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

Section 50.34. "Contents of , nplications. Techni- cal.lnformation," of 10 CFR Par, 50, "Licensing of Production and Utilization Facilitk. ." requires that each applicant for a permit to. con.,truct a nuclear powcr reactor provide a preliminary safety analysis report (PSAR) and that each applicant for a license to opcraic such a facility provide a final safety analysis report (FSAR). Section 50.34 specifies in general terms the inforniation to be supplied in these reports.

A more detailed description, of the information needed by the NRC staff. in its evaluation of applica- tions is given in Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for.

Nuclear Power Plants." Section 12.4. -Dose -As- sessment." of Regulatory Guide 1.70 states that the safety analysis report should provide the estimated W

annual radiation exposure to personnel at the pro?"."

posed plant during normal operations. The purpdse' of the man-rem estimate requirement is to ensuriý..that adequate detailed attention is given during the pr.0,,

liminary design stage (as described in thii PSAR),*.

well as during construction after compltbn of design (as described in the FSAR). to dose-causi fafcti vities to ensure that personnel exposures will be as low as reasonably achievable (Al:ARA). The safety analysis report provides an opoiud ityjor the applicant to demonstrate the adequacy-,b thai'attention and to de-

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

The objective 6(itthguide is to describe a method

acccptabldi.to the NRC stuff for performing an ;is- sessment of 'ollective occupational radiation dose as

• *part of the process of designing a light-water-cooled power reactor (LWR).

B. DISCUSSION

The dose assessment process requires a good work- ing knowledge.of (i) the principal factors contribut- ing tooccupational radiation exposures that oCcur ;t a nuclear reactor power plant and (2) method-s and techniques for ensuring that the occupational radia- tion exposure will be ALARA. In assessing the Col- lective occupational dose at a.pla'ntv.the applicant evaluates each potentially significant 'do.;e-causing activity at that plant. specifically examining such things as design. shieldingp..Iant layout. traffic pat- terns, expected mainiLnancie arind radioactivity sources, with a vievtu: reducing unnecessary expo- sures and considering':the co ti-effecliveness of each dose-reducing method and techniquc. This evaluation process aiid-the dose:.'reductions that nmav he expected to resttI: nre ýtheK' principal objectives of the dose

,,

,The pnpal benefits arising frotm this evaluation process Lccur. during the period of prelimlinary de- sign since many of the ALARA practices are part of the design process. On the other hand. additional benefits can also accrue during advanced design stages and even during early construction s tages. as better evaluation of dose-causing oporaiions are available and further design refinements can be iden- tified. In addition, operations that will need special planning and careful dose control can be identified at the preoperational stage when the applicant can take advantage of all design options for reducing dose.

C. REGULATORY POSITION

'This guide describes the format and content for assessments of the total annual occupational (man-ren) dose at an LWR-principally during the design stage. The dose assessment at this stage should include estimated annual personnel exposures during normal operation and dining anticipated opera- tional occurrences. It should include estimates of the frequency of occurrence, the existing or resulting USNYRC REGULATORY GUIDES

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radiation levels. the manpower requiremients. and the duration of such activities. These estimates can be based on operating experience at similar plants, al- though to the extent possible estimates should include consideration of the design of the proposed plant, in- cluding radiation field intensities calculated on the basis of the plant-specific shielding design.

The dose assessment process and the concomitant dose reduction analysis should involve individuals trained in plant system design. shield design, plant operation. and health physics, respectively. Knowl- edge from all these disciplines should be applied to the dose assessment in determining cost-effective dose reductions.

Plant experience provides useful information on the numbers of people needed for jobs, the duration of different jobs. and the frequency of the jobs. as well as on actual occupational radiation exposure ex- perience. The applicant should utilize personnel ex- posure data for specific kinds of work and job func- tions available from similar operating LWRs. (See Regulatory Guide 1.16. "Reporting of Operating Information-Appendix A Technical Specifica.

tions."

for examples of work and job functions.)

Useful reports on these data have been published by the Atomic Industrial Forum. Inc., and the Electric Power Research Institute. and a summary report on occupational radiation exposures at nuclear power plants is distributed annually by the Nuclear Regulatory Commission.

The occupational dose assessment should include projected doses (luring normal operations. anticipated operational occurrences, and shutdowns. Some of the exposure-causing activities that should be considered in this dose assessment include steam generator tube plugging and maintenance, repairs, inservice inspec- tion. and replacement of pumps, valves, and gaskets, Doses from nonroutine activities that are anticipated operational occurrences should be included in the ap- plicant's ALARA dose analysis. Radiation sources and personnel activities that contribute significantly to occupational radiation exposures should be clearly identified and analyzed with respect to similar expo- sures that have occurred under similar conditions at other operating facilities. In this manner, corrective measures can be incorporated in the design at an early stage.

Tables I through 8 are examples of worksheets for tabulation of data in the dose assessment process to indicate the factors considered. The actual numbers appearing in the dose columns will depend on plant- specific information developed in the course of the dose assessment review.

An objective of the dose assessment process should be to develop:

(I) A completed summary table of occupational radiation exposure estimates (such as Table I).

(2) Sufficient illustrative detail (such as that shown in Tables 2 through 8) to explain how the radiation exposure assessment process was performed, and

(3) A description of any design changes that were made as a result of the dose assessment process.

During the final design stage. (lose assessment can be substantially refined, since at this time details of the design will be known. In particular. completed shielding design and layout of equipment should permit better estimates of radiation field intensities in locations where work will be performed.

As a result of the dose assessment process, it is to be expected that various dose-reducing design changes and innovations will be incorporated into the design.

D. IMPLEMENTATION

The purpose of this section is to provide informa- tion to applicants regarding the NRC staff's plans for using this regulatory guide.

This guide reflects current NRC staff practice.

Therefore, except in those cases in which the appli- cant proposes an acceptable altcrnatlve method for complying with specified portions of the Commis- sion's regulations, the method described herein is being and will continue ito be used in the evaluation of submittals in connection with applications for con- struction permits or operating licenses until this guide is revised as a result of suggestions from the public or additional staff review. For construction permit. the review will focus principally on design consid- erations; for operating license, the review will focus principally on administrative and procedural consid- erations.

TABLE 1 TOTAL OCCUPATIONAL RADIATION

EXPOSURE ESTIMATES

Dose Activity (nian-reinslyear)

Reactor operations and surveillance (see Tables 2 & 3)

Routine maintenance (see Table 4)

Waste processing (see Table 5)

Refueling (see Table 6)

Inservice inspection (see Table 7)

-

Special maintenance (see Table 8)

-

Total man-reins/year

• Occupational exposures from Tables 2 through 8 arc entered in Table I and added to obtain the racility's estimated total yearly occupational dose.

Values shown in Tables 2 through 8 arc typical examples (for BWRs and PWRs) for illustrative purposes only. Actual values can vary. depending on the facility type (BWR or PWR). de- sign. and size.

4

8.19-2

TABLE 2 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE

A verage Exposure dose rate time Activily Imremn/hir)

(hr)

OPERATIONS AND SURVEILLANCE*

Number of Walking Checking:

Containment cooling system Accumulators Pressurizer valves Boron acid (BA) makeup system Fuel pool system Control rod drive (CRD) system:

Modules Controls Filters

0.2

1

1.5

10

5 i

0.5

1

0.2

0.2

0.25

1

0.5

0.5

0.2

0.2 workers

2 Frequetwy I/shift I/day I/day I/day

1/day I/day

1/day Ilshift I/day f)tse (man-rerns/v ear)

0.22

0.36

0.54

0.73

0.36

0.09

0.36

0.27

0.09 Pumps:

CRD

Residual heat removal

1

0.5

0.5

0.5

1!

°

I/day I/day

0.04

0.07 Total

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

OCCUPATIONAL DOSE

Activity Operation of equipment:

Traversing in-core probe system Safety injection system Feedwater pumps &

turbine Instrument calibration Collection of radioactive samples:

Liquid system Gas system Solid system Radiochemistry Radwaste operation Health physics TABLE 3 ESTIMATES DURING NONROUTINE OPERATION AND

A verage Exposure Number dose rate time of (mrem/lr)

(hr)

workers Frequency

2

5

1

2

2

0.5

0.5

0.5

1

8

2

2

2

3

2

3/year I/month I/week I/day I/day I/month

4/year I/day I/week I/day SURVEILLANCE*

Dose (man-rems/yvear)

0.02

0.06

0.05

0.73

1.83

0.03

'0.02

0.73

3.75

1.46

10

5 I0

10

3

1 Total

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

8.19-3

TABLE 4 OCCUPATIONAL DOSE ESTIMATES DURING

A verage

!ýxposure dose rate time Aciivity

( mren/Iir)

( hr)

ROUTINE MAINTENANCE*

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

Mechanical:

Changing filters:

Waste filter Laundry filter Boron acid filter Pressure valves

13A makeup pump BA holding pump Instrumentation and controls:

Transmitter inside containment Transmitter outside containment Standby gas treatment system Radwaste processing system

100

100

100

10

10

10

5

1

2

10

0.5

0.5

0.5

0.5

0.3

0.3

0.5

2

2

20

6/year

10/year

2/year

1/week iU;-4ck

1/%,e:.k

2/weck I/week

2/year

4/year

0.3

0.5

0.1

0.26

0.16

0.16

0.52

0.1

0.02

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 (mrem/hr)

(hr)

workers Frequency (man Activity Dose

-rems year)

Control room Sampling and filter changing Panel operation, inspection, and testing Operation of waste processing and packaging equipment

0.1

10

1

2

3000

4

2

12

2 I/year

1/week I/day I/week

0.3

2.1

0.73

2.5 Total

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

8.19-4

TABLE 6 OCCUPATIONAL DOSE ESTIMATES DURING REFUELING*

A verage dose rate (nrentIhr)

Exposure time (hr)

Number workers Dose Frequenc).

(mn-rntrcslvear)

Activity Reactor pressure vesscl head and intcrnals- removal and installation Fuel preparation Fuel handling Fuel shipping

30

10

2.5

15

60

24

100

15

6 I/year

10.8

2 I/year

0.48

4 L'year

1.0

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 dose rate (in rem Ih r)

Activity Providing access: installation of platforms, ladders.

etc., removal of thermal insulation Inspection of welds Follow up: installation of thermal insulation platform removal and cleanup Exposure time (hr)

30

100

Number of svorkers

4

3

4 Dose'

Freqienc-Y

(mian -rct:sl/v*arj

40

40

I/year I/year I/Ycar

4.8

12.0

6.4

40

40

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.

8.19-5

TABLE 8 OCCUPATIONAL DOSE ESTIMATES DURING SPECIAL

A vero.e L'xiiositrc Nunber hiost rale lime of fivioy (lir-in lir)

(hr)

workers MAINTENANCE "

Fr*'qseitcY

(inuni-renslls/etr)

Servicing of control rod drives Servicing of in-core detectors Replaccment of control blades Dechanneling of spent and channeling of new fuel assemblies Steam generator repairs

50

15 Is

0()

1000

12

10

10

60

4

3

2 I/yea r

1/year I/year I/year

1/year

1.1i

0.3

0.3

1.2

24.0

2

6 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 %kill depcndl on faeiliity desigzn a% wekll a!, %ize and thermal output and nuniher tit fuel assemblics in the rcicior cote.

8.19.6