ML20003D673
| ML20003D673 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/23/1981 |
| From: | Hovey G METROPOLITAN EDISON CO. |
| To: | Barrett L Office of Nuclear Reactor Regulation |
| References | |
| LL2-81-0078, LL2-81-78, NUDOCS 8103300292 | |
| Download: ML20003D673 (14) | |
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'1 Metropolitan Edison Company gj
,9 f,,hf k'j Post Office Box 480 Middletown, Pennsvivania 17057 717 944-4041 March 23,1981 LL2-81-0078 TMI Program Of fice Attn: Lake Barrett, Deputy Director U.S. Nuclear Regulatory Commission c/o Three Mile Island Nuclear Station S
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Dear Sir:
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'dA.307198]mm Three Mile Island Nuclear Statipn, Unit 2 (TMI-2) ;?uxw g Operating License No. DPR-73 3
Docket No.
50-320 Submerged Demineralizer System "Qh
.7 This letter provides our response to your letter NRC/TMI-80-145, da ' %" n W
November 7,1980, and supplements our previous response TLL632, dated December 4, !980.
In our previous letter, we transmitted current SDS decwings to you in response to your request #1.
This letter provides our response to the remainder of your eccment s and requests for additional information.
Additionally, we have provided a copy of our recent Technical Evaluation Report for SDS.
This was subsitted to you on March 11, 1981 under cover of our letter LL2-81-0070.
In our opinion, the submittal of this letter and our TER for SDS provides adequate information to enable your prompt review of this proposed pro-cessing scenario. Your expeditious approval of this request to process containment sump water and RCS water with SDS, polished by EPICOR-II, is requested.
sincerely, G. K. Hovey Vice President and Director, TMI-2 GKH:1h 3q cc:
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. Comment #1 Letter TLL 283 provided a list of piping and instru=ent drawings and general arrangement drawings. An up-to-date listing of these drawings are requested along with the latest revision of the drawing if the drawing has a later revision date than the one provided in TLL 283. Most of the drawings provided were
" Issued 2or Approval", " Approved for Fabrication" drawings should now he available and we request that these drawdhgs be provided.
Response
'Je have provided up-to-date drawings as requested by you under cover of our letter TLL 632 dated Dece=ber 4,1980.
2.1 Comment #1 Conflicting data is available cencerning the estimated amount of water to be processed, the estimated activity in the water, the total activity to be retained in each bed and the total number of each type of bed required. For example:
Some of the discrepancies are undoubted 1v due to changing conditions and better
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information cbtained at later dates. However, the ef fect of this variation
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in data is that the authors of the documents have come to differing conclusions d
I concerning the amount of activity contained in each bed and cotal beds required.
An up to date estimate of the activity to be retained in each bed (zeolite, cation, polisher, and any other bed proposed to be used) and the tot al number of columns of each type of bed is requested. Data that is used to develop this estimate should be clearly stated and justified, including bed size, throughput and techniques to be used to determine bed lor. ding where throughput is limited by bed loading.
Response
Two sources of contaminated wcter can provide input to the Submerged Demin-eralizer System:
(1) water contained in the Reactor Coolant System and (2) 4 water that presently is in the Reactor Containment Building Sump.
With regard to the Reactor Coolant System:
1.
The RCS cold volume is approximately 11,800 cu. ft.
2..
The RCS is full.
3.
Water volume -in the RCS is, therefore, approximately 88,000 gallons.
2.2
-4 RCS sample result s, for a sample taken in February,1981, are given in Table 1.1 of our revised TER.
Those results are given below:
ANALYSIS ANALYSIS TOTAL PERFORMED RESULTS RADIOACTIVITY pH g7.6 N/A Boron 3800 ppm N/A Sodium 1240 ppm N/A 0.066 uCi/ml 22 Ci H-3 Cs-134 3.4 uCi/ml 1132 Ci Cs-137 25 uCi/mi 8347 Ci Sr-89 0.25 uCi/ml 83.27 Ci Sr-90 23 uCi/mi 7661 Ci Sb-125 1.6 x 10-3 uCi/mi 0.53 Ci With regard to the containment sump:
1.
1he volume of water in the containment sump is given in Table 1.1 of our TER.
2.
As specified in the TER, the containment sump water volume is increasing at the rate of.approximately 150 gallons / day. This volume increase tends to' provide for slight dilution of the sump water radionuclide concentration, except for Sr-90.
This slight dilution, however, is not significant.
3.
The containment sump water radionuclide concentrations given in Table'1.1 are from sample results taken in August, 1979. The results presented 4
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ANALYSIS ANALYSIS TOTAL PERFORMED RESULTS RADIOACTIVITY (Decayed to 10-80)
(based on 625,000 gal.)
H-3 0.97 uCi/ml 2295 Ci Sr-89 0.018 uCi/mi 42.6 Ci i
St-90 2.64 uci/mi 6245 Ci sb-125 9.1 x 10-3 uci/mi 21.5 Ci Cs-134 27.2 uCi/mi 64,345 Ci Cs-137 172 uCi/ml 406,890 Ci i
As specified in our TER, our plans are to use a process flow stream as depicted in Figure 1.1.
Utilization of this flow stream will permit ef fective removal of the radienuclides.
Specifically, EPICOR-11 expected ef fluents from processing containment sump water (that source of water with the higher radionuclide concentrations) is given in Table 3.1 of the TER.
Furthermore, Table 3.1 provides the expected effluent concentrations from each bed while processing sump water and is based on information presented in ORNL/TM-7448.
The table does -not depict the use of a strontium-specific media in the cation exchanger.
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, comment #3 J
The system deJign objectives in the TER include reducing concentrations in j
the processed water to levels that meet existing regulatory requirements for release to the enviror; ment. The preliminary projected stream analysis for
, intermediate streams and product water provided in TLL 283 showed that the hroposed system will not meet its design objectives.
ORNL/TM-7448 indicates even more pessimistic projections in Table 17 and provides proposed modifications to improved system performance even though these modifications will not be a
enpugh to meet the system design objective.
In view of the above, indicate your plans to improve system performance. Any proposal which does not meet the i
system design objectives should be thoroughly justified,
Response
The overall objective of decontaminating water at TMI-2 involves the utilization or radwaste processing systems best suited for that purpose. As identified, both TLL-283 and ORNL/TM-7448 indicate that the SDS should be enhanced for ef fective decontamination ef ' specific radioisotcpes and their related species.
We have enhanced SDS to include EPICOR-II polishing of SDS ef fluent.
It is expected that this type of system enhancement will be an ongoing work ef fort.
Furthermore, as more is learned about the reactor building sump water, its contaminant s and the materials selected to recove the contaminants. To this end, a program is in progress designed to optimize the resin selection so j
as to remove various contaminants from the water as these contamiaants are identified. However, it is incorrect to assume that the overall objective of water decontamination cannot be accomplished. The EPICOR II Radwaste System has
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, demonstrated the ability to decontaninate the various radioisotopes and their s pecies. Although EPICOR II has not processed reactor building sump water, a careful review of the EPICOR 11 experience indicates direct comparison of DF capability for antimony, ruthenium, niobium and the cesium and strontium species labeled as being " recalcitrant."
Therefore, ET'.COR II is planned to be opgrated in series with the SDS.
Following passage through the SDS, the water will be pesped through the EPICOR II system for final polishing. The present demand on EPICOR 11 is very slight.
It is expected water collected in the auxiliary building will be stored until an SDS outage occurs or the requirement to process auxiliary building water approaches due to a decrease in tne.available storege capacity.
Table 3.1 of our TER provides the expected performance of this combined SDS/
EPICOR II system cperation.
It should be noted the expected SDS operational capability is based on data from Table 17 of ORNL/TM-7448 report. The Met Ed TLL-283 submittal was taken from preliminsry Oak Ridge National Laboratory (ORNL) results and, therefore, the final ORNL report is considered the preferred reference. The operation of EPICOR-II is not detailed herein as this information has previously been supplied the NRC. With this combination, the objective of water decontanination will be achieved.
The SDS is undergoing optimization and will continue to be optimized even af ter system startup. It is the objective of this program to make SDS fully indepen-dent of EPICOR II while achieving necessary system DF.
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u 4.1 Comment #4 The TER, TLL-283 [in response to question 2(a)], and ORNL/TM-7448 do not all agree in the expected system DF's, in some cases differing by a factor of 100. An updated process flow diagram of the same format as Table 4 in the response to question 2(a) in TLL-283 is requested along with justification of values used.
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Response
t our letter, TLL--283, was based on preliminary information received from ORNL.
ORNL/TM-7448 is the final report and should be considered the reference document.
An updated process flow diagram is incorporated in our TER, Table 3.1. This updated flow diagram incorporates the final values as depicted in ORNL/ TM-7448.
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, Comment #5 The TER indicates that filtration is necessary to achieve designed decontamin-ation factors. ORNL/TM-7448 etates that because of flocculent in the containment sump water, the filters proposed for SDS might be inadequate.
Provide plans to ensure adequate filtering of the process water and the expected radioactivity loading of the prefilter and the final filter based on this updated information.
Based on this loading provide an estimath of the total number of prefilters and final filters needed to process the water.
Response
The SDS TER, Sections 3.1 and 5.1.1 discuss the requirement to provide for SDS influent water filtration. As stated in the TER, the prefilter will provide filtration for particles of 125 micron (nominal) size and the final filter will remove particles down to 10 microns (nominal). This filtration scheme is deemed to be adequate to perform its intended function:
provide hydraulic protection to avoid plugging the zeolite beds.
In our response to your original comments (NRC/TMI-80-89, dated May 16, 1980) we provided estimates of filter radioactivity loadings. These estimates were based on the use of the WG-P-1 pump with flow through the decay heat drop line. If this flowpath is used, these_ estimates remain valid. However, if containment sump water is removed using the surface suction scheme, f ewe r solids will be deposited in the filters because of a lower concentration of solids in the influent to the filter. The same amount of solids would remai ultimately to be ' disposed of from the sump, of course, if the surface suction scheme is used.
It is believed that once the bulk of the water is removed from the containment building sump, the problem of handling such i
residual solids will be eased considerably.
6.1 Con =ent #6 TLL-283 (in the response to question 3) provided the radioactivity loading of the cation bed and the polishing unit for 15,000 gallons of water.
Is the throughput of these colu=ns to be limited to 15,000 gallons? If not, what is the criteria to be used for replacement of these columns? Include in the discussion the ORNL/TM-7448 finding that "very little decontamination, if any, will be obtained in either the organic resin column or in the polishing columns"
- and the TER statement that "the remaining strontium (af ter the zeolite beds) is
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effectively removed by the organic cation resin."
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Response
Recognizing _that the organic ' cation bed and the originally proposed polishing unit were not etfective to accomplish their intended objectives, we have revised our processing plans. In particular, the processing plan revision was =ade in part, as a response to poor performance of these beds, as reported in the final report, ORNL/TM-7448.
As indicated in our TER, we plan to load the cation bed with a strontium-specific cationic-exchange media. This media is expected to be selected in the near' future. - At that time we will advise the NRC of our criteria for replace-ment of.the cation bed.
Furthermore, we have eliminated the previously proposed i
polishing unit, based on the infor=ation provided in ORNL/TM-7448. Our revised plan identifies that we plan to use EPICOR-II as the polishing unit for SDS
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effluents for removal of. recalcitrant species and residual radionucludes.
Table -3.1 of the TERc (which is based on results as presented in ORNL/TM-7448)
. provides information to enable your evaluation of expected system performance.
Furthermore, our planned process flow stream is depicted in Figure 1.1 of the TER.
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- Comment #7 TLL-283 (in the response to question 6) indicated that the processing method for decontamination of the RCS water would be similar to the method planned for the cont ainme nt sump water. ORNL/IM-7448 gave another recommendation concerning how to process the water in the RCS.
In view of this recommendation, provide your plan for processing RCS water.
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Response
RCS processing is planned to proceed by letting down to a Reactor Coolant Bleed Tank at a relatively low flowrate of 5-10 gpm.
During the process of letting down, makeup will be provided at the same rate to maintain a constant invent o ry in the RCS.
The makeup water would be of reactor coolant quality, appropriately borated to meet the required boron concentration as specified in the IHI-2 Recovery Technical Specifications. Processed water is the intended source of makeup water.
Our SDS TER includes RCS processing via SDS.
We have previously requested that approval be granted to process RCS water via EPICOR-II since the present
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contamination levels of the RCS are within the range of radioactive influents for which EPICOR-II has been licensed to operate (1-100 uci/ml).
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On March 13, 1981 members of the GPU technica1 ' staff provided a presentation to NRC personnel concerning RCS processing plans utilizing EPICOR-II exclusively.
Essentially, the mechanism of letdown from and makeup to the RCS remains the irrespective of the processing system, SDS followed by EPICOR-II or same
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EPICOR-II alone. As indicated to the NRC Staff, the option of processing I
the RCS using EPICCR-II would be utilized only if the SDS were unavailable for i
some reason for an extended period.
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- Comment #8 By mid-1981, burial grounds will require such vastes as the polishing unit resin to be solidified prior to disposal. Provide plans for meeting this
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projected requirement for the polishing unit resin.
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Response
4j' The' polishing unit has been deleted fromgthe SDS processing scheme.
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, Comment #9 Provide an accident analysis of dropping a cask containing a loaded zeolite resin liner from the maximum height of crane travel onto (a) the 305' level of the fuel handling building and (b) the SPC system and its supporting compo-nents (eg. N2 system).
Include in the response a summary of the health and environmental effects on the public and on operators in the area and the ef fect I
on the reactor coolant system.
Response
Section 7.5 of the SDS TER prpvides the sucmary requested concerning the drop of a shipping cask containing a loaded zeolite resin liner onto the 305' elevation.
The analysis of the cask drop showed the following:
1.
The effect on plant operators and of f-site is given in Section 7.5 and shows that the public health and safety are not comprised.
2.
A detailed study of cask drops from the maximum height to el 305' between the TMI-l and TMI-2 Fuel Handling buildings shows that by routing the lif ted cask through the safety zones specified in the TMI-l FSAR, no damage which could prevent safe Reactor shutdown / cooling will occur.
3.
The cask drop onto the SPC system will not res ult in failure to maintain continuous R.C. pressure. Existing plant ~ emergency procedures ensure maintaineace of continuous R.C. pressure.
4.
The cask drop on to the N2 support system could conceivably result in the creation of missle hazard if the cask is dropped in a manner that causes the end of one of the N2 bottles to be sheared of f.
The hazard is being studied further. The results of'the analysis will be forwarded when available; the approximate date will be June 1,1981.
10.1 Comment #10 Provide an accident analysis of lifting a loaded zeolite resin liner above
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the pool surf ace.
Response
Section 7.4 of the SDS TER provides this analysis.
It should be noted, however, th at this hypothetical occurrence is considered to be extremely unlikely. The t
l lifting tool for the zeolite vessels has been designed such that, under normal circumstances, a zeolite vessel could be lifted no higher than about 8' below the surf ace of the water.
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