ML20098H275

Forwards Mods to FSAR Sections 1.8,Tables 1.11-1,3.2,3.2.1, 9.1,9.2,9.3,11.2,17.2,17.2-1,260 Series Question Responses & Srai App Item 1 Discussed at Nrc/Pse&G 840718 Meeting. Info Will Be Included in FSAR Amend 8

ML20098H275
Person / Time
Site:	Hope Creek
Issue date:	09/10/1984
From:	Douglas R, Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8409120308
Download: ML20098H275 (60)

Text

{{#Wiki_filter:Pubhc Servce ~l-Electro and Gas Company 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation September 10, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen: HOPE CREEK GENERATING STATION DOCKET NO. 50-354 FSAR CHANGES RESULTING FROM NRC OUALITY ASSURANCE BRANCH OPEN ITEMS Attached is a copy of the complete set of modifications to FSAR Sections 1.8, Table 1.11-1, 3.2, Table 3.2.1, 9.1, 9.2, 9.3, 11.2, 17.2, Table 17.2-1, 260 series question responses and SRAI Appendix-Item (1). These items were discussed at the NRC/PSE&G Ouality Assurance Branch meeting of July 18, 1984, and via telecon between D. Wagner, NRC, J.

Spraul, NRC, and B.

Preston, PSE&G on August 17, 1984. Please note that portions of this material were previously submitted, via letter from R. L. Mitti, PSE&G, to A. Schwencer, NRC, dated August 3, 1984. This information will be included in Amendment 8 to the HCGS FSAR. Should you have any questions in this regard, please contact us. Very truly yours, Odt( p 8409120308 840910 PDRADOCK05000g A QO I I l Attachment The Energy People v.m nm e

A J Mr. Albert Schwencer-2- 9/10/84. C D. H. Wagner (w/ attach.) USNRC. Licensing Project Manager W. H.. Bateman (w/ attach.) USNRC Senior Resident Inspector J. Spraul (w/ attach.) USNRC Quality Assurance Branch N I '. .KT 1'01/02 4

f ._A_TTAC_IDENT, _ FSAR Section l'.8 FSAR Section 9.3 Page 1.8-11: Page 9.3-3

Page-l.8-36 Page 9.3-4 Page'l'.8-49

'Page 1.8-104; .Page l.8-ll5 FSAR Section 11.2 Table 1.11-1 Page 11.2-3 ..FSAR Section 3.2 FSAR Section 17.2 Page 3.2-3_ Page 17.2-1 Page 17.2-5

Tabler3.2-1

'Page 1 of 39 Pags 17.2-14 Table 3.2-l' Page 3 of 39 Page 17.2-19

Table _3.2-1 Page 4 of 39 Page 17.2-20

-Table 3.2-1 Page 5 of 39 Page-17.2-21 Table 3.2-1 Page 6 of 39 Page 17.2-29 Table 3.2-1 - Page 7 of 39 Page 17.2-32 Table 3.2-l' - Page'8 of 39 -Page 17.2-35 Table'3.2 =Page 10 of 39 ' Table 3.2-11 Page 11 of 39 . Table 17.2-1 Page 2 of 2 Table 3.2-1 Page 13 of 39 . Table 3.2-1 Page'14 of 39

Table.3.2-1

--Page 3 5 of 39 260 Series Questions / Responses Table 3.2-1 Page 24 of-39 > Table 3.2-1 ' Page 26 of 39 Page 260.15-1 Table 3.2 Page 28'of 39 Page 260.50-1 Table 3.2 Page 30 of 39 Page 260.60-1 Table.3.2-1 Page 31 of 39 ' Table 3.2-1 Page 32 of 39 -Table 3.2-1 ' Page 33 of 39 SRAI Appendix Table 3.2-1 Page'37 of 39 '(Table 3.2-lf Page 39 of 39 Page SRAI (1)-7 Page.SRAI (1)-10 Page SRAI (1)-11 FSAR Section 9.1 Page SRAI (1)-12 Page SRAI (1)-13 Page,9.1-27 1 FSAR Section 9.2 Page 9.2-32 'Page 9.2-34 Page 9.2-40 SPage 9.2-43 ^7

HCGS FSAR 6/84 s 1.8.1.23 Conformance to Reaulatory Guide 1.23 (Safety Guide 8), 1 Revision 0, February 17, 1972: Onsite Meteoro1ocical Procrams 1~ r ; '. h HCGS complies with Regulatory Guide 1.23. 1.8.1.24 Conformance to Reculatory Guide 1.24 (Safety Guide 24), Revision 0, March 23, 1972: Assumptions Used for Evaluatina the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storace Tank Failure Regulatory. Guide 1.24 is not applicable to HCGS. 1.8.1.25 Conformance to Regulatory Guide 1.25 (Safety Guide 25), Revision 0, March 23, 1972: Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handlina and Storace Facility for Boilino and Pressurized Water Reactors HCGS complies with Regulatory Guide 1.25. 1.8.1.26 Conformance to Reculatory Guide 1. 26, _ Revision 3, February 1976* Ouality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containino Components of Nuclear Power Plants HCGS complies with Regulatory Guide 1.26, with the clarifications outlined below. 'I PSE&G's position is'that equipment that is important to safety is i safety-related and therefore does not distinguish between these terms. PSE&G does recognize the need for the assurance of the i specified operation of certain non-safety-related structures, systems and components, such as fire protection systems, radioactive waste treatment, handling and storage systems, and l Seismic Category II/I items. Such assurance is documented l through the specification of limited quality assurance programs l (described-in Table 3.2-1, footnotes (22), (50) and (52). In in Table 3.2-1 will be neluded items designated ;q " ions

addition, t -jik4 M to ra d ein oper t

%o 4A eA f The exception to Position C.2. is that since the reactor l recirculation' pumps do-not perform any safety function and since failure of the reactor coolant pumps due to seal or cooling water failure does not have serious safety implications, the control rod drive (CRD) seal purge supply and reactor auxiliaries cooling 8 1.8-11 Amendment /

HCGS FSAR 1/84 1.8.1.63 Conformance to Reaulatory Guide 1.63, Revision 2, July 1978: Electric Penetration Assemblies in Containment Structures for Licht-Water-Cooled Nuclear Power Plants Although Regulatory Guide 1.63 is not applicable to HCGS, per its implementation section, HCGS complies with the design, qualification,. construction, installation, and testing requirements of IEEE 317-1976, as modified by Regulatory Guide 1.63, subject to the clarification in Section 8.1.4.12. l 1.8.1.64 Conformance to Regulatory Guide 1.64, Revision 2, June 1976: Quality Assurance Requirements for the Desian of Nuclear-Power Plants 'Ithough Dagniatnry c iA: .0 dos; net cpply te McCS, per ite e

..,---_m ucce --- 14me m4es <*

i+IC.W5hi$Ys5 lbn.3vf& qujL /, s y The ecchitect-cnginecr indicctce whaw Limic design serificctier - ,@th this - e_ m_. < _ _.m..-

e. n,.

.-_.2. _ _3 _ - +w,+ ____it.._.._._ mrm e m a,,.. e nomi s ...y.. 9 etendard 15 e5 niedified end inLec cetcd by "cvisien

• e e"ever, th;-erchita"*-anginaar A i d-cf Ecguleivuy Guid:

.54. u cc. ply with Revicion 2 in Lual iL allcwcd checking of the nne d ri;n cutput decurcnte by the criginater's rup--uieer See Section 17.2 for further discussion of quality assurance procedures and Section 1.8.2 for the NSSS assessment of this Regulatory Guide. 1.8.1.65 Conformance to Regulatory Guide 1.65, Revision 0, October 1973: Materials and Inspections for Reactor Vessel Closure Studs s l Regulatory Guide 1.65 is not applicable. i See Section 1.8.2 for the NSSS assessment of this Regulatory Guide. t-t ,1 4 6 Amendment / 1.8-36 --.z, --,.-,.-_.._.~,..._-....~,-._._._y ._m.,._. .,m_-

HCGS FSAR 8/84 ( l.3.1.87 -Conformance to Regulatory Guide 1.87, Revision 1, June 1975: Guidance for Construction of Class 1 Components in Elevated-Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1593, 1594, 1595, and 1596) Regulatory Guide 1.87 is not applicable to HCGS. 1.8.1.88 Conformance to Regulatory Guide 1.88, Revision 2, _ October 1976: Collection, Storace, and Maintenance of Nuclear Power Plant Quality Assurance Records During the operations phase, HCGS complies with ANSI N45.2.9-1974, as modified and interpreted by Regulatory Guideland_ },'gyg/ NUREG 0800 (Standard Review Plan), Revision 2, SecrThn 17. During the construction and startup phases compliance is subject to the changes listed below. M N M E,I J, ) The architect-engineer indicates that the original HCGS project commitment, via the Bechtel nuclear quality assurance manual (NOAM), was to ANSI N45.2.9 (Draft 11, Revision 0, January 17, 1973) rather than to ANSI N45.2.9-1974. The NOAM was revised to reference the 1974 document, as modified and interpreted by the guide, subject to the following specific changes: a. ANSI Section 2.1, Quality Assurance Record System - Add the following sentence at the end of this section: "The procedures shall include control of records required during completion of the work activity." b. ANSI Section 2.2.2, Nonpermanent Quality Assurance Records - Revise this section to read: " Nonpermanent records are those required to show evidence that an activity was performed in accordance with the applicable requirement but need not be retained for the life of the item and do not meet the criteria listed in Section 2.2.1." c. ANSI Section 3.2.2, Index - Revise this section to read: "The quality assurance records shall be listed in an index. The index shall include, as a minimum, record retention times and the location of the records within the record system. The index system used by organizations for the retention of quality assurance 1.8-49 Amendment 7

~ HCGS FSAR 8/84 II8.1.122ConformancetoReculatoryGuide1.122, Revision 1, February 1978: Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components ~ Although Regulatory Guide 1.122 is not applicable to HCGS, per its implementation section, HCGS complies with it. For furth'er discussion of seismic design, see Sections 3.7 and 3.10. 1.8.1.123-Conformance of Reculatory Guide 1.123, Revision 1, July 1977:-Ouality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants HCGS complies with Regulatory Guide 1.123 puring construction p 4 and startup phases, subject to clarifica ions stated below. item a clarifica ion pplies onlybu//60 During the operations hase,/s regMceswy toa icob vemens of

• rois c o d e i.a p u;/) k c H e e c s d ion gr%a k hs tubere

. nth s,hipmem.s applied Woodsprxpre The architect-engineer indicates that the_ original HCGS project ' commitment was to ANSI N45.2.13 (Draft October 1973) rather than to ANSI.N45.2.13-1976. The architect-engineer NOAM has been revised to reference the 1976 document, as modified by the Regulatory Guide, subject to the following specific changes: Regulatory Guide Section C.2 - This section requires a. the application of elements of the ASME B&PV Code, i l Section III,-Divisions 1 and 2, and Section XI; and l ANSI N45.2.13-1976; specifically, those elements not covered by the ASME B&PV Code for procurement of ASME B&PV Code items and services. The architect-engineer takes exception to the requirement, and has the following alternate position: The application of the ASME B&PV Code requirements above to the procurement of ASME B&PV Code items and services is adequate, based on the-fact that ASME B&PV i- _ Code represents the composite knowledge and experience of a large segment of the nuclear industry,.that the ASME B&PV Code is constantly being reevaluted for adequacy, that addenda are issued frequently, and that, to our knowledge, historical data do not exist that would indicate that the ASME B&PV Code quality l i-1.'8-104 Amendment 7

1 HCGS FSAR 8/84-

Positions C.1.1.2, C.2.1.2, C.3.1.2, and Table 1 of Regulatory Guide 1.143 require that all material specifications for pressure-retaining components within-the radioactive process boundary conform to ASME B&PV Code, Section II.

In addition, they require that piping materials conform to both the ASME and the identical ASTM specification, and they permit substitution of manufacturers' standards, instead of the ASME specification, in the case of pump materials. Although Regulatory Guide 1.143 does not explicitly address in-line process components, sight flow . glasses, Y-strainers, and steam traps procured by the architect-engineer, and the orifice plates and conductivity elements in the NSSS scope of supply do not'have certificates of compliance for l the materials specified. 2Also, the records of shop inspection, l required by Table 1, for the Y-strainers and the steam traps are l_ not available from the supplier. 1 Nevertheless, the quality assurance measures taken provide the reasonable assurance needed to protect the health and safety of the public and that of plant operating personnel. Position C.1.2.1 requires that the designated high-liquid-level conditions should actuate alarms both locally and in the control a room. For all tanks, a high-liquid-level condition actuates an alarm in the radwaste control room only. There are no local alarms since the tank rooms are controlled areas and normally -unmanned. Position C.4.3 requires that process lines should not be less than 3/4 inch (nominal). The crystallizer concentrates and l slurry waste transfer lines to the extruder / evaporators are 1/2 l inch nominal, in order to maintain acceptable flow velocities to i ~ prevent settling in the lines. The fluid flowrates are on the order of one (1) GPM as shown in Table 11.4-7 and on Figure 11.4-9. 1.8.1.144 Conformance to Reculatory Guide 1.144, Revision 1, _I September 1980: Auditino of Quality Assurance Procrams for Nuclear Power Plants I 5 S r,'n /d A"8 f^'** HCGS co li s with Regulatory Guide 1.144/ and comfrutun pk s <, f g,3 J,;,"0w/W s y b ' prin3 $ $styn 9 ,;5 The architect-engineer's quality program for safety-related items during the design and construction phases meets the requirements I of ANSI N45.2.12-1977 as' modified and interpreted by Regulatory + i I l 1.8-115 Amendment 7 I

. -. _.. ~ HCGS FS AR. 6/84 TABLE 1.11-1 (cont), Page g ot Summary FSAR Section (s) SRP Specific SRP , Description of Where _Sestign_ Acceptance Critgda Dif f erences Discussed 17.2.6 1 II-2A1 (R QA program to include com-QA program presently does mitment that the development, not include coussitment that control and use of computer the development, control, ard a programs be conducted use of computer code programs in ordance with the QA program be conducted in accordance and a ription of how the with the QA pecvjram. QA program I he applied. ~, II-3E4 17.2.6 Procedures be establishe' o 10 0 procedu have currer.tly assure that verified comput been es lished to assure codes be certified for use that rified computer codes and that their use be tified for use and that specified. eir use. be specified. \\ II-12.6 17.2.12 calibration of this Primary stan a used to equipment should be perform calibrat are at against standards t ave least q'reater thar. t ccu-I an accuracy of at ast four racy of the devices bein times the requ accuracy calibrated. N of the equ nt being cali-brated , when this is not [ I pos e, have an accuracy t assures the equipment being calibrated will be within required tolerance and that the basis of accep-tance is documented and authorized by responsible manac n..t. t .o

fT. ,t. - f' t e f HCGS FSAR Permit shutdown of the reactor and saintain it in the -b. safe shutdown condition Co'ntain radioactive material. c. A tabulation of quality group classification for each component is shown in Table 3.2-1. Interfaces between components and L' piping of different classifications are indicated on the system piping and instrumentation diagrams (P& ids), which are found in For information on instrument l pertinent sections of the FSAR. A l and electrical equipment classification, see Section 3.10.. l cross reference of system to FSAR figure number is provided in. The code requirements applicable to each quality l Table 1.7-2. group classification are identified in Tables 3.2-2 and 3.2-3. l Quality group classifications have been maintained during design e and construction and are actively maintained during plant operations and system modifications. commensurate with the safety functions performed by the safety-related components, except u where later requirements allow alternative $uality group '5 P ' M " g ;3 (5 'WW A

• M,c* ** d L

classJfications. Tab 'A' t e4m4= Ce 4 ys.p d r4,ma*fs vy Wyd c a.f4 9e r/4s of.Ns s s com/ Coissayauw 4,mgen ter (.rpagp e gep.mp, 4 p.vidad 4 Tuts, *H /. The plant design complies with Regulatory Guide 1.26, with l clarifications as discussed in section 1.8. PLAc6 2 6fAr 'AWAK tions of the radwaste system meet the F roup D (au efined in requirements the radwaste system meeting Regulatory Guide 1.143. may be determined the requir a ity group D a on the appropriate figures in Chapter l 3.2.2.1 SRP Rule Review -In SRP.Section 3.2.2, Subsection II, reference is made to Regulatory Guide 1.26 for determining quality group l classifications of components that are important to safety. !: / Section A and B of this guide imply that all components under the If l, quality groups shown are safety-related, including those listed o under Quality Group D. L. s 6 4 On IKTeS, the Quality" Group D itesis are not considered to " safety-related" or All l these terms are used in other guides and regulations. t 3.2-3

O cy } s.. (~ Ze/Sinf foA f>4CS 32-3 f rR TrosJ S of 1/2 BAD / > 4c rivs wn gr4 nsynGEss~r WM 5cH Ass wirs,w' 7//g gaussssrigS Systgm C 2 Li s gsr ifD By' THE Diffe-r/o" of 7~M4 A y/AGJ Si'o s k Ik 7~/// F/ G un.s of nao ct/A, orsAs'// Tx cl ubwd pefwG VA L v6 s, vsssst 5, r wks, ass se v,pm ts -r, }xs a n.,,,, o a ~,,~~, aw,,, f,4p j,4 gpan m k Is Cw Jg wj a ssvv - u-6/oe/ A. VS r' '?A" ' ! Ai c ,~,7 ::E ^t : ;', GE& A o9-Aq As 6HAroAy Guio e /. /y 3. pr)4 /w7~ 6 asJg wiYk A ECcs& reA, DESi6W com)>lrpg si rd C Lase /Fic.A russ, AS fi'Sc' is ono /- /4 3 IM SSorted fe, T /. / 4 5, c: net -

e HCG 8 PSAR 4/04 l TABLE 3.2-1 Page 1 of 39 l i BCG8 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPOWENTS i Principal Quality Construc-Source 3roup tion OA FSAR of Loca-Classi-Codes and Seiseic Require-Section Supply tion fication Standards Category mente Commente cas can ts (si ces tra l I. Seector.Bretes 4.1 i a. Deactor veneel and head GE A A III-A('8 I Y ('8 l i b. Reactor vessel support skirt GE A NA III-A('8 I Y (98 l c. Reactor veneel appurtenances, GE A A III-A('8 I Y l j pressure retaining portions 4. CRD housing supporte GE A NA III-NF I Y 4 e. Reactor internal structures, GE A NA Mone I Y (888 engineered safety features j f. Reactor internal structures, GE A NA None NA N <sa3 fy other, i g.- Control rode GE A NA None I Y b. Control rod drives GE A NA III-A('8 I Y l j 1. Core support structure GE A NA Mone I Y 1. Power range detector hardware GE A B III-2 I Y tsep 2 k. Fuel assemblies GE A NA None I Y 1. Reactor vessel stabiliser GE A NA III-NF I Y 1 j II. Beclear Boiler Svetem 5.1 1 a. Vessels, level instrumentation GE A A III-1 I Y } condensing chambers i A Vessels, air accumulatore P A,C C III-3 I Y c. Air supply check valves and P A,C C III-3 I Y piping downstream of air supply check valves { d. Piping, safety relief P A C III-3 I Y valve discharge ) e. Piping, main steam, within GE/F A,C A III-1 I Y i outboard isolation valves i f. Piping, feedwater, within P A,C A III-1 I Y j outboerd toolation valves t 9 Piping, main steam, between P C B III-2 I Y tass outboard and outermost 1 isolation valves l A Piping, feedwater, between P C B III-2 I Y tsap j outboard and outermost j isolation valves

y NCGS PS,s 4/8 l,. TABLE 3.2-1 (cont) Page 3 of 39 l l Principal Quality Construc-l Source Group tion OA i FSAR of Loca-Classi-Codes and Seismic pequire-Sectism supply tion fication standards category ments comments asa 8:3 cas esa ces tra j Principal Components IV. CFD Ihrdraulic Systen 4.6.1 D 631.10 NE a. Piping and valves, reactor P C y, y [TM building penetration b. Valves, scram discharge P/GE C B III-2 I Y aea volume lines c. Valves, insert and withdraw lines P/GE A,C B III-2 I Y testees d. Valves, other P/GE C D B31.1.0 NA N c. Pipe cap, water return line GE A A III-1 I Y f. Piping, scram discharge P C B III-2 I Y volume lines c. Piping, insert and withdraw lines P A,C B III-2 I Y h. Piping, other P C D B 31.1. 0 NA N taas 1. Hydraulic control unit including GE C Special (**3 I Y Esas scram accumulater 1 Electrical modules with GE C NA IEEE-279/323 I Y safety function cara k. Cable with safety function P C NA IEEE-279/323 NA Y esos 1. Pumps GE C D None NA N C. Pump motors GE C NA None NA N V. Eneineered Safety Features C. RNA syst 6.3/5.4.7 l 1. Seat exchangers, primary side GE C B III-C & I Y (shutdown cooling, suppression TENA CE*3 l pool cooling, steam condensing) l 2. Heat exchangers, GE C C VIII-1 I Y secondary side TEMA Ct*8 3. Piping, within outermost P C,A 4 III-1 I Y s e,a containment isolation valves (LPCI, shutdown cooling, l head spray) l Amendment 5 l

~ -. HCG 3 FSAR 4/39 l TABLE 3. 2-1 (cont) ,Page 4 of 39 l Principal Quality construc-i Source Group tion QA FSAR of Loca-Classi-Codes and Seismic poquire-Section Supply tion fication Standards Category ments commsets (t) (33 (33 tol (63 ty) 1 Principal C y aata t 4. Piping, beyond outermost P c B III-2 I Y ( s e s -- containment isolation valves l (LPCI, shutdown cooling, suppression pool cooling. -{ head spary, containment i spray, steam condensing) 5. Piping and spray nossles, P A B. III-2 I Y containment spray lines I within outermost isolation valves i 6. Deleted i 7. Pumps (LPCI, shutdown cooling, GE C B P&V-II488 I Y l suppression pool cooling, j head spray, containment spray) { S. Pump motors GE C MA NENA NG-1 I Y 9. Valves, inboard isolation, LPCI GE A A III-1 I Y ,8

• e a line & shutdown return line 4

10. Valves, isolation and within P C.A A III-1 I Y sseatee> j (shutdown suction, head spray) l 11. valves, beyond isolation valves P c B III-2 I Y ssessesa l I (LPCI, shutdown cooling, i suppression pool cooling, l head spray, containment l j spray, steam condensing) l 1 12. slechanical modules with safety GE C NA None I Y function saen l' 13. Electrical modules with GE c NA IEEE-279/323 I Y j safety function tara i 14. Cable with safety function P C NA IEEE-279/323 NA Y t888 15. ECCS jockey pumps P c 8 III-2 I 'Y I g3/. 0 Eff [fa) i 16. Piping and valves, reactor P c WO l J building penetration and i isolation ~ 17. ECCS jockey pump motors P c NA IEEE-323/344 I Y b. Core opsay system: 6.3 1. Piping, within outermost P A,C A III-1 I Y e3 isolation valves 2. Piping, beyond outermost P C a III-2 I Y 8888 i Amendment s l

I e/ed l scos rsAN 1 YABLE 3.2-1 (cont) Foge 5 of 39 l 4 Principal !l Quality Construc-source Group tion QA 1 FSAN of Loca-Classi-Codes and seismi:: Nequire-1 section supply tion fication standards Category mente Comments tas tan cas ces ces tra j Principal components i isolation valves 3. Pumpe GE C B PSV-II(*s I Y 4. Peng motore GE C NA NEMA NG-It's I Y. (**> 5. valves, inboard isolation GE A A III-1 I Y.

• • • 8 6.

Delves, outboard isolation P C A III-1 I Y (*es l and within L Dalves, beyond outermost P C B III-2 I Y asea j oontainment teolation valvee Y S. Electrical modules with GE A,C NA IEEE-279/323 I I sefoty function aees 9. Cable with safety function P A NA IEEE-279/323 NA Y tses 10. ECCS jockey peep F c 5 III-2 I Y 11. ECCS jockey pump motore P c NA IEEE-323/344 I Y l e. Eigh pressere coolant injection 6.3 l (EPCI) systems 1. Piping, within outermost P A,C A III-1 I Y taes l containment isolation valves l 2. Piping, test return line to P C B III-2 I Y I condenente storage tank up l to eeoond isolation valve l 3. Pompe Imain and booster) GE C B PSV-IIt's I Y j 4. EPCI turbine GE C MA VIII-1 I Y tss l l 5. RPCI barometric condenser GE C MA VIII-It's NA N 6. EPCI vacuum pump 6 GE C NA None NA N ) condensate pump l 7. Vacuum pump & condensate pump GE C NA None NA N motore 8. Piping, valve leakoff and P C B III-2 I Y cooling lines to barometric i i condenser l 9. Piping, other P c B III-2 I Y t ae at ess l 10. Valves. containment isolation P A,C A III-1 I Y t oe st ee s s I and within 11. valves, other P C B III-2 I Y t oe st ee s I 12. Electrical modules with GE C NA IEEE-279/323 I Y l safety fonction ters i 1 13. Electrical auxiliary equipment GE C MA Mone I Y l l Am.nd.ent 5 : l

O i acos rse s/s. .l TABLE 3.2-1 (cont) Page 6 of 39 l Principal Quality construc-l Source Groep tion QA FSAR of Loca-Classi-Codes and Seismi Require-section supply tion fication Standards Category ments Comments tsa cas tas toa ses tra Principal Components 14. Cable with safety function P A,C,0 NA IEEE-279/323 NA Y tass 15. ECCS jockey pump P C B III-2 I Y 16. ECCS jockey pump motor P C HA' IEEE-323/344 I Y d. Containment atmosphere control 6.2.5 system 1. Piping and valves, containment P A,C B III-2 I Y tesa l penetration and isolation 2. Containment /drywell monitoring P C,0 B III-2 I Y tses g (Me/Os analyser) 3. Piping and valves, reactor P C B III-2 I Y (*es l building penetrations and l isolation l s 4. Nitrogen system (containment l inerting) : a. Vessels P O D VIII-1 NA N b. Piping 5 valves, reactor P C,R 5.D Emme-[SO) l 4 ; ",/.O ygy jr 3-I 83/ building penetration s 1 solation c. Piping & valves, other P 0,C D B31.1.0 NA N d. Heat exchangers P R D VIII-1 NA N 5. Containment hydrogen recombiner l system: a. Motors P C NA NEMA NG-1 I Y l b. Blowers P C NA None I Y c. Reaction chambers P C B III-2 I Y and spray cooler d. Hydrogen recombiner heaters P C NA NEMA /IEEE-279/323 I Y l e. Deleted l + f. Deleted l g. Piping, containment penetration P A B III-2 I y h. Valves, containment is31ation P C B III-2 I I (*es 1. Piping and valves, other P c B III-2 I Y (*es l Amendmant 5 l

s a .j NCGS PSAR-4/04 { YABLE 3. 2-1 (cont) Page 7 of 39 { Principal Qua lity Construe-source Group tion 04 PSAR of Loca-Classi-Codes and Seismi:: Neguire-Section Supply tion fication Standards Category mente Commente ts) es> tan ces tes ces Principal Componente e. Primary containment leakage 6.2.6 rate testing systems 1. Piping and valves, P C B III-2 I Y teos containment penetration & isolation f. NSIV esaling system: 6.7 1. Valves, outermost toolation P C A III-1 I Y t**9 2. Valves, other, and piping P C B III-2 I Y t**8 3. Electrical modules with safety P C NA IEEE-279/323' I Y l functientevs l VI. Reactor _ core isolgtion cooline 5.4.6 13CIC1 avstems 1. Piping, within outermost P A,C A III-1 I Y esos containment loc 1ation valves 2. Piping, beyond outermost P C B III-2 I Y t ea containment isolation valves 3. Piping, test return to P C B III-2 I Y condenente storage tank up to second isolation valve 4. Piping, valve leakoff S cooling P c 8 III-2 I Y lines to barometric condenser 5. RCIC pump GE C B PSV-II(*3 I Y 6. RCIC barometric condenser GE C NA VIII-It*D NA N 7. RCIC cceedensate pump and GE 'C NA Mone NA N vecoum pump S. Condensate and vacuum pump GE C NA Mone NA N motore 9. Valves, containment isolation P A A III-1 I Y t *

• 88 *e s and within 10.

Valves, other P C B III-2 I Y t se at ess) { 11. RCIC turbine GE C NA VIII-1 I Y t a s 3 [Sy 12. Electrical modules with GE C NA IEEE-279/323 I Y safety function tara 13. Cable with safety function P c NA IEEE-279/323 NA Y tses Amesidsent S O

"O 1 1 j scS3 PSAR 4/34 g. l TABLE 3.2-1 (coatl Page 8 of 39 l l Principal Quality construc-source Group tion OA l FSAR of Imca-Clasel-Codes and seisal: Require-' i section supply tien fication standards Category mente commente j can cas-can ses ses . css primeipal Components I i 1 j 14. EcCa Jockey pump P C B III-2 I Y 1 15. acca jockey pump motor P C NA IEEE-323/344 I y j e i VII. Beactor water cleanup system

5. 4. 8 l

1. Vessels, f11ter/demineraliser GE C C III-3 NA N 2. Beat exchangers, nonregenerative, GE C C III-C/ testa RE*8 II/I (sea g reactor water side 1 3. meat enchangers, nonregenerative, GE c D WIII-1/ Testa R(* 8 II/I s ee s l f cooling water side a 4. Beat enchanger, regenerative GE C 'c C III-C/TE8th R('8 II/I toes g i 5. Piping, within outermost P/GE A, A III-1 I Y 4888 leolation valves 6. Piping, between outermost P c C III-3 I Y csea i feedwater isolation valve l and flow element 7. Piping, beyond outermost P/GE C C III-3 II/I eaeassea j isolation valves or beyond j flow element i 8. Pumps GE C C P6V-III(*3 II/I 4 ee 8 9. Pumpe, filter /demineraliser GE C C PSV-III(*3 NA N 10. Valves, isolation and within P Aec A III-1 I Y taestees 11. Valves, beyond isolation valves P/GE CL C III-3/PSV-III(*3 II/I ( se a l 12. Valves, filter /demineraliser P/GE C C III-3/PSV-III(*3 NA N l i 13. stochanical modules car a GE C MA Mone II/I . ( se s 14. Piping, reactor building P C.R -e D 444-+837.l.0 13$// (50 penetration eD = 7 8 31.l 0 p Qr~ g = U e) .15. volves, reactor tan 11 ding P c isolation 16. Cable with safety function P c NA IEEE-279/323 NA Y tses., 17. Tank, precoat, filter / GE C C API-650 NA N domineralizer 10. Electrical modules with safety GE C NA IEEE-279/323 I Y l 2unctionsae l Amendment 5 l 1 4

_ - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - _. _ _ - _ _ _ _ - _ _ _ - _ _ - - _ _ _ _ - _ _ ~ _. - _. i acs2 rs._ e/w i.;- TABLE 3.2-1 (cont) Page 18 of 39 l t I Principal Quality Construc-source Group tion OA i PfAR of Loca-Classi-Codes and Seismis moquire-i Section supply tion fication standards Category mente comments t es-tas can top tes ses Priacapal carpements I II. maal Peel Con 11am and Cleanus Svetem 9.1.3 l and saran _ meter Clemeno system: a. Wessels, fil E M t=raliser P R D VIII-1 NA N A Present taar P R D API-650 NA N l c. Beat estehengers P C C III-3 I Y C. Ptnel peel coolias pumpe P C C III-3 I y C. Valwee and piping, cooling loop F C C III-3 I T (**s i N s. I f. Walves, other P C,R D/ B31.1.0 NA N ') [C NA7 g [T (4Te) III-3 I C. Piples, makeup F C API-650 l A skimmer merge tanks P c B31.1.0 NA a l 1. Piping, other P c D[ D VIII-1 NA E l 1 Torus unter cleanup peop, P R l holding pumpo I k. Piping and velves, torme unter P C B III-2 I Y l cleanne containment penetration j and imelatten 1. asst eennectAma gemergency cooling) P-C B III-2 I y l m. Fuel peel osoling peep motors P C NA IEEE-323/344 I y ( 3 g 3 f, y,, gjy [gge} Valves o l p;p%3, reef,e f.;jj;,,3 is,faf;,n a ] fen,f,J;,, y, l n. E. andleactive leante Svatene j ) a. Ligeid radamete systems 11.2 l 1. Spent sosia storage tank P R MA 620 NA N 8888 APIM 65o 2. Tasks, atmospheric P R MR API NA n ase l 3. Beat enchangers P R A VIII-1/TEMA C MA 3 6asa g i 4. Piping P C,R J A B31.1.0 NA N 4 se et ses l S. Pumpe P R A B31.1.0/ NA p toeste*s Hyd.I 6. Valves P C,R M4 B31.1.0 NA y iseatses l 7. Vessele P R DP /. VI11-1 NA N 4ea3 8. teoste evaporator GB R /A VIII-1 NA y aasa 9. anschanical modules F/GE R ADC /E B 31.1. 0 NA R taan 10. Instrument and control b3ards GE R NA NEMA 12 NA N (ass i 11. Decontamination solution GE R f A. VIII-1 NA N 4asa l evaporator l c Amendemet S l l t

MC38 PSh.. 4/A [ TABLR 3.2-1 (cont) Page 11 of 39 Principat Quality construc-soeros Group tion QA PsaR of Inca-clasei-codes and sensate Require-secti:>a supply tion fication standards category mente commente cas ses can aos saa ses Priacapel consonante 12. wolves, flee control and P/es R d 4 a31.1.0 NA a esea filter erstem s.2/r a m(ro) 13. Piping End valves, reacter P c 91) m heildiste penetration and B 31. l.0 g locaation g h. Gameses redenote eyetens 11.3 bO NA N 4as8 l 1. Tank, atmospheric P R M, A API 3. Beat==ah==T**e P R JW A VIII-1/ NA 5 t es'8 TsNA c 3. Piping P R d4 331.1.0 NA 3 6se9 l tsea A B31.1.0 NA 5 4as3 l 4. Wolves, fles control P R 5. Delires, other P R 4 B31.1.0 NA g asea g 6. Bart filtere P R 8 VIII-1 NA 3 test g 7. Adeoeher unita P R a A VIII-1 MA N toes l S. chasseel geord bed P R pf A VIII-1 NA a toes e. Solid redneste systems 11.4 IA B31.1.0 NA 3 sees 1. Piptag P R 2. Valves P R .p{ g s31.1.0 NA a sees 3. Pompe P R .sw A. 331.1.0/ NA 3 seeN ses Nyd.1/(**3 g 4. Seeks, ateespberic P R [A API-650/D100/ g64 y sees g VIII-1 g 5. Dessele P R dA VIII-1 NA N 8as8 6. compressore P R M8 (**8 NA N toes l 7. Blasere P R M (**8 NA 3 Seet l g[I p m(So) g S. Piping and valves, reactor P C g) M g he1141mg penetration and 8 J/.I.0 isolation g d. Process and effluent radiological 11.5 mesitoring and sampling systemt 1. usin steam line 338 GE/F C NA IEEE-323/344 I Y

t s' Ncco P. er i j TABLE 3.2-1 (cotti , Page 13 of 39 l Principal Quality construc-Source Group tion QA FSAR of Loca-Classi-Codes and Seismic Nsquire-Section Supply tion fication Standards Category monts consents taa tas cas ses ses tes Principsi Components i 1 3. Espansion tanks P C C III-3 I Y 4. Beat anchangers P C C III-3/TEMA R I Y 5. Pumps P C C III-3 I Y 6. Pump motors P C C NEMA MG-1 I a Y 7. Hydropasumatic accumulators P G C III-3 I Y 8. Electrical modules with safety P C NA IEEE-279/323 I Y l function l c. Reactor aus111 aries cooling

9. 2. 8 system (RACS) 1.

Piping and valves forming P A,C B III-2 I Y tesa part of containment boundary 2. Piping and valves, reactor P C ED e S WI F MF#) I 0 3/./. O building penetration and isolation i 3. Piping and valves, other P A,C,R D B31.1.0 NA N (ses 4. Meat eschangers P C D VIII-1/ NA N TEMA C 5. Pumps P c D B31.1.0/ NA N taes Ryd.I 6. Espension tar.k P C D API-620 NA N d. Turbine auxiliaries cooling 9.2.2 system (TAcS): 1. FiFinq and valves P T D B31.1.0 NA N i 1 l l Amendment 5 l l l l l

Neon Psh, us. g TABLE.3.2-1 (cont) .(age 14 of 39 l' Principal Quality Construc-Source Group tion 04 PsAn of Im-Classi-Codes and seismi:: Require-section supply tion i1 cation standards Category mente Comments Eg) Et3 tal (93 to) 673 Principal components e e. Condeneste and refeeling water 9.2.6 storage and transfer system: 1. Tank, condensate storage P O D D100 NA 8 N l 464-+0 y/t N#) 2. Pipiaq and valves, reactor P c 1P-S 3 /*I. D building penetration and 1 solation P T D Nyd.I NA N 8848 3. Pumps 4. Piping and valves, MPCI, RCIC, P C B III-2 I Y (*es and core spray pump suction 5. Piping ant ealves, MFCI, DCIC, P 0,C C III-3 I Y (*e8 and can seturn line 6. Piping and valves, level P D,C C III-3 I Y 8**8 instrumentation 7. Piping and valves, dike P O C III-3 I Y 8.*e 8 penetrations S. Piping and valves, other P 0,T R D B31.1.0 NA N f. Turbine building chilled 9.2.7.1 water system P T D VIII-1 NA N 1. Tanks 2. Chillers P T D VIII-1 NA N P T C VIII-1/Ni.I NA N 48*8 d 4. Piping & valves, containment P A,C B III-2 I Y (*es 3. Pumps .P c ep -345-4 5 8/./.$ .g-kg seet h"#) penetration & isolation 9, j 5. Piping 6 valves, reactor building penetration 6 ( 6. Piping, other P T,A C.R D B31.1.0 NA N l isolation P T,A,C R D B31.1.0 NA N l S. Cooling coils P T.A,C NA ARI-410 NA N l 7. Valves, other P T NA NENA N3-1 NA N 9. stotors Amendment 5 l

NCGS PSAR 4/ Bq l 4 TABLE 3.2-1 (cont) Page 15 of 39 l 5 4 j Principal Quality Construc-Source Group tion M FSAR of Loca-Classi-Codes and Seismi:: Require-Sectima Supply tion fication Standards Category mente Comments l tan tas sas (as ces tra 2 } Principal Cv=ata j g. Aemiliary building control 9.2.7.2 area chilled teater systems j 1. Chillers P B,G C III-3 I Y l i l 2. Cooling coils P B,C,G C III-3/ARI-410 I Y l P B,G C III-3 I Y l 3. Pumpe i 4. Motore P ,B.G MA IEEE-323/344 I Y l } 5. Piping and valves P B.G,C C III-3 I y asestosa j 6. Yank, head P G C VIII-1 I Y ) b. Potable S sanitary water system: 9.2.4 t P 0,G,B,R D B31.1.0/ RA N taes g l 1. Pumps Hyd.I 1 2. Motore P 0,Geh,R NA MENA NG-1 NA N l 1 3. Piping and valves P 0,G.B R D B31.1.0/MSPC MA N l 1 1. Desineralised water makeup 9.2.3 storage & transfer system: l 1. Tanks P T D API-620 NA N 2. Pumpe P T D Nyd.I MA N taes J. Notors P T NA NEMA NG-1 NA N l i 4. Piping and valves, reactor P C G-D g *g 4-g[ T 4**+[S~0) building penetration 5 isolation 5. Piping and valves, other P All D B31.1.0 NA N i III. Standant Diesel Generator 49$ Auxiliary Syst ems j f a. Puel oil storage and transfer 9.5.4 systems 1. Storage tanks P G C III-3, N195 I Y l 2. Day tanks P G C III-3, N195 I Y 3. Piping and valves, P G C III-3, N195 I Y (*es + f fuel oil system I 4. Pumps, motor-driven fuel oil P G C III-3, N195 I Y transfer Amendment 5 l l 3

1 i i 30GS FSAR 4/94 l l YAaLE 3.2-1 (cont) Page 24 of 39 l Principal Quality Construc-Source Group tion QA FSAR of L>ca-Classi-Codes and Seismic pequire-Section supc'.y tion fication Standards Category monte Commente cas can (si ces ces crs Principal e v ts 6. Containosat a*=ampham P A,B,C NA IEEE-279 I Y l control system 7. unia steam isolation valve P C NA IEEE-279 I Y sealing system 8. Piltration, recirculation, P B,C NA IEEE-279 I Y and ventilation system 9. Reactor building ventilation P C NA IEEE-279 I Y isolation system j 10. unin control room habitability P B NA IEEE-279 I Y and isolation system 11. Essential saxiliary supporting P All NA IEEE-279 I Y systems for engineered safety features control c. Controle and instrumentation 7.4 associated with safe shutdown systems i 1. menetor core isolation GE C NA IEEE-279 I Y l cooling system (RCIC) 2. standby liquid control (SIC) GE C NA IEEE-279 I Y j system 3. RER, reactor shetdown GE C NA IEEE-279 I Y 4 l cocling mode 4. Remote shutdown systems P R NA IEEE-279 I Y l ) 5. Essential auxiliary supporting GE/P All NA IEEE-279 I Y j systems for 'the safe shutdown systems d. Safety-selated display ~ ~ - -7 J l instrumentation kTIAC Ml6 MI A i I t

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J _a a is I = TABLE 3.2-1 (coat 3 P age 26 of 39 -lE' Prinetpel Deality construc-Source Group tion QA FSAR of Ioca-classi-codes and Seise12 Require-sectisn sePply tion fication standards category-ments commente ass cas es> ces ses ers Principal coeronents 'l 1 8. Deleted l 9. Area radiation monitoring P All NA Wone NA N teaa g systems 10. Deleted l l 11. Reactor water cleanup sYeten GE C NA None NA N l 12. Radwaste systems P R NR Mnne NA N l = 13. Puel pool cooling - ' ""_T-- q-P c R None NA JF7 %) l s ICL",....>.e ep m3s % r c. aa a<- we w c. Control compler panels 1. Electrical modales with GE/F B NA IEEE-279/323 I T l safety function 88'3 2. cable with safety function P R NA IEEE-279/323 NA Y esos g h. Local panels and racks 1. Elec leal modules with GB/P All NA IEEE-279/313 I Y l safet function 2. cable th safety function P All NA 18E=.-279/323 Na T gass g IvI. Electric syst a. Engineered safet features (Class 15) ac equipment: 3.3 1. 4.16-kV switchgear P G NA IEEE-308/ I Y 323/344 2. 480-v unit substations P G NA IEEE-308/ I Y 323/344 3. 459-V motor control centers P 3,C,0,W NA 183E-308/ I Y 323/344 i I l L -/ i

i Neas PsAR 4est ) TABLE 3.2-1 (cont) Page 28 of 39 l i Principal 1 Quality construc-source Group tion QA 1 PSAR of Loca-Classi-Codes and Seismic Require-Section Supply tion fication Standards Category ments comments j tas cas cas ces ses ses. j Principal P ts i i I EVII. humillary Systems a. Compressed air (service and 9.3.1 8 instrument) systems: 1 1. Compressors P T MA Mone NA N 2. Pressure vessels, not for P All D VIII-1 NA N safety-related equipment ) 3. Piping and valves, containment P C,A B III-2 I Y (**8 i penetration and isolation i 4. Piping and valves, reactor P C MD W S F/7 W N(f#) l' building penetration and E ##*l O 1 isolation 5. Piping and valves, other P All D B31.1.0 NA N b. Primary containment instrument j gas system: 9.3.6 1 1. Compressors P C B III-2 I Y (*ss 4 2. Pilter housings, dryers, & P c B III-2 I Y coolers (air side) i 3. Coolers (water side) P c C III-3 I Y 4. Aeceiver tanks P C B III-2 I Y 5. Piping and valves, air with P C B III-2 I Y 8**8 { safety function j 6. Piping and valves, cooling water P C C III-3 I Y (*es 7. Piping and valves, air with P A C III-3 I Y (**8 safety function (inside drywell) 4. Piping and valves, containment P A,C B III-2 I Y (**8 penetration and isolation 9. Piping and valves, air, other P A,C D B31.1.0 N N I 10. Motors, compressors P C N/A IEEE-323/344 I Y i i t i Amendment 5 l 1 'l

McGS e 4, g,, That.E 3.2-1 (cont) Page 34 of 39 l~ Principal Quality construc-Source Group tion OA PSAR of Inca-Clasei-codes and Seismic Moquire-Section Supply tion fication standards Category mente Comments sas cas can ses top ses Principal Componente f. Aemiliary boiler syetas: 9.5.9 1. Sanke, bloedoen, chemical P O D VIII-1 NA N feed, feel oil storage and feel oil day 2. Boilere P O D VIII-1 NA N J. Demerator P O D VIII-1 NA N 4. Pumpe P O D VIII-1 NA N 5. Piping and valves, reactor P C C III-3 I T 4 *e a buildiaq penetration and isolation 6. Piping S valves, other P All D B31.1.0 NA N q. Egeipment ame floor drainage 9.3.3 systems 1. Piping, radioactive P C,A,T.R 0 .3i. i.0 NA N 2. Piping, moeradioactive P All D B31.1.0 na N J. Piping and valves, primary P A,C B ITI-2 I Y (*ep containment isolation boundary 4. Piping and valwee, reactor P C $M M 88/e/<0 .2I".[ M[Sd) building penetration and toolation b. Post-accident liquid and gas sample system (PASS) 9.3.2 1. Piping and valves, primary P A,C B III-2 I Y E*es containment toolation and reactor omolant pressere boundary 2. Sabing, reactor building F CR D B31.1.0 NA N l penetration and leolation 3. Piping and valves, other P C,R D B31.1.0 NA n 4. Piping station GE R D B31.1.0 MA N 5. Poet-accident sampler Gs a D m31.1.0 na N S

NCG8 PSI 4/sr l TABLE 3..?-1 (cont) Page 31 of 39 l 4 Principal Quality Construc-Source Group tion On FSAR of Loca-Classi-Codes and Seismi:: Require-Section Supply tion fication Standards Category ments consents i sas cas cas ses ses cas j Principal C P - to i 1. Breathing air: 9.5.10 1. Piping and valves, reactor P C )D ,ar" h) l building penetration 5 isolation i 2. Piping and valwes, containment P A,C B III-2 1, Y (*e3 j penetration and isolation 1 Lighting systems: 9.5.3 j - 1. Ca=p= mts located in P All NA stone II/I (sea safety-related areas ] IVIII. Buildines 3.8 ) j a. Primary containment 3.8.2 1 i 1. Access hatches / locks / doors P C B III-NC I Y 2. Vessel and head P C B III-NC I Y f 3. Penetration assemblies pipes P C B III-2 NC I Y l 1 4. Vent piping P C B III-hc I Y l 5. Vacuum relief valves 6.2.1 P A,C B III-2 I Y E**8 6. Nenorail supports P C NA AISC I Y { 7. Biological shield P A NA AISC/ACI-318 I Y l 8. Coating P A,C .NA Nor.e NA Y sasa f 9. BCCS section strainers P A B None I 'Y { i ) b. Adaillary building - diesel area P G NA AISC/ACI-318 I Y l c. Aus111ary building - control area P B NA AISC/ACI-318 I Y l l j d. Ausiliary building - radweste area P R NA AISC/ACI-318 I Y tasa l e. Turbine building P T NA AISC/ACI-31'8 II/I 4 a a,34 se 3 g f. Administration facility P O NA AISC/ACI-318 II/I 8888 8 4 g. Circulating water pumphouse P O NA AISC/ACI-318 NA N l j Amendsent 5 i

l NCGS PSAR 4/94 I YABLE 3.2-1 (cont) Page 32 of 39 l Principal Quality Construc-Source Group tion QA FSAR of Inca-Classi-Codes and Seismi: Nequire-i Section supply tion fication standards Category mente Comments i (in (33 (s3 <s ces ces Principal e ts i h. Reactor building / including P C NA AISC/ACI-318 I Y l pressure-retaining doors 1. Plant cancelled area P All NA AISC/ACI-318 I Y I

• (f#."A'j"L. 'F "hkk*,*,3 *.EV'ufp.,,%.%u#4;"6)

III. Structures 58 3.8 cees i a. Station service water intake P 0,W NA AISC/ACI-318 I Y l etracture i b. D41sted I c. Diesel generator fuel tank room P G NA Mone I Y j d. Station battery roome P B NA None I Y l e. Spent feel goal, reactor well, 9.1.1, P C NA Mone I Y new fuel vault, dryer 9.1.2 l separator pool, and cask pit i f. Deleted I f g. Unit vent stack, North & South P O NA ACI-307 I SY I h. Condensate storage tank dike P O NA ACI-318 I Y l i. Spent fuel pool liney 9.1.2 P C MA Mone NA N ---- N Missile / jet barriers (concrefe sfmt re) 9.1.1 Skisser surge tanks P C NA m &I-3/P ,Z 35 V3 (*** 1 { k. P A,8,C.G, MA AISC/ACI-318 I Y l R,W 1. structural tackfill P O MA I Y m. Post accident shielding P A,B,C.G, MA ACI-318 I Y R,T n. Beismic Category I slectrical P O NA ACI-318 I Y { duct bank manholes

ECG3 FSAR S/84 TABLE 3.2-1 (comt) Psge 33 of 39 Geneeral Electric Es3 GE = Public Service Electric and Gas Compsay/Bechtel Power Corporation P = (3'3 A = orywell auxiliary buildings control area B = reactor building C = auxiliary buildings standby diesel generator area G = cffsite locale L = outdoors onsite o = auxiliary building: radwsste area R = turbine building T = station servigp water intake structure. W = group classification as defined in Regulatory Guide 1.26. [#43 gfgd 0[- MC [d M, E/N E Ca3 A, B, C, D - NRC qual /(.D* - Quality Group $ l: _M.. Min" in *^M u*M Mide-tr19 3-foreWWus6p 7 --- postions orradmaste systemerDuring-the-operatione-phase. 3he_se_iteme-will-be rFoved dI4 [e VIa/O I l covered by-the-QA-Am-. / \\ fyuid6 /*N, uh - quality group classification not applicable to this equipment. (s3 Deleted to3 Notations for principal construction codes: m III-A, B, C - ASME Boiler and Pressure Vessel Code, Section III, Class A, B, or C III-1, 2, 3, Mc, MG, NF - ASME Boiler and Pressure Vessel Code, Section III, Class 1, 2, 3, or N or Subsections NG or W. P&V-I, II. & III - ASME Pump and Valve for Nuclear Power, Class I, II, & III VIII-1 ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 XI ASME Boiler and Pressure vessel Code, Section XI API-650 American Petroleum Institute, Wel$ed Steel Tanks for Oil Storage API-620 American Petroleum Institute, Recommended Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks C504 American Water Works Association, AWWA 504-70; Section 2 through 19 D100 American Water Works Association, AWWA-D100, Standard for Steel Tanks, Standpipes, Reservoirs and Elevated Tanks for Water Storage B. 31.7 ANSI B31.7, Mulcear Power Piping E31.1.0 ANSI B31.1.0, code for Pressure Piping N195 ANSI N195, Fuel Oil Systems for Standby Diesel Generators SMACHA Sheet Metal & Air Conditioning Contractors National Association, Inc j BEI Heat Exchange Institute i TEMA C&R Tubular Exchanger Manufacturers Assoc, Class C&R j BYD.I Bydraulic Institute AISC American Institute of Steel Cocatruction, Specification for Design Fabrication, and Erection of Structural Steel for Buildings l AISI American Iron and Steel Institute, Specification for the Design of Cold-Formed Steel Structural Members; Design of Light-Gauge Cold-Formed Stainless Steel Structural j i Amendment 7 I j l

i i c. t ucos n. I TABLE 3.2-1 (cont) Page 31 of 3g l i j sass The sendait, trays, and supports for safety-related cables are Seismic category I and Q-listed. t ases revs estety classification is at variance with ANSI N212, which has upgraded this system to Osality eroep s. or AMCA l sets auch publication 2114, auch certified satings Program for Air Performance, i standard 210, Test code 4 for Air flowing Devices, can be used for blower design purposes. it does not fait under l toss This emettom of steam piping is seismically analysed to ensure that loadings asses 11y associated with seismic category I. sans Impact testipq of carbon or low-alloy steels is la accordance with ASME BSPV code, Section VIId, Division 1, paragraph WC5 M. Imw temperature criteria tor carbon or low l alley steele le detinaa se -20*F or below. 12. teos guild to angE B38.2 and New Jersey Administrative code for overhead and Gantry cranes, Title sans The pesar cannwersion system structures are ::castracted in accordance with applicable codes ~ for steen power plante. paagenes, senseM Ms TNost UscohfD )

1. '~^

a_ ww s,wsms _ _- -- - r , are tm:11t to Quality asas dportgune of the edegnemmed^ a the unc Regulatory Guide 1.141. The equipment, piping, and 1 eroep 2 - _ _ _ ^. x_-- -r====*s are fabricated with a amadatory pressure test and welded construction wherever possible. l mogelatory emide 1.143 reduces the seismic design requirements t rom seasmic category I to a l simplified meismic analysis. For further information, refer to NRC Regulatory Guide 1.143. 1 per this project, the radunste area shares a building that includes control and diesel pote generater assas, and therefore is required to he seismic category I. taa s These Pete and associated supporting structures mest be designed to retain structural tmat do not have support and/or pressure integrity during and after a seismic category I event, to retain operability for protection of public saf ety. The basic requsrement is prevention of structural collapee and damage to equipment and structures that are sessaic category I. see s There le ao established standard for commercial peeps. ASME Section VIII, Division 1 and 4 while intended for other aust 331.1.0, power Piping, represent related, available standards that, PJ FM/ applications, are used for guidance and ree-Antions in determininq Quality Group 9 paamp. A## A allonable stressee, steel casting quality f actors, wall thicknesses, materials congetibility and specifications, temperature-pressure environment restrictions, t it t ings, flanges, gaskets, and helting, installation procedures, etc. sees Regulatory emide 1.54 app 1 Lee. i s**3 These devices are supported and analysed to remain functional up to an SSE. and identifiable device (s's & module is an assembly of interconnected components that constitut e or piece of equipment. For esample, electrical modules include sensora, power supplies, and signal processors; sechanical modules include turbines, strainers, and orifices. a===a==at S l

g TABLE 3.2-1 (cont) Page 39 of 39 l cows Doct work is of non-Seismic category I design, but is installed and supported as Seismic category I. cess valve operators on safety related valves that must function are Q-listed and Seismic category I (**3 Equipment is classified in accordance with the conformance statements made in Sections 7.2, 7.3, 7.4, 7.5 and 7.6 in reference to IEEE 279 paragraph 4.4 and IEEE-323. toss The QA Program controls applicable to equipment classified as Seismic II/I are in accordance with Regalatory Guide 1.29 commitments contained in FSAR Section 1.8 toss No QA Program controls applied during Design and construction Phase. QA Programs controls during operation are applied to an extent consistent with the items importance to safety. test On Program controls for the fire protection program, including emergency lighting and communications, are applied to the extent of the ten quality assurance criteria of Appendix A to Branch Technical Position 9.5-1 and to an extent consistent with the item's/ activity's importance to safety. cess The recirculation system piping was built to both ASME Section III and B31.7 codes as required l by the GE 4esign specification. The ASME Section III NPP-1 report requires signatures by a l qualified inspector and also indicates that the pipe was built to the requirements of B31.7. l (**3 Except north radweste area of auxiliary building, since there are no seismic category I I components in this area. I I Tla raadm pressua Jessa0 ederm0 Skubwun LoS tA &l Q&co 100 ak E ncI (SS)Elsi pro 3rawi, W(.lck is codetect hy tle operaficw# G A progreu. (S.) Ang HoJCcJiats n tp ulod 4. *le IM tutil la W W e ela opetafi d QA prmtans. (g9) ContunZeJ o'.srU*~ vbes fS'I*"- ffcf A.c'~ GDC ST sTo $,A *rt(c.vi.,rs.,o/.p se cs s. co Ape 4 <saf' e f YN-p..nery, eo., pon t.h .rl.. m, are Ss ne. p e.-k n (s s) vs. h +;e J 7.ns te v so9 p~fct ane eren!"y 4 a sfmtu<es win be c.a,L.de/ w & 4 yen /:*, QR pr p. (ss) m 3.v y n,y elses fr H Pcr as ecrc %- Le.5 pyt_,+ w.ps, ace t u = e. ers 3-.

l HCGS FSAR 10/83 / requirements. The teru: ::ter cle:nup sy t = piping t: :n.d fre: th; M C sy:te= filter-d:=ineralizer is designM te Sei:=ic c t;;;,ry I rquirc::nt: :t the r+aetcr building penetr: tion:. -d The design of the FPCC and torus water cleanup systems, with respect to the following areas, is discussed in the sections listed below: a. Seismic Category I requirements - Section 3.2 b. Proteccion from wind and tornado effects - Section 3.3 c. Flood design - Section 3.4 d. Missile protection - Section 3.5 Protection against dynamic effects associated with e. postulated rupture of piping - Section 3.6 f. ' Environmental design - Section 3.11. Class 1E power is provided for the safety-related equipment of the FPCC system. Al1 annunciators are provided with non-Class IE uninterruptible power. Class 1E power is provided for the O containment isolation valves in the suction and discharge Ihres for ta tor== t r ci aua =v=t The radiological evaluation of the FPCC system is provided in-Section 12.2. Radiation monitors mounted on the reactor building walls indicate and actuate audible alarms locally and in the MCR. A failure mode and effects analysis o'f the FPCC and torus water cleanup systems is provided in Table 9.1-3. ? The fuel pool evaporation rate, the time. for the pool water to reach 2120F, and the time required to initiate the makeup water, in the event of loss of the FPCCS, are discussed in Table 9.1-18. 9.1.3.4 Inspection and Testino Requirements The FPCC and torus water cleanup systems are preoperationally tested in accordance with the requirements of Chapter 14. The safety-related systems that provide makeup water are periodically tested in accordance with the requirements of Chapter 16. O 9.1-27 Amendment 2

v HCGS FSAR 4/84 suppression pool in the event of a low condensate storage tank level. Redundant low-low level switches have been provided"to allow for automatic switchover of the RCIC pump suction to the suppression pool in the event of a low condensate storage tank level. These. level switches and transmitters are seismically supported on a standpipe located inside the reactor building and are. electrically separated (powered from different Class IE power sources) as shown on Figure 9.2-13. The piping between the CST and the reactor building penetration is heat traced to prevent freezing in cold weather. The heat tracing is powered from a highly reliable battery backed non-1E power source. The heat tracing is provided with an alarm monitoring circuit powered from j l a non-1E battery-backed power supply separate'from the heat tracing power supply. This circuit monitors the heat tracing power supply and thermostat and alarms on loss of either. Heat tracing is not required for that portion of piping inside the . reactor building. 'See Sections 7.3.1.1.1.1 (HPCI) and 7.4.1.1.2 (RCIC) for discussions of the automatic switchover functions. 9.2.7 PLANT CHILLED WATER SYSTEMS i 9.2.7.1 Turbine Buildino Chilled Water System The turbine building chilled water system (CWS) provides chilled water for cooling the drywell, reactor building, turbine building, radwaste area, and service area. In addition, the CWS provides chilled water to the reactor recirculation pump motor air coolers, drywell equipment drain sump cooling coil, sample coolers, and mechanical vacuum pump seal coolers. 9.2.7.1.1 Design Bases l The design bases for the CWS are as follows: a. The CWS is not safety-related, except for the drywell

d ::::t:: buildia; :::1;;;. chilled water penetrations and isolation valves.

These are designed I to meet Seismic Category I requirements and are discussed in Section 6.2.4. i l f 9.2-32 Amendment 5 l

.~ [ ~ HCGS FSAR ~ cperate during' normal plant conditi>ons., The standby chiller is started manually if an operating ch' iller fails. The standby pump is started manually if an operating pump fails. l The glo' sed-loop refrigerant system in each water chiller extracts heat from the circulated water in the evaporator, and rejects the heat to TACS water in-the condenser. A temperature sensor in the chilled. water outlet regulates the refrigerant, flow in the chiller tc maintain a constant outlet water temperature. u Tworedundank'pipingloopsconnectedtotwosetsofcoolingcoils are provided inside the drywell. Changeover from one loop to the other is accomplished from the main control room by changing the position of loop isolation valves. A primary containment isolation signal automatically closes both sets of' isolation valves.. The valves can be reopened from the main control room, if desired, to resume' cooling. water flow. i The CWS is. automatically shut down in the event of LOP. RACS water is available for the drywell by automatically opened motor-operated diversion valves at the CWS/RACS interconnection. These g valves can be operated from the main control room. Chilled water J can be restarted manually when normal power is restored. Chilled water flow through various cooling coils and unit coolers is' automatically controlled by local temperature-actuated valves, cs shown oniFigure 9.2-14. Some air cooling coils have uninterrupted; water flow, and drywell cooling coil valves are controlled from the main control. room. During a refueling operation, chilled water is supplied to the third (standby) reactor building ventilation system (RBVS) cooling unit to increase the ventilation rate. Cooling water for the-drywell may be provided manually to both loops of the cooling L coils in the.drywell air coolers if extra space cooling is required. , l 9.2.7.1.4 , Safety Evaluation l V TheCWShasnosafety-relatedfunction,exceptfortheisolation valves at penetrations through the drywell.:nd th: ::::t;; trildin;. These valves are described in Sect 6cn 6.2.4. 2 1. 9.2-34 b w -wg,w-o .-4r-, ,ee..n,-,nn,-m.,,v nw w - e,,,.,,_+m., .m,_en,ew,-,aana_,- m,,_ _, - -___ _n_w.,_,m,, _ww--,,m-,,

s HCGS FSAR O. " ::t:r,':::ilimry buildipg penetratiene and t e_ f reacter building icelatier velver, rhich ::: derigned beth te Seirei C:t:;;ry I rcquirc:: t: AsuE peau ceg, gectier III, c13rr 3. =ad b. Remove the maximum anticipated heat loads developed by the components-served by the system over the full range of the normal plant operating conditions ar.d ambient ) temperature conditions-c.' Permit the use of corrosion inhibitors to prevent long-term corrosion and organic fouling of the system's piping d.. Serve as a barrier between potentially radioactive systems and the SSWS. 9.2.8.2 gystem Description The RACS consists of two 50%-capacity cooling water pumps with f' two 50%-capacity heat exchangers, one expansion tank, one chemical addition tank, two-100%-capacity booster pumps, and cssociated valves, piping, and controls, as shown on Figures 9.2-16 and 9.2-17. Major equipment design parameters are cummarized in Table 9.2-9. e The RACS system provides demineralized cooling water to nonessential equipment, located in the reactor building i enclosure, the radwaste area, and the'turbi6e building, that can carry radioactive fluids or that require a clean water supply to minimize long-term corrosion. The system is monitored continuously to detect any radioactive inleakage from the equipment being cooled. The service water in the RACS heat exchanger tube side is maintained at a higher pressure than the closeu loop system in the heat exchanger shell side. In the event of tube failure, the service water leaks into the closed l l loop system to preclude the possibility of radioactive release to the environment in the unlikely condition that the RACS cooling l loop becomes radioactive. i i l l 9.2-40

o. HCGS FSAK The expansion tank is connected to the suction side of the pumps and placed at the highest point in the system to accommodate thermal expansion and contraction of the cooling water due to temperature variation, and provides ample net positive suction head (NPSH) to the RACS pumps. The expansion tank nas a capacity of 640 gallons. It also provides necessary makeup water as

• required.

The RACS supply and makeup water are furnisned from~the demineralized water system. Sodium nitrite is periodically added to the system for corrosion prevention, from the 93-gallon chemical addition tank. l The RACS pumps, heat exchangers, chemical addition tank, and expansion tank are located in the reactor building. Valves and piping for the R CS are designed to ANSJ Power Piping Code B31.1, except for the primary containment m

t; /;;;ilicry buildin, penetrations and isolation valves.

Containment penetrations and isolation valves are designed to Seismic Category I requirements and ASME B&PV Code, Section III Class 2. "0 0ter/rerili ry bei,lding p:n:tr:ti:n, 2nd i 12tja

17:: ::: 2 :ign:d t: Sci: i: Cet:;;ry I requir::::ts-and ASM& -B6P8/ C;de, S :t10- !!!, Cl:M.

l Pumps, heat exchangers, and pressur,e vessels are designed to ASME B&PV Code, Section VIII, Division 1. The RACS heat exchangers are designed to TEMA Standard Class R and Heat Exchange Institute .(HEI) standards. The expansi.on tank is designed to the standards of ASME B&PV Code, Section VIII. 9.2.8.3 Safety Evaluation I l l The RACS has no safety-related function and is not required to be operable following a LOCA. Upon a LOCA signal, the RACS heat exchangers are automatically isolated from the balance of the l SSWS, and the RACS pumps are tripped. Each supply and return l header in the drywell has two containment isolation valves that l close automatically upon a LOCA signal. i ( 9.2-43

HCGS FSAR 8/83 Provisions are made to alternate the lead and standby compressor in order to equalize wear. The lead. compressor is in continuous operation and is automatically loaded or unloaded in response to the compressed air system demand. The second service air compressor serves as a standby. The standby compressor starts automatically at a predetermined low air pressure whether caused -by. failure of the lead compressor or an extra demand on the system. Cooling water for the intercooler and aftercooler is provided by the turbine auxiliaries cooling system (TACS). One of the two drying towers of each instedment air dryer package removes moisture from the air stream while the other drying tower of the instrument air dryer package is regenerated with dry air. Moisture is adsorbed, and the regenerated air is expelled to the atmosphere. The towers are alternated on a timed cycle. The emergency instrument air compressor and corresponding air dryer package, which are connected to the Class 1E bus system and maintained in. automatic operation mode, will start operating if the instrument air pressure drops'below 85 psig or both the service air compressors malfunction. In the event of an LOP, power will be manually restored from the SDG by the operator in the main control room. Loading and unloading sequence is regulated by the pressure at the emergency instrument air receiver. Low pressure in the instrument air header shuts a valve in the service air supply header in order to divert all air supplies to the instrument air system. All of the above compressed air system supply equipment is located in the turbine building. 9.3.1.3 Safety Evaluation l The instrument and service air systems have no safety-related function other than the integrity of the piping through the containment penetration. Failure of the systems will not compromise any safety-related system or component or prevent a safe shutdown of the plant. The service and instrument air lines penetrating to the reactor building have a motor-operated valve with a handswitch located in the main control room for containment isolation. Refer to Section 6.2.4 for details of containment isolation design features. Th: ::: vie: cir reector t;ilding p;;;tration : d i;;1stien valv;; ar; d;;ign;d t: 0:i;;i / 9.3-3 Amendment 1 ...-,_-____.,_.,,._- ~ _-.,.-- _ - __--___._-__..____ _ ___ __ _ ____ _ __

HCGS FEAR b-S [ ~ E e 9.3.1.4 Tests and Inspections ~ The containment penetration portions of the compressed air systems are preoperationally tested in accordance wth the require,ments of Chapter 14. The instrument air system is tested 4 in accordance with Regulatory Guide 1.68.3, Preoperational Testing of Instrument Air Systems. Compressors and dryers shall be tested in accordance with ASME and manufacturers' test procedures. 9.3.1.5 Instrumentation Application Instrumentationster m for each instrument air and service air compressor, train to monitor and automatically control each compressor's operation. The compressors are tripped on the following signals: low oil ) pressure, high oil temperature, -high cooling water discharge temperature, high air pressure in the receiver, high outlet air temperature, and high vibration. 'Most of these signals are annunciated in the main control room by common trouble alarms. High air temperature in the aftercooler and moisture separators, low pressure in the air receivers, and high intake filter differential pressure are also alarmed on a local control panel and the main control room by a common' trouble alarm. Instrumentation is also provided'loeally for each instrument mir dryer package train to monitor the packages operation. l Service air compressor and emergency instrument air compressor trouble are individually annunciated and alarmed on the local common service air compressor control panel. These alarms also indicate on the main control room computer, along with the air dryer trouble alarms. 9.3.2 PROCESS AND POST-ACCIDENT SAMPLING SYSTEMS The process sampling system (PSS) is designed to monitor and provide grab samples of both radioactive and nonradioactive fluids used in the normal operation of Hope Creek Generating Station (HCGS). 9.'3-4 ,,,,w- ,---e-pr,- ~ e-mn-v ,a y,,,--e-we. g w--,,c - - - ~, s r r-,e -m ew + e-----,-,,o- -- - mm w--e.m em--w--~~---

---

e-~

w

HCGS FSAR corresponding noble gas release rate of 500,000,Ci/s after 30 minutes decay (design basis). The concentration of radioactivity at the point of discharge shall not exceed concentration limits specified in 10 CFR 20, on an annual average basis. , w: m. g. All piping and equipment in the i e#no N o Category I with the exception of the primary containmenterecterbuildin;icelatiervalver,r}___the at cciated pip 4ng beteren the ieelatten valrec.f p The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the LWMS are discussed in Section 3.2. h. Design features that reduce maintenance, equipment downtime, liquid leakage, or gaseous releases of radioactive materials to the building atmosphere, to facilitate cleaning or otherwise improve radwaste operations, are discussed in Section 12.3. i. All atmospheric liquid radwaste tanks are provided with an overflow connection at least the size of the largest inlet connection. The overflow is connected below the tank vent and above the high level alarm setpoint. It is routed to the nearest" drainage system compatible with its purity and chemical content. Each liquid radwaste tank room'is designed to contain the maximum liquid inventory in case the tank ruptures. i l j. Processed wastes are collected in sample tanks prior to their rouse as condensate quality water or discharged in a controlled manner into the cooling tower blowdown line for dilution before entering the Delaware River. k. The expected and maximum radionuclide activity inventories for LWMS components containing significant. l amounts of radioactive liquids are shown in Tables 11.2-8 and 11.2-9. They are based upon the assumptions given in Table 11.2-1 and upon the following: i 11.2-3

I HCG'S FSAR 4/84 i 17.2 OUALITY ASSURANCE DURING THE OPERATIONS PHASE Public Service Electric and Gas Company (PSE&G) is responsible for assuring tnat the operation, maintenance, refueling and l modification of Hope Creek Generating Station (HCGS) is accomplished in a n.anner that protects public health and safety and that is in compliance with applicable regulatory requirements. To carry out this responsibility, PSE4G has developed and implemented a comprehensive quality assurance program that is applicable to the design, construction, and l testing phases. The description of the quality assurance program provided herein parallels the operational quality assurance program currently being implemented at the Salem Generating Station. l This operational quality assurance program is documented in the nuclear department manual. This description is maintained by nuclear operations quality assurance (NOA). The program provides measures to assure the control of activities affecting the safety-related function of structures, systems, and components. The quality assurance program encompasses fire protection of safety-related areas and other activities enumerated in Regulatory Guide 1.33. A planr.ed monitoring and audit program assures that specified requirements of the operational quality assurance program are met. The program provides coordinated and centralized quality assurance direction, control, and documentation, as required by the NRC criteria set forth in i 10 CFR.50, Appendix B. Applicable NRC Regulatory Guides, codes, kw.hs< d '*dia*#and standards, as well as the policy statements contained in the 9 PSEiC q;;1it; ;;;;r r.:: manual, are used by PSE&G organizations performing activities affecting safety to prepare appropriate implementing procedures. To assess the effectiveness of the PSE&G quality assurance program, independent auditors froe outside the company periodically audit the program for compliance with 10 CFR 50, Appendix B, and other regulatory commitments. Independent audits shall be conducted at least every two years. Reports of such audits are made directly to upper management. i QA policy statements are issued by key management representatives including the Company Board Chairman / President, by the Senior l Vice President - Energy Supply and Engineering and by the Vice President - Nuclear and, as such, are mandatory throughout the i Company. .i I .g. t 17.2-1 Amendment /jl I L i. .__...__.._.._-._-____..._____..__..,..._.__..-._,_._,__...._m ... ~

e. HdGS FSAR 4/84 The PSE&G policies and. organization structure assure that the manager - quality-quality assurance nuclear operations has sufficient organizational freedom and independence to carry out his responsibilities. 17.2.1.1.4.1 Nuclear Operations Quality Assurance Personnel o$ & Msd w,J%;n & qv>Jih usew se. Q s m w t% dis The manager - NOA and engineers reporti g directly to him must each i have a combination of 6 years of experience in the field of quality assurance and operations. At l east 1 of these 6 years of experience must be in the overall implementation of a nuclear power plant quality assurance program.1 A minimum of 1 year and a maximum of 4 of the 6 years of experience may be fulfilled by related technical or academic training. Personnel performino 4 inspections, examinations, and test activitiesFare certified as ^ Level I, Level II, Level III as appropriate to their responsibilities,alsoinaccordancewithRegulatoryGuide1.58,/ as noted. (f.. a.., do yc,v h m ; b a+c.a.) The manager - nuclear operations quality assurance fulfills the above qualifications with the addition of the following: i a. Knowledge and experience in quality assurance, l l b. High level of leadership with the ability to command L the respect and cooperation of company personnel, l vendors, and construction forces i c. Initiative and judgment to establish related policies to attain high achievements and economy of operations. 17.2.1.1.5 Independent Review Groups Three_ advisory. groups are responsible for reviewing and evaluating items related to nuclear safety. The overall responsibilities of these groups are included in the following sections. More detailed descriptions are contained in l ; Section 13.4. ,1 1 The SORC is an in station advisory group. Composed of key e l station personnel, its responsibilities include review of plant 5 Amendment / 17.2-5 l

w d 171s.,n & w < e- $WW WY* 0Y visvH i,,r=- M u esvf =/ & h ~ '" *~' & t" ' Personnel requiring certification are evaluated to establish their qualifications for their respective lavel and discipline. Recertification is based upon demonstrated continued proficiency or requalification, if necessary. Personnel requiring certification in accordance with Regulatory Guide 1.58 are limited to NOA personnel who perform inspection and test activities,emi members of the Operational Test Group (OTG) who perform post-design modification testing '^~~-' rhe ' " ^ ^ ee4e4reeMon personnel receive a periodic training needs assessment to identify additional supportive training needs as well as to evaluate individual post-training performance. The 1 assessment period is three years or less. Inspection and test activities not requiring personnel certification per Regulatory Guide 1.58 include Technical Specification surveillances and periodic inspection and test of fire protection equipment. These personnel are qualified &nd retrained in accordance with applicable requirements cf Regulatory Guide 1.8. Training programs of supporting organizations are described in ( their manuals, which are required to comply with the quality assurance program. t The Nuclear Training Center is responsible for the licensed operator training and retraining, in addition to other technical l and supervisory training programs, including General Employee l Indoctrination, which is required for all personnel having access to the station. 17.2.3 DESIGN CONTROL +_ The design control program includes activities such as field i design engineering, associated computer programs, compatibility f of materials, and accessibility for inservice inspection, maintenance, and repair. During the operations phase, issuance of new drawings and revisions to existing drawings require the implementation of a design change. The nuclear support division procedures, approved by the manager - nuclear operations OA, provide implementation guidance for the intent of Regulatory Guide 1.64 " Quality Assurance Requirements for the Design of Nuclear Power Plants." Within that division, the nuclear engineering section has the following responsibilities: L k& unLlpws& 1ncAu<lce O' sap us] N '"*' (~Th sy (f m4 L twe~ y' Q MM, in cA'd'H k I sk& c e ra d N : SM p urg 8*#c. ve-ene *M~*o #O *VN h rmu d tStO, p cvW h d /m u/ Wd 4w M. Amendmentjifj 17.2-14

i. The designaticn cf tho23 cctivitics rcquiring datoilcd preccdurca is und] by c gnizcnt d:partm3nt h;cds cnd ca o minimum, ccmplica with cppliccblo requirem2nto of R2gulotcry Guido 1.33. j M GS FSAR 4/84 ( ( r I c. Provide right of access for source surveillance and audit by NOA or its agents l d. Provide for required supplier documentation to be submitted to PSE&G or maintained by the supplier, as appropriate e. Provide for PSE&G review a 3 approval of critical proce8ures prior to fabrication, as appropriate. Procurement documents require suppliers and contractors of other than commercial grade items to provide services or components in accordance with a quality assurance program that complies with applicable parts of 10 CFR 50, Appendix B. The requirement for notifying PSE&G of procurement requirements that have not been met is conveyed to the supplier through the standard warranty provision contained in each Purchase Order. In addition, where 10 CFR 21 is imposed, suppliers are required to comply with applicable reporting requirements. 17.2.5 INSTRUCTIONS, PROCEDURES, AND DRAWINGS Organizations engaged in O-and F-designated activities are required to perform these activities in accordance with written and approved procedures, instructions, or drawings, as appropriate. Simple routine activities that can be performed by qualified personnel with normal skills do not require a detailed written procedure. Complex activities require detailed instructions. -9er-defiumd 2r -mmplov " pan tha dae4gdetier of the artic.Ly is C., pan 51Lle d q acL.meni manaQat Procedures include, as appropriate, scope, statement of applicability, references, prerequisites, precautions, limitations, and checkoff lists of inspection requirements, in addition to the detailed steps required to accomplish the activity. Instructions, procedures, and drawings also contain acceptance criteria where appropriate. The general manager - Hope Creek operations is responsible for assuring that station procedures are prepared, approved, and implemented in compliance with the station administrative procedures. Documents affecting nuclear safety are reviewed by 8 17.2-19 Amendment / j

HCGS FSAR 8/84 the station operations review committee (SORC) for technical content, by NOA for quality assurance requirements, and are approved by the responsible station department manager or his designee. The general manager - nuclear support is responsible for issuing specifications, drawings, blueprints, and instruction and technical manuals associated with Q-and F-designated structures, systems, and components. Approved and implemented modifications and design changes are incorporated to these reference documents for the life of the station. Master lists of current editions or revisions of these documents are periodically issued by the general manager - nuclear support to the general manager - Hope Creek operations to periodically assure that only current and approved referenced documents are used at the station. (golS NOA reviews and approves station inspect n plans and procedures that implement the quality assurance pr ram, including testing, calibration, maintenance, modification, and repair. Changes to these documents are also reviewed and approved. In addition, NQA is responsible for review and approval of PSE&G specifications, test procedures, and results of testing. 17.2.6 DOCUMENT CONTROL Instructions, procedures, drawings, and changes thereto are reviewed for inclusion of appropriate quality assurance requirements and are approved by apppropriate levels of management of the PSE&G organizations producing such documents, I end distributed on a timely basis to using locations. Measures cre provided for the timely removal of obsoleted or superseded documents from the using location. Supplier documents are controlled according to contractual agreements with suppliers. mi DIMd MI l The following is a generic listing of ocuments for the j operational phase, showing organization responsibility for review approval, including changes thereto: GoYO a. Design specification - Nuclear upport, NQA l g'flG1tioM b. Designp manuf4cturing,-constr-uct-lon, &nd installaMo draalatp - Nuclear SW supporfy nachdrServMS W Greek Oft 2feWG% N&A ) 17.2-20 Amendment 7

jafMb HCGS FSAR 8/84 p /' deparimod 0(Sanya.dM, c. Procurement documents -/)luclear 6 :11;::, purchasing ppartment,NOA (f nxiear servicesy d. Quelity ersurance-menualW A6V/Se perJ//Jer* /~/J ' d l e. StatT5if'Tdmiraas auve pr----Sures - Anera; ;;;n g er -

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Maintenance, modification, and calibration procedures for Q-and F-designated station work activities - -. _ - - - ;;;Mr "^- -ralhm - Hope Creek Operations, NOA g. Operating procedures - C -. " -ag - Hope Creek Operations, SORC h. FSAR - Nuc1 r G yport N 86'M #N l i. Maintenance, inspection, and testing instruction Reekear-Servi ::, ':^' nuclear ds/dtfinenf impler71 nli!} Or an oga, Won.3, NG A l-modificahos) fefprocedes - nucker oe"'M) S j. 9.Msted te:t pmcedurer -ihmiearses v 6, -- -A l 1.' 0;;ign ;heng: reqEFsts,dtEIoar Supph l l ~In addition, NOA involvement in the work activity includes a review of.nonsafety-related work orders for proper class'.fication prior to conducting the activity and a review of completed safety-related work orders. The establishment and maintenance of a document control system ! L for all instructions, procedures, specifications, and drawings ,'I received from the nuclear department, or prepared at the station for use in operating, maintaining, refueling, or modifying items ,l' and services covered by the quality assurance program, is the ,i; responsibility of-the general manager - Hope Creek operations. The administrative procedures manual describes the control of j specific documents. Control of station practices is included in j the administrative procedures and in department directives + authorized by the responsible station department managers. ,/i Measures are established to assure that the administrative i !' (f procedures and department directives are up-to-date, are properly i l 17.2-21 Amendment 7 l

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%% Tumid fr-witleA.. $*r' I m yis & l HCGS FSAR%pom c4 w 4/84 I e. Critical test sequence f. Acceptance criteria. Test resultsjincluding verification of above items,::: 2;;' 1 Lic are documented and reviewed for acceptability by the qualified department representative. System tests performed following modifications to Q-and F-designated systems require review of test procedures and test results by the SORC., 4 p.s+ ~ d *4& NQA memetreTmr monitori.; .m the conduct of O. Jeel;n ch;ngc - ec::;t;n;; tests to assure compliance with the test procedure. Test results are reviewed for the following: a. Presentation of proper documentation 1 b. Assurance that tests meet objectives c. Identification and reporting of unacceptable results and initiation of corrective measures. 17.2.12 CONTROL OF MEASURING AND TEST EQUIPMENT l Test equipment, instrumentation, and controls used to monitor and measure activities affecting quality and personnel safety are identified, controlled, and calibrated at specific intervals by cognizant nuclear department personnel. Written procedures for meeting these requirements include provisions for: a. Specifying calibration frequency b. Recording and maintaining calibration records c. Controlling and calibrating primary and secondary standards d. Determining methods of calibration e. Tracing use on safety-related items. 8 17.2-29 Amendment /

HCGS FSAR 4/84 repair or "use-as-is" are required to be approved by the responsible. engineering representative. Rework or repair of nonconforming material, parts, or components is inspected and/or retested in accordance with specified test and inspection requirements established by the cognizant engineer, based on applicable amm6m requirements. NQA and the nuclear department review nonconformance reports for l quality problems, including adverse quality trends, and initiate reports to higher management, identifying significant quality problems with recommendations for appropriate action. 17.2.16 CORRECTIVE ACTION Organizations involved in activities covered by the quality i assurance program are required to maintain corrective action programs commensurate with their scope of activity. Noncompliances with the quality assurance program identified by i NQA are documented and controlled by issuing an action request. NQA reviews responses to action requests for adequacy and monitors these action requests through periodic summary and status reports to management. Responses to action requests are based on the four elements of corrective action, which are: l' a. Identification of cause of deficiency b. Action to correct deficiency and results achieved to date l c. Action taken or to be taken to prevent recurrence d. Date when full compliance was or will be achieved. N A For significant conditions adverse to uality not identified by N93I such as 12Rs, NRC/INPO/CMAP findings, is involved in the review of such conditions and provides oversight to assure timely follow-up and close out through monitoring, auditing, and commitment verification. <5 - Items 3 and 4 are optional for noncompliances that do not have a significant effect on the quality assurance program. l l I Amendment)I 17.2-32

HCGS FSAR 4/84 7 d. Indoctrination and training e. Implementation of operating and test procedures f. . Calibration of measuring and test equipment ~ g. Fire protection I h. Other applicable activities delineated in Table 17.2-2. The audit data is analyzed and a written report of the results of i each audit is distributed to appropriate management representatives of the organization (s) audited, as well as to other affected management personnel. Included in the report is a statement of OA program effectiveness. Pe-indica 11y. NOA is audited by independent auditors to verify implementation of the corporate quality assurance pr ram. Reports of these audits are directed to appropriate PSE&G.anagement personnel. l l l at In t m %o y I t L i i-5 17.2-35 Amendment /

~ - -, - -,.. u.. HCGS FSAR N / TABLE 17.2-1 (cont) Page 2 of 2 (e) Area radiation monitoring system operation (f) Process radiation monitoring system operation Meteorologi'al monitoring and data collection (g) c program (h) Packaging and transport of radioactive material per 10 CFR 71 (i) Decontamination. 7. Technical Specification surveillance 8. Performing maintenance ~ 9. Chemical and radiochemical control. B. Additional NRC requirements 1. Technical Specification administrative controls (a) Station operations review committee (SORC) (b) Nuclear review board (NRB) (c) Reportable occurrences. s I 2. Inservice inspection plan 3. Reporting of defects and noncompliance. 9. Moddre.#/ms fw 8 tY' S r**l' Ai M l l l F l l ~$~ Amendment / = - - - - - - - - --,-ve. ---.s-- ,----m-m-- s-en----,-o,-ere---va en-- --,w-

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HCGS FSAR 4/84 m J OUESTION 260.15 The fourth paragraph of FSAR Section 17.2.2 refers to Section 1.8 'for commitments to Regulatory Guides. Section 1.8 primariJy addresses Regulatory Guide commitments during design and construction, and the staff review of the FSAR is concerned with Regulatory Guide commitments during the operations phase. With any proposed' clarifications or exceptions, provide a commitment in the FSAR to the effect that: "During the operations phase of HCGS, PSEEG commits to comply with the regulatory position in..." the appropriate issue of the Regulatory Guide listed on pages 17.1-26 and 17.1-27 (with RG 1.33 replacing RG 1.28) or NUREG-0800 (Rev. 2 - July'1981). For systems, components, and structures covered by the ASME Boiler and Pressure Vessel Code Section III (Classes 1, 2 and 3), the code OA requirements should be supplemented by the specific guidance addressed in the regulatory positions of the applicable Regulatory Guides. (2B3)

RESPONSE

Section 17.2.2 lists regulatory guidance applicable to the QA This list has been revised to include Regulatory Guides program. I 1.116, 1.123, and 1.144. PSE&G will revise section 1.8 to reflect compliance with listed Regulatory Guides which are applicable during the operations phase, along with any clarification, modifications, etc. by Junc 1984. l The code OA requirements are used for the procurement of systems, components and structures covered by the ASME Boiler and Pressure l Vessel Code Section III (classes 1, 2, and 3). The standard QA program controls apply to 0-Listed code items following receipt at the station. gg Q*p // M h gu [ kv[k QQ d f.3 g 4, i // he-4eV % 45#iE C=4 s l<f& fvDcOMYs M e-53a-n3 4 qua 5 GJ t i 8 260.15-1 Amendment ( f

HCGS FSAR 1/84 OUESTION 260.50 (SECTION 17.2) Describe the provisions which assure that when inspections associated with normal operations of the plant (such as routine maintenance, surveillance, and tests) are performed by individuals other than those who performed or directly supervised the work, but are within the same group, the following controls are met: (SRP Section 17.2.10, item 2) The quality of the work can be demonstrated through a a. functional test when the activity involves breaching a pressure retaining item. b. The qualification criteria for inspection personnel are reviewed and found acceptable by QANO prior to initiating the inspection.

RESPONSE

See response to Questions 260.7'and 260.19. Section 17.2JO 'ri,_ -.2-25E has been revised to provide additional information requested. I i I 260.50-1 Amendment /f ,g,,, -.---+-- m w 3- ~ - - - pmw

HCGS FSAR 4/84 l QUESTION 260.60 (SECTION 17.2) Describe those provisions which assure that procedures are established to control altering the sequence of required tests, inspections, and other safety-related operations. Such actions should be subject to the same controls as the original review and approval. (14.3)

RESPONSE

Section 17.2.11 states in part: Test procedures prescribe, as applicable: (d) Critical test sequence ..... Test'results are documented and reviewed for acceptability by the qualified department representative. In addition, station administrative procedures provide for the use of temporary changes F Detail instructions for implementation of temporary changes are provided. (4 k E M % Cs w h // c Mb u-- 4td /" %,,2cd S &e4 Nb 260.60-1 Amendment [ l - = -. -

HCGS.FSAR .Seeba s a.6 The diesel generator combustion air intake and exhaust system is Q-listed (Item VIII.h of Table 3.2-1). a.7 Process radiation monitoring associated with systems re-quired for safety are Q-listed (Item XV.e of Table 3.2-1). a.8 QA program controls were not applied during design and con-struction phase for area radiation monitoring and for process radiation monitoring associated with systema not required for aafety (1 tem XV.f of Table 3.2-1). 0A program controls shall be applied during the operations phase to an extent consistent with the item's importance to safety. a.9, Activities covered by the OA program are delineated in a.10 Table 17.2-1, and included radioactivity sampling, radio-active contamination measureme'nt and analysis, and a.11 personnel monitoring internal under " control of radioactivity". l E.12 Activities covered by the QA.pe,ogram are delineated in Table i 17.2-1 and include instrument storage, calibration and maintenance. ~ a.13 Decontamination +quipment and facilities are not safety-related. Decontamination piping and valves are part of the " Liquid Radwaste System" described in Table 3.2-1. Activities covered by the QA program.are delineated in Table-17.2-1 and include of Radioactivity." personnel decontamination under " control a.14 Activities covered by the QA program are delineated in Table 17.2-1 and include respiratory protection and contamination a.15 control under " control of Radioactivity." a.16 Activities covered by the QA program are delineated in Table l 17.2-1 and include accident-related meterological data L collection equipment under " Meterological Mohitoring and j Data Collection Program." l a.17 Not applicable t'o HCGS, a.18 Structural backfill is Q-listed (Item XIX.1 of revised Table 3.2-1). a.19 The seismic Category I electrical duct bank manholes are Q-listed (Item XIX.n of revised Table 3.2-1). O lint *4==d is et : ":tructure,- a.20 Site ;rrdi9; !? aa* eyster, er ce.T+e..:nt" that ch;uld b; included ir Table _ h_,U__f j_ - _ _ _ '..! # N.3 1D_ b5NN'-#IbINi__

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HCGS FSAR 6/84 for HCGS, an appropriate safety classification will be determined. c.1 The HCGS position on TMI Item I.D.2 is given in Section 1.10. The safeht spfem /s part of M. Qro j Roo m I,,pe/ Aphy Spin / Item XV farameO~ duphy d of revised Table 3.2-1). c.2 The HCGS position on TMI Item II.B.1 is given in Section 1.10. The HPCI, RCIC, ADS, and containment instrument gas systems are 0-listed, as shown in Items V.c, VI, XV.b.1, and XVII.b of Table 3.2-1. The RPV head vent is Q-listed but not Class IE (Item I.c of Table 3.2-1). i c.3 The HCGS position on TMI Item II.B.2 is given in Section 1.10. The post-accident shielding is 0-listed (Item XIX.m of revised Table 3.2-1). c.4 The HCGS position on TMI Item II.B.3 is given in Section 1.10. The post accident sampling system (PASS) is not 0-listed with the exception of the primary containment isolation and reactor coolant pressure boundary piping and valves. c.'5 The HCGS position on TMI Item II.D.3 is given in Section 1.10. The SRV position indication system is 0-listed (Item l XV.d of revised Table 3.2-1). l l c.6 The HCGS position on TMI Item II.E.4.1 is given in Section i 1.10. The dedicated hydrogen control penetrations are Q-listed (Item V d.4.g and h of Table 3.2-1). c.7 The HCGS position on TMI Item II.E.4.2 is given in Section 1.10. Containment isolation valves are O-listed (See Table 3.2-1 under applicable system). c.8 The HCGS position on TMI Item II.F.1 is given in Section 1.10. Accident monitoring instrumentation will be designed in accordance with the guidance provided in Regulatory Guide 1.97, Rev 2. This instrumentation will be reviewed l for classification as 0-listed, and Table 3.2-1 will be modified as necessary. c.9 The HOGS position on TMI Item II.F.2 is given in Section 1.10. No additional instrumentation was identified as a result of this required study, and therefore no changes to Table 3.2-1 are necessary at this time. c.10 The HCGS position on THI Item II.K.3.13 is given in Section 1.10. No change was made to the HPCI and RCIC initiation levels and,.therefore no change to Table 3.2-1 are i necessary. i SRAI (1)-10 Amendment 6

HCGS FSAR 6/84 c.11 The HCGS position on THI Item II.k.3.15 is given in S;ction 1.10. The HPCI and RCIC leak detection systems are,-Q-listed (Item XV.e.2 of Table 3.2-1). c.12 The HCGS position on TMI Item II.k.3.16 is given in Section 1.10. HCGS is reviewing the modifications proposed by the BWROG.to meet the requirements. This review will be completed by December, 1983. Table 3.2-1 will be modified as appropriate. c.13 The HCGS position on TMI Item II.k.3.18 is given in Section 1.10. BWROG response to this THI study is still under evaluation by NRC. HCGS design will.be modified to comply with the NRC's acceptable position. Table 3.2-1 will be modified as appropriate. c.14 The HCGS position on TMI Item II.k.3.21 is given in Section 1.10. No change was made to the core spray and LPCI logic and therefore no change to Table 3.2-1 is necessary. c.15 The HCGS position on THI Item II.k.3.22 is given in Section 1.10. The RCIC suction transfer is Q-listed (Item XV.c.1 of Table 3.2-1). c.16 The HCGS position on THI Item II.k.3.24 is given in Section 1.10. The HPCI and RCIC room unit coolers are Q-listed (Item XIII.c.2 of revised Table 3.2-1). c.17 The HCGS position on THI Item II.k.3.25 is given in Section l 1.10. The recirculation pump sealing cooling water supply i system (RAC and CRD) are not Q-listed (Item XI.c and IV of Table 3.2-1). c.18 The HCGS position on TMI Item II.k.3.27 is given in Section 1.10. See Table 3.2 for listing of existing level instrumentation. c.19 The HCGS position on TMI Item II.k.3.28 is given in Section l 1.10. The ADS valves, accumulators and associated equipment I and instrumentation are 0-listed (Item II.1, II.b, II.c XV.b.1 & 11 and XVII.b of Table 3.2-1). c.20 The HCGS position on TMI Items III.a.1.1/III.a.2 is given in Section 1.10. Activities covered by the QA program are delineated in Table 17.2-1 and include emergency plans under, " combating emergencies and other significant events." c.21 The BCGS position on TMI Item III.a.1.2 is given in Section The Emergency A.y <,. Scilities M Aep.<,i,w spiw(ars 1.10. is s4.=n bi 'It'en XV.d of fable 3.2-1). f Trevhe.{ 1. SRAI (l)-11 Amendment 6 4 e

HCc-S FSAR 6/84 c.22 Activities covered by the QA program are delineated in Table i l 17.2-1 and include inplant I, radiation monitoring under L " Control of Radioactivity." c.23 The HCGS position on TMI ltem II.d.3.4 is given in Section 1.10. This item is not a " structure, system or component" requiring entry in Table 3.2-1. Control cf this activity is provided by appropriate procedures. Chapter 17 describes the Quality Assurance Program coverage of procedural controls. l The following information is provided for additional clarification: a) The nonsafety-related, non-ESF internal cpmponents include the steam dryer, the shroud head and steam separator assembly, the in-core guide tubes and stabilizers, the l i differential-pressure and liquid-control lines inside the RPV, the fuel orifices, and the feedwater spargers. In all BWR 4, 5, and 6 designs, these components are not 0-listed because they are neither required for safe shutdown of the plant, nor would their failure jeopardize the safety functions of other safety-related internal components. During the operating phase of the HCGS, the same high-quality design, procurement, and installation control practices, as were applied during the design and construction phase, will,be applied to any changes to these components. As Section 3.2.1 and notes (13) and (50) for Table 3.2-1 indicate, the quality assurance controls for non-ESF RPV internal components and for seismic Class II/I equipment are described in Section 1.8.1.29. I In addition, the reactor pressure vessel internal structures which are accessible are included in the ISI program, which is covered by the operational QA program. b) Reactor building penetrations are not required to be Q-listed unless the piping system is 0-listed. A non-Q piping system penetrating the reactor building is not required to have a 0-listed penetration, r _ -. _. ir,

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1__ ,n-. - -.__,= c) The spent fuel pool liner does not perform a safety function and therefore is not 0-listed. However, the spent fuel pool does meet the quality assurance requirements of 10 CFR 50, Appendix B, and has been noted as such in Table 3.2-1, Item XIX.e. d) Shore protection of the intake structure does not have a safety function and therefore is not 0-listed (Item XVIII.j SRAI (1)-12 Amandment 6

HCGS FSAR 6/84 of rGvited Tcblo 3.2-1). Howsyst, it ic designed to n flood and seismic event. accommodate desig&cdims +s +k roof dram p+wn " nnj med; ss e) The roof drainage systemAED not 0-listed and Aminot a " structure system or component" that should,be included in Table 3.2-1. Roof drainage cannot adversely impact safety-related equipment because of flood protection measures " A* h y n 4-b bl 85f*4 o f & Q 'fN Y"* y pl.J:4 "hr.. rt a ^hi-Sf

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^ ^ ' E-d !S5% ~ f) The purge (containment inerting) system is described under ~ the containment atmosphere control system (Item V.d.3), not the reactor building ventilation system (Item VIII.c). g) Containment isolation valves used at HCGS meet the requirements outlined in GDCs 54-56 of 10 CFR 50 Appendix A as outlined in Table 6.2-16. h) Table 3.2-1, Item V.a has been revised to clearly identify piping, valves and other equipment used for suppression pool cooling, steam condensing and suction lines for the shutdown cooling modes of the RHR system. 1) There are no nuclear codes and standards applicable to the design and manufacture of the HPCI and RCIC turbines. Approximately 50 to 75 components of the turbines' lubricating oil systems contribute to the electrohydraulic control of the governing valves. Footnotes (11) 555 (48) an/(,Si' provide the applicable quality assurance, documentation, maintenance, and material fabrication information. Process and effluent radiation monitoring systems are listed j) in Item X.d of Table 3.2-1. See Sections 7.6 and 11.5 for the differences between the process radiation monitoring systems and the process and effluent radiation monitoring systems. l k) Table 3.2.1 will be revised to incorporate the Emergency Response Facilities Data Acquisition System (ERFDAS). This system is non-0, non-class IE and non-seismic,except for the Class IE isolation devices supplied with the ERFDAS. 1) The MSIV sealing system consists of valves, valve operators, and piping only; the sealing system is supplied by the instrument gas system (see Item XVIII.b). a) The unit vent stacks are 0-listed as shown in revised Table 3.2-1, Item XIX.g. s SRAI (1)-13 Amendment 6 .}}