ML20199K695

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Forwards Summary,Explanation & marked-up Sections of FSAR for Incorporation Into FSAR Amend 15 After Fuel Load. Submittal Suppls 860303,17,21 & 0404 Transmittals
ML20199K695
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 04/08/1986
From: Corbin McNeil
Public Service Enterprise Group
To: Adensam E
Office of Nuclear Reactor Regulation
References
NUDOCS 8604100210
Download: ML20199K695 (54)
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{{#Wiki_filter:Public Service Electric and Gas Corrpany Ccrbin A. McNeill, Jr. Public Service Electnc and Gas Company P.O. Box 236, Hancocks Bridge NJ 08038 609 339-4800 %ce President - Nuclear April 8, 1986 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Ms. Elinor Adensam, Director Project Directorate 3 Division of BWR Licensing

Dear Ms. Adensam:

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

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

These revisions will be incorporated in FSAR Amendment 15 after fuel load but are being filed now in order to accurately reflect the design and operation of HCGS and support the issuance of an operating license. In addition, an affidavit is provided to affirm that the matters set forth in this transmittal are true and accurate. This submittal supplements similar transmittals from C.A. McNeill to E. Adensam dated March 3, March 17, March 21, and April 4, 1986. 8604100210 860408 {DR ADOCK 05000354 P DR B e= t '\\r

) Director of Nuclear 2 4-8-86 Reactor Regulation Should you have any questions on the subject filing, do not hesitate to contact us. Sincerely, Affidavit Attachments (2) C D. II. Wagner USNRC Licensing Project Manager R.W. Borchardt USNRC Senior Resident inspector

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 i PUBLIC SERVICE ELECTRIC AND GAS COMPANY FINAL SAFETY ANALYSIS REPORT REVISIONS i 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 4 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. l i 1 l Respectfully submitted, Public Service Electric and Gas Company By: N _A ~ Corbin A. McNeill, JA Vice President - Nuclear i Sworn to and subscribed before me,- a Notary Public of New Jersey,nthis $ M_ day of Aprii,1986 my M musYt1980

1 ATTACHMENT 1

SUMMARY

OF CHANGES, ADDITIONS AND/OR MODIFICATIONS 1.8.1.68 + Revision provides additional clarifi-14.2.12.3.24 cation regarding relief valve testing. The revision to Page 14.2-181 supercedes the previous transmittal to the NRC' in a letter from C.A. McNeill to E. Adensam dated March 21, 1986. These revisions reflect various power ascension test program modifications approved by the NRC in a letter to PSE&G dated March 3, 1986. 1.8.1.97 Revision provides a further description 6.2.4.2 of the high-low pressure interface 9A.S.4 evaluation in the fire hazards analysis 2 i GL81-12 and describes commitments made in response (pg. 9A-94,95) to an NRC telecon on April 3, 1986. 7.5.1.3.7.6.c Revision provides a correct reference to the notification requirements for loose parts monitoring. 7.7.1.2.2 Revision reflects the as-built plant F7.7-6 condition. 7.7.1.3.2.3b Revision necessary to reflect the as-built plant condition. This revision impacts SER Section 7.7.1.3 in that the feedwater control system does not contain single element control. F8.2-2 Revision provides the correct Amendment revision of this figure and hence super-l cedes the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adensam dated March 21, 1986. F8.3-5 Revision reflects the as-built plant condition. 9.1.4.2.3.1 Revision provides clarity and ensures ,I consistency with the Technical Specifications regarding shielding requirements for irradiated fuel assemblies when elevated by the fuel hoist or fuel preparation i machines. i l

2 T9.3-3 Revision necessary to correctly reflect pg. 1/3 the as-built plant condition regarding the RWCU inlet sample point. 9.4.2.3 Revision necessary to maintain consistency with the Technical Specifications. T9.4-9 Revision increases the flow rate of service area supply system and chemistry exhaust system to provide clean air to the clothes dryer. F9.5-26 Revision provides additions recommended sh. 1/2 by Colt Industries-Fairbanks Morse Engine Division regarding improvement modifications of the starting air system for the emergency diesel generators j (reference: letter from J.M. Moriarty (Colt) to J.G. Keppler (NRC-Region III) dated February 4, 1986). 10.4.2.2.2 Revision correctly reflects the as-built plant condition. l 11.2.1.1 Revision deletes the liquid radwaste tank dike since Regulatory Guide 1.143 does not require such a design. This revision to Page 11.2-3 supercedes the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adensam l dated March 17, 1986. T11.2-10 Revision correctly reflects the as-tested pg. 2/5 plant condition. T11.2-14 i pg. 1/2 T12.3-2 Revisions necessary to correctly reflect T12.3-3 control room doses, containment leakage T12.3-3a rate and maintains consistency with l pg. 1,3/3 Section 6.4. l 14.2.12.3.19 + Revision reflects modifications to l 14.2.12.3.20 the power ascension program approved 14.2.12.3.23 by the NRC in a letter to PSE&G dated l F14.2-5 March 20, 1986..The revision to Figure l 14.2-5 supercedes the revision transmitted to the NRC in a letter from C.A. McNeill to E. Adensam dated April 4, 1986. 1 l l l l l

i 3 14.2.12.3.27 Revision deletes an incorrect reference to the "mid-power reactor" which was not identified in the Amendment 14 revision to the recirculation flow control startup test. This revision to Page 14.2-187 supercedes the revision transmitted to the NRC in a letter l from C.A. McNeill to E. Adensam dated March 3, 1986. 17.2.5 Revision provides clarification regarding station inspection plans and procedures review and approved by Nuclear Quality Assurance. Q430.35 Revision provides recently completed information on the analysis for design applications of all non-Class-lE devices connected to a Class lE power supply. These revisions impact the SER as noted + These revisions have already been accepted by the NRC in a letter to PSE&G as noted l l I l l

(, i l l I l l ATTACHMENT 2 i i l l l l l l l 1 l l I l l 1 \\ r I 1 I l l l Q

i HCGS FSAR 12/83 l ( p. Appendix A, Paragraph 2.e - Compliance with Regulatory Guide 1.56,- Haintenance of Water Purity in Doiling Water Reactors, is addressed in Section 1.8.1.56. q. Appendix A, Paragraph 4.m - Following f uel load, there is no planned startup test of the MSIV leak control system. The preoperational test demonstrates the operability of the system at design conditions. Testing following fuel load does not contribute any additional meaningful data. - --e. r. Appendix A, Paragraph 5.j - Rod runback and partial scram testing is not performed because the plant does not have this design feature. s. Appendix A, Paragraph 5.n - Although there will be no startup test procedure designated loose parts monitoring, additional data to supplement the preoperational program on loose parts monitoring will be taken as stated in Section 14.2.10. t. Appendix A, Paragraph 5.q - There are no startup tests of the failed fuel detection systems. Preoperational testing and periodic surveillance testing after fuel load ensure the proper operation of radiation monitoring systems used for isolation signals in case of gross fission product release. Data is recorded from these systems and used as baseline data. u. Appendix A, Paragraph 5.s - Although there will be no startup test procedure designated hotwell level control,. operation of the hotwell level control system will be verified using station operating procedures and monitoring hotwell level during Phase III startup testing. v. Appendix A, Paragraph 5.dd - Compliance with Regulatory Guide 1.68.2, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants, is addressed in Section 1.8.1.68.2. g.1. Avrejosy A, PMASEAPH 4.p - HAW DG4M EYSEm R6461-V4,t/6 TGCrn@ MILL 1%. TERRM!Heb AT A TEse (s>EL 'M ( ~ c% 4o0 20% Ct-RATSD THGQMAL Ta&ER Id C&ER TO Hb-t/ICR ADCDtM CckJTPOL C4-GVGT61Y1 hBESSUPG. i.u-4Z Amendment ~3

HCGS FSAR 8/84 B7. Drywell Pressure (HC Category 1; RG Category 1) Position: Implemented. See A6, B9, C8, C10, and D4. B8. Drywell Sump Level (HC Category 3; RG Category 1) Position: Implemented as Category 3. See C6 and Issue 4, Section 1.8.1.97.4.4. d. Maintaining Containment Integrity j B9. Primary Containment Pressure (HC Category 1; l. RG Category 1) Position: Implemented. See A6, B7, C8, C10, and D4. B10. Primary Containment Isolation Valve Position (excluding check valves) (HC Category 1; RG '(reu DicTrA 6 2 4 2)l Category 1) = Position: Impl ement ed/. Redundant indication is x not required on each redundant isolation valve. 1.8.1.97.3.3 Type C Variables l a. Fuel Cladding l C1. Radioactivity Concentration or Radiation Level in i Circulating Primary Coolant (RG Category 1) Position: Not implemented. See Issue 5, Section 1.8.1.97.4.5. l C2. Analysis of Primary Coolant (gamma spectrum) (HC Category 3; RG Category 3) Position: Implemented. d C3. BWR Core Thermocouples (RG Category 1) Position: Not implemented. See B4, B5, and SLI-8121 (December, 1981) (Appendix A to Reference a.8-4). P 1.8-69 Amendment 7

4 HCGS FSAR 9/85 l under maximum differential pressures, steam laden atmospheres, high temperature, high humidity, and radiation. See Section 3.11 for further discussion of environmental qualification and Section 3.9.3 for further discussion of the operability of active components. Generally, the containment isolation system is redundant and physically separated in its electrical and mechanical design, with diversity in parameters sensed for the initiation of containment isolation. Power for the actuation of the two isolation valves in a line is supplied by two redundant, independent power sources without crossties. In general, depending upon the system under consideration, the outboard and I inboard containment isolation valves are powered and controlled by different electrical channels, with the supply source being Class IE ac for both channels. See Chapter 7 for further discussion of the control and instrumentation of the containment isolation system and Section 8.3 for a further discussion of onsite power systems. oR TNCA W#NED The containment isolation system is designed with provisions for administrative control, to ensure that the proper position of all nonpoweredfisolation valves is maintained. All power-operated primary containment isolation valves /have position indicators in i tne main control room. Discussion of instrumentation and yyjH controls for the isolation valves is included in Chapter 7. 'M ! M ItA Dt2 \\ The design of the primary containment isolation system gives consideration to the possible adverse dynamic effects, such as water hammer, nudden isolation valve closure under normal operation, and to thermal expansion in those portions of pipe J between the containment isolation valves. 7 The containment isolation system is designed so that failure of motive power is in the direction of greater safety. Motor-i operated isolation valves remain in their last position upon l failure of electrical power to the motor operator. Air-operated containment isolation valves are spring-loaded to close upon loss of air or electrical power to the pilot-operated solenoid valve. Solenoid-operated isolation valves fail closed upon a loss of electrical power to the solenoid. The 1.68 psig containment pressure setpoint that initiates containment isolation for nonessential penetrations is the minimum compatible with an acceptable plant availability for r power production. ( 6.2-42 Amendment 12

HCGS FSAR 4/84 f 7.5.1.3.7.6 Safety Evaluation l The LPMS is intended to be used for information purposes only and is not a safety-related system. It conforms to Regulatory Guide 1.133. The plant operators use the LPMS to assist in both the detection of anomalous loose parts and determination of their location. The operators do not rely solely on this system or information provided in this system for the performance of any safety-related action. The LPMS equipment selection and installation minimizes the plant personnel exposure to radiation during maintenance, calibration cnd testing consistent with requirements of 10 CFR 20 as discussed in Chapter 12. The NRC will receive: l a. A submittal of the power operations LPMS alert level within 90 days following completion of startup testing. b. A submittal of the powel operations LPMS alert level, if it is changed permanently, c. Notification, if the presence of a loose part is )h' confirmed, in accordance with/the guidel:nec icr r ep o r t :b l e c c c u r r enc-es-th t-ca4-1-4c r p r ompt inctif ica t ion wi t

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~ l.133, " 6:;6-her DEttcTt*J Dic6em Ece Tue 7.5.2 ANALYSIS kwy Sy:wn @ Lar-k/MGR Caxtb EsAc7cm." 7.5.2.1 Performance of Manual Safety Functions Manual safety functions are initially based upon primary information provided by Type A variables as defined by Regulatory Guide 1.97. Safe rod patterns are established by the control room operator from control rod position indications described in Section 7.7.1.1.3. Type A variables and control rod position indications are included in Table 7.5-1. \\ 7.5-22 Amendment 5

i HCGS FSAR Figure 7.7-6. At various control rod patterns, reactor power can be automatically controlled by controlling recirculation flow. This control is effective above approximately 65% of full power for the rod pattern. l An increase in recirculation flow causes the reactor power level to increase. When recirculation flow is reduced the power level _)* ' 1 is reduced. IFigurc ' t illectrote: her the RFCS cperatec i-n conjunction with the turbin; controls for automotic lead I llcuinc. [k Iceer letel exictc cr flew control copobblity, belcw whbehI nutcratic centrcl by flow is not permitted./ If the feedwater flow is below the value that provides minimum required recirculation pump NPSH, recirculation pump speed is automatically limited. The RFCS includes the following: a. RFCS pump drive motor control b. RFCS variable frequency motor-generator (VFMG) set c. RFCS speed control components. l 7.7.1.2.2.1 RFCS Pump Drive Motor Control I Each recirculation pump motor has its own motor-generator set for a power supply. A variable-speed converter is provided between ~ the motor-generator set motor and generator. To change the speed of the reactor recirculation pump, the variable-speed converter varies the generator speed, which changes the frequency and magnitude of the voltage supplied to the pump motor to give the desired pump speed. The RFCS uses a demand signal from either ,I the operator orJthe main plant turbine generater speed aeverMet t (=ech2nin The demand cienal in curelied to/the master controller. A signal from the master controller adjusts the speed setting of the speed controller for each motor-generator set converter. The master controller signal is compared with the actual speed of the generator by the speed controller. The speed controller signal causes adjustment of the speed converter, I resulting in a change of the generator speed until the feedback from the generator equals the master controller signal. The l 7.7-23

HCGS FSAR o reactor power tSange resulting from the change in recirculation flow causes the initial pressure regulator to reposition the ,7,,, turbine control valves. UI the origina ect:nc signa; :c ; j 'turbinc Icad/cpced erree--e-ignal, t4te-turbine responds -te the change in rce tor pcuer level by Odjust4ne-the turMne-<ontee+ t valver until the 10 d/ pecd errer ciens!

rede;;d te-eace.

J l ) 1 7.7.1.2.2.2 RFCS Variable Frequency Motor-Generator (VFMG) Set i The two motor-generator sets and their controls are identical. The motor-generator set can continuously supply power to the pump motor at any speed between approximately 19% and 96% of the drive i motor speed. The motor-generator set is capable of starting the pump and accelerating it from standstill to the desired operating l speed when the pump motor thrust bearing is fully loaded by reactor pressure actino on the pump shaft. The main components of the motor-generator set are a drive motor, a generator, and a i variable-speed converter with an actuation device to adjust the i converter output speea. a. Drive motor - The drive motor is an ac induction motor that drives the input shaft of the variable speed ( converter. The motor can operate continuously at any l speed within the power supply operating frequency and t i voltage range. l I b. Generator - The variable frequency generator is driven by the output shaft of the variable-speed converter. During normal operation, generator excitation is 5 provided by the output of the generator. The excitation of the generator is provided from an auxiliary source during pump startup. 1 t c. Variable-speed converter and actuation device - The variable-speed converter transfers power from the drive motor to the generator. The variable-speed converter actuator automatically adjusts the slip between the converter input shaft and output shaft as a function of 6 i the signal from the speed controller. If the speed controller signal is lost, the actuator causes the speed converter slip to remain as is. Manual reset of l the actuation device is required to return the speed converter to normal operation. 4 s. 2 7.7-24 i

1 l HCGS FSAR l i 7.7.1.2.2.3 RFCS Speed Control Components I The speed control system shown on Figure 7.7-6 controls the variable-speed converters of both motor-generator sets. The motor-generator sets can be manually controlled either )H' individually or jointly. /.^.u t c m e t i c c o n t r o l in accordinc tc thel lturbine contic! mechanism Icad/crced errer signal d The master controller and the speed demand limiter are common to the control of both motor-generator sets. The signal from these two components is fed to two separate sets of control system components, one set for each motor-generator set. The control system components for each motor-generator set are a manual automatic transfer station, a speed controller, a signal failure alarm, a startup signal generator, and two speed limiters. 7.7.1.2.2.3.1 RFCS Master Controller ( i The master controller is a manual / automatic ccatroller thatl v provides a signal to automatically crl manually control both / recircul; tion pumps.r ith an interlock to the init4a4-pr+ssur-e t egulator. During automatic load following, the turbine-ge rator control mechanism supplies a load / speed error sigpa that epresents the mismatch between the steam being supplied to the tur ne-generator and the steam required by the turbfne-generator o maintain constant speed. This load / speed' error signal is su olied to the master controller. Duripq' automatic operation, the aster controller transmits an output signal to the speed contro ersofeachmotor-generator /.The speed controllers adjust .eir generator speed, hynce pump speed, according to the load eed error signal equirement. A pressure setpoint adjustment sign from the tur ne controller goes through the interlock swit on the ster controller, during automatic operation only, to e ip tial pressure regulator. This signal allows an immediate 6sponse by the turbine-generator l to the changed load demand. ca'Re of the pressure setpoint l adjustment, the turbine-geny ator shqm requirements can be met l during the time required)4r a new powt( level to be established by the change in recir M ation flow. Thth ressure setpoint change is effective d1y while a new reacto ower level is being established. As an exampl_ of the use of the pressure regulator o point adjustmen, suppose an increase in plant load is demand by an increar-turbine-generator requirement. The turbine-gent tor l dema s more steam to maintain a constant speed. The turbin i cofmroller allows the initial pressure regulator to adjust to l _yower nuclear system pressure. This temporary decrease in 7.7-25

HCGS ?SAR ( pre diately lets the turbine control valves o the turbine-gener demand while cu ation flow change is being made. Whe lation flow change effects the increase power level, the pr ulator again e nuclear system pressure at its original se 7.7.1.2.2.3.2 RFCS Speed Demand Limiter The speed demand limiter is an adjustable high/ low dual limiter module. It provides a limit on the maximum and minimum motor-generator set speeds that can be demanded by the master controller. Normally, the master controller signal is within the speed demand limiter's limits, and the signal passes through the speed demand limiter to the manual / automatic transfer stations for each motor-generator set. I 7.7.1.2.2.3.3 RFCS Manual / Automatic Transfer Station l The manual / automatic transfer station, one for each motor-generator set, is manually controlled with a transfer switch. While the motor-generator set is being regulated by the master controller, the transfer switch is positioned so that the manual i controller is bypassed, and the master controller signal goes through the manual / automatic transfer station to the speed controller. During startup, the master controller signal is blocked by the transfer switch and the output signal is generated and controlled by the manual / automatic transfer station through adjustment by the operator. 7.7.1.2.2.3.4 RFCS Speed Controller 1 ~ The speed controller, one for each motor-generator set, transmits the signal that adjusts the motor-generator set variable-speed converter. The speed controller compares its setpoint signal to the feedback signal from the motor-generator set tachometer and adjusts its output to the speed :onverter so that the feedback signal from the tachometer equals the setpoint signal. The speed controller setpoint signal is received during automatic operation from the master controller, during individual motor-generator set manual operation from the manual / automatic transfer station, during pump startup from the startup signal generator, and during low feedwater flow from the speed demand limiter. I (- 7.7-26

HCGS FSAR automatic mode. Each level controller contains setpoint deviation meters, an output indicator, a manual output control, manual automatic switching capability, and a manually operated setpoint adjustment: Startup automatic level control - In the startup level a. control mode, measured level is compared to level setpoint with the controller. The resulting signal is conditioned by the proportional plus integral controller circuits and transmitted to the startup level control valve. b. Normal mode automatic level contro.' - The optimum reactor vessel water level is automatically determined by programming the water level according to steam flow. To perform this function, the total steam flow signal adjusts the ratic level controller netpoint.to the optimum reactor vessel water level. The steam programming low load limiter establishes the minimum steam flow signal that can be delivered to the ratio level controller. This is done so that a setpoint cannot be established below the steam separator optimum range. During normal operation (cincle-elerent ed three-element automatic control is orovided for operator use. J W I R-- /r i nc l e-el eren t ccurce ic the re:cter ester level. / The three-element source includes measurements of steam flow, feedwater flow, and reactor water level. The three-element control signal is obtained as follows: The total steam flow signal and the total feedwater flow signal are inputs to a proportional amplifier. The output of this amplifier reflects the mismatch between its input signals and is designated as the steam flow /feedwater flow error signal. If steam flow is greater than fecdwater flow, the amplifier output is increased from its normal value when steam and feedwater fli>ws are equal. The reverne is alno true. This amplifier output signal is trancmitted to a second proportional amplifier that also receives the reactor vessel water level signal. The addition of the reactor vessel water level signal to the steam flow /feedwater flow error signal results in the three-element control signal; which is the input to the ratio level controller. 7.7-33

OAD/ SPEED I ERROR SIGNAL TURBINE CON-i TROL MECH. ADMISSION VALVE + PRESSURE PRES. 1 SET u TO OTHER y POINT INITIAL CHANGE CONTROLLER i + REGULATOR /. MANUAL I I l s. l l' / TURBINE M BYPASS VALVES V REACTOR

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500 KV TRANSMISSION LINES A ~ s n 5 )[ Al KEENEY EEDOM P A SECT 10X h_ 500 4 KV j j TRANSFORMER (4) A T1 T3 N N.C. ' N.O. i N.C. 61x 51x _ N. + U, 13.8 KV \\ SV.lTCHYARD {} N.O. N.C. SLP2 b' SLP1 60x Mx {FMR ( XFMR t I N.C. l N.C. \\l 5@ KV SWITCHYARD L~! 52x '/ N.C. i N.O. N.C. O T4 T2 3<y /~Y w rvn U \\\\\\\\ //// SECT 20m H ww ag w MM 24-500 V Y Y " \\ ^ YY Y Y MAIN TRANSF. / V 13.8 KV FEEDS TO STATION SERVICE TRANSFORMERS N HOPE CREEK GENER ATING ST ATION FINAL SAFETY ANALYSIS REPORT ONE.LINE DIAGRAM FIGURE 8.2 2 AMENDMENT 9.01/85

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w We. _ 's SINGLE LINE METER & J,l,I,U.l.l RELAY DI AGRAM 4.16 KV ~ CLASS 1E POWER SYSTEM ' r.o.w.a FIGURE 8.3 5 SHEET 2 OF 2 Amendment 12,0945 5 1 4 1 3 l 2

HCGS FSAR 10/84 9.1.4.2.3 Fuel Servicing Equipment The fuel servicing equipment described below has been designed in eccordance with the criteria listed in Table 9.1-5. 9.1.4.2.3.1 Fuel Preparation Machine The fuel preparation machine, shown on Figure 9.1-7, is mounted on the wall of the fuel storage pool. It is used for stripping reusable channels from the spent fuel and for rechanneling the new fuel. The machine is also used with the fuel inspection fixture to provide an underwater inspection capability, and with the defective fuel storage container to contain a defective fuel i assembly for stripping of the channel. The fuel preparation machine consists of a work platform, a frame, and a movable carriage. The frame and movable carriage are located below the normal water level in the fuel storage pool., thus providing a water shield for the fuel assemblies being handled. All parts remain underwater when the fuel preparation hug Aenmpf machine removes and installs channels. The carriage on the fuel / preparation machine has permanently installed mechanical uptravel / stops to prevent raisingtfuel above the safe water shield level of 8 feet below the surface. The movable carriage is operated by a foot-pedal-controlled, air-operated hoist. 9.1.4.2.3.2 New Fuel Inspection Stand The new fuel inspection stand, shown on Figure 9.1-8, serves as a support for the new fuel bundles undergoing receiving inspection end provides a working platform for technicians engaged in performing the inspection. The.new fuel inspection stand consists of a vertical guide column (fuel support. structure), a lift unit to position the work platform at any desired level, bearing seats, and upper clamps to hold the fuel bundles in position. 9.1.4.2.3.3 Channel Bolt Wrench The channel bolt wrench, shown on Figure 9.1-9, is a manually operated device approximately 12 feet (3.6 meters) in overall' 9.1-45 Amendment 8

V 9 ) TAaLE 9.3-3 0t/06 l Fege 1 of 3 l INN SAMPLE POINTS-CAPAaILITIES AND DESIGN DATA BWR Sample Points Panet

Pressure, Temperature, Root Valve Automatic Process Sampling
Number, pelg
  • F Piping Isolation A1AAA punitoring Grab Samples skaot Valwee stintaus $ ample systen Location Mas / Norm Maw / Norm Class Provielone Proviolona Performed Fors PAID 6 tocation plos pate Comments Air ejector 10C134 Double too=

Vial Radiation Radiation M26, Sh 1 Viel samplere offges Turb b1de lation on sampler (B3) eee section 11.5 staan jet air ejector y swCU reactor cool-10C251 1300/1102 150/120 RBC mone Wooded Conductlelty Insolubles, 364 4 (C2) 500 al/ min l ant t

et meactor sample Do, Silica, chloridea nel (32)

/ft/s blog station Ph $14$ tank Grab sample from standby control tank Liquid None Boron hA 31 4 See Sectlen 5.3.5 Sumps innade ctat 00C350 175/160 150/100 MBO IrV-F004 mooded saane Radioactivity met Sh 1 500 al/ min (drywell (tour dt) Aun blog 175/160 150/100 IERD NV-F003 sample Conductivity (E2), (G2) /ft/s NV-F015 station pu, Chlorideo M6f, sh 2 (drywell BV-PC20 (D2), (GI) equip drnt { spent fuel puot Aus b1dg 160/100 150/90 HCD Ilone Ikaoded Conductivity Silica and stS4-0 500 at/ min 00C350 160/100 150/90 MCD stone sample ya he avy (G4) /ft/s station elements M54-0 (54) Drywell 1AP 270 IICB Double Incated 44 Moble gases st25-1 sh 8 500 al/ min atacepheric 130 meactor isol reactor blog padio= (F4) /ft/s See section 11.5 blog valves remote nuclide readout spectre Offgas treatment 00C323 Badiation teone am sampling capab411tys system see Section 11.5 Condensate polisher 10C150 220/150 135/124 MBD mone elooded gdi, con-sestallic 500 al/ min l ayates inlet and Turbine sample ductietty tapurittee /ft/s outlet building station

sane, radioactiv-etlica, DO ity caloride R10CU BGnCTM ICCDSI I420lILI 85Cll2O GBC W l*CCLGD OmtcuCnotty {1[DULQlM, M94(C2)

Q)O mf{lnM CCCXAAJT tsll5T R6nC1 tit $ntr1RG DO,54CA, CHlbR:bES /Vf $ BtARDING 3rATickJ pH Amendment 14 T1002729V

HCGS FSAR 11/85 of dampers when high pressure or temperature develops inside the room. The failure of an isolation damper in a closed position results in both a loss of ventilation for the equipment compartment affected, and a high room temperature alarm. Each trouble alarm on the local control panel is alarmed in the main control room as a panel group alarm. Indicating lights on the local panel identify the failed damper, which can be manually reset to the open position. 9.4.2.3 Safety Evaluation The safety design criteria applicable to the safety-related systems are discussed in the following sections: a. Wind and tornado protection - Section 3.3 b. Flood design - Section 3.4 c. Missile protection - Section 3.5 d. Protection against dynamic effects associated with the postulated rupture of piping - Section 3.6 e. Environmental design conditions - Section 3.11. Redundant radi,ation monitors are provided in the exhaust duct of the refueling area and the exhaust. duct of the reactor building. A high radiation signal from any monitor or a LOCA signal automatically isolates the reactor building ventilation penetrations and energizes the FRVS. The exhaust air transit time between the refueling area monitors and the RBVS exhaust system isolation dampers is greater than the combined time of damper closure and the monitor response. The exhaust air transit time is 12 seconds, and the combined monitor response and valve closing time is withinj seconds. The' isolation dampers used for reactor building ventilation penetrations isolation are redundant (two in series powered from 9.4-27 Amendment 13

iKL8 FSAR TABLE 9.4-9 AUI!LIARY SUILDIMG SERVICE AREA EVAC SYSTWI FAA Oesistry Labo-Sere 1ce Area Segelce Area retory Exhaust Itee Supply (SAF) System Exhaust (SAE) Fane (CLE) System Steam Unit Beaters Type Air-handling Fan Fan and filter Steam coil with Steam coil with (blow-through) housing fan fan pahea Ne ber of unite 2 2 2 7 3 Number operating 2 2 1 g Flow rate, each, cfm 20,510 2.9,390 8,S3o Fw Type Cent r if ug al Centrifugal Centrifugal Propeller Propeller Curatity per unit 1 1 1 1 1 se>u r/hp/dr ive 75/ belt 40/ belt 20/ belt 1/4/ direct 1/4/ direct ststic pressure, in. v.g. 8.3 3.5 9.5 NA NA setting coil Type Steam Steam Steam Elect Kumber of colle per unit 1 None None 1 1 Ccpacity, each, atu/h 1,403,500 100,000 25,000 Bumidifier Type Pan (electric co111 Quantity per unit 1 None None None None th/h water evaporation 287 Colling coil Type Chilled water Na ber of coils per unit 1 Hone None None None capacity, each. Sto/h 1,971,000 F11ttre Type Roll type, low tow efficiency effsciency/high BZPA eff1ciency Preocure drop, in. v.g. 0.1-0.5/0.5-l.0 0.2/1.0/1.0-3.0 (cit an-d ir t III I III 85t(2380-45tIII 45-404I2*/99.97tI33 Efficiency (1) ASHRAE Standard 52 76 weight arrestance. (2) ASNRAE Standard 52-76 dust spot. (3) By military standard 242 DCP test method on 0.30-micron particles. T1332934 l

i TRBLG Q.Ll - R RbR. IOlb EM A W OOLY 01/86 NAC SYSTDI PARAMETERS Electric Duct Reheat Colts CLE Booster Fan ,i th y Seheat coil asheat coil asheat coil Reheat coil Beheat coil Seheat coil

  • Fan 3

3 2 4 1 2 1 I 1600 - 1780 210 - 2760 2730 - 3720 800 - 1430 540 630 - 880 900 Centres Mone Mone Bkane None sk>ne None 1 3/4/ Belt 1.5 Electric-fin tube Electric-fin tube Electric-fin tube Electric-fin tube Electric-fin tube Electric-fin tube 1 1 1 1 1 1 24,480 30,720 34,130 15,360 5120 10,240 tione lione stone Mone sk>ne Mone None mone skane Isone neone None Nigh Ef fielemey = = 0.35/1.0 ) 90 - 95123 l l Amendment 14 1 l

FIG,ute 95 -%, sneer L cv a (2) MAIN AIR START VALVES 12b (2) AIR START DISTRIBUTORS \\' .\\. \\\\ ENGINE l [

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-~ l38t J l Sb 3 '8 **T 12a M l r 4a AIR FILTERS l Sa 7 3-WAY RELAY VALVE - ~l j AIR SOU E SELECTOR

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  • ~~I IN LOC AL & RE MOTE

^ 1j 3 8"T E NG. CON T HOL PANE LS (BM) fjy N.C. R ERE 30a e 20 MESH [ BE OR N I .016 WIRE DI A. Q (DISCH. PRESS) V I EW A** N.C.. .a. i A RA BOOST SOURCE SELECHON AND 3.1 1-(4) 2" - 600 lb. RF FLANGES 3 SHUTDOWN SAFETY INTERLOCK RELAY VALVE l DRAIN CONNECTION A3 I 2 3/4 S. W. -.-.13. N. ~, I-. k h l $ A SME SECTIONID,CL.3 EQUIPMENT 19b j MANUAL DRAIN (TYP) 11872210 l HOPE CREEK GENER ATING STATION FINAL SAFETY ANALYSIS REPORT STARTING & CONTROL AIR SYSTEM FIGURE 9.5-26 SHEET 10F 2

HCGS FSAR 11/85 l Shutdown cooling suction valve BC-HV-F009 is powered from Divisio I. In the event that control of this valve is affected by the fire, then an alternate shutdown path can be used. The alternate shutdown cooling mode in lieu of RHR B shutdown cooling mode is as follows: Core spray or LPCI D mode of RHR is used to fill up the reactor until the steam lines are flooded. With one or more SRVs open, water flows out the relief valve and back to the suppression pool. RHR B suppression pool cooling mode cools the suppression pool as before. 9A.5.3 Remote Shutdown Method The remote shutdown method can be found in the FSAR Appendix 9A under response to NRC Generic Letter 81-12, Item 1.e. In addition, depressurization by use of 3 SRVs and use of the B LPCI is available from the RSP. 9A.S.4 Sourious Sianal Analysis Results A complete review of spurious signals per section 9A.1.5.f was performed. The circuits requiring separation were identified and /

fixes, i.e.,

cable rerouting, fire walls, etc., have been included in the plant design. No cable tray fire wrapping is required and only one area of conduit fire wrap is required. No wire cutting or fuse pulling operations are required. The alternate shutdown mode was used in the reactor building and electrical access areas due to the logic and electrical channels associated with the RHR shutdown ccoling valves. These valves are BC-HV-F008, BC-HV-F009, BC-HV-F015A and BC-HV-F015B. In addition, due to the large number of trips associated with the RCIC and HPCI, manual depressurization and low pressure injection systems were relied on. Because of the good cable and equipment separation at HCGS and the above shutdown methods, no manual octions are required to achieve hot or cold shutdown. These shutdown methods have been included in the plant operating procedures. However, it may be desirable to manually establish RHR shutdown cooling or one of the high pressure injection systems to avoid the normally less desirable alternate shutdown path and/or fast depressurization to use LPCI or CS. EEE igh/ low pressure valve interface problems were identified as discussed in the response to Generic Letter 81-12, Item 2. ~ 20ucation za.1 e l' l 9A-32 Amendment 13 l s J

HCGS FSAR 01/86 l ) 2a. Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.

Response

The high-low pressure interfaces which use redundant electrically ~ controlled devices to isolate the primary coolant boundaries are as follows: 1. Reactor Vessel Vent Valves BB-HV-F001 BB-HV-F002 2. RHR Suction Valves BC-HV-F008 BC-HV-F009 ? 3. Reactor Water Cleanup;suctieniValves .\\ BG-HV-F001 b4 M I BG-HV-F004 = 2b., For each set of redundant valves identified in a., verify the redundant cabling (power and control) have adequate physical separction as required by Section III.G.2 of Appendix R.

Response

The required Appendix R, Section III.G.2, separation of the ~ cabling for the above listed sets of valves cannot be provided. The control wiring for each set of valves is separated using modified Regulatory Guide 1.75 separation in the remote shutdown panel, 10C399 and also in the main control room panel &' C050.i T 2c. For each case where adequate separation is not provided, show that fire-induced failures (hot shorts, open circuits, or shorts to ground) of the cables will not cause maloperation and result in a LO'A. C k 9A-94 Amendment 14 f

~ INSERT FOR PAGE 9A-94 BG-HV-F034 BG-HV-F035 4. LPCI Injection / Bypass Valves BC-HV-F017A,B,C,D BC-HV-F146A,B,C,D 5. RHR Return / Bypass Valves BC-HV-F015A,B BC-HV-F122A,B 6. Core Spray Supply / Bypass Valves BE-HV-F005A,B BE-HV-F039A,B i i d i + i 1 c

HCGS FSAR 01/86 l aa.a' foR Rr Da. 4 N do. S ',

Response

gg J In Escn A6 _j mig. Fire-induced f ailures of the cables for the a-eur-1 valves listed in[ item 2a.1,hanc 2. : will not cause maloperation and result in a jlLOCA since t_he valves are normally closed andItheir motor control!'MCC)) power

  1. center fty rackinc cbt the Orc xcrs.;9The RWCUtvalves4are normally open lm(

and their cabling outside the control room, cable spread rooms 's and control equipment room have been verified to be separated by M-HV-DI^* III.G.2.aseparation.f lluGERTl ra>9 Therefore, a fire in the zones where the physical separation is less than Section III.G.2 requires will not cause maloperation and result in a LOCA. See FSAR Section 7.4.1.4. d 9A-95 Amendment 14 \\

INSERT FOR PAGE 9A-95 Therefore, for fires outside these areas, at least one valve will be available to isolate the RWCU system. For a control / diesel area fire which might affect all 4 valves, spurious operation is prevented by removing power to the F034 and F035 valves during normal operation. There are other valves associated with the filter /demineralizer portion of the system which are not identified in Item 2a.3 above because they are located, powered and controlled in the reactor building and the separation of BG-HV-F001 and F004 is adequate to assure isolation from the main control room as discussed above (i.e., F001 and F004 are considered to be the high-low pressure interface for these valves). l l N%]4 l 1 i

~. HCGS FSAR 01/86 air drawn out of the condenser is discharged tc the south plant vent. The mechanical vacuum pumps and their suction valves are actuated remotely from the main control room. If high radiation is detected in the main steam lines (detectors are located in the main steam tunnel between the outboard main steam isolation valves and the main steam stop valves) the pumps are tripped, and the suction valves automatically close. If the seal water flow drops below acceptable limits, the vacuum and seal water pumps are tripped and a low-flow pump trouble alarm actuates in the main control room. A water separator removes any water droplets from the noncondensable gases before discharging the gases to the south plant vent. A seal water pump removes water from the r separator and cycles the water through the seal water cooler and back to the vacuum pump. When working in parallel, the vacuum pumps are designed to evacuate the main condenser from atmospheric pressure to 5 inches of mercury absolute in 120 minutes. 10.4.2.2.2 Steam Jet Air Ejectors After condenser vacuum is established by the mechanical vacuum pumps, and the air inleakage is not greater than 75 scfm, one SJAE is placed in service to maintain the vacuum. The mechanical vacuum pumps are shut down. The mechanical vacuum pumps cannot be run during plant operation due to the radioactive gases that accumulate in the main condenser. The SJAF train is a full-capacity three-stage unit, including three 33%-capacity first-stage ejectors, an intercondencer, one 100%-capacity second-stage ejector, an af tercondenser, and one 100%-capacity third-stage ejector. A redundant SJAE train is provided to maintain condenser vacuum if the first train is not available. The three first-stage ejectors continuously remove G noncondensable gases and entrained steam and discharge them to @fj[f) the SJAE intercondenser. The intercondenser condenses the uma ejector motive steam and the carryover steam. The condensate is 05MM38 returned to the main condenser. The second-stage ejector draws the noncondensable gases and entrained steam from the intercondenser and discharaes them to the aftercondenser. The aF condensate from the aftercondensebddrains to the condencate drain! 'tanN J A third-stage ejector is provided to boost the discharge )' pressure to 11 psig maximum before discharging the noncondensable gases and third-stage motive steam to the gaseous radwaste system. 10.4-7 Amendment 14

HCGS FSAR 10/84 corresponding noble gas release rate of 500,000 uCi/s after 30 minutes decay (design basis). The concentration of radioactivity at the point of discharge shall not exceed concentration limits specified in 10 CFR 20, on an annual average basis. g. All piping and equipment in the LWMS are non-Seismic Category I with the exception of the primary containment. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the LWMS are discussed in Section 3.2. h. Design features that reduce maintenance, equipment downtime, liquid leakage, or gaseous releases of radioactive materials to the building atmosphere, to facilitate cleaning or otherwise improve radwaste operations, are discussed in Section 12.3. i. All atmospheric liquid radwaste tanks are provided with an overflow connection at least the size of the largest inlet connection. The overflow is connected below the tank vent and above the high level alarm setpoint. It is routed to the nearest drainage system compatible J-- with its purity and chemical content. IE:ct liquid fraducate tark rcer in decigned te centair the maximu= l licuid ir.ver.tcr'/ ir care the tcrk rupturOC. j. Processed wastesfare collected in sample tanks prior to their reuse asAc'cncencate: quality water or discharged IDEHMEhAMEEDI in a controlled. manner into the cooling tower blowdown line for dilution before entering the Delaware River. k. The expected and maximum radionuclide activity inventories for LWMS components containing significant amounts of radioactive liquids are shown in Tables 11.2-8 and 11.2-9. They are based upon the assumptions given in Table 11.2-1 and upon the following: l r i ' l l 11.2-3 Amendment 8 l l

i l l 5 !! CGS FSAR 11/85 TABLE 11.2-10 (cont) Page 2 of 5 l Rated Design Rated Flow Ilead TDl! Rated Power Pressure / Temp Compenent Quantity Type (gpm) (ft) (hp) (psig/ F) Code Pumps Waste collector pumps 2 IIor, cent 182 275 40 240/210 Mfg Std Waste surge pump 1 Hor, cent 182 275 40 240/210 Mfg Std Floor drain collector 2 IIor, cent 176 260 40 240/210 Mfg Std pumps Waste neutralizer pumps 2 flor, cent 176 55 5 170/210 Mfg Std Concentrator feed pumps 2 flor, cent 40 100 5 150/210 Mfg Std Concentrated waste pump 2 llor, cent 40 215 15 170/210 Mfg Std l Chemical waste pump 1 Hor, cent 176 150 20 170/210 Mfg Std N l Decontamination cone 1 IIo r, cent gq /M 7.5 150/210 Mfg Std waste pump Detergent drain pumps 2 Hor, cent 25 175 7.5 150/210 Mfg Std Waste sample pumps 2 Itor, cent 75 200/210 Mfg Std Floor drain sample 2 llor, cent 176 310 30 240/210 Mfg Std pumps Waste evaporator dist 2 flor, cent 50 66.7 max 5 170/210 Mfg Std l transfer pumps Waste evaporator cone 2 !!or, cent 50 11.6 1 150/230 Mfg Std waste transfer pumps Waste evaporator 2 llor, cent 50 15.4 max 1 150/210 Mfg Std l recycle pumps l Decontamination solution 1 llor, cent h M 1 150/230 Mfg Std f evap cone waste transfer pump i I Decontamination solution 1 Ifor, cent-50 13.8 cax 1 150/230 Mfg Std l cvap recycle pump waste evaporator 2 Ifor, cent 200 120 15 150/210 Mfg Std .s condensate return pumps Amendment 13 l i

HCGS FSAR 01/86 l TABLE 11.2-14 Page 1 of 2 LIVE VOLUME CAPACITIES OF TANKS AND DESIGN OF PROCESS EQUIPMENT CONSIDERED IN THE CALCULATION OF HOLDUP TIMES Rated Flow Live Volume Rate (gpm) Item Number Capacity, oal. 8 TDH(feet) Tanks Waste collector tank 2 29,000 Floor drain collector tank 2 15,000 / Waste surge tank 1 54,000 Waste neutralizer tank 2 25,000 Concentrated waste tank 2 10,550 i Decontamination solution 1 600 concentrated waste tank Chemical waste tank 1 4,000 Waste and floor drain sample 4 15,550 tanks Waste evaporator distillate 2 1,100 storage tank Decontamination solution 1 30 evaporator condensate return tank Pumps Waste collector tank.pamp 2 182 @ 275 Waste surge tank pump 1 182 8 275 Floor drain collector tank 2 176 8 260 pump Concentrator feed pump 2 40 8 100 Concentrated waste pump 2 40 a 215 fy9 g ggj Chemical waste pump 1 176 a 150 Decontamination concentrated 1 '50 0 140?+-- / waste pump ISE@ $ ' Waste sample tank pump 2 I450 e 375; Floor drain sample tank pump 2 7'176 B 310 Concentrated waste transfer 1 L50 0

2. 0; l

l' 4 pump Waste evaporator distillate 2 50 a 66.7 /39@ 6dNI transfer pump max Waste evaporator concentrated 2 50 a 11.6 waste transfer pump Filters Waste filter 1 180 Floor drain filter 1 180 Amendment 14 l

HCGS FSAR 1/86 l TABLE 12.3-2 POST-ACCIDENT SHIELDING SOURCE TERMSC1) Core Inventory Systems Containing Source Releases Dilution Volume Sources A Noble gases 100% Drywell free volume Drywell atmosphere, Halogens 25% 169,000 ft3 sampling system, Solids 0% H, recombiner H,-0, analyzer B Noble gases 100% Drywell plus torus Torus atmosphere Halogens 25% free volume Solids 0% 310,000 ft3 C Noble gases 100% Reactor coolant Sampling system Halogens 50% plus piping Solids 1% 13,000 ft3 D Noble gases 0% Torus plus Core spray, HPCI & Halogens 50% reactor coolant RCIC water side, Solids 1% 132,000 ft3 RHR shutdown cooling mode, suppression pool cooling & containment spray modes E Noble gases 100% Reactor steam HPCI and RCIC Halogens 25% 9500 ft3 steam side, MSIV Solids 0% sealing systems F Containment learage rate is modelled FRVS recirculation l as 1% per day. FRVS recirculation and ventilation flow is 1.8 reactor building volumes charcoal filters, per hour,.and FRVS ventilation rate reactor building vis k minimum cfl100% of reactor atmosphere / building volume per day. The reactor building free volume is 1.9x106 fta and the refueling floor free volume is 1.4x10* ft3 (1)The initial inventory for the radioisotopes of these sources are given in Tables 12.2-134 through 12.2-136. I Amendment 14 l l

HCGS FSAR 1/86 l TABLE 12.3-3 Page 1 of 2 l

SUMMARY

OF POST-ACCIDENT EXPOSURES TO VITAL AND USEFUL AREAS l Total Dose frem) l Gamma Beta FSAR l Whole Body (3) Skin (3) Thyroid (3) Figure No. Vital Areas Requiring Continuous Occupanev(2) ? ? Y b3h2 control room (*)

c. cr 2

j:_sr:e _2.zr " 12.3-41 .( Technical support center (TSC) 2.9E+0 1.6E+0 3.0E-3 12.3-33,34 Operations support center (OSC) 6.2E-2 1.6E+0 4.3E-1 12.3-41 Guard house 6.3E-1 5.6E-1 1.5E-1 On site vital Areas Requiring l Infrecuent Access t Post-accident sample stationts) 2.1E+0 7.5E-2 2.4E-2 12.3-44 (30 min for taking sample) a Post-accident sample transport 1.2E+0 5.0E-2 1.6E-2 12.3-44 to Pathcs) (20 min for transport-12.3-48 ing sample) Controlled hot chemical lab (s) 3.6E-2 7.5E-2 2.4E-2 12.3-48 I (30 min for analyzing sample) FRVSV RMS post-accident skid 6.1E-1 7.5E-2 2.4E-2 12.3-50 (30 min for taking sample) FRVSV sample transport path 1.3E-2 7.5E-2 2.4E-2 12.3-48 to (30 min for transp9rting sample) 12.3-50 Controlled hot chemical lab 2.4E-2 7.5E-2 2.4E-2 12.3-48 (30 min for analyzing sample) ll Diesel generator and accessories 1.7E-2 1.9E-1 6.2E-2 12.3-39, 40 (1 hour occupancy) Useful Areas (2) l HP/ access control areas 1.1E-2 1.5E-1 4.8E-2 12.3-48, 49 l Emergency assembly points OSC 3.1E-3 5.2E-2 1.7E-2 12.3-41 3 TSC 5.6E-3 4.8E-2 8.6E-5 12.3-33, 34 l control access point 1.1E-2 1.5E-1 4.8E-2 12.3-49 (room 3533) Amendment 14 l

HCGS FSAR 1/86 l TABLE 12.3-3 (cont ' d) Page 2 of 2 l Total Dose (rem) l Gamma Beta FSAR Whole Body (3) Skin (3) Thyroid (3) Figure No. Cafetaria (root 109) 4.1E-2 2.4E-2 7.9E-3 12.3-57A General purpose rocms 4.1E-2 2.4E-2 7.9E-3 12.3-57A (training rooms 103 & 104) Eaintenance shop 4.1E-2 2.4E-2 7.9E-3 12.3-57A (Clean machine fab. shop, rocm 134) ~ Remote shutdown panel 2.2E-3 5.2E-2 1.7E-2 12.3-49 l-Werst yard area 3.8E+0 2.2E+0 7.2E-1 on site l 1 (28 Doses include SRP 6.4 occupancy factors for 30 days.

82) Based on 1-hour occupancy and maximum post-accident dose rate.

Whole body dose includes direct shine and cloud immersion contributions. '3) Beta skin dose is due to immersion, thyroid dose is due to inhalation. 4*) Design basis control room doses;"?re !? er I r c:9 FSAR Section 6.4. For control room cess evaluation, the doses and adose rates for the technical A {g a s ) support center should be usef The estimated whole body doses are taken from FSAR Section 9.3.2.3.2.f, Amendrent 8. The extremity doses for taking, transporting, and an&lyzing j sample are 3.3 rem, 0.07 rem, and 0.1 rem respectively (ibid). I {ARG S/ M b f)) i \\ Amendrent 14 l

.^ 13 CGS FSAR 1/06 l TABLE 12.3-3a Page 1 ot 3 POST-ACCIDENT DOSE FATES TO VITAL AND OTHEP AREAd Dose Pate (t em/h) l l Betatt8 Pime _ dhole Bodyi,5 8 Skin Thyroidtt8 l Vital Areas Pequiring Continuous occuraney. ControlrooM( 1 hour 9.6E-4 1.5E-2 7.7E-5 l 1 day 7.9E-4 2.4E-2 4.0E-5 l 1 week 5.4E-5 1.8E-3 4.2E-6 l 1 mont h

2. 3 E-6 2.1E-4 5.0E-7 l

Techrical support center (TSC) 1 hour 3.9E-3 1.5E-2 7.7E-5 l 1 day 9.2E-3 2.4E-2 4.0E-5 l 1 week 9.9E-3 1.8E-3 4.2E-6 l 1 month 5.3E-3 2.1E-4 5.0E-7 l Orerations supFort center (oSC) I hour 1.9E-3 3.2E-2

1. 6 E-2 l

1 day 7.8E-4 2.5E-2 4.0E-3 l ?,. I week 5.2E-5 1.8E-3 5.5E-4 l 1 mon t h 2.3E-6 2.1E-4 6.6E-5 l Guar dhois sa 1 hour 2.0E-2 1.1E-2 5.6E-3 l 1 day

7. 9 E-3 8.4E-3 1.4E-3 l

1 week 5.2E-4 6.1E-4 1.8E-4 l 1 month 2.3E=5 7.0E-5 2.2E-5 .I Vital Areas Peq1* iring inf reggent Access Post-accident sample station 1 hour

1. 8 E-1 8.9E-2 4.5E-2 1

1 day 1.6E-2 6.8E-2 1,1 R-2 l 1 week 3.5E-3 5.0E-3 1.5E-3 l 1 month 1.0E-3 5.7E-4 1.8E-4 l Controlled hot chemical lab 1 hour 7.0E-3 8.9E-2 4.5E-2 l 1 day 2.8E-3 6.8E-2 1.1E-2 l 1 week 1.9E-4 5.0E-3 1.5E-3 l 1 month 8.1 E-6 5.7E-4 1.HE-4 l Diesel gererator and 1 hour 1.1E-2 1.1E-1 5.HE-2 l 1 accessories 1 day 4.4E-3 8.BE-2

1. 4E-2 l

1 week 2.9E-4 6.4E-3 2.0E-3 l 1 month 1.3E-5 7.3E-4 2.4E-4 l ' d'mer.dment 14 l 1 n

__ ~ _ - _. ~ HCGS FSAR 1/H6 i TABLE 12.3-3a (Con't) Pa ge .1 o r .1 Do9e Rate (rem /h) l l Betatt8 Time vihole Bodyt88 S. kin _Thyroirit i 8 l ~ Femote shutjow7 pane' +1 hour 1.4E-3 3.2E-2 1.6E-2 l 1 day

5. 7 E-4 2.5E-2 4.0E-3 l

1 week 3.BE-5 1.NE-3 3.5E-4 l 1 mor.th 1.7E-6 2.1E-4 6.bE-5 l r korst yard area 1 hour 2.4 1.3 6.7E-1 l 1 day 9.5E-1 1.0

1. lE-1 l

5 1 week 6.3E-2 7.5E-2 2.JE-2 l 1 month 2.RE-3 9.5E-3

2. 7E-3 l

1 (ti Shole body dose includes direct shine and claus immersior. contributions. Beta skin dose is d,> . to irrersion, thyroid dose is due to inhalation. I ~ l (2) Tm DE;5t(:;A) M5ts CO3rglot Roofh OcGMiG Aas Eu4LuATafD 'O MR h 6.Q. M coa %kt 2 0tV) ACCESS 6U4CL4A4Ticd, M h IWA!it5 TGGG9JTiED CU 'IutS TK73tg Eton i h B6 (CED. Amendment le [ l

HCGS FSAR 01/86 .2.12.3.19 Core Power - Void Mode a. Objective The objective of this test is to measure the stabi ty of the core power void dynamic response, and to emonstrate that its behavior is within specifi d ign limits, b. Prer uisites The core ' s maintained in a steady-sta 4 condition prior to s rting this test. ,r c. Test Method The core power void oop mod, that results from a combination of the ne tron inetics and core thermal hydraulics dynamics, i 1 ast stable near the natural circulation end of the ted 100% power rod line. A fast change in the re ti ty balance is obtained by two methods: (1) pre ure r ulator step change, and (2) by moving a ver high wo h control rod one or two notches. Both lo 1 flux and otal core response will be evaluated by onitoring sele ed LPRMs during the transient. d. Acceptanc Criteria Level 1. The ransient response of any system-relate variables to any test input must not diverge. System r ated v riables are heat flux and reactor pressure. Level 2: l The decay ratio of each controlled mode of response should meet the requirements of the General Electric startup test specification. s a t. 14.2-174 Amendment 14 __t

HCGS FSAR 01/86 14.2.12.3.20 Pressure Regulator a. Objectives 1. To determine optimum pressure regulator setting to control transients induced in the reactor pressure control system. 2. To demonstrate the takeover capability of the backup pressure regulator via simulated failure of the controlling pressure regulator. 3. To demonstrate smooth pressure control transition between the turbine control valves and bypass valves. b. Prerequisites Instrumentation has been checked and calibrated. The plant is at a steady-state power level. c. Test Method The pressure setpoint is decreased rapidly and then increased rapidly by about 10 psi. The response of the system is measured in each case. The backup pressure regulator is tested by simulating failure of the operating pressure regulator. The bypass valve is tested by reducing the load limit, which requires the o bypass _ valves to open and control the bypass steam / flow. /At ccrtair tect conditienc, the cetpcint chcngc 1 lperer'rillbeperferredirconjunctierwiththe00re tect - void cde tect. j d. Acceptance Criteria Level 1: 1. The transient response of any pressure control system related variable to any test input must not diverge. i 14.2-175 Amendment 14

HCGS FSAR 01/86 b. Prerequisite The plant has been stabilized at the required power level. c. Test Method Individual main stop, control, and bypass valves are manually cycled and reset at selected power levels. l. The response of the reactor is monitored and the maximum power level conditions for the performance of f this test are determined. The rate of valve stroking and timing of the closed-open sequence are chosen to minimize the disturbance introduced, d. Acceptance Criteria Level 2: Peak heat flux, vessel pressure, and steam flow shall remain below scram or isolation trip settings by a margin consistent with the GE startup test specification. 14.2.12.3.23 Main Steam Isolation Valves a. Objectives 1. To functionally check the MSIVs at selected power ,7 levels,':nd deter.mine the

i=ur perer level they

{ can bc tcsted at individually 2. To determine isolation valves' closure times. 3. To determine reactor transient behavior during and following simultaneous closure of all MSIVs. b. Prerequisites The plant has been stabilized at the required power level. 14.2-179 Amendment 14 ~

HCGS FSAR 01/86 i c. Test Method 1. Individual closure of each MSIV is performed at p selected power levels to verify functional _/ _ performance and to determine closure times. /the is determined for individual L, man;mumpowcricvel _1crure 'zith emple marcir te cerer. 2. A test of the simultaneous full closure of all MSIVs is performed at about 100% power. Operation of the RCIC and HPCI systems and the relief valves is demonstrated. Reactor parameters are monitored r to determine transient behavior of the system during the simultaneous ful'1 closure test. The reactor will immediately scram due to the actuation of the MSIV position switches. Recirculation pumps will trip if Level 2 in the RPV is reached. The feedwater control system will prevent the RPV water level from reaching the steam lines. d. Acceptance Criteria ( Level 1: 1. MSIV closure times shall be as specified in the GE startup test specification. 2. Following the full closure of all MSIVs, vessel pressure and heat flux level shall be as specified in the GE startup test specification. 3. The reactor must immediately scram and the

  • feedwater control system must prevent the water from reaching the main steam lines following full closure of MSIVs from high power.

Level 2: 1. Peak neutron flux, vessel pressure, and steam flow shall remain below scram or isolation trip settings by a margin consistent with the General 14.2-180 Amendment 14 a

HCGS FSAR 01/86 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 l 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) E DUED shall be madeh]as carly in the startup progr: ac BCroR TWGSSUR6 --- m: -> ~:6-&

is..

c--- 2 2.. a-BendEn) @@2% C 2 Z ls...^%.^.L1 % C1 U!1'4"^^1' C1'!1'." L IT'2" l'&12

  1. N M8ME bb$5terhhh5sU'rN^ifnhhhssNry;'lByhNssvalvEr(BPV)

' ID**#* responsec\\is inenitored during the low arc 33ure tests and l / bhe% electrical output response is monitored during IMl ITNGl rated preccurciftesthPlHelgest duration will be about 10 seconds to allow turbine valves and tailpipe sensors to reach a steady state. The tailpipe sensor responses will.be used to detect

bgg, the opening and subsequent closure of each SRV.

The BPVheMiMWe responsegDNill be analyzed for anomalies indicating a restriction in an SRV tailpipe. y Valve capacity will be based on certification by ASME , ~ code stamp and the applicable documentation being 14.2-181 Amendment 14 -4

M 4 l HCGS FSAR 01/86 l d. Acceptance Criteria Level 2 During a simulated main control room evacuation, the ability to bring the reactor to hot standby and 1 l subsequently cool down the plant and control vessel pressure and water level shall be demonstrated using equipment and controls located outside the main control room. I 14.2.12.3.27 Recirculation Flow Control a. Objectives l 1. To determine plant response to changes in the recirculation flow 2. To optimize the setting of the master flow controller i b. Prerequisites The -reactor is operating at steady-state conditions at the required power level. c. Test Method Y ~ With thelmid-pcezer reacter, plant at the mid-power load line, the recirculation speed loops are tested using large plus and minus step changes and the speed J-controller gains are optimized. lAfter the spccd 1 cops ha/c been Optimiced, the cycte= =cy bc switched to the T.acter manual =cdc and the = cater controller optimi=cd. When the plant is tested along the 100% load line, the recirculation system shall be tested by inserting small plus and minus step changes in the local manual and master manual modes. ! ( 1 1 14.2-187 Amendment 14 l

(1) Te( TEST OPEN HEAT on: NO. TEST NAME VESSEL UP 1 2 3 4 5 6 ( 2) Per ( 22) rod 1 Gmical and Ibdiochmical X X X X X X (3) Dyr y 2 Fadiation Measurment X X be( 3 Ebel Ioading X 4 Ebli Q3re Shutdown Margin 5 Cbntrol Ibd trive X X X( 2) X( 2) l ( 2) X( 2) (4)Afg; c1 6 SIM Performance X X 8 IFM Performance 9 LPRM Calibration X X X X (5)Bej 10 APPM Calibration X X X X X X am 11 Process Cm puter X X X X (6) MS 12 RCIC X 13 HPCI X X Rui X XI4) ~ 14 Selected Process 'Ibtp A@ 14 Riter level Ibf Ecg 'Ibnp X X X (7) ID XJ-15 Systen Expansion X X X 4 x) (8)IMj v 16 MTIP T.ccrt ir.r/ 17 Core Performance X X X X X X am [gp IB Stean Production V X X Xl (9) De 19 NC rc rc '/^id "cdc "^=r.cc 3 Pressure Regulator X X X X X X i 21 Feed Sys-Setpoint Changes X X X X X X X (10) AZ X( 5) 1 21 Ebed Sys-Ioss EW Heating X( 6) (11) pe 21 Feedwater Pump Trip l 21 Max EW Punout Capability X( 7) X(10) (12) pe; 1 22 'Ibrbine Valve Surveillance [x y,(Qj 23 MSIV Ebnctional 'Ibst X X(11)' (I h f 23 MSIV Full Isolation X (13) Pc J l 24 Pelief Valves JLg% . ( 20 X eX( 20 X( 20) 1

15) h X(17)

(14) Be X l c suroine 2 rip a waa I D! Ib3ection 26 Shetdown Outside CRC X (15) Ce j 27 Ibcirculation Flow Control X(14) X(18) va I l 28 Recirc-Ole Pump Trip X X 28 EFf Trip 'IWo Pumps X(19) (16) l 3 Pacirc Systm Performance l X Xp X X ay I 3 Wire Pr ?mam x-) INl 3 Pecire Sys Cavitation X (17) 30 Ioss of Offsite Pwr X 31 Pipe Vibration X X X X )( X (18) 29 Pacirc Flow Calibration X X (19)h 32 FAC3 X( 23) I 21I 33 RHR iX X 34 Crywell & Stean 'Ibnnel X X X X ( 2)) Cooling E 35 Gaseous Padwaste X X X 38 SACS Eurformance X X 40 Confinnatory In-Plant Test h X ( 21) ( 22) FSAR 3/7 A d ( 23) O 1

--c (1) Test conditions refer to plant corditions [;)4) DC gog,, in e on Figure 14.2-4 PRgvicostpr TgpjtpJ4gb 6 ( 2) Wrform Test 5, timirg of 4 selected control rods, in conjunction with expected scrams X X (3) Dynamic System Test Case to be conpleted betwen test corditions 1 and 3 X( 2) (4) After recirculation pump trips (natural circulation) X (5) Betwen 80 and 90 percent thermal power, X and near 100 percent core flow X (6) Max FW Runout Capability & Pacirc Pm p knback must have already been cunpleted X (7) Feactor power between 80 and 90 percent X (8) Ibactor power between 45 and 65 percent X X and 75 and 90 percent X (9)' Deleted X (10) At maximum powr that will not cause scran X X(5) X( 6) (11) Perform between test corditions 1 and 3 X(7) (8) X(10) (12)ilisctor mer t,ct cen 40 rd 55 perccat] = (13) p X (13) licactor ;r cr tet.;ccn 50 r.d es pr=nt; e X( 20) X(17) (14) Between test conditions 2 and 3 MLETED/ (15) ICc=rctor load rejectied within bypass

(18) valve capacity X

(16) %cctor peer beteen 50 W 90 parent 2 X Ot 00r0 fl0" l 95 parent - tudi=^ trip-1 (17) Iozd rejection X (18) Between test conditions 5 and 6 9._ X (19) >50% power and >95 core flag c-d perform X( 21) lbcforc '"arbinc 'Irip L^2* P2Wth X (20) Check SRV operability durirg major scran X tests X ( 21) Mrformed during cooldown fran test cordition 6 HOPE CREEK GENER ATING STATION FINAt SAFETY ANAtY$l$ REPORT ( 22) The test number correlates to FSAR Section 14.2.12.3 x where x is the irdicated test nmber. TEST SCHEDULE AND CONDITIONS ( 23) May be perfonned any time test conditions permit. FIGURE 14.2 5 Amendment 14,01/M

HCGS FSAR 11/85 / 21ECH2) l NOA reviews and approves \\% station inspection plans and procedures that implement the OA program, including testing, calibration, maintenance, modification, rework, and repair. Changes to these documents are also reviewed and approved. In addition, NOA is responsio'le for review and approval of PSE&G specifications, test procedures, and results of testing. 17.2.6 DOCUMENT CONTROL Instructions, procedures, drawings, and changes thereto are reviewed for inclusion of appropriate OA requirements and are l, approved by apppropriate levels of management of the PSE&G l organizations producing such documents, and distributed on a timelv basis to using locations. Measures are provided for the timelf removal of obsoleted or superseded documents from the using location. Supplier documents are controlled according to contractual agreements with suppliers. The following is a generic listing of key documents for the operational phase, showing minimum organization responsibility for review and/or approval, including changes thereto: a. Design specification - Engineering and Plant Betterment, NOA b. Design, modification, manufacturing, construction, and installation drawings - Engineering and Plant Betterment, Nuclear Services, Hope Creek Operations, NOA c. -Procurement documents - initiating nuclear department organization, Purchasing Department, Site Services, NOA l d. Nuclear Department Manual - Nuclear Department organizations responsible for implementation, NOA e. Nuclear department second-tier manuals, including station administrative procedures - cognizant department head, NOA f. Maintenance, modification, and calibration procedures for 0, F, and R-designated station work activities - Hope Creek Operations, NOA g. Operating procedures - Hope Creek Operations, SORC 17.2-19 Amendment 13

HCGS FSAR 4/84 l b. A Class 1E MCC supplying, through a Class 1E circuit creaker, the backup power supply to an uninterruptible power supply (UPS) system that supplies a non-Class IE distribution panel. These are shown on Figure 8.3-11. c. A Class 1E MCC and 125 V de switchgear supplying a Class 1E inverter that feeds a non-Class lE distribution panel. This is shown on Figure 8.3-11. For confiaurations a and b above the Class 1E circuit breaker that connects the non-Class lE load to the Class 1E power supply is tripped by the LOCA signal. This circuit breaker is an acceptable isolation device between Class 1E power supplies non-Class IE loads per IEEE 384-1981. For configuration c, the Class 1E inverter that connects the non-Class lE load to the Class IE power supplies is an acceptable isolation device between a Class 1E power supply and non-Class 1E loads, per IEEE 383-1981. Some of the other isolation devices employed at HCGS are optical isolators, Potter Brumfield MDR relays, and other relays. All these devices have been qualified as Class IE isolation devices. The Class 1E cabling is physically separated from non-Class 1E cabling in the raceways and inside electrical equipment per the requirements of Regulatory Guide 1.75, as described in Section 8.3.1.14. The differences are analyzed for acceptance. NSSS: accercment-of NSSS inctrumentation and centrol deficec ic ^ underway to determine Nhether all "c"-Clacc !E cystemt and-componente acccciated with the Clacc 1E cyctemc are qualified-+n acccrdancc with 10 Crn 50.45 rcquirements or to develop a dcccription of design featurec that preclude failure of the-Clarc !E circuit cr equipment- [ 105RT'l l 430.35-2 Amendment 5 l

INSERT FOR PAGE 430.35-2 A detailed analysis for design application for all non-Class 1E devices connected to a Class lE power supply was performed within the guideline requirements of Regulatory Guide 1.75. These analyses include evaluation of the circuit configuration in which the non-Class lE component is used, justification by similarity of this component to an identical Class lE qualified component, seismic withstand capacity of the device and in some cases the potential failure modes of the device and their-affects on the Class lE bus or the safety function. The non-Class lE components / devices so analyzed demonstrate that they do not have any inherent failure mechanism, different-from those of the Class lE component / device, that can affect the Class lE power supply or degrade the plant safety function. Thus these analyses provide adequate justification for the use of above non-Class lE devices on the Class lE power supply. 1 .i --}}

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