ML20031D612
| ML20031D612 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/28/1981 |
| From: | Yeon Kim CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), NUS CORP. |
| To: | |
| Shared Package | |
| ML20031D553 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8110130541 | |
| Download: ML20031D612 (27) | |
Text
/~3 STATEMENT OF YONG S.
KIM
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CONCERNING O'NEILL CONTENTION IIE-3 My name is Yong S.
Kim.
I reside at 210 Hillsboro Drive, Silver Spring, Maryland.
Since January 1, 1963, I have been associated with NUS Corporation, an engineering consulting firm in Rockville, Maryland.
I have primary responsibility within NUS Corporation for reactor physics analysis, including criticality analysis.
My resume, which is attached to this Statement, sets forth my educational background and work experience.
I am the author of the criticality analysis of the proposed expanded snent fuel pool at the Big Rock Point Plant as set forth in Section 4.0 of the Spent Fuel Tick Addition Description and Safety Analysis.
This analysis (hereinafter called the " Application") was submitted by Consumers Power Company on April 23, 1979, to the NRC Staff in support of the proposed expansion of the spent fuel pool capacity.
The detailed supporting information for Section 4.0 is set forth in NUS File G-RA-12, NUS File G-RA-12 Addendum No.
1, and Nemorandum EnSD-FS-194 dated March 18, 1980.
These documents were furnished to the parties and the Atomic Safety and Licensing Board in response to Intervenors' Interrogatories (Set I) 9-22 through 9-28 and 9-30; and as my affidavit, A
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WITED STATES OF MERICA.
NUCLEAR REGULNIORY CIM1ISSION i
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BEFORE TIE ATOMIC SAFETY AND LICENSING DOAPD p
In the Matter of
)
) Docket No. 50-155 OIA l
CONSGERS POWER COMPANY
) (Spent Fuel Pool i
) Modification)
-(Big Rock Point Nuclear Power Plant)
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AFFIDAVIT OF DR. YONG S. IGM i-County of Fbntgcmery ).
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State of bbryland
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I, Yong S. Kim, Executive Engineer, Nuclear Fuel Services i
Department at NUS Corporation, of lawful age, being first duly sworn,
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upon my oath certify that the statments and infomation contained I
in *i; +welve-page Statment concerning O'Neill Contention IIE-3, j
the ctrached resume, and Section 4.0 of the Spent Fuel Rack Addition Desc <ption and Safety Analysis, dated April 1979, are true and correct to the best of my knowledge and belief.
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Executed at Rockville, Furyland.
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. dated March 21, 1980, in support of the responses indicates, I am also the author of these materials.
Based on my educational background and work ex-i perience, I am qualified to address O'Neill Contention IIE-3, which stat J:
The application has not ade-quately analyzed the possibility of criticality occurring in the fuel pool because of the in-creased dr.nsity of storage with-out a gross distortion of the racks.
I.
INTRODUCTION U.
S.
Nuclear Regulatory Commission Standard Review Olan Section 9.1.2 states that the safety function of the spent fuel pool and storage racks is to maintain the i
spent fuel assemblies in a subcritical array during all credible storage conditions.
My Statement defines the terminology related to nuclear criticality, lista the general requirements of a criticality analysis, and discusses my specific analysis.
I show that the general requirements, as applicable to the Big Rock Point Plant, were met in my criticality analysis for the high-density Big Rock Point spent fuel racks.
The condition for a chain reaction in a system such as a wet pool of spent nuclear fuel can be conveniently
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- multiplication factor, represented by the symbol k-effective.
It is defined as the ratio of the average number of neutrons produced by fission in each generation to the total number
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of corresponding neutrons absorbed by fuel, moderator, and 2
other components of the reactor, and leaking out from the reactor, on the average.
If k-offective is unity (one), the l
number of neutrons produced is equal to the total number lost by absorption and leakage, and the chain is stationary j
and is self-sustaining; such a system is said to be critical.
If k-effective is less than unity, the chain is convergent and is not self-sustaining; this is designated as a subcritical system.
Finally, if k-effective is greater than unity, the chain is divergent, and the system is described as super-critical.
The effective multiplication factor can be regarded as being made up of two parts.
One is determined by the composition and arrangement of the materials present in the system, whereas the other is dependent upon its size.
The former, called the infinite multiplication factor and repre-sented by the symbol k-infinite, is the ratio of the average number of neutrons produced in each generation to the i
average number of corresponding neutrons absorbed.
In other words, it is essentially the value k-effective would have if
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It is obvious then that k-effective will always be less than k-infinite.
- Thus, the use of k-infinite instead of k-effective for a finite system such as a pool containing spent nuclear fuel assem-blies is conservative from the safety standpoint.
II.
CRITICALITY ANALYSIS REQUIREMENTS The fission reactions in a nuclear reactor cause the gradual depletion of the fissionable materials in fuel assemblics with time, anc the depleted or burned fuel assem-blies are eventually removed from the reactor and are stored in a spent fuel storage pool generally located at the reactor site.
In the spent fuel pool, the fuel assemblies are inserted into individual storage rack cells or cans that are spaced uniformly throughout the pool to maintain a subcritical condition.
The overall design objectives of a spent fuel storage facility at a reactor site are governed by Section 9.1.2 of the NRC's Standard Review Plan (hereafter referred to as " Standard Review Plan").
Additional guidance for the type and extent of information needed by the NRC Staff to perform the review of licensee-proposed modifications of an operating reactor spent fuel storage pool, and the acceptance O
OV critoria to be used by the NRC Staff in authorizing such modifications, are provided in "OT Position for Review and Acceptance of Spent Fuel Storage and llandling Applications,"
(hereafter referred to as "NRC Guidance").
The specific acceptance criteria applicable to the criticality analysis of spent fuel storage is stated as follows by the Standard Review Plan:
The Safety Analysis Report must show the center-to-center spacing between fuel assemblics in the storage racks is sufficient to main-tain the array, when fully loaded and flooded with non-borated water, in a subcritical condition.
A k-effective of less than about 0.95 for this condition is acceptable.
The NRC Guidance further expands the acceptance criteria for criticality as follows:
The neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all un-certainties under all con-ditions.
The NRC Guidance requires that the k-offective of the fuel storage pool be calculated under the following acts of assumed conditions:
a.
The storage racks shall be designed to contain the most reactive fuel authorized to be stored in the facility without any control rods or any non-A) contained burnabic poison, and the fuel shall be
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assumed to be at the most reactive point in its life.
b.
The moderator shall be assumed to be pure water at the temperature within the fuel pool limits which yields the largest k-effective.
c.
The array shall be assumed to be infinite in lateral extent or to be surrounded by an infinite-ly thick water reflector and thick concrete as appropriate to the design.
d.
Mechanical uncertainties may be treated by assuming
" worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties, c.
Credit shall be taken for the neutron abrorption in structural materials and in solid materials added specifically for neutron absorption.
-f.
Postulated accidents shall include:
(1) dropping of a fuel assembly on top of the racks and any other achievabic abnormal location of a fuel assembly in the pool, (2) dropping or tipping of the fuel cash or other i
I heavy objects into the fuel pool, (3) effect of tornado or earthquake on the-r deformation and relative position of the fuel
- racks,
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(4) loss of all cooling systems or flow under the accident conditions unless the cooling system is single-failure-proof.
g.
A calculational bias shall be determined from the comparison between calculation and experiment.
A calculational uncertainty shall be determined such that the true multiplication factor will be less than the calculated value with a 95 percent probability at a 95 percent confidence level.
The total uncertainty factor on k-effective shall be obtained by a statistical combination of the calculational and mechanical uncertainties.
The k-offective value for the racks shall be obtained by summing the calculated value, the calculational bias, and the total uncertainty.
III. CRITICALITY ANALYSIS The purpose of the criticality analysis is to ascertain that the value of k-effective of the Big Rock Point spent fuel storage pool with its entire fuel racks completely filled with spent fuel assemblies from the Big Rock Point reactor is well below unity, and thus the pool system is subcritical under the normal operating and ab-l normal conditions, including the effects of material and l
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mechanical tolerances, variation in operating parameters, calculational uncertaintics, and potential accident situa-i tions.
The criticality analysis is based on the input information provided by Consumers Power Company for fuel l
types that are to be stored in the newly designed high-density spent fuel storage racks.
A summary of the various j
Big Rock Point fuel types with major fuel parameters is set j
forth in the table attached to my Statement.
This informa-l.
tion was initially provided in handwritten form by Mr. Carl l
Larsen of Consumers Power Company in connection with my 4
response to Intervonors' Interrogatory 9-24 (Set I).
Mr.
Larson's affidavit, dated March 12, 1980, and which is also attached, attests to the accuracy of this information.
The handwritten information was subsequently typed in the form of the attached sic, and it was furnished to the parties and the Atomic Safety and Licensing Board by letter dated April 24, 1980, from Joseph Gallo to William S. Jordan, III.
In the criticality analysis, I considered all the 1
URC Guidance requirements that are applicabic to the analysis of the Big Rock Point spent fuel storage racks.
The detailed I
l information with regard to input design data, method of J
analysis, conditions analyzed, and results of analysis arc.
i presented in Section 4.0 of the Application.
Additional i
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O details are set forth in NUS Analysis File G-RA-12, Addendum No. 1 to Fila G-RA-12, and NUS' Internal Memorandum EnSD-FS-t 194.
The following paragraphs summarize the specific areas of analysis that were-performed by me.
The paragraphs will i
be numbered identically with the corresponding requirements in the NRC Guidance referred to above.
3 (a)
The nominal k-effective (0.8862) of the spent fuel pool was determined for the most reactive fuel i
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assembly under the spent fuel pool configuration l
and the fresh (undepleted) fuel condition was used 1
to represent the most reactive point in its life.
(b )
Pure water was assumed as the moderator in the i
pool, and the temperature of water (212 F) which l
gave the largest pool k-effective was used. lThis
. assumption yields an increase in the nominal k-l cffective of 0,0158..
(c)
To achieve a conservative analysis, all k-effective calculations were performed under the assumption of not only radial (lateral) infinity, but also axial (vertical) infinity.
Hence, k-effective and e
k-infinite are interchangeably used in the analysis, and therefore the results are conservative.
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(d)
All the mechanical tolerances (uncertainties) were I
l considered in-the calculations, and their positive O
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contributions to criticality were included in the determination of the final k-effective.
The i
mechanical tolerances considered were rack can center-to-center spacing, rack can size, rack can wall thickness, eccentric fuel loading, and minimum s
water gap size between rack cans.
In addition, the variations in rack can stainless steel com-3 position and fuel enrichment uncertainty on the k-effective were studied and incorporated into the final k-effective.
These assumptions yield an increase in the nominal k-offective of 0.0036 and 0.0233.
The values shown in (a), (b), and this item (d ),
are summarized in Subsection 4.4 of the Application.
(e)
The neutron absorption in stainless steel cans was taken into account in the nominal k-effective calculation.
Since the Big Rock Point fuel' racks are un-poisoned types, there are no solid materials added specifically for neutron absorption.
(f)
Twa types of fuel handling accidents were con-sidered: (1) a fuel assembly falling on the rack and landing horizontally on top of the rack, and (2) a fuel assembty being-inadvertently brought vertically next to the rack in a water gap between
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the rack assembly and the pool wall.
The re-sulting effects on the pool k-effective were found to be insignificant.
The effect of loss of all cooling system or flow under the accident con-ditions was considered by allowing the rise in coolant temperature and the formation of bubbles in the coolant due to boiling.
Its contribution is accounted for in the determination of the final k-effective (0.0044).
The effect of dropping or tipping of the fuel cask or other heavy objects into the fuel pool on the k-effective was not considered.
These types of accidents are precluded by administrative controls.
These matters are discussed in Subsectinn 4.6 of the Application.
Based on the Structural Analysis of the New Racks set forth in Section 5. 0 of the Application, the effect of earthquake on the deformation and relative position of fuel racks does not cause a significant change in the final k-effective.
The tornado effect has not been evaluated because that aspect of the NRC Guidance has not been imposed on the Big Rock Point spent fuel pool.
(g)
The k-effective uncertainty due to mechanical i
tolerances, including rack can stpir;ess steel
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composition variation, was determined by a sta-tistical combination using the root-mean-square summation method.
The calculational bias due to i
the calculation-experiment difference and the i
uncertainty related to the calculated value for a 95-percent probability at a 95-percent confidence 4-f level were calculated.
The sum of these two i
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values is 0.0167.
The ma::imum k-effective value for the racks was obtained by summing'the calculated values in (a), (b), (d), (f), and (g), and is 0.9500.
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IV.
CONCLUSION The foregoing discussion demonstrates that the j
. maximum k-ef fective of the spent fuel pool with fully loaded spent fuel racks meets the NRC acceptance criteria as applied to the Big Rock Point Plant.
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YONG S. KIM EDUCATION Catholic University of America, Ph.D.,1970 Massachusetts Institute of Technology, M.S., Nuclear Engineering 1961 University of Wisconsin, B.S., Chemical Engineering,1958 REGISTRATION Department of Defense Certified Fallout Shelter Analyst Registered Professional Engineer, State of Maryland,1977 Registered Professional Engineer, State of California,1977 l
EXPERIENCE NUS CORPORATION,1963-Present internuclear Company, 1961-1962 M.I.T. Department of Nuclear Engineering, 1958-1960 N US - Coordinates and pa rticipates in the nuclear criticality analysis activities for the design and licensing of poisoned and unpoisoned spent fuel storage racks. Performed nuclear analysis of advanced once through BWR design incorporating solid moderator as part of ACDA/ DOE uranium resource utilization program. Participated in nuclear criticality analysis and technical review of svern growth related to human intrusion of nuclear waste repository in a domed salt formation.
As Acting Manager of Nuclear Analysis, was responsible for technical work performed by nuclear analysis staff. including core design evaluation and analysis, core follow of operating reactors, computer code development, and training relative to nuclear, thermal hvdraulic, and mechanical behavior of fuelin nuclear reactors. In charge of reactor physics computer program development J
and applications. Adapted numerous nuclear and engineering computer programs to different I
types of digital computers ranging from the large scientific / engineering computers to minicompu.
ters, and performed improvement and modification of existing large computer codes. Managed and participated in the training of nuclear utility engineers in the area of in-core nuclear and thermal-hydraulic analyses and the use and installation of related computer codes. Involved,J*h bid evaluation of various commercial power reactors with regard to nuclear design. Performed a complete nuclear analysis of the shuffled core of a U.S. nuclear merchant ship, NS Savannah.
Performed nuclear design analysis, shielding calculations, and safety analysis of military power reactors, including PM 1, PM 2A, PM 3A, SM 1, SM 1 A, and MH 1 A.
Previously developed computer programs for calculation of radiation doses due to radioactive release frorn nuclear power plants of nuclear rocket latmch sites under 30tmal or accident conditions. Analyzed physics design of the Japan MaterialL sting Reactor, and calculated shield.
I ing requirements for proposed radiation exposure facility at the AFRRI Reactor. Evaluated and compared nuclear fuel costs of proposed power reactors to assist electric utilities in the selection of reactor types. Instructed in NUS Fuel Management Workshop courses.
Developed and programmed NULOC 2, Control Data 6600 code fr.,r multicompartment loss-of.
coolant accident analysis: NUTRIX and NUSIM, Control Data 6600 codes for three-dimensional physics analysis of operating reactors: NADAT 1 Honeywell DDP 516 assemblylanguage code for o
transfer of radiation data; WINDIF, wind diffusion program for evaluation of reactor sites and air I$
pollution for IBM 7090 and Control Data 3600; MOREDO, Control Data 3600 code to compute
- 3li external whole body gamma dose in any population due to reentering nuclear rockets (for Space A
Nuclear Propulsion Office); NEEP, program for IBM 7090 and Control Data 3600 to compute O
NUS COAPOAATION
V YONG S. KIM Page Two effective energy of radioactive isotopes deposited in internal body organs (for SNPO); NURSE, nuclear rocket safety evaluation prograrn for Control Data 3600 (for SNPO); FUELCOST 1, fuel cycle cut crogram for comparative study of fuel costs of different power reactors for IBM 7090 and Control ts ra 3600; EXGAM, code to predict integrated gamma dose for an airborne release of radioactivity for Control Data 3600.
Internuclear - Performed detailed nuclear calculations, including lifetime evaluation for a fully enriched boiling water reactor. Performed nuclear coiculations for the University of Missouri Research Reactor. Analyzed and programmed one-dimensional transport problem with mono-directional sources and an isotropic scattering for radiative transfer applications for IBM-7090 (ISOLATE). Performed heat transfer analysis of small nuclear reat.Mrs for activation analysis applications.
M.I.T. - As laboratory instructor, supervised graduate students performing experiments on reactor physics, radiation detection, and shielding. Made experimental measurement of degrada-tion rate of lucite physical properties in a reactor core; correlated degradation rate with radiation i
dose as part of the organicloop experiment at tne M.I.T. Reactor. Dec ied in pile radiation monitor instruments for this experiment.
MEMBER 3 HIPS American Nuclear Society Society cf Sigma A:
PUBLICATIONS "High Density Spent Fuel Storage Racks Desigr Analysis rieport, Kewaunee Nuclear Plant, Criticality Analysis," NUS 1931 Part D, August 1977.
" Core Analysis Procedures Manual," CD NA-76782, December 1976.
"NUMICE 2, A Spectrum Dependent Non-Spatial Cell Depletion Code " NUS-894, Revision 1 March 1976.
"NUSIM 3, A Digital Computer Program for Three Dimensional Nodal Reactor Simulation," March 1975.
"NUCELL - Cell Spectrum and Depletion Code Based on MUFT and THERMOS," May 1975.
"CYREP-ill, in Core Fuel Management Code " May 1975.
"NULOC 2, NUS Multi Compartment P essure Temperature Temperature History Program in Response to a Loss-of Coolant Accid';nt," NUS 1160, March 1974.
"NUCONTEMPT,NUSVersion of CONTEMPT PS for Prediction of Pressuro Temperature Response to a Loss of Coolant Accident," NUS 1164, March 1974.
"NUTRIX, A Digital Computer Program for Three Dimensional Analysis of Time-Cependent Operat-ing Reactor," NUS-657, March 1970, and Revision 1, March 1972.
l "FLYASH II, A CDC-6600 Computer Code to Calculate Time Dependent isotopic Inventory and Radioactive Disintegration Rates in a Nuclear Reactor," NUS-878. February 1972.
"NUFLOW 1, A Three Dimensional Nodal Core Analysis Computer Program with internally Calcu-lated Core Flow Distnbution," NL5 857, January 1972.
NUS COAPCAATION
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O STATE OF MICHIGAN)
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UNITED STATIS OF AMERICA 1
NUCLEAR REGULATORY C0tMISSION 4
BEFORE THE ATOMIC SAFETY AND LICENSING 30APS In the Matter of
)
CONSUMERS POWER CCMPANY
)
Docket No. 50-155
)
(Spent. Fuel Fool Expansion)
(Big Rock Point Nuclear Plant)
)
)
L A WIDAVIT OF CARL L. LARSEN
,1 I, Carl L. Larsen, of lawful age, being first duly sworn, do state as follows:
1 I am employed by Consumers Power Company as an engineering supervisor in the Generating Plant Modifications Department.
I have overall responsibill:y within the Company for technical, cost, and schedule aspects of :he proposed a
spent fuel pool expansion at the Big Rock Point Plant.
My resume is attached.
I have primary responsibility for answering In:arrogatories 9-32 through 1
9-37, 9-39, 9-40 and 9-41.
In addi:1on I contributed certain infor=ation on Big i
Rock Point fuel, fuel contraces, and storage locations in connection with In:ar-rogatories 9-24 (a), (d) and (f).
In doing so, I talked with Wilma Fogg, an engineering technician in Consumers Power Co=pany's Nuclear Activi:ies Depar:-
sent and with Dave Blanchard, a reactor engineer at Big Rock Point Planc.
I -
also discussed failed fuel and fuel sipping with Dave Blanchard in econection w1:h the response to Interrogatory 9-33.
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To the best of my knowledge and belief, the statements in this affidavit and in the respcases :o :h interrogatories listed above are : rue and correct.
/
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4
_A Ws Carl L. Lar en
/
Subscribed and sworn to before :ne this /4 day of March,1980.
1 s.i 19 Phyllis, Bogart j'
Notary Public, Jackscn County, Michig:n My Comission Expires:
February 24, '982 a
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CARL LEE LARSEN EEPERIE' ICE:
Consumers Power Co=pany since 1979 as Project Manager, responsible 1979 to for several major modifications to operating nuclear pcwer plants.
Present i
Responsibilities include technical, cost and schedule sspects in-cluding vendor selection and construction interface.
1974 to 1979 Gilberr/ Commonwealth Associates, Inc.1973 to 1979.
1978-1979 Senior Licensing Engineer responsible for the prepara. ton of security system design descriptions for a nuclear powet plant, responses.co USNRC questions on fire protection systens and technical suppor for hearings before the ACES concerning a CP stage license application for a nuclear power plant.
1976 - 1978 Responsible for the preparation of responses to USNRC questions during CP review of PSAP, license application.
Participated in technical meetings with USNRC Staff regarding USNRC questions on PSAR.
1975 - 1976 Lead Safety Licensing Engineer responsibla for coordinating the preparation of a Prel1=inary Safety Analysis Report for a nuclear power plant.
Responsibility for the licensability of technical infor=ation for the PSAR.
Coordinated the preparation of plant security docunents and fire protection and energency planning.
Performed NSSS vendor evaluation and liaison between client and NSSS vendor.
1974 - 1975 Performed technical and licensability review of PSAR and related sections of the Environmental Report.
Assisted with shielding design experiments conducted at the University of Michigan.
EDUCATION:
- 3. S. Nuclear Engineering, University of Michigan.
Graduate Engineering Studies, University of Michigan, i
SOCIEEIES:
American Nuclear Society O
~
O'NEILL CONTENTION IIF A.
THE CONTENTION l
Because of the expansion of the spent fuel pool routine releases, and accidental releases similar to those that have already occurred, of effluents will no longer meet the guidelines of Appendix I, Sections II and IV of 10 C.F.R. Part 50 because, in violation of Appendix I, Section IIIA.1, the required calculations do not esti-i mate bio-accumulation factors in a manner appropriate to this site.
B.
MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE ISSUE TO BE HEARD.
i 1.
Issuance of the requested license amendments authorizing an increase in spent fuel storage capacity would not change the fuel used at Big Rock Point or the method of reactor operation.
Therefore, operation of the spent fuel pool with additional stored spent fuel will not introduce different types of radioactive material into the spent fuel pool or into the environment.
(Affi-i davit of Roger W. Sinderman Concerning Bioaccumulation r
Factors at pp. 4, 16).
2.
Therefore, the bioaccumulation factors used in es-J tablishing 10 CFR Part 50 Appendix I limits for each radionuclide released in effluents from the Big Rock 8
Point Plant do not need to be changed due to the proposed increase in spent fuel storage capa' city.
(Affidavit of Roger Sinderman Concerning Bioaccumu-()
lation Factors at p.16).
(
3.
The 10 CPR Part 50 Appendix I limits for Big Rock Point, including the bioaccumulation factors used I
in establishing those limits have been reviewed and approved by the NRC Staff.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at p.3; Evaluation by the office of Nuclear Reactor Regula-tion of the Big Rock Point Plant Waste Treatment Sys-tems with respect to the Requirements of Appendix I I
to 10 CFR Part 50, dated May, 1981).
4.
For atmospheric releases, bioaccumulation factors for Big Rock Point were used in accordance with U.S.
NRC Regulatory Guides 1.109 and 1.111.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at p.7).
5.
There are no unusual food pathways in the Big Rock Point area not considered in Reg';1atory Guides 1.109 and 1.111.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at p.7).
6.
The pathways used to establish Appendix I limits take into account actual land use near Big Rock Point Plant, as determined by annual surveys.
(Affi-i davit of Roger Sinderman Concerning Bioaccumulation I
factors at pp. 7-8).
7.
For purposes of establishing current Appendix I
()
limits for each radionuclide released to Lake
, 1
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Michigan from Big Rock Point, internal exposure i
due to fish ingestion was one of the pathways con-1 l
sidered.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at pp.9-10).
1 8.
Bioaccumulation factors in Lake Michigan organisms for Big Rock Point Appendix I purposes were deter-mined based on research by Environmental Research Group of Ann Arbor, Michigan.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at pp.
1 10-14 and Sinderman Exhibits 4 and 5).
9.
Radioisotopes behave biochemically in a fashion sub-stantially identical to stable isotopes of the same element.
(Affidavit of Roger Sinderman Concerning Bioacc'umulation Factors at p.ll).
a 10.
Determining bioaccumulation factors for non-radio-active isotopes is equivalent to determining bio-accumulation factors for radioactive isotopes of the same elements.
(Affidavit of Roger Sinderman Con-cerning Bioaccumulation Factors at p.ll).
11.
The isotopes released to Lake Michigan from Big Rock Point in 1980 are shown in Table V on page 14 of Mr.
j Sinderman's Affidavit Concerning Bioaccumulation Factors.
12.
The list of elements in Table V is typical of radio-()
O V
active releases from the plant.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors I
l at p. 14).
13.
The appropriate bioaccumulation factors for Lake Michigan fish are shown in Table V.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at p. 14 and Sinderman Exhibits 4 and 5).
14.
These bicaccumulation factors are consistent with NRC Regulatory Guide 1.109.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at pp. 13-14).
15.
Licensee's environmental monitoring program period-ically samples the Lake Michigan aquatic environment to determine the buildup, if any, of radioactive mater-ials released from the Big Rock Point Plant.
The measured concentration of radioactive materials in algae, periphyton, crayfish, shore minnows and other fish are consistent with the bicaccumulation factors contained in U.S. NRC Regulatory Guide 1.109, and the ERG work.
(Affidavit of Roger Sinderman Concerning Bioaccumulation Factors at pp. 14-15).
C DISCUSSION As the affidavit of Roger Sinderman demonstrates, Licensee has done extensive research which confirms that the f ()
l
j bioaccumulation factors used in establishing 10 CFR Part 50 Appendix I limits are indeed appropriate to the Big Rock dointsite.
Further, the proposed increase in spent uel storage capacity does not lead to new and different radio-nuclides being released in plant effluents.
Therefore, no change to the bioaccumulation factors used in establishing Appendix I limits is necessary, and summary disposition is appropriate.
L e
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STATE OF MICHIGAN
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COUNTY OF JACKSON
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V UNITED STATES OF MERICA NUCLEAR REGULATORY COMMISSION l
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD Docket No 50-155 CONSUMERS POWER COMPANY
)
(Spent Fuel Fool Expansion)
(BIG ROCK POINT NUCLEAR PLANT )
AFFIDAVIT OF BrX',ER W SINDERMAN.
My name is Roger W Sinderman. My business address is 19k5 W Parnall Road, Jackson, Michigan. Since Msy 9, 1966, I have been employed by Consumers Power Company, an electric and gas service utility with headquarte s in Jackson, Michigan. I am currently Director of Radiological Services for Consumers Power Company. Prior to this job, I have been involved in various radiation protection and environmental health physics activities at the Comnany's Big Rock Point Plant and the Palisades Plants, as well as at the General Office in Jackson.
I have an MPH and MS in Health Physics from the University of Michigan. My resume is attached.
I am the author of the documents entitled " Testimony of Roger W Sinderman Concerning Bioaccumulation Factors" and " Testimony of Roger W Sinderman Concerning Radiation Doses to the Public From Fuel Stored Near the South Wall of the Spent Fuel Pool".
Where statements of fact are made in this test.imony, they are based on my personal knowledge. Where results of calculations -re discussed or attached as dxhibits, these calculatiens were performed and reviewed for accuracy by employees undi r my supervision and I have also personally reviewed and verif4ed the calculations.
I was a member of the Lake Michigan Utility Study Group which sponsored and reviewed the Environmental Research Group reports which are attached to my testimony on n
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bioaccumulation factors.
I mm personally familiar with the research summarized in those reports.
In addition, those reports are standard reference materials
..-. - -.. =.__-.
1 v-which are available in most major university librarys. I believe I am qualified 4
to interpret such reports by training, experience, and personal familiarity with the way the research was done.
l I swear that this affidavit and the testimony and exhibit $ attached thereto are true and correct, to the best of my knowledge and belief.
CTic,h d
Roger W SinBe'rman
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i SUBSCRIBED AND SWO N TO before me this ay of September, 1981.
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.0GER WILLIAM SIUDEPMAN l
Education:
B.S. Science Engineering, University of Michigan M.S. Health Physics, University of Michigan i
M.P.H. Health Physics, University of Michigan Experience:
Consumera Power Company as Director of Radiological Services June 1981 to responsible for all aspects of radiological control at Consumers Present Power Company nuclear facilities. These responsibilitics include radiation exposure to employees, environmental surveillance, radioactive vaste, effluent control and emergency planning.
I 1974-1981 Consumers Power Company as Corporate Health Physicist i
responsible for all aspects of radiological control at Consumers Power Company nuclear facilities. These i
responsibilities include radiation exposure to employees, environmental surveillance und effluent control.
!i 1973-197h Connumers Power Company as Palisades Plant Health Physic!.
(6 month responsible for radiation protection, effluent and environmental
{
period) centrol at the Palisades Plant.
I 1971-1973 consu:sers Power Company as Environmental Health Physicist responsible for environmental radiological surveillance and control of radiological effluents from the Company's l
r.uclear facilities.
1968-1971 Consumers Power Company as Health Physicist responsible for
^
Dog Rock Point Plant radiological centrol and Palisades Plant construction activities related to radiation protection.
l l
1966-1968 Consumers Power Company as Associate Engineer, General Engineer, l
and Cnemical and Radiation Protection Supervisor at the 1;
Big Rock Point Plant responsitle for Plant radiation protection activities and various enginecring tasks.
i Societies:
Health Physics Society l
j American Public Health Association 4
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