Forwards changes to FSAR Section 18 & other ref sections updating ...

REGULATORY NFORMATION DISTRIBUTION .S iEM (R IDS)

ACCESkON NBR: 8612300217 DOC. DATE: 86/12/19 NOTARIZED: NO DOCKET IFACIL:STN-50-530 Palo VeT de Nuclear Stationd Uni t 3> Arizona Pub 1 i 05000530

AUTH. NAME AUTHOR AFFILIATIONHAYNES> J. Q. Arizona Nuclear Pouer Prospect (Wormerlg Arizona Public Serv

RECXP. NAME RECIPXENT AFFILIATIONKNXGHTONd G. W. PWR Prospect Directorate 7

SUBJECT: Foneends changes to FEAR Section iS h othev vef sections 5'6gupdating info in l esponse to TMI Items II. 8.2> IX. D, 3> II. F. 1

a III.D.3.4.Changes update descriptions to represent cuTrentplant conditions 8t lllill be included in next" FSAR amend.

DISTR IBUTION CODE 8001D COP XES RECEIVED: LTR ENCL SI ZE:TITLE: Licensing Submittal: PSAR/FSAR Amdts 0 Related Correspondence

NOTES: Standardized. plant. M. Davisd h!RR: 1Cg. 05000530

RECIP IENTID CODE/NAME

PWR-8 EBPWR-8 FOBPWR-8 PD7 PDPWR-8 PEICSB

INTERNAL: ACRS 41ELD/HDS3XE/DEPER/EPB 3eNRR BWR ADTS

.LREQ IL 04

/MIB

EXTERNAL: BNL(AMDTS ONLY)LPDR 03NSIC 05

NOTES:

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ADM/LFMBIE FILEIE/DGAVT/GAB 21

'RRPWR-8 ADTSNRR/DHFT/MTBRQN5

DMB/DSS (AMDTS)NRC PDR 02PNL GRUEL> R

COP IESLTTR ENCL

2 21 1

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01 1

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3 3

TOTAL NUMBER OF COPIES REQUIRED: LTTR 37 ENCL 32

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Arizona Nuclear Power ProjectP.o. BOX 52034 4 PHOENIX, ARIZONA85072-2034

December 19, 1986ANPP-39455-JGH/JKR/98.05

Director of Nuclear Reactor RegulationAttention: Mr. George W. Knighton, Project Director

Pt<R Project Directorate 87Division of Pressurized Mater Reactor Licensing — B

U.S. Nuclear Regulatory CommissionWashington, D.C. 20555

Subject: Palo Verde Nuclear Generating Station (PVNGS)Unit 3Docket No. STN 530Changes to the FSAR Concerning TMI Related SectionsFile: 86-D-005-419.05 86-G-056-026

Dear Mr. Knighton:

Attached for your review on PVNGS Unit 3 are FSAR changes to Section 18 and otherreferenced sections. These changes involve updating information in response toTMI items II.B.2, II.D.3, II.F.l and III.D.3.4.

These changes are justified because they update our descriptions to more accur-ately represent the current condition of the plant. Many of the changes areeditorial in nature.

For PVNGS Units 1 and 2, safety evaluations have been completed for implementationof these changes in accordance with the requirements of 10 CFR 50.59. The safetyreviews and evaluations have determined that there are no unreviewed safety ques-tions involved with the changes. These changes will be included in the next FSARamendment.

If you have any questions, please contact Mr. ti. F. Quinn of my staff.

Very truly yours

J. G. HaynesVice PresidentNuclear Production

JGH/JKR/rwAttachment

cc: 0. M. De MicheleE. E. Van Brunt, Jr.E. A. LicitraR. P. ZimmermanA. C. - Gehr

Bbi2300217 Sb1219PDR ADOCK 05000530A PDR

J

..8612300217

PVNGS FSAR &~~~ 4vgd

$0(SITING AND DESIGN

1S.II.B.2 DESIGN REVIEM OF PLANT SHIELDING AND ENVIRONMENTAL

QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS

WHICH MAY BE 'USED IN POSTACCIDENT OPERATIONS

~sitinnMith the assumption of a postaccident release ofradioactivity equivalent to that described in RegulatoryGuides ls3 and 1.4 (i:.e., the equivalent of 50% of the coreradioiodine. 100% of the coze noble gas inventory. and 1% ofthe core solids are contained in the primary coolant). each1icensee shall perform a radiation and shielding-designreview of the spaces around systems that may, as a result ofan accident. contain highly radioactive materials. The

design review should identify the location of vital areas and

equipment, such as the control room, radwaste controlstations, emergency power supplies, motor control centers,and instrument areas, in which personnel occupancy aay be

unduly limited or safety equipment may be unduly degraded bythe radiation fields during postaccident operations of thesesystems.

Each licensee shall provide foz adequate access to vitalareas and protection of safety equipment by, design changes.increased pexmanent or temporary shielding. oz postaccidentprocedural controls.. The design review shall determine whichtypes of corrective actions aze needed foz vital areasthroughout the facility.

VNGS Evaluation

A, radiation and shielding-design review of the spaces aroundsystems that may. as a result of an accident. contain highlyradioactive materials has been conducted.

General design .considerations 'to keep post-accident exposuresALARA are described in section 12.1.2.4. A summary ofthe shieldi'ng design rev'iew results is provided insection 12.3.2+. and a description of the source terms used

18. II.B-11 ',(PROPOSED) NDNENT ~l&

0

II

0

PVNGS FSAR

SITING AND DESIGN

in the post-accident shielding reviev is provided in goC'ection12.2.3. Past-accident radi'ation zones are discussed

in section, 12.3.1.3. and presented as figures 12.3-25 through12.3-45.

The qualification of safety-related equipmen pt is rovided insection 3.11.5.2.The function, operation and design of the radiationaonitozi'ng system are described in section 11.5.

(PROPOSED) AMENDMENT PS 28'TT ~ B 12->6

0

il

0

0

C.

PIGS k'beak(

PROCESS AND EFFLUENT RADIOLOGICAL

MONITORING AND SMPLIHG SYSTENS

would be the radwaste building ventilation sys-tems which are provided with gaseous process

tars to monitor for leakage from the wastegas compressors and waste gas decay tank valves.

Areas in w zc eh' th new and spent fuel is received andstored, specs. xca yf'lythe containment and fuel building,are prove. e wx'd d 'th detectors which indicate and alarmin the presence. of abnormal radiation levels.The location of eac ~di Mo detector—xs—xndz-

draw-cated on 'atmf"II ~ .'~tt Qh 4OA-M.~ (l.5-5o

11.5.1.2 Postulated Accidents

The process, e uen , anffl t d area monitoring systems, collectivelyreferred to as the radiation monitoring system (RMS), aredesigned to per orm ef th following functions in order to meetthe reauirements oi of 10CFR50, 10CFR100, and follow the recom-

'mendations of NUREG 0737 and NRC Regulatory Guides 1.13, 1.97and 8.12 for postulated accidents:

A Provide the capability to alarm and initiate contaxn-ment purge isolation in the presence of hxgh airborne

'adioactivity within the containment which could.poten-tially cause an offsxte dose xn excessess of 10CFR100

limits.B. Provide the cap z i yab 1 ty to alarm and initiate isolation

of the fuel building from the normal ventilation systemand actuation o uef f 1 building essential ventilation inthe unlikely event of a fuel handling accident in thefuel building.

'rovidethe cap x i yab lity to alarm and initiate isolationof the control room normal ventilation system andactuation o cont' control room essential filtration in the

August 1981 11.5-5 Amendment 5

0

il

45,~+«g \'2.K ->s<w Q,pvay

PVNGS FSAR 4 o~+ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES

ARE. AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

1'l.E-KI

12.1.2.4 General Desi n Considerations to Kee Post-AccidentEx osures ALARA

The facility layout will assist in keeping occupationalexposures ALARA even after a design basis accident. Whileexposures will be significantly higher than during normaloperation, required access is provided to vital areas 'and

systems without exceeding 5 rem/hr. Zone maps showing expecteddose rates in the event of a LOCA with sump recizculation areprovided Zone maps for thehypothetical condition of a LOCA with an. intact primary butwith a degraded core are ~re provided ibbRt-. A. discussion of the source terms for these events is .

provided in section 12.2.3. The dose rates projected forthese two sets of drawings do not assume decay beyond thatcorresponding to the onset of recirculation. Even so, virtu-ally unrestricted access will be permitted within the controland diesel generator buildings, as well as portions of theupper floor of the auxiliary building (such as the area of theoperational support center).To provi'de sampling capability with exposures kept ALARA, PVNGS

will incorporate a post-accident sampling system that meetsthe requirements.,of NUREG 0737 and Regulatory Guide 1.97.

IR. ~.B. hRevision 2 as described in section 9.3.2 and

post 4c cl M4 ~i) $

The only other area where„access ~u~ be required is to thehydrogen monitors/recombiners. projected dose rates withoutthe recombiners in operation but at the onset of recizculationare expected to be approximately 10 to 30 rem/hr (sump zecizcu-lation). As the recombiners ~~~ have to be installed ~~

+~~4 ~g 9z 4v~gc.4.lc J. d~z~'.p5~ b"~'PYQQs ~~/~~ $ ~~PI

Sys4

April 1986 12.1-22A Amendment 15

0

PVNGS FSAR

ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES

AR AS L .W IS ASONQBLE ACHIEVAPgE (ALARA)1X 4-o~r 5 mc. a<co ~' '\ no c. e,<m>g~ J ~~j,f, * ~ d4avc 4 vo pfcg,decay to about 1/10 the doses noted above. Thus, the instal-

lation dose rate (assuming sump recirculation) will be lessthan 5 rem/hr. While the dose rate would be greater than5 rem/hr for an intact primary-degraded core event, the recom-biners would not need to be installed since an intact primarywould not be consistent with hydrogen generation inside thecontainment. If hydrogen generation were postulated, thiswould necessitate a break or opening in the primary. Conse-quently, sump recirculation would be available with theconcomitant release of noble gases and dilution by-the refuel-.ing water tank. These consequences would lead to the dosesnoted above for the sump recirculation mode of cooling (i.e.,dose rates less than 5 rem/hr).ESF grade filtered ventilation is provided for auxiliarybuilding rooms below elevation 100 feet (refer to section 9.4).This will reduce airborne sources due to recirculation and/orcontainment leakage. Non-ESF grade filtered ventilation isavailable for use to reduce airborne sources above elevation.100 feet in the auxiliary building (refer to section 9.4). Theuse of non-ESF filtration is acceptable since there are norecirculation components above elevation 100 feet. Thus theonly significant source of airborne activity is containmentleakage. This leakage has already been accounted for in off-site dose analyses which assumed direct containment leakage tothe atmosphere. Secondly, this filter discharges via the plantvent. The plant vent will be monitored in accordance withNUREG 0737 and Regulatory Guide 1.97, Revision 2 to providenotification of decreased filter efficiency.Therefore, considering direct and airborne sources, access canbe provided to those vital areas necessary for control of theplant and personnel exposures will meet GDC 19 and NUREG 0737limits.

Amendment. 5 12.1-22B August 1981

il~

iS~

0

.PUBS'SAR

RADIATION SOURCES

a free volume of the region in vhich the 1eak occursin cm

C {t) ~ airborne concentration of the i radioisotope at.time t in pCi/cm in the applicable region

Prom the above equation, it is evident that thc peak or egui-2ibrium concentration, C i of thc i radioisotope in the

qapplicable region vill.be given by the follaving expression:

C ~ = (L,R') ~ h {P F} ~ / {V AT ) {2)

With high exhaust .rates, this,peak concentration vill be reachedvithin a few hours.

12.2.3 SOURCES USED IN NUREG 0737 POST ACCIDENT SHIELDING

REVIEW

The post-accident shielding review described in sec-tion 12.1.2.4>considered'vo -LOCA events. The first, vas a

LOCA vith recirculation accomplished via the containment.sump. The. second vas a LOCA vith an intact primary vithrecirculation accomplished via the, shutdown cooling system.The following core releases vere used. in thc review:

Source A: Containment airborne:. 100$ noble gases,25$ iadines - see table 12.2-8.

B.

Volumes

A.

B.

Source B: Reactor coolant: 100$ noble gases,50$ halogens, lg solids - see table, 12.2-9.

Source C: Containment sump: 50$ hal'ogens,, 1$ solids-see table. 12.2-10.

used for each source vere:Source A: Containment free valume af 2.6 x 10 ft .6 3

Source B: Reactor coolant system {RCS) volume of9.15 x 10 ft .

Ostj ieibd core. t4'.~~+«+ c.gr+~ ~+ l+ ~~~~+ @ '+sSN 4a h g<w<~ P)~ l5.co. ~a~)

February. 1984 12..2 21 anenament 1>

i

RADZATXON SOURCES

3Source C: The «inimum volume of +ate,r 7.76 x 10 ftpresent at the e oth tia of recirculation. (RCS

+ refueling eater tank + safety injectiontanks. 3

sum recirculation is represented hy sources A

and B {source A +as not. reduced even oughy sources A and so,sources). Theto assume noble gases in both,souia no «echanxsm a

ted in section 12.3.1.3.,results o e sf th hield review are presente xn sec''d

to initial time equals acro, vereDecay curves, norma axeo in ost-the sources as an aid in developing pos-

lans. These curves are presented as ig-'ccidentaccess plans. esc s

ures 12.2-1 {source A), 12.2-2 {source B), .an

(source C).

The. sq%~s nssv~ 4 he.

OPCX& Ilg '4r aae4 eve< are'4~>

)g ~hie tZ.Z- )( .

Amendment 5 12.2 22 August, 1981

0

Ib

~O

PIGS:FSAR

12.2 9

RADIATION SOURCES

/n.g gj,LOCA SOURCE B REACTOR COOLANT

(Curies)

Ãuclide Activity( Nuclide. Activity Nuclide Activity

Se-84Br-84As-85Se-85Br-85Kr-85mKr-85Se-87Br-87Kr-87Br-88Kr-88Rh SSBr-89Kr-89Rb-89Sr-89Br-90

-90"Rb-90Sr-90

Y 90Kr-91Rb 91Sr-91Y-91mY 91

Sr-95Y 95

Zr-95Hb-95Zr-99Hb 99No-99Tc-99mMo-103Tc-103Ru 103Tc 106Ru-106

2.55{+5)1.28(+7)5. 52 (+4)3.02{+5)1.61{+7)3.48(+7)1.21(+6)3.29{+5)2.87{+7)5.91(+7)3.87(+7)3.59(+7)8.61{+5)3.96(+7}1.08{+8)1.12(+6)1 13{+6)3.89(+7)1.16(+8)1.31(+6)6.60(+4)6.57(+4)8.95{+7)1 45(+6)1.49(+6)8 80{+5)1.49{+6)1.64(+6)2.02{+6)1.99(+6)1.95{+6)2.61(+6)2.61{+6)2.69(+6)3 23(+5)1 46{+6)1.53(+6)1.53(+6)6.72{+5)5.27(+5)

Sn-.129Sb-129Te-129mTe 129I 129

Sn-131Sb 131Te-131mTe-131I 131

Xe 131mSn-132Sb-132Te-132

1-132Sn-133Sb-133Te 133mTe-133I 133

Xe-133-Cs-134Sb-134Te-134I 134

Sb»,135Te 135I 135

Xe-135mXe-135Cs 135Cs-136I 137

Xe-137Cs 137Ba-137m

I-138Xe 138Cs-138Xe-140

1.08(+5)2.08(+5)3.68(+4)2.03(+5)8.05( 1)3.50(+5)9.30(+5)1 72{+5)1.00(+6)5.65{+7}6.17{+5)3.08(+5)8.24(+5)1.15(+6}5.80(+7)5.37{+4)1.10(+6)1.42(+6)1.01(+6)1.16(+8)2.26{+8)1.28(+4}5.69{+5)2.14{+6),1.27(+8)1..80(+5)1.09{+6)1 10(+8).6.38 (+7)5 31{+7)'.2;50(-1')1 32(+4)7.20(+7)2 09.(+8)4.84(+4)4.52(+4)4.39{+7)2.05(+8)2.07(+6)1.16(+8)

Cs-140Ba-140La-140Xe-143Cs 143Ba-143La-143Ce-143Pr-143Xe-144Cs 144Ba-144La-144Ce-144Pr-144

2.00 {+6)2. 11(+6 )2.11(+6)2.75(+6)4.52{+5)1.47(+6)1.81(+6)1 83{+6)1.84(+6)4.03{+5)1.25(+5)8.75(+5)1.61(+6)1.25(+6)1.25(+6}

a'. Numbers in parenthesis denote powers of ten

Amendment 5 12 2 24 August 1981

li

0

ThSLE

GCSRPM USED IN POST hCCIDENT SHIELOIHC REVIEW~

Source Type LOCh vith Sump RecirculationLOCh - Degraded Core-

Intact Primary

~ Contaimnent hir~ Hydrogen control system

~ Containment air~ Hydrogen control system

~ Safety Injection Systea~ Containment Spray Systemo Shutdown Cooling Systemo Post-hccident Saepling Systea~ Letdown System~

~ Safety Injection System~ Containnent Spray System~ Shutdmtn Cooling System~ Post-hccident Sawpling System~ Letdown System~c~

a. Where redundant systems exist, both are assumed in usc.b. Radvaste systems not used post-accident.c. Portions up to purification filter inlet.

0

if'

PVSGS FShR

RADIATION PROTECTION

DESIGN FEATURES

12.3.1.2 Desi Radiation Zonin and Access Control

Access into the plant stzuctures and plant yard areasis regulated and controlled.P1ant areai are categorized as design radiation tones accordingto expected maximum radiation levels and anti'cipated personneloccupancy with consideration given toward maintaining personnelexposures as low as is reasonably achievable and within thestandards of 10CFR20. Each design radiation xone defines theradiation level r'ange to which the aggregate of contributingsources must be attenuated by shielding. Each zoom, corridor,and pipeway of every plant building is evaluated for potentialradiation sources during normal, shutdown, spent resin transfer,and emergency operations; for maintenance occupancy reguire-aents; for general access requirements; and for material expo-sure limits to determine appropriate soning. The designradiation xone categories employed, and their descriptions are

~~

~~

~~

~ ~

~~ ~

~ ~~given in table 12.1-1; The specific design soning for eachplant area is shown in figures 12.3-1 through 12.3-20. Fre-quently accessed areas, c.g., corridors, are shielded fordesign radiation Zone 1 or Zone 2 access.

The control of entry or exit of plant operating personneL tocontrolled access areas, and procedures employed to ensure thatradiation levels and allowable working time are within thelimits prescribed by 10CFR20 is described in section 12.5.

4 z~)i2~~/»

)u

12

12.3.1.3 Radiation Zones - Post-Accident

Ra 'on xone maps were developed in accordance wx

HUREG 073 eview potential access throughout the plantpost accident. {Re PVNGS LLIR Section II.S.2.) Twoevents were considered as no 'ection 12.1.2.4 using

sources described in section 12.2. . e events verea LOCA with sump recirculation and an intact prx -degradedcore LOCA. stimated radiation levels in vital areas vereAbased on radi tion sources from the post-accident

February 1984 12.3 11 Amendment: 12

0

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lpga

$0.+ cJj-„~ ~ ~pQ ~. ~~~ i ~ e, c gg.+Cwv ~ v ~ 4 ~C.~ 0737+ ~v'~ .pO~W ~tA.~ CKV. y aW ~jk~. pc~fQC.c.i(~.~'OA'hefacility layout mR assist-in keeping occupataonal

exposures ALARA .even. after a design basis accident. While

exposures vill bc significantly, higher than during normal

operation, required access i's provided to vital areas and

systems vithout exceeding 5 rem/hr. Zone maps showing

expected dose rates in the event of a LOCA vith sump>a.>-s

recirculation are provided as figures . through}4,>-35:

Zone maps for the hypothetical condition of aA

LOCA with an intact. primary but vith a degraded core are}2'3- SQ }g.~-45'r

voided as figures< .. through ...The}2.2.5

source terms correspond to those noted in section

The dose rates projected for these two sets of drawings do

not assume decay beyond that corresponding to the onset ofrecirculation. Even so, virtually unrestricted access

vill be permitted vithin portions of the upper floor of the

auxiliary building (such as thc area of the operational

support center) and the lower levels of the control building.Continuous occupancy villbe permitted in the controlzoom, satellite .Tcchnical Support Center (TSC), TSC,

diesel generator building and emergency operations facility(EOF) as dose rates villbe 15 harem/hr or less.

0

~O

RADIATION PROTECTION,

DESIGN FEATURES ~ goperation of thc following systeas: contaiment, safety pjinjection/shutdown cooliny/containment spray, chemical andvolume control system {up to purification filter QQ.et)rpost-accident sampling, and hydrogen recombincrs. The gaseousxadwaste. system villnot be use& post-accident. Palo Verdedoes not, have a standby gas treatment system or equivalent.

ns a result of this review, piping used for pahgan-pling in the hot lah saiajile .rocaa area ~ lead> tokeep operator doses ALARA.

12-3.2 SHIELDING

Thc bases for the nuclear radiation shielding and the shieldingconfigurations arc discussed in this section.

Amendment 5 12. 3-11A August 1981

0

0

pygGS PSAR

RADIATION PROTECTION

DESIGN FKhTtJRES cPOE~< -sl12.3.2.1 Desi Ob 'cctivcs

The basic objective of the plant radiation shielding, in can-/unction m. a pti 'th program of controlLed personnel access to, andoc ancy of, zadiation areas, is to reduce personnel and popu-occupancy 0 g xa 4

lation exposures to levels that are vit2u.n the oseof 10CFR20 and 10CFR50 and are as low as is reasonably.achiev-able (AZd|RA). Shielding and equipmcnt layout and design areconsidcrcd in ensuring that exposures are kept, AZJQVi duringantxcxpa e p'ci ated ersonnel activities in areas of the plant contain-ing radioactive materials, utilizing the design recommendationsgiven in Regulatory Guide 8.8, Paragraph C.2, where practical.Four plant con ons arditi are considered in thc nuclear radxatxonshielding esxgn:ldi d ' normal, full-power operation; shutdown;

~~

~

spent resin transfer; .and emergency operations (for requiredaccess to safety-related equipment).

.44

The shielding design objectives for the plant during normaloperation including anticipated operational occurrences;. foroperation,shutdown operations; and emergency operations are:

A. To ensure that radiation exposure to plant operating~ ~personnel, contractors, administrators, visitors, and

proximate site boundary occupants are ALARA and withinthe limits of 10CFR20.

B. To assure sufficient personnel access and occupancytime to allow normal anticipated aaintcnance, inspec-tion, and safety-related operations required for eachplant equipment and instrumentation area.

C. To reduce potential equipment neutron activation; andmitigate thc possibility of radiation damage tomaterials.

Amendment, 12 12.3-12 February 1984

1i

il

4+ An analysis of the PVNGS shielding design @as performed to1

determine if TMI level source strengths would inhibit mainten-

ance access or violate 10 CFR 50, Appendix h, General Design

Criterion 19 (GDC 19). The review demonstrated that personnel

radiation exposures in vital area , during post-accident~ +he.

activities villaeet the criteria of HUREG 0737 and"GDC-19

design basis.

The design review of plant shielding discussed fulfillsthe NRC reguirements outlined in N0REG+s0578 end 0737 as

iAwell as Regulatory Guide 1'.97s Revision 2-h

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BorzoiAPPENDIX 12A

RESPONSE: The response is given in amended sec-tions 12.1.1;2, 12.5.1.1, figure 13.1-6, figure 13.1-7,13.1.2.2.2.2, and 13.1.3.1;

The Radiation Protection Section is a separate organizationfrom the Chemistry Section, and the Radiation ProtectionSupervisor reports to the Engineering and Technical ServicesManager who is independent of the Station Operations and

IMaintenance Departments. The Radiation Protection Super-visor has direct access to the Manager of Nuclear Operationsin matters relating to radiological protection and ALARA

programs as authorized in the Station Manual. He is a

permanent member of the Plant Review Board.

UESTION 12A.3 (NRC Question 471.2) (12.5)lt-You shoul'd describe your plan to provide backup coverage inthe event of absence of the RPM, and you should outline thequalifications of the individual who will act as the backup.The December 1979 revi'sion of ANSI 3.1 specified that thetemporary replacement fo" an RPN should have a BS degree inscience or engineering, 2 years. experience in radiation protec-tion, 1 year of which should be nuclear power plant experience,,6 months of which should be onsite. It is our position thatthis experiente be professional experience.

RESPONSE: The response is given in amended sec-tion 13.1.2.2.2.2. The minimum requirements for theposition providing backup coverage in event of absence ofthe 'RPM is discussed in section 13.1.3.1.

lm* . ( a '».3) (12.3)

You should provide information in response to TMI LessonsLearned review, for the following NURZG-0737 areas: II.B.2 PostAccident Shielding and Vital. Area Access; II-B.3 ALARA for

Amendment. 5 12A-2 August. 1981

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/pg (

APPENDIX 12A

Post-Accident Sampling; II.F.1 High Range In-Containment Radia-tion Monitors; .D.3.3 Post-Accident Iodine Sam ling andAnalysis ~ lc gQ l$ .1E.6,2~ lS.KLS, IS.X,F. ( cod IS,X' g.y

RESPONSE: The response is provided inAdditional information is provided in sections 12.1.2.4and 12.3.1.3 for'.Items II.B.2 and II.B.3, in section 11.5 andfigure 12.3-4 for II.F.1, and in sections 9.3.2.2.2 and 11.5X'or III.D.3.3.

(12.1)

Paragraph B, of Section 12.1.2.1.2, stating that "as minimum,shielding is designed to reduce gamma dose rates from sourcesexternal to a radioactive compartment to levels comparable todose rates resulting from equipment within that compartment,"is not clear.Et appears that this could refer to shielding between twoadjacent compartments in which radioactive equipment islocated. If this is the case, then 'shielding should bedesigned to reduce radiation from the operating equipment inone compartment to levels below that which is expected in theadjacent compartment from the shutdown equipment to be main-tained or repaired. Please clarify.

RESPONSE: Paragraph B of section 12.1.2.1.2 has beenrevised to provide the requested clarification.

IL * ~ ! Q" (12.1)

In accordance with Section C.2.e, "Crud Control," of Regula-tory Guide 8.8, it. is our position that consideration shouldbe given to the selection of corrosion resistant, low cobaltcontent alloys to reduce the concentxations of radioactivecorrosion product buildup in systems.

Amendment 5

Section 12.1.2.3, "Equipment General Design Considerations forP

AL2QtA," of your FSAR, should be revised to reflect your designconsiderations foz selection of low cobalt alloys.August 1981 12A-3

'

~~

~Ci

lG.II.D.3 DIRECT INDICATION OF RELIEF AND SAFETY-VALVE POSITION

Position

Reactor coolant system relief and safety valves shall be provided

with a positive indication in the. control room derived from a

reliable valve-position detection device or a reliable indicationof flow in the discharge pipe.

PVNGS Evaluation

PVNGS does not utilize power operated relief valves. The PVNGS

primary code safety valves, located at the top of the pressurizer,are headered into the reactor drain tank (RDT) inside contain-ment. Upstream of the common header each code safety valve ismonitored for seal leakage by an in-line resistive-temperature

.Lc,+~ +«device-(RTD) (refer to FSAR Figure 5.1-1)'.

1ndirect indication of code safety valve leakage is provided

by an increase of RDT pressure and a decrease of pressurizer

pressure and,pressurizer level, monitored by safety-grade

instrumentation.<5

Positive indication of safety valve positionwML —be- providediS

in the control room. Monitoring ~~e provided by an acoustic

monitoring system consisting of an accelerometer (acoustic sensor)

mounted downstream of each valve. The sensing instrumentationlg~~e environmentally qualified to function in a post-IOCA

environment xn accordance with Regulatory Guide 1.89. A plantiS

annunciator alarm ~h-be provided to alarm valve opening.i5

The acoustic monitoring system arill-be powered from a reliable

Augus 1981 l8 II.D.3-1 Amen/ment

<0

i

. ~

instrument, bus with Class IE backup power. The system is~e

des'.gned to meet the requirements of Revision 2 to Regulatory

Guide 1.97. ')

Installation of positive Pressurize'r safety valve positioncv 8

indication and development of emergency procedures +4Xk-be.I

completed prior to fuel loading of PVN Unit l.Pqg 65 u.~i~5

Amen nt 1 l6. II.D-3-2 Aug t 1981

0

Ci

II

2286rDffP-CA/ ~ ~PjgBo&gl

SITING AND DESIGN

18.II.F NSTRUMENTATION .AND CONTROLS

18.II.F 1 'DDITIONALACCIDENT-MONITORING INSTRUMENTATION

A human factor analysis a<ill be performed to ensure that thedisplays and controls added for additional-accident xonitor-ing do'not in rease the potential for operator error (see

section+le.gI.D.1). 1nstallation will,be completed prior tofuel load.

18.II.F 1.'1 '+ NOBLE GAS EFFLUENT MONITOR"-,.

Position

Noble gas effluent monitors shall be insta'lied vith an

extended range designed to function during accidentconditions as ve)l as during normal operating conditions.Multiple monitors aze considered necessary to cover theranges of inter'est.

(1) Noble gas effluent monitors vith an upper range capacityof 10 yCi/cc (Xe-133) are considered to be practicaland should be installed in all operating plants.

(2) Noble gas effluent monitoring shall be provided for thetotal range of concentration extending from normalcondition (as lov as reasonably achievable (ALARA))

. concentrations to a maximum of 10 yCi/cc (Xe-133).Multiple monitors are considered to be necessary tocover the ranges of interest. The range capacity ofindividual monitors should overlap by a factor of ten.

VNGS Evaluation

Section 11.5 provides detailed descriptions of the effluentmonitors installed at Palo Verde Units 1. 2 and 3. Thisincludes the additional aonitors that have been added

specifically to address NUREG-0737 and Regulatory Guide 1.97,Rev. 2 requirements for radiation monitoring. A descriptionof the calibration sources, frequency of calibration. and

18.II.F-1 (PROPOSED) ANENDNENT 15

II

0

2286rSITING AND DESIGN

tl'S.-ltechnique is provided in table .. and sections 11-5.2.1.6.2and 11.5.2.1.6, respectively. The instrumentation is describedin detail in table 11.5-1.

Sampling of effluents meets the criteria of 1WSI N13.1-1969 as

discussed in sections 11.5.2.1.1-.7.2 and 11.5.2.2-1..Section 11.5.2.1.1.7..2 also describes the sampling assembly.

, A description of effluent radiation monitoring is presented~

~ ~ ~ ~n section 11.5.2.1.4. Included in this section areiscussions of monitors located on the plant vent. main

axdI 1 * K I il. +These monitors -operate in con)unction'arith

other monitors, as described in section 11.5, and fulfillthrequirements as outlined in NUREG-0737 and Regulatory'Guide 1.97. Rev. 2.

gf~m {~ %sod ~cu-~ mtvu'hlfS

~pyj~~ ovKA{ QPE 5486~<+

tt <.2-.l ~

PROPOSED) AMENDNENT 15 8.II.F-2

0

0

~ ~ PVNGS FSAR

2286re

SITING AND DESIGN

SAMPLING,AND ANALYSIS OF PLANT EFFLUENTS

o

18. II.F. 1. 2

~i~tonBecause iodine gaseous effluent monitors for the accidentcondition are not considered to be practical at this'tiae,capability for effluent aonitoring, of zadioiodines for the

-.accident condition shall be provided vith saapling conducted =

by adsorption on charcoal or other aedia, followed by onsitelaboratory analysis.;

GS valuat on

section 11.5.2.1.1.7.2.iodine channels, the sampler isPour ~shielding is furnisheddetectors..

is discussed in'For particulate an-'

lead-shielde fi,lter assembly,for all process and ef t;luen.o

JgM8St "I~

Airborne particulate and iodine monitors and samplers.{XJ-SQN-RU-08, -RU-14. -RU-141, -RU-142, -RU-143. -RU-144. andXJ-SQB-RU-145 and -RU-146) sample isokinetically in accordancewith the, principles'nd methods of .ANSI N13.1-1969. Guide to

e

Sampling Airborne Radioactive Materials .in Nuclear Facilities.The particulate .and iodine sample flow is maintained constantover the normal expected'ange of filter paper and/or charcoalcartridge differential pressure by an automatic control system.Local flow, indication and high- and low-sample flow alarmsignals are provided. These signals actuate local alaxms andthe channel .failure alarm. Two particulate monitors(XJ-SQN-RU-08 and -RU-14) are. moving paper filtex type andincorporate microcomputer-controlled step advance, and feedfailure channel failure alarm. Sampling assembly fittings axeprovided which allow grab sampling of the monitoredairstreams. The automatic control system maintains isokineticflow based on a comparison of sample flow to HVAC duct flow.The HVAC duct flow input is performed manually for XJ-SQN-RU-08and -RU-14 and .automatically for XJ-SQN-RU-.,141 through -RU-144.and XJ-SQB-RU-145 and -RU-146.

A flow-integiating .elapsed sample volume indicator is provideddownstream of each particulate and/oz iodine channel. It has

a local digital readout and, is zesettable to zero.

li

0

II

~ ~ ~ ~

Cp0

Monitors are designed to meet a 90$ efficie level for partic-ulates and 90$ efficiency for iodine as required by NUREG-0737 I

I

Table II.F.I 2. -They are also designed to conform eith design It

hasis shielding envelopes for sampling media as discussed ingx.

Iection 12.1.2.4 ~ Monitors are designeD*

-'o

allow personnel to remove, replace, and transport, samplingQtmral ~+ip~ Cgr+eon19

aedia without exceeding the criteria of,M99 of 5 rem whole-

body and 75 rem to the extremities.

0

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II

PVNGS PSAR

2286r

18.II.F 1 3

P~ition

. M~+20$ 'J<~

SITING AND DESIGN

CONTAINMENT HIGH-KWGE RADIATION MONITOR

In containment radiation-level monitors vithof 10 rad/hr shall be installed. A ainiaum8

.aonitors that are physically separated shallHonitors shall be developed and qualified toaccident environment.

a aaxiILum rangeof tvo suchbe provided.function in an

This requirement vas revised in the October 30. 1979 letterfrom H.R. Denton to All Operating Nuclear Pover Plants toprovide for photon-only measurement vith an upper range of10 R/hr.7

VNGS Evaluation

As noted in table 11.5-1, in-containment area monitorsXJ-SQA-RU-148 and XJ-SQB-RU-149 are provided to measure

7y-photon activity vith an upper range of 10 R/hr.

(PROPOSED) hMENDMENT 15 18.II.P-4

Cl

0

2286r'~/

gc /fP. lSITING AND DESIGN'.

18.II.F.1.4 'CONTAINMENT PRESSURE MONITOR

Position

A continuous indication of containment pressure shall be

provided in the control room of each operating reactor.Measurement and indication capability shall include threetimes the design pressure of the containmcnt for concrete,foui times the design pressure for steel. and -5 psig for .allcontainments.

VNCS Evaluation

Wide range containment'ressure measurement is provided as

described in section 7.5.1.1.5, and the Appendix 6A responseto Question 6A.14.

1$Wide range containaent preslure acasureaent„ 'rovidedconsisting of redundant pressure transmitters @hose signals of

lscontainment pressure continuously displayed within theA

control room. Continuous recording> 'rovided for one

channel over the entire range of pressure measurement. The

transmitters 'ocated outside of the containment structureh

and ~measure the containment pressure through sensing linespenetrating the containment structure. The range of the system

sjfrom -5 to 180 psig, three times thc containmenth

'ressure.n,re.

The transmitters 'hysically separated, redundant,A

environmentally qualified to function in a post-LOCA environment

in accordance with Regulatory Guide 1.89, and seismicallyqualified to- function during and following an SSE in accordance

with Regulatory Guide 1.100. The safety grade pressure instru-aentation is powered,from redundant Class 1E buses. The

instruacntation is designed to acct Regulatory Guide 1.97, Rev.. 2 ~

l

0

iO

2286r

P VN US R'hAt( ~~~1 W~ g/

SITING AND'ESIGN

18 II.F.1-5 CONTAINMENT MATER LEVEL MONITOR

gasitionA continuous indication of containment water level shall be

provided in the control zoom for all plants. A narrow rangeinstrument shall'e pzovided for PNRs and cover the rangefrom the bottom to the top of the containment sump. A viderange instrument shall also be provi'ded for PWRs and shallcover the range from..the bottom of the containment to theelevation equivalent to a 600,000 gallon capacity. For SMRs,

a wide range instrument shall be provided'nd cover the rangefrom the bottom to 5 feet above the .normal water. level of thesuppression pool.

VNGS Evaluation

Narrow-range water level instrumentation monitoring thera< Wc.sV~containment-ac~~ sumps. and vide-range containment water

level instrumentation are discussed in section 7.5.1.1.5.eve~narrow rope ~~"~

) ns+~n+oPog.

CP(.< n+AiQ&cqY

Continuous indication. of the radwaste mumps (containment

normal aumps) water level is provided in the control room+

E h sump 'onitored from 6-inches above the-'bottomac

of the sump to 6-inches above the top of the sump by a

sensor environment'ally qualified to function .in a,post-LOCA

environment in accordance with Regulatory Guide 1.89- The's5

xnstrumen a iont tation 'owered from a reliable instrument

bus with Class ZE backup power. The instrumentation isdesigned to meet the requirements of Regulatory Guide 1.97,

Rev. 2.

(PROPOSED) AMENDMENT 15 18. I I .F-6

0

~~dc. rciiqa ledt n ~~~+~~ ~ ~ ~

Continuous contro room indication is provided for contain-

ment vater level~from 6-inches above the top of the rad-

vaste sump to 6-inches above the maximum expected flood

level, providing a total range of ll feet. The sensors

physically separated, redundan't,, environmentallyp.

qualified to function in a post-LOCA environment in

accordance vith Regulatory Guide l.89, and seismically

qualified. to function during and following an SSE in

accordance vith Regulatory Guide l.l00. The safety grade

level instrumentation is powered from redundant Class lE

buses. Recording for one channel is provided in the

control room. The instrumentation is designed to meet

Revision 2 to Regulatory Guide 1.97, Rev. 2.

0

0

FVNGS FSAR

2286r

18 ~ II.F.1.6 CONTAINMENT HYDROGEN MONITOR

~sition

~8)Qatar 4 2,Npp

SITING AND 'DESIGN

luA continuous indication of hydrogen concentration in thecontainment atmosphere shall be provided in the controlroom. Measurement capability shall be provided over therange of 0 to 104 hydrogen concentration under b'oth positiveand negative ambient. pressure.

VNGS Evaluati n

A description of the containment hydrogen aonitoring systemis provided in section 6.2.5.2.2.2. The range and accuracyof the hydrogen analyzer is given in table 7.5-1.

The analyzers,ewe ~ore Cah ~,in standby during normal operation and can providehcontinuous indication of hydrogen concentration in less than

30 minutes after activation from the control room. Thecon

ana1ysers ~operate under containment design conditions frosh

5 to 60 psig, the:containment design pressure. The analyzer

are environmentally qualified to function in a post-LOCA environ-

ent in accordance.vith Regulatory Guide 1.89 and seismically

qualified to function during and following,an SSE in accordance

With Regulatory Guide l.l00. The safety grade hydrogen analyzer

instrument channels are powered from redundant Class lE buses.

the instrumentation is designed to meet Revision 2 to Regulatory

Ouide l.97.

l 6~1C CO~+Q PQ,~gg ~~~~ ~ g g ~~~ ~e~j g/Iroy(445 ~fl'Q Q q~+ ~tet+~ go~~~ewer is a.4o-

18. II .F-7 (PROPOSED) AMENDMENT 15

41

rvau> c>~ JW~gz S8ei

CONTAINMENT SYS~e

Tests have verified that the hydrogen-oxygen recombination isnot a catalytic surface effect associated vith the heaters., butoccurs due to the increased temperature of the process gases.ls the phenomenon is not a catalytic effect, saturation of theunit is not predicted to occur. Results, of testing a prototypeelectric hydrogen recombiner and production unit test resultsare given in reference 1. There is no difference between thehydrogen recombiner units to be installed in PVNGS and theunit for vhich the tests vere conducted.

Pl~I'

6.2.5.2.2.2 8 dro en Nonitorin Subs stem. The hydrogenmonitoring subsystem for each unit consists of two completelyredundant 'trains. Each -train consists of a hydrogen sensor,an electronic subassembly and local and remote readout/alarm .

The electronic subassemblies for trains A and B are housedseparately in cabinets located in the. auxiliary building.+A bottled nitrogen and hydrogen supply is used to calibratethe sensors at those intervals specified in section 16.3/4.

Hydrogen measurement is accomplished by using a thermal con-ductivity ce'l and a catalytic reactor. The sample gas firz-flovs through the sample section of the cell, then pass'esthrough the catalytic converter where hydrogen in the sample

t

is catalytically combined vith free oxygen to form vater vapor,then passes through the reference section of the cell. The

hydrogen content is indicated by the difference in thermal con-ductivity between the sample and reference sides of the cell.Oxygen, in an amount sufficient to combine hydrogen at thehighest range of the analyzer, is added to the sample gas, priorte passing through the sample section of the cell.+ The rangeand accuracy of the hydrogen analyzer is given in table 7.5-1.

+A single failure analysis is given in table 6.2.5-2.~ g~ ~ <~~ IZ. II .F. I. ~ ~ 7PlI re .,m~~~~

''Cv

~+~( pand( Mo~<+ ~ ~'n

CM+ol &parJ pro v'des rocoi~a( oK cad ~~ per.g, ~en

L5 a,6~ Y~I'ndment9 6.2'. 5-10 ugust 198m

4i

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SAFETY RELATEDZl

yISPLAY INSTRUMENTATION

fj~

C.

for manual actuation of the containment combustible gascontrol system. Redundant analog instrument channels .

provide thc required informati'on.

Control room indications are provided to allow,theoperator .to monitor and. evaluate the operation 'ofactive system. components during system operation,including periodic tests and thc post-accident period.Table .7.5-1 lists parameters monitored'n this system.

Control of the containment combustible gas controlsystem is local and indication of system air flow andtemperature is prov'ded at the local panel.

Monitoring of Auxiliary Feedwater System

Refer to section 7.3.1.1.10.7. Information is provide"in the control room to allow the operator to monitorand evaluate the operation of the active system:compo-nents during system operation including periodic testsand the post-accident period. Table 7.5-1 listsparameters monitored in this system.

7.5.1.1.4 CEA Position IndicationRefer to CESSAR Section 7.5.1.1.4.

7.5.1.1;5 Post-Accident Monitoring

Refer to.CESSAR Section 7.5.2.5 and Table 7.5-3.

gag~

7.5.1.1-6 Automatic Bypass Indication on a System Level

A status aonitoring panel in the control room displays theavailability of the CESSAR ESFAS, thc onc-out-of-two ESFAS,

all the ESF systems (including the NSSS ESF systems and thecontainment combustible gas control system), and the automatic

October 1981 7 ~ 5-7 Amendme.-.

0

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aJ0;g~ 4 4e insWn enon deceri4~ ln

<~SR C., also ~~iree as pos+ accrete e<

ni~'I nng fns~oioi4 +on . a~ eis4ngeJ deny >o'ale

gas c&lve.nt paciie4on "monifsrliig eon%'n~t~J'ia4oii moni6iing Con&in~t p'c<sii'e

god r 4AL~ Vh4n

c~W".nm~E +ho )e r on'4» n.'

dghalte 0 eel acoss~'on

pnovtde4 'n e e c4on

discossi~ on'><o~ nievi<<~'+g 'ts p~t~+5c++~ (p 2 5 2 72 ~

$ . ti . & I+.TM>~i'~-

~~~ ~hu.94@~

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0

PVNQS FSAR~IS-UW Vf-~~~

.POPS'8)

PROCESS AND. EFFLUENT RADIOLOGICALMONITORING AND SAMPLING SYSTEMS

nless steel parts (bolts oz

Iampler.

Materials for fasteners of stainu.s) and pump wearing parts (wear rings or seals) are auste-nitic stainless steel or other AS'pecifi'ed material suitablefor the water chemistry an'd/or radiation environment of ihe

11.5.2.1.,1.7.2 Sampling Assembly

11.5.2.1.1.7.2.1 Each process or effluent channel includes a

sampling assembly which consists of a sampler and'he associ-ated piping. fittings, and other components as required totransport the sample through the system. The sampling asse-...b.y

is a closed sealed system and includes a sampling pump, valves,interconnecting piping, filters, fittings. flow and pressuretransducezs. and other local control and instrumentation ele-ments as, required. Samplers, with the exception ofXJ-SQN-RU-141 and XJ-SQB-'RU-145 particulate-iodine samplers..house radiation detection equipmen- and check source(s).Sampler piping and connections are welded except where m=xn-

tenance considerations make flanged oz Swagelok )ointsnecessary. Sampler outlet pipin" connections aze located tcmini'mize cleaning requirements and background buildup due tc

~ ~the adherence of radioactive par=icles to the sample wa 's.For liquid samplers', welding of pressure-containing componentsis performed in accordance with ANSI B31.1. For ESF monitors,welding of pressure-containing components i's performed inaccordance with AWS D1.1-1972 (with 1973 revisions). Meldingaf other equipment is performed in accordance with industryStandards.

.11.5..2.1.1.7.2.2 For liqu'd. and process channels, thesampler is a lead-sh elded steel chamber.. For particulate

an''odinechannels. the sampler is a lead-shielded filter assembly.Four ~~shielding is fuzn.'shed for all process and effluentdetectors.

February 1985 11'-5-25 Amendment

0

DgF4Pvip ~+0PROCESS AND EFFLUENT RADIOLOGICAL

MONITORING AND SAMPLING SYSTEMS

A'rbczne particulate and iodine manitors and samplers.(% - .'.:-:-. -"-., -PJ-14, -RU-141.- -RU-142. -RU-143. -RU-144,:andXJ-'R"-145 and -RU-146) sample i'sokinetically in accordancec''- "-.~ p~™ciples and methods of ANSI N13.1-1969 Cuide toSa-..- '..g Airborne-Radioactive Materials in Nuclear Facilities.T."." paz iculate and iadine sample flow is maintained constantove the normal expected range of filter paper and/ar charcoalcar z-'dge differential pzessuze hy an automatic control system.Local flow indication and high- and low-sample flow ala m

signals are provided. These signals actuate local alarms andthe c?annel failu.".e alarm. Two particulate monitors(XJ-SQN-RU-OB and -RU-14) are maving paper filter type andincc:pc=ate microcomputer-controlled step advance. and feedfailure channel failure alarm. Sampling assembly fittings azeprovided which allow grab sampling of the monitoredairstreams. The automatic control system maintains isakinet.flow based on a comparison'f sample flow to HVAC duct flo«.The 1@AC duct flow input is performed manually for XJ-SQN-RU-05and -'RU-14 and automatically for XJ-SQN-RU-141 through -BU-144,and X -SQ='. RU-145 and -RU-146.

A flow-in:egrating elapsed sample vo'ume irdicatoz is provideddown tream of each particulate and/or iodine channel. It has

a local digi.al readout and is resettable to zero.

Par-icu -e collection effici is greater than for0.3m pa"tic lates. Volatile iodi adsorption ef icien isg:ea e than

11.5.2.1.1.7.3 Detector Assembly. The detector assembly isa complete'y veatherpzoofed assembly. housing a detector,preamplif ier. and radiation check source. The assembly isca@a e o vithstanding the design pressure and temperature ofthe piping system of vhich it is a part. vithout leakage,collapse a the tube walls, oz damage to the detector,

The detector assembly is incorporated in the sampler

~ g P «$ ~~ @..|-.12 +'7NT <~«e, acLs

en ment 1 1 5-26 February 1985ieu.@ate co(

0

0

CC4fA «SL

PVNGS FSAR

PROCESS AND fFFLUEM'ADIOLOGICAL

MONITORING A'ID SAMPLING SVSTQ-"-

shut at 4 to 6 psig to protect the monitor from significantpressure.and temperature transients inside the containment.

5 Additional containment isolation valves (refer to section 6.2.4)"

O shut on .CIAS @hen containment pressure reaches the containment

isolation pressu e setpoint. There ore the CB-E monitor isdes'igned to functio properly sub eeuent to an event vhere

pressure is applied. to the sampler piping.

~11.5.'2.1.4 - Effluent'Radiation Moritoring

11.5.2.1.4.1 Condenser Vacuum Pumo/Gland Seal Exhaust CVSEl2

,

CVHR~ XJ-S N-RU-141 and XJ-S N-FJ-142 Monitor. The vacuum

pumps and gland seal condenser remove gases from the secondary

system. The exhaust is continuously monitored for gaseous—

activity resulting from primary-to-secondary system leakage.

The exhaust is continuously and isokinetically sampled forairborne radioactive particulates and iodines. Since theexhaust is piped to its own separate exhaust from the build-ing. no other airborne mon''tors are provided for the t-=binebuilding. The monitor/sampler provides automatic in tiation o.filtration of condenser vacuum pump/gland seal exhaust whenever

the monitor channel is in a HIGH-HIGH alarm condition.

sji4}

.

Sampled a r is pulled from the c"nde"se vacuum pump/gland

steam exhaust p pin'g at concition of 125: and 100 pe="ent. ~

rola ve humi"" ty. -hea-~ - prov d d t~~w-the —tempe a-

-tu"-e-o --thai=to�—137~-and-an-RH-o -70-percen--at--the-inlet-t~-~ the sample pip ng in or e to preven degradation of par ic"-

late and I-131 sample= f='lters due to excessive moisture. The

do~~stream gas cha"ne'ete;-tor is designed to v'thstand these

sample cond'=ions in continuous qe v ce.

A low'nd a high range moni=or is used to cover a range ofeleven decades with one decade of overlap. Particulate/iodinecartridge samples exist in the low and high range mon'jtor

and're

removed for analysis. High rang cartridge samplers are

shielded.

Amendment 14 11.5-40 February 1985

0

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rosser vn s~~w7 rr~z.l +

~rra wQOPQIE)

y(8~

PZp rrs=W

8

c cP(uenl raJ:Hen~aonitors have complete digital readout and control from the Health

Physics Office and the main control room. The high range monitors

automatically switch to a new particulate/iodine cartridge pair

~ ~

~

~

. vhen e c rrent carted e reaches a preset radiation level.ex=ma~4aXa used minimize absorption of noble gases.

A

Samples are preconditioned as n'ecessary to assure accurate

results without damaging the sample assemblies. Each monitor

is controlled by a remote mi'croprocessor. This microprocessor

is linked by a "daisy chain" to a minicomputer which provides

multiple informational displays on request by the, operator.C

A dedicated alarm status line is maintained on the CRT display..

This status line does not move with each change of CRT displays.

Thus .alarms are provided regardless of the status of the

displays in the Health Physics Office and Main Control Room.

Monitors are provided with an open structural construction

that provides for easy maintenance and good heat dissipation.Backup battery poser is provided to assure continued micro-

processor memory during a loss of external poser sources.

Ãultiple detectors are used to achieve the dynamic range

required. Hard copy readouts are available from dedicated

printers in the Health Physics office and the control roomy. ~+~~ g p 4g geek~ bi.f,[..L + MTt<4d.~ ~

~GvX &&8f'll'S .

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~/7I MV tP ~~~PVNGS FSAR

PROCESS AND EFFLUENT RADIOLOGICAL

MONITORING AND SAMPLING SYSTEMS

a HIGH-HIGH dose rate alarm. Redundancy and diversity areprovided by the fuel building ventilation exhaustOU-SQB-RU-145) gas monitor. Refer to section 7.3 for a dis-cussion of the safety function of the monitor.

lI2

11.5.2.1.5.5 Refuelin Area Monitor Channel "A" RMAA

XJ-S A-RU-33 . The monitor is located on a wall overlookingthe refueling cavity where it monitors for a release of activ-ity due to a fuel handling accident in the containment.

12

11.5.2.1.5.6 power-Access Pur e Area Monitors Channels "A"PAPA and "B" PAPB XJ-S A-RU-37 and XJ-S B-RU-38 . The

monitors are located between the containment power-access .purgeexhaust duct, and the refueling purge exhaust duct just outsidethe containment wall. During power operations, these channelsmonitor the duct for airborne radioactivity concentrationswhich could potentially result in an offsite dose exceeding10CFR100 limits. These monitors perform the safety function ofisolating the containment building purge supply and exhaustducts (initiate CPIAS signals) on a HIGH-HIGH dose rate alarm.Refer to section 7..3 for a discussion of the safety functionsof the monitors.

Q lt. 5'.> l, 5.711.5.2.1.6 Inspection, Calibration and Maintenance

11.5.2.1.6.1 Maintenance. Outdoor sampling systems arehoused in outdoor-type weatherproof enclosures. The enclosuresare designed to permit. performance of all control and routinemaintenance and cleaning operations from the front or top ofthe enclosure. Lifting eyes or other devices are provided forhoisting the unit, to facilitate replacement if it is everreguired. Interior wiring is run in conduit to terminal boardsmounted in junction boxes.

12

February 1984 11.5-43 Amendment 12

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One area monitor with a collimating lead shield is mounted

adjacent to each main steam line in the Main Steam Support

Structure approximately one foot upstream of the atmosp ericx t - .

J I 5—dump valves. Refer To igure '3- These monit rs

aeasure direct dose rates from the main steam line to identifyeffluent from the atmospheric dump, main stcam relief valves,

and auxiliary feedvater pump discharge. An extra 2 inches

of shielding is placed on the containment side of the detector

shield. There are a total of 4 detectors vith one remote

aicroprocessor for each 2 detectors. The ion chamber covers

a range from lmr/h'to 10 /hcn

The detector is designed to operate

in a post-accident environmental condition with a background

of lo R/hg.

0

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KVALsQ C u~ ~.'We'zaNdL

APPENDIX 11A

filters and disposable crud filters are related to their,cor-responding input activities provided in table ll.i-2.

RESPONSE

The folloving assumptions vere utilized in establishingSRS output activities in table 11.4-6c

2 ~

3.

4.

Evaporator. concentrates are solidified and storedin the high'ctivity storage area for one month {i.e.,1 month decay) prior to shipment.

Spent resin beads are stored for 6 months prior tosolidification. Solidified resin is stored in thehigh activity storage area for 1 month (i.e., 1 monthaecay) prior to shipment.

t'.artridge filters are solidified'nd stored in thehigh activity .storage area for one month (i.e., 1 ver t):decay) prior to shipment.

K isposable crud filters are stored for one month(i.e., 1 month decay in the high activity storagearea prior to shipment.

QUESTION 11A.12 (hRC No. 460.18) (ll.5)

be have reviewed your submittal dated April 6, 1981, relatingto il41 Action Plans ZI.F.1, Attachments 1 and 2, and ZIZ.D.1.1of hUK.G 0737. be find your information scant and very inade-quate. Please provide the information on these action itemsas required by NUREG-0737; For guidance, you may refer tosubmittals on these action plans for other PWRs such asSan Gnofre, bnits 2 and 3, and Summer Nuclear Station, vhichhave been found acceptable by the staff.

RESPONSE:

Amendment 14

an expanded discussion of noble gai monitoringand effluent sampling, per Attachments 1 and 2 to hl.REG-0737>

1tem 1Z F l~ is provided<8,.w.F. l ~ \ end l8.T.F.1 0e

'a ruary 196 lib-7

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also addresses leak reduction design aeasuresper ?tea III.D.1.1 of MURED„-0737. Seasureaent and testingof covered systeaa villnot take place until the startup ofthese systeas. Accordingly, expansion of the LLER for opera-tional leak reduction testing can not he, provided until .afteretartup.

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PVNGS FSAR

APPENDIX 12A

Post-Accident Sampling; II.F.1 High RangeQ In-Containment,Radia-tion Monitors; and III.D.3.3 Post-Accident IodineAnalysis.

18.X.+. 3, I p.gl. F. i ~~B.i(i.b.3 >.

RESPONSE: The response is. provided in<1 information is provided in sections 12.1 2 4

and 12.3.1.3 for Item . . . , in section 11.5 andfigure 1 , or II.F.1, and in sections . . -.%Mme 11.5

III.D.3 3.. ('

Q * " " ( Q"""'"Q (12.1)

Paragraph B, of Section 12.1.2.1.2, stating that "as minimum,.

shielding is designed to reduce gamma .dose rates from sourcesexternal to a radioactive compartment to levels comparable to

Q

dose rates resulting from equipment within that compartment,"is not clear.It appears that this could refer to shielding between twoadjacent compartments in which radioactive equipment islocated. If this is the case, then shielding should bedesigned to reduce radiation from the operating equipment inone compartment to levels below that which is expected in theadjacent compartment. from the shutdown equipment to be main-tained or repaired. Please clarify.

RESPONSE: Paragraph B of section 12.1.2.1.2 has beenrevised to provide the requested clarification.

~ r

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In accordance with .Section C.2.e, "Crud Contxol," of Regula-tory Guide 8.8, it is our position that consideration shouldbe given to the selection of corrosion resistant, low cobaltcontent alloys to reduce the concentrations of radioactivecorrosion product buildup in systems.

endment 5

Section 12.1.2.3, "Equipment General Design Considerations -forAIdQQ,," of your FSAR, should be revised to reflect your designconsiderations for selection of,low cobalt alloys.August 1981 12A-3

0

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I) .III.D.3.4 CONTROL ROOM HABITABILITYREQUIREMENTS

Position

In accordance with Task Action Plan item III.D.3.4 and controlroom habitability, licensees shall assure that. control room

operators will be adequately protected against the effects of'I

accidental release of toxic and radioactive gases and that the

nuclear power plant can be safely operated or shut down under

design basis accident conditions (Criterion 19, "Control Room,"

of Appendix A, "General 'Design Criteria for Nuclear Power

Plants," to 10 CFR Part, 50).

PVNGS Evaluation

Potential hazards in the vicinity of the site are discussed inFSAR Section 2.2. The operators in the control room are ade-

quately protected from these. hazards,and the release of radio-

active gases as discussed in FSAR Section 6.4. The required

information provided below is in the format suggested by

Attachment 1,to NUREG 0737 Section III.D.3.4.

INFORMATION REQUIRED FOR 'CONTROL-ROOM HABITABILITYEVALUATION

(1) Control-room mode of operation: ,automatic filtered recir-culation with filtered makeup for pressurization for radio-

logical accident isolation. Automatic filtered recircula-tion without makeup for chemical accident isolation.Manual smoke removal mode (operators alerted by smoke

detector).

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(2) Control-room characteristics:h

(a) air volume control room:

1.6] x 10 ft(b) control-room emergency zone (control room; critical

files, kitchen, washroom, computer room, etc.):

140 ft. elevation, Control Bldg.h

(c) control-room ventilation system schematic with normal

and emergency air-flow rates:

see FSAR Figures 9.4-1 and 9..4-2Z'I) goo

normal rate =MKU ft /min

emergency rate = 28,600 ft /min

(d) infiltration leakage rate:

170 ft /min

(e), high efficiency particulate air (HEPA) filter and

charcoal adsorber efficiencies:

HEPA = 99.97/ of 0.3 micron particles

charcoal ) 95% of particulates and Iodines

(f) closest distance between containment and air intake:

150 ft, approximately

jV.III.D.3.4-2

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(g) layout of control room, air intakes, containment

building, and chlorine, or other chemical storage

facility with dimensions:

see FSAR Figures 1.2-7 and 6.4-1

(h) control-room shielding including radiation streaming

from penetrations, doors, ducts., stairways, etc:

Sec %SAN 4.9 H.3(K)

(i) automatic isolation capability-damper closing time,

damper leakage and area:35

closing time = sec.

leakage = zero leakage (by bubble test)

area = 12.) sg ft (.la>pc,sk A0 ~P~)

(j) chlorine detectors or toxic gas (local or remote):

(k) self-contained breathing apparatus availability(number), refer to FSAR Section 6.4.2,.2.2 Item M.

(1) bottled air supply (hours supply)-, refer to FSAR

Section 6.4.2.2.2 Item M.

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(n) control-room personnel capacity '(normal and

.emergency):

(m) emergency food and potable water supply (how many1

days and how many people):

Presently, the PVNGS control room design has

emergency food and water supply for 6 people-

for 7 days (within the closed control room).

(o) potassium iodide drug supply:

Sufficient potassium iodine will be maintained

in a central location at the station to supply

6 persons for 7 days, as noted in FSAR

Section 6.4.4.3.D.

(3) Onsite storage of chlorine and other hazardous chemicals:

NOTE: No onsite storage of liquid or gaseous

chlorine. It, is stored as sodium hypochlorite

(liquid).

(a) total amount and size of container:

hydrogen:

125,000 SCF 6 2200-2450 .psi is stored

in fourteen steel cylinders

i 3 . III.D.3.4-4

0

0

sulfuric acid:

55,000 gal. in five 11,000-gal. tanks

8,000 gal. in two 4,000-gal. tanks

50,000 gal. in two 25,-000-gal. tanks

22,.000 gal. in two 11;000-gal. tanks .

carbon dioxide:

180 tons in four 7 ft. diameter tanks

(b) closest distance from control room air intake:

hydrogen: >600 ft (and obstructed)

sul'furic acid: closest is >500 ft.most are >3000 ft.

carbon dioxide: >3000 ft.(4) Offsite manufacturing, storage, or transportation facili-

ties of hazardous chemicals

(a) identify facilities within a 5-mile radius; None

(b) distance from control room; N/A

(c) quantity of hazardous chemicals in one container;

N/A

(d) frequency of hazardous chemical transportationtraffic (truck, rail,, and barge); N/A, >5 miles

from site.

(3, III.D.3.4-5

(5) Technical specifications .(refer to standard'echnical

specifications)

(a) chlorine detection system:

,a)~ 4 g,c. (4scLL+igg(Youl ckck s.w lMZj.

(b) control-room emergency filtration system includingthe capability to maintain the control-room pres-

surization at 1/8-in. water gauge, verification ofisolation by test signals and damper closure times,

and filter testing requirements:

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J. Design Basis Ten

PVNG" CESAR0 l/7

BABIYABILITY SYSYESS

The control room essential HVAC system shaD bedesigned to zemain functional during and after a safe

'hutdown earthquake (SSE).

Aiz ducts and their supports shall be Seismic .

Category X.

The control zoom normal HVAC system is described insection 9.4.1.Protection of the habitability systems in the control roomfrom wind and tornado effects is discussed in section 3.3.

~ Flood design is discussed in section'3.4. Missile protectionis discussed in section 3.5. Protection against dynamiceffects associated with the postulated rupture of piping isdiscussed in section 3.6. Environmental design is discussedin section 3.11. 'The fire protection system is discussed insection 9.5.1.Codes and standards applicable to the control room emergencyventilation system are listed in. table 3.2-1. The system isconsistent with,the recommendations of Air Moving andConditioning Association (AMCA) standards and NRC RegulatoiyGuide 1.52, except as noted in section 1.8.

6.4.2 SYSTEM DESIGN

6'.4.2.1 Definition of the Control Room Envelo e

The areas, equipment, and materials to which the control zoom

operator could require access during an emergency are shown infigure 7.5-1. Those spaces requiring continuous or frequentoperator occupancy are also shown in figure 7.5-1. > A layoutdrawing and a description of shielding required to maintainhabitability of the control room during the course of postulated accidents is provided in section 12.3.

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HAB>TABILITYSYSTEMSr t~~g~ )~ ~ppro<iveoklg 15O lcd'- -'stem De6.4.2.2 Ventzlatxon

6.4.2.2.1 General Description

ec '. . overall description of the control

C s stem. The system is shown schematically anfigures 9.4-1 and 9.4-2. gu.Fi re 6.4-1 s ows

out, including eth location of potential radiologica. es ect to the control room azr xntakes.+po p

s are located in sectio 1.2hul 1ng'ld'imensions and locations are oca e io

' toxic gas releases are discussoxic 'd inPotential sources o oxic.section 2.2.3.

abitability zone served by the HVAC systemThe volume of the h z ithe isolation mode is approximate yin the emergency mode or e iso a

1.6 X 10 cubic feet.'a for the air purification systemEnviromental design criteria or e

ost limiting, conditions resulting from anyof the postulated design basis accidents (DBA) an on eiura ' 'e latory Guide 1.52, as dis-duration in accordance with egu

cussed in section 1.8.

al h sically separted high efficiency filtrationTwo identical, p ysxca'th harcoal adsorbers are provide o protrains wz c a

o room.axr fflow and reczrcu a eI t d air flow in the control-l. Section 1.8 presentsonents, are listed in table 6.4-

ce to each posit'.on in Regulatorythe system design conformance,to eac po'd 1 52. The seisin.c c ass'ifications of components,~ Gus e

d' 'able 3.2-1.xnstrumentatxon, andd ducting are given xn tab

'M bi I'~ Req e me ntS ".6.4.2.2.2 Component Descript'.on

'r handling unit contains a fan, pa refilter, aTh tx 1 axA refilter, an activated charcoal filter, a

filter, and a cooling coil.. Pneuma

provided for system isolation purposes.

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under the environmental -conditions associated with theP~ p8t<1 +'t+d DBA. ~ p 3.+. ~ TRI ~<a~ snakc'm~oq

G. Ductwork - p~i»nR A "Cgn~l /bc ~ Ha i~ ill+p,et} u < r&vlMN

'hesystem ductwork and dampers are Seismic Cate-gory I. Ductwork is redundant where required toprovide functional 'support to active components inmeeting the single active failure criteria. Leaktightductwork and isolation dampers are provided whererequired to isolate thc system from unfiltered outsideair.In general conformance with Position C.4 of Regulatory,Guide 1.52, accessibility and adequate working spacefor maintenance and testing operations are provided- inthe design and layout of the air purification systemequipment.

I. Control Access Doors

To minimize inleakage, the control access doors areequipped with self-closing devices that shut the, doorsautomatically following the passage of personnel.Alarms are also provided to annunciate if any of thedoors are open after a changeover to emergency opera-tion. Two sets of doors with a corridor between,acting as an air lock, are provided at each of the twoentrances to the control. room and associated spaces.

J. Isolation Dampers

~I4]n 9SSystem isolation dampers are capable of automaticallyclosin n 1 seconds after receipt of an actuationsignal, as verified hy testing. The isolation dampers>are tested as bubble-tight dampers for zero leakage,as part of the manufacturer's test program.

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atmospheric pressure during emergency operation.The control room essential ventilation systemmaintains the same temperature and humidityconditions when operating in the isolation mode.

Safety Evaluation Three

The control room ventilation system is capable cifremoving sensible and latent heat loads .of 981,552Btu/h and 13,830 Btu/h, respectively, which includesconsideration of equipment heat loads and minimumpersonnel occupancy requirements.

The transfer to essential or isolation operation mode

does not create a hazard for CO2 buildup. In case ofemergency operation, there is a supply of outside air ~ w~z',~„~of 1000 ft /min and the long term equilibrium for C02will remain below 1 part .per thousand for up to50-person occupancy. In case of isolation mode opera-tion, where the control room is sealed, the criticallevel of 3g would be reached at the following timesfor the various occupancies:

6 persons 26.2 days

12 persons 13.1 days

30 persons 5.2 days

Safety Evaluation Four'ood,water, medical supplies> and sanitary facilities

are provided for a minimum occupancy of 6 persons for7 days. Storage locations provided ensure that theabove supplies will not be contaminated as a result ofpostulated accidents.

The supply of food and water is sufficient for a pro-longed occupancy since outside supplies can be pro-vided within the 7-day interval.)(indeed<«g w pakass'~ >oaida d.«~ sepal)q

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PVNGS FSARt'7

APPENDIX 6A

I. Position C.4.b

See Regulatory Guide 1.52 paragraph K response insection 1.8.

Position C 4 c

The control building systems do not have permanently-installed aerosol-injection ports upstream of the fan.Instead of the aerosol-injection ports, access panelsvere provided upstream of the fan.

K. Position C.S.d

See Regulatory Guide 1.52 paragraph M response insection 1.8.

Position C.6.a

See Regulatory Guide 1.52 paragraphs N and 0 responsein section 1.8.

I I .I (MI,5 '0.31 (6.e)In your description of the control. room's protective features,provide the time interval between the time the chlorine concen-tration exceeds 5 ppm at the isolation dampers and the time thedampers are completely closed-.

RESPONSE:

0 ~ X,xlS~W B

5 * . ( '5 I'5 . ) (6.4)List the areas, ecpipment and materials to which the controlroom operator has access during emergency operation, i.e.,during the time the control room is serviced by the emergencyventilation system.

May,1981 6A-5 Amendment- 4 /4

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Ituap cheaiatzy control, in kg or aoles, and the 1ocation of.the TSP baskets. Calculate the post-infection 'aaap pa- Prelim-inary staff. calculationa indicate that the long-tera pH shouldhe at least 8.0 to aeet 10 CPR 100 dose guidelines for the DBh

KKkRESPOMSE: The response villbe provided on the CZSSAR

docket. Also refer to the response to gSSTIOR 6A.30(NRC Question 281.S).

~tSA'29 {I.RC 9 t'50.18) (6-4)Provide the follcn6ng information required for the contzol roomhabitability evaluation:

(1) control zoom shielding including. radiation streamingfrom pcnctrations, doors, ducts, stairways, etc.:

(2) Scl f-contained breathing apparatus availability(number)

(3) bottled air supply (hours supply)

(4) control zoom personnel capacity (normal and emergency)

(5) potassium. iodide drug supply

(6) control room emergency filtration system includingthe capability to maintain the control room pressur-ization at 1/8-inch eater gauge, verification ofisolation by test signals and damper closure times,and filter testing requirements.

RESPONSE:

2)

Thc required information istioq 6.4.2.5The required information isof section 6.4.2.2.2.

provided in scc-a 8 )~>> ~ ~

provided in paragraph M

3.) The required information is provided in paragraph M

of section 6 ~4 ~ 2 ~2 ~2

Amendment 7 '6A 14 December 1981

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AIR CONDITIONING, HEATING, COOLING,

AND VENTILATION SYSTEMS

Normal HVAC S stem —Control, Com uter andAssociated Rooms

The control room complex is located on elevation 140'". The

normal HVAC system provided for the control. and computer -room

includes cooling by an air washer (evaporati've) for the outsideair supply which is common for the total control building and

by a recirculating air conditioning system with cooling coil'sserved by the normal chilled water system described insection 9.2.

Heating is provided by the use of electric zone heaters'locatedin the supply air ducts y g ~

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~q ~ + ) ~~ ggg(+~}g( f jQ q ~)~ YYl+Yl+s ~

9.4.1.2.1 Design Bases

9.4.1.2.1.1 Safet Desi Bases. The normal.HVAC systemprovided for the control and computer room has no safety designbases. Protection of the operator from radioactivity and

poisonous gases is described in section 6.4. The isolation istreated as a part of the essential control room HVAC system.

9.4.1.2.1.2 Power Generation. Desi Bases. The normal HVAC

system provided for the control and computer room complex hasone power generation design basis:

A. Power Generation Design Basis One

The normal HVAC -system shall supply conditioned air tothe control and computer room during normal plant oper-ating conditions to provide personnel comfort and tomaintain a suitable operating environment for equipment.

9.4.1.2.1.3 Codes and Standards. The normal HVAC systemprovided for the control and computer room is designed inaccordance with codes and standards set forth in table 3.2-1.

9. 4-10'

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