ML20196L537
| ML20196L537 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/29/1988 |
| From: | Masnik M Office of Nuclear Reactor Regulation |
| To: | Morris A LANCASTER, PA |
| References | |
| NUDOCS 8807070574 | |
| Download: ML20196L537 (27) | |
Text
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JUN 2 91988
. The Honorable Arthur E. Morris
. Mayor of Lancaster P.O. Box 1559 120-N. Duke Street Lancaster,'PA~ 17605
Dear Panel Member:
Enclosed are. responses to questions raised at the May 26, 1988 TMI-2 Advisory Panel Meeting. Responses have been prepared to written questions submitted by TMIA.and SVA. Questions raised during the meeting are either answered in the enclosure or referenced by page number back to the transcript for the May 26, 1988 meeting.
Our next meeting is' July 14, 1988 at the Holiday Inn, Harrisburg, PA.
The Advisory Panel will meet with the NRC Comissioner's at our new building in Rockville, Maryland on the 16th of August 1988 beginning at 10:30 am. I will provide you with' nore infortnation and a map' at tha July 14th meeting.
Sincerely, original signed by Michael T. Masnik Panel Liaison
Enclosure:
As stated
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l Mr'..Thom'as Gerusky, Director $ Mr. Kenneth L. Miller, Director Bureau of' Radiation Protection .
Division of Health Physics and Dept. of Environmental Resdurces Associate Professor of Radiology l P.O., Box '2063 Milton S. Hershey Medical Center Harrisburg, PA 17120 Pennsylvania State University Hershey, PA 17033 Elizabeth Marshall Thomas D. Smithgall 736 Florida Avenue 2122 Marietta Avenue York, PA 17404 ,
Lancaster, PA 17603 4
5 Niel Wald, M.D.
Professor and Chairman Department of Radiation Health .]
University of Pittsburg A512 Crabtree Hall Pittsburg, PA 15561 jj T
/ Ann D. Trunk Mr. Joel Roth 143 Race Street RD 1, Box 411 Middletown,,PA 17057 Halifax, PA 17032 .
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Dr. John Luetzelschwab ,3 Professor Physics W The Honorable Arthur E. Morris i Dickinson College 4 Mayor of Lancaster !
Carlisle, PA 17013-2896 P.O. Box 1559 :
f, 120 N. Duke Street i j Lancaster, PA 17605 ;
Dr. Gordon Robinson i Associate Professor of j
, Nuclear Engineering i
! 331 Sackett Building University Park, PA 16802
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'N E Frederick S. Rice Personnel Financial Management Inc.
2 Crums Lane Harrisburg, PA 17112 1(
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l OVESTIONS RAISED AT THE MAY 26, 1988 ADVISORY PANEL MEETING, HARRSIBURG, PA Susquehann.a Valley Alliance
- 1. Due to the uncertainties of radionuclide dispersion and deposition followin0 the accident, upon what information is Table 2.4 based?
Response
Because refined modeling methods are not available for accurately analyzing the transport and deposition of the fragmentation debris, or the leaching of soluble materials from the damaged core, a set of assumptions was made regarding the dispersion and deposition of radionuclides in the Ttil-2 facility. These assumptions were based on information available from fuel measurements and contamination measurements throughout the reactor building, as well as on the chemical and physical state of the radionuclides. All assumptions were chosen to ensure that the amount of activity estimated to be in any location either meets or exceeds the amount actually measured in that location. The assumptions cre outlined ir.
Section 2.2 of NUREG-0683, Draft Supplement Ne 3 on pages 2.21 through 2.31.
I 2. By the time of PDMS, will we know the condition of the containment and the damage to it caused by the accident? How will this information be made available to the Public?
Response
l There has been no evidence of any damage to the containment building that would result in any compromise in its ability to contain radiation during PDMS. Worker access is available above i the 305 ft elevation, and no signs of containment degradation hcve been observed. Video access of the 282 ft elevation (the reactor l building basement) has not disclosed any damage to the containment l building.
- 3. While Unit 2 is in PDMS, what research will continue which relates to l
the reactor?
Response
The NRC has no plans for additional research directly related to TMI-2 during the proposed post-dafueling monitored storage period.
- 4. Explain the rationale for delaying clean-up. Delay will have no effect on the long-lived radionuclides. Is the delay then for reasons of technological advances?
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- Response:
- The NRC, in its role as a regulatory agency is evaluating GPU Nuclear's proposal to place the THI-2 facility in post-defueling monitored storage. AccoiJing to the licensee's Technical Plan, (TMI-2 Cleanup Program Post-Defueling Monitored Storage fTP0/TMI-188 Rev. 0), January 1987):
"A monitored storage period following completion of the current cleanup program is beneficial for several reasons.
Occupational dose in the plant will be reduced during monitored storage due to the natural decay of radioactive contamination. Over an extended could be reduced by as much dominant as a factor of 2 isotopes,[m(Sr-90, ore likely only two-thirds].Cs-137)
The period, le occupational dose in radiation zones would be reduced proportionately.
The monitored storage period allows time for continued development of decontamination technology so that the most effective and efficient techniques may be applied. Further reduction in occupational exposures would be achieved through use of advanced robotic technology, automatic cleaning and chemical cleaning techniques, and advanced waste treatment methods.
This monitored storage period also allows for resolution of the current limitation on national waste disposal capabilities so that selection of processes may be less dependent on waste volume production. The result may be further reductions in occupational dose required to accomplish specific tasks."
- 5. How will the number of entries be detennined during PDMS?
Response
The licensee indicated in "THI-2 Cleanup Pr*. gram Post-defueling Monitored Storage - Technical Plan" (TP0/THI-188, January 1987) that entries to the reactor building and auxiliary and fuel handling building would be conducted for purposes of visual inspection,radiationsurvey,andrecording(ofplantconditions (TP0/THI-188). Table 2.3 of that document Table 3.2 of NUREG0683,DraftSupplement3)liststheanticipatedschedulefor initial PDMS monitoring / inspections. The number of entries will be greatest early in PDMS. Although the licensee's plan for the
.nitial frequency of entries is monthly (12 times per year), the licensee indicates that they anticipate "that the initial frequency I will decrease (e.g., quarterly) based on an evaluation of data accumulated auring the initial period." (TP0/TMI-188, p.20).
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- 6. Upon what findings and/or studies does the NRC base its assumption that
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the activity in the top 1/2" of the wall becomes available for ,
resuspension? What allowances are made for the fact that the walls '
might crumble due to stress from age and clean-up activities already undertaken?
Response
Page3.12(section3.2.2.1)ofDraftSupplement3containsthe assumption that the activity in the first 1/2 inch (1.3 centimeters) of the concrete block becomes available for resuspension after the structure has dried for a period of time.
This assumption is based on a study by Arora and Dayal (1986) as referenced in NUREG-0683, Draft Supplement 3. This study indicated that for cesium in a concrete solid, the cesium leach rates were greater when the wet periods were interspersed with dry periods, than when the concrete solid was continuously saturated. The observed enhancement in cesium release with increasing length of dry period is believed to be a result of replenishment of the surface with cesium, migrating from the sub-surface zones during dry periods. In Draft Supplement 3 this phenomenon was bounded, by assuming that up to one-eighth of the radioactive material in the concrete blockwall would migrate to the surface and be available for suspension into the atmosphere. This number is at least several times greater than the amount of radioactive material that is expected to be available for resuspension from the concrete block wall.
The reactor building is a reinforced concrete structure corrposed of a cylindrical wall with a flat foundation mat and a dome roof. The wall thickness of the cylindrical wall is 4 feet and the thickness of the dome is 3 feet, 6 inches. The foundation mat on bearing rock is 11 feet, 6 inches thick with a 2-foot thick concrete slab above the base liner plate. The inside surface of the reactor building is lined with a carbon steel liner with a nominal thickness of 3/8 inches for the cylinder,1/2 inch for the dome and 1/4 inch for the base.
The TMI-2 facility was designed and constructed for a 40-year lifetime from the start of construction (beginning late in 1969) and the staff, in Draft Supplement 3, assumed post-defueling monitored storage period would be com i
i years from the start of construction)plete . in 2009activities The cleanup (a period thatof 40 have occurred or are being proposed for the period before PDMS are relatively nondestructive in nature. The environment to which the walls of the containment have been exposed to since the accident would not cause any significant degradation of the concrete. The NRC staff did not consider the crumbling of walls due to stress from age or cleanup activities credible.
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- 7. What is 10% of each activation product? Upon what information or studies do you make the assumption that 10% of the activation will remain in the reactor building at the end of defueling? (products EIS page 2.27,2.2.1).
Response
Table 2.4 lists the quantity of activation products that were assumed to be present in the facility after defuelirg. Activation products listed in Table 2.4 include manganese-54, iron-55, cobalt-60 and nickel-63. Carbon-14 is also fomed by activation, although a small quantity is formed by fission. Carbon-14 was treated as a fission product because it has been detected in the accident-generated water and is therefore solubic in water and assumed to be distributed in the samp manner that was assumed fcr some of the fission products. Tritium is a'so produced by activation although over 90% of the tritium produced is from ternary fission.
The amount of activation products present at the time of the accident was determined using the ORIGEN-2 computer codes. The amount that would have sen available at the beginning of PDMS in the absence of any cleanup or defueling operations was cciculated by decay correcting the amount present after the accident. The assumption was made that only a small portion of the activation products were removed by sampling or defueling. Of the total quantity of activation products estimated to be present at the end of defueling, 10 percent was assumed to be in a form that would allow for dispersal and could contribute to an offsite dose. The assumption was made that the remaining 90 percent of the activation products were either shipped offsite or were part of the stainless steel of the primary system and therefore unavailable for dispersal.
- 8. The water which will leak into the system has been determined to be 5000 gallons per year. Explain why this amount is so much less than the inleakage for this past 9 years.
Response
According to the environmental evaluation (Letter from F. R.
Standerfer to the NRC, March 11, 1987.
Subject:
Environmental Evaluation for THI-2 Post-Defuelin 4410-87-L-0025, Document ID 0161P)g Monitored inleakage Storage.
of groundwater and precipitation are anticipated to be the major sources of liquids during PDMS. The licensee estimated, based on experience to date and on the anticipated lower frequency of maintenance during PDMS, an annual inleakage of 5000 gallons. Water inleakage currently occurs in the following areas of the plant and is collected as indicated:
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- 1) Fire service penetration; east wall of turbine building at the 300-ft elevation. Drainage is to the turbine l
building sump, water treatment sump, or the condensate regeneration polisher sump.
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- 2) Building joint; between the service ouilding and air intake tunnel. This area does not have sump drainage.
It is pumped periodically, as necessary, to remove inleakage.
- 3) Construction joint; basement of the auxiliary building.
Drainage is to the auxiliary building sump.
- 4) Electrical penetration; southwest corner of the control building area at the 281' elevation. Drainage is to the control building area sump.
Reference (Letter, Standerfer to the NRC, June 23, 1987.
Subject Post-defueling monitored storage environmental evaluation. 4410-87-L-0093, Document ID 0194P. - Referenced in Draft Supplement 3 and available in Public Reading Room).
No inleakage is expect'd e into the reactor building.
The expected annual inleakage of 5000 gallons is much less than the amount (approximately 264,000 gallons) of water that flowed into the reactor building basement during the two years following the accident. The sources of this water included the primary coolant, water from the reactor building spray system and from river water inleakage from the building air coolers. However, the reactor coolant system will be drained before PDMS begins, the building air cooler system has already been drained and the closed loop (does not ne water from the river) cooling system currently in use will be deactivated and drained before PDMS begins.
- 9. Page 3.31, Section 3.3.1.1. Explain those measurements which are being presently undertaken? What is being measured? In what manner will the results affect decisions about RCS decortamination and the future of the fccility?
Response
l The statement in question refers to the measurement of radioactive material located in the reactor coolant system. The amount of radioactive material including the amount of fuel debris is being l
measured in all accessible locations of the reactor coolant system.
The methods that will be used during the decontamination of the reactor coolant system will depend in part on the amount of l radioactive material present in the reactor coolant system and its precise location in the system. For instance, those areas with little or no contamination will require very minor amounts of decontamination, while decontamination efforts in areas that contain large amounts of radioactive material will be more extensive.
The draft supplement on page 3.31 indicated that the selection of methods and processes for additional reactor coolant system decontamination is expected to de)end on the future disposition of the facility and on measurements )eing made ct the present time.
We did not intend to indicate that the results of the measurements would affect decisions on the future of the facility.
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- 10. What would preclude the use of the AGW to clean the RCS7
_ Response:
No action other than disposal of the accident-generated water would preclude its use during the decontamination of the reacter coolant system. For the evaluation in Draft Supplement 3 to the PEIS, it was assumed that the accident-generated water would be processed and removed from the reactor building and the auxiliary and fuel-handling buildirg prior to the initiation of immediate cleanup or post-defueling monitored storage.
- 11. Will the water used for further clean-up coatain chemicals? How will these be removed from the water before the water is released to our drinking water supply?
Response
Water that is used during decontamination and cleanup processes is routinely run through fon-exchange systems (the submerged demineralizer system and the EPICOR II system are ion-exchange i systems) if necessary to filter any radioactive material and !
chemicals that may be present. Any water released to the Susquehanna River or to any other drinking water supply would have to meet the licensee's technical specifications as well as the conditions of the National Pollutant Discharge Elimination System (NPDES) pemit issued by the Commonwealth of Pennsylvania, Department of Environmental Resources (PaDER).
- 12. Page 3.32, Section 3.3.1.2. When do you expect the radiation doses to be low enough to pemit entry into the basement for complete clean-up?
If they are presently too high to pemit entry, does this not rule out the possibility of immediate clean-up as an alternative to be considered?
Response
Entry into the basement would most likely not be considered in areas where the dose rate remained much above 1 R/hr, although, even at these radiation levels, a worker could be allowed to work for a short period of time. High dose rates, however, do not preclude the possibility of cleaning the basement, or the possibility of the immediate cleanup alternative. Dose reduction efforts are currently occurring in the reactor building basement.
These dose reduction efforts, including scabbling of the walls with robots and construction of a manifold for waterflow to leach activity from the concrete block wall, are describad in Draft Supplement 3, in section 2.1.1, pages 2.9 to 2.11.
- 13. hv can the impact of the waste after disposal at either a regional or ot! 'r site be considered outside the scope of this EIS? Delaying clean-up has a major impact on the final resting place for the waste from THI, since the State of Pennsylvania is presently in the process of developing a site.
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Response
The environmental impact of waste disposal at a commercial low-level waste disposal site is the subject of ar environmental evaluation specific to the chosen site, which must be com)1eted before the site can be licensed. Waste streams outside tiose evaluated during the sites environmental evaluation will not be allowed for burial. The environmental evaluation for a regional burial site must be specific to the environmental characteristics of the site, and must also address all types of wastes that will be accepted into it, including wastes from hos)itals and university research labs. Wastes from THI-2 will not )e accepted at a regional site until the site is licensed.
- 14. Page 3.19, footnote a. What are the precautions to be taken to ensure that criticality would not occur?
Response
A variety of precautions are available for use during the cleanup program to ensure that criticality will not occur. These include ensuring that the small quantity of fuel debris remaining after the current defueling efforts is not available in large enough quantities to create any possibility of a criticality. The licensee will provide a criticality analysis that will address each separate quantity of residual fuel in each defined location. The
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criticality analysis will estimate the quantity of fuel remaining, l its location, its dispersion within the location, its physical form l (i.e. film, finely fragmented, intact fuel pellets), its mobility, the presence of any moderating or reflecting material, and its l
potential for a critical event. In this submittal the licensee must demonstrate that the cleanup has progressed far enough such that an inadvertent criticality is precluded.
- 15. Between entries, how will the Licensee know that criticality has not occurred?
Response
Prior to entering PDMS, steps will be taken to ensure that a criticality event is not a credible event (see response to question 14 above). Most of the fuel debris remaining in the THI-2 facility following the current defueling effort would be sealed in piping or enclosed in components. Measurements will be made by the licensee and verified by the NRC and their contractors to ensure that the amount of fuel debris in a given area will not be large enough to cause a criticality. During PDMS the licensee does not i plan to maintain monitoring activities that are specific to identifying an inadvertent criticality event in containment.
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16.
During entries, how will the workers know that criticality is not occurring?
Response
See response to Question No. 15 above. Furthennore, workers will carry radiation survey instruments.
- 17. By what means will the Licensee determine the amount of radioactivity in the reactor prior to purging this radioactivity to the environment?
Response
The radiation monitors located in the purge exhaust and vent stack would be used to ensure that the amount of radioactivity in the effluent is within the. acceptable limits given in the technical specification limits. If the amount of radioactivity in the effluent is above the technical specification limits then the purge exhaust would automatically be returned to the reactor building.
- 18. In the event of an incident at Unit 2, how many workers would be a/ailable at any one time to deal with the emergency-at a time when the workers have been reducei in the first year and then in the second year and thereafter.
Is it possible or likely that workers from Unit I would be drawn te Unit 2 to help deal with an emergency?
Response
According to the licensee (Letter from F.R. Standerfer, Director, THI-2, GPU Nuclear Corporation to W.D. Travers, Director, THI-2 Cleanup Project Directorate, NRC, November 5,1987.
Subject:
Post-Defueling Monitored Storage Environmental Evaluation, NRC Coninent Response) the level of direct employment for the PDMS program would be about 100 to 125 personnel during the transition year following the completion of current defueling activities and about 70-75 personnel thereafter until final cleanup. These workers would be available to deal with an emergency, although the number on site at any one time may vary and is unknown to the NRC.
Currently, fire, security, and medical emergency personnel are
, shared with Unit 1.
- 19. Does GPU Nuclear need an amendment to its license before PDMS is enacted.
Response
Yes. Prior to entering PDMS, an amendment to the THI-2 operating license would be required to align the technical specifications tr plant conditions expected during long-tenn storage.
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Since Unit 2 is in the 100 year flood plain, how will this affect its 20.
License prior to seeking approval for PDMS
_ Response:
As page 4.10 (Section 4.1.3) of Draft Supplement 3 indicates, the island on which both the TMI-1 and TMI-2 reactors are located is not within the 100-year flood plain; however, it is within tha given year) as detemined by the U.S. Army Corps of Engineers.500-year flood Supplement 2,pageA.8andA.9). This will not affect the licensee's ability to seek approval for PDMS for two reasons.
First, the island is diked for flood protection, and the dikes are inspected and maintained by the licensee. Second, THI-2 flood procedures require that flood door panels be installed when the river elevation reaches 302 feet (92 meters). Installation of flood door panels effectively precludes the entry of river water.
- 21. Explain why the estimated occupational doses are so much higher for imediate clean-up.
Response
The occupational ese range that was estimated for the alternative of imediate cleanup (300 to 3100 person-rem) is higher than the occupational dose range that was estimated for delayed cleanup (48 to 1500 person-rem) for the following reasons:
- 1) The 20 year period of post-defueling monitored storage would result in the decay of the principal radionuclides to levels approximately two-thirds the level that would be present l during imediate cleanup.
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- 2) It was assumed that robotics, decontamination and waste i treatment technologies would allow further reduction in occupational dose levels during cleanup following PDMS.
- 22. Explain the subtle difference between the no-action alternative and the Licensee's proposal. What guarantees or laws will preclude the Licensee's PDMS proposal from becoming the no-action alternative?
Response
l Section 3.1.5 of Draft Supplement 3 describes the no-action alternative. Section 3.2 of Draft Supplement 3 describes the licensee's proposal. Section 3.1.5 states that the no-action l alternative would be essentially the same as that described by the l
licensee's PDMS proposal except that heither preparations for PDMS l nor subsequent actions to finish the cleanup would occur.
The NRC will not allow the licensee to place the facility into monitored storage until the necessary requirements for long term storage are met. The NRC regulations require that the license holders at nuclear power facilities take certain steps to assure that 9
the facility will ultimately be decommissioned and equipment, structures,and portions of the facility and site containing
. radioactive contaminants are removed or decontaminated to levels acceptable for unrestricted use of the property.
- 23. Into what areas and how much money will the Licensee or the NRC put into research to develop technology for clean-up following PDMS7 Will the ,
NRC obtain a commitment from the Licensee to finance such development? ;
Response
The NRC has no alcos to develop technology for cleanup following PDMS. This tas( would be left to the licensee. No connitment will be obtained by the NRC from the licensee to finance further development of technology.
- 24. Will all of the waste generated since the on-set of clean-up and up to the placement of the plant in PDMS be removed from the island before the Unit is placed in PDMS7
Response
For the evaluation in Draft Supplement 3, it was assumed that all of the waste associated with decontamination activities since the time of the accident would be removed from the island before Unit-2 is placed in PDMS. As discussed in Draft Supplement 3, some fuel debris would remain in the reactor vessel (page v), and some outside of the reactor vessel (page 2.18, section 2.1.3) and radioactive material will remain in many areas of the reactor building (Section 2.1.2) and some areas of the auxiliary and fuel-handling buildings (Section 2.1.3). The licensee in the document, "THI-2 Cleanup Program Post-defueling Monitored Storage - Technical Plan" (TP0/TMI-186, January 1987) indicated that before the start of PDMS, "Radioactive material will have been removed or contained..."
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THI Alert
- 1. 2.1 The staff noted that, "The primary difference between an undamaged
. reactor at the end of its useful life and the licensee's PDMS proposal is that during PDMS relatively high levels of contamination would remain in the reactor building basement and a small amount of residual fuel would remain in the reactor coolant system [during] storage."
What factual data are these conclusions derived from? How many "undamaged reactors at the end" of their "useful" lives have the NRC dealt with? Were technical experts from these plants consulted? If so, is their input a matter of public record? What other differences exist between these plants sod GPU's PDMS plan? Was embrittlement a factor at these plants? What was the staffing levels at these plants?
Response
The original PEIS (page 2.3; Section 2.1) indicated that "For full cleanup, all cleanup operations would be carried through to the point that the facilities were ready to initiate decomissioning or refurbishment operations." This is the condition of a undamaged reactor at the end of its useful life, at which time after the fuel has been removed, it is ready for decommissioning or refurbishment operations. The statement cited in the above question was meant as
- a comparative statement rather than a quantitative statement. The l comparison between an undamaged reactor at the end of its useful l life, and the licensee's PDMS proposal for the THI-2 reactor
- undamaged reactors was made to indicate that unlike THI-2, undamaged reactors have not had large quantities of radioactive water dumped into their basements and have not had fuel debris dispersed through their reactor coolant system. No comparison of potential for embrittlement or of staffing levels was implied.
The NRC has had considerable experience with reactors that have not had a significant accident before the end of their useful lives.
Examples include Humboldt Bay, Dresden 1, Indian Point 1, Lacrosse, Shippingport, Elk River, and Carolina-Virginia Tube reactor. These reactors differed from each other and from Three Mile Island, Unit 2, in design, operating history, and power levels.
l 2. 2.1.1 The staff argued that, "The reactor containment building is l uniquely designed and constructed to maintain its structural integrity (with almost no leakage) during a wide variety of accidents."
l l How long after an accident was the RCB designed to maintain its integrity? Was it specifically designed to house radioactive waste materials for an indefinite period of time? If not, would not storage of such wastes necessitate a license amen 6 ment?
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Response
The reactor containment building was designed to maintain its integrity during a peak accident pressure of 60 psig allowing only
. 0.2 percent leakage during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, and 0.1 percent leakage per day thereafter. However, the accident that occurred at THI-2 was not an accident of this proportion, having reached a peak accident pressure of 28 psig. The reactor containment building was designed to maintain its integrity for a period of 40 years, whether or not a design base accident occurs. Construction did not begin until late 1969.
The reactor building was not designed specifically to house radioactive waste for an indefinite period of time. The current NRC regulations do notiallow for an indefinite storage of the facility.
- 3. 2.4 How permanent are "pemanent dose reduction techniques?"
Response
The term "pemanent dose reduction techniques" refers to methods that pemanently remove the source of radiation from the area where it is located. This tem was used to distinguish th2se methods from the dose reduction technique of shielding the source of radiation by placing structures on or around it to attenuate the dose rate.
- 4. 2.1.1 "Sectioning and disposal of the reactor internals at i reactor vessel are not considered part of the cleanup because radiation levels expected from these components would be no higher than in a nomal
! reactor nearing the end of its life."
, What are "sectioning and positioning of the reactor internals" part of?
What if radiation levels are incorrect? What exactly are the radiation t
levels of a "nomal reactor at the end of its life?" What constitutes a normal reactor?
Response
i Sectioning and disposal of the reactor internals and reactor vessel
- are considered part of the next phase in the life of a reactor, the l decomissioning or recomissioning process, because this activity I would also occur during decomissioning or recomissioning of a reactor facility that has not undergone a significant accident. In other words, this is not an action that is necessitated in order to clean up the facility as a result of the accident.
Even if radiation levels in the reactor internals and reactor vessel are found to be higher than expected during sectioning and disposal operations, little or no impact would be expected, because additional shielding or distance could be used to reduce occupational dose. However, because of the short length of time l the THI-2 reactor operated (less than 14 months), the quantity of the activation products in the reactor internals and in the reactor 12 l
vessel are less than the quantity that is present in a reactor that has operated for longer than 14 months and much less than the quantity that would be ) resent in a reactor that had operated for
. 40 years. It is not licely that this assumption is incorrect.
Measurements taken on the lower grid rib section and plenum confirm that radiation levels are no greater than-expected on reactor internal components.
The radiation levels emitted from the reactor internals and reactor vessel will vary among nuclear reactor facilities, depending on the material used to construct the vessel and internals, the operating history, and the operating power. The statement in question was meant as a qualitative statement made for the purpose of comparison to explain why certain: activities were considered to be part of the decommissioning or recommissioning process rather than part of cleanup. We do not expect to compare the absolute radiation levels in the reactor internals and reactor vessel of the TNI-2 reactor with the levels in normal reactors at the enc' of their useful lives.
The term "normal reactor" as used in this supplement refers to a reactor that has not undergone a significant accident. This term will be included in the nomenclature section of the final report.
- 5. 2.1.4 What unique problems will the AFHB pose since it 'was not designed to be leak free..." during a "... variety of accidents?" How much, and just exactly what, leaks from the AFHB7 What are the dose levels found in AFHB at the end of its life?"
Response
Buildings in general are not dasigned to be leak free, especially under accident conditions. Because the dose levels expected to be present in the AFHB at the end of the current defueling efforts are expected to be similar to those found in the AFHBs of operating reactors, no unique problems would be posed by the THI-2 AFHB.
The general area dose levels in the AFHB are below 2.5 mR/hr in many areas although they do approach 15 mR/hr in some of the cubicles that contain equipment. This is similar to dose levels that would be found in the AFHBs of operating plants after i operation for 40 years.
l l 6. 2.2.1 Why weren't new calculations taken concerning the number and quantity of remaining radionuclides? Does the NRC or GPU have a comprehensive inventory of the radionuclides released since the accident? Is it possible for radiation levels to shift or relocate from one section of the plant to another? If so, isn't [it] possible that sections designated to have certain radiation levels may now be inconsistent with GPU's endpoint criteria?
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Response
New calculations were made to detemine the number and quantity of radionuclides expected to be in the facility following the current
. defueling efforts. The results are shown in Table 2.4 and the assumptions that were made in support of the calculations are given in sections 2.2.1 through 2.2.3. These calculations were based on the amount of radioactive material present at the time of the accident as given by the ORIGEN-2 computer code.
GPU measures the radioactiva material releases and reports them to the NRC per the requirements of Section 5.6.1.C of Appendix B to the Recovery Technical Specifications. The releases are reported quarterly for gaseous effluent releases, liquid releases, and solid waste and irradiated fuel shipments. However, due to the nature of the accident and the method by which the material has been removed from the reactor and shipped offsite, we can not provide a comprehensive inventory of every radionuclide since the time of the accident.
The following methods will shift or relocate radiation levels from one section of the plant to another;
- 1) Movement of radioactive material by personnel, either advertently or inadvertently,
- 2) Movement of radioactive material by animals or insects,
- 3) Movement of radioactive materials by water or air transport.
The licensee makes at least monthly measurements of the amount of radioactivity present in the THI-2' facility. These measurements are used to detentine the decontamination progress that has taken place to date, and can be used to identify any relocation of radiation levels from one section of the plant to another. These measurements are also used to ascertain whether the endpoint criteria have been met.
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Ouestions from the Transcript Kenneth Miller Director, Health Physics Division Hershey Medical Center
- p. 38 I would like to know, can you do a comparison for us between the contamination levels that will exist at the end of the defueling period with the levels you keep referring to that exist at the end of a useful life of an operating reactor? You're talking about a factor of 2, 10, 10007 Your comment -- you kept referring to the fact that you would like to see the plant cleaned up to the point where it matched the levels of contamination present at the end of the useful life of an operating reactor.
Can you give us some sort of a comparison? Maybe they are already lower and you don't have to do anything -- but I doubt that.
Response
Answered in transcript, pages 39 and 40.
- p. 40 So the required additional cleanup will be strictly concentrated on those areas that are still unreasonably high?
Response
Answered in transcript, pages 40 and 41.
- p. 50 Are the funds currently available to do an imediate cleanup?
Response
Answered in transcript, page 50.
Thomas Gerusky Director, Bureau of Radiation Safety Pennsylvania Department of Environmental Resources
- p. 41 You made a comparison between a four year cleanup and a twenty-four year delayed cleanup with a total exposure comparison for four years and twenty-four years.
Would there be any environmental impact or any exposure to the public following the imediate four year cleanup and after the twenty-four cleanup that has not been taken into consideration in comparing the two?
Response
Answered in transcript, pages 41 and 42.
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- p. 42 But you don't think that the public and we cught to have a feel for what those doses are, wht.t are those environmental impacts are, for the same periods of time until decommissioning, assume you go to decomissioning?
Response
Answered in transcript, page 42.
- p. 42 You're comparing twenty-four years versus four years, and shouldn't you compare 24 to 247 That's what I'm asking. If you're going to an end )oint, shouldn't the end point be the same for the exposures for boti options? ,
Response
Answered in transcript, page 43.
- p. 43 Is that in the document?
Response
Answered in transcript, page 43.
- p. 44 And that's for twenty years?
Response
Answered in transcript, pages 44 and 45.
- p. 50 Do you have an estimate of cost of decommissioning THI 17
Response
The recently enacted decomissioning rule requires around
$100 million to be set aside to assure adequate funds for j decomissioning.
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- p. 51 Is that [ funds currently available to do an imediate cleanup] out of the cleanup fund or out of additional funds that the utility would have to spend on its own.
Response
l Answered in transcript, page 51.
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Arthur Morris, Chairman Mayor of Lancaster, Pennsylvania p:'45 Ws there any attempt done to analyze, or is it part of the study, to analyze the ability of the licensee to finance this cleanup?
Whether they could financially afford to do it imediately or whether, in fact, in twenty years from now whether they'll be able to finance or be afford to do it et the time?
Response
Answered in transcript, page 45.
- p. 104 What kind of financial: responsibility the NRC would hold them to?
Response
The NRC will not require the licensee to set aside funds exclusively for the final cleanup c,f THI-2, however, che recently enacted decomit',ioning r:.le requires around $100 million to be set aside a ensure adequate funds for decomissioning.
Joel Roth Advisory Panel Member, Representirg the Public
- p. 47 Is there any provision that the NRC can make to guarantee that the funds be available at that time?
R$sponse:
Answered in transcript, page 47.
Francis Skolnick, Susquehanna Valley Alliance
. p. 54 The NRC speaks of a twenty year storage period but provides no l rationale for choosing this number.
. Response:
1 Because no inforsation was provided by the licensee as to the length of the storage period, a storage period of 20 years vias assumed because this will approximately coincide with the end of THI-2's operating license in the year 2009.
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- p. 54 The NRC tells us that immediate cleanup would require additional emergency allocations. That's in EIS Page 2.33.
. Response:
Page 2.33 of draft Supplement 3 to the PEIS states, "Imediate cleanup without PDMS could require additional emergency allocations". It has not yet been determined whether or not it will require additional emergency allocations for disposal of waste.
- p. 55 I ask why generate more water when we have already accumulated the major medium for decontamination?
Response
If the accident-generated water is available for use at the time of the final stage of cleanup, no action would areclude its use during decontamination. If it is not availa)1e, an additional source of water would be required. Because we wished to address the impact of storage and final cleanup (the impact of disposal of the accident-generated water was addressed in Supplement 2) we assumed that an additional source of water would be used.
- p. 56 Table 2.4 in the EIS, which shows an estimate of the maximum amount i of radionuclides left and their location... We want to know upon what information this table might be based. Furthermore, we want to have a complete accounting of the radionuclides present in the core at the time of the accident.
Response
The infomation on which the list of radionuclides and their l quantities (asgiveninTable2.3)isbased,isdiscussedin l Section 2.2 of Oraft Supplement 3. That infomation is reiterated as follows:
i 1) The inventory of radionuclides that were estimated
( to be present at the time of the accident was obtained from two separate analyses by two separate groups (GPU and the Electric Power Research Institute)usingtheORIGEN-2computercode.
- 2) The effect that radioactive decay would have had on the inventory of radionuclides between tha time of the accident and the projected completior sf defueling was included. The results are s bwn in Table 2.3 of Draft Supplement 3, which contains a list of the radionuclides that would ha',e inventories of greater than 1 curie on January 1, 1989.
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- 3) The effect cleanup has had on the radionuclide inventory was considered. For instance, a large quantity of the radionuclides have been removed during the defueling process. Krypton-85, a gas, has been vented to the atmos)here. Large quantities of cesium, strontium, and otler water soluble radionuclides have been filtered out of the accident-generated water and disposed of along with the ion-exchange resins in the low-level waste site.
- 4) The assumptions regarding the location of the radionuclides are based on the chemical and physical form of.the radionuclides as discussed in Sections 2.2.1, 2.2.2, and 2.2.3.
Due to the nature of the accident and the method by which the material has been removed from the reactor and shipped offsite, we can not )rovide a complete accounting of every radionuclide since tie time of the accident,
- p. 56 Looking at just two of the radionuclides, tritium, which the NRC failed to mention was an important activation aroduct, and Krypton 85, it is impossible to account for all of bot 1 of these radionuclides.
There were over 8,800 curies of tritium and over 97,000 curies of Krypton in the reactor 3t the time of the accident. How does the NRC end up with less than 1 curie of both tritium and Krypton 857
Response
Sections 2.2.2.1 and 2.2.2.2 of Draft Supplement 3 to the PEIS discuss the assumptions used to estimate the amount of krypton-85 and tritium (respectively) in the reactor after the current defueling process is complete.
Krypton-85, a gas, was released to the reactor building during the accident and was subsequently vented to the atmosphere.
Although some krypton-85 may have remained trapped between and in fuel material in the reactor vessel, during the defueling process this fraction of the krypton-85 was either released into the reactor building and removed through the stack or remained with the fuel material and shipped in the canisters to Idaho. The radiation monitor in the stack has been used to measure all the effluents from the reactor building. During the past several years, no krypton-85 has been measured, indicating that it is no longer present in measureable quantities in the reactor building.
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Tritium was produced within the reactor fuel by several mechanisms including activation and ternary fission. Greater
, than 90 percent of the tritium produced in a pressurized water
. reactor such as THI-2 is produced by ternary fission. Because tritium has the same physical properties as water, once the water is removed from the facility (one of the assumptions in this evaluation was that all the water would be removed from the facility) the tritium has also been removed. It was further assumed that any tritium remaining as dampness in the facility would either exchange with the hydrogen in the air or evaporate during the first few months after removal of the water.
- p. 57 The approval of TMI to become a site for the storage of radioactive waste raises questions about regulatory procedures and, furthermore, the acceptability of this plan to the State of Pennsylvania.
If cleanup were to continue presently, then the waste would go to out of the state sites. If it is delayed, it will largely remain
( within the state.
I I ask how can the NRC dismiss the question of the impact of the waste disposal by sayirg that it would be the subject of an analysis elsewhere? The disposal of waste at TMI is a major issue to be dealt with at this time and it is in keeping with the requirements of the National Environmental Policy Act.
Response
The environmental impact of waste disposal at a comercial low level waste disposal site, is the subject of an environmental evaluation specific to the chosen site, which must be completed before the site can ha licensed. Waste streams outside those evaluated during .ne environmental evaluation for the site will not be allowed for burial. The
, environmental evaluation for a regional burial site must be l
specific to the environmental characteristics of the site, and must also address all types of wastes that will be accepted into it, including wastes from hospitals and university research laboratories. Wastes from THI-2 will not be accepted at a regional site, until the site is licensed.
l l p. 58 How will the NRC deal with the fact that Unit 2 is in the hundred l year flood plain? Will it have to maneuver the regulations in some way that TMI will be exempt from the requirements? Will TMI be able to satisfy the ground water intrusion criteria?
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Response
4 Page4.10(Section4.1.3)ofDraftSupplement3 indicates,the
, island on which both the TMI-1 and TMI-2 reactors are located
, is not within the 100-year flood plain, however, it is within the500-yearfloodplain(0.2-percentchanceoffloodingin any given (see Enigineers year)Supplement as detornined 2, pagebyA.8 theand U.S.
A.9Army)
. Corps of The regulations will not be altered to exempt TMI from requirements.
As indicated on page 3.10, of the PEIS Draft Supplement 3, quarterly ground water monitoring would be continued during PDMS to ensure that little or no out-leakage occurs from plant buildings.
- p. 65 I suppose then another question which I would have to ask is if cleanup is delayed and resumed in whatever period of time, whenever, and they need -- I think its over a million gallons for clean up -- would that water be accident generated water?
Response
The definition of accident-generated water is presented in the nomenclature list of Draft Supplement 3 as follows:
On Fegruary 27, 1980, an agreement executed among the City of Lancaster, Pennsylvania, Metropolitan Edison Company and the NRC defined "accident-generated water" l
as:
- fuel Water handling,that existed and in the buildings containment THI-2 auxiliary,ing includ the primary system as of October 16, 1979, with the exception of water which as a result of decontamination operations becomes commingled with nonaccident-generated water such that the commingled water has a tritium content of 0.025 uCi/mL or less before processing.
. Water that has a total activity of greater than 1 uCi/mL prior to processing except where such water is originally nonaccident water and becomes contaminated by use in cleanup.
- Water that contains greater than 0.025 uCi/mL of tritium before processing.
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. The water generated during final cleanup would not meet the
, first two definitions of accident-generated water, and could meet the third definition if the quanitity of tritium in the water is greater than 0.025 uCi/ml. This translates to 23.7 curies of tritium in the entire 1,000,000 gallons used during cleanup. Because less than 1 curie of tritium is expected to remain ir the facility after defueling and drainage of all liquids, the water generated during cleanup would not meet the legal definition of accident-generated water.
Eric Epstein, Three Mile Island Alert
- p. 75 We would appreciate it;if GPU or the NRC could furnish a complete inventory of where all the radioactive materials have gone since the accident.
Response
Due to the nature of the accident and the method by which the material has bee- removed from the reactor and shipped offsite, we can not provide a complete inventory by isotope of where all the racioactive materials have gone since the accident.
- p. 75 In the document is ventilating the reactor building before each entry the same as purging it?
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Response
Yes,
- p. 75 How will the liquid releases to the Susquehanna River following PDMS differ in composition to the 2.3 million gallons of radioactive water currently stored at THl.
Response
The liquid releases to the Susquehanna River following PDMS l would be recycled through ion exchange columns as necessary, to ensure that the release rates to the Susquehanna River are below technical specificiation limits. The liquid releases would be similar in composition to the accident-generated l water after processing through ion-exchange systems, except that the liquid releases follwing PDMS would contain only trace amounts of tritium. Furtherwore, some of the shorter half-life isotopes (such as manganese-54, cerium-144 and praseodymium-144) would have decayed to negligible levels.
- p. 75 Also, just as a question, and I think I know the answer, is the public entitled to intervene if the indefinite storage option is implemented?
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Response
At the time that the licensee requests an amendment to the
. THI-2 o>erating license allowing PDMS, the public will be
. given t1e opportunity to request a hearing,
- p. 75 The final question, and I think I know the answer to this also. If the cost of the cleanup is figured in 1988 dollars, then estimates for delayed cleanup are imprecise u ! inaccurate.
Response
All costs are figured in 1988 dollars as indicated and discussed on page 3.29 (Section 3.2.6), page 3.42 (Section 3.3.6) and p 5.4 (Section 5.1). The cost estimates in the PEIS are given as, ranges for the purpose of comparison only.
These numbers represent the best estimate of cost at the time the supplement was prepared.
- p. 75 What I was curious is if the NRC factored into the economic costs the costs for retraining and rehiring workers that have been gone for some twenty years.
Response
, The additional cost from retraining workers was addressed on
! page 3.29 (section 3.2.6). It was indirectly factored into l
the cost estimates t, essuming that imme.diate cleanup would require 3 to 4 years c.nd cleanup following PDMS would require 4 years for complet 1.t.
Vera Stuchinski, Chairperson, Three Nile Island Alert
- p. 80 What's to stop GPU from making their own rules?
Response
The NRC will have a continuing onsite presence and will require the licensee to maintain the facility in accordance l with all applicable rules and regulations.
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- p. 82/83 Now, I'd like to ask Dr. Travers why the staff does not consider PDMS in the same manner as storage of the tritiated water. If a l low-level waste site license would be required for storage of the I water, why isn't it required for PDMS7 l Response:
! Answered in transcript, pages 83 to 85.
- p. 86 Do you really feel that that there would be significant decay of the radioactive material within twenty years of any long-lived radionuclides in the reactor?
Response
Answered in transcript, page 86.
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Kay Pickering, Office Coordinator, TMIA No$ questions Ed Trunk, Professor of Mechanical Engineering, Pennsylvania State Univ.
- p. 90 The question was why are we considering this question when we had a timetable before us and we're going down that timetable. Why are we considering this? Why is there a change in the timetable before us right now?
Response
Answered in transcript on page 90. ,
Joyce Corradi, Director, Concerned Mothers and Women
- p. 92 My first question is in reference to what was told to me tonight.
In the presentation by the NRC, they said that in twenty years there would be three million or more people in the area that they were relating to for their dose rate.
I'd like to know where they got their projection and how they got that projection.
Response
Answered in transcript on page 92.
- p. 92 i'd like to know fron; Mr. Standerfer where he got it from and how it was calculated.
Response
Answered in transcript by Frank Standerfer, GPU, on pages 95 and 96.
- p. 96 If, indeed, this is a criteria by which they were using to get dose rates, I should like to know where they came from, the year point end of them, and how valid and updated they are.
Response
Frank Standerfer, GPU, will supply by response by next meeting.
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Debra Davenport, Member, Concerned Mothers and Women p'., 93 I want to know what the licensee plans to do to deal with the
, materials that are directly under the reactor vessel. Is this included in any of the assessments of removai of materials from the plant?
Response
Answered in transcript by Frank Standerfer, GPU, pages 97 and 98.
- p. 94 But what is under the reactor vessel? What is passed -- I know something in the book with the nonles going into the vessel, but what about the tubes leading into the nozzles. What fuel is in there?
So I really question daether we're being told about all the fuel that's in the plant and whether there is a full assessment made on removing those fuels.
Also, I really wonder why, over a long period of time, we repeatedly seem to have a drawback from explaining to the public what might be under the reactor vessel in the basement.
Response
Answered in transcript by Frank Standerfer, GPU, pages 97 and 98.
- p. 97 I want to know, are they going to check that area under the reactor vessel, because this has been an off-again and on-again thing for the past year. Are they going to say what's there?
Response
Answered in transcript by Frank Standerfer, GPU, page 98.
l p. 98 Is it going to be left triere and how much of it is there?
The second one, in the inner core detector tube, is materials from the -- or any materials going under the reactor vessel. When are we going to know about this?
Response
Answered in transcript by Frank Standorfer, GPU, page 98.
l p. 98 Why is it to radioactive down there and you can't get in?
Response
Answered in transcript by Frank Standerfer, GPU, page 98, i p. 98 Why wouldn't it be the same -- as the rest of the --
Response
Answered in transcript by Frank Standerfer, GPU, page 99.
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