ML18018A684

Forwards Responses to Accident Evaluation Branch SER Open Item 236,on Fuel Handling Bldg,Mechanical Engineering Branch Open Item 286 on Leak Rate Testing,Power Sys Branch Open Item 307 & QA Branch Open Item 220

ML18018A684
Person / Time
Site:	Harris
Issue date:	08/12/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LAP-83-376, NUDOCS 8308190223
Download: ML18018A684 (87)

Text

{{#Wiki_filter:b. REGULATORNFORMATION OISTRISUTION 8+EM (RIOS) i<<'GCE'SSION NBR $ 8308190223, DOC ~ DATE! 83/08/12 NOTAR IZED e NO DOCKET FACIL:50 400 Shear on Har ris Nuclear Power Plentr Unit 1< Caroline 05000400 50 401'Shearon Harris Nuclear Power PlantF Unit 2E Carolina 05000401 AUTH ~ NAME AUTHOR AFFILiATION MCOUFFIEgM ~ AD 'Carolina iPower 8 Light Co ~ RECIP NAME RECIPIENT AFFILIATION OENTONrH ~ RE Office of Nuclear Reactor Regul at ionr DirTec tor ~u.

SUBJECT:

Forwards responses o ccident Evaluation Branch Open Item 236Fon fuel handling bldg Mechanical Engineering Branch Open Item 286 on leak rate:testing Power Sys Branch Open Item 307 3 QA Branch Open Item 220 ' IS 7 RI8 UTIO N C 0 D E ~ 8 0 0 1 S C 0 P IE S R E C EI VE D ~ LTR J 'NCl ~(~ 8 IZ E ~, Ie~~eeMee CoM,. TITLE: Licensing "Submittale PSAR/FSAR Amdts 8 Related Correspondence NOTES: RECIPIENT IO CODE/NAME NRR/DL/ADL NRR LB3 LA INTERNAL ~ ElO/HDS1 IE/OEPER/KPB '36 IE/DEQA/QAB '21 NRR/DE/CEB 11 NRR/OE/EQB 13 NRR/DE/MEB 18 NRR/OE/SAB 24 NRR/DHFS/HFE840 NRR/OHFS/PSRB NRR/OS I/AEB 26 NRR/OSI/CPB 10 NRR/OSI/ICSB 16 NRR/OSI/PSB 19 NRR/OS I/RSB 23 RGN2 EXTERNALS ACRS 41 DMB/DSS (AMDTS) LPDR 03 NSIC 05 <<COPIES LTTR ENCl 1 0 1 0 1 0 -3 3 1 1 1 1 2' 1 1 1 1 1 1 1 1 1 1 1 1 1 1 3 t'3 6 6 1 1 1 1 1 RECIPIENT ID CODE/NAM<<E NRR L83 BC KADAMBIF P 01 IE FIlE IK/DEPER/IRB 35 NRR/DE/AKAB NRR/OE/EHEB NRR/DE/GB 28 NRR/OE/MTEB 17 NRR/DK/SGEB 25 NRR/OHFS/LQB 32 NRR/OL/SSPB NRR/DS I /ASB NRR/DS I/CSB 09 NRR/OSI/METB 12 22 REG FILE 04 79 I8 BNL(AMDTS ONLY) FEMA RKP DIV 39 NRC PDR 02 NTIS COPIES LTTR ENCL 1 0 1 1 1 1 1 1 1 0 1 1 2 2 1 1 1 1 1 1 0 1 1 1 1 1 1 1 1 1 1 1 0 1 1 1 1 1 1 1 1 ATOTAL NUMBER OF COPIES REQUIRED: L'J'P 53,ENCl 46

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Carolina Power & Light Company AUG 13 1983 SERIAL: LAP-83-376 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES

Dear Mr. Denton:

Carolina Power & Light Company (CP&L) hereby transmits one original and forty copies of responses to Shearon Harris Nuclear Power Plant Draft: Safety Evaluation Report Open Items. The response numbers are listed on the cover page of the attachment along with the corresponding review branch and reviewer for each response. We will be providing responses to other Open Items in the Draft Safety Evaluation Report shortly. Yours very truly, FXT/ccc (7658FXT) Attachment M. A. McDuffie Senior Vice President Engineering & Construction cc: Hr. E. A. Licitra (NRC) Hr. G. F. Maxwell (NRC-SHNPP) Mr. J. P. OlReilly (NRC-RII) Mr. Travis Payne (KUDZU) Mr. Daniel F. Read (CHANGE/ELP) Chapel Hill Public Library Wake County Public Library 8308190223 8308i2 PDR ADOCK 05000400 PDR Hr. Wells Eddleman Dr; Phyllis Lotchin Mr. John D. Runkle Dr. Richard D. Wilson Mr. G. 0. Bright (ASLB) Dr. J. H. Carpenter (ASLB) Hr. J. L. Kelley (ASLB) 5oo ) 411 Fayetteville Street ~ P. O. Box 1551 e Raleigh, N. C. 27602

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LIST OF OPEN ITEMS, REVIEW BRANCH AND REVIEWER Accident Evaluation Branch/K. Dempsey Open Item 236 Mechanical Engineering Branch/D. Terao Open Item 286 Power Systems Branch/0. Chopra Open Item 307 Quality Assurance Branch/R. Kirkwood Open Item 220 (Partial) 8308190223

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Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 236 It is not clear from the staff analysis of these systems as to whether the Fuel Handling Building (FHB) is to be at a negative pressure during fuel-handling operations. In addition, the applicant failed to give a response time for switchover from the normal mode to the post-accident mode as compared to the travel time from the pool surface to the isolation dampers for the FHB. Furthermore, the location of the radiation monitors around the sides of the pool wall seem to be ineffective in registering a radioactive release from the pool because the pool ventilation system is designed to direct the flow of air over the pool from the sides of the pool to the pool center and upwards to the intake vents. In view of these unresolved

items, the fuel handling accident is an open issue.

~Res esse The FHB will only be held under negative pressure following an event involving the release of radioacivity in the FHB atmosphere. Following an accident such as a fuel handling accident radioactivity released from spent fuel rods will be detected by the radiation monitors located around the spent fuel-pool. These, radiation monitors will then signal the switchover from the normal to the emergency ventilation and filtration system. The switchover time is 30 seconds for the emergency ventilation and filtration system to become fully operational. The isolation of the normal ventilation system is accomplished in < 10 seconds. The total travel time for gaseous radioactivity to travel from the spent fuel pool surface to the isolation damper was conservatively calculated to be 29.7 seconds;

however, the closure time of the isolation damper is

< 10 seconds. Thus no radioactivity is released through the normal filtration path. To initiate operation of the emergency ventilation system and terminate normal ventilation system operation radiation monitors are provided at appropriate locations as shown in Figure 12.3.2-9 (four sets of three for safety function). The radiation detectors are located on the FHB walls and are extended low range GM tube detectors as described in Section 11.5.2.5.2, monitoring the air volume over the fuel pools. In the event of a fuel handling accident the gaseous radioactive material that is assumed to be released will be swept up into the FHB ventilation system. ,The radiation (gamma) that woul.d emanate from the radioactive material in the airflow would cause alarms when the exposure rate exceeded preset limits. The detectors high radiation alarm will actuate switchover from normal FHB ventilation to emergency ventilation system operation. The radiation detectors are described in FSAR Section 11.5.2.5.2 and 12.3.4.1.8.3. The monitor's range is 10 x 10 through 1 x 10 mr/hr with the high alarm set-point at 1 x 10 mr/hr and the alert alarm set-point 2.7 x 10 mr/hr. The capability of the GM tube detectors was based on assuming the activity releases from the accident discussed in Section 15.7.4 which equated to an appropriate line source. This assumed line source provides an exposure rate for the GM tube detectors to monitor. The time required for the accident released activity to provide an exposure rate to exceed the high alarm set-point for the monitors and initiate switchover is such that any released activity will be within required limits as discussed in Section 6.5 and 15.7.4. The FHB operating floor, spent and new fuel pool

areas are provided with two ventilation systems each of which have Particulate,

Iodine, GAS (PIG) airborne effluent monitors monitoring the ventilation exhausts as described in Sections 11.5.2.7.2.2 and 11.5.2.7.2.3.

The FHB normal exhaust is provided with effluent airborne monitoring for indication of airborne activity to operations personnel. Operations personnel have the capability to initiate the FHB Emergency Exhaust system from the control room as described in Section 7.3.1.3.4. The FHB Emergency Exhaust is provided with a PIG monitor for monitoring effluent exhaust downstream of the emergency exhaust systems HEPA-Charcoal filter units. This airborne effluent monitor measures effluent releases during and after a fuel handling accident. Any airborne activity release by the FHB normal ventilation system prior to switchover to the emergency exhaust system will be monitored by the FHB normal exhaust monitors. After switchover the FHB ventilation exhaust will also be monitored. The analyses performed to determine the adequacy of the 30 second switchover time for the FHB ventilation system is described in Attachment A. The two analyses determine the following: 1. The time of radioactivity travel from the spent fuel pool surface to the normal ventilation intake vents isolation damper; (calculated conservatively assuming these dampers remain open). 2. The maximum allowable bypass period following a fuel handling accident. The ventilation system for the Fuel Handling Building (FHB) shown on Figures 9.4.2-1, and 9.4.2-2 shows applicable areas covered by the FHB ventilation system. The control drawing for the ventilation system switchover from normal ventilation to emergency exhaust is shown on Figures 7.3.1-13 and 7.3.1-14. Discussion of the FHB Emergency Exhaust System is provided in Section

6. 5.

Compliance of the FHB with. the requirements of Regulatory Guide 1.52 Revision 2 is shown on Table 6.5.1-2. The FHB operating floor area has GM tube area monitors at appropriate locations on the FHB walls, monitoring the building volume by the spent and new fuel pools. As described in Section 12.3.4.1.8.3 these GLLL tube area monitors will detect gamma radiation emanating from airborne material being drawn up into the FHB ventilation system from a fuel handling accident. When preset levels are reached, a high alarm signal will initiate switchover from normal ventilation to emergency ventilation as described in Section

6. 5.

The FHB normal and emergency ventilation exhausts are monitored by airborne effluent Particulate,

Iodine, Gas (PIG) monitors in the exhaust ducts as described in Sections 11.5.2.7.2.2 and 11.5.2.7.2.3.

The capability to initiate the emergency ventilation system is provided in the control room as described in Section 7.3.1.3.4. The FHB emergency ventilation system switchover time and the FHB normal ventilation dampers isolation time of < 10 seconds is within an acceptable duration that limits offsite doses to less than 10CFR100 limits.

ATTACHMENT A 1. Calculation of the time of radioactivity travel from the spent fuel surface to the normal ventilation intake vents isolation dampers. Time fo'r gaseous radioactivity to travel from the spent fuel pool surface to the isolation damper of the normal ventilation system consists of travel time from the pool surface to the intake header and from there to the isolation damper through the length of ventilation duct. These times are evaluated as follows: A. Travel time from refueling pool surface to exhaust duct The equations of flow for round hoods is obtained from "Industrial Ventilation," 8th edition, by the American Conference of Governmental Industrial Hygienists. The velocity profile is given by: v- ~ 10xX +A

where, V ~ centerline velocity at distance X from hood, ft/min X ~ distance outward along axis, ft (equation is accurate only for limited distance of X, where X is within 1.5D, where D is duct diameter)

OI 236-3

p.

'V ~ ATTACHMENT A (Cont'd) Q = air flow rate, cfm A = area of hood opening, ft2 j ~I D = diameter of round hoods or side of essentially square hood, ft Using Equation (1) above, the average velocity between the hood and any distance X can be obtained as follows: X Vav. =1 Q dX (2) 0 X'-X "vg. = ~XrOO> X'=0 The distance between pool surface and intake header is 44 ft. X = 1,.5D = 1.5 x~ = 1.5 ft 12 (X is evaluated using the smaller side of the intake header) Q = 2,300 cfm 12" x 24" 2 A = ~4< = 2 ft Vavg. in the first 1.5 equals: 2,300 ft of the distance from the intake header tan" 1.5 (10x2 Vavg. 439.2 ft/min Travel time for the first 1.5 ft becomes: 1;5 't.5ft 439 43 Air velocity at 1.5 ft is given by Eq. (1) 2,300 '1 Sr =~i ~~ Conservatively assuming that velocity beyond 1.5 ft does not decrease then travel time required for balance of the distance can be calculated as follows: t = x 60 = 27. 1 sec 1= 94 The travel time, t2, from the intake header to the isolation damper is calculated to be

2. 6 sec.

The total travel time is then: ttotal = tl + t2 = 29.7 sec OI 236-4

ATTACHMENT A (Cont'd) This total travel time is slightly less than the switchover time of 30 seconds. Nevertheless, the 30 second switchover time is considered to be appropriate considering the conservative method used to evaluate the travel time from the pool surface to the intake header. The main point of conservatism is that the velocity of air well beyond 1.5 ft of the intake header was assumed to be constant and applicable at the pool surface. A more realistic analysis (but unwa'rranted for this purpose) of the velocity profile between the intake header and the pool surface would indicate progressively smaller velocities while moving away from the intake header. Consequently, the travel time" of radioactivity would be much longer. The temperature difference between the pool water and the FHB af.r will create convective air flows which enhance dilution of radioactivity released from the pool. Conclusions of the analysis of the maximum allowable bypass period following a fuel handling accident. The analysis of the maximum bypass time was performed assuming uniform mixing of radioactivity in the FHB atmosphere and using the guidance given in Regulatory Guide 1.25. Bypass time is defined as the time period during which gaseous radioactivity is released unfiltered, therefore bypassing the Emergency Exhaust Filtration system. Assuming a two-hour duration for the FHA as prescribed in Regulatory Guide 1.25, a uniform concentration and release rate, the radioactive effluents could be released unfiltered during the initial 36 minute period after the accident without exceeding the dose guidelines for this event. This value of 36 minutes is an indicator of the time margin which is available to switchover to the emergency filtration system and thereby isolating the FHB prior to exceeding the dose guidelines for a fuel handling accident. OI 236-5

Mechanical Engineering BranchlD. Terao Open Item 286

OPEN ITEN 286 3.9.6 ISI - Pumps and Valves pg. 3-41 gUESTION: The applicant must provide the program for the leak rate testing of those valves that provide an interface betwee reactor coolant pressure boundary and low pressure systems.

RESPONSE

CP&L has reviewed the staff's criteria (based on proposed change to Standard Review Plan 3.9.6 Appendix A) for determining which valves are to be leak tested as part of the ASME Section XI program. The attached list reflects all those valves which meet that criteria. Each valve on the list has been reviewed in light of actual operating conditions and the following approved criteria and technical studies: a) MASH-1400, Appendix V, pages V-43, V-44 b) NUREG-0677 Probability of Intersystem LOCA c) EPRI-NP262 PVR Sensitivity to Alterations in.the Interfacing Systems LOCA Those valves which CP&L has determined do not require testing and the justification for their exclusion have been indicated by reference to attached notes.

Valve Number Descri tion SI-V544 Accumulator Injection 'SI"V545 Accumulator Injection SI-V546 Accumulator Injection SI<<V547 Accumulator Injection SI-V548 Accumulator Injection SI-V549 Accumulator Injection RH-V500 RHR Suction RH-V501 RHR Suction RH-V502 RHR Suction RH-V503 RHR Suction SI-V510 High/Low Head Recirc. -Drawing (location) (2165-G809 (D5) G809 (G5) G809 (J5) G809 (D3) G809 (G3) G809 (J3) G824 (I3) G824 (I4) G824 (L3) G824 (L4) N N N Note 1 Note 1 Note 1 Note 1 G808 (B11) Y

Test, Yes No Justification SI-V23 BIT Injection SI-V29 BIT Injection SI-V63 Alternate Injection SI-V511 High/Lc4 Head Recxxc.

SI-V512 High/Low Head Recirc. SI-V513 High/Low Head Recirc. SI-V514 High/Low Head Recirc. SI-V507 High/Low Head Injection SI-V508 High/Low Head Injection SI-V509 High/Low Head Injection SI-V584 Low Head Injection SI-V585 Low Head Injection SI-V586 Low Head Injection SI-V17 BIT Injection G808 (Bll) G808 (B17) G808 (B17) G808 (C17) G808 (B3) G808 (B3) G808 (C3) G810 (C1) G810 (Dl) G810 (El) G808 (D3) G808 (D4) G808 (D5) G808 (D6) Y N N N N N Note 2 Note 2 Note 2 Note 3 Note 3 Note 3 Note 4 Note 4 Note 4 Note 4

Valve Number Descri tion SI-V69 Alternate Injection 'I-V75 Alternate Injection RC-V7 RCS Loop -1 Drain RC-V8 RCS Loop 1 Drain +FSAR drawing numbers may be found Drawin FSAR No. RC-V16 RCS Loop 2'rain RC-V17 RCS Loop 2 Drain RC-U18 RCS Loop 2 Drain RC>>V28 RCS Loop 3 Drain RC-V29 RCS Loop 3 Drain CS-V505 CVCS Normal Charging CS-V504 CVCS Normal Charging CS-V506 CVCS Alternate Charging CS-V507 CVCS Alternate Charging CS-V711 CVCS Pressurizer Spray CS-V70 CVCS Pressurizer Spray CS-L500 CVCS Letdown CS-L501 CVCS Letdown CS-L509 CVCS Excess Letdown CS-L510 CVCS Excess Letdown >"Drawing (location) (2165-Test Yes No G808 (D7) N G808 (D8) N G800 (C3) N 6800 (C3) N G800 (M2) N G800 (M2) N Drawin FSAR No. G800 (M2) N 6800 (D17) N G800 (D17) N .'G803 (C3) N 6803 (C3) N G803 (B3) N G803 (B3) N G803 (D3) N G803 (D3) N G803 (A3) N G803 (A4) N 6803 (C8) N 6803 (C8) N from the following table: Just- =ication Kore 4 Not,e 4 Kore 1 Yo=e 1 Yo=e 1 Yore 1 So-e 1 So-e 1 So=e 1 Kore 4 No-e 4 Xo-e 4 hote 4 Kore 4 Kote 4 Kote 1 hone 1 Note 1 Sot.e 1 G800 6803 6808

5. 1.2-1 9.3.4-1 6.3.2>>1 G809 G810 G824 6.3.2"2 6.3.2-3 5.4.7-1

Note 1 The Event V configuration has never included valves other than check valves. Stem actuated valves need not be included in a program predicated on the prevention of an Interfacing Systems LOCA (ISL) for the following reasons: As applie" to ~PP, stem ac=tated valves have been used to provide R S isalaticn only when their failure to actuate will n"t affect safety during an accident or when there is no fewible use for check valves.

2. 'tem actuated valves provide positive indication of their closed positio=.. This fact, combined with the required periodic stroking. requi=ed for, ASIDE Section XI Category B

valve, provides a determination on their status as barriers. 3. Addressing the concerns of h~G 0677, which included motor actuatec valves, he o=ly set of such valves are those vhich prcvide RHP. suction isolation. As stated in SPlPP's FSAR 3 ~.7.1: , Each motor operated valve is interlocked to prevent its opening if RCS is greater than 425 psig...". This cccnbined vith performing the stroke test only when the RCS is at cold shutdown provides adequate assurance that pressure barrier failure probability v,' remain "extremely small", i.e., a sequential rup=re of hath valves. This is supported by NUREG 0677 (p. 133 and provides an ISL probability of 4.2 "9 x 10 per rea=o year Note 2 These valves (SI-V512, V513) are in series with SI-V510 and SI-V511 which will be leak tested and SI-V587 which is a normally closed MOV. Carolina Power & Light Company has determined that these valves do not require testing based on the following analysis: a. Failure of SI-312 or V513 will cause Interfacing Systems Loca (ISL) only if V510 or V311 respectively and V587 also fail. b. Because V5 87 is a motor opera?ed valve that is stroked tested at cold shut"om and verifiec closed, it will only contribute to ISL by a wptu~e fa'lure. c ~ Leak testing does no-a=feet the rupture failure probability of the HOV. d. Check valves 7510 and %~~11 a e leak tested during start-up after RCS has reached f=ll p=essure. This precludes a leak failure cantrib tion to ISL, since check valve position remaixa consta=t during operation.

Using the above conc'tions, there are tao sets of rupture-rupture failure sequences (V510 and V587, V511 and V587). Using hASH-1400 failure rates and ass~g 1 year test interval, the -8r failure probability is 1.5 x 10 yzeactor year. Note 3 For the series of th ee check valves providing isolation betveen the RCS and RHR, CP&L vill. test SI-V580 and V581 instead of V507, V508, and V509. Note 4 These valves do not. require leak testing for the folloving reasons: a. These valves provide isolation between RCS and the safety injection/charging system vhich is rated at the same or higher pressure. b. A safety injection/charging pump is constantly running during normal/high pressure safety injection, thus pressurizing the inlets to these valves at higher than RCS pressure.

SI".'NPP VALYE TEST PROGRAM UNIT I sn-,-",v: Safety njection (SI) T t p~~ pg+ fg><J 808(7) Page 1 of 2 Ul r0 CA V50' V517 Ol4l R ( 0 0 000 8-5 N-5 L-6 VAt V ~~T=GXY A J 8 o N a C:0 l P l 0 i'O 0 C CA 0 >-z (n~ 4l 0 1 1 1 0 IQ rl 5 iil 0tij FS-1 Ts FS-1 TS FS-1 FL-1 TS U Ir LJ I 4! l( 4l l0 V515 2 R500 2 V505 V506 H-5 F-3 3/4 MO ifO .0 1. 1 a 1 FS-1 FL-1 III~I la7 FS-1 Ts Fs-l TS V23 V29 V507 1 D-3 1 D-4 D-5 FS-1 Fs-a Fs-1 FS-2 FS-2 FS-2 FS-Z. V508 1 V509 1 V501 V3cl V45 v51 v510 V5'l 'l C-3 D-3 D-12 D-13 D-14 c-la CK CK CK CK 2 2 FS-1 FS-1 Fs-1 TN Fs-a FS-1 .""s-a F$ 9 FS-2 rS-2 rs FC r 2

C

SHf HAPP VALV= Tt.S: PROGRAiVl L"AIT C/I CJI QZ ( C 0 C) C7 ttt I C/I ttl I VALVE t7Z l C/I g ~O I 0( JO tt Cl ~ tt/ ttt ttl O CJI C ~\\ ttI C/I c= C/I C/l I tt/ C/I tal CCI O I tJ C ( O z 2 CATEQQRY'tI lOX A B r ~ gyp r p. Self erv Tnjection (SI) &./la i'Re) 810(7) age 1 of 2 O V570 2 M-6 14 V571 2 3-6 14 vp FS-1 TM V572 'M 7 14 'Mp FS-1 V573 H-7 14 MQ 5 V574 2 M-10 14 FS-1 TH '575 ii-10 0 FS-1 V590 2 591 2 R520 2 R521 V576 M-12 N-12 B-4 X 14 14 3/4 3/4 10 Mp FS-1 RL-3 rS-1 V577 2 D-5 10 GA t/0 0 FS-'Jf 587 4 10 "."S-1 jM 578 2 F-4 10 FS V579 C-4 10 CA r5-1 580 '581 r 3 C-3 10 10 CK FS-1 R-9 R-9 ." 5-2 '584 C-1 FS-1 R-9 ."5-2

Si-'NF'P VALVE Tt=ST PRCGRAVl UNIT I SYSi=:C: Sa=e-y Inject'on (SI) D~g Aa rFc) 810(7) Page 2 or" 2 C ssC CA CA C7 c C/l Ol tel 0Z C2: C C ~ 0C CJ ~YE CAi c;C~RY A I 8 C ~sO ~4 \\J CO c 4l O l I 0 5 I C3 <n'~ QJ ~ l Ue I C/lu pcs 0 C: QJl I1 IOR V585 V586 1 X CK PS-1 FS-1 R-9 7S-2 PS-2

RELIEF REQUEST: . VALVES: R-11 SI-V587 CATEGORY: CLASS: FUNCTION: Isolate RHR from SI Recirculation System TEST REQUIREMENT: Exercise valve for operability and 'measure stroke time quarterly. BASIS FOR RELIEF: Opening of this valve during normal operation increases the chance of interfacing systems LOCA. This valve is never used until well into an accident sequence when reverse flow through the reactor is desireable. Manual actuation could be affected without affecting safety. ALTERNATE TEST: This valve will be tested during cold shutdown when RHR is in operation. (7 344NLUlcv)

Power Systems Branch/0. Chopra Open Xtem 307

Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 307 Describe in detail, how the Shearon Harris design meets BTP PSB-1 "Adequacy of Station Electric Distribution System Voltages. ~Res ense The enclosure to this response describes in detail the Shearon Harris design compliance with BTP-PSB-1. The enclosure addresses each position and associated design considerations. Analysis presented in the enclosure establishes the adequacy of electric distribution system voltages to meet all postulated plant conditions and mitigate the consequences of an accident. Distribution system relaying has been designed to minimize the effect of a degraded offsite power system on the plant safety related distribution system. Additional relaying needed to achieve this, as presented in Postion-1 of the enclosure, will be incorporated into the plant. As addressed in Position-4, actual field testing will be conducted prior to plant operation to verify the accuracy of the auxiliary system analysis results within the guidelines presented in the BTP PSB-1. The appropriate test summary will be added to the FSAR in a future amendment.

y ~ ~ ~ ENCLOSURE SHEARON HARRIS NUCLEAR POWER PLANT COMPLIANCE WITH BTP PSB-1 "ADEQUACY OF STATION ELECTRIC DISTRIBUTION SYSTRi VOLTAGES" CAROLINA POWER & LIGHT COMPANY

~ ~ ~ .Position-1 The undervoltage protection for the Shearon Harris safety related Electrical distribution system utilizes tvo (2) independent undervoltage schemes. Each of these protection schemis consists of (3) three undervoltage relays (instantaneous types) and associated timers. One set of undervoltage relays and its timers,,henceforth described as the primary system, consists of undervoltage relays 27-1, 27-2, and 27-3 ~ The second set and its associated timers, henceforth described as the secondary system, consists of relays 27A-L, 27k-2 and 27k-3 Pigure 430.93-1 shoes connection of undervoltage relays to 6.9 kV CLass LE buses LAWh and LB-SB. the Class LE buses. These relays, upon sensing a loss of voltage, automatically disconnect the offsite source from the Class LE buses as described ia PSAR Section 8.3.1.1.2.8 and starts the dieseL generator. Shen the diesel generator has attained rated speed and voltage (within 10 seconds after the'tart signal), the diesel generator incoming breakers to Class LE buses are closed and the Class LE loads are connected to the busses automa'tically by the emergency load sequencer in accordance vith the loading sequence shovn in the PSAR Table 8.3.1-2. Once the loading of the diesel generator has begun, operation of the undervoltage relays is blocked. The secondary system along vith its timers provide protection for the Class LE buses against de raded volta e condition. Tbe secondary relays provide protection to all Class LE loads should voltage on the CLass LE buses drop to a value Lover than the minimum acceptable voltage and persist for a time longer than that can be tolerated by any connected load. The secondary undervoltage relays are connected to two distinct time delay relays. Upon expiration of the first time delay, which is long enough to accommodate the starting of the motor which has the, longest starting time, an aLazm is actuated at the main control board to alezt the operator to the need for action to restore'he system voltage.

Hovever, should a safety actuation signal be present after the expiration of the time delay, automatic tripping actions as described for primary protection is initiated.

Ifno safety actuation signal is present, a further time delay is allowed before the automatic tripping actions are initiated. This second time delay is based on the maximum time for which the most sensitive load can perform its safety functio'n vithout impairment at the degraded voltage. Pigure 430.93-2 shows the undervoltage detection scheme for primary and secondary protection systems and facilities for testing system performance, verifying relay settings and indicating bypass and other abnormal conditions. Contzol svitches at the main control board and pushbuttons at the svitchgear are provided for testing the undervoltage protection system. The test 'includes the undervoltage

relays, the time delay relays, the lockout trip relays and the associated Snterposing relays for the primary and secondary undervoltage systems.

The testing circuit is designed to manually initiate the undervoltage tripping action @bile bLocking the actual tripping of the bus loads.

, ~ 1 ~ <e The test is conducted in the following manner: Switch turned to "U.V.L.O. Test," position at the Main Control Board or the pushbutton is depressed at the switchgear. A normally bright lit white light, located above the switch, will go off. kn alarm at the control room annunciator will sound, indicating undervoltage tr2y. kn alarm at the MCB also sounds after 15 seconds, annunciating the degraded voltage at the bus. A second alarm at the MCB sounds after 60 'seconds, annunciating bus tripping initiation on a degraded bus voltage. Xf a real undervoltage occurs during testing, tritest circuit will be reset automaticaU.y and actual tripping actions and diesel generator startup will be initiated. The undervoltage relays and associated timers are located in the 6.9 kV Class 1E switchgear. The relay contacts are combined in a 2 out of 3 logic to generate loss of voltage or degraded voltage signals and one out of three loge to generate an alarm on loss of instrument transformer potential or to detect any other abnormal condition. The protection scheme~designed'$ consistent with the recommendation of ZZEE-279-1971. The settings of the primary and secondary undervoltage relays and their associated timers are based upon the results of an auxiliary system voltage analysis performed for the Shearon Harris Plant. This analysis demonstrates that the voltage. set points for the primary and secondary relays are such that the Class lE electrical distribution system does not remain connected to the offsite power sources when available voltage degrades to a value where Class lE equipment can not fulfillits intended function. Details of the auxiliary system study are presented in Position-3. The settings of the undervoltage relays take into account the following design capability of safety related loads; all Class .1E moto'rs are designed to operate satisfactorily at 90Z of their rated voltage (i.e. 6.6 kV motors on a 6.9 kV bus). Zn addition, Class 1E motors are designed to start and accelerate their driven equipment at 75X of their rated voltage, which is equivalent to 72X of the nominal bus voltage. Based upon the above design criteria and an analysis of auxiliary system voltage requirements as presented in Position-3, during steady state operation a bus voltage above 88X of the nominal 6900 volts (considering a 2X diop between bus and the motor) could be permitted to remain indefinitely without causing harm to the motors.~ During transient operation, bus voltages lower than 70X of the nominal 6900. volts are a definite indication of loss of voltage.

.The basis of thc time delay settings of the primary and secondary uadervoltage .relays are the capability of safety related 1oads to remain connected to the bus for a degraded voltage coadition as veil'as the permissible maximum time dalay assumed in the FSAR accident analysis for actuation, of engineered safety features systems. Two scenarios are considered to address the adequacy of desiga. The fixst scenario postulates an accident vith a degraded grid voltage condition vhere thc 6900 V Class LE bus voltage degraded to a value where the CLass LE motors caa aot perform their safety related functions i.e. motor terminal voltage is belov the aixdmma contiauous operating level. The second sceaario postuLates an accident vith Class LE bus voltage corresponding to the minimum,allowable stax'ting voltage of the Class LE motors. Por the first scenario, the degraded grid voltage vill bc at a level such that upon initiation of an accident sigaal aad all plant loads are running the voltage at the CLass LE 6900 V bus is belov 89X of the nominal 6900 V. Degraded voltage relays 27A-L, 27h-2 and 27A-3 vill dropout at this level. Since this level of bus voltage could also occur during motor starting transieats, a 15 second time delay is alloved prior to disconnection of the offsite source. This time delay is based on the longest motor starting time. In the absence of an accident signal, a 60 second time delay is allowed prior to initiation of tripping of the offsite source breaker if the degraded voltage condition exists. The 60 second time delay is based on the capability of safety reLated motors to operate for't least one minute at voltages anyvbere ia the raage betveea secondary and primary undervoltage relay settings. If, hovever,'the degraded grid voltage is such that the Class LE bus voltage deteriorates to the level of the setting of the primary (loss of voltage) relays px'ior to 60 seconds, then the primary-. relays initiate the trippiag of the offsite source. The second scenario postulates an accident vith the Class LE bus voltage belov the minimum voltage required for startiag of the safety related motors, 'this vould be a level vhcre the primary relays would drop out. h tine delay of 0.5 seconds is provided prior to disconnection of the offsite source to allov the circuit protective 'device to clear the fault (if the undervoltage conditioa at the CLass LE bus resulted from a fault at the feeder circuit). Attached Table-1 shovs the settings of the primary and secondary relays aad their associated timers. Table-2 and 3 shov the sequence of events and associated times for the tvo scenarios considered above. The numbers in the parenthesis for specific events indicate the maxinum time delay considered in the FSAK accident analysis.

TABLE-1 SHNPP PSAR 3.3 Relay Settings-System Voltage (E) Potential Transf. Primary Potential Transf. Ratio UV Relay Coil Rating Primary Relay Setting Primary Relay Type Secondary Relay Setting Secondary Relay Type 6900 Volts , 7200 Volts 60:1 120 Volts 83 Volts (1) NGV 13B 102 Volts (1) ~.~ ITE-278 TIME SETTING (2 Primary Relays Secondary Relays 0.5 Seconds 15 Seconds (3) 60 Seconds (4) NOTES: (1) Undervoltage relay settings are Dropout settings. Primary relay setting is based on 0.72 E nominal, Secondary relay setting is based on 0.89 E nominal. (2) Time settings are selected so that for undervoltage protection they do not exceed the maximum time delays indicated in the Chapter 15 Accident hnalysis. (3) Setting of first time delay is based on longest starting time of motor, and utilized for alarm at the main control board. Subsequent initiation of SOS will initiate automatic tripping actions. (4) Setting of second time delay is based on the maximum time the safety loads may operate without damage, down to the rating of primary relay settings.

TABLE-2 SE UENCE OF EVENTS UNDER DEGRADED GRID CONDITICH TIME BASE 1* (Seconds) TIME BASE 2** (Seconds) DESCRIPTION OF EVENTS Degraded voltage condition (6900 V Class 1E bus voltage at 89X), SIAS

Signal, DG start signal from SIAS, engineered

,safeguard equipment start signal from SIAS. Degraded voltage relays 27A-l, 2 & 3 dropout. In sequence per PSAR Table 8.3.1-2 10 15 15.05 15.11 15.15 0 I F 05 Safeguard motors started and running. Diesel generator at rated voltage and frequency. Degraded grid voltage relays 27A-l, 27A-2 and 27h-3 and its timers generate trip signals. Offsite source breaker to 6900 V bus lA-SA (1BWB) trips ~ Safeguard motors are shed. Diesel generator breaker closes. 25.00 25.00 10.00 10 00 (26.03) Emergency sequencer energizes and Block fl loads of engineered safeguard loads start. Emergency sequencer Block fl loads (charging/BHSI Pumps) starting.

'TABLE-2 (Cont 'd) TIME BASE l* ~ (Seconds) 30.,0 TIME BASE 2** (Seconds) 15.00 (55.0) DESCRIPTION QP EVENTS Emergency sequencer Block f2 load (RHR/LHSI and containment spray pump are starting. 35;0

20. 0 Emergency sequencer Block f3 loads are starting (service vater pumps)

40. 0 25.00 Emergency sequencer Block f4 loads (component cooling vater pump) are starting.

45. 0 30.00 (94.5)

Emergency sequencer Block $ 5 loads (Auxiliary feedwater pumps) are startiig. - 50.0

55. 0 30.00 40.00 Emergency sequencer Block f6 loads are starting+

Emergency sequencer Block f7 loads are starting.

60. 0

45. 0 Emergency sequencer Block f8 loads are starting.

• Time Base-l indicates length of time from accident occurrence concurrent vith degraded voltage condition.

~Time Base-2 indicates the length of time from accident occurence after it has been established that a degraded giLd voltage condition exists.

SE UENCE OF EVENTS UNDER LOSS OP VOLTAGE CONDITIO DESCRIPTION OP EVENTS ~ 05 Loss of voltage sensed by degraded voltage relays (89Z of below volts for. 60 sec.) 'or by loss of voltage relays (72Z or below for 0.5 seco) ~ 'Offsite feeder breaker to 6900 V buses LL-SA and 1BWB trips, start signal to diesel generator, load shed O.ll Safeguard motors are shed. 10.00 Diesel generator at rated voltage and frequency 10.04 Diesel generator breaker closes. 15.04 (26. 03) Emergency sequencer Block fl loads (charging/HHSI pump) starting. 20.04 25.04 30.04 Emergency sequencer Block f2 loads starting. Emergency sequencer Block f3 loads (service water pump starting).. Emergency sequencer Block f4 loads (component cooling water pump) starting. 35 04 (121. 8) Emergency sequencer Block $5 loads ~ (auxiliary feedwater pump) are starting. 40.04 Emergency sequencer Block f6 loads are starting. 45 04 Emergency sequencer Block f7 loads are starting. 50.04 Emergency sequencer Block f8 loads are s tarting.

• Time Base indicates length of time from loss of voltage occurrence or expiration of 60 sec time delay associated with a degraged voltage.

Position-1 (Cont'd) For the sequence of events tables presented

above, the tolerances on the voltage trip point and the time delay for the loss of voltage

relays, degraded voltage trip point and the time delay for the loss of voltage relays, degraded voltage relays, and associated timers were not included.

The relay voltage trip band willbe plus nr minus 1X of the calibrated trip point, and the time delay relay trip tolerances willbe plus or minus 3%; Due to the magnitude of the differences between the engineered safety features response times tabulated above and the maximum time delays assumed in the FSAR Accident Analyses, inclusion of the relay tolerances will not alter the conclusion stated above, i.e. maximum time delays assumed in the FSAR accident analysis will not be exceeded. The technical specifications will be revised to address instrumentation and actuation devices for protection from degraded offsite power conditions. The technical specifications will address limiting conditions for operations, surveillance requirements, trip setpoints and allowable values for sensors and time delay devices. (7660FXTpgp)

~ ~ TO TO LECT AUX STAAT M JhhRSf TAWN'rrm.earn t ELECTHCPkl.T~ IIO lCNI CLA5$ 1E NlS 1D ION 1E) ICN CLASS 1E CLASS 1E 4 CLASS 1E CNOOV mUS J~ IOIt1~ QANSF 72'/12OV ) ) ) IA ) Pl ~~ CA %5CXXII O NFS '%XIV,CONZ DIESEL CKMl~ LCR 1~ ItOTEs LSCERVtR.TACE INLAYS 2TWo2 L 5 HK IK~VXSa SCLATS LSCf.ltVOLTACE AfLAVS 2TA le 2 4 5 HK ITE-2TII ICLATS SHEARON HARRIS NVCLEAR POWER PLANT Carolina Power 4 L Jght CoepEIny FINAL SAFETY ANALYSIS REPORT UNDERVQLTAGE RELAYING 6.-9KV CLASS 1E BUS 1A"SA OR lB-SB FIGURE 438. 93-i

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TDC, 27 bACKUP RELAY SQUVR ReTY SA FAILURE OR SA 'A TDO LOW OF D.C TQC TDC.

powKR DEGRADED 2-2 LCS5 OF AC VOI.TACEE ~I~ (DKSVJOED VOLThtjE )'LARMS NOTE:

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SITE IEiT SA (RESET) rRESVi RE~mT /TEST OVERRIDE'~6ME '44E P RELY) S&UVS TDC SQUVT ~A C 27 LOSS OF D. C. TDC ~PA UV CIRCUIT ~SA TEST T (l=>C4i FC >~ 6, 2<i'52ESE ~ IA:5~ 4, IS-Sih Ia- +a) CAROLINA POSER 8 LIGHT CO. SHKARON HARRIS NUCLEAR P.R UNiT NO.I 4.9%V CWSS IE bi% JNQNYOLT/QE PROT. F6.430.95-T

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..Position-2 The load shedding feature of all safety related buses (6.9 kV, 480 V) is initiated by a bus uadervoltage condition (seased by loss of voltage or degraded voltage coincident vith accident initiation) and villautomatically strip the busses of theix'oads regardless of the source (offsite or onsite) supplying pover to the bus. The load shedding feature prepares the buses for traas fer from the offsi te pover system to an onsite diesel generator. The basic concern iahercnt in this BTP relates to the possibility of having the load shedding feature inadvertently activate duriag the load sequencing process. The Shearon Harris undervoltage protection system complies vith this position by blockiag the operation of the uadervoltage relays during sequencing of loads on to the diesel generator. (Refer to Pigure 430.93-2). Hovever, the load shedding feature is retained during sequencing of loads folloving aa accident vith offsite pover availablc and also after load sequencing is completed on to thc diesel generator. The goad shedding feature is retained for the above tvo conditions for the folloviag consideration. As shovn ia the attached Sketch 430 93-B1, during sequencing of loads on to the offsite source, at ao time dux'ing thc load sequencing do voltages decrease to a level vhere thc load shedding feature vill be actuated;

hovever, dux'iag or after the sequencing, if a malfunction in thc offsite pover supply circuit occurs.(such as iaadvertent tripping of offsite tie breaker to the safety bus), the load shedding feature will.protect the safety xelat'ed loads from damage by disconnecting the loads from the degraded system.

Secoadly, after load sequencing on to tbe diesel generator is completed, the load shedding feature is reiastated to protect safety related loads (motoxs) from effects of a malfunctions in the diesel generator cix'cuit, such as voltage regulator or excitation system failure. Since the initiating cix'cuit of the load shedding featux'e is derived from undervoltage relays arranged in a coincident logic (2 out of 3 logic), the possibility of an iaadverteat or spurious activation of load shedding is considered remote. Ia addition, an inadadvertent activation of the load shedding feature coastitutes a single failure and the plant can accept this single failure without safety implicatioas. Figure 430.93-3 shows the voltage profile of safety related 6.9 kV buses 1A-SA and 1B-SB during sequeacing of loads on to the dicscl generator. -As can be seen the maximum voltage drop (18.5Z) occurs vhen applying the 3rd block load and recovers within 2 seconds. The degraded voltage relays (27A-l, 27A-2 and 27A-3) vould dropout at that instant. Hovcver, the relays vould reset at 90Z voltage. Since the load shedding fcaturc is aot actuated uatil a time delay of 15 seconds has elapsed from tbe time the undervoltage relays have actuated,

"a sufficient margin exists between the bus voltage profile and relay set point during the loading sequence to prevent inadvertent actuation of the load shedding feature. Therefore, even if a single failure of the blocking feature is postulated the inadvertent actuation of load shedding feature villnot occur. The load sequencing ~ill be periodically tested, at least once per 18 months during shutdown," as required by the technical specification.

0 p O mm ~n ~x ~~M 27 l ~l3 cion+ 0 ~M Sl Q+ BUS VOLTAGE l SEE NOTES BELOW) 1 7 8 / 75% 70% 0 VOLTS .5 EC 8 L AUTOHATIC SEQUENCE TIHERS 5 10 15 20 25 30 35 40 SECONbS BLOCK "1 BLOCK BLOCK 12 13 BLOCK l4 BLOCK BLOCK 05 06 BLOCK BLOCK HANUAL +7 8 BLOCK 3 3 3 3 3 3 3 HAXH. VOLT. DROP (18. SX ZN BLOCK-3) VOLTAGE PROP AND RECOVERY ~ TI% IN SECONDS ne

6. 9KV SAFETY RELATED BUS VOLTAGE PROFILE DURING SEQUENCING OF EMERGENCY LOADS ON TO THE DIESEL GENERATOR D

C: CD Pl NOTESI 1) 2) 3) 4) 5) 6) 7) 8) NOMINAL VOLTAGE 6SOO VOLTS LOSS OF VOLTAGE DEGRADED VOLTAGE RELAYS t 27A io 2i 3) TYPE! ITE 27H DROPS OUT (AT 102V SECONDARY) LOSS OF VOLTAGE RELAYS (27-ii 2. 3) TYPE! NGV13B DROPS'UT t AT 83V SECONDARY) TIWRS ASSOCIATED WITH LOSS OF VOLTAGE RELAYS TIt% OUT AFTER LS SECONDS AND THE DIESEL ACCELERATES TO FULL SPEED WITHIN 10 SECONDS. t OPERATION OF TI&RS ASSOCIATEO WITH DEGRADED VOLTAGE RELAYS MERE NOT CONSIDERED SINCE IN THE TOTAL LOSS OF VOLTAGE EVENT, LOSS OF VOLTAGE TIHERS MILL TRIP PRIOR TO DEGRADED VOLTAGE TIHERS.). DEGRADED VOLTAGE RELAYS RESET WHEN VOLTAGE RECOVERS TO 91. 3% OF 6SOO VOLTS DIESEL GENERATOR BREAKER CLOSES. AUTOHATIC LOADING SEQUENCE COt+KNCE AND LOADS ARE APPLIED IN SEQUENCE SHOWN IN FSAR TABLE-8.3.1-2e

Position-3 The voltage levels at all safety related buses have been optimized and distribution transformer tap settings have been selected for the maximum load and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite po~er source. A computer analysis have been performed to obtain the tap settings of transformers and allowable voltage variations'at safety related buses. Transformer tap settings and bus voltages for various conditions are shown in attached sketches. The analysis consisted of;the following: I 'ase A: Determination of the worst case steady state condition. The postulated plant conditions considered for this case are: Unit auxiliary transformers lA and 1B aze out of service, Unit-1 is operating at full power with all required auxiliary loads connected to starts transformers 1A and lB, the 230 kV switchyard voltage is at minimum anticipated level. Case B: Determination of the ~orst case anticipated transient condition. Underlying assumptions for Case B are as follows: Transient condition subsequent to the steady state condition described in Case-A with (1) LOCA-resulting in sequential starting of LOCA Loads. r (2) Starting of 3000 HP Normal Service Water Pump Motor on the auxiliary bus 1D (or lE) coincident with LOCA load block 81. Case C: Determination of the anticipated light load condition and calculation of the maximum expected voltages at this condition. In addition to the above cases a separate case was studied to determine the minimum acceptable level of Class 1E 6900 V bus voltage (the setting of the degraded voltage zelays) necessary to maintain the normal operating voltages at all safety related equipment down to 208/120 V power panel. Postulated condition for this case are steady state plant full load with accident mitigating loads. The voltage results of computer analyses and associated loading of buses aze shown on the attached one line and voltage profile sketches. For each of the case 'studies two independent analysis are performed simulating each separate connection.to offsite souzce through start up transformer A and B. For example Case-Al reprsents steady state condition with all redundant loads connected to start up transformer A while Case-A2 represents steady state condition with all zedundant loads on start up transformer B.

~ +

. For.208/120 V system, the voltage drops were computed for the worst case motor control center with lowest voltage based on loading and length of cable. The calculated steady state voltages on the distribution system prior to transient are shown on sketches SK-430.93.h Sheets 1 6 2. The steady state motor terminal voltages are above 90Z of the continuous voltage rating of the safety related motors (i.e. equivalent to 86Z of the rated bus voltage i.e. 414 V and 5940 V-for 460 V and 6600 U motors respectively). Voltages at the power panel bus are also maintained above 90Z of nominal bus voltage of 208V. For steady state condition, the voltage time settings of the 6.9 kV Class 1E bus undervoltage (degraded voltage) relays have been selected to avoid disconnection of the safety buses from offsite source during plant running condition as described above. L ~. n The results of the computer analysis depicting the voltages at, the instant of simulaneous motor starting (Case Bl for starting of LOCA, loads and Case B2 for LOCA block 1 concurrent with starting of Normal Service Mater Pump (NSMP) at 6 9 kV bus 1D) are shown on sketches 430.93-81, SHl and 430.93-32, SH 1 6 2. As shown on voltage profile sketches the bus voltages are above the minimum voltage required to maintain minimum motor starting and accelerating voltages of 75Z of motor rated voltage. The 480 V MCC bus voltages aze above the minimum required voltage for Starter coil pickup voltage rating. The voltage time settings of the 6.9 kV bus undervoltage relays (loss of voltage) have been selected to avoid tripping of the safety buses from the offsite source during the transient conditions postulated. Calculated values of voltages at distribution system buses during light load condition (CaseM) are shown on Sketch 430.93-C, SH 1 6 2. This case postulates the unit is in refueling mode with all lighting, HVAC, fuel handling system and security system loads running and 230 kV swicchyard voltage at the maximum expected level of 1.04 p.u. at the 226 kV base. As can be seen from Sheet-2 of Sketch 430.93&, the bus voltages are maintained below 105Z of nominal rating corresponding to 110Z maximum'llowable motor terminal voltage. The result of an analysis showing voltages at various safety related distribution buses for the minimum dropout settings of the bus undervoltage relays (degraded voltage) at th'e 6.9 kV safety bus are shown on Sketch 430.93-9. It can be seen that voltages at all safety related buses are above the minimum required steady state voltage for safe operation of all safety related loads. Based on the results of the above described analyses, it is concluded that the Shearon Harzis station electric distribution system is of sufficient capacity and capability to automatically start and operate all required safety loads within their zequired voltage ratings for the anticipated transient and steady state conditions assuming only off site sources of AC power are available. 12

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4. L tight Coepany CASE B2 YOLTAGE PROFILE HORST CASE TRANSIENT COND.

LOCA BLOCK-1 h NSWP STARTING SKETCH 038. R3-B2 ZT82~

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qr.

Quality Assurance Branch/R. Kirkwood Open Etem 220 {Partial)

Shearon Harris Nuclear Power Plant Draft Safet Evaluation Re ort 0 en Item No. 220* The NRC requested additional information relative to Table 3.2.1-1 of the Shearon Harris FSAR as described below: 1. Table 3.2.2-1, Page 3.2.1-5 Verify'hat the containment air locks, equipment hatch and valve chamber are in accordance with Section III, Article NA-5000, of the ASME Boiler and Pressure Vessel Code. 2. Table 3.2.1-1, Page 3.2.1-4 Verify that the containment liner is in accordance with Section III, Article NA-5000, of the ASME Boiler and Pressure Vessel Code. 3. Table 3.2.1-1, Pages 3.2.1-11 through 13 Components of the letdown and makeup loop of the chemical and volume control system are incorrectly classified Quality Group C. These components which are identified in the attached marked-up sheets should be classified Quality Group B and the reference to note (4) deleted. The Safety Class and component code class are acceptable. 4. Table 3.2.1-1, Page 3.2.1-11 The mixed bed demineralizer, cation bed demineralizer, and the boric acid blender of the chemical and volume control system are incorrectly classified Quality Group D. These components should be classified Quality Group C and the reference to note (4) deleted. The Safety Class and component code class are acceptable., 5. Table 3.2.1-1, Page 3.2.1-12 The Quality Group classification of the boric acid batching tank of the chemical and volume'control system has been omitted. This component should be classified Quality Group D. 6. Table 3.2.1-1, Page 3;2.1-12 The Quality Group classification of the RCP seal bypass orifice of the chemical and volume control system is incorrectly classified Quality Group C. This component should be classified Quality Group A and the reference to note (4) deleted. The Safety Class and component code class are acceptable. 7. Table 3.2.1-1, Page 3.2.1-13 The Quality Group classification of the piping and valves of the chemical and volume control system normally or automatically isolated from parts of the system covered by a, b, or c has been omitted. These components should be classified Quality Group D.

• Note:

The response to the first half of this question was transmitted by our letter dated August 11, 1983.

~ 0

'able 3.2.1-1, Page 3.2.1-15 The Quality Group classification of the recycle evaporator reagent tank of the boron recycle system has been omitted. This component should be classified Quality Group D. 9. Table 3.2.1-1, Page 3.2.1-17 The Quality Group classification of the BIT recirculation

pump, boron injection surge

tank, and the boron injection flush orifice of the safety injection system has been omitted.

These components should be classified Quality Group C and the reference to note (4) deleted. The Safety Class and component code class are acceptable.

10. Table 3.2.1-1, Page 3.2.1-18 The Quality Group classification of the piping and valves of the safety injection system required for supply of boric acid to BIT has been omitted.

These components should be classified Ouality Group C and the reference to note (4) deleted. The Safety Class and component code class are acceptable. Table 3.2.1-1, Page 3.2.1-21 The component code class of the system piping and valves connected to penetrations of the containment penetration pressurization system has been omitted. These components should be classified ASHE Section III, Code Class 2.

12. Table 3.2.1-1, Page 3.2.1-21 The Quality Group classification of the WPB cooling pumps, heat exchanger (tube and shell side) and the piping and valves of the waste process building cooling system has been omitted.

These components should be classified Quality Group D. 13 ~ Table 3.2.1-1, Page 3.2.1-23 The Quality Group classification of the delay coil of the sample system is incorrectly classified Quality Group D. This component should be classified Quality Group B as it is a part of the reactor coolant pressure boundary and the reference to note (4) deleted. The Safety Class and component code class are acceptable.

14. Identify those components in Table 3.2.1-1 to which note (8) is applicable.

15. In addition to Code Cases 1528 and 1355 identified in Section 5.2.1.2 of the FSAR, identify all other ASHE Code Cases (including those that are listed as acceptable in Regulatory Guides 1.84 and 1.85) that were used in the construction of each Quality Group A component within the reactor coolant pressure boundary.

These code cases should be identified by code case number, revision, and title, for each component to which the code case has been applied. 16. ASHE Section III, Division 1, Subsection ND, Class 3 Components, is identified in FSAR Section 3.8 (Page 3.8.1-7) as applicable to components of the containment structure. We find this application of Subsection ND for containment components to be unacceptable. Provide a list of all

A containment components that are constructed to Class 3 of the Code. To be acceptable, containment components should be constructed to Subsection NC or Subsection NE, Class 2 Components. ~Res esse 1. The specification for the containment air locks, equipment hatch, and valve chamber requires compliance with the ASHE Winter 1975 Addenda, Nuclear Power Plant Components, Division 1, Section III, Subsection NA. 2. It is assumed that the NRC intent was for CPSL to state adherence to ASME BGPV Code, Section III, Subsection CA, Article 5000. FSAR Appendix 3.8A, Page 3.8A-1, provides a statement of exception to the requirements of Subsection CA, Article 5000. Compliance with ASHE MPV Code is discussed in Appendix 3.8A of the FSAR. 3. Components in the CVCS forming the high head SIS are classified Quality Group "B" (i.e., Refueling Water Storage Tank, Charging

Pumps, and Injection Header).

The remaining portion of the CVCS forming the letdown and excess letdown paths up to the suction of the charging pumps should be classified Quality Group "C" (Q.G."C") for the following reasons: (1) In accordance with FSAR Section 7.4.1, the excess letdown and normal letdown paths are not required for safe shutdown of the reactor to the cold shutdown condition. s (2) Upon initiation of "S" signal, these letdown paths are automatically isolated. (3) These portions of the CVCS do not meet the criteria of Q.G."B" classification as specified in Regulatory Guide 1.26. (4) The RCP seal water supply/return paths are part of the CVCS, and they are specified in Regulatory Guide 1.26 as a Q.G."C" system. The mere fact that reactor coolant is cycled through the CVCS does not necessitate a Q.G."B" classification. (5) Standard Review Plan, Section 3.2.2 does not provide additional guidance on the CVCS classification. A Q.G'"C" classification of these portions of the CVCS provides adequate assurance of their reliability. This interpretation is in keeping with published guidance. 4. The mixed bed, cation bed, and boric acid blender should be classified Quality Group "D". These demineralizers are sized to clean up fission products introduced during normal plant operations (1% failed fuel), and they would be little help under major accident condition if substantial fuel damage occurred. Additionally, these components are isolated from the RCS upon receipt of an "S" signal. A Q.G."D" classification is in keeping with published guidance and provides an adequate safety classification of these components.

5. 6. The Boric Acid Batch Tank is classified Q.G."D". FSAR Table 3.2.1-1 will be revised in a later FSAR amendment. The RCP Seal bypass orifice is classified Q.G."A". FSAR Table 3.2.1-1 will be revised in a later FSAR amendment. 7. These components will be classified Q.G."D". FSAR Table 3.2.1-1 will be revised in a later FSAR amendment. 8. The recycle evaporator reagent tank will be classified O.G."D". FSAR Table 3.2.1-1 will be revised in a later FSAR amendment. 9. Two redundant heat-tracing lines on the BIT and BIT recirculation system are designed to maintain system temperature above the solubility point for 12 wt. % boric acid solution (FSAR Section 6.3.2.2.2). The BIT is continuously recirculated to maintain a uniform temperature and concentration of solution. Upon receipt of an "S" signal, the recirculation path is automatically isolated and is not required to operate again. System parameters are monitored on the HCB and indicate recirculation/system status. Additionally, the BIT and recirculation system fall under the Technical Specification LCO surveillance program. The classification of the BIT recirculation system as Q.G."C" is unnecessarily conservative, and it does not fall under the Q.G."C" criteria of Regulatory Guide 1.26. Quality Group "D" is a more appropriate classification of this system. 10. The addition of Boric Acid via the BIT flow path is a one time operation in the event of a plant casualty. Addition of Boric Acid to the BIT via the BIT recirculation system is prohibited following an "S"'ignal since the BIT recirculation system is isolated. An existing flow path for providing Boric Acid to the reactor vessel is already classified Q.G."C" (i.e., CVCS Boric Acid Tanks to H.P. charging pumps). It is unnecessarily conservative to classify a second system (i.e., Boric Acid addition to BIT) as Q.G."C". The components listed in Item 10 should retain a Q.G."D" classification. FSAR Table 3.2.1-1 will be revised accordingly in a future amendment. ll~ These components are classified Class 2 in accordance with ASME Code Section III. FSAR Talbe 3.2. 1-1 will be revised in a later FSAR amendment. 12. The WPB Cooling System should not be a contaminated system. Site service water is the heat sink for this system. The WPB cooling pumps, heat exchanger (tube and shell side) and piping/valves should not be classified Q.G."D". They should remain unclassified. 13'here is not delay coil in primary sample system. All piping and valves inside of containment are classified as Q.G."B". All items outside the second isolation valve on each containment penetration is classified QeGo"D" ~

14. Note (8)'pplies to the reactor coolant system auxiliary piping in Table 3.2.1-1. 15. See Attachment l. 16. No components were procured to Subsection ND, ASME Section III, Division 1. Page 3.8.3-7 of the FSAR will be revised to delete reference to Subsection ND and replace with Subsection NC, Class 2 components. The applicable components for mechanical, electrical and instrumentation and controls have the appropriate codes and standards addressed in their respective sections of the FSAR. FSAR Sections 3.2 and 3.8 will be revised in a future amendment to reflect these changes. (7659FXTccc)

ATTACkRKNT 1 UNIT 1 CODES CASES FOR ASME CLASS 1 EQUIPMENT ~Eni ment Code Case BMI Tubing and Coupling Reactor Vessel None 1401 CRDM Steam Generator RC Pumps Pressurizer RC Pipe RTD Bypass Manifold Valves (Class 1) Copes Vulcan W EMD Fisher Control Crosby None

1484, 1493-1, 1355

1528, 1493 1423-1, 1423-2 None 1388-1$

1649 1553-1, 1649

1501, N-3-10 None No Code Form Available as of 6/1/83

ATTACHMENT 1 (Cont'd) UNIT 2 CODE CASES FOR ASME CLASS 1 EQUIPMENT ~Eui ment Code Case BMI Tubing and Coupling Reactor Vessel tm e None None CRDM Steam Generator RC Pumps Pressurizer RC Pipe RTD Bypass Manifold Valves (Class 1) Copes Vulcan W EMD Fisher Control

1484, 1493-1, 1355

1528, 1493 1423-1, 1423-2 None 1649 1553-1, 1649 None Crosby None No Code Form Available as of 6/1/83

TABLE 3,2 ~I-I (Continued) CLASS IF ICATION Of STRUCTURES SVSTEMS ANO COI4%tIENTS S stems and Com onents Safety Class ( I ) Code Oosl n and Construction Codo Solsmlc Class ~Cata or t21 0 oratlons Qua I Ity Qa I Ity Qal Ity Group Assurance Assurance (3) (23) (24) Ross ras b) From tho MS IV up to and 3 Including the last selsmlc restraint In the Turbine Bu I I d I ng c) Oovnstroam of last solsmlc NHS rostralnt In Turblno Bulldlng d) Operators tor Safoty-Rolatod IE Active Valves Soe tbto Seo Hate (I6) (I6) ANSI B3l ~ I IE Soe Note (4) Q Soo Moto (3l) 3 Instrumentat lan IE IE Q See Note ( l5) Steam Generator Blovdovn S stem System Plplng and Valves a) From steam qenurator to 2 ASME I I I and Including containment Isolation valvos b) From containment Isolation 3. ASME I I I valves.to RAB wal I Seo Note (4) Condensato and Feedvator S stem Condensato and Feedvater Pumps Electromagnetic Filter NNS HNS ASME VIII (27 ) R Seo ate (21) (kavqs'k 5('> Sw>>o g ~(.un1i~

TABLE 3,2'-I (Continued) CLASSIFICATIOH OF STRUCTURES SYSTEMS ANO COIPOt(ENTS Oosl n and Construction eratlons Rsaaras S stems and Cnm onents Safety Class ( I) Codo Qua I Ity Qua I Ity Code Solsmlc Quality Group Assurance Class ~Cata or (21 Assurance (31 (231 (F 1 Fire Protection S stem NNS Seo Ibta (Il) - Soo Ibto (IS) Nitro on Su I S stem NNS H dro on Su I S)(stem Fuel Transfer Canal Liner Shl In Cask Pool Liner Reactor Cavlt Liner NNS HA NA Q Soe Note (2l) Q Seo Ihte (21) 0 Seo Note (2l) Reactor Auxl I Iar Dul I din lhcontamlnatlon Liner NHS FHB Cask Mashdcwn Area Liner Q Seo Noto (21) Hain Steam S stem SystemPlplng and Valves a) From the steam generator up to and Including tho. HSIV; a II branch connections from this suction up to and Including tho first normally closed or automatic closuro shutoff volvo (This Includes safety valves and HS PORVs) 2 ASME I I I Crr))~((r, ((( gr D Ses "S R((aa g.bl.l Sqbsn

TAB.E 3,2 ~I-I (Cont lnuod) CLASSIFICATION OF STRUCtURES SYSTEMS ANO COI4'ONEHTS Desi n and Construction 0 oratlons Rssarks S stems and Com ononts Safety Class (I ) Codo 0ua I Ity 0ua I Ity Codo Selsmlc 0uallty Group Assuranco Class ~Cate or (2( Sssoraacs (31 (23( (241 Hydrogen Rocomblnor (Catalytic) HNS ASME III SystomPlplng and Valvos a) tht normal ly or automat Ical ly Isolated from SC-3 component b) Othor ASME I I I NHS 831 I So I I d Waste Process ln S stom t(HS Soo thte (26) Soo Hate (26) a Soe Note (27) Contalnmont Coolln S stem Contalnmont Fan Coolers a) Fans and Caslnqs b) Supply Fan Motor c) Cool lng Col ls d) Ouctt(ork and dampors up to concrete alrshatts o) Ductwork a<<d dampors downstream of concrote alrshafts ~ 2 IE 2 2 NNS ASME I II 8 8 8 8 IE 8 0 0 0 0 Contalnmont Fan Col I Unl ts. NHS Instrument Ion IE IE 0 Soo !ate (15) Contalnmont Ventl lotion S stem Alrborno Aadloactl vlty Romoval NNS Systom A Soo Note (18)

TABLE 3 2,l-l Cont lnuod) CLASS IF ICAT ION OF STAUCTINES STSTEMS ANO COMPONENTS S stems nnd Com onents Safety Class ( I) Code Oosl n and Construction Code Sel sml c eratlons ()jn I Ity ()ua I Ity Qallty 0 oup Assurance Assurance (3) (23) (24) Roeal'k5 Now Fue I Assembly Handling Fixture NNS Now Rod C lustor Control Handling NNS Fixture Lower Internals Storage Stand NIIS Upper Interne ls Storage Stand NNS Load Ce I I Linkage Spent Fuol Storago Racks Rofuellng Cavity Seal Ring lnstrumontatlon LI uld Haste Processln S stem NNS NA IE NNS Soe Note (25) See Note (25)- IE ~ R 4 See fbte (l5) Reactor Coolant 0 nln Tank Pump NNS ASME II I 3 ~ R Reactor Coolant. Oraln Tank Hoot Exchnngor (shel I s I do) 2 ASME III (tubo sldo) NNS ASME VI I I B B O ~ ~ R System Pl ping 4 Va Ives n) Ibt normo I ly or automat Ice I ly Isolated from SC-3 components b) Other ASME III NNS B31 I C P R Gaseous Mnste R ocossln S stem ~ Gas Compressor Gns (@cay Tank NNS ASME I I I 3 ASME III 3 See Ibto (I3) 5-R 4-R

TABLE 3,2 ~I-I (Continued) CLASS IFICATION OF STRUCTUAES SYSTEHS AND COl4%NENTS S stems and Com onents Safoty Class (I) Codo Desi n and Construction Codo Selsmlc Claaa ~Cato or (2t oratlons +a I lty gal Ity gual lty Group Assurance Assurance (3) l23) (24) Remarks Service Mater Booster Pump Hotors IE IE 4 i3 Travo I lng Mater Screens Systom Plplng and Valvos a) Required tor operation of Contalnmont Fan Coolers (Inside contalnmont) b) Required for performance of other safety functions c) tlormally or automatically Isolated from parts of systom covorod by a) d b) above d) Operators for'Safety-Aolatod Act Ivu Valves I Ins trumuntat Ion 2 ASHE I I I 3 ASHE I I I NNS ANSI B3I ~ I IE IE C IE 0 IE 4 Soe Note (3l ) Soo Noto ()5) ~ao 1l I ~ 5 aatoa Sample Heat Exchanger (tube side) NNS (she I I s lde) NNS D 0 Sample Vossol NNS oo-aota-tat P~j ra Gross Fa I lod Fuel Detoctor (GFFD) NNS GFFD Sarnp lo Cooler 3 ASHE I I I Soo Note (4)

TABLE 3.2 ~ I-I (Cont lnued) CLASSIFICATION OF STRUCTINES SYSTEHS ANO CO)4 ONENTS S stems and Com onants Safety Class (I) Coda Oasl n nnd Construction Codo Sais<<lc CCIass ~Cato or I2I orations Quil Ity Qnl Ity gal Ity Group Assuranco Assurance (3) (23) (24) Raaarks System Plplng and Valves a) Not normally or - nutomatlcnll'y Isolated tram safety class components b) Other 3 ASNE III NNS ANSI B3l ~I Soo Note (7) 0 ~ R Seo Note (4) Safat In ection S stem Accumulators Boron InJectlon Tank (BIT) BIT Hoclrculatlon Pump Boron InJactlon Sul go Tank Boron InJactlon F lush Or It Ice 2 ASHE I I I 2 ASNE I II ASHE I I I ASHE I I I 3 ASHE I II Soe Note (7) See Note (4) Soe leyte (4) 2 otIot Ist llydro Test Pump NNS Systa<<Plplng and Valvos a) Port ot RCPB I ASHE III A P

TABLE 5,2.1-1 (Continued) CLASSIFICATION OF STRUCTURES SYSTEHS AND COB'ONENTS S stems and Cam onents Sa toty Class (1) Codo Oosl n and Construction Code Selsml c Class ~oats ol'tt 0 eratlons Otal Ity Qua I Ity gua I Ity Group Assurance Assurance (3) (23) (24) Raaarks Recycle Evap, Feed Filter ASHE III 3 Soo lhto (1) Soo Nota (4) 3 Recycle Evap. Condensate Demlnera I Izor NNS ASHE VI I I Recycle Evan, Roagent Tank NNS ASHE Vill Recycle Ihldup Tonk Vont EJector 3 ASHE III Rocyclo Evap, Condensato F I ltor NNS ASHE VI I I 3 Soo thte (1) See tbto (4) 3 v: a4'0'u Recycle Evap, Concentrate Filter NNS ASHE III Recycle Evaporator Package a) Feed Prehoater, I) Feod Side

2) Steam Sldo b) Gas Stripper c) Submerged Tube Evap, I) Feed Side

2) Steam Side d) Evaporator Condenser I) Olstlllate Sldo

2) Cooling Mater Side ASHE Ill 3

Soo Note (7) NNS ASHE VIII ASHE I I I 3 ASHE I I I I 3 Soo Note (7) I 3 ASHE I I I 3 Soo Note (7) NNS ASHE VIII ASHE III 3 Seo Note (7) Seo Note (4) Seo Noto (4) R Seo Note (4) See Noto (4) Seo thte (4) 3 a

TABLE 3.2'-I (Continued) CLASSIFICATION OF STRUCTURES SYSTEMS ANO CO)4'ONEKTS S stems and Com ononts 'afety Class ( I ) Code Desi n and Construct)on Code Selsmlc Risks ~Cbt8 ol (2t eratlons Qual Ity Qua I Ity (}UaI Ity Group Assurance Assurance (3) (23) (24) Rsssrks System Plplng and Valves a) Part ot RCPB b) Required for reactor coolant letdown and makeup c) Requlrod for provldlng boric acid for the letdoun and makeup loop d) Normally or automatically Isolated from parts of system covered by a, b or c I ASHE I I I 2 ASHE I I I 3 ASHE I I I NNS ANSI 83l ~ I r A ~ 0 C 0 C 0 Soo Ihte (4) Instrumentatlon Operators tor Satqty&elated Activo Valves IE IE IE IE Q Soe Note ( l5) 0 Seo Note (31) Boron Thermal Ro eneratlon Subs stem Hoderat lng IIX (tube s ldu) (shol I side) ASME I I I ASHE I I I 3 - Soe Note (7) 3 Seo Ihte (7) 8 B R R Seo Nota (4) Seo Nato (4) Letdovn Chl I lor NX (tube s ldo) 3 ASHE I I I (sholl side) NNS ASHE Vill Soo Noto (7) Seo Note (4) Lotdovn Reheat NX (tubo side) (shel I side) Thermal Regeneration Demlnerallzor 2 ASHE I II 3 ASHE III 3 ASHE III 2 I 3 Soo Ibte (7) Soo Note (7) C D 0 R Soe Nato (4) Seo lhte (4) Seo Noto (4) Chl lier Pump NNS

TABLE 5.2 ~I-l (Continued) CLASSIFICATION OF STRUCTURES SYSTEHS ANO CORI ONENTS Oesi n and Construction eratlons Rsearks S stems and Cpm onents Satety CI ass (I ) Cede Code Class Sel sml c CatetaCot I (21 Qa I Ity ()ua I Ity gua I Ity Group Assurance Assurance (3) (23) (24) Excess Letdown NX (tube side) (shel I side) Seal Mater NX (tube side) (shell sldo) 2 ASHE III 2 ASHE III 2 ASHE III 3 ASHE III 2 2 8 8 8 8 C 8 C C See Note (4) Soe Note (4) Chemi ca I Hlxlng Tank Chemi ca I H Ixlng Tank Or I I Ice Ebron Heter NNS ASHE VIII NNS NNS ANSI 83) ~ I Boric Acid Tanks fhrlc Acid Fl lter I Boric Acid Transfer Pump 3 ASHE I I I 3 ASHE I I I 3 ASHE I I I Ihrlc Acid Transter Pump Haters IE 8 IE Boric Acid Botching Tank Reactor Coolant Pump ACP) Standpipe NNS ASHE V I I I NNS ASHE VIII RCP Standpipe Or I IIce RCP Seal Bypass Orl flee N)IS I ASHE I I I 0 R aA p See ate (4)

TABLE 3,2,1-I (Cont lnued) CLASS IFICATIOH OF STRUCTlHES SYSTEHS AHD CONPOHENTS ~Sstees end Conaononts Sataty Class ( I) Coda Ihsl n and Construction Code Se I sssl c Close CatolnoCO I21 0 eratlons Oua I Ity Qual Ity Otal Ity Group Assurance Assurance (3) (23) (24) kneel'55 Steno Generator Forglnq Typo A I ASHE I II () Soo Ibte (9) Chetsslcal L Volume Control S step Reqonoratl ve HX Letdo<<n IIX (tube side) (shell side) 2 ASNE I I I 2 ASNE I II 3 ASNE I I I C C 0 0 See Hoto (4) Seo Ibto (4) Hived Bod Datslnerallzor Cation Bed Dealneral lzor Reactor Coolant Fl ltor Vo lumts Contrnl Tqnk Charqlng (lllqh Head Satoty InJectlon) Pusttps ASHE II I Soe lbto (7) 2 ASHE I I I 2 2 ASNE I I I 2 2 ASNE I I I ASHE I II 3 See Hate (7) D .Q9, See Ibto (4) See Note (4) O Soe Ibto (4) Soe Hoto (4) Charging Pultp Notors Sea I Water InJect !on F I Itor Seal Mater Return F liter Boric Acid Blendor Lotdo<<n Or I t lees IE 2 ASHE I I I 2 ASNE I I I 3 ASHE I II 2 ASNE I I I IE Seo lbte (4) C 4 See Hoto (4) 0 III g Seo Rote Itt gC O See ~o-E (>)

TABLE 3,2,l-l (Continued) CLASS IFICATIOM OF STRUCTINES SVSTEMS ANO COl4 OHEHTS S stems and Co ononts Safety Class (I) Code Desi n and Construction Codo Selsmlc Class ~Cata or tst eratlons Qua I Ity Qua I Ity (Iua I Ity Group Assurance Assurance (3) (23) (24) Raaarks Roactor Coolant t4t and Cold Leg P I p lng, F Ittlngs and Fabr Icat Ion ASHE III I I Surge Pipe, Spray Pipe, Flttlnqs, and Fabrlcatlon ASME I I I I Seo Moto (5) Crossovor Log Plplng, Flttlngs I and Fabrlcatlon RTO Bypass Hanlfold Pressur I zer Safoty Valves Pressurizer Poser Operated Relief Valvos and. Black Valves ASHE III I ASHE I I I I ASHE III ASHE III (n a4 0'0 Valvos ot Safety Class I to Safety Class 2 Intortaco ASME I II I Pressurlzor Rol lot Tank HHS ~ ASME VlII Reactor Coolant Thurmovell AuxlllaryRoactor Coolant Plplng (Drains, etc.) I ASHE I I I I ASHE I I I 2 A Q XS Q Pressurizer Rol lot Valve Discharge I Lines (batuoon Prossur I zor lbzz lo and Relief Valve Only) ASHE I I I I

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