Regulatory Guide 8.19: Difference between revisions

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{{Adams
{{Adams
| number = ML13350A224
| number = ML18165A214
| issue date = 05/31/1978
| issue date = 06/14/2018
| title = Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage Man-Rem Estimates
| title = Periodic Review
| author name =  
| author name = Stutzcage E
| author affiliation = NRC/OSD
| author affiliation = NRC/NRO/DSEA
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
| docket =  
| docket =  
| license number =  
| license number =  
| contact person =  
| contact person = Karagiannis H
| document report number = RG-8.019
| case reference number = RG-8.019, Rev 1
| document type = Regulatory Guide
| package number = ML18165A204
| page count = 6
| document type = Regulatory Guidance
| page count = 2
}}
}}
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION                                                                                            May 1978 REGU LATORY GUIDE
{{#Wiki_filter:Regulatory Guide Periodic Review Regulatory Guide Number:                8.19, Revision 1 Title:                                Occupational Radiation Dose Assessment in Light- Water Reactor Power Plants -- Design Stage Man-Rem Estimates Office/Division/Branch:                NRO/DSEA/RPAC
                                            OFFICE OF STANDARDS.DEVELOPMENT
Technical Lead:                        Ed Stutzcage Staff Action Decided:                  Revise
                                                                          REGULATORY GUIDE 8.19 OCCUPATIONALRADIATION DOSE-ASSESSMENT
1.    What are the known technical or regulatory issues with the current version of the Regulatory Guide (RG)?
                                                        IN LIGHT-WATER REACTOR POWER PLANTS
      RG 8.19 was issued in 1979 to describe a method acceptable to the NRC staff for performing an assessment of collective occupational radiation dose to meet the requirements for the "As Low as is Reasonably Achievable" program in 10 CFR part 20,
                                                              DESIGN STAGE MAN-REM ESTIMATES
      Standards for Protection against Radiation, and in 10 CFR part 50.34, "Contents of Applications; Technical Information."
      RG 8.19 is still consistent with the requirements in 10 CFR Part 20, 10 CFR Part 50.34, and RG 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition). Some of the references in the guide are outdated. In addition, references to some regulations and guidance documents need to be updated to account for 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. In particular, the term man-rem should to be changed to the revised term person-rem.


==A. INTRODUCTION==
However, these changes are administrative in nature and do not affect the technical content of the guide or what is expected from licensees or applicants in performing design stage dose estimates.
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
2.    What is the impact on internal and external stakeholders of not updating the RG
      for the known issues, in terms of anticipated numbers of licensing and inspection activities over the next several years?
      Although no licensing actions are anticipated over the next several years, revision of this RG will assist current licensees if they choose to develop procedures for their facility for occupational radiation dose assessment.


====s.     ====
3.    What is an estimate of the level of effort needed to address identified issues in terms of full-time equivalent (FTE) and contractor resources?
      An estimate of the effort needed to revise this RG is between 0.1 FTE and 0.15 FTE. No contractor support is anticipated.


==C. REGULATORY POSITION==
Regulatory Guide Periodic Review
*          The objective 6(itthguide is to describe a method                                        'This guide describes the format and content
4.   Based on the answers to the questions above, what is the staff action for this guide (Reviewed with no issues identified, Reviewed with issues identified for future consideration, Revise, or Withdraw)?
*    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
      Revise.
  * *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==
5.   Provide a conceptual plan and timeframe to address the issues identified during the review.
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,,
The staff plans to develop a draft guide that will be submitted to the Office of Nuclear Regulatory Research by the third quarter of FY20 and to issue it for public comment by the fourth quarter of FY20.
        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.
NOTE: This review was conducted in June 2018 and reflects the staffs plans as of that date. These plans are tentative and subject to change.}}
 
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.
 
8.19-2
 
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.
 
8.19-3
 
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.
 
8.19-4
 
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.
 
8.19-5
 
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}}


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Revision as of 12:25, 30 November 2019

Periodic Review
ML18165A214
Person / Time
Issue date: 06/14/2018
From: Stutzcage E
NRC/NRO/DSEA
To:
Karagiannis H
Shared Package
ML18165A204 List:
References
RG-8.019, Rev 1
Download: ML18165A214 (2)
v • d • e


Regulatory Guide Periodic Review Regulatory Guide Number: 8.19, Revision 1 Title: Occupational Radiation Dose Assessment in Light- Water Reactor Power Plants -- Design Stage Man-Rem Estimates Office/Division/Branch: NRO/DSEA/RPAC

Technical Lead: Ed Stutzcage Staff Action Decided: Revise

1. What are the known technical or regulatory issues with the current version of the Regulatory Guide (RG)?

RG 8.19 was issued in 1979 to describe a method acceptable to the NRC staff for performing an assessment of collective occupational radiation dose to meet the requirements for the "As Low as is Reasonably Achievable" program in 10 CFR part 20,

Standards for Protection against Radiation, and in 10 CFR part 50.34, "Contents of Applications; Technical Information."

RG 8.19 is still consistent with the requirements in 10 CFR Part 20, 10 CFR Part 50.34, and RG 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition). Some of the references in the guide are outdated. In addition, references to some regulations and guidance documents need to be updated to account for 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. In particular, the term man-rem should to be changed to the revised term person-rem.

However, these changes are administrative in nature and do not affect the technical content of the guide or what is expected from licensees or applicants in performing design stage dose estimates.

2. What is the impact on internal and external stakeholders of not updating the RG

for the known issues, in terms of anticipated numbers of licensing and inspection activities over the next several years?

Although no licensing actions are anticipated over the next several years, revision of this RG will assist current licensees if they choose to develop procedures for their facility for occupational radiation dose assessment.

3. What is an estimate of the level of effort needed to address identified issues in terms of full-time equivalent (FTE) and contractor resources?

An estimate of the effort needed to revise this RG is between 0.1 FTE and 0.15 FTE. No contractor support is anticipated.

Regulatory Guide Periodic Review

4. Based on the answers to the questions above, what is the staff action for this guide (Reviewed with no issues identified, Reviewed with issues identified for future consideration, Revise, or Withdraw)?

Revise.

5. Provide a conceptual plan and timeframe to address the issues identified during the review.

The staff plans to develop a draft guide that will be submitted to the Office of Nuclear Regulatory Research by the third quarter of FY20 and to issue it for public comment by the fourth quarter of FY20.

NOTE: This review was conducted in June 2018 and reflects the staffs plans as of that date. These plans are tentative and subject to change.


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