis The basis of this analysis is to determine whether the variable is a key variable or a non-key variable. A key variabic is defined as "a single or minimum number of variables that most directly indicate the accomplishment of safety function or operation of a safety system or radioactive material release".
Example of non-key variables are backup, diverse backup and diagnostic variables. In order to proceed further with this analysis it is necessary to define the three following categories of instrumentation as stated in R.G. 1.97, Rev. 3.
a) Category 1 - provides the most stringent requirements and is-intended for key variables, b) Category 2 provides less stringent requirements and generally applies to instrumentation designated for indicating system operating status, c) Category 3 provides requirements that will ensure that high quality off-the-shelf instrumentation is obtained and applies to backup or diagnostic instrumentation.
Based on the above definitions the relationship between variables types and the instrumentation categories can be summarized as follows:
Category 1 - key variable types A, B, or C Category 2 - key, variable types D or E Category 3 - backup or diagnostic variable types B, C, D, er E (Note: Type A variables cannot be backup or diagnostic.)
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s APPENDIK D TABLE D-1 VARIABLE CATEGORY ANALYSIS SOURCE: Accident Event Sequences (Tree)
MOST STRINGENT RBS UNIT 1 R.G. 1.97 REV. 3 CATEGORY .j VARIABLES CATEGORY DETERMINATION NOTES i Control Rod Position 3 i
3 i.
Neutron Flux 1 1 1 Reactor Vessel Water Level 1 1 Reactor Vessel Pressure 1 1 Core Temperature Not Noted 2 Recirculation Line Flow Not Noted 3 3 Primary Containment Pressure 1 1 Drywell Pressure 1 1 s Primary Containment Radiation 1 1 Suppression Pool Temperature 2 1 Suppression Pool Level 2 1 SRV Position 2 2 4 SRV Discharge Line Flow Not Noted Feedwater Flow 3 3 Feedwater Controller Position Not Noted 3 5 MSIV Positions Not Noted 1 14 Main Steam Line Flow Not Noted 3 5 Containment Isolation Valve Positions 1 1 Turbine Bypass Valve Position Not Noted 3 6 Condenser Hotwell Water Level Not Noted 3 7 Condenser Pressure Not Noted 3 8 Condensate Pump Discharge Pressure Not Noted 3 7 Condensate Storage Tank Level 3 3 Air Ejector Line Flow Not Noted 3 9 Circulating Water Flow Not Noted 3 9 HPCS Flow 2 2 HPCS Key Valve Positions Not Noted 3 10 RCIC Flow 2 2 RCIC Key Valve Positions Not Noted 3 10 Steam Flow to RCIC Turbine Not Noted 3 10 LPCS Flow 2 2 D-3
7 TABLE D-1 (Continued)
MOST STRINGENT RBS UNIT 1 R.G. 1.97 REV. 3 CATEGORY VARIABLES CATEGORY DETERMINATION NOTES LPCS Key Valve Positions Not Noted 3 11 Containment /Drywell Hydrogen Concentration 1 1 SSW Key Valve Positions Not Noted 3 12 RHR HX Inlet / Outlet Temperatures 2 2 SSW Flow Discharge Pressure / Flow 2 2 SLCS Tank Level 2 2 13 SLCS Discharge Pressure 2 2 RPV Boron Concentration (grab) 3 3 RHR Key Valve Positions Not Noted 3 10 Status of IIPCS DG Not Noted 3 Containment Effluent Radioactivity 2 2 Room Temperatures for Detection of Leakage from Containment Breach Not Noted 3 15 Containment Atmosphere Temperature 3 3 NOTES:
- 1. Though diverse backup variables are available following loss of this key variable, a Category 1 classification of this instrumentation is justified because no other instrumentation directly measures the core neutron population.
- 2. The use and category classification of this variable is under consideration per RG 1.97, Rev. 3. An analysis (Reference 6) with thermocouples located at the top of the core shows that for conditions typical of a small LOCA there is a delay of at I ast 10 minutes between the start of core uncovery and the time when the thermocouples read 40*F above saturation. Also, the operation of relief valves during this event will interfere with thermocouple operation and render them useless. Hence, River Bend Station Unit i does not use this instrumentation.
- 3. No category classification is given to recirculation line flow by R.G. 1.97, Rev. 3. This variable will be classified as Category 3 because it is a diverse backup variable to the accomplishment of the safety function recirculation pump trip (RPT). Reactor water level will begin to drop within seconds D-4
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q TABLE D-1 (Continued) following a loss of feedwater in conjunction with a . failure to scram. The operator must verify a rapid RPT followed by initiation of RCIC and HPCS. An unsuccessful or delayed RPT might lead to a degraded core if no immediate operator action is taken,-
- 4. The SRV position indication instrumentation (acoustic monitors) are fully qualified, continuously displayed, and have Class 1E power as per Category 2.
- 5. These variables are not included in R.G. 1.97, Rev.-3. They are classified as Category 3 and are used as diverse backup variables in indication of response to initiating event as well' plant restoration following an event, where required, in all.
major sequences.
- 6. Turbine Bypass Valve Position indication instrumentation is classified as Category 3. For example, the operator will obtain backup indication of availability of the power conversion system following loss of condenser vacuum or feedwater.
- 7. These are backup variables and they will provide backup indication of availability of the power conversion system, feedwater, or condensate pumps for makeup. These instruments are classified as Category 3.
- 8. This variable will be used by the operator as an indication of the initiating event and as backup indication of availability of the power conversion system. The condenser pressure indicating instrument is classified as Category 3.
- 9. The operator will use these diagnostic variables for. the diagnosis of the' initiating event or determination of.the availability of the power conversion system or condenser as a heat sink. The instruments are classified as Category 3.
- 10. The operator will use these~ variables as backup information to.
RCIC/HPCS/RHR flow for ' indication of RCIC,' HPCS and RHR systems :
status or diagnosis of failure _for a111 events other than AE sequence. Therefore, the Category 3 classification _of.this-instrumentation is justified.
11.
LPCS' key valve positions will provide the operator with backup
~
or diagnostic information to LPCS system availability or diagnosis ~of failure. .The instrumentation _is classified as'-
Category 3.
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e l
TABLE D-1 (Continued)
- 12. This backup variable will indicate the ESF system status or diagnose failure and is classified as Category 3 instrumentation.
- 13. The operator will use the SLCS storage tank level variabic in conjunction with SLCS pump discharge pressure indication for an indication of SLCS flow. Hence, a Category 2 classification is given to the SLCS storage tank level instrumentation.
t
- 14. This is a key variable for ascertaining the containment isolation function is performed and is classified as Category .1 instrumentation.
- 15. Room temperature for detection of leakage from containment breach variables are backup variables to the key variable of.
containment radiation and the instrumentation is classified as.
Category 3.
l 1
1 D-6 ' '
f APPENDIX D TABLE D-2
, VARIABLE CATEGORY ANALYSIS SOURCE: Emergency Operating Procedures MOST STRINGENT RBS UNIT 1 R.G. 1.97 REV. 3 CATEGORY VARIABLES CATEGORY DETERMINATION NOTES Reactor Vessel Water Level 1 1 Reactor Vessel Pressure 1 1 Drywell Pressure 1 1 MSIV Positions Not Noted 1 1 Neutron Flux 1 1 2 Suppression Pool Temperature 2 1 Drywell Atmosphere Temperature 2 1 Containment Atmosphere Temperature Not Noted 3 3 Suppression Pool Level 2 1 Containment /Drywell Hydrogen Concentration 1 1 Primary Containment Pressure 1 1 Area Radiation 3 3 Drywell Equipment and Floor Drain Sump Water Level 1 2 4 RHR Flow 2 2 LPCS Flow 2 2 Condensate Storage Tank Level 3 3 RCIC Turbine Speed Not Noted 3 5 Reactor Cooldown Rate Not Noted 3 6 SLCS Tank Level 2 2 7 SRV Position 2 2 8 Control Rod Position 3 3 Cumulative Boron Injected 3 3 Containment /Drywell Oxygen Concentration Turbine Stop Valve Positions Not Noted 3 9 Turbine Control Valve Positions Not Noted 3- 9 Scram Discharge Volume Level Not Noted 3 10 Liquid Effluent Radioactivity Not Noted 3 11 Primary Containment Radiation 1 1 Containment Effluent Radioactivity 2 2 Main Steam Line Radiation Not Noted 3 12 Airborne Radioactivity Releases 2 2 13 D-7 L
TABLE D-2 (Continued)
MOST STRINGENT RBS UNIT 1 R.G. 1.97 REV. 3 CATEGORY VARIABLES CATEGORY DETERMINATION NOTES Suppression Pool Hydrogen /
0xygen Concentration Not Noted 3 -14 Containment Water Level Not Noted 3 .15 l Condenser Pressure Not Noted 3 16 Room Temperatures for Detection of Leakage From Containment Breach Not Noted 3 17 Turbine Bypass Valve Position 3 3 i
NOTES:
t i
i
- 1. See Note 14 of Table D-1.
] 2. Jee Note 1 of Table D-1.
j 3.
Primary containment temperature can be used by the operator as a diverse backup variable to the key variable of drywell temperature or primary containment radiation level. The-instrumentation is classified as Category 3.
- 4. This variable will be used as a backup or diagnostic variable.
to the key variables like drywell pressure, temperature, or radiation level in the event of RCS pressure boundary I
2 breach / leakage. Therefore, the drywell drain sump-level is classified as Category 2 against the R.G.1.97, Rev. 3 recommendation of Category 1. Moreover, this instrumentation serves no useful accident monitoring function other than providing indication and alarm.
.. .t
- 5. RCIC turbine speed is a diverse backup and diagnostic' variable to RCIC system operation and the instrumentation is' classified as Category 3.
I
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- 6. Reactor cooldown rate is a diagnostic variable for the protection of reactor pressure vessel and is-controlled by administrative' procedure. A Category 3 classification is given
- to this instrumentation.
1
- 7. See Note 13 of Table D-1.
4 1
i D .
--,-,,--,---,n., , .,~c.~., - . - - . , . - . , , ,
. . .- , . - , . - , - . , ., , , en- , , ,
1 TABLE D-2 (Continued)
- 8. See Note 4 of Table D-1.
- 9. These are diagnostic variables indicating the status 01 key valves for power conversion system operation and are classified as Category 3 instrumentation.
- 10. The scram discharge volume Icvel will provide the operator with backup information as to the scram initiating event and is classified as Category 3.
- 11. This variable provides diagnostic information on equipment
! performance and is classified as Category 3 instrumentstion.
1
- 12. Main steam line radiation is not considered a key variable insofar as the scope of this report is concerned. The reason being that the usefulness of information by monitoring this variable, in terms of helping the operator in his efforts to prevent and mitigate accidents, has not been substantiated.
Hence, this instrumentation is classified as Category 3.
- 13. Airborne radioactivity release is a key variable and classified as Category 2 by R.G. 1.97, Rev. 3. RBS concurs with this regulatory position.
j 14. This variable is a backup to the key variable of I containment /drywell hydrogen concentration and the variable is classified as Category 3.
- 15. This variable can be used for diagnostic information in the event post-accident containment flooding is required. The instrumentation to be used is classified as Category 3.
- 16. See Note 8 of Table D-1.
- 17. See Note 15 of Table D-1.
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