Text

Forwards Revs to FSAR for Hope Creek Generating Station.Revs Consist of Text Changes to Maintain Consistency W/Tech Specs & to Reconcile as-built Discrepancies.Revs Will Be Included in Amend 15 to FSAR
	ML20141D973
Person / Time
Site:	Hope Creek
Issue date:	04/04/1986
From:	Corbin McNeil; Public Service Enterprise Group
To:	Adensam E; Office of Nuclear Reactor Regulation
References
	NUDOCS 8604080339
	Download: ML20141D973 (103)
	v • d • e

{{#Wiki_filter:- Public Service Dectr;c and Gas Company _ Corbin A. McNelli, Jr. Pubhc Serwce Electric and Gas Company P O. Box 236, H ancocks Bridge. NJ 08038 609339-4800 Vice President - Nuclear April 4, 1986 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Ms. Elinor Adensam, Director Project Directorate 3 Division of BWR Licensing

Dear Ms. Adensam:

FINAL SAFETY ANALYSIS REPORT REVISIONS HOPE CREEK GENERATING STATION DOCKET No. 50-354 Public Service Electric and Gas Company (PSE&G) hereby submits various revisions to the Hope. Creek Generating Station (HCGS) Final Safety Analysis' Report (FSAR). The attached revisions to the HCGS FSAR contain:

1) revisions to maintain FSAR consistency with the Technical Specifications;

2) revisions to reconcile as-built plant discrepancies; and 3) general changes to the FSAR text, tables and figures. provides a brief summary and explanation for each change while Attachment 2 contains the actual marked-up sections of the FSAR.

These revisions will be incorporated in FSAR Amendment 15 after fuel load but are being filed now in order to accurately reflect the design and operation of HCGS and support the issuance of an operating license. In addition, an affidavit is provided to affirm that the matters set forth in this transmittal are true and accurate. This submittal supplements similar transmittals from C.A. McNeill to E. Adensam dated March 3, 1986, March 17, 1986 and March 21, 1986. as*9Bau Pes %A4 i ^ q l

i Director of Nuclear 2 4-4-86 Reactor Regulation Should you have any questions on the subject filing, do not hesitate to contact us. Sincerely, Affidavit Attachments (2) C D.H. Wagner USNRC Licensing Project Manager R.W. Borchardt USNRC Senior Resident Inspector t

i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC AND GAS COMPANY FINAL SAFETY ANALYSIS REPORT REVISIONS Public Service Electric and Gas Company (PSE&G) hereby submits various revisions to the Hope Creek Generating Station (HCGS) Final Safety Analysis Report (FSAR). These HCGS FSAR revisions consist of text changes to maintain FSAR consistency with the Technical Specifications, revisions to reconcile as-built plant discrepancies, and general revisions to the FSAR text, tables and figures. The matters set forth in these revisions are true and accurate to the best of my knowledge, information, and belief. Respectfully submitted, Public Service Electric and Gas Company By: Corbin A. McNeill, J r'. Vice President - Nuclear . Sworn to and subscribed before me, a Notary Public of New Jersey, this gM day of April 1986. . DELOMS0.HA0000 ~ A Notary Public of New Jersey My Commsse Expires Merch 14,ISOS

I- ~ ATTACHMENT 1

SUMMARY

OF CHANGES, ADDITIONS AND/OR MODIFICATIONS 1.8.1.32 + Revisions reflect the resolution of 8.3.2.1.2.3 NRC questions regarding battery charger 8.3.2.1.2.4 design capacity. These revisions have 8.~3.2.2 been previously submitted to the NRC in a letter from C.A. McNeill to E. Adensam on March 26, 1986. 1.8.1.68 Revisions clarify the sequence of testing 14.2.13.1 RPS and ATWS circuits and maintains consistency with the Technical Specifications. 3.9.3.1.12 Revisions reflect the as-built plant T3.9-5v condition and provide a correction pg. 3/3 for an editorial error in Table 3.9-5v. 3.11.1.4 Revisions reflect the replacement T3.11 of Tobar transmitters with Rosemount pg. 10,12,54 transmitters and the appropriate 119/119 discussions regarding the environmental conditions associated with these transmitters. The environmental conditions in areas outside primary containment, but inside the reactor building, are not capable of becoming harsh until after initial criticality. 4.6.3.1.5 Revision clarifies the wording of surveillance tests for the CRD system to maintain consistency with the Tech-nical Specifications. 6.2.5.3.5

+

Revisions reflect additions, deletions 6.7 and changes necessary to support Appendix 6.7.1.1 J Exemption Request #4 - Main Steam 6.7.1.3 Isolation Valves. These revisions 6.7.2 has been previously submitted to the 6.7.2.1 NRC in a letter from C.A. McNeill to 6.7.2.3 E. Adensam dated March 17, 1986. In 6.7.3 addition this information impacta SER 6.7.3.1 Sections 6.2.6 and 6.7. 6.7.3.5 6.7.3.7 6.7.4 6.7.6 T6.7-1 pg. 2,3,5/6 F6.7-1 15.6.5.5.1.2

2 .T6.2-16 Revision maintains consistency pg. 3/33 with the Technical Specifications. This revision supercedes the revision submitted to the NRC in a letter from C.A. McNeill to E. Adensam dated March 17, 1986. T6.2-24. Revision reflects the as-built plant pg. 3/17 condition. 6.4.2.3 Revisions clarify the as-built plant 6.4.3.1 condition. The revision to Page 6.4-5 supercedes the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adensam dated March 17, 1986. 6.'8.1.5 Revision clarifies seismic requirements r L 6.8.2.5 for instrumentation associated with 4 - ' T6.8-2 the FRVS. p0 1/3 y. 7.2.1.1.1 Revision clarifies the as-built plant condition. T7.3-1 Revisions reflect the as-built plant T7.4-1 conditions. 7.6.1.7.2 Revisions identify the as-built plant T7.6-6 conditions, timers and logic functions. F7.6-8 sh. 4/9 7.7.1.1.3 Revision reflects the as-built plant condition. 7.7.1.6.3.5 Revision reflects the as-built plant conditions. 8.3.1.1.2.10 Revision maintains consistency with the Technical Specifications. This revision affects SER Section 8.3.3.1.2.10, page 8-12, in that additional methods are utilized for thermal overload protection of MOVs than identified in~the SER. T8.3-1 Revisions reflect the as-tested plant pg. 2,4/10 condition. These revisions supercede the revisions submitted to the NRC in a letter from C.A. McNeill to E. Adensam dated March 21, 1986. 9.1.3.2.2.4 Revision reflects the correct water quality conductivity limit for the fuel pool filter demineralizer effluent. This revision supercedes the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adensam dated March 17, 1986. !e

3 9.5.1.2.11.1 Revisions reflect the as-built plant 9.5.1.2.26 conditions. Revisions provide the additions necessary 11.2.2.1.3 11.2.2.2.5 to describe the liquid and solid radwaste Fil.2-5 contingency bypass for concentrates, 11.4.2.-1 sludges, and resins. This information , 3.g 1(f ' js 11.4.2.2 was previously discussed in a letter 11.4.2.3 from R.L. Mittl (PSE&G) to W.R. Butler 3. 11.4.2.6 (NRC) dated August 21, 1985. These Fil.4-10 revisions affect various portions of SER Sections ll.2.and 11.4. The revision to Page'11.2-7 supercedes the revision transmitted to the NRC in a letter f rom C.' A. McNeill'to E. Adensam dated March 17, 1986. Tll.2-10 Revisions to maintain consistency with pg. 3,5/5 vendor documents and reflect the as-built plant condition. 11.4.2.4,8 Revision reflects the as-built plant condition. l' ~ 'F14.2-5 Revision accurately depicts the power r.3 ascension program. This revision supercedes g the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adenam dated March 21, 1986. '15.2.4.1.2.1 Revisions provide reference to and 15.2.4.3.2 discussion of the MSIV closure analysis. '15.2.7.2.3 Since the analytical limit falls between T15.2-5 the Technical Specification trip setpoint, 8%, and the allowable value, 15%; it is noted that transient results are not significantly impacted when the analytical MSIV closure scram value of 10% is utilized. These revisions are in accordance with SRP 15.2.4 and reflect similar revisions previously submitted on the Limerick and' Perry dockets. These revisions impact the SER as noted + These revisions have been previously submitted to the NRC as noted w-, ,~ v

L2 4 ATTACHMENT 2 l i l l

y HCGS FSAR 1/84 i k 1.8.1.32 Conformance to Reaulatory Guide 1.32 Revision 2, February 1977: Criteria for Satety-Related Electric Power Systems tor Nuclear Power Plants Although Regulatory Guide 1.32 is not applicable to HCGS, per its implementation section, HCGS complies with IEEE 308-1974, as bibblandorsed and modified by Regulatory Guide 1.32, subject to the clarification of Position %C.1.d and C.1.f. = .1 /WSad1 Position C.1.d of Regulatory Guide 1.32 references Regulatory Guide 1.75. HCGS compliance to this Regulatory Guide is discussed in Section 1.8.1.75. Position C.1.f of Regulatory Guide 1.32 references Regulatory Guide 1.9. HCGS compliance to this Regulatory Guide is discussed in Section 1.E.1.9. See Chapter 8 for further discussion of the electrical system and Section 1.8.2 for the NSSS assessment of this Regulatory Guide. 1.8.1.33 Conformance to Regulatory Guide 1.33, Revision 2, February 1978: Quality Assurance Procram Requirements (Operation) HCGS complies with ANSI N18.7-1976/ANS-3.2, as endorsed and modified by Regulatory Guide 1.33. The contents of the plant operating procedures will comply with the applicable requirements of Section 5.3 of ANSI /ANS-3.2-1982. See Section 17.2 for further discussion of quality assurance during plant operation. 1.8.1.34 Conformance to Regulatory Guide 1.34, Revision 0, December 28, 1972: Control of Electroslag Weld Properties Regulatory Guide 1.34 is not applicable to HCGS because the process is not used. l See Section 1.8.2 for the NSSS assessment of this Regulatory Guide. l 1.8-18 Amendment 4 [.

INSERT FOR PAGE 1.8-18 HCGS complies with Position C.l.b of Regulatory Guide 1.32 as discussed in Section 8.3.2.2.

HCGS FSAR 01/86 Criteria for Engineered Safety Feature Atmospheric Cleanup System of Light-Water-Cooled Nuclear Power Plants, is addressed in Section 1.8.1.52. k. Appendix A, Paragraphs 1.k(2) & (3) - Preoperational testing of personnel radiation monitoring and survey equipment or laboratory equipment is not performed. Calibratior tests are performed prior to core load in accordance with station procedures. 1. Appendix A, Paragraphs 1.m(4) & 1.0(1) - Regulatory Guide 1.104 was withdrawn by the NRC on 8/22/79. During preoperational testing, the cranes will be verified to function in accordance with specifications. The controls, interlocks, and travel limits of the reactor building and fuel handling crancs are verified, m. Appendix A, Paragraph 1.n(11) - Compliance with Regulatory Guide 1.80, Preoperational Testing of Instrument Air Systems, is addressed in Section 1.8.1.80. salTIAL

#E"l'Y n.

Appendix A, Paragraph 2.c - The reactor protection I system will be functionally checkedt prior tofEEEI f JsEM using station surveillance and calibration W ACf0RDWCG WaH UE HCGS Iprocedures. The reactor protection system is shown to TEFGdL operate in conjunction with the control rod drive $wicATIM startup test, described in Section 14.2.12.1.8.

Also, j

the reactor protection system is verified to operate following scheduled transient tests such as MSIV 7mq isolation and turbine tripj. LCAb y Ec3ecTIcM J o. Appendix A, Paragraph 5.0 -lPertions of leck detect 4en l gcecrned by Technical Spec 4f-leat4 enc will be-f unctiona11y chceked-dusrt-pedor-to-fue-1-load-us1-ng-atation survei11anee-and-ca14bret-len-procedures Setpoints related to leak detection high steam flow in HPCI and RCIC are determined and set as stated in Sections 14.2.12.3.12 and 14.2.12.3.13. Normal operation of leak detection systems, such as drywell equipment drain sump pump will be accomplished using station operating procedures. 6 1.8-41 Amendment 14

HCGS FSAR 01/86 w. Appendix A, Paragraph 5.gg - The ATWS subsystems are thoroughly checked out logically and functionally during the preoperational test program, as described in Sections 14.2.12.1.2.c.6, 14.2.12.1.3.c.3, 14.2.12.1.4.c.4, 14.2.12.1.8.c.9, 14.2.12.1.9.c.7, and,f' RECIRCLtLAYAl 14.2.12.1.10.c.4. JPortions of ATh's governce by SfZ9n 7EOd5 Technic;l Spccifications will be functionally checked

dD THg juct prier te fuel lead usinc ctation curveillance and t

m/ calibratic crecedurec./ The recirculation pump trip y_ T(d2g myg h (RPT) is tested as part of the/tencrater and turbinc? trips that are performed in Phase III testing. x. Appendix A, Paragraph 5.ii - Hope Creek design does not incorporate the recirculation flow control valve. 1.8.1.68.1 Conformance to Regulatory Guide 1.68.1, Revision 1, January 1977: Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants HCGS complies with the intent of Regulatory Guide 1.68.1. For further discussion of the initial test program, see Section 14. 1.8.1.68.2 Conformance to Regulatory Guide 1.68.2, Revision 1, July 1978: Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants HCGS complies with the intent of Regulatory Guide 1.68.2. For further discussion of the initial test program, see Chapter 14. I ~ ~ 1.8-42a Amendment 14

HCGS FSAR a. The operating conditions for each SLC pump and motor are functionally tested by pumping demineralized water through a closed test loop. Each SLC pump is capable of injecting the net contents of the SLC tank into the reactor in not less than 50 minutes and not more than 125 minutes. Each pump is capable of injecting flow into the reactor against a pressure of zero psig up to the initial setpoint pressure of the reactor safety / relief valves (SRVs). b. The design conditions for each SLC pump include: 1. Flow rate: 43 gpm i \\ { @ H0 F) 2. Available NPSH, maximum: 12.9 psi % 3. Maximum operating discharge pressure: 1190 psig 4. Ambient conditions: Y 1 g g y g.p; Temperature:\\l50 to 10'"r; [2o ro q5 % l Relative humidityh 20 tc ^ 0 *c ! 5. The normal plus upset conditions which control the pump design include: 0.9 Gees.) 0.04 (Ext.) (a) Design pressure 1400 psig (b) Design temperature _'C 7 ' ' ni n7r 1500F J' (c) OBE ,S - (d) suctica nczz;c ;cccc t 70 peuns;, o -Mo - '90 foot-pounds 'c) Diccharge nozzle leads F - 370 poundo, o M 0 foc bpounds u "here: Fn and M are 30 defined ir Table 2.9-Sn. a 3.9-66

a ..b g t HCGS TSAP 1/Mb l T\\flLE 3.9-Sv (co-t) l'a qe i nt J l Q Attow.itete ca lcul at ers(2) Componeat ard p g Criterion Loidini controtting Stress ..:.-,, f y g Noar1e Load Det1*i1ttor.: W DS g l'orces are ir. (lb) and moments The normal plus upset cordit to-F = Allowable valise of F i g are 17 (tt-1t3 the allowable loads ir.clude: wner all momeats tre zero. I comb 1 rat ion of forces and moments (Las) g are as follows: Design pressure and tempera-M. = Allowable val'ae of M 3 ture, dead weight aid thermal wher all f orces are rero.g l expars tor, and operating basis [fT-fES) l F. F M earth 1uake. l g g + 1 1 Suetto, norrie: l P_ Fe M 4,580 Rg M%l o Fe = gg M. = n, ou in = 4r,71 1 ' ym Di nCII.t r tl" l rorrie: 7;H50 F.'s afico ll Fe = M M. M. n, w> n;r :3,#qi i = i The emerger.cy 3r t ault ed Sitct i o, I cond it or. loads te cl u'1e

norrie:

l M F a %$ l Fe 6,6Mu = i Mo = 1't,

s u

+ N[,,q q l Design pressure and temperature, 7 l dead weight and thermal Otacharge norrie: l expar s tor., and sat e shutdown

r. = 9,s20 m

rg. t8:3 I ,5-4 e q. '3y carthiuske. M. IN,464 = g Y Where Fg (Ib) s the maximum of the three ort hogo. gal f orces F,, l F,, F,) an is the mutmum of any of the three orthogoral moments l H2N M for the same refernce coordinates. F, and Me for i M,, Psy,nd*)taulted conditions are base values given above. upset a I I Note l <st the calculated stresses for the energency or f aulted cor ditions are less than their g corresponding allowable stresses f or the noreal plus upset condit tor ; tneretore, rormat plus l upset condition is r.ot evaluated. l i (2.) CAttitLATEb UMDS UXteC EVAUU1TSD A% ACC W By (-EdAL ELECWIC TEC LETTER GB-9L.>l DGTEb JAOrAP1/ 3I,l%% ~ Amenoment is l t t M

HCGS FSAR 09/85 \\ Radiation exposures to components will be minor and will be due to two_ sources: radiation shine, and immersion in airborne radioactivity released in a controlled manner from the reactor building. The TID from both sources will be less than 100 rads for 180 days. 3.11.1.3 Excluded Systems - NSSS and Non-NSSS Table 3.11-8 shows the systems designated as seismic Category I in Table 3.2-1 that are to be excluded from the HCGS Environmental Qualification Program. The table identifies each system with correlation to Table 3.2-1 and the reason for exclusion from the HCGS Environmental Qualification Program. 3.11.1.4 Environmental Conditions - NSSS and Non-NSSS The environmental conditions shown in Table 3.11-1, and the associated figures may be changed at a later date because of continuing evaluations that are being performed on a case by case basis. k llhCd2r --* 3.11.2 OUALIFICATION TESTS AND ANALYSES 3.11.2.1 NSSS Safety-Related Class 1E Electrical Equipment Harsh Environment Oualification Components of the nuclear steam supply system (NSSS) Class 1E electrical equipment are qualified in accordance with the environmental qualification criteria and guidelines specified as Category II in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," dated December 1979 (for comment), and IEEE-323-1971, "IEEE Standard for Qualifying class IE Equipment for Nuclear Power Generating Stations." However, the HCGS environmental qualification program has upgraded qualification of equipment to the requirements of NUREG-0588, Category I and IEEE 323-1974. Components of the NSSS Class 1E electrical equipment in a harsh environment are qualified by test, or a combination of testing and analysis. Those components used in more than one system, which can be or are located in different plant areas, are (\\-- ' 3.11-4 Amendment 12

INSERT FOR'PAGE 3.11-4 The environmental conditions that components may be exposed to in areas outside primary containment, but inside the reactor _ building, are not capable of becoming harsh _until after initial criticality is achieved.

? .] p p-1/96 HCGS FSAs Page lo of 119 TABLE 3.18-5 EOUIPtiENT SFLFCTED FOR MARSH FNV!pONNFMT QUALIFICATION PSID N-II-I ( SY2 TENS SAFETY AUX Cool SYSTEM EG PAN TN! ACTION ID NO. EDCATION FOUIP. PLAN fQUIP. P.O. Note (St NPL NO. COMPONFNT D Ll1G. FtFV. NOTR (Il Not ti(,2 5 FF98 prr. No, K001 1-FG-PtFT-24054 Diff. Press Trane, peactor 102 No No in IS'5A K001 l-FG-PDT-2485R Ditt. Press Trans. Reactor 102 No No 155 l*5A Noul I-FG-PUT-2485C Diff. Press Trane. Desctor 102 No No 155,8 % Mun! l-rG-PDT-24u5n Oltf. Prese Trane. Reactor IU2 No No 155 J20lO l-FG-HS-2485A2 Hano Switch peactor 102 No No 25 J2010 3 - 8.G-H S-2 4 8 5 8 2 Hand Switch peactor IUJ No No 25 J2010 1-EG-MS-200$C2 Hand Switch peactor 102 No fm 25 JJOIO l-EG-MS-248502 Hand Switch peactor 102 No No 25 P1050 1-FG-HV-24914 Contr. Velve peector 102 peo No 144 P3050 1-tG-tS-24934 Limit Switch peactor 102 No No 144 P3050 1-FG-HV-249tM Contr. Valve peactor 102 No No 144 P9050 1-EG-IS-Jetle Limit Switch peector 102 No No 144 Pl050 !=rG-HV-2494A Contr. Velve peactor 102 No No I44 P1050 3-tG-IS-24944 Limit Switch peactor 102 No No 144 P)D50 3-tG-HV-24948 Contr. Velve peactor 104 No No 144 P1050 1-FG-IS-2494M Limit Switch poector 102 No No 844 PID50 8-tG-HV-24944 Contr. Velve peactor 102 No No 144 P l"O l-FG-sp-24964 Limit nettch peactor 102 too

=

I44 i PIS50 8-FG-HV-2496R Contr. Velve Peettor 102 No 144 P1050 3-EG-IS-24969 Limit Switch posctor 102 No No 144 A-P3050 1-tG-uV-2496C contr. Valve menctor 102 No No I44 PIOSO l-EG-IS-2496C Limit Switch Reactor 102 No No 144 P1350 I-EG-NV-2496D Contr. Valve peector 102 No No 144 P1050 l-FG-IS-24960 Limit Sultch peactor 102 Ito No 144 J30l0 l-FG-LT-2500A Level Trans. Reactor 2OI No No im J 23 00 3*EG-LT*huoN Level Trane. Reactor 208 ISO Mn IQ J10lO l-EG-LT-250cc Level Trane. Reactor 208 No No 29 J1280 3-FG-LT-2500D Level Trans. Reactor 208 No No 29 / Amendsent 14 I

e-n. I 1/P(. NCGS FSAp

age 3 2 of !!9 TABLE 3.33-5 EOulPNENT SFLECTED FOR NApSN FMVIROceNFNT QUALIFICATION P610 N-II-I SYSTrup SAFETT AUX COOL SYSTEN FG PAN TMI ACTION I D MO.

14 CATION EQUIP. PLAN EQUIP. P_. O. Note 151 NPL NO. COM PONENT DLDC. ELEV. NOTE (1) leOTE 12) EESS BEF. NO. P)D50 1-EG-HV-2522C Control Velve peactor 802 No No 147 P30%O I-FG-IS-2522C Limit Sultch peactor 102 No No 147 P3050 l-EG-MV-2522D Control valve poector 102 No No 147 P3050 1-EG-25-25220 Limit Switch peector 102 No No 147 J20lO l-rG-MS-2522D3 Mand Switch peector 102 No No 25 J10lO I-t G-PtTT-25 2 94 Press. Diff. Trane. Reactor 162 No No 29 J1GIO l-EG-Pt7r-25 299 Press. Diff. Trans. Reactor 162 No No 29 J4560 1-EG-TE-2515A Teep. Element peactor 102 fee No 43 J5560 l-EG-TE-25358 Temp. Element peector 102 Yee No el J)DIO I-EG-FT-25444 Flow Trene, peactor 102 No No E ' l$Q J10lO l-FG-FT-2544n Flow Trans, peector 102 No No A.**- isTA J)olo I-FC-FT-2544C Flow Trane. Reactor 102 peo No 4 gggA J10lO l-rG-FT-2544D Flow Trane, peactor 102 No No R gg J10lO l-FG-FT-254941 Flow Trene. Reactor 102 Yes No Jo JinIO l-EG-FT-2549BI Flow Trane. Reactor 102 Yes No 29 J)DIO l-rG-FT-454983 Flow Trene. Reactor 102 No No 29 P9050 1-EG-MV-79234 Contr. Valve peactor 162 No No 144 P3050 3-EG-IS-7928A Limit Switch peector 162 feo No 144 7 P)DSQ 1-EG-MV-79219 Contr. Valve peactor 162 00 0 No 144 P10%O l-EC-IS-79219 Limit Switch peactor 162 soo too 544 P3050 l-rc-MV-7922* Contr. Velve peactor 162 leo No let b P30%O l-EG-25-7922A Limit Switch peector 162 No soo 144 P3050 1-EG-*w-79229 Contr. Velve peactor 162 Iso No 144 P3550 3-EG-29-79220 L!stt Switch peector 162 No No 144 Amendment 14

c .~ n ~ ( ) l/P6 NCCS FS As p g. 5 4. og glg TARI.E 3.II-5 FOUIPNFNT SFLFCTFD FOR HApSN ENVIpONNFMT OffALIFICATION P&lD N-51-1 SYSTIN: RF%IISUAL NFAT RFrerWAL SYSTEM DC PAN TNI ACTION I D seO. LOCATION EQUIP. FLAN FOUIP. P.O. NOTE I5) NPL No. Cope >000E NT RLDC. FLFV. BeOTE II) NOTE I2) FFSS DFF. NO. an05 3 - pC-PT-N 0 5 6 N ERI Press. Trane peactor $4 No No 155

001 8-MC-PT-N057 Ell Frees. Trane, peactor 77 aso No 106 uM'0 5 4-RC-PT-N0504 Eli Press. Diff. Trane, peactor 77 No No 1%%

N006 l-DC-PT-N05en Ell Prete. Dif f. Trane, peactor ?? No No 155,13TA mt'0 8 I-DC-PT-N050C Eli Press. Di f f. Tr ane. peactor 77 No feo 155 N001 8-DC-PT-N050D Ell Frees. Diff. Trane, peactor 77 No No ISS,IS$k "003 4-RC-FOT-N060A Ell Frees. Diff. Trane. peactor 77 soo No 106 N005 1-PC-PDT-WO609 Ell Press. Diff. Trane, peactor 77 No feo 106 F64849 I-RC-TE-N0274 Temp. Elemt. Reactor 77 fee

m 34 F43159 I-RC-TE-NO27n Temp. Elemt.

peactor 77 Tee No 34 J55a0 I-MC-TE-4401 Temp. Elemt. peactor 77 00 0 No 43 JI180 3-RC-FT-4415 Flow Trans. Peactor 77 00 0 No 29 Pluto I-nC-Nv-4419 Control Velve peactor 77 peo No 127 P1980 I-MC-IS-4439 Limit Switch peactor 77 No No 127 JIolO I-SC-FT-4464A Flow Trane, peactor 77 fee No 29 Jtolo I-DC-FT-44415 Flow Trane. Desctor 77 fee No 79 JIolo I-pC-FT-44624 Flow Trane. peactor 102 Tee No IM Jiolo I-pC-FT-44628 Flow Trane. Peactor ?? Tee No Amoadment I4 6

o s t 3 3/M

MS FSA" TA98.R ).85 %

rage.119 of 189 wrtTr s (Il Thle colven identlflee the equigment located in a harsh environment required fn. Poet accident monitoring== defined in D.G.I.st and Psas section 1.4.1.97,1. 429 Thle ento== identiftee the equipment located la a herah environment identitled og int Action Plan Pquipment weisru e737 og defined in 8 33 Ste=e that are generic ete purchemed and Joed throughout the plant and may be used in the dea March Environment. lel Electrle Penettet ten Ateegblies lor egetrol and last rument ation c ircult e include inline spIlCe Connnector h t t 9, til stagigneter for purposee et equipment quellficet ton accountabilit y only.timit or poeltlen switch tog nuohere have been created by v othemettee or the inett went I nde s. Faceptione to thle occur when poeltlen seltchegThese t og numbero may not be found in PalD's, u ter meal leggSI P.o. are mounted on manuel welseg or are In that coes the t og nuntpero may be found in the Snetrument indes and appIlcohle PalD'e. e (CD Trenomitters with EESS Def. No. 29 wb t! be changed to Rosemount !!53e Traneelttere ut!!! sing output code

A*

Electronic Circuit noorde prior to initial cr!Licality, e ben 4=ent 14 i i i 1

1 HCGS FSAR 1/85 l 4.6.3.1.5 Surveillance Tests The surveillance requirements for the CRD system are described below: a. Sufficient control rods are withdrawn, following a refueling outage when core alterations are performed, to demonstrate with a margin of either 0.38% ak/k or 0.28 ak/k as appropriate that the core can be made suberitical at any time in the subsequent fuel cycle with the most reactive operable control rod fully withdrawn and all other operable rods fully inserted. b. Each partially or fully withdrawn control rod is exercised one notch at least once each week. In the jF"' [ event that operation is continuing witNOrrrz: cr -- 2 onEr #wn044RE wees >: vec cui ci cerv:r:2 this test is performed at Count (. Bob, [ least once each day. The weekly control rod exercise test serves as a periodic check against deterioration of the control rod system and also verifies the ability of the CRD to scram. If a rod can be moved with drive pressure, it may be expected to scram, since higher pressure is applied durino scram. The frecuency of exercising the _ control rods, under the conditions ofhttrcz cr 2;rr:_=

contral rado valucd out of scrvic 2 provides even

,)P'further assurance of the reliability of the remaining control rods. c. The coupling integrity is verified for each withdrawn control rod as follows: 1. When the rod is first withdrawn, observe discernible response of the nuclear instrumentation 2. When the rod is fully withdrawn the first time, observe that the drive will not go to the over-travel position. Observation of a response from the nuclear instrumentation during an attempt to withdraw a control ( 4.6-42 Amendment 9 l

HCGS FSAR T 6.2.5.3.5 Results The analysis was undertaken to determine the capabilities of the recombiners to control oxygen concentration inside the primary containment and to determine the impact of MSIV air inleakages on recombiner operation post-LOCA. a. Oxygen / hydrogen concentrations - With an initial containment oxygen concentration of 4%, the recombiner would be required to start at 1.52 days post-LOCA when the oxygen concentration exceeds 4.5%. The oxygen concentration initially decreases due to the increase of temperature and steam in the containment and then increases due to the additional air inleakage from the MSIV. At 1.52 days, it decreases again due to the recombiner operation; but as the containment hydrogen concentration is depleted, the oxygen concentration eventually builds up again due to the continuous MSIV leakage. At 39 days post-LOCA, the oxygen concentration exceeds 5%. The hydrogen concentration is negligible, however, thereby maintaining the ability to prevent hydrogen burning. In cases with a THI-type design situation where large quantities of hydrogen are released, the hydrogen would continue to recombine with the oxygen from the MSIV air inleakage for a longer time until all the H, is recombined. This is possible as the oxygen removal rate by the recombiner exceeds the total oxygen addition rate by radiolysis and MSIV air inleakages. b. Containment pressure - The recombiner is limited to operate at full flow only at pressures less than 30 psia. For a design basis accident (DBA)-LOCA, the containment pressure is below 27 psia after 1.25 days, which is before the recombiner is required to operate due to reaching a 4.5% oxygen concentration. The increase in pressure,r'due to the MSIV air inleakage at the#c'.

r.re: :c ; reaches 27 psia at 137 days after CallBmMU) the accident.

It does not reach 45.7 psia (31 psig) or EdMI C" half of the peak design pressure according to 96ECFH SRP 6.2.5, Section II.4, until approximately 450 days RsP. ALL after the accident. These time-periods are well beyond - Ax4R the point when iodine releases due to venting to enmo maintain the pressure would result in significant Ops 4M offsite doses. Figures 6.2-39 and 6.2-40 are plots of OLW'S i 6.2-83 t

HCGS FSAR 10/84 the post-LOCA containment pressure versus time up to 180 days after the accident. The discussion above indicates that the recombiners are adequate to control the oxygen concentration inside the primary containment post-LOCA. The results presented are based on the following conditions: 1. 44 initial oxygen concentration 2. 4-1/24 oxygen concentration for recombiner actuation Fot AU. RUR scfh air inleakagep:: :s: y mb 3. m4:0 STER 4M UNS CfWlBsNED 4. 150 scfm maximum recombiner flow. Figures 6.2-32 and 6.2-33 show the hydrogen and oxygen concentrations versus time in the drywell.. Figures 6.2-34 and 6.2-35 show the concentrations in the suppression pool air space. All four figures show the concentrations for the case using no recombiners and for the case with one train of the recombiner system operating at its design flow of 150 cfm. Figures 6.2-36 and 6.2-37 show the cumulative total oxygen and hydrogen generated inside the containment. As indicated by the oxygen and hydrogen concentrations in the previous figures, the recombiner capacity of 150 scfm exceeds the hydrogen and oxygen generation rates at all times during the accident. Both hydrogen and oxygen concentrations are monitored following a postulated LOCA. The hydrogen recombiners will be started under either of the following conditions: a. Oxygen concentration is greater than St and hydrogen concentration reaches 3.5%, or b. Hydrogen concentration is greater than 4% and oxygen concentration reaches 4.54. Table 6.2-21a provides the peak hydrogen and oxygen concentrations inside the drywell and suppression chamber immediately following recombiner actuation. ( 6.2-84 Amendment 8

TA6Li 6.2 16 (C%fAlgalNT Pl%(TAAfl0N5 Valve length of %J'@er Pipe frOM Contateent hRC General and/or Valve Cont. to Ponetratton Line Line Cesign (SF Ortftce Valve Valve Arraagement(2) Type C Outside % rt*e Isolated Flutd Site, 19 Crtterton System (ll) Pl at e Type (l) loc at ion PLID(81 fest Valves, ft P-Il GCIC furtnne 5 team 4 55 %o FC-V001 GT laside

6/l Yes Ste m Supply 4

FC-V002 GT Qutside Yes 1.0 1 FC V048 GS Inside Yes P 12 Main Ste n st e ri/ 3 55 %o A8-V019 GT Inside 9/J fes Crain mater 3 A8-V040 GT Outside Yes 0.5 P-!) Scare P-14 %ot usef P-15 Sp are P-16 Sp are P-l? 3eactor Pec tre mater 3/4 55 to 65-5V 4110 G8 Inside 10!K Ves .atee Sample 3/4 BS SV 4111 G8 Outside Yes 12.2 P 13 S t aa-Se y Sodium pen. I 1/2 55 Yes BH V029 CK Inside !!/L Yes Liaald taterated 1 1/2 3H V028 SCK Outside Yet 12.6 Jantrol solutton 1 1/2 BH-V054 SCK Outside Yes 10.6 F 10 acctre Ps a .ater 3/4 55 %o 88 V043 CK Instde 12/K Seal eater 3/ 4 8F V098 G5 Cutside 15.8 P-23 hectrc P ep mater 3/ 4 55 %o 88-V041 CK Instae 12/K 'e Seal mater 3/ 4 BF-V099 G8 Outstde 23.1 P-21 15! Access Feaetration P 22 Cry. ell Parge Gas 26 56 res GS V001 BF Outside 13/M ves lalet Veat 6 G5 V02) 8F Outside Yes 43.6 26 05 V021 SF Outside Yes 41.6 24 35-V020 af Outside Yes <242 e 4 '.5 - V )34 GT Outside fes 4 35-V005 GT Outside res 10.7 e 23

ry. ell Pu ge G4s 76 56 fes GS-1024 BF Outside 14/4 fes r

Ltlet vent 26 GS V076 BF %tside ves 5.3 2 GS v025 G8 Outstde ves 25.7 4 GS-V002 Gi histde Yes 4 GS v003 GT Outside fes f.4 P.24A ha Ccetatn. .ater 15 56 ves 6C v019 GT Outstde 15/0 fes cat Soray 16 8C-V018 GT Qutside 'es 6.0 fl0'J2115-01/

] r i. i l hE6.216 M b.M*S Esf PENCTRAf!0NS Page 3 et 33 Length of t l Ptpe from Power Cont. to Prieary Secondary hormal $nutdown Post. Failure Contalw ent Valve yee C Outside Mode of Method of valve valve Accident valve Isolation Closu e Power r Pest Valves, ft. Oper et ton ( 3 ) Ac t uat ion ( ll) Pos t t ion ( 4 ) Position (10) Pestttoa(g) Fosttien Stenal(5) Twe.$ Sourc et 6) scars st ?) ( AC motor Manual O C 0 A$ 15 hone hA 0 [s.s tes ves 1.0 AC motor Manual O C 0 A5 !$ home hA e .o.s AC motor Manual C C C A$ 15 hone h4 0 c.t j .Ves AC motor Manual 0 C C A$ !$ 8.0.t.F.G.L 30 A t ves .Ves 0.5 AC motor Manual 0 C C A5 !$ 8.0.t.F.G E 30 0 t lj. m 1 e ) $pring Man al 5 e C C C A.8 l$ a t ( ,Ves u

tes 12.2

$prin; Manual go C C C A8 it O t 1 na h4 s Flow hoat C C C 44 Ves tes 12.6 Flom Manual (13) 0 0 0 C home ha A s Yes 10.6 Flow Manual (13) 0 0 0 C hone h4 0 s ha h4 t Flow hone O C C h4 .Ves 15.8 AC motor Manual 0 C C A$ !$ M.E 4$ D t Flow hone O C C hA

A hA t

Yes 23.7 AC motor Manual 0 C C A$ 11 M.A 4$ 0 t s $pring Manual C C C C A.N.I l$ A t.v.s Ves tes 43.6 5pelng Manual C C C C A.M,1 1% U t.s its 43.6 $pring Manual C C C C A.M 1 I$ O t.s tes (242 Sp-teg Manual C C C C A.M.! l$ D t.s AC motor Manual C 0 C A$ li A.M.I as s . s. s Ves tes 10.7 AC motor Manual C 0 C A$ !$ A.M.I 45 0 s.s $pring Manual C C C C A.M.I l$ A t.s tes tes 1.3 $pelng Manual C C C C A.M.I Il 0 t.s tes 25.7 Spring Manual C C C C A.M,1 l$ 0 f.s FC motor Manual C 0 C A$ 15 A.M.I 4$ A s.: Ves

tes 7.4 AC motor Manual C

0 C A$ IS A.M.I a$ C ss AC motor Manual C 0 C A$ l$ hoae ha 8 s Tes tes 6.0 AC motor Manual C 0 C A$ 15 hone h4 e s i Waewet la vieso l l l L.

HCGS FSAR 1/H6 {- TABLE 6.2+24 (cor.t ) Page J of 17 Inboarf1 Isolation ~ ' Outtoarti Isolatior Penet PSID Test Barrier Descriptsor/ furrier Dr'Mertption/ T Valve Numtw'r - Flot en valve trueber Notes Numt er Nu*ber

System Description

y i P 29E M-59 Instr gas to drywell C F L-V0 29 3 FL-VU27 P 29 M-13 RACS supply C ED-V020 ED-V019 ED-V021 P 30 M-13 RACS return C ED-V022 P 31 M-15 Breathing air C FG-VOI6 4 KC-VO le e P 32 Spa re A P 33 Spare A P 34A M-59 Prot:e guide tube C SE-V026 12 SE-V0 21 1,10 P 3er M-59 Probe qui:le tube .C SE-V027-12 SE-V022 7,1ts P jac M-59 Probe guide tute C SE-V029 12 SE-V02J 7,10 F 1 j g : M ) P 34D M-51 Pro g tabe C SE-V029 12 SE-V024 t,10 i m 1 P 34E M-59 Probe quiSe tube C SE-V030 12 S E-VU 2) 7,10 ? 3tf M-59 Tip purge system C SE-V006 sE-Vu0e l I 2 P 34G Spare A P 35A-D M-47 CR3 insert A BF-V139 7,13 (Typical of 185 HOUs) BT-IV-1Je-7.13 BFM123 7,13 9 120 7,1J N P 3EP-3 M-47 050 withdrawal A tr-rV-121 7,12,13 [ (Typical of 195 HC39) Br g 122 7,12,13 or 121 7,12,11 [ i P 3BA M-97 Chilled water to drywell C GB-V083 e GB 'J0F0 coolers n GB-PSV-9%22M 7,1F I P 39e M-87 chilled water from drywell C GP-V094 e CB-Ve71 coolers A GE-PSV-9523B 7,17 RL-V002 P 39 M-59 Irstrument g4s suctior C FL-V001 ptL-VU49 } P 23' M-55 EPCI turbine cubauet C (W) TD-V006 e,12 FD-Voce 7 I FD-V007 8,12 l Ame, omen t 14 t

I HCGS TSAR a. Two 100%-capacity air handling units, including low and high efficiency filters, fanc, chilled water cooling coils, and electric heating coils, are provided for use during normal operation or following an accident. Electric pan humidifiers are provided for use during normal operation. b. Two 100%-capacity return air fans are provided for use during normal plant operation ar.J foilowing a design basis accident (DDA). c. Two 1004-capacity emergency air filtration units are provided for use following an accidont. Each unit har its own fan, loy effictency prefi.iters, elefer2e heatino coils, upstream high efficienc'y particulate air (HEPA) filters, charccal adsorbers, and dowr. stream iiEPA l filters. d, An exhaust fan is provided to exhaust air from the i control room, and toilet facilities durinc normal i eperntion. i Two separ~ te 'outside air intakeskete provided for use l WITH PLr+1dji~ e. a l 7pgygg A during normal operati=on or following a DBA, CunOYni HIS$ilB ! EE f. The dar. puts for control rocm isolation purposes are i bubble-tight with a closure time of 5 secondu maximum. 4 6.4.2.3 LeaktichtncEg Control r6cm enveInpe construction joint 3 and penetrations for cabic, pipe, HVAC doct, HVAC equipment, dampers, and doors are designed cpecifically fer leaktightness. 6 list of potential leak pathn to the control room As provided in Table 6.4-1, along with the type of material, jpirt, or penetration. Periodic tests to verify control room leaktightnes's are discussed in Section 6.4.5. l The cpnirol room envel ed A r_an outicakage rate j,of leord ban 1000 cf* p e_tfi gonnty ositive prensuref @r A p Mggit ge relative Eo the 'controT roca at 3acenl. ateas. he 1070 cfm _cgj((. h leakage rate is equivalent to 1.1 control roem volume changes per r .w l (g I 6.4-5 l +- _ _. ,w yy

HCGS YSAR 1/85 6.4.2.5 Shieldino Desian Control room shieldino is discussed in Sectica 12.3.2. Section 12.3.2 iden'tiftes the radiation sources shfelding zones, and shielding thickness. 6.5.5 SYSTEH OPERATIONAL PROCCDURES 6.4.3.1 Normal Ooeration During normal plant operation, the control room supply (CRS system, control room return air (CRRA) fan, and control area exhaust (CAE*, system maintain the design conditions in the control room envelepe. The CRS/CRRA system consists of redundant, 1005.-capacity units, each supplied by a separate Class 1E power cource. Ea0h CPS system consists of an outside air intake 1 cover and plenum, tornado damper, smoke detector, f radiation censpr, radiation monitor, outside air intake isolation dampers, ASHRAE 55% dust spot low efficiency prefilters, 80% to 85% ASHRAE duct spot high efficiency filters, humidifier, electric heating coil, supply tan, and chilled water coil. See Section 9.2.7 for a description of the chilled water system ductwork, controls, and monitcring. A fixed amount of outside /q[ air, 3000 cim, is provided to r,atisfy ventilation, exhaust, and prescurization requirements. The p.citrol Lpom envelope in maintained at _ ~ WTATEF07+TMEt pos i t i v e pr en s u r ef, w i t h respect to its ad3ncent areas during normal plant operaElon, by ~21~ supplying more air to than exhausting from the control room CF AT, envelope. Air is returned from the air conditioned space through hMOT IA the CR,".A fan to the CRS system supply unit. The excess air ir. RM14? ) 6Ags exhausted outdoors by the CAE f an and by exfiltration. 6.4.3.2 P_ost-Accident Operation The control room envelope heating, ventilation and air conditioning UNAC) Systems are designed to ensure habitability during any design basis radiological accident. Upon receipt of a reactor vessel low water leve) (L1)/high drywell prespute signal from a loss of coolant accident (LOCA), or high radioactivity in the cutside air intake, the norcal outside air intake isolation damper for the CRS system closes, the CAE fan stops, the CAE exhoust { isolation darpers close, and the control room emergency filter (CREF) train connected to the operating CRS unit starts automatically to filter the 1000 cfm emergency outside air intake l for room pressurication, when the mode r. witch is in the outdoor air position. The redundant CREF train remains on standby. The CREF ( 6.4-7 Amendment 9

HCGS FSAR / \\ 6.7 MAIN STEAM ISOLATION VALVE SEALING SYSTEM DED64 The main steam isolation valve (MSIV) sealing system limits the BASG leakage of fission products through the MSIVs following a\\ loss-petitwgJr of-coolant accident NLOCA). This is accomplished by pressurizing I NbE-the sections of the main steam lines between the inboard MSIVs AK i and the outboard MSIVs, and between the outboard MSIVs and the main steam stop valves (MSSVs) to a pressure above that of the DB A -- reactor pressure vessel (RPV). Leakage through the inboard MSIVs into the primary containment continues throughout the post-accident period, but does not jeopardize primary containment integrity, as discussed in Section 6.2.5. 6.7.1 DESIGN BASES 6.7.1.1 Safety Criteria The following criteria represent system design, safety, and performance requirements imposed on the MSIV sealing system:* a. The MSlv sealing system is designed with sufficient capacity and capability to limit the leakage from the main steam lines for as long as postulated accident conditions require primary containment integrity to be maintained. b. The MSIV sealing system conforms to Seismic Category I requirements. c. The MSIV sealing system is capable of performing its safety function considering effects resulting from a [DBA9 r LOCA, including missiles that may result from equipment failures, dynamic effects associated with pipe whip and jet forces, and normal operating and accident-caused local envice>nmental conditions consistent with the design basis accident (DBA). d. The MSIV sealing system is capable of performing its p,ggggg safety function;

: nt': Lerm and an assumed single j 0 4 A M CA active Tailure, including failure of any one of the MSIVs to close.

e 6.7-1

e HCGS FSAR ( e. The MSIV sealing system is designed so that the integrity or operability of the main steam lines or the MSIVs is not affected by a single active component failure. f. The MSIV sealing system is capable of performing its safety function following a loss of offsite power (LOP) coincident with apLOCA. g. The MSIV sealing system is remote-manually actuated and MpqhxHnEY', designed to permit actuation within#20 minutes after a LOCA. This time period is consistent with loading ,f IDBA-; requirements of the Class IE electrical buses and reasonable times for operator action. h. The MSIV sealing system controls include interlocks to prevent inadvertent operation of the MSIV sealing system. In particular, interlocks are provided to prevent multiple valve openings that would result in blowing high pressure steam to the building volume whenever the pressure in the connecting main steam lines exceeds the MSIV sealing system initiating i pressure. 1. The MSIV sealing system, including instrumentation and circuits necessary for the functioning of the system, is designed in accordance with standards applicable to an engineered safety feature (ESF) system. j. The MSIV sealing system is designed to permit testing of the operability of controls and actuating devices during power operation to the extent practical, and to permit testing of the complete functioning of the system during plant shutdowns. k. The MSIV sealing system is designed so that any effects resulting from the use of air or nitrogen as a sealing medium will not affect the structural integrity or operability of the main steam lines or MSIVs. / 6.7-2

HCGS FSAR r k, 6.7.1.2 Regulatory Acceptance The piping and components of the inboard MSIV sealing system from the main steam line drain line connection, to and including the inboard MSIV sealing system isolation valve, are Quality Group A, as supplemented by Appendix A of Regulatory Guide 1.96. The rest of the piping and components are Quality Group B, with the exception of the test line from valves HV-6055A and HV-6055B to valve HV-6057, which is Quality Group D. The piping and components of the outboard MSIV sealing system are classified as Quality Group B. The portion of the primary containment instrument gas system, which supplies sealing gas to the MSIV sealing system, is a Quality Group B. All applicable codes and addenda used in the design of the MSIV sealing system are discussed in Section 3.2. The overall system conforms to Regulatory Guide 1.96, as discussed in Section 1.8. 6.7.1.3 Leakace Rate Reouirements The design features employed with this system are established to reduce the dose rate of radioactive materials released to the environment following a#LOCA. IDBA i imposed on the MSIv sealing system toiLeakage control requirements are i Eliminate the potential for MSIV leakage that would a. otherwise bypass filtration, recirculation, and ventilation (FRVS) system filtration b. Handle technical specification leakage rates I Provide features to allow leakage rate verification c. test at every refueling shutdown d. Allow limited component operability tests during normal operation. t h 6.7-3 1

.s HCGS FSAR The design and operational requirements imposed on the MSIV sealing system relative to the foregoing criteria are established to: Y ~ 94 g pp a. Allow MSIV leakage rates of up to g 5 '- rd M r/ ir :: = main steam lines gygggg) Fbt AU. 5 raat b. Ensure and limit total plant radiation dose impacts below 10 CFR 100 guidelines. 6.7.2 SYSTEM DESCRIPTION The main steam isolation valve (MSIV) sealing system is designed to eliminate the release of fission products through the MSIVs that would bypass filtration, recirculation, and ventilation system (FRVS) filtration after afLOCA. This is accomplished by i pressurizing the sections or tne main steam lines between the inboard and the outboard MSIVs, and between the outboard MSIVs and the main steam stop valves (MSSVs), to a pressure above that of the reactor pressure vessel (RPV). Sealing gas is supplied from two independent primary containment instrument gas receivers. The primary containment instrument gas system components are located in the reactor building and consist of two 100%-capacity compressor trains. The MSIV sealing system piping and instrument diagram (P&ID) is shown on Figure 6.7-1. The primary containment instrument gas system P&ID is shown on Figure 9.3-11. As indicated on Figures 6.7-1 and 9.3-11, two independent subsystems (inboard and outboard) are provided to accomplish the leakage control function. The inboard subsystem receives power from Class IE electrical channel D and the outboard subsystem gets its power from Class 1E electrical channel C. 6.7.2.1 Inboard Subsystem The inboard MSIV sealing system injects gas into the main steam lines between the inboard and the outboard MSIVs. Upon initiation, gas from the gas receiver of the primary containment instrument gas system passes throughaa pressure differential control valve that maintains system pressure 5 psi above the reactor vessel pressure,la flow element,j a motor-6.7-4

HCGS FSAR 5/85 i / operated isolation valve, and then a piping system that divides into four lines, one for each main steam line. The gas then passes through a check valve, motor-operated isolation valve, and finally into the main steam line. Pressure sensors, sensing seal gas line, and reactor vessel pressures are used in regulating the pressure differential between the reactor vessel and the seal gas system at 5 psi. In case of gross leakage through the inboard MSIV, a timer is used for reclosing the pressure differential control valve. A flow indicator in the main control room is used to monitor the flow through the flow element. Pressure sensors are also provided for monitoring the main steam line pressure between the inboard and the outboard MSIVs. The pressure is monitored in the main control room. High or low pressure in th_e main steam line causes an alarm in the main 'I b m SIcontrol room. pu-er u r are provided such that the motor-operated isolation valves connected to each main steam line can not be opened when the main steam line pressure /is above @ psig. TE IW-gggggg Check valves installed in the air supply lines provide backups 'g> for the interlocks. H94 6.7.2.2 Outboard Subsystem The outboard MSIV sealing system connects to the main steam lines between the outboard MSIVs and the MSSVs. The outboard subsystem is similar to that of the inboard MSIV sealing system except for instrumentation. The differences are the two pressure transmitters that monitor the pressure between the outboard MSIVs and the MSSVs. These pressure transmitters are used by the isolation valves of the four branches of the outboard MSIV sealing system for their common interlocks since they are located in the main steam line drain header, which is common to the four main steam lines. 6.7.2.3 System Operation During normal plant operation, the MSIV sealing system is in a standby mode with the gas receivers of the primary containment 6.7-5 Amendment 10

HCGS FSAR 5/85 W SEAub instrument gas system charged to a pressure ofIl05 psig, so that gas supply is available upon demand. ' D6A-ltrAJ INEDI Within\\20 minutes af ter a h:::-ef-:: 1:nt :::ident (LOC?), ,3 r/ ::. p ey nign dryvell precrure an Ucr leu r 2cter e:ter--

Icvci, the system E3Wmanually initiated from the main control inapri roomd af ter the operator verifies that the MSIVs and thfeSSVs are ' wet gg j

A 1 shut and the main steam line pressure has dropped to E ppsig or 050 ~ l less. Upon initiation, and if all p eare satisfied, the ,g4yggggg j motor-operated isolation valves canTbe opened, allowing gas from the primary containment instrument gas receivers to enter the main steam lines. INSEner / 0 The MSIV sealing system is designed based on ahm4Eirer precruro ir une meir stea-line er e psig at tNe time of syster initistic te e maximur precrure of 20 prig-This pre"ents the lifting of th: MS!? dich. The MSI" car be unseated by a back preccure differenti:1 cf 25 ::1-A flow element measures the gas flow through each subsystem and provides a signal to a flow switch and a timer. Whenever the flow exceeds the setpoint of the flow switch, the timer is started. The timer will run for a specified amount of time unless the flow drops below the flow setpoint. If the timer should run out it will close the pressure differential control valve. Thus, grcss leakage from the system is prevented because of this automatic action of the timer. Each individual steam line is capable of being isolated. Either the inboard or the outboard subsystem is capable of performing the leakage control function. Pressures in the system and in the RPV, as well as the gas flow, are monitored in the main control room. The inboard and the outboard MSIV sealing system are remote-macually initiated separately.from the control room. The outbaard subsystem is initiated after the inboard MSIV sealing system has been started. i 6.7-6 Amendment 10 m

.s. INSERT A FOR PAGE 6.7-6 ~ 'I (1) if offsite releases exceed 10 times those allowed by 10CFR20 (i.e. the Alert level of the ECGS), and (2) if the RPV is depres-surized (i.e. below 20 psig), and only INSERT B FOR PAGE 6.7-6 DBA LOCA for breaks 2 ft* and larger in which RPV depressurization will' occur in less than 20 minutes. Hence the maximum containment pressure following this LOCA will be approximately 17.5 psig (Figure 6.2-7) while the minimum pressure can be O psig. N 1 t i _.. ~ _ _

.s HCGS FSAR 8/84 t 6.7.2.4 Ecuionent Reouired The following equipment / components are provided: a. Piping - Process piping is carbon steel pipe throughout and it is designed and constructed to ASME B&PV Code, as discussed in Section 3.2. b. Valves - Motor-operated, air-operated, relief, and check valves c. Instrumentation - The requirements and criteria for the MSIV sealing systet instrumentation are discussed in Chapter 7. The remainder of the piping and components are discussed in Section 9.3.6. 6.7.3 SYSTEM EVALUATION An evaluation of the capability of the main steam isolation valve peg _ (MSIV) sealing system to control the release of radioactivity LoCA from the MSIVs following aT2nc -:f-cccI:nt crident ' Lec.". ' h a s been conducted. The results(of this evaluation are presented in the following sections. 6.7.3.1 Functional Protection Features l The equipment in the two independent subsystems (inboard and outboard) are physically separated. The equipment is designed to operate under the expected environmental conditions appropriate to the equipment location. The MSIV sealing system equipment is protected by separation and barrier from exposure of the system components to internal ID0A. missiles caused by eauipment failure (see Section 3.5.1), pipe breaks, and jet forces caused by the\\LOCA event. Equipment is located in the reactor building, hence the effects of the design I basis recirculation line break and postulated external missiles l would not impact the system ability to function. Furthermore, the primary containment instrument gas system equipment that . supplies gas to the MSIV sealing system is located in the reactor ~ ,q building outside the steam tunnel. 6.7-7 Amendment 7

.\\ HCGS FSAR The use of Class 1E power sources to power the components of the system ensures system operation during loss of offsite power (LOP). 6.7.3.2 Effects of Sinole Active Failures The MSIV sealing system functions following an active component failure (including any one MSIV failure to close) by virtue of two redundant subsystems. The subsystems are independently powered from two separate divisions of the Class 1E power supply. The effects of other failure modes are evaluated in Section 6.7.3.6. 6.7.3.3 Effects of Seismic-Induced Failures The MSIV sealing system is designed to operate during and following the application of Seismic Category I design loads in conjunction with operating loads associated with LOCA. 6.7.3.4 Isolation Provisions Containment isolation is provided by automatic closure of the inboard MSIVs and inboard MSIV sealing system isolation valves. The inboard MSIV sealing system isolation valves have interlocks to prevent their opening when the pressure in the main steam line is greater than-the design rating of the sealing system. One out of two valves satisfies containment isolation requirements. Isolation (separation) of electrical components is discussed in Chapter 8. 6.7.3.5 Leakage Protection Evaluation The MSIV sealing system is designed to limit the release of radioactive materials to the environment following a LOCA. The system accomplishes this function through the use of the equipment described in Section 6.7.2. I (M%bOHATEdj The manual initiation of /the system e carried out withi 20 minutes following:- grc:rene, provided the setpoints of the

'4 pg4.gfjj pemistdMR5N ae*a))are satist ied/

.t is possible, due to MSIV closure / @D CN:STE R6EASGS Ack4 THE ECG 6.7-8 LGVEL.

ll .s HCGS FSAR 5/85 sequence, to have high pressure between the inboard and outboard MSIVs. If such pressure were to exist, there would be no need to activate any portion of the MSIV sealing system when the pressure between valves is higher than the reactor vessel pressure, since ggggg) steam trapped between the valves when they closed would serve as 2F FG4 the sealing medium. Once the pressure has/;:::y ;L%the MSIV AND sealing system M$ activated to effect leakage Control. snsnu f y,urtEs) newAsGS tsAca w EC6 AUg7 If there is gross leakage from the system, a timer closes the

aggi, pressure differential control valve.

The system would then be manually secured by the operator. Thus, the system detects high main steam line pressure and prevents system actuation. It also prevents excessive MSIV or main steam line leakage through the automatic closure of the pressure differential control valve by the timer.

Thus, containment pressurization is precluded.

The dose contribution of the main steam lines, considering the MSIV sealing system, is evaluated in Section 15.6. 6.7.3.6 Failure Mode and Effects Analysis The consequences of component malfunctions are shown in Table 6.7-1. The failure modes and effects analysis of the MSIV sealing system instrumentation is discussed in Section 7.3.2. 6.7.3.7 Influence on Other Safety Features The MSIV sealing system is powered from the_ Class 1E power sources. The load is estimated to be about 40 kW. -(The leakage from the inboard MSIVs is discussed in Section 6.2.5. men e' :=f assumed to leak at a rate of FFP51gefh. In addition, this system, by exhausting leakage steam and exhaust % )3wsp] sregg gases, does not introduce or expose the steam piping or the uNES valves to thermal or mass loadings different from those ARE experienced in normal isolation valve service and therefore cannot affect or degrade the sealing ability of the HSIVs. 6.7-9 Amendment 10

n HCGS FSAR 5/85 r

6.7.3.8 Radiological Evaluation The activity released to the environment through the inboard MSIV leakage, and the resulting offsite dose consequences, are evaluated in Section 15.6.5. 6.7.4 INSTRUMENTATION REQUIREMENTS The instrumentation necessary for control and status indication of the main steam isolation valve (MSIV) sealing system is classified as essential. It is designed and qualified, in accordance with applicable IEEE standards, to function under Seismic Category I andfLOCA environmental loading conditions lfTA 1 appropriate to its installation with the control circuits designed to satisfy the mechanical and electrical separation criteria. See Section 7.3.1 for a control and instrumentation description. 6.7.5 INSPECTIONS AND TESTING l Preoperational tests for the MSIV sealing system are discussed in s Chapter 14. During normal plant operation, the main steam isolation valve (MSIV) sealing system can Be tested. A simulated reactor pressure vessel (RPV) pressure signal is introduced from the control equipment room in place of the pressure transmitter output from the RPV. This actuates the pressure differential control valve, admitting gas from the gas receiver. Gas passes through the flow elements, then through the 1-inch test line and back to the primary containment gas compressor suction, or to the drywell when the compressor is not working. This will verify the modulating pressure differential control valve operation, and whether or not the flow element and other components in the test line they are functional. The containment isolation valves are closed during these tests. Tests are conducted for complete functioning of the whole system during extended plant shutdowns or refueling. An actual signal output from the RPV can be used to verify the operability of controls and the actuating devices in the MSIV sealing system. mm 6.7-10 Amendment 10 . o.

? s HCGS FSAR t t With the pressure transmitter isolated from the RPV, overlap testing can be performed after shutdown with the primary pressure still above the setpoint. = Following the performance of the local leak rate test of the main steam lines, a mathematical correction will be made to the sum of the leak rates for all four lines in order to account for the density effects between the test pressure and operational pressure. A density correction must be made since a higher mass flow rate (i.e. higher leakage) exists at the systems maximum operational pressure (25 psig). The LLRT results will be multiplied by a density correction factor of 1.48 (see the calculational summary contained in Reference 6.7-1), and compared to the combined 46 scfh acceptance criteria for all four main steam lines (Technical Specification 3.6.1.2). 6.

7.6 REFERENCES

\\- 6.7-1 Letter from PSE&G (C.A. McNeill) to the NRC'(E. Adensam), " Main Steam Isolation Valve Sealing System", dated March, 1986. - Calculation of the LLRT Correction Factor Utilized When Determining the Leakage Rate Through the Main Steam Lines. i { s j 6.7-11 l I l

IiCGS FSAR TADLE 6.7-1 (cont) Page 2 of 6 Plant Operating Component Failure Effect of Failure Failure Mode Effect of Pailure Male System component Mode on t he System Detection on Plant operation Emergency Individual motur-d. Any one valve The subsystem associated System pressure No effect (LOP or operated tsolatton tails to open with the failed valve will is monitored in LOCA & LOP) valve not f unction, 1,u t the rest the main control of the subsystem can re-room. Low pres-main functional sure is annunciated in the main control room b. Any one valve One out of two valves System pressure No effect f ails to close satisfies isolation by is monitored in on demand virtue of single active the main control component failure criteria. room. Iligh pres-Check valves are installed sure is annunciated as backups for isolation in the main control in addition to interlocks room Emergency Check valve Any one valve is The inboard subsystem can None No effect (LOP or stuck open function as required LOCA & LOP) Emergency Common rotor-a. Fails to open The inboard subsystem System pressure or No effect (LOP or operated isolation will not activate flow is monitored in LOCA 6 LOP) valve the main control room. High or low pressure is annun-ciated in the main control room b. Fails to close one out of two valves satis-System pressure No effect on demand fies isolation by virtue of is monitored in the single actAve comlonent main control room, failure criteria. Check liigh or low pressure valves serve as backups is annunciated for isolation in the main control room C-j a y N

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~ IICGS f SAR TADLE 6.7-1 (cont) Page 3 of 6 Plant Operating Component Failure Effect of Failure Failure Mode Effect of Failure Ma h-System Component Mode On the System Detection on Plant Orw ration Emergency Pressure differ-a. Fails to open The inboard sut system fails System pressure 24 0 effect (LOP or ential control to function and flow rate are IDCA & LOP) valve monitored in the main control room. Low pressure is annunciated in the main control room b. Falls to clow Tlic inboard subsystem will System pressure is flo cf f cct at t imer setpoint be presnurized to a j [monitoredinthe main control room. I maximum pressure ottgg psig and isolatet. l liigh pressure is annunciated in the 1 1 -

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1 Outtoard Subsyntemtaa Emergency Outboard MSIV Any one valve The inboard outsystem asso-The system pressure tio effect (LOP or fails to close ciated with the failed is monitored in the IDCA & IDP) outboat:d MSIV will activate main control room. until the timer runs out. Low pressure is and then isolates. tJot annunciated in the enough pressure will main control room. develop between the inboard MSIV and the MSSV. The subsystem reactivates until proper seal Fressure is established The outboard subsystem can supplement the necessary seal pressure for leakage control Emergency MSSV Any one valve The outboard subsystem System pressure No effect (LOP or fails to close activates but isolates is monitored in LOCA & LOP) af ter the timer times out. the main control The operator can then secure room.. the affected subsystem. The Low pressure is rest of the sulsystem can annunciated in the remain functtonal main control room

i m. IICGS FSAR 8/83 TABLE 6.7-1 ( ccer. ) Page 5 ot 6 Plant Operating Component Failure Ettect of Failure ka11ure Mode Ettact at failure 1' _.0052.__ Svat9D_f2EDQD2tl1. _dQde QU_1AIS_EYD1 W--- LV12E1AQU. QD.MJU1.QMKdtADD Scr ;rn i .rrrrrre reli=+~ Faile +a er _r The e_tb^ rd r ""- _' t e r Errter ci merer 1r Fr c

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HCGS FSAR 12/83 discussed in Table 6.8-2. The system is preoperationally tested in accordance with the requirements of Chapter 14, and periodically tested during power generation in accordance with the requirements of Chapter 16. 6.8.1.5 Instrument Recuiremants The FRVS recirculation units can be actuated manually from the main control room. Each FRVS train is designed to function automatically upon receipt of an ESF system actuation signal. The status of system equipment, including indication of fan motor readiness, filter pressure drops, air temperatures, and air flow rates, is displayed in the main control room during both normal and accident conditions. D6 S44U3/ Tables 6.8-2 and 6.8-5 addresses the extent to which the RGUUED recommendations of Regulatory Guide 1.52 are followed with M 1005 respect to instrumentation. All instrumentation \\is qualified to Seismic Category I requirements. Redundancy and separation of the instrumentation is maintained, equal to the redundancy and separation of the equipment. The following alarms are annunciated in the main control _ room: a. Fan failure b. Heater failure (high humidity alarm) High pressure drop across the upstream HEPA filter bank c. d. High pressure drop across all filter banks (a group alarm) e. Preignition charcoal temperature f. Ignition charcoal temperature 6.8-7 Amendment 3 f

HCGS FSAR 12/83 6.8.2.3 Safety Evaluation The FRVS ventilation system is Seismic Category I design and located in a Seismic Category I, tornado-protected structure. It has redundant automatic startup signals and manual actuation from the main control room. It is normally in a standby mode and is available for testing, inspection, and maintenat:ce. The FRVS ventilation system is designed to preclude direct exfiltration of contaminated air from the reactor building following an accident or abnormal occurrence that could result in high airborne radiation in the reactor building. Equipment is powered from separate Class IE buses, and all power circuits meet IEEE-279 and -308 requirements. Redundant components are provided to ensure that a single failure does not impair or preclude system operation. FRVS failure mode and effect analysis is presented in Table 6.8-5. 6.8.2.4 Tests and Inspections Tests and inspections are described in Table 6.8-6 and Section 9.4.2.4. Conformance with Regulatory Guide 1.52 is discussed in Table 6.8-2. The system is preoperationally tested in accordance with the requirements of Chapter 14, and periodically tested during power generation in accordance with the requirements of Chapter 16. 6.8.2.5 Instrument Requirements Each FRVS ventilation system unit is designed to function automatically upon receipt of an ESF system actuation signal. The FRVS ventilation system can also be actuated manually from the main control room. The status of system equipment, including indication of fan motor readiness, filter pressure drops, air temperature, and air flow rates, is displayed in the main control room during both normal and accident plant conditions. Wd SARTry pgggg) Tables 6.8-2 and 6.8-15 address the extent to which the pgggggg recommendations of Regulatory Guide 1.52 are followed with respect to instrumentation. All instrumentationkis qualified to Seismic Category I requirements. Redundancy and separation of the instrumentation 6.8-13 Amendment 3

HCGS FSAR 82/83 TADLE 6.8-2 FRVS COMPLIANCE WITU HECOMMENDATIONS OF UEGULATORY GUIDE 1.52(aa paq. 3 og 3 Complied With _ ___ DEGGE1P119D 9EJ "lCESDGE.- BUluld19EI.E9Eil190 Y Hil!9 E!:GiIEL83 d119DJYfr125 XsD111d1A90.DIA125 EffEAa 2 SIEigg_ggsigg_ggligg.ig Position a Yes Table 6.8-1 Table 6.8-l Figure 9.4-4 Figure 9.4-5 P.sition b Yes Dwq P-0044-1 Dwq P-0045-1 ais.sile protection tea 114 P-0046-1 separate the zedundant l VfQ(llfEDj units, from each other and t rom ad jacent zotating equipment Pasition c Yes All comporients are All co monents are Seismic Category I Seismic category 1 Position d NA la NA Located outside prAmary containment, but inside the reactor building Position e Yes Tables 6.8-J Tables 6.8-4 and 6.1-1 and 6.1-1 P3sition f Yes 30,000 cfm 9000 cfm Flow rates, each train Position g Yes Reco rded : Recorded: system flow rate system flow rate PD across 1irst PD acrono combined filters pgg g HEPA tilter Bldq PD Alarms: Alarma: SkE 'IIII6 MIbI Section 6.8.1.5 Section 6.8.2.5 TGRRGdi SAFETV Tablu 6. 8-5 Taole 6. ts-5 ggg gg4 Position h Yes Section 7.3 and Section 7.3 and 6.8.1.5 6.8 2.5 Position i WA WA NA No permanent bypass arranqvment installed Position j Yea Sectaon

1. 8, response Sectaoa 1. 0, response to hequlatory Guade 1.52 to hsquiatory Guide 1.52 Position k Yes Section 6. fl. I. 2 Section 6.U.2.2 Positaon i Yes Sectson 6.0.1.4 Section. 4. tl. 2. 4 Amunda,ent 3

HCGS FSAR energized (initiate scram) when trip logic channels Al or A2 and B1 or B2 are both tripped. Sensor logic trip channel inputs to the RPS, causing reactor scram, are discussed in the following paragraphs. 7.2.1.1.1 Neutron Monitoring System Neutron flux is monitored to initiate a reactor scram when predetermined limits are exceeded. Neutron monitoring system (NMS) instrumentation is described in Section 7.6. The NMS sensor channels are part of the NMS and not the RPS; however, the NHS logics are part of the RPS. Each NMS-intermediate range monitor (IRM) logic receives its signal from one IRM channel and each average power range monitor (APRM) logic receives its signal from one APRM channel. The output logics of the APRM and IRM are combined to actuate a logic of one of the four RPS trip logic channels. The NMS logics are arranged so that failure of any one logic cannot prevent the initiation of a high neutron flux trip or high thermal trip. There are eight NHS logics associated with the RPS. Each RPS trip logic channel receives inputs from two NHS logics. INSH T = f For th: i-itial fuel Ic2d, "igh high flur trip inputc frer r2ch reurce rar;r enitor (Sou' are pre"ided te produce e non incid:nt reacter ""S trip Fe!!c"ing the iriti 1 fuel Icadin, this nunccincident trip ic rcr cod. The NMS trip logic contacts for IRM and APRM can be bypassed by selector switches located in the main control room. APRM channels A, C, and E bypasses are controlled by one selector switch and channels B, D, and F bypasses are controlled by a second selector switch. Each selector switch will bypass only one APRM channel at any time. IRM channels A, C, E, and G and channels B, D, F, and H are 4 bypassed in the same manner as the APRM channels. 7.2-3

i INSERT FOR PAGE 7.2-3 A bypass of the noncoincident neutron monitoring trips, i.~e. any SRM, IRM, or APRM, is provided to permit reactor operation whenever the noncoincident neutron monitoring system trip function is not required. The noncoincident neutron monitoring trip function is required only during the initial criticalities with the initial fuel loading, suring core rearrangements, and during certain shutdown margin demonstrations when in the source range of operation (low power). Shorting links installed in the RPS provide this bypass ~ function.

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.s v :. EensOr Esnsor Instrument 71 0 Functicn Instrument Rance l i Peactor vessel high water Level transritter -150 to +60 level H?CI turbine in. v.g. E'PCI turbine er.hsust high Pressure O to 200 pressure transmitter psig HPCI systa: pump high/lew Pressure 30" Hg Vac to-l suction' pressure transmitter 25 psig/ 30* Hg Vac to O psig Eeactor.vassel lov water Level transmitter -150 to +60 level in, w.g. ll Pri:ary containment Pressure O to 10 psig (drywell) high pressure transmitter HPCI pump minitur ficw Flow transmitter 0 to 700 gpm l HPCI system stec: supply Pressure O to 200 psig l low pressure transmitter l HpCI pump discharge flow Flcw indicator 0 to 6000 gp: controller controller Condensate storage tank Level transmitter

10 low level in w.g.

Suppression pool high water Level transmitter 10 level in w.g. Turbine overspeed Electronic NA turbine governor controller T' I ._1 HFt1 Tuxsw evHausr huge o ro ao pse;, TJAPHRA6M H@ Pi%%dRE 'T405Me7 milt ewh h, 6 - A er.0 ent 14 l

HCGS FSAR 1/86 l TABLE 7.4-1 REACTOR CORE ISOLATION COOLING INSTRUMENT RANGES RCIC Function Instrument Rance l lDAWRASHI Turbine exhaust \\high Pressure 0 - 30 psig 8 l pressure transmitter RCIC system pump high/ Pressure (High) 30" Hg vac to low suction pressure transmitter 85 psig (Low) O to 30" Hg Ab. l Reactor vessel high/ low Level (High)-150/0/+60 inches (2)l water level transmitter 0-60 psig ll (Low)-150/0/+60 inches RCIC system steam supply Pressure 0 - 200 psigEEI l low pressure transmitter Turbine overspeed Centrifugal device v 0-1500psigEEE[ l ,~ RCIC system pump Pressure discharge pressure high transmitter gyq IMNIHVi RCIC4 flow Flow trans-0- gpm l mitter W 'Suggcession pesi level Level 110 !!,O' l -t ra ncm i t-t-ee Condensate storage tank Level NA low-low level switch Flow Flow 0 - 700 gpm transmitter = (1) With zero reference 527.5 inches above vessel zero. ) lt:? Lhe plent Technic &l Specificaticnc for inctrumentf See -s e tpc i n t and 0110wabIc >clucc. Tunews evsAccr ur6u h O - aco esis MUR6 MGtn(TTER N Amendment 14 l

HCGS FSAR 11/85 g f. [ } (7 6-(,-' actuated as shown in Table 7".' St Energization of the RRCS ARI valves depressurizes the scram air header independent of the logic and vent valves of the RPS system as shown on Figure 7.6-9. Tne valve are sized to allow insertion of all control rods to begin within 15 seconds. Positive position (open or closed) is indic.ated for all eight (8) RRCS ARI valves at the RRCS control panels. Additional immediate RRCS response to the initiation signals includes recirculation system pump motor breaker trip r \\. 7.6-14b Amendment 13

e HCGS PSAR 01/86 immediately if reactor high pressure is received or 9 seconds after a low water level (L2) signal is received. The high pressure initiation signal will initiate a feedwater runback af ter 25 seconds whether the feedwater pumps are in "outo" or " tr.a n u a l" if the APRM not downscale trip sigr.al is present. Should power not be downscale 230 +5 secenUs from the beginning of the ATWS event, the RWCU system ~will be isolated, and the SLC system will be automatically initiated. Ten minutes after SLC initiation, the RRCS, except for ARI function, can be reset provided that power as measured by the APRMs is dcwnscale, RRCS actuation parameters have reset and the RRCS manual reset pushbutton switches located in the main control room are de-pressed. ) The RRCS is continually checked by a solid state microprocessor-based self-test system. This self-test system checks the RRCS ( sensors, logic, and protective devices and itself). Nuclear boiler system instrumentation is provided to monitor reactor vessel high dome pressure and low RPV water level. The sensors, transducers, and trip units are Class 1E, independent of the RPS, and environmentally qualified to perform their l-protective function during ATWS events. Y The APRMs provide a downscale trip signal to the RRCS permissive logic. This signal is Class lE and contains all available channels of input. Each RRCS channel, except for ARI function, can be manually reset - by denressino the RRCS reset pushbutton switches (four, one for APRM 'each tripped ch nnel) pr vided thathERCS actuation parameters have reset and 10 minutes has elapsed since initiation of the POh6E 5 j i - SLC system. When the RRCS is reset, the following seal-in t osignals are bloken: [ a. 5:cF initiati'r 2r7 RWCU system isolation l 1-b. Low water level ( L2) recircula tion pump tripb ' O tcd l [10. 2

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g hecirculation pump, trip 5(3nd)ffigh reactor pressure) d. feedwater runback g un22;. M 7.6-15 Amendment 14

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l TA?LE 7.6-f, 8 R tCS PE1P LC'JIC PESPONSE -. ~ ~ _ = ~ m - PPCS f r)GIC PE3FOME - After Atter '/5 l 4eemds AnC tecouts Ano' 10 SecMe 2J0 4 5 sec9eds l RPCS Initiati9n IDw htar tevel APM WCt 7tter Ami APP 9 10 Mir utes Ar.d l t Siqnal I wediate 2 Si g l Presr A . teowr.sca l e _IiMttation Not (W.scalc 7Ji'_t_* Gecofids j k i Sic system liit taa*1on keset possir19 Peactor High ARI N Pilr:Mck AFI reeet Pres su re 8ecirc Pump perr.lesive NWCf f i nolo * *tti s1 t t; A t t est itn

totot Trip available 10 mil'ute rige als retve Serrt 30, 25, t.inar P artec cJr.treo 4

ed 210 + 5 sec j timers Sirsyster.inte19tionhebet Peacloc Water' ARI AFI.restet. roesitte I.ow level 2 Recire Pomp permisolve RWCtt isolation 11 anstration c l MJtor Trip available 10 mir.ute sagraLe rave . w l tirer staeted 'cleateo 1 Stert 9, 30, med 430 t 5 see timers l ARI reset SIC nystedt initiat ion Igece-t possible Namial API Initiatton statt 33 and peratssite PW'N agoLation 12 anitiatson 230 t S second svailabla 10 minute signais riave i timers tirer started cleared i 1 .) J r a ,, e, _.. l

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ISHEET 4) ~~- = ' " ' '. ' 544E 190C HOPE CA!EK 1 CENE R ATING STATION 1 F'N AL SAFETY AN ALYSIS REPORT REDUfJDANT REACTIVITY CONTROL SYSTEM IfJITIATION LOGIC _.. T I 9 l FIG'JRE 7.58 'O I 5HEET 4 OF 9 ArmeMmm: 12. 09/B5 I

m A. .a l HCGS FSAR A '\\ (no switch closed) are legitimate. Any other combination of switches is flagged as a fault. If the data passes all of these tests, it is: 2 a. Decoded and transmitted in multiplexed form to the displays in the main control panel b. Loaded into a memory to be read by the computer as required. As soon as the rod's identity is processed, the next rod's identity is generated and processed and so on for all of the rods. When data for all rods has been gathered, the cycle repeats. The RMCS is totally operable from the main control room. Manual operation of individual control rods is possible with a jog switch for control rod insertion, withdrawal, or settle. Rod position indicators, described below, provide the necessary information to ascertain the operating state and position of all control rods. Conditions that prohibit control rod insertion are alarmed with the rod block annunciator. 7.7.1.1.3 RMCS Control Room Displays The rod information display on the reactor control panel is patterned after a top view of,the reactor core. The. display is designed to allow the operator to acquire information rapidly by scanning. Across the face of the full core display are marks indicating the location of the rods in the core. Next to each mark are several indicators that illuminate to indicate specific conditions associated with the rod. These indicators are as follows: Indicator Lamp Color Meanino Q te Rod selected XX-YY Drift Rod is drifting i Accum Amber Accumulator trouble Trip Blue Scram valves have opened Full-in Green Rod is full in Full-out Red Rod is full out During operation, all rods either fully withdrawn or fully inserted are indicated on the full core display with full-in or { 7.7-14

E HCGS FSAR 4/84 stop and control valves close upon loss of control system power or hydraulic pressure. 7.7.1.6.3.5 PRTGS Turbine-Generator to Reactor Protection System Interface The RPS initiates reactor scram when any monitored plant condition requires it. Two such conditions are: main stop valve closure and turbine control valve fast closure when reactor power is above 30% of rated power. The main stop valve closure signal bhil is generated before the main stop valves have closed more than This signal originates from position switches that sense,_, stop-valve motion away from fully open./ :One limit switen Isr 7,n provided for each ot the main stop valves. The switches are closed when the stop valves are fully open, and open within un tT. 10 milliseconds after the setpoint is reached. The switches are 3dNDEI electrically isolated from each other and from other turbine AOE plant equipment. The turbine control valve fast closure signal is generated by four hydraulic oil pressure sensocs that are located on each control valve and that sense hydraulic oil pressure decay as an indication of fast control valve closure. The switches are closed when the valves are open, and open within 30 milliseconds after the control valves start to close in a fast closure mode. To avoid reactor scram due to main stop valve closure or turbine control valve fast closure when power is below 30% of rated power, two independent sensing lines are provided from the turbine shell pressure taps in the high pressure turbine and are connected to pressure switches to supply power level signals to the RPS. The pressure taps are located to provide a pressure signal proportional to turbine steam flow. The pressure taps are shared with other instrumentation sensors. All sensors have individual shutoff valves. 7.7.1.6.3.6 PRTGS Turbine-Generator to Main Steam Isolation System Interface Four independent main condenser vacuum sensors provide an isolation signal to the NSSS main steam isolation valves (MSIVs). Condenser vacuum transmitters and trip units are discussed in Section 7.3.1.1.2. 7.7-45 Amendment 5

HCGS FSAR /k,I are provided with LTD, short time delay (STD), and ground tripping devices. Holded-case circuit breakers provide inverse-time overcurrent and/or instantaneous short-circuit protection for all connected loads in the 480-V MCC. For motor circuits, the molded-case circuit breakers are equipped with an adjustable instantaneous magnetic trip function only. Motor thermal overload protection is provided by thermal overload relays mounted in the starter. The molded case breakers for nonmotor feeder circuits provide thermal inverse-time overcurrent protection and instantaneo jus i short-circuit protection. 4The thermal overload contact offW ImO$hsl safety-related, motor-operated valves (MOVs is connected in series with the starter seal-in contact for manual operation. This overload contact is bypassed either by a signal that requires automatic operation of the MOV or by a contact of the control pushbutton as long as it is pressed./ 8.3.1.1.2.11 Testing of AC System During Power Operation Testing of the ac system during power operation is discussed in Section 8.1.4.21. ~ 8.3.1.1.3 Standby Power Supply The standby power supply for each of the four safety-related load groups consists of one SDG complete with its auxiliaries, which include the cooling water, starting air, lubrication, intake and exhaust, and fuel oil systems. The sizing of the SDGs and the loads assigned among them is such that any combination of three out of four of these SDGs is capable of shutting down the plant safely, maintaining the plant in a safe shutdown condition, and mitigating the consequences of accident conditions. Each SDG is rated at 4430 kw for continuous operation and at 4873 kw for 2 hours of short-time operation in any 24-hour period. The continuous rating of the SDG is based on the maximum total load required at any one time. Each SDG is connected exclusively to its dedicated 4.16-kV Class 1E bus. Each of the four Class 1E power supply channels feed loads in its own dedicated load group. No provisions exist for parallel operation of the SDG of one channel with the SDG of a redundant channel. The SDGs are electrically isolated from each other. Physical separation for fire and missile protection is provided between SDGs by housing them in separate rooms of a Seismic Category I ~ 8.3-13

INSERT FOR PAGE 8.3-13 The thermal overload protection device for various other safety-related MOVs is bypassed continuously and can be temporarily placed in force when the valve motors are undergoing periodic or post maintenance testing. In addition, those thermal overload protection devices that are normally in force during plant operation are bypassed under accident conditions. This thermal overload protection design meets the requirements of Regulatory Guide 1.106, Position C.l. The motor starter circuitry for a limited number of safety-related MOVs has no integral bypass device. For these cases, the trip setpoint of the thermal overload protection device is established high enough that the safety-related action of the MOV is completed. This thermal overload protection design meets the requirements of Regulatory Guide 1.106, Position C.2.

HCGS FSAR 11/85 (' 8.3.2.1.2.3 Class IE Battery Chargers The battery chargers are full-wave, silicon-controlled rectifiers. The chargers are suitable for float-charging their respective lead-calcium batteries. The chargers operate from a 480-V 3-phase, 60-hertz power supply. The chargers are supplied from MCCs of the same channel as the battery system channel it supplies. Battery chargers associated with a battery are capable of supplying the largest combined demand of the various continuous steady-state loads plus charging capacity to restore the battery from the charge state at the completion of their design duty cycle (design minimum charge) to the fully charged state within 12 hours. 1*'seerl 8.3.2.1.2.4 Class lE Battery Loads The loads supplied by each Class lE battery system, along with its length of operation during a loss of all ac power, are shown in Tables 8.3-7 through 8.3-10 anq/ricures 0.2 0 and 0.2 10.1 3_- \\ / lFc,uRG l 9.3-6. i Loads are divided among different battery systems so that each system serves loads that are identical and redundant, are different from but redundant to plant safety, or are backup equipment to the ac-driven equipment. 8.3.2.1.2.5 Separation and Ventilation For each Class IE de system, the battery bank, chargers, and de switchgear are located in separate compartments of the Seismic Category I auxiliary building. The battery compartments are ventilated by a cystem that is designed to preclude the possibility of hydrogen accumulation. Section 9.4 describes the ventilation system in the battery rooms. Redundant de systems are separated to minimize the likelihood of a single hazard causing the loss of more than one channel. (%. 8.3-42 Amendment 13 \\ s

INSERT FOR PAGE 8.3-42 a Tables 8.3-7 through 8.3-10 show various transient and steady state battery loads grouped into specific time increments. The listed load levels are not continuous steady state loads for the entire time period listed. The indicated load levels are maximum current levels experienced during that particular time increment, and some are of a shorter duration than the actual time increment in which they appear. Loads which are not considered'as continuous steady state loads are momentary loads su6h as switchgear control operations, motor operated _ valve operations, motor starting currents and various inrush currents. Momentary loads are supplied from the battery when such loads exceed the maximum output of the battery charger. In addition, inverters are not considered as continuous steady state loads because they are normally supplied from AC power sources and not from the battery charger.

m HCGS FSAR 1/84 Thus sufficient independence and redundancy exists between the Class 1E de systems to ensure performance of minimum safety functions, assuming a single failure. Spare battery chargers are provided to replace any of the Class 1E chargers. Independence of redundant de systems is discussed in Section 8.3.2.2.a. I IMSGET, d. Regulatory Guide 1.41, Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments, March 1973 - The Class 1E de systems have been designed in accordance with Regulatory Guides 1.6 and 1.32, and testing capabilities are provided in accordance with the guidance of Regulatory Guide 1.41. These systems are tested as follows: 1. Testing of the ac power system, including an acceptance test of battery capacity, is performed before unit operation in accordance with the requirements described in Chapter 14. 2. The charger, battery connections, and charger supply are verified for proper assignment of ac load groups. 3. Class 1E de systems are functionally tested along with the associated ac load groups by disconnecting and isolating the other ac load groups, its ac power sources, and the associated de system. Each test includes simulation of an engineered safety features (ESF)' actuation signal, startup of the SDG and the load group under test, sequencing of loads, and the functional performance of the loads. During these tests, the ability of the de system to perform its intended functions, e.g., control of SDGs and Class 1E ac switchgear, is verified. 4. During the testing of the Class 1E de system and i its associated ac load group, the buses and loads of the de systems associated with other ac load 8.3-45 Amendment 4

F-INSERT FOR PAGE 8.3-45 d. Regulatory Guide 1.32, Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants, February 1977 - Position C.l.b states that the capacity of_the battery charger supply should be based on the largest combined demands of the various continuous stehdy-state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the plant during which these demands occur. For the 125V-dc system, HCGS complies with this position by clarifying uninterruptable power supply inverter (UPS) operation with respect to DC bus loading. Following a loss of all AC power, the UPS inverters are powered directly from their respective battery. Upon restoration of AC power, the UPS inverters are powered from the same Class lE AC bus as their associated battery charger. During normal plant operation with AC power available, the UPS inverters are powered from either one of their two associated 480-V (MCC) power supplies. As a result, with respect to battery charger calculated design capacity, the UPS inverters are not included as an electrical load on the DC bus. The design continuous steady-state load is defined as only those electrical loads which are supplied solely by the batteries. Based on the above clarification, HCGS complies with Regulatory Guide 1.32, Position C.l.b. i

/ HCGS FSAR 1/84 (; groups not under test are monitored to verify the absence of voltage, indicating no interconnections of redundant de systems. 4: f Regulatory Guide 1.93, Availability of Electric Power 7 Soucces - Compliance with Regulatory Guide 1.93 is discussed in Section 1.8. f. Regulatory Guide 1.128, Installation Design and gp ' Installation of Large Lead Storage Batteries for Nuclear Power Plants, October 1978 - Compliance with Regulatory Guide 1.128 is discussed in Sections 1.8. h g. Regulatory Guide 1.129, Maintenance, Testing, and p Replacements of Large Lead Storage Batteries for Nuclear Power Plants, February, 1978 - Regulatory Guide 1.129 endorses IEEE 450-1975, with clarifications. Recommended practices of IEEE-450 for maintenance, testing, and replacement of batteries are followed for the Class 1E batteries and are discussed i in Chapter 16. I i f. IEEE 308-1974, IEEE Standard Criteria for Class 1E / Electric Systems for Nuclear Power Stations - The Class IE de system provides de electric power to the Class IE loads and for control and switching of the Class 1E systems. Physical separation and redundancy is provided to prevent the occurrence of common mode failures. The design of the Class 1E de system includes the following: 1. The de system is separated into four independent channels. 2. The safety actions by each group of loads are independent of the safety actions provided by each group's redundant counterparts. 3. Each de subsystem includes power supplies that consist of one battery bank, and one or two battery chargers, as required. 8.3-46 Amendment 4

HCGS FSAR 1/Hb. l FABLE 8.3-1 (cont) Page 2 or to Number Connected To Class 1E Loading Sequenceses Operatirg Distribution System Time Time

Rating, kW Diesel Runep Min From Min From Item Description Equipment No. each,_hp eachtS8 A

C D D No. DBAtass No. LO Pt s s 8 Class 1E toads 16. RB FRVs recirculation system 1A-V213 thru 150 120 2 1 2 1 4 19 s fans 1F-V213 30 stri 17 Control room supply fans 1A-VH203 40 32 1 1 1 30 s 1 30 s 18-VH403 18. 208Y/120-V ac XFMRS to power 10X201,202,203 75(4 4 4 4 4 12 13 s 12 13 s dist panels 204 sets) 10K411,412,413 414 10X421,422,423, 424 10X501,502,503, 504 19. Deleted 20. Intake structure exhaust fans 1 A, B, C, D-V50 4 40 32 1 1 1 1 2 13 sits 8 2 13 s t s ' 8 1 21. Control room chilled water 1A-P400 60 48 1 1 1 65 s 1 b3 s circulating pumps I B-P400 22. Control room supply unit 1A-VH403 90 1 1 1 60 s 1 60 s heating coils 1B-VH403 GEE 2h 23. Control room water chillers 1A-K400 680 506 1 1 1 GEEEE 1 19-K 4 00 MX) s (E) H30 s (4) 24. Diesel generator room recire 1 A-V412 tnru 125 100 2 2 2 2 3 Jo st** 3 30 s syster fins 1H-V412 25. Primary containment instrument 1 A-K20 2 15 12 I' 1 1 10 min 1 30 ntn gas compressor 1B-R202 26. Battery chargers, 250-v de 10D423 12 1 1 2 IJ s 2 13 s 100433 27. Control area battery room 1A-V410 5 4 1 1 1 60 s 1 60 s exhaust fans 19-V410 29. PB FFVS recirculation unit 1A-VH213 tnru 100 2 1 2 1 4 19 s unit heating coils I F-VH 213 30 st** Amendment 14 l

e' u HCGS FSAR 5/85 l TABLE 8.3-1 (cont) Page 4 of 10 Number Connected To Class IE Loading sequencetas Operatir.g Distribution System Tims Time

Rating, kW Diesel Buses Min From Min From Item Description Egiprnent No. each,_hp eachtel A

C B D No. DBA(883 No. IAP(833 Class 1E loads 45. Deleted. l 56. 480-V power supply to Class IE 1AC488, 1BC488 4 1 1 1 1 2 13 s 1 13 s chiller panels 1AC491, 1BC491 47. Traveling screens 1A-S501 5 4 1 1 1 1 3 55 s 3 55 IB-S501 1C-s501 1D-S501 48. ECCS jockey pump 1 A-P228 to 8 1 1 1 1 3 13 s 3 13 s I B-P228 1C-P228 ID-P228 49. Motor-driven diesel generator I A-P4 0 2 2 1.6 1 1 1 1 3 13 s 3 13 s l fuel oil standby pumps 1 B-P40 2 1C-P40 2 1D-P402 50. Standby liquid control pump 1 A-VE261 45 1 1 1 15 min 1 15 min roca duct heaters 1B-VE261 51. 480-V power supply to hydrogen I A-C200 1 1 1 1 13 s 1 13 e and oxygen analyzer panels I B-C200 52. 250-V de battery room duct 10-VE418 10 1 1 13 s 1 13 s l heaters 53. 125-V de diesel area battery 1A-VE420 21 1 1 1 1 3 13 s 1 13 s room duct heaters IB-VE420 4 1C-VE420 1 D-VE420 54. BPCI pump room duct heater 10-VE260 11 1 1 13 e 1 13 s 55. RCIC pump room duct heaters 10-VE259 7 1 1 13 s 1 13 s 56. 250-V de battery room duct 10-VE417 8 1 1 13 s 1 13 s l heater 57. Class 1E panel room water 1A-K403 268 19 8 1 1 1 1 chillers 1B-K403 ggg g Amendment 10 l

HCGS FSAR 01/86 j The filter-demineralizer system also services the torus water cleanup system for the purification of suppression pool water. The stainless steel filter-demineralizer vessels are of the pressure precoat type. A tube nest assembly consisting of the tube sheet, clamping plate, filter elements, and support grid is inserted as a unit between the flanges of the vessel. The filter elements are stainless steel and are mounted vertically in the vessel. Air scour connections are provided below the tube sheet, and vents are provided in the upper head of each vessel. The filter elements are installed and removed througn the top of each vessel. The holding elements are designed to be coated with powdered ion exchange resin as the filtering medium. The fuel pool filter-demineralizers maintain the following effluent water quality specifications: Y $ Iz0 j Specific conductivity at 250C, micromho/cm N30. ! l pH at 250C 6.0 to 7.5 Heavy elements (Fe, Cu, Ni), ppm <0.05 Silica (as SiO,), ppm <0.05 Chloride (as Cl-), ppm <0.02 Total suspended solids, ppm 90% removal to a minimum of 0.01 ppm INRDENT MdD gpugyr The influent and effluent water of the spent fuel pool filter wggg demineralizer is continuously monitored by on-line pH and Agg conductivity instrumentation. In addition, grab samples of the ~aanalyzed weekly for C1, suspended solids, silica, and gamma isotopics, and monthly for the heavy elements. Decontamination factors (df) of > 10 are expected for any Cl-and suspended solids and > 5 for isotopes of I and Co. Resin beds will be regenerated and/or replaced when these df's are not achieved. The spent fuel pool demineralizer will be operated as required to maintain radiation levels on the refueling platform less than 2 mrem /hr. 9.1-31 Amendment 14 l

HCGS FSAR 1/86 isolation from the ventilation system are actuated by a signal from total flooding the carbon dioxide system control panel. 9.5.1.2.11.1 Carbon Dioxide Total Flooding Systems Carbon dioxide total flooding systems are provided for plant areas or equipment, as listed in Table 9.5-2. Low pressure carbon dioxide system operation is initiated by attainment of a fixed high temperature. A temperature sensor initiates the following automatic operation sequence for total flooding: a. A local predischarge alarm is sounded to allow personnel to evacuate the area. The alarm condition is registered on the fire protection status panel in the main control room. \\ Electrothermal devices actuated by \\ the circuit shut all fire dampers, b. After completion of a delayed action timer circuit sequence, the master and selector valves open, releasing carbon dioxide to the hazard area. / Upon completion of the discharge cycle, the timer closes the master and selector valves. The alarm condition is maintained until the system is reset manually. Control pilot valves located outside the hazard area may also be operated manually to activate the system. A supervised 1/4-inch ball valve, with a valve position alarm in the main control room serves as a defeat valve in the pilot line and can be used to deactivate the system when personnel occupy the room. The carbon dioxide storage capacity is adequate to permit two separate discharges within the largest single protected area. The local manual defeat valve for the CO, systems will not be routinely locked due to personnel safety considerations. However, actuation of the manual defeat mechanism alarms in the Main Control Room on the fire protection status panel. Operations or fire brigade personnel will investigate and rectify any unanticipated CO, system defeat. 9.5-19 Amendment 14

HCGS FSAR 09/85 dike area or the sump. However, the dike has sufficient capacity to hold 110 percent of the contents of the fuel oil day tank. Each diesel generator room drains via normally closed isolation valves to a common drainage sump pump basin that has a sump pump capable of discharging 100 gpm. The normal ventilation system can be used for manual smoke ventino. Each supply and return duct is provided with ETL-f# operated fire dampers which are used to isolate the roommeseeed IhMENI the carbon dioxide total flooding system is actuated. ( 9.5-30 Amendment 12 I \\

HCGS FSAR tower blowdown. Off-standard quality water can be recycled to the floor drain collector tanks or to the waste neutralization tanks to be processed in the regenerant wasterbubsystem. If the treated wastes meet the standards forf!condenrate water Mu22 tyrA - / used in the plant, and if the water inventory permits their /kHoduugD] recycle, the processed floor drain waste can be discharged to the CST for plant reuse. 11.2.2.1.3 Regenerant Waste Processing Subsystem The regenerant waste subsystem collects wastes from the regeneration process for the condensate and radwaste demineralizers and the high conductivity drain sumps in the radwaste area of the aux,iliary building and the turbine building. These wastes are collected in the waste neutralizer tanks, where they are neutralized and, if required, buffered with solutions of sodium phosphate before being processed through the waste evaporators for concentration. The distillate resulting from the evaporation process is returned to the waste collector tanks. The waste evaporator concentrate is collected in the concentrated waste tanks for chemical pretreatment and is transferred to the solid waste management system (SWMS) for solidification and offsite disposal. In addition, concentrate is transferred from the decontamination solution concentrated waste tank to the concentrated waste tanks. ]l0Q3F[ 5-11.2.2.1.4 Chemical Waste Processing Subsystem Chemical wastes collected in the chemical waste tank consist of laboratory wastes, decontamination solutions, and sample rack drains. After accumulating in the chemical waste tank, these wastes are neutralized to a pH value of 7 to 10 and, if required, ~ buffered wLth a solution of sodium phosphate, then processed by evaporation through the decontamination solution evaporator. The chemical wastes are normally evaporated to reduce volume. The concentrate is discharged to the decontamination solution concentrated waste tank for radioactive decay, then transferred to the concentrated waste tanks. The vapors produced during this evaporation are sampled and discharged through the south plant vent. When the radioactivity concentration is low, a cross-connection with the floor drain subsystem allows the chemical wastes to be processed through the floor drain filter and demineralizer and then diluted with the cooling tower blowdown prior to discharge to the Delaware River. I 11.2-7

.~. 1 l INSERT FOR PAGE 11.2-7 A bypass of both waste evaporators is provided to allow for. processing of the waste collected in the waste neutralizer tanks (by either the crystallizer or a portable radwaste system) in the event that the waste evaporators are unavailable. 4 i [ l I 1 I 1

c HCGS FSAR 1/84 11.2.2.2.4 Demineralizers The equipment and floor drain demineralizers are mixed-bed type. The strong acid and base mixed-bed resin gel-type (1:1 ratio by volume) is regenerated when the effluent conductivity exceeds a preset conductivity limit, as well as upon high differential pressure. Exhausted resins are sluiced to the radwaste demineralizer regeneration system for resin regeneration and reuse. The carbon steel demineralizer vessels are rubber-lined. Fine mesh strainers are provided in the vessel discharge and in the downstream piping to prevent resin fines from being carried over to the sampling tanks. When the resins cannot effectively be regenerated, they are discharged to the spent resin tank for radioactivity decay and then transferred to the SWMS for solidification and offsite disposal. Each demineralizer vessel is located in a separate shielded room to minimize exposure to personnel during operation and routine caintenance. The demineralizer vessel is designed per ASME B&PV Code, Section VIII, Division I. 11.2.2.2.5 Radwaste Evaporators a. Waste evaporators - Two waste evaporators are arranged in parallel for simultaneous operation as well as backup to each other. The waste evaporators are a forced-circulation, submerged-tube design to minimize scale deposits and have a design feed rate of 40 gpm each. Heating steam is provided from the plant heating steam system. The evaporator bottoms are discharged to a concentrated waste tank from which it is fed to volume reduction equipment in the SWMS. The distillate from the evaporators is recycled to a waste collector tank. A grab sample of the distillate is periodically taken for chemical analysis. WSC.gr r 11.2-10 Amendment 4

INSERT FOR PAGE 11.2-10 A bypass of both waste evaporators is provided to allow -for processing of the waste collected in the waste neutralizer. tanks (by either the crystallizer or a portable radwaste system) in the event that the waste evaporators are unavailable.

r ? i IICCS FSAR 11/95 i TABLE ~1.2-10 (cont) Page 3 of 5 Rated Design Rated Flow itcad TDit Rated Power Pressure / Temp Component Quantity Type (9Pm) (ft) _ (hp) (psi <J/ F) Code f Decontamination solution 2 flor, cent 10 140 2 150/210 Mfg.Std evaporator condensate return pumps Waste evaporator 2 flor, cent 9000 35 150 64/274 Mfg Std recirculation pumps Waste filter aid pump 1 Positive 154 gph, max 150 psig 1 150/250 Mfg Std displacement max Floor drain filter aid 1 Positive 154 gph, max 150 psig 1 150/250 Mfg Std Pump displacement max Chemical addition pump 3 Positive gph, max 100 psig 2 100/200 Mfg Std displacement max hk Radwaste demin regen 2 Positive 90 gph, max 110 psig 1/2 110/200 Mfg Std l acid pump . displacement max Radwaste demin regen 2 Positive 90 gph, max 110 psig 1/2 110/200 Mfg Std caustic pump displacement max waste filter holding 1 Hor, cent 75 50 3 150/220 Mfg Std Pump waste precoat pumps 2 Mor, cent 325 50 7 1/2 150/150 Mfg sto Floor drain filter 1 Nor, cent 75 50 3 150/220 Mfg Std holding pump a 4 Amendment 13

i McGS FSAR TABLE 11.2-10 (cont) Page 5 of 5 -Type Fated Flow Degign Diam / Height Mat er ial E.sch Egalpment Pressure / Temp coer= ment Quantity 'fitt Iype/ Number, J 1get,_,,,g3LangLet _,,,1p3 q/*FI cme Ba3 waste dominera11:er 1 one complete ASME, Sect VI!! regeneratiosi system: regeneration per 7 tours 125/150 ASME, Sect VIII 8 O Catton vessel 1 7' CS SA-515 Gr - g. Vert /cy1 70/ rubber-Div I l lined Anton vessel 1 7' CS SA-515 Cr - 125/150 ASME, Sect Vit!' vert /cyl 70/rubter-Div I lined 150/200 ACME, Sect vt!! caustic dilution 1 s'/1C' CS SA-515 hot water tank Vert /cyl Gr 70 Diw I

FIGURE fl.2- $~ SHser I oF Z EQUIPMENT DR, EQUIPMENT DR AIN COLLECTOR TANK -O vWELL a REACTOR stDo E QUIPME N T D A AIN $UwS A EQUIPMENT -Tu; sine stoo EOuir ORAIN $ UMP -t AsTE CO* CENT.R ATORs CONoENsERs - p2 ORAIN FILTER ~ k -OFF ST AND ARD RECYCLE - KE ACTOR W ATER CLE ANUP SYSTEM - CLE AN ur PH Ast SEP AR ATOR DECANT - LOW CONouCTivtTV RINSE COND DEMIN EOulPMENT - LOW CONouCTtytTY RINSE M ADWASTE DEMIN -cADWAsTE stoo EOuir oRAiN suws DRAIN COLLECTOR -CF F ST AND ARD RECYCLP TANK WASTE SLUDGF WASTE SURGE TANK ai FLOOR DR AIN r COLLECTOR TANK FLOOR DRAIN - CONoENsATE sTOR Act TAmit R ADW AsTE ARE A - FLOOR 0R AIN SUMPS DRYWELL & RE ACTOR BLDG 7 FILTER - FLOOR DRAIN SUMPS FLOOR DR AIN OFF STANDARD -[ECYCLE W ASTE SLUOGE PHASE SEP DECANT - TUR*NE SLOG FLOOR OR 444 $Uw$ RHR SYSTE M FLOOR DRAIN COLLECTO R TANK WASTE SLU' REGENEF WASTE - M-NEUTR ALIZER TANK - TUR8INE BLDG HIGH CONO SUMP R ADW ASTE ARE A HIGH 4 ,VASTE - CONO SUMP W ASTE CONCE NTR ATOR$ W ASTE SURGE T ANR - noWasTE DEMIN REGEN SYSTE M CONDENSATE RE GEN 7.lh EVAPORATORS system WASTE b( NE U T R AllZ E R TANK Ang: INFORMATlW GJ SIREAP.4 MMf4 f ob BYPASS IDeGFIEro By A 4 15 GiWN portcy W TAst6 112-2.

EOUIPMENT DR AIN PROCESSING SYSTEM i6 TO COND EQUIPMENT STORAGE TK DRAIN SAMPLE TANK EEUIPMENT A EQUIPMENT ~ 0; AIN FILTER Q DRAIN DEMIN TO COOLING 2 DRAIN SAMPLE ^g TOWER EOulPMENT BLOWOOWN LINE TANK CASTE SLUDGE TANK FLOOR DR AIN PROCESSING SYSTEM a TO COOLING TOWER B LOW DOWN FLOOR DRAIN SAMPLE TANK (~ LINE ~ + 'r F LOOR DR AIN A FLOOR DR AIN A DEMIN FILTER 1 CONDENSATE FLOOR DRAIN ^ ^ TANK WASTE SLUDGE TANK REGENER ANT WASTE PROCESSING SYSTEM WASTE EVAPORATOR - WASTE (EOUlPMENT DR AIN) 4 DISTILLATE COLLECTOR T ANKS l TANKS f NASTE ~ ' ORATORS HOPE CREEK GENERATING STATION CONCEN TR A TE D A FINAL SAFETY ANALYSIS REPORT W ASTE Ob g - g ) STORAGE TANKS \\ VOLUME REDUCTION & SOLIDIFICATlON SYSTEM LIQUID WAST E M AN AGEMENT PROCESS FLOW DIAGRAM FIGURE 11.2 5 SHEET 1 OF 2 A mentiment 14.01/86

TABLE (1.2-E SHart 2 er 2. CHEMICAL WASTE (DECONTAMIN AT VENTh - LABOR ATORY WASTES -FUEL POOL F/D DR AINS -WASTE FILTER DRAINS - FL DR FILTER DRAINS SOLN WAST TANK r EVAPORATOC DETERGI - PE RSONN EL DECONTAMIN ATION - LAUNDRY DR AINS r -t DETERGENT DETERGENT DETERGEl "A" DRAIN FIL -O Ncrilir : INFORMATIOJ OJ ST'PEAM MWY1M ttEDTIFIED BY h b IS GIVEM IN TAsts 11. 2-2

RR WR:RMATm cWJ ONTAMINATION SOLUTION) PROCESSING SYSTEM VENT DECON SOLUTION DECON CONCENTRATED VOLUME REDUCTION Arc SOLIDIFICAllON SOLN g 3VAPORATOR WASTE TANK DETERGENT WASTE PROCESSING SYSTEM OISCHARGE CHANAL y DETERGENT X DR AIN FILTER ) ) DECON SOLN EVAPORATOR B t HOPE CREEK GE NER ATING STATION FIN AL SAFETY AN ALYSIS REPORT LIQUID WASTE M AN AGEMENT PROCESS FLOW DIAGRAM FIGURE 11.2-5 SHEET 7 OF 2

HCGS FSAR 10/84 asphalt in the extruder / evaporators. The mixture of waste and asphalt is deposited in 55-gallon drums and is allowed to stand, cool and solidify. The drums are then capped, checked for externa.1 contamination, and stored for shipment offsite. Dry trash is sorted and compacted, if appropriate, by a box compactor and stored for shipment offsite. lwaarri

== Piping and instrument diagrams for solid radwaste management are shown on Figures 11.4-1 thrcugh 11.4-9. Layout of the packaging, storage, and shipment areas of the solid waste management system are shown on Figures 1.2-20, 1.2-21, and 1.2-22. Equipment and floor drainage systems are discussed in Section 9.3.3. 11.4.2.2 Resin Slurries The SWMS receives filter media, waste sludge and/or resin slurries from the wa3te sludge phase separator, the cleanup phase separators and the spent resin tank. These slurries are pumped from the colid radioactive waste collection subsystem directly to the centrifuge feed tank on a batch basis. A centrifuge feed tank decant pump is provided to remove excess water once the centrifuge feed tank is filled and, the slurry batch is allowed to settle. The water is pumped to the waste l sludge phase separator. The decant pump will not be used during normal operation when the centrifuge is operable. However, it is provided for use, if desired, during centrifuge bypass operations. Slurries in the centrifuge feed tank are recirculated using a progressing cavity pump. Recirculation together with tank agitation produces a homogeneous mixture. While the slurry is recirculated, the pH is monitored at a local instrument rack. An alarm in the control panel is actuated when the pH deviates from the preset range. Caustic can be added to the recirculation loop to adjust the pH, when necessary. When the slurry batch has reached a homogeneous state, a sample is taken to determine the solids concentration. After the solids concentration has been determined, the operator sets the slurry or centrifuge feed metering pump cpeed on the control panel. For normal operation, a slip stream is taken off of the recirculation loop and metered to the centrifuge. This is done by using a centrifuge feed metering pump. The rate of metering ( 11.4-4 Amendment 8 i

INSERT FOR PAGE 11.4-4 A bypass of the SWMS is provided upstream of the cryutallizer and centrifuge feed tank. This bypass enables the processing of solid wastes by a portable system, as discussed in Section 11.4.2.6, should the SWMS become unavailable for any reason.

~ HCGS FSAR 4/84 to the centr.fuge is initially set by the operator at the process l control panel, cepending on the slurry solids concentration. Tne centrifuge then separates the catrier water from the resin / filter media sludge and returns the water by gravity drain back to the waste sludge phase separator. The remaining waste solids are discharged with approximately 60% by weight moisture to extruder / evaporator "A" via a vertical chute. While normal operation consists of the centrifuge feeding to extruder / evaporator "A" as described above, both extruder / evaporators can be fed slurries of sludges and resins. Slurry feeding is intended to be used as a backup process scheme. In the event that the centrifuge feed system is unavailable, the slurry stream can be fed directly to both extruder / evaporators by taking a slip stream from the recirculation loop slurry metering pumps. Sampling is performed to establish the feed rate. (N$RT} 11.4.2.3 Concentrates The waste crystallizer receives concentrated regeneration and miscellaneous chemical waste on a batch basis directly from the concentrated waste tanks located in the liquid waste management system (LWMS). The concentrated waste pumps in the LWMS are used to feed concentrates into the crystallizer recirculation pump suction piping. The concentrates are concentrated further in the crystallizer to a slurry of approximately 50S. by weight Na,SO.. The concentrates are recirculated by the crystallizer recirculation pump through the crystallizer heater and into the crystallizer vapor body. The heated concentrates entering the vapor body flash, and part of the liquid evaporates. The vapor leaving the vapor body passes through the crystallizer entrainment separator. In the normal operating mode, the vapor is condensed and returned to the liquid radioactivo waste system upon meeting quality requirements. In the alternate, accelerated processing mode, the cryctallizer mechanical vapor con. pressor is used to compress the vapor from the separator to the conocncina pressure of the crystallicer heater. This compressed vapot It used as the primary heating medium in the crystallizer heater, where as the normal operating mode uses only plant heating steam. Upon reaching 50% by weight Na,SO. concentration, the concentrated waste is discharged to the crystallizer bottoms tank i for storage until the initiation of the solidification process. s Prior to solidification and while circulating, the pH of the 11.4-5 Amendment 5 a

HCGS FSAR 01/86 bottoms batch is monitored and may be adjusted with NaOH from the caustic addition skid. The contents of the bcttoms tank are maintained at approximately 1800F by electrical strip heaters attached to the bottom head of the tank. Except f or the vapor body recirculation pipe, all pipe carrying crystalliter concentrates are maintained at 1800F by electrical heat tracing. l Maintenance of this temperature prevents dissolved solids in the concentrate fror crystallizing. The bottoms tank recirculation pumps provide a stream of 50V concentr.ates to the concentrate metering pump (s) for cne or both extruder / evaporators where the remaining water is evaporated while the waste is mixed with asphalt. [WSWT 11.4.2.4 Solidification, Packacino, and Drum Handlino 11.4.2.4.1 General Extruder / evaporator "A" receives dry cake discharge from the centrifuge or slurry feed from the centrifuge feed tank as discussed in Section 11.4.2.2 or concentrates from the crystallizer bottoms tank as discussed in Section 11.4.2.3. Extruder / evaporator "B" receives slurry feed from the centrifuge feed tank as discussed in Section 11.4.2.2, and concentrates from the crystallizer bottoms tank as discussed in Section 11.4.2.3. The extruder / evaporators mix the waste streams with asphalt at approximately I gpm and 3250F. At this temperature, all l renaining water is evaporated. The extruder / evaporators through their kneeding and mixing action also compact the waste and csphalt, producing a denser product. The waste and asphalt mixture is deposited into a 55-gallon drum. The solid radwaste monorail hoist places empty drums on the turntablos, which position the drums for filling under the oxtruder/ evaporator discharge ports. The same monorail hoist removes filled drums from the turntables. The filled drums are placed on a conveyor and guided to a capping / swipe station. At this location, the drums are caretd and swiped. Radiation readings are also te The drums are then conveyed to the truck bay where they are...her loaded into a cask for chipping or placed in the temporary storage area located in the north part of the auxiliary building by the storage area bridge crane. t 11.4-6 Amendment 14 1

t INSERT FOR PAGE 11.4-5 A bypass of the SWMS is provided upsteam of the centrifuge feed tank. This bypass enables the processing of resins by a portable system, as discussed in Section 11.4.2.6, should the downstream portion of the SWMS become unavailable for any reason. INSERT FOR PAGE 11.4-6 A bypass of the SWMS is provided upstream of the crystallizer. This bypass enables the processing of concentrates by a portable system, ac discussed in Section 11.4.2.6, should the downstream portion of the SWMS become unavailabic for any reason, t

HCGS FSAR \\ Six CCTV cameras provide visual access to the following areas: E _ % )oan, tilt, and room camerag for viewing W a. Tr n e r arr er " conveyor 7 pg c[. Two fixed-focus cameras, one in each vent hood for viewing container filling d g. one pan, tilt, and zoom camera mounted on the overhead crane for viewing storage area and truck bay c[. One fixed-focus camera with crosshair indicator mounted on overhead crane for viewing target plates to assist in drum grab alignment and drum placement in storage ff. Four CCTV monitors with selector controls to be viewed / on any monitor. 6 All cameras are radiation-hardened and all camera lense 6 are nor. browning. 11,4.2.4.9 Electrical / Control System The entire volume reduction and solidification process is controlled remotely from the radwaste control room. This ellows I the operator to perform the processing and drum handling functions from a single low radiation area. System controls provide manual start and automatic stop for major processing functions. Signals from strategic temperature ar.d flow points are cor,nected to o'larms and safety interlocks. Alartt.s indicate off-design conditionc, and any predetermined deviction will automatically shut down the process. This includes shutdown of the system for i lack of sufficient asphalt flow to the extruder / evaporator, A separate drum handling console is provided to centrol all drurr handling operations, exclusive of capping and swiping functions. l Controls for the overhead crane, monorail, and CCTV system, as we)) ac four CCTV monitors, are included in this conso3e, ^ g L 11.4-11 ~q e- ~ e

HCGS FSAR 01/86 In addition, a remote drum capping panel is provided to allow local operation of the drum conveyor and capper. 11.4.2.5 Trash The low-level dry waste materials, i.e. clothing, plastics, and HEPA filters, are processed by a hydraulically-operated box compactor. Containers of approxitately 100-cubic feet made of metal and/or plywood lined with galvantzed steel are used for storing and shipping the compacted trash. The box compactor is equipped with an external HEPA filtration system to provide a negative air flow into the container during the compacting operatter. Noncompactible trash, i.e., tools and compenents, is packaged in a suitable-sized container which meets DOT requirements. e }).4.3 References 11.4-1 Werner & Pfleiderer, Radwaste Volume Reduction and I Solidification System, Topical Report No. WPC-VRS-001, Revision 1, May 1978. i 11.4-2 Public Service Electric & Gas Company Hope Creek f Generating Station, Pcocess Control Program, 4 Revision 0, July 1985 (as Bubmitted to W. Butlet-(NRC) from R.L. Mittl (PSE&G) in a letter dated August 21, 1985). ( 11.4-12 AmendT.ent 14 w=

~ INSERT'FOR PAGE 11.4-11 'b. Two fixed-focus cameras-with wide angle lenses for n viewing the fill stations. INSERT FOR PAGE 11.4-12 11.4.2.6 Porte.ble Dewatering / Solidification System Permanent flanged connections are provided on the-south wall of'the RWMS truck bay to enable processing of concentrates, filter media, waste sludge and/or resin slurries by'a portable dewatering / solidification system. This provides maximum system flexibility and minimizes radiation exposure in the event that key portions of the SWMS become unavailable 'for any reason. The following.SWMS flanged connections are provided in truck bay: Concentrates feed Resin / sludge feed Decant return Condensate return Service air supply Vent filter connection Space is provided outside of the truck bay,-(to the west) for a temporary control panel with an adjacent 480 VAC pcwsr supply connection. Check valvoa are supplied in all feed /suppiy lines.

CRYSTALLilER DISitLL ATE PunaPS /\\ 0 A-7-776. 08-7-774 QF y' nASTE COLLECTOR TANa$ ) OA-T -302 08-T-302 CRYST ALLilE R 100-E -M23 & 3 ) CONCENTRATED WASTE 1r ASSOCIATED WOtuast y TANKS OA -T-324 REDUCTION EOuiPastNT On - T - 12 4 [ 3 2 \\ SOTTOMS T ANK FROM LIOulO RADWASTE FLOW osacRAas FIGURE 1 2-5 C R YS TAL LilE R 00- T-M F ) RECtRCULATIO90 ~ CRv5 FALL 12ER807T4 CLE ANUP SACKW ASH T ANE RECBRCULATet RECE RVING TANK PUm8PS C A-#-2TT 10- T -21 F OS -P -JF T CL E ANUP BACRW ASH S E T RAasSFE R PUnaPS h 1 A 214,15-7-214 T": C A WASTE StuOGE 2 't PM A5E SE P AR A TOR CENTRefucE FEED C d YANa OECANT.Pumar 00-T-sie ) 0~- g W ASTE SLUDGE DISCHARGE hos n8seG Puee 00-7-322 FLOOR ORAa4 COLLECTOR 1r TANa$ OA-T 307 co - T-mf ( CENTRIFUGE FEEo T ANE 00-T-294 SLURRY ME 7 Puesp OA-7 W ASTE SLUDGE DECANTPUMP ) CL E ANuP PM ASE 00-7-323 SEPARATOR ~- CENTRIFUGE FEED mez T ANE RECIRCUL.A. TION [ OA-T-316 PUMP 00-P. ) stuRRv ME TC Puesp Os-P. CtE ANuP Pw ASE Sg*R]O," CLEANUP (ENTRepuCE FEEc ut TE RiNo C E N T,R I SLUDGE rumps OA-7-wF Oe-r-mF og,, DISCH A RG E [ MIXING PUMP WASTE COLLECTOR T ANu$ O A-T -XI2. 05-7-302 g CEE ANUP DECANT PUtdP OM-391 SPENT ME $14 TANK 00- T - 32 3 ) SPE N 7 m E $1M PutsP 00-7-317 NossCOas4USTIOLE WASTE TOOT $ RUset R APPARE L SOE Cone ACTOR 100 F T 3 cL As3 Fit TE RS 00-S-907 CONTAINERS PvC EOutPMENT s COMPACTOR FILTR ATION SYSTEM 00 VH 319 i w sw 9

~

TOR TANKS CONCENTR ATE MET E RING De -T -302 PUMP 5 OA-P-E t OC-f-381 H

f1 CONCENTRATE METERING - Q' PUMPS OS-F-M 1. 00-P-381 hd 71 CCTST ALL12E R BOTTOMS T ANK RECIRCULATION PUMPS OA-P-377 08-9 371 VENT HOOD O A-4 -34 E X TRUDE R/E V APOR A TOR SS-G ALL ON OA-5-20 e# DRUME SLURRV ME TERING n 6 W OA P-3BS h e - FLOOR [ ,.e.- DRAIN yeny HOOD SLUM TV METE RING OS 4-M PUMP 0e-9-3n3 CET TRIPUGE E xTRuCE R/EvaTOR ATOR e SS-CALLON 00-8-3ba On-s-360 ORunes b A

FLOOR DRAIN E XTR UDE R/E V APOR ATOR i,

VENT HOOD FILTER OB-F.333 E NThuOf RI E V APOR ATE R VE NT MOOD BLCWWE R. E M T RUOE R/E V APORAT04 C VENT HOOO FILTE R Q OA. - :: EXTRUDER / EVAPOR ATE A VENT HOOO stourtR NOTE SINFORM ATION ON STRE AM Ag NUneERs sOENitesEO ev A O es newf N iN T AaLE is e-T h l HOPE CREEK LTION NORM GENER ATING STATION PLANT VENT FINAL SAFETY ANALYSIS REPORT SOLID WASTE MANAGEMENT SYSTEM

2. CD4ECTIOiJS RR A TORTA8tB IS-PROCESS FLOW DIAGRAM WhTERIM(3ff.0LJbtVI TION GVSTEH N WO M N NN2b -

FIGURE 11.410 Amendment 5,4/84

HCGS FSAR 01/86 N. typical media, use of which is also not considered necessary to verify conformance to the design. m. Appendix A, Paragraphs 1.m (4) and 1.0 (1) reference Regulatory Guide 1.104. Comment: Regulatory Guide 1.104 was withdrawn by the NRC on 8/22/79. During preoperational testing, the cranes will be verified to function in accordance with specifications. The controls, interlocks, and travel limits of the reactor building and fuel handling cranes are verified. n. Appendix A, Paragraph 1.n (11) references Regulatory Guide 1.80 1.68.3). Refer to 14.2.13.5 for comments. o. Appendix A, Paragraph 2.c, concerns functional testing of the reactor protection system. Comment: The reactor protection system will be functionally checkedf prior tot =uci Icac using station surveillance and calibration procedures. The reactor i p g protection system is shown to operate in conjunction MAL 7gue,cg, with the control rod drive startup test, described in El#0AWD' %g,cgg .Section 14.2.12.1.8. Also, the reactor protection system is verified to operate following scheduled transient tests such as MSIV isolation and turbine tripk /GENERATtR LCAD gggenya p. Appendix A, Paragraphs 2.d and 5.0, concerns leakage detection of the reactor coolant system. Comment: Y "crticas c Leak Deteet-len-governed - by Techni-ce+ f Specificetions will be funct-lona44y-ehecked-just-pr4er to fuel leading ucing sta4Lon-sur-ve411:nce and J calibration proccdurec./ Setpoints related to Leak Detection high steam flow in HPCI and RCIC are verified and set as stated in Sections 14.2.12.3.12 and 14.2.12.3.13. Normal operation of leak detection systems, such as drywell equipment drain sump pump will be accomplished using station operating procedures. 14.2-202 Amendment 14

HCGS FSAR 1/84 ~ ( '. s Although there will be no startup test procedure designated hotwell level control, operation of the hotwell level control system will be verified using station operating procedures and monitoring hotwell level during Phase III startup testing. w. Appendix A, Paragraph 5.g.g concerns the testing to determine the operability of equipment provided for ATWS. Comment: The ATWS subsystems are thoroughly checked out logically and functionally during the preoperational test program, as described in Sections 14.2.12.1.2.c.6,/' 14.2.12.1.8.c.9, 14.2.12.1.3.c.3, 14.2.12.1.4.c.4, 14.2.12.1.9.c.7, and 14.2.12.1.10.c.4. / Portions of-ATWS go>erned by Techtvka1 Specificat-ionc s'i11 be f u n c t i o n a l l y c h e c k c d j u s t - pr-ior--t-o-f+el-load - u s i n g ctation sur;cillance anc calibratier proceducee. Add i t i onal-4y, lRPT pump trips, which are ATWS related, are accomplished during Phase III testing, as discussed in Section 14.2.12.3.28.c. x. Appendix A, Paragraph 5.i.i concerns reactor coolant flow control valve closure. Comment: HCGS design does not incorporate the recirculation flow control valve; however, the runback of the reactor recirculation pumps for cavitation protection and loss of feedwater pump is accomplished during Phase III testing, as discussed in Section 14.2.12.3.28.c. 14.2.13.2 SRP II.b, Reculatory Guide 1.20, Revision 2, May 1976: Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testina HCGS complies with Regulatory Guide 1.20, with the clarification that the HCGS reactor internals were tested in accordance with the provisions for nonprototype Seismic Category I plants. The results of the vibration assessment program are found in GE Licensing Topical Report NEDE-24057. For discussion of the preoperational flow test and inspection (, program, see Section 3.9.2.6. 14.2-204 Amendment 4

1 TEST OPETJ HEAT !JO. TEST !WT VESSEL UP 1 2 3 4 5 6 ( 22) ! 1 Chenical and Ibdiochenical X X X X X 2 Fadiation Nasurenent X X X 3 Etel loadim X 4 Eull Qare Shutdown Paruin l 5 Control Icd Irive X X X( 2) X( 2) x( i l X( 2) 6l SI'! Nrfcc anct X l X 8 l; IFri Mrforrance 1 91 LPR'i Calibration X X X X 10 APR'i Calibration X X X X X X 11 Process Caguter X X XI X X 12 FCIC X X 13 EPCI X X 14 Selected Process 'Iup M X X(4) h L 14 Water IcVel mf Irg 'Itstp I X X X MP 15 Systen Expansion X X X +-- X us

1: L:=:- t a in tv-1 i

17 Cbre Ierformance X X X X X X 18 Stean Prtduction X 19 Cbre Pwr-Nbid Mode Tesponse X X 2) Pressure kgulator X X X X X X 21 Ebed Sys-Setpoint Changes X X X X X X X 21 Feed Sys-Ioss EW Heating X( 5) 21 Feede ter Ptr:p Trip X( 6) X( 7) 21 Fax FK Rmout Capability 22 'Iurbine valve Surveillance X(8) X(10) 23 MSIV Ebnctional 'Ibst X X(11I X(12) X(13) 23 MSIV Ebll Isolation X 24 21ief Valves ( 20 ,X( X( 20) 25 "Iurbine Trip & Ioad d X(15)b (16 X(17) x Pcjection j Shutdown Outside CRC / X 27 Fecirculation Flow Cbntrol X(14) XIl0) 2 Recirece Ptrnp Trip X X 2 RPT 'Itip 'Iko PLI ps X(19) r g gg< _2B Pecire Systen Ibrforrance X X X X L Z' R~cire n--- R:nhad,1 1 [ 3 meirc Sys Cavitation 30 Ioss of Offsite Pwr X h X 31 Pipe Vibration X X X X 29 Pecirc Flow Calibration X X 32 PhuJ X( 23) h23) XI 21I 33 RHR 4 34 Drywell & Stean 'Ibanel X X j X X X Cooling 35 Gascous Radm ste X X X 38 SACS Performance X X 40 Cbnfirratory In-Plant 'Ibst h X x, x(24) [ xcacol l l l l l 1

(1) Test conditions refer to plant conditions [2d RCic. M,IF M on Figure 14. N PREVouSL7 N 6 " '^ - ( 2) Perform Wst 5, timiry of 4 selected control rods, in con 3 unction with expecttd scram ' X tX (3) Dprnic Syste:n Test Case to be co pleted l betwen test coMitions 1 and 3 l ly( 2) (4) Af ter recirculation pump trips (natural circulation) X (5) Betwen 80 ard 90 percent thermal pur, i X and near 100 percent core floa !X i (6) Pax FW Punout Capability & Fecire Pxp Pr.back :~ast have altrady been empleted X (7) Fleactor power between 80 ard 90 percent X X (8) mactor' power betwen 45 aM 65 percent X and 75 a:d 90 perrent X (9) Deleted X X (10) At maximtra power that.will not cause scran X( 5) X( 6) (11) Perfom betwen test oorditions 1 and 3 XC7) X(10) (12) Feactor power betwen 40 ard 55 percent X (13) Icactor power between 60 and 85 perwnt X( 20) X(17) (14) Betwen test conditions 2 ard 3 lTuRatuc Trip / (15)!mrmr leM MetM within bypass ) valve capacity ] (10) reacter p~rr MM=r 50 'M 90 p^roent-j[ (lQ DE(J-it-l./ l X Ot 00r0 fl? ' l 95 PrTt

"Ni^
ZIP I (17) frni rejection X

(18) Between test conditions 5 and 6 X (19) >50% tower ard >95 core flcuf eM-par 4orm X( 21) pore

==.e Trip ; m@t-im X ( 2)) Check SW operability durirg major scran X tests X ( 21) Performed durirg cooldown fran test t cordition 6 Hoet CREEK GENER ATING STATION FINAL SAFETY ANALYSIS REPORT (22) 'Ibe test ntster correlates to FSAR Section 14.2.12.3 x where x is the irdicated test ntster. TEST SCHEDULE AND CONDITIONS ( 23) Pay be perforned any time test corditions pemit.~ FIGUFiE 14 5 Ame

ent 14.01.10

HCGS FSAR 15.2.4 MAIN STEAM ISOLATION VALVE CLOSURES 15.2.4.1 Identification of Causes and Frequency Classification 15.2.4.1.1 Identification of Causes Various steam line and nuclear system malfunctions or operator actions can initiate main steam isolation valve (MSIV) closure. Examples are low steam line pressure, high steam line flow, high steam line radiation, low water level, or manual action. 15.2.4.1.2 Frequency Classification 15.2.4.1.2.1 Closure of All MSIVs This event is categorized as an incident of moderate frequency. To define the frequency'of this event as an initiating event and not as the by-product of another transient, only the following contribute to the frequency: I a. Manual action (intended or inadvertent) f l ) } b. Spurious signals, such as low pressure, low reactor water level, and low condenser vacuum c. Equipment malfunctions, such as faulty valves. Depending on reactor conditions, closure of one MSIV may cause immediate closure of all the other MSIVs. If this occurs, it is also included in this event category. During the MSIV closure, position switches on the valves provide a reactor scram if the valves in three or more main steam lines are less than 90% opend except for interlocks that permit proper plant startup.

However, protection system logic permits the test closure of one MSIV without initiating scram from the position switches.

(SE W 1, T4as s2-s) ( 15.2-16 l

E - HCGS FSAR 9/85 levels. Credit is taken for the operation of the pressure and flux signals to initiate a reactor scram. All plant control systems are assumed to be functional. 15.2.4.2.3 The Effect of Single Failures and Operator Errors Mitigation of pressure increase is accomplished by initiation of the reactor scram via MSIV position switches and the RPS. SRVs also operate to limit system pressure. All of these aspects are designed to the single failure criterion and additional single failures would not alter the results of this analysis. Failure of a single SRV to open is not expected to have any significant effect. Such a failure is expected to result in less than a 20 psi increase in the maximum vessel pressure rise. The peak pressure still remains below 1375 psig. The design basis and performance of the pressure relief system is discussed in Section 5.2.2. 15.2.4.3 Core and System Performance 15.2.4.3.1 Mathematical Model The computer model described in Section 15.1.2.3.1 was used to simulate these transient events. 15.2.4.3.2 Input Parameters and Initial Conditions These analyses have been performed, unless otherwise noted, with plant conditions as tabulated in Table 15.0-3. The MSIVs close in 3 to 5 seconds. The worst case, which is the 3-second closure time, is assumed in this analysis. Position switches on the MSIVs initiate a reactor scram when the valves are less than 90% openn Closure of these valves inhibits I steam flow to the feedwater turbines terminating feedwater flow. i \\ (3i5hk%1,) ~' let&E IS.;L-S' 15.2-19 Amendment 12 s

HCGS FSAR The following is the sequence of operator actions expected during the course of the event when no immediate restart is assumed. The operator will: a. Verify that all rods are in, following the scram b. Verify HPCI and RCIC initiation c. Verify that the main steam safety / relief valves (SRVs) open on reactor high pressure d. Verify that the reactor recirculation pumps trip on reactor low-low level e. Secure HPCI when reactor level and pressure are under control f. Continue operation of the RCIC system until decay heat diminishes to a point where the RHR system can be put into service. g. Monitor the turbine coastdown and break the vacuum as necessary h. Complete the scram report and survey the maintenance requirements. 15.2.7.2.3 System Operation Loss of feedwater flow results in a proportional reduction of vessel inventory causing the vessel water level to drop. The first corrective action is the low level, L3, scram actuation. ^~ RPS respondsMwi u i r' 1 second after this trip to scram the g Atcur reactor. The low level, L3, scram function meets the single failure criterion. Containment isolation, when it occurs, would also initiate a main steam isolation valve (MSIV) position scram signal as part of the 15.2-33 i

HCGS FSAR TABLE 15.2-5 SEQUENCE OF EVENTS FOR MAIN STEAM ISOLATION VALVE CLOSURE (FIGURE 15.2-5) Time, s Event 0 Initiate closure of all MSIVs. 0.3 MSIVs reach 90% open. 0.3 MSIV position trip scram is initiated.O 2.6 Group 1 SRVs are activated. 2.6 High pressure pumptrip is initiated. 2.6 Group 2 SRVs are activated. 2.7 Group 3 SRVs are activated. 8.5 Group 1 SRVs are closed. 1 Nore.1 : T9s 9o*/o dF9J VAWE RAS USED R* Tus Has Ces< Awtysis. Tm use er An 8s% a:sa vruc RR TE. Tb5Toa Frppa wouto WX RAVE A Sf60fF1ChJT tmFMCr 04 The T2MSIEMT S WAILTS. i ~

HCGS FSAR 10/83 l' t Fuel loading errors, undetected by in-core instrumentation following fueling operations, may result in undetected reductions in thermal margins during power operations. No detection is assumed, and therefore, no corrective operator action or automatic protection system function occurs. 15.4.7.2.2 Effect of Single Failure and Operator Errors This analysis represents the worst case, i.e., operation of a misplaced fuel bundle with three SAFs or SOEs, and there are.no additional operator errors that could make this accident more severe. Refer to Section 15.9 for further details. 15.4.7.3 Core and System Performance Analysis methods for this event are discussed in Section S.2.5.4 of GESTAR II (Reference 15.4-3). Results of analyzing the worst fuel bundle loading error are reported in Table 15.4-7. As can be seen, the MCPR remains well I above4the point where boiling transition would be expected to E j ~nEF occunt ane-;ne man: mum LHG? deec not exceed the 't plastic strain] FYJ4R !!! Pit for the clad.1 Therefore, no fuel damage occurs as a result' SAFETY of this event.

UMfr, 15.4.7.4 Barrier Performance An evaluation of the barrier performance is not made for this event since it is a very mild and highly localized event.

No change in the core pressure will be observed. p 15.4.7.5 Radiological Consecuences An evaluation of the radiological consequences is not required for this event since no radioactive material is released. 15.4-14 Amendment 2

HCGS FSAR 10/83 TABLE 15.4-7 RESULTS OF MISPLACED BUNDLE ACCIDENT M Swery lJ m fT

1. kiInitial CPR without misplaccdi

[ l' m.._mi_s tuinir= cPP witMImisplaced W-l ff.p } 1 \\ MCFR 2. FoR. / bundle A l l3. f.CPil for event O.10! l Amendment 2

.S 4 HCGS FSAR 8/84 b c. Leakage from the main steam isolation valve sealing system (MSIVSS). nit is assumed that the MSIVs will leak at a[ rate of 46 3GFW /C 707 cfm 'II.5 CTU for each of the four valves; for mye AOL the first 20 minutes of the accident. At that time, pbaR operator action to initiate the MSIVSS will eliminate mMMSqpe further leakage. As discussed above, after the initial LMES 175 seconds, the leakage is released to the reactor building. The iodine is assumed to be 91S. elemental, 5% particulate, and 4% organic. The FRVS will maint'ain the reactor building at a negative pressure (-0.25 inch w.g.) by exhausting air according to the following equation: E(t) = 336 + 5637 exp. (-1.18t) l where: E(t) = exhaust rate, cfm t = time after the building reaches -0.25 inch w.g., h (assumed to be 175 s after LOCA) l ( The FRVS provides for filtered recirculation and filtered exhaust. A description of the FRVS design is provided in Section 6.8. A discussion of the mathematical modeling of the FRVS is in Section 15A.6.2. Fission product activity cirborne in the reactor building and the activity released to the environment based on the above assumptions are given in Tables 15.6-15 and 15.6-16, respectively. 15.6.5.5.1.3 Radiological Results Dose conversion factors for iodine are taken from Regulatory Guide 1.109 and breathing rates during the accident are taken from Regulatory Guide 1.3 as presented in Appendix 15A. The l whole body-dose is calculated using the dose conversion factors l for~the semi-infinite cloud model discussed in Regulatory Guide 1.109. k I 15.6-20 Amendment 7 t .}}

ML20141D973

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Dear Ms. Adensam:

SUMMARY

System Description

7.6 REFERENCES

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ML20141D973

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Dear Ms. Adensam:

SUMMARY

System Description

7.6 REFERENCES

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