ML17297A706
| ML17297A706 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/21/1981 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| ANPP-18711-JMA, NUDOCS 8108260173 | |
| Download: ML17297A706 (193) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR;810 8260173'OC ~ DATE:: 81/08/21 NOTARIZED: NO DOCKEi >>
¹ FACILE; STN 50 528 Palo Verde Nuclear StationP Unit 1P Arizona Publ i 0 (LLL5?@
STN 50-529 Palo Verde Nuclear StationP Unit 2P Arizona Publ.i 05000529 STN 50 530 Palo Verde Nuclear StationP Unit 3P Arizona Publi 05000530 AUTH',NAi>>lE AUTHOR AFF ILIAT'ION VAN BRUNTP EI, E' Arizona Public Service Co, REC IP ~ iVAMEI RECIPIENT AFFILIATION TEDESCOPR ~ L>>, Division of Licensing
SUBJECT:
For wards preliminary FSAR sections re radiation prote'ction; Matl will be, incorporated into FSAR Amend 5, DISTRIBUTION CDDEi'OO IS COPIES RECEiIVED;LTR TITLE(! PSAR/FSAR AMIDTS and Rel ated Correspondence j ENCLIZE"'OI>>D NOTES:Standardized Plant ~ 1 cy:C'rimes 05000528 Standardized Plant. 1 cy'.C Grimes 05000529 Standardized Plant.i cy:C Grimes 05000530 RECIPIENT COPIES RECIPIENT COPIES ACTION! '/0 ID CODE/NAMEi LIC LI CENSNG BR ¹3 LA LTTR ENCL" 1
1 0
0 ID CODE/NAME LiIC BR KERRIGANP J
¹3 BC
~ 00 LTTR ENCL~
1 0
1 INTERNAL': ACCID EVAL AUX SYS BR 27 il 1 1 1 1 ~
CONT'YS BR26'HEM EiVG BR 1 1 BR 09 1 1 CORE PERP BR 10. 1 1 EFF TR SYS BR12 1 1=
EMRG PRP DEV 35 1 1 EMRG PRP LIC 36 3 3 EQUIP QUAL BR13 3 FEMA REP DIV 39 1 LE'6 GEOSCIENCES HYD/GEO BR I
LI C QUAL BR MECH ENG BR 28 30.
32",
18 2
2 3
1 1
2 2
3 1
HUM FACT ENG 00 I8C SYS BR L'IC GUID BR 33 MATL< ENG BR MPA 16 17 1
'1 1-1 1 '0 1
1 1
1 OELD 1 0 OP LI C BR 34 1 1 PO'WER SYS BR 19 1 PROC/TST REV 20 1 1 QA BR 21>> 1 1 SSESS BR22 1 1-REAC SYS BR 23>> 1 1 EG L 01 1 1 SIT ANAL BR 2Q 1 1 STRUCT NG BR25 1 1 EXTERNALi: ACRS NRC'DR 02'6 41 1
16 1
LPDR NSIC 03 05 1>>
1 ~
NTIS 1 1 Ug p.y )oA]
REQUIREDt'TTR ~
Q5 ENCL:
5f TOTAL NURBER OF COPIES
r a tl 0
O O MufHIE Mt RUIEVPitt5BI OXSEGM!HW STA. P.O. BOX 21666 - PHOENIX, ARIZONA 85036 August 21, 1981 ANPP-18711 JMA/KWG Mr. R. L. Tedesco Assistant Director for Licensing Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 0)
(Jl Subj ect: Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 Docket Nos. STN-50-528/529/530 File: 81-056-026 G;1.10
Reference:
Your Letter to E. E. Van Brunt, Dated July 31, 1981;
Subject:
Summary of Radiation Protection Working Meeting
Dear Mr. Tedesco:
Please find attached the preliminary FSAR sections which should adequately respond to all open items identified in your referenced letters on radiation protection. This material will be incorporated in FSAR Amendment 5.
If you have any questions, please contact me.
Very truly your E. E. Van Brunt, Jr.
APS Vice President, Nuclear Projects ANPP Project Director EEVBJr/KWG/pc Attachment pool S
cc: J. Kerrigan (w/a)
P. Hourihan (w/a)
A. C. Gehr (w/a) pi~< ),o~ (/(
gfgl Si08260f73 Bf082f!"
n 11' W ~
A III II'I II 4,~
I
Mr. R. L. Tedesco August 21, 1981 ANPP-18711 JMA/KWG Page 2 sate STATE OF ARIZONA )
)
COUNTY OF MARICOPA)
I, Edwin E. Van Brunt, Jr., represent that I am Vice President Nuclear Projects of Arizona Public Service Company, that the foregoing'e document has been signed by on behalf of Arizona Public Service Company with full authority so to do, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.
~u-Edwin E. Van Brunt, Jr.
\ %4U Sworn to before me this day of , 1981.
Not y Public My Commission expires:
r zeP2-
~ ~
4c A
E J Pp;~
+Plllpw ~ z
PVNGS FSAR Liquid samples are taken from both RCS hot legs, containment sumps, auxiliary building sumps and the ESF A&B mini-flow line.
All samples are routed to a liquid input header. After sample selection, isotopic analysis is performed. The sample is then.
depressurized and cooled to allow chemical analyses to be performed. At this point a syringe grab sample can be taken, or the sample can be discharged to the RDT or EDT. Upon completion of the analysis, the source is isolated, and the system is then purged with demineralized water, then nitrogen gas.
Gas samples are taken from containment air via the containment hydrogen control system. Samples are routed to a gas input header. Isotopic analysis is performed then the sample is depressurized and cooled to STP conditions in order to perform 02 analysis. A syringe grab sample can be taken or the sample is returned to the containment. The normal hot lab counting room at the 140-foot elevation in the auxiliary building is shielded to provide low background post accident. The counting chamber can be purged with instrument air or bottled gas. When the analysis is complete, the source is isolated, and the system is purged with nitrogen gas.
Liquid samples will provide information on isotopic content, gross gamma, pH, chloride concentration, dissolved oxygen, dissolved hydrogen and boron. Gas samples will provide information on isotopic content, gross gamma, gaseous oxygen, and hydrogen (from hydrogen monitor of the containment hydrogen control system).
@R(MESH g~~ PQQ,~t Amendment 5 9.3-31A August 1981 08-04-81
gi 8 PVNGS FSAR PROCESS AUXILIARIES
,~ "6 '~,0P 'Q~Q
"'9.3.2.2.2 Post-Accident Liquid samples are taken from both RCS hot legs, containment sumps, auxiliary building sumps and the ESF A8d3 mini-flow line.
0 All samples are routed to a liquid input header. After isotopic analysis is performed. The sample is then.
sample'election, depressurized and cooled to allow chemical analyses to be performed. At this point a syringe grab sample can be taken, or the sample can be discharged to the RDT or EDT. Upon completion of the analysis, the source is isolated, and the system is then purged with demineralized water, then nitrogen gas.
Gas samples are taken from containment air via the containment hydrogen control system. Samples are routed to a gas input header. Isotopic analysis is performed then the sample is depressurized and cooled to STP conditions in order to perform 02 analysis. A syringe grab sample can be taken or the sample is returned to the containment. The normal hot lab counting room at the 140'foot elevation in the auxiliary building is shielded to provide low background post accident. The counting chamber can be purged with instrument air or bottled gas. When the analysis is complete, the source is isolated, and the system 0
is purged with nitrogen gas.
Liquid samples will provide information on isotopic content, gross gamma, pH, chloride concentration, dissolved oxygen, dissolved hydrogen and boron. Gas samples will provide information on isotopic content, gross gamma, gaseous oxygen, and hydrogen (from hydrogen monitor of the containment hydrogen control system).
f)/ flip j;i',:j,j()',,"itj,.,l',
I Amendment 5 9.3-31A August 1981 08-04-81
%g PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS
~ ~
unlikely event that radioactivity is introduced into the control building intake plenum.
D. Provide long-term post-accident monitoring of ventila-tion exhaust from the auxiliary building ESF equipment areas following a loss-of-coolant accident.
+*I E. Inform the control room operator of the occurrence and approximate location of an abnormal radiation increase in a zone adjacent to the containment containing pip-ing, electrical or hatch penetrations. &
Inform the control room operator and personnel in the immediate vicinity of the monitor of an abnormal radiation increase inside buildings where access is required to service equipment important to safety post-accident.
G. Provide long-tenn post-accident monitoring of effluents from the plant vent, fuel building vent, main condenser vent, and the main steam relief and atmospheric dump valves.
11.5.2 SYSTEM DESCRIPTION 11.5.2.1 Continuous Process Effluent and Area Radiation Monitorin and Sam lin The requirements of the system design bases for continuous monitoring are satisfied by an integrated, microcomputer-based system of 52 monitors, including a total of 73 detector chan-nels with their associated sampling and auxiliary equipment.
Section 11.5.2.l.l provides a description of system hardware including design features such as instrumentation, types and .
locations of readouts, annunciators, and alarms, provisions for emergency power'supplies, and provisions for decontamina=
tion and replacement. Section 11.5.2.1.2 provides information 11.5-6 Amendment 5 0?-01-81 August 1981
PVtIGS FSAR lHCKViENT~~.
PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS EOIT SECTION would be the radwaste buildi 'pgjon sys-tems which are provided with 'p,rocessi: '<ljG g g 198 i monitors to monitor for leakage from Sea~ e gas compressors and waste gas decay tank valves.
B- Areas in which the new and spent fuel is received and stored, specifically the containment and fuel building, are provided with detectors which indicate and alarm in the presence of abnormal radiation levels. I C. The location of each area radiation detector is indi-cated on the radiation zoning and access control draw-ings, figures 12.3-1 through 12.3-20.
11.5.1.2 Postulated Accidents The process, effluent, and area ponitoring systems, collectively referred to as the radiation monitoring system (RMS), are designed to perform the following functions in order to meet the requirements of 10CFR50, 10CFR100, and follow the recom-mendations of NUREG 0737 and NRC Regulatory Guides 1.13, 1e97 and 8.12 for postulated accidents:
A. Provide the capability to alarm and initiate contain-0 ment purge isolation in the presence of high airborne radioactivity within the containment which could poten-tially cause an offsite dose in excess of 10CFR100 limits.
B. Provide the capability to alarm and initiate isolation of the fuel building from the normal ventilation system and actuation of fuel building essential ventilation in the unlikely event of a fuel handling accident in the fuel building.
C. Provide the capability to alarm and initiate isolation
\
of the control room normal ventilation system and actuation of control room essential filtration in the August 1981 11.5 5 Amendment 5 07-01-81
LIOe I 0 EDIT SECTION sapctf Irscct lou 4 I ~ ool
~N 004 CCOO Uv 4$ e-'-+- LIISIID I
N-OSS-Stoa.ts Lteselo aao Naac
~ 1 1 I lal TPU T lcaoco I 0'4 I
~ WW ICT AUG 11 1981 Uv sapcfv acct tor ~ ~ >> oSt ccoo Q 4$
I UV Ie 0414COA Q
~ OOI To Cof la ~ ~ I
~ N.OSl-ccoa-ts-L IOII0 OILS SAIPLC 10 Not Las UV IS 5 ~
I'4 41 s N.ots ccoa 0 OC5 NO, I.CC SP OOI I C' I I a
$4 <<.oto-Pcs ~ OOI 10 40T L'V IL I I
$$ CCDAAPCSA
~>> oN cctla
-o pCD a Lc Dolce LINc ool IP ~ ISI coacs
, I CD SC5 I I
I ~ O ~ I atp
+ca acs Uv 44
" Q-g I I
I V04'S Oe
~ t a sr IOS 4 r'
A NA 0 15
~ VILL voas a 1 aSN UV ~ca acs 4$
Ip'ooe
~N OI\ ICSA IOI ccwtallrCNT pcclrc slop 4 I~ ~ ~ I 0J " ic ~ Il ~ 4
~caats voas Uv ool
~r.olt ICSa
~ Q
~4 CoutAIIPKNT aaora5TC suua I C ~ 21 P UV
~5
~N 01$ ICDa 0 WIILIAST OLDC SIWPS ool ID ~ $1 W
t DCulucalaaltco rater w.ool ~ >>.Daa.na-cs. ~
Sw VILL eras 5aelpt 2 I ~
II C crea
<<.014-IODa sstcTrou INOTC It
~ I5 I~ ~ 0 crass Coa 50 Csols
~ W I 4 aura
~N 014 IOOa 50 0u Sw raLL It ap oot Uv ~ I I I I IOD I to Ca as voLINK cor,leoa Ta la m oot
~N 04$ PCDO 6 LP Uv 25 < PCL SUIPLt N.OIT Wa-Ipse
~ ool CataI PKNT alr
~ 1 ~ I I 41 Cr Cs ~
CC~PCsa lofts 5 pa'I NKaj ~ al~~ aragvLj P-Qt
-e Caacs LP W
Sl 9q "'" -- CPACS Volt LAO
~C 'I t vol ~ 012 ID gS art CaaCS sw raLL Su I
I 'I I LCD'ooa TIC STSTCU 055ICNA104 55 15 TO Patctot ALL LINC Vaavt alKI INSIIIUIKIT IAPOCPS t <<s Srou IW TPISDPA ~ lro IPI ts5 OINCPVI55 L ~ I INDICalCO B+- L cases C:-g N ot ~ 0004.4 Tec stparatlou cooup Otslcraloe aarlro Iwacss DTNcprl5c IlolcalcD.
N Is pptctoC ALL I,INC Nvlotps Swww cw Tres 10 cl5 aas Saeptt It ~ I erou cf Lao
~ 1 uv I $$
N 01 ~ PCDA Uv CCDO&a CCWTAIMKNT ala ~ OOI L ~0 1YM=SSP-002, Rev 0 .
IC 41 I~ 04 2 SCOP ICSA 5P OOI Cat Caas supLC
~ ~ ' I TO Nol Las Palo Verde Nudear Generating Station 4al OUTPUT ptaotr FSAR POS~CCrOZm: SmP r
~
Cas INPUT Kaotr 2 SYSTRt4'~~," ".,~ '..',.
~@CHANGE Vtrooa SVPPL I CD
, Figure 9.3-2A August 1981 Amendment 5 08-06-81
Il J
EolT SE~~~O~ PVNGS FSAR'CREMENThL I',UG 14 1S81 PROCESS AND EFFI UENT RADIOLOGICAL 0UX MONITORING AND SAMPLING SYSTEMS
~g concerning redundancy, diversity, and independence of com-ponents. Sections 11.5.2.1.3, 11.5.2.1.4, and 11.5.2.1.5 pro-vide a description of the function and location of each process, effluent, and area monitor. Section 11.5.2.1.6 provides a description of provisions for calibration, main-tenance, and inspection.
Figure 11.5-1 is a basic block diagram of the RMS.
Table 11.5-1 is a tabulation of basic information describing each of the continuous process, effluent, and area radiation monitors and sampler, including monitor location, design environmental limits, design background dose rate, type of monitor and measurement made, sampler and/or detector type, calibration isotope, range of activity concentrations or dose rates to be monitored and expected concentrations or dose August 1981 11.5-6A Amendment 5 6-22-81
I I(), ',"!'""": !( ...,;,vms Fma
.PROCESS AND EFFLUENT RADIOLOGICAL MONITORING,AND SAMPLING SYSTEMS
~h This page intentionally blank Amendment 5 '1.5<<6B August 1981
+~-81
Sampler/Monitor Location Designated Location (Instr. Tag No.} Quantity For Environmental Design Background Detector hc (Par Dwg. Ref.)(4) Per Unit Qua 1 if ication (a) (mR/h Co-60) Sampler Type Tye (b) Med NON ESF MONITORS Essential Cooling Water Auxiliary Bldg. 2.5 Off-Line/Liquid Y Scint (ECW) System Mon).tors (XJ-SQN-RU-2 and XJ SQN-RU-3)
(Figuze 9.2 3)
Steam Generator Biowdow)n Auxiliary Bldg. 2.5 Off-Line/Liquid Y Scint (SGB) Monitors (XJ-SQN-RU-4 and XJ-SQN-RU-5)
Nuclear Cooling Water Outside (Yard) 0.5 Off-Line/Liquid Y Scint (NCW) System Monitor XJ-SQN-RU-6)
Figure 9.2-4 Sheet 1 of 3)
Refer to section 11.5.2.1.1.6.1 for ESF monitors, section 11.5.2.1.1:6.2 for non-ESF monitors.
"I" T scintillation detector coupled vitA photomultiplier tube
"()" - NAI Plastic phos.phor () scintillation detector coupled vith photomultiplier tube "SCA" - Single channel analyzer 4G-M" - Geiger-Mue lier detector "Ion" - Ion Chambe:z 4y-Y4 - Area radia,tion W Time delay from introduction of radiation to sampler until indication of that level, for a clean filter ci
- e. "INSTR": 120 VAC Non IE Instrument Pover
"(M)": Motor: 480 VAC Non IE Power "VITAL 'A'": 120 VAC Vital Instrument Power, Channel "h" "VITAL 'B'": 120 VAC Vital Instrument Pover, Channel "B"
- f. Automatic actions initiated on HIGH-HIGH alarm only. HIGH alarm annunciates but does not initiate a cont
- g. Blower Motor: 480 VAC Class IE Pover, Train 4A4
- h. ~ Blover Motor: 480 VAC Class IE Pover, Train "B" i~~ "LMD4 - less than minimum detectable
- j. Ba-133 is the calibration isotope for I-131.
- k. Kr-85 is the calibration isotope for Xe-133.
- 1. Area monitors are shovn on radiation xone diagrams, figures 12.3-1 through 12.3-20.
- n. Detector and Annunciator are located in Containment. Miczopzocessor is located in Auxiliary Building.
PVNGS FSAR W%NKW PROCESS AND EFFLUENT RADIOLOGICAL SECT(p',
LUI I, MONITORING AND SAMPLING SySTEMS AlJG $ ~ 198 Table 11.5-1 CONTINUOUS PROCESS AND MONITORING (Sheet EFFL351'ADIATION 1 of,ll) )5
Response
Expected .Alarn Tine at Min.
Calibration, Range> Concentrations Setpoint Detect/)e Power Automatic Actions edge) Nuclide (pci/cn ) (uci/cn3) (vci/cn~) Conc. Supply( ~ ) Initiated(<)
-10 LMD(i) 6 Min. Instr. Alarm only.
)s Cs-137 10 2 x 10 1 Cs-137 ss C0-137 10-'-10 1 x105 1 Min. Alarm only.
Ms-137 ss Cs-137 10 6 10 1 2x10 1 Min. Alarm only.
Cs-137 iition.
I action.
I2 August 1981 11.5-7 Amendment 5 6-22-81
Jj
'j I
il I
I j
1
pAf-Qf;EEOC PROCESS AND EFFLUENT RADIOibGC AROIT S'CTIOH E
MONITORING AND SAMPLING SYSTEMS Table 11.5-1 AUG l 41881 S
)Cp 'ND EFPLUE%T RADIATIO BRA '(Sheet 2 of ll) l5
Response
Expected Alarm Time at Min.
Range Concentrations Setpoint Detectable Power Automatic Actions (uci/cm3) (uci/cm3) (uci/cm3) Conc. (d) Supply<e) Initiated(f) 10 -10 2xlo Cs-137 1 Min: Instr. (m) Alarm and divert aux. stm. conden-sate to liquid radwaste system.
5xlo -5xlo 3 x 10 1 Hour Instr. (m) Alarm only.
Cs-137 lo 11-lo e x 1O10 e 1 x 10 8 Hours Alarm only.
I-131 I-131 0" 9 x 10 Xe-133 1.8 x 10 Xe-133 1 Min. Instr. (m) Alarm only.
lo e-lo 1 2xlo Xe-133 1 Min. Instr. (m) Alarm only.
10 -10 LND/1 x 10 2 x 10 1 Min. Instr. Alarm and initi-Kr-85 Kr"85/ ate close of 1.5 x 10 the waste gas Kr<<85 discharge valves.
10-9-10-~ 1.2 x 10 15 Min. Instr. (m) Alarm only.
Cs-137 1D -1D 2.4 x 10 15 Min. Alarm only.
I-)31
-10 1 7.2 x 10 1 Min. Alarm only.
Xe-133 Amendment 5 11.5-8 August. 1981 8-06-81
I I'
r'
~ \\ y I
II I P
t
'l J
i I
PROCES S AND EFFLUENT RAD IOIIOG CALEOIT SKTl0f<
MONITORING AND SAMPI ING SYSTE1I1$ -
6 1 4 198l
)g8)
Table 11.5-1 lN OUT CONTINUOUS PROCESS AND EFFLUENT RADIATION MONITORING {Sheet 3 of ll) I5
Response
Alarm Time at Min.
Range Expected'oncentrations Setpoint Detectable Power Automatic Actions (vci/cm3) (vci/cm3) (vci/cm3) Conc. (~) Supply <<) Initiated (<)
5xlp-115xlp-6 2.4 x 10 1 Hour. Instr. (m) Alarm only.
Cs-137 10 6-10 1 2 x 10 1 Min. Instr. (m) Alarm only.
Xe-133 1 10+4 mR 10 15 mR/h 30 Sec. Instr. Alarm only.
lp 1-10+4 mR 15 mR/h 30 Sec. Instr. Alarm only.
>44 lnR
.5 mR/h 30 Sec. Instr. Alarm only.
lp 1 10+4 mR 5 0.8 mR/h 2.5 mR/h 30 Sec. Instr. Alarm only.
lp-1 lp4 mR 1.5 mR/h 2.5 mR/h 30 Sec. Instr. Alarm only.
5 1(-1 lp4 mR 60 mR/h 100 mR/h 30 Sec. Instr. Alarm only.
1i 10-1 104 mR 1.5 mR/h 2.5 mR/h 30 Sec. Instr. Alarm only.
K lp-l lp4 mR 5 0.5 mR/h 15 mR/h 30 Sec. Instr. Alarm only.
August 1981 -11.5-9 Amendment 5
g ~ ~
~ > u ~
r C
Sampler/Monitor Location Designated Location (Instr Tag Nope ()uantity For Environmental Design Background Detector Ac (Paz Dug. Ref.) a Per Unit ()uslification (a) (mR/h Co-60) Sampler Type Type(b) Mea NON-ESF MONITORS (cont)
Central Calibration Facility Area (CFA) l (Unit 1 Outside (Yard) N/A N/A G M Monitor only)
(XJ-SQN-RU-24)
Controlled Machine Shop Radvaste Bldg. N/A N/A G-M Area (MSA) Monitor (XJ SQN-RU-25)
Sample Room Area (SRA) Auxiliary Bldg. N/A N/A G-M Monster (XJ-SQN-RU-26)
Waste Solidification Raduaste Bldg. N/A N/A G-M System Process Contiol Area (PCA) Monitor (XJ-SQN-RU-27)
(XJ-SAN-RU-28)
PROCESS AND EFFL(JENT RADIOLOGICAL MONITORING AND SAMPLING SYSRW ~E Teble ll l 5 AUG 1 1981 CONTINUOUS PROCESS AND EFFLUENT OUT RADIATION MONITORING (Sheet 4 of. ll) )5
Response
Expected Alarm Time at Bin.
sit Calibration'uclide Range Concentrations Setpoint Detect/)e power Automatic Actions radiac) (uci/cm3) (aci/cm3) (aci/cm3) Conc. Supply(e) Initiated (*>
Co<<60 10 1-104 mRh 0.5 mR/h 15 mR/h 30 Sec. Instr. Alarm only.
Co-60 10 1 104 mR 0.5 mR/h 15 mR/h 30 Sec. Instr. Alarm only.
E 1 104 mR Co-60 15 mR/h 30 Sec. Instr. Alana only.
Co-60 1 104 mR h Variable Variable 30 Sec. Instr. Alarm and initi-ate shutdown of waste solidifi-cation sequence.
fill Amendment 5 11. 5-10 August l981 6-22-81
0 PVNGS FSAR PROCESS AND gP MONITORINg CONTINUO'ADIATION N
Sampler/Monitor Response Location Designated Location Expected Alarm Time at Min.
(Instr. Tag No.) {}uantity For Environmental Design Background Detector Activity Calibration Range Concentrations Setpoint Detectable Power (Pal Dwg. Ref.)(1) Per Unit {}ualification( (mR/h Co" 60) Sampler Type ! Type (b) Measured(c) Nuclide (uci/cm3) (y ci/cm3) ( u ci/cm3) Conc. (d) Supply(e)
ESF MONITORS "A" Contxol Room Venti- Control Bldg.
Envir.)
0.5 Off-Line/Gas Gross P Kr-SS(k) 10 -10 2xlo 3. Min. Vital "A"(
lation Intake (CRUI-A) (Normal Xe-)33 Monitor (XJ-SQA-RU-29)
(Figure 6.4-1)
"B" Control Room Venti- Control Bldg. 0.5 Off-Line/Gas Gross P Zr-SS(k) 10 -10 2 x 30 1 Min. ItBII ("
lation Intake (CRVI-B) (Normal Envir.) Xe-133 Monitor (XJ-SQB-RU-30)
(Figure 6.4-1)
"A" Fuel Pool Area Fuel Bldg. N/A N/A G-M Co-60 10 -10 h 0.5 mR/h 2.5 mR/h 30 Sec. Vital "A" (FPA-A) Monitor (Normal Envir. )
(XJ-SQA-RU-31)
"A" Refueling Machine Containment N/A N/A Co-60 lp-1 lp4 mR 0.5 mR/h 2.5. mR/h 30 Sec. Vital "A" (Normal Envir.) h Area (RMA-A) Monitor (XJ-SQA-RU-33)
"B" gyntainment Build- Auxiliary Bldg.
(Normal Envir.)
2.5 Off-Line/Gas Gross P Kr-SS(k) 10 -10 2xlo 1 Min. Vit:
ing Refueling Purge Exhaust (CBPE-B) Monitor (XJ-SQB-RU-34)
(Figure 9.4-13)
"B" containment Build- Auxiliary Bldg. 2.5 Recirc/Moving Gross P Cs-137 1O '-1O 4 1.6xlo 3.2 K 10 1 Min. Vital ing 'tmosphere (CB-B) (Normal Envir.) Paper Particu-Monitor (XJ-SQB-RU-1) late Filter Cs-137/LMD 2 x 30 (Figure 9.4-12 Cs-137 Sheet 1 of 3) Recirc/Fixed Volatile '-1O 4 y/SCA Ba-3.33 1O 2 x 10 9/ 15 Min.
Charcoal or I-131 2 x ~0 Silver Xeolite Cartridge Recirc/Gas ~ Gross P Kr-SS(k) lo 6-lo 1 2xlo XE-133/ 4 x 0 ~
1 Min.
Xe-la~
2xlo Xe-133
'ital "A" Power-access Purge Auxiliary Bldg. <2.5 N/A G-M Co-60 10 -10 mR 0.75 Sec. "A" Area (PAPA-A) Monitor (LOCA Envir.)
(XJ-SQA-RU-37) 1 mR "B" Power-access Purge Auxiliary Bldg. <2.5 N/A G-M Co-60 lp 104 h 0.75 Sec. Vital "B" Area (PAPA-B) Monitor (LOCA Envir.)
(XJ-SQB-RU-38)
/
August 1981 11.5-11 8-06-81
,Ig l'
'I I
~
t i'
q)
PVNGS FSAR INCPCMrt>TAL PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMCDlr SRTIOff Table ll.5-1 AUG > 4]88~
NUOUS PROCESS AND EFFLUENT ON MONXTORING (Sheet 7 of ll) )5
Response
Expected Alarm Time at Min.
Range Concentrations Detectgfe Power Automatic Actjons (pci/cm3) (y ci/cm3) {act/cm~) Conc. Supply {e) Initiatedl~
10 - 10 R/h 10 mr/h 30 seo Instr. (m) Alarm only.
10 to 3 9 x'10 2 x 10 1 Min. Instr. (m) Alarm - After XE-133 XE-133 proper overlap, shifts to high range on increas-ing radiation level.
N/A N/A x 10 to 10 1 Min. Instr. Alarm - After 10 proper overlap, shifts to low range on decreas-ing radiation level.
10 10 1 Min. Shifts to another sample cartridge.
10 1 Min. Shifts to another sample cartridge.
10 to 3 7.2 x 10 1 Min. Instr. Alarm - After XE-133 proper overlap, shifts to high range on increas-ing radiation level.
10 to 10 1.2 x 10 15 Min. H/A CS-137 August 1981 11.5-12A n nna en< S 8-06-81
lfCp
~ '
4
(.
4l pp E MEN/IIL PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEM)olT SKTIOf)
Table ll.5-1 AUG i 4 )~~i CONTINUOUS PROCESS AND EFFLUENT OUg RADIATION MONITORING (Sheet 8 of ll) )5
Response
Expected Alarm Time at Min.
Range Concentrations Setpoint Detectable Power Automatic Actions (y ci/cm3) (pci/cm3) (pci/cm3) Conc.(d) Supply (e) Initiated(f) 0 to 10 2.4 x 10 15 Min.
I<<131 3 x 10 5 to 10 1 Min. Instr. (m) Alarm - After
,il x 10 proper overlap, shifts to low range on decreas-ing radiation level.
10 10 1 Min. Shifts to another sample cartridge.
10 10 1 Min. Shifts to another sample cartridge.
10 to 3 2 x 10 1 Min. Vital "Bu(") Alarm - Initiate XE-133 Fuel Building Essential Venti-lation (FBEVAS)
- After proper overlap, shifts to high range on increasing radiation level.
N/A N/A 3 15 to 101 1 Min. Vital "B"( Alarm - After 10 y XE 133 proper overlap, shifts to low range on decreas-ing radiation level.
Amendment 5 11.5-12B August l981 8-06-81
I" g
l
It~f'aq',"
PVNGS FSAR IMppF MENTAL PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTER AUG 14 )981 CONTINUOUS PROCESS AND EFFMHHX OUT RADIATION MONITORING (Sheet 9 of 11) )5
Response
Expected Alarm Time at Min.
Range Concentrations Setpoint Detectable Power Automatic Actions (yci/cm3) (vci/cm3) (pci/cm3) Conc.(d) Supply (e) Initiated(f) 10 LMD 10 15 Min. Shifts to another sample cartridge.
10 10 15 Min. Shifts to another sample cartridge.
1 R/hr 2,R/hr 30 Sec. Vital A Alarm only.
to HCAA Vxta 10 > R/hr B (HCAB) 30 Sec. Vital A Alarm only.
PCMA Vxta B (PCMB)
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Instr. (m) Alarm only.
to4 Sec.'0 10 R/hr I II
.1 R/hr .2 R/hr Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr'o
.2 R/hr 30 Sec. Instr. (m) Alarm only.
R/hr August 1981 11".5-12C Amendment 5 )5 8-06-81
I II
~ ~I ' I "g
'I / ~
PVNGS FSAR tNCREMLNTAL PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS~
Table 11.5-1 AUG 3. 41981 CONTINUOUS PROCESS AND EFFLUEV OUT RADIATION MONITORING (Sheet 10 of ll))5
Response
Expected Alarm Time at Min Range Concentrations Setpoint Detectable Power Automatic Actjons (yci/cm3) (yci/cm3) (yci/cm3) Conc.(<) supply(e) Initiated'L<
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr'1 R/hr .2 R/hr 30 Sec. Instr. Alarm only.
to4 10 R/hr (m)'nstr.
.1 R/hr .2 R/hr 30 Sec. (m) Alarm only.
Cg .2 R/hr 30 Sec. Instr. (m) Alarm only.
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr Amendment 5 11.5-12D August 1981 )5 .
8-06-81
~ ~
>'lit N
~
(
PVNGS FSAR i'A t'.MENTAL PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS Table 11.5-1 i4 CONTINUOUS PROCESS AND EFFLUE5fR OUT RADIATION MONITORING (Sheet ll of ll)
Response
Expected Alarm Time at Min.
n Range Concentrations Setpoint Detectable Power Automatic Actions (vci/cm3) (yci/cm3) (yci/cm3) Conc.(d> Supply <<) Initiated(>)
.1 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr - (m) Alarm only.
to4 10 R/hr
.2 R/hr 30 Sec. Instr. (m) Alarm only.
.1 R/hr .2 R/hr "
30 Sec. Instr. (m) Alarm only.
to4 4 10 R/hr
.1 R/hr .2 R/hr 30 Sec. Instr. (m) Alarm only.
to 10 R/hr
.1 R/hr 2 R/hr 30 Sec. Instr. (m) Alarm only.
to4 10 R/hr 0
August 1981 11.5-12E Amendment 5 8-06-81
~,
cs I-
~ 'f 0
I
PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS
- 0 rates, alarm setpoint, provisions for power supplies, and auto-matic actions initiated. EOI7 $ ~g)pg Bases for the ranges listed in table 11.5-1 are as follows: AUG i j ~gg~
A. For process monitors, the ranges include: fN~g
- 1. Maximum calculated concentrations during normal operations and anticipated operational occurrences.
- 2. The highest sensitivity commercially available when purchased in order to detect process system leakage and airborne contamination as early as possible.
B. For effluent monitors, the ranges include:
Maximum calculated concentrations for normal 1
operations, anticipated operational occurrences, and postulated accidents.
- 2. Minimum concentrations that must be detected in order to allow automatic and/or operator actions to avoid exceeding Technical Specifications for the release of radioactivity.
C. For area monitors the range extends from a minimum value of the radiation Zone I upper limit to a maximum of the saturation limit of commercially available ion chamber detectors.
l.. order to satisfy the above criteria, the condenser vacuum pump/gland seal exhaust -gas monitor, 'fuel building vent, and plant vent are provided with overlapping high and low range detectors each with a range of at least eleven decades.
i \
Bases for the setpoints provided in table 11.5-1 are as follows:
A. For ESF monitors and non-ESF process monitors, the setpoint shown is as close to the normally expected concentrations as calculated error and statistical
(
~ considerations will allow, in order to notify the August 1981 11.5-13 Amendment 5
'-14-81
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 4
operator and/or initiate automatic action as soon as possible'in the presence of process system leakage, airborne contamination, or a postulated accident condition.
B. For non-ESF effluent monitors, the setpoint shown is the applicable 10CFR20 limit for the release of radioactivity.
C. For non-ESF area monitors, the setpoint shown is the upper dose rate limit of the zone in which the monitor is. located....,
11.5.2.1.1 General Description Refer to figure 11.5-1. In order to optimize the reliability, flexibility, maintainability, and detection sensitivity and accuracy of the RMS, it is completely digital in nature (except for backup analog recording of ESF monitors).
~ I ~ ~g ol V * ~
11.5.2.1.1.1 " Field Unit.'t each location where radiation is sampled and/or monitored by one or more channels, the sampling/
detecting/auxiliary equipment; along with the control microcom-puter, is a single assembly and is referred to as a "field unit". Each field unit is separately powered and is capable of automatic continuous stand-alone operation, even if communica-tions with the control room are interrupted. This capability includes: r A. Acquisition and storage of radiation levels and any other information normally stored within the field unit microcomputer. Data storage capacity for this purpose provides for complete storage of the preceding twenty-four 10-minute averages, twenty-four hourly averages, and twenty-four daily averages of radiation level and the complete files of critical parameters (setpoints, conversion constants, etc.) for all channels within the field unit.
11.5-14 9-11-79
PVNGS FSAR iNCREMEN1 AI, PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS
$ 0lT SEOTlOH (0 B. Control of field unit functions no the field unit microcomputer.
, y AUG i." ~0~>
A plug-in receptable to which can be attached por- OUT C. a table indication and control {PIC) unit or post-accident monitor portable indication and control
{KEPIC) unit to provide complete local control capability.
The PIC unit is a rugged, lightweight, self-contained electronics package, suitable for handcarrying. The unit includes sufficient digital readouts, handswitches and circuitry to provide the following capabilities to the operator when the PIC unit is plugged into any RMS field unit microcomputer.
When a key-operated LOCAL-REMOTE switch located on the. PIC unit is in the REMOTE position, the PIC operator is able to request and receive indication of the currently stored value of 1) any critical
(
~ parameter, or 2) radiation level information, for any channel operated by the connected microcomputer.
I
- 2. When the PIC LOCAL-REMOTE switch is in the LOCAL position, the PIC operator is able to completely control all functions of the connected microcom-puter normally exercised at remote display and control units. Whenever the LOCAL-REMOTE switch is in LOCAL, the affected microcomputer is auto-matically disconnected electrically from REMOTE indication and control.
D. For each area monitor, a permanently mounted local indication and control {LIC) unit or post-accident monitor 'local indication and control {KELIC) unit.
When the channel is operating, this unit indicates radiation level in mR/H and visually and audibly (0 annunciates the presence of a HIGH-HIGH 'level alarm.
August 1981 11.5-15 Amendment 5 y
6-22-81
. PROCESS AND EFFLUENT RADIOLOGICM
~:~ r~~gp 11.5.2.1.1.2 gp g~
MONITORING AND SAMPLING SYSTEMS
'Communications and Remote Indication and 4i Control 11.5.2.1.1.2.1 ESF Monitors. Each ESF field unit has its own dedicated cable over which it communicates digitally with a dedicated microcomputer-controlled Remote Indication and Con-trol (RIC) module located in one of the RMS control zoom cabi-nets. Complete remote indication and control of each safety-related field unit, is exercised at the RIC. Each RIC automatically outputs a signal to the balance-of-plant ESF .
actuation system whenever a HIGH-HIGH radiation level setpoint is exceeded at the field unit. In this way each ESF monitor is completely independent from any other monitor.
11.5.2.1.1.2.2 Non-ESF Monitors. Most of the non-ESF field units are connected together for communications in a single "daisy-chain" loop configuration along with the centralized microcomputer-controlled Display and Control Unit (DCU). The DCU console is located in the Radiation Protection office. The DCU microcomputer and each of the field units are nodes in the loop. Each field unit in turn:communicates digitally with the DCU by passing messages from node to node around the loop.
The equipment in the loop is designed such that field unit-DCU communications is single-failure-proof to industry standards.
In a parallel, identically configured loop with the DCU are two interface modules, located in the ESF control room cabinets (one in the "A" cabinet, and one in the "B" cabinet), which are connected, through qualified isolation devices, to each ESF monitor. These interface modules cannot interfere with the :-.
operation of the ESF monitors, but are automatically fed radiation level and alarm status data for each safety monitor.
In this manner, complete radiation level and alarm status is normally available for display at the DCU for both ESF and non-ESF monitors. >
Amendment 5 11.5-16 August 1982.
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAIBEM~N~AIc:
MONITORING AND SAMPLING SYSTEMS The PCA monitors reguire local indication only and are not con-PKI',II'JI,!I'~
General S stem Performance '3,.5.2.1.1.3 QlJf 11.5.2.1.1.3.1 Measurement Capability. RMS process channels have the capability to measure and display radiation levels over a five .decade (at least eleven decades for the condenser 5
vacuum pump/gland seal exhaust monitor, fuel building vent, and plant vent.) range specified in table 11.5-1 for each channel. RMS area channels measure and display radiation levels over a five decade range, from 10 -1 to 10 mr/h for normal operation and 10 -1 to 10 4 R/h for post-accident 5 operation.
A. For calibration purposes, minimum detectable levels (lower limit of range) for all process channels are detected under sample conditions-of standard tem-perature and pressure.
B. Minimum detectable levels for all process channels are detected in the presence of a Co-60 external radiation field of a level specified for each process channel xn table 11.5-1.
C. Minimum detectable levels for all process channels are detected at a minimum statistical confidence level above background of 95 percent.
D. Response times are as follows:
- 1. For area channels, the time delay between actual
'introduction of the minimum detectable level of dose rate at the detector and remote display and/or local indication of this level is less than thirty seconds.
August 1981 11.5-17 Amendment 5 6-22-8l
PVNGS FSAR PROCESS AND EFFIUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS
- 2. Foi Hcjuxd and gas channels, the time delay between actual introduction of the minimum detectable level of activity concentration to the sampler and remote display and/or local indication of this level is less than one minute.
- 3. For particulate and I-131 channels, the time delay between actual introduction of the minimum detectable level of activity concentration to the sampler and remote display and/or local indication of this level less than the interval specified for each of these channels in Table 11.5-1.
- 4. Response time of every channel is approximately inversely proportional to radiation level at the detector for levels greater than the required minimum detectable level, down to a minimum response time of 0.5 seconds for all channels.
11.5.2.1.1.3.2 System Accuracy A. For area channels, displayed dose rate level is accu-rate to within +20 percent of the actual dose rate level at the detector through the range specified above and through an incident radiation energy range of .1 to 3.0 MeV. Above 3.0 MeV displayed dose rate
, is accurate within -20 percent, but may read higher than +20 percent accurate,. up to 7.0 MeV.
B. For process channels, displayed activity concentration level is accurate to within +25 percent of the actual activity concentration level present in the sampler through the range specified above.
11.5.2.1.1.3.3 Non-Saturating Design. Each RMS radiation channel. has a non-saturating design so that it indicates a level higher than its design range upper limit when exposed to a radiation level up to 100 times this limit.
11.5-18
PVNGS FSAR ,]~ggpg",pY, t SiL PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.1.4 Performance of ESF Monitors. E S tl channel performs the following specific functi A. Provides continuous remote indication at the control room cabinet of current radiation level and channel jgg)J status. . ~rot gl'f~
B. Provides upon demand remote indication at the control room cabinet of the currently stored value of any channel critical parameter. Refer to sec-tion 11.5.2.1.1.5.3 for the list of critical parameters.
C. Provides complete remote manual control of the channel functions specified in section 11.5.2.1.1.5.6. This control is exercised through operation of the RIC unit.
Provides continuous analog hardcopy recording of chan-nel radiation levels for "A" channels only. This is accomplished by processing digital outputs through digital>>to-analog converters to an analog multipoint stripchart recorder.
Provides an output control signal to the BOP ESFAS whenever a HIGH-HIGH radiation level or channel failure is present. Each of these signals is brought to a relay with open to alarm, fail open contacts connected to terminal strips located in the associated control room cabinet.
Provides the following isolated digital outputs, wired to the interface module referred to in sec-tion 11.5.2.1.1.2.2:
- 1. Current radiation level
- 2. Channel alarm status
< l><i4)fVr~
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAI MONITORING AND SAMPLING SYSTEMS 11.5.2.1.1.5, Performance of Non-ESF Monitors. Non-ESF equip-ment performs. the following specific functions:
Provides continuous video displays upon request at the com-munication console in the control room and the DCU in the radiation protection office of the associated Unit. These displays provide real-time information to help the operator to monitor and control radiation levels within and effluent from each Unit.
Refer to 11.5.2.1.1.7.9. Normally the display blanks auto-matically after a preset adjustable time interval if no com-mands are given. Displays appear upon demand from the associated keyboard and include the following:
11.5.2.1.1.5.1 Display I. This display indicates simulta-neously the operational and alarm status of every channel (except the PCA and FTA monitors) in the associated Unit.
Specifically, status conditions indicated by Display I for each associated channel include:
- 1. Channel operation normal (free from failures) in remote control, alarms not inhibited, check source not activated, radiation level less than both high level trip points, radiation level rate of increase less than trip point.
- 2. Channel or attendant auxiliary equipment failure, including loss of external power, failure of the adjacent interconnecting cable, or deliberate transfer to ZOCAL control. Refer to sec>>
tion 11.5.2.1.1.1.
- 3. Channel alarm inhibited.
- 4. Channel check source activated.
- 5. Channel radiation level above the HIGH trip point.
- 6. Channel radiation level above the HIGH-HIGH trip point.
Amendment 5 11.5-20 August 1981 6-22-81
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGKAR~M~~~~~
MONITORING AND SAMPLING SYSTEMS keyboard. following remote manual controls are accom-
g)
~
~ The 5 plished via either of the DCU keyboards:
~ ~
K.LI s A.
B.
Remote manual activation of the check so Remote manual filter step (if applicable).
~ ~ ~
paper filters are operated in the fixed/stepped mode.
~,
Moving .
lg Jiff~
C. Remote manual startup or shutdown of detectors and/or sampling auxiliary equipment for maintenance.
D. Remote manual acknowledgment'of radiation level HIGH, HIGH-HIGH, or HIGH RATE alarms.
E. Changes to the file of critical parameters for each channel, including addition or deletion of parameters and changes to parameter values.
A key lock switch, which assures positive administrative control over access to these remote manual controls, is mounted on each DCU keyboard (control room and radiation
)s 11.5.2.1.1.5.6 Annunciation and Recording. Provides, on a real time basis, annunciation and chronological record logs of all events which affect the status of non-ESF RMS equipment.
A change in the status of any channel in the associated Unit causes the selected display to flash an alpha-numeric message "CHANGE OF STATUS" until the operator acknowledges the message from the keyboard. Each DCU keyboard also outputs an audible alarm, until acknowledged at the keyboard, whenever any chan-nel changes its status to status conditions 5, 6, or 7.
The control room DCU typer .acts as an automatic status logger for all RMS channels. The following occurrences are logged, along with date and time of occurrence.
A. A change in status, including channel identifier and an alphanumeric description of the change in status.
(0 August 1981 11.5-23 Amendment 5 6-22-8l
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS B".WIPIExercise
"'.;~'if)<If of ll,lilt/>lfjf a remote manual control as defined above, including the channel identifier and an alphanumeric description of the control function.
C. A change in the operational status of the DCU or of the REM system. Refer to section 11.5.2.1.1.8.
The capability is provided in the REM system for automatic update, upon restoration, of all normally logged events which occur during a temporary (up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) shutdown for mainte-nance, or malfunction, of the typer.
Each DCU transmits via data link to the REM system the follow-ing data:
A. Hourly averages of radiation level inputs from all channels.
B. Time periods for vhich the above data is of question-able validity or not available.
This data vill be automatically transferred to disk memory and eventually to large capacity magnetic tape for long-term storage.
11.5.2.1.1. 6 Environmental uglification 11.5.2.1.1.6.1 Each ESF monitor is environmentally qualified in accordance with section 3.11 per location as specified in table 11.5-1.
11.5.2.1.1.6.2 ESF monitors and RMS control room cabinets are designated Seismic Category I and are seismically qualified as described in section 3.10. This equipment is also classified as Quality Class Q, and a quality assurance program has been implemented for this equipment.
11.5.2.1.1.7 Desi n and Fabrication Details 11.5.2.1.1.7.1 Materials of Construction. Materials of construction for pressure-bearing surfaces wetted by process Amendment 5 11. 5-24 August 1981 6-22-81,
PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND S I f
A preamplifier is provided integral with the d KltOh reliable transmission of a high-quality signal to the channel g $
microcomputer. The overall response is fast enough to resolv406 ~ g Q8) iver 0 a full-height pulse pair spaced 1 microsecond apart. u<~UT Scintillation detectors are beta- or gamma-sensitive for analysis of photopeaks up to 3.0 MeV. Photo-detectors'uitable multipliers are shielded to prevent gain changes due to orienta-tion or stray magnetic fields. The case is made of a corrosion-resistant material suitable for continuous outdoor operation.
Geiger-Mueller detectors are halogen quenched beta-gamma sensi-tive tubes of sufficient size and wall thinness to detect the minimum specified dose rate.
Ion chambers used inside the containment are fully qualified to post LOCA conditions in accordance with IEEE 323-1974 and IEEE 344-1975. Ion chambers used outside the containment are qualified to the zone they are placed in and cover the full range of P-y dose for that area.
11.5.2.1.1.7.4 Motors 11.5.2.1.1.7.4.1 The liquid sampler pump (RTI monitor only) and each airborne sampler blower motor is a standard 460 vac,
'I three-phase motor.
11.5.2.1.1.7.4.2 Motor controllers include auxiliary contacts for use in the failure circuit of the respective radiation monitoring channels.
11.5.2.1.1.7.4.3 Remote motor control is provided via each DCU Keyboard for non-ESF field units and via each RIC unxt for
~
ESF field units. Local motor control switches and indicating lights at the field unit are provided.
11.5-27
. August l981 07-01-S1 Amendment 5 r
~ ~ I )5 f~g '
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.1.7.5 'ower Supply 3 11.5.2.1.1.7.5.1 Electrical power having the following char-acteristics is made available as listed in table 11.5-1 for each monitor.
~ Grounded, 60 Hz, 3-phase, 480 V-ac +10/ {for motors only)
~ Grounded, 60 Hz, 1-phase, 120 V-ac +10/
11.5.2.1.1.7.5.2 Loss of Electrical Power Failure and Recover Modes. To protect data normally stored in field unit microcomputers, minimize system downtime following power fail-ures, and, in general, to minimize the impact of power failures on the real-time performance of the RMS, field units and the DCU have the following characteristics:
A. A loss of power lasting for a period of 50 milliseconds or less does not affect the performance of any monitor or DCU.
B. For a short<<term {less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) sustained loss of power to a field unit, RIC, or DCU microcomputer, the affected microcomputer:
- l. Senses the impending loss of power and conducts an orderly shutdown,
- 2. Upon restoration of power, restarts immediately and automatically, except {for DCU) manual entry of date/time, and functions normally without loss of any data stored in memory at the time of power loss.
C. For a long-term (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or greater) sustained loss of power, loss of data stored in read-write memory can occur. For field unit microcomputers, sufficient program is located in read-only memory circuits, so that, upon subsequent restoration of power, recovery to full 11.5-28 Amendment 5 August 1981 07-01-81
rREMEN AL PROCESS AND EFF T RADIOLOGICAL MONITORING functional status can be effected either remotely at an gg SLCTlof1 operable DCU in the associated unit, or, if local "":""~
control is taken, at the field unit itself using a PIC AUGQ i. 83 (or KEPZC) unit. For DCU microcomputers, sufficient 'll~
Ig program is located in read-only memory circuits so that upon subsequent restoration of power, recovery to full functional status of any DCU is effected automati-cally, except for manual entry of date/time, by the DCU requesting a program and data download from the RMS system.
11.5.2.1.1.7.5.3 Battery power supplies provided for read/
write memory are rechargeable and provide a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of backup power to the read/write memory.
11.5.2.1.1.7e6 Output Relays. Alarm output relays are pro-vided in the field unit for the RTI, CVSE, MGDT, and PCA non-ESF monitors. They are also provided in the RIC unit for the CRVI-A, CRVI-B, FPA-A, FBVE-B; PAPA-A, and PAPA-B ESF monitors which actuate BOP ESFAS circuits. These relays initiate the automatic control actions listed in Table 11.5-1. Each of the relays has fail open contacts. The relays are deenergized in the presence of a HIGH-HIGH alarm, regardless of whether the monitor is in LOCAL or REMOTE control.
11.5.2.1.1.7.7 Primary Coolant Activity Monitoring Subsystem (PCAMS) Interface Unit. This unit includes analog-to-digital converters, a microcomputer, PIC receptacles, and other equip-ment necessary to receive the following signals from each of the two channels of the primary coolant activity monitoring subsystem (described in CESSAR Section 9.3.4) and communicate them to the associated non-safety communications loop for DCU display.
~ Analog level (0-10 Vdc) input a HIGH radiation alarm contact input 11.5-29 August 1981 Ov-O1-81 Amendment 5
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEM S This interface cabinet,.
unit is located in the non-ESF RMS control room O
11.5.2.1.1.7.8 ESF Monitor Interface Units. Each of these two units includes isolation devices microcomputer(s), PIC receptacle(s), and other equipment necessary to receive the following digital information from each of the ESF channels and communicate them to the non-ESF communications loop.
e Radiation level
~ CHANNEL FAILURE alarm
~ CHANNEL TEST signal
~ HIGH radiation alarm
~ HIGH-HIGH radiation alarm These interface units are located, one each, in the CHANNEL "A" and CHANNEL "B" sides of the ESF RMS control room cabinet.
11.5.2.1.1.7.9 Display and Control Unit (DCU). The DCU includes a microcomputer with sufficient auxiliaries to control the following DCU peripheral devices:
Two video monitor CRTs with keyboards Associated controllers and equipment required for communication with the associated communications loops.
The DCU includes a hardware data link which can trans-mit information both to and from the REM system.
A status logger typer with a keyboard for the control room communications console.
O; 11.5-30 Amendment 5 August 1981 07-01-81
PVNGS FSAR lXGREMEBTAL PROCESS AND EFFT ICAL NONITORINQ AND glTK" for the control room DCU are arranged in a group. gj
"-'omponents
~'.. %81 The CRT and keyboard and the typer are mounted as a group in OUT the communications console in the main control board area. ))~
l~
11.5.2.1.1.7.10 ESF RMS Control Room Cabinet. The RIC units output relays, digital-to-analog converters and recorders are mounted in a control room cabinet assembly. The cabinet also houses the ESF monitor interface units (refer to sec-tion 11.5.2.1.1.7.S). The freestanding NEMA 12 panel assembly is divided into two halves, one for channel "A" equipment the other for channel "B" equipment, separated by a fire-proof wall. Cabinet cooling is provided independently to each side.
The cabinet and its contents are designated seismic category I and are designed, constructed, and anchored accordingly. This cabinet is located adjacent to the non-ESF RNS control room cabinet away from the main control hoard area.
11.5.2.1.1.7.11 Remote Indication and Control (RIC) Units.
Each RIC unit is of modular design and rack mounted. It is microcomputer controlled and connected by standard dual twisted wire pairs (TWP) to its associated field unit micro-computer. The RIC unit includes a front panel with sufficient digital readouts, handswitches and circuitry to satisfy the functional requirements of section 11.5.2.1.1.4. The RIC unit provides IEEE 323 qualified data recording.
11.5.2.1.1.7.12 Digital-to-Analog Converters. A digital-to-analog signal converter is provided with each ESF channel.
This converter continuously converts a digital output from the channel RIC to an analog signal suitable for input to the multipoint recorder.
August 1981 11.5-31 Amendment 5 g-22-81
PVNGS FSAR "l5-"jI> p'~ltd~ f(pic
~ ., )
PROCESS AND EFFIUENT RADIOIOGICAK MONITORING AND SAMPIING SYSTEMS The converter output signal has sufficient resolution so that recorder accuracy is within +5 percent with respect to the radiation level indicated at channel RIC unit.
11.5.2.1.1.7.13 Multipoint Recorder. The multipoint recorder meets the seismic category I requirements except that the recorder output does not meet accuracy requirements during a seismic event. The recorder is not qualified to IEEE 323 since data storage is provided at the RIC's (refer to section 11.5.2.1.1.7.11). The recorder has adequate capacity to record all "A" safety channels. Recorder accuracy for each channel is within +5 percent with respect to the radiation level indicated at the channel RIC unit. The recorder incor-porates selectable channel indication, selectable channel.
printout, and selectable chart and print speeds.
11.5.2.1.1.7.14 Non-ESF RMS Control Room Cabinets. This cabinet is located adjacent to the ESF RNS control room cabinet.
It houses the primary coolant activity monitoring system inter-face unit and the microcomputer, display generator, and commu-nications hardware of the communications console in the control room. This cabinet is designated seismic category I and is designed, constructed, and anchored accordingly.
11.5.2.1.1.7.15 Portable Monitor Connection Boxes. At eighteen locations throughout the Unit, portable monitor con-nection boxes are hardwired into the non-safety communications loop. These connection boxes contain receptacles for plug-in jacks attached to portable area monitors and a movable airborne monitor, any of which can be temporary connected to the loop=
for remote display and control during in-service inspection and other radioactive maintenance. When no jack is connected to=a box, it is designed to appear as a short in the communications loop.
Amendment 5 11.5-32 August 1981 6-22-81
PVNGS PSAR 00 PROCESS AND E AL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.1.7.16 Portable Area Monitors. Each of the three portable area monitors includes a single area channel. The complete detector assembly and electronics package is mounted together and is capable of being hand carried. The detector is connected to the electronics package and is detachable for separate mounting. The electronics package is connected to EO)T SKTIOf<
plug-in jacks.
<UG l 41981 11.5.2.1.1.7. 17 Movable Airboine Monitor. The movable lN OUy airborne monitor includes a particulate channel, an I-131
'channel, and a gas channel. The complete sampling assembly, detectors, auxiliary equipment and electronics package is mounted together on a cart. The electronics is connected to plug-in jack(s).
11.5.2.1.1.8 DELETED Amendment 5 11.5-33 August 1981 8-14-81
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.2 Redundancy, Diversity, and Independence 11.5.2.1.2.1 In order to satisfy the design bases for postulated accident conditions defined in section 11.5.1.2.2, ESF channels are designed to monitor area and/or airborne radiation levels at key locations and provide alarm outputs to the balance of plant engineered safety features actuation system (BOP ESFAS). In keeping with this function, ESF chan-nels are designed in accordance with IEEE Standards 279, 308, 323, 336, 338, 352, 379, 383, and 384, except as noted in sec-tion 11.5.2.1.1.7.13.
11.5.2.1.2.2 ESF monitors and their associated RIC units form independent and redundant sensor channels A and B for three safety feature actuation signals, CREFAS, FBEVAS, and CPIAS as follows:
CRVI-A: CREFAS - channel A CRVI-B: CREFAS - channel B FPA-A: FBEVAS - channel A FBVE-B: FBEVAS - channel B PAPA-A: CPIAS - channel A PAPA-B: CPIAS << channel B 11.5.2.1.2.3 Sin le Failure. Each pair of ESF monitors is designed to accommodate any random single failure without precluding the initiation of a safety features actuation signal August 1981 .11.5-34 Amendment 5 8-06-81
PVNGS FSAR g 'IF,N "~L PROCESS AND EFFLUENT RADIO MONITORING AND SAMPLING SYSTEMS 11.5.2.1.3.13
~ ~ ~ Containment Buildin Atmos here Monitor
~
ply pg,g;.I Channel "B" CB-B ~ The containment building atmosphere .is AUG continuously monitored for particulate, iodine, and gaseous
~
activity. The sample is drawn from the containment building ~ JT~
in a closed system, is monitored outside the containment, and then is returned to the containment building atmosphere after it. passes through the samplers. The particulate and gaseous channels serve as two methods of RCPB leak detection in accor-dance with Regulatory Guide 1.45. This monitor is designated seismic Category I.
In addition to the three radiation channels a hygrometer which measures dewpoint temperature is provided for the CB-B monitor.
The hygrometer is microcomputer-controlled and is configured to be an extra channel of the CB-B monitor. The hygrometer channel measures dewpoint temperature over the range of 0-200F. Indi-cated dewpoint is accurate to within +.4F with respect to actual dewpoint of the sampled air at a rate of change of dew-I point up to +3F/sec. This channel can he used as a method of RCPB leak detection.
In order to obtain a representative sample of containment air, the sample line inlet is located on the operating level between two of the normal cooling units intakes. This location also facilitates RCPB leak detection by these monitors.
Available monitor sensitivities allow the particulate and I-131 channels of the CB-B monitor to detect maximum permissible concentrations allowed by 10CFR20 in the containment building within one hour for Cs-137 and within eight hours for I-131.
The CB-B monitor is located just outside the containment build-ing. It samples the containment atmosphere through piping penetrations. Isolation valves at the monitor automatically
(
4 September 1980: 11.5-39 Amendment 2 r ~ t< ~
PVNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS shut at 4 to 6 psig to protect the monitor from significant pressure and temperature transients inside the containment.
Additional containment isolation valves (refer to section 6.2.4) shut on CIAS when containment pressure reaches the containment, isolation pressure setyoint; Therefore the CB-B monitor is designed to function properly subsequent to an event where pressure is applied to the sampler piping.
11.5.2.1.4 Effluent Radiation Monitoring 11.5.2.1.4.1 Condenser Vacuum Pum Gland Seal Exhaust CVSE 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 for airborne radioactive particulates and iodines. Since the exhaust is piped to its own separate exhaust from the building, no other airborne monitors are provided for the turbine building. The CVSE monitor/sampler provides automatic initiation of filtration of condenser vacuum pump/gland seal r exhaust whenever the CVSE channel is in a HIGH-HIGH alarm condition.
Sampled air is pulled from the condenser vacuum pump/gland steam exhaust piping at conditions of 125F and 100 percent relative humidity. Heaters are provided to raise the tempera- .
ture of the air to 137F and an RH of 70 percent at the inlet to the sampler piping in order to prevent degradation of particu-late and I<<131 sampler filters due'o excessive moisture. The downstream gas channel detector is designed'to withstand these sample conditions in continuous service.
A low and a high range monitor is used to'cover a range of eleven decades with two decades of overlay. Shielded particu>>
late and iodine samples exist in the low and high range moni-
'f tor and are removed for analysis.
Amendment 5 11.5-40 August 1981 07-01-81
I PVNGS FSAR INCBEMENAL PROCESS AND CAL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.4.2 Plant Vent PV Monitor. The plant vent exhaust is continuously and isokinetically monitored for particulate, EOG >-'g(G!!
X-131, and gaseous activity. UI; A low and a high range monitor is used to cover a range of .
< g
)gl fourteen decades with two decades of overlap. Shielded particulate and iodine samples exist in the low and high monitor and are removed for analysis. The low range iodine and particu-late samples also have a 5 decade detection range.
11.5.2.1.4.3 Fuel Buildin Ventilation Exhaust Monitor Channel "B" FBVE-B . This gaseous channel monitors the fuel building ventilation exhaust for release of activity due to a fuel handling accident. The monitor performs the safety function of isolating the normal ventilation system and activating the essential ventilation system (initiates a FBEVAS signal) on a HIGH-HIGH activity alarm. Redundancy and diver-sity are provided by the fuel pool area monitor (FPA-A) that
' also actuates the essential ventilation system. Refer to section 7.3 for a discussion of the safety function of the FBVE-B monitor. During normal operation the only significantly abundant isotope which would be released from the fuel build-ing is tritium; therefore, particulate and I-131 monitors are provided, but not recpired.
A low and a high range monitor is used to cover a range of eleven decades with two decades of overlap. Shielded particu-late and iodine samples exist in the low and high range monitor and are removed for analysis.
11.5.2.1.5 Area Radiation Monitoring One function of area radiation monitors (except the PAPA-A and PAPA-B monitors) listed in Table 11.5-1 is to indicate and alarm August 1981 11.5-41 Amendment 5 6>>22-81
. ~P. 1 K'-'bf~j ~), I 3 p PVNGS FSAR PROCESS AND EFFLUENT RADIOIOGICAL MONITORING AND SAMPLING SYSTEMS locally and remotely the area dose rate to ensure proper per-sonnel radiation protection. Several of the area monitors also perform other additional functions or have unique characteristics:
11.5.2.1.5.1 Central Calibration Facilit Area CFA Monitor.
The CFA monitor is located in a small outbuilding in the yard of Unit 1 which is shared by all three Units as a central calibra-tion facility.
11.5.2.1.5.2 Waste Solidification S stem Process Control Area PCA Monitors. These two monitors are not connected to any of the communications loops of the RMS. They are permanently in LOCAL control and are integral parts of the control system for the Waste Solidification System. Refer to section 11.4.2.3.1 for a description of their functions.
11.5.2.1.5.3 Deleted 11.5.2.1.5.4 Fuel Pool Area Monitor, Channel "A" FPA-A .
The FPA-A monitor is located on a wall overlooking the fuel pool where it monitors for a release of activity due to a fuel handling accident in the fuel building. The monitor performs the safety function of isolating the normal ventilation system and activating the essential ventilation system on a HIGH-HIGH dose rate alarm. Redundancy and diversity are provided by the fuel building ventilation exhaust (FBVE-B) gas monitor. Refer to section 7.3 for a discussion of the safety function of the FPA-A monitor.
11.5.2.1.5.5 Refuelin Area Monitor, Channel "A" RMA-A .
The RMA-A monitor is located on a wall overlooking the refuel-ing cavity where it monitors for a release of activity due to a fuel handling accident in the containment. The PAPA-A monitor is the channel "A" CPIAS sensor.
Amendment 5 11.5-42 August 1981 6-22-81
PVHGS FSAR CREMENTAL PROCESS AND EF ML L MONITORING AND SAMPLING SYSTEMS 11.5.2.1.5.6 Power-Access Pur e Area Monitors, Channels "A" gg age>
and "B" PAPA-A AND PAPA-B . The PAPA-A and PAPA-B monitors QG i are located between the containment power-access purge exhaust , ~~q and the refueling purge exhaust duct just outside the 2-'uct, containment wall. During power operations, these channels monitor the duct for airborne radioactivity concentrations which could potentially result in an offsite dose exceeding 10CFR100 limits. These monitors perform the safety function of isolating the containment building purge supply and exhaust ducts (initiate CPIAS signals) on a HIGH-HIGH dose rate alarm. Refer to l2 section 7.3 for a discussion of the safety functions of the PAPA-A and PAPA-B monitors.
11.5.2.1.6 Inspection, Calibration and Maintenance 11.5.2.1.6.1 Maintenance. Outdoor sampling systems are housed in outdoor-type weatherproof enclosures. The enclosures are designed to permit performance of all control and routine
( maintenance and cleaning operations from the front or top of the enclosure. Lifting eyes or other devices are provided for hoisting the unit, to facilitate replacement if it is ever required. Interior wiring is run in conduit to terminal boards mounted in junction boxes.
Instrument air is supplied for flushing of gaseous or airborne monitors. Demineralized water is supplied for'flushing all liquid monitors.
The interior of the sample chamber has a surface finish as required to minimize contamination by absorption or adherence of radioactive material thereto and to facilitate cleaning by
. use of a cleaning solution. The interior of all samplers is readily removable to allow ultrasonic decontamination or replacement.
RIC units and other equipments are slide<<mounted in the control room cabinets.
August l981 11.5-43 Amendment 5 6-22-81
PVNGS FSAR PROCESS AND EFFIUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.2.1.6.2 Calibration and Ins ection.' remotely operated check source is provided with each detector assembly. The check source isotope has a half-life greater than 10 years, with emission(s) in the energy range and of the same type as that being monitored, and is usable as a convenient operational and gross calibration check of the associated detection and readout equipment. The check source strength provides a count rate of a minimum of 1.5 decades above background.
A burn-in test and isotopic calibration of the complete radia-tion monitoring system are performed at the supplier's factory.
Field calibration sources, with their decay curves, are pro-vided with the system hardware.
Further isotopic calibrations are not required, since the geometry cannot be altered significantly within the sampler.
Subsequent calibration of samplers is then performed based on a known correlation between the detector responses and field calibration standards.
This single-point calibration confirms the detector sensitivity.
The field calibration is performed by removing the detector from the shield assembly and placing the calibration source on the sensitive area of the detect'or.
The radiation monitoring system channels are checked and inspected periodically. Grab samples are collected for iso-topic analysis in accordance with schedule in table 11.5-3.
Setpoint checks are done on a monthly basis, and calibration is performed at each refueling shutdown or indication of equipment malfunction. Instruments are serviced as required; Field calibration of the indicated channels is performed fol-lowing any equipment maintenance that can change the accuracy of the instrument indication. It is also done whenever the-check source indicates an abnormal response. Setpoints are also checked during equipment .calibration.
Amendment 5 11.5<<44 August 1981 6-22-81
lNCREME¹AL PVNGS PSAR PROCESS GICAL MONITORING STEMS E)NA 11.5.2.2 Routine Sam lin (gjG 3- ~ )98>
The requirements of the system design bases for routine continu-(jQT ous and discrete sampling of radioactivity are satisfied by of liquid, gaseous, and airborne samplers, laboratory i'ystem equipment for sensitive radio-chemical analyses, and a program of procedures for obtaining and analyzing representative samples when and where appropriate. This section provides a detailed description of system hardware and procedures in general, including the types of sample nozzles and other sampling equip-ment used, the procedures to obtain representative samples, and
'nalytical procedures. Table 11.5-2 is a tabulation of basic information describing each of the sampling locations, including August 1981 11.5-45 Amendment 5 6-22-81
Table 11.5-2 ROUTINE SAMPLXNG Expected Expected Types of Effluent I ~
Process Sample Concentragions Releases Sampling Location Basis for Location Selection Plowrate Composition (sCi/cms) (Refer to Table 11.5-3)
Maste gas decay Determination of identity 0-50 standard H2 ~ N2 10xlD Batch releases of tanks and quantity of radio-nuclides releasedg calibra-ft3/min KR>>85 fission and activation gases and tritium tion check of WGDT monitor.
Containment Determination of identity 30i000 Containment 2.3 x 10 Batch releases of building atmo- and quantity of radio- standard atmosphere Xe-133 fission and activation sphere (CB-B) nuclides to he released ft~/min LND I-131 gases and tritiumg monitor during containment purget preaccess 1.6 x 10"9 releases of airborne calibration check of CB-B 2,000 mini- C-137 radioactive monitor purge particulates, releases of airborne radio-active iodines Plant vent Determination of identity l<Di610 Ventilation Continuous releases of monitor and quantity of radio- standard exhaust air fission and activation nuclides being releasedg ft3/min gases and tritiums calibration check of PV including releases of airborne monitor 30'00 radioactive standard parti culatest releases ft3/min of airborne radio-containment active iodines purge Condenser vacuum pump/gland seal Determination of and identity quality of radio-2960 standard Condenser Gases 9 i x 10 Xe 133 Continuous releases fission and of activation exhaust gas nuclides heing releasedg ft3/min gases and monitor calibration check of CVSE of airborne tritium'eleases monitor radioactive particulatesg releases of airborne radio-active iodines II - ~ ~ II
PVNGS FSAR NCREMEf4TAL ENSURING THAT OCCUPATIONAL N EXPOSURES ARE AS LOW AS IS REASONABL )
12.1.1.2
~ ~ 0 eration Policies EDlT SECTlOW The Station Manual is one of the major means of promulgating
. AUG 1l 1981 the operational ALARA policy. This policy is. demonstrated in the radiation protection program, the training program, and station procedures.
Besides describing management's commitme'nt to ALARA, the Radiation Protection Division of the Station Manual designates the station personnel who have the responsibility and author-ity to implement ALARA. The Radiation Protection Supervisor has the responsibilities of the onsite Radiation Protection Manager, described in Regulatory Guide 8.8 and section 13.1, and is responsible to implement the radiation protection pro-gram and to direct steps to prevent unnecessary radiation exposures. He reports through the Engineering and Technical Services Manager and has direct access to the Manager of Nuclear Operations as shown in section 13.1.2. The supervising radia-tion physicists report to the Radiation Protection Supervisor and assist in the day-to-day operation of each unit radiation protection program. The Radiation Protection Supervisor super-vises the radiation protection personnel who perform the various radiation protection surveys. The responsibilities and author-ity of the supervisory positions discussed above and the quali-fications of the personnel who fillthem are discussed in sections 13.1.2 and 13.1.3, respectively.
It is the responsibility of the Training Manager and the Radiation Protection Supervisor to implement radiation protec-tion training for company employees and contractors commen-surate with the requirements of 10CFR20. The Radiation Protec-tion Supervisor verifies that personnel follow the radiation August 1981 12.1-3 Amendment 5 08-06-81
, ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES
""'""'<.'j.j;""":"'-'-"
ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) c,y ij protection procedures designed to ensure that exposures are maintained ALARA. To ensure compliance with this policy, the Radiation: Protecti'on Supervisor and supervising radiation physicists reporting to him are charged with the responsibility to promptly advise higher management of practices which exceed their authority to correct. In addition, periodic reviews (at least annual) of the ALARA program, including review of radia-tion exposure records and operating procedures, are conducted by offsite Health Physics personnel.
Prior to startup, personnel requiring access to radiation areas will be trained in radiation protection procedures and techniques. These personnel will be tested to verify that they understand how these procedures relate to the performance of their jobs. They also will be retrained in radiation protection procedures and techniques on a 3-year cycle and will be tested annually. Contractors who work in a restricted area of one of the units after initial fuel loading of the unit will receive radiation protection training commensurate with the performance of their jobs and will be tested on this knowledge. Construction personnel will be instructed in site emergency procedures in the event of an emergency at one of the operating units. For more detail see section 12.5.
Prior to startup of the first unit, station procedures to be used for work which involves significant personnel radiation exposure will be reviewed to verify that the procedures adhere t
to the ALARA philosophy. Revisions to station procedures involving significant personnel radiation exposure will also receive an ALARA review. System or station modifications affecting personnel radiation exposure will also be reviewed to see that the ALARA concept is applied.
)
Amendment 5 08-06-81'ugust 12.1-4 1981
PVNGS FSAR ENSURING THAT OCCUPATIONAL I,NCREMENTg,
/
AS LON AS IS REASONABLE ACH i'RE The Radiation Protection Supervisor will'eriodically survey Epg SEggw)5 station operations to identify situations in which exposures can be reduced. AUG 11 198l 12.1.2 DESIGN CONSIDERATIONS This section discusses the methods and features by which the policy considerations of section 12.1.1 are applied.
Provisions and designs for maintaining personnel exposures ALARA are presented in detail in sections 12.3.1, 12.3.2, and 12.5.3.
12.1.2.1 General Desi Considerations for Shieldin and ALARA Ex osures General design considerations, shielding, and methods employed to maintain in-plant radiation exposures ALARA con-
\
sistent with the recommendations of NRC Regulatory Guide 8.8, Section C.3, have two objectives:
A. Minimizing the necessity for and amount of personnel time spent in radiation areas.
B. Minimizing radiation levels in routinely occupied plant areas in the vicinity of plant equipment expected to require personnel attention.
Plant operating personnel are protected by radiation shielding wherever a potential radiation hazard may exist. The shielding performs the following additional functions:
A. Assist in limiting radiation exposure to plant control room personnel to within the limits of 10CFR50, Appendix A, Criterion 19 in the unlikely event of an accident.
r B. Protect certain components from excessive activation or excessive radiation exposure.
Facilitate access for maintenance of components.
August 1981 12.1-5 Amendment 5
",: ', ,.'."~~'l,'ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 1 In order to maintain exposure ALARA', a radiation zone classifi-cation system has been developed and used. Plant areas have been classified in accordance with itable 12.1-1.
12.1.2.1.1 Neutrons Neutron shielding design is as follows:
A. Neutron shielding is designed to ensure that the area does not become a scattering source, producing excess-ive doses in other regions.
B. The shielding is designed to ensure that neutron activation does not result in doses exceeding the permitted shutdown dose rates in the region.
'8 Table 12.1-1 RADIATION ZONE CLASSIFICATION Zone Dose Rate Designation (mrem/h) Allowed Occupancy Less than 0.5 Uncontrolled, unlimited t access (plant personnel) 0.5 to 2.5 Controlled, limited access, 40 h/wk to unlimited 3 2.5 to 15 Controlled, limited access, 6 to 40 h/wk 15 to 100 Controlled, limited access, 1 to 6 h/wk Over 100 Normally inaccessible, access only as permitted by radiation protection personnel 1 h/wk 12.1-6 8-9-79
PVNGS PSAR 1NCRF MENTAL ENSURING THAT OCCUPATIONAL I S ARE AS LON AS IS REASONABLE ACHIEVABL 1
C. Neutron radiation damage limits of equipment aie not exceeded unless provisions are made for periodic EDIT SECTION replacement. AUG 1 1 j98>
',i4 OUT 12.1.2.1. 2 Gamma Radiation Gamma radiation shielding design is as follows:
A. Shielding is designed .to reduce gamma dose rates throughout the plant to levels consistent with expected occupancy during normal operation as specified by the designated radiation zones.
As a minimum, shielding is designed to reduce gamma dose rates from sources external to a radioactive compartment to levels comparable to dose rates resulting from equipment within that compartment.
This design ensures that radiation levels in a compart-ment undergoing maintenance that are due to operating equipment in adjacent compartments will be the lesser of the zone 3 limit (15 mrem/hr) or the operational dose rate of the equipment under repair.
C. Shielding is provided to attenuate radiation from sources external to equipment compartments so that expected maintenance can be performed without exceed-ing exposure limits.
D. Shielding is designed to reduce gamma radiation after reactor shutdown to levels which allow access for required maintenance operations.
E. Gamma radiation damage limits for the equipment are ~
P not exceeded unless provisions are made for periodic replacement.
(
August 1981 12.1-7 Amendment 5 07-01-81
q jl'$ P ygl i Ik PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.2 Facilit La out General Desi Considerations for ALARA A. facility general design considerations to minimize the amount of personnel time spent in a radiation area include:
- 1. The layout for personnel access, routing of piping, and location of components is in a Amendment 5 12.1-7A August 1981 06-29-81
PVNGS FSAR ll 0%ri 'll+VlC INTA)
ENSURING T OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW I
@ CHIEVABLE (ALARA)
TOIT SEGTl0tl AUG 1 j. j98jl li] GUT This page intentionally blank August 1981 12.1-7B Amendment 5 06-29-81
',,'ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 1
'ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) manner to minimize personnel radiation exposure during both operation and maintenance. Access control and traffic patterns have been evaluated to assure that radiation exposures are maintained as low as reasonably achievable. In the interest of maintaining ALARA doses, access to a given radiation zone does not require passing through a higher radiation zone.
Radioactive components of the same system are grouped together as practical to minimize radioactive piping runs. Ion exchangers and spent resin collection system components are located as close to the radwaste solidification area as practical. Radioactive wastes are stored in shielded enclosures separated from normally accessible areas.
Due to the desirability of low background radia-tion levels, the counting room has been located away from highly radioactive components.
Due to frequent access requirements, control panels, readout devices, and transmitters, where possible, are located in low radiation zones (Zone 2 or less) in order to minimize operator exposures.
In order to reduce radiation exposures due to sampling operations, sample stations are isolated insofar as practical from other radioactive equipment, and exposed sample piping is minimized.
Primary and secondary system samples taken within the containment are piped to the plant laboratory to minimize the need to access the containment.
12.1-8 8-9-79
ENSURING THAT OCCUPATIONAL RADIAT XPOSURES ARE AS LOW AS IS REASONABLE ACHI radioactive and nonradioactive piping are separated within the penetration area, with ED1T SEGTlGH pr 's~ " or "t'z tion f t por y AUG111981 shielding for maintenance purposes.
GUT 34 In general, piping, electrical, and HVAC penetra-tions through radiation shields are designed to minimize radiation streaming into accessible areas. The following considerations apply:
- a. Cross-sectional areas of penetrations are minimized.
- b. Where practical, penetrations are oriented so that there is an offset between 'the radiation source and accessible areas.
c ~ If such an offset orientation is impractical, the penetration is located as far above the floor elevation as possible to minimize
. d.
direct exposure of personnel.
If the above methods are not utilized, alternative means are employed. For piping penetrations, the alternatives include grouting the annular space of the penetra-
" tion or utilizing shield collars around the piping at the openings of the penetration to reduce radiation streaming. For ventila-tion and electrical penetrations, use of baffle shields to eliminate streaming into accessible areas is acceptable.
12.1.2.3 E i ment General Desi Considerations for ALARA The following guidelines have been considered in the design of radioactive equipment, as a means of minimizing operator exposure 12 ~ 1~17 8-9-79
PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES I ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) by minimizing both equipment contamination and service time l required:
A. Reliability, durability, constWction, and design features of equipment, components, and materials to reduce or eliminate the need for repair or preventive maintenance.
B. Servicing convenience of anticipated maintenance or potential repair, including ease of disassembly and modularization of components for replacement or removal to a lower radiation area for repair.
C. Provisions, where practical, to remotely or mechanically operate, repair, service, monitor, or inspect equipment (including inservice inspection in accordance with American Society of Mechanical Engineers (ASME) Sec-tion XI).
D. Redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high.
Equipment general design considerations directed toward minimiz-ing radiation levels in proximity to equipment or components requiring personnel attention include:
A. Provision for draining, flushing, or if necessary, remotely by cleaning equipment and piping containing radioactive material.
B. Design of equipment, piping, and valves to minimize the buildup of radioactive material and to facilitate flushing of crud traps. The use of cobalt containing alloys in systems exposed to primary coolant has been minimized except where very hard erosion resistant surfaces are necessary (e.g., stellite seat materials t
on valves). C-E supplied components exposed to primary coolant have less than 0.2/ cobalt, nominally. Refer also to CESSAR-FSAR Table 5.2-2.
)
Amendment 5 12.1-18 August 1981 06-29-81
ENSURING THAT OCCUPATIONAL RADI ARE AS LOW AS ZS REASONABLE ACHIEVAB igA) ')/
C Provisions for minimizing the spread of contamination "CRTS CHION into equipment service areas, including direct drain p,UG ] y ~g8~
connections.
D. Provisions for isolating equipment from radioactive process fluids.
E. Provision for a spent fuel pool cleanup system to maintain the radiation level of the fuel pool area within the Zone 2 limit. See table 12.1-1 for the description of radiation zones.
F. Heat: exchangers have been provided with corrosion-resistant tubes with tube-to-tube sheet joints fabri-cated to minimize leakage. Impact baffles are provided and process fluid velocities are limited as necessary to minimize erosive effects. Provisions are made for removal of the tubes for maintenance.
G. Pumps in radioactive service have been purchased with mechanical seals to reduce seal servicing time.
Smaller pumps are provided with flanged connections for ease in removal. Pump casings are provided with drain connections for draining the pump for maintenance.
H. Water is used to fluidize tanks from which resin is transported. Resin tanks incorporate integral self-cleaning screens in overflow connections to retain resins within the tank. Overflow connections for radioactive tanks are piped to the liquid radwaste system (LRS) to facilitate radwaste processing.
Filters are supplied with the means to either remotely P backf lush the filter or to perform cartridge replace-ment with remote tools.
Demineralizers pre designed to remotely remove spent resins hydraulically and replace new resins from a remote location. Resin strainers have been designed for full system pressure drop.
August 1981 12.1-19 Amendment 5 06-29-81
~
ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LON AS II IS REASONABLE ACHIEVABLE (ALARA)
Evaporators are provided with chemical addition connec-tions to allow the use of chemicals for descaling operations.
The reactor head laydown area has been designed to sub-stantially reduce both the dose to those changing 0-Ring seals and personnel working in adjacent areas. An interior shield wall separates the 0-Ring changeout from the radiation field under the hemispherical head.
An exterior shield wall isolates the activated flange and external surfaces from adjacent work activities.
Frequently operated valves of highly radioactive systems are designed for remote operation. Motor operators, air operators, and reach rods are provided where necessary. The criteria for selecting valve operators are as follows:
- 1. In radioactive areas Zones 3, 4, and 5, valves which are operated frequently; for example, on a weekly basis or more often, are equipped with a remote actuator such as a simple reach rod, electric motor actuator, or pneumatic actuator and position indicator. These valves are con-trolled from the applicable control station or operating aisle.
- 2. Valves which are operated occasionally; for example, between one and twelve times a year, are classified as follows:
a ~ Valves which are located in radiation areas of 100 mr/h (Zone 4) or less during operation may be manually operated directly.
- b. Valves which are located in higher radiation areas (Zone 5) have been equipped with a simple reach rod, or, if a simple V
i 12.1-20
'-9-79
ENSURING THAT OCCUPATIONAL RADIATION EXFOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.4 General Desi Consideratio &Accident Ex osures ALARA ;81'~<
ED% SC(;at The facility layout will assist in keeping occupational AUG18 )c~:
exposures ALARA even after a design basis accident. While exposures will be significantly higher than during normal @ QU.
operation, required access is provided to vital areas and systems without exceeding 5 rem/hr. 'Zone maps showing expected dose rates in the event of a LOCA with sump recirculation are provided in Section II.B.2 of the LLIR. Zone maps for the hypothetical condition. of a LOCA with an intact primary but with a degraded core are also provided in Section II.B.2 of the LLIR. A discussion of the source terms for these events is provided in section 12.2.3. The dose rates projected for these two sets of drawings do not assume decay beyond that corresponding to the onset of recirculation. Even so, virtu-ally unrestricted access will be permitted within the control and diesel generator buildings, as well as portions of the upper floor of the auxiliary building (such as the area of the operational support center).
To provide sampling capability with exposures kept ALARA, PVNGS will incorporate a remote, automated, post-accident sampling system that meets the requirements of NUREG 0737 and Regulatory Guide,1.97, Revision 2. This system can be operated from the radiochemistry counting room, the control room, or the techni-cal support center. Backup grab sample capability will be provided in the hot lab sample room using microchemistry tech-niques to keep the volume of source as small as possible.
The only. other area where access might be required is'to the hydrogen monitors/recombiners. Projected dose rates without the recombiners in operation but at the onset of recirculation are expected to be approximately 10 to 30 rem/hr (sump recircu-lation). As the recombiners do not have to be installed until
( ~
August 1981 12.1-22A Amendment 5 8-17-81
PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ASi,LOW AS IS REASONABLE ACHIEVABLE (ALARA) at least one,day after the DBA, dose rates will drop due to decay to about 1/10 the doses noted above. Thus, the instal-lation dose rate (assuming sump recirculation) will be less f \
than 5 rem/hr. While the dose rate would be greater than 5 rem/hr for an intact primary-degraded core event, the recom-biners would not need to be installed since an intact primary would not be consistent with hydrogen generation inside the containment. If hydrogen generation were postulated, this would necessitate a break or opening in the primary. Conse-quently, sump recirculation would be available with the concomitant release of noble gases and dilution by the refuel-ing water tank. These consequences would lead to the doses noted above for the sump recirculation mode of cooling (i.e.,
dose rates less than 5 rem/hr).
ESF grade filtered ventilation is provided for auxiliary building rooms below elevation 100 feet (refer to section 9.4).
This will reduce airborne sources due to recirculation and/or containment leakage. Non-ESF grade filtered ventilation is available for use to reduce airborne sources above elevation 100 feet in the auxiliary building (refer to section 9.4). The use of non-ESF filtration is acceptable since there are no recirculation components above elevation 100 feet. Thus the only significant source of airborne activity is containment leakage. This leakage has already been accounted for in off-'ite dose analyses which assumed direct containment leakage to the atmosphere. Secondly, this filter discharges via the plant vent. The plant vent will be monitored in accordance with NUREG 0737 and Regulatory Guide 1.97, Revision 2 to provide notification of decreased filter efficiency.
Therefore, considering direct and airborne sources, access can be provided to those vital areas necessary for control of the plant and personnel exposures will meet GDC 19 and NUREG 0737 limits.
Amendment 5 12.1-22B August 1981
/ EDIT SECTION PVNGS FSAR WCRFMpNTAL AUG 11 l98l KGKZIQN SOURCES V = free volume of the region i '
leak occurs C (t) =
in cm airborne concentration j<>.
of the i radioisotope gg/f at time t in pCi/cm in the applicable region From the above equation, it is evident that the peak or equi-i librium concentration, CEEqi of the
~ i radioisotope in the applicable region will be given by the following expression:
(2)
With high exhaust rates, this peak concentration will be reached within 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 two LOCA events. The first was a LOCA with recirculation accomplished via the containment sump. The second was a LOCA with an intact primary with recirculation accomplished via the shutdown coding system.
The following core releases were used in the review:
A. Source A: Containment airborne': 100% noble gases, 25% iodines - see table 12.2-8.
B. Source B: Reactor coolant: 100% noble gases, 50% halogens, 1% solids - see table 12.2-9.
C. Source C: Containment sump: 50% halogens, 1% solids-see table 12.2-10.
Volumes used for each source were:
A. Source A: Containment free volume of 2.6 x 10 6 ft3 .
B. Source B: Reactor coolant system (RCS) volume of 9.15 x 10 ft3 August 1981 12.2-21 Amendment 5 06-29-81
PVNGS FSAR RADIATION SOURCES C. Source C: The'inimum volume
~P 0" of water, 7.76 x 10 ft y ...i,~ fi i>P!l, present at the time of recirculation. (RCS
+ refueling water tank + safety injection tanks.)
A LOCA with sump recirculation is represented by sources A and C. An intact primary-degraded core LOCA is represented by sources A and B (source A was not reduced even though there is no mechanism to assume noble gases in both sources). The results of the shield review are presented in section 12.3.1.3.
Decay curves, normalized to initial time equals zero, were developed for the sources as an aid in developing post-accident access plans. These curves are presented as fig-ures 12.2-1 (source A), 12.2-2 (source B), and 12.2-3 (source C).
Amendment 5 12.2-22 August 1981 07-01-81
1NCREMBJTAL PVNGS FSAR AUG y y ]98~
11'UT ACES TABLE 12.2-8 LOCA SOURCE A - CONTAINMENT AIRBORNE (Curies)
Nuclide Activity( )
Kr-85m 3.48(+7)
Kr-85 1.21(+6)
Kr-87 5.91(+7)
Kr-88 8.59(+7)
Kr-89 1.08(+8)
Kr-90 1.16(+8)
Kr-91 8.95(+7)
I-129 4.03(-1)
I-131 2.83(+7)
Xe-131m 6.17(+5)
I-132 2.90(+7)
I-133 5.80(+7)
Xe-133 2.26(+8)
I-134 6.35(+7)
I-135 5.53(+7)
Xe-135m 6. 38 (+7)
Xe-135 5.31(+7)
I-137 3.60(+7)
Xe-137 2.09(+8)
I>>138 2.19(+7)
Xe-138 2.05(+8)
Xe-140 1.16(+8)
Xe-143 2.75(+6)
Xe-144 4.03(+5)
( ~ a. Numbers in parenthesis denote powers of ten August 1981 12.2-23 Amendment 5 /s 07<<01>>81
~~
)~y I RADIATION SOURCES TABLE 12.2-9 LOCA SOURCE B - REACTOR COOLANT (Curies)
Nuclide Activity Nuclide Activity Nuclide Activity Se-84 2.55(+5) Sn-129 1.08(+5) Cs-140 2.00(+6)
Br-84 1.28(+7) Sb-129 2.08(+5) Ba-140 2.11(+6)
As-85 5. 52 (+4) Te-129m 3.68(+4) La-140 2.11(+6)
Se-85 3.02(+5) Te-129 2.03(+5) Xe-143 2.75(+6)
Br-85 1.61(+7) I-129 8.05(-1) Cs-143 4.52(+5)
Kr-85m 3.48(+7) Sn-131 3.50(+5) Ba-143 1.47(+6)
Kr-85 1.21(+6) Sb<<131 9.30(+5) La-143 1.81(+6)
Se-87 3.29(+5) Te-131m 1.72(+5) Ce-143 1.83(+6)
Br-87 2.87(+7) Te-131 1.00(+6) Pr-143 1:84(+6)
Kr-87 5.91(+7) I-131 5.65(+7) Xe-144 4.03(+5)
Br-88 3.87(+7) Xe-131m 6.17(+5) Cs-144 1.25(+5)
Kr-88 3.59(+7) Sn-132 3.08(+5) Ba-144 8.75(+5)
Rb-88 8.61(+5) Sb-132 8.24(+5) I a-144 1.61(+6)
Br-89 3.96(+7) Te-132 1.15(+6) Ce-144 1.25(+6)
Kr-89 1.08(+8) I-132 5.80(+7) Pr-144 1.25(+6)
Rb-89 1.12(+6) Sn-133 5.37(+4)
Sr-89 1.13(+6) Sb-133 1.10(+6)
Br-90 3.89(+7) Te-133m 1.42(+6)
KR-90 1.16(+8) Te-133 1.01(+6)
Rb-90 1.,31(+6) I-133 1.16(+8)
Sr-90 6.60(+4) Xe-133 2.26(+8)
Y-90 6.57(+4) Cs<<134 1.28(+4)
Kr-91 8.95(+7) Sb<<134 5.69(+5)
Rb-91 1.45(+6) Te-134 2.14(+6)
Sr-91 1.49(+6) I-134 1.27(+8)
Y-91m 8.80(+5) Sb-135 1.80(+5)
Y-91 1.49(+6) Te-135 1.09(+6)
Sr-95 1.64(+6) I-135 1.10(+8)
Y-95 2.02(+6) Xe-135m 6.38(+7)
Zr-95 1.99(+6) Xe-135 5.31(+7)
Nb-95 1.95(+6) Cs-135 2.50(-1)
Zr-99 2.61(+6) Cs-136 1.32(+4)
Nb-99 2.61(+6) I-137 7.20(+7)
Mo>>99 2.69(+6) Xe-137 2.09(+8)
Tc-99m 3.23(+5) Cs-137 4.84(+4)
Mo-103 1.46(+6) Ba-137m 4.52(+4)
Tc-103 1.53(+6) I-138 4.39(+7)
RG-103 1.53(+6) Xe-138 2.05(+8)
Tc-106 6.72(+5) Cs-138 2.07(+6)
Ru-106 5.27(+5) Xe-140 1.16(+8)
- a. Numbers in parenthesis denote powers of ten Amendment 5 12.2-24 August 1981 07-01-81
TABLE 12.2-10 LOCA SOURCE C - CONTAINMENT SUMP (Curies)
Nuclide Activity{ Nuclide Activity Nuclide Activity Se-84 2.ss(+s) Sn-129 1.08(+5) Cs-140 2.00(+6)
Br-84 1.28(+7) Sb-129 2.08(+5) Ba-140 2.11(+6)
As-85 5. 52 (+4) Te-129m 3.68(+4) La-140 2.11(+6)
Se-85 3.02{+5) Te-129 2.03{+5) Cs-143 4.52(+5)
Br-85 1.61(+7) I-129 8.05(-1) Ba-143 1.47(+6)
Se-87 3.29(+5) Sn-131 3.50(+5) La-143 1.81(+6)
Br-87 2.87(+7) Sb-131 9.30(+5) Ce-143 1.83(+6)'.84(+6)
Br-88 3.87(+7) Te-131m 1.72(+5) Pr-143 Rb-88 8.61{+5) Te-131 1.00{+6) Cs-144 1.25(+5)
Br-89 3.96(+7) I-131 5.65(+7) Ba-144 8.75(+5)
Rb-89 1.12(+6) Sn-132 3.08(+5) La>>144 1.61(+6)
Sr-89 1.13(+6) Sb-132 8.24(+5) Ce-144 1.25(+6)
Br-90 3.89(+7) Te-132 1.15(+6) Pr-144 1.25(+6)
Rb-90 1.31(+6) I-132 5.80(+7)
Sr-90 6.60{+4) Sn-133 5.37(+4)
Y-90 6.57(+4) Sb-133 1.10(+6)
Rb-91 1.45(+6) Te-133m 1.42(+6)
Sr-91 1.49(+6) Te-133 1.01(+6)
Y-91m 8.80(+5) I-133 1.16(+8)
Y-91 1.49(+6) Cs-134 1.28(+4)
Sr-95 1.64(+6) Sb-134 5.69(+5)
Y-95 2.02(+6) Te-134 2.14(+6)
Zr-95 1.99(+6) I-134 1.27(+8)
Nb-95 1.95(+6) Sb-135 1.80(+5)
Zr-99 2.61(+6) Te-135 1.09{+6)
Nb-99 2.61{+6) I-135 1.10(+8)
Mo-99. 2.69(+6) Cs-135 2.50(-1)
Tc-99m 3.23(+5) Cs-136 1.32(+4)
Mo>>103, 1.46(+6) I-137 7.20(+7)
Tc-103 1.53(+6) CS-137 4.84(+4)
RG-103 1.53(+6) Ba-137m 4.52(+4)
Tc-106 6.72{+5) I-138 4.39(+7)
RQ-106 5.27(+5) Cs-138 2.07(+6)
- a. Numbers in parenthesis denote powers of ten August 1981 12.2-25 Amendment 5 07-01-81
I 1 E
1 PL j~
I(
!3c l'
'll ff V I
<Ij Rg
'1,~
I
EDIT SE(ETIO! J (MeY/sec) Vs time (days) lUG y y ]AU)
(MeV/sec) t p
.01 4 s 42.3 Our 100.0 10.0 CO C7 tlJ I
1.0 0.1 0.01 1 10-1 10.2 10 10'eV/sec (MeV/sec) t o Palo Verde Nuclear Generating Station FSAR DECAY CURVE - SOURCE A (Sheet 1 of 2)
Figure 12.2-1 August 1981 Amendment 5 8-06-81
)
l" f
'(1'i
)
fg}l cE('J(0 (MeV/sec)
(MeV/sec)
Vs time (days) ~Ub l1 198 t
W 23Ct %70 Oi~q 10 10+
Me V/sec (MeV/sec) ot Palo Verde Nuclear Generating Station ri</ii: FSAR DECAY CURVE - SOURCE A (Sheet 2 of 2)
Figuze 12.2-1 August 1981 Amendment 5 I
8-06-81
l C)
~
iNCREVENTAL Fly sE(;y~
(Me V/sec)
(MeV/sec)
Vs time (days) AUQg 3i t e 0.01 4t4 180. lN Ooy 100.0 1.0 0.1 0.01 1oo 1O-'0-2 MeV/sec 10 (MeV/sec) ~
Palo Verde Nuclear Generating Station FSAR DECAY CURVE - SOURCE B Pigure 12.2-2 August 1981 Amendment 5 8-12-81
~ ~
lNCREMENTAL EMT SEKTIOH (MeV/sec) Vs time (deys)
(MeV/sec) < I UG ] 1198l
.01 cs <180. )( Ply 100.0 10.0 CO
<<C CI tlJ l
1.0 0.1 0.01 10.2 10 10 10'eV/sec (MeV/sec) tm Palo Verde Nuclear Generating Station 0 FSAR DECAY CURVE - SOURCE C Figure 12.2<<3 August 1981 Amendment 5
C~
PVNGS FSAR INCREMENTAL RADIATION PROTECTION
,e 12.3.1.2 Radiation Zonin and Access Control DESIGN FEATURES Access to areas inside the plant structures and plant yards PD)T. SECTION QUQ ].J )98l is regulated and controlled by radiation zoning and access control (section 12.5.1.3). Each radiation zone defines the radiation level range to which the aggregate of contributing sources must be attenuated by shielding.
Plant areas are categorized as radiation zones according to expected maximum radiation levels and anticipated personnel occupancy with consideration given toward maintaining personnel exposures as low as is reasonably achievable and within the standards of 10CFR20. Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal, shutdown, spent resin transfer, and emergency operations; for maintenance occupancy .requirements; for general access requirements; and for material exposure limits to determine appropriate zoning. The radiation zone categories employed, and their descriptions are given in table 12.1-1. The specific zoning for each plant area is shown in figures 12.3-1 through 12.3-20. Frequently accessed areas, e.g., corridors, are shielded for Zone 1 or Zone 2 access.
The control of entry or exit of plant operating personnel to controlled access areas, and procedures employed to ensure that radiation levels and allowable working time are within the limits prescribed by 10CFR20 is described in section 12.5.
12.3.1.3 Radiation Zones - Post-Accident Radiation zone maps were developed in accordance with NUREG 0737 to review potential access throughout the plant post accident. (Refer to PVNGS LLIR Section II.B.2.) Two events were considered as noted in section 12.1.24 using the .sources described in section 12.2.3. The events were a LOCA with sump recirculation and an intact primary-degraded core LOCA. Estimated radiation levels in vital areas were based on radiation sources from the post-accident August 1981 12.3-11 Amendment 5
'8-17-81
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES operation of the following systems: containment, safety injection/shutdown cooling/containment spray, chemical and volume control system (up to purification filter inlet),
post-accident sampling, and hydrogen recombiners. The gaseous radwaste system will not be used post-accident. Palo Verde does not have a standby gas treatment system or equivalent.
As a result of this review, piping used for backup grab sam-pling in the hot lab sample room area will be lead wrapped to keep operator doses ALARA. Note that the primary post-accident sampling method is a remote, automatic sy'tem.
12.3.2 SHIELDING The bases for the nuclear radiation shielding and the shielding configurations are discussed in this section.
Amendment 5 12.3-11A August 1981 08<<18-81
RADIATION PROTECTION T SECTlQH FEATURES
'0 8Q Iggpf AUG 11 OUr 1981 This page intentionally blank.
Og I
~
ij August 1981 12.3-11B Amendment 5 06-29-81
PVNGS FSAR RADiATION PROTECTION DESIGN FEATURES 12.3.2.1 Desi n Ob'ectives The basic objective of the plant radiation shielding is to reduce personnel and population exposures, in conjunction with a program of controlled personnel access to and occupancy of:.)
radiation areas, to levels that are within the dose regulatiorjp of 10CFR20 and are as low as is reasonably achievable {ALARA)=.-.
within the dose regulations of 10CFR50. Shielding and equip-ment layout and design are considered in ensuring that exposures are kept ALARA during anticipated personnel activities in areas of the plant containing radioactive materials, utilizing the design recommendations given in Regulatory Guide 8.8,
~ ~ ~ ~
Paragraph C.2, where practical.
Four plant conditions are considered in the nuclear radiation shielding design: normal, full-power operation; shutdown; spent resin transfer; and emergency operations {for required access to safety-related equipment).
The shielding design objectives for the plant during normal operation, including anticipated operational occurrences; for shutdown operations; and emergency operations are:
To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10CFR20.
B. To assure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspec-tion, and safety-related operations required for each plant equipment and instrumentation area.
C. To reduce potential equipment neutron activation and mitigate the possibility of radiation damage to materials.
i 12.3-12 7-30-79
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURESAUB 1 1 198t D. To ensure that the control room will be sufficiently IN OUT shielded so that the direct dose plus the inhalation dose (calculated in chapter 15) will not exceed the limits of 10CFR50, Appendix A, General Design Criterion 19.
I 12.3.2.2 General Shieldin Desi Shielding is provided to attenuate direct radiation through walls and scattered radiation through penetrations to less than the upper limit of the radiation zone for each area shown xn figures 12.3-1 through 12.3-20. The shielding requirements for plant areas are presented in figures 12.3-1 through 12.3-20. Design criteria for penetrations are consistent with the recommendations of Regulatory Guide 8.8, and are discussed in section 12.3.1.1.2.
Should dose rates in excess of zone criteria occur after PVNGS I
begins operation, PVNGS will add shielding or revise access to the affected area so as to ensure proper access control.
pfg The material used for most of the plant shielding xs ordinary concrete with a minimum bulk density of 140 lb/ft3 . Whenever poured-in-place concrete has been replaced by concrete blocks, design ensures protection on an equivalent shzeldzng basxs as determined by the density of the concrete block selected.
Concrete radiation shields are designed following the recommendations of Regulatory Guide 1.69 as discussed zn section 1.8. Water is used as the primary shield material for areas above the spent fuel storage area.
12.3.2.2.1 Containment Shielding Design Derring reactor operation, the containment protects personnel occupying a djacent jac plant structures and yard areas from radiation originating in the reactor vessel and prima~ loop components. The concrete containment wall, together with the reactor vessel and steam generator compartment shield walls, August 1981 12.3-13 Amendment 5 08-06-81
e
/
~
r JIB ~ ~ ~ \4 PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES reduces radiation levels outside the containment to less than 0.5 mrem/h. The containment shield is reinforced, prestressed concrete completely surrounding the nuclear steam supply 0,
system. The wall is 4 feet thick, and the dome varies from 4 feet at the springline to 3 feet 6 inches at 'the top.
For design basis accidents, the containment shield together with the control room shielding reduces the plant radiation intensities from fission products inside the containment to acceptable levels, as defined by 10CFR50 Appendix A, General Design Criterion 19, for the control room.
Where personnel and equipment hatches or penetrations pass through the containment wall, additional shielding is provided to attenuate radiation to the required level defined by the outside radiation zone during normal operation and shutdown, and to acceptable levels as defined by 10CFR50 during design basis accidents.
12.3.2.2.2 Containment Interior Shielding Design During reactor operation, many areas inside the containment are Zone 5 and normally inaccessible. However, shielding is provided to reduce dose rates to approximately 15 mrem/h or less in the areas that require potential access during power operation. These are the Zone 3 areas of the containment shown in figures 12.3-1 through 12.3-20.
The main sources of radiation are the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping. The reactor vessel is shielded by the concrete primary shield, reactor cavity shield, and by the concrete secondary shield which also surrounds all other primary loop components. Air cooling is provided to prevent overheating, dehydration, and degradation of the shielding and structural properties of the primary shield and reactor cavity shield.
)
Amendment 5 12.3-14 August 1981 8-06-81
PVNGS FSAR lNCREMENTAL RADIATION PROTECTION EOll $ EC7(pM The primary shield is a large mass of reinforced concrete AUG )] )gs) surrounding the reactor vessel and extending upward from the tN pUy containment floor to form the walls of the fuel transfer canal. The concrete thickness is 7 feet up to the height of the reactor vessel flange where the thickness is reduced to 4 feet.
The reactor cavity shield, an annular mass of concrete, is below the reactor vessel nozzles between the vessel and the primary shield as shown in figure 12.3-21. This shield minimizes neutrons streaming from the annulus between the reactor vessel'nd the primary shield. The bottom of the cavity shield is located at about elevation 96 feet. The top of the cavity shield is located at about elevation 100 feet.
Estimates of neutron dose rates vary spatially over the operat-ing level from 50 to 60 mrem/hr at airlock and near pre-access filter unit to approximately 100 to 500 mrem/hr at locations that can view the CEDN cable structure and up to 1 to 2 rem/hr at locations along the edge of the refueling canal. Neutron dose levels outside the steam generator compartments at eleva-tions below the operating level are expected to be negligible due to shielding by the concrete operating level floor.
The primary shield and reactor cavity shield meet the following objectives:
A. In conjunction with the secondary shield, to reduce the radiation level from sources within the reactor vessel and reactor coolant system to allow limited access to the containment during normal, full-power operation.
B. After shutdown, limit the radiation level from sources within the reactor vessel, to permit limited access to
~0 reactor coolant system equipment.
C. To limit neutron flux activation of component and
, ~ structural materials.
August 1981 12.3-15 Amendment 5 08-06-81
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES The regenerative heat exchanger of the letdown portion of the chemical and volume control system is located in a shielded compartment that is normally Zone 5, a restricted access area.
e Shielding is provided for it consistent with its postulated maximum activity (section 12.2.1) and with the access "j and zoning requirements of adjacent areas.
After shutdown, the containment is accessible for limited periods of time and all access is controlled. Areas are (i
5 Amendment 5 12.3-1SA August 1981 06-29-81
PVNGS FSAR lNCREMFNTAL EDIT SECTION RADIATzON PRpTECTzpN D N FEATURES AUG $ ]. ig8) 0Ur gpss/
This page intentionally blank.
~
o t~
August 1981 12.3-15B Amendment 5
,06-29-81
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES
\
surveyed to e'stablish allowable working periods. Dose rates are. expected to range from 0.5 to 1000 mrem/h, depending 4) on the location inside the containment (excluding reactor cavity). ". These dose rates result from residual fission products, neutron-activated materials, and corrosion products'.-
in the reactor coolant system.
Spent fuel is the primary source of radiation during refueling.
Because of the extremely high activity of the fission products contained in the spent fuel elements and the proximity of Zone II areas, extensive shielding is provided for areas surrounding the spent fuel pool and the fuel transfer canal to ensure that radiation levels remain below zone levels specified for adjacent areas. Water provides the shielding over the spent fuel assemblies during fuel handling (refer to 5
figure 12.3-24). Furthermore, substantial structural barriers
~ ~
~
to limit, access in the vicinity of the fuel transfer tube during fuel handling operations have been provided. Refer to 5
figure 12.3-22 for the fuel transfer shielding arrangement.
The secondary shield is a reinforced concrete structure C')
surrounding the reactor coolant equipment, including piping, pumps, and steam generators.
This shield protects personnel from the direct gamma radiation resulting from reactor coolant activation products and fission products carried away from the core by the reactor coolant.
In addition, the secondary shield supplements the primary shield by attenuating neutron and gamma radiation escaping from the primary shield. The secondary shield is sized to allow limited access to the containment during full power operation.
The thickness of secondary shield walls is 4 feet.
12.3.2.2.3 Auxiliary Building Shielding Design During normal operation, the major components in the auxiliary building with potentially high radioactivity are those in the Amendment 5 12.3-16 August 1981 08-06-81
/pe)IATION PROTECA8gEMENTAL DESIGN FEATURES EolT SEGTtOM chemical and volume control system, the shutdown cooling AUG 3.1 198j system, the fuel pool cooling and cleanup system and the lw OUy primary sampling system.
Shielding is provided as necessary around the following equip-ment in the auxiliary building to ensure that the radiation zane and access requirements are met for surrounding areas.
A. Letdown heat exchangers and piping B. Purification, preholdup, and deborating ion i
exchangers C. Chemical and volume control tank D. Charging pumps and piping E. Shutdown cooling heat exchangers F. Chemical drain tanks and pumps G. CVCS and radwaste filters H. Spent fuel pool cleanup ion exchangers and filters I. Spent resin tanks and piping J. Gas stripper K. Seal injection heat exchanger L. Boronometer M. Process radiation monitor N. Seal injection filters
- 0. Crud filters, tank, and pump Shielding is based upon operation with maximum activity conditions as discussed in sections 11.1, 11.2, 11.3, and 12.2.1.
Zhpending on the equipment in the compartments, the access Aries from Zones 2 through 5. Corridors are shielded to Allow Zone 2 access. Operator areas for valve galleries are limited to Zone 3 access. Frequently operated valves in high radiation areas are provided with reach rods extending to Zone 2 or Zone 3 areas.
August 1981 12.3-17 Amendment S 6-22-81
~,(>
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES Removable sections of block shield walls, or concrete hatches with offset gaps to reduce radiation streaming are provided for replacement of ion exchangers, backflushable filter, pumps, and heat exchangers.
12.3.2.2.4 Fuel Building Shielding Design Concrete shield walls surrounding the spent fuel cask loading and storage area, fuel transfer and storage pools, and fuel transfer tube between the containment and fuel transfer pool are sufficiently thick to limit radiation levels outside the shield walls in accessible areas to Zone 2. Access to the fuel transfer tube through the concrete radiation shield is provided by a heavy concrete hatch through the roof of the shield as shown in figure 12.3-23. The hatch is labeled to caution maintenance personnel that there are potentially lethal radiation fields during fuel transfer.
Water in the spent. fuel pool provides shielding above the spent fuel transfer and storage areas. The relationship between dose rate over spent fuel during transfer and depth of cover-ing water is shown in figure 12.3-24. Radiation levels at the fuel handling equipment are limited to 2.5 mrem/h.
The spent fuel pool cooling and cleanup (SFPCC) system (section 9.1} shielding is based on the maximum activity discussed in section 12.2 and the access and zoning requirements of adjacent areas. Equipment in the SFPCC system to be shielded includes the SFPCC heat exchangers, pumps and piping. (SFPCC filters and ion exchangers are located in the auxiliary building.)
12.3.2.2.5 Radwaste Building Shielding Design Radwaste systems are principally located in the radwaste building. Additionally, the boric acid concentrator and the boric acid concentrate ion exchanger are in the radwaste building.
Amendment 5 12.3-18 August 1981 08-06-81
PVNGS FSAR tNCREMENTAL RADIATION PROTECTION DESIGN FEATURES Eo)T SECTION E. Design of potentially contaminated equipment incorporates features that minimize the potential AUG 11 198l for airborne radioactivity during maintenance lM OUT operations. These features may include flush connections on pump casings for draznxng and flushing the pump prior to maintenance, or flush t onn piping connections pi systems that could become highly radioactive.
12.3.3.3.2 1'o Gux'd e znes o Control Airborne Radioactivity A. fl The axr ow xs 'irected from areas with lesser potential for contamination to areas with greater potential for contamination.
B. In building compartments with a potential for contamina-tion the exhaust is designed for greater volumetric flow than is supplied to the area to mxnzmxze the amount of uncontrolled exfiltration from the area.
C. Air cleaning systems criteria for emergency systems are discussed under Regulatory Guide 1.52 in Section 1.8. System design for both normal and emergency systems is described ~n section 9.4.
D. Means are prove'd e d df to buildings upon xndxca won isolate the containment and fuel radioactive contamination to prevent the discharge of contaminants to the envxronmen annd minimize in-plant exposure.
E. Means are prove'd e d to isolate the control room to minimize xn 1 ea Rage of contaminated air to the operator.
F Suitable containment xsolatzon valvess are installed to that the containment integrity is maintained.
See additional discussion in sections 3.1.47,.
and 6.2.4.
12.3-25 7-30-79
PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES G. Redundancy and Seismic Category I classification features are provided for components of the HVAC systems required for safe shutdown of the reactor plant.
8 12.3.3.3.3 Guidelines to Minimize Personnel Exposure from N
HVAC Equipment Access and service of ventilation systems in potentially radioactive areas are provided by component location to minimize operator exposure during maintenance, inspection, and testing as follows:
A. Ventilation equipment rooms for outside air supply units and building exhaust system components are located in radiation Zone 2 and are accessible to the operators. Work space is provided around each unit for anticipated maintenance, testing, and inspection. Filter-adsorber units generally are consistent with the recommendations of Regulatory Guide 1.52 for access and service requirements. (Refer to Section 1.8).
B. cvocal HVAC equipment that services the normal build-ing requirements is located in'reas of low contamination.
12.3,4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION Section 11.5 describes the function, operation, design, and locations of the radiation monitoring system (RMS).
The RMS provides 33 fixed area and 14 fixed airborne monitors per unit as noted in section 11.5. Additionally, there are 18 portable monitor connection boxes located through-out each unit that can connect any of the three portable area or one movable airborne monitors available for each unit for use during extended maintenance. Thus, a portable monitor can be
)
Amendment 5 12.3-26 August 1981 7-7"81
'3 I
0
EDlT $ ECTlPg
>0~11 198>
I IPilfl 1N~UT MHT 2 RPWIT)
IQQ pi 8
j MAM II t.. W-y I ~
SVi UC.
, ~. It, AREA RADIATION MONITORS
$ ~
ELEV.
AUXIUARYSLDC.
Al PT. 0 M. LEVEL 0 MONITORS I'REA AREA RADIATION RADIATION IIONITORO I
'%i ~I'+":IIIV' OO
)Fp' t
Qi Ifae
+ TOK I
AREA RADIATION MONITORS
~ ~
2 CONTAINMENT CLDC IO01
~s REACT N VESOEL 21 CAVITY COOLING PANS PPE CHASE 22 WET LAY
~ ~ 20 21 CONTROL CLOG. NNMAL AHU5'TER 21 26 'fRANSfCRMOIS 16 L4iifiNCPAiifl OiSiRiCU.i0ii PANfl IP DG CLOG OUTSIDEAiRW)SUER UNIT 11 GAfE n ~TAMILCLOSE ESK SWGR PMl ESSENTIAL AHU'5 20 AUX, STEAM CONDENSATE RECEIVER TANK 16 2UNCTKNI DOK n ESSENTML NKLED WATER DPV~TKH PUMPS 30 CHEMCAL NAiN TANKS ID PiPE TRENCH 5 PKTICATES CONTROLLED ACCESS 2K CHf MKALADOifeN TANKS 31 fSSENIIAL AIR CONDITCHING UNITS 25 CONTIKKRXM ESSENTML 32 ECW CHf MICAL ATOTON TANKS AND/DR ROLL UP DOOR AHUS f SSENTML CHILLED WA'IER ECW PUMPS .13.N.SD~1 0 AUXIUART 5 LDC. ELEY. 51 FT. 6 M. LEVEL C 21 ESSENTIAL ~IS EXPAASKPI 1ANK 34 ECW PUM1 ROOM ESSENTIAL LC.U. ECW RAOIATiOil MOhiTORS REV 36 REACfN MAKE UP WATER PUMPS CONC ACID MAKEUP PUMPS 31 35 5'iUTDOWN COOUNG HEAT EXCHANGERS Palo Verde Nuclear Generating Station 31 GAS STRiPPER 40 CAS Sf coif R COii THOL CASiNET FSAR A'I GAS 5'f I@OPER KNSTAUMEIi'I RACK i RADIATION ZONES (OPERATZON) BETWEEN EL 40~-0" & 1QQ'-0" CHANGE Figure112. 3-1 August 1981 Amendment 5 8-06-81 0 , e ,0 ~ . (\ t ttp t ,I)t (t h AREA RAOIATION MONITORS (NCIIK66KPLYI(!I EDIT SECTION,~'UG 1 ( )a N 0>> ~ 1'!I~UT ~ Qi ~ IP V>>ll I i' Nt IPttt I NP tati 1 1 - - O'. - D(S;- i ( I~ I II ~o ~ I >> I I $ >>i rj (lI) 43 I TURBINE BLDIL I 6 0>>>> ~ Q>> I Q. , I --=.'-. (8)- Doe1 (!oO OQOQOoOQ.
- j, 4 i>>wIP
- , ---0, 0-.
2 CHlctttQc i, ~ I! jp ~ ~ ~ tatttl'tA 8EIEKI 4 b(bg AREA RADIATION Di~lf I MONITORS ~ I-" c"'. CONTROL AND COBRIDOA BLDG CBI. lc 't>> i Q>>QQ>>>>>>i iQQ' D C. Eoltp. RM OtAV>>EL -C-INES EL -A-NO. CONTLN MENT 0 . I 1 I IL 0 C. EQUIP RM CHAN>>if "D" O.C. EGVip RM CHAN>>EL l -A-20 21 22 DIESEL GEN. TRA!N DtcSEL GEN. TRON HIGH VOLT. CV6CLE -8" 20 tl 22 RKACIQP VESSEL 6'siAM Gf>>ERAIQR NQ. I '5 STKEAM GEVERAtoa No. 2 O.C. Eottl'. RM CHAK>>EL "8- 23 CONTROL PAhEL 23 ILC. PUMP>>Q. 18 ~Qi Iji 2C>> 'I BATTERY RM, CttANVEL -C" BATTERY RM. CHAV>>EL "Q' 2l NVORAL GRQ. TRANSP. 2l 25 8 C. PUMP ho. 1A A C. PUMP liO. 28 I I Qi( II 0 ~>>>> Q.
- 0
~0 27 25 29 BAT TERY AM. OtlkhEL RM. CHAN V EL ESF SWGIL RM TRAIN "A" ESC SWGIL AM. TIVVV "8- "A'ATi(RY No. 20 AAQWASTE LOW ACTIVITYSPENT Rf5IN TK. HtGH ACTMIYSPENT RfstN IK. '1 26 27 26 29 R.C. PUMP lio. 2A SAFOY INIECttok TK. NQ. IA SAFETY IIMECTOI TK. NO. IB SAFOY INIECTCN TK. NO. 26 I q CONTROL BLD(L DP 30 31 REMQT E St!UTOOW>> RM, OUTSIDE AIR CHASE 22 23 SPEVT Rfstk TRAVSFER PUMP 30 SAFOY IV)ECIKW TK. No. 2A ~ ~ DLI SVIV. OFF Cf 24 INTERCEPTOR DECO>>. SPRAY (CQIH I 31 32 PRKSSVattfa C 5 8. (AYDQWN Aa(A 46fih S3 34 FOREMAV OFFICE IVS'f REPAIR SHOP 25 26 ~ fOOL RM. ANO SUPPLY SIQR. lRS PECYCLE ANQ MONITOR IK. PUMP 33 34 U.G S. (AYQOWN AREA EMEPGKVCY lOCK I' 35 VITAL IkST. STGR. 27 las WASTE HQLO.UP PUIIP 35 FVKL VPENDER Aho CARRIAGE 36 MACH. Sttop 28 Las CO>>CEATAATE Moktfoa PUMP 36 RKGE>>ERAIIVE It(AT EXCH i Qi 37 SIQA. 29 LAS CQNCEhTRATE MOVITOR TK. 37 TOP. 5'tORAGE RAO( 30 NKUTRAUZlhG SUMP 36 CEA ELEVATOR 31 HOLDUP IA>>K 39 CEA OSPOSAL CA>> J 0>> I >> 32 R.W, FEED SCQ 40 FUEL STNLM RACK t~p3. 33 10 TON TAA>>SFER CAAf 41 TEAM>> GALlERV NORMAL EIH. FAN 34 55 GAL DRUM STQR. (HIGH LEVEL) 0~! ~ 35 Cot 5'foA. '342 tENQQN GALLERY NORMAL SUPPLY FAll TEMP. CO>>sfa Dp>>G. DtfSEL CEN 36 CEMEliT STDA, u 24" 0 MLN FEEQWAIER ~ UC. 37 TRVCK loAIXNG BAY 45 INOX>>K CHASE AREA RAOIATION 36 f MPIY 55 GAL DRUM 5'foR. 39- 55 GAL DRUM STOR. (LOW LEVfl) 80 100 FEET MONITORS DRY WASTE CDMPACTOA $0 2 ESOVPPION APE CHASK G RAP H I C SCA LE NO. TURBINE HCH COhQVCTIVITY RACK FUEL BLOC 3 ELEC OuSE NQ. AIIXIUAAT AUX F.W. PUMP HYAC. OuSE HtGH SQO VM ANALYtfa TRA>>SF(A TUBE 20 Ht RECOMBtkf 21 . BLQthQ(NN Sf ACK POWER AST Pkl. O FAILED FUEL CQNTAikfa 5 SUMP 21 ESSENTIAL ACU tt RAD Moh"Toa 22 HVAC Ak>>VC. 80. SEOVE>>CE CONT. PVU .,6 VOLT. AEG. Stoa. RACKS (A.CODE AST, SVRV. SAMPLESI 6 7 ELEVATOR Mo>>QRL" 23 flfC PfhET R4 TRLN "8" 23 2l sEAY ivio 145T Ata GOMp pf WAStE COLL TK. 23 TOOL SION 6 VALVE CAlLERY ('ATON TK SPENT FUEL stoa RACKS 24 COVCEV faATQA PA>>EL 9 MCC OE>>DIES UP U N 0. 25 AFIERÃXEA Akto>> TK 25 CASK 25 Itt AECOMBthfa CG>>TROL PA>>EL 10 LOAD CEVIER l>>D'CATES ONE WAY 000>> QR AIR RKC, Mtx Aho HOLD TK 26 CASK LOAot>>G Pit 26 Ht coMPRESSQR 27 A!R DRYER lQW IOS SUMP 27 II SFWICHGEAR COVTRQLLfo FROM ASIDE QF 27 Ht PUKE EXH. 26 HTR, DAV. 'fK. I '52 HIGH TOS 5UMP ts STORAGE RM CASK PERSON>>fl BARRIER LAYQQWN lt EOV'P OR ACCESS OPE>>i>>G BLDG GATE 28 POST LOCA AVALYKEA AMMONIAADDII. PUMP 13 TAAVSM'TIER RICK DH ESSENTIAL AIR FILTRATK)>> V>>II 29 CVCS CHEM A(OT, 29 30 HYDRAZINE Aoott PUMP tt 55 SI CO>>TRQL PNL. CAUS T IC PNL. 30 U 9 A A>>Q LQWfR CAVSH Co>>E LAYOOW>> ll VOLtAGE Rf G IIICCATES Y CE 30 OIAAGI>>G PUMPS 15 ltELAY CAB 31 STM. PKG EXil Dak IK. 1 AOO SKID 3I ILIL RAtLS 31 La Icx>>QLLVG Vhtts Hla. Dak PUMP 55 16 LiGttTAG PA>>(L 32 'faANSFER SWITOt S2 33 F W PUMP TURB. -I 56 if CQ>>0. POLISH. DEMIN, TK FUTURE WA'Ls 32 33 SPENT FVEI. POOL HEAT EXCH SPENT fUEL (OMVG PUMP 17 Dsf it'Bt!t to>> PA>>f L 13.M%DR 002 REV 0 CAVD PUMP 16 30>>CTI(>>i BOI 34 CRUD TANK 1 W. PUMP TURB. 56 COND PUMP 5TRAthfa 3l POOL CLEANU'P PUMP 19 Ptpf faf>>OI 35 EHC POWER UNtt Coho PVMP5TRAI>>fa 35 hEW FUEL ELEV. 35 SEAL thl It(AT EXCH. 59 36 LOQQw>> HEAT EXOL 36 FW PVMptURBLO CENT. 6!I R R Act(5 t>>VtfA. BOA Pa@ Verde Nffctear Generating Station 37 HOPWELL Coho A>>ALYIERS 37 ECW>>fit fxCH 36 COVQ(hsf R "A" PORTA ~~ CQV IOL CK>>f fa FSAR 36 SCR Potttta CO>>TROLLER 39 Co>>Of>>SEA "8" fvas CQOU>>Gwta pvMP ) 39 H(AI(R PA>>fls CQV;i(NSER -C- Swtf IiGGVI RM lo lCHILVG TAA>>SF 41 VACUUM PVMPS BA'flfRYRM 41 PfhO LAR 42 FtlfURE SPACE (COND IUBE CLNG EQVtp Met>> SIM Uk(5 ~2 ELEC. PEhET RM TRAIN "C- I RADIATiON ZONES (OPERATION); 1 AT EL 100'-0" CHANGE Piglsre 12.3-2 Al:glsst 1981 Amendment 8-06-81 0:-- ~' i fNCREMEgr AREA RADIATION ~ MONITORS PtIn SE!,"tloN f,i RVG11 1981 <<Si Qr l!<~UT i Oi a airah Cr i'I TURBINE BLDG. I ~ Z"Paav'-. 0a g, <<aM>> 0 I ~ f ~ 4 I I . AREARADIATION-
- MONITORS I i Qi 0i'i 0.
I 'gl-EIE Ll ., I;A t,llf, itl O ~gfflffEIII'af ~ I' I 0 CONTROL 0a I BLDG. 0 li O )i i0 > I CHANGE ! Figure 12.3-3 August 1981 Amendment 5 8-06-81 Jff D c I ~ T " ygS" 'm I )I )2' f I .lNCREMpg~@ 'AREA RADIATION 'MONITORS
- HIGH RANGE ARMS f (AREA RADIATION MONI OIIS EDlT SE'eTION. l AUG] g l
j N~g i%6Aiihhti%%% .Xc ' X ~N". I~Cg Qp Il i ~ 'J Qi I 0$ Qi 8 Qi Qp iip IÃ. I; '.- '""'-:::-" ": I'~C'P 'i-;.:;P~ . 4 C(NaliN N lh1 8(DG. 8VP STRVC. E ~ I ~ Op i ~ ZX -I I AREA RADIATION. OI H Op MONITORS ..~ '<< ~ ~ '".'Y(UKIUART8L(C. (GJQS0 0 r..:g Pp .P p Qi I p ~p CONTAINMENT SAFETY INJECTKJN TANK EXHAUST MDNHOA
- . @ RIDJi,:ESJ 40 '(GJ'. C 0. ~ 21 MISSILE SCf LO 21 f WAST GAS DICAY 'TAhK
~ W% Ait255 NORlllL AFU 22 WASTE GAS SuRGE LANK p STEAM GEKERATOR hTJ. 1 EVAPOPATORSIJRFACE CONDENSER I CEDM NORMAL ADJ UMT AND FAT EVAPORATOR VAPOR 800y FVEL 25 WAStE GAS COhtppcSSOR ~ i 4 ~ H2 ANO 02 Ahl(YffR ~ ELEVATOR MACHINE RM. Oi QI TRANSFER SYSTf M CDKTROL COKS(KE , ~:4',' 4 ~ HYDRAVLCPOWER PACK Qp CLOSED HIM LIFT RC ASSEM8' AND GEOM DfCON AREA 20 EKNALtST MOIJaOR C(paROL 8(DG. 0 30 REACTOR Ac&0 (AYDOWN AND SEAL RING STORAGE 21 SUPERVISOR OFFICE ~OI OO OJj lt((NO ~ 31 32 STEAM GENERATOR ho. 2 CEA CHANGE PLTF M 22 23 KITCHEN PAIi1RY 202 I 24 WOMEN"5 TOILET PRESSVRITEA REF UEL MACH. 25 MEh'5 T(KLET 26 CONFEIUNrof ROOM U ii i TUR8INE 27 28 COMPUTER ROOM VISTRVMfhT RfPANI ROOM 0 E3 II 20 HP 29 STORER(OM ~4 TORSI'P 21 T URLNE. 30 COJaJKR ROOM 22 Rf NEATER DRAIN TANK (NESEL 23 NOSMAL AJR HAIioliNG UKTT 24 Mlih TUR6. LO M5T ElIMVilTOR 20 R(SW FRL (TYP6 PLCS) 25 EKNAVST AID. MONaoR 21 EKHAVST MONITOR 26 GENE RA(OR 22 DOMfS'IC HOT VRTER HfATER 27 COLLE CTDR HOUSE 1. HATCH 23 SAMPLE RACKS 2. PIPE CHASE 74 TOILETS 3 ( DIESEL GENERATOR ELEC. CHASE SHOWERS-4 HVAC. CHASE 25 I 5. AREA RADIATION 26 ORYIhG 20 EKHlVST SILE NCE R 6. SUMP flfVATOR 27 Mfh"S lOCKER AxtMS ND. 1 AND 2 21 CON SVST ION AIR Vi1AKE MONITOR 28 WOM(h 5 LOCKER ROOM 22 ESSENTIAL EXHAttST FAN 7. MoliORAIL I 29 CLOTHWG 23 N(NIMll.EKHAJSl FAN 8. VALVE 6ALlfRY 9. ISSVEi'0 JANITOR 24 AIR DRYER MCC 31 CLEAN STORAGE l 25 FUEL DIL OAY TANKS 10. LOAD CfhTER 32 25 START(h~p AIR PACKAGE 11. SWITCIPJEAR I STORAGE 3 33 COT ROOM
- 12. EOIJ!P. OR ACCESS OPENING 34 LUNA RXSl FUEL 13, TRANSVaTER RACK I SPECIAL H. P. OF FICE I~ . VOLTAGE REG 15, RELAY Cl8 I
36 HEALTH PHYSICS'7 20 SPEh1 FUR FOOL Hi51. RESP. MATJT. ISSUE 21 SPEh1 FUEL HlhtJLING MADIJiE 15. LIGHTING PANEL
- 17. DISTR 8IRKHJ PANEL 38 OFF Cf 22 CASK LOADOIG PIT I& JVKCTKJN 80K I
39 SAMPLING ARfA 23 NfW FUEL STORAGE COVERS
- 19. PIPE TREN H 40 FIRST AKJ 24 CASK LIFTING RJG (AYDOWN AREA
~I Pf RSOKNE L DfC(tN ARfA 2 25 DECON PIT DENOTfS UI' N 0. GRAPHIC SCALE 42 H P. SAMPLE CC(ihT, ROOM 26 OECOh PIT COVER ~ D NiDICAYES Okf-WIYDOOR OR Cot(TAO((EO l 43 HOT LA8 27 GATE STORAGE P F AOM IN&XOC 8JJG 44 SAV R f R(OM H3(O AND SINK 28 NlW 1Vfl INSPECTKJN PIT I 45 FLAMVA8icSTORAGE 29 NfW FUEL COATAINERS LAYDOIYNAREA 64ATE ~6 CYLLNDER STORArf 30 Upfko(R HYD. (NEV VNl( I POOL C(OLING PIJ&CATKJIJ PANEl INDICATES C(34TRDLLED 47 48 PIPIhGt AREA COVNTING ROOM1 31 CANAL ACCESS AND/DR ROLL UP 23-RTCDR~ REV 0 49 TO(R Aho EOJIP'OOM 33 IIEW FUEL ELEVATOR DOOR 50 " D(CON Sl Of Coh. ROOM EM(ICE<<CYSHOIJERAND fYE WASH Verde Nuclear Gener<<lng S>t'~ 'alo 54 COLO(48" COKIE(OR 5 FSAR NSOZSTZON ZONES (OPERATION) EL 140'-0" S 200'-0 'ETWEEN CHANGE 12.3-4 -'igure August 1981 Amendment 5 ' 08-18-81 m R (~ J I' ~ f ( ~ J ap ~ k pl g 1 ,A 0 . aL E ...- ---lNCREMENTAg EDIT SECTION pra/u ZTd'am AUG 11. l98) p yp zP/cs) IN~UT C4 t S/IW <Cu , sF~ WiA/ /47 1 ~ ~ P ~ 'I ,. c. opzAp pz. Ec. /4o'-i)" 4 .:II ~ r~ I I I FC ZC-/2d'-ll' L ~ r 0 e c;j: 'n!'A ~ AJL7ZZC E A'OV'4'D Fg. $ L. /Od-'D gC. /0/'" ~ 0 c~ r t,~.r~gr .r, re Ml', J I Cl 4 Q5A CZAR ','/ "0 'rI~ ~ ~ ~ I I c.. I I I I I Ir I~ ~ 4 ~ ~ ' JC dz. BT-0" I 4 I I I ~ I f ~ ~:f r w-~ I I I I I I +~ ) ( c-> FZ. E'C . 8P '- 0" t IL ~, I I f,p
- Ou'alo I
0 M1 ~ ~ 4 ~ 1 ~ L W -.r. 4 REACTOR CAVITY RIEL" r', ~,'c /~"CUE Verde Nuclear Generating Station ..~srgu~ruzA ra 4 f FSAR REACTOR CA".:TY SHIEM Figure: 2. 3-21 August 1981 Amendment ~ 8-06-81 i' ~%K ~ ~ ~v a.-A ) I' tNCaEMKN~4L E01T SECTION AUGIE tSst lM 0UT ~ ~ ~ ~ e ~; ~ ~ ~ ~ ~ ~ ~ ~ 4~ ', ~ 0 ~,0 ~ ~ FUEL TRANSFER 5'" ~ ~ CHANNEL ' (REFUELING )~ POOL) ~~ ~ 0 ~ 0' 'TEEL ~ ~ ~o ~, 0 PLATE I ~ O ~ 'O, ~ I, I l TRANSFER l TUBE 12" I I ~ ~ I I l l l 3" SEISMIC GAP ! 'ONTAINMENT I ~~ O ~ I WALL ~ ~ ~ ~ I I '"';:I I Palo Verde Nuclear Generating Station FSAR FUEL TRANSFER SHIELDING (IN CONTAINMENT) Figure 12.3-22 August 1981 Amendment 5 wc"~~ it ~ 'F \ f INC~ gpss SECS' ~UG y y )98') OUTSIDE REMOVABLE CONCRETE ACCESS PLUGS 8'-O .'A t+N <-Op CONTAINMENT ,4 ~ 0; .P. Pfw ,r ~ 5' ~~ \ 3l+It oo FUEL, ~ < '-S'.'ONTAINMENT w 0 ~ a.or 'FUEL :ei '.4 1'o 0~ BUILDING jy ~ '.% ' v'k: e, ~ BUILDING ~ TRANSFER TUBE TRANSFER TUBE .Q'0 %. ~ 0'g ~ 5'-B'i+ll 't+tf p' ~ T k'. i:o+ 0 p ~ SEISMIC GAP 0 r.. ~. ~ SEISMIC GAP 3'-IP STEEL PLATE 0, ~ ~ ~ he.t AUXILIARYBUILDING PaIo Verde Nuckar Generatxni, Stabon GROUND PLAN VIEW FSAR FUEL TRANSFER TUBE ELEVATION VIEW ZNSPECTZON FACZLZTY Figure 12. 3-23 August 1981 Amendment 5 8-06-81 0 il, f f t I 10 or 104 10 102 mRem hr 10' 10 10.1 102 2 4 8 8 10 12 14 16 18 20 22 FEET OF COVERlNG WATER Palo Verde Nuclear Generating Station NOTES: a. Dose from 4,100 MWt APR fuel assembly with maximum radial and axial peaking ~Y/4<. FSAR
- b. Minimum 9 feet of water cover maintained (assumes one foot freeboard clearance over storage racks and VERTXCAL DOSE RATE pool at low water level alarm poly PROM 1 SPENT FUEL, ELEMENT <a>
- c. Feet above active portion of assembly e Figure 12.3-24 August 1981 Amendment 5
" jJej .o w g(p~ ~ r'.~" 1>CREMENTAL PVNGS FSAR EoiT. SECTION AUG 18 198l 12.4 DOSE ASSESSMENT lN OUT, 12.4.1 RADIATION EXPOSURES WITHIN THE PLANT Within-each accessible area in the plant, the peak external dose rate due to direct radiation is considered as the maximum dose rate for which the area is zoned (section 12.3.2). These dose rates are not expected to occur during normal operation because the plant shielding is based on maximum coolant activities corresponding to 1% defective fuel cladding. The annual average isotopic concentrations of fission products are expected to be much less than the maximum. Therefore, the actual dose rates in a given zone are expected to be signifi-cantly less than the maximum calculated dose rate in that zone. Another source of radiation exposure within the. plant comes from airborne radionuclides. Radiation exposure from this source to occupational workers in the accessible areas of the plant is usually insignificant in comparison with the exposure to direct radiation from radioactive sources. Under certain circumstances {e.g., operational access within the containment), doses from airborne activity may be a major fraction of the allowed limits. Based on operating plant experience, numerous design features are incorporated into the design of PVNGS to minimize plant personnel exposure. The PVNGS design reflects attention to ALARA detail due to the ALARA reviews conducted during design as noted in appendix 12B. Specific design measures taken to minimize doses during maintenance and operation include the following: A. Radioactive systems are designed such'that the components which remove fission and corrosion products are placed in the process stream as early as practicable. To reduce personnel exposure time at valve stations, motor-operated or pneumatic-operated valves are used where practical. Where manual valves are used, Amendment. 5 12.4-1 August 1981 +ji~4~ i ~ ) PVNGS FSAR DOSE ASSESSMENT provisions are made, when necessary, for shielding the operator from the valve by use of shield walls and valve stem extensions. C. Gauges, instrumentation, and sampling stations that require frequent visual inspections are located in corridors or on local or central control boards to minimize exposures. D. For valve maintenance, provisions are made for drainage of associated equipment. E. Temporary and portable shielding is provided for use by plant personnel during maintenance. F. Radioactive liquid and gas piping is routed to minimize radiation exposure to plant personnel. G. Other specific design features to minimize exposures are described in section 12.3. These included a remote filter handling system, reactor cavity shielding, and a delay in CVCS piping to allow for decay of nitro-gen-16 prior to letdown fluid leaving containment. The following examples of design changes are typical of those made due to ALARA reviews: I A. Permanently installed breathing air connections inside the containment. B. Installation of jib cranes inside containment to speed maintenance. C. Permanent. shielding and remote cleaning has been provided *at the reactor vessel head laydown area. From operating plant experience for pressurized water reactors (PWR). between the years 1970-1974, the distribution of person-nel and man-rem doses according to functions for light-water reactors is presented in tables 12.4-1, 12.4-2, and 12.4-3 (1) Based on this operating data and on reference 2, the total O. Amendment 5 12.4-2 August 1981 8-17~81 INGREMKfGN PVNGS FSAR EDIT SEGTIGN DOSE ASSESSMENT man-rem dose from PVNGS is estimated to be 193 man-rem/year-AUG i 3 198] unit for station personnel. A survey of man-rem dose to coh=DU tract maintenance personnel based upon the 1979 data of NUREG 0713 is shown in table 12.4-3A. The table indicates that, on average, contractor doses are approximately 285 man-rem per nuclear power plant (BWR and PWR, combined), annually. 5 This tabulation does not reflect credit for the dose reduction measures designed into PVNGS (refer to section 12.3) and may be considered an upper bound of contractor exposure. Thus, total annual exposures for station and contract personnel are projected to be less than (193 + 285 =) 478 man-rem/unit. A breakdown of this estimated man-rem by job .category and radia-tion zone is presented in tables 12.4-4 and 12.4-5. The assumptions used in determining these doses are presented in table 12.4-6. 12.4.2 RADIATION EXPOSURE OUTSIDE THE PLANT 12.4.2.1 Construction Worker Doses The annual individual external dose and the immersion dose received by workers at units under construction from operating units have also been estimated. These estimates, based on 2000 h/yr exposure, are summarized in table 12.4-7 for Unit 3. August 1981 12.4-2A Amendment 5 5 8-06-81 PVNGS FSAR DOSE ASSESSMENT This page intentionally blank Amendment 5 12.4-2B August 1981 6-30-81 PVNGS FSAR lNCHENiENTAL DOSE ASSESSMENT Table 12.4-2 YEARLY AVERAGES AND GRAND AVERAGE FOR NUMBER OF PERSONNEL,d6 i ~ ~~ AND MAN-REM DOSES FOR OPERATING PWR PLANTS {a) Qg~ Total Total Average Average No. of No. of Man-Rem No. of Man-Rem Year Units Personnel Dose Personnel Dose 1970 985 844 493 422 1971 750 822 250 274 1972 1 603{b) 2,408 401 482 1973 4,601 3i620 657 517 1974 13 7,653 5i365 413 1970-74 30 15 592(c) 13,059 538 435 a~ This table is based on the data given in table 12.4-1.
- b. The entry corresponds to 4 plants only, since no information on personnel is available for Point Beach, Unit 1.
c~ The entry corresponds to a total of 29 plants only. Unit 3 was chosen as it will have the highest dose rate during construction of any of the PVNGS units. The total man-rem dose to construction personnel, based on the estimated labor requirements shown in table 12.4-8 are summarized in table 12.4-9. 12.4-5 PVNGS FSAR DOSE ASSESSMENT Table 12.4-3 DISTRIBUTION OF PERSONNEL AND MAN-REM DOSES FOR VARIOUS FUNCTIONS OF OPERATING LIGHT-WATER REACTORS Percentage of Percentage of Work and Job Function Total Personnel Total Man-Rem Routine operations 19.2 14 and surveillance Routine maintenance 34.5 45 Inservice inspection 1.44 3 Special maintenance 28.71 20 Radwaste handling 2.08 Refueling 14.07 14 a~ The basis for this table is the information provided in Table 5 of reference 1 for 39% of the total exposures from lightwater reactors in 1974. This includes PWRs and BWRs. Amendment 5 12.4-6 August 1981 EDIT SECT)08 PVNGS FSAR INCREMENTAL I',UG 10 >SM 0 UT. DOSE ASSESSMENT IN Table 12.4-3A MAN-REM EXPOSURES TO CONTRACTOR PERSONNEL, 1979 Work and Job Function Total Man-Rem~ Average Man-Rem~ Routine operations and 487.3 11.6 surveillance Routine maintenance 1794.2 42.7 Inservice inspection 1602.4 38.2 Special maintenance 7023.9 167.2 Radwaste handling 311.5 7.4 Refueling . 768.7 18.3 Total 185.4
- a. From Table 8 of NUREG 0713
- b. 42 units. Refer to Table 5 of NUREG 0713 August 1981 12.4-6A Amendment 5 8-%g-81
PVNGS FSAR DOSE ASSESSMQNT Table 12.4-4 ESTIMATED ANNUAL PVNGS MAN-REM DOSES BY JOB CATEGORY ; Percentage of Category Man-Rem/Unit Total- Man-Rem/Unit Operations N 99.1 51.3 Maintenance: Routine (42.1) (21.8) In-Service (2.8) (1.5) Inspection Special ~16. 7 9.7 Total (63.6) 63.6 (33.0) 33.0 Radwaste handling 9.9 5.1 Refueling 20.4 10.6 Total 193. 0 100.0
- a. Breakdown of maintenance exposure is by the relative ratios of routine and special maintenance and inservice inspection shown in table 12.4-3.
- 5) Amendment 5 12.4-6B August 1981 8-06-81
Table 12.4-5 ESTIMATED ANNUAL GAMMA DOSE TO PVNGS PERSONNEL (3 UNITS) Expected Expected Mumber of Manhours of occupancy per Year Average B Job Cat o Estimated Dose Rate Maintenance c Annual Radiation in Zone Radvaste Total Man-Rea Zone (mrem/hr) Operation Routine Inservice Inspection Special Handling Refueling Manhours Exposure't b 0+13 606, 140 86,030 5,735 38,235 16,200 34,400 786,740 102.3 0+63 25,800 14,294 953 6,353 5,800 5,400 58,600 36 ' 4.0 18,832 17,500 1, 167 7 g 777 3,620 7,472 56,368 225.5 25o0 2,612 1,244 83 553 380 660 5,532 138.3 100.0 616 50 22 0.0 68 760 76.0 Total 654,000 119, 118 7,942 52,940 26,000 48,000 908,000 579,0 (3 unit) oa 297.3 126.2 8.5 56.2 29.7 61.1 579.0 579,0 I O ~ 4 of Ch 9.7 5.1- 10.6 I I Man-Rem 51.3 21.8 1.5 100 tO W 193.0 per unit
- a. Contractors not included. Refer to table 12.4-3A. ts
- b. Detailed breakdovn by job classification presented in table 12.4-6.
- c. Breakdovn of maintenance exposure is by the relative ratios of routine and special maintenance and inservice inspection shovn in table 12.4-3.
cn coo CD n CO Table 12.4-6 RADIATION ZONE OCCUPANCY BY JOB CLASSIPICATIONS (Sheet 1 of 2) Percentage of Time Spent in Zone I Classification of Personnel Number Category (b) Station Management 0 100 Support Services, Training and Security 114 0 100 1 0 97 1.3 0.5 0.2 Scheduling and Licensing 0 97 1.3 0.5 0.2 uality Department 13 97 1.3 0.5 0.2 ngineering and Technical Services 3 0 97 1.3 0.5 0.2 13 0 100 20 0 95 1.5 0.5 Refer to figure 13.1-6. The numbers in this table represent typical expected personnel. 0 = Operation M = Maintenance RH = Radwaste Handling R = Refueling P JNGS FSAR SEcTtON ED@ Bvc PUG 10 ]90l 1cpF'N , 12.5 RADIATION PROTECTION PROGRAM OUT. 12.5.1 ORGANIZATION 12.5.1.1 Pro ram Administration II The Palo Verde Nuclear Generating Station (PVNGS) organization is presented in section 13.1.2. The Radiation Protection Supervisor is responsible for station radiation protection pro-gram administration. He is also responsible for ensuring that station operations meet the radiation protection requirements of 10CFR19, 10CFR20, and 10CFR50 Appendix I. Commitment to the recommendations of Regulatory Guide 1.8 is discussed in section 1.8. The commitment to the philosophies embodied in Regulatory Guides 8.2, 8.8, and 8.10 and the authority to implement them are discussed in section 1.8. The Radiation Protection Supervisor will implement the radiation protection program within the units. He will prepare necessary reports and procedures within his area of responsibility. The Radiation Protection Supervisor and the personnel reporting to him will conduct the daily functions associated with the radiation protection program including radiation surveys. and associated sample collection and analysis. Backshift radiation protection surveillance will be provided by radiation protection technicians. Responsibilities and authority of the Radiation Protection Supervisor and the Supervising Radiation Physicists and the qualifications of the personnel holding these positions, are discussed in section 13.1. August 1981 12. 5-.1 Amendment 5 8-17-81 PVNGS FSAR ~ ~ RADIATION PROTECTION PROGRAN 4) I 12.5.1.2 Pro ram Ob'ectives Objectives of the radiation protection program are to ensure that personnel exposure to radiation and radioactive materials is within the, requirements of 10CFR20 and that such exposure is kept as low as is reasonably achievable (ALARA). Further-more, the objective is to control station effluent releases which fall under the restrictions of 10CFR20.106 and 10CFR50 Appendix I and to ensure that these releases do not.exceed the limits of the,'station radiological effluent technical specifications. 12.5.1.3 Radiation Protection Pro ram The station radiation protection program will be officially initiated when appropriate portions are implemented to receive radioactive material licensed to APS and will be in effect continuously jthereafter until the units are decommissioned. This program ',consists of rules; practices, and procedures that are used to accomplish objectives stated above in a practical and safe manner. ~ The radiation protection program will ensure that: A. Personnel permitted access to radiation controlled areas receive appropriate radiation protection training. '. B. Appropriate access control techniques and protective clothing are used to limit external contamination. C. Respiratory protection equipment, is used where needed to 1'imit internal contamination. Radiation restricted areas are segregated and appropri-ately posted to limit radiation exposure. E. Instruments and equipment are properly calibrated so that accurate radiation, contamination, and airborne activity surveys can be performed. ~, 12.5-2 7-30-79 PVNGS FSAR lNCAFM<<n At RADIATION PROTECTION PROGRAM r"" personnel dosimetry devices are supplied.!.~iT S"~~~0~~ F. '~t'ppropriate G. An internal dose assessment program (whole body countKi'!: i. ing and/or bioassay) is conducted. Incoming and outgoing shipments of radioactive materials are properly handled. I. Necessary measures are performed to keep exposures ALARA. A more detailed discussion of the procedures used to implement this program is contained in section 12.5.3. The program .also ensures that appropriate effluent release samples are collected and analyzed consistent with the recom-mendations of Regulatory Guide 1.21 to verify that the station has an acceptable effect on the environment, and there-fore on offsite personnel. The radiation protection program will be periodically reviewed ~ ~ ~ ~ ~ ~ as discussed in section 12.1.1.2. ~ ~ ~ ~ ~ ~ 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES The radiation protection equipment, instrumentation, and facilities in each unit are identical and include an access control area, radiation protection office, a radiation protec-tion laboratory complex, first aid room, several personnel It decontamination facilities, locker rooms, protective clothing and respirators, air samplers, fixed and portable radiation detectors, and personnel dosimeters. Chemistry and radio-chemistry laboratories and a radiochemistry counting room are located at the access control point of each unit. 12.5.2.1 Radiation Protection E i ment Equipment for personnel protection and contamination control is ~ provided and described in the following sections. ~ ( 12.5-3 .8-9-79 PVNGS FSAR RADIATION PROTECTION PROGRAM 12.5.2.1.1 Respiratory Protection Equipment Various types of respiratory protection equipment are provided for use to protect against airborne radioactive contamination as prescribed in the Station Manual. Typical respiratory protection equipment includes: A. Pressure demand air line respirator: A pressure demand, full-facepiece air line mask which has U. S. Bureau of Mines (USBM) or National Institute of Occupational Safety and Health (NIOSH) approval. B. Air supplied hood with constant flow air supply which has NIOSH approval. C. Air mask: A pressure demand, full-facepiece self- , contained mask which has USBM approval or NIOSH approval. D. Ultra filter respirator: A full-facepiece filter mask which has USBM approval or NIOSH approval. 12.5.2.1.2 Protective Clothing Various types of protective clothing are stocked at the plant ~ ' to protect against contamination. Typical clothing includes: A. Body protection Gab coats ~ Coveralls 3~ Plastic suits B. Head protection Surgeons caps 2~ Cloth hoods 3 ~ Plastic hoods C. Hand protection 2. Disposable gloves Surgeons gloves ~, Amendment 5 12.5-4 August 1981 8-06-81 PVNGS FSAR )NCREMF.NTp,i RADIATION PROTECTION PROGRAM surveys for radiation or contamination control. Most of these instruments are stored so that they are easily accessible to ~ ~ ~ ~g-l.~0(q personnel in the units. pV<;" ",'" 'l0,; ) Thb criteria for selection of these instruments were to obtain accurate and reliable instrumentatzon that could be easily 0'. serviced and that would cover the entire spectrum of radiation measurements expected to be made at the station during normal operation, shutdowns, and accident conditions. These instruments are calibrated semi-annually when in use. The calibration is normally performed in the station calibration facility. Sufficient quantities of each type of instrument will be obtained to permit calibration and maintenance without diminish-ing the radiation protection capability. Quantities of the below listed types of instruments shall be maintained on hand to provide the listed minimum quantities for each unit, after allowing for instruments out of service (in need of calibra-tion or repairs). These quantities are sufficient to provide for maximum expected instrument usage and unusual occurrences: A. Six ion chamber type dose rate meters used to measure beta-gamma exposure dose rates ranging up to 250 R/h. B. Eight portable gamma dose rate meters used to measure gamma dose rates ranging up to 1000 mr/h. C Six very high range gamma dose rate meters with a range up to 1000 R/h. D. Six beta-gamma G-N survey meters. Two alpha survey meters with a range up to 500,000 cpm. Two neutron survey meters with readout, covering the range of 0 to 200 mrem/h. Two fast-slow neutron counters covering the range of 0 to 500,000 cpm. August 1981 12.5-7 Amendment 5 8-06-81 t w~),~ Jg) )gal ~ ~ I ' PVNGS FSAR ~ v "I, ~ RADIATION PROTECTION PROGRAM 12.5.2.2.3 Personnel Monitoring Instruments Personnel monitoring instrumentation is provided to determine external and internal contamination levels and radiation doses received by personnel. The criteria for selection of dose measuring devices were to have devices that could be quickly and accurately evaluated by station personnel (thermoluminescent dosimeters) and that could be easily read by the individual (self-reading dosimeters). The criteria'for selection of external contamination measuring equipment were to have devices available at checkpoints and other areas that could be used to determine the location of contamination (friskers) and at the normal exit from the con-trolled area that do not. require any action by personnel being checked (portal monitors). The principal criterion for selec-tion of the whole-body counting system was to have a system readily available to supply information concerning internal exposure levels. The friskers, portal monitors, and thermoluminescent dosimeter (TU)) readers are calibrated electronically and/or with a source at least semiannually. A calibration check on the friskers and portal monitors is performed monthly and on the TM readers daily. A calibration check of the self-reading dosimeters is performed at least semi-annually and complxes with the regulatory position of Regulatory Guide 8.4. The whole-body counting system is calibrated at least semiannually using a phantom containing various radionuclides. A check of the calibration of the whole-body system is performed by computer each time an analysis is performed. Quantities of each type of device will be obtained to permit . calibration and repair without diminishing the radiation pro-tection supplied. The devices and minimum numbers of each include: A. Count rate meters that are used as friskers to detect beta-gamma external contamination. They are normally 0, used with G-M detectors. (At least 30). Amendment 5 12.5-8 August 1981 08-06-81 PVNGS FSAR ED1T SECTlOf4 IN pUG 18 '1981 RADIATION PROTECTION PRIG IN OU.T. B. One portal monitor used to check for beta-gamma external contamination at the exit of the controlled area of each unit and two in the guard house. The monitor consists of an instrument console and a portal. The portal contain's several liquid scintillation detector channels to provide head-to-foot detection capability. A sensor activates the counting circuit when a person steps into the portal. Visual and audible alarms are provided. (A back up set is available to substitute for an inoperative portal monitor). C. Self-indicating dosimeters of various ranges. These dosimeters may include some electronic alarming and/or rate dependent types for use when deemed necessary by the Radiation Protection Supervisor. Dosimeters shall be tested and used in accordance with the recommenda-tions of the regulatory position of Regulatory Guide 8.4. (At least 300 dosimeters). D. One automatic and three manual TLD reading systems. These systems are used in the in-plant determination of whole-body exposure and job-related exposure. E. One whole-body counter, located onsite. The multichannel analyzer and computer are programmed to analyze the data and report the radionuclides detected together with the percent body (or thyroid) burden. F. Containers for collection of urine samples (normally used for tritium) and for fecal samples (possibly used under accident conditions) will be available. These samples are sent to a vendor for analysis. (at least 600 containers). 12.5.2.2.4 Area Radiation Monitoring (ARM) System The ARM system provides readout and alarms locally and in the control room as described in section 11.5. August 1981 12.5-9 Amendment 5 8-06-81 PVNGS FSAR i ~ t h~ Prr war ~ ~ RADIATION PROTECTION PROGRAN 12.5.2.2;5 Air Sampling and Monitoring Instrumentation Air sampling and monitoring instrumentation is used to deter- .-,.i e mine the levels of radioactivity in airborne effluents to comply with 10CFR50 and 10CFR20, and to determine the levels of airborne radioactivity in in-plant areas, where personnel are likely to be exposed, in compliance with 10CFR20. The criteria for selection of the various types of equipment were: A. To install fixed constant air monitors (CAMs) on the effluent paths with alarms in the control room so that automatic (or operator) action can be taken to correjct abnormal situations. ll B. To install fixed CANs, for in-plant determination, in areas where airborne activity was expected to occur or where it would need to be determined during an emergency. C. To use portable for" in-plant determinations, to ~ CAMs, monitor work areas where airborne activity levels could I be high. D. To use portable air samplers to determine airborne R activity at some jobsites during maintenance and normal operation. The monitors're calibration checked at least every refueling shutdown using check sources that are related to initial calibration. As long as the check source response does not change significantly, a complete calibration will not be performed. Air sample flowrates are also checked at least quarterly using a flowrate meter or manometer. The fixed air monitors are described with the radiation moni-toring system (RMS) in section 11.5. Typical portable air sampling and monitoring instrumentation is listed below: A. Portable having gamma scintillation detectors ~, CANs to measure gross gamma or iodine activity. These monitors have a strip chart recorder and visual and audible alarms. t Amendment 5 12.5-10 August 1981 08-06-81 PVNGS FSAR >l<GRN)CtGAL RADIATION PROTECTION PROGRAM ( Additional clothing issue and storage areas, personnel and L-,-,,i g(g)O.~ gK! equipment decontamination areas, and health physics equipment , > p,Ui-storage are located at the controlled. entrance area for the IH fuel building and containment. An emergency shower and eyewasn ~ ~ ~ are also located there. A single, central calibration facility is located near the con-densate storage tank outside and north of Unit 2. The facility serves the entire station and contains a neutron source and shielded calibration range. Smaller, semi-portable sources for RMS detector calibration are also stored here when not in use. TLD badge and other personnel dosimeters are also cali-brated in this facility. A radioactive laundry system is housed in the laundry and decontamination facility which is located near the radwaste building of Unit 1, and is a common facility for three units. The system uses four "RADKLEEN" dry cleaning machines, spe-cially designed by Health Physics Systems, Inc., for decon-tamination of cloth and rubber protective clothing, and utilizes Dupont's Valclene dry cleaning solvent. The system consists of cleaning chamber, solvent tank, still drying fan, evaporator refrigeration compressor, and several filters. The contains two separate filtration systems. One set of I'ystem micron filters, arranged in the parallel banks, filters solvent before it enters the cleaning drum. The other set of filters, arranged in series, comprises a recirculation filtration sys-tem which continually removes contamination from solvent in the contaminated sump, then returns the solvent to the clean sump o The system is a closed cycle "zero release" system. All con-tamination is retained within the machine until it is removed for proper waste disposal. Both cleaning and drying cycles are performed in the same machine which eliminates the pos-sibility of any accidental contamination during transfer between washers and dryers. Radioactive waste is 'trapped by August 1981 12.5-13 Amendment 5 06-29-81 PVNGS FSAR RADIATION PROTECTION PROGRAM special sized filters. Aqueous or hydrocarbon soluble con-tamination is separated during solvent, distillation and re-condensation. The contamination is transferred to the Unit 1 Liquid Radwaste System. Contaminated filters and residue are removed from the machine and drummed for proper disposal. The machine has a dry weight loading capacity of 30 pounds and'is capable of operating in automatic and manual modes through a complete cleaning and drying cycle and is designed to operate on a continuous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day duty cycle. 12.5.3 PROCEDURES Radiation protection procedures are established to keep person-nel radiation exposures ALARA and within the limits of 10CFR20. These procedures are discussed in section 13.5.2. Policy and operational considerations for maintaining personnel radia-tion exposures are discussed in sections 12.1.1 and 12.1.3. 12.5.3.1 Radiation and Contamination Surve s ~ ) Radiation protection personnel. normally perform routine radia-tion and contamination surveys of accessible areas of the units. These surveys consist of radiation dose rate measure-ments and/or contamination smears as appropriate for the speci-fic area. Air samples are routinely taken in accessible portions of controlled areas. Surveys related to specific activities may be performed if necessary prior to, during, or after activities that would be expected to produce additional significant radiation exposure to individuals. Survey pro-cedures and routine survey schedules are provided in the Station Manual. ~ ) Amendment 5 12.5-13A August 1981 06-29-81 PVNGS FSAR INCREMENTAL RADIATION PROTECTION PROGRAM E;::: r sECT}ot.'A<vVI This page intentionally blank. August 1981 12.5-13B Amendment 5 06-29-81 PVNGS FSAR RADIATION PROTECTION PROGRAM 12.5.3.2 Procedures and Methods to Maintain Ex osures ALARA Operating, maintenance, and radiation protection procedures are reviewed, as discussed in section 12.1.1, to identify situations in which potential exposures could be reduced. Such ALARA considerations include: A. Restricted Areas Restricted Areas as defined in 10CFR20.3 (14) are established at the protected area fence and access is controlled at that point for the purpose of protecting individuals from exposure to radiation and radioactive materials. B. Controlled Areas Procedures establish permanent and temporary controlled areas within the restricted areas where access is further administratively controlled for the purpose of limiting exposure of individuals to radiation and radioactive materials. Within controlled areas, radia-tion and high radiation areas are identified and segre-gated in accordance with 10CFR20.202 and 10CFR20.203, utilizing access control and posting of radiation and high radiation areas. High radiation areas having dose rates greater than 100 mrem/h, but less than 1000 mrem/h will be barricaded with a rope, gate, or other suitable unlocked barrier. Access to high radiation areas having dose rates greater than 1000 mrem/h will meet the requirements of 10CFR20.203(c). C. Radiation Exposure Permits Procedures require Radiation Exposure Permits for entry into areas where the dose rate exceeds 15 mrem/h, or = which contain removable contamination levels in excess 0'2.5-14 8-9-7S 'PVNGS FSAR ~HCBFViiwt Al RADIATION PROTECTION PROGRAM is responsible for monitoring himself and his clothing when he crosses a local control point or the main access control porn/ p c t=CT)QN as discussed in section 12.5.2. If contamination above allow-,,- ", .~ able limits is found, the individual is decontaminated using OU f~ facilities previously described in section 12.5.2. 12.5.3.5 Airborne Activit Ex osure Control Shen airborne radioactivity is detected in excess of the limits of 10CFR20.203(d), the area is posted as an airborne radio-activity area, and access is controlled in accordance with section 12.5.3.2. Occupancy is restricted or respiratory protection equipment is provided to maintain exposures within the limits of Appendix B, Table 1, Column 1 of 10CFR20, if. personnel entry'is required into areas where the source of airborne radioactivity cannot be removed or controlled. An air sampling program is used to ensure that appropriate respiratory protective equipment is specified on the Radiation Exposure Permit. The respiratory protection program is organized to conform to the applicable portions of ANSI Z88.2-1969. Effectiveness of the respiratory protection program is evaluated by various types of bioassay analyses or nasal smears, or respirator facepiece interior smears. Respiratory equipment discussed in section 12.5.2 is available near the main access control point to permanent controlled areas. Supplementary emergency respiratory equipment is available in the control room and in emergency kits. The following controls are incorporated in the program: A. Each respirator user is advised that he may leave an airborne'radioactivity area for psychological or physical relief from respirator use. Each user shall 12.5-17 8-9-79 PVNGS FSAR RADIATION PROTECTION PROGRAM leave the area or any in the case of respirator malfunction other condition that might cause reduction in ! the protection afforded the user. B. Air samples and surveys are made to identify the presence of airborne radioactivity and to estimate individual exposures so-that selection of appropriate respiratory equipment can be made. C. Procedures are established to ensure correct fitting, use, maintenance, and cleaning of respirator equipment. Each individual qualified to use respiratory protection equipment receives a quantitative fit test annually and performs a qualitative fit test prior to use of respiratory protection equipment. 12.5.3.6 Personnel Monitorin Station employees, contractor personnel, support personnel, and visitors are required to wear thermoluminescent dosimeters (TLD) or self-reading dosimeters when in a controlled area. In addition, job dosimeters are issued to individuals working ~ ? under a Radiation Exposure Perm'.t. The exposure readings of these job dosimeters are used for specific ALARA job exposure evaluation as well as to indicate current individual exposure status. Use of neutron dosimeters complies with the recommenda-tions of Regulatory Guide 8.14. The Station Manual requires bioassays, including whole body counting, consistent with the recommendations of Regulatory Guides 8.9 and 8.26. The type of determination and the frequency of determination depends upon the work environment of the individual and the work situation. Job dosimeters are used for specific ALARA job exposure evalua-tion as well as to indicate current individual exposure status. Personal dosimeters are evaluated on a monthly basis and are ~; Amendment 5 12.5-18 August 1981 8-06-81 PVNGS FSAR INCREMENTAL EDlT SECTlGN AUG 13 1981 TABLE OF CONTENTS lg OUT ~Pa e Question 12A.1 (NRC Question 460.17) (12.2) 12A-1 l4 Question 12A.2 (NRC Question 471.1) (12.1) 12A-1 Question 12A.3 (NRC Question 471.2) (12.5) 12A-2 Question 12A.4 (NRC Question 471.3) (12.3) 12A-2 Question 12A.5 (NRC Question 471.4) (12.1) 12A-3 Question 12A.6 {NRC Question 471.5) (12.1) 12A-3 Question 12A.7 {NRC Question 471.6) (12.2) 12A-4 Question 12A.8 (NRC Question 471.7) (12.3) 12A-5 Question 12A.9 (NRC Question 471.8) (12.3) 12A-5 Question 12A..10 (NRC Question 471.9) (12.3) 12A-5 Question 12A.11 (NRC Question 471.10) (12.4) 12A>>6 Question 12A.12 (NRC Question 471.11) (12.5) 12A-6 Question 12A.13 (NRC Question 471.12) (12.5) 12A-7 Question 12A.14 (NRC Question 471.13) (12.5) 12A-7 Question 12A.15 (NRC Question 471.14) (12.5) 12A-8 Question 12A.16 {NRC Question 471.15) (1.8) 12A-8 August 1981 12A-i Amendment 5 08-06-81 ~ ~ I PVNGS FSAR EOlT SEGTlON 6 i "" ( 2 * . ( Q t'2.17) QUTi APPENDIX 12A (12.2) Provide radionuclide inventories of the refueling water tank and the reactor makeup water tank referred to in FSAR Sec-2 tion 12.2.1.7 (Section 12.2.1.7 states that. these tank inven-tories are described in Section 12.2.1.1.5.1; however, these are not described in Section 12.2.1.1.5.1). RESPONSE: The response is given in amended section 12.2.1.7. 2
- 12 .2 (2 Q t.'71.1) (12.1)
Section 12.1.1.2, paragraph two states, in part, that: The Radiological and Chemistry Supervisor has the responsibilities of the onsite Radiation Protection Manager, described in Regu-latory Guide 8.8 and Section 13.1." Based on information contained in the draft document "Criteria for Utility Management and Technical Competence" it is our ( position that the Radiation Protection Group be a separate organization from the Chemistry Group. Additionally, in accordance with Regulatory Guide 8.8, it is our position that the Radiation Protection Manager (RPM equal to the Radiation Protection Supervisor) should have access to the Assistant Plant Manager in radiation protection matters. In matters relating to radiological health and safety, the RPM has direct responsibility to both employees and management that can best be fulfilled if he is independent of station divisions, such as operations, maintenance or technical support, whose prime responsibility is continuity or improvement of station operability. Revise Subsections 12.1.1.2, 12.5.1.1, 13.1.2.2.2.2, 13.1.3.1, Figure 13.1-6, and Figure 13.1-7 and other FSAR Sections where appropriate to reflect compliance with Regulatory Guide 8.8, ~ ~ ~ ~ C.1.b. (3) and Criteria for Utility Management and Technical 4 ~ ~ ~ ~ ~ ( Competence. August 1981 12A-1 Amendment 5 08-17-Sa PVNGS PSAR Ã99J3Q54> APPENDIX 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, ( .0) 13.1.2.2.2.2, and 13.1.3.1. The Radiation Protection Section is a separate organization from the Chemistry Section, and the Radiation Protection Supervisor reports to the Engineering and Technical Services Manager who is independent of the Station Operations and Maintenance Depar'tments. The Radiation Protection Super-visor has direct access to the Manager of Nuclear Operations in matters relating to radiological protection and ALARA programs as authorized in the Station Manual. He is a permanent, member of the Plant Review Board. You should describe your plan,to provide backup coverage in the event of absence of the RPN, and you should outline the qualifications of the individual who will act as the backup. The December 1979 revision of ANSI 3.1 specified that the temporary replacement for an RPM should have a BS degree or engineering, 2 years experience in radiation protec-in'cience tion, 1 year'f which should be nuclear power plant experience, 6 months of which should be onsite. It is our position that, this experienCe be professional experience. RESPONSE: The response is given in amended sec-tion 13.1.2.2.2.2. The minimum requirements for the position providing backup coverage in event of absence of the RPM is discussed in section 13.1.3.1. (12.3) You should provide information in response to TMI Lessons Learned review for the following NUREG-0737 areas: II.B.2 Post Accident Shielding and Vital Area Access; II.B.3 ALARA for ~, Amendment 5 12A-2 August 1981 08-17-81 7)0~ PVNGS FSAR pp)T Spc I( tZCREMENTAL gGl GyG $ 8$ APPENDIX 12A N l.N Post-Acci'dent ~ Sampling; II.F.1 High Range In-Containment Radia- ~ ~ tion Monitors; and III.D.3.3 Post,-Accident Iodine Sampling and Analysis. RESPONSE: The response is provided in the PVNGS LLIR. Additional information is provided in sections 12.1.2.4 and 12.3.1.3 for Items II.B.2 and II.B.3, in section 11.5 and figure 12.3-4 for II.F.1, and in sections 9.3.2.2.2 and 11.5 for III.D.3.3. I" *' " ( Q"" "" (12.1) Paragraph B, of Section 12.$ .2.1.2, stating that "as minimum, shielding is designed to reduce gamma dose rates from sources external to a radioactive compartment to levels comparable to dose rates resulting from equipment within that compartment," is not clear. It appears that this could refer to shielding between two adjacent compartments in which radioactive equipment is located. If this is the case, then shielding should be designed to reduce radiation from the operating equipment in one compartment to levels below that which is expected in the adjacent compartment'rom the shutdown equipmenti to be main-tained or repaired. Please clarify. RESPONSE: Paragraph B of section 12.1.2.1.2 has been revised to provide the requested clarification. I * .6( Q 6' <12.1) In accordance with Section C.2.e, "Crud Control," of Regula-tory Guide 8.8, it is our position that consideration should be given to the selection of corrosion resistant, low cobalt content. alloys to reduce the concentrations of radioactive corrosion product buildup in systems. Section 12.1.2.3, "Equipment General Design Considerations for ALARA," of your FSAR, should be revised to reflect your design considerations for selection of low cobalt alloys. August 1981 12A-3 Amendment 5 08-18-81 PVNGS FSAR !~ <i ' 12A ~'j APPENDIX RESPONSE: The response is provided in the revised sec-tion 12.1.2.3. lt * " ". ( Q"" "" (12.2) Very high radiation levels can occur in the vicinity of spent fuel transfer tubes; therefore, all accessible portions of the transfer tubes must be shielded during fuel transfer. Please address the manner in which shielding, access control and radiation monitoring will be incorporated into the radiation protection program to prevent workers from receiving very high exposures during transfer of spent fuel from the reactor to the spent fuel pool through Me fuel transfer tubes. Provide appropriate figures (e.g., plan and elevation) that shows the shielding arrays for all direct, gamma radiation and steaming pathways from the spent fuel during the transfer. On the same figure show the location of any administrative controls by barriers, signs, audible and visual alarms, locked doors, etc. Use of removeable shielding for this purpose is accept-able. Shielding shall be such that the resultant, contact radiation levels shall be no greater than 100 rads per hour. ~ In addition, normal high radiation area access controls must be implemented for access to areas below 100 rads/hr. All accessible portions of the transfer tubes shall be clearly marked with a sign stating that potentially lethal fields are possible during fuel transfer. If removable shielding is used for the fuel transfer tubes, it must also be explicitly marked ', as above. In other than permanent shielding is used, local audible and visible alarming radiation monitors must be installed to alert personnel if temporary shielding is removed during the transfer operations. Your drawings 12.3-6, 12.3-2, and 9.1-5 do not show sufficient shielding details. Sec-tion 12.2.1.1.4 of your FSAR should be revised accordingly. Amendment 5 12A-4 August 1983. 08-17-81 PVNGS FSAR EO)T $ ECT[Og R'CAB/i~~"'UG18 4.i APPENDIX 12A .OUT, RESPONSE: The response is provided in the 'revised sec-tions 12.3.2.2.2 and 12.3.2.2.4. Figures 12.3-22 and 12.3-23 have been added to show shielding details. I!" *' t"" Q "t' *t (12.3) In Section 12.3.2.2.2, "Containment Interior Shielding Design," you should describe shielding used to minimize neutron steam-ing from the annulus between the reactor pressure vessel and the vessel shield, and provide an estimate of neutron dose rate in containment accessible areas. RESPONSE: The response is provided in the revised sec-tion 12.3.2.2.2. lt" * " 9 t" Q""t" * (12.3) In accordance with Regulatory Guide 1.70, "Standard Format and Content, of Safety Analysis Reports for Nuclear Power Plants!', Revision 3, Section 12.3.1, it is our position that FSAR should include plant layouts showing shield wall thickness. If the addition of shield thicknesses on the reduced size drawings would cause the thicknesses to be unreadable, the shield thicknesses of major radioactive equipment should be provided in a separate table. Section 12.3.2 of your FSAR should be revised as. above, or plant general arrangement drawings identifying shield thick-nesses should be provided. RESPONSE: Scaled drawings depicting shield thicknesses have been transmitted to the NRC under separate cover. UESTION 12A.10 (NRC Question 471.9) (12.3) In Section 12.3.2.2.2 (on page 12.3-16 of your FSAR), the sec-ond sentence of the second paragraph states that "Water August 1981 .12A-5 Amendment 5 08-17-81 PVNGS FSAR APPENDIX 12A provides- the-shielding over spent fuel assemblies during fuel handling." This Section should be revised by specifying the minimum water shield above the spent fuel during fuel handling. C RESPONSE: The response is provided in figure 12.3-24. (12.4) Section 12.4 and Table 12.4-5 should be expanded to include the expected average personnel exposures from special mainte-nance and inservice inspection, in accordance with Regulatory Guide 1.70, Revision 3, and Regulatory Guide 8.19, "Occupa-tional Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage Man-Rem Estimates." Section 12.4 should also include a description of any changes made during planning or design review for the purpose or reduc-ing,the projected dose rates. (Actual exposure data from similar operating plants, operated in similar manner may be used for the dose assessment for unpredictable activities, but should be corrected for improvements in plant design and operating procedures.) [See NUREG-0713, "Occupational Radia-tion Exposure at Commercial Nuclear Power Reactors 1979, Annual Report."] Section 12.4 of your FSAR should be revised to provide the above information. RESPONSE: The response is provided in the revised sec-tion 12.4'. (12.5) In Section 12.5.2.1.1, paragraph B, states that typical respiratory protection equipment includes: B. "Two-piece plastic suits with constant flow air supply." According to our information, there are no NIOSH approved air supplied suits.'ince 10 CFR 20.103 requires that all respiratory 0; Amendment 5 12A<<6 August 1981 08-17-81 0'%evmrN PVNGS FSAR APPENDIX 12A protective devices used for radiation protection purposes be NIOSH approved, these devices should be deleted from your FSAR. EolT. S .Tlof. RESPONSE: The response is given in amended sec- 'UG1 1981 tion 12.5.2.1.1. A number of vendors supply air supplied that are including Defense Apparal 0U.Tj hoods NIOSH approved, whose air supplied hood has NIOSH approval TC-19C>>120. Q" *' " ("" Q" ' "* (12.5) Section 12.5.2.2.2 describes the types of portable radiation detection instruments and the minimum quantities which will be available for each unit. It is our position that the number of portable radiation survey instruments (especially those which are most frequently used by radiation protection personnel, 0-5000 mR/hr), be increased to reflect following considerations: (a) three units can be ' shut down for repairs at the same time (maximum usage of survey instruments), (b) a number of instruments out of service (in need of calibration or repairs), and (c) a number of spare, operational instruments should be always available for use in unusual occurrences. Based on your analysis, you should, revise the list of instruments or justify the proposed number of portable radiation detection instruments. RESPONSE: The response is given in amended sec-tion 12.5.2.2.2. Q" *' ".(* ("" Q""Q (12. 5) In Section 12.5.2.2e3 describing various types of Personnel Monitoring Instruments (such as count rate meters, portal monitors, self-reading dosimeters, automatic and/or manual TM reading systems, etc.), the quantities of these instruments should be specified, and the words "Sufficient quantities" should be removed. August,1981 12A-7 nmendment 5 . I~ 08-17-81 PVNGS FSAR APPENDIX 12A RESPONSE: The response tion 12.5.2.2.3. is given in amended sec-(12.5) In this Section 12.5.2.3, Facilities Related to Radiation Protection, you should describe your provisions for onsite laundry of protective clothing. RESPONSE: The response is provided in the revised sec-tion 12.5.2.3. Q * . ( Q (1.8) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," specifies in Sec-tion 1.8, Conformance to NRC Regulatory Guides, that: "The SAR should include a table indicating the extent to which the applicant intends to comply with all applicable NRC regulatory guides and the revision number of those guides. For each applicable regulatory guide, the table should identify those sections of the SAR to which the guide applies and should indicate any proposed exceptions to the regulatory position." Because this Regulatory Guide 1.70 has been accepted in your FSAR, Section 1.8, it is our position that the following Regu-latory Guides should be also included in your FSAR, Sec-tion 1.8 or you should describe equivalent alternatives: R.G. 8.4, "Direct Reading and Indirect Reading Pocket Dosimeters," R.G. 8.14, "Personnel Neutron Dosimeters," R-G. 8.19, "Occupational Radiation Dose Assessment in Light Water Reactor Power Plants Design Stage Man-Rem Estimates," 0 Amendment, 5 12A-8 August 1981 08-17<<81 PVNGS FSAR APPENDIX 12A R.G. 8.21, "Measuring, Evaluating and Reporting Radioactivity ~MT -~ in Solid Waste and Releases of Radioactive 5laterials in .Liquid and Gaseous Effluents from Light-Water- AUG13 i981 Cooled Nuclear Power Plants," and gl t~i ~ ~r) 4 R.G. 8.26, 'pplications of Bioassay for Fission and Activa-tion Products. RESPONSE: The response is provided in the revised sec-tion 1.8. {The request to address Regulatory Guide 8.21 is assumed to be for Regulatory Guide 1.21's the title given is that of Regulatory Guide 1.21. Regulatory Guide 8.21 applies only to processing and manufacturing plants.) August 1981 12A-9 Amendment S 08-06-81 I'l I I l) leCREMENTAL PVNGS FSAR ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1.2.2.2.2 Radiation Protection Su ervisor. The Radiation Protection Supervisor is responsible to the Engineering and 5 Technical Services Manager for the preparation, coordination, and conduct of the station chemistry and radiological programs, including delineating the operating philosophy and procedures EoiT<<>ION for maintaining occupational radiation expos'ures as low as is reasonably achievable. His position corresponds to "Radiation Protection Manager" as discussed in Regulatory Guide 1.8. OUT Reporting to the Radiation Protection Supervisor are the Supervising Radiation Physicists for each unit. The Radiation Protection Supervisor is responsible for control of radiation exposures to personnel, maintenance of related records, conduct of surveillance, and approval of radioactive waste disposal activities. A Supervising Radiation Physicist, is designated as backup to provide coverage in event of absence of the Radiation Protection Supervisor.. The Supervising Radiation Physicist shall have a minimum of five years experience, of which two years shall be at a professional level. 13.1.2.2.2.3 0 erations En ineerin Su ervisor. The Opera-tions Engineering Supervisor is responsible to the Engineering and Technical Services Manager for mechanical and electrical engineering support, including monitoring station performance and the inservice inspection program. 13.1.2.2.2.4 Com uter Su ervisor. The Computer Supervisor is responsible to the Engineering and Technical Services Manager for coordinating station computer activities, including hard-ware and software; 13.1.2.2.2.5 Chemistr Su ervisor. The Chemistry Supervisor is responsible to the Engineering and Technical Services Manager for the conduct of the water chemistry program and coordinates with the Radiation Protection Supervisor on radiation exposures and contamination problems associated with the chemistry program. August 1981 13-1-17 Amendment 5 08-'05-'81 ~ l',j't'0 +,q$ Vi ~'", PVNGS FSAR ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1.2.2.2.6 Licensin Su ervisor. The Licensing Supervisor is responsible to the Engineering and Technical Services Manager providing licensing support for station activities. 'or 13.1.2.2.3 Operations Superintendent The Operations Sup'erintendent is responsible to the Plant Manager of Nuclear Operations for the safe, reliable, and efficient operation of power block equipment and systems. He directs the Operating Supervisor of each unit. 13.1.2.2.3.1 Unit 0 eratin Su ervisor. The Unit Operating Supervisor is responsible to the Operations Superintendent for the conduct of the unit operations in a safe and efficient manner in accordance with technical specifications and station lt instructions. He supervises the activities of the unit s operating personnel. The Unit Operating Supervisor will pos-sess a Senior Reactor Operator license. 13.1.2.2.3.2 Shift Su ervisor. The Shift Supervisor is responsible to the Unit Operating Supervisor for the safe, reliable, and efficient operation of the unit during his assigned shift. He directs the activities of the operators on his shift, coordinates maintenance activities performed while he is on duty, and ensures compliance with required radio-logical control procedures. Should the Shift Supervisor be absent or incapacitated, the Shift Foreman will assume his responsibilities. The Shift Supervisor will possess a Senior Reactor Operator License. 13.1.2.2.3.3 Shift Foreman. The Shift Foreman is a backup to the Shift Supervisor and supervises shift personnel in conduct of operations as assigned. He will possess a Reactor Operator License or a Senior Reactor Operator License. Amendment 5 13-1-18 August 1981 08-05-81 PVNGS FSAR )NERF.M.FNYI L ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1.2.2.3.4 Control 0 erator. The Control Operator maniphojTSECTIG!l lates the reactor plant controls. He will possess a Rea'ctogU~ ~ 3 ~9@ Operator License. IW nur 13.1.2.2.3.5 Auxiliar 0 erator. The Auxiliary Operator is responsible, under the direction of the Shift Supervisor and Shift Foreman, for operating auxiliary systems and assisting in the refueling of the plant as directed. 13.1.2.2.4 Maintenance Superintendent The Maintenance Superintendent is responsible to the Plant Manager for the performance of preventive maintenance and repairs on plant systems and equipment. Reporting to the Maintenance Superintendent are'the Mechanical Supervisor, Electrical Supervisor, Instrumentation and Control Supervisor, and the Station Services Supervisor. The Mechan-ical Supervisor and the Electrical Supervisor are responsible ' for mechanical and electrical maintenance, respectively. Instrumentation and Control Supervisor is responsible for The calibration and maintenance of instruments and controls. The Station Services Supervisor is responsible for providing sup-port services such as crane operations, vehicle repairs, carpentry, and painting. 13.1.2.2.5 Scheduling and Licensing Supervisor The Scheduling and Licensing Supervisor i's responsible to the Plant Manager for providing scheduling and licensing support for station activities. 13.1.2.2.6 Quality Supervisor The Quality Supervisor is responsible to the Plant Manager. He is responsible for the onsite Operating Quality Assurance Program and reviews, audits, and observes station operations (0 to ensure compliance with licensing requirements. August 1983. 13.1-19 Amendment 5 08-06-81 PVNGS FSAR ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1.2.2.7 Support Servi:ces Manager The Support Services Manager is responsible to the Plant Man-ager for maintaining the station central personnel file, technical file, and record file. He directs the activities of the Security Director, Office Supervisor, Materials Super-visor, Training Director, and Safety Director. The Training Director is responsible to the Support Services Manager for the preparation, coordination, and conduct of the station training program. He directs the activities of the Simulator Supervisor and the nuclear plant instructors. 13.1.2.3 0 eratin Shift Crews An operating crew for each unit will normally consist of a Shift Supervisor (who will possess a Senior Reactor Operator license), a Shift Foreman, and a Control Operator (both of whom will possess Reactor Operator licenses), and two Auxiliary Operators. The minimum shift crew composition for various modes of operation is shown in table 13.1-2. During refueling operations, when the reactor core configura-tion is being altered, a Senior Reactor Operator or a Senior 0't Reactor Operator Limited to Fuel Handling will directly supervise the fuel handling activities and will have no other concurrent duties. least one qualified Radiation Protection Technician shall be assigned to each shift in each unit, with fuel loaded in the reactor. During refueling operations or when numerous radiation protec-tion activities are in progress, additional Radiation Protection Technicians will be assigned and supervision provided, as needed. ~, r Amendment 5 13.1-20 August 1981 08-06-81 INCREIMENTAl. PVNGS FSAR ORGANIZATIONAL STRUCTURE, OF APPLICANT Table 13.1-2 E0J7 S".Cr,'ox MINIMUM SHIFT CREW COMPOSITION AUG 1 8 ]g8i (FOR EACH UNIT) ji4 0uy Applicable Modes License Category 1, 2, 3 a 4 5 a 6 1 (c) SOP (SRO) OP (RO) 1 Non-Licensed ao Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unex-pected absence of on duty shift crew members provided immediate action is taken to restore the shift crew. composition to within the mini-mum requirements of this table. b Operational modes are as defined in the Techni-cal Specifications. C~ Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising Core Alterations. 13 ' ~ 3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL 13.1.3.1 Qualification Re uirements The recommendations of Regulatory Guide 1.8, Personnel Selection and Training, are used as the basis for establishing minimum qualifications for nuclear power plant personnel. 13.1-21 9-11-79 PVNGS FSAR ORGANI ZATIONAL STRUCTURE OF APPLICANT The minimum requirements for station personnel described in section 13.1.2 are keyed to ANSI N18.1-1971 as follows: ANSI N18.1-1971 Position Position h Manager of Nuclear Plant Manager (4.2.1) Operations Engineering and Tech- Technical Manager (4.2.4) nical Services Manager Nuclear Supervisor Reactor Engineering 6 Physics (4.4.1) Licensing Supervisor Supervisor not requiring AEC Iicense (4.3.2) Operations Engineering Supervisor not requiring AEC License Supervisor (4.3.2) Chemistry Supervisor Radiochemistry (4.4.3) Radiation Protection Radiation Protection Manager (Note a) Supervisor Supervising Radiation Radiation Protection (4.4.4) (Note d) Physicist Maintenance Maintenance Manager (4.2.3) Superintendent Maintenance Control Supervisor not requiring AEC License Center Supervisor (4.3.2) Instrumentation and Instrumentation and Control (4.4.2) Control Supervisor Plant Instrumentation Repairman (4.5. 3 ) and Control Technician Station Services Supervisor not requiring AEC Iicense Supervisor (4.3.2) Mechanical Supervisor Supervisor not requiring AEC License (4.3.2) Plant Mechanic Repairman (4.5.3) Electrical Supervisor Supervisor not requiring AEC License (4.3.2) Amendment 5 13.1-22 August 1981 O8-O6-81 PVHGS FSAR ll~CaLMFXTAL ORGANIZATIONAL STRUCTURE OF APPLICANT ANSI N18.1-1971 Position xiii i girl (O>> Position Para ra h No. AUG y 1.) j Plant Electrician Repairman (4.5.3) h4 Qlfl Administrative ,Supervisor not requiring AEC License Services Manager (4.3.2) Operations Operations Manager (4.2.2) No License Superintendent Operating Supervisor Supervisor Requiring AEC License (4.3.1) (Note c) Shift Supervisor Supervisor Requiring AEC License (4.3.1) Assistant Shift Supervisor Requiring AEC License Supervisor Nuclear Operator III Operators (Licensed) (4.5.1) Nuclear Operator I Operator (Not Licensed) (4.5.1) 6c II Operations Quality (Note b) Assurance Manager ( Training Manager ~ ~ Supervisor not requiring AEC License (4.3.2) Security Manager Supervisor not requiring AEC License (4.3.2) NOTES: a~ The Radiation Protection Supervisor will meet the recom-mendations of the regulatory position of Regulatory Guide 1.8 for the Radiation Protection Manager.
- b. The Operations Quality Assurance Manager shall have at l4 least 6 years experience in the field of quality assur-ance. At least 6 months of the 6 years shall be in nuclear quality assurance. ' minimum of 2 years of the 6 years shall be related technical or academic training.
A maximum of 4 years of this 6 years may be fulfilled , by related technical or academic training. August 1981 13.1-23 Amendment 5 08-06-81 PVNGS FSAR ORGANIZATIONAL STRUCTURE OF APPLICANT
- c. 'he Operating Supervisor shall have a minimum of 6 years experience, of which a minimum of 2 years shall be nuclear power experience.
13.1.3.2 ualifications of Plant Personnel Resumes of the initial appointees to key plant managerial and supervisory positions through the Shift Supervisor level are included in appendix 13B. Amendment 5 13.1-24 August 1981 08-06-81 ~ PVNGS PLANT EDJT SEC ',MANAGER ttUG 1 9 2/2/2 'SCHEDULEItS TECHNICAL i SCHEDULING 5 UCENSING ASSISTANT PLANT 4UALfTY QUALITY EN 6 INEE IIS SUPE ltVISOR ~NAGER SUPERVISOR PERSONNEL ENGlkEEIIING C SUPl'0RT OPERATIONS MAINTENANCE TECHNICALSERV. SERVICES SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT MANAGER 1 1 I1/1 3 - 1/ OPE RATIONS INSTRUMENTATION fIUCLEAR TRAINING SECURITY OPERATIN G ENGINEERING 4 CONTROL MECHANICAL SUPERVISOR DIRECTOR. DIRECTOR SUPER/ISOR SUPERVISOR SUPERVISOR SUPERVISOR S 5/5/5 $ IUCLEAR MECHANICALE RADIATION IkSTR UMENTATION MECHANICAL SECURITY INSTRUCTORS ' E LECTR ICAL SHIFT ENGINEERING PROTECTION 4 CONTROL MAINTENANCE ENGINEERING FORCE SUPERVISOR .PERSONNE L SUPERVISOR PERSONNEL PERSONNEL PERSONNEL 4/-I- 5/5/5 ftAD IATION SUPPORT STATION PR OT ECTION COMPUTER SIMULATOR SHIFT ELECTRICAL SERVICES SERVICES PERSONNEL SUPERVISOR PERSONNEL FDREMAN SUPERVISOR STAFF SUPERVISOR 5/5/5 COMPUTER STATION ELECTRICAL CHEMISTRY CONTROL PERSOkkEL SERVICES MAINTENANCE SUPERVISOR OPERATOR PERSONNE L PERSONNEL R XEGEIID: 4/4I4 12 UNIT 1 MlkIMUM PERSOkNEL MAYBE 10 /10 /10 UklT2 NUMBER OF TRAkSFERRED FROM CHEMISTRY UklT 3 PERSONNEI. Ik UNIT TO UNIT DEPENDING AUXILIARY PERSONNEL CLASSIFICATION ON SITE WORK OPERATOR PER UNIT CONSIDERATIONS Palo Verde Nuclear (~crating Station X- TOTAL NUMBER OF PERSONNEL FSAR IN CLASSIFICATIOk CLASSIF ICATIOII PALO VERDE NUCLEAR GENERATING STATION KE TITLE OPERATING ORGANIZATION CHART LICENSE REQUIREMENT C- SENIOR REACTOR OPERATOR LICENSE Figure 13.1-6 R- REACTOR OPERATOR LICENSE August 1981 Amendmeiit 9 > 8-18-81 l t .'.S( ) A j. V JNCREigENTpg ~I BENI AND IUAIITKII 5f!SS- EDIT SECTI PIIESENT NUQSKII MINIMUMNUMBER OF BPLOYKES z 4 1 Z OF EMPLOYEES AUQ f 8 ~88~ ~ LANTMANAGKI5 ASSISTANT PLANT MANAGER 1 IM~Uy ENGINEEIIJNG AND TECHNICALSERVICES SUPKRINTENDKNT 1 NUCLEAR SUPERVISOR 1 1 NUCLEAR ENGINEERING tERSDNNEL 4I QR RZ RADIATIONPROTECTION SUPERVISOII 1. ,1 IIADIATIONPROTECTION PERSONNEL CHEMISTRY SUPERVISOR 1 CHEMISTRY PERSONNEL 12' OPERATIONS ENGINEERING SUPERVISOR OtERATIDNS ENGINEERING tERSONNEL 2 7; COMPUTER SUPERVISOR 1 COMPUTER PEINONNEL 3 I7i OtERATIONS SUPERINTENDENT 1 OPERATING SUPERVISOR 3I SHIFT SUPERVISOR 15 SHIFT FOREMAN CONTROL OPERATOR 5% AUXILIARYOPERATOR 10 10 -30 i MAINTENANCESUPERINTENDENT 1 MECHANICALSUPERVISDR 1 MECHANICALMAINTENANCEtERSONNEL 18 54 I ELECTRICAL SUPERVISOR 1 KLECTRICALMAINTENANCEPERSDNNEL 58! INSTRUMENTATION AND CONTROL SUPERVISOR 1 INSTRUMENTATIONAND CONTROL tERSONNEL 3 27'I STATION SERVICES SUPERVISOR SCHEDULING AND LICENSING SUPERVISOR SCHEDULERS AND TECHNICAL ENGINEERS 2 3= DUALITYSUPERVISOR 1 I QUALITYtERSONNEL 1 ~6' SUPPORT SERVICES MANAGER 1 TRAINING DIRECTOR 1 1 INSTR U CTORS 2 ~4! SIMULATOR tERSONNEL 41 !4! SECURITY DIRECTOR 1 1 UNIT 2 UNIT3 FUEL LOAD Fab Verde Nuclear Generabng Station FSAR. Figure 13.1-7 August l98l ,'mendment 5 ',. "~ j ~ P. % ~~P 1 4 ~L~ e., EDIT 8ECTIOII 1 NCREM ENTAt PVNGS FSAR f,UG )) )98 IN OUT APPENDIX 13A TABLE OF CONTENTS Question 13A-1 Question 13A-2 (NRC (NRC I Question 471.16)'(13.1) Question 471.17) (13.3) ~ ~Pa 13A-1 13A-2 e ( ~ August 1981 13A-i Amendment 5 )6 07-01-81 P \~ I' fNCREMF.N'P t VN S F gUG~~~9~~ APPENDIX 13A ( Section 13.1.2.3, Operating Shift Crews, second paragraph, (13.1) states that: "At least one member of each shift operating crew will be trained in the station radiation protection procedures and ...." In addition, Figure 13.1-6 shows that six radiation protection personnel (technicians) will be assigned to each unit. NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants" requires that a radiation protection technician, whose qualifications are described in ANSI 18.1, shall be onsite at all times. Section 13.1.2.3, as written, would allow a designated member of the shift crew (e.g., reactor operator) to act as a health physics technician if he is qualified to implement radiation protection proce-dures. It should be noted that this qualification is no longer acceptable to the staff after the reactor is at power. Only an assigned health physics technician will be acceptable. In addition to the minimum of one HP technician per shift, in choosing a sufficient number of personnel, the'ollowing should be considered: (a) one radiation protection technician will be assigned to each shift; (b) one (probably the Radia-tion Protection Foreman), could be in charge, during the absence of the Radiation Protection Supervisor; (c) three units could be shut down for repairs simultaneously (resulting in numerous radiation protection activities); (d) training and retraining; and (e) routine absences. Therefore, Section 13.1.2.3 and Figure 13.1-6 should be revised accordingly.
RESPONSE
The response is given in amended section 13.1.2.3 and figure 13.1-6.
August 1981 13A-1 Amendment 5 8-06-81
)) * . ) Q t'7) ~ 17) (13.B)
The Station Personnel Resume, Appendix 13.B, for the Radiation Protection Supervisor (Mr. J.W. McDuffee) does not indicate whether a BS degree was received from the Phoenix College (1960-61). Also, the duration of his professional level experience has not been stated in all his activities. It is our position that the Radiation Protection Supervisor have the qualifications listed in Regulatory Guide 1.8 for a Radiation Protection Manager.
Please update the resume of the Radiation Protection Supervi-sor, and specify the type of degree received and expand the description of his experience to allow evaluation against the criteria in Regulatory Guide 1.8.
RESPONSE
An update resume for Mr. McDuffee will be provided to the NRC under separate cover along with updated resumes of other supervisory personnel described in section 13.1.3.2.
Amended section 13.1.3.1 requires the Radiation Protection Supervisor to meet the qualification recommendations of the regulatory position of Regulatory Guide 1.8 for the Radiation Protection Manager.
Regulatory Guide 1.8 requires the Radiation Protection Manager to have bachelor's degree or the equivalent in a science or engineering subject. To provide clarification, our definition of "equivalent" in the context of Regulatory Guide 1.8, is as follows:
Four years of formal schooling in science or engineering,
- b. Four years of applied radiation protection experience at a nuclear facility, c ~ Four years of operational or technical experience/
training in nuclear power, or
- d. Any combination of the above totaling 4 years.
Amendment 5 13A-2 August 1981 8-06-81
<O17 8EG7lOH INCREMENTAL PVNGS FSAR AUG 11 1981 IN OUT APPENDIX 13A It should be noted that the above requirement is in addition to the requirements for five years of professional experience in applied radiation protection as specified in the guide.
0 August 1981 -13A-3 Amendment 5 8>>06-81
~ ~ ~
~
l
PVNGS FSAR +~V(CITAL PROCESS AND EFFL()ENT RADIOLOGICAL MONITORING AND SAI4PLING SYS4&9; SECTl0%
Table 11.5-1 AV(": 1 ~ 19BI CONTINUOUS PR(X'ESS AND EFFLUENT --~ .Tj RADIATION MONITORING (Sheet 4 of. 11)
Sampler/Monrtor Response Location Designated Location Expected Alarm Time at Min.
(Instr. Tag No.) Quantity For Environmental Design Background Detector Activity Range Concentrations Setpoint Detectable Automatic Actions (Pal Dwg. Ref.) (r) Per Unit Qualification(a) (mR/h Co-60) Sampler Type Type (b) Measu=ed(c)
Calibration'uclide (uci/cm3)
Power (uci/cm3) (u c i/cm3) Conc.(d) Supply(e) Initiated(
NON-ESF MONITORS (cont)
Central Calibration 1 Outside (Yard) N/A N/A Co-60 10-1 104 mR 0.5 mR/h 15 mR/h 30 Sec. Instr. Alarm only.
Facility Area (CFA) (Unit 1 h Monitor only)
(XJ-SQN-RU-24)
Controlled Machine Shop Radwaste Bldg. N/A 8-( Co-60 10-1 104 mR 0.5 mR/h 15 mR/h 30 Sec. Instr. Alarm only.
Area (MSA) Monitor h (XJ-SQN-RU-25)
Sample Room Area (SRA) Auxiliary Bldg. N/A N/A 8-) Co-60 10-1 104 mR 15 mR/h 30 Sec. Instr. Alarm only.
Monitor h (XJ-SQN-RU-26)
Waste Solidification lp-1 lp4 Radwaste Bldg. Alarm and initi-N/A N/A mR System Process Contiol Co-60 Variable Variable 30 Sec. Instr.
Area (PCA) Monitor ate shutdown of waste solidifi-(XJ-SQN-RU-27)
(XJ-SQN-RU-28) cation fill sequence.
Amendment 5 11.5-10 August 1981 I<- ~k 6-22-81 ~
$ ~33K4G') I
/
PVNGS FSAR ggMNKNTM PROCESS AND EFFLUENT RADIOLOGIST MONITORING AND SAMPLING SYSTEMS AU(i 1 j )981 Table 11. 5-1 CONTINUOUS PROCESS AND EFFLUX%'T}
RADIATION MONITOPING (Sheet 1 of 11) )5 Sampler/Monitor Response Location Designated Location Expected .Alarm Time at Min (Instr. Tag No.} Quantity For Environmental Design Background Detector Activity Calibration, Range> Concentrations Setpoint Detect~)je Automatic Actions (P&I Dwg. Ref.)(I) Per Unit Qualification(a) (mR/h Co-60) Sampler Type TYpe 'b) Measured(c) Nuclide ()jci/cm (pci/cm3) (uci/orna) Conc.
Power Initiated(f)
) Supply NON-ESF MONITORS Essential Cooling Water Auxiliary Bldg. 2.5 Off-Line/Liquid I Scint Gross Cs-137 lo 6-'lo 1 LMD(i) 2xlo (ECW} System Monitors 1 Min. Instr. Alarm only.
(XJ-SQN-RU-2 and Cs-137 XJ-SQN-RU-3)
(Figure 8.2-3)
Steam Generator Slowdown Auxiliary Bldg. 2.5 Off-Line/Liquid I Scint Gross Cs-137 10-6 10-1 2 x- x Io-'I Min. Instr, Alarm only (SGB) Monitors (XJ-SQN-RU-4 and XJ-SQN-FU-5)
Cs-~-: . s-137 1
Nuclear Cooling Water Outside (Yard) 0.5 Off-Line/Liquid I Scint Gross I Cs-137 lo 6-lo I 2xlo 1 Min. Instr. Alarm only.
(NCW) System Monitor (XJ-SQN-RU-6} Cs-137 (Figure 9.2-4 Sheet 1 of 3)
- a. Refer to section 1 1.5.2.1.1.6.1 for EsF monitors, section 11.5.2.1.1.6.2 for non-ESF monitors.
"I" - NAI I scintillation detector coupled with photomultiplier tube
- Plastic phosphor () scintillation detector coupled with photomultiplier tube "SCA" - Single channel analyzer "G-M" - Geiger-Mueller detector "Ion" - Ion Chamber
- c. a()-I" - Area radiation
- d. Time delay from imtrodudtion of radiation to sampler until indication of that level, for a clean filter con(ition.
"INSTR": 120 VAC Non IE Instrument Power a(M)a: Motor: 489 VAC Non IE Power "VITAL 'A'": 120 VAC Vital Instrument Power, Channel "A" "VITAL '8'": 120 VAC Vital Instrument Power, Channel "8"
- f. Automatic actions initiated on HIGH-HIGH alarm only. HIGH alarm annunciates but does not initiate a control -action.
- g. Blower Motor: 480 VAC Class IE Power, Train "A"
- h. ~ Blower Motor: 489 VAC Class IE Power, Train "B" "LMD" - less than minimum detectable
- j. Ba-133 is the calibration isotope for I-131.
- k. Kr-85 is the calibration isotope for Xe-133.
- l. Area monitors are shown on radiation zone diagrams, figures 12.3-1 through 12.3-20.
- n. Detector and Annunciator are located in Containment. Microprocessor is located in Auxiliary Building.
August 1981 11.5-7 6-22-81 I
w MZ. (, ol'73 oZ Amendment 5
~
%iT SECTiC MONITORING AND SAMPLING SYSTEMS AU61 0 SS Table 11.5-1 (J
CONTINUOUS PROCESS AND EFFLUE@
RADIATION MONITORING (Sheet 6 of 11) )5 Sampler/Monitor Response Location Designated Location Expected Alarm Time at Min.
(Instr. Tag No.) Quantity For Environmental Design Background Detector Act)arty Calibration Range Concentrations Setpoint Detectable Power Automatic Actions (PsI Dwg. Ref.) (I) Per Unit Qualification(a7 (mR/h Co-60) Sampler Type Type (b) Measued (c) Nuclide (uci/cm3) (uci/cm3) (uci/cm3) Conc.(d) Supply(e) Initiated(f)
M~00 MONITORS Portable Area Mo, itors N/A N/A G-M Co-60 I(-1 104 mR variable Variable 30 Sec. 120 VAC Alarm only.
h convenience outlets Movable Airborne Monitor N/A 1.0 Recirc/Moving Gros') Cs-137 5xlo -5xlo Variable Variable 15 Min. 120 VAC Alarm only.
Paper Particu- convenience late Filter outlets N/A Recirc/Fixed I/SCA Volatile Ba-133 5xlo 5xlo -5 Variable Variable 15 Min. Alarm only.
Charcoal or 1-131 Silver Xeolite Cartridge N/A Gros Recirc/Gas
()
Kr-85 lo 6-lo I variable Variable 1 Min. Alarm only.
11.5-12 August 1981 Amendment 5 g(<qzdo~ ls- -3