ML20069H694

Revised Testimony of GP Lahti on Contentions 39 & 109 Re Radiological Impacts Associated W/Radioactive Contaminant Release to Groundwater

ML20069H694
Person / Time
Site:	Byron
Issue date:	04/04/1983
From:	Lahti G COMMONWEALTH EDISON CO., SARGENT & LUNDY, INC.
To:
Shared Package
ML20069H668	List: ML20069H665 ML20069H690 ML20069H694 ML20069H705
References
ISSUANCES-OL, NUDOCS 8304060320
Download: ML20069H694 (11)

Text

Edison 04/04/83 g- Revised Testimony UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL

) 50-455 OL (Byron Nuclear Power Station, )

Units 1& 2) )

SUMMARY

OF TESTIMONY OF GERALD P. LAHTI CONCERNING CONSOLIDATED CONTENTIONS 39 AND 109 Mr Lahti is an engineer, employed by Sargent &

Lundy, the Byron Station Architect-Engineer. He is Assistant Division Head of the Nuclear Safeguards and Licensing Division in charge of shielding and radiological safety. His testimony addresses the radiological impacts associated with assumed releases to the groundwater of radioactive contaminants.

Mr. Lahti first discusses the design basis event analyzed in the Byron FSAR which pertains to groundwater releases. Based on the travel times for transport of radionuclides through the groundwater and the dilution of 1

radionuclides with the groundwater, Mr. Lah.i concludes that the radiological consequences associated with the design basis event would not exceed the limits established by 10 CFR Part i

20. M r . Holish's testimony addresses the assumptions and

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calculations regarding the applicable dilution factor and the l

travel time of contaminants through the groundwater.) Mr.

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Lahti also discusses the consequences of releases to the groundwater associated with a core melt scenario. Although a detailed assessment and analysis of core melt events has not been performed for Byron, Mr. Lahti concludes that because of the travel times involved, interdictive measures to mitigate radiological consequences could be taken.

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

COMMONWEALTH EDISON COMPANY ) Docket No. 50-454-OL

) 50-455-OL (Byron Station, Units 1 and 2) )

TESTIMONY OF GERALD P. LAHTI CONCERNING CONSOLIDATED CONTENTIONS 39 and 109 Q.l. State your name and present occupation.

A.l. My name is Gerald P. Lahti. I am Assistant Division Head of the Nuclear Safeguards and Licensing Division in charge of Shielding and Radiological Safety at Sargent & Lundy in Chicago, Illinois.

Q.2. State your educational and professional qualifications.

A.2. I received a BSCE in Civil Engineering from Wayne State University in 1959. I received a MSE (Nuclear Engineering) from the University of Michigan in 1960, and completed ~ additional part time course work in Mechanical and Nuclear Engineering at the University of Delaware, University of Toledo and Case Western Reserve University. From 1960 to 1963, I was employed by E. I. duPont deNemours & Co.,

ll i-l Inc., where I mathematically analyzed and designed polymer transfer systems and extrusion dies.

! From 1963 to 1973, I was a member of the National Aeronautics and Space Administration staff at Lewis Research Center, Cleveland, Ohio. There, I l evaluated radiation hazards and designed radiation shields for nuclear reactors considered for power or

propulsion systems in space vehicles. In 1968 I assumed supervisory responsibilities in this area.

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l In 1973 I joined Sargent & Lundy and I have been i

f employed continuously since that time in the Shielding and Radiological Safety Section. This Section, which is under my supervision, evaluates i and designs all radiation shielding and other radiation protection features incorporated in nuclear power plant design. I am also responsible for assessing the radiological impact of radionuclides released during normal and abnormal power plant operations.

I am a Registered Professional Engineer in the State of Illinois and a member of the American Nuclear Society (ANS) and Health Physics Society. I am a

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, . - -__ . - - .._ _- = . ._- . _ _ .. - . ._.. .

l past Chairman of the AMS's Radiation Protection and 1

Shielding Division.

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Q.3. What is the scope of your testimony?

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] A.3. My testimony addresses Consolidated Contentions 39 and 109, which state:

Since the ground water system underlying the Byron site has not been characterized adequately, the consequences of radionuclide releases to the underlying aquifer cannot be ,.

predicted with confidence. In consequence, no proper NEPA analysis can of this important subject can be made. . . .

Specifically, my testimony addresses the possible consequences of. accidental release of radionuclides to the ground water of the Byron Station resulting i from postulated accident scenarios.

0.4. What accident scenarios have you considered i

regarding the manner in which radioactivity is released to the groundwater?

A.4. I have considered two accident scenarios. The first l is a design basis event which is postulated to I .

l demonstrate the suitability of the site. It involves postulating the rupture of one of the 125,000 gallon boron recycle holdup (BRH) tanks located in the auxiliary building and subsequently ,

releasing radioactivity into the ground water.. The 1

second scenario involves postulating a core melt l

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_4-accident as discussed in the Byron Final Environmental Statement which was prepared by the NRC Staff.

0.5. Please describe the BRH tank rupture scenario.

A.S. The postulated design basis accident involving the BRH tank is discussed in Section 2.4.13.3 of the Byron FSAR. I prepared the radiological portion of this section and I am incorporating it as part of this testimony. The BRH tank is located 54 feet below grade in the auxiliary building. The floor at this elevation consists of eight feet of concrete.

Thus, the base of the concrete floor is 62 feet below the ground surface.

l The postulated design basis accident assumes that 125,000 gallon tank is full, ruptures and that its

! radioactive contents percolate onto the foundation bedrock through a .1 inch wide crack assumed to have developed along the entire width of the auxiliary 1

building. This crack is also assumed to have penetrated through the eight foot thick concrete l floor.

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! Q.6. What parameters are used in the evaluation of the consequences of the postulated accident involving j the BRH tank?

A.6. The leak rate of the BRH tank fluid through the crack in the auxiliary building floor is calculated

! -8 to be 2.03 x 10 cfs per foot of crack. Thus, the contaminated water percolates slowly to the founda-

, tion bedrock. .Thereafter, the contaminated water mixes with the groundwater resulting in dilution of the radioactive concentration by a factor of 2,200.

i The calculated travel time of the contaminated flow i

from the point of release to the nearest well is estimated to be 30.49 years. Both the travel time and the dilution factor have been developed by Mr.

Holish and are discussed in his testimony.

l I should also emphasize that the contaminated l

groundwater, driven by the hydraulic gradient, is first assumed to travel through the relatively I

( impermeable grouted rock mass underlying the plant and then through ungrouted rock. For reasons of i

, conservatism, no further dilution of the groundwater

, stream in either a transverse direction or in the e

direction of flow is assumed. Furthermore, no removal of radioactive contamination either by f

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--va'OmM' r ' " -P-- e a- *WCT -

weer*'

. - - - . .- - - . - _ - = _ .-_ - - . -- _ __ -

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adsorption (physical trapping) or absorption 4

(chemical ion exchange) is assumed during this time.

j Q.7. Have you determined the radiological consequences which result from this postulated accident?

l A.7. Yes. Obviously the principal concern associated l with the accident ccenario I have just described relates to possible consumption of contaminated water by the pnblic. Since it would take approximately 30 years for contaminated water to reach the nearest well, which is owned by j Commonwealth Edison, the vast majority of the nuclides released by the postulated accident would decay to negligible levels. The nuclides with longer half-lives which would not have decayed to l such levels are Cs-134, Cs-137 and H-3. However,

, because of the initial dilution of the BRH tank

fluid in the groundwater, the concentrations of these nuclides, as depicted on Table I of my i

testimony, would be well within the limits of i

P allowed concentrations of radionuclides in water to unrestricted areas established by Column 2 of Table II of Appendix B to 10 C.F.R. Part 20. Thus, the consequences of this postulated accident would not represent a public threat.

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i l Q.8. Have you considered the effect of a core melt i

i accident on-the groundwater underlying the Byron

site?

A.8. Yes, in a general sense , I have. This issue has not

! received my detailed consideration due to the extreme unlikelihood of such an event occurring as I

l explained in Mr. Klopp's testimony.

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, First, in terms of ground water impacts, it must be l j  !

j remembered that we believe that the majority of the I radioactive materials released as a result of a postulated core melt would likely be released to the containment atmosphere or be plated out on cooler i

f surfaces within containment. The remainder of the i fission products would remain in a slag-like l

material mix of uranium, structural metals and l concrete. Assuming this mass melted through the l

l concrete base mat its heat would likely flash any encountered ground water to steam, providing little source to the groundwater. Later, when sufficiently cool, the ground water could slowly leach fission products from the remaining mass.

Under these circumstances, and given the travel times involved, it would be possible to take measures to interdict the groundwater flow, by w--- y-.wew -- ,

• gw y ,y --?y- --er--*y e-M -p*~----

pumping to control groundwater gradients or grouting, so as to mitigate any radiological consequences.

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TABLE 1 Radionuclide Concentrations in the Nearest Well Due to a Boron Recycle Holdup Tank Rupture Allouable -

Limits 10 CFR 20 Appendix B Tank Concentration Table II, Concentrations

• at well Column 2 Nuclide JT Ci/g X Ci/ml*** A( Ci f ul Rb-88 3.7-2** 0 3.0-5 Rb-89 2.1-3 0 3.0-6 Mo-99 5.3-2 0 2 .' ) - 4 I-131 2.5-2 0 3.0-7 I-132 2.8-2 0 8.0-6 I-133 4.0-2 0 1.0-6 I-134 5.6-3 0 2.0-5 I-135 2.2-2 0 4.0-6 Cs-134 2.3-2 3.7-10 9.0-6 Cs-136 2.8-10 0 9.0-5 l

l Cs-137 1.5-2 3.4-6 2.0-5 Cs-138 9.8-3 0 3.0-6 l Ba-137m 1.4-2 0 3.0-6 l

l H-3 3.5+0 2.9-4 3.0-3 1

FSAR Table 2.4-20

• • -2 The notation 3.7-2 means 3.7 x 10

• • • After decay of 30.49 years and dilution by a factor of l 2200 l

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