Text

Forwards marked-up FSAR Sections,Revised to Maintain Consistency W/Tech Specs & to Reconcile as-built Plant Discrepancies.Revs Will Be Incorporated Into Amend 15 After Fuel Load But Filed Now to Support Issuance of OL
	ML20210D765
Person / Time
Site:	Hope Creek
Issue date:	03/21/1986
From:	Corbin McNeil; Public Service Enterprise Group
To:	Adensam E; Office of Nuclear Reactor Regulation
References
	NUDOCS 8603270051
	Download: ML20210D765 (155)
	v • d • e

{{#Wiki_filter:. 4 i Public Service Electric and Gas Company Corbin A. McNeill, Jr. Public Service E ectric and Gas Company P.O. Box 236,Hancocks Bndge NJ 08038 609339-4800 %ce President - Nuclear -March 21, 1986 f i l i Director of Nuclear Reactor Regulation United States Nuclear Regulatory. Commission j 7920 Norfolk Avenue j Bethesda, Maryland 20814 Attention: Ms. Elinor Adensam, Director l -Project Directorate 3 Division of BWR Licensing i

Dear Ms. Adensam:

FINAL SAFETY ANALYSIS REPORT REVISIONS HOPE CREEK GENERATING STATION DOCKET N0. 50-354 Public S3rvice Electric and Gas Company (PSE&G) hereby submits various revisions to the Hope Creek Generating i Station (HCGS) Final Safety Analysis Report (FSAR). -The attached revisions to the HCGS FSAR contain:

1) revisions i

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 i

each change while Attachment 2 contains the actual marked-up sections of the-FSAR. These revisions will be incorporated 3 t 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. j In addition, an affidavit is provided to affirm that the matters set'forth in this transmittal are true and accurate. 1 I This submittal supplements similar transmittals from C.A. j McNeill to E. Adensam dated March 3, 1986 and March 17, 1986. I 327 0 [0 $ p A .--,m-,,.-,_,.-_ _ _. _,. -, _.. _ ~, ..,.__--_._.__..-,_,_.__...,_.m_._..___._ _.... ~ ~

Director of Nuclear 2 3-21-86 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

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 t PUBLIC SERVICE ELECTRIC AND GAS COMPANY FINAL SAFETY ANALYSIS REPORT-REVISIONS t 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. F 1 111, Jh - Vice President - Nuclear Sworn to and subscribed before me, a Notary Public of New Jersey ~, this J/M day f Marchr1986. 2 b2M 3 9s s DELORIS 0. HA00EN~ A Notary Public of New Jersey.. My Commssion Expres March 141m

L' ' ATTACHMENT 1

SUMMARY

OF. CHANGES, ADDITIONS AND/OR MODIFICATIONS j 1.2.4.1.1 Revisions incorporate various references 1 l 1.2.5 to GESTAR. l.3.1, 1.3.3 T1.3-1, sh. 1,'2,3/6 T1.6-1, sh. 1-9/10 i-4.1.2.1.3 4.1.3.2 e 4.1.3.3 4.1.4.3 l 4.1.5, 4.2 4.3.2.4.2 4.3.2.5.1 4.3.5 l 4.4.2, 4.4.8 T4.4-1,10 l 6.3.6 15.0 i T15.0-3, pg. 3/3 i T15.0-4 [ F15.0-1 l 15.3.5, 15.4.2 l 15.4.10 l T3.ll-5 Revisions to these two tables are necessary. l pg. 6,10,16,21, to meet' latest safety'and hazard analysis 23,28,30,32-34, requirements for?the Electrical Harsh 37,38,42-46, Environmental Program. The revisions 53,54,58,61-66, to T3.11-5, pg. 30/119 and T3.11-6, pg. 2/3 70,71,80,83, supercede those transmitted to the NRC-84,95,12.1, in a letter from C.A. McNeill to E. Adensam 118/119 dated March 17, 1986. T3.ll-6 pg. 2/3 4.6.3.1.5 Revisions provide the commitments to Q410.26 implement the-requirements of'NUREG-0803 i regarding scram discharge volume piping l . breaks. These revisions address. License l Ccndition #6 contained-in~the Hope Creek Draft' Low Power License (NRC } letter to PSE&G. dated March 17, 1986.) I 7.3.1.1.1.2 Revision provides a description of keylocked hand switches for local manual control of certain ADS valves. i

2 7.4.1.1.2. Revision deletes the reactor vessel high water level automatic trip of the RCIC turbine and closure of the throttle valve. In actuality, the high water level trip is used to initiate closure of the RCIC steam supply valve to shut off the steam to the turbine and stop RCIC operation (pg. 7.4-3, Am. 14), Hence this revision is necessary to accurately reflect' system design as summarized in the FSAR. This revision affects SER Section 7.4.1.1, pg. 7-37. 8.1.1 Revisions describe the addition of. 8.2.1.4 the island substation and 13-kV feeders 8.2.1.5 to the substation. F8.2-2 8.3.1.1.1 8.3.1.1.2.10 F8.3-1,2 T8.3-1 Revisions reflect as-tested plant conditions. i pg. 2,4,10/10 F8.3-16 Revision necessary to maintain consistency sh. 7/8 with Technical Specification 4.8.2.1.d. T9.1-10 Revision deletes footnote (3) from Seismic Category I for Item 9 - Vacuum Breaker Valve Removal Hoist. This hoist is normally stored outside torus i but brought in, installed and used i during cold shutdown. Since sh'utdown cooling would not be lost if the hoist fell off the monorail during cold shutdown, the as-built hoist does not have to i have seismic' restraints. T9.3-5 Revision necessary to clarify surveillance pg. 1/2 test prerequisites. 13.5 Revisions reflect the current station 13.5.1 review process for procedures as reflected 13.5.2.1.5 in the Technical Specifications and 17.2.2 plant administrative procedures. Revisions also reflect current procedure tit]as. 14.2.12.3.3 Revisions reflect various power ascension 14.2.12.3.16 test program modifications approved 14.2.12.3.24 by the NRC in a letter to PSE&G dated 14.2.12.3.25 February t, 1986. Revisions to Figure i 14.2.12.3.28 14.2-5 supercede those provided.to 14.2.12.3.32 the NRC in a letter from PSE&G dated F14.2-5 March 3, 1986.

= 3 14.2.12.3.3.c Various revisions necessary to accurately 14.2.12.3.6.b reflect the startup test program. 14.2.12.3.8.b These revisions. supplement similar 14.2.12.3.29.d changes previously submitted to the 14.2.12.3.33.b NRC from PSE&G on February 4, 1986 (FSAR Amendment'14). 14.2.12.3.12 Revision removes the reference to pump oil temperature for the RCIC system as such a temperature element is not, nor need to be, provided in accordance with General Electric requirements.

These revisions impact the SER'as-noted I

These revisions have already been accepted by the NRC in a letter to PSE&G as noted.

7 i 1 l 4 i i l i .i 1

ATTACHMENT 2 1

i c HCGS FSAR The arrangement of structures on the site is shown on Figure 1.2-1. The general arrangement for the major power block , structures is shown on Figures 1.2-2 through 1.2-11. The j ,e'quipment arrangement for these structures is shown on j Figures 1.2-12 through 1.2-43. i 1.2.4 SYSTEM DESCRIPTION 4 i, { A summary of the system description for Hope Creek Generating Station (HCGS) is' provided below. i 1 i d 1 1.2.4.1 Nuclear System I i a 1 i The nuclear system includes a direct-cycle, forced-circulation, General Electric (GE) boiling water reactor (BWR) that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power conditions is shown on Figures 10.1-1 and 10.1-2. 1 P 1 1.2.4.1.1 Reactor Core and Control Rods ) f flNSERT,i O 4 nel f or the rcactor core consists-of-s-1-ight4y enriched ueefthem-i dic ide pelletc ccaled in 2irc 1cy-2-t+bec.' Thece tubes-for--fuel reds) ere essembled inte individual fuel assemblies. Each f uel acccmbly hac ceveral fuel red: with gedelinia, Cd,0,, -ized ir c0 lid celetier with the "O,. The Gd,0, ic burn:ble= paiser that diminishes the reactivity of the fresh fuel. !-t-le-considerably depleted as the feel reaches-the end Of--its-f-i-est-p 2 c"cle: / .A W Gross reactivity centrol of-the core is achieved by-movab % Mtter-entry control-rede. The control-rede are cruciform-in sh=re end are located throughout the latt-ice Of fuel :ssembl-les. Tha r entrol rods are positioned by-individual centrol-rod-dri-ves MRDs4,- J. cen:ervat-i-ve--14mi4--of-pl-ast4c-stra1-n-is-the de ign-- cr4terion- .used-for-f-uel-rod-c-l add i ng-f a i4 ur e. The pe k linear hect: i generatton-for-steady-state-operat-ion-is-wel4-below the feel deme;e li-it, ever late-in-plant life. E::perie'nce-hec sheer that the-control--rods-a re-no t-sus cept-i b le-to-di st or-t-l on -a nd-h a v e-an. 1 1.2-17 t 1

-INSERT FOR PAGE 1.2-17 The reactor core and control rods are described in Section-I and Appendix A, Subsection A.l.2.2.3.1 of Reference 1.2-1.

l \\ HCGS FSAR 01/86 freerage life expectancy many timcc the rccidence t4mc of a fucl' fic2dinc. 1.2.4.1.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structures, steam separators and dryers, jet pumps, control rod guide tubes, distribution lines for the feedwater, core sprays, core differential pressure and liquid control lines, in-core instrumentation, and other components. The main connections to the vessel include steam lines, coolant recirculation lines, feedwater lines, control rod drive (CRD) and in-core nuclear instrument housings, core spray lines, core differential pressure l line, jet pump pressure sensing lines, water level l instrumentation, and CRD system return lines (capped). l l The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure in the steam space above the separators is 1020 psia. The vessel is. fabricated of low alloy steel and is clad internally with stainless steel (except for the top head, l which is not clad). [. \\ The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the i fuel rods. The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steam lines. Each steam line is provided with two automatic containment isolation valves in series; one on each side of the primary containment barrier. 1.2.4.1.3 Reactor Recirculation System The reactor recirculation system consists of two recirculation i pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. Each loop has one motor-driven recirculation pump powered and controlled by a dedicated motor-generator set located outside the primary containment. Recirculation pump I speed can be varied, to allow some control of reactor power level through the effects of coolant flow rate on the moderator void content. ( i 1.2-18 Amendment 14 a : = = -

I I HCGS FSAR V( $~ effluents from the plant that are potentially radioactive are 1 monitored. 1.2.4.8.2 Area Radiation Monitors 1 Area radiation monitoring systems alert plant and main control j room personnel of excessive gamma radiation levels at.various 4 locations.within the plant. 1.2.4.8.3 Site Environs Radiation Monitors I Radiation monitors are provided outside the plant structures to monitor radiation levels. The data obtained from these' monitors are used to compute the onsite and offsite radiation levels due to the plant operations. 1' 1 1.2.4.9 Shieldina ] 1 Shielding is provided throughout the plant, as required, to reduce radiation levels to operating personnel and the general public within the applicable limits set forth in 10 CFR 20, 10 CFR 50, and 10 CFR 100. It is also designed to protect certain plant components from radiation exposures that could i result in unacceptable alterations of material properties or j activation. l t 1* ba b

1. 2.- l

" Gee Esc 7eic booARb Appocanoo YGNXA. UEL, IktLLLDIMS THG ' UMmED STATES f %DS- 04011 - P - A - 9 AMD NED6 - di)hol 1 - P-A us. i j i 1 ; l 1.2-43 f 1-. _~:.._

S HCGS FSAR r \\ 1.3 COMPARISON TABLES I 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS This section highlights the principal design features of the plant and compares its major features with those of other boiling water reactor (BWR) facilities. The design of this facility is based on proven technology obtained during the development, i design, construction, and operation of BWRs of similar types. The data, performance characteristics, and other information presented here represent a current, firm design. The following tables summarize the plant design characteristics of the Hope Creek Generating Station (HCGS), the Hatch Nuclear Plant, the Limerick Generating Station, and the Susquehanna Steam Electric Stations = Table No. System 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics 1.3-2 Comparison of Power Conversion System Design Characteristics 1.3-3 Comparison of Engineered Safety Features and Auxiliary Systems Design Characteristics 1.3-4 Comparison of Containment Design Characteristics 1.3-5 Radioactive Waste Managerment Systems Design Characteristics i 1.3-6 Comparison of Structural Design Characteristics 1.3-7 Comparison of Instrumentation and Electrical { Systems Design Classifications l ~ 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION (FSAR) All of the significant changes that ha've been made in the facility design since submission of the PSAR are listed in I Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR that describes the changes and i the bases for them. = l l.a.3 mmesuces; "66UGP ELECTPJC, $TMhAf'b kFRJCAT1CM M WCTOR i\\ l.b-I ~ ~ ' h, "JVIkt~LLLDlG5 TH6." llM fTGD OT47BS M60F, " hETCG-l 340ll-P- A 4 AMD MGts-24ou - P-A us.3,3_3 I Ts nta twA N Tes 13-I c movitsD su %meGas l.3-I. -n-- ~

j i HCGS FSAR 1/86 l TABLE 1.3-1 Page 1 of 6 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CH ARACTERISTICSE

8 l

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", Nove Ser-1975. NEDE-20944-P BWR/4 and BWR/5 Fuel Design, i Proprietary Versions, October 1976. i i 1 NEDE-20944-P-1 BWR/4 and BWR/S Fuel Design, 4.3 l 1 Amendment 1, (only BWR/4&S,) 4.4 i Amendment 7 4 1 l HCGS FSAR 8/84 r t \\_ TABLE 1.6-1 (cont) Page 3 of 10 Report Referenced in Number Title FSAR Section l January 1977. p 1 / NEDE 21156 Cupplcrcntal Information for Plant 4.f MOdificatiOr to Ellri".2tc Significant I-Corr "ibration, January 1975. NEDE-21354-P BWR Fuel Channel Mechanical Design 3.9 and Deflection, September 1976. NEDE-23014 HEX 01 User's Manual, July 1976. 15.2Y _L L T.U..L r a. n^ I

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n- -wnan.a_me i. _- % M. _%e n h. E. E.1 1.M. E. . _W b m. 42 ^ 4 3' RWISIO3 ) NEDE-24011-P-A General Electric Standard Application /4.4,g ~f I for Reactor Fuel \\;;2tes apprece: 63;150,\\

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^ K 15.3,IT 4 \\ rec:::cr. NEDE-240ll-P-A-US General Electric Standard 4. 4, 0 Application for Reactor Fuel, United [ States Supplement'9;12ters appreved s T NEDE-24222 Assessment of BWR Mitigation of ATWS 15.8 l (NUREG-046J Alternate No. 3), Volume 1, May 1979; Volume 2, December 1979. NEDE-2:226-P Evaluation-cf Control Blade Life with !.T Pctential Loss of D,C, Deccmber 1979. NEDE-24834 Hanford 2 Crimped CRD Hydraulic 3.6 l Withdrawal Lines, (Proprietary). NEDO-10029 An Analytical Study on Brittle 5.3 Fracture of GE-BWR Vessel Subjected to the Design Basis Accident, July 1969. NEDO-10173 Current State of Knowledge, High 11.1 Performance BWR Zircaloy-clad UO, i Fuel, May 1970. NEDO-10200 A Core Flcu Distribub+cn in "Oder.a 1.0 2 B0!!!ng-h'ater -Reacter Oc Measured in -- Montice!!c, October 1976. Amendment 7 l t j HCGS FSAR 1/84 l (~

t

}' TABLE 1.6-1 (cont) Page 4 of 10 Report Referenced in Number Title FSAR Section NEDO-10320 The General Electric Pressure 6.2 Suppression Containment Analytical j Model, April 1971; Supplement 1, May 1971. / I NEDO-10349 Analysis of Anticipated Transients 15.8 { Without Scram, March 1971. NEDO-10505 Experience with BWR Fuel Through 11.1 September ~1971, May 1972. NEDO-10527 Rod Drcp Accident An:1ysic fcr L:rge

15. (

Ociling 'Jeter Reacters, March 1072; j Supplement 1, July 1972, Supplement 2, Januar; '??2. NEDO-10585 Behavior of Iodine in Reactor Water 15.6 During Plant Shutdown and Startup, i August 1972. ( ~ NEDO-10602' Testing of Improved Jet Pumps for the 3.9 BWR/6 Nuclear System, June 1972, t i NEDO-10739 Methods for Calculating Safe Test

6. 3,-

i Intervals And Allowable Repair Times 15.9 i for Engineered Safeguard Systems, b' January 1973. l r NEDO-10801 Modeling the BWR/6 Loss-of-Coolant .1.5 2 Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness, l March 1973. NEDO-10802 Analytical Methods of Plant Transient 4.1, 4 Evaluations for General Electric 15.1 15.0, Boiling Water Reactor, February 1973. 4 i NEDO-10846 BWR Core Spray Distribution, April 1.5 1973. a NEDO-10871 Technical Derivation of BWR 1971 -11.1 i Design Basis Radioactive Material ) Source Terms, March 1975. l NEDO-10899 Chloride Control in BWR Coolants, June 5.2 1973. l j Amendment 4 l i HCGS FSAR 8/84 1 i 4 p TABLE 1.6-1 (cont) Page 5 of 10 i Report Referenced in r Number Title FSAR Section M i NEDO 1005E Cencral Electric E9' Thermal ^nalycic 15.0 l l Bacic (CET^.E) Da t ::, Correlatic.S, l / and Decigr ^pplicaticr, Nc'> ember '9'2. ] ,l' i NEDO-10958-A General Electric BWR Thermal Analysis 1.5, 4.4 i Basis (GETAB): Data, Correlation, and i Desian Application, January 1977. l ) NEDO-11209-0A Nuclear Energy Business Operations 1.8 i Boiling Water Reactor Quality Assurance Program Description, Latest NRC-Accepted Revision NEDO-12037 Summary of X-Ray and Gamma-Ray Energy 12.3 and Intensity Data, January 1970. 1 NEDO-20231 Emergency Core Cooling Tests of an 1.5 Internally Pressurized, Zircaloy-Clad, 8X8 Simulated BWR Fuel Bundle, l, December 1973. 'N EDC Ocneral Electric Sciling h'ater 15.0 l -20300 elcad A.pplicatier for-Reacter Generic o own r,m1 uq. 3one i NEDO-20566 General Electric Company Model for .1. 5, 3.9, Loss-of-Coolant Accident Analysis in 6.3 Accordance with 10 CFR 50, Appendix K, April 1977 NEDO-20626 Studies of BWR Designs for Mitigation 15.8 of Anticipated Transients Without Scrams, October 1974. NEDO-20651 BWR Radiation Effects Design Curve, 5.3 March 1975. NEDO-20922 Experience With BWR Fuel Through 11.1 September 1974, June 1975. NEDO 200?? BF." " and EF"./5 Fuel Dec!gr., 44, 4.27 = October 1976. 4.3, f.t NEDO 20053 Three Dimension 01 Sciling 'ater 15.6- -Reactor Ccre Simulatcr, May 1076, Amendment 7 ) I HCGS FSAR 8/84 g TABLE 1.6-1 (cont) Page 6 of 10 l l Report Referenced in i Number Title FSAR Section -NEDO-20952 > Three-Di enc:enal BF" core Simulater, j -January '9-~ I NEDO 20^0' Peach Better ^tcmic cuer Statier 4 4 o ~~ Unitt 2 and 2, Safety /.nclycic 9epcrt for blant "Odific tiOnc t0 Elininate i Significant In-Corc '/ibrations, L October 1^'5. NEDO-21142 Realistic Accident Analysis for 15.4, General Electric Boiling Water Reactor-15.6, The RELAC Code and User's Guide, 15.7 January 1978. NEDO-21143 Conservative Radiologice'. Accident 15.4, l Evaluation The CONAC01 Code, March 15.6, 1976. 15.7 i l NEDO-21159 Airborne Release from BWRs for Envi-11.1, 12.3 i f ronmental Impact Evaluations, March 1976. j NEDO-21159-2 Airborne Releases from BWRs for 12.3 Environmental Impact Evaluations, 1977. NEDO-21506 Stability and Dynamic Performance of 4.1/, 0.?! the General Electric Boiling Water ~ Reactor, January 1977. NEDO-21660 Experience with BWR Fuel through 11.1 December 1976, July 1977. l NEDO-21778-A Transient Pressure Rises Affecting 5.3 i Fracture Toughness Requirements for BWRs, December 1978. l NEDO-21821-2 Boiling Water Reactor Feedwater 5.3 Nozzle /Sparger Final Report 7 (Nonproprietary), August 1979. jg,/ i NEDO 237SS-1 Fuel Red Preprescurination, 4.2 Amendment sy '9?S-u j NEDO-24057-A Assessment of Reactor Internals 1.5, 3.9 Vibration in BWR/4 and BWR/5 Plants, November 1977. s ~. Amendment 7 ) i 5 HCGS FSAR 8/84 1 r TABLE 1.6-1 (cont) Page 7 of 10 Report Referenced in Number Title FSAR Section NEDO-24154 Qualification of the 4.1, 5.2 One-Dimensional Core Transient S 5. u,; 15.1 l Model For BWR, October 1978. 4 f j NEDO-24988 Analysis of Generic BWR Safety / 5.2 Relief Valve Operability Test i Results, October 1981. I I i OTHER REFERENCED REPORTS AEEW-R-705 An Investigation Into the Effects 4.4 of Crud Deposits on Surface Temp-erature, Dry-Out, and Pressure Drop, with Forced Convection Boiling of Water at 69 Bar in an Annular Test' Section, 1971. AE n!L-700 Void "casurcacnts in the Regica cf ?.T Subeccled and Lee Ouality Boiling, Part !!, ^pr!! '955. AI-75-2 Thermal Hydrogen Recombiner System 6.2 3 for Water-Cooled Reactors, Revision 2, July 1975. AI-77-55 Thermal Hydrogen Recombiner System 6.2 j for Mark I and II Boiling Water i Reactors, September 1977. ANL-5522 The Effect of Pressure on.Sciling d.t Denci ty ir Mel t iple "ec-t-angulae ,jL Channel, February 1955. ANL 5521 Sciling--Dencity in Vertica! Rec-4 4 tangular Multichannel Sectione with i -Natur a l Ci r-sul a t i on, November 1956-. ) ANL-6285 Peeer-to-Vel d Tr-ans&sr--Funst i enc, 0.4 July 196!. 1 1 ANL-6948 Condensation of Metal Vapors: Mercury 6.2 and the Kinetic Theory of Condensation, i October 1964. l Amendment 7 r .~ . ~- l i i i HCGS FSAR 8/84 i e t

\\

TABLE 1.6-1 (cont) Page 8 of 10. Report Referenced in Number Title FSAR Section !l 4,,F MSI-110; Vapor Fcrmation and Ochacicr in Boil:nc " cat Trarrfer, Februar" 1957 2 J BHR/ DER 70-1 Radiological Surveillance Studies 11.1 at a Boiling Water Nuclear-Power Reactor, March 1970. i J LMI-1103 Vapor Formation and Scha/ict ir it,)"' Ecilinc 4 cat Trancfer, rebruarv 1957 ~ CF 59-6-47 Removal of Fission Product Gases from 11.3 (ORNL) Reactor Offgas Streams by Adsorption, i 1959. 1 EPRI NP-495 Sources of Radiciodine at Boiling 12.2 j Water Reactors, February 1978. ORNL-3041 SDC, A Shielding-Design Calculation 12.3 ]' for Fuel-Handling Facilities, i March 1966. l\\ j '- ORNL-4585 Morse - A Multigroup Neutron and 12.3 4 Gamma-Ray Monte Carlo Transport j Code, September 1973. 1 1 ORNL-4628 Origen - The ORNL Generation and 12.3 1 Depletion Code, May 1973. I i ORNL-4932 Radioactive Atoms Supplement 1, 12.3 ) August 1973. I l ORNC-NSIC-23 Potential Metal-Water Reacti'on in 6.2 1 Light-Water-Cooled Power Reactors, l August 1968. l ORNL-RSIC-10 A Survey of Empirical Functions Used 12.3 to Fit Gamma-Ray Buildup Factor, February 1966. I l ORNL-RSIC-21 Neutron and Gamma-Ray Albedos, 12.3 i February 1968. 1 I ORNL-TM-4280 The DOT 3 Two-Dimensional Discrete 12.3 Ordinates Transport Code, September 3 1973. e ( Amendment 7 l t M.. HCGS FSAR 8/84 (~ \\ TABLE 1.6-1 (cont) Page 9 of 10 ~ Report Referenced in Number Title FSAR Section FC-4290 Hydrogen Evolution from Zinc Corrosion 6.2 under Simulated Loss-of-Coolant Accident Conditions, August 1976. STL 272 2S Kinetic Studien Of Meterogencouc

4. 4 '[

~ a t er Dea cters, ^pril '965. l WASH-1258 Numerical Guides for Design Objectives 11.3 and Limiting Conditions for Operation ~ to Meet the Criterion as Low as Practicable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents. WCAP-8776 Corrosion Study for Determining 6.2 Hydrogen Generation from Aluminum and Zine During Post-Accident Conditions, 1976. i WAPD-TM-918 Thermal and Hydraulic Effects of Crud 4.4 l \\ Deposited on Electrically Heated Rod Bundles, September 1970. h'APO ST 10 .. Method Of Predicting Steady-State i f Boiling Vapor Fractienc ir Reactor Ccelant Charnelc, June '0E0. BECHTEL POWER CORPORATION REPORTS BC-TOP-4A Seismic Analyses of Structures and 3.7 Equipment for Nuclear Power Plants, Revision 3, November 1974. l BC-TOP-3A Tornado and Extreme Wind Design 3.3 l Criteria for Nuclear Power Plants, Revision 3, August 1974. j 1 BC-TOP-9A Design of Structures for Missile 3.5 l j Impact, Revision 2, September 1974. BN-TOP-1 Testing Criteria for Integrated 6.2 Leakage Rate Testing of Primary ) Containment Structures for Nuclear i f' Power Plants, November.1972. Amendment 7 g i HCGS FSAR 5/85 See-Section 1.8.2 for the' NSSS assessment of this Regulatory Guids. 1 i 1.6.1.107 Conformance to Reculatorv Guide 1.107,- Revision 1, Fecruarv 1977: Qualifications for Cement Groutina for Prestressino 7encons in containment Structures l Regulatory Guide 1.107 is not/ applicable to HCGS. I 1.E.1.108 Confermance to Reculatorv Guide 1.108, Revision 1, r j Aucust 1977: Periodic Testino of Diesel Generator Units j Used as Onsite Electric Power Systems at Nuclear Power j Plants i HCGS complies with Regulatory Guide 1.108, with the following 4 exception: 1 a. During the preoperational test phase, following the diesel 24-hour full load test, the proper design

(

accident loading sequence will be demonstrated by the test described in Section 14.2.12.1.47. This test will verify the ability of the SDG to start and accept the sequenced design loads as specified in Table 8.3-1. This test will. provide ECCS flows to the reactor vessel. l.WSEErl r s ] c For periodic testing reqdired by the Hope. Creek Technical Specifications, the test.per this regulatory position will be p'erformed during shutdown. This test will simulate, separately, a loss ~of offsite power, and 4 a loss of offsite power plus a LOCA condition, to, verify the SDGs' ability to start and accept the sequenced des'ign loads. i 1.8.1.109-Conformance to Reculatorv Guide 1.109, Revision 1, October 1977. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents >for the Purpose of Evaluatino Compliance with 10 CFR.Part 50, j Appendix I j v y HCGS complies with Regulatory Guide ~1.109. t 11 e 1.8-97 Amendment 10 i o ,.---.--.---m.-. --.n.,.-...,,..+-n.,--,,.,,-- ~..-v.....a 1 INSERT FOR PAGE 1.8-97 b. The criteria regarding sustained load levels of 100% 'and 110% can be demonstrated when those significant' parameters being measured have stabilized to acceptable values. Although the-110% load le' vel should be maintained i for 2 hours, reduced runtimes at 110% load levels are not regarded as an inadequate demonstration as defined by Position C.2.c provided: Runtime is sufficient to stabilize significant parameters being measured at the'110% load level and additional runtime (continuous beyond 22 hours) at the 100% load level is available to adequately demonstrate the diesel generator's load - carrying capability on an extended basis to compensate for the reduced runtime at the 110% load level. 1 J Y N 1 i d ? .n,. .w., 1/86 HCGS rS AR Page 6 of 819 TA R t.E 3,Il-5 FOUIPMENT SEl.ECTFD FOR HARSH l'NVIRONMENT OUALIFICATION P&lD M-II-I SYSTFM SAFFTY Aux COOL SYstrN EG FAM TNI ACTION I D NO. LOCATION FOUIP. PLAN fOulP. P.O. Note (51 N PL NO. COMPONENT BIDG. FIFV. NOTr (I) NO7F (2) FFSS PFr. NO. J2010 IA-C-201 SACS Control Panet A Peactor 102 No No 26,27,68,69, D J2010 IR-C-201 SACS Control Panel p peactor 102 No No 26,27,Q: 26,27 3 , -)O J2010 IC-C-201 SACS Control Panel C peactor 102 No No q % J20lO I D-C + 201 SACS Control Panet D Reactor 102 No No 26,27,' ,ffg#,b I:11200 IA-P-210 SACS Pump Notor Reactor 102 No No I Ell 2DO IH-P-210 SACS Pump Notor peactor 102 No No I Ell 200 IC-P-2IO SACS Pump Notor peactor 102 No No 1 Fil2OO ID-P-210 SACS Pump Notor Reactor 102 No No I P1010 4-EG-SV-2290A Solenoid valve Reactor 54 No No 124 P30lO l-rG-tS-2290A Poettion Switch Peactor 54 No No 125 P3010 1-t:G-SV-2 2908 - Solenoid valve peactor 54 No No 124 P30lO l-EG-rs-22908 Position Switch Reactor 54 No No 125 P3DIO l-EG-SV-2290C Solenoid Valve Reactor 54 No No 124 P3010 3-EG-IS-2290C Poettion Switch Reactor 54 No No 125 P9030 1-EG-SV-22900 Solenoid valve peactor 54 No No 124 P3010 l=EG-!S-22900 Poaltion Switch peactor 54 No No 125 P3010 1-EG-SV-2290E Solenoid valve peactor 77 No No 124 PIDIO I-EG-25-2290E Position Switch peactor 77 No No $25 P1010 3-EG-SV-2290r Position Switch peactor 77 No No 124 P1030 1-EG-IS-2290r Position Switch peactor 77 'No No 125 P1010 1-EG-SV-2290G Solenoid valve peactor 54 No No 124 P1010 1-EG-2S-2290G Position Switch peactor 54 No No 125 e m Amendment 14 w m .} t 1/P6 HCGS FSAR Page 10 of I19 TAHLE 3.11-5 FOUIPNENT SFI.FCTFD FOR HARSH FNVlpONMFNT QtfALIFICATION P&lD N-1l-I i SYSTFNs SAFtTY AUX COOL SYSTEN FG PAN TNI ACTION ID NO. LOCATION f:OU I P. PLAN FOUIP. P.,0 Note (5) NPL NO. CONPONENT plDC. F l. F V. NOTE (1) NOTfQ2 ) I'FSS RFF. NO. i N001 1-FG-PDT-2485A Diff. Press Trans. Reactor 102 No No 1%s,#5TA N001 1-FG-PDT-2485R Diff. Press Trans. peactor IU2 No No 155,857 4 N003 1-FG-PDT-2485C Diff. Press Trans. Scactor 102 No. No 15 5, l$5 A t Nuol I-FG-PDT-24BSD Ditf. Press Trans. Reactor 102 No No 155 J20lO l-t'G-H S - 2 4 0 5 A 2 Nano Switch peactor 102 No No 25 J2010 1-fG-HS-2485R2 . Hand Switch peactor lut No No 25 J20lO l-EG-HS-2485C2 Hand Switch peactor 102 No No 25 J20!O l-EG-NS-2485D2 Hand Switch peactor 102 No No 25 P1050 1-FG-HV-24914 Contr. Valve peactor 102 No No 144 r1050 1-FG-2S-2491A Limit Switch peactor 102 No No I44 I P905Q 1-EG-HV-24918 Contr. Valve peactor 102 No No 144 Plu50 1-FG-2S-44918 t.imit Switch peactor 102 No No 144 P1050 I-FG.HV-2494A Contr. Valve peactor 102 No No 144 P1050 l-FG-2S-24944. Limit Switch peactor 102 No No 144 P1050 1-EG-HV-24948 Contr. Valve peactor 104 No No 144 Pl050 1-FG-IS-24948 Limit Switch peactor 102 No No 144 P1050 1-EG-HV-2496A Contr. Valve peactor '302 No No 144 PID50 1-FG-IS-2496A Limit Switch peactor 102 No No 144 P1050 1-EG-HV-24968 Contr. Valve peector 102 No No 144 P9050 1-EG-25-24968 Limit Switch peactor 102 No No 144 P10%0 1-EG-HV-2496C Contr. valvo peactor 102 No No 144 P9050 1-tG-Is-2496C Limit Switch peactor 102 No No 144 Pl050 1-EG-HV-2496D Contr. Valve peactor 102 No No 144 P9050 1-FC-IS-2496D Limit Switch peactor 102 No No 144 J1010 1-EG-LT-25084 Level Trans. Reactor 201 No No 29 J)Ul0 l-EG-LT-2)UuH Level Trans. Peactor 201 No No 29 J1010 1-EG-LT-2508C level Trans. peactor 208 No No 29 J1010 l-FC-LT-2500D Level Trane, peactor 20I No No 29 s Amendment 14 i I 9 t' l 4 i O l I 4 f 1 o tn 6 e m 1 d 6 n e J 9 g e / a m P A 1 O. N F F R D S f S 7 6 E 2 P F 1 NP OI IU) TQ2 C l'( A NE IAT MLO o N TPN N O I T AC I ) F 1 I L P( A I E J t MUT Q AOO o PFN N T N E M N O 5 I R R1 V v A1 N E S F ( 7 F3 NI 7 H O SE S I GL R T CR A A r HA H C. to T O R LC c O F a F L e B R D I' TC F L F S T N E M P I UQ E ev l a T V N E l s N o N r t M n O o C C O. N L P M R E T A N M D 8 E 0 I 4 I 1 L A 5) V R E O ( H N Ne C I Do t N M M I N 1 s M S 10 f T O. 0 Y 1 S P P ? i ) / "/% 4 l i 1/n HCGS FS AR rage 21 of 819 TABLE 3.11-5 FQulP74fMT SFLFCTFD FOR HApSH FNVipONMFNT QtJALIFICATION pg l D N-2% 0 NYSTFNs PLANT LEAK DETECTidN SK PAM TNI ACTION ID NO. LOCATION FOUIP. PLAN l'Olll P. P.O.__ Not_, I5I NPL NO. CONIONENT Mi ttC. FLFV. OdOT M I) PMIT E (2) FFSM DFF. NH. J40lO l-tG-LSH-23594 Level Sw. High peactor 102 No No 42 j J48)O l-l'G-1,$H-2 3 59R Level Sw. High peactor 102 No No 42 i J48 BO l-FG-l%H-23644 (evel Sw. High peactor 102 No No 42 l' J4830 3-l'G-LSH-23648 Level Sw. High peactor 102 No - No 42 J40)O l-tD-8.NH+2365A Level Sw. High peactor '77 No peo 42 Je s se I -I D-e.Sh-215 55 Level Sw. High peactor 77 No No 42 J48to l-fD-LSH-2365C Level Sw. High peactor 77 .No No 42 J4810 1 - BD-t E-415 5 - 1 Level Element peactor 54 No. No 42 J48)O l-BD-LSH-4 8 51 -1 Level Sw. liigh peactor 54 No No 42 J48)O l-P D-LE-4151 - 2 Level Element peactor 54 No No 42 J40)Q I-BD-LSH-4151-2 tsvol Sw.'High peactor 54 No No 42 J48)Q I-nc-LE-4403Al Level Element peactor 54 No No 42 J48)O l-BC-tSM-4401A1 Level Sw. High peactor 54-No No 42 J48)O l-tic-LE-4 4 0 342 Level Element peactor 54 No No 42 J40)O l-8C-LSH-4 4 0 3 A 2 Level Sw. High peactor 54 No No 42 J4810 1-HC-LE-4401sl t.evel Element peactor 54 No No 42 J4Ql0 l -BC-IA H - 4 4 0101 Level Sw. High peactor 54 No No 42 J48)O l -hC-L E-4 4 0 3'.,? . Level Element peactor 54 No No 42 J 4 810 l-RC-t$H-440)D2 Level Sw. High peactor 54 No No 42 Y l * '^* W 'it? ' L:- t :=, "if 5' mae_ea_angn_n en-m $m =;,/ e - -- n _n a n n g a j '?O "C-L" '!t*-2

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l p. 5' -JC 30 4=acm.LEm 41513.2 Leva LEleme-t e r rt.a - ?4 a \\ l h { t I t I Amendment 14 l l ..a- =.- r-t ) 8 i 1/H6 d HCCS FSAS pagu }) og lly TAHLE ).18-5 EQUIPMt'NT St.ttCTED FOR HApSH ENVl>ONMFNT OuALIFICATION Ph lli M 16-I SYS?FNs RADIOL 4M;lCAL MONITONING STS 1 sr PAM TMJ ACTION B.OC AT I ON EQUIP. Pt.AN 4QulP. 1[3 w. l P.O. Note (5) MPl. NO. COMIONI.NT hl ld;. F QV. NOT M ll Nulf,12) 6 t %S ke t. No. M004 l-SP-ut-N006A Dll Dadiation Element Reactor 145 No No 3/8 n001 I -S P-D E -N 0 0 6 tl Dil Radiation Element peactor 145 No No lJ1 noot 1-SP-RE-N006C Dil Radiatlon Element Reactor 145 No No 1/6 y p.M005 l-SP-DE-N006D Dil Radiation Element Reactor 145 No Ne l J/l 1 Ja' O 3 -i!- - e &J 5 ^. " M Gt4ea-5 3 r ---i =r

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li8 N - a t? :e !.sa-as-48 4 34 andletten s! at meeste: 43" ++ 4% -o+? :;; rP-as-44HC Sedleteen-5&:::-i a eesee---44s u i1 --dt M J " ^^ 0 chbestee/Ece--C54 C f. M 4e-+3l ?o J)?JO l-SP-st-442%A B ON Ch ambe r De t ec t o r Neactor 145 Tes Yes )$ JIP)O l-SP-pf:-4A2%H ION Chamt>cr Det ector peector 145 Yes fes )$ J37)u l-SP-pt%4H56A p F t: Two Channel f asct Monitor Beactor 208 No No 36 J I 7 to l-SP-ut-485614 pt r Two t'hannel fasct Monttnr penetor 20! No No 36 J11)u l -s P-p r-4 n 5 6C p t t: Two t'hannel inac t Monitor peactor 201 No' No 16 84130 1 -S P-k r-4 4 % 1 A n H t: Two (!hanne l thict Monttnr Heactor 178 No No 36 3331Q 1 -S P-N F-4 8 5 7tt WHE Two Channel tasct Monitor peactor 178 No No 16 JillQ I-SP-RE-4H570 HME Two Channel inJc t Nonitor Reactor 17R No No 16 38740 1-SP-FF-4RilA FV4 Sensor Reactor 133 Yes No 17,3 h 3174J t-SP-FF-4Rtin DVA fie n gor Reactor 173 Yes No 17,)A I o Amcss.las nt 14 m _._. _ _... _ _. - - tj g l /f46 l. NCGS FS AR Page 28 of lig TABLE 1.11-5 EQUIPMENT SEtECTPD FOR NApSH FNVIRONMFNT OUALIFICATION PglD M-4l-1 - SYSTt M ' NUCLEAR RO I Lt.R AH PAN TNI ACTION tt No. LOCATION FOUIP. PLAN FQUIP. P.O. Notej 5) NPL NO. COMPONFNT BLDG. ELFV. NOTEj l) NOTE 121 FFSS,pFF.,NO. P1DIAQ' l-AR-HV-F067D. Control Valve peactor 102 No No 137 P301DQ .I-AR-ZS-F067D Limit Switch peactor 102 Yes No 137. P 10100 1-AR-HV-F071 Control Velve peactor 102 No No 139 P101a0 1-AR-2S-F071 Limit Switch peactor 102 No No 139 P3020 1-AE-25-F0744 Limit Switch peactor 102 Yes No 114 P1020 l-AE-2S-F074R Limit Switch peactor 102 Yes No 134 Nool l-AR-PDT-N0864 R21 Press Diff. Trans. Reactor 77 No No 155,l m M001 4 - AR-PDT-N08 68 R2I Press Diff. Trans. Peactor 77 No No. 155 MOUR l-AR-Pf tT-N0 0 6C R21 Press Diff. Trans. Peactor 77 No No 155 Mont 1-AR-PDT-NO86D R21 Press Diff. Trans. Reactor 77 No No 155 M0pl. I-AH-PDT-NOS7A R21 Press Diff. Trans, peactor 77' No No 155 MODI. l-Att PDT-N087R R2l Press Diff. Trans. Reactor 77 No No 155 Q001 l-AR-l'DT-N09 7C R21 Press Diff. Trans, practor 77 No No 155 l M001 l-AR-PDT-NOA7D R21 Press Diff. Trans. Peactor 77 No No 155 MODI l-AR-PDT-NOR8A H23 Press Dlff. Trans. Reactor 77 No No 155 '? MuGI l*AR-PDT-NORRR R2l Press Diff. Trans. Reactor 77 No No 155 Mont 1-AR-PDT-NORRC R21 Press Diff. Trans. Reactor 77 No No 155 M003 l-AR-PDT-N088D H21 Press Diff. Trans. Reactor 77 No No 155 Muul 1-AR-PDT-N0994 921 Press Diff. Trans. Reactor 77 No No 155 Mont 1-AR-PDT-N089R 921 Press Diff. Trans. Peactor 77 No No 155 4 Moul I-AR-PDT-N089C R21 Press Diff. Trans. psactor 77 No No 155 Mont I-AR-PDT-N089D 921 Press Diff. Trans. Reactor 77 No No 15% i 4 1 t i t. 1 Amendment 14 i I i i 1/P6 HCGS FS A R Page 30 of !!9 TABLE 3.18-5 ) EQUIPMcNT ssLECTED FOR NARSH FNVIRONMENT QUALIFICATION P6tD M-41-1 f SYSTFMs NUCLEAR Boll.ER AB PAN TMI ACTION I D NO. LOCATION EQUIP. PLAN F.OUIP. P. O., Note 15) MPL NO. COMPONFNT BLDG. ElEV. NOTE (1) NOTE (2) EFSS DFF. 98 0. J5560 1-AR-TE-364AA Temp. Elemt. Reactor 54 Yes No 43 J4560 3-An-TE-364RR Temp. Elemt. Reactor 54 Yes No 41 J%%60 1-AR-TE-3648C Temp. Elemt. Deactor 54 Yes No 4I .t M &O l-AR-TE-3640D Temp. Elemt. Reactor 54 Yes No 41 j Mont 1-SN-SV-36524 R21-F083 Solenold valve l'e sc t or 128 No No 159 Mont I-Sg-SV-16528 R21-F013 Solenoid Valve peactor 121 No No 159 M001 1-SN-SV-36534 821-F013 Solenoid valve peactor 121 No No 159 i M005 l-SN-SV-36539 928-F013 Solenoid valve peactor 128 No No 159 E005 3-SN-SV-36544 B21-F013 Solenold Valve peactor 123 No No 159 M001 I-SN-SV-3654R R21-F013 Solenold valve peactor 121 No No 159 R001 l-SN-SV-3655A B21-F013 Solenoid valve peactor 121 No No 159 R005 1-SN-SV-3655a 821-Fol3 Solenoid valve peactor 121 No No 159 M001' l-SN-SV-36564 B23-F013 Solenoid valve peactor 128 No No 159 } R001 l-SN-SV-3657A R 21-Fo l 3 Solenold valve peactor 121 No No 159 R001 l-SN-SV-16504 823-FO!3 Solenoid valve peactor 121 No No 159 M001 3 -SN-S V-36 594 B21-F013 Solenoid valve peactor 121 No No 159 M001 l-SN-SV-3660A B21-F013 Solenold valve peactor 123 No No 159 1 n001 3-SN-SV-3661A B21-F013 Solenoid valve peactor 121 No No 159 1 M001 l-SN-SV-3662A 821-F013 Solenoid valve peactor 121 No No 159 i M001 1-SN-SV-36634 R21-F013 Solenoid Valve peactor 121 No No 159 l} Mool I-SN-SV-3664A 921-F013 Solenoid valve peactor 121 No No 159 MIO1 1-SN-SV-3665A B21-FOI) Solenoid valve peactor 128 No No 159 008 l-SN-SV-36658 R28-F013 Solenoid Valve peactor 121 No No 159 4000 I-AR-XE-4507A Acouet t e Elemt. Reactor 121 Yes. Yes 58 J80!O l-An-MT-45074 Acouette Trans. peactor 102 Yes Yes 50 J2SCO' l-AR-ME-4507R Acoustic Elemt. Reactor 121 Yes Yes - 51 Jn000 l=AR-MT-45078 Acouette Trene. Reactor 102 Yes Yes 50 J80!O I-An-NE-4507C Acoustic Elemt. Reactor 121 Yes Yes 58 JOOOO l-AR-NT-4 507C Acoustic Trane. Reactor 102 Yes Yes 50 J0000 1-AR-3E-4507D Acouette Elemt. Reactor 121 Yes Yes 51 J8000 1-AR-NT-4507D Acoustic Trane. Reactor 102 Yes Yes 50 J8000 3-AR-XE-4507E Acouette Elemt. peactor 121 Yes Yes 51 J0000 1-AS-RT-4507E Acouette Trane. Reactor 102 Yes Yes 50 p J0000 3-AB-XE-4507F Acoustic Eleet, peactor. 123 Yes Yes 51 '-{ P30@ l-AE-HV-48W owrax. Wws W 102 @l 4 Amendment 34 l l m ...a. ... - ~, _ - _ -... _ _ .~_-n ~ ~ 1 1/F6 HCCS FS A R 'Page 32 of 139 TARI.E 3.11-5 E00lPMENT SELECTFD FOR NAPSH FNVIRONMFNT OUALIFICATION P68D M-42-1 NYSTFMr NtfCEFAR BOILER VTSSFL INSTRU i pf4 i PAM TMI ACTION ID NO. (DCATION EOUIP. PLAN FOUIP. P.O. Note 15) M PL NO. COMIONENT PLOG. FIFV. NOTE (ll. NOTF 12) FFS9 PFF. NO. M001 l-SE-EAM-2002A C51 ~ Voltage Preamp Reactor 102 No No 154 Mn01 3-SE-FAM-k002R CSI Voltage Preamp Reactor 102 No' No 154 I, . MOui l - S E-F. AM - R0 0 2C CSI Voltage Preamp Reactor 102 No No 154 Mool 4-SE-FAM-R002D C51 Voltage. Preamp Reactor 102 No No 154 Mn01 1-SE-EAM R002E C51 Voltage Preamp . Reactor 102 No No 154 Mont I Sr-F.AM-R002F C51 Voltage Preamp Reactor 102 No No 154 M001 l-S f:- F AM-5 00 2G C51 Voltage Preamp Re act or 102 No No 154 M00I !=sF-FAM-R002N C51 Voltage Preamp Reactor 102 No No 154 ' M0nl 1 -st -pf:-N00 l A CSI Radiation Elevet. Reactor 121 Yes No 162 Mnol I SE-Pf:-N00 th C51 Radletion Elemt. Reactor 121 Yes No ^162 ' Moni 1-SE-RE-N00lc C51 Radiation Elemt. Pesctor 121 Yes No '162 Mool 3-SE-RE-N001D CSI Radiation Elemt. Reactor 121 Yes No 162 Mont l-Sr-RF-N002A CSI Radiation Elemt. Reactor 121 Yes No' 303 Moni 1-SE-RE-N002R CSI Radiation Elemt. Reactor 121 Yes No 103 Mong I-SE-pr-N002C C5I Radiation Elemt. Peactor 121. Yes No 10 3 ~ Mnol 1 -S f:- R E-N 00 2 D CSI Radiation Elemt. Peactor 121 Yes No 101 Mn01 1-SE RE-N002E C5I Radiation Elemt. Reactor 128 Yes No 103 Mont l-SE-RF-N002F C51 Radiation Elemt. Reactor 121 Yes No 103 Monl I-$F-RE-N002G C58 Radiation Elemt. Reactor 121 Yes No 103 Mn01 I-SP-PF-N002N C5I . Radiation Elemt. Reactor 121 Ye9 No 103 1 Mool 1 - H H - PT-N 0 5C., C71 Pressure Transmitter Reactor 162 No. .No 155 Mnnt l-pp-PT-N050s C71 Pressure Transmitter Reactor 162 No No 'I S $ r unul I-na-PT-N050C Cll Pressure Transmitter Reactor-162 No No 155 Moni I -n te-PT-N0 500 C7! Pressure Transmitter Reactor 162 No No 155 Mnul 1-19 p-PT-N 0 7 84 B21 Pressure Transmitter Reactor 77 No No 155, IS% . M001 1-en-PT-N0789 R21 Pressure Transmitter Reactor 77 No No 155 - M001 I-Hn-PT-N078C B21 Pressure Transmitter Reactor 77 No No 155 4 Amendment 14 l I E ~.. - - _ - _ - - -. -. - -. - + p \\. e 1/86. HCGS FSAR Paq. 33 of 189 TARI.F 1.11-5 FOUI PMENT SFLETTFD FOR HARSH FNVIRONMFNT QUALIFICATION P6s M-42.I SYST)Mr NHOLFAR HOll.FR VESSFL INSTRU MH PAM TMt ACTION 8 81 NO. 8.OCATION FOUIP. PLAN FOUIP. P.O. Note 15) MPL NO. COMPONFNT stDC. FLFV. NOTF (l) NOTF 12) FFSS RFF. No. Mnol 1 - an. py. N 0 78 D R21 Pressure Transmitter Reactor 77 No No 155 Nol 1-Hn-t.T-N0804 R21 Level Transmitter peactor 77 No Yee 155,l$$4 M001 1 - H h-l.T-N 0 8 0 H R2l Level Transmitter peactor 77 No Yes 155 M001 1 - H H = t.T -N 0 8 0C B21 Level Transmitter peactnr 77 No Yes 155 r M001 l-DH-LT-NOROD B21 Level Transmitter Peactor 77 No Yes 155 155,1 W A Moot 't-SM-LT-NORIA R21 Level Transmitter peactor 77 No Yes MODI l-SM-LT-NO9IB R28 level Transmitter Reactor 77 No Yes 155, JQ A MODI I -SM -l.T-N O S I C H21 Level Transmitter Peactor 77 No Yes 155 Mnot I -SM -l.T-N04 t D R28 Level Transmitter Reactor 77 No Yes 155 M001 1 - HR -l.T-N O R $4 p23 level Transmitter peactor 77 Yes Yes 155 Mont 4 -HH-1.T-NO A 5R R21 Level Transmit t er Reactor 77 Yes Yee 155-Mont I-sp-PT-N0904 H28 Pressure Transmitter Reactor 77 No No js5 (WA Moul l-Bn-PT-N090B 828 Pressure Transmitter Reactor 77 No No .355 ' Moni I-PD-PT-N090E B21 Pressure Transmitter peactor 77 No No 155 M001 1-HB-PT-N090F B21 Pressure Transmitter Reactor 77 No No 155,1$5A Mool 1-nF-PT-N090J 821 Pressure Transmitter peactor '77 No No 155 M001 l-HF-PT-N090E B21 Pressure Transmitter ~peactor 77 No No 155,' MODI l-MF-PT-N090N H21 Pressure Transmitter Reactor 77 No No 155,155% M001 l-BF-PT-N090P H21 Pressure Transmitter Reactor 77 No No 155,g g M005 1-BR-LT-N091A 921 Level Transmitter peactor 77 Yes Yes 155 15 5,, MyA M001 1-SD-LT-N0915 R21 Level Transmitter Beactor 77 Yes Yes M001 1-Sn-LT-N091C R21 Level Transmitter peactor-77 No Yes 155,lSTA~ M001 1-SB-LT-N091D 821 Level Transmitter peactor 77 No Yes 155 MODI l-SR-LT-N091E B21 Level Transmitter peactor 77 No 7es 155 M004 l-SM-LT-N091P 828 Level Transmitter Peactor 77 No Yes 155 M001 l-SB-LT-N09tC 821 . Level Transmitter Reactor 77 No Yes 155 155,l$ % Cool 1-Sn-LT-M091N 821 tevel Transmitter Reactor 77 No Yes Q001 l-R9-PT-N0944 B21 Pressure Transmitter peactor 162 No No 155 NdDl l-86 *PT* NOM D B28 N llTM RE Vo lgg" Mcol I B& pr nloMw B2r N N6W m Ig7 i Ame ndmen t 14 I t I ..m m. ~ _.. _ ...m__.m -~ ~-r - ~,. .,_m m --. - g _,4 m m _.m... 4. e, i - ? 1/06 NCGS FSAp Page 34 of !!9 TARLE 3.ll=5 EOulPMENT SELECTED FOR NApSH ENVIRONMENT OuAttrtCArloM P6tp M-42-1 NYSTtM3 NUCIEAR ho ll.Elt VESSEL INSTRU HR PAM TMI ArYION 10 No. LOCATION EQUIP. Pl>w EQulP. P.O. _ Note 15) M PL NO. COMPONENT RLDG. FIEV. NOTE (1) NCIE (2) E'ESS REF. NO. M001 l-RH-PT-N094n B21 Pressure Transmit ter peactor 162 No No 155 M001 3-PD-PT-N094C R21 Pressure Transmitter peactor 162 No No 155 l M001 I-RH-PT-N094D B28 . Pressure Tronomitter peactor 162 No No 155 Mool 1-pH-PT-N094E H28 Pressure Transmitter ' Reactor 162 No No 155 Moni I-HH-PT-N094F B28 Pressure Transmitter peactor 162 No. No' 355 'Mont 1-pn-PT-N094C B28 Pressure Transmitter peactor 162 No No 155 M001 1-pp-PT-N094H B2l Pressure Transmitter peactor 162 No No 155 M001 1-SN-LT-N0958 821 Level Transmitter peactor 77 No. Yes .355 . Cont l-SN-l.T-N095D R23 Level Transmitter peactor 77 No Yes 155 ( M001 1-an-LT-N097D B21 - Level Transmitter peactor 77 No No 155 1. MOUI l-nn-LT-N097N R21 t.evel Transmitter peactor 77 No No 155 M005 1 - H R - l.T-N4 0 2 A M23 Level Transmitter peactor 77 No No l l e, 85TA M 001. I-BR-l.T N402R 921 Level Transmitter peactor 77 . No No ll4 i M004' 3-Dn-LT-N402E B2B . Level Transmitter peactor 77 No No 114,8 SEA lle 1 M004 3-hn-LT-N402F B21 Level, Trenami tter peector 77 No No MODI l-en-PT-N403A p23 Press Transmitter peactor 77 No No 110,15 5 4 M005 1 - H M - PT-N 4 0 ln 321 Press Transmitter peactor 77 No No 110 M001 3-pn-PT-N40]E R28 Press Transmitter peactor 77 No No 110,l55A 1! 0001 1-R D-PT-N 4 0 3 r B28 Press Transmitter peactor 77 No No !!0 Moni I-Rn-pE-12Dl93-pil Radiation Elemt. Reactor 120 Yes No 107 Mont I-an-RE-13n191 n!! podiation Elemt. Reactor' 120 Yes No 107 Munt .3-sn-pE-14Dl9) Ril Radiatton Elemt. Reactor 120 Yes No 107 s MODI l-An-DE-15Dl91 Bil padiation Elemt. Reactor 120 Yes No 107 Mont 1-kn-RE-16Dl93 Bil Radiation Elemt. Peactor 120 Yes No 107 i M001 l-RM-DE-2lDl9) Dll Radiation Elemt. peactor 120 Yes No 107 MOOR l-MM-pE-22Dl9) Bil' padiation Elemt. Reactor 120 Vee No 107 j' i i Amendment 14 L I i L t -. - _ _ _ ~.... _,.. _ m m__,_._ ..__m_ m_ _. =. -. __m., (? i 1/pe; j NCCS FSAR page 37 of 119 TAplE 1.83-5 EQUIPMENT SELECTED ~FOR HAPSH FNVIpoNNFNT OUALIFICATION P&lD M-43-1 4 575TI M s REACTOR ptCIRC SYS MH PAM TMI ACTION g I D No.. LorATION FOUIP. PLAN EQUIP. P.O. NOTF (5) NPL NO. COMPONENT plDC. F t.I V. NOTEJI) NOTE 12) FFSS DFF. NO. COnl I-nn-FT-N0144 pil Flow Trans. Reactor 77 Tes No 155 g 574 1 Nool 1-pa-FT-N0145 mit Flow Trans. Reactor 77 No No 155 M001 l-en-FT-N014C H31 Flow Trane, peactor 77 No No 155 g N001 1-pH-FT-NO]4D p)! Flow Trans, peactor 77 No No 155,l5CA N008 3-hR-FT-N024A pli Flow Trans. Peactor 77 Yes No 155 M9ul 1-en-FT-N024R R33 Plow Trans. Reactor 77 No No 355 M001 3-pH-FT-M024C B3B Flow Trans. Reactor 77 No. No 355 Nont I-en-FT-Nud4D n1l Flow Trans. Reactor 77 No No 155 P lu lRO l-HF-HV-3800A Control Valve peactor 54 No No l }9 l P IG IDO l-pr-25-38004 Limit Switch peactor 54 Yes No 139 P 10 l00 1-pF-HV-3900R Control Valve peactor 102 No 139 P10340 1-RF-IS-3000m Limit Switch peactor 102 Yes No 139 J6010 1-en-SV-4310 Solenold valve peactor 121 No No 46 t J6010 3-pR-2S-4380 Positled Switch peactor 128 Yes No 46 JG 010 l-en-SV-4llt Solenotrl valve peactor 145 No No 46 JG010 l-BR-2S-4Jil Position Switch peactor 145 Yes No 46 1' 1 L 1 I i f l i I }. t ].V l. j. Armendment 14 l-I' i t t r -.. - ~ - -. - ~ _ _ - -. ~.. -. -... - - - - - n...,.. - ~ ~.... ~ - -.. -. - ~.,. ~ ~. ps t [ 1/96 NCGS FS As page 30 of Ilg TARLF 3.11-5 FOUIFNENT SELFCTFD FOR HApSH FNVIRONNFNT OUALIFICATION P6fD t N-44-1 SYSTIN: RFACTOR NATFD Cl. FAN-UP BG PAN TN! ACTION ID No. LOCATION FOUIP. PLAN FOUTP. r P.O. NOTF (5) NPL NO. COf4PONFNT plDG. FLFV. NOTF (1) NOTF (2) FFSS PFF. NO. { P1020 1-BC-HV-F001 Contr. Valve peactor 145 No No 132 P1020 I-pG-2S-F001 Limit Switch peactor 145 Yes No 132 P9020 1-PG-HV-F004 Contr. Valve peactor 145 No No 132 P102Q I-MG-2S-F004 Limit Switch peactor 145 Yes No 132 Pl020 3-RG-HV-F034 Contr. Valve peactor 77 No No 132 Pl020 1 -RG-2 S-F0 34 Limit S=ltch. Reactor 77 No No 132 Pluto I-nc-HV-F035 Contr. Valve peactor 77 No No 132 I3020 l-BG-25-F035 Limit Switch peactor 77 No No 132 PtolU 1-AE-HV-F0 39 Contr. Valve peactor 102 No ~ No 132 P in 20 1-AF-2S-F039 Limit Switch peactor 302 Yes No 132 N003 3-BG-FT-Nd62A G33 Flow Trane, peactor 77 No No 155 N001 3-pG-FT-N012D G13 Flow Trane, peactor 102 No No 155 N003 3-pG-FT-N036A G33 Plow Trana, peactor 77 No No 155 155, esrA N003 l-PG-FT-NO36D G33 Flow Trane. Reactor 77 No No N001 1-DG-FT-N041A C33 ' Flow Trane. Reactor 102 No No 155 Nuol I-BG-FT-N043D G13 ' Flow Trans. Reactor 102 No No 155 6 i 6 Amendment 14 I l l. li II s .m .~.m..=_...m. . ~.. .m ._m.u_. m_ i. _.. ~ r"- f$ i.. 4 1/96 HCCS FSAR rage 42 og 119 TABLE 3.11-5 / EQUIPME'NT SFIECTED FOR HARSH FNVIRONMFNT OtIALIFICATION Pa, I D M-47-l NYNTIM CONTROL. ROD DRIVE HYD-PART R j HP PAM TNI ACTION In NO. LOCATION FOUIP. PLAN FOUIP. P.O. NOTF (5) NPL NO, COMPONFNT RLDC. IIFV. NOTE (1) NOTF ( 2[__.__FFSS RFF. NO. Mont I-RD -I.S-N0 l in Cll level Switch Reactor 102 No No 151 -..i i.u. ,<m... i s, cii e_.._i e_ =25 L' : "5;55 Eri" in, u-5^3 ^;, - T Mira l I-HP -IS=NU $ N Cl3 Level Swatch Reactor 102 No No ll) -Mndt 1-RF-IS-N0llH Cll Level Switch Reactor 104 No No MaiU I l - nF-f.V-819 Cl3 Solenoid Valve Reactor l et 2 No No - 122 Typical of 18% Solenoid Valves i 5 l l

4 s

A le l i Ii l-4 i- 'i-Amendment le l* Ii li

i t'

6 r

u i? ( 10iI l u --__.7 -..,.-~n.- _ ~,. -., _...- _ .-.__..._..~._,.n_.- ,w__.. ... -, - - _.. - -. _ +....... f 1/86 ~; 1 HCGS FSAR pag, g) og glg TARIE 1.ll-5 a FQUIPMENT SFI FCTFD FOR HARSH FNVipONMFNT QUAL.IFICATION P6tD M-4 4 - 8 NV9f f Ms STANDMY 1.!QtflD CONTPOL BH PAN TM1 ACTION ID NO. IIK'ATION FOUIP. PLAN FQUIP. P.D. NOTt 15) MPL NO. CONPONFNT B LtC. FIFV. NOTF-(l) NOTE (28 FFSS RFF. NH. Mont IA-r-Jus C4 l C001 Pump Mot or peactor 162 No No 104 Mudt IR-P-20s C41 C001 . Pump Notor peactor 162 No No 104 Moul 1-itH-av-F004A C41 Emplosive Control Valve peactor 162 No No 102 J Moul I-PH-XV-F004R Cet Euploalve Control valve peactor 162 No No 102 t P itt l AQ I-HH-HV-F0064 Control Valve peactor 145 No No 139 Pl0)AO l -IIH - 2 S - P 00 6 4 f.imit Switch peactor 145 Yes No I l9 I P lp lay 1-BH-HV-F006R Control Valve peactor 145 NQ No 139 P 10 9 AQ l

ft H - 2 S - F00 6 R Limit Switch peactor 145 Yes No I)9 M'Hil 1-HH-PT-N0044 C41 Prese Trans.

- Peactor 162 Yes No 135 Mihil 1-RH=PT-N004R C4) Press Trans. Reactor 162 Yes No 115 MHol l-HH-LT-N0104 C48 Level Trans. Reactor 162 No . No ll) i Munt 3-DH-LT-N010R C41 . Level Trans. Reactor 162 No No !!) Muul I-Mn-La-N010f C41 Level Trans. Reactor 162 No No ll) Mnot I-PH-LT-N010F Cet level Trans. Reactor 162 ' No No ll) MC 38 1-BH-flS-S 4 A C414 Hand Sw. Reactor 162 No No 152 Mool 1-PH-HS-S4R CelA Hand Sw. Reactor 162 No No 152 i Mito l 1-PH-LT-N082 Cet Level Trane, peactor I62 veg No 18]t} I I I .) .I + 2' i 1 1 -1 4 Ame ndment 14 \\ $t ik 'f i 4 -. _ _.. _ _ _..= ~.--~ -. ~.. -. - _. g i: I/ff 6 NCCS FS A R Page 45 of 119 TABLE 1.ll-5 FQUIPMFNT SEIFCTFD FOR HApSH FMVipON4FNT QtlAl.lFICATION P&lD M-49-1 SYSTEM: RFACTop CCRE ISOL COOL. pp v PAM TMI ACTION ID NO. LOCATION FOUIP. PLAN FOUIP. P.O. NOTF I5) M Pl. NO. COM PONENT p l.lf.. flFV. NOTE J ) NOTE _j2) FFSM PFF. NO. P lu loo 4-FC-IS-F025 Position Switch peactor 54 No No 141 )- P90100 1-FC-SV-F026 Solenoid Valve peactor 54 No No 140 P l0100 1-FC-25-F026 Poettion Switch peactor 54 No No 141 Plale l*pD-HV-F0ll Control Valve peactor 54 No No 126 Pipl0 l-pD-2S-F0ll Lielt Switch peactor 54 Yes No 126 , Mnio I-tc-sv-FU5e Selepold Valve peactor 54 No No 44 Y ..-..ri-1,-;; A;m; m..

..m i M
L n,

P in lQ 1-fC-HV-F05v Control Valve peactor F7 No No 126 Plotg .3-FC-IS-F059 Limit Switch peactor 77 Yes No 126 [ P 10 lOQ 4-FC-HV-FU60 Control valve peactor 54 No No 138 'l3 P901QQ l-FC-IS-F060 Limit Switch peactor 54 Yes No 318 P901Q 4-FC-HV-F062 Control Valve peactor 54 No No 127 lInig 4-FC-IS-F062 Limit Switch peactor 54 Yes No 127 P ID i*.0 1-FC-HV-F076 Control Valve peactoa 100 No No 139 l , Pl05Q 3-PC-HV-F084 Control Valve peactor 77 No No 127 P30140 1-FC-2S-F076 Limit Switch peactor 100 Yes No 139 l Pitslo 'l-FC-IS-F004 Limit Switch peactor 77 Yes No 127 M001 3 - p D-F T-NUD I ESI F!nw Trene peactor 54 Yes No 155 1 M001 3-FC-PT-N007 E5I Press. Trene peactor 54 No No 106 MODI l-FC-LSH-Nol0 ESI Level Switch High peactor 54 No No 150 lj M005 1 - p D-PT-N 050 E51 Press. Trans peactor 54 No No 155 9 MODI l - p D-FT-N0 5 I E5I Flow Trane peactor 54 No No 155 M001 l-rC-Pus-NU57M t.)! Press. Diff. Trans peactor 77 No No 155 Nno! l-PC-PDT-N0570 E51 Press. Dif f. Trans peactor 77 No No 155 I' i .!li li i-E l !? p e s Amendusent le u I' I i l 5 I l '} l i i l l I 1/96 HCCS FS AR I"9' 4' "I II' TARLE 1.81-5 FOutrNrNT SFLFCTFD FOR NAPSH FNVIPONMFNT QUALIFICATION P&lD NYS TI M e ppACTOR CORE lSol COOL. M-49-1 Hit PAM TN! ACTION IO No. LOCATION EQUIP. PLAN EQUIP. P.O. NO7 5; (%) N P(. NO. COMPONENT BLDG. FIFV. NOTE (1) DNIT E (2) FF9S PFF. No, M001 l-FC-PT-N058R E55 Press. Trans Reactor 77 No No 155 M001 1-PC-PT-N058D E55 l'r e s s. Trans peactor 77 No No 155gt w A N003 3-SC-PT-N058F E5I Frese. Tr ane Peactor 17 No No 155 Nuol 3-SC-PT-N058N E51 Press. Trans Peactor 77 No No 1%% .fl0IO I-N D-t T-415 7 Pr ess. Tr ena. Reactor S4 No No Q J10!O l-9C-07-4158 Plow Trane. Reactor $4 No No 19 Amendment 14 i f,- ss \\ "\\ ) t I f 1/P6 HCCS FSAF p,g, gy og ggg TARIE 3.51-5 l ~FOtllPMENT SElFCTFD FOR NApSN FNVIRONMENT OffALIFICATfDN P&ID "~ NYNTIM: RENIDUAL NEAT RFMOVAL Sr$ TEM PC PAM TMI ACTION t il No. LOCATION EQUIP. P LAN FQUIP. P_. O. Nt}TF J 5j,__ MPL NO. COMf0NENT BLDC. EIFV. NOTEj l) NOTEj 7) FFSS DFF. NT). P lp lAQ l-HC-IS Fl464 Position Switch peactor 100 Yes No 143 P ID I AO l-HC-SV-Fl46M Solenoid Valve peactor 100 No No 140 Pl0 IAQ I-pr-2N-F146R Position Sultch peactor 100 Yes No 14l P f u LAO l*HC-SV-Fl460 Solenoid Valve peactor 100 No No 140 PlotAO 1 nCazS-rl46C Position Switch -Beactor 100 Yes No 145 P103AQ I-RC-SV-Fl460 Solenoid Valve peactor 100 No No 140 PIDIA0 1-RC-IS-F146D Position Switch peactor 100 Yes No 141 Mool lanC-FT-N013 Ell Flow Transmitter peactor 107 No No 106 Mont 1-ac-tT-Nol5A Fl! Flow Transmitter peactor 77 Yes No 155 Mn01 l-nc-FT-N015n Ell Flow Transmitter peactor 77 Yen No dQb Mool l-HC-FT-N015C Ell Flow Transmitter peactor 54 Yes No 155 M001 3-DC-FT-N015D Ell Flow Transmitter peactor. 54 Yes No 155 M005 1-PC-FT-N0524 Ell Finw Transmitter peactor 77 No No 155 8504 Mn0I l-pC-FT-N052n Ell Flow Transmitter peactor 77 No No 155 3 Mool 1-DC-FT-N052C Ell Flow Transmitter peactor 54 No No 155 N001 l-MC-FT-N052D EII Flow Transmitter peactor 54 No No 155 M001 I-pC-PT-N 0 51 A Ell Press. Trans peactor 54 No No 106 Mnol 8-PC-PT-N051n Ell Press. Trans peactor 54 No No 306 M005 l-BC-PT-N 0 51C Ell Press. Trans peactor 54 No No 106 Nool I-PC-PT-N053D Ell Press. Trans peactor 54 No No 106 M001 1 -nc-PT-N 0 5 59 Eli Press. Trans psactor 54 No No 155 Mont 1-pC PT-N055D Ell Press. Trans peactor 54 No No 155,8$$)k ' 155 M005 l-nc-PT-N055F Ell Press. Trans peactor 54 No No Mnot l-MC-PT-N055N Ell Fress. Trans peactor 54 No No 155 MH05 1-DC-PT-N0568 Ell Pr ess. Trans peactor 54 No No 155 Mool 1-HC-PT-N0560 Eli Press. Trans peactor 54 No No 155 M005 l-ec-PT-N 0 56 F Ell Press. Trans peactor 54 No mi 155 i Ame ndme nt 14 ..-._..m _m 1/P6 8 HCCS FSAR Page 54 of 119 - TARLE 1.ll-5 FOl'IPMENT SFLFCTFD FOR HApSM FNVIRONMFNT OuALIFirATION P&tD M-51-1 hvsffM RFsinuAL NFAT PFMOVAL SYSTFN RC PAN TN! ACT!DN I D NO. P.O. NOTEJ ) LOCATION EQUIP. PLAN FQ1t i P. N PL NO. COMPONENT DLDC. FLFV. NOTU l l NOT W 2l FFSS,_pF E NO,. Mool I-nC-PT-N0564 Ell Press. Trane peactor 54 No No 155 Mool I-nC-PT-N057 Eli Prese. Trans. Reactor 77 No No 106 { F# 01 3-DC-PT-N058A Ell Press. Diff. Trang. Peactor 77 No No 155 M001 1-HC-PT-N0598 Eli Press. Di f f. Trane. Reactor 77 ,No No 155,8W A Mnot I-pc-Py-N058C Ell Press. Di f f. Trans. Reactor 77 No No 155 N005 1-mC-PT-N058D Ell Prese. Diff. Trana, peactor 77 No No 155,lW A N001 l-DC-FDT-N0604 Ell Press. Diff. Trans. Reactor 77 No No 106 Nuol 3-pC-PDF-N060R Eli Press. Diff. Trane. Reactor 77 No No 106 t th48%9 8 hC-7E-N027A Temp. Elemt. Reactor 77 Yes No. 34 F6H159 l-MC-TE-N027R Temp. Elemt. Reactor 77 Yes No 34 ,8 % 60 1-pC-TE-4408 Temp. Elemt. Reactor 77 No No 4) J1050 1-PC-FT-4435 Flow Trane, peactor 77 No No 29 Pinto 1-nC-NV-4419 Control Valve peactor 77 No No 127 P1010 1-8C-25-4439 Limit Switch peactor 77 No No 127 I Jln10 l-SC-FT-4468A Flow Trans. Reactor 77 Yes No 29 Jinto I-MC-FT-44618 Flow Trane. Reactor 77 Yes No 29 86010 1-DC-FT-4462A Flow Trans. Peactor 102 Yes No 29 .41010 l-MC-FT-4462R F3tw Trane, peactor 77 Yem No 29 b 1: o Lt t I Amendment 14 i [ t t J _ _ _... - ~. -. ~ . _.. ~. - - - _. _. -, - -........ .... ~ _ _..... -. - -. - - .r ' i 3/86 Nccs rsas rege se of Ile TABIE 1.31-5 EOurPMENT SELECTED PON NARSH FNVIPONMFNT OUALIFICATION 'P6ID SYSTIN: CORE srpAv N-52-1 RE 4 PAN TN! ACTION ID NO. LOCATION EQUIP. Pl.AN EX)tlIP. P.O. NOTE (5) NPL NO. COMPONENT RLDC. FIFV. NOTy_I l NOTE (21 EESS.NEP, NO. N001 3-BE-FT-N05tA Et= rene. peactor 54 No No 155 N001 l pr-rf-N051n E23 Flow Trane, i dp g peactor 54 No No 155 N004 3-RE-PT-N0544 E21 Press Trane, penetor 54 No No 155. fwA

001 1-st-PT-N0548 E21 Press Trang.

Peactor 54 No No 155;iSTA Nr 08 l-RE-PT-N0558 E23 Press.' Trane. Reactor 54 No No 155 i R001 1-DE-PT-N0550 E2I Press. Trans. Peactor 54 No No 155

001 1-ME-PT-N055r E21 Press. Trane, peactor 54 No No 155 Mont 1-pr-PT-N055H E21 Press. Trane, peactor 54 No No 155 N001 3-RE-PDT-N056 E21 Press. Diff. Trane, peactor 77 No No 3 55, tWA g

i 5 f r b i 'I l Amendment 14 f b 1/fl6 HCCS FSAR Page 61 of !!9 TABLE 3.18-5 l EQUIPNENT SELtCTFD FOR MARSH FNVIPONMrNT OUA1.IFICATION P6ID l N-55-8 SYSTFM HIGH PRt'NS COOL. INJFCTION HJ 4 PAN TNI ACTION 'ID NO. LOCATION FOUIP. PLAN FOUIP. P.O. NOTF 1%) NPL NO. COMPONFNT R LDf;. FIFV. NOTE (1) NOTE (2) FFSS REF. NO. Pl020 1-FD-HV-F001 Control valve peactor 54 No No I)) P3020 1-Fluts-F001 Limit Switch peactor 54 No No III b Pl020 l-FD-HV-F002 Control Valve peactor 102 No No 132 l P)d20 l -F D-IS-F00 2 Limit Switch peactor 100 Yes No 132 Pl020 FD-HV-F0 01 Control Valve peactor 102 m) No 132 P3020 1-FD-IS-9003 Limit Switch peactor 102 Yes No 132 P1010 1-nJ-HV-F004 Control Valve peactor 54 No No J26 P30lO l-RJ-IS-F004 Limit Switch peactor 54 No No 126 I' l O 2Q 1-RJ-HV-F006 Control valve peactor 102 No No 1)) t P3020 1-BJ-IS-F006 Limit Switch peactor 102 Yes No 13) P1020 1-RJ-HV-F007 Control Valve peactor 54 No No I)) ( P)020 1-RJ-IS-F007 Limit Switch peactor 54 No No 33) Pl02Q l-BJ-HV-F000 control valve peactor 77 No No 33) r P3020 5-RJ-3S-F000 Limit Switch peactor 77 No No I)) PIO2O l-A P-HV-F0 l l Control Valve peactor 77 No No 133 P3020 1-AP-IS-F0ll Limit Switch peactor 77 No No I)) P1020 1-8J-HV-F012 Control valve peactor 54 No No I)) i PID20 1-RJ-IS-F012 Limit Skitch peactor 54 Yee No 13) P 301CO l-F D-S V-F0 2 0 Solenoid Valve peactor 54 No No 140 P 30 LOO l-FO-IS-F020 Position Switch peactor 54 No No 141 P10 lOQ I -F D-SV-F0 29 Solenoid valve peactor 54 No No 140 P303DO l-FD-IS-F029 Poettion Switch peactor 54 No No I41 P90I0 5-RJ-HV-F042 control Valve peactor 54 Yes No 126 P301Q I-RJ-IS-F042 Limit Switch peactor 54 Yes No 126 36080 1-FD-9V-F054 Solennitt valve peactor 54 No No 44 M n:m

;0,:

L _ ;. ;m..

_ ; n ;.

n _ ; L. ', 0 P)040 4-FD-HV-F074 Control Valve peactor 77 No No 126 ' Amendment 14 .m .____m_._. [ g 1/P6 HCCS FSAR Page 62 of 139 TABLE 1.11-5 EQUIPMENT SFLFCTED FOR HAPSN l'NVIPONMFNT OUALIFICATION P6fD SYSTfMs HIGH PPESS COOL. INJECTION M-55-1 BJ PAM TMI ACTION ID No. LOCATION FOUIP. PLAN FOUIP. P O. NOTE (5) NPL NO. COMPONFNT RLDG. EtFV. NOT Q ll NOTE _(2) FFSS RFF. NO. g P30lO l-FD-IS-F071 Limit Switch peactor 77 Yes No 126 P10lO l*FD-HV-F075 Control Valve Reactor 77 No No 127 i P3010 1-FD-IS-F075 Limit Switch Reactor 77 Yes No 127 P1010 l-FD-HV-F079 Control valve Reactor 77 No No 127 P3010 1-FD-IS-F079 Limit Switch Reactor 77 Yes No 127 P10 3AO l-FD-HV-F300 Control Valve Reactor 100 No No 139 P101AQ 3-FD-IS-FIDO Limit Sultch Reactor 100 Yes No 139 Mont I - F D-FT-N0 0 8 E45 Flow Trane, peactor 54 Yes No 155 MODI l-FD-PT-N013 Eel Press. Trene. Reactor 54 No No 106 N001 1-FD-LSN-N014 E41 Level Switch M10h Reactor 54 No No 150 M005 1-RJ-PT-N050 E41 Press. Trane. Reactor 54 No No 155 i. MODI l-BJ-FT-N058 E43 Flow Trans. Reactor 54 No No 155 l 8:001 1-F D-PDT-N 0 57 A E45 Prosa. Diff. Trans. Reactor 77 No No 155 i K001 1-F D-PDT-N05 7C Ett Press. Diff. Trans. Reactor 77 No No 155 ^ N001 1-FD-PT-N050A E41 Press. Transmit ter Reactor 77 No No 155 M001 1-FD-PT-r059C Ett Press. Transmitter Reactor 77 No No 155 M005 3-FD-PT-Nn58E E41 Press. Tronomitter Reactor 17 No No 155 N003 l-F D-PT-N 0 50 G Eel Press. Transmitter Reactor 77 No No 155 lWA E005 1-BJ-LT-N0624 Ett Level Transmitter peactor 54 No No 120 j Ecol I-DJ-LT-N062E Ett Level Trans. Reactor 54 No No 120 9 J1010 1-sJ-PT-4771 Press Trans. Reactor 54 No No @ IssA J1080 1-BJ-LT-4001 Level Trane. Reactor 54 Yes Yes 28 P30340 3-BJ-NV-4003 Control valve Reactor 54 No No 139 P301&O l-RJ-IS-4001 Limit Switch peactor 54 Yee No 139 P101AQ 1-BJ-HV-4004 Control valve Reactor 54 No No 139 P103A0 I-nJ-IS-4004 Limit Switch peactor 54 Yes No 139 M001 1-RJ-LT-4005-1 Level Trans. Reactor 54-Yes No 155 i j Amendment 14 i i t s 4 3/86 NCC9 FSAR Page 63 of 119 TARLE 1.18-5 EOUIPMENT SELECTFD FOR MARSN FNVIPONMENT OttaLIFICATION P61D N-55-1 SYSTFM N ICJt PRFSS COOL. INJFCTION RJ PAN TNI ACTION ID NO. LOCATION EQUIP. PLAN fX)UI P. P.O. NOTF 15) M PL NO. COMPONENT RLDG. ELFV. NOT U I) NOTE (21 FFSS RFF. NO. C004 6-NJ-LT-4905-2 level Trane. Reactor 54 No No 155,lTTA P in )QQ 1 -RJ -f4V-4 8 6 5 Control Velve Reactor 54 No No 139 P1034Q 1-RJ-25-4865 Limit Switch Reactor 54 Yee No 1 39 P in )RO l-ILI-NV-4966 Control Valve Reactor 54 No No I)9 PtolQg I-RJ-25-4966 Limit Switch Reactor 54 Yes No 139 Pip 2O l-RJ-HV-9278 Control valve Reactor 102 No No 1)) P1020 l-RJ-2S-9278 Limit Switch Reactor 102 Yao No 133 s-Amendment 14 l / \\, f I ./ 1/R6 NCCS FS A p Page 64 of 819 TABLE 3.11-5 FOUIPMENT SELECTFD FOR. MARSH FNVIPONMENT QUALIFICATION P6fD NYSTFMs NPCI Pt'MP TURRINE M-56-1 FD PAM TMT ACTION I D NO. IN ATION EQUIP. PLAN FOUIP. P.O. NOTFJ5) MPL NO. COMPOMFMT RLDC. FLFV. NOTE (1) NOTF (2) FFSS pFF. NO. R004 20-S-2ll E41-C002 MPCI Turbine peactor 54 No No 123 M005 lo-P-233 E41-C002 Auu. Oil Pump Motor peactor 54 No No 123 M001 80-P-235 E41-C002 MPCI Vec Tank Cned. Pump peactor 54 No No 121 E001 10-P-214 E41-C002 MPCI Ctand Seal Cord. Vac. Pump peactor 54 No No 123 MUDI No Tag No. E41-C002 MPCI Turbine Governor peactor 54 No No 123 R0320 IA-P-228 Jockey Pump Notor. ECCS peactor 54 No No 80 i J6010 1-FD-SV-F025 Solenoid Valve Peactor- $4 No No 44 J60lO l -F D-IS-F0 25 Poettion Switch peactor 54 No No 45 Plo)AQ 1-F D-S V-r0 26 Solenoid valve peactor 54 No No 140 P103AO I-FD-35-F026-Poettion Switch peactor 54 No No 141 P103AO l-RJ-HV-F059 Control Valve peactor 54 No No 138 P10340 3-DJ-IS-F059 Limit switch peactor 54 No No 138 R003 1-pJ-PT-M052 Ett Press. Trane, peactor 54 No No 106 R001 3-NJ-PT-N053 Ett Press. Trane, peactor 54 No No 155, lSSA 155 MODI l-FD-PT-N0554 Ett Press. Trans. Reactor. 77 No No , ISS A e i 1 Amend = nt 14 L i i i _n 4 4 ) .1/R6 NCCS FSAR Page 65 of 119 TABLE 3.11-5 FOUIPMENT SELFCTFD FOR NARSH FNVlpONMENT OtlAl.lFICATION P&lD l M-56-1 S YSTD M s NPCI Pt'MP TUpplNE FD PAM TNI ACTION i ID NO. LOCATION FQUIP. PLAN FQUIP. M1.. NOTE (5) MPL NO. COMPONENT plDG. FIEV. NOTE (1) NOTE (2) FFS S., p(F. NO t M001 3 - t D-PT-N055C E45 Press. Trans. Reactor 77 No No 155, l5% M003 4-PD-PT-N055E E4l Press. Trane. Reactor. 77 No No 155 M001 3-FD-PT-N055G E45 Pr ese. Trans. Reactor 77 No No 155,l M Q001 l-FD-PT-N056A E45 Pro s. Trane. Reactor 54 No No 155 M001 1-FD-PT-NOS6E F41 Prese. Trane, peactor 54 No No 155 M001 3-FD-Fv-4879 E41 Flow Control valve Actuator peactor 54 No No 12) Mool 3-FD-Fv-4880 E43 Flow Control Valve Actuator peactor 54 No No 123 M001 1-FD-LSH-4890 Ett Level Switch Migh peactor 54 No No 32) .31010 3-pJ-PT-4891 Press Trans. Reactor 77 No No 29 0001 1 - F D-l.S L-4 9 0 ) E41-C002 Level Switch Low peactor 54 No No 123 M001 1.FD-PSH-4905 E43-C002 Press Switch High peactor 54 No No 32) i M001 3-FD-25-4907 E48-0002 Limit Switch peactor 54 No No 12) MODI l-FD-PSL-4908 E41-C002 Press. Switch Low peactor 54 No No 123 1 MODI l-FD-TS-4909 E48-C002 Temp Switch peactor 54 No No $2) N001 3-FD-PDSM-4910 E41-C002 Press Diff. Swltch High peactor 54 No No 123 M001 1 -F D-LS H L-4 912 E48-C002 Level Switch High Low peactor 54 No No 123 N001 1-FD-PS-4913 E41-C002 Press. Switch peactor 54 No No 32) M003-1-FD-SE-4919 E41-C002 Speed Elemt. Reactor 54 No No 32) .l P9014 l-F D-HV-4 9 2 2 Contr. Valve peactor 54 No No I )9 P i l I Amendment 14 i e . _ _ _ ~,.. _ ~. - .., ~. - -.. .. ~ 4 \\ y 1/96 HCGS FSAR Page 66 of 119 TABLE 1.II-5 EQUIPMENT SELFCTED FOR HARSH FNVIRON4FNT 00ALIFICATION P6lD N-%7-1 SYSTENs CONT 4fN4ENT ATMOS. CONTRL. GS PAN TNI ACTION I n NO. LOCATION EQUIP. Pl.AN EQUIP. P.O. NOTE 15) NPL NO. COMPONENT RLDG. ELEV. NOTE (1) NfYT E ( 2) FFSS RFF. NO. J1590 1A-C-200 H2/02 Analyser PNL A Reactor 162 No Yes 30 J3590 IH-C-200 H2/02 Analyser PNL R Reactor 162 No Yes 30 J1590 BC-C-200 H2/02 Analyser Heat Trace PNL Re act or 162 No No il ^ J1590 ll>C-200 H2/02 Analyser Heat Trace PNL Reactnr 162 No No 31 Ji%90 l HIS-T t:-0 3 51 Te merature Element Reactor 145 No Yes )! J1590 ldis-TE-0152 Temperature Element peactor 145 No Yes II Jt590 IdaM-TE-015) Temperature Element Reactor 77 No Yes 31 J3590 1415-TE-0354 Temperature Element peactor 162 No Yes )) j J1599 l-CS-TE-0355 Temperature Element Reactor 102 No Yes 31 i J)S90 I-Ch-TE-0156 Temperature Element peactor 162 , No Yes it i J1590 1415-TE-0)S7 twaperature Element Reactor 162 No Yes 11 J1590 , ' 142 5-T E-0 3 5 8 Temperature Element Reactor 162 No Yes 34 1 590 Id;S-TE-0159 Terparature Element Reactor 162 No Yan )) n50 14:s-SV-4950 Solenold Valve Reactor 145 No No 145 4 P30%O !+GS-25-4950 Position Switch Reactor 145 Yes No 146 PlotAO 14;S-ttv-4 951 Control valve Reactor. 145 No No I)9 P10 BAO l -G s-2 5-4 9 51 Limit Switch Reactor 145 Yes No 139 P)050 1-CS-SV-4952 Solenoid Valve Reactor 145 No No 145 8 Pl0SO 14;S-IS-4952 Position Switch Re act or 145 Yes No 146 PIG 3AO . I -CS -HV-4 9 5 54 Control Valve Reactor 132 No No 119 P IG ) AO IMGS-IS-49554 Limit Switch Reactor 132 Yes No 119 P101A0 1 -C S-HV-4 9 5 5R Control Valve Reactor 162 No No 139 P 301 A0 1-GS-IS-4955n Limit Switch Reactor 162 Yes No 119 Pl050 4-CS-SV-4956 Solenold Valve Reactor '102 No No 145 1 P3050 l < S-IS-4 9 5 6 Position Switch Reactor 102 Yes No 146 P1050 l-GS-SV-4958 Solenoid Valve Reactor 77 No No $45 P1050 1 -G S-I S-4 9 5 8 Position Suitch Reactor 77 Yes No 146 i P303A0 1 -G S -NV-4 95 9 A Control Valve Reactor 77 No No 119 JMQ (.G-TC-O M Tei,,lm47kes em RsACTot. l(.2. No Mrs 3ll Amendment le .s I ,m 1 g y +. ve ' i t s o 1/P6 HCGS FSAR Page 7 tp og I!9 TARLE 1.18-5 EQUIPMENT SELECTED FOR HARSH FNVIRONMENT OUALIFICATION P6tD Mf'iTFM s CONTAINMFNT ATMOS. CONTRL. US PAN TMI ACTION ID NO. LOCATION EQUIP. PLAN EQti t P. P.O NOTE 15) M P L NO. COMPONENT R LIW'. ELEV. Nf4E (1) NOTE QI FESS REF. NO. g Ji%90 l-GM-SV-5086Al Solenold Valve Reactor 162 No No 30 JI59g IM;S-sv-5086A2 Solenoid valve Reactor 162 No No 30 JI590 l <.S-SV-5086RI Solenold Valve Reactor 162 No No 10 .31%90 14;S-SV-508682 Solenold Valve Reactor 'l62 No No 10 J5590 14;S-HS-5087A Hand Sw. Reactor 162 No No 30 J1590 l-GS-SV-5087Al Solenold Valve Reactor 162 No No 30 JI590 IH;N-SV-5087A2 Solenold Valve Reactor 162 No No 30 JI590 I d;S-HS-50 8 7R Hand Sw. Reactor 162 No No 30 .85590 id;%-SV-5087HI Solenoid Valve Reactor 162 No No 10 .35590 1-GS-sv-5087n2 Solenoid Valve Reactor 162 No No 30 JI590 l<.5-TAH-50924 Temp. Alarm High Reactor 162 No No 30 JI%90 1 -G S-T A L-509 24 Temp. Alarm low Reactor 162 No No 30 .fl590 1-GM-TAH-5092R Te q. Alarm High Reactor 162 No No 10 JI590 l d:S-T A L-509 2R Temp. Alarm tow Reactor 162 No No 10 JI59Q I4;S-ThN-5092A Ta g Sw. Nigh Reactor 162 No No 10 J159Q 1415-TS L-509 2 A Temp. Sw. Low Reactor 162 No No 30 JI590 l <.N-TSH-5092n Temp. Sw. High Reactor 162 No No 10 JtS90 l -CN-TS L-50 9 2H Temp. Sw. tow Reactor 162 No No 10 J 1590 l <.S-FAl.-50944 Flow Alarm High Reactor 162 No No 50 J1590 l <.5-F A L-50 9 4 8 Flow Alarm Low Reactor 162 No No 30 Jt590 l <.S-PDS-5094Al Press. Diff. Sw. Re ac tor 162 No No 10 J1590 l <.S-PDS-5094A2 Pres 9. Diff. $w. Reactor 162 No No 10 JIS00 4 -G S-P DS-509 4 R I Press Diff. Sw. Reactor 162 - No No. 30 J1590 l <.S-PDS-5094R2 Press Dif f. Sw. Reac tor 162 No No 30 P lu l AO 14;S-HV-5741A Control Valve Reactor 132 No No 139 P 10 lGO 14;S-2S-5743A Limit Switch' Reactor 132 No No 119 P10100 1 -GS-NV-5 7418 Control Valve Reactor 132 No No 139 P301A l <.S-IS-57434 Limit Switch Reactor 132 No No 119 k i - - --,0

T:::t ;

l Amendment 14 T i I i ~ .+.- i b 1/86 HC1;S FSAR Page 78 of 189 TARLE 1.11-5 f EctflPMENT SEl.ECTED FOR H ARSH FNVIRONMENT Otf ALIFICATION P6tD F M-58-8 l SYNTEMI CONTAIN4ENT HYDRUEjEN RECONdlNATION SYSTEN l PAM TNI ACTION i n Nt). LOCATION EQUIP. PLAN EQUIP. P_. O. NOTE 15) MPL NO. COM PON ENT

RLDG, ELEV.

NOTE II) NOTE (2) FESS REF. NO. QD47A0 lA-S-205 Hydrogen Recombiner peactor 162 No No 52.5).56 thris 64 MG4700 IR-N-205 Hyttrogen Recombiner Reactor 162 No No 5 2,51.5 6 t hr u 6 4 M047AO .lA-C-215 Power Panet Reactor 162 No No 54 QU 4 700 IA-E-235 Hester peactor 162 No No 56 Q04700 IA-V-235 Fan Motor peactor 162 No No 52 j QUE700 lH-C-215 Power Panel peactor 162 No No 54,W,6Q,67 N04740 lH.F.- 2 3 5 Heater Reactor 162 No No 56 M04700 IH-V-215 Fan Motor peactor 162 No No 52 P1010 1415-HV-50%04 Contr. Valve peactor 145 No Yes 'l27 Pl010 l-CS-25-505HA Limit Switch peactor 145 Yes No 127 P10lO Id;5-HV-50500 Contr. Velve peactor 102 No Yes 127 P1010 IdiS-IS-5050n Limit Switch peactor 102 Yes No 127 P3010 INGS-HV-50524 Contr. valve peactor 145 No Yes 127 P1010 1-CS-IS-50424 Limit Switch peactor 145 Yes No 1.? ? P1010 1415-H V-505 28 Contr. Valve peactor 102 No Yes 127 P9010 1-05-2S-50528 Limit Switch peactor 102 Yes No 127

P3010 l <. 5-H V-50 51 A Contr. Valve Reactor 77 No Yes

$27 P1010 I-GS-25 50514 Limit Switch peactor 77 Yes No 127 P1680 l d;S-stV-505 3n Contr. Valve peactor 77 No Yes i 'PldIO l<is-25-5051n Limit Switch peactor 77 Yes No lj P9010 l-GS-HV-50544 Contr. Valve peactor 77 No Yes I?? P lo t o l-CS-IS-50544 I.imit Switch peactor 77 Yes" No 127 i P1080 1-CS-HV-50548 Contr. Valve Reactor 77 No

Yeg 127

,w P9010 1-GS-IS-5054n Limit Switch peactor 77 Yes No. 127 P In lRO l-Rf-HV-50554 Contr. Valve peactor 54 No No 119 P 10 lOQ I-NC-IS-5055A Limit' Switch peactor 54 No No 119 P101AO l-HC-HV-5055n Contr. Valve peactor 77 No No lit 7a i 1 Amendment 14 1 4 i 4 I t r R' .l [ 'd 2 I/00 HOGS PSAR Page 80 of 119 .j TARLE 3.15-5 4 l ggg g pM ENT SELECTED FOR H ARSH ENVIRONMENT QUALIFICAT!uN P& I D j NYNrFMs Mal 4 MTMAM IMH VLV NEAL SYN RP 6 P AN ' TN! ACTION in No. LOCATION EQUIP. PLAN EQUIP. P.O. NOTE,151,,_,, MP L NO. COMPONFNT RLDG. ELry. NOTE (I) NOTE 12) EESS REF. No. 1 4-P t0.l Au l -w r-H V-S e l5A Contral valve Reactor 102 No No 139 i Mont I-An-py-5R15A Preeg Trang. Reactor 102 Yes No 15 i P ld l AQ I-NP-24-5R15A Limit Mwitch Reactor 102 Yes No I)9 j P l0 lAQ I-EP-HV-5955n Control Valve Peactor 102 No no 139 M00% l-AR-PT-5915R Press Trang. Reactor 102 No No 15 i Pl0lAQ-l-KP-23-5495R Limit Switch Reactor 102 No No 139 P lo tAQ - 3-KP-HV-5416A Control Valve Reactor 102 No No 139 M001 l-AR-PT-54164 Prege Trang. Reactor 102 Yes No 155 ? P 10 9 AQ 3-WP-25-5816A Limit switch Reactor 102 Yee No 139 l Plq]AQ 1-EP-HV-5416H Control Valve Reactor 102 No No 139 annt 3-AR-PT-5R16R Press Trang. Reactor 102 No No 15 j P ID I AQ I -E P= 7 S-5 816 R f.imit Switch Reac t or 102 No No 139 .j P l0 3AQ 1 -E P-HV-5 R 17A Cnntrni Valve Reactor 102 No no 139 t M001 3-AR-PT-5817A Press Trang. Reactor 102 Yes No 15 P IG I AQ I-er-2s-5837A Limit Switch peactor 102 Yes No 13 1, Pint 4Q 1-aP-HV-5937R Control Valve Reactor 102 No No 139 4 i i ' i, l i, s ( 1 } 1 ? j Ame nde-ee n t 14 l t 4 'l. i I ? d6 f W- <j Q3 1 I 1/P6 } HCCS FSAR Page 8) of 189 ) TARLE 3.11-5 4- '8 EQUIPMENT SELECTFD FOR HApSH FNVIRONMENT OUALIFICATION P6ID M-83-1 SYSTFM REACTOR BLDC. StfPPLY CONTROL. DIAC. l CR PAM TM1 ACTION ID NO. IDCATION EQulP. PLAN EQUIP. P.O. NOTE 15l MPL NO. COMPONENT Rt.DC. F l.FV. NOTE (I) NOTE (2) FESS BEF. NO. M7860 . LAC-04) Heater Control Panel peactnr I)2 No No 100 f63 o M7R60 IPC-043 Heater Control Panel peactor 178 No No 100 ;lb3 MiW60 100-043 Heater Control Panel peactor 132 No No 100,163 j M7860 I tt-0 4 ) Heater Control Panel peactor 162 No No 100, f 63 M7860 IAC-044 Heater Control Penel peactor 162 No No 100,l63 81,l63' M7p60 BPC-044 Heater Control Panel peactor 17A No No 100 MPalQ IA-VH-208 Unit Cooler peactor 54 No No MillO lR-VH-208 Unit Cooler peactor 54 No No 8t M7810 lA=VH 209 Unit Cooler peactor 54 No No 81 M7 BIO IR-VM-209 Unit Cp ler peactor 54 No No 81 M7tl0 IA-VH-210 Unit Cooler peactor 54 No No 81 MillQ IR-VH+440 Unit Cooler peactor 54 No No St MFilQ IC-Vtt-210 Unit Cooler peactor 54 No No 81 M7tl0 ! D-VH-210 Unit Cooler peactor 54 No No 81 MillO lE-VH-210 Unit Cooler peactor 77 No No 81 j M7Il0 Ir-VH-280 Unit Cooler peactor 77 No No 81 a M7810 IG-VH-280 Unit Cooler peactor 54 No No 88 hi n7tlQ lH-VH-210 Unit Cooler peactor 54 No No 81 8 MPIIQ IA-VH-2tl . Unit Cooler peactor S4 No No 81 I M 7 t IQ - lH-VH-Jll Unit Cooler peactor 54 No No St M11IQ IC-VH-218 Unit Cooler - peactor 54 No No 81 t M1130 ID-VH-Jll Unit Cooler peactor 54 No No Ol' M7tlQ IE-VH-2tl Unit Cooler peactor 54 No No 81 Mill 0 IF-VH-211-Unit Cooler peactor 54 No No 81 M7tlQ IG-VH-2tl Unit Cooler' peactor 54 No No 81 J M7tt0 IH-VH-231 Unit Cooler peactor 54 No No 81 M 7 t lO IA-V-21) Fan & E-H Actuator .peactor 132 No No 83 M 7190 19-V-21) Fan & E-H Actuator peactor 178 No No p1 4 l' Amendaeont 14 o h .. - - - - - ~. -.. - _. -. . _. - ~, ~. _... -. -. -. _ - - ~. - -. f-r.. l f ( -) 1 .q 1/96 NCGS FSAR TABLE 1.11-5 'Page R4 of 189 .a ,f EOu!PMENT SELECTFD FOR NARSH FNVlpONMFNT OUALIFICATION P6tD M-83-I j NYSTtMs RFACTUp.RLDC. SUPPLY CONTROL DIAC. 'cm 'l 8 PAN TMI ACTION 1 ID NO. LOCATION EQUIP. PLAN F.QU I P. I P.O. NOTF (5) M PL NO. COMPONENT BLDC. ELEV. NOTE (Il NOTE (2) FFSS REr. NO. 1j 887110 IC-V-283 Fan & E-H Actuator Coactor 132 .No No 83 M7110 ID-V-21) Fan & E-H Actuator peactor 162 No No 81 M7110 IE-V-213 Fan 6 E-H Actuator peactor 162 No No 81 M7130 IF-V-213 Pan 6 E-H Actuator peactor 178 No No 83 M7tio IA VH-214 Unit ~ Cooler peactor 102 No No 81 4 i M7810 i n-Vit-214 Unit Cooler peactor 102 No No 81 ,{ M7180 IC-VH-214 Unit Cooler peactor 102 No No 81 g } M7tio ID-VH-214 Unit Cooler peactor 102 No No 81 M700A0 IA-C-281 Unit cooler Ctrl Pnt. Reactor 102 No No 91,97,98 ) ,j M780A0 In-C-281 Unit Cooler Ctrl Pnt, peactor 102 No No 98,97,98 66-M780A0 BC-C-281 Unit Cooler Ctrl Pnt. Reactor 102 No No 91,97,98 {- M78000 ID-C-281 Unit Cooler Ctrl Pnt. Reactor 77 No No 91,9 7,9m A 66 i l I i Amendinen t 14 I l l i j y 4.' 0 I t 1/86 j HCCS FS A R Page 95 of 189 TAplE 3.11-5 j EQUIPMENT SELECTED FOR HApSH FNVlpONMFNT OUALIFICATION P61D SYSTIMt RFAC HLDG ERN CONTpL e N-84-1 i Cu I PAM TMI ACTION I ID No. LOCATION EQUIP. PLAN EQUIP. f P.jl.. NegE,(5) MPL NO. COM PON FNT ptDC. FIFV. NOTE (1) NOTE (2) EFSS REF. NO. M 7 p 00 - 140-045 Heater Control Panel peactor 145 No No 100. Ko3 M7060 IRC-045 Heater Control Panel peactor 145 No No 100 i l63 M 7 410 IA-V-206 Fan & EH Actuator peactor 145 No ko 83 l M7110 IM-V-206 Fan & FH Actuator peactor 145 No No 83 M72CO l-CU-SV-9414Al Solenoid valve peactor 178 No No 88 M 7 200 1-CR-SV-9414A2 Solenold valve peactor 178 No No 88 M72RO l-Cu-IS-9414A Poettion Switch peactor 178 Yes No 89 1 M72CO l-CU-SV-9414pl Solenoid valve peactor 178 No No 88 i Q72RQ 3-CU-SV-9454p2 Solenoid Valve peactor 178 No No 88 1 M7280 l-CU-2S-94148 Position: Switch peactor 178 Yes No 89 JJOBO l-CU-FT-9425A Flow Trane, peactor 145 No No 29 0700A0 1-CU-PDT-9425A Press. Diff. Trans. Reactor 145 No No 99 M700A0 1-CH-TE-9425A Temp. Elemt. peactor 145 No No 93 g M 7 3 70 1-CH-H D-9 4 2 5 A l Hand Damper Actuator poactor 162 No No 87 1 07t?O l-CU-HD-9425A2 Hand Damper Actuator peactor 145 No No 87 4 M7170 l-CU-F D-9 4 2 5 A 3 Flow Damper Actuator peactor 145 No No 87 Lj M7170 4 -CU-FD-94 25A5 riow Damper Actuator peactor 145 No No 87 g J)oIO l-CO-FT-9425M Flow Trans. Reactor 145 No No 29 'l Q7R000 1-GU-PUT-94255 Press. Diff. Trans. Reactor 145 No No 99 M70000 I-Cu-TE-9425R Temp. Elest. Reactor 145 No No 93 M787Q 1-CU-HD-9425BI Hand Damper Actuator peactor 162 No No 87 j M7170 1 -CU-H D-9 4 25 R 2 Hand Damper Actuator peactor 145 No No 87 M7870 l-GU-F D-94 25R ) Finw Damper Actuator peactor I45 No No 87 M7370 1 -CU-F D-9 4 2 5p5 Flow Damper Actuator peactor 145 .No No 87 M7830 1-CU-FD-94264 . Flow Desper Act uator peactor-145 No No 84 078000 1-CU-FT-9426A Flow Trans. Reactor 145 No No 99 i M70000 1-CU-FSL-9426Al Flow Sw. Low poactor 145 No No 92 j Q780A0 1-CO-PDT-9426Al' Press. Diff. Trans. Peactor 201 No No 99 g M78000 I-CU-FSL-9426A2 Flow Sw. Low peactor 145 No No 92 ] 070000 1-GU-FDT-9426A2 Press. Diff. Trans. Reactor 102 No No 99 i i l Ame ndme nt 14 I. 3 5 ) i t j 1/P6 NCES FaiAR Page !II of I19 TAnLE 3.11-5 + EOUIPNFNT SFLFCTED FOR M ARSEN FNVI RONNFNT OU ALIFICATION P&In N/A svstrN: m:NFRIC PAN TNI ACTION I D No. LOCATION EQUIP. PLAN EQHIP. P.O. 9H vit l %) VEN ty)R COM M)N E NT R LII;. E l.E V. Nr1T E (1) N(FT E ( 2) EfSS REF. NO. J 6 j El700 Rl4 Brand-Rem Co, Tr l, and Twina n t al Cable Note (3) N/A N/A 9 Elfoo pl6 Brand-Res Co, Trl, and Twinasial Cable Note ( 3) N/A N/A 9 Fl?UV RS8 Brand-Res Ce, Trl, and Twinesial Cable Note (3) N/A N/A 9 El?DO R59 Brand-Reu Co, Trl, and Twinesial Cable Note ( )) N/A N/A 9 El?OO R62 Hrand-Rex ' Co. Tr l, and Twinas ial Cable Note (3) N/A N/A 9 El?pa0 RMI' pochbestos Co, Trl, and T inesial Cable Note ( 3) N/A N/A 12,l2g Elfucy RM2 Rockbestos Co. Tr l, and Twinasial Cable Note (3) ) Fl1000 RN7 Rockbestoe 300V Shielded Inst, Cable Note (3) ' N/A N/A 12 N/A N/A 10 El10A0 Ro6 Rockbestos Co, Trl, and Twinasial Cable Note (3) N/A N/A 10 1 Fl? pao pr9 Rockbeston Co. Trl, and Twinaulat Cable Note (3) N/A N/A 10 f El7000 RGil Rockbestos Cu, Trl, and Twinasial Cable Note (1) N/A N/A. 10 El?000 m22 Rockbestos Co. Trl, and Twinasial Cable Note ( 3) N/A N/A 10 J El7000 MI) Rockbestos Co. Trl, and Twinasial Cable Note (3) N/A N/A ll,Q6 i .El704Q $110 Rockbestoe 600V Control / Power Cable Note ()) N/A N/A 11 El1000 SI4 Rockbestos Co Tr l, and Twinesial cable Note (3) N/A N/A II,QB { El7000 536 Rockbestos Co, Trl, and Twinaulat Cable Note (3) N/A N/A II,828 g2B El10A0 SIO pockbestos Co, Trl, and Twinaula! Cable ' Note (3) N/A N/A II, El?OCQ VE4 Rockbestos Co Trl, and Twinasial Cable Note ( 3) N/A N/A 12 A,826 Elfl0 104 Eaton 600V Shielded Instru Cable Note ( 3) N/A N/A !) El780 3 71 Eaton 600V, Shielded Instru Cable Note (3) N/A N/A I) Elfl0 Ex32 Eaton 600v Shielded Instru Cable Note (3) N/A N/A I) El?IO ' Do l Eaton 600v Shielded Instru Cable Note (3) N/A N/A j) El?lo D)4 Eaton 600V Shielded Instru Cable Note ()) N/A N/A j) El?t0 Do% Eaton 600V Shielded Instru Cable Note (3) N/A N/A I) Elfl0 ts e Eaton 600V Shielded Instru Cable Note (3) N/A N/A 13 1 1 Amendment 14 ? i' _. - -.. -.. ~. .= r~q p q t/06 NCGS FSAR Page 110 ot 119 i 'l TARLE 3.ll-5 EQUIPNFNT SELECTED FOR H ARSN ENVIRONMENT OUALIFICATION Palp N/A (. NY STI:4s.GrNERIC a PAN TNI ACTION .l - I D NO. 1OCATION EQUIP. PLAN E0ulP. P.O. NOTE (5) VENDOR COMPONENT RLtlG. Et.rv. NOTE (1) NOTE (2) EESR R F. F. NO. F433200-1-lF Rsychee Corp. Nigh Voltage Terminatione Note (3) N/A N/A 19 F4758%0-1-IF N-NCR Raychem Corp. Motor Connection Ett Note (3) N/A N/A 20 l F40479Q-l-IF NCS F-N RayChem Corp. Cable Breakout Kit Note (3) N/A N/A 21 F4147tO-1-lF Raychen Corp. Cable End Sealing Ett Note (3) N/A N/A 21 F4 85610-1-IP NPKS.4 PEN.NPRC Raychee Corp. Cable Splice Assy Note (3) N/A N/A 23 F4855tO-t-IF Raychem Corp. Thornofit Insulation Sys Note ( 3) N/A N/A 22 F582160-l=lF ' N-21009-O n Conas Corp. Electric conductor Seal Assy Note (3) N/A N/A 24 4 i 4._ F61553 N - 188% COMx h. Hccamen % Gute Any Nore(3) N/A N/A 33 4 - til17 .I- 'l f 1 f i t i i f-1 l t Amendiment 14 f t 4 I i i 1 ,/ O, ^ 5 q i i i 1 J l f, HCGS FSAR i 1/86 TABLE 3.11-6 Page 2 of 3 1; SAFETY-RELATED EQUlFWENT LOCATED IN A HARSH ENVIRONMENT EXEMPTED I FRO 4 ENYlRONMENTAL OUAllFICATION REQUIRENENTS i EQUIPMENT TAG NO. NPL No. DESCRIPTION REASON 1-A E-HV-F039 Motor Operated Valves These motor operated valves %are not quellfled for submergence 1-CB-HV-F071 I Motor Operated valves caused by a feedwater line treak In the steam tunnel. They have t:aen provided with l-KP-HV-58294,8 Motor Operated Valves primary and backup IE tus protectivo devices located in the herard free area, = f 1-KP-HV-4834A,8 Notor Operated valves j 1-kP-HV-5835A B Motor Operated valves lAPJD SDL643tD VAL 4/6Sl I l-KP-HV-5836A,B Motor Operated Valves 1-KP-HV-S837A,8 4 tor Operated Valves ^ ^ "" f ^' ? " "";7x ^;;r;;;; 7;;as ' *? ""-FM ?9 "-+e- ^- e? M Ye!re: - ' '" "" 'M?C t?r W:?:dMhn ' " "" T ^^. ? ht Cw tu %Im No Tag No. C11-F010 Pus 1 tion Switch These MANCO Ilmit switches perform no safety functions. Failure modes and effect No Teq No. Cll-Foll Position Switch analysis has shown that there are no possible fallure podes which can adversely No Tag No. C11-F180 Position Switch ef fect the IE power supply. No Tag No. Cll-F18I Position Switch i 1-BE-SV-F006A E21 Solenold Valve These solenold valves and position switches perform no safety functions. However, No Tag No. E21-F006A Position Switch because of their association with a IE poww supply, they have taen provided with 1-8E-SV-F0068 E21 Solenold Yelve primary and tackup protective devices. No Tag No. E21-F0068 Position Sultch I-BC-SV-F041A Ell Solenoid valve } 'No Tag No. Ell-F041A Position Switch j 1-BC-SV-F0418 Ell Solenold Yelve l No Tag No. Ell-F0418 Position seltch 4 1 8C-SV-r04 c Eli Solenold valve ] No Tag 2 Ell-F04tC Position Switch Nnendment 14 f j 1-AE-W-4iW HarteOsengn Wwg3 [ 1-KL-POV-5825A Sourmo VAuM a-kt-PDV-582KB htEmtb {/Atts HCGS FSAR 8/83 "g ' w) ~ ~ from its-drive withou6'distuebing the remainder of the control system. The bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel open. 4.1.3.2 Description of Control Rods M! A description of the control rods ishgIU34 in Section 4.2.2.1. l f 4.1.3.3 Sucolementary Reactivity Control The initial-and reload-core control requirements are met by use of the conbined effects of the movable control rods, supplementary burnable poison, and variation of reactor coolant flow. A description of the supplementary burnable poison is """=,\\ i n S ec t i on. 4. 2. Y 4.1.4 ANALYSIS TECHNIOUES 4.1.4.1 Reactor Internal Components Computer codes used for the analysis af the internal components are as follows: a. MASS b. SNAP (MULTISHELL) c. GASP d. NOHEAT e. FINITE f. DYSEA g. SHELL 5 j 4.1-5 Amendment 1 HCGS FSAR 8/83 (s e. Mechanical reactivity control permits criticality checks during refueling and provides maximum plant safety. At any time in i.ts operating history, the core is designed to be subcritical with any one control rod fully ' withdrawn. f. The selected control rod pitch represents a practical value of individual control rod reactivity worth, and allows adequate clearance between control rod drive (CRD) mechanisms below the pressure vessel for ease of maintenance and removal. 4.1.2.1.2 Core Configuration The reactor core is arranged as an upright circular cylinder containing a large number of fuel cells and is located within the reactor vessel. The coolant flows upward through the core. The core arrangement (plan view) and the lattice configuration are given in Section 4.3. 4.1.2.1.3 Fuel Assembly Description lR99W8\\ Descriptions of the fuel assembly and the fuel rods are\\w: r, in Section 4.2. f 4.1.2.1.4 Fuel Assembly Support and Control Rod Location A few peripheral fuel assemblies are supported by the core plate. Otherwise, individual fuel assemblies in the core rest on fuel support pieces mounted on top of the control rod guide tubes'. Each guide tube, with its fuel support piece, bears the weight of four fuel assemblies and is supported by a CRD penetration nozzle in the bottom head of the reactor vessel. The core plate provides lateral support and guidance at the top of each control rod guide tube. The top guide, mounted inside the shroud, provides lateral support and guidance for each fuel assembly. The reactivity of the core is controlled by cruciform control rods and their associated mechanical-hydraulic drive system. The control rods occupy alternate spaces between fuel assemblies. Each independent CRD enters the core from the bottom, accurately positions its associated control rod during normal operation, and 4.1-3 Amendment 1 I r i i HCGS'FSAR 8/83 4.1.4.3 Reactor Systems Dynamics I l The analysis techniques and computer codes used in reactor REM 4J5 systems dynamics are described in Section 4 of References 4.1-10, I 0F THE 4.1-10A, and 4.1-10B. .Section 4.4.4:e dire provideph ccm= ct: stability analysis for the reactor coolant system (RCS). '[ 4.1.4.4 Nuclear Encineerino Analysis r The analysis techniques are described and referenced in Section 3 of Reference 4.1-1. s 4.1.4.5 Neutron Fluence Calculations 4 1 Vessel neutron-fluence calculations are carried out using a one-dimensional, discrete-ordinates, Sn transport code with general anisotropic scattering. t Il, Thin code'is a modification of a widely used discrete-ordinates code that solves a wide variety of radiation transport problems. The program solves both fixed source and multiplication problems. i Slab, cylinder, and spherical geometries are allowed with various i boundary conditions. The fluence calculations incorporate, as an initial starting point, neutron fission distributions prepared i from core physics data as a distributed source. Anisotropic scattering is considered for all regions. The cross' sections are prepared with 1/E-flux-weighted, P matrices for anisotropic i scattering, but do not include resonance self-shielding factors. Fast neutron fluxes at locations other than the core midplane are calculated using a two-dimensional, discrete-ordinates code. The i two-dimensional coJe is.an extension of the one-dimensional code.. 1 '4.1.4.6 Thermal-Hvdraulic Calculations i J l j A description of the thermal-hydraulic models is given in Section 4 of Reference 4.1-1. i 4.

1.5 REFERENCES
i i 4.1-1 " General Electric Standard Application for Reactor l Fuel," including the " United States Supplement," [ s 4.1-17 Amendment 1 l l L
/ HCGS FSAR 8/83 s, -H \\ [ NEDE-24011-P-A and_NEDE-24011-P-A-US7 : rc Lancravcd revision); ,4( 4.1-2 L. Beitch, "Shell Structures Solved Numerical'ly by Using a Network of Partial Panels," AIAA Journal, Volume 5, No. 3, March 1967. i 4.1-3 E. L. Wilson, "A Digital Computer Program For the Finite Element Analysis of Solids With Non-Linear Material Properties," Aerojet General Technical Memo No. 23, Aerojet General, July 1965. 4 4.1-4 I. Farhoomand and E. L. Wilson, "Non-Linear Heat Transfer Analysis of Axisymmetric Solids," SESM Report SESM71-6, University of California at i Berkeley, Berkeley, California, 1971. 4.1-5 J. E. McConnelee, " Finite-Users Manual," General Electric TIS Report DF 695L206, March 1969. 4.1-6 R. W. Clough and C. P. Johnson, "A Finite Element k Approximation For the Analysis of Thin Shells," International Journal of Solid Structures, Vol. 4, 1 1968. i 4.1-7 "A Computer Program For the Structural Analysis of Arbitrary Three-Dimensional Thin Shells," Report No. GA-9952, Gulf General Atomic, 1969. 4.1-8 A. B. Burgess, " User Guide and Engineering Description of HEATER Computer Program," General l Electric, NEDE-20731-02 March 1974. .1' 4.1-9 .L. J. Young, "FAP-71 (Fati~gue Analysis Program) Computer Code," GE/NED Design Analysis Unit R. A. Report No. 49, January 1972. 4.1-10 L. A. Carmichael and G. J. Scatena, Stability and Dynamic Performance of the General Electric Boilina Water Reactor, NEDO-21506, January 1977. k 1 \\ \\ 4.1-18 Amendment 1 ? l
__ _... -. = - _ _ _ - _. _ _ i h l HCGS FSAR 10/84 4
i i
j 4.2 FUEL SYSTEM DESIGN !IlNSERT' l he fuel system design for the HCGS is identical to that which i t NRC reviewed and approved for GESSAR II (Reference 4.2-1 l Meth and criteria used to evaluate fuel system perform e are i also id
ical to those used for GESSAR II, except for ne t
i evaluation combined fuel-lift loadings from a sa shutdown j earthquake an loss-of-coolant accident. See S tion 3.9.1.4.10 and T e 3.9-See for the results o the fuel-lift i { evaluation. The re ts of the NRC review Section 4.2 of l GESSAR II documented i eferences 4.2-2 nd 4.2-3 are therefore j applicable to the HCGS. s 4.
2.1 REFERENCES
I I 4.2-1 General Electri tandard S ty Analysis Report, ~' Docket No. 5 47 ) 4.2-2 NUREG-9, " Safety Evaluation Repor elated to the L Fin Design Approval of the GESSAR II 6 Nuclear and Design", April, 1983 4.2-NUREG-0979 (Supplement No. 1 ), " Safety Evaluatio 1 Report Related to the Final Design Approval of the j GESSAR II BWR/6 Nuclear Island Design", July, 1983 l 3 1 ~ I t i l l i i i I f I 1 i 4.2-1 Amendment 8 l A n 1
,~ INSERT.F.OR PAGE 4.2-1 l The format of this. section corresponds to Standard Review Plan 4.2 in NUREG-0800. Most of the information is presented by reference to GESTAR II (Ref. 4.2-1). ) 4.2.1 DESIGN BASES Reference to design bases are given in_ Subsection A.4.2.1 of GESTAR II (Ref. 4.2-1). 4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS I Reference to the fuel system description and design drawings ] are given in Subsection A.4.2.2 of GESTAR II (Ref. 4.2-1). 4 4.2.2.1 Reactivity Control Assembly (Control Rods) The control rod description is given in Subsection 2.2.4 and is shown in Figures 2.6a, 2.6b, and 2.7 of NEDE 20944-P-1 ) (Ref. 4.2-2). 4.2.2.2 Reactivity Control Assembly Evaluation j The control rod evaluation is given in Subsection 2.3.3 of NEDE 20944-P-1 (Ref. 4.2-2). t j 4.2.3 DESIGN EVALUATION Compliance with the design bases is discussed in Subsection i A.4.2.3 of GESTAR II (Ref. 4.2-1), with the exception that Paragraphs 4.2.3.2.9 and 4.2.3.3.5 appear as below. 1 ] 4.2.3.2.9 Mechanical Fracturing Evaluation ) All mechanical breaking under normal operation and abnormal i ,perational transients is bounded by the analysis for LOCA i plus SSE given in Section 3.9.1.4.10. i l 4.2.3.3.5 Structural Deformation Evaluation } Results of the Hope Creek specific SSE plus LOCA analysis ) are documented in Section 3.9.1.4.10. 4 } 4.2.4 TESTING, INSPECTION AND SURVEILLANCE PLANS l Descriptions of fuel assembly testing, inspection, and surveillance are referenced in Subsection A.4.2.4 of GESTAR ] II (Ref. 4.2-1). 1 l
INSERT FOR PAGE 4.2.1 (Continued) 4.
2.5 REFERENCES
4.2-1 " General Electric Standard Application for Reactor Fuel", including the " United States Eupplement", NEDE-240ll-P-A-7, and NEDE-240ll-P-A'7-US. 4.2-2 "BWR/4 and BWR/5 Fuel Design", NEDE-20944-P-1 (Proprietary) and NEDO-20944-1, October 1976, and " Amendment 1", January 1977.
_ = _. -.. HCGS FSAR 1/84 I (_ i 4.3.2.4.2 Reactivity Variations Information on reactivity variations is referenced in Subsection A.4.3.2.4.2 of Reference 4.3-1. The combined effects of the individual constituents of reactivity are. accounted for in each { g,3 y Re g g i n T a b l et"-~'--".. b l 4.3.2.5 Control Rod Patterns and Reactivity Worths l Control rod patterns and reactivity worths are discussed in i Section 3.2.5 of NEDE-20944-P-1 (Reference.4.3-2). Typical control rod patterns and the associated power distributions are presented in Appendix A of Reference 4.3-2. These control rod 4 patterns are calculated with the BWR' Core Simulator. Qualification for this model is discussed and referenced in 4 Section 3.1 of Reference 4.3-1. i ____n.~__.4.4... asu ~ ,.a...s. Scrar rc ctivity is calculated 00 dcccribed in Section S.2 of 2 Reference '. 2 ' and ir disc" Seed in Sectier 3.2.5.2 Of Refererce 1 1 23 I i j 4.3.2.6-Criticality of Reactor During Refueling i 4.3.2.7 Stability f 4'.3.2.7.1 Xenon Transients I 4.3.2.7.2 Thermal Hydraulic Stability 4.3.2.8 Vessel Irradiations t l The neutron fluxes at the vessel have been calculated using the j one-dimensional, discrete-ordinates, transport code described in Section 4.1.4.5. The discrete-ordinates code is used in a distributed source mode with cylindrical geometry. The geometry l describes six regions from the center of the core to a point i beyond the vessel. The core region is modeled as a single, i j homogenized cylindrical region. The coolant water region between i 4.3-2 Amendment 4 4 1
HCGS FSAR 8/83 ( ( 4.
3.5 REFERENCES
l ~ '~ 4.3-1 " General _ Electric Standard Application for Reactor Fuel," including the " United States Supplement," _E NEDE-24011-P-Ahand NEDE-24011-P-A'US.Ll :::t1 l 22 revec revis Ort : ,4 / 4.3-2 "BWR/4 and BWR/5 Fuel Design," NEDE-20944-1 (Proprietary) and NEDO-20944-1, October 1976, and " Amendment 1," January 1977. (*.. _ _ i ('s. 4.3-4 Amendment 1
.=. l HCGS FSAR 1/84 i 4.4 THERMAL AND HYDRAULIC DESIGN Most o' the information in Section 4.4 is provided in the j licensing topical' report GESTAR II (Reference 4.4-1). The section numbers in Section 4.4 directly correspond to subsection numbers of Appendix A of GESTAR II. The differences are discussed below. 4.4.1 DESIGN BASES i The thermal and hydraulic design bases are referenced in Section A.4.4.1 of Reference 4.4-1. The design steady-state operating limit for the minimum _ critical power ratio (MCPR) and 1 the linear heat generation rate (LHGR) are given in Table 4.4-1. J 4.
4.2 DESCRIPTION
OF THERMAL AND HYDRAULIC DESIGN OF THE REACTOR CORE i A description of the thermal and hydraulic design of the reactor core is referenced in Section A 4.4.2 of Reference 4.4-1. Any additions or differences are given in the appropriate section below. j An evaluation of plant performance from a thermal and hydraulic standpoint is provided in Section 4.4.3. 4 ] 4.4.2.1 Summary Cc rer i een' '[SUmmbH H V A%;; p;rinef, of the thermal and hydraulic design parameters-l .this resctor with those of reactors of similar desied is given in Table 4.4-1. Np i 4.4.2 Linear Heat Generation Rate 4.4.2 Volo Fraction Distribution 4 The core average and maximum exit void fractions in the core at rated condition are given in Table 4.4-1. The axial distribution 4 of core void fractions for the average radial channel and the maximum radial channel (end-of-node value) for the core are given L 7 i i 4.4-1 Amendment 4 H.H. 2.1 CRmcAL Pawe &c
l l HCGS FSAR 1/84 in Table 4.4-3. The core average and maximum exit value are also provided. Similar distributions for steam quality are provided in Table 4.4-5. The core average axial power distribution used to produce these tables is given in Table 4.4-4. 4. 4. 2.g Core Coolant Flcw Distribution and Orificino Pattern 4.4.2 Core Pressure Drop and Hydraulic Loads 4.4.2 Correlation and Physical Data GE has obtained substantial amounts of physical data in support of the pressure drop and thermal-hydraulic loads. This information is given in Appendix B of Reference 4.4-1 where responses are provided to NRC questions on Section 4 of GESTAR II. 4.4.2 Thermal Effects of Operational Transients 4.4.2 Uncertainties in Estimates 4.4.2P)Flux Tilt Considerations The inherent design characteristics of the BWR are particularly well suited to handle perturbations due to flux tilt. The stabilizing nature of the moderator void coefficient effectively damps oscillations in the power distribution. In addition to this damping, the in-core instrumentation system and the associated on-line computer provide the operator with prompt and reliable power distribution information. Thus, the operator can readily use control rods or other means to limit effectively the undesirable effects of flux tilting. Because of these features and capabilities, it is not necessary to allocate a specific peaking factor margin to account for flux tilt. If, for some reason, the power distribution could not be maintained within normal limits using control rods, then the operating power limits would have to be reduced as prescribed in Chapter 16. 4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM f 4.4-2 Amendment 4
HCGS FSAR 1/84 4.4.7 SRP RULE REVIEW Acceptance criterion II.8 of SRP Section 4.4 specifies, in part, that the effects of crud should be accounted for in the thermal-hydraulic design, and also that process monitoring provisions be capable of detecting a three percent pressure drop in the reactor coolant flow. In general, the critical power ratio (CPR) is not affected as crud accumulates on fuel rods (References 4.4-2 and 4.4-3). l Therefore, no modifications to GEXL are made to account for crud deposition. For pressure drop considerations, the amount of crud assumed to be deposited on the fuel rods and fuel rod spacers is greater than is actually expected at any point in the fuel lifetime. This crud deposition is reflected in a decreased flow area, increased friction factors, and increased spacer loss coefficients, the effect of which is to increase the core pressure drop by approximately 1.7 psi, an amount which is large enough to be detected in monitoring of core pressure drop. It should be noted that assumptions made with respect to crud deposition in core thermal-hydraulic analyses are consistent with established water chemistry requirements. More detailed discussion of crud (service-induced variations) and its uncertainty is found in Section III of Reference 4.4-4. l 4.
4.8 REFERENCES
4.4-1 General Electric Standard Application for Reactor [T41 Fuel," including the " United States Supplement," i NEDE-240ll-P-A\\and_NEDE-24011-P-ALUS.5Ictcct)-3_ - pprcved revisionH 4.4-2 R.V. McBeth, R. Trenberth,-and R. W. Wood, "An Investigation Into the Effects of Crud Deposits on Surface Temperature, Dry-Out, and Pressure Drop, with Forced Convection Boiling of Water at 69 Bar in an Annular Test Section'," AEEW-R-705, 1971. 4.4-3 S.J. Green, B.W. LeTourneau, and A.C. Peterson, " Thermal and Hydraulic Effects of Crud Deposited on Electrically Heated Rod Bundles," WAPD-TM-918 September 1970. 4.4-4 General Electric, " General Electric Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application, "NEDO-10958A, January 1977. 4.4-9 Amendment 4
1 HCGS FSAR ~ TABLE 4.4-1' Page 1 of 2 THERMAL AND HYDRAULIC DESIGN CH'ARACTERISTICS OF THE REACTOR CORE HCGS WNP-2 NMP 2 General Goeratina Conditions (251-764)(251 704)(251 764) Reference design thermal output, MWt 3293 tita }393 Power level for engineered safety 3436 &+&& ibbEb features, MWt Stean flow rate, at 419.90F final 14.159 14.30'2'14.30'i' feedwater temperature (FFWT), millions Ib/h Core coolant flow rate, millions lb/h '100.0 Fears Fears Feedwater flow rate, millions lb/h 14.127 -14.26 14.27 System pressure, nominal in steam 1020 rett b&FE dome, psia System pressure, nominal core design, 1035 FG}5 +039 psia Coolant saturation temperature at core 549 5+9 &+9-design pressure, CF Average power density, kW/ liter 48.7 49.15 4^.15 Maximum linear heat generation rate, 13.4 +4r4 &lr4 kW/ft Average linear heat generation rate, 5.34 Er+G Eva9-kW/ft Core total heat transfer area, ft2 74,841 't,7El I4 784 7 Maximum heat flux, Btu /h-ft: 361,600 96+7500 301,000 Average heat flux, Btu /h-ft2 144,100 145,100 45,050 Design operating minimum critical power (see Chapter 15) ratio (MCPR) b Core inlet enthalpy, at 526,1 !5 2 7. 0 ' ' ? 5 2 7, C ' '4 419.90F FFWT, Btu /lb
i j HCGS FSAR I, i l TABLE 4.4-1 (cont) page 2 of 2 l 1 i i Core inlet temperature, at 531.6 533.0': 2532.C
19.90F FFWT, OF I
l l Core maximum exit voids within 77.1 { i assemblies, % r Core average void fraction, 0.419 0.41S 0.?08 active coolant Maximum fuel temperature, OF 3435 3435 3435 i Active coolant flow area per 15.824 15.82d 15.924 j assembly, in.2 Core average inlet velocity, ft/s 6.41 5.SS 5.9' ] Maximum inlet velocity, ft/s 6.803 '.23 S.00-1 Total core pressure drop, psi 21.25 24.74 23.7 l Core support plate pressure drop, psi 16.82 20.22 10.28-Average orifice pressure drop 3 1 Central region, psi 7.16 5.03 5.05-Peripheral region, psi 14.53 10.54 15.0-Maximum channel pressure loading, psi 10.88 12.20 12.10 i Average-power assembly channel pressure loading (bottom), psi 9.61 10.02-j Shroud support ring and lower shroud pressure loading, psi 22.87 27.52 Upper shroud pressure loading, psi 6.05 7.2 \\ i St f5 at
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) 1, ? l i l I l i i 4 i i I
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lg' HCGS FSAR rod indicates indirectly that the rod and. drive are coupled. The over-travel position feature provides a positive check on the coupling integrity, for only an uncoupled drive can reach the over-travel position. d. During operation, accumulator pressure and level at the~ normal operating value is verified. Experience with CRD systems of the same type indicates that weekly verification of accumulator pressure and level is sufficient to ensure operability of the accumulator portion of the CRD system. e. At each refueling outage, each operable control rod is subjected to scram time tests from the fully withdrawn-position. Experience indicates that the scram times of the control rods do not significantly change over the time interval between refueling outages. A test of the scram times at each refueling outage is sufficient to . identify any significant lengthening of the scram times. IfEBEb i 4.6.3.1.6 Functional Tests The f unctional testing program of the CRDs consists of the 5-year maintenance life and the 1.5X design life test programs, as described in Section 3.9.4. 4 'There are a number of failures that can be postulated on the CRD, { but it would be very difficult to test all possible failures. A partial test program with postulated accident conditions and imposed single failures is available. The fo'llowing tests with imposed single f ailures have been performed to evaluate the performance of the CRDs under these conditions: 1 i a. Simulated ruptured scram line test 4.6-43
i INSERT FOR PAGE 4.6-43 i f. Prior to startup after the first refueling outage, PSE&G shall: 1. Confirm that the leak rates, loading conditions and material prcperties for the scram discharge volume piping system are bounded by the limiting values for those parameters identified in the May 10, 1984 BWR Owners Group letter, i 2. Comply with BWR Owners Group recommendations for 1 leak detection capability, j 1 3. Comply with the applicable generic secondary containment-Emergency Procedure Guidelines, and 4. Provide assurance that the expected radiation fields and contact exposure levels at the scram discharge { volume piping systems in the facility will not impair the performance of routine tests, inspections, and post-scram reset walkdowns. l i v i f
HCGS FSAR 5/85 automatically realign from system flow test modes to the emergency core cooling mode of operation following receipt of an automatic initiation signal. The core spray and LPCI systems begin injection into the reactor pressure vessel (RPV) when reactor vessel pressure decreases to system discharge shutoff pressure. HPCI injection begins as soon as the HPCI turbine-pump is up to speed. The injection valve is open, since the HPCI system is capable of injecting water at full flow into the RPV over a pressure range from 200 psig to reactor pressure specified in mode A of Figure 6.3-3. 6.
3.6 REFERENCES
6.3-1 General Electric, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Accendix K, NEDE-20566-P, November 1975. 6.3-2 H. M. Hirsch, Methods for Calculatina Safe Test Intervals and Allowable Recair Times for Encineered Safecuard Systems, NEDO-10739, General Electric, January 1973. 6.3-3 General Electric, " General Electric Standard Application for Reactor Fuel," including the l-7\\ l " United States Supplement," NEDE-24011-P-Ahand NEDE-24011-P-ALUS. Octest accroved rcvision11 1 6.3-43 Amendment 10
i I HCGS FSAR 5/85 / RHR pump A or C or core spray pump A, and RHR pump l A or C or core spray pump C. After receipt of the initiation signals and after a delay provided by time delay relays, each of the two solenoid pilot air valves for all ADS valves are energized. This allows pneumatic pressure from each ADS valve. accumulator to act on the air cylinder operator of its respective ADS valve. Each ADS trip system timer can be reset manually to delay system initiation. If reactor vessel water level is restored by the HPCI system prior to the end of the time delay, ADS initiation will be prevented. A manual inhibit switch is provided in each division of the ADS initiation logic. By placing this switch in the inhibit position, the operator will inhibit automatic depressurization. This will be indicated by a white status light and an annunciator window in the main control room. If the ADS has already begun and the initiation signal is sealed in, the inhibit switch will not break the seal-in, and the operation of the ( ADS will not be terminated. The ADS trip system B actudtes the A solenoid pilot valve on each ADS valve. Similarly, the ADS trip system D actuates the B solenoid pilot valve on each ADS valve. Actuation of either solenoid pilot valve causes the ADS valve to open to provide depressurization. Manual initiation of the ADS trip systems or individual ADS valves is possible from the main control room. To manually initiate an ADS trip system, the control room operator must actuate two armed pushbutton switches, one for each of the logic channels associated with that trip system. Manual initiation bypasses the ADS trip system time delay and all the trip logic. The control room operator can manually open an individual ADS valve l by depressing one of the two pushbutton switches (one for each pilot solenoid) that will bypass the trip logic and energize the associated pilot solenoid allowing air to open the valve.f [W2%Tl c. ADS testability - The ADS has two complete trip systems, one in trip system B and one in trip system D. 7.3-9 Amendment 10
INSERT FOR PAGE 7.3-9 In addition, controlled access (key-locked) hand switches provide local manual control for certain ADS valves. l l l 4
i HCGS FSAR 01/86 detect low water level in the CST. Either switch can automatically cause suction transfer. To prevent losing suction to the pump, the two suction valves are interlocked so that one suction path must be open before the other closes. See Figure 7.4-2, Sheets 4, 5, and 6 for valve operation logic. One of the RCIC pump suction automatic switchover level switches is also used to provide CST low-low level indication at the remote shutdown panel (See Section 7.4.1.4.5.2). The RCIC turbine is functionally-controlled as shown on the RCIC functional control diagram (FCD), Figure 7.4-1. The turbine governor control system limits the turbine speed and adjusts the turbine steam control valve so that design pump discharge flow rate is.obtained. The flow signal used for automatic control of the turbine is derived from a differential pressure measurement across a flow element in the RCIC system pump discharge line. The turbine is automatically tripped and.the throttle valve ~ closed if any of the following conditions are detected: ( a. Turbine overspeed b. High turbine exhaust pressure c. Low pump suction pressure d. Auto-isolation signal 1. High area temperature 2. Steam line high differential pressure or instrument line break 3. Steam supply pressure low s4 4. Exhaust diaphragm high pressure. Pe. Emectum... ; high -eimm le.el ' Instrument ranges for the RCIC system controls and ( instrumentation are listed in Table 7.4-1. 7.4-4 Amendment 14 q-
HCGS FSAR sources to the HCGS switchyard, all of which are physically [ independent sources of offsite power to the HCGS unit. They are: \\ a. A tie of approximately 2112 feet between the HCGS and Salem switchyards b. A tie of approximately 42.9 miles to the New Freedor switching station c. A tie of approximately 30.1 miles to the Keeney switching station. The maximum winter capacity of each of these lines is 3500 MVA. The maximum summer capacity of each of these lines is 3220 MVA. The 500-kV switchyard provides preferred power through its interconnections with two sets of two station power transformers, to the 13.8-kV ring bus as shown on Figure 8.3-1. Station power transformers T1 and T4 each supply two 13.8/4.16-kV and one 13.8/7.2-kV station service transformer. Station power transformers T2 and T3 each supply one 13.8/4.16-kV station /J.8-4V/ service transformer Egg one. n o =/ = =6tation lighting and 208-/Jo V 7 AWb OME 1 power transformer,/ /In the event a station power transformer is 13.9/13.S kV unavailable, alternate feed is made available to the.affected gggp buses. MastpTla! TRNISR9G. The offsite power systems and their interconnections are described in Section 8.2. 8.1.-2 ONSITE POWER SYSTEMS The onsite power system for the unit consists of two major categories: a. Class 1E power system - The Class 1E power system supplies Class 1E loads that are necessary for safe and orderly shutdown, for maintaining the plant in a safe shutdown conditien, and for mitigating the consequences of an accident. A limited number of non-Class 1E loads important to the power generating equipment' integrity are also supplied from the Class 1E power system. These non-Class 1E loads are listed in Table 8.3-1. 8.1-2
HCGS FSAR 11/85 ( of the 500-kV bus. Direct stroke lightning protection is provided by overhead 19/89 Alumoweld shield wires. The control and status indication for the 500-kV main step-up transformer disconnect switch is installed on a breaker control relay rack in the switchyard control house. The switchyard control house will provide an auxiliary switch contact for input to generating station computer systems via a data input / output (I/0) cabinet for status indication. For safety reasons, the control switch for the transformer disconnect switch is provided with a lock-in handle. The generating station control room operator must release the key in his possession to permit operation of this control switch. l 8.2.1.4 Switchyard The 500-kV switchyard, located to the east of the Hope Creek plant, is designed with tapered tubular steel structures and rigid aluminum bus work. This yard consists of two breaker-and-a-half bays containing five SF-6 circuit breakers connected to two 500-kV main buses, 10X and 20X, as shown on Figure 8.2-2. Bus 10X is protected by primary and backup differential relays. Breaker failure relaying detects a failure-to-trip or failure-to-interrupt condition at the line terminal and trips associated breakers necessary to isolate the line. The 500 kV and 13.8.kV' -( circuit breakers are pneumatically operated and each breaker has sufficient stored air for a minimum of three operations without compressor actuation. Compressor motors are supplied by dual ac feeds from separate panels in the switchyard. The control room and the switchyard control house have independent and simultaneous control of the 500 kV circuit breakers. The electric system operation center, located'in
Newark, N.J.,
has limited control of the line breakers 51X, 60X, and 61X and the tie breaker 50X, and no control of the generator breaker 52X. Restoration of the 500 kV lines after a LOP would-generally consist of the following procedural steps: The system load dispatcher would be contacted to verify availability of 500 kV circuits. Verify 4 kV and 7.2 kV non-1E bus infeed breakers are opened. 500fl%W Y Verification of'""""' "" transformer and 13 kV ring bus breaker positions aligned to restore offsite power. I The load dispatcher is contacted for final clearance to reclose l 500 kV breakers. l i 1 \\._ q 8.2-3 Amendment 13 _ _., _ _ _ _. _.. w _.._ _ _._,
k HCGS FSAR 01/86 Once 500 kV power.is reestablished, 4 kV and 7.2 kV power is '~ provided to the respective non-1E buses, loading of these non-1E buses can then commence. Final transfer of Class 1E loads from-the standby to the preferred power source can be made when plant conditions are stable. Generating station auxiliary services are supplied via two i 13.8-breaker bays by four 500/14.4 kV, 42/56/70-MVA, oil-immersed, self-cooled / forced-air-forced-oil-cooled-(OA/FOA/FOA) $fgjf#g three-phase transformers connected to the 500-kV busses 10X and I 20X, asLshown on Figure 8.2-2. Station power transformers T1 and bowb l T4 each supply two 13.8/4.16-kV and one 13.8/7.2-kV station gamuney l service transformers. The remaining two transformers, T2 an T3, nucRe*2 \\(each supply one 13.8/4.16-kV station service transformer one / g,g gg
32.5 tv/zer'" station light and power transformer / Each 13.8-kV i 2$-poV breaker bay consists of three breakers in series,.
To prevent paralleling of the transformers, one of the breakers is normally I open. This breaker is closed in case one of the transformers is-out of service. As shown on Figure 8.2-2, there are six 13.8-kV, 1500-MVA oil circuit breakers. Breaker failure' protection detects the failure to trip or failure to interrupt conditions at the line terminals g and electrically isolates faulty equipment. Primary and backup, relay protection on the 500/14.4-kV station-power transformers is i provided by the use of harmonic restraint differential relays. T The 13.8-kV system is ungrounded and connected to the delta side ~ of all station power and station service transformers. rTo detect a phase-to-ground fault in the system, a 22.5 tv/::c V:#arounded-wye grounding transformer is installed on the secondary side of i3.z kV[ each station transformer. The neutrals of the grounding 12o V transformers are connected to neutral resistors and relays for phase-to-ground fault detection and annunciation. To locate the fault, a current transformer is installed on each' of the 13.8-kV cables. The current transformers are connected.to j-a milliammeter via selector switch to measure the residual cable current and locate the faulty cable. This configuration of the offsite power system, with provisions for periodic testing, is in full conformance with NRC GDC 17 and 18 of Appendix A to 10 CFR Part 50, which is further discussed in 3 Section 8.3.1. 2.1. I HCGS complies with Regulatory Guide 1.32. Clarifications and- -l 1 ] exception are noted in Section 1.8. { ~ m l 8.2-4 Amendment 14 ---,--.#1-* rv*- --*-r-- ^7-- '~
HCGS FSAR 11/85 Station light and power transformers SLP 1 and SLP 2 are tapped from 13.8-kV bus sections S3 and S8, respectively. These transformers provide service power to the switchyard via distribution panels located in the switchyard and in the control house.4 Control power for the protection of switchyard equipment is supplied by two 125-V de switchyard service batteries (regular and backup), equipped with two full-capacity chargers each. In the event of a relay operation, the relays can be reset and the equipment returned to service within one hour. This scheme ensures.that primary and backup relay protection of switchyard electrical equipment will not be lost. The two batteries, de distribution cabinets and battery chargers are electrically independent andflocated more than 8 feet apart in the switchyard pfgi' control house. Control cables from the de distribution cabinets to regular and backup protection racks run in separate cable trays. 8.2.1.4.1 13.8-kV Supply Station power is supplied from the 13.8-kV switchyard via multiple runs of 15-kV, 2000-kcmil power cable in polyvinyl chloride (PVC) conduit. The PVC conduit runs are encased in concrete and run underground from the 13.8-kV feeder positions to in-plant station service transformers. These duct banks are routed to minimize the possibilities of simultaneous failure under operating, postulated accident, and environmental conditions. 8.2.1.5 System Monitoring PSE&G transmission lines and rights of way are patrolled at least five times each year to ensure that the physical and electrical integrity of transmission line supports, hardware, insulators, and conductors are acceptable for safe and reliable-service. This periodic transmission line patrol is conducted by 15(AJD SUBSTATic4 TRAMSFO2Hees SLP3 ano SLP4 APE TAF$Ub Fhl 13.9-kV bus FECTicNS S4 AND S9 RGSR5CDU6LV, NJD P20VICE SGResCE VCR (3 9-kV lSLAMD TarfR DGTRtbLcroon SYSTEM VIA \\3.9-kV SWrTCHEEAR LC GTEb I4 THe MCG3 StocTCHVARD. 8.2-4a Amendment 13
1 HCGS FSAR 11/85 k helicopter and ground patrols. Climbing inspections of structures are performed at least every 3 years depending on the age of the line. Monitoring of the offsite power sources in the plant control room is provided for by a hard-wired, console-mounted, mimic bus arrangement that shows the status of/ station power,# station servicet and ctaticr I c.h t enc poec: transformers. Potential lAUDI indication of the 500/13.8 kV systems and status indication of the transformer secondary and bus tie disconnect switches are provided by the plant computer systems. Control and status indications of all 500-kV and 13.8-kV breakers are also shown. Annunciation accompanies status changes of circuit breakers, loss of potential, transformer trouble, fire protection system actuation, carrier equipment failure, and fault recorder failure. The switchyard fault recorder inputs include phase currents, l voltages, and carrier information for all three Hope Creek switchyard offsite power sources. Inputs to the plant fault recorder include the following: a. Voltage and current information on the generator and main transformer b. Voltage information on the 13.8-kV bus sections 1, 5, 6, and 10 c. Voltage information on the station service transformers d. Voltage.information on all 7.2-kV and 4.16-kV buses. The plant computer system displays additional offsite power system information for the operator on CRTs. Each display is a mimic bus arrangement similar to the hard-wired mimic bus and includes the status of switchyard power circuit breakers (PCBs). The main generator output leads to the 500-kV switchyard are monitored in the control room. A mimic bus arrangement provides control and status indication of the synchronizing PCB. Potential indication and monitoring of current, watts, volt-amperes reactive (VARs), watthours, and voltage are provided. (, Annunciation accompanies an abnormal change in the status of the synchronizing PCB and failure of the supervisory system. 8.2-5 Amendment 13
l 500 KV TRANSMISSION LINES ^ s h h " Abb!TIOJ " KEENEY EEDOM i m SECT 10X h 500-13 KV TRANSFORMER y (4) uM T-3 T.I N. N.O. .C. 61x 51x 0 0 / 13.8 KV \\ l SWITCHYARD {} N.O. N.C. SLP2 M SLP1 XFMR ( 3 XFMR ~~ ~ {}N.O. N.C. \\ 500 KV / SWITCHYARD O O 52x
/
N.C. N.O. N.C. 'O T-4 T-2 3>< evw evn '5 (b 5t: I JJJ SECT 20x E3 ww og sa mm U Y l'l' 24 500 Y Y Y Y MAIN TRANSF. \\ / V 13.8 KV FEEDS TO STATION SERVICE TRANSFORMERS HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSIS REPORT r' i ONE LINE DIAGRAM FIGURE 8.2 2 AME'NDMENT 8,10/84 . _ _ =. _...
l HCGS FSAR 11/85 ( 8.3 ONSITE POWER SYSTEMS The onsite power-systems consist of ac and de power systems. E.3.1 AC POWER SYSTEMS 8.3.1.1 Description The onsite ac power systems include a Class 1E system and a non-Class IE system. Figure 8.3-1 is the single line drawing of both the systems. The onsite ac power is defined in Section 8.1.2. l 8.3.1.1.1 Non-Class 1E AC Power System The non-Class 1E portion of the onsite power system supplies ac power to non-Class 1E loads. A limited number of-non-Class 1E \\ loads, important to the power generating equipment integrity, are supplied from the Class IE distribution system through isolation devices. These non-Class 1E loads are listed in Table 8.3-1. The offsite power for the plant is fed through the 500-kV system via the 13.8-kV yard ring bus. Two separate buses, 10X and 20X, of the 500-kV switchyard, feed the 13.8-kV ring bus via 500 GND. Y/14.4 kV station power transformers, T1, T3 and T2, T4, respectively, as shown on Figure 8.3-1. Physically independent routing of buc 10X and 20X feeders from sections 1 and 2 of the 500-kV switchyard to the station power transformers T1, T3 and T2, T4, minimizes the-likelihood of simultaneous failure of the two 500-kV buses. The 13.8-kV ring bus provides auxiliary power during startup, normal operation, shutdown, and post-shutdown operation of the unit. Station power transformers T1 and T4 each e,mo-feed two 13.8-4.16-kV and one 13.8-7.2-kV station service 2s/2>-y wD our \\ transformer. Station power transformers T2 and T3 each feed one i 13.pl3.8ld/ \\ 13.8-4.16-kV station service transformer oned:: cc zur = _.. - ~ bagD station lighting power transformer,4 Two 7.2-kV and six 4.16-kV S498mn non-Class 1E buses are supplied from the above eight station WJMA2HR service transformers. The 7.2-kV buses are 10A110 and 10A120, and the 4.16-kV buses are 10A101, 10A102, 10A103, 10A104, 10A501, and 10A502. The configuration of the non-Class IE power system is described below for the normal operation of the 13.8-kV ring ( bus and all the 4.16-kV and 7.2-kV non-Class 1E buses. 8.3-1 Amendment 13 i l I
HCGS FSAR 11/85 be closed only if one of the infeed breakers of the double-ended unit substation-is open. The 480-V unit substations feed 480-V motor control centers l (MCC), motors of 100- to 250-horsepower rating, and 480-V power l panels. MCCs supply power to motors'of up to 75 horsepower rating, battery chargers, 480/277-V power distribution panels, and 480 and 208/120-V power distribution panels. Uninterruptible power supply (UPS) panels of 120 V ac supply the security system, public address system, NSSS computer, BOP computer, etc. The distribution panels feed miscellaneous loads such as lighting, space heaters, and unit heaters. The non-Class 1E equipment ratings are listed below: a. Transformers 1. Main step-up transformer: 3-1 e, 362.5/406 MVA each, FOA 550C/650C, 24-500 kV; impedance 16% on 362.5 MVA base, NLTC +5% in 2-1/2% step 2. Station power transformer: 4-3 4, 42/56/70 MVA each, OA/FOA/FOA, 650C, 500-14.4 kV, impedance 5.1% on 42 MVA base NLTC +5% in 2-1/2% step l 3. Station service transformers: 2-3 e, 15/20/25 MVA, OA/FOA/FOA 550C, and 16.8/22.4/28.0 MVA, OA/FOA/FOA 650C, 13.8-7.2 kV GNDY, impedance 5.5% on 15 MVA base. HV-LTC= -15% to +5% 4-3 e, 17.41/23.21/29 MVA, OA/FOA/FOA 550C and-19.5/26/32.5 MVA, OA/FOA/FOA 650C, 13.8-4.16 kV, GNDY, impedance 7.7% on 17.41 MVA base. HV-LTC= -15% to +5% 2-3 e, 14.7/19.6 MVA OA/FA 550C, 16.5/21.95 MVA l OA/FA 650C, 13.8-4.16 kV, GNDY, impedance 5.48% on l 14.7 MVA base, HV-LTC= -15% to +5% 4. Station lighting and power transformer: 2-3 e, 500 kVA,pu::ca 208 V GNDY/120 V D39 Col J. lsu2mTj z 8.3-3 Amendment 13 J o
q HCGS FSAR 11/85 b. Switchgear ~ 1. 7.2-kV switchgear: 1200/2000 A continuous rating, 500 MVA 3 e class, 35,000 A interrupting rating at 8250 V (maximum rated voltage) 2. 4.16-kV switchgear: 1200/2000 A continuous rating 350 MVA 3 e class, 42,400 A interrupting rating at 4760 V (maximum rated voltage) I NSEEr ', e c. 480-V unit substations 1. Transformers: (a). 1000 kVA, 3 e, 4160-480 V (b) 1500 kVA, 3 e, 4160-480 V i 2. Bus: 2000 A continuous rating for 1000 kVA unit substations, 3200 A continous rating for 1500 kVA unit substations 3. Breakers (metal clad): 30,000 A d. 480-V MCCs control centers 1. Horizontal bus: 800 A continuous rating, 42,000 A ~ bracing 2. Vertical bus: 300 A continuous rating, 42,000 A bracing 3. Breakers (molded case); 150 A and 250 A frame sizes, 25,000 A, symmetrical rms interrupting rating \\ %s 1 8.3-4 Amendment 13 4
f 4 i ]. INSERT FOR PAGE 8.3-3 i a i i-5. Island Substation Transformer: 2-35, 12/16 MVA,.OA/FA 65*C, 13.8-13.8-kV Y/7970 V, impedance 5.89% (SLP3), 5.94% (SLP4) on 12 MVA base at 13.8 kV. .i i INSERT.FOR PAGE 8.3-4 i i i 3. 13.8 kV switchgear: 1200'A continuous rating, 19,300 A short circuit-(symmetrical) rating. i 4 4 i s A I t i 4 .t i 6 1 i I j
HCGS FSAR Each MCC cubicle derives its 120-V ac control power from a control power transformer located within the cubicle. 8.3.1 1.2.10 Electric Circuit Protection Systems Protective relay schemes and trip devices on the primary and backup circuit breakers are provided throughout the power system in order to: a. Isolate faulted equipment and/or circuits from unfaulted equipment and/or circuits b. Prevent damage to equipment c. Protect personnel d. Minimize system disturbances 'f \\_ e. Maintain power continuity of power supply in the unaffected part of the system. The short-circuit protective system is analyzed to ensure that the various adjustable devices are applied within their ratings and set to be coordinated with.each other to attain selectivity necessary to isolate a faulted area quickly with a minimum of disturbance to the rest of the system. Major types of protection measures employed include the following: a. Differential relaying - Differential reldying schemes are provided for the main generator, main generator-ISLAMD N m in transformer. station power transformers, station V gg service transformers,hfi3.S r[ c2M ec fro-a cccm ectT
~' #'"
" #'C ""t
3 0"3
'O CTO '/ 4.16-kV TF4]SFeRMc45' buses, SDGs, motors above 3000 horsepower rating, and 4.16-and 7.2-kV buses. These schemes provide high'- speed disconnection by opening appropriate breakers to prevent severe damage in case of faults occuring within the bounds of the ar'eas served by these relays. b. Overcurrent relaying - Each Class 1E 4.16-kV bus (, incoming feeder circuit breaker is equipped with three 8.3-11 4
m .m ______m m.. 1 e 'h HCGS FSAR 1/86 ~ l TABLE 8.3-1 (cont) Page 2 or 10 Number Conr.ected To Class IE Loading Fcquencetra Operatirq Distribution System Time Time
Rating, kW Diesel Buses Min From Min From Item Description Equipment No. each, hp eachts8 A
C H D N o. DBAtt88 No. LO P( 8 8 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 st r 8 17. Control room supply f ans 14-VH 4 0 3 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) 10X411,412,413 . 414 10X421,422,423, 424 10X501,502,503, 504 19. Deletel 20. Intake structure exhaust fans 1 A, B, C, D-V50 4 40 32 1 1 1 1 2 13 st858 2 13 st**8 l 21. Control room chilled water 1A-P400 60 48 1 1 1 65 s 1 ed s circulating pumps 1B-P400 22. Control room supply unit 1A-VH403 90 1 1 1 60 s 1 60 s heating coils 1B-VH403 60 23. Control room water chillers .1A-K400 680 506 1 1 1 1 60 s I B-K 4 00 24 Diesel generator room recirc 1A-V412 thru 125 100 2 2 2 2 3 30 s8 3 30 s syster fans 1H-V412 1 25. Primary containment instrument 1A-K202 15 12 I' 1 1 30 man 1 30 min gas compressor 1B-K202 ij' 26. Battery chargers, 250-V de 10D423 1) 1 1 2 13 s 2 1J s 10D433 a i4-27. Control area battery room 1A-V410 5 4 1 1 1 60 n 1 60 s } exhaust fans 10-V410 28. RB FRVS recirculation unit 1A-VH213 thru' 100 2 1 2 1 4 19 s } unit heating coils 1 F-VH 213 30 nt*8. t, Amendment.14 I i 1
8 HCGS FSAR 5/85 l TABLE 8.3-1 (cont) Page 4 of 10 Number Connected To.C1aas 1E Loadin M ejuence(28 Operatir.g Distribution System Tlw Time
Rating, kW Diesel Buses Min From Min Frow lirm Dencription Equip _ne nt No. each, hp each(88 A
C H D No. DBM 8 8 3 tio. IDP( 8 8 8 r Class 1E loads 45. Deleted. l 46 480-V power supply to class 1E 1AC488, 1BC488 4 1 1 1 1 2 13 s 1 13 s l chiller panels IAC491, IDC491 l 47. Traveling screens 1A-S501 5 4 1 1 1 1 3 55 a 3 55 IB-S501 1C-S501 1D-S501 48. ECCS jockey pump 1 A-P228 10 8 1 1 1 1 3 13 e. 3 13 s 18-P228 1C-P228 1 D-P228 49. Motor-driven diesel generator 1 A-P40 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 1A-VE261 45 1 1 1 15 min 1 15 min room duct heaters 1B-VE261 51. 400-V power supply to hydrogen 1A-C200 1 1 1 1 13 e 1 13 s and oxycen analyzer panels 1B-C200 52, 250-v de battery room duct 10-VE418 10 1 1 13 e 1 13 s l' heaters ~ 53. 125-V de diesel area battery 1A-VE420 21 1 1 1 1 3 13 a 1 13 s room duct heaters 1B-VE420 1C-VE420 1 D-VE420 54 HPCI pump room duct heater 10-VE260 11 1 1 13 a 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 268 198 1 1 1 1 75 s chillers 1B-K403 fgg(jg Amendment 10 l
,/- t HCGS FSAR 1/86 l FABLE 8.3-1 (cont). Page 10 or 10 (*) Loading sequence is based on availability of three standby diesel gererators and their associated electric power distribution systems. (3) DBA = Design basis accident LOCA = toss-of-coolant accident l LOP = Loss of of f site power DBA = LOCA + LOP i I (*3 MOV's maximum stroking time will vary f rom 20 to 70 seconds except. tor the main stesm stop l valves, with a stroking time of 120 seconds. l MOV loads are not included in this table and the diesel generator loading tables that follow because of their small magnitude and short period of operation. (s) " Operating Rw" is taken as 0.8 motor hp rating f or motors 250 hp and smaller (*3 During a DBA, any two core spray pumps and three RHR pumps can be manually tripped af ter 10 minutes from the occurrence of LOCA, depending upon the load or each standby diesel generator. Either the A or the B RHR pump must be retained in service af ter 10 minutes trom the occurrence of LOCA. (73 Buses A and B each have twa FRVS recirculating f ans and two unit heating coils connected to them. Buses C and D each have oie FRVS recirculating f an and one unit neating coil connected to them. In the case of a DBA, one fan and one unit heating coil will start on each or buses A, B, C, and D and the remaining f ans and heating coils will start in diesel buses A and B at the times shown in the loading chart. (e) Deleted. (') In case the lead f an f ails to start, the lag f an will start automatically at 95 seconds. (son Deleted. ( a s s Upon the occurrence of a LOCA, non-Class 1E loads are tripped by LOCA signals in 3 seconds by tripping the unit substation circuit breakers feeding ti,e non-Class 1E MCes and motors. These loads can te reenergized manually at 10. minutes after the occurrence of LOCA. (t** Two redundant 25-hp pumps are provided. (13 8 Times shown are from the occurrence of LOCA or loss ot of f site power. (t*3 Loads are not sequenced but power is required to be available to the loads in the event of LOP or be available'within 25 seconds af ter an ATRS initiation signal. The time shown (13 sec) is ' when power is available af ter LOP. (858 Loads are not sequenced but are controlled by process signals. For SDG loading purposes, these l l loads are assumed to start and run af ter 13 seconds from the DBA/LOCA event. l & ~IIHG MbtCATED IS SECLLang: T1H6 f5G2LAJRGb $ MMU W W IM N h W IF CutLLEP.s ARG M Md To W & W, M M MW% 6 O SW& Amen dmen t is l [ t
( e N <0 o r! y $,I e. E -_8 m __8 a. 8 _g e e W m _-o us ---8 a b E W 2 E _g w N R --_g p2 t. m N I n _g .4* U --m s 8 W U g 8 o LOAD IN AMPERES HOPE CREEK BATTERY 10D 421 GENERATING STATION FINAL SAFETY. ANALYSIS REPORT ('- CLASS lE 250 VDC ( BATTERY LOAD PROFILE FIGURE 8.3-16 SHEET 7 OF 8 AMENDMENT 8,10/84
P6U2E Q.l-IO (Cam %m 4 OM' Ites Equipment Tag rioor Elev loc rig Col humoer Crane or Botat Nurber Building (ft1 haber At 1 Ama tor building polar crane 10a200 Reactor 231 1.2 32 u-V 2 Personnel air lock hoist 105217 Reactor 102 1.2-24 P-R 3 ancirculation pump motor 1AR201 anactor 102 1.2-28 sa-boist 15E201 (Drywell) Sa-4 Reactor water clean-up filter / 1AR220 seactor 174-4 1.2-31 R-Q desineraliser boist 185220 5 EPCI pump and turbine hoist 1AE211 anactor 54 1,.2-25 w-v 138211 5 3CIC pump and turbine hoist 10H212 Beactor 54 1.2-25 w-v; 7 Main steam tunnel underhung crane 108214 Deactor 102 1.2-24 P-Qs 108223 4 Inboard MSIV hoist 105203 menctor 102 1.2-25 Q-a s (Drywell) 9 vacuum breaker valve removal 105207 seactor 54 1.2-27 u-Vs hoist (Torus) to Main steam line relief valve 10n202 menctor 135-4 1.2-29 removal hoist (Drywell) 11 Turbine building. bridge crane 108102 Turbine 137 1.2-15 E-rs 12 reedwater heater removal hoist 1AR103 Turbine 102 1.2=14 3-3g 18:1103 13 May equisment removal hoist 108104 Turbine 171 1.2-17 a.ra 14 Petor-generator set hoist 0A5105 Turbine 137 1.2.-15 su-a Oss105 i T10021717
4 (. ECGS FSAR YABI2 9.1-1C VY ICAD BACLING SY3YYF.S CAYA St:MMAJtY 01/86 Page 1 of 4 Is Ioad Over haz Vert Is Icad Over Safety-RelatedI33 Capacity Lift Seitaic Cesagn Sa f e ty-Rela ted (5) Equisme nt.on Exclusion ftonal (ft in) Cat I Standardf21 Equipment ? Next f.ower Elev Criterton(Il 23R 150 main 129-0 Yes a, b Yes Yee kne 10 aux 233 30 16-3 No(3) e, 4 leo Yes hne '4-20L 24 12-0 2133 c, d Yes Yes leone .9-171 17R 10 26-0 No o No No 3 21R 4 9-10 No I3) c, d Yes NA None 187 3 9-0 NoI3I c, d Yes NA C 2C ' 2-1/2 (1 4) NoI3I a, dIII Yes Yes kne 203 2 16-0 NOI3) 4 Yes Yes None
-22A 2
7-0 d Yea No C 1 32-2 stoI33 c, d No ye c .p 220 72-3 No e, b No No 3 main meta 45 122-0 aux aux 22 24 12-6 MO d W No 3 i3 15 37-0 No e, d No No 3 29 15 16-5 No e, 4 No No 3 Amendment 14 Q-T; 53R -2cn 36-5 l l l l
m-_ .,_m g.. HCGS FSAD 11/95 l TWLE 9. 3-5 Paqo 1 of 2 STANDBY LIQUID CONTROL SYSTEM OPEP ATING PP FSSifR F/TEMPEP ATilPE CO"DITIONS Test MMantia Standby Modeta) circulation Teet In ject ion Test t
8 Ope ra t ir.q Modet t 8 -
Pressure, Temperature, Pressure, Temperature, Pressure, Te mpe ra tu re, Pressure, Terperature, Piping psigt3D OF psigt88 0F
_psigt88
F
_psiqt88 UP 4 Pump suction Storage tank 70/110t
8 rest tank 70/110t**
Test tar k 70/110t*8 Storage tant 70/1105** static head static static static head headtS) headtS8 Purg discharge to 0 70/110 0/1190 70/100 40 (plus 70/100 40 (plus 70/110 explcsive valve reactor reactor inlet stat ic head) static head) to 1253 l Explosive valve Reactor static 70/110 Reactor 70/100 <40 (plus 70/100 <40 (plus 70/110 outlet to, but head to 1140t*3 static head reactor reactor not ir clud ing,. to 1140t** static head) etatic head) Potor-operated to <1255 l stop check globe i valves Motor-operated head to 1140 t
3 static head static static head Reactor static 70/ 560t

8 Reactor 70/560(F)
Feictor 5 12st*8 Peacter 70/*>butF8 stop check glotes i valves to core t3 1140t*3 headt*8 to 11ss o t
8 spray line (t) The pump flow rate is zero (pump not operating) during the standby enode, at 43 gpm durir q tha test modes (one pu:rp cperation) and is at rated system flow during the operatieg mode (two purp operation).
t *
Peactor to be at 0 psig and\\125*F before chaigisg from tha stardby mode to the iajectio, test mode.
j (83 Fressures tabulated represent pressure at the potats idertified below. To obtain pressure at iM e r me: State potrts ir p .the system, the pressures tabulated rust be adjusted for elevattor. ditterence a-d pressura drop bet weer nuen trter-g trediate points and the pressure points identified below: Piping Pressure Point 1 Furp suction Pump suction tlange inlet 4 rump discharge to explosive valve inlet Pump discharge tlange outla+ Explosive valve' outlet to, but not ircluding, motor-Explosive valve outlet cicrated stop check globe valves i Amer dmea t 13 l
HCGS FSAR 11/85 f 13.5 PLANT PROCEDURES Plant procedures are prepared by the plant staff, support organizations, or contract organizations under the direction of the General Manager - Hope. Creek Operations and implemented by the Operations Manager, Maintenance Manager, Technical Manager, and Radiation Protection / Chemistry Manager. Plant procedures are prepared for applicable safety-related activities delineated in Regulatory Guide 1.33 and provide the controls necessary to comply with applicable Regulatory Guides as listed in Section 1.8.
GEr I e
\\ Preparation of plant procedures necessary for fuel load has begun \\ and is scheduled to be completed ci mon nc: prior to the event. SE[Er-s> Review of t procedures :::cct a nc nucicar cc:c'ty/ and changes thereto l p are performed by the/ Station Operations Review Committee (SORC) wastr / and approved by the individual department managers for Hope Creek O Operations. Procedures are periodically reviewed and revised when changes are i necessary or desirable. Similarly, procedures are reviewed and revised when necessary following the completion of system design changes ~or equipment modifications. Subsequent to an abnormal occurrence, e.g., an unexpected plant transient, significant operator error, etc, the appropriate procedure (s) receive a review. The purpose of this review is to ascertain whether the procedure may have contributed to the cause of the abnormal. occurrence or was adequate in its capacity to mitigate the consequences. Circumstances may develop during maintenance, operation, or testing of the station systems when an existing' instruction or procedure is not entirely applicable as written or otherwise interferes with performance. In these instances, the existing procedure may be temporarily changed. All such changes to procedures are made so as not to change the intent of the SSERT existing procedure and are reviewed and approved by two members D,_ of station management, k.nowledgeable in the area affected, prior to its implementation.\\ Tor changer to precedure "9:cr may 1 affect the operational status of plcnt system Or equipment, On c, )L' 1 trember chall hcid an SnO licence, e 13.5-1 Amendment 13 ^- n
h 5 INSERT A FOR PAGE 13.5-1 l 4-A technical review and control system utilizing qualified reviewers functions to perform periodic or routine review j-of procedures. i i INSERT B FOR PAGE 13.5-1 1 1 L Station Administrative-Procedures, changes thereto and 4 implementing l INSERT C FOR PAGE 13.5-1 that involve a significant safety issue 1 INSERT D FOR PAGE 13.5-1 1 4 t The members of station management must be knowledgeable in the area affected and at least one must hold an SRO j ' license. i 1, - i 4 5 4 1
HCGS FSAR 11/85 13.5.1 ADMINISTRATIVE PROCEDURES Station administrative procedures are written to provide stationwide direction in areas that are common to all station departments. Administrative procedures are prepared using the following format: a. 1.0 Purpose b. 2.0 References c. 3.0 Definitions d. 4.0 Responsibilities e. 5.0 Procedure f. List of Attachments and Forms Furthermore, Section 5.0 may be subdivided, as appropriate, to facilitate the use of a specific procedure. Additional administrative procedures may be written as required. The following is a list of station administrative procedures: a. SA-AP.22-001 Preparation and Approval-of Station Procedures MD I This procedure describes the methods used in %V;Sb4 l ypreparation (format and organization), indexing, / revie" and accreva2 of all HCGS orocedures. Furthermore, thic procedure defincc the level of revice,J ' cnd cpprovc1 required for each individual precedure er - y .- _ _ _....<
w-....
.m. c -+. + m.
4.. - am-4,.,
4m.. _____m. _m. functional unit precedure, and whether er not thc-pcecedure effectc 0, R or F dccignated itc.T.a. b. SA-Ap.22-002 Station Organization and Operating. [. Practices \\ l 13.5-2 Amendment 13 1 1
_ _._ =.. d f HCGS FSAR 7/85 pc
[
establishes the administrative controls required for implementing-and maintaining such a program. These l controls are generally applicable but may be altered.by the various departments using approved implementing i j procedures to meet their specific needs. [ld&GRTn l y p. SA-AP.ZZ-024 Radiological Protection Program l y An overview of the radiation protection program is i provided, which includes pertinent information and instructions to plant staff, as_ dictated by 10 CFR 19. Additional requirements and instructions which affect 4 all or a majority of the station staff'are also i specified. These include: ALARA requirements, dosimetry and exposure limits, emergency exposure criteria, radiation work permit guidelines, and radiological protection requirements for visitors. i W,/. SA-AP.ZZ-025 Station Fire Protection Program l This procedure provides a program for prevention, detection, and control of fire hazards, the safeguarding of life, and prevention'of property damage 1 or loss due to fire at HCGS. i x y. SA-AP.ZZ-026 Nuclear Mutual Limited and Jurisdiction Boiler and Machinery Inspection Program This procedure identifies the requirements of, and ) assigns responsibilities for, complying with inspection j and testing requirements of the station's insurance j program. y x. SA-AP.AA-027 Station Inservice Inspection Pregram l y j This procedure establishes administrative r.ontrols, identifies general requirements, and assigns . responsibilities for inservice inspection (ISI) commitments of the HCGS. i 2//. SA-AP.ZZ-028 Reporting of Defects and Noncompliances l i
\\
t 13.5-7 Amendment 11 n-n, .,--y -e-,e .e 1 r-,.-4 ,e-w--.-.-> ,.vv w y-
,.,y----
,----v.--,, y
INSERT FOR PAGE 13.5-7 u. SA-AP.ZZ-023 Scaffolding Program This procedure provides administrative controls for the erection and use of scaffolding in the plant. l l
HCGS FSAR 7/85 g This procedure describes the action necessary to implement NRC 10 CFR 21, Reporting of Defects and Noncompliances. act g. SA-AP.22-029 Radioactive Waste and Material Control l y This procedure delineates responsibilities for the use, unconditional release, collection, processing, handling, packaging, inspection, receipt, storage and shipping of radioactive waste and material. bb.,ps. SA-AP.22-030 Station Response and Commitment. Control Program This procedure establishes a program for tracking, implementing review, and providing resolution to various action items brought to the attention of the Hope Creek Generating Station OC ybb. SA-AP.22-031 Station Housekeeping Program l This procedure delineates the housekeeping responsibilities and controls used to ensure the cleanliness of facilities, materials, and equipment used at HCGS. $PFESLF i cd.fcc. SA-AP.22-032\\tevisions.cnd Chanacs td Station wo Procedures / h 1&CEDUPE Revisioos This procedure describes the methods used ink llNSERf TOCOTT0nding, SC Mell SS 1mplOmOntin9, p0rm2nOnt Ond JJh' temporary (Or-the-cpet) changer te :pprc'Jed precedurc0 . :nd documentc. Cf.ydd. SA-AP.ZZ-033 Station Security Program l The requirements of the security plan, as described in Section 13.6 applicable to station personnel, are discussed in this procedure. In addition, description of responsibilities and authorities, employee and visitor access control criteria, security badging ( 13.5-8 Amendment 11 j
INSERT FOR-PAGE 13.5-8 ie g and approving newly created and changed station
HCGS FSAR 11/85 l } system, physical security system, etc, are also provided. Security. documents contain safeguard information and thus their distribution is strictly limited. I ff. yee. SA-AP.22-035 Station Reporting Requirements This procedure serves as a reference for the station staff to indicate which reports are required by federal, state, and local agencies, and provides the administrative controls necessary for ensuring that such reports are prepared, reviewed, and submitted correctly. gg.P&&. SA-AP.22-036 Phase III Startup Test Program l This procedure defines the responsibilities and procedures used during Phase III - Plant Operational Testing, which includes initial fuel load and the tests requiring the application of nuclear. heat. The testing 1 activities will b~e listed with a brief overview of ~' their scope. This procedure will be deleted at the start of the first refueling. hh. 799 SA-AP.ZZ-037 Environmental Control Program 1 This procedure provides an overview of the basic philosophies, policies, and objectives of HCGS to systematically control environmental' conditions to avoid accidental discharges, thereby minimizing j environmen.tal impact of plant operations. i E ykk. SA-AP.22-038 Control of Marerials Usage Program l This procedure establishes controls for the use and J disposal of those chemicals that-can be harmful to personnel or cause damage to plant systems or equipment. i i l JJ ykk. SA-AP.22-039 Receipt of New Fuel l f-t' i i ] 13.5-9 Amendment 13
I 1 i l l HCGS FSAR-11/85 ( This procedure identifies major refueling tasks and defines those departmental duties and responsibilities necessary to conduct refueling operations. kk. hvi. SA-AP.22-040 Master Equipment List This procedure defines the requirements for the control and use of the Hope Creek Master Equipment List. fl. ykk. SA-AP.ZZ-041 Confined Space Entry This procedure specifies the methods for personnel protection for entering confined spaces, am.S4. SA-AP.ZZ-042 Station Field Questionnaires This procedure describes the use of the FO to request Site Engineering Support. nn. tem. SA-AP.ZZ-043 Operability and Maintainability Enhancement This procedure establishes a mechanism to document a review of station design to ensure compatability with operating requirements and recommending improvements to enhance operability and maintainability of the as-built plant. SA-AP.22-044 Station Aids Program oo. Pen. This procedure describes the approval, documentation, and review requirements to ensure Station Aids are current, complete, and necessary. pp. foe. SA.AP.22-045 Respirator Protection D'EAM ~ This procedure provides methods for personnel protection related to the inhalation of both [' radiological and nonradiological agents. ( 13.5-10 Amendment 13 _O
HCGS FSAR 11/85 _k_ gg.ppp. SA-AP.22-046 Radiological Access Control This procedure describes the methods, policies, guidelines, limits and requirements for Restricted Area and Radiological Control Area access. fr. Pgg. SA-AP.22-047 Operating Experience Evaluation This procedure establishes a program which provides rapid dissemination of information pertaining to industry operating experience. ss.Fss. SA-AP.22-048 Station Performance and Reliability Monitoring This procedure establishes a program for monitoring and I trending plant process data to identify reductions in (~, unit efficiency or component performance, and to further evaluate the root cause and recommend appropriate corrective actions. ib.pss. SA-AP.ZZ-049 Conduct of Refuelingkind Corc Altcrationg l ,4 This procedure describes the administrative controls and departmental responsibilities during fuel _ handling and core alterations. UM htt. S -AP.22-050 Station Hetesw Program l This procedure establishes guidelines for determining post-maintenance retest procedures. VV. few. SA-AP.22-051 Leakage Reduction Program l This procedure describes the program for reducing leakage from systems located outside containment that ,-~ would contain highly radioactive fluids following an accident. 13.5-10a Amendment 13
HCGS FSAR 11/85 hM/. f**. SA-AP.22-052 Chemistry Control Program l This procedure establishes the program for ensuring that the water chemistry requirements for the NSSS and NSSS support systems are maintained within vendor and industry standards. In addition to these station administrative procedures, 1 operationally oriented administrative procedures provide guidelines for the operations senior shift supervisors and their shift crews, as well as procedures for night order book usage and control. Operations administrative procedures meet the j requirements of 10 CFR 50.54(i), (j), (1),.and (m). i Figure 13.5-1 indicates the main control room area designated as j "at the~ controls," the area restricted to licensed personnel and-the limitations of the reactor operator while manipulating the controls. 13.5.2 OPERATING AND MAINTENANCE PROCEDURES The operating and maintenance procedures meet the relevant requirements as discussed in Section 1.8. It is planned that most operating and maintenance procedures will be completed at least three months prior to fuel load and will b'e available for review in advance draft form at least six months prior to fuel load. This will provide sufficient lead time to ensure that plant personnel can become familiar with them. Where practical the preoperational testing phase will be used'to demonstrate the adequacy'of the operating procedures. 13.5.2.1 Main Control Room Ooeratino Procedures The following categories delineate those procedures that are performed primarily within the main control room. Operator familiarization with these procedures is acquired through initial, requalification and replacement training programs. Furthermore, these procedures will be utilized in simulator training. 13.5.2.1'.1 System Operating Procedures The procedures for startup, operation and shutdown of safety-- related Bh'R systems at HCGS will be called System Operating Procedures (SOPS). SOPS will be developed to cover the operating activities listed in Regulatory Guide 1.33, Appendix A, item 4 (',' and will include the following procedures: m 13.5-10b Amendment 13
1 HCGS FSAR 6/84 Reactor Pressure Control System Malfunction l l Reactor Vessel Level Control System Malfunction -l j Neutron Monitoring System Malfunction l 1 Fuel Pool Cooling and Cleanup System Malfunction l i Standby Liquid Control System Initiation l 13.5.2.1.4 Alarm Response Instructions l-Alarm response instructions guide operators in their response to i main control room alarm conditions. The alarm system at HCGS consists of control room overhead annunciators, console pushbutton alarms, computer (digital) alarms,.and local /back panel alarms. A color code system (red, amber and white) is utilized for prioritizing control room overhead annunciators. The computer and local panel alarms are associated.with:an ( overhead annunciator. The priority of these-alarms would be'the same as the associated overhead alarm. The alarm response procedures will be.available in the main control room.for the operators use. These procedures will be compiled in a manner which-is consistent with the alarm' system f layout in the control room. For example, the overhead annunciator response procedures will'be indexed by window box identification number. 13.5.2.1.5 Temporary Procedures l Temporary procedures.may be issued to direct operations during activities such as testing, refueling, maintenance, and modifications, and to provide guidance in unusual situations not ,A,within the scope of existing procedures. Temporary procedures. rece nnl in effect inc =peritted perinem nr ? -- er y require the same review.and approval process as other plant procedures,. including independent review, as described in Section 13.4. l l 13.5.2.2 Additional'Operatino and~ Maintenance Procedures The following categories Jelineate th'ose procedures that are performed primarily outside the limits of the main control room. .A i 13.5-17 Amendment.6 i -<.--q__m___7_.,.__.__
HCGS FSAR. 01/86 ( b. Prerequisites Section 14.2.10 (initial fuel loading) describes the prerequisites for commencing fuel loading. c. Test Procedure The fuel loading procedure. includes tests performed /ANoi durino the fuel loading evolution, including subcriticality checksA6 shutdown margin demonstration 'and control red functional tes t: J - /4 d. Acceptance Criteria Level 1: The partially loaded core shall be suberitical by at least 0.38% AK/K with the analytically determined strongest rod fully withdrawnf 14.2.12.3.4 Full Core Shutdown Margin a. Objective The test objective ~is to demonstrate that the reactor will remain subcritical throughout the first fuel cycle with the most reactive control rod fully withdrawn. l b. Prerequisites The core is fully loaded and in the xenon-free condition. c. Test Method The shutdown margin demonstration is performed by withdrawing selected control rods until criticality is reached. The empirical data are used to correct calculated values to obtain true shutdown margin. 14.2-156 Amendment 14 ( m _
INSERT FOR PAGE 14.2-156 or by at least 0.38% delta k/k with the reactivity equivalent of the strongest rod added by the withdrawal of other control rods. i
HCGS FSAR 01/86 [ times following planned reactor scrams as detailed on l Figure.14.2-5. In addition, proper response of the CRD flow control valve will be verified. d. Acceptance Criteria Level 1 The normal withdrawal speeds and-scram times shall meet the requirements of the GE startup test specifications. Level 2 The friction test results should meet the requirements of the GE startup test specifications. 14.2.12.3.6 Source Range Monitor Performance a. Objective The test objective is to demonstrate that the neutron-sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. b. Prerequisites Y Fuel loading is complete, treutron ecurce: 5:ve bec.1 Aut>l jujinct:12ec, :nc all control rods (hcre beed inserted, jfhe CRD system is operational. pgg (~ 14.2-158 Amendment 14
q HCGS.FSAR 01/86 e i, c. Test-Method With'the neutron sources-installed, source range monitor count rate data is taken and compared to'the required signal count and. signal count-to-noise count ratio. Source range data is taken during rod withdrawals to the point of-criticality. Rods will be withdrawn in accordance with a-pre-established withdrawal sequence. Movement of rods in a prescribed sequence is monitored by the RWM and RSCS which prevent out of sequence movement. d. Acceptance Criteria Level 1 There must be a neutron signal count-to-noise count ratio of at-least two and'a minimum neutron count rate. of 0.7 counts /second on the required operable SRMs. t 14.2.12.3.7-Rod-Sequence Exchange This-Test Has Been Deleted 14.2.12.3.8 Intermediate Range Monitor Performance a. Objective The test objective is to determine the IRM system overlap with the SRMs and APRMs 'and adjust the IRMS as
required, b.
Prerequisites The reactor is critical and the IRM gains have been optimized. r cer.ccr >:::c=.: .14.2-159' Amendment 14
HCGS FSAR 01/86 f above the. reactor pressure to simulate the largest. expected pipeline pressure drop. This CST testing is hWG8)l done to demonstrate general system operability and for makingut225 controller adjustments. Reactor. vessel injection tests follow to complete the-controller adjustments and to demonstrate automatic starting from a cold standby condition. " Cold" is defined as a minimum 72 hours without any kind of RCIC l operation. Data will be taken to determine-;he RCIC high steam flow isolation trip setpoint while injecting at rated flow to the reactor vessel. After all final controller and system adjustments have been determined, a defined set of demonstration tests must be performed. Two consecutive reactor vessel injections starting from cold conditions in the automatic mode must satisfactorily be performed to demonstrate system reliability. Following these tests, a set of CST injections are done to provide a benchmark 1 for comparison with future surveillance tests. After the auto start portion of certain of the above tests is completed, and while the system is still operating, s=all step disturbances in speed and flow command are input (in manual and automatic mode respectively) in. order to demonstrate satisfactory ~ stability. This is to be done at both low (above minimum turbine speed) and near rated flow initial conditions to span the RCIC operating range.. A demonstration 9of extended operation of up to two hours (or until rur; anc turbine oil temperature is stabilized) of continuous running at rated flow conditions.is to be scheduled at a conven.ient time during the startup test program. Depressing the manual initiation pushbutton is defined as autoriatic starting or automatic initiation of the RCIC' system. d. Acceptance Criteria ~ Level 1: l 1. Following automatic initiation,. the pump discharge flow must be equal to or greater than rated-flow as specified in Section 5.4.6 within the time specified-by the GE startup test specification. i 14.2-165 Amendment 14 3 4
l HCGS FSAR 01/86 which are based on not exceeding ASME Section III Code stress values. These specified displacements will be used as acceptance criteria in the appropriate startup test procedures. Level 2: l 1. All hangers and snubbers shall be within.their normal operating range. 2. The displacements at the established transducer locations shall not exceed the expected values as provided by the piping designer. W. 2.12.3. It [IXaER=D] .2.12.3.16 TIP Uncertainty a. Objective Th test objective is to demonstrate the repr.ucibility of the TIP system readin b. Prerequisi s l The core is at s ady-state p er level with equilibrium xenon, o as to equire no rod motion or change in core flow ma' tain power level during data acquisition by the TIP stem. c. Test Method 1. Core er distribution data e obtained during the ower ascension test progra Axial power d ribution data are obtained at ach TIP ocation. At intermediate and'high power levels, several sets of TIP data are tained to determine the overall TIP uncertainty. 2. TIP data are obtained with the reactor operat 9 with a symmetric rod pattern and at steady-stat conditions. The total TIP uncertainty for the 14.2-171 Amendment 14 n
HCGS FSAR 01/86 ( test is calculated by averaging the total TIP uncertainty determined from each set of TIP a. T IP uncertainty is made up of rand oise and geomet components, d. Acceptance Criteria Level 2: total TIP uncertainty shall be within the spe ed . limits required in the GE startup test specification. 14.2.12.3.17 Core Performance a. Objective The test objective is to evaluate the principal thermal and hydraulic parameters associated with-core behavior, b. Prerequisites The plant is operating at a steady-state power level. ~ c. Test Method With the core operating in a steady-state condition, the core performance evaluation is used to determine the following principal thermal and hydraulic parameters associated with core behavior: l. Core flow rate 2. Core thermal power level 3. MLHGR i l 4. MCPR i 14.2-172 Amendment 14 a
HCGS FSAR 01/86 'l Electric startup test specification requirements l when individually testing the MSIVs. 2. The RCIC and HPCI systems shall function in accordance with the GE startup test specification following the MSIV closure from high power. 14.2.12.3.24 Relief Valves a. Objectives 1. To demonstrate proper operation of the main' steam relief valves and verify that there are no major blockages in the relief valve discharge piping. 2. To demonstrate their leaktightness following operation. b. Prerequisites The reactor is on pressure control with adequate bypass or main steam flow to maintain pressure control throughout the relief valve opening transient. c. Test Method A functional test of each safety relief valve (SRV) y/ i Ng shall be madetac carly 1.- the startup prograr me / MTMW 10 *DM. practical. Thic tect 10 normally perforr d during the c:--. u_,..._ -.. m _
n. w -
4. u_. , =. _2 a: RATEDWM
"!., ". 'y::. !.L. '3
~ 7. i
".;, _'_'~..,~, r~ Bypass va1 vesK ( EPV )
pygg, f ggp' responselir T.^nitered during the Ice precrure tests a=d helectrical output M rate _.; pressure testR& % response is monitored during/estduration-willbe I%Ei lo seconds to allow turbine valves and tailpipe sensors to reach a steady state. The tailpipe sensor responses will be used to detect the opening and subsequent closure of each SRV. The BPVWEiE MWe responseQill be analyzed for anomalies indicating a restriction in an SRV tailpipe. Valve capacity will be based on certification _by ASME code stamp and the applicable documentation being 14.2-181 Amendment 14 [ 1
'HCGS FSAR 01/86 available in the onsite records. Note that the nameplate capacity / pressure rating assumes that the flow-is sonic. This will be true if the back pressure is not excessive. A major blockage of the-line would not necessarily be offset and it should be determined that none exists through the BPV/ response signatures. Vendor bench test data of the SRV capacity and setpoint is evaluated during~preoperatinal testing. The acoustic monitoring subsystem will be-monitored during. the relief valve test program and planned reactor
trips, d.
Acceptance Criteria Level 1: 1. There should be positive indication of steam discharge during the manual actuation'of each valve. Level 2: Decay ratio for pressure control variables is as 1. specified in the GE startup test specification. 2. The temperature measured by thermocouples on the discharge side of the valves should return'to the temperature recorded before the valve was open as required in the GE startup test specification. Y 3. During the Erduced red reted _rrrrrure functional tests, steam flow through each relief valve as compared to average relief valve flow is as specified.in the GE startup test specification. (-x_ l 14.2-182 Amendment 14 .+ .s
Y HCGS FSAR 01/86 i 14.2.12.3.25 Turbine Trip and Generator Load Rejection a. Objective The test objective is to demonstrate the proper response'of the reactor and its control' systems following trips of the turbine and generator. b. Prerequisites Power testing has been completed to the extent necessary for performing this test. The plant is stabilized at the required power level. c. Test Method j' This test is performed atim+HNem different power levels [TWol m in the power ascension program. For the turbine trip, the main generator remains loaded for a time so there is no rise in turbine generator speed, whereas, in the generator trip, the main generator output breakers open t i i 14.2-183 Amendment 14 l ^
HCGS FSAR 01/86 C and residual steam will cause a momentary rise in turbine generator speed. This speed will be monitored during each test. Y At test condition 3, a turbine trip will bc initiated 1 manually frer the centr:1 recr ' At test condition 6, a generator trip (load rejection) will be initiated by simulating a condition that will cause the generator output breakers to open. Duringf transientR_it is IDE' i expected that the reactor will scram and the T recirculation pump trip (RPT) breakers will open. It is not expected the HPCI or RCIC will initiate. Reactor water level, pressure, and heat flux will be monitored. The action of relief valves will be monitored. [TuRsmE Y AI=enera:Or, trip will be performed at low power such that nuclear boiler system steam generation is just within bypass valve capacity. The purpose of this test is to demonstrate scram avoidance. IBcm V Duringin11 inrec1 transients, main turbine stop, control, and bypass valve positions will be monitored. m g' Prior to the low power /,N, trip, bypass valve capacity will be determined. ), d. Acceptance Criteria Level 1: 1. For turbine and generator trips at power levels greater than 50%, the response times of the bypass valves shall be as specified in the GE startup test specification. 2. Feedwater control system settings must prevent flooding the main steam lines. 3. The reactor recirculation pump drive flow coastdown shall be as specified in the GE startup test specification. 4. The positive change in vessel dome pressure and heat flux must not exceed the limits specified in the GE startup test specification. 14.2-184 Amendment 14
1 HCGS FSAR 5/85 rs ) 5. The total time delay from start of turbine stop valve motion or turbine control valve motion to complete suppression of electrical arc between the fully open contacts of the RPT circuit breakers shall be less than the limit'specified in the GE startup test specification. Level 2: 1. The bypass valve' capacity shall be equal to or greater than that required by the GE startup test specification, which compares bypass valve capacity to the accident analysis. 2. There shall be no MSIV closure during the first three minutes of the transient and operator action shall not be required during that period to avoid the MSIV trip. 17hRBwE l- [ t 3. For the;;;cr.creten trip within bypass valves capacity, the reactor shall not scram for initial thermal power valves within that bypass valve capacity and below the power level at which trip scram is inhibited. 4. Low water level recirculation pump trip, HPCI and RCIC shall not be initiated. 5. Feedwater level control shall avoid loss of feedwater due to high level trip during the event'. l i 14.2-185 Amendment 10 n - ~,
1 1 l HCGS FSAR 01/86 () 1. To determine transient responses and steady-state I conditions following recirculation pump trips at selected power levels 2. To obtain recirculation system performance data 3. To verify that cavitation in the recirculation system does not occur in the operating region of the power / flow map. Tc ferify the adequacy Of the recirculation ' rur.b: k t: ?.itigste scre.? uper lor: Of On: feeductcr pump. 4(E To verify that the feedwater control system can control water level without causing a turbine trip / scram following a single recirculation pump trip. SE To demonstrate the adequacy of the recirculation pump restart procedure at the highest possible power level. b. Prerequisites The reactor is operating at steady-state conditions at required power level, c. T.est Method Single pump trips are performed at test condition 3 and 6. Dual pump trip is demonstrated at test condition 3. The.one-pump trip tests are to demonstrate that water level will not rise enough to threaten a high level trip of the main turbine or the feedwater pumps. The dual pump trip verifies the performance of the RPT circuit and the recirculation pump flow coastdown prior to the high power turbine generator trip tests. Single pump trips are initiated t y tripping the pump motor breakers. Adequate margins to scrams and capability of ~ the feedwater system to prevent a high level trip will be monitored. The two pump trip will be initiated by 14.2-189 Amendment 14 _n
HCGS FSAR 01/86 simultaneously tripping both recirculation RPT breakers using a test switch. The. recirculation pump restart demonstrates the adequacy of the restart operating procedure at the highest possible power level. At several power and flow conditions, and in conjunction with single pump trip recoveries, recirculation system parameters are recorded. At test acadition 2 and at near r:ted recircul: tion flee, 2 10cc Of a fccdwater pump i: rimul:ted. This 10 done prier t Or ctu:1 ft:de:ter pump trip to determine the 2dequ y Of recirculatic. pump run5: k fe:ture in pre /enting : Scr:r. 1 { While at test condition 3, it will be demonstrated that the cavitation interlocks which runback the recirculation pumps on decreased feedwater flow are i adequate to prevent operation where recirculation pump 3 or jet pump cavitation can occur. i d. Acceptance Criteria Level 1: 1. During recovery from one pump-trip, the reactor shall not scram. 2. The two pump drive flow coastdown time constant following a dual recirculation pump trip is as specified in the GE.startup test specification. Level 2: 1. Neutron flux and heat flux scram, and reactor water high level trip avoidance margins are as specified in the GE startup test specification. 2. System performance parameters, including core flow, drive flow, jet pump M-ratio,. core delta-pressure, recirculation pump efficiency and jet pump nozzle and riser plugging criteria are as specified in the GE startup test specification. o \\.. 14.2-190 Amendment 14
HCGS FSAR 01/86 j 3. Runback logic shall have settings adequate to prevent operation in areas of potential cavitation, y/ ' 4. Thc recirculaticn pumps shall runback upon & tr ip' l Of the runbacP circuit. J 14.2.12.3.29 Recirculation System Flow Calibration a. Objective The test objective is to perform a complete calibration of the installed recirculation system flow instrumentation, including specific signals to the plant process computer. b. Prerequisites (~.. The reactor is operating at steady-state conditions. \\' The initial calibration of the recirculation system flow instrumentation has been completed. c. Test Method During the testing program at operating conditions required for rated flow'at rated power, the jet pump flow instrumentation is adjusted to provide correct flow indication based on the jet pump flow. The flow-biased APRM/RBM system is adjusted to correctly follow core flow based on drive flow. Additionally, the total core flow and recirculation flow signals to the process computer will be calibrated to read these two process variables. d. Acceptance Criteria Level 2: 1. Jet pump flow instrumentation shall be adjusted such that the jet pump total flow anc::::: r; * ' gg,g77; recorder provides a/ core flow indication at rated-conditions. 14.2-191 Amendment 14
-w HCGS FSAR 01/86 y d. Acceptance Criteria Level 1: The piping displacements at the established locations shall not exceed the limits specified by the piping designer, which are based on not exceeding ASME Section III Code stress values or ANSI B31.1 values. These acceptable vibration levels will be used as acceptance criteria in the appropriate piping vibration startup test procedures. 14.2.12.3.32 Reactor Water Cleanup System a. Objective The test objective is to demonstrate the operation of the RWCU system. b. Prerequisites The reactor.has been operated at a near rated s-temperature and pressure long enough to achieve a steady-state condition. c. Test Method With the reactor at rated temperature and pressure, process variables 7are recorded during steady-state IMI operation inWLassenmodes of operation of the RWCU system: blowdown. not
nce;J and normal.
!The bettc= nc;d drain fica indicator wi44 be calibrated by taking fler from the better dr:in only and using the R'AC'J-2 ~~ cycter inlet flee indicater :: : Otand:rd te compara againct. d. Acceptance Criteria Level 2: 1. The performance data recorded during operation in the listed modes shall be acceptable as specified (~ by the GE startup test specification, y 14.2-194 Amendment 14 l l l r--
HCGS FSAR 01/86 Y 2. .90 calibrate better head flee indicater againet' RWC" ficw indicatcr if the deviatier ic greater than > luec cpecified ir the CE ctartup tect cpecificatienc. 2S. Pump vibration as measured on the bearing housing A and coupling end shall be less than or equal to the values specified in the GE startup test specifications. 14.2.12.3.33 Residual Heat Removal System a. Objectives 1. To demonstrate the ability of the RHR system to remove residual and decay heat from the nuclear system, so that refueling and nuclear system servicing can be performed ( 2. To demonstrate the capability of the RHR system to ( reduce the suppression pool temperature below the established limit immediately following a blowdown. b. Prerequisites Preoperational testing has been completed. f?he teet 1 'procedur; hac been reviewed, apprcred, :nd relenced for\\ itectinc J Instrumentation has been checked or calibrated as appropriate. c. Test Method Two modes are tested to verify system capability under actual operating conditions. The modes to be tested are suppression pool cooling and shutdown cooling. During the operations, the heat transfer rate is controlled to maintain acceptable cooldown rates. Data are recorded and reviewed to verify the satisfactory operation of the RHR system within design limits. / \\ 14.2-195 Amendment 14
TEST OPEN HEAT I NO. TEST NAME VESSEL UP 1 2 3 4 5 6 ( 22) 1 Gonical ary3 Ibdiochenical X X X X X X 2 Padiation Fbasurcraent X X X i X 3 Fuel loading 4 Eb11 Core Shutdown Margin X 5 Oantrol Ibd Drive X X X( 2) X( 2) ( X( 2) 6 Sm Ierfomance X 8 Im Perfomance X X 9 LPRM Calibration X X X X 10 APRM Calibration X X X X X X 11 Process Co puter X X X(3) X X 12 PCIC X X 13 HPCI X X 14 Selected Process 'Ibg h X XI4) 14 W1ter Icvel Pef Leg 'Itrnp / X X X 15 Systan Expansion X X X X X ,5 .n u =:rt'in wl 0 4 /~~ 17 03re Perfomance X X X X X X 18 Stean Production X 19 (bre Pwr-Void Mode response X X 2 Pressure Pegulator X X X X X X 21 Feed Sys-SetIx) int Changes X X X X X X X 21 Feed Sys-Ioss EW Heating X( 5) 21 Feedwater Ptrnp Trip X( 6) 21 Max EW Puncut Capability X(7) i 22 'Ibrbine. Valve Surveillance X(8) X(10) I11) X(12) X(13) 23 MSIV Functional Test X X 23 MSIV Full Isolation X 24 Felief valves X( 20) X( 4 X( 20) 25 'Ibrbine Trip & Ioad X 15)l (16 X( 17) X Pcjection 26 Shutriown Outside CRC X 27 Peciraalation Flow (bntrol X(14) X(18) 2 Pecirc-01e Pump Trip X X 2 RPT Tri m Pumps X(19) 28 Pecire System Performance i X X X X iZ Itcire n m Prhack-l [ 3 Ibcire Sys Cavitation 30 Ioss of Offsite Pwr X 31 Pipe Vibration X X X X X 29 Pecirc Flow Calibration X X 32 IWQ3 X( 23) 33 RHR X( 23) X( 21) 34 Drywell & Stean 'Iunnel X X X X 03oling 35 Gaseous Padwaste X X X 38 SACS Perfomance X X 40 Cbnfimatory In-Plant Test X FSAR 3/7 I
(1) 'Iust conditions refer to plant conditions on Figure 14.2-4 6 ( 2) Perform 'Ibst 5, timing of 4 selected control rods, in conjunction with expected scrams ~ X X (3) Dynamic System ' Inst Case to be canpleted between test mMitions 1 and 3 X( 2) (4) After recirculation pump trips (natural circulation) X (5) Betseen 80 aM 90 percent thermal power, X and near 100 percent core flow X (6) mx FW Runout Capability & Fecirc Pu p R:nback :xst have already been cmpleted X (7) Fuactor power between 80 and 90 percent X X (8) Ibactor power between 45 and 65 percent X aM 75 and 90 percent X (9) D31eted X X (10) At maximan power that will not cause scran X( 5) X( 6) (11) Ierform between test conditions 1 and 3 X( 7) ) X(10) (12) Feactor power between 40 and 55 percent 3) X (13) Icactor powr between 60 and 85 percent X( 20) X( 17) (14) Between test conditions 2 and 3 lTuRBrus TPap / (15)' " ^^+~ '^u -^4 within bypass 3) valve capacity y '10) Itacter pmer b^tren 60 W 90 p^rc^nt,[ X Ot C0r0 fl S' 1 95 F9rT nt - tu 6 1 T 'rIF i (17) Ioad rejection X (18) Between test mnditions 5 and 6 X (19) >50% powr and >95 mre flow, and perfonn X( 21) before 'Iurbine Yrip & Ioad Pejecticn X ( 2)) Geck SIU operability during major scran X tests X ( 21) Ferfonned during cooldown fran test condition 6 HOPE CREEK GENER ATING STATION ( 22) 'Ibe test ntrber correlates to FSAR Section FIN AL SAFETY AN ALYSIS REPORT 14.2.12.3 x where x is the iMicated test ntrber. TEST SCHEDULE AND CONDITIONS ( 23) my be perfomed any time test conditions permit. FIGURL 14 2 5 Ame,eent 14. 01,M i r
HCGS FSAR 01/86 l CHAPTER ~15 15 0.l-15.O.4 LEll?M=D ACCIDENT ANALYSES TABLE OF CONTENTS Section Title 15.0 GENERAL 15.0.1 Analytical Cbjective / 15.0.2 .n:1ytic:1 Categorice ^ 15.0.3 Transient and Accfdent Event Evaluation l 15.0.3.1 Identificetion cu Causes and Trcquency Clcccific: tion-15.0.3.2 Sequence of Evcats and System: Operation: 15.0.3.3 Ccre and Cystc; Performance 15.0.3.4 Ecrrier Performance-45.0.3.5 Radiclogical Concequcncc 15.0.4 Nucicar Safety Operation:1 Analyci 'NSOA) Esiationshis i 15.0.5 References 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 15.1.1 Loss of Feedwater Heating 15.1.1.1 Identification of Causes and Frequency Classification 15.1.1.2 Sequence of Events and Systems Operation 15.1.1.3 Core and System Performance 15.1.1.4 Barrier Performance 15.1.1.5 Radiological Consequences 15.1.2 Feedwater Controller Failure - Maximum Demand _ 15.1.2.1 Identification of Causes and Frequency Classification 15.1.2.2 Sequence of Events and Systems Operation 15.1.2.3 Core and System Performance 15.1.2.4 Barrier Performance 15.1.2.5 Radiological Consequences 15.1.3 Pressure Regulator Failure - Open 15.1.3.1 Identification of Causes and-Frequency Classification 15.1.3.2 Sequence of Events and' Systems' Operation 2 15.1.3.3 Core and System Performance 15.1.3.4 Barrier Performance 15.1.3.5 Radiological Consequences 15.1.4 Inadvertent Main Steam Relief Valve Opening 15-i Amendment 14 l
HCGS FSAR 01/86 l CHAPTER 15 FIGURES Ficure Title 15.0-1 k! Typical Power /F1cu "apl 15.0-2 Scram Position and Reactivity Characteristics l 15.0-3 Minimum Operating CPR Limit l 15.1-1 Loss of 1000F Feedwater Heating, Auto Flow Control l 15.1-2 Loss of Feedwater Heater, Manual Flow Control 15.1-3 Feedwater Controller Failure, Maximum Demand, with High Level Turbine Trip l 15.1-4 Pressure Regulator Failure l 15.2-1 Generator Load Rejection Trip, Reactor Scram l Bypass-On 15.2-2 Generator Load Rejection Trip, Reactor Scram, Bypass-Off 15.2-3 Turbine Reactor, Trip Scram, Bypass and RPT-On l 15.2-4 Turbine Trip, Reactor Trip, Bypass-Off, RPT-On l 15.2-5 Three-Second Closure of All Main Steam Line ll Isolation Valves with Position Switch Reactor Trip 15.2-6 Loss of Condenser Vacuum at 2 I_nches per Second l 15.2-7 Loss of All Grid Connections 15.2-8 Loss of All Feedwater Flow 15.2-9 ADS /RHR Cooling Loops 15.2-10 Activity Cl Alternate Shutdown Cooling Path Utilizing RHR Loop B 15.2-11 Activity C2 Alternate Shutdown Cooling Path Utilizing RHR Loop A 15.3-1 Trip of One Recirculation Pump l 15-xix Amendment 14 l
HCGS FSAR Chapter 15 ACCIDENT ANALYSES DELETED 15.0 GE: ERAL 5-In this chapter, the effects of anticipated process disturbances d postulated component failures are examined to determine the'r a Co. sequences and to evaluate the capability built into the plc it to c ntrol or accommodate such failures and events. The scope of the situations analyzed includes anticipat d sexpected) operational occurrences, e.g., loss of ele rical load; abnorm 1 (unexpected) operational transients .at induce system operati,ns condition disturbances; postulatrd accidents of 1:w probability, e.g., the sudden loss of integr' y of a major corponent; and hy,othetical events of extremely ow probability, e.g., an anticipate d transient without the ope.ation of the entire control rod..ive (CED; system. / 15.0.1 ANALYTICAL OBJ TIVE Tne cpectrum of postulated int iat'.g events is divided into categories based upon the type disturbance and the expected frequency of the initiating occ' ence. The limiting events in each ccmbination of category < df quency are quantitatively analyzed. The plant safety nalysis evaluates the ability of the plant to operate within re latory gut elines, without undue risk to the public health and afety. 15.0.2 ANALYTICAL ATEGORIES Transient and a cident events contained in this r'nort are discussed in iidividual categories as required by,, gulatory Guide 1.70, "evision 3. The results of the events a.e summarized in Table 1 .0-1. Each event evaluated is assigned to ne of the following applicable categories: { Decrease in core coolant temperature - Reactor ves I water /roderator terperature reduction results in an increase in core reactivity. This could lead to fuel c l a +i i n a % T a n e. 15.0-1
HCGS FSAR \\ b. Increase in reactor pressure - Nuclear system pressure increases threaten to rupture the reactor coolant pressure boundary (RCPB). Increasing pressure also collapses the voids in the core moderator, thereb" in-creasing core reactivity and power level, which threaten fuel cladding due to overheating. c. Dec ease in reactor core coolant flow rate - A redus"lon in the core coolant flow rate reatens to overhe t the cladding as the coolant be omes unable to ,p adequate' remove the heat generated the fuel. d. Reactivity a. power distributio anomalies - Transient events include-in this categor are those that cause rapid power inc ases due to creased core flow disturbance event Increased core flow reduces the void content of th_ modera' r increasing core y~ reactivity and power leve I e. Increase in reactor cla.t inventory - Increasing ( coolant inventory c uld rem 'I t in excessive moisture carryover to the ruin turbin. feedwater turbines, etc. l f. Decrease in r actor coolant inven ory - Reductions in coolant inv tory could threaten t fuel as the coolant be omes less able to remove he heat generated in the c .e. g. Radi active release from a subsystem or co. onent - Lc" of integrity of a radioactive containm t c;mponent is postulated, h Anticipated transients without scram (ATWS) - To determine the capability of plant design to accommo-te an extremely low probability event, a multisystem maloperation situation is postulated. 15.0-2
i HCGS FSAR 'N 5.0.3 TRANSIENT AND ACCIDENT EVENT EVALUATION 15.0. .1 Identification of-Causes and Frecuency Classific ion i i Situation. and causes that lead to the analyzed initia ing event i are descri 'd within the categories designated earli The frequency of ccurrence of each transient or accide t is 4 summarized, ba ed upon currently available operat' g plant j history for the vent. Events for which inconci sive data exists are discussed sep rately within each event sec ton. f j 1 Each initiating' event ithin the major gr ps is assigned to one i of the following freque y groups: l l a. Incidents of moder te fre ency - These are incidents i that can occur from nc during a calendar year to once in 20 years for a par 'cular plant. They are~ referred i to as " anticipated ( p cted) operational transients." i j b. Infrequent incid nts - Thes are incidents that can occur occasion ly during the' life of a particular ~ i plant, spanni g once in 20 yea to once in 100 years. They are re' tred to as " abnorm (unexpected) operation 92 transients." j c. Limit'ng faults - These are incidents at are not t expa ted to occur but are postulated be use their co sequences can result in the release of ignificant ...ounts of radioactive material. They are eferred to j as " design basis accidents (DBAs)." 1 l l Normal operation - Operations of high frequency a .not j discussed here but are examined along.with a'., b., nd j
c. above in the nuclear systems operational analyses 'n Section 15.9.
1 4 i i / + i i 15.0-3 1 I i . ~ ~.. n. - - - -,
HCGS FSAR 12/83 I 15.0.3.1.1 Unacceptable Results for' Incidents of Moderate / Frequency-Anticipated (Expected) Operational Transients The pilowing are considered to be unacceptable safety resulte for inKidents of moderate frequency, i.e., anticipated (expe ted) cperatio.al transients: A gelease of radioactive material to the env' ons that a. exc eds the limits of 10 CFR 20 b. Reactor operation induced fuel cladding failure c. Nuclear sys em stresses in excess o that allowed for the transien classification by a licable industry codes d. Containment stress in exce of that allowed for the transient classific ion by applicable industry codes. 15.0.3.1.2 Unacceptable Resul for Infrequent Incidents-Abnormal (Unexpe ed Operational Transients The following are considere to be unac ptable safety results for infrequent incidents, .e., abnormal perational transients: a. Release of r dioactivity that resul s in dose consequenc that exceed a small fra tion of 10 CFR 10 i.e., no more than 10% of he 10 CFR 100 limits .b. Fue damage that precludes resumption of no. mal op ration after a normal restart c. Generation of a condition that results in conseg ntial loss of function of the reactor coolant system (RC ) d. Generation of a condition that results in a consequential loss of function of a necessary containment barrier 15.0-4 Amendment 3
1 HCGS FSAR i / e. Nuclear system stresses in excess of that allowed for the transient classification by applicable industry i codes 1 1 Containment stresses in excess of that allowed f r the 2 j transient classification by applicable industry codes. j 15.0.3.1.3 Unacceptable Results for Limiting Faul s - Design i Basis Accidents, (DBAs) i The following ar considered to be unacceptable afety results for limiting faul (DBAs): i a. Radioactive.aterial release tha results in dose j consequences hat exceed the gu'deline values of 10 CFR 100 l i b. Failure of fuel cl
ding, hich would cause changes in-d core geometry such at re cooling would be inhibited c.
Nuclear system strese s 'n excess of those allowed for j the accident classi icati n by applicable industry Codes 4 d. Containment st esses in excess f those allowed for the accident cla ification by appli able industry codes when contai ent is required 4 e. Radiati exposure to plant operation personnel in the main c ntrol room in excess of 5 rem w le. body, 30 rem inhal tion, and 75 rem skin. 15.0.3.2 ouence of Events and Systems Operations Each tr nsient or accident is discussed and evaluated in rms of: a. The sequence of events from initiation to final stabilized condition 15.0-5
i f t HCGS FSAR 4/84 I b. The extent to which normally operating plant instrumentation and controls are assumed to function i The extent to which plant and reactor protection systems (RPSs) are required to function d. The credit taken for the functioning of normally l op ating plant systems e. The ope ation of engineered safety systems hat is required 4 f. The effect o a single activs failur (SAF) or a single operator error (SOE). 15.0.3.2.1 Single Failur or Operat Errors ] l This section discusses the appli ti of SAF and SOE to analyses \\. l of the postulated. events. Single tive component failure (SACF) criteria have been required and s e ssfully applied on past NRC-approved docket applications to BA c tegories only. Regulatory Guide 1.70, Revision 3, infers hat SA and SOE requirements [ should be applied to transie events, i cluding high, moderate, and low probability occurre ces, as well accident or very low probability situations. Transient evaluations ave been judged against iteria of one SACF or one SOE as e initiating event with no F assumptions added to the prote ive sequences although a great ajority of these protective equences use safety systems that n accommodate SAC aspects. Even under these postulate
events, the plant dam e allowances or limits are very much th same as those for no ial operation.
l The ori.nal categorization of events was based on' frequency f the in' iating event alone, and thus the al.lowance or. limit w acco ingly established based on that frequency level. .If add
ional assumptions and conditions (initial event and single c oponent failure (SCF) and/or SOE), were to be introduced,.the vents would be shifted to a lower frequency category.
Less. N ../ 15.0-6 Amendment 5
I l HCGS FSAR 4/84 i/ \\ restrictive limits or allowances would be' applied to the results ji of the analyses of transients and accidents. l Most vents postulated for consideration are already the resul 1 of single equipment failures or SOEs that have been postulate i during Any normal or planned mode of plant operations. The ypes of operat'onal single failures and operator errors conside ed as I initiating events and subsequent protective sequence cha enges ) are identif ad in the following paragraphs. i. i 15.0.3~2.1.1 itiating Event Analysis i I l To initiate an even one of the following acti ns must occur: i a. The undesired pening or closing f any single valve (a check valve is ot assumed to e ose against normal flow) J j b. The undesired starti or topping of any single l component I l c. The malfunction or m ope tion of any single control device d. Any single ele rical component failure e. The nonmee anistic break of a_proce s piping line or process i strument line. f. Any E. Operator e or.is defined as an active deviation from. itten operatin procedures or nuclear plant standard operating practic An SOE is the set of actions that is a direct conse ence of a single erroneous decision. This set of a ions is l'..ited as follows: a. Those actions that could be performed by one person s ,f< l 15.0-7 Amendment 5
HCGS FSAR 8/84 i / / b. Those actions that would have constituted a correct procedure had the initial decision been correct hose actions that are subsequent to the initi o erator error and have an effect on the des' p c. ned op ation of the plant, but are not necessa ly dire tly related to the operator error. Examples of SOEs a as follows: a. An increase i power above the ep ablished flow control power limits b control rod witKdrawal in the specified sequences b. The selection and cou let withdrawal of a single control rod out of seg ce I c. An incorrectIcalibr tion o an average power range -monitor (APRM) k_ d. Manual isola 'on of the main ste lines as a result of operator mi interpretation of an a em or indication. 15.0.3.2.1.2 S gle Active Component Failure o Single Operator tror Analysis These fail es include: The undesired action or maloperation of a single tive component b. Any SOE where operator errors are as defined in Section 10.0.3.2.1.1. C 15.0-8 Amendment 7
- ~ HCGS FSAR 8/84 15.0.3.3 Core and System Performance / S.ction 4.4 describes the various fuel failure mechanisms. Av idance of unacceptable results a. and b. in in Section 4.4. .4 for ~ ncidents of moderate frequency is verified statisticall with nsideration given to date, calculation, manufacturin, and cperati g uncertainties. An acceptable criterion is that 9.9% of the f el rods in the core are not expected to.experie e boiling t nsition. For more detail, see Reference 15. -1. This criterion i met by demonstrating that incidents of m erate frequency do ot result in a minimum critical power tio (MCPR) less than 1.0 The reactor steady-state critical ower ratio (CPR) operating limit is derived by determining t decrease in MCPR for the mos limiting event. All other eve s result in { smaller MCPR decr ses and are not reviewed in epth in this chapter. The MCPR uring significant abnorma events is calculated using a t nsient core heat trans er analysis computer program. The compute program is based on multinode, single-channel, thermal-hydrau ic model that re ires simultaneous solution of the partial ifferential eg tions for the conservation of mass, ene y, and mome um in the bundle, and that accounts for axial var ower generation. The primary inputs to the model(ation in b clude physical description of the bundle, and channel inlet flow and enthalpy, pressure, and power generation as functions of time A detailed description of th analy ical model may be found in Appendix C of Reference 15. -1. Main aining MCPR greater than the safety MCPR limit is sufficient, but not necessary, condition to ensure that no fuel damage occurs. This-is discussed further in S tion 4.4. For situations in ich fuel damage is sustal ed, the extent of damage is determi ed by correlating fuel energ content, cladding temperature, fu rod internal pressure, and c1 dding mechanical characteristic. These corr lations are substantiated by fuel rod fai re tests and are scussed in Sections 4.4 and 6.3. 15.0 .3.1 Input Parameters and Initial Conditions for Analyzed Events (' In general, the events analyzed within this section have values for input parameters and initial conditions as specified in 15.0-9 Amendment 7
HCGS FSAR 8/84 Table 15.0-3. Analyses that assume data / inputs different from hese~ values are designated accordingly in the appropriate event 'scussion in this chapter. These transient analyses presuppose, fo the end-of-cycle-one (EOC-1) conditions being simulated, th the lant design includes a recirculation pump trip (RPT) actua ed by either fast closure'of the turbine control valve or closur of the main stop valve. An EOC-1/RPT system is par of the cur nt plant design. 15.0.3.3.2 Initial Power / Flow Operating Constraint The basis for mo t of the transient safety.analys is the thermal power at he 100% rated core flow corres nding to 105% nuclear-boiler-rat (NBR) steam flow. This op rating point is the apex of a bound operating power / flow ma that in response to any classified abn rmal operational trans' nts, yields the minimum pressure and t ermal margins of any operating point within the bounded map. As shown on Figur 15.0-1, the apex of the bounded power / flow ma is point A; t upper boundary is the design flow control line, ich is 104.. rod line A-D'; the Iower boundary is the zero wer line -J'; the right boundary is the rated pump speed line -H'; a d the left boundary is either the minimum pump speed ine -J or the natural circulation line D'-J'. N_ This power / flow map, A-D'-J'- A, r resents the acceptable operational constraints for normal peratio'nal' transient evaluations. Any other constraint t t can truncate the bounded power / flow
map, e.g.,
the recire ation valve and pump avitation regions, ~ the licensed power I'mit, and other restrict 1 ns based on pressure and therm margin criteria, must be bserved. For instance, if the censed power is 100% NBR, th power / flow map is truncated by e line B-C and reactor operati must be confined withi the boundary B-C-D'-J'-J-L-K-B. the maximum operating pow
level, e.g.,
point F, has'to be lim ted to satisfy pres ure margin criteria, the upper constrai on power / flow s correspondingly reduced to the rod line, e.g., line F-G', whi intersects the power / flow coordinate of the ew operati basis. In this case, the operating boundaries re F-G'-J' J-L-K-F. Operation is not allowed at any point al ng line -M, removed from point F, at the derated power but at i red ed flow. If, however, operating limitations are impose by j Ge eral Electric Boiling Water Reactor Thermal Analysis Basis ' ~ ' ( ETAB), Reference 15.0-1, derived from transient data with an peratin~g basis at point A, the power / flow boundary for 100% NBR (, 15.0-10 Amendment 7
HCGS FSAR 8/84 i '\\ licensed power is B-C-D'-J'-J-L-K-B. /This power / flow boundary is uncated by the MCPR operating limit, for which there is no di ect correlation to a line on the power / flow map. Operation s all wed within the defined power /ficw boundary and within the cons aints imposed by GETAB. If operation is restricted to point " by the MCPR operating limit, operation at point M i allowec provided that the MCPR limit is not violated. Consequentl the upper operating power / flow limit of reactor is predicate on the operating basis of the analysis and the corresponding onstant rod pattern line. This boun ary can be truncated by th licensed power and the GETAB ope ting limit. Certain localized ents are evaluated at oth than the above-mentioned conditions These conditions are cluded in the appropriate event dis ssion in this chapte 15.0.3.3.3 Results The results of analytical eva uatio s are provided for each event. In addition, critical ra.eters are shown in Table 15.0-1. From the data in ble 15.0-1, an evaluation can be made of the limiting event f that particular category and parameter. Table 15.0-2 provi es summary of applicable accident analysis results. 15.0.3.3.4 Regulator Guide 1.49; G eral Compliance or Alternat Approach Assessm nt For commitment, rev' ion number, and scope, e Section 1.10. Regulatory Guid 1.49 requires that the proposed icensed power level be restr'cted to a reactor core power level f 3800 megawat thermal or less, and that analyses a d evaluations in support the application should be made at 1.02 imes the proposed l' censed power level. The compliance is sho in Table 15. -3. 15.0. .4 Barrier Performance D ring transients that occur with no release of coolant to the _ontainment, only RCPB performance is considered. Ifreleaseto}3 15.0-11 Amendment 7
I HCGS FSAR 8/84 ( l \\(he containment occurs as in the case of 1imiting faults, then c allenges to the containment are evaluated as well. \\ Conta' ment integrity is maintained as long as internal pr sures remain elow the maximum allowable values. The design in rnal pressure are as follows: a. Dr well (prime.cy containment), 58 ps'ig b. Suppre sion chamber (primary containme ), 58 psig c. Reactor bu'1 ding enclosure, 6.5.in es of water. Damage to any of the ra 'oactive materia barriers as a result of accident-initiated fluid 'mpingement an jet forces is considered in other sections'of the F where me hanical design features of the systems and components a e descr' ed. DBAs are used to determine the size and streng h reg irements of the essential nuclear system components. A omp rison of the accidents considered in this section with ose used in the mechanical ( design of equipment reveals eit e no difference or less severe stresses than those assumed fo me anical design. 15.0.3.5 Radiolocical Con ecuences Consequences of radioa tivity release duri g the following three types of events are nsidered-a. Incide s of moderate frequency.(anti pated opera onal transients) b. I requent incidents (abnormal operational ransients) c. Limiting faults (DBAs).. For 11 events whose consequences are limiting, a detailed qu ntitative evaluation is presented. For nonlimiting events, a alitative evaluation is presented,.or the results are i eferenced from a more limiting or enveloping case. ( ]- b 15.0-12 Amendment 7 l .. ~. -...a n
HCGS FSAR 8/84 ( \\e or limiting faults (DBAs), two quantitative analyses are v c sidered: a. The first is based on conservative assumptions considered acceptable to the NRC for the purp es of unding the worst-case event and determini the a quacy of the plant design to meet 10 CF 100 gui lines. This analysis is referred t as the "desi basis analysis." b. The second based on realistic sumptions considered ~ to reflect ex ected radiologica consequences. This analysis is re cred to as th " realistic analysis." Results for both are shown to e w' hin NRC guidelines. 15.0.4 NUCLEAR SAFETY OPE IONA ANALYSIS (NSOA) RELATIONSHIP Section 15.9 is a compr ensive, total-p nt, system-level, _,( qualitative failure m e and effects analy 's (FMEA),~ relative to (/ all the Chapter 15 e ents considered, the pr ective sequences used to accomodate he events and their effect and the systems involved in the otective actions. Interdepend ncy of analysis and cro'ss-referral of pro ctive actions i an integral part of this chapter and appendi Con ined in Section 15.9 is a summary' table that classifies ev nts by frequency only, i.e., not just within a given catego uch.au; " decrease in core. coolant temperature. " 15.
0.5 REFERENCES
,e 15.0-1 ' Gen Electric, General
Electric BWR Therm Analysis is (GETAB):
Data, Correlation 7"and Desian Applichtron, NEDO-10958 and 'EDE-1095 8, ((ELEH5DI November 1973. j 15.0-2 General Ele c, Qualification o_ e
s.,
One-Di ional Core Transient Model oiling Jiatt5r Reactors, NEDO-24154, October 1978. N 15.0-13 Amendment 7
HCGS FSAR 8/84 DETETEDj N.. '1 d, Analytical' Methods o 15.0-3 R Transient Eva u trent eral Electric Boillna Wate General W pri1 1973. 64-4 "(DER 4( F_4EtziaC STAubA@ APR CATrcy fibe PeAcroe Fua.,' inclubsoG we " Uuasu 5r4765 SttPReNs7Jr, # N6DE c24ol1 - P-A- -? AMD NGCG e2(lOII-p- A - V - us. t [' g s
4 3
UD 15.0-14 Amendment'7
b ~ HCGS FSAR ( TABLE 15.0-3 (cont) '(Page 3 of 3) 36. Pressure setpoint of RPT, psig 1101 ?qp1;coble tc cvent.c onclyced using model dcccribed in \\ (1) 9eference 15.0.3. The ine.rtia time constant is defined by the expression: (2) t= 2r Jnn g To where: inertia time constant, s t = pump motor moment of inertia, Ib-ft2 Jo = rated pump speed, rps n = gravitational constant, ft/s2 g = pump shaft torque, ft-lb. To = _ h TRAostans simmTeb N we ODYM ccmwasR ~ mtsL, wts tourr IS CALCutA76b BY ODYN ~ v'. - " " - - ^ ~ " -,,m
HCGS FSAR f t TABLE 15.0-4 Required Operating Limit MCPR Values OLMCPR-OLMCPR Pressurization Events (Option A)(2) (Option B)(1) Load rejection with bypass 1.14 1.06 Load rejection without bypass 1.18 1.10 Turbine trip with bypass 1.12 1.06 Turbine trip without bypass 1.17 1.09 Feedwater controller failure with bypass 1.20 1.17 Nonpressurization Events OLMCPR Loss of feedwater heating, MFC 1.20 Rod withdrawal error (RBM=106%) 1.20 ( Rk1 (1) Option A and B include adjustment factors as specified c3yyg mercrc=cc ::.c. : 4 6 <N v
t, jag N_- 120 A B 100 C E m 2 p 8 DEslGN FLOW CONTRO LINE M E 80 h TYPICAL POWER / FLOW DER ATE LINF c. .J< ba=w l 60 D D'i 40 G' E C 3 8 E S O D L o a k U-K a 20 - N ATUR AL flRCULATION Q g Q C C CAVITATION REGION MINIMUM PUMP SPE ED J H H g g g J O 0 20' 40 60 80 100 120 CORE F LOW (% NSRI HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSIS REPORT j. t \\L u,= eom e me,.,.= EM F1GURE 15.01
HCGS FSAR 8/83 15.3.4.5 Radiolocical Consecuences s The radiological consequences of this event are the same as discussed in Section 15.3.1.5. i 15.3.4.6' SRP Rule Review a SRP 15.3.3 - 15.3.4 acceptance criterion II.10 states that analysis for reactor coolant pump rotor seizure and reactor coolant pump shaft break events should include assumptions of turbine trip and coincidental loss of offsite power (LOP) and coastdown of undamaged pumps. Coincidental LOP and ;urbine trip are not assumed in the HCGS analysis but would, if included, produce consequences ~less severe than those of Section 15.2-6. The turbine trip or, indirectly, the loss of offsite power, will initiate reactor scram and rapid power reduction. The severity of pump shaft seizure or pump' shaft break without assuming LOP is f evidenced by the fast coastdown of core flow,.which reduces the f\\- thermal margin significantly before a reactor-scram is initiated by an L8 signal. l 15.
3.5 REFERENCES
l 15.3-1. General Electric, General Electric Standard Aeolication for Reactor Fuel, including the United States suoclement, NEDE-24011-P'A and NEDE-24 011-P-ArOS. E ! s ter t approved reviricr?2 (,P. x_ 15.3-16 Amendment 1 c -7 y e -r -y y v
HCGS FSAR 10/83 /< \\' 15.4.1.2.4 Barrier Performance An evaluation of the barrier performance is not made for this event since there is no postulated set of circumstances for which . this error could occur. 15.4.1.2.5 Radiological Consequences An evaluation of the radiological consequences is not required for this event since no radioactive material is released from the fuel. 15.4.2 ROD WITHDRAWAL ERROR - AT POWER This event is described in Ecction 5.2.1.5 on Reference 15.4-3. Analysis specific to the first cycle was performed. The limiting rod pattern is shown on Figure 15.4-1, and a summary of the results is presented in Table 15.4-22. For the selected rod block monitor setpoint, the change in the critical power ratio ( (aCPR) is 0.132 and the maximum linear heat generation rate (MLHGR) is ~5.4 kw/ft. 15.4.3 CONTPOL ROD MALOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR) This event is~ covered by the evaluation cited in Sections 15.4.1 and 15.4.2. 15.4.4-ABNORMAL STARTUP OF IDLE RECIRCULATION PUMP 15.4.4.1 Identification of Causes and Frequency Classification 15.4.4.1.1 Identification of Causes Abnormal startup of an idle recirculation pump results directly from the operator's manual initiation of pump operation. It assumes that the remaining loop is already operating. W 15.4-5 Amendment 2
HCGS FSAR 10/83 s 15.4.10 REFERENCES 15.4-1 C. J. Paone, Bank Position Withdrawal Secuence, NEDO-21231, September 1976. 15.4-2 Deleted 15.4-3 General Electric, General Electric Standard Aeolication For Reactor Fuel, including the United FTF-4 States Suoniement, NEoE-240ii-e A, ana NEDE-24oii-F-a>US.nLatect ^pprovec Merisier ; 1 15.4-4 Stancavage, P.P. and Morgan, E.J., Conservative Radiolooical Accident Evaluation - The CONACOI Code, NEDO-21143, March 1976. 15.4-5
Nguyen, D.,
Realistic Accident Analysis - The RELAC Code, NEDO-21142, October 1977. General Electric,[COncr21 Electric StOnd2rd ' pol l CO t 10nl for P,COctor FuOl, including th0 UnitOd StatOO Supplcmcnt NEDE 24011 P A and NEOE 24011 P A US, 'latect approved revicic~r}. 1 15.4-6 N.R. Horton, W.A. Williams, K.W. Holtzclaw, Analvtical Methods for Evaluatina the Radioloaical Asoects of General Electric Boilina Water Reactors, APED-5756, March 1969. 15.4-7 Nuclear Regulatory Commission, Standard Review Plan, NUREG-75/037, Washington, D.C., November 24, 1975. q J 15.4-22 Amendment 2
HCGS FSAR 11/85 t v. Regulatory. Guide 1.146, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants w. BTP 9.5-1, Appendix A, Guidelines for Fire Protection for Nuclear Plants Docketed Prior to July 1, 1976. Commitments to Regulatory Guides, with respect to revision level, exceptions, etc, are contained in Section 1.8. Substantive changes to the OA program described herein will be j submitted to the NRC within 30 days of implementation. Nonsubstantive changes will be identified in the annual FSAR updates. The overall OA program is described in the Nuclear Department manual. This description is prepared and maintained by NOA. PSE&G organizations performing activities affecting nuclear safety prepare and maintain implementing; procedures and instructions. These procedures and instructions, and subsequent revisions thereto, are subject to.NOA review and approval to the extent necessary to verify compliance with the OA Program and the applicable quality-related Regulatory Guides and standards identified above. NOA will monitor the preparation and issuance of required procedures to assure that they are in place prior to implementation of activities needed to support systems turnover to station operations and that all required procedures are in place bt lecct 00 davciprior to fuel load. 7 The General Manager - Hope Creek Operations has' inst _ituted and will maintain an administrative procedures (AP) manual for Hope Creek Generating Station (HCGS). The station APs and all subsequent revisions thereto are prepared l by the technical staff, are reviewed by the Technical Engineer, Technical Manager, NOA and SORC, and are approved by the General Manager - Hope Creek Operations. Regulatory Guide 1.33 requires that plant activities affecting quality-related items and services be conducted in accordance with written administrative controls prepared by management. The procedures and instructions by which plant activities are 17.2-9 Amendment 13 L ~; -.
I HCGS FSAR 11/85 performed are prepared by the responsible station organization as' required by station APs,' reviewed by the organization responsible for the activity, reviewed by NOA for quality requirements, reviewed by the SORC (for procedures affecting safety), and approved by the department manager. In the absence of a department manager, procedures will be approved by the assistant general manager or his designee. Procedures cannot be implemented unless the review / approval process is accomplis!2d. Station APs provide a means to accommodate on-the-spot changes to l subtier implementing procedures. The routine practice for revising a procedure is to repeat the original review and approval sequence. ---*-Implementation of the OA program is verified by means of independent inspections, monitoring, and audits conducted by NOA. NOA reviews and analyzes problems affecting safety that occur during the operational phase. Items subject to review include: a. Documented nonconformances occurring at the vendor's facility and those during receiving, storage, installation, test, and operation, e.g., Deficiency Reports, Nonconformance Reports, Licensee Event Reports, etc b. Documented corrective actions taken on significant noncompliances and.on audit findings c. NRC inspection findings, notifications, bulletins, etc. The General Manager - Nuclear Quality Assurance, or his designee, has tne authority to stop work through the issuance of a stop work order where continuance of an activity would seriously compromise safety or constitute a persistent'and deliberate failure to correct a serious deficiency. Designees include the Manager - Station Quality Assurance for activities conducted at the station and the Manager - QA Engineering and Procurement for supplier activities. NOA reports significant problems affecting the quality assurance program to respective management along with: a. Measures taken to improve OA program controls l Tue QA Pacces saatt Be in1Reviemse 90 C4YS TRioR To kt1EL LOAb. 17.2-10 Amendment 13 .~i.. -
HCGS FSAR 10/84 ( OUESTION 410.26 (SECTION 4.6) Provide the information requested in our generic letter 81-34, dated August 31, 1981, regarding NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of Bh? Scram-Syster Piping." / I
RESPONSE
L HCGS is participating in the Bh'ROG activities related t'o the scrar discharge pipe intecrity. The Bh'ROG's final response to the NRCMc neing - prepare icr NRC revie-and 2ppr-cval. I. SCCS p;;.m specific responce will bc prohi-thin 60 dayc of NRG-r ec c l u t t e r o f t h e E*@DS per i t i on,---HCGS v i l l implement any-required fir by the end of--the next refueling cutage which ic at Icact 12 mounthe af ter "RC recolution. P e n d i n g '".s t e r i a l availchility, thin schedule may change wi th NRG-approva-1,, .l HM D=9J PiDVt@ WHILG Ts= }KCS PJJT SPEC 1RC WM EE t; SHov3!3 l'd ( 65t7to 4.6.5. l. T.-f. l / i 1s. 410.26-1 Amendment B .}}

ML20210D765

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Dear Ms. Adensam:

SUMMARY

1.5 REFERENCES

2.1 REFERENCES

2.2 DESCRIPTION

2.5 REFERENCES

3.5 REFERENCES

4.2 DESCRIPTION

4.3 DESCRIPTION

4.8 REFERENCES

3.6 REFERENCES

0.5 REFERENCES

3.5 REFERENCES

RESPONSE

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ML20210D765

Text

Dear Ms. Adensam:

SUMMARY

1.5 REFERENCES

2.1 REFERENCES

2.2 DESCRIPTION

2.5 REFERENCES

3.5 REFERENCES

4.2 DESCRIPTION

4.3 DESCRIPTION

4.8 REFERENCES

3.6 REFERENCES

0.5 REFERENCES

3.5 REFERENCES

RESPONSE

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