Regulatory Guide 8.19

Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage Man-Rem Estimates

ML13350A224
Person / Time
Issue date:	05/31/1978
From:	NRC/OSD
To:
References
RG-8.019
Download: ML13350A224 (6)

U.S. NUCLEAR REGULATORY COMMISSION May 1978 REGU LATORY GUIDE

OFFICE OF STANDARDS.DEVELOPMENT

REGULATORY GUIDE 8.19 OCCUPATIONALRADIATION DOSE-ASSESSMENT

IN LIGHT-WATER REACTOR POWER PLANTS

DESIGN STAGE MAN-REM ESTIMATES

A. INTRODUCTION

ing knowledge.of (i) the principal factors contribut- Section 50.34. "Contents of , nplications. Techni- ing tooccupational radiation exposures that oCcur ;ta cal.lnformation," of 10 CFR Par, 50, "Licensing of nuclear reactor power plant and (2) method-s and Production and Utilization Facilitk. ." requires that techniques for ensuring that the occupational radia- each applicant for a permit to. con.,truct a nuclear tion exposure will be ALARA. In assessing the Col- powcr reactor provide a preliminary safety analysis lective occupational dose at a.pla'ntv.the applicant report (PSAR) and that each applicant for a license to evaluates each potentially significant 'do.;e-causing opcraic such a facility provide a final safety analysis activity at that plant. specifically examining such report (FSAR). Section 50.34 specifies in general things as design. shieldingp..Iant layout. traffic pat- terms the inforniation to be supplied in these reports. terns, expected mainiLnancie arind radioactivity sources, with a vievtu: reducing unnecessary expo- A more detailed description, of the information sures and considering':the co ti-effecliveness of each needed by the NRC staff. in its evaluation of applica- dose-reducing method and techniquc. This evaluation tions is given in Regulatory Guide 1.70, "Standard process aiid-the dose:.'reductions that nmav he expected Format and Content of Safety Analysis Reports for. to resttI: nre ýtheK' principal objectives of the dose Nuclear Power Plants." Section 12.4. -Dose -As- sessment." of Regulatory Guide 1.70 states that the safety analysis report should provide the estimated ,, pnpal benefits arising frotm this evaluation

,The W

annual radiation exposure to personnel at the pro?"." process Lccur. during the period of prelimlinary de- sign since many of the ALARA practices are part of posed plant during normal operations. The purpdse' of the man-rem estimate requirement is to ensuriý..that the design process. On the other hand. additional adequate detailed attention is given during the pr.0,, benefits can also accrue during advanced design liminary design stage (as described in thii PSAR),*. stages and even during early construction s tages. as well as during construction after compltbn of design better evaluation of dose-causing oporaiions are (as described in the FSAR). to dose-causi fafctivities available and further design refinements can be iden- to ensure that personnel exposures will be as low as tified. In addition, operations that will need special reasonably achievable (Al:ARA). The safety analysis planning and careful dose control can be identified at report provides an opoiud ityjor the applicant to the preoperational stage when the applicant can take demonstrate the adequacy-,b thai'attention and to de- advantage of all design options for reducing dose.

• scribe whatever,ý.esigaandý'rocdural changes have resulted from tlikidose assessment proces

C. REGULATORY POSITION

• The objective 6(itthguide is to describe a method 'This guide describes the format and content

• acccptabldi.to the NRC stuff for performing an ;is- for assessments of the total annual occupational sessment of 'ollective occupational radiation dose as (man-ren) dose at an LWR-principally during the

• *part of the process of designing a light-water-cooled design stage. The dose assessment at this stage power reactor (LWR). should include estimated annual personnel exposures during normal operation and dining anticipated opera-

B. DISCUSSION

tional occurrences. It should include estimates of the The dose assessment process requires a good work- frequency of occurrence, the existing or resulting USNYRC REGULATORY GUIDES Commnwta bh~uftilbe swnt to It'. Stitievhsy of the Comnfnjvtsn.US Nu'ti-A. Areq, Fligullator Guefnw et lisued to deeehbaahu~natke&aiia&te to me pubic mqethods taint Comm~ts.t~n. Wath,,nqtun OZ. 20651j. Att..ntion Outbhethi; ..... 5in...

aameotabl. to th*.NAC sMoll al .nnplamefiting specifi~c owls of the. Commtuoon's ofoied.

igguitotiotti.1dodlineate tectinsquet ted by She%falli nevaluoloqg tftiloc tsobiems The quittine.0-wsu"Io the tnilslwni t-, fw,..el 0tn-w,,

or: poinulated accidents. at to PtoneS. ouicdance t0 moiticents. Rtegulatory Guirks are not gsastnuten kw regulationst. andS copitpance vvitf them it not rotsuired. 1.pow" fli ' &JPNfwcf.

Mfithods aenc volutiott1 diffleten from thotselt out in the VuKde¶"nit be etcl1i 2. Research omtiTest Reatolw 7. tfin'itu awleit they provide a bouitfor the findigtraquisiteto the iknce or conttinuance. 3. FuelsandMairriAls Fdcatie 9. occu",iifmrufttefaltil of*&offitt of tkiceMeby the Cammts,.nn.. .Etn~nd ~l ~a Aflitmut At.oms Comment s and iueUl antoifor improvements in thewequidles we eescousepd! at 0eeal n ~n tt'to, 5 eea timeW1. aredQus~t e.~t be revised, as uopoatovito. to aco.rmmodate cornmertis and Aestuests Irv singte caione ol tivuemitpen lwh4,ch may to. me.'mslu.uJI to. Ito ut..r to #effect nowa inliomatirnn cit e.miernrce. Howevrr. common%%antt Ithi i quidt~it men rton autctflonwlc dirlmithitstro- 1- ttot n%-91P..nnesoil iw,ottrqnet . sfo. ti raceid v.fttin~ about two rrinoftlt after its iuMSce, tvill be pt~itcultidv useful inl iftnu~nns dsicukl be nudfe in oakn w fqit.the US. Nurf"~ 6feqr.tutautsCtsnc -nnn.

esetustin,1the neted lot an eary reCvisici, Whehnhsfltm,0,C. M05$t. Attentiosi Doecois. 0-%o.nn itI Dii-otrent Custuro

radiation levels. the manpower requiremients. and the radiation exposure estimates (such as Table duration of such activities. These estimates can be I).

based on operating experience at similar plants, al- (2) Sufficient illustrative detail (such as that though to the extent possible estimates should include shown in Tables 2 through 8) to explain how consideration of the design of the proposed plant, in- the radiation exposure assessment process cluding radiation field intensities calculated on the was performed, and basis of the plant-specific shielding design. (3) A description of any design changes that The dose assessment process and the concomitant were made as a result of the dose assessment dose reduction analysis should involve individuals process.

trained in plant system design. shield design, plant During the final design stage. (lose assessment can operation. and health physics, respectively. Knowl- be substantially refined, since at this time details of edge from all these disciplines should be applied to the design will be known. In particular. completed the dose assessment in determining cost-effective shielding design and layout of equipment should dose reductions. permit better estimates of radiation field intensities in Plant experience provides useful information on locations where work will be performed. 4 the numbers of people needed for jobs, the duration As a result of the dose assessment process, it is to of different jobs. and the frequency of the jobs. as be expected that various dose-reducing design well as on actual occupational radiation exposure ex- changes and innovations will be incorporated into the perience. The applicant should utilize personnel ex- design.

posure data for specific kinds of work and job func-

D. IMPLEMENTATION

tions available from similar operating LWRs. (See The purpose of this section is to provide informa- Regulatory Guide 1.16. "Reporting of Operating tion to applicants regarding the NRC staff's plans for Information-Appendix A Technical Specifica.

using this regulatory guide.

tions." for examples of work and job functions.)

Useful reports on these data have been published by This guide reflects current NRC staff practice.

the Atomic Industrial Forum. Inc., and the Electric Therefore, except in those cases in which the appli- Power Research Institute. and a summary report on cant proposes an acceptable altcrnatlve method for occupational radiation exposures at nuclear power complying with specified portions of the Commis- plants is distributed annually by the Nuclear sion's regulations, the method described herein is Regulatory Commission. being and will continue ito be used in the evaluation of submittals in connection with applications for con- The occupational dose assessment should include projected doses (luring normal operations. anticipated struction permits or operating licenses until this guide is revised as a result of suggestions from the public or operational occurrences, and shutdowns. Some of the exposure-causing activities that should be considered additional staff review. For construction permit. the review will focus principally on design consid- in this dose assessment include steam generator tube erations; for operating license, the review will focus plugging and maintenance, repairs, inservice inspec- principally on administrative and procedural consid- tion. and replacement of pumps, valves, and gaskets, erations.

Doses from nonroutine activities that are anticipated operational occurrences should be included in the ap- TABLE 1 plicant's ALARA dose analysis. Radiation sources and personnel activities that contribute significantly TOTAL OCCUPATIONAL RADIATION

to occupational radiation exposures should be clearly EXPOSURE ESTIMATES

Dose identified and analyzed with respect to similar expo- sures that have occurred under similar conditions at Activity (nian-reinslyear)

Reactor operations and surveillance other operating facilities. In this manner, corrective (see Tables 2 & 3) *

measures can be incorporated in the design at an Routine maintenance (see Table 4)

early stage.

Waste processing (see Table 5)

Tables I through 8 are examples of worksheets for Refueling (see Table 6)

tabulation of data in the dose assessment process to Inservice inspection (see Table 7) -

indicate the factors considered. The actual numbers Special maintenance (see Table 8) -

appearing in the dose columns will depend on plant- specific information developed in the course of the Total man-reins/year dose assessment review. *Occupational exposures from Tables 2 through 8 arc entered in Table I and added to obtain the racility's estimated total An objective of the dose assessment process should yearly occupational dose.

Values shown in Tables 2 through 8 arc typical examples (for be to develop: BWRs and PWRs) for illustrative purposes only. Actual values can vary. depending on the facility type (BWR or PWR). de- (I) A completed summary table of occupational sign. and size.

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TABLE 2 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE OPERATIONS AND SURVEILLANCE*

A verage Exposure Number dose rate time of f)tse Activily Imremn/hir) (hr) workers Frequetwy (man-rerns/vear)

Walking 0.2 0.5 2 I/shift 0.22 Checking:

Containment cooling system 1 1 I/day 0.36 Accumulators 1.5 I/day 0.54 Pressurizer valves 10 0.2 I/day 0.73 Boron acid (BA) makeup system 5 0.2 1/day 0.36 Fuel pool system i 0.25 I/day 0.09 Control rod drive (CRD) system:

Modules 1 1 1/day 0.36 Controls 0.5 0.5 Ilshift 0.27 Filters 0.5 0.5 I/day 0.09 Pumps:

CRD 0.5 0.2 1! I/day 0.04 Residual heat removal 0.2 I/day 0.07

°

Total

• 'Te data shown are for illustrative purposcs only and would be expected to vary significantly from plant ,; plant.

TABLE 3 OCCUPATIONAL DOSE ESTIMATES DURING NONROUTINE OPERATION AND SURVEILLANCE*

Average Exposure Number dose rate time of Dose Activity (mrem/lr) (hr) workers Frequency (man-rems/yvear)

Operation of equipment:

Traversing in-core probe system 2 2 2 3/year 0.02 Safety injection system 5 I/month 0.06 Feedwater pumps &

turbine 1 I/week 0.05 Instrument calibration 2 I/day 0.73 Collection of radioactive samples:

Liquid system 10 0.5 I/day 1.83 Gas system 5 0.5 I/month 0.03 Solid system I0

10 0.5 4/year '0.02 Radiochemistry 1 2 I/day 0.73 Radwaste operation 3 8 3 I/week 3.75 Health physics 1 2 2 I/day 1.46 Total

• The data shown arc for illustrative purposes only and would be expected to vary significantly from plant to plant.

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TABLE 4 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE MAINTENANCE*

Average  !ýxposure Number Aciivity dose rate

( mren/Iir)

time (hr)

of workers Freeiuenc)v Dose (mnimz-reinlfl/eur) 0

Mechanical:

Changing filters:

Waste filter 100 0.5 6/year 0.3 Laundry filter 100 0.5 10/year 0.5 Boron acid filter 100 0.5 2/year 0.1 Pressure valves 10 0.5 1/week 0.26

13A makeup pump 10 0.3 iU;-4ck 0.16 BA holding pump 10 0.3 1/%,e:.k 0.16 Instrumentation and controls:

Transmitter inside containment 5 0.5 2/weck 0.52 Transmitter outside

1 2 I/week 0.1 containment Standby gas treatment system 2 2 2/year 0.02 Radwaste processing system 10 20 4/year 1.6 Total

• The data shown are for illustrative purposes only and would be expected to vary significantly from plant to plant.

TABLE 5 OCCUPATIONAL DOSE ESTIMATES DURING WASTE PROCESSING*

A verage Exposure Number dose rate time of Dose Activity (mrem/hr) (hr) workers Frequency (man -rems year)

Control room 0.1 3000 I/year 0.3 Sampling and filter changing 10 4 1/week 2.1 Panel operation, inspection, and testing 1 2 I/day 0.73 Operation of waste 2 12 2 I/week 2.5 processing and packaging equipment Total

• The data shown are for illustrative purposes only and would be expected to vary significantly from plant to plant.

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TABLE 6 OCCUPATIONAL DOSE ESTIMATES DURING REFUELING*

A verage Exposure Number dose rate time Dose Activity (nrentIhr) (hr) workers Frequenc). (mn-rntrcslvear)

Reactor pressure vesscl head and intcrnals- removal and installation 30 60 6 I/year 10.8 Fuel preparation 10 24 2 I/year 0.48 Fuel handling 2.5 100 4 L'year 1.0

Fuel shipping 15 15 2 I/year 0.45 Total

• The data shown are for illustrative purposes only and would be expected to vary significantly from plant to pla'ni.

Most work functions performed during rcfueling. and the associated occupational dose received, will vary depending on facility design (BWR or PWR), reactor pressure vessel size. and number of fuel assemblics in the reactor core. For a detailed description of prc- planned activities, time. and manpower schedule, refer to the "'critical path for refueling task%.*' which should he available from the Nuclear Steam Supply System tNSSS) supplier.

TABLE 7 OCCUPATIONAL DOSE ESTIMATES DURING INSERVICE INSPECTION'

A verage Exposure Number dose rate time of Dose'

Activity (in rem Ih r) (hr) svorkers Freqienc-Y (mian -rct:sl/v*arj Providing access: installation of platforms, ladders.

etc., removal of thermal insulation 40 30 4 I/year 4.8 Inspection of welds 40 100 3 I/year 12.0

Follow up: installation of thermal insulation platform removal and cleanup 40 40 4 I/Ycar 6.4 Total

• The data shown are for illustrative purposes only and would be expected to vary significantly from plant to plant.

Estimates should be based on average yearly values over a 10-year period. Variations are expected as a consequence of reactor size, design, number of welds to be inspected yearly. and the degree of equipment automation available for remote camination of welds.

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TABLE 8 OCCUPATIONAL DOSE ESTIMATES DURING SPECIAL MAINTENANCE "

A vero.e L'xiiositrc Nunber hiost rale lime of fivioy (lir-in lir) (hr) workers Fr*'qseitcY (inuni-renslls/etr)

Servicing of control rod drives 50 12 3 I/yea r 1.1i Servicing of in-core detectors 15 10 2 1/year 0.3 Replaccment of control blades Is 10 I/year 0.3 Dechanneling of spent and channeling of new fuel assemblies 0() 60 2 I/year 1.2 Steam generator repairs 1000 4 6 1/year 24.0

Total

• Thc data shown are for illustrative ;Iurptisc only and would he epected to vary significantly front plant to plant.

Nto%t prcplanned (or riwlinet rnt~enanicc ajoivities durink. otitage arc de-,ritcd in the -critical path fo'r refueling task-,".which

%hould be availabule fromn the NSSS supplier, and ire performed in parallel with the critical path refueling tasks to %horiten reactor outage time Actual d,.'e %killdepcndl on faeiliity desigzn a%wekll a!, %ize and thermal output and nuniher tit fuel assemblics in the rcicior cote.

8.19.6