ML20052D823
| ML20052D823 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/04/1982 |
| From: | Barlit V, Bridenbaugh D, Goldsmith M SUFFOLK COUNTY, NY |
| To: | |
| References | |
| ISSUANCES-OL, NUDOCS 8205070243 | |
| Download: ML20052D823 (69) | |
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NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD t ) In the Matter of ) i ) LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 0.L. (Shoreham Nuclear Power Station, Unit 1) ) ) ) f PREPARED DIRECT TESTIMONY OF MARC W. GOLDSMITH, DALE G. BRIDENBAUGH AND VERA BARLIT ON BEHALF 0F SUFFOLK COUNTY N pts b REGARDING SUFFOLK COUNTY 15 g3 ,3 2 CONTENTION 26 - ALARA 4 J U li \\ g s\\ May 4, 1982 5%%fo p$ g ilt T
6 / Y I Sumary Outline of Suffolk County i Contention 26 Testimony
- Suffolk County contends that LILCO has not adequately demonstrated that Shoreham meets the requirements of 10 CFR 20.1(c) with regard to provisions for maintaining occupational radiation exposure as low as is reasonablyachievable(ALARA).
The Nuclear Regulatory Corrmission (NRC) regulates its licensees to restrict an individual's exposure to within prescribed limits. In ~ addition the NRC requires that nuclear plant cperators maintain collective and individual exposures as far below the limit as is reasonably achievable. This means that every activity at a nuclear facility involving exposure to radiation should be planned to minimize unnecessary exposure to individual workers and to the worker population. 4 According to the NRC, this can be done by: establishing an ALARA program; properly' designing f acilitics and selecting equipment; establishing a radiation control program, plans, and procedures; and 1 making support equipment, instrumentation and f acilities available. Maintaining occupational exposures ALARA is a requirerrent of 10 CFR 20.l(c). The specific concerns relate to implementing an ALARA program at Shoreham. These concerns arise from inadequate features and procedures to minimize occupational radiation exposures primarily during maintenance. Specifically there are: 1 inadequate design features to separate and isolate equipment; l
- / ASLB Memorandum and Order, March 15, 1982, pg. 30.
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s l f f inadequate materials and equipment selection procedures that do not appear to have followed ALARA considerations in certain j cases; inadequate procedures to control the water chemistry to minimize crud buildup; inadequate procedures to monitor and control occupational radiation doses; and inadequate means to provide action if in-p'lant radiation levels ~ l significantly exceed U.S. plant averages. This testimony discusses the importance of maintaining occupational radiation exposures to the ALARA standard, the elements of the ALARA program at Shoreham, the specific inadequacies listed above, and recomendations relative to LILC0's program. These recomendations i include implementing or improving procedures, design features, and programs to minimize occupational radiation exposure at Shoreham. 4 Exhibits
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Regulatory Guide 8.8, "Information Relevant to Ensuring That j Occupational Radiation Exposures at Nuclear Power Stations Will l Be As Low As Is Reasonably Achievable", Revision 3, June 1978. 4 1 2. Final Safety Analysis Report, Figures 12.3.1-16, 5.1.3-1, and 12.3.1-18. 3. Final Safety Analysis Report, Table 12.4.3-1, Volume 12. l 4. NUREG-0713, " Occupational Radiation Exposure at 2mmercial Nuclear Power Reactors 1980", Volume 2, December 1981, cover page and page 18. 5. " Compilation and Analysis of Data on Occupational Radiation Exposure Experiences at Operating Nuclear Power Plants", SAI Services, September 1974, cover page and page 12. 6. Radiation Buildup Effects, by Dr. William R. Dehollander, March 1980, cover page and page 24. 7. Radiation Buildup Effects, by Dr. William R. Dehollander, March 1920, cover page and page 13. 8. NUREG-0855, " Health Physics Appraisal Prograa:', March 1982, cover page and abstract.
- / ASLB Memorandum and Order, March 15, 1982, p. 30.
UNITED STATES OF AMERICA NUCLEAR REGULATORI COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ) In the Matter of ) LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 0.L. (ShorehamNuclearPowerStation, Unit 1) ) ) ) PREPARED DIRECT TESTIMONY OF MARC W. GOLDSMITH, DALE G. BRIDENBAUGH AND VERA BARLIT ON BEHALF 0F SUFFOLK COUNTY REGARDING SUFFOLK COUNTY CONTENTION 26 - ALARA I. INTRODUCTION This testimony was prepared by Marc W. Goldsmith, Dale G. Bridenbaugh,andVeraBarlit.1/ A statement of our qualifications and i experience has been Asoarately provided to this Board. 1/ The primary / secondary authors of each section of this testimony are as follows: Sections I and II -- M. W. Goldsmith /V. Barlit Section III and IV -- M. W. Goldsmith /D. G. Bridenbaugh/V. Barlit M. W. Goldsmith was responsible for overall coordination.
f 1 Testimony on S.C. Contention 26: ALARA Q. Would you please state the contention on which you are testifying? A. The Suffolk County Contention 26 reads as follows: Suffolk County contends that LILC0 has not adequately demonstrated that Shoreham meets the requirements of 10 CFR 20.1(c), with regard to provisions for maintaining occupational radiation exposure as low as is reasonable achieveable (ALARA). Demonstration of compliance is inadequate as follows: (a) Plant and equipment design has not been shown to be optimally developed for minimization of radiation exposure during maintenance of the plant by: (1) Selection of low cobalt materials; (ii) Separation or isolation of various components and piping systems; (iii) Provisions for flushing or decontamination; (iv) Equipment layout and arrangement for ease and automation of maintenance and refueling; and (v) Condenser design utilizing the minimum number of shell connections. (b) Procedures have not been developed and emplaced to provide: (i) Limitation of iron-cobalt buildup in the primary system through water chemistry control; (ii) Monitoring and control of individual and plant total annual occupational radiation doses, with invididual exceedance of three rem per quarter and five rem per year occurring only on an emergency basis and requiring special management approval; and 2
l r (iii) Acticn to reduce radiation levels and/or exposure if the in-plant totals significantly exceed U.S. plant i I averages. Q. What is the purpose of your testimony? i A. The purpose of this testimony is to discuss the importance of l maintaining occupational radiation exposures to the as low as l reasonably achievable' standard (ALARA). Thus the testimony discusses ALARA, elements of the ALARA program at Shoreham, and specific i concerns and recommendations with respect to LILC0's program. ( Specific concerns discussed are: f i t o design of shielding and cubicles for radioactive maintenance; l o selection of equipment and materials; l o procedures for water chemistry, decontamination and, personnel doses; and, [ o the ALARA monitoring program. I. Introduction - The Importance of ALARA l l l Q. Why is it important to maintain occupational radiation exposures as low as is reasonably achievable? l 4 i l A. It is generally accepted by the scientific community that any i exposure to radiation is cumulative and ultimately may cause harmful j biological effects. Immediate or prompt effects are very unlikely since large exposures would normally occur only if there were a i radiation accident. However, a significant and widely recognized concern is the delayed incidence of cancer. The chance of delayed cancer is believed to depend on how much radiation exposure a person { receives. Therefore, every reasonable effort should be made to keep l exposures low and therefore to lessen the probability of concerns. l The regulations specifically recognize this concern by making ALARA part of the regulations in 10 CFR 20.l(c). I i L 3 =.
f Q. How are workers exposed to radiation in nuclear powar plants. A. Radiation exposure can result from external sources or from ingestion of materials (absorbed or inhaled) that are radioactive. Radiation doses to personnel in nuclear power stations are predominantly from external exposure, i.e., from radiation received from sources external to the body, although careful control of ingestion, particularly during maintenance operation is essential. The intensity of the external radiation field is determined primarily by four factors: the quantity of radioactive material; the characteristics of the emitted radiation; shielding between the radiation source and the receptor; and, distances between source and receptor. Important parameters in determining doses from external exposures are: the length of time the receptor remains in the radiation field; and the intensity of the radiation field. Internal exposure to radiation results when radioactive materials are taken into the body by breathing, ingestion, or absorption through the skin. Radioactive materials usually concentrate in different organs in the body, resulting in a dose to a specific organ (i.e., radioactive iodine concentrates in the thyroid gland). Internal exposure is controlled by limiting the release of radioactive material into the air and by carefully monitoring the work area for airborne radioactivity and surface contamination. Q. How is exposure to radiation regulated? A. The Nuclear Regulatory Commission (NRC) regulates its licensee's to restrict an individual's exposure to within prescribed limits. The radiation dose limits for both the public and workers are set forth in the Code of Federal Regulations, Titt 10, Part 20 " Standards for i l I 4
~ ~ Pr:tection Against Radiation." The standards currently limit external occupational radiation dose to 1-1/4 rems / in any 2 calendar quarter or specified 3-month period for whole body exposure. Limits for internal exposure are set by calculating the quantity of radioactive material that has been taken into the body and the total organ dose that would result. Then, based on limits established for particular body organs similar to 1-1/4 rems in a calendar quarter for whole-body exposure, the regulations specify maximum permissible concentrations (MPC) of radioactive material in the air to which a worker can be exposed for 40 hours per week. Individual internal dosages are added to external exposure dosages. Exposure to radiation that results from radioactive materials taken into the body is measured, recorded, and reported to the worker separately from external dose. The internal dose to the whole body or to specific organs does not at this time count against the 3 rems per calencar quarter limit. Q. Are there regulations limiting collective doses to all station personnel? A. No. Upper limits are placed on individual doses, but not on collective doses. The collective occupational dose (man-rems) is the sum of all occupational radiation-exposures received by all the workers in an entire worker population. For example, if 100 workers each receive 2 rems, the individual worker dose is 2 rems and the collective dose is 200 man-rems. The total additional risk of cancer and genetic effects in an exposed population is assumed to depend on the collective dose. However, the NRC does not impose collective dose limits. 2/ The rem is the unit used for dose comparison and stands for roentgen ~ equivalent man. A rem is equal to the absorbed radiation dose multiplied by a factor that takes into account the way a particular radioactive particle distributes energy in tissue, thus including its effectiveness in causing harm. 5
l. l, l Q. How d:es NRC provide guidance on controlling collective dosts? A. In addition to providing an upper limit on an individual's permissible radiation exposure, the NRC also requires that nuclear plant operators maintain exposures as f ar below the limit as is reasonably achievable. This also means that every activity at a nuclear facility potentially involving exposure to radiation should be planned to minimize unnecessary exposure to individual workers and thus to the worker population. According to the NRC,M a job that involves exposure to radiation should be done only when it is clear that the benefit justifies the risks assumed. Section 20.1(c) of 10 CFR Part 20, " Standards for Protection Against Radiation," states that " persons engaged in activities under licenses issued by the NRC...[should]...make every reasonable effort to maintain radiation exposures as far below the limits specified in Part 20 as is reasonably achievable. The Series 8 Regulatory Guides provide specific guidance on the Staff's suggested means of meeting the Part 20 requirements. Thus, Regulatory Guide 8.8, "Informatidn Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As Low As is Reasonably Achievable" (Exhibit
- 1) provides information relevant to attaining goals and objectives for planning, designing, constructing, operating, and decommissioning to meet the criterion that exposures will be ALARA.
II. The Establishment of an ALARA Program Q. What are the goals of ALARA as set by NRC7 A. As stated in Regulatory Guide 8.8, 3/ Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As Low As is Reasonably Achievable." 6
"tha goals of the effert to maintain occupational radiation exposures ALARA are (1) to maintain the annual dose to individual station personnel as low as is reasonably achievable and (2) to keep the annual integrated (collective) dose to station personnel (i.e., the sum of annual doses (expressed in man-rems) to all station personnel) as low as is reasonably 4 achievable. The NRC staff believes that the stated objectives are attainable with current technology and with good operating practices. Q. How can these goals be met? A. According to the NRC Staff, as set forth in Regulatory Guide 8.8, these goals can be met by: I 1. Establishing a program to maintain radiation exposures ALARA. "To attain the integrated effort needed to kaep exposures of station personnel ALARA, each applicant and licensee should develop an ALARA program that reflects the efforts to be l taken by the utility, nuclear steam supply system vendor, and architect-engineer to maintain radiation exposure ALARA in all phases of a station's 1ife." 2. Properly designing facilities and selecting appropriate equipment. " Design concepts and station features should reflect consideration of the activities of Station personnel (such as maintenance, refueling, inservice inspections, processing of radioactivewastes, decontamination,anddecommissioning) that might be anticipated and that might lead to personnel exposure to substantial sources of radiation. Radiation protection aspects of decommissioning should be factored into planning, designing, construction, and modification activities. Station design features should be provided to reduce the anticipated exposures of station personnel to these sources of radiation to the extent practicable. 7
I t Specifications for cquipment sh3uld rcflect the objectives of the ALARA program, including considerations of reliability, serviceability, limitations of internal accumulations of radioactive material, and other features addressed in this guide. Specifications for replacement equipment also should reflect modifications based on experience gained from using the original equipment." 3. Establishing a radiation control program, plans, and procedures. "A substantial portion of the radiation dose to station personnel is received while they are performing services such as maintenance, refueling, and inspection in high radiation areas. Objectives discussed previously can provide station design features conducive to an effective program to maintain occupational radiation exposures ALARA. However, an effective program also requires station operational considerations in terms of procedures, job planning, recordkeeping, special equipment, operating philosophy, and other support." 4. Making support equipment, instrumentation and facilities available. "A radiation protection staff with f acilities, instrumentation, and protective equipment adequate to permit the staff to function efficiently is an important element in achieving an effective program to maintain occupational radiation exposures ALARA. The selection of instrumentation and other equipment and the quantities of such equipment provided for normal station operations should be adequate to meet the anticipated needs of the station during normal operations and during major outages that may require supplemental workers and extensive work in high radiation areas." 8
ALARA objcctives sh uld b3 considered throughout the planning, designing, constructing, operating, maintenance, and decommissioning of a LWR station to provide reasonable assurance that exposures of station personnel to radiation will be ALARA. Q. What are the goals of ALARA as specified by LILCO? A. According to LILCO, as set forth in the FSAR Chapter 12, the plant and equipment design criteria and the use of appropriate procedures ensure that Shoreham meets the requirments of 10 CFR 20.1(c) and maintains occupational radiation exposure ALARA. However, LILC0's assertion cannot be accepted out-of-hand because the FSAR discussion, entitled " Radiation Protection", Chapter 12, describes LILCO's program to meet ALAP (as low as practical), a concept that was superceded by ALARA in 1979. Thus as stated in the FSAR: It is the policy of the Long Island Lighting Company to keep occupational radiation exposure (ORE) as low as practicable (ALAP), consistent with plant construction, maintenance, and operation requirements, and within the applicable regulations. This ALAP policy applies to total man-rems accumulated by all personnel, as well as to individual exposures. LILC0 management will make every effort to provide the environment for this policy to function in a proper manner. Management's commitment to this policy is reflected in the design of the plant, the careful preparation of plant operating and maintenance procedures, the provisions for review of these procedures and for review of equipment design to feed back the results of operating experience, and most importantly, the establishment of an extensive, on-going training program. Sufficient training will be provided for all personnel so that each individual is capable of carrying out his responsibility for maintaining his own exposure ALAP consistent with discharging his duties and also for observing rules adopted for his own safety and that of others. The development of the proper attitudes and awareness of the potential problems in the area of health physics, by proper l training of all glant personnel, is the cornerstone of LILC0's policy._4./ 4/ FSAR, Volume 12, Section 12.1.1.1., pg. 12-1-1. l l 9
~ Q. D::s ALARA repr:stnt a strengthengd policy over thm ALAP? Please e explain. A. ALARA represents a strengthened policy over ALAP. ALARA represents an affirmation of the use of a quantification rationale (benefit-cost analysis) in regulation. ALARA, "...has an improved structure, (over ALAP) with numbers to refer to during the design process, limits for protection of the most-exposed individuals, and clear directions for measures to reduce the j population dose..." "ALARA is one of the most important of numerous design factors built into NRC regulations for licensing [ new f acilities and also has implications for 'back fitting' [ existing plants as well."5/ Q. Is there agreement on the ALARA goals? A. It appears that the goals set forth for ALARA are similar for all i parties. The area of disagreement relates to the implementation of the program to meet the ALARA goals. As discussed above the, multiple goals are interrelated and dependent on proper implementation to achieve the objective of low worker doses. As discussed in the subsequent sections, implementation of the program at Shoreham needs to be improved in several areas. III. Specific Concerns Relative to the ALARA Program Implementation at Shoreham Q. What are the concerns relative to maintaining Shoreham L i occupational radiation exposures ALARA? i S/ Considerations of Health Benefit-Cost Analysis for Activities Involving Ionizing Radiation Exposurs and Alternatives EPA i 520/4-77-003, 1977, p. 93. l 6 h I r 10
A. Ccnccrns arisa from inad;quate design featurcs and procedures at Shoreham to minimize radiation exposures during maintenance. Specifically, design features to separate and isolate equipment are inadequate; materials and equipment selection does not appear to have followed ALARA considerations in certain cases; procedures either do not exist or are inadequate to control the water chemistry at Shoreham to minimize crud buildup; procedures to monitor and control occupational radiation doses are not adequate, and there are inadequate means to take positive action if in-plant radiation levels significantly exceed U.S. plant averages is not present. Therefore,10CFR20.l(c)withrespecttoALARAisnot met at Shoreham. Q. What are some of the specific design concerns relative to maintaining exposures ALARA at Shoreham? A. The equipment configuration and layout associated with the Mark II containment at Shoreham results in a cramped and crowded configuration. Examples of problems created by this condition (1) the main steam isola, tion valves (with extremely are: difficult access in the pipe tunnel and dose contributions from the multiple valves and pipes); (2) the common location of the recirculation pumps, recirculation piping, and the transport carriers for control rod drives to the CRD maintenance area; and, (3) location of Safety Relief Valves, piping snubbers, main steam line nozzles, etc. high in the drywell. Access to this equipment is required during the refueling / maintenance outage and all such work is normally carried out simultaneously (or as near to simultaneoutly as can be achieved). Much of this work is often done while refueling is underway, and workers in the drywell, particularly the upper regions could receive non-ALARA or even overexposures from improperly handled or dropped fuel. The problems described above are apparent on the marked-up figares taken from the FSAR appended as Exhibit 2. l i 11 l l
Q. What is th2 reas:n for this problem? A. Plant arrangement was primarily dictated by economics of plant construction cost and that evaluation of ALARA aspects a decidedly secondary consideration. Use of the pressure suppression containment concept was successful in reducing containment size, and therefore, cost, but it resulted in minimum space for maintenance activities. Q. Do these space problems potentially cause greater radiation exposure? A. It would certainly appear so. The highest exposure tasks are the same ones that are the most closely crowded together. A copy of Table 12.4.3-1 from the FSAR is appended as Exhibit 3. It shows that the highest exposure rate jobs are expected to be MSIV maintenance, recirculation pump maintenance, ISI-drywell piping, and CRD work. These are precisely the tasks impacted by the cramped drywell, and their close proximities will compound the effect. Q. Could some of these deficiencies be corrected without a complete redesign of the plant? A. In some cases they could. For example, the use of additional block shield walls (either permanent or semi-permanent shielding) can be increased and such walls could be described in the radiation protection program. In the formal ALARA program review of equipment access including identification of additional catwalks and ladders that may be needed. Temporary shielding and necessary monorails and equipment handling hoists for maintenance l should be identified prior to operation of the plant and installed in the appropriate locations before the system is made radioactive. The ALARA program should include evaluation of such provisions. 12 i
Q. Why is equipment s21ecticn so important to the minimization of occupational radiation exposures? A. In operating nuclear power plants, it is not the operators - but the maintenance personnel - who are subjected to the highest levels of radiation exposure. Exhibit 4 from NUREG-0713,6/ provides a detailed sumary of the distribution of collective dose by work function and personnel types for BWRs, PWRs and all LWRs. I It shows'that contract workers performing special maintenance at LWRs incur the largest portion of the collective dose. At BWRs, { workers involved in these activities received 80.8% of the l cumulative dose for BWRs, an increase of about 10% from the 1980 value, while at PWRs these workers received 70.6% of the cumulative dose, an increase of only 3.6% over the 1980 value. The portions of the collective dose received by workers during inservice inspection and refu'eling at BWRs are 3.3% and 5.2% respectively; at PWRs such workers received 8.2% and 7.1%, respectively, of the collective dose. Overall, contractor personnelreceived68.4%ofthecollectivedose(about10%more than in 1980), and the station and utility employees received the remaining 31.6% at LWRs. t It is possible by careful equipment selection and attention to layout to reduce exposure received during maintenance. There are four ways in which this can be accomplished: 1. Reduce the amount or frequency of maintenance required; 2. Shorten the time spent working in a radioactive environment; 3. Reduce the radiation level in the work area; it includes, use i of shielding and reducing the coolant activity level and/or deposited activated corrosion produces; and 4. Reduce the number of workers required. i -6/ NUREG-0713, Occupational Radiation Expousre at Comercial Nuclear Power Reactors 1980, Volume 2, December 1981, p.18. 13
1 Q. Are there other areas wh:;ra shielding is inadequate for maintenance? A. It is difficult to make a definitive statement because the type of detailed analysis necessary to judge post-shutdown radiation levels has not been performed. The maintenance worker is subject to the radiation fields existing after shutdown during refueling outages when the majority of maintenance and inspections are performed. The shielding diagrams showing radiation levels (FSAR Section 12.3) show some limited post-shutdown levels. However, no derivation of the 1 post shutdown levels is presented. It is, therefore, difficult to determine the adequacy or accuracy of the post shutdown levels in order to place a relative importance of one action or activity over another. Specific concerns are pieces of equipment that are doubled or tripled up in cubicles. This has the potential to cause higher doses than if equipment was individually segregated. Specific equipment that could present problems are: the fuel pool clean-up pumps; pumps in the radwaste building; and, the reactor coolant water clean up heat exchangers. This equipment requires periodic maintenance and shared cubicle arrangement has the potential to complicate the maintenance process. This could add to worker exposure. Q. What is a specific concern relative to equipment design at Shoreham? AccordingtoanEPRIreport,U one source of radiation is A. radioactive crud which collects on the condenser surfaces. Much of this comes from the location of inlet connections on the condenser shell. Adding a dump tank to collect all lines before they enter the condenser would reduce this problem. It would also have distinct advantages: y Limiting Factor Analysis of High Availability Nuclear Plants (BWR), Volume 1, August 1979, pg. 2.2-68 1 14
1. It would rcduce the amount of radioactiva crud on the tube surf aces, thus reducing the field strength; 2. It would reduce the number of connections to the condenser shell and thus reduce the number of potential air leaks that would require location and repair and hence exposure to radioactivity (these leaks are largely at valves in the connecting lines); and, 3. It would reduce the number of tube failures due to either direct impingement or broken-off baffles. LILC0 states it has minimized the condenser shell connections.at Shorehsn.E/ However, LILC0 has not added a dump tank to collect these connections to provide the above advantages. Therefore, radiation dosages wil: not be minimized during condenser maintenance operations and additional maintenance may be required. Q. Why is it important to select materials which would minimize occupational radiation exposures? A. Erosion and corrosion of materials particularly in the primary system are large contributors to the radiation fields in the plant. Selection of materials with good wear and corrosion resistance would help minimize this problem. In addition, certain materials have alloying elements that have long half-lifes when activated. These alloys should be avoided where possible to reduce the long half-life activation products. Q. What is the specific concern relative to the selection of materials at Shoreham? l A. Reg. Guide 8.8 suggests the use of materials bearing low nickel and low cobalt concentrations for primary coolant piping, vessel internal p/ Information received through the discovery process. l 15
i e surf aces, and oth:r comp:ntnts that are in contact with primary l coolant. This suggestion was made because materials containing I nickel or cobalt produce long active half-life corrosion products, l that accumulate in the system resulting in an increase in radiation. Information submitted to date by LILC0 does not indicate that LILC0 selected materials are low in nickel and cobalt. For example, NRC Information Request 331.1 states: "This section [FSAR Section 12.1] discusses theoretical approaches to the reduction of radioactive crud; however, except for flushing tap provisions, no other commitments are described. Expand this section to include a listing of specific techniques used to l minimizeradiatJonfromactivatedcorrosion products..."9 The response provided by LILC0 to the NRC request is incorporated in Section 12.1.3.1.2 of the FSAR. A review of that Section of the FSAR indicates only that 1) piping system geometry, 2) elimination of crud pockets, 3) avoidance of low flow velocities, and 4) utilization of full flow condensate demineralization are utilized to minimize the buildup of cobalt 60. No indication is given demonstrating that any attempt was made to specify low cobalt or low nickel materials and a review of the piping material section in the FSAR (Section 5.2.3) gives no indication that any consideration was given to ALARA reconenendations in the selection of reactor coolant pressure boundary material. This section in fact indicates that stellite hard facing material is one of the materials of construction exposed to reactor coolant. Stellite is coninonly used for valve seats and discs and contains a high percentage of cobalt, which could potentially result in a high accumulation of corrosion products, causing an increase in plant radiation. In addition, the careful selection of materials would help to minimize maintenance resulting from corrosion failures such as pipe cracking. 9/ FSAR, Volume 15, p. 331-1. 16
l Q. What are tha sp;cific concerns relative to the implementation of such i procedures at Shoreham? A. Water chemistry control procedures which would optimally minimize corrosion products at Shoreham have not been documented. As can be seen from Exhibit 5, annual exposures at BWRs were found to increase with plant ag2, increasing by a factor of two to three each year for the first three to five years. The observed increases in annual exposure are believed to have resulted from an increase in radiation fields during these years, which are due to the accumulation of activated corrosion products. Annual exposures at newer BWRs are expected to follow the same pattern. Reduction of radiation fields can be accomplished by reducing activated corrosion products. The radiation field buildup after initial operation at the BWR plants is caused when corrosion products generated in the primary coolant system are carried into the reactor and activated on the fuel surface. They are then deposited on the piping and components in the primary circuit. To reduce the external radiation fields, it is important that the mechanism of transport, activation, deposition, and release of corrosion products be thoroughly understood, and that effective countermeasures be taken. According to an EPRI Journal report,5 " plant monitoring has established Co-60 as the major contributor to BWR radiation fields." This isotope is produced by activation of natural cobalt (Co-59) which is released to the coolant through wear and corrosion of alloys within the system. Cobalt is a major material in wear-resistant alloys, and is present as a low-level impurity in numerous iron-nickel alloys. Presently, efforts are underway at EPRI to identify the significant sources of cobalt in BWR materials. Because it is impossible to eliminate cobalt in some materials and undesirable to do so in other materials, cobalt will remain in the R/"BWRRadiationControl",EPRIJournal, October 1980,pg.51-52. 17
j. system as a low-lcvel, inpurity. Thsrsfore, cobalt control techniques are being developed. With better identification of cobalt sources, corrosion products can be minimized by selecting alternative materials for replacement when needed. Iron corrosion products are believed to play a fundamental role in cobalt activation and transport. Iron is released from corroding surfaces and is transported through the feedwater system to the reactor. The control of iron is important because it is only via the medium of iron oxide that cobalt-59 (the natural isotope) can reside in the high neutron flux within the core long enough to become cobalt-60. Dissolved cobalt absorbs on iron oxide and is carried to the fuel surface. The oxide adheres to the heated zircalloy fuel rod, thus holding the cobalt-59 in the flux for a period of time sufficient to activate the cobalt to cobalt-60. According to a G.E. report on radiation field buildup, b ow long the iron remains on h the fuel and thus how much cobalt-60 is released to the water is a function of the quality of the iron oxide. J According the G.E. Report it is believed that very small particle size iron oxide (colloidal, less than 0.2 microns) adheres firmly to the fuel and virtually never leaves the surface again. Thus, the cobalt-60 can be high but it never leaves the fuel; so it cannot deposit on the piping. Middle size iron oxide particles (0.2 - 1 micron) are small enough to adhere to the fuel rod but large enough so the bonds are weak. After some residence time on the rod, the particles can be released again into the water, carrying the cobalt-60, and thus making cobalt-60 available to the piping. Therefore, this middle size iron oxide should be controlled. Larger size iron oxide particles (greater than 1 micron) are usually erosion generated and are generally harmless with respect to radiation buildup as they usually remain as sediment. lj/ " Radiation Effects of Water Quality", by Dr. William R. Dehollander of G.E., March 1980. 18
Q. How can the activaticn of corrosion products be minimized. A. According to the G.E. report it is believed that the variable oxygen content in the feedwater assists in the release of middle size iron oxide from the feedwater piping. Therefore, a controlled oxygen level should be attained to minimize middle size iron oxide particles. The most desirable constant value is specific reactor system dependent and should be determined based on an analysis of iron oxide released as a function of the controlled oxygen level. However, results have shr)wn that corrosion was minimized between a bond of 20 PPB to 200 PPB. The rate of corrosion of stainless steel, is a linear function with conductivity (see Exhibit 6). Therefore, measures should be taken to keep the conductivity of the reactor water low. Q. Have any nuclear plants beer successful in controlling oxygen, iron and conductivity levels? A. Countemeasures against corrosion were successfully carried out at the Tsuruga Nuclear Power Station in Japan. The Tsuruga plant has the largest operating experience in Japan - comercial operation started in March 1970. In 1973, a steep increase in the iron concentration due to an increase of corrosion product generation in the feedwater system was experienced. A comprehensive program review of plant chemistry was conducted at Tsuruga and countermeasures against corrosion were taken. Since Tsuruga's sixth annual inspection, the rate of iron feed from the feedwater has been reduced at Tsuruga by the following measures: 1. the control of oxygen dissolved in feedwater ( 60 PPB to minimizecorrosionproducts). 2. flushing of feedwater system before startup (to keep the conductivity low); 19
3. improvement cf p;rformance of condensate demin:ralizers (resin exchange to keep the conductivity low); 4. hot draining off of condensate water and feedwater during long outages (to keep the conductivity low). The following results were obtained since the countermeasures were taken: 1. radiation field strength resulting from accumulated corrosion products dropped by a factor of two in two years; (from 250 mr/h to 150mr/h,seeExhibit7); 2. iron input to the feedwater dropped by a factor of ten resulting in a decroase of corrosion products; 3. soluble and insoluble iron particles were reduced resulting in a decrease or corrosion products; and 4. Cobalt-60 on the inside of the primary system piping was reduced. Q. Could a similar water chemistry control program be implemented at Shoreham to minimize corrosion buildup? A. Yes, preventive measures such as those taken at the Tsuruga plant could be taken at Shoreham. Procedures for' preventing corrosion during long outages could be implemented along with procedures to flush the feedwater system before startup and measures could be taken to control the oxygen dissolved in the feedwater within the 20-200 PPB range. Q. What procedures are involved in having an adequate radiation monitoring program? A. Station technical and supervisory personnel, working closely with radiation protection personnel, can reduce exposure. Procedures to reduce exposures include planning activities of personnel who must enter radiation areas, by studying the actions and procedures of individuals working in such areas, and by conducting postoperation l l 20
d; briefings en projects rssulting in substantial exposures to identify how procedures might be modified to reduce exposures on I subsequent similar tasks. Training programs for all station f personnel can establish and reinforce the principles of radiation protection as applied to specific job functions. By making personnel aware of the methods and the special equipment and protective equipment available to them, potential radiation doses can be reduced. I Q. What are the specific concerns relative to the radiation monitoring [ program at Shoreham? A. It does not appear that LILC0 management and organization is fully committed to a meaningful ALARA program. This is apparent from LILCO's response to NRC Request 331.31 contained in the Final Safety f Analysis Report. It admits that Shoreham's radiation protection { program does not fully comply with Regulatory Guide 8.8, Revision 3, [ June 1978.El A significant deficiency is the f ailure of the Shoreham radiation protection program to include a specific measurement sytem. Such a program is very important. In a recent i report documenting an appraisal of the 1 t water reactor health physics programs throughout the country,- the NRC reported on i deficiencies in ALARA programs they had reviewed at existing plants. 4 Included in the deficiencies cited was the following: "There were no apparent measurable goals set for the ALARA effort; there.ias no management system developed that would indicate the degree of success of ALARA thatis,ifthegoalhasbeenachieved."_gfortundertaken, l R/FinalSafetyAnalysisReport,p.331-31, Revision 18, June 1980. l 13/ Exhibit 8, NUREG-0855, Health Physics Appraisal Program, March 1982. 14/ NUREG-0855, p. 19. l l l i 21
1 i In tha sugg:sted actions for improving health physics programs, the NRC stated that one of the most frequent omissions in audit programs is performance audits. They found that most radiation protection programs include functional audits which determine 1 whether selected activities are performed and if they are performed at the proper frequency, but assessment of the effectiveness of the program is frequently left out of the audit. l This noted deficiency is related specifically to Reg. Guide 8.8 C.l.b (1)(b) and (c). These provisions call for that ensuring effective measurement systems are established in the radiation protection program and that the measurement system results are reviewed on a periodic basis and corrective actions taken. LILCO's f ailure to incorporate this critical step in the radiation protection program at Shoreham is a significant deficiency and one that should be corrected. Q. Does it appear LILC0 is correcting this deficiency (lack of measurement program) in the ALARA procedures? A. Yes, there is evidence of some additional work underway. We have recently obtained a copy of a draft procedure entitled "ALARA Goals and Measurement". This draft has not been approved and is dated March 26, 1982. It does specify that ALARA goals shall be set by the LILC0 Section Heads. Q. If this procedure is adopted, will that correct the lack of measurement deficiency? A. No it will not. The draft procedure requires only that goals be set and records kept and reviewed. It lacks definition of action steps. It even shows that acceptance criteria are not applicable. Q. Does this lack of action criteria represent a significant deficiency? 22
l A. Yes it do;;s. Tha apparent reluctance of LILC0 to idsntify action level guidelines but instead to apparently rely on the Review Operations Committee for corrective action does not signify a formal comitment to the program. This lack of comitment was also a significant finding in the health physics appraisal program documented in NUREG-0855. That appraisal identified lack of management support as a significant weakness that existed at approximatelyone-thirdofthefacilitiesinspected.El The lack of management support was reflected in several ways as reported by the NRC. At some facilities, the radiation protection manager did not report directly to the station manager and this apparently contributed to causing the quality of radiation protection to be significantly less at some plants. Q. Does the radiation protection manager report directly to the station manager at the Shoreham Plant? A. LILCO's response to question 331.31 in the FSAR indicates that the responsibility for health physics rests with the HP engineer and HP supervisor. It goes on to state however that the HP engineer "has direct recourse to th"e plant manager" which indicates to me that this is an optional path rather than a direct reporting function. This indirect reporting path is confirmcd in Section 13 of the FSAR. Q. What other deviations from Regulatory Guide 8.8, Rev. 3 exist? A. As reported by LILCO's response to Request 331.31, health physics technicians are presently assigned to work only on the day shift from Monday thru Friday. This does not comply with RG 8.8, Rev. 3,B.(1). The Reg. Guide recommends that health physics _1_5/ NUREG-0855, p. 5. 5 I 23
protection sh:uld be provided for each shift cperating crew. This incidently was discovered to be a common problem at approximately one-third of the U.S. operating plants as reported in NUREG-0855. [ That report recomended that adequate staff be provided so that the assigned responsibilities can be accomplished. We certainly agree that HP coverage should be present at all shifts unless it is known that no unusual situation may arise that no maintenance work is underway. Q. What would satisfy the concerns you have expressed relative to maintaining occupational radiation as low as is reasonably achievable at Shoreham? A. The following are the minimum steps needed for maintaining occupational radiation as low as is reasonably achievable at Shoreham: 1. Increase the use of shield walls in areas previously discussed; 2. Document that a thorough study of equipment access will be conducted; 3. Identify temporary shielding, monorails, and equipment handling hoists for maintenance purposes prior to operation; 4. Document procedures to reduce cobalt bearing materials during normal equipment replacement programs. 5. Install a dump tank connecting all the lines going into the condenser; 6. Implement water chemistry control procedures similar to the Tsuruga plant. Procedures for preventing corrosion during long outages could be implemented along with procedures to flush the feedwater system before startup and measures could be taken to control the oxygen disolved in tne feedwater within 20-200 PPB range. ? 24
7. Imprava th3 radiation protection prcgram at Shareham by: a. including a specific measurement system into the program. b. ensuring that effective measurement systems being established and the measurement system results are reviewed on a periodic basis and corrective actions taken; c. committing to setting specific numerical goals -(level guidelines)bymanagement; d. assigning health physics technicians to each shift operating crew. 8. It appears that the future elements of the radiation protection program are beginning to be developed for Shoreham. Since they are in such a preliminary state, we recommend a detailed review by experienced outside consultants after the program is nearly completed. 25
l EXHIBIT 1 Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" Revision 3, June 1978. 4
e Revision 3 /p "'eg*. U.S. NUCLEAR REGULATORY COMMISSION June 1978 %) REGULATORYGU DE OFFICE OF STANDARDS DEVELOPMENT .g l ( REGULATORY GUIDE 8.8 INFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT NUCLEAR POWEft STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE A. INTRODUCTION high dose rates.2 An ad hoc committee of the Na-tional Council on Radiation Protection and Meas-Paragraph 20.l(c) of 10 CFR Part 20, " Standards urements (NCRP) (Ref.1) chose in 1959 to make the for Protection Against Radiation." states that licen-cautious assumptions that a proportional relationship aces should make every reasonable effort to maintain ex sts between dose and biological effects and that exposures to radiation as far below the limits speci' the effect is not dependent on dose rate. Essentially, fled in Part 20 as is reasonably achievable. This this amounts to assumptions of a nonthreshold, guide provides information relevant to attaining goals " linear" (straight line) dose-effect relationship. and objectives for plannmg, designing, constructing, e l operating, and decommissioning a light-water reactor The International Commission on Radiological (LWR) nuclear power station to meet the criterion Protection ICRP), the Federal Radiation Council that exposures of station personnel' to radiation dur-(FRC) whose functions now reside in the Environ-ing routine operation of the station will be "as low as mental Protection Agency (EPA), and committees of is reasonably achievable" (ALARA). This guide is the National Academy of Sciences / National Research l also responsive to the admonition of the Federal Council (NAS/NRC) have used this hypothesis to es-Radiation Council (now EPA) that occupational radi-timate conservatively the number of possible biologi-i ation exposures be maintained ALARA. Major' acci-cal effects that statistically may be associated with l dent situations and emergency precedures are not exposures to radiation. within the scope of this guide. The NAS/NRC Biological Effects of lon..izmg l Much of the information presented in this guide Radiation (BEIR) Committee (Ref. 2) reiterated that also is applicable to nuclear power stations other than the assumptions of a nonthreshold linear relationship l those cooled with light water. The applicable goals between dose and biological effects independent of and objectives should be used for all nuclear power the dose rate should be applied for radiation protec-I stations until more specific goals and objectives are tion purposes. This recommendation has been available for other types of power reactors. adopted by EPA (41 FR 28409) for the purpose of estimating the potential human health impact of low levels of ionizing radiation. The radiation protection
- 8. DISCUSSION goal s to reduce doses wherever and whenever rea-The relationship between radiation dose and sonably echievable, thereby reducing the risk that is biological effects is reasonably well known only for assumed (for radiation protection purposes) to be doses that are high compared with current annual Proportional to the dose.
dose limits and only when such deses are delivered at in 1973, the ICRP (Ref. 3) stated: a 1.ines indicate substantive changes from previous issue. "Whilst the values proposed for maximum permis- ' " Station personnel," as used in this guide, includes all per-sees working at the station, whether full. time or part. time and 8 Throughout this guide the word " dose" will allude to " dose whether employed by the licensee or by a contractor for the equivalent," the term used for radiation protection purposes, i
- licensee, with the unit espressed in " rems."
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l l 1 o sible doses are such as to involve a risk which is doses among broad job categories and among the small compared to the other hazards of life, equipment systems or components that represent sub-l nevertheless, in view of the incomplete evidence stantial sources of exposures. Doses to station per-on which the values are based, coupled with the sonnel are influenced by many variables, including knowledge that certain radiation effects are irrever-the ability of fuel elements to retain fission products,
- f sible and cumulative, it is strongly recommended the extent of deposition of activated corrosion prod-l that every effort be made to reduce exposure to all ucts throughout the primary and auxiliary coolant sys-types of ionizing radiation to the lowest possible tems, the reliability of other specific equipment, the level."
station layout, and radiation protection programs. l Merely controlling the maximum dose to individu-If design reviews or inspections had revealed that als is not sufficient; the collective dose to the group radiation exposures at nuclear power stations were (measured in man-rems) also must be kept as low as unavoidable or that the cost of reducing the expqsures la reasonably achievable. " Reasonably achievable" would be unreasonable, the exposures might be con-is judged by considering the state of technology and sidered ALARA by definition. However, this has not the economics of improvements in relation to all always been the case, and this guide is intended to the benefits from these improvements. (However, a assist in achieving a status wherein exposures are comprehensive consideration of risks and benefits considered to be ALARA. will include risks from nonradiological hazards. An A majw patbn of the radiation exposure of sta-action taken to reduce radiation risks should not re-tion pusonnel is aceived during maintenance, rad-j suit in a significantly larger risk from other hazards.) waste handh,ng, m, service inspection, refueling, and Under the linear nonthreshold concept, restricting nonroutine operations (Ref. 6). The decommissioning the doses to individuals at a fraction of the applicable process also has a potential for substantial exposures j limit would be inappropriate if such action would re-to personnel. Effective design of facilities and selec-l sult in the exposure of more persons to radiation and tion of equipment for systems that contain, collect, would increase the total man rem dose. The radiation store, process, or transport radioactive material in l protection 3 community has recognized for many any form will contribute to the effort to maintain I years that it is prudent to avoid unnecessary exposure radiation doses to station personnel ALARA. to radiation and to maintain doses ALARA. In addi-Products of erosion or corrosion (.i.e., " crud,,*) tion to reduced biological risks, the benefits of such that become mobile and are activated constitute an practices may include avoidance of costs for extra imPortant (perhaps principal) source of radiation with per:;onnel to perform maintenance activities and respect to the exposure of station personnel. (Crud,s l i avoidance of nonproductive station shutdown time accumulated in and transported by the coolant. Some 3 caused by restrictions on station personnel working in C mPonents of the crud become radioactive when radiation areas. passing through the reactor core. Migration of crud to Annual collective radiation dose equivalents % other systems occurs with coolant or steam. Specific ceived by personnel working at an LWR nyM radionuclides that have been identified in crud and power station have ranged from less than W mu that can contribute substantially to the radiation i rems to over 5,000 man-rems (Refs. 4 ud % rw source are Co-58, Co-60, Mn-54, Zn-65, and Zr-95.) a. cally, annual collective dose equivalem: n e 1 DPosures,ot station personnel who serv. ice equip-400 to 1,000 man rems at LWR statiora that tiave ment contammated by crud can generally be reduced j been in operation from 2 to 14 years and have substantially by mmimizmg the formation of crud and generating capacities ranging from less than 100 by designing or modifying equipment to minimize to-MWe to 800 MWe. In view of the anticipated growth gtsons where crud can deposit and accumulate. Pro-of nuclear power stations over the next few decades visions fw isolating components and flushing with and the radiation exposure experience to date, addi-crud removing tid such as demineralized water can j tional efforts to reduce radiation doses to nuclear often reduce accurelations prior to activities such as j power station personnel are warramed. maintepance or equipment replacement. The wide range in collective radiation doses to sta-Stati n and equipment layout also can affect the tion personnel among the various stations appears to Potential for radiation exposures. Exposures at wes be primarily a function of doses received in mainte. where multiple radiation sources exist sometimes can j nance operations in radiation areas. Some data are be reduced by additional separation of individual available to permit estimates of the distribution of sources. Adequate space for case of maintenance and other operations can perrnit the tasks to be completed
- ne seen "rediation protection," as used in ibis seide is con-sidered to be sysosynous with the term " applied health more quickly, thereby reducing the length of expo.
physics";i.e., the development and implemenestion of methods sad procedures necessary to evaluate radiation hazards and to
- " Crud"is corrosion and erosion products and other solids that provide protection to man and his environment from unwar.
are formed by chemical and physical reaction between the reac-rested esposere. tot coolant and structural materials. ) 8.8 2 1
l o sures. Shielding by structural materials, equipment, ing at the judgment, but it should not be the decisive and auxiliary or permanent shields can reduce expo-factor in all cases. sures by isolating radiation sources. Where equip-The nuclear steam supply system (NSSS) vendor, ment components constitute a substantial radiation the desi ner, the architect.cngineer (A/E), the con-8 source that cannot be effectively reduced m place, stmetor, and the operator of the nuclear power facil-features that permit the removal of such components sty each have responsibilities related to the effort of for maintenance at remote locations often can be ef-maintaining occupational radiation exposures fective in reducing exposures. The use of remote-ALARA. Rus, coordination and cooperation are es-handling features also can reduce exposures of station sential to achieving th,ese goals an,d objectives of personnel in certain instances. maintaining occupational radiation exposures Station technical and supervisory personnel, work-ALARA. Ing closely with radiation protection personnel, can nis guide is written primarily for the applicant or reduce exposures by planning activities of personnel licensee. However, the designer, the A/E, and the who must enter radiation areas, by studying the ac-constructor will find many of the guide's consid-tions and procedures of individuals working in such erations helpful in the design and construction proc-areas, and.by conducting postoperation debriefings ess to ensure that their efforts are consistent with the on projects resulting in substantial exposures to iden-needs of the applicant or licensee to maintain radia-tify how procedures might be modified to reduce ex-tion exposures ALARA. posures on subsequent similar tasks. Training pro-grams for all station personnel can establish and rein-Specific design or operational objectives for main-force the principles of radiation protection as applied taining ratiation exposures ALARA are suggested by to specific job functions. By making personnel aware the parameters that determine the magnitude of doses of the methods and the special equipment and protec-to station personnel, both as individuals and as a tive equipment available to them, potential radiation group. Doses to personnel in nuclear power stations doses can be reduced, are predominantly from external exposure, i.e., from radiation sources external to the body. However. The concept of maintaining occupat.ional radiation there also exists a potential for doses from internal CKPosures ALARA does not embody a specific num-exposures, i.e., from radioactive materials taken into erical guideline value at the present time. Rather, it is se body. a philosophy that reflects specific objectives for radi-ation dose management in: Important parameters in determining doses from extemal exposms are (1) the length of time that the
- 1. Establishing a program to maintain occupational receptor remains in the rautation field and (2) the in-radiation exposures ALARA; tensity of the radiation field. Some degree of expo.
- 2. Designing facilities and selecting equipment; sm of station pusonnel cannot be avoided during
- 3. Establishing a radiation control program, plans, the operation and maintenance of nuclear power sta-and procedures; and tions. However, there are many ways by which the
- 4. Making supporting equipment, instrumentation, exposures and resultant doses can be lowered by re-and facilities available, ducing the time mterval of the exposure and the in-When an adequate data base, including economic tensity of the radiation field. The intensity of the information, is available, the criteria for keeping an-radiation field is determined by (1) the quantity of nual collective doses to station personnel ALARA radioactive material, (2) the nature (i.e., characteris-might be derived or selected in numerical terms.
tics) of the emitted radiation, (3) the nature of the However, a data base of operating experience and shielding between the radiation source and the re-cost information to provide quantitative guidance for ceptor, and (4) geometry (e.g., distances and establishing such criteria is not available at this time, dimensions). and the criteria for meeting the provision of para-i graph 20.l(c) of 10 CFR Part 20 must therefore take Parameters important in determining doses from the form of qualitative guidance (e.g., goals, objec. internal exposures are (1) the quantity of radioactive material taken into the body, (2) the nature (isotopi-tives, and statements of good practice). cal and body deposition characteristics) of the material, De NRC staff has not performed a cost-benefit and (3) the time interval over which the material f analysis for each of the considerations discussed or is retained by the body. The principal modes by presented in Section C of this guide. This guide pre-which radioactive material can be taken into the body sents goals and objectives that were selected to are (1) inhalation (2) ingestion (3) skin absorption, satisfy the principles, philosophy, and criteria for and (4) injection through wounds. At nuclear power maintaining occupational radiation exposures stations, radioactive materials are generally confined. ALARA. Attaining these goals and objectives will but some dispersion within the station is unavoidable require good engineering judgment on a case-by-case and constitutes the source of (1) contaminated air and basis. A cost-benefit analysis may be helpful in arriv-liquids that present the potential for intake by inhala-8.8 3
~ s tion and absorption and (2) contaminated surfaces Attaining the following objectives to the extent that present the potential for intake by ingestion and practicable throughout the planning, designing, con-through cuts or abrasions in the skin. Absorption structing, operating, maintenance, and decommis-generally is not an important intake mode at nuclear sioning of an LWR station will be considered to pro-power stations except for tritium, which can be ab-vide reasonable assurance that exposures of station i sorbed through the skin. personnel to radiation will be ALARA. The methods i are deliberately stated such that considerable flexibil-Consequently, the basic variables that can be con-I '*" ". sed in the manner by which the objectives J trolled to limit doses from internal exposures are can be achieved. Differences among stations might those that limit (1) the amount of contamination (2) necessitate further innovation in methods used to I the dispersal of the contamination, and (3) the length 5******' } of time that personnel must spend in contaminated areas. Protective equipment can keep the intake of
- 1. Program for Maintaining Station Personnel 4
the contaminant to a minimum. Physical and chemi-Radiation Doses ALARA j cal methods can be used to hasten the elimination of To attain the integrated effort needed to keep expo-l radioactive material taken into the body; however, sures of station personnel ALARA. each applicant g because of the risks associated with the use of these and licensee should develop an ALARA program that methods, they are reserved for very serious cases reflects the efforts to be taken by the utility, nuclear where the probability of experiencing biological ef-steam supply system vendor, and architect-engineer fects is quite substantial, e.g., large mtakes such as to maintain radiation exposure ALARA in all phases those that might occur in serious accident situations. or,,g,gjon.s life. This program should be in written Objectives stated in this guide for maintaining occu-form and should contain sections that cover the gen. i pational radiation exposures ALARA are derived by erally applicable guidance presented in this guide, as j considering the parameters that affect dose, the vari-a minimum, and more specific guidance as required ables that exist in the station design features, and the to address the particular LWR that is the. subject of variables that can be provided by station administra. the licensing action. This program may be combined tive actions. Section C Regulatory Position, states with the station's radiation protection manual, safety j objectives in a manner that encourages innovation by analysis report, or other documents or submittals. It j permitting considerable flexibility on the part of the need not be an independent document. ] utility, the NSSS vendor, the designer, the construe-
- a. Establishment of a Program To Maintain Oc-tor, and the A/E. However the regulatory position cupational Radiation Doses ALARA l
also describes a large number of specific concerns that should be addressed in meeting the goals and ob-(1) A management policy for, and commitment { jectives. to, ensuring that the exposure of station personnel to radiation will be ALARA should be established. a i C. REGULATORY POSITION (2) The policy and commitment should be re-flected in written administrative procedures and m-The goals of the effort to maintain occupational structions for operatior.s involving potential expo-radiation exposures ALARA are (1) to maintain the sures of personnel to cadiation and should be re-annual dose to individual station personnel as low as flected in station design features. Instructions to de-is reasonably achievable and (2) to keep the annual signers. constructors, vendors and station personnel l integrated (collective) dose'to station personnel (i.e., specifying or reviewing station features. systems, or i the sum of annual doses (expressed in man-rems) equipment should reflect the goals and objectives to j to all station personnel) as low as is reasonably maintain occupational radiation exposures ALARA. i achievable. (Few utilities design or build their nuclear power sta- } The NRC staff believes that the stated objectives tions; but as customers of designers and builders, utilities should expect the designers and builders to j are attainable with current technology and with good j operstmg practices. The costs for attaining thesa ob-be respoasive to their needs and instructions.) i jectives have not been established and are expected to
- b. Organization, Personnel, and Responsibilities vary widely depending on the features of the specific power reactor facility and the method selected to ac-(1) In view of the need for upper-level manage-complish the objectives. The favorable cost-benefit ment, support, responsibility and authority for imple-ratio for achieving some of these objectives may be meming the program to maintain occupational radia-l obvious without a detailed study. For other objec.
tion exposures ALARA should be assigned to an m. tives, however, a cost benefit study might be re-dividual (or committee) with organizational freedom "5ute development and implementation. Respon-I? *ilities and authorities should include: quire.d to determine whether the objectives are res-sib. sonably achievable. Doses to station personnel can affect station availability, and this factor should be (a) Ensuring that a corporate program that in-g' considered in assessing the cost-benefit ratio. tegrates management philosophy and regulatory re-I 8.8-4 i
e I quirements is established, with specific goals and ob-(3) The Radiation Protection Manager (RPM) l jectives for implementation included; (onsite) has a safety fur.ction and responsibility to both employees and management that can be best ful-(b) Ensuring that an effective measurement filled if the individual is independent of station divi. system is established and used to determine the de-si ns, such as operations, maintenance, or technical gree of success achieved by station operations with support, whose prime responsibility is continuity or i . regard to the program goals and specific objectives; improvement of station operability. The RPM should (c) Ensuring that the measurement system re-have direct recourse to responsible management per-j - suits are reviewed on a periodic basis and that correc-sonnel in order to resolve questions related to the tive actions are taken when attainment of the specific conduct of the radiation protection program. objectives appears to be jeopardized; (The specific responsibilities given here for (d) Ensuring that the authority for providin3 the RPM are illustrative and not intended to be all-procedures and practices by which the specific goals inclusive with respect to the ALARA program or ef-and objectives will be achieved is delegated; and fort. They do not include any of the responsibilities (c) Ensuring that the resources needed to in areas other than ALARA efforts.) achieve goals and objectives to maintin occupational radiation exposures ALARA are made available. Responsibilities of the RPM with respect to a program to maintain occupational radiation exposures In view of the responsibilities required to 1m-ALARA should include: plement a program to maintain occupational radiation exposures ALARA, the individual (or committee) (a) Participating in design reviews for selected for this function might also be chosen to facilities and equipment that can affect potential radi-coordinate the effort among the several corporate ation exposures; functional groups (such as the operations, mainte-3 nance, technical support, engineering, safety, and (b) Identifying locations, operations. and con-radiation protection groups) and to represent the cor. ditions that have the potential for causing significant porate interests in dealing with the NSSS designer, exposures to radiation; vendor. A/E, and builder during the design and con-(c) Initiating and implementing an exposure struction phases. If the expertise for perforJntng this control program; function is not within the corporation when the sta- ) tion is in the design stage, consultants who possess (d) Developing plans, procedures, and 4 I the required expertise should be used. The utility methods for keeping radiation exposures of station l should obtain assurance that available data and ex-personnel ALARA; perience obtained from similar nuclear power stations (e) Reviewing, commenting on, and recom- + are considered and reflected in the work of the NSSS mending changes,n job procedures to maintain espo-i designer, vendor A/E, and builder so as to provide sures ALARA: features in the new statim that permit an effective j ALARA program. (f) Participating in the development and ap-Proval of training programs related to work in radia-(2) The Plant Manager (Superintendent or tion areas or involving radioactive materials; a j equivalent) is responsible for all aspects of station operation, including the onsite radiation protection (g) Supervising the radiation surveillance pro-1 program. gram to maintain data on exposures of and doses to station personnel, by specific job functions and type Responsibilities of the Plant Manager with re-spect to a program to maintain occupational radiation exposures ALARA should include: (h) Supervising the collection, analysis, and evaluati n f data and inf rmati n attained from I (a) Ensuring support from all station person-radiological surveys and monitoring activities;, g . (i) Supervising, training, and qualifying the (b) Participating in the selection of specific radiation protection staff of the station; and goals and objectives for the station; (j) Ensuring that adequate radiation protection (c) Supporting the onsite Radiation Protection e verage is provided for station personnel dun,ng all Manager (RPM) in formulating and implementing a w rking hours. station program in maintaining occupational radiation exposures ALARA; and (d) Expediting the collection and dissemina-
- Dais coti.cted during omisses can indienie trends or radiation tion of data and information concernm, g the program buildup in equipment ihai can permit estimates of probable radi-to the corporate management.
anion levels to be encountered during subsequent outases. 8.8-5
Qualifications
- needed for the RPM job, as keep it current. Station personnel whose duties do not well as those needed for other positions in organiza-require entering radiation areas or working with tions operstmg nuclear power stations, are presented radioactive materials should receive sufficient in-in Regulatory Guide 1.8, " Personnel Selection and struction in radiation protection and station rules and Training."
regulations to understand why they should not enter such areas.
- c. Training and Instruction Training programs that have as their goal an in-A training program in the fundamentals of radia.
crease in craft skills provide a broader base of knowl-tion protection and in station exposure control proce. edgeable station personnel available to service dures should be established. It should include in, equipment in radiation areas and permit the services structing all personnel whose duties require (1) work-to be performed more reliably and more efficient 1y. This can promote lower individual and collective ing with radioactive materials, (2) entering radiation dose levens. areas, or (3) directing the activities of others who work with radioactive materials or enter radiation
- d. Review of New or Modified Designs and i
areas. The training nreram also should include suf-Equipment Selection ficient instruction ir the: biological effects of expo-sures to radiation io permit the individuals receiving (1) Since several groups within a utility (e.g., the instruction to understand and evaluate the signifi. maintenance, operations, radiation protection, tech-cance of radiation doses in terms of the potential nical support, engineering, and safety groups) are in-terested in station desigri and equipment selection, risks. the utility should ensure that these groups are The training should be commensurate with the adequately represented in the review of the design of duties and responsibilities of those receiving the in-the facility and the 5:lection of equipment. A coordi-structions, as well as with the magnitude of the po-nated effort by the several functional groups within tential doses and dose rates that can be anticipated. the utility is required to ensure that station features Personnel (including contractor personnel) who direct w 11 permit the goals and objectives of the ALARA the activities of others should be familiar with the program to be achieved. Although the A/E and desig-licensee's radiation control program and should have ners greatly influence station design features, utilities the authority to implement the licensee's commitment should not delegate all responsibilities for station de-to ensure the radiation exposures of station personnel sign review and equipment selection to the NSSS de-will be ALARA. signer, vendor, or A/E. The training program should include instruction (2) Design concepts and station features should on (1) radiation protection rules for the station and reflect consideration of the activities of station per-(2) the applicable Federal regulations. Copies of sonnel (such as maintenance, refueling, inservice in-these rules and regulations should be made available spections, processing of radioactive wastes, decon-to those receiving the instructions. The training pto-tamination, and decommissioning) that might be an-gram should be approved by the RPM and presented ticipated and that might lead to personnel exposure to by competent instructors. The information presented substantial sources of radiation. Radiation protection in the training program should be reviewed periodi-aspects of decommissioning should be factored into cally and modified, where necessary, to reflect con-planning, designing, construction, and modification temporary techmques and adjustments based on ex-activities. Station design features should be provided perience in station operations. Instruction of station to reduce the anticipated exposures of station person-personnel should stress the importance of exposure-nel to these sources of radiation to the extent practic-reduction efforts by every individual and should em-able. phasize the need for feedback of information obtained (3) S ecifications for equipment should reflect P when similar tasks were performed previously. the objectives of the ALARA program, including l i Stats personnel should receive instruction at considerations of reliability, serviceability, limitations periodic intervals to reinforce their knowledge and of nternal accumulations of radioactive material, and other features addressed in this guide. Specifications for rep acement equipment also should reflect mod-l
- consideration hu been given to peer group certification,i.e.,
certification of health physicists by the Arnetican Board of ifications based on experience gained from using the Health Physics (ABHP), as representing evidence of adequate original equipment. qualifications for ILPM candidates. While the staff believes that peer group tertification is desirable, the present ABHP certifica.
- 2. Facility and Equipment Design Features tion is not necessarily specifically applicable to applied health physics or.3distion protection needs in nuclear power stations.
Radiation sources within a nuclear power station However, the staff is discussing with the ABHP the prospects differ appreciably with respect to location. intensity, D for a special certification progreen specifically directed toward and characteristics. The magnitude of the dose rates the needs of radiation protection personnel at nuclear power that results from these sources is dependent on many stations. I 8.8 6 l
factors, including the facility and equipment design, areas and control over the movement of sources of layout, mode and length of operation, and radiation radiation within the station. Where high radiation source strength and characteristics. areas (>100 mrem /h) exist, i 20.203 of 10 CFR Part i 20 requires that station design features and adminis-To provide a basis for design, the quantity and trative c ntr 15 provide effective ingress control, ease isotopic composition of the radioactive material that f egress, and appropriate warning devices and can be anticipated to be contained, deposited, or ac-n tices. Access control of radiation areas also should cumulated in the station equipment should be esti-re lect the following considerations: mated. Fission product source terms should be esti-mated using these bases: (1) an offgas rate of (1) Extraordinary design features are warranted 100,000 gCi/see after 30 minutes delay for BWRs to avoid any potential dose to personnel that is large and (2) 0.25% fuel cladding defects for PWRs. Acti-enough to cause acute biological effects and that vation source terms, including activated corrosion could be received in a short period of time. Positive products, should be based on measurements and ex-control of ingress to such areas, permanent shielding, perience gained from operating stations of similar de-source removal, or combinations of these alternatives t sign. ANSI N237-1976 (Ref. 7) is based on such ex-can reduce the dose potential. i perience and provides information that can be used as (2) Administrative controls such as standard a basis for estimating activation source terms. WN.a operating measurements are used, extrapolation of Operating procedures can be effective in preventing data to equilibrium conditions may be needed to esti-inadvertent exposures of personnel and the spread of contamination when radioactive material or contami-mate ultimate activation source terms. Neutron and prompt gamma source terms should be based on ap-nated equipment must be transported from one station location to another and when the route of transport plicable operating experience and reactor core physics calculations. through lower radiation zones or " clean" areas can-not be avoided. ALARA program objectives are preser ted below for each of several station features or functions. Each (3) Station features such as platforms or walk-statement of objective is followed by a number of ways, stairs, or ladders that permit prompt accessibil-specific concerns or suggestions that should be ad. ity for servicing or inspection of components located dressed. in higher radiation zones can reduce exposure of per-sonnel who must perform these services.
- a. Access Control of Radiation Areas
- b. Radiation Shields and Geometry To avoid unnecessary and inadvertent exposures of personnel to radiation, the magnitude of the poten-Radiation shields should be designed using the tial dose rates at all locations within the station design basis 2ssumptions explained in regulatory po-should be estimated during station design. Actual sition 2 and conservative assumptions for geometries.
dose rates should be measured periodically during Calculational methods known to provide reliable and operation to determine current exposure potentials.
- accurate results (i.e., methods and modeling tech-Zones associated with the higher dose rates should be niques that have been demonstrated to give accept-3 kept as small as reasonably achievable consistent able accuracy in analyses similar to the problem of with accessibility for accomplishing the services that concern) should be used to determine appropriate must be performed in those zones, including equip-shield thicknesses. Shield design features should re-ment laydown requirements. Radiation zones where flect the following considerations to maintain occupa-station personnel spend substantial time should be de-tional radiation exposures ALARA:
signed to the lowest practical dose rates. (1) Exposure of personnel servicing a specif.ic (It is common practice to identify " radiation component (such as a pump, filter, or valve) to radia-zones" within a nuclear power station. The zone des-tion from other components containing radioactive ignations are established to reflect the design material can be reduced by providing shielding be-maximum dose rates that may exist in areas within tween the individual components that constitute sub-the station where station personnel must have access stantial radiation sources and the receptor. to perform required services. Several systems for de-(2) Where it is impracticable to provide perma-signating, ' radiation zones,, currently exist among nent shielding for individual components that consti-the utihties, and ANSI Committee 6.7,s developing tute substantial radiation sources, the exposure of i a standard that should prove useful in attaining com-Personnel maintaining such components can be re-mon designations and terminology in this matter. To duced (a) by providing as much distance as practica-avoid ambiguity, no reference to radiation zone num-ble between the serviceable components and the sub-bers is made in this guide at this time.) stantial radiation sources in the area and (b) by pro-A system should be established to permit effec. viding temporary shields around components that tive control over personnel access to the radiation contribute substantially to the dose rate. 8.8 7
= o (3) Potential exposure of station personnel to (10) Floor and equipment drains, piping, and radiation from certain systems containing radiation sumps that are provided to collect and route any con-sources can be reduced by means of a station layout taminated liquids that might leak or be spilled from that permits the use of distance and shielding between process equipment or sampling stations can become j the sources and work locations. These systems in-substantial radiation sources. The drain lines can be clude (but are not limited to) the NSSS and the reac-located in concrete floors, concrete ducts, columns, for water cleanup, offgas treatment, solid waste or radwaste pipe chases to provide shielding. These treatment, and storage systems, as well as systems systems can also become a source of airborne con-infrequently containing radiation sources such as the tamination because of the potential for gases to form standby gas treatment and residual heat removal sys-in, and be released by, such systems (see regulatory tems. position 2.d(6)). Radiation from an operating BWR turbine can
- c. Process Instrumentation and Controls constitute a substantial source of exposure for con-Appropriate station layout and design features struction personnel or others who have access to the site for extended periods of time if insufficient shield.
should be provided to reduce the potential doses to ing is provided. Personnel who must operate, service, or inspect sta-tion instrumentation and controls. The following con-(4) Streaming or scattering of radiation from lo-siderations should be reflected in selecting the station cally shielded components (such as cubicles) can be features: reduced by providing labyrinths for access. However, such labyrinths or other design features of the cubicle (1) The exposure of personnel who must manu-should permit the components to be removed readily ally operate valves or controls can be reduced from the cubicle for repair or replacement where such through the use of " reach rods" or remotely operated valves or controls. However, these devices c:n re-work is expected or anticipated. Single scatter labyrinths may be inadequate if the cubicle contains a quire lubncation and maintenance that can be the substantial radiation source. source of additional exposures, and these factors should be taken into consideration. (5) Streaming of radiation into accessible areas through peneirations for pipes, ducts, and other (2) The exposure of personnel who must view or shield discontinuities can be reduced (a) by means of Operate instrumentation, monitors, and controls can layouts that prevent substantial radiation sources be reduced by locating the readouts or control points in low radiation zones. within the shictd from being aligned with the penetra-tions or (b) by using " shadow" shields such as (3) Instrumentation must satisfy functional re-shields of limited size that attenuate the direct quirements, but the exposure of personnel can be re-radiation component. Streaming also can occur duced if the instruments are designed, selected. spec-through roofs or floors unless adequate shielding en-ified, and located with consideration for long service closes the source from all directions. life, ease and low frequency of maintenance and (6) The exposure of station personnel to radia-calibration, and low crud accumulation. Operating tion from pipes carrying radioactive material can be experience should be recorded, evaluated, and re-reduced by means of shielded chases. flected m the selection of replacement instrumenta-tion. (7) Design features that permit the rapid removal and reassembly of shielding. insulation, and other (4) The use of instrumentation that contains min-material from equipment that must be inspected or imal quantities of contaminated working fluid (e.g.. serviced periodically can reduce the exposure of sta-Pressure transducers rather than bellows-type pres-tion personnel performing these activities. sure gauges) can reduce the potential for exposure at the readout locations. (8) Space within cubicles and other shielding to provide laydown space for special tools and ease of
- d. Control of Airborne Contaminants and Gase-servicing activities can reduce potential doses by ous Radiation Sources permitting the services to be accomplished expediti-Station design features should be provided in all ously, hus reducing exposure time, g
g (9) The exposure of personnel who service com-of radioactive material in air to levels well below the ponents that constitute substantial radiation sources values listed in Appendix B. Table 1, Column I of 10 or are located in high radiation fields can be CFR Part 20. Effective design features can minimize minimized by removing the components and trans-the occurrence of occasional increases in air contami-porting them to low radiation zones where shielding nation and the concentrations and amounts of contam-and special tools are available. Design features that inants associated with any such occasional increases. permit the prompt removal and installation of these Designs that permit repeated, identified releases of C components can reduce the exposure time. large amounts of radioactive materials into the air 8.8-8
o spaces occupied by personnel are contrary to a pro-trunks" without imbalancing the ventilation system. gram to maintain occupational radiation exposures in areas where contaminated equipment must be A1. ARA.. opened infrequently, portable auxiliary ventilation systems featuring blowers, HEPA @ters, and acti-Station design features should provide for pro-vated charcoal filters (where radioiodine might be an-tection against airborne rad.ioactive material by ticipated) on carts can be used effectively. Portable means of engineering controls such as process, con-auxiliary ventilation systems should be tested fre-tamment, and ventilation equipment. The routine quently to verify the efficiency of the filter elements provision of respiratory protection by use of indi-in their mountings. When the efficiency has been ver-vidually worn respirators rather than engmeered de-ified, the system may be exhausted to the room or the sign features is generally unacceptable. The use of ventilation exhaust duct without further treatment and respirators, however, might be appropriate in certain thus imbalance of the permanent ventilation system nonroutine or emergency operations when the app! - an be avoided. cation of engineering controls is not feasible or while such controls are being installed. (4) Machining of contaminated surfaces (e.g..
- elding, grinding, sanding, or scaling) or " plug-The approved use of respirators is subject to the ging, of leaking steam generator or condenser tubes requirements of i 20.103, " Exposure of Individuals can be substantial sources of airborne con amination.
to Concentrations of Radioactive Matenals in Air in These sources can be controlled by using auxiliary Restricted Areas,, of 10 CFR Part 20 and to regula-y,,,;;,,;,, ,y,,,,,, tory guidance on acceptable use. (See Regulatory Guide 8.15. " Acceptable Programs for Respiratory (5) Sampling stations for primary coolant or Protection," and NUREG-0041, " Manual of Res-other fluids containing high levels of radioactive ma-piratory Protection Against Airborne Radioactive Ma-terial can constitute substantial sources of airborne terials" (Ref. 8).) Design features of the station venti-contamination. Such sources can be controlled by lation system and gaseous radwaste processing sys-using auxiliary ventilation systems, tems should reflect the following considerations: (6) Wet transfer or storage of potentially con-(1) The spread of airborne contamination within taminated components will minimize air contamina-tion. This can be accomplished by keeping contami-the station can be limited by maintaining air pressure nated surfaces wet, by spraying, or, preferably, by gradients and airflows from areas of low potential airborne contamination to areas of highe'r potential keeping such surfaces under water. contamination. Periodic checks would ensure that the design pressure differentials are being maintained.
- e. Crud Control (2) Effectively designed ventilation systems and Design features of the primary coolant system, gaseous radwaste treatment systems will contain the selection of construction materials that will be in radioactive material that has been deposited, col-contact with the primary coolant, and features of lected, stored, or transported within or by the sys-equipment that treat primary coolant should reflect tems. Exposures of station personnel to radiation and considerations that will reduce the production and ac-to contamination from ventilation or gaseous rad-cumulation of crud in stations where it can cause high waste treatment components occur as a result of the exposure levels. The following items should be con-need to service, test, inspect, decontaminate, and re-sidered in the crud control effort:
place components of the systems or perform other (1) Production of Co-58 and Co-60, which con-duties near these systems. Potential doses from these stitute substantial radiation sources in crud, can be systems can be minimized by providing ready access seduced by specifying, to the extent practicable, low. to the systems, by providing space to permit the ac-nickel and low-cobalt bearing materials for pn, mary tivities to be accomplished expeditiously, by separat-coolant pipe, tubing, vessel internal surfaces, heat ex-ing filter banks and components to reduce exposures changers, wear matenals, and other components that to radiation from adjacent banks and components, are in contact with primary coolant. Alternative mate-and by providing sufficient space to accomodate aux-nals for hard facings of wear materials of high-cobalt iliary ventilation or shielding of components, content should be considered where it is shown that (3) Auxiliary ventilation systems that augment these highWt materials contribute to the overall ex-the permanent system can provide local control of posure levels. Such consideranon should also take into airborne contaminants when equipment containing account potential increased service / repair require-potential airborne sources is opened to the atmos-ments and overall reliability of the new material in phere. Two types of auxiliary ventilation systems relation to the old. Alternative materials for high. have proved to be effective. In areas where contami-nickel alloy materials (e.g., Inconel 600) should be noted equipment must be opened frequently, dampers considered where it is shown that these materials con-and fittings can be provided in ventilation ducts to tribute to overall exposure levels. Such consideration permit the attachment of flexible tubing or " elephant should also take into account potential increased 8.8-9
i = service / repair requirements and overall reliability of limit the spread of contamination from leakage of I the new materials in relation to the old. liquid systems. I j (2) Loss of material by erosion of load bearing (3) Accumulations of crud or other radioactive j i hard facings can be reduced by using favorable material that cannot be avoided within components or j geometrics and lubricants, where practicable, and by systeris can be reduced by providing features that using controlled leakage purge across journal sleeves w 11 permit the recirculation or flushing of fluids with to avoid entry of particles into the primary coolant. the capacity to remove the radioactive material l (3) Loss of material by corrosion can be reduced through chemical or physical action. The fluids con-by continuously monitoring and adjusting oxygen taining the contaminants will require treatment, and this source should be considered in sizing station I concentration and pH in primary coolant above 250*F and by using bright hydrogen-annealed tubing radwaste treatment systems. and piping in the primary coolant and feedwater sys-(4) Continuity in the functioning of processing or ventilation systems that are important for control-j (4) Consideration should be given to cleanup ling potential doses to station personnel can be pro-systems (e.g., using graphite or magnetic filters) for vided during servicing of the systems if redundant removal of crud from the primary coolant during op-components or systems are available so that the com-j eration. ponent (with associated piping) being serviced can be isolated' 1 (5) Deposition of crud within the primary cool-ant system can be reduced by providing laminar flow (5) The potential for contamination of " clean 3 and smooth surfaces for coolant and by minimizing services" (such as station service air, nitrogen, or j crud traps in the system to the extent practicable. water supply) from leakage from adjacent systems containing contaminants can be reduced by separating i
- f. Isolation and Decontamination piping for these services from piping that contains radioactive sources. Piping that carries radioactive I
Potential doses to station personnel who must sources can be designed for the lifetime of the sta-I service equipment containing radioactive sources can tion, thus avoiding the necessity for replacement (and be reduced by removing such sources from the attendant exposures) and lessening the potential for equipment (decontamination), to the extent practica-containjnation of clean services if it is impracticable ble, prior to servicing. Serviceable systems and com-to provide isolation through separate chases. ponents that constitute a substantial radiation source should be designed, to the extent practicable, with (6) Surfaces can be decontaminated more ex-features that permit isolation and decontamination. peditiously if they are smooth, nonporous, and free Station design features should consider, to the extent of cracks, crevices, and sharp corners. These desira-j practicable, the ultimate decommissioning of the ble features can be realized by specifying app,oriate facility and the following concerns: design instructions, by giving attention to finishing (1) The necessity for decontamination can be re-work during construction or manufacture, and by duced by limiting, to the extent practicable, the de-using sealers (such as special paints) on surfaces position of radioactive material within the processing where contamination can be anticipated. (ANSI i equipment-particularly in the " dead spaces" or N101.2 provides helpful, guidance on this matter " traps" in components where substantial accumula-(Ref. 9).) tions can occur. The deposition of radioactive mate-4 I rial in piping can be reduced and decontamination ef-(7) Where successful decontamination of impor-I forts enhanced by avoiding stagnant legs, by locating tant systems could be prevented by an anticipated failure of a critical component or feature, additional ) connections above the pipe centerline, by using slop-ing rather than horizontal runs, and by providing features that permit alternative decontamination ac. i drains at low points in the system. tions can be provided. l (2) The need to decontaminate equipment and (8) Contaminated water and deposited residues j station areas can be reduced by taking measures that in spent fuel storage pools contribute to the exposure will reduce the probability of release, reduce the at accessible locations in the area. Treatment systems amount released, and reduce the spread of the con-that remove contaminants from the water can perform taminant from the source (e.g., from systems or more efficiently (a) if intake and discharge points for components that must be opened for service or re-the treatment systems are located to ptovide enhanced j placement). Such measures can include auxiliary ven-mixing and to avoid stagnation areas in the pool and l tilation systems (see regulatory position 4.b), treat-(b) if pool water overflows and skimmer tanks are j ment of the exhaust from vents and overflows (see provided. Fluid jet or vacuum-cleaner. type agitators f' regulatory position 2.h(8)), drainage control such as can help reduce the settling of crud on surfaces of the curbing and floors sloping to local drains, or sumps to pool system. 8.8-10
a
- g. Radiation Monitoring Systems (a) Using full-ported valves constructed such that the slurry will not interfere with the opening or r " built in" monitoring systems tha' closing of the valve and Central give information on the dose rate and concentratior.
of airborne radioactive material in selected station (b) Avoiding cavities in valves. areas can reduce the exposure of station personnel (3) The deposition of resin and sludge that would who would be required to enter the areas to obtain the data if such systems were not provided. These sys-occur if elbow fittings were used can be reduced by using pipe bends of at least five pipe diameters in tems also can provide timely information regarding radius. Where pipe bends cannot be used, long radius changes in the dose rate or concentrations of airborne radioactive material in the areas. (The instaliation of elbows are preferred. a central monitoring system is easier and less expen-(4) Smoother interior pipe surfaces at connec-sive if it is a part of the original station design.) The tions (with attendant reductions in friction losses, de-selection or design and installation of a central pcsition of material, and tendencies to " plug") can monitoring system should include consideration of be achieved by using butt welds rather than socket the following desirable features: welds and by using consumable inserts rather than backing rings. (1) Readout capability at the main radiation pro-tection access control point; (5) Where the use of tees cannot be avoided, line I 55*s can be reduced if the flow is through the run (2) Placement of detectors for optimum coverage (straight section) of the tee, and accumulations of ma-of areas Sef.10); terial in the branch of the tee can be reduced by (3) Circuitry that indicates component failure; orienting the branch horizontally or (preferably) (4) Local alarm and readout; .(6) Slurry piping is subject to plugging that may (5) Clear and unambiguous readout; require backflushing from the tank and equipment iso-(6) Ranges adequate to ensure readout of the lation valves and pressurizing with water, nitrogen, or highest anticipated radiation levels and to ensure air to " blow out" plugged lines. However, the use of positive readout at the lowest anticipated levels; and pressurized gas for blowing out lines can present a po-tential contamination source and may not be effective (7) Capability to record the readout of all sys- ,n relieving plugged I;nes. tems. (7) Water, air, or nitrogen for sparging can be used to fluidize resins or sludges in storage tanks, The
- h. Resin and Sludge Treatment Systems use of gases, however, presents a potential source of
~. Systems used to transport, store, or process re. airborne contamination and tank rupture from over-sins or slurries of filter sludge present a special Pressures. hazard because of the concentrated nature of the (8) The spread of contamination by the loss of radioactive material. Design features for resin-and resin or sludge through overflows and vents can be re-sludge. handling systems should reflect this :oncern duced by using screens, filters, or other features that and the following specific considerations: will collect and retain solids. However, such features (1) The accumulation of radioactive material in generally require cleaning by remote flushing, by rapid components of systems used to process resin and replacement, or by other means to reduce exposures sludges can be reduced by: during servicing. (a) Reducing the length of piping runs; Consideration should be given to ANS N197 " Design and Performance of BWR Liquid Radioactive (b) Using larger diameter piping (to minimize Waste Processing Systems (N18)" (Ref. I!); ANS plugging); 55.1, " Design Criteria for the Solid Radwaste Proc-(c) Reducing the number of pipe fittings; essing System f BWR, PWR, and HTGR"(Ref.12); and ANS N199, "PWR Liquid Waste System Design (d) Avoiding low points and dead legs in pip-(N18)" (Ref.13). These standards cover 'ome as-ing; pects of slurry systems. (e) Using gravitational flow to the extent prac-t ca and I. Other Features (f) Minimizing flow restrictions of processed Stat on layout and station tasks should be re-
- I*I" viewed to identify and provide special features that (2) The need for maintenance and the presence complement the ALARA program. Station design of intense local radiation sources can be reduced by:
should reflect consideration of the following concerns: 8.8-11
a (1) The selection of radiation-damage-resistant (9) The sources of radiation such as sedimentation materials for use in high radiation areas can reduce the that occurs in tanks used to process liquids containing need for frequent replacement and can reduce the radioactive material and residual liquids can be re-probability of contamination from leakage. duced when servicing by draining the tanks. The de- / sign can include sloping the tank bottoms toward out-(2) The use of stainless steel for constructing or lets leading to other reprocessing equipment and, lining components, where it is compatible with the cre Practicable, providing built-in spray or surge process, can reduce corrosion and can provide options ~ for decontamination methods. (10) Spare connections on tanks or other compo-(3) Field-run piping that carries radioactive mate-nents located in higher radiation zones may be desira-rial can cause unnecessary exposures unless due con-ble to provide flexibility in operations. Exposures of sideration is given to 6e routing. Such unnecessary personnel can be avoided if these connections are pro-exposures can be avoided if the routing is accom-vided as a part of the original equipment rather than by l plished under the cognizance of an mdividual famihar subsequent modification of the equipment in the pres-with the principles of radiation protection or if,a de-ence of radiation, tailed pipeng layout is provided, i.e., if the piping is not field-run. (11) Inspections to satisfy the ASME Code (Ref.
- 14) and regulatory requirements can result in expo-(4) Where filters or other serviceable compo' sures of station personnel to radiation. Many of the nents can constitute substantial radiation sources, ex-objectives presented above will aid in reducing poten-posures can be reduced by providing features that tial exposures to personnel who perform the required permit operators to avoid the direct radiation beam inspections. Station features and design should, to the and that provide remote removal, installation, or ser-extent practicable, permit inspections to be accom-vicing. Standardization of filters should be consid-plished expeditiously and with minimal exposure of cred.
personnel. The effort to maintain occupational radia-tion exposures ALARA can also be aided by prompt (5) The servicing of valves can be a substantial accessibility, shielding and insulation that can be source of doses to station personnel. These doses can quickly removed and remstalled, and special tools and be reduced by providing adequate working space for instruments that reduce exp* sure time or permit re-o easy accessibility and by locating the valves in areas mote inspection of components or equipment contain-that are not in high radiation ficids. mg potential radiation sources. (6) Leakage of contaminated coolant from the (12) Components can be removed from process-primary system can be reduced by using h,ve loaded ng systems more expeditiously if adequate space is valve packings and bellow seals. provided in the layout of the system and if the inter. c nnecti n5 Permit Prompt disconnects. (7) Potential doses from servicing valves and from leakage can be reduced by specifying and instal-(13) Station features that provide a favorable ling reliable valves for the required service, by using working environment sue.h as adequate lighting, venti-radiation-damage-resistant seals and gaskets, and by lation, working space, and accessibility (via such i using valve back seats. The use of straight-through means as working platforms, cat walks, and fixed lad-valve configurations ua..wi the buildup of accumu-ders) can promote work efficiency. h d mh pd
- m f
lations in internal crevices and the discontinuities that rep ace lamps. Poin high radiation areas can be reduce exist m valves of other configurations. In most cases, l valves can be installed in the " stem-up" orientation by using extended service lamps and by providing de-to facilitate maintenance and to minimize crud traps. I The desired features are reliability, good perform-sign features that permit the servicing of the lamps from lower radiation areas. ance, and the ability to be maintained infrequently and rapidly. (15) An adequate emergency lighting system can reduce potential exposures of station personnel by (8) Leaks from pumps can be reduced by using can-Permitting prompt egress from high radiation areas if ned pumps where they are compatible with the service the station lighting system fails. needs, provided that lower personnel exposures can be l achieved thereby. If mechanical seals are used on a pump in a slurry service, features that permit the use of flush water to clean pump seals can reduce the ac-A substantial portion of the radiation dose to station cumulation of radioactive material in the seals. Drains personnel is received while they are performing serv-on pump housings can reduce the radiation field from ices such as maintenance, refueling, and inspection in this source during servicing. Provision for the collec-high radiation areas. The objectives that were pre-y tion of such leakage or disposal to a drain sump is sented in regulatory position 2 can provide station de-i appropriate. sign features conducive to an effective program to i 8.8 12
1 e maintain occupational radiation exposures ALARA. needed to perform the required services in the radia-However, an effective program also requires station tion areas. Such a program would address conditions operational considerations in terms of procedures, job that require a special work permit or other special pro-pianning, recordkeeping, special equipment, operating cedures. philosophy, and other support. His section deals with the manner in which the station administrative efforts (8) A work permit form with an appropriate for-can influence the variables of (1) the number of per-mat can be useful for recording pertinent information sons who must enter high radiation areas or contami-concerning tasks to be performed in high radiation L nated areas, (2) the period of time the persons must areas. so that the information is amenable to cross-remain in these areas, and (3) the magnitude of the referencing and statistical analysis. Information of i l potential doso. interest would include the following items:
- a. Preparation and Planning (a) Designation of services to be performed on specific components, equipment, or systems; i
Before entering radiation areas where sigmficant doses could be received, station personnel should have (b) Number and identification of' personnel the benefit of preparations and plans that can ensure working on the tasks; the exposures are ALARA while the personnel are per-forming the services. Preparations and plaas should re-(c) Anticipated radiation, airborne radioactive material, and contaminatior levels, based on current ficct the following considerations: surveys of the work areas, and date of survey; (1) A staff member who is a specialist in radiation protection can be assigned the responsibility foi con-(d) Monitoring requirements such as continuous tributmg to and coordinating ALARA efforts in sup-air monitoring or sampling equipment; port of operations that could result in substantial indi-(e) Estimated exposure time required to com-vidual and collective dose levels. plete the tasks and the estimated doses anticipated (2) To provide the bases for planning the activity, fr m the exposure; surveys can be performed to ascertain information with (f) Special instructions and equipment to respect to radiation, contamination, airborne radioac-minimize the exposures of personnel to radiation and tive material, and mechanical difficulties that might be contamination; nountered while performing services. (g) Protective cloth.ing and equipment require-(3) Radiation surveys pru.ided in conjunction ments; with inspections or other activities can define the na-ture of the radiation fields and identify favorable loca-(h) Personnel dosimetry requirements; tions where personnel may take advantage of available (i) Authorization to perform the tasks; and t shielding, distance, geometry, and other factors that affect the magnitude of the dose rate or the portions of (j) Actual exposure time, doses, and other in-the body exposed to the radiation. formation obtained during the operation. (4) Photographs of "as installed" equipment or (9) Consideration of potential accident situations components can be valuable for planning purposes and or unusual occurrences (such as gross contamination can be augmented by additional photos taken during leakage, pressure surges, fires, cuts, punctures, or the surveys. The use of portable TV cameras with tap-wounds) and contingency planning can reduce the po-ing features has considerable merit as both an opera-tential for such occurrences and enhance the capability tional aid and a teaching aid. for coping with the situations expeditiously if they oc- '(5) ne existing radiation levels frequently can be reduced by draining, flushing, or other decontamina-(10) Portable or temporary shielding can reduce tion methods or by removing and transporting the dose rate levels near " hot spots" and in the general component to a lower radiation zone. An estimate of area where the work is to be performed. the potential doses to station personnel expected to re-sult from these procedures is germane in selecting (II) Portable or temporary ventilation systems or among ahernative actions. contamination enclosures and expendable floor cover-(6) A preoperational briefing for personnel who ings can control the spread of contamination and limit will perform services in a high radiation area can en-the intake by workers through inhalation. sure that service personnel understand the tasks about to be performed, the information to be disseminated, (12),' Dry runs,, on mockup equipment can be and the sPecial instructions to be presented. U5eful f r tram 8n8 Personnel, identifying problems that can be encountered m the actual task situation, and (7) A program can be implemented to provide ac-selecting and qualifying special tools and procedures cess control and to limit exposures to those persons to reduce potential exposures of station personnel. 8.8 13
^ o j (13) Adequate auxiliary lighting and a comforta-
- c. Postoperations ble environment (e.g., vortex tube coolers for supplied Observations, experience, and data obtained dur-air suits) can increase the efficiency of the work and ing nonroutine operations in high radiation zones N-thus reduce the time spent in the higher radiation should be ascertained, recorded, and analyzed to iden-zones.
tify deficiencies in the program and to provide the (14) Radiation monitoring instruments selected bases for revising procedures, modifying features, or and made available in adequate quantities can permit making other adjustments that may reduce exposures accurate measurements and rapid evaluations of the during subsequent similar operations. radiation and contamination levels and changes in (1) Formal or informal postoperation debriefings leveht when they occur. Routine calibration of instnt-of station personnel performing the services can pro-ments with appropriate sources and testing can ensure vide valuable information concerning shortcomings in operability and accuracy of measurements. preoperational briefings, planning, procedures, special (15) Performing work on some components inside tools, and other factors that contributed to the cause of i disposable tents or, for less complicated jobs, inside doses received during the operation. i commercially available disposable clear plastic glove (2) Dose data obtained during or subsequent to an bags can limit the spread of contamination. Such operation can be recorded in a preselected manner as measures can also avoid unnecessary doses resulting part of a " Radiation Work Permit" or similar pro [ tram from the need to decontammate areas to permit per-(see regulatory position 3.a(8)] so that the data are sonnel access or to allow for entry with less restrictive amenable to statistical analyses. J protective clothing and equipment requirements. I I '" "* " # "##" E *
- "S*
- "I 5
(16) Careful scheduling of inspections and other nent failures that resulted in the need for servicing m tasks m. high radiation areas can reduce exposures by high radiation areas can provide a basis for revising permitting decay of radiation sources during the reac-specifications on replacement equipment or for other tor shutdown period and by chmmating some repeti-modifications that can improve the component reliabil-tive surveys. Data from surveys and experience at-ity. Such improvements can reduce the frequency of i tamed in previous operations and current survey data servicing and thus reduce attendant exposures. can be factored into the scheduling of specific tasks. (4) Information gained in operations can provide
- b. Operations a basis for modifying equipment selection and design features f new facilities.
During operations in radiation areas, adequate supervision and radiation protection surveillance (5) Summaries of doses received by each category should be provided to ensure that the appropriate pro-of maintenance activity can be reviewed periodically cedures are followed, that planned precautions are ob-by upper management to compare the incremental re-served, and that all potential radiation hazards that duction of doses with the cost of station modifications might develop or that might be recognized during the that could be made. operation are addressed in a timely and appropriate
- 4. Radiation Protection Facilities, Instrumentation, manner.
and Equipment (1) Assigning a health physics (i.e., radiation ( safety or radiation protection) technician the responsi. A radiation protection staff with facilities, in-j bility for providing radiation protection surveillance strumentation, and protecuve equipment adequate to for each shift operating crew can help ensure adequate Permit the staff to function efficiently is an important j radiation protection surveillance. element in achieving en effective program to maintain i occupational radiation exposures ALARA. The selec-(2) Personnel monitoring equipment such as tion of instrumentation and other equipment and the j direct-reading dosimeters, alarming dosimeters, and quantities of such equipment provided for normal sta-l personal dose rate meters can be used to provide early tion operations should be adequate to meet the antici-l evaluation of doses to individuals and the assignment pated needs of the station during normal operations of those doses to specific operations (see Regulatory and during major outages that may require supplemi.n-Guides 1.16. " Reporting of Operating Informa-tal workers and extensive work in high radiation areas. tion-Appendix A Technical Specifications," and (Accident situations are not considered in this guide.) 8.4, " Direct Reading and Indirect Reading Pocket Station design features and provisions should reflect the Dosimeters"). following considerations: j (3) Communication systems between personnel in
- a. Counting Room high radiation 4r.:ies and personnel who are momtoring the operation in other locations can permit timely ex-A low radiation background counting room is
'li changes of information and avoid unnecessary expo-needed to perform routine analyses on station samples f sures to monitoring personnel, containing radioactive material collected from air, wa-8.8-14 l l
4 I e ter, surfaces, and other sources. An adequately (6) Portal monitors. equipped counting room would include:
- d. Protective Equipment (1) Multichannel gamma pulse height analyzer Utility-supplied protective equipment selection (Regulatory Guide 5.9, " Specifications for Ge(Li) should include consideration of :
Spectroscopy Systems for Material Protection Measurementr,-Part 1: Data Acquisition Systems," (1) Anticontamination clothing and equipment j { provides guidance for selecting Ge(Li) spectroscopy that meet th: requirements of ANSI Z-88.2 (Ref.15) systems); for use in atmospheres containing radioactive mate-rials or the National Institute of Occupational Safety (2) Low-background alpha-beta radiation propor-and Health,s (NIOSH) " Certified Personal Protective tional counter (s) or scintillation counter (s); 1 Equipment List, and current supplements from i (3) End-window Geiger-Muller (G-M) counter (s); DHEW/PHS (Ref.16). and (2) Respiratory protective equipment, including (4) A liquid scintillation counter for tritium a respirator fitting program that satisfies the guidance ? analyses. Analyses of bioassay and environmental of Regulatory Guide 8.15 and NUREG-0041 (Ref. 8). t samples and whole-body counting (see Regulatory-
- S8PPort Facilities l
Guide 8.9, " Acceptable Concepts, Models, Equa-l tions, and Assumptions for a Bioassay Program") call Design features of radiation protection support for additional equipment and laboratory space if the facilities should include consideration of: f analyses are performed by station personnel tather than (1) A portable-instrument calibration area de- ) by other specialists through contractual arrangements. signed and located such that radiation in the calibra-
- b. Portable Instruments tion area will not interfere with low-level monitoring f
Portable instruments needed for measuring dose r counting systems; i 4 rates and radiation characteristics would include: (2) Personnel decontamination area (this facility should be located and designed to expedite rapid (1) Low-range (nominally 0 to 5 R per hour) ion CItanuP of Personnel and should not be used as a I chambers or G-M rate meters; multiple-purpose area or share ventilation with 3 (2) High-range (0.1 to at least 500 R per hour) ion food-handling areas) with showers, basins, and in-chambers;' stalled "frisker" equipment; l (3) Alpha scintillation or proportional count rate (3) Facilities and equipment to clean, repair, and meters; decontaminate personnel protective equipment, m nit ring instruments, hand tools, electromechani-l (4) Neutron dose equivalent rate meters; l cal parts, or other material (highly contaminated tools j (5) Air samplers for short-term use with particu-or other equipment should not be decontaminated in late filters and iodine collection devices (such as acti-the area used to clean respiratory equipment); vated charcoal cartridges); and (4) Change rooms that (preferably) connect with (6) Air monitors with continuous readout fea-the personnel decontamination area and a control sta- ) tures.? tion area equipped with sufficient lockers to accom-j
- c. Personnel Monitoring Instrumentation m date permanent and contract maintenance workers who may be required during major outages; Personnel monitoring instrumentation selection should include consideration of:
(5) Control stations for entrance or exit of per-sonnel into radiation-and contamination-controlled .l (1) G M ' Friskers" for detecting low levels of access areas of the station such as the personnel en-radioactive material; trance to the containment buildings and the main en-trance to the radwaste processing areas; these cetrol (2) Direct-reading low range (0 to 200 mR) and stations also may be used as the control point for mtermediate-range (0 to 1000 mR) pocket dosimet-radioactive material movements throughout the sta-I ers (see Regulatory Guide 8.4); tion and for the storage of portable radiation survey (3) Alarm dosimeters; equipment, signs, ropes, and respiratory protective 89ui ment; P (4) Film badges and/or thermoluminescent i dosimeters (TLD); (6) Equipment to facilitate communication be-
- "*""'****'"8
"* * * *** U "; ""d (5) Hand and foot monitors; and (7) Sufficient office space to accommodate the ' Vwieble slutn utpoint features on thew intrunwnts can be temporary and permanent radiation protection staff, valuable in providing a warning when unespected substantial } changes in dose rate or air concenstation occur. Permanent records, and technical literature. 8.8-15 i
e D IMPLEMENTATION protection design presented in the applicant's final safety ana1ysis report will be reviewed against regula-The purpose of this section is to provide informa-tory position 2 of this guide and differences from the tion to applicants and licensees regarding the NRC rec mmendations of the guide will be identified (par-f. staff's plans for using this regulatory guide. ticularly for plants designed before Regulatory Guide This guide reflects current NRC staff practice in 8.8 was issued). However, no substantive design license application reviews. Therefore, except in changes will be required at the operating license stage those cases in which the applicant proposes an ac-unless the design change can prevent substantial ceptable alternative method for complying with speci-man-rem exposures that cannot be prevented by pro-fled portions of the Commission's regulations, the cedural measures and the design change is consistent methods described herein are being and will continue with the cost-effectiveness principle of maintaining to be used in the evaluation of submittals for con-occupational radiation exposures ALARA. struction permits and operating license applications M b Ws ide until this guide is revised as a result of suggestions may be substituted for those stated herein, provided from the public or additional staff review. they satisfy the criterion "as low as is reasonably At the operating license review stage, the radiation achievable" of paragraph 20.l(c) of 10 CFR Part 20. REFERENCES
- 1. Ad Hoc Committee of the National Council en
- 9. ANSI N101.2, " Protective Coatings (Paints) for Radiation Protection and Measurements. " Somatic Light Water Nuclear Reactor Containment Radiation Dose for the General Population," Science Facilities." Copies may be obtained from the Ameri-131, 482 (1960).
can National Standards Institute,1430 Broadway, New York, N.Y.10018.
- 2. "The Effects on Populations of Exposure t
- 10. ANS/HPS 56.8, " Location and Design Criteria Low Levels of lonm,ng Radiation.
National for Area Radiation Monitoring Systems for LWRs," Academy of Sciences / National Research Council, gg* DHEW Contract PH-43-64-44 November 1972.
- 11. ANS N197, " Design and Performance of BWR r
- 3. International Commission on Radiological Pro-Liquid Radioactive Waste Processing Systems tection (ICRP), " Implications of Commission Rec-(N18)." Copies may be obtained from the American ommendations That Doses Be Kept As Low As Read-Nuclear Society, 555 North Kensington Avenue, La ily Achievable " ICRP Publication 22, Pergamon Grange Park, Illinois 60525.
Press,1973. Copies may be obtained from Pergamon
- 12. ANS 55.1, " Design Criteria for the Solid Press, Maxwell House. Fairview Park Elmsford, Radwaste Processing System of BWR, PWR, and New York 10523.
HTGR." Copies may be obtained from the American i. Nuclear Society, 555 North Kensington Avenue, La
- 4. C. A. Pelletier et al., " Comp.lation and Analy-Grange Park, Illinois 60525.
ses of Data on Occupational Radiation Exposure Ex:
- 13. ANS N199, "PWR Liquid Waste System De-perienced at Operating Nuclear Power Plants,.
sign (N18)." Copies may be obtained from the Atomic Industrial Forum,1974 American Nuclear Society, 555 North Kensington
- 5. T. D. Murphy. N. J. Dayem. J. Stewart Bland, Avenue, La Grange Park, Illinois 60525.
and W. J. Pasciak, " Occupational Radiation Expo-
- 14. Section XI. ASME Boiler and Pressure Vessel sure at Light Water-Cooled Power Reactors,1%9-Code and Addenda. Copies may be obtained from the l
1975." NUREG.0109. U.S. Nuclear Regulatory American Society of Mechanical Engineers, United Commission. August 1976. Copies may be obtained Engineering Center,345 East 47th Street, New York, from the National Technical Information Service, N.Y.10017. Springfield, Va. 22161.
- 15. ANSI Z-88.2, " Practices for Respiratory Pro-
- 6. NUREG-0322. " Ninth Annual Occupational tection." Copies may be obtained from the American National Standards Institute,1430 Broadway New Radiation Exposure Report,1976." Copies may be York, N.Y.10018.
obtained from the National Technical Information
- 16. NIOSH, " Certified Personal Protective Service, Springfield, Va. 22161.
Equipment List," July 1974, and supplements by
- 7. ANSI N237, " Source Term Specification."
DHEW/PHS. Published by U.S. Department of i I Copies may be obtained from the American Nuclear Health, Education, and Welfare. Public Health Serv-Society 555 North Kensington Avenue, La Grange ice, Center of Disease Control National Institute of Park, Illinois 60525. Occupational Safety and Health. Copies are available
- 8. Copies of NUREG 0041 may be obtained from from the Office of Technical Publications, National the National Technical Information Service, Institute of Occupational Safety and Health, Post Of-Springfield, Va. 22161, fice Building, Cincinnati, Ohio 45202.
8.8 16
EXHIBIT 2 Final Safety Analysis Report Figures 12.3.1-16, 5.1.3-1, and 12.3.1-18. l l
9 0 y b~ APPLICANT'S FINAL SAFETY ANALYSIS REPORT SHOREHAM NUCLEAR POWER STATION UNIT 1 1 ) h 7p g,gg ~' VOL.12 ~. ~
~ ' e i-l lJ ^$kh / cE Ford e,- EACTOR l @ NEW i i FUEL \\ u
- g. g _W' /
.OUD $RVs SHR VAULT h r" / r EL,,0..- FUEL POO COOLING ggg HEAT HG. p REACT Ra193 ~ e 0-h .e--- 7 r-e i j E <,, e... -f !.....? 7g g *'i L: f ACTIVE; [ *01/ Qhh} '\\ H ) '() EL ta'-7
- ll /
.. o. g .q // CONTROL M r.ei@ t Roos- ') E L 4 0' 0* -.m We EL W1 1;- + 4 '- 6* 1 LEGEND.
- m E OuRING NORM AL OPE RATION.
- Ug","a",%',*" "C'*'"
FI G.12.3.1 - 16 SHIELDING ARRANGEMENT AND m ouR.. ORMAL OPERAtl0N m ac 7 inuroc.= RADIATION ZONE DESIGNATIONS FOR THE REACTOR BUILDING SECTION 1-1 SHOREHAM NUCLEAR POWER STATION - UhlT 1 FINAL SAFETY ANALYSIS REPORT I
g I i ,P 79-O' ( ell 4 5 kb MAIN e f.' ~ e ~ STEAM ** (EL1h TEL 13 2*-S* 8 13 2*- 5* ',t. / ( gt. FEEDWATER M 105t7} t) fg. EL96,-ilj,,- aP- --89'] I TQS. l (ELj 5 ( EL 87'-0* a { FEEDWATER SYSTEM 93'-2h* 7 . -es. 22* RECT RCr I I ~ " =s LOOP 28" RECIRC l EL 76* 4 k' N) OS - ' -~**s f EClRC.- (EL82' 6* [ MAIN STEAM PU M P E goj i S EL62' 8*A -
- c. m a
r= m ?:t f \\p.R .,.91I t \\ w... a w l ';:) q l-PLATE R i -ER PfPES i g (TYP) i I =_gp - --mure:-- - -_7 SUPPRESSION EL 20*-O" ] D' f... h*d ~*
- ,?
EL 8'-O* ? :' ' ". ~.
- D.:.
NOT E St ALL 3IMENSIONS ARE APPROIlWAf f EL.0' 0*
- Sta LtvEL FIG. 5.1.3-1
[ REACTOR COOLANT SYSTEM l ELEVATION DRAWING t l SHOREHAM NUCLEAR POWER STATION-UNITI l FINAL SAFErY ANALYSIS REPORT
a t i DRYWELL EQUIPMENT HATCH i i ~ i / kI PERSONNELN HATCHL y{ @i-2.MT l @ @ O '\\ 20 - \\ / \\ / E L.78'-7" / \\ s\\. @M p RECIRC. RECIRC. PUMP' a/ ~ W 7 - 2( PUMP EL. 63'-0 I ( ) ~4'-6" 1 t v. FIG.12.3.1-18 ) SHIEL0 LNG ARRANGEMENT AND i RA01 AT10N ZONE DESIGN ATIONS FOR THE REACTOR BUILDING SECTION 3-3 SHOREHAM NUCLEAR POWER STATION - UNIT pg 9
EXHIBIT 3 Final Safety Analysis Report, Table 12.4.3-1, Volume 12.
Y APPLICANT'S FINAL SAFETY ANALYSIS REPORT SHOREHAM NUCLEAR POWER STATION UNIT 1 N dEf9 l VOL.12 P
o SNPS-1 FSAR TABLE 12.4.3-1 ESTIMATES OF ANNUAL DOSES C 10 l Nork Area Productive Task Exposure Rate Work Time Man-Exposure (2) (Sub-Task) (mrem /hr) (hr/ man) Power (man-rem) i CRD Change Out (ISC Techs) 123 11 2 2.9 (Mechanics) 123 78 12 20.9 CRD Repair (Mechanics) 40 78 3 9.6 ISI-Drywell Piping (Mechanics) 150 48 16 31.6 (NDT Techs) 150 11.1 2 3.6 Refueling (Mechanics) 20 130 5 13.4 Recirc. Pump Maint. (ISC Techs) 205 0.4 1 0.09 (Mechanics) 75 70 12 21.5 Radwaste Syst.'Maint. (Mechanics) 150 75 12 36.7 Main Steam Relief Valve Maint. (I6C Techs) 75 6 1 -0454 (Mechanics) 100 40 6 17.8* LPRM Maint. (ISC Techs) 100 12 4. 6.1 MSIV Maint. (Electricians) 100 2 2 0.o2 (Mechanics) 250 25 12 30.9 Snubber Inspection / Repair (Mechanics) 120 20 6 9.4 RWCS Maint. I (Electricians) 250 5 2 2.6 (Mechanics) 200 13 3 8.2 i 1 of 2 Revision 18 - June 1980
SNPS-1 FSAR TABLE 12.4.3-1 (CONT'D) Work Area Productive Task Exposure Rate Work Time Man-Exposure C 2 ) [ (Sub-Ta sk) (mrem /hr) (hr/ man) Power (man-rem) ( RHR System Maint. l (Mechanics) 86 24 2 5.0 Turbine W;rk L (Mecha .) (NDT Techs) 2 180 8 3.1 Special Maint. (Laborers) 100 125 20 53.8 (Mecnanics) 300 125 34 83.3 l Health Physics 65 i Personnel f Outage S Routine 437(3) Operations ~ Routine ".aint. (Mechanics) 146 TOTAL 583 (*) Data obtained from Atomic Industrial Fornm, In'c.; " Study of l the Effects of Reduced Radiation Exposure Limits on the i Commerical Nuclear Power Industry"; July 1979. Estimates given are based on a 3 rem / quarter occupational dose limit.. (a) An ingress / egress
- factor, discussed in the AIF Study, is included.
(3) Although the estimate for the selected. tasks given. is 427 l man-rem, a total estimate of 437 man-rem is estimated in the AIF Study for all outages and routine operation. t i 2 of 2 Revision 18 - June 1980 l l
s 4 EXHIBIT 4 NUREG-0713. " Occupational Radiation Exposure at f Comercial Nuclear Power Reactors 1980" Volume 2, December 1981, cover page and page 18. 1 9 i y. y.
I NUREG-0713 Vol. 2 Occupationa Radiation Exposure a: Commercial \\'uc ear 3 ower Reactors 198C Annual Report Manuscript Completed: December 1981 Date Published: December 1981 B. G. Brooks Licensee Operations Evaluation Branch Office of Management and Program Analysis U.S. Nuclear Regulatory Commission Washington, D.C. 20555 / ~.
- 4 x....-
4
~ TABLE 8 ANNUAL COLLECTIVE DOSES BY WORK FUNCTION AND PERSONNEL TYPE 1980 Work Function Station Employees Utility Employees Contract Workers Se Others Total per Functio-10_[.LlyG__W A TJ R R E A C T O R S REACTOR OPERATIONS 8 SURVEILLAtlCE 1615.1 5.8 % 82.1 0.3 % 421.0 1.5 % 2118.2 7.i E ROUTINE MAINTENANCE 2487.9 8.9 % 1448.2 5.2 % 7984.7 28.6 % 81920.8 42.! % IllSERVICE IHSPECTION 108.4 0.4 % 142.9 0.5 % 655.8 2.4 % 907.1 3.I E SPECI AL MAIllTEHANCE 745.2 2.7 % 606.4 2.2 % 9262.5 33.2 % 10614.1 38.* 8 WASTE PROCESSING 524.2 1.9 % 12.7 0.0 % 343.7 1.2 % 880.6 3.i REFUELING 557.6 2.0 % 67.0 0.2 % 814.6 3.0 % 1439.2 5.; 3 TOTALS 6038.4 21.7 % 2359.3 8.4 % 19482.3 69.9 % 27880.0 100. ERESLU_R_HID WATfR REACTORS REACTOR OPERATIDHS 8 SURVEILLANCE 1784.0 7.6 % 95.2 0.4 % 822.6 3.5 % 2701.8 11.5 % ROUTIHE MAINTEHAHCE 1719.t 7.3 % 627.2 2.7 % 4003.5 17.0 % 6349.8 27.: % IliSERVICE INSPECTION 158.6 0.7 % 126.9 0.5 % 1637.6 7.0 % 1943.1 8.~ SPECIAL MAINTENANCE 1019.6 4.3 % 1042.0 4.4 % 8215.6 34.9 % 10277.2 43.i % WASTE PROCESSING 333.9 1.4 % 31.0 0.1 % 264.7 1.1 % 629.6 2. i *. REFUELING 603.3 2.6 % 319.4 1.4 % 731.5 3.1 % 1654.2 7.* T01ALS 5620.5 23.9 % 2241.7 9.5 % 15675.5 66.6 % 23535.7 100.: % ALL._LIQHT W RER_RFACTOR$ REACTOR OPERAIIDHS 8 SURVEIL L AtlCE 3399.1 6.6 % 177.3 0.3 % 1243.6 2.4 % 4820.0 9.I % ROUTIHE MAINTENANCE 4207.0 8.2 % 2075.4-4.0 % 18988.2 23.3 % 18270.6 35.3 % IllSERVICE INSPECTION 267.0 0.5 % 269.8 0.5 % 2293.4 4.5 % 2830.2 5.5 % SPECIAL MAINTEllANCE 1764.8 3.4 % 1648.4 3.2 % 17478.I 34.0 % 20891.3 40.1 % WASTE PR0fAS%ING 858.1 1.7 % 43.7 0.1 % 608.4 1.2 % 1510.2 3.I REFUELING 1160.9 2.3 % 386.4 0.8 % 1546.1 3.0 % 3093.4 6.: TOTALS 19658.9 22.7 % 4601.0 8.9 % 35157.8 68.4 % 51415.7 100.I % i i l 1
9 EXHIBIT 5 " Compilation and Analysis of Data on Occupational Radiation Exposure Experiences at Operating Nuclear Power Plants", SAI Services September 1974, cover page and page 12. 0 5 I f
c., p y. Wi 'g I s er..
- =e wn ~
r l Comai a: ion and Analysis i of Data on n.. -
- ,W '
r Occupationa w y q' _ Radiation Exposure Experiencec a. Opera ~:ing Nuclear Power P an~:s { ( i ~ .i x. -) i Al .- I t.; i .ll
- . i r.,
+ F i' j' <D l
- /
t I '-) i l i 's -5, j Prepared by SAI Services ~.' A Division of Science Applications,Inc. w, 4 ( %.c. ; r '. 1 ~; \\ NNs .I ,r. 9
- a. w.. w
+ Qja, (%. 'i f. ;,t)j.$ @..
- Q:-
- 9. ;v-a^
g W si Atomic industrial Forum, Inc. 2 . ;= M.!k@p : ! I i I
Figure 4-1 Annual Total Body Exposura As A Function of Plant Aga 0 BWRs 3' 5 1000 /
- 5 33 m Os a6,
$ a. O ". 14 %x-X le V/ / 4 100 g h j
- TWO UNITS z
<:;;, /
- UNIT 1 ONLY I
I I I I I I I I I I 10 7f / PWRs i 1000 " O[12 / h'* a g ',/' wo/ / 2 si a 3 l 0) [\\g,Co x + x /\\ + c / /N /W-E 100 / \\ a 4 s f Y a 2
- g to
+ 5 i< i l l i I I I I I I l l g 1 2 3 4 5 6 7 8 9 10 11 12 13 YEARS OF OPERATION 12 Symbols represent various plants j
EXHIBIT 6 Radiation Buildup Effects, by Dr. William R. Dehollander March 1980, cover page and page 24. l
l l RADIATION BUILDUP EFFECTS F f r 0F l I WATER QUALITY r l l i t t .i li;, i ,i t l i 11 L l;i' BY DR. WILLIAM R. DEHOLLANDER I i i b ~ 0 MARCH 1980 .i 'j 'i .i li l ,1 SET IV i )
- ~~ - ^^ '~ - g- ~ MEAN CORROSION OF CARBON STEEL !N ~ AIR-SATURATED V/ATER AT 125 F RATES R ION RELEASE ~ VELOCITY CONDUCTIVITY FT/SEC mg/dm -mo mg/dm2 2 -mo 0.1 0 347 278 1 16 12 3 9 3 6 2 2 3.6 0 930 ~ l 515 1 839 46 3 303 52 f 6 7 12 I
- KEEP WATER CIRCULATING AND AVOID STAGNANT WATER e KEEP WATER CONDUCTIVITY LOW c u, J --t i
\\
O i i EXHIBIT 7 Radiation Buildup Effects, by Dr. William R. Dehollander March 1980, cover page and page 13. 1
+.- ,l b y = t RADIATION BUILDUP EFFECTS L e ir i OF l I: WATER QUALITY ll l l I I., ii 6 P ji t li BY 1 DR. WILLIAM R. DEHOLLANDER i !l i 6 .I. i. f 'i'i 1: 1 i t ..l ~ P MARCH 1980 I m !I L ~ t L t SET IV l m
~ c T ... r RADIATION FIELD BUILDUPON RECIRCULATION PIPING 1000
- RATE CAN BE LOW
[3 osx1Eil7u'"" 5$n'$'M e LEVEL CAN BE LOWERED 900 - l e OSKAR5HAMN 2 V Mites, ORE 1 800 o sonous nony, ..ourien to e snowns Fenny.3 + FuxusniuA.: o aAnstnAcx i o ruxusnauA.3 700 ooAo ci,,is. . ninc..Ats i Q OuAD ct,Its 2 9 HAMAOK A 1 """""5""' ' 5"'"^"' 150 mR/h 600 . isuouc^ PER EFPY + DOSE 500 RATE (mR/h) 400 + v v v v 300 ^ + s' + 4 i 200 s A ',g /+ ' y s l ot
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e o EXHIBIT 8 NUREG-0855, " Health Physics Appraisal Program" March 1982, cover page and abstract. 0 6 l l l 1 l
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i 9 O ABSTP/CT The accident at Three Hile Island in March 1970 and subseouent investigations identified, among other items, serious concerns involving several aspects of the rediation protection program. Significantly, some concerns involved areas not addressed by regulations or facility technica1 specifications. This in turn led to initiation of a ma,ior effort to evaluate the adecuacy and effective-ness of radiation protection programs at all currently operating nuclear power facilities during calendar year 1980 by the Office of Inspectier and E.1forcement (IE), Nuclear Regulatory Commission. This inspection effort was termed an appreisal since it was structureci to facilitate an integrated look at the total radiation protection program, delve into matters for which explicit regulatory ~ requirements did not exist, and emphasized evaluation of capability and per-formance rather than ccmpliance with regulations. This report discusses the results of the 48 appraisals and the anticipated regulatory arltions that may be taken to further address the concerns. iii}}