ML20033B490
| ML20033B490 | |
| Person / Time | |
|---|---|
| Site: | Atlantic Nuclear Power Plant  |
| Issue date: | 11/27/1981 |
| From: | Anderson C, Rachel Johnson, Rossi C Office of Nuclear Reactor Regulation |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20033B480 | List: |
| References | |
| NUDOCS 8112010462 | |
| Download: ML20033B490 (32) | |
Text
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g}d UNITED STATES OF AMERICA NUCLEAR REGULATORY COM!11SSION
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BEFORE THE AT0f1IC SAFETY AND LICENSING BOARD In the flatter of
)
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0FFSH0RE POWER SYSTEMS
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Docket flo. STN 50-437
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(Floating Nuclear Power Plants)
)
NRC STAFF'S RESPONSES TO NOVEftBER 12, 1981 BOARD QUESTIONS 2 THROUGH 7 Q.2. With respect to unrosolved generic issues in Appendix C to Supplement 4 to the SER, the Staff concluded that (a) as to some issues there is reasoncble assurance that the FNPs can be manufactured before these generic issues have been resolved without endangering the health and safety of the public, (b) as to some issues that FNPs can be nanufactured before these generic issues have been resolved, without endangering the health and safety of the public, (c) as to some issues that the FNPs can be manufactured and operated before these generic issues have been resolved, without undue risk to the health and safety of the public, and (d) as to some issues that a satis-factory solution will be incorporated in the FNPs before they are placed in operation, and, therefore, there is reasonable assurance that FNPs can be manufactured and operated, without undue risk to the health and safety of the public? What were the reasons for wording these conclusions differently?
A.2. The difference in the wording of these conclusions was inad-vertent. There was no intent to reach different conclusions for the different issues.
A single conclusion equally applicable to each of the issues is that there is reasonable assurance that the FNPs can be nanu-factured and operated before these generic issues have been resolved, without undue risk to the hulth and safety of the public.
Author - Clifford Anderson 8112010462 811127' PDR ADOCK 05000437 T
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1 Q.3. In light of the information required to be fnenished by Gulf States Utilities Co. (River Bend Station, Units 1 and 2),
ALAB-444, 6 NRC 760, 775 (1977), with regard to the Staff's
. discussion of unresolved safety issues, Task flumbers A-12 and A-47, in Supplement 4 to SER, what alternative courses of action might be available should the programs for each not produce the envisaged results?
A.3. These two Unresolved Safety Issues (USI) are managed by the Generic Issues Branch of the Division of Safety Technology in the Office of Nuclear P,eactor Regulation. They will be dealt with separately below.
A technical progran aimed at the resolution of the A-12 safety issue was completed in 1979. The proposed resolution was published in NUREG-0577, For Comment. As noted in that document, there are two separate failure modes of support structures to be considered, brittle fracture _ under accident loads and lamellar tearing. The two are discussed separately in the following text.
1.
Brittle Fracture The basis for prevention of brittle fracture under accident loads, was discussed in detail in 14UREG-0577, For Comment. The recommended resolution for new plants was to require that the ductile-to-brittle fracture mode transition temperature in the steels selected for a given support structure be signiricantly lower than the lowest service tempera ture.
Conservatively, that would be the appropriate ambient tempera ture.
Such an approach is nothing more than the appreciation of what now is regarded as common knowledge in materials engineering.
Nunerous failure analyses (of pressure vessels, bridges, surface ships, penstocks, etc.) have shown that failure to adhere to the recommendation is the most common root cause of brittle fracture. tioreover, there is no need to consider technical alternatives since the proposed resolution is
. effective, easy to implement in new plants and involves a minimal incremental cost. The proposed resolution (i.e., NUREG-0577, For Comment) was reviewed and approved by the NRC Staff and management. A final, modified, version is being prepared to address the public's comments with publication planned for January 1982.
2.
Lamellar Tearing The NRC Staff determined that the well-known phenonenon of lamellar tearing has been found to be the root cause of only one known failure, based on an extensive literature survey.
Also, it was noted that lamellar tearing is a construction problem (it is tatally unexpected as a result of operational loadings) and as such should be detected and repaired as a result of routine pre-service inspections. The Electric Power Research Institute has expressed its intent to conduct research which will lead to a resolution of the lamellar tearing problem as a generic basis.
The Staff has concluded that the research program will resolve the outstanding questions regarding lamellar tearing, and has determined that continued licensing and operation are justified during the course of the progran. This is based on the knowledge that lamellar tearing is not as urgent a problem as previously contemplated (based on the lack of service failures) and the Staff's conclusion that the likelihood of support failure due to lamellar tearing is low.
The latter conclusion is drawn from the knowledge that applied stresses during operation are low and the probability of an initiating event (imparting large stresses to a torn joint) is very low. The Staff also considers lamellar tearing to be a
', lower order failure mechanism than others that are possible in heavy weldments (e_.S., weld toe cracking).
Based on these considerations, the Staff detemined that lanellar tearing as a generic issue could be separated from the A-12 generic task.
Furthemore, the Staff has concluded that action by licensees and applicants regarding lamellar tearing may be deferred until the research program has been completed.
If this program should provide unexpected and unfavorable infomation regarding residual strength of lamellar-torn joints, the Staff will take aporopriate action such as requiring inspection (and repair, if necessary) of applicable supports.
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The purpose of Task A-47 is to either confim that existing criteria for non-safety grade control systems are adequate or to propose additional criteria for implementation.
As noted in Supplement 4 of the SER, systematic reviews of the safety systems are perfomed with the goal of ensuring that control system failure (single or multiple) will not defeat safety system action.
This is accomplished by verifying that adequate isolation devices between safety and non-safety systems have been provided, or that adequate physical separation between these systems is provided to preclude the propagation of non-safety equipment faults to the protection systen.
These reviews are performed utilizing, in whole or in part, the guidelines and criter!a identified in the Standard Review Plan (NUREG 75-087), Section 7.7.
In addition, a specific set of hypothetical accidents are analyzed to demonstrate that plant trip and/or safety system equipment actuation occurs with sufficient capability and on a time scale such that the
, potential consequences to the health and safety of the public are within acceptable limits. These conservative analyses are performed for a wide
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range of postulated events even though specific actual events might not j
be represented by the assumptions made in the analyses.
With the emphasis on the availability of post-accident instrumentation, the Staff review of FNP at the operating license stage will also evaluate the designs to assure that control system failures will not deprive the operator of information required to maintain the plant in a safe shutdown condition after an accident.
For recent operating license reviews the applicants have been requested to evaluate their control systems and identify any contrcl system whose malfunction could seriously impact safety. The applicants have been requested to identify the use (if any) of common power supplies, and the use of common sensors or common sensor impulse lines whose failure could have potential safety significance. The results of these reviews are (or will be) documented in the Safety Evaluation Reports on a case-by-case basis.
Also, current operating license applicants have been requested to review the possibility of consequential control system failures which could exacerbate the effects of high energy line breaks (HELB) and adopt design changes or new operator procedures where needed, to assure that the postulated events would be adequately mitigated. The Staff reviews are documented in the Safety Evaluation Reports on a case-by-case basis.
In addition, as part of the recent OL licensing reviews, applicants have been required to perform evaluations on a case-by-case basis to ensure the adequacy of plant procedures for accomplishing shutdown upon
. loss of power to any single electrical bus supplying power for the instruments and control systems.
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These reviews have not as yet identified the need for any additional generic requirements.
Based on these ongoing case review activities and experience on operating plants, the Staff has concluded that FNP can be nanufactured and operated before this generic issue has been resolved without undue risk to the health and safety of the public.
In the event that Task A-47 identifies some potentially significant scenarios or criteria related to control systems that have not explicitly been considered in the design and operation of FNPs, corrective measures will be required at that time.
Corrective neasures could include,.for example, a requirement for additional safety system functions in order to provide protection for control system failures.
Authors - Richard E. Johnson (A-12)
Charles E. Rossi (A-47)
Q.4. What changes in the magnesium oxide core ladle would be necessary to provide complete containment of a molten core? Is such a design feasible for the proposed floating nuclear plant?
A.4. It is important to first recognize that the requirements for the FNP core ladle came about through the environmental review rather than the safety review.
In order to assure that floating nuclear plants would be comparable to land based plants in the highly unlikely event of a core melt accident, it was necessary to provide increased resistance to a molten core penetrating the FNP hull structure.
It was determined that in order to ensure such comparability, it was necessary to delay melt-through of the FNP core for several days. Through its investigation, the
. Staff detemined that such a result could be accomplished by installation of a passive sacrificial bed of a refractory material like lig0. Thus, the Staff's Condition 4 of FES-III, NUREG-0502, dated December 1978, was based on the goal of delaying core melt-through, rather than pemanent retention.
The changes that would be required in the FNP core ladle design to provide pemanent retention of core debris have not been evaluated in detail by the NRC Staff. Therefore, we are not. in a position to state with any degree of specificity what changes would be required, including concommitant ramifications on the overall FNP containment and hull structure design. Accordingly, without having performed such detailed evaluations, we are not able to make a sound judgment on the feasibility of such a design for the proposed FNPs.
Our very preliminary general view on the design concept for pemanent retention of core debris in FNPs is that pemanent retention would necessitate incorporating a coolant system to dissipate heat from the extremities of the core ladle refractory material. This heat dissipation capability must be high enough to prevent either melt-through of the core ladle material, or exceeding temperature thresholds where the supporting structures would fail mechanically. The type of coolant, mode of heat transfer (e.g., natural convection or forced convection) and optimum thickness of the refractory core ladle material are questions that need detailed evaluatinn.
In addition, as reviously mentioned, the v
concommitant ramifications on the overall FNP containment and hull structure design would also require detailed evaluation.
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Author - Andrew R.11archese Q.5.A.
Explain in greater detail the differences between the density driven case the Kulacki-Goldstein case for heat transfer in the molten pool.
A.5.A.
The Kulacki-Goldstein correlation was obtained from experiments in a horizontal fluid layer with uniform volumetric energy sources bounded horizontally by two _ rigid planes of constant and equal temperature. The experiments did not consider the influence of a melting interface nor was there any mass addition to the horizontal fluid layer.
The experiments were restricted to calculating thermal convection due to density differences caused by thermal gradients in the fluid layer. The Farhadieh-Baker correlation (called density driven in the SER Supplement 3) was obtained from an experimental study, which was conducted on a melting systen consisting of a liquid layer overlying a solid substate. The liquid and solid when molten were mutually misible.
For experiments where the overlying liquid was more dense than the molten solid the system was found to be gravitationally unstable.
Consequently, the heat transfer coefficient for this system was found to be an increasing function of the density ratio of the liquid and melted solid.
The Farhadieh-Baker correlation results in a mas ked enhancement of heat transfer relative to the Kulacki-Goldstein correlation.
Author William Trevor Pratt Q.5.B.
Describe qualitatively the behavior at the interface between the magnesium oxide and the molten uranium oxide in the core ladle, particularly at the bottom of the pool.
A.S.B.
The nature of the interface between the molten core debris and the lig0 core ladle in a passive system changes with time during the thermal transient.
Immediately upon contact of molten UO and cold fig 0 a 2
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crust will be formed of solidified U0 at the interface. Later, as the 2
itgG begins to heat up, melting of the crust will commence at the lig0-UO2 utectic t'emperature.
It should be noted that the fig 0 nicrostructure.
consists of grains held together by a binding phase that nelts at slightly lower temperature then the grains. Thus, the binding phase will be attacked initially by the U0. The molten layer will initially be a 2
slurry consisting of solid particles suspended in a liquid having the utectic composition.
As the (10 melts and accumulates at the interface, 9
the temperature required to melt additional 110 will increase. This is 9
because UO nust diffuse through the molten layer to reach the Mg0 2
surface. Thus, the melting rate becomes a function of the diffusion process.
As the molten layer is diluted with t10, the melting 9
temperature will approach the nelting point of pure Mg0.
with increases in temperature, eventually the slurry will disappear. Heat transfer at the interface during the " slurry" phase is an extremely complex problem, the interfacial temperatures and temperature distribution within the slurry being the primary unknowns. The presence of slag materials at both the botton and the top of the pool may further complicate the heat transfer processes.
It may be possible, however, to model heat transfer through the slurry layer as simple conduction in a slab while using the density driven correlation between the pool and the upper surface of the slurry.
Authors William Trevor Pratt and David G. Swanson Q.5.C.
Discuss the possibility that pits will fom in the pool bottom and grow preferentially.
. A.S.C.
This concern was addressed in the SER supplement 3 and is really in two parts.
First, there was a concern that during the manufacturing process, impurities in the lig0 would segregate to the boundaries of the lig0 grains in the brick. The UO would then 2
preferentially attack this intergranular naterial at somewhat lower teaperatures than the nelting point of pure lig0.
This would produce a thin reaction layer between the attack plane and the fully liquid pool.
This phenomena is discussed in more detail in the response to Question 5.B.
Specimens from snall scale molten UO -lig0 tests (with a 3 inch 2
diameter) at the Aerospace Corporation did not show signs of pitting.
Preferential penetration of the low melting intergranular binding phase was observed but was quite linited.
The only available data for large scale 1190 structures is fran the steel industry.
In Basic 0xygen Furnaces, where iron oxide particles can attain temperatures as high as 4500'F, lig0 liners erode at the rate of 24 inches /1200-2000 hours of operation (approximately 1200-2000 heats lasting an hour each).
A second concern related to the possibility that UO could flow 2
between the lig0 bricks. Should such a localized attack penetrate and dislodge a few bricks, then the concern was that brick flotation could occur. The ladle uses a tongue-in-groove design to ninimize the effect cf localized attacks.
Section V.H of the SER supplement 2 discusses industrial experience related to localized attacks.
Authors - Villian Trevor Pratt and David G. Swanson
1 Q.5.D.
Is' the range of measurements described in the Kulacki-Goldstein paper appropriate to the situation in the molten core ladle?
A.S.D.
We noted above that the Kulacki-Goldstein correlation was obtained from experiments in a horizontal fluid layer bounded by rigid planes. This is obviously not the case in the core ladle.
At the bottom and sides of the ladle the UO will be melting the-fig 0.
The U0 and lig0 2
2 are misible and the UO is more dense than the lig0.
Consequently, heat 2
transfer'to the bottom and sides of-the ladle more closely resemble the conditions in the Farhadieh-Baker experiments. This is why the Farhabieh-Baker correlation (called density driven in the SER supplement 3) was the preferred correlation for heat transfer to the bottom and sides of the ladle (used in all but 2 of the~ cases in the SER supplement 3). However, upward heat transfer in the molten pool was to the underside of a crust. This configuration we consider to nore closely resemble the Kulacki-Goldstein experimental system so that upward heat transfer was assumed be given by the upward facing Kulacki-Goldstein correlation.
Finally, it was noted in the SER supplement 3 that the Farhadieh-Baker correlation was obtained by reference to a horizontal melting systen and that application of this correlation to the side of the ladle (vertical melting system) was uncertain.
Further experimental.
work was suggested in this area.
Author - William Trevor Pratt Q.5.E.
Table 5 of SER Supplement 3 reflects that two of the analyzed sixteen cases show that the pool freezes.
Explain in qualitative terms why that happens.
A.5.E.
The two cases that predict pool freezing in the SER supplement 3 assumed the ablating temperature of the lig0 to be at the
, melting point of the 110-U0 eutectic (4136*F) rather than at the nelting 9
2 point of the pure Mg0 (5072*F).
These two cases were an attempt to scope the uncertainty associated with the !!g0-UO interface discussed above.
2 The temperature of the molten pool tends to asymtotically approach the assumed ablating temperature of the 110-U0 interface. The bulk freezing 9
2 temperature of the lig0-UO mixture follows the idealized phase diagram 2
shown in Figure 8 of the SER supplement 3.
If a low lig0 ablating temperature is assumed and small quantities of molten lig0 are added to the UO2 (the pool freezing point remains high) then the pool temperature transient could result in solidification of the 110-U0 mixture. This 9
2 was observed in two cases in the SER supplement 3.
Author - William Trevor Pratt Q.6. Explain why convective heat transfer by air was not considered by the Staff in its thermal analysis of the molten pool.
Following a core melt scenario, would there be any air exchange between the chamber of the core ladle and the balance of the containment?
A.6. Heat transport by radiation from the molten pool to structures above the pool would dominate any natural convection cooling effect by air. A mechanism for removing heat by convection from the ' reactor cavity to the upper containment would exist, but the Staff has not evaluated this effect at this time.
By not including this effect, the themal threat to the ladle and structures above the ladle is maximized. One of the principal aims of Supplement No. 3 (NUREG-0054, dated February 1980) was to assess the thermal performance of the core ladle. However, the Staff indicated in Section V.F. that as part of a longer-term effort associated with the final design evaluation of the FNP core ladle, we L
. would perfom more detailed integrated studies that will in;olve modeling the containment and core ladle as a system to ascertain the system
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response to various core melt sequences.
Such s+udies would include the effect of heat transport by air exchange between the core ladle region and the upper containment building.
Authors - William Trevor Pratt and Andrew R.11archese Q.7. What changes would be required in the OPS design to meet the requirements of the proposed final rule on hydrogen control which, as modified by the Staff at the Commission's request, is now being considered by the Commission?
A.7. Proposed rule 10 C.F.R. 50.34(F)(3)(v) requires an applicant for a manufacturing license or construction permit to:
Provide preliminary design information at a level of detail consistent with that normally required at the construction pemit stage of review sufficient to denonstrate that:
(A) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the ASt1E Boiler and Pressure Vessel Code,Section III, Division 1, Subsub-article NE-3220, Service Level C Lirr1ts, except that evaluation of instability is not required, considering pressure and t.ead load alone.
For concrete containments by meeting the requirements of the ASf1E Boiler Pressure Vessel Code,Section III, Division 2 Subsub-article CC-3720, Factored Load Category, considering pressure and dead load alone) during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting agent.
As a minimum, the specific code requirements set forth above appropriate for each type of containment will be met for a combination of dead load and an
14 -
internal pressure of 45 psig. Modest deviations from these criteria will -be considered by the staff, if good cause is shown by an applicant. Systems necessary to ensure containment integrity shall also be demonstrited to perfom their function under these conditions.
(B)
(1) Containment structure loadings produced by an inadvertent full actuation of a post-accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis accident loadings will not produce stresses in steel containments in excess of the limits set forth in the AS!!E Boiler and Pressure Vessel Code,Section III, Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Subsubarticle CC-3720, Service Load Category), (2) The containment has the capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respectively) the pressure cal-culated to result from carbon dioxide inerting.
This rule is intended to insure containment integrity in the event of a release of hydrogen and its subsequent burning or inerting following a 100% metal / water reaction. The burning of hydrogen or containment atmosphere inerting as hydrogen control methods would produce a pressure rise in the containment atmosphere which, prior to the Tiil accident, was not a design consideration.
The Floating Nuclear Plant containment is designed to withstand several different load combinations associated with ASi1E code service levels. The plant containment design internal pressure is 15 psig using Service Level A limits of the ASME Boiler and Pressure Vessel Code, Y
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As set forth above, the proposed rule requh es a minimum containment internal design pressure of 45 psig,in combination with dead l'oad, Service Level C Code limits. The Code provisions for instability are not imposed for the 45 psig, Service Level C condition.
Service Level A is that nomal operating condition which applies factors of safety consistent with the expectation that the c"ents to which this level is assigned will actually occur. That fs, they represent the performance of normal service functions. 5fnce the occurrence of such events has been anticipated and fully evaluated in the design, no operational action is required should the event occur.
Service Level C is an emergency operating condition resulting from the occurrence of low probability phenomena for which safe shutdown of the plant is required.
The CP/ML requirement of 45 psig, Service Level C, produced two changes in the OPS containment design, namely (1) between elevation 162'2" and 199'4" the steel shell thickness was increased from 7/8" to 1" and (2) between elevation 199'4" and 224'0" the steel shell thickness was incresed from 1/2" to 1".
These are the only changes necessary to satisfy the 45 psig, Service Level C requirements.
A recent change was made to the OPS containment design requirements to add an additional load combination, namely 25 psig, in combination with ambient tem 9erature and dead load, Service Level A stress limits.
This resulted in two additional changes to the containment shell, namely the stiffening of (1) the equipment access hatch and (2) the containment dome in the knuckle region just above the top of the cylinderical section.
Author, Harold E. Polk
PROFESSIONAL QUALIFICATIONS
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OF CLIFFORD ANDERSON EXPERIENCE April 1980 to Present:
I am presently a Task Manager with the Generic Issues Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation, Headquarters (Washington, D.C.), U.S. Nuclear Regulatory Commission.
For the past five years I have served as the Task Manager of the NRC program to resolve the A-8 Unresolved Safety Issue (USI), " Mark II Containment Pool Dynamic Loads."
I have recently been designated Task Manager of the A-48 USI, " Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment."
As task manager, my primary duties are to provide the technical management over the NRC technical staff and contractor technical assistance leading to the resolution of these issues. Other duties include the preparation and coordination of the appendices for many of the Safety Evaluation Reports which address the status of USIs with respect to particular plants.
February 1973 to April 1980:
I served as a Senior Containment Systems Engineer within the Containment Systems Branch in the Office of Nuclear Reactor Regulation. My primary duty was to perform safety reviews of containment systems for specific plants.
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August 1970 to February 1973:
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I was employed by Nuclear Fuel Services as a Senior Engineer in the thermal-hydraulic section systers group.
I worked on a variety of projects including the development of BWR and PWR reload fuel, the design of spent fuel shipping casks and the expansion of the NFS West Valley reprocessing plant.
October 1965 to August 1970:
I I wu Ked at Hillman Associates, an engineering consulting company in Columbia, Ma ryland.
I worked first as an engineer and later as a group leader in the heat transfer and the thermal hydraulic analysis section of the reactor power and development department.
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- August 1965 to October 1965:
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I worked as an engineer in the nuclear safety division at the CANEL-Connecticut Advanced Nuclear Engineering Laboratory in !!iddletown, Connecticut.
June 1962 to August 1964:
While I was a graduate student at North Carolina State in Raleigh, North Carolina, I worked half time as a lab instructor teaching physics laboratory to freshman and sophomore engineering students.
I also served a sunmer research internship during the summer of 1962 at Brookhaven National Laboratory.
EDUCATION My educational background includes 40 credit hours of graduate work at North Carolina State University at Raleigh, North Carolina, in the nuclear engineering department during the period of September 1962 through August 1964 and a Bachelor of Scien:e in Nuclear Science from the State University of New York at Fort Schuyler obtained during the period of September 1958 through June 1962.
A list of other special training I have received is attached.
HONORS AND PUBLICATIONS In 1978 I received a high quality certificate fron the U.S. Nuclear Regulatory Commission for work associated with the generic review of pressure suppression cortainments.
A list of publications and patents I have authored or co-authored is attached.
/
. PUBLICATI0llS AND PATENTS
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" Mark II Containment Program Load Evaluation and Acceptance Criteria,"
11UREG-0808, August 1981
" Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," NUREG-0487, October 1978 A Technical Update on Pressure Suppression Type Containments in U.S. Light Water Reacto.- Nuclear Power Plants," NUREG-0474, July 1978 (Major Contributor)
" Transportation of Irradiated Pu-Recycle Fuel," Presented at 1971 Boston ANS Meeting (Coauthor)
Patent Applications P1288 and P1290 - Spent Fuel Shipping Containers,1971 New Form Technical Specifications as Applied to First Generation Nuclear Power Plants," Reactor Safety,1968 (Coauthor)
" Thermal Bowing in Pin-Type Fuel Elements," Reactor Safety,1968 (Coauthor)
" Predicted Performance of Flat Plate Solar Thermoelectric Energy Conversion Panels," Presented at the Thermoelectric Specialists Conference, May 1966 (Coauthor)
"Ax-TNT, A Code for the Investigation of Fast Reactor Excursions and Blast Waves from a Spherical Charge of TNT," Pratt and Whitney Aircraft (CANEL),
TIM 950, 1965 OTHER SCHOOLS OR TRAINING Westinghouse PWR Systems Course at Westinghouse Training Facility, Pittsburgh, PA, 9/24/73-10/5/73 Safety of Light Water Cooled Nuclear Power Plants, Northwestern University, 9/8/75-9/12/75 Fast Reactor Safety Course, Massachusetts Institute of Tecnnology, 7/26/76-7/30/76 Task Force and Project Management, Civil Service Comission, S/1/78-5/5/78 BWR Technology Course (R204B), July 15-25,1983 BWR Simulator Prep Course (101B), August 6-8, 1980 BWR Simulator Course (603B, August 11-15, 1980
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PROFESSIONAL QUALIFICATION OF DR. RICHARD E. JOHNSON I am a senior engineer in the position of Unresolved Safety Issue (USI) Task Manager in the Generic Issues Branch of the Division of Safety Technology, Office of Nuclear Reactor Regulation (NRR), U. S.
Nuclear Regulatory Cmmission.
Currently, I am the manager of Tasks A-11 and A-12, " Reactor Vessel Materials Toughness" and " Fracture Toughness ~
and Potential for Lamellar Tearing of Steam Generator and Reactor Coolant Pump Supports," respectively. Also, I am the principal materials reviewer for Task A-10. "BWR Feedwater and CRD Hozzle Cracking," and for the generic issue of reactor pressure vessel themal shock, which has been
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recmmended as a new U5I.
Since 1976, I have worked as a materials engineer in NRR dealing with a great variety of operating plant problems.
For more than 25 years I have been involved in materials engineering aspects of failure and failure prevention, becming a nationally-recognized expert in engineering fracture nechanics. My doctoral dissertation
- dealt with the subject of fracture mechanics and over the years, I,have been active professionally in the ASTM (working with Committee E-24 on f racture nechanics), the AS'1E (working with the subgroup on evaluation of Section XI,' Boiler and Pressure Vessel Code Conmittee), as well as being a member of the AIME (Metallurgical Society) and the Society of Sigma Xi.
Fran time to time, I have taught and lectured on the subjects of mechanics of materials and fracture mechanics including a period 'when I was Adjunct Professor of' Engineering at Wright State University. A list of publications would include more than one hundred items, the most recent of which is NUREG-0744, " Resolution of the Reactor Vessel Materials Toughness Safety Issue."
- "The Influence of the System Gemetry on Fracture Toughness" GD
CHARLES E ROSSI 4. ; ;
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I have been with the' U. S. Nuclear Regulatory Comission (NRC) since m
s October 1980.
Since August,1981 I have been a Section Lead'er in the -
Instrumentation and Control Systems Branch, Division of Systems Integration.
Office of Nuclear Reactor Regulation.
I am responsible for supervising the review of nucl' ear power plant instrumentation and contr'o1 system designs for compliance with regulatory criteria.
From October 1980 to August 1981 I was a Principal Reactor Engineer in the Instrumentation and Control Systems Branch.
I performed the operating license revied of the Callaway ~and Wolf Creek instrumentation and control system designs, the review of construction permit applicant responses to Three Mile Island Lessons Learned Items ielated to instrumentation and' control systems, and the review of licensee responses to recomendationf made by Babcock and Hilcox resulting from failure modes and effects analyses of the Integrated Control System.
I have a.Ph.D degree (1969) and M.E degree (1967) in Applied Physics from Harvard University, a M.S degree (1962) in Physics from George Washington University and a B.A degree Magna cum Laude Highest Honors (1958).in Engineering and Applied Physics fr.pm Harvard University.
I have a certificate from a six month reactor engineering course given b'y the Bettis At,omic Powir Laborato'ry (1960).
I was elected to Phi Beta Kappa in 1958 and Sigma Xi in 1962.
From June 1958' to July 1962 I served as a comissioned officer in the United States Navy.
I was assigned to Naval Reactors, U. S. Atomic Energy Comission, whire I reviewed and approved test and operating procedures for submarine nuclear power plant fluid systems and reactor instrumentation and
' control systems designs for the pressurized water reactor at'Shippingport, PA.
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2rca. ;epcember 1966 to ibvember 1977 I hold professional and ma'nagement positions in the Nuclear Energy Systems division of the Westinghouse -
Electric Corporation. As a manager I* supervised.the preparation of system functional design requirements for nuclear reactor plant systems which affect
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plant control, protection, and transient. performance.
In addition to reactor control and protection systems, these systems included emergency feedwater systems, emergency boration systems, and steam dump systems. For four years I was the lead engineer responsible for establishing functional requirements for reactor control and protection systems used in the Westinghouse 3 loop
- nuclear reactor plants and for performing ac'cident apalyses of these plants for safety analysis reports submitted to the Atom'ic Energy Comission.
From November 1977 to October 1980 I was Systems and Civilian Applicatio,ns Program Manager in the Office of Inertial Fusion at the U. S. Department of Energy.
In this position, I provided technical and administrative direction for studies of 'the corrnercial applications.
of inertial confinement fusion.
I am a member of the American Nuclear Society and past member of thb IEEE Standards Comittee on Safety Related Systems.
I have authored or co-authored over ten technical articles for presentation at conferences or publication in journals.
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PROFESSIONAL QUALIFICATIONS OF WILLIAM TREVOR PRATT EXPERIENCE September 1976 - Present:
I am presently group leader (principal investigator) of the Accident Analysis Group, Division of Engineering and Risk Assessment, Brookhaven flational Laboratory. As Group Leader, my primary duties are to provide technical management over BNL staff and provide technical assistance to the Office of Nuclear Reactor Regulation (NRR), U.S. Nuclear Regulatory Commission (NRC). Other duties involve the safety review of specific reactor plants and the preparation and coordination of Technical Evaluation Reports, which contain a description of the work accomplished. At present I am primarily involved in an evaluation of the containment response of the Zion and Indian Point Reactor Plants to postulated core meltdown accidents.
I was involved in an evaluation of a proposed core ladle to be installed in floating nuclear power plants.
I have also worked on post accident heat renoval analysis for LMFBRs (namely FFTF and CRBR).
1975-1976:
Gibbs & Hill, Inc., New York, N.Y.
10001, U.S.A. Mechanical Engineer--Nuclear Menber of Development Group:
preparation of Standard B0PSAR based on 3800 MWt PWR Reactor Island.
Responsible for analysis of PWR water systems (Component Cooling Water, Containment Spray, Auxiliary Feedwater, spent fuel pool cooling and service water systems). Worked on analysis of suppression pool loadings due to BWR safety relief blowdown.
1974-1975:
Babcock & Wilcox, Ltd., Renfrew, Scotland - Mechanical Engineer: Thermal design of industrial boiler plant including associated equipment.
Development of a number of computer codes including an evaluation of flue gas properties at elevated pressure and temperature.
1973:
United Kingdom Atomic Energy Authority Reactor Group, Dounreay, Scotland -
Professional and Technical Officer, Grade II: 11 ember of technical section of the Dounreay Fast Reactor (DFR); technical appraisals in the areas of core thermal hydraulics and reactor containment.
Preparation of safety working party report (27) on an appraisal of the DFR containment leakage test program.
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t 1966-1967:
Whessoe, Ltd., Teesside, England - Design Engineer: Thermal and mechanical design, selection of materials of constrcction, preparation of technical reports and specifications for shell and tube heat exchangers.
1959-1966:
Head Wrightson Ltd., Teesside, England: Standard technical apprenticeship.
EDUCATION I obtained a Doctor of Philosophy Degree in mechanical engineering from the University of Strathclyde, Glasgow, Scotland. The graduate work was done during the period of September,1969 through December,1972 under the direction of Prof. Simpson, Head of the Departnent of Thermodynamics and Fluid Mechanics.
I also obtained a Bachelor of Science in Mechanical i
Engineer fran the University of Strathclyde.
The undergraduate studies were done during September,1967 through July,1969.
I attended Teesside Polytechnic, England fran September,1961 through April,1965 and obtained a higher national diploma in mechanical engineering.
H0 HORS AND Pl0LICATIONS I was awarded the Prescott Scholarship by the Institution of Mechanical Engineers, London in 1968. A list of publications I have authored or co-authored is attached.
. PlBLICATIONS (OPEN LITERATURE) 1.
W.T. Pratt, " Flash Evaporation in the Downcomer of a Natural Circulation Loop," Ph.D. Thesis, University of Strathclyde, Glasgow (1974).
2.
W.T. Pratt and R.A. Bari, " Impact of H Combustion on Degraded Core 9
Accidents in PWR Contaimments," presented at a Workshop on the Impact of Hydrogen on Water Reactor Safety, Albuquerque, January 1981.
3.
W.T. Pratt and R.A. Bari, "PWR Containment Response During a Postulated Core Meltdown Event," Trans. Am. Nucl. Soc. 38, 460 (1981).
4.
W.T. Pratt and R.D. Gasser, " Analysis of a Passive Ex-vessel Core Retention Device During a Postulated Core Melt Event," Proceedings of the ANS/ ENS Topical Meeting on Thermal Reactor Safety, Vol.1, pp. 218-225, April,1980.
5.
W.T. Pratt and R.D. Gasser, " Effects of Steel on a Core Meltdown in a Sacrificial Bed," Trans. Am. Nucl. Soc. 34,492(1980).
6.
R.D. Gasser and W.T. Pratt, " Thermal Response of a Molten Pool with Stefan Type Boundary Conditions," ASME Paper, #80-HT-9, July 1980.
7.
R.D. Gasser and W.T. Pratt, " Containment Response to Postulated Core Meltdown Accidents in the Fast Flux Text Facility," Nuclear Technology, Vol. 47, pp. 282-307, February 1980.
8.
J.K. Long, A.Q. Marchese and T.P. Spets (NRC), R.D. Gasser and W.T. Pratt (BNL), " Radiological and Containment Analyses for a Postulated Fast Reactor Melt-through Accident with Containment Venting," Proceedings of the International Meeitng on Fast Reactor Safety Technology, Vol. III, pp. 1251-1260, August 1979.
9.
R.D. Gasser and W.T. Pratt, " Analysis of LMFBR Containment Response to a Core-Disruptive Accident," Trans. Am. Nucl. Soc. 28,460(1978).
10.
J.J. Pyun, R.D. Gasser, W.T. Pratt and R.A. Bari, "Ex-vessel Containment Response to a Core Meltdown," Proceedings of the 3rd PAHR Information Exchange. ANL-78-10, pp. 327-334, November, 1977.
11.
K.R. Perkins, R.A. Bari and W.T. Pratt, "In-Vessel Natural Circulation During a Hypothetical Loss-of-Heat-Sink Accident in the Fast Flux Test Facility," ASME Paper, #79-WA/HT-66, December 1979.
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. 12.
R.A. Bari, W.T. Pratt and K.R. Perkins, " Phenomena and Scenarios Related to a Loss-of-Heat-Sink Accident (with Scram) in a LMEBR,"
Proceedings of the International Meeting on Fast Reactor Safety T2chnology, Vol. II, pp. 665-674, August 1979.
13.
R.A. Bari, H. Ludewig, W.T. Pratt and Y.H. Sun, " Accident Progression for a Loss-of-Heat-Sink with Scram in an LMFBR," Huclear Technology, Vol. 44, pp. 357-380, August 1979.
14.
R.A. Bari, H. Ludewig, W.T. Pratt and Y.H. Sun, "An Assessment of the Loss-of-Heat-Sink Accident with Scram in the LMFBR," Prescated at the International Meeting on Nuclear Power Reactor Safety, B 2sel s,
October 1978.
15.
R.A. Bari, H. Ludewig, W.T. Pratt, Y.H. Sun, "Recriticality Considerations for the Loss-of-Heat-Sink Accident with Scram," Trans.
Am. Nucl. Soc. 28_,471(1978).
16.
R.A. Bari, M.A. Klenin, W.T. Pratt and Y.H. Sun, " Meltdown Phase for an LMFBR Loss-of-Heat-Sink During Shutdown," Trans. Am. Nucl. Soc. 26, 347 (1977).
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PlBLICATIONS (LABORATORY REPORTS) 17.
W.T. Pratt and R.A. Bari, " Containment Response During Degraded Core Accidents Initiated by Transients and Small Break LOCA in the Zion / Indian Point Reactor Plants," NUREG/CR-2228, July 1981.
18.
W.T. Pratt and R.D. Gasser, " Thermal Analysis of a Floating Nuclear Power Plant Core Ladle," BNL-HUREG-27037, December 1979.
19.
S.S. Tsai, R.D. Gasser and W.T. Pratt, " Sodium Fires Evaluation for the Fast Flux Test Facility," BNL-NUREG-24635, August 1978.
20.
R.D. Gasser and W.T. Pratt, " Containment Response to Postulated Core Meltdown Accidents in the Fast Flux Test Facility," BNL-NUREG-24141_R, August 1978.
21.
S.S. Tsai, R.D. Gasser and W.T. Pratt, " Containment Design Basis Accident for LMFBRs:
Review of Methods," BNL-NUREG-23221, September 1977.
22.
R.D. Gasser, S.S. Tsai, D.C. Albright and W.T. Pratt, " Containment Design Basis Accident Analysis for the Clinch River Breeder Reactor,"
BNL-NUREG-25561, July 1977.
23.
K.R. Perk' ins, W.T. Pratt and R.A. Bari, " Evaluation of In-vessel Natural Circulatin during a Hypothetical Loss-of-Heat-Sink Accident in teh Fast Flux Test Facii 'ty,"
BNL-NUREG-26565, August 1979.
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R.A. Bar, H. Ludewig, W.T. Pratt and Y.H. Sun, " Accident Progression for a Loss-of-Heat-Sink with Scram in an LMFBR," NUREG/CR-0427, BNL-liUREG-50910, October 1978.
25.
R.A. Bari, H. Ludewig, W.T. Pratt and Y.H. Sun, " Material Relocation and Recriticality Assessment for the Loss-of-Heat-Sink Accident in the Li1FB R,"
BNL-NUREG-23432, November 1977.
26.
R.A. Bari, M.A. Klenin, W.T. Pratt and Y.H. Sun, " Preliminary Assessment of the Meltdown Progression of the Loss-of-Heat-Sink Accident with Scram in the Ll1FBR," BNL-liUREG-23137, August 1977.
27.
W.T. Pratt, "A Critical Appraisal rf the DFR Containment Leakage Test e
Programme," DFR/SWP/P212, May,1973.
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i Resume David G. Swanson Education:
Ph.D.
Nuclear Chemistry, Purdue University,1969 B.S.
Chemical Engineering, Nrrthwestern University, 1964 Experience:
1980-present:
Applied Science Associates, Inc.
Palos Verdes Peninsula, Co.
As a consultant to the U.S. fbclear Regulatory Commission (NRC), Dr. Swanson has worked on reactor safety studies for the Zion and Indian Point reactors. This work focused on the review and development of concepts for containment of molten core debris following a core meltdown event.
1973-1980:
The Aerospace Corporation, El Segundo, California Dr. Swanson served as an adviser and consultant to the U.S. Nuclear Regulatory Commission (NRC), and has worked on reactor safety studies for the NRC. These studies have been concerned with the evaluation of materials interactions at high temperatures.
Particular emphasis has been placed on the phenomena associated with a core meltdown event. The interactions of concrete and refractory materials with molten core debris and liquid sodium were extensively studied. These investigations included a major experimental program and the review of work performed elsewhere in this area. DOE advanced reactor programs and NRC research programs in the areas of sodium-cooled fast reactors, the floating nuclear power plant and the high-temperature gas-cooled reactor have been reviewed and evaluated for NRC. Other activities for NRC have included materials evaluation and characterization.
Defense related activities have included experiments to study the response of materials to high temperature and radiation environments.
Computer programs have been developed to calculate the response of materials to radiation by means of radiation transport and hydrodynamics. Other activities have included studies of the aging of materials in earth satellites, the generation of contaminants from materials in satellites and issues concerned with the location of satellites in orbit.
In addition, defense related studies included photon-deposition calculations, electron-transport and heat-transfer calculations.
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1968-1973:
Sandia Laboratories, Albuquerque, New flexico Radiation " transport and hydrodynamic calculations were perfomed for nuclear weapons environments and nuclear weapons effects experiments were conducted. The response of materials to nuclear weapons environments was calculated and studied experimentally. The experimental studies were concerned with the response and survival of materials when subjected suddently to a very high temperature environment. Other activities included heat transfer calculations themodynamic calculations, and chemical kinetics. All of these activities were related to the design of nuclear weapons.
Past NRC-Related Experience For the past seven years, the principal investigator has assisted the Office of Nuclear Reactor Regulation in the review and evaluation of the materials interactions that can occur as a result of postulated core meltdown accidents. During this period, considerable expertise in reactor naterials evaluation and analysis has been developed. Assistance has been provided on reviews for the Floating Nuclear Plant (FNP), the Zion and Indian Point plants, the Fast Flux Test Facility (FFTF), the High Temperature Gas-Cooled Reactor (HTGR) and the Clinch River Breeder Reactor Plant (CRBRP).
Technical assistance has been provided in the review and evaluation of information submitted by the FNP applicant (Offshore Power Systems) in the area of materials interactions associated with the delay of core melt penetration.
This has included detailed technical assessments of molten core interactions with concrete and with sacrificial materials for use in the preparation of NRC safety and environmental review documents.
In these studies, the effects of mechanical and thermal shock, brick floatation, eutectic formation, chemical interactions and slag attack have been investigated for the material (Mg0) proposed for use in the core ladle.
In the safety review of the FFTF program, technical assistance was provided in the review and evaluation of materials issues. The interaction of liquid sodium with basalt and magnetite concrete and the commercial insulating and fire-resistant firebricks employed in the cell liners was studied.
Technical assistance in this area also included an examination of the problems associated with a core disruptive accident, including the interactions between molten core debris and concrete and firebrick. The problem included the rate of penetration by sodium.and molten core debris into concrete and firebrick. Other issues concerned cracking and spallation, the potential for adverse chemical reactions and the extent of gas evolution.
In addition, technical assistance has been provided for safety issues concerned with the potential for a postulated core melt accident in the proposed CRBRP.
For this effort, the problems associated with molten core-limestone concrete and sodium-limestone concrete interactions were
- assessed and evaluated.
Potential sacrificial naterials for proposed core retention concepts were examined. 1101 ten core debris and liquid sodium interactions with Itg0 were studied extensively.
In the area of High Temperature Gas-Cooled Reactors (HTGR), the loss of graphite strength due to oxidation induced by water vapor and the presence of contaminants has been exanined.
Review and assessment of certain programs at General Atonic has been provided; this has included examination of various core catcher concepts for Gas-Cooled Fast Reactors.
Other core retention proposals have been evaluated. These have included the use of borax and high-alumina cement.
A close interaction was maintained and materials analysis input has been provided to the UCLA post-accident heat removal (PAHR) effort and the Brookhaven National Laboratory (BNL) nodeling activities concerned with post-accident containment. This work has included a determination of the critical parameters influencing the extent of core debris penetration into either concrete cr high-temperature sacrificial materials and factoring soil parameter; 4to the above modeling efforts.
Review and comments have been provided concerning the RES Confirmatory Research Program at Sandia Laboratories. This evaluation has been concerned with the materials interactions associated with the retention of sodium and molten core debris.
DOE Base Technology programs in the area of materials interactions concerned with the retention of sodium and debris have been reviewed and evaluated. A close interaction with DOE contractors has been maintained in order to assist NRR in incorporating these results into its safety environmental evaluations of advanced reactor systems.
The principal investigator has conducted experiments to study the interactions of both concrete and fig 0 with molten core debris and liquid sodium.
Over forty experiments with molten UO,, and molten steel were performed, including experiments in which the 00 or steel was maintained in a molten state on top of either concrete or lig0.2 The interactions between the materials were thoroughly analyzed and have been the subject of a number of published reports.
During the course of seven years of work for the Office of Nuclear Reactor Regulation in this area, the principal investigator has acquired expert knowledge in the area of nuclear reactor accident scenarios which may lead to core melting. He has also acquired a general knowledge of NRC licensing procedures, regulations, equirements and safety criteria.
The principal investigator. holds a Ph.D. in nuclear chemistry and is a licensed professional nuclear engineer in California.
HAROLD E. POLK PROFESSIONAL QUALIFICATIONS STRUCTURAL ENGINEERING BRANCH
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DIVISION OF ENGINEERING OFFICE OF NUCLEAR REACTOR REGULATION I am a Senior Structural Engineer in the Structural Engineering Branch, Division of Engineering, Office of Nuclear Reactor Regulation, U. S.
Nuclear Regulatory Commission, Washington, D. C..
I am responsible for reviewing safety analysis reports with r'egard to structures and seismic analysis for nuclear power plants assigned to me.
I joined the Division of Engineering in November 1974.
I have served as Structural Reviewer for the safety reviews of many plants including Hartsville Nuclear Power Station, Black Fox Station, Arkansas Nuclear One Unit 2 Yellow Creek Nuclear Plant and New England Power 1 and 2 project.
I hold a Bachelor of Civil Engineering (1958) from North Carolina State College, and did graduate study at North Carolina State College in Structural Engineering.
My 23 years experience includes 8 years of aircraft stress analysis and flight performance with the Boeing Company on the Minuteman Missile Program and the Apollo Project which landed
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the first men on the moon.
I joined the NRC staff after completing over 4 years of seismic dynamic analysis of nuclear power plants with the Bechtel Power Corporation of which the last 2 years was a Supervisor of 'a Seismic Analysis Group.
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4 2-I am currently a member of the American Nuclear Society, ANS2.2/2.10 Working Grcup on Seismic Instrumentation and a member of American Institute of Steel Construction, AISC Nuclear Specification Task Committee III.
I am a registered Professional Engineer in the states of Maryland (PE 8075) and Florida (18643).
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