ML20196B065
| ML20196B065 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/20/1998 |
| From: | Colburn T NRC (Affiliation Not Assigned) |
| To: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| References | |
| TAC-MA0246, TAC-MA246, NUDOCS 9811300247 | |
| Download: ML20196B065 (4) | |
Text
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November 20, 1998 Mr. James W. Langenbach, Vice President and Director-TMI GPU Nuclear, Inc.'
P. O. Box 480 Middletown, PA 17057
SUBJECT:
CONTROL ROOM HABITABILITY FOLLOWUP REQUEST FOR ADDITIONAL INFORMATION FOR THREE MILE ISLAND NUCLEAR STATION, UNIT NO.1 (TMI-1) (TAC. NO. MA0246)
Dear Mr. Langenbach:
The staff has been reviewing your submittal of March 24,1998, which provided a control room habitability dose assessment analysis for the TMI 1 control room during the maximum hypothetical accident. You also provided additional information by letter dated June 19,1998, as j
requested during our May 27,1998, meeting with your staff at NRC headquarters. On November 9,1998, the NRC staff held a teleconference with members of GPU Nuclear, Inc., to discuss the status of NRC's review of the subject licensing action. During our November 9,1998, conference call with your staff, our staff provided a listing of areas of concem with respect to your submittal.
The purpose of this letteris to formalize those concems and provide recommendations for their resolution. The staff's concems relate primarily to the meteorological assumptions made in your submittals. As agreed to by your staff, we request you respond within 60 days receipt of this letter or sooner if convenient, or notify us by telephone to negotiate an alternative response date.
If you have any questions, please contact me at (301) 415-1402.
l Sincerely, Original signed by:
i-l Timothy G. Colbum, Senior Project Manager 1
Project Directorate 13 l
Division of Reactor Projects - 1/11 l
Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/ encl: See next page
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J.Langenbach Three Mile Island Nuclear Station, Unit No.1 m:
Michael Ross Robert B. Borsum i
Director, O&M, TMl B&W Nuclear Technologies
. GPU Nuclear, Inc.
Suite 525 P.O. Box 480 1700 Rockville Pike Middletown, PA 17057 Rockville, MD 20852 John C. Fomicola William Domsife, Acting Director Director, Planning and Bureau of Radiation Protection Regulatory Affairs Pennsylvania Department of GPU Nuclear, Inc.
Environmental Resources 100 Interpace Parkway ~
P.O. Box 2063 Parsippany, NJ 07054 Harrisburg, PA 17120 Jack S. Wetmore Dr. Judith Johnsrud Manager, TMI Regulatory Affairs National Energy Committee GPU Nuclear, Inc.
Sierra Club P.O. Box 480 433 Orlando Avenue Middletown, PA 17057 State College, PA 16803 Emest L. Blake, Jr., Esquire Peter W. Eselgroth, Region 1 Shaw, Pittman, Potts & Trowbridge U.S. Nuclear Regulatory Commission 2300 N Street, NW.
475 Allendale Road Washington, DC 20037 King of Prussia, PA 19406 Chairman Board of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120 Chairman Board of Supervisors of Londonderry Township R.D. #1, Geyers Church Road Middletown, PA 17057 Wayne L. Schmidt Senior Resident inspector (TMI-1)
U.S. Nuclear Regulatery Commission P.O. Box 219 Middletown, PA 17057 Regional Administrator Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 -
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i Reauest for Additional Information TMI-1 Control Room Habitability As discussed during a November 9,1998, telephone call between the NRC staff and GPU Nuclear, Inc., the NRC has two concerns regarding assumptions used in the control room X/Q calculations used in support of Topical Report TR-121, "TMI-1 Control Room Habitability for Maximum Hypothetical Accident." The following summarizes our recommendations regarding i
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these concems and provides a basis for each recommendation.
' When the ARCON96 diffuse source option is used to estimate X/Q values for an assumed i
vertical area source, the initial diffusion coefficient inputs, o, and o,, should be no larger than the width of the, area source divided by six and the height of the area source divided by six, respectively.
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p Use of the diffuse source option in the ARCON96 methodology assumes an area source release of known dimensions In the case under reNiew for Three Mile Island, it is not known if an actual release would approximate an area source nor what the dimensions of the source are likely to be. An assumption could be made that the postulated release can be approximated by a Gaussian plume having dimensions similar to the containment building cross-sectional area.
Initial' diffusion coefficients and virtual point sources have historically been used to represent area sources. The initial diffusion coefficients are estimated based on the dimensions of the area source. However, there is little experimental verification of the methods for estimating the initial values. Consequently, the values are usually selected using arguments based on the Gaussian plume assumptions. For example, assuming (1) that any actual plume at the location of release has not yet experienced reflection, (2) is essentially contained completely within the cross-sectional area, and (3) is as wide as the area source, leads to an initial horizontal diffusion coefficient that is equal to the width of the area divided by six. Similar arguments can be used to estimate initial values for vertical diffusion coefficients.
Wind tunnel experiments for a single power plant and selected meteorological conditions provide the only known set of experimental data for evaluating initial diffusion coefficients. In the experiments, tracer material was released simultaneously from about 100 points distributed uniformly over the vertical surface of the containment building and concentrations were measured at numerous downwind locations in the building complex. The ARCON96 diffusion algorithms were used to calculate X/Q values for the conditions modeled in the wind tunnel and compared with the wind tunnel X/Q values. Two sets of initial diffusion coefficients were used when making the calculations using the ARCON96 methodology. In one case the width of the area source was divided by four and the height by two, and in the other case both the height and width were divided by six. Using initial diffusion coefficients equal to the building width and height both divided by six provided the better fit to the data.
Thus, given the uncertainties identified above, use of the area source dimensions divided by six for estimating initial diffusion coefficient values seems reasonable for NRC regulatory purposes.
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1 n( l The NRC also recommends that when using ARCON96, the user should calculate X/Q values for each postulated release location, without assuming that the effluent is caught in recirculation flow such that the resultant plume assumes the dimensions of the neighbonng structure. The user should also justify assumptions regarding release characteristics such as whether a release is considered to be a point or area source from the postulated release location. For the Three Mile Island maximum hypothetical accident assessment, separate X/Q calculations should be made for the containment building, auxiliary building and borated water storage tank.
GPU Nuclear, Inc., has used test data from other building complexes to subjectively estimate building wake recirculation at Three Mile Island Unit 1. While building wake recirculation does occur, flow around buildings is very complex and dependent on specific building configurations and conditions. In the case of the auxiliary building, it is the staff's opinion that there is not sufficient quantitative justification specific to Three Mile Island Unit 1 to demonstrate that under 95 percentile X/Q conditions the effluent will flow back to and assume the dimensions of the full containment height and width prior to undergoing additional dispersion assumed in applying the ARCON96 methodology. Further, the ARCON96 methodology was developed from field test measurements made around building complexes ar,)d implicitly includes consideration of effects generated by neighboring buildings.
With respect to non-meteorological assumptions made in the analysis supporting your submittal, the staff notes that you have several other licensing applications before the NRC for review or to be submitted in the near future which have or will contain some dose assessment analysis (either onsite or offsite) in support of those applications. The staff reminds you that the assumptions made in those applications should be consistent regarding source term, postulated leakage, etc., or if different, clearly identify where they are bounding with respect to other applications already before the staff for review. This will facilitate our review of your submittals.
The staff also appreciates your staff's offer to meet with the reviewers to discuss the various submittals and will contact your staff to arrange the meeting after your pending applications have been received.