ML20042E503
| ML20042E503 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/09/1990 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20042E501 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 028, 028-R01, 28, 28-R1, GL-82-33, NUDOCS 9004230321 | |
| Download: ML20042E503 (133) | |
Text
{{#Wiki_filter:- t se v' ser OYSTER CREEK RESPONSE TO USNRC REGULATORY GUIDE 1.97 l l TOPICAL REPORT # 028 (Revision 1) -- l
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PROJECT NUMBER: 5300-40215 Revision 0 Revision 1 M. A. ALAMMAR J. Roumes/K.' R...Eibon -I K. R. EIBON , , January 9', 1986 January 11, 1990 i i APPROVALS: ' 1 Av YA
- MANAGER, SAFETY ANALYSIS & PLANT CONTROL DATE .
dt i f0 i MANAG EP&I # IlATE i
.>e N DIRECTOR, SYSfEMS ENGINEERING 4-r-r.
DATE i
\
. Audauuc- 4/6/@
DIRECTOR, ENGINEERING & DESIGN DA'TE l
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i 9004230321 900 ,39 M"g PDR ADOCK PDR -;#'$ P ,
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E i > -TR 038 Rev. 1 Page 2 of 71 ABSTRACT This Topical Report describes how the Oyster Creek Nuclear Generating Station .{' meets the requirements of US NRC Reg. Guide 1.97, " Instrumentation for ... Nuclear Power Plants ... Following an Accident", Revision 3. Table 1 of this j- , report is based upon previous GPUN submittals and the BWR Owners Group' I^ Position, " Response to Reg._ Guide 1.97," July 1982, Ref. 2. This report , recommends that various DC instruments be required for monitoring post-accident .i' conditions.
-j Recommendations for displaying those parameters for-which instrumentation L' currently does not exist or upgrading existing instrumentation which is not in compliance are given in Table III.
The purpose of Revision 1 is to update TR 028 to reflect changes to plant hardware since January 1986. It also includes clarifications requested by the NRC staff concerning the previous submittal and incorporates findings of an interdisciplinary, internal review of the OC Reg. Guide 1.97 instrumentation.
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This revision also brings Table III into closer agreement with-the recommended format of NUREG 0737, Supplement 1. t TRP009/2
TR 028 i
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TABLE OF CONTENTS *
'r EMn3 ABSTRACT ... . . . . .. . . . . . . . . .. . . , . . . . . . . . .. 2 l-
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TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . ' . . .. 3-t
1.0 INTRODUCTION
. . . - . ... . . . . . . . . . . . . . . . . . ~ . . .- 4 e
L 2.0 METHODOLOGY . . . .. . . . . . . . .. . . . . . . . .. .-. . . . 4' .g
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2.1 NRC Position . . . . .. . . . . . . . . . . . .~. . . . . 4 !
'F 2.2 BWROG Position . . . . . . . . . . . . . . .. . . .'. .- . 5 2.3 GPUN Position . . . . . . . . . . . . . . . . . . . . . . . 5 L e
i 2.4 Definitions . . . . . . . . . . . . . . . . . . . . ... . . .6 ' 3.0 RESULTS . . . . . . . . ... . . .,. . . . . . . . .. . . . .. 7
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4.0 CONCLUSION
. . . . . . . . . . . . . . . . . . . . . . . . ... . 8-5.0 RECOMMENDATIONS . . . . . . . . . . . . . . . .. . . . . . . 8-
6.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . .. .. 8 -'
i t 7.0 TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 8 i l TABLE I Proposed Minimum Parameter Set for Oyster Creek . . . 9 i TABLE II Introduction'. . . . . . . . .. . - . . . . . .. . - . . . 26 i 6capliance Status of Minimum Parameter-Set /, ' for Oyster Creek . . . . . . . .. . . . . . . . . . .- 35 TABLE III Reg. Guide 1.97 Instrument List and Actions Required i to Satisfy RG 1.97' Requirements for OCNGS Minimum Parameter Set'
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ATTACHMENT - BWROG Position on Reg. Guide 1.97 Revision 2 l TRP009/3 l l
TR 088 Rev. 1 I
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1.0 INTRODUCTION
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USNRC Regulatory. Guide 1.97, Revision 2, was issued in December, 1980. The intent of Regulatory Guide 1.97 is to ensure that all light water -7 reactors are properly instrumented to assure proper responses during and l following an accident. Regulatory Guide 1.97 provides methods which are { acceptable to the NRC staff to comply with the commission's requirements for instrumentation necessary to monitor and' assess plant conditions- , during and following an accident.. 4 In addition, Generic Letter 82-33 required that all utilities provide a.
, schedule for unplementing Regulatory Guide 1.97. On April 15, 1983,~ GPU. -
Nuclear advised the NRC that GPU would perform a plant specific study for oyster Creek and submit the results of this study, including a schedule of implementation by May 1, 1984. . The purpose of this report is to provide the following information forf , oyster Creek. Nuclear Generating Station-(OCNGS):
- Specify the minimum parameter set and ~ instrumentation requirements /*,
necessary to meet the intent of R.G. 1.97. (Revision 3 of Regulatory Cuid2 1.97 has been reviewed and used in Revision 1 of this report.) , [ - Determine the status of the existing instrumentation'against the R.G. 1.97 design and qualification requirements.
- Provide justification for the variables and instruments that do not-comply with the requirements of R.G. 1.97 but are considered adequate - ,
for OCNGS or identify these as deficiencies with proposed l /, resolutions, 1 l - Provide explanations for items which exceed NRC requirements for . oyster Creek. ! 2.0 METHODOLOGY 2.1 NRC Position The NRC in R.G. 1.97, Revision 2, Table 1 identified-a minimum list of "BE, "C", "D" and "E" type parameters for' light water reactors and their required ranges. Type "A" variables were not'specified because they will depend on the specified planned-manual operator actions required during Reactor Accidents. The NRC has stated that Type "A" variables would be determined on a plant specific basis. I-l l TRP009/4 i t i -. .-. - . . . .. - - . , _ - - . . . . . - - - __w.. .. _ - . - . - - . , . - . . - . - -
. __ - _ ~ _ _ _ _ . . . . . _ _ ._ . _ _ _ _ _ _ . _ . _ _ _ _ _ .
I TR 028
' ' R;v. 1 .j Page.5 of 71 2.2 BWROG Position on R. G. 1.97. Revision 2 The recommendations of the BWROG are documented in a report entitled "BWR owners Group Position'on NRC Regulatory Guide l.97, Revision -
2." The BWROG identified five variables as Type,"A" and proposed six' additional variables as potential Type "A". 'Also, the BWROG stated I its position with regard to Types "B", "C", "D", and "E" variables . j and design categories as identified in R.G. 1.97, Table 1. The BWROG j* q also used symptom oriented EOPs and critical safety function approach' J in the formulation of their position on R.G. 1.97. i A later subcommittee of'the BWROG has submitted a report to the NRC staff concerning the Neutron Monitoring, variable B-1. GPUN is [, following the progress of this report as stated in Table III.- t 2.3 GPUN Position , GPUN concurs with the recommendations of the BWROG except as noted! in Table I of this report. This report goes beyond NRC and BWROG j' recommendations for Type B, C, D and E variables. Considerations during OCNGS variable selection were ! 4; 06snarios that could lead to core damage or radiation release U l to the public, which may not have been considered as a basis in the analysis included in the Facility Design and Safety Analy-i sis Report, (i.e., the TMI-2 Accident on March 28, 1979); and i therefore, could have been overlooked. , I b) Another parameter selection technique employed was a: critical - p safety function (i.e., reactivity, core: cooling, reactor i coolant system integrity, and containment integrity)' approach. /,. The critical function. approach was also used in development of the Emergency operating Procedures (EOPs) and their entry ; conditions. c) In the original selection of Type "A" variables, GPUN-con-- sidered' Design Basis Accident Events to ensure that safety /, systems' success during a DBA does not require operator action based upon parameters other than those selected.~ d) A, review of the original (event-oriented) emergency procedures for any specified manual actions proved futile. Most procedures simply stated " ensure all automatic functions have taken place" and assume most everything functions properly. In the case of the refueling accident DBA, no specific procedure currently. exists. Symptom Based Emergency operating Procedures were later used as a source of planned manual actions.' The current Oyster Creek EOPs are based upon Revision 4 of the BWR Owners Group /. Emergency Procedure Guidelines. TRP009/5
1 TR 028 Rev. 1 Page:6 of 71 J e l 2.4 Definitions The information available to the Control Room Operator is separated j* , by usage into instrument Types by Reg. Guide 1.97. ! The following definitions of variable Types are taken directly.from R.G. 1.97, Revision 2. Tvoa "A" "Those variables to be monitored that provide the primary I information required to permit the control Room Operator to take the specified manually controlled actions for which no t automatic control is provided and that are required for safety systems to accomplish their safety function for-desian basis accident events." R.G. 1.97 calls these <{ variables plant specific and requires plants to specify them as needed. - Tvom "B" "Those indications that provide information to indicate whether plant safety functions are being accomplished. e" Tvom "c" "Those variables that provide information to indicate the potential for being breached or the actual breach of.the-barriers to fission product release, i.e.,.-fuel cladding / ! primary coolant pressure boundary, and containment." 4 Tvom "D" "Those variables that provide'information to; indicate the 3 operation of' individual safety systems and other systems- , important to safety.". :
.y.-
( Tvoa "E" "Those variables to be monitored as required'for use-in i determining the magnitude'of the release of radioactive-materials and in continually asseesing such releases." The instrumentation is further classified into categories which set the design and qualification criteria, e.g. redundancy, seismic /-. 't
; requirements. The most important. instruments are thus held to higher standards in order to improve reliability, cateaorv i "Provides the most stringent requirements and is intended for key variables";i.e.,'"provides for full j,
- qualification, redundancy and real time display".
cateaorv 2 "Provides less stringent requirements and generally' , applies to instrumentation-designed for indicating system operating status." "does not (of itself). include seismic qualification, redundancy or con- /, tinuous display" (Type "D" and "E" key variables).. cateaorv 3 "High-quality commercial grade equipment."
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TRP009/6
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TR'028 Rev. 1 Page-7-of 71 These definitions provide the functional requirements for the design and qualification criteria for instruments identified in this report. , s NOTE: Variables that could be considered A And B And C variables, e.g. ;
. Reactor Water Level,-are repeated only where necessary. They_'
have been placed~for clarity only in their highest type'and- , category. .i! t 3.0 RESULTS [ The results of this report are summarized in Tables ~I, II, and III. A. Table I - Proposed Minimum Parameter Set for. Oyster Creek Table-I lists the variables and categories proposed by R.G. 1.97 andi the BWROG's and-GPUN's-position on these variables. Table I also-contains the RWROG's recommended and= proposed Type;"A" variables. Explanation for inclusion or exception to the.R.G. 1.97 or BWROG's , recommendation or inclusion of' additional variables are contained in~ Table I as required. , The basis for developing Table I of this report was R.G. 1.97, Revision 3, the BWROG response to Reg. Guide 1.97'(attached) and'the li ; Oyster Creek Emergency Operating Procedures. (See'Section 2.3) B. Table II - Compliance Status of Minimum Parameter Set'for Oyster-Creek ll /. ; Table II develops the minimum parameter set for. Oyster Creek based on Table I, the comments'of the-BWROG and a detailed review of the-Oyster Creek Control Room in light of emergency procedure: !- requirements. Table II provides the status of these parameters. based on the major requirements listed in Table one of R.G. 1.97; range, power supplies, redundancy,. environmental and seismic qualification. Table II also contains an introduction discussing effective OC design- I criteria and how modern design standards have been: applied to a BWR2 /* Mark .1, constructed in-1968. j C. Table III - Regulatory Guide 1.97 Instrument List and Actions l Required to satisfy-R.G. 1.97 Requirements for' OCNGS Minimum Parameter Set - [- , Table III details the instruments used to monitor the R.G. 1.97 parameters identified in Table II and addresses deficiencies of the. existing instrumentation. TRP009/7 l l 1 w ,v-, w
4-8' . TR O28. Rev. 1 Page 8 of 71 1
4.0 CONCLUSION
The required parameter set for Oyster Creek is developed in Table I. The. status of compliance for each proposed parameter is identified in- ^ Table II. It is concluded that 'several instruments need upgrading while others are to be accepted as is. These items.are-identified ~and discussed in Tables II and III. 5.0 RECOMMENDATIONS It is recommended that the parameter instrumentation requiring upgrading. . identified in Table III be performed in accordance with the oC integrated schedule.
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6.0 REFERENCES
- 1. Reg. Guide 1.97, Revision 2, December 1980
- 2. BWROG Response to R.G. 1.97, Revision 2, December, 1980 (Attached)
- 3. Reg. Guide 1.97, Revision 3, May 1983
- 4. GPUN TDR 578 - Torus to Drywell Bypass Area Evaluation
- 5. GPUN - E.Q. Master List Rev. 4
- 6. NUREG 0737,. Supplement la Clarification of TMI Action Plan
- 7. TDR 904, OC Equipment Qualification
- 8. Topical Report 065, OCNGS SPDS Certifications !d
- 9. Standard Technical Specifications for Boiling Water Reactors
- 10. Oyster Creek Generic cable Specification, SP-9000-41-005 +
- 11. . Emergency Operating Procedures, EMG-3200 Series f 7.0 TABLES I
- l. A) Table I - Proposed Minimum Parameter Set for Oyster Creek
- B) Table II - Compliance Status of Minimum Parameter Set' for Oyster '
Creek /. c) Table III - Regulatory Guide 1.97 Instrument List and ' Actions Required to Satisfy R.G.~1.97 Requirements;for OCNGS Minimum Parameter Set-L r TRP009/8 L l . t l-
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e T2 028 Rev. 1 Page 10 of 71 TABLE I PPOPOSED MINHELM PARAMETER SET FOR OYSTER CREEE Recougmended By WRC BWROG GPUN Parameter Description Y/W Cet Y/N Cet Y/W Cet Explanation A-2 RPV Water Level - 1 Y 1 Y 1 Besides being on EOP entry condition menuet
- _ l [. '
operator actions are to restore and meintain water tevet to:
- 1) Control core sprey to prevent flooding ~
the vessel.
- 2) Teke menuel actione if sefety systems i do not function property.
- 3) Ensure levet is low enough to prevent severe water houser prior to operation of isoletion condensers.
The safety functions are: 3 e) Adequete core cooling.
' b) Reactor cootent systent integrity.
i Y 's Yes, Recommended for laptementation et Att Plants N s No, Not Reconuended for Iaptementation et Att Plants
, 'P = Potentfot,' Implement if Plant Specific Evaluation Warrants TRP009A/2
_ -, , - _ _ -., _ .. . _ - , c- , , -, _ .. _. _ . _ . .. .
70 028 Gev. 1 Page 11 of 71 TA8LE I PROPOSED MIN!9RM PARAMETER SET FOR OYSTER CREEK Recomenended By NRC SWROG GPUN Parameter Description' Y/N Cet Y/N Cet Y/N Cet Explanation A-3 Torus Water Temperature - 1 Y 1 Y 1 Torus water temperature is en indication of: a) Exceeding design temperature timits. b) Adequete heet absorption capability in the torus. c) ECCS pump net positive suction heed (NPSN). Besides being en E0P entry condition menuet Operator Actions are:
. l/.
a) Initiate torus cooling. b) Scree the reactor or initiate liquid poison during en ATUS.- c) . Emergency depressurization of the reector to stay within the heet capacity teaperature timit. d) Close a stuck open EMRY. B The safety functions ere:
- 1) Containment integrity.
e 2)- Reactor cootent it'tegrity. ~
.-I Y = Yes, Reconsnended for taptementation et Att Plants ; -
N = No, Not Recomunended for laptementation et Att Plants P = Potentist, Implement if Plant Specific Evolustion Warrants TRP009A/3
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TR 028 Rev. 1 Page 12 of 71 TABLE I PROPOSED MINIMUM PARAMETER SET FOR OYSTER CREEK Recomunentled By NRC BWROG GPtM Parameter Description Y/N Cat Y/N Cat Y/A Cat Emptanation A-4 Torus Water Levet Besides being an E0P entry condition manuel l[- 1 Y 1 Y 1 operator actions are: a) Maintaining torus tevet below the torus load timit curve or depressurire the reactor to below the toad timit in the E0Ps. b) A&3 water as necessary to ensure ERCS equipment has adequate NPSH. c). Limit an increasing levet to ensure an adequate torus vent path. Safety. functions are: a) Contairinent integrity. b) Adequate core cooting. A-5 Drywett Pressure - 1 P 1 Y 1 The BWROG reconssends, Ref. 2, that drywell pressure be considered a Type A variable for those plants noj having automatic initiation of drywett spray. OC has an automatically ;
- initiated contairunent spray system. -
- If OC deletes the auto-start provisions of the contairwent spray controt logic, this vertebte will be re-examined. .
Y = Yes, Reconenended for taptementation at Att Plants
' N = No, Not Recommended for Ipplementation at Alt . Plants P = Potential, Inplement if Plant Specific Evaluation Warrants' TRP009A/4
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TR 628
, Rev. 1 ., .
Page 13 of 71 TA9tE I PRor0 SED MIW19Ut PARAMETER SET FOR OYSTER CREEK Recomumended 9y WRC BWROG GPUN Parameter Description Y/W Cet Y/N Cet Y/W Cet Explanation A-6 Drywett combustible Gas - 1 P 1 Y 1 Besides being en E0P entry condition menuet , actions are: l/ Concentration (0 2 ' "2I e) If hydrogen exceeds minianse detectable
~
tevels or explosive limits, institute hydrogen control procedures. Safety functions are: a) Adequete core cooling. b) Primary conteirament integrity. - A-7 condensate Storage Tert - 1 P 1 N - Only proposed for those plants with WPCI or Level RCIC systems with automatic suction transfer (i.e. from CST to torus). These systems do [. ' not exist et Oyster Creek (see 0-2). See WWROG discussion in Appendia A, Pope 12 of Reference 2.- Y = Yes, Recomumended for taptementation et Att Plants - N = No, Not Reccausended for Implementation at All Ptents
, P = Potentist, Isplement if Plant Specific Evetuation verrants -
TRP009A/S -
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w TO 028 Rev. 1 Page 14 of 71 TABLE I PROPOSED MININtM PARAMETER SET FOR OYSTER CREEIC Rectamended By NRC SWROG GPUN Parameter Description Y/N Cet Y/N Cet Y/N Cat Explanation A-8 Emergency Diesel - 1 P 1 N - Only nonessential loads would be edded to the Generator toad EDG efter e design besis occident e.g. RSCCW, service water, drywell fans and on air compressor. These loods are not needed [g lamedletely following a reector eccident and are not necessery to soitigste Dee!gn Besis events. (Also see D-22. Not in Tebte It because of CPUN position.) A-9 Resetor Building - 1 P 1 N - See SWROG discussion in Appendix *A", Pege 12 Flood tevet of Ref. 2. Capacity for operetor to p g reactor *ullding stap to torus does not exist and this is "en aid not the accomptishment of a safety function". Not in Tabte II because of GPUN position. ' Y = Yes,- Reccaumended for laptementation et All Plants N
- Wo, Not Recommended for laptementation et All Plants
, . P = Potentiel, luplement if Plant Specific Evolustion Werrants
- 1RP009A/6
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TQ 0?8 Rev. 1 Page 15 of 71 TABLE I PROPOSED MIN!8RM PARAMETER SEY FOR OYSTER CREEK Recomumended By NRC OWROG GPUN Parameter Description Y/N Cat Y/N Cat Y/W Cat Explanation B-1 Neutron Flux Y 1 Y 2 See Table til g/ B-2 Control Rod Position Y 3 Y 3 Y 3 B-3 RCS soren Concentration Y 3 Y 3 Y 3 B-4 RPV Water Levet Y 1 Y 1 Y 1 B-S BWR Core Thermocouples Y 1 N - N - Requirement deleted in #evision 3 of Reguletory Guide 1.97. B-6 RPV Pressure Y 1 Y 1 Y 1 B-7 Drywell Pressure Y 1 .Y 1 Y 1 See A-5, 8-9 and 8-12. Drywell pressure is the key indication of conteirunent integrity and aids the operator.In determining menuel actions for rion-tiesign bests , events. For these ftmetions the occuracy of /. the existing wide range conteifunent pressure instrtsments is acceptsble, Ref. 7. ' Should menuet initiation of conteirunent sprey be necessary, a norrow range conteltument pressure instrument is located at the conteirunent 'sprey control station, see 8-12.- . l . Y = Yes, Reconenended for 19 t ementation at All Plants N = No, Not Rh..s.M for- Inplementation at All Plants
. P = Potentlat, !fnplement if Plant Specific Evaluation Warrants TRP009A/7 _
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T2 028 Cev. 1 Page 16 of 71 TABLE I PROPOSED MIN 18RSI P M TER SET FOR OYSTER CREEK Recommended By NRC Blm0G GPUN Parameter Description Y/W Cet Y/N Cet Y/N Cet Emptenation 1
- 8-8 Drywell Stmp Levet Y i Y 3 Y 3 See Attechnent, issue 4 B-9 Primary Containment Y 1 Y 1 Y 1 See A-5, 3-12, 5-7. Conteirament pressure con Pressure conservatively be considered dryuell pressure [-
because es shown in Ref. 4, dryueli pressure envelopes torus pressure. > 8-10 Drywett Isolation valve Y 1 .Y 1 Y 1 Position B-11 Reactor Water Level Cold ' Reference Leg Tenp - - - - Y -2 2 See P 4 (Drywelt Tenperature) 8-12 forus Pressure - - - . Y 2 See AS, 8-9 h Y = Yes, Recomunended for Implementetlen et Att Plants a N = No, Not Recommended for Isptementetion at Ati Ptonts
, P. = Potentist, Isplement if Plant Specific Evoluotion Worrants TRP009A/8-
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_ ___ _ _ - _ _ - - ._ _ . - . ~ --- - - TO 028 CW.1 Pege 17 of 71 TA8tE I NED MINIMlm PARAMETER SET FOR OYSTER CREEE Rectosuended 8v NRC BuROG GPuW Parameter Description Y/N Cet Y/4 Cet Y/N Cet Explanetton C-1 Radioactivity Conc. or Y 1 N - N - BuROG Issue 5 of Reference 2. Radiation Levet in Primary Cootent C-2 Anotysis of Reactor Y 3 Y 3 1 3 Cootent (Grab Semple) C-3 But Core Thermocouples N - N - N - See 8-5, deleted by Rev. 3 of Reg. Guide 1.97. l/* C-4 Dryvett Area Nigh Y 3 Y 3- Y 3 Radiation C-5 Drywell Supp and Equipment Y 1. Y 3 Y 3 BuROG Issue 4. Use eterm, teek rate indicator, Drain Tank Levet flow Integretors and pump rim time integretors. See Table III and 8-8. ' C-6 Conteirunent Effluent Y 3 Y 3 Y- 3 variebtes C-6 and C-8 are considered to be.
. Radioactivity (Noble Geses) ' essentietty equivalent'et Oyster Creek because of physicet configuration.
. 4 Y = Yes, Recomumended for Implementation et Ait Plants N = No, Not Reconumended for Imptementation at At1 PiontS P = Potentiel, Implement if Plant Specific Evetustion Werrants TRP009A/9
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TR 028
. Rev. 1 .,
Page 18 of 71 TA8tE I PROPOSED MIN 19UI PARAMETER SET FOR OYSTER CREEK
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Recomumended By NRC OWROG GPUN Parameter Description Y/N Cet Y/N Cet Y/N Cet Explanation C-7 Radiation Exposure Rete N - N - N - Reg. Guide 1.97 Rev 3 deletes reepsirements. See BWROG issue 6 of Reference 2. This verfeble is not a good indicator for detection of breech due to streedy high redietion levels. C-8 Effluent Radioactivity 7 2 Y 2 Y 3 See C-6, the Reactor and Turbine tuilding fromeReactor and Turbine effluent monitors are both located outside Building (Noble Geses) secondary conteirament in mild enviroruments. c-9 Reactor Building Pressure - . Y 2 Reactor building pressure (etmospheric d/p) is the best indication of secondary conteirument integrity and proper operation of the Stantby Gas Treatment System. See Val C-10 Rx Building Tenperature - - - - Y 3 White currently 'en E0P entry cond'*lon, this is e relatively insensitive indication of primary conteirament teakage. Reector. building tours and radiation monitors will better detect primary contelnment piping teoks. 1 Y e Yes, Reconnended for Inplementation et Att Plants . N = No, Not Recomunended for !splementation at Att Plants
. P = Potentiet,' Impteps if Plant Specific Evetuation Warrants
'TRP009A/10 i .
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TR 028 tev. 1 Page 18 of 71 TABLE I PPOPOSED M!ntpRM PARAfETER SET FOR OYSTER CREEK Recomumended 8v NRC BWROG GPUN Parameter Description Y/N Cat Y/N Cat Y/N Cat Explanation C-7 Radiation Exposure Rate N - N - N - Reg. Guide 1.97 Rev 3 deletes rewirements. See 9WROG issue 6 of Reference 2. This variable is not a good indicator for detection of breach d;e to stready high radiation levels. C-8 Effluent Radioactivity Y 2 Y 2 Y 3 See C-6, the Reactor and Turbine Building from Reactor and Turbine effluent monitors are both located outside Building (Noble Gases) secondary conteirusent in mild envirersnents. C-9 Reactor Building Pressure - - - - Y 2 Reactor building pressure (steospheric d/p) is the best indication of secondary contairunent integrity and proper operation of the Stanrhy Gas Treatment System. See D-21 C-10 Rx Building Temerature - -
. - Y 3 White currently on EOP entry condition, this is a relatively insensitive indication of primary contalrunent teskage. Reactor building tours and radiation monitors witI better detect primary conteirunent piping teoks.
i Y = Yes,' Reconunended for Implementation at All Plants. ' N = No, Not Reconenended for Implementation at Att Plants
. 1P = Potential,'I mtemant if Plant Specific Evaluation Warrants.
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TR 028 Cev. 1 Page 19 of 71 TABLE I PROPOSED MIN!RM PARAMETER SET FOR OYSTER CREEK Recomumended By NRC Sut0C GPUN Parameter Description Y/N Cet Y/N Cet Y/N Cat Emptenation D-1 Main Feeduster Flow Y 3 Y 3 Y 3 0-2 Condensate Storage Tank Y 3 Y 3 Y 3 ' Level D-3 Torus Spray Flow Y 2 N - N - See SWROG issue 7 of Reference 2. Also see D-16, Table III. D-4 Drywell Atmosphere Temp. Y 2 Y 2 Y 2 Reactor water tevet cold reference leg temperatures are indicative of drywelt temperatures. . Due to their tocations and Oyster Creek post-accident expected drywelt turbulence, these thennoccuptes are representative of dryvett / temperature. This is eteo confirmed try ooservation of normat operat_ing temperatures. See 3-11. D-5 Drywell Spray Flow Y 2 N - N - See BWROG issue 7 of Reference 2. See 0-16 (Cont. Sprey Flow) i D-6 Main Steare Line Y 2 N - N~ - Not recomumended if not part 'of plant design Isotation Valve Leakage .(see Reference 2, pg. 20)
.i Control System Pressure Y = Yes, Recomunended for Implementation et Att Plants ~
E i No, Not Recomunended for Implementation at Att Plants
.
- Potential, . Implement if Plant Specific Evoluotion Warrants - '
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Y2 028 Rev. 1 Page 20 of 71 TABLE I PROPOSED MINI 9EM PARAMETER SET FOR OYSTER CREEK Recomumended By NRC SWROG GPUN Parameter Description Y/N Cet Y/N Cet Y/N Cet Explanetion
- D-7 Velve Monitoring System The Oyster Creek VMS is en scoustic monitoring (VMS) - vesset Pressure system which indicates volve position, shut or reliefs not shut, in the control room. Direct operator EMRV'S (5) Y 2 Y 2 Y. 2 oction witt close en open EMRV. Nowever, the Oyster Creek code safety relief volves are Code Safety (16) - - - -
Y 3 completely automatic and con not be over vetves ridden. D-8 Isolation Condenser Y 2 Y 2 .Y 2 Shett-side Water Levet D-9 Isolation Condenser Y 2 Y 2 Y 2 System vetve Position D-10 RCIC Flow Y 2 N - N - System not instetted et Oyster Creek, i D-11 MPCI Flow 'Y . 2 ' N' - N - System not instetted'st Oyster Creek. + i D-12 Core Spray Flow Y 2. Y 2 Y- 2 D-13 LPCI Flow Y- 2 . N ,. - :N - System not instetled at Oyster Creek. 0-14 SLCS Flow Y- 2 Y 3 Y ~'3 See SWROG Issue 9 of Reference 2. ~
/. ,
. D-15 SLCS Storage Tank tevet Y' '2' Y 3 .Y 3 'See 9WROG tssue 10.
D-16 . Conteirunent Spray System Y 2 Y. 2- Y 2 Flow Y = Yes, Recesumended for Isplementation et Att Plants N = No, Not Recomumended for taptementation et All Plants-P = Potentiet, laptement if Ptont Specific Evetuntion Werrants' TRP009A/12
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m. l-EU 028 R ev. 1 l Page 21 of 71 TABLE I l PROPOSED MINIM.M PARANETER SET FOR OYSTER CREEK l Recomumended By
-t I
NRC BWROG GPUN Parameter Description Y/N Cat Y/N Cet Y/W Cat Emptenation D-17 Containment Spray Y 2 Y 2 Y 2 Primary indicator of the proper operation Heat Exchanger Outlet of the conteinnent sprey system's function Tegerature of delivering cooling water spreys to the i drywelt/ torus that is necessery for leng-time decoy heet removet.during e LOCA. D-18 Emergency Service Water Y 2 Y 2 N - Verification of proper operation of the con-Temperature to Containment toinment sprey NX is obtained from conteinment Sprey Nest Exchanger sprey system flow, tegerature date and ESW flow. Not in Table II because of CPUN { position. 0-19 Emergency Service Water Y 2 Y 2 Y 2 Ftow to Centainment Sprny 0-20 Nigh Radioactivity Y ' 3 Y 3 Y 3 In Re&este Contret Room only. Tank Levet D-21 Emergency Ventitetion Y 2 Y 2 N - Each air duct penetration to the Reector Danper Position Building is provided with a series polr of isolation dampers. The most direct and expeditious method of evolusting secondary containment integrity and Sten 6y Ges Treatment System (SGTS) performance is to monitor the [.
' Reactor Building pressure. . A negetjve pressure of 0.25 inches of water indicates that ett' isolation denpers and eccess doors in the Reactor Sullding are closed and the SGTS is '
operati,ng normally. Also see C-9.
/ Y = Yes, Recomunended for Implementation et Att Plants N = No, Not Reccaanended for Implementation et Att Plants P = Potentist, taptement if Plant Specific Evaluation Warrants -
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TR 028 Rev. 1 Page 22 of 71 TAetE I PROPOSED MINIMJN PARAMETER SET FOR OYSTER CREEK Recomumended Sv ' NRC BWROG GPUN
- Parameter Description Y/N Cat Y/W Cat Y/N Cat Explanotion i
D-22 Status of Stancby Power Y 2 Y 2 Y 2 oyster Creek has no safety grade hydroutIc - l ; /, and other Energy Sources system. Att safety functions of Instrument Inportant to Safety Air are accomplished by tocet occumulators. D-23 Turbine Bypass Vatve - - Y 3 Y 3 SWROG issue 11 of Reference 2. Position D-24 Condenser Hotwett level - -
.Y 3 Y 3 BWROG !ssue 11 of Reference 2.
D-25 SBGTS Fan end Valve - - - - Y 2 Indication 0-26 MSIV Position - - - - Y 1 See 8-10.
- l /,
0-27 Main Steam Flow - -
. - Y 3 Indication of heat toed on mein heet sink. It would verify that steem is going to condenser vice torus or through a pipe break and is etso an automatic MSIV isolation signet.
D-28 Core Spray Isolation - - -' - - Y 3 Needed to quetify core sprey flow es flow to Valve Position .the reactor vs. through the test line, relief + valve or line breek. D-29 CRD Flow - - - Y 3' CRD flow and pressure'ere regsired to essess the operability of this system to provide high pressure cootent make te and scram occumuletor supply. Y = Yes, Reconumended for taptementation at Att Pl' ants ~ N = No, Not Recomumended for taptementation et Att Plants P = Potential, Implement if Plant Specific Evatustion Warrants , tRP009A/14 p r"p ,i-#-{ f' W e Y *' i* g= y F V'" T' h= y * %-N *$ $"'swC" $e-- Ny@ - yh a,_4d"-9'ML_1m # '*Nw-g- '
TR 028 - Rev. 1 Page 23 of 71 TA8tE I PROPOSED MIN!9RM PARAfETER SET FOR OYSTER CREM Recomumended 8v NRC SWROG GPUN Parameter Description Y/N Cat Y/W Cat Y/N Cat Explanation 0-30 CRD Pressure - - - - Y 3 CRD flow and pressure are required to assess (Accumulator Charging) the operability of this system to provide high . pressure coolant make tp and scram accumulator supply. D-31 Reacter - - - - Y 1 Vacuum breaker position is useful for ~ Building to Torus ensuring the torts will perform its steam !- Vacuum Breaker Position suppression ftmetion. See B-10. D-32 Reactor Recirculation - - Y 3 Y '3 Recircutation flow is a contingency
. Flow /
variable used for ATWS mitigation. s 1 4 Y = Yes,'Recomumenc>d for laptementation at Att Plants
- N = No, Not Recomumended for Inplementation at All Plants P = Potentist, laptement if Plant Specific Evaluation Warrants TRP009A/15-y - -,4-- wi-T- -. ># -
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Te 028 Rev. 1 Page 24 of 71 TABLE I PROPOSED NINffRSI PARMETER SET FOR OYSTER CREEK Recommended By NRC BWROG GPUN 4 Parameter Description Y/N Cat Y/W Cat Y/N Cet Explanation E-1 Dryvett Area Radiation Y 1 Y 1 Y 1 liigh Range E-2 Reactor Building Aree Y 2 N - N - See 8WROG Issue 12 of Reference 2 and E-3. Rediation E-3 Reector and Turbine Y 3 Y 3 Y 3 See 8WROG Issue 13. A combinetton of instelled Building Exposure Rete sensors and portable radiation survey instruments and sanpters will be used if post-accident access is necessery. E-4 Stack Noble cas Y 2 Y 2 Y 3 The noble ses monitor for the stock effluent
' Concentration and consists of high quotity, comunerclet grade Flow Rate equipment. In the event it faits, post-accident manitoring wItI be accompiished by manuet sanpting and laborotory onetyees.
E-5 Stack and Turbine 'Y 3 Y 3 Y' 3 Building Particulate and Halogen Concentration E-6 Radiation Exposure Meters - - - - -~ .- Deleted by NRC Errata, July,1981 E-T Airborne Radiohalogens Y 3 Y 3 Y 3 and Particulate . Y = Yes, Reconumended for laptementation at All Plants-N = No, Not Recousmended for laptementation et All Plants P = Potential, Implement if Plant Specific Evaluation Worrants i TRP007A/16
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TO 028 Gev. $ Page 25 of 71 TABLE I PROPOSED MINIfRM PARAfETER SEY FOR OfSTER CREEE Recouemended By NRC SWROG GPUN Parameter Description Y/N Cet Y/N Cet Y/N Cet Explanotion E8 Plant Environs Radiation Y 3 Y 3 Y 3 E-9 Plant and Environs Y 3 Y 3 Y 3 Radioactivity E-10 Wirv3 Direction Y 3 Y 3 Y 3 E-11 Wind Speed Y 3 Y 3 Y 3 E-12 Estimation of Y ,3 Y 3 Y 3 Atmospheric Stability E-13 Reactor Cootent and Y 3 Y 3 Y 3 Delete simp semple, see teWROG 1ssue 14 Drywell Stmp Sample Use torus sample insteed. ': E-14 Drywett Nydrogen/0xygen Y 3 Y 3 Y 3 and Genne Spectrum E-15 Turbine Building Noble Y 2 Y 3 Y 3 See E-4. Gas Concentration
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- Y = Yes, Recommended for Implementation et Att Plants N = No, Not Reccennended for Implementation et All Plants
. P = Potentist, laptement if Plant Specific Evolustien Werrants TRP009A/17 m y v. ,, ord="- M-u*-- P q *p' *++.-% * * ' - .'
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Page 26 of 71 j TABLE II - INTRODUCTION , OYSTER CREEK DESIGN AND OUALIFICATION CRITERIA i Oyster Creek was designed and constructed in the late 1960's prior to the l issuance of General Design Criteria and nuclear industry standards such as IEEE ; Standards or ANSI design / quality assurance criteria documents. As these' ; criteria were issued, they were applied as appropriate during installation ~of plant modifications. The design criteria and support documentation for individual instrument loops are therefore dependant upon the age of that , equipment.
- Regulatory Guide (Reg. Guide).l.97 recommends that specific criteria documents ,
l be applied to post-accident monitoring instrumentation, particularly category l~ ! instruments. Because of the age of oyster Creek, documentation is not always ~! available or is difficult to obtain for instrumentation loops as is now required. The purpose of this. Introduction is to delineate those arean in which Oyster Creek does not comply with modern standards. Standards applicable t when those instruments were installed and current Oyster Creek criteria are identified in reference documents. ' t Table III also contains the specifics of Oyster Creek compliance with /. Regulatory Guide 1.97 recommanded design criteria. The Table III format has l' been changed to closer agreement with the format of NUREG 0737, Supplement 1. ' Scheduled hardware modifications to bring Oyster Creek into closer agreement with P.egulatory Guide 1.97 recommendations are also noted.
'~
i 1.0 EQUIPMENT QUALIFICATION A. REGULATORY GUIDE 1.97: - The instrumentation should be qualified in accordance with Reg. Guide 1.89 and the methodology described in'NUREG-0588. f E L POSITION: ; Oyster Creek was constructed prior to issuance of Reg. Guide 1.89. - Environmental qualification of equipment in harsh environments is-identified by inclusion of equipment in the OC Master List. This program was instituted in response to: requirements of 10CFR50.49 > requirements. New or newly identified Category 1 and 2 equipment will ,be qualified per the CPUN Environmental Qualification program and Integrated Living Schedule. , GPU Nuclear Corporation has established and carried out a program to ensure the environmental qualification of safety related equipment in accordance with 10CFR50.49. TRP009B/l
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I i TR 028 R v. 1 i Page 27 of 71 l 1.0 EQUIPMENT QUALIFICATION (Cont'd) During Inspection 86-08 conducted from March 24 to 27, 1986 this i program was evaluated and it was concluded that GPUN had implemented i
- a program to meet the requirements of 10CFR50.49. Furthermore, all open items have been addressed and closed out.
l The methodology and criteria for the definition of systems and components was developed bused upon regulatory guidance and included i Regulatory Guide 1.97 post accident monitoring equipment as { appropriate. l As was indicated in our letter of November 26, 1985 and during the . EQ Enforcement conference of October 20, 1988; GPUN utilized a combination of walkdowns, procurement records and original , construction information to identify EQ equipment. Components ~ . without Tag Number's, common items, such as cables, splices and terminal blocks were originally identified using procurement records, evaluating original plant construction information and modification documents. A reverification walkdown was subsequently conducted which provided reasonable confidence that common items ' used at Oyster Creek had been identified and qualified. In addition, EQ programmatic controle in the area of maintenance, procurement and modifications have been established to identify, and qualify any discrepancies. ! I ' l As such, GPUN is in compliance with 10CFR50.49 as required under Reg. Guide 1.97. / , i ne r e a on who e ranges are required '.1 extend beyond those calculated in the most severe DBA . . . c.Ymid be qualified i using . . . AN8-4.5. l , Qualification applies to the complete instrumentation channel from + sensor to display. If the instrumentation channel is to be used in computer based display . . . qualification applies . . . to and , including the channel isolation device. GPUN POSITION The entire instrument channel has been examined for the oyster Creek Reg. Guide 1.97 instrument loops. All field sensors located in , harsh environments are qualified. The.0yster Creek control room is a mild ar.eigonment. Original plant instrument loops were installed , using commercial grade cable in 1968. Thus, some cables installed , prior to February 2, 1983'were qualified to DOR Guidelines. A portion of this cable has been field verified to give reasonable assurance of qualification. The sampling method was chosen because , of the age of the existing cable and because the construction / documentation files of the late 1960's are in most cases incomplete. In more recent modifications better cable .I tracea'oility cxists.- TRP009B/2 I k
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l TR 088 Rev. 1 i Page 28 of 71 ! 1 1.0 EQUIPMENT QUALIFICATION (Cont'd) i C. REGULATORY CUIDE 1.97 ] The seismic portion of qualification should be in accordance with Reg. Guide 1.100. Instrumentation should continue to read with the I required accuracy following, but not necessarily during, a safe shutdown earthquake. CPUN POSITION: I oyster Creek was constructed prior to the issuance of Reg. Guide 1.100. The Oyster Creek FSAR identifies all instrumantation proposed as Reg. Guide 1.97 Category 1 instrumentation as Seismic Class 1. However, appropriate documentation is not available for i original plant instrumentation. The instrumentation which was subject to later modifications have in most cases documentation to support seismic Class 1 design. The original plant equipment seismic design will be reviewed using SQUG methodology where applicable, otherwise analysis or other qualification methods will be used. Since there is reasonable assurance of the adequacy of the seismic design for the original plant equipment, as it is identified in the FSAR as Class 1, the review and qualification documentation upgrade will be accomplished in accordance with the same schedule as the USI A-46 equipment. j/, 2.0 REEMNDANCY REGULATORY GUIDE 1.97 No single failure within either the accident-monitoring instrumentation, its auxiliary supporting features, or its pcwer sources concurrent with the failures that are a condition or result of a specific accident should prevent the operators from being presented the information necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following the accident. Where failure of one accident-monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant. This may be accomplished by p oviding additional independent channels of information ' of the same variable (addition of an identical channel) or by providing an independent channel to monitor a different variable that bears a known relationship to the multiple channels (addition of a diverse channel). Redundant or diverse channels should be electrically
- independent and physically separated from each other and from equipment not classified important to safety in accordance with Reg. Guide 1.75,
" Physical Independence of Electric Systems," up to and including any
- isolation device. Within each redundant division of a safety system, ,
l redundant monitoring channels are not needed except for steam generator i level instrumentation in two-loop plants. l TRP0098/3 l -
- TR 088 Rev. 1 Page 29 of 71 2.0 REDUND&NC1f (Cont'd)
CPUN POSITION oyster Creek was designed and constructed prior to the issuance of Reg. Guide 1.75. CPUN has not committed to backfitting Oyster Creek to obtain compliance with Reg. Guide 1.75. The current GPUN QA Plan requires that separation and independence of existing circuits be maintained or improved per the criteria existing at the time of the original instrument installation. (An example of nonapplicability of physical separation criteria is the Oyster Creek control Room particularly the original main control / display panels. Due to space limitations six inch (6") separation between channels is unattainable in these panels.) However, since the Oyster Creek Control Room is continually manned, and has installed fire detection and suppression devices and is supplemented by an Appendix R Remote Shutdown Panel, the risk resulting from lack of strict separation is alleviated. Redundant Control Room displays are provided for all Category 1 instruments with redundant channels. Those redundant channels that are electrically independent from each other are noted in Table III. It should be noted that independence is based upon drawings and other design documents. Hand-over-hand walkdowns of these instrument channels have not been conducted in /, preparing this report. These walkdowns are not practical due to inacesssibility of old cables and conduits. A few category 1 instruments are separated in accordance with Reg. Guide 1.75. These instrument channels were installed after the New Cable i Spread Room was built. This allowed separation of Division 1 and Division 2 circuits. Table III contains further details. 3.0 POWER SOURCE RECULATORY GUIDE 1.97: The instrumentation should be energized from station standby power sources as provided in Reg. Guide 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants," and should be backed up by batteries where momentary interruption is not tolerable. CPUN POSITION Oyster Creek station standby power sources consist of two emergency diesel generators and the station batteries. All Category 1 instruments are powered from buses that receive power from one or more of these-sources. The only instruments which require battery backup, i.e., cannot sustain a momentary loss of power, are those that use volatile software which are provided with internal batteries. The DC FSAR Section 8.3 gives a more detailed discussion of Reg. Guide 1.32 and the associated IEEE Standard 308. Homentary is thus taken to be seconds, time for tbc diesel generators to energise the vital power buses. A few instrument 4, e.g. Reactor Water Level narrow range, will operate through a sustained AC power interruption since they are powered by station batteries. TRP009B/4 1
i TR 028 f Rev. 1 , Page 30 of 71 ! 4.0 canuusL AT&11ARILITY [ REGULATORY CUIDE 1.97: The instrumentation channel should be available prior to an accident , except as provided in paragraph 4.11, " Exception" as defined in IEEE i Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating stations," or as specified in the Technical specifications. l GPUN POSITION: ! In conjunction with the BWR Owner's group (Ref. 9), GPUN takes exception to the recommendatien that the instrument channel out of service time , for the post-accident monitoring instrument should be the same as the r system which it is monitoring. The proposed SWR standard Technical specificatione require only that post-accident monitoring equipment must i be available to be operational prior to start-up. 3 41ternative instru- l ments must be available to substitute for the non-cperational Reg. Guide l 1.97 channel, Ref. 9.) 1 5.0 QUALITY AESURANCE , REGULATORY GUIDE 1.97: The recommendations of the following regulatory guides pertaining to quality assurance should be followed: Reg. Guide 1.28 " Quality Assurance Program Requirements (Design and !- Construction)" - Reg. Guide 1.30 " Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment" Reg. Guide 1.38 " Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for j Water-Cooled Nuclear Power Plants" Reg. Guide 1.58 " Qualification of Nuclear Power Plant Inspection, l Examination, and Testing Personnel" Red. Guide 1.64 " Quality Assurance Requirements for the Design of
, Nuclear Power Plants
- Reg. Guide 1.74 " Quality Assurance Terms and Definitions" l Reg. Guide 1.88 " Collection, Storage, and Maintenance of Nuclear Power
- l Plant Quality Assurance Records" l
l Reg. Guide 1.123 " Quality Assurance Requirements for Control of I Procurement of Items and Services for Nuclear Power Plants" Reg. Guide 1.144 " Auditing of Quality Assurance Programs for Nuclear Power Plants" l TRP0098/5
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l l I TR 028 Rev. 1 l Page 31 of 71 ! 1 5.0 OU1LITY assumanca (Cont'd) ! i Reg. Guide 1.146 ' Qualifications of Quality Assurance Program Audit Personnel for Nuclear Power Plants" j Reference to the above Reg. Guides (except Reg. Guides 1.30 and 1.38) is i being made pending issuance of a revision to Reg. Guide 1.28 that is i under development (Task RS 002-5) and that will endorse ANSI /ASKE ; NQA-1-1979, " Quality Assurance Program Requirements for Nuclear Power i Plants.' ' i GPUN POSITION: ! See the GPUN Operational QA Plan, Appendix C, for_the current GPUN ! position concerning compliance to these Reg. Guides. As stated in Appendix C, compliance does not imply backfitting and/or retroactive ! compliance for construction prior to issuance of the QA plan. 6.0 DISPIAY AND RECORDING ! REGULATORY GUIDE 1.97: Continuous real-time display should be provided. The indication may be ! on a dial, digital display, CRT, or strip chart recorder.
), ,
Recording of instrumentation readout information should be provided'for at least one redundant channel. If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders. Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand. Intermittent displays such as data loggers and scanning ; recorders may be used if no significant transient response information i is likely to be lost by such devices. GPUN. POSITION: l continuous real time display is provided in the Control Room for all f l Category I Reg. Guide 1.97 variables. In addition, for those Category i 1, 2 and 3 variables where direct and immediate trend or transient information is essential for operator action, continuous recording l capability is provided on redundant dedicated recorders. The only - exceptions are noted in Table III as deficiencies. A computer record, J which can be displayed on demand, exists for those variables which do , not require immediate operator action, but provide long-term trend
- l information of system performance. i Per Ref. 8, GPUN has defined which parameters will be displayed on the ,
Safety Parameter Display System (SPDS). Reg. Guide 1.97 Type A i parameters are displayed in SPDS. However, not all Category 1 instruments are on the SPDS display. , TRP009B/6
- ER 038 Rev. 1 l Page 32 of 71 (
)
i 7.0 RANGE ggGULATORY GUIDE 1.97: If two or more instruments are needed to cover a particular range, l overlapping of instrument span should be provided. If the required . sange of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should l be used. , i GPUN POSITION: i separate Reg. Guide 1.97 instruments have been installed for cases where , the sensitivity of the normal operating range instrument precludes { covering the entire recommended span. ; 8.0 EQUIPMENT IDENTIFICATION ; REGULATORY GUIDE 1.97: Types A, B and C instruments designated as Categories 1 and 2 should be ; specifically identified with a common designation on the Control Panels so that the operator can easily discern that they are intended for use under accident conditions. /* . GPUN POSITIONt ' The philosophy of the EOPs is to provide symptomatic guidance for the operator for a wide spectrum of events including multiple failures and operator errors and for the operator to utilize whatever instruments are available to determine plant conditions. Since it is not possible, in advance, to determine all specific failure modes of instruments under all possible conditions, no distinction between R.G. 1.97 and non-R.C. 1.97 instruments is made in the EOPs. In fact, doing so could inadvertently mislead the operator into relying (unjustifiably) on an erroneous R.G. 1.97 instrument when not warranted by plant conditions. . Therefore, in order to preserve the operational flexibility present in , the ECPs, no specific identification of R.G. 1.97 instruments will be made. I < l 9.0 JNTERFACES l REGULATORY GUIDE 1.97: The transmission of signals for other use should be through isolation devices that are designated as part of the monitoring instruments and that meet the provisions of this document. l l GPUN POSITION: ( Dedicated isolation devices are used for Reg. Guide 1.97 instrument loops that input the new plant computer. Some original instrument channels at oc employ fuses. A further discussion of isolation criteria employed at OC can be found in a Generic Specification, Reference 10. TRP009B/7 1
l TR 028 ] Rev. 1 l Page 33 of 71 l
.0.0 SElLVICTM02 TESTING, AND CALIBRATION REGULATORY CUIDE 1.97: .
servicing, testing, and calibration programs should be specified to ; maintain the capability of the monitoring instrumentation. If the j , required interval between testing is less than the normal time interval J between plant shutdowns, a capability for testing during power operation should be provided. Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the ' access to such removal means. - The design should facilitate administrative control of the access to all l setpoint adjustments, module calibration adjustments, and test points. i Periodic checking, testing, calibration, and calibration verification , < should be in accordance with the applicable portions of Reg. Guide. ? 1.118, " Periodic Testing of Electric Power and Protection Systems," pertaining to testing of instrument channels. (Notes Response time [ 1 testing not usually needed.) The location of the isolation device should be such that it would be I accessible for maintenance during accident conditions. cPUN POSITION: . . f.~' Testing and calibration for Reg. Guide 1.97 instruments is included in the overall oyster Creek instrument calibration program. sufficient access and capability to perform tests and maintenance during power operations exists. These include means to remove channels from service, administrative means to control such removal and test points. Checks, calibrations and testing are in accordance with station procedures, not j necessarily Reg. Guide 1.118. 3 11.0 EUMAN FACTOR 5 PEGULATORY GUIDE 1.97: The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctiening components or modules. The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications potentially confusing to the operator. Human factors analysis should be used in determining type and-location of displays. s TRP009B/8
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TR 028 ; Rev. 1 , Page 34 of 71 i To the extent practicable, the same instruments should be used for l accident monitoring as are used for the normal operations of the plant i to enable the operators to use, during accident situations, instruments l with which they are familiar. } i GPUN POSITIC!is Human Factors analysis is used in determining the type, design and ; i location of displays for Reg. Guide 1.97 variables. This analysis included a task analysis of the Emergency operating Procedures to ensure ,! that the ranges and precision of the available instrumentation is adequate for the procedures, and that the information and controls , needed to implement the procedures are available to the operator. To ! the extent that instrument ranges allow, the same instruments used for accident monitoring are used for normal operations of the plant. 12.0 DIRECT MRkBUREME3fT REGUI.hTORY CUIDM 1.97: To the extent practicable, monitoring instrumentation inputs should be i from sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by' analysis to Provide unambiguous information.
, GPUN POSIllggs oyster Creek Reg. Guide 1.97 instruments employ direct measurement.
Parameterr. that do not have dedicated sensors and are calculated using
- other measurements are unambiguous or not needed for verification of safety function. ,
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*- 3 I
. - - . - - . , , ,,. . . . . - - - - ~ - . , + - - ,w, . . . ~ - , . . , , , , , - . , ~ . ,
l TO 028 Rev. 1 Pege 35 of 71 TABLE II COMPLIANCE STaitlS OF MINteRM PARAETER SET FOR OTSTER CREEK EWyt#0NMENTAL SEISmtc SUALITY 0.C. REG. Eft 0E PCMER CR VARIASLES SUALIFICATI04 OUALIFICATI(M ASSURANCE REDUNDANCY RANCE RANGE SUPPLY DISPLAY (NUTE 2) (NOTE 3) (NOTE 1) A-1 RPy Pressure Comply See intro- Cesgdy 0-1500psig See intro- N/A See Tatde Yes (Category 1) duction ductten III A-2 RPV Weter Levet Comply See Intro- Comply See intro- -144" TAF N/A See Table Yes (Category 1) duction duction +185* TAF tit [. (See Note 4) A-3 Torus Water Temperature Couply Comply Compty See Intro- 40*-240*F N/A See Tetde Yes (Category 1) duct ten iII A4 Torus water Level Comply See Intro- Couply See Intro- 10*-360" N/A See Tatde Yes (Ca*egory 1) duction duction inches III a A-5 Drywatt Pressure Comply Comply Comply See Intro- 0-260 psig N/A See Tetde Yes duction III
-A-6 Orywell Combustible Comty Cgly Cosgdy 0-302(N 2 IP'I' See Intro- "#" I" I'Id' Y
Gas Concentretion duction 0-251(02 ) 131 (N , 0 I 2 2 (Category 1) n TRP0090/1
-- e i e ,- e-, ,- . , , , - , we- + = m.,- , , --,mww , ~#., e + . . - - , , , . . , , . ,%-.---. .E
TO 028 4 Rev. 1 Page 36 of y1 TA8LE II COMPLTANCE STATUS OF ItININUM PARAiETER SET FOR OYSTER CREEK i ENVIRONNENTAL SEIS8 TIC SUALITY 0.C. REG. G710E POIER CR ; VARI A8tES QUALIFICATION OUALIFICATION ASSURANCE REOUISANCY RANGE RANGE SUPPLY DISPLAY ' (NOTE 2) (NOTE 3) (NOTE 1)
/.
8-1 Neutron Ftta - - - - - - See TABLE III - ------- ------ ---- 10' -100 ' --------- Percent Pouer 8-2 Control Red Position N/A N/A N/A N/A Cogty Futt in or N/A Yes , (Category 3) not fait in 8-3 RCS Sottbte Soren N/A N/A N/A- N/A Comply 0-1000 ppm N/A N/A Concentration (Sanple) (Sample) (Category 3) i 8-4 RPV Water tevet - - - - - - - - - - See A - - - - ------------------- Core Segiport See A-2 See A-2 ' (category 1) ptete to center tine of MSL 8-6 RPV Pressure - - - - - - - - - - See A - - - - ------ ------------ 15pele- See A-1 See A-1 (Cetegory 1) 1500peig 8-y Drywell Pressure Compty See Introduction Comply See Intro- 0-260 0 to Design See Table Yes /- (Category 1) duction psig' Pressure III 8-8 Orywett Sump Level N/A N/A N/A N/A N/A 9etteur to N/A No " l (Category 3) /. top 8-9 Primary Conteirament '
- - - - - - - - - - See A - - - - ------------------- 10pele- See A-5 See A-5 Pressure (Category 1) Design f.
- - Pressure TRP009C/2 4
d
, , , . - -c , - -.-g ,- , --,9 e- -- . -.r.-- , . - , .+% - + . . .--,..-.w. s,v.-3.,,,e.y . . - , - , e -3 , - . - , , . - ,
10 028 Nev. 1 Pege 37 of 71 TABLE II COMPLIANCE STATUS OF 9eintetst PAIUWETER SET FOR OYSTER OtEEK ENVIROWNENTAl. SEISetIC SUALITY 0.C. NEG. GU10E POIER CR VARIA9tES GUALIFICATION OUALIFICATIG8 ASSIMANCE PEDUWANCY WANGE N SW7tY DISPLAY (NOTE 2) (NOTE 3) (NOTE 1) 8-10 Primary conteltummt Compty Sae introduction Closed - See Intro- Closed- Closed - See Note 10 Yes Isoletion velve Not Closed duction & Not Closed Not Closed Position (Excluding Note 14 Check vetves)
/,
(Category 1) 8-11 Reactor Water Levet tempty N/A Comply N/A 0-600*F N/A N/A Yes Reference Leg Temp (Category 23 N-12 Torus Pressure Compt.y N/A CWy N/A 60 pel N/A N/A Yes (Category 2) O e TRPOO9C/3
% .- , c w c ,m = . e -m - - e e+- = a . w.-. . .e +e,-~w- - - -.* , . - + , - - -#_ , ,-w .- ,- .
TR 028 Rev. 1 Page 36 of 71 TABLE II CtMPLIANCE STATUS OF RINI8RM PARMETER SET FW OTSTER CREEK ENVIRouMENTAL SEISMIC SimLITY 0.C. REC. Eff9E PERER CR vnRIAeLEs 004LIFICAfl04 OtmLIFICATION ASSURAuCE REDUWANCY RANCE RANE MY DISPLAY
]/.
C-2 Anotysis of Primary N/A N/A N/A N/A Comply 10 tCI/Wt to N/A N/A , Cootent (Sample) (semple) 10 Cf/M I (Gausne spactrism) (Category 3) C-4 Drywel1 Aree Nigh N/A N/A N/A N/A 1-10 R/hr 1R/h- N/A ves /. Railotion (Category 3) 10'R/hr C-5 Drywett prain s m W/A N/A N/A N/A see setteerto W/A Alone is Levet Tabte I top provided !' (Cetegory 3) In Reessete (Idantified and CR thidantified Leskoge) O e TRP009C/4
TQ 028 Rev. 1 Page 39 of 71 TABLE II CONPLIANCE STATUS OF 8t!N!8Ut PARINETER SET FOR OYSTER CREEK ENv!ROWNENTAL SEISMIC NUALITY 0.C. REC. GtfiDE POWER CR VARIA9tES OtmLIFICATION OUALIFICATION ASSURANCE NEDUNDANCY #ANCE RANGE SUPPLY DISPLAY C-6 Conteirummt Ef flumt N/A N/A N/A N/A tempty 10-6,g,fc,, ,f , ,,, Radioectivity-Noble 10 uCl/cc Geses (Cattgery 3) C-8 Ef flumt Radioactivity N/A N/A N/A N/A Comply 10-6uCl/cc- N/A Yes Noble Geses 1[uti/cc 4 (Category 3) C-9 Rx Bldg. Pressure N/A N/A N/A N/A Comply N/A N/A Yes (Category 2) C-10 Rx Stdg. T6gerature N/A N/A N/A N/A Cgy N/A N/A Yes (Category 3)
TR 028 Rev. 1 Page 40 of 71 TABLE II CtpFLIANCE STATUS OF 8tist* Epi PARfuETER SET FOR OmER rww 1 EWWIRONMENTAL set 9ttC EUALITY O.C. RES. EU19E PtNEW Of VARIA9tES SUALIFICATION SUALIFICATION ASSURANCE REDURANCY RANEE RANK StPPLY DISPLAY (#0TE 2) (NOTE 3) (NOTE 1) 0-1 Mein Feeduster Flees N/A N/A N/A N/A Comply 9-119E design m/A Yes (Cetegory 3) fleet D-2 Cendmsete Storage w/A m/A N/A N/A Setteus to top m/A 0-43 ft. Yes - /- Tar 4t Levet (Cetegory 3) D-4 Dryesatt Atnesphere Comply N/A Cesgd y Coupty 0-600*F 40*F- Comply !* Yes Temperature
&&C*F (Cetegory 2) 0-T Vetwe Monitoring Systent (VMS)-
Vessel Pressure netlef EMRV'S (5) Comply m/A tempty N/A Closedf Closed-not n/A Yes [. (Cet. 2) met closed etened, or 4 0 to 58 pois Co
- Sofety (16). W/A N/A N/A N/A Cleoed/ cleoed-not s/4 Yes Velm (Cet.3) mot closed closed, or 0 to 50 psis TRP009C/6
+- - w .o_. er'- y *
.w--g y e er ***a- w- w w--N v t- -y .e, -g-* .- y* g=-ge -g-e, -*w<,-w-e p. w-+---- mm-t-we*-+w-- -'we+-- --y-
ft 028 Rev. 1 Page 41 ef 71 ' TABLE I! > COIPL!aNCE STATUS OF 9tlWI8R5t PARApETER SET FOR CTSTER CREEK f EWFIROWEEWTAL SEISet!C SUALITY O.C. REG. tRf10E PCMER CR VARIAOLES SUALIFICAff04 OURLIFICATION ASSURANCE REDUISANCY RANGE #ANE StrPLY DISPLAY (WFE 2) (soft 3) (80TE 1) t D-8 Isoletion Cendmser Co mty W/A Cagly W/A 0-10 ft. Top to Comply Yes ' System Shett-Side meter betteus [ tevet (Cetegory 2) 0-9 Isoletion Con &nser Comty W/A Comty N/A OperV Opmer See note 10 Yes Syst s volve Position Closed Closed (Cetegory 2) l D-12 Core Spray Systm CMy W/A Cegly N/A 0-5000 0-110E Cegty Yes ' flew spo desi p flow (Category 2) 0-14 SLCS Flow N/A W/m w/A N/A See Tebte 0-110E p/A Yes (Cetegory 3) desi m flow III (IfuHrect) f, Onty fIow/ No flow Ind. D-75 StCS Storage Tank W/A. W/A N/A N/A Cogty 90ttom to N/A Yes /, tevet top (IP-44) (Cetegory 3, sea .
, Note 15)
D-16 Conteirvent Sprey Comply N/A Ce gly 5/A 0-7500 0-110E Ceeply Yes Flow p desi p flew (Category 2) [ 0-17 Centeirvent Sprey Maet Cogty N/A Comply W/A 0-500*F. 32-350*F Cagly Yes Exchangar Outlet Te g. (Cetw 2) O TPr097C/T
.t-- 4 +gr+_ 7- w-,y.9. g- g,g y v3 am w. & W m
- y wrwv G.sy-9*-ey-sg cy -r -grc e w y1g = +wgy- = ,se a , m spy._g_y ,i - p sp,, a- * ,w,.eg, .y--..-. y
TR 028 Rev. 1 Pege 42 of T1 18_LI 8 l! CSPt!ANCE STATUS OF NtWISEft PAANETER SET FOR OfSTER CREEE ENVI#0muEWTAL SEISettC OUALITY 0.C. RES. GUIDE PtBER CR 4 VARIAstES SUALIFICAT! tut SUALIFICATION ASSURANCE WEDUWAWCT RAWEE RauGE SUPPLY DISPLAY (sofE 2) (NOTE 3) (NOTE 1) l-0-1? Emprpmey service ------------- See Tebte 111 - - - - - - - - - - - - - - - - 100E s/A no Ueter Flou to design flour Centeirwumt Sprey #*et Exchangers (Cetegory 2) D-20 nigh medieectivity m/A m/A s/A s/A Cogty Top to e/A trafication tiottid Tank Levat bottom in Rothseste (Category 3) CR D-21 Esprgency Ventitetion ------------- Deleted Per Tebt e ! - - - - - - - - - - - - - - - - - - - - - - - - - - - - ----- Dauper Position 7 D-22 Stetus of Sten & y Power Co gty N/A Cogty N/A Conty Voltages - N/A Yet and Other Energy Currents Sources taportant to Pressures Sefety (Cetegory 2) 0-23 Turbine sypess Vetve w/A u/A s/A s/A open/Close w/A N/A Tes Position , (Cetegory 3) D-24 Cor+nser Motwall W/A W/A N/A N/A 0- @ N/A N/A Yes tevet (Category 3) TFr009Cfs r -- m w,y u ew g-- - p - = - - + - - ,y*3 6 -w - q g- m- ---vme,-w a e.r- ye , c ,+e y ,,e3 e p-%3 m , , - a % s --s.,-. ,w. = w -
TR 028 Rev. 1 Pope 43 of 71 . TABLE 11 C%DIPLIANCE STATUS OF 3t151849t PinfWETER SET FOR OYSTER CREEE ENVIROWNEWTAL SEISmit tumLITT O.C. WEG. GUISE POMER CR VARIA9tES SUAL1FICAT10m eumLIFICATION ASMIRANCE sEDUteAuCT #AeGE WANGE 27 PLY 8ISPLAY (wcTE 2) (WOTE 3) (acte 1) D-25 S8GTS Fen and Velve Comply W/A Czuupty W/4 OrVOff N/A N/A Yes Indicetion (F es) (Cet* gory 2) OpmrClosed (Volves) 0-26 MstV Position ---------------------- See 8 - - - - - - - - - - - - - - - - - - - - - - - - - - !- (Cetegory 1) 0-27 Mein Steem Flow N/A 6 N/A N/A N/A 0-4x19 ,f , ,f , ,,, (Cetegory 3) ItsW9 r 0-28 Core Sprey Isoletion N/A W/A N/A N/A Opm/Close N/A N/A Yes-Velve Position (Category 3) D-29 CRD Flow W/A N/A W/A N/A 0-100gpo e/A s/4 Yes (Category 3) D-30 CRD Pressure N/A W/A W/A N/A 0-2990psig W/A N/A Yes (Acessuuteter Chorging) (Category 3) D-31 Rs stdg. to Torus ---------------------- See 8 * - - - - - - - - - - - - - - - - - - - - - - - - - /, 17eevtse Srecher , Position (Aterne) (Category 1) D-32 Rx Recirculation N/A N/A N/A N/A N/A N/A N/A Yes Flow (Cetagory 3) TRP00?t/9 ,,i ..r v, -
- .~. - m- e- ~ m., ,y . . , , , < - = ,_ym-- - ,i, , -, e %- --,-ew , . , . - ~%__e,--. -+y,, ,,
TR 028 Rev. 1 Pege 44 of 71 TABLE II COMPLIANCE STAftf5 0F Wistutm PARARETER SET FOR OfSTER CREEK ENV!WoumENTAL SElsstIC SUALITY 0.C. REG. ERff9E PolEt CR WARIASLES OUALIFICATION OUALIFICATION ASSURANCE REDUNDABCY RAsEE RANGE StPPLY DISPLAY (NOTE 2) (NOTE 3) (soff 1) E-1 Dryw it Aree Redietion Comply Ceeply Comply See introduc- 1-10 it/hr-10 See Tebte Yes Migh Renga tien R/ter III (Cetegory 1) E-3 Reector and Turbine n/A n/A m/A m/A note 12 10 m/hr m/A Yes l/. Sullding Exposure to TO R/hr Rete (Category 3)
-6 E-4 Steck noble Ges W/A N/A N/A N/A Comply 10 uci/cc- N/A No /.
Concentration and ate 7 to ucl/cc-Flew mate 0-110E vent (Cetegory 3)
- sips flem E-5 Stock and Turbice N/A W/A N/A N/A Comply 10' acf/cc- m/A N/A 2
suftding Perticutete (sempte) 19 uct/ec and Netoym Concentretion 0-199E vent (Cetegory 3) (se gte) desipt ftow E-T Airterne mediohstogens N/A N/A N/A N/A Comply- 19 utt/cc- N/A W/A and Petticutetes (Portable 19 utf/cc Portable Sempting Sampte) (Category 3) E-8 Plant and Environs N/A N/A N/A - W/A Comply 19' R/hr- 9/A N/A Redietlen (Portable (Portable 10 R/hr instrumentetion) Serpte) Photons
. (Category 3) 19' m/hr I
19 R/hr FTTA radietten and tow ener w Photons tRP009C/10
' ^ "
it 028 Rev. 1 Pege 45 of 71 iA8tE II COMPLIANCE STATUS OF 881415U4 PARfmETER SET FOR OYSTER CREEIC ENVIRONNENTAL SEISutC NUALITY 0.C. REG. GUIDE POMEN CR VARIA9tES SUALIFICATION NUALIFICATION ATSURANCE REDUlieANCY RANCE RANGE SUPPLY DISPLAY (NOTE 2) (NOTE 3) (NOTE 1) f, E-9 Plant end Envirens N/A N/A N/A N/A Comply Instti-chemet N/A N/A Radiesetivity (Portebte (Portstrie Gemme-esy . Instrisemtetien) Semple) Spectreseter (Cat w y 3) E-10 Wiev$ Direction N/A N/A N/A N/A Comply 0-3 W N/A Yes (Category 3) i E-11 Wind Sp M N/A N/A N/A N/A Comply 0-h N/A Yes (Category 3) (67 sch) E-12 Estimetien of N/A N/A N/A N/A Cengpty *$*C- N/A Yes Atmosphn ic Stability 10*C ! (Category 3) Accuracy i per 50 meters E-13 Reecter Cootent N/A N/A N/A N/A Ceuply Grab Sesgpte N/A N/A and cryu tt Sts, (Sampte) Sample (Cetegory 3) E-14 Drywett N , 0 N/A N/A N/A N/A Comply Grab Sempte 4/A N/A 2 2 and Gegen Spectrise (Semple) (Categwy 3) (Sanple) E-15 Turbine suilding Noble N/A N/A N/A N/A Campty 4 10 uef/cc N/A N/A f, GeS to th (Category 3) uti/cc i i IRP00?C/11 c- .,-g.c,.-., y ,- , . , - v yg w
*r a pg4, - e p ..-e.:.-+ -,e-w -% - ,- a.e v., . . ., -. # __+x- _ --wew
TR 024 Rev. 1 Page 46 of 71 NOTES.FOR TABLE...II
- 1) GPU Nuclear has a documented quality assurance plan which, as a minimum, satisfies the requirements of 10CFR50, Appendix B.
GPU Nuclear's quality assurance plan has been reviewed and approved by the NRC. Any instrumentation which has to be replaced because of technical inadequacies (performance or environmental qualification) will be procured per the requirements of GPU Nuclear's quality assurance plan, which may or may not include the quality assurance requirements specified in USNRC Regulatory Guide 1.97, Revision 2. The instruments in compliance with technical and environmental qualification requirements but lacking quality assurance documentation shall be considered on a case by case basis and may be used for compliance with R.G. 1.97 quality assurance requirements. Lack of quality assurance documentation for existing instrumentation will not per se require any modification to be performed at Oyster Creek.
- 2) Equipment located in a mild environment is not included in the scope of /,
the E.Q. rule, 10CFR50.49(c)(3).
- 3) The seismic qualification of current post-accident monttoring instrumentation complies with the oyster Creek licensing basis as stated in the FSAR. Equipment upgraded to satisfy range and environmental qualification criteria will be evaluated for seismic considerations.
Where documentation for seismic qualification of existing equipment does not exist, the equipment will be considered on a case by case basis. GPUN is developing a program of seismic equipment qualification dependent upon resolution of Unresolved safety Issue A-46, seismic Qualification of l f. Equipment in Nuclear Power Plants. Also see Table II - Introduction. I TRP009D/1 i
l, i
'
- TR 028 i Rev. 1 l Page 47 of 71 (Notes fer Table II) (Continued)
- 4) The Reg. Guide 1.97 reactor water level system consists of two inde-pendent systems to monitor the entire range of reactor level. These systems have redundant indication in the control room. Qualified water I j' level instruments indicate water level from five inches below the bottom of the fuel to five inches below the isolation condenser steam lines. ,
This is the DC limiting condition for high reactor water level control. ,
- 5) Deleted l /-
- 6) The drywell atmosphere temperature is monitored at two different l j*
locations in the drywell. Each channel is recorded in the control room. I c 0
- 7) GPUN takes exception to the high range requirement of 10 uci/cc at ,
oyster Creek. l
- 8) Deleted
- 9) Deleted
, /.
- 10) Deleted
- 11) Deleted
- 12) Worst case maximum ranges of type E-3 variables will be addressed on a case-by-case basis rather than using an arbitrary range of 104 R/hr.
- 13) Deleted l/
- 14) Redundancy is given by one indicator on each of the two drywell isolation valves.
- 15) GPUN takes exception to application of Category 2 design criteria to the Standby Liquid control System level indication. As stated in the BWROG ,
report, Issue los-
- 1. The design basis of the Standby Liquid System assumes no concurrent loss of coolant accident, therefore the working environment of the-SLCS is mild. This is in agreement with the GPUN position as stated in Ref.*5, the RQ Master List. /,
- 2. The SLCs design basis recognizes that the system has less importance to safety than the RPS or ESF systems.
- 3. The ATwS rule, 10CFR50.62, does not require qualification of the SLCS motors and pumps. At oc these are in the vicinity of the SLCS tank level indication circuits. ,
For the above reasons, GPUN feels that the SLCS tank level indicator is - correctly graded Category 3. i TRP009D/2
a . TR 028 Rev. 1 Page 48 of 71 TABLE III Reg. Guide 1.97 Instrument List and Actions to Satisfy Reg. Guide 1.97 Requirements for OCNGS Minimum Parameter Set NOTE: Table III contains only Category 1 and Category 2 parameters. Category 3 instruments are commercial grade instrument channels and do not require the quality assurance and design criteria listed in Table III.
/.
The deficiencies listed in Table III that are undergoing evaluation will be resolved by either: 1) inclusion of a modification into the DC Integrated Schedule or 2) providing a justification for the deficiency. A follow-on submittal concerning these resolutions will be transmitted as discussed in the cover transmittal letter. 1 i i ( TRP009E/1
TR 028 Rev. 1 Page 49 of 71 TARLR III variables RPV Pressure RG 1.97: Category 1 A-1, B-6 RG 1.97: Type A, 8 Sensor Tag Number PT-55-IA0092A, PT-56-IA00928 Display type and Analog Indicators: Tag Number PI-622-0849, PI-622-0950 Instrument Range 0-1500 peig Environmental Qualification Included in 50.49 Program Master List Seismic Qualification Included in CPUN SQUG Program Quality Assurance GPUN QA Operational Plan Redundancy and Redundant sensors in Reactor sidg. 51'-3" Locatione Instrument Rack RK03
/.
Power Supply Diesel Generator 1 and 2 through RPS Panels 1 and 2 (See deficiency $1 below) Location of Display control Room Panel SP/6P Computer Point None Deficiencies:
- 1. The RPS power supply will be lost on a Lose of Off-site Power event until the RPS motor generators are manually restarted. This effect was discovered ,in 1989 during a Loss of off-site Power test. This power supply deficiency will be corrected per the OC Integrated Schedule pending investigation of alternative modifications to existing power supplies.
- 2. Currently only narrow range reactor pressure is recorded on a strip chart recorder. Installation of wide range pressure recorders to meet Category 1 criteria is'being evaluated.
Comments:
- 1. Wide range reactor pressure is displayed in the SPDS. The two 0-1600 poig transmitters are not environmentally qualified. These transmitters are part of the Balance of Plants foodwater control system.
TRP009E/2 , j
- TR 028 Rev. 1 Page 50 of 71 TABLE ...I II Variables RPV Water Level (Fuel Zone) RG 1.97: Category 1 A-2, B-4 RG 1.97; Type A,8 Sensor Tag Number DPT-4-IA0090A, DPT-6-IA0091B DPT-5-IA0091A, DPT-7-IA00908 Display Type and Recorders LR-1-IA00908 Tag Number Analog Indicators LI-1-IA0094A, LI-2-IA00945 Instrument Range -144" to +180" TAF { Top of Active Fuel)
Environmental Qualification Included in 50.49 Program Master List Seismic Qualification Included in GPUN SQUG Program or A-46 Methodology Quality Assurance CPUN QA Operational Plan r j, Redundancy and Redundant sensors in Reactor Bldg. 51'3" Locations Instrument Rack RK03 Power Supply Diesel Generator 1 and 2 through RPS Panels 1 and 2 Location of Display Control Room Panel SF/6F Computer Point 1813.0, 1813.1 Defielencies
- 1. The RPS power supply will be lost on a Loss of Off-Site Power event until the RPS motor generators are manually restarted. This effect was discovered in 1989 during a Loss of Of f-site Power test. This power supply deficiency will be corrected per the OC Integrated Schedule pending investigation of alternative modifications to existing power supplies.
- 2. The pressure transmitters' power supplies are not independent from Balance of Plant transmitter loops. Installation of a dedicated Fuel Zone power supply is being evaluated.
TRP009E/3
. TR 028 Rev. 1 Page $1 of 71 TARLE III variable RPV Water Level (Fuel Zone)
A-2, B-4 (Cont.)
- 3. Only one of two Fuel Zone RPV Water Level channels is currently recorded on a strip chart recorder. Installation of reactor water level recorders to-meet Category 1 criteria is being evaluated.
Comments:
- 1. During a Loss of Off-Site Power event, Fuel Zone microprocessor software will be preserved by internal batteries until power is restored.
- 2. The two widest range Fuel Zone level transmitters share a common reactor vessel penetration. However, they are redundant due to a " pipe within a
/.
pipe" arrangement. This commonality does not affect the top 125 inches of ' Fuel Zone instrument range for which separate vessel penetrations and transmitters are provided. Modification to correct this problem is impossible since no other reactor vessel instrument penetration exists that can monitor this low level.
- 3. The Fuel Zone microprocessor detects measurement errors due to reference leg flashing, turbulence due to core spray or recirculation flow, over-ranging and faulty analog inputs. When these conditions are detected, the microprocessor will either shift to a wider range or turn OFF the display so as not to mislead the operator.
TRP009E/4
. . TR 038 Rev. 1 Page 52 of 71 i
TARLE III Variables RPV Water Level (Narrow Range- RG 1.97: Category 1 - A-2, B-4 Yarway) RG 1.97: Type A, B i Sensor Tag Number LT-RE-0005/19A/B Diaplay Type and Analog Indicators: , Tag Number LI-RE-0021A/B, ' LI-622-1640, 1634 Instrument Range 85" - 185" TAF (top of Active Fuel) Environmental Qualification Included in 50.49 Program Master List Seismic Qualification Included in GPUN SQUG Program or A-46 Methodology Quality Assurance CPUN QA Operational Plan ; Redundancy and Redundant sensors in Reactor Bldg., Locations Instrument Racks RK01, RK02 RB 51'-3" Elevation !, T Power Supply Diesel Generator 1 and 2 through RPS I panels 1 and 2 or Sattery B and C Location of Display control Room Panel SF/6F,-18R/19R Computer Point 1808.1, 1808.0 r Deficiencies
- 1. The narrow range Yarway reactor water level instrumentation is not recorded on a strip chart recorder. Installation of reactor water level recorders to meet Category 1 criteria is being evaluated.
t Comments:
- 1. The RPS power supplies are supplemented by DC to AC converters which_ supply 115V AC from statior. batteries during a Loss of Off-site Power event. This +
! happens prior to the RPS panels being re-energized by the Diesel Generators. l TRP009E/5
- TR 028 Rev. 1 Page 53 of 71 IABLB.III variable: Torus Water Temperature (Bulk) RG 1.97: Category 1 A-3 RG 1.97: Type A Sensor Tag Number TE-664-0030A/B, Display Type and -0031A/B, Tag Number -0032A/B,
-0033A/B,
-0034A/S,
-0035A/B Digital' Indicator TI-664-0042 A/B,
-0043 A/B Instrument Range 400- 240*F Environmental Qualification Included in 50.49 Program Master List !*
Seismic Qualification Comply Quality Assurance GPUN QA Operational Plan Redundancy and Redundant sensors inside thermowells about Locations the Torus Room Power Supply Diesel Generator 1 and 2 through RPS panels 1 and 2 or Battery B and C Location of Display control Room Panel 1F/2F, 18R/19R Computer Point X00674, X00675 Comments:
- 1. The RPS power supply is supplemented by DC to AC converters which supply 115V AC from station batteries during a Loss of off-Site Power event. This happens prior to the RPS panels being re-energized by the Diesel Generators.
TRP009E/6
TR 028 , Rev. 1 i Page $4 of 71 TARLK_III f Variable: Torus Water Level RG 1.97: Category 1 , A-4 RG 1.97: Type A sensor Tag Number LT-0037, LT-0038 ; Display Type and Digital Indicator : Tag Number LI-243-0002 A/B , Instrument Range 10" - 360" , Environmental Qualification Included in 50.49 Program Master List Seismic Qualification Included in CPUN SQUG Program or A-46 Methodology Quality Assurance GPON QA Operational Plan Redundancy and Redundant sensors in Reactor Bldg. ' Locations Torus NW Corner Room Elevation 19'-6" ; Power supply Diesel Generators 1 and 2 through RPS Panels 1 and 2 (See Deficiency #1 below) Location of Display Control Room Panel 1F/2F, 16R , computer Point X00047, X00048 Deficiencies >
- 1. The RPS power supply will be lost on a Loss of Off-Site Power event until .
the RPS motor generators are manually restarted. This effect was ; discovered in 1989 during a Loss of Off-Site Power test. This power supply deficiency will be corrected per the OC Integrated Schedula pending investigation of alternative codifications to existing power supplies. ; t TRP0092/7
TR 038 Rev. 1 Page 55 of 71-ThBLE III Variable Drywell Pressure RG 1.97: Category 1 A-5, B-7, B-9 RG 1.97: Type A, B Sensor Tag Number PT-0053, 0054 Display Type and Two pen recorders: , Tag Number PR-0053 PR-0054 l
\
Instrument "L. ys 0-260 psia l Environmental qualification Included in 50.49 Program Master List
-i
) Seismic Qualification Included in GPUN SQUG Program or A-46 Methodology
- s. >
Quality Assurance GPUN QA Operational Plan Redundancy and Redundant sensors in Reactor Bldg. 51'3" Locations
.c Elevation / _j
.i 1
Power Supply Diesel Generator 1 and 2 through RPS-panels 1 and 2 (See Deficiency #1-below) f j- Location of Display control Room Panel 16R Computer Point XOOll7, X00ll8 -j Defielencies: i
- 1. The RPS power supply will be loot on a Loss of Off-Site Power event until the RPS motor generators are manually restarted' This effect was l
discovered in 1989 during'd Loss of off-Site Power test. This power' supply deficiency 'will be corrected per the OC Integrated. Schedule'pending investigation of alternative modifications to existing power supplies.-
- 2. The drywell pressure indicators are-limited accuracy. recorders located on a Control Room back panel. Upgrading the front panel containment pressure
' control instrumentation is scheduled and also best accomplished.during the- i containment spray automatic initiation modification.. This is due to space constraints at the containment spray station, Panel 1F, and uncertain ,
requirements of a manual containment spray configuration. Currently, OC uses non-Reg. Guide 1.97 pressure switches to automatically start the s containment spray pumps on high drywell pressure. 1
;1
+
i
.TRP009E/8
* . TR 088 l Rev. 1- t Page 56 of 71, TABLE III ;
'1 variable: Drywell Combustible Gas RC 1.97: Category 1 'i (Hp , 02 ) Concentration RG 1.97: Type A, C. '
A-6 Sensor Tag Number IT-0001A/B Display Type and Recorder AR-0001, 0002 Tag Number Analog Indicator Panel 16R' Instrument Range 0 - 30% (H2 ) 'I O - 25% (02 ) Environmental-Qualification Included in 50.49 Program Master List seismic Qualification Comply Quality Assurance GPUN QA operational' Plan Redundancy and Redundant sensors in Reactor Bldg. 75'3" Locations Elevation Local Panels A and B Power Supply Diesel Generator 1 and 2 through RPS , Panels'l-and 2 (See-Deficiency #1 below) Location of Display Control Room Panel 16R Computer Point X00076 (H2 ) 02480.0 (H2 ) X00077 (02 ) X00078 (02 ) 4 Deficiencies: - ~
- 1. The RPS power supply will be lost on a Loss of Off-Site Power event'until the RPS motor generators are manually restarted. -This effect was discovered in 1989 during a Loss of Off-Site Power test. This power supply deficiency will be corrected per the oc Integrated Schedule pending investigation of alternative modifications to existing power supplies.
E TRP009E/9 2
- - ~
, TR 028 Rev. 1 Page 57 of 71 TABLE III A
Variables Neutron Monitoring RG 1.97: Category 1 B-1 R3 1.97: Type B Sensor Tag Number See comments Display Type and Tag Number Comments:
- 1. USNRC Regulatory Guide 1.97 states that neutron monitoring' instrumentation should satisfy Category I design-requirements. GPUN's original position-was that neutron monitoring instrumentation should-satisfy _ category II design requirements. GPUN since has withdrawn this position and advised the NRC that we are participating in the GE BWROC activity to downgrade the design requirements for neutron monitoring.
The BWROG submitted its position in March 1989. The NRC staff responded to-this position paper on January 29, 1990. Subsequent discussion between the BWROG and NRC staff has left unresolved the question of which accident scenarios and design constraints should be applied to post-accident neutron monitoring systems. As of this writing, GPUN remains uncommitted with regard to neutron monitoring until the BWROG issue is resolved. ! Independent of the BWROG effort, GPUN has performed a technical and cost ~ evaluation of the wide range neutron monitoring system offered by General' Electric, GPUN is considering installing the GE wide range neutron l monitoring system at Oyster Creek to address operational and' maintenance 1 problems associated with the existing IRMs and-SRMs. This system, if installed, will also meet post accident neutron monitoring requirements of Regulatory Guide 1.97. The technical evaluation showed that it was-not feasible to install the Gammametrics ex-core neutron monitoring system due to physical space constraints inside the olological shield wall. 1 Further evaluation will include plant testing to determine system compatibility with the existing drywell coaxial cable penetration assemblies; impedance matching at these penetrations may make a new j wide-range monitoring system infeasible. GPUN is also evaluating the experience of Japan Atomic Power Corporation which installed.the first j eight channel wide-range neutron monitoring system in 1989. In February, l GPUN participated in a utility group investigation of a WRNM bulk purchase. The next bulk purchase meeting is scheduled.for April 1990.- GPUN is also awaiting NRC approval of the GE Licensing Topical Report on the Wide Range Neutron Monitoring system, NEDO 31439~. Approval of this document is required prior to final GPUN project commitment. TRP009E/10
i TR.038 ' Rev. l' ' Page 58'of 11 . E TABLE III ' i b variable: Drywell Isolation Valve Position RG 1.97: Category 1 B-10 RG 1.97: Type B Sensor Tag Number See Comment Section and' Display Type and next page Tag Number , Instrument Range Open/ Closed Em ironmental Qualification Included in.50.49 Program Master-List a seismic Qualification Included in GPUN A-46 Program or . - SQUG Methodology Quality Aseurance GPUN QA Operational Plan l Redundancy and Comment 1 , Locations Power Supply See next page-Location of Display control Room Panels,,llF l Comments:
- 1. Redundancy of valve position indication is provided through redundant valves that satisfy the function of containment isolation.
- 2. The list of containment isolation valves is extracted from tho'OC. Technical 1 Specifications and FSAR per GPUN memo. 5350-89-524. 4 TRP009E/11 ,
, .m , - - , , , - , -
n ._ -
. .TR 038'- ;
Rev. 1 ! Page.59 of 71 ; TABLE III Variables Drywell Isolation valve Position RG 1.97: Category 1 B-10 RG 1.97: Type B (Continued) CONTAINMENT ISOLATION VALVES VALVE FUNCTION / VALVE DESIGNATION POWER SUPPLY PANEL ! Main Steam Iselation Valves Panel DC-D/DC-F V-1-0007, 0008, 0009, 0010 (AC RPS1,2)' (NSO3A, NS03B, NSO4A, NSO4B) Reactor Building Closed Cooling Valves MCC 1B21A/1A21 j ' (V-5-0147, V-5-0166, V-5-0167) Instrument Air Valve CIP-3 and-DC-D , (V-6-0395) i Emergency Condenser Vent Valves MCC DC-1/DC-2 (V-14-0001, V-14-0005, V-14-0019, V-14-0020) i Reactor Cleanup Valves MCC 1AB2/DC-B V-16-0001, V-16-0002, V16-0014, V16-0061) !' - Shutdown Cooling Valves MCC 1AB2 (V-17-0019, V-17-0054) Drywell Equipment Drain Tank Valves VACP-1 (V-22-0001, V-22-0002) Drywell Sump Valves VACP-1 ' (V-22-0028, V-22-0029) Drywell and Torus Atmosphere Control Valves l (V-27-0001, V-27-0002, V-27-0003,:V-27-OOO4) VACP-1 (V-28-0017, V-28-0018) VACP-1: -i l (V-28-0047), VACP-1 .. (V-23-0013, V-23-0014, V-23-0015, V-23-0016) VACP-1 ! (V-23-0017, V-23-0018, V-23-0019, V-23-0020)~ ' VACP-1 (V-23-0021, V-23-0022) 'VACP-l'
~ '
Torus to Reactor Building Vacuum Relief Valves VACP-1 (V-26-0016, V-26-0018) L l l' i j TRP009E/12 i i l t
- -- i, , . . .
i I
*
- TR 028 Rev. 1 Page 60 of 71 TABLE III a
I variable: RW Level Cold Reference RG 1.97: Category-2 Leg Temperature; B-ll, D-4 RG 1.97: Type B,.D
.1 Sensor Tag Number TE-130-0450, 0451 j Display Type and l l
Tag Number Recorders TR-IA0055 j Instrument Range 0 - 600 F Environmental Qualification Included in 50.49 Program Master List seismic Qualification N/A Quality Assurance GPUN QA Operational Plan-Redundancy and N/A Locations /', Power Supply Diesel Generator 1 and 2 through ' Instrument. Panel 4-C-Location'of Display control Room Panel BR Computer Point X00009, X00010 Comments:
- 1. There is no procedure for testing the thermocouples after installation.
However, these thermocouples are analog devices which read approximately the same ambient temperature. Since they are monitored. daily and continuously trended the likelihood of an undetected thermocouple failure , is minimal. These thermocouples are in a contaminated area.inside the drywell and a significant radiation dose would be incurred during testing. For these reasons in place thermocouple calibration is not justified. l TRP009E/13
TR 028 Rev.'1 Page 61-of'71~ TABLE III variables Drywell Area Rad High Range RG 1.97: . category 1. C-4, E-1 RG 1.97: ' Type C;E i Sensor Tag Number RE-0790, 0791 Display Type and Analog. Indicator Tag Number RI-0790, 0791 1 7 Instrument Range 1 - 10 R/Hr (Rads per Hr). Environmental Qualification Included in 50.49 Program Master List seismic Qualification Comply Quality Assurance GPUN QA Operational Plan l Redundancy and Redundant sensors inside.the Drywell l Locations at Elev.-40'-0" and Elev. 47'-0* , 1 Power Supply Diesel Generator 1 and'2 through RPS Panel 1 and 2 Location of Display Control Room Panel 2R Computer Point 10001, X0002 l
'I Deficiencies: .)
.i
- 1. The RPS power supply will be lost on a Loss of Off-Site Power event until.
the RPS motor generators are manually restarted. This effect was; j discovered in 1989 during a Loss of Off-Site Power test.. This power supply ' 7 deficiency will be corrected per the OC' Integrated Schedule pending ~l investigation of alternative modifications to existing power supplies. ; 1 L 1 TRP009E/14
~i
t t-
.-- . TR 028'.
Page 62 of 71 TABI2 III Variables Reactor Building Pressure RG 1.97: Category 2 ;
-C-9 RG 1.97: Type C Sensor Tag Number .DPIT-822-1102, 1103, 1104, 1105'
~ Display Type and Analog-. Indicator:
Tag Number 11R-0010, 11R-0011 r i
~
Instrument Range -1 psig to 1 peig (proposed, Comment'1) ~ -- Environmental Qualification See Comment 2 Seismic Qualification N/A Quality Assurance GPUN QA Operational Plan Redundancy and N/A . Locations /, L Power Supply Ventilation Panels P-14, 15 .; Location of Display Control Room Panel 11R Computer Point None l Comments:
- 1. Engineering is developing a modification to refurbish this, instrument.,
This work is included in the OC Integrated Schedule for the next refueling " outage, 13R
-t
- 2. DPIT 822-1102, 1103, 1104, 1105 are totally pneumatic transmitters powered' by the plant Instrument Air System.
i TRP009E/15
TR D28 i Rev. 1 Page 63 of 71 TABLE III variables EMRV's Position Indication RG 1.97: Category 2 , D-7 RG 1.97: Type D 9 Sensor Tag Number MS-VE-0017 thru 21 e l- Display Type and Analog Indicators:- Tag Number MS-VI-0017 thru 21 P Instrument Range closed /Not closed y Environmental Qualification Included in CPUN 50.49 Program Master List Seismic Qualification N/A , L Quality Assurance GPUN QA Operational Plan Redundancy and N/A Locations Power Supply- Diesel Generator 1 and 2 through Instrument Panel 4A L ,- Location of Display Control. Room' Panel 1F/2F & 15R Computer Point 735.0, 735.1, 735.2, 735.3, 735.4' . r 1 i Comments: None TRP009E/16 ;
1 i
, JR 028' ,
Rev. 1
- Page 64'of 71 TABLE III Variable: Isolation condenser Shell side RG 1.97:' Category 2 Water Level,.D-8 RG 1.97: Type D Sensor Tag Number LT-IG-0000 A,B Display Type and- Analog Indicators: ;
Tag Number LI-IG-0007 A,3 - f s Instrument Range 0 - 10 Ft. Environmental Qualification Included in 50.49-Program Master List Seismic Qualification N/A Quality Assurance CPUN QA operational Plan Redundancy and N/A Locations Power Supply Diesel Generator 1 and-2 through Instrument Panel 4B Location of Display control Room Panel'i:/2F i Computer Point 108.0, 208.0 Comments: None
'I w
TRP009E/17
i TR 038 ' Rev. 1 Page 65 of 71 TABLE III Variable: Isolation Condenser System. RG 1.97: Category 2 Valve Position; D-9 RG 1.97: Type D Sensor Tag Number V-14-0001, 0005, 0019, 0020, Display Type and 0030, 0031, 0032, 0033, 0034, Tag Number 0035, 0036, 0037 Red / Green Lights , Instrument Range Open/ Closed Environmental Qualification Included in 50.49 Prcgram Master' List i Seismic Qualification N/A -e f. Quality Assurance GPUW QA Operational Plan Redundancy and N/A Locations Power Supply Valve Motor power supply also energizes-valve position i indication i Location of Display Panels IF/2F and 11F Computer Point 211.0, 111.0, 111.2, 211.2, 101.0, 102.0, 202.0, 201.0, 109.0, 209.0, 110.0, 210.0 211.1, 111.1, 111.3, 211.3- 1 ! ~ 101.1, 102.1, 202.1, 201.1' [ . 109.1,.209.1, 110.1, 210.1- j' i ! Comments: None l l TRP009E/18 l
~;
i TR 088. ;
.'~ ~
Rev.~1 ]
-Page'66 of 71 j e
TABLE III i
~
Variables Core. Spray Flow RG 1.978 Category 2 D-12 RG 1.97: Type D Sensor Tag Numbe'r . FIT-RV0026A,B Display Type.and Analog Indicators: Tag Number FI-RV0027A,B , Instrument Range 0 - 5000 GPM L Environmental Qualificatien Included;in 50.49 Program Master List seismic Qualification N/A Quality Assurance GPUN QA. Operational' Plan [3 Redundancy and N/A Locations Power Supply Diesel Generator 1 and.2 through Instrument Panel 4B , Location of Display Control' Room Panel 1F/2F Computer Point 00312.0, 00412.0-Comments: None r, t TRP009E/19 4 m .
. g TR 028 Rev. 1 Page 67 of'71-TABLE III Variables Containment Spray System Flow RG 1.97: Category 2 D-16 RG 1.97: Type D q
Sensor Tag Number FT-IP0003A/B-Display Type and Analog Indicators: Tag Number FI-IP0004A/B
-Red / Green Lights Instrument Range 0 - 7500 GPM, Open/Close Environmental Qualification Included in 50.49 Program Master List i
seismic Qualification N/A Quality Assurance CPUN QA Operational Pla_n-Redundancy and N/A [. Locations t Power Supply Diesel Generator 1 and;2 through IP-4B Location of Display Control Room Panel"1F/2F Computer Point 908.0, 810.0
- Comments
- 1. This instrutnent measures-containment spray flow to the.drywell.. A small portion of containment spray flow, approximately 5%,_is bypassed to the torus. This flow is indicated by the position of these. torus cooling ,
valves which are included in the EQ master' list. { V'-21-18, Containment Spray Bypass Valve V-21-15, Containment Spray Bypass Valve
.- i TRP009E/20-
TR 028 Rev. 1~ Page 68 of 71 ZARLE III , I variables containment Spray System Heat RG 1.97: Category 2 , Exchanger Outlet Temperature RG 1.978- Type D D-17 , sensor Tag Number TE-0040 B,C Display Type and Recorders Tag Number TR-IP0001 Digital Indicators-1F-TI-0002 Instrument Range 00 -500 F Environmental Qualification Included'in 50.49 Program Mast'er List Seismic Qualification N/A Quality Assurance GPUN QA Operational Plan Redundancy and N/A Locations Power Supply Diesel Generator.1 and 2 through IP-4 Location of Display Control Room Panel 1F/2F ! Computer Point 808.1, 808.0 - i Comments: None
- TRP009E/21 4
- , - - ._...,_.,.._y. ,_,, ,
.- , .TR 028 Rev. 1 ';
Page 69 of 71 TABLE III variables Emergency Service Water Flow RG 1.97: Category 2 D-19 RG 1.97: Type D Sensor-Tag Number' - Display Type and Tag Number See Comments i Instrument Range Environmental Qualification 't Seismic Qualification Quality Assurance Redundancy and Locationn . Power Supply Location of Display i computer Point l s comments:
- 1. Instrumentation does not currently exist in the control Room. Existing local annubars will be provided'with control room read out. This ,
modification is part of the OC Integrated Schedule and currently scheduled for installation during the 13R refueling outage. l l l ?: i l- TRP009E/22 l I I
o TR 028 i Rev. l' Page 70 of 71 , TABLE III variables Status of Standby Power & RG 1.97: Category 2 l Other Energy Sources; D-22 RG'1.97: Type D Sensor Tag Number -See Comments Display Type and Analog Meters Tag Number , Instrument Range "B" Battery Current 0-1200 amps, Voltage 0-150V . .
"C" Battery Current 0-600 amps, voltage 0-150V, Emergency' Diesel' Gen. #1 0-4000 KW- 3 Emergency Diesel Gen. #2 0-4000 KW Environmental Qualification Located in mild environments p
.f. .
Seismic Qualification N/A-Quality Assurance GPUN QA Operational' Plan Redundancy and N/A Locations Power Supply See Comments: , Location of Display Control Room Panel 8F, 9F L Computer Point "B" Battery Voltage 02387.1
~
"C" Battery Voltage 238104.0' "1" Diesel KW 02374.0 "2" Diesel KW. 02385.0 u
comments: 4
- 1. - Instrument channels consist of current and voltage coils with control Room readout. They are unpowered, passive circuits.
TRP009E/23 - d i
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* - TR 028 j Rev. 1 _
l Page 71 of 71 1 f TABLE III Variables SBGTS Fan & Valve Indication RG 1.97: Category 2' ! D-25 RG 1.97: Type D. t sensor Tag Number EF-1-8, EF-1-9, V-28-0023, Display Type and 0024, 0026, 0027, 0028, 0030, 0048 Tag Number Red / Green Lights Instrument Range On/off, Open/ Closed Environmental Qualification Included in 50.49 Program Naster-List Seismic Qualification N/A a [. Quality Assurance GPUN QA Operational Plan Redundancy and N/A . Locations
?
Power Supply Valve and fan motor power-supplies also energize position indications , Location of Display Control Room Panel 11R Computer Point None Comments None 9 d TRP009E/24
- a. .
ATTACIMENT *
. TOPICAL REPORT # 028, (REVISION 1)
I- OYSTER CREEK RESPONSE TO USNRC REGULATORY GUIDE 1.97 I l BWR OWNERS GROUP-1 l' Poeltion on NRC Regulatory Guide 1.97, Revision 2 1 t I I I I l ;
- l
\
l . July 1982 s __._____m.
i
-t A 18 .g l
84.:
. : v.2 J
DISCLAIMER ;
'i The positions reported herein are con .
sensus responses to the. requirements of NRC Regulatory Guide 1.97, Revision 2, December. 1980, and as such do not necessarily express in every particu , ' lar the several positions of the par-- ; ticipating utilities.-
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5
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.. i CONTENTS I[
Page
- 1. INTRODl*CTION 1-I Sponsoring Utilities 3
- 2. Sk'R OWERS GROUP POSITION STATEMENT ' S General Position Statement 5-
! Implementation of Design Changes 9 ( 3. . PROPOSED TYPE A VARIABLES 10 variables Identified as Type _A 11 ! Potential Type A Variables- 12 4 PLANT VARIABLES FOR ACCIDENT MONITORING 14-Type A Variables 16 Type B Variables 17 Type C Variables 18 d Type D Yariables 20 Type E Variables -23 ( 5. SUPPLDIENTARY ANALYSES 25' l (Issues 1 - 14) ;
't
- 6. CONCLUSIONS .
58 l APPENDIX A: THERMAL CONDUCTIVITY OF IN-CORE THERMOCOUPLES IN BOILING WATER REACTORS 63 i APPENDIX B: Sk'R VARIABLES (TABLE-1, RG 1.97) 96 APPENDIX C: ABBREVIATIONS 107 i l i l l 1
-i i
i s 1 1 111 I
l . , s. ig < g ..
- 1. INTRODUCTION
.l.
Following the publication of Regulatory Guide 1.97, Revision 2 by the U. S. Nuclear Regulatory Commission in ; December 1980, the BWR Owners Group (BWROG) established-a f comittw co' review and evaluate the. regulatory positions described therein.I i The intent of RG 1.97 is to ensure that all light-water- f cooled nuclear power plants are instrumented as necessary to [ measure certain prescribed variables.and systems during.and j after an accident. The principal purpose of the BLTOG RG 1.97., Committee was to evaluate the safety effects and the feasibil-icy of implementing the proposed-regulstory positions--particu-larly those defined in Table 1,.RG 1,97. Twenty-four (2!.) donstig and two (2) forei;;n utilities i l aupported the Committee's efforts. Seventeen (17) of these utilities provided representatives to serve on the committee.
}
A subcommittee of the RG 1.97 committee was: formed (Feb.:1982) 9 to address the issue of inadequate Core cooling (ICC) detection. Meetings of the committee. commenced in April'1982 and con-cinued c'a rough July 1982. The sponsoring utilities and their-I representatives who served on the BWROG RG l'.97 Conmittee are
~
i identified at the end of this section. The committee's work was devoted primarily to discussions -
- of specific task assignments, to presentations of-committee-and contractor-generated data related to RG 1.97 ' requirements, l and to the formulation of recommendations based on the commit- O tee's reviews and analyses. Besides conducting its,own studies, the committee contracted other analytical work to Roy l & Associates, Inc. ; S. Levy, Inc.; and the General- Electric - -l
! Company. ! s ll I As used throughout this report, RC 1.97 refers to RG 1.97,
~
Revision 2, December 1980. l- 1
~
, .co-;-o.1 4 m A --.*;m=.- ' "- N#, '
A'sumary statenent of the Owners Group position relative to RG 1.97 requirements is presented in Sec. 2; some proposed Type A variables, which are unspecified- ir. RG 1.97, are defined i in Sec. 3; a detailed Ovners's position statement on.a variable- , by-variable basis is provided in Sec. 4; and abstracts ofLthe-supporting analyses and studies are contained in Sec.'5.- Per - cinant contractor reports, a copy of Table 1 from RG 1.97,.and a list of abbreviations are presented in the appendices.
+
A ( 2
5: . Sponsoring Utilities The. sponsoring utilities of the BWROG RG 1.97 Committee, their assigned contacts or committee members, and consultants
- are identified below.
Committee Meebership (hames of the working members of-the committee are in italics.) Boston Edison Company RICH ST. ONCE: s'ERRY ?.!:E2W3RI h Cincinnati Gas & Electric Company W:LLIA:' COC?E?; ROCER THONCT ~ Cleveland Electric Illuminating Company g RA: :A:::::: Detroit Edtson Company
. ,~.; ;p5;;.
Georgia Power Company
' " ': '"r"' (ICC chairman) (from_ Southern Company Services Inc.) -
I Gulf States Utility Company MATEE RA}0!AN; Fli?LLI?S ?CRTER i l Iowa Electric Light and Power Company l
?CS?EF (FCC:) 3ALA3 (Chairman)
Jersey Central Power & Light Company l i/ANIJ CHARJOS; PAUL PROCACCI; ABDUL R. BAIG_ Long Island Lighting Company I rice!: P: ??? Mississippi Power & Light Company SAM HOBBS; EU.%'d 3RCh7;
~
Northeast Utilities MARIO BLAT lCATLCR
- q Northern Indiana Public Service Company ADAM SHAMBAZI l
Pennsylvania Power & Light Company
./OliN BARTOS; DAN CARDIN0BE Philadelphia Electric Company 175 30f TRS; RICK OGITIS I Power Authority of the State of New York G. RAi/GARA ; J. STREET l i
3
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- g.
Public Service Electric and Cas Company : RICHARD O'CONNELL : Tennessee Valley.Autherity K.\THRYN ASHLIY: ? *! i'?.* I'".!,4 fli E ; I?ashington Public Power Supply System .; ARUN JOSHI; 32D CllTI.70 TON q i Supporting Utilities ( Carolina Power f. Light Company ,
'f Centrales Nucleares Del Norte (S.A.) :
Commonwealth Edison Company : ! Ente, Nazionale'per l' Energia Electrica Illinois Power Company - Nebraska Public Power District Niagars !!ohawk Power Corporation
.krthern States !'c.ter Company I
CPRI/NSAC C. Dan t?ilkinson, program manager (replaced by Robert Kubik for - report coordination in Feb. 1982) Consultants General Electric Company ; S. Levy, Inc. -; Roy and Associates I. - 4 I; I i -
' f E .
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.l *
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- 2. IBWR OWNERS GROUP POSITION STATEMENT The BWROG position on SRC Regulatory Guide.l.97 Revision 2, is presented in the following statement. The statement reflects the intent of the regulatory positions set forth in 5
g RG 1.97 but includes alternatives and" deviations that relate to specific instrumentation requirements and to the particulars of their implementation. : The statements that follow in this section are general positions on the requirements specified in the designated para-graphs of RG 1.97. A detailed position' statement on a variable-by-variable basis is presented in Sec. 4, and supplementary 9 data are provided in Sec. 5 and in the appendices. l ; I General Positiori Statement i BWROG concurs with the intent of RC 1.97, Revision.2. The intent of the regulatory guide is to ensure thn; necessary and sufficient instrumentation exists at each nuclear power ' station for assessing plant and environmental conditions during . and followit.g an accident, as required by 10 CFR Part.50, I Appendix A and General Design Criteria-13, 19 .and 64. Imple-mentation of RG 1.97 requirements-is recommended-except.in those instances in which deviations from the letter of the i 1 guide are justified technically and when they can be'imple-mented without disrupting the general intent of the Guide. ' l Incessassing RC 1.97, the Owners Group has drawn upon information contained in several applicable documents, such
~
as ANS 4.5, NUREG/CR-2100, and.the BWROG Emergency Procedures ;
.] Guidelines, and on data derived from other analyscs and stud- l ies. The E ars Group believes that literal ec iliance with the provisions of the guide, because of their specific nature, 1 I' is not appropriate. Some RG 1.97 requirements call for exces- 'I sive ranges or inappropriate categories. Other requirements 5
Yh could adversely affect operator judgeent under certain condi-l tions. For exampic, rese,3rch by S. l.evy, Inc., shows that core ~ _ thermocouples will provide ambiguous information to.BWR opera-tors. The Owners Group intends to. follow the criteria used' by the NRC for establishing Category 1, 2, and 3 instruments, although it should be noted that Category 2' instruments could h-!
~!
vary.widely between utilities, because of various plant-unique features. is appli-The following Owners Group compliance statement cable to the regulatory positions defined in RG l'.97, Revision -- 2 (the paragraph numbers cited correspond to those in RC 1.97). ,
.y
- 1. Aceident-Monitorinf Instrumentaeion i T a r . 1.1 : The B'.9 t.hmers Group concurs with this defini- !
4 tian. ' N r J . .' : Th La Owner 'Creup concurs with this defini-tien. Par. 1.3: Instrunents used for accident monitoring to " meet the provisions of RC 1.97 shall have the proper sensitivity, ~ the con-range, transient response, and, accuracy to ensure that I trol room operator is able to perform his role in bringing the plant to, and maintaining it in, a safe shutdown condition and in assessing actual or possible releases of radioactive mate-rial following an accident. Each utility shall assess its plant accident-monitoring instrumentation' system. Accident-monitoring instruments that are required to be l .; environmentally qualified will be qualified to the-requirement The seismic of NUREG-0588 and Memorandum and Order CLI-80-21. a qualification of instruments will be based on individual assessments performed by each utility. Each plant will comply with the quality assurance require-ments, using its approved quality assurance program, as described in the FSAR or elsewhere. This would ensure that accident-monitoring instruments comply with the applicable requirements i of Title 10 CFR 50, Appendix B. l I
"% {
6 v .
'f
3 - f-4 Each plant program for periodic checking, testing, cali-
- brating, and calibration verification of accident-monitoring instrument channels (RC 1.118) shall he in accordance with the l utility's commitment, as specified in the FSAR, or elsewhere.
Par. 1.3.1: A third channel of instrumentation for-Category 1 instruments.will be provided only if a failure of one accident-monitoring channel results in information ambi-guity that would lead operators to defeat or fail to accomplish a required safety function and if one of the following meas-ures cannot provide the information:
- 1. Cross-checking with an' independent channel that '
monitors a different variable bearing a known relationship to the variablo being monitored.
- 2. Providing the operator ' t:, the sap::btlity of pct- ,
turbing Re measured variable to doctrnine which channel has
]
failed by observing the response on each instrunont.- '
- 3. The use of portable instrumentation for validation.
Categcry 1 instrument channels, which are designated as i being part of a Class IE system. will' meet the more-stringent design requirements of either the system or the regulatory guide. The requirements for physical-independence-of electrical systems (RG 1.75) shall be based on each plant!s commitments i in the FSAR, or elsewhere. ! Par. 1.3.2: The Btm Owners Group concurs with the regu-- { latory position for Cctegory 2 instrumentation, except as modified by Par.1.3 above. Par. 1.3.3: The BWR Owners Group concurs with the regu- d latory position for-Category 3 instrumentacion. ; l Par. 1.4: To assist the control room operator, identifi- ! cation of instruments designated as Categories 1 and 2'for ! f variable types A, B, and C should be made with due considera- # tion of human factors engineering. This-position is taken to clarify the intent of RC 1.97, which specified that these s I 7 l
ih . Il instruments be easily disi vened ! r use during accident condi-
- i tions (see Issue 1, 5cc. 5). .
{l, Par. 1.5: m: Stl!: & ners Group con:urs with the regula-tory position taken in this section, except as modified by l Par. 1.3 above. [ Par. 1.6 It is the position of_BhTOG that in_ terms of ; accident monitoring at a EkT facility, Table 1 of RG 1.97 does. ;
- not represent a minimum number of variabies and does not neces-sarily represent correct variable ranges or instrumentation ,
categories. ,. Each S'..? facility shall assess its compliance with the intent of-nG 1.97 by establishing a list of accident-monitoring , variables applicabic te its om plant. The classification of t un' : uwatut un um .:
. t^ c.:. w r , t; . var tam u a Category;1, ,
. ' , Or 3 .ii.a L 1 !.s t' s omn: t aas:s tit'i ' h i. in',:ent and method used
. . +
*.. 97.
in
- The.BkT. Owners Group position on the implementation of ,
each variable described in Table 1 of RG 1.97 and in other applicable documents is presented in Sec. 4.
- 2. Svstems Operation Monitoring and Effluent Release Moni-toring Instrumentation The Bb7 Gwners Group position stated in Par. 1.3'above is applicable to the Type D and E variables described in.
1 RG 1.97. Par. 2.1: The Bk"R Owners Group concurs with these: I definitions. , Par. 2.2: The Bb7 Owners Group concurs wich'this regula-tory position.- Par 2.3: The Bb7 Owners Group concurs with this regula-tory position. Par. 2.4: The Bh1 Owners Group concurs with this regula- , l tory position. g Par. 2.5: The Bh3 Owners Group position as stated in Par. j 1.6 above is applicable to this regulatory _ position.
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._lmplementation of Dnsign Changes 1 p.
The:8WR Owners Croup recoments that the implementation into each plant demi;n of additt.ocal design changes, ss required l by RC 1.97, be integrated with the implementation of other con-trol room design improvements. A relationship exists between identf.fying accident-monitoring
. variables, developing operating procedures, reviewing control room human factors engineering, and incorporating design changes into the plant. BbT.00 believes that an integrated approach l precludes the use of a specific implementation date for all BWR plants. In this regard, the Owners Croup recommends titat imple-nentation dates should be sclieduled on a plant-by-plant basis.
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- 3. PROPOSED TYPE A VARIABLES I
< . Fegulatory Cuide 1.97, kevision 2 designates all Type A U vartaoles as plant-specific, thereby defining none in particu-1er. The Guide defines Type A variables as Those variables to be monitored that provide primary information required to permit the control room operator to take specific manu-- ally 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. Regulatory Guide 1.97 defines primary information as "informa, tion that is essential for the direct accomplishment of the g' spectfled safetr functions." Variable.* assoc!.e ed with con- 5 t i n p n c.* .w t 1<* n e t h a ' .x ,, be ident!!1a in *.titten procedures aru .wl .ud ! M- thi, Jelinition of primary information.
.% part of their review of RG 1.97, the SL*R owners tinder-took the tas,k of developing and analyzing a group of variables that were determined to be potential candidates for inclusion in RG 1.97 as specific Type A variables. The variables identi '
find by the Owners Group are generic in nature, and the appli-cability of a given variable to a particular facility should be determined on an individual utility basis. In the summary that follows, two groups of variables are defined: (1) proposed Type A variables and (2) potential Type A variables. The variables listed are based on the BVR Owners , Group Emergency Procedure Guidelines (EPC's). Although all of-the operator actions a ncified below may not be required to ensure that safety systems fulfill their safety functions in terms of design-basis events, they are nonetheless included in ' the interest of completeness. 4 4 10 ' 1
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' Variables identifled as Type A !
(The variables listed here are also included in the tabu-lation of Sec. 4.) j Variable A1. RPV Pressure l t Operator actions (1) Depressurize RPV and maintain safe ! cooldown rate by any of several systems, such as main turbine bypass valves, isolation condenser, HPCI, RCIC, and WCU: (2) initiate isolation condenser; (3) manually open one SRV to { reduce pressure to below SRV setpoint if any SRV is cycling. Safety function (1) Core cooling; (2) maintain reactor coolant system integrity. Variable A2. RPV Vater Level Operator action: Festore and maintain RPV vater level. l Safety function: Core cooling l Variable A3. Suporession Peel Water Temperature f Operator actient (1) Operate available suppression pool , cooling system when pool temperature exceeds normal operating limits; (2) scram reactor if temperature reaches limit for ; scram; (3) if suppression pool temperature cannot be maintained . below the heat capacity temperature limit, maintain RPV pressure ! below the corresponding limit; and (4) attempt to close any $ stuck-open relief valve. Safety function: (1). Maintain containment integrity and (2) maintain reactor coolant system integrity. Variable A4. Suppressian Pool Water Level Operator action: Maintain suppression pool water level f within normal operating limits: (1) transfer RCIC suction from the condensate storage tank (CST) to the suppression pool 1 in the event of high suppression-pool level; and. (2) if suppres-sion pool water level cannot be maintained below the suppression pool load limit, maintain RPV pressure below corresponding i limit. Safety function: Maintain containment integrity.
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. . _ . . . _ . , _ _ . __ .. _ . u
k I Variable 35. P re.ec il r s v ur., operator action:. C.ontrol primary containment pressure-by ans of several sys:$ne such as centainment pressure con-trol svstems, suppres. ion peel sprays, drr.cil sprays. Safety function: (1) Maintain containment integrity and (2) maintain reactor coolant system integrity. Potential Type A Variables (The following is a list of possible Type A variables.to be determined at each plant; they are not included in Sec. 4.)_ Variable 1. Condensste Sterace tank Level Operator action: Transfer HPCI or RCIC suction or both frem. C81 Le suppre. ien rec 1. Pim*cssien: N'T "as recon.cnded autor.ctic suction trans-for for HPC: and RC;C. This variable is not a Type A variable if the automatic su: tion transfer is not installed. Variable 2. Emergen " Diesel Cenerator_{EDC) Load Operator action: Control loading of the EDC's. Discussion: Some plants have a planned manual action to verify the loading on the EDC's before any other safety-relcted loads are added. If no planned action is necessary, this vari-able is not type A. Variablo 3. P,aaetor Suildine Flood Level Operator action: Initiate pump-back of sump to suppression pool. Discussion: Uater can accumulate in the reactor building during long-term cooling with any postulated leakage. The flood-level indication vould alert the operator to a problem, but this indication is an aid to and not the accomplishment of a safety function. 12
I . Variable /.. Drwell Tenserature
- Operator action: Initiate sprays, reactor water level compensation.
Discussion: This variable may be needed for reactor-water-level compensation. Note: Although the EPC's mention drywell ! temperature, the drywell pressure is the key variable for con- 1 tainment integrity; dr>vell temperature is a secondary consid-eration. This issue vill be addressed by the ICC subcommittee. ) Variable 5. Suppression Pool Pressure Operator action: Initiate suppression pool sprays. ' Discussion: The suppression pool sprays are not used in safety analysis. Although the EPC's use suppression pool pres- l ,l sure to initiate suppression pool spray, centainment pressure may be used to approximate the suppression pool pressure. l Variable 6. Ox" ten or Hvdrogen Concentration I Operator action: If containment atnosphere approaches the i combustible limits, initiate combustible gas control systems, 2
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0xygen for inerted and hydrogen for non-inerted containments.
]
Safety function: Maintain containment integrity. . j I ; I I. :
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- 4. PLANT VARIABLES FOR ACCIDENT MONITORING :
Bh*ROG position- an the implementation of the variables listed in Table 1 of RC 1.97 and on the assignment of-design and qualification criteria for the instrumentation proposed for their measurement is summarized in the tabulation that follows. , in brief, the measurement of the five variable types provides the follewing kinds of information to plant operators . during and after an accident:. (1) Type A--primary information, on the basis of which operators take planned specified manually controlled actions: (2) Type B--information about the accom- ; l l plishment of plant safety functions: (3) Type C--information
}
about the breschint of bstriers to fission product release; (4> .g v .t--f e.ier:mion sbout the op.ratien of individual safety > systems; and (3) iyrv L--information about the magnitude of the, , relusse of radieactive esteria18. , The three categories shown for the variables define the design and qualification criteria for the instrumentation that ; is to be used for their measurement. Category 1 imposes the -l most stringent requirements; Categories 2 and 3 impose pro- l t gressively less stringent requirements, j The categories are also related (in RG 1.97) to " key , variables." Key variables are defined dif ferently for the . different variable types. For Type B and Type C variables, i t the key variables are those variables that most directly bdicate the a:::r.rlis:r:cnt of a safcty fw:ction; instrumenta-tion for these key variables is designated Category'1. Key variables that are Type D variables are defined as chose vari-ables that most directly fudice:0 :;te operation of a safety ayste ; instrumentation for these key variables is usually : Category 2. And key variables that are Type E variables are j defined as those variables that most directly indicate the relcaN cf Pa:fict::ti 'c materia!; instrumentation for these key e 14 :
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variables is also usually Category 2. A complete discussion ! I of the variable types and instrumentation design criteria is ' presented in RC 1.97. i ! lt should be noted that the Type A variables listed below are being proposed for inclusion in RG 1.97 on the basis of l analyses conducted by the Owners Group (Sec. 3). Table 1 of ; RG 1.97 designates all Type A variables as plant specific and thus defines none in particular. j The variables are listed here in the same sequence used in Table 1, RC 1.97; however, for convenience in cross-I referencing entries and supporting data, the variables are ,. j designated by letter and number. . For example, the sixth B-type l variable listed in RG 1.97 is denoted here as variable B6. (A copy of Table 1 from RG 1.97 is provided in Appendix C.) BL* ROC's position is show for each variable and for its inscru.entation design criteria'and category. (The latters s.7 i and M
' preceding the category numbers identify the'0wners Group and RC 1.97, respectively.) in general, there are three kinds of responses or re' commendations: (1) implement the variable I
and required instrumentation in accordance with the regulatory position stated in Table 1, RG 1.97 (2) implement, with quali- {
-I fying exceptions or revisions; and (3) do not implement.
As necessary, the positions of BWROG are elaborated or substantiated in the Supplementary Analyses section (Sec. 5) ! or in supplementary documents provided in the appendixes. .YJ te C & P:fsren:cs te rk J.::: b. Sec. 6 are r~ait by citing the ;
$48k? n;4Mb8PS th:t 2ff4CP in the uffer corngt Of the pages IM Etc. 3.
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i Type A Variables L. ~ The fo11oving Type A variables are recom. ended by the . i M ers Group (00) for inclusion in RG 1.97 as type A. (See Sec. 3.) A1. Reactor pressure (00 Category 1) RECO W.ENDATION: Implement. See B6, C4, and C9. A2. Coolant level in reactor (OC Category 1) REC 0!D!ENDATION: Implement. See 84 A3. Suppression pool water tar.perature (OC Category 1) RECO'0IENDATIOS: Implement. See D6. A4 Suppression pool water level (OG Category 1) REC 0!D!ENDA' TION: Implement. See C7 and D5. i Dr.well pre **ure W c ategory 1) .'
'5.
vtc0';:::'DA; a 1epler.wn t . ey;c '. fer plants , without autosta.rting drwell sprst. See C7, 59, CB, ' C10, and .% . ; h i o 4 9 e
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l Type B Variables
?':e .i:i t:1 d:ntv:t Bl. Neutron Flux (OG Category 2; RC Category 1)
RECOMMENDATION: Implement, but as Category 2 with alarm and reduced range, in accordance with data in Issue 2.
- 82. Control Rod Position (OG Category 3; RC Category 3)
RECOMMENDATION: Implement B3. RCS Soluble Boron Concentration (sample) (OG Category 3; RG Category 3) RECOMMENDATION: Implement CCP0 C00!O';t B4 Coolant tevel in Reactor (OG Category 1; RG Category 1) ,, .' R ECO:t:ENDATION: Do not imple.ent. See A3, C3, and ; a, <
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- 1. sue 3. -
B5. St.'R Core Thermocouples (RG Category 1) RECOMl:ENDATION: Do not implement. See C3 and Appendix A. int:ini't.c .*ca:::r C cla t System ?>1tegrity B6. RCS Pressure (OG Category 1; RG Category 1) RECOMMENDATION: Implement. See A2, C4, C9, and Issue 3. B7. Drywell Pressure (OG Category 1; RC Category 1) RECOMMENDATION: Implement. See A6, B9, CS, C10, and D4 BS. Drywell Sump Level (OG Category 3; RG Category 1) RECOMMENDATION: Implement as Category 3.- See C6 and Issue 4. Maintaining Contai ment Isttegrity B9. Primary Containment Pressure (CG Category 1; RG Category 1) RECOMMENDATION: Implement. See A6, B7, C8, C10, and D4. B10. Primary Containment Isolation Valve Position (excluding check valves) (00 Category 1; RG Category 1) RECOMMENDATION: Implement. Redundant indication is not required on each redundant isolation valve. 17 ~
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! i o i Type C Variables ;
<.: : llr:g l
C1. Radioactivity Concentration or Radiation Level in Circu- gl lating Primary Coolant (RG Category 1) [t RECO)0!ENDAT10N: Do not implement. See issue 5. C2. Analysis of Primary Coolant (gama spectrum) (00 Category
- I 3; RG Category 2)
RECOFBtENDATION: Implement C3. BW Core Thertnocouples (RG Category 1) . RECO)DtENDATION: Do not implement. See B5 and Appendix A.
.% .: n ar :c.::.a>:: lv..c::a- l : w ..aw;.
RCS Pressure (OG Category 1: RC Category 1) C!. . ( Ern'a'rND/iTIT:: * :10-ent. ,4 e 42. E6. and C9. l-L .t . rl: ::n iNin:innsnt Arwa :::.distic . (W Category 3; ls hr Catescry 31 R LC0:.
. tE.DAT N:. : Impic.ent. Site El.
Co. Drywell Drain Sumps Level (identified and unidentified leakage) (OG Category 3; RG Category 1) f RECO)B!ENDATION: Implement as Category 3. See B8 and Issue 4 l C7. Suppression Pool Water Level (OG Category 1; RG Category 1) REC 0totENDATION: Implement. See AS and D5. g CS. Drywell Pressure (OC Category 1; RG Category 1) RECOFDIENDATION: Implement. See A6, B7, 59, C10, and D4 Tea:t. sn~:e):: C9. RCS. Pressure (OG Category 1; RG Category,1) RECOMMENDATION: Implement. See A2, B6, and C4 l C10. Primary Containment Pressure (OG Category 1; RC Category 1) l RECOMMENDATION: Implement. See A6, B7, 89, CS, and D4 N' C11. Containment and Drywell H Concentration (OG Category'1; a-RC Category 1) E RECO)B!ENDATION: Implement i I 1, ' g
. S
1' l C12. Containment and Drywell Oxygen Concentration (for i inerted containment plants) (OG Category 1; RC Category 1) # RAC0PetENDATION: Implement. See A1. i C13. Containment Effluent Radioactivity--Noble Cases (from ;
. identified release points including Standby Cas Treatment l System Vent) (0C Category 3; RG Category 3) i REC 0FDIENDATION: Implement d14. Radiation Exposure Rate (inside buildings or areas, e.g., !
auxiliary building, fuel handling building, secondary i cotitainn.ent, which are in direct contact with primary containment where penetrations and hatches are located) (RG Category 2) REC 010!ENDATION: Do not implement. See E2, E3, and issue 6. , l C15. Effluent Radioactivity--Noble Cases (from buildings as indicated above) (OG Category 2; RC Category 2) i 4 RECO:2CNDATION: !=picment , I ! l , f I ; 4 i . I : I : w 19
b Ei i ljgg. Type D Variables ; q \ X4 ' r r . '. .~ '-
- s. . m 2V mi t '
D1. Main Teedwater Flow (OG Category 3; RG Category 3) RECobetENDATION: Implement l'l D2. Condensate Sterage Tank Level (OG Category 3; RG'Categot'y Ef 5:
- 3) ;
REC 0tNENDATION: Implement Frbu.") Concab:~:tr:t-Relatcd Ey: tem l D3. Suppression Spray Flow (RG Category 2) REC 0tt!ENDATION: Do not implement. See Issue 7. D4. Drywell Pressure (0C Category 2; RG Category 2) ' REC 0te!ENDATION: Implement. See A6, 57, B9, C8. and C10. l
"$. Sup;1ren v Dn t'.e3 ',la ter 1.evel (00: Category 2; RG Category M0!EN;'.i rIO . : Implement. See A5 and C7. ,
t D6. Suppression Pool L'ater Temperature (OG Category 2; RG Category 2) RECOSDIENDATION: Implement. lr
- Both local and bulk temperature. See A4.
D7. Drywell Atmosphere Temperature (OG Category 2; RG Cate-gory 2) REC 0tetENDATION: Implement. Sec Issue 8. g: 3+ DS. Dr>vell Spray Flow (RG Category 2) See Issue 7. REC 0t&tENDATION: Do not implement.
.t.*.dn Scw: Sptt-D9. Main Steamline isolation valves' Leakage Control System g s.;
Pressure (OG Category 2; RG Category 2) l Rg4D MENDATION: Implement if system is part of plant design. g 3( D10. Primary System Safety Relief Valve Position, Incleding ADS or Flow Through or Pressure in' Valve Lines 1 (OG Category 2; RG Category 2) REC 0F0!ENDATION: Implement I,
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I f:;'aty Sy::tema l l Dil. Isolation Condenser System Shell-Side Water Level I (OG Category 2; RG Category 2) l RECOMMENDATION: Implement if system is part of plant I design. l
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' D12. Isolation Condenser System Valve Position (OG Category 2;
.I RG Category 2) ; RECOMMENDATION: Impicment if system is part of plant l e design. ; i D13. RCIC Flow (OG Category 2; RG Category 2) I RECOMMENDATION: Implement. See Issue 9. D14. HPCI Flow (00 Category 2; RG Category 2) , RECOMMENDATION: Implement. See Issue 9. l D15. Core Spray System Flow (OG Category 2; RG Category 2) RECO'D:ENDATION: Implement. See Issue 9. ; D16. LPCI System Flow (OG Category 2; RG Category 2) REC 0!DiENDATION: Implement. See Issue 9. 017. SLCS Flow (OG Category 3; RG Category 2) , i RECOMMENDATION: Implement as Category 3. Avait ATWS resolution. See Issue 9. D18. SLCS Storage Tank Level (OG Category 3; RG Category 2) RECOMMENDATION: Implement as Category 3. Await ATWS ,, resolution. See Issue 10. ! i, , Residual Heat Re~ oval (RHR) Systems
! D19. RHR System Flow (OG Category 2; RG Category 2) ,
l RECOMMENDATION: Implement D20. RRR Heat Exchanger Outlet Temperature (OG Category 2; ,
, RG Category 2) i I
RECOMMENDATION: Implement ,i : l Cooling Water System D21. Cooling Water Temperature to ESF System Components (OG Category 2; RG Category 2) i RICOMMENDATION: Interpret as main system flow and ' implement, , t 21
i I ( D22. Cooling L'ater Flow to EST System Corpenents (0( Category 2; RC Category 2) RECT >9(ECATION: Interpret as main system f Jow and l'. i implement,
.%du:::c Sy:tcm D23. High Radioactivity Liquid Tank Level (OG Category 31 g RG Category 3) l*
RECOMMEWATION: Implement l PentiL2tist hett 0 ff D24. Emergency Ventilation Damper Position (OG Category 2; RG Category 2) l5 RECO MENDATION: Interpret as meaning dampers actuated ' under accidJnt conditions and whose failure could result in radioactive discharge to the environment. Control ;
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room damper position should be indicated.. Implement. i N5. Status of Standby Power and Other Energy Sources Important to Safety (hydraulic, pneumatic) (OG Category 2; RC Category 2) l' RECO)DIENDATION: Implement; on-site sources only. I] (:.'c ec: 'hc addition cf the following 0-type v:.riales is . rc:: mended by 3': ROC; see Issue it, Sc:. b.) ] D26. Turbine Bypass Valve Position (OG Category 3) , , RECOW!ENDATION: Add to RG 1.97. See Issue 11. ; D27. Condenser Hotwell Level (OG Category 3) REC 0!MENDATION: Add to RG 1.97. See Issue 11. 02S. Condenser Vacuum (OC Category 3) REC 0!MERATION! Add to RG 1.97. See Issue 11. E, !' D29. Condenser Cooling Water Flow (OG Category 3) RECOMMENDATION: Add to RG 1.97. See issue 11. l D30. Primary Loop Recirculation ylow (OG Category 3) REC 0FMENDATION: Add to RG 1.97. See Issue 11.
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Type E Variables .l Containment Radi:ti:n 'l El. Primary Containment Area Radiation--High Range (OG Category 1; RG Category 1) i RECOMMENDATION: Implement in accordance with l NUPEG-0737 commitment. See C5. l E2. Reactor Building or Secondary Containment Area Radiation (RG Category 2 for Mark I and II containments; OG Category I 1 and RG Category 1 for thrk III containments) ; RECOMMENDATION: Do not implement for Mark I and II con- ; tainments. Implement for Mark III containments. See C14, ! I E3, and Issue 12. Area Radiation I 1 I E3. Radiation Exposure Race (inside buildings or areas where i access is required to service equipment important to . ; safety) (OG Category 3; RG Category'2) I I REC 0tDIENDATION: Implement as Category 3. using existing instrumentation. See C14, E2, and Issue 13. "l Airborne Radioactive Materials Released from ?: ant j E4 Noble Gases and Vent Flow Rate (OG Category 2; RG Cate- ) gory 2) l RECOMMENDATION: Implement - E5. Particulates and Halogens (OG Category 3; RG Category 3) , ) 1 RECOMMENDATION: Implement ] l Dwirons Radiation and Radicactivity l l E6. Radiation Exposure Meters (continuous indication at fixed I locations) l REC 0f0fENDATION: Deleted. See NRC arrata of July 1981. ! 1 E7. Airborne Radiohalogens and Particulates (portable sampling 1 with on-site analysis capability) (OG Category 3; RG Cata, ' gory 3) j RECOMMENDATION: Implement E8. Plant Environs Radiation (portable instrumentation) j (OG Category 3; RG Category 3) ! RECOMMENDATION: Implement (portable equipment) j E9. Plant and-Environs Radioactivity (portable instrumenta-tion) (OG Category 3; RG Category 3) RECOMMENDATION: Implement (portable equipment) I 23 l
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.: ,~gr4 C10. Wind Direction s ic Ca t e p r y 3 ; 5'.C Ca t e;o ry 3 )
*:ECOMEND.iile::: Implement Ell. Wind Speed (00 Category 3; RG Category 3)
T,ECOM ENDATION: Implement-E12. Estimation of Atmospheric Stability (OG Category 3;
- RG Category 3)
RECotttENDATIO!; Impivment , 4 .e '.i . r : ,c t - * -* e,- c.: d;,!.!:p (h Apcic C:rdi*,ity Or...cie:) i E13. Prin.4ry Coolant and Susp (OC Category 3--Primary Coolant only; RO Category 3) ' RECom ENDATION! Implement Primary Coolant. Do not impicment Sump. See Issue 14 e
;i . itintainnent '. h (0C Category 3; R0 Catv;ery 3) i, EDT *. ;'.*:': AT ; ", : hplement . .
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- 5. SUPPLEMENTARY ANALYSES These supplementary analyses support positions cited in I Sec. 2 (Issue 1) and Sec. 4 (Issues 2-14).
I Contents Issue 1. Instrument Identification l Issue 2. Variable B1 Issue 3. Trend Recording Issue 4 Variables 88 and C6 Issue 5. Variable C1 I Issue 6. Variable C14 Issue 7 Variables 03 and 08
!ssue 8. Variable D7 Issue 9. Variables D13-D17 Issue 10. Variable D18 Issue 11. Variables D26-D30 Issue 12. Variable E2 Issue 13. Variable E3 Issue 14. Variable E13 l
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'E . i' 1 J 688UE 1. INSTRU MENT IDENTlfelC ATION i:~ - [ issue Definition 1 Regulatory Guide 1.97 specifies, in par 1.4.b, the Il , following: "The instruments designated as Types A, B, and C and Categories 1 and 2 should be specifically identified on l 1 the control panels so that the operator can easily discern g that they are intended for use under accident conditions." 5 Ii Discussion t The objective of this regulatory position is the achieve- l, e ment of good human f actors engineering in the presentation of information to the contrcl room operator. This objective is f best achieved by evaluating current practices and procedures l i that provide for identifying instruments in a manner that sids ' the operator; redundant labels would tend to distract the oper-ator and cause confusion. The Control Room Design Review of -; the BWR Owners Group has the charter to provide a basis for assuring proper identification of accident instrumentation with consideration for curtant information for safe plant f shutdown, operational training, and procedures. I;
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Conclusion
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. Tastruments designated as Categories 1 and 2 for monitor-T ing variable types A, .B. and C should be identified in such a manner as to optimize applicable human factors engineering f and presentation of information to the control room operator.
This position is taken to clarify the intent of RG 1.97, which specifies that these instruments be easily discerned for use during accident conditions.
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d' ,I ' ISSUE 2. VARIABLE B1 B1: Neutron Flux leaue Definition The measurement of neutron flux is specified as.the key variable in monitoring the status of reactivity. Neutron flux is classified as a Type 8 variable, category 1. The specified range is 10-6 percent to 100 percent full power (SP.M APRM). The stated purpose is " Function detection; accomplishment of mitigation." i Olscussion The lower end of the specified range, 10-6 percent f'ull power, is intended to allow detection of an approach to criti-cality by some undefined and noncentrollable mechanism after shutdown. In attempting to analyse the performance of the neutron-flux monitoring systems, a scenario was postulated to obtain the required approach to criticality. Basically, it assumes an increase in reactivity from loss of boron in the reactor water. I The accident scenario incorporates the following factors:
- 1. The control rods fail (completely or partially) to insett, and the operator actuates the standby liquid control system (SLCS).
- 2. The SLCS shuts the reactor down.
- 3. A leak in the primary systein results in an outgo of borated water and its replacement by water that contains no boron.
- 4. A range of leak rates up to 20 gpm was considered (see Table 1).
I w 7 I
{ l Calculations were made to evaluate the rise in neutron .I i: l population as a function of different leak rates. The cal- '
'culations were made for a shutdovn neutron level of 5 = 10-8 ,
percent of full power.. The choice of 5 = 10-6 is based on ll measurements at two nuclear plants. The shutdown level was ! assumed to have a negative reactivity of 10 dollars, an l-i assumption that is representative of a shutdown with all rods , inserted. The results of the calculations are presented in Table 1. The numbers in the table refer.to the time in hours
-l required to increase the flux by 1 decade. For example, with a leak of 5 gpm, it takes 100 hr to increase the power from 5 = 10-8 percent to 5 = 10-7 percent, and 10 hr to increase ,
it from 5 = 10-7 percent to 5 = 10-6 percent. The reactor is suberitical and the neutron level is given , by { Neutron level = 5 = M, where S is the source strength and M is the multiplication. which is given by M = 1/(1 - k). 9 Tor k = 0.9, M is 10; for k = 0.99, M is 100 and so forth. , For criticality, the denominator approaches 0, as k approaches 1.0. Thus, the calculation model used the above equation to calculate relative neutron flux levels for a suberitical reac-ter until the recctor was near critical; then the critical.
'i equation of power with excess reactivity was used. Reactor power is directly proportional to neutron level.
The increase in reactivity toward criticality can be turned around by actuating the SLCS. It is assumed that oper-l cth.gprocedures provide for refining the SLCS tank soon after its coruction. A second actuation of the SLCS would cause a ' decrease in reactivity because of the high concentration of ~ boron in the injected SLCS fluid relative to that in the leak-ing fluid (nominally 400 ppm). The sensitivity of the detector .< must allow adequate time for the operator to act. Ten minutes I
~
28
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1 . . \ ,
l 3' - a is considered sufficient time for operator action for accident ! prevention and mitigation. f l : ! Table 1 shows that the detector sensitivity (i.e., lower l range) requirement is a function of leak rate and therefore l of reactivity-addition rate. On the basis of a 20-spm leak, . rate, Table 1 shows that a detector that is on scale within 3 decades of the shutdown power would allow 0.18 hr (10.8 min) - for operator action before reactor power increased another l decade. A total of 0.36 hr (21.6 min) would be available for operator action from the time the indicator comes on scale to the time reactor power reaches 0.5 percent of full power. An I alarm would be provided to warn the operator when the 'teutron l flux starts to increase beyond a plant-specific set-point. The 20-gpm leak rate, which was assumed to continue for 27.75 hr. was used to define the sensitivity of the detector. It should be noted that the assumed leak rate, extended over , the 27.75-hr period, would result in a loss of inventory so large that it could not in reality go undetected-by the oper-ator. Moreover, reactivity-addition caused by this gradual boron depletion is unlikely, since boron concentration is sampled and measured periodically. Again, the improbable [ 20-spa leak rate was used only to obtain a mechanistic and l conservative approach for selection of instrument sensitivity. An absolute criterion for the lower range must include , consideration of the neutron source level. The use of the ; neutron level 100 days after shutdown is conservative. There : l is high* probability that conditions would be stable and con-l tro11abis.2 days after the emergency shutdown, for the core- l decay heat is at a low level and the boron monitoring system - should be functioning by that time. The actual neutron level l will vary with fuel design, fuel history, and' shutdown con- ; trol strength. Measurements of shutdown neutron flux (with all rods inserted) at two Bt!R reactors show readings of 30 to 80 counts /sec (1000 counts /see corresponds to 10-" of full 29 ~ l :
i power). Measurements on other Bia reactors and for different =i fuel histories would show some variation, but those variations , would be small compared with a criterion that is concerned ; with units of decades. Regulatory Guide 1.97 classifies the instrumentation for l measuring a variable as Category 1 on the basis of (1) whether it is a key variable (defined in Sec. G), and (2) its importance to safety. Neutron flux is the key variable for measuring , reactivity control, thus meeting the requirement of criterion (1). The degree to which this variable is important to safety is another consideration. The large number of detectors (i.e., g
. P i source-range monitors and intermediate-range monitors) that are driven into the core soon after shutdown makes it highly ;
probable that one or more of the existing N>15 detectors will - be inserted. On the other hand, there is little probability that there would be, simultaneously, a need for this measure- ! ment (in terms of operator action to be taken) and an acci-
- dent environment in which the hHs would be rendered inoperable.
Further, the operator can always actuate the SLCS on loss of - instrumentation. Although some upgrading of the current hh5 may be appro-priate to improve system reliability and its ability to survive a spectrum of accidents, a rigorous Category 1 requirement is not justified when the purpose and use of the measurement are , analyzed as they relate to the criterion of "importance to safety." A Category 2 classification of this variable fully meets the intent of RG 1.97.
- Four alternative design approaches to meeting the neutron flux requirements of RG 1.97 have been identified. All four alternatives would provide indication over the range recom-mended by BWROG, using state-of-the-art electronics for dis-playing the detector reading. A particular utility can choose a suitable alternative, based on its own. design evaluation.
l The principal features of the four alternatives are presented l below. > 30 ,
~
_Alternativo 1. The first alternative provides for upgrading two or more of the source-range monitors (SRM's). The upgrading includes the connecting cable inside the drywell and the power source for the SRM drives. At least two SRM's would have dual roles of accident instrumentation and normal start-up; these two SRM's would be withdrawn a lesser dis-tance from the core than the SRM in the current design. It is estimated that in its fully withdrawn position, the cur-rent SRM will detect about 10~3 or 10-5 percent of full power. This sensitivity can be increased by using a withdrawn posi-tion that is less than the present 2-2.5 ft from the core.. A withdrawn position that produces 10 percent depletion in 5 years was used as a guide to the ecrimum allowed burn-up of the sensor. This position below the core would give the SRM . a detection capability of about 2a 10~7 percent of full power. The SRM drives need not be upgraded, because the upgraded detector system would be adequate, even if the drive did not move the SRM detector. (An upgraded power source for the drives improves the probability of insertion.) The success of this alternative--which uses the four SRM's for normal start-up--depends on a design modification to accommodate the new cable (the key concern is the flexibility of the cable, for the detector moves about 10 ft; this movement is accommo-dated in the cable loop) and on the design of a limit switch or a detent mechanism to hold the drive tube in the required intermediate position. Alternative 2. The second alternative is to replace two or more SRN systems with upgraded systems. The full SRM system, including the drives, would be upgraded. This approach would require input from a potential equipment supplier in order to estimate schedules, cost, and overall effect of the upgrading. Whereas the first alternative uses upgraded cables and power supply (which are commercially available). this 1 31
b i approach would require additional engineering to achieve an upgraded drive system as well. A Category 1 drive system is a developmental item. g In the third alternative, fixed in-core i Alternative 3. detectors are used. The system uses SR.M-type detectors as stationary detectors that are positioned close enough (as dis-cussed above) to the core to meet the lower range requirements. New cables are needed to meet the requirements of the accident environment. This system would provide dedicated " accident
'I monitors" in two of the intermediate-range monitor (IRM) tubes or in two local-power range-monitor (LPFJt) tubes. It may be f easible to put five detectors in the LPFJI cube or, if space is limited, the bottom detector of the LPFJ! string could be replaced with the " accident" detecter. k'ich this approach the four movable SFJ:'s would continue te be available for normal functions.
Alternative !.. In the final alternative, out-of-core detectors, which are being qualified for use in pressurized water reactors (PkT s), are used. Considerations of this ongoing PWR qualification program for Category 1 instrumen-tation and the lack of any effect on the current neutron moni-toring system (NMS) make this alternative an attractive one. The key question is whether these out-of-core detectors can meet the lower range requirement, for the detectors are posi-tioned outside the RPV shield wall. A test is needed to demonstrate that the neutron count at this location is ade-quate. Based on calculations of neutron flux made for a Bk'R at full power (see Fig. 1) and on current detector design practices, the out-of-core detector may be feasible. Other effects, such as attenuation by water that is at a lower tem-perature (than the full-power operating temperature) and by boron in the water, need to be considered. s 32
- . - s
Conclusion k A range from 5 = 10-5 percent of full power (within 3 decades of the neutron flux level 100 days after shutdown) to 100 percent of full power is recommended.. An alarm is also i recommended that would alert the operator of a rise in neutron flux. It is concluded that a Category 2 classification is responsive to the intent of RG 1.97, as are the four alterna- - tives, provided that the design program resolves the specific ! design concerns identified in the Discussion. t 4 33 1 i
k TABLE 1. REL\TIVE SEUTRON TLUX VERSt'S TIMEa Leakage rate, gpm (ramp rate, c/ min) Percent 5(0.15) 20(0.60) ot 1(0.03) e power * { I 4 E. 4 I a 5x 10-8 -555 500 -111 100 -27.75 25 , 5x 10 " -55 50 -11 10 ~2.75 2.5 5 = 10-6 -5 5 -1 1 -0.25 0.25 5 = 10-5 o o o 5 = 10-' O.A 0.8 0.36 0.36 0.18 0.18 l 5 = 10-3 1.33 0.53 1 0.51 0.15 . 0.25 0.07 5 = 10-7 1.59 0.26 { 0.62 0.11 l 0.31 0.06i , 5 = 10-1 1.30 0.21 1 0.72 0.10 0.36 0.05 5 = 10* 1.89 0.09 0.80 0.0S 0.40 0.04 s " Shutdown flux = 5 = 10-8 percent of power. b t = total number of hours; o = hours for neutron flux to increase [, by one decade. E' l' I I I I I I 34 g.
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I g' 5 ISSUE 3. TREND RECORDING a! 5; I lesue Definition The purpose of addressing Issue 3 is to determine which variables set forth in RG 1.97 require trend recording. - i Discussion . Regulatory Guide 1.97, par. 1.3.2f, states the general j requirement for trend recording as follows: "Where direct and j issnediate trend or transient information is essential for f operator information or action, the recording should be con- f
~
tinuously available for dedicated recorders." Using the BWR , owners Group Emergency Procedures Guidelines (EPG's) as a basis, the only trended variables required for operator action are : reactor water level and reactor vessel pressure, f Conclusion on a generic basis, only reactor water level (variable -
- 54) and reactor vessel pressure (variable B6) require trend .[
recording; however, other variables may be necessary on a- l plant specific basis, i g! h I: I I! i
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ISSUE 4. VARIABLES B8 AND CS B8: Drywell Sump 1.evel C6: Drywell Drain Sumps Level lasue Definition Regulatory Guide 1.97 requires Category 1 instrumentation , to monitor drywell sump level (variable B8) and drywell drain sumps level (variable C6). These designations refer to the drywell equipment and floor-drain' tank. levels. Category 1 instrumentation indicates that the variable being monitored
- is a Key variable. In RG 1.97 a. key variable is defined as
. that single variable (or minimum, number of vr-tables) that most directly indicates the accomplishment of a safety function. . . ." The following- discussion supports the Bk'R Owners Group alternative position that drywell sump level and drywell drain-sumps levels should be classified as Category 3.
instrunantation. Discussion The Bk'R Mark 1, II, and III drywells .have two drain sumps. One drain is the equipment drain sump which. collects identi-fied leakage; the other is the floor drain sump, which. collects unidentified leakage. Although the level of the drain sumps can be a direct indi-
.f cation of breach of the reactor coolant system pressure. boundary,.
the indication-is not unambiguous, because there is water in-those sumps during normal operation. There is other instru-mentation required by RG l.'97 that would indicate leakage.in the drywell:
- 1. Drywell pressure--variable B7, Category l'
- 2. Drywell temperature--variable D7, Category 2'
~
37 l . j%s ~ ~ N w
is; t , 4 3.. Primary containment area' radiation--variable C5, i Categothi)
]
' ThhEryve11-sumplevelsignalneitherautomaticallyini-t1ates safety-related systems nor alerts the operator to the ;
need to take safety-related actions.- Both sumps have level s detectors that provide only the following nonsafety indications:
- 1. Continuous level indication .(some: plants)
- 2. Rate of rise indication?(some-plants)
- 3. High-level alarm (starts first sump pump)'
- 4. High-high-level alarm-(starts'second sump pump) '
In addition, timers are used in most' plants to indicate the f duration of sump-pump operation and thereby. permit the amount , of leakage to be estimated. Regulatory Guide 1.97 requires instrumentation to function during and'after an accident. The drywell s. ump systems are _, deliberately isolated at the primary containment penetration upon receipt of an~ accident signal to establish containment- ' integrity. This1 fact renders the drywell-sump-level signal irrelevant. Therefore, by design, drywell-level instrumenta-tion serves no useful accident-monitoring function. The Emergency Procedure Guidelines use .the RPV level and' 9 the drywell pressure as entry conditions for the Level Control Guideline.- A small line break will cause the drywe11' pressure to increase before a noticeable increase in the sump level. Therefore, the drywell sumps will provide a " lagging" versus e i "early" indication of a leak.
.n .
s,# H Conclusion I Based on the above considerations, the BWR Owners Group believes that the drywell-sump level and drywell-drain-sump 4 level instrumentation should be classified as Category 3, "high-quality off-the-shelf instrumentation." .. 38 i i
, 1 ISSUE 5. . VARI ABLE C 1 C1: Radioactivity Concentration or Radiation Level in Circulating Primary Coolant t t
' lasue Definition 1
-I Regulatory Guide'l.97 specifies that'the status of the fuel cladding be monitored during and after an accident. .The -
I specified variable'to accomplish this monitoring is-variable a I Cl--radioactivicy concentration or radiation level in circulat- [ ing primary coolant. The range is given as "1/2 Tech Spec ; I Limit to 100 times Tech Spec Limit, R/hr." In Table 1 of RC 1.97, instrumentation for measuring variable C1.is desig- , 7;
, nated as Category 1. The purpose for monitoring;this. variable is given as " detection of breach," referring. -in this case, r
to breach of fuel cladding.
.. (
Discussion y l
.The usefulness of the information obtained by monitoring the radioactivity concentration or radiation' level in the cir-culating primary coolant, in terms of-helping the operator in d;
his efforts to prevent and mitigate accidents, has.not'been i substantiated. The critical actions that must be taken to , I- prevent and mitigate a gross breach of fuel cladding are (1)
, shut down the reactor and (2) maintain water level. MonitoringL
, variable C1, as directed in ~RG 1.97, will- have no- influence on either of these actions. The purpose of this monitor falls in-the category of "information that the barriers to release of ;
radioactive material are being challenged" and " identification
!- of degraded conditions and their magnitude, so the operator can ! take actions that are.available to mitigate the consequences." ,
i 39
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" LipE :
Additional operator' actions to mitigate the consequences of'f,uel Bi ; barriers being challenged, other than those based on Type A and I B variables, have not been' identified. ' Regulatory Guide 1.97 specifies measurement of the radio- [ activity of the circulating primary coolant as;the key variable in w>nitoring fuel cladding status during isolation of the_ NSSS. The word's " circulating primary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows f past the core. A basic criterion for a valid measurement of j the specified variable isTchat the coolant being' monitored,is coolant that is in active contact with the fuel. that is,= flow-Pi ing past the failed fuel. Monitoring the active coolant (or a ' The-post-sample thereof) is the dominant consideration. accident sampling system (PASS) provides a representative f j, sample which can be monitored. j The subject of concern in the RG'l.97 requirement is assumed to be an isolated NSSS that it shutdown. This assump-- l tion 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.. f l Further, the post-accident sampling system ((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. Conclusion The designation of instrumentation for measuring variable
'I Cl should be Category 3, because no planned operator actions are identified and no operator actions are anticipated based on this variable serving as the key variable. Existing Cate-gory 3 instrumentation is adequate for monitoring fuel cladding fI a
f status. y e
o ISSUE 6. VARIABLE C 14 C14: Radiation Exposure Race issue Definition . . l Variable C14 is defined in Table 1 of RG 1.97 as follows:
" Radiation exposure rate (inside buildings or areas, e.g., - -
auxiliary building, fuel handling building, secondary contain-: ment), which are in direct contact with' primary containment where penetrations'and hatches are located." The reason for: monitoring variable C14 is given.as " Indication of breach." Discussion The use of local radiation exposure race monitors- to detect . breach or leakage through primary containment penetrations is - impractical and unnecessary. In general, radiation exposure rate in the secondary containment will be largely a function- 1 of radioactivity in primary containment and in she fluids flowing in ECCS piping, which will cause direct radiation. . shine on the area monitors. Also, because of the amount of I piping and the number of electrical penetrations and hatches 1 and their widely scattered locations, local. radiation. exposure . rate monitors could give ambiguous indications. The proper , way to detect breach of containment is by using the plant; I noble gas effluent monitors. l
] ,
I Conclusion i I
.y I Using radiation exposure rate monitors to cietect primary 1 ' 4 containment breach is neither feasible nor necessary. Other 'l
- 41. . .;
. . A
s g' W : c- . . l means of breach detectien. chat are better suited to this function (as described above), are available. Therefore, it-is the position of the BWR Ovners Group that this parameter - not- be implemented. I; .
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[ . . d VARIABLES D3 AND D8 l
%: ISSUE.7.-
N D3: Suppression Spray Flow : DS: Dryvell Spray T1ow ; lasue Definition , Regulatory Guide 1.97 specifies flow measurements of suppression chamber spray.(SCS) (variable D3) and drywell > spray (variable D8) for monitoring the operation of the , primary containment-related systems.. Instrumentation for ' measuring these variables is designated Category 2', with at l range of 0 to 110 percent of design. flow ~. These flows relate to spray flow for controlling pressure.and temperature of the. drywell and suppression chamber. - - Discussion - i The drywell sprays can be used to control the pressure I and temperature of the drywell. The residual ~ heat removal ~. 'r (RHR) system-flow element is used for measuring drywell flow in most designs. [ The suppression pool sprays can be used to control-the pressure and temperature in the suppression chamber. The operator controis-pressure and temperature by adjusting sup-pressiost hbamber spray flow. The RER system flow element is. -? used for low indication in most designs. .Some plants ~have a flow element in the. branch line to the sprays.- The suppres-- I sion chamber spray operates in parallel with the drywell spray: ' and is regulated with a. throttling valve. The flow is deter- ? mined by the position of the throttling valve that is in the : branch line that feeds the containment spray lines. These. ; valve positions are indicated in= the control room. The i n 43
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E; '
't ef fectiveness of these flows can be. verified by pressure and j
tempereture changes of the drywell and' the suppression chamber. - r a Conclusion g-!. The current plant designs, in conjunction with operating practice, provide for operator information that is.suf ficient for determining the existence of . spray flows to the 'dryvell E '- 5 -- and suppression chamber without the use of a-dedicated flow-. ,
.s measuring instrument.
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ISSUE 8. VARIABLE D7
.. L~%e'.
.,? D7: Drywell Atmosphere Temperature issue Definitlon Regulatory Guide specifies drywell atmosphere temperature (variable D7,' Category 2) as one of the key variables in monitoring individual safety systems. The temperature range is specified as 40*F to 440'F. .
Discussion The evaluation of this' issue addressed requirements that - g call for direct operator action based on variable;D7,.that is, temperature and the associated variable of pressure. The BWR Emergency Procedure Guidelines (EPC's) provide guidelines for control of containment pressure and temperature. Classifica-- tion of this variable should be done on a plant-specific basis- I with full consideration for EPG requirements. Temperature-monitoring hardware inside the drywell may l not be qualified to the accident conditions:specified in ~ RG 1.97; the primary item of concern is the cable inside the drywell. { , 4 g 1 Conclusion ; v., frecommends implementation of variable D7 require-
- I. -
ments asspecified in RG 1.97. 9 1 4 O i l- ~ 45 1
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< Is ISSUE 9. VARIABLES D13.D1 N. -g D13: RCIC Flow gl D14: HPCI Flow D15: Core Spray System Flow - [l ~
D16: LPCI System Flow j D17: SLCS Flow
?
Issue Definition Regulatory Guide 1.97~ specifies flow measurements of the l reactor' core isolation cooling (RCIC) 'I following systems: (variable D13),'high-pressure cool' ant' injection (HPCi) (vari- , able D14), core spray-(CS) (variable D15), low-pressure coolant
' injection (LPCI) (variable D16), and istandby liquid control- ..l i
(SLC) (variable D17). The purpose is for monitoring the-oper-acion of individual safety.' systems. Instrumentation for meas-4 uring these variables is designated as Category 2; the range is specified as 0 to 110 percent.of design flow.. These vari-ables are related to flow into the reactor pressure vessel (RPV). - Discussion , The RCIC, HPCI, and CS systems each have one branch line- - the test line--downstream of the flow'-measuring: element., The- *
~
test line is provided with a motor-operated valve-that'is nor-mally closed (two valves in series in the case of the HPCI). Further, the valve in the test'line closes' automatically when
~
the emergency system is actuated, thereby ensuring;that indi-- cated. flow is not being diverted -by the cast line. Proper valve position can be verified by a direct . indication'of valve position. Although the LPCI has several branch lines located downstream of each flow-measuring element, each of those l a
., , , - - +, .- , . ,,-4 ,- , , . ,, ,
1 -
. lines is normally closed. Proper valve position'can be veri- >
fied by a direct indication of valve positidn. ; a l- Fodfalloftheabovesystems,there:nrevalidprimary -l indicatoth.'other than flow measurement to verify the per- . formance of th'e emergency system; for example, vessel water ! level. The SLC system is manually initiated. Flow-measuring devices were not provided for this systeta.- The pump-discharge ' { header pressure, which is indicated in the control room 'will indicate SLC pump operation. ~Besides the discharge header- - l pressure observation, the operator can verify the proper functioning of the SLCS by monitoring the following: 1.- The decrease in the level of the boric acid storage
- 2. The reactivity change in the reactor as measured by .
neutron flux
- 3. The motor contactor indicating; lights - (or motor cur-rent) 4 Squib valve continuity' indicating lights
- 5. The open/close position indicators of check valves (available in some plants) .l The use of these indications is believed to be a valid altarna-tive to SLCS flow indication. '
l- Conclusion The flow-measurement schemes for the RCIC, HPCI, CS, and LPCI are adequate in that they meet :he' intent of RC 1.97.- Monitoring.the SLCS can be adequately done by measuring. vari - s ables other than the flow.
- l
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47
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y n: ISSUE 10. - VARIABLE D18 l D18: SLCS Storage Tank Level I lasue Definition f ! Regulatory Guide 1.97 lists standby liquid-control system (SLCS) storage-tank level as a Type D variable with Category 2 design and qualification criteria. l . Discussion The symptomatic Emergency Procedure Guidelines (EPG), Revision 1, as presently approved do not consider ATWS condi- , tions; however,' the EPG committee of the Bk'R Owners Group has l: l been developing a draft reactivity control guideline in which procedures are described for raising the reactor water level .. based on the amount of boron injected into the _ vessel, as i. - indicated by the SLC tank level. Additionally, the operator is required to trip the SLC pumps before a low SLC ' tank level . 1' is reached, thereby preventing damage to.the pumps that-would; render them useless for future injections during the scenario.'-
- Regarding the instrumentation category requirement ~for - variable D18, RG 1.97 indicates that it is a: key variable in monitoring SLC-system operation. Regulatory Guide-1.97 also states that in general, key Type-D variables be designed and ,
a
- qualified to Category 2 requirements.
IN; applying these requirements of the Guide to this instrumentation, the following are noted: 1
'I q
i 48
I a
.4.
I. .
. e current design basis for the SLCS assumes a need for aar 'rnative method of reactivity control without a con-current"1Is$-of-coolantaccidentorhigh-energylinebreak.
1, The environment in which the SLCS instrumentation must work is therefore a " mild" environment for qualification purposes.-
- 2. The current design basis for the SLCS recognizes that the system has an importance-to safety that is less than=
the importance to safety of the reactor protection system and + the engineered safeguards systems.- Therefore,'in accordance l with the graded approach to quality assurance specified=in RG 1.97, it is unnecessary to apply a full quality-assurance - program to this instrumentation. Based on a graded approach to safety, this variable is . 7 1 more appropriately considered a Category 3 variable. Conclualon ! SLCS storage-tank-level instrumentation should meet ' l Category 3 design and qualification criteria. : l It is realized that the resolution of the-ATWS issue may j include substantial changes to the'SLCS design criteria. At l , i that time, the SLCS instrumentation should be reevaluated to
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.i ISSUE 11. VAF41 ABLES D26.D30--
D26: Turbine Bypass Valve Position- ;
.D27: ' Condenser Hoewell Level, j Condenser Vacuum' (
D28: ' D29: Condenser Cooling Water Flow D30: Primary Loop Recirculation
.I;;
lasue Definition ltl Regulatory Guide 1.97 states chat "The plant designer' , should select variables and information' display channels
~
, l required'by his design to enable the control room personnel BI to ascertain the operating status of each-individual safety-system and other sy' stems important'to safety to that' extent r necessary to determine if each system is operating or.can be placed in operation. . ..
The purpose of this analysis'was to determine whether certain othat D-type-variables should be added to Table 1, RG 1.97. I! Discussion , I! Regulatory Guide 1.97 addressed safety systems.and systems. important to safety to mitigate consequences of an accident.- I] Another list of variables has been compiled for thetBWR in NUREG/CR-2100 (Boiling Water Reactor Status Monitoring during
~
Accident Conditions, Apr. 1981). That report andT.a: companion report, NUREG/CR-1440 (Light Water Reactor Status Monitoring during Accident Conditions, June 1980), address plant systems not important to safety, as well as systems that are important to safety. In particular, these reports consider the potential role of. the turbine plant in mitigating certain accidents. These two reports were reviewed in determining whether any f variables should be added to the RG 1.97 list. l. 50 ~, 2 j
~- . - _ - - , . - . - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
The.NUREC evaluations used a-systematic eproahh to derive- 'I a var e list. .The basic approach of the analysis weg to focus or those accident conditions with which the operator is most likely to be confronted and on those accident conditions that result'in the most serious consequences,:should the oper-
. Stor fail to accomplish his required tasks. These studies .
used probabilistic event trees and the sequences of-the. Reactor
- Safety Study (WASH 1400) and similar studies. The events in I 3 each sequence that involved operator action.were identified.-
Also, events were added to the event tree to include additional' operator actions that could mitigate the accidtnt.. The event , ; !. tree defines a series of key plant states that could evolve as ! the accident progresses'and as the operator. attempts to respond. l Thus the operator's 'nformational i needs are linked =to tNese { i plant states.
~
NUREC/CR-2100 is a Bk'R evaluation undertaken to address. appropriate operator actions, the information needed to take. I those actions. and the instrumentation necessary--and suffi- [ cient-to provide' the required information. ' The sequences evaluated were
- 1. Anticipated transient followed by loss of decay-heat
[ removal [ l 2. Anticipated transients without scram (ATWS)
- 3. Anticipated transient together with' failure of '
HPCI, RCIC, and low-pressure ECCS
- 4. rge loss of coolant accident _(LOCA) with failure of emergency core-cooling systems ,
_g 5NSmallLOCAwithfailureofemergencycore-cooling jy systems The RG 1.97 list is based on accidents that result in an - isolated NSSS. The NUREG documents considered accidents that ; l could be prevented or mitigated by using water inventory and ' the heat sink in the turbine plant. 4 L ~
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Concluslon Five-of the 15 variables identified in the NUREG, but not in RG 1.97, are recommended as Type D, Category 3 additions to the RG 1.97 list. Four of these variables are in the . turbine plant: the. turbine bypass valve position, condenser I hotwell level, condenser vacuum. and condenser cooling water- , flow. These variables provide a primary measure of the status- l of a heat sink or water inventory in the turbine plant.: The turbine-plant systems are not - to be classed as "saf ety systems" or as systems important to safety.. The addition of reactor primary-loop recirculation flow is also recommended. i y i j i i 52
. ,1
. . . _j
-- - - - - - - - - - - . - -- - . - . ~ .. -
g7 14-
.r 'lSSUE 12. VARIABLE E2 i I: f* ,
, petor.BuildingorSecondaryContainmentRadiation
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lasue Definition > l Regulatory Guide 1.97 specifies that " Reactor building or secondary containment area radiation" (variable E2) should be monitored over the range of 10-1 to '10" R/h for Mark I and ' II containments, and over the range of 1 to 107 R/hr for Mark l III containments. The classification for Mark I and II.is . Category 2; for Mark III, the classification is Category 1.- 1 Discussion f* . . As discussed in the variable C14 position statement-I (Issue 6), Secondary Containment' Area Radiation is an inap-l propriate parameter to use to' detect,or' assess primary con- l I cainment leakage. However, for the Mark III containment, the reactor building is essentially part of the primary contain-'
- ment and it is appropriate to moditor that building volume as specified in RG 1.97.
t Concluslon-
*-hQ' the position of BWROG that the specified reactor; a 3 ,
g build '. area radiation monitors be. installed on Mark III i
~
containments, but that these monitors should not.be required for plants with Mark I and II containments. i I : l ; 53
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't ISSUE 13. VARIABLE E3 ,
1: E3 ! . Radiation Exposure Rate- .; i
- l. '
Issue Definition 1 lf
,r Regulatory Guide 1.97 specifies in Table 1, variable E3, that radiation exposure rate (inside buildings or areas where ,
l accese is required to service equipment important to safety). .. l be monitored over the range of 10-1 to 10" R/hr for detection. , l I and for long-of significant releases, for release assessment, term surveillance.
.Ip l Discussion ,,
l la general, access is not required to any area' of the - j secondary containment in order to service equipment important to safety in a post-accident situation. .If and when accessi- ' bility is reestablished in- the long- term, it will be done by a combination of portable radiation survey instruments.and post-accident sampling of the secondary containment atmosphere.- The existing lower-range (typically 3 decades lower than the RG 1.97 .f range) area radiation monitors would be used only in those instances in which radiation levels were very mild. ! 1 Conclusion It is BWROG's position that unless plant-specific design requires access to a harsh environment area to service safety-related equipment during an accident, this parameter should - 1 be modified to allow credit for existing area radiation moni-tors. That is, this parameter should be reclassified as 1
-i s
54 3
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'Cate ith a lower range to be_ selected on a plant-spec is.
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O[ i~ ISSUE 14, VARI ABLE E13 B El'3 : Primary Coolant and Sump IL lasue Definition : Regulatory Guide 1.97 requires. installation of the capa- - bility for obtaining grab samples (variable E13) of the con-
^
tainment sump, ECCS pump-room sumps, and other similar.:auxi 11ary building sumps for the purpose of release assessment, , verification, and analysis. .i
. Discussion l take into The need for sampling a particular sump must u; I
account its location and the design of-the planc in which it '> I is l'.s talled. For all accidents in which radioactive material' 3l would be in the primary containment sump of a BWR Mark I or I! Mark 11 containment, this' sump will be isolated'and'will over-
.s flow to the suppression pool. A suppression pool sample can-l
' therefore be used as a valid alternative to a containment-sump. '. i sample. The analysis of ECCS pump-room sumps and other similar auxiliary building sump liquid samples can-be used for release 1 assessment, as suggested in RC 1.97 only for those designs in which potentially radioactive water can be pumped out of a-controlled area to an area such as radwaste. .For designs in , which sump pump-out is not allowed on a high-radiation or an LOCA signal, or in which the water is pumped to the suppression pool, a sump sample does not contribute to release assessment. For these designs, the use of the subject sump samples for verification and analysis is of little value; a' sample of the , suppression pool and reactor water, as required-by other .; 56 l 6
, , . - . - - _ . e .. - - -
ik
-. portions of RG 1.97 provides a much better measurement for these purpos -
^A.
Conclusion
- l. A suppression-pool sample can be used as,an alterna-
.: tive to a primary containment-sump' sample for plants with Mark I or II containments.-
- 2. The analysis of ECCS pump-room sumps and other similar '
auxiliary building sumps is a consideration only-if the water , l is pumped out of the reactor-building (e.g., pumped to radwaste)~. For_ designs in which sump pump-out'is not allowed on a receipt of an accident signal, or in which the water is pumped _to the: suppression pool, analysis is not necessary. Provisions-for sump sampling and analysis should be in accordance with'each A' i utility's response to NUREG-0737. n 1
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- 6. -CONCLUSIONS i
The BWR Owners Group RC 1.97 Committee completed an
-i ' extensive analysis of the regulatory positions proposed in
. NRC Regulatory Guide 1.97, Revision 2. The principal-goal-of the comittee was to formulate the position of the' BWR ,
{ Owners Group relative to RG 1.97 requirements. Toward that'
~
end, the comittee developed--on the basis of studies con-f ducted by its own' representatives and its contractors--a series of positions with respect to interpreting.and.imple- I menting the various provisions of RG~1.97-. .
;. The Owners Group concurs with the intent of'RG 1.97, d which is to ensure that each BUR f acility -is suf ficiently instru:nented to make possible the timely and effective l assessment of plant and environmental conditions'during and '
following an accident. l The Owners Group also recommends implementing the partic - ular variables and instrumentation requirements of RG 1.97, except in those instances when deviations from the RG l'.97 positions are-indicated, are desirable, are'in-accord'with the intent of RG 1.97, and are technically' justifiable. -The' exceptions noted by the Owners Group are generally derived' from the incompatibility of an RC 1.97 requirement with the intent of RG 1.97; from evidence that the implementation of-an RG 1.97 position would not accomplish its intended objec-
] tive or that the consequence of its' implementation would be
{
, undesirable from a safety point of view; or from the availa--- a if bility of more effective or more practical ways of' achieving.
a particular monitoring activity. I l 59 f 't
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