ML19212A413

Reg Guide 8.19, Occupational Radiation Dose Assessment in LWR Power Plants - Design Stage Man-Rem Estimates

ML19212A413
Person / Time
Site:	Grand Gulf
Issue date:	05/31/1978
From:	NRC OFFICE OF STANDARDS DEVELOPMENT
To:
Shared Package
ML19209A046	List: ML19209A045 ML19209A048 ML19209A050 ML19212A413
References
REGGD-08.019, REGGD-8.019, NUDOCS 7910010824
Download: ML19212A413 (9)

Text

ATTACIDtENT

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U.S. NUCLEAR REGilLATORY COMMISSION May 1978 Q(.%.

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OFFICE OF STAfJDARDS DEVELOPf.1ENT REGULATORY GUIDE 8.10 P00[

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I OCCUPATIONAL RADIATION DOSE ASSESSMENT IN LIGHT-WATER REACTOR POWER PLANTS DESIGN STAGE MAN-REM ESTIMATES A. INTRODUCTION ing knowledge of (1) the principal factors contribut-Section 50.34," Contents of Applications; Techni-ing t ccupational radiation exposures that occur at a cal Information," of 10 CFR Part 50, " Licensing of nucle r reactor power plant and (2) methods and Production and Utilization Facilities," requires that techniques for ensuring that the occupap,onal radia-each applicant for a permit to. construct a nuclear tion exposure will be ALARA. In assessing the col-power reactor provide a preliminary safety analysis lective ouupational dose at a plant,[the.appheant report (PSAR) and that each applicant for a license to evaluates each potentially significant dbye-causing

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operate 'such a facility provide ~a final safety analysis activity at that plant, sp.ecifiqally examemng such g

th,ings as design, shiep!tnge?pfan'elap,d radio out, traffic pat-report (FSAR). Section 50.34 specifies in general tcrns, expected ma5nteqance,, arf l

terms'th'e iriformation to be supplici.1 isthise repsts.-

sources, with a view'toaeducing unnecessary expo--

A more detailed description of the information sures and consiTerind $e c@t-effectiveness of each

-needed by the NRC staff in its evaluatiorr of applica-dose reduciig r'nc't!Iod arid tIchniqne. This evaluation i

tions is given in Regulatory Guide 1.70, " Standard - ggocess Mthe dos'e'mductions that may be ~ expected Format and Content-of Safetr Analysis Reports-for' to retutt3tPthelprincipaf obscrives of the dose -

Nuclear Power Plants." Section _12.4, " Dose As-asshssthent.%' #

sessment," of Regulatory Guide 1.70 states that the aA safety anal' sis report should provide the estimated MgT$pnocipal benefits arising from this evaluatio y

annual radiation exposure to personnel at the pp.

1procesj occur during the period of prehminary de-posed plant during normal operations. The purpose of ~.

sigp smce many of the ALARA practices are part of the man-rem estimate requirement irto ensure'that the design process. On the other hand, additional adequate detailed attention is given durih[the'ph benefits can also accrue during advanced design liminary design stage (as described in th' PSAR),is stages and even duririg early construction stages, as e

well as during construction after completion of design better evaluation of dose-causing operations are (as described in the FSAR), to dose-causindirIvities available and further design ~refinemerits can be iden-to ensure that personnel exposures will be as low as tified. In :iddition, operations that will need special reasonably achievable (ALAly). 'The safety analysis P anning and careful dose con.rol can be identified at l

report provides an op,p5nuriitK'or the applicant to the preoperational stage when the applicant can take demonstrate the adequacy.of that attention and to de.

advantage of all design options for reducing dose.

scribe whateverEisrhand.hocedural changes have resulted from th ose'alsessment process.

C. REGULATORY POSITION N

The obycty'e of hirguide is to describe a method This guide describes the format and content acceptable-to the NRC staff for performing an as-for assessments of the total annual occupational sessment ol{ollective occupational radiation dose as (man-rem) 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 during anticipated opera-B. DISCUC3lON tional occurrences. It should include estimates of the The dose assessment process requires a good work-frequency of occurrence, the existing or resulting USNRC REGULATORY GUiOES commems eowd be ** to t'e secre'arv of t** c-ss.oa. u s. wcw, =~ca Repast,ry Gwes are iwed to descr.te and rnat e evadab'e to the pt.bt< raet*'<Ms

o'Y Co** ***oa. W a.N~p ea. O C. 20555. Attear oa Dochenes ard Sav ce aaeptente to too fs AC statt of waptea eatq i.veofic par *s of the Ccnames oa's 8'***"-

regw!ations, to det.aeese tM ha=3=es used by the staff n eveiwating scocif.c secteeras i

The gw.ces er e nssued m the f ollomewig tea twoad d.v.s.oas or postwisted accdems or to prowde gudence to apolic.aats. flegwtatorv Gudes are not mehstitutes for regulat. oat, ams coraptiaare mth mera s not rab,ved t pw n,,,.o,s 6 Pr odxts

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Methods and totutio^s d.tf eteal from those,et out m the guides mil be acce9t-

2. Aegearch a-d Test Reactors F. Tramcertation ab6e J they prowde a tasis f or the f.ad.ags reyw.s.te to the.sm.aan-e oe coat.awance

3. F wees and weer.a s F acJ.t.es 9 Cresaonaat 6'essth e

of a peera.e or 1 ceme by the Cammes.oa.

4 E av roarrwatal and Sat'ag 9 A

• t' wat 't e* e**

V Coawasatt aad swwest oas for wvmeove aears a t*ese gu des are eacow *,od at all r

fir ies. sad gudes m.sf t.e revised. as socropriate, to accor,mcdate comraeats and

,,,,,,,, f or s.

coo.es of %.o gses Laa.cn mas tie e radweedi or br plae-to eeffect riew.ab.wmeemaa or e m per ac e.

How=ever, commeats on tNs gwce,ef m,,,

o,,n,,,,,,,,e d.str.but soa em f or s.rf e c op es of twe,ee gxes m sseebc v

esce wed eth.a about ten tricaths af ter.ts imaace, mil be perteewlerty wie've sa

d. sens shoi.td be Nde.a aret.ag to ree U S %ciear pepeeary Comm ss oa.

ewes et.ag the need for en eerry rev.s.on.

Wase mg+oa. O C.

23 55. astem ma D.rw'or. 0. s.oa of Cocsw e nt Coatros u

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yadiation levels, the manpower requin.,nents, and the radiation exposme estimates (such as Table duration ef such actisities. These estimates can be 1 ),

bascd on operating experience at similar plants, al.

thiu;1 Fro the extent possible estimat'es should include (2) Sufficient illustrative detail (such as that b,

shown in Tables 2 through 8) to explain how consideration of the design of the proposed plant, in-the radiation exposure asses ment process ciuding 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, dose assessment can

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_. 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.

pennit better estimates of radiation field intensities in Plant experi&ce orovides useful information on I e ti ns where work will be performed.

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 vall as on actual occupational radiation exposure ex.

changes and innovations will be incorporated into the perience. The a[5plicant should utilize personnel ex-de sign'.

posure data for specific Linds of work and job func-D. IMPLEMENTATION tions available from similar operating LWRs. (See Regulatory Guide 1.16 " Reporting of Operapng The purpose of this section is to provide informa-l Information-Appendix A Technical Specifica-tion to applicants regarding the NRC staff's plans for ti o n s,' f_or examples of work and job functions.).

using this regulatory guide.

Useful reports on these data have been published by.

This guide reflects current NRC staff practice.

the Atomic Industrial Forum, Inc., and the E!ectric Therefore, except in those cases in which the appli.

Power Research Institute, and a summary report on cant propoces an acceptable alternatSe 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 to be used in the evaluation N

The occupational dose assessment should include of submittals in connection with applications for con-projected doses during normal operations, anticipated struction permits or operating licenses until this guide operational occurrences, and shutdowns. Some of the is revised as a result of suggestions from the public or exposure. causing actisities that should be considered additional staff review. For construction permit, the m this dose assessment include steam generator tube review will focus principally on design consid-plugging and maintenance, repairs, inservice inspec-erations; for operating license, the review will focus tion, and replacement of pumps, valves, and gaskets.

principally on administrative and procedural coasid-erations.

Doses from nontoutine 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 identified and analyzed with respect to similar expo-Dou sures that have occurred under similar conditions at

"""b"d other operating facilities. In this manner, correctise r opersons and sudance measures can be. incorporated.in the design at an (see Tables 2 & 3)

1. 'IY 5' 3' Routine maintenance (see Table 4)

Waste processing (see Table 5)

Tables I through 8 a.re 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 coasidered. The actual numbers Special maintenance (see Table 8) appearing in the dose columns will depend on plant-specific information develop d in the course of the Total mannen' shear dose assessment review.

occupational esposures from Tables 2 through 8 are entered in Table I and added to obtain the facility's estimated total An objective of the dose assessment process should

1. 1. NuIsE Tabe's 2 through 8 are typicat examples (for be to desefop:

aw as 3.M Pw 0 for iMustrMise purpnes only. Acual values N

(1) A completed summary table of occupational can sary. deperding on the facility type (BwR or PWR). de-M, sign, and size.

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TAB!.E 2

' OCCUPATIONAL DOSE ESTIMATES DunlNG ROUTINE OPERATIONS AND SURVEILLANCE

• A verage Exposure Number L.

Jose rate time

- of Dose

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Activity (mrem /hr) *

(hr) workers Frequency (man-rtins/ year)

Walking 0.2 0.5 2

1/ shift 0.22

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Checking:

Containment cooling system 1

1 l'

1/ day 0.36 Accumulators 1.5 1

1 1/ day 0.54 Pressurizer valves 10 0.2 1

1/ day 0.73 Boron acid (BA) makeup system 5

0.2 1

,. 1/ day

, 0.36 Fuel pool system 1

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0.25 1

_ 1/ day.

0.09 Control rod drive (CRD) systeni:

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Modules 1

1 1

1/ day O.36 Controls 0.5 0.5 1

1/ shift 0.27 Filters 0.5 0.5 1

1/ day 0.09 1i Pumps:

l CRD 0.5 0.2 1

1/ day 0.04 Residual heat removal 1

0.2 1

1/ day 0.07

• The data shown are for iLstra6ve purposes onty and would be cupected to vary significandy from plant to prar,t.

(n TABLE 3 OCCUPATIONAL DOSE ESTIMATES DURING NONROUTINE OPERATION AND SURVEILLANCE

• A verage Exposure Nun ber Jose rate time of Dose Activity (mrem!hr)

(hr) workers Frequency (man-rems / year)

Operation of equipment:

Traversing in-core probe system 2

2 2

3/ year 0.02 Safety injection system 5

1 1

1/ month 0.06 Feedwater pumps &

turbine 1

1 1

1/ week 0.05 Instmment calibration 2

1 1

1/ day 0.73 Collection of

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radioactive samples:

Liquid system 10 0.5 1

1/ day 1.83 Gas syst:m 5

0.5 1

1/mor.th 0.03 Solid system 10 0.5 1

4/ year 0.02 Radiochemistry 1

I 2

1/ day 0.73 Radwaste operation 3

8 3

1/ week 3.75 IIcalth physics 1

2 2

1/ day 1.46 Total

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• The data shown are for i!!ustrative purposes only and woutd be expected to vary significantly from plant to plant.

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TADLE 4 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE MAINTENANCE

• Average Exposure Number

.h

?

Dose Jose rate time of Activity (mrem thr)

(hr) workers frequency (man-rems / year)

Mechanical:

Changing filters:

Waste filter 100 0.5 1

6/y ear 0.3 Laundry filter 100 0.5 1

10/ year 0.5 Boron acid filter 100 0.5 1

2/ year 0.1 Pressure valves 10 0.5 1

1/ week 0.26 BA makeup pump 10 0.3 1'

1/ week 0.16 BA holding pump 10

, 0.3 1

1/ week 0.16 Instrumentation and controls:

Transmitter inside containm nt 5

0.5 2

2/ week 0.52 Transmitter >utside l

containme it 1

2 1

1/Acek 0.1 Standby gas 'reatment 2

2 2

2/ year 0.02 system Radwaste,>rocessing 10 20 2

4/ year 1.6 system Total

~ 90-

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

TABLE S OCCUPATION AL DOSE ESTIMATES DURING WASTE PROCESSING

• Average Exposure Number doserate time of Dose Activity (mrem /hr)

(hr) workers Frequency (man-rems / year)

Control room 0.1 3000 1

1/ year 0.3 Sampling and filter changing 10 4

1 1/ week 2.1 Panel operation, inspection, and testing 1

2 1

1/ day 0.73 Operation of waste 2

12 2

1/ week 2.5 process,ing and, p.ckaging equipment Tc:al

'The dau shown uc for illustratisc purpeses only ar.d would be expected to vary significantly frem pf ant to plant.

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TABLE 6 j

OCCUPATIONAL DOSE ESTIMATES DURli1G HEFUELING*

P00R ORIBlWMAverage

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E.sposure Number dose rate time of Dose Activity (mrem /hr)

(hr) workers Frequency (man rems / year)

Reactor pressure vessel head and internals---

removal and installation 30 60 6

1/ year 10.8 Fuel preparation 10 24 2

1/ year 0.48 Fuel handling 2,5 100 4

1/ year 1.0 Fuel shipping 15 15 2

1/ year 0.45 Total l

• ne data shown are for i!!ustrative purposes only and would be espected to vary significantly from plant to plant.

3 Most work functions perfemed daring refueling, and the associated occupational dose received, will vary depending on facility design l

(BWR or PWR), reactor pressure vessel size, and cumber of fuel asse:nblies in the reactor core. For a detailed description of pre-planned activit:es, time, and manpower schedule, refer to the " critical path for refueling tasks," which should be available from the Nucicar Steam Supply System (NSSS) supplier.

./t TABLE 7 OCCUPATIONAL DOSE ESTIMATES Dl'7; rig INSERVICE INSPECTION

• Average E.tvosure Number Jose rate time of Dose Activity (mremlhr)

(hr) workers Frequency (man-remslyear)

Providing access: installation of platforms, ladders, etc., removal of thermal insulation 40 30 4

1/ year 4.8 Inspection of welds 40 100 3

1/ year 12.0 Follow up: installation of thermal insulation platform removal and cleanup 40 40 4

1/ year 6.4 Total

• The data shown are for i!!ustratise purpmes only and would be espected to vary significantly from plant to plant.

Estimates should be based on ascrage 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 equiprnent auternation as ailable for remote examinanon of

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welds.

I 8.19-5 1D??

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(ABLE 8 OCCUPATIONAL L.iE ESTIMATES DURING SPECIAL M.

4TENANCE*

Average Exposure Number u

Jose rate time of Dose 5

(mremlhr)

(hr)

. workers Frequency (man-rems / year)

G

. Activity Servicing of control rod drives 50

- 12 3

1/ year 1.8 Servicing of in-core detectors 15 10 2

I/ year 0.3 Replacement of control blades 15 10 2

1/ year 0.3 Dechanneling of spent and channeling of new fuel assemblies 10 60 2

1/ year 1.2 Steam generator repairs 1000 4

6 1/ year 24.0 Total

/

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

Most preplanned (or routine) mair:tenance activit:es during outige are described in the critical path for refueling tasks " which g

should be available from the NSSS supplier, and are performed in parallel with the critical path refueling tasks to shorten reactor i

outage time.

Actual dose will depend on facility design as well as size and thermal output and number of fuel assemblies in the reactor core.

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M5ING PAFQ 4

ATTACIIMENT S / 3/ / 7 43 DEf t0flSTRATION OF FUNCTIONAL CAPABILITY FOR PASSIVE PIPIf1G COMP 0NEtiTS MECHANICAL ENGINEERING BRANCH DIVISION OF SYSTEliS SAFETY I.

DEFINITIONS Functional Capability - Capability of piping components to deliver rated flow and retain dimensional stability when the design and service loads, and their resulting stresses and strains, are at prescribed levels.

Piping Components - These items of a piping system, such as tees, elbows, bends, pipe and tubing and branch connections constructed in accordance with the rules of Section III of the ASf1E Code.

(Systems designed to ANSI B31.1 or B31.7 are included as appropriate)

Piping System - A group of connected piping components and other associated Code components (i.e. pumps, valves, vessels) performing jointly a specified plant function, or in the case of multi-functional systems, more than one function.

II.

SITUATIONS IN WHICH FUf4CTIONAL CAPABILITY IS ASSURED UITHOUT FURTHER PROOF A.

Class 1 Piping Components:

1.

Functional capability may be considered assured without further proof for any class 1 piping component when the level "A" or "B" or "C" limit is used with Equation 9 of NB-3650 provided Do/ t1 50, where Do is the outside diameter and t is the wall 1033

i 2-KORKING PAPEl thickness of the piping component.

The level "C"

limit to be satisfied for the above verification procedure is:

1.5 Sy for austenitic piping components and 2.25 Sm for ferritic piping components.

2.

For tees and branch connections the level "D" limit may be used with equation 9 of flB-3650 without additional requirements for functional verification, provided Do/t<50.

The level "0" limit to be satisfied for the above verification procedure is:

2.0 Sy for austenitic piping components and 3.0 Sm for ferritic piping components, B.

Class 2/3 Piping Components:

1.

Functional capability may be considered assured for Class 2/3 piping components for Levels A and B limits in Equation (9) of NC-3652.1 or ND-3652.1 provided Do/t<50.

2.

For tees and branch connections, level "C" limits may be used without additional requirements for functional verification (.

However, for elbows or bends, the following additional require-ments shall be met whenever level "C" limits are specified:

(a) Use (0.8 B ) instead of (0.75 i) 2 (b) Use (1.5 S ) or (1.8 S ) whichever is lower for the right-h hand-side of Equation (9).

1033..;

' EORKING PAPQ In each of the above cases, Do/t shall be equal or less than 50.

3.

Class 2/3 piping components may be evaluated as Class.1 piping components for verifying functional capability, provided the rules and limits as specified in Item II. A. above are met.

III.

SITUATIONS IN WHICH FUNCTIONAL CAPABILITY REQUIRES ADDITIONAL DEMONSTRATION A.

Class 1 Piping Components:

1.

Piping components other than tees and branch connections, such as elbows, pipe bends and straight pipe, using level "D"

limits as defined in Item II.A.2.

2.

Any piping components with Do/t>50.

B.

Class 2/3 Piping Components:

1.

Straight pipe using level "C"

limits.

2.

Elbows or pipe bends which can not meet the requirements specified in Item II.B.2 above when level C limits are specified.

3.

Tees and branch connections when level "D" limits are specified.

4.

Any piping components with Do/t>50.

5.

Level "D" limits shall not be used for elbows, pipe bends and straight pipe.

l 1033