View source for HL-5729, Submits Proposed Rev to Plant QA Program as Described in Chapter 17 of Unit 2 Fsar,Per 10CFR50.54(a)(3).Rev Would Relocate to Controlled Document Other than Fsar,Units 1 & 2 Lists of SR SSC Comprising Items Covered Under QA Program

{{Adams
| number = ML20202G396
| issue date = 01/27/1999
| title = Submits Proposed Rev to Plant QA Program as Described in Chapter 17 of Unit 2 Fsar,Per 10CFR50.54(a)(3).Rev Would Relocate to Controlled Document Other than Fsar,Units 1 & 2 Lists of SR SSC Comprising Items Covered Under QA Program
| author name = Sumner H
| author affiliation = SOUTHERN NUCLEAR OPERATING CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000321, 05000366
| license number = 
| contact person = 
| document report number = HL-5729, NUDOCS 9902050161
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 34
}}

=Text=
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: Vee President      _.
Op. rete.g Comp y,l.c;                                                                !
                       . Hatch Project Support               40 inverness Parkway                                                                  ?
Post Office Box 1295 Birrningharn, Alabarna 35201 '
Tel 205.992.7279 Fax 205.992.0341                                                             -
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January.27, 1999'                                                                                 M                    ,
Energy to Serve YourWorM"      i
                                                                                            ~
                        ~ Docket Nos. 50-321                                                                                      HL-5729 50-366
                        ; U.S. Nuclear Regulatory Commission
                        . A'ITN:- Document Control Desk '
         -               . Washington, D.C. 20555 Edwin liHatch Nuclear Plant Revision to Ouality Assurance Proaram Ladies and Gentlemen:                                                                                                     l l
: In accordance with the provisions of 10 CFR 50.54(a)(3), S mthern Nuclear Operating Company l
(SNC) hereby proposes a revision to the Plant Hatch Quality Assurance (QA) Program as                                     !
described in Chapter 17 of the Unit 2 Final Safety Analysis Report (FSAR). SNC proposes to
                        . relocate to a controlled document other than the FSAR the Unit I and Unit 2 lists of safety-related
                        ._ structures, systems, and components that comprise the items covered under the QA Program.
Enclosure 1 provides 'a detailed description of the proposed changes and SNC's justification for                          i making the changes. Enclosure 2 provides the corresponding marked-up FSAR pages.                                          i i
The NRC recently approved a consolidation of Quality Assurance Programs for all Entergy sites jj/
                                                                                      ~
l' into one Quality Assurance Program. This SNC proposalis consistent with the Entergy methodology for the identification and control of safety-related structures, systems, and components to which the 10 CFR 50 Appendix B QA program applics.
Based on the guidance of 10 CFR 50.54(a)(3)(iv), SNC understands the enclosed proposed change                      Ofh l will be regarded as accepted by the Commission upon receipt of a letter to that effect from the                      /
                        ' appropriate reviewing office, or 60 days after submittal to the Commission, whichever occurs first.
Should you have any questions in this regard, please contact this office.
Respectfully submitted,
                                      /'
H. L. Sumner, Jr.
CLT/JDR/eb -                0500J5

==Enclosures:==
: 1. Description of and Justification for Proposed Changes to QA Program
: 2. ' Corresponding Marked-Up FSAR Pages cc: (Sde next page.)
9902050161'990127<                             ^
            -PDR          ADOCK~05000321-np ,       P                               PDR                  ,
e

  . o U. S. Nuclear Regulatory Commission Page 2
       -January 27, 1999 cc:   Southern Nuclear Operatina Company Mr. P. H. Wells, Nuclear Plant General Manager Document Management- A2.001 IJ.S. Nuclear Reaulatory Commission. Washinaton. D.C.
Mr. L. N. Olshan, Project Manager - Hatch U.S. Nuclear Regulatory Commission. Region 11 Mr. L. A. Reyes, Regional Administrator Mr. J. T. Munday, Senior Resident Inspector - Hatch i
i l
l l
l K
       . HL-5729 -

 ,     ._ _ ,           _         . _ _ _ . . . ___.y.._.                      _ _ _ . _ - _ _ _ . _ . _ _                _. ._ .. ..
                      ,                                                  Enclosure 1 Edwin I. Hatch Nuclear Plant Description of and Justification for Proposed Changes to the Ouality Assurance Pronram Desriation of Proposed Channes Sou:hcrn Nuclear Operating Company (SNC) proposes to revise the Plant Hatch Quality Assurance (QA) Program description in the Hatch Unit I and Unit 2 FSARs. SNC proposes to relocate the current Unit 1 "Q" Index Checklist and Unit 2 List of Safety-Related Structures, Systems, and                      -
Components from the FSAR QA Program sections, controlled under 10 CFR 50.54(a), to Volume 3 of 4
the System Evaluation Document (SED), the Safety Components List (SCL), a licensee document to be controlled under 10 CFR 50.59.
He specific proposed changes to the QA Program are:
: 1. Delete table D.9-1, "Q" Index Checklist, from Appendix D of the Unit 1 FSAR.
: 2. Delete table 17.2-2, List of Safety-Related Structures, Systems, and Components, from the Unit 2 FSAR.
: 3. Revise section D.9 of the Unit 1 FSAR to remove mention of tables D.9-1 and 17.2-2.
: 4. Revise section 17 2.2.I of the Unit 2 FSAR to:                                                                ,
l
                          . Remove mention of tables D.9-1 and 17.2-2.
l
                          . Identify the criteria, specifically safety-related, used to identify the items to which the QA            l Program applies.
                          . Remove current criteria previously used in assigning methods or levels of QA.
                          . Incorporate by reference into the QA Program section of the Unit i and Unit 2 FSARs the SCL.
The specific proposed change to the SCL is to consolidate the Unit I and Unit 2 QA Program tables from the FSARs into the SCL. Any differences between the safety-related structures, systems, and components (SSCs) in the tables and the SCL will be reconciled and evaluated under the provisions of 10 CFR 50.59. This reconciliation will be completed prior to deletion of the Unit I and Unit 2 QA Program tables from the FSARs.
1HL-5729                                                  El-1

Enclosure 1 Description of and Justification for Proposed Changes to the Quality Assurance Program Backaround Unit 2 FSAR section 17.2.2.1 states that the items to be covered under the QA Program are designated as safety-related items and are listed in FSAR table 17.2-2. Similarly, the Unit I list of safety-related SSCs is in Unit 1 FSAR table D.9-1. These tables have existed in the QA Program description since receipt of the operating licenses for Plant Hatch Units 1 and 2.
On July 8,1983, the NRC issued Generic Letter 83-28 (GL 83-28), " Required Actions Based on Generic Implications of Salem ATWS Event." In the generic letter, the NRC stated that issues related to reactor trip reliability and general management capability would be addressed. The generic letter required specific actions on the part oflicensees. Among these actions was a .
requirement to assure that all components necessary for accomplishing required safety-related functions are properly identified in documents, procedures, and information-handling systems used to control safety-related plant activities.
Georgia Power Company (GPC), the operating licensee of Plant Hatch until March 22,1997, at which time the license responsibilities were transferred to SNC, developed the SED which consists of three volumes, in response to GL 83-28. Volumes I and 2 contain brief descriptions of designated systems in Unit I and Unit 2, respectively, that may contain safety-related components.
Volume 3, the SCL, contains the list of safety-related components. Identification of these components is based on a functional analysis of the system and the individual component.
                  - More specifically, the SCL identifies the components necessary for accomplishing the safety-related functions associated with the four design basis accidents (DBAs) (e.g., main steam line break, control rod drop accident, loss of coolant accident, and refueling accident.) These accidents are the most limiting with respect to their radiological consequences. In addition, the scope of the SCL includes identification of any components which, although not required for accomplishing a safety-related function, were treated as safety-related by a previous licensee commitment.
By {{letter dated|date=September 28, 1988|text=letter dated September 28,1988}}, GPC designated the SED as the vehicle that would satisfy the GL 83-28 requirement to identify all components necessary for accomplishing required safety-related functions.
Reason for Proposed Channes
                  . Beginning in 1988, with the designation of Volume 3 of the SED, the SCL, as the vehicle that identifies the components for accomplishing safety-related functions, programmatic control of the scope of the QA program has been adequately maintained using the SCL and the lists of safety-related SSCs in the FSAR QA Program description. In the succeeding i1 years, although the lists in the SCL and FSAR QA Program description were updated to varying degrees, the emphasis has been on maintaining the accuracy of the SCL. As well, an increasing number ofinconsistencies between the two lists have been identified. As a result, a corrective action to address these inconsistencies and clarify the discussion of program scope in the QA Program description was
;                    identified.
l-                   HL 5729 '                                                        El-2

Enclosure 1 Description of and Justification for Proposed Changes to the Quality Assurance Program He review of the inconsistencies, particularly in the lists of safety-related SSCs in the QA
       ~ Program, and the consideration of the information described previously in the Background section, have led SNC to propose the previously described QA Program changes.
Justification for Proposed Channes Appendix B to 10 CFR 50, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, states in the introduction that the pertinent requirements of the appendix apply to all activities affecting the safety-related functions of SSCs. In Section II, Quality Assurance Program, Appendix B requires the identification of SSCs covered under the QA Program.
In addition to the above regulatory requirements, the following NRC documents provide insight into methods ofimplementation of the regulatory requirements:
* NUREG 0800, " Standard Review Plan," (SRP) section 17.2, Quality Assurance During the            i Operations Phase, Revision 2, July 1981, e   NUREG 0800, " Standard Review Plan," (SRP) section 17.3, Quality Assurance Program Description, Revision 0, August 1990, e    Generic Letter 83-28, and i
e   NRC Inspection Program Manual.                                                                    i l
SRP section 17.2 looks to review the list of SSCs covered by the QA Program, and refers to the acceptance criteria in SRP section 17.1, which requires a commitment in the QA Program that all activities affecting SSCs important to safety will be subject to controls of the QA Program. SRP section 17.3 states that the QA Program is to include the criteria used to identify the items and activities to which the QA Program applies. In GL 83-28, the NRC required all licensees to properly identify all components necessary for accomplishing safety-related functions in documents, procedures, and information-handling systems. NRC Inspection Manual, Inspection Procedure 35740-30, QA Program (QA/QC) Administration, refers to chapter 17 of the FSAR as describing or referencing the SSCs covered by the QA Program. The Inspection Procedure anticipates lower-tier document implementation of the QA Program requirements.
      . The documents mentioned above indicate that there is a requirement in 10 CFR. 50 Appendix B to include in the QA Program the criteria for identifying the SSCs covered by the QA Program and to maintain a list of such SSCs. Further, the documents show that, although there is a requirement to have and maintain a list of SSCs covered by the QA Program, there is no requirement or expectation on the part of the NRC that the list reside in the QA Program description.
Relocation of the lists covered by the QA Program to the SCL results in changes to the list being controlled under 10 CFR 50.59. Component information in the SCL is considered an input into the design control process, and revisions to the SCL are typically created by the plant design change process. Plant design changes are evaluated in accordance with 10 CFR 50.59, as such, a change HL-5729                                             El-3

Enclosure i Description of and Justification for Proposed Changes to the Quality Assurance Program to the SCL will receive the same 10 CFR 50.59 treatment as a design change. Accordingly, as                  ,
SCL revisions are part of the design change control process, evaluation under 10 CFR 50.54(a)                l would not be applicable.                                                                                    I l
Unit 2 FSAR section 17.2.2.1 currently lists criteria for assigning levels of QA. The criteria will
       ' be replaced with criteria for identifying the SSCs covered by the QA Program. These criteria are            l administratively and procedurally controlled. The proposed change meets the requirement that the            i criteria for identifying SSCs to which the QA Program applies be included in the QA Program description.
Basis for Concludine Annendir B Criteria Are Still Met                                                      j l
SNC complies with 10 CFR 50 Appendix B by stating in the Unit 2 FSAR, Appendix A, that SNC                  I commits to conforming to Regulatory Guide 1.33, " Quality Assurance Program Requirements -
(Operation)." This commitment makes the 10 CFR 50 Appendix B requirements apply to all safety-related SSCsc Regulatory Guide 1.33 endorses ANSI 18.7, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," and ANSI /ASME N45.2,
         " Quality Assurance Program Requirements for Nuclear Facilities." Both of these ANSI standards provide guidance for achieving compliance with all 18 criteria of 10 CFR 50 Appendix B. The proposed change does not compromise conformance with Regulatory Guide 1.33 or the QA Program as approved by the NRC.
Fundamentally the proposed changes will enhance assurance that programmatic control of the scope of the QA program will be maintained in the future under a consolidated listing of safety-related SSCs in the SED, specifically the SCL. Such consolidation assures inconsistencies are addressed initially and enhances the ability to maintain list accuracy and completeness since there will now be a single list. Although there is a requirement to have and maintain a list of SSCs which are covered by the QA program, there is no requirement or expectation on the part of the NRC that the list reside in the QA Program description. Therefore, migration of the list of SSCs covered by the QA program from the QA Program description to the SCL does not diminish compliance with the criteria of 10 CFR 50 Appendix B.
HL-5729                                           El-4

l Enclosure 2 Edwin 1. Hatch Nuclear Plant                                                        )
Revision to Quality Assurance Program Correspondina Marked-Up FSAR Panes The following marked-up Unit I and Unit 2 Final Safety Analysis Report pages are included in this enclosure:
Unit i Section                                      Page(s)
H List ofTables                                xlvii Appendix D List ofTables                     vi Appendix D-9                                 D.9-1 Table D.9-1                                  Sheets 1-9 Unit 2 Section                                      Pane (s)
List ofTables                                lvii Chapter 17                                   iv 17.2.2.1 and Insert                           17.2-8, 17.2-9 Table 17.2-2                                 Sheets 1-9 HL-5729                                      E2-1

   ._         , ~.. _ - . _ _ __..- _ ._.. . _ . _ ._ _ - _._ _ . _                                               _
l l
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                                                                                             .HNP.-1-FSAR                                          1 l.
   ,=
I                                                                                                                                                   1
                                                                                 ' LIST OF TABLES - (CONTINUED)                                     l I
                          -Table'No.                                                          Table' Title D . 5 -l'                           -Quality Assurance Functiras Required.                                       )
e D.2 1                                 G" Inder    C.'.   :klint
;                                 E.1-1                                Deposition' Velocity Coefficients E.1-2                                 Diffusion Coefficients
;                                 E.1-3                                Shielding--' Dose Reduction Factors L-                            1E.2-1                                 ' Percent Occurrence of Good Observations From the                            >
l l'
BNL Site for Various Directions and Windspeeds                               i
                                . . .                                  (6464 h During-1963)                                                        1
                                                                      '(deleted)
E.2-2                                                                                                              l
                              -E.2-3                                   (deleted)                                                                    '
l~                               E.2-4                                .(deleted);
'-                            - E '. 2 - 5                            '(deleted)'                                                                  i
                                                                     ' Percent Occurrence of Good Observations From the l
l                                E.2-6 i
!                                                                     BNL Site for Various Directions and Windspeeds                                !
(6464 h During 1963),All Stabilities                                        i l                             :E.2-7                                   1963 BNL Environmental" Monitoring- Monthly Average                        1 l                                                                     Argon-41. Radiation Levels,. Station' Locations (mR/ week)'
L                               E.2-8                                 Average Annual Gamma Dose (mR/ year) ;for BGRR 1963 -                        ;
E
'"                                                                     Predicted'and Observed              . .
E.3-1;                                Noble'Radiogas. Release Rate to.Envipons From 12-Bed.
L                                                                     Recombiner/ Charcoal System at 40.sf / min Flow (mci /s)                     i l                               E.3-2                                ' Noble Radiogas Release Rate to Environs From-                               i
[                                                                     Gland-Seal System-(mci /s)-
;                               E . 3 -3:                             Annual Average < Gamma. Detector Dose at Ground Level-r From Continuous Release Rate of 6930 mci /s From L'
l 12-Bed RECHAR System; (mrem / year) .
E.3                                Annual Average. Gamma Detector Dose at' Ground Level
                                                                    'From Continuous Release of 433 mci /s From Gland-Seal L
                                    ._                                System (mrem / year):
l,'
                            ~E.3-5                                    Annual Average Gamma Dose at. Site Boundary From
              ~                                                       Continuous Building Ventilation Release of 278 mci /s                       1 j
p? f p                                                                                                                                                   !
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[.                                                                                                                                                .)
REV 1                 7/83
;)
~                                                                                                                   -REV 3                 7/85 REV'S                 7/87 REV 9A               10/90     i
;                                                                                                xlvii               REV 15B               7/97 l l
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l                    -

  .. -       . .     -.    . . . . . . - .             -       . . . - - . .~ - .      , _ .     .... . . . .
1 HNP-1-FSAR-D                                             1 LIST OF TABLES Table No.                           Table Title D.5-1                Quality Assurance Functions Required D.9 1                -Q' Index Check.tist                                                   j O     i l
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vi                       REV 3      7/85             l

HNP-1-FSAR-D                                                                      '
O                                                                                                                                                 \
U                D.9        OUALITY ASSURANCE PROGRAM-OPEDATLQHE The operations quality assurance program for Edwin I. Hatch Nuclear Plant-Unit 1 (HNP-1) is the same as that for HNP-2.
Hc, wever, minor administrative and operational differences Q               caused by differences in design limits and different standard and code commitments exist. The operations quality assurance program l
is described in section 17.2 of HNP-2-FSAR.
Check                                                                                              '
                .T [ "Q"        In                       t is a li ing of tho                         structures Aystems             nd comp ents that ontribute                               the prevent n of or Os             miti            on of un e risk t the healt consegue es of acci nts which c nd safety                the public d cause the HNP- "Q" dex Che           ist is cop ained in                           le D.9-1, a                the HNP-              Q" /
Index          ecklist is,c6ntained i table 17.2-2 f the HNP- FSAR/ j t
1 v                                                                                                                                                  1 l
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                                   .0         Buildinas and Structures i
n..
1.1   Intake Structure (common to both units)                                                                          -
I 2  Main Stack.(common-to both units)
   }A .g                                      1.   ' Buildings
                                                      .3.1      Reactor Building (including                                     ent-fuel storage pool and new-fuel storage v lt) 1.,   2     Control Building (common t both units) 1.3.        Diesel Generator Building common to both units) 1.3.4       Parts of Other Structure                                     Housing and/or upporting Class 1 Equi ent
: 1.   .4'1
                                                                        .             Condensate                        orage. Tank Enclosure 1.3 4.2               Liquid Nit ogen Storage Tank and                                                  ,
Foundatio                                                                         !
1.3.4.3               Diesel G nerator Fuel Oil Storage                                                 i Tanks 1.4   Pr. mary.Contain                  nt 12 . 0         Mechanical Systems and                             nonents 2.1   Nuclear Boiler S                     tem 2.1.1       Reactor            ressure                      asel (RPV) 2.1.1 1               Vessel 2.1.    .2            Head 2.1 1.3               Head Studs
      -                                                         2.  .1.4              Control Rod Ho sings (including guide
{                                                                                tubes)
N2 -                                                        .1.1.5              Stabilizers 2.1.        Internals 2.1.2.1               Dryer / Separator.Holddo n Bolts
   .[
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2.1.2.2 2.1.2.3 Feedwater Sparger Core Shroud l
2.1.2.4              Jet Pump Assemblies 2.1.2.5               Core Support Structures 2.1.2.6               Flow Baffles, Guides, and Or ices 2.1.2.7               Core Spray Sparger 2.1.2.8               Incore Flux Monitor Housing ls,/7                              .                          2.1.2.9              Power-Range Monitors 2.1.2.10             Control Rods (inside RPV)
                                                                                                                                     "'_ . , e .m. -, , $. hs

i._--.          . , _ . - . . . . _ _ .              , _ .      _ . . _ - _ _ . . . . _ _ . _ . . _       . . .        4 _.         -. _ .
I l
r~r                                                                ::"" l-FSAR-D j )
i TADLE D.3-1 (6nssA 4 UF 9) 2.1.3   Valves 2.1.3.1 . Steam Isolation Valves 2.1.3.2 Feedwater Check Valves 2.1.3.3 Safety Relief Valves
        . ,                                          2.1.3.4 ' Safety Relief Valve Dis                                arge Vacuum
        /~N                                                             Breakers                                                             ;
(_)                '
2.1. Pipe and Fittings                                                                        ;
2.1.4.1            Up to Second Cont nment Isolation Valve External t Drywell 2.1.4.2            SRV Discharge H aders 2.1.5   Ins       umentation and C                         trol (essential)                     :
2.2      Recirculation System Recircul tion Pum a and Motors (structural 2.2.1-integrity ost-d sign basis accident-(DBA))
7-s                                   2.2.2   Isolation                  Ive            (suction valve - structural                   :
integrity on y
:(                                      j2.2.3   Pipe and Fitt gs                                                                        ,
2.2.4   Instrumentat o and Controls (essential)
                                            '2.2.5   Recirculati n P p Trips                                                                 ;
2.3      Control Rod Driv                System
                                            '2.3.1   Hydrau         c System                                                                 j 2.3. .1            Valves (essent'al)                                                   !
2.3       .2       Scram Discharge Header.                                              ;
: 2.  .1. 3          Pipe and Fitting (essential) j       f                               2.3.2    ydraulic Control Unit (HCU                                                              ;
                                                    .:2.3.2.1           HCU Package 2.3.2.2            Foundations and Boltin                                                ,
2.3.2.3. Pipe and Fitting l
                                               .3.3 ~ Rod Drive Mechanisms                                                                    l
            )
2.3.3.1             Support Assembly 2.3.3.2            Piston Mechanism                                                      i
                                                    ~2.3.3.3            0-Rings.and Seals                                                     j 2.3.3.4            Couplings and Latches                                                  i Instrumentation and Controls'(partial)
(                                      2.3.4
                                                                                                                                              )
REV 3       7/       l REV 8 .7/9 l
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                                            -MMP-1-E5AR-v -
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TAOLE O.0-1 ( SIIEET 3 OF 9 )
                     .4 Standby Liquid Control System                                       j 1
,r 3                    2.4.1    Storage Tank                                               l
()  '
0.4.2 4.3 Pumps and Motors Explosive Valves j
: 2.   .4  Accumulators                                               )
2.4 5    Relief Valves 2.4. Pipe and Fittings (partial) gy                   2.4.7    Instrumentation and Controls (es ential) i     1
   \   /
2.5 Residual Heat Removal (RHR) System 2.5.1    Hea   Exchangers 2.5.2    Pump    and Motors 2.5.3    Relie Valves 2.5.4    Check V lves 2.5.5    Control alves 2.5.6    Pipe and 'ttings (pa tial) 2.5.7    Instrumenta ion and ontrols (essential) 2.6 Core Spray System rN                     2.6.1    Pumps and Motor
(_)                    2.6.2 2.6.3 Control Valves Check Valves 2.6.4    Pipe and Fit ings 2.6.5    Instrument ion and       ntrols (essential) 2.7 High-Pressure C       lant Injectio     (HPCI) System 2.7.1    Turbin    and Pumps 2.7.   .1   Turbine i
2.7 1.2     Steam Valves
: 2.  .1.3    Pumps                                          i i
I~'l                    2.7.2      alves L-]                                                                                         1 2.7.2.1     Control Valves                                 l 2.7.2.2     Relief Valves 2.7.2.3     Check Valves 2     .3 Pipe and Fittings (partial) f~')
x--                       .7.4   Instrumentation and Controls (essential                   J l
l l
1
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2Ee 3 7,2 3 X

H!!? 1 F3AR-D TABLE D.O 1 (O"ECT 4 OF 0) l
       )   ,-   ,
: 2. 8' Reactor Core Isolation Cooling (RCIC) System 2.8.1      Turbine and Pump p._
2.8.1.1       Turbine 2.8.1.2       Steam Valves 2.8.1.3       Pump 2.8.9      Valves
   -                               2.8.2.1       Control Valves 2.8.2.2       Check Valves l
                                      .8.2.3     Relief Valves 2.8.3      Pi      and Fittings (par al) 2.8.4      Inst umentation and C trols (essential)                  :
2.9   Reactor Buildi g Ventilatio            System i
tilatio     System                         j 2.9.1      Normal V l
2.9.1.1          ta     Dampers (forming secondary       I co     inment boundary)                     1
()
s 2.9.1.2       Ex    ust Dampers (forming secondary       )
  '~
c9 t inment boundary) 2.9.1.3       chfeg ard Equipment Emergency Coolers 2.9.1.4         nstru entation and Controls-(for items 2.   .1.1 through 2.9.1.3 l
2.10. Standby Gas T eatment System i
2.10.1      Preheaters 2.10.2      Fi    ers                                               ;
2.10.3     F   s 2.10.4       alves 2.10.5       ucting (filter housing and       ortions forming
/~
(_3)                                boundary of secondary.contai ment)
Instrumentation and Controls ( ssential)
                                                         ~
2.10.
2.11  Fue Storage, Refueling, Handling and S rvicing System r^x-                    2   1.1    Fuel
(_)                        .11.2    New-Fuel Racks 2.11.3 Spent-Fuel Racks 2.11.4 Control Rod Racks 2.11.5 Damaged Fuel Racks 2.11.6' Spent-Fuel Cask
,3
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TABLE D.5-1 (5HEEI 5 GF 5)
            \-                     2.11.7  Refueling Platform
('/
(_ .                         2.11.8  RPV Head Strongback
                                 '2.11.9   Dryer / Separator Sling
                                    .11.10  Service Platform 11.11 Bridge Crane
['
t 2.12      Tra   raing Incore Probe (TIP) System (                   sential) 2.12.1   Tube Indexer 2.12.2     torage Cask 2.12.3     (sposal Cask 2.12.4   Cable Drive 2.12.5   She r Valves 9
2.13      Radiation Mon toring 2.13.1   Neutron onitoring System (partial) 2.13.2   Process R diatio Monitoring System (partial) 2.14. Reactor Water Clean                     stem j ' )-                          2.14.1   Pipe and Fit              gs (partial) 2.14.2   Instrument 3 io            and Controls (essential) 2.14.3   Isolation valve                                                   l 2.15      Plant Service W ter Pumps nd Motors, Associated Piping, Valves Instrumenta ion and Controls (essential),
and Heat Exch ngers 2.16' Plant Instr ment Air System (a                     required for safe shutdown) 2.17      Main Co rol Room Control Panels                      d Devices for                :
f-                             Seismi Class 1 Equipment                                            ,
kJ                   2.18      Pipe Supports and Hangers for Seismi                    Class 1 Pipe 2.19      Di sel Generator Building Ventilation S stem
    ,s                  2.20        tation Battery and Emergency Switchgear                    coms 1                       Ventilation System (s)
' . V(
: 2.        Drywell Pneumatic System (as required for sa
                                 ' shutdown) 2.22 . Nitrogen Makeup System (as required for DBA miti ation)
()                   2.23      Drywell Purge System (as required for DBA mitigati                       )        l REV 3    7/S5       I
                                                                                             - REV 5    7/90       ]

l "NP-1-F5ax-u (m,)     ,
             .                 T.^.BLE O . 0 - 1 (5HEEI 6 06 9) 24 Cranes and Hoisting Equipment gs                2.24.1   Cranes and Hoists in Reactor Buildin
'(~j                         (including supports) (structural in grity required post-design basis earthqu e) 2.25  Fu    Pool Cooling and Cleanup 2.25. Pipe and Fittings
()
   <s 2.25.2   Valves 2.26  RHR Servi e Water System 2.26.1   Cro s-Connect Piping Sys em (within second auto    tic isolation val e)                             l 2.26.2   Pipin 2.26.3   Pumps 2.26.4   Pump Mot rs 2.26.5   Isolation Valves 2.26.6   Other Valv 2.26.7   Electrical        dul s (with safety function) 2.26.8   Cable (with s f ty function)
   ,-~
(            2.27 Control Room Ventilat'o         System I
2.27.1   Pre-Filters and'Fi'ters 2.27.2   Fans 2.27.3   Valves 2.27.4   Dampers /
2.27.5   Ductin 2.27.6   Louve 2.27.7    Inst mentation and Contr is (essential)                  l 2.27.8   Hea rs, Coolers, and Con nsing Units 2.28 Refueling     loor Ventilation System I   ')            2.28.1    Intake and Exhaust Dampers
<k/                 2.28.2    Instrumentation and Controls (evsential) l              2.29  Inta e Structure Ventilation System 2.30  Le k Detection System l,i a
ik l           2.31   hip Restraints and Jet Impingement Barriers 2.3   Primary Containment Isolation System                              i 2.32.1   Valves                                                   ,
(''T               2.32.2   Instrumentation and Controls                             l L-)                                                                                  l l
                                                                      .m. , , , . , ,
1

p\
  /            .
nup.1_rgan_n TABLE L.5-1 (5HEET 7 0" 3)              .
('Y
    '%/
                  .33   Reactor Protection System                                                         )
1 2.33.1    Instrumentation and Controls 2.34   Post-Accident Monitoring System t''           2.35  'T  us Drainage and Purification Syste
   \wsl)                                  .
l H2.35 1    Isolation Valves                                                        l 2.35. Pipe and Fittings (essent         1)                                    1 2.35.3    Instrumentation and Cont ols (essential)                                l l
2.36   Sampling    ystem 2.36.1    Pri ary Containment 2 0 2 Analyzers                                     )
2.36.2    Valv s (essential)                                                      l 2.36.3    Pipe nd Fittings (essential) 2.36.4    Instru entation nd Controls (essential) 2.37   Reactor Building Clos            Cooling Water System r                    (inside containme t -          to external containment                            l
,.(   S)                isolation valves) 3.0 Electrical 3.1  Switchgear Asso iated Wit            Engineering Safeguards 3.1.1    4160-V Emergency Bus s                                                     l 3.1.    .1 4.16-kV:Switc gear Bus (1E R22-S005)                            l 3.1.1.2    4.16-kV Switch ear Bus (1F R22-SOO6)                            l
: 3.  .1.3   4.16-kV Switchg ar Bus (1G R22-SOO7) 3.1.2     OO-V Emergency Load Cente s and Bus Ducts
[% )\..
3.1.2.1    600-V Load Centers        us 1C (R23-SOO3) 3.1.2.2    600-V Load Centers B s 1D (R23-SOO4) 3.1.2.3    600-V Bus Ducts Assoc ated with Load Centers
    '- -         3.2    witchboards and Panels 1
3.2.1    Control Boards 3.2.1.1    Reactor and Containment Cooli g and f
   -~                                     Isolation Board V
REV 3 een o
                                                                                            /05
                                                                                          ,\e   n
                                                                             -nu,       w e f

MMD-1-FS?.R O-TABLE L.5-1 (snest e Of v) 1           1                          3.2.1.2            Power Range Neutron Monitoring                                                 binet 3.2.1.3            Channel A Primary Isolation an Reactor Protection System Vertical Bo d 3.2.1.4            Channel B Primary Isolation                                                 d RPS Vertical Board 3.2.1.5             Steam,.Feedwater Condensat                                              ,
:                                             Circulating, and~ Service ater
                   .-                                              Benchboard 3.2.1.6 Emergency Diesel Gener or No.'1 (BB+VB) 3.2.1.7.' Emergency Diesel Gene ator No. 2 (BB+VB)
                                                  .2.1.8 Emergency Diesel Gen rator No. 3 (BB+VB) 3.2.2      Pr         ective Relay Board l
3.2.2 1            Channel A RHR R                        ay Vertical Board 3.2.2.             Channel B RHR                     elay vertical Board-V                                   3.2.2.3            HPCI Relay Ve tical Board 3.2.2.4               CIC Relay V rtical Board 3.2.2.5                      oard Is ation Valve Relay Vertical
               ~ '
B          rd
                                               '3.2.2.6 Ou oard solation Valve Relay Vertical                                                                  ;
  ;                                                                Boa                                                                                         :
3.2.2.7 Chann                      A CS Relay Vertical Relay Board 3.2.2.8 Chann                      B CS Relay Vertical Board 3.2.2.9 Auto B1 ckdown Relay Vertical Board
                                               -3.2.2.10 He ing, Ventilating, and                                                              .
                                                                                                                                                             -)'
A -Condi ioning Control Boards.
11-P654 nd H11P657) 3.2.2.11 ocal Panel Continuing Essential System Contr 1s'or Components 3.2.3      Motor Control Centers ( CCs) t           '
                                               .3.      .3.1       600-ac' Essential MCCs (reactor building) 3       .3.2       250-dc Essential. CCs (reactor building cO(
D
                                                  .2.3.3.~600/208 V-ac Esse                                     ial MCCs (diesel building) 600/208 V-ac Essent                               1 MCCa (control
             ,                                  3.2.3.4 building)
: 3.   .4    de Switchgear.in Control Buildi 3.2.4.1            250 V-dc Switchgear Bus                                                  (R22-sol 6)        !
3.2.4.2            250 V-dc Switchgear Bus 1                                                (R22-sol 7)        i l
l RZ'l o    /vu  i~
                                                                                                                                                     \s i
                                 - . _.    -      _           -             _.         . - - -                         . _ . . _ _ . . _ - .           _. 1

_      m   _      _ . ~ . . . . _ _ _ _ _ .__ . . ~ . . _ . - . _ . _ . . . _ _ . . . , _ _ _                                        . ,_ _.__ _ ..             . . . . .
~
                                                                                                    - uup-1-yg,*,n_ y
                                                                                                                             -                                                 i
                                                                                  - T.^.SLE 0. 3 -1. ( SiiEEI y vr yJ 3.3              Raceways Associated with Engineering Safeguards
: 3.         .1           Conduit l Supports 3.3 2                   Cable Tray Supports 3.3.                    Pull Boxes.and Junction Boxes
     .                                                3.3.4                   Underground Ducts, Fittings, and E casement M                               3.4              Cables A                          ociated With Engineering Safe uards                                                   e 3.4.l'                   Inst ument Cables 3.4.2                   Emerg cy 600-V and 208/120-                   Power and Control Cables
:                                                     3.4.3                   4160-V                wer Cables 3.5              dc Equipment 3'.5.1                  Battery and                 ce'ssori s 3.5.1.1               125/25 -V tation Batteries 3.5.1.2               125-V E     gency Diesel. Batteries                        l-i f .O '
.:\/-                                                 3.5.2                   Battery Charger                n ts                                              l              :
3.5.2.1               Batt ry Cha    er' Units for Station                                      -
Ba eries                                                   l              ,
3.5.2.2                 attery Charger Units for Emergency                       l Diesel Batteries                                           l 3 .' 6           Generators 3.6.1                   Eme               ency Diesel Generator Set
     ...                                              3.6.2                   Ne            ral Grounding Resistors for Emergency Diesel G nerator Sets                                                                                   l 3.7              Misce                    aneous Electrical Items                                                                         i 3.7                     ac/dc Essential Distribution Cabinet                    (control and diesel buildings)
Drywell Penetrations O                                              3 7.2
                                                          .7.3                Lighting and Miscellaneous' Distribution Transformers for Emergency Services l
l
                                                                                                                                 - ALV 3              //53
                                                                                                                                  - RCV 0             7/00 lI.                                                                                                                                                                          1
                       .                        ,                   ,                         ,     <                          -          -                . -              -l

   ?                                                       HNP-2-FSAR LIST OF TABLES (CONTINUED)
Table No.                            Table Title
                     '15.1-40         Main Steam Line Break Accident (NRC Analysis)
Activity Released to the Environment (Ci)
                  .15.1-41'           Main Steam Line Break Accident (NRC Analysis)
Radiological Effects of Puff Release (Release Height of 30 m) 15.1-42        Main Steam Line Break Accident Parameters Tabulated for-Postulated Accident Analyses
                     '15.1-43         Fuel-Handling Accident (Realistic Analysis)                          ,
Activity Airborne in the Refueling Area (Ci)                          '
15.1-44        Fuel-Handling Accident (Realistic Analysis)
Activity Released to the Environment (Ci)                            l 15.1       Fuel-Handling Accident-(Realistic Analysis)                         ~
Radiological Effects (H = 120 m)
                   . 15.1-46        Fuel-Handling Accident (NRC Analysis) Activity Airborne in the Refueling Area (Ci)                                 i 15.1-47        Fuel-Handling Accident (NRC Analysis) Radiological Effects 15.'l-48        Fuel-Handling Accident - Parameters Tabulated for                   l Postulated Accident Analyses                                        ;
15.1-49         Summary of Results Turbine and Generator Trips                       l Without Bypass (No Direct Scram, No RPT) 15.1-50         Event Acceptance Limits                                       l 15A-1           Moderate- and High-Energy Lines 15A-2            Blowdown Data for High-Energy Line Breaks 15A-3            Equipment Required and/or Preferred for Use in
                        .             Reactor Shutdown Following a High-Energy Line Break Outside Primary Containment 15C-1            BWR Operating States 15C-2            Events Applicable in Each BWR Operating State 17.2-1           Georgia Power Company E. I. Hatch Nuclear Plant QA/QC Procedures l''.2 2          1.ict Of C:f;t-i ";l;t;d Statee., Sysi.mu , cui c~~cacar e A.38-1            (deleted)
I REV 3   7/85
        "                                                                              REV 7   7/89 e
REV 9A 10/90 REV 9C  7/91 lvii                     REV 15A 2/97 I)

HNP-2-FSAR-17 LIST OF TABLES Table No.                                     Table Title
{'
17.2-1        Georgia Power Company E. I. Hatch Nuclear Plant
                                                     .QA/QC. Procedures 17.2-2         List of fif=&y-Ral= tad Structurec, Oysteme, and
                                    .                 Cc=;cT. Tt=
(
r k
         .[..
L y
                                                                                                                                          ,.      i c.
REV 7                7/89 iv             REV 9C               7/91 s - ---
                    . . . i . . .          .

  +
HNP-2-FSAR-17                                     -
17.2.1.2.10     Plant Training and Emergency Preparedness Manager The plant training and emergency preparedness manager reports to the PSAGM and is responsible for the preparation, scheduling,                          ^
presentation, and administration of training programs at Plant                  '
Hatch to ensure that employees can accomplish their job functions and that the programs comply with Nuclear Regulatory Commission (NRC) requirements.      He is responsible for training programs to qualify students for required NRC licenses; for maintenance and modification of the simulator; and for training programs in plant               t' safety, plant operations, security, health physics, chemistry, and              i fire brigade training.       Onsite activities relating to the emergency preparedness program are also the responsibility of this position, including associated training.
17.2.1.2.11      Nuclear Safety and Compliance Manager The nuclear safety and compliance manager reports to the NPGM and is responsible for mana Specifications program,ging   and the for PRB,  administering identifying   programthecommitments.
Technical He is also responsible for coordinating the Operating Experience Review Program and for NRC/SAER open item control.
f r .
17.2.1.2.12       Plant Modifications and Maintenance Support Manager The plant modifications and maintenance support manager reports to the PSAGM and is responsible for the implementation of major design change requests. He is also responsible for maintenance support contractors.
17.2.2      QA PROGRAM DURING PLANT OPERATIONS 17.2.2.1      Policies. Goals, and Obiectives The policies, goals, and objectives of the QA program during plant operations are summarized in the introductory paragraph of                                ,
this section.f The it        s to be     vered under he QA p ' gram a l
PEEsi ated s safe -related ' ems and are isted f gener                    i g     ance 'n tabl      7.2-2.      e HNP-1 lis of safe -relat                 I l
truct es, syc ems, and omponents c                                             {
be foun in tab      D.9-1        \       I the      P-1-FS    . The l'    is prepar      and mai ained c rrent b th A/E d ing plant esign, cons uction,                d opera on.                         i visio     will be    cumented th ugh the
* quired         AR upda e                   '
l        proce     . Additi al guidance o be con dered in assigni j
(meth ds or leve s of QA are s follows.             ;
OSee, insert CLhhed*
                                                                                               ~
:                                                         REV 5 7/87 REV 9C     7/91 REV 6 7/88 REV 11B    1/93 REV 7 7/89 REV 14D    7/96 17.2-8       REV 8 7/90 REV 15B    7/97    l

l l
INSERT FOR HNP-2-FSAR SECTION 17.2, PAGE 17.2-8 I
The items covered under the QA Program are the safety-related structures, systems, and components. The HNP-1 and HNP-2 lists of safety related structures, systems, and components are contained in Volume 3 of the System Evaluation Document (SED),
Safety Components List (SCL). The SCL is prepared and maintained current by the A/E.
The criteria used for identifying the safety-related structures, systems, and components are described in the SCL and are administratively and procedurally controlled by the A/E.
The SCL is incorporated by reference in the HNP-1 and HNP-2 FSARs, and changes to the SCL are evaluated under the provisions of 10 CFR 50.59.

i L
HNP-2-FSAR-17 l              .
T     impo       ce          malf ction             failur     of th    item l
o pla      safe      or r labili              .
V
                            . Th    desi     and f         ricati         compl    ity or     niquen s of e ite   .
                            . The    eed f       spec              contr   s and s        eilla  e av ocess      d equ         ent.
The egree           which f ctiona complia ce can                     e d onstra d by ins etion                        test.
l l                              The      lity his        ry and          egree of standa        izat' n of l                     q__      the    tem.                                                                       a The QA program requires that safety-related activities be carried out under suitably controlled conditions, including use I                of approved procedures and appropriate equipment, proper environmental conditions, and assurance that all prerequisites have been satisfied. Quality policies, QA manuals (QAMs), and                                          I procedures are mandatory requirements, which must be implemented and enforced. These safety-related activities and other activities designated by management as outlined in the HNP QAM include but are not limited to:
Administrative controls and reporting requirements.
Health physics and radiation protection.
Radiological environmental monitoring and the Environmental Protection Plan.
Plant chemistry.
Radioactive waste control.
Reactor and plant operations.
i Materials control (procurement, storage, and issue).                                      i QC (NDE,-inspection, etc.).
Surveillance program.                                                                     J Measuring and test' equipment calibration and control.                               l i                        -
Balance-of-plant instrument calibration.
Security Plan and procedures.
Emergency Plan and procedures.
Design changes and' plant modification control.
Departmental training programs.
i Maintenance programs.
Fuel handling and special nuclear material control.
Inservice inspection program.
l Procedure control and review.
i i      :
REV 2 7/84 REV 3 7/85    REV 7      7/89 L
N                                                                         REV 4 7/86    REV 10C 7/92 REV 5 7/87    REV 14D 7/96
                                                                  .17.2-9        REV 6 7/88    REV 15B 7/97 j <k

IC"-      O "NF-2-F5ax-17 TALLE 17 . 2 2 (C"EZT 1 0" 0)   -
             -   Li5- GF 5AirAi-RELAAEL 51AUCTURCO, SYSTE"S, ANO COMF NENTS I. Rea    or System Reacto       vessel Reactor essel support skirt Reactor v sel appurtenances, pressure              taining portions control ro drive (CRD) housing suppor s Reactor inte al structures, engineer d safety features (ESFs)
Control rods CRDs Core supply struct re Power range detecto hardware Fuel assemblies-II. Nuclear Boiler System Vessels, level instrumen          ion condensing chambers Vessels, air accumulato         ,   r supply check valves, piping downstream of air su ly c ck valve
                      -Piping, relief valve ischarge Piping, main steam,       ithin oute ost isolation valve Pipe supports, mai steam Piping,. other wit n outermost is         ation valves Piping, instrume ation beyond out           ost isolation valves Relief valves-Valves, main s eam isolation valves           SIV)
Valves, other isolation valves and w hin Valves,-inst umentation beyond outermos isolation valves
                      -Mechanical odules, instrumentation, wit safety function Electric 1 modules with safety function
(.                    Cable,       ith safety function III.-   Recir ulation System Pi ng P' ing suspension, recirculation line 1pe restraints, recirculation line
.(~  .
umps (structural integrity post-design basis accide t
                          -(DBA))
Valves Motors, pump (structural integrity post-DBA)
Electrical modules with safety function Cable,.with safety function
~(
REV 9C    7/51-
                                                                           - RE" 13A  10/94 -
l

                  .. .             ..      ~   . . - .   . . , . . - .
                                                                          . -   - - . .        . - . _ - . . . - - - . . - ~ _ -                    ..    .      .-
                                     ,                                        UNF-2-E5AR-1/
TAOLC 17.2--2 (SHEET 2 Or 0) -
IV. CRD Hydraulic System-lves, isolation,-water return line Va ves, scram discharge volume lines L
Val es, insert'and withdraw lines j                                            =Pipi         ,. water return line-within isolat                                    n valves l
l                     Pipin             scram discharge volume lines
                                             -Piping, insert and withdraw lines Hydrauli control unit                          .
1 Electrica modules, with safety f ction I
Cable, with safety function
* V. Engineered Sa ety Features                                                                                             ,
A.        Residual Hea          Removal (           R) System                                                    E i
Heat exchanger              pri ary side Heat exchangers, se ondary side                                                                              ,
Piping, within o                rmost containment                        isolation         valves Piping, beyond o t rmost containment                                     isolation         valves l
Containment spr y l'ne piping within                                     isolation         valve Containment s ay li e piping beyond                                      isolation          valve            i Pumps Pump motors                                                                           . .
Valves, is lation, low-                   essure coolant injection line l                                                       Valves,.          olation, other Valves,          eyond isolation v lves
                                                       ' Mechan' cal. modules Elect ical. modules, with saf                              y function Cabl         , with safety function 1
i i                    -l    .
j l-                                                                                                                                                                    l 3
l,
'                                                                                                                                                                    )
LU   -                                                                                                                                                          1 1
i I'L '
s
                                                                                                                                                            -\
                                                                                                                                     - _    w     a m                                                                                                                                                      j a

lHC-2-25AR-li TAOLC 17.2-2 (5HEEi 5 vE 9)
B. Core. Spray                                               l Piping, within outermost isolation va es Piping, beyond outermost isolation v lves, Pumps Pump motors Valves, containment isolation a       within
    -l'                     lves,: beyond outermost _ conta   ent isolation valves E   ctrical modules, with' safe y function Ca e, with safety. function C. .High- ressure Coolant Inj ction                           l Piping,    ithin outermo     containment isolation valve Piping,     yond outerm t containment isolation valve, other t n return est line, turbine gland-seal line and rain po line Pumps Pump 1 turbines Valves,'within o termost-isolation valve
                        ' Valves, beyond- solation valves, motor operated Valves,-other Electrical m ule        with safety function Electrical      xilia    equipment Cable, wit safety        nction VI. Safe Shutdow    Systems A. Reacto   Core Isolation     oling (RCIC) System           3 Pipi g,  within outermost c 'tainment isolation valves
            .             Pip'ng, beyond outermost co ainment isolation valves Pu ps
: j.                 V lves, containment isolation nd within
       *L                  alves, other RCIC turbine Electrical modules, with safety       nction Cable, with safety function L
+

          .        w
                       .                                     M"E 2-ESAR-17 TABLE 17.z-4 (bMEET Ja OF-9)
B. Standby Liquid Controi System.(SLCS)
SLC tank                                                                      1 Pump       .
ump motor.
V Ives, explosive                                                            .]
Va es, isolation and withi                                                     i I
Val s, beyond isolation v ves                                                 i Pipin ,'within isolation alves
                                      . Piping, beyond isolation valves Electri 1. modules, wi                         safety function Cable, wi h safety fu tion
                         -VII.   ' Fuel ~ Handling a d Storage.
A. Reactor Servi                ipment Fuel preparati            machine General'purp e.             apple (common to both> units)
Control rod rappl (common to both units) l
: Dryer and         parator sling and reactor pressure vessel                  l
                                         .(RPV) s ongback ( ommon to.both units)                                      !
B. Refueli 'g : Equipment 1
Refue ing equipment asse                        ly platform                   l Brid e crane
                                                                                                                     ~
i
            - ~
    ..                                                     .                                        \
    .: ,n
:'N..
4 b

_,    gMN-2[?AE-17 d;                                            TART.R 17.2-2 (RMERT'4 OF 9)_--
C.   -St'orage Equipment 1
                                                 . Fuel storage racks.
c                                 Defective fuel.' storage l container
           ,                                     ' Defective fuel' storage-rack r
                              ' VII        -Radioactive ~ Waste Systems A.   . Liquid Radwaste' System-gc i'                                      Piping, containment penetratio Valves, containment'isolatio
                                           'B.         actor Water. Cleanup Syst Pi ng,'within reactor:c o.lant pressure boundary.
Val s,' reactor coolan pressure boundary isolation val es and within IX'. Fuel Pool'C oling and C ennup System Valves.and pi            ng, m   eup y,                             .X. Water Systems
       .t.
                                           'A. RHR Service' a er System         .,
Cross-con         et-pi ng to RHR system, within second
                                                      -automa c' isola on valve Piping,       therowith afety. function Pumps?
                                                  ' Pump: .otors-
                                                  .Valv s, isolation
              ,                                   . Val es,-other E    ctrical. modules,' with afety function ble,- with safety func,tio
       ;p;.t -.
        .                                    B. Reactor Building Closed Cool            g Water System
           ,                                       Piping.and valves forming'part            f containment boundary C. Plant Serv' ice Water Syst'm   e
    < gt. . .>                                     Piping and valves, with safety.func on 4                                           Pumps       .t.
  ,                             ,                  Pump motors             .
Heat exchangers,1with safety function Electrical, modules, with safety      '
(unction Cable,.with safety function
       ..                                          .         .-'                    '  \
3(.                                  .D.   ' Torus Drainage ~and Purification System Pipi           and      l
  ,9                  ,
                                                , yalves.within Containment Boundary                                1
                                                  - '                                                 ~h
                                                                                                      --   5 \,
                              -c                                                                                      )

l l
                                       -HNP-9-ESAR- 1/                                l
      .                                                                               1
                             - TP.SLC 17.2-2 '3HEET 5 OF 0)    -
l l
l XI. Diesel Generator Systems                                        l l Day tanks Pumps, fuel oil system Pump motors, fuel oil system                                      l 1
iesel generators                                                 i E ectrical modules, with safety fun tions Ca  es, with safety functions Die 1 fuel storage tanks                                          l Diese lube oil system Diesel starting air system Piping    d valves from receiv r to diesel receivers XII. Heating, V    tilation, and   ir-Conditioning (HVAC) Systems    l l A. Control     ilding                                            i
: 1. Control room    VAC Common f r U    ts 1 and 2. See Unit 1 list.
,                       2. Battery ro     HVAC Fans Fan mot rs Damper Ductw rk
: 3. HV     equipment room C     on for Units 1 a     2. See Unit 1 list.
B. Rea   or Building 1  Reactor building HVAC I
          ~
Piping connected to standby as treatment' system (SGTS)
Piping penetration (emergency ore cooling system (ECCS) pump rooms)
Isolation valves
: 2. ECCS pump rooms Motors Fans Coils, cooling Ductwork RE" 4  7/A6 ,
                                                                             \

E u.u n _ o _ r_ e.n
: o. _ ,_ n. __
                                  -- IABLL li.z-Z (SHEET 6 OF 9)
C. Fuel Handling Area
                                                                                               ~
: 1. Spent-fuel pool exhaust system Valves / dampers,, isolation Ductwork Reactor Containment (Drywell)
: 1. Containment cooling system i
Ductwork Dampers
: 2. ombustible gas contr 1 system Mo ors                                                      "
Fan
                                  .Adso   er/scrubbe             units-Hydro n recomb'ners Piping, contaiiment penetration
                                 ; Valves,    ont nment isolation Containme t/ rywell monitoring system
                            '3.  'SGTS Motors
_ Fans Pref   ters.                                                 I De   sters-H   h-efficiency pa ticulate air filters                     )
sorber units uctwork Dampers I
Piping Valves E    Diesel Generator Building HVAC
      .                       Motors I                           Fans Dampers                                                          ;
II. Main Steam and Power Conversion Systems                            l
   ,;                    A. Main Steam System.                                               l t-Main steam piping to turbine stop valves a d branch
                             -line piping (> 2 in. ) up to and including- f rst valve REV '   7/AA    ;
                                                                                     \
s

c.
                                           -- IIIC ES AR- 17
_ TABLE 17.z-z (SHEET 7 or v)
                      'B. Condensate and Feedwater System Reactor feedwater piping and valves, RPV             outermost isolation valve j
XI  Instrumentation and Control Systems
                         . Reactor Instrumentation
: 1. Reactor protection system (R S)
All portions that must op ate to control and safely shutdown the reac or to a hot shutdown condition
: 2. N utron monitoring s' tem Pip g, traversing incore probe (TIP)
Valve    ,  isolatio    , TIP subsystem Electr' cal modu s, intermediate range monitor (IRM) and av age power range monitor (APRM)
Cable, I         an APRM, with safety function
(                       3. Nonnuclear        nstrumentation All port ns hat input to the RPS All por ons t at input to the ESFs actuation syst B. ESFs Act ation System All po tions C. ESFs Systems (Controls and           nstrumentation Required for Safety Associated with            ch Actuated System)
E CS ontainment isolation system Containment purge systemss Emergency diesel generator system Main steam line break detection s           tem D. Controls and Instrumentation Associa ed with Safe
(
Shutdowd Systems                '
             ,             Rod drive control system Fluid system contacts for safe shutdown control room habitability system Combustible gas monitors system
. I,f Drywell, Pneumatic System                                       l REV =  7/00 -
REV S  7/o L
                                                             ,                        \

      -                    ..          - _ ~ .       - . - -                   . .       .         . -. ..             - -.-.-.- .                      .. - ~         - - - ..                                    .-.-
                        ,            1
               <            r
          ~
u.u.n _ . e. .e .c.s.n._.s n.
                                                                                    ' TAOLE 17.2-2 (ancsA_.6 OF 9)                                    ~
J
_                                  E.       Instrumentation Associated with Other Sys ems
              ;[                                                -Required for Safety.                                                                                                                                      ,
: Spent-fuel pool cooling system.                                                                                                                          '
-                                                               ' Fuel 1 handling area ventilation isola ion system Control 1 room panels; .
                   ,.                                           ' Turbine building flood detection                                              struments
              'y                                                                                                                                                                                                           .
F.
[
ocess-Radiation Monitoring 5y tem                                                                                                                ;
XV. 'Electri al Systems n
A~ . 'ESFs                c Equipment Essenti                   4.16-kV buses Essential 600-V load ce                                           era                                                                                    a Essential 00-and 600/                                       8-V control centers-                                                                          l l
B.      ESFs'dc Equi                      ent l
i 125-V and 250-V 1
tion and diesel generator batteries                                                                         i and racks, bat                    ry charges                                                                                                      j
                                                                .125 - 250-V swi                               ear and distribution panels                                                                                q C,     ' Electric Cabl s.for                               SFs-Equipment                                                                                            !
~
5-kV power ables 600-V pow                    ' cables Control                   d instrumentati n cables                                                  ,
                                           .            D.     :Miscel aneous Electrical Reac or containment building                                               ectrical penetration semblies
              .- '                                               Co duit supports, safety relate
^ Mk                                                                    ay supports, underground ducts, fittings and                                                                                                       i encasement, safetyLrelated Emergency lighting syst, ems ,
Emergency 1 communications systems
                                                             ' Diesel ^ generator
* c
                                                               ' Remote shutdown system o_ g.#                                                                                                                g XVI      AuxiliaryfSystems-                                                 g                                                                                                i A.. ' Compressed Air Systems i,c,-                                               Accumulators for safety-related equipment
               ~'                                               Piping and valves, with safety function l
8
                                                                                       ,.s                                                                       RF.V 4
                                                                                       ~
_ """ 0           %80 7 f ^0-
    ^
                                                                                   -     ~
                                                                                                                                                                                            \                           .
                                           , W-                   ,,- ..,. ,         w      a    ,           , ,         . , , , .     -&   ,    - ,,                           .n,.,,,-.     . . , , . , . - , . .

,        , ~ - .               .      -      -      .        . _ . . . . - - - .            _    ~ . . -       . - _ _
MMP-2-FS.^R--17            -
l
                                                                                                                                'l TABLE 17.2-2 (6nEEI 9 OF 9)
B.
                                                ~
Nitrogen Inerting System Storage tank Makeup' supply piping and valves
                                 ' Ambient vaporizer
           .                      Piping and valves forming part o                     containment C. S .pling Systems ~
Pip gg and valves on ASME II                     - 1 Systems Pipi         and valves on ASME                I - 2 Systems                                    ;
Pipin and valves on ASME II - 3 Systems Piping nd valves, conta'nment penetration, isolation                                            )
Post lo -of-coolant ac ident radiation monitoring system D. Reactor Bui ding an           Containment Chilled Water System Containment i ola ion valves Buildings                                                                                      I XVII.
Primary containm           t l
Access watche flocks doors Liner plate               .
Penetration assemblies                                                                             I Coating Diesel' gene ator. building                common to both units)
Control b4 lding (common to both units)                                                               ,
Reactor         ilding                                                                               ]
Vacuu relief valves                                                                               i Acce s hatches / locks / doors Cra.e supports Int    e structure (common to both units)
XVIII. S ructures                                                                                     l Condensate storage tank foundation a d enclosure Diesel generator fuel tank vault Station battery rooms-
  ,                          . Spent-fuel pool and liner Unit vent stack (common'to both units)                       -
                                                                                                                                  )
Containment coatings-                                                                                j E
XIX. Miscellaneous Mechanical Whip ~ restraints and supports for Seismic Clas                             1 piping REV      7/96   _
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