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Forwards Design Review Studies Re short-term TMI Lessons Learned Task Force Requirements.Proposes Mods to Plant Sys. Nine Oversize Tables & Two Oversize Diagrams Encl
	ML19257A504
Person / Time
Site:	Peach Bottom
Issue date:	01/02/1980
From:	Daltroff S; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001040494
	Download: ML19257A504 (55)
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PHl!.ADELPHIA ELECTRIC COMPANY 23O1 M ARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 SHIELDS L. DALTROFf'

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ELactacPm uctom January 2, 1980 Re: D ock et N os.

50-277 50-278 Harold R.

Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Design Review Studies Required by Short Term Lessons Learned - Peach Bottom Nuclear Power Station

Dear Mr. Denton:

Short Term Lessons Learned (NUREG 0578) identified design reviews of specific plant safety systems to be p erf ormed by each licensee.

We have completed the category A review items, except for two aspects of the plant shielding review (2.1.6a) which will be completed by January 31, 1980.

The reviews compared existing plant design with the p os t accident criteria providad in the NRC clarification letter of October 30, 1979.

We are hereby transmitting the results of these studies in the following enclosures to this letter.

1687 194 :

Emergency Power Supply (2.1.1) :

Containment Isolation (2.1.4) :

Dedicated Hydrogen Control Penetrations (2.1.5a) :

Reactor Coolant Leakage Control Program (2.1.6a) :

Plant Shielding (2.1.6b) :

Post Accident Samp ling Capability (2.1.8a) p% :

Interim High Range Effluent Monitors (2.1.8b) :

Technical Supp ort Center (2.2.2b) g.5/

Additionally, several p rop os ed imp rovements to plant systems are identified.

These modifications will be completed by January g. g g 1,

1981 in accordance with the NRC implementation schedule YW'

(,0 c.

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8001040

'WW ' i ^

L Mr.

H.

R. Denton Page 2 presented in NUREG 0578.

The prop osed modifications are as follows:

1.

Provide diverse automatic isolation signals to the radioactive gas s amp le line is olation valves.

2.

Install additional containment penetration isolation valves on the ins t ru me nt nitrogen suction line and the radioactive gas sample U.nes.

3.

Provisions will be made t o t ak e liquid and gas samples in an area accessible during p ost accident conditions.

4.

A permanent Technical Support Center that is habitable during p os t accident conditions will be established.

We trust this letter is responsive to your requirements.

If additional inf ormation is necessary, please advise us.

Very truly yours,

/

O kW

'J Enclosures 1687 195

D ock et Nos. 50-277 50-278 Item 2.1.1, Emergency Power Supply In the November 14, 1979, NRC (D.

G.

Eisenhut) letter to the GE BWR Owners Group (T.

D.

Keenan), we were advised that the Owners Group position on this item was acceptable provided that emergency power was available to p rovide a long-term s ource of air for air-operated relief valves.

We have reviewed the Peach Bottom Atomic Power Station design and verified that the instrument air compressors are p owered f rom the emergency diesels.

f

. ~...

1687 196

ENCIASURE 2 Itan 2.1.4, (bntalment Isolaticn A detailed review of the contalment isolation provisions at Peach Bottczn has been c:xopleted. This review fulfilled the design review requiranents of NURIT,-0578, Section 2.1.4.

The following is a h<rription of the degree of ocmpliance with thn fmr cxantalment isolaticm criteria given in NUREG-0578:

1)

Diversity of Isolation Signals - (2:ntalment isolaticn valves currently provided with autcznatic closure logic receive isolaticn signals as indicated on the attached table. These valves receive diverse, safety-grade isolaticn signals in accordance with SRP 6.2.4, with the following excepticns:

a)

FWQJ sucticn and return line isolaticn valves (70-12-15, 18, 68) are currently provided with the following signals to initiate valve closure:

Beactor Iow Water Ievel (0")

PHCU High Flow (300%)

BWCU mn-Regen. HX High Ttstp. (200'F)

Standby Liquid Cbntrol Systen operatico The latter two are process signals rather than ccm-taiment isolation signals. These isolaticn signals are provided to sv 7 the core in case of a possible break in the Peactor Water Cleanup Systan, to protect the ion exchange resin frm damage due to high tatperature, and to prevent the raroval of baron by the icn exchange resin.

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The RWCU systan is described in FSAR Secticn 4.9.

Closing times of the IECU isolaticn valves have been chosen in order to prevent the reactor vessel water level frm falling below the top of active fuel if a break were to occur in any of the FHCU lines. Diverse isolaticn signals are supplied to isolate the PWCU in the unlikely event of such a line break. The system is intentionally left in service whenever the abwe isolation signals are not activatr_3 in order to provide continuous purification of a portion of the recirc-ulation flow.

b) 1he HPCI and RCIC turbine exhaust line drain isolation valves (10-4240, 4241, 4247, 4248) are presently closed upon trip of their respective turbines. These drain lines are pro-vided to eliminate the accurtulaticn of cendensate and the consequent piping loads due to water slug carryover. Since 1687 197

\\

f the drain lines are provided to increase the availability and ra. liability of their respective systans, tney are not isolated under IDCA cmditions.

Instead, the isolation valves in these lines are closed when their respective systan has otxtpleted its function.

c)

Isolation valves in the vacuum breaker lin2s which connect to the HPCI and RCIC turbine exhaust lines (KN244, 4244A) are not provided with diverse isolaticm signals.

'Ibe vacutan breaker linas were provided to eliminate transient conditicms in the turbine exhaust lines during systen shutdowns. Withcntt these vacutzn breakers, pressure fluctuaticms may occur which cause slanning and cycling of the exhaust line check valves, severe pipe and torus vibrations, and water alty carryover.

Closure of these valves is currently initiated by a ocnbination of low reactor pressure and high drywell pressure.

^

'Ihus, these vacutzn breaker lines are isolated following a IDCA when their respective systans have ccatpleted their functicn.

2)

Classificaticn of Systans - All lines penetrating primary ccmtainment have been reviewed and classified as essmtial or non-essential. Essential lines are defined as those essential to w=psy reactor shut &un, reactor core cooling, ccntaiment heat rmoval and post-IOCA cxxobustible gas control.

'Ihe claaaificatim of each isolaticn valve is indicated on the attached table.

3)

Isolation of Non-Essential Lines - ttn-essential lines are presently provided with autanatic isolation valves which are closed either by contalment isolation signals or reverse flow (i.e, check valves, excess flow check valves), with the follcaing exceptions:

a)

'Ihe TIP drive line isolation valves are not provided directly with autcznatic isolation signals. 'Ihe TIP's are autctnatically withdram on Beactor Iow Water Ievel (0") or High Drywell Pressure (2 psig). Each drive line is then isolated by a ball valve which is autanatically closed cm TIP position.

7. shear valve in each drive line provides isolation in tFa event that a ball valve fails to close or a drive cable fan a to retract. 'Ihe shear valve is operated fran the control rocxn. 'Ihe TIP systen and its isolatico provisions are further described in FSAR Secticns 7.5.9 and 5.2.3.5, respectively.

'Ihe above descrihxi arrangemsat of isolaticn valves and signals meets the intent of the NFC design requirenents.

1687 198

t b)

LPCI and (bre Spray test valves (AO-10-163A, B and AO-14-15A, B) are not provided with autcxmtic isolation. These 1" valves are located in parallel with the check valves which provide inboard isolation in the IPCI and Cbre Spray injection lines. These rnrmally-closed, fail-closed valves are operated cnly to equalize pressures to permit stroxing of the check valves. These valves are opened oy a push bottm which utilizes a nmentary omtact to open the valve for testing. The existing design meets the intent of the NIC design requirements.

c)

The IIRT instrument omnecticms and Service Air supply lines are provided with dual locked-closed, manual globe valves for isolation. We do not feel that it is reaanry to provide isolaticn signals to these valves.

1his is in accordance with 10 CFR 50, Appeniix A, GDC 56 and SRP 6.2.4 Acceptance Criteria 3.f.

d)

The isolaticn valves cm the Reactor Building Cboling mter (10-2372, 2374) and Drywell Chilled Water Systms (M>-2200A, B and 2201A, B) do not receive autcmatic isolaticn signals. It is our pocition that these valves should not be autmatically isolated since their mntinued use will tend to mitigate the ccnsequences of an accident.

In additim,10 CFR 50, Appendix A, GDC 57 allows the use of a remote-manual valve on lines such as these that are neither part of the reactor coolant pressure boundary nor emnected directly to the ocntainment abughu.e.

e)

The PMim-tive Gas Sanple line isolation valves (SV-4966 A-D) are not provided with autcmatic isolation signals. These valves are normally closed but may be opemi by the operator during cxmtainment isolaticn to measure radiaticn levels in contalment. Modifications will be made to provide diverse autcxmtic isolaticn and keylocked bypass during the first scheduled outages of each tmit in 1980, subsequent to the availability of equipnent. Tb assure that these valves remain closed prior to inpleentation of this modification, the fuses providing power to these valves wil.1 be rmoved. The fuses will only be reinserted under shift supervisor cxmtrol to obtain containment radiation measurments during cmtainment isolaticn.

4)

Reset of Isolation Signals - We have ccupleted a design review of the control systms for all autmatic isolation valves to assure that resetting their isolation logic will not result in autmatic reopening of the valves. The regtured logic changes wre ocupleted on Unit 3 during an outage which began on Dec e ber 6, 1979. Similar nodifications will be made on Unit 2 during an outage scheduled to begin on, or before, January 1,1980.

DRWWWB;mvD/7 1687 199

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PF.e'G BOFKH A'ICMIC POWER STATICH Gntalment Isolaticn Valves tK7IES:

Valve M.zmbering: All valve nuters apply to both units, except 4 digit nuters. Ebr those valves, the first digit designates the unit 2 or 4 - Unit 2; 3 or 5 - thit 3.

Valve Types:

GB Globe DCV Diaphram (tntrol valve GT Gate VB Vacinxn Breaker CK Geck XV Explosive Valve BL Ball RO Bestricting Orifice B

Batterfly BCK Bc.ll Geck SV Solenoid IG Hydraulic (bntrol Unit RV Relief XKV -

Excess Flow Check Valve SCE Stop Check h

Isolation Signals:

Group I A.

Reactor Iow Uater Ievel (-48")

B.

High Steam Idne Flow (140%)

C.

High Steam 'Ibnnel 'Itmp. (2000F)

D.

Iow Steam Line Pressure (850 psi in Run mode)

E.

High Steam Idne Radiation (3 x normrd)

Group II A.

Reactor Iow Water Ievel (0")

B.

High-Drywell Pressure (2 psig)

C.

I5CU High Flow (300%)

D.

R5CU ncn-regen, heat exch. high tap. (2000F)

E.

High Peactor Pressure (shutdown cooling - 75 psig)

F.

High Reactor Pressure (600 psig)

G.

Standby Liquid Cbntrol Systan Operaticn*

Group III A.

Reactor Iow Water Ievel (0")

B.

High Drywell Pressure (2 psig)

C.

Reactor Bldg. High Radiation (16 mr/hr)

D.

Refueling Floor High Rad. (16 mr/hr)

Group IV A.

RCIC Steam Line Righ Flow (300%)

B.

RCIC Steam 'Ibnnel High Temp. (2000F)

C.

RCIC Steam Line Iow Pressure (50 psig)*

D.

High Drywell Pressure (2 psig)

E.

RCIC Steam Line Isolated Group V A.

HPCI Stm. Line High Flow (300%)

B.

HPCI Stm. Tunnel High Temp. (2000F)

C.

HPCI Stm. Line Iow Pressure (100 psig)*

D.

High Drywell Pressure (2 psig)

E.

HPCI Stm. Line Isolated F.

Reactor Iow Water level (-48")

1687 200

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, Giulp VI Peactor Iow Water Ievel (-144")

High Drywll Pressure (2 psig)

Group VII IPCI Initiatim:

Reactor Iow Water Ievel (-144")

High Drywell Pressure (2 psig)

Reactor Iow Pressure (450 psig)

IN Bernote Manual M

Manual (local cnly)

IC Iccked Closed

'nzrbine Trip

Tr FMP. -

Push Button, msnentary contact opens valve for test Process Signals h

Are diverse actuaticn signals provided in accordance with SRP 6.2.47 isolates mly if in shutdown cooling node.

Modification in g vg r.ss to cap this line.

Line is ESF and Essential unless in shutdown cooling node.

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Diversi Number Line Description Fluid Size (Yes /No).(Yes/No)

Number Location Troe 2 No.

Group (Signal)

(Yes/Nd l

N-17 R!iR Head Spray water 6"

No

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II (A,B,E)

Yes MO-10-33 Outside GT II (A,B,E)

Yes N-18 Drywell F1. Dr. Pump water 3"

No No MO-20-82 Outside DCV 8

II ( A,B)

Yes Discha;ge MO-20-83 Outside DCY II (A,B)

Yes N-19 Drywell Equip. Dr.

water 3"

No No MO-20-94 Outside DCV 8

II (A,B)

Yes Pump Discharge HO-20-93 Outside DCV II ia,B) ya N-21 Service Air Supply air 1"

No No Inside GB 9

If n.a.

Outside GB LC n.a, N-22 Inst. Cas Supply air 1"

No No Outside CK 10 10-29691 Outside DCV II( A,B)

Yes N-23 RBCW to Recire. Pumps water 4"

No No MO-2373 Outside GT 11 ftM n.a.

N-24 RBCW from beirc. Pumps water 4"

No No MO-2374 Outside GT 11 HH n.a.

N-25 Drywell & Torus-Purge sir 16" No No A0-2505 Outside B 12 III Yes

& 205B Purge Supply -purge

& 20" No No A0-2519 Outaide B III Yes

-purge No No A0-2520 Outside B III Yes

-purge No No A0-2521A Outside B III Yes

-purge No No A0-2521B Outside B III Yes

-Na purge No No A0-2523 Outside DCV III Yes

-Na Furge No No Outside CK(2)

III Yes

-Vacuum Relief Yes Yes A0-2502A Outside B

FR n.a.

-Vacuum Relief ves Yes 9-26A Outside VB

-Inst. (preasure)

Yes Yes Outside GB N-2{

Drywell Purge Exh. -CAD air 18" Yes Yes A0-2509 Outside DCV 13 III Yes

-CAD Yes Yes 10-2510 Outside DCV III Yes

-purge No No A0-2506 Gatside B III Yes CO

-purge No No A0-2507 Outside B III Yes N

-inst. gas No No A04235 Outside DCV III Yes

-cacs sample No No SV-2671G Outside SV III Yes

-caes sample No No SV-2978G Outside SV III Yes N

-cad sample Yes Yes SV 4960B Outside SV RM n.a.

C

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U

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-Inst. (Press Yes Yes Outside CB

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Valve @ Valve Fig. Imlaticn@

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tb tb Inside BO 16 Plate Pressure 25, 27 Outside XFW IF28A, B Inst. Lines - RW water / steam 1"

Yes Yes Inside IO 16 C,E,P Icvel/ Pressure C-tb C-tb 17A,lSA, (11tside XK.V ll,13A,15A FF28D Inst. Line - RW steam 1"

tb tb Inside IO 16 Ileal Pressure 23 Outside XKV N-29A,D, Inst. Lines - RW water / steam 1"

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tb

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mter 1"

tb e

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water 1"

tb (4

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. Valve

'falve Fig.

Isclation @ Diverse @

Valveh Pene.

Line ESP Essential Ntsnber Line Description F1t.id Size (Yesho)

(Yesho)

Nister Incation !vpe b No.

Group (Signal)

(Yes/No)

N-213A Torus Drain (with level water Yes Yes Outside G3 15

3 inst.)

'-215 Inst. Line - Unit 2, air 1"

Yes Yes Outside 33 15 Tcrus Level N-216 EPCI Min. Flow water 4"

Yes Yes 23-62 Outside CK 32 N-218A Inst. Gas Supply air 1"

No No Outside CK 10 A0-2968 Outside DCV II(A,B)

Yes N-218B CACS Sample Line air 1"

No No SV-26711 Outsi-SV 26 III Yes SV-2978A Outa' SV III Yes N-218C IL3T Ccnnection air 3/4" No No Outside GB(2) 18 14 n.a.

.%219 Torus Purge Exhaust-caos air 18" No No A0-2311 Outside 3 24 III Yes

-caos No No AO-2512 Outside 3 III Yes

-CAD Yes Yes A0-2313 Outside DCV III Yes

-CAD Yes Yes A0-2314 Outside DCV III Yes

-cacs anal.

No No SV-2671P Outside SV III Yes

-cacs anal.

No No SV-2976F Outside SV III Yes

-CAD anal.

Yes Yes SV-4960A Outside SV EM n.a.

-CAD anal.

Yes Yes SV-961A Outside SV F11 n.a.

Ne no SV-9361 Outside SV R>l r.a.

-rad.kas-inst. press.)

Yes Yes Outside JB

_.01C Va n..:: Punp Disch, air 2"

No No 13-10 Outside SCK 35

.-221 13-33 Outsida CK

!-223

.E I h cine Drain uater 2"

Yes Yes 23-13 Outside SCK 35 23-56 Outside CK

-224 Ovre Spa,, Test Line - uater 10" Yes no 10-14-26A Outside 33 36 VI Yes Unit 2 No

o Outside CK(3)

Yes Yes14-65A,C Outside CK

.:-22)

.CIC & D r s Water unter 6"

No Yes FD-13-4)

Outside GT 37 FJ!

n.a.

1 uleant-p S ct.

No Eo NO-1,-70 Outside 3r II(A,B)

Yes No No 10-1--71 Outside Or II(A G)

Yes to D RV-12-72AtoD Outside 3V

]

..-226A A.at Ptr.p S..ction water 2,"

Yes Yes 10-12-13AtcD Outside GT 38 FJI n.a.

L-227

'"?C1 P=p Sse tion ue.ter 16" Yes Yes M0-23-58 Outside JT 39 V(A,B,C)

Yes co N

N O

l Pene.

Line ESF Fasential Valve @

Valve Valva Fig.

Iselation @ Diverse h POrher line Description Fluid Size (YesMJo)

(Yes /fic)

Y rbe--

Loc a ti en ime (2) No.

G:oup (Sicsi) (Yesolo)

N-223 Core Spray Pimp Suction water 16" Yes Yes

!D-14-7AtcD Outsida GT 40 m

n.a.

AtcD N-229 Core Spray Pts:p Min water 4"

Yes Yes14-663,D Outside CK 41 Flow - Lhit 2 Outsida CK(2)

N-230 RCIC Ptap Min. Flow uater 2"

No Yes 13-29 Outside CK 42 N-233 i?CI Test Line - Unit 2 unter 4"

Yes No

}D-23-31 Outsida GT 43 V(D.F)

Yes N-234 Cara Spray Test uater 10" Yes No

}D-14-26B Outsida GB 44 VI Yes Line - Unit 2 Outside CK(2)

?.-234A Cere Spray Test unter 10" Yes No

!D-14-2G Outsida CB 44 VI Yes Line - Unit 3 Outsida CK(3)

..-Zj,3 0,:re Spray Zast.

ster 1d" Yes No ID-1L26A Outside GB 44 VI Yes Line - Unit 3 Outsida CK(2)

N-235

-PCI Teat Line - Unit 3

..a ter 4"

Yes I;o ID-23-31 Outsida GT 43 V(D,F)

Yes N-236A Care Spray Pmp i.in.

i; ster 4"

Yes Yes14-663,D Outsida CK 45 31cw - : nit 3 N-236B Care Spray Pu p Min.

.:ater 4"

Yes Yes14-661,C Outside CK 41 Flew - Unit 3 No

.'io Oatsida CK(2) 2.-250 Inst. Line - Dtit 3, air 1"

Yes Yes Oatside GB 15 Iarus Leval 1F:==

N C

NAM

{[] [))Q CChiTAiWMEWT 150LATiOW VALVE ARRC2 ' ~T.

LEGEND LETTERS DE SCR\\ 1"T ich!

AO AIPs OPEPsATED VALVE MO MOTOR OPEPsATeo VALVE TC TEST COWNECT\\oW Ps C 5 PsEACTOF5 CoolAMT SYSTEM TCh TESTABLE CHEC6 VALV::.

SV SoLEwolO OPERATED VALVE Ps0 FLOW RESTRtCTING ORtFice LC VALV8 LOCW&D CLOSED CV CONTROL VALVE XV EX PLOStVE VALVS SYMPaOLS Ib0.7 2lI lOl BALL VALVE ch FLOW L\\M\\T\\MC,(OREXCESS FLOW CHECW VLV.)

M CHECW VALVE MOTOR Dxd GLOSE NA\\ yE pq ACTOATED VALVE h

sotewogo

- ^xC GATE VALVE d

DI APMS, AM pq ACTuh,TED VALVE gdc LTuATED VALVE Flow

'I RESTR\\c7\\wd CFacicE

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'y PATTEEF\\.,y VALyE

~~

A D'I T 2RLt 4 ACTUATED MORMALLY NO W ALLY

,g.

CouTsotVALvE x

OPEH N

CLOSED hlOTES:

E

STRANER,

l. VALVES AB; S40v/W IM POS\\T \\oM CORRESPOND \\MC, To MORMAL P0v/EM O PE R AT t O A..

. '90TTEP VW'S Do NOT MORMALLY PERFORM A COMTAIMMEMT \\ SOL. FOM

PBAPS C.; s A 5. tMNT 150LAT\\ON VALNE AF?. 2 ~

MS COMTAIMMENT rP r

n Ao

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TC A

MA\\M u

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MO MO DR AI MS l

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STEAM i

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M AI M STEAM LINE (k1 T A TO O)

,. FIG.l STEAM DRAltd5 (M-6)

Oh

,e

- - ~

TCK MO s

RCS CONTAINMENT ' "

MO L

I4 3HPCI r-L,j TC TC

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s TC T

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3 FEEDWATEPg TCK M_O l.1 TC

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RCS CONTAIN MENT MO MO TC C)<!

EJ C)<3 3

1687 2 0 V

DRAlld TO Flc,.1.

RCIC STM SUPPLY (M 10)

MPCI STM SOPPLY (N il)

, FIG.3

i.

RCS CONTAINMENT Mo MO TC u

2 m

2 ym Gd y m U

~

F1G. 4 RH R SauTOOWN COOLING SocTION(Ml?.)

RCS coNTA1WMENT MO TcW A

Ex0

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TC ->

AO

lG. 5 R H R SHOTcoWN COOLING RETORN(M-G A cose seaAy oiscsAsae (w-le Air.s) 1687 2I4

.. _. iT! ~ - ~ ~

RCS CONTAlN MENT Mo MO TC E3 C)<l 3

RO

F-t3 o

I NSTR. LtME

g; ob INSTR.Ltwe ao RWcu PUMP sucTtoN (u-143 FtG. co IwsTaumaw7 uwe (a-so o,e)

RC5 CowTAtWMENT Mo Mo 7c

's 'i

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FIG.7 RHR HEAD S P R A'Y (N l l)

RCS CONTA)WMENT DRY WELL DRAIN SUMP AO AO TC A

A D<

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lAl 3

3 Ed TC 1687 2 h-DRYWELL EQulR DRN.50MP DISCH. (M 1b) klG. 6 DRYWELL FLOOR ORM. SUMP DtSCM- (M \\ J)

=. _................. _.

RCS CONTAINMENT LC LC em 4

w2 e

O w,

L. J k

i r,

o TC FIG. J SERVICE AtR SUPPLY (M-10 RCS CoNTAIMMEMT go T.C D G--i. 'I E-]

t/

a lL\\l 3

TC FIG.lO IMST. G AS SoPPLY (N 1LN 52F,N-118A)

RCS CONTAIMMEMT MO

[ *',.

[]

3 SOPPLY MO

[*'

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3 RETURN t

1687 216 F)G, il F3 g C W (N 'L'6 4 W-14)

AO RPV COMTAINMEWT

, Mt PURGE SUPPLY AO AO A

A Pusse N 25{3

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TC 1687 217 4

DPNWELL E TOPsuS PURGE SUPPLIES mz

a-w 4 x-zee

O RAD. G AS

, AN ALY M F.

RCS CONTAINMENT 2

SAMPLE 8 8 CAD X

pg pq 3 ANALYT E P SAMPLE CACS

-[*O

> <0

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SAMPLE TC Ao Ao O

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[]

5Q bH SGTS AO AO A TC O

CAD dO!

!A H E%HAusT AO A

INST. G AS l6 H COMP.

SucTtow

w; 3 INST.

} ()87 i

FtG.13 j

DPNwELL Puac,e ExNAusT (w-2co) i

RCS COMTAtMMENT T\\P

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)q e

MECHAMitM F\\G.14 TlP PURC3E (N- % p)

RCS COMTAINMENT L A =C-

] LEVEL

) NSTR.

Pool

[Ma';-

(M-20GA,B)(M 7.15)

FIG.15 IwsTauuewT yyeS (u.11S A) (N-250) 1687 219

=.......;..............

~.

RCS COMTAtwuswT 8

DeG-O INSTRUMEMT ll

/-

(M-t 1 E,F) (M-26 A To F)

(N19 A.D.E4F)(WSo A To F)

(M-37.AEA9F)(N 33 ATOD)(W-34 IuSTRUMEWT uwes (4 40AToD)(N 50A,B,C)(M-StE)(M-57.E)

FtG. %

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l

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INSTROMEMT s

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RCS cowTAtuMewT

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INsTaoMeMT 1687 220 Ar is itsi caswecTious (u-37. C,R) (M16 C)

PCS CONTANt e EMT c

[pj INSTRuu :(~

(W SS F) (N-4G A,6)

-16.19 IMSTROMENT LINE (N 49 B,C,E 4F)(W \\O7.6)

RCS CouTAIMMEMT

@ l3 s

EJ IO' D<:

xv TIP DRNES (u-M A To E) lG._7.0 RCS CONTAINMENT TC TC

/

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!._ j o

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CRD RETORN

( N-%)

h Pc5 cow 7A gwrew-Tc CAC

'M " $ 4 7 4 2NJECTiON MO KO

<;r es

)

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^

Rcs cowTAIWMENT Ao AO A

Tc A

>d J LI-LJ lJ Ll 2

i REctRC Loop SAMPLE

( M-4l )

Fic,.2.4 MAiW STEAM SAMPLE (M-57)

Rc5 coWTAINMEWT TC yg u

1687 222 Fic,. 25 STAWo-Sy upoto coWTaot (W - 41)

RCS couTAu MEMT

( 7 9 s c A t_ )

r- - -> HNDRAULiC CONTROL UWlT i

l TO GCRAM I

ld!

N u D\\SCH. HEADER I

, FROM DRWE

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I N -

V N

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c-@'

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3 l84 x8 "WTR. HEADER WiTHDaml l

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i WTR. HEADER 0iSERT i

l PROM a

l o ACCUMULATOR i

CH AKC,\\ MC, E^V L___.,

BALL CMECW VA\\.VE \\WTE.GRAL \\N1.TH DK\\VE OW\\T CMECKS FLOW \\M TWO D\\ RECT \\ONS.

'687 223 (W b^( ATO D)

, tG. 'i'2.

CRD INS"i*,T 4 WITHDMAWAL (N >2>A To D)

RCS CoNTA\\MMeNT

@t @

To gas

<]

>0 C)<hSM AN A'.Y 2.ER a

Tc Tc (M 51 A,8)

F 1G. 'LG CACS SAMPLE Likl55 (NtioB)

RCS CouTA1WMERT To RAD D,"

a GAS SAMPLER

,TO CAD 9

m 2

r * ~-

gas r,

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n A Aty7.Eg

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Tc Tc

, INSTR.(u-7.03 ONLY) fig. 27 CACS / CAD SAMPLE LlWES (N-51c) (Wlod RCS CONTAINMEWT FROM GAS y

XaM pg AN A LY7.ER n

TC TC 1687 224 M, 'L B CAC5 SesMPl=E RET 9R.H (M-51D)

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2 (M-%, W -54,N-55,N-%)

F t G.19 DRWELL CHILLED \\ VATER RC%

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C lNhr-LA Tc X

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__ Toaug Vacuum e>aaAnea (N tos A)

RCS CONTA)MMEMT q

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spaAy EJ

4

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->4-PW R W

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S U P P.

FLOW POOL

_{y (iNes FIG. 'St RH R TORUS SPRAY

( N 2.11 A, B )

RHR TESTJi POOL COOLlMG (M 110 A,E)

RCS CONTA\\NMENT E:

M 2 PUMP DISCHARGE

~

SUPP.

Pool I.

1687 226 FIG. 32.

HPCI MIN. F LOW (N-7 iG) t_

?

T.N"AiMMEMT AO AO A

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n g*

igi ijgi 7

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a EXHAUST

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Docket Nos. 50-277 50-278 Item 2.1.5A, Dedicated Hj[ Control Penetrations Peach Bottom Atomic Power Station utilizes a Containment Atmospheric Dilution (CAD) Sys tem f or p os t-LOCA combus tible gre control.

This system is described in FSAR Supplement 1,

Respqnse to Question 14.7.

A review of the Peach Bottom design has been completed.

We have verified that the containment isolation provisions for all purge and vent lines are single failure proof during CAD system operation.

The following small interconnecting lines do not pres ently meet this criteria:

Penetrations System Iso. Valves N-26 Instrument Nitrogen Suction A0-4235/5235 N-26 Radioactive Gas Sample Lines SV-4966B/5966B N-51C Radioactive Gas Sample Lines SV-4966C/5966C N-203 Radioactive Gas Sample Lines SV-4966D/5966D N-219 Radioactive Gas Sample Lines SV-4966A/5966A Modifications are being developed to add additional isolation provisions to these lines.

We expect that these modifications will be completed by January 1, 1981.

The use of pressure control valves in the 2" diameter CAD system purge lines will assure that flow requirements (a p p roxi ma t ely 100 scfm intermittently _for an average of 10 scfm). _ _ _

will be met.

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System Test Frequency I

f RWCU Pressurize during operational Hydro - Inspect Each Operational Scram Discharge for leaks.

Hydro.

I a) RWCU Backwash While system i'a not in operation - inspect Visual Each to Phase for leaks or evidence of leaks which occur during Refuel Cycle Separator operation.

(Not in operation for ALARA) or b) Cleanup Phase Take area airborne samples - satisfactory airborne l

Separator l levels verifies system integrity.

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Cleanup S l' Discharge ;udge mix f

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ENCIASURE 5 Design Review of Plant Shieldirg and Environmental Qualification of Equipnent for Spaces / Systems Wiich May be Used in Post-Accident Operations, NUREG-0578 Section 2.1.6.b Our review of accessibility has been separated fran our review of equipnent qualification for post-accident corxlitions. Our equipnent qualification review is being done in response to Bulletin 79-01, and, as w have iMimted in our previous letters, the results of this review will be subnitted on or before January 31, 1980. We results of our review of post e W nt accessibility is discussed below.

Arms Requiring Continuous %=ncy I.

Control Roan - We shielding design and ventilation smem of the main control roon were evaluated for loss-of-coolant, min steam line break, control rod drop, and refueling accidents. Se results of these evaluations are in the FSAR. We source term specified in H. R. Denton's letter to All Operating Nuclear Power Plants, dated October 30, 1979, was determined not to affect the results in the FSAR.

We FSAR dose calculations assuned a release of 100% of the noble gases, 50% of the halogens, and 1% of the solids contained in the reactor to the primary containment, source terms based at TID-14844, and 0.635% primary containment leak rate and a secondary contairunent ventilation rate of 100% per day.

The integrated dose in the main control room for 30 days continuous m m ncy was calculated to be 2.0 rem. If a primary containment leakage rate of 2.0% is assumed, the 30 day dose would be 4.95 rem.

The additional dose due to 1% of the containment leakage bypassing the standby gas treatment filters will be determined by January 31, 1980. Wis assunption of bypass was not required for the FSAR.

2.

Technical Support Center - We location and description of the Technical Support Center is discussed in our response to NUREG-0578, Section 2.2.2b.

Dose calculations wre performed to determine shielding and ver*:ilation requirements. We calculations assumed a design basis la. of-coolant accident with 0.5% primary containment leakage rate and with 1% of the containment leakage bypassing the standby gas treatment system filters. W e Unit 1 centrol rocm was nodeled to include the control rean, conputer roan, switchgear roon and office area since there would be significant air flow comfunic5Eion between these volumes. We intake airflow was varied from 100 cfm to 20,000 cfm.

1687 236

. 'Ibe 180 day anaaa without shieldirq ard charcoal filters are the following.

P1 tune shine Thyroid Airborne whole body Direct Shine

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3.49 x 10 aan 1.174 x 10 Rem 7.09 x 10 Rem 13.1 asa As a result of these calculaticns, a filter train consisting of preheater, prefilter, HEPA filters ard charcoal filters will be N.

'Dnis systen will remove 99% or more of the iodine; thus, the resultant 180 dose to the thyroid will be substantially reduced ard within acceptable limits. Shielding will be aMad to the Unit 1 Nilding's wall cn the side facing the Unit 2 reactor buildirs to bring the dose fran direct shine within acceptsble limits. Six inches of concrete or equivalent will reduce the 180 day dose fras direct shine to 2.6 rem.

Areas Requiring Infrequent Access 1.

Secondary Otmtairment - Within seccndary containment are motor centrol centers, aampla stations and the instrunents that measure reactor water level. Dose calculaticms were peu.f +

" to determine

-=ihility, shielding requirenants, and avin=nt qualification.

In rocas that en*ain E03 piping, a Mw14ng asstmption was made that the dose fran shine would be determired at the surface of a 20 foot 1cng, 24 inch in diameter pipe containing the source term mIw-i fiarl in H. R. Denton's Ocechar 30, 1979 letter, as referenced above. 'Ihe airkarne dose due to leakage frcm primary containment was assumed tn be uniform through cut all recms and areas within secondary en*airunent. Post-accident access to rocas containing ECX:S eqniP nt is not anticipated. 'Ihe motor control centers, sample stations and the reactor water level instrunents are in areas Ware the dose is overwhelmingly due to airknrne. The ganma dose rate in air was calculated to increase with time until 4-8 hours when it levels off and ts.n to decrease with time.

Hours After Accident Dose rate (Rad / hour) 24 35 1

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. 1he primary cot *ai-ant leakage rate was===W to be 0.50% in the calculation to detemine the above doses; however, the integrated leakage rate test performed cn Unit 2 in 1976 showed a measured leakage rate of 0.006% and the test perfvuoi on Unit 3 in 1977 for Unit 3 showed a inaaaned leakage rate of 0.29%. 1hus, if a design basis loss of coolant Waant occurs, the dose rate in secordary cor* air-et would be slower to rise and pnaaihly not as high as shown above.

(The Technical Specification limit the allcwable leakage rate to 0.5% of the primary mntalrunent volune per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 49.1 paig.)

1he General Electric Caupany is evaluating the effects.of an accident on the reactor vanaal level instrunentation as part of NURai-0578 itan 2.1.

.b.

Included in this evaluaticn is the determination of whetkc access is required to the reactor vessel level instrument racks to backfill the instrunent line for the reference leg of the instrumentaticn. This evaluation is expected to be ccmplete at the end of this year. If - as is required, we will provide a means of backfilling which is operable fran an accessible area. This wifimtion, if maaary, will be conpleted prior to January 1, 1981.

We believe the probability that an operator would have to go to an essential motor cx:ntrol center cfter an accident is very law. If maa would be required, we estimate that an operator may have to 4 minutes spend a maximum of 8 minutes in secaldary contairrnent:

at the notar control center and 4 minutes for entry and exit.

Entry into secordary containment within the first h hour after the accident with a stay time of 8 minutes could result in a dose of 4-5 rem bassi cn a primary c:ntainment leakage rate of 0.500%.

Access after the first h hour after an accident would be limited to shorter stay times or delayed until after roughly 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the accident. Even if a single failure occurs at a motor control center, we have the capability of removing decay heat; thus, an entry into secondary ecntainment wculd not be maury.

The capability of cbtaining primary coolant and primarv containment samples is diammand in our response to NUREG-0578, Section 2.1.8a.

2.

Radwaste Panel - The radwaste panel is located in the radwaste building. Dose calculations to determine the dose wpent frcm direct shine have been atmplgted. She total 180 day dose fran ren. Dose calculations to detemine ganma radiation is 7.7 x 10 the dose wwent fran airborne and determination of corrective actica, if any, should be cxmplete by January 31, 1980.

1687 238

. 3.

Radihicat r =Mratory

'the radimhamical laboratory is located in the errhine building. '1he dose went fran direct shine has been Gp. 1he total 180 day dose fran gansna radiatim is ent and 8.78 x 10 rern. Dose calculations for the a4

rne w de*aminatim of corrective action, if any, should be crupleted by January 31, 1980.

GUI/mtd 3/3 1687 239

Docket Nos. 50-277 50-278 Item 2.1.8A, Japroved Post-Accident Sampling Capability A review of post-accident sampling capabilities at Peach Bottom has been completed.

Provisions and p rocedures currently exist f or sampling and analysis of pos t-accident gases and liquids through the use of the existing Reactor Water Cleanup sample lines and Containment Radioactive Gas Monitors at radiation levels permitted by plant shielding design.

All existing sample p oints are within the reactor building.

We can not assure that pers onnel exp osures during the sampling and analysis operations will be within the NRC's criteria for the source terms given in NUREG-0578, Item 2.1.6.b.

Major sampling system modifications to minimize pers onnel exp osure are currently being evaluated.

Provisions will be made to take approp riate samples f rom p oints outside of the reactor building.

Several alternative sampling system designs are currently being evaluated.

continuing'to review the effect of background We are radiation levels in the counting laboratory.

We have not been able t o es timate airborne levels in the turbine building location within the period of time allowed.

These analysis will be completed by January 31, 1980 and approp riate corrective actions taken.

1687 240

Docket Nos. 50-277 50-278 Methods for Estimating Releases Interim High Range Effluent Monitors (Item 2.1.8.b)

The interim method for estimating releases will be accomplished by January 31, 1980 with the installation of high range noble gas effluent monitors.

The system will basically consist of an NMC Model GA-2TO wide range gamma area monitor in a fixed geometry on the sample line with readout in the control room.

A system will be installed on the off-gas stack sampling line and the Units Two and Three reactor building vent s amp ling lines.

The shielding review for Item 2.1.6B has shown that access to the iodine cartridges cannot be assured f or certain p os tulated a ccidents.

Based on reviews to date, it is expected that the reactor building vent iodine cartridges will be relocated to the radwaste building roof or the turbine building fan rooms.

Modi fica tions to assure access will be implemented by January 1, 1981, as required by the NRC s ch edule f or NUREG 0578, Section 2 1.6b and 2.1.8b modifications.

The interim high range noble gas sampling locations are s h ow n on Figures 1 and 2.

Each sampling line will have an upstream filter and charcoal cart ridge to mitigate particulate and iodine buildup in the monitor.

The fixed geometry will be c ons t ru c t e d of a stainless steel gas chamber and detector holder, as sh own in figure 3, surrounded by a lead shield to reduce background interference.

Control room readout will be on a three p en recorder.

The system will be connected t o saf eguard p ower.

The calculated range of the sy, stem from volume s ou rce calculations will be from 1.4 x 10 to 1.4 x 10+ uC1/cc (Xe-133).

Calibration factors and procedures using a noble gas source will be completed by January 31, 1980.

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t Docket Nos. 50-277 50-278 Onsite Technical Support Center, NUREG 0578, Section 2.2.2b Interim Technical Support Center The second floor of the office complex of Unit No. I at Peach Bottom has been converted f or use as an interim technical Support Center (TSC) for use by plant management, technical engineering support personnel and NRC personnel in the event of an accident.

The arrangement of the interim (TSC) is as sh own on Figure 3.

The TSC has direct communication with the control room, Philadelphia Electric Company's c orp ora t e offices, and the NRC's headquarters in Bethesda, Mary lan d, and regional office in King of Prussia, Pennsylvania.

The TSC als o contains t e ch nical information in the form of plant drawings, procedures, and descriptions, and radiation monitoring equipment.

A closed circuit television link with Unit No. 2 and 3 control room will be provided f or display of plant parameters.

Installation of the CCTV system is expected to be completed by the end of th's firs t quarter, 1980.

Permanent Technical Support Center A permanent TSC will be established on the third floor of the office complex of Unit No. I by January 1, 1981, as sh own of Figure 1.

Location The location of the Unit No. I building relative to the Peach Bottom Units No. 2 and 3 is sh own on Figure 2.

The Unit No. I building is immediately outside the Units No. 2 and 3 protected area.

Size The permanent TSC is of sufficient size t o h ouse 25 pers ons.

Communications, technical inf ormation and radiation monitoring equipment will be as described f or the interim TSC.

1687 245

Mr. H. R. Denton Page 2 Instrumentation A closed circuit television sysiam, including p rovisions for video tape recording, will be ins talled f or display of plant parameters.

Power Supply The TSC will be normally f ed f rom non-essential 480 V buses.

Emergency power will be provided by either a new skid mounted diesel generator or the existing Unit No. 1 diesel generator.

Since these emergency sources are independent of Unita 2 and 3, we will not be degrading Units 2 and 3 safety-related power sources.

4 During operation of the TSC, if the normal s ou rce is lost, the dies el generat or will be automatically started.

This may result in a 10-15 second interruption at the TSC during diesel generator start-up.

There will be no loss of stored data on the video tape du.e to this interruption.

Structural Integrity Locating the TSC in the Unit No. I building establishes it in a well built building, designed in accordance with sound engineering p ractices.

Habitability The TSC ventilation system will be designed in accordance with Reg. Guide 1.52, General Design Criterior 19 of 10 CFR 50, Appendix A, requested by NRC letter dated October 30, 1979, sections 2.2.2.b.

1.

Ventilation will be p rovided through redundant intake systems, designated A and B.

System A will p rovide clean filtered outside air under normal conditions.

System B will provide radiologically clean outside air during accident conditions.

System B will contain a filter train consisting of preheater, prefilter, HEPA filters and charcoal filters.

System B will be manually initiated.

2.

Monitoring systems will be p rovided in the TSC to continuously indicate airborne radioactivity concentration.

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SUBJECT:

Dear Mr. Denton:

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