ML20067D348
| ML20067D348 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 11/23/1982 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20067D346 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8212130221 | |
| Download: ML20067D348 (92) | |
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UNITED STATES OF AMERICA NUCLEAR REGULAT0RY C0MMISSION
- In the. matter of ~ -) * - : ' ~ - -
General Electric Company ) Docket No. STN 50-447 -
Standard Plant )
AMENDMENT NO. 10 TO APPLICATION FOR REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No. 10 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No.10 consists of two parts, a non proprietary portion and a portion considered by the General Electric Company to be proprietary.
The pages considered to be proprietary are so marked and are transmitted
. under separate cover.
Amendment No. 10 further amends GESSAR II by furnishing Appendix 1D; Assessment of Regulatory Guide 1.97, Revision 2 against GESSAR II Design.
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<c- Respectfully submitted,
, General Electric by: s/G. G. Sherwood -
Glenn G. Sherwood, Manager -
Nuclear Safety & Licensing Operation t
l STATE OF CALIFORNIA ) ss:
l COUNTY OF SANTA CLARA)
On this 23 day of November in the year 1982, before me, Karen S.
Vogelhuber, Notary Public, personally appeared Glenn G. Sherwood, personally proved to me on the basis of satisfactory evidence to be the person whose name is subscribed to this instrument, and acknowledged that he executed it.
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s/K. S. Yogelhuber NOTARY PUBLIC, STATE OF CALIFORNIA Santa Clara County My Commission Expires December 21, 1984 175 Curtner Avenue San Jose, CA 95125 csc/I110814-1 8212130221 821123 PDR ADOCK 05000447 ,
, A PDR
e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of )
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General Electric Company ) Docket No. STN 50-447 Standard Plant )
AMENDMENT NO. 10 TO APPLICATION FOR REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding,
- hereby files Amendment No. 10 to the 238 Nuclear Island General Electric l
Standard Safety Analysis Report (GESSAR II).
Amendment No. 10 consists of two parts, a non proprietary portion and a portion considered by the General Electric Company to be proprietary.
The pages considered to be proprietary are so marked and are transmitted under separate cover.
O Amendment No. 10 further amends GESSAR II by furnishing Appendix 10; Assessment of Regulatory Guide 1.97, Revision 2 against GESSAR II Design.
Respectfully submitted, General Electric by: s/G. G. Sherwood Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation STATE OF CALIFORNIA ) ss:
COUNTY OF SANTA CLARA)
On this 23 day of November in the year 1982 before me, Karen S.
Vogelhuber, Notary Public, personally appeafeTdenn G. Sherwood, personally proved to me on the basis of satisfactory evidence to be the person whose
. name is subscribed to this instrument, and acknowledged that he executed it.
l' s/K. S. Vogelhuber NOTARY PUBLIC, STATE OF CALIFORNIA Santa Clara County 4 My Commission Expires December 21, 1984 175 Curtner Avenue San Jose, CA 95125 csc/1110814-1 1
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 10 INSTRUCTIONS FOR FILING AMENDMENT NO. 10 0
A tab is also included for Appendix 1D.
' Remove and insert the pages listed below. Dashes (----) in the remove or insert column indicate no action required.
Remove Insert Summary Table of Contents iv iv Appendix 1D
New Appendix 1D O
O Amendment 10 November 23, 1982
GESSAR II. 22A7007 238 NUCLEAR ISLAND Rev. 0
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SUMMARY
TABLE OF CONTENTS Chapter /
Section Title Volume 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
1 1.1.1 Type of License Required
.l.1.2 Identification of Applicant 1.1.3 Number of Plant Units 1.1.4 Description of Location 1.1.5 Type of Nuclear Steam Supply System 1.1.6 Type of Containment 1.1.7 Core Thermal Power Levels 1.1.8 Scheduled Completion and Operation Dates 1.2 GENERAL PLANT DESCRIPTION 1 1.2.1 Principal Design Criteria 1.2.2 Plant Description 1.3 COMPARISON TABLES 1 1.3.1 Comparisons with Similar Facility Designs 1.3.2 Comparisons of Final and Preliminary Information 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1 1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design J
1.4.3 Nuclear Steam Supply System Supplier 1.4.4 Turbine-Generator Vendor i 1.4.5 Consultants I 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1 1.5.1 Current Development Programs 1.5.2 PSAR Commitment Items 1.6 MATERIAL INCORPORATED BY REFERENCE 1 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7.2 Piping and Instrumentation Diagrams 1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1 1.8.1 Compliance Assessment Method 1.9 STANDARD DESIGNS 1 1.9.1 Interfaces 1.9.2 Exceptions APPENDIX 1A RESPONSE TO TMI-RELATED MATTERS 1A APPENDIX 1B ASSESSMENT OF UNRESOLVED SAFETY ISSUES 1A APPENDIX IC RESPONSE TO NRC ADDITIONAL GUIDANCE lA APPENDIX 1D ASSESSMENT OF REGULATORY GUIDE 1.97, -
O REVISION 2 AGAINST GESSAR II DESIGN 1A O
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t l I i GESSAR II 22A7007 )
238 NUCLEAR ISLAND Rev. 10 f
O i
- APPENDIX ID I ASSESSMENT OF REGULATORY GUIDE 1.97, REVISION 2 ;
! AGAINST GESSAR II DESIGN l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 O
(_) APPENDIX 1D ASSESSMENT OF REGULATORY GUIDE 1.97, REVISION 2 AGAINST GESSAR II DESIGN CONTENTS Page 1D.1
SUMMARY
1D.1-1 1D.l.1 I:.troduction ID.1-1 1D.1.2 Objective 1D.1-2 1D.1.3 Conclusions 1D.1-2 1D.2 ASSESSMENT 1D.2-1 1D.2.1 Implementation Position 1D.2-1 1D.2.2 Type A Variable Assessment 1D.2-1 1D.2.3 Variable Assessments 1D.2-2 7s ID.2.3.1 Neutron Flux 1D.2-4 k- ID.2.3.2 Control Rod Position Indication ID.2-4 1D.2.3.3 Reactor Coolant Boron Concentration 1D.2-5 1D.2.3.4 RPV Water Level Indication 1D.2-5 1D.2.3.5 RPV Pressure 1D.2-7 1D.2.3.6 Drywell Sump Level 1D.2-8 1D.2.3.7 Drywell Pressure ID.2-9 1D.2.3.8 Containment Area Radiation Monitoring ID.2-10 1D.2.3.4 Suppression Pool Water Level 1D.2-10 1D.2.3.10 Primary Containment Isolation Valve Position 1D.2-11 1D.2.3.11 Primary Containment Temperature 1D.2-12 1D.2.3.12 Primary Containment Pressure ID.2-12 1D.2.3.13 Drywell/ Containment Hydrogen Concentration 1D.2-13 1D.2.3.14 secondary Containment Area Radiation 1D.2-13 1D.2.3.15 Secondary Containment Noble Gas Effluent 1D.2-13 1D.2.3.16 Containment Noble Gas Effluent 1D.2-13 10.2.3.17 Suppression Pool Temperature ID.2-13 O
V 1D-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 CONTENTS (Continued) gg Page 1D.2.3.18 Drywell Air Temperature 1D.2-14 1D.2.3.19 Coolant Radiation 1D.2-14 1D.2.3.20 Coolant Gamma Sample 1D.2-15 1D.2.3.21 MSIV Leakage Control System Pressure ID.2-16 1D.2.3.22 RHR System Flows 1D.2-16 ID.2.3.23 RHR Heat Exchanger Outlet Temperature /
RHR Service Water Flow 1D.2-17 1D.2.3.24 RCIC/HPCS/LPCS Flow -
ID.2-18 1D.2.3.25 Standby Liquid Control (SLC) Flow / Pressure 1D.2-18 1D.2.3.26 SLC Tank Level 1D.2-18 1D.2.3.27 SRV Position Indication 1D.2-19 1D.2.3.28 Feedwater Flow 1D.2-19 1D.2.3.29 Condensate Storage Tank Level 1D.2-19 1D.2.3.30 ESF Cooling Water Flow / Temperature 1D.2-20 1D.2.3.31 Radwaste High Radioactivity Tank Level 1D.2-20 1D.2.3.32 Emergency Vent Damper Position ID.2-20 h 1D.2.3.33 Standby Power Sources ID.2-21 1D.2.3.34 Ventilation Flow Rate 1D.2-22 1D.2.3.35 Particulate / Halogen Release 1D.2-22 1D.2.3.36 Environs Radioactivity Monitoring 1D.2-22 1D.2.3.37 Meteorology 1D.2-22 1D.2.3.38 Post-Accident Sampling 1D.2-22 1D.3 REFERENCES 1D.3-1 1DA ATTACHMENT A TO APPENDIX 1D - IMPLEMENTATION POSITION FOR REGULATORY GUIDE 1.97, REVISION 2 1DA.1-1 1DB ATTACHMENT B TO APPENDIX ID - 238 NUCLEAR ISLAND TYPE A VARIABLE ASSESSMENT 1DB.1-1 1DB.1 INTRODUCTION IDB.1-1 1DB.2 APPROACH 1DB.1-1 1DB.3 RESULTS 1DB.1-1 1D-ii
GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev. 10 CONTENTS (Continued) i 4
Page 1DC ATTACHMENT C TO APPENDIX 1D - TECHNICAL DESCRIPTION FOR ENHANCED LEVEL INSTRUMENTATION SYSTEM IDC.1-1 1DC.1 INTRODUCTION
- 1DC.1-1 1DC.2-1
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IDC.2 TECHNICAL DESCRIPTION 1DC.2.1 System Description 1DC.2-1 1DC.2.2 Theory of Operation 1DC.2-2
', IDC.3 CODES, STANDARDS AND REGULATIONS 1DC.3-1
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4 22A7007 GESSAR II 238 NUCLEAR ISLAND Rev. 10 TABLES Table Title Pag 3 1D-1 Variables Assessed for Regulatory .'
i Guide 1.97 Assessment of
238 Nuclear Island 1D.4-It
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ID-2 Summary of Information Indicating Degree of Compliance of 238 Nuclear Island with 1
Regulatory Guide 1.97, Revision'.2 1D.4-5' 1D-3 238 Nuclear Island Type A Variables . ID f4-9~
_ f 1DA-1 Variables Assessed for Regulatory I. <
j Guide 1.97 Assessment of 238 Nuclear Island ^$ 1DA.1-25 1DA-2 Regulatory Guide 1.97 Summary of Requirements 1DA.1-32 1DB-1 238 Nuclear Island Anti:ipated Operational Transients 1DB.2-1 1DB-2 238 Nuclear Island Abnormal Operational
1DB.2-3 1DB-3 238 Nuclear Island Design Basis Accidents 1DB.2-4 f
1DB-4 238 Nuclear Island Special Events ". - 'lCB.2-5
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ILLUSTRATIONS f
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E APPENDIX 1D ,
as s ASSESSMENT OF c I
, REGULATORY GUIDE 1.97, REVISION 2 r AGAINST,GESSAR II DESIGN < .,
, l , lf-0 1D.1
SUMMARY
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- r. 2 This report provides an_ assessment of the 238 Nuclear Island - /,( r.
design described by GESSAR I'I against the guidelines of' Regulatory Guide 1.97, Revision 2, dated December 1980, " Instrumentation for Light Water Cooling Nucicar Power' Plant to Assess Plant and Envir-onmental Conditions During, and Following an Accident." Any odi-fications th.GESSAR reg,uired by implementation of changes I
idef.tified by this a.7se,ssment are under review.
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1D.1.1 Intro'dut ion /
Regulatory Guide 1.97 describes a method acceptable-to the NRC staff for compliance with the; Commission's regulation to provide instrumen%6tiontoponitorplantvariablesinsystemsduringand followingan-[accidEnti St[ light water cooled nuclear power plants.
Although the ibtent of the legulatory Guide has beeft, met,'several r - '
exceptions have been made to the Regulatory Guide as.vritten. E i ; />
These exceptions a#4 c.hoir ' cases are discussed further in this report. The'se excepYihns ure viewed as acceptablN means to mple- ,,
ment the, guide's intent.
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c All varipbles identified by th's regulatory guide ar/ identifie517 in ,
- !his t report. Table 1D-1 summarizes these variables'. Uowever, ,'
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sincctsome v'priables are outside of the 238 Nuclear Island', scope,
- they,are left for the Applicant to assess. j. -
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4 GESSAR II 22A7007
, 238 NUCLEAR ISLAND R;v. 10 1D.l.1 Introductio (Continued)
O Since there are othcr modifications being made to address post-TMI changes (Appendix 1A), the assessment makes use of the exist-ence of these changes and does not repeat a description of them here. j 1D.l.2 Objective This assessment is conducted in three stages. The first stage
. develops an implementation position which defines the specific requirements against which eact variable is assessed. The imple-menta tiorr' p/osition is included as Attachment A and is discussed 4
in ST/ftection 1D.2.1.
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'4 The second stage of the assessment defines the Type A variables.
, These variables are identified from a review of design basis accidents in Chapter 15 and the Emergency Procedure Guidelines.
Amoredetailed}discussionoftheTypeAvariablesisgivenin Subsection 1D.2.2, k /6 a.,
The third stage oEi the assessment is a review of each variable
// l shown on Table lD-1. For this review, statements are provided f to either justify the current design or to provide a conceptual
, definition of changes necessary to comply with the regulatory guide.
- These discussions may be found in Subsection 1D.2.3.
., s 1D.l.3 Conclusions
/ Table 1D-2 summarizes the results of the assessment. The degree of compliance is shown by the information in the table and the accompanyiq.q remarks.
1 Table 1D-2 shows that the 238 Nuclear Island, as modified by the recommenda,tions under review as a result of this assessment, fully complier with the intent of Regulatory Guide 1.97, Revision 2.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2 ASSESSMENT.
I 1D.2.1 Implementation Position The implementation position defines the requirements of the regulatory guide as modified by GE to represent an acceptable means to implement the intent of the regulatory guide. The implementation document is included as Attachment A.
The implementation position is organized in the same manner as Section C of the regulatory guide. Where other documents have been referenced in the guide, tho reference material is included instead. Where an exception or modification to the guide has been made, the information has been included in brackets. Discussion of these exceptions is included with a discussion of the variable in Subsection 1D.2.3.
The interpretation of Type A variables has been expanded to acknowledge variables considered by the Emergency Procedure Guide-A)
(_ lines (Reference 1). Emergency actions, such as manual, scrams or depressurization, were included in the review to identify' Type A variables.
Clarification to the quality assurance requirements was added to reference the General Electric Quality Assurance Program.
1D.2.2 Type A Variable Assessment Type A variables are fundamentally plant parameters needed to alert the control room operators to take safety actions by manually initi-ating a system or function which otherwise would not be automatically initiated in the course of an event. The regulatory guide does not specify Type A variables; rather it requires that each plant develop its own list of Type A variables from a review of each plant design.
For this assessment, the list of Type A variables was identified from a review of accidents described in Chapter 15 and a review of
(} the Emergency Procedure Guidelines (Reference 1). The event descriptions of Chapter 15 and the Plant Nuclear Safety Operational 1D.2-1
GESSAR II 22A7007 238 NUCLEAR ISLhMD Rev. 10 1D.2.2 Type A Variable Assessment (Continued)
Analysis (NSOA) of Appendix 15A were reviewed to determine the plant systems which require manual initiation and the key vari-ables associated with manual initiation of those systems. The Emergency Procedure Guidelines prepared by General Electric for the BWR Owners' Group were also reviewed to identify any other variablee requiring safety action and not identified by the review of the FSAR. A summary of the Type A variables are identified through this process as shown in Table 1D-3. Details of the Type A variable assessment are provided in Attachment B.
1D.2.3 Variable Assessments This section summarizes the results of the individual variable assessments concentrating on deviations identified between the existing design of the 238 Nuclear Island and the implementation position for the regulatory guide. ggg Strict compliance with the regulatory guide is not recommended in all cases. In some cases, an acceptable alternate has been pro-posed which meets the intent to have meaningful post-accident indi-cations. For some parameters, this can be met by adopting alternate variables to those specified in the regulatory guide or by specify-ing combinations of other variables. Another approach taken is to take exception to the guide where a reasonable justification can be provided. Finally, where a design change is indicated, the basic conceptual design is described as under review. When these reviews are completed, the resolution will be included in future amendments.
Many of the variables assessed do not meet the qualification requirements specified by the regulatory guide for Category 1 and Category 2 instruments. These requirements specify that the instrumentation "be qualified in accordance'with Regulatory Guide g 1.89... and the methodology described in NUREG-0588" for the 1D.2-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 10 1D.2.3 -Variable Assessments (Continued)
O environmental qualifications of safety-related electrical equip-ment. In addition, the Regulatory Guide requires that "the seismic portion of qualification should be.in accordance with Regulatory Guide 1.100."
The equipment requiring qualification is summarized in Tables 3.10-1 [MPL] (for seismic qualification) and Tables 3.11-9 [MPL]
(for Environmental Qualification). The environmental qualification is to be in accordance with IEEE 323-1974 methods and the envelopes summarized in Tables 3.11-2 through 3.11-8. The seismic qualifica-tion is to be in accordance with IEEE 344-1975 methods. A summary of the qualification methods and results of the qualification evaluation is to be provided by the Applicant. Tables 3.10-1 and 3.11-9 have been reviewed in the course of this assessment to deter-mine whether additions to the equipment lists are needed because of the requirements of Regulatory Guide 1.97. Where changes are needed, a. statement to that effect is included in the individual paragraphs that follow. Qualification of these additional instru-ments is the responsibility of the Applicant.
Where performance requirements are required to be specified -(imple-mentation position paragraphs C.1.b and C.2.4) , the accuracy for various components of each instrument channel have been assessed as l they apply to normal operation. Assessment of any potential instru-I ment accuracy variation due to post-accident environments and their relationship to the accuracy needs of the instruments during post-accident periods has not been addressed as part of this appendix.
Such an evaluation should be made as part of the environmental qualification program for the post-accident instrumentation and is the Applicant's responsibility to address. As a minimum, Type A variables should be evaluated in this regard.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.1 Neutron Fluy. g The current neutron monitoring system consists of the source range monitor (SRM), intermediate range monitor (IRM), and the power range monitor (LPRM/APRM). While significant redundancy exists in the design, the current design does not meet the require-ment from Subsection C.l.3.1.b to be "... electrically independent and physically separated from each other and from equipment not classified important-to-safety in accordance with Regulatory Guide 1.75." In addition, because the SRM and IRM must be inserted into the core following a reactor scram and because the drive equipment is not likely to survive a post-accident environment for seismic disturbance, the SRM and IRM do not meet the requirements for avail-ability to monitor the variable during the time of interest.
General Electric is currently evaluating the design of a system which will comply with the criteria specified in Regulatory Guide 1.97. The system, called the Wide Range Neutron Monitoring lll (WRNM) System, will be a replacement for the SRM and IRM indica-tions. Subsequent review and acceptance by the NRC will be required after the system design is finalized.
1D.2.3.2 Control Rod Position Indication Control Rod Position Indication is classified as Category 3 by the Regulatory Guide because it serves as a backup to neutron flux.
Its post-accident monitoring function is only to verify function of a reactor protection system and, consequently, this function is only required for a brief period of time. Because of these reasons, the equipment function time and local environment are not specified for the control rod position indication as called for by the imple-mentation position. No recommendations for change are made for Control Rod Position Indication.
O 1D.2-4
, GESSAR II 22A7007 238. NUCLEAR ISLAND Rev. 10
() Reactor Coolant Boron Concentration 1D.2.3.3 i
Boron concentration in the reactor water is determined by analysis of a reactor water sample obtained from the post-accident sample station (refer to Subsection 11D.2.3.38). Recommended boron analy-sis procedures are included with the operation and maintenance manual supplied with.the sample station. Actual' procedures are
! left for.the applicant to provide.
1D.2.3.4 RPV Water Level Indication i i
The existing post-accident. water level indication system consists of three wide-range instruments (calibrated for full pressure) ccvering the range -160" to +60" and two fuel zone instruments (calibrated for atmospheric pressure) covering the range -326" to
-116". Narrow-range indication is not included in the assessment.
.() The RPV water level indication system is the primary variable indicating the availability of adequate core cooling and is con-sidered acceptable, without diverse methods of indication, pro-vided adequate redundancy and unambiguity are provided over the
- entire range of interest.
The range for coolant level in the reactor specified on Table 1D-1 of the implementation position has becn modified to require indi-cation only "above normal reactor water level" rather than the
" centerline of main steam line" as specified in the regulatory guide. This change was made because the high reactor water level transient has been shown to be of not great concern (Subsec-tion 1A.23) especially considering the existence of improved high i reactor water level trips on HPCS. Furthermore, since the presence of reactor water on the normal range assures adequate core cooling, higher ranges of indication are not needed for this purpose.
O 1D.2-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.4 RPV Water Level Indication (Continued) llh The requirement in Table 1D-1 of the regulatory guide for BWR core thermocouples has been replaced with a requirement that the reactor water level indication should extend from below the core plate to above the pressure vessel Level 1 as adequate indication of core cooling. Through work by General Electric and BWR Owners' Group (References 2 and 3), the NRC staff has recently retracted the requirement for BWR core thermocouples (Reference 4), but they have requested that an investigation continue to identify an acceptable diverse means of indicating adequate core cooling.
This investigation is still in progress. Work by General Electric and the BWR Owners' Group (Reference 3) has provided the NRC staff with extensive information on the relationship between the reactor water level and adequate cooling of the core. These studies have shown that as long as at least one of the water delivery systems is available and flowing to the reactor pressure vessel, that ade-quate core cooling exists. Indication of these flows is already lll required by the Regulatory Guide. Consequently, it is General Electric's view that no instrumentation other than RPV level indi-cation is required to assure indication of adequate core cooling.
Subsection C.l.3.1.b states that additional instrumentation should be provided to allow the operator to determine the actual condi-tions of the plant "when failure of one accident monitoring channel results in information ambiguity." Three independent channels of indication are considered acceptable to meet this provision.
During a small-break accident which results in increasing drywell temperatures, the accuracy of the RPV water level indication varies significantly. This effect is described in detail in Reference 2.
A small drywell break could lead to ambiguity in all instrument ranges either because the redundant channels would not agree (if lll one failed) or because of increasing drywell temperatures and its 1D.2-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.4 RPV Water Level Indication (Continued) effect on vessel level indication. In addition there could be ambiguity in-the overlap region between fuel zone and wide-range indications due to differing calibration conditions and the dry-well temperature effects.
t These drywell temperature effects are minimited by specifying a minimum vertical elevation drop in the drywell. However, they are not eliminated and some disagreement between channels can be expected.
In addition to not having three independent. channels the fuel zone instruments also use nondivisional power as. compared with station
, standby power as required by hubsection C.l.3.1.c.
In order to resolve the redundancy and ambiguity problems discus-sed above, and because of the importance of RPV water level to the verification of adequate core cooling, a three-divisional, unam-biguous system of measuring RPV water level is under review. The Enhanced Water Level Indication (ELI) System, which is described in detail in Attachment C, provides three Class lE channels of water level indication spanning the full range of required reactor water 4 level measurement from the bottom of the core support plate to the top of the current wide indicators. The system uses microprocessors to provide temperature and pressure compensation and to provide a variety of alarms for off normal conditions such as reference ,
leg boiling / flashing or leaking equalizer valves.
1D.2.3.5 RPV Pressure Two recorded channels of RPV pressure indication are provided in the current design by pressure transmitters connected to the same 1 g- reference leg as is used for the reactor vessel water level indi-i k/ cation. Depending on the location of the break, a small-break i
1D.2-7
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.5 RPV Pressure (Continued) h accident in the instrument line, combined with a power supply failure .in the redundant channel, could prevent the operator from having unambiguous indication of reactor pressure. (Nonsafety-related indic ation of both reactor water level and reactor pres-sure might be available associated with the feedwater control system indication.) Because RPV pressure is associated with required Manual Safety actions (Type A) unambiguous indication should be made available to meet Subsection C.l.3.1.b of the guide. The ELI design discussed in Attachment C also includes digital indication of a third channel of reactor pressure. This third digital channel has a useful operational benefit for veri-fication of low pressure ECCS injection and for RHR shutdown cool-ing system initiation permissives.
1D.2.3.6 Drywell Sump Level O
The current design consists of twu drywell sumps: equipment drain sump for identified leakage, and floor drain sump for unidentified leakage. The monitoring instrumentation is shown in Figures 7.6-12c and 7.6-12d.
An exception is made to the regulatory guide as written for the design category on this variable. Rather than Category 1, General Electric's position is that Category 3 design requirements are more appropriate for the following reasons:
Indication of drywell sump level provides post-accident monitor-ing of drywell leakage and may be an early, indication of a very small drywell break. However, it is primarily a backup variable to other indications of drywell breaks such as drywell pressure (Subsection 1D.2.2.7) or drywell radiation level (Subsec-tion 1D.2.3.8). As such a lower design classification is appropriate.
1D.2-8
GESSAR II 22A7007 238 MUCLEAR ISLAND Rav. 10 1D.2.3.6 Drywell Sump Level (Continued)
Both drywell sump dischargec are automatically isolated by the primary containment isolation design (Subsection 6.2.4). Thus the only period of meaningful post-accident information is prior to the sump filling to the top. Because of this relatively short duration for any but very small breaks, providing instrumentation to meet the higher design categories is not considered necessary.
In comparison with Category 3 design requirements, the existing design is incomplete in terms of range of indication only for_the equipment drain' sump. The floor drain sump recorder adequately monitors sump level and rate of change of level from the bottom to the top of the sump as recommended by the guide. A modification is under review which would add sump level indication to the equip-ment drain sump similar to that used on the floor drain sump.
(Figure 7.6-12c).
1D.2.3.7 Drywell Pressure Two channels of drywell pressure indication are provided in the control room with a range consistent with that identified in the regulatory guide. These instruments are separate from the narrow-range instruments used for the reactor protection system.
These two channels provide adequate redundancy to provide unam-biguous indication, even in the event of failure of one of the channels, since diverse indication is provided to indicate a breach of the reactor coolant pressure boundary. These diverse ,
indications include RPV Pressure (Subsection 1D.2.3.5) and 1 Suppression Pool Water Level (Subsection 1D.2.3.9).
No modifications to the 238 Nuclear Island design Gre necessary.
The pressure transmitters and indicators, however, should be added O
1D.2-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.7 Drywell Pressure (Continued)
O to Tables 3.10-1 (T41) and 3.11-9 (T41) to ensure their qualification.
1D.2.3.8 Containment Area Radiation Monitoring Area radiation levels at different locations in the containment are defined in Section 12.3. The specific location and I :e of area radiation monitors is the responsibility of the Applis ant to provide.
1D.2.3.9 Suppression Pool Water Level Indication of water level in the suppression pool is provided by four physically separated (one per pool quadrant) sensors powered from Division 1 and 2 Class lE power. Additional discuncion may be found in Subsections 6.2.7.5, 7.3.1.1.6, and 7.3.2.6.1.
An exception to the range specified in the regulatory guide is taken for this variable. The 238 Nuclear Island design provides indication from about -16'9" elevation to -4'9" elevation. This 12 ft range extends from about 1 foot above the top drywell vent to near the top of the weir wall. For Type C variables (to detect a breach of the coolant pressure boundary) the guide speci-fies that the range extend from "the bottom of ECCS suction line to 5 feet above normal water level."
This lower level corresponds to about a -27 ft elevation and is about 9 ft below the top drywell vent. Monitoring water level to this low level is considered unnecessary because of the presence in the 238 Nuclear Island design of the upper pool dump feature and the physical arrangement of the plant design which precludes lower water levels.
O 1D.2-10 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 10 1D.2.3.9 Suppression Pool Water Level (Continued)
O)
\~
For Type D variables (to monitor operation) the guide specifies
" top of vent to top of w7ir well." The 238 Nuclear Island design is consistent with the. range specified to monitor operation of the system.
The upper range is an appropriate upper bound since higher levels would start to overflow into the drywell. Containment flooding, as specified by the Emergency Procedure Guidelines (EPG) (Reference 1, step SP/L-3.5), specifies a maximum primary containment water level which is somewhat higher than the range provided by the current design of level indication. However, in practice, sources of water from external to the primary containment would not be used as called for in EPG step SP/L-3.1. Thus the likelihood of reaching the.
maximum containment water level is very remote and higher ranges of water level indication are not needed.
() 1D.2.3.10 Primary Containment Isolation Valve Position The primary containment isolation valves reviewed in this assess-ment were identified from a review of Table 6.2-25. All valves providing a containment isolation function (except check valves) were included.
A detailed review of each valve to ensure that valve position indi-cation is provided was not conducted because there are other requirements related to the NRC's Standard Review Plan 6.2.4 which require the availability of control room indication of primary con-tainment isolation valve position. Valve position indication is, in general, provided by limit switches attached to the valve opera-tors which provide a signal to open and close indicating lights on the control room panel.
Containment isolation valve limit switches should be included or
() referenced by Tables 3.10-1 and 3.11-9 to ensure qualification of the switches.
1D.2-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 10 1D.2.3.10 Primary Containment Isolation Valve Position (Continued) g Implementation position Section C.l.c which is taken from ANSI 4.5 deals with the duration time for equipment qualification. This paragraph states that " shorter qualification times are acceptable if equipment replacement or repair can be accomplished within an acceptable out-of-service time." The primary containment isola-tion valve position indicating lights in the control room are easily replaceable in a short period of time and a redundant (open or close) indication is provided. Because of these considerations, the valve position indicating lights are viewed as not requiring any specific qualification to meet the intent of the regulatory guide.
1D.2.3.ll Primary Containment Temperature This variable is not specified by the regulatory guide. However, it has been included as a Type A variable because initiation of the ggg containment spray system is specified by the Emergency Procedure Guidelines (Reference 1), step CN/T) in response to a high indica-tion of this variable. The 238 Nuclear Island design provides two channels of indication on the control room BOP benchboard. The design is shown in Figure 9.4-6.
No modifications are needed to this design. However, the instru-monts including displays should be included in Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.12 Primary Containment Pressure Containment pressure is indicated by two channels of indication as shown in Figure 9.4-6.
The instrumentation provides redundant indication up to twice the design pressure of the containment building. This range is con-sidered adequate to accomplish the functions specified in the Emergency Procedure Guidelines especially con idering that a diverse indication of containment temperature is provided.
1D.2-12
r GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 I^'i V
1D.2.3.12 Primary Containment Pressure (Continued)
Two channels of pressure instrumentation are adequate in case of a single failure because diverse indications to monitor containment integrity are provided.
No modifications are needed for this instrumentation. However, the instruments including displays should be included in Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.13 Drywell/ Containment Hydrogen Concentration Drywell and containment hydrogen concentration may be determined by analysis of samples obtained from the post-accident sample station (Subsection 1D.2.3.38). On-line instrumentation is the responsibility of the Applicant to provide.
1D.2.3.14 Secondary Containment Area Radiation
( ])
Area radiation levels in the secondary containment are defined in Section 12.3. The specific location and range of area radiation monitors tre the responsibility of the Applicant to provide.
1D.2.3.15 Secondary Containment Noble Gas Effluent
! Applicant to provide.
1 1D.2.3.16 Containment Noble Gas Effluent l
l Applicant to provide.
I 1D.2.3.17 Suppression Pool Temperature Two temperature sensors per quadrant of the suppression pool are j ( )f provided in the 238 Nuclear Island design with control room 1
1D.2-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.17 Suppression Pool Temperature (Continued) 4.ndication and recording. The instruments are discussed in Subsections 7.6.1.11 and 7.6.1.12.
The sensors are located in the upper third of the suppression pool and thus provide a conservative indication of suppression pool temperature for use on the Emergency Procedure Guidelines.
No modifications are needed to these instruments. The instruments including the di splays should be included in Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.18 Drywell Air Temperature Two Class lE channels of drywell temperature indication are pro-vided on the control room BOP benchboard (P870) as shown in Fig-ure 9.4-5.
The range of the display (up to 400 F) is adequate to carry out the functions prescribed by the Emergency Procedure Guidelines (Reference 1).
No modifications to the design are needed. The instruments including displays should be included on Tables 3.10-1 ar.d 3.11-9 to ensure qualification.
1D.2.3.19 Coolant Radiation No instrumentation is provided in the current design to monitor radioactivity levels in the primary coolant and no changes to the plant design are planned.
The specified range for the potential instrument (1/2 Technical Specification Limits (TSL) to 100 times TSL) suggests that the 1D.2-14
-GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. .10 1D.2.3.19 Coolant Radiation (Continued)
(~/)
purpose of this instrument is to assess coolant radiation level during routine plant operation. The current design provides sampling capability for reactor coolant as described in Subsec-tion 9.3.2 and provides offgas and mainstream line process radi-ation measurement as discussed in Section 11.5 for detection of fuel cladding branches.
The value for the technical specification limit has not been established by the staff in standard technical specifications for BWR/6. Subsection 16.3/4.4.5, however, indicates that the TSL is 2 pCi/g Iodine-131 equivalent. On-line reactor coolant monitoring of this level of coolant activity may be impractical during normal operation because of the additional contributions to the detector from other isotopes such as circulating Nitrogen-16 or Cobalt-60 deposited on reactor coolant piping.
b) v Furthermore, should a significant breach of the fuel cladding occur the expected levels of iodine in the reactor coolant would far exceed the upper range specified by the regulatory guide for this instrument. The samples provided by the post-accider', sam-ple system-(Subsection 1D.2.3.38) will provide quantif'. cation of
- the coolant radioactivity level. Under these condicions, an on-l line monitor would serve no useful purpose.
~
l 1D.2.3.20 Coolant Gamma Sample The radioactivity content of the reactor ws.ter is determined by i
analysis of a reactor water sample obtained from the post-accident i
i sample' station (Subsection 1D.2.3.38). Recommended procedures to I
l determine the gross activity in the coolant are included with the I
cperation and maintenance manual supplied with the sample station.
Actual procedures are the responsibility of the Applicant.
(
u 1D.2-15 l
i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.21 MSIV Leakage Control System Pressure h The current d'osign uses a Class lE positive leakage control system as described in Section 6.7. Proper system function is monitored and recorded by air system flows rather than system pressure as specified by the regulatory guide. System isolation automatically occurs on high flow or low differential pressure between the RPV and the pressurized lines.
The flow monitoring instrumentation is considered adequ,ite to meet the intent of the regulatory guide to indicate proper system function. No changes are planned for this system.
1D.2.3.22 RHR System Flows The RHR System serves a variety of functions among them being low pressure coolant injection, containment spray, suppression pool cooling, and shutdown cooling. The valving arrangements (refer to llh Figure 5.4-12a) required to achieve these different functions of the RHR System occur downstream of the flow element and flow trans-mitter which is used to indicate RHR System flow. This instrument channel therefore provides the operator with flow indication during any of these operating modes for the RHR System.
From an operational point of view, proper functioning of the con-tainment spray mode of the RHR System is provided by the contain-ment temperature (Subsection 1D.2.3.ll) or containmer.t pressure instrumentation (Subsection 1D.2.3.12). Should the containment spray mode be used, it is anticipated that the operators would only initiate flow long enough to decrease these containment parameters at which time flow in the containment spray mode would be termi-nated. Thus, the primary indicator of proper containment spray mode operation is the containment pressure or temperature indica-tion rather than the RHR System flow.
9 1D.2-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10
/N 1D.2.3.22 RHR System Flows (Continued)
(_)
Indioolion of proper LFCT operation is provided by a combination of RHR System flow and valve position indication associated with
, the injection lines. In addition to the RHR System flow and valve position, an increasing trend on reactor water level indi-cation (Subsection 1D.2.3.4) provides an operator with knowledge of proper LPCI mode function.
No change to RHR System flow indication is planned as a result of this assessment.
1D.2.3.23 RHR Heat Exchanger Outlet Temperature /RHR Service Water Flow As shown in Figure 5.4-12a the RHR heat exchanger outlet tempera-ture is monitored by temperature elements on the RHR System side
(~} and on the service water side of the heat exchanger.
s-Display of these temperatures is on a commercial quality "Westron-ics" recorder located in the control room. This instrument does not meet the qualfication standards indicated for Category 2 instruments and and is not included on Table 3.11-9-E12 as needing qualification.
4 As an alternate to RHR heat exchanger outlet temperature, the RHR service water flow is monitored and displayed in the control room, as shown on Figure 5.4-12a. Monitoring of this variable is an acceptable alternate to outlet temperature since it is an unam-biguous indication of heat remeval from the RHR primary side especially considering that heat exchanger fouling is provided for s in the design. These instruments are included in the List of Specified Instruments (Table 3.ll-9-E12).
Monitoring RHR service water flow also satisfies the need for ESF flow monitoring (Subsection 1D.2.3.30). The RHR service water flow 1D.2-17 i
e, , - -
GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 10 1D.2.3.23 RHR Heat Exchanger Outlet Temperature /RHR Service Water Flow (Continued) indication has been assumed to satisfy part of that requirement for purposes of this study.
1D.2.3.24 RCIC/HPCS/LPCS Flow Each of these systems is indicated on the control room by a single channel flow indication system meeting the range specified. No changes are planned for these instruments. The LPCS flow indi-cator (E21-R600) should be added to Table 3.11.9 (E21) to ensure its qualification.
1D.2.3.25 Standby Liquid Control (SLC) Flow / Pressure No flow indication is provided in the 238 Nuclear Island design.
The positive displacement SLC pumps are designed for constant flow as described in Section 9.3. This flow is periodically tested to g ensure it is at the rated value. Any flow blockage or line break would be indicated by abnormal system pressure (high or low) follow-ing SLC initiation. The 238 Nuclear Island design includes a safety-related channel of SLC pressure. Changing neutron flux (subsec-tion 1D.2.3.1), SLC pressure, and squib value position are con-sidered adequate to verify proper system function. Because of these reasons no flow indication is needed and no modifications are planned. SLC pressur; and squib valve position are thus considered adequate to meet the regulato'y guide as defined in Section C.2.2 of the implementation position. The SLC control room pressure indicator should be included on Tables 3.10-1 and 3.11-9 to ensure qualification.
1D.2.3.26 SLC Tank Level The 238 Nuclear Island design consists of commercial quality air-powered " dip tube" instrumentation which monitors SLC tank level over the range specified by the regulatory guide. The instrumenta-tion is shown in Figure 9.3-5.
1D.2-18
I GESSAR II 22A7007 238' NUCLEAR ISLAND Rev. 10 i
, 1D.2.3.26 SLC Tank Level (Continued)
.An exception is taken.to the design. category specified by the regulatory guide. ' Category 3 requirements are considered adequate for this variable since it serves a backup function to other parameters:used to monitor system operation-(Subsection 1D.2.3.25).
f No modifications are needed for the SLC tank level indication channel to meet Category.3 design requirements.
The air supply to the SLC level indication is provided from the instrument air distribution system shown in Figure 9.3-2. The isolation requirements of this air line are currently under review in connection with the design to accommodate anticipated transients without scrams to ensure its availability.
1D.2.3.27 SRV Position Indication Postive indication of SRV open and closed position is provided by r the SRV open/ closed monitoring system and is discussed in Section lA.24.
No modifications to the system as described are planned. The pressure switches and computer logic should be added to Table 3.11-9 (B21) for proper qualification.
1D.2.3.28 Feedwater Flow Feedwater ficw indication is provided by recording the current design as shown in Figures 5.1-3d and 7.7-6.
l No modification to the instruments as described are needed.
However, since this variable does not indicate-the " operation of a safety system or other systems important to safety" its classi-fication as a Type D variable is not justified.
1D.2.3.29 Condensate Storage Tank Level
~
Applicant to provide.
1D.2-19
{
l _ _- _ - _ - _ _ _ , _ . _ . . _ _ . -- -- . _ - -- .- - . - - - - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.30 ESF Cooling Water Flow / Temperature gg Cooling water flows and temperatures at all ESP components is indicated by local instruments for the purpose of achieving system balancing (Section 9.2). Qualified control room indication of total system flow and temperature should be provided by the appli-cant. For compliance with this part of the regulatory guide these instruments are outside of the 238 Nuclear Island scope of supply.
The RHR heat exchanger flows and temperatures are discussed in Sub-section 1D.2.3.21.
1D.2.3.31 Radwaste High Radioactivity Tank Level The current Radwaste System design is described in Section 11.2.
Level indication of three high conductivity tanks and two low conductivity tanks is provided by recorders in the radwaste control room.
No modifications are necessary to this instrumentation to meet the Regulatory Guide 1.97 criteria. However, since the Radwaste Sys-tem is neither a safety system nor a system important to safety, its classification as Type D is not justified.
1D.2.3.32 Emergency Vent Damper Position Indicating lights showing the vent damper position on all supply and exhaust paths uced during normal operation for the Containment, Auxiliary Fuel buildings are provided on the control room BOP benchboard. Radwaste building ventilation supply and exhaust damper position is indicated in the radwaste control room.
No modifications are needed to comply with the criteria of the regulatory guide. The position indicator switches should be included on Tables 3.10-1 and 3.11-9 to ensure qualification.
O 1D.2-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.33 Standby Power Sources
~ !
The 238 Nuclear Island design includes control room indication of electrical, air, and liquid power supply systems.
The electrical power supply instrumentation is shown on Figures 8.3-2, 8.3-3, and 8.3-15. Safety-related control room display (voltage and amperes) of three 6.9 kV ESF buses and four 480V Class lE buses supplied from 6.9 kV feeders (voltage only) is provided.
In addition, the design includes voltage display of four Class lE 125V direct current buses. Control room indication of the elec-trical operation (amperes, voltage, watts, vars, and frequency) of each of the three diesel generators is also provided. No changes are needed to provide indication of the electrical system status during post-accident periods.
Air supplies to safety-related valves and functions are provided by the Pneumatic-Supply System discussion in Section 6.8. Non-(^T safety-related valves and instruments are powered by the com-pressed air systems described in Subsection 9.3.1.
Only the safety-related portion of the Pneumatic Supply System needs to be monitored under post-accident conditions. The safety-related functions are identified on Table 6.8-1. Two channels of control room indication are provided on the 238 Nuclear Island design. The pressure transmitters and indications should be included j on Tables 3.10-1 and 3.11-9 to ensure qualification.
l The air supply to the Air Positive Seal Isolation Valve Leakage Control System and the water supply to the Water Positive Seal Isolation Valve Leakage Control System are described in Sub-section 6.5.3.3.
The Positive Seal Leakage Control Systems provides a backup func-tion to the Primary Containment Isolation System which is monitored
(~}
v by the individual isolation valve position indications (Subsec-tion 1D.2.3.10). For this reason Category 3 design requirements are applied to these channels.
1D.2-21 L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1D.2.3.34 Ventilation Flow Rate gg Post-accident ventilation from the auxiliary, fuel, and shield building air spaces is through the Standby Gas Treatment System (SGTS). All other exhaust flows (except for those from the Rad-waste Building) are isolated. Consequently only SGTS flow needs monitoring.
No instrument to specifically monitor SGTS flow is provided in the 238 Nuclear Island design except for isometric probes on the SGTS vent which are the responsibility of the applicant to provide.
The design does provide safety grade indication of SGTS damper position on the control room BOP benchboard: Radwaste ventilation damper position is provided in the Radwaste Control Room. These position indications are considered adequate to meet-the intent of the regulatory guide. ggg 1D.2.3.35 Particulate / Halogen Release Applicant to provide.
1D.2.3.36 Environs Radioactivity Monitoring Applicant to provide.
1D.2.3.37 Meteorology Applicant to provide.
1D.2.3.38 Post-Accident Sampling Post-accident samples are obtained from the post-accident sample station (PASS) designed to meet NUREG 0737 requirements. The PASS is described in Subsection lA.21. ll) 1D.2-22
Y GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 4
1
() 1D.2.3.38 Post-Accident Sampling (Continued) j An exception is taken to the areas requiring sampling specified j by the regulatory guide. Samples of drywell, containment, auxil-lary or fuel building sumps are not provided for in the current design and no modifications ~are planned. This position is justi-'
fied because no useful information such as the extent of core damage, or for release assessment would be obtained through such samples.
Flow to these sumps is primarily from reactor water leakage either planned or unplanned from breaks. The PASS samples reactor water directly thus provides an indication of the extent of possible core damage. Leaks of radioactive liquid from the systems which could cause increases in sump level are primarily detected by area and process radiation monitors (Subsections 1D.2.3.8, 1D.2.3.14 and 1D.2.3.15). These indications are a direct indication of poten-() tial gaseous offsite impact.
These sumps are automatically isolated by a containment isolation signal (Subsection 6.2.4). Since any potential release is' con-tained by this action, liquid release outside the contained areas i is not likely. Because of these reasons samples of the sumps are f not necessary.
I In addition to being unnecessary, sump samples may be imprachical
! because of the small sample volume (0.1 ml) and the high particu-I late content expected in any sump. Without extensive filtering, clogging of the liquid sample ball valve is a possibility.
i
()
1D.2-23/lD.2-24 1
i GESSAR II 22A7007 4
238 NUCLEAR ISLAND Rev. 10
() 1D.3 REFERENCES
- 1. -Emergency Procedure Guidelines - BWR l-6, January 1981
] (NEDO-2 4 934 ) .
- 2. Review of BWR Reactor Vessel Water Measurement, April 1980
_ (NEDE-2 4 801 ) (Proprietary).
- 3. Additional Infor1..ation Required for NRC Staff Generic Report-on Boiling Water Reactors Volume 1, August 1979 -(NEDO-24708A) .
- 4. NRC Staff Recommendations on the Requirements for Emergency Response Capability, March 10, 1982.
i 1
4 4
t
(:)
I i
j 1D.3-1/lD.3-2 i
% j .,
GESSAR II
'V22A7007 238 NUCLEAR ISLAND Y -
, Rev.Ji0 r^
e s g:
,, 3
..O~ Table' 1D-1 7 ;' . n%
VARIABLES ASSESSED FOR REGULATORY GU10E'1.97 ASSESSMENT ~
'OF.238 NUCLEAR ISL,AND # . L4 f -
./
u -i
- Discussion.
Variable Type * 'dat3 dory * ' SubsSction
~~
~
Reactivity Control
-%i c
?
/
3' Neutron Flux .
A,B 1 T 1D.2.3.I F (value,. rate, trend) e'
< ; -'-d y~
Control Rod Position- B 4 3 'C q
10.2.3.2 ,
r g
Boron Concentration B ,3 y 1D.2.3.3 (sample) , *. .
~
- x y '
Core Cooling 4-h.,, ,
d*?N Coolant Level in the A,B,C 1 1D.2.3'.'4 5 .
Reactor (value, trend)
"d" y 4 bt
, p p Maintaining Reactor Coolant ," s F '
F-
\d System Integrity
-- r
- - '* s 1D.2..,3.5 RCS Pressure A,B,C' l.
(value + alarm) .(? ,.
vm Drywell Sump Leve1 B,C ,3 .3 1D.2.376- '
- f. .
(value + alarm) ~s i y
Drywell Pressure B,C,D '1 , 2 .'1D'I.3.7 J' i'[
(value + alarm) , , , 24 9
- ~
y Primary Containment E 1- 9' 7 1D.' 2. 3. 8 '
f .iw Area Radiation s'
C 3 '. b > ^
~
- g , w _. ,
i Suppression Pool A,C,D 1,2 cg ^yD.2.3'.9
- Water Level # F c-a ' ~ '&
s V v ,
Maintaining Containment .y , -
Integrity i J [V.' F- itit ..y Primary Containment B 1 3 1D.2.3,.10t '
g Isolation Valve Position .t .
(Excluding Check Valves). '(R[
,' ,w ,
p Primary Containment A 1 1D.2.3.11 + ,'
! Temperature l *As defined by Attachment A , _
t 1D.4-1 4 -,,j wm e -- + - - , . -
. . , , ,, ,v,. ,- e- , - .
s GESSAR II 22A7007 238 NUCLEAR ISLAUD R;v. 10 Table lD-1 -
, VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued)
Discussion Variable Type * ,; Category
- Subsection Maintaining Containment Integrity 3 (Con tinued)
A,B,C 1_ 1D.2.3.12 Pressure ~~
(va lue , rate, tr6nd, s.; '
+ glarm)
Dryweli/ Containment A,C * ^
l 1D.2.3.13 Ilydrogen Concentration >
~lvalue)
'..m.-
Sfcq[da2y Containment C,E -2' 1D.2.3.14
- Area Hadi~ation
'~'
(valuel Secondary Containment C,E 2 - 1D . 2. 3.15 Noble Gas Effluent -
~
3 - 1D.2.3.16 O Noble Gas Effluent Suppression Pool , A,D 1,2 16.'2.3.17 Temperaturn. ,.
n ,-t ,
jf 1D-2.3.18 Dry,well Air Temperature A,D 1,2 Fuel Cladding Barrihr .
Monitgring , -
CoolabbRadiation NA '
NA 1 D '. 2. 3 419 I
(value +! alarm) -[
Coolant GeNma C ', ,
3 1D.2.3[20 (1 sample /?6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) " '
results within 72 hr System Operation (
, ,. . n' Ur, o Main Steam Line Isolation D , .
I# ~ 2 1D.2.3.21 Valve Leakage Control \. '
System Pressure
/
Containment;9 pray D 2 1D.2_3.22 Flow 'i '
.?-
- As defined by' Attachment A j
k 3 x p 1D.4-2 , .
- e. , i.
s *
. . =_
- # a .
.1
);; ,
) /- GESSAR'II - '22A7007
< 238 NUCLEAR ISLAND R v. 10
/ .a .
s -
Table 1D-1 *
(
VAREABLES ASSESSED'FOR REGULATORY GUIDE 1.37 ASS 3SSMENT OF 238 NUCLEAP ISLAND (ContinU2d) )~
Discussion Variable Type
- Catagory* -Subsection
), o. .
Gifhem Operation '
Continued)
- e.' a Residual Heat Removal D 2 1D.2.3.22 (RHR) Sy, stem Flow I' P
RHR Service-Water Flow D 2 1D.2.3.23 Low Pressure Coolant D 2 1D.2.3.22 Injed' tion System Flow ,
R"eactor Core Isolation D 2 1D.2.3.24 Coiling System Flow
.s . ~
Hich Pressure' Coolant D 2 1D.2.3.24 Spjay. System Flow
~
'^
Core, Spray System D 2 1D.2.3.24
- Flow Standby Liquid Control D 2 1D.2.3.25 ,
System (SLCS) Flow o
SLCS Storage Tank Level D 3 1D.2.3.26 .
SRV Position D 2 1D.2.3.27 ~
Fee 6waterFlow$ t D 3~ 1D.2.3.28 9
'I CST Level D 3 1D.2.3.29 1D . 7,. 3. 3 0 ESF Cooling Water Flow D' 2 g ESF Cooling Water D' ( 2 ,rlD.2.3.30 Temperature , .
High Radioactivity Tank D 3 1Dc2.3.31 Level ,
1 Emergency Vent Damper D 2 l Position ;f D. 2.3.32 4
- Standby Energy Status D 2 1D.2.3.33
- As defined by Attachment A 1
1D.4-3 4
/
GESSAR II 22A7007
. /. 238 NUCLEAR ISLAND R r>v . 10 a t
' , -' - g I ,) Table 1D-1 VARIAB:IES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT n OF 238 NUCLEAR ISLAND (Continued)
/' - Discussion Va iable Type
- Category
- Subsection Effluent Monitoring SGTS Ventilation Flow E 2 1D.2.3.34 Rate Other Ver.tilation Flow E 3 1D.2.3.34 Rdtes ,
~3 Particulate /IIalogen E 3 1D.2.3.35 Release (sample) ,
Environs Radioactivity E 3 1D.2.3.'6 Monitoring -
Meteorology /
E 3 1D.2.3.37 i
Post-Accident'!'ampling 3 E 3 1D.2.3.38 (sample) . .
O f
/
(I i
I
- As defined by Attachment A O
1D.4-4 I
1 .
3-i J T
I i
l' i Table 1D-2 4
SUMMARY
OF INFORMATION INDICATING DEGREE OF COMPLIANCE OF 238 NUCLEAR ISLAND l WITH REGULATORY GUIDE 1.97, REVISION 2
)
0"'l l i cation Quality Control Room Comments /
i VarAable Type Environment Seismic Assurance Fedundancy Range Power supply Display Notes i
Reactivity Control
) Neutron Flux A,B Power Range RG 1.89 RG 1.100 Yes 4 ch nnels 1012nv to 10I4nv IE Unint DC Recordsr A j int Range RG 1. 8 9 PG 1.100 Yes 4 channels 10'nv to 10II nv RPS Recorder A,8 Source Range RG 1. 8 9 RG 1.100 No 4 channels 10 nv to 10'av RPS Recorder A,8 j control Rod Position B N/A N/A N/A 1 chanrel Full in/ full out Untnterruptible Full core i, Boron Concentration 8 N/A N/A N/A N/A 0-1000 Ppm N/A N,4 C Core Cooling RPV Water Level A,B,C CD Wide Range RG 1.89 RG 1.100 Yes 3 channels -160* to +60* IE Recorners/Indte D g
Fuel Zone RG 1.89 RG 1.100 No 2 channels -326" to -116* Inst Bus Recorders /1%!ac D.E CQ 3
In-Core Thermocouples N/A N/A N/A N/A N/A N/A N/A OM
,U D pg Reactor Coolant Integrity i g RPV Pressure A,B,C RG 1.89 RG 1.100 Yes 2 channels 0-1500 psig IE Recorders E,J %%
DW Equip Dr Sump 1mvel 8,C N/A N/A N/A 1 channel High-high Inst Bus Alarm F,L gg f
! DW Flow Dr Sump Flow B,C N/A N/A N/A 1 channel D-5 gpm Inst Bus Recorder L (/1 F4 B C.D RG 1.100 Yes IE A.J Drywl1 Pressure RG 1.89 2 channels 0 to 30 psig Indicator
~
Pri Cont Radiation C,E *
- Supp Pool Water level A.C.D RG 1.89 RG 1.100 Ye s 1 channel per -16*9" to -4'9" 1E Recorder /indic A U
! each of 4 quadrants containment Integrity
) Isolation Valve Position B RG 1.89 RG 1.100 Yes 2 channels per Open/ closed IE Indicator lights A d each of 2 l,
values
! Cont Temperature A RG 1.89 RG 1.100 Yes 2 channels 6 to 300*F 1E Indicator A Cont Pressure A.B.C RG 1.89 RG 1.100 Yes 2 channels 0 to 30 psig IE Indicator A.N,J
' DW/ Cont Hydrogen A,C * * * * * *
- a See Cent Area Rad C,E N/A N/A N/A * * * *
, See Cont Effluent C.E Pri Cont Effluent C N/A N/A N/A * * * *
' Supp Pool Temp A,D RG 1.89 RG 1.100 Yes 2 channels per 0 to 300*F 1E Indicator / A NM I each of 4 Recorder O N
! quadrants 4 M
, * -a q DW Temper,ture AD RG 1.89 RG 1.100 Yes 2 channels 0 to 400*F 1E Indicator A.H g HO o -J
- Applicant to provida.
I l
I e
i
Table 1D-2
SUMMARY
OF INFORMATION INDICATING DEGREE OF COMPLIANCE OF 238 NUCLEAR ISLAND WITH REGULATORY GUIDE 1.97, REVISION 2 (Continued)
C2alification 1st C trol C ntsJ Variable Type Environment Selsete Assurance Red ndancy Range Pnwer Sur ply Disrtay No+cs Fuel Cladotnq Parrier Cooient Radiation N/A N/A N/A N/A N/A N/A N/A C Coolant Gesuna (samplel C N/A N/A N/A eda N/A N/A C System Olie re t ion MSIV LCS Flow D RG 1.89 RG 1.100 Yes Inboard / 0 to 15 scfm IE Recorder Outboard System per Each Steam Line
- Drain M Cont Spray Flow D N/A N/A N/A h/A N/A N/A See RHR W 8' 10w G RHR Flow D RG 1.t9 RG 1.100 Yes I channel per 0 to 10,000 gpm IE Indicator each of 1 O loops CM W RHR Service Mater Flow RG 1.89 RG 1.100 Yes I channel per 0 to 10,000 gpa 1E Indicater G O V) g D pg
' s MM A See RMR > lO g LPCI Flow D N/A N/A N/A B/A N/A N/A riow g g RCIC Flow D RG 1.8 9 RG 1.100 Yes I channel 0 to 800 gFm IE InJ1cator yy HPCS Flow D RG 1.89 RG 1.100 Yes I channel O to 10,000 gpm IE Indicator V)
LPCS Flow D RG 1.99 RG 1.100 Yes I channel 0 to 10,000 gpm II Indicator U SLC Pressure D RG 1.d9 RG 1.100 Yes I channel O to 1800 psig IE Indicator A.N SLC Tark Level D N/A N/A N/A 1 channet 0 to 5000 gal Inst Bus Indicatar L U SRV Position D RG 1.89 RG 1.100 Yes 3 channels Open/ Closed IE Indicator LagSt s A per SRV 0
Feedwater Flow D N/A N/A N/A 1 channel per 0 to 20x10 lb/hr Instrument Bus Recorder each of 2 loops cond Storage Tank Level * * * * * *
- D E5F Cocling Nater Flow D * * * * * *
- ESF Cooling Nater D Temperature
.tagh Rad Tank Level D N/A N/A N/A 1 channel per Bottom to top last Bus Recorders 1 each of 5 tanks Emerg vent Damper D RG 1.89 RG 1.100 Yes 2 channels per Open/ Closed IE Indicator Lights A Position each of 2 dampers pee each of 11 paths N O N
< l**
- Applicant to provide.
- Q O
HO OM O O O
1 1
l t
O O .
1
! o l Table ID-2
SUMMARY
OF INFORMATION INDICATING DEGREE OF COMPLIANCE OF 238 NUCLEAR ISLAND .
WITH REGULATORY GUIDE 1.97, REVISION 2 (Continued) i 4 ,
- IA****" Gaality control Roon comunents/
Variable Type Environment Seismic Assurance Redundancy B a n<re Power Surply Display Notes I a
system Operation (Contieuedt f
Standby Power Status D RG 1.89 RG 1.100 Yes 1 chanrel per voltage, it Indicators A f q bus amye res l Standby Air status D RG 1.89 RG 1.100 1 channel per o to 300 psig IE Indicator A each of 2 }
}' divisions j Standby LCS APS D N/A N/A N/A 1 channel per O to 150 psig IE Indicator L Pressure each of 2
} divisions y i
Standby If5 WP5 O N/A N/A h/A 1 channel per 0 to 150 psig 1E Indicator L La3 l Pressure each of two ID :
j divisions {
2O i Effluent McMiterint CM i
U SCTS Vent riow E RC 1.99 RG 1.100 Yes 2 channels Open/ Closed IE Indicator Laghts R OE I i per each of 2 UE A dampers per MM '
I each of 2 >M i
4 trains M '
N/A N/A Open/ Closed H
i Raowaste Vent Flow E N/A 2 channels Inst Bus Indacator Lights 1.E.L gg ,
j per damper g i j Part/uaio en sample E * * * * * *
- t* i Environs nonitoring E * * * * * * * !
Meteorology * * * * * * *
{
i E O l Post-Accident Sample E N/A N/A N/A 2 points
- Inst Bus N/A N per area
-3 sampled r i
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I ,
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- Arp1teant to provide. I l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 10 Table 1D-2 NOTES A. Qualification extends from sensor to display. Addition of instruments to Table 3.10-1 or Table 3.11-9 required.
B. Sensor drive mechanism not qualified or qualifiable to RG 1.89 or RG 1.100 standards. No modification planned. A wide-range instrument is under development as a replacement for these instruments.
C. Sampling is by the post-accident sample station. Recommended analytical procedures provided.
D. An enhanced water level indication is under review as a replacement for these instruments.
E. An additional channel of display is under review.
F. Modification of equipment drain sump instrumentation is under review.
G. This variable is an alternate to RHR heat exchanger outlet temperature. gg H. This instrumentation provides adequate information for the plant operator to carry out actions defined by the Emergency Procedure Guidelines.
I. Display is in the Radwaste Building Control Room.
J. Two channels of pressure instrumentation are considered ade-quate to provide unambiguous post-accident indication.
K. Indication of damper positica is sufficient indication of system flow.
L. Design to Category 3 requirements is acceptable for this instrument since it serves a backup function to other variables.
M. SLC pressure provides adequate indication of system operation in combination with other variables.
N. Analytical procedures and laboratory facilities to be provided by Applicant.
O 1D.4-8
-. . = . - - .- . - . _ . _ _ . . . - - - . - . --. - . . - _ _ . . - - . - - .
f i
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 i
t t-l Table 1D-3 4 238 NUCLEAR ISLAND TYPE A VART..LLES J
Neutron Flux I RPV Water Level j RPV Pressure 1
] Suppression Pool Temperature i
Suppression' Pool Water Level Drywell/ Containment Hydrogen Concentration 4
Drywell Air Temperature i
l Containment Pressure i
j Containment Air Temperature
(:) '
i l
t l
l L
I l l i
I 1D.4-9/lD.4-10 ,
- l 1
s.-- , , - . . . . - , - . . , - . - . . - . . - , _ . _ , . . . . _ , - . . . - . - , , . - , . _ . _ _ - , - _ ~ . . - . ~ _ _ ., , . . . . . - . . . _ .-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 O
ATTACllMENTS TO APPENDIX 1D O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 O
ATTACllMENT A TO APPENDIX ID l I IMPLEMENTATION POSITION FOR REGULATORY GUIDE 1.97, REVISION 2 O
l 4
O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 v
ATTACHMENT A TO APPENDIX 1D IMPLEMENTATION POSITION FOR REGULATORY GUIDE 1.97, REVISION 2 This attachment addresses the criteria for assessment of variables against the requirements
- of Regulatory Guide 1.97, Revision 2. The organization of this implementation position follows Regulatory Guide 1.97, Revision 2, Section C, ' Regulatory Position," point by point. Some rewording has been made for clarity.
Where other documents are referenced by paragraph in the regula-tory guide, the referenced material is included.
Additional clarifying information or modifications to Regulatory
(")
L./
Guide 1.97 to reflect an exception to the regulatory guide text are included in brackets [ J.
This attachment constitutes an engineering and licensing position for assessment of 238 Nuclear Island variables.
Table lDA-1 lists the variables addressed and requirements unique to tha. variable. Table 1DA-2 summarizes the generic require-mento applicable to each category of variable. It should be noted that some of the requirements are Applicant's scope, as indicated in Table 1DA-2 and are to be addressed separately.
- Although the term " requirement" is used throughout this docu-('T ment for clarity, it is recognized that Regulatory Guide 1.97 provides guidance and not requirements.
(_/
1DA.1-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Position C.1 - Accident Monitoring Instrumentation g
[The following criteria, in addition to Positions C.l.1 through C.2.5, should be implemented):
C.I.a - Regulatory Guide Definition - Type B Variables Type B Variable: Those variables that provide information to indicate whether plant safety functions are being accomplished.
Plant safety functions are (1) reactivity control, (2) core cool-ing, (3) maintaining reactor coolant system integrity, and (4) maintaining containment integrity (including radioactive effluent control). Variables are listed with designated ranges and category for design and qualification requirements. Key variables are indicated by design and qualification Category 1.
Interpretation / Implementation Position For the BWR, the Type B variables identified in Table 1DA-1 O
satisfy the provisions of Regulatory Guide 1.97 and should be implemented as identified in Table 1DA-1 of this implementation position with the exceptions noted. Exceptions are bracketed, [ ] .
C.l.b - Performance Requirements The determination of the performance requirements for Type A, B, and C variable measuring instruments for accident monitoring instrumentation channels should include, as a minimum, identifi-cation of:
- 1. Range of the process variables to be measured and monitoring instrumentation [per Table 1DA-1]
- 2. Required accuracy of measurement [per Table 1DA-1]
O 1DA.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.1.b - Performance Requirements (Continued)
(
- 3. Required response characteristics, if applicable [per Table 1DA-1]
- 4. Time interval during which the measurement is needed (function time)'
- 5. Local environment (s) in which the instrument channel components must operate
- 6. Requirements for rate or trend information [per Table 1DA-1]-
- 7. Any required spatial distribution of sensors
- 8. Any requirements for group displays of related O information v
C.l.c - Qualification Duration Requirements The qualification duration of Types A, B, and C, Category 1 and Category 2 information display channels should be:
(1) For Type A variable monitoring instruments - the dura-tion for which the instrumentation is required for manual operator actions to bring the plant to " cold shutdown" following a design basis accident event.
(2) For Types B, C, D, and E variable monitoring instru-mentation - at least 100 days following a DBA event unless shorter duration times can be justified and l documented.
r u
4 l 1DA.1-3 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.l.c - Qualification Duration Requirements (Continued)
Shorter qualification times are acceptable if equipment replacement or repair can be accomplished within an acceptable out-of-service time, taking into consideration the environment where the equipment is located and the information needs of the operator.
Position C l.1 - Regulatory Guide Definition Type A Variables Those variables that are to be monitored provide the primary information required to permit the control room operator to take the specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events. Primary information is information that is eFaential for the direct accomplishment of the specified safety functions; it lll does not include those variables that are associated with con-tingency actions that may be identified in written procedures.
Interpretation / Implementation Position Type A variables are limited to those variables which are neces-sary (primary) to alert the control room operator of the need to perform preplanned manual actions for safety systems to perform their safety functions, such as (1) initiating safety systems (e.g., hydrogen mixing, main steam leakage control system) and/or (2) changing safety system lineups, as necessary, to permit the systems to perform safety function (e.g., suppression pool cool-ing, containment spray, etc.), for which no automatic system controls are provided and are required to mitigate the conse-quenc. of specific design basis accidents (DBAs) which threaten the health and safety of the public, as defined in this safety llh 1DA.1-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10
() Interpretation / Implementation Position (Continued)
Analysis Report (SAR) , to bring the plant to a safe condition (e.g., cold shutdown).
Type A variables do not include variables (1) which may indicate whether a specific safety function is being accomplished (Type B) or (2) which may indicate the need for contingency or corrective actions, resulting from the failure of the plant (Type C) or sys-tem (s) (Type D) to respond correctly when needed, or (3) which may indicate to the operator that it is desirable to change /
modify the operation / alignment of systems important to safety to maintain the plant in a safe condition after plant safety has been achieved. [ Emergency actions specified by Emergency Procc-dure Guidelines (EPGs) in response to specific operating limits should, however, be considered.)
() Position C.l.2 - Regulatory Guide Definition Type C Variables Those variables that provide information to indicate the poten-tial for being breached or the actual breach of the barriers to fission product releases. The barriers are (1) fuel cladding, (2) primary coolant pressure boundary, and (3) containment.
Interpretation / Implementation Position Type C variables identified in Table 1DA-1 are considered ade-quate to satisfy the provisions of Regulatory Guide 1.97 and should be implemented as identified in the Regulatory Guide 1.97 Table 1DA-1 with the exceptions noted in Table 1DA-1 of this implementation position. Exceptions are bracketed, [ ].
f I
1DA.1-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Position C.1.3 - Design and Qualification Criteria C.1.3.1 - Design and Qualification Criteria - Category 1 C.l.3.1.a Instrumentation should be qualified in accordance with Regulatory Guide 1.89 and the methodology described in NUREG-0588. Qualifi-cation applies to the complete instrument channel from sensor to display, where the display is a direct-indicating meter or record-ing device. Qualification applies to the instrument channel from the sensor to and including the channel isolation device, where the instrumentation channel signal is to be used in a computer-based display, recording, and/or diagnostics program, unless it can be shown that failure of the isolation device cannot jeopar-dize the function of the Class 1E instrument channel.
The signal isolation device should be located where it is acces- ggg sible for maintenance during accident conditions.
The portion of the instrumentation channel requiring seismic qualification should be qualified in accordance with Regulatory Guide 1.100 (including hydrodynamic loads). Instrumentation should continue to read within the required accuracy following, but not necessarily during, a safe shutdown earthquake (SSE)
(including hydrodynamic loads).
Instrumentation whose ranges are required to extend beyond those ranges calculated in the most severe design basis accidcnt event for a given variable should be qualified as follows:
The qualification environment for Type C information dis-play channel components shall be based on the plant unique design basis accident events, except those components directly subjected to the monitored variable environment llg 1DA.1-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 e C.1.3.1.a (Continued) k_x) shall be qualified to the assumed maximum range for the monitored variable. The monitored variable shall be assumed to approach this peak by extrapolating the most severe ini-tial ramp associated with the DBA events. The decay for this variable shall be considered proportional to the decay for this variable associated with the DBA events (see Fig-ure 1DA-1). No additional qualification margin needs to be added to the extended range variable. All environmental envelopes, except those portaining to the variable measured by the information display channel, shall be associated with the DBA events.*
C.1.3.1.b Sufficient instrumentaticn channels should be provided to assure that no single failure (1) within the accident-monitoring chan-nel, or (2) its auxiliary supporting features, or (3) its power sources, concurrent with the failures that are a condition or results of a specific accident (consequential damage) will pre-vent the operators from being presented the information necessary for them to (1) determine the safety status of the plant and (2) bring the plant to and maintain it in a safe condition follow-ing an accident.
- The above environmental qualification requirement for Type C equipment does not account for steady state elevated levels that may occur in other environmental parameters associated with the extended range variable. For example, a Type C sensor measur-ing containment pressure must be qualified for the measured process variable range (i.e., three times design pressure for concrete containments) , but the corresponding ambient tempera-ture is not mechanically linked to that pressure. Rather, the ambient temperature value is the bounding value for desigr, basis accident events analyzed in Chapter 15. The extended range requirement is to ensure that the equipment will continue to provide information if conditions degrade beyond those postu-(~N lated in the safety analysis. Since Type C variable ranges are
(_) non-mechanistically determined, extension of associated param-eter levels is not justifiable and has therefore not be required.
1DA.1-7
GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 10 C.l.3.1.b (Continued) ggg Additional instrumentation channels should be provided to allow operators to determine the actual condition of the plant when failure of one accident monitoring channel would result in infor-mation ambiguity (viz., redundant displays disagree) that could 1 cad operators to defeat or fail to accomplish the required safety function. Additional instrumentation channels should consist of either (1) providing another independent channel monitoring the same variable (redundancy), or (2) providing an independent channel to monitor a different variable that bears a known rela-tionship to the multiple channels (diversity).
Redundant or diverse channels should be electrically and physic-ally separated from each other and from equipment not classified important to safety in accordance with Regulatory Guide 1.75.
At least one instrument channel shou]d be displayed on a direct-indicating or recording /.evice (e.g., strip chart recorder).
lll Redundant [or diverse] monitoring channels are not needed within each redundant division of a safety system.
C.l.3.1.c Instrumentation channels should be energized from station standby power sources as provided in Regulatory Guide 1.32 and should be backed up by batteries where momentary interruption is not tol-erable (uninterruptible power) .
C.l.3.1.d Instrumentation channels should be available prior to an accident except as provided in Paragraph 4.11, " Exemption," as defined in IEEE Standard 279 or specified in Technical Specifications.
lll 1DA.1-8
i GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 10 C.l.3.1.e O
j The recommendations of the following Regulatory Guides pertaining to quality assurance should-be followed:
i-l l Regulatory Guide 1.123 Regulatory Guide 1.38 4
Regulatory Guide 1.144 Regulatory Guide 1.58 Regulatory Guide 1.146 Regulatory Guide 1.64 Regulatory Guide 1.28 Regulatory Guide 1.74 Regulatory Guide 1.30 Regulatory Guide 1.88 ,
[ Quality assurance per NEDO-11209, NEBG BWR QA Program descrip-tion is acceptable.]
1 C.l.3.1.f Continuous instrumentation channel readout should be provided (this may be by [either an analog or digital meter] or recorder).
[ Interruption of instrumentation readout is acceptable where interruptible power is acceptable. (See Position C.l.3.1.c.)]
l!
I Overlapping instrument spans should be provided when two or more instruments are required to cover a particular range.
C.1.3.1.g Recording of instrumentation readout information should be pro-l
! vided as follows:
l (1) the information should be continuously available on l
l dedicated recorders [e.g., strip chart recorders),
where direct and immediate trend or transient infor-t mation is essential for operator information or l
i 1DA.1-9 l
l l
. - _ _ _ , _ - - _ . - . . ~ - - _ _ , _ - - - _ _ - _ - - , - . _ _ _ . - - _ , . . . _ _ . _ , - - _
GESSAR II 22A7007 238 NUCLEAR ISLAND Ray. 10 C.l.3.1.g (Continued)
(2) the information should be continuously updated, stored in a computer memory and displayed on demand where the information is not essential for operator information or action; or (3) intermittent displays such as data loggers and scan-ning recorders should be used if no significant transient response information is likely to be lost by such devices.
C.1.3.2 - Design and Qualification Criteria - Category 2 C.l.3.2.a Instrumentation should be qualified in accordance with Regulatory Guide 1.89 and methodology described in NUREG-0588. (Qualifica-tion applies to the complete instrument channel from sensor to display where the display is a direct indicating meter or record-ing devices.] Qualification applies from the sensor [to and including] the isolator / input buffer where the channel signal is to be processed or displayed on demand, unless it can be shown that failure of the isolator / input buffer cannot jeopardize the function of Class 1E instrument channel.
The signal isolation device should be located where it is access-ible for maintenance during accident condition.
The portion of the instrumentation channel requiring seismic qualification should be qualified in accordance with Regulatory Guide 1.100 (including hydrodynamic loads) when the instrumenta-tion is part of a safety-related system.
O 1DA.1-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 10 e~s C.1.3.2.b Instrumentation channels should be energized from a high-reliability power source, not necessarily standby power, and i
should be backed up by batteries where momentary interruptions are not tolerable (noninterruptible power).
C.1.3.2.c i
- Instrumentation channels should be available with the out-of-service intervals based on
(1) the normal Technical Specification requirements on out-of-service for the system it serves where applicable, or (2) where specified by other requirements.
C.1.3.2.d The recommendations of the Regulatory Guide pertaining to quality assurance, identified in Position 1.3.1.e, should be followed, I where applicable, considering the importance to safety of the instrumentations under consideration. Since some instrumenta-tion is less important to safety than other instrumentation, the quality assurance requirements that are implemented should:
(1) provide control over activities affecting quality to the extent consistent with the importance to safety of the instrumentation, and I
(2) be determined and documented by personnel knowledge-abic in the use of the instrumentation.
O l 1DA.1-ll
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 10 C.1.3.2.e The instrumentation readout should be on an individual meter /
recorder (e.g., strip chart recorder) or processed for display on demand by a CRT or other appropriate means.
C.1.3.2.f Effluent radioactivity monitoring, area radiation monitoring, and meteorology monitoring instrumentation readouts should be:
(1) continuously available on dedicated recorders (e.g.,
strip chart recorders) where direct and immediate trend or transient information is essential for opera-tor information or action, or (2) continuously updated, stored in a computer memory, and displayed on demand, if not essential to the operator, and (3) displayed by dial, digital, CRT or strip chart recorders.
C.1.3.3 - Design and Qualification Criteria - Category 3 C.l.3.3.a Instrumentation should be of high-quality commercial grade and should be selected to withstand the specified service environment.
C.1.3.3.b The method of display may be dial, digital, CRT or strip chart recorder indication.
1DA.1-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 10
(} C.1. 3.3.b (Continued)
Effluent radioactivity monitoring, area radiation monitoring, and meteorology monitoring instrumentation readouts should be:
(1) continuously availabic on dedicated recorders (e.g.,
strip chart recorders) where direct and immediate trend or transient information is essential for opera-tor information; or (2) continuously updated, stored in computer memory, and displayed or demand, if not essential to the operator; and (3) displayed by dial, digital, CRT, or strip chart recorders.
(} Position C.1.4 - Additional Criteria - Categories 1 and 2 C.1.4.a Category 1 and Category 2 equipment should be designated as part of the accident-monitoring instrumentation [ Type A, B, C] or systems operation and effluent-monitoring instrumentation 1
l [ Type D, E].
Transmission of signals from such equipment for other uses should be through isolation devices that are designated as part of the monitoring instrumentation and that meets the provisions of Regulatory Guide 1.97.
C.1.4.b Category 1 and 2 instruments designated as Types A, B, and C
() should be [ located visually accessible to the operator] and 1DA.1-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.1.4.b (Continued)
O specifically identified on control panels so the operator can casily discern that they are intended for use under accident conditions.
Position C.1.5 - Additional Criteria - Categories 1, 2, and 3 1.5.a Servicing, testing, and calibration programs [ consistent with the requirements of Position 1.3] should be specified to maintain the capability of the monitoring instrumentation. The capability for testing during power operation should be provided for those instruments where the required interval between testing is less than the normal time interval between generating station shutdowns.
[ Applicant to provide programs.]
1.5.b The instrumentation design should facilitate administrative control over removing channels from service.
[ Applicant to provide administrative controls.]
1.5.c The instrumentation design should facilitate administrative con-trol of access to setpoint adjustments, module calibration adjustments, and test points.
[ Applicant to provide administrative controls.]
O 1DA.1-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1.5.d
}
Iluman factors analysis should be used in determining the type and location of displays and the monitoring instrumentation.
design should minimize the development of conditions that would cause motors, annunciators, recorders, alarms,'etc. to give ambiguous indication potentially confusing to the operator.
[See Chapter 18 regarding conduct of human factors review.]
C.l.5.e Instrumentation should be designed to facilitate the recogni-tion, location, repair, or adjustment of malfunctioning com-ponents or modules.
C.1.5.f O Monitoring instrumentation should be from sensors that directly measure the desired variables whenever possible. Indirect measurements should be made only when they can be shown by analy-sis to provide unambiguous information.
C.1.5.g The same instruments should be used for accident monitoring as those used for normal plant operations to enable the operators to monitor instruments with which they are most familiar during accident situations. Separate instruments should be used only when the required " accident" range would result in a loss of instrumentation sensitivity in the normal operating range.
I 1DA.1-15
GESSAR II 22A7007 238 NUCLEAR ISLAND R0v. 10 C_.l.5.h g; Instrumentation channel periodic checks, testing and calibraticn verification should be in accordance with the applicable portions of Regulatory Guide 1.118 (Note: Instrumentation response time testing for post-accident instrumentation is (not required as it is peak values of the variables that are of interest rather than the rate of peak value achievement].
(Applicant to provide programs.]
Position C.l.6 - Additional Criteria - Type B and C Variables C.l.6.1 - Type B Variables In conjunction with Table 1DA-1 the following should be consid-ered as a minimum number of instruments, and their respective ranges, for accident monitoring instrumentation. ggg C.l.6.1.a - Reactivity Control Monitoring The measured variable should be neutron flux or combinations of other variables, if properly justified, to indicate accomplish-ment of control of reactivity in the core. If neutron flux is used, the measurements should extend from (10-6] to [10+2]% of full reactor power. Current value, rate and trend information should be available in the control room [for the primary variable).
C.I.6.1.b - Core Cooling Monitoring Reactor vessel water level monitoring should ue provided to indi-cate the accomplishment of core cooling. Current value and trend information should be available in the control room.
O 1DA.1-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.l.6.1.c - Reactor Coolant System Integrity
{
The measured variables should include reactor pressure, drywell pressure and drywell sump level to indicate the accomplishment and maintenance of Reactor Coolant Pressure Boundary (RCPB) integrity. Current value ir. formation should be available in the control room.
C.l.6.1.d - Reactor Containment Integrity i
The measured variables should include reactor containment pres-sure and remote operated containment isolation positions-(closed /not closed) to indicate the accomplishment and mainte-nance of reactor containment integrity. Current value of pres-sure and valve position status should e available in the control room.
C.1.6.1.e - Radioactive Effluent Monitoring
/}
The measured variabic should be noble-gas monitoring of the planned release points to indicate accomplishment of radio-active effluent control. The information should be available to the control room operator.
C.l.6.2 - Additional Criteria, Type C Variables l
In conjunction with Table 1DA-1, the following should be con-sidered as a minimum number of instruments, and their respective l ranges, for accident monitoring instrumentation. In addition,
! the instrumentation (primary variable] should detect and alarm
! the following with the current value information in the control room.
() ;
i 1DA.1-17 i
c GESSAR'II 22A7007 238 NUCLEAR ISLAND Rav. 10 C.l.6.2.a - Fuel Clad Barrier Monitoring ,
These variables indicate a breach iq the fuel cladding barriers -
(i.e., an in-core fuel cladding breach capable of releasing more than 1 percent of fuel cladding gap and plenum activity of the core). The measured variable should be reactor coolant /
system radioactivity (gross gamma).
r i '
Operator sampling of reactor coolant should be used as a means Ed,'
verify the measured variable. Sampling provisions should per- ,N s c.
mit one sample to be taken every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with analyses results ' ,. ,,
available within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the sampling. ,
n ,
C.1.6.2.b - Reacter Coolant Pressure Boundary Monitoring These variables indicate a breach in the reactor coolant system that produces a loss of reactor coolant inventory in excess of normal coolant makeup capability. The measured variables used to detect a RCPB bre2ch shall span the full-spectrum of design basis accident event ureak sizes. The measured variable should include drywell pressure and drywell sump Jovel with an accuracy of 20% of span and a response time of less'than 10' seconds fdr input step change of 10%. Current value shoul 3 be provided /In' 9 the control room. '.
C.l.6.2.c - Primary Reactor Containment Pressure <Boondary Monitorinq ,
t
, 1 These variables indicate a breach in the Primary Containment -.
\,
Pressure Boundary that is capable of producing radiation ,' '
releases in excess of Title 10, Code of Federal Regulations.
O 1DA.1-18
V,9
,-g T'e 238 MUCLEAR GESSAR II ISLAND 22A7U~07-r- ,, R v. 1 -
t _
r'; C.1.6.2.cb- Primary Rgac$nr, Containment Pressure Bound 9ry ,
(_) i Monitoring (Continued)
~
' ~~~
j-
^
, , [ ,,s +.
x Part 100 "P.nactor Site critoria" at the ed!vsion.
s . ; ,
area boundary ys using TID-14844, " Calculation of Distence Pdctors for Power and
- ', A -
?
T Test Reactor Sites, March 23,-1982,", sourd' Bar.us. .
w: ~ . , , ,
1;s ,-
s ,
u N g
- s The measured variable should be primary durAainment,s. pressure and
, ~
secondary containme.t air space radiatiop'emonitoring.for gross * %
gamma. The containment pressurl;e-mopitoring,,instrunient ,w.nge ~
, e ,
should be the range specifled it? Table llDAil with as displby chan-nel accuracy within i20% of span and a redppqse t elme of lays sthan 10 cec for an input step change .of 10 perce'nt ok[ span. The Sec-ondary containment air space' radiation monitoring instru$gptation range should be as specified, on; Table 1DA-1 with a.' display 7han-m .. -
nel accuracy as specified with a display channel redponsh time of less than 10 seconds for an input step change of 10% of span.
6
,s Current value should be provided and alarmed on the control room.
(-) SecNndary containment air space radiation detectors should respond to gamma radiation photos within any edtrgy range from 60 kev to 3 MeV with an energy response accuracy e,r 1200 at any _
~
spec 3fic photon energy from 0.1 MeV to 3 MeV. Overall Eystem accuracy should be within a factor og 2 over tho' entire range.
1 1 + .<
C.1.6.2.d - Potential Breach of the Final Fission' Product Barrier i ,.
I .<, i x The ncasured variable should be (primary) containment pressure, primary reactor containment hydrogen co,ncentration and RCPB pJessure. The primary containment prgssure monitoring instru-mhntation, range shoud be as cpecified on Table lDA-1 with a dis-play channei a ,' 7.my (of] 110% of span and a response time of less than'l second for an input step change of 110% of span.
Current value should be provided in the' control room. The pri-mary containment hydrogen meditoring instrumentation range
.O
(,) should be as specified in Taule 1DA-1 with a display channel i 1DA.1-19
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.l.6.2.d - Potential Breach of the Final Fission Product Barrier (Continued) accuracy of 110% of span. Initial and subsequent samples should be available at sufficiently short intervals to allow the opera-tor to monitor the value of hydrogen concentration in the contain-ment and take appropriate timely action as required. Current value should be provided in the control room. The RCS pressure monitoring instrumentation range should be as specified on Table 1DA-1 with a display channel accuracy within 10% of span and a response time of less than 1 second for an input step change of 10% of span. Current value should be provided in the control room.
Position C.2 - Systems Operation Monitoring and Effluent-Release Monitoring Instrumentation C.2.1 - Definitions O
- a. Type D, those variables that provide information to indicate the operation of individual safety systems or other systems important to safety,
- b. Type E, those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and in continually assess-ing such releases.
Interpretation / Implementation Position for Types D & E Type D and E variables identified in Table 1DA-1 are considered adequate to comply with this position and should be implemented as identified in Table 1DA-1 with the exceptions noted in Table 1DA-1 of this implementation position. Exceptions are bracketed, [ ].
O 1DA.1-20
GESSAR II 22A7007
~238 NUCLEAR ISLAND Rev. 10 i C.2.2 The plant designer should select the appropriate variables and inforniation display channels [from.the list of identified vari-ables on Table 1DA-1 and Type D variables] required by his design to enable the control room operating personnel to:
- a. Ascertain the operating status of each individual safety system and other systems important to safety to that extent necessary to determine if each system is operating or can be placed in operation to help mitigate the consequences of an accident [(Type D)].
- b. Monitor the effluent discharge paths and environs with the site boundary (of his plant] to ascertain if there have been significant releases (planned or unplanned) of radioactive materials and to continually assess such releases [(Type E)].
- c. Obtain required information through a backup or diagnosis channel where a single channel may be likely to give ambiguous indication.
C.2.3 The process for selecting system operation and effluent release variables should include the identification of:
- a. For Type D -
(1) the plant safety systems and other systems important to safety that could be operating or that could be placed in operation to help miti-gate the consequences of an accident; and O 1DA.1-21
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 C.2.3 (Continued) (2) the variable or minimum number of variables that indicate the operating status of each system identified in (1) above.
- b. For Type E -
(1) the planned paths for effluent release; (2) plant areas inside buildings where access is required to service equipment necessary to miti-gate the consequences of an accident; (3) on-site location where unplanned releases of radioactive materials should be detected; and (4) the variables that should be monitored in each located identified in (1), (2), and (3) above. C.2.4 - Performance Requirements The performance requirements for system operation monitoring (Type D) and effluent release monitoring (Type E) information display channels should include, as a minimum, identification of: (1) range of the process variable and monitoring instrumentation [per Table 1DA-1]; (2) required accuracy of measurement [per Table 1DA-ll; (3) required response characteristics, if applicable [per Table 1DA-1]; O 1DA.1-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 () C.2.4 - Performance Requirements (Continued) , (4) time interval in which the measurement is needed (function time); 2 (5) the local environment (s) in which the instrument channel components must operate, (6) requirements for rate on trend information [per Table 1DA-1]; (7) any requirements to group displays of related information; and (8) any required spatial distribution of sensors. C.1.5 - Design and Qualification Requirements O The design and qualification criteria for systems operation (Type D) and effluent release monitor * 'vpe E) instrumenta-tion should be taken from the iapprc, licable] criteria provided in Positions 3.3 and 1.4 for ' 1, 2, and 3 instruments. Table 1 of this implementation position is considered adequate to comply with the instrumentation and instrument range criteria of Regulatory Guide 1.97 with the exceptions identified in Table 1DA-1 of this. implementation position. Exceptions are bracketed, [ ]. i i C) 1DA.1-23/lDA.1-24
O O O Table 1DA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND System Variable No. Range Type Cat.egory Remarks Reactivity Control
-6 Neutron Flux C51 10 % to 100% A,B 1 Function detection; (value, rate, trend) full power accomplishment.of shutdown (Primary Coolant) ,
, Control Rod Position Cl2 Full in or not B 3 Verification N full'in (Back-up Variable] $ z s Boron Concentration D24 0 to 1000 ppm B 3 Verification gQ p (sample) !Back-up Variable] & u)
. m un g >>
a Core Cooling NN Coolant Level in the B21 Bottom of core A,B 1 Function detection; Reactor (value, trend) support plate accomplishment of & of [above nor- mitigation [ Primary 6 mal reactor Variable] water level] B21 [Below core B,C 1 Potential for breach' plate to above [ Primary Variable] reactor [ Alternate to In-core level 1] TC) ^ E~< ' 5 o O i w m,
Table IDA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued) System variable __No. Range Type Category Remarks Maintaitaing Reactor Cool-ant System Integrity RCS Pressure B21 15 psia to A,B,C 1 Function detection; (value + alarm) 1500 psig accomplishment of 10%; 1 sec/ mitigation. Poten-10% response tial for breach. g [ Primary Variable] g g Drywell Sump Level E31 Bottom and B,C [3] ya o (value + alarm) top of sump oM . 20%; 10 sec/ E$ 7 w 10% response $$ m Drywell Pressure T41 1,2 Function detection; 0 to [30 psig] B,C,D gH (value + alarm) 20%; 10 sec/ accomplishment of e 10% response mitigation. @ o Primary Containment -- 0 to 10 R/hr E Area Radiation 1 to 10 C 3 Release assessment R/hr Detection of breach Suppression Pool Water P50 [ Top drywell. A,C,D 1,2 Detection of breach Level vent to weir wall] Maintaining Containment Integritj{ Primary Conta3.nment All Closed B 1 Accomplishment of xw Isolation Value Position Not Closed isolation CN (Excluding Check Valves) [ Primary Variable] Ib (value) g$ ow Primary Containment T41 40 F to A 1 Temperature 250 F e O O
_ _ _ . -- - - . _ . , _ __ _. -. . ._ . ~ . O O O Table IDA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued) System Variable No. Range Type Category Remarks i^ Maintaining Containment i Inegrity (Continued) Primary Containment T41 [0 to 30 psig] A,B,C 1 Function detection; Pressure accomplishment of (value, rate, trend, + mitigation; poten-alarm) tial or actual breach.[ Primary u i10%; 1 sec/ " Variable] 10% Z y response $$
. . Na H Drywell/ Containment -
O to 30% i10% A,C 1 Potential for breach >> b Hydrogen Concentration [ Applicant to i
'd (value) provide'. ] ((
4 Secondary Containment - 10~ to 10 C,E 2 Indication of breach Area Radiation R/hr Release assessment U i (value) 60 kev to [ Applicant to 3 3 MeV; 20% provide]. 0.1 MeV to 2 Mev-2 x total range; 10 sec/10%
- response -
-6 Secondary Containment -
10 c/cc C,E 2 Potential for breach g Noble Gas Effluent to 10-3 pc/ cc [ Applicant to 0y provide]. g wo O -4
Table 1DA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMEN1' OF 238 NUCLEAR ISLAND (Continued) System Variable No. Range Type Category Remarks Maintaining Containment Integrity (Continued) Primary Containment - 10jpu/ccto C 3 Potential for breach Noble Gas Effluent 10 pu/cc [ Applicant to provide.] Suppression Pool G51 30 to 230 F A,D 1,2 g Temperature w y o Drywell Air Temperature T41 40 to 440 F A,D 1,2 ya o tu .# Fuel Cladding Barrier E$ 7 w Monitoring gy Coolant Radiation - 1/2 to 100 [NA] [NA) [Not needed] $[ (value + alarm) times tech- { nical spec 1- z fication U limit (TSL) 50%; 5 min /10% response Coolant Gamma - 10 pCi/gm to C 3 Detection of breach (1 sample /6 hours) 10 Ci/gm results w/in 72 hr System Operation gg ow Main Steam Line Isolation E33 [0 to 110% D 2 [ Alternate to pressure] .# 3 Valve Leakage Control maximum flow] g O System [ Flow] o4 O O O
O O O Table 1DA-1 VARIABLES ASSESSED FOR REGULATORY GUTOE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued) System Variable No. Range Type Category Remarks System Operation (Continued) Containment Spray Flow E12 0 to 110% D 2 [See RHR flow.] design flow Residual Heat Removal E12 0 to 110% D 2 " (RHR) System Flow design flow co RHR [ service water flow] E12 [0.to 110% D 2 [ Alternate to heat 2 o design flow] exchanger temperature.] $0 e us - to u2 H Low Pressure Coolant E12 0 to 110% 'D 2 -[See RHR flow.] >> b Injection System Flow design flow HH W u) H Reactor Core Isolation E51 0 to 110% D 2 & Cooling System Flow design flow $ o-High Pressure Coolant E22 0 to 110% .D 2 [ Spray] System Flow design flow [ Low Pressure] Core Spray E21 0 to.110% D 2 System Flow design. flow Standby Liquid Control C41 0 to 110% D 2 [ Alternate to flow.) Sytem (SLCS) [ Pressure] design flow SLCS Storage Tank Level C41 Bottom to D [3] ,y top aw
.I -a SRV Position B21 0 - 50 psig D 2 g o. ~t ow i
1
Table 1DA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued) System Variable No. Range Type Category Remarks System Operation (Continued) Feedwater Flow C34 0 - 110% D 3 design flow CST Level - Bottom + top D 3 Indication of avail- y able water [Appli- w cant to provide.] 2 o ESF Cooling Water Flow - 0 to 110% D 2 [ Applicant to bO , design flow provide.] E$ Y w ESF Cooling Water - 32 to 200 F D 2 [ Applicant to Temperature provide.] $U High Radioactivity - Bottom to top D 3 $ Tank Level Emergency Vent Damper X73 Open/ Closed D 2 Position T41 X63 Standby Energy Status P53 Power, Air D 2 P60 P61 Effluent Monitoring g ow SGTS Ventilation Flow - [Open/Not Open] E 2 Release assessment < q
.# O HO OM O O O
O O O Table 1DA-1 VARIABLES ASSESSED FOR REGULATORY GUIDE 1.97 ASSESSMENT OF 238 NUCLEAR ISLAND (Continued) System Variable No. Range Type Category Remarks Effluent Monitoring (Continued) Other Ventilation Flow - [Open/Not Open) E [3] Release assessment
-3 Particulate / Halogen -
10 c/cc to E 3 Release assessment Release (sample) 102 pc/cc [ Applicant to provide.1 - g Environs Radioactivity - Various E 3 Release assessment - g o Monitoring [ Applicant to ya provide.] Og'
. to m yy 7 Meteorology -
Various E 3 Release assessment w [ Applicant to g provide.] mH t< Coolant, Post-Accident Sampling - E 3 Release assessment 2 U (sample) containment [ Sump sample not air needed.] l l l M
? >,
O HO O4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Table 1DA-2 REGULATORY GUIDE 1.97
SUMMARY
OF REQUIREMENTS CATEGORY l Requirement 1.3.1 a. Whole channel (including indicator) Qualification to Regulatory Guide 1.89, Regulatory Guide 1.100 (equivalent to IEEE 323-1974; IEEE 344-1975)
- b. Unambiguous Indication with Single Failure
- c. Station Standby Power
- d. Available Prior to Accident
- e. Quality Assurance Applied
- f. Continuous Indicator
- g. Variable Recorded (dedicated records if trend essential, otherwise computer memory is permitted) 1.4 a. Isolation Devices for Non-Accident O
Monitoring
- b. Identified as to Post-Accident Use (Type A,B,C, only)
Servicing, Testing and Calibration
- 1.5 a.
Program
- b. Removal from Service Administrative Control
- c. Setpoint, Calibration, Testing Administrative Control
- d. Human Factors Applied to Type and Location of Display
- e. Malfunctioning Components Maintainable
- f. Direct Variable Measurement
- g. (Normal Operation Instrumentation Used ,
- h. Testing and Calibration per Regulatory Guide 1.118
- Applicant to provide. llh l
l 1 1DA.1-32
GESSAR II 22A7007 - I"') ;
" " ~
238 NUCLEAR ISLAND Rev. 10
\_/ -
Table IDA-2 REGULATORY GUIDE 1.97
SUMMARY
OF REQUIREMENTS (Continued) CATEGORY 2 Requirement 1.3.2 a. Whole Channel (including display) Qualified to Regulatory Guide 1.89 and Regulatory Guide 1.100
- b. High Reliability Power Source (not necessarily station standby)
Available Per Technical Specifications.
- c.
- d. Quality Assurance Applied
- c. Individual Display or CRT
- f. Variable Recorded (NG Effluent and Area Rad only)
() (dedicated recorder if trend essential, otherwise computer memory) 1.4 a. Isolation Devices for Non-Accident Monitoring
- b. Identified as to Post-Accident Use (Types A,B,C, only)
Servicing, Testing and Calibration
- 1.5 a.
Program
- b. Removal from Service Administrative
- Control Setpoint, Calibration and Testing
- c.
Administration Control
- d. Confusing Indication Minimized; Human Factors Applied
- e. Malfunctioning Components Maintainable
- f. Direct Variable Measurement
- g. Normal Operation Instrumentation Used
- h. Testing and Calibration per Regulatory Guide 1.118 O
- Applicant to provide.
1DA.1-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Table 1DA-2 REGULATORY GUIDE 1.97
SUMMARY
OF REQUIREMENT? (Continued) CATEGORY 3 Requirement 1.3.2 a. High Quality Commercial Grade
- b. Individual Display or CRT Variable Recorded (NG Effluent, area rad only) 1.5 a. Servicing, Testing und Calibration
- Program
- b. Removal from Service Administrative
- Control
- c. Setpoint, Calibration, and Testing
- Administrative Control
- d. Confusing Indication Minimized; Human Factors Applied c.
f. Malfunctioning Components Maintainable Direct Variable Measurement lll
- g. Normal Operation Instrumentation Used
- h. Testing and Calibration per Regulatory
- Guide 1.118
- Applicant to provide, g 1DA.1-34
GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 10 O MAGNITUDE OF EXTENDED RANGE MEASURED PROCESS VARI ABLE ASSUMED MEASURED PROCESS VARIABLE MAXIMUM ENVELOPE V ARI AB LE - RANGE
/ I\
/ I i O / i \ DESIGN BASIS ACCIDENT
[ l \ EVENT MEASURED PROCESS VARI ABLE ENVELOPE
\
/ l
/ \
/ \
/ \
INITIAL M AXIrAUM \
% R ATE OF k VARIA8LE g CHANGE
\
\
TIME Figure 1DA-1. Typical Environment Qualification Envelope for Type C Instruments I l 1DA.1-35/lDA.1-36
4 4 , i 22A7007
~~ . GESSAR.II ! 238 NUCLEAR: ISLAND Rev. 10 i
I 1 i
- i. !
I 1 i f i- ATTACHMENT B i TO APPENDIX'lD ; t i j j: . t t. I 1 i s > s l. 238 NUCLEAR ISLAND l TYPE A VARIABLE ASSESSMENT 4 4 i i O , t {. I i l 1 i r f t I' L l 1 I I E i k L
GESSAR II 22A7007 238 NUCLEAR ISLAND' Rnv. 10
~0' ATTACHMENT B TO APPENDIX ID 238 NUCLEAR ISLAND TYPE A VARIABLE ASSESSMENT 1DB.1 INTRODUCTION This attachment describes the basis for selecting the Type A variable list used as a basis for the assessment against the pro-visions of Regulatory Guide 1.97, Revision 2.
1DB.2 APPROACH Regulatory Guide 1.97, Revision 2, defines Type A variables as "Those variables . . . that . . . permit the control room operator to take the specific manually control (s.fety) actions for which no automatic control is provided . . . for design basis accident events." The identification of the Type A. variables are derived from two sources: GESSAR II, Chapter 15, and the Emergency Procedure Guidelines developed by General Electric for the BWR Owners Group. 1DB.3 RESULTS Chapter 15 contains discussions of numerous events not all of which are design basis accidents. Appendix 15A is a Plant Nuclear Safety Operational Analysis (NSOA) which addresses these same events in the following categories: Normal Operations Anticipated Operational Transients Abnormal Operational Transients Design Basis Accidents Special Events [ 1DB.1-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 1DB.3 RESULTS (Continued) Variables associated with normal operations are excluded from further investigation because those activities are planned actions which would not normally be expected to cause a threat to the general public. Because the Probabilistic Risk Assessment (Section 15D.3) shows that the risk to the general public is dominated by transients rather than design basis accidents, all of the above categories (except normal operations) were considered to determine what parameters required operater action. Tables lDB-1 through 1DB-4 list the events considered and the primary variables associated with called-for manual action. The manual action variables are taken from either the NSOA or the Chapter 15 event descriptions. A review of Tables 1DB-1 through 1DB-4 shows that about half of g the identified events required no operator action; the safety actions required are automatically initiated. The required manual actions are summarized in Table 1DB-5 along with the associated variables. The Emergency Procedure Guidelines were also reviewed to deter-mine if there are other variables not specifically identified by Chapter 15 which are associated with required operator actions. Table lDB-5 includes these additional variables and actions which result from a review of the following guidelines: RPV Control Containment Control Some of these variables, especially those related to emergency action, might be considered beyond the scope of the regulatory l guide by virtue of requiring " contingency actions that may also lll 1DB.1-2
A GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 4 ss J 1DB.3 RESULTS (Continued) be identified in written procedures." However, they are included with the list of Type A variables because they define conditions i which require action on the part of operators to respond to safety-related conditions. The final list of Type A variables was derived from the variables ! indicated on Table 1DB-5. Only Reactor Water Temperature (T RPV was deleted since a direct relation between vessel pressure and temperature exists in a boiling water reactor. i 4 O i O 1DB.1-3/lDB.1-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 O)
\- Table 1DB-1 238 NUCLEAR ISLLND ANTICIPATED OPERATIONAL TRANSIENTS Manual Action
- Event Variables 4
i 7 Scram none 8 Loss of Instrument Air none l
'9 Inadvertent HPCS Pump Start none 10 Inadvertent Recirc Pump Start- none 11 Recirc Flow Control Failure (increasing) none 12 Recirc Flow Control Failure (decreasing) T gp, P RPV 13 Recirc Pump Trip -
none () 14 Inadvertent MSIV Closure none 15 Inadvertent SRV Opening T gp, P ! RPV I 16,17 Continuous Rod Withdrawal none [ 18 Loss of Shutdown Cooling T gp, P ppy 19 Shutdown Cooling Increased Flow none 20 Loss of Feedwater T gp, P ppy 21 Loss of Feedwater Heating none l l 22 Feedwater Controller Failure (maximum demand) none [ 23 Pressure Regulatory Failure. (open) none 24 Pressure Regulatory Failure (closed) T gp, P RPV t 25 Turbine Trip (w/ bypass) none 26 Loss of Condenser Vacuum T , SP' RPV 1 27 Generator Trip (w/ bypass) none i 1DB.2-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rcv. 10 Table 1DB-1 238 NUCLEAR ISLAND ANTICIPATED OPERATIONAL TRANSIENTS (Continued) Manual Action
- Event Variables 28 Loss of On-Site AC Power T gp, P RPV' RPV 29 Loss of Off-Site AC Power T gp, P RPV' RPV O
*0ther than for monitoring LRPV and PRPV, securing IIPCS/RCIC when RPV level is controlled, transferring mode control switches as appropriate, or verifying protective actions (see Table 1DB-6 lg for definition of symbols) .
1DB.2-2
. . - . -. .. - .. - - . . - - ~ ~ . . . .- . .. . . -
- GESSAR II
- 22A7007 238 NUCLEAR ISLAND Rev. 10 1
i j C:) Table 1DB-2
- 238 NUCLEAR ISLAND l ABNORMAL OPERATIONAL TRANSIENTS-4 Manual Action
- Event Variables 4 30 Generator Trip (w/o bypass) none 1
31 Turbine Trip (w/o bypass) none 1 32 Improper Fuel Loading none 33 Recirc Pump Seizure T SP' RPV' 34 Recirc Pump Shaft Break T gp, P RPV i i *See Table 1DB-1. ] I () O 1DB.2-3 i
t GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Table 1DB-3 238 NUCLEAR ISLAND DESIGN BASIS ACCIDENTS Manual Action
- Event Variables 35 Control Rod Drop T gp, P ppy 36 Fuel Handling Accident (outside containment) None 37 Loss of Coolant (inside ,
containment) H", P ppy, L ppy, T gp 38 Loss of Coolant (outside containment) T gp, P ppy, L RPV 39 Instrument Line Break (outside drywell) T gp, P RPV' RPV 40 Feedwater Line Break (outside containment) T gp, P ppy, L RPV 41 Gaseous Radwaste System Leak ** O 47 Augmented Offgas Treatment System Failure ** 43 Radioactive Liquid Waste System Failure None 44 Liquid Containing Tank Failures None 45 Fuel Handling Accident (inside containment) None
*See Table 1DB-1. ** Isolation of line based on channel cross-check, alarm, area radiation, process radiation, area temperature, or leak detec-tion system alarms. Action is for a normal operation suutdown.
Isolation is not a required safety action. O 1DB.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 7 O Table 1DB-4
- 238 NUCLEAR ISLAND SPECIAL EVENTS
, Manual Action
- Event Variables 46 Shipping Cask Drop none i ,
47 Anticipated Transient w/o Scram Tgp,PRPV' ! 48 Shutdown Outside Control Room- T gp, ,P RPV 49 Shutdown w/o Control Rods Tgp,Pppy, 1 l *See Table 1DB-1. (:) i l 3 I I l' i l t 4 1DB.2-5
--.,,.,.,-,-.n. .,,...,..r,a,,n-- - _m.,_,,nn,w.,.,.--,-_,,,-,-.,.-......n-,,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 Table lDB-5 238 NUCLEAR ISLAND
SUMMARY
OF MANUAL ACTIONS Manual Action Variable Source
- Initiation of Suppression Pool Cooling T gp F, E Initiation of Shutdown Cooling P ppy F, E Manual Depreosurization Pppy,LRPV F, E Initiation of 11 2 Recombiners H F c
Initiation of Leakage Control Systems F PDW'bRPV Initiate Standby Liquid Control 0,T gp F, E Emergency Action ** if Exceed: E Heat Capacity Limit Pppy,Tgp,LSP Suppression Pool Load Limit Lgp,PRPV Reference Leg Boiling Limit TC' h DW,Tppy Maximum Drywell Temperature T DW Maximum Coctainment Temperature T C Maximum Containment Pressure P C' SP Pressure Suppression Limit P ,L gp Initiation of Containment Sprays T C' C' SP Initiation of Containment Venting P E C l l *E = EPG F = FSAR
** Scram, Emergency Depressurization and/or RPV Flooding.
l O l 1DB.2-6
i i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 10 ( Table 1DB-6 DEFINITION OF SYMBOLS T - Suppression Pool Temperature SP T - Drywell Temperature DW T - Containment Temperature C T ppy - Reactor Water Temperature P - RPV Pressure RPV P - ntainment Pressure C L - RPV Level RPV L gp - Suppression Pool Level 0 - Neutron Flux H - rywe / cntainment Hydrogen Concentration C i s i I 4 l 1 4 i (:) 1DB.2-7/lDB.2-8
t GESSAR II 22A7007 4 238 NUCLEAR ISLAND Rev. 10 ) s O i, : 4
- ATTACHMENT C TO APPENDIX 1D [
I i a 2 TECHNICAL DESCRIPTION FOR l ENHANCED LEVEL INSTRUMENT SYSTEM i i l. ) ATTACHMENT C is PROPRIETARY and is provided O umder eegerete cover. l I i 1 a
- O i
) 1
. _ . _ - . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . ~ _ . , _ _ _ _ . _ . -_ _._ _,,_._ _. -__}}