ML20024B717
| ML20024B717 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/23/1983 |
| From: | Boegli J, Branagan E, Serbu R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20024B704 | List: |
| References | |
| NUDOCS 8307110270 | |
| Download: ML20024B717 (20) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COPHISSION l
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of.
Docket Nos. 50-413 50-414 (Catawba Nuclear Station, Units 1 and 2)
AFFIDAVIT OF JACQUES S. BOEGLI.
EDWARD F. BRANAGAN, JR., AND RICHARD JOHN SERBU IN SUPPORT OF
SUMMARY
DISPOSITION OF DES CONTENTION 19 1.
I, J.S. Boegli, being duly sworn, do depose and state:
I am an employee of the U.S. Nuclear Regulatory Comission in the Effluent Treatment Systems Branch, Office of Nuclear' Reactor Regulation.
I am responsible for the review and evaluation of radioactive waste treatment and effluent control systems and for the calculation of effluent source terms for nuclear power reactors. My professional and educatien qualifications are attached to this statement.
I certify that I have personal knowledge of the matters set forth herein with respect to the above areas for which I am responsible, and that the statements made are true and correct to the best of my knowledge.
2.
I, Edward F. Branagan, Jr., being duly sworn, do depose and state:
I am a Health Physicist in the Radiological Assessment Branch, Division of Systems Integration within the Office of Nuclear Reactor.
Regulation. A copy of my professional qualifications is attached.
I certify that I have personal knowledge of the matters set forth herein 8307110270 830708 PDR ADOCK 05000413 0
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with respect to assessmant of the impact from exposure of the public to radioactive effluents from spent fuel stored at Catawba. The statements made are true and correct to the best of my knowledge.
3.
I, Richard John Serbu, being sworn, do depose and state:
I am an employee of the U.S. Nuclear Regulatory Comission (NRC). My present position is Health Physicist, Radiological Assessment Branch, Division of Systems Integration within the Office of Nuclear Reactor Regulation.
A copy of my professional qualifications is attached.
I certify that I have personal knowledge of the matters set forth nerein with respect to assessment of occupational exposures to on-site personnel and that the statements made are true and correct to the best of my knowledge.
4.
This affidavit addresses DES Contention 19, which reads as follows:
" Failure to evaluate the environmental costs of operation of Catawba as a storage facility for spent fuel from other Duke facilities compromises the validity of the favorable cost-benefit balance struck at the construction permit phase of this proceeding. Since the CP stage hearing, Duke Power has considerably expanded the Catawba spent fuel pool capacity and provided for denser storage of irradiated fuel. FSAR Table 1.2.2-1.
Applicants intend to use Catawba for storage of irradiated fe::1 from the McGuire and Oconee nuclear facilities of l
Duke Power Company. FSAR 9.1.2.4; OL Application, pp.11-12."
In admitting this contention, the Licensing Board stated that "... the primary focus of DES 19 would be on the environmental effects of routine releases from such [0conee and McGuire] transhipped fuel during ncmal operations at Catawba." (Memorandum and Order, February 25,1983,p.9).
5.
In the FES, the Staff analyzed "the environmental costs of operation of Catawba as a storage facility for spent fuel from other Duke facilities" in the following manner. The major environmental
- pathways of exposure of humans were considered. Tables D.1 and D.4 included i
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3-releases from spent fuel from Catawba and the spent fuel that is expected to be stored at Catawba from Oconee and McGuire.
In its review, the Staff determined that the releases of radioactive materials from fuel storedintheCatawbaSpentFuelStorageFacility(SFSF)wereverysmall fractions of the total releases from normal operations of the entire Catawba facility. Similarly, dose commitments to a maximally exposed individual and to the population from operating Catawba included releases from the storage at Catawba of spent fuel from Catawba, Oconee and McGuire.
(SeeTablesD.6,D.7.D.8). Finally, estimates of dose to workers from norwal handling of spent fuel casks from Oconee and McGuire were evaluated at Sec. 5.9.3.1.2 of the FES (p. 5-19). The Staff concluded that the systems as now designed and built are capable of controlling effluent releases, including those from stored spent fuel from other Duke facilities, to meet the dose-design objectives of Appendix I to 10 CFR 50.
In addition, the estimated doses to individual members of the public and to the general population from exposure to all effluents from the facility were very small fractions of the annual doses from exposure to background radiation. Further, the Staff, in its review, determined that estimated doses to individual members of the public and to the general population from exposure to effluents from the SFSF were very small fractions of the estimated doses from exposure to all effluents.
6.
This affidavit sets forth in more detail than in the FES the sources.and amounts of routine releases of radioactive materials, and i
resultant dose commitments both onsite and offsite which may be expected from the SFSF.
In the following analysis the Staff has evaluated the environmental impact associated with the receipt and storage of spent
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_4 fuel in the SFSF by addressing (1) the types of releases from the SFSF
.(liquid, gaseous,- solid) leading to possible exposure of the public, and (2) the possible occupational doses to workers associated with spent fuel storage and fuel handling. The Staff's environmental evaluation encompasses the contribution of the expanded spent fuel storage facility, at full capacity, and includes the contribution attributable to receipt and' storage of Oconee and McGuire spent fuel at Catawba.
Routine Releases 7.
The amount of radioactivity which will be released into the environment from the SFSF, and the amount of radioactivity which may be attributed to the storage of Oconee and McGuire spent fuel in the Catawba SFSF, may be determined based on the capacity of the SFSF, the release and transport mechanisms that result in the appearance of radio-active material in liquid and gaseous streams, and the plant-specific design features used to treat and store radioactive material from fluid sireams by collection on media for disposal as solid waste. Such estimates of routine radioactive releases are called the " source term" and are derived by techniques presented in NUREG-0017. " Calculation of Releases of Radioactive Materials in Gases and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code) using inplant measurements at many of the operating nuclear power plants. The SFSF at Catawba, Unit 1 and Unit 2, are independent, as required by 10 CFR Part 50, Appendix A GDC No. 5.
However, they have similar design features such that the routine releases umy be calculated on a per unit basis.
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Capacity of the SFSF 8.
In detemining the routine releases per unit, the Staff con-sidered'the SFSF to be full at its design capacity of 1418 fuel assemblies.
Although the environmental statement published upon consideration of the Catawba construction pemit application was base'd on a SFSF with 265 fuel assemblies, the FES and Safety Evaluation Report published for the-operating license application are based on the increased capacity. This increase would allow for 193 feel assemblies from off-loading a fully i
loaded core at Catawba at any time, plus storage space for approximately 258 fuel assemblies from Catawba that are less than five years out-of-core, and storage space for approximately 967 fuel assemblies that would be over five years out-of-core. Applicants have proposed to store spent fuel from Oconee and McGuire at Catawba which would be at least five years out-of-core. However, for the purpose of calculating routine releases, the Staff considered for the FES that the Catawba fuel assemblies and those from McGuire and Oconee to be equivalent after five years out-of-core since the assemblies are approximately the same size (8.4 x 8.4 x 144 inches), the same materials (Zircaloy-4/Inconel 718),
contain the same U0 fuel (about 1150 pounds UO / assembly) are used to 2
2 about the same burnup rate (33 MWD /Kg U) and would have approximately the
. same fission product inventory after the same amount of time in storage.
Therefore, the origin of fuel assemblies over five years out-of-core would not impact the routine releases. Although the date at which maximum capacity would be reached would be advanced by the storage of McGuire and Oconee fuel assemblies in the SFSF, the Staff bases for routine releases considered a full SFSF at any time during the operating i
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life of the Catawba Nuclear Station. Therefore, the routine releases based on a full SFSF will not be increased by the shipment of five year old fuel assembli s from McGuire and Oconee, and the routine releases at a time where the SFSF is below capacity will be less than at full
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capacity.
Release Mechanisms and Treatment Provisions 9.
The release mechanisms for routine releases are the same for the underwater storage of fuel assemblies at all facilities. During the
. movement and storage of fuel assemblies in the SFSF, both volatile and nonvolatile radioactive materials may be transferred to the SFSF pool water from the outer surface of the fuel assemblies or from defects in the fuel assembly cladding. Most of the outer surface seterial consists of activated corrosion products, such as Co-58, Co-60, Mn-54 and Fe-59, which are nonvolatile. The Staff estimates that this outer surface material. constitutes about 0.001 Ci/ assembly and that most of the material is insoluble. The spent fuel pool cleanup system removes the insoluble material transferred to the pool water by continuous recirculation through filters and removes any soluble material by demineralization. Most of the surface material is removed during the first few months of storage in the pool water such that there would be little contamination of the SFSF by assemblies shipped from McGuire or Oconee that have been stored in their respective spent fuel pools at least five years prior to shipment to Catawba. None of the surface corrosion products are volatile since they are salts and metal oxides.
- 10. The radioactive materials that may be transferred to the SFSF pool waters from cladding defects are generally nonvolatile fission a
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products, such as Cs-134, Cs-137, Sr-89 and Sr-90. The abundance of fission products transferred into the SFSF water is dependent on the fission yield, the time since irradiation in core, the size of any cladding leak and the temperature of the cladding. The Staff
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estimates that the fission products transferred to the pool water due to cladding defects constitutes about 0.01 C1/ assembly at the tima of unloading from the core, less than 0.001 C1/ assembly after about 20 days in storage, and an undetectable amount after several years in storage.
These soluble and insoluble radioactive materials are continuously removed from the pool water by the demineralizers and filters in the spent fuel pool cleanup system. Measurements at operating-plants have shown that most defects or pinhole perforations in the fuel cladding are self-sealing when the cladding temperature is relatively cool, approximately 180*F. This self-sealing condition occurs in about 20 days after a fuel assembly is removed from the core, and together with radio-active decay, greatly reduces the net effect of nonvolatile fission products in the SFSF pool water. Since Oconee and McGuire spent fuel received and stored at Catawba would be at least 5 years out-of-core, the contribution of radioactive materials due to defects in such fuel would be undetectable.
- 11. The release mechanism for volatile fission products is the same as for. the nonvolatile; however, their solubility in the SFSF pool water sbst be considered. Generally the SFSF water is maintained below 140*F to reduce the cladding temperature, reduce water evaporation and to increase the solubility of gases and thereby contain most of the volatile materials. Radioiodines and most of the noble gases are 4
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reduced by radioactive decay, together with the self-sealing condition of the fuel cladding. Tritium produced in the reactor coolant and within the fuel assembly is not a significant nuclide, since in the case of Catawba, there is no major mixing of reactor coolant water cr fuel
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ac...nbly transfer cask water into the SFSF. Operating experience has demonstrated that after 4 to 6 months there is no significant release of volatile fission products from fuel assembifes and the only significant noble gas nuclide attributable to long term storage of fuel assemblies would be Kr-85, which is at undetectable concentrations in the plant effluent after two years out-of-core.
- 12. The SFSF pool water is recirculated and continuously cooled, filtered and demineralized. The treatment removes radioactive materials from the water by filter and exchange media such that nonvolatile materials are collected for disposal as solid waste, and solids disposal L
is to a licensed burial site. During routine servicing and maintenance operations, excess water from the SFSF is treated in a similar manner by the liquid radwaste t*eatment' system. There are no releases of SFSF pool water. Radioactive materials in gaseous effluents from the SFSF are collected by the fuel building ventilation system, treated by filters and absorbers for particulate and radiciodine removal (if any),
monitored and released to the atmosphere via the plant vent.
(These effluents are discussed further in the section entitled " Calculated Release.s.and Dose Impact".)
- 13. Prior to the publication of the FES, the Staff reviewed the release mechanisms for radioactive materials from spent fuel assemblies.
and the provisions for treating the liquid, gaseous and solid waste
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generated by the operation of the Catawba SFSF. The Staff found that the n ! ease mechanisms were not altered by the number of fuel assemblies in the SFSF, and the treatment provisions were designed and installed adequately to meet the requirements of a full SFSF. Since there are no detectable transfers of radioactive material from the fuel assemblies to the pool water after five years of storage, there would be no difference I
between a full SFSF containing only Catawba spent fuel assemblies and a full SFSF containing fuel assemblies from McGuire and Oconee. The pool water radioactivity would be essentially the same.
Calculated Releases and Dose impact
- 14. There are estimated to be essentially no liquid releases from the SFSF since it is a closed recirculation / treatment system. Therefore, radioactive materials in liquid effluents calculated for the FES and the SER for Catawba, Unit Nos. I and 2 did not include SFSF releases, and the proposal to store fuel assemblies from McGuire and Oconee would not change this conclusion.
- 15. Solid radioactive wastes, generated by the Catawba SFSF pool water filter and demineralizer treatment system are packaged and shipped to a licensed burial site. The Staff estimated that the volume of solid waste generated by disposal of filters and demineralizer exchange media from the spent fuel cleanup system would amount to about 6 cubic feet per year per unit at Catawba, containing approximately 0.1 Ci/ cubic feet.
The environmental impact of the transportation and disposal of these low level wastes are accounted for by the generic values in Table S-3,10 CFR 51.20, as stated in the FES. As a conservative estimate, the Staff assumed that the amount of solid waste generated by the storage of
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. McGuire and Oconee fuel assemblies at the Catawba SFSF may be increased by 6 cubic feet per year per unit.- There would be no increase in activity of this solid waste since there is no increase in the radioactivity of the pool water. The annual average amount of solid waste to be shipped from each unit at Catawba is' estimated to be about 20,000 cubic feet per year. The total solid waste volume generated by the storage of McGuire and Oconee fuel assemblies at Catawba is estimated to be less than 0.1% of the preceding value and would not have any-significant environmental impact not already considered in the FES and the SER for the Catawba Nuclear Station.
- 16. Due to the release mechanism for volatile fission products from fuel assemblies after several years out-of-core (described in paragraph 11, above), it is estimated that there would be no measureable releases of noble gases in the plant effluent. Using computer models, the Staff calculated in the FES, Table D.1 on page D-5, that the annual average release of Kr-85 for the auxiliary building stack, which includes releases from the SFSF, would be less than 1 Ci/yr averaged over the 30 year operational life of the plant, with the SFSF less than at full l
capacity some of this time. The Staff conservatively estimated that if the SFSF were at full capacity all of the time, the maximum routine release of Kr-85 would be less than 1 Ci/ year or no more than 0.5% of l
the total annual release of Kr-85 from either unit.
17.. Therefore, if fuel assemblies from McGuire and Oconee were shipped after 5 years out-of-core and stored in the Catawba SFSF a conservative estimate of the gaseous releases would be less than 1 curie per unit per year of Kr-85. The estimated doses to the total body and
. I skin of a maximally exposed individual are estimated to be much less than 0.1 mrem / year. The Staff's method for calculating the total body dose impact is provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I."
Revision 1, October 1977. This dose is not significant when compared to the approximately 100 mrem / year that an individual receives from natural background radiation and small compared to the fluctuation in the annual dose that an individual would receive from natural background radiation.
- 18. The dose to the total body of the population within a 50-mile radius of the Catawba Nuclear Station (estimated to be about 1,700,000 persons in the year 2000, FES, p. D-9) due to nonnal operation of the SFSF, with the storage of McGuire and Oconee fuel assemblies, is estimated to be less than 0.1 man-rem / year. This dose is a very small fraction of the annual dose of about 160,000 person-rems (FES,p.D-9) that this population would receive from natural background radiation.
Thus, the Staff concludes that nonnal operation of the SFSF including the proposed storage of fuel assemblies from McGuire and Oconee in the Catawba SFSF will not have any significant impact on exposures offsite.
Occupational Dose 19.
In addition to the effluent pathways, the DES /FES and SER considered occupational doses which might be associated with the handling and storage of spent fuel, including fuel transferred from other facilities. The SER, in Section 12.5, considered occupational doses and ALARA/ radiation protection practices associated with fuel l
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s handling and storage operations at the SFSF at Catawba, which was evaluated by the NRC Staff as follows:
The storage of spent fuel at Catawba has been evaluated in accordance with Standard Review Plan Sections 12.2 and 12.3-12.4.
The applicant has provided information which satisfactorily meets with our guidance and positions in the Standard Review Plan, including the requirements in 10 CFR Part 50, Appendix A; General Design Criteria-61; 10 CFR Part 70, 70.24; and Regulatory Guide 8.12, and is therefore acceptable. Additionally the radiation protection program, organization, and policies have been evaluated as indicated in Sections 12.1 and 12.5 of the Safety Evaluation Report, and are acceptable for the transfer and storage of spent fuel from Oconee and McGuire. No significant additional occupa-tional doses should result from the storage of additional spent fuel at Catawba, since direct doses from stored fuel provide only a fractional contribution to spent fuel pool area dose rates in comparison to radioactivity in pool water. Similarly, dose increases due to the handling of spent fuel casks at Catawba would contribute only a very small fraction to the total projected dosc for the facility.
- 20. The Staff evaluated and found ecceptable Applicants' estimates of worker doses associated with spent fuel pool operations. These estimates are based on work area dose rates, the nature and location of work to be perfomed, and the time spent performing work in the pool area. Applicants' estimates reflect source tems, dose rates, work, and work times which are consistent with those measured and observed for similar facilities and operations throughout the industry.
- 21. Dose rates of 1 mrem /hr to 2 mrem /hr are typically measured at the surface of spent fuel pools, primarily due to the presence of 60 58 contaminants such as Co and Co in the pool water. These contaminants are introduced into the spent fuel pool water as fuel is moved about in the pool'and activated corrosien products (which typically adhere to in-core fuel assemblies) are disturbed and transfer to the water.
However, as stated in paragraph 13, supra, Oconee and McGuire fuel, being stored over five years prior to storage at Catawba, would not
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transfer defectable amounts of such containments to the Catawba SFSF.
- Therefore, while occupational dose rates from operations at Catawba are expected to reflect industry experience, generating dose rates in the range of 0.5 mrem /hr to 5 mrem /hr--and most probably generating dose
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- rates of 1 mrem /hr to 2 mrem /hr during stable storage periods with the pool cleanup system in operation--no significant amount of these dose rates would be attributable to the Oconee and McGuire spent fuel proposed to be transshipped to Catawba.
- 22. Due to the shielding effect of water, the f; a assemblies themselves will contribute little to the overall dose rate at the surface of the spent fuel pool - a fraction of the dose rate originating from the pool water itself. A specific water level for radiation 4
shielding is not required by regulations. Design guidelines for spent fuel pools, ANSI N 210-1976, " Design Objectives for Light Water Reactor l
l Spent Fuel Storage Facilities at Nuclear Power Stations," recommends that radiation dose to personnel in normally occupied areas of spent fuel pools be maintained as low as reasonably achievable below 2.5 mrem /hr whole body dose during nomal operations. This design guideline can be met by maintaining a minimum of 10 to 12 feet of water over fuel in storage--enough water to reduce the gama dose rate from fuel assemblies to the range from 5 mrem /hr to 0.5 mrem /hr at the pool surface. However, dose rates at the surface of the pool with a typical water depth of 24 feet-are on the order of only 10-6 mrem /hr due to direct radiation from recently discharged fuel assemblies.
- 23. At the CP stage Applicants made the following comitment:
During all phases of spent fuel transfer, the gama dose rate at the surface of the water is 2.5 mr/hr or less. This is
accomplished by maintaining a minimum of 10 feet of water above the top of the fuel pellets in the fuel assembly during all handling operations.
(Source: Applicant's PSAR, Section 9.1.4.3.4, " Radiation Shielding")
To provide further assurance that a minimum 10 feet of water is maintained above the fuel, keeping dose rates below 2.5 mrem /hr, the Applicant has committed to provide limited maximum lift height for handling equipment used to raise and lower spent fuel.
(Source:
Applicant'sFSAR,Section9.1.4.1,"DesignBases"). The Applicants' assumptions and bases for shielding design and operations applicable to the SFSF were evaluated by the Staff, and the Applicants' design methods, including the use of source terms, cross section data, shield and source geometries, and radiation transport calculatioral schemes, were found to be consistent with accepted practice.
(Source: Catawba SER,Section12.3.2). The storage of spent fuel from McGuire and Oconee at Catawba does not impact the Applicants' commitments or the Staff's findings, since fuel storage assessments were based on recently discharged fuel'(and full capacity) rather than five year old spent fuel (and less than full capacity).
- 24. As noted in the FES (Table D.9), the Staff has estimated 480 person-rems as the total body dose to plant workers for a year of operation for a single unit at Catawba. Normal fuel handling operations in the fuel handling building for Catawba are expected to result in an average. total body dose of about 1.5 person-rems per year per unit. As stated in Section 5.9.3.1.2., additional handling of spent fuel from Oconee and McGuire is estimated to result in a total body dose of 0.029 person-rem per spent fuel shipment, or 8.7 person-rems for 300' shipments
(the maximum number of shipments per year proposed by Applicants).
Doses from fuel handling are thus only a small fraction of total dose j
for the Catawba facility.
25.
In summary, the Staff evaluated in the FES the environmental impacts of spent fuel stored in the Catawba SFSF, as expanded since the original construction pemit application. This evaluation included the operation of the Catawba SFSF as a storage facility for spent fuel from Oconee and McGuire. The Staff's evaluation is contained in the FES, at pp. 5-19, 9-7, 9-8, 9-12, 9-13, and in Appendix D.
A more detailed exposition of that analysis has been presented above.
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- 26. The conclusions of the Staff's evaluation are as follows:
(a) The releases of radioactive material from fuel stored at Catawba, including fuel from Oconee and McGuire, are estimated to be l
very small fractions of the total releases from normal operations at i
Catawba.
j (b) The Catawba effluent treatment systems as now designed and built are capable of controlling effluent releases, including releases from stored spent fuel from Oconee and McGuire, to meet the dose-design objectives of Appendix I to 10 CFR 50.
(c) The doses to individual members of the public and the general population exposed to effluents from fuel stored at Catawba are very small fractions of the annual doses from background radiation.
(d). Occupational doses attributable to spent fuel storage and handling operations, including handling and storage of spent fuel received from Oconee and McGuire are a small fraction of the total worker dose for the Catawba facility.
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(e) As a result, the proposed operation of the Catawba SFSF has been fully evaluated, to include receipt and storage of Oconee and McGuire spent fuel, and fcund to have a small impact on the environment.
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=r Edward F. Branagan,~ Jr.
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LLALA Richard Jo(serbu Subscribed and sworn to before me this SLS day of M wt.
_, 1983 NM Notary Public 4
My Comission Expires:
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.w J. S. Boeg11 Professional Qualifications Is n
.0ffice of Nuclear Reactor Regulation My name is J. S. Boeg11.
I am a lead nuclear engineer in the Effluent Treatment Systems Branch in the Office of Nuclear Reactor Regulation.
I attended Case Institute of Technology and was granted a B.S. in i
Chemical Engineering from Indiana Technical College in 1951.
In 1952. I received a M.S. Degree in Chemical Engineering from Kansas State College and in 1955 to 1956 I completed advanced courses in chemical and nuclear engineering at the University of Michigan and applied Health Physics training at the Oak Ridge National Laboratory.
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From 1953 to 1973 I was employed by the National Aeronautics and Space Administration and held positions as research engineer in heat and mass transfer, design engineer in nuclear reactor coolant, utilities.
. ventilation and radwaste systems, process systems super isor, and-technical consultant at the NASA Plum Brook Reactor in Ohio.
In July 1973, I joined the NRC (formerly AEC) as a senior nuclear engineer in the Effluent Treatment Systems Branch.
I evaluate nuclear power plant systems and equipment for fission product removal, treatment of gas, liquid and solid radioactive waste, and radiation safety proposals
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as provided by license applicants to meet applicable NRC regulations.
The dut.ies involve development analytical models and performing calcula-tions on the effectiveness of proposed radwaste systems, studying 5
technological improvements and developing criteria governing radwaste processing, monitoring, shielding and handling.
I evaluate the impact of radioactive eff1 dents on the environs and prepare the radwaste section 4
of the Environmental Statement for nuclear facilities.
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EDWARD F. BRANAGAN, JR.
l OFFICE OF NUCLE.G REACTOR REGULATION PROFESSIONAL QUALIFICATIONS
. From April 1979 to the present. I have been employed in the Radiological Assess-ment Branch in the Office of Nuclear Reactor Regulation of the U. 5. Nuclear Regulatory Commission (NRC). As a Health Physicist with the Radiological Assessment Branch, I am responsible for evaluating the environmental radio-logica: impacts resulting from the operation of nuclear power reactors. In particular, I en responsible for evaluating radio-ecological models and health effect models Mr use in reactor licensing.
In addition to my duties involving the evaluation of radiological impacts from nuclear reactors, my duties in the Radiological Assessment Branch have
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included the following:
(1) I managed and was the principal author of a report entitled " Staff Review of 'Radioecological Assessment of the Wyh1 Nuclear Power Plant"' (NUREG-D668); (2) I serve as a technical contact on an NRC contract with Argonne National Laboratory involving development of a
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computer program to calculate health effects from radiatior; (3) I serve as the project manager on an NRC contract with Idaho National Engineering Laboratory involving estimated and measured concentrations of radionuclides in the environment; (4) I serve as the project manager on an NRC contract with Lawrence Livermore Laboratory concerning a literature review of values for parameters in terrestrial radionuclide transport models; and (5) I serve as the project manager on an NRC contract with Dak Ridge National Laboratory concerning a statistical analysis of dose estimates via food pathways.
From 1976 to April 1979, I was employed by the NRC's Office of Nuclear Materials Safety and Safeguards, where I was involved in project management and technical work.
I served as the project manager for the NRC in connection with the NRC's estimation of radiation doses from radon-222 and radium-226 releases from uranium mills, in coordination with Dak Ridge National
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2 Laboratory which served as the NRC contractor. As part of my work on NRC's Generic Environmental Impact Statement on Uranita Milling (GEIS) I estimated health effects from uranium mill tailings. Upon publication of the GEIS, I presented a paper entitled " Health Effects of Uranium Mining and Milling for Commercial Nuclear Power" at a Conference on Health Implications of New Energy Technologies.
I received a B.A. in Physics from Catholic University in 1969, a M.A. in Science Teaching from Catholic University in 1970, and a Ph.D. in Radiation Biophysics from Kansas University in 1976. While completing my. course work for my Ph.D., 1 was an instructor of Radiation Technology at Haskell Junior College in Lawrence, Kansas. My doctoral research work was in the area of DNA base damage, and was supported by a U.S. Public Health Service traineeship; aqy doctoral dissertation was entitled " Nuclear Magnetic Resonance Spectroscopy of Gamma-Irradiated DNA Bases."
I am a member of the Health Physics Society.
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UNITED STATES OF AMERICA i
HUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD RICHARD JOHN SERBU PROFESSIONAL QUALIFICATIONS health physicist with the Radiation 1 am presently assigned as aProtection Section of the Radiological Assessm Integration, Office of Nuclear Reactor Regulation. U. 5. Nuclear Regulatory Commission.
I graduated from the State University College of New York at Pot:d:= w Bachelor of Arts Degree in Chemistry.the field of radiation protection reactors since June 1973.
From June 1973 to April 1980. I held positions as Project Engineer. Do Health Physics; Manager Radiological Monitoring; Project Engineer. Radi j
' Training; Radiological Controls Supervisor; and Instructor. Chemistry logical Controls at Knolls Atomic Power Laboratory. included dev This programs, operational health physics /ALARA programs, and dosimetry pro
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includes broad experience in all aspects of reactor health physics / radiation tection; familiarity with reactor systems; radiation protection aspects of reac startup; radiation protection for maintenance and refueling /overha.ul; che Since April of control programs; and compliance with established requirements.
1980. I have been with the Nuclear Regulatory Commission as a >;diolo In this capacity. I am responsible for the review and evaluation of radiation tection/ALARA (As Low As Reasonably Achievable) aspects of nuclear pow facility equipment and design, planning and procedure programs, an practices which are employed by nuc. lear reactor licensees and licen in meeting the standards for protection against radiation of 10 CFR Part 20 1
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